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Sample records for river elec pwr

  1. Withlacoochee River Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Withlacoochee River Elec Coop Jump to: navigation, search Name: Withlacoochee River Elec Coop Place: Florida Phone Number: 352-567-5133 Website: www.wrec.net Twitter: https:...

  2. East Mississippi Elec Pwr Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    search Name: East Mississippi Elec Pwr Assn Place: Mississippi Phone Number: Meridian Office: 601-581-8600 -- Quitman Office: 601-776-6271 -- DeKalb Office: 601-743-2641 --...

  3. South River Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Elec Member Corp Jump to: navigation, search Name: South River Elec Member Corp Place: North Carolina Phone Number: (910) 892-8071 Website: www.sremc.com Twitter: https:...

  4. Red River Valley Rrl Elec Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec Assn Jump to: navigation, search Name: Red River Valley Rrl Elec Assn Place: Oklahoma Phone Number: 1-800-749-3364 or 580-564-1800 Website: www.rrvrea.com Twitter:...

  5. Raft River Rural Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Raft River Rural Elec Coop Inc Place: Idaho Service Territory: Idaho, Utah, Nevada Phone Number: 208-645-2211 Website: rrelectric.com Facebook: https:www.facebook.compages...

  6. East River Elec Pwr Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Facebook: https:www.facebook.compagesEd-the-Energy-Expert431620883566287?refts&frefts Outage Hotline: (605) 256-8057 or (605) 256-8056 or (605) 256-8059...

  7. Singing River Elec Pwr Assn (Mississippi) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    9,647.445 93,322.028 60,225 3,117.42 30,825.248 8,207 692.763 8,259.846 11 13,457.628 132,407.122 68,443 2008-06 9,059.584 86,892.462 60,106 3,046.146 30,089.083 8,193 709.428...

  8. Red River Valley Coop Pwr Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Red River Valley Coop Pwr Assn Jump to: navigation, search Name: Red River Valley Coop Pwr Assn Place: Minnesota Website: www.rrvcoop.com Facebook: https:www.facebook.comRRVCPA...

  9. Pearl River Valley El Pwr Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Valley El Pwr Assn Jump to: navigation, search Name: Pearl River Valley El Pwr Assn Place: Mississippi Phone Number: Columbia: 601-736-2666 -- Hattiesburg: 601-264-2458 -- Purvis:...

  10. Fall River Rural Elec Coop Inc (Wyoming) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Website: www.fallriverelectric.com Facebook: https:www.facebook.comFallRiverREC Outage Hotline: 1.866.887.8442 (After Hours) Outage Map: outage.fallriverelectric.como...

  11. Choctawhatche Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Choctawhatche Elec Coop, Inc Jump to: navigation, search Name: Choctawhatche Elec Coop, Inc Place: Florida Phone Number: (850) 892-2111 Website: www.chelco.com Twitter: https:...

  12. Washington Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Washington Elec Member Corp Jump to: navigation, search Name: Washington Elec Member Corp Place: Georgia Phone Number: 478-552-2577; 1-800-552-2577 Website: washingtonemc.com...

  13. Intermountain Rural Elec Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Rural Elec Assn Place: Colorado Website: www.irea.coop Twitter: @IREAColorado Facebook: https:www.facebook.comIntermountainREA Outage Hotline: 1-800-332-9540 References:...

  14. Mountrail-Williams Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Mountrail-Williams Elec Coop Jump to: navigation, search Name: Mountrail-Williams Elec Coop Place: North Dakota Phone Number: Williston Office- 701-577-3765 -- Stanley Office-...

  15. Hess Retail Natural Gas and Elec. Acctg. (Delaware) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. (Delaware) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: Delaware References: EIA Form EIA-861 Final...

  16. Hess Retail Natural Gas and Elec. Acctg. (Connecticut) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. (Connecticut) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: Connecticut Phone Number: 212-997-8500...

  17. Hess Retail Natural Gas and Elec. Acctg. (District of Columbia...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. (District of Columbia) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: District of Columbia References:...

  18. Public Service Elec & Gas Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec & Gas Co (Redirected from PSEG) Jump to: navigation, search Name: Public Service Elec & Gas Co Abbreviation: PSEG Place: New Jersey Year Founded: 1903 Phone Number:...

  19. Upson Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Name: Upson Elec Member Corp Place: Georgia Website: www.upsonemc.comUpson%20EMC%2 Facebook: https:www.facebook.comupson.emc Outage Hotline: 706-647-5475 References: EIA...

  20. Northern Virginia Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    NOVEC) Jump to: navigation, search Name: Northern Virginia Elec Coop Place: Manassas, Virginia References: EIA Form EIA-861 Final Data File for 2010 - File1a1 SGIC2 EIA Form...

  1. Northern Virginia Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Northern Virginia Elec Coop Place: Manassas, Virginia References: EIA Form EIA-861 Final Data File for 2010 - File1a1 SGIC2 EIA Form 861 Data Utility Id 13640 Utility Location...

  2. New England Hydro-Tran Elec Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    New England Hydro-Tran Elec Co Jump to: navigation, search Name: New England Hydro-Tran Elec Co Place: Massachusetts Phone Number: 860 729 9767 Website: www.nehydropower.com...

  3. Big Horn County Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    County Elec Coop, Inc Jump to: navigation, search Name: Big Horn County Elec Coop, Inc Place: Montana Phone Number: (406) 665-2830 Website: www.bhcec.com Outage Hotline: (406)...

  4. Central Hudson Gas & Elec Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Gas & Elec Corp Jump to: navigation, search Name: Central Hudson Gas & Elec Corp Place: New York Phone Number: 845-452-2700 or 1-800-527-2714 Website: www.centralhudson.com...

  5. HHH FEC Cooperation Mach Elec Co Ltd | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    HHH FEC Cooperation Mach Elec Co Ltd Jump to: navigation, search Name: HHH-FEC Cooperation Mach.&Elec. Co., Ltd Place: Weihai, Shanghai Municipality, China Zip: 264209 Sector:...

  6. Clearwater-Polk Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Clearwater-Polk Elec Coop Inc Jump to: navigation, search Name: Clearwater-Polk Elec Coop Inc Place: Minnesota Phone Number: 218-694-6241 Website: www.clearwater-polk.com Outage...

  7. Hess Retail Natural Gas and Elec. Acctg. (Maine) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. (Maine) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: Maine Phone Number: 1-800-437-7645 Website:...

  8. Brown County Rural Elec Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Rural Elec Assn Jump to: navigation, search Name: Brown County Rural Elec Assn Place: Minnesota Phone Number: 1-800-658-2368 Website: www.browncountyrea.coop Outage Hotline:...

  9. Virginia Mun Elec Assn No 1 | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec Assn No 1 Jump to: navigation, search Name: Virginia Mun Elec Assn No 1 Place: Virginia Website: www.mepav.org References: EIA Form EIA-861 Final Data File for 2010 -...

  10. Joe Wheeler Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Joe Wheeler Elec Member Corp Jump to: navigation, search Name: Joe Wheeler Elec Member Corp Place: Alabama Phone Number: (256) 552-2300 Website: www.jwemc.org Twitter: @jwemc...

  11. Deep East Texas Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Deep East Texas Elec Coop Inc Jump to: navigation, search Name: Deep East Texas Elec Coop Inc Place: Texas Phone Number: 1-800-392-5986 Website: www.deepeast.com Facebook: https:...

  12. Mora-San Miguel Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Mora-San Miguel Elec Coop, Inc Jump to: navigation, search Name: Mora-San Miguel Elec Coop, Inc Place: New Mexico Phone Number: 575-387-2205 (Mora) -- 505-757-6490 (Pecos) Website:...

  13. Barrow Utils & Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Utils & Elec Coop, Inc Jump to: navigation, search Name: Barrow Utils & Elec Coop, Inc Place: Alaska Phone Number: 907-852-6166 Website: www.bueci.org Outage Hotline: After Hours:...

  14. Rich Mountain Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Rich Mountain Elec Coop, Inc Jump to: navigation, search Name: Rich Mountain Elec Coop, Inc Place: Arkansas Phone Number: 1-877-828-4074 Website: www.rmec.com Outage Hotline:...

  15. Cavalier Rural Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Rural Elec Coop, Inc Jump to: navigation, search Name: Cavalier Rural Elec Coop, Inc Place: North Dakota Phone Number: 701-256-5511 Facebook: https:www.facebook.compages...

  16. Harrison County Rrl Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Harrison County Rrl Elec Coop Jump to: navigation, search Name: Harrison County Rrl Elec Coop Place: Iowa Phone Number: 712-647-2727 Website: www.hcrec.coop Outage Hotline:...

  17. Harrison Rural Elec Assn, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Harrison Rural Elec Assn, Inc Jump to: navigation, search Name: Harrison Rural Elec Assn, Inc Place: West Virginia Phone Number: 304.624.6365 Website: www.harrisonrea.com...

  18. Panola-Harrison Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Panola-Harrison Elec Coop, Inc Jump to: navigation, search Name: Panola-Harrison Elec Coop, Inc Place: Texas Phone Number: (903) 935-7936 Website: www.phec.us Facebook: https:...

  19. Nelson Lagoon Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Lagoon Elec Coop Inc Jump to: navigation, search Name: Nelson Lagoon Elec Coop Inc Place: Alaska Phone Number: (907) 989-2204 Website: www.swamc.orghtmlsouthwest-a Outage...

  20. East End Mutual Elec Co Ltd | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    End Mutual Elec Co Ltd Jump to: navigation, search Name: East End Mutual Elec Co Ltd Place: Idaho Phone Number: (208) 436-9357 Website: www.electricunion.orgcompany- Outage...

  1. Public Service Elec & Gas Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec & Gas Co Jump to: navigation, search Name: Public Service Elec & Gas Co Abbreviation: PSEG Place: New Jersey Year Founded: 1903 Phone Number: 1-800-436-7734 Website:...

  2. Sioux Valley SW Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    SW Elec Coop Jump to: navigation, search Name: Sioux Valley SW Elec Coop Place: Colman, South Dakota References: EIA Form EIA-861 Final Data File for 2010 - File1a1 SGIC2 EIA...

  3. Copper Valley Elec Assn, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Valley Elec Assn, Inc Jump to: navigation, search Name: Copper Valley Elec Assn, Inc Place: Alaska Phone Number: Copper Basin: 907-822-3211 or Valdez: 907-835-4301 Website:...

  4. Wayne-White Counties Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Wayne-White Counties Elec Coop Jump to: navigation, search Name: Wayne-White Counties Elec Coop Place: Illinois Phone Number: (618) 842-2196 Website: waynewhitecoop.com Facebook:...

  5. Tipmont Rural Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    search Name: Tipmont Rural Elec Member Corp Abbreviation: Tipmont REMC Address: 403 S Main St Place: Linden, Indiana Zip: 47955 Phone Number: 800-726-3953 Website:...

  6. Hess Retail Natural Gas and Elec. Acctg. (Pennsylvania) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    Pennsylvania) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: Pennsylvania References: EIA Form EIA-861 Final Data File for 2010 - File220101...

  7. Denton County Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    County Elec Coop, Inc Place: Texas Service Territory: Texas Website: www.coserv.com Outage Hotline: (800) 274-4014 Outage Map: outagemap.coserv.comexternal References: EIA...

  8. Central Valley Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Coop, Inc Jump to: navigation, search Name: Central Valley Elec Coop, Inc Place: New Mexico Phone Number: (575) 746-3571 Website: cvecoop.org Outage Hotline: (575) 746-3571...

  9. North Central Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Central Elec Coop, Inc Place: Ohio Website: www.ncelec.org Twitter: @NorthCentralEC Facebook: https:www.facebook.comNorthCentralElectric Outage Hotline: 419-426-3072 ...

  10. Buckeye Rural Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Buckeye Rural Elec Coop, Inc Place: Ohio Website: www.buckeyerec.commain Facebook: https:www.facebook.combuckeyerec Outage Hotline: 1-800-282-7204 References: EIA Form EIA-861...

  11. Bailey County Elec Coop Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec Coop Assn Place: Texas Phone Number: (806) 272-4504 Website: www.bcecoop.com Facebook: https:www.facebook.combcecoop Outage Hotline: (806) 272-4504 References: EIA Form...

  12. Comanche County Elec Coop Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Comanche County Elec Coop Assn Place: Texas Website: www.ceca.coophome.aspx Facebook: https:www.facebook.comCECA.coop Outage Hotline: 1-800-915-2533 References: EIA Form...

  13. New England Elec Transm'n Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Transm'n Corp Jump to: navigation, search Name: New England Elec Transm'n Corp Place: New Hampshire References: EIA Form EIA-861 Final Data File for 2010 - File1a1 EIA Form 861...

  14. Southern Pine Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    search Name: Southern Pine Elec Coop, Inc Place: Alabama Phone Number: Atmore Office: 251.368.4842; Brewton Office: 251.867.5415; Evergreen Office: 251.578.3460; Frisco...

  15. South Louisiana Elec Coop Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    search Name: South Louisiana Elec Coop Assn Place: Louisiana Phone Number: Houma Office: (985) 876-6880 or Amelia Office: (985) 631-3605 Website: www.sleca.com Facebook:...

  16. Blue Ridge Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Blue Ridge Elec Member Corp Place: North Carolina Phone Number: 1-800-448-2383 Website: www.blueridgeemc.com Twitter: @blueridgeemc Facebook: https:www.facebook.comBlueRidgeEMC...

  17. French Broad Elec Member Corp (Tennessee) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    French Broad Elec Member Corp Place: Tennessee Phone Number: (828)649-2051 or (828)688-4815 or (800)222-6190 or (828)682-6121 Website: www.frenchbroademc.com Twitter:...

  18. French Broad Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    French Broad Elec Member Corp Place: North Carolina Phone Number: (828)649-2051 or (828)688-4815 or (800)222-6190 or (828)682-6121 Website: www.frenchbroademc.com Twitter:...

  19. Hess Retail Natural Gas and Elec. Acctg. (Maryland) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    Maryland) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: Maryland References: EIA Form EIA-861 Final Data File for 2010 - File220101 EIA Form...

  20. Hess Retail Natural Gas and Elec. Acctg. (Massachusetts) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. Place: Massachusetts Phone Number: 212-997-8500 Website: www.hess.com Twitter: @HessCorporation Facebook: https:www.facebook.com...

  1. Hess Retail Natural Gas and Elec. Acctg. (Rhode Island) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    Rhode Island) Jump to: navigation, search Name: Hess Retail Natural Gas and Elec. Acctg. Place: Rhode Island References: EIA Form EIA-861 Final Data File for 2010 - File220101...

  2. Hess Retail Natural Gas and Elec. Acctg. (New Hampshire) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. Place: New Hampshire Phone Number: 1-800-437-7645 Website: www.hess.com Twitter: @HessCorporation Facebook: https:www.facebook.com...

  3. Morgan County Rural Elec Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    search Name: Morgan County Rural Elec Assn Place: Colorado Website: www.mcrea.org Twitter: @MorganCountyREA Facebook: https:www.facebook.compagesMorgan-County-Rural-Ele...

  4. Heartland Rural Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec Coop, Inc Place: Kansas Phone Number: (800) 835-9586 Website: www.heartland-rec.com Twitter: @HeartlandREC Facebook: https:www.facebook.comHeartlandREC Outage Hotline:...

  5. Elec District No. 5 Maricopa C | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    District No. 5 Maricopa C Jump to: navigation, search Name: Elec District No. 5 Maricopa C Place: Arizona Phone Number: (480) 610-8741 Outage Hotline: (480) 610-8741 References:...

  6. North Georgia Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    navigation, search Name: North Georgia Elec Member Corp Place: Georgia Phone Number: Dalton: (706) 259-9441; Fort Oglethorpe: (706) 866-2231; Calhoun: (706) 629-3160; Trion:...

  7. Rich Mountain Elec Coop, Inc (Oklahoma) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Inc (Oklahoma) Jump to: navigation, search Name: Rich Mountain Elec Coop, Inc Place: Oklahoma Phone Number: 1-877-828-4074 Website: www.rmec.com Outage Hotline: 1-877-828-4074...

  8. Oliver-Mercer Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Oliver-Mercer Elec Coop Inc Place: North Dakota References: Energy Information Administration.1 EIA Form 861 Data Utility Id 14088 This article is a stub. You can help OpenEI...

  9. Grundy County Rural Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Elec Coop Place: Iowa Phone Number: 319-824-5251 Website: www.grundycountyrecia.com Outage Hotline: 1-800-390-7605 Outage Map: www.iowarec.orgoutages References: EIA Form...

  10. Panola-Harrison Elec Coop, Inc (Louisiana) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Louisiana) Jump to: navigation, search Name: Panola-Harrison Elec Coop, Inc Place: Louisiana Phone Number: (318) 933-5096 Outage Hotline: (318) 933-5096 References: EIA Form...

  11. Delaware County Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Delaware County Elec Coop Inc Place: New York Phone Number: (607) 746-9283 or Toll Free at (866) 436-1223 Website: www.dce.coop Facebook: https:www.facebook.compages...

  12. Paulding-Putman Elec Coop, Inc (Indiana) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Jump to: navigation, search Name: Paulding-Putman Elec Coop, Inc Address: 401 McDonald Pike Place: Paulding, Ohio Zip: 45879-9270 Service Territory: Indiana, Ohio Phone Number:...

  13. A Resource assessment protocol for GEO-ELEC | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Resource assessment protocol for GEO-ELEC Jump to: navigation, search OpenEI Reference LibraryAdd to library Report: A Resource assessment protocol for GEO-ELEC Authors...

  14. Sam Rayburn G&T Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Sam Rayburn G&T Elec Coop Inc Jump to: navigation, search Name: Sam Rayburn G&T Elec Coop Inc Place: Texas Phone Number: (936) 560-9532 Outage Hotline: (936) 560-9532 References:...

  15. Steuben Rural Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Steuben Rural Elec Coop, Inc Place: New York Phone Number: 607-776-4161 or 800-843-3414 or 716-296-5651 or 800-883-8236 Website: www.steubenrec.coop Outage Hotline: 1-866-430-4293...

  16. Duck River Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    10,355 1,290 21,368 18 13,312 170,739 70,025 2008-01 8,728 110,789 59,691 2,848 31,132 10,373 1,150 18,079 18 12,726 160,000 70,082 References "EIA Form EIA-861 Final...

  17. MHK Projects/Homeowner Tidal Power Elec Gen | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Homeowner Tidal Power Elec Gen < MHK Projects Jump to: navigation, search << Return to the MHK database homepage Loading map... "minzoom":false,"mappingservice":"googlemaps3","typ...

  18. Keosauqua Municipal Light & Pwr | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Keosauqua Municipal Light & Pwr Jump to: navigation, search Name: Keosauqua Municipal Light & Pwr Place: Iowa Phone Number: 319-293-3406 Website: villagesofvanburen.comdirecto...

  19. Hess Retail Natural Gas and Elec. Acctg. (New York) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    Hess Retail Natural Gas and Elec. Acctg. Place: New York References: EIA Form EIA-861 Final Data File for 2010 - File220101 EIA Form 861 Data Utility Id 22509 This article is a...

  20. Preliminary study on direct recycling of spent PWR fuel in PWR...

    Office of Scientific and Technical Information (OSTI)

    Preliminary study on direct recycling of spent PWR fuel in PWR system Citation Details ... conference on advances in nuclear science and engineering, Bali (Indonesia), 14-17 ...

  1. PWR AXIAL BURNUP PROFILE ANALYSIS

    SciTech Connect (OSTI)

    J.M. Acaglione

    2003-09-17

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B&W 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001).

  2. VERA Core Simulator Methodology for PWR Cycle Depletion (Conference...

    Office of Scientific and Technical Information (OSTI)

    VERA Core Simulator Methodology for PWR Cycle Depletion Citation Details In-Document Search Title: VERA Core Simulator Methodology for PWR Cycle Depletion Authors: Kochunas, ...

  3. The RenewElec Project: Variable Renewable Energy and the Power System

    SciTech Connect (OSTI)

    Apt, Jay

    2014-02-14

    Variable energy resources, such as wind power, now produce about 4% of U.S. electricity. They can play a significantly expanded role if the U.S. adopts a systems approach that considers affordability, security and reliability. Reaching a 20-30% renewable portfolio standard goal is possible, but not without changes in the management and regulation of the power system, including accurately assessing and preparing for the operational effects of renewable generation. The RenewElec project will help the nation make the transition to the use of significant amounts of electric generation from variable and intermittent sources of renewable power.

  4. Northeast Missouri El Pwr Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Pwr Coop Jump to: navigation, search Name: Northeast Missouri El Pwr Coop Place: Missouri Phone Number: 573-769-2107 Website: www.northeast-power.coop Outage Hotline: 573-769-2107...

  5. Polk County Rural Pub Pwr Dist | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Polk County Rural Pub Pwr Dist Jump to: navigation, search Name: Polk County Rural Pub Pwr Dist Place: Nebraska Phone Number: (888) 242-5265 Website: www.pcrppd.com Outage...

  6. Sam Rayburn Municipal Pwr Agny | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Municipal Pwr Agny Jump to: navigation, search Name: Sam Rayburn Municipal Pwr Agny Place: Texas Phone Number: 936-336-3684 or 936-336-5666 Website: www.cityofliberty.orgGOVERNME...

  7. Central Montana E Pwr Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    E Pwr Coop Inc Jump to: navigation, search Name: Central Montana E Pwr Coop Inc Place: Montana Phone Number: 406-268-1211 Website: www.cmepc.org Outage Hotline: 406-268-1211...

  8. Improving fuel-rod performance. [PWR; BWR

    SciTech Connect (OSTI)

    Ocken, H.; Knott, S.

    1981-03-01

    To reduce the risk of fuel-rod failures, utilities operate their nuclear reactors within conservative limits on power increases proposed by nuclear-fuel vendors. Of particular concern to US utilities is that adopting these limits results in an industrywide average plant capacity loss of 3% in BWR designs and 0.3% in PWR designs. To replace lost BWR capacity by other generating means currently costs the utilities $150 million annually, and losses for PWRs are about $20 million. Efforts are therefore being made to identify the factors responsible for Zircaloy degradation under PCI condition and to improve nuclear-fuel-rod design and reactor operation.

  9. South Mississippi El Pwr Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    South Mississippi El Pwr Assn Place: Mississippi Phone Number: 601.268.2083 Website: www.smepa.coop Outage Hotline: 601.268.2083 References: EIA Form EIA-861 Final Data File for...

  10. Effects of Multiple Drying Cycles on HBU PWR Cladding Alloys

    Office of Energy Efficiency and Renewable Energy (EERE)

    The purpose of this research effort is to determine the effects of canister/cask vacuum drying and storage on radial hydride precipitation in high‐burnup (HBU) pressurized water reactor (PWR)...

  11. Grand Valley Rrl Pwr Line, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Valley Rrl Pwr Line, Inc Place: Colorado Website: www.gvp.org Twitter: @GVRuralPower Outage Hotline: 970-242-0040 Outage Map: www.gvp.orgcontentoutage-map References: EIA Form...

  12. Impact of High Burnup on PWR Spent Fuel Characteristics (Journal...

    Office of Scientific and Technical Information (OSTI)

    Reducing the burden of management of spent nuclear fuel is important to the future of nuclear energy. The impact of higher pressurized water reactor (PWR) fuel burnup is examined ...

  13. Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR...

  14. PWR representative behavior during a LOCA

    SciTech Connect (OSTI)

    Allison, C.M.

    1981-01-01

    To date, there has been substantial analytical and experimental effort to define the margins between design basis loss-of-coolant accident (LOCA) behavior and regulatory limits on maximum fuel rod cladding temperature and deformation. As a result, there is extensive documentation on the modeling of fuel rod behavior in test reactors and design basis LOCA's. However, modeling of that behavior using representative, non-conservative, operating histories is not nearly as well documented in the public literature. Therefore, the objective of this paper is (a) to present calculations of LOCA induced behavior for Pressurized Water Reactor (PWR) core representative fuel rods, and (b) to discuss the variability in those calculations given the variability in fuel rod condition at the initiation of the LOCA. This analysis was limited to the study of changes in fuel rod behavior due to different power operating histories. The other two important parameters which affect that behavior, initial fuel rod design and LOCA coolant conditions were held invarient for all of the representative rods analyzed.

  15. Quality Assurance Review of ISOTOPE and ORIGEN Decay Masses for PWR Fuel (51 GWd/MTU)

    SciTech Connect (OSTI)

    Gastelum, Jason A.

    2011-03-28

    This memorandum documents the comparison of ISOTOPE decay mass calculations for PWR 51GW fuel with analogous calculations in ORIGEN.

  16. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide...

    Office of Scientific and Technical Information (OSTI)

    the CONFU assembly exhibits negative reactivity feedback coefficients comparable in ... NUCLEAR FUELS; PWR TYPE REACTORS; REACTIVITY COEFFICIENTS; REPROCESSING; SIMULATION; ...

  17. Swing-Down of 21-PWR Waste Package

    SciTech Connect (OSTI)

    A.K. Scheider

    2001-05-04

    The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design.

  18. Leak before break application in French PWR plants under operation

    SciTech Connect (OSTI)

    Faidy, C.

    1997-04-01

    Practical applications of the leak-before break concept are presently limited in French Pressurized Water Reactors (PWR) compared to Fast Breeder Reactors. Neithertheless, different fracture mechanic demonstrations have been done on different primary, auxiliary and secondary PWR piping systems based on similar requirements that the American NUREG 1061 specifications. The consequences of the success in different demonstrations are still in discussion to be included in the global safety assessment of the plants, such as the consequences on in-service inspections, leak detection systems, support optimization,.... A large research and development program, realized in different co-operative agreements, completes the general approach.

  19. Analysis of PWR RCS Injection Strategy During Severe Accident

    SciTech Connect (OSTI)

    Wang, S.-J. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, K.-S. [Institute of Nuclear Energy Research, Taiwan (China); Chiang, S.-C. [Taiwan Power Company, Taiwan (China)

    2004-05-15

    Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plant (NPP) with large, dry containment. The severe accident management guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG.The purpose of this work is to analyze the RCS injection strategy of PWR system in an overheated core condition. Power is assumed recovered as the vessel water level drops to the bottom of active fuel. The Modular Accident Analysis Program version 4.0.4 (MAAP4) code is chosen as a tool for analysis. A postulated station blackout sequence for Maanshan NPP is cited as a reference case for this analysis. The hot leg creep rupture occurs during the mitigation action with immediate injection after power recovery according to WOG SAMG, which is not desired. This phenomenon is not considered while developing the WOG SAMG. Two other RCS injection methods are analyzed by using MAAP4. The RCS injection strategy is modified in the Maanshan SAMG. These results can be applied for typical PWR NPPs.

  20. Robotic inspection of PWR coolant pump casing welds

    SciTech Connect (OSTI)

    Pratt, W.R.; Alford, J.W.; Davis, J.B.

    1997-12-01

    As of January 1, 1995, the Swedish Nuclear Inspectorate began requiring more thorough inspections of cast stainless-steel components in nuclear power plants, including pressurized water reactor (PWR) reactor coolant pump (RCP) casings. The examination requirements are established by fracture mechanics analyses of component weldments and demonstrated test system detection capabilities. This may include full volumetric inspection or some portion thereof. Ringhals station is a four-unit nuclear power plant, owned and operated by the Swedish State Power Board, Vattenfall. Unit 1 is a boiling water reactor. Units 2, 3, and 4 are Westinghouse-designed PWRs, ranging in size from 795 to 925 MW. The RCP casings at the PWR units are made of cast stainless steel and contain four circumferential welds that require inspection. Due to the thickness of the casings at the weld locations and configuration and surface conditions on the outside diameter of the casings, remote inspection from the inside diameter of the pump casing was mandated.

  1. Chemical behavior of fission products in the ORNL fission product release program. Supplement. [PWR; BWR

    SciTech Connect (OSTI)

    Collins, J.L.; Osborne, M.F.; Lorenz, R.A.

    1983-01-01

    Tests data are presented for BWR and PWR rods in test HI-4 and test HI-5. Operating conditions fission product release data are included.

  2. Design study of long-life PWR using thorium cycle

    SciTech Connect (OSTI)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-06

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that {sup 231}Pa better than {sup 237}Np as burnable poisons in thorium fuel system. Thorium oxide system with 8%{sup 233}U enrichment and 7.6{approx} 8%{sup 231}Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1%{Delta}k/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53%{Delta}k/k and reduced power peaking during its operation.

  3. Lakes_Elec_You

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Lakes, Electricity & You Why It's So Important That Lakes Are Used To Generate Electricity Why We Can Thank Our Lakes For Electricity Because lakes were made to generate electricity. Back in the mid-1940s, Congress recognized the need for better flood control and navigation. To pay for these services, Congress passed laws that started the building of federal hydroelectric dams, and sold the power from the dams under long-term contracts. Today these dams provide efficient, environmentally

  4. Florida Nuclear Profile - Crystal River

    U.S. Energy Information Administration (EIA) (indexed site)

    Crystal River1" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration date" 3,860,0,"--","PWR","application/vnd.ms-excel","application/vnd.ms-excel" ,860,0,"--" "Data for 2010" "1 Unit was offline in 2010 for repairs." "-- Not applicable.

  5. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    SciTech Connect (OSTI)

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These may be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section, similar in

  6. ACHILLES: Heat Transfer in PWR Core During LOCA Reflood Phase

    Energy Science and Technology Software Center (OSTI)

    2013-11-01

    1. NAME AND TITLE OF DATA LIBRARY ACHILLES -Heat Transfer in PWR Core During LOCA Reflood Phase. 2. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS N/A 3. CONTRIBUTOR AEA Technology, Winfrith Technology Centre, Dorchester DT2 8DH United Kingdom through the OECD Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France. 4. DESCRIPTION OF TEST FACILITY The most important features of the Achilles rig were the shroud vessel, which contained the test section, and the downcomer. These maymore » be thought of as representing the core barrel and the annular downcomer in the reactor pressure vessel. The test section comprises a cluster of 69 rods in a square array within a circular shroud vessel. The rod diameter and pitch (9.5 mm and 12.6 mm) were typical of PWR dimensions. The internal diameter of the shroud vessel was 128 mm. Each rod was electrically heated over a length of 3.66 m, which is typical of the nuclear heated length in a PWR fuel rod, and each contained 6 internal thermocouples. These were arranged in one of 8 groupings which concentrated the thermocouples in different axial zones. The spacer grids were at prototypic PWR locations. Each grid had two thermocouples attached to its trailing edge at radial locations. The axial power profile along the rods was an 11 step approximation to a "chopped cosine". The shroud vessel had 5 heating zones whose power could be independently controlled. 5. DESCRIPTION OF TESTS The Achilles experiments investigated the heat transfer in the core of a Pressurized Water Reactor during the re-flood phase of a postulated large break loss of coolant accident. The results provided data to validate codes and to improve modeling. Different types of experiments were carried out which included single phase cooling, re-flood under low flow conditions, level swell and re-flood under high flow conditions. Three series of experiments were performed. The first and the third used the same test section but the second used another test section

  7. 21-PWR WASTE PACKAGE WITH ABSORBER PLATES LOADING CURVE EVALUATION

    SciTech Connect (OSTI)

    J.M. Scaglione

    2004-12-17

    The objective of this calculation is to evaluate the required minimum burnup as a function of initial pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This calculation is an example of the application of the methodology presented in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). The scope of this calculation covers a range of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the loading curve determination. This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel applications. The intended use of these results will be in establishing PWR waste package configuration loading specifications. Limitations of this evaluation are as follows: (1) The results are based on burnup credit for actinides and selected fission products as proposed in YMP (2003, Table 3-1) and referred to as the ''Principal Isotopes''. Any change to the isotope listing will have a direct impact on the results of this report. (2) The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste package is defined, then this report would need to be reevaluated based on the alternative materials. This calculation is subject to the ''Quality Assurance Requirements and Description'' (QARD) (DOE 2004) because it concerns engineered barriers that are included in the ''Q-List'' (BSC 2004k, Appendix A) as items important to safety and waste isolation.

  8. Comparison of PWR-IMF and FR fuel cycles

    SciTech Connect (OSTI)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj |; Necas, Vladimir

    2007-07-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  9. CASL - PWR Reactor Vessel Multi-Physics CFD Model

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    PWR Reactor Vessel Multi-Physics CFD Model Jin Yan*1, Yiban Xu1, Andrew Petrarca1, Zeses Karoutas1, Emre Tatli1, Emilio Baglietto2, Jess Gehin3 1Westinghouse Electric Company LLC 2Massachusetts Institute of Technology 3Oak Ridge National Lab *Correspondence to: yan3j@westinghouse.com A complete 3D SolidWorks CAD model of Watts Bar Unit 1 was constructed based on drawings. A single fuel assembly CAD model including all geometrical details was created based on the Westinghouse V5H 17x17 fuel

  10. Waterside corrosion of Zircaloy fuel rods. Final report. [PWR

    SciTech Connect (OSTI)

    Garzarolli, F.; Jung, W.; Schoenfeld, H.; Garde, A.M.; Parry, G.W.; Smerd, P.G.

    1982-12-01

    There is an economic incentive to extend average fuel-rod-discharge burnup to about 50 GWd/t. For these higher burnups it is necessary to know if increased waterside corrosion of the cladding will influence fuel-rod performance. For this reason, EPRI sponsored a joint program with C-E and KWU with the objective of investigating PWR waterside corrosion. This final report presents and discusses the results of various subtasks that comprised this project. In the review of corrosion data and models in the literature it was concluded that the PWR environment enhances the corrosion rate by about three times that expected from ex-reactor tests. A large number of fuel rods were characterized in both spent-fuel-pool and hot-cell campaigns. Chemical, physical and microstructural attributes of irradiated and unirradiated oxide films were measured. These included determinations of chemical composition, crystal structure, microstructure, density, specific heat, thermal conductivity, and post-irradiation autoclave corrosion behavior. Procedures used to calculate the fuel-rod surface temperature were reviewed. A model has been developed to predict in-reactor corrosion behavior.

  11. Westinghouse VANTAGE+ fuel assembly to meet future PWR operating requirements

    SciTech Connect (OSTI)

    Doshi, P.K.; Chapin, D.L.; Scherpereel, L.R.

    1988-01-01

    Many utilities operating pressurized water reactors (PWRs) are implementing longer reload cycles. Westinghouse is addressing this trend with fuel products that increase fuel utilization through higher discharge burnups. Higher burnup helps to offset added enriched uranium costs necessary to enable the higher energy output of longer cycles. Current fuel products have burnup capabilities in the area of 40,000 MWd/tonne U or more. There are three main phenomena that must be addressed to achieve even higher burnup levels: accelerated cladding, waterside corrosion, and hydriding; increased fission gas production; and fuel rod growth. Long cycle lengths also require efficient burnable absorbers to control the excess reactivity associated with increased fuel enrichment while maintaining a low residual absorber penalty at the end of cycle. Westinghouse VANTAGE + PWR fuel incorporates features intended to enhance fuel performance at very high burnups, including advances in the three basic elements of the fuel assembly: fuel cladding, fuel rod, and fuel assembly skeleton. ZIRLO {sup TM} cladding, an advanced Zircaloy cladding that contains niobium, offers a significant improvement in corrosion resistance relative to Zircaloy-4. Another important Westinghouse PWR fuel feature that facilitates long cycles is the zirconium diboride integral fuel burnable absorber (ZrB{sub 2}IFBA).

  12. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    V. Delabrosse

    2003-02-27

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  13. 21-PWR Waste Package Side and End Impacts

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-29

    The objective of this calculation is to determine the structural response of a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities and initial angles between the waste package and the unyielding surface is studied. The scope of this calculation is limited to estimating the area of the outer shell (OS) where the residual stress exceeds a given limit (hereafter ''damaged area''). The stress limit is defined as a fraction of the yield strength of the OS material, Alloy 22 (SB-575 N06022), at the appropriate temperature. The design of the 21-PWR waste package used in this calculation is that defined in Reference 8. However, a value of 4 mm was used for the gap between the inner shell and the OS, and the thickness of the OS was reduced by 2 mm. The sketch in Attachment I provides additional information not included in Reference 8. All obtained results are valid for this design only. This calculation is associated with the waste package design and was performed by the Specialty Analyses and Waste Package Design Section. The waste package (i.e. uncanistered spent nuclear fuel disposal container) is classified as Quality Level 1.

  14. Advanced Integral-type Small-size PWR - SMART

    SciTech Connect (OSTI)

    Chang, Moon H.; Zee, Sung Q.; Kim, Keung K.; Kim, Si-Hwan

    2004-07-01

    SMART - an advanced integral-type small-size PWR has been developed for the dual purpose applications of seawater desalination and small-scale power generation. SMART, rated thermal power of 330 MW, adopts various new and innovative design features along with proven PWR technologies, and the non-site-specific basic design has been completed. Highly enhanced safety and reliability, improved performance, simplicity, and modularization are those of strongly emphasized design philosophies which were applied to the system design. The inherent safety improving design and passive safety design features are the key safety design features uniquely characterizing SMART compared to the conventional loop-type reactors. Various transients and accidents were analyzed for the SMART basic design, and the results confirm that safety is assured with margins in any postulated transient and accident. Further, the level-1 full power PSA shows that the core damage frequency is about two orders of magnitude less than those of the existing conventional PWRs. Based on the safety and reliability analyses, it is found that design optimization will further improve the level of safety. Various design verification efforts have been carried out and others are currently underway. (authors)

  15. VERA Core Simulator Methodology for PWR Cycle Depletion

    SciTech Connect (OSTI)

    Kochunas, Brendan; Collins, Benjamin S; Jabaay, Daniel; Kim, Kang Seog; Graham, Aaron; Stimpson, Shane; Wieselquist, William A; Clarno, Kevin T; Palmtag, Scott; Downar, Thomas; Gehin, Jess C

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  16. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    SciTech Connect (OSTI)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  17. PWR loss of feedwater ATWS: analysis and sensitivity study

    SciTech Connect (OSTI)

    Shier, W.G.; Lu, M.S.; Levine, M.M.; Diamond, D.J.

    1983-01-01

    The incident at the Salem Nuclear plant has presented a renewed interest in the analysis of the consequences of anticipated transients without scram (ATWS). This paper presents the results of an analysis of a complete loss of feedwater ATWS for a typical 4-loop PWR. The loss of feedwater transient was selected since previous analyses have shown that this transient produces one of the more limiting overpressure conditions in the primary system. These results provide a detailed analysis of this transient using current analytical techniques and show the sensitivity to several important parameters and plant modeling techniques. The RELAP5/MOD1 computer code has been used for this analysis. The code version is designated as Cycle 13 with additional modifications provided by both INEL and BNL.

  18. Design study of long-life PWR using thorium cycle (Journal Article...

    Office of Scientific and Technical Information (OSTI)

    life PWR core because it gives reactivity swing less than 1%Deltakk and longer ... long time operation with reduced excess reactivity as low as 0.53%Deltakk and reduced ...

  19. Conceptual design study of small long-life PWR based on thorium...

    Office of Scientific and Technical Information (OSTI)

    The optimization of 350 MWt small long life PWR result small excess reactivity and reduced ... on advances in nuclear science and engineering, Denpasar, Bali (Indonesia), 16-19 Sep ...

  20. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGES-Beta [OSTI]

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  1. Study of a Station Blackout Event in the PWR Plant

    SciTech Connect (OSTI)

    Ching-Hui Wu; Tsu-Jen Lin; Tsu-Mu Kao [Institute of Nuclear Energy Research P.O. Box 3-3, Longtan, 32500, Taiwan (China)

    2002-07-01

    On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station Blackout (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies. (authors)

  2. Identification and evaluation of PWR in-vessel severe accident management strategies

    SciTech Connect (OSTI)

    Dukelow, J S [Pacific Northwest Lab., Richland, WA (United States); Harrison, D G [Jason Associates, Idaho Falls, ID (United States); Morgenstern, M [Battelle Human Affairs Research Center, Seattle, WA (United States)

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents.

  3. Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

    SciTech Connect (OSTI)

    Rempe, J. L.; Knudson, D. L.; Lutz, R. J.

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  4. Analysis of Potential Hydrogen Risk in the PWR Containment

    SciTech Connect (OSTI)

    Deng Jian; Xuewu Cao [Shanghai Jiaotong University, Shanghai (China)

    2006-07-01

    Various studies have shown that hydrogen combustion is one of major risk contributors to threaten the integrity of the containment in a nuclear power plant. That hydrogen risk should be considered in severe accident strategies in current and future NPPs has been emphasized in the latest policies issued by the National Nuclear Safety Administration of China (NNSA). According to a deterministic approach, three typical severe accident sequences for a PWR large dry containment, such as the large break loss-of-coolant (LLOCA), the station blackout (SBO), and the small break loss-of-coolant (SLOCA) are analyzed in this paper with MELCOR code. Hydrogen concentrations in different compartments are observed to evaluate the potential hydrogen risk. The results show that there is a great amount of hydrogen released into the containment, which causes the containment pressure to increase and some potential in-consecutive burning. Therefore, certain hydrogen management strategies should be considered to reduce the risk to threaten the containment integrity. (authors)

  5. Assessment of PWR waterside corrosion models and data. Final report

    SciTech Connect (OSTI)

    Cox, B.

    1985-10-01

    The published data on waterside corrosion of PWR fuel cladding and unfuelled components have been reviewed, and the models used to assess the data have been studied. All corrosion models use too simplified a view of the corrosion process to obtain other than a general trend for the actual oxidation data. The in-reactor post-transition oxidation of the Zircaloys appears to be heavily dependent on water chemistry variations both between reactors, and along the length of an individual fuel rod. Crud deposition may be one primary cause of this, perhaps by allowing the independent development of the water chemistry within the crud layer, as much as by its effect on cladding surface temperatures. However, the effect of the thickening of the oxide film, which permits the development of an independent water chemistry inside the oxide, leading to an accelerating oxidation rate at large oxide thicknesses, seems to be the most important factor. It is concluded that a spectrum of results ranging from essentially no in-reactor enhancement of the oxidation rate to a sizeable enhancement (>10) may be seen depending upon the thickness of the oxide films, the water chemistry of the reactor, and crud deposition. A post-irradiation test that may help to distinguish between the factors involved has been suggested. 105 refs., 38 figs.

  6. Corrosion fatigue characterization of reactor pressure vessel steels. [PWR; BWR

    SciTech Connect (OSTI)

    Van Der Sluys, W.A.

    1982-12-01

    During routine operation, light water reactor (LWR) pressure vessels are subjected to a variety of transients that result in time-varying stresses. Consequently, fatigue and environmentally-assisted fatigue are mechanisms of growth relevant to flaws in these pressure vessels. To provide a better understanding of the resistance of nuclear pressure vessel steels to these flaw growth processes, fracture mechanics data were generated on the rates of fatigue crack growth for SA508-2 and SA533B-1 steels in both room temperature air and 288/sup 0/C water. Areas investigated were: the relationship of crack growth rate to prior loading history; the effects of loading frequency and R ratio (K/sub min//K/sub max/) on crack growth rate as a function of the stress intensity factor range (..delta..K); transient aspects of the fatigue crack growth behavior; the effect of material chemistry (sulphur content) on fatigue crack; and growth rate; water chemistry effects (high-purity water versus simulated pressurized water reactotr (PWR) primary coolant).

  7. Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

    SciTech Connect (OSTI)

    Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G; Salko, Robert K; Evans, Thomas M; Turner, John A; Belcourt, Kenneth; Hooper, Russell; Schmidt, Rodney

    2014-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.

  8. Study on Equilibrium Characteristics of Thorium-Plutonium-Minor Actinides Mixed Oxides Fuel in PWR

    SciTech Connect (OSTI)

    Waris, A.; Permana, S.; Kurniadi, R.; Su'ud, Z.; Sekimoto, H.

    2010-06-22

    A study on characteristics of thorium-plutonium-minor actinides utilization in the pressurized water reactor (PWR) with the equilibrium burnup model has been conducted. For a comprehensive evaluation, several fuel cycles scenario have been included in the present study with the variation of moderator-to-fuel volume ratio (MFR) of PWR core design. The results obviously exhibit that the neutron spectra grow to be harder with decreasing of the MFR. Moreover, the neutron spectra also turn into harder with the rising number of confined heavy nuclides. The required {sup 233}U concentration for criticality of reactor augments with the increasing of MFR for all heavy nuclides confinement and thorium and uranium confinement in PWR.

  9. Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

    SciTech Connect (OSTI)

    Putney, J.M.; Preece, R.J.

    1993-06-01

    An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3.

  10. In-core and ex-core calculations of the VENUS simulated PWR benchmark experiment

    SciTech Connect (OSTI)

    Williams, M.L.; Chowdhury, P.; Landesman, M.; Kam, F.B.K.

    1985-01-01

    The VENUS PWR engineering mockup experiment was established to simulate a beginning-of-life, generic PWR configuration at the zero-power VENUS critical facility located at CEN/SCK, Mol, Belgium. The experimental measurement program consists of (1) gamma scans to determine the core power distribution, (2) in-core and ex-core foil activations, (3) neutron spectrometer measurements, and (4) gamma heating measurements with TLD's. Analysis of the VENUS benchmark has been performed with two-dimensional discrete ordinates transport theory, using the DOT-IV code.

  11. Effects of Lower Drying-Storage Temperatures on the DBTT of High Burnup PWR

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Cladding | Department of Energy Effects of Lower Drying-Storage Temperatures on the DBTT of High Burnup PWR Cladding Effects of Lower Drying-Storage Temperatures on the DBTT of High Burnup PWR Cladding The purpose of the research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor cladding alloys during cooling for a range of storage temperatures and hoop

  12. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented. 6 references, 4 figures.

  13. Radionuclide release from PWR spent fuel specimens with induced cladding defects

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1984-03-01

    Radionuclide releases from pressurized water reactor (PWR) spent fuel rod specimens containing various artificially induced cladding defects were compared by leach testing. The study was conducted in support of the Nevada Nuclear Waste Storage Investigations (NNWSI) Waste Package Task to evaluate the effectiveness of failed cladding as a barrier to radionuclide release. Test description and results are presented.

  14. Performance evaluation of two-stage fuel cycle from SFR to PWR

    SciTech Connect (OSTI)

    Fei, T.; Hoffman, E.A.; Kim, T.K.; Taiwo, T.A.

    2013-07-01

    One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with an average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)

  15. Proceedings: 1983 Workshop on Secondary-Side Stress Corrosion Cracking and Intergranular Corrosion of PWR Steam Generator Tubing

    SciTech Connect (OSTI)

    1986-03-01

    Participants in this international workshop discussed research investigating mechanisms and propagation rates of intergranular corrosion in PWR steam generators. Laboratory test results, which have been consistent with power plant experience, permitted preliminary definition of corrosion rates in alloy 600 tubing.

  16. DOSE RATES FOR WESTINGHOUSE 17X17 MOX PWR SNF IN A WASTE PACKAGE (SCPB: N/A)

    SciTech Connect (OSTI)

    T.L. Lotz

    1997-01-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to estimate the dose rate on and near the surface a Multi-Purpose Canister (MPC) PWR waste package (WP) which is loaded with Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel. The 21 PWR MPC WP is used to provide an upper bound for waste package designs since the 12 PWR MPC WP will have a smaller source term and an equivalent amount of shielding. the objectives of this evaluation are to calculate the requested dose rate(s) and document the calculation in a fashion to allow comparisons to other waste forms and WP designs at a future time.

  17. Conceptual design study of small long-life PWR based on thorium cycle fuel

    SciTech Connect (OSTI)

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-09-30

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of {sup 233}U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation.

  18. Optimization of small long-life PWR based on thorium fuel

    SciTech Connect (OSTI)

    Subkhi, Moh Nurul; Suud, Zaki Waris, Abdul; Permana, Sidik

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  19. Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3. International Agreeement Report

    SciTech Connect (OSTI)

    Preece, R.J.; Putney, J.M.

    1993-07-01

    A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser.

  20. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  1. PCI-related cladding failures during off-normal events - draft. [PWR; BWR

    SciTech Connect (OSTI)

    Van Houten, R.; Tokar, M.; MacDonald, P.E.

    1984-05-01

    Pellet-cladding interaction (PCI) has long been identified as a fuel rod failure mechanism during power increases in both pressurized and boiling water reactors, and commercial guidelines have practically eliminated such failures during standard operations. A question remains regarding the possible formation of through-wall cladding cracks during several types of postulated off-normal reactor events involving power increases. This report includes preliminary findings for reactor events of the type addressed by Chapter 15 of the NRC Standard Review Plan. Specifically, the BWR turbine trip without bypass, PWR control rod withdrawal error, subcritical PWR control rod withdrawal error, BWR control blade withdrawal error, and the PWR steamline break are analyzed on the joint bases of peak rod power, power increase, ramp rate, and duration at elevated power. These Chapter 15 events are compared to numerous test reactor results and to other relevant investigations, and tentative conclusions on transient severity and data base adequacy are presented. Progress in developing computer codes for predicting PCI-induced fuel rod failures is also discussed. 49 references.

  2. A Study on the Conceptual Design of a 1,500 MWe Passive PWR with Annular Fuel

    SciTech Connect (OSTI)

    Kwi Lim Lee; Soon Heung Chang

    2004-07-01

    In this study, the preliminary conceptual design of a 1500 MWe pressurized water reactor (PWR) with annular fuel has been performed. This design is derived from the AP1000 which is a 1000 MWe PWR with two-loop. However, the present design is a 1500 MWe PWR with three-loop, passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. The preliminary design parameters of this reactor have been determined through simple relation to those of AP1000 for reactor, reactor coolant system, and passive safety injection system. Using the MATRA code, we analyze the core designs for two alternatives on fuel assembly types: solid fuel and annular fuel. The performance of reactor cooling systems is evaluated through the accident of the cold leg break in the core makeup tank loop by using MARS2.1 code. This study presents the developmental strategy, preliminary design parameters and safety analysis results. (authors)

  3. Analysis of a double-ended cold-leg break simulation: THTF Test 3. 05. 5B. [PWR

    SciTech Connect (OSTI)

    Craddick, W.G.; Pevey, R.E.

    1982-09-01

    On July 3, 1980, an experiment was performed in the Oak Ridge National Laboratory Thermal-Hydraulic Test Facility that simulated a double-ended cold-leg break pressurized-water reactor (PWR) accident. Analysis of the experiment revealed that nuclear fuel rods exposed to the same hydrodynamic environment as that which existed in the experiment would have departed from nucleate boiling both earlier and later than the fuel rod simulator (FRS), depending on the size of the gap between the nuclear fuel pellets and cladding and on the initial power of the nuclear fuel rod. Comparison of the results of the current experiment, which used an FRS bundle with geometry similar to 17 x 17 PWR fuel assemblies, to the results of earlier experiments, which used an FRS bundle with geometry similar to 15 x 15 PWR fuel assemblies, revealed no differences that can be attributed to the difference in geometries.

  4. Thermal Response of the 21-PWR Waste Package to a Fire Accident

    SciTech Connect (OSTI)

    F.P. Faucher; H. Marr; M.J. Anderson

    2000-10-03

    The objective of this calculation is to evaluate the thermal response of the 21-PWR WP (pressurized water reactor waste package) to the regulatory fire event. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation (Attachment IV) is that of the potential design of the type of waste package considered in this calculation. The procedure AP-3.12Q.Calculations (Reference 1), and the Development Plan (Reference 24) are used to develop this calculation.

  5. Neutronics and safety characteristics of a 100% MOX fueled PWR using weapons grade plutonium

    SciTech Connect (OSTI)

    Biswas, D.; Rathbun, R.; Lee, Si Young; Rosenthal, P.

    1993-12-31

    Preliminary neutronics and safety studies, pertaining to the feasibility of using 100% weapons grade mixed-oxide (MOX) fuel in an advanced PWR Westinghouse design are presented in this paper. The preliminary results include information on boron concentration, power distribution, reactivity coefficients and xenon and control rode worth for the initial and the equilibrium cycle. Important safety issues related to rod ejection and steam line break accidents and shutdown margin requirements are also discussed. No significant change from the commercial design is needed to denature weapons-grade plutonium under the current safety and licensing criteria.

  6. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  7. The unsteady 2-D numerical analysis for thermal stratification in surge line of PWR

    SciTech Connect (OSTI)

    Youm, H.K.; Park, M.H.; Jin, T.E.

    1995-12-01

    In this paper, the unsteady 2-dimensional model for thermal stratification in a pressurizer surge line of PWR plant is proposed to numerically investigate, the heat transfer and flow characteristics. The dimensionless governing equations are solved by using the SIMPLE (Semi-Implicit Method for Pressure Linked Equations) algorithm. The results are compared with simulated experimental results of TEMR Test. The time-dependent temperature profiles in the fluid and pipe wall are, shown with the thermal stratification occurring in the horizontal section of the pipe. The corresponding thermal stresses are, also presented.

  8. Steam-generator chemical-cleaning Demonstration Test No. 3 in a pot boiler. [PWR

    SciTech Connect (OSTI)

    Fink, G.C.; Helyer, M.H.; Key, G.L.

    1983-04-01

    Steam generators in pressurized water reactor (PWR) plants have experienced tubing degradation and support structure damage by a variety of corrosion mechanisms related to the accumulation of secondary side corrosion products. The Steam Generator Owners Group (SGOG) and the Electric Power Research Institute (EPRI) have sponsored a program to develop a process for the chemical removal of steam generator corrosion product accumulations. In this report, the contractor describes the results of a pot boiler demonstration test of the SGOG/EPRI Mark III Chemical Cleaning process.

  9. Steam-generator chemical-cleaning process development. Final report. [PWR

    SciTech Connect (OSTI)

    Schneidmiller, D.; Stiteler, D.

    1983-04-01

    As a result of work sponsored by the Steam Generator Owners Group (SGOG) and managed by the Electric Power Research Institute (EPRI), a process for chemical removal of iron- and copper-bearing sludges and tube-to-support plate crevice corrosion product deposits from the secondary side of pressurized water reactor (PWR) steam generators has been developed. The process has undergone extensive pilot-scale testing and has shown to be effective for the removal of both synthetic and actual steam generator corrosion product deposits. This report documents the results of UNC Nuclear Industries' participation in the SGOG chemical cleaning development program.

  10. VERA Modeling and Simulation of the AP1000 PWR Cycle 1 Depletion

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    CASL-U-2015-0302-000 VERA Modeling and Simulation of the AP1000 PWR Cycle 1 Depletion L3:VMA.AMA.P11.06 David Salazar, Westinghouse Fausto Franceschini, Westinghouse September 30, 2015 L3:VMA.AMA.P11.06 Official Use Only ii Protected under CASL Master NDA CASL-U-2015-0302-000 REVISION LOG Revision Date Affected Pages Revision Description 0 09/30/2015 All Initial issuance Document pages that are: Export Controlled ____________No______________________________________ IP/Proprietary/NDA

  11. Application of LBB to high energy piping systems in operating PWR

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  12. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  13. Application of the MELCOR code to design basis PWR large dry containment analysis.

    SciTech Connect (OSTI)

    Phillips, Jesse; Notafrancesco, Allen; Tills, Jack Lee

    2009-05-01

    The MELCOR computer code has been developed by Sandia National Laboratories under USNRC sponsorship to provide capability for independently auditing analyses submitted by reactor manufactures and utilities. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. To assess the adequacy of containment thermal-hydraulic modeling incorporated in the MELCOR code for application to PWR large dry containments, several selected demonstration designs were analyzed. This report documents MELCOR code demonstration calculations performed for postulated design basis accident (DBA) analysis (LOCA and MSLB) inside containment, which are compared to other code results. The key processes when analyzing the containment loads inside PWR large dry containments are (1) expansion and transport of high mass/energy releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment pressure reduction due to engineered safety features. A code-to-code benchmarking for DBA events showed that MELCOR predictions of maximum containment loads were equivalent to similar predictions using a qualified containment code known as CONTAIN. This equivalency was found to apply for both single- and multi-cell containment models.

  14. Study of a transient identification system using a neural network for a PWR plant

    SciTech Connect (OSTI)

    Ishihara, Yoshinao; Kasai, Masao; Kambara, Masayuki [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Mitsuda, Hiromichi; Kurata, Toshikazu; Shirosaki, Hidekazu [Inst. of Nuclear Safety System, Inc., Kyoto (Japan)

    1996-08-01

    This paper presents the procedure and results of a system for identifying PWR plant abnormal events, which uses neural network techniques. The neural network recognizes the abnormal event from the patterns of the transient changes of analog data from plant parameters when they deport from their normal state. For the identification of abnormal events in this study, events that cause a reactor to scram during power operation were selected as the design base events. The test data were prepared by simulating the transients on a compact PWR simulator. The simulation data were analyzed to determine how the plant parameters respond after the occurrence of a transient. A method of converting the pattern of the transient changes into characteristic parameters by fitting the data to pre-determined functions was developed. These characteristic parameters were used as the input data to the neural network. The neural network learning procedure used a generalized delta rule, namely a back-propagation algorithm. The neural network can identify the type of an abnormal event from a limited set of events by using these characteristic parameters obtained from the pattern of the changes in the analog data. From the results of this application of a neural network, it was concluded that it would be possible to use the method to identify abnormal events in a nuclear power plant.

  15. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    SciTech Connect (OSTI)

    Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F.

    2013-07-01

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  16. Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

    SciTech Connect (OSTI)

    Wagner, J.C.

    2002-10-23

    This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

  17. Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5

    Energy Science and Technology Software Center (OSTI)

    1999-06-02

    CONTEMPT4/MOD6 describes the response of multicompartment containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program can accommodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures, and mass and energy inventories due to inter-compartment mass and energy exchange taking into account user-supplied descriptions of compartments,more » inter-compartment junction flow areas, LOCA source terms, and user-selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers, and BWR pressure suppression systems. CONTEMPT4/MOD6 also provides analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation to accommodate degraded core type accidents.« less

  18. Phenomenon analysis of stress corrosion cracking in the vessel head penetrations of French PWR`s

    SciTech Connect (OSTI)

    Pichon, C.; Buisine, D.; Faidy, C.; Gelpi, A.; Vaindirlis, M.

    1995-12-31

    During a hydrotest in 1991, a leak was detected on,a reactor vessel head (RVH) penetration of a French PWR. This leak was due to a phenomenon of Primary Water Stress Corrosion Cracking (PWSCC) affecting these penetrations in Alloy 600. The destructive and non-destructive examinations undertaken during the following months highlighted the generic nature of the degradations. In order to well understand this phenomenon and implement the most suitable maintenance policy, a large scale scientific program was decided and performed jointly by Electricite de France and FRAMATOME. The paper will present all the results obtained in this program concerning the parameters governing the PWSCC. In particular the following fields will be developed: (1) the material, its microstructure in line with the manufacturing and its susceptibility to PWSCC; (2) the stresses and their evaluations by measurements, mock up corrosion tests and Finite Element Analysis (FEA); (3) the effect of surface finish on crack initiation; and (4) the crack growth rate. This phenomenon analysis will be useful for evaluating the risk of PWSCC on other Alloy 600 areas in PWR`s primary system.

  19. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    V. DeLa Brosse

    2003-03-27

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  20. Maximim Accelerations On The Fuel Assemblies Of a 21-PWR Waste Package During End Impacts 

    SciTech Connect (OSTI)

    T. Schmitt

    2005-08-17

    The objective of this calculation is to determine the acceleration of the fuel assemblies contained in a 21-Pressurized Water Reactor (PWR) spent nuclear fuel waste package impacting an unyielding surface. A range of initial velocities of the waste package is studied. The scope of this calculation is limited to estimating the acceleration of the fuel assemblies during the impact.

  1. Development of a model for predicting intergranular stress corrosion cracking of Alloy 600 tubes in PWR primary water. Final report

    SciTech Connect (OSTI)

    Garud, Y.S.

    1985-01-01

    A preliminary mathematical model developed in this study may make it possible to predict stress corrosion cracking on the primary side of PWR steam generator tubing. The study outlines a comprehensive testing program that will provide the operational and experimental data to further develop and verify the model.

  2. Grid-to-rod flow-induced impact study for PWR fuel in reactor

    DOE PAGES-Beta [OSTI]

    Jiang, Hao; Qu, Jun; Lu, Roger Y.; Wang, Jy-An John

    2016-06-10

    The source for grid-to-rod fretting in a pressurized water nuclear reactor (PWR) is the dynamic contact impact from hydraulic flow-induced fuel assembly vibration. In order to support grid-to-rod fretting wear mitigation research, finite element analysis (FEA) was used to evaluate the hydraulic flow-induced impact intensity between the fuel rods and the spacer grids. Three-dimensional FEA models, with detailed geometries of the dimple and spring of the actual spacer grids along with fuel rods, were developed for flow impact simulation. The grid-to-rod dynamic impact simulation provided insights of the contact phenomena at grid-rod interface. Finally, it is an essential and effectivemore » way to evaluate contact forces and provide guidance for simulative bench fretting-impact tests.« less

  3. Decay Heat of Major Radionuclides for PWR Spent Fuels to 10,000 Years

    SciTech Connect (OSTI)

    J.S. Tang

    2001-12-20

    The objective of this calculation is to determine decay heat of a pressurized-water reactor (PWR) spent nuclear fuel (SNF) assembly with four different initial-enrichment and burnup characteristics. The major contributing radionuclides to the decay heat are also identified and graphically presented. The scope of this calculation is limited to the time period of the first 10,000 years after discharge from reactors. The results of this calculation will be used to evaluate the effects of the projected commercial spent nuclear fuel (CSNF) inventory on the repository design based on revised nuclear energy forecasts. This calculation was performed in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (BSC (Bechtel SAIC Company) 2001). AP-3.12Q, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the repository design activity.

  4. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    SciTech Connect (OSTI)

    Herer, C.

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  5. On the explanation and calculation of anomalous reflood hydrodynamics in large PWR cores

    SciTech Connect (OSTI)

    Rodriguez, S.E.

    1985-01-01

    Reflood hydrodynamics from large-scale (1:20) test facilities in Japan have yielded apparently anomalous behavior relative to FLECHT tests. Namely, even at reflooding rates below one inch per second, very large liquid volume fractions (10-15%) exist above the quench fronts shortly after flood begins; thus cladding temperature excursions are terminated early in the reflood phase. This paper discusses an explanation for this behavior: liquid films on the core's unheated rods. The experimental findings are shown to be correctly simulated with a new four-field (vapor, films, droplets) version of the best-estimate TRAC-PF1 computer code, TRAC-FF. These experimental and analytical findings have important implications for PWR large-break LOCA licensing.

  6. Effect of coolant chemistry on PWR radiation transport processes. Progress report on reactor loop studies

    SciTech Connect (OSTI)

    Brown, D.J.; Flynn, G.; Haynes, J.W.; Kitt, G.P.; Large, N.R.; Lawson, D.; Mead, A.P.; Nichols, J.L.; Woodwark, D.R.

    1986-05-01

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. There are strong indications that the in-core deposition behavior of corrosion product species is not fully accounted for by the solubility model based on nickel ferrite; boric acid plays a role apart from its influence on pH, and corrosion products are adsorbed to some extent in the zirconium oxide film on the fuel cladding. In DWL, soluble species appear to be dominant in deposition processes. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. 13 figs.

  7. Fuel-rod response during the large-break LOCA Test LOC-6. [PWR

    SciTech Connect (OSTI)

    Vinjamuri, K.; Cook, B.A.; Hobbins, R.R.

    1981-01-01

    The large break Loss of Coolant Accident (LOCA) Test LOC-6 was conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory by EG and G Idaho, Inc. The objectives of the PBF LOCA tests are to obtain in-pile cladding ballooning data under blowdown and reflood conditions and assess how well out-of-pile ballooning data represent in-pile fuel rod behavior. The primary objective of the LOC-6 test was to determine the effects of internal rod pressures and prior irradiation on the deformation behavior of fuel rods that reached cladding temperatures high in the alpha phase of zircaloy. Test LOC-6 was conducted with four rods of PWR 15 x 15 design with the exception of fuel stack length (89 cm) and enrichment (12.5 W% /sup 235/U). Each rod was surrounded by an individual flow shroud.

  8. LOCA rupture strains and coolability of full-length PWR fuel bundles

    SciTech Connect (OSTI)

    Mohr, C.L.; Hesson, G.M.

    1983-03-01

    The LOCA Simulation Program tests sponsored by the United States Nuclear Regulatory Commission are the first full-length nuclear-heated experiments designed to investigate the deformation and rupture characteristics as well as the coolability of nuclear-heated fuel under accident conditions. The results of the seven tests preformed in the program using 32-rod full-length PWR fuel bundles have shown that for a wide range of flow blockage condtions no significant reduction in coolability of the fuel bundle could be found. These results have been confirmed by data from out-of-pile electrically-heated experiments. Although there is a difference between nuclear and electrically-heated test data, the conclusion is still the same. Coolability of a deformed bundle during reflood is dominated by the dispersion of droplets in the deformed zone which provides adequate cooling and which is not reduced by the deformation of the fuel rod cladding.

  9. Source term experiment STEP-3 simulating a PWR severe station blackout

    SciTech Connect (OSTI)

    Simms, R.; Baker, L. Jr.; Ritzman, R.L.

    1987-05-21

    For a severe PWR accident that leads to a loss of feedwater to the steam generators, such as might occur in a station blackout, fission product decay heating will cause a water boiloff. Without effective cooling of the core, steam will begin to oxidize the Zircaloy cladding. The noble gases and volatile fission products, such as Cs and I, that are major contributors to the radiological source term, will be released from the damaged fuel shortly after cladding failure. The accident environment when these volatile fission products escape was simulated in STEP-3 using four fuel elements from the Belgonucleaire BR3 reactor. The primary objective was to examine the releases in samples collected as close to the test zone as possible. In this paper, an analysis of the temperatures and hydrogen generation is compared with the measurements. The analysis is needed to estimate releases and characterize conditions at the source for studies of fission product transport.

  10. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR

    SciTech Connect (OSTI)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed.

  11. IMPACT OF FISSION PRODUCTS IMPURITY ON THE PLUTONIUM CONTENT IN PWR MOX FUELS

    SciTech Connect (OSTI)

    Gilles Youinou; Andrea Alfonsi

    2012-03-01

    This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called PuFP. Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.

  12. Probability and consequences of a rapid boron dilution sequence in a PWR

    SciTech Connect (OSTI)

    Diamond, D.J.; Kohut, P.; Nourbakhsh, H.; Valtonen, K.; Secker, P.

    1995-11-01

    The reactor restart scenario is one of several beyond-design-basis events in a pressurized water reactor (PWR) which can lead to rapid boron dilution in the core. This in turn can lead to a power excursion and the potential for fuel damage. A probabilistic analysis had been done for this event for a European PWR. The estimated core damage frequency was found to be high partially because of a high frequency for a LOOP and assumptions regarding operator actions. As a result, a program of analysis and experiment was initiated and corrective actions were taken. A system was installed so that the suction of the charging pumps would switch to the highly borated refueling water storage tank (RWST) when there was a trip of the RCPs. This was felt to reduce the estimated core damage frequency to an acceptable level. In the US, this original study prompted the Nuclear Regulatory Commission to issue an information notice to follow work being done in this area and to initiate studies such as the work at BNL reported herein. In order to see if the core damage frequency might be as high in US plants, a probabilistic assessment of this scenario was done for three plants. Two important conservative assumptions in this analysis were that (1) the mixing of the injectant was insignificant and (2) fuel damage occurs when the slug passes through the core. In order to study the first assumption, analysis was carried out for two of the plants using a mixing model. The second assumption was studied by calculating the neutronic response of the core to a slug of deborated water for one of the plants. All three types of analyses are summarized below. More information is available in the original report.

  13. The CASTOR-V/21 PWR spent-fuel storage cask: Testing and analyses: Interim report

    SciTech Connect (OSTI)

    Dziadosz, D.; Moore, E.V.; Creer, J.M.; McCann, R.A.; McKinnon, M.A.; Tanner, J.E.; Gilbert, E.R.; Goodman, R.L.; Schoonen, D.H.; Jensen, M.

    1986-11-01

    A performance test of a Gesellschaft fuer Nuklear Service CASTOR-V/21 pressurized water reactor (PWR) spent fuel storage cask was performed. The test was the first of a series of cask performance tests planned under a cooperative agreement between Virginia Power and the US Department of Energy. The performance test consisted of loading the CASTOR-V/21 cask with 21 PWR spent fuel assemblies from Virginia Power's Surry reactor. Cask surface and fuel assembly guide tube temperatures, and cask surface gamma and neutron dose rates were measured. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Limited spent fuel integrity data were also obtained. Results of the performance test indicate the CASTOR-V/21 cask exhibited exceptionally good heat transfer performance which exceeded design expectations. Peak cladding temperatures with helium and nitrogen backfills in a vertical cast orientation and with helium in a horizontal orientation were less than the allowable of 380/sup 0/C with a total cask heat load of 28 kW. Significant convection heat transfer was present in vertical nitrogen and helium test runs as indicated by peak temperatures occurring in the upper regions of the fuel assemblies. Pretest temperature predictions of the HYDRA heat transfer computer program were in good agreement with test data, and post-test predictions agreed exceptionally well (25/sup 0/C) with data. Cask surface gamma and neutron dose rates were measured to be less than the design goal of 200 mrem/h. Localized peaks as high as 163 mrem/h were measured on the side of the cask, but peak dose rates of <75 mrem/h can easily be achieved with minor refinements to the gamma shielding design. From both heat transfer and shielding perspectives, the CASTOR-V/21 cask can, with minor refinements, be effectively implemented at reactor sites and central storage facilities for safe storage of spent fuel.

  14. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect (OSTI)

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  15. Radionuclide release from PWR fuels in a reference tuff repository groundwater subsquently changed to Radionuclide release from PWR fuels in J-13 well water

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-04-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: (1) fuel rod sections split open to expose bare fuel particles; (2) rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; (3) rod sections with water-tight end fittings and two 200-{mu}m diameter holes through the cladding; and (4) undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested in deionized water. Selected initial results are also given for Turkey Point fuel specimens tested in J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water.

  16. Analysis of results from a loss-of-offsite-power-initiated ATWS experiment in the LOFT facility. [PWR

    SciTech Connect (OSTI)

    Varacalle, D.J. Jr.; Koizumi, Y.; Giri, A.H.; Koske, J.E.; Sanchez-Pope, A.E.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by loss-of-offsite power, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a scaled safety relief valve (SRV) representative of those in a commercial PWR, while reactor power was reduced by moderator reactivity feedback in a natural circulation mode. The experiment showed that reactor power decreases more rapidly when the primary pumps are tripped in a loss-of-offsite-power ATWS than in a loss-of-feedwater induced ATWS when the primary pumps are left on. During the experiment, the SRV had sufficient relief capacity to control primary system pressure. Natural circulation was effective in removing core heat at high temperature, pressure, and core power. The system transient response predicted using the RELAP5/MOD1 computer code showed good agreement with the experimental data.

  17. Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT facility. [PWR

    SciTech Connect (OSTI)

    Grush, W.H.; Woerth, S.C.; Koizumi, Y.

    1983-01-01

    An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data.

  18. SAS2H Generated Isotopic Concentrations For B&W 15X15 PWR Assembly (SCPB:N/A)

    SciTech Connect (OSTI)

    J.W. Davis

    1996-08-29

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

  19. Development and Application of Laser Peening System for PWR Power Plants

    SciTech Connect (OSTI)

    Masaki Yoda; Itaru Chida; Satoshi Okada; Makoto Ochiai; Yuji Sano; Naruhiko Mukai; Gaku Komotori; Ryoichi Saeki; Toshimitsu Takagi; Masanori Sugihara; Hirokata Yoriki

    2006-07-01

    Laser peening is a process to improve residual stress from tensile to compressive in surface layer of materials by irradiating high-power laser pulses on the material in water. Toshiba has developed a laser peening system composed of Q-switched Nd:YAG laser oscillators, laser delivery equipment and underwater remote handling equipment. We have applied the system for Japanese operating BWR power plants as a preventive maintenance measure for stress corrosion cracking (SCC) on reactor internals like core shrouds or control rod drive (CRD) penetrations since 1999. As for PWRs, alloy 600 or 182 can be susceptible to primary water stress corrosion cracking (PWSCC), and some cracks or leakages caused by the PWSCC have been discovered on penetrations of reactor vessel heads (RVHs), reactor bottom-mounted instrumentation (BMI) nozzles, and others. Taking measures to meet the unconformity of the RVH penetrations, RVHs themselves have been replaced in many PWRs. On the other hand, it's too time-consuming and expensive to replace BMI nozzles, therefore, any other convenient and less expensive measures are required instead of the replacement. In Toshiba, we carried out various tests for laser-peened nickel base alloys and confirmed the effectiveness of laser peening as a preventive maintenance measure for PWSCC. We have developed a laser peening system for PWRs as well after the one for BWRs, and applied it for BMI nozzles, core deluge line nozzles and primary water inlet nozzles of Ikata Unit 1 and 2 of Shikoku Electric Power Company since 2004, which are Japanese operating PWR power plants. In this system, laser oscillators and control devices were packed into two containers placed on the operating floor inside the reactor containment vessel. Laser pulses were delivered through twin optical fibers and irradiated on two portions in parallel to reduce operation time. For BMI nozzles, we developed a tiny irradiation head for small tubes and we peened the inner surface around J

  20. Examination of spent PWR fuel rods after 15 years in dry storage.

    SciTech Connect (OSTI)

    Einziger, R.E.; Tsai, H.C.; Billone, M.C.; Hilton, B.A.

    2002-02-11

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited prestorage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission gas

  1. Examination of Spent PWR Fuel Rods After 15 Years in Dry Storage

    SciTech Connect (OSTI)

    Einziger, R.E.; Tsai, H.C.; Billone, M.C.; Hilton, B.A.

    2002-07-01

    Virginia Power Surry Nuclear Station Pressurized Water Reactor (PWR) fuel was stored in a dry inert atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory (INEEL) for 15 years at peak cladding temperatures decreasing from about 350 to 150 deg. C. Prior to the storage, the loaded cask was subjected to extensive thermal benchmark tests. The cask was opened to examine the fuel for degradation and to determine if it was suitable for extended storage. No rod breaches had occurred and no visible degradation or crud/oxide spallation were observed. Twelve rods were removed from the center of the T11 assembly and shipped from INEEL to the Argonne-West HFEF for profilometric scans. Four of these rods were punctured to determine the fission gas release from the fuel matrix and internal pressure in the rods. Three of the four rods were cut into five segments each, then shipped to the Argonne-East AGHCF for detailed examination. The test plan calls for metallographic examination of six samples from two of the rods, microhardness and hydrogen content measurements at or near the six metallographic sample locations, tensile testing of six samples from the two rods, and thermal creep testing of eight samples from the two rods to determine the extent of residual creep life. The results from the profilometry (12 rods), gas release measurements (4 rods), metallographic examinations (2 samples from 1 rod), and microhardness and hydrogen content characterization (2 samples from 1 rod) are reported here. The tensile and creep studies are just starting and will be reported at a later date, along with the additional characterization work to be performed. Although only limited pre-storage characterization is available, a number of preliminary conclusions can be drawn based on comparison with characterization of Florida Power Turkey Point rods of a similar vintage. Based on this comparison, it appears that little or no cladding thermal creep and fission

  2. TREAT source-term experiment STEP-1 simulating a PWR LOCA

    SciTech Connect (OSTI)

    Simms, R.; Baker, L. Jr.; Blomquist, C.A.; Ritzman, R.L.

    1986-01-01

    In a hypothetical pressurized water reactor (PWR) large-break loss-of-coolant accident (LOCA) in which the emergency core cooling system fails, fission product decay heating causes water boil-off and reduced heat removal. Zircaloy cladding is oxidized by the steam. The noble gases and volatile fission products such as cesium and iodine that constitute a principal part of the source term will be released from the damaged fuel at or shortly after the time of cladding failure. TREAT test STEP-1 simulated the LOCA environment when the volatile fission products would be released using four fuel elements from the Belgonucleaire BR3 reactor. The principal objective was to collect a portion of the releases carried by the flow stream in a region as close as possible to the test zone. In this paper, the test is described and the results of an analysis of the thermal and steam/hydrogen environment are compared with the test measurements in order to provide a characterization for analysis of fission product releases and aerosol formation. The results of extensive sample examinations are reported separately.

  3. The stress corrosion cracking behavior of alloys 690 and 152 WELD in a PWR environment.

    SciTech Connect (OSTI)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2009-01-01

    Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10{sup -11} m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25--870 C.

  4. Pressure vessel fracture studies pertaining to the PWR thermal-shock issue: experiment TSE-7

    SciTech Connect (OSTI)

    Cheverton, R.D.; Ball, D.G.; Bolt, S.E.; Iskander, S.K.; Nanstad, R.K.

    1985-08-01

    Thermal-shock experiment TSE-7 was conducted for the purpose of investigating the behavior of surface flaws under pressurized-water reactor (PWR) overcooling-accident conditions. This experiment was the eighth in a series of thermal-shock experiments conducted for this purpose with large steel cylinders (A 508, class-2 chemistry; 991-mm OD x 152-mm wall x 1.2-m length) as a part of the Heavy-Section Steel Technology (HSST) Program. The initial flaw for TSE-7 was a shallow, semielliptical, inner-surface, axially oriented, sharp crack located at midlength of the test cylinder. The thermal shock was applied to the inner surface only, and this was accomplished by effectively dunking the test cylinder, initially at approx.93/sup 0/C, into a large volume of liquid nitrogen. The specific purpose of TSE-7 was to determine whether, in agreement with analysis, a short and shallow surface flaw, in the absence of cladding, would extend on the surface to effectively become a very long flaw as a result of severe thermal-shock loading. During the experiment, there were three major initiation-arrest events. The first event consisted of some radial propagation and very extensive surface extension, with many bifurcations taking place. The second and third events consisted primarily of radial propagation. A fourth initiation event was prevented by warm prestressing. These results were in good agreement with predictions. 50 refs., 77 figs., 13 tabs.

  5. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E.

    1997-04-01

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  6. Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Assessment of Radionuclide Monitoring in the CSRA Savannah River NERP Research ... Upcoming Seminars The Savannah River Ecology Laboratory is a research unit of the ...

  7. Central Wisconsin Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    https:www.facebook.compagesCentral-Wisconsin-Electric-Cooperative268841143249085?refaymthomepagepanel Outage Hotline: 800-377-2932 References: EIA Form EIA-861 Final...

  8. Illinois Municipal Elec Agency | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Yes Activity Buying Transmission Yes Activity Buying Distribution Yes Activity Wholesale Marketing Yes This article is a stub. You can help OpenEI by expanding it. Utility...

  9. Western Massachusetts Elec Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Green Button Access: Implemented Green Button Landing Page: www.wmeco.comResidential Green Button Reference Page: www.wmeco.comResidential References: EIA Form EIA-861 Final...

  10. 2005 Elec. Safety-rev1.pmd

    Energy Savers

    5 Electrical Safety During Excavations and Penetrations ... Abbreviations Used in This Report CFR Code of Federal ... LLNL Lawrence Livermore National Laboratory NNSA National ...

  11. Cumberland Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Schedules Grid-background.png Average Rates Residential: 0.1060kWh Commercial: 0.1120kWh Industrial: 0.0733kWh The following table contains monthly sales and revenue data...

  12. Cumberland Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    EIA Form EIA-861 Final Data File for 2010 - File1a1 Energy Information Administration Form 8262 EIA Form 861 Data Utility Id 4624 Utility Location Yes Ownership C...

  13. Northwestern Wisconsin Elec Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Yes Activity Distribution Yes Activity Wholesale Marketing Yes Activity Retail Marketing Yes This article is a stub. You can help OpenEI by expanding it. Utility Rate...

  14. Northern Virginia Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 13640 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes RTO PJM Yes Activity Distribution Yes Alt Fuel Vehicle Yes Alt Fuel Vehicle2 Yes This...

  15. Rutherford Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    1-800-521-0920 or 1-800-228-9756 or 1-800-228-5331 Outage Map: www.remc.comstorm-centerouta References: EIA Form EIA-861 Final Data File for 2010 - File1a1 Energy...

  16. CEPAN method of analyzing creep collapse of oval cladding. Volume 5. Evaluation of interpellet gap formation and clad collapse in modern PWR fuel rods

    SciTech Connect (OSTI)

    Adams, W.M.; Fisher, H.D.; Litke, H.J.; Mordarski, W.J.

    1985-04-01

    This report presents the results from a review of interpellet-gap formation, ovality, creepdown and clad collapse data in modern PWR fuel rods. Conclusions are reached regarding the propensity of modern PWR fuel to form such gaps and to undergo clad collapse. CEPAN, a creep-collapse predictor approved by the NRC in 1976, has been reformulated to include the creep analysis of cladding with finite interpellet gaps. The basis for this reformulation is discussed in detail. The model previously used in the calculation of the augmentation factor, a peak linear heat rate penalty due to the presence of interpellet gaps within the fuel rod, has been modified to incorporate gap-formation statistics from modern fuel. Finnally, the benefits of the limited gap formation and the CEPAN reformulation for the licensing of modern PWR fuel rods are evaluated.

  17. On-line PWR RHR pump performance testing following motor and impeller replacement

    SciTech Connect (OSTI)

    DiMarzo, J.T.

    1996-12-01

    On-line maintenance and replacement of safety-related pumps requires the performance of an inservice test to determine and confirm the operational readiness of the pumps. In 1995, major maintenance was performed on two Pressurized Water Reactor (PWR) Residual Heat Removal (RHR) Pumps. A refurbished spare motor was overhauled with a new mechanical seal, new motor bearings and equipped with pump`s `B` impeller. The spare was installed into the `B` train. The motor had never been run in the system before. A pump performance test was developed to verify it`s operational readiness and determine the in-situ pump performance curve. Since the unit was operating, emphasis was placed on conducting a highly accurate pump performance test that would ensure that it satisfied the NSSS vendors accident analysis minimum acceptance curve. The design of the RHR System allowed testing of one train while the other was aligned for normal operation. A test flow path was established from the Refueling Water Storage Tank (RWST) through the pump (under test) and back to the RWST. This allowed staff to conduct a full flow range pump performance test. Each train was analyzed and an expression developed that included an error vector term for the TDH (ft), pressure (psig), and flow rate (gpm) using the variance error vector methodology. This method allowed the engineers to select a test instrumentation system that would yield accurate readings and minimal measurement errors, for data taken in the measurement of TDH (P,Q) versus Pump Flow Rate (Q). Test results for the `B` Train showed performance well in excess of the minimum required. The motor that was originally in the `B` train was similarly overhauled and equipped with `A` pump`s original impeller, re-installed in the `A` train, and tested. Analysis of the `A` train results indicate that the RHR pump`s performance was also well in excess of the vendors requirements.

  18. Analyses of High Pressure Molten Debris Dispersion for a Typical PWR Plant

    SciTech Connect (OSTI)

    Osamu KAawabata; Mitsuhiro Kajimoto [Japan Nuclear Energy Safety Organization (Japan)

    2006-07-01

    In such severe core damage accident, as small LOCAs with no ECCS injection or station blackout, in which the primary reactor system remains pressurized during core melt down, certain modes of vessel failure would lead to a high pressure ejection of molten core material. In case of a local failure of the lower head, the molten materials would initially be ejected into the cavity beneath the pressure vessel may subsequently be swept out from the cavity to the containment atmosphere and it might cause the early containment failure by direct contact of containment steel liner with core debris. When the contribution of a high-pressure scenario in a core damage frequency increases, early conditional containment failure probability may become large. In the present study, the verification analysis of PHOENICS code and the combining analysis with MELCOR and PHOENICS codes were performed to examine the debris dispersion behavior during high pressure melt ejection. The PHOENICS code which can treat thermal hydraulic phenomena, was applied to the verification analysis for melt dispersion experiments conducted by the Purdue university in the United States. A low pressure melt dispersion experiment at initial pressure 1.4 MPas used metal woods as a molten material was simulated. The analytical results with molten debris dispersion mostly from the model reactor cavity compartment showed an agreement with the experimental result, but the analysis result of a volumetric median diameter of the airborne debris droplets was estimated about 1.5 times of the experimental result. The injection rates of molten debris and steam after reactor vessel failure for a typical PWR plant were analyzed using the MELCOR code. In addition, PHOENICS was applied to a 3D analysis for debris dispersion with low primary pressure at the reactor vessel failure. The analysis result showed that almost all the molten debris were dispersed from the reactor vessel cavity compartment by about 45 seconds after the

  19. Radionuclide release from PWR fuels in a reference tuff repository groundwater

    SciTech Connect (OSTI)

    Wilson, C.N.; Oversby, V.M.

    1985-03-01

    The Nevada Nuclear Waste Storage Investigations Project (NNWSI) is studying the suitability of the welded devitrified Topopah Spring tuff at Yucca Mountain, Nye County, Nevada, for potential use as a high-level nuclear waste repository. In support of the Waste Package task of NNWSI, tests have been conducted under ambient air environment to measure radionuclide release from two pressurized water reactor (PWR) spent fuels in water obtained from the J-13 well near the Yucca Mountain site. Four specimen types, representing a range of fuel physical conditions that may exist in a failed waste canister containing a limited amount of water were tested. The specimen types were: fuel rod sections split open to expose bare fuel particles; rod sections with water-tight end fittings with a 2.5-cm long by 150-{mu}m wide slit through the cladding; rod sections with water-tight end fittings and two 200-{mu}m-diameter holes through the cladding; and undefected rod segments with water-tight end fittings. Radionuclide release results from the first 223-day test runs on H.B. Robinson spent fuel specimens in J-13 water are reported and compared to results from a previous test series in which similar Turkey Point reactor spent fuel specimens were tested on deionized water. Selected initial results are also given for Turkey Point fuel specimens tested on J-13 water. Results suggest that the actinides Pu, Am, Cm and Np are released congruently with U as the UO{sub 2} spent fuel matrix dissolves. Fractional release of {sup 137}Cs and {sup 99}Tc was greater than that measured for the actinides. Generally, lower radionuclide releases were measured for the H.B. Robinson fuel in J-13 water than for Turkey Point Fuel in deionized water. 8 references, 7 figures, 9 tables.

  20. Severe accident modeling of a PWR core with different cladding materials

    SciTech Connect (OSTI)

    Johnson, S. C.; Henry, R. E.; Paik, C. Y.

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  1. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    SciTech Connect (OSTI)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A.

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  2. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    SciTech Connect (OSTI)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made.

  3. Containment pressurization and burning of combustible gases in a large, dry PWR containment during a station blackout sequence

    SciTech Connect (OSTI)

    Lee, M.; Fan, C.T. (National Tsing-Hua Univ., Dept. of Nuclear Engineering, Hsinchu (TW))

    1992-07-01

    In this paper, responses of a large, dry pressurized water reactor (PWR) containment in a station blackout sequence are analyzed with the CONTAIN, MARCH3, and MAAP codes. Results show that the predicted containment responses in a station blackout sequence of these three codes are substantially different. Among these predictions, the MAAP code predicts the highest containment pressure because of the large amount of water made available to quench the debris upon vessel failure. The gradual water boiloff by debris pressurizes the containment. The combustible gas burning models in these codes are briefly described and compared.

  4. First interim examination of defected BWR and PWR rods tested in unlimited air at 229/sup 0/C

    SciTech Connect (OSTI)

    Einziger, R.E.; Cook, J.A.

    1983-01-01

    A five-year whole rod test was initiated to investigate the long-term stability of spent fuel rods under a variety of possible dry storage conditions. Both PWR and BWR rods were included in the test. The first interim examination was conducted after three months of testing to determine if there was any degradation in those defected rods stored in an unlimited air atmosphere. Visual observations, diametral measurements and radiographic smears were used to assess the degree of cladding deformation and particulate dispersal. The PWR rod showed no measurable change from the pre-test condition. The two original artificial defects had not changed in appearance and there was no diametral growth of the cladding. One of the defects in BWR rod showed significant deformation. There was approximately 10% cladding strain at the defect site and a small axial crack had formed. The fuel in the defect did not appear to be friable. The second defect showed no visible change and no cladding strain. Following examination, the test was continued at 230/sup 0/C. Another interim examination is planned during the summer of 1983. This paper discusses the details and meaning of the data from the first interim examination.

  5. Experiment data report for LOFT large-break loss-of-coolant experiment L2-5. [PWR

    SciTech Connect (OSTI)

    Bayless, P.D.; Divine, J.M.

    1982-08-01

    Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large (approx. 1000 MW(e)) commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected.

  6. Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants. Case study: PWR during routine operations

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Haggard, D.L.

    1986-01-01

    The purpose of this report is to present data in support of evaluating the impact of fuel cladding failure events on occupational radiation exposure. To determine quantitatively whether fuel cladding failure contributes significantly to occupational radiation exposure, radiation exposure measurements were taken at comparable locations in two mirror-image pressurized-water reactors (PWRs) and their common auxiliary building. One reactor, Unit B, was experiencing degraded fuel characterized as 0.125% fuel pin-hole leakers and was operating at approximately 55% of the reactor's licensed maximum core power, while the other reactor, Unit A, was operating under normal conditions with less than 0.01% fuel pin-hole leakers at 100% of the reactor's licensed maximum core power. Measurements consisted of gamma spectral analyses, radiation exposure rates and airborne radionuclide concentrations. In addition, data from primary coolant sample results for the previous 20 months on both reactor coolant systems were analyzed. The results of the measurements and coolant sample analyses suggest that a 3560-megawatt-thermal (1100 MWe) PWR operating at full power with 0.125% failed fuel can experience an increase of 540% in radiation exposure rates as compared to a PWR operating with normal fuel. In specific plant areas, the degraded fuel may elevate radiation exposure rates even more.

  7. LOFTRAN/RETRAN comparison calculations for a postulated loss-of-feedwater ATWS in the Sizewell 'B' PWR

    SciTech Connect (OSTI)

    Papez, K.L.; Risher, D.H.

    1983-05-01

    The loss-of-main-feedwater transient without reactor trip (scram) has received particular attention in pressurized water reactor (PWR) anticipated transient without scram (ATWS) analysis primarily due to the potential for reactor coolant system over pressurization. To assist in the licensing of the U.K. PWR, Sizewell 'B', comparative calculations of a loss-of-feedwater ATWS have been performed using the Westinghouse-developed LOFTRAN loop analysis code and the Electric Power Research Institute/ Energy Incorporated-developed RETRAN-01 code. The calculations were performed with and without the emergency boration system (EBS), which is included in the Sizewell reference design. Initial results showed good agreement between the codes for the major features of the transient, but also a time shift in the transient profiles at the time of the pressurizer pressure peak. This was found to be due to differences in the steam generator modeling, which resulted in a difference in the onset of the very rapid degradation in heat transfer as the steam generators approach dryout. When the same model was used in both codes, very good agreement was obtained. Remaining differences in the results are attributed primarily to differences in the boron injection models, which resulted in an over-prediction of the core boron concentration in the RETRAN calculation. The results with an EBS indicate that the peak pressurizer pressure is relatively insensitive to variations in modeling.

  8. In-Vessel Retention of Molten Core Debris in the Westinghouse AP1000 Advanced Passive PWR

    SciTech Connect (OSTI)

    Scobel, James H.; Conway, L.E.; Theofanous, T.G.

    2002-07-01

    In-vessel retention (IVR) of molten core debris via external reactor vessel cooling is the hallmark of the severe accident management strategies in the AP600 passive PWR. The vessel is submerged in water to cool its external surface via nucleate boiling heat transfer. An engineered flow path through the reactor vessel insulation provides cooling water to the vessel surface and vents steam to promote IVR. For the 600 MWe passive plant, the predicted heat load from molten debris to the lower head wall has a large margin to the critical heat flux on the external surface of the vessel, which is the upper limit of the cooling capability. Up-rating the power of the passive plant from 600 to 1000 MWe (AP1000) significantly increases the heat loading from the molten debris to the reactor vessel lower head in the postulated bounding severe accident sequence. To maintain a large margin to the coolability limit for the AP1000, design features and severe accident management (SAM) strategies to increase the critical heat flux on the external surface of the vessel wall need to be implemented. A test program at the ULPU facility at University of California Santa Barbara (UCSB) has been initiated to investigate design features and SAM strategies that can enhance the critical heat flux. Results from ULPU Configuration IV demonstrate that with small changes to the ex-vessel design and SAM strategies, the peak critical heat flux in the AP1000 can be increased at least 30% over the peak critical heat flux predicted for the AP600 configuration. The design and SAM strategy changes investigated in ULPU Configuration IV can be implemented in the AP1000 design and will allow the passive plant to maintain the margin to critical heat flux for IVR, even at the higher power level. Continued testing for IVR phenomena is being performed at UCSB to optimize the AP1000 design and to ensure that vessel failure in a severe accident is physically unreasonable. (authors)

  9. Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1

    Energy.gov [DOE]

    Results of testing employing surrogate instrumented rods (non-high-burnup, 17 x 17 PWR fuel assembly) to capture the response to the loadings experienced during normal conditions of transport indicate that strain- or stress-based failure of fuel rods seems unlikely; performance of high-burnup fuels continues to be assessed.

  10. Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    in 1997 and replaced with two other areas, both located in the Savannah River swamp. ... on the natural levy that parallels the Savannah River. Area: 1 2 3 4 5 6 7 8 9 10 11 ...

  11. River Corridor - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    River Corridor Richland Operations Office Richland Operations Office River Corridor B Reactor 300 Area 324 Building 618-10 and 618-11 Burial Grounds C Reactor D and DR Reactors F ...

  12. River Corridor Achievements

    Energy.gov [DOE]

    Washington Closure Hanford and previous contractors have completed much of the cleanup work in the River Corridor, shown here.

  13. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    SciTech Connect (OSTI)

    Bi, G.; Liu, C.; Si, S.

    2012-07-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade {sup 233}U-Thorium (U{sub 3}ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade {sup 233}U extracted from burnt PuThOX fuel was used to fabrication of U{sub 3}ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U{sub 3}ThOX mixed core, the well designed U{sub 3}ThOX FAs with 1.94 w/o fissile uranium (mainly {sup 233}U) were located on the periphery of core as a blanket region. U{sub 3}ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U{sub 3}ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U{sub 3}ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U{sub 3}ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U{sub 3}ThOX loading on the periphery of core has no

  14. Simulation of German PKL refill/reflood experiment K9A using RELAP4/MOD7. [PWR

    SciTech Connect (OSTI)

    Hsu, M.T.; Davis, C.B.; Behling, S.R.

    1981-11-01

    This paper describes a RELAP4/MOD7 simulation of West Germany's Kraftwerk Union (KWU) Primary Coolant Loop (PKL) refill/reflood experiment K9A. RELAP4/MOD7, a best-estimate computer program for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This study was the first major simulation using RELAP4/MOD7 since its release by the Idaho National Engineering Laboratory (INEL). The PKL facility is a reduced scale (1:134) representation of a typical West German four-loop 1300 MW pressurized water reactor (PWR). A prototypical scale of the total volume to power ratio was maintained. The test facility was designed specifically for an experiment simulating the refill/reflood phase of a Loss-of-Coolant Accident (LOCA).

  15. Development code for sensitivity and uncertainty analysis of input on the MCNPX for neutronic calculation in PWR core

    SciTech Connect (OSTI)

    Hartini, Entin Andiwijayakusuma, Dinan

    2014-09-30

    This research was carried out on the development of code for uncertainty analysis is based on a statistical approach for assessing the uncertainty input parameters. In the butn-up calculation of fuel, uncertainty analysis performed for input parameters fuel density, coolant density and fuel temperature. This calculation is performed during irradiation using Monte Carlo N-Particle Transport. The Uncertainty method based on the probabilities density function. Development code is made in python script to do coupling with MCNPX for criticality and burn-up calculations. Simulation is done by modeling the geometry of PWR terrace, with MCNPX on the power 54 MW with fuel type UO2 pellets. The calculation is done by using the data library continuous energy cross-sections ENDF / B-VI. MCNPX requires nuclear data in ACE format. Development of interfaces for obtaining nuclear data in the form of ACE format of ENDF through special process NJOY calculation to temperature changes in a certain range.

  16. Impact Analysis of a Dipper-Type and Multi Spring-Type Fuel Rod Support Grid Assemblies in PWR

    SciTech Connect (OSTI)

    Song, K.N.; Yoon, K.H.; Park, K.J.; Park, G.J.; Kang, B.S.

    2002-07-01

    A spacer grid is one of the main structural components in a fuel assembly of a Pressurized light Water Reactor (PWR). It supports fuel rods, guides cooling water, and maintains geometry from external impact loads. A simulation is performed for the strength of a spacer grid under impact load. The critical impact load that leads to plastic deformation is identified by a free-fall test. A finite element model is established for the nonlinear simulation of the test. The simulation model is tuned based on the free-fall test. The model considers the aspects of welding and the contacts between components. Nonlinear finite element analysis is carried out by a software system called LS/DYNA3D. The results are discussed from a design viewpoint. (authors)

  17. The influence of dissolved hydrogen on primary water stress corrosion cracking of Alloy 600 at PWR steam generator operating temperatures

    SciTech Connect (OSTI)

    Jacko, R.J.; Economy, G.; Pement, F.W.

    1992-12-31

    PWR primary coolant chemistry uses an intentional dissolved hydrogen concentration of 20 to 50 ml (STP)/kg of water to effect a net suppression of oxygen-producing radiolysis, to minimize corrosion in primary loop materials and to maintain a low redox potential. Speculation has attended a possible influence of dissolved hydrogen on the kinetics of initiation of Primary Water Stress Corrosion Cracking (PWSCC) behavior of Alloy 600 steam generator tubing. Three series of experiments are presented for conditions in which the level of dissolved hydrogen was intentionally varied over the hydrogen and temperature ranges of interest for steam generator operation. No significant effect of dissolved hydrogen was found on PWSCC of Alloy 600.

  18. Determination of the threshold values for corrosion fatigue crack growth rate of pressure vessel steels in PWR primary water

    SciTech Connect (OSTI)

    Haenninen, H.E.; Arilahti, E.; Ehrnsten, U.

    1992-12-31

    Corrosion fatigue crack growth rates over a range of frequencies from 10 Hz to 0.00001 Hz in two materials that have exhibited low-rate (A508 Class 3) and high-rate (A533B) crack growth behaviour at 288{degrees}C were studied at 200{degrees}C in PWR primary water. The frequency values above which marked environmental enhancement was observed were determined. Also the threshold values in terms of {Delta}K{sub th}, above which the marked environmental enhancement was observed in the crack growth rate, were determined both for A533B steel and Soviet pressure vessel steels with certain test parameters. Based on the extensive fractography the crack growth rate results are discussed mechanistically.

  19. The toughness of irradiated pressure water reactor (PWR) vessel shell rings and the effect of segregation zones

    SciTech Connect (OSTI)

    Bethmont, M.; Frund, J.M.; Housin, B.; Soulat, P.

    1996-12-31

    To establish the integrity of pressure water reactor (PWR) vessels it is necessary to determine the toughness of A508Cl.3 steel at the end of its life, that is after thermal aging and irradiation embrittlement. In safety analyses the toughness can be deduced from a reference curve set forth in the code (ASME or RCC-M). The validity of the reference curve has been verified for several years for unirradiated French reactor vessels. Tests were performed on specimens taken from materials having heterogeneities in chemical composition. For most of the test results the reference curve is a lower bound. To solve te problem of determining the toughness of SA 508 Cl.3 steel after irradiation and in the presence of possible heterogeneities, the toughness results were gathered. The synthesis shows that the RCC-M code curve is conservative.

  20. Safety and licensing issues that are being addressed by the Power Burst Facility test programs. [PWR; BWR

    SciTech Connect (OSTI)

    McCardell, R.K.; MacDonald, P.E.

    1980-01-01

    This paper presents an overview of the results of the experimental program being conducted in the Power Burst Facility and the relationship of these results to certain safety and licensing issues. The safety issues that were addressed by the Power-Cooling-Mismatch, Reactivity Initiated Accident, and Loss of Coolant Accident tests, which comprised the original test program in the Power Burst Facility, are discussed. The resolution of these safety issues based on the results of the thirty-six tests performed to date, is presented. The future resolution of safety issues identified in the new Power Burst Facility test program which consists of tests which simulate BWR and PWR operational transients, anticipated transients without scram, and severe fuel damage accidents, is described.

  1. The effects of parameter variation on MSET models of the Crystal River-3 feedwater flow system.

    SciTech Connect (OSTI)

    Miron, A.

    1998-04-01

    In this paper we develop further the results reported in Reference 1 to include a systematic study of the effects of varying MSET models and model parameters for the Crystal River-3 (CR) feedwater flow system The study used archived CR process computer files from November 1-December 15, 1993 that were provided by Florida Power Corporation engineers Fairman Bockhorst and Brook Julias. The results support the conclusion that an optimal MSET model, properly trained and deriving its inputs in real-time from no more than 25 of the sensor signals normally provided to a PWR plant process computer, should be able to reliably detect anomalous variations in the feedwater flow venturis of less than 0.1% and in the absence of a venturi sensor signal should be able to generate a virtual signal that will be within 0.1% of the correct value of the missing signal.

  2. Experiment operations plan for the MT-4 experiment in the NRU reactor. [PWR

    SciTech Connect (OSTI)

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700/sup 0/F).

  3. Office of River Protection - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Office of River Protection Office of River Protection Office of River Protection Office of River Protection Email Email Page | Print Print Page |Text Increase Font Size Decrease...

  4. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    SciTech Connect (OSTI)

    Loeb, Andreas; Runge, Hartmut; Stanke, Dieter; Bertholdt, Horst-Otto; Adams, Andreas; Impertro, Michael; Roesch, Josef

    2013-07-01

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  5. Technical basis for the initiation and cessation of environmentally-assisted cracking of low-alloy steels in elevated temperature PWR environments

    SciTech Connect (OSTI)

    James, L.A.

    1997-10-01

    The Section 11 Working Group on Flaw Evaluation of the ASME B and PV Code Committee is considering a Code Case to allow the determination of the conditions under which environmentally-assisted cracking of low-alloy steels could occur in PWR primary environments. This paper provides the technical support basis for such an EAC Initiation and Cessation Criterion by reviewing the theoretical and experimental information in support of the proposed Code Case.

  6. Savannah River National Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Savannah River National Laboratory srnl.doe.gov SRNL is a DOE National Laboratory operated by Savannah River Nuclear Solutions. At a glance 'Tin whiskers' suppression method Researchers at the Savannah River National Laboratory (SRNL) have identified a treatment method that slows or prevents the formation of whiskers in lead-free solder. Tin whiskers spontaneously grow from thin films of tin, often found in microelectronic devices in the form of solders and platings. Background This problem was

  7. River and Plateau Committee

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    of Energy River Turbine Provides Clean Energy to Remote Alaskan Village River Turbine Provides Clean Energy to Remote Alaskan Village August 18, 2015 - 10:36am Addthis River Turbine Provides Clean Energy to Remote Alaskan Village Alison LaBonte Marine and Hydrokinetic Technology Manager To date, Ocean Renewable Power Company (ORPC) is the only company to have built, operated and delivered power to a utility grid from a hydrokinetic tidal project, and to a local microgrid from a hydrokinetic

  8. Experiment operations plan for the TH-2 experiment in the NRU reactor. [PWR; BWR

    SciTech Connect (OSTI)

    Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Freshley, M.D.

    1983-06-01

    A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for TH-2--the second experiment in the series of thermal-hydraulic tests conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. The major objective of TH-2 was to develop the experiment reflood control parameters and the procedures to be used in subsequent experiments in this program. In this experiment, the data acquisition and control system was used to control the fuel cladding temperature during a simulated LOCA by using variable reflood coolant flow.

  9. Great River (1973)

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Volume One Film Collection Volume Two 75th Anniversary Hydropower in the Northwest Woody Guthrie Videos Strategic Direction Branding & Logos Power of the River History Book...

  10. River of Power (1987)

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Volume One Film Collection Volume Two 75th Anniversary Hydropower in the Northwest Woody Guthrie Videos Strategic Direction Branding & Logos Power of the River History Book...

  11. Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    of bottomland hardwoodfloodplain forest communities of a southern river swamp system. ... or urban waste discharge, or power plant cooling effluents. Area: 1 2 3 4 5 6 7 ...

  12. Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    location of the Savannah River Ecology Laboratory, is one of the original ten SREL habitat reserves and was selected to complement the old-field habitatplant succession studies ...

  13. Lower Colorado River Authority | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    of Lower Colorado River Authority's communications requirements Lower Colorado River Authority (134.07

  14. about Savannah River National Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    The EDM capability at the Savannah River National Laboratory (SRNL) is unique to the Savannah River Site. It allows for very fine, precise cutting of metal without destroying ...

  15. Savannah River Field Office | National Nuclear Security Administration |

    National Nuclear Security Administration (NNSA)

    (NNSA) Savannah River

  16. Reactor Safety Research Programs. Quarterly report, July-September 1984. Volume 3. [PWR; BWR

    SciTech Connect (OSTI)

    Edler, S.K.

    1985-02-01

    This document summarizes work performed by Pacific Northwest Laboratory from July 1 through September 30, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Results from an instrumented fuel assembly irradiation program being performed at Halden, Norway, are reported. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Severe fuel damage accident tests are being conducted in the NRU Reactor, Chalk River, Canada.

  17. Reactor safety research programs. Quarterly report, January-March 1984. Vol. 1. [PWR; BWR

    SciTech Connect (OSTI)

    Edler, S.K.

    1984-06-01

    This document summarizes work performed by Pacific Northwest Laboratory from January 1 through March 31, 1984, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Results from an instrumented fuel assembly irradiation program being performed at Halden, Norway, are reported. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data on analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Fuel assemblies and analytical support are being provided for experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada.

  18. Improvement of the thermal margins in the Swedish Ringhals-3 PWR by introducing new fuel assemblies with thorium

    SciTech Connect (OSTI)

    Lau, C. W.; Demaziere, C.; Nylen, H.; Sandberg, U.

    2012-07-01

    Thorium is a fertile material and most of the past research has focused on breeding thorium to fissile material. In this paper, the focus is on using thorium to improve the thermal margins by homogeneously distributing thorium in the fuel pellets. A proposed uranium-thorium-based fuel assembly is simulated for the Swedish Ringhals-3 PWR core in a realistic demonstration. All the key safety parameters, such as isothermal temperature coefficient of reactivity, Doppler temperature of reactivity, boron worth, shutdown margins and fraction of delayed neutrons are studied in this paper, and are within safety limits for the new core design using the uranium-thorium-based fuel assemblies. The calculations were performed by the two-dimensional transport code CASMO-4E and the two group steady-state three dimensional nodal code SIMULATE-3 from Studsvik Scandpower. The results showed that the uranium-thorium-based fuel assembly improves the thermal margins, both in the pin peak power and the local power (Fq). The improved thermal margins would allow more flexible core designs with less neutron leakage or could be used in power uprates to offer efficient safety margins. (authors)

  19. Presentation of the MERC work-flow for the computation of a 2D radial reflector in a PWR

    SciTech Connect (OSTI)

    Clerc, T.; Hebert, A.; Leroyer, H.; Argaud, J. P.; Poncot, A.; Bouriquet, B.

    2013-07-01

    This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of characteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Donnees et Aide a l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R and D choices made at Electricite de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow. (authors)

  20. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect (OSTI)

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  1. Effects of Zircaloy oxidation and steam dissociation on PWR core heat-up under conditions simulating uncovered fuel rods

    SciTech Connect (OSTI)

    Viskanta, R.; Mohanty, A.K.

    1986-04-01

    The studies described in this report identify the regimes of slow transients in a partially uncovered core of a PWR. The threshold height and onset time for oxidation of the cladding of a fuel rod have been evaluated. The effects of oxidation in increasing the decay heat load, component temperature, reduction of cladding thickness and generation of hydrogen have been estimated. The condition for steam starvation has been determined. At high uncovered core heights, typically say 2.8 m for a geometry simulating the TMI-2 type of reactor, the solid and coolant temperatures can reach the limits of steam dissociation. The effects of radiation heat exchange between cladding and coolant, Zircaloy oxidation, steam dissociation, gap conductance between fuel and cladding and system pressure on the heatup of fuel rods have been investigated. The time for uncovering a certain core height is taken as the independent parameter. It is seen that if the uncovering process is allowed to continue beyond 9 minutes corresponding to an uncovered height of 1.9 m, onset of cladding oxidation can be a reality. These values provide a guideline for the response time of the emergency core cooling systems. 10 refs., 22 figs.

  2. In-plant test and evaluation of the neutron collar for verification of PWR fuel assemblies at Resende, Brazil

    SciTech Connect (OSTI)

    Menlove, H.O.; Marzo, M.A.S.; de Almeida, S.G.; de Almeida, M.C.; Moitta, L.P.M.; Conti, L.F.; de Paiva, J.R.T.

    1985-11-01

    The neutron-coincidence collar has been evaluated for the measurement of pressurized-water reactor (PWR) fuel assemblies at the Fabrica de Elementos Combustiveis plant in Resende, Brazil. This evaluation was part of the cooperative-bilateral-safeguards technical-exchange program between the United States and Brazil. The neutron collar measures the STVU content per unit length of full fuel assemblies using neutron interrogation and coincidence counting. The STYU content is measured in the passive mode without the AmLi neutron-interrogation source. The extended evaluation took place over a period of 6 months with both scanning and single-zone measurements. The results of the tests gave a coincidence-response standard deviation of 0.7% (sigma = 1.49% for mass) for the active case and 2.5% for the passive case in 1000-s measurement times. The length measurement in the scanning mode was accurate to 0.77%. The accuracies of different calibration methods were evaluated and compared.

  3. Tests with Inconel 600 to obtain quantitative stress-corrosion cracking data for evaluating service performance. [PWR

    SciTech Connect (OSTI)

    Bandy, R.; van Rooyen, D.

    1982-09-01

    Inconel 600 tubes in pressurized water reactor (PWR) steam generators form a pressure boundary between radioactive primary water and secondary water which is converted to steam and used for generating electricity. Under operating conditions the performance of alloy 600 has been good, but with some occasional small leaks resulting from stress corrosion cracking (SCC), related to the presence of unusually high residual or operating stresses. The suspected high stresses can result from either the deformation of tubes during manufacture, or distortion during abnormal conditions such as denting. The present experimental program addresses two specific conditions, i.e., (1) where deformation occurs but is no longer active, such as when denting is stopped and (2) where plastic deformation of the metal continues, as would occur during denting. Laboratory media consist of pure water as well as solutions to simulate environments that would apply in service; tubing from actual production is used in carrying out these tests. The environments include both normal and off chemistries for primary and secondary water. The results reported here were obtained in several different tests. The main ones are (1) split tube reverse U-bends, (2) constant extension rate tests (CERT), and (3) constant load. The temperature range covered is 290 to 365/sup 0/C.

  4. CRC DEPLETION CALCULATIONS FOR THE RODDED ASSEMBLIES IN BATCHES 1, 2, 3, AND 1X OF CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Kenneth D. Wright

    1997-09-03

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain rodded fuel assemblies from batches 1, 2, 3, and 1X of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A rodded assembly is one that contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) for some period of time during its irradiation history. The objective of this analysis is to provide SAS2H calculated isotopic compositions of depleted fuel and depleted burnable poison for each fuel assembly to be used in subsequent CRC reactivity calculations containing the fuel assemblies.

  5. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 1, 2, AND 3 OF CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Kenneth D. Wright

    1997-07-29

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 1, 2, and 3 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  6. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 4 AND 5 OF CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Kenneth D. Wright

    1997-07-30

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 4 and 5 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  7. CRC DEPLETION CALCULATIONS FOR THE NON-RODDED ASSEMBLIES IN BATCHES 8 AND 9 CRYSTAL RIVER UNIT 3

    SciTech Connect (OSTI)

    Michael L. Wilson

    2001-02-08

    The purpose of this design analysis is to document the SAS2H depletion calculations of certain non-rodded fuel assemblies from batches 8 and 9 of the Crystal River Unit 3 pressurized water reactor (PWR) that are required for Commercial Reactor Critical (CRC) evaluations to support the development of the disposal criticality methodology. A non-rodded assembly is one which never contains a control rod assembly (CRA) or an axial power shaping rod assembly (APSRA) during its irradiation history. The objective of this analysis is to provide SAS2H generated isotopic compositions for each fuel assembly's depleted fuel and depleted burnable poison materials. These SAS2H generated isotopic compositions are acceptable for use in CRC benchmark reactivity calculations containing the various fuel assemblies.

  8. River and Harbors Act

    Energy.gov [DOE]

    Section 10 of the Rivers and Harbors Act of 1899 (33 U.S.C. 403) prohibits the unauthorized obstruction or alteration of any navigable water of the United States.

  9. Savannah river site

    National Nuclear Security Administration (NNSA)

    at the Savannah River Site (SRS) to supply and process tritium, a radioactive form of hydrogen that is a vital component of nuclear weapons. SRS loads tritium and non-tritium...

  10. RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

    SciTech Connect (OSTI)

    David Andrs; Ray Berry; Derek Gaston; Richard Martineau; John Peterson; Hongbin Zhang; Haihua Zhao; Ling Zou

    2012-05-01

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

  11. Rod consolidation of RG and E's (Rochester Gas and Electric Corporation) spent PWR (pressurized water reactor) fuel

    SciTech Connect (OSTI)

    Bailey, W.J.

    1987-05-01

    The rod consolidation demonstration involved pulling the fuel rods from five fuel assemblies from Unit 1 of RG and E's R.E. Ginna Nuclear Power Plant. Slow and careful rod pulling efforts were used for the first and second fuel assemblies. Rod pulling then proceeded smoothly and rapidly after some minor modifications were made to the UST and D consolidation equipment. The compaction ratios attained ranged from 1.85 to 2.00 (rods with collapsed cladding were replaced by dummy rods in one fuel assembly to demonstrate the 2:1 compaction ratio capability). This demonstration involved 895 PWR fuel rods, among which there were some known defective rods (over 50 had collapsed cladding); no rods were broken or dropped during the demonstration. However, one of the rods with collapsed cladding unexplainably broke during handling operations (i.e., reconfiguration in the failed fuel canister), subsequent to the rod consolidation demonstration. The broken rod created no facility problems; the pieces were encapsulated for subsequent storage. Another broken rod was found during postdemonstration cutting operations on the nonfuel-bearing structural components from the five assemblies; evidence indicates it was broken prior to any rod consolidation operations. During the demonstration, burnish-type lines or scratches were visible on the rods that were pulled; however, experience indicates that such lines are generally produced when rods are pulled (or pushed) through the spacer grids. Rods with collapsed cladding would not enter the funnel (the transition device between the fuel assembly and the canister that aids in obtaining high compaction ratios). Reforming of the flattened areas of the cladding on those rods was attempted to make the rod cross sections more nearly circular; some of the reformed rods passed through the funnel and into the canister.

  12. Evaluation of the thermal-hydraulic response and fuel rod thermal and mechanical deformation behavior during the power burst facility test LOC-3. [PWR

    SciTech Connect (OSTI)

    Yackle, T.R.; MacDonald, P.E.; Broughton, J.M.

    1980-01-01

    An evaluation of the results from the LOC-3 nuclear blowdown test conducted in the Power Burst Facility is presented. The test objective was to examine fuel and cladding behavior during a postulated cold leg break accident in a pressurized water reactor (PWR). Separate effects of rod internal pressure and the degree of irradiation were investigated in the four-rod test. Extensive cladding deformation (ballooning) and failure occurred during blowdown. The deformation of the low and high pressure rods was similar; however, the previously irradiated test rod deformed to a greater extent than a similar fresh rod exposed to identical system conditions.

  13. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect (OSTI)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  14. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 2: primary coolant loop model. Final report

    SciTech Connect (OSTI)

    Eberhardt, A.C.

    1981-09-01

    This report describes the Zion Station reactor coolant loop model developed by Sargent and Lundy Engineers for Lawrence Livermore National Laboratory as part of its Load Combination Program. This model was developed for use in performing seismic time history analyses of an actual pressurized water reactor (PWR) system. It includes all major items affecting the seismic response of a 4-loop Westinghouse nuclear steam supply system: the components, supports, and interconnecting piping. The model was further expanded to permit static analysis of dead weight, thermal, and internal pressure load conditions.

  15. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5: probabilistic fracture mechanics analysis. Final report

    SciTech Connect (OSTI)

    Harris, D.O.; Lim, E.Y.; Dedhia, D.D.

    1981-08-01

    The purpose of the portion of the Load Combination Program covered in this volume was to estimate the probability of a seismic induced loss-of-coolant accident (LOCA) in the primary piping of a commercial pressurized water reactor (PWR). Such results are useful in rationally assessing the need to design reactor primary piping systems for the simultaneous occurrence of these two potentially high stress events. The primary piping system at Zion I was selected for analysis. Attention was focussed on the girth butt welds in the hot leg, cold leg and cross-over leg, which are centrifugally cast austenitic stainless steel lines with nominal outside diameters of 32 - 37 inches.

  16. Probability of pipe fracture in the primary coolant loop of a PWR Plant. Volume 2. Primary Coolant Loop Model. Load Combination Program, Project I final report

    SciTech Connect (OSTI)

    Eberhardt, A.C.

    1981-06-01

    This report describes the Zion Station reactor coolant loop model developed by Sargent and Lundy Engineers for Lawrence Livermore National Laboratory as part of its Load Combination Program. This model was developed for use in performing seismic time history analyses of an actual pressurized water reactor (PWR) system. It includes all major items affecting the seismic response of a 4-loop Westinghouse nuclear steam supply system: the components, supports, and interconnecting piping. The model was further expanded to permit static analysis of dead weight, thermal, and internal pressure load conditions. 7 refs., 42 figs., 9 tabs.

  17. Wing River Wind Farm | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    to: navigation, search Name Wing River Wind Farm Facility Wing River Wind Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner Wing River...

  18. Sioux River Ethanol LLC | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Ethanol LLC Jump to: navigation, search Name: Sioux River Ethanol LLC Place: Hudson, South Dakota Zip: 57034 Product: Farmer owned ethanol producer, Sioux River Ethanol is...

  19. Office of River Protection - Hanford Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Office of River Protection Office of River Protection About ORP ORP Projects & Facilities Newsroom Contracts & Procurements Contact ORP Office of River Protection Email Email Page...

  20. Raft River Geothermal Facility | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Facility Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Raft River Geothermal Facility General Information Name Raft River Geothermal Facility Facility Raft River...

  1. Flambeau River Biofuels | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Flambeau River Biofuels Jump to: navigation, search Name: Flambeau River Biofuels Place: Park Falls, Wisconsin Sector: Biomass Product: A subsidiary of Flambeau River Papers LLC...

  2. Sky River Wind Farm | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Wind Farm Jump to: navigation, search Name Sky River Wind Farm Facility Sky River Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner...

  3. Savannah River Site Waste Disposition Project

    Office of Environmental Management (EM)

    Terrel J. Spears Assistant Manager Waste Disposition Project DOE Savannah River Operations Office Savannah River Site Savannah River Site Waste Disposition Project Waste ...

  4. Savannah River | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    River Savannah River Following are compliance agreements for the Savannah River Site. Also included are short summaries of the agreements. Natural Resources Defense Council Consent Decree, May 26, 1988 (705.68 KB) Natural Resources Defense Council Consent Decree, May 26, 1988 Summary (40.89 KB) Savannah River Site Consent Order 99-155-W, October 11, 1999 (196.38 KB) Savannah River Site Consent Order 99-155-W, October 11, 1999 Summary (46.6 KB) Savannah River Site Consent Order 85-70-SW, November

  5. Schlumberger soundings in the Upper Raft River and Raft River...

    Open Energy Information (Open El) [EERE & EIA]

    soundings in the Upper Raft River and Raft River Valleys, Idaho and Utah Jump to: navigation, search OpenEI Reference LibraryAdd to library Report: Schlumberger soundings in the...

  6. Performance Spec. for Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shipping Port Spent Fuel Canisters

    SciTech Connect (OSTI)

    JOHNSON, D.M.

    2000-03-14

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders.

  7. Lower Colorado River Authority | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    from Lower Colorado River Authority on Smart Grid communications requirements Lower Colorado River Authority (349.31

  8. Reese River Geothermal Project | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Geothermal Project Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Development Project: Reese River Geothermal Project Project Location Information...

  9. Savannah River Site Robotics

    ScienceCinema (OSTI)

    None

    2016-07-12

    Meet Sandmantis and Frankie, two advanced robotic devices that are key to cleanup at Savannah River Site. Sandmantis cleans hard, residual waste off huge underground storage tanks. Frankie is equipped with unique satellite capabilities and sensing abilties that can determine what chemicals still reside in the tanks in a cost effective manner.

  10. Savannah River Site Robotics

    SciTech Connect (OSTI)

    2010-01-01

    Meet Sandmantis and Frankie, two advanced robotic devices that are key to cleanup at Savannah River Site. Sandmantis cleans hard, residual waste off huge underground storage tanks. Frankie is equipped with unique satellite capabilities and sensing abilties that can determine what chemicals still reside in the tanks in a cost effective manner.

  11. EA-1981: Bonneville-Hood River Transmission Line Rebuild, Multnomah...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    81: Bonneville-Hood River Transmission Line Rebuild, Multnomah and Hood River Counties, Oregon EA-1981: Bonneville-Hood River Transmission Line Rebuild, Multnomah and Hood River ...

  12. Hood River Passive House

    SciTech Connect (OSTI)

    Hales, D.

    2013-03-01

    The Hood River Passive Project was developed by Root Design Build of Hood River Oregon using the Passive House Planning Package (PHPP) to meet all of the requirements for certification under the European Passive House standards. The Passive House design approach has been gaining momentum among residential designers for custom homes and BEopt modeling indicates that these designs may actually exceed the goal of the U.S. Department of Energy's (DOE) Building America program to reduce home energy use by 30%-50% (compared to 2009 energy codes for new homes). This report documents the short term test results of the Shift House and compares the results of PHPP and BEopt modeling of the project.

  13. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  14. PWR FLECHT SEASET 21-rod bundle flow blockage task data and analysis report. NRC/EPRI/Westinghouse Report No. 11. Appendices K-P

    SciTech Connect (OSTI)

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  15. Corrosion and hydriding performance evaluation of three zircaloy-2 clad fuel assemblies after continuous exposure in PWR cores 1 and 2, at Shippingport, PA

    SciTech Connect (OSTI)

    Hillner, E.

    1980-01-01

    Three original Zircaloy-2 clad blanket fuel bundles from the pressurized water reactor (PWR) at the Shippingport Atomic Power Station were discharged after continuous exposure during Cores 1 and 2. Detailed visual examination of these components after approx. 6300 calendar days of operation (51,140 EFPH) revealed only the anticipated uniform light gray (post-transition) corrosion products with no evidence of unexpected corrosion deterioration, fuel rod warpage, or other damage. All corrosion films were found to be tightly adherent to the underlying cladding. An extensive destructive examination of a selected fuel rod from each of three fuel bundles produced appreciably greater end-of-life rod average oxide film thickness when compared with corresponding values produced from a set of empirical equations generated from the out-of-pile (autoclave) testing of Zircaloy coupons.

  16. Estimating pressurized water reactor decommissioning costs: A user`s manual for the PWR Cost Estimating Computer Program (CECP) software. Draft report for comment

    SciTech Connect (OSTI)

    Bierschbach, M.C.; Mencinsky, G.J.

    1993-10-01

    With the issuance of the Decommissioning Rule (July 27, 1988), nuclear power plant licensees are required to submit to the US Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. This user`s manual and the accompanying Cost Estimating Computer Program (CECP) software provide a cost-calculating methodology to the NRC staff that will assist them in assessing the adequacy of the licensee submittals. The CECP, designed to be used on a personnel computer, provides estimates for the cost of decommissioning PWR plant stations to the point of license termination. Such cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial costs; and manpower costs. In addition to costs, the CECP also calculates burial volumes, person-hours, crew-hours, and exposure person-hours associated with decommissioning.

  17. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR (pressurized-water reactor) during an outage

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Kenoyer, J.L.

    1987-08-01

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions.

  18. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 1. Summary, Load Combination Program. Project I final report

    SciTech Connect (OSTI)

    Lu, S.; Streit, R.D.; Chou, C.K.

    1981-06-01

    This report summarizes work performed to establish a technical basis for the NRC to use in reassessing its requirement that earthquake and large loss-of-coolant accident (LOCA) loads be combined in the design of nuclear power plants. A systematic probabilistic approach is used to treat the random nature of earthquake and transient loading and to estimate the probability of large LOCAs that are directly and indirectly induced by earthquakes. A large LOCA is defined in this report as a double-ended guillotine break of the primary reactor coolant loop piping (the hot leg, cold leg, and crossover) of a pressurized water reactor (PWR). Unit 1 of the Zion Nuclear Power Plant, a four-loop PWR, is the demonstration plant used in this study. To estimate the probability of a large LOCA directly induced by earthquakes, only fatigue crack growth resulting from the combined effects of thermal, pressure, seismic, and other cyclic loads is considered. Fatigue crack growth is simulated by a deterministic fracture mechanics model with stochastic inputs of initial crack size distribution, material properties, stress histories, and leak detection probability. Results of the simulation indicate that the probability of a double-ended guillotine break, either with or without earthquake, is very small (on the order of 10/sup -12/). The probability of a leak was found to be several orders of magnitude greater than that of a large LOCA, complete pipe rupture. A limited investigation involving engineering judgment of a double-ended guillotine break indirectly induced by an earthquake is also reported.

  19. Crack growth rates and metallographic examinations of Alloy 600 and Alloy 82/182 from field components and laboratory materials tested in PWR environments.

    SciTech Connect (OSTI)

    Alexandreanu, B.; Chopra, O. K.; Shack, W. J.

    2008-05-05

    In light water reactors, components made of nickel-base alloys are susceptible to environmentally assisted cracking. This report summarizes the crack growth rate results and related metallography for field and laboratory-procured Alloy 600 and its weld alloys tested in pressurized water reactor (PWR) environments. The report also presents crack growth rate (CGR) results for a shielded-metal-arc weld of Alloy 182 in a simulated PWR environment as a function of temperature between 290 C and 350 C. These data were used to determine the activation energy for crack growth in Alloy 182 welds. The tests were performed by measuring the changes in the stress corrosion CGR as the temperatures were varied during the test. The difference in electrochemical potential between the specimen and the Ni/NiO line was maintained constant at each temperature by adjusting the hydrogen overpressure on the water supply tank. The CGR data as a function of temperature yielded activation energies of 252 kJ/mol for a double-J weld and 189 kJ/mol for a deep-groove weld. These values are in good agreement with the data reported in the literature. The data reported here and those in the literature suggest that the average activation energy for Alloy 182 welds is on the order of 220-230 kJ/mol, higher than the 130 kJ/mol commonly used for Alloy 600. The consequences of using a larger value of activation energy for SCC CGR data analysis are discussed.

  20. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    SciTech Connect (OSTI)

    Mueller, Don E.; Marshall, William J.; Wagner, John C.; Bowen, Douglas G.

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  1. Look to the River Columbia River Opens New Opportunities for...

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Volume One Film Collection Volume Two 75th Anniversary Hydropower in the Northwest Woody Guthrie Videos Strategic Direction Branding & Logos Power of the River History Book...

  2. Plumas-Sierra Rural Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    PlumasSierraREC Outage Hotline: (800) 555-2207 Outage Map: www.psrec.coopservice-area.ph References: EIA Form EIA-861 Final Data File for 2010 - File1a1 EIA Form 861 Data...

  3. Wright-Hennepin Coop Elec Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    3,398.952 34,434.725 40,549 1,975.209 23,239.907 4,920 388.083 5,035.346 51 5,762.244 62,709.978 45,520 2008-04 3,544.862 38,665.009 40,564 1,688.916 21,647.519 4,965 376.604...

  4. Arkansas Valley Elec Coop Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    3,460 776 15,071 13 7,532 101,140 52,031 2009-01 5,376 71,871 48,524 794 11,779 3,462 709 14,853 13 6,879 98,503 51,999 2008-12 4,441 59,651 48,487 736 11,437 3,460 541 15,069...

  5. Big Sandy Rural Elec Coop Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    www.bigsandyrecc.com Twitter: @bigsandycoop Facebook: https:www.facebook.compagesBig-Sandy-RECC142216049157162 Outage Hotline: 888-789-7322 Outage Map:...

  6. Central Texas Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    https:www.facebook.compagesCentral-Texas-Electric-Cooperative520773011297941?reftntnmn Outage Hotline: 1-800-900-2832 References: EIA Form EIA-861 Final Data File for...

  7. Southwest Iowa Rural Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    888-220-4869 Website: www.swiarec.coop Facebook: https:www.facebook.comswiarec?refhl Outage Hotline: (888) 220-4869 Outage Map: www.iowarec.orgoutages References: EIA...

  8. Four County Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    www.facebook.compagesFour-County-Electric-Membership-Corporation188316197857616?reftntnmn Outage Hotline: (888)368-7289 Outage Map: gis.fourcty.orgpubmap.html...

  9. Butler County Rural Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Iowa Phone Number: 888-267-2726 Website: www.butlerrec.coop Twitter: @ButlerCountyREC Facebook: https:www.facebook.combcrec Outage Hotline: 888-267-2726 Outage Map:...

  10. Golden Valley Elec Assn Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    36,891 1,877.823 13,175.914 6,327 8,171.591 69,159.555 448 14,556.482 111,832.096 43,666 2009-01 5,677.62 38,170.143 36,902 2,140.742 15,217.149 6,337 8,864.82 76,857.948 449...

  11. New York State Elec & Gas Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    76,528 8,823 123,751 1,276 219 2,550 4 97,366 787,691 737,183 2008-01 71,181 538,900 666,439 24,517 211,875 77,744 4,058 66,989 1,321 204 2,341 4 99,960 820,105 745,508...

  12. Maquoketa Valley Rrl Elec Coop | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Place: Iowa Phone Number: 319-462-3542 or 800-927-6068 Website: mvec.com Twitter: @mvecia Facebook: https:www.facebook.comMaquoketaValleyElectricCooperative Outage Hotline:...

  13. Osage Valley Elec Coop Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Outage Hotline: 660-679-3131 or 800-889-6832 Outage Map: ebill.osagevalley.comomsouta References: EIA Form EIA-861 Final Data File for 2010 - File1a1 EIA Form 861...

  14. Canadian Valley Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    anadianValleyElectric Outage Hotline: (855)875-7166 Outage Map: ebill.canadianvalley.orgomso References: EIA Form EIA-861 Final Data File for 2010 - File1a1 EIA Form 861 Data...

  15. Claverack Rural Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Outage Hotline: 1-800-326-9799 or 570-265-2167 Outage Map: ebill.claverack.comomsoutage References: EIA Form EIA-861 Final Data File for 2010 - File1a1 EIA Form 861...

  16. Monroe County Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Hours (618) 939-7171 or (800) 757-7433 or (866) 567-2759 Outage Map: ebill.mcec.orgomsoutageMap References: EIA Form EIA-861 Final Data File for 2010 - File1a1 EIA Form...

  17. Dakota Valley Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    1,101.081 16,334.085 5,106 390.564 6,162.524 588 1,168.972 25,110.935 332 2,660.617 47,607.544 6,026 2008-12 1,130.851 17,821.033 5,108 429.98 6,905.622 589 861.853 26,018.826...

  18. Verdigris Valley Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    ) TOT SALES (MWH) TOT CONS 2009-03 3,334 39,732 29,287 620 6,280 4,308 487 5,668 607 4,441 51,680 34,202 2009-02 3,065 36,726 29,285 456 4,469 4,299 405 4,606 607 3,926...

  19. Southern Indiana Gas & Elec Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    104,006.182 18,545 11,514.897 200,402.234 101 33,244.465 421,608.6 146,543 2008-02 12,607.003 129,571.861 128,066 9,445.235 104,704.602 18,561 11,374.157 198,519.29 100...

  20. Jones-Onslow Elec Member Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    56,736 2,076 24,111 4,679 10,879 107,097 61,415 2008-07 8,471 79,614 56,654 1,971 22,607 4,668 10,442 102,221 61,322 2008-06 6,356 61,755 56,244 1,774 20,036 4,651 8,130 81,791...

  1. Southern Indiana Gas & Elec Co | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    EIA Form EIA-861 Final Data File for 2010 - File1a1 Energy Information Administration Form 8262 EIA Form 861 Data Utility Id 17633 Utility Location Yes Ownership I...

  2. Columbia Basin Elec Cooperative, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    EIA-861 Final Data File for 2010 - File1a1 EIA Form 861 Data Utility Id 4005 Utility Location Yes Ownership C NERC Location WECC NERC WECC Yes Activity Transmission Yes...

  3. Guadalupe Valley Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    2010 - File1a1 EIA Form 861 Data Utility Id 7752 Utility Location Yes Ownership C NERC Location TRE NERC ERCOT Yes Activity Transmission Yes Activity Buying Transmission Yes...

  4. Houston County Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    2010 - File1a1 EIA Form 861 Data Utility Id 8898 Utility Location Yes Ownership C NERC Location TRE NERC ERCOT Yes Activity Transmission Yes Activity Distribution Yes Activity...

  5. Navasota Valley Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Website: www.navasotavalley.com Facebook: https:www.facebook.comnavasotavalley Outage Hotline: 1-800-443-9462 Outage Map: outages.navasotavalley.com:85 References: EIA...

  6. Central Florida Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Inc Place: Florida Phone Number: 1-800-227-1302 or 352-493-2511 Website: www.cfec.com Outage Hotline: 1-800-227-1302 or 352-493-2511 Outage Map: www.cfec.comoutage-mapsite...

  7. Mountain View Elec Assn, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Inc Place: Colorado Website: www.mvea.coop Facebook: https:www.facebook.comMVEAInc Outage Hotline: (800) 388-9881 Outage Map: outage.mvea.org References: EIA Form EIA-861...

  8. Roosevelt County Elec Coop Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    3,749 3,650 146 1,178 965 1,478 13,978 1,594 2,017 18,905 6,209 2008-06 357 3,560 3,638 132 1,092 961 1,347 13,188 1,579 1,836 17,840 6,178 2008-05 292 2,856 3,639 128 1,070 964...

  9. Southern Pine Elec Power Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    87,734 60,157 3,222 30,360 4,536 5,087 63,820 24 18,195 181,914 64,717 2008-05 6,897 62,132 60,058 2,887 27,862 4,522 4,430 56,228 24 14,214 146,222 64,604 2008-04 6,581 59,423...

  10. Calhoun County Elec Coop Assn | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Facebook: https:www.facebook.compagesCalhoun-County-REC173498466069004?skwall Outage Hotline: 800-821-4879 Outage Map: www.iowarec.orgoutages References: EIA Form...

  11. Paulding-Putman Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 14599 Utility Location Yes Ownership C NERC Location RFC NERC RFC Yes RTO PJM Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  12. Northeast Texas Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 13670 Utility Location Yes Ownership C NERC Location SPP NERC SPP Yes RTO SPP Yes Operates Generating Plant Yes Activity Generation Yes Activity Buying...

  13. Pioneer Rural Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 15054 Utility Location Yes Ownership C NERC Location RFC NERC RFC Yes RTO PJM Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  14. Northern Neck Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 13762 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes RTO PJM Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  15. Cookson Hills Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    861 Data Utility Id 4296 Utility Location Yes Ownership C NERC Location SPP NERC SPP Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  16. Western Farmers Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 20447 Utility Location Yes Ownership C NERC Location SPP NERC SPP Yes RTO SPP Yes Operates Generating Plant Yes Activity Generation Yes Activity Transmission Yes...

  17. Upshur Rural Elec Coop Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 19579 Utility Location Yes Ownership C NERC Location SPP NERC SPP Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  18. Jefferson Davis Elec Coop, Inc | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Data Utility Id 9682 Utility Location Yes Ownership C NERC Location SERC NERC SERC Yes RTO SPP Yes Activity Distribution Yes This article is a stub. You can help OpenEI by...

  19. RegIntlElecTrade_Eng_final.PDF

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    North America Regulation of International Electricity Trade prepared by North American Energy Working Group December 2002 2 The North American Energy Working Group The North American Energy Working Group (NAEW G) was established in spring of 2001 by the Canadian Minister of Natural Resources, the Mexican Secretary of Energy and the U.S. Secretary of Energy, to enhance North American energy cooperation. The Group is led by officials from Natural Resources Canada, the Mexican Secretariat of

  20. Employment | Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Employment Openings are posted on the UGA Human Resources website. To search for employment opportunities at SREL, select Department #267 (Savannah River Ecology Laboratory). UGA HR

  1. Smith River Rancheria- 2006 Project

    Office of Energy Efficiency and Renewable Energy (EERE)

    Smith River Rancheria has a strong commitment to becoming energy self-sufficient, reduce their energy costs, and stimulate economic development in the community.

  2. North Sky River | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Sky River Jump to: navigation, search Name North Sky River Facility North Sky River Sector Wind energy Facility Type Commercial Scale Wind Facility Status In Service Owner NextEra...

  3. River and Plateau Committee

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    December 2011) Page 1 Area RAP Committee Area of Interest Issue Manager(s) (*denotes lead) Other interested committee members Focus/Product For FY2012 Framing Questions/Issues (Articulated by Issue Managers) Cross- cutting River Corridor 100 & 300 Areas * 100 B/C Area * 100 K Area * 100 N Area * 100 D & H Areas * 100 F Area * 300 Area Shelley Cimon Dale Engstrom* Liz Mattson Jean Vanni Gerry Pollet Bob Suyama Wade Riggsbee 6 RODs RI/FS and Proposed Plans to be issued between now &

  4. River and Plateau Committee

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    2/15 River and Plateau Committee Priorities for advice on FY17 budget Not in priority order, numbering refers to last year's related advice points, per DOE response  (#1) The Board strongly urges DOE-Headquarters (HQ) to request full funding from Congress to meet all legal requirements of the ongoing cleanup work in FY 2016 and 2017 in addition to the following specific requests.  (#5) The Board advises DOE-RL to restore funding for removal and treatment of thousands of stored containers

  5. Independent Activity Report, Savannah River Remediation- July 2010

    Energy.gov [DOE]

    Savannah River Operations Office Integrated Safety Management System Phase II Verification Review of Savannah River Remediation

  6. Enforcement Letter, Westinghouse Savannah River Company- April 19, 2004

    Energy.gov [DOE]

    Issued to Westinghouse Savannah River Company related to Employee Reprisal at the Savannah River Site

  7. Enforcement Letter, Westinghouse Savannah River Company- November 14, 2003

    Energy.gov [DOE]

    Issued to Westinghouse Savannah River Company related to Criticality Safety Violations at the Savannah River Site

  8. Wild and Scenic Rivers | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Scenic Rivers Jump to: navigation, search Retrieved from "http:en.openei.orgwindex.php?titleWildandScenicRivers&oldid612228" Feedback Contact needs updating Image...

  9. Independent Oversight Inspection, Savannah River Site Office...

    Energy Savers

    Office - December 2009 Independent Oversight Inspection, Savannah River Site Office - December 2009 December 2009 Inspection of Nuclear Safety at the Savannah River Site Office and ...

  10. Savannah River Remediation (SRR) Expanded Staff Meeting

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Savannah River Remediation Delivering the Mission Dave Olson President and Project Manager ... Liquid Waste Operations contractor Savannah River Remediation LLC * Began work in ...

  11. Independent Oversight Inspection, Savannah River Site - January...

    Energy Savers

    Independent Oversight Inspection, Savannah River Site - January 2010 January 2010 Inspection of Emergency Management at the Savannah River Site This report provides the results of ...

  12. Green River Biodiesel Incorporated | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Biodiesel Incorporated Jump to: navigation, search Name: Green River Biodiesel Incorporated Place: Houston, Texas Zip: 77056 Product: Biodiesel project developer and...

  13. Beijing Haohua Rivers International Water Engineering Consulting...

    Open Energy Information (Open El) [EERE & EIA]

    Haohua Rivers International Water Engineering Consulting Co Ltd Jump to: navigation, search Name: Beijing Haohua Rivers International Water Engineering Consulting Co.Ltd. Place:...

  14. Flambeau River Biofuels Demonstration-Scale Biorefinery

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    in Wisconsin (NewPage Corporation in Wisconsin Rapids and Flambeau River Papers, LLC in Park Falls). NewPage and Flambeau River have demonstrated successful collaboration on...

  15. Grand River Dam Authority | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Dam Authority Place: Oklahoma Phone Number: 918-256-5545 Website: www.grda.com Twitter: @okgrda Facebook: https:www.facebook.compagesGrand-River-Dam-Authority...

  16. Independent Oversight Activity Report, Savannah River Site -...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    - February 2014 Independent Oversight Activity Report, Savannah River Site - February 2014 February 2014 Operational Awareness Visit of the Savannah River Site...

  17. River Hydrokinetic Resource Atlas | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    dress":"","icon":"","group":"","inlineLabel":"","visitedicon":"" Hide Map Language: English River Hydrokinetic Resource Atlas Screenshot References: EPRI1 River Atlas2 The...

  18. Savannah River Remediation, College Create Job Opportunities...

    Office of Environmental Management (EM)

    Remediation, College Create Job Opportunities for Graduates Savannah River Remediation, ... "With ongoing missions at the Savannah River Site and construction at Plant Vogtle and ...

  19. Kings River Conservation Dist | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Kings River Conservation Dist Jump to: navigation, search Name: Kings River Conservation Dist Place: California Phone Number: 559-237-5567 Website: www.krcd.org Facebook: https:...

  20. Voluntary Protection Program Onsite Review, Washington River...

    Energy Savers

    Washington River Protection Solutions, LLC, Hanford - Feb 2014 Voluntary Protection Program Onsite Review, Washington River Protection Solutions, LLC, Hanford - Feb 2014 February...

  1. Hood River Passive House

    SciTech Connect (OSTI)

    Hales, David

    2014-01-01

    The Hood River Passive Project was developed by Root Design Build of Hood River Oregon using the Passive House Planning Package (PHPP) to meet all of the requirements for certification under the European Passive House standards. The Passive House design approach has been gaining momentum among residential designers for custom homes and BEopt modeling indicates that these designs may actually exceed the goal of the U.S. Department of Energy's (DOE) Building America program to "reduce home energy use by 30%-50% (compared to 2009 energy codes for new homes). This report documents the short term test results of the Shift House and compares the results of PHPP and BEopt modeling of the project. The design includes high R-Value assemblies, extremely tight construction, high performance doors and windows, solar thermal DHW, heat recovery ventilation, moveable external shutters and a high performance ductless mini-split heat pump. Cost analysis indicates that many of the measures implemented in this project did not meet the BA standard for cost neutrality. The ductless mini-split heat pump, lighting and advanced air leakage control were the most cost effective measures. The future challenge will be to value engineer the performance levels indicated here in modeling using production based practices at a significantly lower cost.

  2. Hood River Passive House

    SciTech Connect (OSTI)

    Hales, D.

    2014-01-01

    The Hood River Passive Project was developed by Root Design Build of Hood River Oregon using the Passive House Planning Package (PHPP) to meet all of the requirements for certification under the European Passive House standards. The Passive House design approach has been gaining momentum among residential designers for custom homes and BEopt modeling indicates that these designs may actually exceed the goal of the U.S. Department of Energy's (DOE) Building America program to reduce home energy use by 30%-50% (compared to 2009 energy codes for new homes). This report documents the short term test results of the Shift House and compares the results of PHPP and BEopt modeling of the project. The design includes high R-Value assemblies, extremely tight construction, high performance doors and windows, solar thermal DHW, heat recovery ventilation, moveable external shutters and a high performance ductless mini-split heat pump. Cost analysis indicates that many of the measures implemented in this project did not meet the BA standard for cost neutrality. The ductless mini-split heat pump, lighting and advanced air leakage control were the most cost effective measures. The future challenge will be to value engineer the performance levels indicated here in modeling using production based practices at a significantly lower cost.

  3. Tensile and Fatigue Testing and Material Hardening Model Development for 508 LAS Base Metal and 316 SS Similar Metal Weld under In-air and PWR Primary Loop Water Conditions

    SciTech Connect (OSTI)

    Mohanty, Subhasish; Soppet, William; Majumdar, Saurin; Natesan, Ken

    2015-09-01

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in September 2015 under the work package for environmentally assisted fatigue under DOE’s Light Water Reactor Sustainability program. In an April 2015 report we presented a baseline mechanistic finite element model of a two-loop pressurized water reactor (PWR) for systemlevel heat transfer analysis and subsequent thermal-mechanical stress analysis and fatigue life estimation under reactor thermal-mechanical cycles. In the present report, we provide tensile and fatigue test data for 508 low-alloy steel (LAS) base metal, 508 LAS heat-affected zone metal in 508 LAS–316 stainless steel (SS) dissimilar metal welds, and 316 SS-316 SS similar metal welds. The test was conducted under different conditions such as in air at room temperature, in air at 300 oC, and under PWR primary loop water conditions. Data are provided on materials properties related to time-independent tensile tests and time-dependent cyclic tests, such as elastic modulus, elastic and offset strain yield limit stress, and linear and nonlinear kinematic hardening model parameters. The overall objective of this report is to provide guidance to estimate tensile/fatigue hardening parameters from test data. Also, the material models and parameters reported here can directly be used in commercially available finite element codes for fatigue and ratcheting evaluation of reactor components under in-air and PWR water conditions.

  4. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    SciTech Connect (OSTI)

    De Rosa, Felice [ENEA, Italian National Agency for New Technologies, Energy and the Environment (Italy)

    2006-07-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when {delta}Tsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  5. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    SciTech Connect (OSTI)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  6. Concept Paper Savannah River Nuclear Solutions, LLC Savannah River Site

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Paper Savannah River Nuclear Solutions, LLC Savannah River Site Aiken, SC 29808 Michael S. Navetta, PE Manager- Energy Park Initiative (803) 952-8806 michael.navetta@srs.gov U.S. EnergyFreedomCenter PREDECISIONAL DRAFT Today We Can Start To Unshackle America Decades of debate for ending America's dependence on foreign fossil fuels, climate change and environmentally positive energy has produced a myriad of technologies that independently offer a partial solution. Applying existing technologies

  7. Savannah River Site | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Savannah River Site Savannah River Site Savannah River Site | June 2011 Aerial View Savannah River Site | June 2011 Aerial View Savannah River Site (SRS) has mission responsibilities in nuclear weapons stockpile stewardship by ensuring the safe and reliable management of tritium resources; by contributing to the stockpile surveillance program; and by assisting in the development of alternatives for large-scale pit disassembly/conversion capability. SRS also manages excess nuclear materials and

  8. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.; Wright, J.B.

    1980-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  9. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (1.4 kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.

    1981-09-01

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.4 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a stainless steel canister representative of actual fuel canisters, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel near-surface drywell tests being conducted at E-MAD, the spent fuel deep geologic storage test being conducted in Climax granite on the Nevada Test Site, and for five constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  10. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 5. Probabilistic fracture mechanics analysis. Load Combination Program Project I final report

    SciTech Connect (OSTI)

    Harris, D.O.; Lim, E.Y.; Dedhia, D.D.

    1981-06-01

    The primary purpose of the Load Combination Program covered in this report is to estimate the probability of a seismic induced LOCA in the primary piping of a commercial pressurized water reactor (PWR). Best estimates, rather than upper bound results are desired. This was accomplished by use of a fracture mechanics model that employs a random distribution of initial cracks in the piping welds. Estimates of the probability of cracks of various sizes initially existing in the welds are combined with fracture mechanics calculations of how these cracks would grow during service. This then leads to direct estimates of the probability of failure as a function of time and location within the piping system. The influence of varying the stress history to which the piping is subjected is easily determined. Seismic events enter into the analysis through the stresses they impose on the pipes. Hence, the influence of various seismic events on the piping failure probability can be determined, thereby providing the desired information.

  11. The role of Hydrogen and Creep in Intergranular Stress Corrosion Cracking of Alloy 600 and Alloy 690 in PWR Primary Water Environments ? a Review

    SciTech Connect (OSTI)

    Rebak, R B; Hua, F H

    2004-07-12

    Intergranular attack (IGA) and intergranular stress corrosion cracking (IGSCC) of Alloy 600 in PWR steam generator environment has been extensively studied for over 30 years without rendering a clear understanding of the essential mechanisms. The lack of understanding of the IGSCC mechanism is due to a complex interaction of numerous variables such as microstructure, thermomechanical processing, strain rate, water chemistry and electrochemical potential. Hydrogen plays an important role in all these variables. The complexity, however, significantly hinders a clearer and more fundamental understanding of the mechanism of hydrogen in enhancing intergranular cracking via whatever mechanism. In this work, an attempt is made to review the role of hydrogen based on the current understanding of grain boundary structure and chemistry and intergranular fracture of nickel alloys, effect of hydrogen on electrochemical behavior of Alloy 600 and Alloy 690 (e.g. the passive film stability, polarization behavior and open-circuit potential) and effect of hydrogen on PWSCC behavior of Alloy 600 and Alloy 690. Mechanistic studies on the PWSCC are briefly reviewed. It is concluded that further studies on the role of hydrogen on intergranular cracking in both inert and primary side environments are needed. These studies should focus on the correlation of the results obtained at different laboratories by different methods on materials with different metallurgical and chemical parameters.

  12. Raft River Idaho Magnetotelluric Data

    SciTech Connect (OSTI)

    Gregory Nash

    2015-05-13

    Raw magnetotelluric (MT) data covering the geothermal system at Raft River, Idaho. The data was acquired by Quantec Geoscience. This is a zipped file containing .edi raw MT data files.

  13. Dayao County Yupao River BasDayao County Yupao River Basin Hydro...

    Open Energy Information (Open El) [EERE & EIA]

    Dayao County Yupao River BasDayao County Yupao River Basin Hydro electricity Development Co Ltd in Jump to: navigation, search Name: Dayao County Yupao River BasDayao County Yupao...

  14. Savannah River Site: Plutonium Preparation Project (PuPP) at Savannah River

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Site | Department of Energy Site: Plutonium Preparation Project (PuPP) at Savannah River Site Savannah River Site: Plutonium Preparation Project (PuPP) at Savannah River Site Full Document and Summary Versions are available for download Savannah River Site: Plutonium Preparation Project (PuPP) at Savannah River Site (13.39 MB) Summary - Plutonium Preparation Project at the Savannah River Site (53.49 KB) More Documents & Publications EIS-0220: Final Environmental Impact Statement

  15. Snake River Plain Geothermal Region | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Survey At Snake River Plain Region (DOE GTP) Micro-Earthquake At Snake River Plain Geothermal Region (1976) Reflection Survey At Snake River Plain Region (DOE GTP)...

  16. Voluntary Protection Program Onsite Review, Savannah River Remediation...

    Office of Environmental Management (EM)

    River Remediation, Llc, Liquid Waste Contract, Savannah River Site - November 2014 Voluntary Protection Program Onsite Review, Savannah River Remediation, Llc, Liquid Waste ...

  17. Savannah River Site - L-Area Southern Groundwater | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    L-Area Southern Groundwater Savannah River Site - L-Area Southern Groundwater January 1, ... InstallationName, State: Savannah River Site, SC Responsible DOE Office: Savannah River ...

  18. Savannah River Site - C-Area Groundwater Operable Unit | Department...

    Office of Environmental Management (EM)

    C-Area Groundwater Operable Unit Savannah River Site - C-Area Groundwater Operable Unit ... InstallationName, State: Savannah River Site, SC Responsible DOE Office: Savannah River ...

  19. Savannah River Site - R-Area Groundwater Operable Unit | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    R-Area Groundwater Operable Unit Savannah River Site - R-Area Groundwater Operable Unit ... InstallationName, State: Savannah River Site, SC Responsible DOE Office: Savannah River ...

  20. BLM Humboldt River Field Office | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    River Field Office Jump to: navigation, search Name: BLM Humboldt River Field Office Abbreviation: Humboldt River Address: 5100 E. Winnemucca Blvd. Place: Winnemucca, Nevada Zip:...

  1. PP-366 Twin Rivers Paper Company, Inc. | Department of Energy

    Office of Environmental Management (EM)

    6 Twin Rivers Paper Company, Inc. PP-366 Twin Rivers Paper Company, Inc. Presidential Permit authorizing Twin Rivers Paper Company, Inc. to construct, operate, and maintain ...

  2. Independent Oversight Activity Report, Office of River Protection...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Office of River Protection - May 2013 Independent Oversight Activity Report, Office of River Protection - May 2013 May 2013 Operational Awareness Visit at the Office of River...

  3. EA-1692: Red River Environmental Products, LLC Activated Carbon...

    Office of Environmental Management (EM)

    2: Red River Environmental Products, LLC Activated Carbon Manufacturing Facility, Red River Parish, LA EA-1692: Red River Environmental Products, LLC Activated Carbon Manufacturing ...

  4. EA-273 Rainy River Energy Corporation | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Rainy River Energy Corporation EA-273 Rainy River Energy Corporation Order authorizing Rainy River Energy Corporation to export electric energy to Canada. PDF icon EA-273 Rainy ...

  5. Savannah River National Laboratory (SRNL) | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    (Redirected from Savannah River National Laboratory) Jump to: navigation, search Logo: Savannah River National Laboratory Name: Savannah River National Laboratory Place: Aiken,...

  6. Natural Gas Resources of the Greater Green River and Wind River...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Natural Gas Resources of the Greater Green River and Wind River Basins ... Resource Type: Technical Report Research Org: National Energy Technology Laboratory, ...

  7. Green River, Utah, Disposal Site Fact Sheet

    Office of Legacy Management (LM)

    Green River, Utah, Disposal Site This fact sheet provides information about the Uranium Mill Tailings Radiation Control Act of 1978 Title I disposal site near Green River, Utah. This site is managed by the U.S. Department of Energy Office of Legacy Management. Location of the Green River, Utah, Disposal Site Site Description and History The Green River disposal site is about 0.5 mile east of the Green River and 1.5 miles southeast of the city of Green River, Utah. The site consists of an

  8. THE SNAKE RIVER PLAIN AQUIFER THE SNAKE RIVER PLAIN AQUIFER

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    aquifer THE INL & THE SNAKE RIVER PLAIN AQUIFER THE SNAKE RIVER PLAIN AQUIFER underneath the Idaho National Laboratory is one of the most productive groundwater resources in the U.S. Each year about 2 million acre-feet of water is drawn from the aquifer. Approximately 95 percent of the water withdrawn from the aquifer is used for irrigation, 3 per- cent for domestic water, and 2 percent for industrial purposes. The aquifer is the primary water source for more than 280,000 people in

  9. Snake River Geothermal Consortium FORGE Logo | Department of Energy

    Energy.gov (indexed) [DOE]

    Logo More Documents & Publications Snake River Geothermal Consortium FORGE Logo Snake River Geothermal Consortium FORGE Map Snake River Geothermal Consortium FORGE Logo Milford, Utah FORGE Logo Snake River Geothermal Consortium FORGE Logo West Flank FORGE Logo

  10. Grays River Watershed Geomorphic Analysis

    SciTech Connect (OSTI)

    Geist, David R.

    2005-04-30

    This investigation, completed for the Pacific Northwest National Laboratory (PNNL), is part of the Grays River Watershed and Biological Assessment commissioned by Bonneville Power Administration under project number 2003-013-00 to assess impacts on salmon habitat in the upper Grays River watershed and present recommendations for habitat improvement. This report presents the findings of the geomorphic assessment and is intended to support the overall PNNL project by evaluating the following: 􀂃 The effects of historical and current land use practices on erosion and sedimentation within the channel network 􀂃 The ways in which these effects have influenced the sediment budget of the upper watershed 􀂃 The resulting responses in the main stem Grays River upstream of State Highway 4 􀂃 The past and future implications for salmon habi

  11. Enforcement Letter, Savannah River Ecology Laboratory- June 7, 2000

    Energy.gov [DOE]

    Issued to Savannah River Ecology Laboratory related to Radioactive Material Control Deficiencies at the Savannah River Site

  12. Conceptual Model At Raft River Geothermal Area (1988) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Conceptual Model At Raft River Geothermal Area (1988) Exploration Activity Details Location Raft River...

  13. Conceptual Model At Raft River Geothermal Area (1977) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Conceptual Model At Raft River Geothermal Area (1977) Exploration Activity Details Location Raft River...

  14. Field Mapping At Raft River Geothermal Area (1977) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Field Mapping At Raft River Geothermal Area (1977) Exploration Activity Details Location Raft River...

  15. Geophysical Method At Raft River Geothermal Area (1975) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Geophysical Method At Raft River Geothermal Area (1975) Exploration Activity Details Location Raft River...

  16. Field Mapping At Raft River Geothermal Area (1980) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Field Mapping At Raft River Geothermal Area (1980) Exploration Activity Details Location Raft River...

  17. Microearthquake surveys of Snake River plain and Northwest Basin...

    Open Energy Information (Open El) [EERE & EIA]

    microearthquakes; Nevada; North America; passive systems; Pershing County Nevada; Raft River; reservoir rocks; seismic methods; seismicity; seismology; Snake River plain;...

  18. Core Analysis At Raft River Geothermal Area (1981) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Core Analysis At Raft River Geothermal Area (1981) Exploration Activity Details Location Raft River...

  19. Field Mapping At Raft River Geothermal Area (1990) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Field Mapping At Raft River Geothermal Area (1990) Exploration Activity Details Location Raft River...

  20. Conceptual Model At Raft River Geothermal Area (1987) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Conceptual Model At Raft River Geothermal Area (1987) Exploration Activity Details Location Raft River...

  1. Conceptual Model At Raft River Geothermal Area (1990) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Conceptual Model At Raft River Geothermal Area (1990) Exploration Activity Details Location Raft River...

  2. Conceptual Model At Raft River Geothermal Area (1983) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Conceptual Model At Raft River Geothermal Area (1983) Exploration Activity Details Location Raft River...

  3. Aeromagnetic Survey At Raft River Geothermal Area (1981) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Aeromagnetic Survey At Raft River Geothermal Area (1981) Exploration Activity Details Location Raft River...

  4. Core Analysis At Raft River Geothermal Area (1976) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Core Analysis At Raft River Geothermal Area (1976) Exploration Activity Details Location Raft River...

  5. Geophysical Method At Raft River Geothermal Area (1977) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Geophysical Method At Raft River Geothermal Area (1977) Exploration Activity Details Location Raft River...

  6. Exploratory Well At Raft River Geothermal Area (1977) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Exploratory Well At Raft River Geothermal Area (1977) Exploration Activity Details Location Raft River...

  7. Exploratory Well At Raft River Geothermal Area (1975) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Exploratory Well At Raft River Geothermal Area (1975) Exploration Activity Details Location Raft River...

  8. Tracer Testing At Raft River Geothermal Area (1983) | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Tracer Testing At Raft River Geothermal Area (1983) Exploration Activity Details Location Raft River...

  9. Aeromagnetic Survey At Raft River Geothermal Area (1978) | Open...

    Open Energy Information (Open El) [EERE & EIA]

    search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Aeromagnetic Survey At Raft River Geothermal Area (1978) Exploration Activity Details Location Raft River...

  10. Platte River Power Authority | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    search Name: Platte River Power Authority Place: Colorado Website: www.prpa.org Facebook: https:www.facebook.comPlatteRiverPower Outage Hotline: 1-888-748-5113 References:...

  11. Research | Savannah River National Environmental Park

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Over the years research on the Savannah River NERP has provided many insights into human ... research on the SRS and the establishment of the Savannah River Ecology Laboratory (SREL). ...

  12. Savannah River Site - TNX | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Savannah River Site - TNX January 1, 2013 - 12:00pm Addthis US Department of Energy Groundwater Database Groundwater Master Report InstallationName, State: Savannah River Site, SC ...

  13. Overview | Savannah River National Environmental Park

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Park Overview The Savannah River Site is an 803-km2 Department of Energy (DOE) facility ... of South Carolina, near Aiken, SC. The site is bordered on one side by the Savannah River. ...

  14. Savannah River Site | National Nuclear Security Administration...

    National Nuclear Security Administration (NNSA)

    Savannah River Site NNSA operates facilities at the Savannah River Site to supply and process tritium, a radioactive form of hydrogen that is a key component of nuclear weapons. ...

  15. Preliminary Notice of Violation, Westinghouse Savannah River...

    Energy.gov (indexed) [DOE]

    River Company - EA-97-11 Preliminary Notice of Violation, Westinghouse Savannah River Company - EA 98-09 Preliminary Notice of Violation, Kaiser-Hill Company, LLC - EA-1999-06...

  16. An Inside Look at River Corridor

    Energy.gov [DOE]

    In the seventh chapter of The Handford Story, the Energy Department takes a look at the River Corridor -- a 50-mile stretch of the Columbia River that flows through the Hanford site in southeast...

  17. River Corridor Closure Project Partnering Performance Agreement

    Energy.gov [DOE]

    WCH and DOE have a mission to complete the clsoure of the Hanford River Corridor by 2015.  Early and efficient completion of this work scope law the River Corridor Closure Contract (DE-AC06...

  18. Red River Biodiesel Ltd | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Ltd Jump to: navigation, search Name: Red River Biodiesel, Ltd. Place: Houston, Texas Zip: 77006 Product: Red River operates a biodiesel plant in Houstion, Texas with a capacity of...

  19. First Savannah River Shipment Arrives At WIPP

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    First Savannah River Site Shipment Arrives At WIPP CARLSBAD, N.M., May 10, 2001 - The U.S. ... from DOE's Savannah River Site has arrived at the Waste Isolation Pilot Plant (WIPP). ...

  20. New Savannah River Site Deputy Manager Named

    Energy.gov [DOE]

    AIKEN, S.C. – DOE’s Savannah River Operations Office selected Terrel “Terry” J. Spears as the deputy manager of the Savannah River Site (SRS) this month.

  1. Columbia River Component Data Evaluation Summary Report

    SciTech Connect (OSTI)

    C.S. Cearlock

    2006-08-02

    The purpose of the Columbia River Component Data Compilation and Evaluation task was to compile, review, and evaluate existing information for constituents that may have been released to the Columbia River due to Hanford Site operations. Through this effort an extensive compilation of information pertaining to Hanford Site-related contaminants released to the Columbia River has been completed for almost 965 km of the river.

  2. Categorical Exclusion Determinations: Savannah River Operations Office |

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Department of Energy Savannah River Operations Office Categorical Exclusion Determinations: Savannah River Operations Office Categorical Exclusion Determinations issued by Savannah River Operations Office. DOCUMENTS AVAILABLE FOR DOWNLOAD February 29, 2016 CX-014636: Categorical Exclusion Determination Micro-encapsulation of PuO2 Surrogates in a Low-water Cement-based Waste Form CX(s) Applied: B3.6 Date: 02/29/2016 Location(s): South Carolina Offices(s): Savannah River Operations Office

  3. Publications | Savannah River National Environmental Park

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Savannah River NERP Program Publications The Snakes of the Savannah River Plant with Information about Snakebite Prevention and Treatment. J. Whitfield Gibbons. 1977. SRO-NERP-1. 26 p. [Download PDF] The Reptiles and Amphibians of the Savannah River Plant. J. Whitfield Gibbons and Karen K. Patterson. 1978. SRO-NERP-2. 24p. [Download PDF] The Freshwater Bivalve Mollusca (Unionidae, Sphaeriidae, Corbiculidae) of the Savannah River Plant, South Carolina. Joseph C. Britton and Samuel L. H. Fuller.

  4. Savannah River Needs Assessment | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Savannah River Needs Assessment Savannah River Needs Assessment June 23, 1998 This Needs Assessment for former Savannah River Site construction workers was developed for the purpose of collecting existing information relevant to exposure and health outcomes among former workers, utilizing this information to develop viable methods for contacting these former workers, and providing an initial determination of the most significant worker hazards, problems, and concerns for the site. Savannah River

  5. Bayer Material Science (TRL 1 2 3 System)- River Devices to Recover Energy with Advanced Materials(River DREAM)

    Energy.gov [DOE]

    Bayer Material Science (TRL 1 2 3 System) - River Devices to Recover Energy with Advanced Materials(River DREAM)

  6. Ecotoxicology | Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Savannah River NERP Research Opportunities Field Sites / Data Research Facilities Low Dose Irradiation Facility Tritium Irrigation Facility Microsatellite Development Education Graduate Undergraduate Radioecology Curriculum Outreach Outreach Program SREL Herpetology Information for Visitors SREL Conference Center Maps and Directions Airports and Lodging Security Requirements Reprints Employment Make a Gift University of Georgia logo University of Georgia search SREL GO powered by Google

  7. Smith River Rancheria- 2005 Project

    Office of Energy Efficiency and Renewable Energy (EERE)

    The Smith River Rancheria located just south of the Oregon border, on the coast, knows that developing a power-generation facility is a crucial part of its multifaceted plan for both economic development and reduction of energy costs. The activities associated with this grant will include the basic steps necessary to determine whether a generation facility will, in fact, benefit the tribe.

  8. HANFORD SITE RIVER CORRIDOR CLEANUP

    SciTech Connect (OSTI)

    BAZZELL, K.D.

    2006-02-01

    In 2005, the US Department of Energy (DOE) launched the third generation of closure contracts, including the River Corridor Closure (RCC) Contract at Hanford. Over the past decade, significant progress has been made on cleaning up the river shore that bordes Hanford. However, the most important cleanup challenges lie ahead. In March 2005, DOE awarded the Hanford River Corridor Closure Contract to Washington Closure Hanford (WCH), a limited liability company owned by Washington Group International, Bechtel National and CH2M HILL. It is a single-purpose company whose goal is to safely and efficiently accelerate cleanup in the 544 km{sup 2} Hanford river corridor and reduce or eliminate future obligations to DOE for maintaining long-term stewardship over the site. The RCC Contract is a cost-plus-incentive-fee closure contract, which incentivizes the contractor to reduce cost and accelerate the schedule. At $1.9 billion and seven years, WCH has accelerated cleaning up Hanford's river corridor significantly compared to the $3.2 billion and 10 years originally estimated by the US Army Corps of Engineers. Predictable funding is one of the key features of the new contract, with funding set by contract at $183 million in fiscal year (FY) 2006 and peaking at $387 million in FY2012. Another feature of the contract allows for Washington Closure to perform up to 40% of the value of the contract and subcontract the balance. One of the major challenges in the next few years will be to identify and qualify sufficient subcontractors to meet the goal.

  9. Savannah River Plant/Savannah River Laboratory radiation exposure report

    SciTech Connect (OSTI)

    Rogers, C.D.; Hyman, S.D.; Keisler, L.L. and Co., Aiken, SC . Savannah River Plant); Reeder, D.F.; Jolly, L.; Spoerner, M.T.; Schramm, G.R. and Co., Aiken, SC . Savannah River Lab.)

    1989-01-01

    The protection of worker health and safety is of paramount concern at the Savannah River Site. Since the site is one of the largest nuclear sites in the nation, radiation safety is a key element in the protection program. This report is a compendium of the results in 1988 of the programs at the Savannah River Plant and the Savannah River Laboratory to protect the radiological health of employees. By any measure, the radiation protection performance at this site in 1988 was the best since the beginning of operations. This accomplishment was made possible by the commitment and support at all levels of the organizations to reduce radiation exposures to ALARA (As Low As Reasonably Achievable). The report provides detailed information about the radiation doses received by departments and work groups within these organizations. It also includes exposure data for recent years to allow Plant and Laboratory units to track the effectiveness of their ALARA efforts. Many of the successful practices and methods that reduced radiation exposure are described. A new goal for personnel contamination cases has been established for 1989. Only through continual and innovative efforts to minimize exposures can the goals be met. The radiation protection goals for 1989 and previous years are included in the report. 27 figs., 58 tabs.

  10. Savannah River Laboratory Decontamination Program

    SciTech Connect (OSTI)

    Rankin, W.N.

    1991-01-01

    Savannah River Laboratory (SRL) has had a Decontamination and Decommissioning (D D) Technology program since 1981. The objective of this program is to provide state-of-the-art technology for use in D D operations that will enable our customers to minimize waste generated and personal exposure, increase productivity and safety, and to minimize the potential for release and uptake of radioactive material. The program identifies and evaluates existing technology, develops new technology, and provides technical assistance to implement its use onsite. This program has impacted not only the Savannah River Site (SRS), but the entire Department of Energy (DOE) complex. To document and communicate the technology generated by this program, 28 papers have been presented at National and International meetings in the United States and Foreign Countries.

  11. Savannah River Laboratory Decontamination Program

    SciTech Connect (OSTI)

    Rankin, W.N.

    1991-12-31

    Savannah River Laboratory (SRL) has had a Decontamination and Decommissioning (D&D) Technology program since 1981. The objective of this program is to provide state-of-the-art technology for use in D&D operations that will enable our customers to minimize waste generated and personal exposure, increase productivity and safety, and to minimize the potential for release and uptake of radioactive material. The program identifies and evaluates existing technology, develops new technology, and provides technical assistance to implement its use onsite. This program has impacted not only the Savannah River Site (SRS), but the entire Department of Energy (DOE) complex. To document and communicate the technology generated by this program, 28 papers have been presented at National and International meetings in the United States and Foreign Countries.

  12. Raft River geoscience case study

    SciTech Connect (OSTI)

    Dolenc, M.R.; Hull, L.C.; Mizell, S.A.; Russell, B.F.; Skiba, P.A.; Strawn, J.A.; Tullis, J.A.

    1981-11-01

    The Raft River Geothermal Site has been evaluated over the past eight years by the United States Geological Survey and the Idaho National Engineering Laboratory as a moderate-temperature geothermal resource. The geoscience data gathered in the drilling and testing of seven geothermal wells suggest that the Raft River thermal reservoir is: (a) produced from fractures found at the contact metamorphic zone, apparently the base of detached normal faulting from the Bridge and Horse Well Fault zones of the Jim Sage Mountains; (b) anisotropic, with the major axis of hydraulic conductivity coincident to the Bridge Fault Zone; (c) hydraulically connected to the shallow thermal fluid of the Crook and BLM wells based upon both geochemistry and pressure response; (d) controlled by a mixture of diluted meteoric water recharging from the northwest and a saline sodium chloride water entering from the southwest. Although the hydrogeologic environment of the Raft River geothermal area is very complex and unique, it is typical of many Basin and Range systems.

  13. U. S. Department of Energy Savannah River Operations Office - Tenant

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Organizations Organizations Tenant Organizations Federal Offices National Nuclear Security Administration External Link Office of Environmental Management External Link U. S. Forest Service - Savannah River External Link Contractor Organizations Savannah River Nuclear Solutions External Link Savannah River National Laboratory External Link Savannah River Remediation External Link Centerra University of Georgia - Savannah River Ecology Laboratory External Link Shaw AREVA MOX Services External

  14. Flambeau_River_Biofuels.pdf | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Flambeau_River_Biofuels.pdf Flambeau_River_Biofuels.pdf Flambeau_River_Biofuels.pdf Flambeau_River_Biofuels.pdf (26 KB) More Documents & Publications Pacific Ethanol, Inc Flambeau River Biofuels Demonstration-Scale Biorefinery NewPage Demonstration-Scale Biorefinery

  15. The Columbia River System : the Inside Story.

    SciTech Connect (OSTI)

    United States. Bonneville Power Administration.

    1991-09-01

    The Columbia Ricer is one of the greatest natural resources in the western United States. The river and its tributaries touch the lives of nearly every resident of the Northwest-from providing the world-famous Pacific salmon to supplying the clean natural fuel for over 75 percent of the region's electrical generation. Since early in the century, public and private agencies have labored to capture the benefits of this dynamic river. Today, dozens of major water resource projects throughout the region are fed by the waters of the Columbia Basin river system. And through cooperative efforts, the floods that periodically threaten developments near the river can be controlled. This publication presents a detailed explanation of the planning and operation of the multiple-use dams and reservoirs of the Columbia River system. It describes the river system, those who operate and use it, the agreements and policies that guide system operation, and annual planning for multiple-use operation.

  16. DOE Selects Savannah River Remediation, LLC for Liquid Waste Contract at Savannah River Site

    Energy.gov [DOE]

    Washington, D.C.  -The U.S. Department of Energy (DOE) today announced the award to Savannah River Remediation, LLC as the liquid waste contractor for DOE's Savannah River Site (SRS) in Aiken,...

  17. Deep drilling data, Raft River geothermal area, Idaho-Raft River...

    Open Energy Information (Open El) [EERE & EIA]

    data, Raft River geothermal area, Idaho-Raft River geothermal exploration well sidetrack-C Jump to: navigation, search OpenEI Reference LibraryAdd to library Report: Deep drilling...

  18. Savannah River Site Vegetation Map | Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Set-Aside Program SREL HOME Savannah River Site Vegetation Map swatch 1. Industrial swatch 2. Open water swatch 3. Bare soil / bare surface swatch 4. Sparse herbaceous vegetation swatch 5. Grasses and forbs swatch 6. Shrubs, grasses, and forbs swatch 7. Disturbed and revegetated in 1997 swatch 8. Marsh / aquatic macrophytes swatch 9. Young, open-canopy loblolly pine swatch 10. Open-canopy loblolly pine swatch 11. Young, dense-canopy loblolly pine swatch 12. Dense-canopy loblolly pine swatch 13.

  19. Preliminary Notice of Violation, Westinghouse Savannah River...

    Office of Environmental Management (EM)

    concerning the unnecessary radiation exposure of three Westinghouse Savannah River Company (WSRC) personnel and the subsequent falsification of radiation dose records. ...

  20. Enforcement Letter, Westinghouse Savannah River Company - November...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Company related to repeated criticality safety violations in 2002 and 2003 at the H-Canyon facility at DOE's Savannah River Site. PDF icon Enforcement Letter, Westinghouse...

  1. The Columbia River System Inside Story

    SciTech Connect (OSTI)

    2001-04-01

    The Columbia River is one of the greatest natural resources in the western United States. The river and its tributaries touch the lives of nearly every resident of the Pacific Northwest—from fostering world-famous Pacific salmon to supplying clean natural fuel for 50 to 65 percent of the region’s electrical generation. Since early in the 20th century, public and private agencies have labored to capture the benefits of this dynamic river. Today, dozens of major water resource projects throughout the region are fed by the waters of the Columbia Basin river system.

  2. Big Rivers Electric Corp | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    (270) 827-2561 Website: www.bigrivers.com Facebook: https:www.facebook.compagesBig-Rivers-Electric-Corporation142180855818082?rf154289971250771 Outage Hotline: (270)...

  3. Big River Resources LLC | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Resources LLC Jump to: navigation, search Name: Big River Resources LLC Place: West Burlington, Iowa Zip: 52655 Product: Dry-mill bioethanol producer with a cooperative structure....

  4. Cuivre River Electric- Energy Efficiency Rebate Programs

    Energy.gov [DOE]

    Cuivre River Electric Cooperative, through the Take Control & Save program, offers rebates for cooperative members who purchase efficient geothermal and dual fuel heat pumps, and electric water...

  5. Rebecca Sharitz | Savannah River Ecology Laboratory

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Faculty & Scientists SREL Home UGA Plant Biology Rebecca Sharitz Savannah River Ecology ... We are also conducting studies on the population biology and conservation of rare plants, ...

  6. Lumbee River EMC- Residential Energy Efficiency Program

    Energy.gov [DOE]

    Lumbee River EMC (LREMC) offers rebates to its residential customers who purchase and install qualified energy efficient products or services. Rebates are available for:

  7. Savannah River Laboratory monthly report, November 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-12-31

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  8. Savannah River Laboratory monthly report, November 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-01-01

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  9. Savannah River Laboratory monthly report, July 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-01-01

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  10. Savannah River Laboratory monthly report, July 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-12-31

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation; tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  11. Brazos River Authority | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Place: Texas Phone Number: 1-888-922-6272 Website: www.brazos.org Facebook: https:www.facebook.compagesBrazos-River-Authority126719790675809?frefts Outage Hotline:...

  12. Independent Oversight Review, Savannah River Operations Office...

    Office of Environmental Management (EM)

    - July 2013 July 2013 Review of the Employee Concerns Program at the Savannah River ... O 442.1, DEPARTMENT OF ENERGY EMPLOYEE CONCERNS PROGRAM Independent Activity ...

  13. Two Rivers Water & Light | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Water & Light Jump to: navigation, search Name: Two Rivers Water & Light Place: Wisconsin Phone Number: (920) 793-5550 Website: trwaterandlight.com Facebook: https:...

  14. River Valley Technology Center | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Valley Technology Center Jump to: navigation, search Name: River Valley Technology Center Place: United States Sector: Services Product: General Financial & Legal Services (...

  15. Savannah River Laboratory monthly report, October 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-12-31

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separations operations; environmental concerns; and waste management. (FI)

  16. Lower Colorado River Authority | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

    Name: Lower Colorado River Authority Place: Texas Website: www.lcra.orgPagesdefault.asp Twitter: @lcra Facebook: https:www.facebook.comlowercoloradoriverauthority Outage...

  17. Savannah River Laboratory monthly report, August 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-12-31

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  18. Savannah River Laboratory monthly report, August 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-01-01

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  19. Gila River Indian Community- 2012 Project

    Energy.gov [DOE]

    The Gila River Indian Community (GRIC) will conduct feasibility studies of potential renewable energy projects on its lands in south central Arizona.

  20. Power of the River History Book

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Power-of-the-River-BPA-History-Book Sign In About | Careers | Contact | Investors | bpa.gov Search News & Us Expand News & Us Projects & Initiatives Expand Projects &...

  1. Savannah River Laboratory monthly report, October 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-01-01

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separations operations; environmental concerns; and waste management. (FI)

  2. Preliminary Notice of Violation, Westinghouse Savannah River...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Achievable (ALARA) deficiencies that contributed to unplanned worker uptakes and the spread of contamination at DOE's Savannah River Site. PDF icon Preliminary Notice of...

  3. New Columbia River Estuary purchases benefit salmon

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    the mouth of the Columbia River to permanently protect riverside habitat for Northwest fish and wildlife, including threatened and endangered salmon and steelhead. The...

  4. Brochure: Federal Columbia River Power System (FCRPS)

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    these agencies maximize the use of the Columbia River by generating power, protecting fish and wildlife, controlling floods, providing irrigation and navigation, and sustaining...

  5. Savannah River Laboratory monthly report, September 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-01-01

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  6. Savannah River Laboratory monthly report, September 1991

    SciTech Connect (OSTI)

    Ferrell, J.M.

    1991-12-31

    This document details monthly activities at the Savannah River Laboratory. Topics addressed are reactor operation, tritium facilities and production; separation operations; environmental concerns; and waste management. (FI)

  7. Independent Oversight Review, Savannah River Site - September...

    Office of Environmental Management (EM)

    at the Savannah River Site Environmental Management Nuclear Facilities This report provides the ... Management Evaluations, within the DOE Office of Health, Safety and Security. ...

  8. Beijing Changjiang River International Holding | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

    100761 Sector: Services Product: Beijing Changjiang River International Holding is a Chinese emissions broker and services company. Coordinates: 39.90601, 116.387909 Show Map...

  9. ENVIRONMENTAL SCIENCES; SAVANNAH RIVER PLANT; ENVIRONMENTAL EFFECTS...

    Office of Scientific and Technical Information (OSTI)

    5 audit of SRP radioactive waste Ashley, C. 05 NUCLEAR FUELS; 54 ENVIRONMENTAL SCIENCES; SAVANNAH RIVER PLANT; ENVIRONMENTAL EFFECTS; RADIOACTIVE EFFLUENTS; EMISSION; HIGH-LEVEL...

  10. PIA - Savannah River Remediation Accreditation Boundary (SRR...

    Energy Savers

    PIA - Savannah River Nuclear Solution IBARS Srs Site Apps. Accreditation Boundary PIA - WEB Physical Security Major Application Occupational Medical Surveillance System (OMSS) PIA, ...

  11. DOE to Extend Savannah River Nuclear Solutions Contract at Savannah River Site to September 2016

    Energy.gov [DOE]

    Aiken, SC -- The Department of Energy’s (DOE) Savannah River Operations Office today exercised its option to extend the current Savannah River Site Management and Operating contract with Savannah River Nuclear Solutions, LLC (SRNS) for an additional 38 months, from August 1, 2013 to September 2016.

  12. PWR FLECHT SEASET 21-rod-bundle flow-blockage task: data and analysis report. NRC/EPRI/Westinghouse report No. 11, main report and appendices A-J

    SciTech Connect (OSTI)

    Loftus, M.J.; Hochreiter, L.E.; Lee, N.; McGuire, M.F.; Wenzel, A.H.; Valkovic, M.M.

    1982-09-01

    This report presents data and limited analysis from the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). The tests consisted of forced and gravity reflooding tests utilizing electrical heater rods with a cosine axial power profile to simulate PWR nuclear core fuel rod arrays. Steam cooling and hydraulic characteristics tests were also conducted. These tests were utilized to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 163-rod flow blockage bundle tests.

  13. Snake River Geothermal Consortium FORGE Map | Department of Energy

    Energy.gov (indexed) [DOE]

    Map More Documents & Publications Snake River Geothermal Consortium FORGE Map Snake River Geothermal Consortium FORGE Logo Idaho National Laboratory Phase 1 Report Snake River Geothermal Consortium FORGE Map Milford, Utah FORGE Map

  14. EA-273-A Rainy River Energy Corporation | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    -A Rainy River Energy Corporation EA-273-A Rainy River Energy Corporation Order authorizing Rainy River Energy Corporation to export electric energy to Canada. PDF icon EA-273-A ...

  15. Hood River Passive House, Hood River, Oregon (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2014-02-01

    The Hood River Passive Project was developed by Root Design Build of Hood River Oregon using the Passive House Planning Package (PHPP) to meet all of the requirements for certification under the European Passive House standards. The Passive House design approach has been gaining momentum among residential designers for custom homes and BEopt modeling indicates that these designs may actually exceed the goal of the U.S. Department of Energy's (DOE) Building America program to "reduce home energy use by 30%-50%" (compared to 2009 energy codes for new homes). This report documents the short term test results of the Shift House and compares the results of PHPP and BEopt modeling of the project. The design includes high R-Value assemblies, extremely tight construction, high performance doors and windows, solar thermal DHW, heat recovery ventilation, moveable external shutters and a high performance ductless mini-split heat pump. Cost analysis indicates that many of the measures implemented in this project did not meet the BA standard for cost neutrality. The ductless mini-split heat pump, lighting and advanced air leakage control were the most cost effective measures. The future challenge will be to value engineer the performance levels indicated here in modeling using production based practices at a significantly lower cost.

  16. Louisiana Nuclear Profile - River Bend

    U.S. Energy Information Administration (EIA) (indexed site)

    River Bend" "Unit","Summer capacity (mw)","Net generation (thousand mwh)","Summer capacity factor (percent)","Type","Commercial operation date","License expiration date" 1,974,"8,363",98.0,"BWR","application/vnd.ms-excel","application/vnd.ms-excel" ,974,"8,363",98.0 "Data for 2010" "BWR = Boiling

  17. Smith River Rancheria- 2003 Project

    Energy.gov [DOE]

    Implement a planning effort that addresses current and future energy needs and results in a long-term sustainable plan for energy self-sufficiency on the Rancheria. The Smith River Rancheria, located in northern Del Norte County, California about four miles south of the Oregon border on the Pacific Ocean, composes 184.4 acres. This effort is intended to promote energy self-sufficiency and assist in meeting the tribe's goal to preserve and protect the aboriginal territory of the Tolowa People.

  18. EA-1692: Red River Environmental Products, LLC Activated Carbon

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Manufacturing Facility, Red River Parish, LA | Department of Energy 2: Red River Environmental Products, LLC Activated Carbon Manufacturing Facility, Red River Parish, LA EA-1692: Red River Environmental Products, LLC Activated Carbon Manufacturing Facility, Red River Parish, LA June 1, 2010 EA-1692: Final Environmental Assessment Construction and Start-Up of an Activated Carbon Manufacturing Facility in Red River Parish, Louisiana June 11, 2010 EA-1692: Finding of No Significant Impact Red

  19. Comments of the Lower Colorado River Authority | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Lower Colorado River Authority Comments of the Lower Colorado River Authority Comments of the Lower Colorado River Authority on Implementing the National Broadband Plan by Studying the Communications Requirements of Electric Utilities to Inform Federal Smart Grid Policy Comments of the Lower Colorado River Authority (54.12 KB) More Documents & Publications Lower Colorado River Authority Lower Colorado River Authority NBP RFI: Communications Requirements- Comments of Meeker Cooperative Light

  20. Arctic river flood plains are home to hidden carbon

    U.S. Department of Energy (DOE) all webpages (Extended Search)

    Arctic river flood plains are home to hidden carbon Arctic river flood plains are home to hidden carbon In the race to account for how carbon moves through Arctic ecosystems, especially as they warm, scientists may be overlooking one major component: river flood plains. September 27, 2016 The Colville River runs across northern Alaska. The Colville River runs across northern Alaska. Arctic river flood plains are home to hidden carbon In the race to account for how carbon moves through Arctic

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    National Nuclear Security Administration (NNSA)

    Savannah River Site FY 2016 FY 2016 Performance Evaluation Plan, Savannah River Nuclear Solutions, LLC FY 2015 FY 2015 Performance Evaluation Report, Savannah River Nuclear Solutions, LLC FY 2015 Performance Evaluation Report, Fee Determination Letter, Savannah River Nuclear Solutions, LLC FY 2015 Performance Evaluation Plan, Savannah River Nuclear Solutions, LLC FY 2014 FY 2014 Performance Evaluation Report, Savannah River Nuclear Solutions, LLC FY 2014 Performance Evaluation Report, Fee

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    Energy.gov [DOE]

    Slide Presentation by Bonnie Barnes, Savannah River Remediation. Work Planning and Control at Savannah River Remediation.

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    Open Energy Information (Open El) [EERE & EIA]

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    Energy.gov (indexed) [DOE]

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    Energy.gov [DOE]

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    Open Energy Information (Open El) [EERE & EIA]

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    Energy.gov (indexed) [DOE]

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    National Nuclear Security Administration (NNSA)

    12 Savannah River Nuclear Solutions, LLC, PER Summary SUMMARY OF FY 2012 SAVANNAH RIVER NUCLEAR SOLUTIONS, LLC, AWARD FEE DETERMINATION Total Available Fee Total Fee Earned % ...

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    U.S. Department of Energy (DOE) all webpages (Extended Search)

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    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Passive Groundwater Cleanup Measures Save Savannah River Site Millions of Dollars Passive Groundwater Cleanup Measures Save Savannah River Site Millions of Dollars November 25, ...

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    U.S. Department of Energy (DOE) all webpages (Extended Search)

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  17. FY 2006 Washington Savannah River Company, LLC, PER Summary ...

    National Nuclear Security Administration (NNSA)

    6 Washington Savannah River Company, LLC, PER Summary SUMMARY OF FY 2006 WASHINGTON SAVANNAH RIVER COMPANY, LLC, AWARD FEE DETERMINATION Total Available Fee Total Fee Earned % ...

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    Energy Savers

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    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Practices at the Savannah River Site Disposal Practices at the Savannah River Site Full Document and Summary Versions are available for download PDF icon Disposal Practices at the ...

  20. Dennis Yates Of Savannah River Operations Named 2013 Facility...

    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Dennis Yates Of Savannah River Operations Named 2013 Facility Representative Of The Year Dennis Yates Of Savannah River Operations Named 2013 Facility Representative Of The Year ...

  1. DOE - Office of Legacy Management -- Savannah River Swamp - SC...

    Office of Legacy Management (LM)

    Savannah River Swamp - SC 01 FUSRAP Considered Sites Site: Savannah River Swamp (SC.01 ) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site ...

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    Energy.gov (indexed) [DOE]

    Savannah River Operations Office. PDF icon NNSASROONEPA-APS-2013.pdf More Documents & Publications 2012 Annual Planning Summary for Savannah River Operations Office 2010 Annual ...

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    Office of Environmental Management (EM)

    Construction Workers Achieve Safety Milestone at Savannah River Site Construction Workers Achieve Safety Milestone at Savannah River Site April 29, 2014 - 4:23pm Addthis Savannah ...

  4. Microsoft PowerPoint - Allison - Savannah River Presentation

    Office of Environmental Management (EM)

    September 30, 2009 September 30, 2009 JEFFREY M. ALLISON, MANAGER JEFFREY M. ALLISON, MANAGER Savannah River Operations Office Savannah River Operations Office The State of ...

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    Office of Environmental Management (EM)

    Site Agreement Name Savannah River Site Federal Facility Agreement Under Section 120 of ... with past and present activities at the Savannah River Site are thoroughly investigated ...

  6. DOE Order 435.1 Performance Assessment Savannah River Site |...

    Energy Savers

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    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

    Emergency Management - Office of River Protection K Basin Sludge Waste System CRAD, Emergency Management - Office of River Protection K Basin Sludge Waste System May 2004 A section ...

  9. Vermont Watershed Management Rivers Program Website | Open Energy...

    Open Energy Information (Open El) [EERE & EIA]

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    Office of Environmental Management (EM)

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  11. Tianlin Baile River Hydropower Co Ltd | Open Energy Information

    Open Energy Information (Open El) [EERE & EIA]

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  12. Yingjiang County Binglang River Hydroelectric Power Co Ltd |...

    Open Energy Information (Open El) [EERE & EIA]

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    Open Energy Information (Open El) [EERE & EIA]

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    Open Energy Information (Open El) [EERE & EIA]

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    Open Energy Information (Open El) [EERE & EIA]

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    Open Energy Information (Open El) [EERE & EIA]

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    Office of Energy Efficiency and Renewable Energy (EERE) (indexed site)

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    Open Energy Information (Open El) [EERE & EIA]

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    Open Energy Information (Open El) [EERE & EIA]

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    Open Energy Information (Open El) [EERE & EIA]

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