PUMA test program for SBWR

Ishii, M; Revankar, S T; Dowlati, R

Title: PUMA test program for SBWR

Full Record
Other Related Research

Abstract

The objective of the PUMA integral test program is to obtain confirmatory test data for the SBWR Developed by the General Electric-Nuclear Energy Company. The program was initiated in July 1993 under the sponsorship of the NRC. The SBWR has a simplified coolant circulation system and a passive emergency cooling system. The engineered safety systems and safety-grade systems in the SBWR are: (1) the Automatic Depressurization System (ADS), (2) the Gravity-Driven Cooling System (GDCS), (3) the Passive Containment Cooling System (PCCS), (4) the Isolation Condenser Systems (ICS), and (5) the Pressure Suppression Pool (SP). The GDCS and PCCS are new designs unique to the SBWR and do not exist in operating BWRs. The ICS is similar to those in some operating BWRs. The PCCS is designed for low-pressure operation for the containment cooling, but the ICS is capable of high pressure operation as well to cool the reactor pressure vessel. The PUMA design was completed based on an extensive scaling analysis. The PUMA facility having 1/4 height and 1/400 volume scales is constructed. Various facility characterization tests and instrumentation and data acquisition system checks are performed presently. The facility acceptance test will be performed in November and integral tests willmore » be initiated.« less

Authors:

Ishii, M; Revankar, S T; Dowlati, R ^[1]

Purdue Univ., West Lafayette, IN (United States); and others

Publication Date:: Fri Mar 01 00:00:00 EST 1996

Research Org.:: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Brookhaven National Lab. (BNL), Upton, NY (United States)

OSTI Identifier:: 269714

Report Number(s):: NUREG/CP-0149-Vol.1; CONF-9510156-Vol.1
ON: TI96009154; TRN: 96:016820

Resource Type:: Conference

Resource Relation:: Conference: 23. water reactor safety information meeting, Bethesda, MD (United States), 23-25 Oct 1995; Other Information: PBD: Mar 1996; Related Information: Is Part Of Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings; Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)]; PB: 265 p.

Country of Publication:: United States

Language:: English

Subject:: 22 NUCLEAR REACTOR TECHNOLOGY; 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; ENGINEERED SAFETY SYSTEMS; TESTING; BENCH-SCALE EXPERIMENTS

Citation Formats


                    Ishii, M, Revankar, S T, and Dowlati, R. PUMA test program for SBWR.  United States: N. p., 1996. 
        Web.

Copy to clipboard


                    Ishii, M, Revankar, S T, & Dowlati, R. PUMA test program for SBWR.  United States.

Copy to clipboard


                    Ishii, M, Revankar, S T, and Dowlati, R. 1996.  
        "PUMA test program for SBWR".  United States.  https://www.osti.gov/servlets/purl/269714.

Copy to clipboard


                    
@article{osti_269714,

  title        = {PUMA test program for SBWR},

  author       = {Ishii, M and Revankar, S T and Dowlati, R},

  abstractNote = {The objective of the PUMA integral test program is to obtain confirmatory test data for the SBWR Developed by the General Electric-Nuclear Energy Company. The program was initiated in July 1993 under the sponsorship of the NRC. The SBWR has a simplified coolant circulation system and a passive emergency cooling system. The engineered safety systems and safety-grade systems in the SBWR are: (1) the Automatic Depressurization System (ADS), (2) the Gravity-Driven Cooling System (GDCS), (3) the Passive Containment Cooling System (PCCS), (4) the Isolation Condenser Systems (ICS), and (5) the Pressure Suppression Pool (SP). The GDCS and PCCS are new designs unique to the SBWR and do not exist in operating BWRs. The ICS is similar to those in some operating BWRs. The PCCS is designed for low-pressure operation for the containment cooling, but the ICS is capable of high pressure operation as well to cool the reactor pressure vessel. The PUMA design was completed based on an extensive scaling analysis. The PUMA facility having 1/4 height and 1/400 volume scales is constructed. Various facility characterization tests and instrumentation and data acquisition system checks are performed presently. The facility acceptance test will be performed in November and integral tests will be initiated.},

  doi          = {},

  url          = {https://www.osti.gov/biblio/269714},
  journal      = {},
number       = ,

  volume       = ,

  place        = {United States},

  year         = {Fri Mar 01 00:00:00 EST 1996},

  month        = {Fri Mar 01 00:00:00 EST 1996}

}

Copy to clipboard

Conference:

View Conference (15.81 MB)

Other availability

Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that hold this conference proceeding.

Save / Share:

Export Metadata

Save to My Library

Similar records in OSTI.GOV collections:

System code requirements for SBWR LOCA predictions

Journal Article Rohatgi, U; Slovik, G; Kroeger, P - Transactions of the American Nuclear Society

The simplified boiling water reactor (SBWR) is the latest design in the family of boiling water reactors (BWRs) from General Electric. The concept is based on many innovative, passive, safety systems that rely on naturally occurring phenomena, such as natural circulation, gravity flows, and condensation. Reliability has been improved by eliminating active systems such as pumps and valves. The reactor pressure vessel (RPV) is connected to heat exchangers submerged in individual water tanks, which are open to atmosphere. These heat exchanger, or isolation condensers (ICs), provide a heat sink to reduce the RPV pressure when isolated. The RPV is alsomore »« less
Evaluation of the effects of initial conditions on transients in PUMA

Conference Parlatan, Y; Jo, J; Rohatgi, U

A Simplified Boiling Water Reactor (SBWR) is the latest Boiling Water Reactor (BWR) designed by the General Electric (GE). Major differences between the SBWR and the currently operating BWRs include the use of passive gravity-driven systems in the SBWR for emergency cooling of the vessel and containment. In order to investigate the phenomena expected during a Loss of Coolant Accident (LOCA), Nuclear Regulatory Commission (NRC) has sponsored an integral scaled-test facility, called Purdue University Multidimensional Integral Test Assembly (PUMA). The facility models all the major safety-related components of SBWR. Two PUMA initialization calculations were performed to assist the Purdue Universitymore »« less
Full Text Available
Evaluation of the effects of initial conditions on transients in PUMA

Conference Parlatan, Y; Jo, J; Rohatgi, U

Major differences between the SBWR and the currently operating BWRs include the use of passive gravity-driven systems in the SBWR for emergency cooling of the vessel and containment. In order to investigate the phenomena expected during a Loss of Coolant Accident (LOCA), NRC has sponsored an integral scaled-test facility, called Purdue, University Multidimensional Integral Test Assembly (PUMA). The facility models all the major safety-related components of SBWR. Two PUMA initialization calculations were performed to assist the Purdue University in establishing test initialization procedures. Both calculations were based on the initial conditions obtained from SBWR LOCA simulation. In the base case,more »« less
GIRAFFE test results summary

Technical Report Yokobori, S.; Arai, K.; Oikawa, H.

A passive system can provide engineered safety features enhancing safety system reliability and plant simplicity. Toshiba has conducted the test Program to demonstrate the feasibility of the SBWR passive safety system using a full-height, integral system test facility GIRAFFE. The test facility GIRAFFE models the SBWR in full height to correctly present the gravity driving head forces with a 1/400 volume scale. The GIRAFFE test Program includes the certification tests of the passive containment cooling system (PCCS) to remove the post-accident decay heat and the gravity driven cooling system (GDCS) to replenish the reactor coolant inventory during a LOCA. Themore »« less
https://doi.org/10.2172/269712

Full Text Available
System code requirements for safety analysis of SBWR

Journal Article Andersen, J; Shiralkar, B - Transactions of the American Nuclear Society

The simplified boiling water reactor (SBWR) being developed by General Electric Nuclear Energy is an advanced boiling water reactor relying on natural circulation during normal operation and passive safety features. The major elements of the passive safety features are the automatic depressurization of the reactor pressure vessel (RPV) following a loss-of-coolant accident (LOCA) through safety/relief valves and depressurization valves, the gravity-driven coolant system (GDCS), and the passive containment cooling system (PCCS) for residual heat removal. These passive safety systems, although based on existing technology, have generated new requirements for the computer codes used in safety and design analysis. TRACG ismore »« less

Similar Records