skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC., RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT, JULY-AUGUST-SEPTEMBER 1960

Abstract

A Fortran code (Kernmat) for the computation of the effective multiplication factor and absorptions in the components of a heterogeneous reactor based on the heterogeneous methods of small source theory was developed. Refinements were made to incorporate in the HYDNA code the calculation of the void fraction in the local boiling region snd the transition region between local and bulk boiling. A study was made to determine the heat sources in the CVTR and the spatial distribution of these sources. The CVTR Phase I-B tests were run in Loop D. These tests were primarily intended to provide a comparison between a three and four thermal baffle design and to further assess the problem of by-pass leakage between the baffles and the pressure tube wall. A new arrangement for sealing the thermal baffles to the pressure tube wall was evolved for the Phase II tests. The new concept employs a ball-cone type seal in the pressure tube extension. This new seal concept also eliminates the need of integral latch arms in the shield plug and permits it to be made in one piece. A full-size Type 410 stainless steel--Zircaloy-2 Conoseal joint cracked while undergoing thermal cyclic tests. Careful inspection revealed thatmore » a circumferential crack (270 deg ) had developed between the 410 stainless steel snd the 309 stainless steel buttering on the sleeve. It was concluded that microcracks existed in the assembly prior to test in the region of failure and were instrumental in the failure. The thermal baffle test facility was adapted with high pressure fittings and flanges to allow testing of thermal baffle arrangements at higher temperatures. The autoclave selected for the facility will also be modified. To aid in the evaluation of a ceramic baffle as an alternate to the present CVTR bsffle design, three samples were formed by swagging a tube filled with ZrO/sub 2/ powder. The samples were defected in the clad and put in an autoclave for high temperature water tests. Analysis of the water after the tests indicated that very small quantities of ZrO/sub 2/ were deposited in the water. Fabrication of the model U-tubes, heater rods, the venturi, orifice plates and void fraction chambers for the hydrodynamic stability tests was completed. Fabrication and installation of the apparatus for messuring the thermal resistance between the fuel pellets and the cladding was completed. Thermocouples were found to be out of tolerance and some were returned to the manufacturer for an error analysis or repair. All data from the first series of mixing studies were analyzed statistically. A fuel assembly with fuel rods wire wrapped on a 12-inch pitch produced slightly more mixing than a fuel assembly having fuel rods wrapped on a 15-inch pitch. The in-pile loop at WTR was completed and tests were performed in preparation for placing the loop in operation. The performance of the loop was satisfactory including 58 of the 64 loop thermocouples. The results of the Conoseal gasket crevice corrosion experiments to date indicate no serious problem at the 300 to 567 deg F CVTR operating conditions. Because of difficulties developed in air cooling the side and bottom thermal shields, it was recommended that these shields be cooled by water coils sandwiched between slabs of steel. The critical experiment has thus far given information concerning control rod worth, heavy water worth, critical size of a 2.0% enriched, half-height core, critical size of a 1.1% enriched full-height core, coolant void coefficients, flux plots and the worth of black and grey control rods in a tworegion, full- height core. (auth)« less

Publication Date:
Research Org.:
Westinghouse Electric Corp. Atomic Power Dept., Pittsburgh
OSTI Identifier:
4019613
Report Number(s):
CVNA-65
NSA Number:
NSA-15-017828
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-61
Country of Publication:
Country unknown/Code not available
Language:
English
Subject:
REACTOR TECHNOLOGY; BOILING; COMPUTERS; NUMERICALS; PROGRAMMING; ZONES

Citation Formats

. CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC., RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT, JULY-AUGUST-SEPTEMBER 1960. Country unknown/Code not available: N. p., 1961. Web.
. CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC., RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT, JULY-AUGUST-SEPTEMBER 1960. Country unknown/Code not available.
. Tue . "CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC., RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT, JULY-AUGUST-SEPTEMBER 1960". Country unknown/Code not available.
@article{osti_4019613,
title = {CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC., RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT, JULY-AUGUST-SEPTEMBER 1960},
author = {},
abstractNote = {A Fortran code (Kernmat) for the computation of the effective multiplication factor and absorptions in the components of a heterogeneous reactor based on the heterogeneous methods of small source theory was developed. Refinements were made to incorporate in the HYDNA code the calculation of the void fraction in the local boiling region snd the transition region between local and bulk boiling. A study was made to determine the heat sources in the CVTR and the spatial distribution of these sources. The CVTR Phase I-B tests were run in Loop D. These tests were primarily intended to provide a comparison between a three and four thermal baffle design and to further assess the problem of by-pass leakage between the baffles and the pressure tube wall. A new arrangement for sealing the thermal baffles to the pressure tube wall was evolved for the Phase II tests. The new concept employs a ball-cone type seal in the pressure tube extension. This new seal concept also eliminates the need of integral latch arms in the shield plug and permits it to be made in one piece. A full-size Type 410 stainless steel--Zircaloy-2 Conoseal joint cracked while undergoing thermal cyclic tests. Careful inspection revealed that a circumferential crack (270 deg ) had developed between the 410 stainless steel snd the 309 stainless steel buttering on the sleeve. It was concluded that microcracks existed in the assembly prior to test in the region of failure and were instrumental in the failure. The thermal baffle test facility was adapted with high pressure fittings and flanges to allow testing of thermal baffle arrangements at higher temperatures. The autoclave selected for the facility will also be modified. To aid in the evaluation of a ceramic baffle as an alternate to the present CVTR bsffle design, three samples were formed by swagging a tube filled with ZrO/sub 2/ powder. The samples were defected in the clad and put in an autoclave for high temperature water tests. Analysis of the water after the tests indicated that very small quantities of ZrO/sub 2/ were deposited in the water. Fabrication of the model U-tubes, heater rods, the venturi, orifice plates and void fraction chambers for the hydrodynamic stability tests was completed. Fabrication and installation of the apparatus for messuring the thermal resistance between the fuel pellets and the cladding was completed. Thermocouples were found to be out of tolerance and some were returned to the manufacturer for an error analysis or repair. All data from the first series of mixing studies were analyzed statistically. A fuel assembly with fuel rods wire wrapped on a 12-inch pitch produced slightly more mixing than a fuel assembly having fuel rods wrapped on a 15-inch pitch. The in-pile loop at WTR was completed and tests were performed in preparation for placing the loop in operation. The performance of the loop was satisfactory including 58 of the 64 loop thermocouples. The results of the Conoseal gasket crevice corrosion experiments to date indicate no serious problem at the 300 to 567 deg F CVTR operating conditions. Because of difficulties developed in air cooling the side and bottom thermal shields, it was recommended that these shields be cooled by water coils sandwiched between slabs of steel. The critical experiment has thus far given information concerning control rod worth, heavy water worth, critical size of a 2.0% enriched, half-height core, critical size of a 1.1% enriched full-height core, coolant void coefficients, flux plots and the worth of black and grey control rods in a tworegion, full- height core. (auth)},
doi = {},
url = {https://www.osti.gov/biblio/4019613}, journal = {},
number = ,
volume = ,
place = {Country unknown/Code not available},
year = {1961},
month = {10}
}

Technical Report:
Other availability
Please see Document Availability for additional information on obtaining the full-text document. Library patrons may search WorldCat to identify libraries that may hold this item. Keep in mind that many technical reports are not cataloged in WorldCat.

Save / Share: