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Information Notice No. 92-21: Spent Fuel Pool Reactivity Calculations
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 March 24, 1992 NRC INFORMATION NOTICE 92-21: SPENT FUEL POOL REACTIVITY CALCULATIONS Addressees All holders of operating licenses or construction permits for nuclear power reactors. Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to potential errors in reactivity calculations for spent fuel pools. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. Description of Circumstances On February 14, 1992, the NRC was notified by Northeast Utilities of a discrepancy between reactivity calculations performed for the Millstone, Unit 2, spent fuel pool by ABB Combustion Engineering (CE) and the licensee's contractor (Holtec). The licensee has indicated that the keff calculated by Holtec was approximately 5 percent higher than that previously calculated by CE. The NRC has recently learned that Houston Lighting and Power (HLP) has identified a discrepancy between the reactivity calculations performed for the South Texas, Unit 1, spent fuel pool by Pickard, Lowe and Garrick (PLG) and the licensee's contractor (Westinghouse). The licensee has indicated that the keff calculated by Westinghouse was approximately 2 to 2.5 percent higher than that previously calculated by PLG. Boraflex is utilized as a neutron absorber between spent fuel pool rack cells in both the Millstone, Unit 2, and South Texas, Unit 1, spent fuel pools. Discussion The computer code analyses performed by CE to predict neutron transport for the Millstone, Unit 2, spent fuel storage racks used the two-dimensional, discrete ordinates code DOT. CEPAK was used to generate the neutron cross sections for DOT. The computer code analyses performed by Holtec used KENO (Monte Carlo method). The source of the discrepancy between the CE and Holtec calculations has been attributed by CE to two approximations made in the generation of neutron cross sections. First, a transport cross section was used by CE as 9203180053 . IN 92-21 March 24, 1992 Page 2 of 2 an approximation for the total cross section. While this approximation is valid for most materials, it is not valid for materials having large thermal cross sections. Therefore, applying this approximation to regions containing a strong neutron absorber (such as Boraflex) results in an overestimation of the neutron absorption and a corresponding lower calculated keff in that region. Second, a geometric buckling term corresponding to a sparsely populated and weakly absorbing (unpoisoned) array was utilized by CE as an approximation of buckling in the highly absorbing configuration. This approximation, however, is not valid for the specific configuration found in the Millstone racks where the assembly pitch is small and the fuel assembly is completely surrounded by a strong neutron absorber. After these approximations were corrected, the results of the CE analyses were in good agreement with Holtec's. The original computer code analyses performed by PLG to predict neutron transport for the South Texas, Unit 1, spent fuel storage racks used the two-dimensional diffusion theory code PDQ. LEOPARD was used to generate the cross sections for PDQ. Computer code analyses performed by Westinghouse utilized KENO (Monte Carlo method). The lower value of keff calculated by PLG has been attributed by HLP to the inaccuracies inherent in using diffusion theory to predict neutron attenuation through a thin region that strongly absorbs neutrons (such as Boraflex). Both the CE and PLG methodologies had been benchmarked against criticality experiments that have been reported to closely represent the characteristics of the spent fuel storage racks. However, it should be noted that the number of criticality experiments that included a strong neutron absorber (such as Boraflex) was limited. This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager. Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical contacts: Jack Ramsey, NRR (301) 504-1167 Larry Kopp, NRR (301) 504-2879 Attachment: List of Recently Issued NRC Information Notices .
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