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HomeNRC LibraryDocument CollectionsGeneric CommunicationsInformation Notices1992

# 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
.
```

*Page Last Reviewed/Updated Friday, May 22, 2015*

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