Information Notice 1992-21, 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 discrep- ancy 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 k
calculated by Holtec was
approximately 5 percent higher than that previou¶ly calculated by CE.
The NRC has recently learned that Houston Lighting and Power (HLP) has identi- fied a discrepancy between the reactivity calculations performed for the South
Texas, Unit 1, spent fuel pool by Pickard, Lowe and Garrilck (PLG) and the
licensee's contractor (Westinghouse). The licensee has indicated that the k f
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
eutron ros sections.
First, a transport cross section was used by CE as
(
92318005
2_
I
, I
IN 92-21 March 24, 1992 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 k
in that region.
Second, a geometric buckling term corresponding to a sparsly 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 k f calculated by
PLG has been attributed by HLP to the inaccuracies inherefltin 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..
/r~les E.Rossi, Drector
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
..
I
Attachment
IN 92-21 Match 24, 1992 v
Page 1lof I
LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
92-20
92-19
92-18
92-17
92-16
92-15
Inadequate Local Leak Rate
Testing
Misapplication of Potter &
Brumfield MDR Rotary Relays
Potential for Loss of Re- mote Shutdown Capability
during A Control Room Fire
NRC Inspections of Pro- grams being Developed at
Nuclear Power Plants in
Response to Generic
Letter 89-10
Loss of Flow from the
Pump during Refueling
Cavity Draindown
Failure of Primary System
Compression Fitting
Uranium Oxide Fires at Fuel
Cycle Facilities
Relap5/Mod3 Computer Code
Error Associated with the
Conservation of Energy
Equation
Inadequate Control Over
Vehicular Traffic at
Nuclear Power Plant Sites
03/03/92
03/02/92
02/28/92
02/26/92
02/25/92
02/24/92
02/21/92
02/18/92
02/18/92
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
All fuel cycle and uranium
fuel research and development
licensees.
All holders of OLs or CPs
for nuclear power reactors.
All holders of OLs or CPs
for nuclear power reactors.
92-14
92-02, Supp. 1
92-13 OL = Operating License
CP = Construction Permit
IN 92-21 March 24, 1992 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 k
in that region.
Second, a geometric buckling term corresponding to a spar!ety 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 k f calculated by
PLG has been attributed by HLP to the inaccuracies inhereft 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.
Chares E Rossi
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
- SEE PREVIOUS CONCURRENCE
OFC
- DOEA:OEAB
- ADM:RPB
- SC:DOEA:OEAB:C:DOEA:OEAB :C:DST:SRXB :C:DOEA:OGCB :D:DOEA
___
_ _ __________-____-_
___-___________
NAME :JRamsey*
- JMain*
- DFischer*
- AChaffee*
- RJones*
- CBerlinger* :C~ otL
- D
ischer
- Ahf
- Roe*-Z;
jz
DATE :03/09/92
- 02/20/92
- 03/09/92
- 03/09/92
- 03/09/92
- 03/13/92
- 3//'t/92
OFFICIAL RECORD COPY
Document Name: IN 92-21
IN 92-XX
March xx, 1992 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 k
in that region.
Second, a geometric buckling term corresponding to a sparSiy 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
utililized KENO (Monte Carlo method). The lower value of ke
calculated by
PLG has been attributed by HLP to the inaccuracies inherent
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) is 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
- SEE PREVIOUS CONCURRENCE
OFC
- DOEA:OEAB
- ADM:RPB
- SC:DOEA:OEAB:C:DOEA:OEAB
- C:DST:SRXB
- C:DOEA:OGCB :D:DOEA
_____ _______
_------------: ------------:------------: ------------ :--------------r
NAME :JRamsey*
- JMain*
- DFischer*
- AChaffee*
- RJones*
- CBerlinger :,CRossi
DATE :03/09/92
- 02/20/92
- 03/09/92
- 03/09/92
- 03/09/92
- //a9t W
/ /92
OFFICIAL RECORD COPY
Document Name:
DIFFUSION THEORY IN
IN 92-XX
March xx, 1992 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 k
in that region.
Second, a geometric buckling term corresponding to a sparEty 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
utililized KENO (Monte Carlo method). The lower value of k
calculated by
PLG has been attributed to the inaccuracies inherent in usiRg diffusion theory
to predict neutron attenuation through a region that strongly absorbs neutrons.
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) is 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
OFC :DOEA:OEAB
- ADM:RPB
- S D A:OEAB:C:DQEA:OEAB :C:DST:SRXB :C:DOEA:OGCB :D:DOEA
--- w
An--a
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---------
NAME : eivJMatn
&
scher
- AChaffee
- FJ hes
- CBerlinger
- CRossi
DATE : 5/i /92
- 02/20/92
- 3/j /92
- 3/ 7/92
3 /1/92
- / /92
OFFICIAL RECORD COPY
Document Name:
DIFFUSION THEORY IN