ML20082G240
| ML20082G240 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/12/1991 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-64793, NUDOCS 9108160269 | |
| Download: ML20082G240 (3) | |
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t August 12, 1991 U.
S.
Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555
Subject:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Pressurized Thermal Shock Adntional Information (TAC 64793)
This letter provides our response to your request for additional information, dated July 12, 1991, related to the pressurized therma) shock issue described in UFSAR Section 5.3.3.6.
The design basis neutron transport calculation was carried out in R, O geometry using the DOT two-dimensional discrete ordinates of Reference 1 and the SAILOR cross-section library of Reference 2.
The SAILOR library is a 67 group coupled neutron-gamma ray ENDFB-IV based data set produced specifically for light water reactor calculations.
In these design basis
- analyses, anisotropic neutron scattering was treated with a P
expansion of the cross-sections and the angular discretization 3
was modeled with an S order of angular quadrature.
8 The spatial core power distribution utilized in the design basis computation was derived from statistical studies of long term operation of Westinghouse 3-loop plants.
Inherent in the development of this reference core power distribution was the use of an out-in fuel management strategy; i.e.,
fresh fuel on the core periphery.
For the peripheral assemblies, a
20 uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used to conservatively bound the expected power distributions.
Due to the use of this bounding spatial distribution, the results from the design basis calculation establish conservative exposure projections for reactors of this design.
Since it is unlikely that actual reactor operation would result in the implementation of a power distribution at the nominal + 2a level for a large number of fuel cycles. and, because of the implementation of low leakage fuel management strategies, the fuel cycle specific calculations shoul1 result in exposures well below the conservative projections using design basis values.
The difference between the conservative design basis calculation and fuel cycle specific calculations performed in conjunction with the evaluation of Surveillance Capsule U
is illustrated in Tables 6-? and 6-3 of Reference 3.
9108160269 910812,z (h'
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Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Pre'asurized Thermal Shock Additional Information (TAC 64793)
Page 2 Further discussion of the methods used in both design basis and plant specific neutron transport calculations is provided in WCAP-12406.
I Sincerely,
{.
Vice President Nuclear Group cc:
Mr.
J.
Beall, Sr. Resident Inspector Mr.
T.
T.
Martin, NRC Region I Administrator Mr.
A.
W.
DeAgazio, Project Manager Mr.
M.
L.
Bowling (VEPCO)
Beaver-Valley Power Station, Unit No. 2 Docket No.--50-412, License No. NPP-73
. Pre'ssurized Thermal Shock Additional Information (TAC 64793)
Page 3
References:
1.
Soltesa,.
R.G.,
et.
al.,
" Nuclear Rocket Shielding
- Methods, Modification, Updating,- and Input Data Preparation - Volume 5 -
Two Dimensional Discrete Ordinates Transport Technique",
WAN L-PR- ( LL) - 0 3 4, Vol.
5, August 1970, 2.
-SAILOR RSIC DATA LIBRARi COLLECTION DLC-76,
" Coupled Self-Shielded, 47
- Neutron, 20 Gamma
- Ray, P3, Cross Section Library for Light Water Reactors.
3.
- Yanichko, S.E.,
et. al.,
" Analysis of capsule U from the Duquesne Light Company Beaver Valley Unit 2
Reactor Vessel Radiation Surveillance Program", WCAP-12406, September 1989.
-..