ML20234E965
| ML20234E965 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 09/15/1987 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8709220583 | |
| Download: ML20234E965 (3) | |
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2C (412) 393 6000 One Oxford Centre ktbburgI, P September 15, 1987 15279 U. S. Nuclear Regulatory Commission
./ Attn:
Document Control Desk j
Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Sixth Refueling Outage Plant Modifications Gentlemen:
i This letter provides notification of the following Unit 1 plant modifications that will be incorporated during the forthcoming refueling outage.
This outage will begin in December 1987.
- 1. Components in the barrel / baffle region of the reactor vessel will be modified from the current downflow design to an upflow design.
The upflow modification has been implemented at a
number of other Westinghouse nuclear plants to eliminate the baffle jetting problem that we observed during the last refueling.
- 2. Fuel assembly thimble plugs will be removed.
Thimble plug removal has been implemented at the Prairie Island Nuclear Plant and was accepted by the NRC as described in the letter and SER dated October 18, 1985.
Westinghouse has performed the required analyses to determine the effect of the above modifications and in addition has performed a LOCA reanalysis to justify operation with a steam generator plugging level of up to 10%.
The new Large Break LOCA analysis (FSAR 14.3.2) was performed with the latest approved version of the Westinghouse Large Break Evaluation Model (see Reference 1).
The results indicate that considerable margin will remain available to the limits defined in 10 CFR 50.46 and Appendix K.
The new Small Break LOCA analysis (FSAR 14.3.1) was performed with the latest approved version of the Westinghouse Small Break Evaluation Model (see References 2,
3 and 4).
These results also i
indicate margin to the limits defined in 10 CFR 50.46 and Appendix K.
The Containment Long Team Mass and Energy Release and Containment Subcompartment Analyses (FSAR 14.3) were performed to determine the impact on the current FSAR analyses.
The results indicate there is no additional impact on the FSAR containment analyses.
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Sixth Refueling Outage Plant Modifications Page 2
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i The Steam Generator Tube Rupture analysis (FSAR 14.2.4) was l
performed to evaluate the radiological consequences.
The results indicate. the conclusions reported in the FSAR analysis remain unchanged.
The blowdown hydraulic loads resulting from a LOCA are considered in FSAR 14.3.3 and Appendix B-3.
A WCAP will be issued describing the hydraulic forces analysis and the subsequent structural evaluation.
The Long Term Cooling evaluation (FSAR 14.3.2) was performed.
The results indicate an insignificant effect on the mass average boron concentration required to maintain core subcriticality following a large break LOCA.
The post-LOCA hot leg recirculation switchover time is dependent on power
- level, and the RCS, RWST and accumulator water volumes and boron concentrations.
Since these parameters will not be affected there is no impact on the post-LOCA hot leg recirculation switchover time.
We discussed these modifications and analyses with the NRC Project Manager and it was agreed that this notification would be provided to advise the NRC of these changes in plant design and licensing basis since the results of the analyses will be used to update the FSAR.
Based on the preliminary evaluations performed, we believe that these changes may be done in accordance with 10 CFR 50.59 since the potential impact of the modifications on the FSAR analyses for each of the LOCA related accidents has been evaluated and it was determined that in all cases the effect of the modifications did not result in exceeding any design or regulatory limits.
This preliminary conclusion and its' supporting bases are subject to a final review and determination by the site safety committees prior to the start of the modifications.
This letter is intended to provide preliminary notification of our planned actions.
Should you have any questions concerning the details of the above modifications, please contact me or members of my staff for additicnnl information.
Very truly yours, Sie er Vice President Nuclear Operations
\\
1 s
BDOver-Valley'Powsr Station, Unit No'. 1 j
Docket No. 50-334, License'No. DPR-66 j
Sixth Refueling Outage Plant Modifications j
Page 3 j
j cc:
Mr'.JP. Tam,-Project Manager Mr. F. I. Young, NRC Sr. Resident Inspector Mr. L. Prividy, NRC Resident Inspector Mr. W
-T. Russell, NRC Region I Administrator VEPCO j
References:
1.
WCAP-10266-P-A Rev.
2 Addendum 2, Besspiata, J.
J.,
et al.,.
"The 1981 Version of the Westinghouse ECCS Evaluation Model1 Using the BASH Code", March 1987.
2.
WCAP-10054-P-A (Proprietary), Lee, H.,
et al., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP. Code",
August 1985.
L 3.
WCAP-10079-P-A (Proprietary),
'WCAP-10080-A' f
(Non-Proprietary),
- Meyer, P.
E., "NOTRUMP-A Nodal Transient Small Break and General Network Code", August 1985.
4.
WCAP-11145-P-A, Rupprechet, S.
D.,
et al.,." Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code", October 1986.
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