ML17264A384
| ML17264A384 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 02/27/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17264A382 | List: |
| References | |
| NUDOCS 9603050497 | |
| Download: ML17264A384 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 T
V U
8 TH OFFICE OF NUCLEAR REACTOR REGULATION R LATING TO R
FERENCE OF THE NOTRUHP SHALL-BREAK LOSS-OF-COOLANT ACCIOENT ANALYSIS HETHOOOLOGY B
H ROCH ST R
GAS ANO EL CTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT OOCKET NO. 50-244 I.O
~INT ONCT ON By letter -dated June 19,
- 1995, Rochester Gas and Electric Corporation (RG&E),
the licensee for theNR.
E. Ginna Nuclear Power Plant (Ginna), submitted documentation discussing the evaluation model to be used for a small-break loss-of-coolant accident (SBLOCA) at the Ginna plant.
As part of the documentation, RG&E provided information to show that the evaluation model (NOTRUHP) for the SBLOCA model (Westinghouse WCAP-10054-P) could appropriately be applied to the Ginna two-loop upper plenum injection design.
- Also, as part of the documentation, RG&E provided information to describe inputs and assumptions that would be used in applying NOTRUHP to the Ginna plant.
- 2. 0 EVALUATION In its submittal of June 19,
- 1995, RG&E (1) described a typical SBLOCA
- scenario, (2) referenced the NOTRUHP evaluation model (EH), (3) discussed how the NOTRUHP EH includes correlations to appropriately analyze SBLOCA events for the Ginna design, (4) 'identified specialized inputs and assumptions used for the two-loop Ginna design, and (5) provided a small-break spectrum of analytical results for the Ginna plant that was calculated by using the NOTRUHP EH.
Although RG&E indicates that the results provided were performed with another version of NOTRUHP (known as COSI), which is under continuing NRC review, the submittal demonstrates the applicability of any version of NOTRUHP to the Ginna plant.
The calculational results, using this as yet unapproved version of the NOTRUHP EH, exhibit such a large margin to the Ginna large-break LOCA spectrum that they amply demonstrate that current large-break LOCA analyses continue to bound the SBLOCA analyses for the Ginna plant.
~NN ON On the basis of its review, the NRC staff finds that the NOTRUHP SBLOCA evaluation model described in any approved version of Westinghouse WCAP-10054-P is acceptable for use in Ginna SBLOCA analyses, and approved versions of this methodology may be referenced in Ginna licensing documentation, including the Ginna core operating limits report.
Enclosure 9603050497
'&0227 PDR ADQCK 05000244 P
The staff finds that the SBLOCA results provided in the licensee's submittal demonstrate that large-break LOCA analyses currently bound SBLOCA analyses for the Ginna plant.
Assumptions such as high-pressure injection pump performance in the submitted analyses may be used to determine plant surveillance
- criteria, and so on, even though the results were performed with an unapproved version of the NOTRUHP EH (known as COSI).
This conclusion is primarily due to the large margins used in the submitted analyses that have been qualitatively demonstrated.
Principal Contributor:
Frank Orr b
Date:
Dr. Robert C. Mecred
'Ace President, NucleM Operations, Rochester Gas and Electric Corporation 1i 89 East Avenue Rochester, NY 14649 i
i'UBJECT:
R.
E.
GINNA NUCLEAR POWER PLANT SHALL-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS MODEL (TAC NO. M92764)
Dear Dr. Mecredy:
By letter dated June 19,
- 1995, Rochester Gas and Electric Corporation (RG&E),
the licensee for the R.
E. Ginna Nuclear Power Plant (Ginna), submitted documentation discussing the evaluation model to be used for a small-break loss-of-coolant accident (SBLOCA) at the Ginna plant.
As part of the documentation, RG&E provided information to show that the evaluation model (NOTRUMP) for the SBLOCA model (Westinghouse WCAP-10054-P) could appropriately be applied to the Ginna two-loop upper plenum injection design.
- Also, as part of the documentation, RG&E provided information to describe inputs and assumptions that would be used in applying NOTRUHP to the Ginna plant.
On the basis of its review, the NRC staff finds that reference to an approved version of NOTRUMP is acceptable and that this reference is suitable for inclusion in the Ginna core operating limits report or other licensing documentation.
A copy of the related Safety Evaluation is enclosed.
This action closes TAC No. M92764.
Sincerely, Original signed by:
Docket No. 50-244
Enclosure:
As stated cc w/encl:
See next page Allen R. Johnson, Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation DISTRIBUTION:
Docket File PUBLIC PDI-1 R/F SVarga JZwolinski LMarsh AJohnson SLittle
- RJones, SRXB
- RCooper, RI.
LDoerflein, RI JWiggins,'RI I
DOCUMENT NAME:
G:iGINNAiGI92764.SE To receive a copy of this document, indicate in the box:
"C" = Copy wi'thout attachment enclosure "E"
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No co OFFICE NAME DATE LA:PD1-1 SLittl 02/0 /96 E
PM:PDl-1 AJohnson/rsl 02/
/96 D:PDT-i LMsrsh 02/2-'7/96 0
c a
Recor Copy' 02/
/96 02/
/96
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