ML17261A754
| ML17261A754 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/24/1987 |
| From: | Kober R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Stahle C NRC, NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML17261A755 | List: |
| References | |
| NUDOCS 8801070036 | |
| Download: ML17261A754 (9) | |
Text
REGULA1 Y INFORMATION D ISTR IHUTIO
'YSTEM
( R IDS )
ACCESSION NBR: 880i070036 DOC. DATE: 87/i2/24 NOTARIZED:
YEB DOCKET FACIL 50244 Robert Emmet Qinna Nuclear Planti Unit 1>
Rochester G
05000244 AUTH. NAME AUTHOR AFFILIATION K0B ER i R. M.
Rochester Qas 5 Electric Corp.
RECIP. NAME RECIPIENT AFFILIATION BTAHLE. C.
Document Control Branch (Document Control Desk)
SUBJECT:
Forwards proprietary MCAP-ii668 h nonproprietary WCAP-ii678I "LOFTTR2 Analysis of Potential Radiological Consequences Following Steam Generator Tube Rupture at RE Qinna Nuclear Power Plant. " 4lCAP-ii668 withheld (ref i0CFR2. 790).
DISTRIBUTION CODE:
P*OID COP ES R CEIVED: LTR + ENCL ~
SIZE:
TITLE: Proprietary Review Distribution-Operating Reactor NOTES: Licensee Exp date in accordance with iOCFR2i 2. 10'P(9/i9/72).
05000244 RECIPIENT ID CODE/
ME PDi-3 LA STAHLEI C INTERN*L: *SOD/DO*~
OGC/HDS2 COPIES LTTR ENCL i
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AEOD/DSP/TPAB 4 ~ i i
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/i7iilllI7/I I!Illili&l II!ESN L41< Owlaa ROCHESTER GAS AND ELECTRIC CORPORATION D 89 EAST AVENUE, ROCHESTER, N.K f484g-pppg ROGER W, KOBER VICE PIIESIDENT ELECTRIC PRODUCTION December 24, 1987 TELCPHDNE AREA CODE 7ld 546-2 IOO U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Mr. Carl Stahle PWR Project Directorate No.
1 Washington, D.C.
20555
Subject:
Maximum Coolant Activity Technical Specification R.E.
Ginna. Nuclear Power Plant Docket No. 50-244
Dear Mr. Stahle:
Enclosed are:
l.
One (1) copy of Application for Withholding, CAW-87-123, dated December 3,
1987 accompanying Affidavit, and Proprietary Information Notice.
2.
One (1) copy of WCAP-11668, "LOFTTR2 Analysis of Potential Radiological Consequences Following a Steam Generator Tube Rupture at the R.E. Ginna Nuclear Plant",
November 1987 (Proprietary).
3.
One (1) copy of WCAP-11678, "LOFTTR2 Analysis of Potential Radiological Consequences Following a Steam Generator Tube Rupture at the R.E. Ginna Nuclear Plant",
November 1987 (Non-Proprietary).
L g he
, ClO A5 aa OA NC a
oo.
mA NCLQ.'ochester Gas and Electric Corporation'RG&E) letter from R.W. Kober to J.A. Zwolinski dated October 9, 1985 transmitted the results of an evaluation of the potential radiological con-sequences due to a steam generator tube rupture (SGTR) for the R.E.
Ginna Nuclear Power Plant.
The purpose of the evaluation was to increase the Ginna Technical Specification primary coolant activity limit from 0.2 to 1.0uCi/gm.
This evaluation assumed a
steam generator tube plugging level of 10-o.
Since the time of this evaluation the safety analysis for Ginna has been reevalu-ated assuming a
15% steam generator tube plugging level.
The previous evaluation using a
10% steam generator tube plugging level was presented in WCAP-10884 (Proprietary)/WCAP-10885 (Non-Proprietary).
In the process of updating the results for 15': tube plugging, an error was discovered, in the previous calculations for 10% tube'lugging.
It was determined that the Cygne>F:ggu MO, f Sb& 876 Qt'3
reactivity insertion due to reactor trip was underestimated in that analysis, which resulted in an erroneously high core heat generation rate following reactor trip.
Since the calculated mass releases to the atmosphere are dependent upon the core heat generation following reactor trip, the offsite doses which were calculated based on the mass releases are incorrect.
Because of this error, RGGE requests the NRC to remove WCAP-10884 (Proprie-tary)/ WCAP-10885 (Non-Proprietary) from Docket No. 50-244 for the R.E. Ginna Plant and. replace it with the attached WCAP-11668 (Proprietary)/WCAP-11678 (Non-Proprietary).
Since SGTR analysis has been updated for 15% tube plugging, WCAP-10884 (Proprietary)/WCAP-10885 (Non-Proprietary) for 10%
steam generator tube plugging will not be revised..
- However, a corrected SGTR analysis has been performed to determine the offsite radiation doses based on 10% tube plugging and the results are compared with the calculated offsite doses for 15%
tube plugging in Table 1.
As seen from Table 1, the increased tube plugging results in a slight increase in the offsite doses for Case 1 which is the limiting case, but the effect of tube plugging on the results for Case 2 is negligible.
The results of this evaluation demonstrate that the offsite radiation doses for a SGTR will be less than the allowable guideline values specified in Standard Review Plan 15.6.3 based on the Standard Technical Specification limit on primary coolant activity.
Xt is noted that the LOFTTR1 program was used to perform the previous SGTR analysis for 10% tube plugging presented in WCAP-10884 (Proprietary)/WCAP-10885 (Non-Proprietary),
whereas the LOFTTR2 program was utilized to perform the analysis for 15% tube plugging.
The LOFTTR1 program was developed as part of the revised SGTR analysis methodology and was used for the SGTR evaluations in WCAP-10698-P-A and Supplement 1 to WCAP-10698-P-A.
However, the LOFTTR1 program was subsequently modi-fied to accommodate steam generator overfill and. the revised program, designated as LOFTTR2, was used for the evaluation of the consequences of overfill in WCAP-11002 "Evaluation of Steam Generator Overfill Due to Steam Generator Tube Rupture Accident".
The LOFTTR2 program is identical to the LOFTTR1 program, with the exception that the LOFFTR2 program has the additional capability to represent the transition from two regions (steam and water) on the secondary side to a single water region as overfill occurs, and the transition back to two regions again depending upon the calculated secondary conditions.
Since the LOFTTR2 program has been validated against the LOFTTRl program, the LOFTTR2 program is also appropriate for performing licensing basis SGTR analyses.
Since the completion of WCAP-11668 (Proprietary)/WCAP-11678 (Non-Proprietary)
Westinghouse has identified a generic issue regarding a fundamental assumption in the radiological analysis.
Specifically, the analysis assumes the steam generator tubes do not uncover following a reactor trip due to a SGTR.
Current evidence indicates that this may not be a conservative assumption.
0
A fundamental assumption of the radiological analyses for accidents involving steam release from the secondary side is that elemental iodine, transferred from the primary side via tube leaks, will partition between the steam generator water and steam, such that the iodine concentration in the steam is a small fraction of the water concentration.
The most conser-vative leak location is at the top of the tube bundle.
- Hence, uncovery of the bundle (leakage or rupture site) potentially creates a direct activity release path to the environment i.e.,
there will be neither dilution, by the secondary side water, nor partitioning of the'ctivity carried by the leakage flow.
For the Ginna-specific analysis the doses for Case 2 on Table 1 are conservatively estimated to increase by a factor of 3 for the accident initiated spike and a factor of 6 for the pre-accident spike (Case 1 is unaffected by this issue because the steam generator tubes do not uncover).
The resulting estimated doses are still within allowable guidelines.
A methodology to more precisely calculate the doses due to tube uncovery will be developed by the Owners Group or a subgroup of utilities.
When this methodology is available, the doses in WCAP-11668 will be revised.
As noted previously, an increase in the steam generator tube plugging level from 10% to 15't only results in a slight increase in the calculated radiation doses for Case 1 and a negligible change for Case 2.
The relative comparison between the results for 10% and 15-o tube plugging are not affected by the potential for tube uncovery for an SGTR.
Xn addition, it is estimated that the calculated offsite doses for 15-o tube plugging will rem'ain well below the allowable guidelines after. the effect of the potential uncovery is conservatively assessed.
On this basis, it is concluded that the consequences of an SGTR for Ginna will be acceptable with 15% tube plugging.
As this submittal contains information proprietary to West-inghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information.
The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.
Accordingly, it is respectfully requested that the infor-mation which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations.
Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse affidavit should reference CAW-87-123 and should be addressed to R.A. Wiesemann, Manager Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P.O.
Box 355, Pittsburgh, Pennsylvania 15230-0355.
Ver truly yours, v@ k/
Roger W. Kober Enclosures
TABLE 1 Comparison of Offsite Doses for 10% and.
15% Steam Generator Tube Plugging Case 1
Doses (Rem)
Case 2*
SG Plugging Level 10%
15-s SG Plugging Level Allowable 10%
15%
Values
- 1. Accident Initiated.
Iodine S ike Exclusion Area Boundary (0-2 hours)
Thyroid Low Population Zone (0-8 hours)
Thyroid 25.6 1.6 26.4 1.7 3.8 0.3 3.8 0.3 30 30
- 2. Pre-Accident Iodine S ike Exclusion Area Boundary (0-2 hours)
Thyroid Low Population Zone (0-8 hours)
Thyroid 6.2 6.4
- 99. 2 102. 1
- 22. 1 1.4 22.1 1.4 300 300 Case 1
worst single failure for overfill Case 2
worst single failure to maximize steam releases
- As defined in Standard.
Review Plan 15.6.3 and 10CFR100