ML19321A857

SER for Facility,Suppl 3 to NUREG-75/095

ML19321A857
Person / Time
Site:	05000516, 05000517
Issue date:	01/17/1977
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0057, NUREG-57, NUDOCS 8007240299
Download: ML19321A857 (13)

Text

Safety

< s- ~==

Evaluation Itcpori a

...,yto*m Y.".to"o' rGIGted to construction of 0ffice of N ci ar p

Jamesport Nuclear Power Station, Units 1 and 2 syn so.s,7 anuarY 1977 Long Island Lighting Company Supplement No. 3 ge

^

r

}

Available from National Technical Infonnation Service Springfield, Virginia 22161 Price: Printed Copy $3.50 ; Microfiche $3.00

. 4 i

t vt V

' (.

'2

+

+-

r--w-

-r-c

-~-ew--

NUREG - 0057 (Supplement 3 to NUREG-75/095)

January 17, 1977 SUPPLEt!ENT NC. 3 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF LONG ISLAND LIGHTING COMPANY NEW YORK STATE ELECTRIC & GAS CORPORATION JAMESPORT NUCLEAR POWER STATION UNITS 1 & 2 DOCKET N05. STN 50-516 AND STN 50-517

TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

1-1 6.0 ENGINEERED SAFETY FEATURES.........

6-1 6.2 Co n ta i nme n t Sy s t ems..............................................

6-1 6.2.1 Containment Functional Design.............................

6-1 6.2.7 Containment Backpressure Evaluation for ECCS Performance....

6-1 6.3 Emergency Core Cooling System (ECCS)..............................

6-2 l

6.3.3 Perfo rmance Evalua tion...........................

6-2 LIST OF APPENDICES APPENDIX A Upda ted Ch ronol ogy..............................

A-1 APPENDIX B Er ra ta for Su ppl emen t 2............................

B-1 1

1 1

f i

4

1.0 INTRODUCTION

.The Nuclear Regulatory Commission's (Commission) Safety Evaluation Report (NUREG 75/095) in the matter of the application by Long Ifeand Lighting Company and New York State Electric & Gas Corporation (herein referred to as the applicant) for

- licenses to construct and operate the proposed Jamesport Nuclear Power Station Units 1 and 2 (Jamesport Unit 1 and 2 or facility) was issued on October 6 1975. In this Safety Evaluation Report, the Nuclear Regulatory Commission staff indicated: (1) certain staff positions which would be made conditions of the construction permit unless the applicant made commitments to meet these require-ments, and (2) certain outstanding issues for which additional information would i '

be required to permit the staff to conclude that the facilit design meets applicable commission requirements.

The purpose of this supplement is to update the Safety Evaluation Report by providing the staff's evaluation of additional information submitted by the applicant since the issuance of Supplement 2 to the Safety Evaluation Report.

i The issuance of this supplement finishes the staff's review of all outstanding issues, which were listed in the initial Jamesport Safety Evaluation Report. In addition, corrections or additions have been made to the Safety Evaluation Report in certain specific areas. Each section in this supplement is numbered the same as the sections of the Safety Evaluation Report that is being updated and, unless otherwise indicated. is supplementary to and not in lieu of the discussion in the Safety Evaluation Report.

Our evaluation of the analysis of the emergenc core cooling system has confirmed that the system complies with the Final Acceptance Criteria. In Supplement No.

1, 2.and in this supplement, we have discussed each of the outstanding issues 4

identified in Section 1.8 of the Safety Evaluation Report and have indicated a favorable resolution of each matter. Therefore, we reaffirm our conclusions as set forth in Section 21.0 of the Safety Evaluation Report.

t 1-1

6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 ContainmentIgnctionalDesign The applicant reUised his containment analysis and mass / energy release data for the postulated \\oss-of-coolant accident analysis in Amendment Ten [ dated August 20, 1976]

of the Prelimin7ry Safety Analysis Report. The results of the applicant's analysis using revised redis/ energy release rates obtained from usir.g the LOCTIC computer code indicate a peak containment pressure of 38.2 pounds per square inch which nearly matches the previous fjgure of 38.3 pounds per square inch. We have performed a confirmatory analysis using the CONTEMPT-LT MOD 26 computer code in checking the pressure and temperature response for the containment. The applicant did not provide sufficient information for us to determine that the mass / energy release rates used in the LOCTIC code were acceptable. The staff therefore used the approved Westinghouse Topical Report WCAP-8312A (Westinghouse mass and energy release data from containment design) (19) in our confirmatory analyses. The WCAP-8312A rates were adjusted two percent higher in order to match Jamesport power level and coolant energy. The staff requires that (for a construction permit) the contlinment peak pressure analysis indicate a pressure at least ten percent less than the design pressure. Our analysis resulted in a peak containment pressure of 38.7 pounds per square inch,This pressure is about fourteen percent less than the design pressure of forty-five pounds per square inch. We, therefore, conclude that the containment analysis is acceptable.

The Safety Evaluation Report previously stated that we were unable to conclude on the operability of safety related instrumentation inside the containment following a postulated main steam line break accident because the peak calculated containnent atmosphere temperature exceeded the containment de:ign temperature, for a short period of time. However, the applicant has comitted to accept the generic resolution of the Westinghouse program to qualify Class IE instrumentation for the temperature transient involving the peak temperature. This program is presently being reviewed by the staff. Westinghouse is currently performing tests on instrumentation and is expected to finalize these tests shortly. Westinghouse will then make a report of these findings to the staff. Therefore, there is reasonable assurance that a generic resolution will be finalized before a decision on issuing the Jamesport operating licenses. There is ample time for any modifications, which may be necessary as a result of the staff's review, to be incorporated into the Jamesport design before an operating license is issued.

6-1

6.2.7 Containment Backpressure Evaluation for Emergency Core Cooling System Performance Appendix K to 10 CFR Part 50 of the Commission's regulations rec,uires that the effect of operation of all the containment installed pressure reducing systems and processes be included in the emergency core cooling system evaluation. The Safety Evaluation Report stated that we would require the minimum containment pressure ant'ysis for the emergency core cooling system evaluation be done using Branch Technical Position Containment Systems Branch 6-1 (minimum containment pressure model for PWR ECCS performance evaluation). This technical position provides a method for detennining containment parameters for plants at the construction permit stage. And it is based on information obtained from nuclear power plants already built. The applicant has presented the containment parameters that were used in his emergency core cooling system evaluation. These agree with the Containment Systems Branch position 6-1 containment parameters. We, therefore, conclude that the minimum containment pressure analysis was done in an acceptable manner, and that this matter 1:; esolved.

r 6.3 Emergency Core Cooling System 6.3.3 Performance Evaluation Analyses have been perfonned for Jamesport, Units 1 and 2, to determine the conse-quences of a postulated loss-of-coolant accident and to assure emergency core cooling system adequacy. These analyses were performed with an evaluation model approved by the NRC that conforms with Appendix K to 10 CFR 50.46 (1, 2, 3, 4, 5,18)

Westinghouse studies (6) indicated that small breaks are not limitirg for Westing-house plants. This was confirmed for Jamesport in Reference 7 and included a spectrum of small breaks.

Westinghouse generic sensitivity studies (8) have shown guillotine breaks in cold leg piping (DECLG) to be the most limiting large breaks for Westinghouse four loop (17 x

17) plants. Loss-of-coolant accident analyses for Jamesport (9,10) confirmed this finding and identified a Moody discharge coefficient of 1.0 to be most limiting. All of the above analyses were performed with evaluation models conforming to Appendix K.

In Aug"st,1976 Westinghouse (11) notified the NRC that measured initial upper head fluid temperatures were higher than those used in prior analyses (cold leg tempera-ture). In conjunction with this disclosure Westinghouse presented sensitivity studies (12) demonstrating that use of tne hot leg temperature (Thot) rather than the cold leg temperature (Tcold) in loss-of-coolant accident analyses produced a higher calculated peak clad temperature but did not alter the worst break type (DECLG).

1 The applicant has re'erenced the above generic study and its previous loss-of-coolant accident analyses to show that the 1.0 DECLG is the most limiting large break. The November.15,1976, submittal (7) noted that there is more than 500 degrees Fahrenheit margin in peak clad temperatore for the limiting small break and that the small break peak clad temperature sensitivity to the upper head temperature is less than 25 6-2

degrees Fahrenheit; thus, small breaks are not limiting when the hot leg temperature is assumed. The staff agrees with the applicants assumption and also concludes that small breaks are not limiting, because the calculated clad temperatures are well below 2200 degrees Fahrenheit.

The applicant has reanalyzed (13,14) the worst break (1.0 DECLG) using the assumption of the hot leg temperature in the upper head to determine the consequences for this design basis accident. The calculated peak cladding temperature is 2195 degrees Fahrenheit which is below the 2200 degrees Fahrenheit criterion specified in 10 CFR 50.46(b).

Maximum calculated local cladding oxidation is 8.7 percent which is below the 17 percent criterion given in 10 CFR 50.46(b). Calculated core wide hydrogen generation is less than 0.3 percent which is below the one percent criterion specified by 10 CFR 50.46(b).

This analysis was performed assuming a power peaking factor of 2.235 and 102 percent reactor power (in accordance with Appendix K,Section I, A of 10 CFR Part 50).

The worst single failure, identified by references 6 and 16 to be the loss of a low pressure safety injection pump, was assumed for the loss-of-coolant accident analyses.

The emergency core cooling system (including the single f tilure of valves, Safety Evaluation Report Section 7.3.3) was reviewed for.cortpliance with the single failure criterion and found acceptable. Staff review of containment parameters, heat sinks and analysis methods is discussed in Section 6.2.7 of the Safety Evaluation Report. The manner of detemining minimum containment pressure was found to be in confomance with Branch Technical Position Containment Systems Branch 6-1, and is, therefore, acceptable.

The emergency core cooling system is realigned upon receipt of a low level sigr:al from the refueling water storage tank to the recirculation mode as discussad in Section 6.3.2 of the Safety Evaluation Report. Simultaneous hot leg and cold ieg injection will be used during long term cooling to preclude excessive boron concentration builuup.

Equipment and procedures have been providad to accomplish the above phases of long term cooling and have been reviewed and found acceptable.

I Westinghouse has generically determined that, twenty-four hours after a postulated loss-of-coolant accident, the emergency core cooling system should be realigned for simul-taneous hot leg injection. This realignment is used to preclude boron concentration buildup in the core. However, this time will be calculated specifically for Jamesport when data becomes available at the operating license stage of plant construction.

Equipment and procedures will be provided (as described in Appendix 6.3A of the James-port Preliminary Safety Analysis Report) to detect, alarm and identify emergency core cooling system passive failures, which may occur outside the containment. This equip-ment and procedures provide assurance that a failure could be isolated within thirty minutes of their occurrence.

The results of the loss-of-coolant accident analy.as show levels of cladding oxidation and peak cladding temperature. The analysis also indicates a cladding temperature decline from its peak of 2195 degrees Fahrenheit.- The emergency core cooling system

'6-3

design for long term cooling ensures that continued cooling will be provided to the core. Therefore, core geometry will remain capable of being cooled, which satisfied the requirements of 10 CFR 50.46(b).

Conclusions The following conclusions have been drawn from review of the Jamesport loss-of-coolant accident analysis:

1)

The analysis was performed with evaluation models approved for conformance with Appendix K to 10 CFR 50 (5,18).

2)

Calculated fuel performance is within the limits of 10 CFR 50.46(b).

3)

Provisions are made for long term cooling and maintenance of coolable core geometry as required by 10 CFR 50.46(b).

4)

Power peaking factor is limited to 2.235.

Therefore, the staff finds the Jamesport. Units 1 and 2, loss-of-coolant accident analysis acceptable.

l 1

l 6-4

REFERENCES 1.

Bordelon, F. M., Massie, H. W., and Zordan, T. A, " Westinghouse ECCS Evaluation Model -

Sumary," WCAP-8339, July 1974.

2.

Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model - Supplementary Infor' nation,"

WCAP-8471 (Proprietary) and WCAP-8472 (Nonproprietary), April 1975.

3.

Eicheldinger, C., " Westinghouse ECCS Evaluation Model - October 1975 version," WCAP-8622 (Proprietary) and WCAP-8623 (Nonproprietary) November 1975.

4.

Letter from C. Eicheldinger of Westinghouse Electric Corporation to D. B. Vassallo of the Nuclear Regulatory Comission, Letter Number NS-CE-924 dated January 23, 1976.

5.

Letter from D. B. Vassallo (NRC) to C. Eiche1dinger (W), May 30, 1975.

6.

Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340, July 1974 (Proprietary) and WCAP-8356, July 1974 (Nonproprietary).

7.

Letter, J. Weismantle (LILCO), to R. Boyd (NRC), JNRC-182, and Enclosure, November 15, 1976.

8.

Johnson, W. J., Massie, H. W. Jr., and Thompson, C. M., " Westinghouse ECCS-Four Loop Plant (17 x 17) Sensitivity Studies," WCAP-8566. July 1975.

9.

Jamesport, Units 1 and 2, Appendix K Analysis (using October 1975 Model and Tcold),

July 27, 1976 and August 19, 1976, submitted by (LILCO) J. Weissmantle to R. Boyd. (NRC)

10. Letter, Weismantle (LILCO), to R. Boyd (NRC), JNRC-170, " Additional Appendix K Informa-tion," August 19, 1976.

11. Letter from C. Eicheldinger of Westinghouse Electric Corporation to V. Stello of the Nuclear Regulatory Commission, letter number NS-CE-1163, dtd. August 13, 1976.

12. Beck, H.

S., Kemper, R. M., " Westinghouse ECCS - Four Loop Plant (17 x 17) Sensitivity Studies wit & !',per Head Fluid Temperature at Thot," WCAP-8865, October 1976.

13. Weismantle, John A. (LILCO), letter to R. Boyd, S'tC, JNRC-180, November 1,1976.

14. JNRC-174, October 8,1976. LILC0 letter, J Weissmantle to R. Boyd(NRC).

15. Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model - Supplementary Information,"

WCAP-8471-P-A, April 1975 (Proprietary) and WCAP-8472-A, April 1975 (Nonproprietary).

6-5

)

REFERENCES (Continued)

16. Safety Evaluation Report, Jamesport, Units 1 and 2. U. S. NRC, NUREG-75/095, October 1975.

17. Safety Evaluation Report, Jamesport, Units 1 and 2, Supplement No.1 U. S. NRC (Suppl. I to NUREG-75/095, April 1976).

18. Letter from D. Vassallo (NRC) to C. Eicheldinger (W), fiay 13, 1976.

19. Approval Letter from D. Vassallo (NRC) to C. Eicheldinger (W), thrch 13, 1975 on WCAP 8312.

l

[

6-6

APPENDIX A UPDATED CHRONOLOGY July 22, 1976 LILCO letter (J. A. Weismantle to R. Boyd) notification of forthcoming changes on steam generator design.

July 27, 1976 LILCO letter (J. A. Weismantle to R. Boyd) Appendix K and Containment evaluations.

August 6, 1976 NRC letter (R. Boyd to A. Wofford), Procedural changes.

August 19, 1976 Amendment 10 to Preliminary Safety Analysis Report submitted.

August 19, 1976 LILCO letter (J. A. Weismantle to R. Boyd) additional infor-mation on design break calculation.

August 26, 1976 LILC0 letter (J. A. Weismantle to R. Boyd) new information on upper head temperature calculation.

September 30, 1976 NRC letter (R. Boyd to A. Wofford) fire protection.

October 8, 1976 LILC0 letter (J. A. Weismantle to R. Boyd) information on limiting break.

November 4,1976 Amendment 11 to Preliminary Safety Analysis Report submitted.

November 12, 1976 LILCO latter (J. A. Weismantle to R. Boyd) LILCO response to fire protection letter.

f November 12, 1976 Meeting held with LILC0 on SER supplement schedule.

1 A-1

APPENDIX B ERRATA FOR SUPPLEMENT 2 Page 20-2 Supplement 2 Para 20.3.1, line 6, change "9.93 to 2.93;" line 7, change

" Refueling Bonds" to " Refunding Bonds."

Page 20-2, Para 20.2.2, line 6, change -- " ingrained in the public" to -- " ingrained in public"--.

Page 20-3, 2nd para, third lire from end, change " common equity ef fects" to " common equity affects"--.

3rd para, first line, char.ge "Since a lengthly" --- to "Since a lengthy"--.

Page 20-8, first para, third line, change "it can be ssen" to -- "it can be seen." --;

fif th line, change "adegate" to " adequate" and "--; "the indenture" to, "the indenture."

Para. 20.4, third line, change "It is inteded" to "It is intended." ---

Para. 20.5, fifth line, change "We do no consider" to "We do not consider."

Third line, change -- " industry pratices" to -- " industry practices."

l 1

4 l

i B-1 l