ML20137P259
| ML20137P259 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 01/24/1986 |
| From: | Olshan L Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8602040451 | |
| Download: ML20137P259 (25) | |
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R 3o UNITED STATES g
E' f kghg o NUCLEAR REGULATORY COMMISSION
.E WASWNGTON, D. C. 20555
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m u en Docket Nos.: STN 50-454 and STN 50-455 APPLICANT:
COMMONWEALTH EDISON COMPANY FACILITIES:
BYRON STATION, UNITS 1 AND 2
SUBJECT:
MEETING
SUMMARY
- REVIEW 0F BYRON LC0 RELAXATION PROGRAM On December 3 and December 6, 1985, meetings were held in Bethesda, Maryland to discuss the review of the Byron LC0 (Limiting Conditio1s for Operation)
Relaxation Program. Members of the NRC, Brookhaven Natio.al Laboratory (BNL),
Commonwealth Edison (CECO) and Westinghouse were present. Attendees at the meetings are listed in Enclosures 1 and 2.
CEC 0 submitted its LC0 Relaxation Program in May 1984 to.iustify extending the allowable outage times (A0T) for certain equipment from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days. The CEC 0 submittal included a Probabliste Risk Assessment (PRA) done for Byron to evaluate the change in risk to the public with the increased A0T. NRC was assisted in its review of the PRA by BNL.
At the December 3 meeting, BNL presented the results of its review using the slides provided in Enclosure 3.
The most significant result of the BNL re iew was that BNL estimated the core melt frequency for Byron 1 to be about 10~j/ year.
For this estimate, BNL assumed one unit operation, loss of service water as an i
initiating event and the probability of a seal LOCA, given loss of service water, is 0.5.
The main contribution to this estimate was the two-pump Essential Service Water (ESW) System on Byron 1.
Loss of ESW results in a loss of reactor coolant pump seal cooling and loss of cooling to ECCS pumps. Thus, when considering the probability of the running ESW pump failing, the standby ESW pump failing to start, the subsequent induced seal LOCA (from loss of reactor coolant pump seal cooling) and the inability to mitigate the LOCA (from loss ofg/ year.ooling to ECCS pumps), BNL estimated the core melt frequency to be abo 10-Most of the December 3 meeting was spent discussing ways to lower the estimated core melt frequency for Byron 1.
The two major areas of discussion were the success criterloc for the cooling tower fans and interim modes of operation until Byron 2 is lice.n ed. When Byron 2 is licensed, its ESW pumps can be cross-tied to Byron 1 if needed. This ability to cross-tie ESW systems signi-ficantly reduces the core melt frequency.
CECO pointed out that the success criteria for the cooling tower fans assumed in its original submittal, and subsequently used in the BNL analysis, was too conservative. Westinghouse and BNL agreed to provide a new estimate of core melt frequency using a revised success criterion for the cooling tower fans.
8602040451 860124 PDR ADOCK 05000454 P
PDR t
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1
D::: 1986
- In addition to the discussion on the core melt frequency, BNL also presented the results of its review concerning CEC 0's request to increase the A0Ts on certain equipment. BNL indicated that the increase in core melt frequency when increasing the A0T from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days is negligible for the following systems: containment heat removal systems (containment spray pumps and fan coolers), ECCS (charging pumps, SI pumps and RHR pumps), and component cooling. For the auxiliary feedwater pumps, the effect of the increased A0T is slichtly greater, and for the diesel generators and ESW pumps the effect is greater still.
(See Enclosure 3 for additional details).
During a conference call on December 4, Westinghouse stated that results of recent tests conducted in France indicated that the assumptions used concerning the induced seal LOCA were conservative.
Thus, on December 6 another meeting was held to follow-up the discussions held on December 3 and 4.
Slides used by CEC 0 are included in Enclosure 4.
CECO began the meeting with a brief description of the ESW system for Byron and summary of PRA results determined by Westinghouse and BNL.
In its discussion, CEC 0 demonstrated that the original success criterion for cooling tower fans was, in fact, too conservative.
CECO then committed to have at least one of the Byron 2 ESW pumps "available" for cross-tie to Byro.n 1 before Byron 1 enters Mode 4; i.e, subcritical and reactor coolant temperature greater 200*F. Byron 1 had been shutdown since October 25, 1985 for reasons unrelated to this discussion.
BNL then presented its results considering the revisad success criterion for cooling tower fans and having a Byron 2 ESW pump available. However, BNL also felt that the failure rate of the ESW pumps assumed in the Westinghouse analysis was not conservative enough. Thus, assuming the revised success criteria for the cooling tower fans and an increased value of for the failure rate f the ESW pumps, BNL estimated the core melt frequency to be still about 10-g/ year.
However, having a Byron 2 ESW pump ava{/ year.lable reduces the core melt f by about a factor of ten, to about 10~
A member of the NRC asked whether Byron met the single failure criteria. The NRC member postulated the loss of one ESW pump to be the transient and the single failure to be the failure of the other pump to start. This scenario, as previously discussed, could lead to core melt. Westinghouse responded by quoting from ANSI /ANS-58.9-1981, " Single Failure Criteria for Light Water Reactor Safety-Related Fluid System," which was written by a group comprised on industry and NRC representatives. Paragraph 4.5 of this document states, "Where the initiating event is the postulated failure of one or more redundant trains of a dual purpose safety-related fluid system, i.e., one required to operate during Condition I as well as to shut down the reactor and mitigate the consequences of the initiating event, a single failure in the remaining l
, i
NN 2 41986
. train, or trains, of that system shall not be assumed, provided that the system is designed to Seismic Category I standards, is capable of being powered from both off-site and onsite sources, and is constructed, operated, and inspected to quality assurance, testing, and inservice inspection standards appropriate for safety classes 1, 2 or 3."
This statement is reiterated in paragraph 3.2.1.d. of ANSI / ANSI-51.1-1983. Westinghouse concluded that the scenario postulated by the NRC member was not appropriate and that the Byron ESW system met the single failure criteria.
Finally, Westinghouse briefly discussed the results of recent testing in France. Westinghouse believes.that the test results demonstrate that the assumption in the PRA study regarding the probability of a induced seal LOCA, given loss of seal cooling, is conservative. Westinghouse believes that the test results indicate that there may be sufficient time to take corrective action, such as repairing an ESW pump, before the induced seal LOCA leads to core melt.
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L. N. 01shan, Project Manager PWR Project Directorate #5 Division of PWR Licensing-A
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Mr. Dennis L. Farrar Byron Station Commonwealth Edison Company Units 1 and 2 cc:
Mr. William Kortier Ms. Diane Chavez Atomic Power Distribution 528 Gregory Street Westinghouse Electric Corporation Rockford, Illinois 61108 Post Office Box 355 Pittsburgh, Pennsylvania 15230 Regional Administrator, Region III U. S. Nuclear Regulatory Commission Michael Miller 799 Roosevelt Road Isham, Lincoln & Beale Glen Ellyn, Illinois 60137 One First National Plaza 42nd Floor Joseph Gallo, Esq.
Chicago, Illinois 60603 Isham, Lincoln & Beale Suite 840 Mrs. Phillip B. Johnson 1120 Connecticut Avenue, N.W.
1907 Stratford Lane Washington, D. C.
20036 Rockford, Illinois 61107 Douglass Cassel, Esq.
Dr. Bruce von Zellen 109 N. Dearborn Street Department of Biological Sciences Suite 1300 Northern Illinois University Chicago, Illinois 60602 DeKalb, Illinois 61107 Ms. Pat Morrison Mr. Edward R. Crass 5568 Thunderidge Drive Nuclear Safeguards & Licensing Rockford, Illinois 61107 Sargent & Lundy Engineers 55 East Monroe Street Ms. Lorraine Creek Chicago, Illinois 60603 Rt. 1, Box 182 Manteno, Illinois 60950 Mr. Julian Hinds U. S. Nuclear Regulatory Commission Byron / Resident Inspectors Offices 4448 German Church Road Byron, Illinois 61010 Mr. Michael C. Parker, Chief i
Division of Engineering Illinois Department of Nuclear Safety 1035 Outer Park Drive Springfield, Illinois 62704
-..-.x.
MEETING NOTICE DISTRIBUTION 1
0ocket:or Central File NRC Participants NKC PDR Local PDR L. Olshan PD#5 Reading File ORAS
- v. Benaroya-A.
Buslik H. Denton S.
Israel Project Manager L. Olshan J. Jackson OELD J. Milhem E. Jordan C. Moon B. Grimes J. Rosenthal J. Partlow (Emergency Preparedness only)
A.
Spano Receptionist (Bldg. where meeting is being held)
C. Berlinger ACRS (10)
OPA F. Congel W.
Gammill N. Olson J. Milhem Resident Inspector V. Noonan Regional Administrator T. Novak J.
Shapaker B.
Sheron i
t 4
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cc: Licensee / applicant & Service List 1
I
m ENCLOSURE 1 REVIEW 0F BYRON LEO RELAXATION PROGRAM DECEMBER 3, 1985 NRC Brookhaven National Laboratory V. Benaroya N. Cho A. Busiik T. Chu S. Israel R. Youngblood J. Jackson J. Milhem C. Moon Commonwealth Edison L. Olshan K. Ainger J. Rosenthal D. Farrar A. Spano T. Miosi
- 13. Nelson J. Pausche M. Snow T. Tramm Other Westinghouse i
S. Bazo, Bechtel S. DiTommaso M. Oper S. Sancaktar D. Sharp a
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ENCLOSURE 2 REVIEW 0F BYRON LC0 RELAXATION PROGRAM December 6, 1985 NRC Brookhaven National Laboratory N. Anderson R. Youngblood V. Benaroya C. Berlinger A. Buslik F. Congel Commonwealth Edison W. Gammill K. Ainger J. Jackson D. Farrar J. Milhem J. Pausche V. Noonan M. Snow T. Novak T. Tramm L. Olshan J. Rosenthal J. Shapaker B. Sheron A. Spano Westinghouse T. Burnett F. Cadek Other C. Campen K. Green, Sargent & Lundy S. Dilammaso G. Swindlehurst, Duke (WOG)
M. Oper S. Sancaktar D. Sharp W. Tauche Y
ENCLOSURE 3 BASIC CONSIDERATIONS CORE DAMAGE FREQUENCY FOR ONE UNIT OPERATION LOSS OF SERVICE WATER INITIATOR SEAL LOCA
- PROBABILITY OF SEAL LOCA GIVEN LOSS OF SERVICE WATER; 0 5--0 0
- NO RECOVERY FROM SEAL LOCA FAULT TREE LINKING APPROACH B0UNDING CALCULATIONS AND SENSITIVITY ANALYSIS ON A0T SYSTEM INPORTANCE WITH RESPECT TO A0T t
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BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC(Illl J
INITIATOR FREQUENCY FOR LOSS OF SERVICE WATER SYSTEt1
- ESTIMATE EXPECTED NUMBER OF SYSTEM FAILURES DURING ONE YEAR INTERVAL USE THE FAULT TREE
- MODELING REPAIRS OF AN ALTERNATING SYSTEM REQUIRES RENEWAL THEORETIC CONSIDERATIONS
- FOR CUT SETS I EAB+IEBA T
1 T
[
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BA 0
- } x,Ti, + } x 6, 3
WHERE T = 1 YEAR
- BNL RESULTS
- CASE 1: 1 53X10-3/RY
- CASE 2: 1 14X10-2/RY
- WCAP-10526 SUPPLEMENT RESULTS
- 3 DAY LCO: 1 96X10-*/RY
- 7 DAY LCO: 3.39X10-*/RY BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(Illl
Model 0 - Model in WCAP-10526 with modifications in Section 4.2 (except Event Tree 17) and data in Tables 4.1 and 4.2.
Model 1 - Model 0 plus Event Tree 17 (loss of service water).
Model 2 - Model I with Pr(S -0A) = 0 where S -0A is the seal LOCA event 2
2 given loss of service water.
Model 3 - Model I with no maintenance unavailability contributions from CCWS and ESWS.
Case A0T1 A0T2 1
All systems None 2
None All systems 3
All others DGs 4
All others ESWS 5
All others CHRS (CF and CS) 1 6
All others Chg and SI 7
All others RHR 8
All others AFWS 9
All others CCWS BROOKHAVEN NATIONAL LABORATORY l} lj l A5500ATED UNIVERSITIES, INC.(1Ill
COMPARISON OF MEAN CORE DAMAGE FREQUENCIES FROM ONE UNIT OPERATION CASE WCAP-10526 MODEL 0 MODEL 1 MODEL 2 MODEL 3 1
1 41(-4) 1 88(-4) 1 01(-3) 1.07(-4) 2.77(-4) 2 8 24(-4) 7 15(-3) 7.72(-4) 4 73(-4) 3 3 32(-4) 1 15(-3) 1 18(-4) 4 21(-4) 4 6 18(-4) 6.73(-3) 4 85(-4)
NA" 5
1 88(-4) 1 01(-3) 1 07(-4) 2 80(-4) 6 1 88(-4) 1 01(-3) 1 07(-4) 2 77(-4) 7 1 94(-4) 1 01(-3) 1 13(-4) 2 84(-4) 8 2 26(-4) 1 07(-3) 1.73(-4) 3 18(-4) 9 1.88(-4) 1 01(-3) 1 07(-4)
NA**
10*
1.67(-4) 2 52(-4) 1 63(-3) 1 74(-4) 2 97(-4)
- IR== 34 HOURS FOR ALL SYSTEMS.
- NOT APPLICABLE.
BROOKHAVEN NATIONAL LABORATORY l}l)l A5500ATED UNIVERSITIES, INC.(llll
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BYRON ESSENTIAL SERVICE WATER SYSTEM PRA EVALUATI0NS WESTINGHOUSE BROOKHAVEN NATIONAL LAB CORE DAMAGE FREQUENCY FROM BYRON PRA 2 x 10-5 (ASSUMES 2 UNIT OPERATION WITH SYSTEM CROSS-TIES)
CORE DAMAGE FREQUENCY FROM BYRON LCORP 1.4 x 10-4 1.9 x 10-4 (ASSUMES SINGLE UNIT OPERATION LOSS OF ESW NOT AN INITIATOR)
LOSS OF ESW PROBABILITY 1.9 x 10-4 1.6 x 10-3 PROBABILITY OF RCP SEAL LOCA
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.5 CORE DAMAGE FREQUENCY WITH LOSS OF ESW 1 x 10-4 8 x 10-4 INITIATOR E
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INITIATORS ONE UllIT OPERATION NO CROSS-TIES e
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DESIGN DATA 1
i WET BULB 78 *F WATER TERMPERATURE (LV TOWER) 98 *F 1
SX WATER TEMP (SYSTEM DESIGN) 100 *F i
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HEAT LOAD 1.4 x 10
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1 1^ 3 - PUMP, 1 - UNIT OPERATION i i l i. J THE TOTAL COREMELT FREQUENCY IS ESTIMATED TO BE IN THE RANGE OF 3 = 6 x 10-5 8 x 10-5/ YEAR Q I ' l' l i i Y + 4 i i i 1 ~ j-I s 1 r vse-ww~,- , ~, .e m m.wa,n,, -.e,-,- -.,,, - - - - - ,,,:,,,.n,,. w ,,,,n-- m., w - w-m. ---r, er.,-, .-m s-vwm m,1m e er,-,
r g 3 4 1986 . train, or trains, of that system shall not be assumed, provided that the system is designed to Seismic Category I standards, is capable of being powered from both off-site and onsite sources, and is constructed, operated, and inspected t'o quality assurance, testing, and inservice inspection standards appropriate for safety classes 1, 2 or 3." This statement is reiterated in paragraph 3.2.1.d. of ANSI / ANSI-51.1-1983. Westinghouse concluded that the scenario postulated by the NRC member was not appropriate and that the Byron ESW system met the single failure criteria. Finally, Westinghouse briefly discussed the results of recent testing in France. Westinghouse believes that the test results demonstrate that the assumption in the PRA study regarding the probability of a induced seal LOCA, given loss of seal cooling, is conservative. Westinghouse believes that the test results indicate that there may be sufficient time to take corrective action, such as repairing an ESW pump, before the induced seal LOCA leads to core melt. N l. hjhQ L. N. Olshan, Project Manager PWR Project Directorate #5 Division of PWR Licensing _A ____,fg____/PDe5 0FC :PWR-A NAMPhE01shan DATE : I /N/86 0FFICIAL RECORD COPY t !}}