ML20023D671
| ML20023D671 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/18/1983 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8306020317 | |
| Download: ML20023D671 (2) | |
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Docket No. 50-312 ORB #4 Rdg DEisenhut OELD NSIC
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Mr. Ronald J. Rodriguez EJordan
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.Sacranento Municipal Utility District ACRS-10
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Executive Director, fluclear JTaylor 6201 S~ Street e SMiner
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,,h P. O. Box 15830 RIngram Sacramento, Califorata 95813 Gray File EBlackwood s-H0rnstein r.t *
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Dear Mr. R0drigue7:
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[l Enclosed is N gNeric Safety Evaluation (SE) supporting continued operation a
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of PWRs which have Dresser safety valves on the primary systen. Rancho
,Seco Unit No. '< is one such plaM/.(
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\\ The review was init'iated when Electric' Power Research Institute testing
} ghowed that certain ring settings did not allow Dresser safety valves to discharqe full rated flow.
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ye concluded tfiat, the Rancno Seco valve ring settings were adequate to neet the require $nininum capacities for the liniting transient for interin operation prior ',o the current refueling shutdown. Any further questions on this issue may be addressed to your Project f4anager.
Sincerely,
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Joha F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing
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Sacramen'.o Municipal Utility Rancho Seco, Docket No. 50-312 Dis tric.t 3~
cc w/ enclosure (s):
David S. Kaplan, Secretary and Christopher Ellison, Esq.
General Counsel Dian Grueuich, Esq.
Sacramento Municipal Utility California Energy Commission District 1111 Howe Avenue 6201 S Street Sacramento, California 95825 P. O. Box 15830 Sacramento, California 95813 Ms. Eleanor Schwartz California State Office 600 Pennsylvania Avenue, S.E., Rm. 201 Sacramento County Board of Supervisors Washington, D. C.
20003 827 7th Street, Room 424 Sacramento, California 95814 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Mr. Robert H. Engelken, Regional Administrator U. S. Nuclear Regulatory Commission, Region V Washington, D. C.
20555 1450 Maria Lane, Suite 210 Resident Inspector / Rancho Seco Walnut Creek, California 94596 c/o U. S. N. R. C.
14410 Twin Cities Road Herald, CA 95G38 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Regional Radiation f.epresentative EPA Reoion IX Alan S. Rosenthal, Chairman 215 Fremont Street Atomic Safety and Licensing San Francisco, California 94111 Appeal Board U. S. Nuclear Regulatory Commission Mr. Robert B. Borsum Washington, D. C.
20555 Babcock & Wilcox Nuclear Power Generation Divisior.
Dr. John H. Buck Suite 220, 7910 Woodmont Avenue Atomic Safety and Licensing Bethesda, Maryland 20814 Appeal Board U. S. Nuclear Regulatory Commission Thomas Baxter, Esq.
Washington, D. C.
20555 Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Christine H. Kohl Washington, D. C.
20036 Atomic Safety and Licensing Appeal Board i
Herbert H. Brown, Esq.
U. S. Nuclear Regulatory Commission l
Lawrence Coe Lanpher, Esq.
Washington, D. C.
20555 Hill, Christopher and Phillips, P.C.
1900 M Street, N.W.
California Department of Health Washington, D. C.
20036 ATTN:
Chief, Environmental Radiation Control Unit Helen Hubbard Radiological Health Section P. O. Box 63 714 P Street, Room 498 Sunol, California 94586 Sacramento, CalifornD. 95814
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UN!TEC STATES i 4.,/ i NUCLEAR REGULATORY COMMISSION 3
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WASmCTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORT OF CONTINUED OPERATION BY PWR'S WITH DRESSER PRIMARY SAFETY VALVES
1.0 INTRODUCTION
As part of TMI Action II.D.1, " Relief and Safety Valve Testing," Electric Powar Research Institute (EPRI) conducted tests on certain models of Dresser safety valves (Models 31739A and 31709NA).
The EPRI tests showed that the 31739A Model did not allow full rated flow at some ring settings.
Also, the tests showed that at appropriate ring settings, full rated flow i
did occur. On October 14, 1982, Duke Power Company declared the Dresser safety valves on Oconee Unit 2 to be inoperable.
Its decision was based on a comparison of the plant's actual ring settings with the EPRI test data and on discrepancies between actual as-found ring settings and the documented ring settin'gs.
In light of the EPRI test data and Duke Power Company's discovery, the NRC gathered information on the safety significance of these problems from PWR vendors, PWR regulatory response groups, and licensees using Dresser safety valves. The safety of each plant with Dresser primary safety valves has been reassessed by an analysis by one of these sources (Refs. 1-11, 13, 15-
- 17. 27-29). Also, some information was gathered from these groups in tele-phone conversations (Refs. 14,18-26).
The EPRI data is presently being reviewed on a generic basis by the staff.
Staff review of the plant specific reports based on the generic EPRI data has also begun. These reviews are underway as part of the staff's evalu-ation of TMI Action Item II.D.1, " Relief and Safety Valve Testing." The staff will evaluate the ring settings recommended by EPRI, Dresser, and PWR NSSS vendors as part of the 11.0.1 review to assure that those recommended settings do provide full rated relief capacity for the various Dresser primary safety valves.
2.0 EVALUATION The purpose of this SE is to provide a justification for continued operation of PWR's which have Dresser Safety valves on the primary system until final resolution of II.D.1 (see Table 1 for a listing of these plants).
The'NRC staff evaluation of this concern for operating PWR's was done in First, the staff determined the minimum primary saftey valve two steps.
relief capacity needed to meet ASME code limits for the most limiting transients. This step was done generically in most cases with separate consideration given to the general design of tiach PWR vendor - B&W, Westinghouse, and CE., In the second step, the staff determined if the Dresser valves at each plant could provide that minimum relief capacity required to satisfy the most limiting transient.
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2.1 Evaluation of Relief Capacity Needed by Transient Analysis Babcock & Wilcox Plants (B&W)
Design analysis used to size primary safety valves on B&W plants is discussed in B&W Topical Report BAW-10043. The limiting event for
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establishing primary safety valve cepacity is the control rod with-drawal accident during startup. The limiting event for establishing required safety valve capacity on the secondary system is the turbine trip event from full power (100% load rejection). Parametric analyses indicate that 50% of rated primary safety valve capacity and 90-95%
of rated secondary safety valve capacity are required to limit peak pressures below the code upset limit (i.e.,110% of design pressure).
FSAR analyses of overpressure transients have been reviewed for B&W plants with Dresser Primary Safety Valves and show results which are consistent with the generic analyses. Peak RCS pressures determined in SAR evaluations of the low power control rod withdrawal event are generally 100-150 psia below the code upset limit of 2750 psia.
In these analyses, energy deposi-tion in the RCS is maximized with respect to reactivity addition rate and primary-to-secondary heat transfer coefficient.
In the limiting analysis i
for peak pressure the power excursion is terminated by a reactor trip on high primary pressure. A high RCS pressure trip setpoint between 2400-2450 psia is assumed in both the safety valve sizing study and FSAR evaluations.
This latter assumption is conservative with respect to actual settings in B&W operating plants today. After the accident at TM1, all B&W operating plants reduced their high RCS pressure trip setpoint from 2400-psia to 2300 psia and raised their PORV lif t setting to 2450 psia. A reduction in high RCS pressure trip setpoint of 100 psia would be expected to significantly reduce the relief capacity requirements.
B&W stated that an evaluation of moderate frequency events indicated that approximately 10% of the capacity of one safety valve would be sufficient overpressure protection. This evaluation excluded the subcritical control rod withdrawal accident and in some cases, took credit for operation of the non-safety grade steam dump and by-pass system. The results of the analyses (i.e.,10% as required capacity) were based on a calculation of peak pres-
' rizer surge flow during the icss of offsite power transient. Al though a 10% valve capacity requirement may have been appropriate during the period immediately following discovery of this problem, when no B&W plants were susceptible to subcritical rod withdrawal accidents and while the dump and bypass system on operating plants were operable, the staff concludes that a 50% valve capacity requirement is appropriate for justification of continued operation for B&W plants until final resolution of Tit! Item II.D.l.
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. Combustion Engineering Plants (CE)
The design basis event for sizing primary and' secondary safety valves in CE plants is the 100% load rejection from full power. Sizing calculations have historically taken no credit for power decay following reactor trip.
In the FSAR analyses of this event, pressure excursions are terminated solely by blowdown through primary and secondary code safety valves and power reduction, following reactor trip on high primary pressure.
Peak primary pressures calculated in current cycle safety analyses of the load rejection event are shown in Table 1.
Except for San Onofre 2-3, peak pressures are shown to be well below the code upset limit (i.e., 2750 psia).
Parametric analyses of the load rejection event for plants other than San Onofre 2-3 have been performed recently by CE in response to the EPRI valve test results. CE has reported that based on these analyses less than 25% of rated valve capacity is required, depending on the plant, to keep the RCS pressure from exceeding the code upset limit during the design basi's load rejection transient. These analyses were performed on a plant specific basis and used present plant operating data for inputing core power, core flow, core inlet temperature, primary and secondary pressure and pressurizer level. Except for the use of a zero moderator temperature coefficient, assumptions were consistent with plant safety analyses. Plant safety analyses normally assume a slightly positive moderator temperature coefficient for conservatism. A positive coefficient means that power would increase as the water temperature increased in the transient, thereby, adding more energy to be dissipated through the safety valves. A zero coefficient means that power would not increase as temperature increased; this would result in a less severe transient. The assumption of a negative coefficient is realistic for these plants. Therefore, except for San Onofre 2-3 and Calvert Cliff 1, the staff concludes that a 25% valve capacity requirement is appropriate for justification of continued operation for CE plants until final resolution of TM1 Item II.D.1.
San Onofre 2-3 As shown in Table 1, the San Onofre 2 design has no excess margin relative ta RCS pressure limits. The reasons for the difference between San Onofre 2-3 and the CE 2560 MWt plant class are two-fold: (1) San Onofre 2-3 runs at a power level 33% higher than the 2560 MWt class plant, but has a primary coolant volume which is less than 10% larger; (2) the initial opening pressure of the steam generator safety valves is set 100 psi higher on San Onofre 2-3 (1100 psia versus 1000 psia).
The elevated secondary safety valve lift pressure coupled with a low steam generator operating pressure results in a significant delay in secondary safety valve lift following a full load rejection. The calculated opening times for secondary and primary safety valves in San Onofre 2-3 FSAR analyses are 10.0 and 10.1 seconds,
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respectively.
In the Calvert Cliffs updated FSAR analyses, for example, these times are 3.7 and 9.8 seconds; resulting in significantly more of the energy in the primary system being dissipated through the steam generators.
Based on this evaluation, the staff concludes that San Onofre 2-3 requires full rated safety valves relief _ capacity to prevent overpressure.
Calvert Cliffs 1 Baltimore Gas and Electric submitted the results of limiting transient analyses done specifically for Calvert Cliffs 1.
Those analyses showed that using conservative calculations,10% of the relief capacity of the safety valves would be required to avoid RCS overpressure. The licensee also performed realistic transient analyses which showed that no relief capacity from the safety valves would be necessary. Additionally, the Power Operated Relief Valves (PORV) are operational at Calvert Cliffs 1, and the PORY block valves are open. The staff concludes that 10% of valve capacity is needed to prevent overpressurization. This relief capacity may be provided by the combination of the PORV's and the safety valves in order to justify continued operation until the safety valves are adjusted in the Fall 1983 outage.
Westinghouse Plants (W)
The only Westinghouse units with Dresser Primary Safety Valves are North un' 1 and 2.
For Westinghouse units, the required primary safety valve capacity is detennined based on the 100% load rejection event with credit for secondary safety valve operation and main feedwater flow, but no credit for power decay following reactor trip.
i Bounding analyses of the load rejection event with reactor trip indicate that 40% of rated primary safety valve capacity is needed in North Anna 1 and 2.
The peak calculated pressure in the North Anna 100% load rejection event (updated FSAR analyses) is 2575 psia which supports this conclusion.
The most limiting event analyzed for North Anna is the Seized RCP Shaf t event with two loops operating and one loop isolated. As shown in Table 1 the peak calculated pressure for this event assuming 100% valve capacity is H 50 psia. This is the pressure that is permitted by the ASME Code Upset i.i mi t.
In analysis of the Seized Shaf t event for the more limiting two-loop Westinghouse plant design peak RCS pressure is shown to remain below 2800 psia when no primary safety valve actuation is assumed, i.e., when pressure suppression is via negative scram reactivity insertion and secondary safety l
valve lift only. This is well below the 120% of design pressure permitted by the ASME Code and accepted by the staff for low probability overpressure t
events in the Standard Review Plan. Therefore, the staff concludes that 40% of rated relief capacity is needed from the primary safety valves to prevent overpressurization.
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TABLE 1 RATED VALVE PEAK RCS***
CAPACITY LIMITING PRESSURE REQUIRED VALVE CAPACITY PLANT (LBS/HR-MWt)
TRANSIENT (PSIA)
(% OF RATED)
OCONEE 1/2/3 232 Subtritical
<2600 50 4
Rod Withdrawal CRYSTAL RIVER 3 234 Subcritical
<2600 50 Rod Withdrawal ARKANSAS 1 334 Subcritical
<2600 50 Rod Withdrawal RANCHO SECO 3G9 Subcritical
<2700 50 Rod Withdrawal
~ PALISADES 235 100% Load
<2400 25 i
Rejection CALVERT CLIFFS 1/2 220 100% Load 2550 25 Rejection 10*
MILLSTONE 2 220 100% Load 2573 25 Rejection MAINE YANKEE 228 100% Load 2689 25 Rejection SAN ONOFRE 2' 298 Feedwater 2860 100 Line Break NORTH ANNA 1/2
> 420 Seized RCP
<2750 40**
Shaft
- 10% is the plant specific calculation for Calvert Cliffs 1.
C* Based on Design Basis Loss of Load Transient.
oo*For 100% Rated Capacity.
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L 2.2 Evaluation of Relief Capacity Currently Provided by Dresser Valves B&W The following B&W operating plants utilize Dresser safety valves on the primary system: Crystal River 3, Oconee 1-3, Arkansas Nuclear One 1, and Rancho Seco. The Dresser safety valve ring settings at each of these plants have been adjusted according to recommendations by EPRI and B&W.
The reconmended settings are intended to provide full rated relief capacity based on the results of the EPRI tests. The staff has made a preliminary review of these recommendations and concludes that these settings wiil allow the valves to provide at least 50% of full rated relief capacity.
Westinghouse The only Westinghouse plants which utilize Dresser safety valves on the primary system are North Anna 1-2.
The licensee has informed the staff of the existing ring settings and stated that the settings provide adequate flow. The ring settings on Unit I have alreach been adjusted for full capacity based on Westinghouse recommendations; the Unit 2 ring settings will be adjusted during the Spring 1983 refueling outage.
The Wastinghouse recommendations are based on the EPRI tests. The staff has made a preliminary review of the existing ring settngs on both Units and concludes that these settings will provide at least 40% of full 3
rated relief capacity.
CE The following CE plants utilize Dresser safety valves on the primary system:
Maine Yankee, San Onofre 2-3, Palisades, Millstone 2, and calvert Cliffs 1-2.
The licensees for these plants all informed the staff of the existing ring settings and stated that the existing ring settings provide adequate relief 3
capacity. The ring settings at San Onofre have been adjusted to provide full rated relief capacity based on the EPRI data. The settings on Calvert i
Cliffs Unit 1 will be adjusted to the recommendations of CE based on the EPRI data the next time the plant is in cold shutdown long enough to perform 1
the adjustment; refueling is scheduled for Fall 1983.
4 The staff has made a preliminary review of these settings and has made the following conclusions.
1.
The ring settings at Palisades, Maine Yankee, Millstone 2, and Calvert Cliffs 2 will allow the valves to provide at least 25% of full rated relief capacity.
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The ring settings at San Onofre 2 & 3 will allow the valves to relieve at their full rated capacity.
3.
The existing ring settings on Calvert Cliffs 1 are not adequately documented; however, regardless of setting, the valves would provide some relief capacity. The PORV's would provide relieving capacity equivalent to at least 10%
of the full rated capacity of the safety valves. The avail-able safety valve capacity in combination with the PORY capacity provides adequate interim protection until the safety valve rings are adjusted.
3.0 CONCLUSION
In Section 2.1, the staff evaluated the minimum primary safety valve relief capacity needed by each plant to meet ASME code limits for the most limiting transients. Except for two plants ti.ese evaluations were done on a generic basis for the NSSS vendors - Westinghouse, CE and B&W. Plant specific evaluations were provided for San Onofre 2-3 and Calvert Cliffs 1.
Table 1 shows the results of these evaluations.
In Section 2.2, the staff evaluated the relief capacity allowed by the current ring settings at the plants with Dresser primary safety valves.
These evaluations were not intended to estimate the existing relief capacity, but to assure that adequate capacity exists to meet the minimums
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- in Table 1.
Except for Calvert Cliffs 1, the staff determined that the e.tisting settings will. provide adequate valve relief capacity to meet the required minimum capacities given in Table 1.
For Calvert Cliffs 1, the staff concluded that the combination of the safety valves and the PORV's provided the necessary minimum capacity. Therefore, the staff concludes that this evaluation provides adequate justification for continued operation of these plants. Each utility with plants using Dresser primary safety valves is studying the EPRI data in concert with vendors and owners groups.
This justification is intended to support continued operation until the proper settings for full rated relief capacity are ascertained by these studies and the rings are adjusted. Plant shutdown solely to adjust the rings is not warranted. If readjustment is necessary, justification for continued operation has been provided until the first outage of sufficient length to allow ring adjustments af ter the proper ring sectings have been determined.
4.0 ACP!OWLEDGEMENT This evaluation was prepared by M. Caruso, F. Cherny, and R. Emch.
Date: APR 18 IS83
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REFERENCES 1.
Letter from J. H. Taylor (B&W) to Dr. R. C. DeYoung, Director IE; October 15, 1982.
2.
Letter from Hal B. Tucker (Duke Power Company) to Harold R. Denton (NRC).
Re: Oconee Nuclear Station; October 14, 1982.
3.
Letter from Dr. Patsy Y. Baynard (FPC) to Mr. Darrell G. Eisenhut (NRC)
Re: Crystal River Unit 3; November 1,1982.
4.
Letter from D. Hukill (GPU) to J. F. Stolz (NRC), Re: Three Mile Island Nuclear Station, Unit 1; October 28, 1982.
5.
Letter from John R. Marshall (APL) to Mr. J. F. Stolz (NRC)
Re: Arkansas Nuclear One, Unit 1; October 19, 1982.
6.
Letter from John H. Mattimoe (SMUD) to J. F. Stolz (NRC); Re: Rancho Seco Nuclear Generating Station Unit 1, Safety Valve Operability; October 19, 1982.
7.
Letter from John J. Mattimoe (SMUD) to J. F. Stolz (NRC): Re: Rancho Seco Nuclear Generating Station Unit 1, Safety Valve Operability; October 21, 1982.
8.
Letter from David J. VandeWalle (RRG for CE Plants) to Mr. Darrell G. Eisenhut (NRC); Re: Pressurizer Safety Valve Operability; October 18, 1982.
9.
Letter from E. P. Rahe (Westinghouse Electric Corp.) to Mr. Darrell G. Eisenhut (NRC); October 15, 1982.
- 10. Babcock and Wilcox Topical Report, " Overpressure Protection for Babcock and Wilcox Pressurized Water Reactors," BAW-10043; May 1972.
- 11. Cooper, K., et al., " Topical keport: Overpressure Protection for-Westinghouse Pressurized Water Reactors," Westinghouse NES Report, WCAP-7769, Revision 1; June 1972
- 12. NUREG-0800, USNRC Standard Review Plan, Rev. 2, 1981, Sections 15.2.8 and 15.3.3.
13.
E. M. Burns et al., " Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program,"
WCAP-10105, June 1982.
- 14. Telephone conversation between J. Kingseed and R. Mitchell (Combustion i
Engineering) and M. A. Caruso (NRC),11:30 EST, November 30, 1982.
- 15. Letter from J. H. Taylor to R. C. DeYoung dated December 23, 1982.
1
REFERENCES (Cont.)
- 16. Letter from Arthur Lundvall, Jr. (BG&E) to Robert Clark (NRC);
Re: Calvert Cliffs Units 1 and 2, January 6,1983.
- 17. Letter from Author Lundvall, Jr. (BG8E) to Robert Clark (NRC);
Re: Calvert Cliffs Units 1 ano 2, January 17, 1983.
- 18. Telephone conversation between representatives of SMUD and
. Richard Emch, Jr., et al. (NRC); Re: Rancho Seco; November 30, 1982.
- 19. Telephone conversation between representatives of VEPC0 and Richard Emch, Jr. et al. (NRC); Re: North Anna Units 1 and 2; November 30, 1982.
- 20. Telephone conversation between representatives of BG&E and Richard Emch, Jr., et al. (NRC); Re:
Calvert Cliffs Units 1 and 2; November 30, 1982.
- 21. Telephone conversation between representatives of APL and Richard Emch, J r., et al. (NRC) Re: Arkansas Nuclear One, Unit 1; November 30, 1982.
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- 22. Telephone conversation between representatives of NE Nuclear and E. Conner (NRC); Re: Millstone Unit 2; December 16, 1982.
??. Telephone conversation between representatives of CPC and T. Wambach (NRC); Re: Palisades; December 16, 1982.
- 24. Telephone cony'rsation between representatives of Maine Yankee Atomic Power and D. iaffe (NRC); Re: Maine Yankee; December 17, 1982.
- 25. Telephone conversations between representatives of BG&E and Richard Emch, Jr., et al. (NRC); Re:
Calvert Cliffs Units 1 and 2; December 28 and December 30, 1982.
- 26. Telephone conversation between representatives of B&W and Richard Emch, Jr., et al. (NRC); November 30, 1982.
- 27. Letter from K. P. Baskin (SCE) to F. Miraglia (NRC); Re: San Onofre 2; June 29, 1982.
- 28. Letter from W. G. Counsil (NU) to D. G. Eisenhut (NRC); Re: Millstone 2; January 6,1983,
- 29. Letter from W. L. Stewart (VEPCO) to Harold Denton (NRC); Re: North Anna Units 1 and 2; December 7,1982.
ATTACHMENT 2 PLANT SPECIFIC INFORMATION OF DRESSER PRIMARY SAFETY VALVES Crystal River 3 This plant uses the smaller of the two Dresser safety valves tested in the EPRI program Model #31739A. The licensee has changed the ring settings to those which are considered " appropriate" as determined in the EPRI program.
Based on its review of information provided by Babcock & Wilcox, the staff has concluded that the mvised settings mcommended by B&W will allow the valves to provide at least 50% of their ASME rated capacity.
Oconee 1-3 All valves have been adjusted to provide full rated flow based on recomendations by B&W, EPRI, and Dresser. We conclude that the valves as presently set will provide at least 50% of rated relief capacity.
Arkansas Nuclear One - Unit 1 The plant utilizes two Model 31759A Dresser valves.
This model of valve was not specifically tested in the EPRI program and is between the sizes of the two that were. It is closer in size to the smaller of the two tested valves, the 31739A. The licensee has received two very similar recommendations for optimum ring settings for this valve from both EPRI and B&W. The ring settings have been adjusted to the B&W recommended optimum values.
The staff has evaluated the B&W recommended settings and concluded that the ANO-1 valves set to B&W recommendation will achieve at least 50% of their ASME rated capacity.
Ranch Seco The plant is currently shutdown for refueling; the licensee has committed to adjust the ring settings as recommended by B&W before restart. The plant utilizes two Model 31759A Dresser valves. As noted above for ANO-1, this model of valve was not specifically tested in the EPRI program.
These adjustments to the recommended settings are intended to assure that the valves will attain 100% of their ASME rated capacity. The staff concludes that the valves will provide at least 50% of rated relief capacity when they are reset to the B&W recommendations.
k
. Westinghouse NSSS North Anna 1-2
.These pla 's each utilize three Model 31759A Dresser valves. This model valve was nct specifically tested in the EPRI program.
The existing ring settings have been reviewed and evaluated by Westinghouse.
The licensee provided "as found" ring settings for the Unit 1 safety valves and the optimum settings recommended by Westinghouse to assure the valves could achieve full ASME rated capacity. The licensee noted that the Unit i valves had already been reset to the Westinghouse recommended settings.
The licensee has stated that the Unit 2 valves have the same settings that the B and C valves on Unit 1 had in the "as found" condition. For Unit 2, the licensee has committed to reset the valves to the Westinghouse recommended. settings at the next refueling outage in Spring 1983.
I The staff has reviewed the "as found" ring settings provided by the licensee
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and the Westinghouse recommended settings. The licensee advised that Westinghouse had concluded that valves with the "as found" settings would be able to achieve 70-90% of ASME rated capacity and that valves with the recommended settings should be able to relieve at their full ASME rated capaci ty.
Based on staff review of the information available from the EPRI test program, the licensee and Westinghouse, the staff concludes that there is not sufficient basis for it to agree with the Westinghouse conclusion of 70-90% rated relieving capacity for the current Unit 2 settings.
However, the staff has concluded that valves with either the "as found" settings or the Westinghouse recommended settings will provide at least 40% of their ASME rated capacity.
Combustion Engineering NSSS San Onofre 2-3 These plants each utilize two Model 31709NA Dresser safetf' valves. This model valve is the same as one of the models that was tested in the EPRI program. The valve rings have been set at one of the settings that provided acceptable performance in the EPRI program.
Based on review of the EPRI data to date, the staff has concluded that there is sufficient assurance that t e valves would relieve at their full ASME rated relieving capacity.
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! Maine Yankee This plant utilizes three Model 31709KA safety valves. This model of valve was not one of the two models tested in the EPRI program.
It is' smaller in size than both of the tested valves. Information was provided by EPRI and Dresser justifying how the test results from the tested valves can be extrapolated to this valve. Also, the licensee advised the staff of the existing ring settings for the Maine Yankee safety valves. While revised ring adjustments arc being determined the licensee is following the recommen-dations of Combustion Engineering Infobulletin 82-09. The CE bulletin recommended (1) continuing study of the EPRI data, (2) alerting operating personnel of the problem, and (3) verification and adjustment of the ring settings consistent with EPRI results on a schedule consistent with the licensee's overall program to upgrade valve performance. Based on review of the EPRI data and the reported licensee ring settings, the staff concludes that the existing settings will provide at least 25% of rated relieving capaci ty.
Pali sades This plant utilizes Dresser Model 31739A safety valves. The licensee is following the recommendations of Combustion Engineering Infobulletin 82-09.
The staff has been advised by the licensee of the existing ring settings for the Palisades pressurizer safety valves. The next refueling outage is cheduled for Fall 1983; ring readjustment, if necessary to achieve full rated capacity, could be performed during that outage. Based on review of the EPRI data and the existing ring setting, the staff concludes that the existing settings would provide at least 25% of rated relieving capacity.
Millstone 2 This plant utilizes Dresser Model 31739A safety valves. The licensee is following the recommendations of Combustion Engineering Infobulletin 82-09.
The NRC staff has been advised by the licensee of the existing ring settings for the Millstone 2 pressurizer safety valves. Also, the licensee has informed the staff that it intends to adjust the rings settings to the settings recommended by CE to provide full rated capacity in Spring 1983.
Based on review of the EPRI data and the existing ring settings, the staff concludes that the existing settings would provide at least 25% of rated relieving capacity.
Calvert Cliffs 2 This plant utilizes two Dresser Model 31739A safety valves. The valves on Unit T have been adjusted in accordance with settings recommended by the valve manufacturer, Dresser. The licensee has advised the staff of the revised ring settings.for the valves on Unit 2.
The staff has reviewed these settings against the EPRI data and have concluced that they would provide at least 25% of ASME rated capacity.
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. Calvert Cliffs 1 This plant utilizes two Dresser Model 31739A safety valves. The licensee provided information on the present ring settings of the safety valves at Calvert Cliffs 1 and on the relief capacity of the safety valves as pre-sently set.
The. licensee has committed to adjust the valve rings on Unit 1 to the recommended settings the next time the plant is in cold shutdown 'long enough to perform the adjustment; refueling is sche & led for Fall 1983.
The licensee can not document the present ring settings of the Unit 1 safety valves. Plant methods and procedures and as well as information from Dresser indicate that the adjustment rings should have the " standard" Dresser settings; however, the documentation is not conclusive. The licensee provided information based on input from CE, Dresser, and EPRI discussing the relief capacity of Dresser valves at various settings and concluded that the valves would provide at least 10% of rated relief
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capacity regardless of ring settings. The licensee also indicated that if the valves have " standard" ring settings, the valves will provide at least 25% of rated relief capacity.
Based on review of the EPRI data, the staff concludes that the valves..ill provide at least 25% of rated relief capacity if they have " standard" ring settings. However, since there is no written record of the existing ring settings, the staff concludes that the licensee does not know the valve settings. Also, Duke Power Company found that some of the valves at Oconee (which were supposed to have " standard settings) had settings which would allow less relief capacity than the " standard" settings. Therefore, the staff concludes that the valve ring settings are probably very close to the " standard" Dresser settings; however, it is possible that the settings are worse than the " standard" settings. Based on its review of the licensee's submittals and the EPRI data, the staff also concludes that the valves on Calvert Cliffs 1 will provide some relief capacity regardless of ring settings, probably as much as 10% of rated relief capacity.
However, the staff finds that the evidence which suggests that the " worst case" capacity would be at least 10% of rated capacity is not conclusive.
Each PORV at Calvert Cliffs 1, has a capacity which is 50% of the full rated capacity of one safety valve or 25% of the combined-Tated capacity of both safety valves; therefore, the PORV's have more than adequate capacity to prevent overpressure during the limiting transient (10%).
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