ML20246F292
| ML20246F292 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/08/1989 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20246F298 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NYN-89057, NUDOCS 8905120191 | |
| Download: ML20246F292 (7) | |
Text
-
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New Hampshire George s. Thomas
- Vice Presidenf-Nuclear Production Yankee 1
NYN-89057 May 8, 1989 l
United States Nuclear Regulatory Commission Washington, DC 20555 q
Attention: Document Control Desk
References:
a) Facility Operating License NPF-56, Docket No. 50-443 b) PSNH Letter, NYN-87136 dated November 23, 1987, "NUREG-0737, Task II.D.1, Performance Testing of Relief and Safety Valves," G. S. Thomas to USNRC
Subject:
NRC Request for Additional Information Regarding NUREG-0737, Item II.D.1 Gentlemen
~In Reference (b), New Hampshire Yankee (NHY) responded to several NRC questions regarding the applicability of performance testing of safety and relief valves by the Electric Power Research Institute (EPRI) and the analyses performed on the Seabrook Station safety and relief valve piping systems.
The NRC Staff has recently requested further additional information from NHY regarding safety and relief valve load combinations,' inlet water conditions and pressure settings. Responses to these additional information requests are contained in the Enclosure.
Should you have further questions concerning this response, please contact Mr. Robert E. Sweeney in our Bethesda Office at (301) 656-6100.
Very truly yours, 7
J '7 0 George,S.
Thomas Enclosure 8905120191 890508 0
.PDR ADOCK 05000443 i
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New Hampshire Yankee Division of Public Service Company of New Hampshire P.O. Box 300
- Seabrook, NH 03874
- Telephone (603) 47M521 i
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j United-States. Nuclear Regulatory Commission May 8, 1989
~ Attention: Document Control Desk Page 2 c'c :
Mr. William T. Russell Regional Administrator United States. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 i
Mr. Victor Nerses, Project Manager Project Directorate I-3.
United States Nuclear Regulatory Commission
. Division of Reactor Projects Washington, DC 20555 Mr. Dav'id G. Ruscitto NRC Senior Resident Inspector P.O.. Box 1149 I
Seabrook Station, NH 03874 l
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1 I
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~
D ENCLOSURE TO NYN-89057 l
1 Ouestion 1 i
a.
Confirm that the safety valve and relief valve discharge loads were i
combined with earthquake loads.
b.
At what service limit were they combined?
c.
How does Seabrook's loading combinations compare to those in Electric Power Research Institute (EPRI) Report, "PWR Safety and Relief Valve i
Test Program Guide for Application of Valve Test Program Results to Plant-Specific Evaluation" Interim Report, July 1982, Tables 2A and 2B.
Response
a.
As delineated in Attachment 1 hereto, the safety valve and relief valve discharge loads (TR) were combined with the earthquake loads (OBE and SSE). As discussed in NHY letter NYN-87136 [ Reference (b)) in response to RAI 11.E. stresses from individual load cases were, in general, conservatively combined using absolute summation. Justification was also provided to allow the use of Square Root Sum of the Squares (SRSS) meth'd of combination as an option for dynamic loads.
(Sae NOTE 6 of hereto.)
- b. hereto (Table 2A) has been annotated (shown in parentheses) to delineate the service limits used by NHY.
It can be seen that NHY utilized more conservative service limits than required by Table 2A.
c.
All the piping downstream of the pressurizer safety and relief valves is seismically designed and supported: therefore, a comparison of load combinations is made in Attachment 2 hereto with Table 2A only. hereto (Table 2A) has been annotated (shown in parentheses) to delineate the corresponding load combinations used by NHY in the design of the discharge system piping downstream of the j
valves.
The applicable service limits used by NHY are also shown (in parentheses)
Note, the load combinations, presented in Attachment 1 hereto and Attachment 2 hereto, do not differentiate between the relief valve and safety valve transient loadings.
The term TR is used 1
generically to represent transient loading for the piping system under
- review, i l 1
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(
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a 4
l Ouestion 2 l
l In demonstrating safety relief valve operability in relatively cooler 0
0 regimes (i.e., 569 F to 584 F), how does Seabrook consider effects of its inlet liquid water conditions? Are these inlet conditions still valid?
Does WCAP-11677 provide details that show considerations of inlet liquid water conditions at these temperatures and their effects? Please provide WCAP-ll677 or relative portions, if possible.
Response
The inlet water conditions used to determine operability of safety relief valves at Seabrook Station are identified in WCAP-ll677. The site specific conditions in this report are still valid. A copy of WCAP-ll677 is attached hereto as Attachment 3.
Question 3 Reaffirm that Seabrook's safety valve settings were the factory settings used in EPRI test.
Response
The pressure settings were not changed from factory settings of 2,500 psia.
These settings were tested by the EPRI test, arid the applicability of the test to the Seabrook safety valves was discussed in detail in NHY letter SBN-969 dated March 17, 1986.... -
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' ' TABLE 7 ANSI B31.1 - NNS,-I SEISMICALLY DESIGNED PIP!NG,SYSTDiS
-' LOAD COMBINATIONS ~6 STRESS LIMITS **
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4 CONDITION
~ **' COMBINATION CATEGORY
- LIMITS COMBINATION
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Ravision 1 g&t, e nt.2_ y& Jr 3-TABLE 3 l
DEFINITIONS OF LOAD ABBREVIATIONS
=SustainedLoadsDuringNormalPlantdperation N
SOT
= System Operating Transient III SOT
= Relief Valve Discharge Transient
~
g S07
= Safety Valve Discharge Transient 3
SOT
= Max (50Tg 50T ): or Transition Flow F
g
'OBE
= Operating Basis Earthquake SSE
= Safe Shutdown Earthquake MS/FWPB = Main Steam or Feedwater Pipe Break DBPB
= Design Basis Pipe Break LOCA
= Loss of coolant Accident May also include transition flow, if determined that (1) required operatirg proceduras could lead to this con-l dition.
(2)
Although certain transients (.for example loss of load) which are classified as a service level B conditions may actuate the safety valves, the extremely low probability of actual safety valve actu-ation may be used to justify this as. a service level C condition
=
with the limitation that the plant will be shut down for examination after an appropriate number of actuations (to be determined on a i
plant specific basis).
j NOTE:
Plants without an FSAR may use the proposed criteria contained in Tables 1-3.
Plants with an FSAR may use their original design basis in conjunction with the appropriate system operating transient definitions in Table 3r or they may use the proposed' criteria con-tained in Tables 1-3.
l l