ML20101S286
| ML20101S286 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/09/1992 |
| From: | Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20101S288 | List: |
| References | |
| NUDOCS 9207170038 | |
| Download: ML20101S286 (5) | |
Text
/
i PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK DLVD.
July 9, 1992 WAYNE, PA 19087 5691 Docket Nos: 50-277 (ris)saosooo 50-278 License Not DPR-44 DPR-56 U.S. Nucienr Regulatory Commission Attn:
Document Control Desk Washington, DC 20555
Subject:
Peach Bottom Atomic Power Station, Units 2
and 3,
Technical Specification Change Request
Reference:
1)
J.W.
Shea, USNRC to G.
J.
Beck, PECo, Request for Additional Information (RAI), date June 11, 1992 2)
G.
J.
- Beck, PECo to
- USNRC, Technical Specification Change Request, date May 18, 1992.
By letter dated May 18, 1992 Philadelphia Electric Company (PECo, reference 2) requested a revision to the Peach Bottom Atomic Power Station, Units 2 and 3 Technical Specifications regarding the frequency of inspection and replacement of Main Steam Safety Valves and Relief Valves.
After reviewing this submittal the NRC staff concluded that some additional information was required and issued a Request for Additional Information (RAI, reference 1).
The specific requests are repeated along with our response to each request.
In addition, as requested in a telephone conversation between G.
J.
Siefert and J.
W.
Shea the revised Technical' Specification page 147 for both units is attached for your review.
Reauest 1:
In the safety discussion, PECO concluded that no time-based failure mechanism is evident from the review of the as-found Surveillance Test (ST) data since 1987.
The licensee is requested to provide information on any Safety or Relief Valve (SRV) ST failures seen during that period, including magnitude and direction of failures and a comparison of the observed setpoint drift with applicable ASME Boiler and Pressure Vessel 20se requirements and guidelines.
l.
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9207170030' 920709 l
yDR ADOCK 05000352 PDR
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x ResD:
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/Y, lists PBAPS Main Steam Safety (SV) and Relief o lves (RV) with as-f ound first-pop set pressures outside Technical b p elfication 2.2.1 tolerances occurring since 1987.
This data is
.c C4M industry-typical, e
p.
As documented in General Electric Licensing Topical Report c-31753P,
" BWN C In-Service Pressure R311ef Technical y
~ation Rev.sien",
Febloary
- 1990, SVs and RVs have
, ally experi'nced some difficulty in eating the 11%
4 s
tolerance criterion following a
cycle of reactor 4
3 e r.
son.
The ilt tolerance used to develop the Technicul
- O tcations stems from the original ASME accer Lance criterion
+
.aw valves or for returning ss1ves to service.
U n ;; the 11%
. clon as an indicator of acceptable in-service performance is g
m
'stic.
ANSI /ASME has acknowledged this and has moc'.fied the in - s.: 'vice testing criterion from ilt to
+3%
per OM-1-1981, "kequirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices."
A review of Attachment 1 shows that 10 of the 12 valves lifted within the current ASME OM-1 neceptcnce criteria of +3%.
OM-1 was first incorporated in ASME XI Section IWV by reference in the Winter 1985 addenda.
ASME B & PV Code Case N-415, f
M "Altern tive Rules for Testing Pressure Relief Devices",
also 9.a allows
- he use of OM-J as an alternative to the requirements in earlier edi' Mns of ASME XI Section IWV for pressure rc li ef devices.
Reauest 2:
Discuss the significcnce of the f ailures described in quastion 1 when compared to the SRV performance assumed in the Updated Final IMfety Analysis Report, specifically setpoint tolerance.
Responce 2:
As part of their safety design basis (UFSAR paras, 4.4.2.1, 4 - 4.6) the RVs/SVs are required to limit peak vessel pressure to the ASME Upset limit of 110% of design pressure (1.1 x le50 psig =
1375 psig).
The overpressure analyses in the UFSAR (Chapters 4, 14, and Appendix K Cxhibit VI) utilize a +1% tolerance en the setpointe of the RVE and SVs.
Technical Specification section 2.2.1 specifies an RV/GV setpoint tolerance of 11%.
The occurrences of as-found setpoints out of Technical Soecification tolerances listed in Attachment I would not have resulted in peak veesel pressare exceeding 1375 psig during an overpressure transient.
Occurrence number 3 will be addressed since it is the limiting case based on number of valves involved and magnitude of drift.
Three RVs had high first pop set pressures and the two SVs had low first pop set pressures.
The as-found setpoints for the remaining i
l 8 RVs are as follows:
1 valve with no as-found data due to excessive body-to-base joint leakage during as-found testing; 2 1
valves- -
unable to locate data; 5
valves within Technical Specification -tolerances
(-0.7,
+0.6,
+0.6,
+0.2,
+0.2%).
The three high RV setpoints would not have resulted in peak vessel pressure exceeding 1375 psig during an overp-essure transient based on the following:
a.
The nuclear system pressure relief system has significant excess capacity as evidenced by the following results from UFSAR Appendix K Exhibit VI for closure of all MSIVs:
- Pedk Vessel Case Press. psic 1.
High neutron flux scram, all RVs/SVs functioning 1260 2.
MSIV pos, suitch scram /only 2 of 13 RVs/SVs functioning
<1375 3.
High neutron flux scram /only 7 of 13 FVs/SVs functioning
<1350 The Case'2 result demonstrates the significant excess capacity available for the expected direct scram.
Case 1, closure of all MSIVs with high neutron flux scram, is the bounding reload overpressure analysis banis. For the case 3 variation, even if 6 of the 13 RVs/SVs were not functioning, peak vessel pressure would remain below the code allowable.
The high RV setpoints in cccurrenct 3 are bounded by case 3, even if the 3 RVs with unknown setpoints are assumed to not function.
Appendix K' Exhibit'VI analyses are based on the original RV nominal setpoints of 1080, 1090,-and 1100.
.Use o'
the case-2 and 3 results for qualitative evaluation of excess pressure relief capacity is acceptable since the Chapter 4
overpressure analysis results for case 1 based on the current 1105, 1115, and 1125 RV setpoints is equivalent - case 1 = 1260 psig, b.
A correlation has been made between an increased RV set point, the delay in RV operation es d the ef fect on the maximum vessel pressure.
The RVs can op o at a later time in the transient and still protect the system from overpressure.
Preliminary overpressure analyses were recently performed to evaluate the
'~
effect of increased delay time on the hVs.
A delay time of 0.6 seconds was used, whereas a delay time of 0.4 seconds is normally used in the reload analyses.
Due to system pressure ramping at the time of RV actuation (approximately 150 psi /sec), this is equivalent to a setpoint increase of between 2% and 3% on all 11 RVs.
This resulted in a 19 psig increase in peak. system pressure for closure of all MSIVs with high neutron flux scram. -Sufficient margin (>90 psi) remained to the 1375 psig code limit, c.
The' low setpoints of the 2 SVs would have resulted in their opening sooner in the transleht, helping e-reduce the peak system pressure.
Picase feel free to contact us if you have any additional questions or concerns.
i Very truly yours, i
f-G.
Beck Manager-Licensing cc:
T. T. Mart.i n, Administrator, Region I, USNRC J._J.
Lyash, USNRC Senior Residet.c Inspecter, PBAPS l-I
i L:-
COMMONWEALTH OF PENNSYLVANIA ss.
COUNTY OF CHESTER D.
R. Helwig, being first duly sworn, deposes and says:
That he is Vice President of Philadelphia Electric Company; the Applicant herein;'that he has read the attached Response to a Request For Additional Information for Peach Bottom ' facility Operating Licenses DPR-44 and DPR-56, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.
j Vice PresidQJ1t G
Subscriber' and sworn to before me this /bH day 1992' P-.
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