ML17263A252
| ML17263A252 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/26/1993 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| GL-89-04, GL-89-4, TAC-M83287, NUDOCS 9305050262 | |
| Download: ML17263A252 (13) | |
Text
ACCELERATFD DOCUlVIENTDIST UTION SYSTEM REGULAT INFORMATION DISTRIBUTION STEM (RIDS)
ACCESSION NBR:9305050262 DOC.DATE: 93/04/26 NOTARIZED: NO DOCKET FACIE:!50'-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
05000244 AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.
Rochester Gas
& Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.
Project Directorate I-3
SUBJECT:
Updates NRC on status of open issues in NRC 921020 SE of inservice test program relief requests covering pumps valves. Actions
& proposed alternatives considered guidance in GL 89-04.Relief requests encl.
DISTRIBUTION CODE:
A047D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: Inservice/Testing/Relief from ASME Code NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
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>'IIIIi' IIIIIIIIII ROCHESTER GAS AND ELECTRIC CORPORATION ROBERT C MECREDY Vice President Cinne Nucteer Production
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o 89 EAST AVENUE, ROCHESTER N. Y. 14649-0001 TELEPHONE AREACODE 71B 546 2700 April 26, 1993 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Allen R. Johnson Project Directorate I-3 Washington, D.C.
20555
Subject:
Inservice Testing (IST) Program for Pumps and Valves 1990 1999 Third 10-Year Interval, Revision 1
R.E.
Ginna Nuclear Power Plant Docket No. 50-244 Ref. (a):
(b):
(c):
(d):
(e):
(g)-
Letter from A. R. Johnson, NRC, to R.
C. Mecredy, RG&E,
Subject:
Safety Evaluation IST Program Relief Requests and Response to Anomalic.s (TAC M83287) dated October 20, 1992 Letter from R.
C Mecredy, RG&E, to A. R. Johnson,
- NRC, dated October 29, 1991,
Subject:
Inservice Testing (IST)
Program for Pumps and Valves 1990-1999 Third 10-Year Interval, Revision 1
Letter from R.
C Mecredy, RG&E, to A. R. Johnson,
- NRC, dated April 23, 1992,
Subject:
Inservice Testing (IST)
Program for Pumps and Valves 1990-1999 Third 10-Year Interval, Revision 1
Letter from A. R. Johnson, NRC, to R.
C. Mecredy, RG&E, dated April 15, 1991,
Subject:
R. E. Ginna Nuclear Power Plant Inservice Testing (IST)
Program for Pumps and Valves 1990-1999 Third 10-Year Interval Letter from A. R. Johnson, NRC, to R.
C. Mecredy, RG&E, dated April 20, 1993,
Subject:
Issuance of Amendment No.
52 to Facility Operating License No.
DPR-18 Letter from R.
C. Mecredy, RG&E, to A. R. Johnson,
- NRC,
Subject:
SWSOPI Inspection Report 50-244/91-201, dated Sept.
23, 1992 NUREG-0821, Integrated Plant Safety Assessment for R. E.
Ginna Power Plant, Final Report, dated Dec.
1982
Dear Mr. Johnson:
The purpose of this letter is to update the status of the open issues identified by the NRC Safety Evaluation (SE) of October 20, 1992 (Ref. a).
The SE addresses Rochester Gas and Electric (RG&E)
Corporation's IST Program for pumps and valves.
Our actions and proposed alternatives have considered the guidance delineated in NRC Generic Letter 89-04.
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By letters dated October 29, 1991 (Ref. b) and April 23, 1992 (Ref.
c),
RG&E responded to the original SE, dated April 15, 1991 (Ref.
d) and nineteen anomalies identified in the Technical Evaluation Report (TER).
The NRC responded that three issues required further response by RG&E.
These issues are addressed as follows:
1 ~
Section 2.1 of Ref.
(a) provides the acceptance for Relief Request PR-10, which allowed measurement of discharge pressure of the D/G fuel oil transfer pumps rather than inlet (suction) and differential pressure.
As clarification to the NOTE in Section 2.1.3 of Ref.
(a), the Relief Request PR-10 has been revised to reflect that "pump differential
- pressure, as required by the
- Code, will provide no useful data."
Attachment 1 represents the revised relief request PR-10 for the Emergency Diesel Generator Fuel Oil Transfer Pumps (PDG02A and PDG02B) to accurately reflect the requested alternate test method to measure pump discharge pressure in lieu of differential pressure.
2.
In Relief Request VR-7 RG&E proposed to verify valve remote position indication for the pressurizer safety valves 434 and 435 by moving the valve's coil and observing the appropriate response of the control room indication.
This position indication calibration is per the plant Technical Specifications and is performed annually during each plant refueling outage.
Setpoint testing, as required by Table IWV-3510-1 of the
- Code, is every 5
years.
This relief was
- granted, as documented in section 3.6.1 of Ref. (d), provided that valve position indication is verified to accurately reflect obturator position during the valve setpoint testing.
This provision was identified as Anomaly 11.
RG&E provided resolution to this anomaly in Attachment 1 of our letter dated April 23, 1992 (Ref. c).
In Enclosure (2) to the SER dated October 20, 1992 (Ref. a) it, was indicated that our resolution did not fully address all the concerns of the anomaly.
The concern identified related to ensuring that the interface between the position indication and the Pressurizer Safety Valves (434 and 435) is verified; i.e.
that verification of the position of the valve stem/disk is represented by the position indication at the valve.
RG&E has reviewed this concern and submits the following information:
Procedural controls employed for the calibration of the position indication are adequate to provide sufficient assurance that accurate position indication will occur during safety valve actuation.
The calibration procedure verifies that the position indication lights trip at the appropriate position of safety valve obturator motion.
Following completion of the calibration procedure when the position indicators, which employ linear variable differential
- 2. (cont.)
transducers (LVDTs), are mounted on the safety valves, the proper clearance is obtained to ensure obturator motion will be accurately represented.
This is solely a
mechanical interface and the fitup tolerances are monitored to ensure proper operation.
RG&E has since decided to perform yearly setpoint testing using an outside vendor and steam test loop in lieu of in-house testing using nitrogen as was done in the past.
However, this new test method does not affect the basis for relief request VR-7. It would impose unnecessary hardship and increased cost to attempt to verify position indication during setpoint testing at the test facility, since it would require the fabrication and installation of a
portable position indication test assembly which employs the same type of LVDT, without providing a commensurate increase in quality or safety benefit.
Monitored system parameters also provide assurance of valve position.
Should a pressurize'r safety valve change position during power operation, further assurance of position indication is provided by the continuous monitoring by licensed operators of plant parameters affected by this event.
Indications of pressurizer safety valve opening include, but are not limited to, changing pressurizer
- level, increasing pressurizer relief tank
- level, increasing "tailpipe" temperature downstream of the affected pressurizer safety
- valve, and increased makeup to the Reactor Coolant System.
Therefore, RG&E maintains that this anomaly should be considered resolved and that the current method of calibration, performed on a
yearly basis during each refueling, ensures that position indication
= accurately reflects obturator position.
3 ~
In Relief Request VR-18 RG&E proposed to utilize quarterly fail-safe testing during the diesel generator testing as a
means of demonstrating that the four rapid-acting D/G fuel oil transfer solenoid valves (5907,
- 5907A, 5908, 5908A) properly operate.
Relief was requested from the stroke time measurement since the valves are totally enclosed and have no visible indication of valve position.
NRC agreed (Section 3.12.1 of Ref. d) that it is impractical to measure the stroke times of these valves, because there is no way to determine when a valve receives an actuation signal or when it completes its travel.
These rapid-acting valves stroke almost instantly.
It was acknowledged that it would be burdensome and costly to replace the valves with valves having position indication.
NRC granted interim relief and also requested that some means be developed to measure the full-stroke time of these valves to monitor valve condition and detect degradation.
This was identified as Anomaly 14.
In response
3.
(cont. )
to RG&E's proposed resolution provided in a letter dated April 23, 1992 (Ref. c),
NRC granted additional interim relief, since it was agreed that the monthly diesel generator operational tests and quarterly fail-safe test of the valves ensure that these valves are operating properly.
However, it was additionally requested that some means of detecting valve degradation be provided such as non-intrusive diagnostic techniques, or that enhanced preventative maintenance be conducted.
RGGE has reviewed the NRC concern and options discussed and submits the following information:
a 0 b.
C ~
d.
RGErE has evaluated these SOVs as part of the Ginna Station Reliability Centered Maintenance Program.
The results of this evaluation determined that these SOVs do not justify performance of enhanced preventative maintenance based on the service conditions and failure history.
No history of failure exists for these SOVs in the EDG fuel oil system.
The service conditions of the valves are not conducive to initiating failures since they are located in a controlled environment.
A search of the current Nuclear Plant Reliability Data System conducted for the affected SOV model revealed a
low number of failures (four), over a
20 year period, none of which were in fuel oil systems.
These SOVs are rated for continuous
- duty, even though they are normally de-energized.
Also, since they are D.C.
- powered, they are not subject to accelerated electrical degradation due to current inrush.
In 1991, Engineering Work Request (EWR) 4526 completed a
system upgrade and setpoint verification for the EDG fuel oil system.
As part of EWR 4526, the operation of these SOVs with respect to maintaining fuel oil day tank level was evaluated.
A design analysis was performed which demonstrated that adequate time (in excess of one hour) was available for operators to respond to day tank level alarms and mitigate any adverse effects in the event of SOV failure.
e.
RGGEs position relative to non-intrusive diagnostic test methods and the equipment necessary to support such
- tests, has not changed.
Based upon the very low number of failures documented and the limited number of solenoid valves requiring testing at Ginna Station, the high cost of such equipment has not been justified.
Application of a 2-second time limit for stroke time testing to determine operability of the valve as established in GL 89-04, Attachment 1, Position 6,
does not appear to provide an acceptable means of detecting degradation, prior to failure, in view of the rapid valve response expected for these type of valves and the experience that if the valve does not operate promptly, it usually will not operate at all.
Consequently, RG&E does not believe there is a benefit to applying this criteria.
Based upon our examination of the failure history in NPRDS and at Ginna Station, the mild service conditions of the valves, and the capability to manually respond to a failure, even if it should
- occur, RG&E maintains that the alternative requirements in Relief Request VR-18 are adequate to ensure the operability of these SOVs.
We, therefore, recommend
- anomaly, 14 be considered resolved.
The following discussion identifies three groups of valves whose inservice test requirements have been changed.
For the first group (Item 1. below), this has resulted in our submittal of a new relief request.
For the second group (Item 2.
below),
we incorrectly represented several valve classifications and test requirements.
For the third group (Item 3.
below),
a recent Technical Specification amendment has resulted in a change to the type of test we will perform.
1 ~
In the RG&E letter dated April 23, 1992 (Ref.
c),
our resolution to Anomaly g4 for Relief Request GR-6 was withdrawn indicating our intent to,perform stroke time measurement of several valves quarterly in lieu of annually during refueling shutdowns.
Originally, Relief Request GR-6 sought relief for the quarterly stroke timing of hand control valves 142,
- 4297, 4298, 4480 and 4481.
Resolution was that valve 142 would be stroke time tested during cold shutdown per cold shutdown justification CS-23 and that valves 4297, 4298, 4480, and 4481 would be exercised quarterly but not stroke time tested.
Our evaluation of an acceptable repeatable test method has resulted in the need to submit a
new relief.request VR-30.
This relief request is submitted as attachment 2.
Relief is requested to stroke time the Turbine-Driven Auxiliary Feedwater Pump Discharge Flow Control Valves 4297 and 4298 and the Motor-Driven Auxiliary Feedwater Pumps A and B Discharge Bypass Flow Control Valves 4480 and 4481annually during the refueling shutdown.
Relief is requested
- since, in order to stroke time these valves, they must be rendered inoperable.
This would require entering a Limiting Condition for Operation (LCO) during power operation.
Additionally, valves 4297 and 4298 perform their safety function by opening and they are normally open during power operation.
Valves 4297 and 4298, in the event of electrical or pneumatic failure, will fail open to perform their safety function.
Valves 4480 and 4481
- 1. (cont.)
perform their safety function by closing and they are normally closed during power operation.
They fail closed in the event of loss of power or air.
RG&E proposes to measure and evaluate stroke times during refueling shutdowns and exercise and fail-safe test them quarterly.
2.
In the RG&E letter dated September 23, 1992 (Ref. f), as a
result of changing the quarterly service water pump test method as requested by Relief Request PR-12, it-was stated that the twelve manual butterfly valves serving the inlets and outlets to the four containment recirculation'an coolers and the two reactor compartment coolers (valves 4627,
- 4628, 4629,
- 4630, 4635,
- 4636, 4641,
- 4642, 4643,
- 4644, 4757, and 4758) would be reclassified from Category A-Active to Category A-Passive.
It was also stated that a position indication test would be added for all twelve valves.
The correct
'requirements for these valves are as follows:
a)
Cooler Outlets (4629,
- 4630, 4636,
- 4643, 4644, and 4758) are Category A Passive and require a valve leak rate test in accordance with ASME,Section XI, IWV-3420 each refueling shutdown.
No position indication test is required since these valves do not have remote position indication.
This is consistent with the staff position presented in Section 4.22.3 of NUREG-0821'(Ref. g).
b)
Cooler Inlets (4627,
- 4628, 4635,
- 4641, 4642, and 4757) and no longer required to,be in the IST Program.
These valves presently have a requirement for quarterly stroke testing.
- However, since these valve's only safety function is to remain
- open, RG&E has imposed administrative limits to maintain these valves locked-open.
Based on this
- change, there are no IST requirements for the valves.
These is no safety consequence as a result of this change to the IST program.
3
~
The seat leakage test requirements for the steam generator blowdown and blowdown sample containment isolation air-operated valves 5735,
- 5736, 5737, and 5738 are changed from an 10CFR50 Appendix Z type test to ASME Section XI, IWV-3420 test.
The steam generator blowdown containment isolation manual valves 5701,
- 5702, 5733, and 5734 are deleted from the IST Program.
These changes are in accordance with Amendment No.
52 to the Ginna Station Technical Specifications as approved by the NRC by letter dated April 20, 1993 (Ref. e).
RG&E will forward the revised Appendix C to the Ginna Station Quality Assurance Manual, Inservice Pump and Valve Testing Program for the 1990-1999 Interval reflecting these changes under a
separate submittal in the near future.
KAM/282 Attachment Very truly yours,
/
Robert C. Mecre y xc:
Mr. Allen R. Johnson (Mail Stop 14D1)
Project Directorate I-3 Washington, D.C.
20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector
Attachment 1
RELIEF REQUEST NO.
PR-10 (Revised)
SYSTEM'/G Fuel Oil Transfer System PUMPS:
Diesel Fuel Oil Transfer Pumps (PDG02A, PDG02B)
SAFETY CLASS:
FUNCTION:
Various TEST REQUIREMENT:
The test quantities shown in Section XI, Table IWP-3100-1 (differential pressure in particular) shall be measured or observed and recorded.
BASIS FOR RELIEF The D/G fuel oil transfer pumps are positive displacement type pumps.
The measurement of pump inlet (suction) pressure provides no useful data for evaluation of pump performance or for detecting pump degradation.
Therefore, pump differential pressure, as required by the Code, will provide no useful data.
ALTERNATE TESTING:
Pump discharge pressure shall be measured in lieu of differential pressure per OM-1987, Part 6 Table 2.
Attachment 2
RELIEF REQUEST NO.
VR 30 (New)
SYSTEM:
VALVES:
CATEGORY:
SAFETY CLASS:
FUNCTION:
- 4297, 4298,
- 4480, 4481 3
Valves 4297 and 4298 open to provide and control auxiliary feedwater flow from the turbine-driven auxiliary feedwater pump to the steam generators.
TEST REQUIREMENT:
Valves 4480 and 4481 are bypass flow control valves that close on an SI signal to isolate the steam generators.
During startup, they are used to provide better flow control to the steam generators.
Stroke time of power operated valves shall be measured per IWV-3413 and evaluated per IWV-3417.
BASIS FOR RELIEF:
These valves are hand control valves which operate using a variable set air signal.
They do not have a typical control switch.
Position indication is not directly indicated, only the control air signal is indicated.
Manual activation of these valves is not possible in the present configuration.
Lifting of leads or jumpers for the valve controls would be necessary.
Stroke timing during power operation would require rendering these valves inoperable and entering a Limited Condition for Operation (LCO).
Valves 4297 and 4298 perform their safety function by opening, are normally open and fail open.
Valve 4480 and 4481 perform their safety function by
- closing, are normally closed and fail close.
ALTERNATE TESTING:
Measurement and evaluation of stroke
.times shall be performed during refueling shutdowns.
These valves will be exercised and fail-safe tested quarterly.
0