ML20056F526
| ML20056F526 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/11/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20056F525 | List: |
| References | |
| NUDOCS 9308300009 | |
| Download: ML20056F526 (7) | |
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UNITED STATES j
NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D C. 205 %-4001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
1.0 INTRODUCTION
By letter dated June 16, 1993, as supplemented July 30, 1993, the Power Authority of the State of New York (the licensee) submitted a request for changes to the James A. FitzPatrick Nuclear Power Plant, Technical Specifications (TSs).
The requested changes would provide a one-time extension of the current intervals for various surveillances that are required i
to be performed once each operating cycle. Specifically, the requested changes would extend the current intervals for bench checking and disassembling safety / relief valves in accordance with TSs 4.6.E.1 and 4.6.E.2, l
and for functionally testing 10-percent of each snubber type in accordance i
with TS 4.6.1.3.
The requested changes would also extend the current interval l
for testing instrument line excess flow check valves in accordance with TS 4.7.D.I.b.
The licensee requested a one-time extension until the end of the next refueling outage, currently scheduled to start in January 1995. The only planned plant shutdowns between now and the next refueling outage are two, 3-week maintenance outages scheduled for September 1993 and April 1994.
Following the January 1995 refueling outage, the regular schedules specified in the TSs for these surveillances would be resumed. This safety evaluation only addresses the changes proposed for the currert safety / relief valve and excess flow check valve surveillance intervals. The proposed change to the snubber surveillance interval remains under review by the NRC staff and will be the subject of future correspondence. The July 30, 1993, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION One-time extensions of the normal surveillance intervals were requested by the licensee to allow related testing to be performed during the next scheduled refueling outage. Typically, an 18-month surveillance interval, with a 25 percent extension allowed by TS 4.0.B.1, coincides with an operating cycle from one refueling outage to the next. However, the previous refueling outage lasted for 14 months, from November 1991 to January 1993. For surveillance requirements that can only be accomplished during a refueling outage, the intervals will exceed the normal 18-months for an operating cycle, even with 9308300009 930811 l
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\\ I the 25 percent extension. The first such surveillance will be due in August 1993. To meet the currently required schedules, a lengthy plant shutdown after one-third of the operating cycle would be necessary.
2.1 Safety / Relief Valves The safety / relief valves (SRVs) provide overpressure protection for the reactor vessel and main steam lines. TS 4.6.E.1 requires that at least half of the SRVs be setpoint tested or replaced each operating cycle. A commitment in Licensee Event Report (LER)92-016 further requires that all SRVs, rather l
than half, be setpoint tested each operating cycle. The current scheduled surveillance due dates for the eleven SRVs are between April 3, 1993, and July 18, 1993.
When the 25 percent interval extension as provided for in TS 4.0.B.1 is applied, the due dates are extended to between August 18, 1993, and December 2, 1993. Similarly, TS 4.6.E.2 requires that at least one SRV be 1
disassembled and inspected once each operating cycle. Applying the 25 percent j
interval extension, this surveillance is due on August 30, 1993. The requested one-time TS change would extend the current interval for these SRV surveillance requirements until the refueling outage that is scheduled to begin in January 1995.
The licensee provided test data information which indicates several occurrences of significant setpoint drift in the high direction. Similar industry information indicates that for the plant model Target Rock 2-stage SRV, there can be significant setpoint drift high.
The SRV setpoint affects i
the self-actuation of the valves on high pressure; however, the setpoint does not affect the Automatic Depressurization System or the manual actuation capability. To support the increased surveillance interval described above, i
the licensee performed an analysis to demonstrate that the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) pressure limit of 1375 psig (or 110% of design pressure) would not be exceeded for a condition where nine SRVs drifted high to 1195 psig and two SRVs did not open at all. This would reasonably bound the observed setpoint drift observed in i
the plant data. The analysis also confirmed that the actuation of the SRVs at the higher setpoints would not adversely affect the High Pressure Coolant Injection System, the Reactor Core Isolation Cooling System, the Primary Containment Integrity, the fuel thermal limits, or Emergency Core Cooling System / Loss of Coolant Accident performance.
The staff requested additional information in a letter dated July 6,1993, to determine the effects which the additional time period between surveillances would have on the SRV setpoint drift and to determine the licensee's action plan for reducing the severity of the setpoint drift. The licensee's response dated July 30, 1993, provided SRV setpoint test data for the period of 1983 to 1993. A review of this data indicates that there would not be any increased tendency toward additional drift for longer than normal surveillance periods in the range of the period being requested. Also, during the 14-month period of the last refueling outage, the plant was not operating and no degradation 4
is expected to have occurred. The corrosion induced bonding which causes most
f.-
) I of the setpoint drift takes place in an operating environment and SRV setpoint drift during the extended period between surveillances should be similar to that experienced in normal surveillance intervals.
In order to reduce pilot i
disk-to-seat sticking, the licensee has committed to install new pilot disks containing a catalyst material in half of the valves during the 1995 outage.
The new disks with the catalyst material will reduce the large concentrations
)
of oxygen in the pilot valve area.
If this method fails to reduce the drift problem, the licensee has agreed to consider implementing the addition of additional pressure-sensing power actuation equipment.
During the current operating cycle, the licensee has determined that one SRV is leaking. However, leakage of either the main disk or the pilot disk is not expected to significantly affect the valve setpoint.
l The staff has determined that the licensee has performed an adequate analysis and evaluation to demonstrate the acceptability of the requested one-time extension of the SRV surveillance intervals. Therefore, it is acceptable for i
the licensee to perform the SRV setpoint testing, disassembly and inspection required by TS 4.6.E.1, TS 4.6.E.2, and the LER 92-016 commitments during the refueling outage scheduled to begin January 1995. Thereafter, these i
surveillances will be required at the regular TS intervals.
i 2.3 Instrument line Excess Flow Check Valves i
Instrument piping which is connected to the reactor coolant system, and which exits the primary containment, is designed with excess flow check valves. The instrument piping ends at the instrument connection. Also, each line contains a 0.25-inch restricting orifice inside primary containment and a manual isolation valve upstream of the excess flow check valves. The end of each line is inside the reactor building, but outside primary containment. The 1
excess flow check valves in these lines isolate reactor coolant in the event an instrument-leaks or breaks. The talves isolate by closing on forward flow and thus prevent reactor coolant leakage outside the containment. There are 81 instrument line excess flow check valves installed in the Fitzpatrick plant.
Instrument line excess flow check valves are tested once each operating cycle to verify proper operation as required by current TS 4.7.D.I.b.
The following table lists the test dates and results of this testing for the last 12 years:
Test Date Test Results 08/02/80 100% passed 02/09/82 100% passed 08/23/83 100% passed 05/08/85 100% passed 04/02/87 100% passed 10/28/88-100% passed 06/02/90 97.5% passed 10/05/92 96.3% passed i
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Of the two valves that failed the acceptance criteria in 1990, one failed j
again in 1992. The repeated failure was experienced by valve 29EFV-348. The licensee did not identify any industry data for these valves, but the plant data indicates that the valves have a high reliability.
Even though a 1
degrading trend may be developing, the low likelihood of failure of an instrument line and its associated excess flow check valve during the 14-to month extension does not warrant shutdown to perform the overdue surveillance.
Excess flow check valve testing is performed in conjunction with the reactor pressure vessel (RPV) system leakage test near the end of each refueling l
outage to meet inservice inspection requirements of ASME Code,Section XI.
l The RPV system leakage test is performed at a pressure of approximately 1000 l
psig and demonstrates the integrity of the pressure vessel and the Class 1 l
primary piping. When the pressure is above 600 psig, the excess flow check valves testing can be completed in approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Preparation time I
to perform valve lineups and to pressurize the reactor vessel is approximately 2 days. Depressurization and valve lineup restoration requires an additional i
day.
Therefore, if the plant was shutdown to solely perform the overdue l
surveillances, testing the excess flow check valves would require approximately 4 days. The effort would be essentially the same to test only a percentage of the valves.
RPV system leakage tests are neither planned nor required during the 3-week maintenance outages scheduled for September 1993 and April 1994. Without a one-time change to the TSs, the 3-week outages would have to be extended by approximately 4 days to complete the excess flow check valve testing.
Based on {l) the low failure rate of the valves, (2) the capability to isolate the lines with manual valves upstream of the excess flow check valves in the event an instrument experiences a problem, (3) the extensions to the maintenance outages that would be required to perform testing, and (4) the resumption of the regular schedule following the January 1995 refueling outage, the requested one-time extension of the current interval for the surveillance requirement of TS 4.7.D.I.b is acceptable. However, the licensee should, as a minimum, visually examine the condition of valve 29EFV-34B during the 3-week maintenance outage in September 1993.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released
i offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public coment on such finding (58 FR 36444). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributors:
P. Campbell C. Hammer Date:
August 11, 1993 a
f I
e
I Mr. Ralph E. Beedle August 11, 1993 Mr. Jack Gray of your staff advised me during a telephone discussion on i
August 10, 1993, that this condition of the SE was acceptable. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Reaister notice.
i Sincerely, l
)
Original signed by.
l John E. Menning, Project Manager Project Directorate-I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Enclosures 1.
Amendment No.195 to DPR-59 2.
Safety Evaluation cc w/ enclosures:
4 See next page i
Distribution:
See attached sheet l
l 5
- See previous concurrence f
0FFICE PDI-1:LA PDI-1:PM
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i DATE OFFICIAL RECORD COPY FILENAME: G:\\FITZ\\ FIT 86766.AMD 1;