ML20087K743
| ML20087K743 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/31/1995 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20087K720 | List: |
| References | |
| JAF-RPT-RWR-004, JAF-RPT-RWR-00493-R1, JAF-RPT-RWR-4, JAF-RPT-RWR-493-R1, NUDOCS 9508240122 | |
| Download: ML20087K743 (73) | |
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- i FITZPATRICK - 24 MONTE OPERATING CYCLE NUCuH STD_4 SUPPLY SYSTEMS SURVEIT 7"CE TEST IMPROVEMENTS JAF-RPT-RWR-00493, Revision 1 May, 1995 i
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24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS TABLE OF CONTENTS I.
EXECUTIVE-
SUMMARY
1 II.
PURPOSE 2
III. GENERAL SYSTEM SAFETY FUNCTION 2
IV.
SURVEILLANCE AND MAINTENANCE EVALUATION 4
V.
SUMMARY
AND CONCLUSIONS 25 VI.
REFERENCES 29 ATTACHMENT A SAFETY EVALUATION ATTACHMENT B NSSS OPERATIONAL OCCURRENCE REPORTS ATTACHMENT C SAFETY / RELIEF VALVE DRIFT ANALYSIS ATTACHMENT D DRIFT WORKSHEET FOR ISP-78 ATTACHMENT E MOV WORK REQUEST EVALUATION ATTACHMENT F TECHNICAL SPECIFICATION CHANGES l
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24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS I.
Executive Summary The Fitzpatrick plant will be operating on a 24 month operating cycle.
This' longer-cycle length >has a direct effect on' surveillance testing and maintenance activities that are currently performed on a 18 month or refuel outage basis.
At Fitzpatrick, the' Nuclear Steam Supply System (NSSS) is routinely. inspected, tested, and maintained to provide high-reliability.
This system is subject to tests which verify the operability of several subsystems such as: the Recirculation System, the Main Steam System up to the main steam isolation valves, and the Feedwater System from the feedwater isolation valves.
In addition, preventive maintenance (PM) is periodically performed on individual components.
Test frequencies are mandated by the plants' technical specifications, operatic a' requirements, and inservice inspection schedules.
Maintenance activities are based on-operational feedback and manufacturer's recommendations.
This study evaluates the changes to maintenance and surveillance requirements to support a nominal twenty four month operating cycle.
Justification is provided, where appropriate, to support test extensions.
Our evaluations conc!.ude that 1) the current surveillance test. intervals'can be safely extended to support a nominal 24 month operating cycle, and 2) maintenance activities can be accommodated by the longer cycle.
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24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS II.
Puroose The Fitzpatrick plant will be operating on a 24 month operating cycle.
To avoid either an 18 month surveillance outage or an extended mid-cycle outage, changes are required to the Nuclear Steam Supply System surveillance test intervals prescribed by the Fitzpatrick Technical Specifications.
Substantiating the impacts of the longer-cycle length on the NSSS surveillance, maintenance, and test activities requires a comprehensive review of the system, its individual components, and the integrated effect of all test and maintenance activities on operability.
III.
General System' Safety Function The Nuclear Steam Supply System consists of three subsystems: the Recirculation System, the Main Steam System up to the. main steam isolation valves, and the Feedwater System from the feedwater isolation valves.
These systems combine to ensure there is a sufficient flow of coolant through the reactor vessel in all modes of operation.
Their specific function is described below.
The Recirculation System provides variable coolant flow to the reactor core for the adjustment of reactor power level.
Adjustment of the core coolant flow rate changes reactor power output, thus providing a means of following plant load demand without adjusting control rod position.- The i
system is designed to provide a slow.coastdown of flow so that fuel thermal limits cannot be exceeded as a result of malfunctions.
The arrangement of the system piping is such that a failure cannot compromise the integrity of the floodable inner volume of the reactor vessel.
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The Recirculation System consists of two recirculation loops external to the reactor vessel which provide a piping path for the driving flow of water to the reactor vessel I
jet pumps.
Each external loop contains one variable speed, motor-driven recirculation pump and three motor-operated gate valves which are provided to facilitate pump maintenance.
Each pump discharge line'contains a venturi-type flow element which provides a coolant flow input signal for the Reactor Protection System.
The
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recirculation loops are part of the reactor coolant pressure boundary and are located totally inside the primary containment structure.
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J4 24 MONTil OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS The recirculated coolant consists of saturated water, returned from the steam separators and dryers, which has' been cooled by incoming feedwater.
This water passes down r
' the. annulus between the reactor vessel inner wall and the core shroud.
A portion of the coolant' exits from the vessel and passes through the two external recirculation loops to become the driving flow for the jet pumps.
Each of the two external recirculation-loops discharges high pressure flow into an external manifold from which individual recirculation inlet lines are routed to the jet l
pump risers within the reactor vessel..The remaining portion of the coolant mixture in the annulus becomes the jet pump driven flow.
This flow enters the jet pumps at the suction inlet and is accelerated by the. driving flow.
The flows, both driving and driven, are mixed in the jet pump throat section, pass down through the jet pump diffuser and into the vessel bottom head and up through the.
Core.
The Main Steam System provides high pressure steam from the-boiling water reactor to the turbine-generator unit.
.The l
Main Steam System consists of four high pressure carbon steel lines from the reactor through the main steam isolation valves to the turbine stop and control valves.
The main steam isolation valves automatically close to_.
isolate the Reactor Coolant Pressure BoundaryL(RCPB) in the event of a pipe break downstream from the_ valves.
This automatic closure: 1) ' prevents damage to the fuel barrier by limiting the loss of. reactor coolant, 2) limits the release of radioactive ~ materials by closing the-RCPB in i
case of a gross release of radioactive materials from the reactor fuel to the reactor coolant and steam, and 3) limits release of radioactive materials by closing the i
primary containment barrier in case of a major leak from-the Reactor Coolant System inside the primary containment.
Two isolation valves are installed on each main steam line; one is located inside and the other is located outside the primary containment.
In the event of a main steam line break occurs inside the primary containment, closure of the isolation valve outside the containment acts to seal the primary containment itself.
The MSIVs function in conjunction with the steam line flow restrictors to limit
.the coolant blowdown rate from the reactor. vessel, thereby preventing core damage and excessive release-of radioactivity during a primary steam line break outside the primary containment.
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i 24 h10NTil OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS The main steam piping is designed to carry steam from the reactor vessel through the primary containment to the main steam turbine.
Four steam lines are utilized between the reactor and the turbine.
The'use of these multiple lines permits turbine stop valve and main steam isolation valve tests during plant operation with a minimum amount of load reduction.
To. fully achieve this objective, the.four steam lines are headered through a bypass valve chest upstream of the turbine stop valves.
_This also ensures that the Turbine-Bypass System is connected to the active steam
- 3 lines.
Drain lines are connected to the low points of each.
main steam line inside and outside the drywell.
Each drain line is connected to a high pressure-drain manifold located on the condenser shell through an. orifice and a drain valve mounted in parallel.
The orifice permits continuous draining of the steam lines' low points.
Relief valves are located on the main steam piping.
An air accumulator is provided for each relief valve.
It will hold the valve open for 30 minutes or will permit five actuations following failure of air supply to the accumulators.
The Feedwater System provides the required flow of feedwater to the reactor with sufficient margin to continue flow under all anticipated transient conditions.
The system includes equipment and piping _ downstream of condensate outlet feedwater heater isolation valves that provides the flow path from the Condensate System to the reactor vessel.
This equipment includes two fifty percent capacity turbine-driven centrifugal reactor feed pumps, arranged in parallel to take suction from the fifth point feedwater heater outlets and discharge through the sixth point feedwater heaters and then into the reactor vessel.
IV.
Surveillance Test and Maintenance Activity Evaluation The operability of systems and components required by the plant's safety analyses is established by the surveillance requirements contained in the Technical Specifications.
Surveillance testing, by definition, can only identify that a component or a system is incapable of performing its safety function (i.e.,
Preventive maintenance, however, reduces the number of failures found during plant operation or during testing.
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24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST 151PROVEMENTS Nuclear Steam Supply System test' maintenance, and inspection activities were thoroughly evaluated-to determine the impacts of a 24 month operating cycle.
The longer cycle length' requires an extension of the following tests:
1.
ST-22B 2.
ST-39H 3.
ST-39K 4.
ST-39N 5.
ISP-3-8 6.
ISP-78 7.
ISP-92-1 8.
ISP-92-2 Manual Safety Relief Valve Operation and Valve Monitoring System Functional Test (IST)
Reactor Vessel Operational Pressure Test (ISI)
Shutdown Cooling suction Operational Pressure Test (ISI)
Outboard MSIV Simulated Loss of Instrument Air Drift Test-Reactor Level Indication Instrumentation'(02-3LI-58A)
Calibration Recirculation Pump Temperature Instrument. Calibration Reactor Vessel Safety / Relief Valve Monitoring System (VMS)
Functional Test Reactor Vessel Safety / Relief Valve' Thermocouple Functional Test 5
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- I 24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM
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SURVEILLANCE TEST IMPROVEMENTS A review of the NSSS-preventive maintenance _ activities shows the following maintenance activity currently scheduled for a 1
frequency of 18 months: Reactor Vessel Safety / Relief Valve l
Maintenance (MP-2.4), Main Steam Isolation Valve Maintenance i
(MP-29.1), Main Steam Line Isolation Valve Closure (RPS)
I Position Switch Calibration (MP-29.2), Limitorque Motor Operators - Model SMB/SB Preventive Maintenance, Inspection, and Functional Testing (MP-59.21), Limitorque Motor
-l Operators - Model SMB and SB Preventive Maintenance, j
Inspection, Lubrication, and Testing on Baselined MOVs (MP-59.51), Lubrication of Electric Motors (Without Disassembly) with Grease Lubricated Bearings (MP-101.04),
and Replacement of ASCO Series NP8323 Solenoid Valves (MP-200.1).
All other scheduled NSSS maintenances are not affected by the extension to the nominal 24 month fuel cycle.
1.
Surveillance Test Chances The decision to extend surveillance test intervals considers:
1.
the function of the test in determining overall system i
availability, 2.
the integrated effect of testing and maintenance activities i
on system operability, and 3.
the burden of testing at power.
For example: testing that could lead to a plant transient, testing that results in i
unnecessary equipment wear, or testing that leads to radiation exposure of plant personnel.
These three considerations are applied to an evaluation of Fitzpatrick Nuclear Steam Supply System (NSSS) surveillance j
tests.
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This evaluation also included a study of' specific surveillance l
histories, operational occurrences, and maintenance programs to determine if equipment operability problems are being identified in a timely fashion (References 1, 2, & 3).
Surveillance test data for the past six years (1986-1991) was analyzed for each
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test affected by the extended surveillance interval.
Operational Occurrence Reports involving NSSS components were also analyzed for the last 6 years; a summary of this review is included in Attachment B.
The maintenance review sought to confirm that j
recurring or symptomatic problems affecting opetability are
- j currently being corrected without relying on surveillance tests to identify performance degradation.
Justifications are included for each specific test extension.
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24 h!ONTH OPERATING CYCLE NUCLEAR STEAh1 SUPPLY SYSTEh!
yURVEILLANCE TEST 151PROVEMENTS The following NSSS surveillance tests are performed once per operating cycle or once per refueling outage.
- 1. Manual Safety Relief Valve Operation and Valve Monitoring System Functional Test - IST (ST-22B)
The Pressure Relief System consists of eleven safety / relief valves (SRVs), all of which are located on the main steam lines within the drywell between the reactor vessel and the first main steam isolation valves.
The SRVs can be either automatically actuated by excess steam pressure or opened manually through remote switches.
Their purpose is to prevent overpressurization of the reactor coolant system, which could lead to reactor coolant pressure boundary failure and result in an uncontrolled release of fission products.
Each pilot-operated SRV consists of two principal components: a pilot stage assembly and the main stage assembly.
The SRVs have two modes of operation: spring and relief.
In the spring mode, the spring loaded pilot valve i
opens when steam pressure at the valve inlet overcomes the i
spring force holding the pilot valve closed.
In the relief mode, valves are opened manually or automatically using pressurized nitrogen.
All SRVs can be operated from a remote panel located outside the control room.
i This surveillance procedure consists of manually actuating each SRV to verify proper mechanical function, and also ensure that no blockage exists in the valve discharge line (tailpipe).
Valve operation is verified by observing effects on temperature and acoustic monitors in the SRV tailpipe (Reference 4).
The first consideration in SRV surveillance extension is on-line testing as a determination of SRV system integrity.
The Safety Relief Valve Monitor Instrument Check (ST-22I),
performed on a monthly basis, demonstrates the operability of SRV tailpipe acoustic and temperature monitoring systems by both recording noise levels from each SRV accelerometer and recording thermocouple temperature.
This monthly surveillance ensures that a leaking or partially open SRV will be detected during normal operation (Reference 5).
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24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM.
SURVEILLANCE TEST IMPROVEMENTS Another factor to consider for extension is the current requirement that some pilot assemblies be removed and tested each refueling outage.
The bench testing includes setpoint, leakage, and pilot disk sticking tests.
If necessary, refurbishment is performed in accordance with manufacturer recommendations.
The final consideration for surveillance extension with the nominal 24 month operating cycle is the past performance of the SRVs.
Operating occurrence reports from 1986 to 1991 were reviewed to determine SRV reliability.
This analysis showed that setpoint drift was a concern.
The problem is generic to the industry and is not cycle length dependent.
FitzPatrick has been responsive to the issue, and has submitted new SRV Technical Specifications changes to the NRC in early 1990.
These changes included an adjustment of the maximum permissible setpoint tolerance from one percent.
to three percent (Reference 6).
SRV setpoint drift has been tracked at Fitzpatrick with respect to valve length of service, and does not exhibit an increasing drift trend with a longer service interval.
Extension of SRV l
surveillance testing with the 24 month operating cycle is therefore not precluded since SRV drift is not cycle length i
dependent.
Results from this study are included in Attachment C.
The Manual Safety Relief Valve Operation and Valve Monitoring System Functional Test can be safely extended for the following reasons: 1) leaking or partially open SRVs are readily detected during normal operation, 2) a l
review of past performance shows that SRVs are mechanically reliable with the exception of setpoint drift, and there has been no evidence of tailpipe blockage.
Regarding j
setpoint drift, trending with respect to length of service j
revealed SRV drift not to be cycle length dependent.
1 Subsequent to'the initial issuance of this report (revision 0,
June, 1992), the Authority has assessed potential' drift over a 60 month period since proposed Specification 4.6.E.1 allows _a maximum test frequency of 60 months.
It was 6
concluded that SRV drift, as measured by the surveillance data, is not dependant on the time between surveillances (Reference 42).
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b 24 510NTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEh!
SURVEILLANCE TEST IN1PROVEA1ENTS 2.
Reactor Vessel Operational Pressure Test - ISI (ST-39H)
The purpose of this surveillance is to verify the integrity of the reactor vessel and attached Class'I piping systems.
The leaktightness of the reactor coolant pressure boundary is a major concern due its to importance in protecting e
acainst the relez,se of radioactive material.
Using the c<>ntrol rod drive system.and the reactor water cleanup r,ystem, the reactor vessel and attached piping is filled and pressurized.
During pressure holds at 500 psig and 1000 psig, leakage inspections are conducted (Reference 7).
The first factor in pressure test extension are currently applicable code requirements.
In accordance with the requirements of ASME Section XI, Class I system leakage pressure testing must be performed prior to the reactor going critical following a refueling outage.
Section XI does not stipulate a maximum duration between refueling outages; so test extension is not precluded.
The frequency requirement for leakage pressure tests is instead based on the repair, replacement, and/or the disassembly and reassembly of the class I system pressure boundary (Reference 8).
Another consideration in surveillance extension is the measurement of leakage from the reactor coolant pressure boundary during normal plant operation.
An increase in.
j leakage would be readily detected by the drywell sump monitoring system and/or the drywell continuous atmosphere radioactivity monitoring system.
In addition, excessive leakage would affect drywell temperature and pressure, which are continuously monitored (Reference 9).
The Reactor Vessel Operational Pressure Test can be safely extended for the following reasons:.1) ASME Section XI requirements do not preclude pressure test extension, 2) the refueling outage is the most appropriate ticc to conduct this test because the RCS has been disassembled and reassembled, and 3) leakage from the reactor coolant 1
pressure boundary would be readily detected.
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,w 24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEN1 SURVEILLANCE TEST IN1 PROVEN 1ENTfi
- 3.. Shutdown Cooling Suction Operational Pressure Test ISI (ST-39K)
The purpose of this test is to verify the integrity of the shutdown cooling suction line between 10MOV-17-and 10MOV-18.
The procedure consists of pressurizing the class g
I piping between the two valves to 1005 psig and holding l
for a 10 minute interval.
A visual examination.of piping for evidence of leakage is then performed.
The test is completed satisfactorily if there is no leakage observed from the welded pressure boundary between 10MOV-17 and 10MOV-18 (Reference 10).
The first consideration in extension of this surveillance with the longer operating cycle is the burden of testing at l
power.
The Class I piping between the two valves is isolated from the recirculation system during normal' operation.
Both motor operated isolation valves are interlocked with an RCS signal such that they cannot be opened unless the reactor pressure is less than 75 psig.
Therefore, it is impractical to perform this leak test at times other than the refueling outage (Reference 11).
Another factor in test extension relates to currently applicable code requirements.
In accordance with the requirements of ASME Section XI, Class I_ system leakage i
pressure testing must be performed prior to the reactor going critical following a refueling outage.
Section XI does not stipulate a maximum duration between refueling outages; therefore not precluding test extension.
The frequency requirement for leakage pressure tests is instaad based on repairs, replacements, or the disassembly and reassembly of the Class I system pressure boundary (Reference 8).
Another consideration in surveillance extension is the measurement of leakage from the' shutdown cooling piping during the system operation.
Leakage inside the drywell would be readily detected by the drywell sump monitoring system and/or the drywell continuous atmosphere radioactivity monitoring system.
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24 MONTII OPERATING CYCLE
~ NUCLEAR STEAM SUPPLY SYSTEM n
SURVEILLANCE TEST IMPROVEMENTS
- The Shutdown Cooling Suction Operational Pressure Test can be safely extended for the following reasons: 1) the
. refueling outage is the most appropriate time to conduct this test because the components have been disassembled and reassembled, 2) the longer test interval is consistent with the requirements of ASME Section XI, and 3) leakage from the pressure boundary during shutdown cooling operation would be' detected.
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4.
Outboard MSIV Simulated Loss of Instrument Air Drift Test (ST-39N)
L The purpose of this surveillance test is to verify the integrity of the outboard MSIV instrument air supply check valves to maintain a sufficient supply of instrument air in the accumulators to maintain the outboard MSIVs open upon a loss of instrument air.
The test also verifies the integrity of the instrument air accumulators to maintain the outboard MSIVs open during the postulated loss of the
. instrument a r system.
Leaktightness of the accumulators i
is verified by isolating the instrument air supply and measuring the time interval from the start of depressurization until the MSIVs close under spring force'.
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Leakage is acceptable if the MSIVs remain open for 30 minutes or longer (Reference 12).
The fail-safe position of the valve is closed, but since the accumulator is being tested for leakage over a period of time, the most practical'way to test accumulator integrity is by having the accumulator keep the MSIV open.
The MSIVs are spring loaded, pneumatic, piston-operated valves designed to fail closed on loss of pneumatic pressure to the-valve operator.
Each valve operator is.
actuated by a dual solenoid pilot valve - one solenoid powered by AC, the other by DC.
An accumulator, located close to each isolation valve, provides pneumatic pressure for the purpose of assisting in valve closure, in the event-of failure of the pneumatic supply pressure to the valve operator system.
Also, each MSIV has 4 spring guide shafts i
each containing three stacks of emergency closing springs.
These springs are compressed when the valve is open and
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expand when the air pressure is either vented or lost from under the piston.
This exerts a downward force pushing the valve stem and disk to close the valve.
The springs alone are capable of shutting the valve in 3 to 5 seconds.
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24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM i
SURVEILLANCE TEST IMPROVEMENTS I
The first consideration with surveillance extension relates to industry recommendations.
GE SIL-477 recommends testing-each MSIV actuator and accumulator, which assists closure-of the MSIVs following loss of air, for leaktightness'each refueling outage (Reference 13).
James A.
FitzPatrick i
Maintenance Procedure 29.1, Section 7.5.32, performs-this recommended leak test of the accumulator assembly each refueling outage.
This ensures both theLintegrity of the accumulators and their ability to assist spring force in MSIV closure when-instrument air is lost (Reference 14).
The extension of MP-29.1 with the nominal 24 month 3
operating cycle is discussed in the maintenance evaluation section later in the report.
A further factor MSIV accumulator operability is the quality of instrument air and its importance in accumulator-and check valve integrity.
In NRC Generic Lotter 88-14, the poor quality of the instrument air was found to be directly'related to pneumatic component malfunctions (Reference 15).
The Authority's response to Generic Letter 88-14 included a review of general instrument air procedures and surveillances, as well as an analysis of the i
outboard MSIVs in a loss of air scenario.
Some changes as a result of this review included: Loperating procedures ensuring that the air parameters are monitored and recorded during normal operation along witt a testing-program that checks air quality by measuring dew point and particle size I
(Reference 16).
These practices ensure that air remains at i
a high quality and meets industry standards, thereby lowering the likelihood of failure for MSIV pneumatic components, such as the failure of the MSIV: accumulator check valve to isolate on low air' supply pressure.
Another consideration in surveillance extension is the burden of testing at power.
The fail-closed testing of the outboard MSIVs cannot be performed during power operation since access to the steam tunnel is required.- Also, the temporary modification of the instrument air system to i
i simulate the loss of air may cause unplanned plant transients (Reference 11).
A review of occurrence reports from 1986 to 1991 revealed no failures associated with'MSIV pneumatic components.
MSIV accumulators have been consistently leaktight and routinely meet the acceptance criteria.
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24 510NTII OPERATING CYCLE NUCLEAR STEAh! SUPPLY SYSTEh!
SURVEILLANCE TEST IN1 PROVEN 1ENTS The Outboard MSIV Simulated Loss of Instrument Air Drift Test can be safely extended for the following reasons: 1)
.the integrity of the accumulators is ensured each refueling outage through post-maintenance testing, 2)-the quality of instrument' air meets current industry' standards, and thereby lowers the likelihood of MSIV pneumatic component failure, 3) the MSIVs cannot be tested during power operation, or.during cold shutdown when the drywell is inerted, and 4) a review of past MSIV accumulator performance has shown them to be leaktight and the check valves functioning properly.
- 5. Reactor Level Indication Instrumentation (02-3LI-SUA)
Calibration (ISP-3-8)
The purpose of this procedure is to perform a calibration of the local Reactor Water Level Indicating Switch.
This wide range Barton indicator provides operators with reactor water level information during shutdown from outside the.
control room and is calibrated once every operating cycle (Reference 17).
This calibration can he extended with the longer fuel cycle if the potential incresse in instrument drift does not affect the capability to' achieve a safe plant shutdown.
Safe shutdown from outside the control room uses reactor water level control based on the Emergency Operating Procedures (EOPs).
Allowances are provided in the EOPs for instrument inaccuracies and environmental effects as necessary.
For example, density compensation is required due to the increase in instrument water leg temperature associated with post accident conditions in the drywell.
These allowances, which are based on post-accident conditions contain ample margin to accommodate any increased drift in the reactor water level indicator.
Based on discussions with the vendor, drift is expected to be minimal (less than 1 percent of span over the lifetime of the switch).
Due to this small drift, the vendor also stated that drift is included in the reference accuracy (Reference 18).
Therefore, extension of this calibration interval with the nominal 24 month operating cycle is allowable since the drift of the indicating switches is not affected by length of service.
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,e, 24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS Another consideration in extension are problems with calibrating the reactor level instrumentation during power i
operation.
On-line calibrations of reactor water level instruments is considered risky because valve manipulations and/or an induced hydraulic transient can result in a scram (References 19 & 20).
Therefore, it is safer to extend the calibration interval with the longer fuel cycle than to calibrate the instrument every 18 months during power operation.
The Reactor Level Indication Instrumentation Calibration can be safely extended for the following reasons: 1) the potential increase in drift associated with the longer calibration interval is accommodated by the large allowances for harsh environmental effects on instruments in the EOPs, 2) drift is minimal, and 3) calibration of the i
instrument at power could cause unplanned plant transients.
- 6. Recirculation Pump Temperature Instrument Calibration (ISP-78)
The purpose of this procedure is to demonstrate the operability of the Recirculation Pump Temperature Instrument Channels of the Recirculation System.
There are four temperature transmitters calibrated as part of the i
procedure: 02-TT-157A, B,
C, &D (Reference 21).
Past calibration data was analyzed to predict drift associated with the longer calibration interval of 30 months (24 months plus 25%).
Field data from five calibration dates were evaluated to establish drift values by comparing the "as-found" channel conditions to the previous calibrations "as-left" conditions.
The expected drift for the resistance temperature detectors (RTDs) over a year will not exceed 0.2 percent of the operating temperature range (Reference 22).
From this manufacturer data, the vendor allowable drift (VDA) for the instrument was calculated for 30 months as 1.096% of span.
This allowable drift was extrapolated using the " square root sum of the squares" methodology.
An analysis of past drift data was performed to find the best estimate drift (BED).
The BED, 0.696% of span, is an average of the past drift values, and is a good predictor of future drift for these instruments.
Since the best estimate drift is lower than the vendor allowable drift, the calibration of'the Recirculation Loop Suction Temperature can be safely extended with the nominal 24 month operating cycle.
The drift calculations are included in Attachment D.
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24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS -
- 7. Reactor Vessel Safety / Relief Valve Monitoring System (VMS) Functional Test (ISP-92-1)
The purpose of this procedure is to perform a functional check of the accelerometers that are located on the SRVs.
The procedure-is applicable.to the following eleven instruments: 02VMS-071A through L.
The surveillance consists of lightly tapping the SRV tailpipe close to the accelerometer.while listening to the audio output.
The resultant acoustic response is then observed and recorded at the Relay Room panel for both the primary and redundant accelerometer (Reference 23).
The first consideration in the SRV monitoring surveillance extension is on-line testing as a determination of system operability.
The Safety Relief Valve Monitor Instrument Check (ST-22I), performed on a monthly basis, demonstrates the operability of safety relief valve tailpipe acoustic and temperature monitoring systems by measuring both noise levels and temperatures from each SRV (Reference 5).
Another factor in test extension is that the surveillance cannot be completed during power operation.
The required tapping on SRV tailpipes to test accelerometer response j
requires access to the normally inerted drywell.
The final consideration for extension is the past performance of the. equipment in testing.
The previous four i
refueling outages test results were reviewed to evaluate system reliability.
Out of 88 accelerometer readings, both primary and redundant for 11 instruments in four tests, there were only 3 instances where the' accelerometer failed to register.
In none of these cases were both the primary and redundant accelerometers inoperable.
Therefore, l
operators would have been able to detect SRV operation.
The Reactor Vessel Safety / Relief Valve Monitoring System-Functional Test can be safely extended with the nominal 24 month operating cycle for the following reasons: 1) the monthly surveillance of the SRV accelerometers ensures i
their operability, 2) the test must be perforraed during shutdown conditions since access to the drywoll is required, and 3) a review of past performance for the acoustic monitors showed their reliability.
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- 24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS 8.
Reactor Vessel Safety / Relief Valve Thermocouple Functional Test (ISP-92-2)
The purpose of this procedure is to functionally. check the' thermocouples' located on the safety / relief valves.
This procedure is. applicable to the following eleven 1
instruments:.02-TE-112A & B and 02-TE-113A thru J.
The surveillance consists of recording the ambient temperature recorded by the thermocouple, then placing a bag of ice on the thermocouple.
The response of the thermocouple to the lower temperature is correspondingly recorded.
The ice is.
removed, and the thermocouple monitored to ensure its return to ambient conditions (Reference 24).
The first consideration in SRV thermocouple surveillance extension is on-line testing as a determination of system operability.
The Safety Relief Valve Monitor Instrument Check (ST-22I), performed on a monthly basis, demonstrates the operability of safety relief valve tailpipe acoustic and temperature monitoring systems by measuring both noise levels and temperatures from each SRV monitor (Reference 5).
Another consideration for test extensicn is that the surveillance cannot be completed during power operation.
The procedure requires drywell access, and is therefore not practical to test at power.
The. final consideration for extension is the past performance of the equipment in testing.
A review of the 1
operating occurrence reports from 1986 to 1991 revealed no failures with SRV thermocouples.
The Reactor Vessel Safety / Relief Valve Thermocouple Functional Test can be safely extended with the nominal 24 month operating cycle for the following reasons: 1).the Li monthly surveillance of the SRV thermocouples ensures their j
operability, 2) the test must be performed during shutdown-conditions since access to the drywell is required, and 3) a review of past performance for the SRV thermocouples showed their reliability.
i 16
s
'f.
24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IN1 PROVEN 1ENTS
- 2. Maintenance Activity Chances
- 1. Reactor Vessel Safety Relief Valve Maintenance (MP-2.4)
This procedure describes the maintenance and inspection of the Reactor Vessel Safety / Relief Valves.
This includes solenoid removal & replacement, pilot removal &
replacement, base removal & replacement, and the maintenance of the main stage assembly.
In addition,.the pilot assembly is bench tested, with refurbishment performed, if necessary, in accordance with the manufacturer recommendations.
Also, maintenance includes inspection of the pilot valve stem labyrinth seal area (Reference 25).
The major concern with SRV performance is their tendency to drift beyond the range of allowable lift settings.
A review of the operating occurrence reports from 1986 to 1991 revealed that SRV performance has been good other than drift.
This problem with SRV setpoints is a known issue and is currently being addressed by the BWROG.
Also, a Technical Specification change package was submitted to the NRC to increase the maximum permissible setpoint tolerance from one percent to three percent, consistent with ANSI /ASME OM-1-1981 (Peference 6).
With reference to the extension to a nominal 24 month operating cycle, the longer frequency between the surveillances is expected to have no effect on SRVs from a mechanical standpoint.
Concerning the issue of setpoint drift, an analysis of SRV actual lift settings with respect to length of service was performed to identify any drift trends.
The results show that there is no increasing drift trend with a longer length of service.
Therefore, SRV.
drift is not time dependent and will not be affected by the extension to a 24 month operating cycle.
The SRV drift trend, prepared by G. Ottman - JAF Technical Services, is shown in Attachment C.
The final factor in extension of this maintenance activity is the burden of performing the maintenance during power operation.
SRV maintenance must be completed when the reactor is in the cold shutdown condition with the drywell de-inerted and ventilated.
i I
i 17
24 510 NTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS The Reactor Vessel Safety Relief Valve Maintenance can be safely extended with'the nominal 24 month operating cycle for the following reasons: 1) a review of past operational occurrence reports found no problems with SRVs other than drift.from the' lift setting, 2) an analysis of SRV setpoint drift with respect to length of service shows no increasing L
drift trend in longer intervals, and 3) maintenance of the l
SRVs at times other than the refueling outage is impractical.
- 2. Main Steam Line Isolation Valve Maintenance (MP-29.1)
This procedure describes the routine maintenance performed on the Main Steam Isolation Valves.
This procedure is.
applicable to the following valves and their subcomponents:
29AOV-80A, B,
C,
& D and 29AOV-86A, B,
C,
& D.
This maintenance procedure includes the inspection, cleaning, and lubrication of the MSIVs, as well as post-maintenance testing which includes a springs-only fast close test (Reference 14).
l The quarterly performance of the MSIV Fast Closure Test (ST-1B) ensures the integrity'of MSIV mechanical 1
components.
MSIVs are closed and opened by their control switch.
The time interval between placing the MSIV control switch to the closed position and the red open indicating l
light illuminating is measured with a calibrated stopwatch.
All MSIVs should have a closing time of 3 to 5 seconds.
l Problems with this on-line surveillance would require an investigation into the cause, with maintenance performed if necessary (Reference 26).
This fast-close stroke test ensures the mechanical' operability of the MSIVs.
- However, the fast-close stroke test is unable to readily detect ~
spring fatigue or degradation since closure is accomplished.
with both spring and pneumatic force.
The issue of MSIV spring integrity is discussed in GE SIL No. 477 (Reference 13).
The letter recommends that a sprines-only full' stroke closing test be performed to confirm that stem packing friction does not prevent MSIV j
closure in the event that pneumatic force is lost.
This I
testing is to be performed following refueling outage MSIV u
18
24 510NTII OPERATING CYCLE NUCLEAR STEAh1 SUPPLY SYSTEh!
SURVEILLANCE TEST IN1 PROVEN 1ENTS leak rate testing, or whenever the packing chamber is adjusted.
The justification for extending the springs-only testing portion of this maintenance activity relies on the fact that problems with stem packing friction is a known issue which is closely maintained and monitored.
Procedures ensure valve packing adjustments do not increase stem friction force beyond the spring force following an outage, while the monthly fast-close test would detect excessive increases in stem friction.
A review of the MSIV work request history was performed to determine if MSIV operability problems were time-dependent, and also check the reliability of the MSIV springs.
No time-dependent or spring-related MSIV failures were found.
All MSIV actuators were rebuilt in the 1988 refueling outage.
The Main Steam Line Isolation Valve Maintenance can be safely extended for the following reasons: 1) the on-line fast stroke testing ensures MSIV operability and would identify equipment problems except for the integrity of the springs, 2) MSIV springs-only closing capability is ensured by the post-maintenance testing, while of on-line fast-stroke test would detect excessive increases in spring force; also, GE SIL No. 477 recognizes that springs-only testing should be performed following leakage testing and valve packing adjustments, and 3) a review of past MSIV work requests found no evidence of time dependent or spring related failuras.
- 3. Main Steam Line Isolation Valve Closure (RPS) Position Switch Calibration (MP-29.2) i l
The purpose of this procedure is to perform a calibration of the Main Steam Isolation Valve 10% Closure Limit Switches of the Reactor Protection System.
This is applicable to the following switches: 29PNS-80A, B,
C,
&D and 29PNS-86A, B,
C, &D (Reference 27).
The first consideration for extension of this maintenance procedure is the existence of on-line monitoring and testing to determine MSIV limit switch operability.
Since the switches are periodically monitored in the control room, a broken switch or any significant movement in switch position would be readily detected by operators.
Also, the Main Steam Isolation Valves Limit Sv ;ch Instrument Functional Test (ST-II) is performed monthly to ensure the operability of MSIV closure limit nwitches through the cycling of the MSIV to the 10 pert.ent closed position.
The 19
24 h!ONTH OPERATING CYCLE NUCLEAR STEAAI SUPPLY SYSTE51 SURVEILLANCE TEST 151 PROVE 5fENTS failure of this surveillance would alert operators to investigate the switches, and perform the required maintenance, if necessary (Reference 28).
Another consideration for maintenance extension with.the longer operating cycle is the fact that the calibration must be performed while the plant is in a shutdown condition.
Realigning of the limit switches during power operation could subject technicians to increased radiation exposure.
The Main Steam Line Isolation Valve Closure (RPS) Position Switch Calibration can be safely extended with the nominal 24 month fuel cycle for the following reasons: 1) periodic control room monitoring of MSIV limit switches and the monthly performance of the MSIV limit switch functional test ensure equipment operability, and 2) performance of the calibration at power could subject technicians to increased radiation exposure.
- 4. Limitorque Motor Operators - Model SMB/SB Preventive Maintenance Inspection and Functional Testing (MP-59.21) 5.
Limitorque Motor Operators - Model SMB and SB Preventive Maintenance Inspection,. Lubrication, and Testing on Baselined MOVs (MP-59.51)
These Limitorque Motor Operators preventive maintenance procedures are currently performed once every 18 months and include the following: external inspection, motor inspection, motor and gear house grease inspection, worm gear immersion inspection, "as-found" torque switch settings, and' post-work testing.
Two separate procedures are written since the MOVs at FitzPatrick are undergoing baseline diagnostic testing using the VOTES test system.
At the current time, all MOVs have not been baselined; therefore, separate maintenance procedures are necessary.
The MOV preventive maintenance procedure for baselined MOVs includes "as-found" limiter plate setting,."as-found" geared limit switch settings, and "as-left" geared limit switch settings (Reference 29 & 30).
1 5
20 1
% e 24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM '
SURVEILLANCE TEST IMPROVEMENTS
'The following NSSS MOVs are maintained every 18 months:
"A" and "B" Recirculation Pump Discharge Valves (02MOV-53A &
02MOV-53B), Inside Main Steam Drain. Isolation Valve-(29MOV-74), Outside Main Steam Drain Isolation Valve (29MOV-77).
The Reactor Head Vent to Main Steam Line Limitorque Valve--(29MOV-102).Was listed in the maintenance procedure, but has since been disabled and is currently in manual mode..
These procedures consist primarily of maintenance on.the
.Limitorque actuator, but also includes the separate issue Of valve stem cleaning.and lubrication.
.The justifications for each are explained below.
Concerning actuator maintenance, the Limitorque Type SMB Instraction and Maintenance Manual states that a schedule should be made to periodically inspect and lubricate all t
Limitorque equipment.
The manual recommends an initial 18 month maintenance frequency with a provision for altering i
the interval based on operating experience, frequency of i
operation, and operating conditions (Reference 31).
This was affirmed by Limitorque Maintenance Update 92-1 which
' restated that the maintenance' interval depends on inspection histories (Reference 32).
MOV actuator performance will be reviewed periodically to determine the t
adequacy of the preventive maintenance.
Motor Operated Valve surveillance and maintenance is currently being evaluated at FitzPatrick as a result of Generic Letter 89-10 (Reference'33).
The generic letter work is now in progress, and.will ultimately provide greater assurance of MOV operability.
The MOVs are located both inside and outside of the drywell.
The outboard valve (29MOV-77) is operated infrequently and is installed in a controlled environment (i.e. dust free and moderate temperature).
The inboard valves (02MOV-53A & 53B, and 29MOV-74) are also operated infrequently.
However, inboard valves are subject to moisture intrusion and the effects of higher temperatures on oils and lubricants.
An important consideration in the extension of this preventive maintenance activity involves past performance.
This is less of a concern for the four MOVs covered under this report since they have been'recently overhauled or replaced.
In the 1990 refueling outage, 29MOV-74 was replaced.
In the 1992 refueling outage, 02MOV-53A & B are.
scheduled for overhaul, while 29MOV-74 is scheduled for 21
-. o 24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS replacement.
Also, by the end of the 1992 refueling outage, all MOVs will be baselined using the VOTES diagnostic system.
An important consideration in MOV operability is the quality and reliability of the lubrication used in the valves and actuators.
To allow extension of this preventive maintenance, the lubricants utilized in the actuators must not significantly degrade over the nominal 24 month operating cycle.
The main gear case is lubricated with Exxon Nebula EP-0.
This lubricant is a calcium complex-gelled mineral oil plus additives that is effective over a wide range of load, speed, temperature, and moisture conditions.
The operating limit for the Nebula grease ranges from 200' to 300' F.
The geared limit switch is lubricated with either Exxon Beacon 325 or Mobil Mobilgrease 28.
The Beacon 325 is an ester-based, lithium soap-gelled product that is formulated for use over a wide temperature range.
Its operating limit is from 200* to 300* F.
The Mobilgrease 28 is a synthetic hydrocarbon based, clay-gelled product that is also designed for use over a wide temperature range.
Its operating limit is from 200' to 325' F.
The motor bearings are permanently lubricated at the factory for a 40 year life (Reference 34).
Since the temperature inside the drywell is required to be 140' F or less during plant operation (Reference 35),
and the auxiliary buildings are also kept at a controlled temperature, MOV lubricant degradation due to temperature effects during normal operation is not likely to occur.
The valve stem maintenance includes the cleaning of the valve stem threads below the operator and applying Nebula EP-0.
Extending the lubrication interval could increase the possibility of poor MOV performance from an increase in stem friction.
Justification for the extension of valve stem lubrication relies on the ability of MOV stroke testing to detect valve stem lubrication problems.
As stated previously, 2 MOVs are stroke tested quarterly, while two are stroked during cold shutdown.
Such testing, along with past experience showing an 18 month lubrication frequency to be acceptable, allows the lubrication interval to be safely extended with the 24 month operating cycle.
Also, EPRI-NMAC is currently researching the area of valve f
stem / stem nut lubrication.
Upon completion of the program, the valve stem inspection and lubrication may be modified to require a different lubricant, a different lubrication interval, or a combination of both (Reference 34).
i i
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(
24 510NTII OPERATING CYCLE NUCLEAR STEAh! SUPPLY SYSTEh!
SURVEILLANCE TEST Ih1PROVEh1ENTS The integrity of the MOVs that perform a containment isolation function, 29MOV-74 & 77, is confirmed by the containment isolation valve leakage test, which is performed every refueling outage.
This surveillance ensures that each MOV meets its individual leakage acceptance criteria (Reference 36).
Increased valve leakage may indicate common deficiencies such as seat leakage or valve stem packing leakage, but also the early l
stages of actuator degradation.
This would allow maintenance to laa performed on the af fected MOV before actuator operability concerns arise.
Also, the FitzPatrick Inservice Testing (IST) Program includes the quarterly stroke testing of 29MOV-74 & 77, and the cold shutdown stroke testing of 02MOV-53A & 53B (Reference 11).
A failure of the valves to meet surveillance acceptance criteria would be investigated and corrected, if necessary.
Another measure of MOV operability is the VOTES diagnostic system, which can detect the rate of MOV loading phenomenon l
during flow / differential pressure tests.
The system can measure changes in MOV performance, and also help adjust torque switch settings (Reference 37).
Baseline testing will be completed for all MOVs by the end of the 1992 refueling outage.
In the future, deviations from the baseline will be evaluated and corrective maintenance performed as necessary to ensure continued MOV operability.
The final consideration in the extension of MOV preventive maintenance are the conditions required for its completion.
Three of the MOVs are inside the drywell, therefore requiring access for maintenance.
Since the drywell is inerted during normal operation, it is not practical to i
perform this maintenance activity other than during a refueling outage.
The Limitorque Valve Operator Preventive Maintenance can be safely extended for the following reasons: 1) the Limitorque manual recommends that the initial preventive maintenance frequency should be every 18 months, but does not preclude adjustment of the maintenance interval based on acceptable operating frequency, conditions, and experience, 2) work performed in conjunction with Generic Letter 89-10 program will enhance reliable operation of the MOVs, 3) the MOVs evaluated are recently overhauled or I
replaced, 4) the lubricants used in the MOV actuators have
)
a good record of past performance with no evidence of cycle l
dependent degradation, 5) on-line and cold shutdown stroke 23
{-
- V 24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS I
testing, along with good past experience, ensure reliable valve stem / stem nut lubrication, 6) leak testing of the two MOVs that perform a containment isolation function could reveal early actuator degradation, 7) IST stroke testing, both quarterly-and cold shutdown, help to ensure the operability of all four MOVs, 8) baseline diagnostic testing (VOTES) is being performed on all MOVs and will increase confidence in valve operability, and 9) performance of the maintenance during operation is impractical since access to the to the normally inerted drywell is required for three of the MOVs.
6.
Lubrication of Electric Motors (Without Disassembly) with Grease Lubricated Bearings (MP-101.04)
This maintenance procedure consists-of the periodic l
lubrication of electric motors (without disassembly) with grease lubricated bearings.
It applies to electric motors that have grease lubricated ball or roller bearings.
The motors included in the maintenance activity that fall under the category of NSSS include the Recirculation MG Set Lube Oil Pumps (02-184P-2A1, 2A2, 2A3, 2B1, 2B2, & 2B3) l (Reference 38).
i The maintenance is scheduled to be performed each refueling outage, but is currently being completed more frequently for the Recirculation MG Set Lube Oil Pumps.
The Lubrication Evaluation Data Sheets that specifically apply to the greasing of motor bearings in the Recirculation MG Set Lube Oil Pump revealed that greasing is performed annually with 0.6 oz. of Mobilith AW-2 grease per bearing (Reference 39).
This falls well within vendor recommendations that the motor bearings be greased at 18 4
month intervals.
Therefore, extension of this maintenance j
activity with the nominal 24 month operating cycle is not necessary.
- 7. Replacement of ASCO Series NP8323 Solenoid Valves (MP-200.1) i The purpose of this maintenance procedure is to specify the i
replacement installation requirements (which serves as the procedure for the performance of environmental qualification maintenance and preventive maintenance) for the ASCO Series NP8323 solenoid valves installed on the 24 1
- +
24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST INIPROVEMENTS i
Main Steam Isolation Valve actuators.
This procedure is applicable to the-ASCO three-way solenoid valves with the following plant identification numbers: 29SOV-80A2 through D2 and 29SOV-86A2 through D2.
The' maintenance frequency of 18 months has been-determined by the EQ program Aging Data i
Base (Reference 40).
I The primary consideration in replacement interval extension with the 24 month operating cycle is the EQ life of the l
equipment. ~The ASCO Series NP8323 Solenoid valves have a qualified life of 2.6 years for the inboard valves (29SOV-80A2 through D2) and 1.7 years for the outboard valves (29SOV-86A2 through D2) due to the combined effects i
of a normal and accident environment (Reference 41).
Since the EQ life of the solenoid valves on the inboard MSIVs (29SOV-80A2 through D2) is greater than 30 months ~(2 years + 25%), their replacement can be safely extended with the longer operating cycle.
It is also recommended that the replacement of the solenoid valves'on the outboard.
MSIVs (29SOV-86A2 through D2) be either: 1) performed annually in conjunction with the mid-cycle and refueling outages since the EQ life is less than 30 months, or 2) the EQ life be' reevaluated for possible extension to-30 months.
V.
Summary and Conclusions-To support the 24 month fuel cycle, extension of the following once per refueling outage and once per operating cycle Nuclear Steam Supply System surveillance test intervals is proposed.
- 1. The Manual Safety Relief Valve Operation and Valve t
Monitoring System Functional Test (ST-22B) can be safely extended for the following reasons: 1) leaking or-partially open SRVs are readily detected during normal operation, 2) a review of past performance shows that SRVs are mechanically reliable with the exception of i
setpoint drift, and there has been'no evidence of tailpipe blockage.
Regarding setpoint drift, trending with respect to length of service revealed SRV drift not to be cycle length dependent.
t 25
m o.
24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS
- 2. The Reactor Vessel Operational Pressure Test (ST-39H).
1 can be safely extended for the following reasons: 1) i
.ASME Section.XI requirements do not preclude pressure test extension, 2) the refueling outage is the most' appropriate time to conduct this test because the RCS has been disassembled and reassembled, and 3) leakage from the reactor coolant pressure boundary would be readily detected.
- 3. The Shutdown Cooling Suction Operational Pressure Test (ST-39K) can be safely extended for the following reasons: 1) the refueling outage is the most appropriate i
time to conduct this test because the components have been disassembled and reassembled, 2) the longer test interval is consistent with the requirements of ASME Section XI, and 3) leakage from the pressure boundary
{
during shutdown cooling operation would be detected.
- 4. The outboard MSIV Simulated Loss Of Instrument Air Drift.
Test (ST-39N) can be safely extended for the following reasons: 1) the integrity of the accumulators'is ensured each refueling outage through post-maintenance testing,
- 2) the quality of instrument air meets current industry standards, and thereby lowers the likelihood of MSIV pneumatic component failure, 3) the MSIVs cannot be tested during power operation or cold. shutdown when the drywell is inerted, and 4) a review of past MSIV accumulator performance has shown them to be leaktight and the check valves functioning properly.
- 5. The Reactor Level Indication Instrumentation Calibration (ISP-3-8)'can be safely extended for the following reasons: 1) the potential increase in drift associated with the longer calibration interval is accommodated by the large allowances for harsh environmental effects on instruments in the EOPs, 2) drift is minimal, and 3) calibration of the instrument at power could cause unplanned plant transients.
- 6. The Recirculation Pump Temperature Instrument Calibration (ISP-78) can be safely extended with the nominal 24 month operating cycle since the best estimate l
drift is lower than the vendor allowable drift.
r 26
24 MONTH OPERATING CYCLE i
NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS.
i
- 7. The Reactor Vessel Safety / Relief Valve Monitoring System Functional Test-(ISP-92-1) can be safely extended with the nominal 24 month operating cycle for the following-reasons: 1) the monthly surveillance of the SRV
~
accelerometers ensures their operability, 2) the test.
must be performed during shutdown conditions since access to the drywell is required, and 3) a review of past performance for the acoustic monitors showed their reliability.
- 8. The Reactor Vessel Safety / Relief Valve Thermocouple Functional Test (ISP-92-2) can be safely extended.with the nominal 24 month operating cycle for the following reasons: 1) the monthly surveillance of the SRV thermocouples ensures their operability, 2) the test must be performed during shutdown conditions since access to the drywell is required, and 3) a review of past performance for the SRV thermocouples showed their reliability.
- 9. The Reactor Vessel Safety Relief Valve Maintenance (MP-2.4) can be safely extended with the nominal 24 month. operating cycle for the following reasons: 1) a review of past operational occurrence reports found no problems with SRVs other than drift from the lift setting, 2) an analysis of SRV setpoint drift with respect to length of service shows no increasing drift trend in longer intervals, and 3) maintenance of the SRVs at times other than the refueling outage is impractical.
- 10. The Main Steam Line Isolation Valve Maintenance
. l (MP-29.1) can be safely extended for the following i
reasons: 1) the on-line fast stroke testing of the MSIVs would identify equipment problems except for the integrity of the springs, and 2) GE SIL No. 477 recommends that MSIV testing and maintenance be performed every refueling, not precluding maintenance activity extension.
- 11. The Main Steam Line Isolation Valve closure (RPS)
Position Switch Calibration (MP-29.2):can be safely extended with the nominal 24 month fuel cycle for the following reasons: 1) periodic control room monitoring of MSIV limit switches and the monthly performance of the MSIV limit switch functional test ensure equipment operability, and 2) performance of the calibration at power could lead to unplanned plant transients.
27
~
m.-- j "e
24 MONTII OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS l
- 12. The Limitorque Valve Operator Preventive Maintenance (MP-59.21 & MP-59.51) can be safely extended for the following reasons: 1) the Limitorque manual recommends that the initial preventive maintenance frequency should be every 18 months, but does not preclude adjustment of the maintenance interval based on acceptable operating c
frequency, conditions, and experience, 2) work performed i
in conjunction with Generic Letter 89-10 program will enhance reliable operation of the MOVs,.3) the MOVs.
t evaluated are recently overhauled or replaced, 4) the lubricants used in the MOV actuators have a good record of past performance with no evidence of cycle dependent degradation, 5) on-line and cold shutdown stroke testing, along with good past experience, ensure reliable valve stem / stem nut lubrication, 6) leak testing of the two MOVs that perform a containment isolation function could reveal early actuator t
degradation, 7) IST stroke testing, both quarterly and I
cold shutdown, help to ensure the operability of all four MOVs, 8) baseline diagnostic testing (VOTES) is being performed on all MOVs and will increase. confidence in valve operability, and 9) performance of the maintenance during operation is impractical since access j
to the to the normally inerted drywell is required for three of the MOVs.
- 13. The Lubrication of-Electric Motors (Without Disassembly) with Grease Lubricated Bearings-(MP-101.04) is not' necessary to extend with the longer operating cycle
.i since the maintenance activity is currently being performed on an annual basis.
t i
- 14. The Replacement of ASCO Series NP8323 Solenoid Valves i
(MP-200.1) has two recommendations concerning extension due to different component locations.
Since the EQ life l
of the solenoid valves on the inboard MSIVs (29SOV-80A2 through D2) is greater than 30 months (2 years + 25%),
i their replacement can be safely extended with the longer operating cycle.
It is also recommended that the i
replacement of the solenoid valves on the outboard MSIVs.
(29SOV-86A2 through D2) be either: 1) performed annually in conjunction with the mid-cycle and refueling outages since the EQ life is less than 30 months, or 2) the EQ life be reevaluated for possible extension to 30 months.
i 28 i
)
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s 6
M A
i 24 510NTII OPERATING CYCLE NUCLEAR STEAh1 SUPPLY SYSTE51 SURVEILLANCE TEST 151 PROVE 51ENTS l
Technical Specification changes will be required by the extension of certain-tests.
The-proposed changes to the
-i Fitzpatrick Technical Specifications are given in~
Attachment F.
All other Technical Specifications surveillance requirements-evaluated for extension in this' report state that tests are to be completed during refueling.
These cases require no changes to the Technical Specifications.
VI.
References j
1.
Index of Operations Surveillance Test Procedures, January 23, 1992.
i i
2.
Operating Occurrence Report Logs from 1986 to 1991.
j 3.
Maintenance Department Preventive Maintenance Schedule,-September 24, 1990.
4.
Manual Safety Relief Valve Operation and Valve Monitoring System Functional Test (IST) (ST-22B),
Revision 19, October 17, 1990.
5.
Safety Relief Valve Monitor Instrument Check l
(ST-22I), Revision 9, December 12, 1991.
6.
James A.
FitzPatrick Nuclear Power Plant, i
" Proposed. Change to the Technical Specifications Regarding Updated SRV Performance Requirements and l
Miscellaneous Changes," JPN-89-084, December 20, 1989.
i 7.
Reactor Vessel Operational Pressure Test (ISI]
(ST-39H), Revision 15, June 6, 1990.
I 8.
ASME Boiler and Pressure Vessel Code,Section XI, Summer 1983 Addenda.
9.
James A. FitzPatrick Nuclear Power Plant, Central
{
Control Room Logs.
l I
10.
Shutdown Cooling Suction operational Pressure Test (ST-39K), Revision 3, June 6, 1990.
j 11.
James A. Fitzpatrick Nuclear Power Plant, Second Ten Year Interval Inservice Testing Program, May 5,
1991.
29
mm
- L 24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEh!
k SURVEILLANCE TEST INIPROVEN1ENTS 12.
Outboard MSIV Simulated Loss of Instrument Air p
Drift Test (ST-39N), Revision 3, May 2, 1990.
L.
13.
GE Service Information Letter (SIL) No. 477, " Main Steam Isolation Valve Closure," January 3, 1989.
14.
Main Steam Isolation Valve Maintenance (MP-29.1),
Revision 9, June 28, 1990, 15.
NRC Generic Letter No. 88-14, " Instrument Air Supply Problems Affecting Safety Related Equipment," August 22, 1988.
16.
NYPA Letter JPN-89-007, " Response to Generic Letter 88-14 Instrument Air Supply System Problems Safety Related Equipment," February 17, 1989.
17.
Reactor Level Indication Instrument Calibration (ISP-3-8), Revision 0, September 21, 1988.
18.
Memorandum DC-92-001, "Barton Flow Indicator Switch," March 27, 1992.
19.
Prompt Reportable Occurrence, LER 90-001 20.
Prompt Reportable Occurrence, LER 90-026 I
21.
Recirculation Pump Temperature Instrument Calibration (ISP-78), Revision 11, December 4, 1991.
j 22.
Rosemount Inc., " Series 78S, 88S Platinum I
Resistance Temperature Sensors," Product Data l
Sheet 2389, 1984.
j l
23.
Reactor Vessel Safety / Relief Valve Monitoring l
System (VMS) Functional Test (ISP-92-1), ' Revision 1,
September 10, 1986.
24.
Reactor Vessel Safety / Relief Valve Thermocouple Functional Test (ISP-92-2), Revision 1, June 3, 1987.
25.
Reactor Vessel Safety / Relief Valves Maintenance (MP-2.4), Revision 3, July 18, 1991.
30 l
a m*
.I
- 24 MONTH OPERATING CYCLE-NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST IMPROVEMENTS L
26.
MSIV. Fast' Closure (ST-1B), Revision 14, August 21, 1991.
27.
Main Steam Line Isolation Valve Closure (RPS)
Position Switch Calibration (MP-29.2), Revision 0, March 26, 1989.
28.
Main Steam Isolation Valves Limit Switch Instrument Functional Test (ST-II), Revision 15, i
June 12, 1991.
29.
Limitorque Motor Operators - Model SMB/SB Preventive Maintenance Inspection and Functional Testing (MP-59.21), Revision 10, November 20, 1991.
i 30.
Limitorque Motor Operators - Model SMB and SB Preventive Maintenance Inspection, Lubrication and Testing on Baselined MOVs (MP-59.51), Revision 1, December 12, 1991.
31.
Limitorque. Type SMB Instruction and Maintenance Manual: Bulletin SMBI-82C, 1982.
32.
Limitorque Maintenance Update 92-1, Section 6, "Limitorque 18 Month Lube Inspection."
33.
NRC Generic Letter 89-10, " Safety Related j
Motor-Operated Valve Testing and Surveillance,"
j June 28, 1989.
34.
Electric Power Research Institute (EPRI), Nuclear Maintenance Applications center, " Lubrication Guide," Revision 1, July, 1991.
j 35.
Operating Procedure F-OP-53, "Drywell Ventilation and Cooling," Revision 5, May 19, 1988.
36.
Type B and C LLRT of Containment Penetrations (IST) (ST-39B), Revision.25, January 7, 1992.
37.
Data Acquisition of Limitorque Valves'Using VOTES (MP-59. 3 6), Revision 3, January 8, 1992.
38.
Lubrication of Electric Motors (w/o disassembly) with Grease Lubricated Bearings (MP 101.04),
4 Revision 14, May 8, 1991.
31
4 W-4 24 MONTH OPERATING CYCLE NUCLEAR STEAM SUPPLY SYSTEM i
' SURVEILLANCE TEST IMPROVEMENTS l
39.
Lubrication Evaluation Data-Sheet for Recirculation MG Set Lube Oil Pump - Motor, October 11, 1985..
40.
' Replacement of ASCO Series NP-8323 Solenoid Valves (MP-200.1), Revision 1, May 30, 1990.
41.
Environmental Qualification File No. 3503,. System Component Evaluation Worksheet regarding the following components: 29SOV-80A2 through D2 &
2950V-86A2 through D2.
42.
Memorandum EP-RF--95-060, " Extension of the Surveillance Interval for JA7 Safety Relief Valves 3
in Support of the 24 Month Operating Cycle", dated March 7, 1995.
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- Nuclear Steam ' Supply System
- Surveillance Test Extensions'-
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(-i:
FIT 2 PATRICK - 24 MONTH OPERATING CYCLE i
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-i ATTACHMENT A i
SAFETY EVALUATION I
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24 MONTH OPERATING CYCLE l
NUCLEAR STEAM SUPPLY SYSTEM SURVEILLANCE TEST EXTENSIONS l
I.
Safety Evaluation i
ne proposed changes have been reviewed in accordance with the requirements of 10 CFR 50.59 and 10 CFR 50.92. These changes, which extend the test intervals do not involve an unreviewed safety question nor do they constitute a Significant Hazards Consideration.
1.
He probability of occurrence and the consequences of an accident or malfunction of safety-related equipment previously evaluated in the safety analysis report will not be increased.
Changes are proposed to increase the surveillance test interval (STI) with the nominal 24 month fuel cycle for the following surveillances: Manual Safety Relief Valve Operation and Valve Monitoring System Functional Test - IST (ST-22B), Reactor Vessel Operational Pressure Test - ISI (ST-39H), Shutdown Cooling Suction Operational Pressure Test - ISI (ST-39K), Outboard MSIV Simulated los~s of Instrument Air Drift Test (ST-39N), Reactor 12 vel Indication Instrumentation Calibration (ISP-3-8), Recirculation Pump Temperature Instrument Cahbration (ISP-78), Reactor Vessel Safety / Relief Valve Monitoring System (VMS) Functional Test (ISP-92-1), and Reactor Vessel Safety / Relief Valve Thermocouple Functional Test (ISP-92-2). These changes extend the STIs. They do not involve any hardware j
modifications. Here is no increase in (1) the probability of an accident occurring, (2)
+
the consequences of an accident, and (3) the consequences of equipment malfunction.
However, increasing the STIs may affect the probability of equipment malfunction.
Regarding the probability of equipment malfunctions:
j The Manual Safety Relief Valve Operation and Valve Monitoring System e
Functional Test (ST-22B) can be safely extended with the longer operating cycle.
Leaking or partially open SRVs are readily detected during normal operation, and a review of past performance shows that SRVs are mechanically reliable with the exception of setpoint drift. Regarding setpoint drift, a safety evaluation was prepared by GE that defined an upper limit for SRV opening pressure well above current allowable (1%), proposed allowable (3%), and expected worst case (considering the increased emphasis on SRV operability) setpoint drift.
FitzPatrick recognizes the complexity of the SRV drift issue and is working to correct operability problems by participating in the BWROG SRV setpoint drift program and submitting SRV Technical Specifications changes. Also, all SRVs will be removed for bench testing over the operating cycle until the setpoint drift issue is resolved which exceeds Technical Specificatiens and ANSI /ASME OM-1-1981 s
requirements.
ne Reactor Vessel Operational Pressure Test (ST-39H) and the Shutdown e
Cooling Suction Operational Pressure Test (ST-39K) can be safely extended with the longer operating cycle. leakage from the pressure boundaries during normal
-l or shutdown cooling operation would be detected. Also, ASME Section XI requirements do not preclude pressure test extensions, and the refueling outage is 1
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/24 MONTH OPERATING CYCLE l
LM
. NUCLEAR STEAM SUPPLY SYSTEM -
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[h SURVEILLANCE TEST EXTENSIONS N
the most appropriate time to conduct the testing since the systems have been 4
' disassembled and reassembled.
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be safely extended with the longer operating cycle. He surveillance has no effect on the fail-safe position (closed).of the MSIVs; keeping the MSIV open during the.
loss of instrument air is the test condition. The quality of instrument air meets.
current industry standards which lowers the likelihood of MSIV pneumatic
'l component failure. Also, the MSIVs cannot be tested during power operation since the air supply originates from a common header which would require the impractical isolation of all four isolation valves for testing..
~
j ne Reactor level Indication Instrumentation Calibration (ISP-3-8) can be safely '
1i e
extended with the longer operating cycle. The drift is minimal, and the potential increase in drift associated with the longer calibration interval is accommodated by
_ the large allowances for harsh environmental effects on instruments in the EOPs.-
. Also, calibration of the instrument at power could cause unplanned plant transients.
ne' Recirculation Pump Temperature Instrument Calibration (ISP-78) can be l
e safely extended with the longer operating cycle since the projected drift (BED 30) is y
lower than the expected drift based on vendor data (VDA30),
q ne Reactor Vessel Safety / Relief Valve Monitoring System Functional Test (ISP-'
e 92-1) and the Reactor Vessel Safety / Relief Valve Hermocouple Functional Test 1
(ISP-92-2) can be safely extended with the longer operating cycle. De monthly surveillance of the SRV acoustic monitors and thermocouples, along with their reliable past performance ensures system operability. Also, the tests must be performed during shutdown conditions since access to the drywell is required.
i 2.
De possibility of an accident or malfunction of a different type than evaluated l
previously in the safety analysis report is not created. -
1 r'
De proposed changes extend STIs. He proposed changes do not change the manner in which the Nuclear Steam Supply System functions. ' An evaluation of past
}
' equipment performance and a study of on-line testing show the longer STIs will ra degrade Nuclear Steam Suply equipment. Therefore, the proposed changes do not l
create any new failure moces or a new accident.'
1
-1 3.
The margin of safety as defined in the basis for any technical p:'ution is not H
reduced.
De proposed changes do not reduce the margin of safety as defined in the basis for -
i any Technical Specifications. The proposed changes extend STIs. Evaluation of the l
past performance of the equipment indicates that the effects of extending the STIs would not involve a significant reduction in a margin of safety.
1 2
% o Nuclear Steam Supply System Surveillance Test Extensions FITZPATRICK - 24 MONTH OPERATING CYCLE l
I ATTACHMENT B NSSS OPERATIONAL OCCURRENCE REPORTS
.24M OPERATING CYCLE
- NUCLEAR STEAM SUPPLY SYSTEM,.
OPERATIONAL OCCURRENCES (1986 - 1991) q
- 86-015 MSIV 29AOV 80B Limit Switch did not reset.- 86-035 "A" Feedwater Return Valve Gasket Leak (Bonnet). Valve since replaced. -
l 86-043 MSIV 90% Limit Switch Out of Calibration.86-081 "B" Recirculation Discharge Bypass Valve motor not operating valve. Valve since removed.86-086 MSIV 29AOV 86A 90% Limit Switch did not reset.86-094 Outboard Main Steam Line Drain - Relay Tripped (sprayed by water).86-193 Reactor Recirculation Pump Suction Temperature Setpoint Drift.87-010 MSIV 29AOV-86A failed leak rate test.87-019 SRV Setpoint Drift.87-034 Reactor Recirculation Pump Motor "A" Tripped during performarx:e of RCIC Auto Actuation Test.
j 87-068 MSIV 29AOV 86C Fant Close Time Out of Specification. (2.5 sec. compared to i
Tech Spec. valve 3 - 5 sec.).88-061 SRV 02RV-71F, L & H found to lift at > 1% above setting.
l i
88-148 RWR Temperature Detectors 10TE. HSA & B found not in service.88-149 SRVs 02-RV-71D & J found set > 1% above nominal.-
88-166 MSIV 29AOV-80A 90% Limit Switch not resettmg.
1 i
89-013 SRVs 02-RV-71C, K & B left > 1% above setting.89-033 SRV 02RV-71K inadvertently opened by operator during surveillance.89-054 MSIV 29AOV-80A 90% Limit Switch failed to reset during monthly surveillance ST-11.
89-095-RWR Loop Temp Recorder 02TR-165 found not calibrated.89-134 RWR Inboard Sample Valve 02 2SOV 39 failed closed.89-165 Flow Unit 02FT 110F reading 20% when actual flow at 42%.89-179 MSIV 29AOV-86C would not open after fast close test.,
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-24M OPERATING CYCLE l
NUCLEAR STEAM SUPPLY SYSTEM i
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OPERATIONAL OCCURRENCES (1986 - 1991) l n
89-18f Feedwater Flow Transmitters out of calibration.
1 89-215' Reactor Scram during SRVTesting.89-240 Feedwater Flow Instuments A & B (anomalous test results) no compensation for static pressurization.89-253 SRVs 02RV 71E & F found > 1% lift setting limit.
1
+
90-073 MSIV 29AOV-86A reopened with switch in closed position.
i 90-080 MSIV Position Switch 29PNS-80C2 (90%) may have exceeded EO life.
s-90-142
. MSIV 29AOV-80D would not close fully using slow close switch.
90 174 SRV 02RV 71K pilot assembly as found setting below nameplate during Wyle labL f
test.90-176 MSIVs 29AOV-80B & 86B inoperable due to missed surveillance test for springs-j only closing time.
l l
90 194 "A" Recirculation loop rapid temperature increase of 160' to 320' in 3 seconds.
l 90 228 SRV 02RV-71J pilot assembly setpoint found with 2.3% deviation.
l 90 239 Limit Switch 29PNS-80A2 did not reset when MSIV full open.
l r
90-245 RWR Pump 02 2P-1B run back to minimum speed while at 100% power.90-294 -
RWR Outboard Sample Valve 02 2AOV-40 did not indicate closed during surveillance ST-1C.
l 90-299 RWR Inboard Sample Valve 02 2SOV-39 leaking past seat approximately 1.2 liters / min with valve closed (02 2SOV-40 inoperable also).
4 i
90-332 Startup Flow Control Valve wide open.91-037.
Recirculation MG Pump trip on 1/26/91 and 2/4/91.
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-Nuclear Steam Supply System' Surveillance Test Extensions l
.i FITZPATRICK - 24 MONTH OPERATING CYCLE i
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5 ATTACHMENT C t
SAFETY / RELIEF VAINE DRIFT ANALYSIS t
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w SRV SETPOINT DATA - 1985 TO 1992-PREPARED BY: GERRY OTTMAN 5/16/92 14 l
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' Nuclear Steam Supply System Surveillance Test Extensions l
p; FITZPATRICK - 24 MONTH OPERATING CYCLE I
l-ATTACHMENT D DRIFF WORKSHEET FOR ISP-78
iISP-78 DATA FOR AVERAGE (30 MONTHS) = 0.696 - % SPAN (18 MONTHS)-
- 0.5'06 4 SPAN
- RO2EMOUNT 414H3BF1 STD DEV (30 MONTHS) = 0.537 4 SPAN (18 MONTHS)~
= 0.444 % SPAN 0.95 % SPAN.
i RECIRC IDOP Al PUMP INLET TEMP AVG +STD (30 MONTHS) = 1.233 % SPAN (18 MONTHS)
=
AVG +3STD(30 MONTHS) = 2.307 % SPAN (18 MONTHS)
= 1.838 % SPAN l
BED (30 MONTHS) = 0.696 4 SPAN (18 MONTHS)
= 0.506 4 SPAN VDA (30 MONTHS) = 1.096 % SPAN (18' MONTHS)
= 1.077 % SPAN.
YES BED 18 < VDA18 =
YES BED 30 < VDA30
=
02-TT-157A 30 MONTHS 30 MONTHS 18 MONTHS DATE DRIFT (mV)
DAYS DRIFT (mV)
% SPAN
% SPAN
'29-SEP-1986 1.94 558 2.28 1.518 1.2939
,09-APR-1988 0.24 530 0.29 0.195 0.1601 SEP-1989 0.28 253 0.51 0.340 0.1868 01-JUN-1990 0.23 292 0.4 0.266 0.1534 20-MAR-1991
____ _______ TT-157B
'29-SEP-1986 1.16 563 1.35 0.902 0.7737 14-APR-1988 0.29 525 0.36 0.238 0.1934 21-SEP-1989 0.22 253 0.4 0.267 0.1467 01-JUN-1990 0.79 292 1.37 0.914 0.5269 20-MAR-1991 02-TT-157C 29-SEP-1986 1.92 563 2.24 1.492
-1.2806 14-APR-1988 0.23 525 0.28 0.189 0.1534 21-SEP-1989 0.48 253-0.87 0.583 0.3201 01-JUN-1990 1.24 292 2.15 1.434 0.8271 20-MAR-1991 02-TT-157D 29-SEP-1986 1.87 563 2.18 1.454 1.2472 14-APR-1988 0.31 525
-0.38 0.254 0.2068 21-SEP-1989 0.11 253 0.2 0.133 0.0734 01-JUN-1990 0.83 292 1.44 0.960 0.5536 20-MAR-1991 NOTE:
- 1) NO DATA POINTS DELETED USING THE OUTLIER TEST i
- 2) CAL-TOL = CALIBRATION TOLERANCE EXPRESSED AS % SPAN
..-.m.
-.. - - ~. - -
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Nuclear Steam Supply System '
Smveillance Test Extensions f
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- FI12 PATRICK - 24 MON 114 OPERATING CYCLE i
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A1TACHMENT E i
MOV WORK REQUEST EVALUATION l
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MOV WORK REQUEST EVALUATION 29MOV 74
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Work Request #063588 Datc: 4/1/90 i
Fiche:N/A Deficiency: Valve tagincorrect.
Corrective Action: Fabricate and install new tag. VCS-90 was etched out and VGW-90 etched on tag.
Work Request #007210 Issue Date: 10/17/88 Fiche:5921:110-130 Deficiency: Routine PM on Limitorque motor operator per MP-59.21.
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Conective Action: Replaced reducers and fittings.
Work Request #051110 Issue Date:3/9/90 Fiche:10890:44 Deficiency: Connected MOV to power supply in training for procedure checkout 59.36 DRAIT.
Corrective Action: Connected replacement MOV 74 to power for procedure 59.36.
Work Request #056946 Issue Date: 9/20/89 Fiche:5380:6-94 Deficiency: While performing MOVN13 testing, went to stroke valve manually and declutch lever, would not stay in manual. Valve was electrically operable.
Corrective Action: Adjusted lever IAW MP 59.22. PWT SAT.
Work Request #060929 Issue Date: 11/12/88 Fiche:5316:89-115 Deficiency: Valve actuator driving valve into backseat before motor trips.
Corrective Action: Adjust open limit switch to 10% Cycled, stroked, timed SAT.
Issue Date: 11/20/88 Fiche: 5288:202-211 Deficiency: Position indicator showed valve in dual position while indicator at switch indicated open.
Corrective Action: Reset rotor to open contacts.
l Work Request #062871 Issue Date: 11/11/88 Fiche: 5342:140-166 Deficiency: Replace melamine torque switch with new fibrite torque switch.
Corrective Action: Installed switch to meet EQ requirements via 10CFR21.
Issue Date: 10/18/88 Fiche:5316:139-147 Deficiency: While PM on operator, found flex for motor leads was broken where it enters motor compartment.
(:
'M
' r Corrective Action: Disconnected and reconnected motor leads. Replace flex. PWT SAT.
Work Request #071512 Issue Date: 4/20/90 i
Fiche: 10932:59-68 Deficieocy. N/A Corrective Action: Installed VOTES strain gage per MP-59.42.
. Work Request #073424 Issue Date: 4/4/90 Fiche:N/A Deficiency: N/A Corrective Action: Implemented installation procedure for MOD F1-87-128, Replacement of MOV-74.
Work Request #072976 Issue Date: 4/2/90 Fiche:10905:41 Deficiency: MOV 74 has packing leak, causing MOV-74 and MOV 77 to fail LLRT.
Corrective Action: Valve replaced per F1-87128.
29MOV 77 Work Request #087487 Issue Date: 10/2/91 Fiche: N/A Deficiency: Inspect valve for potential problems affecting valve's ability to open as discussed in INPO OE.
475 and JAF OE REVIEW 910472. Complete by 10/31/91.
J Corrective Action: N/A Work Request #063286 Issue Date: 10/31/88 i
Fiche: 5319:4-59 Deficiency: Body to bonnet leak found during hydro testing.
Corrective Adion:l'ixed. PWT SAT.
' i I
Issue Date: 5/2/90 Fiche:5447:41-44 Deficiency: During PWT for PM, seal-in function of open control circuit failed to operate.
Corrective Action: Inspeded and retested, problem could not be found - self corrected.
j Work Request # 082273 issue Date: 3/12/91 Fiche:1707:4820-4823 DeSciency: Provide insulation removal and reinstallation support for walkdown inspection. -
J Corrective Action: lasulation removed, walkdown, replaced originalinsulation.
02MOV-53A Work Request #087609 issue Date: 12/19/91 Fiche:N/A
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+ Deficiency: Performed as found baseline on MOV.,
Corrective Action: N/A Work Request #007154 Issue Date: 10/7/88 Fiche:5315:216-231 Deficiency: Routine PM per 59.21 Corrective Action: SAT.
' Work Request #060784 Issue Date:8/24/88.
Fiche:10671:161-175 Deficiency: Check backseated valve for damage.
Corrective Action: Closed out per TTS-88-0873.
l Work Request #063850 Issue Date: 10/13/88 Fiche: 5229.223 229 "
~j Deficiency:N/A Corrective Action: Open limit switch cover to allow EO and NRC inspection.
Work Request #064600 Issue Date:10/30/88 Fiche:5320:190-204 Deficiency: Fureasive sealleakoff.
Corrective Adion: Torqued packing per MP 59.9. PWT SAT.
Work Request #070398 Issue Date: N/A Fiche:5442:16-27 Deficiency: Valve backseated on 1/12/90 for leakage. Need to repack valve.
Corrective Action: Repacked per 59.20.
t
. Work Request #071515 Issue Date:4/9/90 Fiche:N/A Deficiency:N/A s
. Corrective Action: Replace packing put in Jan 90 forced outage with graphite packing.
- Work Request #071536 Issue Date:8/8/90
. Fiche: 1703:2128-2131 Deficiency: Triple packing not sealing well, put in 5 ring graphite live load packing. -
Corredive Action: Completed.
t Work Request #071606 Issue Date:1/12/90 Fiche:5428:5542 Deficiency: Eledrically backseat valve due to possible leakage.
. Corrective Action: Valve backseated per MP-59.20. See WR#71665.
Work Request #071665 Issue Date:6/14/90
- t
- Fiche: 10975:174-175 i
Deficiency: Do engineering evaluation of valve.
Corrective Action: Memo JMD 90-109,500 cycle fatigue failure.
Work Request #071667 Issue Date: 1/20/90 Fiche: 5428:169 224 Deficiency: Repack valve since electrically backseated.
Corrective Action: Current trace per MP-59.21. MOVATS current trace Cl-011.4 amps.
Work Request #071703 Issue Date: 1/19/90 Fiche: 5428:97-102 Deficiency: Electrically backseat valve.
Corrective Action: See WR#071665.
Work Request #083211 Issue Date: 10/7/91 Fiche: N/A Deficiency: Nameplate in error. Replace per F1-87-061-114.
Corrective Action: Done.
02MOV-53B Work Request #088146 Issue Datc:11/22/91 Fiche: N/A Deficiency: Installlive load packing on valve. Reduce potential to backseat.
Corrective Action: Done.
Work Request #090840 Issue Datc: 12/19/91 Fiche: N/A Deficiency: Perform as found baseline on MOV Corrective Action: N/A Work Request #007153 Issue Date: N/A Fiche: 5292:173-191 Deficiency: Perform routine PM-59.21.
Work Request #060785 Issue Date:8/23/88 Fiche: 10671:176-190 Deficiency: Check backscated valve for damage.
Corrective Action: Memo II3-88-0873.
Work Request #063850 Issue Date: 10/13/88 Fiche: 5279:223-229 Deficiency: Open 10% limit switch cover for EQ and NRC inspection.
Corrective Action: Done.
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Work Request #' 64595 -
0 Issue Date: 10/30/88 W
Fiche:5321:4 '
- Deficiency: Weeping drip from packing.
Corrective Action: Torqued packing per MP-59.5. PWT SAT.
.. Issue Date: N/A Fiche: 1703:2128-2131 -
Deficiency: Triple packing no good. Change to 5 ring graphite packing.
Corrective Action: Ref: D1-90-055.
. Work Request #073789. ~
Issue Date:9/5/90 Fiche: 1703:3415-3439 Deficiency: Troubleshoot problem with runback signal.
Corrective Action: Fixed. ' ~
' Work Request #083211 Issue Date: 10/7/91 Fiche:N/A Deficiency: Nameplate in error. Replace per F1-87-061-114.
Corrective Action: Done.
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Nuclear Steam Supply System 4
. Surveillance Test Extensions FIT 2 PATRICK - 24 MONTH OPERATING CYCLE i
i i
ATTACHMENT F-l TECHNICAL SPECIFICATIONS CHANGES
J Al'N t'I* -
T.ible 4.1 2 ( con t.
- d )
RE ACT98_ PR97ECT I DH..EYETKti_iSCF AM ). I NSTR. MENT..C At. I D N AT I OM t
IIIMI)Sai_CALI BB AT EDM f.F EQUENC I EE_IDR_REACTOP _ PRQIECTIDH _ I NSTRQMCNT. CHAMU ELS in%rument Chamael GI9pP__(1).
CallhteLion..fil_.__.. _
__.tilaisuam frequency (2)
Turbine Control valve-Fest A
Standard Pressure Source Once/ operating cycle Closure Oil Pressure Trip Turbine Stop Valve Closure A
Note (5)
Note (5) l It9TFS EQR_ TABLE b.l.-2 1.
A description of three groups is included in the Bases of this Specification.
2.
Calibration test is not required on the part of the system that is not required to be operable.
i s t r i.g.ml, b at us t
is required prior to retura to service.
3.
The current source provides an instrument channel allgement. Calibration using a radiation source shall f.e m.i to each refueling outage.
Response time is not a part of the routine instrument channel test but will be checked once per operatineg cycle.
4.
am 2 4 ~..dL s.
Actuation of these switches by normal means will be performed f-ri,y_ th:
- f :!!:
- tr, c.
ObMA 5.
Calibration shall be performed utilising a water columna or similar device to provide assurance that dasaa.ge 6.
to a float or other portions of the float assembly will be detected.
7.
Sensor calibration once per operating cycle. Master / slave trip unit calibration once per 6 months.
1 AmendmentNo.f,
, 136 47
..-NPP l
TABLE 4.2-8 (cont'd)
MINIMUM TEST AND Call 8 RATION FREQUENCY FOR ACCIDENT MONITORING INSTRUMENTATION ins:rument instrument instrument instrument FunctionalTest Calibration Fraam Check 15.
Core Spray Flow N/A Once/ Operating Cycle Once/ day 16.
Core Spray Discharge Pressure N/A Once/ Operating Cycle Once/ day 17.
LPCI (RHR) Flow N/A Once/ Operating Cycle Once/ day 18.
RHR Service Water Row N/A Onos/ Operating Cycle Once/ day 19.
Safety /RelHd Valve Position Indicator C c/(^~;MJe 0,*
N/A Once/ month (Primary and Secondary) yge 4g_g /
Sec. z.4 w n f 20.
Torus Water Level (narrow range)
Isr-4z -r.
N/A Once/ Operating Cycle Once/ day 21.
Drywell-Torus Differential Pressure N/A Once/ Operating Cycle Once/ day t
Amendment No.136, I 9(
h
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/
JAFMPP 3.5 (cont'd) 4.5 (cont'd) 9.
Automatic Depressurisatloa SystnR_ihDS)
D.
Automatig_Dg2[232M[i34tign Jyaten_.iADS1 The/ DS shall be operable whenever the reac-1.
Surveillance of the Automatic Depressurisa-A 1.
tor pressure is greater than 100 psig. and tion System shall be performed ex i;; ; _.: evee7 1/fge.f4f irradiated fuel la la the reactor vessel and
- t' ; ;,;;; as follows
prior to reacter startup from a cold condi-tion, escept as specified belows a.
A simulated automatic initiation which c)24 opens all pilot valves.
N ST-ZZS m.
From and after the date that one of the l
sevem safety / relief valves of the ADS is b.
Manually open each safety / relief-valve made or feued to be inoperable for any while bypassing steam to the condenser reason while it is required, continued and observe a 110% closure of the turbine reactor operation is permissible only bypass valves. to verify that the safety /
during the succeedix* 30 days unless relief valve has opened.
repairs are made and provided that during such time the EPCI System is operable.
JJ A c.
A simulated automatic initiation which is inhibited by the override switches. % ST zzA b.
From the time that more than one of the sov.a saf.t,/,e n.f.aives of ti,e ads are made or fomed to be looperable for any g
g reason, contiamed reactor operation is
/
permissible durlag the succeeding 24 h'rs.
unless repairs are made and provided, that i
Amendment No.
,a
, 134 119
4 l
m JAFNPP 3.6 (cont'd) 4 0 IC'"*'d3 I
D.
Safety and Safety / Relief Valves E.
Safety and Safety / Relief Valves
- 1. During reactor penser operating
- 1. At least one half of att
.o.ditie
..e,rior to.tarte,
.afetyereuer valoe.au w onc,,
bench checked g
from a cold condittee, or whenever checked o r reactor coolant pressure le greater valves - :_ __-
.;!. The than a*
- ' re and temperature safety / relief valve settinge (4P-2.4 greater then 212*F, shall be set as required in Specification.
the safety made of all 2.2.a.
All valvee shall be tested every safety /retter valves ebe11 be
- _;
- ;:::ter; ;,;!::.
operable, emcept as ayecified by (g a g entId.
specificatten 3.6.E.2 The Amtsumatie Depreeeurination System Valves shall be operable as required by eyecification 3.5.D.
Amen h nt No.
,7 A1d. 134 142a m
. 1.-- - -
i JAFNPP 3.6 (cont'd) 4.6 (cont'd) 1
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2.
At least one safety / relief valve shall be disassembled and 2.
inspected c.;;/g;..;;,de.
a.
From and after the date that the safety valve function CPery a4 go.fc,
$1P-2 4 of one safety / relief valve is rnade or found to be inoperable, conhnued operation is permissable only during the usarcaarhng 30 days unless such valve is made operable sooner.
b.
From and alter the time that the safety valve function on two safety / relief valves is made or found to be snoperable, continued reactor operation is i
permssible only dunng the succooomg 7 days unless such valves are sooner made operable.
l 3.
If c:pardenhan 3.6.E.1 and 3.6.E.2 are not met, the reactor 3.
The integrity of the nitrogen system and coi6pcr.c-nts shsR be placed in a cold condton within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
which provide manual and ADS. actuation of the 4.
Low power physics teshng and reactor operator trammg safety re W alves shall be demonstrated at W once shaN be. peretted with enoperable components as every specalled in item B.2 above, provided that reactor W 4.
An annual report of safety /refief vave failures and temperature is <212"F and the reactor vessel is vented or ct
- -y will be sent to the NRC in accordance with the reactor vesselheadis removed.
Sechon 6.9.A.2.b.
i i
I t
Amendment No. yI,76,130, t34 143
1-123 Mw street '
ly, WM3 Plams. New York 10601
,s r m 914 6816200 >
W
~
A niewYorkPbwer-Memorandum
& Authonty August 16, 1994 NED-AP-94-379 TO:
T. Noskalyk FE00ts A. Petrenko
)
SUBJECTS JAF SAFETY RELIEF YALVES SETPOINT CEANGE General Electric Report GE-NE-187-50-1191,. prepared for JAFNPP l
Power Uprate and ' issued on ' Nov. - 1991, assumed that the setpoint
(
change for the SRV single setpoint of 1110 psig was implemented.
l This change was recommended by GE in the " Updated SRV Performance i
Requirements" report NEDC-31697P, prepared for JAFNPP in April, 1989. However, this recommendation to change SRV setpoints from 1090, 1105 & 1140 psig to a common setpoint of 1110 psig, was not implemented.
Following is the NED/IEC review of documents and concurrence of the original setpoint change. Attached is the MCM 8 Setpoint' Change l
Request, originated by System Engineering (SE) on. 6/29/94. Note, that'although "there is a change to the facility as described in the FSAR", no Nuclear Safety Evaluation is needed for this MCM, for l
the following reasons.
A detailed Safety Evaluation was included in the submittal of the Technical specification SRV specification change to the NRC. The conclusion of the last safety Evaluation revision, submitted via-JPN-94-013 on March 2, 1994, includes all pertinent statements to insure that "the changes, as proposed, do not-constitute an unreviewed safety question as defined in 10 CFR50.59".
I was informed on 8/12/94 by SE that " site will go ahead with change to SRV pilots during upcoming outage, using 1110 psig setpoints". It is assumed that tolerances, furth9r discussed below, will be applicable for both, the-current and uprate power-set-points. Should the power uprate take place in 1996, the submitted MCM 8 forms will be revised for a setpoint of 1145 psig i 3%.
There is no reason to repeat all the conclusions, recommendations' and changes since the initial. Technical Specification change
[
submittal via JPN-89-084 in December 20, 1989. However, it is worthwhile to summarize pertinent events since that time.
The initial submittal of Technical Specification change request specified that "at least 9 of 11 SRVs shall have a nominal setting r c l
l Excellence Innovation Integrity Teamwork i
i.
l 1
JAF SAFETY RELIEF VALVES SETPOINT CRANGE i
G2 1110 psig with allowable setpoint error of i 3 percent". This submittal was based on the recommendations of GE report NEDC-31697P. Two tolerances are quoted in' Table 1-1 of this report.
- First, the tolerance beyond which valve refurbishment and additional testing is required, was specified as 3%, for both the original three and the new single setpoints. Second, the tolerance on the as-left SRV setting prior to returning a valve to service, was specified as 1%, also for both case setpoints.
Updated SRV Performance Requirements for JAFNPP report NS-31617P-2 of March 1993, did not change any of the above statements. The NRC i
sta change pf valve setpoint tolerance to i'
.However, the NRC did not accept the
. philosophy of an upper limit as means to further reduce the number of LERs. " Evaluation to determine the need for an LER aust be made i
for setpoints when_ drift outside i 3% is found". The subsequent i
SESbsitaal: specification" change"appliottiqm aggopq$jW on f ana,, sent to th,L,,NRC_. on March 2,
1994 vii T,"
rasti i
aMTeemed the above NRS'copoornJh. Approval of these documents is expected in August 1994.-
There are no Analytical Limits associated with these SRVs.
Technical Specification Section 2.2 specifies nominal setting for 9 of 11 SRVs, as discussed above. GE has calculated a nominal SRV setpoint for the original and power uprate conditions, to satisfy both overpressure limits and simmer margin. Therefore, no formal Instrument Loop Setpoint calculation has to be performed. It is only a past Drift (DR), which has to be evaluated.
' The SRV pilots have been refurbished for a setting of i 1%, before they were returned to the site.
This setting is defined as j Calibration Tolerance (CT), which may be used in place of Reference Accuracy (RA) in the setpoint uncertainty calculation (Ref.3). The 8
fmaximumpermissiblesetpointtoleranceofi3%,beyondwhichvalve refurbishment is required,, sh uld have a margin when.considering CT and Drift (DR). The.
indicaje that the aetgal;dr.ify for the JAFNPP SRVs did coed 1.40 W 3 a 24-
{monthperiod-seenote1below.The oo @$pge"tri A"or 0.74% o'r.~S.2 ~~ T nty will be 2.264, M ' the.
15aving' h KE
, to avoid unnecessary LERs-see
- Attachment to this meno. This applies for the present and 24-month
' rsfueling cycle.
Notes:
1.The past actual drifts were established using Wyle Laboratories Group records. The standard deviation and average was calculated using % average of As-Found and As-Left (refurbished) data, as a measure of past drifts of various SRV pilots. With sample size of j
20, the 95/95 Tolerance Factor is 2.75 (Ref.1).
j ;
l 3
~.
JAF SAFETY RELIEF YALVES SETPOIIrf CEANGE One set of data with average real drift of 3.75% (see Att.page 3),
was not used. JAFNPP is switching to new disc material which will preclude effects of the first " pop" during the test, causing the highest deviation from the sample. As mentioned before, past drift did not exceed 1.4%; most drifts were below 1.0%.?
2.It is assumed that Measurement & Test Equipment Error (MTE) is included in the Wyle Labs documented test data.
To conclude, NED/IEC concurs with the following:
single SRV 1110 and 1145 psig,(power the,aprimum p.setpoints ofermissible setpoint tolerance of f,J" t~1h 1.
iWfof Both it~ Aimi"24-monta"cyclag'.
^
2.
"GE'aSalys[s,} presently,tyo of eleven SRVs.nay be without adversely af f actihy~ HPCI", ' RCIC,' Pi'imary
- C'
'nment integrity, Fuel Thermal Limits and ECCS/LOCA performance. It is recommended that these items be re-visited before implementing power uprate.
3.
All eleven SRVs should be replaced with the refurbished and racertified pilots, if necessary, prior to startup of each.
refueling outage. This is twice as many SRVs as required by the Technical Specification.
This will provide greater assurance to satisfy paragraphs 1 & 2 above. However, not all pilots need to be sent out to Wyle Labs. As an example, replacing five previously set and certified spare pilots and sending out six for testing and new settings, will satisfy requirement for replacement of all SRV pilcts. When the SRVs with spare pilots (previously reset and certified) were installed and exercised after each plant startup, no SRV leakage was ever detected. It may be assumed that these pilots do not drift while in storage.
l The above discussion covered the single setpoint issue only.
Additional " Items potentially affected by SRV Setpoint Change",
identified by SE (see attached list), are not addressed here for the following reasons. Technical Specification request for single setpoint change has been completed.
FSAR change and EOP Calculations-if any-will be addressed after the installation of l
single setpoint pilots. Specific SRV setpoints are not quoted in FSAR. Editorial changes will be needed on pages 14.5-2, -12 & -13, where the text refers to a " lowest safety / relief valve group and setpoints". This implies that there is more than.one group of SRV setpoints.
MOV analysis is currently run for power uprate conditions. Since the 1110 psig single setpoint envelopes the existing setpoints, there is no need to run this program for this pressure-see NED-AP-94-377. I
4
~
JAF SAFETY RELIBF YALVES SETPOINT CRANGE Should there be changes to HPCI, RCIC, RPV Transient or ATWS system -
responses, they will be addressed after the single setpoint of 1110 psig is approved for implementation. The same applies for Torus
- Loading, operator Training Lesson Plans and Plant Equipment Database. SRV Surveillance Test MST-102.04 will be updated for applicable setpoint change. It is further recommended that these issues be revisited and further analyzed for the Plant Power Uprate conditions.
References:
1.
Probability and Statistics for Engineers and Scientists, 3-rd Edition by R.E.Walpole and R.H.Myers, 1985 MacMillan Publ.Co.
2.
Statistics for Management by R.I. Levin, 4-th Edition.
3.
ISA-dRP67.04, Part II, Methodologies for the Determination of Satpoints for Nuclear Safety-Related Instrumentation (Draft 10).
4.
MST-102.5 Rev.00,RV Safety / Relief Valve Setpoint Verification.
y M
.s w / 4 Ye# d u rx l
- *f~ f*.t hf j$ (g, 7<forcefl'W A
977J-o $-en S.
hbu/*4os.) f(E4 yf Aa a
~~
/ J,jAo brah 20,jee' AndrM Petrenko n:e prs-f9044 G 4
Sr.I&C Engineer, NED/IEC
^
cc:R.Fredrickson D. Ruddy J. Gray J.Ellmers T.Savo
't i
G.Ottman Records-11 J. - - _ _ - -
4 1TTACEMENT 4.1 anTronre==an mancast TAF-PV-ao$ra, mu no,s 2W-7/A+L u.aification No.:
D IP3 saxr o
scR wumber:
Chanae 6N MMed Mr all (l SWs to ///O pSk7 Description of Change:
Reedted b I W Re h l Ouhme b k Spet A m b eri Requested Implementation Date/ Reason:
equesi cae<4)v at Mc b inmwnb iwte M T.S. [See abeW tetetewee lish) bdPaN bftvnan 6f26fW Requested By:
(Originator)
(Date)
SETPOIET CEAEEE - Check as appropriate gh category:
3 category I O category M O Non-Cat I (IP3)
O Cat II/III (JAP) setpoint Type:
O1 52 03 04 05 0 mapedited 5 Permanent O Temporary I Approved O Disapproved Reason for Disapproval d-Ad.h th chs/w /Afd M 6bs/w srx7 son-a-ps
'(Systeh Engineer)
(Date)
(Sys Eng Ngr/Sup)
(Date)
(Calc. No.)
(Rev.)
hv70 -#1b/f/d-AsM/r W7&Acv
//f4/k/
h $Y4.--
Ib'f (Sponsor organisation)
S r
inser)
(Datd)
(Calculation Soview)
(Date)
Y2$$
(Mgf70up,WPO-NED/ SED /TSD)
'(Date)
(System Engineer)
(Date)
R&PETT AED EN ETRrJIRTim AFFLIC&BILITE RETIM XEA EQ
- 1. Is there a change to the facility as described in the PSARt B
O Sffrowr.s*V khur ssa A 77; Hs* Met td.c-th."f St&J Justifications umf Fwri-arnwr S>ev HwPs To o O
M
- 2. Is there a change to a procedure as described in the PSAR7 Justification O
JE
- 3. Is there a test or esperiment act described in the PSARr Justification:
- 4. Is there a change to the Technical Specifications or epting Licenser JE O
Justifications v?bA7/b CRW M M M/*Wffv/A/Aff 7s 1975 Ot? - sf1 A77. 72w-4 A-onn. urah 'AfAb' C '99f e O
lt
- 5. Is there a change to the Environmental Protection Plant M
- ' ' ~ ~ ~ ~
- ~ -
Justification:
' 6*
.2 O
8
- 6. Will there be an adverse effect on the environment?
Justificatical O Answers to all questions are 30. No potential for an unreviewed Safety or Environmental Quest exists.
El Answers to Questica 2 or 3 marked *
. Preparation of a Safety Evaluation is I
required. Safety Eva sation No. NDr
.fX NEh-M 37y, R Answer to Question 4 marked 'YE4*.
Preparation of a license==aad==at is required.
Letter No. J7N-14-#/L /egnor2H4 O Answer to guestion s of a marked fas. Preparation of an Environmental Evaluation required.
avironmental Evaluation No.
JV/4/94 8Lf l N 9 4 L NMY (Per$ breed By)
(Date)
'(Reviewed By(
(Date) umanamuser arPaarms.
(PORC Meeting. No.)
(PORC Chairman)
(Date)
(Resident Manager)
(Data)
ATTACIIGMT 4.s1 (COM'T1 SETPOINT N GE manUEST sCR Dumber: MA W D
arranzrun surroIsr cammme O Approved O Disapproved Reason for Disapproval (Requested By)
(Date)
(Safety / Environ / Tech Review)
(Data)
(Res Ngr, setpoint Types 1 & 2)
(Date)
(Ieplementing Document)
(Shift sup, Setpoint Types 3, 4 & 5)
O camCxLIarze Reasons (Requested By)
(Date)
(Speer Organisation)
(Data)
(System Engineer)
(Date)
(PORC Mtg No.)
(PORC Chairman)
(Date)
(Resident Manager)
(Date)
Setpoint Change Document Update Sheets:
Transmitted:
Received:
Attached:
Setpoint Change Document Revision Lists: Transmitted:
Received:
Attached:
DOC m AFFBCrEn Name/ Number Ocarability Closecut fff A*1'7. 41f 7 ef" / 7fAff ?pf/Av 7/d//,P O
O f / / SC f f b A Y S W.ff f? ! w 'r C Yd q d O
O O
O O
O O
O O
O l
C M IM (Implemented-work Doc. No.)
(Date)
(Job Supervisor)
(Date)
Data satry Request - Initiated:
Returned:
(Data)
(Date)
Required Documents Updated or Modification Follow List (NFL):
(Date)
(Date) l Completed:
(setpoint Coordinator)
(Date)
NYPA FORM MCM-8, ATTAC1 DENT 4.1 (REVISION 0)
Page 2 of 2
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ATTACEMENT TO NED-AP-94-379 WYLE LABORATORIES GROUP REPORTS Collected at JAFNPP on 8/10/94 Cartidge 5/g 305,0,at 1105 11 psig,.Na#adnitarar21/42.
i Test Run /
As Found Test Run #
Refurbished To Notes f
1 1119psig 3
1113 Runs 1 & 2 2
1111 4
1112 Adjusted 3
1108 5
1105 4
1107 6
1102 Avg.1111psig Avg.1108psig
,Q uns on 1
1101 2
1071 3
1110 Run2, instr 3
1065 4
1105 failure.
4 1066 5
1104 l
Avg.1078psig Avg.1105psig f
Past Drift =1105-1111=-6.Opsig/1105)100%=-0.54%
1 i
Cartidge S/N 1087 at 1140 11 psig, March 21 & 25/92.
l Test Run #
As Found Test Run #
Refurbished To Notes 1
1132psig 3
1139psig 2
1142 4
1139 l
3 1134 5
1140 4
1129 6
1135 7
1137 8
1133 Avg.1134psig Avg.1137psig Runs on Dec.12/89 E'. April 23/90. /
1 1194*
3 1152psig
- Notice of 2
1121 4
1144 anomaly.
3 1115 5
1140 Runs 1 & 2 4
1108 6
1146 Adjusted.
1 7
1143 Avg.1114.7psig Avg.1145psig i
Past Drift =(1145-1134)/1145)100%=0.96%
RunsonN6Q(!E20/h'7/
10 1141psig
- Instr.
3 1120 11 1140 Malfunct.
4 1113 12 1137 5
1127 13 1141 Avg.1119.5psig Avg.1139.75psig Past Drift =(1139.75-1114.7)/1139.75)100%=2.2%
WYLE LABORATORIES GROUP REPORTS collected at JAFNPP on 8/10/94 Cartridge % D M 1140 1 11 psig,ApD&R}1'[$4T.'JUri,'131/92 Test Run #
As Found Test Run #
Refurbished To 1
1149 psig 1
1141 2
1134 2
1147 3
1128 3
1148 4
1134 4
1151 Avg. 1136.25 psig 5
1149 Both Runs 1
1148 psig 8
1150 2
1140 9
1148 3
1139 10 1144
' {-
4 1134 11 1158 Avg.1147.5 psig Past Drift =(1147.5-1136.25)/1147)100%=0.98%
Cartridge S/N 1111 at 1140 1 11 psig, June 7 & 20/90 Test Run #
As Found Test Run #
Refurbished To 1
1144 psig 1
1137 2
1142 2
1133 3
1145 3
1130 4
1143 4
1130 Avg. 1143.5 psig Both Runs g'j[Q)/,87 /
1 1153 psig 10 1141 2
1143 11 1140 3
1147 12 1138 4
1144 13 1140 5
1143 Avg.1139.75 peig Past Drift =(1139.75-1143.5)/1143.5)100%=0.33%
i -
.d WYLE IJAORATORIES GROUP REPORTS Collected at JAFNPP on 8/10/94 Cartridge 5/N 1012 at 1140 i 11 psig, March 18 & 21/92 Test Run #
As Found Test Run #
Refurbished To 1
1156 psig*
3 1149 2
1103 4
1146 3
1102 5
1145' 4
1186 6
1145 Avg. 1097 psig 7
1142
- Notice of Anomaly Runs on Oct. 10 & 33/88 1
1142 psig i
1141 2
1127 2
1139 3
1127 3
1141 4
1132 4
1138 Avg.1139.75psig Past Drift =(1139.75-1097)/1139.75)100%=3.75%*
- Assuming leakage, more than 3 (Std Dev.), not used.
Runs on Feb. 5 & 16/87.
1 1131 psig 8
1151 2
1198 9
1146 3
1100 10 1150 4
1104 11 1146 Avg.1148.25psig Past Drift =(1148.25-1132)/1148.25)100%=1.4%
Cartridge S/N 1013 at 1140 1 11 psig, March 19 & April 8/90 Test Run #
As Found Test Run #
Refurbished to 1
1227 psig*
4 1139 2
1157 5
1143 3
1160 6
1145 4
1157 7
1146 Avg. 1158 psig Runs of April 20 & 24/90 1
1147 3
1144 2
1170 4
1146 3
1176 5
1138 4
1157 6
1149 Avg.1162.5psig Avg. 1144.25psig
- Notice of Anomaly Past Drift =(1144.25-1158)/1158)100%=1.19%
i NYLE IABORATORIES GROUP REPO@TS Collected at JAFNPP on 8/10/94
]
Runs on M & g g t.,1/87, 1
1171 1
1143 2
1145 2
1145
-l 3
1140 3
1150 4
1135 4
1150 Avg.1147.0 Past Drift =(1147-1162.5)/1162.5)100%=-1.33%
Cartridge S/N 1045 at 1140111psig, August 8/90 & Jan. 20/92 l
Test Run #
As Found Test Run #
Refurbished To 1
1167psig*
1 1151 2
1142 2
1151 3
1146 3
1146 4
1149 4
1147 Avg. 1145.67psig
- Note of Anomaly Runs on May 4'&' August 11/86/
1 1151psig 3
1141 i'
1149 4
1145 3
1147 5
1142 4
1146 7
1149 i
Avg.1144.25psig r
Past Drift =(1144.25-1145.67)/1145.67)100=0.12%
Cartridge S/N 1950!at 1150.i 10 psig, March 20 & """'il 6/9$
Test Run #
As Found Test Run #
Refurbished T6 j
1 1122psig*
1 1199 2
1101 2
1101 3
1102 3
1197 4
1102 4
1194 5
1104 Avg. 1101.67psig Avg. 1099psig
- Notice of Anomaly Runs on Dec. g j { K y i1Tif/90/
1 1089psig*
1 1097 2
1098 2
1101 3
1095 3
1101 4
1086 4
1107 Avg.1093psig Avg.1101.5psig
- Notice of Anomaly I
WYLE LABORATORIES CROUP REPORTS Collected at JAFNPP on 8/10/94 Past Drift =(1101.5-1101.67)/1101.67) 100%=-0.015%
Runs on % g t? q yS;J/
1 1204PSIG*
4 1110 2
1107 5
1109 3
1105 6
1108 4
1105 7
1105 8
1105 Avg.1105.67psig Avg.1107.4psig
- Notice of Anomaly Past Drift =(1107.4-1193)/1174.4) 100%=1.30%
Runs on % {[hy l
None -------
1 1109 2
1115 3
1105 4
1111 Avg.1110.5psig Past Drift =(1110-1105.67)/1110) 100%=0.23%
Cartridge S/N 1051 at 1090 10 psig, Dec.12/89 & Feb. 3/92 Test Run #
As Found Test Run #
Refurbished To l
1 1080 1
1086 2
No Data 2
1094 3
1080 3
1091 4
1093 Avg.1080psig 5
1089 Runs on A$st*TFF3~i/37 1
1102 1
1095 2
1106 2
1080 3
1105 3
1095 4
1102 4
1092 Avg.1090.5psig j
Past Drift =(1090.5-1080)/1090.5)100%=0.96%
)
WYLE LABORATORIES GROUP REPORTS 9
Collected at JAFNPP on 8/10/94 Cartridge.*asarfign7af 1040t lipsig,:_Ilsrch 19'E' 31/92 '/
7 Test Run #
As Found Test Run #
Refurbished To 1
1139 3
1145 2
1125 4
1143 3
1126 5
1144 4
1130 6
1148 Avg.1130psig 7
1149 8
1143 Runs on April, M [9a.g 1
1147 7
1140 2
1156 8
1145 3
1150 9
1145 4
1144 10 1146 Avg.1149.25psig Aug.1144psig Past Drif t=(1144-1130)1144)100%=1.22%
Runs on OUNY $1 T EI/88 1
1165 4
1144 2
1156 5
1148 3
1156 6
1137 4
1155 7
1137 Avg.1158psig Avg.1141.5psig Past Grift =(1141.5-1149.25)/1149.25)100%=0.67%
1 1137 12 1145 2
1137 13 1146 3
1136 14 1146 4
1136 15 1145 Aug.1145.5psig Past Drift =(1145.5-1158)/1158)100%=1.08%
Cartridge M(djM at 1140 i 11psig, Q@Fi21/92/
Test Run #
As Found Test Run #
Refurbished To 1
1139 5
1142 2
1138 6
1140
{
3 1135 7
1139 4
1134 8
1140 Avg.1136.5psig i
i WYLE LABORATORIES GROUP REPORTS Collected at JAFNPP on 8/10/94 Runs on Dec. 14/88 & April 1/90 1
1156 5
1136 2
1153 6
1130 3
1146 7
1137 4
1139 8
1141 9
1143 Avg.1139.2psig Past Drift =(1139.2-1136.5)/1139.2)100%=0.24%
Cartridge S/N 1056 at 1105 i 11psig, April 21 & May 9/90 Test Run #
As Found Test Run Refurbished To 1
1104 1
1115 2
1101 2
1114 3
1110 3
1113 4
1107 4
1110 Avg.1105.5psig Runs on May 2 & July 14/88 1
1110 2
1111 2
1184 3
1113 3
1181 4
1108 4
1183 5
1111 Avg.1089.5psig Avg /1110.75psig Past Drift =(1110.75-1105.5)/1105.5)100%=0.48%
1 1143*
9 1104 2
1104 10 1108 3
1107 11 1102 4
1104 12 1101 Avg.1005 Aug.1103.75psig
- Notice of Anomaly Past Drift =(1103.75-1089.5)/1103.75)100%=1.29%
Cartridge S/N 1062 at 1090 1 11psig, June 7 & 21/90 Test Run /
As Found Test Run Refurbished To 1
1077*
1 1097 2
1176 2
1098 3
1178 3
1102 4
1174 4
1099 Avg.1076psig
- Notice of Anomaly..
.3..
WYLE LABORATORIES GROUP REPORTS Collected at JAFNFP on 8/10/94 Runs on May 3 & August 17/88 1
1172
- 10 1090 11 1089 12 1085 13 1083 l
- Severe Leakage Avg.10r,6.75psig i
Past Drift =(1086.5-1076)/1086.5)100%=0.99%
cartridge S/N 1088 at 1090 1 10 psig, Dec. 21/88 & August 18/89 Test Run #
As Found Test Run Refurbished To 1
1156psig*
1 1104 2
1121 2
1108 3
1124 3
1184 4
1130 4
1091 5
1086 6
1094
- Notice of Anomaly Avg.1094.5psig Runs on June 16 & 19/94 2
1084 1
1010110* Inlet Pressure 3
1088 2
1010110*
4 1088 3
1010 10*
5 1085 4
1010110* -
Avg.1086.25psig Past Drift =(1094.50-1086.25)/1094.5)100%=0.75%
i l l
M E. -
. i l
k ATTACEMENT TO.NED-AP-94-379 l
LExpected: Drift, based _on~the saple size'is:
DR=2.75(STD DEV)+ AVG j
The Past. Drift data for the Standard Deviation, are from_the above.
data' bank'and are summarized here. Note:it is a pilot cartidge air operator being refurbished. Then, the pilots are returned to the site to be installed on Safety Relief Valves as needed..
Cartidge S/N x
x-R (x-R)*
Notes 1080 0.54 0.264 0.070 1087 0.96 0.156 0.024 1087 2.20 Not used.
>3(Std Dev) 1110 0.98 0.176 0.031 1111 0.33 0.474 0.225 1012 1.40 0.596 0.355 1013 1.19 0.386 0.149
)
1013 1.33 0.525 0.276 1045 0.12 0.684-0.468 l
1050 0.015 0.789
-0.622 1050 1.30 0.496 0.Z46 1050 0.23-0.574 0.329 1051 0.96 0.156 0.024 1052 1.22 0.416 0.173 i
1052 0.67 0.134 0.018 1052 1.08 0.276 0.076 1053 0.237 0.567 0.321 1056 0.48 0.324 0.105 l
1056 1.29 0.486 0.236 1062 0.99 0.186 0.034 1088 0.75 0.054 0.003 Then, SUN x = 16.072 Mean x = 16.072/20"= 0.804 Std Dev = SUM (x-x) /n-1) = (3.745/19) = 0.446 Then, the expected Drift will be: DR= 2.75(0.446)+0.804=2.03%.
and Channel Uncertainty is:
- f., -s :
1.(CT + DR*) =(1.0*+ 2.03* f =
.26% or 25.
psig, rounded to 25.0 psig. The margia is:
- s. o.~
. 3.0 - 2.26 =i 0.74% or i 8.21 psig, or about 8.0 psig.
i As-Found-Zone (AFZ) is: 1,110 1 25psig, or (1,085-1,135)psig.
- Leave-As-Is-Zone (LAIZ),to be used for refurbished pilots,is:
1,110 i 11psig, or (1,099-1,121)psig.
The Range is 1,110 i 33 psig, or (1,077-1,143)psig.
-g-m
--w p