ML20197H386
| ML20197H386 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 12/22/1997 |
| From: | Salas P TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-94117, TAC-94118, NUDOCS 9712310266 | |
| Download: ML20197H386 (16) | |
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Tennessee Vaney Authority, Post Offce Ekx 2000, Soity-Daisy, Tennesase 37379-2000 December 22, 1997.
U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washingtc..,
D.C.
20555 Gentlemen:
In'the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SQN) - REVISED INSERVICE TEST (IST)
PROGRAM, SECOND TEN-YEAR INTERVAL
References:
- 1. NRC letter to TVA dated March 20, 1996,
" Inservice Testing Program Relief Requests, Second Ten-Year Interval Sequoyah Nuclear Plant Units 1 and 2 (TAC Nos. M94117 and M94118)"
- 2. TVA letter to NRC dated November 21, 1995, "Sequoyah Nuclear Plant (SON) - American Society of Mechanical Engineers (ASME) -
Section XI Programs for the Second Inspection Interval, Units 1 and 2" The purpose of this letter is to provide amended relief requests contained in SON's Second Ten-Year Interval IST Program.
This information is being provided in response to Referenco 1.
By Reference 2, TVA submitted SON's Second Ten-Year Interval-
'IST. Program along with thirteen relief requests for NRC review.. By. Reference 1, NRC issued the results of their review in a safety evaluation report ::(SER).
As described in the SER, SON Relief Requests RP-03 and RP-05 were approved and authorized for use with certain conditional. requirements as: stipulated'in the SER for each relief request.
SQN Relief Requests RP-07, RV-05, and RV-06 were denied for use.
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TVA evaluated:-the Referencsci 1 SER and has enclosed amended Rellef1 Requests RP-03,_RP-05,-RP-07,-'RV-5, and:RV-6 for NRC'-
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Lconside. tion and-: approval.
In addition,: a : new relief l reque.
'RV-7) is-provided for review and approval.
"Pleaseidirect-questions-'concerning-this issue _to me at.
(423)i843-7170 or J. D. Smith-at'(423).843-6672.-
LSincerely,
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Pedro Salas.
Licensing and Incustry'A'ffairs-Manager Enclosures-a cc (Enclosures):
.Mr. R. W.~Hernan,.Priject Manager.
Nuclear Regulatory-Commission-f One White Flint, North
-11555 Rockville Pike Rockville,-Maryland 20852-2739-
- NRC
- Resident Inspector Sequoyah Nuclear. Plant
- 2600-Igou Ferry Road Soddy-Daisy,-Tennessee-37379-3624
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. Regional Administrator
- U.S.--Nuclear Regulatory Commission Region-II Atlanta __ Federal Center 61~Forsyth St.,
SW, Suite 23T85
-Atlanta,: Georgia 30303-3415 t
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l ENCLOSURE TENNESSEE NALLEY AUTHORITY SEQUOYAH NUC: LEAR PLANT (SQN)
UNITS 1 AND 2 Inservice Test Relief Requests - Second 10-Yiar Interval PUMP RELIEF REQUEST - RP-03 PUMP RELIEF REQUEST - RP-05 PUMP RELIEF REQUEST - RP-07 VALVE RELIEP REQUEST - RV-5 VALVE RELIEl' REQUEST - RV-6 VALVE RELIEF REQUEST - RV-7 E-1
n.
SEQUOYAE NUCLEAR PLANT UNITS 1'AND 2 IST REQUEST-FOR RELIEF, RP,.
AFFECTED COMPONENT:
Boric Acid Transfer' Pumps TEST
- PEQUIREMENT:-
OM-6, Paragraph 4. 6.1.2 (a), requirec that the full-scale. range of each analog instrument shall not_be greater than three times the reference value.
BASIS FOR RELIEF:
These pumps have low-suction pressure regairements where the pressure has been measured as low as.
1.5 psig with typical' suction pressure readings of
'2 to 5 psig.
To meet the requirements of OM-6, Paragraph 4.6.1.2(a), special low-range pressure gauges would have to be purchased.
A multiplication of three times the reference pressure value is 4.5 psig and the maximum allowable error of 2% would be 0.09 psig.
Using a
-15 prig gauge during testing would provide a maximum allowable' error of 0.30 psig.
The 0.21'psig difference in accuracy'of the two gauges is negligible.
In addition, a typical 15 psig suction pressure gar e has subdivision increments of 0.05 psig, which ould allow pt 71se readings to one half of this increment or 0.025 psig.
-The Boric Acid Transfer Pumps have a differential pressure of 80 to 90 psid and typical discharge pressure readings are in the range of 90 to 105 psig.
The discharge pressure is measured with 150 to 300 psig gauges with a minimum accuracy of
+/- 1%.
This exceeds code accuracy requirements.
The discharge pressure is the_ controlling value in the differential pressure measurement.
The gauge used to measure the discharge pressure would provide accurate and repeatable readings necessary
-to determine acceptable pump _ performance.
Considering'the accuracy in readability (i.e.,
0.025 psig for a 15 psig suction gauge), the Jeffect on the determination of an accurate
-differential pressure reading is negligible.
This_ relief _ request is based on guidance provided in the SER-dated October 23, 1987, and was approved in SON's first 10-year interval.
E-2
- SEQUOYAN NUCLEAR PLANT-UNITS 1 AND 2'
_ IST REQUEST FOR ; RELIEF, RP 03 (continued)
" ALTERNATIVE TEST!
Pump testing will be performed using 15 psig suction gauges in'lieufof_ gauges required by OM-6, Paragraph 4.6.1.2(a).
This~ relief request is based upon NUREG-1482, Section 5.5.1.
Note: 'This_ relief request originally addressed the shutdown board room chilled water pumps..- Relief for these pumps is no longer required.
The original decision to include these pumps was-based upon information obtained through operational experience.
During baseline pump testing, it was determined that the suction pressure was significantly higher than information obtained.
through operational experience such_that code required gauges could be used.-
CONCLUSION:
Based upon the.above. discussion, the alternative-test provides an acceptable level of quality and safety... Authorization to implement the proposed alternative is requested in accordance with 10CFR50. 55a (3) (1).
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8EQUOYAN NUCLEAR PLANT UNITS 1 AND 2' IST REQUEST FOR RELIEF, RP-05 AFFECTED
- COMPONENTS :.
Containment" Spray Pumps TEST:
--REQUIREMENT:
OM-6, Paragraph 4.6.1.2 requires digital instruments to be calibrated such that the reference value does not exceed 70% of the'
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calibrated range of-the instrument.
RASIS FOR
-RELIEF:
Portable digital ultrasonic flow equipment is used to measure flow for the_ Containment spray' Pump tests with the current maximum allowable' reference value for the flow of.4,940 gpm.
Following the calibrated range requirements of OM-6, a reference value of 4,940 gpm would require digital instrumentation with a calibrated range of 7,064 gpm.
A1 flow-rate of 7,064 gpm is equal to a velocity of 45.3 feet per second in the 8-inch Schedule 40 piping of the Containment Spray System.- Per theJspecifications provided by the-ultrasonic flow equipment manufacturer, the maximum flow velocity measurement capability is 40 feet per second or-6,237 gpm.
The ultrasonic flow equipment has a manufacturer's stated accuracy of1+/- 1% and is calibrated to
+/- 2% of the calibrated range.
This exceeds the OM-6 accuracy requireinent of +/- 2% and has proven acceptable for use in determining flow measurements.
Calibration of ultrasonic equipment to.7,064 gpm would not provide any greater assurance that_the equipment.is in calibration at the reference value than calibrating the instrumentation-to the reference value.-
SON does not have installed flow instrumentation or other means to measure flow in the_ Containment Spray System _to the required accuracy other than through the use of ultrasonic portable equipment.
The inability to use ultrasonic equipment would requireia modification'to the piping system with E
4 SEQUOYAN NUCLEAR PLANTiUNITS.1 AND--2 IST, REQUEST FOR RELIEF, RP-05 (continued).
no-appreciable increase in accuracy.
An installed i
flow measuring _ device in the Containment Spray-System would not enhance the detection of pump degradation over that-presently provided by the ultrasonics flow equipment.
ALTERNATIVE-TEST:
Calibrate ultrasonic flow instrumentation such that the reference value for flow (which is the.
set parameter) does not exceed 95% of the-calibrated range.
CONCLUSION: _
Based upon the above discussion, the alternative test provides an acceptable level of quality and safety.
Authorization to implement the proposed alternative is requested in accordance with 10CFR50. 55a (3) (1).
E-5
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SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 IST ka'tUEST FOR' RELIEF, RP-07 This relief request is withdrawn at this time.
E-6
SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 IST REQUEST FOR RELIEF, RV-5 SYSTEM:
Essential Raw Cooling Water VALVES:
67-719h, 67-719B, 67-720A, 67-720B, 67-739A, 67-739B, 67-740A, 67-740B CLASS:
3 CATEGORY:
C-Active TUNCTION:
These check valves open to admit air into the essential raw cooling water (ERCW) pump column to allow the water in the pump column to drain back to the pump pit upon stopping the pump.
Draining the water from the pump column protects the pump motor from excessive starting torque during motor / pump starts.
These valves also remain open for a period of time during pump starts to provide a vent path for the air in the pump columns and discharge head.
The check valves close to provide a flow boundary after the pump starts and water reaches the pump discharge head and check valve.
The valves are located in a horizontal run of pipe and are installed upside down (i.e.,
the valve bonnets face downward) so that gravity will assist in opening the valve.
IMPRACTICAL REQUIREMENT:
The code requirement to verify that valves full stroke open.
BASIS FOR RELIEF:
There is no required flow rate for these valves and no practical way to determine the flow rate through these small diameter valves.
The rules of OM Part 10 and NUREG-1482 were developed with liquid flow in mind rather than compressible gaseous flow.
Attempting to measure air flow rate this small will result in very inaccurate and unrepeatable results.
Additionally, the nature of the flow through these valves is such that it will not be at a steady state long enough to quantify.
The flow will rapidly accelerate to a maximum and E-7
-SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2
.IST REQUEST FOR RELIEF, RV (continued) then steadily decrease at the driving force of the water column lovel above the river elevation
-decreases.
The use of a smoke-test is a method
-that provides positive means to determine that the valve goes to the open_ position.
Tho' valve only needs to open to provide.a vent path so that water drains out of the pump column when the pump is stopped.
Draining water from the pump column ensures that unnecessary starting torque is not placed on the pump motor during motor / pump starts.
Disassembly of the valves-every refueling outage is not a practical means of verifying valve performance since ERCW pump maintenance is not_ tied to refueling outage evolutions.
SON has eight ERCW pumps that can be taken out of service at any time.
-The smoke test provides a-positive means of the valves opening function and can be perfor-9d quarterly as required by the code.
Consequently, the smoke test demonstrates that the valve safety.
function is fulfilled.
ALTERNATE TESTING:
Verify that the valves open after stopping the pumps by use of a smoke test to verify that the
. valves stroke open.
This will be indicated by the smoke being drawn into the piping by the vacuum caused by the water in the pump column when it starts to drain back to the pump pit.
The closing function of the valves is demor.strated during each pump test.
FREQUENCY FOR ALTERNATE TESTING:
Once per quarter.
CONCLUSION:
Based upon the above discussion, the alternative test provides an acceptable level of_ quality and safety.
Authorization to implement the proposed alternative is requested in accordance with 10CFR50.55a (3) (1).
E-8
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!SEQUOYAE NUCLEAR PLANT UNITS-1'AND 2 IST REQUEST FOR" RELIEF,nRV -
SYSTEN:
-Reactor Coolant-System VALVES:-
FSV-68-396 and-FSV-68-397__
CLASS:-
1 CATEGORY:
-B-Active FUNCTION:
Valves are opened manually to provide _a_ reactor vessel head vent path to vent noncondensables.from the head during an accidenttto promote natural-circulation, and to prevent < gases from impeding-reactor coolant-circulation flow through the core.
IMPRACTICAL REQUIREMENT;- - Quarterly 3troke time test.
BASIS FOR-RELIEF:
These valvas are'one-inch Target Rock solenoid valves that have no position indication and are totally enclosed (seal welded bonnet) which prevents visual confirmation of valve position.
This valve design creates the inability to measure the time that it takes the valve-to stroke.
These valves are throttle valves with a thumbwheel control that positions'the valve at 0%, 25%, 50%,
and 100%.
These valves are fast acting valves with a stroke time of less than two-seconds and a stroke of approximately one quarter of an inch.
An enhanced maintenance program of disassembly and inspection was considered.
This method was not considered appropriate for the-following reasons.
First, the valves are located inside containment.
In addition, frequent disassembly can lead to operational. problems due to distortion of the valve parts caused by the repetitive welding process to-reinstall'the seal weld..This is.not: considered acceptable for the purposes of testing and could lead to premature replacement of the valves.
3 E-9
SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 IST REQUEST _FOR RELIEF, RV-06 (continued)
Secondly, once the valve is opened and the internals of the valve are examined, the condition of the internal parts would not provide any additional indication of acceptable valve operation than the acoustic monitoring.
TVA believes there is no feasib.1 method for measuring the stroke time.
In aduition, an enhanced maintenance program would not provide additional assurance of acceptable valve operation.
The alternative method described below provides an acoustical monitoring method to determine acceptable valve operation.
ALTERNATE TESTING:
TVA proposes to verify that the valve operates properly through the use of acoustic monitoring.
An acoustic monitoring signal of the system noise is taken prior to opening the valvo.
The valve is opened by operating the thumbwheel controller and another acoustic signal is ebtained at the full-open position.
The 'ralve is then closed and another acoustic signal le obtained at full closed.
The initial acoustic signal at the full-closed position is compared to the second acoustic signal taken at the full-closed position.
Comparative values provide assurance that the valve is moving to the correct position and that the valve is operating acceptably.
FREQUENCY:
Every refueling outage for acoustic monitoring.
NOTE:
This is similar to relief which was approved for SON's First 10-Year Interval Inservice Test Program and follows the guidance given in NUREG-1482, Section 4.2.8.
CONCLUSION:
Based upon the above discussion, the alternative test provides an acceptable level of quality and safety.
Authorization to implement the proposed alternative is requested in accordance with 10CFR50.55a (3) (1).
E-10
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d SEQUOYAH NUCLEAR PLANT UNITS l'AND 2 IST REQUEST FOR RELIEF,_RV-7.
- SYSTEM:
Chemical'and Volume Control-VALVE:
Manual Valve _62-546 CLASS:.
2 CATEGORY:
B-Active FUNCTION:
In the closed position,=this valve prevents reactor coola.it. pump (RCP) seal inject' ion flow from bypassing the seal injection filters and provides the safety-related function of containment 4
isolation. -The valve is capable of being opened to:
provide RCP seal injection flow in the event that both seal injection filters are blocked.
- IMPRACTICAL REQUIRENENT:
Manually exercise the valve to its safety position (closed for containment isolation).
BASIS.
FOR RELIEF:
Opening this manual valve bypass around the seal injection filters allows-the introduction of impurities into the RCP seals which, in the~past, has caused' damage to the pump seals and resulted in increased seal leakage.
Cycling this valve during unit refueling could allow impurities to bypass the seal injection filters and be pumped forward to the pump seals upon the return of unit operation.
TVA considers risk of damage to the RCP seals to be higher than'the-probability that this valve would be required to close for containment isolation.
System Operating Instructions require the valve in the bypass line to P1 isolated during normal operation, thereby 'solating the bypass line from the supply line, because this valve is normally closed and is not_normally operated, the probability'of the valve.being open at the time containment. isolation is required, combined with the probability of its failure to close, is sufficiently? low.
E-ll
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SEQUOYAR NUCLEAR PLANT UNIT 8.1 AND 2
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v IST REQUEST FOR RELIEF, RV-7 (continued)
Manual Valve'62-546_is part of a TVA Class B American Society _of MechanicalLEngineers_(ASME)
Class 2 closed system-outside of containment.-
.During normal reactor operation, RCP seal injection is provided through one of two seal injection filters, which are in parallel along with the seal injection filter bypass line.
Downstream of the filters, the supply line separates into four lines at the containment penetration to feed each RCP individually.
At the containment penetration 1are inboard check. valves and outboard manual isolation needle-valves.
The system design also provides~a-
.second check valve that is not missile protected, downstream of each inboard check valve in close proximity of each RCP.
In each of the four seal injection lines, the outboard valves are not capable of remote operation and receive no isolation _ signal for automatic closure.
The four manual valves at the containment penetration are-not accest's:e postaccident due to high radiation dose rates in'the vicinity-of their location.
Valve 62-546 (along with 62-549 and 62-550),was-approved by NRC as exemption to the GDC 55 design requirements for containment isolation as one of.
the outboard containment isolation valves, which are capable of being closed quickly to isolate the seal injection containment penetration for containment isolation.
These valves, which isolate the seal water injection filters and the bypass lines, may be operated-by reach rods extending from the concrete cubicle housing to the valves.
Postaccident seal injection is supplied by the high-head Anjection pumps (i. e., centrifugal charging pumps), which also provide the seal injection _ flow and normal charging flow in nonaccident conditions.
Under normal, transient, and accident-conditions, at least one of the centrifugal-charging pumps is in operation providing emergency core cooling system / charging flow /sealiflow as required. Therefore, a water seal L
will-be provided at a. pressure greater than 1.1 P.
L with at least aE30-day water supply to preclude air leakage out of containment.
These redundant E
isolation provisions (i.e.,
the inboard check valves, the closed system, the water seal, and the seal injection. filter E-12
. =. _.
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-t SEQUOYAN NUCLEAR PLANT UNITS 1 AND.2 IST. REQUEST FOR RELIEF, RV-7 (continued) isolation valves)Lprovide assurance that no single failure'could result-in-release of containment atmosphere to the environment.
-In accordance-with 10CFR50 Appendix J, Paragraph III.C.3,'the seal injection penetrations are not a potential containment atmosphere leak path and do not require a Type C leak test with air or nitrogen.
Additionally, a water leak test is not required since a continuous supply of seal water is provided from the containment sump.
The exemption to GDC 55, noted above for nonautomatic containment isolation valves, was approved the NRC on December 14, 1987 (reference TAC Nos. 64623, 64389).
The leak rate testing has been evaluated and found acceptable in NUREG-1232, Volume 2, Section 3.6.
Therefere, the risk associated with testing this valve does not provide a commensurate increase in safety.
ALTERNATE
. TEST:
None CONCLUSION:
Based upon the above discussion, the proposed alternative test would provide an acceptable level of quality and safety.
Authorization to implement the proposed alternative is requested in accordance with 10CFR50.55a (3) (1).
E-13
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