ML20236Y209
| ML20236Y209 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/07/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20236Y205 | List: |
| References | |
| NUDOCS 9808110254 | |
| Download: ML20236Y209 (16) | |
Text
Wen p k UNITED STATES j
t NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006H100',
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE INSERVICE TESTli4G PROGRAM RELIEF REQUESTS TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NUMBERS 50-327 AND 50-328
1.0 INTRODUCTION
The Code of Federal Regulations,10 CFR 50.55a, requires that inservice testing (IST) of certain American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 pumps and valves be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (the Code) and applicable addenda, except where relief has been requested and granted or proposed alternatives have been authorized by the Commission pursuant to 10 CFR 50.55a (f)(6)(i), (a)(3)(i), or (a)(3)(ii). In order to obtain authorization or relief, the licensee must demonstrate that: (1) conformance is impractical for its facility; (2) the proposed alternative provides an acceptable level of quality and safety; or (3) compliance would result ir. a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Section j S0.55a (f)(4)(iv) provides that inservice tests of pumps and valves may meet the requirements j set forth in subsequent editions and addenda that are incorporated by reference in 10 CFR j
50.55a(b), subject to the limitations and modifications listed, and subject to U.S. Nuclear Regulatory Comm:n@n (NRC) approval. NRC guidance contained in Generic Letter (GL) 89- 1 04, " Guidance on Developing Acceptable Inservice Testing Programs," provided alternatives to the Code requirements determined to be acceptable to the staff and authorized the use of the attematives in Positions 1,2,6,7,9, and 10 provided the licensee follow the guidance i delineated in the applicable position. When an alternative is proposed which is in accordance with GL 89-04 guidance and is documented in the IST program, no further evaluation is required; however, implementation of the alternative is subject to NRC inspection.
The Tennessee Valley Authcrity (TVA or the licensee) submitted amended Relief Requests RP-03, RP-05, RP-07, RV-05 and RV-06, as well as new Relief Request RV-07, in a letter l dated December 22,1997, in addition, the licensee submitted new Relief Request RV-08 in a l letter dated April 16,1998. Section 50.55a authorizes the Commission to grant relief from ASME Code requirements or to approve proposed altematives upon making the necessary findings. The NRC staff's findings with respect to granting or not granting the relief roouested j or authorizing the proposed attemative as part of the licensee's IST program are contained in !
this Safety Evaluation (SE). I l
EELOSURE :
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The second 10-year interval for Sequoyah Nuclear Plant Units 1 and 2 (SON) was estimated by the licensee to have begun on December 15,1995. The estimated date was attributed by the licensee to be due to an extended outage at SON. The licensee did not provide what they believe the exact second ten-year interval start date is, in either of their submittals, as requested by the staff. Therefore, this issue is still pending. The current IST program is based on the requirements of the 1989 Edition of ASME Section XI.
2.0 PUMP RELIEF REQUESTS 2.1 Relief Recuest RP-03 The licensee has requested relief from the full-scale range instrument requirements of OM-6, Paragraph 4.6.1.2(a), for the boric acid transfer pumps. The licensee has proposed to use 15 psig gauges to take pump suction pressure measurements.
2.1.t Licensee's Basis for Reauestina Relief The licensee states: l l
These pumps have low-suction pressure requirements 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.
U!.ing a 15 psig gauge during testing would provide a maximum allowable error of 0.30 psig. The 0.21 psig difference in r ccuracy of the two gauges is negligible. In addition, a typical 15 psig suction pressure gauge has subdivision increments of 0.05 psig, which would allow precise 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 of11%. 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 effect on the determination of an accurate differential pressure readi>.g is negligible.
2.1.2 Altemative Testina The licensee proposes:
Pump testing will be performed using 15 psig suction gauges in lieu of gauges required by OM-6, Paragraph 4.6.1.2(a).
This relief request is based upon NUREG-1482, Section 5.5.1.
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l 2.1.3 Evaluation l Table 2 of OM-6 specifies that differential pressure shall be taken during each safety-related l pump test. Differential pressure may be determined by using a differential pressure gauge or
! calculating the difference between the discharge and suction pressure gauge. The Code requires that the full-scale range of each analog instrument shall not be greater than three times the reference value of the instrumented pump parameter. The reference value for the
. suction side of the boric acid transfer pumps is 1.5 psig while the range of the suction pressure gauge is 0 to 15 psig which exceeds the range allowed in the Code. In an NRC staff SE dated March 20,1996, the licensee's attemative was awhorized for a period of 180 days to allow the licensee to further investigate the burden of using gauges that meet the Code requirements and more fully assess the impact on the final measurement of differential pressure and resubmit their proposed alternative. It should be noted that the licensee's resubmittal was 15 months after the interim period expired.
The ficansee stated that the discharge pressure of the boric acid transfer pumps is in the range of 90 to 105 psig and the discharge pressure gauge is measured with gauges that read from 150 to 300 psig with a minimum accuracy of *1%. It should be noted that if a 300 psig gauge is used with a boric acid transfer pump which typically has a discharge pressure value of 90 psig, then this gauge would also not be in compliance with the Code requirements. The licensee should ensure that this condition is not currently in existence at SQN.
The licensee has proposed to use auction pressure instrumentation with a range of 0 to 15 psig.
For a discharge pressure of 90 psig with the *1% accuracy of the 150 psig discharge pressure gauge, coupled with the installed suction pressure instrumentation of 0 to 15 psig for a 1.5 psig reference value, an estimate of the possible variance in the differential pressure reading for the boric acid transfer pumps can be made. When compared to the Code allowable variance with instruments with an accuracy of 12% that are three times the actual reference values, the actual variance is approximately 1.6 psig where the Code-allowed variance is 5.5 psig.
Therefore, the alternative provides an acceptable level of quality and safety because the actual variance is less than the Code-allowed variance.
The staff takes issue with the licensee's statement that boric acid transfer pump suction gauges that meet the requirements of the Code are "special," implying that the manufacturing and accuracy requirements are extraordinary and significantly more expensive than other analog instrumentation at the plant. In fact, the suction pressures for the boric acid transfer pumps at SON appear typical for other similar plants and do not present a unique burden. In addition, the i staff has stated that the installation and replacement of instruments is generally not an undue burden. The position referenced by the licensee in NUREG-1482, Section 5.5.1, applies to
< earlier licensed plants but not for the purchase of replacement instruments that can be procured i to meet the current Code requirements.
2.1.4 Conclusion The proposed alternative to the full-scale range instrument requirements of OM-6, Paragraph 4.6.1.2(a), for the boric acid transfer pumps is authorized pursuant to 10 CFR 50.55a (a)(3)(i) based on the alternative providing an acceptable level of quality and safety. The licensee
4 should ensure that the, discharge pressure instrumentation is also in accordance with the Code full-scale range requirements. This 'ssue may be reviewed in a future NRC inspeci!on.
2.2 Relief Reauest RP-05 The licensee has requested relief from the digital instrument range requirements of OM-6, Paragraph 4.6.1.2 (b). for the containment spray pumps. The licensee has proposed to calibrate the ultrasonic flow instrument such that the fixed reference value does not exceed 95% of the calibrated range.
2.2.1 Licensee's Basis for Reauestina Relief The licensee states:
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 digitalinstrumentation with a calibrated range of 7,064 gpm. A 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 the specifications 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 h a manufacturer's stated accuracy of.*1% and is calibrated to 12% of the calibrated range. This exceeds the OM-6 accuracy requirement of 12% 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 tFa 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 require a modification to the piping system with no appreciable increase in accuracy. An installed 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.
2.2.2 Altemate Testina The licensee proposes:
Calibrate ultrasonic flow instrumentation such that the reference value for flow (which is the set parameter) does not exceed 95% of the calibrated range.
2.2.3 Evaluation The Code requires that reference values for digital instruments not exceed 70% of the calibrated range of the instrument. As stated in the March 20,1996, SE, the basis for the 70%
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5 reference value requirement for digital instruments is not elaborated or, in the Code. However, an exception exists for vibration instrumentation. The staff authorized the licensee's proposed attemate testing provided that the licensee's flow reference value would not exceed 95% of the calibrated range of the instrument and that the flow rate was the fixed reference value. The staff requested that the licensee update this relief request within 1 year of the issuance of the safety evaluation.
The licensee's revised relief request was included in their submittal dated December 22,1997.
The licensee stated that the pump flow rate, which is the fixed reference value, will not exceed j 95% of the calibrated range of the digital flow meter. However, this flow meter is calibrated to an accuracy of *2%, not *1% as the licensee stated in their original submittal dated j November 21,1995. In a phone conversation with the licensee on June 22,1998, SQN staff stated that their original submittal was not sufficiently clear in this area. J I
The requirements of OM-6, Paragraph 4.6.1.2 (b), are similar to the requirements of the 1989 i Edition of ASME Section XI, Article IWA-5000, System Pressure Tests. Paragraph IWA- 1 5264(b) requires that digital pressure measuring instruments shall be selected such that the intended maximum test pressure shall not exceed 70% of the calibrated range of the q instrument. The apparent purpose of this requirement in Section XI for the inservice inspection
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hydrostatic test is that when the test pressure, which is introduced by an outside source, exceeds the intended test pressure, an assessment can be made of the material stresses on ]
the pressure retaining components that were exposed inadvertently to the excessive test pressure. The language of this Paragraph appears to have been carried over to OM-6 and
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applied to both pressure and ficw instrumentation.
In this specific case, the flow measurement is the fixed reference value. Therefore, provided l that the current reference value is not adjusted, the actual flow variation required to be j measured by the flow meter is between 93 and 97% of the calibrated range. In addition, '
system pressure is provided by an individual containment spray pump, not an external source. .
Therefore, the proposed attemate testing provides a reasonable assurance of operational readiness.
The licensee should use caution when implementing this alternate testing. For instance, if the testing procedure is changed such that the differential pressure is the fixed reference value and the flow measurement becomes the variable reference value, then the current instrumentation would not be acceptable because the flow acceptance criteria requires the variable to be measured to 110% of the variable reference value. This would be out of the calibrated range of the instrument. Similarly, if the flow measurement remains the fixed reference value, but because of pump maintenance the fixed reference value changes, this change should be assessed against the calibrated range of the instrument.
2.2.4 Conclusion The proposed attemative to the Code digitalinstrument range requirements for the containment j spray pump ultrasonic flow meter is authorized pursuant to 10 CFR 50.55a(a)(3)(i) based on the i altemative providing an acceptable level of quality and safety.
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3.0 .yALVE RELIEF REQUESTS 3.1 Rehef Request RV-5
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The licensee has requested relief from the valve obturator movement requirements of OM-10, Paragraph 4.3.2.4(a), for the essential raw cooling water check valves listed below. The l licensee has proposed to verify the open function by use of a smoke test and the closure function by the absence of flow in the vent line dyring the pump test.
The valves affected are:
67-719A 67-719B 67-720A 67-720B 67-739A 67-739B 67-740A 67-740B A description of the function of these valves provided by the licensee is included below: j 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
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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 i valves are locate.d 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. .
1 3.1.1 Licensee's Basis for Reauesting Rehef j
The licensee states:
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 r.ir flow rate this small will result in very inaccurate and unrepeatable results. Additionally,' the nature of the flow through these valvet is such i
' that it will not be at a steady state Jr.g enough to quantify. The flow will rapidly <
accelerate to a maximum and then steadily decrease as the driving force of the water column level above the river elevation decreases. The use of a smoke test is a 9,vthod that provides positive means to determine that the valve goes to the open position. The 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 coNmn ensures that unnecessary starting torque is not placed on the pump motor during motor / pump staris.
Disassembly of the valves every refueling outage is not a practical means of verifying valve performance since ERCW [ Essential Raw Cooling Water) pump maintenance is not tied to refueling outage evolutions. SQN has eight ERCW pumps that can be taken
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out of service at any time. The smoke test provides a positive means of the valves opening function and can be performed quarterly as required by the code.
Consequently, the smoke test demonstrates that the valve safety function is fulfilled.
3.1.2 Altemate Testina
'ihe licensee proposes:
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 demonstrated during each pump test.
The licensee stated that this testing will be performed once per quarter.
3.1.3 Evaluation These check valves have a safety function to close to prevent the diversion of flow from the ERCW system. In addition, these check valves have an open functien to protect the pump motor from excessive starting torque during a purnp start. The Code requires that valves with
. an active safety function be exercised quarterly, Paragraph 4.3.2.4(a) of the Code also states that 'other positive means" may be used to verify the necessary check valve obturator movement.
The licensee testing to verify closure was not sufficiently described in their submittal for evaluation. In a phone conversation on June 22,1998, the licensee stated that closure of each check valve was verified during the pump start of each inservice test. Observable flow through the vent line would he an indication that the check valve associated with that particular ERCW pump did not close. The licensee also stated that this testing was conducted at conditions which would be exp ected when these pumps were required to perform their safety related function. This acceptance criterion constitutes another positive means to verify check valve closure and is, the.efore, acceptable. The smoke test also provides an acceptable indication that the check vr;ves have moved to their open position. Therefore, the proposed attemative provides an af,ceptable level of quality and safety.
3.1.4 Conclusion The proposed alternative to the Code valve otJurstor movement requirements of OM-10,
! Paragraph 4.3.2.4(a), for the ERCW check valves is .nuthonzed pursuant to 10 CFR
- 50.55a(s)(3)(i) based on the alternative providing an .icceptable tevel of quality and safety.
1 3.2 Releef Request RV-06
' The licensee has requested relief from the power-operated valve stroke time requirements of OM-10, Paragraph 4.2.1.4, for the reactor vessel head vent valves ESV-68-396 and
8 FSV-68-397. The licensee has proposed to verify that the valve operates properly through the use of acoustic monitoring.
3.2.1 Licensee's Pasis for Reauestina Relief The licensee states:
These valves are one-inch Target Rock solenoid valves that have no position indication and are totally enclosed (seat welded bonnet) which prevents visual confirmation of valva 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 disassernbly 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.
Secondly, once the valve is opened and the internals of the valve are examined, the condition of the internal parts would not provide any additionalindication of acceptable valve operation than the acoustic monitoring. TVA believes there is no feasible method for measuring the stroke time.
In addition, 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.
3.2.2 Aiternate Testina_
The licensee proposes:
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 valve. The valve is opened by operating the thumbwheel controller and another accustic signal is obtained at the full-open position. The valve is then closed and another acoustic signal is 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.
The licenseo stated in their submittal that acoustic monitoring would be performed on each valve every refueling outage.
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3." .3 Evaluation The Code requires that power-operated valves be stroke timed either in the open direction, the closed direction, or both directions depending on the required safety function of the valve. The ;
reactor vessel head vent valves have a safety function in the open direction to vent non-condensable gases from the reactor vessel and a safety function to close to prevent flow .
diversion from the reactor coolant system. As the licensee stated, these valves are rapid acting solenoid valves which stroke in less than 2 seconds and have no position indication. Stroke timing these valves using conventional methods would be a hardship without compensating I increase in safety because tiie valves would have to be replaced.
In the staff's SE of March 26,1996, this relief request was denied because the acoustic monitoring method was not described clearly and the means for monitoring degradation in the l valve could not be evaluated. NUREG-1482, Section 4.2.8, recommends that the actual testing technique evaluate actual disk movement and not just movement of the pilot valve or valve stem. The licensee stated in their submittal of December 22,1997, that their method of acoustic monitoring conformed with the guidance provided in Section 4.2.8. The licensee's i testing methodology was further discussed in a phone conversation on June 22,1998. The !
licensee stated that acoustic flow noise is monitored before and after the valve stroke. Then the traces are compared to ensure that the valve has changed position. Monitoring system flow noise by acoustic methods would demonstrate that the valve disk has changed position.
Therefore, the proposed alternative test method provides an acceptable level of quality and safety.
3.2.4 Conclusion The proposed alternative to the Code power-operated valve stroke time requirements of OM-10, Paragraph 4.2.1.4, for the reactor vessel head vent valve .. dSV-68-396 and FSV-68-397 is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) based on the Determination that compliance with
- the specified requirements results in a hardship without a compensating increase in the level of quality and safety. This issue may be reviewed in a future NRC inspection. !
3.3 Relief Reauest RV-7 The licensee has requested relief from the exercise requirements of OM 10, Paragraph 4.2.1.2, for manual valve 62-546 in the chemical and volume control system. The licensee has proposed no altemate testing for this valve.
3.3.1 Licensee's Basis for Reauestina Relief The licensee states:
Opening this manual valve bypass around the seal injection filters allows the introduction of impurities into the RCP [ reactor coolant pump) seals which, in the past, has caused damage to the pump seals and resulted in increased sealleakage. Cycling L___ -
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10 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 be isolated during normal operation, thereby, isolating 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.
Manual Valve 62-546 is part of a TVA Class B ASME Class 2 closed system outside of containment. During normal reactor operation, RCP seal injection is provided through one of two sealinjection filters, which are in parallel along with the sealinjection 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 are inboard check valves and outboard manualisolation 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
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injection linec, the outboard valves are not capable of remote operation and receive no isolation signal for automatic closure. The four manual valves at the cc,ntainment penetration are not accessible 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 [ General Design Criterion) 55 design requirements for containment isolation as one of the outboard containment isolation valves, which are capable of being closed quickly to isolate the sealinjection 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 injection 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 / seal flow as required. Therefore, a water seal will be provided at a pressure greater than 1.1 P, with at least a 30-day water supply to preclude air leakage out of containment. These redundant isolation provisions (i.e., the inboard check valves, the .:fosed system, the water seal, and the seal injection filter isolation valves) provide assurance that no single failure could result in release of containment atmosphere to the environment. In accordance with 10CFR50 Appendix J, Paragraph Ill.C.3, the seal injection penetrations are not a potential containment atmosphere leak pam 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 [by] 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. Therefore, the risk associated with testing this valve does not provide a commensurate increase in safety.
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4 11 3.3.2 Altemate Testina The licensee has proposed no attemate testing.
3.3.3 Evaluation The code requires that active valves be exercised to their safety function every three months.
Exercising can be deferred to cold shutdowns or refueling outages without requesting relief from the NRC if the testing requirements during power operations or cold shutdowns are ,
impractical to perform. The Code does not have provisions to exempt testing of active j
- components.' Valves which maintain their obturator position and are not required to change !
obturator position to perform their safety function (defined in the Code as passive valves) do not - )
require exercise testing. '
The licensee has stated that the manual reactor coolant pump seal injector bypass valve has an active safety function. Therefore, the Code requires that it be exercised. Although the licensee has obtained an exemption from performing leak testing of this containment isolation valve, this exemption does not apply to verifying that the valve can perform its active safety function. The only method to demonstrate this valve's active safety function is by performing an exercise test.
Therefore, proposing no attemative testing for this valve is unacceptable.
The staff recognizes that there are significant consequences to stroking this valvw in certain 3
plant operational modes. Such exercising may introduce contamination to the reactor coolant pump seals which could s lgnificantly affect the operation of the plant. However, degradation of an active manual valve cannot be determined without performing an exercise test, The licensee should make an assessment of the practicability of testing during power operations
- and cold shutdowns, in addition, the licensee should determine what plant modes will allow exercising this valve in a practicable manner. If certain plant modes are determined to exist that may not be in accordance with the Code requirements (i.e., maintenance which leaves this line depressurized at an interval greater than refueling outages) the licensee should submit a new relief request. The licensee may also elect to reexamine whether these valves are in fact
- passive valves Instead of active valves.
~ 3.3.4 Conclusion Relief is denied. The licensee should either perform the required Code teaung or consider the impracticality of testing and the scope requirements of manual RCP sesi injector bypass valve 62-546 and revise their IST program or submit a new relief request as necessary.
3.4 Relef Request RV-8
~ The licensee has requested relief from the exercise and valve obturator movement requirements of OM-10, Paragraphs 4.3.2.2 and 4.3.2.4 respectively, for containment spray L . check valves 72-547,72-548,72-555, and 72-556. The licensee has proposed to test these
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, valves at least once'every 10 years in conjunction with Technical Specification (TS) suweillance requirement (SR) 4.6.2.1.1.d which is the containment spray nozzle verification test.
12 3.4. Licensee's Basis for Recuestino Relit i The Licensee states:
i The containment spray system is used during accident conditions to limit containment pressure by spraying cooler water into the containment vessel atmosphere. This spraying action reduces the temperature and pressure inside the containment vessel.
To perform the pressure reduction function,' containment spray system piping continues from the containment spray check valves, that are inside the containment vessel, to the open spray nozzles on the spray headers. Because the spray nozzles are open, it is not possible to perform flow tests on the check valves during any mode of plant operation.
Flow testing would result in a spray of borated water into the containment vessel atmosphere and wetting of equipment in the building. Therefore, TVA has applied the code-accepted, altemate testing method of disassembly and inspection of these valves.
1 Temporary scaffolding is used to perform check valve disassembly ar d inspection of !
check valve internal components. These check valves are located near the containment spray and the residual heat removal spray headers in the top of the reactor building's f
1 containment vessel. No permanent access to the valves exists. The valves are reached by construction of scaffolding approximately 30-foot high. The scaffolding is installed on top of and secured to the polar crane. Scaffold erection and use is typically a challenge to industrial safety. Scaffold erection, use, and disassembly forinspection of the containment spray check valves is a significant industrial safety hazard.
The check valves are either 8 inch or 12 inch,150 pound per square inch gauge rated f valves, manufactured by Aloyco. The check valves and associated piping are )
constr ucted of stainless steel material; therefore, no degradation to valve internals is expected. A review of past test instruction performances indicated that there have been no failures or any indication of adverse problems, j 3.4.1. Additional Justification for the Grantino of Relief Provided by the Licensee l The licensee states: ,
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The valves are required for the purpose of containment isolation. The check valves are not local leak-rate tested as required by Definition H, Criterion 3, in 10 CFR 50, Appendix J, where the system is required to operate intermittently under oostaccident conditions. These spray lines are water sealed and are not potential ccatainment atmosphere leak paths A water leg is maintained in each riser between the closed outboard containment isolation valves and the spray header. These outboard valves are leak-rate tested with water to ensure a 30-day inventory as required by TS SR 4.6.12.b. NRC found this acceptable as documented in NUREG-1232., Volume 2, Section 3.6. TVA submitted ar; exemption to Appendix J to rely on the remote manual valve outside containment, the seal water system, and the closed containment spray system outside containment as the basis for not leak-rate testing the inboard check valves to Appendix,! requirements. NRC approved this exemption on September 22, i 1988. I I
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i 13 A review of past test instruction performances for valve disassembly and inspection indicated that there have been no failures or any indication of adverse problems.
Disassembly and inspection of these check valves does not provide any additional information to address degradation. A review of industry experience us,ing the manufacturer for a key word search indicates there are no industry problems with valves of the same manufacturer. A review of EPRI's [ Electric Power Research Institute's]
" Application Guidelines for Check Valves in Nuclear Plants," yielded no additional concerns for this type of valve application. The SQN TS SR for testing the spray header and nozzles for unobstructed flow paths is consistent with the standard TS. The SQN TS SR requires air or smoke flow testing every 10 years for verification of unobstructed spray nozzles. This indicates that NRC understands the low risk for degradation in this section of the system.
3.4.2 Alternate Testina The licensee proposes:
Containment Spray Check Valves 72-547, 72-548, 72-555, and 72-556 will be demonstrated operable in conjunction with the spray nozzle verification test required by TS SR 4.6.2.1.1.d.
In demonstrating the containment spray system to be operable TS SR 4.6.2.1.l.d requires "At least once per 10 years verify each spray nozzle is unobstructed." To perform this test, the water leg is drained between the inboard and outboard containment isolation valves. An air compressor, with heated air capability, is connected to the system upstream of the check valves. Hot air is blown through f.e riser piping, the inboard containment check valves, and the header nozzles. A thermographic camera is used to verify unobstructed flow through the nozzles. Flow through the nozzles would also verify flow through the check valves.
3.4.3 Evaluation ,
These check valves have an active safety function to open to facihtate flow from the containment spray pumps to the containment spray headers. The Code requires that valves with an active safety function be exercised quarterly. If exercising these check valves quarterly is impractical, the Code has provisions which allow this exercise requirement to be satisfied either at cold shutdown or refueling outages. To demonstrate the necesary obturatcr movement , each check valve must open to the position required to fulfill its safety function.
Gener'c Letter 89-04, Position 1, states that the full-stroke open position of a check valve may
. be verified by passing the maximum required accident condition flow through the valve. Any fuw rate less than this would be considered a partial-stroke exercise. Paragraph 4.3.2.4(a) of the Code also states that "other positive means" may be used to verify the necessary check valve obturator movement.
Paragraph 4.3.2.4(c) of the Code allows licensees to use disassembly and inspection of each check valve every refueling outage to demonstrate the necessary valve obturator movement.
14 Position 2 of GL 89-04 allows licensees to group check valves (with the same manufacturer, mr I, size, and service conditions) and stagger their inspection frequency when it is a burden to espect all valves in the group every refueling outage. Position 2 also allows, in cases of extreme hardship, the licensee to defr testing to every other refueling outage. One example which may be considered as extre i s hardship is offloading the reactor core to perform the disassembly of a check valve. In general, extreme hardship is associated with a unique burden which exists at a specific plant that results in the disassembly at the frequency specified in Position 2 cverly burdensome. '
There are no means to test these check valves with flow without introducing a spray of water inside the containment. The licensee is currently disassembling and inspecting the containment spray check valves in accordance with Position 2. The licensee has stated that inspection of ;
these check valves requires the construction of temporary scaffolding which is an industrial I hazard. While it is recognized that improper erection of scaffolding can create a work hazard, this activity is common at pressurized water reactor plants and does not present a unique burden to disassemble these check valves.
The licensee has proposed to perform the necessary Code testing by taking credit for TS SR 4.6.2.1.1.d, which is the containment spray nozzle verification test. This test is conducted once
. every 10 years to ensure that the spray nozzles are unobstructed. The spray nozzle test may constitute a partial-stroke exercise because the check valves would have to open to allow the smoke to pass to the nozzles. However, this test would not constitute a full-stroke exercise of the check valves because the position of the check valve obturator would not be known since the licensee does not perform non-intrusive testing of these check valves or has indicated that these check valves are equipped with local position indication to determine obturator position.
In addition, the smoke test does not constitute other positive means as described in the Code because this is not an activity that occurs during any mode of plant operation where system response could be measured (i.e., pressure differential across a valve). With regards to the exemption of leakar.,e testing for these valves, this testing would not contribute any information to the determination of the ability of the check valve to perform its active safety function.
Finally, the 10-year test frequency is significantly longer than the Code required testing frequency. Therefore, the proposed testing is determined not to be an acceptable attemative to the Code requirements.
The licensee's basis for requesting relief includes information on maintenance history and industry experience with this particular valve. Although this information does not support the proposed attemate testing, part of this information may support an attemative test to use the
' 1996 Addenda to the 1995 Edition of Code Subsection ISTC, Appendix 11, " Check Valve Condition Monitoring Program." One such proposed attemative was granted to the Wolf Creek Operating Corporation, licensee for the Wolf Creek Generating Station, in a safety evaluation dated Novamber 26,1997. The staff approval of this attemptive contained a number of conditions which are described in the referenced SE.
1 i
3.4.4 Conclusion Relief is denied. The licensee should continue testing in accordance with GL 89-04, Por.ition 2, as described in Sequoyah Relief Request RV-1.
J.
l 15 4.0 ACTION ITEMS 4.1 Second 10-Year interval Start Date The licensee has not provided what they believe is their exact second 10-year interval start date in either of their submittals as requested by the staff. This issue is still pending as it was not addressed by the licensee in either submittal.
4.2 Relief Reauest RP-03 it appears that the potential exists for 300 psig discharge pressure gauge to be used
' with a boric acid transfer pump which typically has a discharge pressure of 90 psig. j This gauge would not be in compliance with the Code requirements. The licensee '
should ensure that this condition is not currently in existence at Sequoyah.
4.3 Relief Reauest RV-7 Relief Request RV-7 was denied to exempt from exercise testing of manual reactor coolant pump seal injector bypass valve 62-546 because the licensee proposed no alternative testing for this valve. The licensee should either perform the required Code testing or consider the impracticality of testing and the scope requirements of this valve and revise their IST program or submit a new relief request as necessary.
4.4 Relief Reauest RV-8 Relief Request RV-8 was denied to test the containment spray check valves 72-547, 72-548,72-555, and 72-556 at least once every 10 years in conjunction with TS SR 4.6.2.1.1.d, which is the containment spray nozzle verification test, because the proposed testing was not an meeptable alternative to the Code requirements The licensee should continue testag in accordance with GL 89-04, Position 2, as described in SQN Relief Request RV-1.
5.0 CONCLUSION
j The staff concludes that the relief requests as evaluated by this SE provide reasonable ;
assurance of operational readiness of the pumps and valves in question to perform their safety- 1' related functions. The staff has determin&d that grsnting of relief requests and authorizing
. alternatives (RP-05, RP-07, RV-05, and RV-06) pursuant to 10 CFR 50Ea (f)(6)(i), (a)(3)(i), or
'(a)(3)(ii) is authorized by law and will not endanger life or property, or the corr, mon defense and security and is otnerwise in the public interest. In making this determination, the staff has ,
considered the impracticality of performing the required testing and the burden en the licensee 1 if the requirements were imposed. Relief Requests RV-07 and RV-08 were denied as
- discussed above.
Principal Reviewer: J. Colaccino l
Date: August 07,'1998 1
Mr. J. A. Scalice SEQUOYAH NUCLEAR PLANT Tennessee Valley Authority cc:
Senior Vice President Mr. Pedro Salas, Manager Nuclear Operations Licensing and Industry Affairs Tennessee Valley Authority Sequoyah Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Soddy Daisy, TN 37379 Mr. Jack A. Bailey Mr. J. T. Herron, Plant Manager Vice President Sequoyah Nuclear Plant Engineering & Technical Services Tennessee Valley Authority Tennessee Valley Authority P.O. Box 2000 6A Lookout Place Soddy Daisy, TN 37379 1101 Market Street Chattanooga, TN 37402-2801 Regional Administrator U.S. Nuclear i .agulatory Commission Mr. Masoud Bajestani Region 11 Site Vice President 61 Forsyth Street, SW Sequoyah Nuclear Plant Suite 23T85 Tennessee Valley Authority Atlanta, GA 30303-3415 P,0. Box 2000 Soddy Daisy, TN 37379 Mr. Melvin C. Shannon Senior Resident inspector General Counsel Sequoyah Nuclear Plant Tennessee Valley Authority U.S. Nuclear Regulatory Commission ET 10H 2600 Igou Ferry Road 400 West Summit Hill Drive Soddy Daisy, TN 37370 t Knoxville, TN 37902 Mr. Michael H. Mobley, Director Mr. Raul R. Baron, General Manager TN Dept. of Environment & Conservation Nuclear Assurance Division of Radiological Hedth ;
Tennessee Valley Authority 3rd Floor, L and C Anr>ex !
SM Lookout Place 401 Church Street l 1101 Market Street Nashville, TN 37243-1532 Chattanooga, TN 37402-2801 l 1
County Executive !
Mr. Mark J. Burzynski Manager Hamilton County Courthouse Nuclear Licensing Chattanooga, TN 37402-2801 ,
Tennessee Valley Authority !
4X Blue Ridge 1101 Market Street l Chattanooga, TN 37402-2801 l
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