ML20117J656

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Forwards Response to NRC Questions Re Code Relief Requests for Third 10-Year Interval for IST of ASME Code Class 1,2 & 3 Pumps & Valves
ML20117J656
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 05/23/1996
From: James Knubel
GENERAL PUBLIC UTILITIES CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
6710-96-2136, NUDOCS 9606030057
Download: ML20117J656 (27)


Text

l GPU Nuclear Corporation Nuclear s - 44, s

  • P.O. Box 480 Middletown, Pennsylvania 17057-0480 (717) 944-7621 Writer's Direct Dial Number:

May 23,1996

. 6710 2136 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1) I Operating License No. DPR-50 )

Docket No. 50-289 l Inservice Testing of ASME Code Class 1,2, and 3 Pumps and Valves (IST) - Response to NRC Questions Regarding Code Relief far the Third Ten Year Interval GPU Nuclear submitted the TMI-1 IST program description for the third 10-year IST interval on September 21,1995. Included were requests for relief from Code requireraents which includa 1) Code requirements that have been determined to be impractical for 7MI-l in accordance with 10 CFR 50.55a(f)(5)(iii) and 2) alternatives proposed in accordece with I (50.55a(a)(3).

The NRC's comments on our relief requests were discussed in several conference calls which included Mr. Joseph Colaccino of NRR. The attachment includes a revision or restatement of the original relief requests followed by the NRC comment and the GPU Nuclear response.

Revised test appears in bold; deleted text appears with a line through it.

Sincerely, o

JIKnubel Vice President and Director, TMI Enclosure cc: Administrator, Region I TMI-1 Senior Project Manager

c. TMI Senior Resident Inspector 9606030057 960523 l PDR ADOCK 05000289 P PDR I

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6710-96-2136 Attachment Page 1 of 24 RELIEF REQUEST NO. P 1 (REVISED)

Tae No. Component Typ_c MU-PI A Makeup & Purification Pump "A" Centrifugal MU-PIB Makeup & Purification Pump "B" Centrifugal MU-Plc Makeup & Purification Pump "C" Centrifugal Code Section from Which Relief is Requested

1. OM-6,15.6, " Duration of Tests," requirement for a run time of at least two minutes after reaching stable pump conditions before obtaining data, and
2. OM-6, {5.2, " Test Procedure," requirement for testing at a single reference point.
3. C'" 5, "4.5.4, v!S= ! r =cc;;.. ....;;; nq.:.c.x;;;. ,

1 A_Iternate Test Description As permitted by NRC GL 89-04, Position No. 9, the pumps will be full flow tested each refueling outage (see justification). The refueling outage test wil8 include measurement of stable flow rate, differential pressure, and vibration. Pump testing will be performed with the system lined up to pump to the RCS through different flow path combinations to provide pump data at various flowrates.

Run time through each flow configuration may be less than the two minutes required by OM-6. Due te ic shcr: duratica cf ic;:ing, a be;; ;ffen wi!! be u;;d ic tak: "ibration data =d a!! poi._ ... ,7 .~.

be obtained.

Basis for Relief Request The amount of time that the Makeup Pump injects at full flow to the RCS must be limited. Pumping to the RCS will raise pressurizer level and a plant transient can occur. Run time therefore must be limited.

Pumping time is limited to a total of approximately 5 min for all flow configurations. Because of the short time available for a test run, throttling to a specific reference point can not be accomplished. l i

The pump is run with several different valve lineups to verify that flowrate and head are equal to or higher than accident design requirements. Flow rate and pressure measurements for each lineup is  !

compared with previous test data. Acceptance is based on meeting or exceeding accident flow and head requirements. This meets the intent of the code. The test is similar to that described in NUREG 1482, 65.2 except for the following:

1) A manufacturers curve is not used. Comparison is with the FSAR Safety Analysis curve and I previous full flow tests. II
2) A five point curve is not used. The pump will operate at several different points 7-end
3) Vibratica i :aken during4hc quanctly tes =d :c i; extent :irn; a!!cws during de fu!! f!cw test.

These tests demonstrate pump operability and meet the intent of the code.

6710-96-2136 Attachment Page 2 of 24 NRC Comments on Relief Request P1 and GPU Nuclear Response:

NRC Comment 1:

Has there been any maintenance performed on any of the three pumps?

GPU Nuclear Response:

In addition to routine maintenance, GPU Nuclear has installed balance discs on the inboard and outboard bearing of each pump which has improved the vibration levels.

1 NRC Comment 2:

Has the FSAR curve been validated?

GPU Nuclear Response:

Yes the FSAR curves have been validated.

NRC Comment 3:

All vibration data, including data in each direction during full flow testing, must be collected. i Licensee's proposed testing is unacceptable.

GPU Nuclear Response:

GPU Nuclear has decided to withdraw the request for relief from the requirements to take all of the vibration data.

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. l 6710-96-2136 j Attachment l Page 3 of 24 j RELIEF REQUEST NO. P 2 (REVISED)  !

Tan No. Component Iyp_q l EF-P1 Turbine Driven EF Pump Centrifugal EF-P2A Motor Driven EF Pump "A" Centrifugal EF-P2B Motor Driven EF Pump "B" Centrifugal l l

I Code Section from Which Relief is Reauested I i

Regarding refueling interval tests, relief is requested from: l

1. OM-6, 65.6, " Duration of Tests," requirement for a run time of at least 2 minutes after reaching stable pump conditions before obtaining data, l
2. OM-6, 65.2.c, requirement to compare flow rate and pressure to their respective reference l valuesvand 1 l
3. OM 6, i1.6.d.a, requiremen: :c : ke vib=:ica ==.;uremer.:: cn =ch acecnib!: pump h =ing.

Alternate Test Descriotion TMI-l Tech Specs requires a test each refueling to demonstrate EFW pumps can pump water from the Condensate Storage Tanks (CSTs) to the Once Through Steam Generators (OTSGs). Exeept-fer

=:: :v=y 3rd = fueling cutage, $c pump wi!! be =opped = seen = accident d=ign Scw=:: is achieved.

Evry 3rd =fu !ing (n=:ing !!R in 9/95), $c :est-wn! demc=:=: ful! Scw fc =ch pump b=cd on a =frence value. During $c fe!! Scw :=t, $c pump wi!! be Mcpped b=cd en OTSO ! vel when 1000 ga!!c= h= been :==fered. Run time while pumping to the OTSG in !$= := may be less than two minutes as required by OM-6. B=: ffer: wi!! be made :c take vib=:!ca data for each pump be=ing, but be==: cf $; ;hc-: dum:icn of $c =fue!!ng :=:, .: =y act be pczib!;:0 obtain 1 a!! dea.

During the next refueling outage (12R), GPU Nuclear will verify the pump curve to be the valid accident design curve in accordance with the guidance in NUREG-1482, 65.2, except that only three points on the curve will be taken. Thereafter, each refueling outage, a full flow test of all three pumps will be performed eve j 3.-d =f;;Sg cc:cg: in which accident design flow and differential pressure for each pump will be verified by running the pumps only long enough to take stable flow, differential pressure, and vibration data; pressure and flow will be compared to the reference curve. This test verifies acceptable flow rate and differential pressure of the pumps.

In between full Scv- :=::, $c refue!!ng int =va! :=: w!!! demc=:=: accid =: d=ign Scw=::.

Basis for Relief Reauest The EFW pumps are only used for emergency operation. They are not used for startup, shutdown, or normal plant operation. EFW flow to the OTSGs is limited by the cavitating venturis. Since the pumps operate only for test, no significant degradation is expected.

Since the refueling interval tests transfer lower quality water to the OTSGs, the number and duration of tests must be limited to minimize routine exposure of the OTSGs to lower quality water.

Minimizing test duration is necessary to limit the amo.mt of water injected into the OTSGs where corrosion damage promoted by O 2 can occur. Throttling to a reference value of flow / differential pressure as well as waiting out a minimum of two minutes run time lengthens the amount of time the EFW pumps are running while pumping oxygenated water into the OTSGs.

. l 6710-96-2136 Attachment Page 4 of 24 Justification for our request is based on limiting the amount of oxygenated water that is pumped into the OTSGs to minimize the potential for steam generator tube degradation. The feedwater ,

oxygen concentration limit for normal plant operation is 5 ppb; and TMI-1 normally maintains )

feedwater oxygen concentration less than 1 ppb. At the lower temperatures during shutdown, hydrazine is unable to react and scavage the oxygen. So to help lower the oxygen content, prior to the test a nitrogen blanket is applied to the CSTs through spargers. We can typically lower l the oxygen concentration of the water to approximately 200 - 300 ppb. This range is still many I times higher than the limit recommended by The Electric Power Research Institute (EPRI)'. I We believe that performing a shorter test will not compromise our ability to adequately j demonstrate EFW pump operability and meets the intent of the Code.

Becc.u;; cf the shcr: tes: duratica, it n: y not be pczib!: te obtain " ibratica da:a.

EFW pump quarterly tests verify the pumps are operational, start on demand, and generate the required discharge pressure. During the quarterly test, vibration data will be taken on each bearing while pumping through the recirculation line.

NRC Comments on Relief Request P2 and GPU Nuclear Response:

NRC Comment 1:

Test frequency and methodology are unacceptable as proposed. No basis for not full flow l testing each pump every refueling outage. Unsure of how other B&W plants perform this I testing. Perhaps licensee should make inquiries.

GPU Nuclear Response:

GPU Nuclear contacted the other B&W plants and found that others, some of which have full flow test loops, are performing a refueling interval test as required by the code. W, are withdrawing our request to take less vibration data than that required by the code. This l reduces the extent of the relief requested to the issue of run time and throttling to a reference value. We are requesting relief to limit the quantity of oxygenated water that is pumped into the OTSGs. This will minimize the potential for steam generator tube degradation.

NRC Comment 2:

All vibration data must be collected during pump testing.

GPU Nuclear Response:

GPU Nuclear withdraws the portion of this request regarding relief from the requirements to take all of the vibration data.

' EPRI TR-102134, Final Report, "PWR Secondary Water Chemistry Guidelines -

Revision 3," May 1993.

6710-96-2136 Attachment Page 5 of 24 RELIEF REQUEST NO. P 3 (REVISED)

Tn No. Comoonent Tygg NR-PI A Nuclear Service River Water Pump "A" Centrifugal NR-PIB Nuclear Service River Water Pump "B" Centrifugal NR-PIC Nuclear Service River Water Pump "C" Centrifugal Code Section from Which Relief is Reauested Relief is requested from OM-6, 55.2, " Test Procedure," item (d) regarding the determination of flow l rate.

Alternate Test Description Flowrate for individual pumps will be measured at refueling outages.

Basis for Relief Reauelt The test flow instrumentation for this system is located in the common discharge from all three pumps. l The piping configuration does not facilitate installation of individual pump flow measuring devices.. l GPU Nuclear has not been successful in attaining acceptable accuracy or repeatability using individual annubar flow instruments.

To read total NR pump now (NR-FI-290), for any pump combination including two pump operation, it is necessary to isc!9te makeup water to the circulating water nume by closing the 30" butterfly valves (NR-V4A and NR-VS). This directs all NR Pump now to the Nuclear Service IIeat Exchangers and reduces the temperature of Nuclear Services Closed Coolihg Water which cools the Reactor Coolant Pump (RCP) seal return coolers. This results in a decrease in Makeup System supply water temperature including the supply for Reactor Coolant Pump Seal water.

The resulting nuctuations in Reactor Coolant Pump Seal leakoff now rate reduces the performance of the pump seals and adds to the risk of eventual RCP seal damage.

During normal plant operation, at least two of the three pumps are in operation. Operation of only one Nuclear Service River Water Pump is not a!!cwed becar.: of re!!abili:y concern; and could jeopardize plant equipment due to system heat loads for-a-large p=t of the y = If all but one NR Pump were secured for the purpose of testing, this would cause significant temperature variations in safety related components, add significant operator burden to assure adequate cooling of the many plant components that would be affected, and result in some operational risk.

Individual NR Pump flow rate measurement is impractical during plant operation or during Cold Shutdowns of short duration. The quarterly test to measure differential pressure and vibration as

, well as the refueling test to verify head and now rate greater than the accident design will continue to assure operability of the NR Pumps.

NRC Comments on Relief Request P3 and GPU Nuclear Response:

NRC Comment:

There is not enough information to evaluate this relief request. The licensee has not addressed the effect of measuring flow rates for two pumps will have on their Code acceptance criteria,

, GPU Nuclear Response:

Response to this comment is provided in the revised text of the relief request.

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6710-96-2136  :

Attachment Page 6 of 24 l

RELIEF REQUEST NO. P 4 (NEW RELIEF REQUEST 2)

Tan No. Comoonent Iyg

DH-PIB Decay Heat Pump "B" . Centrifugal Code Section from which Relief is Reauested Relief is requested from OM-6, 66.1, " Acceptance Criteria" which requires doubling the test frequency until the cause is determined and corrected for vibration readings in the alert range (0.325 ips).  ;

t Alternate Test -

The alert range for DH-PIB will be raised to vibrations greater than 0.400 ips (vs greater than 0.325 ips)in the vertical direction. The alert range for the horizontal direction will remain at vibration levels greater than 0.325 ips.

Basis for Relie(

OM-6 requires doubling test frequency when the overall vibration amplitude is greater than 0.325 ips during quarterly testing. The Code assumes that the equipment has degraded to the point where more i frequent monitoring and possibly a repair are warranted. There is no consideration for test conditions, '

vibration history, or equipment maintenance history. These pumps are tested each refueling at both medium and high flow rates where the vibration levels have always been lower with the majority of ,

vibration occurring at the vane pass frequency.

Consideration of vibration amplitudes was not part of the original acceptance criteria for many of the pumps procured for earlier nuclear plants. As a result, some pumps purchased in the late 1960's and early 1970's had inherently high vibrations. During low flow conditions, typical of IST testing, vibration amplitudes are at their highest. TMI plant data, shop testing of the spare pump by GPUN, and conversations with several pump vendors indicate that it is not unusual to experience vibrations in excess of 0.325 ips with the older pumps, especially at low flow conditions. Provided there is a successful long term operating history and provided there is no significant change in vibration amplitude or spectra, there is no reason to suspect equipment degradation at these vibration levels.

TMI's DHR pumps are one specific example and show the type of evaluations that we perform for pumps that exceed the alert limit.

TMI's DHR pumps are early edition API 610 process pumps. .They have operated with occasionally high vibration since 1974. This includes extensive operating time between 1979 and 1985 (TMI's extended shutdown which lasted approximately 61/2 years). The pumps have not failed, there is no unusual degradation in hydraulic performance, and seal and bearing life are normal. Vibration amplitudes average 0.293 ips (standard deviation of 0.100) with the highest vibration occurring at the lower flow IST conditions. Because of normal variation in vibration response and measurement, measured vibration exceeds 0.325 ips about once per year during low flow IST operation. However, there is no upward trend in the data and the greatest majority of vibration has always been at vane pass frequency.

GPU has discussed these relatively high vibration readings with several vendors who manufactured API pumps. The vendors stated that high vibrations are expected with early edition API 610 pumps, 2

This new specific relief request has been added as discussed in the response to NRC comments on generic Relief Request No. PG 1 which has been withdrawn.

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6710-96-2136 Attachment Page 7 of 24 particularly at the low flow rates encountered during inservice testing. Additionally, GPU Nuclear has shop tested TMI's spare DH Pump and found the vibration readings were almost identical to TMI's j two inservice pumps. During the shop test, vibration data were recorded at many different flow rates.

At flow rates equal to and below the IST flow rate, vibrations occasionally exceeded 0.325 ips. This pump was inspected prior to and after the shop test to assure no degradation had occurred. The spare  ;

pump is the same model number; it was purchased around the same time as the inservice pumps and l has never been used.

Figures 1 and 2 show vibration data from test < of the DH Pumps (DH-PI A and DH-PIB, respectively) since 1987. Points shown represent the highest vertical overall amplitude since amplitudes in the vertical direction are higher than the horizontal vibration amplitudes. The vibration spectrum is essentially all at vane pass frequency with no IX or 2X harmonics that could indicate pump problems. l These data present a clear case for raising the allowable vibration levels for DH-PIB. Several attempts-have been made to reduce the vibration levels for the Decay Heat Pumps, including pump motor alignment, strengthening the backfoot, and adding lead weights to the back end of the pump.

Based on the successful operating history of DH-PIB, no step changes or trends in vibration data as shown on Figure 1, extensive vibration analysis, shop testing, and vendor input, GPU Nuclear does not consider the vibration amplitudes of DH-PIB unacceptable. Replacing or modifying the pumps to reduce vibrations only to assure they do not occasionally exceed 0.325 ips could be unnecessary.

Further, doubling the test frequency would result in running the pumps more often at low flow conditions and would provide no useful information. Therefore, this request to allow raising the alert range to greater than 0.400 ips (vs greater than 0.325 ips) is justified for DH-PIB.

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6710-96-2136 Attachment i Page 9 of 24 eib=:ica :=!yi, :. hep : :ing, =d v= der input, CPU Nu !=r dc= no: :c=ider $c vib=:ica n.p!hud= cf TM!*: cpem ing DH pump; u=ce:ptab!;. Rep! ing er modifying 10 pump; ic redue vib=:ic= cn!y :c =;ur $cy do nc cceric=?!y acced 0.325 ip; ;culd be u= :==ry. Further, doub!!ng Se t=: frequency =cu!d :=u!: in ru= ng 1 pump: more often :: !c: Oc /high vibra:ien ,

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NRC Comments on Reliet Wquest PG1 and GPU Nuclear Response:

i NRC CoCunent:

Relief requests to raise the vibration alert range chould be done on a pump necific basis. The 4

relief request (s) will be evaluated for each indisidual pump. Information on pump vibration

history (including, pump, specific bearing, vibration direction and dates) can be provided in either raw or graphical form. A discussion of the investigat!on into the high pump bearing vibration j levels, including any discussion with the manufacturer and any actions taken, should be included.

. 1 A number of these relief requests have been granted. Typically, the alert range will be increased 1

in the specific direction to a level that exceeds historical vibration levels for that particular pump.

GPU Nuclear Response:

GPU Nuclear will limit the request at this time to Decay Heat Pump 1B (DH-PIB). This generic relief request is therefore withdrawn and a new specific relief request No. P 4 has been added for raising the vibration alert rance for DH-PIB from 0.325 ips to 0.400 ips in the vertical direction.

See New Relief Rcq se:,t No. P 4.

~  :

6710-96-2136 Attachment Page 10 of 24 RELIEF REQUEST NO. PG 2 (WITHDRAWN)

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NRC Comments on Relief Request PG2 and GPU Nuclear Response: )

i i

NRC Comment:

Licensee not allowed to request relief to use future editions of the Code without including related ,

requirements. l GPU Nuclear Response:

This relief request is being withdraw at this time. GPU Nuclear understands that this request is ,

premature pending generic action by ASME and NRC to approve pending code. l

s  :

6710-96-2136 Attachment Page 11 of 24 RELIEF REQUEST NO. PG 3 (REVISED)

Code Section from Which Relief is Reauested Relief is requested from OM-6, 66.1, " Acceptance Criteria," regarding doubling of test frequency in the alert range, and declaring a pump inoperable if vibration, flow, or differential pressure is in the alert or required action range.

Alternate Test DescriptiqD In lieu of the requirements in paragraph 6.1, " Acceptance Criteria," in OM-6 for pumps whose hydraulic and/or vibration data falls into the required action range of Table 3 " Ranges for Test Parameters," the requirements of ISTB 6.2.2, " Action Range," of the of the 1995 Edition of the ASME OM Code will be implemented for the inservice testing of safety related pumps. The related requirements of ISTB 4.6, "New Reference Values," will also be implemented for the inservice testing of safety related pumps.

If measured pump parameters fall within the alert range, the test frequency will be doubled until the cause of the deviation is found and corrected or an analysis of the pump is performed and new reference values established. If measured test parameters fall within the required action range, the pump shall be declared inoperable until either the cause of the deviation is determined and the condition corrected or an analysis of the pump is performed and new reference values established.

The analysis will include a comparison of the test results with the required design parameters, an evaluation of previous data to establish a trend, an investigation into the reason for the parameter change, and if necessary the collection of additional data. To be successful, the evaluation must conclude that the condition does not impair the ability of the pump to perform its safety function. The evaluation will be maintaine.) in the test records.

Basis for Relief Reauest An analysis of the pump condition can demonstrate that the pump can furnish its design function especially for those pumps with large margins above their design requirement. Doubling test frequency for pumps is only intended to get more data. For pumps that are normally standby, the degrading mechanism should not be active when the pump is off. Doubling test frequency may not establish any additional information.

Based on a successful operating history with no step 1hanges or abnormal trends, data which may occasionally fall in the alert range may not be unacceptable. Modifying or replacing the pumps to reduce vibrations only to assure that they do not occasionally fallin the alert range could be unnecessary. Further, doubling the test frequency would result in running the pumps more often at low flow conditions and would provide no useful information.

The deviation of pump hydraulic and/or vibration data which falls in the required action range of Table 3, " Ranges for Test Parameters," (consisting of Figure 1 and Tables 3a and 3h of the errata to OMa-1988 contained in OMb-1989) is an indication of degradation of pump performance.

However, such a deviation does not address the ability of the pump to perform its intended safety function or the rate of degradation. It may be possible through analysis of past perforer.ance data from the subject pump or other similar pumps to ensure that the pump remains capahic of performing its intended safety function until the next scheduled surveillance test. Such an analysis could prevent the unnecessary shutdown of the unit to perform repairs to the pump.

6710-96-2136 l Attachment Page 12 of 24 l

These provisions were permitted by earlier editions of ASME Section XI, Subsection IWP, and are i currently permitted by the ASME OM Code ISTB-1995, 666.2.2 and 4.6.

l l

NRC Comments en Relief Requet PG3 and GPU Nuclear Response:

NRC Comment:

This can be done per 10 CFR 50.59. Similar relief requests have been denied. See also PG2.

GPU Nuclear Response:

GPU Naclear has learned that similar relief has been granted other licensees. Having contacting another licensee with a similar request, the above has been revised to incorporate the intent of a request which has been approved by the NRC.

l 1

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671b'96-2136 Att'. hment Page 14 of 24 NRC Comments on Relief Request VI and GPU Nuclear Response:

NRC Comment:

Is valve MU-V79 required to be exercised in your IST program since MU-V78 is locked closed?

GPU Nuclear Response:

No, MU-V79 (a check valve is in series with hand operated locked-closed globe valve MU-V78) has no open safety function. Also, a small amount ofleakage past MU-V78 and MU-V79 would be of little consequence since the leaking fluid would remain in the Makeup and Purification System. Increased leakage would be evident by an increase in Makeup Tank level. Therefore this valve does not require leak rate testing. Since the discharge pressure of the running Makeup Pump provides the closing force, MU-V79 cannot be exercised during operation. It is physically impossible to open MU-V79 during normal system operation because RCS makeup and RCP seal injection flows are necessary for steady state plant operation. Exercising MU-V78 valve would provide no meaningful information since the valve has no open safety function and MU-V79 can not physically open or be orened during normal plant operation. GPU Nuclear has concluded that these valves do not need to be inc uded in the TMI IST program. Therefore, this relief request is being withdrawn.

i 6710-96-2136 Attachment .

Page 15 of 24  ;

i l

l RELIEF REQUEST NO. V 2 (WITHDRAWN) l l

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6710-96-2136 Attachment Page 16 of 24 NRC Comments on Relief Request V2 and GPU Nuclear Response:

NRC Comment:

Is valve DH-V50 required to be exercised in your IST program?

GPU Nuclear Response:

No. DH-V50 has no nuclear safety function in the open position but has a closed safety function to prevent fluid from the Reactor Building Sump from entering the nonseismic portion of the Spent Fuel System when Decay Heat Removal is in the Low Pressure Injection (LPI) mode with suction from the Reactor Building Sump. SF-V-44, a diaphragm valve in series with DH-V-50 and outside the ISI boundary, is also normally closed. Either of these valves serves to maintain LPI system integrity.

The head from the BWST is greater than the head DH-V50 and/or SF-V44 would er -rience  !

during accident conditions, Maintaining BWST level, which is alarmed, verifies that DH-V50 and/or SF-V44 are in the closed position and not leaking. GPU Nuclear has concluded that DH-V50 does need not need be included in the IST program. Therefore, this relief request is being withdrawn.

i l

6710-96-2136 4

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6710-96-2136 Attachment Page 18 of 24 NRC Comments on Relief Request V3 and GPU Nuclear Response:

NRC Comment:

Do the valves listed in these relief requests (V3, V4, and V5) meet the guidance for extending valve inspection frequency due to extreme hardship as described in GL 894)4, Position 2 and NUREG-1482, pages A-7 through A-157 GPU Nuclear Response:

Yes, An evaluation has been prepared in accordance with NUREG 1482, " Guidelines for Inservice Testing at Nuclear Power Plants," NRC Staff Position 2, Page A-8, which states that 4 disassemble / inspection can be longer that once every six years, if the following information is I developed i

1. Disassemble and inspect each valve in the valve grouping and document in detail the condition of each valve and the valve's capability to be full-stroked.
2. A review of industry experience, for example, as documented in NPRDS regarding the same ,

type of valve in similar service. l

3. A review of the installation of each valve addressing the "EPRI Applications Guidelines for Check Valves in Nuclear Power Plants" for problematic locations.

The evaluation,8 which is documented in the IST program, allows one valve (BS-V52A or BS-52B) to be disassembled / inspected every other refueling outage, alternating between A and B valves. Therefore, relief is not needed. l 1

3 GPU Nuclear Memorandum 3310-96-0007, dated March 13, 1996.

6710-96-2136 Attachment Page 19 of 24 RELIEF REQUEST NO V 4 (WITHDRAWN) l T ,. k. is i

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. .- l 6710-96-2136 Attachment Page 20 of 24 NRC Comments on Relief Request V4 and GPU Nuclear Response:

NRC Comment:

Do the valves listed in these relief requests (V3, V4, and V5) meet the guidance for extending valve inspection frequency due to extreme hardship as described in GL 89-04, Position 2 and NUREG-1482, pages A-7 through A-15?

GPU Nuclear Response:

Yes. An evaluation has been prepared in accordance with NUREG 1482, " Guidelines for Inservice j Testing at Nuclear Power Plants," NRC Staff Position 2, Page A-8, which states that i disassemble / inspection can be longer that once every six years, if the following information is  !

developed:  ;

1. Disassemble and inspect each valve in the valve grouping and document in detail the condition of each valve and the valve's capability to be full-stroked.
2. A review of industry experience, for example, as documented in NPRDS regarding the same type of valve in similar service.
3. A review of the installation of each va!ve addressing the "EPRI Applications Guidelines for Check Valves in Nuclear Powo. Plants" for problematic locations.

The evaluation,d which is documented in the IST program, allows one valve (BS-V30A or BS-30B) to be disassembled / inspected every other refueling outage, alternating between A and B valves. Therefore, relief is not needed.

I l

l l

1 4

GPU Nuclear Memorandum 3310-96-0007, dated March 13, 1996.

I y ,. l 6710-96-2136 Attachment l Page 21 of 24 l l

I i

RELIEF REQUEST NO. V 5 (WITHDRAWN)

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7 6710-96-2136 Attachment i Page 22 of 24 I

1 NRC Comments on Relief Request V5 and GPU Nuclear Response: )

l NRC Comment: l Do the valves listed in these relief requests (V3, V4, and V5) meet the guidance for extendmg valve inspection frequency due to extreme hardship as described in GL 89-04, Position 2 and NUREG-1482, pages A-7 through A-15?

GPU Nuclear Response: )

Yes. An evaluation has been prepared in accordance with NUREG 1482, " Guidelines for Inservice i Testing at Nuclear Power Plants," NRC Staff Position 2, Page A-8, which states that disassemble / inspection can be longer that once every six years, if the following information is developed:

1. Disassemble and inspect each valve in the valve grouping and document in detail the condition of each valve and the valve's capability to be full-stroked. I
2. A review of industry experience, for example, as documented in NPRDS regarding the same type of valve in similar service. j l
3. A review of the installation of each valve addressing the "EPRI Applications Guidelines for j Check Valves in Nuclear Power Plants" for problematic locations.

The evaluation,' which is documented in the IST program, allows one valve (CO-V175A or CO-V175B) to be disassembled / inspected every other refueling outage, alternating between A and B valves. Therefore, relief is not needed.-

I 5

GPU Nuclear Memomndum 3310-96-0007, dated March 13, 1996.

6710-96-2136 l

Attachment Page 23 of 24 RELIEF REQUEST NO. VG 1 (UNCHANGED)

Code Section from Which Relief is Reauested Generic relief is requested from OM-10, 664.2.2.3(a) and 4.1 regarding test frequency.

Alternate Test Description 1

These tests will be performed each refueling outage instead of the Code specified frequency "at least I once every two years." l Basis for Relief Reaug The refuel cycle for 5 MI-l is nominally two years. Several of the valves requiring leak testing cannot be tested with the plant operating. If, due to an intermediate outage (s), the refueling cycle exceeds two years, the code requirement could require a shutdown simply to test the certain valves.

This is impractical. Testing each refueling is a reasonable alternative.

Typically, valve position verification is done more frequently than once every two years. Some i valves must be stroked to verify position. Of these, several cannot be stroked with the plant I operating. As described above, the refuel cycle may extend beyond two years. Position verification using the code specified frequency, could cause the plant to be shut down. Position verification at least every refueling is a reasonable alternative to at least every two years.

Refueling interval testing has been extended to accommodate the 24 month refueling interval.

This change was made by a license amendment for all safety related equipment required to be tested each refueling interval. In some cases the regulations have been changed to accommodate 24 month refueling intervals without the need for an exemptions However, for IST position verification tests, the ASME Code has not been changed and therefore written specific relief is required. The fact that this does not represent a significant change is sufficient to justify this relief.

NRC Comments on Relief Request VG1 and GPU Nuclear Response:

NRC Comment:

Licensee should list valves that are within the scope of this relief request.

GPU Nuclear Response:

The valves included within the scope of this relief request are those IST valves listed in Table B-1 where a Position Verification Test is required at the 2-year frequency and are either inaccessible during plant operation or where position verification is only appropriate during the refueling interval test Although the list may change with changes to the IST program in accordance with 10 CFR 50.59, the current list of these valves is as follows:

AH-VIA/B/C/D CF-V20A/B DH-V4A/B HM-V4A/B MS-VI A/E /C/D CA-V4A/B CM-V1 DH-VSA/B HR-V22A/B MS-V2A/R CA-V5A/B CM V2 DH-V6A/B HR-V23A/B MS-V8A/B CA-V189 CM-V3 EF-V2A/B IC-V2 MU-V2A/B CF-VI A/B CM-V4 HM-VIA/B IC-V3 M U-V3 CF-V2A/B DH-VI HM-V2A/B IC-V4 M U-V10 CF-Vl9A/B DH-V2 HM V3A/B IC-V6 M U-V12

6710-96-2136 Attachment

.Page 24 of 24 MU-V14A/B NR-VIA/B/C RB-V2A RC-V43 WDL-V49 MU-V16A/B/C/D N R-V2 RB-V7 RC-V44 WDL V50 MU-V18 NR-V4A/B RC-RV2 RR-VIA/B WDL-V61 MU V20 NR-V6 RC-V2 RR-V3A/B/C WDL-V62 MU-V25 NS-V4 RC-V4 RR-V4A/B/C/D WDL-V303 MU-V26 NS-VIS RC-V28 RR-V5 WDL-V304 MU-V36 NS-V35 RC-V40A/B RR-V10A/B WDL-V534 MU-V37 NS-V52A/B/C RC-V41A/B WDG-V3 WDL-V535 MU-V51 NS-V53A/B/C RC-V42 WDG-V4 I

l 1

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DH-P-1 A Vibration Data .

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08/11/87 12/23/88 05/07/90 09/19/91 01/31/93 06/15/94 10/28/95 03/11/97

DH-P-1B Vibration Data '1 (Flowrate is ~925gpm)

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