ML20043J013

Proposed Tech Specs Re LPCI Pump Flow Rate Requirements

ML20043J013
Person / Time
Site:	FitzPatrick
Issue date:	06/21/1990
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20043J011	List: ML20043J010 ML20043J012 ML20043J013
References
NUDOCS 9006270195
Download: ML20043J013 (10)

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J ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CH W WATE l REQUIREMENTS I (JPTS-90 005) l l

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50 333 l DPR 59 )

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JAFNPP 3.5 (cont'd) - 4.5 (cont'd)

2. From and after the date that one of the Core Spray 2. When it is determmed that one Core Spray System is Systems is made or found inoperable for any reason, moperable, the operable Core Spray System, and both continued reactor operation is pamibsible dunng the LPCI subsystems, shall be venfied to be operable succeeding 7 days unless the system is made operable immediately. The remarung Core Spray System shall be earlier, provided that dunng the 7 days all actrve ven6ed to be operable daily thereaRer.

components of the other Core Spray System and the LPCI System M N wh

3. LPCI System testmg shall be as spacJ.=d in 4.5.A.1a, b, c,

3. Both LPCI subsystems of the RHR System shall be d, f and g except that each RHR pump shall deliver at least cperable whenever irradiated fuel is in the reactor and prior 8,910 gpm agamst a system head cu W.J ig to a l to reactor startup from a cold condition, except as

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reactor vessel to pnmary contamment differential pressure specified below. of greater than or equal to20 psid. l

a. From the time that one of the LPCI subsystems is a. When it is determoed that one LPCI subsystem is

made or found to be inoperable for any reason, inoperable, the operable LPCI subsystem and both i

continued reactor operation is po. nissible dunng the Core Spray Systems shall be venfied to be ope able succeedmg 7 days unless that subsystem is made immediately and daily thereafter.

operable earlier provided that dunng these 7 days i

the operable LPCI subsystem and both Core Spray Systems shall be operable.

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l Ar.mdment No. *f,4[,9[1[,1[,1[

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JAFNPP 4.5 (cont'd) 3.5 (cont'd) .

2. When it is determoed that one RHRSW pump of the Should one RHRSW pump of the cuiipcs.ords required in 2.

3.5.B.1 above be made or found inoperable, conhnued components requred in 3.5.B.1 above is inoperable, the remaming components of the contamment coohng mode reactor operation is permissible only dunng the succeedire 30 days provided that dunng such 30 days al: subsystems shall be venlied to be operable immediately rerrsrnrg curipcs.ords of the curdaii rient coolog mode and daily thereafter.

yste are ym. 3. When one contamment cooling subsystem loop beoones Should one of the contamment cooling subsystems inoperable, the redundant contamment cooling subsyst un l 3. loop shaN be W to be operable immediately and '.mily beconw inoperable or should two of the RHRSW pumps thereaner. When two of the RHRSW pumps 1,ecome become inoperable, continued reactor operation is permissible for a penod not to exceed 7 days.

inoperable, the remammg components <f As h 4 .

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4. li the requirements of 3.5.82 or 3.5.B.3 cannot be met, the reactor shall be placed in a cold cundition wittwn 24 hr.

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5. Low power physics testing and reactor operator trammg shall be pemitted with reactor coolant temperature

<212"F with an inoperable component (s) as specired in 3.5.8 above.

' Amendment No. , 1f,If,1

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ATTACHMENT ll SAFETY EVALUATION FOR PROPOSED TECHN RDING l LPCI PUMP PLOW RATE REOUIREMENTS (JPTS 90005) 1 I

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59 9

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,- Attachment 11 SAFETY EVALUATION Page 1 of 6 I. DESCRIPTION OF THE PROPOSED CHANGE The proposed changes to the James A. FitzPatrick Technical Specifications revice

-t Specifications 3.5.A.3.a. 4.5.A.3., on page 114, and 3.5.B.3 on page 116.

Two changes are pryd The first, deletes the expired 14 day Umiting Conditions for p, Operability (LCOs) on the 'A' side Low Pressure Coolant injection (LPCI) and Containment Cooling Subsystem.

The second, reduces the surveillance test flow rate acceptance value for each RHR pump from 9900 gpm to 8910 gpm.

Both the 14-day LCO change and the RHR pump flow rate change were previously approved on a temporary basis in Amendment 153 (Reference 1).

A. Deletion of the 14-day LCOs

1. Specification 3.5.A.3.a. Page 114 Delete the asterisk (*) at the end of this specification and remove the following note:

• LPCI subsystem 'A" may be Inoperable for a 14 day period. This -

temporary LCO condition exists until the end of Cycle 9.

2. Specification 3.5.B.3, Page 116 Delete the asterisk (*) at the end of this specification and remove the following note:

• Containment Cooling subsystem 'A' may be Inoperable for a 14 day period. This temporary LCO condition exists until the end of Cycle 9.

B. RHR Pump Flow Rate v

1. Specification 4.5.A.3., Page 114 l

Replace the RHR pump f'uw rate acceptance value of 9900 gpm with 8910 gpm.

L Delete the dagger (t) at the end of this specification and remove the following note:

tFor the remainder of Cycle 9, RHR pumps 'A' and 'C' shall each deliver at least 8,910 gpm against a system head corresponding to a reactor vessel to primary containment differential pressure of greater than or equal to 20 psid.

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,. Attachment ll SAFETY EVALUATION i Page 2 of 6

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11. PURPOSE OF THE PROPOSED CHANGE A. Deletion of the 14-day LCOs l The deletion of the temporary 14 day LCO conditions eliminates extraneous and out of date ,

Information from the technical specifications.  !

B. RHR Pump Flow Rate The Emergency Core Cooling System (ECCS) for the RtzPatrick plant consists, in part, of the Low Pressure Coolant injection (LPCI) mode of the RHR system. LPCI is comprised of two subsystems (loops) with each LPCI subsystem consisting of two motor driven RHR pumps, and piping and valves to transfer water from the suppression pool to the reactor vessel, The RHR pumps are tested in accordance with Section XI of the ASME Boller and Pressure  !

Vessel Code (Reference 3) and Specification 4.5.A.3 to ensure that adequate emergency core cooling capacity is available. The current criterion is that flow for each pump must be at least 9000 gpm against a system head corresponding to a reactor vessel to primary containment differential pressure of at least 20 paid.

At 0900 gpm, the RHR pump (s) are on the steep part of their head capacity curves (i.e., near runout flow). In this region of the pump curve, small variations in flow result in non-proportional larger variations in pressure. The lower minimum allowable RHR pump flow rate (8910 gpm) will reduce the impact of flow measurement uncertainties. Future inservice pump tests will be more repeatable and pump degradation more accurately detected, inservice Testino of the RHR sM in a letter dated February 9,1C00 to the NRC (Reference 2), the Authority described an .

y apparent decreasing trend in the measured differential pressure (dp) across the "A* and "C" l RHR pumps when tested at a flow rate of 9900 gpm. After investigating instrumentation, valving, and other possible causes for the negative trend, the source of the problem appeared to be internal to the RHR pumps. Temporary technical specification changes were requested, and approved, to allow pump refurbishment during power operation.

The "C" RHR pump was overhauled in March,1990 during the temporary LCO extension granted via Amendment 153 to the FitzPatrick operating license (Reference 1). No evidence of pump degradation was noted and post work testing showed pump performance to be ,

unchanged after the overhaul. The pump supplier, Byron Jackson, believes the pump is '

performing as designed and the apparent degradation in 'A' and "C" pump performance to be the result of either divergence of flow into branch lines upstream of the flow orifice, flow measuroment uncertainties, or a combination of both.

During the 1990 refueling outage, the Authority has systematically tested and evaluated potential root causes of the apparent reduction in dp associated with LPCI loop 'A". Tho -

evaluations focused on components common to the "A* and 'C' pumps during

. performance testing. System hydraulic considerations such as potential suction blockage, -

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, Attachment ll SAFETY EVALUATION Page 3 of 6 air entrainment, flow divergence between the pumps and the flow orifice, and test i measurement unoortainties were extensively evaluated. These evaluations conclude that the RHR pumps are performing satisfactorily and that the test equipment consistently under-predicts flow through the flow orifice (i.e., indicated flow is less than actual flow). Under.

predicting flow causes pump testing to be conducted at a higher flow rate which results in a lower pump dp indication. Test results are worse near runout conditions because a given i difference betwoon actual versus indicated flow results in a disproportionate change in j indicated dp. >

Analysis of historical surveillance data indicates two discreet downward step changes in j apparent pump performance. The first, occurring in January 1986, corresponded to a change in pump discharge pressure test instrumentation. The second, occurring in August i 1989, corresponds to a change in the control room indicator used to read pump flow. Since August 1989, RHR pump (s) performance has shown no downward trends.

1 L Section XI of the ASME Boller and Pressure Vessel Code (Reference 3) describes safety

related pump inservice testing requirements, inservice tests on the RHR (LPCI) pumps are conducted at nominal motor nameplate speed by varying system test line resistance until measured flow equals the preestablished reference value. Other test quantitles, as specified in Table IWP 31001 of the Code, are then measured or observed to confirm acceptable .

pump operation, j lt has been the Authority's practice to conduct RHR pump flow rate tests at the flow rate i specined in Surveillance Requirement 4.5.A.3. This allows direct verification of pump performance versus the ECCS Appendix K analysis assumptions regarding LPCI flow rate at l ' a given pressure. The lower minimum RHR pump flow rate of 8910 gpm will allow a l

corresponding change to the RHR pump inservice (surveillance) test procedures.

lil. IMPACT OF THE PROPOSED CHANGE i A. Deletion of the 14 day LCOs Deleting the 14 day LCO conditions is an editorial change. These changes can not impact the capability of the emergency core cooling systems or the containment cooling l mode of RHR.

[ B,. RHR Pump Flow Rate Amendrnent 153 to the FitzPatrick Technical Specifications changed'the surveillance test l flow acceptance value for the 'A' and 'C' RHR pumps from 9900 gpm to 8910 gpm (a decrease of 10%). The Authority's application for Amendment 153 (Reference 2) was L limited to the *A" LPCI subsystem and for the duration of operating Cycle 9 because of the l exigent nature of the amendment application.

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The justification for a 10 percent flow reduction was the RtzPatrick SAFER /GESTR Appendix K LOCA analysis (Reference 4) supplemented by a r.ifety evaluation (Reference

, Attachment 11 SAFETY EVALUATION Page 4 of 6 5). Together, these analyses substantiate continued operation with an allowable LPCI pump flow rate of 8910 gpm. The key issues were:

. Sensitivity studies performed with the SAFER /GESTR Appendix K LOCA models demonstrate an increase in fuel peak cladding temperature (PCT) of 88'F for a 10%

reduction of all ET.S flow rates. Since the current limiting licensing PCT is more than 8000F below the 22t#F allowable limit, the RtzPatrick plant continues to meet the requirements of 10 CFA 50.46 and 10 CFR 50, Appendix K with over 500 F margin, o GE has estimated a slight increase in PCT (less than or equal to 80 F) for the worst case Appendix R fire assuming a 10% decrease In LPCI flow rate. Since the calculated fuel PCT for this event was 10130 F, which is well below temperatures associated with fuel cladding damage, the FitzPatrick plant continues to meet the requirements of 10 CFR 50, Appendix R.

e An evaluation of the other operating modes of RHR (i.e., suppression pool cooling, decay heat removal, and containment spray cooling) shows that the RHR pumps operating at 8910 gpm exceed the design and FSAR flow rates for these modes, in other words, e there is no effect on the other RHR operating modes because the design of the AHR pumps is based on the more limiting requirements of the LPCI mode.

The resultant impact of changing the RHR pump surveillance test flow acceptance value from 9000 ppm to 8910 ppm is twofold.

First, there is an insignificant decrease in the safety margins associated with ECCS performance following a LOCA and with reactor inventory makeup capability during postulated Appendix R events.

Secondly, there is an increase in margin between the allowable value specified in Technical .

Specification Section 4.5 A.3 and the ASME Section XI inservice test (IST) reference values.

In addition, testing the RHR pumps at 8910 gpm will eliminate problems inherent in testing the pumps near runout flow conditions. The present test flow rate of 9900 gpm is approximately 30% greater than the pump's design point of 7700 gpm and only 6% below runout flow conditions (10,500 gpm).  ;

Operation of the plant in accordance with the proposed amendment is not a safety concern. The conclusions of the plant's accident analyses as documented in the FSAR or the NRC Staff's SER are not altered by these changes to the Technical Specifications. The Authority intends to revise the licensing basis SAFER /GESTR ECCS performance analysis prior to start up following the 1991 refueling outage. The updated analysis will contaln LPCI flow changes and other planned plant changes.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with the proposed amendment would not involve a significant hazarda consideration as defined in 10 CFR 50.92, -

since it would not:

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Attachment 11 SAFETY EVALUAT)ON Page 5 of 6

1. Involve a significant increase in the probability or conseque@t of an accident previously evaluated. The LPCI (RHR) system is designed to mitigate tre consequences of analyzed accidents and is normally in the standby try>de. The proposed changes have no effect on the probability of oocurrence of previously evaluated accidents.

The effect of a reduction of the RHR pump flow rates has boon fully analyzed. These analyses demonstrate that the consecuences of postulated socidents remains well within the acceptable limits established in the FitzPatrick FSAR and applicable federal regulations.

The 880F ed increase in peak clad temperature is not significant with respect to the existing F margin to the 2200F acceptance critoria.

The proposed change which daistes the temporary 14 day LCO conditions eliminates extraneous and out of date information from the technical specifications. This change is an editorial change and can not impact the capability of the emergency core cooling systems or the containment cooling mode of RHR.

. 2. create the possibility of a new or different kind of accident from those previously evaluated.

The proposed changes do not Involve hardware changes at the FitzPatrick plant. No actions taken as a result of the proposed changes can initiate any type of accident.

3. Involve a significant reduction in the margin of safety. The effect of a 10% reduction in the RHR pump flow rate has been fully analyzed. Although the calculated fuel PCT has increased by 88 F, this is not significant with respect to the 600 F margin to the ECCS acceptancecriterlaof 2200 F.

V. IMPLEMENTATION OF TNE PROPOSED CHANGE Implementation of the proposed change will not impact the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the change impact the environment.

t fl. CONCLUSION The changes, as prW. do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they;

a. wl!! not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report;

b. will not increase the possibility for an accident or malfunction of a type different from any evaluated previously in the safety analysis report; e ,

c, will not reduce the margin of safety as defined in the basis for any technical specification; and

o. Involves no significant hazards consideration, as defined in 10 CFR 50.92.

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!: SAFETY EVALUATION Page 6 of 6 Vll. REFERENCES

1. Amendment No.153 to the James A. FitzPatrick Operating Uconse, dated February 28,1990, temporarily extends the allowable out of service time for one LPCI and Containment Cooling subsystem from 7 days to 14 days and reduces the RHR pump flow rate allowable value from 9000 gpm to 8910 gpm.

2. NYPA Letter (JPN 90-013), J.C. Brons to NRC, dated February 9,1990, requesting changes to the RHR Pump Technical Specifications on an exigent basis.

3. ASME Boller and Pressure Vessel Code, Sect!on XI,1980 Edition through Winter 1981 Addenda.

4. James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LOCA ' ass of Coolant Accident Analysis, NEDC 31317P, dated October,1986.

5. James A. FitzPatrick Nuclear Power Plant Nuclear Safety Evaluation For a 10% Decreaso in LPCI Flow, JAF SE 90024, dated February 5,1990, prepared by General Electric Co. and approved by the New York Power Authority.

6. James A. FlizPatrick Nuclear Plant Updated Final Safety Analysis Report, Chapter 4.8 (Residual Heat Removal System) and Chapter 6 (ECCS).

7. James A. RtzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972, and Supplements.

8. Sensitivity of the James A. FitzPatrick Nuclear Power Plant Safety Systems Performance to Fundamental System Parameters, Proprietary General Electric Report, MDE 83 0786, dated July,1986.