ML20106C954

Proposed Tech Specs Re Change Flow Reduction of 10% in Core Spray Pump Surveillance Requirements

ML20106C954
Person / Time
Site:	FitzPatrick
Issue date:	09/28/1992
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20106C816	List: JPN-92-060, Forwards Application for Amend to License DPR-59,revising Tech Spec Flow Requirement for Core Spray Pumps & Associated Bases ML20106C952 ML20106C954
References
NUDOCS 9210060464
Download: ML20106C954 (18)

Text

B ATTACHMENT I to JPN-92-060 PROPOSED TECHNICAL SPECIFICATION CHANGE FLOW REDUCTION OF 10% IN CORE GPRAY PUMP SURVEILLANCLB10VIREMEN_T1 (JPTS-89-039) l l

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59 9210060464 920928 PDR ADOCK 05000333 P PDR

JAFNPP --

3.5 (cont'd) 4.5 (cont'd

b. Flow Rate Test - Once/3 Months Core spray pumps snali deliver at least 4,265 gpm [

against a system head -

corresponding to a reactor vessel pressure greater than or equal to 113 psi above primary containment pressure.  ;

t

c. Pump Operability Once/ month

d. Motor Operated Once/monin Valve .

e. Core Spray Header .

op Instrumentation Check Once/ day _.'

< Calibrate Once/3 nonths .

Test Once/3 months

f. Logic System Once/eact.

Functional Test ope ating cycle

g. Testaolo Check Tested for Valves operability any time the reactor is-in the cold condition-exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />,if operability tests have -

not been performed during the preceding 31 days.-

Amendment No. pd,1[9

. i JAFNPP .

?

3.5 (cont'd) 4.5 (cont'd)

F. ECCS-Cold Condition F. ECCS-Cold Condition Surveil;ance c.i the low pressure ECCS systems required

1. A minimum of two low pressure Emergency Core by 3.5.F.1 and 3.5.F.2 shall be a fcilows:

Cocn.ng subsystems 'shall be operable wheneve.

irradiated fuel is in the reactor, the reactor is in the mm a Dowrate test at least mce wery 3 rhonh cold condition, and work is being performed with the n the required Core Spray pump (s) and/or the RHR potential for draining the reactor vessel.

pump (s). Each Core Spray pump shall deliver at le st 4,265 gpm against a system head l

2. A minimum of one low pressure Emergency Core corresponding to a reactor vessel pressure greater Cooling subsystem shall be operable whenever than m equal to 113 psi above primary containment irradiated fuel is in the reactor, the reactor is in the pressure. Each RHR pump shall deliver at least t cold condition, and no work is beino oerformed with L91 gpm ga nst a system head corresponding to a the potential for draining tiie reactor 5essel.

eactor vessel to primary containment differential pessw 20 psid.

3. Emergency Core Cooling subsystems are not required to be operable provided that the reactc- 2. Perform a monthly operability test on the requiied vessel head is removed, the cavity is flooded, the Core Spray and/or LPCI motor operated valves.

spent fuel pool gates are removed, and the water 3. Once each shift vuify the suppression pool water level above the fuel is in accordance with level is greater than or equal to 10.33 ft. w+1enever .

Specification 3.10.C.

tne low pressure ECCS subsystems are aligned to the suppression pool.

4. With the requirements of 3.5.F.1,3.5.F.2, or 3.5.F.3 not satisfied, suspend core alterations and all 4. Once each shift verify a minim5 .i e' 324 inches of operations with the potential for draining the reactor water is available in the Condenssle Storags Tanks vessel. Restore at least one system to operable (CST) whenever the Core Spray System (s) is aligned status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish Secondary to the tanks.

Containment integrity within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Amendment No. l f ,f, $[,1[4 m

JAFNI-P e.5 BASES The testing laterval for the Core and Containment Cooling The RCIC flow rate is described in the UFSAR. The flow rates Systems is based on a quantitative reliability analysis, industry to be delivered to the reactor core for HPCi, the LPCI mode of practice, judgement, and practicality. The Emergency Core RHR, and CS are based on the SAFER /GESTR LOCA analysis.

Cooling Systems have not been designed to be fu!!y testable The ffow rates for the LPCI mode of RHR and CS arc modified during operation. For example, the core spray final admission by 310 percent reduction from the SAFER /GESTR iOCA va!ves do not open until reactor pressure has fallen to 450 psig; analysis. The reductior's are based on a sensitivity analysis thus, during operation even if high drywell pressure were (General Electric MDEd3-07E5) performed for th: parameters simulated, tne final valves would not open. In the case of the used in the SAFER /GESTR analysis. [

HPCI, automatic inU.iation during power operation would result '

i,n pumping cold water into the reactor vessel which ,s The CS surveillance requirement includes an allowance for i not desirabic. system leakage in r Jdition to the flow rate required to be delivered to the reac* 3r core. The leak rate from the core spray The systems will be automatically actuated during a refueling piping inside the r 3 actor but outside the core shroud is outage. In the case of the Core Spray System, condensato assumed in the UF5 AR and includes a known loss of less th6n storage tank water wi!I be pumped to the vessel to verify the 20 gpm from the 1/ A inch diameter vent hoie in the core sprq operability of the core spray header. To increase the availability T-box connection in each of the loops, and in the B loop, a of the individual components of the Core and Containment potential additional loss of less than 40 gpm from a c?arnshell looling Systems the components which make up the system repair whose structural wc'd covers only 5/6 of the e., instrumentation, pumps, valve operators, etc., are tested circumference of the pipe. Both of these identified sources of more frequently. The instrumentation is functionally tested leakage occur in the space between the reactor vessel wall and each month. Ukewise, the pumps and motor-operated valves the core shroud. Therefore flow lost through these leak are also tested each month to assure their operability. The sources does not contribute to core cooling.

combination automatic actuation test and monthly tests of the The surveillance requirements to ensure that the discharge

. pumps and valve operators is deemed to be adequate testing of these systems. piping of the core spray, LFCI mode of the RHR, HPC!, and RCIC Systems are filled provides for a visual observation . hat With compor*ents or subsystems out-of-service, overa!! core water flows from a high point vent. This ensures that and containment cooling reliability is mainta'ned by verifying the operability'of the remaining ccoling equipment. Consistent with the definition of operable in Section 4.0.C, demonstrate means conduct a test to show; verify means that the i associated surveillance activities have been satisfactori!y performed within the specified time interval.

Amendment No. )d 1[

132

Attachment 11 to JPN 92 060 SAFETY EVALUATION FOR PROPOEED TECHNICAL SPECIFICATION CHANGE FLOW REDUCTION OF 10% IN CORE SPRAY Pl;MP S38VELIAAEGE3EQUIREMRJL1JPTS 89-Q3SJ I. DISCRIPTION OF THE PROEQSED31dAJ_GES The proposed chance to the James A. FitzPatrick Technical Specifications revises the coro spray pump flow rato requirements and the Bases for the coro spray pump requirements. The enanges to the Technical Specifications are addressed below.

Minor changes in format, such as type font, margins or hyphenation, are not described in this submittal. These changes are typographical in nature and do not af fect the content of the Technical Specifications.

Eano 113. SntcdicMon 4.5. A.1A Replace the valuo "4,62S gpm" with the value "4,265 gpm."

Pano 122. Specification 4.5.F.1.

Replace the value "4,625 gpm" with the value "4,265 gpm."

Eagy 132. Bases 4.E Add paragraphs four and five that read as follows:

~

"The RCIC flow rate is described in the UFSAR. The flow rates to be delivered to the rouctor core for HPCI, the LPCI modo of RHR, and CS are based or' the k SAFER /GESTR LOCA analysis. The tiow rates for the LPCI mode of RHR and CS are modified by a 10 percent reduction from the SAFER /GESTR LOCA analysis.

Tiie reduct ons are based on a sensitivity analysis (General Electric MDE-83-0786) 2 performed for the paramoters used in the SAFER /GESTR analysis. ,

The CS surveillance requirement includes an allowance for system leakage in addition to the flow rate required to be delivered to the reactor core. The leak rate from the core spray piping inside the reactor but outsido the core shroud is assumed in the UFSAR and includes a known loss of less than 20 g?m from the 1/4 inch diarneter vent hole M the core spray T-box connection in each of the loops, and in the B loop, a potential additional loss of less than 40 Opm from a clamshell repair whose structural weld covers only 5/6 of tne circunforence of the pipe. Both of these identified sources of leakage occur in the sp.re between the reactor vessel wall and the core shroud. Therefore, flow lost through "Teso leak sources does not contrioute to core cooling."

11. P_t)3 POSE OF THE PRQP_QSLD_QHANGE.S The CS System is one of several Emergency Core Cooling Systems (ECCS) used to mitinate the consequences of loss-of coolant accidents (LOCAs). Core rpray is comprised of two subsystems (independent loops) with each subsystem consisting of a 100 percent capacity motor driven pump, piping, valses and a sparger to transfer water from the suppression pool to thc reactor vessel. The A and 8 core spray lines

Attachment 11 to JPN 92 060 SAFETY EVALUATION Page 2 of 9 enter the reactor vessel through two nozzles located 1E0o apart to provide physical separation. Each nozzle has a thermal sa.3ve that is welded into a T box. Two pipes are run from the T box to form a semicircular header with a downcomer at each end.

The downcomer has an elbow where the spray lines pass through the upper part of the shroud and into the spray sparger. This configuration is shown on Figures 1 and 2.

The core snray pumps are tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (Reference 1) and Technical Specifications 4.5.A.1.b and 4.5.F.1 to ensure that adequate emergency core cooling capacity is available. The current requirement in the Technical Specifications is that core spray pumps deliver at least 4,625 gpm against a system head corresponding to a reactor vessel pressure _

greater than or equal to 113 psi above primary containment pressure. The surveillance test should also account for system leakage that is not delivered to tne -,

cnre. Surveillance testing is conducted in accordance with the IrwService Testing (IST) program.

The purpose of this change request is to reduce the flow requirement for core spray surveillance testing. The reduction in the CS flow requirements is intended to allow the IST program pump performance band to be used to provide the potential for increased system availability. The change will also clarify the testing requirements for system leakage.

Ill. SAFETY IMPLICATIONS OF THE PROPOSED CH AN_QEJ The CS system is an emergency core coolin0 systems (ECCS) used to raitigate the consequences of loss of coalant accidents (LOCAs) and to provide inventory makeup in the alternate shutdown cooling mode in the event that the suction path from the reactor becomes unavailable for shutdown cooling or reactor inventory is lost. The -

surveillance testing required by Technical Specifications 4,5 A.1.b and 4.5.F.1 is intended to verify the capability of the core spray pump to deliver to the core the flow assumed in the SAFER /GESTR LOCA analysis (Reference 2). The proposed reduction in core spray pump flow rate will not af fect plant safety because the CS system can perform its required functions at the reduced flow rate while accounting for the system leakage. These considerations include:

1. C_gre Sorav Functional Reauirement A sensitivity analysis (Reference 3) was performed based on the SAFER /GESTR LOCA analysis (Reference 2) to assess the conservatism in current and proposed Technical Specification requirements for ECCS components. The sensitivity analpis varied component per!ormance requirements (e.g , Diesel Generator startup time, pump flow rates, valve stroke times, etc) to determine the sensitivity of the SAFER /GESTR LOCA analysis results for the design basis accident (i.e., recirculation line break). The flow rates for CS, Low Pressure Coolant Injection (LPCI) and Hi0h Pressure Coolant Ir.jection (HPCI) were reduced by 10% over their emire range in the analysis. For CS, the reduction was equivalent to a minimum rated flow of 4,163 Opm to the spray nozzles at a reactor vessel pressure equal to 113 psi above containment pressure.

E Attachment 11 to JPN 92-060 SAFETY E'lALUATION Page 3 of 9 The SAFER /GESTR LOCA analysis supplemented by the sensitivity analysis and an independent safeiy evaluation (Reference 4) justify operation with the reduced CS pump flow rate. The key issues in this determination were as follows:

• Sensitivity studies performed with the SAFER /GESTR LOCA models demonstrate an increase in fuel peak cladding temperature (PCT) of less than 120 F during the postulated design basis accident Jue to a reduction in all of the parameters. Since the current limiting licensing PCT is more than 600 F below the 2200 F allowable limit, the reactor core continues to meet the requirements of 10 CFR 50.46 and 10 CFR 50, Appendix K with a margin of 500 F. The statistical upper bound PCT remains at least 150 F less than the Appendix K case and will meet the 1,600 F limit of Reference 5.

• Sensitivity studies performed with the SAFER /GESTR LOCA models demonstrate thas 88 F of the increase in fuel peak cladding temperature (PCT) is attributable to a 10% reduction of all ECCS flow rates. This leaves a safety margin of more than 500 F.

• An increase in temperature will result in a small increase in the metal water reaction for thn limiting break accidents. The results of an earlier LOCA analysis (Reference 6) are moro limiting than the results of the sensitivity analysis so the metal water requirements of 10 CFR 50.46 are still met. The containment evaluation of UFSAR Section 14.6.1.3 is also bcunding.

• There is no increase in the PCT for the worst case Appendix R fire due to a reduction in core spray flow. The Appendix R analysis (references 7 and 8) assumed operation of one RHR pump in the LPCI moce. The other RHR pumps and the core spray pump were assumed to be inoperable.

• The requirements for inventory makeup to mitigate the consequences of inadvertent draindown while shutdovsn are bounded by the LOCA. The limiting double ended guillotine break of the recirculation line (4.17 sq. f t.) is lar0er th.in any opening associated with draindown. Upda+ed Final Safety Analysis Report (UFSAR) Section 14.6.1.3, indicates that a single core spray system is capable of long term cooling for the LOCA and it is, therefore, adequate for draindown in the cold condition. Since the sensitivity analysis has used the same methodology as the current LOCA analysis, a single CS system at reduced flow is suitable for the cold condition.

2. Eystem Leakane System leakage is the difference between CS pump flowrate and CS flowrate inside the core shroud. The CS flowrate used in the LOCA analysis (References 2

, and 3) is the CS flowrate inside the core shroud.

When the FitzPatrick plant was being designed, leakage was postulated to occur from the thermal sleeve between the T box and vessel nozzle and a quarter inch vent hole in the T box that allowed for release of non condensable pases (Reference 9).

1

_ _ ._ _ _. - . . _ . _ . _ - _ _ _ _ _ _ . . . m. _ _ . _. _

Attachment il to JPN 92 060 SAFETY EVALUATION Page 4 of 9 The leakage requirement included in this proposed Tcchnical Specification change is based on an assessment of the actual system leakage (References 9,"O and 11). The assessment was part of the analysis used to validate CS flowrate after repair of a crack in the core spray piping outside the shroud on the "B" ioop. The assessment identifies the elimin=' ion of thermal sloovo leakage before plant operation and calculates the upper mund leakage from the upper T box vent hole (0.251.05 inch) as less than 20 gpm. The crack in the "B" loop core spray piping was repaired by wolding a clam shell on the upper riser outside the shroud.

The weld :: overs only 5/6 of the circumference of the pipe (attachment 7 to Reference 11) and calculatio is in Reference 11 conservatively conclude that leakage from the unwolded sector is less than 40 gpm.

Based on the a. ;vo, the required CS .flowrate must allow for a leakage of 20 gpm and 60 gpm to 'A" and "B" loops, respectively. Since 4,163 gpm is required for delivery to the core (see item 1), the reduced flow requirement bounds the calculated maximum leak rate.

CS pump surveillance testing to moet Technical Specifications 4.5.A.1.b and 4.5.F.1 and the IST requirement is currently performed in accordance with Surveillance Test Procedure ST-3P (Reference 12). Surveillance Procedure ST-3P is currently adequate to demonstrato the ability of the core spray pump. Allowable ranges for test quantities are specified in accordance with Tablo IWP-3100-2 to ASME Section XI.

Approval of the proposed Technical Specification change will allow ST-3P to be revised for testing at a reduced flow rato.

Operation of the plant in accordance with the proposed amendment will not be a safety concern. The of fect of the reduction in CS pump flow is_a decrease in the margin betwoon the calculated PCT and the allowable limit. Secondly, there is an increase in margin between the requirements proposed for the Technical Specifications and the ASME Section XI inservice test (IST) reference values. These safety considerations were previously identified in a request to reduce LPCI pump flow by 10% (Reference 13) which was approved as Amendment 171 (Reference 14). The conclusions of the plant's accident analyses as documented in the UFSAR and the NRC staff's SER at operat!ng licenso stage are not altered by these changes to the Technical Specifications.

The Authority has revised the licensing basis SAFER /GESTR LOCA Analysis (Reference 15) as part of the power uprate evaluation (Reference 16). The SAFER /GESTR 1.OCA Analysis for power uprate used lower pump flow rates than found in the wasitivity analysis. When approved, the updated analysis will provide the basis for new ECCS flow requirements. Further reductions in flowrate can be requested at that time.

l t-

Attachment 11 to JPN.92-000 SAFETY EVALUATION Pa00 5 of 9 IV. EV ALU ATION _QLSMNIFIC ANT H AZ ARD S_CONSipjifMllDfj Operation e' the FitzPatrick plant in accordance with the proposed Amendment would not involve a si0nificant hazards consideration as defined in 10 CFR 50.92, since it would not-

1. involve a significant increase in the probability or consequences of an accident pwviously evaluated.

The CS system is designed to mitigate the consequences of analyzed accidents and is normally in the standby mode. The proposed changes reduce flowrate whiclis a reduction in the performance condition required to respond to an -

accident. This does not effect the manner in which the CS system is tested or its function. Therefore, the changes havo no effect on the conditions which could initiate an accident.

The offect of a 10% reduction in the CS pump flow rate has been analyzed using approved methodolo0y. The PCT merease of 88 F has no significant effect on the existing mar 0i n to the 2200 F acceptance criteria. Slight increases in metal water reaction occur. These are not of safety significance because the prior LOCA analyses used a higher PCT and the UFSAR evaluations are based upon metal water reactions due to more severe conditions. There are no changes to the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR). The change, therefore, does not ef fect continued compliance with 10 CFR 50.46.

2. create the possibility of a now or dif f erent kind of accident from any accident pmviously evaluated.

The decrease in flowrate for the CS pump is a decrease in the performance _

requirement for the system. Conditions that could lead to an accident are not changed. There are no changes to the manner in which tests are conducted, no changes to system design and no changes to operatin0 procedures that could result in a new or different kind of accident.

3. involve a significant reduction in a mar 0 i n of safety.

The of fect of a 10% reduction in the CS pump flow rate has been analyzed using approved methods. Margin of safety is provided by the conservatisms required in Appendix K and by L conservative application of the approved GESTR-LOCA and SAFER Models in NEDE-23785. These margins are not effected by this change.

V. iMPLEMEJI_ATION QF THE PROPQSED CHANGES Implementation of the proposed changes will not adversely affect the ALAHA or Fire Protection Programs nor will the changes af fect the environment. There will be no plant modifications or testing chan0es that can have an ef fect on either the programs or the environment.

Attachment il to JPN 92-060 '

SAFETY EVALUATION Pace 6 of 9 vi. .c. ORC W S10E The changes, as proposed, do not constitute an unroviewed safoty question as defined in 10 CFR 50.59. That is, they:

1. will r.ot change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report;

2. will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report;

3. will not reduce the margin of safety as defined in the basis for any technical specification; and The changes involve no si0nificant hazards consideration, as defined in 10 CFR 50.92.

vil. BEfEBEficES References relied upon to prepare the Technical Specification chan00 request:

1. ASME Boiler and Pressure Vessel Code,Section XI,1980 Edition through Wintor 1981 Addenda.

2. NEDC-31317P, " James A. Fitz natrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis " dated October 1986 including Errata and Addenda (Proprietary).

3. GE "Sencitivity of the James A. FitzPatrick Nuclear Power Plant Safety Systems Performance to Fundamental System Parameters" dated July 1936 (MDE 0_786) Proprietary.

4. James A. FitzPatrick Nuclear Power Plant, Nuclear Safety Evaluation _JAF-SE ~

146, Rov. O, " Evaluation of a.10% Reduction in LPCS Pump Surveillance Flow Rate," dated September 1992.

5. NRC letter, C. O. Thomas, to GE dated June 1,1984 regarding acceptance for referencing of Licensing Topical Report NEDE-23785, Revision 1, Volume 111 (P),

'"The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident."

6. GE " Loss-of Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant, NEDO-21662-2, dated July 1977.

l 7. GE " Analysis To Extend Operator Action Timo For Alternate Shutdown Panels in

!- Support of FitzPatrick Compliance To Appendix R" dated November 1985 (MDE-8137-0585, Revision 2).

Attachment il to JPN-92-000 SAFETY EVALUATION Page 7 of 9 4

8. NYPA lotter (JPN-85-090), J. C. Brons to NRC, dated December 17,1985, providing additionalinformation on an exemption request from Section ill.L to Appendix R of 10 CFR 50 ro0arding altomato shutdown capability.

9. GE " Core Spray Lino Crack Analysis for James A. FitzPatrick Nuclear Power Plant" dated October 1988 (EAS 64-03243) Propriotary.

10. James A. FitzPatrick nuc! car safoty evaluation JAF-SE 88-190, " Repair of In-vessel Coro Spray Lino Using e '"olded Clamshell Sloove," dated October 14, 1988.

11. NYPA lotter (JAFP 88 0965), R. J. Converso to NRC, dated October 21,1988, providin0 additional information on an internal vossol Coro Spray System pipo crack.

12. James A. FitzPatrick surveillanco proceduto ST-3P "Coro Spray Flow Rato and Valvo inservico Test," Revision 9, dated December 4,1991.

13. NYPA lotter (JPN 90 049), J. C. Brons to NRC, dated June 21,1990, requesting a change to the Technical Specifications to reduco LPCI pump flow requirements.

14. NRC lotter, B. C. McCabo to NYPA, dated July 1,1991, regardin0 iss once of amendment 171 to the Technical Specificetions.

15. GE NEDC-31317P, " James A. FitzPatrick Nuclear Pow

• Plant SAFER /GESTR-LOCA Loss-of Coolant Accident Analysis," Revision 1, dated November 1991 (Propriotary).

16. NYPA lottor (JPN 92-028), R. E. Boodlo to NRC, dated June 12,1992, requestin0 an Amendment to the Technical Specifications to allow power uprato.

Background documents not specifically referenced.

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report Sections 6.3, 7.2, 7.3, 8.3 and Chapters 5 and 14, Revision 5 dated throu0h January 1992.

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

3. James A. FitzPatrick Nuclear Power Plant, " Inservice Testing Program for Pumps and Valves," dated May 1,1991, i

4. James A. FitzPatrick Nondestructivo Examination Proceduro I.B.I.P. 2, " Visual

! Examination of the Roac*or Vossel and Internals," dated May 14,1991.

I 4

. . . . . . . . _ . ~ _ . _ _ . _ . _ _ . . . - _ _ _ . . . __ _ __ .

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Attachment il to JPid 92-0GO SAFETY EVALUATION Page 8 of 9 FIGURE 1 l

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Page 9 of 9 FIGURE 2

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9 ATTACHMENT 111 to JPN-92-060 PROPOSED TECHNICAL SPECIFICATION CHANGE FLOW REDUCTION OF 10% IN CORE SPRAY PUMP SURVEILLANCE REQUIREMENTS MARKUP OF TECHNICAL SPECIFICATION PAGES (JPTS 89-039)

New York Power Authority l JAMES A. FITZPATRICK NUCLEAR POWER PLANT

! Docket No. 50-333 DPR-59 l

n

~

JAFNPP 4.5 BASES The testing Interval ior the Core and Corce.inmerit Coding With components or subsystems out-of-semce, overa!! core Systems in ba::sd cn a quarrJiative reliabi!!ty analysis, industry and containment evAng reliability is maintained by verifying the operability of the em#ning cooling equipment. Consistent l practice, jodgement, and practicality. The Emergency Core Coding Systems have not been desgned to be fu!!y testable with the definition of operable in Section 4.0.C demonstrate during operation. For ewe, the core spray final admission rneans conduct a test to show; venty means that the vWves do not open until reactor pressure has fallen to 450 psig; associated surveillance activities have been satisfactonly thus, during operation even if high drywell pressure were performed within the specified time interval.

simulated, the final vWves would not open. In the case of the MThe survei!!ance requirements tc ensure that the discharge HPCI, automatic initiation during power operation would resuit / i ing of the core spray, lPCI mode of the RHR, HPCI and in pumping cold water into the reactor vessel which is not . RCiC Systems are filled provides !or a visual observation that Mrade. water *ows from a high point vent. This ensures that The systems will be automatica!!y actuated during a refueling x outage. In the cm of the Core Spray System, condensate storage tank vWer will be pumped to the vessel to verdy the s y- 1f\5eX4 A operabinty of the core sprey header. To increase the availability of th( IndivMual components of the Core and Containmert Coding Systems the compcoents which make up the system i.e., ene rurwitation, pumps, valve operators, etc., are tested more fr'x;uently. The instrumentation is functionally tested each month. Ukewise, the pumps and motor-operated valves are also tested each month to assure their operability. The combination automatic actuation test and monthly tests of the pumps and va?ve operators is deemed to be adequate testing of these systems.

Amendment No.14,148 132 w, -

l INSERT A The RCIC flow rato is described in the UFSAR. The flon rates to be delivered to the reactor core for HPCI, the LPCI modo of RHR, and CS are based on the SAFER /GESTR LOCA analysis. The flow ratm for the. tPCI modo of RHR and CS are modified by a 10 percent reduction from the CA%R/G;:MR LOCA ana!ysis. The reductions are based on a sensitivity analysis (General Electric MDE 83-0786) performed for the paramotors used in the SAFER /GESTR analysis.

1ho CS surveillanco requirement includes an allowance for system leakago in addition to tho flow rato required to be delivered to the reactor core. The leak rato from the coro spray piping inside the reactor but outside the coro shroud is assumed in the UFSAR and includes a known loss of less than 20 gpm from the 1/4 inch diamotor vont hole in the coro spray T box connection in cach of the loops, and in the B loop, a potential _

additional loss of loss than 40 gpm from a clamshell repair whoso structural wcld covers ,

only 5/G of the circumference of the pipe. Both of these identified sources of leakago ,

occur in the space betwcon the reactor vessel wall and the coro throud. Therefore flow lost through those leak sources dt,es not contribute to coro coolirg. 6, D

l 1

_ _ _ _ _ _ _ _ _ _ _ - - _ - ____-__-_ _ _ __ _ ___