JPN-91-064, Proposed Tech Spec Changes Re ASME Section XI & Emergency Svc Water Pump Surveillance Testing

Proposed Tech Spec Changes Re ASME Section XI & Emergency Svc Water Pump Surveillance Testing

ML20086E042
Person / Time
Site:	FitzPatrick
Issue date:	11/19/1991
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20086E026	List: JPN-91-064, Proposed Tech Spec Changes Re ASME Section XI & Emergency Svc Water Pump Surveillance Testing ML20086E021 ML20086E032
References
JPN-91-064, JPN-91-64, NUDOCS 9111270033
Download: ML20086E042 (46)

Text

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ATTACHMENT I to JPN 910G4 PROPOSED TECHNICAL SPECIFICATIOtt CH ANGES ASME SECTION XI AND ESW PUMP SURVEILLANCE TESTitlG (JPTS 90023) l 1

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

+111270033 93333.p DPR 59 fL"' ADOCK 0n00o233 m

JAFNPP 4

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Pagg 1.0 Dofinitions 1 UMITING SAFETY SAFETY LIMITS SYSTEM SETTINGS 1.1 Fuoi Cladding integrity 2.1 7 1.2 Reactor Coolant System 2.2 27 SURVEILLANCE UMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.0 Gonoral 4.0 30 3.1 Nactor Protection System 4.1 30h l 3.2 Instrumentation 4.2 49 A. Primary Containment Isolation Functions A. 49 B. Coro and Containment Cooling Systems - Initiation and B. 49 Control C. Control Rod Block Actuation C. 50 D. Radiation Monhuring Systoms -Isolation and initiation D. 50 Functions E. Drywoll Loak Detection E. 54 F. Survoillanco Information Roadouts F. 54 G. Rocirculation Pump Trip G. 54 H. Accident Monitoring Instrumentation H. 54

1. 4kV Emergoney Bus Undervoltage Trip 54 3.3 Reactivity Control 4.3 88 A. Reactivity Umitations A. 88 B. Controf Rods B. 91 C. Scram Insortion Times C. 95 D. Reactivity Anomatios D. 96 3.4 Standby Uquid Control System 4.4 105 A. Normal Operation A. 105 w B. Operation With inoporablo Components B. 106 C. Sodium Pontaborato Solution C. 107 3.5 Core and Containment Cooling Systems 4.5 112 A. Coro Spray and LPCI Systems A. 112 B. Containment Cooling Modo of the RHR System B. 115 C. HPCI System C. 117 D. Automatic Deprossurization System (ADS) D. 119 E. Reactor Coro isolation Cooling (RCIC) System E. 121 Amendmont No. 1

,f,

JAFNPP 3.0 Continued 4.0 Continued D. Entry into an OPERATIONAL CONDITION (mode) shaft not be D. Entry into an OPERATIONAL CONDITION (mode) sha!! not be made unless the cordtions of the Umiting Condition for made un! css the Surveillance RegtswTent(s) associated with Operation are met without reliance on provisions contained in the Unwng Condition for Operation have been p6fwn ed within the ACTION statements unless otherwise excepted. This the appfcable surveii!anc;e interval or as otherwise specified.

provision shall not prevent passage through OPERATIONAL CONDITIONS (modes) required to comply with ACTICN requirements, i

E. When a system, subsystem, train, component or devce is E. Surveillance Requirements for inservice inspection and testing of determined to be inoperable soiety tm, its (,Tegecy components sha!! be applicable as followc:

power source is inoperable, er solely because its normal pcwer source is inoperable, it may be considered OPERABLE for the 1. Inservice inspection of wrigonents and inservice testing purpose of satisfying the requrrements of its appicable Umiting of purnps and valves sha!! be performed in accordance Condition for Operation, provided- (1) its corresponding normal with Section XI of the ASME Boiler and Pressure Vessel or emergency power source is OPERABLE; and (2) all of its Code and app!icable Addenda as required by 10 CFR 50, redundant system (s), subsystem (s), train (s), coirpent(s) and Section 50.55a(g), except where specific written relief has device (s) are OPERABLE, or likewise satisfy the requirements of been requested of the Commission pursuant to 10 CFR 50, this specification. Unless both conditions (1) and (2) are Section 50.55a(g)(6)(i).

satisfied, the unit shall be placed in COLD SHUTDOWN within i the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This specification is not applicable when in Cold Shutdown or Refuel Mode.

4 Amendment No. ,

20a

JAFNPP 4.0 Conthued

2. Surveillance intervals specified in Section XI of the AS?/E Boiler and Prr , sure Vessel Code and applicable Addenda for the inservice inspection and testing actmties required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications-ASME Boiler and Pressure Vessel Code and epfI-;.able Required frequencies for Addenda terrninologf for performinginservice inservice inspection and inspection and testing

! testing activities activities Weekfy Atleast once per 7 days Monthly At least once per 31 days

! Quarterly or every 3 rrasdir3 Atleast once per 92 days Senr. annually or every 6 Atleast once per 184 days l months Every 9 months Atleast once per276 days Yearly or annuaRy Atleast once per 366 days i

3. The provisions of Specification 4.0.B are applicable to the atxne required frequencies for performing inservice inspection and testing activities

4. Performance of the above inservice inspection and testing activities sha!! be in addition to other specified Surveil:ance l Requirements.

5. Nothrng in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requiremeres of any Technical Specification.

. Amendment No. % %,

30b l

JAFNPP l

30 BASES A. This specification states the applicability of each specification D. Con M The intent of this provision is to insure that facility operation is

, in terms of defined OPERATIONAL CONDITION (rnode) and is provided to delineate specifically when each specification is not trutsated with either required equipment or systems .

applicable. inoperable or other limits bemg exceeded.

8. This specification defines those conditions necessary to Exceptions to this provision may be made for a limited number i constitute compliance with the terms of an individual Umiting of specifications when startup with inoperable equipment  ;

Condition for Operation and associated ACTION requirement. would not affect plant safety. Thesa exceptions are stated in t'w i ACTION statements of the appropnate specifications. [

., C. This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION E. This specification defineates what additional conditions must 4 statements and whose occurrence would violate the intent of be satisfied to permrt operation to continue, consistent with the  :

I the specification. Under the terms of Specification 3.0, the ACTION statements for power sources, when a normal or facility is to be placed in COLD SHUTDOWN within the euupa,y power source is not OPERABLE. It specifica!!y .

following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It is assumed that the unit is brought to the prohibits operation when one division is inoperable because its  ;

required OPERATIONAL CONDITION (mode) within the normal or emergency power source is inoperabic at a required times by prompt'y initiating and carrying out the system, subsystem, tra;n, ccopenent or device in another i appropriate ACTION statement. division is inoperable for aTther reason. j I

D. This specification provides that entry into an OPERABE The provisions of this specification perrrut the m.,' TION CONDITION (mode) must be made with (a) the full statements associated with individual systems, subsystems, [

complement of required systems, equipment or components trains, components or devices to be consistent with the  !

OPERABW and (b) all other parameters as specified in the ACTION statement of the associated electrical power source. It f Umiting Conditions for Operation being met without regard for allows operation to be govemed by the time allowable deviations and out of service provisions contained in j the ACTION statements. l L

t Arnendmerf No. jd, aoc  :

i l t

JAFNPP 3.0 BASES - Continued E Continued E Continued limits of the ACTION s'atement associated with the Umiting As a further example, Specification 3.9A requires in part that Condition for Operation for the normal or emergency power two 115KV lines and reserve station trimisrcss be available.

source, and not by the individual ACTION statements for each The ACTION statement provides a 7 day out-of-service time system, subsystem, train, coccponent or device that is when both required offsite circuits are not OPERABLE If the determined to be inoperable solely because of the inoperability definition of OPERABE were applied without consideration of of its normal or emergency power source. Specification 3.0.E, all systems, subsystems, trains, ccTpcacnts and devices suppliwi by the incpwade normal For example, Specification 3.9A requires in part that both p w se, M of N oMe his, Wd also M j emergency diesel generator systems be OPERABLE The inopwaNe. Ms M dcwe Mng h We ANN ACTION statement provides for a ~/ day cut-of-service time when s une & d a@@ M Howwer, h emergency diesel generator system A or 8 is not OPERABE. If p ms d WMm 3.0E W h Ume W fa the definition of OPERABW were applied without consideration opeaSm to M mW

• h AGON sWemW of Spccification 3.0.E. all sys* ems, subsystems, trains, fa mopeaNe nmna! power m WM WM h components and devices supplied by the inoperable emergency ther specified conditions are satisfied. In this case, this would power source, diesel generator system A or 8, would also be inoperable. This would dictate invoking the appficable ACTION mean Ma me h, on N wnsgqpow m M M statements for each of the apo!icable Limiting Conditions for ERABE (as must M h womno Wed W h eim p g po w mee) and au rh systems, Operation. However, the prwisions cf Specification 3.0.E subsystems, trains, ecccycnents and devices in the other permrt the time limits for continued operation to be corcistent m must M TME or Rwn Ms4 S#catim with the ACTION statement for the inoperable emergency diesel 3.0.E % M ca@e d Weg Mr Mgn Wons W generator system instead, provided the other specified an emagg pm mco TMS In ch M, conditions are satisfied. If they are not satisfied, shutdown is wnsgency pm m A W B e M @ME required in accordance with this specification. and a!! redundant systems, subsystems, trairs, components anc' devices in both divisions Amendment No. jd,

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JAFNPP J.0 BASES - Continued E. Coritinued must also be OPERABLE. If these conditions are not satisfied, shutdown is required in accordance with this specification.

In Cold Shutdown and Refuel Modes, Specification 3.0.E. is not applicable, and thus the individual ACTION statement for each applicable Umiting Condition for Operation in these OPERATION AL CONDITIONS (modes) must be adhered to.

l Amendment No. f, aoe i

JAFNPP 4.0 BASES D. This specification ensures that surveiltance ac&vities associated A. This specification provides that surveillance activities necessary to insure the Umiting Conditions for Operation are met and will be with a Umiting Condition for Operation have been performed within the specified time interv prior to entry into an applicable performed during the OPERATIONAL CONDITIONS (modes) for OPERATIONAL CONDITION (mode). The intent of this provision which the Umiting Conditions for Operation are applicable. is to ensure that survemance acbviEcs have been satisfac'ai!y Provisions for additiona! surveillance actrvities to be performed demonstrated on a current basis as required to meet the without regard to the applicable OPERATION /L CONDITIONS OPERABlUTY requiremerra of t% Umiting Condition for (modes) are provided in the individual Surveitlance Requirements.

Op d on.

B. The provisions of this specification provide allowable tolerances Under the terms of this specification, for example, du ing initial for performing surveiitance activities beyond those specified in the normal surveillance interval. These tolerances are necessary to plant start-up or foi!owing extended p0 ant outage, the applicable survei!!ance achvities must be performed within t% stated provide operational flexibility because of scheduling and surveillance interval prior to placing or retuming the system or performance considerations.

5pmm G status.

C. The provisions of this specification set forth the criteria for determination of compliance with the OPERABluTY requirements E. This speci5 cation ensures that inservice inspection of co npanents and inservice testing of purrps and valves will be performed in of the Umiting CondiSons for Operation. Under this criteria, accordance with a periodicaffy updated version of the plant equipment, systems or ccingonents are assumed to be

• Inservice Testing Program

• and the "Weid and Support inservice ,

OPERABLE if the associated survei!!ance actrvities have been I satisfactorily performed within the specified time interval Nothing Inspection Prow aT.* to compty with Section XI of the ASME Boiler and P essure Vessel Code and Addenda as regired by 10 CFR in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found 50 W The plant programs identify classifications required by the or known to be inoperable although sti!! meeting the Survei!!ance ASME code. Request for relief from any of the above regturements is provided in wnting to the Commission and is not a Requirements.

part of these Technical Speci5 cations.

This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessef Code and applicable Addenda. This clarificatbn is provided to ensure consistency in surveillance intervais throughout these Technical Soecifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

Amendment No. 4/. [, [,1)h,1[ = I

JAFNPP 4.0 Continued Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precede.n over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specification 4.0.D to perform surveillance activities prior to entry into an OPERATIONAL CONDITION or other specified appiscability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which a!!aws pumps to be tested up to one week after retum to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of perforn-ing its specified function is declared snoperable and takes precedence over the ASME Boiler and Pressure Vessel provision which aflows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared inoperable.

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Amendment No.

3Og i

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JAFNPP 3.7 UMITING CONDITIONS FOR OPERATION 4.1 SURVEILLANCE REOUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicability- Applicability-Applies to the instrumentation and associated devices wtiich initiate the Applies to the surverflance of the instrumentation and associated reactor scram, devices which irnttate reactor scram.

Objective: Objective:

To assure the operability of the Reac*or Protection System. To speerfy the type of freqwncy of survedlance to be applied to the protection instninw&Uon.

4 Specification" Specification-  !

A. The setpoints, minimum number of trip systems, minimum A. Instrumentation systems shall be functiona!!y tested and number of instrument channels that must be operable for each calibrated as indicated in Tables 4.1-1 and 4.1-2 respectively.

position of the reactor mode switch shall be as shown on Table  !

3.1-1. The design system response time from the opening of the  !

sensor contact to and including the opening of the trip actuator >

contacts sha!! not exceed 50 msec. .

8. Minimum Critical Power Ratio (MCPR) B. Maximum Fraction of Urniting Power Densrty (MFLPD) ,

During reactor power operation, the MCPR operating limit shall The MFLPO sha!I be determined daily dunng reactor power ,

not be less thEn that shown in the Core Operating Umits Report. operation at >25% rated thermal power and the APRM high flux i

= cram and Rod Block trip settings acqusted if necessary as

1. During Reactor power operation with core flow less than specified ir the Core Operating Umits Report.

j 100% of rated, the MCPR operating limit sha!! be multiplied i by the uppopiate K, as specified in the Core Operating  ;

Umits Report.

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l Amendment No.

! 30h I

l

~

JAFNPP 3.4 UMITING CONDITIONS FOR OPERATION 4.4 SURVElu.ANCE REQUIREMENTS 3.4 STANDBY UQUID CONTROL SYSTEM 4.4 STANDBY UQUID CONTROL SYSTEM Applicability: ApplicaMity-Applies to the operating status of the Standby Uguid Control System. Applies to the periodic testing requirements for the Standby Uguid Cortrol System.

Objective: Objective To assure the availability of a system with the capability to shut down the To verify the operab0ity of the Standby Uqtad Control System.

reactor and maintain the shutdown condition without control rods.

Specification- Specificatiort A. Normal Cperation A. Nonna! Operation During periods when fuel is in the reactor and prior to startup The operability of the Standby Uguid Ccxtrol System shall be from a cold condition, the Standby Uquid Control System sha!! verified by performance of the following tests:

be operable except as specified in 3.4.B below. This system need not be operable when the reactor is in the cold condition, 1. At least once every three months-all rods are fully inserted and Specification 32.A is met.

l Deminera!ized water sha!! be recycled to the test tank.

Pump minimum flow rate of 50 gpm shall be duriws ded l against a system head of > 1,275 psig.

2. At least once dunng each operating cycle -

Manua!!y initiate the system, except the explosive valves and Amendment No.1)d, 1- 105 i

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^

JAFNPP

AVIS requirements are satisfied at all concentrations above 10 The relief valves in the Standby Uguid Control System protect weight percent for a minimum enrichment of 34.7 atom percent the system piping and positive displacement pumps, which are of B-10. nominally designeo for 1,500 psig, from overpressure. The pressure relief valves discharge back to the standby liquid Figure 3.4-1 shows the permissible region of operation on a control pump suction line.

sodium pentaborate solution volume versus concentration graph. This curve was developed for 34.7% enriched B-10 and i a pumping rate of 50 gpm. Each point on this curve provides a Operation with Inoperable CWowts B.

minimum of 660 ppm of equivalent natural boron m the reactor vessel upon injection of SLC solution. At a solution volume of Only one of two standby liquid control pumping circuits is i 2200 gallons, a weight concentration of 13 % sodium neMM W wh if a cid is @abig he is m pentaborate, enriched to 34.7% boron-10,is needed to meet immediate threat to shutdown capability, and rocctor operation shutdown requirements. The maximum storage volarne of the may continue during repairs. Assurance that the remaining solut 4780 gations which is the net overflow volume in the spiem will ph its fe is obtainM W miyng m operability in the operable circuit at least daily.

Boron concentration, isotopic enrichment of boron-10, solution temperature, and volume are checked on a frequency

adequate to assure a high reliability of operation of the system C. Sodium Pentaborate Solution should it ever be required.

I To guard against precipitation, the solution, including that in the The only practical time to test the Standby Uguid Control pump suction piping, is kept at icast 107 above saturation System is during a refueling outage and by initiation from local temperature. Figure 3.4-2 shows the saturation temperature stations. Components of the system are checked periodica!!y including 107 margin as a function of sodium pentaborate as described above and make a functional test of the entire solution cern,entration. Tank heater and heat tracing system

, system on a frequency of more than once cach refueling are provided to assure compliance with this requirement. The outage unnecessary. A test of explosive charges from one set points for the automatic actuation of the tank heater and manufacturing batch is made to assure that the charges are heat tracing system are established based on the solution satisfactory. A continuous check of the firing circuit continuity concentration. Temperature and liquid level alarms for the is provided by pilot lights in the control room. system annunciate in the control room. Pump operabiiity is checked on a frequency to assure a high reliability of operation of the system should it ever be required.

i Amendment No. [,1[,1[,

109

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

b. Flow Rate Test - Once/3 months l Core spray pumps shall deliver atIcast 4,625 gpm against a system head corresponding to c reactor vessel pressure greater than or equalto 113 psi above primary containment pressure. ,

c. Core Spray Header l l

Ap Instrumentation Check Once/ day Calibrate Once/3 months Test Once/3 months

d. Logic System Once/ operating ]

FunctionalTest cycle I

c. Testable Check Tested for operability [

Valves any time the reactor is in the cold condition exceed:ng 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 92 days. l Amendment No. pd,1fs 113

JAFNPP 3.5 (cont'd) 45 (cont'd)

From and after the date that one of the Core S,xay 2. When it is determirnv" that one Core Spray System is 2.

inoperabic, the operaore Core Spray System, and both Systems is made or found inoperable for any reason, continued reactor operation is permissible during the LPCI subsystems, sha!1 be verified to be operable succeeding 7 days unless the system is made operable immediately. The remaining Core Spray System sha!! be earlier, provided that during the 7 days all active verified to be operable daily thereafter.

components of the other Core Spray System and the LPCI System shall be operable.

Both LPCI subsystems of the RHR System sha!I be 3. LPCI System testing sha!! be as specified in 4.5.A.1.a. b, d, 3.

and o except that each RHR pump sha!! deliver et least i opcrab!c whenever irradiated fuel is in the reactor and prior to reactor startup from a cold condition, except as 8,910 gpm against a system head corresponding to a reactor vessel to primary containment differential pressure specified below.

at greater than or equal to 20 ps.d

a. From the time that one of the LPCI subsystems is

a. When it is determined that one LPCI subsystem is mada or found to be inoperable for any reason, continued reactor operation is permissible during the inoperable, the operable LPCI subsystem and both succeeding 7 days unless that subsystem is made Core Spray Systems shal1 be verified to be operable operable earlier provided that during these 7 days immediately and daily thereafter.

the operab!e LPCI subsystem and both Core Spray Systems sha!! be operabic.

Amendment No. 1/, pd, p[,1[,1[,1jd, [1, 114 I

9 k

i:1

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

5. All recircu!ation pump discharge valves sha3 be operable 5. A!! recirculation pump discharge valves shal be tested for prior to reactor startup (or closed if permitted elsewhere in operabirdy any time the reactor is in the cold condition these specifications). 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 92 days. l l S. If the requirements of 3.5A cannot be met, the reactor shall be placed in the cold condition within 24 hrs.

Containment Cooling Mode (of the RHR System) 8. Containment Coofing Mode (of the RHR System)

B.

1. Both subsystems of the containment cooling mode, each 1. Subsystems of the containment cooling modo shall be including two RHR and two RHRSW pumps, shall be demonstrated operable by performing-operable whenever there is irradiated fuel in the reactor a. a pump operability and flow rate test on the RHR vessel, prior to startup from a cold condition, and reactor. umps per Survei!!ance Requirement 4.5.A.3.

g coolant temperature >212*F except as specified below.

l b. a flow rate test at least once every 3 months I demonstrating a flow rate of 4000 gpm for each l RHRSW pump and a total f ow rate of 8000 gpm for two RHRSW pumps operating in paraitet.

c. During each five-year period, an air test sha!! be l performed on the containment spray headers and nozzles.

Amendment No. [,[ 1[,1[,1[,1[, 1153

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

DEUETED C. HIGH PRESSURE COOLANT INJECTION (HPCf SYSTEM)

C. HIGH PF. ESSURE COOLANT INJECTION (HPCI SYSTEM)

Survei!!ance of HPCI System shall be performed as follows provided a reactor steam supply is available. If steam is nnt availabio at the time the surveillanco 1est is scheduled to be performed, the test sha!! be performed within 10 days of continuous operation from the time stectr. ~oec

• 7.as available.

1. HPCI System testing shall be as specifi6d in 4.5A.1.a, b, d

1. The HPCl Systerii shall be operable whenever the reactor and o except that the HPCI pump shall deliver at least pressure is greater than 150 psig and reactor coolant temperature is greater than 212*F and irradiated fuel is in 4,250 gpm against a system head ctrresponding to a reactor vessel pressure of 1,120 psig to 150 psig.

the reactor vessel, except as specified belove.

l Amendment No. [,[,1F[7, . 11e

JAFNPP 4.5 (Cont'd) 3.5 (Cont'd)

E. Reactor Core Isolation Coofing (RCIC) System E. Reactor Core Isolation Cooling (RCIC) System The RCIC System shall be operable whenever there is 1. RCIC System testing sha!! be performed as fo!!ows 1.

provided a reactor steam supply is available. If stear.: is irradiated fuel in the reactor vessel and the reactor pressure is greater than 150 psig and reactor coolant not available at the time the survei!!ance test is schedded ,

i to be performed, the test shall be performed within ten {

temperature is greater than 212*F except from the time that the RCIC System is made or found to be inoperable days of continuous operation from the time steam l for any reason, continued reactor power operation is becomes avaliable.

permissibio during the succeeding 7 days unless th item Rwg-system is made operable eart:er provided that during these Once/ operating

a. Simulated Automatic ,

7 days the HPCI Systemis operable.

^ " }

2. If the requirements of 3.5.E cannot be met, the reactor T t 3 sha!! be placed in the cold condition and pressure less

b. Flow Rate Test - Once/3 months than 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The RCIC pump shall deliver

3. Low power physics testing and reactor operator training atleast 400 gpm at a system shall bc permitted with anoperable componen+s as head corresponding to a specified i,n 3.5.E2 above, provided that reactor coolant reactor pressure of 1120 psig temperature is <212'F. to 150 W

c. Testable Check Tested for operability l Valves 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 tes's have not been performed I

during the proceding 92 days. l

d. Logic System Once/ operating l FunctionalTest cycie

• Automatic restart on a low water Icvel signal which is subsequent to a high water level trip.

Amendment No. % 1)d,1)6, 121

i JAFNPP 3.5 (cont'd) 4.5 (cont *d)

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2. When it is determ!ned that the RCIC System is hwobie at a time when it is required to be operabic, the HPCI System shaft be verified to be operable immediately and da!!y thereafter.

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Amendment No. f,1[, . 121a

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

ECCS-Cold Condition F. ECCS-Cold Condition F.

Survei!!arre of the low pressure ECm systems required by 3.5.F.1 and 3.5.F2 shall be as foitows:

A minimum of two low pressure Emergency Core Coo!ing 1. Perform a !!cwtate test at least once every 3 months on the 1.

subsystems shall be operable whenever irradiated fuel is in the required Core Spray pump (s) and/or the RHR pump (s). Each reactor, the reactor is in the cold condtion, and work is being Core Spray pump shafi deriver at least 4,625 gpm against a performed with the potential for draining the reactor vessel system head corresponding to a reactor vessel pressure greater than or ecual to 113 psi above pnmary containment pressure.

Each RHR pump shal1 deliver at least 9900 gom against a system head corresponding to a reactor vessel to primary containment differential pressure of > 20 psid.

I A minimum of one low pressure Emergency Core Cooling 2. Once each shift verify the suppression pool water level is greater l 2.

than or equal to 10.33 ft whenever the low pressure ECCS subsystem sha!! be operable whenever irradiated fuel is in the reactor, the reactor is in the cold cond; tion, and no work is being subsystems are aligned to the suppression pool.

performed with the potential for draining the reactor vessel. l

3. Once each shift verify a minimum of 324 irches of water is ]

3. Emergency Core Coo!ing subsystems are not required to be availab!e in the Condensate Storage Tanks (CST) whenever tM operable provided that the reactor vessel head is removed, the Core Spray System (s) is aligned to the tanks.

cavity is f' coded, the spent fuel pool gates are removed, and the water level above the fuel is in accordance with Specification 3.10.C.

4. With the requirements of 35.F.1,3.5.F.2, or 3 5.F.3 not satisfied, suspend core afterations and all operations c h the potential for draining the reactor vessel. Restore at least one system to operable 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 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.[1[,1[, 122

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

G. Maintenance of Filled Discharge Pipe G. Maintenance of Filled Discharge Pipe Whenever core spray subsystems, LPCI subsystems, HPCI, or TM following surveittance requirements shall be adhered to, in RCIC are required to be operable, the discharge piping from the arder to assure that the discharge piping of the core spray pump discharge of these systems to the last block vafve shaft be subsystem, LPCI subsystem, HPCI, and RCIC are fi!!ed-filled.

l l 1. From and after the time that the pump discharge piping of 1. Prior to the testing of the LPCI subsystem and core spray l the HPCI, RCIC, LPCI, or Core Spray Systems cannot be sitsystem, the discharge piping of t.We systems sha!! be maintainedin a filled vented from the high pouit, and water flow observed.

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)

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Amendment No.1[,

122a l

JAFNPP

~4.5 BASES The testing interval for the Core and Contamment Cooling With components or subsystems out-of-service, overa!! core Systems is based on a quanthative reliability analysis, industry and containment cooling reI%aity is maintained by verify:ng practice, judgement, and practicality. The Emergency Core the operabmty of the remaining cooling equipment. 'M'ent Cooling Systems have not been designed to be fuity' testable with the definition of operable in Section 4.0.C, de.T.cashate during operation. For example, the core spray final admission means conduct a test to show; verify means that the valves do not open until reactor pressure has fallen to 450 psig; associated. survei!!ance activities have been satisfaciony

- thus, during . operation even if high drywell pressure were performed within the specified time intervai.

simulated, the final valves would not open. In the case of the HPC!, automatic initiation during power operation would result The surveillance requiremerta to ensure that the discharge ing cold water into the reactor vessel which is not piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC Systems are fi!!ed provides for a visual otnervation that water flows from a high point vent. This ensures that The systems will be automatically actuated durir; a efueling outage. In the case of the Core Spray System, condensate

,. storage tank water will be pumped to the vessel to verify the operability of the core spray header. Individual ccaripenents of the Core and Containment CooF.ng Systems (e.g.,

j instrumentation, pumps, valve operators, etc.) are tested more j frequently. The instrumentation is functional!y tested each i month. The pumps and motor-operated vanes are tested once l every 3 months to assure their operability. The combination of

! automatic actuation tests and quarterly tes's of the pumps and

vane operators is adequate to demonstrate availability of these l

systems.

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Amendment No. [,1f6, 132

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

. StructuralIntegrity F. StructuralIntegnty The structural integnty of the Reactor Coolant System shall be 1. The requirements of Specrfication 4.0.E are applicable. l maintained at the level required by the original acceptance standards throughout the life of the Plant.

1 M awed h kWm Nmis Wred W those high stressed circumfererfJa! piping joints in the mrJn sicam and feedWater lines larger than 4 inches in diameter, where no restraint against pipe whip is prowJed.

The augmented in-service inspection pcgfrsT. sha!!

consist of 100 pc, cunt inspection of these welds per inspectioninterval.

3. An inservice inspection Program for piping identified in the NRC Genenc Letter 88-01 sha!! be implemented in accordance with NRC s*J positions on schedules, methods, peiss sol, arid sample expansion included in this Genenc Letter, or in acordance with attemate measures approved by the NRC s'd.

G. Jet Pumps G. Jet Pumos Whenever the reactor is in the startup/ hot standby or run Whenever there is recirculation flow with the reactor in the modes, a!l jet pumps shall be operable. If it is determined that a startup/ hot standby or run rnodes, jet pump operability sha!! be jet pump is inoperable, the reactor shall be placed in a cold checked daily by venfying that the foi:owing condi'Jons do not condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. occur simuftaneousty-Amendment No. g,1p,1/6, a

JAFNPP 3.7 (cont'd) 4.7 (cont'd) breaker is sooner made operable, provided that the repair procedure does not violate primary containment integrity.

5. Pressure Supprestion Chamber - Drywell Vacuum Breakers 5. Pressure Suppression Chamber - Drywell Vacuum Breakers

a. ' When primary containment integrity is required, all drywell a. Each drywell suppression chamber vacuum breaker shall suppression chamber vacuum breakers shall be operable be exerc: sed through an opening - dosing cycle quarterly. l

. and positioned in the fu!!y closed position except during testing and as specified in 3.7.A.5.b below.

b. One drywell suppression chamber vacuum breaker may b. When it is determined that one vacuum breaker is be non-fully closed so long as it is deterrnined to be not inoperable for fu!!y closing when operability is required, the more than 1* open as indicated by the position lights. operable breakers sha1 be exercised immediately, rid every 15 days thereafter until the inoperable valve has been retumed to normal senrice.

c. One drywell suppression chamber vacuum breaker may c. Once each operating cycle, each vacuum breaker valve be determined to be inoperable for opening. sha!! be visua!!y inspected to insure proper maintenance and operation.

d. If specifications 3.7.A.5.a. b, and c cannot be met, an d. A leak test of the drywe!I to suppression chamber structure orderly shutdown will be initiated, and the reactor shall be shall be conducted once per operating cycle; the placed in a cold condition. acceptable leak rate is <025 in. water / min, over a 10 min period,with the drywcIl at 1 psid.

Amendment No. %,

178

JAFNPP 3.7 (cont'd) 4.7 (cont'd)

e. Leakage between the drywell and suppression c. Not applicable chamber shall not exceed a rate of 71 sefm as monitored via the suppression chamber 10 rnin pressure transient of 0.25 'n. water / min.

f. The self actuateu vacuum breakers shall opm when f. Not applicable subjected to a force equivalent to 0.5 psid acting on the valve disc.

g. From and dier the date that one of the pressure g. During each refueling outage cach vacuum breaker suppression chamber /drywell vacuum breakers is shall be tested to deteridne that the force required made or found to be inoperable for any reason, the to open the vacuum breaker does not exceed the vacuum breaker sha!! be locked closed and reactor force specified in Specification 3.7./L5.f.

operation is permissible only during the succeeding seven days unless such vacuum breaker is em made operable, provided that the repair procedure dces not violate primary containment integrity.

Amendment No.[ %,

179

JAFNPP 3.11 (cont'd) 4.11 (cont'd)

D. Emergency Service Water System D. Emergency Service Water System To ensure adequate equipment and area cooling, both 1. Surveillance of the ESW system shall ba pwformed as 1.

ESW systems shall be operable when the requirements of follows:

specification 3.5.A and 3.5.8 must be satisfied, except as Frequency Item specified below in specification 3.11.D.2.

a. Simulated Automatic Once/ operating Actuation Test Tjcie

b. Ihv Rate Test- Once/3 months Each ESW pump shall deliver at least 1607 gpm to its respective loop against a total developed system head equal to or greater then the ASME Section XI actionlevelon the pump curve. f

c. ESWInstrumentation l Check Once/ day Calibrate Once/3 months Test Once/3 months

d. LogL' System Once/ operating FunctionalTest cycle Amendment No. [ 1p, 240

JAFNPP 4.11 (cont'd) l 3.11 (cont'd)

\ \

From and after the time that one Emergency Service Water 2. ESW wi!! not be supplied to RBCLC system urk.g testing. i

2. l System is made or found to be inoperable for any reason continued reactor operation is permissible for a period not to exceed 7 days total for any calendar month, provided that:

- the operable Emergency Diesel Generator System is demonstrated to be operable immediately and daily thereafter; and

- all Emergency Diesel Generator System emergency loads are verified operable immediately and daily thereafter.

3. If specification 3.11.D2 cannot be met an orderly shut down shall be initiated and the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No. ,

241

JAFNPP 3.11 & 4.11 BASES Main Control Room Ventilation System 8. Crescent Area Ventilation A.

One main control room emergency ventilation air supply faa Engineering analyses indicate that the temperature rise in safeguards compartmerfs without adequate ventilation flow or provides adequate ventilation flow under accident conditions.

Should one emergency ventilation air supply fan and/or fresh air cooing r is such that continued operation of the safeguards filter train be out of service during reactor operation, a repair time equipment or associated auxiliary couipment cannot be assured.

cf 14 days is allowed because during that time, a redundant 100% capacity train is required to be cperable. C. Battery Room Ventitation The 3 month test interval for the main control room emergency Engineering analyses indicate that the temperature rise and ventitation air supply fan and dampers is sufficient since two hydrogen buildup in the battery, and battery charger redundant trains are provided and neither is normally in compartments without adequate ventilation is such that pera h continuous operation of equipment in these compartments A pressure drop test across each filter and across the filter cannot be assured.

system is a measure of filter system condition. DOP injection measures particulate removal efficiency of the high efficiency D. Emergency Service Water System particulate fitters. A Freon-112 test of the leakage test. Since the The ESWS has two 100 percent cooling capacity pumps, each filters have charcoa! of known efficiency and holding capacity for powered from a separate standby power stpply. The ESWS elemental iodine and/or methyl iodine, the test also gives an indication of the relative efficiency of the installed system. supplies take water to the cooling systems of the emergency diesel generators and other components required to function Laboratory analysis of a sample of the charcoal filters positively fo!Iowing an accident. The system can 8:so supply components demonstrates halogen remova! cfficiency. These tests are of the RBCLCS. Performance of the Surveillance Requirement I conducted in accordance with manufacturers' flow rate test will demonstrate pump hydrauiic capability when recommendations. meeting the ASME Section XI requirements.

The purpose of the emergency ventilation a.ir supply system capacity test is to assure that sufficient air is supplied to the main control room so that a slight positive pressure can be maintained, thereby minimizing in-leakage.

l Amendment No. , ,

243

Attachment 11 to JPN 91-064 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES ASME SECTION XI AND ESW PUMP SURVEILLANCE TESTING (JPTS-90-023)

1. DESCRIPTION OF THE PROPOSED CHANGES This application for an amendment to the James A. FitzPatrick Technical Specifications addresses throo associated issues: Emergency Service Water (ESW) pump surveillance testing; incorporation of ASME Section XI, and; editorial corrections.

A. ESW Pump Surveillanco Requirements / Flow Rato

1. On page 240, Surveillanco Requiroment 4.11.D.1.b, replaco: r

" Flow Rato Test - ESW pumps shall deliver at least 3,250 gpm against a system head corresponding to a total purnp head of 180 psi, as determined from the pump certification curve by measuring the pump shutoff head which shall be 1117 psi."

with

• Flow Rato Test Each ESW pump shall deliver at least 1607 gpm to its respectivo loop against a total developed system head equal to or greater than the ASME Section XI action level on the pump curvo.'

2. On pago 243, Bases Section 3.11 & 4.11 D., replace:

"The ESWS utilizes lako water to the cooling system of the emergency diesel generators. The system will also supply water -

to those components of ;he RBCLCS which are mquired for omergency conditions during a loss of power condition. Theso includo ECCS pumps and area unit coolers" with "The ESWS supplies lake water to the cooling systems of the emergency diesel generators and other components required to function following an accident. The system can also supply compononts of the RBCLCS. Performanco of the Surveillanco Requiremont flow rato test will demonstrato pump hydraulic capability when meeting the ASME Section XI requirements.'

B. Incorporation of ASME Section XI

1. Reviso page i to show Specification 3.1, Reactor Protection System, located on page 30h to reflect the renumbering of pages in item 2.

Attachment 11 to JPN 91-064 SAFETY EVALUATION Page 2 of 19

2. Renumber existing pages 30b,30c,30d,300, and 30f to read 30c,30d,300,30f, and 30h, respectively. The changes described in the following items 3 and 4 refer to these renumbered pages and indicate where new pages are inserted.

3. Add a new Surveillance Requirement 4.0.E by revising page 30a and adding a new page 30b which include the following:

• E. Surveillance Requirements for inservice inspection and testing of components shall be applicable as follows:

1. Inservice inspection of components and inservice testing of pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been requested of the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).

2. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

i ASME t3 oiler and Prosure Vessel Code and applicable Addenda Required frequencies for terminology for inservice inspection performing inservice inspection and testing activities and testing activities Weekly At least once per 7 days Monthly At least once por 31 days Quarterly or every 3 months At least once per 92 days Samiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days

3. The provisions of Specification 4.0.B are applicable to the above required frequencies for performing inservice inspection and testing activities.

4. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

5. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification."

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Attachment 11 to JPN-91-064 SAFETY EVALUATION Page 3 of 19

4. Add a new Bases Section 4.0.E by revising page 30f and adding a new page 30g which include the following:

" E. This specification ensures that inservice inspection of components and inservice testing of pumps and valves will be performed in accordance with a per!odically updated ve: ;ion of the plant " Inservice Testing Program" and the " Weld end Support inservice inspection Program" to comply with Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. The plant programs identify classifications required by the ASME code. Request for relief from any of the above requirements is provided in writing to the Commission and is not a part of these Technical Specifications.

This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.

Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specincation 1.0.D to perform surveillance activities prior to entry into an OPERATIONAL CONDITION or other specified applicability condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes procedence over the ASME Boils and Pressure Vessel provision which allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared inoperable."

5. On page 105, Surveillance Requirement 4.4.A.1, replace the phrase *At least once per month" with "At least once overy three months."

6. One page 109, Bases Section 3.4.A, delete the sentence:

" Experience with pump operability indicates that monthly testing is adequate to detect if failures have occurred."

Attachment 11 to JPN.91064 i SAFETY EVALUATION Pago 4 of 19

7. On page 113, Surveillanco Requiremont 4.5.A.1; a) Doloto Surveillanco Requiroment 4.5.A.1.c.

b) Deloto Surveillanco Requirement 4.5.A.1.d.

c) Replace "31 days" with "92 days" in Surveillanco Requirement 4.5.A.1.g.

d) Renumber Surveillance Require- .1.0, f, and g to read 4.5.A.1.c, d, and e, respectivolv

8. On pago 114, Survoillanco Requiremont 4.5.A.3, replaco:

"LPCI System testing shall be as specified in 4.5.A.1.a b, c, d, f, and g ..."

with "LPCI System testing shall be as specified in 4.5.A.1.a b, d, and o ..."

9. On page 115a, Surveillanco Roquirement 4.5.A.5, replaco "31 days" with "92 days."

10. On page 115a, Surveillance Requirement 4.5.B.1; a) Delete Surveillance Requirement 4.5.B.1.b.

b) Delete Surveillance Requirement 4.5.B.1.c.1.

c) Renumber Surveillance Requirements 4.5.B.1.c.2 and d to read 4.5.B.1.b and c, respectively.

11. On pago 117, Surveillanco Requirement 4.5.C.1, replace:

• HPCI System testing shall be as specified in 4.5.A.1.a, b, c, d, f, and g . ."

with "HPCI System testing shall be as specified in 4.5.A.1.a, b, d, and o . .*

, Attachment 11 to JPN 91-064 SAFETY EVALUATION Pago 5 of 19

12. On pago 121, Survoillanco Requiremont 4.5.E.1; a) Doloto Survoillance Requiremont 4.5.E.1.b.

b) Deleto Surveillanco Requiremont 4.5.E.1.c.

c) For Survol!!anco Requlroment 4.5.E.1.d roplace:

• Flow Rato Once/3 months" with

" Flow Rato Test - Onco /3 months The RCtC pump shall deliver at least 400 gpm at a system head corresponding to a reactor pressure of 1120 psig to 150 psig.'

d) Replace "31 days" with '92 days" in Survoillance Requiremont 4.5.E.1.o.

e) Renumber Surveillanco Requirements 4.5.E.1.d, o, and f to road

! 4.5 E.1.b, c, and d.

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13. On page 121a, delete the sentenco:

"The RCIC pump shall deliver at least 400 gpm for a system head corresponding to a reactor pressure of 1,120 psig to 150 psig."

14. On page 122, Surveillarico Requirement 4.5.F; a) Delete Surveillanco Requirement 4.5.F.2.

b) Rer. umber Surveillanco Requirements 4.5.F.3 and 4 to road 4.5.F.2 '

and 3.

15. On page 122a, deloto the words "Every month" from Surveillanco Requiremont 4.5.G.1.

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Attachment 11 #a JPN 91-064 SAFETY EVALUATION Page 6 of 19

16. On page 132, Bases Section 4.5, replace in the second paragraph:

"Likewise, the pumps and motor-operated valves are also tested cach month to assure their operability. The combination automatic actuation test and monthly tests of the pumps and valvo operators is doomed to be adequato testing of those systems,"

with "The pumps and motor operated valves are tested onco overy 3 months *.o assure thc4 operability. The combination of automatic actuation tests and quarterly tests of the pumps and valve operators is adequate to demonstrate availability of those systems."

17. On page 144, Surveillanco Requirement 4.6.F.1, replace:

"Nondestructivo inspections shall be performed on the ASME Boiler and Pressure Vessel Code Class 1,2 and 3 components and supports in accordance with the requirements of the wcld and support inservice inspection program. This inservice inspection program is based on an NRC approved cdition of, and addenda to,Section XI of the ASME Boiler and Pressure Vessel Code which is in effect 12 months or less prior to the beginning of the inspection interval."

with "The requirements of Specification 4.0.E are applicabic."

18. On page 178, Surveillance Requirement 4.7.A.5.a, replace the word " monthly" with the word " quarterly."

19. On page 240, Surveillance Requirement 4.11.D.1; a) Delete Survoillanco Requirement 4.11.D.1.c.

b) Delete Surveillance Requirement 4.11.D.1.d.

c) Renumber Surveillance Requirements 4.11.D.1.0 and f as 4.11.D.1.c-and d, respectively, and move from page 241 to 240.

Attachment 11 to JPN 91-064 SAFETY EVALUATION Page 7 of 19

20. On pago 241, move Surveillanco Requirements 4.11.D.1.0 and I to pago 240 as noted in item 19.c.

C. Editorial Corrections 21

1. On pago 30a, Specification 3.0.D, replace the word "thru" v '. 'through" 'i 5

D

2. On page 105, Surveillanco Requirement 4.4.A.1, replace the word "vorified" with ;M

" demonstrated." 6t;'

3. One pago 113; Lh p,

7 ,

a) Replace tho word " Months" with " months

• in Survoillanco Requirement 4.5.A.1.b.

b) Replace the phrase "Once/cach operating cyclo" with

" Onco / operating cycle"in Surveillanco Requiroment 4.5 A.1.1 (ronumberod as 4.5.A.1.b).

4. On page 115a; a) Replace ":" with a " " in Specification 3.5.B.1.

b) Replace the word " verifying" with " demonstrating" in Surveillanco Requirement 4.5.B.1.c.2 (renumbered as 4.5.B.1.b).

5. On pago 122a, Specification 3.5.G.a renumber specification "3.5.G.a" as specification *3.5.G.1.*

6. On pago 132, Bases Section 4.5 in the second paragraph, replace:

"To increase the availabilliy of tho individual components of the Coro and Containment Cooling Systems the components which make up the system i.e., instrumentation, pumps, valvo operators, etc., are tested more frequently."

with

" Individual components of the Coro and Containment Cooling Systems (e.g., instrumentation, pumps, valvo operators, etc.) are tested more frequently."

Attachment 11 to JPN-91-0G4 SAFETY EVALUATION Pago 8 of 19

7. On pago 179, Survoillanco Requiroment 4.7.A.S g, doloto the phraso:

... and each vacuum breaker shall bo inspocted and verified to moet design requiromonts."

y 8. On page 240, Surveillanco Roquiremont 4.11.D.1; E6 3 a) Replace the phrase "Each operating cycle" with *Cnco/ operating

g (

cycle" in Specification 4.11.D.1.a.

$zy .b b) Replaco the ptvase "Once/cach operating cyclo" with 7' p g " Onco / operating cyclo" in Specification 4.11.D.1.1 (renumbered as "W

1 7 %

4.11.D.1.d).

11. PURPOSE OF THE PROPOSED CHANGES A. ESW Pump Survoillanco Requirements / Flow Rate During the August 21,1990 ESW onforcement conference (Reference 5) the Authority identified the limitations of the " shut off head" ESW pump survoillanco tost currently required by the FitzPatrick Technical Specifications. At that mooting, the Authority committed to preparo and submit a Technical Specification chango to require an improved ESW pump tost, Reference 5.

In Referenco 7, the Authority clarified it's commitment and stated that the test requiroments would reflect the appropriato portions of Section XI of the American Society of Mechanical Engineert (ASME) Boiler and Pressuro Vossol (B&PV) Codo and the FitzPatrick inservico inspection and test programs based upon revised flow requirements. Reference 8 provided the scheduto for completion of this action. This ar plication satisfies that commitment, in preparing the proposed Technical Specification chango, the Authority has revised pump flow rate requirements. The flow rate was based on an ovaluation of the minimum required flow to safety related components supported by the ESW during a Design Basis Accident (DBA) using an clovated take temperature, Raforence 6.

B. General Incorporation of ASME Section XI This portion of the amendment submittalimplements ASME B&PV Code Section XI as a surveillance requirement in addressing the inspection and testing of ASME B&PV Codo class 1,2, and 3 components as established by the applicablo sections of 10 CFR 50.55a(g). The purpose of this chango is to eliminato unnecessary testing at power consistent with NRC Commission policy, Reference 11, by consolidating portions of the Technical Specification surveillance test program, inservico Test Program, and Wold and Support inservico inspection Program. The changes will

Attachmont il to JPN 91-064 l SAFETY EVALUATION Pago 9 of 19 assure adequato testing for operability while oliminating component wear due to excessive testing.

This chango replaces the monthly Technical Specification surveillanco requiremont for pumps and valvos with the James A. FitzPatrick ASME B&PV Section XI Inservico Test Program, Referenco 9, in a manner consistent with the Standard Technical Specifications, Roference 10. This chango also revises other Surveillanco Requirements to be consistont with the requirements of ASME Section XI (o.g.,

methodologics for determining referenco data, acceptablo cabbration frequencies, testing of specific paramotors, acceptanco critoria, etc.). The effect will be to climinato unnecessary testing of safety related pumps and valves, particularly during power operation.

C. Editorial Corrections Various editorial or administrativo changes to pages which were the subject of this amendment submittal are made to improve the consistency and clarity of the Technical Specifications.

Ill. SAFETY IMPLICATIONS OF THE PROPOSED CljANGES A. ESW Pump Surveillanco Requirements / Flow Rate The ESW system consists of two independent supply loops cach with an emergency service water pump to provido cooling to the Emergency Coro Cooling System (ECCS) components and other vital equipment required for a safo reactor shutdown. '

In the event of failure of one of the two omergency pumps, the remaining loop can provido sufficient cooling water to support opuation of the minimum required ECCS equipment during a DBA.

The present surveillanco requirement for a flow rato test of the ESW pumps specifics a minimum pump dischargo pressure at zero flow for cach ESW pump (i.e., shut off head test). The proposed surveillance requiremont will overcome the shortcomings of the current test by demonstrating the capability of the pumps to provide flow to the system and by minimizing the wear attributable to shutoff head testing.

The proposed surveillance test will be performed with the ESW pumps aligned to provido flow to components that are required following a design basis accident. The acceptanco criteria used for the proposed surveillance test woro derived considering throo factors: 1. a rocalculation of minimum system flow requirements; 2. an ovaluation of the system hydraulic characteristics, and; 3. the development of procedurcs for the inservico testing (IST) program.

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Attachment 11 to JPN-91-064 SAFETY EVALUATION Page 10 of 19

1. System Flow Requirements The proposed survoillanco test identifies a minimum flow rate of 1607 gpm.

This value is based on a recent calculation of the ESW system flow I requirements necessary to remove heat following a DBA, Reference 6. The revised calculation demonstratos that heat removal requirements are mot when flow to components required to function following the DBA (i.e., omorgency diosol generator jacket, electric bay coolers, crescent area coolers, cable tunnel coolers, control room air handling units, and relay room air handling units) is 1605 and 1607 gpm for trains A and B, respectively.

The minimum flow raio in the proposed surveillanco test is less then either the existing FSAR (soo Tablo 9.71) flow rato requirement of 2915 gpm or the present Technical Specification of 3250 gpm. 2915 gpm is based on full single loop cooling and individual component flows higher than necessary to perform the required function, Reference 6. 3250 gpm is based on the cooling requirement for all components supplied by the ESW and provides a reasonablo allowance for normal pump degradation from the head capacity design curvo.

The proposed minimum flow rate is not inconsistent with either the FSAR or the current technical specification since it represents only that cooling flow to components required for an accident and reduced flow rates, based on recalculated flows from Reference 6, to thoso required components.

A minimum flow rato of 1607 gpm is acceptable for demonstrating pump flow.

2. System Hydraulics The proposed surveillance test identifics the ASME Section XI action lovel on l the pump curve as the basis for determining pump operability when providing ,

the minimum required flow. The proposed acceptance critoria was based on an i ESW test, Reference 24, which demonstrated that each ESW pump could l

provide minimum flow to the components required following the DBA while also supplying RBCLCS components. The RBCLCS components which wero l isolated during this test will romain isolated during normal power operation.

l Changes to the requirements to isolate those RBCLCS components will require rotesting. Calculations, Reference 25, based on test data have further demonstrated that the ESW pumps have margin to operato below the ASME i Section XI action lovel on their pump curves and stili deliver minimum flow to l

components required for the DBA when the RBCLCS components are aligned.

A surveillance test where the minimum flow rato is supplied to required components while the pump is performing at or above the lower action level on the pump curve is acceptable to demonstrate tha hydraulic capability of tho l

pump. A flow test which includes ESW injection into the RBCLCS cannot be performed during plant operation.

. Attachment il to JPN 91064 SAFETY EVALUATION Pago 11 of 19

3. Flow Test Proceduro in addition to the shut off test performod as required by Survoillanco Roquiremont 4.11.D.1.b, the Authority has started to perform additional pump tests, References 12 and 14, that measure pump flow, differential pressure (dp),

and flow through most required components. Those tests moot ASME Section XI, Reference 13, by indicating the ESW pump hydraulic condition. Those tosts will be used to meet the proposed surveillance requirement.

The tests which will bo used to moot the proposed surveillanco requiroment deliver flow to all components required after a DBA except the Control Room and Relay Room air handling units (AHUs). Those AHUs are glycol coolod during normal operation and circulation of normal or emergency service water through the system is minimized to keep them clean and to avoid flushing of glycol to Lako Ontario. The flow path in the proposed survoillance tost will include the Control Room chillor and the Chiller Room AHU which require a flow rate slightly above that of the Control Room and Relay Room AHUs (226 gpm and 254 gpm for trains A and B, respectively as opposed to 200 gpm).

The proposed tests demonstrato minimum ESW flow through required components or their equivalent and are required to provido a pump flow and head that meet, as a minimum, the action level on the pump curve. They are therefore sufficient to demonstrato pump operability, The proposed pump survoillance test exceeds current service water system surveillanco l

requirements of the Standard Technical Specifications, Reference 10. The pump testing proceduro results in an improved indication of pump and system operability whilo reducing stress to the pumps.

The proposed revision to the surveillance requirements will be performed as part of the current program for testing under ASME Section XI and will provido assuranco of the hydrauFc condition of both the pump and system to moet plant accident requirements, The system alignment for testing will includo all components required for a DBA except those which are aligned manually under accident conditions. The t proposed test will demonstrate the capability of the system to perform its intended

! function.

B. ' incorporation of ASME Section XI This amendment adds the requirements and criteria of ASME Section XI into the Technical Specifications as now requirements, removes the surveillance test requirements for pumps and valves that have been replaced by the Section XI program, and revises the testing frequency to be consistent with ASME Section XI.

This change is both administrative and technical in nature. The replacement of i multiple individual test requirements with a single requirement (Section 4.0.E) is an l administrative change which has a negligible impact on plant operations and safety, i

l i

Attachment ll to JPN 91-0G4 SAFETY EVALUATION Page 12 of 19 The extension of the specified surveillance intervals from monthly to quarterly is a technical change.

The FitzPatrick Technical Specifications contain, in part, monthly pump and valve surveillance test requirements for the following systems:

Standby Uquid Control System (4.4.A)

Core Spray System (4.5.A)

Residual Heat Removal System (4.5.B)

High Pressure Coolant injection System (4.5.C)

Reactor Core Isolation Cooling System (4.5.E)

Emergency Core Cooling System (4.5.F) )

Reactor Coolant Systems Structural Integrity (4.6.F) '

l Drywell Vacuum Breakers (4.7.A.5)

Emergency Service Water System (4.11.D)

These FitzPatrick Technical Specifications generally require that pumps and valves be tested once por month. These month!y surveillance tests (i.e., a pump functional test and a valve stroke test), demonstrate system availability by operating the starting circuits and verifying proper equipment operation, have been replaced. They are replaced by the requirements imposed by the new Surveillance Requirement,4.0.E, which incorporates the FitzPatrick inservice testing program, and will result in a quarterly testing cycle in place of the existing monthly tests. Retained are the pump functional tests which establish pump hydraulic operability by confirming an established discharge flow rate or discharge pressure. These tests and tests on other components (e.g., injection line testable check valves, recirculation pump discharge valve, and drywell/ torus vacuum breakers) have been revised to require quarterly testing.

In the late 1960's General Electric (GE), used simplified probabilistic risk techniques to establish a logical basis for both surveillance test intervals and the allowable outage times which are contained in BWR technical specifications. GE Report APED-5736, Reference 16, and a 1968 article from the magazine Nuclear Safety, Reference 17, provide an in-depth discussion of these modeling techniques. These two documents were used in the Bases Sections of the FitzPatrick Technical Specifications as a rationale for the test intervals specified and as a basis for past technical specification requirements on testing redundant systems when in a degraded LCO condition.

These studies established the connection between system availability as a function of failure rates, repair times, and the duration between operability tests. They concluded that frequent system testing would provide greater assurance of system operability since the likelihood of detecting a component suffering from degradation prior to failure was increased.

The testing requirement that resulted from these studies did not increase system availability since a system is classified as being unavailable while tests are being conducted. A trade-off exists between the confidence in a system's operability due to frequent testing and a system's availability due to less frequent testing. This approach did not recognize that a component which is repeatedly tested would experience

, Attachment li to JPN-91-064 SAFETY EVALUATION Page 13 of 19 further degradation compared to a component which is in a static condition awaiting operation.

Following issuance of the FitzPatrick operating license, both the Standard Technical Specifications and the ASME Code were revised to require quarterly pump and valve testing. These changes were based, in part, on concerns for accelerated component aging due to excessive testing and on a better understanding of the relationship between test frequency and component / system availability. These changes eliminated bnnecessary monthly tests which are a burden on plant personnel and result in unnecessary additional wear and tear on the components and equipment in the safety systems, and also reduced the risk of plant transients associated with testing at power.

A reduction in testing would therefore provide the benefits of reducing system unavailability and the associated possibility of a plant transient during such testing at power and reducing component degradation due to extensive testing and the need for down time during component maintenance. Additionally, the ASME tests measure changes in pump and valve performance. Degradation can be detected and corrective action (i.e., further testing, repair, etc.) implemented to provide continuous assurance that safety equipment can fulfill their intended functions. A review of the FitzPatrick FSAR and the Technical Specifications indicates no design basis licensing critoria which would preclude this surveillance test extension.

10 CFR 50.55a(g) requires that the plant's inservice testing program be revised at 120 month intervals. The revised program must use, to the extent practicable, the testing requirements contained in the latast edition and addenda of the ASME Code that is in effect 1 year prior to the 120 month interval. The wording of the proposed Section 4.0.E is general enough to accommodate changes to the inservice test program without requiring future technical specification changes.

The proposed Section 4.0.E is consistent with the Standard Technical Specification requirement that Technical Specification requirements take procedent where they are more stringent. However, the proposed Section 4.0.E does differ from the Standard Technical Specification requirement to comply with ASME Section XI except where relief has been granted. The proposed Section 4.0.E allows deviations from the code where relief has been requested in writing from the Commission. This deviation reflects current practice. Changes are discussed with the NRC staff and formally proposed long before they are formally approved.

The proposed revision to the Technical Specifications is consistent with the Standard Technical Specifications with the exception of the Suppression Chamber to Drywell Vacuum Breaker System. This system limits vacuum in the drywell to meet the drywell-wetwell boundary design differential pressure requirement during negative pressure transients or post accident atmospheric cooldown. Overall operability is based on an "n + 1" design capacity. Standard Technical Specification 4.6.4.1.b contains a monthly stroke test for the suppression chamber to drywell vacuum breakers but establishes no technical basis for requiring the surveillance requirements of these valves to be more frequent than other swing check valves. These valves are 30 inch diameter swing check valves with a counterweight to ensure that the valve

l .

I Attachmont 11 to JPN 91064 SAFETY EVALUATION Page 14 of 19 remains seated until a pressure differential of 0.5 psid exists across the seat in addition to the stroko test requiremont, the valves are currently subject to visual inspections to assure proper maintenanco and operation as well as an operability test every operating cyclo. Revising the frequency of the stroko test to agree with the current quarterly stroke test required by the ASME Section XI program is justifiablo based upon the similarity to other swing check valves testod to this frequency and the lack of any history of poor vacuum breaker performanco at FitzPatrick.

C. Editorial Corrections Changos identified in Section 1 of this amendment submittal as editorial or administrativo changes can bo subgrouped as:

1. Typographicai/ Punctuation Corrections The spelling correction in itom I.C.1 and the punctuation correction in item I.C.4.a will not alter the safety evaluation of the Technical Specifications in any way.

2. Editorial Changes Editorial changes have boon made that clarify the Technical Spocifications.

They includo improvement of word usago (items IS.3. and I.C.7), correction to numeration (item I.C.5), and grammatical corrections (item I.C 6). In all three cases, the changes mado to the Technical Specifications do not entail any changes which would alter the conclusions of the plant's accident analysos as documented in the FSAR or the NRC staff's SER.

3. Administrativo Changes The changes identified in items I.C.2 and I.C.4.b correct the usage of the words

" verify" and " demonstrate" as established in the Technical Specifications by Amendment 148, Reference 20. They climinate the need for redundant and unnecessary surveillance tests that result from overlapping requirements and are consistent with the Authority's interpretation of these words.

These changes update Technical Specification pages that were under review when Amendment 148 was approved and represent an improvement to the consistency of the Technical Specifications.

I.

l Attachment 11 to JPN-91-064 SAFETY EVALUATION Page 15 of 19 IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordanco with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:

1. involvo a significant increase in the probability or consequences of an accident previously evaluated.

The changes identified in this proposed amendment cvise the pump surveillance testing for the ESW system, consolidating surveillance testing for various systems (e.g., ECCS, HPCI, ESW, etc.) with Section XI of the ASME B&PV Code as a basis, and make editorial corrections. None of these changes involves a hardwaro modification to the plant, a change to system operation, a change to the manner in which the system is used, or a chango in the ability of the system to perform its intended function.

The change to the ESW pump surveillance tost represents an improvement in the test process. The proposed testing will measure actual pump flow through the system using a system alignment that will not prevent the system from performing its required function, if required. The proposed flow test required that a new performance criteria be established.

These critoria were developed using recalculated minimum ESW flow requirements as well as the results of a system flow test and calculations which assured that the pumps, operating below tho action level on the pump curve, could provido the minimum flow while other non-safety components were aligned. Procedures control system operation consistent with the performanco criteria. This change allows for an improved demonstration of the ESW pump capability to meet system performance requirements under DBA conditions.

The use of Section XI of the ASME B&PV Codes as a basis for establishing surveillance testing and acceptanco criteria will not alter existing accident analyses. This has boon acknowledged and accepted by the NRC given it's usage in the Standard Technical Specifications, Reference 10. The change to surveillance testing frequencies reduces testing at power, increases the availability of systems important to the mitigation of a DBA, and minimizes component degradation due to excessive testing.Section XI testing tracks component performance allowing identification of component degradation.

The editorial changes are strictly non technical in nature with no impact to existing analyses.

They clarify the Technical Specifications by improving the legibility of this document and updato it to incorporate changes previously approved that were missed due to the nature of the amendment process. These changes, by their nature. do not have any affect.

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

The proposed changes involve no hardware changes, no changes to the operation of the systems, and do not change the ability of the systems to perform their intended functions.

l

1 L ,

Attachment 11 to JPN 91064 SAFETY EVALUATION Page 16 of 19 The procedures for testing the ESW pumps as required by the proposed surveillance requirement include those used to meet IST requirements. The system alignment was considered in system design. The flow rate used to establish the acceptance criteria for the new ESW test is based on current accident analyses.

The use of ASME Section XI as the basis for testing involves no testing alignments or practices not previously used as part of either the IST program or testing performed to Technical Specification requirements.-

The editorial changes have no effect on plant practices.

3. involve a significant reduction in the margin of safety.

There are no hardware modifications, changes to system operations, or effect on the ability of systems to perform their intended function associated with the proposed changes.

The revised surveillance test for the ESW system reduces stress on the system pump so the system remains capable of meeting its DBA commitment. The revised flow rate reduced the flows required for the ESW system to meet its design requirement following a DBA by removing conservatism from calculations to reflect system performance. The capability of the pumps to maintain the minimum flow is the subject of the tests that enhance the demonstration of system and pump operability and reduce pump wear.

The proposed changes to add Section XI of the ASME B&PV Code and remove individual Surveillance Requirements in the Technical Specifications does not relax any controls or limitations. The resulting reduction in test frequency, while reducing the possibility of detecting a degraded component prior to failure, is offset by the increased availability of systems important to plant safety and an associated reduction in component degradation due to excessive testing. Additionally, the ASME testing program evaluates components for degraded performance and will identify such degradation early.

There are no safety margins associated with the editorial corrections.

V. IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed chariges will not adversely affect the ALARA or Fire Protection Program at the FitzPatrick plant, nor will the changes impact the environment. The results of these changes are expected to reduce the dose to plant personnel since the number of tests performed in close proximity to radiological sources will be reduced. The proposed change will not change the testing process currently in place to meet ASME Section XI requirements and therefore can have no impact on the Fire Protection program or the environment.

I

,' Attachment 11 to JPN 91-OS4 SAFETY EVALUATION Page 17 of 19 VI. CONCLUSION This change, as proposed, does not constitute an unreviewed safety questial as defined in 10 CFR 50.59. That is, it:

a. will 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;

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

d. involves no significant hazards consideration, as defined in 10 CFR 50.92.

Vll. REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Sections 4.8,6.0, and 9.7.1.

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

3. ASME Boiler and Pressure Vessel Code,Section XI,1980 Edition through Winter 1981, Articles lWP and IWW.

4. NRC Inspection 50-333/90 May 27 - Jun 301990 - Routine Inspection Plant Activities Violation - Deviation & Unresolved item, dated August 2,1990.

5. NRC letter, C.W. Hehl to W. Fernandez, Received October 9,1990, "Results of the August 21,1990 Enforcement Conference," (Inspection Report 50-333/90-04).

6. James A. FitzPatrick Nuclear Power Plant Safety Evaluation for Clarification of Design Basis Requirements for tne JAFNPP Emergency Service Water System, JAF-SE-90-067, Revision 1, March 6,1991.

7. NYPA letter, W. Fernandez to U.S. NRC, dated April 15,1991, (JAFP-91-0228), " Update of the Status of Activities for the Emergency Service Water System."

8. NYPA letter, R.E. Beedle to the U.S. NRC, dated September 6,1991, (JPN 91045), " Revised Schedule for improved Emergency Service Water System Pump Technical Specifications."

9. Inservice Testing Program for James A. FitzPatrick Nuclear Power Plant, Second inservice interval, Revision 4, dated May 1,1991.

e

, Attachment 11 to JPN-91-OG4 SAFETY EVALUATION Pago 18 of 19

10. NUREG-0123, " Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)", Revision 3, Section 4.7.1.2.d.2, dated Fall 1980.

11. SECY-88-304, " Policy issuo Regarding Staff Actions to Reduce Testing at Power", dated October 26,1988.

12. James A. FitzPatrick Nuclear Power Plant Operations Surveillance Test Proceduro, ST-8N, "ESW Pump inservico Test (IST)," Revision 3, dated December 6,1990.

' 13. ASME Section XI Interpretation: X1 179-19;

Subject:

Section XI, Division 1, Operability I Umits of Pumps,IWP 3210.

I

14. James A. FitzPatrick Nuclear Power Plant Operations Survoillanco Test Proceduro, ST-80,

" Testing of the Emergency Service Water System (IST)," Revision 5, dated July 11,1991.

15. NRC letter, D.B. Vassallo to LW. Sinclair, dated October 29,1982, (JAF 82-270), transmits Amendment 71.

16. APED-5738, " Guidelines for Determining Safo Test Intervals and Repair Times for Engineered Safeguards," dated April 1969.

17. " Reliability of Engineered Safety Features as a Function of Testing Frequency,"

Nuclear Safety Vol. 9, No. 4, July - August,1968.

18. USNRC Standard Review Plan, NUREG-0800, Soction 3.9.6, " Inservice Testing of Pumps and Valves," Revision 2, dated July 1981.

19. NYPA Performance Engiacering Memorandum JPEM 91-001, dated June 13,1991, "ESW Pump Performance Testing and Operability Requirements."

20. NRC letter, D. LaBarge to J.C. Brons, dated January 3,1990, (JAF 90-002) transmits Amendment 148.

21. NYPA letter, P.J. Early to U.S. NRC, dated June 26,1979, (JPN-79-C36), " Proposed Emergency Service Water Pump Testing Technical Specifications."

22. NYPA letter, J.C. Brons to U.S. NRC, dated March 30,1990, (JPN 90-027), " Final Response to NRC Generic Letter 89-04 Regarding Guidance on Developing Acceptablo inservico Testing Programs."

23. NYPA letter, J.C. Brons to U.S. NRC, dated May 31,1989, (JPN-89-034), " Low Pressure Coolant injection Pump Flow Surveillance and Demonstrate / Verify Terminology."

24. James A. FitzPatrick Nuclear Power Plant Temporary Operating Proceduro, TOP-117, " Full Flow Testing of the Emergency Service Water System," dated June 1990.

. Attachment 11 to JPN 91064 SAFETY EVALUATION Pago 19 of 19

25. James A. FitzPatrick Nuclear Power Plant Calculation, [[::JAF-090-102|JAF-090-102]], *ESW Pump Minimum Hoad Requiroments," Revision 0, dated June 10,1991.

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