ML20005F088

Proposed Tech Specs Permitting Continued Plant Operation W/ Two Safety Relief Valves out-of-svc & Changing Setpoints for All Eleven Safety Relief Valves to Single Nominal Setpoint

ML20005F088
Person / Time
Site:	FitzPatrick
Issue date:	12/20/1989
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20005F085	List: JPN-89-084, Forwards Application for Amend to License DPR-59,revising Tech Specs to Permit Continued Plant Operation W/Two Safety Relief Valves out-of-svc & to Change Setpoints for All Eleven Safety Relief Valves to Single Nominal Setpoint ML20005F087 ML20005F088
References
NUDOCS 9001120275
Download: ML20005F088 (23)

Text

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4TTACHMENTI

s PROPOSED TECHNICAL SPECIFlCATION i

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L' NewYork Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59 900112o275 891220 ADOCK0500g33 PDR P

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.JI JAFNPP 22 REACTOR COOT. ANT SYSTEM 12 REACTORCOOLANTSYSTEM APPUCABfUTY:

APPUCABluTY:--

Applies to trip setWrgs of the instruments and devices which are Apphes to limits on reactor coolant system pressure.

pronded to prevent the reactor r=lant system sWety limits from teng exceeded.

OBJECWE:

OBJECTIVE To define the level of the process verleblos et which automatic To estatWesh a limit below which the integrity of the Reactor pre:= achon is initiated to prevent the safety Ilmits from Coolant System is not thrs:A due to an overpressure being exceeded.

condition.

SPECIFICATION.

SPECIFICATION:

The Umihng Safety System setting shall be specmed 1.

The reactor vessel dome pressure shall not exceed 1.325 l

below-1.

psig at any time when irradiated fuel is present in the reactor vessel.

A.

Reactor coolerit high pressure screm sheff be

< 1.045 peig.

B.

At loest 9 of the 11 reactor cooient system sdety/rdlet volves shall have opening pressures less than or equel to an upper Ilmit value of 1195 psig.

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JAFNPP 12 and 22 BASES The reactor coolant presswe boundary integrity is an important The lirrutmg vessel overpressure transient event is a mein steam barrier in the prevention of uncontrolled release of fission isolation valve closure with flux scram. This event was analyzed products. It is essential that the integnty of this boundary be withen NEDCG1697P.

• Updated SRV Performance Requirements protected by establishing a pressure limit to be observed for all for the JAFNPP*, assummg 9 of the 11 SRVs were operable with operating conditions and whenever there is irradiated fuel in the opeik,9 pressures less than or equal to an upper limit value of i

reactor vessel.

1195 psig. The resultant peak vessel pressure for the event was o M hs than % vesd presswe M M of 1,375 The pressure safety limit of 1,325 psig as measured by the vessel

@ (Se cwrM M a6 W e response W steam space pressure indicator is equivalent to 1,375 psig at the e steam he wh h 4 h scram WJ h lowest elevation of the Reactor Coolant System. The 1,375 psig uppw v

d 1195 @ es N W opening presswe up to value is derived from the design pressures of the reactor M

h

, answning 2 m pressure. vessel and reactor coolant system piping. The We N presswes W h respective design pressures are 1250 psig at 575'F for the uppe (119

) enswe M N ASME % M on peak reactor vessel,1148 psig at 568*F for the recirculation suction r

pesswe b M.

piping and 1274 psig at 575'F for the discharge piping. The pressure safety limit was chosen as the lower of the pressure A safety limit is applied to the Residual Heat Removal System transeents pemwtted by the at=* ahle design codes: 1965 (RHRS) when it is operating in the shutdown cooling mode.

ASME Booler and Pressure Vessel Code, Section lit for pressure When operating in the shutdown cooling mode, the RHRS is vessel and 1969 ANSI B31.1 Code for the reactor coolant system included in the reactor coolant system.

piping. The ASME Boiler and Pressure Vessel Code permits pressure transents up to 10 percent over design pressure (110%

x 1,250 = 1,375 psig) and the ANSI Code permits pressure transients up to 20 percent over the design pressure (120% x 1,150 = 1,380 psig). The safety limit pressure of 1,375 psig is i

referenced to the lowest elevation of the Reactor Coolant System.

Amendment No.

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- TABLE 42-2 FAnt'd)

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newfusund TEST Af00 CAUBRADON FREQUEMCY FOR CORE AND CONTAINMENT COOUNG SYSTERAS Loge System Funchonal Test Ffeguency

1).

core Spray Subsystem MM Once/6 months 2)

Low Pressure Coolant injechon Subsystem M (9).

Once/6 months 3)

Containment Coohng Subsystem (9)-

Once/6 months 4)

HPCI Subsystem M (9)

Once/6 months

-5)

HPCI Subsystem Autoisolation M (9)

Once/6 months 6)

ADS Subsystem M (9)

Once/6 monIhs 7) acic Subsystem Autoisolation M (9)

Once/6 months l

NOTE-See listing of notes following Table 42-6 for the notes referred to herein.

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JAFNPP

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35 (cont'd) 45 (cont d)

D.

- Automatic Depressurization System (ADS)

D.

Automatic Depressurization System (ADS) 1.

The ADS shall be operable with at least 5 of the 7 ADS 1.

Surveillcnce of the Automatic Depressurization System valves operable:

shall be performed dunng each operating cycle as follows:

a.

whenever the reactor pressure is greater than 100 a.

A simulated automatic initiation which opens all pilot psig and irradiated fuel is in the reactor vessel, and valves.

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b.

A simulated automahc initiation which is inhebited by b.

prior to reactor startup from a cold condition.

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An erGrmat No. f,p4, )[L 119

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A 3.5 (cont'd).

45 (cont'd)

I 2.

If the requirements of 35.D.1 cannot be met, the reactor shall be placed in the cold condition and pressure less l

than 100 psig within 24 hr.

3.

Low power physics testing and reactor operator training shall be permitted. with inoperable ADS curpurents, provided that reactor coolant temperature is <212*F and the reactor vessel is vented or reactor vessel head is removed.

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JAFNPP 3.5 BASES (cont'd)

D.

Autornatic Depressurization System (ADS)

C.

High Pressure Coolant injection (HPCI) System N relief valves of the ADS are a backup to the HPCI

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The High Pressure Coolant injection ~ System is provided to subsystem. They enable the Core Spray or LPCI Systems to adequately cool the core for all pipe breaks smaller than those provide protection agam;t the small pipe break in the event of for which the LPCI or Core Spray Systems can protect the core.

HPCI failure, by depressunzing the reactor vessel rapidly enough to Me h Cue $ay a W Systans. h cae pay -

The HPCI meets this requirement wdhout the use of a-c electrical and/or LPCI provde sufficient flow of coolant to limit fuel clad power. For the pipe breaks for wtuch the HPCI is intended to taperatures to well below clad fragmentation and to assure that function, the core never uncovers and is continuously cooled cae W ranaens M and thus no clad damage occurs. Refer to Section 6.5.3 of the FSAR.

The ADS has sufficent excess capacity such that only five of the seven h are r@M opwaW Mng pows opwah (see Low power physics testing and reactor operator training with mmance Weeds fa inoperable component (s) will be conducted only when the HPCI System is not required, (reactor coolant temperature $2127 and coolant pressure $150 psig). If the plant paramaters are below Loss of three ADS valves reduces the pressure relieving the point where the HPCI System is required, physics testing and capacsty, and, thus, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action to a cold condition with operator training w31 not place the plant in an unsafe condition.

reactor pressures less than 100 psig is specrfied.

Operability of the HPCl System is recoired only whan reactor Low power physecs testing and reactor operator training with pressure is greater than 150 psig' and reactor coolant inoperable curyonents. will be conducted only when that temperature is greater than 212*F because cc40 spray and low coiryonent or system is not required, - { reactor coolant pressure coolant injection can protect the core for any size pipe temperature $212"F and rsactor vessel vented or the reactor break at low pressure.

vessel head removed). With the reactor coolant temperature 5212*F and the Reactor vessel vented or the Amendment No.1p

.128

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

E Safety / Relief Valves E

Safety / Relief Va*ves 1.

During reactor powrr operating conditions and prior to 1.

At least 5 of the 11 safety / relief valves shall be berr.h

- startup from a cold condition, or whenever reactor coolant checked or replaced with bench checked valves once pressure is greater than atmosphere and temperature each operating cycle. All valves shall be tested every two l

greater than 212*F, the safety mode of at least 9 of 11 operating cycles. The testing shall CmiTOrrstrate that the 11 safety / relief valves shall be operable. The Automatic safety / relief valves actuate at 1110 psig 35 Depressurization System valves shall be operable as required by specification 3.5.D.

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JAFNPP 6

4 3.6 (cont'd) 4.6 (cont'd) 2.

If Specification 3.6.E.1 is not met, tte reactor shall be 2.

At least one safety / relief valve shall be disassernbled and placed in a cold condition within 24 hr.

mspected once/ operating cycle.

3.

Low power physcs testing and reactor operator trammg 3.

The intagnty of the nitrogen system and cv...pivo is shall be permitted with inoperable cer.ponents. as whch provde manual and ADS actushon of the l..

specified in Specification 3.6.E.1 above, provided that safety / relief valves shall be demonstrated at least once reactor coolant temperature is <212"F and the reactor every 3 months.

vessel is vented or the reactor vessel head is removed.

4, j

4.

The provisions of Specification 3.0.D are not amR ai-Ac.

steam to the condenser and observe a >10% closure of the turbine bypass valves, to venfy that the safety / relief valve has opened. This test shall be performed at least once each operating cycle wdhen the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of continuous power operation at a reactor steam dome pressure of > 940 psig.

Amendment No.f,

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JAFNPP' 3.6 and 4.6 BASES (cont'd) systems and the Mark I torus loadmg analyses. Based on i

E.-

Safety / Relief Valves safety / relief valve testing experience and the analysis l

The safety / relief valves (SRVs) have two modes of operation; referenced above, the safety / relief valves are bench tested to l

the safety mode or the relief mode. In the safety mode (or denivrotiate that in-service opening pressures are within the spring mode of operation) the spring loaded pilot valve opens nominal pressure setpoints 3% and then the valves are when the steam pressure at the valve inlet overcomes the returned to service witti opening pressures at the nominal I

spring force holding the pilot valve closed. The safety mode of setpoints 15 In this manner, valve integrity and the margin operation is required during pressurization tren oia Rs to ensure to the upper limit value specified in 2.2.1.B are maintained from I

vessel pressures do not exceed the reactor coolant pressure cycle to cycle.

safety limit of 1,375 psig.

The analyses with NEDC-31697P also provide the safety basis in the relief mode the spring loaded pilot valve opens when the for which 2 SRVs are permitted inoperable during continuous spring force is overcome by nitrogen pressure wtuch is power operation. With more than 2 SRVs moperable, the provided to the valve through a solenoid operated valve. The margin to the reactor vessel pressure safety limit is segnificantly solenoid operated valve is actuated by the ADS logic system reduced, therefore, the plant must enter a cold condition within (for those SRVs which are included in the ADS) or manually by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> once more than 2 SRVs are determined to be the operator from a control switch in the main control room or inooerable. (See reload evaluation for the current cycle).

at the remote ADS panel Operation of the SRVs in the relief A manual actuation of each SRV is performed to verify that the for the ADS is discussed in the Bases for Specification valves

• are mechanically funchonal and that no blockage exists in the valve discharge line. Adequate reactor steam dome Experiences in safety / relief valve testing have shown that pressure must be available to pufcain this test, in accordance failure or deterioration of safety / relief valva.s can be adequately with the manufacturer's recunnviendations, to avoid damaging detected if at least 5 of the 11 valves are bench tested once per the valve. Therefore, plant start-up is allowed and sufficient operating cycle so that all valves are tested every two operating time is provided after the required pressure is achieved (940 cycles. Furthermore, safety / relief valve testing experience has psig) to perform this test.

demonstrated that safety / relief valves which actuate within power s teshng and reactor operator training with 3% of the design pressure setpoint are considered operable inoperabic components w,ll be conducted only when the i

(see ANSI /ASME OM-1-1981). The safety bases for a sinofa nominal valve opening pressure of 1110 psig are described'in safety / relief and safety valves are I

NEDC-31697P, " Updated SRV Performance Requirements for the JAFNPP!' The single nominal setpoint is set below the reactor vessel design pressure (1250 psig) per the requirements of Articic 9 of the ASME Code - Section III, Nuclear Vessels. The setting of 1110 psig preserves the safety I

margins associated with the HPCI and RCIC turbine overspeed Amendment No. f,1 152

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ATTACHMENT 11 J

SAFETY EVALUATION FOR PROPOSED N EA W 5 F CIR I R G ER REM OP5ATEFERVptRT5RRAVCE REQUIREMENTS (JPTS 89 017) 1 i

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

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• , 1 SAFETY EVALUATION Page 1 of 12 1.

DESCRIPTION OF THE PROPOSED CHANGES This application for an amendment to the James A. FitzPatrick Technical Specifications proposes new Safety / Relief Valve (SRV) performance limits to take credit for the currently installed SRV capacity. Other changes, unassociated with SRV performance, clarify selected portions of the Technical Specifications and correct minor typographical and editorial errors.

A.

New SRV Performance Umits Four changes to the existing SRV performance limits are proposed:

. The first permits continued plant operation with two SRVs out of service. Since 7 of the 11 SRVs at FitzPatrick are also ADS (automatic depressurization system) valves, this reduces the number of ADS valves required to be operable to five. Current specifications permit only one SRV out of service for thirty days.

. Secondly, the setpoint for all eleven SRVs are changed to a single nominal setpoint.

Current specifications stagger the setpoints from 1090 to 1140 psig, i

. The third change increases the maximum permissible setpoint tolerance from one to three percent.

. Fourth, the Umlting Safety System Settin) (LSSS) is defined as 1195 psig.

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l A comparison of the changes in performance requirements is summarized as follows:

P_crformance Requirement Present Umit New Umit 1.

SRV opening pressure

+1% of Setpoint 1195 psig l

required to prevent l

overpressurization of the reactor coolant system (TS 2.2.1.B) 2.

Maximum SRV opening

+1% of Setpoint 1195 psig pressure used in otherlicensing basis analyses (FSAR Ohapter 14) 1 3.

Nominal SRV Setpoint 2 @ 1090 psig 11 @ 1110 psig 2 @ 1105 psig 1

7 @ 1140 psig 4.

Setpoint tolerance

+1% of Setpoint

+3% of Setpoint 5.

Number of SRVs and ADS valves 0

2 assumed to be out of service (TS j

3.5.D and 3.6.D)

V

' 1 SAFETY EVALUATION j

Page 2 of 12 e

The specific changes to the FitzPatrick Technical Specifications, which incorporate those

(_

now SRV performance limits, are detailed below:

1.

Specification 2.2.1.B. page 27; change,

' Reactor coolant system safety / relief valve nominal settings shall be as follows:

Safety / Relief Valves 2 valvos at 1090 psig 2 valves at 1105 psig 7 valvos at 1140 psig The allowable setpoint error for each safety /rollef valvo shall be 1 1 percont.'

to read:

"At least 9 of the 11 reactor coolant system safety /rolief valvos shall have opening pressures less than or equal to an upper limit value of 1195 psig."

2.

Bases Soction 2.2.1.8, page 29; delete the last paragraph (begins with *The numerical distribution...") and change the fourth paragraph (bogins with "The current reload analysis...") to rcad:

'The limiting vessel overpressure transient event is a main steam isolation valvo closure with flux scram. This event was analyzed within NEDC-31697P, " Updated SRV Performanco Requiroments for the JAFNPP,"

assuming 9 of the 11 SRVs were operable with opening pressures less than or equal to an upper limit value of 1195 psig. The resultant peak vessel pressure for the ovent was shown to be loss than the vessel pressure code limit of 1,375 psig. (See current reload analysis for the reactor response to the main steam isolation valve closure with flux scram event.) The upper limit value of 1195 psig is the SRV opening pressure up to which plant performance has been analyzed, assuming 2 SRVs are inoperable.

4 Thorofore, SRV opening pressures below the upper limit (1195 psig) ensure that the ASME Code limit on peak reactor pressure is satisfied."

3.

Specification 3.5.D.1, page 119 and 120; replace specification with the following:

The ADS shall be operable with at least 5 of the 7 ADS valves operable:

a.

whenever the reactor prosve is grnator than 100 psig and irradiated fuel is in the reactor vec,ai, and

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SAFETY EVA111ATION i

Page 3 of 12 I

b. prior to reactor startup from a cold condition.

I 4.

Speelfication 3.5.D.3, page 120; delete the cross-reference to action statements 3.5.D 1.a and 3.5.D.1.b and add ' ADS." The revised specification reads i

• Low power physics testing and reactor operator training shall be permitted with inoperable ADS components, provided that reactor coolant q

temperature is <212 F and the reactor vessel is vented or reactor vessel j

head is remove 3."

i 5.

Specification 4.5.D.2, page 120. Delete this specification.

6.

Bascs Section 3.5.D, page 128; change the second paragraph (begins with,

• Redundancy has been provided...") to read as follows:

• The AD3 has sufficient excess capacity such that only five of the seven valves are required operable during power operation (see NEDC 31697P,

' Updated SRV Performance Requirements for the JAFNPP').

Loss of three ADS valves reduces the pressure relieving capacity, and, thus, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action to a cold condition with reactor pressures less than L

100 psig is specified.*

7.

Specification 3.6.E.1, on page 142a; delete the words

• Safety and" from the title, change the word *all" to *at least 9 of 11,* and delete the phrase "except as specified by Specification 3.6.E.2.* The revised specification shall read as follows:

• During reactor power operating conditions and M:a to startup from a cold condition, or whenever reactor coolant pressure k greater than atmosphere and temperature greater than 212"F, the safety moo'e of at least 9 of 11 safety / relief valves shall be operable. The Automatic Depressurization System valves shall be operable as required by specification 3.5.D.*

l-8.

Specification 4.6.E.1, page 142a; delete the words

• Safety and" from the title, change

• one half of all* to *5 of the 11,* delete the cross reference to Specification 2.2.B and add the revised valve actuation setpoints. The revised specification shall read as l

follows:

i "At least S of the 11 safety / relief valves shall be bench checked or replaced with bench checked valves once each operating c)cle. All valves shall be tested every two operating cycles. The testing shall demonstrate

- that the 11 safety / relief valves actuate at 1110 psig 3%"

9.

Specification 3.6.E.2, page 143; delete this specification.

l 10.

Specification 3.6.E.3, page 143; delete the cross reference to specification 3.6.E.2 and renumber this specification to be 3.6.E.2 i

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a Attachment ll SAFETY EVALUATION L

Page 4 of 12 l

11.

Specification 3.6.E.4, page 143; change the cross reference from *ltem B.2* to

" specification 3.6.E.1" and renumber this specification to be 3.6.E.3.

12.

Bases Section 3.6 and 4.6, page 152; delete the first paragraph (begins with

• Experiences in safety valve...') and change the third paragraph (begins with "The safety function is...') and fourth paragraph (begins with "It is realized that...") to read:

Experiences in safety / relief valve testing have shown that failure or deterioration of safety / relief valves can be adequately detected if at least 5 of the 11 valves are bench tested once per operating cycle so that all valves are tested every two operating cycles. Furthermore, safety / relief valve testing experience has demonstrated that safety / relief valves which actuate within 13% of the design pressure setpoint are considered operable (see i

ANSI /ASME OM 1 1981). The safety bases for a single nominal valve opening pressure of 1110 psig are described in NEDC 31697P, " Updated SRV Performance Requirements for the JAFNPP" The single nominal setpoint is set below the reactor vessel design pressure (1250 psig) per the requirements of Article 9 of the ASME Code Section lil, Nuclear Vessels.

The setting of 1110 psig preserves the safety margins associated with the HPCI and RCIC turbine overspeed systems and the Mark i torus loading analyses. Based on safety / relief valve testing experience and the analysis referenced above, the safety / relief valves are bench tested to demonstrate i

that in service opening pressures are within the nominal pressure setpoints t3% and then the valves are returned to service with opening pressures at the nominal setpoints 21%. In this manner, valve integrity and the margin to the upper limit value specified in 2.2.1.B are maintained from cycle to cycle.

The analyses with NEDC 31697P also provide the safety basis for which 2 SRVs are permitted inoperable during continuous power operation. With more than 2 SRVs inoperable, the margin to the reactor vessel pressure safety limit is significantly reduced, therefore, the plant must enter a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> once more than 2 SRVs are determined to be inoperable. (See reload evaluation for the current cycle).

B.

Miscellaneous Administrative Charpps Five miscellaneous changes are provided to clarify terminology, correct typographical errors, remove a surveillance requirement which should have been deleted as part of Amendment 130, to clarify when SRV manual actuation is performed, and to delete a duplicate specification.

1.

Terminology Clarifica lons t

a.

Specification 1.2.1, page 27; change the phrase

• reactor coolant system pressure

• to " reactor vessel clome pressure."

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Attachment il i

SAFETY EVALUATION l

Page 5 of 12 b.

Bases Section 3.6.E and 4.6.E, page 152, change the second paragraph to read:

l The safety / relief valves (SRVs) have two modes of operation; the safety mode or the relief mode. In the safety mode (or spring mode of operation) the spring loaded pilot valve opens when the steam i

y pressure at the valve inlet overcomes the spring force holding the pilot valve closed. The safety mode of operation is required during 6

pressurization transients to ensure vessel pressures do not exceed the p

reactor coolant pressure safety limit of 1,375 psig.

In the relief mode the spring loaded pilot valve opens when the spring A

force is overcome by nitrogen pressure which is provided to the valve through a solenoid operated valve. The solenold operated valve is actuated by the ADS logic system (for those SRVs which are included in the ADS) or manually by the operator from a control switch in the main control room or at the remote ADS panel. Operation of the SRVs in the relief mode for the ADS is discussed in the Bases for Specification 3.5.D.

2.

Typographical Corrections L

l a.

Bases Section 1.2 and 2.2, page 29, second paragraph; change the *

• signs to 5

"= " signs.

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b.

Specification 3.5.D.2, page 120; delete the *," after *100 psig."

3.

Amendment 130 Change l

6.

a.

Specification 4.2.B. Table 4.2 2, page 80; delete item 8, " ADS Relief Valve Bellow Pressure Switch."

5 4.

SRV Manual Actuation Test i

a.

Specification 4.5.D 1.b, page 119; move this specification to new Section 4.6.E.4 (page 143) and add "This test shall be performed at least once each operating cycle within the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of continuous operation at a reactor steam dome pressure of >,940 psig, b.

New Specification 3.6.E.4, page 142; add "The provisions of Specification 3.0.D are not applicable."

c.

Bases Section 3.6.E and 4.6.E, page 152; add:

"A manual actuation of each SRV is performed to verify that the valves

• are mechanically functional and that no blockage exists in the valve discharge line. Adequate rea for steam dome pressure must be available to perform this test, in accordance with the manufacturer's recommendations, to avoid damaging the valve. Therefore, plant start up is allowed and sufficient time is

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SAFETY EVALUAT50N Page 6 of 12 provided after the required pressure is achieved (940 psig) to perform this test."

i 5.

Duplicate Specification o

a..

Specification 4.6.E.4, page 143, delete the following:

An annual report of safety / relief valve failures and challanges will be sent to the NRC in accordance with Section 6.9.A.2.b.

1 u.

PURPOSE OF THE PROPOSED CHANGES l

A.

New SRV Performance Umits i

n Existing Specifications 2.2.1.B and 4.6.E require the SRVs to open with staggered setpoints l

and with a tolerance of 1% Specification 4.6.E limits plant operation with one SRV out of service to thirty days. Considering the existing SRV capacity, and recent experiences with 1

SRV setpoint drift, these specifications unnecessarily restrict plant operation based on very

]

conservative SRV performance limits. The proposed changes will reduce forced outages

]

and decrease maintenance and surveillance testing costs; without impacting safety or plant performance.

)

l A detailed analysis of these changes has been performed for the Authority by the General Electric Company. The results of these analyses are summarized in a report entitled

• Updated SRV Performance Requirements for the James A. FitzPatrick Nuclear Power l'

Plant" (NEDC 31697P) (Since this report contains proprietary information, copies are being l

transmitted under a separate cover.) NEDC 31697P predicts plant response assuming that the following new SRV performance limits were adopted:

t

. relaxation of the11% nominal valve nameplate setpoint tolerance 1o13%,

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. operation with any two SRVs or ADS valves inoperable, L

setting all 11 SRVs at a single nominal nameplate setpoint, and

. the Umiting Safety System Setting for the pressure relief system is defined to be 1195 psig.

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NEDC 31697P demonstrates that sufficient margin still exists in the reactor vessel l'

overpressure protection, fuel thermal limits, and torus loading analyses if these changes are L

instituted.

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Attachment ll SAFETY EVALUATION i

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. Single Nominal SRV Setpoint The adoption of a single setpoint for all eleven SRVs will reduce the quantity of spare SRV top-works that must be kept on hand.

l Consistent with the Mark l Contalnment Short Term Program initiatives, Amendment No. 43 1

(Reference 14) implemented staggered SRV setpoints which limit the number of valves which could experience consecutive actuation following an isolation transient. The Authority has since performed a Mark l Containment Long Term Program assessment (References 15,16 and 17) which demonstrated that the allowable containment loads are not exceeded due to multiple SRV actuations. A single nominal setting of 1110 psig is J

selected to preserve the safety margins assumed in the containment loading calculations.

Setpoint Tolerance I

Operating experience at the Fitzpatrick plant and at other BWRs has shown that SRV setpcint drift exceeds the setpoint tolerance (See LERs85-009,87 004,88-004, and 88 010,

)

i References 1,2,3, and 4 respectively). Implementing a 3% setpoint tolerance will lessen i

the number of valve refurbishments, minimize the number of valves requiring confirmatory

]

testing, and reduce the quantity of reportable events.

I l

Two SRVs Out of Service The excess installed SRV capacity permits two ADS valves or SRVs to be Inoperable during continuous power operations. This will reduce the number of forced outages due to valve inoperability.

Limiting Safety System Setting (LSSS)

Establishing the LSSS as the upper limit opening pressure of 1195 psig is consistent with Standard Technical Specifications and the definition of a Umiting Safety System Setting (Specification 1.H).

B.

Miscellaneous Changes 1.

Terminology Corrections a.

This change more clearly specifies where in the reactor coolant system the pressure safety limit of 1325 psig should be measured. Use of the vessel steam dome pressure Indicator is consistent with Bases Section 1.2.

b.

This change more clearly defines the methods of SRV actuation. The terminology changes are consistent with the revised wording of Specification 3.6.E.1 and the ADS Bases section.

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L SAFETY EVALUATION l

Page 8 of 12 l

i 2.

Typographical Corrections a.

The change to Bases Section 1.2 and 2.2 and Specification 3.5.D.2 correct i

typographical errors.

)

3.

Amendment 130 Change a.

The change to Specification 4.2.B deletes the requirement to perform logic j

functional testing on the ADS bellows pressure switch. This change was inadvertently omitted from Amendment 130 (Reference 11).

4.

SRV Manual Actuation Test a.

Relocates the SRV manual actuation test to the proper Technical Specification Section.

b.

This change clarifies that manual actuation of the SRVs must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of achieving the required test pressure of > 940 psig. ThN is consistent with plant and industry practices and has been requested by the NRC Resident inspector, f

c.

Specification 3.0.D and 4.0.D require the successful completion of all survolliance testing prior to plant start up. This change is in accordance with Standard Technical Specifications and eliminates a literal Inconsistency within the Technical Specifications.

5.

Duplicate Specification a.

Specification 4.6.E.4 is deleted because it is redundant to Specification 6.9.A.2.b.

Reporting requirements are not surveillance tests and are properly located in Section 6 of the Technical Specifications.

111.

IMPACT OF THE PROPOSED CHANGES A.

Now SRV Performance Requirements NEDC 31697P (Reference 5) considered the affects of these changes on eight plant performance issues. The paragraphs below summarize the results of this t.nalysis. Refer to NEDC 31697P for complete details of each analysis.

l Vessel Overpressure Umits: An upper limit opening pressure of 1195 psig results in a 50 psi margin to the ASME Code upset reactor vessel pressure limit of 1375 psig.

Fuel Thermal Umits: The revised SRV performance requirements have no impact on fuel thermallimits.

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.* 1 SAFETY EVALUATION L

Page 9 of 12 LOCA/ECCS Performance: The peak cladding temperatures of the ECCS/LOCA analysis are insensitive to SRV opening pressure increases. Operation with 2 SPVs or ADS valves out of service has an insignificant impact on ECCS/LOCA performance.

HPCl/RCIC Operability: The revised SRV performance requirements have an Insignificant impact on HPCl/RCIC performance. However, the margin to the 125%

mechanical overspeed trip for the HPCI and RCIC turbines is reduced by SRV opening pressure increases. The selection of a nominal setting of 1110 psig preserves the turbine overspeed margin by limiting the required turbine speed to 101% of nameplate rating.

Containment Response and Integrity: The revised SRV performance requirements have no impact on the calculated peak containment pressures and temperatures. An increase in SRV opening pressures to 1195 psig and the resultant increase in SRV discharge loads do not exceed containment structure stress allowables for the limiting load combinations. A nominal setting of 1110 psig is selected to preserve the safety margins included in the Mark l Plant Unique Load Definition Report.

SRV Simmer Margin: The selection of a 1110 psig single setpoint provides a 110 psi simmer margin and does not increase the occurrence of pilot valve leakage as compared to the current nominal settings (see Section 5.2 of NEDC 31697P).

Therefore, the probability of a stuck open relief valve is not increased.

10CFR50 Appendix R-Alternate Shutdown Capability: Post fire shutdown capability remains within 10CFR50 Appendix R limitations. The revised SRV performance requirements increases the previously analyzed duration of fueluncovery by 40 seconds. The resultant peak cladding temperature remains below the temperature at which cladding perforations are expected.

Setpoint Drift to Minus 3%: A revised tolerance of 3% permits setpoint drift down to 1077 psig. This value is 32 psi above the high pressure scram setpoint and provides a sufficient cushion above normal reactor operating pressures.

NEDC-31697P concludes that the changes in the pressure relief system performance requirements do not have a significant safety impact on vessel overpressure margin, fuel thermal limits, LOCA/ECCS performance, HPCl/RCIC operability, containment response, or containment integrity. Furthermore, the performance changes have an insignificant impact on Alternate Shutdown System (10CFR50 Appendix R) performance, simmer margin, and downward setpoint drift.

These new performance limits are primarily administrative changes. The sctpoint tolerance of + 3% and the upper limit va!ue (Umiting Safety System Setting) of 1195 psig are changes reflecting an ASME testing criterion change (Reference 10) and changes to design basis analyses.

Actual physical changes to the plant are minimal. The physical changes are continuous operation with 2 SRVs/ ADS valves out of service and revised setpoints to 1110 psig.

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I 4 J' 1 SAFETY EVALUATION Page 10 of 12 B.

Miscellaneous Administrative Change These changes are purely administrative in nature. They do not involve a plant modification; nor do they impact any procedural or administrative controls.

IV.

EVALUATION OF SiONIFICANT H_AZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with the proposed amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, sinco it would not:

1.

involve a significant increase in the probability or consequences of an accident previously evaluated. A bounding analysis [NEDC 31697P, " Updated SRV Performance Requiromonts for the James A. Fitzpatrick Nuclear Power Plant") of the revised SRV performance requirements considered plant operation with 9 of 11 SRVs operable and with a common valve actuation pressure of 1195 psig. The analysis demonstrates that a 50 psi margin exists between the maximum anticiphted pressure and the American Society of Mechanical Engincors (ASME) Code upset reactor vessel pressure limit of 1375 psig. The analyses of NEDC 31697P also demonstrate that the new SRV performance limits have no significant impact on thermal limits, ECCS/LOCA performance, HPCl/RCIC operability, containment response, containment integrity, or 10CFR50 Appendix R alternate shutdown capability.

The analyses also considered simmer margin and downward setpoint drift.

The five miscellaneous changes clarify terminology, correct typographical errors, remove a surveillance requirement which should have been deleted as part of Amendment 130, clarify when SRV manual actuation is performed, and delete a duplicate specification. Those changes are purely administrative in nature and, as such, do not impact previously evaluated accidents or equipment malfunctions.

2.

create the possibl!!ty of a new or different kind of accident from those previously evaluated.

The now SRV performance limits are primarily administrative changes. The only physical changes involve recalibration of SRV setpoints and operation with 2 SRVs/ ADS valves out-of service. The operation and function of the pressure relief system are unaffected. No now failuro modes are introduced.

The proposed miscellaneous changes are purely administrative in nature and, as such, do not create the possibility of an accident or malfunction.

3.

Involvo a significant reduction in the margin of safety. The new SRV performanco limits slightly reduce the existing margin to vessel overpressure and the margin to the 125%

mechanical overspeed trip for the HPCI and RCIC turbines. However, the reduction in the overpressure margin is insignificant (approximately 25 psi) and the plant's response to transients and accidents remains well within the limits established in Section 111, Division I of the ASME Boller & Pressure Vessel Code and the regulatory limits established in General Design Criteria (GDC) 15, Standard Review Plan Section 5.2.2, and FSAR Section 4.4. The reduction in turbine overspeed margin is negligible (less than 1%), because it is within the allowable tolerance of the trip settings.

1 1

e Attachment ll SAFETY EVALUATION Page 11 of 12 t-The proposed miscellaneous changes are purely administrative in nature and do not involve a reduction in safety margin.

V.

IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of the proposed changes will not impact the Al. ARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.

VI.

CONCLUSION The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 I

CFR 50.59. That is, they:

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 different type than any evaluated previously in the safety analysis report; c.

will not reduce the ma! gin of safety as defined in the basis for any technical specification; i

and d.

Involves no significant hazards consideration, as defined in 10 CFR 50.02.

Vil.

REFERENCES 1.

Ucensee Event Report 85-009, Main Steam Safety Relief Valves Found Out of Tolerance During Test.

2.

Ucensee Event Report 87 004, Main Steam Safety Relief Valves Found Out of Tolerance.

3.

Ucensee Event Report 88 004, Reactor Safety / Relief Valve Setpoint Drift.

4.

Ucensee Event Report 88 010, Reactor Safety / Relief Valv6 Sc+ point Drift.

5.

NEDC 31697P, Updated SRV Performance Requirements for the James A. Fitzpatrick Nuclear Power Plant, April 1989.

6.

James A. FitzPatrick Nuclear Power Plant Updated Final Safety Arialysis Report, Section 4.4

• Pressure Relief System," Section 4.7 " Reactor Core Isolatica Cooli g System," Section 6.4 "High Pressure Coolant injection System,' and Section 14 " Safety Analyses."

7.

USAEC ' Safety Evaluation of the James A. FitzPatrick Nuclear Power Plant" (SER), dated November 20,1972.

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y th..

F y

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1 Attachment ll

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SAFETY EVALUATION L

Page 12 of 12 k

8.

_ USAEC " Supplement 1 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power

-1 Plant" (SER), dated February 1,1973.

9.

USAEC " Supplement 2 to the Safety Evaluation of the James A. FitzPatrick Nuclear Power Piare (SER), dated October 4,1974.

_10.. ANSl/ASME OM-1 1981, Requirements for Inservice Performance Testing of Nuclear Power

. Plant F,est.ure Relief Devices, s*

11.

Amendment 130 to the James A. Fitzpatrick Operating Ucense, May 31,1989.

.12.-

HPCI Turbine Instruction Manual., Teiry Steam Turbine Company, VPF # 2300-61 13.

RCIC Turbine Instruction Manual, Terry Steam Turbine Company, VPF # 2059-49-2 14.

Amendment 43 to the James A. Fitzpatrick Operating Ucense, November 22,1978.

15.

" Plant Unique Analysis Report of the Torus Suppression Chamber for JAFNPP," Teledyne Eng!neering Services, TR 5321 1, Revision 1, September 1984.

4 16.- ' ' Plant Unique Analysis Report of theTorus Atte.ched Piping for JAFNPP," Teledyne Engineering Services, TR 53212, Revision 1, November 1984.

17.'

NRC Letter JAF 84-364, dated December 12,1984, " Post Implementaticr. sudit Review of u

Unique Analysis Report for Mark l Containment Long Term Program - Program Found Accepteble."

J18.. ASME Boller & Pressure Vessel Code, Section 111 - Rules for Construction of Nuclear Vessels,1965 Edition with Addenda through Winter 1966.

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