ML20073L751

Amend 217 to License DPR-59,modifying SRV Performance Limits

ML20073L751
Person / Time
Site:	FitzPatrick  (DPR-59-A-217)
Issue date:	09/28/1994
From:	Michael Case Office of Nuclear Reactor Regulation
To:
Shared Package
ML20073L758	List: ML20073L751 ML20073L761 ML20073L768
References
NUDOCS 9410130280
Download: ML20073L751 (12)

Text

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o UNITED STATES j

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D.C. 2055!K)001 j

POWER AUTHORITY OF THE STATE OF NEW YORK l

l DOCKET NO. 50-333

?

l JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No. 217 i

License No. DPR-59 i

1.

The Nuclear Regulatory Commission (the Commission) has found that:

j A.

The application for amendment by Power Authority of the State of New York (the licensee) dated December 20, 1989, as supplemented i

January 16, 1990, January 3, 1992, January 30, 1992, May 5, 1993, and May 26, 1993, and superseded March 2, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as 1

amended (the Act) and the Commission's rules and regulations set forth j

in 10 CFR Chapter I; i"

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the e

Commission; C.

There is reasonable assurance (i) that the activities authorized j

by this amendment can be conducted without endangering the health and i

safety of the public, and (ii) that such activities will be conducted i

in compliance with the Commission's regulations; i

D.

The issuance of this amendment will not be inimical to the common

{

defense and security or to the health and safety of the public; and i

~

E.

The issuance of this amendment is in accordance with 10 CFR Part i

51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-59 is hereby i

amended to read as follows:

s i

9410130280 940928 PDR ADOCK 05000333 p

PDR t

' (2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as I

revised through Amendment No.217,.are hereby incorporated in the license. The licensee shall operate the facility in accordance with i

the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance to be implemented prior to startup from the next refueling outage.

FOR THE NUCLEAR REGULATORY COMNISSION l

)

Michael J.

ase, Acting Director Project Directorate I-l Division of Reactor Projects - I/II l

Office of Nuclear J.eactor Regulation l

Attachment:

[

Changes to the Technical l

Specifications Date of Issuance: September 28, 1994 l

i I

l

i ATTACHMENT TO LICENSE AMENDMENT NO,217 FACILITY OPERATING LICENSE NO. DPR-59 3

DOCKET NO. 50-333 Revise Appendix A as follows:

Reraove Paaes Insert Paaes 27 27 29 29 80 80 119 119 120 120 128 128 142a 142a 143 143 152 152 a_

JAFNPP 1.2 REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM APPLICABILITY:

APPLICABILITY:

Applies to limits on reactor coolant system presswe.

Applies to trip settings of the instruments and devices which are provided to prevent the reactor coolant system safety limits from being exceeded.

OEUECTIVE:

OBJECTIVE:

To establish a limit below which the integrity of the Reactor Coolant To define the level of the process vanables at which automatic System is not threatened due to an overpresswe condition.

protective action is initiated to prevent tne safety limits from being exceeded.

SPECIFICATION-SEECIFICATION:

l 1.

The reactor vessel dome presswe shall not exceed 1,325 psig 1.

The Limiting Safety System setting shall be specified below:

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

A.

Reacter coolant high presswa scram shall be s1,045 psig.

B.

At least 9 of the 11 reactor coolant system safety / relief 1

valves shall have a nominal setting of 1110 psig with l

an allowable setpoint error of 3 percent.

1 4

i Amendment No. M, M, #5,M,'M, 217, i

27

JAFNPP 1.2 and 2.2 BASES The reactor coolant pressure boundary integrity is an important barrier The limiting vessel overpressure transient event is a main steam in the prevention of uncontrolled release of fission products. It is -

isolation valve closure with flux scram. This event was analyzed i

essential that the integrity of this boundary be protected by within NEDC-31697P, " Updated SRV Performance Requirements for cstablishing a presswa limit to be observed for all operating conditions the JAFNPP," assuming 9 of the 11 SRVs were operable with opening and whenever there is irradiated fuel in the reactor vessel.

pfessures less than or equal to 1195 psig. The resultant peak vessel pressure for the event was shown to be less than the vessel presswe

[

The pressure safety limit of 1,325 psig as measured by the vessel code limit of 1375 psig. (See current v.doad analysis for the reactor stism space pressure indicator is equivalent to 1,375 psig at the response to the main steam isolation valve closure with flux scram lowet elevation of the Reactor Coolant System. The 1,375 psig value event). The value of 1195 psig is the SRV opening pressure up to is dcf "xl from the design pressures of the reactor pressure vessel which plant performance has been analyzed, assuming 2 SRVs are and reactor coolant system piping. The respective design pressures inoperable. Therefore, SRV opening pressures below 1195 psig ars 1250 psig at 575 *F for the reactor vessel,1148 psig at 568 *F for ensure that the ASME Code limit on peak reactor pressure is satisfied.

the recirculation suction piping and 1274 psig at 575 *F for the discharge piping. The pressure safety limit was chosen as the lower A safety limit is applied to the Residual Heat Removal System (RHRS) of the pressure transients permitted by the applicable design codes:

when it is operating in the shutdown cooling mode. When operating in 1965 ASME Boiler and Pressure Vessel Code, Section lit for pressure the shutdown cooling mode, the RHRS is included in the reactor vessel and 1969 ANSI B31.1 Code for the reactor coolant system coolant system.

piping. The ASME Boiler and Pressure Vessel Code permits pressure transients 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 referenced to the lowest elevation i

of the Reactor Coolant System.

4 Amendment No. E,K,38(,J80',217 29

JAFNPP TABLE 4.2-2 (Cont'd)

MINisdM TEST AND CALIBRATION FREQUENCY FOR CORE AND CONTAINMENT COOLING SYSTFaas Logic System Functional Test Frequency 1)

Core Spray Stbsystem (7) (9)

Once/6 months L

2)

Low Pressure Coolant Irjection Stbsystem (7) (9)

Once/6 months 3)

Containment Cooling Subsystem Once/6 months 4)

HPCI Subsystem (7) (9)

Once/6 months 5)

HPCi Subsystem Auto isolation (7)

Once/6 months 6)

ADS Subsystem (7)(9)

Once/6 months 7)

RCIC Stbsystem Auto isolation (7)

Once/6 months I+

NOTE:

See notes following Table 4.2-5.

i 4

Amendrnent No..<T,E,105,J8f,28f, 217, 80

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

D.

Aidnmatic Deoraccuri7ation System (ADS)

D.

Aidtvnntic Deorancuriration System (ADS) 1.

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

Surveillance of the Automatic Depressurization System valves operable:

shall be performed during 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 irradiMed fuel is in the reactor vessel, and

valves, b.

prior to reactor startup from a cold condition.

b.

A simulated automatic initiation which is inhibited by the override switches.

Am.sndment No. M, M, M4, 217, 119

[

t f

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

I 2.

If the requirements of 3.5.D.1 cannot be met, the reactor 2.

A logic system functional test.

shall be placed in the cold condition and pressure less ! San l

100 psig within 24 hr.

a.

When it is determined that two valves of the ADS are l inoperable, the ADS subsystem actuation logic for the operable ADS valves and the HPCI subsystem shall be verified to be oper4le immediately and at least weekly thereafter.

b.

When it is determined that more than two relief / safety valves of the ADS are inoperable, the HPCI System shall be verified to be operable immediately.

I' 3.

Low power physics testing and reactor operator training shall be permitted with inoperable ADS components, provided that reactor coolant temperature is s212 F and the reactor vessel is vented or reactor vessel head is l

removed.

4.

The ADS is not required to be operable during hydrostatic pressure and leakage testing with reactor coolant temperatures below 300 F and irradiated fuel in the reactor vessel provided all control rods are inserted.

t 1

i 1

Amendment No. 35,.14tr,1PJ,.207, 217,

)

120

- ~ - -

,.n_

JAFNPP I

I 3.5 BASES (cont'd)

C.

HM Praecure Cantant iniactim (HPCI) System D.

Automatic Deoraccurimiim System (ADS)

I

~

i The Hign Pressure Coolant injection System is provided to The relief valves of the ADS are a backup to the HPCI adequately cool the core for all pipe breaks smaller than those subsystem. They enable the Core Spray or LPCI Systems to for which the LPCI or Core Spr::y Systems can protect the core.

provide protection against the small pipe break in the event of HPCl failure, by depressurizing the reactor vessel rapidy enough i

The HPCI meets this requirement without the use of a-c electrical to actuate the Core Spray or LPCI Systems. The core spray power. For the pipe breaks for which the HPCI is intended to and/or LPCI provide sufficient flow of coolant to limit fuel clad I

function, the core never uncovers and is continuously cooled and temperatures to well below clad fragmentation and to assure that thus no clad damage occurs. Refer to Section 6.5.3 of the core geometry remains intact.

FSAR.

The ADS has sufficient excess capacity such that only five of the Low power physics testing and reactor operator training with seven valves are required operable during power operation (see inoperable component (s) will be conducted only when the HPCI

. NEDC-31697P, " Updated SRV Performance Requirements for System is not required (reactor coolant temperature s212 Y the JAFNPP").

andcoolant pressure s150 psig). If the plant parameters are below the point where the HPCI System is required, physics Loss of three ADS valves reduces the pressure relieving testing and operator training will not place the plant in an unsafe 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 l

condition, reactor pressures less than 100 psig is specified.

[

Operability of the HPCI System is required only when reactor Low power physics testing and reactor operator training with pressurn 11 greater than 150 psig and reactor coolant inoperable coinpenents will be conducted only when that tempera is greater than 212 Y because core spray and low component or system is not required (reactor coolant pressure coolant injection can protect the core for any size pipe temperature s212 Y and reactor vessel vented or the reactor break at low pressure.

vessel head removed). With the reactor coolant temperature s212 Y and the Reactor vessel vented or the Amendment No. JOT, 217, 128

JAFNPP 3.6 (cont'd) 4.6 (cont'd) l E.

Safetv/ Relief Valves E.

Safetv/ Relief Valves

[

1. During reactor power operating conditions and prior to startup

1. At least 5 of the 11 safety / relief valves shall be bench l

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

212 F, the safety mode of at least 9 of 11 safety / relief valves operating cycles.* The testing shall demonstrate that the 11 shall be operable. The Automatic Depressurization System safety / relief valves actuate at 1110 psig 13%

valves shall be operable as required by specification 3.5.D.

• The current surveillance interval for bench checking safety / relief valves is extended until the end of R11/C12 refueling outage scheduled for January,1995. This is a one-time extension, effective only for this surveillance interval. The next surveillance interval will begin after the completion of the bench chech testing and after the safety / relief valves are declared operable.

Amendment No. M.E.E.R.136, MWf,196, 217, 142a i

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

2. If Specification 3.6.E.1 is not met, the reactor shall be placed in

2. At least one safety / relief valve shall be disassembled and a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

inspected once/ operating cycle.*

l

3. Low power physics testing and reactor operator training shall

3. The integrity of the nitrogen system and components which be permitted with inoperable componente, as specified in provide manual and ADS actuation of the safety / relief valves l

Specification 3.6.E.1 above, provided that reactor coolant shall be demonstrated at least once every 3 months.

temparature is s212 7 and the reactor vessel is vented or the reactor vessel head is removed.

l

4. The provisions of Specification 3.0.D are not applicable.

4. ManuNiy open each safety / relief valve while bypassing steam to the condenser and observe a 210% closure of the turbine bypass valves, te verify that the safety / relief valve has opened.

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 power operation at a reactor steam dome pressure of z940 psig.

5. The safety and safety / relief valves are not required to be operable during hydrostatic pressure and leakage testing with rnactor coolant temperatures between 212 7 and 300 Y and irradiated fuel in the reactor vessel provided all control rods are inserted.

t The current surveillance interval for disassembling and inspecting at least one safety / relief valve is extended until the end of R11/C12 i

refueling outage scheduled for January,1995. This is a one-time j

extension, effective only for this surveillance interval. The next surveillance interval will begin upon completion of this surveillance.

Amendment No. 45,-?F,.taa,.$a( $?f,.W5, 217, 143

5 JAFNPP 3.6 and 4.6 BASES (cont'd)

E.

Safety /Rahaf Valves The safety / relief valves (SRVs) have two modes of operation; of Article 9 of the ASME Code - Section ill, Nuclear Vessels.

the safety mode or the relief mode. In the safety mode (or The setting of 1110 psig preserves the safety margins associated spnno mode of operation) the spring loaded pilot valve opens with the HPCI and RCIC turbine overspeed systems and the i

when the steam pressure at the valve inlet overcomes the spnng Mark I torus loading analyses. Based on safety / relief valve force holdng the pilot valve closed. The safety mode cf testing experience and the analysis referenced above, the operation is required during pressurization transients to ensure safety / relief valves are bench tested to demonstrate that vessel pressures clo not exceed the reactor coolant pressure in-service opening pressures are within the nominal pressure safety limit of 1,375 psig.

setpoints 13% and then the valves are retumed to service with opening pressures at the nominal setpoints 11%. In thie mner, in the relief mode the spring loaded pilot valve opens when the

. valve integnty is maintained from cycle I; cycle.

spnng force is overcome by nitrogen pressure which is provided to the valve through a solenoid operated valve. The solenoid The analyses with NEDC-31697P also provide the safety basis operated valve is actuated by the ADS logic system (for those for which 2 SRVs are permitted inoperable during continuous SRVs which are included in the ADS) or manually by the power operation. With more than 2 SRVs inoperable, the margin operator from a control switch in the main control room or at the to the reactor vessel pressure safety limit is significantly reduced, remote ADS panel. Operation of the SRVs in the relief mode for 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 /> the ADS is discussed in the Bases for Specification 3.5.D.

once more than 2 SRVs are determined to be inoperable. (See reload evaluation for the current cycle).

Exponences in safety / relief valve testing have shown that failure or deterioration of safety / relief valves can be adequately detected A manual actuation of each SRV is performed to verify that the if at least 5 of the 11 valves are bench tested once per operating valves are mechanically functional and that no blockage exists in cycle so that aN valves are tested every two operating cycles.

the valve discharge "ne. Adequate reactor steam dome pressure Furthermore, safety / relief valve testing experience has must be available to perform this test, in accordance with the I

demonstrated that safety / relief valves which actuate within 13%

manufacturer's recommendations, to avoid damaging the valve.

of the design pressure utpoint are considered operable (see Therefore, plant start-up is allowed and sufficient time is provided ANSI /ASME OM.1-1981). The safety bases for a single nominal after the required pressure is achieved (940 psig) to perform this i

valve opening pressure of 1110 psig are desenbod in test.

NEDC-31697P, " Updated SRV Performance Requirements for 1

the JAFNPP." The single nominal setpoint is set below the Low power physics testing and reactor operator training with reactor vessel design pressure (1250 psig) per the requirements inoperable components will be conducted only when the l

safety / relief and safety valves are r

j Amendment No. E, Ja(, 217, 4

152

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