ML20244D014

Amend 26 to License NPF-57,increasing Surveillance Test Intervals & Allowable out-of-svc Times for Reactor Protection Sys

ML20244D014
Person / Time
Site:	Hope Creek
Issue date:	06/05/1989
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML20244D016	List: ML20244D014 ML20244D020
References
NUDOCS 8906160032
Download: ML20244D014 (11)

Text

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1 p* Klog'o UNITED STATES

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,o PUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY l

DOCKET NO. 50-354 1

HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 26 License No. NPF-57 1.

The Nuclear Regulatory Commission (the Commission or the NRC) has found-

)

that-

.i A.

The application for amendment filed by the Public Service Electric &

)

Gas Company (PSE&G) dated February 6,1989, as supplemented by letter j

dated May 4, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Consission; C.

There is reasonable assurance: (i)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and j

safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR l

Chapter I; 1

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

.q E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of

,j the Commission's regulations and all applicable requirements have been satisfied.

.. 1 2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby l

amended to read as follows:

i (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 26, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PSE86 shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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ADOCK 05000354 PNU

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This license amendment is effective as of its date of' issuance.

FOR THE NUCLEAR REGULATORY' COMMISSION-C.-

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Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 5, 1989 h

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ATTACHMENT TO LICENSE AMENDMENT NO. 26

' FACILITY OPERATING LICENSE NO. NPF-57 l

DOCKET NO. 50-354-

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'l Replace the_ following pages of the, Appendix "A" Technical Specifications 'with the attached pages.

The revised pages'are identified by Amendment number and-1 contain vertical lines indicating the area of change. Overleaf pages provided i

to maintain document completeness.*

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Remove Insert I

3/4.3-1 3/4 3-1 3/4 3-2*,

3/4 3-2*-

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3/4 3-5 3/4'3 I 3/4 3-6*

3/4 3-6*

l 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 8 3/4 3-1 B 3/4 3-1 I

B 3/4 3-2*

B 3/4 3-2*

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t 3/4.3 INSTRUMENTATION 1

3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protection system instrumentation channels shown in Table 3.3.1-1 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.

APPLICABILITY: As shown in Table 3.3.1-1.

1 ACTION:

With the number of.0PERABLE channels less than required by the Minimum-a.

l OPERABLE Channels per Trip System requirement for one trip system, place j

the inoperable channel (s) and/or that trip system in the tripped condiC tion

• within twelve hours.

The provisions of Specification 3.0.4 are not

[

applicable.

b.

With the number of OPERABLE. channels less than required by the Minimum l

OPERABLE Channels per Trip System requirement for both trip' systems, place at least one trip system ** in the trhped condition within one hour and take the ACTION required by Table 3.3.1-1.

j SURVEILLANCE REQUIREMENTS l

4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL

.j CONDITIONS and at the frequencies shown-in Table 4.3.1.1-1.

I 4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated autoraatic operation of all channels shall be performed at least once per 18 months.

4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.

l "An inoperega channel need not be placed in the tripped condition where this l

would caud6.the Trip Function to occur.

In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the ACTION required by l

Table 3.3.1-1 for that Trip Function shall be taken.

• • If more channels are inoperable in one trip system than in the other, place the trip system with more inoperable channels in the tripped condition, except when this would cause the Trip Function to occur.

l HOPE CREEK 3/4 3-1 Amendment No. 26 i

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. REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the' tripped condition provided at least one OPERfSLE channel in the same trip system is monitoring that parameter.

(b) This' function shall be automatically bypassed when the reactor mode switch is in the Run position.

(c) Unless adequate shutdown margin has been demonstrated per Specifica-tion 3.1.1, the " shorting links" shall be removed from the RPS circuitry prior to and during the, time any control' rod is withdrawn *.

(d) The non-coincident NMS reactor. trip function logic is such that all channels go to both trip systems. Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per the Trip System are 4 APRMS, 6 IRMS and 2 SRMS.

(e) - An APRM channel is inoperable if there are less than 2' LPRM inputs per.

level or less than 14 LPRM inputs to ~an APRM channel.

(f) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.

(g) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.-

(j) This function shall be automatically bypassed when turbine first stage pressure is < 153.3 psig** equivalent to THERMAL POWER less than 30% of RATED THERMAE POWER. To allow for instrument accuracy, calibration, and drift, ytpoint of < 132.4 psig** is used.

(k) AlsoactiirtestheEOC-RPTsystem.

• Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

• • Initial setpoint.

Final'setpoint to be determined'during the startup test

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4 3/4.3 INSTRUMENTATION V

1 BASES

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3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically. initiates 'a reactor scram to:

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. Preserve the integrity.of the fuel cladding.

I a.

i b.

Preserve'_the integrity of the reactor coolant system.

Minimize the energy which must,be adsorbed following 'a. loss-of-coolant.

O c.

accident,'and d.

Preventinadvertent.. criticality.

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This specification provides the limiting' conditions forl operation: necessary :

i to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of-service because of main--

l tenance.,When necessary, one channel may be made inoperable for:briefl intervals

)

to conduct required surveillance.

The reactor protection. system is made.up of two independent trip? systems.

There are.usually four channels to monitor each parameter with-two channels in j

3 each trip system. :The outputs of the channels in'a trip system are combined.

1 in a logic so that.either channel will trip that: trip system. The. tripping of

~j both trip systems will produce a reactor scram. :The system meets the-intent

~j of IEEE-279 for nuclear power plant protection systems.-

Specified surveillance d

intervals and surveillance and maintenance. outage. times have been determined 1

in accordance with NEDC-30851P, " Technical Specification ~.-Improvement Analyses'-

for BWR Reactor Protection. System," as approved by the NRC_and documented in-j.

the SER (letter to T. A. Pickens from A. Thadani dated' July 15,1987)..The bases for the trip settings of the RPS are discussed in the bases for Specifi-cation 2.2.1.

y The measurement of recponse time at the specified frequencies'provides:-

assurance that the protective functions associated with each channel:are com-q plated within-the time limit' assumed in the safety analyses.

No' credit'wasi taken for these channels with response times indicated as not; applicable..

i Response ties may be demonstrated by any series of sequential.: overlapping?

or total channel test measurement, provided such tests demonstrate the total-channel response time.as defined. _ Sensor response. time verification may be.

demonstrated by either. (1) inplace, onsite or offsite test measurements, or

?

(2) utilizing replacement sensors with certified response times' l

HOPE CREEK B 3/4 3-1 Amendment No. 26 I

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INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION 1

This specification ensures the effectiveness of the instrumentation used 3

to mitigate the consequences of accidents.by prescribing the OPERA 81LITY trip j

setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for.brief intervals to conduct required surveillance.

Some of the trip settings may have tolarances explicitly stated where both the high and low values are critical and may.have a substantial effect on safety.- The set-points of other instrumentation, where only the high or low and of the setting l

have a' direct bearing on safety, are established at a level away from the normal-operating range to prevent inadvertent actuation of the systems involved.

j Ex;ept for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the' sensors are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move.

For A.C, operated valves, it is assumed W. the.

A.C. power supply is lost and is restored by startup of the emergency ofesel generators.

In this event, a time of 13 seconds is assumed before the. valve starts to move.

In addition te the pipe break, the failure of the D.C. operated valve is assumed; thus.the signal delay (sensor response) is concurrent with the 10 second diesel startup. The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunc-tion with the 13 second delay.

It follows that checking the valve speeds and the 13 second time for emergency power establishment will establish the response time for the isolation functions.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION-INSTRUMENTATION

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The emergency core cooling' system actuation instrumentation is provided to initiate appens to mitigate the consequences of accidents.that are beyond l

the ability ef/the operator to control. This specification provides the OPERABILI1Y W rements, trip setpoints and response times that will ensure i

effectiveness V the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used

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to send the actuation signal to more than one system at the same time.

1 Operation with a trip set less conservative than its Trip Setpoint but l

within its specified Allowable Value is acceptable on the basis'that the difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety analyses, i

HOPE CREEK B 3/4 3-2

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