ML20125B624

Amend 160 to License DPR-16,deleting auto-start Logic of Containment Spray Sys by Plant Mods to Be Performed in 14R Refueling Outage

ML20125B624
Person / Time
Site:	Oyster Creek
Issue date:	12/04/1992
From:	Stolz J Office of Nuclear Reactor Regulation
To:	GPU Nuclear Corp, Jersey Central Power & Light Co
Shared Package
ML20125B627	List: ML20125B624 ML20125B628 ML20125B641
References
DPR-16-A-160 NUDOCS 9212100088
Download: ML20125B624 (12)

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UNITED STATES

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GPU NUCLEAR CORPORATION AE JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 160 License No. DPR-16 1.

The Nuclear Re9 latery Commiscian (the Commission) has found that:

A.

The application for amend.nent by GPU Nuclear Corporation, et al.,

(the licensee), dated February 19, 1992, complies with the standards-and requirements of the Atomic Enerr3 Act of 1954, as a:nended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the y

provisions of the Act, and the rules and regulations of the-Comission; C.

There is reasonable assurance (i) that the activities-authorized by thfs amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliancs with the Comission's regulations; D.

The -issuance of this amendment will not be-inimical to the comon-

, defen*e and security or to the health and safety of the public; and E.

.The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all appihble requirements have been satisfied.

-9212100088 921204 PDR ADOCK 05000219 P.

PDR l

4 f 2.

Accordingly, the' license is amended by changes to the Technical Specifications as indicated in the cttachment to this license amendment, and paragraph 2.C.(;) of Facility Operating License No. DPR-16 is hereby amended to read as follows:

(2) Igchnical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.160, are hereby incorporated in the license.

GPU Nuclear Corporation sha11' operate the facility in-accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Jo n F. Stolz, Direct P ject Directorate I-4 ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 4, 1992 L

-e ATTACHMENT TO LICENSE AMENDMENT NO.160 FACILITY OPERATING LICENSE N0 DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indic: ting the areas of change.

Remove Insert Pages 1.0-3 Pages 1.0-3 3.1-4 3.1-4 3.1-4a 3.1-11 (Table 3.1.1) 3.1-11 (Table 3.1.1) 3.1-14 3.1-14 3.1-15 3.1-15 3.4-8 3.4-8 4.1-9 (Table 4.1.2) 4.1-9 (Table 4.1.2) 4.4-2 4.4-2 h

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1.14 SECONDARY CONTAINMENT INTEGRITY Secondary containment integrity means that tha reactor building is closed and the following conditions are met:

A.

At least one door at each access opening is closed.

B.

The standby treatment system is operable.

C.

All reactor building ventilation system automatic isolation valves are operable or are secured in the closed position, i

1.15 (DELETED) 1.16 RATED FLUX Rated flux is the neutron flux that corresponds to a steady state sower level of 1930 NW(t).

Use of the term 100 percent also refers to tie 1930 thermal megawatt power level.

1.17 REACTOR THERMAL POWER-TO WATER 4

Reactor thenaal power-to-water is the sum of (1) the instantaneous integral over the entire fuel clad outer surface of the product of heat transfer area increment and position dependent heat flux and (2) the instantaneous rate of energy deposition by neutron and gamma reactions in all the water and core components except fuel rods in ti,e cylindrical volume defined by the active core height and the inner surface of the core shroud.

1.18 PROTECTIVE INSTPyMENTATION LOGIC _EEFINITIONS A.

Instrument Channel An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.

B.

Trio System A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping of a single trip system (e.g., initiation of a core spray loop, automatic depressurization, isolation of an isolation condenser, offgas system isolation, reactor building isolation, standby gas treatment and rod block) or the coincident tripping of two trip systems (e.g., initiation of scram, isolation condenser, reactor isolation, and primary containment isolation).

0YSTER CREEK 1.0-3 Amendment No.: 10,160 Change 7

4 particula: protection instrument is not required; or the plant is placed in the protection or safe condition that the instrument initiates.

This is accomplished in a normal manner without subjecting the plant to abnormal operations conditions. The action and out-of-service requirements apply to all instrumentation within a particular function, e.g.,

if the requirements on any one of the ten scram functions cannot be met then control rods shall be inserted.

The trip level settings not specified in Specification 2.3 have been included in this specification. The bases for these settings are discussed below.

The high drywell pressure trip setting is 1 3.5 psig.

This trip will scram the reactor, initiate core spray, initiate primary containment isolation, initiate automatic depressurization in conjunction with low-low-low-reactor water level, initiate the standby gas treatment system and isolate the reactor building.

The scram function shuts the core down during the loss-of-coolant accidents.

A steam leak of about 15 gpm and a liquid leak of about 35 gpm from the primary system will cause drywell pressure in reach the scram point; and, therefore, the scram provides protection for breaks greater thar. the above.

High drywell pressure provides a second means of initiating the core spray to mitigate the consequences of loss-of-coolant accident.

Its trip setting of 13.5 psig initiates the core spray in time to provide adequate core cooling. The break size coverage of high drywell pressure was discussed above.

Low-low water level and high drywell pressure in addition to initiating core spray also causes isolation valve closure.

These settings are adequate to cause isclation to minimize the offsite dose within required limits.

l It is permissible to make the drywell pressure instrument channels j

inoperable during performance of the integrated primary containment l

1eakage rate test provided the reactor is in the cold shutdown condition. The reason for t'ais is that the Engineered Safety Feature,,

l which are effective in case af a LOCA under these conditions, will st,ill be effective because they will be.Tctivated (when the Engineered Safety Features system is regiired as identified in the technical specification of the system) f,y low-low reactor water level.*

The scram discharge volume has two separate instrument volumes utilized to detect water accumulation. The high water level is based on the design that the water in the SDIV's, as detected by either set of level l

instruments, shall not be allowed to exceed 29.0 gallons; thereby, i

permitting 137 control rods to scram. To provide further margin, an l

accumulation of not more than 14.0 gallons of water, as detected by l

either instrument volume, will result in a rod block M an alarm.

The l

accumulation of not more than 7.0 gallons of water, as detected in either instrument volume will result in an alarm.

0YSTER CREEK 3.1-4 Amendment No: 20, 73, 79, 112, 149, 152,160

• Correction:

11/30/87

].-

3 2

Detailed analyses of transients have shown that sufficient protection is 4

provided by other scrams below 45% power to permit bypassing of the turbine trip and generator load rejection scrams.

However, for operational convenience, 40% of rated power has been chosen as the setpoint below which these trips are bypassed. This setpoint is coincidcnt with bypass valve capacity.

A low condenser vacuum scram trip of 20 inches Hg has been provided to protect the main condenser in the event that vacuum is lost. A loss of condenser vacuum would cause the turbine stop valves to close, resulting in a turbine trip transient.

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t 0YSTER CREEK 3.1-4a Amendment No: 20,73,79,112,149,15%

• Correction:

11/30/87 160

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TfBLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONT'D)

Reactor Modes Min. No. of Min. No. of in which Function Operable or Instrument Must Be Doerable Operating Channels Per Trip

[ tripped]

Operable Action Function Settina Shutdown Refuel Stattug Run Trio Systems Trio Systems Reautred*

0. Core Soray Consider the

1. Low-Low Reactor *"

X(t)

X(t)

X(t)

X 2

2 respective Water Level core spray loop ' inoperable, and comply with

~

2. High Drywell s 3.5 psig X(t)

X(t)

X(t)

X 2(k) 2(k)

Pressure

3. Low Reactor 1 285 psig X(t)

X(t)

X(t)

X 2

2 Pressure (valve permissive)

E. Containment Sorav Comply with Technical Specification 3.4 F. Primary Containment Isolation

l. High Drywell 1 3.5 psig-X(u)

X(u)

X(u)

X 2(kj 2(k)

Isolate Pressure containment or place in

2. Low-Low Reactor 1 7'2"'above X(u).

X(u)

X(u)

X 2

2 cold shutdown Water Level top of condition active fuel l

OYSTER CREEK 3.1-11 Amendment No.: 44 '79, 112.

Change 4 16d Correction: 5/11/84

TABLE.3.1.1 PROTECTIVE' INSTRUMENTATION REQUIREMENTS (CONT'D)

Reactor Modes Min. No. of Min. No. of in.which Function Operable or Instrument Must Be Operable Operating-Channels Per a

Trip'

[ tripped]

Operable Action Function LSettina.

. Shutdown Refuel Startuo Run Trio Systems Ir_in Systems Reautred* '

6. IRM Uoscale.

<'108/125 fullscale '

X X

2.

3

7. a) water level 514 gallons X(z)

X(z).

X(z)-

1

'l per high scram-instrum.

i discharge

. volume j

volume North b) water level Ls 14 gallons.

X(z)

X(z)

X(z) 1 1 per high-scram-volume instrum.

discharge.

volume South s

L. Condenser Vacuum Puno Isolation

1. High Radiationi s 10 x Normal lDuring Startup and 2

2 Insert-l l'

in Main' Steam background

'Run when vacuum pump 1 Control Rods l

Tunnel operating.

[

i

~M. Diesel Generator Time delay-after i

load Seouence energization of j.

Timers

' relay

.15%

X.

X X

X

'2(m) 1(n).

Consider the pump l

1. CRD pump 60 sec inoperable and comply with Spec. 3.4.D (see E

Note q) i l

1 i

OYSTER CREEK 3.1 Amendment No.: 15 44, 60, 63 16d i

________4,--,___.

. [

f TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONT'D) f Reactor Modes Min. No. of Min. No. of in'which Function Operable or Instrument i

Must Be Operable Operating Channels Per Trip

[ tripped]

Operable Action Function Settina Shutdown Refuel Startuo Run Trio Systems Trio Systems Egauired*

2. Service Water 120 sec.

15% X X

X X

2(o) 2(p)

Consider the pump i

Pump (aa)

(SKlA) inoperable and 10 sec. i 15%

comply within (SK2A).

7 days (See (SK7A)

Note q)

(SK8A)

3. Closed Cooling 166 Sec.i 15% X X

X X

2(m) 1(n)

Consider the pump Water Pump (bb) inoperable and comply within 7 days (See Note q)

N. Loss of Power

a. 4.16KV Emergency **

X(ff)

X(ff)

X(ff) X(ff) 2 1

Bus Undervoltage (Loss of Voltage).

b. 4.16 KV Emergency **

X(ff)

X(ff)

X(ff) X(ff) 2 3

See' note ee Bus underveltage (Degraded Voltage)

OYSTER CREEK 3.1-15 Amendment No.: 15, 60, i

73, 80 160

The containment spray system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. Actuation of the containment spray system in accordance with plant emergency operating procedures ensures that containment and torus pressure and temperature conditions are within the design basis for containment integrity, EQ, and core spray NPSH requirements._ The flow from one pump in either loop is more than ample to provide the required heat removal capability (2). The emergency service water system provides cooling to the containment spray heat exchangers and, therefore, is required to provide the ultimate heat sink for the energy release in the event of a loss-of-coolant accident. The emergency service water pumping requirements are those which correspond to containment cooling heat exchanger performance implicit in the containment cooling description.

Since the loss-of-coolant accident while in the cold shutdown condition would i

not require containment spray, the system may be deactivated to permit integrated leak rate testing of the primary containment while the reactor is in the cold shutdown condition.

The control rod drive hydraulic system can provide high pressure coolant injection capability.

For break sizes up to 0.002 ft, a single control rod drive pump with a flow of 110 gpm is adequate for maintaining the water level nearly five feet above the core, thus alleviating the necessity for auto-relief actuation (3).

The core spray main pump compartments and containment spray pump compartments were provided with water-tight doors (4).

Specification 3.4.E ensures that the doors are in place to perform their intended function.

Similarly, since _ a loss-of-coolant accident when primary containment integrity is not being maintained would not result in pressure build-up in the drywell or torus, the system may be made inoperable under these conditions. This prevents possible personnel injury associated with contact with chromated torus water.

References 1

1.

NEDC-31462P, "0yster Creek Nuclear Generating Station SAFER /COREC00L/GESTR-LOCA Loss-of-Coolant Accident Analysis,"

August 1987.

2.

Licensing Application, Amendment 32, Question 3 3.

Licensing Application, Amendment 18, Question 1 4.

Licensing Application, Amendment 18, Question 4 5.

GPUN Topical Report 053, " Thermal Limits with One Core Spray Sparger" December 1988.

6.

NEDE-30010A, " Performance Evaluation of the Oyster Creek Core Spray _

Sparger", January 1984.

7.

Letter and enclosed Safety Evaluation, Walter A. Paulson (NRC) to P. B.

Fiedler(GPUN), July 20, 1984.

8.

APED-5736, " Guidelines for Determining Safe Test Intervals and' Repair Times for Engineered Safeguards", April 1969.

0YSTER CREEK 3.4-8 Amendment No.: 153,160 n

,a e

TABLE 4.1.2 MINIMUM TEST FRE0VENCIES FOR TRIP SYSTEMS Trio System Minimum Test Freouency

1) Dual Channel (Scram)

Same as for respective instru-mentation in Table 4.1.1

2) Rod Block Same as for respective instru-mentation in Table 4.1.1

3) DELETED DELETED

4) Automatic Deoressurizatign, Each refueling outage each trip system, one at a time

5) MSIV Closurg, each closure Each refueling outage logic circuit independently (1 valve at a time)

6) Core Sorav, 1/3 mo. and each refueling each trip system, one at a time outage.

7) Primary Containment Isolation. each Each refueling outage closure circuit independently (1 valve at a time)

8) Refuelina Interlocki Prior to each refueling operation

9) Isolation Condenser Actuation Each refueling outage and Isolation, each trip circuit independently (1 valve at a time)

10) Reactor Buildino Isolation Same as for respective and SGTS Initiation instrumentation in Table 4.1.1 1

11) Condenser Vacuum Pumo Isolation Prior to each startup

12) Air Eiector Offaas line Isolation Each refueling outage

13) Containment Vent and Purge Isolation 1/20 mo.

0YSTER CREEK 4.1-9 Amendment No.: 108, 116, 144,160

C.

Containment Coolina System lits f.ttninCX

2. Motor-operated valve operability Every 3 months l

3. Pump compartment water-Once/ week and after each entry tight doors closed O.

Emercency Service Water System

1. Pump Operability Once/ month. Also after major maintenance and prior to startup following a refueling outage.

E.

Control Rod Drive Hydraulic System

1. Pump Operability Once/ month. Also after major maintenance and prior to startup following a refueling outage.

F.

Fire Protection System

1. Pump and Isolation Once/ month. Also after major valve operability maintenance and prior to startup following a refueling outage.

Bases:

It is during major maintenance or repair that a system's design intent may be violated accidentally.

Therefore, a functional test is required after every major maintenance operation.

During an extended outage, such as a refueling outage, major repair and maintenance may be performed on many systems.

To be sure that these repairs on other systems do not encroach unintentionally on critical standby l

cooling systems, they should be given a functional test prior to startup, i

Motor operated pumps, valves and other active devices that are normally on standby should be exercised periodically to make sure that they are free to operate.

Motors on pumps should operate long enough to approach equilibrium temperature to ensure there is no overheat problem. Whenever practical, valves should be stroked full length to ensure that nothing impedes their motion.

Engineering judgment based on experience and availability analyses of the type presented in Appendix L of the FDSAR indicates that testing these components more often than once a month-over a lon'g period of time does not significantly improve the system reliability.

Also, at this frequency of testing wearout should not be a problem through the l

life of the plant, l

During tests of the electromatic relief valves, steam from the reactor vessel will be discharged directly to the absorption chamber pool.

Scheduling the tests in I

conjunction with the refueling outage permits the tests to be run at low power, prior to 5 percent power, enhancing the safety of the plant by assuring EMRV operability before higher power levels are reached.

1 OYSTER CREEK 4.4-2 Amendment No.: 109,160

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