Text

Amends 96,96 & 93 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Revising Tech Specs to Upgrade Engineering Safety Features Ventilation Filter Sys Surveillance Requirements
	ML20003E710
Person / Time
Site:	Oconee
Issue date:	04/01/1981
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Duke Power Co
Shared Package
ML20003E711	List: ML20003E710; ML20003E717; ML20003E719;
References
	DPR-38-A-096, DPR-47-A-096, DPR-55-A-093 NUDOCS 8104100112
	Download: ML20003E710 (31)
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50-269 OCONEE NUCLEAR STATI0ft,_, UNIT N0. 1 Af1ENDf1ENT TO FACILITY OPERATING LICEf!SE

'Anendmen* 'lo.96 License fio. DPR-38 l.

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

A.

The applications for ameridment by Duke Power Company (the licensee) dated May 1, 1979, February 16,_1981 and March 6,.-1981, ccmply with the.

standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the pro-visions of the Act, and the rules and regulations of the Co.taission; C.

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

The issuance of this amendment will not be inimical to the comnon defense 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 Conmission's regulations and all applicable requirements have been satis-fied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of l

Facility Operating License No. DPR-38 is hereby amended to read as follows:

3.8 Technical Specifications The Technical Specifications contained in Appendices A and S, as revised through Amendment No. 96 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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

i FOR THE $UCLEAR REGULATORY C0!Vi!SSION

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50- 270 OC0t;EE NUCLEAR STATI0ft, UNIT NO. 2 AtENDI'ENT TO FACILITY OPERATING LICENSE A.mendment tio. 96 '

License llo. DPR-47 1.

The Nuclear Regulatory Co=nission (the Comission).has found that:

A.

The applications for ame:idment by Duke Power Company (the licensee) dated May 1, 1979, February 16, 1981 and March 6, 1981, comp'/ with the standards and requirements of the Atomic Energy Act of 19:4, as amended (the Act), and the Co=nission's rules and regulations set'forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the pro-visions of the Act, and the rules and regulations of the Commission;

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There is reasonable assurance (i) that the activities authorize.d by this amendnent can be conducted without endangering the health and safety of-the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance.of this amendment will not be inimical to the comnon defense and security or to the health and safety of the public; and E.

The issuance nf this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satis-fied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Facility Operating License No. DPR-47 is hereby amended to read as follows:

3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 96 are hereby incorporated in the ifcense.

The licensee shall cperate the facility in accordance with the Technical Specifications.

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s This license acendment is effective as of the date of its is;uance.'

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3 FOR THE NUCLEAR REGULATORY C0!C11SSION

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Changes ~ to the -Technical i

Specifications Cate of Issuance: - April 1,1981 s

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NUCLEAR REGULATORY COT.**.11SSION

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50-287 OC0!:EE NUCLEAR STATI0ft, UtilT !!0. 3 A!!Ef;DME!iT TO FACILITY OPEPATit;G LICENSE Amendment'!!o.93 License rio. DPP-55 1.

The Nuclear Regulatory Commission (the Conmission) has fcund that:

A.

The applications for amendment by Duke Power Company 'the licensee) dated f4ay 1, 1979, Feoruary 16,-1981 and March 6, 1981, cc.: ply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the pro-visions of the Act, and the rules and regulations of the Cocmission; C.

There is reasonable assurance (i) that the activities authorized by this amendnent can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comnission's regulations; D.

The issuance of this amendment will not be-inimical to the comnon defense 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-Co: mission's regulations and all applicable requit

. ente % ve been satis-fied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachnent to this licerse anendment and paragraph 3.B of Facility Operating License fio. OPR-55 is hereby amended to read as follows:

3.8 Technical Specifications The Technical Specifica'tions contained in Appendices A and B, as revised through Amendment tio. 93 are hereby incorporated in the license.

The license.e shall operate the facility in accordence with the Technical Specifications.

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This license amendment is effective as of the-date of its issuance.

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FOR THE NUCLEAR REGULATORY COMMISSION i

Jo n F. Stolz, Chief 0 erating-Reactors Branch #4 ivision of Licensing -

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Attachment:

Changes to the Technical Specifications Date of Issuance: April 1, 1981 J

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ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO. 96 TO DPR-38 AMENDMENT NO. 96 TO DPR-47 AMENDMENT NO. 93 TO DPR-55 00CKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of *-

.gpencix ~ "A" Technical Specifications with the attached pages. Tt.e revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.

REMOVE PAGES INSERT PAGES 3.5-4 3.5-4 3.5-5 3.5-5 3.5-5a 3.5-Sa 4.1-8 4.1-8 4.4-10 4.4-10 4.4-11 4.4-11

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4.4-12 4.5-1 4.5-1 4.5-2 4.5-2 4.5-5 4.5-5

4.5-6 4.5-6 4.5-10 4.5-10 4.5-11 4.5-11 4.5-12 4.5-12 l

4.6-1 4.6-1 4.5-2 4.6-2 4.6-3 4.6 4.7-1

4. 7-1 4.7-2 4.7-2*

4.10-1 4.10-1 a.12-1 4.12-1 4.14-1 4.14-1 4.14-2 4.14-2 i

4.19-1 d.19-1

No change on this. page; provided for-convenience only.

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TABLE 3.5.1-1 INSTRUMENTS OPERATING C 'DITIONS (A)

(B)

(C)

Minimum Minimum Operator Action If Conditions Operable Degree of Of Column A and B Functional Unit Channels Redundancy Cannot Be Met 1.

Nuclear Instrumentation 1

0 Bring to hot shutdown within.

Intermediate Range 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (b)

Channels 2.

Nuclear Instrumentation 1

0

. Bring to hot shutdown within Source Range Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (b)(c) 3.

RPS Manual Pushbutton 1

0 Bring to hot shutdown within.

12' hours

'. RPS Power Range 3(a) 1(a)

Bring to hot shutdown'within Instrument Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5.

RPS Reactor Coolant 2(d) 1 Bring to hot shutdown within

.Te=perature Instrument 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Channels

6..RPS Pressure-Tenperature 2(d) 1 Bring to hot' shutdown within-Instruments Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 7.

RPS Flux Imbalance 2

1 Bring to hot shutdown within Flow Instrument Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 5.

RPS Reactor Coolant Pressure t

i a.

High Reactor Coolant 2

1 Bring to hot shutdown within Pressure Instrument 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

Channels b.

Low Reactor Coolant 2

1 Bring to hot shutdown within D essure Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 9.

RPS Power-Number of Pumps 2

1 Bring to hot shutdevn within Instrument Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

10.

RPS High Reactor Building 2

1 Bring to hot shutdcwn within Pressure Channels 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Trip System (9)y Reactor RPS Anticipator l

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Loss of Turbine 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

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Loss of Main Feedwater 2

1 Bring to het shutdown within

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3.5-4 I

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TA3LE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (cont'd)

(A)

(3)

(C)

Minimum Minimum Operator Action If Conditions Operable Degree of Of Column A and 3 Tusctional Unit Channels Redundancy Cannot 3e Met 12.

ESF High Pressure Injection System and Reactor Buildina Isolation (Non-essential systems) a.

Reactor Coolant 2

1 Bring to hot shutdown within Pressure Instru-12 hours (e) nent Channels b.

Reactor Building 2

1 Bring to hot shutdown within 4 PSIG Instrument

~ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Channels c.

Manual Pushbutton 2

1 Bring to hot shutdo-:n within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e) 13.

ISF Low Pressure In-

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jection System a.

Reactor Coolant 2

1 3 ring to hot shutdown within Pressure Instrument 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Channels b.

Reactor Building 2

1 3 ring to hot shutdown within.

4 PSIG Instru=ent 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Channels c.

Manual Pushbutton 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e) 14 ESF Reactor Building l

Isolation (Essential Systems)

& Reactor Building Cooling System a.

Reactor Building 2

1 Bring to het shutdown within 4 PSIG Instrc=ent 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Channel b.

Manual Pushbutton 2

1 Bring to hot shutdown wi'.hin

'12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e) 15.

ISF. Reactor Building i

Spray system I

a.

Reactor Building 2

1 3 ring to hot shutdown within High Pressure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

Instrument Channel Anendnen:s Nos. 95, 95, 93 3.5-5

TABLE 3.5.1-1 INSTRUMENTS OPERATING CONDITIONS (cont'd)

(A)

(B)

(C)

Minimum Minimum Operator Action If Conditions Operable Degree of of Column A and 3 Functional Unit Channels Redundancv Cannot Be Met b.

Manual Pushbutton 2

1 Bring to hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (e)

's.

Turbine Stop Valves 2

1-3 ring to hot shutdown within Closure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (f)

(a) For channel testing, calibration, or maintenance, the minimum number of operable channels ay be two and a degree of redundancy of one for a maximum of four hours.

Cr) When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

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(:) When 1 of 2 intermediate range instrument channels is greater than 10 a=ps, hot shutdown is not required.

(d) Single loop operation at power (after testing and approval by the NRC/ DOL) is not permitted unless the operating channels are the two receiving Reactor Coolant Temperature from operating loop.

(e) If minimum conditions are not met within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months af ter hot shutdown, the unit shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(f) One operable channel with :ero minimum degree of redundancy is allowed for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before going to the hot shutdown condition.

(;) This requirement is applicable as follows:

i Unit 1 - following Summer 1981 refueling outage Unit 2 - following Fall 1981 refueling outage Unit 3 - immediately upon the effective date of this license amendment 2-i ents Nos. 96, 96, 93 3.5-5a gi-w u-e-

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1 Channel De._s_c_r.ipt ion Cherk Test Calibrate Hemarks m

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49. E,nergency feedwater M0 flA RF 8

Flow Indicators 4

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50. PORV and Sa fety Valve 110 flA RI' Position Indicators 1

51. RPS Anticipalory NA M0 RF

<d Reactor Trip System j

. Loss of Turbine i

52. RPS Anticipatory NA H0 RF Reactor Trip System Loss of Main feedwater 1

to ES - Each Shift Qtt - Qisarterly 11A - Daily AN'- Annually WE. Weekly I10 - flonthly PS - Prior to startup, if not 'performent previous week NA - Not Applicable HF - Heiueling-Outage 4

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i 4.4.3 Hydrogen Purge System A;pli.cability A; plies to the Reactor Building Hydrogen Purge System.

Cbjective Is verify that the Reactor Duilding Hydrogen Purge System is operable.

5;ecification 4.4.3.1 In-place Testing a.

During each refueling outage, an in-place system test shall be performed. This test shall demonstrate that under sipulated emergency conditions, the system can be taken from storage and placed into operation within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b.

This refueling outage test shall consist of:

1.

Visual inspection of the system.

2.

Hock-up of the system to one of the three Reactor Buildings.

3.

Flow seasurement using flow instruments in the portable purging station.

4 Verification that the pressure drop across the combined EEPA filters and charcoal ads'orber banks is less than six inches of.

water at the system design flow rate (:10%).

5.

Verification of the operability of the heater at rated power when tested in accordance with ANSI N510-1975.

4.4.3.2 Operational Perfor=ance Testing a.

The testing requirements of this section =ay be perfor=ed withcut hooking-up the system to cne of the Reactor Buildings.

b.

. Monthly, the hydrogen purge system shall be operated with the heaters on for at least ten hours.

During each refueling outage, the hydroger purge system c.

fans shall be snown to operate at desien flow (t105) when tested in accordance with 445I N510-1975.

d.

Leak tests using DOP or halogensted hydrocarbca, as appropriate shall be perfor=ed on the hydrogen purge filters:

1 During each refueling outage; 2.

After each ec=plete or partial replace:ent of EEPA filter back or charcoal adsorber bank; Anend:ents Nos. 95, 96, 93 4.4-10

1 4

3.

After any~ structural maintenance on the system housing; 4

-After painting, i'.re, or chemical release in any venti-lation zone communicating with the system.

The results of the DOP and halogenated, hydrocarbon tests on EEPA e.

filters and charcoal adsorber banks shall show > 99% DOP removal and

> 99% halogenated hydrocarbon removal, respectively, when tested in accordance with ANSI N510-1975. Otherwise, the filter system shall-be declared inoperable.

f.

Durirg each refueling outage, following 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, or after painting, fire, or chemical release in any ventilation zone communicating with the system, a carbon sample shall be removed from.the Reacter Building purge filters for laboratory analysis. Within 31 days of removal,.

this sample shall be verified to show >90% radioactive methyl iodide removal when tested'in accordance with ANSI N510-1975 (1300C, 95% R.H.).

Otherwise, the filter system shall be declared inoperable.

4.4.3.3 H Detector Test 2

Hydrogen concentration instruments shall be calibrated each l

refueling outage with proper consideration to coisture effect, Amendments'Nes. 96, 96, 93 4.4-11

lases Pressure drop across the ec=bined high efficiency particulate air (EEPA) filters

_l and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that.the filters and adsorbers are not clogged by excessive as:unt of foreign matter. A test frequency of once per year establishes systes.

erfor: acce capability.

EEPA filters are installed before the charcoal adsorbers to prevent clogging of l

the iodice adsorbers. The charcoal adsorbers are installed to reduce the poten-tial release of radioicdice. Bypass leakage for the charcoal adsorbers and

articulate removal efficiency for HEPA filters are determined by halogenated hydrecarbon and DCP respectively. The laboratory carben sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditices. Operations of _ the fans significantly different frem the design-flev vill change the removal efficiency of the EEPA filters and charcoal adsor-hers.

If the performances are as specified, the ' calculated doses would be less than the guidelines stated in 10 CyR 100 for the accidents analyzed.

The frequency of tests and sample a:alysis are necessary to show that the EEPA filters and charcoal adsorbers can perform as evaluated. Replacement adsorbent shculd be qualified according to the guidelines of Regulatory Guide 1.52.

The charcoal adsorber efficiency test procedures should allow for the removal of one ads:rber tray, emptying of _ one bed from the tray, mixing the adsorbent thoroughly a:d obtaining at least two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.

If the iodine re:: val efficiency test results are unacceptable, all adsorbent in the system shculd be replaced. Any EEPA filters found defective should be replaced with filters qualified purauant to Regulatory Positica C.3.d of Regulatory Guide 1.52.

t Cperatien of the system every conth will de'constrate operability of the filters and adsorber system. Operation for ten hours is used to reduce the moisture built up c the adsorbent.

If painting, fire or chemical release occurs during system operation such that the EEPA filter or charcoal adsorber could b'ecome centaminated fro = the fumes, che:icals or foreign materials, the same tests and sa:ple analysis should be

erfor: ed as required for operational use.

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!. ani:ents Nos. 95, 96, 93 4 4-12 l

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4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR BUILDING COOLING SYST?i PERIODIC TESTING

. 4.5.1 Emergency Core Cooling Systems Applicability Applies to periodic testing requirements for the Emergency, Core Cooling Systems.

Objective To verify that-the Emergency Core Cooling Systems are operable.

Specification 4.5.1.1 System Tests 4.5.1.1.1 High Pressure Injection System a.

During each refueling outage, a system test shall be conducted to demonstrate that'the system is operable. A test signal-will be applied to demonstrate actuation of the High Pressure Injection System for emergency core cooling operation.

b.

.The test will be considered satisfactory if control board indication verifies '. hat all components have responded to'the actuation signal properly; all appropriate pump breakers shall have opent2 or closed and all valves shall have completed their travel.

4.5.1.1.2 Low Pressure Injection System a.

During each refueling outage, a system test shall be cenducted co demonstrate that the system is operable. The test shall be performed in accordance with the procedure summarized below:

(1) A test signal will be applied to demonstrate actuation of the Low Pressure Injection System for emergency core cooling operation.

(2) Verification of the engineered safety features function of the Low.

Pressure Service Water System which supplies cooling water co the low L

pressure coolers shall be made to demonstrate operability of the Coolers.

i b.

The test will be considered satisfactory if control ooard indication verifies that all components have responded to the actuation signal properly; all appropriate _ pump breakers shall have opened or closed, and all valves shall have completed their travel.

i 4.5.1.1.3 Core Flooding System i

During each refueling outage,- a system test shall be conducted to demon-s.

l strate proper operation of the system. During pressurization of the i

Amendments Nos. 96, 96, 93 4.5-1 i

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The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened.

5.1.2 Component Tests

. 5.1.2.1 Pumps quarterly, the high pressure and low pressure injection pumps shall be started and

perated to verify proper operation. Acceptable performance will be indicated if the pump starts, operates for 15 minutes, and the discharr.e pressure and flow are within: 10 percent of a point on the pump head curve.

(Figures 4.5.1-1 and 4.5.1-2)

'.5.1.2.2 Valves - Power Operated a.

Quarterly, each Engineered Safety Features valve in the Emergency Core Cooling Systems and each Engineered Safety Features valve associated with-emergency core cooling in the Low Pressure Service Water System shall be tested to verify operability, b.

The acceptable performance of each power-operated valve will be that motion is indicated upon actuation by appropriate signals.

c.

During each refueling outage, low pressure injection pump discharge (engineered safety features) valves, lov pressure injection discharge throttling valves, and low pressure injection discharge header crossover valves shall be cycled manually to verify the manual operability of these power-operated valves.

3ases The Emergency Core Cooling Systems are the principle reactor safety features in the event of a loss of coolant accident. The removal of heat from the

ore provided by these systems is designed to limit core damage.

The High Pressure Injection System under normal operating conditions has one pump operating. At least once per month operation is rotated to another

_igh pressure injection pump. This verifies that the high pressure injection pucps are operable.

The requirements of the Low Pressure Service Water System for cooling water are more severe during normal operation than under accident conditions.

..atatton of the pump in operation on a monthly tasis 'terifies that two pumps.

are operable.

The low pressure injection pumps are tested singularly for operability by

pe;ing the borated water. storage tank cutlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped frem the borated water storage tank thrcugh each of the injection lines and h a C /. to the tank.

4.5-2 Anendments Nos. 96, 96,'93

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1000 2000 3000 4000 5000 Capacity,gpm LOW PRESSURE INJECil0N FUMP C4ARACTERISTiCS

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kM', OCONEE NUCLEAR STAT Figure 4.5.1-2 4.5-5

'.5.2 Reactor Building Cooling Systems A;;1icability Applies to testing ' the Reactor Building Cooling Systems.

Objective

!o verify that the Reactor Building Cooling Systems are operable.

.cecification

.5.2.1 System Tests

.5.2.1.1 Reactor Building Spray System a.

During each refueling outage, a system test shall be conducted to demonstrate proper operation of the system. A test signal will be applied to demonstrate actuation of the Reactor Building Spray Sys-tem (except for reactor building inlet valves to prevent water enter-ing noznles). Water will be circulated from the borated water stor$ge tank through tha reactor building spray pu=ps and returned through the test line to the borated water storage tank.

i.

Station compressed air will be introduced into the spray headers to verify the availability of the headers and spray nozzles at least every five years.

The test will be. considered satisfactory 'if visual observation and control board indxcation verifies that all co=ponents have respended to the actuation signal properly; the appropriate pu=p breakers shall have closed, and all valves shall have completed their travel.

.5.2.1.2 Reactor Building Cooling Syst4m s.

During each refueling outage, a system test shall te conducted to de=en-

}

strate proper operation of the system.

The test shall be performed in accordance with the procedure st=marized below:

(1) A test signal will be applied to actuate the Reactor Building Cooling System for reactor building cooling operation.

Arend:ents Nos. 96, 96, 93 4.5-6

4.5.3 Penetration Room Ventilation System Applicability Applies to testing of the Penetration Rcom Ventilation System.

Objective

. To verify that the Penetration Rocm Ventilation System is operable.

~

Specification 4.5.3.1 Operational and Performance Testing Monthly, each train of.the Penetration Rocm Ventilation System a.

shall be operated for at least 15 minutes at design flow +10%.

b.

~

During each refueling outage, it shall be demonstrated that:

1.

The Penetratica Room Ventilatien System fans operate at design flow ( 10%) when tested in accordance with ANSI N510-1975.

2.

The pressure drop across the co=bined HEPA filters and char-coal adsorber banks is less than six inches of water at the system design flow rate (t 10%)

3.

Each branch of the Pentration Room Ventilation System is capable of automatic initiation.

4 The. bypass valve for filter cooling is manually operable.

Leak tests using DOP or halogenated hydrocarbon, as appropriate c.

shall be perfor=ed on the Penetration Roco purge filters:

1-During each refueling outage; 2.

After each complete or partial replacement of a HIPA filter bank or charcoal adsorber bank; 3.

After any structural maintenance on the system housing; 4

After painting, fire, or chemical release in any ventilation zone cec =unicating with the system.

d.

The results of the DOP and halogenated hydrocarbon tests en HEPA filters and charcoal adsorber banks shall show 199% DOP removal and 199% halogenated hydrocarbon removal, respectively, when tested in accordance with ANSI N510-1975.

Amendments Nos. 96, 96, 93 4.5-10

During each refueling outage, following 720 baurt of'sys.teg e.

operation, or after painting, fire, or chemical release in any ventilation zone communicating with the system, a carbon-sample shall be removed frem the Reactor Building purge filters for la5 oratory analysis. Within -31 days of removal,.

this sample shall b'e verified to show >90% radioactive methyl iodide removal when tested in accordance with

'NSI N510-1975 (1300C, 95% R.H.).

Otherwise, the filter system shall be declared inoperable.

2ases j'

?ressure drop across the combined high efficiency particulate air-(EEPA) fil-l

ers and charcoal adsorbers of less than six inches of water at the system de-sign flow rate will indicate that the filters and adsorbers are not clogged by excessive a=ounts of foreign matter. A test frequency of once per operating rycle establishes system performance capability.

IEPA filters are installed before the charcoal adsorbers to prevent clogging of j

the iodine adscrbers. The charcoal adsorbers are installed to reduce the poten-

ial release of radiciodine. Bypass leakage for the charcoal adsorbers and

articulate removal efficiency for HEPA filters are determined by halogenated.

hydrocarbon and DOP respectively. The laboratory carbon' sample test results indicate a' radioactive methyl iodide renoval efficiency for expected accident

enditions. Operation of the fans significantly different from the design ficw will change the removal efficiency of the HEPA filters and charcoal adsor-bers.

If the performances are as specified, the calculated doses would be less than the guidelines stated in.10 CFR 100 for the accidents analyzed.

The frequency of tests and sample analysis are.necessary to show that the EEPA i

filters and charcoal adsorbers can perform.as evaluated. Replacement adsorbent thculd be qualified according to the guidelines of Regu'.-tory Guide l.52.

The.

harcoal adsorber efficiency test procedures should alle. for the removal of one adsorcer tray, e=ptying of one bed from the tray, mixing the adsorbent ther-

ughly and obtaining at least two samples. Each sample should be replaced.

Any EIPA filters found defective should be r,eplaced with filters qualified

ursuant to Regulatory Positi: n C.3.d of Regulatory Guide 1.52.

peration of the system every month will demonstrate operability 'of the filters and adsorber system. Operation for 15 minutes de=enstrates operability and mini-ni:es the moisture build up during testing.

f painting, fire or chemical release occurs during system operation. such that

he EIPA filter or charcoal adsorber could become contaminated from the fumes,

hemicals or foreign materials, the same tests and sample analysis should be

erfer: ed'as required for operational use.

'er:nstration of the automatic initiation capability is necessary to assure systes performance capability.

J r

I

._ end ants ::es. 95, 96, 93 4.5-11

\\;

9 4.5.4 Low Pressure Injection System Leakage Applicability Applies to Low Pressure Injection System leakage.

Objective To maintain a preventive leakage rate for the Low Pressure Injection System which will prevent Significant off-site exposures.

Specification 4.5.4.1 Acceptance Limit The maximum allowable leakage from the Low Pressure Injection System cocponents (which includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.

4.5.4.2 Test During each refueling outage, the following tests of the Low Pressure In-l jection System shall be conducted to determine leakage:

a.

The portion of the Low Pressure Injection System, except as specified in (b), that is outside the containment shall be tested either by use in normal operation or by hydrostatically testing at 350 psig.

b.

Piping from the contain=ent e=ergency su=p to the low pressure injection pump suction isolation valve shall be pressure tested at no less than 59 psig.

c.

Visual inspection shall be made for excessive leakage from components of the system. _Any excessive leakage shall be measured by collection and weighing or by another equivalent method.

Bases The leakage rate limit for the Low Pressure Injection System is a judgment value based on assuring that the c:epent 5 can be expected to operate with-out =echanical failure for a period on tc.rder of 200 days after a loss of coolant accident. The test pressure.(350 psig) achieved either by nor=al system operation or by hydrostat:cally test:ng, gives an adequate margin over the highest pressure within the system after a design basis accident.

Similarlf, the pressure test for the return lines fr:m the centain=ent to the Low Pressure Injection System (59 psig) is equivalent to the design pressure of the containment. The dose to the thyroid calculated as a result of this leakage is 0.76 re: for a two-hour exposure at the site bocadary.

REFERENCE ISAR, Section 14.2.2.4.4 Amendments Nos. 96, 95, 93

' 5-12

b

'.6 EMERGENCY POWER PERIODIC TESTING Aerlicability Applies to the periodic testing surveillance of the emergency power sources.

" b _dective

~ o verify that the emergency power sources and equipment will respond promptly and properly when required.

icecification e.6.1 Monthly, a test of the Keowee Hydro units shall be performed to verify proper operation of these emergency power sources and associated equip-ment. This test shall assure that:

a.

Each hydro unit can be automatically started frem the Unit I and 2 control room.

b.

Each hydro unit can be synchronized through the 230 kV overhead circuit to the startup transformers.

c.

Each hydro unit can energize the 13.8 kV underground feeder.

d.

The 4160 volt startup transformer main feeder bus breakers and standby bus breaker shall be exercised.

.6.2 a.

Annually, the Keowee Hydro units will be started using the emergency start circuits in each control room to verify that each hydro unit and associated equipment is available to carry load within 25 sec-onds of a simulated requirement for engineered safety features.

b.

Promptly follewing the abovepannual test, each hydro unit will be l

loaded to at least the combined load of the auxiliaries actuated i

by'ESG signal in one unit and the auxiliaries of the other two units in hot shutdown by synchronizing the hydro unit to the off-site power system and assuming the. load at the maximum practical rate.

.6.3 Monthly. the Keowee Underground Feeder Breaker Interlock shall be l

verified to be operable.

.6.;

During each refueling outage, a simulated emergency transfer of the l

4160 volt main feeder buses to the startup transformer (i.e.. CT1, CT2 l

or CT3) and to the 4160 volt standby buses shall be made to verify l

proper operation.

.6.5 Quarterly, the External Grid Trouble Protection System logic shall be tested to demonstrate its ability to provide an isolated power path between Keowee and Oconee.

i i

.6.6 Annually and prior to planned extended Keowee outages, it shall be.

demonstrated that a Lee Station ecmbustion turbine can be started and i

i l

A:end:ents Nos. 96, 96, 93 4.6-1

t connectad to;the 100 kV line.

It shall be demonstrated that the 100 kV line can ba :eparated from the rest of the system and supply power to the 4160 volt main feeder buses.

4.6.7 At least once every 18 months, it shall be demonstrated that a Lee station combustion turbine can be started and connected to the isolated 100 kV line and carry the equivalent of the maximum safeguards load of one Oconee unit (4.8 MVA) within one hour.

4.6.8 Annually, it shall be demonstrated that a Lee station combustion turbine can be-started and carry the equivalent of the maximum safeguards load of one Oconee unit plus the safe shutdown loads of two Oconee units on the system grid.

4.6.9 Batteries in the Instrumentation and Control, Xeowee, and Switching Station shall have the following periodic inspections performed to assure maximum battery life. Any battery or cell not in compliance with these periodic inspection requirements shall be corrected to meet the requirements within 90 days or the battery shall be declared inoperable, a.

Weekly verify that:

(1) The electrolyte level of each pilot cell is in between the minimum and maximum level indication marks.

(2) The pilot cell specific gravity, corrected to 7,7 F and full electrolyte level, is 1 1.200.

(3) The pilot cell float voltage is 1 2.12 VDC.

(4) The overall battery float voltage is 1 125 VDC.

b.

Quarterly verify that:

(1) The specific gravity of each cell corrected to 77 F and full electrolyte level, is 1 1.200 and is not less than 0.010 below the average of all cells measured.

(2) The voltage of each cell under float charge is 1 2.12 VDC.

(3) The electrolyte level of each connected cell is between the mini =um and =aximum level indication marks.

c.

Annually verify that:

(1) The cells, end-cell plates and battery racks show no visual indication of structural damage or degradation.

(2) The cell to cell and terminal connections are clean, tight and coated with anti-corrosion grease.

Amendments Nos. 96, 96, 93 4.6-2

e 4.6.10 Annually, a one hour discharge service test at the required maximum load shall be made on the instrument and control batteries, the Keowee batteries, and the switching station batteries.

4.6.11 Monthly, the operability of the individual diode monitors in the.Instru-ment and Control Power System shall be verified by imposing a simulated diode failure signal on the monitor.

4.6.12 Semiannually, the peak inverse voltage capability of each auctioneering diode in the 125 VDC Instrument and Control Power System shall be measured and recorded.

Bases The Keowee Hydro units, in addition to serving as the emergency power sources for the Oconee Nuclear Station, are power generating sources for the Duke system requirements. As power generating units, they are operated frequently, normally on a daily basis at loads equal to or greater than required by Table 8.5 of the FSAR for ESF bus loads. Normal as well as emergency startup and operation of these units will be from the Oconee Unit 1 and 2 Control Room. The frequent starting and loading of these units to meet Duke system power requirements assures the continuous availability for emergency power for the Oconee auxiliaries and engineered safety features equipment. It will be verified that these units will carry the equipment of the maximum safeguards load within 25 seconds, including instrumentation lag, after a simulated re-quirement for engineered safety features. To further assure the reliability of these units as emergency power sources, they will be, as specified, tested for automatic start on a monthly basis from the Oconee control room. These tests will include verification that each unit can be synchronized to the 230 kV bus and that each unit can energize the 13.8 kV underground feeder.

Tse interval specified for testing of transfer to emergeucy power sources is based on maintaining maximum availability of redundant power sources.

Starting a Lee Station gas turbine, maration of the 100 kV line from the

^

remainder of the system, and char' of the 4160 volt main feeder buses are s

specified to assure the continu and operability of this equipment. The one o

hour time limit is considered tue absolute maximum time limit that would be required to accomplish this.

REFERENCE FSAR Section 8 i

Amendments Nos. 96, 96, a 93 4.6-3 i

L

4.7 REACTOR CONTROL ROD SYSTEM TESTS -4.7.'l Control Rod Trip Insertion Time Test Applicability Applies to the surveillance of the centrol rod trip insertion time.

Objective To assure the control rod trip insertion time is within that used in the safety analyses.

Specifiestion The control rod insertion time shall be measured at either full flow or no flow conditions as follows:

For all rods following each removal of the reactor vessel head, a.

b.

For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and c.

For all rods at least once following erch refueling outage.

l The maximum control rod trip insert on time for an operable control rod drive mechanism, except for the Axial Power Shaping Rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconds at reactor coolant full flow conditions or 1.40 seconds for no flow conditions. For the APSFs it shall be demonstrated that loss of power will not cause rod movement.

If the trip insertion time above is not met, the rod shall be declared inoperable.

Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The specified trip time is based upon the safety analysis in FSAR Chapter 14 A rod is considered incperable if the trip insertion time is greater than the specified allowable time.

REFERENCES (1) FSAR, Section 14 (2) Technical Specification 3.5.2 Amendnents !!cs. 95, 96, & 93 4.7-1

.7.2 Cont rol Rod Procram Veri fication 4;:icability Iqpl;us to surveillance of the control rod systems.

": i e :t f've

~c verify that the designated cont rol rod (by core position l'through 69) is cperating in its programmed funct ional position and group.

hrough 12, Groups 1-8)

(Rod 1

acification

.7.2.1 Whenev'er the control rod drive patch panel is locked '(af ter in-spection, test, reprogramming, or maintenance).cach control rodt drive mechanism shall be selected from the control room and exercised by a movement of approximately two inches to verify that the proper rod has responded as shown on the unit computer printout-of that rod.

.7.2.2 Whenever power or ins trumentation-cables to ' the control rod drive.

asse-blies atop the reactor or at the bulkhead are disconnected or removed, an independent verification check of_their reconnection shall be performed.

4 Isses w

I2:5 centrol rod has a relative and an absolute position indicator system.

C ne set of outputs goes to the plant computer, identified by a unique number

'l thrcugh 69) associated with only one core position. The other set of

a: puts gees to a prograemable bank of,69 edgewise meters'in the control rocm.

n the event that a patching -error is made in the patch panel'or cc ne: tors in the cables leading to th'e control ;oo drive assemblies or-to the control room meter bank are improperly t;ansposed upon reconnection,

nsse errers and transpositions will be discovered by a comparat've check b ;- :

(1) selecting a specific rod frcm or.e group (e.g., Rod 1 in Regulating Orc:p 6), (2) noting that the program-approved cora position for this red f he ;rcup (assume the approved ccre position 'is No. 53), -(3) exercising

e selected rod and (4) noting that.the computer prints out both absolute relctive position response for the, approved core position (assumed to be

sition No. 53) and that the proper meter. responds in the control' room

is;1ay bank (assumed to be Rqd 1 in Group 6) for both absolute and relative -

r e:3r p:sitions. This type of comparative check'will not assure Ldetection

[

c f. imprcperly connected cables inside the reactor building.

For these,'it is n a:1ssary for a tecr nsible person, other than the one doing the work, to va:if;. by appropriate means that each cable has been matched to the proper.

::rcl~ rod drive sssembly.

i

(

4.7-2 L

k_

e 4.10 FIACTIVITY ANOMALIES Applicability Applies to potential reactivity anomalies.

Objective To require the evaluation of reactivity anomalies of a specified magnitude occurring during the operation of the unit.

Specification Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be periodically compared with the predicted value.

If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one per-cent in reactivity, an evaluation as to the cause of discrepancy shall be made and reported to the Nuclear Regulatory Commission.

Bases To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between

)

fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. When full power is reached initially, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point. As power operation proceeds, the measured boron concentration is compared with the predicted con-centration and the slope of the curve relating burnup and reactivity is compared with that predicted. This process of normalization should be com-pleted after about 10% of the total core burnup. Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated. Any reactivity anoma.y greater than 1% would be unexpected, and its occurrence would be thoroughly investigated and evaluated.

The value of 1% is considered a safe limit since a shutdown margin of at least 1% with the most reactive rod in the fully withdrawn position is always maintained.

Amendments tics. 96, 96, 3 93 4.10-1

o I

'.12 CONTROL ROOM FILTERING SYSTEM A;plicability A; plies to control room filtering system components Cbjective

!) eerify that these systems and components will be able to perform their design functions.

f;erification 4.12.1 Operating Tests 2/s em tests shall be performed quarterly. These tests shall consist of v;sual inspection, a flow measurement at t.he outlet of each unit and pressure drop neasurements across each filter bank. Pressure drop across pre-filter shall not exceed 1 inch H. ' and pressure drop across EEPA shall not exceed inches E 0.

Fan motors shall be operated continuously for at.least one h:ur, and all louvers and other mechanical systems shall be proven operable.

a.12.2 Filter Tests Curing each refueling outage, for the Unit I and 2 and the Unit 3 control recm a: in place leakage test using DOP on EEPA units and Freon-112 (or equivalent)

charcoal units shall be performed at design flow on each filter train.

Re-

val of 99.5 percent DOP by each entire EEPA filter unit and removal of 99.0 percent Freon-112 (or equivalent) by each entire charcoal adsorber unit shall c:nsti:ute acceptable performance. These t&sts nurt also be perfor=ed after 4:y maintenance which cay affect the structural integrity of either the filtra-tion system units or cf the housing.

32ses

,s'

!ie purpose of the Control Room Filtering System is to limit the particulate a:3 gasecus fissica products to which the control area would be subjected c; ring an accidental radioactive release in or near the Auxiliary Suilding.

The system is designed with two 100 percent capacity filter trains each of w:ich consists of a prefilter, high efficiency particulate filters, charcoal fil ers and a booster fan to pressurine the control room with outside air.

5 n:e these systems are not normally operated, a periodic test is required :o i:sure their operability when needed. Quarterly testing of this system will s::v tha:-:he system is available for its safety action. During :his tes:

te system will be inspected for such things as-water, oil, or other foreign.

naterial, gasket deterioration, adhesive deterior:: ion in the HIPA units, and u: usual or excessive no:se or vibration when the fan motor is running.

Refueling outage' testing will verify the efficiency of the charcoal and abso-l

'.:e fil:ers, s

Ami : i :s "os. 95,' 9 5, & 93

4.14 RI.ACTCR SUIDING PURGE FILTERS AND SPENT yCEL PCOL VENTILATICN SYSTEM Applicability Applies to testing of the Reactor Building purge filters for Units 2 and 3 and the respective spent fuel pool ventilatica systems.

Cbfective To verify that the Reactor Building purge filters will perform their design functic: and that when used with the respective spect fuel pool ventilation syste=, uill reduce the off-site dose due to a fuel handling accident.

Specificiatic:

4.14.1 Operati::al and Perforsacee Testing Mc thly, each train of the spect fuel pool ventilatica systes a.

shall be operated throu-h the respective Reactor Building purge filters for at least 15 :::utes at design flev i 10%.

b.

During each refueling outage, the spent fuel pool ventilatics fans shall be shewn to opera:e at design flew i 10% when tested in accordance with ANSI 5510-1975.

c.

Leak tests using DOP or halogenated hydrocarbon, as appr:priate, shall be perfor=ed c the Reacter 3cilding purge filters:

1.

Durinz each refueling outage; 2.

After each c:=plete or partial replacement of HIPA filter back or charc:a1 adscrber back; 3.

After a=y structural caintenance ce the syste= h:using; 4

After painting, fire, or che=ical release in any ventila-tica rene ec==unicati:g with the syste=.

d.

The results of the DOP and halogenated hyd:ccarbc tests :

EEPA filters and charccal adscrber banks shall shev > 99% DCP reseval and 1 99% haloge:ated tyd::carbc2 re== val, respectively, whe: tested in accordance with ANSI N510-1975.

I Ouring each refueling cutage, folicving 720 hcurs Of syste= cpera-g tica, er af:er painting, fire, er che ical re: ease in any venti-g la:ica

---a

- unicating with the system, a c arben sample shall be i

re:cred frc= :he Reac:cr 3uilding purge filie a fer labcrc cry analvs is. "ithin 31 davs of re val, this sa:ple shall be verifiec :n shov 190; radicactive : ethyl i:dide re=cval vie:

tested in ecordance with ANSI N510-1975 (130', 95% R.R.).

Ctherwise, the filter syste= shall be declared incperable.

Amendments Ncs. 95 ~, gs, & 93

o lases The Unit 2 Reactor Building purge filter is used in.the ventilation system for the com=on spent fuel pool for Units 1 and 2.

The Unit 3 Reactor Building purge filter-is used in the Unit 3 spent fuel pool ventilation system. Each filter is ec:structed with a prefilter, an absolute filter and a charcoal filter in series. The high efficiency particulate air (EEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The char-coal adsorbers are installed to reduce the potential release of radiciodine.

Iypass leakage for the charcoal adsorbers and particulate removal efficiency for EIPA filters are determined by halogenated hydrocarbon and DOP respectively.

The laboratory carben sample test results indicate a radioactive methyl iodide re= oval efficiency for expected accident conditions. Operation of the fans sig=ificantly different from the design flow will change the removal efficiency cf the EEPA filters and charcoal adsorbers.

If the perform ~ances are as specified, tne doses for a fuel handling accident would be minimized.

Tne frequency of tests and sample analysis are necessary to show that the EEPA filters and charcoal adsorbers can perform as evaluated. Replacement adsor-bent should be qualified according to the guidelines of Regulatory Guide 1.52.

The charcoal adsorber efficiency test procedures should allow for the removal cf cne adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two'sacples. Each saeple should be replaced.

Any HIPA filters found defective should be replaced with filters qualified p :suant to Regulatory Position C.3.d of Regulatory Guide 1.52.

Operation of the spent fuel pool ventilation system every month will demonstrate cperability of the fans, filters and adsorber system.

f painting, fire or chemical release occur's during system operation such that the EIPA filter or charcoal adsorber could beco=e contaminated from the fumes, ene=icals or foreign materials,,the same tests and sample analysis should be performed as required for operational use.

(

' Ari : tr.ts rios. 96 96, & 93 4.14-2

t 6

4.19 FIRE PROTECTION AND DETECTION SYSTEM Applicability Applies to the fire protection and detection systems which' protect systems and equipment required for safe shutdown.

Objective To verify the operability of fire protection and detection systems.

Specifications 4.19.1 The High Pressure Fire Protection System components shall be tested as follows:

Item Frequency (a) High pressure service water pump Monthly functional test (b) System functional test Every 18 months (c) High pressure service water pump Annually capacity test to verify flow of 3000 gpm (d) System Flow Tes't in Accordance with Every 3' years Chapter 5, Section 11 of the Fire Drotection Handbook, 14th Edition, hTPA (e) Alignment of fire protection valves Monthly (f) Sprinkler systems in safety related areas 1.

Systen functional test Each refueling 2.

Inspection of spray headers Annually

3.

Inspection of spray nozzle Annually *

(g) Fire hose stations 1.

Visual inspection Monthly

2.

Maintenance inspection Annually

3.

Partial opening of fire hose Every 3 years station valve 4.

Hose Hydrostatic test at least Every 3 years 50 psig greater than the maximum pressure at the station

This frequency applies only for areas which are normally accessible during operation.

If an area is inaccessible during operation, inspections shall be performed in those areas during each refueling outage.

Amendments fios. 96 95, & 93 4.19-1

,

ML20003E710

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