ML20084U078

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Proposed Tech Specs Reflecting one-time Extension of Performance Intervals of Certain SRs
ML20084U078
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 06/09/1995
From:
CENTERIOR ENERGY
To:
Shared Package
ML20084U065 List:
References
NUDOCS 9506130081
Download: ML20084U078 (13)
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Text

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PY-CEI/NRR-1954L Attachm nt 2 '

Page 1 of 5 APPLICABILITY i SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other cofiditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.  ;

4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowabl-e extension not to exceed 25 percent of.the specified surveillance interval.

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4.0.3 FailuretopeIforsaSurveillanceRequirementwithintheallowedsurveil-lance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPEPABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified -

that a Surveillance Requirement has not beemperformed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to pennit th'e completion of the surveillance when the allowable outage time' limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment. ,

4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condi-tion shall not be made unless the Surveillance , Requirement (s) associated with the Limiting Condit. ion for Operation have been performed within the applicable surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. -

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and

. Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Comission pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Bqbr and Pressure Vessel Code and applicable Addenda shall be applicablPhis follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice '

terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days ,

arly or annual (

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x TABLE 4.3.3.1-1 (Continued) ,

7 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REOUIREMENTS E '

CHANiiEL OPERATIONAL Z CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

- TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED C. DIVISION 3 TRIP SYSTEM

1. HPCS SYSTEM
a. Reactor Vessel Water Level -

Low, Level 2 S Q R(*) 1, 2, 3, 4*, 5*

b. Drywell Pressure-High ## S Q R(') 1, 2, 3
c. Reactor Vessel Water Level -

High, Level 8 S Q R(*) 1, 2, 3, 4*, 5*

d. Condensate Storage Tank Level -

Low S Q R(*) 1, 2, 3, 4*, S*

w e. Suppression Pool Water i Level - High S Q R(*) 1, 2, 3, 4*, 5*

w .f . HPCS Pump Discharge Pressure -

4, High S Q R(*) 1, 2, 3, 4*, 5*

  • g. HPCS System Flow Rate - Low S Q R(*) 1, 2, 3, 4*, 5*
h. Manual Initiation ## NA R NA 1, 2, 3, 4*, 5*

D. LOSS OF POWER

1. 4.16 kV Emergency Bus Under- NA NA 1, 2, 3, 4**, 5**

voltage (Loss of Voltage) l

2. 4.16 kv Emergency Bus Under- S M R(dc)s I, 2, 3, 4**, 5**

voltage (Degraded Voltage)

  1. Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 100 psig.

g- ## The injection function of Drywell Pressure - High and Manual Initiation are not required to be m>m g OPERABLE with indicated reactor vessel water level on the wide range instrument greater than the #"?

Level 8 setpoint coincident with reactor pressure less than 450 psig. *%@

g

  • When the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3. "iit
    • Required when ESF equipment is required to be OPERABLE. P,,

O (a) Calibrate lu,"@g Oc - Il Veheh odg;c. j/

h(cqtt Sl(ST eat FuNCTioh>4LTEST

- me) be eiderded lo II e ccykkee e 1 -

. PY-CEI/NRR-1954L

.- Attcchment 2 l Page 3 of 5 l INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION i

LIMITING CONDITION FOR OPERATION i

3.3.7.5 The accident monitoring instrumer.tation channels shown in Table 3.3.7.5-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3.7.5-1. ,

ACTION:  !

With one or more accident monitoring instrumentation channels inoperable, take '

the ACTION required by Table 3.3.7.5-1.

SURVEILLANCE REQUIREMENTS -

r 4.3.7.5 Each of the above required accident monitoring instrumentation channels shall be demonstrated OPERABLE by performan of the CHANNEL CHECK .

and CHANNEL CALIBRATION operations at the frequencie I wn in Table 4.3.7.5-1.

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I 4 CH AWEL t AttiWATIc0 mug l>< cdewJeA as i.d<,A;%dby ad ,ta) % J39, q,3,;,p.g PERRY - UNIT 1 3/4 3-77  !

m E TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMcNTATION SURVEILLANCE REOUIREMENTS 5 APPLICABLE

, Cl CHANNEL OPERATIONAL

- INSTRUMENT Ci m CALIBRATION CONDITIONS

1. Reactor Vessel Pressure M R 1, 2, 3 i i
2. Reactor Vessel Water Level I
a. Fuel Zone M R 1,2,3 l
b. Wide Range M R 1, 2, 3 )
3. Suppression Pool Water level M R 1, 2, 3 '
4. Suppression Pool Water Temperature M R 1, 2, 3
5. Primary Containment Pressure M R 1, 2, 3
6. Primary Containment Air Temperature M R 1, 2, 3
7. Drywell Pressure M R 1, - 2, 3
8. Drywell Air Temperature H R 1, 2, 3
9. Primary Containment and Drywell Hydrogen Concentration Analyzer and Monitor NA Q* 1,2,3 l R 10. Safety / Relief Valve Position Indicators M R 1,2,3
11. Primary Containment /Drywell Area Gross Y Gamma Radiation Monitors M R** 1, 2, 3 8 12. Offgas Ventilation Exhaust Monitor # M R 1, 2, 3  !
13. Turbine Building / Heater Bay Ventilation j Exhaust Monitor # '

M R 1, 2, 3 i

14. Unit 1 Vent Monitor # M R 1,2,3
15. Unit 2 Vent Monitor # M R 1,2,3
16. Neutron Flux

> a. Average Power Range M R 1, 2, 3 2 b. Intermediate Range M R 1, 2, .3 i c. Source Range M R 1,2,3 2 17. Primary Containment Isolation Valve M RC *I 1, 2, 3 A Position l P m>m  ;

8 *Using sample gas containing: 4 ;; 7 l

a. One volume percent hydrogen, balance nitrogen. *#Q
b. Four volume percent hydrogen, balance nitrogen.
  • T t*
    • The CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including R, g 5 the detector, for range decades above 10 R/hr and a one point calibration check of the detector w aP below 10 R/hr with an installed or portable gamma source. "O Y

g#Ll9h-4 guec4a surwe m w e a<a b w m ,,p t<a a s .,3 & $$l, rd,it.,g.

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PY-CEI/NRR-1954L Attachment 2 f Page 5 of 5

CONTAINHENT SYSTEMS u

SE VEILLANCE REOUIREMENTS (Continued)

1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within 0.25 L,. The formula to be used.is:

[L, + L,, - 0.25 L,] s L, s [L, + L,, + 0.25L,] where L, = supplemental test result; L, - superimposed leakage; L., - measured Type A leakage.

2. Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.
3. Requires the quantity of gas injected into the primary containment or bled from the primary containment during the supplemental test to be between 0.75 L, and 1.25 L,.

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d. Type B and C tests shall be conducted with gas at P,* at intervals no greater than 24 months except for tests involving:
1. Air locks,
2. Main steam line isokation valves,
3. Valves pressurized with fluid from a seal system, ,
4. All containment isolation valves in hydrostatically tested lines which penetrate the primary containment, and-
5. Purge supply and exhaust isolation valves with resilient material seals.
e. Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.
f. Main steam line. isolation valves shall be leak tested at least once per 18 months.
g. Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J of 10 CFR 50 Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P, and the seal system capacity is adequate to maintain system l pressure for at least 30 days.
h. All containment isolation valves in hydrostatically tested lines which penetrate th ary containment shall be leak tested at least once per 18 months uired WLMvf**u PERRY - UNil I 3/4 6-5 AMENDMENT NO. 5,70,22,32.M,57

l PY-CEI/NRR-1954L Attachment 3 Page 1 of 2 SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment involves no significant hazards considerations are included in the Commission's Regulations, 10 CFR 50.92, which state that the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any previously evaluated, or (3) involve a significant reduction in a margin of safety.

The proposed amendment has been reviewed with respect to these three factors and it has been determined that the proposed change does not involve a significant hazard because:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed TS change requests one-time only extensions of the surveillance intervals related to: a) ASME Section XI valve leak rate, stroke and timing, and position indication testing; b) Accident Monitoring Instrumentation related to valve position indication testing; c) Division 1, 2, and 3 Degraded Voltage and Undervoltage instrumentation LSFT; and, d) leak rate testing for hydrostatically tested containment isolation valves.

Based on the discussion in the License Amendment Request which shows:

i) The extension of the interval for ASME Section XI stroke and timing, leak rate measurement and position indication testing requirements are acceptable based on results of past testing which indicates a margin to TS limits will be maintained;

11) The extension of the interval for Position Indication Calibration as specified in Table 4.3.7.5-1, Item 17 is acceptable based on the testing results from the past two refueling outages that indicate no failures have occurred; iii) LSFT interval extension for the Division 1, 2, and 3 Degraded Voltage and Undervoltage instrumentation is acceptable based on the NRC Safety Evaluation Report (Peach Bottom Atomic Power Plant, Units 2 and 3, dated August 2, 1993) which supported extension of the interval for l LSFT from 18 to 24 months. This was based on the small probability of relay or contact failure relative to mechanical component failure probability and, therefore, the increase in LSFT interval represented no significant change in the overall safety system unavailability; and, iv) The extension of the interval for hydrostatic leak testing of containment isolation valves is acceptable based on the consistently low l past leak rate data which is a small percentage of the TS limits. l Therefore, from the above it is shown that the proposed changes vill not significantly increase the probability of an accident previously )

evaluated.

E PY-CEI/NRR-1954L ,

.. Attcchment 3 Page 2 of 2' i

t

2. The proposed change vould not create the possibility of a new or  ;

different kind of accident from any accident previously_ evaluated. l The proposed TS change requests one-time extensions of the surveillance intervals for ASME Section XI valve testing, instrumentation l

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calibration, instrument channel LSFT, containment isolation valve hydrostatic leak rate testing. The proposed changes do not necessitate 1:

a physical alteration to the plant (no new or different. type of i equipment will be installed). In that the requested extension durations  !

are small as compared to the overall interval allowed by TS, NRC and industry evaluations support extension of LSFT, and past testing results- :

provide confidence of no effect on equipment availability by extending the surveillance interval, the change does not create the possibility of i a new or different kind of accident from any accident previously evaluated.

3. The proposed change vill not involve a significant reduction in the margin of safety.

The proposed TS change requests-a one-time extension of the surveillance intervals for ASME Section XI valve testing, instrumentation calibration, instrument channel LSFT, and containment isolation valve ',

hydrostatic leak rate testing. The proposed changes do not necessitate ,

a physical alteration to the plant (no new or different type of l equipment vill be installed). In that the requested extension durations  ;

are small as compared to the overall interval allowed by TS, NRC and industry evaluations support extension of LSFT, and past testing results provide confidence of no effect on equipment availability by extending the surveillance interval, the change does not involve a significant reduction in the margin of safety.

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PY-CEI/NRR-1954L Enclosure 1 Page 1 of 3 Justification for Surveillance Interval Extension Inservice Inspection and Testing of ASME Code Class 1, 2, and 3 Components Technical Specification 4.0.5 Technical Specification Surveillance Requirement 4.0.5 requires inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves to be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR50, Section 50.55a(g).

The table below lists the particular valves and corresponding ASME Section XI tests requiring extension and identifies when the equipment is required to be OPERABLE. The table headings are PI for Position Indication Test and LD for Leak Rate Test.

Component PI Test LD Test OP CON Frequency Due Frequency Due 1D17-F079A 2Y 3-08-96 1,2,3 1D17-F079B 2Y 3-08-96 1,2,3 1D23-F010A 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F010B 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F020A 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F020B 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F030A 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F030B 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F040A 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F040B 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1D23-F050 2Y 2-11-96 2Y 2-11-96 1,2,3,4,5 1E21-F005 2Y 2-18-96 1,2,3,*

1G43-F050A 2Y 2-10-96 2Y 2-10-96 1,2,3 1G43-F050B 2Y 2-10-96 2Y 2-10-96 1,2,3 1G43-F060 2Y 2-10-96 2Y 2-10-96 1,2,3 1G50-F272 2Y 3-23-96 1,2,3,*

1G50-F277 2Y 3-23-96 1,2,3,*

IM17-F055 2Y 4-08-96 2Y 4-08-96 **

1M17-F065 2Y 4-08-96 2Y 4-08-96 **

1P52-F200 2Y 5-16-96 1,2,3,*

1P52-F646 2Y 5-16-96 1,2,3,*

1P57-F015A 2Y 2-21-96 1,2,3,*

1P57-F015B 2Y 2-21-96 1,2,3,*

1P57-F020A 2Y 2-21-96 1,2,3,*

IP57-F020B 2Y 2-21-96 l',2,3,*

  • When handling irradiated fuel in the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

PY-CEI/NRR-1954L

.- Enclosure 1 Page 2 of 3 In general, the request for an extension of the surveillance interval is necessary for two reasons: either the valve cannot be tested during plant f operation, or the location of the valve is inaccessible during plant operation. I Performance of the test vould require a plant shutdown, in the first case, or l significant personnel exposure, in the second case.

In order to cover the most limiting case, a nominal 111 day extension to the surveillance interval is requested to permit the ASME Section XI tests to be next performed during the respective system outages in RFO-5. Without the requested extension, the earliest performance of these surveillances is 111 days before the most conservative projected end of RFO-5. The outage management schedule for RFO-5 represents the best effort to avoid multiple system ,

outages / manipulations and considers shutdown risk in addition to Technical i Specification requirements. i Position Indication (PI) Tests constitute the majority of the tests being extended. The justification for the extension of the surveillance interval of Position Indication Tests is based upon the prior success rates of this type of test. This type of test confirms that the valve positions indicated in the 4 control room agree with actual valve positions as determined by local valve position indication or by the valve's ability to control pressure in the >

associated test volume. RFO-3 and RFO-4 test results for the valve positions repeatedly indicated excellent success rates with zero failures. There is no reason to believe a nominal 111 day extension of the surveillance interval vould i adversely affect this trend.

The justification for the extension of the surveillance interval of the Leak  !

Rate (LD) tests is based on the prior success rates of the particular valves.  ;

These ASME Section XI leak rate tests are applicable to the following valves:

1D23-F010A, Suppression Pool Level A Dry Leg Isolation Valve 1D23-F010B, Suppression Pool Level B Dry Leg Isolation Valve  ;

1D23-F020A, Containment Pressure A Containment Isolation Valve i 1D23-F020B, Containment Pressure B Containment Isolation Valve 1D23-F030A, Containment Dryvell A Differential Pressure-Containment Leg ID23-F030B, Containment Dryvell B Differential Pressure-Containment Leg l 1D23-F040A, Containment Dryvell A Differential Pressure-Dryvell Leg 1D23-F040B, Containment Dryvell B Pifferential Pressure-Dryvell Leg 1D23-F050, Suppression Pool Level r Dry Leg Isolation Valve i 1G43-F050A, Suppression Pool Level A Vet Leg Isolation Valve  :

1G43-F050B, Suppression Pool Level B Vet Leg Isolation Valve

  • 1G43-F060, Suppression Pool Level C Vet Leg Isolation Valve 1H17-F055, Containment Vacuum Breaker A Containment Isolation Valve 1H17-F065, Containment Vacuum Breaker B Containment Isolation Valve i

These valves were Leak Rate tested in both RFO-3 and RFO-4. The results of that j testing is presented below in units of SCCM. The pair of values for ID23-F050 i represents the as-found and as-left values.

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_ , _ . - __ ______O

i PY-CEI/NRR-1954L ]

L. Enclosura 1 i Fage 3 of 3 I Valve Limit RFO-3 RFO-4  !

1D23-F010A 1000 10 2 1D23-F010B 1000 2 2 1D23-F020A 1000 10 6.41 1D23-F020B 1000 10 3.32  !

1000 10 2 l 1D23-F030A 1D23-F030B 1000 10 2 1D23-F040A. 1000 10' 45.61 .

1D23-F040B 1000 10 2 1 1D23-F050 1000 2 1339/2 ,

1G43-F050A 1890 0 0 i 1G43-F050B 1890 0 0 1 1G43-F060 1000 2 0 ]

1M17-F055 1000 2 2 H 1M17-F065 1000 2 , 2

'Although the as-found value for 1D23-F050 exceeded the acceptance criterion, ,

this data point is considered not representative of the general leak rate  !

expectations. Based on data from two cycles of operation, the leak rate for any i given valve is expected to be on the order of tenths of one percent of the -j corresponding limit. The post-naintenance as-lef t value for 1D23-F050 was at t the lover limit of the monitoring equipment. One other leak rate (1D23-F040A) ,

showed an increase between RFO-3 and RFO-4. However, a linearly extrapolated

  • leak rate for 1D23-F040A projected to the end of the requested surveillance i interval extension vould still be only a very small fraction of the limiting  !

value. Based on this information, significant degradation is not expected  ;

during the extension of the surveillance interval for a nominal period of 111 days until the end of RFO-5.

Four.other LD tests (not addressed by the above justifications) for valves 1E21-F005, 1E21-F006, 1E51-F065 and 1E51-F066 require extension. Extension of  ;

the surveillance interval for Technical Specification Surveillance Requirement

  • 4.0.5 is justified on the same basis as that provided for extending Surveillance Requirement 4.4.3.2.2 (addressed in Enclosure 85 to letter FY-CEI/NRR-1890L, i dated March 24, 1995). Although the justification discussed extensions until the start of RF05, the leak rate testing results for these valves also justify extension of the TS 4.0.5 requirements through the end of the outage.  ;

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PY-CEI/NRR-1954L  !

Enclosure 2-Page 1 of 1 Justification for Surveillance Interval Extension ,

Logic System Functional Test of  :

Emergency Core Cooling System i Technical Specification 4.3.3.2  ;

Table 4.3.3.1-1, Items D.1 and D.2 Technical Specification Surveillance Requirement 4.3.3.2 requires LOGIC SYSTEM

  • FUNCTIONAL TESTS (LSFT) and simulated automatic actuation of all channels of the Emergency Core Cooling System at least once per 18 months. The 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) and the 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) functions are included in Table 4.3.3.1-1 as Items D.1 and D.2, respectively. These functions are required to be OPERABLE in .

OPERATIONAL CONDITIONS 1, 2, 3, 4** and 5**, where ** means when the ESF '

equipment is required to be OPERABLE.

Extension of the surveillance interval is requested to permit the LSFTs to be next performed during the respective Divisional outages in RFO-5. These surveillances were last performed during RFO-4. Performance of these surveillances during operations vould result in increased probability of an inadvertent plant trip. Vith the maximum allovable extension of Technical '

Specification 4.0.2, performance of these surveillances would be required 143 days before the most conservative projected end of RFO-5. The outage management schedule for RFO-5 represents the best effort to avoid multiple system outages / manipulations and considers shutdown risk in addition to Technical Specification requirements. l The justification for extension of the surveillance interval is based upon the l NRC Safety Evaluation Report (dated August 2, 1993) related to the extension of  !

the Peach Bottom Atomic Power Station, Unit Numbers 2 and 3, surveillance intervals from 18 to 24 months:

" Industry reliability studies for boiling water reactors (BVRs), prepared by the BVR Ovners Group (NEDC-30936P) show that the overall safety systems' reliabilities are not dominated by the reliabilities of the logic system, but by that of the mechanical components, (e.g., pumps and valves), which are consequently tested on a more frequent basis...Since the probability of a relay or contact failure is small relative to the probability of mechanical component .

failure, increasing the logic system functional test interval represents no i significant change in the overall safety system unavailability."  !

l The evaluation above is applicable to PNPP and the extension of the surveillance ,

interval requested for PNPP is bounded by the interval accepted on the Peach Bottom docket. Therefore, the nominal 143 day extension of the surveillance interval is justified.

J PY-CEI/NRR-1956L Enclosure 3 Page 1 of 1 Justification for Surveillance Interval Extension Channel Calibration of Accident Monitoring Instrumentation Technical Specification 4.3.7.5 Table 4.3.7.5-1, Item 17 Technical Specification Surveillance Requirement 4.3.7.5 requires periodic CHANNEL CHECKS and CHANNEL CALIBRATIONS to demonstrate the OPERABILITY of Accident Monitoring Instrumentation. The Primary Containment Isolation Valve Position is included in Table 4.3.7.5-1 as Item 17. These functions are required to be OPERABLE in OPERATIONAL CONDITIONS 1, 2 and 3.

A nominal 47 day extension of the surveillance interval is requested to permit the CHANNEL CALIBRATIONS to be next performed during the respective system outages in RFO-5. These surveillances were last performed during RFO-4.

Vithout the requested extension, the earliest performance of these surveillances, including the 4.5 month maximum allovable extension of Technical Specification 4.0.2, is 47 days before the most conservative projected start of RFO-5. Since the valves cannot be exercised or accessed in OPERATIONAL CONDITIONS 1, 2, or 3, performance of these surveillances vould necessitate a reactor shutdown.

The justification for extension of the surveillance interval is based upon the prior success rate of CHANNEL CALIBRATIONS for the primary containment isolation valve positions of concern. These surveillance requirements are satisfied by the performance of Position Indication Tests in accordance with ASME Section XI tests. These tests confirm that the valve positions indicated in the control room agree with actual valve positions as determined by local valve position indication or by the valve's ability to control pressure in the associated test volume. RFO-3 and RFO-4 test results for the valve positions of concern repeatedly indicated excellent success rates with zero failures. There is no reason to believe an increase of 47 days in the surveillance interval vould adversely affect this trend.

Therefore, the nominal 47 day extension of the surveillance interval is justified.

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PY-CEI/NRR-1954L Enclosure 4 Page 1 of 1 Justification For Surveillance Interval Extension  :

Technical Specification 4.6.1.2.h Valve Leak Testing Uydrostatically Tested Containment Isolation Valve Technical Specification (TS) Surveillance Requirement (SR) 4.6.1.2.h requires that containment isolation valves in hydrostatically tested lines which penetrate primary containment be leak tested at least once per 18 months (with a 4.5 month extension of the interval per Technical Specification 4.0.2). In addition, TS SR 4.6.1.2.h specifies that the measured combined leakage of containment isolation valves in hydrostatically tested lines be limited to less than or equal to 1 gpm times the total number of valves in this classification.

The hydrostatically tested containment isolation valve in Penetration 106 (1E51-F068) is located in the Reactor Core Isolation Cooling Turbine Exhaust line which is not accessible for testing in Operational Conditions 1, 2 or 3.

The surveillance interval, including the allovable extension of Technical Specification 4.0.2, for this valve vill expire 11 days prior to the most conservative projected start date for RFO-5.

During RFO-3 and RFO-4, valve 1E51-F068, a 12-inch motor operated gate valve, was hydrostatically leak tested and indicated leakage of 0.0 gpm. This leakage is well below the 1 gpm acceptance criteria as stated in TS SR 4.6.1.2.h. Based ,

on past leak rate test data and its repeatability, extension of the surveillance  !

interval for this valve for a nominal period of 11 days is justified. l l

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