ML20094B664

Proposed Tech Specs Extending Containment Sys Surveillance Test Intervals to Accommodate 24 Month Operating Cycles

ML20094B664
Person / Time
Site:	FitzPatrick
Issue date:	10/25/1995
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20094B623	List: JPN-95-046, Forwards Application for Amend to License DPR-59,revising Ts,Re Extension of Containment Sys Surveillance Test Intervals to Accommodate 24 Month Operating Cycles ML20094B647 ML20094B664 ML20094B678
References
NUDOCS 9511010184
Download: ML20094B664 (34)

Text

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l JAFNPP .

3.7 : LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Anolicability: Anolicability:

Applies to the operating status of the primary and secondary Applies to the primary and secondary containment integrity.

containment systems. 5 Obiective: Obiective:

To assure the integrity of the primary and secondary containment To verify the integrity of the primary and secondary containment

, cystems. systems.

4

- Specification: Specification:

A. Primary Containment A. Primary Containment

. 1. . The level from the bottom of the torus and temperature 1. The torus water level and temperature shall be monitored .l n.ne of the water in the torus shall be maintained within the as specified in Table 4.2-8.

@$ following limits whenever the reactor is critical or g whenever the reactor coolant temperature is greater than The accessible interior surfaces of the drywell and above D- 212*F and irradiated fuel is in the reactor vessel: the water line of the torus shall be inspected once per 24 8 months for evidence of deterioration.

@ a. Maximum level of 14.00 feet.

, . o, Whenever there is indication of relief valve operation or i gg b. Minimum level of 13.88 feet. testing which adds heat to the suppression pool, the pool oo temperature shall be continuously recorded until the heat l200

The torus water level may be outside the above addition is terminated. The operator will verify that

~

3 limits for a maximum of four (4) hours as a result of required operability testing of HPCI, RCIC, RHR, CS, average temperature is within applicable limits every 5'-

minutes. In lieu of continuous recording, the operator and the Drywell-Torus Vacuum Relief System, shall log the temperature every 5 minutes.

i c. Maximum water temperature Whenever there is indication of relief valve operation with -

the temperature of the suppression pool reaching 160*F (1) During normal power operation maximum or more and the primary coolant system pressure greater water temperature shall be 95 F. than 200 psig, an external visual examination of the torus l shall be conducted before resuming power operation.

i Amendment No. 1 S, 3S, iS,1 S0,181,190,197 165

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JAFNPP ,

' 3.7 - (cont'd) : 4.7 (cont'd) {

breaker is sooner made operable, provided that the repair procedure does not violate primary containment integrity.

5. Pressure Suppression Chamber - Drywell Vacuum 5. Pressure Suppression Chamber - Drywell Vacuum Breakers Breakers

- a. When primary containment integrity is required, all a. Each drywell suppression chamber vacuum breaker drywell suppression chamber vacuum breakers shall shall be exercised through an opening - closing cycle be operable and positioned in the fully closed monthly.

4 position except during testing and as specified in 3.7.A.5.b below.

b. One drywell suppression chamber vacuum breaker b When it is determined that one vacuum breaker is may be non-fully closed so long as it is determined inoperaiale for fully closing when operability is to be not more than 1' open as indicated by the required, the operable breakers shall be exercised position lights. immediately, and every 15 days thereafter until the i inoperable valve has been retumed to normal service.

c. One drywell suppression chamber vacuum breaker may be determined to be inoperable for opening.

c. Once per 24 months, each' vacuum breaker valve shall be visually inspected to insure proper l'}

maintenance and operation.

d. Deleted d. A leak test of the drywell to suppression chamber structure shall be conducted once per 24 months; the acceptable leak rate is sO.25 in. water / min, over a 10 min period, with the drywell at 1 psid.

Amendment No. 131,192 178

O JAFNPP .

3.7 (cont'd) 4.7 (cont'd)

e. Leakage between the drywell and suppression . e. Not applicable -

chamber shall not exceed a rate of 71 scfm as monitored via the suppression chamber 10 min pressure transient of 0.25 in. water / min. -

f. The self actuated vacuum breakers shall open f. Not applicable when subjected to a force equivalent to 0.5 psid acting on the valve disc.

g. From and after the date that one of the pressure g. Once per 24 months, each vacuum breaker shall be' suppression chamber /drywell vacuum breakers is tested to determine that the force required to open made or found to be inoperable for any reason, the the vacuum breaker does rmt exceed the force vacuum breaker shall be locked closed and reactor specified in Specification 3.7.A.5.f and each vacuum operation is permissible only during the succeeding breaker shall be inspected and verif'e d to meet seven days unless such vacuum breaker is sooner design requirements.

made operable, provided that the repair procedure  ;

does not violate primary containment integrity. ,

Amendment No. 3,

• 31 4

179 i ,

JAFNPP .: ,

!' 3.7 (cont'd) ~ 4.7 (cont'd) l l

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B. Standbv Gas Treatment System B. Standby Gas Treatment System l

' 1. Except as specified in 3.7.B.2 below both circuits of the 1. Standby Gas Treatment System surveillance shall be Standby Gas Treatment System shall be operable at all performed as indicated below:

times when secondary containment integrity is required.

a. Once per 24 months, it shall be demonstrated that:

(1) Pressure drop across the combined high-efficiency and charcoal filters is less than 5.7 in. of water at 6,000 scfm, and (2) Each 39kW heater shall dissipate greater than 29kW of electric power as calculated by the following expression:

P =}5EI where:

P= Dissipated Electrical Power; E = Measured line-to-line voltage in volts (RMS);

i = Average measured phase current in amperes (RMS).

t Amendment No. 10, 20, 3S, SS, SS, 81, 03,101,-324-181

JAFNPP

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4.7 (cont'd)-

, b. At least once during each scheduled secondary l containment leak rate test, whenever a filter is changed, whenever work is performed that could

. affect the filter system efficiency, and at intervals l not to exceed six months between refueling outages, it shall be demonstrated that:

I (1.) The removal efficiency of the . particulate filters -

~ is not less than 99 percent based on a DOP

- test per ANSI N101.1-1972 para. 4.1.

(2.) The removal efficiency of each of the charcoal  ;

filters is not less . than 99 percent based on a s Freon test.-

c. At least once each yr, removable charcoal cartridges .

shall be removed and absorption capability shall be '

demonstrated.  ;

d. Once per 24 months, automatic initiation of each branch of the Standby Gas Treatment System shall be demonstrated.

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i Amendment No. 44 182

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JAFNPP e r

3.7 (cont'd) - 4.7 (cont'd)

e. Once per 24 months, manual operability of the '

bypass valve for filter cooling shall be demonstrated.

i. f. Standby Gas Treatment System Instrumentation. '

Calibration-i differential Once/ operating ,

pressure Cycle switches

2. From and after the date that one circuit of the Standby 2. When one circuit of the Standby Gas Treatment Gas Treatment System is made or found to be inoperable System becomes inoperable, the operable circuit - .i for any reason, the following would apply: shall be verified to be operable immediately and daily j thereafter.  ;

a. If in Start-up/ Hot Standby, Run or Hot Shutdown  !

mode, reactor operation or irradiated fuel handling is permissible only during the succeeding 7 days unless such circuit is sooner made operable, t provided that during such 7 days all active components of the other Standby Gas Treatment Circuit shall be operable.

b. If in Refuel or Cold Shutdown mode, reactor operation or irradiated fuel handling is permissible -

only during the succeeding 31 days unless such t circuit is sooner made operable, provided that during such 31 days all active components of the ,

other Standby Gas Treatment Circuit shall be operable.  !

3.- If Specifications 3.7.B.1 and 3.7.B.2 are not met, the 3. Intentionally Blank reactor shall be placed in the cold condition and irradiated fuel handling operations and operations that could reduce the shutdown margin shall be prohibited.

Amendment No. 10, SS,

• iS,154 ,

183 i

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JAFNPP ,

3.7 (cont'd) - 4.7 (cont'd)

c. Secondary containment capability to maintain a 1/4

. in. of water vacuum under calm wind conditions with a filter train flow rate of not more than 6,000 cfm, shall be demonstrated once per 24 months prior to refueling.

D. Primary Containment isolation Valves D. Primary Containment isolation Valves

1. Whenever primary containment integrity is required per 1. The primary containment isolation valves surveillance 3.7.A.2, containment isolation valves and all instrument shall be performed as follows:

line excess flow check valves shall be operable, except as specified in 3.7.D.2. The containment vent and purge a. Once per 24 months, the operable isolation valves valves shall be limited to opening angles less than or that are power operated and automatically initiated  ;

equal to that specified below: shall be tested for simulated automatic initiation and for closure time.

Valve Number Maximum Openino Anale 27AOV-111 40* b. At least once per operating cycle, the instrument 27AOV-112 40 line excess flow check valves shall be tested for 27AOV-113 40 proper operation. l 27AOV-114 50 27AOV-115 50 c. At least once per quarter: .

! 27AOV-116 50 t 27AOV-117 50 (1.) All normally open power-operated isolation 27AOV-118 50 valves (except for the main steam line and i Reactor Building Closed Loop Cooling Water i System (RBCLCWS) power-operated isolation valves shall be fully closed and reopened. l Amendment No. 151,173,195  :

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1. 4.7 BASES (cont'd) L -

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budding isolation valves, leak-tightness of the reactor - The in place testing of charcoal filters is performed using building and performance of the Standby Gas Treatment _ Freon or equivalent, which is injected into the system _. ,

System. Functionally. testing the initiating sensors and _ upstream of the charcoal filters. Measurements of the '

i associated trip channels demonstrates the capability for Freon' concentration upstream and downstream of the'. ,

s  :

automatic ~ actuation. . Performing these tests prior to . charcoal filters is made. The ratio of the inlet and outlet

[ . refueling ~will demonstrate secondary containment. concentrations gives an overall indication of the leak : .

! capability prior to the time the primary containment is Ttightness of the system. Although this is basically a' leak '

. opened for refueling., Periodic testing gives sufficient "

D test, since the filters have charcoal of known efficiency

. t confidence of reactor building integrity and Standby Gas and holdmg capacity for elemental iodme and/or methyl i Treatment System performance capability. iodine, the test also gives an indication of the relative efficiency of the installed system. :  :

The' test frequencies are adequate to detect equipment

- deterioration prior to significant defects, but the tests are High-efficiency particulate filters are installed to minimize -

• not frequent enough to load the filters, thus reducing their potential release of particulates to the environment. An reserve capacity too quirdiy. That the testing frequency is efficiency of 90 percent is adequate to retain particulates [

. adequate to detect deterioration was demonstrated by the ' that may be released to the reactor building followng an -

- tests which showed no loss of filter efficiency after 2 yr.

_ accident. This will be demonstrated by in-place testmg *

._ of operation in the rugged shipboard environment on the- with DOP as testing medium. ,

, NS Savannah (ORNL 3726). Pressure drop tests across  !

filter sections are performed to detect gross plugging or-leak paths through the filter media. Considering the relatively short time that the fans may be run for test purposes, plugging is unlikely, and the test interval of once +

per 24 months is reasonable. Duct heater tests will be  :

conducted once per 24 months. ' Considering the j

{ simplicity of the heating circuit, the test frequency is ,

sufficient.

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' t 195  :

b a

_.m _a-- _ m.___-____--- _ _ . _ _ _ _ _ . _ _ _ _ _ .-_ _ _ _ _ _ . - - _ _ _ _ _ - _ . _ _ _ _ _ _- _--_______m 1 _ r 1 w-

JAFNPP .

4.7 BASES (cont'd)

The test interval for filter efficiency was selected to minimize in order.to assure that the doses that may result from a steam  ;

plugging of the filters. In addition, retention capacity in terms line break do' not exceed the 10CFR100 guidelines, it is of milligrams of iodine per gram of charcoal will be necessary that no fuel rod perforation resulting from the demonstrated. This will be done by testing the charcoal once accident occur prior to closure of the main steam line isolation a year, unless filter efficiency seriously deteriorates. Since valves. Analyses indicate that fuel rod cladding perforations shelf lives greater than 5 yr. have been demonstrated, the test would be avoided for main steam valve closure times, including -  ;

interval is reasonable. instrument delay, as long as 10.5 sec.  ?

i D. Primary Containment isolation Valves For Reactor Coolant System temperatures less than 212'F, the ,

containment could not become pressurized due to a  !

The large pipes comprising a portion of the Reactor Coolant loss-of-coolant accident. The 212*F limit is based on ,

System, whose failure could result in uncovering the reactor preventing pressurization of the reactor building and rupture of core, are supplied with automatic isolation valves (except the blowout panels.  !

those lines needed for Emergency Core Cooling Systems operation or containment cooling). Valve closure times are The primary containment isolation valves are highly reliable,  ;

adequate to prevent loss of more coolant from the have low service requirement, and most are normally closed. ,

circumferential rupture of any of 'hese lines outside the Power operated primary containment isolation valves that can l containment than from a steam line rupture. Therefore, be cycled during normal plant operations are cycled periodically isolation valve closure times are sufficient to prevent per the ASME Section XI Inservice Testing Program. Valves  ;

uncovering the core. that can not be cycled during normal plant operations are tested  ;

once every 24 months. The initiating sensors and associated trip channels are periodically checked to demonstrate proper l' response. This combination of testing adequately ver' is operability of power operated and automatically initiat- vimary ,

containment isolation valves.

i Amendment No. 496-196 ,

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JAFNPP

'4.7) BASES -- (cont'd)

The main steam line isolation valves are functionally tested on -

L a more frequent interval to establish a high degree of reliability.

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'The ' primary containment is penetrated by several small i~ - diameter instrument lines connected to the reactor coolant system. Each instrument line contains a O.25 in restricting orifice inside the primary containment and an excess flow check valve outside the primary containment.

L The RBCLCWS valves are excluded from .the quarterly surveillance requirements because closure of these valves will eliminate the coolant flow to the drywell air and recirculation pump-motor coolers. Without cooling water, the drywell air and equipment temperature will increase and may cause damage to the equipment during normal plant operations.

Therefore, testing of these valves would only be conducted in the cold condition.

l A list of containment isolation valves, including a brief description of each valve is included in Section 7.3 of the updated FSAR.

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!- l l ' Amendment No. 151,173,203 197 a

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ATTACHMENT 11 to JPN 95 046 Safety Evaluation For Proposed Changes to Technical Specification Containment Systems Surveillance Test Intervals to Accommodate 24-Month Operatina Cveles (JPTS-95-001D)

L New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

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Attachment ll to JPN-95-046 Containment Systems SAFETY EVALUATION Page 1 of 11

1. DESCRIPTION OF THE PROPOSED CHANGES 1.- Page 165, Specification 4.7.A.1, change the inspection of the drywell accessible interior surf aces from "each operating cycle" to "once per 24 months." The revised specification reads:

"The accessible interior surfaces of the drywell and above the water line of the torus shall be inspected once per 24 months for evidence of 4

deterioration."

2. Page 178, Specification 4.7.A.5.c, change the visual inspection of the vacuum breaker valves from "each operating cycle" to "per 24 months." The revised specification reads:

"c. Once per 24 months, each vacuum breaker valve shall be visually inspected to insure proper maintenance and operation."

3. Page 178, Specification 4.7.A.5.d, change the surveillance interval of the leak test of the drywell to suppression chamber structure from "once per operating cycle" to "once per 24 months." The revised specification reads:

"d. A leak test of the drywell to suppression chamber structure shall be conducted once per 24 months, the acceptable leak rate is u0.25 in, water / min, over a 10 min period, with the drywell at 1 psid."

4. Page 179, Specification 4.7.A.S.g, change the testing interval of the vacuum breaker valves from "during each refueling outage" to "Once per 24 months." The revised specification reads:

"g. Once per 24 months, each vacuum breaker shall be tested to determine that the force required to open the vacuum breaker does not exceed the force specified in Specification 3.7.A.5.f and each vacuum breaker shall be inspected and verified to meet design requirements."

5. Page 181, Specification 4.7.B.1.a, change the Standby Gas Treatment System surveillance requirements for filter pressure drop and heater power dissipation from "once per operating cycle" to "once per 24 months." The revised specification reads:

"a. Once per 24 months,it shall be demonstrated that:"

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JAttachment ll to 'JPN-95-046 Containment Systems SAFETY EVALUATION Page 2 of 11

6. Page 182, Specification 4.7.B.1.d, change the Standby Gas Treatment System automatic initiation surveillance from "once each operating cycle" to "once per 24 months." - The revised specification reads:

"d. Once per 24 months, automatic initiation of each branch of the Standby Gas Treatment System shall be demonstrated."

7. Page 183, Specification 4.7.B.1.e, change the Standby Gas Treatment System manual operability of the bypass valve for filter cooling surveillance from "once per operating cycle" to "once per 24 months." The revised specification reads:

"e. Once per 24 months, manual operability of the bypass valve for filter cooling shall be demonstrated."

-8. Page 185, Specification 4.7.C.1.c, change this surveillance interval from "at each refueling outage" to "once per 24 months." The revised specification reads:

"c. Secondary containment capability to maintain a 1/4 in. of water vacuum under calm wind conditions with a filter train flow rate of not more than 6,000 cfm, shall be demonstrated once per 24 months prior to refueling."

9. Page 185, Specification 4.7.D.1.a, change the Primary Containment isolation Valves simulated automatic initiation and closure time test from "once per operating cycle" to "once per 24 months." The revised specification reads:

l "a. Once per 24 months, the operable isolation valves that are power operated I and automatically initiated shall be tested for simulated automatic initiation I and for closure time."

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10. Page 185, Specification 4.7.D.1.b, delete footnote indicated by "*" in Specification 4.7.D.1.b since the conditions allowing a test extension have expired.

11. Bases page 195, second paragraph fourth sentence, change "once per operating cycle" to "once per 24 months." The revised bases read ,

"Considering the relatively short time that the fans may be run for test purposes, I' ' plugging is unlikely, and the test interval of once per 24 months is reasonable."

13. -Bases page 195, second paragraph fifth sentence, change "once during each operating cycle" to "once per 24 months." The revised bases reads:

" Duct heater tests will be conducted once per 24 months."

I Attachment 11 to JPN-95-046 )

Containment Systems j SAFETY EVALUATION  !

Page 3 of 11 ]

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14. Bases page 196, first column first line, move the words "by in place testing with DOP as testing medium." to second column last sentence on page 195 to make the bases easier to read. This is an editorial change only and does not change the wording of the Technical Specification Bases.

15. Bases page 196, second column last paragraph, and Bases page 197, first line, delete current wording and replace with the following:

" Power operated primary containment isolation valves that can be cycled during normal plant operations are cycled periodically per the ASME Section XI Inservice Testing Program. Valves that can not be cycled during normal plant operations are tested once every 24 months. The initiating sensors and associated trip channels are periodically checked to demonstrate proper response. This combination of l testing adequately verifies operability of power operated and automatically initiated primary containment isolation valves."

11. flyBPOSE OF THE PROPOSED CHANGES This application for amendment to the James A. FitzPatrick Nuclear Power Plant t Technical Specifications proposes to extend the surveillance test intervals for the Containment Systems to accommodate 24 month operating cycles. These changes will elirninate the need to shut the plant down mid-cycle to conduct these surveillances. Extended surveillance intervals are identified in the proposed 4

Technical Specifications as being performed "once per 24 months." These changes

' follow the guidance provided by Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate 24-Month Fuel Cycle,"

(Reference 1).

Containment system surveillance tests conducted once per operating cycle were evaluated to confirm that the surveillance intervals could be safely extended. The evaluation (Reference 2) included a detailed study of containment system surveillance history. Surveillance test data was analyzed for components affected by the extended operating cycle. The evaluation concluded that the identified containment systems surveillance tests can be safely extended to accommodate a 24 month operating cycle.

Surveillance intervals associated with 10 CFR 50 Appendix J required primary containment leakage rate testing are not being extended at this time. These requirements will be addressed in an upcoming amendment application to adopt the new Appendix J Option B teak rate testing requirements. The Standby Gas Treatment (SBGT) system instrument calibrations required by SR 4.7.B.1.f and the Excess Flow Check Valve testing required by SR 4.7.D 1.b are not addressed in this amendment application. They will be addressed in the upcoming 24 month cycle Instrumentation systems amendment request.

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Attachment ll to JPN-95-046 Containment Systems i

SAFETY EVALUATION Page 4 of 11 Containment Systems Objective and Descriotion The design objective of the Containment systems is to provide the capability, in conjunction with other engineered safeguards, to limit the release of radioactive materials so that off site doses from a postulated design basis accident (DBA) are below the guideline values of 10 CFR 100.

The containment systems at the FitzPatrick Plant consist of the Primary and Secondary Containment and their support systems. The Primary Containment system consists of a drywell, pressure suppression chamber (torus) which stores a large volume of water (suppression pool), the connecting vent system between the drywell and suppression pool, isolation valves, vacuum relief system , and provisions for containment cooling. The Secondary Containment totally encloses the Primary Containment, the refueling and reactor servicing areas, the new and spent fuel storage facilities, and other reactor auxiliary systems. The Secondary Containment serves as the containment during reactor refueling and maintenance operations when the primary containment is open, and as an additional barrier when the primary containment is functional. The Secondary Containment consists of the Reactor Building, the Standby Gas Treatment (SGT) system, the main stack and the Reactor Building isolation and Control System.

The SGT system provides the secondary containment with a negative pressure to control reactor building leakage, and provides filtration for removal of particulates and iodines prior to release through the main stack. The system consists of two redundant full capacity air filtration trains. Each train consists of a demister, profilter, an electric heater, a high efficiency particulate absolute (HEPA) filter, an activated charcoal adsorber and a HEPA after filter. With the reactor building isolated, each fan has the capacity necessary to reduce and maintain the reactor building at a minimum sub-atmospheric pressure of 0.25 inches of water.

111. SAFETY IMPLICATION OF THE PROPOSED CHANGES Once per operating cycle surveillance requirements (SRs) for the Containment systems are provided in Technical Specifications 4.7 A.1,4.7 A.2,4.7.A.5, 4.7.B.1,4.7.C.1 and 4.7.D.1. The purpose of these tests and the potential safety implications of the extended surveillance interval are discussed below.

Primarv Containment Interior !nsoection (SR 4.7.A.1)

This surveillance requires a once per operating cycle inspection of the primary containment surfaces for signs of deterioration. Specifically, the accessible interior surfaces of the drywell and above the water line of the torus are inspected for evidence of deterioration such as scaling, rusting, and/or paint chipping which could affect the structuralintegrity of the primary containment. During plant operation all surfaces required to be inspected are inaccessible. Therefore, a plant shutdown is required to perform this surveillance.

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Attachment ll to JPN-95 046 Containment Systems SAFETY EVALUATION Page 5 of 11 The surveillance interval of once per operating cycle is based on the accessibility to -

the containment interior, not on a specific time based requirement related to - i l

4 expected degradation rates. .The surfaces subject to inspection are coated to minimize corrosion with a painting system that has been demonstrated to be acceptable for use in reactor containments. These surfaces are in an inerted i environment, which helps to reduce the corrosion rate. In the event that excessive .

. corrosion is found, this condition would be evaluated for acceptability for the next 4

operating cycle or the condition corrected.

1 Conclusion SR 4.7.A.1 Based on the proven life of the cor.tainment coating system and the normally.

-inerted environment these surfaces are exposed to during normal plant operations, this surveillance interval can be safely extended to accommodate a 24 month operating cycle. Historical surveillance test data from the past ten years supports 4

~t his conclusion.

Suporession Chamber to Drvwell Vacuum Breaker Tests (SR 4.7.A.5.c and 4.7.A.5.a)

SR 4.7.A.5.c requires a once per operating cycle visual inspection of the suppression chamber to drywell vacuum breaker valves to insure proper I

maintenance and operation. SR 4.7 A.5.g requires that once each refueling outage, each vacuum breaker be force-tested and inspected and verified to meet design requirements.

L The Primary Containment is equipped with five drywell to torus vacuum breaker

! valves which prevent primary containment pressure from dropping below its  ;

vacuum rating relative to the external design pressure. The vacuum breaker valves  ;

draw noncondensables from the torus to prevent the drywell vacuum rating from )

being exceeded. These valves are 30 inch diameter swing check valves equipped i with counterweighted arms which open when torus pressure exceeds drywell pressure by 0.5psid They are seated with increased pressure on the drywell side, j i'

The vacuum breaker valves are exercised through an opening - closing cycle monthly. Each valve is operated manually using the counterweight lever and

srnooth valve operation is verified. The remote full open and full closed indications e are verified during this test. The vacuum breakers are tested periodically in i accordance with the requirements of the American Society of Mechanical Engineers (ASME)Section XI Inservice Testing (IST) Program. This test verifies operation and setpoint of the vacuum breakers by opening the valves with a calibrated torque i

wrench. It also verifies maintenance and operation of the valves by a visual inspection. This combination of on line testing is sufficient to detect operability

, . problems with the valves.

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Attachment ll to JPN-95-046 Containment Systems 1

, SAFETY EVALUATION  !

Pag'e 6 of 11 ')

Conclusion SRs 4.7 A.5.c and 4.7.A.5.g -

Based on existing on-line functional testing, this surveillance interval can be safely extended to accommodate a longer operating cycle. Historical surveillance test data j supports this conclusion. l i

' Drvwell to Torus Structure Leak Test (SR 4.7 A.5.d)  ;

This SR requires a once per operating cycle leak test of the drywell to torus structure. This testing verifies that bypass leakage, through the suppression chamber to drywell vacuum breakers, is limited to 71 standard cubic feet per minute (SCFM). This provides assurance that steam released to the drywell during ,

a postulated LOCA will flow to, and be condensed by, the water in the suppression l pool. The leakage limit of 71 scfm of nitrogen is approximately one-tenth the maximum allowable bypass capacity.

A review of test results from 1987 to 1994 revealed one test failure which was caused by a valve required to establish test conditions. The test was completed l

successfully after the valve was repaired and drywell bypass leakage was within the acceptance criteria.

In addition, the pressure in the drywell is kept at least 1.7 psi greater than the ,

I pressure in the torus to ensure that appropriate torus and support system safety margins are maintained following a postulated DBA. This differential pressure is monitored once each shift. An abrupt differential pressure drop would alert operators of possible bypass leakage. This condition would be investigated and appropriate corrective actions taken.

Conclusion SR 4.7.A.5.d The past leak test results and on-line monitoring of containment differential pressure ensures with a high degree of confidence that extension of the test interval to support a 24 month cycle will not adversely affect drywell bypass leakage.

SGT Combined Filter Differential Pressure and Heater Power Tests (SR 4.7.B.1.a)  !

SR 4.7.B.1.a specifies once per cycle surveillance testing to be performed on the SGT System. These surveillances determine combined filter differential pressure and measure heater power.

SR 4.7.B.1.a.(1) requires a demonstration that the pressure drop across the combined HEPA and charcoal filters is less than 5.7 inches of water at 6,000 scfm.

This test is performed to detect gross plugging of the filter media.

- The system is normally in a standby condition, therefore gross plugging or fouling

-of the HEPA filters and charcoal adsorbers is minimized. Individual filter differential pressures are monitored during periods of system operation. In addition, alarms are provided in the control room to alert operators of high filter differential pressures during system operation, in the event of abnormal differential pressures, the cause

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Attachment ll to JPN-95-046 l Containment Systems ,

SAFETY. EVALUATION Page 7 of 11 would be investigated and deficiency corrected. The SGT system has dual filter trains ensuring system availability in the event of failure of one filter train.

. SR 4.7.B.1.a.(2) requires a demonstration that each SGT system heater dissipates

! greater than 29kW of electric power. Each SGT system train is equipped with an  ;

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electric heater which is energized during system operation. These heaters maintain F relative humidity at the charcoal filters below 70% in order to assure the efficient

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<j removal of iodine in the charcoal filters.

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The SGT system is normally in a standby condition with the heater elements de-energized. This decreases the possibility of heating coll damage or failure due to a foreign material impingement and minimizes wear on the heating ele:'ents and control circuits. The operation of the heater is verified during system operation and

any abnormal indications would be observed and the cause corrected. The 29 kW ]

heating requirement is easily met by the larger heating capacity of the 39 kW l

- heaters. .This large capacity of the heaters compared to the heating requirement

provides enough margin so that minor degradation can be accommodated.

Conclusion SR 4.7.B.1.a i Based on the redundant design of the SGT system and monitoring of individual filter differential pressure and electric heater operation during periods of system i operation,'these surveillance intervals can be safely extended to support a 24 month operating cycle. Historical surveillance test data from the past ten years supports this conclusion.

. SGT Svstem Simulated Automatic Initiation (SR 4.7.B.1.d)

This SR requires a once per operating cycle demonstration of the automatic initiation of each train of the SGT system. This ensures, in conjunction with other system tests, that the SGT system is capable of performing its design safety

- function. System instrumentation is periodically tested on-line. Major system j components are tested on-line in accordance with the ASME Section XI IST

. Program requirements. During this testing, system motor operated valves are

! cycled and the fans are also started. The SGT system has redundant filter trains and is normally in the standby condition.

Conclusion SR 4.7.B.1.d

.This surveillance can be safely extended since system instrumentat ion and mechanical components are tested periodically on-line. Historical surveillance test

. data supports this conclusion.

1 1,

t l

i i,

. , + - , -

l- .

Attaunment ll to JPN-95-046

, Containment Systems SAFETY EVALUATION Page 8 of 11 SGT Manual Suction Cross tie Bvoass Valves for Filter Coolino Ooerability (SR , 4.7.B.1.e)

This SR requires a once per operating cycle demonstration of the operability of the manual suction cross-tie bypass valves for filter cooling. The SGT filter trains are cross-tied at the fan suctions to allow cool air to be drawn over th,e charcoal filters '

of the inactive filter train. The air flow provides cooling of the charcoal filter from a fire safety standpoint and also prevents iodine desorption from charcoal filters as the result of elevated temperatures. There is little possibility of valve failure in a closed position (that could cause a no-flow condition in the inactive train) since the valves are kept open during power operation. The valves are only operated during the once per cycle surveillance testing and when isolating a train for maintenance.

- Upon restoration of a SGT train following maintenance, the valves would be verified open during the valve lineup. Since the valves are infrequently operated, they are unlikely to wear out due to fatigue.

Conclusion SR 4.7.B.1.e This surveillance interval can be safely extended to support a 24 month operating cycle because these manually operated volves are normally open during plant operation, are infrequently operated, and are verified open upon restoration of a SGT train per the valve lineup. Historical surveillance test data from the past ten years supports this conclusion.

Secondarv Containment Caoability Testina (SR 4.7 C.1.c)

This SR requires that the secondary containment capability to maintain a 0.25 in of water vacuum under calm wind conditions with a SGT filter train flow rate of

.g;. 6,000 cfm, be demonstrated at each refueling cutage prior to refueling. This testing demonstrates the proper operation of the reector building isolation valves, leak tightness of the reactor building and performance of the SBGT system.

During normal operation, reactor building differential pressure is controlled by the Reactor Building isolation and Control System. This system maintains differential pressure at greater than negative 0.25 inches of water. Reactor building differential pressure and system flow rate is monitored once per shift in the control room. The likelihood of leakage during power operation is minimal due to the passive design of the reactor building Performance of work that could affect secondary containment penetrations and components is administratively controlled. A review of past surveillance test results (1988 through 1993) indicated that the SGT has been able to maintain the reactor building differential pressure within the test acceptance criteria.

l 1

i

0 Attachment ll to JPN-95-046 Containment Systems SAFETY EVALUATION Page 9 of 11 Conclusion 4.7.C.1.c This SR interval can be safely extended to support a 24 month operating cycle because there is a low likelihood of reactor building leakage during power operation due to its passive design and there is sufficient administrative controls of work on secondary containment penetrations. Historical surveillance test data from the past ten years supports this conclusion.

Power Ooerated Primarv Containment isolation Valve Simulated Automatic initiation and Closure Time Testina (SR 4.7.D.1.a)

This SR requires a simulated automatic initiation and closure time test for each operable power operated primary concainment isolation valve once per operating cycle. The test verifies the ability of the system to perform its design automatic function by confirming proper operation of electrical and mechanical components.

Closing times for these valves are verified to ensure that they will close fast enough to restrict the release of radioactive material to the environs well below the guidelines of 10 CFR 100.

The Primary Containment and Reactor Vessel Isolation Control System (PCRVICS) provides timely protection against the consequences of accidents involving the release of radioactive materials from the fuel or the reactor coolant system. The system initiates automatic isolation of appropriate process lines which penetrate the primary containment whenever monitored variables exceed pre-selected limits.

Isolation is accomplished by primary containment isolation valves, which are highly reliable and have low service requirements. The majority of these primary containment isolation valves are normally closed during plant operation. The PCRVICS is designed with a high probability that when any essential monitored variable exceeds the isolation setpoint, the event results in automatic isolation. The system is designed such that no single failure within the PCRVICS prevents an isolation action when required.

Power operated primary containment isolation valves which can be cycled during normal plant operations are cycled, and stroke times measured periodically on-line in accordance with the ASME Section XI IST Program. Primary Containment isolation l initiating and actuation logic are periodically tested to verify proper response. This l combination of on-line testing adequately verifies operability of initiating logic and mechanical components.

Conclusion 4.7.D.1.a Based on existing on-line testing, high reliability of these valves, and the redundant design of the PCRVICS the simulated automatic initiation and stroke time testing intervals can be safely extended to support a 24 month operating cycle. Historical surveillance test data supports this conclusion.

Attachment il to JPN-95-046 Containment Systems SAFETY EVALUATION Page 10 of 11 Technical Soecification Bases Chanaes The proposed Bases changes on Technical Specification page 195 change wording relating to "once per operating cycle" to "once per 24 months. The bases changes clarify the new surveillance intervals and do not propose new or different system design limits. As such, there are no safety implications in these proposed bases changes.

The proposed bases change on page 196 deletes the numerical value of the failure probability that a line will not isolate. The failure probability of 1.1 X 10-7 that a line will not isolate was contained in the original FitzPatrick Plant Technical Specification Bases and was representative of a 12 month operating cycle. The methodology used to derive this value was contained in the Technical Specification Section 4.2 Bases and Figure 4.2-1, " Test interval Vs. Probability of System Unavailability," located on page 87. This figure ard section were deleted by Amendment 227 to the FitzPatrick Technical Specifications. This statement was overlooked during the development of amendment 227, and should have been deleted.

IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not: l

1. involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not involve any physical changes to the plant, do not alter the way the containment systems function, and will not degrade the i performance of the containment systems. The type of testing and the corrective i actions required if the subject surveillances fail remains the same. The proposed  !

changes do not adversely affect the availability of the containment systems or l affect the ability of the systems to meet their design objectives. A historical review l of surveillance test results indicated that there was no evidence of any failures I which would invalidate the above conclusions. )

2. create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not modify the design or operation of the plant and therefere no new failure modes are introduced. No changes are proposed to the type and method of testing performed, only to the length of the surveillance interval. Past equipment performance and on-line testing indicate that longer test intervals will not degrade the containment systems. A historical review of surveillance test results indicated that there was no evidence of any failures which would invalidate the above conclusions.

Attachment il to JPN 95-046 Containment Systems SAFETY EVALUATION Page 11 of 11

3. involve a significant reduction in a margin of safety.

Although the proposed changos will result in an increase in the interval between surveillance tests, the impact on system reliability is minimal. This is based on more frequent on-line testing and the redundant design of the containment systems.

A review of past surveillance history has shown no evidence of failures which would significantly impact the reliability of the containment systems. Operation of the plant remains unchanged by the proposed containment system surveillance test interval extensions. The assumptions in the Plant Licensing Basis are not impacted.

Therefore the proposed changes do not result in a significant reduction in the margin of safety.

V. IMPLElENTATION OF THE PROPOSED CHANGE implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes affect the environment.

VI. CONCLUSlQN Based on the discussion above, the identified containment surveillance tests can be safely extended to accommodate a 24 month operating cycle. The assumptions in the FitzPatrick licensing basis are not invalidated by performing the containment surveillances at the bounding interval limits (30 months) to accommodate the 24 month operating cycle.

The Plant Operating Review Committee (PORC) and the Safety Review Committee (SRC) have reviewed these proposed changes to the Technical Specifications and have concluded that they do not involve an unreviewed safety question, or a significant hazards consideration, and will not endanger the health and safety of the public.

Vll. REFERENCE _4

1. Generic Letter 91-04, " Changes in Technical Specification Surveillance Intervals to Accommodate 24-Month Fuel Cycle."

2. NYPA document JAF-RPT-MULTI-01116, " Containment Systems Surveillance Test Extensions," dated May 27,1994.

3. NYPA memorandum, S. Haskell to F. Lake (JSED-95-0395) dated October 13,1995 regarding Supplementary Data for Surveillance Test Extension Evaluation.

ATTACHMENT lli to JPN-95 046 Markup of the current Technical Specification pages Extension of Containment Systems Surveillance Test intervals to Accommodate 24-Month Operatina Cycles (JPTS 95-001D)

- New York Power Authority JAMES A FITZPATRICK NUCLEAR POWER PLANT Docket No. 50 333 DPR-59

,, , -- -,-e- - . ., --, - - -

____.______.._._____._._.___..___._.-_.___.__..m._ _ -.

i JAFNPP 3.7 LIMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REfMMREMENTS -

3.7 ' CONTAlfGENT SYSTEMS 4.7 CONTAINMENT SYSTFRE -

Annlicability:

Annlicability:

Applies to the operatine status of to primary and secondary containment systems.

~

Applies to the primary and secondary cwitainment integrity.

i.

Objective

Objective:

l To aneure the integrity of the primary and secondary containment systems. To verify the integnty of the primary and secondary containment systems.

Sancification:

. R Soecification:

!  ! A. Primary Containment  !!

A. Primary Containment

'hl'

1. The level from the bottom of the torus and temperature 1. The torus water level and temperature shall be monitored f j

4 of the water in the torus shen be maintained within the as specified in Table 4.2-8.

foNowing limits whenswer the reactor is critical or whenever the reactor cocient temperature is greater then 'i

' The ama==iMa interior surfaces of the drywell and above 212*F and irradiated fuel is in the reactor vessel: the water line of the torus shen be inspected i goperating cyci) for evidence of deterioration.h l a.- Maximum lowel of 14.00 feet. wee _ ew me means Whenever there is indication of relief velve operation or l b. henimum level of 13.88 foot. -

testing which adds heet to the suppression pool, t'w pool The torus water lowel may be <=*eida the above temperature shen be continuously recorded until tt a heat i addition is terminated. The operator will verify that l limits for a maximum of four (4) hours as a result of average temperature is within applicable limits eve y 5 required operability testing of HPCI, RCIC, RHR, CS,: j minutes. In lieu of continuous recording, the open tor and the DryweN -Torus Vacuum Relief System. shalllog the temperature every 5 minutes.

c. Maximum water temperature Whenever there is indication of relief valve operation with (1) During normal power operation maximum the temperature of the suppression pool reaching 160*F water temperature shen be 95*F.

or more and the primary coolant system pressure prester

[ then 200 psig, an extemet visuel examination of the torus c shell be conducted before resumeng power operati(n.  ;

i1 A.T-.ni-..i No. /,/, , %b,Nk,pb,197

.us

~

JAREPP 3.7 (cont'd) 4.7 (cont'd) brooker is sooner made operable, provided that the repair procedure does not viciste primary i containment intepity.

Pressure Suppression Chamber - Drywe5 Vacuum 5. Pressure Suppression Chamber - DryweE Vacuum -

5.

Breakers Breakers When primary containment integrity is required, aE a. Each drywe5 suppression chamber vacuum breaker >

a.

sheE be exercised through an opening - closing cycle drywe5 suppmesion chamber vacuum bremsers sheE he operable and positioned in the fugy closed monthly.

positica except durin0 testin0 and as specified in 3.7.A.5.b beloor.  !

b. One drywe5 suppremeien chamber vacuum breaker b. When it is determined that one vacuum breaker is may be non-fuGy closed so Song as it is determined inoperable for fugy closing when operability is to be not more then 1* open as irh by the required, the operable breakers sheE be exercised

" . n ' - ',, and every 15 days thereafter until the position lights.

k--; " volve has been retumed to normat sWWICo.

I pa- 24 msH%s

c. One drywe5 suppression chamber vacuum breaker c. OnceM osermann cvbIvecuum bre-ker f .

may be determined to be inoperable for opening.

valve sheE be visueEy inspected to insure prey

! Q maintenance and operation.

i 24 mw%s !l

)

,d. A leek test of the drywell to suppression chamberJ l

i (l d. Deleted

. structure shall be conducted once per@aQ "

l the acceptable leek rate is 50.25 in.

waterknin, over a 10 min penod, with the drywell at 1 poid.

i

Am.ndm.nt No. J/4,192 118

_ _ - _ _ _ . _ _ - . . . - . - . - . - - - - - - -- - - - - - - - - - - - . - - - - - . - - - - - - - - ~ - < - ~ ~ - a. - - -

I JAFMPP

• 3.7 (coat'd) 4.7 (cont *d)

e. Leakage between the drywell and suppression e. Not applicable chamber shall not exceed a rate of 71 scfm as monitored via the suppression chamber 10 mia pressure transiost of 0.25 in. weter/mla.

f. The self actuated vacuum breakers shall open f. Not applicable O when subjected to a force equivalent to 0.5 psid acting on the valve disc.

om par :14 mw& d ,

f g..From and after the date that one of the g. During each refueling outage]each vacuum l

pressure-suppression chamber /drywell vacuum breaker shall be tested to determine that the breakers is made or found to be inoperable for force required to open the vacuum breaker does any reason, the vacuum breaker shall Le locked not exceed the force specified la Specifica-closed and reactor operation la permissible tion 3.7.A.S.f and each vaca w breaker shall only durlag the succeedlag seven days maless be inspected and verified to meet design such vacueus breaker is sooner made operable, requirements.

provided that the repair procedure does not violate primary containment integrity.

t b

i L

I Amendment No. [ ,134 '

JAFNPP 3.7 (co:st'd) 4.7 (cont'd) c <

l -

Standbv Gas Treatment System B. Standby Gas Treatment System B.

1. Except as specified in 3.7.B.2 below both circuits of the 1. Standby Gas Treatment System surveillance shall be Standby Gas Treatment System shall be operable at all performed as indicated below:

times when secondary containment integrity is required. ONcs ev 24 modns

a. At least once per operating ch it shall be demonstrated that:

(1) Pressure drop across the combined high-efficiency and charcoal filters is less than +

5.7 in. of water at 6,000 scfm, and

! (2) Each 39kW heater shall dissipate greater than 29kW of electric power as calculated by the l following expression:

P=/TEI where: l P= Dissipated Electrical Power: l E= Measured line-to-line voltage in volts (RMS);

i = Average measured phase current in amperes (RMS). .

Amendment No. 10, 28, 2S, SS, SS, 81, 83,1 S1, 221 181 i

a e e C A e,,A,-4. E. A: E,L-1 . 4-Am. m.- 4 _,,4u ma 1., m_m __A., ,&

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4.7 (cont'd)

J/ M F - ,

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' b. At least. Once' during each scheduled secondary coptain nent - leak rate tent,, is shenever a filter changed, whenever work is p.rforined that could affcet system the f il t.cr -

~ ~ efficiency,-and at inteirvals '

r:ot to exceed nix nonths in tsecen - r ef uelitig out.igr o ,

~-

shall :be demount.ra t ed it that:

removal efficiency.

~

(1.) The of the par t. icula te

, filters is not. -lenu t han IDP 99 percent b::cd on s

., test per Af!SI  !!101.1-Ir>72

' tura. 4.1.

S &

(2.)The rc:noval effielency e of each of the' charcoal fi'Iters is not.lcss than.

99 percent based on a Trcon tent.

I ', . I', '

c. At least once- cach yr, removabic charcoal shall be recoved i

cartridges and ab:: ort. tion capability ,

shall be derutintrated.

! d. . At Icast once per operating [

g cyclefuttyn itic Atittiation

)% each branch of the ot Standby t;an Treatment. System l i shall be denwirintrated. '

t I

- Amendment No. 10 -

182-t

____.---.-n-_-. -_ - - . . _ _ - _ _ __

_ _ _ _ _ _ _ _ . - _ _ - - - _ _ . - - _ - _ . - _n- -x- - --- - _ - _- _ . _ - . _ - _ , - . _ ___ _ _ -_ w =.- -- _ -__ - _ e w -- __ - - - - _- a _ . - ---J

- - - _ _ _ - - - _ . _ - - - - - - - - _ . - - - - . - - - - - - - - ~ - - - - - - -

% 4 JAFNPP 3.7 (cont *c4 4.7 (cont'd)

e. At least once per operasano cvckR manued operatmhty i i of the tqpass valve for tiller cochn0 sheE be .

demonstrased.

r. Standby Gas Treatment System instrumentahon cambrason-

. dinerential once/operahng pressure cycle l switches l 2. From and after the dele that one circuit of the Stancey Gas l Troelmont System is made or lound to be inoperable for 2. When one circuit of the Standby Gas Treatment System any reason, Wie fotowing wondd apply: becomes enoperable, the operable circuit shall be venlied to be operable immedasely and daily thereaker.

a. N in Startg/ Hot Standby, Run or Hot Shtedown

• mode, reactor operation or irrasseted fuel hentsing is l permiselbes only charing the succescAng 7 days unises aucit circuit is sooner made oper%

provided that durin0 such 7 days all active  !

componenes of the other Standby Gas Treatment l Circuit sheB be operabia. '

l

b. N in Refuel or Cold Shutdown mode, reactor

• operation or irracSeted fuel hencSing is permisable only during the ==-e ng 31 days unless such  !

circuit is sooner made operable, provided that  !

during such 31 days aR active components of the cIher Standty Gas Trealment Circut shen be  ;

j operable.

3. If Mahnns 3.7.B.1 and 3.7.B.2 are not met, the raar4ar shen be placed in the cold condition and irrartasart 3. W M ( t' fuel hancSing operations and operahons that could red"ce -

the shutdown margin shall be prohibited.

l i

Amendment No. )#,ps, pdf 154 183 ,

t

JAFMPP

• y.

3.7 (cont'd) 4.7 (cont'd) -

c. Secondary contamment capability to maintain a 1/4 in. of water vacuum under calm wind conditons with a filter train flow rate of not more than 6,000 cim, shaN be demonstratedCat each>u Gruelmg outagWe to refuelmg j c m e-r an w %

D. Pnmary Contananant laalasian Vaivan D. Primarv Corde lee 8% Vd= '

1. Whenever primary containment insegnty is required per 1.

The pnmary contamment isoladen valves sunmilance 3.7.A.2, containneent isolation vseves and aN instrument shat be performed as foNows-Ene excess Sour check valves shot be operable, d) on - 49 m oa N ,

except as speclSed in 3.7.D.2. The containment vent a. _t least once per operatmo cycle) the operable and purge wafves sheB be Emited to openug angles isolaten valves that are power operated and less Wien or espel to Wiet spectBed below; automabcaNy wetated shaN be tested for g w " _ =- n.=___.A--- -

sumdated automabc watahon and for closure time.

4(f

~

27AOV-111 .

27AOV-112 40' b. At least once por operating cycle. the instrument 27AOV-113 4(f line excess flow valves shaN be tested for 27AOV-114 50' proper operation.*

l 27AOV-115 50' 27AOV-116 Sif c. At least once por quarter-27AOV-117 50' '

27AOV-118 50' (1.) AN normaNy open power-operated isolabon valves (except for the main stream line and Reactor Buildng Closed Loop Coolmg , ,

Water System (ROCLCWS) power-operated isolagen valves) shaN be futy closed and reopened. d

[ The ont sunellance interval testing instrument line excess flow check valves is extended until the end of the R11C12 refuelmg outage scheduled for January,1995. This is a )

one-time extensen, effective only for this sunellance interval. The next survedlance interval win begm upon complehen of this survedlance.

(

Amendment No.164. Hl4, 195 185

. 4 4.7 BASES (cont'd) JAFHPP building isolation valves, leak- may be run for test purposes, yum .j tightness of the reactor building and the test g and performance of the Standby Gas plugging interval of is unlikely,foperatina once per cycle > - -

Treatment System. Functionally is reasonable. Duct heater tests " '

testing the initiating sensors and will be conducted once (ur'ing eac)h -

associated trip channels per .2%A operating cycle) Considering the demonstrates the capability for simplicity of the heating circuit, j automatic - actuation. Performing the test frequency is sufficient.

these tests prior to refueling will t demonstrate secondary containment he in place testing of charcoal capability prior to the time the filters is performed using Freon or i primary containment is opened for equivalent, which is injected into refueling. Periodic testing gives the systen upstream of the charcoal i sufficient confidence of reactor filters. Measurements of the Freon ,

building integrity and standby Gas concentration upstream and Treatment System performance downstream of the charcoal filters

capability. is made. The ratio of the inlet and. I outlet concentrations gives an h e test frequencies are adequate to overall indication of the leak i detect egufipment deterioration prior tightness of the system. Although to significant defects, but the this is basically a leak test, since tests are eot frequent enough to the filters have charcoal of known load the filters, thus reducing efficiency and holding capacity for

their reserve capacity too quickly. . elemental iodine 'and/br- methyl

! That' the  : testing frequency is iodine, the test also gives an i adequate to detect deterioration was indication of .the relative demonstrated by the tests which efficiency of the installed -system. i showed no loss of filter efficiency ~

after 2 yr. . of operation in the High-efficiency particulate filters ,

rugged shipboard environment on the are installed to minimize gotential NS Savannah: (ORNL 3726) . Pressure release of particulates to the drop tests across filter sections environment. An efficienew of are performed to detect gross 90 percent is ' adequate to retr.in plugging or leak paths through the particulates that may be released to filter media. Considering the the reactor building following an relatively short time that the fans accident. This will be demonstrated t y W p acl e_. 4-esbNg u A b c9 < 5 j Ss% rneAem. 4 195 .t i

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n eggmom ra,e r

. (c t e oeoric N m e ts r,e es c v l

s t

. S i y hlu eev eyef i

f n

E T p o d yl f

i r hyonr P TScl opal s r nnu e S

A B

h i i i d

m n

7 .

e 4

D m I l A

t. ..

9 JAFNPP  ;

4.7 BASES (cont'd) ,

operatukty results in a more reholde syst The main steem line isolption volves are functioneNy tested on a more frequent interwel to esteldeh a high doores of i

- rolleidity.

l The primary containment is penetrated try several amen f

. ehemoter instrument lines connected to the reactor coolant

. system.~ Each instrument Nne contains a 0.25 in. restricting i

, orifice inside the primary containment and an excess Sow check valve outside the pnmery containment.

i The RSCl.CWS wolves are ==<*=8ad from the guerterly surweigence requirements w=== closure of these valves win -

eliminate the cociant flow to the drywet air and recircadetion -

f pumgHnotor coolers. Without cocEne weser, the dryweN air I and equipment temperature wiB increase and may cause s demoso to the equipment during nonnel plant operations.  :

C I **""*"'**'*******************'**'" '

the cold conscion.

A list of containment isolation volves, including a brief  ;

(l doeceiption of each velve is included in Section 7.3 of the l

updated FSAR.

bl '

i l

I i

Amendment No.1[,[3, 203 197  ;

i

l ATTACHMENT IV to JPN-95-046 4

Reference 2 NYPA Report JAF-RPT-MULTI-01116 Rev.1 Containment Systems Surveillance Test Extensions 1 I

1 1

2 4

l

! New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59