ML20066E814

Proposed Tech Spec Changes Deleting Table 3.7-1, Primary Containment Isolation Valves

ML20066E814
Person / Time
Site:	FitzPatrick
Issue date:	01/16/1991
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20066E784	List: JPN-91-002, Forwards Application for Amend to License DPR-59,deleting Table 3.7-1, Primary Containment Isolation Valve, from Tech Specs ML20066E798 ML20066E814
References
NUDOCS 9101220286
Download: ML20066E814 (15)

Text

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ATTACHMENT l PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING THE DELETION OF TABLE 3.71

" PRIMARY CONTAINMENT ISOLATION VALVES" JPTS 89 027 NewYork Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 9101220286 910116 PDR ADOCK 05000333 P PDR

JAFNPP UST OF TABLES (Cont'd)

Tablo Title Pyo 4.2 8 Minimum Test and Calibration Frequency for Accident Monitoring 86a Instrumentation 4.6-1 Comparison of the James A. RtzPatrick Nuclear Power Plant Inservico 157 Inspection Program to ASME Inservice inspection Code Roquirements 4.6 2 Minimum Test and Calibration Frequency for Drywoll Continuous 162a Atmosphoto Radioactivity Monitoring System 4.7 1 Minimum Test and Calibration Frequency for Containmont Monitoring 210 Systems 4,7 2 Exception to Typo C Tests 211 3.12 1 Water Spray / Sprinkler Piotected Areas 244]

3.12 2 Carbon Dioxido Protected Areas 244k 3.12 3 Manual Firo Hoso Stations 2441

-4.12 1 Water Spray / Sprinkler System Tests 244q 4.12 2 Carbon Dioxido System Tests 244r 4.12 3 Manual Firo Hose Station Tests 244s 6.2 1 Minimum Shift Manning Requirements 260a 6.10-1 Component Cyclic or Transient Umits 261 1

Amendment No. M,%1% 1[,1[1[,

vi

JAFNPP 32 BASES in addition to reactor protection instrumentation which initiates Actuation of primary containment valves is initiated by a reactor scram, protective instrumentation has been provided protective instrumentation shown in Table 32-1 which senses which initiates action to mitigate the consequences of the conditions for which isolation is required. Such accidents which are beyond the operator's ability to control, or instrumentation must be availab!e whenever primary terminates operator errors before they result in serious containment irugrity is required.

consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation De istmmentation did initiates pig @ tem isolation is connected in a dual bus arrangement.

funct, ion, initiation of the Core Cocling Systems, Control Rod Block and Standby Gas Treatment Systems. The objectives of The low water level instrumentation set to trip at 177 in. above the s,occifications are to assure the effectiveness of the the top of the active fuel closes all isolation valves except those protective instrumentation when required, even during periods in Group 1. Details of valve grouping are given in the JAF when portions of such systems are out of service for FSAR section 7.3. For valves which isolate at this level, this trip maintenance, and to prescribe the trip settings required to setting is adequate to prevent uncovering the core in the case assure adequate performance. When necessary, one channel of a breiin the largest line. l may be made inoperable for brief intervals to conduct required The low-low reactor water level instrumentation is set to trip functional tests and calibrations.

when reactor water level is 126.s in. above the top of active Some of the settings on the instrumentation that initiate or fuel. This trip control core and containment cooling have to!crances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Amendment No. , ,

ss

JAFNPP .

Venturis are provided in the main steam lines as a means of initiates the HPCI and RCIC and trips the recirculation pumps.

The low-low-low reactor water level instrumentation is set to trip measuring steam flow and also limiting the loss of mass when the water level is 18 in. above the top of active fuel. This inventory from the vessel during a steam line break accident.

The primary function of the instrumentation is to detect a break trip activates the remainder of the ECCS subsystems, closes in the main steam line. For the worst case accident, main the main steam isolation valves, main steam line drain valves steam tine break outside the drywell, a trip setting of 140 and reactor water sample ""o isolation valves, and starts the percent of rated steam flow in conjunction with the flow limiters cmergency diesel genere .s. These trip level settings were and main steam line valve closure, limits the mass inventory chosen to be high enoug.. to prevent spurious actuation but >

loss such that fuel is not uncovered, fuel temperature peak at low enough to initiate ECCS operation and primary system isolation so that post-accident cooling can be accomplished approximately 1,000T and release of radioactivity to the environs is bebw 10CFR100 guidelines. Reference Section and the guidelines of 10CFR100 will not be exceeded. For 14.6.5 FSAR.

large breaks up to the complete circumferential break of a 24 in. recirculation line and '.vith the trip setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. Reference paragraph 6.5.3.1 FSAR.

The high drywell pressure instrumentation is a diverse signal for malfunctions to the water level instrumentation and in addition l

to initiating ECCS, it causes isolation of Groups B and C isolation valves. For the breaks discussed above, this instrumentation will generally initiate ECCS operation before the low-low-low water level instrumentation; thus the results given above are applicable here also. Details of the isolation valve closure group are given in the JAF 5 SAR section 7.3. The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents.

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Amendment No. gd,4[, f, If, se

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 at each refueling outage prior to refueling.

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

1. The primary containment isolation valves surveillance shall

1. Whenever primary containment integrity is required per be performed as follows:

3.7.A.2, containrnent isolation valves and all instrument line l

excess flow check valves shall be operable, except as a. At least once per operating cycle, the operatic j specified in 3.7.D.2. The containov.nt vent and purge isolation valves that are power operated and valves shall be limited to opening as igtes less than or equal automatically initiated shall be tested for simulated to that specified below- automatic initiation and for closure times as specified in the JAF FSAR section 7.3. l Valve Number Maximum Opening Angie 27AOV-111 40* b. At least once per operating cycle, the instrument line 27AOV-112 40 excess flow check valves shall be tested for proper 27AOV-113 40* cperation.

c. At least once per quarter *

(1.) All normally open power-operated isolation valves (except for the main stream line and 27AOV-118 50 Reactor Building Closed Loop Cooling Water System (RBCLCWS) power-operated isolation valves) shall te fu!!y closed and reopened i Amendment No.1 ,

JAFNPP.

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

(2.) With the reactor at reduced power level, trip main steam isolation valves and verify closure time.

d. At least twice per week, the main steam line power-operated isolation valves shall be exercised by -

partial closure and subsequent reopening.

c. The RBCLCWS isolation valves shall be fully closed and reopened any time the reactor is in the cold

. condition exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />,if the valves have not

2. With one or more of the containment isolation va!ves listed been fully closed and reopened during the preceding in the JAF FSAR section 7.3 inoperable, maintain at least y~ ~

one isolation valve operable in each affected penetration that is open and:

2. Whenever a containment isolation valve listed in the JAF FSAR section 7.3 is inoperable, the position of at least one

a. Restore the . inoperable valve (s) to operable status other valve in each line having an inoperable valve shall be with,n i 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or recorded d-N.

b. Isolate each affected penetration vdthin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the closed position. Isolation valves ,

closed to satisfy these requiremerits may be i l

reopened on an intermittent basis under j administrative control; or I c. Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or a blind flange.

3. If Specifications 3.7.D 1 or 3.7 D2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in cold condition within 24 hrs.

Amendment No.1/ $ ,1[,

186 l

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JAFNPP 3.7 BASES (cont'd)

A list of containment isolation valves, including a brief of the containment. Closure of one of the valves in each line would be sufficient to maintain the integrity of the Pressure description of each valve is included in the tpdated JAF FSAR Suppression System. Automatic initiation is required to section 7.3.

f I minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.

The containment isolation valves on the containment vent and purge lines may be open for safety related reasons. Safety related reasons include, but are not limited to, the foi!owing:

inerting or de-inerting primary containment; maintaining containment oxygen concentration; maintaining drywell and suppression pool atmospheric pressures; and maintaining the differential pressure between the drywell and suppression pool.

These valves have been modified to limit the maximum angle of opening as shown in 3.7.D.1.

Nine remote manual isolation valves have been added to the l Reactor Building Closed Loop Cooling Water System l

(RBCLCWS) in order to comply with 10 CFR 50 Appendix A I GDC 57; These valves are air operated (with solenoid pilot valves), normally open, and are designed to fail "open'on loss of electrical power or "as is' upon loss of instrument air. Each AOV is provided with a Seismic Class I accumulator tank to allow operation of the valves upon loss of instrument air up to 2 full valve cycles. The fail-open design permits continued operation of the system to supply water to the recirculation pump-motor coolers and drywell coolers during normal operation and as necessary under accident conditions. If there is a postulated accident, and indications of leakage from

' RBCLCWS appear, the operator will selectively close the AOV's affected to provide containment isolation.

Amendment No.1jd, 192

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p

'JAFNPP c ~+

4.7 BASES (cont'd) i

- operability results in a more reliable system.. .,

The main steam line isolation valves are functionally. tested on a'

. more frequent interval to' establish a high degree of reliability; The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant

.. - system..' Each instrument line contains a 0.25 in. . restricting I. .

orifice inside the primary containment and an excess flow check valve outside the primary containment. - '!

The RBCLCWS valves are excluded from the quarterly .

l 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.

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

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1 ATTACHMENT 11 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING THE DELETION OF TABLE 3.71

• PRIMARY CONTAINMENT ISOLATION VALVES" JPTS-89-027 r.

'k:

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

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Attachmont ll SAFETY EVALUATION Page 1 of 5

1. DESCRIPTION OF THE PROPOSED CHANGES The proposed chango to the James A. FitzPatrick Technical Specifications dolotos Tablo 3.71 (ontitled, " Primary Containment Isolation Valvos" on pagos 193 through 209) along with tho associated notes and any references to this tablo. References to a similar tablo in the updated FitzPatrick Final Safety Analysis Report (FSAR) replaces the doloted tablo.

Specifically, this amendment proposos the following changos to the FitzPatrick Technical Specifications:

1. Table 3.71 and its associated notes currently occupy twelva pagos extonding from page 198 to 209. All twelvo pagos have boon replaced with a singlo page bearing the words *PAGES 198 THROUGH 209 HAVE BEEN DELETED." A note on the bottom of the pago will a! ort the roador that the 'Next pago is 210."

Those twelvo pages previously contained the following:

Pages 198 through 209 Tablo 3.7-1, " Primary Containment Isolation Valves."

Pago 207 " Notes for Tablo 3.7 t, Isdation Signal Codes " listed and described the isolation signal codos.

Pages 208 and 209 " Notes for Tablo 3.71," listed fiftoon explanatory notes associated with Table 3.7-1.

2. In Section 3.2 on pago 55 the phraso,'Detalls of valve grouping and required closing timos are given in Specification 3.7" has boon icvised to road, "Dotails of vaivo grouping aro given in the JAF FSAR F,cctio'17.3." The phrase " assuming a 60 second valve closing timo" and the sentence following the phrase have both boon deloted.

3. In Section 3.2 on page 56 the phrase, %os Specification 3.7 for isolation valvo closure group" has boon revised to rond, "Dotails of the isolation valve group are given in the JAF FSAR section 7.3." Tta parase "it causes isolation of Groups B and 3" has boon replaced with "it causes .m.et;on of Groups B and C."

4. Table 3.71 has been doloted from the L st of Tables on page vi,

5. In Section 3.7.D.1 on pago 185, the rDrie, "specified in Table 3.71" has boon dolotod. The next sentence has berm (to eted.

6. In Section 3.7.D.2 on pago 186, the o i n o " listed in Table 3.7-1" has boon replaced with " listed in the JAF FSAR section ', c.l."

Attachment ll SAFETY EVALUATION Page 2 of 5

7. 1n Soction 4.7.D.1.a on page 185, the phraso "and for closuro times as specified in Tablo 3.71' has boon revised to road, "and for closure timos as specified in the JAF FSAR section 7.3"

8. Soction 4.7.D.2 on page 186, the phraso " listed in Table 3.71" has boon replaced with *llstod in the JAF FSAR section 7.3."

9. A referenco to tho tablo of containment isolation valves in the updattd JAF FSAR 1 section 7.3 has boon added to Basos 3.7 and 4.7 (on pagos 192 and 197 respectively).

II, PURPOSE OF THE PROPOSED CHANGES The purpose of this chango is to reduce the administrativo resources required by both the NRC and the Authority to maintain an accurate and up to-aato table of containment isolation valves. The elimination of lists of components has boon identified as a generic improvement

-by both Industry and NRC programs.

The Authorit/ endorsed technical specification reform activities in November 1985 (Reference 2), Over four years ago, both the NRC's Technical Specifications improvement Project Report (Reference 3) and the Atomic industrial Forum's Report (Roforence 4) ondorsed the idea of using the FSAR as an appropriato placo for this type of information.

Rathor than preparing and submitting a Technical Specification amendment request for each Tabic 3.71 alteration, the Authority will maintain the table of containment isolation valves In s'io FitzPatrick updated FSAR. This will assure that the tabic is periodically updatoc without the administrativo burdon of an operating licenso amendment.

. FSAR Table 7.3.1, entitled " Process Piping Ponotrating Primary Containment," was included in the FSAR updato issued in July 1986 (Reference 1). This now table arranges entries by containment penetration, better describes the penetrations' function and clearly identifics the associated containment isolation valvo(s). Isolation signals, valvo closure timo, and normal status are includod on the FSAR version of this table. The now tablo also has a noto section similar to the one in the Technical Specifications, but the FSAR notes have boon clarified.

Other improvements make the table more useful and easier to uso. A different format climinatos the need to reduce the table photographically and unnecessary columns in the tablo have boon doloted.

Tab!o 4.7 2 ontitled, " Exception to Type C Tests," (pages 211,212,213,213a, and 213b) lists for Typo C tests exceptions for certain containment penetrations and containment isolation valves. This tablo is not doloted as part of this amendment application and will be retained as part of the Technical Specifications.

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s Attachment 11-

,M' SAFETY EVALUATION i Pago 3 of 5 lili ~ EFFECT OF THE PROPOSED CHANGES The dolotion of the table of containment isolation valvos is purely administrativo in nature and will not degrado safety, This amendment does not alter or remove any operability or survolllanco requiromonts currently in tho FitzPatrick Technical Specifications.

10 CFR 50 contains adequato requirements applicable to containment isolation valvos to assure the safo operation of the FitzPatrick plant whether or not a list of containment isolation valves is included in the Technical Specifications.

Both 10 CFR 50.59, 'Changos, tests and experiments" and 10 CFR 50.71(o), "Maintenanco of records, making of reports" already contain provisions which requiro the Authority to inform the NRC of changes to the plant 10 CFR 50.59 permits the Authority to:

"(i) make changes in the facility as described in the safety analysis report, (ii) make changos in the proceduros as described in the safety analysis report,...without prior Commission' approval unless the proposed change, tom or experiment involves a

- change in the technical specifications incorporated in the licenso or an unroviewed safety. question." s

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10 CFR 50.59'roquires that the Authority annually submit to the NRC 'a report containing a _

brief description _of such changes, tests or experiments, including a summary of the safety

' ovaluation of each.".>This part also requires that the Authority completo a safety evaluation for cach chango to assure that the chango does not involve an "unroviewed safety

_ question."L10 CFR 50.59 (a) (2) defines throo critoria to be applied to determino if a change

" involves an unroviewed safoty. question."

- 10 CFR 50.71(e) ' requires the Authority to reviso the FitzPatrick FSAR cach year "to assuro 4 that the information included in the FSAR contains the latest matorial developod."

LiO_CFR 50.59 and.10 CFR 50.71 provido adequato assurance that changes to the plant that - .

result in changes to this table will be evaluated by the Authority and reported to the NRC.

Two other power reactor licensees have boon granted similar amendments by the NRC to

!doloto the table of containmont isolation valves from Technical Specifications (References 5 -

and 6),-

- IV,' EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with this proposed

-amendment would not involve a significant hazards consideration, as defined in 10 CFR L 50.92, since the proposed changes would not:

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-Attachmont il SAFETY EVALUATION >

Pago 4 of 5

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1. ~ Involve a significant increase in the probability of an accidont or consequence q previously evaluated - The relocation of this information from the Technical Spocifications to the FSAR is purely an administrativo chango. It will havo_ no offect on how the plant is maintained or operated nor does it alter the plant's design.l Fodoral rogulations 10 CFR 50.59 and 10 CFR 50.71 afroady contain .

provisions that requiro the Authority to complete a safety ovaluation of any changos to the plant, to report those changes annually, and to updato the FSAR. l 2.: croato the possibility of a now or different kind of accident from those previously evaluated. The rolocation of the containmont isolation valvo tablo

. does not involve a modification to the plant or a change in the procedures used for plant operation.-

3. Involvo a significant reduction in the margin of safety. A similar tablo has boon provided in the updated FitzPatrick FSAR. The FSAR is revised in accordanco -

with the provisions of 10 CFR 50.71(o). This amendment Joos not attor any oporability or survolllanco requin omonts currently in the FitzPatrick Technical eSpecifications.

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^ V.- lMPLEMENTATION OF THE PROPOSED CHANGES j b

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, Implomontation of the proposed changos will not impact the Al ARA or Firo Protection l

Programs at the FitzPatrick plant, nor will the changes impact the environmont.

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= VI. CONCLUSION i Thoso changos, as proposed, do not constituto an unroviewed safety question as defined in a 10 CFR 50.59. That is, they:

- a. will not increaso the probability of occurronce or the consequences of an accident or- 1

= malfunction of equipment important to safety previously evaluated in the safety analysis report;-  ;

b. will not increase the possibility for an accident or malfunction of a different typo froni - ,

. any evaluated previously in the safoty analysis report; j c. will not educe the margin of safety as defined in the basis for any techn_ical

' specification; and c  ! d. Involve no significant hazards consideration, as defined in 10 CFR 50.92.

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Attachment ll

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SAFETY EVALUATION Page 5 of 5

- Vll. REFERENCES

1. NYPA letter, J.C. Brons to H.R. Denton dated July 22,1986 (JPN 86 032) regarding 1986 annual FSAR updato.

2. NYPA lotter, J.C. Brons to H.R. Donton dated November 15,1985 (JPN 85 083 IPN-85 060) regarding NYPA ondorscmont of technical specification rotorm activities.

3. NRC Technical Specifications improvement Project Final Report, "Rocommendations for improving Technical Specifications," dated September 30,1985.

4. Atomic Industrial Forum, "Tochnical Specifications Improvomonts," Subcommittoo on Technical Specification improvements, dated Octobor 1,1985.

5. NRC letter, T.V. Wambach to D.C. Shelton, dated April 13,1990, regarding the do!otion of Tablo 3.6.2, Containment Isolation Valvos in its entirety from the Davis Bosso Technical Specifications.

6. NRC lotter, H. Silver to W.S. Wilgus, dated May 22,1989 regarding the dolotion of the Containment Isolation Valvo Tablo from the Crystal River Unit 3 Technical Specifications.

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