ML20211F300

Proposed Changes to Tech Specs to Achieve Conformance W/ Acceptance Criteria of NUREG-0737 & Generic Ltrs 83-02 & 83-36

ML20211F300
Person / Time
Site:	FitzPatrick
Issue date:	10/20/1986
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20211F259	List: ML20211F256 ML20211F275 ML20211F300
References
RTR-NUREG-0737, RTR-NUREG-737 GL-83-02, GL-83-2, GL-83-36, JPN-86-47, NUDOCS 8610310122
Download: ML20211F300 (13)

Text

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JAFNPP TABLE OF CONTENTS (cont'd)

Page D.

Emergency Service Water System 240 E.

Intake Deicing Heaters 242 3.12 Fire Protection Systems 4.12 244a A.

High Pressure Water Fire Protection System 244a B.

Water Spray and Sprinkler Systems 244e C.

Carbon Dioxide Systems 244e D.

Manual Fire Hose Stations 244f E.

Fire Protection Systems Smoke and Heat Detectors 244g F.

Fire Barrier Penetration Seals 244g 5.0 Design Features 245 5.1 Site 245 5.2 Reactor 245 5.3 Reactor Pressure Vessel 245 5.4 containment 245 5.5 Fuel Storage 245 5.6 Seismic Design 246 6.0 Administrative Controls 247 6.1 Responsibility 247 6.2 Plant Staff Organization 247 6.3 Plant Staff Qualifications 248a 6.4 Retraining and Replacement Trainin9 248a 6.5 Review and Audit 248a 6.5.1 Plant Operating Review Committee (PORC) 248a 6.5.2 Saf ety Review Committee (SRC) 250 6.6 Reportable Occurrence Tction 253 6.7 Safety Limit Violation 253 6.8 Procedures 253 6.9 Reporting Requirements 254a 6.10 Record Retention 254g 6.11 Radiation Protection Program 255 6.12 Industrial Security Program 258 6.13 Emergency Plan 258 6.14 Fire Protection Program 258 6.15 Environmental Qualification 258a Amendment No. [, f[, [, }[, yI, %

iii 8610310122 861020 PDR ADOCK 05000333 P

PDR

JAFNPP TABLE 3.2-8 ACCIDENT MONITORING INSTRUMENTATION NO. OF MINIMUM NO. OF CHANNELS PROVIDED OPERABLE CHANNELS MEASUREMENT INSTRUMENT BY DESIGN REOUIRED ACTION RANGE 1.

Stack High Range Effluent 2

1 B

10-1 to 107 mR/hr.

Monitor 2.

Turbine Building Vent High 2

1 B

10-1 to 107 mR/hr.

l Range Effluent Monitor 3.

Radwaste Building Vent High 2

1 B

10-1 to 107 mR/hr.

Range Effluent Monitor 4.

Containment High Range 2

1 A

1 to 108 rads /hr.

Radiation Monitor

• 5.

Containment Pressure 2 wide range 1

A 0 to 250 psig 2 narrow range 1

A

-5 to +5 psig 6.

Drywell Level 2

1 A

22 to 100 ft.

(H O) 2 (H O) 7.

Suppression Pool Level 2

1 A

1.7 to 27.5 ft.

2 8.

Reactor Vessel Pressure 2

1 A

0 to 1500 psig 9.

Drywell Hydrogen 2

1 A

0 to 30% H2 Concentration Monitor j

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• At 450 R/hr, closes vent and purge valves Amendment No.

77a

JAFNPP TABLE 3.2-8 (cent'd)

ACCIDENT MONITORING INSTRUMENTATION i

NOTES FOR TABLE 3.2-8 With the. number of operable channels less than the required minimum, either restore the inoperable channels A.

to operable status within 30 days, or: (1) initiate an alternate method of monitoring the appropriate parameter (s), or (2) be in a cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With the number of OPERABLE channels less than required by the minimum channels OPERABLE requirements, B.

initiate the alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and: 1) either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or 2) prepare and submit a Special Report to the Commission within 14 days following the event outlining the cause of the inoperability, the action taken, and the plans and schedule for restoring the system to OPERABLE status.

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4 Amendment No.

77b t

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+

JAFBPP t

3.5 (Cont'd) 4.5 (Cont'd)

E.

Reactor core Isolation Coolina (RCIC) System E.

Reactor Core Isolation Coolina (RCIC) System 1.

The RCIC System shall be operable whenever 1.

RCIC System testing shall be. performed as there is irradiated fuel in the reactor follows provided a reactor steam supply. is vessel and the reactor pressure is greater available.

If steam is not available at.the than 150 psig and prior to a reactor startup time the surveillance test is scheduled to from a cold condition, except from the time be performed, the test shall be performed.

that the RCIC System is made or found to be within ten days of continuous operation from inoperable for any reason, continued reactor the time steam becomes available.

power operation is permissible during the succeeding 7 days unless the system is made Item Frequency operable earlier provided that during these 7 days the HPCI System is operable.

a.

Simulated Automatic Once/ operating.

Actuation (and cycle 2.

If the requirements of 3.5.E cannot be met, Restart *) Test the reactor shall be placed in the cold condition and pressure less than 150 psig b.

Pump Operability Once/ month within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

Motor Operated Once/ month 3.

Low power physics testing and reactor opera-Valve Operability tor training shall be permitted with inoper-able components as specified in 3.5.E.2 d.

Flow Rate Once/3 months

above, provided that reactor coolant e.

Testable Tested for opera-temperature is.I212*F.

Check Valves bility 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 operability tests have not been performed during the pre-ceding 31 days.

f.

Logic System Once/ operating

• Automatic restart on a low water level signal which Functional Test cycle is subsequent to a high water level trip.

Amendmentf6 121

6.

In addition to items 1, 2 & 3 above, two tdditienzl opsrztcra chall be readily available on site whenever the reactor is in other than cold shutdown.

During cold shutdown, an additional operator shall be readily available on site.

7.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

8.

In the event of illness or absenteeism up to two (2) hours is allowed to restore the shift crew or fire brigade to normal complement.

9.

A Fire Brigade of five (5) or more members shall be maintained on site at all times. This excludes two (2) members of the minimum shift crew necessary for safe shutdown and any personnel required for other essential functions during a fire emergency.

10. A Shift Technical Advisor shall be on site and readily available to the control room except during the cold shutdown or refuel mode.

11. Administrative procedures shall be developed and implemented to limit the working hours of unit staff Who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and maintenance personnel who are working on safety-related systems.

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week While the plant is operating.

However, in the event that unforeseen proolems require substantial amounts of overtime to be used or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />

, straight, excluding shif t turnover time.

b.

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor mere than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period, all excluding shift turnover time.

c.

A break of at least eight hours should be allowed between work periods, including shift turnover time, d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Resident Manager or the Superintendent of power, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting.

Amendment No. 2[ h, [,

248 s

f f

(A) ROUTINE REPORTS AND REPORTABLE OCCURRENCES (Continued)

1. STARTUP REPORT (Continued) b.

Startup Reports shall be submitted within (1) 90 days following completion of the startup test program, or (2) 90 days following resumption or commencement of commercial power operation, or whichever is earliest.

If the Start-up Report does not cover both events, i.e.,

completion of startup test program and resumption or commencement of commercial power operation, supplementary reports shall be submitted at least every three months until both events are completed.

2. ANNUAL REPORTS' a.

Annual Occupational Exposure Tabulation A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receivi g exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions,.1/ e.g.,

reactor operations and surveillance, in-service inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totaling less than 20%

of the individual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

E b.

Annual Report of S/RV Failures and Challenges An annual report of safety / relief valve challcnges will be submitted prior to March 1 of each year.

3.

MONTHLY OPERATING REPORT A report providing a narrative summary of facility operating experience, major safety-related maintenance, and other pertinent information, should be submitted no later than the 15th of each month following the calendar month covered to the USNRC Director, Of fice of Management Ir 'armation and Program Control.

1/Thistabulationsupplementstheregarcementsoff20.407of10CFRPart20.

Amendment No. 3/'

j 254-b

6.19 POSTACCIDENT SAMPLING PROGRAN A program shall be established, implemented, and maintained which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

A) Training of personnel, B) procedures for sampling and analysis, C) provisions for maintenance of sampling and analysis L

1 Amendment go, 250e

ATTACHMENT II TO JPN-86-47 PROPOSED TECHNICAL SPECIFICATIONS CHANGES RELATED TO NUREG-0737 ITEMS (JPTS-83-10,JPTS-84-ll) 1, t

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NEW YORK POWER AUTHORITY t

JAMES A.

FITZPATRICK NUCLEAR POWER PLANT I

DOCKET NO. 50-333 I

DPR-59 i

7 Section I - Description of the Proposed Cnanges The proposed changes to the Technical Specifications are shown in Attachment I to the Application for Amendment.

The changes are as follows:

The Table of Contents on page 111 has been retyped to restore margins.

In addition, the following changes are incorporated.

Page numbers refer to the retyped revision of the Table.

The page numbers for sections 6.3, 6.4, and 6.5 are changed from "248" to "248a".

The following line is added:

6.15 Environmental Qualification 258a.

New pages 77a and 77b contain a newly created Table 3.2-8,

" Accident Monitoring Instrumentation."

Tne following item is added to Section 6.2 (page 248):

11. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Adequate shift coverage shall be maintained without routine heavy use of overtime.

The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is operating.

However, in the event that unforeseen problems require substantial amounts of overtime to be used or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a.

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time, b.

An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period, all excluding shift turnover time.

c.

A break of at least eight hours should be allowed between work periods, including shift turnover time.

d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

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1 Any deviation from the above guidelines shall be authorized by the Resident Manager or the Superintendent of Power, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting.

The following item is added on page 121 :

f.

Logic System Once/ operating Functional Test cycle Section 4.5.E.1 of page 121 is revised to read " Simulated Automatic Actuation (and Restart) Test *."

On the bottom of page 121 the following note is added: ** Automatic restart on a low water level signal which is subsequent to a high water level trip."

On page 254-b, item 2 is reformatted to read as follows:

2.

ANNUAL REPORTS a.

Annual Occupational Exposure Tabulation A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure 1/e.g.,

according to work and job functions, Tnservice reactor operations and surveillance,

}

inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignment to various duty functions may be estimates based y

on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling less than 20% of the individual total dose need not be l

accounted for.

In the aggregate, at least 80% of l

the total whole body dose received from external sources, shall be assigned to specific major work functions.

b.

Annual Report of S/RV Failures and Challenges An annual report of safety / relief valve failures

(

and challenges will be submitted prior to March 1 l

of each year.

l New page 258e adds new Section 6.19 on the "Postaccident Sampling Program."

Section II - Purpose of the Proposed Changes The changes are necessary as a result of the following NUREG-0737 items:

I.A.l.3.1 Limit Overtime II.B.3 Postaccident Sampling capability II.F.1.1 Noble Gas Effluent Monitor II.F.1.3 Containment High-Range Radiation Monitor II.F.1.4 Containment Pressure Monitor II.F.1.5 Containment Water Level Monitor II.F.1.6 Containment Hydrogen Monitor II.K.3.3 Reporting Safety Valve and Relief Valve Failures and Cnallenges II.K.3.13 Reactor Core Isolation Cooling (RCIC) Automatic Restart Section III - Impact of the Proposed Changes In accordance with the requirements of 10 CFR 50.92, the enclosed application is judged to involve no significant hazards based upon the following information:

The additional instrumentation described in these proposed Technical Specification changes should improve safety at the FitzPatrick plant by providing additional information which is helpful to the operator in assessing plant conditions following an accident.

The Commission has provided guidance concerning the application of the standards for making a "no significant hazard considerations" determination by providing certain examples in the Federal Register (F.R.) Vol. 48, No. 67 dated April 6, 1984, page 14870.

Example vii states:

"A change to make a license conform to changes in the i

regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations."

This change to the Technical Specifications reflects modifications made as a result of NUREG-0737.

Therefore, this example is judged to apply to this application.

The proposed changes on page iii are editorial in nature and considered to match Commission example (i),

"A purely administrative change to Technical Specifications; for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature."

The proposed new Item II in Section 6.2 on page 248 was identified by the NRC as a change needed to be consistent with NUREG-0737 Item 1.A.l.3.1, Limit Overtime.

The proposed cnange is considered to match Commission example (vii).

Since'this change will serve to limit overtime for plant staff who perform safety related functions, the proposed change is determined to involve no significant hazard considerations.

The proposed changes on page 121 are intended to incorporate in Technical Specifications the requirements of NUREG-0/37 Item II.K.3.13 and Generic Letter 83-02 for automatic restart of the RCIC System on low-low reactor water level following trip of the system on high reactor water level.

The other change proposed on page 121 adds a logic system function test for the Reactor Core Isolation Cooling System.

This test had been omitted from the original Technical Specifications.

The change proposed on page 254b was identified by the NRC as needed to be consistent with NUREG-0737 Item II.K.3.3,

" Reporting Safety and Relief Valve Failures and Challenges."

The proposed change is considered to match Commission example (vii).

Section IV - Evaluation of Significant Hazard Considerations The following is a more detailed 50.92 evaluation for this application:

(1)

The proposed license amendment would not involve a significant increase in the probability or consequences of an accident proviously evaluated, because the proposed accident monitoring instrumentation does not affect previous accident probability analyses.

(2)

The proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated, because no new mode of failure is introduced.

(3)

The proposed amendment would not involve a significant reduction in the margin of safety, because additional information provided to the operator following an accident would enhance the safety of the plant.

Limits on overtime for Plant staff who perform safety-related functions would also enhance the safety of the plant, as would increased reporting requirements and surveillance testing.

Section V - Implementation of the Changes Implementation of the changes, as proposed, will not impact the ALARA or fire protection programs at FitzPatrick, nor will the changes impact the environment.

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Ssction VI - Conclusion The incorporation of tnese enanges:

(a) will not change the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report-1 (b) will not increase the possibilicy of an accident or malfunction of different type than any evaluated previously in the Safety Analysis Report; (c) will not reduce the margin of safety as defined in the basis for any Technical Specifications; (d) does not constitute an unreviewed safety question as defined in 10 CFn 50.59; and (e) involves no significant hazard considerations, as defined in 10 CFR 50.92.

Section VII - References (1)

James A. FitzPatrick Nuclear' Power Plant Final Safety Analysis Report (FSAR).

(2)

James A.

FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).

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