ML20093M376

Forwards Responses to Re Environ Qualification of safety-related Equipment

ML20093M376
Person / Time
Site:	FitzPatrick
Issue date:	10/15/1984
From:	Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Vassallo D Office of Nuclear Reactor Regulation
References
JPN-84-67, NUDOCS 8410220095
Download: ML20093M376 (32)

Text

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' 123 Main Street White Plams, New Yark 10601 914 681.6200

1. > NewWrkPbwer Memorandum

& Authority October 15, 1984 JPN-84-67 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Equipment Qualification; Response to Questions

Reference:

NRC letter, D. B. Vassallo to J. P. Bayne, dated August 20, 1984.

Dear Sir:

The referenced letter transmitted questions on equipment qualification efforts in progress at the FitzPatrick plant.

Responses to these questions are enclosed as Attachment 1 and 2 to this letter.

If you have any questions, please do not hesitate to call Mr. J. A. Gray, Jr. of my staff.

Very truly yours, J. P. Bayne a /'^~P First Executive Vice President Cntef Operations Officer cc: Office of the Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 i

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4 Attachment I to JPN-84-67 '

Response,to NRC< Questions-on Equipment Qualification p ,

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.a October 15, 1984 -

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)- 'New York Power Authority' James'A. FitzPatrick Nuclear Power Plant '

Docket No. 50-333 ,

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uNRC Question 1. Submit all applicable justifications for continued operation (JCOs) that are

.p ( 2 currently being relied upon.

-p NYPA Response 1.. eVia' Reference 1 the Authority transmitted

'JCOs for those equipment items requiring

- corrective action to. establish

. environmental. qualification. In Reference 2 we were notified that these JCOs were

' adequate except those for the junction boxes to protect terminal-blocks and splices. The Authority provided a JCO for this equipment in Reference 3. In that

- letter, the Authority updated the status of the JCOs deleting those which were'no longer necessary and adding new ones which

- had been found to be required. Enclosure s 1 l'to this'1etter presents the currently applicable JCOs.

4 NRC QUESTION 2. For each JCO associated with equipment that,is assumed to fail, provide confirmation that no significant degradation of any safety function or

.mils eading:information to the' operator will occur as a result of failure of equipment under.the accident environment resulting from a design basis event.

,NYPA RESPONSE 2. I'n Attachmen't 2, each JCO associated with equipment that is assumed to fail, provides confirmation-that no significant degradation of any other safety function, or misleading information to the operator d' (other than'that resulting directly from the failed component), will occur as a result of failure of equipment under the environment resulting from a design basis accident.

NRC QUESTION 3. Confirm that in performing the review of the methodology to identify equipment within the scope of 10 CFR 50.49(b)(2) the following steps have been addressed:

a. A list was generated of safety-related electric equipment as defined in paragraph (b)(1) of 10 CFR 50.49 required to remain functional during or following

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4 design-basis Loss of Coolant Accident (LOCA) or High Energy Line Break (HELB)

Accidents. The LOCA/HELB accidents are the only design-basis accidents which result in significantly adverse environments to electrical equipment which is required for safe shutdown or accident

/ mitigation. The list was based on reviews of the Final Safety Analysis Report (FSAR), Technical Specifications, Emergency Operating Procedures, Piping and Instrumentation Diagrams (P& ids), and electrical distribution diagrams;

b. The elementary wiring diagrams of the safety-related electrical equipment identified in Item "a" were reviewed to

' identify any auxiliary devices electrically connected directly into the control or power circuitry of the safety-related equipment (e.g., automatic trips) whose failure due to postulated environmental conditions could prevent required operation of the safety-related equipment and;

c. The operation of the safety-related systems and equipment were reviewed to identify any directly mechanically connected auxiliary systems with electrical components which are necessary for the required operation cf the safety-related equipment (e.g., cooling water or lubricating systems). This involved the review of P& IDS, component technical manuals, and/or systems descriptions in the FSAR.

d. Nonsafety-related electrical circuits indirectly associated with the electrical equipment identified in Item "a" by common power supply or physical proximity were considered by a review of the electrical design including the use of applicable industry standards (e.g., IEEE, FEMA, ANSI, UL, and NED) and the use of properly coordinated protective relays, circuit breakers, and fuses for electrical fault protection.

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NYPA RESPONSE 3.a. This item was addressed in Enclosure 3 to Reference 2. That Enclosure provided the

" Summary of Methodology for Identifying Electrical Equipment within the Scope of 10 CFR 50.49." The first item in the Enclosure refers to paragraph (b)(1) of 10 CFR 50.49 and states:

"The safety design basis of safety systems as described in the FitzPatrick FSAR were revieved along with existing plant emergency operating procedures. Based on this review systems and components required to remain functional in order to mitigate postulated design basis events were identified. The environmental conditions at the specific location of the safety-related equipment was then determined. Safety-related equipment determined to experience postulated harsh environments as a result of these events were included in the 10 CFR 50.49 listing."

NYPA RESPONSE 3.b. This item was addressed in the second paragraph of Enclosure 3 to Reference 4.

That paragraph states:

"A review of the electrical elementary diagrams for the safety-related egipment identified under paragraph (b)(1) was performed. This review confirmed the application of the original plant design criteria for electrical separation of safety-related electrical equipment and circuit coordination / protection schemes.

As a result, no non-safety related electrical equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions were identified."

NYPA RESPONSE 3.c. The operat*.on of the safety-related electrical equipment identified in Item "a" was reviewed to identify any directly mechanically connected auxiliary systems which have electrical components which are necessary for the required operation of the safety-related equipment. As a result, no such equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions was identified.

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NYPA RESPONSE 3.d. The response to this question is the same as A3.b NRC QUESTION 4. Provide confirmation that all design basis events which could potentially result in a L

harsh environment, including flooding

.i outside containment, were addressed in identifying safety-related electrical f equipment within the scope of 10 CFR 50.49 i

(b)(1).

i NYPA RESPONSE 4. This question was answered in Reference 4 which states:

"The Authority has identified all design-basis events which could result in a potentially harsh environment, including flooding due to a high-energy line break (HELB) outside containment."

NRC QUESTION 5. Confirm that the electrical equipment within the scope of 10 CFR 50.49(b)(3) is all R.G. 1.97 Category 1 and 2 equipment or that justification has been provided for any such equipment not included in the environmental qualification program.

NYPA RESPONSE 5. This question was answered in Reference 4, Enclosure 3. The third item in the Enclosure refers to paragraph (b)(3) of 10 CFR 50.49 and states:

" Regulatory Guide 1.97, Rev 2

" Instrumentation ... to Assess Plant and Environments During and Following an Accident" was reviewed as it applies to Boiling Water Reactors (BNR). Instruments were then identified which were presently installed in the FitzPatrick Plant and which meet the required design criteria.

If those instruments required environmental qualification (categories 1 and 2), the associated components were included in the 10 CFR 50.49 component listing."

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dy;- #.

R-fCrrnc'i f l.. PASNY letter , J .P. Bayne to T. A. . Ippolito, ,

dated September 29, 1981 (JPN-81-78).

2. NRC letter, D.B. Vassallo to J.P. Bayne,. dated April 19, 1983.

3. NYPA letter, J.P. Bayne to D.B. Vassallo, dated June 6, 1983 (JPN-83-52).

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4. NYPA letter , J . P . Bayne to D. B. Vassallo, dated June 15, 1984 (JPN-84-36).

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Attachment 2 to JPN-84-67  !

Response to NRC Questions on Equipment Qualification  ;

Dated October 15, 1984  :

i New York Power Authority. .

James A. FitsPatrick Nuclear Power Plant Docket No. 50-333 l l

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o JAMES A. PITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM:. NUCLEAR BOILER (02)

COMPONENT I.D.: 02VMY-71A-L SRV ACOUSTICAL MONITORING 02VME-71A-L SYSTEM The Safety Relief Valve (SRV) Acoustical Monitoring System was installed in January 1981 in response to NUREG-0578. NRC direction provided to the Authority required immediato installation of equipment to be followed by a qualification program.

The acoustical monitoring system has completed a qualification test program which was commissioned by a utility group. During the test program, the equipment was exposed to test conditions more severe (temperature, pressure, radiation) than the maximum postulated accident conditions at JAPNPP. A comparison of the test program results to the JAPNPP plant specific requirements is currently being performed. If required, the JAPNPP valve monitoring system will be modified to bring the installed system into the tested configuration.

In addition, the part of the system which is located in a harsh environment is fully redundant. Each SRV is equipped with two acoustical sensors and ase.ociated preamplifiers. If one sensor channel were to fail, the other sensor channel can be connected to the system cabinet in the Relay Room. In addition, each SRV discharge line is also equipped with temperature sensors which indicate an open SRV. Temperature readouts and alarms are provided in the Control Room.

Based on the redundant equipment used in the system (two acoustical sensors / preamplifiers and temperature sensors) to accomplish the required safety function, and pending the assessment of the test program results, continued operation la justified.

l JAMES A. FITZPATRICK NUCLEAR POWER PLANT l ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: NUCLEAR BOILER VESSEL INSTRUMENT (02-3 )

l COMPONENT I.D.: 02-3AU-278 (A-D) - REACTOR HIGH PRESSURE i ANALOG TRIP UNIT (RPS) l ROSEMOUNT 510 DU Following postulated Reactor Building HELB's, it is unlikely l that reactor trip would be required based on high reactor pressure. If required for this postulated accident, this unit 1 would perform its intended function of providing a trip signal I to the normally energized (fail-safe) protection logic.

Redundant trip units are provided and are located at different I instrument racks locations.

Qualification type test data on file supports qualification for harsh environment parameters for HELB and LOCA with the exception of thermal aging. Harsh environment parameters for temperature, pressure, radiation, and humidity are enveloped by the test data. The aging concern can be partially resolved by the periodic surveillance testing which is performed to verify trip unit operation and calibration.

The function of this component is performed in the initial phases of postulated design basis accidents. The type-testing i noted above provides a high degree of confidence that this component will perform its intended design function. In addition, redundant trip units are located at different instrument rack locations. There is also diverse instrumentation which can initiate a reactor trip during a postulated Reactor Building HELB.

Based on the validity of the partial test data in support of qualification, continued operation is considered justified pending raplacement of these items.

References l l

a. Rosemount Report 3768A " Qualification Test Summary for 1 the Trip / Calibration System Rosemount Model 510DU', dated March 10, 1976.

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JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: NUCLEAR BOILER VESSEL INSTRUMENT (02-3)

COMDONENT I.D.: 02-3PT-178 (A-D) REACTOR HIGH PR89SURE TRANSMITTER (RPS)

ROSEMOUNT 1151GP 4

This pressure transmitter provides a trip signal on high reactor pressure to the Reactor Protection System. The Rosemouat 1151GP transmitter has been fully type tested to IEEE 323-1971 for harsh environment parameters of pressure, radiaticn, temperature, and humidity at more severe levela than experienced in the specific JAP locations. The only outstanding qualification issue is aging of the transmitter's electronica.

This concern can be partially resolved due to the periodic surveillance testing which is performed to verify transmitter operation and calibration.

The function of this component is performed in the initial phases of postulated design basis accidents. It will experience a harsh environment following postulated Reactor Building HELB's. The type testing noted above provides a high degree of confidence that this unit will perform its intended function.

In addition to redundant sensors located at different instrument locations, there is also diverse instrumentation which can initiate a reactor trip for tnis postulated event.

Based on the validity of the partial test data in support of qualification, continued operation is considered justified pending replacement of these itema.

References a) Rosemount, Inc. Report 127227, Rev. B 'Huclear Service Qualification Testing Interim Report, Model 1151DP -

Differential pressure Transmitter" b) Rosemount, Inc. Report 117415, Rov. A " Qualification Tests for Rosemount Pressure Transmitter, Model 1102" c) EDS Assessment Report 0900-001-005 " Evaluation of Class IE Transmitter (Rosemount Model 1151GP)*

d) General Electrical Environment Qualification Record-DV File 145C3240 (Rosemount Model 1151DP)

hl: JAMES A. FITZPATRICK NUCLEAR POWER PLANT.

%- s 1 ENVIRONMENTAL-QUALIFICATION OF' SAFETY-RELATED EQUIPMENT ,

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-+ LJUSTIFICATION FOR CONT 1NUED OPERATION' '

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c SYSTEM:. NUCLEAR BOILER VESSEL., INSTRUMENT (02-3)

COMPONENT I.D.: 202-3 LITS-73' REACTOR-WATER LEVEL INDICATING '

< -SWITCH .

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HThis unit provides a reactor water level permissive ~ signal to

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. .thefRHR System;and provides-indication of water level to a

-Control, Room. indicator. -This unitcis qualified to perform its short term-function:of-providing reactor water level permissive i Lsignal to the-RHR System. However, test data does-not support

.  ; full qualification for providing' reactor > water: level indication over.~an extended post-accident time: frame after a postulated

.HELB in:the Reactor Building.

Type-test data of. identical' equipment at_ elevated temperatures

- and: humidity in' conjunction with a radiation threshold analysis

,o Dof. internal components supports qualification for its Reactor

Building location-following postulated LOCA's inside

_ ' containment. . Type. test data supports.short termfoperation following postulated Reactor' Building HELB's. This is:

considered acceptable -based on other water 11 eve 1' indications

-available in the. Control; Room and the ability to access the

. Reactor 1 Building'for. repairs'within;a short time' frame'following

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a postulated-ReactornBuilding HELB.

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. Based on'the validity of partial) test'-daEa and the availability

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1 of alternate equipment to perform _the safety function, continued _

4 operation is considered. justified.pending replacement of this 2 itemi

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a) E YarwayiReport[No'. 3232-3155

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(Yarway: Report:No.- 5628-3509 1

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' JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF. SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: CONTROL ROD DRIVE (CRD)-(03)

COMPONENT I.D.: 03SOV-117; -118; -140A, B; -31A, B SCRAM AIR PILOT SOLENOID VALVES The solenoid valve performs its safety related function immediately upon initiation of the postulated accident and prior to failure-if any. In addition, the solenoid valve is required to.de-energize in the performance of'its safety related 1 function.,' i.e., the valve is " fail safe". Based on the completion of the. solenoid valve's safety related function prior to significant' exposure to the accident environment, failure of this. device would not cause degradation of any safety function or provide misleading information to the operator. An ongoing qualification program has identified type test data for identical solenoid valves for high temperature and humidity conditions.

Based on the validity of partial test data obtained for the

. solenoid valves and that failure of these items will result in no degradation of any safety function, continued operation is

-considered justified pending finalization of the qualification

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documentation.

References a) General Electric Qualification' Report, NEDC 30602-P, dated April, 1984 " Scram Pilot-Solenoid (ASCU HVA-90-405)

Environmental Qualification Study for NYPA"

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JAMES'A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF-SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION

' SYSTEM: REACTOR CORE' ISOLATION COOLING COMPONENT'I.D.: 13LS-12 BAROMETRIC CONDENSER TANK LEVEL SWITCH This switch performs:its safety related design function in a mild environment. However, the existing system-logic design incorporates common electrical fusing for this item and harsh

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environment electrical equipment requiring qualification. -The

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main concern for this switch.is that it does not lose its-insulation resistance to ground when exposed to harsh-

- environment, de-energizing other safety related. equipment.

In a postulated HELB, isolation of affected lines'is deP.ected and the. isolation signal provided within a matter of seconds.

The. probability:that this switch could develop a significant-ground within this'short. time is extremely small.

Failure'of_this' switch)will result in no significant degradation of any safety function, and misleading information to the operator will not. occur since_the RCIC' system is not. relied upon

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- to mitigate theLsubject design basis event.

Based on this summary, continued operation-is-considered ijustified pending: replacement of this item.

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JAMES A...,FITZPATRICK NUCLEAR POWER. PLANT.

. ENVIRONMENTAL QUALIFICATION.OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION 1

SYSTEM: .CONTAINMENTFSYSTEM (16)

' COMPONENT I.D.: 16-lRTD-107, -108 - DRYWELL AMBIENT TEMPERATURE DETECTOR The materials in the RTD's which may be subject to deterioration due to the harsh environment are the RTD mandrel, the lead wire insulation and the terminal blocks. However,'considering that all materials used in the construction of the RTD's have

. temperature ratings greater'than the maximum postulated accident temperature and that the terminal block is enclosed in a gasketed exlosion proof head, the RTD's can be expected to function after exposure to the harsh environment from a postulated accident.

In addition, there are a large number of thermocouples also measuring drywell ambient temperature, which provides a large measure of redundancy and diversity.

Based'on the ratings of the materials used in the construction of the:RTD's in. conjunction with the large measure of redundancy in the system to accomplish the safety function, continued operation is justified pending item replacement.

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' /- ENVIRONMENTAL QUALIFICATION'OF SAFETY-RELATED EQUIPMENT

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,  : JUSTIFICATION-FOR CONTINUED OPERATION

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' SYSTEM: RADIATION' MONITORING-(17)

( ~_ COMPONENT I .D.': .a.-17RE-50A, B---STACK EXHAUST EFFLUENT MONITOR (LOW RANGE)

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17RE-53A, B STACK EXHAUST EFFLUENT 17RT-53A, B~ MONITOR (HIGH RANGE)

lb.-17RE-431, -432 - TURBINE BLDG. EXH.

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EFFLUENT MONITOR (LOW RANGE) 17RE-434A, B - TURBINE BLDG. EXH.

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3 17RT-434A, B EFFLUENT MONITOR (HIGH 3 RANGE)

c. 17RE-458A, B - RADWASTE BLDG. EXH.

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EFFLUENT MONITOR (LOW

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& ~ 17RE-463A,~B RADWASTE BLDG. EXH.

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. 17RE-463A, B-EFFLUENT MONITOR (HIGH RANGE)

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SThese'. monitors provide for measurement-of1 post-accident plant-

- n , effluents'forurelease1 assessment. This, equipment is-located

-remoteifrom.the; plant areas expe'riencing direct postulated r Se ,

faccident environments.: However, sample streamiradiation' levels

_ - - lcan~resultiin higher:than normal radiation levelsiin the' area of' M ^ '

.this instrumentation... Both.the:highland 1ow range units are 2 specifically designedito: measure.the-required range of

? t ' radioactivity.J The--high range units.are:also shielded to--

gprotect their electronic components:from'the accident stream. .

~ radiati'on..

g . Based:on this. data,1 continued operation ~ is considered justified . .

M, 4 pending' completion of.re-analysis..of the shielding design for

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units. -

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_4F - ENVIRONMENTAL QUALIFICATION-OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR' CONTINUED OPERATION LSYSTEM: RADWASTE~(20)

. L; COMP ONENT1 I .D. :L '20PNS-83, 20PNS-95 DRYWELL FLOOW DRAIN SUMP-OUTBOARD ISOL. : VALVE POSITION SWITCH

-Theiradwaste position switches.are required.to function-short,

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term 1for= indication.of primary-containment: isolation following a cpostulated,LOCA.. Th'e1 valves on which'these switches are located -

ni -fail;safelon lossiof power'. - Therefore, therefcan.be high

, . confidence in1the position of these1 valves without relying on x - .the functioning of_these position switches. The exposure to

,this, accident environment requires the switches to operate in an

elevated. radiation environment. No'significant temperature or

_ pressure increases are postulated for the location of.the Radiation

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. position (of-the. position' switches due to.LOCA event.

. is:the:only-harsh environment accident parameter due to the switches' location in the East Pipe Tunnel. Component materials

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'ofLthe NAMCO: limit switch Lave been identified and qualification

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1 .c documentation located. The. qualification data has been i evaluated perfDOR: guidelines andLby. applying Arrhenius techniques.; Results of this evaluation'. indicate'that the inon-metallic components have greater-tha'n 9 x-103. years of expecte'd: life lat the< maximum pipe tunnel temperature.of 104*F

'l" except for Buna-N'. The,Buna-N components have.an expected life.

.ofigreater than 11 years, i i

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- Additionally, a radiation analysis -performed on: the component

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materia 1s.shows that the radiation threshold for each 7 "non-metallic material is greaterLthan the postulated total.

y lintegrated. dose..

,  : Based on the time-temperature :and radiation analysis performed in~ conjunction with the validity of; partial test data,-continued operation'is: considered _ justified' pending replacement.

References '

-1)-(Masoneilan-International' Report:No. 1003 .

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JAMES A.'FITZPATRICK NUCLEAR POWER PLANT-

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ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT

, s 1 JUSTIFICATION.FOR CONTINUED OPERATION F

SYSTEM: ' JRADWASTE-(20)

COMPONENTTI.D.: l20SOV-83,-20SOV95 DRYWELL PLOOR' DRAIN SUMP OUTBOARD ISOL. VALVE PILOT SOLENOID-

_-! - 1The Radwaste= solenoid _ operated valves are required to function short termJfor primary containment isolation during a. postulated N - ' _LOCA. The exposure to this accident environment requires the a valves 1to operate in.an elevated radiation. environment.. A

.Jadiationlexposure-analysis indicated that the radiation.. _

ohreshold for the device exceeds the postulated _ total integrated-

,. .  ; dose. lio-significant temperature or pressure increases are L ' postulated for the location of the' solenoid valves due to the r LOCA: event.

In addition, the solenoid. valve is required to de-energize in the performance of__its safety related function, i.e.,:the valve

Ls " fail safe" . ' Based on the short term operational requirement c .of'the. valve, and.that failure of-this device would not cause

.. degradation offany_ safety function, continued operation is considered justified pending replacement.

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JAMES A. FITZPATRICK NUCLEAR' POWER PLANT ENVIRONMENTAL' QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: HIGH~ PRESSURE COOLANT INJECTION.

COMPONENT I.D.: 23LS-99 GLAND SEAL COND. HOTWELL LEVEL SW.

23LS-100 GLAND SEAL COND. HOTWELL HIGH LEVEL SW.

This switch _ performs its safety related. design function in a mild environment. However, the existing system logic design incorporates common electrical fusing for this item and other

' harsh environment electrical equipment requiring qualification.

.'The main concern for this switch is that it does not lose its

' insulation resistance to ground when exposed to harsh environment, de-energizing other safety related equipment.

In a postulated HELB, isolation of affected lines is detected and the isolation signal provided within seconds. The probability that this switch could develop a significant ground within this short time-is extremely small.

Failure of this switch will result in no significant degradation of any' safety function and misleading information to the operator will not occur since the HPCI system is not required to mitigate a Reactor Building HELB design, basis event.

Based on this summary, continued operation is considered justified pending replacement of this item.

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JAMES A. FITZPATRICK NUCLEAR-POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: PRIMARY CONTAINMENT' ATMOSPHERE CONTROL / MONITORING (27)

COMPONENT I .D. : 27E/P-103A,B N2 FLOW TO CONTAINMENT ELECTRO-PNEUMATIC CONVERTER FOR 27FCV-103A, B

These insdruments are utilized to control the flow of nitrogen to'the containment following a postulated LOCA inside

-containment. . This~ equipment is not exposed to the direct LOCA environment but to secondary environmental effects in the Reactor Building (radiation, elevated temperature).

-Alternate methods are available'for establishing nitrogen flow to-the containment for venting purposes should this electro-pnuematic converter fail. The alternate methods utilize fully-qualified equipment in the Reactor Building in conjunction with manual control of the nitrogen flow from a mild environment (CAD Building).

Since>the safety function can be accomplished by alternate

. -equipment,. continued operation is considered justified pending replacement.

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. . JAMES A..FITZPATRICK. NUCLEAR POWER PLANT

- ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT.

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~J USTIFICATION FOR-CONTINUED OPERATION .

! SYSTEM:L - PRIMARY CONTAINMENT 4ATMOSPHERN CONTROL / MONITORING

_ J COMPONENT ?I .'D.': 2702-AZ-101A,B AND 27DWA-HTA, HTB -

. PRIMARY CONTAINMENT OXYGEN ANALYZER- ,

AND HEAT TRACING SYSTEM -( BECKMAN,T..YLOR)

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Ifithe containment:02 analyzers were assumed to fail as a. result-

_of postulated environmentaliconditions inithe-Reactor Building

- due.:to a LOCA inside containment _ (radiation, elevated

temperature . 120*F, humidity), the determination to vent the

, m containment of accumulated gases is made_ based on hydrogen concentration.only. .Thisfis allowed by the revised emergency

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procedures based on;the' BWR Emergency Procedure Guidelines.

Thiss is fur'ther-justified'because JAP operates with an-inerted 4:1 containment (nitrogen), therefore,;the only source.of oxygen is

. Esmall'and'due-to radiolytic reaction. The containment is ~

tcompletely. isolated from all: sources of outside air. - In-addition, all instrument piping-uses nitrogen.instead of air for '

4 incrmal; operations..

Furthermore, the containment atmosphere can also be sampled

-ucing the Post-Accident Sample System (PASS), and;the sample analyzed in the laboratory toidetermine oxygen content.

. Based.on'the availability:of~other qualified equipment.(hydrogen analysis):and PASS system, continued plant operation'is-justified pending 02 analyzer replacement.

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JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: PRIMARY CONTAINMENT ATMOSPHERE CONTROL / MONITORING (27)

COMPONENT.I.D.: . 27RTD-101( A-D)- SUPPRESSION POOL TEMPERATURE SENSORS (RTD)

" - These instruments are utilized to provide a secondary indication of suppression pool temperature following postulated design basis. events.. This instrumentation provides redundant indication to the fully-qualified 16 RTD suppression pool

- temperature monitoring system.

Based on-the use of these indicators as a secondary source of torus ~ temperature, continued operation is justified pending replacement.

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i JAMES A. FITZPATRICK NUCLEAR POWER PLANT o-ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION SYSTEM: REACTOR? BUILDING VENTILATION (66)

COMPONENT I.D.: 66UC22A-K FAN MOTORS The Crescent' Area' Unit' cooler motors are Severe Duty Motors mounted within totally enclosed air-over enclosures (TEAO). The motors are designed to operate in a continuous ambient of 150 F with 100% relative humidity. The maximum temperature in the Crescent ' Area af ter _ a postulated LOCA is 110*F and for a HELB a

temperatoreftransient above 150*F for 10 minutes occurs.

However, the motors will not experience these temperatures as

'they are.in-duct' mounted downstream of the cooling coils. The maximum integrated radiation exposure in Crescent Area is 6.9x106 ' R. Testing of similar motors with same class insulation shows no significant degradation of insulation due to

.these levels of radiation.

Based on the validity of partial radiation test data in conjunction with the analysis that shows the equipment will not

.tue exposed.to conditions more severe than its. design ratings, continued operation is justified pending replacement, i

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JAMES A. FITZPATRICK NUCLEAR POWER PLANT

. ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT i JUSTIFICATION FOR CONTINUED OPERATION SYSTEM:- DRYWELL' COOLING SYSTEM - (68 )

COMPONENT I.D.: ~68TE-20li hru'212 t DRYWELL 68TE-301 thru 310 THERMOCOUPLES Component'aterials.of m the' thermocouple have been identified, and a preliminary evaluation of time-temperature and radiation effects performed.-

.- The materials in these thermocouples consist of metal. .

(Cu-Const), ceramic insulators, and a pressed asbestos terminal Lblock with material trade name "Hemit". The ceramic insulators-are aging an'd radiation insensitive and the "Hemit" material which is functional up to 400*C is also listed as aging and

.radiationt insensitive.

Inl addition, there are a total of. twenty-six (26) thermocouples-

sensing drywell air temperature which provides a large measure

-of' redundancy..

-Based on the ratings ~of.the materials.used in the construction of the thermocouples, no-degradation to equipment operation should-occur as a result of the postulated accident condition.

Therefore,1 continued: operation is justified pending replacement.

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. JAMES.A.:FITZPATRICK NUCLEAR POWER PLANT

w.  ! ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT f -

JUSTIFICATION.FOR CONTINUED OPERATION G5 . ,

x SYSTEMil . ELECTRICAL' POWER'-(71) ~

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COMPOWENT L I .D. :- 71ACA5, B5 1120 VAC' DISTRIBUTION PANEL PT-71ACA4, B5-INCLUDING 600/120 VAC TRANSFORMER v

These Lpanels' power two _(2 )' electrical' loads which may be required'to mitigated postulated design basis accidents.

LOAD a .- 1 27NS-CA',' CB Nitrogen Instrument Supply. Cabinet

.b. -71INV-3A,3B LPCI Independent Power Supply Control Power 1

1. ; .27NS-CA,-CB1- These; panels provide 120VAC. power to various insturments and control components associated with the. nitrogen Containment Air Dilution (CAD) , System. This equipment.would'only be required to performiits intended design function -following a - postulated LOCA inside

-primary: containment.. These electrical panels 7

would~not be exposed to the direct accident environment but would be exposed to secondary-environmental effects on elevation 300' of

-the Reactor Building. This accident.

f environment would consist of a mild increase '

in temperature;(110*F maximum), a mild humidity transient, and radiation (3.0x105 s .

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rads).

4. ..

Distribution 1 breakers of similar design have

'been shown by type testing to withstand

radiation doses 'o f 4.4x10 ' rads. . In saddition,-since the electrical circuits are loaded to a maximum of 80% of their trip.

rating by design _(40-104F), operation at a maximum temperature,of;110*F will not trip the breaker..

'2.. 171INV-3A,3B .This load is the control power supply for the LPCI independent power supply charger /

inverter logic. This power source is only

-required for startup of the LPCI charger /

tinverter. Once output voltage is

- established, failure of the external power source to the control logic will will not

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affect inverter operation.

-Based on'.the above reasons, continued operation is considered justified pending relocation of these loads to a distribution panel' in a mild' environment.

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JAMESIA. FITZPATRICK' NUCLEAR POWER. PLANT g.

ENVIRONMENTAL QUALIFICATION'OF SAFETY-RELATED EQUIPMENT

' JUSTIFICATION FOR CONTINUED OPERATION

_ cSYSTEMi- : ELECTRICAL POWER SYSTEM (71)

COMPONENT:I.D.: 71L15, Ll6, SWITCHGEAR-(DISCONNECT, TRANSFORMER,

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BREAKERS) -' GENERAL ELECTRIC AKD-5 The AKD-5 switchgear.has been qualified by-a combined test and i

ana' lysis. program to> accident environmental conditions for pressure,-temperature,Jand humidity-which envelope those

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conditions' postulated for.the JAF plant locations. Thermal agingLand' radiation have been addressed by analysis per DOR Guideline requirements'.-

The G.E.: transformer-with Class,220*C insulation system has been

' stested in a 904 humidity environment.and.has a maximum hot spot temp'raturel e rating:of 220*C (150*C rise). The JAF transformers

~

are loaded to 50% of design rating and the rated nameplate KVA of theseitransformers operate with an 800C average winding frise.

Thus,,the. insulation. winding system'is being operated

. appreciably.-below its. rating. As a result, postulated ambient temperature. increases due to.a HELB (less than 70*C) will;not affect.the' operability of this component. ~ Internal cabinet humidityeis.-lower than the external? cabinet humidity due to the

~

Eheat generated 'in the1 transformer.' ' Thermal aging and radiation

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chavepeenaddressedbyanalysisper~DORGuidelinerequirements.

Based on the validity.of this partial test data in support of qualification, continued operation is justified.

Aeferences

1) Wyle Test Report 17655-SW6-1 dated August 30, 1984 -

-Nuclear Environmental Assessment Report on G.E. SNitchgear Type AKD-5 for Use in-Nine Mile Point Unit 1 Nuclear Power Plant 2)- ' Patel Report PEI-TR-92-4-130- Final Assessment Report on General Electric AKD-5 Switchgear Used-in JAFFNPP 3). . General-Electric NEDE-30303, November, 1983 - JAFNPP Switchgear Environmental Evaluation of AKD-5 Switchgear Equipment-(L15 and Ll6)

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.. JAMES.A;>FITZPATRICK NUCLEAR POWER ~ PLANT

ENVIRONMENTAL QUALIFICATION 10F SAFETY-RELATED "QUIPMENT

JUSTIFICATION FOR CONTINUED OPERATION . ,

" SYSTEM: ELECTRICAL

• SYSTEM (71) l COMPONENT I.'D.: 71INV-3 A,L- 3 B and 71 BAT-3 A', 3B LPCI INDEPENDENT POWER SUPPLY.

CHARGER / INVERTER-Justihication,forcontinued-operationisprovidedbasedon I.-

the;following::

e a

~. iPrimary' Containment LOCA (large) - This equipment is

, located remote from.the direct harsh environment of this: accident (and w'ould perform its intended design

' function;of.-providing power to the LPCI valve bus

' prior to theElocal; temperature, radiation, or_ humidity F -

lsignificantly-exceeding. normal conditions. The required operating time is-less than 3 minutes for operation of. reactor' recirculation loop isolation valves and~ opening ofEthe LPCI injectionLvalves.

Eb.- Primary ContainmentLLOCA (small and intermediate) -

This equipment is located remote from the direct

accident environment.. Although the required operating

.timeLis significantly longer for this accident -(6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> maximum), no significant accident radiation exposure is expectedidue to-its_ elevation in the' Reactor Building and minimal fuel damage that is

. postulated-for this accident. The long term temperature ~does not.significantly increase above normal,(110*F).

c. Reactor Building'HELB A method of plant depressurization and cooldown following a postulated HPCI or RCIC-steam line break is described in

'NEDO-24297, Revision 1 ("High-Energy.Line Break Evaluation forithe James A. FitzPatrick Nuclear Power Plant" - dated October, 1980), Section 6.2.2. This method of plant cooldown requires manual

-depressurization'of the_ reactor using the Automatic Depressurization System (ADS)-while restoring and maintaing water level using one of the two Core Spray Pumps. Based on this analysis, there would be some core heat-up. however, there would be considerable margin to the 10CFR50, Appendix K limit of 2200'F peak clad temperature (PCT).

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, JAMES [A. FITZPATRICK NUCLEAR POWER PLANT

= ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION

-SYSTEM: ' ELECTRICAL SYSTEM (71)'

' COMPONENT I .D. : 71INV-3 A , 3 B and 71 BAT-3 A , 3B LPCI INDEPENDENT POWER SUPPLY  ;

CHARGER / INVERTER-Following a RWCU line break, HPCI and RCIC Systems

' located in the Crescent Area experience an insignificant change in environmental conditions (5'F

. rise for less than 30 seconds, and a 0.5 psig pressure

. rise 1for less than 30 seconds). Therefore, these systems will remain functional to provide high pressure' cooling. Following depressurization using RCIC or HPCI,_ reactor inventory can be maintained

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using one of two Core Spray Pumps. Refer to NEDO-24297, Revision 1 (Section 6.2.4).

'II . In addition, post-HELB temperature / pressure analyses are extremely conservative.

III. Based on.the:above analysis, continued operation is considered justified pending completion of installation of environmental enclosures aroung this equipment.

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JAMES A._FITZPATRICK NUCLEAR POWER PLANT.

ENVIRO 3 MENTAL QUALIFICATION OP SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION

! SYSTEM: ELECTRICAL' POWER SYSTEM (71)

COM?ONENT I.D.: AC AND'DC MOTOR CONTROL CENTERS (MCC) -

(MCC-151, -152 -153, -155, -161, -162,

-163, -165 AND BMCC-1,-2, -3, -4 AND -6 GENERAL ELECTRIC 7700 SERIES MOTOR CONTROL CENTERS-

' Qualification t'est data applicable to these General Electric Motor Control Centers-(MCC's) has been identified and a plant specific qualification report developed which meets DOR Guideline requirements. -This report is being finalized.

Reviews of drafts of this report confirms that the specific JAF -

MCC equipment is qualified to the postulated design basis event and environments for the-specific equipment locations. This qualification program. consisted of the followi.,g phases:

1. Preliminary assessment of applicable G.E. MCC test data resulted in the identification of environmental qualification testing applicable to the JAVNPP MCC's and the postulated harsh environments.

'2. Subsequently, a Phase I engineering report (G.E.

NEDC-30322-P) was prepared for fifteen specific MCC components installed in the JAF G.E. 7700 Series Motor Control Centers. This report provided'further substantiation of the qualification of the subject MCC's to the JAF normal and accident environmental conditions.

3. A final qualification _ test report (G.E. NEDC-30694-P) for the JAP Motor Control Centers was completed by General Electric in draft form in August, 1984, and has been reviewed by the Authority, and its qualification consultant, and an independent reviewer. Comments from this review are presently being resolved or incorporated.

Final issuance'of'this: report is_ scheduled for November, 1984 Based on the valid test data obtained and evaluations performed, continued operation is justified pending qualification report completion.

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.. . JAMES?A.'FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION .<.

OF'. SAFETY-RELATED' EQUIPMENT

, . JUSTIFICATION-FOR CONTINUED' OPERATION

. " SYSTEM: ' ' MISCELLANEOUS'

~

1. COMPONENT I.DJ: ' -CURTIS (TB) TERMINAL BLOCK IN A

' ' GASKETED STEEL JUNCTION BOX-2

~ IComponent? materials of the Curtis terminal blocks have been

. identified and qualification documentationion'similar equipment

. located., TheLphysical design of these blocks and. materials are

!similar to.the1gualifiedLGeneral Electric; terminal' blocks. The.

5 iCurtis '_ blocks are .one: piece, imolded,12 point (30A)'manuf actured ofigeneral. purpose blac,k, phenolic.- They are designed for.a continuous; rating of 250*F and:have a.thre'shold for radiation j

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damage greater thancl.0 x 106 ,R. '

TheLqual'ification documentation of the General Electric terminal oblocks shows that the. equipment; performed successfully under j test: conditions'(temperature, pressure, and radiation).more  !

Esevere thancthe postulated accident' conditions at-JAFNPP. These l

-terminal blocks are located outside of-containment at JAFNPP.

1 Based on'the. validity.off. test data on similar equip'ent

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-(materialsiand-design) " continued operation!is.justif,ied. . .

l References

" Environmental Qualificationjof: 'TerminaliBlocks/ Boxes

.1) '

'-EB-25", Report 50-213 c

-

2 ) . Limitorque Report BO119

,y_ -- 3 ) Westinghouse. Report 1 PEN-TR-77-83, " Test Report on the

~

- Effect of a'LOCA on the Electrical Performance.of Four Terminal Blocks"

'4 )-. " Rad'iation Hardness o'f_Termimal Blocks",-Westinghouse' Memo No. 76-lCC-QUhEQ-M24 ,

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~ : JAMES AT FITZPATRICK NUCLEAR ' POWER PLANT ;

/ : z BNVIRONMENTAL.~ QUALIFICATION,OF SAFETY-RELATED' EQUIPMENT 4

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'in JUSTIFICATION FOR CONTINUED OPERATION- '

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LSYSTEM: MISCELLANEOUS

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-'COMPONENTII .Dli' : CINCH JONES (TB)-TERMINAL BLOCK IN A

. GASKETED STEEL JUNCTION BOX i%

'N Component' materials ofjth'e Cinch Jones: terminal blocks have been

+

identified and. qualification: documentation'on similar equipment f located. 'The  : physical design 1of these blocks and materials are

. Esimilar"to' the qualified General. Electric terminal blocks. The LCinch' Jones. blocks are'one piece, molded,112. point.(30A)

' manufactured'of: general; purpose black phenolic. - They.are M.,. designed'for a continuous rating of 250*F and have a threshold t a "- .for' radiation damage. greater than 1.0 x 106 R.

M The4 qualification documentation of the General' Electric ter'minal

• -  :-blocks shows thatithe_ equipment performed.successfully under test 1conditionsf(temperature, pressure, and radiation) more W :se, vere-than,the postulated accident conditions at JAFNPP.- These

terminal blocks are located outside of-containment at JAFNPP.

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,m -Based on the-validity.of test date on similar equipment

, ' : /+ ; (materials and design), continued operation is justified.

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. References

1) " Environmental Qualification of Terminal' Blocks / Boxes f-EB-25, Report 50-213 s.[ .

L2)' Limitorque-Report.B0119 H, . .

E 3). Westinghouse Report PEN-TR-77-83, " Test Report on the F Ta[$ ' '

Effect of a LOCA on the Electrical Performance of Four Terminal Blocks"-

4) " Radiation Hardness of Terminal Blocks", Westinghouse Memo No. 76-lCC-QUAEQ-M24-4

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JAMES A.-FITZPATRICK NUCLEAR POWER PLANT ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT JUSTIFICATION FOR CONTINUED OPERATION V

SYSTEM: MISCELLANEOUS L COMPONENT I.D.: STATES (TB) TERMINAL BLOCK IN A GASKETED STEEL JUNCTION BOX Component materials of the States terminal blocks have been identified and partial qualification documentation located. The materials have~been evaluated per the DOR guidelines and by applying Arrhenius techniques. Results of the evaluation indicate that the_ lowest expected life of the terminal blocks is 266 years at the maximum Reactor Building temperature, 104*F.

The qualification documentation shows that the terminal blocks have been successfully irradiated to a level of 2.2 x 108 rads gamma.

Though qualification documentation was not located for States terminal blocks showing testing to envelope the postulated peak i temperature and pressure requirements, there is documentation to demonstrate the qualification of similar terminal blocks. The-t physical design of these blocks and materials are similar to qualified General Electric terminal blocks.

The qualification documentation on the General Electric terminal l blocks shows that the equipment performed successfully under l test conditions more severe than the maximum postulated accident conditions at JAFNPP.

!' Based on the validity of partial' test data on the States terminal block and the successful performance of similar I

terminal blocks under. test conditions more severe than the postulated accident conditions at JAFNPP, continued operation is justified.

t References

( 'l) " Test Report on States Terminal Blocks and Test Switches", Report No. 15809-82, Rev. 2 I '2) " Environmental Qualification of Terminal Blocks / Boxes

-EB-25", Report 50-213 r

3) Limitorque Report B0119 h

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