ML20092P674

Forwards Programs,Plans & Schedules Re Generic Ltr 83-28, Generic Implications of Salem ATWS Events. Procedure for post-trip Evaluation Encl

ML20092P674
Person / Time
Site:	FitzPatrick
Issue date:	06/29/1984
From:	Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Vassallo D Office of Nuclear Reactor Regulation
Shared Package
ML20092P675	List: ML20092P674 ML20092P677
References
GL-83-28, JPN-84-42, NUDOCS 8407090243
Download: ML20092P674 (23)

Text

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123 Main Street Whte P'sns NewWrk 1(M01 914 681.6200 -

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4# Authority June 29, 1984 JPN-84-42 Director of Nuclear Reactor Regulation U.b. Nuclear Regulatory Commission Washington, D.C.

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 Generic Letter No. 83-28 Required Actions Based on Generic Implications of Salem ATWS Events

References:

1.

NRC Generic Letter No. 83-28, D.

G.

Eisenhut to All Licensees, dated July 8, 1983.

2.

NYPA letter, J.

P. Bayne to D. G.

Eisenhut, dated September 6, 1983 (JPN-83-80) regarding same subject.

3.

NYPA letter, J. P.

Bayne to D.

B. Vassallo, dated November 9, 1983 (JPN-83-92) regarding same subject.

Dear Sir:

Reference 1 requested information,-plans and schedules related to the generic implications of the February, 1983 Salem ATWS events.

Reference 2 provided the Authority's preliminary schedule for submitting the requested information.

1

-m l\\.

9 In Reference 3, the Authority responded to the extent practical ano coramitted to further supplement our response.

Attached is our final response to Generic Letter No. 83-28.

It aescribes those programs anc procedures currently in effect at FitzPatrick.

It also cescribes our plans and schedules for iteras not aadressed by current programs.

Solae or the information included in the attachment duplicates prior submittals.

This was done to consolidate all elements of our response in a single document.

It you have any questions or require additional information, please contact !!r.

J.

A.

Gtay, Jr. of my staff.

Very truly yours, 2.

J.

P.

Bayne Executive Vice President 14uclear Generation State ot 14ew York County of Westchester Subscrioeu and Sworn to before me this Jf day of 1984.

JEANN! 1A LUNA NCTARY PUSLIC, STATE OF NEW YORK NO. 644614303 CUALIFIED IN WESTCHESTER TY TERM EXPIRES MARCH 30th 19.

110tary Public cc:

Ottice or the Resident Inspector U.S.

Nuclear Regulatory Commission Lycoming, New York 13093-

my NEW YORK PtAIER AUTHORITY dataes A.-FitzPatrick 14uclear Power Plant (JAFNPP)

,.ae s:

RESPONSE TO GENERIC LETTER 83-28 Attachlaent 1 to JPN-64-42 1

Tnis uoculaent:oescribes the current procedures and programs for each itela laentitlea in NRC Generic Letter 83-28. A description of the Authority's position or plans is also provided for those items not currently accressec in proceaures or programs. An estimated colapletion cate is also provided, where sufficient information exists to pro]ect a cate.

Response to Itela 1.1 - Post Trip Review (Program Description and k

Fruceaure) l]

The JAFDPp post trip review program and procedure is contained in Operations Department Stanalng Order (ODSO) 23 titled " Post Trip Review." 'it:Is ODSU accreses Items 1.1.1, 1.1.2, 1.1.4, 1.1.5 and 1.1.6 ot Generic Letter 83-26.

A copy of ODSO 23 is included as Appena1x A.

1.1.3 The qualitications and training of the personnel responsible for colapleting 0060 23 is oescribea below:

Shirt Technical Advisor (STA)

Tne STA (or the Irx21vidual fulfilling the STA role) has been trained and is qualitied to the requirements of NUREG 0737.

'l i This incluoes training in the analysis and determination of the causes of oft-normal conditions such as mitigating core

caraa9e.

operations Superintendent and Shift Supervisors Tue Operations Superintendent and Shift Supervisors are qua11ried in accordance with the requirements of ANSI 18.1 (1971) with respect to training and experience and hold Senior. Reactor Operator (SRO) licenses.

In addition, trainire is currently being conoucted to upgrade the Shift Supervisor's level of training to meet or.. exceed the " Draft

(

Coralssion kolicy tor Engineering Expertise on shif t" (Federal p

Register Volume 4b, No. 143, July.25, 1983).

G 1

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Other Shirt Personnel utner sultt personnel are, as a mininum, qualified in m

accorcance with the requirements of ANSI 18.1 (1971) with fi

, respect.to training arxt experience.

In addition, a number of h.

these personnel hola Reactor Operator (RO) or SRO licenses.

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! Response to Itera 1.2 - Post-Trip ReYiew, Data and Information.

h Lopeu1111.y i.

1. 4.1 '

l Capability tor Assessing Sequence of Events

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i 1.2.1.1 The sequence of events function is provided by the l-

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plant process cor.gsuter's Sequence of Events log.

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The Sequence at Events log is initiated (printed) l Dy a change in state of any of 116 digital (on-off) inputs.

1.2.1.2 Currently 161 puints are available of which 45 are k.

cesignated as spares. 'Ihe 116 points in use at this tilae ore listed below:

TABLE 1 Process Computer Points Currently In-Use it POINT ID DESCRIPTION D000 RFP A SUCTION PRESSURE DOUl RFP B SUCTION PRESSURE DOO2 RFP A DISCHARGE PRESS D003 RFP B DISCHARGE PRESS D013 115KV SOUTH BUSS UV RLY D014 115KV NORTH BUSS UV RLY Duld llbKV BRKR 10012 OPEN D0lb 115KV BRKR 10012 CLOSED

[

D017 ll5KV BRKR 10022 OPEN l

Dul8 115KV BRKR 10022 CIDSED D019 345KV BRKR 10042 OPEN I

DO20 345KV BRKR 10042 CIDSED DU21 BREAKER FAILURE 10012 DO22 BREAKER FAILURE 10022 DU23 BREAKER FAILURE 10042 DO24 BREAKER FAILURE 10052 DU25 34bKV NMPT LINE fl0 P

DO26 345KV EDIC LINE #1 D027 345 BUSS v

DO26 ll5KV I4HT HSE HLL LINE

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DU29 11bKV LINE 'IO NINE MILE DO30 345KV DRKR 10052 OPEN DO31 RESERVE STATION TFR T2 UV D032 RESERVE STATION TFR T3 UV lL D033 345KV BRKR 10052 CIOSED l

DO34 MOIST SEP A III LVL TRP D035 tDIST SEP B HI LVL TRP bi D036 SOllIC DETECIOR RV-2-71A D037 SONIC DETEC10R RV-2-71B D036 SONIC DETECIOR RV-2-71C D039 SCtlIC DETECIOR RV-2-71D DO40 SONIC DETECIOR RV-2-71E I

D042 SOtlIC DETEClOR RV-2-71F f

D041 SONIC DETECIOR RV-2-71G

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DO42

, SONIC DETECIOR RV-2-71H D043 SONIC DETECIOR RV-2-71J DU44 SONIC DETEClOR RV-2-71K j,

D045 SOllIC DETECIOR RV-2-71L Dub 2

'lUN EHC PANEL 24 VDC PWR 0054 MAIN 'IURBINE TRIP t:

0053

'IUNE BACK UP OVERSPD TRIP s

3 '

t r

l D056 TBt!E IDSS OF 125 VDC TRP Dub 7 7URBINE MANUAL TRIP D058 TBNE EXII HOOD HI T TRIP Dub 9 IG1 cot 3D VACUUli A TRIP Dubu IIM TIr1E BRG OIL PRESS DU62 TBNE TRIP HI VIBRATION D063 TBNE TRIP-IOSS STAT CLNT

}

Dub 4 TBNE TRP-THRST BRG HEAR D065 TBNE TRP-SHAFT PMP PRES I

Dubb TBN TRP-EMERG TRP FLUID f

D067 1UN TRP-HYD FLUID PRESS

)

D0bb TBN TRP-IDSS SPEED FDBK

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D500 SDIV Al W LEVEL SW SCRAM I

Dbul SDIV B1 W LEVEL SW SCRN4 f

DSO2 SDIV A2 W MILG TRP SCRAM

[-

DbO3 SDIV B2 W ANIO TRP SCRAM N

D504 MAIN STEAM LINE CHNL Al Dbub 11AIN STEAM LINE CER1L B1

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Dbu6

!!AIN STEAM LINE CIR1L A2 DbO7

!!AIN STEA!! LINE CHNL B2 i;

D506 COrnt!T HIGH PRESS CH Al 8

Dbu9 CONTNf HIGH PRESS CH Bl D510 CONT!!T HIGH PRESS CH A2 g

Db11 wtJT!ff HIGH PRESS CH B2

(.

Dbl 2 REACIOR CIR1L Al HI PRESS Dbl 3 REACIOR CHNL B1 HI PRESS Dbl 4 REACIOR CIR1L B2 HI PRESS I,

Dbib REACIOR ID iffR LVL CH Al Dbl 7 REACIOR ID WfR LVL CH B1 D516 REACTOR ID UfR LVL CH A2 D$19 REACIOR LO 1(fR LVL CH B2 Db20 PSL A-1 HIGH RADIATIOt1

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Db21 P.SL B-1 HIGH RADIATION l

Db22 11SL A-2 HIGH RADATION D523 MSL B-2 HIGH RADIATION Db24 NEUT M0t1 SYSTEM CHNL Al i

Db25 11EUT MON SYSTEti CIR1L A2

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D526 11EUT mot 1 SYSTEM CER1L B1 Ip Db27 11EUT 10N SYSTEM CIR1L B2 L

Db26 SDIV Al E LEVEL SW SCRAM Db29 SDIV B1 E LEVEL SW SCRAM Db30 11ANUAL SCRAM CHAIR 1EL A Db31 MANUAL SCRAM CHMRJEL B DbJ2 REACTER SCRAM CHAIRIEL A i

Db33 REACTOR SCRAM CHANNEL B Db34 BCfrH SCRAM CHAIR 1ELS ASB Db3b SDIV A2 E ANLG TRP SCRA!!

i-Db30 SDIV B2 E ANIL TRP SCRA!!

Db3o TSV FAST CLOSURE CIR1L Al Db39 TSV FAST CIDSURE CHNL B1 Db40 TSV FAST CIOSURE CIR1L A2 Db41 TSV FAST CIDSURE CHNL B2 Db42 1Y.N FAST CIOSURE CIRIL Al Db43 TCV FAST CLOSURE CHNI B1 F

Db44 7CV FAST CIOSURE CIR1L A2 A

$xp A h

F d

V Db45 TGV FAST CIOSURE CHI 1L B2

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Dd4b AVRll CHNL A UPSCALE LVL l

D547 APR!l CHNL B UPSCALE LVL

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Db46 -

APRM CHNL C UPSCALE LVL

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D549 APRM CHNL D UPSCALE LVL 3

D550 APRM CHNL E UPSCALE LVL j

Db51 APRM CHNL F UPSCALE LVL

1. 2..t. J T1:ae alscrimination between events is one millisecond; (i.e.,

ror two events occurring within less than one millisecond of eacu other, their sequence cannot be determined.)

~

1.2.1.4 The Sequence of Events log is printed on a printer in the control ruou. A facsimile of an actual printout of data point Db32 (which was. generated as a result of part of a routine survelliance test) is shown below:

TIME STA'IUS (nour, (on-off Iainutes trip-reset ano seconos) 111LLISECUNDS EOINT ID DESCRIPTIOtl etc.)

T 093247 640 SEQ D532 REAC'IOR SCRAM CHANNEL A TRIP 0932bu 7A9 SEu Db32 REAC'IOR SCRAM CHANNEL A RSET A.2.1.5 The process computer does not retain (store) Sequence of Events uata except as part of its providing data printout.

T-Once the data has been printed, it is no longer available witnin the cor.puter. The Sequence of Events printout "hara-copy" is retained as part of the post-trip review recorus.

ht 1.4.1.0 twer for the process conputer and Sequence of Events printer is suppliea trom the Uninterruptable Power Supply (UPS) systeu. The UPS is described in FSAR Section 8.9.

I 1.2.4 Capaullity ot Assessing the Time History of Analog Variables j.

t Equigient useu to assess the time history of analog variables I~

i consists of the plant process couputer's Post Trip Img l-runction anu a number or strip chart recorders.

(See Table 2 I

anu Appenuix B).

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1.2.2.1 Briet Description ot Equipment lost Trip Log

t F

'nie Post Trip Log contains 20 selected plant inputs

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t' and is automatically initiated upon occurrence of precetined plant tr.ips.

It'can also be initiated on operator deiand.

i Strip Chart Recorders t,

A numoer ot strip chart recorders are available to fL the operator _to assist in assessing the time I

history ot analog variables.

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'1.2.2.2 Parameters Monitored, Sanple Rate Er Selection Basis i

_4_

T t

_.__.______________________________.______,_._i___.______m_____

Table 2

'l POST TRIP IOG POI!7fS BU32 APRM A FLUX LEVEL

%MJR BU33 april B FLUX LEVEL

% Pia B034 APRM C FLUX LEVEL

%PWR BU35 APRM D FLUX LEVEL

%Ptn B036 APRM E FLUX LEVEL

%BJR BU37 APRM F FLUX LEVEL

%PWR BU441UfAL CORE FIdi M)/HR BU45 CURE DiFFEREllTIAL PRESS E

bu47 FDi/IR LOOP A FIoW M6/HR BU46 FDiffR 100P B FIDi 11#/HR Bub 3 REACTER WATER LEVEL Il1CH bub 4 '11/fAL STEAM FIUJ

!!#/HR bub 7 REAC10R PRESSURE PSIGN L

bub 2 RX FW INLET Al TEMP DEGP F204 MAItJ STEM HEADER PRESS MU17 DRYuhLL PRESS (ABSOLUTE) 11019 '1UR \\(fR LVL (-72/+72)

INCH MU2O Mt \\/fR-AT (14tM-IJff=95) 20036 TB BYPASS VLV IOSITION %

'10040 htbIlJE SPEED RPf1 Post Trip Loy cata is sanpled at 2 second intervals by the process L

col.puter, delection or pararaeters and sar.ple rate were based on the recorrenaations of the NSSS vendor and limited by the insta11eu esulpent. Eacn of the 20 parameters raonitored is m

Aintiteu to 120 data points. The 2 second sample rate is consluereu optuaura ror the equipi.ent currently installed, f

btrip Cnart Recorcers Strip cndrt recorcer data is recorded continously. Chart paper speea is normally one (1) inch per hour and each chart is

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9enerally cate/tir.e star. ped daily tor reterence. Neutron i

1.onitoring recorcers (07-PR-46 A,B,C,&D, 07-R-45) and the reactor water level recoruer (06-LH/PR-97) may.also be

{y operated with a paper speed of one (1) inch per minute. This p

reuture is only used during plant startup and scheduled shutdown.

3 Appenulx B lists the strip chart recorders (and the associated p.

Forolaetets) udeu to ueterraine the cause et unscheduled reactor shutoowns. Tne parameters listed in' Appendix B are used to

. col.plete the Udta recordire for the post trip review procedure

{p (Appencix A).

1.2.4.4 Duration or T11ae History Post Trip Loy, i

Post Trip Log data is continuously stored and.

f; upaated at 2 second intervals in.a portion of the:

1 process cor.puter iaet.ory and remains in the memory ror 2 nanutes. Thus, raemory contains the raost i

f r

-S-E.

+,

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recent 60 cata bits for each Post Trip Log parameter prior to a plant trip.

Upon occurrence of a pre-selected plant trip condition the plant

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process computer program prints the Post Trip Ing.

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The Post Trip Log contains data for 2 minutes prior to the trip and data for 2 minutes after the trip

(

at 2 second intervals totaling 120 data entries for l

each ot the 20 Post Trip Ing data points.

Strip Cnart Recorcers

]I As previously described, strip chart recorders provide a continuous record of the neasured parameters value before and y

atter a trip.

0 1.2.2.4 Data Format lost Trip Log Post Trip Log format is shown below:

f n 1D n ID Pt ID 1

2...........

20 Tne ME ME y

1 2 ***********

20 J

1XXXXX 1XXXXX 1XXXXX c

4 T11.e VALUE VALUE VALUE l20 y

2...........

20 t

1XXXXX 1XXXXX 1XXXXX

.l L'

s strip chart Recorders strie cuart recotuer format is typical of most strip. chart recorders useu in the inaustry today, (i.e.,,a strip chart approximately 5 inches

[-

wiue with 11nes ano numerals indicating the scale, with two color pen traces on the chart.)

1. 2. 2. :)

Data Retention Capability i

I Post Trip Log As noted in the response to Item 1.2.1.5, the cas.puter does not retain (store) data used for i

tunctions (such as Post Trip Log) except as'part of its data printout. Once this data is printed, it is T

e P _

b

f.

no longer available within the coi:puter. The Post Trip Log printout is retained as part of the post trip review records.

Strip Chart Recordings Strip chart recordings are also retained as part of the plant Records Retention Program discussed above.

1.2.2.6 Power Source

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Fo6t Trip Log Power tor the process cor.puter and printer for the Post Trip Log is from the Uninterruptable Ibwer supply (UPS) system which is described in FSAR Section 6.9 Strip Chart Recorders Akpenoix B lists the power source for each strip chart recorder that may be used in assessing the time history ot analog variables used to determine the cause of an unscheduled reactor shutdown.

j 1.2.3 lh auultion to the Sequence or Events printout, Post Trip Ing printout and strip chart recordings listed in response to 1teias 1.2.1 and 1.2.2, the following information is generally available to assist in determining the cause of an,fuscheouled reactor shutdown.

~

1.

Position indication for containment isolation valves

~

and nulaerous other valves associated with Emergency Core Cooling Systems (ECCS), Reactor Water Recirculation (RNR), Main Steam, Reactor Water Cleanup EWC), Reactor Core Isolation Cooling 1

(RCIC) and balance or plant systems such as f

Feedwater, Condensate, Service Water and Emergency Diesel Generators.

t; 2.

Irmication of system flow, system pressure, motor

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arperes, voltage, generator voltage, amperes and treguency, turbine speed (RPM) and similar indication of systera initiation, operation or trip _ as i.

appropriate tor the systeu(s) of concern.

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J.

Annunciators in the Control Room and/or at local h

panels for plant system (s).

y i-4.

Protective relaying targets (flags) indicating relay

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operation and/or transuission system and, off-site p.

power oscillograph.-

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b.

Protective _ systela and ECCS logic indicators, b.

Process computer alarm printout, periodic 1ogs, special lo9s and viceo displays.

- 7~-

b a

E_

7.

Statements relating to individual personnel actions, involvement or observations. Such statements are L

recorded as part of the critique which is conducted I

as part of post-event evaluation if the event I

involved any ot the following:

h

- A coLplex evolution

- Obvious personnel error i

ECCS actuation Scrau or main steam isolation

- Event cause or effects are not immediately evident ihe event report alone does not completely oescribe the event

%i These inoicators (1 through 7 above) provide supplemental sources of information. the primary sources of y

intormation used to assess the cause of an unscheduled shutuown are the Sequence of Events og, Post Trip Log and strip chart recorcings listed in response to Items 1.2.1 ana 1.2.2.

1. 4. 4e beneaule tor Planned Changes lhe Authority has contracted for the design, purchase and installation of a new coLputer system which will include a Satety Parameter Display System (SPDS) including certain oisplays associated with Emergency Operating Procedures ano the eventual replacement of the existing process coLputer. This coLputer may also include some dis satisty the requirements of Regulatory Guide 1.97. plays to Detallea specitications and a description of the system as it relates to Regulatory Guide 1.97, SPDS and Emergency Operating Procedures Love not been finalized.

ihe completion date for the new system will be provided in accorcance witu our 140 REG 0737 Supp.1 commitments. The Authority considers the currently installed Sequence of Events Loy, Post Trip Log and strip chart recorders to be j"'

auequate for post-trip review in the interim.

t<esponse to iteu z.1 - tsulpuent classitication and Vendor Interface tueactor Trle systeu conponents)

Tue Louponent quality Assurance (QA) Category List in existence at the daues A. FitzPatrick liuclear Power Plant has been reviewed. She review incluueu vertrication that all coLponents in the Reactor Protection Systeu (RPS) (System b) are presently classified as QA Cate90ry I except for tne RPS 110 tor Generator Sets which are classitieu as @ Category II.

COLponents in the RPS System are protecteu trou botor Generator Set malfunctions, such as over-voltage, unuer-voltage, anu under-treguency conditions, by electrical protection asseublies which are classified Category I.

ihe m Category Classitication of other systems, such as Reactor Vessel Instruuentalon (Systeu 02-3), lieutron Monitoring (System 07)-.

'.f m

1

+

. anu Process naulation Monitoring (System 17) which corprise part of ji

. the " Reactor Trip Function *, has tren reviewed. All or part of these j )

Q steus are classitieu as QA Category I, indicating that those f

portions ot the systems which are associated with the " Reactor Trip Function" are properly classified. Reactor trip function components are classitled QA Category I.

Tue Auttiority has performed a review of the documents, procedures and I.

intorlaation handling systems used in the plant to control L

satety-relateu activites, including maintenance work requests (work L

oruers), parts replacement and plant modifications. The documents, proceuures anu intoruation handling systems concerned are controlled unuer the LA'Piogram, or are identified as safety related and require

' review by the Plant Operations Review Comittee. This provides i

assurance that'maintensiice, parta replacement and modification work is properly classitieu as (A Category I when required.

Venuor Intertuce - A tormal Operating Experience Review Program is in errect anu includes review of, and response to, General Electric BUR Service Intornation Letters (SILs). BWR SILS are used to document recometaMu cuanges in equipnent anc procedures, as well as convey Intonation concerning unique operating conditions and experiences at i

bwa plants. The review anu ihplementation of SILs is recorded and fed p

back to the General Electric Company (GE) using a standardized SIL p

btatus Response form.

Periou1cally, a SIL Index is issued, assuring r

that all applicable inrormation has been received. The General

[

Electric Company SIL program, in con] unction with the Operating Experience Reviev Program, assures that complete, current and contro11eu ushs venuor_.intormation is appropriately referenced or incorporateu into plant specitic procedures. Accordingly, the program h

ensures that reactor trip system venaor information is controlled

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turoughout the life ot the plant. No changes to the program are planneu.

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in uuultion, General' Electric lias established a reporting system to hanule sorety concerns that corplies with the requirements of

{-

10 CFx 21.

General Ele'etric also has established several other Intotuation systems which are described below.

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Ur ent Col.gauni, cations A procecure for handling urgent communications

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s to u w owner / operators has been established by GE for use in providing

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rost notitication or satety concerns. These communications are f

usually in the toru ot a short letter which provides a brief explanation ano.auvice'or precautionary measures to be o!Ocev to p

avolu potential operational hazaras. Dee to their urg'.d av 3,

tuese cot.saunications are sent to operating plants by la m etrective metnou',(i.e., telex, telecopy, cable, specia; mail uu.1dling, f

-ecc.).

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.y t

dervice auvice Letters - These.docutaents are issued by GE product

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.uepartuents otner than the San Jose based Nuclear Energy Product t

Departuents anu arc useo to provide notification of product problems anu/or service intormation on a broad. range of GE consumer and

~inaustrial grouucts.~ Those' Service'Adv ae Letters that are recognized.

- i-uy.the issutry product department as! applying to devices used in t

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- nuclear plants are specially icentitied and are flagged for alstrioution to all nuclear plants.

Turbitte Intor1.ation Letters (TILS) - TILs are issued by GE's Large bteala Turuine Generator Departnent to provide descriptions of product i

probielas or 11aprovements ano to reco:::aend modifications that will

(

miti ate probleias or iraprove product performance.

9 Operation ana Maintenance Manuals

'1hese documents are issued by all GE proauct uepartlaents to provide instructions for installation, l

operation anu 1.aintenance et GE designed repairable equipment and

[

systelas.

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Applicat.lon internation Docuraents - These documents describe potential t

operating proulems ano provice aesign change or operating recol.naenuations to Iaitigate or avoid thera. These documents are pr11.ar11y aimed at requisition plants, but are also forwarded to operating plants when they have any applicability to those plants.

L if Urgent Connaunications, Service Advice Letters, TILS and/or Application,

l Inrori.ation Docut.ents, are piocessed upon receipt using the existing L

Operat1129 Experience Review Program.

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'112e Operatiss Experience Review Progrant also provides for the

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systeiaatic review of inaustry operating experience documents such as NxC lhE bulletins, I4RC Isb Inforlaation Notices, Licensee Event Reports, inaustry newsletters, INPO Significant Event Reports (SERs),

INPU 619niticant Operating Experience Reports (SOERs) and the intormation receivea on the INPO Nuclear Network (formerly NOTEPAD).

Further the Authority actively participated in the INPO sponsored L

Nuclear Ut11ity Task Action Contaittee (NUTAC) which was formed to auoress Section 2.2.2 ot A neric Letter 63-28. Wnile the Vendor Equipment Technical Intormation Program (VETIP) discussed in the NUTAC repott specitically auclesseo itaprovement of non-NSSS vendor Intoriaation excnange, it is anticipated that much of the VETIP

[

Intori.iation W111 be ulrectly applicable to NSSS supplied equipment (incluaing Reactor Trip Systera Components). It is anticipated that intortation receivea as a' result or the VETIP will also be processed utilizing the existing operating Experience Review Program.

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Plant btanuity Oroer (PbO) No. 26 entitled " Operating Experience FeeubacK", luentitles responsibilities for review, feedback and incorporation or operating experience inforraation into training prograsas. PbO 26 rurther. requires audits by the Quality Assurance I

uepartixnt. Tt.e NxC start has received PSO 28 and found it to meet l'

the!requirelaents ot Itela I.C.5 of NUREG 0737 (NRC May 21, 1982 letter,

'D.u. Vussallo to L.W. Sinclair.)

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Response to iteu 2.2 - Equipment.Classitication and Vendor Interface LPIvytala tor all berety-Relateo Conponents)

[

E 2.2.1.1 burin 9 the original classification of components at JAFNPP,-

the criteria tor iaentitying components as safety-related witnin system classified as safety-related was as follows:

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QA Category I is uetinea as those plant systems, or portions ot systems, structures, and equipment whose

-railure 'ot Laitunction would cause a release of rauicactivity that would endanger public safety. This category also incluaes equipment which is vital to a I

sate snutoown of the plant and the removal of decay and

[

sensible heat, or equipent which is necessary to nitigate consequences to the public of a postulated

[

accident.

l'

'Abis cerinition was interpreted to Lean those structures, systems, ano conponents that:

t

. Are necessary to assure the integrity of the reactor

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coolant pressure boundary.

. Are necessary to assure the capability to shutdown the i

reactor and maintain it in a safe shutdown condition.

. Are necessary to assure the capability to prevent or

_i mitigate the consequences of accidents which could g

result in potential off-site exposures corparable to the.

T guiceline exposures of 10 CFR 100.

. Contain or r.ay contain radioactive material and whose tailure would result in conservatively calculated t

potential ott-site coses which are more than 0.5 rem to the whole body or its equivalent to any part -of the body.

b Tuis sal,e criteria is presently used tor identifying components as j

sarety'relateo anu are currently included in the Engineering Design i

Proceuures in use at JAFl4PP.

I, 2.4.1.4 The JAF14PP presently has a Corponent Quality Assurance i

Category List which identities the safety-related conponents within sarety-related systems. This list is issued and

- contro11eo by the 'JAFI4PP Quality Assurance Department. The list was'oeveloped by a consultant.-~The consultant had a statt resident at JAFI4PP and used the criteria given in response to 2.2.1.1 above.

In audition to the stated criteria, the consultant used the following data:

JAFI4PP Final' Satety Analysis Report- (FSAR)

Plant drawings proviaed by the-Architect / Engineer (A/E).

Systeu cescriptions proviaed by the A/E Instrur.ent -lists provided by the A/E -

_Tecnnical manuals provided by the NSSS vendor-Venaor Lanuals and instructions which were provided by-the equipent -venoors The consultant's statt pertormed walk-throughs'of the plant-and~

-veritieu; Installation, nate plate cata, ratings,-and other -

-intornation, as part of the oeve}oment of the list. : Typical entries on the Alst ate as rollows:-.

[TyYe'

-Catewory.(technical)

Cor@onent' Description

. COI+onent 14uuber t.l 4

11

y e

T' Data Reterence xeuarks Wality Assurance Category Tue 11sts were coupiled and cross checked by the consultant. They were then transmitted to the Site Quality Assurance Department for l

review and concurrence. lieubers of the Quality Assurance staff l

reviewed the lists for completeness and accuracy.

Comments were returneo to the consultants, as necessary, and when all comments were j

resolvea, the wality Assurance Department concurred with the lists.

I

. Revisions anu/or changes to the list are controlled in accordance with approved plant procedures.

Slince this upcate was coupleteo, many plant uodifications have been insta11eu involving new equipment. Sorae of these raodifications affect tue w1&onent Wality Assurance Category List. Since only minor revisions ano additions have been made to the list since that update i

was coupleteu, new components installed since 1978 raay not appear on tue 11st. As described in Sections 2.2.1.3.A and 2.2.1.3.B, Quality Assurance persolusel review plant uoairication records to determine the,

category.

I A prograra to rurther 11. prove anu upcate the Component Quality i

Assurance Category List is describeu in Section 2.2.1.6.

(

t 4.2.1.a.

A contro11eu copy ot the Component Quality Assurance

[

category List is issuea to the Superintendent of each major plant uepartuent requiring the inforraation. The Quality Assurance Category

[

or plant couponents is thus readily available to personnel requiring

[

the intor!.idtlon.

I t_

Elant Acuinistrative Procedures require that the Plant

)

Operations Review Committee (PORC) review Administrative Proceuutes una other cocuments which atfect nuclear plant satety, or it. pact on the environment. Procedures requiring L

6 RC review are icentitled by an asterisk (*) after the title, thereby alerting personnel as to which procedures involve sarety related considerations. The following h

procecures are in etfect at JAFlJPP which describe the e

Y controis and requirel.ents which apply to_ safety-related

{.

activities.

l' Auministrative Procedures V

Work Activity Control Procedures

{

n Rules or Practice f

Wality Assurance Program n;

w ality Assurance Procecures i.

In ~auultion, the tollowing departraents raaintain controlled 1-ceporti. ental proceoures which govern the conduct of safety-related work.

,l' Operations q

f'

- Instrur.ent.6 Control Radiolo9 cal & Environraental Services i

Iep 6

b,_

7 11aintenance Tecnnical bervices (uality Assurance Traitsing l

Furtner, tne tollowire controls are useo at JAFNPP.

A.

A work nequest Event Dericiency (WRED) form must be completed to 1111tiote corrective 1.utritenance. The WRED is routed through the guu11ty Control Department which reviews and verifies the Quality Assuratice Category ot the involved component. The URED is also udrhed to itioicate it Quality Control Inspection is required.

In J

y auultion, ror corrective uaintenance performed on safety-related i

(Category I) cot.ponents, or other work requiring QC inspection, I

the use or a Work Tracking Form (WfF) is required. A UfF is used to properly preplan, track, control and document corrective i

naintenance anu provices sign-otts for the department peforming l

[

the activity, guality-Centrol personnel and the Operations Depar tLent.

I a

b.

The procureuent ut Itaterials is initiated by a Purchase Requisition.

Purchase Requisitions must be routed to Quality Control wnich verities the Quality Assurance Category of the I.aterial, cenotes if QC Receiving Inspection is required, and specirles the requirea documentation, test reports, etc.,

i which must be included on the requisition. The requisition 18 then routea to guality Assurance which checks and verifies i

the QC intormation, includes any further requirements to be

[

luposeo on the vencor and uenotes the method used to qualify i

tne verxior. When the Purchase Order (P.O.) for Category I Material is typed, the P.O. (with a copy of the requisition) is routed to the Quality Assurance Department to verify that the requitet.ents or the requisition have been correctly

[

entereo on the Purchase Order, y

C.

Tne rollowire occuments are required to have a Quality Assurance review and sign-off prior to implementation:

Aaulnistrative Procedures Ergineering Design Procedures 140ultication Control ForIr/3 tiocitication Documentation Tracking Forms h

11oultication QA/gG and Design Requirenents fiodification Installation Procedures Preoperdtional Tests and Test Results Procurement Specifications 2.2.1.4 Tne 1.wayer.ent controls to verity that procedures for preparation, validation and I;outine use of the information nonu11ty systeu have beyn tollowed, are as follows:

A.

The Plant Operation Review Committee reviews plant proceuures anu enanges thereto, proposed tests and experiments, g:

C

{ L l

~_.

P ano prog >seu changes or nodification to plant systems, that I

a

atteet-nuclear satety.

O n

ti B.

Mie plant uepartr.ents utilize an internal departmental review.

d n

U C.

The wality Assurance Department implements an audit program p

to provice a comprehensive, independent evaluation of quality

[

relatea procedures and activities to assure that they are in colpliance with the Authority's establishea program

(

requireuents.

D.

Mle LA 6 it Depart!.ent, under the direction of the safety i

Review Coimlittee (SRC) Chairman and the Executive Vice President-Nuclear Generation, coordinates efforts to schedule an INFO or a Joint Utility Management Audit Group audit.

(Tue Authority is a participant in a group of utilities for the purposes ot pertorming independent assessments of QA activities.) The scope ot the INFO or Joint Utility Manager.ent Aucit Group includes, as a minimum, the activities.

Perror!.eu Dy the Authority's QA & R Department.

In addition,,

areas outside at this scope may be assigned by the SRC Chairien or the Executive Vice President-Nuclear Generation.

The total audit program covers the 18 criteria of Appendix B to 10 CFR 50, within a 24 month period.

2.2.1.6 During I & E II:spection 84-11 and a visit by NRC heacquarters personnel tram June 18, 1984 to aune 22, 1964, NRC personnel reviewed the Equipment Classitication Program at JAFNPP. As a result of this review, NRC personnel noted the following:

The exiting Component wality Assurance Category list contains inconsistencies.

r The existing Couponent Quality Assurance Category

[.

list is oitticult to use.

i.

The Equipt.ent Classification Program does not

" trigger" generic reviews or updates of similar cor.ponents when a cos:ponent is identified as

}.

'L(

requiring review or reclassification.-

t Tne nutnority agrees with these tindings and cornits to review the

[

Ooi.ponent Quality Assurance Category List.for completeness and-accuracy. Consiuerin9 the resource requirements,-the desirability to cuange the toruat to enhance ease ot.use, and the advisability of

[

inte9tatity the proyrum with other activities (cuch as preventive

(;

maintenance, corrective maintenance, parts procurement, and plant 1.ouitications), tue Authority expects.to conplete this work by_

beces.aser J1, 1965.

Venoor interrace r

2.2.4 L'

As noteu in response to section 2.1, a formal Operating i

Experience Review program is currently in effect. " Ibis prograu is baseu on the recor.Dendations ot the ~ Institute of Nuclear Power Operatins (IN10) and provides for formal

[.[

uocui.entea review ot operating experience documents from both i

1

- Authority 111 tertial arid external sources including IIRC, IllPO at:0 - the NSSS venuor. The JAFIJPP is also an active putticipalit in tue tauclear Plant Reliability Data System j

(bPRDS). 14PRDS is an inoustrywide system managed by INPO for I.aonitorin9 the pertornatice of selected syteus and components at nuclear power plants. IUPO's Signiticant Event Evaluation anc Intoruution tietwork (SEE-IN) attenyt to ensure that the i

culaulative learning process from operating and maintenance experience is ettective and that lessons learned are reported i

ano corrective action taken in a timely manner to improve 0

plant satety, reliauility arx1 ovailability.

[

n Tne lua; sponsored 11UTAC which was formed to plan a program to address k

bection 2.2.2. nas col.pleteu its work. copies of the NUTAC report k

were receivec by the participating utilities in early April, 1984. As

[

part or this pro 9 tau, the 14UTAC uade several recora.lendations that f

woulu enhance and inprove both the NPRDS and SEE-IN.

Entlancements recommenaeu by hUTAC tor NPRDS include:

Expand cetinition of components to better describe mechanical '

col.ponents.

I-f II+rovea tailure reporting guidance in areas of: analyzing

[

role ot piece parts in tailures; inadequate vendor

{

intortation as a tallure cause; and inproved failure analysia reports.

Develop 1tsternal methocs to assure clear and complete llPRDS reports.

Proviue tollow-up NPRDS. reports, ki I

Auc1 tion or new systems and corponents to program scope.

f bilal Aarly, the bt:,E-1N prograu expansion recounendations include:

I Reports preparea tor potential tailures due to Caulty or missing venoor intorLation or other equipment technical

{.

Intorudtion.

^.;

'?

broduen pro 9tata to il. prove ability to trend HPRDS data.

1; dince both programs are aamir:i.stered by INFO, the Authority alone L

carunot assure that eacn.or these enhancenents will be fully il+1el.entea. Furthermore, the IMPO staff responsible for executing these pto9tuus Ley suggest alternate means for enhancing NPRDS and dEE-IN.

p-uoseu upon our pteluaibary review ot t'he report, the ' Authority endorse uuTAC's alternate approach tor implementing the guidance of Section 2.2.1. or beneric Letter 63-26.

The Authority plans to request IMPO' to incorporate NUTAC's recola.erxL tions.

t f

rv 4pp Vo U g.

=

I Solae or 14ttihC's reconraencations will requite that new or revised

_ proceuures ar.u ptogralas be prepared und iupler.cuted by the Authority.

. Prelitainary chaises, Dased upon the available guidance, will be tj 11aplelaenteu by October 31, 1964 We expect that turther changes may

}

oe necessary as the enhanced programs evolve.

f 7

Respolise to 3.1 - lost 11aintenance Testing (Reactor Trip System

(

wl.nyonents I

J.1.1 he Authority has reviewed JAFtJPP test and maintenance procecures and Technical Specifications to assure that post-Daintenance operabilitity testing of safety-related F

components in the reactor trip system is required. Work

[

Activity Control Procedure (UACP) 10.1.1 and Operations

(-

Departuent Standing Order 16 require post maintenance L

testing ano provide guidance in the conduct of such testing to assure that the equipment is capable of

[

pertoruing its safety tunctions before being returned to J

service.

h In auc1 tion, we have been reviewing our methods for

)

}'

oetermining the cegree and extent of post maintenance testirN. As a result ot this review and a recent flRC g

Inspection (140.64-11 conducted June 16-22, 1984),

written guiuance tor post raintenance testing will be prepared ano inplemented by September 30,19o4. This i

new 9ulcance will traprove the consistency and thoroughness with which post maintenance testing is conoucted.

J.1.4 We Authority will tormalize the Vendor Technical !!anual Controls system and library at JAFilPP. his system will collect venuor equipment technical information (technical manuals, instructions, service advice etc.)

in an organizeu, easily retrievable and controlled tashion. Access to this library will be controlled by k'

c laeans et a aucunent " check-out" procedure. This library i

will be established by December 1, 198..

{

Atter tuis progran has been it.plemented and the library tormalized, we will initiate a review of vendor tecnnical laanuals (or instructions) to verify that approprate vendor and engineering recommendations are incorporated in test and uaintenance procedures, and

[

technical specitications. Completion of this review, y

revision or test and maintenance procedures is expected f

to be completed by December 1, 1965.

j J.1.J A review ot Technical Specia.1 cations has been conducted i

-to cetermine it any post naintenance test requirements

_[

uegraue sarety. As a result et this review, the Authority is not aware of any post' maintenance requirei.ents in the Tecluiical Specifications which ueyrace satety. _ he Authority notes, however, that lA%G octivities to inprove Technical Specifications may at sosae tuture date ioentity reconsnendated changes to tests required by Technical Specifications. These i g o

e changes taay luvolve' surveillance treguency, allowable

[

p

out-ot-service. intervals or recommended post maintenance

{

tests, based on probabilistic risk assessment techniques which consiaer all or some of the considerations noted k

in Section 4.b.3.

Response to 3.2 - Post Maintenance Testing (All Other Safety-Related coa.ponents y

3.4.1 The review ot test and Maintenance procedures and Technical Specifications discussed in the response to 3.1.1 1s applicable to all safety-related equipment.

3.4.2

'1he tormal review of vendor technical manuals and the revision or proceuures and/or Technical Specifications uiscusseu in response to 3.1.2 is applicable to all satety-relateu equipent.

3.2.3 The review ot Technical Specifications discussed in response to 3.1.3 is applicable to all safety-related equlgent. The IMHOG activities discussed are also applicable to all sarety-related equipment.

~

4.1 Reactor Trip'bystem Reliability (Vendor-Related Modifications) 4.2 Reactor Trip bystela xeliability (Preventative Maintenance and burvelliance Prograta tor ueactor Trip Breakers) 4.J xeuctor Trip Systela Reliability ( Automatic Actuation of Shunt Trip Attacni.ent ror westirshouse anu Bhw Plants)

~

/C 4.4 xeactor Trip bysteu Reliability (Iraarovements In Maintenance and Test Proceuures tor uhw Plants)

{j i'

Response to 4.1, 4.2, 4.J anu 4.4 r

Tne out.es A. FitzPatrick uuclear Power Plant is a boiling water i

reactor uesigned by General Electric.. Therefore, Items 4.1, 4.2,_

h 4.a anu 4.4 are not applicable.

Response to 4.5 - Her.ctor Trip System Reliability (System Functional

.h Testines g_ 's 4.b.1 neuctor Trip Systeu reliability is demonstrated by

![

cot.pletion ot the tests and calibrations required by

[.

Tecnnical bpecitication Table-4.1-1 in. conjunction with Tecnnical Specitication required control rod scram time testity wuica periouically cemonstrates function and 2

reliabllity'ot the entitie Reactor Trip System. Backup scrau. valves are apt required to be' tested by Technical bpecitications ano the system design does not permit.

on-line tunctional test. The: Authority does not believe

.that any signiticant inprovement in system reliability s

woulu be achieveu it the system were modified to. permit.

(

on-line tunctional test and such testing was t

periouically pectormed.

F e

'r.

a r

.- -w--

' ~ - ^ ' ' ' ^ ^ ~ ~ ^ ^ ~ ~ ~ ~ ^ ^ ^ ~ ~ ~ ^ ~

Y

=

4.5.2 As an-alternative to on-line test of backup scram I

valves,. the Authority proposes to iraplement tunctional

,Ij testing or the backup scram valves once each refueling

[

cycle willie the plant is shutcown as part of the JAFi1PP i

test program. The Authority notes the IIRC found

[

acceptable sirailar testire of backup scram valves in 140 REG 0979, April 1963, entitled " Safety Evaluation Report relateu to the Final Design Approval of the GESSAtt II, WR-6 Iluclear Island Design."

h i

4.b.3 Tue Authority has pertorued a review of Technical Specitications to cetermine if the test intervals specirleu are consistent with achieving high reliability.

'1he review did not consider:

Personnel errors during testing Reduced redundancy during testing, or Uncertainty in coimon raode tailure rates

\\

a In general, our review did not reveal excessive testing L

(as inuicateu by excessive component " wear-out") or intreguent testing (as indicated by high failure rates.)

'ihe Authority in a January 16, 1984 letter (JPti-84-01, g

J.p. Bayne to D.B. Vassallo) comented on the effects of ulesel generator cola tast starts. While no forced outages have resultec from ciesel generator testing, the lore-teru it. pact on reliability and availability are known to be negative. We have not quantified the degree or these negative etfects.

L

'1he Authority is participating in the developent of a IAet wners' Group (WROG) Technical Specification II.proveraents program. 'Ihis program will include the -

uevelopent or a raethod-for the review of intervals for on-line runctional testing required by Technical i

Spec 1tications. A generic riethodology will be developed to show the sensitivity of system unavilability to changes in:

L

' Cora nent railure rates Cor.. on raode tailure rates j

Reduceu reuuncancy curing testirs j

.- Huuan error rates during testing, and Coraponent " wear-out" rates caused by testing.

'1he. Authority plans to apply the. results of this program

.to JAF11PP.

'1he scheuule ter this BWHOG ac'tivity is currently being -

preparea by the Technical Specitication Inprovements

- L Col.nalttee. - Accordingly, the Authorit schedule tor-this work at this time. y cannot provide a 1

x

-- w m

~.

3

1; s

1 f

J k

II 3

a 1

i NEW YWK KMER AUIRORITY 4

daI*s A. FitzPatrick-Nuclear Ibwer Plant bl

.3

. (,

L

.c l

b Appendix A to J114 4 2 i :.

U Operations Departraent Standing Order No.23 L

lost Trip Review ih L,

i L

f' i

n 6

e-f._ :

r E..--..

A

In?trument ho.

Parameter kecorder Range Power Source i

'07-PR-46A, b, (Red) AERM B, D,

& F or i

C&D IhM b, D, F,

&G 0 to 125 Note 6

{

(Black) APRM A, C,

& E or

{'

IRM A, C, E,

&H 0 to 125 Note 6 i

07-R-45 (Red) SR, B or D 0.1 to 106 CPS Note 6 f

(Black) SRM A or C 0.1 to 106 CPS Note 6 Notes for Appendix B

[

r 1.

120V AC Safeguard Control and Ins +tument Bus B1 (71-ESS-B1).

Power is k

derived from the "B" Emergency I

.r 4 KV Bus (Bus 10600). See FSAR Figure 6. 9-1.

[

2.

120V AC Safeguard Control and Instrument Bus Al (71-ESS-Al).

Powe,r is

?

derived from the "A" Emergency Power Bus (Bus 10500). See FSAR Figure L.9-1.

a 3.

120V AC Common Control and Instrument Bus 9 ( 71-AC-9).

Power is i

derived f rom eith the "A" or "B" Normal Power Bus (Bus 10300). See FSAR Figure b.9-1.

[l 4

120V AC haergency Control and Instrument Bus A2 (71-AC-A2).

Power is derived f rom the "A" Emergency Power 4 KV Bus (Bus 10500). See FSAR Figure 6.9-1.

I; 5.

120V AC Emergency Control and Instrument Bus B2 (71-AC-B2).

Power is I

derived f rom the "B" Emergency Power 4 KV Bus (Bus 10600). See FSAR

[

Figure b.9-1.

~

L b.

120V AC Uninterruptable Power Supply (UFS).

The UPS motor generator is driven by either AC or DC power. AC power is derived f rom the "B"

[

Emergency lower 4 KV Bus (Bus 10600) and tha DC power is f rom the "A"

{

125V DC Station Battery. Maintenance power (used when the motor generator is out ot service) is derived f rom the "A" Emergency Power 4 l

KV bus (bus 10500). See FSAR Figure 8.9-1.

I h

f p

a f:

k

+

e Y

~

I

(

e t

?i s

c

~*

NEW YORK POWER AUTilORITY L

JImus A. FitzPatrick Nuclear Power Flant r

Appendix B to JPN 4 2 f'

Strip Chart hcorders i

I-Instrument No.

Parameter hcorder Rango Ibwe r Source f:

Ob-Lh/PR-97 (Eed) Reactor Pressure 0 to 1200 psig Note 6 f

(Black) Nactor Water tevel 164. 5 to 224. 5 inches above TAF Note 6 I

t t

06-FR-96 (Red) Feactor Steam Flow 0 to 12x106 lbs/hr.

Note 6

{.

(Black) Feedwater Flow 0 to 12x106 lbs/hr.

Note 6 g.

c I

Ob-kR/FR-96 (Nd) Nactor Pressure 800 to 1100 poig Note 6 (Black) Turbar.a Steam Flow 0 to 10x106 lbs/hr.

Note'6

{

02-3-LR-9b hactor hater Level

-100 to +200 Inches (Fuel Zone) below (~) or above

(+) TAF Note 1

{-

f lu-FR-143 (Ped) LPCI Loop A Flow 0 to 25x103 gpm Note 2 (Black) LCPI Loop B Flow 0 to 25x103 Note 1 gpm I

02-FR-163A&D (hd) Pacirc. Loop A Flow 0 to 70x103 gpm Note 3 (Black) Recirc. Loop D Flow 0 to 70x103 Note 3 gpm 27-PR-ll5Al&A2 (N d) Primary Contain, p

Pre ssure O to 250 poig Note 4 (Green) Irimary Contain.

j Prussure

-5 tp +5 psig Note 4 t

t i

27-kR-115bl&D2 ( hed ) Primary Contain.

Pre ssure O to 250 psig Note 5 (Green) Priraary Contain.

Fressure

-5 to +5 psig Note 5 f

I 02-3-FH/th-Sb (Ied) Loro Diffurential Pressure O to 25 paid Note 6 f

lbs/hr.

Note 6 f

(Black) Core Flow 0 to 90x106 I

t 6-PR-blA&D (Nd) Reactor Urcosure O to 1500 psig Note 4

[

(Green) hactor Prussuru 0 to 1500 psig Note 5

[

v 23-Lh-202A &

(Nd) Suppression kool 203A level

1. 7 to 2 7. 5 f oo t Note 4 (Green) Drywell Level.

22 to 106 feet Note 4 1

25-LR-202D &

(Rod) Suppression Pool 203b Levul

1. 7 to 27. 5 feet Note 5 (Gree n) Drywell Level 22 to 106 feet Note 5

+,

b y

.