ML20196L123

Proposed Tech Specs Pages,Revising TS to Include Oscillation Power Range Monitor (OPRM) Sys

ML20196L123
Person / Time
Site:	Hope Creek
Issue date:	05/17/1999
From:	Public Service Enterprise Group
To:
Shared Package
ML20196L121	List: LR-N990137, Application for Amend to License NPF-57 to Revise TS to Include Oscillation Power Range Monitor (OPRM) Sys,In Preparation for Plans to Fully Enable OPRM Trip Function During Ninth Refueling Outage,Scheduled for Apr 22,2000 ML20196L123
References
NUDOCS 9905260196
Download: ML20196L123 (22)

Text

i Document Control Desk LR-N990137 Attachment 3 LCR H99-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)

TECHNICAL SPECIFICATION PAGES WITH PROP _0 SED CHANGES The following Technical Specifications for Facility Operating Licensu No. NPF-57 are affected by this change request:

Technical Specificai!on_ Pag _e Existing Specifications 3.4.1.1 3/4 4-1 3.4.1.1 Action b. 3/4 4-2 3.4.1.1 Action c. 3/4 4-2 3.4.1.1 Action d 3/4 4-2a 4.4.1.1.1.c 3/4 4-2a 4.4.1.1.1.d 3/4 4-2a 4.4.1.1.4 3/4 4-2b Figure 3.4.1.1.1 3/4 4-3 Bases 3/4.4.1 B 3/4 4-1 Bases 3/4.4.1 B3/44-2 6.9.1.9 6-20, 6-21 ,

New Specifications 3/4.3.10 3/4 3-108 (New Page)

Bases 3/4.3.10 B3/4.3-9 (New Page)

Bases 3/4.3.10 B3/4.3-10 (New Page)

Bases 3/4.3.10 B3/4.3-11 (New Page)

Bases 3/4.3.10 B3/4.3-12 (New Page)

Bases 3/4.3.10 B3/4.3-13 (New Page) l l

1 f

9905260196 99051734 DR ADOCK 0

3/4.4 . REACTOR C001 ANT $YsTEM i

3/4.4.1 RECIRCULATION SYSTEM RECIRC'JLATION LCCPS LIMITING CCNDITION FOA CPERATICN 3'4.1.1 wit i Two reactor coolant system sectreulation loops shall be in operation ffi k.

To L core flow (reater th or equal o 45% of ytted core f' w, or EN4AL 3.4.1.1-L. PCWER ess than o equal to he limit specified in Agure

/ _

APPLICA5ILITY:, CPERATICNAL CONDITICNS l' and 2'.

ACTICur I

4.  !

With one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

• as Place the rectreulation flow control system ta the Local Manual mode, and be Reduce THERMAL PCWER to s 70% of RATED THERMAL PCWER, anc I c1 Increase the MINIMUM CRITICAL POWER RATIO (MCPRISafety Limit per Specification 2.1.2, and [

I d) Reduce the Maximum Average Planar Linear Heat Generation Rate iMAPLNGR) limit to a value of 0.86 times the two recirculation loop limit per specification 3.2.1, and et OELETED.

f)

Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and g) Perform surveillance requirement 4.4.1.1.2 if THERMAL POWER is s 389 of RATED THERMAL POWER or the rectreulation loop flow in the operating loop is s 50% of rated loop flow.. 4 .

2. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the Average Power Range Monitor lAPRM) scram Trip setpoints and Allowatte values to those applicable for single recirculation loop operation per Specifications 2.2.L and 3.2.2: otherwise, wtch the Trip setpoints and Allowable Values associated with one trip system not reduced to those applicatie for single recirculation loop operation, place the affected trip system in the tripped condition and wichtn the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, reduce the Trip setpoints and Allowae;e Values of the affected channels to chose applicatie for single rectreulation loop operation per specifications 2.2.1 and 3.2.2.

3. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the APRN Control Red Stock Trip

'see special Test Exception 3.10.4. ,

MCPE CREEE 3/s 4.t Amenament No.107

REACTI3 CEIDL&ET SYSTEM ACTION (Centinued) -

%) points and Allowable Values to those applicable for single tec.irculation loop operation per specifications 3.2.2 and 3 5 6; otherwise, with the Trip setpoint and Allowable Values associated with one trip function not reduced to those applicable for single recirculation loop operation, place at least one affacted channel in the tripped condi, tion and within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, ' reduce the Trip setpoints and Allowable values of the affected channels to those applicable for single recirculation loop operation per specifications 3.2.2 and 3.3.6.

4. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the Red Block Monitor Trip setpoints and Allowable values to those applicable for single recirculation loop operation per specification 3.3.6; otherwise, with the Trip setpoints and Allowable values aseeciated with one trip

• function not reduced to those applicable for single recirculation loop operation, place at least one affectos channel in the tripped condition and within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, reduce the Trip setpoints and Allowable Values of t,he remaining channels to those applicable for single recirculation loop operation per 6pw ification 3.3.6.

5. The provisions of specification 3.0.4 are not applicable. l

6. Otherwise he in at least 30T SEUTDcWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l

b. _W ith no reacter coolant system restreula**a= 1- in operation,  !

Cd 2 "X'7 X 'a ",e'a' Tie N L 'M* W X M ;"X l initiate measures to pleos the unit in at least sTARTUP within 6 - -

-r. and in S- .ioin o. n t . -,s.

- - ~ __

c. Wit er -e coolant recirculat loops operatter. total core il ese than 45% t greater an 404' l of rated low and pcWER gree han the 1 specified Figure 3.4.1 -1

1. ermine the and LPRN 1evels (Surveill 4.4.1.1.4):

least ones 4 homes, a) b) within 30 tes after t letion of ~

T ERMhL increase e at least 5% TED THE j

t

2. With AFRK or LpRN neutron fl Lee levele I

gr er than three ' s their abliehod ine note i evels A and C one LPRK ring per os actant F ,

Detesto det es A and C of LPRK str in the coat of the re e 1d be /

)

men tored. _

Amendment Mc ,3 l EOFB CREEK 3/4 4 2

r_

I RFm s

MFf105 (Oostiaued) 1- 1.

- 15 ms ..se La te serrective to restere the Lee levels to with he required lima withia 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

' reasing core flew a greater than el of rated core [

r by reducing POWER to less t er equal to t 'liatt epeelfied in Fi .4.1.1-1.

d. ith een er two er seelaat roeirculaties 1 La operaties and al core flew I them er equal 40% and ,

TEMhL FOWR l ter them the 1  % specified la pure 3.4.1 71-1, withia 18 as initiate ser ive asties to se THERN&C POWER to less t er equal to the imit spoeified Figure 3.,4 4f f.1-1 or  !

ineresse flow to than 404 wi 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

'/

ssRveru.Ascs Racermamwes

)l )

4.4.1.1.1 With ese roaster seelaat system restreulaties leep act La operation at least sees per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> verify that

a. Reester TEBANAL POWER is s 70% of RATED TERNAL 70NRR, and

h. The restreulaties fisw sentrel system is is the Lesal Manual mode, and -

1

e. The speed of the operatinf treulatidh pump is less than or equal l to 90% of rated peep e -

d.

limit flew is ter than Led La F when TEERMhL

.4.1.1-1.

is gree than the

/

(

l 4.4.1.1.2 With see remeter seelaat system restrestaties leep met La j operaties, withis as more than 15 siestes prior to either TEREMhL POWER l incrosse or roeiresisties Loop flow incrosse, verify that the following '

differential taptuse requireamste are met if TM 90MR is s 38% of [  !

RhT5 TM pm er the restreulatien keep flew La the operating  !

roeire.t a== 1eep is a 50% of rated leap flows l

a. d 145'F betuses remeter vessel eteme spese emelaat and bottom head i drata ties scolant, and

h. 5 50'F hetuses the reester ecolant withis the leap met La eyeration and the sosiast La the remeter pressere vessel, and

e. d 50'F hotuosa the remeter eestaat withis the leap met in operation and the operating Leop. -

The differentiai temperature requiremente er spesifisettees 4.4.1.1.2h and 4.4.1.1.Se de est apply whos 'the loop est is operaties is testated free the remeter pressure vessel.

I Amenement 38e.63 l MDB M 3/4 4-Sa o

anacrea enes hart arq souvezLLance magatangeres (onettened) .

.........a 4.4.1.1.3 demometrated 0p334353Baet puey MB est oseep tube mechanical and electrical with everspeed setpointe lose them er equal to 109% and 107%, respectively, of rated eere flow, at least sees per is meethe.

4.4.1.1.4 Retah a haeolies -

I the regione f and LNES flus asies value w thin which eeni Le required (

within 3 e of ester regies for whi i: sties 2.4.1.1, ce c) base 1La has prov heen perfemmet La taring te requi unless refoe regies stase the astage.

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t#PE MIR 3/4 4 3 Anneent W. 3 April 7,1987 S * ,

e lla.A BRACIDE.. CDDIAlff SY9TERI thMS y - . . . . . . . .

2/4.4.1 RacIncirt.attamm The impact of stagle roeirculatien leep eparaties upea plant safety is aseeeeed and shows that stagle leep operaties le posinitted if the IIcyt fast eladdiaq Safety Limit Le increased as ested by Speettisation 2.1.2, APIst seram and control red bleek seapointe are adjusted as ested La fattes 2 2.1-1 and 3.3.4 2 respoetisely. alhPLEAR limite are deeressed by the fester gives La spostficatios 3.2.1, and ascrt operating limite are adsusted per Speettiestion 3/4.2.3.

hedittaamily, serveillasse en the pump speed of the speseting restrostaties loop is imposed to emelede the poseihuity of emessolve some intereale vibraties. The servet11asse en differential temperstases below 38%

TWegmL peutR or 50% rated restreulaties leep flee to to attigets the undee thessel strees en vessel seestes, roeirculating pung and veneel besten heed during the esteeded operaties of the siegte reeirestation leap mode. ,

An inoperable jet pump is set la itself a estfisiest reassa to declare a roeirestaties loop inoperable, but it does, is ease of a design =

haste-eesideas, inermee the ble= deus area and restsee the espahiuty of .

refleeding the sere, thus, the requiremost for ohntdeus of the fasuity with a ses pas , Laeyerable. Jet pump iausse saa he detested by maatterlag set pump performones en a prescribed schedmie for significant degradation.

Resireulation loop flew einestok 1&aLte ese La esep11ames with the acca IJIIE analysis design eriteria for tus resironisties leap opesemies. The 11atte will ensure as adequate sees flow assetdeus from either roeiresisties keep fouewise a IAch. In the ease where the sienstat limite seemet he maintaiand euring tuo loop operaties, eestimmed speestaan La pommitted La a single roeirculatkan leep made.

In order to protest undee strees em the vessel messies and betten head region, the restremistian leap temperatures shall he withia 50*F et seek other prior to startop of en idle toep. The loop tempeestese meet alas he within le'F of tin remeter pressure vesent ecolant temperatese to psevent thessal shook to the roeironistian pump and resta==tamtam sessaas. Sudden egna1Leattom et a temperatere differesse > 148'F hetuosa the seester veseel Instten head eestaat and the ecclaet La the upper segies ed the reester vessel i by teoreasing esee flew rete weald essee undes strees La the seester vesest

• • • • • • " W p, .tah,.

.he em vs .f e.d inel L. to ever the escala. c sa tigt operaties th easy power / fles se of * . J. Leg , a annu -', 4- 2'Jihy limit eparati

. --Seen '--- l eye mentres i eeen esiste that I

tiens (e ., red , peuer 1 1 _- '

ase pod , arget

/esetzen n limiti sooinet tag la e Lppet fles se levels he most while 9 "'*- _

I sept cIm B 3/4 4-1 homedeses os. 63 ll

t

. 3/4.4 REACTOR COOLANT SYSTEM _

RAlts Staellity tests at on region of the power ting 8Wis were reviewed

(( - -

should be performee/ flow e determine a generte l for determining the onservation decayap in which surveillance e' ,

plant variability o decay eric region for s has been detarmin ratto with efilance to accounto of 0.4 re and fuel designs. t for the plant tc was-of rated core f to correspond to i

3.4.1.1 1. and a THERMAL

1. greater than that speelffed l Pla l region specific calculati <

r monitoring neute e can be perforced cons flux noise levels. In this determine an appifcable In atise can be reduc since plant to p1 case t degree of {

is case, adequate verie111ty 1 decay ratio greater an rgin or will equal betoassure 0.8 by senttering d be oliefnetec.

regten whicn nas Neutron flum of iteit cycle n tron flux escillatyt ns.ist limits are als established te neutron flux SWR cares leally operate flus notse 1 eis of 1-125 of ratse caused by rendes selling naise. and f1utron

• Typical  !

the range power (peak-low to high recircu tien peak) have been loop flew ring both si rted for recircul ion loop operation. and dual bound

' fuel se found values areof constesfed in the theeutronignflus 1/sechanical gnificantly not a levels wh are at perating SWRs have d to be egligible conse of GE swa -

ace. In addi Ifait cycle oscillations nstrated that a stati11ty rol en, stability tasts cycles of 5-10 times t typical ccur they resul in peak ,te5 pea neutron neutron fium flut values. Therefore, act it neutron taken to ficient to flux noise e ly le els esceeding th e (3) times the ypical value e

ensure detection of att cycle nest su f-fluz escill ons.

Typically, magnitude as e utron flux noise evels show a gr 1 increase flow is increas absolute (constant cent I red pattern with two reacter recir latten toeps in o noise level tained at a spect c core flow c ration. Therof re, the baself neutron flu fisws. To sintain a rease e vartetten be be appifed eve a range of c re of the f range, the range ver which a the law f1 and high f1 and exceed 205 of rated core fl with two reti fic teseline i applied she d not from tests and operating p nts indicate latten leepe i sporetten, ata will result in appremias a range of of rated ce flow operation with two resi lation y a 50E increes in neutron f1 noise level ring leeps, saline data s Id be taten er the naminus red line at wh line date takeri et lower red lines (f.the majority e operation willHowe e base- .

tower power) w I result in a c,onser-vettve level atvalue a givensince theflow.

core neutron flum no se, level is p lonal to power 3/4.4. 2 SAFETY /RELItF VAlvtl '

the reacter coelant systes from being pressurized above 1375 psig in accordance with the ASME Code.

A total of 13 OpfA48LE safety /re Het

ADMINISTRAT!YE CONTROLS

• r -- -- ....... ...

ahalaanrz1qi arg5mmt antJiaan masapf (continued)

The radieestive effluent release report shall slee imelude as aseeeement of radiation desee to the likely east espeeed MRImm er TM PUN.!C free roaster releases and other nearby uranius fuel eyele seersee (Leetoding desee free primary efflueet pathueye and direst radiaties) for the provisue la consecutive meethe to show confeseanse with 40 crt 190, Revireemental Radiation Proteetles Standards for Nuclear Pouer Operaties. Aseeptable methode for calculattag the does costributies frem liquid and gaseems offluente are given in Regulatory Guide 1.109, Rev. 1.

The radl===tive afflueet release report shall Laelude the following l infasmaties for each elace of setid waste (as defined by 10 cra 81) shipped efinite euring the repeat periode

a. container volume,

b. Total serie quantity (specify whether detessiaed by asseuremsat or estieste),

s. Primeipal rada-lide (specify whether de*===aamd by measqrement er estimate),

d. Type of waste (e.g., spost rocia, eespeet dry waste, eveparater

• hettame),

e. Type of oestainer (e.g., Lah, Type A, Type e, Lasgo quantity), and

f. seitdifiesties agent (e.g., commet, uses f====tdahyde).

The radienstive efflueet release report shall imelude emptammad releases from l the site to the sumasTR$CTW ARRh of radienstive materiale La gemeeue and liesid offluente en a guarterly heele. s The radioactive efflueet release soport sha11 Laelude any shaages to the l F200508 ODETROL PRoeRhlt (Pct), err $t1E DOSE ChtAWEATEGE MANWhL (cDeu) er radiosetive waste eyetess made during the reporting period.

nEeMILE GrEntf!Es mEsenra 4.9.1.8 heetime reporte of operettag statisties and ekstdeun espertense shall he omhmitted en a monthly beste to the U.S. Waelear Regulatory casumiseLes ce Desament Centeel Desk, "==h8=Th==, D.C. 20585, with a sory to the Usume Administrater, Regien 1, ao later them tho'19th et eesh meeth fellowing the calendar aseth severed by the report.

enest nommasman LIntres anseer 8.g.1.9 Core operettag limite shall be established and h *=d in the PASGS generated COM OPMkTIBB LIEtf8 Mp0Rt before seek releed eyeLe er any remalaiag part of a releed syste for the felleeing Technical spesificattene 3/4.3.1 AVERheB ptAIIng r m ma g ht a g es 3333 3/4.3.3 IRINEINE M ITI M L 901 5 RkfSS 3/4.3.4 Ltumam s h; em m erges ames f ,3.10 Osc'd M on Power by AAcm.%. ( OMM

. .m. . e.d et . 67 11 I I

r ADMINISTRATIVE CONTROLS come oprawins LIi.iiti urEkT (Continued) g ee;

'the analytical' methods used.to determine the cor those previously reviewed and approved by WRC in operating limits shall be approved revision)*, E-24011-P-A (the latest

• W ##80 General Electric Standard

[assTAR II)/deley P 14slead 3tN65-A 4 lication for Reactor Fuel

• Seed Meth r' M*N8N8%

d,am*a twt #4Mliv Daher Sed Suppes .

19t6.

The core operating limits shall be determined so that all applicable limit W IH*"""/

(e.g., fuel theriaal mechanical limite, core thermal-hydraulic mli its, ECCS s analysis limits) of the safety analysis are met. limits, nuclear limits suc The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle , to the NRCInspector.

Resident Document Control Desk.with copies.to the Regional Administrator and SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Comunission, Document control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, within the time period specified fore each t' port .

6.9.3 Violations of the requirements of the fire protection program described in the Final Safety Analysis Report which would have adversely affected the ability to achieve and maintain safe shutdown in the event of a fire shall be subm to the U.S. Nuclear Regulatory conmission, Document control Desk, Washington i

I

, DC 20555, with a copy Report System withinto30the USNRC Administrator, Region 1, via the Licensee Event days.

1 6.10 mannem RETeMTTfier Code of Federal Regulations, the following records shall be retai least the minimum period indicated.

SPECIAL REDQRTS 6.10.2 The following records shall be retained for at least 5 years a.

-Records and logs of ut.it operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair, and nuclear replacement of principal items of equipment related to safety, c.

All REPORTABLE EVENTS submitted to the Comunission.

d.

Records of surveillance activities, inspections, and calibrations required by these Technical Specifications.

e.

Records 5.8.1. of changes inade to the procedures required by Specification f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector lealg tests and results.

• For Cycle.k 9, License htasNo.117 evaluated in the Safety avaluation dated perth 9 1999to support noPs CRast s-21 Amendment No.W.117 l l

F 3/4.3 INSTRUMENTATION

3/4.3.10 OSCILLATION POWER RANGE MONITOR LIMITING CONDITION FOR OPERATION 3.3.10 Four channels of the OPRM instrumentation shall be OPERABLE *. Each OPRM channel period based algorithm amplitude trip setpoint (Sp) shall be less than or equal to the Allowable Value as specified in the CORE OPERATING LIMITS REPORT APPLICABILITY

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTIONS

a. With one or more required channels inoperable:

1.** Place the inoperable channels in trip within 30 days, or 2.** Place associated RPS trip system in trip within 30 days.

b. With OPRM trip capability not maintained:

1. Initiate alternate method to detect and suppress thermal hydraulic instability oscillations within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

2. Restore OPRM trip capability within 120 days,

c. Otherwise, reduce THERMAL POWER to less than 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.3.10.1 Perform CHANNEL FUNCTIONAL TEST at least once per 184 days.

4.3.10.2 Calibrate the local power range monitor once per 1000 Effective Full Power Hours (EFPH) in accordance with Note f, Table 4.3.1.1-1 of TS 3/4.3.1.

4.3.10.3 Perform CHANNEL CALIBRATION once per 18 months. Neutron detectors are excluded.

4.3.10.4 Perform LOGIC SYSTEM FUNCTIONAL TEST once per 18 months.

4.3.10.5 Verify OPRM is enabled when THERMAL POWEn is > 30% RTP and recirculation drive flow s the value corresponding to 60% of rated core flow i once per 18 months 4.3.10.6 Verify the RPS RESPONSE TIME is within limits once per 18 months on a STAGGERED TEST BASIS. Neutron detectors are excluded.

• When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated ACTIONS may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, provided the OPRM maintains trip capability.

• • An inoperable channel or RPS trip system need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases initiate an alternate method to detect and suppress thermal hydraulic instability oscillations.

HOPE CREEK 3/4 3-108 Amendment No.

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INSTRUMENTATION BASES-3/4.3;10 Oscillation Power Range Monitor'(OPRM)

General Design Criterion 10 (GDC 10) requires the reactor core and associated coolant,: control, and protection systems to be designed with appropriate margin to: assure that acceptable fuel design limits are not exceeded during any condition of normal operation, including the affects of anticipated operational occurrences.. Additionally, GDC 12 requires the reactor core and associated coolant, control, and protection systems to be designed to assure

_that power oscillations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. The OPRM System provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel minimum critical power ratio (MCPR) safety limit.

References 1, 2, and 3 describe three separate algorithms for detecting stability related. oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. The OPRM System hardware implements these algorithms in microprocessor based modules. These modules, installed in local power range monitor (LPRM) flux amplifier slots in the Neutron Monitoring System (NMS) cabinets, execute the algorithms based on LPRM inputs and generate alarms and trips based on these calculations. These trips result in tripping the Reactor Protection System (RPS) when the appropriate RPS trip logic is satisfied, as described in the Bases for Specification 3.3.1, "RPS Instrumentation." Only the period based detection algorithm is used in the safety analysis. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations.

'The period based detection algorithm-detects a stability related oscillation based on the occurrence of a fixed number of consecutive LPRM signal period confirmations followed by the LPRM signal amplitude exceeding a specified

.setpoint.' Upon detection of a stability related oscillation, a trip is Egenerated for that OPRM channel.

The OPRM system consists of 4 OPRM trip channels, each channel consisting of two OPRM modules. Each OPRM module receives input'from LPRMs. Each OPRM module also receives-input from the RPS average-power range monitor (APRM) power and flow signals to automatically enable the trip function of the OPRM module.

Each OPRM module is continuously tested by a self-temt function. On detection of any OPRM module failure, either a Trouble alarm or INOP alarm is activated.

The OPRM module provides an INOP alarm when the self-test feature indicates that the OPRM module may not be capable of meeting its functional requirements.

It has been shown that BWR cores may exhibit thermal-hydraulic reactor instabilities in high power and low flow portions of the core power to flow operating domain. .GDC 10 requires the reactor core and associated coolant, control, and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition.of. normal operation, including the effects of anticipated operational occurrences. GDC 12 requires-assurance that power oscillations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed.

The OPRM System provides compliance with GDC 10 and GDC 12 by detecting the onset of oscillations and suppressing them by initiating a reactor scram.

This~ assures that the MCPR safety limit will not be violated for anticipat,ed oscillations.

I HOPE CREEK B 3/4 3-9 Amendment No.

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INSTRUMENTATION

-BASES The OPRM Instrumentation satisfies Criteria 3 of the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, idated July 22, 19931(58 FR 39132).

Four channels of the OPRM System are required to be OPERABLE to ensure that stability related oscillations are detected and suppressed prior to exceeding the MCPR. safety limit. Only one of the two OPRM modules' period based detection algorithm is required for OPRM channel OPERABILITY. The highly

' redundant ~and low minimum number of required LPRMs in the OPRM cell design ensures that large numbers of cells will remain operable, even with large numbers'of LPRMs bypassed.

1 The OPRM instrumentation is' required to be OPERABLE in order to detect and

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suppress-neutron flux oscillations in the event of thermal-hydraulic instability. -As described in References 1, 2, and 3, the region of anticipated oscillation is defined by THERMAL POWER h 30% RTP and

' recirculation drive flow 5 the value corresponding to 60% of rated core flow.

Therefore, the OPRM trip is required to be enabled in this region. However, to protect against anticipated transients, the OPRM is required to be OPERABLE with THERMAL POWER > 25% RTP. This provides sufficient margin to potential instabilities as a result of a loss of feedwater heater transient. It is not necessary for the OPRM to be OPERABLE with THERMAL POWER < 25% RTP because instabilities are not anticipated to result from any expected transients below l

this power. ,

ACTIONS a.1 and a.2 Because of the reliability and on-line self-testing of the OPRM instrumentation and the' redundancy of the RPS design, an allowable out of service time of 30 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the OPRM instrumentation still maintains OPRM trip capability (refer to Required Actions b.1 and b.2). The remaining OPERABLE OPRM channels continue to provide trip capability and provide operator information relative to stability activity. The remaining OPRM modules have high reliability. With this high reliability, there is a low probability of a subsequent channel failure within the allowable out uf service time..In addition, the OPRM modules continue to perform on-line self-testing and alert the operator if any further-system degradation occurs.

If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the OPRM channel or associated RPS trip system must be placed in the tripped condition per required actions a.1 and a.2. An inoperable channel or RPS trip system need not be placed in the tripped condition where this would cause the Trip sunction to occur. In these cases an alternate method to detect and suppress thermal hydraulic instability oscillations must be initiated. Placing the inoperable OPRM channel in trip

. tor the associated RPS trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the OPRM channel (or RPS trip system) in trip (e.g., as in the case where placing the' inoperable channel in trip would result in a full scram), the alternate method of detecting and suppressing thermal hydraulic instability osciliations is required. This alternate method is described in Reference 5. It consists of increased operator awareness and monitoring for neutron flux oscillations when operating in the region where oscillations are possible. If indications of oscillation, as described in Reference 5, are observed by the operator, the operator will take the actions described by procedures, which include l initiating a manual scram of the reactor.

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HOPE CREEK B 3/4 3-10 Amendment No. l l

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cINSTRUMENTATION BASES ACTIONS b~.'1 and b;2 Required action b.1 is intended to' ensure that appropriate actions are taken if multiple,' inoperable, untripped OPRM channels within the same RPS trip system result in not maintaining OPRM trip capability. OPRM trip capability is considered to be maintained when sufficient OPRM channels are OPERABLE or in trip (or'the associated RPS trip system is in trip), such that a valid OPRM signal will generate a trip signal in both RPS trip systems. This would require.both RPS trip systems to'have'one OPRM channel OPERABLE or in trip.(or the associated RPS trip system in trip). ,

Because of the low probability of the' occurrence of an instability, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is an acceptable time.to initiate the alternate method of detecting and suppressing thermal hydraulic. instability oscillations described in Reference

5. .The alternate method of detecting and suppressing thermal hydraulic instability oscillations would adequately address detection and mitigation in the event of instability oscillations. Based on industry operating experience with actual instability oscillation, the operator would be able to recognize instabilities during this time and take action to suppress them through a manual scram. In addition, the OPRM System may still be available to provide alarms to the operator-if the onset of oscillations were to occur. Since plant operation is minimized in areas where oscillations may occur, operation for 120 days without OPRM trip capability is considered acceptable with implementation of the alternate method of detecting and suppressing thermal hydraulic instability oscillations.

ACTION col With any Required Action-and associated Completion Time not met,-THERMAL POWER must be reduced to < 25% RTP with 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Reducing THERMAL POWER to < 25% RTP places the. plant in a region where instabilities cannot occur. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER < 25% RTP

'from full; power conditions in an orderly manner and without challenging plant

. systems,

.SR 4.3.10.1-A CHANNEL FUNCTIONAL TEST is performed on_each required channel to ensure that the entire channel will perform the intended-function. A Frequency of 184 days.provides an acceptable level of system average availability over the Frequency and'is based on the reliability of the channel (Ref. 7).

SR 4.3.10.2

-LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local I flux profile for appropriate representative input to the OPRM System. The -

1000 EFPD Frequency is based on operating experience with LPRM sensitivity -

changes. This surveillance is satisfied in accordance with' Note f, Table 4.3.1.1-1 of TS 3/4.3.1.

SR 4.3.10.3-l The CHANNEL CALIBRATION is a complete check of the instrument loop. This test verifies the channel responds to the measured parameter within the necessary  ;

range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with O

the plant specific setpoint methodology. l I

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. HOPE CREEK B 3/4 3-11 Amendment No.

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t INSTRUMENTATION BASES Calibration of the channel provides a check of the internal reference voltage and the internal processor clock frequency. It also compares the desired trip setpoints with those in processor memory. Since the OPRM is a digital system, the internal reference voltage and processor clock frequency are, in turn, used to automatically calibrate the internal analog to digital converters. The l Allowable Values are specified in the CORE OPERATING LIMITS REPORT. I As noted, neutron detectors are excluded from CHANNEL CALIBRATION because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 1000 EFPD LPRM calibration  ;

using the TIPS (SR 4.3.10.2). i The Frequency of 18 months is based upon the design objective that the OPRM operate over a complete fuel cycle, as a minimum, without requiring calibration.

SR 4.3.10.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods and scram discharge volume (SDV) vent and drain valves in Specification 3.1.3.1,

" Control Rod OPERABILITY" overlaps this Surveillance to provide complete testing of the assumed safety function. The OPRM self-test function may be ,

utilized to perform this testing for those components that it is designed to monitor. l The 18-month Frequency is based on engineering judgment and reliability of the components. Operating experience has shown that these components usually pass the surveillance when performed at the 18 month Frequency. l SR 4.3.10.5 This SR ensures that trips initiated from the OPRM system are not inadvertently bypassed when the capability of the OPRM system to initiate an RPS trip is required. The trip capability of the OPRM system is only required during operation under conditions susceptible to anticipated T-H instability oscillations. The region of anticipated oscillation is defined by THERMAL POWER 2 30% RTP and recirculation drive flow s the value corresponding to 60%

of rated core flow.

The trip capability of individual OPRM modules are automatically enabled based i on the APRM power and flow signals associated with each OPRM channel during normal operation. These channel specific values of APRM power and j recirculation drive flow are subject to surveillance requirements associated '

with other RPS functions such as APRM flux and flow biased simulated thermal power with respect to the accuracy of the signal to the process variable. The OPRM is a digital system with calibration and manually initiated tests to verify digital input including input to the auto-enable calculations.

Periodic calibration confirms that the auto-enable function occurs at l appropriate values of APRM power and recirculation flow signal. Therefore, l verification that OPRM modules are enabled at any time that THERMAL POWER 2 30% RTP and recirculation drive flow s the value ccrresponding to 60% of rated core flow adequately ensures that trips initiated from the OPRM system are not inadvertently bypassed, l

i HOPE CREEK B 3/4 3-12 Amendment No.

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l INSTRUMENTATION l l

BASES The trip capability of individual OPRM modules can also be enabled by placing the module in the non-bypass (Manual Enable) mode. If placed in the non-bypass or Manual Enable mode the trip capability of the module is enabled and this SR is met. The frequency of 18 months is based on engineering judgement .

and reliability of the components. l k

SR 4.3.10.6 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis (Ref. 8). The OPRM self-test function may be utilized to perform this testing for those I components it is designed to monitor. The PPS RESPONSE TIME acceptance I criteria are included in Reference 8.

As noted, neutron detectors are excluded from RPS RESPONSE TIMF testing  !

because the principles of detector operation virtually ensure an instantaneous J response time. RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is based upon operating experience, which shows that random failures of instrumentation components causing serious time degradation, but not channel failure, are infrequent.

REFERENCES:

1. NEDO-31960, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology," June 1991.

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2. NEDO 31960, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology," Supplement 1, March 1992.

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3. NRC Letter, A. Thadani to L. A. England, " Acceptance for Referencing of Topical Reports NEDO-31960, Supplement 1, 'BWR Owners Group Long-Term Stability Solutions Licensing Methodology'," July 12, 1993.

4. Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors," July 11, 1994.

5. BWROG Letter BWROG-9479. " Guidelines for Stability Interim Corrective Action," June 6, 1994.

6. NEDO-32465, "BWR Owners Group Reactor Stability Detect and Suppress Solution Licensing Basis Methodology and Reload Application," May 1995.

7. CENPD-400-P, Rev 01, " Generic Topical Report for the ABB Option III Oscillation Power Range Monitor (OPRM)," May 1995.

8. GE-NE-A13-00381-04, " Reactor Long-Term Stability Solution Option III:

Licensing basis Hot Bundle Oscillation Magnitude for Hope Creek, " March 1999.

9. OG96-630-169, " Guidelines for Stability Option III Enable Region,"

September 1996 HOPE CREEK B 3/4 3-13 Amendment No.

,. - j Docum:nt Control D:sk LR-N990137 Attachment 4 LCR H99-01 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF 57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS) {

CONFORMANCE WITH NRC ACCEPTED TOPICAL REPORTS AND ASSOCIATED NRC SAFETY EVALUATION REPORTS The NRC, in their letter (Bruce A. Boger Director Division of Reactor Controls and l

Human Factors to Robert A. Pinelli, BWROG, dated August 16,1995) which transmitted the Safety Evaluation Report accepting the Asea Brown. Boveri Combustion Engineering (ABB-CE) Option lli OPRM system as a permanent long term solution for the thermal-hydraulic stability issue, required each licensee referencing the OPRM system licensing topical report, CENPD-400-P-A, to address the plant specific issues identified in the SER and to identify and justify any deviations from CENPD-400-P-A and the associated SER.

The plant specific items as stated in Section 5.0 of the SER are restated below along with PSE&Gs response.

1) Confirm the applicability of CENPD-400-P, including clarifications and reconciled differences between the specific plant design and the topical report design descriptions.

The Hope Creek OPRM conforms with the Safety Evaluation Report by the O'fice of Nuclear Reactor Regulation, CENPD-400-P, Generic Topical Report for the ABB Option Ill Oscillation Power Range Monitor. The Hope Creek installation and implementation of OPRM is consistent with CENPD-400-P and the associated SER.

There are no deviations from Topical Report Sections 2.3 and 3.0.

At Hope Creek, the " plant process computer", as used in the context of the SER is the Control Room Information Display System (CRIDS).

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2) Confirm the applicability of BWROG topical reports that address the OPRM and i associated instability function, set points and margins.

The Boiling Water Reactor Owners' Group (BWROG) topical reports which address the OPRM and associated instability functions, set points and margins are NEDO-  ;

31960 "Long Term Stability Solutions Licensing Methodology" od its supplement and NEDO-32465 " Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications" (References 1,2 and 5 of CENPD-400).

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' Docum:nt Control Dssk LR-N990137

' Attachment 4.

LCR H99-01 o

In the Safety Evaluation Report (SER) accepting NEDO-31960, the NRC specified five conditions required for implementation of Option illin any type of BWR. Each of these five conditions has been reviewed and the following confirmations of the '

. applicability of NEDO-31960 and its supplement to the proposed implementation of the Option ill solution at Hope Creek are provided.

- i) . All three algorithms described in NEDO 31960 and its supplement are used.

These algorithms are the amplitude _ based or high-low detection algorithm; the

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. growth rate or high oscillation amplitude algorithm and the period based algorithm. The OPRM executes these algorithms and generate alarms and RPS

. trips based on the results.

ii) t The validity of the selected scram set points has been confirmed using the initial application and reload review methodology described in NEDO-32465. This C

methodology uses a combination of generic representative plant and plant specific analyses and includes an uncertainty treatment that accounts for the number of failed sensors.

iii)- The selected bypass region outside of which the detect and suppress action is deactivated is defined in the proposed Technical Specifications of Attachment 3.

iv)- The automatic protective function of the OPRM when fully implemented will be a full reactor scram.

v). The LPRM assignment grouping proposed to be _ implemented conforms with LPRM assignments shown in Appendix D of NEDO-32465. The NRC SER-concluded that the initial Application and Reload Review methodologies of

. NEDO-32465 are acceptable. These methodologies specify that the LPRM assignment is a key assumption.- The discussion of the initial application methodology in section 5 of NEDO-32465 specifically identifies the LPRM assignments in Appendix D as examples of the expected LPRM assignments a utility may choose. Therefore, it is concluded that since the LPRM assignment proposed to be implemented at Hope Creek is consistent with an example presented in Appendix D of NEDO-32465, and Appendix D LPRM assignments lwere cited as possible LPRM assignments in the initial application methodology which the NRC found acceptable, the proposed LPRM assignment for Hope Creek is acceptable. No changes to current requirements for a minimum operable number of LPRM detectors for accurate power distribution monitoring  !

' and APRM operability were proposed in NEDO-31960 and found acceptable by the NRC in the associated SER. The proposed implementation at Hope Creek is consistent with this accepted NRC position.

i In the Safety Evaluation Report (SER) accepting the Boiling Water Reactor Owners'

. Group (BWROG)" Reactor Stability Detect and Suppress Solutions Licensing Basis 2

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Docum:nt Control D:sk LR-N990137 Attachment 4 LCR H99-01 Methodolo'gy for Reload Applications" (NEDO-32465), the NRC concluded that the proposed methodology should produce set point values that will result in a very low likelihood of exceeding CPR safety limits during instability events. The description of i the confirmation set points of the period based algorithm in Section 3.4 of NEDO-32465 identifies that a range of values for the period tolerance and corner frequency was established to allow the OPRM system to be tuned for the unique LPRM noise characteristics of each plant. A specific range of values of the period tolerance and corner frequency demonstrated against available plant data was listed in Table 3-1 of NEDO-32465.

The range of period tolerance set point values listed in Table 3-1 of NEDO-32465 does not include the value of the set point proposed for implementation of Option lli at Hope Creek. Based on testing performed during cycle 8 it has been determined that a period tolerance of 50 milliseconds is needed to avoid spurious trips and alarms of the OPRM system for the Hope Creek plant. The proposed value of 50 milliseconds is consistent with the range identified for the period tolerance set point in section 6.2 of Supplement i to NEDO-31960.

In an October 21,1997 letter to the NRC from Southern Nuclear Company (SNC),  ;

additional information was provided that demonstrated that the Option til firmware

. successfully recognized, processed and generated output trips over a range of values for the period tolerance and the corner frequency for instability event data from the Swiss Plant KKL (SNC letter HL-5501). Included in the range of values of the period tolerance set point considered was the value of 50 milliseconds. The capability of the Option 111 firmware as prescribed in NEDO-31960 and its supplement to reliably perform its intended function to provide continuous confirmations upon transition from stable reactor operation to a growing reactor instability has been verified and the proposed deviation from the range of values for the period tolerance listed in Table 3-1 of NEDO-32465 is justified.

3) Provide a plant-specific Technical Specification (TS) for the OPRM functions consistent with CENPD-400-P, Appendix A.

Provided as Attachment 3

4) Confirm that the plant-specific environmental (temperature, humidity, radiation, '

electromagnetic and seismic) conditions are enveloped by the OPRM equipment environmental qualification values.

The Hope Creek installation and implementation of OPRM is consistent with the Equipment Qualification described in section 3.3 of the SER. The OPRM components ,

are those subject to the ABB-CE environmental qualification program. The OPRM l

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Docum:nt Control D:sk LR-N990137 Attachment 4 LCR H99-01 equipment'is installed in the main control room. The OPRM System, along with the Replacement Bulk Power Supply and the Dual Voltage regulator are qualified to perform their Class i E safety function for continuous operation for the following Environmental conditions in the Hope Creek main control room as listed in the HCGS Environmental Design Criteria, D7.5, Table 6 Normal Operatina Conditions:

Temperature: Maximum 78 'F; Average 76 'F; Minimum 74 *F Relative Humidity: Maximum 50% Minimum 40%

Radiation: 2x10E2 RADS TID Abnormal Conditions Temperature 78'F Relative Humidity: Maximum 50%; Minimum 40%

Pressure .25/ .25 in WC Temperature and humidity qualification of the OPRM module was performed by test.

The OPRM is designed to continuously operate in the following environment, while meeting all performance requirements: -

Normal Ambient Temperature: 40 'F to 120 *F 1

Abnormal Ambient Temperature: 140 'F for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Humidity: 30% to 95% RH non-condensing Voltage: 90 to 140 VAC Frequency: 47.5 to 63 Hz.

The OPRM system is designed to provide a high degree of immunity form EMI/RFI and to minimize generated EMI/RFI that may interfere with devices connected to it, devices that share a common AC supply, and devices located in the same enclosure. The l OPRM system is designed and tested to meet the Electrostatic Discharges (ESD) requirements of IEC 801-2, Level 4. Fast transient withstand capability is demonstrated for all power input and ou'.put and all process input and output circuits,

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signal common and protective earth connections based on IEC 80104, Level 4 (4 kV on j power and ground,2kV on process signals) as described in IEC 801-4, Sections 7.3.1 and 7.3.2. OPRM circuitry is located inside a metal enclosure. All external power, 1 inputs and outputs, pass through filters which, together with the metal enclosure, provide an EMI boundary. These features, when combined with grounding and cable ,

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T Document Control D:sk LR-N990137 Attachment 4 LCR H99-01 separation ^in accordance with PSE&G standards and restrictions on welding end portable transceiver use in the main control room area, ensure the OPRM system is protected from the effects of electromagnetic interference.

The following two devices were seismically qualified as documented in Document no.

R1581, " Seismic Simulation Test for the Oscillation Power Range monitor (OPRM)','

PN11426-1, -2 and the Digital isolation Block (DIB), PN 11427- 1, Wyle Laboratories Seismic Simulation Test Report No. 44758-1, Revision A. This test used Required response Spectra (RRS) which envelopes and satisfies the requirements of the base testing reported for other OPRM devices above in Document No. 0000-lCE-37700, Rev. O. The SSE RRS (5% damping) run from 4.5 g (rigid range) to 10g (resonant response range) with the transition at 35 Hz. The CBE RRS (5% damping) run from 3g (rigid range) to Sg to 5.6g (resonant response range with a transition at 35 Hz). The devices were tested on the same test-mounting frame as used for the first group of devices noted above. The OPRM module is rack mounted and the DIB is DIN rail mounted.

Four TECs are mounted on Unistrut (two per span) located very low (6" above the base) spanning bays 1 and 5. This device is being dynamically qualified in accordance with a DBE Test Response Spectrum peaking at 9 g at 2% damping. The requirements for this device are not as severe as other devices because of the low and rigid mounting configuration. Seismic data for testing of the TEC isolator is based upon the generic RRS indicated in the Qualification Test Report (Non-OPRM Equipment),

Document Number 00000-ICE-37700, Rev.0.

Buchanan Terminal blocks are mounted on Unistrut and located very low (6" above the base) spanning bays I and 5. The terminal blocks are generally considered to be seismically insensitive and rugged (See EPRI report TR105849, Generic Seismic Technical Evaluation of Replacement items for Nuclear Power Plants).

Din Rail and Unistrut mounting is utilized in panel H 11-P608. The panel is 150" wide by 36" deep by 90" high. Each of the five bays is 30" wide. The total panel weight is approximately 4500 lbs. It is anchored to the floor with a minimum tie down of 20 5/8" diameter bolts (or full plug weld for all twenty holes). With the exception of the OPRM module and switch, all devices are mounted on new Unistrut (N3300) or DIN (35/15) rail (or combinations). The longest Unistrut support beam is less than 30"(the width of a single bay). A simply supported beam, conservatively center loaded with 10 pounds )

I (more than any device combination on any beam) will have a strong axis fundamental frequency of 84 Hz and a weak axis fundamental frequency of 44 Hz. This rigidity I assures against local amplification of seismic inputs for such internally mounted devices.

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Docum:nt Control D:sk LR-N990137 Attachment 4 LCR H99-01

5) Confirni that administrative controls are provided for manually bypassing OPRM channels or protective functions, and for controlling access to the OPRM functions.

The Hope Creek OPRM installation and implementation is consistent with the SER. l Hope Creek procedures provide administrative control for placing individual OPRM modules in manual bypass. When the OPRM modules are not in manual bypass, the OPRM protective function is automatically bypassed or automatically activated when l

' the reactor power and recirculation flow are in the appropriate regions of the reactor i power / flow map to require automatic bypass or activation respectively. The OPRM as installed and implemented automatically enables its pre- trip and trip alarm outputs ,

upon entry into the high power, low core flow region of the power / flow operating map.

6) Confirm that any changes to the plant operatcr's main control room panel have l received human factor reviews per plant-specific procedures. i The Hope Creek OPRM installation and implementation includes activation of the main I control room overhead annunciator if the OPRM has been manually bypassed or deliberately rendered inoperative. Keylock access is necessary to manually bypass an OPRM module. Changes to OPRM software require both keylock access and a password. Procedural requirements control placing an OPRM module in bypass and

- verifying restoration.

The Hope Creek OPRM installation and implementation includes an operator interface via the CRIDS computer, Control Room annunciators that signal system status and/or problems, and the C PRM front panel LED'S. Alarms which are wired to the Annunciator Logic Cabinet and displayed on the Control Room Overhead Annunciator Panel (window boxes) include the TRIP ENABLE, ALARM, TRIP, AND BYPASS /INOP/ TROUBLE. Human Factor Engineering principles consistent with the HCGS Annunciator System Study have been applied in selecting the annunciator i location and groupings. In addition, OPRM modules are provided with local indicators. i Local LED indicators provide indication for ALARM, TROUBLE, INOP, TRIP, TRIP ENABLED, and READY. ALARM, TRIP, and TROUBLE LEDs are latched until reset I locally.

Other than the plant specific items addressed above, there are no deviations from CENPD-400-P-A and the associated SER.

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