Text

Proposed Tech Specs,Revising Fuel Cycle Specific Tech Spec Sections 3/4.2,B3/4.2,5.3.1 & 5.3.2 & Adding New Definition & Administrative Requirement
	ML20247C310
Person / Time
Site:	Hope Creek
Issue date:	05/18/1989
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML19302D914	List: ML20247C304; ML20247C310;
References
	NLR-N89055, NUDOCS 8905240392
	Download: ML20247C310 (37)
	v • d • e

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l ATTACHMENT 2 HCGS LICENSE CHANGE REQUEST 89-12, NLR-N89055 REMOVAL OF CYCLE SPECIFIC PARAMETER LIMITS FROM THE TECHNICAL SPECIFICATIONS

'AFFECTED TECHNICAL SPECIFICATION PAGES s

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g g I , 0500 {

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3 DEFINITIONS i

SECTION

1. 0 DEFINITIONS PAGE 1.1 ACTI0N...............................................j............... 1-1 1.2 AVERAGE PLANAR EXPOSURE.............................................- 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................... 1-1

1. 7 1.4 CHANNEL CALIBRATION..................9............................... 1-1 C s hsiss 1.5 CHANNEL CHECK..................................... >

1-1

@b Of6 DAM 4 1. 6 CHANNELFUNCTIONALTEpT............................................. 1-1 u nits REfo C 1.7 ALTERATION......................................................

CORE i ; . 1-2 g AM\DRE MAXIMUM FRACTION OF LIMITING POWER DEN 51TY. . . . . . . . . . . . . . . . . . . . . . [2

/ <

/./o -(\ M\'t RIT ICA L POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . >f .....

/.i( 4 \ 1W$0S E EQUIVALENT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,1

/ ..? .

/.a._,((\1.W-AVERAGE DISINTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . /

N-

/M ,(\\hWMERGENCY CORE COOLING 5YSTEM (ECCS) RESPONSE TIME. . . . . . . . . . . . . . . . . . .

'.62.'

//N(\\hWND-0F-CYCLE REC!RCULATION PUMP TRIP SYSTEM RESPONSE TIME......... 1-). .. .

/./5,(\\1.M)FR ACTION OF LIMITING POWER DEN 51TY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

ut, g\ yAIWRACTION OF RATED THERMAL, P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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>f I./1g (\\h WV)tEQUENCY N0TATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . FI l l((\\L11\BENTIFIED LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

/.ff (\\ %1 W 50LATION SYSTEM RESPONSE TIME....................................... 1;3-

/ 20 (\%.\T9%QMITING CONTROL RCD PATTERN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..1. .5'.

/.2J~(\\ WQ\)!NEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

/

i.27-y0DA EWDGIC SYSTEM FUNCTIONAL TE5T. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-f

/.23,Q.@!mM FRACTION OF LIMITING POWER DEN 51TY. . . . . . . . . . . . . . . . . . . . . . . . . . .

M

/. 2L (\\A.NT4 WieER(5) 0F THE PUBLIC. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N

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/. 25,(X V\tOINIMLM CRITICAL POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . M HOPE CREEK $ heMMC At gas

INDEX DEFINITIONS SECTION DEFINITIONS (Continued) 3 PAGE

/ 2L( \1htSNQFF-GAS RADWASTE TREATMENT SYSTEM. . . . . . . . . .y. . . . . . . . . .

/.2 7 (\ h200FF SITE DOS E CALCULATION MANUAL. . . . . . . . . . d ..........

/.Sf (W.tNfERABLE - OPERABILITY. . . . . 9 /

......................................... 3-&

/.27 , (\WRDERATIONAL CONDITION - CONDITION. . . . . . . .d. . . . . . . . . . . . . . . . . . . . . .l#I .....

/ . 3 0 , C\1\29 % Y S I C S T E ST,5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

/ 3/ o ( 4324.) ESSURE BOUNDARY LEAKAGE. . . . . . . . . . . . . . . . . . 1,,-5 ..........-

/4z.

f\1.11\bRIMARY CONTAINMENT INTEGRITY. . . . . . . . . . . . . . . . .F5' ..........

/. 33_p 't .'12\ AR0 CESS CONTRO L PR0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ./. . . . . . . . . . . .

p

/6/ ,(\ ND33GURGE- PURG I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .p4- ...........

f 35 ,(\ Nh34 DATED THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . p& .......

%d D j.s(, _A \1.%)EACTOR PROTECTION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . .Pti . . . . .W ..

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/. 3 '/A\1MlOtEPO RTAB L E EVENT. . . . . . . . . . . . . . . . . . . . . . . . . . .N. . . . . . . . .

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/.3r 00 0ENSITY.......................................................... 1-6 -

/.3LC\ L%3ECONDARY CONTAINMENT INTEGRITY. . . . . . . . . . . . . . . . p? ...........

/40 g N1M K1HUTDOWN MARGIN...................................................... F7'

/g/ J\ l.%li1TE 80VNDARY. . . . . . . . . . .

........................................... N -

/.4t J \ 4 4 D SOLIDIFICATION....................................................... 1-J.

1.t$ ,(\ M M h300RC E CHEC K. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F1f 1.% _, M SPIRAL /

REL0AD........................................................ 1-8 l

/

I 4(\ 4h40S P I RAL UNL0AD. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . )S- . . . .l . . . . . . .

l.%(5.\Ih40 STAGGERED TEST BAS IS. . . . . . . . . . . . . . . . . . . . F. .8'. .l . . . . . . ,

1 4 L/\ WOTHERMAL P0WE R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l . . . . . . .

/

).q h 6\4MNTUR8INE BYPASS SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . .9'. . l. . . . . . )

HOPE CREEK 11 Amendment No. 14 l -

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! INDEX DEFINITIONS SECTION DEFINITIONS (Continued)

PAGE

/Yf 4:48 UNIDENTIFIED LEAKAGE................................................. 1-p l

/SO UNR'STRICTED

. E AREA.................................................... 1-9' l Tc -

' hL&O VENTI.LATION EXNAUST TREATMENT SYSTEM' .1-Ir. .- l. . . . . .

r. 5p MA V ENT I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .rg'/ ....

TABLE 1.1, SURVEILLANCE FREQUENCY NOTATION................................ 1-10 TABLE 1.2, OPERATIONAL CONDITIONS......................................... 1-11.,

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HOPE CREEK 111 Amendment No. 14 !

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-INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRfr4ENTS SECTION j PAGE

/

3/4.0 APPLICABILITY............................y(.................. 3/4 0-1 .

/

3/4.1 REACTIVITY CONTROL SYSTEMS ,r 3/4.1.1 SHUTDOWN MARGIN............... ....................... 3/4 1-1 I 3/4.1.2 3/4 1-2 REACTIVITYANOMALIES......./..*............................

3/4.1.3 CONTROL RODS Control Rod Operability,.................................. 3/4 1-3 Control Rod Maximum cam Insertion Times................. 3/4 1-6 Control Rod AverageJacram Insertion Times................. 3/4 1-7 Four Control Rod G oup Scram Insertion Times.............. 3/4 1-8 Control Rod Scrad Accumul ators. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-9 Control Rod Drive Coupling................................ 3/4 1-11 Control Rod Position Indication........................... 3/4 1-13 Control Rod Drive Housing Support......................... 3/4 1-15 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Rod Worth Minimizer........................................ 3/4 1-16 Rod Sequence Control System............................... 3/4 1-17 Rod Block Monitor......................................... 3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................. 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Solution. Volume /

Concentrate on; Requi rements. . . t . . .c. ... . . . . .. 3/4<1-21' 3/4.2 POWER DISTRIBUTION LIMITS' 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................ 3/4 2-1 "eximua Average Planar Linear tic:t Generation R:t 6 (MAPHEGR)-versus-Average Plance Exp0:ere'- -

Figure 3.2.1-1 h itial Core Tuci Typ: PSCISC71f......... 3/4 2-2 (De[cled ' .

Figure 3.2.1-2 Initici C r: Fu:1 Type "2C:200' Y........ 3/4 2-3 l HOPE CREEK v 1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Figure 3.2.1-3 I ntfal-Gore fuel Type P8010103.h ....... 3/4 2-4 (Delefed Figure 3. 2.1-4%itial Core Fuel Type PBCIB2499. . . . . . . . 3/4 2-5 Figure 3.2.1-5 Initic1 Core fuel Type PGCIB278 # c........ 3/4 2-6 3/4 2.2 APRM SETP0lNTS............................................ 3/4 2-7 3/4.2.3 MINIMUM CRITICAL POWER RATI0.............................. 3/4 2-8

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Figure 3.2.3-lbinimum Critical Power Rctic (MCPR) "'

verce: ct Rated c1c d.................... 3/4 2-10 Oclefech Figure 3.2.3-2'K f Factor................................ 3/4 2-11 Table 3.2.3-1E MCPR Fcedwater Heating Cepecity"'

Adjust entS ............................. 3/4 2-12 3/4.2.4 LINEAR HEAT GENERATION RATE............................... 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................. 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation........................... 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times..................................... 3/4 3-6 Figure 4.3.1.1-1 Reactor Protection System Surveillance Requirements........................... 3/4 3-7 3/4J3.2' I SOLATION: ACTUATION ' INSTRUMENTATION?. '. . . . . . . . . . .... . . . . . .. .. 3/4 3-9 '

Table 3.3.2-1 Isolation Actuation Instrume.ntation....... 3/4 3-11 Table 3.3.2-1 Isolation Actuation Instrumentation Setpoints................................. 3/4 3-22 Table 3.3'.2-3 Isolation System Instrumentation Response Time...................................... 3/4 3-26 Table 4.3.2.1-1 Isolation Actuation Instrumentation Surveillance Requirements................. 3/4 3-28 HOPE CREEK vi

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INDEX l i

ADMINISTRATIVE CONTROLS SECTION PAGE l

i ANNUAL REP 0RTS............................................ 6-17 )

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_ - ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT. . . . . . . . 6-18 e

C00E MU SEMIANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT.... 6-19 A N D O P E R A D 'Y6 MONTHLY OPERATING REP 0RTS................................. 6-20 6.9.2 SPECIAL REP 0RTS........................................... 6-20 6.10 RECORD RETENTION.............................................. 6-21 6.11 RADIATION PROTECTION PR0 GRAM.................................. 6-22 6.12 HIGH RADIATION AREA........................................... 6-22 6.13 PROCESS CONTROL PROGRAM (PCP)................................. 6-23 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM)........................ 6-24 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASE0US, AND SOLID WASTE TREATMENT SYSTEMS................................. 6-24 k

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l HOPE CREEK xxvi

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor -

pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs, IRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative 4- position. j f.fpe e g,p g "

CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY yggf ghpgThe CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD) shall be j,7) highest value of the FLPD which exists in the core.

, CRITICAL POWER RATIO l

(WMNThe CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the l

f,jf f assembly which is calculated by application of the GEXL correlation to I cause some point in the assembly to experience boiling transition,

-divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 (WNtkDOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gfff gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for'this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

l-AVERAGE DISINTEGRATION ENERGY MW shall be the average, weighted in proportion to the concentration of I j,ft f each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME ML.W1he EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time f,f 3 f interval,from when the monitored parameter exceeds its ECCS actuation-set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

HOPE CREEK 1-2

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Insert A CORE DESIGN AND OPERATING LIMITS REPORT 1

1.8 The CORE DESIGN AND OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for i

the current operating reload cycle. These cycle-specific l core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Plant operation within these limits is addressed in individual specifications.

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DEFINITIONS l

l END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME

(//KM/Jhe END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the 9 9) fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

I

a. Turbine stop valves, and

b. Turbine control valves.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

FRACTION OF LIMITING POWER DENSITY

(\ L%)The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at j45) a given location divided by the specified LHGR limit for that bundle type.

FRACTION OF RATED THERMAL POWER 6/T.Af(/Jhe FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured

/.g, f THERMAL POWER divided by the RATED THERMAL POWER.

FREQUENCY NOTATION

((\1\16\7he FREQUENCY NOTATION specified for the performance of Surveillance jgf Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE (II/1///)DENTIFIED LEAKAGE shall be:

HI) a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both spe-cifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME

(\\1uB\}he ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint et the channel I4f sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL ROD PATTERN

(/ A/VJ/A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting 120 f value for APLHGR,.LHGR, or MCPR.

HOPE CREEK 1-3

DEFINITIONS LINEAR HEAT GENERATION RATE R 132RNOINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit fy, y length of fuel rod. It is the integral of the transfer area associated with the unit length.

heat flux over the heat LOGIC SYSTEM FUNCTIONAL TEST G//TJ212) LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, l'g3 f etc, i.e. , all relays and contacts, all trip units, solid sta' ' logic elements, of a logic circuit, from sensor through and includ',a the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTlunAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic, system is tested.

MAXIMUM FRACTION OF LIMITING POWER DENSITY 6N1)dY\7he MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be highest

/.23 f value of the FLPD which exists in the core.

i MEMBER (S) 0F THE PUBLIC C/h/f30 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally j,zy f associated with the plant. This categoryAlso does.not include employees of excluded from this category the ttility, it contractors or vendors.

are /e. sons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recre-ational, occupational or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO C\ \ lib (CIhe MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which f,737 f exists in the core.

OFF-GAS RADWASTE TREATMENT SYSTEM

(//ff 90An 0FF-GAS RADWASTE TREATMENT SYSTEM is any system designed and installed 72g Jb to reduce radioactive gaseous effluents by collecting reactor coolant sys-tem offgases from the main condenser evacuation system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

OFFSITE DOSE CALCULATION MANUAL l C\ h\id)he OFFSITE DOSE CALCULATION MAH1: iL (0DCM) shall contain the current method-F ology and parameters used in the calc 4:at; n of offsite doses due to radio-

/.27 / active gasecus and liquid effluents, + t e calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the radiological environmental monitoring program.

l HOPE CREEK 1-4 i i

DEFINITIONS OPERABLE - OPERABILITY

(\%N\$ system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s) and

/"f when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION - CONDITION

(//l//E6n OPERATIONAL CONDITION, i.e. , CONDITION, shall be any one inclusive

,' g f combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS

(\\D2'9\)HYSICS TESTS shall be those tests performed to measure the fundamental

/,30 f nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE

(//1/,MffRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault j,g j in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY

(\\UAl\fRIMARY CONTAINMENT INTEGRITY shall exist when:

I' a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE primsry containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deact b.ated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. Each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.

~

e. The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

HOPE CREEK 1-5

DEFINITIONS PROCESS CONTROL PROGRAM Q/MM/Jhe PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure p that the SOLIDIFICATION or dewatering and packaging of radioactive wastes results in a waste package with properties that meet'the minimum and stability requirements of 10 CFR Part 61 and other requirements for trans-portation to the disposal site and receipt at the disposal site. With SOLIDIFICATION, the PCP shall identify the process parameters influencing SOLIDIFICATION such as pH, oil content, H 2 O content, solids content ratio of solidification agent to waste and/or necessary additives for each type of anticipated waste, and the acceptable boundary conditions for the process i parameters shall be identified for each waste type, based on laboratory scale and full scale testing or experience. With dewatering, the PCP shall include an identification of conditions that must be satisfied, based on full scale testing, to assure that dewatering of bead resins, powdered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of the low-level radioactive waste disposal site.

PURGE - PURGING l Ci\LWURGE or PURGING shall be the controlled process of discharging air or gas l

j'g f from a confinement to maintain temperature, pressure, humidity, concentra-tion or other operating condition, in such manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 6f/1/iORATED THERMAL POWER shall be a total reactor core heat transfer rate to p35f the reactor coolant of 3293 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1

l (\ WilS] REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from f, g f when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or I

total steps such that the entire response time is measured. 3 REPORTABLE EVENT

(//VN) REPORTABLE EVENT shall be any of those conditions specified in f Secti"on 50.73 to 10 CFR Part 50.

ROD DENSITY (s\\%N\JROD DENSITY shall be the number of control rod notches inserted as a f raction of the total number of control rod notches. All rods fully

/.fr f inserted is equivalent to 100% R0D 0ENSITY.

1 HOPE CREEK 1-6

I I I *-

DEFINITIONS 1

SECONDARY CONTAINMENT INTEGRITY U/1/34MEC040ARY CONTAINMENT INTEGRITY shall exist when:

1 a. All secondary containment penetrations required to be closed during accident corditions are either: I

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve or damper, as applicable secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

. b. All secondary containment hatches and blowout panels are closed and sealed.

c. The filtration, recirculation and ventilation system is in compliance with the requirements of Specification 3.6.5.3.

d. For double door arrangements, at least one door in each access to the secondary containment is closed.

e. For single door arrangements, the door in each access to the secondary containment is closed, except for normal entry and exit.

f. The sealing mechanism associated with each secondary containment i penetration, e.g., welds, bellows or 0-rings, is OPERA 8LE.

g. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

SHLJTDOWN MRGIN V/4/)9$NLITDOWN MRGIN shall be the amount of reactivity by which the reactor is

+ suberitical or would be subcritical assuming all control rods are fully

/.YO/ inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

SITE SOUSARY

(\\ M4eLThe 5"ITE SOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled, by the licensee, f

l HOPE CREEK 1-7

.,,,.-g ....m-e= g e * * * * * * * ** * '*

i DEFINITIONS SOLIDIFICATION

.(\\1\WMOLIDIFICATION shall be the immobilization of wet radioactive wastes such

/M j as evaporator bottoms, spent resins, sludges, and reverse osmosis concen-trates as a result of a process of thoroughly mixing the water type with a solidification' agent (s) to form a free standing monolith with chemical and physical characteristics specif.ed in the PROCESS CONTROL PROGRAM (PCP).

SOURCE CHECK

(// h/4 GA SOURCE CHECK shall be the qualitative assessment of channel response gj when the channel sensor is exposed to a source of increased radioactivity.

SPIRAL RELOAD

(\\hW SPIRAL RELOAD is a core loading methodology employed to refuel the core

/. 44 / loaded after a complete core unload. During a SPIRAL RELOAD the fuel is to be into individual control cells (four bundles surrounding a control blade) in <.sW ral fashion centered o, an SRN moving outward. Before initiating a SPIRAL RELOAD, up to four bundles may be loaded in the four bundle locations immediately surrounding each of the four SRMs to obtain the required channel count rate.

(///MWA SPIRAL UNLOAD is a core unloading methodology employed to defuel when

/,4/5/# the complete core is to be unloaded. The core unload is performed by first removing the fuel from the outermost control cells (four bundles surrounding a control blade). Unicading continues in a spiral fashion by removing fuel from the outermost periphery to the interior of the core, synnetric about the SRMs, except for the four t 11es around each of the four SRMs. When sixteen or less fuel bundler ..<e in the core, four around each of the four SRMs, there is no need to maintain the re-quired channel count rate.

STAGGERED TEST BASIS tiA%4M STAGGERED TEST BASIS shall consist of:

l

/.'/C I a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER

(//1/44BHERMAL POWER shall be the total reactor core heat transfer rate to the l reactor coolant.

HOPE CREEK 1-8 Amendment No.14

DEFINITIONS TUR$!NE SYPASS SYSTEN RESPONSE TIME f//1/4(Ahe TURBINE BYPASS SYSTEM RESPONSE TIME consists of two Idi /f ~vals: a) time from initial movement of the main tureine stop valve or con.

trol valve until 80% of the turbine bypass capacity is established, and b) the time from initial movement of the main turbine stop valve or control valve until initial movement of the tureine bypass valve. Either response time may be measured by any series of sequential, overlapping, or total steps such that the entire response time is asasured. 1 L

UNIDENTIFIED LEAKAGE y

).'n s / A/4F/9 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE. l UNRESTRICTED AREA 7,goj (g\M9dn UNRESTRICTED AREA shall be any area at or beyond the SITE 80UNOARY acc u to which is not controlled by the licensee for purposes of protec- 'l tion of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUW ARY used for residential quarters or for industrial, coenercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM

(// Y.Mf) VENTILATION EXHAUST TREATHENT SYSTEM shall be any system designed and l l.61 f installed to reduce Oaseous radiciodine or radioactive asterial in particu-late form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing todines or particulate from the gaseous exhaust streas prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING

(\M.hlENTING shall be the controlled process of discharging air or gas from a l l 6 2- confinement to maintain temperature, pressure, humidity, concentration or other operatin0 condition, in such a manner that replacement air or gas is not provided or required durin does not imply a VENTING process.g VENTING. Vent, used in systes names, e

e HOPE CREEK 1-3 Amendment No. 14 l FD 1 see 9

3: : : .i: ::3 ::!;-~:N .:w '3 3: : . :. . E::. I :.aN .: .:NEAR -EA' 3ENE:A:0N Ra E

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'- 9;;re s 2. 2.1-1. 2. 2.1-2, 0. 2.1 -0, 0. 2.1 0, enc 3. 2.1- SP The lim ts shall not exceso tne '

, ,;-;, 3,2, - , 3, g g OPE M N ,;-2, 0.2.1-2 ene 0.2.1-5'sna11 ee reca:e: ::

l.TACg 5 salue of 0.86 times accirculation loop operation, tne two. recirculation loop operation limit wnen ir s ;;e ff(DR'I~

,_ AD8LICAE!LITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greate- t in or eccai to 25% of RATED THERMAL POWER.

AC-*0N:

Wi2.1-

'- -+ 2. tt an4 APLNGR exceeding the limits ef--4gure-3:2.1-1, 0. 2.1-2, 0. 2.1- 0.

APLHGR t er 3.2.1-5, initiate corrective action within 15 minutes anc rest: e less tnan 25% of RATED THERMAL POWER within th SURVEILLANCE REQUIREMENTS A.2.1

) e*4 m in: All APLHGRs shall be verified to be equal to or less than the limits

'r : 9;um 3. 2.1-1, 0. 2.1-2, 3. 2.1-3, 0. 2.14 ; .: 0. 2.1- 5-:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .,

t.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at f least 15% of RATED THEMMAL POWER, and c.

Initially and-at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

1 HOPE CREEK 3/4 2-1 Amenoment No. 3

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i DOWER 0:5'R*BUT:CN LIMITS i

3/4.2.3 HN:wuM CRITICAL POWER RAT!O

. uIT*NG : ND: :0N FOR OPERATION

~ gpec;f?e in tac CORE !

3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater gggg AgD tnan tne % fR 1imit dshewa ia rigu e 2.2.2-1 er c'gure 2.2.2-2, :: :::!'::t!: A t4me: the

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t 4 = 0.86 seconds, control rod average se m insertion time limit to notch 39 per Speciff tion 3.1.3.3,.

N g = 0.688 + 1.65[ 1 b ),

t J (0.0 n

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

of RATED THERMAL POWER.

HOPE CREEK 3/4 2-8 Amendment No. 15

POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION ACTION:

a With tne end-o[-cycle recirculation pump trip system inoperable per Spe-

.N cification 3.3.4.2, operation may continue and the provisions of Speci-fication 3.0.4 are not applicable provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is

' determined to be greater than or equal to the "C"R 1,icit :: : f un:tf:n :f-4 r[ g g Nc 7p; /

e th: :.:r ;: ::r-- W :h:r '- 9 ; r: 2.2.2-1 r 9 ;r: 2. 2. 2-E, n e un the CORE j :::11 :t!:, EOC-RPT inoperable 7 svever44 :: the K7 :h:wn '- "f;r: 3.2.2-2.4 gy jS)0 / b. Witn MCPR less than the applicable MCPR 14ett-sh:wn f r rigere: 3.2.3-l'-

) and 0.2.0-2, ti::: -the K' :h:wn in "ign: 3. 2. 3-O'- initiate corrective OPERATM l action within 15 minutes and restore MCPR to within the required limit l lM IT'S within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

( (2EPORf _

SURVEILLANCE REQUIREMENTS 4.2.0 "C"R, with;

a. : - 1r0-pe4er-to-perfe- :n:: Of- the initial : rr.e-t4*:c-meeetreements-fer the cycle fa :::M :: ef t' 5;;;if t::ti:n '.1. 3.2, Or

b. ; :: deft;hd in 0;;;ificati:n 0.2.0 u::d to det ;in; the limit w+ thin 72 h gr: Of th: : nclu:ica Of each-scree-t4me-sveve4444nc+

test -required-by S;;;i'i;::ti:n t.1. 2. 2,-

%p)cPR-- -+shall be determined to be equal to or greater than the applicable MCPR Heit

F4 sete mte: e f :: rigue: 2.2. -1 :r ei;=: :.2.2-2, ti=:: tne x, n:wn in 9ttre 2. 2. 0-3; "- :

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a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of t THERMAL POWER increase of at leastr150of ' RATED THERMAL POWER, and...

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R00 PATTERM for MCPR.

d. The provisions of Specification 4.0.4 are not applicable.

HOPE CREEK 3/4 2-9 Amendment No.15

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VERSUS TAU AT RATED FLOW EOC-2000 MWD /ST TO EOC

/

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HOPE CREEK 3/4 2-11 Amendment No. 15 l

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POWER DISTRIBUTIP LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE

~

LIMITING CONDITION FOR OPERATION

+Lekm;V 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.Ow/fte APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. !

ACTION:

With the LHGR of any fuel rod exceeding the limi , initiate corrective action within 15 minutes and restore the LHGR to withi the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED HERMAL POWER within the next i 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . g g;,oj M g,(ggg 9gg,g,j -

AND OPE (2AT~l/V6 LIMITS 2EPol?r, SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than the limiti

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL R0D PATTERN for LHGR.

d. The provisions of Specification 4.0.4 are not applicable.

4 i

l l HOPE CREEK 3/4 2-13 l

3/4.2 POWER DISTRIBUTION LIMITS BASES -

The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondaH1.,

on the red to rod power distribution, within an assembly. The peak clad temperatu*e is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the hi powered rod divided by its local peaking factor. The limiting value for APLHGR is ism 4 4n\Ftoutes\3.VN-L 1.2N-2.\ A AIth 1.VNk4Naed LY.X4k The calcu'lational procedure used to establish the APLHGR (showtopimeN

~

[pa4N\

accident \WKIN%1.MM.

analysis. TheMh2NN4VmA A AMlis analysis was performed based using on a General 1oss-of-coolant Electric (GE) cal-culational 10 CFR 50.

models which are consistent with the requirements of Appendix K to A complete discussion of each code employed in the analysis is, pre-sented in Reference 1.4 u1TTe a

nces p Inis nalysis compare T.o pr vious ipes n be rok down sfollgws.

I ut anoe

/

1. Corr cted periza on Cal ulation Coeffi ents i the poriz tic co elati n used n the FLOOD e de were orrect .

2. ncor rated re acc ate byp ss areas '

The b ass eas i the top uide w e recal ulated ing a no e accur e te nique

3. rrecte guide e the' 1 resist ce. j

4. Corre ed hea capacityofreaft intern s he node . / f g(tET6

' Effc.fice iM THE ...SfE M ED WTM CoArc DOrl b d AND CCM SC3iG4 IWD of ERATi+36 wit 5 C Ats uoG UNTJ 65 f o W. .

%GCRL i

I i

HOPE CREEK B 3/4 2-1

, , POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATEY(Continued Mode / Chang /- '

Co CCFL essure fferent' 1 - 1 p - Ince orate at pressu flo from th bypass lower enum mu- overco e assur tion drop i core. a1p

2. I porate C pres e SAFE- FLOOD e trans e assum ion - T assump on us in essure t nsfer w> n the pr ssure i increa ng was cha ed.

T eseAchan few sofare he1chated esbelo affee

. the acci ent cale ation ir espect e of N FL. '

a. I ut Chan

1. B ak Areas The DBA reak are was cale ated m e acc rately.

b. Mo 1 Chance ~

1. Imp 'ved Radi ion and C duction 1culat n-I orporat' n of C STE 05 f heatup c culation.

A st of th significa plant in t para ters accid t ana.lysi is prese ed in Bas the lo of-c lant Table 3.2. .

For plant operation with single recirculation loop, the MAPLHGR limits

/Theconstantfactor0.86isderivedfromLOCAanalysisinitia loop operation to account for earlier transition at the limiting fuel noce compared to the standard LOCA evaluations.

3/4.2.2 APRM SETPOINTS The fual cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The flow biased simulated thermal power upscale scram setting and the flow biase neutron flux upscale control rod block trip setpoints must be adjusted to ensure that the MCPR does not become less than the fuel cladding Safety Limit or that

> 1% plastic strain does not occur in the degraded situation. The scram set- l points and rod block setpoints are adjusted in accordance with the formula in Specification 3.2.2 whenever it is known that the existing power distribution would cause the design LHGR to be exceeded at RATED THERMAL POWER.

5fECiFiED IN THE Cui Dess%n Anb cAMMg uM.Ts 7Ifc.tT HOPE CREEK B 3/4 2-2 Amendment No. 3

1 1

Esses *a:'e 2 3 2 ; *.

5::N::::N- lND.- Da# N E E45 "': -E t.:53 : :: :. NT C:: EN* ANA.#$!5 )

l

'a-- a amete-s

rb T*ERMAL s PO'wER .

p 3430 Mwt" w. .cn c:r es:: :s

\ \ i t

to 105% of atec steam :-

vessel $ team Outout . ,l . . . . .

. Ia.87 x .05 lem/nr n':-

corre nas to 105% e' ate:

N ,

stea flow x

vessel Steam Dome Pressure. ! . .. 1 psia Oesign Basis Re culation Liop Break Area for:

a. Large Breakt. 4.1 ft2;

b. Small Breaks 09fty

Fuel Parameters: '

~

" "_ f ":L J;T ,

i n i i 1.=t SP CATION DESIGN MINIMUM NE EAT AXIAL FUEL BUNDLE CRITICAL NER iTI RATE PEAKING POWER FUEL TYPE GEOMETRY 'kw/ FACTOR RAT!7 Indtial Core 8x8 [ .3. 4 k 1. 4 1.20**

l A more detailed listin of input of :each model d its source is presentec in Section II of Refe nce 1 and su ,section 6.3. of the FSAR.

"This power level ets the Append EKrequiremen heatup calcula n assumes a bund a power consiste f 1025. The core the hignest with operation of red red at 102% o its Technical 5 fication LINEAR HEAT GENEAAT RATE limit. '

g,a

- For si recirculation loop ope' ation, loss of nucleat '

at 0.1 s after LOCA regardli ss of initial MCPR. oiling is assumec 3.

t

-(

(

t.:.,

/

Aw HOPE CREEK B /4 2-3 * .

j Amendment No.3 l

/

i h6L6(6 ENflUE f/16E '

o

~

30'wER :: 3*3!!L*:CN _!WI*$

BASES 3'A 2 w:N'utM :R!?! CAL DOWER RATIO

-me recuired operating limit MCPRs at steacy state ocerating concitions as specified in Specification 3.2.3 are cerivea f rom tne established fuel c'accing integrity Safety Limit MCPR, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation =1tn t.ne initial condition of the reactor being at the steady state operating limit, 't is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any2.2.

Specification time du' ring the transient assuming instrument trip setting given in To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational trarsient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained.

The evaluation of a given transiellt begins with the system initial parameters shown in FSAR Table 15.0-3 that are input to a GE-core dynamic behavior transient computer program. The code used to evaluate pressurization events is described in NEDO-24154(3) and the program used in non-pressurization events is described in NEDO-10802(2) The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC code described in NEDE-25149(4) .

The principal result of this evaluation is the reduction in MCPR caused by the transient.

/ mi Tep pose of t at the tha rat flo th reg red K fac r of igu ce fl co itio

3. 3-3 to less han fine perati g 0% of ated j

PR st pr uct th MCPR nd t K ctor. he K f tor ass re t t e5 ety mit PR ner se ans nt sul ng on a to gene tor eed ntrol ailurf1 11 t be tolafdduri a {

j Th K g f ctor na e plie to th ual nd au osati flow ntrol des he fa ors al sh in igur 3.2.3 an are h1 abl to a P 2, 8 /3 a BWR/ reawer ors, devel ed ge rica y eK actor wer f

l d ive usi t fl cont I li se cor espon ng t RATED ERMA POWE at ate cor fl .

l F I th man 1 f1 con el e, t K f tors re ca ulat d su that or e in flow ate, s 11 ted b the g j sco tube set int d the cor espo ding HE alo the ated ow co rol 1 e, t li ting l

b die' rol tive ower as a usted ntil he MCP chang wi dif rent re low Th rati of t e MCP cale ated a giv n poi of re f ow, di iced j y le o rati lim MCP dete ines eK.f )

/

HOPE CREEK B 3/4 2-4 Amendment No. 15 l

'o POWER DISTRIBUT!CN LIMITS BASES "INIMUM CRITICAL PCWER RATIO (Continued) e au 'omati flow contr 1 mod , tne same roce re . 5 fc op att in I

/a D i o a e cept he i itial powe distr outio was stabl shed uch *.at t e CPR was qual to tb oper ting imit CPR th mal flow.

RAT THE AL P ER a rat d he fact rs s wn i Fig'ur 3.2. -3 ar cons rvati e for the ner 1 I limi MCP Ele ric ant pera on b cause he o rati of 5 ecif' ati n

3. 3 i the me a the rigin 1. 2 oper ing mit PR u d fo th neri deri atto of K . .

At THERMAL POWER levels l'ess than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed ano the moderator void content will be very small. For all oesignated control red patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are* very slow when there have not been significant power or control rod changes. The require-ment for calculating MCPR when a limiting control rod pattern is appro' ached ensures that MCPR will be known following a change in THERMAL POWER or power

, shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NEDO-10802, February 1973.

3. Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors, NED0-24154, October 1978.

4. TASC 01-A Computer Program for the Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.

HOPE CREEK B 3/4 2-5 Amendment No. 15

)

DESIGN FEATURES m

5.3 REACTOR CORE The fuel aggedfieg pa ffe reacNr core are des;cr; Led in t/c Co2E DESIGN MD FUEL ASSEMBLIES J OPERATING LIMrTS REPOET 5.3.1 The reactor core shall contain 764 fuel assemblies ith :::h fuel-e-4ssembly containing-62-fal red and tee water red c!:d 'ith Zirc 1cy-2."-

E::F feel red :h:11 hav : n =feal-ec-t4ve-fuel-length-of-15G-inche:."- The initial core loading shall have a maximum average enrichment of 1.90 weight percent U-235. Reload fuel,shall be similar in physical design to the initial core loading.

a5 descr,% eof it; fle CORE DESIGAl CONTROL R00 ASSEMBLIES M0 OPENTTA/G l.rMETS RsPoRT 5.3.2 Thereactorcoreshallcontain185controlrodassemblies,l:::ht cen:f; ting of : crc 4f: : :rr y Of :teinle:: : teel tube; centaining 143' inch:: Of beren ::rbid:, 2 0, pr d:7 :;rr : t O : crucif: = 2.:p:d '-

4

tcinle:: : teel :h::th a 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of:

1. 1250 psig on the suction side of the recirculation pump.

2. 1500 psig from the recirculation pump discharge to the jet pumps.

c. For a~ temperature of 575"F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 21,970 cubic feet at a nominal steam dome saturation temperature of 547*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

HOPE CREEK 5-4

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

The radioactive effluent release report to be submitted within 60 days after January 1, of each year shall also include an assessment of radiation cases to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent path-ways and direct radiation) for the previous 12 consecutive months to show con-formance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribu-tion from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.

The radioactive effluents release shall include the following information for each class of solid waste (as defined by 10 CFR 61) shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Type of waste (e.g., spent resin, compact dry waste, evaporator bottoms), .

e. Type' of container (e.g., LSA,' Type A, Type B, Large Quantity), and

f. Solidification agent (e.g. , cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to the UNRESTRICTED AREA of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP), OFFSITE DOSE CALCULATION MANUAL (00CM) or radio-active waste systems made during the reporting period.

MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Occu-ment Control Desk, Washington, D.C. 20555, with a copy to the Regional Admin-Td6E 2T istrator of the Regional Office no later than the 15th of each month following the calendar month covered by the report.

8 NPECIALREPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator of the Regional' Office of the NRC within the time period specified for each report.

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 ,

submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator of the Regional Office of NRC via the Licensee Event Report System within 30 days.

HOPE CREEK 6-20 Amendment No. 21 9

l Insert B:

CORE DESIGN AND OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the PSE&G generated CORE DESIGN AND OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in the most recently approved version of the General Electric Standard Application for Reactor Fuel (GESTAR II) . The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE DESIGN AND OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

l o ,

l' 1;

1 ATTACHMENT 3 HCGS LICENSE CHANGE REQUEST 89-12, NLR-N89055 REMOVAL OF CYCLE SPECIFIC PARAMETER LIMITS FROM THE TECHNICAL SPECIFICATIONS CORE DESIGN AND OPERATING LIMITS REPORT (CONTAINS GE PROPRIETARY INFORMATION)

_ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ . _ _ . . . _ __ _ _ _ _ _

ML20247C310

Text

References:

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