ML20217E045

Proposed Tech Specs Supporting Use of ABB-CE Fuel & Reload Analyses Beginning with Upcoming Cycle 10

ML20217E045
Person / Time
Site:	Hope Creek
Issue date:	09/30/1999
From:	Public Service Enterprise Group
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ML20138G239	List: ML20217E037 ML20217E045
References
NUDOCS 9910190058
Download: ML20217E045 (28)

Text

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2.0 SAFETif CHITS AND LIMITING SAFETY SYSTEM SETTINGS

=.................. 2........................................................,

2.1 11FETY LIMIT 3 Tur m_L POWER Low pressure or Low Flow 2.1.1 THERMAL POW'a shall not exceed 25% of RATED THERMAL POWER with~the reactor vessel steste dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACIEEt With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification s.7.1.

'RE7 LACE Ww It45E'it.x A TN#84AL POWER. Hioh Pressure and Micrh Flow _p-x

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APPLICARILIT7: OPERATIONAL CONDITIONS 1 and 2.

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REACTOR COOLANT SYSTEM PRRERtIRR 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICARILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

&GII.QEt With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 peig, be in at least NOT SHUTDOWN with reactor coolant' system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

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i HOPE CREEE 2-1 Amendment No.

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t DocumInt Central Desk LR-N99429 AttachmInt 3 LCR H99-09 lNSERT A 2.1.2 With reactor steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow:

The MINIMUM CRITICAL POWER RATIO (MCPR) for GE fuel shall be 21.10 for two recirculation loop operation and shall be 21.12 for single recirculation loop operation.

The MCPR for ABB/CE fuel shall be 21.10 for two recirculation loop operation and shall be 21.13 for one recirculation loop operation.

INSERT B With reactor steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow and the MCPR below the values for the fuel stated in LCO 2.1.2, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

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I The fu cladding, reactor pressure ve isel and primary system the princip 1 barriers to the release of r iping are ioactive materials to e her g

environs, safety Limits are established protect the integrity o these pd g barriers dt. ring normal plant operations d anticipated transients. The fuel dding integrity safety Limit is set a ch that no fuel damage is ca culated to r if the limit is not violated.

ecause fuel damage is not di tetly observ le,g:g_en-ba_ck_approachisuse to establish a safety Limit si rh

• hat the MCP is

_ lj j M _150dfor two, fo single rsci recitenlati_on loop op_ oration _and_h operatiodf % M,~:Es: 9--J f 0; 3s n ;~

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gq :n;;g wuservative margin relativei to the consitions required to - ~

1 cladding integrity. The fuel cladding is one of the physical maintain r barriers w ch separate the radioactive materials from the environs.

The integrit of this cladding barrier is related t9 its relative freedom from perfor ions or cracking. Although some corrosion or use related cracking may during the life of the cladding, fission produc'; migration from this oc rce is incrementally cumulative and continuously measurable.

s Fuel cladding erforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

Thase conditions represent a significant departure from the condition nraad-d bz da.4 crn 1.or nimanad_ ooeration.

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%,h ot The use of the applicable NRC-approiTeTeritical power :,r:2_1"irr i valid for all critical power calculations performed at reduced pressures colow g

785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 pai. Analyses show that with a bundle flow of 28 x 108 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 2e x 108 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

HOPE CREEK B 2-1 Amendment No.

l

1 SAFETY LIMITS BASES 2.1.2 THERMAL POWER. High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel daange is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure free nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to 8WR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient liett.

However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.95 of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and in the procedures used to "

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General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A

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(The approved revision at the time the reload analyses are performed.

The approved revision number shall be identified in the CORE OPERATING LIMITS REPORT.)

2.

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1 HOPE CREEK g23 Amendment No. 42

Bases reble 82.1.2 1 O

u Ciar.1 tris vie 0 1. r Orr.R 1.110.

0F rHE FUEL CLADOING SAFETY LIM!ra Standard Deviati Quantity (5 of P int)

Feedwater Flo" 6

Feedwater Temperature 0.76 Reactor Pressure

0. 5 Core Inlet Temperature 0.2 Core Total Flow Two Recirculation Loop operatfon 2.5 Single Recirculation Loop Opdration 6.0 Channel Flow Area 3.0 Fricd on Factor Multiplier

' 10.

Channel Friction Factor Multiplier 5.0 TIP Readings Two Recircul ion Loop Operation 8.7 Single Reciremiation Loop Operation 9.1 R Factor 1.6 l

Critical Power 3.6

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i "The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of goedrent power symmetry for the reacter core. The values herein apply to both two recirculation loop operation and single recirculation loop operation, except as noted.

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HOPE CREEK S 2-3 Amenennt No.15 W l 51088

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

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O LIMITING CONDITION FOR OPERATION 3.2.; All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE 1n exce the J '

s 'arified in iha ennF C PERATING LIM SR R

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sect c;;1 tica 1:q :;;r tie.- H=4t wh*' ia <iagle recirculati:n 1::; - :ratioW APPLICA8ILITY:

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

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ACTION:

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

O SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the lieits specified in the CORE OPERATING LIMITS REPORT:

l a.

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

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b.

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

c.

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

d.

The provisions of Specification 4.0.4 are not applicable.

O HOPE CREEK 3/4 2-1 Amendment No. 34 00( 2 0 1999 l

i POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT. e T) t=( v B

where:

TA = 0.86 onds, control rod avera e cram insertion time lie to notch 39 per Spec ication 3.1.3.3, N,

g = 0.672 + 1.6 y 3 (O.

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n I

t,y,=

i=1 Nt y

i n = number o surveillance tests perfomed to

.te in cycle, b

th N=n t' of active control rods measured in t i

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s welliance test, tg = average scran time to notch 39 of all rods measure in the i surveillance test, and

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th N = total number of active rods measured in Specif Mation APPLICABILITY:

OPERATIONAL COMITION 1, when THERMAL POW 241s greater than or equal to 25%

of RATED THERMAL POWER.

ACTION:

l a.

With the end-of-cycle recirculation purp trip system inop:1rable per Spe-l 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,. N PR is i

determined to be greater than or equal to the E0C-RPT inoper ele limit specified in the CORE OPERATING LIMITS REPORT.

I HOPE CREEK 3/4 2-3 Amendeont No.47

POWER DISTRIBUTION LIMITS LIMITING C0lGITION FOR OPERATION (Continued)

ACTION: (Continued) b.

With MCPR 1ess than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT, initiate correctiv POWER to less than 25% of RATED THERMAL POWER SURVEILLANCE REQUIREMENTS 4.2.3 MCPR _ th:

t = 1.0 pr for the cy% performance of the initial scram time measuriene ts cle7 ccordance with Specifica* e 4 f 3.2, or b.

I as defined in Specificat ed to determine the limit r.

within 72 houri-of-th~

ast M d by Specification 4.1.3.2, conclusion of eactriccan time surve N

shall be determined to be equal to or greater than the applicable MCPR Ifait specified in the CORE OPERATING LIMITS REPORT:

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

b.

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

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is c.

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operating with a LIMITING CONTROL R00 PATTERN for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

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O h0PE CREEK 3/4 2 4 Amendment No.34 g,.;.

OCT 2 0 889

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4 3 /_4.' 4 REACTOR OCOLANT, SYSTEM '

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3 3 / 4. 4.~ 1 RECIPCULATION-SYSTEM

.PECIRCULATION LOCPS

-LIMITING CCNDITION FOR OPERATION'

3. 4.1.1=-

Two reactor. coolant system recirculation loops,shall be in operation withs.

Total core flow greater than or equal to 454 of rated core flow, a.

or tb.

THERMAL PCWER lets thart or equal to the limit specified in Figure 3.4.1.1-1.-

' APPLICABILITY:' OPERATIONAL CONDITIONS 1* end 2*.

$(ed be% Yo tb Q( DfHn10 Air fV\\

ge gy fg teMC.4 ACTION:

o cey?OS a.: -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:

.al Place the recirculation flow control system in the Local Manual mode,..and b)

Reduce THERMAL POWER to s 70% of RATED THERMAL POWER, and z

.h c)

!IncreaseLthe MINIMUM CRITICAL POWER RATIO (MCPR) Safety t

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Limit per Specification 2.1.2, and d)

Reduce the Maximum ~ Average Plana ear He t Generation a

PatA iMAPLHGR) limit to-a 3aluar f 0. 00 nm..

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G. u.4 f *2 Y"'

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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;38% of PATED THERMAL POWER or the recirculation loop. flow in the' operating loop is s'50% of rated loop flow.

L

'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 (APRM) 2..

Scram Trip setpoints and Allowable Values to those applicable for single recirculation' loop operation-per Specifications l-2.2.1 and'3.2.2 otherwise, with the Trip Setpoints and Allowable. Values-assoc!.sted with one trip system not reduced to

.those applicable for s.,ngle recirculation loop operation, place the affected trip syster 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 />, rsduce the Trip Setpoints and Allowable Values of the affected channels to those applicable for single 1

recirculation 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 APRM Control Red diock Trip

• See Special Test Exception 3.10.4.

HOPE CREEK 3/4 4-1' Amendment No.107

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REAC"TIVITY CONTROL SYSTEMS p

BASEE v

3/4,1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertic a sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous,.. scattered patterns of control rod. withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there ia no possible rod worth l

which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RWM to be OPERABLE when THERMAL POWER is less th 3 or equal to 10% of RATED THERMAL POWER provides adequate control.

I The RWM provides automatic supervision to assure that cut-of-sequence rods l

will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the t hn a of h a t_opical r y Reference 1, aid swo suppl __ nt.,

.-__er:20er ? =ad & Julditionarl-pertinent 1

anely.is is also w.. seined in T--

'- as 17-te *h=

==fer-ce 4 topical re ert.A v

The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of 3

control rods.

HOPE CREEK B 3/4 1-3 Amendment No.105 e

b l-l REACTIVITY CONTROL SYSTEMS i

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RASES.

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rate, solution concentration or boron equivalent to meet the ATWS Rule must not invalidate the original system design basis. Paragraph (c) (4) of 10 CFR 50.62 states that:

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"Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron control equivalent in

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control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution (natural boron enrichment)."

The described minimum system parameters (82.4 gym, 13.6 percent concentration and natural boron equivalent) will ensure an equivalent i

injection capability that exceeds the A' dis Rule requirement, The stated l

minimum allowable pumping rate of 82.4 gallons per minute is met through the l

simultaneous operation of both pumps.

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O HOPE CREEK B 3/4 1-5 Amendment No.105 l

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3/4.2 POWER DISTRIBUTION LIHITS BASES fM" c Tht tpee4f4cationutf_this section assure that the :aWTdHTng 4

emperature following the co<tuTat wi i no excgad -t ident CFR 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak claddfng temperature 7 Fro 11owing ah nt is primarily a function of the average heat generation r

'of all ce the ro f a fuel assembly at any axial location and is de M od power distribution within an assembi nt only secondari y on the rod is calculated as ng a LHGR for the highest po rod which is equal to or e peak clad temperature i

less than the design R corrected for den cation.

This LHGR times 1.02 is used in the heatup co lon with exposure dependent steady state gap conciuctance and rod-to rod lo ng factor.

The Technical Spectfication AVERAGE PLANAR LINEAR HEAT G RATE (APLHGR) is this LHGR of the highest powered rod divided by i ocal peak.' factor.

The lieiting value for APLHGR is specified in the OPERATING LIMI"5 ORT.

l The ca ational procedure used to establish APLHGR speciffed in the CORE OP ING LIMITS REPORT is based on a loss-of-coo accid ~nt analysis.

Theprtlysis was performed using General Electric (GE) calc e

tional models Q

which are consistent with the requirements of Appendix K to 10 50.

A complete disqussion of each code employed in the analysis is presented in Reference 1.

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HOPE CREEK B 3/4 2-1 g Amendment No. 34

/

Document C:ntrol Desk LR-N99429 Atta:hm:nt 3 LCR H99-09 INSERT C The specifications in this section help assure that the fuel can be operated safely and reliably during normal operation. In addition, the limits specified in these specifications help ensure that the fuel does not exceed specified safety and regulatory limits duringanticipated operational occurrences and design basis accidents. Specific, t these limits:

1. Ensure that the limits specified in 10CFR50.46 are not exceeded following the postulated design basis loss of coolant accident.

2. Ensure reactor operations remains within licensed, analyzed power / flow limits.

3. Ensure that the MCPR Safety Limit is not violated following any anticipated operational occurrence.

4. Ensure fuel centerline temperatures remain below the melting temperature and peak i

cladding strain remains below 1% during steady state operation.

JNSERT D The AVERAGE PLANAR LINEAR HEAT GENERAT_ ION RATE (APLHGR)is a measure of the average Linear Heat Generation Rate (LHGR) of all the fuel rods in a fuel assembly at any axial location. The Technical Specification APLHGR is the LHGR of the highest-powered rod divided by its local peaking factor. Limits on the APLHGR are specified to ensure that the fuel design limits are not exceeded. The limiting value of the APLHGR limit is specified in the

. CORE OPERATING LIMITS REPORT. The calculation procedure used to establish the APLHGR is based on a loss-of-coolant accident analysis. The post LOCA peak cladding temperature (PCT) is primarily a function of the APLHGR and is dependent only secondarily on the rod to rod power distribution within an assembly. The analytical modes used in evaluating the postulated loss-of-coolant accidents are described in References 1 and 2. These models are consistent with the requirements of Appendix K to 10CFR50.

For plant operation with a single recirculation loop, a lower value for the APLHGR limit is specified in the CORE OPERATING LIMITS REPORT. This lower value accounts for an earlier transition from nucleate boiling which occurs following a loss-of-coolant accident in the single loop operation compared to two loop operation.

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POWER DISTRIBUTION LIMITS O

BASES i

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) lant specified in eration with s ngle recirculation loop, the MA LIMITS REPORT are mul Tniits i

constant factor 0.86 is derived y 0.86.

The l

ro loop operation to account-_for-ee initiated from single l

er transition a

% 1 nod c

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3/4.2.2 APRM SETPOINTS The fue', cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGL 4t. RATED THERMAL POW The flow biaseu simulated thermal power-upscale scram setting and the flow bia 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-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.

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HOPE CREEE B 3/4'2-2 Amendesnt No. 34 L

1 POWER DISTRISUTION LIMITS BASEf 3/4.2.3 MINDRSD CRITICAL POWER RATIO -

22~-

The required operating limit MCPRs at stea@ state operettng conditions as specified in Specificat' en 3.2.3 are derived from the established fusi cladding integrity Safety Limit MCPR, and an analysis-of abnormal operational transients. For any abnormal operating transient analysis, evaluation with the initial condition of the reactor being at the stea% state operating iteit, it is required that the resulting MCPR does.not decreatenbelow the Safety Limit MCPR at any time during the transient assuming instrument. trip setting given in Specification 2.2

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...yg 3,

1 To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnores) operational transient,.the most lietting tran-sients have been analyzed to determine which result in the largest. reduction in I

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 reentred minimus operating limit MCPR i

of Specification 3.2.3 is obtained.

g The evaluation of a given transient tietins wi'~}h r

tas initial pare-i meters shown in FSAR Table 15.0-3 that are 'nput to re $namic behavier ',

transient computer program. The codes used to eval ransienta art discussed in Refe 2.

W purpose of the K, facter specified in the CORE OPERATINE LIMITS is to operating liMts at other than rated core flow conditions.

ess than 100K o tad flow the required MCPR is the product of the operati limit MCPR and the K, r.

The K factors assure that the Safety L t

will not be violates dur a flow, increase transient resulting f a motor-generato speed control failure.

Kg acters may be applied to manual and auta-f natic flow control sedee.

The K, factors values specif1 the OPERATING LIMITS REPORT were developed generically and are applicabl all SWR /2, SWR /3 and SWR /4 reactors.

The K, factors were derived using low rol line corresponding to RATED THERMhL POWER at rated tore f1 The K,ined as a f facters are det ned in the following manner:

change in CPR is detere as of core flew along the rated power contre) line.

Then, for a gf scoop taba setpe'at in the manual flow contre ting mode, the IEPR flow is established that would give the Safety tICPR if the flow wee increased to the scoop tee setpoint.' The rette e t reduced flow to the operating liett MCPR is the K7 facter at that

flow, JW I M SE?LT E HOPE CREEK B 3/4 2-3 Amendment No. 42

Document Centr:I De k LR-N99429 Atta:hm:nt 3 LCR H99-09 INSERT E The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (MCPR, and MCPR, res sectively) to ensure adherence to fuel design limits during the worst transient wit 1 moderate frequency that is postulated in Chapter 15.

Flow dependent MCPR limits (MCPRr) are determined by steady state methods using a

)

three dimensional BWR simulator code (Reference 2). MCPR, curves are provided based on the maximum credible flow runout transient (i.e., runout of both loops).

i A three dimensional BWR simulator code and a one dimensional transient code (Reference 2) determine power dependent MCPR limits (MCPRp). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram limits are bypassed, high and low MCPRp operating limits are provided for operation between 25% of RATED THERMAL POWER and the bypass power levels.

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l POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) l For operation in the automatic flow control mode, the same procedure is j

employed except the McPR at low flow is established such that the MCPR is equal to the operating limit MCPR at RATED THERMAL POWER and rated core flow.

The K factors specified in the CORE OPERATING LIMITS REPORT are conserva-tivebecauletheoperatinglimitMCPRsofSpecification3.2.3areequaltoor greater than the original 1.20 operating limit MCPR used for the generic deriva-tion of K.

g At THERMAL POWER 1evels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump-speed and the modera-tor void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates 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 l

at 255 of RATED THERMAL POWER 1evel with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this I

power level will be shown to be unnecessary. The daily requirement for calculat-ing MCPR when THERMAL POWER is greater than or equal to 255 of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have t

not been significant power or control rod changes. The requirement for calculat-ing MCPR when a limiting control rod pattern is approached ensures that MCPR l

will be known following a change in THERMAL POWER or power shape, regardless of l

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 gone ration even if fuel pellet l

I densification is postulated.

References:

l 1.

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

l 2.

{er t.r T mne -,,,...-..asion sw sm use 5= nuisau analyses an i.i. __.

i The approved revision number shall be identified in the CORE OPERATING l

LIMITS

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" T 6 p c42.e o c_c S A 4 3

TE? o 1iLT 4* r 3o m iw 4

\\.,OA h r 3 EAcTorf I EioAD b E L 1-HOPE CREEK B 3/4 2-4 Ameniteent No. 42 L

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3/4.4 MEACTOR COOLANT SYSTEM ansEs O

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3/4.4.1 RECIRCULATION SYSTEM-The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety Limit is increased as noted by specification 2.1.2, APRM scram and control rod block setpoints are adjusted as-noted in Tables 2.2.1-1 and 3.3.6-2 respectively. MAPLEGR limits are decreased by the factor given in specification 3.2.1, and MCPR operating limits are adjusted per specification 3/4.2.3.

Additionally, surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive-core internals vibration. The surveillance on. differential temperatures below 38%

THERMAL POWER or 50% rated recirculation. loop flow is to mitigate the undue thermal stress on vessel nossles, recirculating pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not in itself a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design.

basis-accident, increase the blowdown area and reduce the capability of reflooding the core, thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCs LOCA analysis design criteria for twst. recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nossles and bottom head region, the recirculation loop temperatures shall be within 50'P of each other prior to startup of an idle loop. The loop temperature must also be within 50'P of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nossles. Sudden equalisation of a temperature difference > 145'P between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

The objective of plant and fuel design is to provide stable operation with margin over tra normal operating domain. However, at the high power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod pattern, power shape). To pgovide assurance that neutron finx limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flum noise levels should be monitiored while operating in this region.

1 BOPE CREEK B 3/4 4-1 Amendment No. 63 ll

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3/4.a REACTOR 000LANT SYSTEM

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5 tact 11ty tests at operating BWRs were reviewed to determine a generic _

region of the power / flow map in which surveillance of neutron flux noise lese should be perforued.

A conservation decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant te plant variability of decay ratio with core and fuel designs.

kas been determined to correspond to a core flow of less than or equal to 4 %ghn This geneM c re of rated core flow and a THERMAL POWER greater than that specified in Figure 3.4.1.1-1.

Plant specific calculations can be perforts region for monitr-r neut- - flux noise levels to determine an applicable M this case the degree of conservatism can educe-ice plant to plar..

riability would be eliminate::.

In this case, ade a decay ratio gree.r than or equal to 0.2.e marr will be assured by conitoring the region whica nas of limit cycle neutron flux oscillations. Neutron flux noise limits are also establis neutron flux noise caused by random W 11ng and flow noise.BWR cores typically ope Typical neutron flux noise levels of 1-125 of rated L. er (peak-to peak) ha.e been reported for the range of low to high recirculati:.a loop flow during both single and dual recirculation loop operation.

bound these values are considered 19 the thermal / mechanical design oN fuel and are found to be of neglig', e consequence. In addition, sta rty teTts l

at operating BWRs have demonstrates that when stability related neutron flux limit cycle oscillations occur they result in peak-to peak neutron flux limit O

i cycles of 5-10 times the typical values.

Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the tyo Mat value are suf-ficient to ensure early detection of limit cycle reutron f *.a oscillations.

Typically, neutron flux noise levels show a m dual inrease in absolute magnitude as core flow is increased (constant control rod patte m ) with two reactor recirculation loops in opera 9 on. Therefore, the base::ne neutron f 6 noise level obtained at a specific. e flow can be applied over a range of : ore flows.

To maintain a reasonable var:ation between the low flow and high flow ead of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in operation.

Data from tests and operating plants indicate that a range of 20E of rated core flow will result in apprcximately a 505 increase in neutron flux noise level during l

operation with two recirculation loops. Baseline data should be taken near the maximumrodlineatwhichthemajorityofoperationwilloccur.

However, base-line date taken at lower rod lines (i.e., lower power) will result in a conser-vative value sir.co the neutron flux noise level is proportional to the power l

level at a given core flow.

I 3/4.4.2 SAFETY / RELIEF VALVES i

l The safety valve function of the safety / relief valves operates to prevent i

the reactor coolant system from being pressurized above the Safety Limit of 1*'5 psig in accordance with the ASME Code.

A total of,13 OPERA 8LE safety / relief I

HOPE CREEK 3 3/4 4-2 Amendeont No. 3 APR 7 sp

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ADMINISTRATIVE CONTROLS.

CORE ODERATING LIMITE REPORT (Continued)

The analytical methods used.to determine the core operating limits shall be those previously reviewed and approved by NRc in R.,a-hvA4-7-J. 'the late:N approvetrrevistor0T'- Genera 141ectric-StandakApplisation-for-Reactor-Fuel A l

NE8M M "

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4 The core ornating limits'shall belietermined so that all applicable limits (e.g., fr.41 thermal-mechanical limits-, core thermal-hydraulic limits, ECCS limie, michar. limits. such as shutdown margin, and transient and accident armlysis limits) of the safety analysis are met.

The CORE 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.

i SPECIAL REDORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Comnission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, 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. Nucitar Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the USNRC Administrator, Region 1, via the Licensee Event

/

Report System withl'1 30 days.

6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

SPECIAL REPORTS 1

6.10.2 The following records shall be retained for at least 5 years:

Records and logs of unit operation covering time interval at each a.

power level.

b.

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

All REPORTABLE EVENTS submitted to the Commission.

c.

d.

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

Recordas of changes made to the procedures required by Specification e.

6.8.1.

f.

Records of radioactive shipments.

Records of se led source and fission detector leak tests and results.

O1

• 75YvrET.. avalumsed-te tM 9-fety Wal"-ti-datadmygggto-support L.-

e nse Amen 5 SEEP.t. no.117 W

HOPE CREEK 6-21 Amendment No. W.Il7 l

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l NEN

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MAJOR CHANGES TO RAD!a**TIVE LIOUIO. GASE005 " SOLIO WAST 5Yu r.nz (continues).

A stumery of the evaluation that led to the detemination that e.

the change could be made in accordance with 10 CFR 50.59;

[

I b.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; A detailed description of the equipment, components and c.

processes involved and the interfaces with other plant systems; d.

An evaluation of the change, which shows the predicted.c releases of radioactive materials in liquid and gaseouer effluents and/or quantity of solid unste that differ frase-those previously predicted in the license application and ;

amendeants thereto; An evaluation of the change, which shows the expected maximuun e.

exposures to individual in the unrestricted area and to the noneral population that differ from those previously estimated

< n the license application and amendments thereto; I

f.

A comparison of tre predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; g.

An estimate of the exposure to plant operating personnel as a result of the change; and h.

Documentation of the fact that the change was revieued and found' acceptable by the 50RC.

2.

Shall become effective upon review and acceptance by the 50RC.

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6 E 6 V\\ C. E t c.e wed - %. 7. A, " W e t e_ me S a Oa Te pA i.

G % d u s tM e e Te a ch e s '~R e t oa d.

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Metc -won 7-A C l e4.e s t' ye t d v evi s. W F o d ( C3E STAR -[g ehc d L f E tt ct ri e b ^asx:us A

Gent 2At rE u t,or HOPE CREEK 6-25 Amendment No. 52 t

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Docum:nt C:ntrol Desk LR-N99429 Atta*hm:nt 5 LCR H99-09 l

NON-PROPRIETARY VERSION OF ATTACHMENT 1 TO LCR H99-09 L.

Documsnt Crntrsi Desk -

LR-N99429 Attachm:nt 5 LCR H99-09 HOPE CREEK GENERATING STATION REVISIONS TO THE TECHNICAL SPECIFICATIONS (TS)

BASIS FOR REQUESTED CHANGE:

i

' The changes proposed in this request support the operation of Hope Creek in Cycle 10

)

with a mixed core of General Electric and Asea Brown Bovieri/ Combustion Engineering

)

(ABB/CE) fuel. The proposed changes implement appropria_tely conservative Safety Limit Minimum Critical Power Ratios (SLMCPRs) and revise the Technical

~ Specifications (TS) for Average Planar Linear Heat Generation Rate, Minimum Critical Power Ratio, and Recirculation System to reflect the use of approved methodologies for calculating core operating limits.

REQUESTED CHANGE:

'As shown in Attachment 3 of this letter, TS 2.1.2 is being modified to: 1) define a 1.10 SLMCPR for dual recirculation loop operation to be applicable to the GE9B fuel and to add a 1.10 dual recirculation loop operation SLMCPR applicable to the ABB/CE fuel; 2)

' define a 1.12 SLMCPR for single recirculation loop operation to be applicable to the GE9B fuel and to add a 1,13 single recirculation loop operation SLMCPR applicable to the ABB/CE fuel; and 3) to remove the cycle specific footnote for the Safety Limit applicability. The Bases for TS 2.1, " Safety Limits", will be revised to reflect the new-MCPR limits for dual and single recirculation loop operation and to include the ABB/CE bases. Administrative changes are being made to TSs 3/4.2.1,3/4.2.3 and 3/4.4.1 and their associated Bases, and to TS 6.9.1.9 to reflect the use of approved ABB/CE methodologies for calculating core operating limits for Cycle 10 and subsequent cycles.

JUSTIFICATION OF REQUESTED CHANGES:

The proposed changes contained in this submittal will revise the Technical Specifications to reflect the new SLMCPR values established by ABB/CE for Hope Creek and to reflect the use of approved methodologies for calculating core operating limits. As stated previously, these plant specific evaluations were performed by ABB/CE for Hope Creek, Reload 9, Cycle 10 using NRC approved methods.

Introduction -

For Hope Creek, the Fuel Cladding Integrity Safety Limit is set such that no mechanistic fuel damage due to fuel cladding overheating is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in the onset of transition boiling have been used to mark the beginning of the region where fuel

' damage could occur. Although it is recognized that the onset of transition boiling would,

c 3 Docum:nt Centrol De:k LR-N99429 Attachm:nt 5 LCR H99-09 L

not necessarily result in damage to BWR fuel rods, the critical power at which boiling

. transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity safety limit protecting the cladding from overheating is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The new SLMCPRs for l

Cycle ;10 at Hope Creek are 1.10 for dual recirculation loop operation and 1.12 for single recirculation loop operation for the GE fuel and 1.10 for dual recirculation loop operation and 1.13 for single recirculation loop operation for the ABB/CE fuel.

Hope Creek Cycle 10 SLMCPR Analysis The approved methodology described in Reference 1 was used to establish the Cycle 10 SLMCPRs for the GE9B and ABR/CE SVEA-96+ fuel. Specifically, the treatment of SLMCPR in mixed cores involving non-ABB/CE fuel (())

j In accordance with this methodology, dual recirculation loop and single recirculation loop SLMCPRs of 1.10 and 1.12, respectively, (()). Hope Creek Cycles 7 through 9 were composed entirely of GE9B fuel. (()). The General Electric methodology for performing the Cycle 8 and 9 plant-and cycle-specific calculations for Hope Creek are j

addressed in References 3 through 5.

f The SVEA-96+ SLMCPR for Hope Creek Cycle 10 is based on a Reference Core loading pattern and state point depletion strategy realistically representing current plans for Cycle 10. (())

Specific single recirculation loop SVEA-96+ SLMCPR calculations were also performed for Hope Creek Cycle 10. Single recirculation loop SLMCPR calculations were performed (()) The calculation uses the same procedure as for dual recirculation loop operation, except (())

This evaluation of the SVEA-96+ SLMCPR (()) established maximum calculated dual recirculation loop and single recirculation loop SLMCPRs of 1.092 and 1.123, respectively. These limiting values were both established at a (()) Therefore, these calculations support SVEA-96+ dual recirculation loop and single recirculation loop

' SLMCPRs to two significant figures of 109 and 1.12, respectively, and could be used to conservatively represent the entire cycle. However, bounding SVEA-96+ dual l

recirculation loop and single recirculation loop SVEA-96+ SLMCPRs of 1.10 and 1.13,

- respectively, have been selected to reduce the probability of having to increase the l

SVEA-96+ SLMCPR in future cycles.

l.

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1 Docum:nt Centrol De:k LR-N99429 Attachm:nt 5 LCR H99-09 L Evaluation of the Hope Creek Cycle 10 SLMCPR value Table 1 provides a comparison between the proposed GE9B assembly and SVEA-96+

> SLMCPRs and relevant input parameters with the corresponding parameters in Cycle 9.

The Cycle 10 SLMCPRs were established by conservatively applying the approved methodology described in Reference 1 for the GE9B and ABB/CE SVEA-96+ fuel. (())

In general, for a given set of uncertainties, the calculated safety limit is dominated by two key parameters: (1) flatness of the bundle pin-by-pin power /R-factor distributions; and (2) flatness of the core bundle-by-bundle CPR distributions. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR. There is no reason to conclude that the Cycle 10 bundle and fuel rod CPR distributions are more limiting (i.e. flatter) than the corresponding Cycle 9 distributions. (())

Table 1: Comparison of Cycles 9 and 10 Safety Limits

- Quantity, Description Cycle 9 Cycle 10 i

i Number of bundles in core 764 764 Number of GE9B Assemblies in Core -

764 532 Number of SVEA-96+ Assemblies in Core None 232 Projected Cycle Burnup (End of Full Power 372 458 Life)

Effective Full Effective Full Power Days Power Days Limiting cycle exposure point See Proprietary

(()) Effective Full Information in Power Days

- Reference 2 GE9B dual recirculation loop SLMCPR 1.09 1.10 GE9B single recirculation loop SLMCPR 1.11 1.12 SVEA-96+ dual recirculation loop SLMCPR N/A 1.10 SVEA-96+ single recirculation loop SLMCPR N/A 1.13 l

The Cycle 10 SLMCPR calculations are based upon a core consisting of:

e

-- 70% GE9B fuel: 196 once burned assemblies (2.80% enriched),236 twice burned assemblies (176 at 3.27% enriched and 60 at 2.98% enriched),100 assemblies bumed for three cycles (48 at 3.25% and 52 at 3.24% enriched).

- 30 % SVEA-96+ fuel: 232 fresh assemblies (3.25% enriched) 1 Docum:nt C2atrol Desk LR-N99429 Attachn :nt 5 LCR H99-09 Administrative Changes Reflecting Fuel Vendor Transition The proposed changes to the Average Planar Heat Generation Rate (APLHGR),

Minimum Critical Power Ratio (MCPR), Recirculation Loop Limiting Condition for Operation (LCO) Action Statements, and references to fuel vendor analyses and reports are considered to bs administrative in nature since the Core Operating Limits Report

-(COLR) wi.'l continue to be used to appropriately control and limit the bounds of plant operation with slow control rods or during single recirculation loop operation, and the COLR will still be developed in accordance with NRC approved methods. Similarly, the revised references to the fuel vendor throughout the Technical Specifications are also considered to be administrative in nature since they reflect the current status of NRC approval of methodologies utilized by PSE&G and the fuel vendor to develop operating and safety limits for the fuel and core designs.

Consistent with NRC requirements for the application of ABB/CE approved methods, conservative multiplicative factors will be applied with the ABB/CE developed CPR correlation for the calculation of GE9B Operating Limit Minimum Critical Power Ratios (OLMCPRs). NRC Condition 7 in the Safety Evaluation Report for Reference 1 specifies that:

"The ABB/CE methodology for determining the operating limit maximum [ sic] critical power ratio (OLMCPR) for non-ABB/CE fuel as described in CENPD-300-P and additional submittals... is acceptable only when each licensee application of the methodology identifies the value of the conservative adder to the OLMCPR. The correlation applied to the experimental data to determine the value of the adder must be shown to meet the 95/95 statistical criteria. In addition, the licensee's submittal must include the justification for the adder and reference the appropriate supporting docmentation."

A multiplicative factor of (()). This factor has been established on the following basis.

Section 4.4 of the Standard Review Plan specifies that an acceptable approach to meet the 95/95 criterion on boiling transition is to establish a limit (the Safety Limit MCPR, or

' SLMCPR) such that at least 99.9% of the fuel rods in the core would not be expected to experience boiling transition during normal operation or anticipated operational occurrences. Accordingly, conservative SLMCPR values have been established for the GE9B and ABB/CE fuel for Cycle 10 which assure that this criterion is satisfied. (())

(())

ABB/CE has established a CPR correlation which will be used to establish the GE9B fuel Operating Limit MCPR (OLMCPR) in accordance with the methodology described in Reference 1. (()).

Document Ccntrol De2k LR-N99429 Attachm:nt 5 LCR H99-09

(()) PSE&G has also imposed an additional reduction of 0.01 to be incorporated with the

(()) PSE&G has required the reduction to the multiplicative factor to ensure increased conservative margin for the application of the approved ABB/CE methodology to the mixed core of ABB/CE and GE fuel at Hope Creek.

Therefore, the conservative SLMCPR in conjunction with the conservative multiplicative

. factor to be applied in the GE9B OLMCPR provide assurance that more than 99.9% of the fuel rods avoid transition boiling during normal operation as well as during postulated anticipated operational occurrences.

ENVIRONMENTAL IMPACT.

The proposed TS changes were reviewed against the criteria of 10CFR51.22 for environmental considerations. The proposed changes do not involve a significant

' hazards consideration, a significant change in the types or a significant increase in the amounts of effluents that may be released offsite, or a significant increase in the individual or cumulative occupational radiation exposures. Based on the foregoing, PSE&G concludes that the proposed TS changes meet the criteria given in

' 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an EnvironmentalImpact Statement.

cob CLUSIONS,:

Based on all of the facts, observations and arguments presented above, PSE&G concludes that a dual recirculation loop SLMCPR of 1.10 for both GE and ABB/CE fuel in Cycle.10 core are justified and are conservative. Similarly, PSE&G concludes that single recirculation loop SLMCPR values of 1.12 and 1.13 are conservative for the GE98 and SVEA-96+ fuel in Cycle _10, respectively, and are, therefore, justified.

' Changes to TSs 3/4.2.1,3/4.2.3 and 3/4.4.1 and their associated Bases are being made to reflect the use of approved ABB/CE methodologies for calculating core operating limits and are administrative in nature. An additional administrative change is also being made to TS 6.9.1.9 to reflect the addition of approved ABB/CE methodologies.

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r Docum:nt C:ntrol De:k LR-N99429

. Atta:hm:nt 5 LCR H99-09

REFERENCES:

1. Licensing Topical Repott, Reference Safety Report for Boiling Water Reactor Reload Fuel, CENPD-300-P-A, July 1996.

2. PSE&G Letter LR-N98404, LCR H98-06, REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS; SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR); HOPE CREEK GENERATING STATION, FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354, August 25,1998.

3. Licensing Topical Report, GeneralElectnc BWR Thermal Analysis Basis (GETAB):

Data, Correlation and Design Applicction, NEDO-10958-A, January 1977.

4. General Electnc Standard Application for Reactor Fuel, NEDE-24011-P-A-13-US, August 1996.

5. Genera / Elecido Fuel Bundle Designs, NEDE-31152P, Revision 6, April 1997.

6-