ML20066B755

Proposed Tech Specs Re Rev to Definitions & Bases Sections of Tech Specs

ML20066B755
Person / Time
Site:	Hope Creek
Issue date:	12/28/1990
From:	Public Service Enterprise Group
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Shared Package
ML20066B751	List: ML20066B749 ML20066B755
References
NUDOCS 9101080327
Download: ML20066B755 (13)

Text

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Ref: LCR 90-09

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A'ITACllMENT 1 PROPOSED TECIINICAL SPECIPICATIONS AND BASES CilANGE 9

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l PROPOSED CilANGE TO THE TECIINICAL SPECIFICATIONS

. FACILITY OPERATING LICENSE NPF-57 IlOPE CREEK GENERATING STATION _ -

DOCKET NO. 50-354 ref: LCR 90-09 1

DESCRIPTION OF THE CHANGE As shown on the marked-up Technical Specifications (TS) and BASES pages in-

-Attachment 2, PSE&G requests that DEFINITION 1.10, CRITICAL POWER RATIO, and BASES Sections B2.1.1, D2.1.2, B3/4.2.3, Bases Table'B2.1.2-2, and certain References in those Bases Sectic,s be revised.

l REASON FOR-THE-CPANGE The proposed changes-are administrative in nature, making the Hope Creek Generating Station (HCGS)-TS definition for CRITICAL POWER. RATIO;more generic with_ regard to the critical power correlation. . This will permit the-use of new NRC-approved fuel designs and their-associated NRC-approved correlations.

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without requiring amendment of this section each time. ' Additionally, the; BASES r

for THERMAL POWER and MINIMUM CRITICAL POWER RATIO are modified ~to reference the . latest approved version .of the GE Standard Application for~ Reactor Fuel (GESTAR II) and to more clearly define-how certain MCPR factors are determined.

JUSTIFICATION FOR THE CHANGE. '

l

-Since the fuel design and its supporting analysis methodologies'have toibe previously reviewed and approved bylNRC before use in the HCGS reactor, this y change-to.a DEFINITION-ia essentially adLinistrative tu nature. This proposed  ;

amendment will preclude =TS DEFINITION. -evisions every time there are minor- . l changes in the fuel ma ifacturer's' critical power ~ correlations'to support their new fuel design features. Provided those changes are reviewed and approved by the NRC, the more generic reference, " applicable NRC-approved critical power correlation, 'is- appropriate.

The changes to tha BASES, which deviate from the standard TS language are--

clarificatione provided to.PSE&G'by General Electric Company for-inclusion in

-this amendment request.

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, 1 10CFR50.92 SIGNIFICANT IIAZARDS CONSIDERATION ANALYSIS - .

PSE&G has, pursuant to 10CFR50.92, reviewed the proposed amendment to determine whether our request involves a significant hasards-consideration. .We have- y

. determined that:

The operation of 11000 Creek Generatino Station (HCGS) in accordance with the proponed chance will not involve a sionificant increase in the probability or consecuencen of_an accident previous 1v evaluated.

, The. proposed-amendment does not involve a physical or proc'edural._ change for any structure, component.ortsystem that.affects the probability or consequences of; l any accident or-malfunction of equipment limportant,to safety previously

evaluated in the Updated Final Safoty Analysis Report (UFSAR).. In order to

' install any new fuel design in the IICOS reactor, the change =in fuel design and

! supporting correlations will have been previously reviewed and approved by the NRC and the limiting transients previously evaluated in the SAR.will-have been j re-analyzed for each reload design. New core operating limits will-have been generated and documeated in the CORE OPERATING LIMITS ~ REPORT-(referenced in the

! Technical Specifications) to-ensure that allfsatety criteria were met for'all analyzed accidents and limiting transients. Therefore; the CRITICAL POWER ,

i RATIO definition will always be correct - in that the CPR correlation being l c

used will have been approved by the NRC as part of any new fuel design:

approval, b

} The operation of_llope-Creek Generating Station-(IICGS) . in accordance with the- ~i j proposed chance-will not create the possibility of a new or-different kind of j accident from any previously evaluated.

There are.no physical changes to the. plant or to the manner in which the plant is operated involved in the proposed revision. The' proposed. change ~will. define n . CRITICAL POWER RATIO as the ratio'of-that power in an-assembly whichtis-calculated'by application of the " applicable NRC-approved" critical' power correlation" to cause some point _in-the assembly'to' experience _ boiling c transition, divided by the-actual-assembly power. The previous-definition:

F _ specified General Electric?s "GEXL" correlation-which has-'been modified to include considerations for a-high performance _ spacer (ferrule type)-' design in_the GE9 fuel.- The new correlation used is-termed,

'lGEXL-plus" . This proposed amendment will preclude TS DEFINITION revisions every timo there are minor changes in the fuel manufacturer's. critical power-

. correlations to support their new fuel design features.- Provided-those changes are reviewed and approved by the NRC, the more. generic reference, " applicable.

NRC-approved critical: power correla' U n", _ no new 'or :dif f erent accident, from any previously evaluated, is creste ay this. broader. definition.

The operation of Ilope Creek Generating Station (llCGS) in accordance with the proposed chance does not involve a significant reduction in a maroittof safetyt For each core loading, chapters 4 and '15, which contain information aaout the fuel design and the results of safety analyses, are re-evaluated. ".his process ensures that the fuel system design, nuclear-design, thermal /hydrr.ulic design and the conclusions of the original core analysis remain valid for the accidents and limiting transients previously evaluated Ic the CAR. The proposed revision will merely redefine, in broader terms, the defini'lon t of critical power ratio and will not cause a change in any margin of safety.

Conclusioni Based upon the foregoing evaluation, we have determined that this proposed change does not invojve a Significant flazards Consideration.

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l Ref: LCR 90-09 ATTACHMENT 2 INSERTS AND NARKED-UP PAGES

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i INSERTS FOR' PROPOSED' CHANGES INSERT 1 applicable NRC-3pproved critical power INSERT 2 performed at reduced 2 INSERT 3 a statistical model that'contines all of the uncertainties in operating- l parameters and in the proceduces useo to calculate critical power.

Calculation of the Safety Limit MCPR is defined in Reference.l. The required inputs to the statistical model are the uncertainties listed in Bases. Table B2.1.2-1.

INSERT 4

' Reference 1 1

! 1. General Electric Standard Applicationfor Reactor Fuel,- NEDE-240ll-P-A (latest approved revision) .

INSERT 5 This page. intentionally left. blank-

, ~ . _ . . _ . . . _____.___ _ _ _. _ _ . _ -. . . ~

4 I

INSERTS - Cont'd INSERT 6 The codes used to evaluate transients are discussed in Reference 2.

INSERT 7 operating limit

, INSERT 8 The Kt factors are determined in the following manner: The change in ,

CPR is determined as a function of core flow along the' rated power flow control line. Then, f or a given scoop tube setpoint in the. manual flow :

control operating mode; the MCPR-at reduced flow is established that-would give-the Safety Limit MCPR-if the core flow was-increased to the scoop tube setpoint. The ratio of the.MCPR at reduced flow to the operating limit MCPR is the Kt factor at that reduced flow.

-i INSERT 9 is 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.

INSERT 10 are equal to or greater than

, DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION'shall be the addition, removal, relocation or movement of fuel, scarces, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed =ind fuel in the vessel. Normal movement of the SRMs, IRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATI0l:S shall not preclude completion of the moveinent of a component to a safe conservative position.

CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY

1. 8 The CORE highest MAXIMUM value FRACTION of the FLPD OF LIMITING which exists POWER DENSITY (CMFLPD) sh in the core.

CORE OPERATING LIMITS REPORT 1,9 The CORE OPERATING LIMITS REPORT is the unit specific document that provides core operating limits for the current operating relenor cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Plant o limits is addressed in individual specifications peration within these CRITICAL POWER RATIO **1 1.10 The CRITICAL POWER RATIO (CPR) shall be the rat f at power in the assembly which is calculated by application of the correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which'alone would produce the same thyroid dose as the quantity and l

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."

E-AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average, weighted in proportion to the concentration of each radionuclide 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 C00 LING SYSTEM (ECCS) RESPONSE TIME

! 1.13 The EMERGENCY CDPF COOLING SYSTEM (ECCS) RESPONSE TIME shall be I_ interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor entil the'ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, l

pumo discharge pressures reacn their required values, etc. Times shall i include diesel generator starting and sequence loading delays where applicable.

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

HOPE CREEK 1-2

2'. 3AFE*v L'9173 BASES 2.0 *NTRCOUCTION The fuel cladding, reactor pressure vessel ano primary system Dioing are the principal barriers to the release of radioacti,ve 'sterials to the environs. Safety Limits are established to protect the integrity of these carriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observaDie, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation. MCPR greater than 1.07 for two re-circulation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separatt the radioactive materials from the environs. The integrity of this cladding tarrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incre-mentally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.

While fission procuct 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. These conditions represent a signia ficant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or low Flow INSERT 1 INSERT 2 1_ . -

The us of the[,,,jTcorrelation is not valid for all critical power calculations flow. Theref ressures below 785 psig or core flows less than los of rated other means. Thisthe fuel cladding integrity Safety Limit is established by 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 alwa flow of 28 x 10{s be greater than 4.5 psi. Analyses show that with a bundle 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 28 x 103 lbs/hr. Full scale ATLAS test cata taken at pressures from 14.7 paia to 800 psia indicate that the fuel assemely criti-cal 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 i

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.1.2 THEFM. 90hER Hion Pressure anc Mich Flow The fae; :taccing* integrity Safety Limit 1, .et sucn that so Nel dama;e is calculatec to occur if the limit is not violneo . Since tne carameteas

=nien result in fuel damage are not directly ceservaele curing reactor oceration.

the thereal and hydraulic conditions resulting in a cenarture from nucleate boiling nave been used to mark the beginning of the region where fuel casage could occur. Althougn it-is recogniaod that a departure free nucleate boilie; would not necessarily result in damage to SWR fuel rods, the critical' power at which boiling transition is calculated to occur has been adopted as a conver'ent limit. However, the uncertainties in monitoring the core operating state ano in the procedures used to calculate the critical power result in an uncertairy in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel asseemly for whicn more than 99.5 of the fuel rocs in the core are expected to avoic 0011199 trensition considering the power distribution witnin the core and all uncerta'~

ties.

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!a s es 'ac's 62. ' .. 2+ 2 NOMINAL Vat.Uf! : PARAMET!PS USED !

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!RM L 20WER 3323 W Core 81 0w 108.5 'M r i

Dome Pro re 1010.4 psig nnel Flow Area 0.1049 ft 2 .

R Factor High enrichment 1.043 Medium enrichment 1.029 Low enricament - 1.03?

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POWER DISTRIBUTION LIMITS '

B_ASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions d as specified in Specification 3.2.3 are derived from the established fuel 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 steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any 2.2.

Specification time during 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 transient, the most limiting tran-sients have been analyzed to determine which result in tNe' 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.

INSERT 6 The evaluation of a given trans nt begins with the system initial parameters shown in FSAR Table 15.0-3 t at are input to a GE-core dynami_cy behavior transient computer progra'm. he _: u Yd t0 cvC.uttf PT:2 0fiz; tion

- , -e event; %i:--desc+4 bed-4n-NEDO-24HF and-the-program-used-in-non pmsur-ha eveM s 4 described i" ME00-10802O The cutput: cf thi program along i.ith-On: ini~tial MCPn :cra :L '... m ,ci 'further analy::: cf the ther all, -iti"g bundle "ith the ci g!c ch:nnel tran icnt ther.21 hydraulic T^,5C : d d:::r! bed-

!^ NECE-2 Sit 9 'h; principal rc: ult cf thi cv;1Naticr i: the reducti0r i" MCPo aused b" the trantientm w

The purpose of the K, factar specified in the CORE OPERATING LIMITS i REPORT is to define, operating limits at other than rated core flow INSERT 7 conditions. At less than 100% of rated flow the required MCPR is the product of theMMCPR and the K factor. The K factors assure that the Safety Limit MCPRwillnotbeviol$tedduringafl$wincreasetransientresultingfroma motor generator speed control failure. The K 7 factors may be applied to both 1

manual and automatic flow control modes.

I j The Ky factors values specified in the CORE OPERATING LIMITS REPORT were developed generically and are applicable to all BWR/2, BWR/3 and BWR/4 reactors.

The K THERMdlPOWERatratedcoreflow, factors were derived using the flow control line

.m - mm,~m. ,m e,,+

.mm m m, m_ ,m.. , _ _ -. __.__ _ e_-- -.

th t-f0- the n imum f'a rate, a; limited by the pump 3; cop tub ;;t P0 int and th; rrc ponding T"ERMAL 'CWE" alcag the rated ficy contr0! Ijne, th0

-iting bund!:':'rcistive pc er 53 edju3ted until the "CPR chenge3 ith di'ferent cre '10x:. The ratic cf th MCPn calculated at a given point of cere "Om di"id0d by th0 Operating ' Wit MCP9, deter-i"0 m -/ th0 "7 w

HOPE CREEK B 3/4.2-3 Amendment No. -

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j POWER DISTRIBUTION LIMITS  !

BASES MINIMUM CRITICAL POWER RATIO (Continuae) INSERT 9 For _op_eration- in the a'utomatit fjqw- control _ mode._the same procedu -.e i

_ ;O:MC "CT' .ee -w.OCt  ; te ;t.. _ ' 'ep,

.W.rW;r ti~;tli";.;T ett a; l i;.i t "CP" .; G 2G ;;^.= CM f.K Of 7,e

".'.T " T"!P"X P'".'!" =d r:t:d th = :1 ' 5 .

The K tors ified in the ERATI MITS REPORT are conservati(e because the operating limit MCPRs o peci cat on . e original 1.20 operatina 1imit MCPR used for the generic derivation of ,

f, .IN8ERT 10 -

3 At THERMAL POWER levels less than or equal to 25% of RATED THERMAL Pcwtn, the . reactor will be operating at minimum recirculation pamp speed and the moderator void content will be very small. car all designated control: rod t patterns which may be employed at this point, op6 eating plant experience indi-cetes that the resulting MCPR value is in excess of requirements by a considerable -

m gin. Durin  !

be made at 25%g initial start-up testing of the'plaat, a MCPR evaluation will of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated suc'i 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- i ment for calculating MCPR when a limiting control rod pattern is approached 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 Accord .nce with 10 CFR .50, Appendix K, NEDE-20566, November 1975.

'J L "L. 'N ; ',;, " ' " M ""101 ' " . "' * """ " " i _ ' ' " " ' ' " " ' ' ' "? " " ' " " ' 'A" " "

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Channel, T; C ai;;l Deac.i p ien,. .CC Cl?0, Jen.er,7 1000.

1 HOPE CREEK B 3/4'2-4

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