ML20042G242

Proposed Tech Specs Re Fuel cycle-specific Parameters

ML20042G242
Person / Time
Site:	Oyster Creek
Issue date:	05/07/1990
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20042G230	List: ML20042G229 ML20042G234 ML20042G242 ML20042G243
References
NUDOCS 9005110264
Download: ML20042G242 (11)

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. OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 i DOCKET NO. 50-219 i TECHNICAL SPECIFICATION CHANGE REQUEST NO. 180' ,;

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. Applicant hereby requests the Commission to change Appendix A to the.above captioned license as indicated below. Pursuant to 10CFR50.91, an analysis

.concerning the determination of no significant hazards considerations is also presented:.

l.- Sections to be Chanced 1.0 " Definitions", 3.10 " Core Limits" and 6.0 " Administrative Controls". I e

2. Extent of Chance , ,

i Removal of APLHGR, LLHGR and MCPR cycle-specific limits. Add the requirements to develope and maintain a core' operating limits report and ,

submit a copy to the NRC upon issuance. Definitions supporting the change  !

-,- are also provided. +

3. Chances Recuested As indicated on the attached revised Technical Specification pages 1 1,1.0-4, 2.0-7, 3.10-1, 3.10-2, 3.10-3, 3.10-4, 6-17 and new page 6-17a.

Existing pages 3.10-5 through 3.10-12 have been deleted, j!

4.- Discussion i

Generic Letter 88-16 provides guidance for-Technical Specification changes-concerning cycle-specific limits and is. intended to eliminate an unnecessary resource burden for both the licensee and the NRC.- Consistent

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with Generic Letter 88-16, TSCR No. 180 modifies the license-such that core operating parameters are maintained within the limits provided in the Core Operating Limits Reports (COLR). The core operating limits will be documented in the COLR before each-reload cycle or any remaining part of a s l reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

TSCR No. 180 also establishes reporting requirements.and provides definitions supporting the proposed changes. The COLR, including any mid-cycle revisions or supplements there to, will be provided upon issuance, for each reload cycle, to the NRC. Definitions concerning p . average planar linear heat generation rate, local linear heat generation-l, rate and COLR have been added.

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The proposed changes do not alter the manner in which the core will be operated nor the method by which the limits are established.

L 9005110264 900507 PDR ADOCK 05000219 p PDC TSCR-180 l , 1

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5. Determination .

L GPU Nuclear has. determined that operation of the Oyster Creek Nuclear Generating Station in accordance with the proposed technical ,

specifications does not involve a significant hazard. The changes do' nots ,

l..  : Involve a significant increase in the probability or the consequence of an accident previously evaluated. There are no changes to plant configuration, availability of safety systems, -

the manner in which the safety systems are initiated or the way

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the plant is operated that will increase the probability ~or consequences of an-accident.

2. Create the possibility of a new or different kind of accident from any previously evaluated.. The proposed change does.not alter the plant configuration, nor does it change the-availability of safety systems or the manner in which they respond to initiating events. As such, the-possibility of a new-or different kind of accident from any previously evaluated is not created.

3. Involve a significant reduction in a margin of safety. The-proposed limits will be based upon analysis results which~were performed in accordance with methods and procedures approved by  !

NRC for use at Oyster. Creek, thus the margin of safety is not reduced.

i TSCR-180 4',.

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y TABLE OF CONTENTS Section'11 Definitions Page 1.1- operable 1.0-1 i 1.2 Operating 1.0-1 1.3 Power operation 1.0-1 1.4 Sta'rtup Mode 1.0-1 1.5 Run Mode 1.0-1 Shutdown condition 1.6 1.0-1 1.7 Cold shutdown 1.0-2 1.8 Place in Shutdown Condition 1.0-2

'1.9 Place in Cold Shutdown Condition 1.0-2 1.10- Place in Isolation Condition 1.0-2 1.11 Refuel Mode 1.0-2 1.12 Refueling Outage 1.0-2 1.13 Primary containment Integrity 1.0-2 1.14 secondary Containment Integrity 1.0-2 1.15 Deleted 1.0 1.16 Rated Flux 1.0-3 1.17 Reactor Thermal Power-To-Water 1.0-3 1.18 Protective Instrumentation Logic Definitions 1.0-3 1.19 Instrumentation Surveillance Definitions 1.0-4 1.20 FDSAR 1.0-4 1.21: Core Alteration 1.0-4 1.22- Critical Power Ratio 11.0-4 1.23 Staggered Test Basis 1.0-4 1.24 Surveillance Requirements 1.0-5 1.25 Fire Suppression Water System 1.0-5 1.26s Fraction of Limiting Power Density (FLPD) 1.0-5 1.27 MaxLmum Fraction of Limiting Power Density (MFLPD) 1.0-5 1.28 Fraction of Rated Power'(FRP) 1.0-5 1.29 Top of Active Fuel (TAF) 1.0-5 1.30 Reportable Event 21.0-5 1.31 Identified Leakage 1.0-6 1.32 Unidentified Leakage 1.0-6 1.33 Process Control Plan 1.0-6' 1.34 Augmented Offgas System (AOG) 1.0-6 1.35 Member of the Public '1.0-6 1.36 offsite Dose Calculation Manual 1.0-6 1.37 Purge 1.0-6 1.38 Exclusion Area 1.0-6 1.39 Reactor Vessel Pressure Testing 1.0-7 1.40 Substantive Changes 1.0-7 1.41 Dose Equivalent I-131 1.0-7 1.42 Average Planar Linear Heat Generation Rate 1.0-7 1.43 Core Operating Limits Report 1.0-7 1.44- Local Linear Heat Generation Rate 1.0-7 i i

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j' Ca[ :1; 19 ~. __ INSTRUMENTATION' SURVEILLANCE DEFINITIONS.

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[:, 'A. LChannel Check u ..7

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A qualitative' determination.of acceptable operability by observation.

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of channel' behavior during operation.4: This determination shall D include, where possible,icomparison'of'the channel'with other

, independent channels measuring the'same'. variable. '

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[-f ,r. 'B5'  ; Channel Test-

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t Injection-of a simulated signal into the channel to verify its 3

' proper response including, where applicable,~ alarm and/or trip-

'. initiating l action.

C.. Channel Calibration nj g

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Adjustment of channel output such that.it. responds, with acceptable Lg

-c - range'and accuracy,:to known. values of.the parameter which the-channel measures.'- Calibration shall' encompass the entire channel l, including equipment ' actuation, alarm or trip. j 2 D.; Source Check

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A SOURCE' CHECK is.the qualitative assossment-of channel response-when the channel sensor is exposed to a source of radioactivity.,

. 1.20- FDSAR Oyster Creele Unit No. 1 Facility Description and Safety Analysis; Report as .

amended by revised pages and figure changes contained-in Amendments 14, 31: i and 45.* q q

1.21 CORE ALTERATION  !}

q A core' alteration is the addition, removal, relocation or other manual movement of fuel or controls in the reactor' core. Control rod movement with the control rod drive hydraulic'aystem is not defined.as a-core

-alteration.

-l 1.22 CRITICAL POWER RATIO .h

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The critical power ratio is the ratio of that. power in a fuel assembly which^1s calculated, by application of an NRC approved CPR' correl'ation, to' j cause.some point in that. assembly to experience boiling transition divided; ]

by the actual assembly operating power. ] 1

.1.23 STAGGERED TEST BASIS A Staggered Test' Basis shall consist of:

A. A test schedule for n systems, subsystems, trains or other .f designated components obtained by dividing-the specified test a interval into n equal subintervals. .;

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• Per Errata dtd. 4-9-69 .

OYSTER CREEK 1.0-4 Amendment No. >

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- [, 5 '1.39 REACTOR VESSEL PRESSURE TESTING System pressure testing required by ASME Code Section XI, Article IWA-5000, including system leakage and hydrostatic test, with reactor vessel completely water solid, core not critical  !

and section 3.2.A satisfied.

1.40 SUBSTANTIVE CHANGES SUBSTANTIVE. CHANGES are'those which affect the activities associated with a document or the document's meaning or intent.

Examples of non-substantive changes ares (1) correcting i spelling,'(2) adding (but not deleting) sign-off spaces, (3) blocking in notes, cautions, etc., (4) changes in corporate and personnel titles which do not reassign responsibilities and which'are not referenced in the Appendix A Technical Specifications,-and (5) changes in nomenclature or editorial changes which clearly do not change function, meaning or intent.

1.41 DOSE EOUIVALENT I-131 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 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 E-7 of. Regulatory Guide 1.109, " Calculation of .,

Annual Doses to Man from Routine Releases of Reactor Effluences for the Purpose of Evaluating Compliance with 10 CFR Part 50

-Appendix I".

1.42 AVERAGE PLANAR LINEAR HEAT GENERATION RATE l'

i The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod for all .

, the fuel rods in the specified' bundle at the specified height

l. divided by the number of fuel' rods in the fuel bundle at that height.

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, 1.43 CORE OPERATING LIMITS REPORT l

l The Oyster Creek CORE OPERATING LIMITS REPORT (COLR) is the j' document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating

. limits shall be determined for each reload cycle in accordance with Specification 6.9.1.f. Plant operation within these operating limits is addressed in individual specifications.

1.44 LOCAL LINEAR HEAT GENERATION RATE The_ LOCAL LINEAR HEAT GENERATION RATE (LLHGR) shall be applicable to a specific planar height and is equal to the AVERAGE PLANAR LINEAR GENERATION RATE (APLHGR) at the specified height multiplied by the local peaking factor at that height.

OYSTER CREEK 1.0-7 Amendment No.:

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s 3.10 CORE LIMITS ,

. Acolicability: Applies to core conditions required to meet the Final Acceptance Criteria for Emergency Coro Cooling Performance.-

Obiective: To assure conformance to the peak clad temperature limitations during a postulated loss-of-coolant accident as specified in 10 .

-; CFR 50.46 (January 4, 1974) and to assure conformance to the [

operating limits for local ~ linear heat generation rate and-minimum critical power ratio.

Specification:

'A. Average Planar LHGR During power operation the maximum AVERAGE PLANAR LINEAR HEAT GENERATION. RATE (APLHGR) for each fuel type as a function of exposure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated to restore operation to within the prescribed limits. If the APLHGR is not returned to i within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period surveillance and corresponding action shall continue until reactor. operation is within the prescribed limits at which time power operation may be continued.

B. Local LHGR During power operation, the Local LINEAR HEAT GENERATION RATE (LHGR) of any rod in any fuel assembly, at any axial location shall not exceed the maximum allowable LHGR limits specified in the COLR.

If at any time during operation it is determined by normal surveillance that the limiting value of LHGR is being exceeded, action shall be initiated to restore operation to within the prescribed limits. If the LHGR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition L within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed limits at which time power operation may be continued.

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OYSTER CREEK 3.10-1 Amendment No.

TSCR-180

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C. Minimum Critical Power Ratio (MCPR)

During steady state power operation the MINIMUM CRITICAL. POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit as specified in the COLR. 't

[ s When APRM status changes due to instrument failure (APRM or LPRM input-failure), the"MCPR requirement for the degraded condition shall be met within.a time interval of eight (8) hours, provided-that the' control rod block is placed in operation during this it.terval. .

For core' flows other than rated, the nominal value for MCPR' q shall be increased by a factor of k g, where kg is as shown the COLR. ,

e If at any time during power operation it.is determined by

- h__ -normal surveillance that the limiting value for MCPR is being_

exceeded for: reasons other than instrument f ailure, action shall-be initiated to restore operation to within the prescribed limits. If the steady state MCPR is not returned to

'within the' prescribed limits within two [2] hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue until-reactor operation is within the prescribed limit at which time power operation may be continued.

Bases:

The Specification for average planar LHGR assures 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. The analytical. methods and assumptions used in evaluating the fuel design limits are presented in FSAR Chapter 4.

LOCA analyses are performed for each fuel design at selected exposure points to determined APLHGR limits that meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is.

performed using GE calculational models which are consistent with the requirements of 10 CFR 50,' Appendix K.

The PCT following a postulated LOCA is primarily a function of the average heat generation rate of all the rode of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly.

Since expected location variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than i 20*F relative to the peak temperature for a typical fuel design, tha limit on OYSTER CREEK 3.10-2 Amendment No.:

TSCR-180

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\,d tha cycrag) lin20r hatt g:n:rction rtto-ic cufficient to assure.that calculated temperatures are below the limits specified in 10 CFR 50.46. -

The maximum average planar LHGR limits for the various~ fuel types currently being used are provided in the COLR. Both ,

four and five-loop operating limits are shown. Technical Specification 3.3.F permits four-loop operation provided-that the idle loop is not isolated from the reactor vessel. The results of the GE LOCA analysis for'four-loop operation with the idle loop not isolated provides MAPLHGR r values identical to the five-loop limits at all exposure levels.

'i LOCA analysis performed for ENC fuel permits four-loop _ .{

operation provided the fifth loop has its discharge valve closed and its bypass and suction valves open.

Fuel-design evaluations are performed to demonstrate that the cladding 1% plastic strain and other fuel _ design limits are not exceeded during anticipated operational occurrences for operation with LHGR's up to the operating limit LHGR.

The analytical methods and assumptions used in evaluating the anticipated operational occurrences to establish the operating limit MCPR are presented in the FSAR, Chapters 4, 6 and 15 and in Technical Specification 6.9.1.f. To assure that the Safety Limit MCPR is not exceeded during any moderate frequency transient event, limiting transients have been analyzed to determine the largest reduction in Critical Power Ratio (CPR). The types of transients evaluated are pressurization, positive reactivity insertion and coolant temperature decrease. The operational MCPR limit is selected to provide margin to accommodate transients and uncertainties in monitoring the core operating state, manufacturing, and in the critical power correlation itself. This limit is derived by addition of the CPR for the most limiting transient to the safety limit MCPR designated in Specification 2.1. ,

The APRM response is used to predict when the rod block occurs in the analysis et the rod withdrawal error transient. The transient rod position at the rod block and corresponding MCPR can be determined. The MCPR has been evaluated for different APRM responses which would result from changes in the APRM status as a consequence of i bypassed APRM channel and/or failed / bypassed LPRM inputs.

The steady state MCPR required to protect the minimum transient CPR for the worst case APRM status condition

. .y (APRM Status 1) is determined in the rod-withdrawal error transient analysis. The steady state MCPR valves for APRM status conditions 1, 2, and 3 will be evaluated each cycle. For those cycles where the rod withdrawal error transient is not the most severe transient the MCPR Value for APRM status conditions 1, 2, and 3 will be the same and be equal to the limiting transient MCPR value.

OYSTER CREEK 3.10-3 Amendment No.

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• Thb time intCrval of-Eight (8) hour 2 to adju t th3 ctaidy.

state of:MCPR to account-for a degradation in the APRM-

-- status is justified on the basis of instituting a control rod block which' precludes the possibility of experiencing 1a-rod withdrawal error transient since rod withdrawal is physically prevented. This time-interval is' adequate to

-allow the operator to either increase the MCPR to the appropriate value or to upgrade the status of the APRM l system while in a condition which prevents the. possibility. .

of this transient occurring.

1 Transients analyzed each fuel cycle will be evaluated with.

respect to the operational MCPR limit'specified:in the COLR. I The purpose of the kg factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of.the operating.

~ limit MCPR and the kg factor. Specifically, the kg factor provides the required. thermal margin to protect y against a flow increase transient.

The kg factor curves, as shown in the COLR,.were developed generically using the flow control line ,

corresponding to rated thermal power at rated core flow.  !

For the manual flow control mode, the kg factors were calculated such that at the maximum flow state (as limited by.the pump scoop tube set point).and the corresponding core power (along the rated flow control-line), the limiting bundle's relative power was adjusted'until the MCPR was alightly above the Safety Limit. .Using this relative bundle power, the MCPR's were calculatedJat-different points along the rated flow control line .}

corresponding'to different core flows. 'The ratio of the -

MCPR calculated at a given point of core flow, divided by the operating limit MCPR determines the value of k g.

OYSTER CREEK 3.10-4 Amendment No.:

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. ,(4)' c cummary'of meteorologic 1 d ta call cted during ths yrcr

.shall be included-in the report submitted within 60 days after. ,

January-1 of each year. Alternatively, summary meteorological data may be retained by GPU Nuclear and made available to the NRC upon request.

e. Annual Radiolocical Environmental Report: A report of radiological environmental surveillance activities during each year shall be submitted before May 1 of the following year. Each' report shall include the following information required in Specification 4.16 for radiolog1 cal environmental surveillance:

(1) a summary description of the radiological environmental j monitoring program, (2) a map and a table of distances and directions'(compass azimuth) of locations of sampling stations from the reactor,-  ;

l (3) results of analyses of samples and of radiation measurements, (In the event some results are not available, the reasons shall be explained in the report. In the event the missing results are obtained, they shall be reported to'the NRC as soon as is= reasonable.)

J (4) deviation (s) from the environmental sampling schedule in-Table 4.16,1.

(5) identification of environmental samples analyzed when instrumentation was not capable of meeting detection capability in Table 4.16.2.

(6) a summary of the results of the land use survey.

(7) a summary of the-results of licensee participation in an NRC approved inter-laboratory crosscheck program for. environmental samples.

-(8) results of dose evaluations to demonstrate compliance with 40 CFR Part 190.10a.

f. Core Operatino Limits Report (COLR)

Core operating limits shall be established and documented in the COLR 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 as described in the COLR. 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 COLR, 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.

OYSTER CREEK 6-17 TSCR-180

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Basis 6.9.1.E An annual report of radiological environmental surveillance activities includes factual data summarizing results of activities required by the surveillance program. In order to aid interpretation of the data, GPUN may choose to submit annlysis of trends and comparative non regional radiological environmental data. In addition, the. licensee may choose ,

to discuss previous radiological environmental data as well as the observed radiological environmental impacts of station operation (if any) on the environment. ,

6.9.2 REPORTABLE EVENTS The submittal of Licensee Event Reports shall be accomplished in accordance with_the requirements set forth in 10 CFR 50.73.

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OYSTER CREEK 6-17a TSCR-180