ML20072Q387

Proposed Tech Specs,Consisting of Rev 1,Suppl 1 to Tech Spec Change Request 180,removing Cycle Specific Parameter Limits

ML20072Q387
Person / Time
Site:	Oyster Creek
Issue date:	12/13/1990
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20072Q386	List: ML20072Q382 ML20072Q387 ML20072Q390
References
NUDOCS 9012210191
Download: ML20072Q387 (5)

Text

. . -_ .-

h ENCLOSURE Technical Specification Change Request No. 180 Supplement I Revision I l

t I

f i

I 9012210191 901213 PDR ADOCK 05000219 P PDR

C .' Minimum Critical Power Ratio (MCPR) o 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.

When APRM status changes due to instrument failure (APRM ot 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 interval.

For core flows other than rated, the nominal value for MCPR shall be increased by a factor of kg , whero kg is specified in the COLR.

If at any time during power operation it is determined by normal surveillance that the limiting value for MCPR is.being exceeded for reasons other than instrument failure, 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.

LOTA 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 calculacional 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 rods 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 l fuel design, the limit on l

l OYSTER CREEK 3.10-2 Amendment No.:

1

__ - ~~

l the averags linsar heat generation rate is sufficisnt to l - assure that calculated temperatures are below the lim;ts l specified in 10 CFR 50.46.

The maximum average planar LHGR limits for the various fuel types currently being used are speellied in the COLR. The MAPLHOR limits for both five-loop and four-loop operation with the idle loop unisolated are shown. Four-loop l operation with the idle loop isolated (suction, discharge l and discharge bypass valves closed) requires that a MAPLHGR multiplier of 0.98 be applied to all fuel types.

Additional requirements for isolated loop operation are given in Specification 3.3.F.2.

Fuel dosign evaluations are performed to demonstrate that the cladding 11 plastic strain and other fuel design limits are not exceeded during anticipated operational occurrences for operation with LHORs up to the operat.ng 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 trancients 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 transiente and uncertainties in monitoring the core operating state, manufacturing, and in the critical power correlation itself. This limit is derived by addj. tion 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 of the rod withdrawal error l transient. The transient rod position at the rod block and corresponding MCPR can be determined. The MCPq has been evaluated for different APRM. responses uhich would result from changes in the APRM status as a consequence of bypaosed 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 (APRM Status 3) is determined in the red withdrawal error l transient analysis. .The steady state MCPR values 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 l be equal to the limiting transient MCPR value.

l OYSTER CR3EK 3.10-3 Amendment No.

a

d. The LOCAL LINEAR HEAT GENERATION RATE-(LLHCR)=for Specification 3.10.B.-

and shall be documented in the COLR.-

1

2. The analytical methods.used' to determine the core operating = )

limits shall be those previously-reviewed and approved by the l NRC, specifically those described in the following documents.  ;

l

a. GPU Nuclear (GPUN) Topical Report'(TR) 020, Methods for the .

Analysis of Boiling Water Reactors Lattice physics,-(The approved revision at the time' reload analyses are performed ]

shall be identified in the COLR.) -

b. GPUN TR 021, Methods-for the Analysis of Boiling Water Reactors Steady State physics, (The approved revision at the time reload analyses are performed shall besidentified_in

+he COLR.)

c. GPUN TR 033, Methods for the Generation of. Core Kinetics Data for RETRAN-02, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

d. GPUN TR 040,-Steady-State and Quasi-Steady-State Methods' Used in the Analysis of Accidents ar.d= Transients, (The

. approved revision at the time reload. analyses.are. performed shall be identified in the COLR.)

e. GPUN TR 045, BWR-2 Transient Analysis Model Using the Retran i Code, (The approved revision at the-time reload analyses are-performed shall be identified in the COLR.)"

f. NEDE-21462P and NEDE-31462, Oyster Creek Nuclear Generating Station SAFER /CORECOOL/GESTR-LOCA Loss-of-Coolant Accident Analysis, (The approved revision at_the time reload analyses <

are: performed shall be11dentified in the'COLR.)

g. NEDE-240ll, General Electric Standard Application for.

Reactor Fuel,1(The approved revision at'the time reload-e.nalyses are performed shall be-identified in the COLR.)

h. NEDE-24195, General Electric Reload: Fuel . Application for Oyster Creek, (The approved revision at the time reload analyses are performed shall be identified-in the COLR.) _.

i. XN-75-55-(A); XN-75-55, Supplement 1-(A); XN-75-55,'

Supplement 2-(A),-Revision 2, " Exxon Nuclear Company WREM-Based.NJP-BWR ECCS Evaluation Model and Application to .

the Oyster Creek Plant," April 1977 '

{

i l

OYSTER CREEK 6-12a Amendment No. l

. . ~ - - .. - - . . . . .- - - _ - .__-. - - . . -

+

j. XN-75-36(NP)-(A); XN-75-36(NP), Supplemsnt 1-(A), " Spray Cooling Heat Transfer Phase Test Results, ENC - 8x8 BWR-Fuel 60 and 63 Active Rods, Interim Report," October 1975

3. The core operating limits anall be determined such that all applicable limits (e.g., fuel thern - mechanical limits, core thermal-hydraulic limits, ECCS limi :, nuclear limite such as shutdown margin, transient analysis :imits, and accident l analysis limits) of the safety analysis are met.

4. The CORE OPERATING LIMITS REPORT, including any mid-cycle l revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator _and Resident Inspector.

Basist 6.9.1.e An annual report of radiological environmental surveillance activities includeh factual data summarizing results of activities required by the surveillance program. In order to aid interpretation of the data, GPUN may choose,to submit analysis 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.

l l

l OYSTER CREEK 6-12b Amendment No.: 84, 108, 134

. _ . . - m - ... e .m -

. _ _ . _ _ . . .