ML20062B655

Proposed Tech Specs Re Min Critical Power Ratio Safety Limits

ML20062B655
Person / Time
Site:	Hatch
Issue date:	10/16/1990
From:	GEORGIA POWER CO.
To:
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ML20062B653	List: HL-1276, Application for Amend to License NPF-5,changing Tech Specs Re Min Critical Power Ratio Safety Limits ML20062B655
References
NUDOCS 9010260009
Download: ML20062B655 (4)

Text

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ENCLOSURE 3-PLANT HATCH - UNIT 2 NRC-DOCKET 50-366 OPERATING LICENSE NPF-5 REQUEST TO REVISE TECHNICAL SPECIFICATI0NSL >

MCPR SAFETY LIMITS l

PAGE CHANGE INSTRUCTIONS EiLqt Instruction 2-1 Replace B 2-1 Replace B 3/4 2-3 Replace l

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HL-1276 01113 E3-1 9010260009 901016 PDR ADOCK 05000366 '

P PDC ,

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS ,

THERMAL POWER (Low Pressure or Low Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow. '

less than 10% of rated flow. {

APPLICABILITY: CONDITIONS 1 and 2. l ACTION:

Hith 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 />. .!

THERMAL POWER (High Pressure and High' Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 for two-loop recirculation or 1.07 for single-loop recirculation l operation with the reactor vessel steam' dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: CONDITIONS 1 AND 2.

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ACTION: s Hith MCPR less than 1.06 for two-loop recirculation or 1.07 for single-loop l recirculation operation and the reactor vessel steam dome pressure-greater than 785 psig and core flow greater than 10% of rated' flow, be in at least HOT SHUTDOWN within-2 hours. ,

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REACTOR C001. ANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: CONDITIONS 1, 2', 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUT 00HN with l reactor coolant system pressure 1 1325 psig within'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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HATCH-UNIT 2 2-1 Proposed. TS/0389q/242-'105-q

. 2.1 SAFETY LIMITS i BASES l

2.0 The fuel cladding, reactor pressure vessel and primary system j piping are the principal barriers to the release of radioactive materials '

to the environs. Safety Limits are established-to protect the integrity  ;

of these barriers during normal plant operations and anticipated tran- j sients. 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 observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than~1.06 for two-loop operation and 1,07 for single-loop operation. These limits represent a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier 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-incrementally cumulative and-continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occurifrom reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measuraole 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.

bolling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

The evaluations which justify normal operation, abnormal transient, and accident analyses for two-loop operation are discussed in detail _in Reference 1. Evaluation for single-loop operation demonstrates that two-loop transient and accident analyses are more limiting than single-loop (Reference 2).

l 2.1.1 THERMAL POWER (Low Pressure or Low Flow) '

The use of the NRC-approved transition bolling correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 101,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 essential'y all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10' lbs/hr. bundle pressure drop.is "

nearly independent of-bundle power and has a value-of 3.5 psi. Thus, the

• i bundle flow with a 4.5 psi driving head will be greater than 28 x:1_0*

lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 i

psia indicate that the fuel assembly critical power at -this flow is approximately 3.35 MHt. With the-design peaking factors, this corresponds to HATCH - UNIT 2 B 2-1 Proposed'TS/0391q/242-106

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t POWER DISTRIBUTION LIMITS-B_ASES 3/4.2.2 APRM SETPOINTS  ;

This section deleted.

I 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state gerating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06 for two-loop operation and 1.07 'l for single-loop operation, and an analysis of abnormal operational transients (Reference 1). For any abnormal operating transient analysis evaluation with the initial condition of the' reactor being at the steady. state operating limit (specified in the CORE OPERATING L7MITS REPORT), it is required that the.

resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limits,are not violated during any anticipated abnormal operational transient, the.most limiting transients have been analyzed to determine which results in the largest-reduction in CRITICAL POWER RATIO (CPR). The type of transients eval m ted were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

Details of how evaluations are performed, the methods used, and how the ,

MCPR limit is adjusted for operation at less than rated power and' flow '

conditions are given in Reference 1 and in the CORE OPERATING LIMITS REPORTS.

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L HATCH - UNIT 2 8 3/4 2-3 Proposed'TS/0392q/242-106