ML20137M596

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Proposed Tech Specs,Representing Changes to TS 2.1.2,action a.1.c for LCO 3.4.1.1 & Bases for TS 2.1
ML20137M596
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 03/31/1997
From:
Public Service Enterprise Group
To:
Shared Package
ML20007F971 List:
References
NUDOCS 9704080128
Download: ML20137M596 (5)
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Text

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I ATTACHMENT 3

  • LR-N97187 HOPE CREEK GENERATING STATION "

FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 r SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SINCPR) CHANGES ,

TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. NPF-57 are affected by this change request:  ;

Technical Specification Page 2.1.2 2-1 l Bases 2.0 B 2-1 3.4.1.1 3/4 4-1 l

9704000128 970331 PDR ADOCK 05000334 P PDR ,

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2.0 SAFETY LIMITS AND LIMITING SAFETY. SYSTEM SETTINGS

,J 2.1 SAFETY LIMITS -

THERNAL 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: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dose 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 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 TheMINIMUMCRITICALPOWERRATIO(MCPR)shallnotbehlessthan ~

. with

! two recirculation loop operation and shall not be less than M ith single i recirculation loop operation, in both cases with the reactor vessel steam dose  !

pressure greater than 785 psig and core flow greater than 10% of rated flow.  !

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. I ACTION: ,g With MCPR less than 4 ith two recirculation loop operation or less than 1 M ith single recirculation loop operation and in both cases with the reactor vessel steam done pressure greater than 785 psig and core flow greater 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 6.7.1.

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

APPLICA8Ek!TY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vesse) '

steam done, above 1325 psig, be in at least HOT 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. -

HOPE CREEK 2-1 Amendment No.15

2.1 SAFETY LIMITS BASES  :

2.0 INTRODUCTION

4

. The fuel cladding, reactor pressure vessel and primary system piping

are the principal barriers to the release of radioactive materials to the i environs. Safety Limits are established to protect the integrity of these
barriers 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 observable, a step-back approach is used to establish a Safety Limit such that

, the MCPR is not less tharyl-efTor two recirculation loop operat n and for single recirculation loop operation. MCPR greater than . g for two re-i l .~ og circulation loop operation and #c007 or single recirculation loop operation b8 represents 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 incre-

. mentally cumulative and continuously measurable. Fuel cladding perforations, i 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 j signal a threshold beyond which still greater thermal stresses may cause gross i 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 signi-i ficant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Lu Pressure or Low Flow l The use of the applicable NRC-approved critical power correlation is not 4 valid for all critical power calculations performed at reduced pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding i integrity Safety Limit is established by other means. This is done by estab-lishing 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 psi. Analyses show that with a bundle flow of 28 x los 1bs/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 10s 1bs/hr.

_3 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.

4~

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. 42

. 3/4.4 REACT 6R COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING COKDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:

a. Total core flow greater than or equal to 45% of rated core flow, or .

1

b. THERMAL POWER less than or equal to the limit specified in Figure l 3.4.1.1-1.

I l APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 . 1

&QIIDE: .-

a. With one reactor coolant system recirculation loop not in operations ,

l

1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s: I a) Place the recirculation flow control system in the Local Manual mode, and  !

I b) Reduce THERMAL POWER to s 70% of RATED THERMAL POWER, and c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit 57 0.01 Le 1.00# per specification 2.1.2, and d) Reduce the Maximum Average Planar Linear Heat Generation Rate l (MAPLHCR) limit to a value of 0.86 times the two recirculation

(' loop limit per specification 3.2.1, and e) DELETED. l f) 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 j

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

2. 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) scram Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per specifications 2.2.1 and 3.2.2; otherwise, with the Trip setpoints and Allowable values associated with one trip system not reduced to those applicable for single recirculation loop operation, place the affected trip system 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 />, reduce the Trip setpoints and Allowable Values of the affected channels to those applicable for single 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 Rod Block Trip i See Special Test Exception 3.10.4.

HOPE CREEK 3/4 4-1 Amendment No. 63 l

. i s <

ATTACHMENT 4 LR-N97187 HOPE CREEK GENERATING STATION <

FACILITY OPERATING LICENSE NPF-57 DOCKET NO.'50-354 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR) CHANGES AFFIDAVIT AND BASIS FOR WITHHOLDING INFORMATION CONTAINED IN LCR H97-05 FROM PUBLIC DISCLOSURE 1

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