ML20149G430

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Proposed Tech Specs Implementing Appropriately Conservative Safety Limit Min Critical Power Ratio
ML20149G430
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 07/16/1997
From:
Public Service Enterprise Group
To:
Shared Package
ML19317C513 List:
References
NUDOCS 9707230204
Download: ML20149G430 (3)
v • d • e


Text

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2.0 SAFETY LIMITS ANO LIMITING SAFETY. SYSTEM SETTIN05 ~~

1 2.1 SAFETY LIMITS

.- THERMAL DOWER, low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with tne 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:

4 With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam does pressure less than 785 psig or core flow less than 10% of rated flew, be in at least HOT SHUTDOWN within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />'3 and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow bI J

2.1.2 The NINIMUM CRITICAL POWER RATIO (MCPR) shall not be less han th two recirculation loop operation and shall not be less than-1 ith single recirculation loop operation, in both cases with tlie reactor vessel stvan does pressure greater than 785 psig and core flow greater than 10% of rated flew. l l

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: .go With MCPR less than t0Mh two recirculation loop operation or less than

> 1 8 with single recirculation loop operation and in both cubes with the reactor vessel steam dome pressure greater than 705 psig and core flow greater than 10%

\.\1 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 CO^LANT SYSTEM PRESSURE 2.1.3 The reactor coolant systes pressure, as esasured 'n the reactor vessel.

steam done, shall not exceed 1325 psig.

APPLIt' N : OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION: -

a With the reactor coolant system pressurw, as esasured in the reactor vessel steam done, ainove 1325 psig, be in at least HOT SHUTOOWN with reactor coolant systea 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 9707230204 970716 PDR ADOC'A 05000354 P PDR

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2.1 SAFETY LIMITS i

BASES

2.0 INTRODUCTION

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The fuel cladding, reactor pressure vessel and primary systna piping >

i are the principal barriers to the release of radioactive materials to the j environs. Safety Limits are established to protect the integrity of these l

barriers during normal plant operations and anticipated transients. The fuel

cladding integrity Safety Limit is set such that no fuel damage is calculated j to occur if the limit is not violated. Because fuel damage is not directly M 1- yo observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than,;h0FYor two recirculation loop operation and h08 lLg for single recirculatton loop opera on. MCPR greater than h0Cfor two re- un i circulation loop operation an or single recirculation loo $ operation i

represents a conservative marg n relative to the conditions required to maintain l

fuel. cladding integrity. The fuel cladding is one of the physical barriers j

which separate the radicactive 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 l 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 thersal stresses which occur from reactor operation l

significantly above design conditions and the Limiting Safety System Settings.

I While fission product migration from cladding perforation is past as measurable l

as that from use related cracking, the thermally caused cladding perforations i signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding l 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-4 i

ficant departure from the condition intended by design for planned operation.

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2.1.1 THERMAL POWER. Low Pressure or Low Flow

' The use of the applicable NRC-approved critical power correlation is not valid for all critical power calculations performed at reduced pressures below

, 785 psig or core flows less than 10% of rated flow. Therefore, This is the fuel cladding done by estab-integrity Safety Limit is established by other means.

lishirg a limiting condition on core THERMAL POWER with the following basis.

Since the pressaire 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 Thus,

'_ drop is nearly independent of bundle power and has a value of 3.5 psi.

the bundle flow with a 4.5 psi driving head will be greater than 28 x 10s 1bs/hr.

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, 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 TilERMAL ?0WER for reactor pressure below 785 psig is conservative.

B 2-1 Amendment No. 42 ,

HOPE CREEK

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I RECIRCULATION LOOPS 4

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LIMITING CONDITION FOR OPERATION j maammmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmmma 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation

with

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

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

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 .

, ACTION: _

, s. With one reactor coolant system recirculation loop not in operation:

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!. 1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

1' a) Place the recirculation flow control system in the Local Manual mode, and l Reduce THERMAL POWER to $ 70% of RATED THERMAL POWER, and j b) j c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) safety Limit l

- by 0.01 te L[per specification 2.1.2, and d) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) liniit to a value of 0.86 times the two recirculation loop limit per specification 3.2.1, and

) e) DELETED.

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j 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 i

$ 38% of RATED THERMAL POWER or the recirculation loop flow in l

i the operating loop is s 50% of rated loop flow.

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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) l scram Trip setpoints and Allowable values to those applicable 4

for single recirculation loop operation per specifications j 2.2.1 and 3.2.2r otherwise, with the Trip setpoints and j

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 ger 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 Elock Trip

' See special Test Exception 3.10.4.

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

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