ML20217N791

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Revised Pages 2-1,3/4 4-1 & B2-1 to Replace Previously Submitted Pages Contained within Attachment a of 971215 Proposed Change to TS to License NPF-49
ML20217N791
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 04/24/1998
From:
NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML20217N786 List:
References
NUDOCS 9805050433
Download: ML20217N791 (8)
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ATTACIIMENT A NIAGARA MOIIAWK POWER CORPORATION

, LICENSE NO. NPF-60 l

DOCKE'1 NO. 50-410 l

l Proposed Channt to Tech lical Specifications

' Pages 2-1, 3/4 4-1, and B2-1 have been retyped in their entirety with marginal markings to l

indicate changes. These retyped pages 2-1,3/4 4-1, and B2-1 replace the retyped pages 2-1, 3/4 4-1, and B2-1 contained within Attachment A of NMPC's December 15, 1997 submittal.

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9805050433 990424 PDR ADOCK 05000410 P PDR a 1

.. 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow i

2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the i reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated l flow. {

l APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. I ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flaw 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 I Specification 6.7. j THERMAL POWER. Hiah Pressure and Hiah Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR)* shall not be less than 1.09 with l two recirculation loop operation and shall not be less than 1.10 with single recirculation I loop operation with the reactor vessel steam dome pressure greater than 785 psig and core ,

flow greater than 10% of rated flow. j APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTIOl3:

With MCPR* less than 1.09, with two recirculation loop operation or less than 1.10 with l single loop operation, 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 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 steam dcme, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1,2,3, and 4. l ACTION: ,

With the reactor coolant system pressure as measured in the reactor vessel steam dome j 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.

l B_EACTOR VESSEL WATER LEVEL  !

t 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

  • MCPR values are applicable to Cycle 7 operation only. l NINE MILE POINT - UNIT 2 21 Amendment No Id t -

.. /4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REClRCULATION SYSTEM 1

RECIRCULATION LOOPS' l l

LIMITING CONDITIONS FOR OPERATION

- 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
b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1* AND 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:
1. Within four hours:

a) Place the recirculation flow control system in the Loop Manual (Position Control) mode, and b) Reduce THERMAL POWER to $70% of RATED THERMAL POWER, and, c) Increase the MINIMUM CRITICAL POWER RA'lO (MCPR)* *

  • Safety Limit by 3 0.01 to 1.10 per Specification 2.1.2, and, '

d) Reduce the Maximum Average Planar Linear Heat Generation Rate  !

(MAPLHGR) limit per Specification 3.2.1, and, j l

e) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and l

Rod Block Monitor Trip Setpoints and Allowable Values to those applicable l for single recirculation loop operation per Specifications 2.2.1, 3.2.2 and

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3.3.6.  ;

i f) Reduce the volumetric drive flow rate of the operating recirculation loop to j s41,800*

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t l *- See SpEcial Test Exception 3.10.4. l 1

    • ~ Th,'s value represents the volumetric recirculation loop drive flow which produces 100%

core flow at 100% THERMAL POWER.

1 MCPR values are apphcable to Cycle 7 operation only. l I l

l NINE MILE oOINT - UN T 2 3/4 4-1 Amendment No. [d, //

Goneoted4uly4h4000 l

. 2.1 BASES FOR SAFETY LIMITS i l

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1.0 INTRODUCTION

The fuel cladding, reactor pressure vessel, and primary system 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 transients. The fuel cladc'ing integrity Safety Limit is set so 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 so that the MCPR* is not less than 1.09 for two recirculation loop operation and 1.10 for single recirculation loop operation. MCPR* greater than 1.09 for two recirculation loop operation and 1.10 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers that 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 that occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. Although fission product migration from cladding perforation is just as measurable 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 that would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or low Flow The use of critical power correlations 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, 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 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 103 lb/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 l head will be greater than 28 x 103lb/hr. Full-scale ATLAS test data taken at pressuros 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 THERMAL POWER for reactor pressure below 785 psig is conservative.

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  • MCPR values are applicable to Cycle 7 operation only. l NINE MILE POINT UNIT 2 B2-1 Amendment No. /d, dd

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ATTACIIMENT B NIAGARA MOIIAWK POWER CORPORATION LICENSE NO. NPF-69 l l

DOCKET .NO. 50-410 1

I Marked-up Cony of Proposed Channe to Current Technical Snecifications The current version of pages 2-1, 3/4 4-1, and B2-1 of the NMP2 Technical Specifications have been hand marked-up to reflect the proposed changes. These marked up pages 2-1,3/4 4-1, and B2-1 replace the previously submitted marked-up pages 2-1, 3/4 4-1, and B2-1 contained within Attachment C of NMPC's December 15,1997 submittal.

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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 done pressure less than 785 psig or core flow less than 10% of-rated flow.

APPLICA8ILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: .

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam done 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 require- ,

ments of Specification 6.7. 1 THERMAL POWER, High Pressure and High Flow

. I l.09 2.1.2 TheMINIMUMCRITICALPOWERRATIO(MCPRbg shall not be ' than?-1-W with two recirculation loop operation and shall not be less than with s ngle recirculation loop operation with the reactor vessel steam pressure greater  ;

than 785 psig and core flow greater than 10% of rated flow, i APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

Sith_MCP ess than with two recirculation loop operation or less than h feGEi/with single loop operation, 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.

3EACTORCOOLANTSYSTEMPRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam done,'shall not exceed 1325 psig.

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

ACTION:

With the reactor coolant system pressure as measured in the reactor vessel 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 cosply with the requirements of Specification 6.7.

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

  • Meht values are a plic46/s. -h Cycle 7 o en4ien enly .

NINE MILE P0 INT - UNIT 2 2-1 Amendment No. g

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t 3/4.'4 REA'CTOR COOLANT SYSTEM i 3/4.4.1 RECIRCULATION' SYSTEM

-RECIRCULATION LOOM i

LIMITING CONDITIONS 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-  !
b. THERMAL' POWER within the' unrestricted zone of Figure 3.4.1.1-1. I APPLICA8ILITY: OPERATIONAL' CONDITIONS 1* and 2*.  !

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ACTION:

a. With one reactor coolant system recirculation loop not in operation:
1. Within four hours:. I a) Place the recirculation flow control system in the Loop i Manual'(Position Control) mode, and '

b) Reduce THERMAL POWER to_1 70% of RATED THERMAL POWER, and, c) . Increase the MINI ICAL POWER RATIO (MCPR)fSafety Limit'by 0.01 to per Specification 2.1.2, and, 1.1 d) Reduce the Maximum erage Planar Linear Heat Generation Rate 7

_(MAPLHGR) limit per Specification 3.2.1, and, f V_

e) Reduce the Average Power Range Monitor (APRM). Scram and Rod

' Block-and Rod Block Monitor Trip Setpoints'and Allowable Values

.to'those applicable for single recirculation loop operation per

. Specifications 2.2.1, 3.2.2 and 3.3.6.

f). Reduce the volumetric drive flow rate of the operating recirculation' loop to's 41,800**:gpe.

  • See Special-Test Exception 3.10.4.
    • This value represents the volumetric recirculation loop drive flow which

. produces 1 05 core flow at 100% THERMAL POWER.

^ ^ ^ i; .. _ ....., . ..-.. ., ..., . ... .,. ...; tr.r;.;t tt; "' :t :;:::tk.;-

cye?:.

%%%. McPC values are ay(IIcable. % Cyele. 7 - ~oftra% onIy.

NINE MILE POINT - UNIT 2 3/4 4-1 AmendmentNo.16,//

Ci.,iiai.ed My C, 1 ;;0

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'2.1 BASES FOR SAFETY LIMITS _]

2.

1.0 INTRODUCTION

The fuel cladding, reactor pressure vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are l established to protect the integrity of these barriers during normal plant operations and . '

y an!.icipated transients. The fuel cladding integrity Safety Limit is set so that no fuel damage is calculated to' occur if the limit is not violated. Because fuel damage is not -

directly observable, a ek approach is used to establish a Safety Lmit so that the_

l.M WCPR"is not less tha _ for tw ecirculation operation andWMNor single 7.70

~ recirculation 1.000 operation! MCP rester than for two recirculaTson loop /,ay peration an or single recirculation loop opedtibn represents a conservativ margin relati@ve to the conditions required to maintain fuel cladding integrity. T cladding is one of the physical barriers that 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 that occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. Although fission product migration from cladding perforation is just as measurable 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 that would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of critical power correlations is not valid for all critical power calculations N i performed at reduced pressures below 785 psig or core flows less than 10% of rated i flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. v 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 essentially all elevation I 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' lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow ,

with a 4.5fpsi driving head will be greater than 28 x 108 lb/hr. Full scale ATLAS test l

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 THERMAL POWER for reactor pressure below 785 psig is conservative.

M Mcf& valuu are ay Mble. I 40 Cye.it. 7 N. y NINE MILE POINT - UNIT 2 B2-1 Amendment No. 34 /J(

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