ML20153G959

Corrected Amend 97 to License DPR-61

ML20153G959
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/28/1988
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML20153G944	List: ML20153G942 ML20153G959 ML20153G967
References
NUDOCS 8805120069
Download: ML20153G959 (3)

Text

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ATTACHNENT TO LICENSE AMEN 0HENT NO. 97 FACILITY OPERAT!NG LICENSE NO. OPR-61 DOCKET NO. 50-213 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of c.hange.

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g. g l-3 1-3 1-6 1-6 2-1 2-1 Figure 2.2-1 Figure 2.2-1 Figure 2.2-2 Figure 2.2-2 22 2-2 2-3 2-3 2-5 2-5 2-7 2-7 2-8 2-8 3-3 3-3 3-4 3-4 3-4a 3-da 3-4b 3-4b 3-4c 3-4c 3-4d 3-4d

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SECTION 2.0 SAFETY LIMITS AND MAXIMUM SAFETY SETTINGS 2.1 Introduction 2-1 2.2 Reactor Core 2-1 2.3 Reactor Coolant System Pressure 2-4 2.4 Protective Instrumentation 2-5 SECTION 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Introduction 31 3.2 Reactor Coolant System Activity 3-2 3.3 Reactor Coolant System 33 3.4 Combined Heat up, Cooldown and Pressure Limitations l 35 3.5 Chemical and Volume Control System 3-8 3.6 Core Cooling Systems 3-10 3.7 Minimum Water Volume and Boron Concentration in the Refueling Water Storage Tank 3-11 3.8 Turbine Cycle 3-12 3.9 Operational Safety Instrumentation and Control Systems 3 14 3.10 Reactivity Control Systems 3 16 l 3.11 Container.nt 3 18 3.12 Station Service Power 3-21 3.13 Refueling 3 25

3.14 Primary System Leakage 3-27 3.15 Intee.tionally Left Blank 3-27 3.16 Intentionally left Blank 3-28 3.17 Power Distribution Limits 3-30 3.18 Intentionally Left Blank 3-32 3.19 Hydraulic Snubbers 33S 3.20 Intentionally Left Blank 3 39 ]

3.21 Safety Related Equipment Flood Protection 3 40a 3.22 Fire Protection Systems 3 41 3.23 Post Accident Instrumentation 3 45 3.24 Special Test Exceptions 3-47 l

!! Amendment No. 97 Corrected

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,)i RASES 2.2.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation 'to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and, therefore, THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to ONB through the W 3 correlation. The W 3 DNB cor*elathn has been developed to predict the DNB flux and the location of DNB for axially uniform ahd nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The minimum value of the DNBR during steady. state operation, normal operational transients, a~nd anticipated transients is limited to 1.30. This value corresponds to a 95 v ecent probability at a 95 percent corfidence level that DN', will not occur and is chosen as an 0ppropriate margin to 'JNB for all operating conditions.

The curves of Figures 2.21 and 2.2-2 show loci of points 6f l THERMAL POWER, pressurizer pressure and core inlet temperature for which the minimum DNBR is no less than 1.30, and the core outlet void fraction is no greater than 0.32.

These curys are based on nuclear enthalpy hot channel factors, F of 1.60 and 1.64 for four loop and three loop operation,#r ,upectively. An allowance is included for an increase in Fy at reduced power.

These limiting hot channel factors are higher than those calculated at full power for the range fron all control rods fully withdrawn to maximum allowable control rod insertion.

This insertion limit is described in Specification 3.10.2.6 and 3.10.2.7. The required reduction in power level as Figures 3.101 and 310 2 insures that the DNB dictated b{ ways greater than 1.30.

ratio is a.

22 Amendment 97 Corrected

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