ML20247Q326

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Proposed Tech Specs,Revising Min Critical Power Ratio Safety Limits
ML20247Q326
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 09/18/1989
From:
GEORGIA POWER CO.
To:
Shared Package
ML20247Q324 List:
References
NUDOCS 8909280167
Download: ML20247Q326 (4)


Text

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4 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 I and 2.

ACTION:

With 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 1.04 for two-loop recirculation or 1.05 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.

ACTION:

With MCPR less than 1.04 for two-loop recirculation or 1.05 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

REACTOR COOLANT 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:

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

p 6? 89091e i fCK05000366 eDc HATCH-UNIT 2 2-1 Proposed TS/0286q/198-77 I

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J 2.1 SAFETY LIMITS BASES 2.0 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 tran-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 1.04 for two-loop operation and 1.05 for single-loop operation. These limits represent a conservative margin relative to the corditions 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 ftom 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 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 cracki.7, 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 boiling, 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

3. Evaluation for single-loop operation demonstrates that two-loop transient and accident analyses are more limiting than single-loop, Reference 4.

2.1.1 THERMAL p0WER (Low Pressure or Low Flow)

The use of the GEXL correlation is not valid for all critical power calculations at 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, '

l the core pressure drop at low power and flows will always be greater l

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 bundle flow with a 4.5 psi driving head will be greater than 28 x 108 lbs/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 THERMAL l

POWER for reactor pressure below 785 psig is conservative.

HATCH - UNIT 2 B 2-1 Proposed TS/0287q/198-77

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Bases-Table B 2.1.2-1~

I UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT
  • Standard Deviation Quantity. (% of Point)

Feedwater Flow 1.76 Feedwater. Temperature 0.76

~ Reactor Pressure 0.5' T

Core Inlet Temperature' O.2 Core Total' Flow 2.5 Channel Flow Area 3.0~

Friction Factor Multiplier - 10.0 Channel Friction Factor Multiplier 5.0 1TIP Readings 8.7 R Factor 1.5 Critical Power 3.6 s

I^

  • The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

l HATCH - UNIT 2 B 2-4 Proposed TS/0287q/198-21 s.__-__.___._-_.----------A------ --

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I . .

POWER DISTRIBUTION LIMITS

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' BASES 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 operating-conditions

.as specified in Specification 3.2.3 are derived from the established fuel claddingLintegrity Safety Limit MCPR of 1.04 for two-loop operation and 1.05 l~

-for. single-loop operation, and an analysis of abnormal operational transients as described in References 1 and 3. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, 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 exceeded 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 evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

HATCH - UNIT 2 B 3/4 2-3 Proposed TS/0288q/198-77

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