ML20008D863

Proposed Tech Specs,Sections 2.0 Through 5.0,changing Safety Limits,Reactor Protection Sys Instrumentation Setpoints, Power Distribution Limits & Design Features Re Cycle 2 Operation W/Reload 1 Fuel Inserted

ML20008D863
Person / Time
Site:	Hatch
Issue date:	10/17/1980
From:	GEORGIA POWER CO.
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ML20008D862	List: ML20008D861 ML20008D863 ML20008D864
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NUDOCS 8010230372
Download: ML20008D863 (11)

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ATTACHMENT 3 NRC DOCKET 50-366 2

OPERATING LICENSE NPF 'EDWIN I. HATCH NUCLEAR PLANT UNIT 2 PROPOSED TECHNICAL SPECIFICATIONS j The proposed amendment to the Unit 2 Technical Specifications, Appendix A 4

.to the Operating License, would be incorporated as follows:

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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 r0WER 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: CONDfTIONS 1 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 1.07 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 1.07 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 recctor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: CONDITIONS 1, 2, 3 and 4.

ACTION:

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 < 1325 psig with 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

HATCH-UNIT 2- 2-1 t

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 Li'mit such that the MCPR is not less than 1.07. MCPR

> 1.07 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, fis-sion 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 signifi-cantly above desigt; conditions and the Limiting Sa~fety System Settings.

While fission product migration from cladding perforation is just as measurabla 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 deterioratian. 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 mndition intended by design for planned operation.

2.1.1 THERMAL POWER (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, 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 10d lbs/hr, bur:dle 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.10 3lbs/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 chan 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure blow 785 psig is conservative.

I 1

HATCH - UNIT 2 B 2-1

Bases Table B 2.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

• Standard Deviation Quanti ty (% of Point)

Feedwater Flow 1.76 Feedwater Temperature 0.76 ReactGr Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow 2.5 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings 8.7 R Factor 1.6 Critical Power 3.6

• 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.

HATCH - UNIT 2 B 2-4

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System . Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor co.ol, ant system are prevented from exceeding their Safety Limits. Operation with a trip set less conservative than its Trip Setpoint, but within its specified Allowable Value, is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

1. Intermediate Range Monitor, Neutron Flux The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10, ranges. Thus, as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems..

The most significant source of reactivity changes during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM's are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRf1 channel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 1%

of RATED THERMAL POWER, thus maintaining MCPR above 1.07. Based on this analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

2. Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and the Safety Limits. The margin accommodates the anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system. Temperature coefficients are small and-control rod patterns are constrained by the RSCS and RWM.

MATCH - UNIT 2 B 2-9

1 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR lit; EAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 1

3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as,a function of AVERAGE PLANAR EXPOSURE shall not exceed the linlits ih'own in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, or 3.2.1-4.

APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.

1 ACTI0fi:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3 or 3.2.1-4, initiate corrective action within 15 minutes and continue

. corrective action so that APLHGR is within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THE'RMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limit determined from Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, or 3.2.1-4:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have

been established, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

HATCH -. UNIT 2 3/4 2-1

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.t O 5.000 10.000 15.000 20,000 25,000 30,000 AVERAGE PLAflAR EXPOSURE (mwd /t)

FUEL TYPE P8DRB284LA itAXIMUM AVERAGE PLAtlAR LlflEAR llEAT GENERATI0ri RATE (MAPI.llGR)

VERSUS AVERAGE PLANAR EXPOSilRE FIGURE 3.2.1-2

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 . The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than the value shown below multiplied by the Kr shown in Figure 3.2.3-1.

a. 1.24 from beginning of cycle 2 to end of cycle 2.

APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% RATED THERMAL PDWER.

ACTION:

With MCPR less than the applicable limit determined from Figure 3.2.3-1, initiate corrective action within 15 minutes and c'ontinue corrective action so that MCPR is equal to or greater than the applicable limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined .to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and l

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for MCPR.

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i HATCH - UNIT 2- 3/4 2-6

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3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of coolant accident will not exceed the 22000F limit specified in the Final Ac. 3tance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.

3/4.2.1 AVERAGE PdAAAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature follow-ing the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification APLHGR is this LHGR of the highest powered rod divided by its local peaking factor.

The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.1-4.

The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.1-4 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses performed with Reference 1 are: (1) the analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.1-4; (2) fission product decay is computed assuming an energy release rate of 200 MEV/ fission; (3) pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; and (4) the effects of core spray entrainment and counter-current flow limitation as described in Reference 2, are included in the reflooding calculations.

A list of the significant plant input parameters to the loss-of-coolant accident analysis it presented in Bases Table B 3.2.1-1.

HATCH - UNIT 2 B 3/4 21

d POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER. The scram setting and rod block functions of the APRM instru-ments or APRM readings must be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation. The scram settings and rod block settings or APRM readings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

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 cladding integrity Safety Limit MCPR of 1.07, and an analysis of abnormal operational transients. 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 l insertion, and coolant temperature decrease. '

The limiting transient which determines the required steady state i MCPR limit is the load rejection trip with failure of the turbine bypa:,s. l This transient yields the largest A CPR. When added to the Safety Limit MCPR of 1.07 the required minimum operating limit MCPR of Specification 3.2.3 is obtained.

l HATCH - UNIT 2 B 3/4 2-3

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5.-0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall-be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 4-5.1.2 The low popuiation zone coincides with the exclusion area and is also shown in Figure 5.1.1-1.

5.2 CONTAINMENT i

CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a series  :

of vertical right cylinders and truncated cones which form a drywell.

4 This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a steel pressure vessei in the shape

of a torus. The primary containment has a total minimum free air volume

of 255,978 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a. . Maximum design internal pressure 56 psig.

1

b. Maximum allowable internal pressure 62 psig.

c. Maximum internal tempature 3400F.

4 d. Maximum external pressure 2 psig.

. 5.3 REACTOR CORE 1

FUEL ASSEMBLIES 1

5.3.1 The reactor core shall contain 560 fuel assemblies with each fuel assembly containing 62 fuel rods and '2 water rods clad with Zircaloy -4.

Each fuel . rod shall have a nominal active fuel ' length of 150 inches and

-contain a maximum total weight percent of 3341 grams uranium. The initial _ core loading shall have a maximum average ' enrichment of 1.87 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum average enrichment

of 2.90 weight - v cent U-235.

-HATCH - UNIT'2 5-1 L

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