ML19289F539
| ML19289F539 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/05/1979 |
| From: | Whitmer C GEORGIA POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7906110155 | |
| Download: ML19289F539 (11) | |
Text
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June 5, 1979 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN 1. HATCH NUCLEAR PLANT UNIT 2 PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS Gentlemen:
Pursuant to 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Georgia Power Company hereby proposes an amendment to the Technical Specifications (Appendix A to the Operating License).
The proposed change will substitute the use of the Core Maximum Fraction of Limiting Power Density (CMFLPD) for that of the 3bximum Total Peaking Factor (MTPF) in the determination of the Average Power Range Monitor (APRM) scram and rod block trip settings.
In a letter submitted to the Nuclear Regulatory Commission on October 18, 1978, Georgia Power Company requested this same change for Unit 1.
As mentioned in that letter, the replacement of MTPF with CfFLPD in the determination of the APRM scram and rod block trip settings will result in a more realistic approach to establishing and maintaining those settings during power operation.
This is due to the existence of assemblics of different active fuci lengths and with different numbers of rods in the Unit 1 core.
In the Unit 2 core, however, the assemblies are all the same length and contain the same number of rods; therefore, this change has no affect on establishing and maintaining APPJ1 scram and rod block settings. For this situation, CMFLPD and MTPF yield identical results.
The primary reason for changing the Unit 2 Technical Specifications is to provide for consistent procedural usage between units.
The reactor engineer will be responsible for both units and therefore con-sistent usage will eliminate a potential source of confusion during operation.
The Plant Review Board and the Safety Review Board have reviewed the proposed change to the Technical Specifications and have determined that it does not constitute an unreviewed safety question.
The proposed change will result in the use of the most limiting node in determining APRM settings; plant hardware is not modified by the proposed change.
Thus, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed have not been increased.
Similarly, 2230 336 7906110/ $ ~
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Director of Nuclear Reactor Regulation l
U. S. Nuclear Regulatory Commission June 5, 1979 Page Two because plant equipment is not changed, no new modes of failure will be introduced. Information required for power ascension and normal operation, which will be provided by CMFLPD, will more accurately reflect the reactor condition; thus, margins of safety are not reduced by the proposed change.
Yours very truly, NC 3
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s Chas. F. Whitmer MRD/RDB/mb Attachments (1) Proposed Determination of Amendment Class (2) Proposed Change to Technical Specifications Sworn to and subscribed before me this 5th day of June, 1979
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Ruble A. Thomas George F. Trowbridge, Esquire 2230 337 i
a ATTAOIMENT 1 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNIT 2 PROPOSED DETERMINATION OF AMENDMENT CLASS Pursuant to 10 CFR 170.12 (c), Georgia Power Company has evaluated the attached proposed amendment to Operating License NPF-5 and have determined that:
a)
The proposed amendment does not require the evaluation of a new Safety Analysis Report or rewrite of the facility license; b)
The proposed amendment does not contain several complex issues, does not involve ACRS review, or doas not require an environmental impact statement; c)
The proposed amendment does not involve a compicx issue, au.nviron-mental issue or more than one safety issue; d)
The proposed amendment does involve a single issue; namely, the substitution of CMFLPD for MTPF in the determination of APRM scram and rod block trip settings; c)
The proposed amendment is therefore a Class III amendment.
2230 338 4
e
ATTAQlMENT 2 NRC DOCKET 50-366 Ol'ERATING LICENSE NPF-5 EDWIN I. HATOI NUCLEAR PIRIT 1.711T 2 PROPOSED CHANCES TO TECHNICAL SPECIFICATIONS The proposed changes to the Technical Specifications (Appendix A to Operating License DPR-57) would be incorporated as follows:
Remove Page Insert Page I
I II II 1-4 1-4 1-6 1-6 B 2-10 B 2-10 3/4 2-5 3/4 2-5 B 3/4 2-3 B 3/4 2-3 2230 339
INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1-1 ACTI0N.....................................................
1-1 AVERAGE PLANAR EXP0SURE............................
AVERAGE PLANAR LINEAR llEAT GENERATION RATE.................
1-1 1-1 CIIANNEL CALIBRATION..............
CllANNEL CllECK..............................................
1-1 CilANNEL FUNCTIONAL TEST....................................
1-1 CORE ALTERATION............................................
1-2 C RITICAL P OWER RATI 0.......................................
1-2 DOSE EQUIVALENT I-131......................................
1-2
'ii-AVERAGE DISINTEGRATION ENERGY............................
1-2 EMERGENCY CORE COOLING SYSTDI (ECCS) RESPONSE TIME.........
1-2 FREQUENCY N0TATION.........................................
1-3 IDENTIFIED LEAKAGE.........................................
1-3 ISOLATION SYSTDI RESPONS E TIME.............................
1-3 LIMITING CONTROL ROD PATTERN...............................
1-3 LINEAR HEAT GENERATION RATE................................
1-3 LOGIC SYSTEM FUNCTIONAL TEST...............................
1-3 CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY...........
1-4 MINIMUM CRITICAL POWER RATI0...............................
1-4 OPERABLE - OP ERABILITY.....................................
1-4 OPERATIONAL C0NDITION......................................
1-4 PHYSICS TESTS..............................................
1-4 2230 340 llATCll-UNIT 2 I
INDEM DEFINITIONS SECTION 1.0 DEFINITIONS (Continued)
PAGE PRESSURE BOUNDARY LEAKAGE..................................
1-4 PRIMARY CONTAINMENT INTEGRITY..............................
1-5 RATED TilERMAL P 0WER........................................
1-5 REACTOR PROTECTION SYSTEM RESPONSE TIME....................
1-5 REPORTABLE OCCURRENCE......................................
1-5 R0D DENSITY................................................
1-5 SECONDAP.Y CONTAINMENT INTEGRITY............................
1-6 S ilUTDOWN MARGI N............................................
1-6 STAGGERED TEST BASIS.......................................
1-6 TilERMAL P0WER..............................................
1-6 FRACTION OF LIMITING POWER-DENSITY.........................
1-6 UNIDENTIFIED LEAKAGE.......................................
1-6 TABLE 1.1, SURVEILLANCE FREQUENCY N0TATION......................
1-7 TABLE 1.2, OPERATIONAL CONDITIONS...............................
1-8 2230 341 IIATCll - UNIT 2 II l
DEFINITIONS CORE MAXIMLH FRACTION OF LIMITING POWER DENSITY 1.18 The CORE !LiXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD) shall be the largest FLPD which exists in the core for a given operating condition.
MINIMUM CRITICAL POWER RATIO 1.19 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.
OPERABLE - OPERABILITY 1.20 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).
Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of per-forming their related support function (s).
OPERATIONAL CONDITION 1.21 An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.
PIIYSICS TESTS 1.22 PlIYSICS TESTS shall be those tests performed to measure the fun-damental nuclear characteristics of the reactor core and related instru-mentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE 1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.
2230 342 IIATCil-UNIT 2 1-4
DEFINITIONS SECONDARY CONTATNMENT INTEGRTTY 1.29 SECONDARY CONTAINMENT INTEGRITY shall exist when:
1.29.1 All secondary containment ventilation system automatic isolation dampers are OPERABLE or secured in the isolated position.
1.29.2 The Standby Gas Treatment System is OPERABLE pursuant to Specifi-cation 3.6.6.1.
1.29.3 At least one door in each access to the secondary containment is closed.
1.29.4 The sealing mechanism associated with each penetration (e.g.,
welds, bellows or 0-rings) is OPERABLE.
SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be subcritical from its present condition assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e.
68 F; and xenon free.
STAGGERED TEST BASIS 1.31 STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.
The testing of one system, subsystem, train or other designated components at the beginning of each subinterval.
THERMAL POWER 1.32 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
i FRACTION OF LIMITING PONER DENSITY 1.33 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the ratio of local LHGR for any specific location on a fuel rod divided by the design LHGR.
UNIDENTIFIED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
2230 343 ilATCll-UNIT 2 1-6
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Average Power Range Monitor (Continued)
Of all the possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power increase.
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shut-down before the power could exceed the Saf ety Limit.
The 15% neutron flux trip remains active until the mode switch is placed in the RUN position.
The APRM flow biased trip system is calibrated using heat balance data taken during steady state conditions.
Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation; i.e., the thermal power of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer. Analyscs demonstrate that with only the 120% neutron flux trip setting, none of the abnormal operational transients analyzed violates the fuel Saf ety Limit and there is substantial margin from fuel damage.
Therefore, the use of the flow referenced trip setpoint provides adequate margins of safety.
The APRM trip setpoint was selected to provide adequate margin for the Safety Limits and yet allows operating margin that reduces the possibility of unnecessary shutdown.
The APRM gains must be adjusted by the specified formula in Specification 3.2.2 in order to maintain these margins _when the CMFLPD exceeds the fraction of RATED THERMAL POWER.
3.
Reactor Vessel Steam Dome Pressure-High liigh pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to in-crease the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counter-acting the pressure increase by decreasing heat generation. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips.
The setting provides for a wide margin 2230 344 IUCDQI - UNIT 2 B 2-10
POL'ER DISTRIBUTIO" LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITIOf FOR OPERATION 3.2.2 The flow biased APRM scram trip setpoint (S) and rod block trip set-point (SRB) shall be established according to the following relationships:
S < (0.66W + 54%)
SRB 1 (0.66W + 42%)
where:
S and S re in Percent of RATED THERMAL POWER, RB W = Loop recirculation flow in percent of rated flow.
In the event of operation with a CORE MAXIFRDI FRACTION OF LIMITING POWER DENSITY (CMFLPD) greater than the FRACTION OF RATED THERMAL POWER (FRCTP),
Core Mi thermal (FRCTP = 2436 M1 thermal), the APRM gains shall be adjusted such that:
APRM readings > 100% x CMFLPD.
APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.
ACTION:
ding the allowable value or APRM readings < 100% x CMFLPD With S or S exc RB initiate corrective action within 15 minutes and continue correceive action so that S and S or APRM readings as appropriate are within the required limitswithin2$oursorreduceTIIERMALPOWERtolessthan25%ofRATEDTHERMAL 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 4.2.2 The CMFLPD shall be determined and the APRM readings adjusted, as required:
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 TilERMAL POWER has been increased by at least 15% of RATED TilERMAL POWER and steady state operating conditions have been established, and 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 c.
operating with a CMFLPD > FRCTP.
2230 545 llATCl! UNIT 2 3/4 2-5
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 power.
The APRM readings are adjusted in accordance with this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power dis-tribution, to ensure that an LHGR transient would not be increased in the degraded condition.
3/4.2.3 MINIMU'I CRITICAL POWER RATIO The required operating limit MCPRs at steady state opercting conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, 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 is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result 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.
The limiting transient which determines the required steady state MCPR limit is the load rejection trip with failure of the te:bine bypass.
This transient yields the largest A MCPR. When added to the Safety Limit MCPR of 1.06 the required minimum operating limit MCPR of Specification 3.2.3 is obtained.
2230 346 HATCH - UNIT 2 B 3/4 2-3