ML20114B940

Amend 105 to License DPR-57,revising Tech Specs to Reflect Hardware Mods & Other Items That Are Part of Aprm,Rod Block Monitor & Tech Spec Improvement Program

ML20114B940
Person / Time
Site:	Hatch
Issue date:	12/31/1984
From:	Lainas G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20114B941	List: ML20114B940 ML20114B949 ML20114B962
References
TAC-54432, NUDOCS 8501290649
Download: ML20114B940 (43)

Text

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[aus 7o UNITED STATES

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g NUCLEAR REGULATORY COMMISSION 5

- j-WASHWGTON, D. C. 20555

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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA l

DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.105 License No. DPR-57 1.

The Nuclear Regulatory Comission (the Commission) has found that:

A.

The application for amendment by Georgia Power Company, et al.,

(the licensee) dated February 6, 1984, as supplemented September 6, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I;.

B.

The facility will-operate in conformity with the application, the provisions of the Act, and the rules and regulations of the-Comission~;

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted i.

in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security.or to the health and safety of the public; and

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E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical-Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:

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2 Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.105, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

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FOR THE NUCLEAR REGULATORY COMMISSION k,

Gus C. Lainas, Assistant Director for Operating Reactors Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December 31, 1904 l.

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ATTACHMENT TO LICENSE AMENDMENT NO.105 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

Remove Insert 9

1.1-1 1.1-1 1.1-2 1.1-2 1.1-3 1.1-3 1.1-4 1.1-4 1.1-10 1.1-10 1.1-12 1.1-12 1.1-13 1.1-13 1.1-14 1.1-14 1.1-17 1.1-17 Figure 1.1-1 3.1-1

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3.- 1-1 3.1-4 3.1-4 3.1-5 3.1-5 3.1-6 3.1-6 3.1-7 3.1-7 3.1-12 3.1-12 3.1-17 3.1-17 3.2-15 3.2-15 3.2-16 3.2-16 3.2-16a 3.2-17 3.2-17 3.2-65 3.2-65 3.3-5 3.3-5 3.3-15 3.3-15 3.3-19 3.3-19 Figure 3.6-5 Figure 3.6-5 3.6-10 3.6 -

3.11-1 3.11-1 3.11-2 3.11-2 l

3.11-2a 3.11-2a 3.11-3 3.11-3 3.11-4a 3.11-4a 3.11-5 3.11-5 3.11-6 3.11-6 Figure 3.11-1(Sheet 6)

Figure 3.11-1 (Sheet 7)

Figure 3.11.3 Figure 3.11.3 i

Figure 3.11.4 Figure 3.11.4 Figure 3.11.5 Figure 3.11.5 Figure 3.11.6 Figure 3.11.6 Figure 3.11.7

SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINC6 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability Applicability The Safety Limits established to pre-The Limiting Safety System Settings serve the fuel cladding integrity apply apply to trip settings of the instru-to those variables which monitor the ments and devices which are provided to fuel thermal behavior.

prevent the fuel cladding integrity Safety Limits from being exceeded.

Objective Objective The objective of the Safety Limits is The objective of the Limiting Safety to establish limits below which the System Settings is to define the level integrity of the fuel cladding is of the process variables at which auto-preserved.

matic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.

Specifications Specifications A. Reactor Pressure > 800 osia and Core A.

Trio Settings Flow > 10% of Rated The limiting safety system trip set-The existence of a minimum critical tings shall be as specified below:

power ratio (MCPR) less than 1.07 shall constitute violation of the

1. Neutron Flux Trip Settings fuel cladding integrity safety limit.

a. IRM High High Flux Scram Trio Setting B.

Core Thermal Power Limit (Reactor The IRM flux scram trip setting Pressure < 800 psia) shall be 5 120/125 of full scale.

When the reactor pressure is 5 800

b. APRM Flux Scram Trip Setting psia or core flow is less than 10% of (Refuel or Start & Hot Standby rated, the core thermal power shall Mode) not exceed 25% of rated thermal power.

When the Mode Switch is in the REFUEL or START & HOT STANDBY position, the APRM flux scram C.

Power Transient trip setting shall be 5 15/125 of full scale (i.e., 5 15% of rated To ensure that the Safety Limit estab-thermal power).

l lished in Specification 1.1.A and 1.1.B is not exceeded, each required

c. APRM' Flux Scram Trip scram shall be initiated by its Settings (Run Mode) expected scram signal. The Safety I

Limit shall be assumed to be exceeded (1) Flow Referenced Simulated when scram is accomplished by a means Thermal Power Monitor Scram other than the expected scram signal.

Trip Settinz When the Mode Switch is in the RUN position the APRM flow referenced simulated thermal powe'r scram trip setting shall be:

Amendment No. 27, 38. A2, 52, 5),105 1.1 1

SAFETY LIMITS LXMITING SAFETY SYSTEM SETTXNGS 1.1.D.

Reactor Water Level (Hot or Cold 2.1.A.1.c.(1)

Flow Referenced Simulated Shutdown Condition)

Thermal Power Monitor Trip Setting (Run Mode) (Continued)

Whenever the reactor is in the Hot or Cold Shutdown Condition with S s 0.58W + 62%

irradiated fuel in the reactor vessel, (Not to exceed 117%)

the water level shall be > 378 inches above vessel invert when fuel is where:

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seated in the core.

S = Setting in percent of rated thermal power

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(2436 MWt)

W = Loop recirculation flow-rate in percent of rated (rated loop recirculation flow rate equals 34.2 x 10' lb/hr) l.

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(2) Fixed APRM High High Flux Scram Trip Settina (Run Mode)

The APRM fixed flux scram trip setting shall not be allowed to exceed 120% of rated thermal power.

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1*1-2 Amendment No. 27, $2, J2, JS, $p, 7),

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SAFETY LIMITS LXMXTING SAFETY SYSTEM SETT1NGS 2.1.A.1.d APRM Rod Block Trip Setting This section deletad.

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2.1.A.2.

Reactor Vessel Water Low Level Scram Trip Setting (Level 3)

Reactor vessel water low level scram

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trip setting (Level.3) shall be 2 8.5 inches (narrow range scale).

3.

Turbine Stop Valve Closure Scram Turbine stop valve closure scram

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trip setting shall be 5 10 percent valve closure from full open. This scram is only effective when tur-bine steam flow is above that corresponding to 30% of rated core thermal power, as measured by turbine first stage pressure.

1*I'3 Amendment No, 7), Jt, Jp, 7), Jp),105

l SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.A.4 Turbine Control Valve Fast Closure Scram Trip Setting Turbine control valve fast closure scram trip shall initiate within 30 milliseconds

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of start of control valve fast closure. Fast closure is sensed by measuring electro-hydraulic control oil line pressure which decreases rapidly upon generator load rejection and just prior to fast clos'ure of the control-valves. This scram is only effective when turbine steam flow is above that corres-ponding to 30% of rated core thermal power as measured by turbine first stage pressure.

5.

Main Steam Line Isolation Valve Closure Scram Tric Set-i tinR Scram trip setting from ma'in steam line isolation valve closure shall be s 10 percent i

valve closure from full open.

This scram is effective in the Run Mode.

j 6.

Main Steam Line Isolation Valve Closure on Low Pressure Main steam line isolation valve closure on low pressure at inlet to turbine valves i

shall. occur at 2 825 psig, while in the Run Mode.

7.

Main Steam Line Isolation Valve Closure on Low Con-denser Vacuum Main steam line isolation valve closure on low condenser va-cuum shall occur at 2 7 inches Hg vacuum.

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l Amendment No. Jp, Jp),105 l

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BASES FOR LIMITING SAFETY SYSTEM SETTINGS

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2.1 FUEL CLADDING INTEGRITY i

The abnormal operational transients applicable to operation of the HNP-1. Unit have.been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 2537 MWt. The analyses were based upon plant operation in accordance with the operating map given in Figure 3-1 of Ref. 8.

In addition, 2436 MWt is the licensed maximum power level of HNP-1, and this represents the maximum steady-state power which shall >ot knowingly be exceeded.

Conservatism is incorporated in the transient analyses in estimating the con-trolling factors, such as void reactivity coefficient, control rod scram worth, i

scram delay time, peaking factors, and ' axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient 1

results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic performance. Results obtained from a i

General Electric boiling water reactor have been compared with predictions made

~by the model. The comparisons and results are summarized in Reference 1.

The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25% greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 80% of the total scram worth'of the control rods. The scram delay time and rate of rod insertion allowed by the analyses are conservatively set ' equal to the longest delay and slowest inser-tion rate acceptable by Technical Specifications. Active coolant flow is equal to 88% of total core flow. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the i

early portion of the negative reactivity insertion. The rapid insertion of negative reactivity is assured by the time requirements for 5% and 25% inser-tion. By the time the rods are 60% inserted, approximately four dollars of negative rea~ctivity have been inserted (see Figure 7-1, NEDO-21124-7) which strongly turns the transient, and accomplishes the desired effect. The times for 50% and 90% insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the.

ultimate fully shutdown steady-state condition.

For analyses of the thermal consequences of the transients, a MCPR equal to or greater than the actual operating limit MCPR is conservatively assumed to exist prior to initiation of the transients.

i This choice of using conservative values of controlling parameters and ini-tiating transients at the design power level, produces more pessimistic answers than would result by using expected values of control parameters and analyzing at higher power levels.

Steady-state operation without forced recirculation will not be permitted, except during startup testing. The analysis to support ope"ation at various i

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1**~1 Amendment No. 27,33. #2 105 M

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BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1.A.1.a. IRM Flux Scram Trio Settina (Continued) tism was taken in this analysis by assuming that the IRM channel closest to.the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited tc one percent of rated power, thus maintaining MCPR above 1.07.

Based on the above analysis, the IRM provides protection against local control rod withdrawal errors i

and continues withdrawal of control rods in sequence and provides backup protection for the APRM.

b. APRM Flux Scram Trip Settina (Refuel or Start & Hot Standbv Mode)

For operation in the startup mode whi~. the reactor is at low pressure, the APRM scram setting of 15 percint of rated power provides adequate thermal margin between the setpoint and the safety limit,'.25 percent of rated. The margin is adequate to accommodate anticipated maneuvers asso-cisted with power plant startup. Effects of increasing pressure at zero or low void content are minor, 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 to be

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uniform by operating procedures backed up by the rod worth minimizer and i

the Rod Sequence Control System. Worth of individual rods is very low i

j in a uniform rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause.of sig-nificant power rise. 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 percentage of rated power, the rate'of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.

In an assumed uniform i

rod withdrawal approach to the scram level, the rate of power rise is no 4

more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed 1

i the safety limit. The 15 percent APRM scram remains active until the mode ssitch is placed in the RUN position. This switch occurs when reactor pressure is greater than 825 psig.

c. APRM Flux Scram Trio Settinas (Run Mode)

The APRM T1ux scram trips in the run mode consist of the flow referenced simulated thermal power monitor scram setpoint and a fixed high-high neutron flux scram setpoint.

In the simulated thermal power monitor,

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t the APRM flow referenced neutron flux signal is passed through a filter-i ing network with a time constant which is representative of the fuel dy-I namics.

This provides a flow referenced signal that approximates the average heat flux or thermal power that is developed in the core during transient or steady-state conditions. This prevents spurious scrams, which have an adverse effect on reactor safety because of the resulting j

thermal stresses.

Enseples of events which can result in momentary neutron flux spikes are momentary flow changes in the recirculation system flow, and small pressure disturbances during turbine stop valve i

and turbine control valve testing. These flux spikes represent no hazard to the fuel since they are only of a few seconds duration and less than 120*. of rated thermal power.

1.1-12 Amendment No. 17, 78, $2. 52, 85,105 i

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BASES FOR LIMITING SAFETY SYSTEN,5JTTING5 2.1.A.1.c APRM Flux Scram Trip Settings (Run Mode) (Continued)

The APRM flow referenced simulated thermal power monitor scram trip setting at full recirculation flow is adjustable up to 117% of rated power.

This reduced flow referenced trip setpoint will result in an earlier scram during slow thermal transients, such as the loss of 100*F feedwater heating event, than would result with the 120% fixed high neutron flux scram trip. The lower flow referenced scram setpoint therefore decreases the severity (ACPR) of a slow theragl transient and allows lower Operating Limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the cycle.

The APRM fixed high-high neutron flux scram trip, adjustable up to 120%

l of rated power, does not incorporate the time constant, but responds directly to instantaneous neutros flux. This scram setpoint scrams the reactor during fast power increase transients if credit is not taken ~

for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced scram.

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Reactor vessel Water Low Level Scram Trio Settino (Level 3)

The trip setting for low level scram is above the bottom of the separator skirt. This level is > 14 feet above the top of the active fuel. This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in FSAR Section 14.3 show that a scram at this level adequately protects the fuel and the pressure barrier. The designated scram trip setting is at least 22 inches below the bottom of i

the normal operating range and is thus adequate to avoid spurious scrams.

!t Amendment No. 27, pp, 52, pg, pg, pp, 1.1 13 7),19),105 w-ur

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BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1. A.3. Tuhbine Stop Valve Closure Scram Trip Settinas -

The turbine stop valve closure scram trip anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram trip setting of 510 percent of valve closure from full open, the resultant increase.in surface heat flux is limited such that MCPR remains above 1.07 during the worst case transient that assumes the turbine bypass is closed. This scram is' bypassed wh'en turbine steam flow is below that corresponding to 30% of isted thermal power, as measured by turbine first stage pressure.

4. Turbine Control Valve Fast Closure Scram Trio Settina This turbine control valve ~ fast closure scram anticip'ates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability cf the turbine bypass. The Reactor Protection System initiates a scram when fast closure of the control valves is initiated by the fast acting solenoid valves. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protection system. This trip setting, a nominally 50% greater closure time and a different valve characteristic from that-of the turbine stop valve, combine to produce transients very similar and no more severe than for the stop valve. This scram is bypassed when turbine steam flow is below that corresponding to 30% of rated thermal power, as measured.by turbine first stage pressure.

5. Main Steam Line Isolation Valve Closure Scram Trio Settino The main steam l'ine isolation valve closure scram occurs within 10% of valve movement from the fully open position and thus anticipates the neutron flux and pressure scrams which remain as available backup pro-tection. This scram function is bypassed automatically when the Mode Switch is not in the RUN position.

6. Main Steam Isolation Valve Closure on Low Pressure The low pressure isolation of the main steam lines at 825 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel, which might result from a pressure regulator failure causing inadvertent opening of the control and/or-bypass valves.

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1 1 1-14 Amendment No, 37, 33, Jp, (3, pg, Jp),

105 l

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BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.1.C. References

1. FSAR Section 3.7.5.3, Performance Range for Normal Operation.

2. Linford, R.

B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NEDO-10802, Feb., 1973.

.3. FSAK Section 3.6.6, Nuclear Evaluations

4. FSAR:'Section 14.3, Analysis of Abnormal Operational Transients 5.'FSAR Section 7.5, Neutron Monitoring System 1

6.1FSAR Section 14, Plant Safety Analysis

7. "Edwin I.. Hatch Nuclear Plant Unit 1 Channel Inspection and Safety Analysis with Bypass Flow Holes Plugged," NEDO-21124, Nov., 1975.

8. " Average Power Range Monitor,' Rod Block Monitor and Technical Specifications Improvement (ARTS) Program'for Edwin I. Hatch Nuclear Plant, Units 1 and 2,"

NEDC-30474-P December 1983.

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' heendment No.105 1,1,17 ll' i

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM (RPS) 4.1 REACTOR PROTECTION SYSTEM (RPS)

Apolicability Applicability The Limiting Conditions for Operation The Surveillance Requirements asso-associated with the Reactor Protection ciated with the Reactor Protection i

System apply to the, instrumentation and System apply to the instrumentation s

associated devices which initiate a and associated devices which initiate reactor scram, a reactor scram.

Obiective Objective The objective ot' the Limiting Condi-The objective of the Surveillance tions for Operation is to assure Requirements is to specify the type the operability of the Reactor and frequency of surveillance to be Protection System.

applied to the protection instrumen-tation to assure operability.

Specifications Specifications A.

Sources of a Trip Signal Which A.

Test and Calibration Recuirements j

Initiate a Reactor Scram for the RPS Tha instrumentation requirements RPS instrumentation systems and associated with each source of a associated systems shall be func-scram signal shall be as given in tionally tested and calibrated as Table 3'.1-1.

indicated in Table 4.1-1.

The action to be taken if the number When it is determined that a of operable channels is not met for channel has failed in the unsafe both trip systems is also given in condition, the other RPS channels Table 3.1-1.

that monitor the same variable shall be functionally tested B.

Core Maximum Fraction of immediately before the trip system Limiting Power. Density (CMFLPD) containing the failure is tripped.

The trip system containing the This section deleted, unsafe failure may be placed in the untripped condition during the period in which surveillance testing is being performed on the other RPS channels.

7.

Core Maximum Fraction of Limiting Power Density (CMFLPD)

This section-deleted.

i Amendment' No. 73,105

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Table 3.1-1 (Cont'd) e

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. Scram Operable 4 JNumber Source of Scram Trip Signal Channels Scram Trip Setting Source of Scram Signal is (a)

Required Per Required to be Operable D

Trip System Except as Indicated Below (b)'

5 High Drywell Pressure 2

5 2 psig Not required to be operable when

.g primary containment integrity is o

not required. May be bypassed when necessary during purging for containment inerting or deinerting.

w 6

Reactor Vessel Water Level -

2

> 8.5 inches Low (Level 3) 7 Scram Discharge Volume High Permissible to bypass (initiates High Level control rod block) in order to reset RPS when the Mode Switch is

a. ~ Float Swltches 2

171 gallons b.

Jhermal Level Sensors 2

171 gallons 8

APRM Flow Referenced. Simulated 2

S 10.58 W + 62%

Thermal Power Honitor (Not to exceed 117%)

Tech Spec 2.1.A.I.c(1)

Fixed High High Neutron 2

S $_ 120% Power Flux Tech Spec 2.1.A.I.c(2)

Inoperative 2

Not Applicable An APRM is inoperable if there are less than two LPRM inputs

-per level or there are less than 11 LPRM inputs to the APRM channel.

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!I' Table 3.1-1 (Cont'd)

A y

Scram Operable ljtksber Source 'of Scram Trip Signal Channels Scram' Trip Setting Source of Scram Signal is

. Required P)r Required to be Operable g

(a)

Trip System Except as indicated Below

-(b)

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8 APRM Downscale 2

> 3/125 of full scale The APRM downscale trip is active only when the Mode Switch is in RUN. The APRM downscale trip is automati-cally bypassed when the IRM instrumentation is operable and not tripped.

15% Flux 2

< l5/125 of full The APRM 15% Scram is auto-P'

' icale Tech Spec natically bypassed when the T'

2.1. A. I.b -

Mode Switch is in the RUN v'

position.

9 Main Steam Line Radiation 2

< 3 times normal Not required if all steam P

~5ackground at rated lines are isolated, thermal power.

10 Main Steam Line Isolation 4

< 10% valve closure Automatically bypassed when Valve Closure

~From full open the Mode Switch is not in Tech Spec 2.1.A.5 the RUN position. The design permits clo:ure of any two lines without a scram being initia ted.

11 Turbine Control Valve 2

Within 30 milli-Automatically bypassed when Fast Closure seconds of the start turbine steam flow is below of control valve that corresponding to 30% of fast closure rated thermal power as measured Tech Spec 2.1.A.4.

by turbine first stage pressure.

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Table 3.1-1 (Cont'd) 3.

Scram Operable S

Number Source of Scram Trip Signal Channels Scram Trip Setting Source of Scram Signal is Required gy (a)

Required Per to be Operable Except as Indicated

+

Trip System Below

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(b) c5 c'

12 Turbine Stop Valve

'. 4 s10% valve closure Automatically bypassed when Closure from full open turbine steam flow is below that Tech Spec 2.1.A.3 corresponding to 30% of rated thermal power as measured by turbine first stage pressure.

Notes for Table 3.1-1 The column entitled " Scram Number" is for convenience so that a one-to one relationship can be established a.

j between items in-Table 3.1-1 and items in Table 4.1-1.

l b.

There shall be two operable or tripped trip systems for each potential scram signal. If the number of F

cperable channels cannot be met for one of the trip systems, that. trip system shall be tripped. However, d

7 one trip signal channel of a trip system may be inoperable for up to two (2)' hours during periods of required survelliance testing without tripping the associated trip system, pr< ided that the other remaining channel (s) monitoring that parameter within that trip system is (are) operable.

For SCRAMS 1 thru 7 and 8 APRM 15% Flux. if the number of operable channels is not met for both trip systems; initiate insertion of all control rods capable of being moved by control rod drive pressure and complete their Insertion within fo'er (4) hours.

For SCRAM 8 (APRM High Trips, inoperative, and Downscale), if the number of operable channels is not met for both trip systems; initiate insertion of all ' control. rods capable of being moved by control rod drive pressure and complete their insertion within four hours or reduce power to the IRN range and go to the l

START & HOT STANDBY position of the Mode Switch within eight hours.

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For SCRAMS 9 and 10, if the number of operable channels is not met for both trip systems; reduce. turbine i

load and close main steam line: Isolation valves within eight hours or initiate insertion of all control

)j rods capable of being moved by control rod drive pressure and complete their insertion within four hours.

For SCRAMS 11 and 12, if the number of operable channels is not met for both trip systems, reduce reactor power to 25% of rated thermal power of less within eight hours.

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S Table 4.1-1 I

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Reactor Protection System (RPS) Instrumentation Functional Test, Functional Test Minimus Frequency, and Calibration Minimum Frequency -

2 P

3 cran Instrument Functional Test Instrument Calibration-Number Source of Scram Trip Signal

Group, Minimum Frequency Minimum Frequency (a)

(b)

(c)

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1 Mode Switch in SHUTDOWN A

Once/ Operating Cycle Not Applicable 2

Manual Scram A

Every 3 months Not Applicable 8

3 IRM High High Flux C

Once/ Week during refueling Once/ Week and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of Startup (e)

Inoperative C

Once/ week during refueling Once/ Week and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of Startup (e)

P 4

Reactor Vessel Steam Dome D

Once/ Month Once/ operating cycle y

Pressure - High w

5 High Drywell Pressure A

Once/ Month (f)

Every 3 months 6

Reactor Vessel Water Level -

D Once/ Month (g)

Once/ Operating Cycle Low (Level 3) 7 Scram Discharge Volume High High 4

Level

.a.

Float Switches A

Once/ Month (f)

(h) b.

Thermal Level Sensors 8

Once/ Month (f)

Once/ operating cycle 8

APRN Fixed High-High Flux B

Once/ Peek (e)

Twfee/ Week l

Inoperable R

Once/ Week (e) gg,1ce/ Week Downscale 3

Once/ Week (e)

Twice/ Week Flow Referenced Simulated 3

Once/ Week (f)

Once/ Operating Cycle l

1 Thermal Power Monitor I

15% Flux ~

C Within 24 Ifours of Startup (e)

Once/ Week

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3.1.A.4.

Reactor Vessel Steam Dome Pressure - High (Continued) setting also protects the core from exceeding thermal hydraulic limits as a result of pressure increases from some events that occur when the reactor is operating at less than rated power and flow.

5.

High Drywell Pressure Pressure switch instrumentation for the drywel,1 is provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high

- drywell pressure scram is provided at the same setting (s2 psig) as the core standby cooling systems initiation to minimize the energy which must be accommodated during a loss of coolant accident. The instrumentation is a backup to the reactor vessel water level instrumentation.

6.

Reactor Vessel Water Level 6 Low (Level 3)

The bases for.the Reactor Vessel Water Level-Low Scram Trip Setting (Level

3) are discussed in the bases for Specification 2.1.A.2.

7.

Scram Discharge Volume High High Level The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. A part of this piping is an instrument volume which is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must bt accommodated during a scram. During normal operation the discharge volume is empty; however, should the discharge volume fill with. water, the water discharged to the piping from the reactor could not be accommodated which would result in a slow scram time or partial or no control rod insertion. To preclude this occurrence, level switches have been provided in the instrument volume which scram the reactor when the volume of water reaches 71 gallons. As indicated above, there is suffi-cient volume,in the piping to accommodate the scram without impairment of the scram times"or amount of insertion of the control rods. This function shuts i

~

the reactor down while sufficient volume remains to accommodate the dis-charged water and pre ~cludes the situation in which a scram would be required but not able to perform its function adequately.

8.

APRM Three APRM instrument channels are provided for each protection trip sys-tem. APRM's A and E operate contacts in one trip logic and APRM's C and E operate contacts in the other trip logic. APRM's'B, D and F are arranged similarly in the other protection trip system. Each protection trip sys-tem has one more APRM than is necessary to meet the minimum number're-quired per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.

i Flow Referenced Simulated Thermal Power Monitor and Fixed High-a.

High-Neutron Flux The bases for the APRM Flow Referenced Simulated Thermal Power i

Monitor and Fixed High-High Neutron Flux Scram Trip Settings are discussed in the bases for Specification 2.1.A.1.c.

j l

Amendment No. $7, JA4,105 3.1-12 n.-

-r_..

m

c BASES FOR SURVEILLANCE REQUIREMENTS 4.1.A.

Test and Calibration Requirements for the RPS (Continued)

~ Group C devices are active only during a given portion of the operational cycle.

For example, the IRM is active during startup and inactive during full power operation. Thus, the only test that is meaningful ts the one

~

performed just prior to shutdown or startup; i.e., the tests that are per-formed just prior to use of the instrument.

1 Calibration frequency of the instrument channel is divided into two cate-gories: They are as follows:

1.

Passive type indicating devices that can be compared with like units on a continuous reference.

11.

Vacuum tube or semiconductor devices and detectors that drift or lose sensitivity.

Experience with passive type _ instruments in generating stations and substa-tions indicates that the specified calibrations are adequate. For those devices which employ amplifiers, etc., drift spec,1fications call for drift to be less than 0.4%/ month; i.e.,

in the period of a month a drift of

.4% could occur and still provide for adequate margin.

For the ADRM system, drift of electronic apparatus is not the only consideration in determining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven (7) days. Calibration on this frequency assures plant operation at or below thermal limits.

The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. This is compensated for in the APRM system by calibrating twice a week using heat balance data and by calibrating individual LPRM's every 1000 effective full power hours using TIP traverse data.

Group D devices consist of analog transmitters, master trip units, slave trip units, and other accessories. The general description of the ATTS devices is provided in Reference 3.

As evidenced by NEDO-21617-A, the NRC has approved the following surveillance frequencies for ATTS equipment:

1 1.

Once per shift for channel check 2.

Once per month for channel functional test 3.

Once per operati.ng cycle for channel calibration B.

Maximum Fraction of Limitina Power Density (MFLPD)

This section deleted.

1 Amendment No. 7), JAMF.105 3,g 37

O l{

. Table 3.2'.7 NEllTRON HONITORINC INSTRINfENTATION WHICH INITIATES CONTROL ROD RLOCKS'

s t

l Ref.

Trip Required if No. -

Instrument Condition Operable Trip Setting Remarks (a)

Nomenclature Channels per Trip

- g System

.y 1

un 1

SPM Inoperative 2(b)(c)(d)

Not applicable Inoperativs trip produced by l

switch not in operate, power supply voltage low, and circuit i

boards not in circuit.

Not fully inserted 2(b)(c)(d)

Not fully inserted This function is bypassed when l

the count rate is > 100 cps or the IRMs are on range 3 or above.

E' Y

Downscale 2(b)(c)(d) 2: 3 counts /sec This function is bypassed when l

[

the count rate is > 100 cps or the IRMs are on range 3 or a bove,.

Upscale 2(b)(c)(d) s 105 counts /sec.

1 2

IRN Inoperative 3(b)(d)

Not applicable.

Inope :stive trip produced by j

switch not in operate, power supply voltage low,.or. circuit boards not in circuit.

+

Not fully inserted 3(b)(d)

Not fully inserted Only required in the Refuel and l

l Start & Hot Standby Modes.

Downscale-3(b)(d) 2: 5/125 of full scale Trip bypassed when IRN on Range 1.

I High Flux 3(b)(d) s108/125 of full scale 3

' APRM Inoperative 2(b)(e)

Not applicable Inoperative trip produced by switch not inoperate, power supply voltage low, or circuit boards not in circuit.

~

O

i 7

Table 3.2-7 (Continued)

~*

EI O,.

4 gp Required Operable w.

Ref.

Trip Channels I*

No.

Condition per Trip

g (a)

Instrument Nomenclature S ystem Tri.p Setting Remarks e

3 APRM Downscale 2(b)(e) 2: 3/125 of full scale Not required while performing low power physics test at atmospheric je pressure during or after refueling at power levels not to exceed 5 MWt.

2I 12% Flux 2(b)(e) s 12/125 of full scale This function is bypassed when the l

l Mode Switch is placed in the RUN ui position.

Upscale 2(b)(e) s 0.58 W + 50%

W is the loop recirculation flow l

P8 3

rate in percent of rated. Trip

[y level setting is in percent of rated power. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.

4 RBM Inoperative 1(e)(f)

Not applicable Inoperative trip produced by switch l not in operate, circuit boards not in circuit, fails to null, less 4

than required r. umber of LPRM inputs for rod selected.

Downscale 1(e)(f) 294/125 of full scale l

.I

~

Table 3.2-7 (Continued)

I Required 5

operable Ref.

Trip Channels af flo.

Condition per Trip

~

(a)

Instrument Nomenclature Systen Trip Setting-Remarks j

4 RilH lipscale 1(e)(f) m l.ow Trip 1 117/125 of' full scale There are three upscale trip levels.

Sctroint (LTSP)

Only one is applied over a specified operating core thermal power range..

Intermediate Trip S.111.2/125 of' full scale All RBH trips are automatically by-Setpoint (ITSP) passed below the low power setpoint.

,The upscale LTSP is applied between liigh Trip S.107.4/125 of. full scale the low power and the intermediate Setpoint (IITSP) power setpoints. The upscale ITSP is applied between the intermediate i

power setpoint and the high power setpoint. The upscale llTSP is applied above the high power set-i point.

w 4

.n d.

Power Range Not EE Setpoints applicable Low Power S.30% rated c' ore thermal Power range setpoints ' control the 4

Setpoint (LPSP) power enforcement of the appropriate up-scale trips over the proper core Intermediate Power S.65% rated core thermal thermal power ranges. The power Setpoint (IPSP) power signal to the RRH is provided by the APRN.

Illgh Power S 85% rated core thermal Setpoint (IIPSP) powe r J

l Rypass Time Not 5.2.0 seconds RBH bypass time 'd61ay is set low Delay (td2) applicable enough to assure minimum rod move-ment while upscale t rips are by-passed.

1 5-Scram liigh Water 1(g) 5.18 gallons Discharge Level Volume 4

4 e

a Notes for T.shi'e 3.2-7 9-3 a.

The column entitled "Ref. No." is only for convenience so that =a one-to-one relationship can be established between items in Table 3.2-7 and items in Table. 4.2-7.

E b.

For the START & HOT, STANDBY position of the Mode Switch, there shall be two operable or tripped systems 23 for each potential trip condition.

If the requirements established by the column cannot be met for one of the two trip systems, the condition,may exist-for up to.seven days provided that during that time g-the operable system is functionally tested immediately and daily thereaf ter; if this condition lasts longer us than seven days, the system shall be tripped.

If the requirements _est.ablished by this column cannot be for both trip systems, the systems shall be tripped.

met One of the four SRM inputs may be bypassed.

c.

d.' The SRM and IRM blocks need not be operable in the Run Mode. This function is bypassed when the Mode Switch is placed in the RUN position.

~

The APRM and RBM rod blocks need not be operable in the Start &' Hot Standby ' Mode (Except 12% APRM Rod Block),

e.

f.- The RBM is only required when core thermal power is > 30% and th'e limiting condition defined in Section j

3.3.F exists.

l

[,

g.

This trip is Operable in Power Operation and Hot Standby Mode, and Refuel Mode when any control rod is j,

withdrawn. Not applicable to control rods removed per Specification 3.10.F.

u

{

i f

,a l

i e

BASES FOR LIMITING CONDITIONS FOR OPERATIONS 1

3.2.G.3.b. Downscale A downscale indication of f 3/125 full scale on an APRM is an indication that the instrument has failed or the instrument 'is not ' sensitive enough.

j In either case, the instrument _will not respond to changes in control rod

)

motion and thus, control rod motion is prevented. The downscale trip is set 2 3/125 full scale.

c.12% Flux (Refuel and Start & Not Standby Modes)

~

This rod block anticipates the reactor scram which would occur at 15%

l

- rated thermal power (flux), thus preventing the scram by arresting rod i

movement. Thus the operator. is afforded a chance to evaluate the oper-l ating conditions and take suitable action before a scram is incurred.

d. Upscale (Flow Referenced)

An PRM rod block trip setting is flow referenced and prevents a signi-ficant reduction in NCPR, especially during operation at reduced flow.

4. RBM The RBM rod block function provides loca1' protection of the core; i.e., the prevention of boiling transition in a local region of the core from a single 4

rod, withdrawal error. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or calibration. This time period does not significantly increase the risk of an inadvertent con-trol rod withdrawal. The RBM function is required only when a limiting rod pattern for RWE (Section 3.3.F) exists.

1

a. Inoperative i

This red block assures that no control rod is withdrawn (above 30% power) i unless the RBM channels are in service or are properly bypassed.

- ~

b. Downscale i

This rod block as'sures that the RBM's are on scale in the power range or are properly bypassed.

c. Upscale' This rod block prevents the erroneous withdrawal of a single worst case control rod so that local fuel damage does not result. The RBM upscale setting is chosen so that no local fuel damage can occur from a single

. control rod withdrawal error during power range operation.

4 H. Radiation Monitorina Systems Which Limit Radioactivity Release (Table 3.2-8)

1. Off-Gas post Treatment Radiation Monitors i

Two air ejector off-gas post treatment radiation monitors are provided in a two from two logic arrangement for the purpose of isolating the off-gas line

[

from the main stack. Each monitor system has three upscale trips at differ-ont radiation levels namely HI, HI HI'and HI HI HI.

Additionally, a down-scale trip is provided which results from various inoperative conditions of the scnitor channel.

Isolation of the off gas line outlet and drain valves 3.2-65 Amendment No. 105

..m

.e e

4 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RECUIREMENT3 3.3.7.

Operation with a Limiting Control 4.3.F Ooeration with a Limitine control Rod Pattern (for kod Withdrawal Rod Pattern (for Rod Withdrawal Error. RWE)

Error. RWE)

A Limiting Rod Pattern for RWE exists During operation when a Limiting when:

Control Rod Pattern for RWE exists and only one RBM channel is 1.

Thermal power is below 90%

operable, an ingtrument functional of rated and the MCPR is less test of the RBM shall be performed 5

than 1.70, or prior to withdrawal of the control rod (s). A Limiting Rod Pattern for 2.

Thermal power is 90% of rated RWE is defined by 3.3.F.

or above and the MCPR is less than 1.40.

During operation with a Limiting Control Rod Pattern for RWE and when core thermal power is 2 30%,

either:

1.

Both RBM channels shall be oper-G.

Limitine the Vorth of a Control Red able, or Below 20*. Rated Thermal

• Power 1

2. 'If only'one RBM channel is oper-

1. Red Vorth Minimizer YRVM1 j

able, control rod withdrawal shall be blocked within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or Prior to the start of control rod -

withdrawal at startup, and as soon 3.

If neither RBM channel is oper-as automatic initiation of the RWM able, control rod withdrawal shall occurs during rod insertion while be blocked.

shutting down, the capability of the Rod Vorth Minimiser to properly i

G.

Limiting the Vorth of a Control Rod fulfill its function shall be veri-Belew 20% Rated Thermal Power fled by the following checks.

I 1.

Rod Worth-Minimizer (RVM)

a. The correctness of the control rod withdrawal sequence input to l

Whenever the reactor is in the Start the RWM computer shall be'yeri-l

& Hot Standby or Run Mode below 20%

fied.

1 rated thermal power, the Rod Worth Minimizer shall be operable or a

b. The RWM computer on line diag-second licensed operator shall nostic test shall be successfully verify that the operator at the performed.

reacter console is following the

• • control red program.

c. Proper annunciation of the selec-tion error of at least one out-of-sequence centrol red in each fully inserted greup shall be verified.

d. The rod block function of the RWM shall bi verified by withdrawing

)

or inserting an out-of-sequence control rod no more than to the 4

block point.

I Amendment No. 27,78,#2,52,105 3 3-3 I

. BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.3.F Ooeration with a Limiting Control Rod Pattern (for Rod Withdrawal Error, RWE) i Surveillance Requirements:

i A limiting control rod pattern for RWE is a pattern which, due to unrestricted f

withdrawal of any single contrcl rod, could result in violation of the MCPR Safety Limit.

Specification 3.3.F defines a limiting control rod pattern for RWE. During use of such patterns when both RBM channels are not ogerable, it is judged that testing of the RBM system prior to withdrawal of control rods to assure its operability will assure that improper withdrawal does not occur. Reference NEDC-30474-P (Ref. 17) for more information.

G.

Limiting the Worth of a Control Rod Below 20%' Rated Thermal Power

1. Rod Worth Minimizer (RVM)

Limiting Conditions for Operation:

Th.e Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS) restrict withdrawals and insertions of control rods to pre-specified sequences. All patterns associated with these sequences have the charac-teristics that, assuming the worst single deviation from the sequence, the i

drop of any control-rod. from the ' fully inserted position to the position of the control rod drive would not cause the reactor to sustain a power excursion resulting in any pellet average enthalpy in excess of 280 calories per gram. An enthalpy of 280 calories per gram is well below the level at which rapid fuel dispersal could occur (i.e., 425 calories per gram).

Primary system damage in this accident is not possible unless a significant amount of fuel is rapidly dispersed. Reference Sections 3.6.5.4, 3.6.6, 4

7.14.S.3, 14.4.2, and Appendix P of the FSAR, and NEDO-24040.

l l

A i

'l l

Amendment No. 27, $2, J2,105 3,3,33 I

i

a-e s

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.3.J. References (Continued) l 12.

FSAR Section 7.14.5.3, Rod Worth Minimizer Function 13.

FSAR Section 7.7.5, Rod Sequence Control System 14.

FSAR Section 3.6.4.1, Control Rods

~

15.

FSAR Question 3.6.7, Amendment 24

16. FSAR Appendix P, Rod Sequence Control System (RSCS)

.' i

" Average Power Range Molnitor, Rod Klock Monitor and Technical Specification 17.

Improvement (ARTS) Program for Edwin I. Hatch Nuclear Plant, Units 1 and 2,"

NEDC-30474-P, December 1983.

l.

t Amendment No. 105 3.3-19 i

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r EAmenhent No. 105 e

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e I,I:11TIh'G CONDITIONS FOR OPERATION SURVEILLANCE REnUIRE:1ENTS 3

4.6.I.

Jet Pm es (Continueo)

2. The indicated value of ccre flow rate varies fra the value derived frm locp ficw measurenents by more than lot.

l

3. The diffuser to lower plenum differential pressure reading on an individual jet pmp vary fran the mean of all jet pmp differential pressures by more than 104.

3.5.J.

Recirculation Pmp Speeds 4.6.J.

Recirculation Pmo Sc_ eeds

1. Core thermal power shall not exceed Recirculation pump speeds shall be 14 of rated themal power without recceded at least once per day.

forced. recirculation.

[2. Operation with' a single recirculation J

.p ep is permitted for 24' hours.unlessi the recirculation pump is sooner ~made operable. With one recirculation pump not in operation, initiate action within 15 minutes or continue action to reduce reactor power to or below the limit specified in Figure 3.6-5 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If the p mp cannot be made operable or the limit of Figure 3.6-5 cannot be met within the required time, the reactor shall be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. Followird one pmp operation the discharge valve of the low speed pm p' may not be opened unless the speed of the faster pmp is less than 50% of i

its rated speed.

i K.

Structural Intecrity of Primarv K.

Structural Intecrity of Primary Systen Eouncarv System Souncarv_

D e structural integrity of the A preservice inspection of i

primary system bouncary shall be accessible components listed ia maintained at the level required to Table 4.6-1 will be conducted I

assure safe operation throughout the before initial fuel loading to life of the unit. S e reactor shall establish a preservice base for be maintained in a cold shutdown later inspections. Se Condition until each indication of a nondestructive inspections listed defect has been investigated and in Table 4.6-1 shall be performed evaluated.

as specified. The results obtained fran empliance with this specification will te evaluated after 5 years and the conclusions of this evaluation will be reviewed with the NRC.

. Amendment No. N. H.105 3.6-10

_. _ _. _ -, ~ _

~

l LIMITING CONDITIONS FOR OPERATI0ii SUllVEILLAT4CE REQUIREMENTS 3.11 FUEL RODS 4.11 FUEL RODS Applicability Applicability The Limiting Conditions for Operation The Surve'111ance Requirements apply associated with the fuel rods apply to to the parameters which monitor the those parameters which monitor the fuel rod operating, conditions.

fuel rod operating conditions.

Objective Objective The Objective of the Limiting Condi-The Objective of the Surveillance tions for Operation is to assure the Requirements is to specify the type performance of the fuel rods, and frequency of surveillance to be applied to the fuel rods.

Specifications Specifications A. Averste Planar Linear Heat Genera-A. Averate Planar Linear Heat Genera-

_ tion Rate (APLHGR) tion Rate (APLHGR)

During power operation, the APLNGR The APLHGR for each type of fuel as for all. core locations shall not a function of average planar

', exceed the appropriate APLHGR limit exposure shall be determined daily for those core locations. The APLHGR during reactor operation at 2 25%

limit, which is a function of average rated thermal power.

planar exposure and fuel type, is the appropriate value from Figure 3.11-1, sheets 1 through 5, multiplied by the smaller of the two MAPFAC factors de-l termined from Figure 3.1171, sheets i

6 and 7.

If at any time during oper-l ation it is determine'd by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the pre-j scribed limits.

If the APLHGR is not returned to within the pre-scribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power with-in the next four (4) hours. If the limiting condition for operation is restored prior to expiration of the specified time interval, then further progression to less than 25% of rated thermal power is not required.

B. Linear Heat Generation Rate (LHGR)

B. Linear liest Generation Rate (LHGR) i During power operation, the LHGR as' The LHGR as function of core a function of core height shall not.

height shall be checked daily dur-exceed the limiting value shown in ing reactor operation at 2 25%

1 Figure 3.11-2 for 7 x 7 fuel or the rated thermal power.

limiting value of 13.4 kw/ft for 8 x 8/

8 x 8R fuel.

If at any time during Amendunt No. 51 A7, $7, 87, FS,105

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LIMITING CONDITIONS TOR OPERATION SURVEILLANCE REQUIREMENTS 3.11.B Linear Heat Generation Rate (LHGR)

(Continued) operation it is determined by normal surveillance that the limiting value for LHGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the LHGR is not returned to within the prescribed limits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power i

within.the next four (4) hours.

If

}

i the limiting condition for operation is restored prior to expiration of the specified time interval, then i

further progression to less than 25%

of rated thermal power is not required.

C.

Minimum Critical Power Ratio (MCPR) 4.11 C.1 Minimum Critical Power Ratio (MCPR) 4 1

The minimum critical power ratio (MCPR)

MCPR shall be determined to be shall be equal to or greater than the equal to or greater than the operating limit MCPR (OLMCPR), which applicable, limit, daily during 4

i is a function of scram time, core reactor power operation at 2 25%

power, and core flow. For 25*. $

rated thermal power and following

)

power < 30%, the OLMCPR is given in any change in power level or dis-Figure 3.11.7.

For power 2 30%,

tribution that would cause opera-l the OLMCPR is the greater of either:

tion with a limiting control rod pattern as described in the bases 1.

The applicabli limit determined for Specification 3.3.F.

l from Figure'3.11.3, or 4.11.C.2 Minimum Critical Power Ratio Limit j

2.

The applicable limit from either Figures 3.11.4, 3.11.5, TheMCPRlimitatratedflowanh or 3.11.6, multiplied by the rated power shall be determined for K factor determined from each fuel type, 8X8R, P8X8R, 7X7 P

Figure 3.11.7, where:

from figures 3.11.4, 3.11.5, and 3.11.6 respectively using:

t = 0 or

• ave *B

, whichever is a.

t=1.0 prior to initial scram

• A*B, greater t ue measurements for the cycle, performed in accordance T = 0.90 sec (Specifications 3.3.C.2.a A

with specifications 4.3.C.2.a scram time limit to 20% insertion 4

or from fully withdrawn) b.

t as defined in specification i

-h 3.11.C.

3 = 0.710+1.65 1

(0.053) [Ref.10]

The determination of the limit

}

T n

must be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> IN g of the conclusion of each scram i=1 time surveillance test required

~

by specification 4.3.C.2.

1 3.11-2 Amendment.No. 31, $2, if,16, 86,105 t

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.11. C. Minimum Critical Power Ratio (MCPR)

(Continued) n I Ntgg t,y, = i=1 n

I N g

i=1 n = number of surveillance tests performed to date in cycle N = number of active control rods g

th measured in the i surveillance test i

t = Average scram time to 20*. in-g sortion from fully withdrawn of all rods measured in the ch i

i surveillance test, and, 1 = total number of active rods measured in 4.3.C.2.a.

+

If at any time during operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. 'If the steady i

state MCPR is not returned to within the prescribed Jimits within two (2) hours, then reduce reactor power to less than 25% of rated thermal power within the next four(4) hours.'If the Limiting Condition for Operation is restored prior to expiration of the specified time interval, then i

further progression to less than j

25% of rated thermal power is not required.

D. Reportina Requirements If any of the limiting values iden-tified in Specifications 3.11.A.,

B., or C. are exceeded, a Reportable j

cccurrence report shall be submitted.

j If the corrective action is taken, as described, a thirty-day written report will meet the requirements 1

of this specification.

l Amendment No. pg,105 3.11-2a l

_,.,_.m..___

...-,e

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

?

3.11 FUEL RODS

.A. Averste Planar Linear Heat Generation Rate (APLHGR)

This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 30, Appendix K, even considering the postulated effects of fuel pellet densific,ation.

The peak cladding temperature following a postulated loss-of-coolant acci-dont is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent second-arily on the rod to rod power distribution within an assembly. Since ex-pected local variations in power distribution within a fuel assembly affect l

the calculated peak clad temperature by less than + 20*F relative to the

. peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures conform to 10 CFR 50.'46.

The limiting value for APLHGR at rated conditions i

is shown in Figures 3.11.1. sheets 1 thru 5.

l A flow dependent correction factor incorporated in to Figure 3.11-1 (sheet 7) is

)

applied to the rated conditions APLHGR to assure that.the 2200*F PCT limit is complied with during LOCA initiated from less than rated core flow.

In i

addition, other power and flow depsndent corrections given in Figure 3.11-1 (sheets 6 and 7) are applied to the rated conditions APLHGR limits to assure that the fuel thermal-mechanical design criteria are met during abnormal

}

transients initiated from off-rated conditions.

j The calculational procedure used to establish the APLHGR shown in Figures 3.11.1, sheets 1 thru 5, is based on a loss-of-coolant accident analysis.

The analysis was performed using General Electric (GE) calculational models 1

which are consistent with the requirements of Appendix K to 10 CTR 5_0.

A

}

complete discussion of each code employed in the analysis is presented in l

Reference 1.

Differences in this analysis as compared to previous analyses performed with Reference 1 are:

(1) The analyses assume a fuel assembly planar power consistent with 102*. of the MAPLHGR shown in Figure 3.11.1; 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 i

during the flow stagnation period; (4) The effects of core spray entrainment i

{

and counter-current flow limiting 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 is presented in Table 1 of NEDO-21187I3)

Further i

discussion of the APLHGR bases is found in NEDC-30474-PIII).

i i

e i

l Amendment No. If, 21, 33, #2, 96, 105 3.11-3 1

ii BASES FOR LIMYTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11.C.

Minimum Critical Power Ratio (MCPR)

(Continued)

The purpose of the MCPR, and the Kp of Figures 3.11.3 and 3.11.7, respectively, is g

to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power, the required MCPR is the larger value of the MCPRg and MCPRp at the existing core flow and power state. The MCPRgs are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCPR s were calculated such that for the maximum core flow rate and the corres-t ponding m_RMAL POWER along the 105% of rated steam flow control line, the limiting bundle's relative power was adjusted until t.he MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calculated at different j

points along the 105% of rated steam flow control line corresponding to different i

core flows. The calculated MCPR at a given point of core flow is defined as MCPR.

f i"

The core power dependent MCPR operating limit MCPRp is the power rated flow MCPR operating 1Leit multiplied by the K factor given in Figure 3.11.7.

p The K s are established to protect the core from transients other than core flow p

increases, including the localized event such as rod withdrawal error.

The K s p

were determined based upon the most limiting transient at the given core power 4

r level.

(For further information on MCPR operating limits for off-rated conditions, j

reference NEDC-30474-P. ' 8 8')

l i

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1 1

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i Amendment No.,42,105 3.11-4a i

BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11.C.

Minimum Critical Power Ratio (MCPR) (Continued)

I l

1 t

I i

1 1

1 i

At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content j

will be very small. For all designated control rod patterns which may be en-i ployed at this point, operating plant experience and thermal hydraulic analy-sis indicated that the resulting MCPR value is in excess of requirements by l

a considerable margin. With this low void content, any inadvertent core flow

)

increase would only place operation in a more conservative mode relative to i

MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at the 25*. thermal power level with minimum recirculation pump speed.

4 The MCPR margin'will thus be demonstrated such that future MCPR evaluations below this power level will be shown to be unnecessary. The daily require-i ment for calculating MCPR above 25*. rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant j

power or control rod changes. The requirement for calculating MCPR when a j

limiting control rod pattern is approached ensures that MCPR will be known

~

i following a change in power or power shape (regardless of magnitude) that j

could place operation at a thermal limit.

)

.D.

Reoortina Reauirements i

The LCO's associated with monitoring the fuel rod operating conditions are required to be met at all times, i.e. there is no allowable time in which the plant can knowingly exceed the limiting values for AFLHGR, LNGR, and i

MCPR.

It is a requirement, as stated in Specifications 3.11.A 3, and C that if at any time during steady state power operation, it is determined that the limiting values for APLHGR, LNGR, or MCPR are exceeded, action is then initiated to restore operation to within the prescribed limits.

l This action is initiated as soon as normal surveillance indicates that i

an operating limit has been reached. Each event involving operation beyond I

a specified limit shull be reported as a Reportable Occurrence.. If the specified corrective action described in the LCO's was taken, a thirty-day I

written report is acceptable.

1 3.11-3 Amendment No. M, M,105

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BASES FOR LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3.11.E. References 1.

General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566-P,, November,1975.

2.

General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to USAEC by letter, G. L. Gyorey to V. S,tello, Jr., dated December 20, 1974.

i 3.

Edwin I. Hatch Nuclear Plant Unit 1 Emergency Core Cooling System i

Analysis - Appendix K Requirement With Modified Low Pressure Coolant Injection System, NEDO-21187, Supplement 1, April, 1976.

4

" Fuel Densification Effects on General Electric Boiling Water Reactor 4

Fuel", Supplements 6, 7, and 8, NEDM-10735, August, 1973.

4 5.

Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 16, 1974 (USA Regulatory Staff).

6.

Communication:

V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27,1974 I

7.

"Edwin I. Hatch Nuclear Plant Unit 1 Channel Inspection and Safety

}

Analysis with Bypass Flow Holes Plugged". NEDO-21124-1, July, 1976.

8.

R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February, 1973 (NEDO-10802).

9.

General Electric Boiling Water Reactor Reload No. 1 Licensing Amendment for the Edwin I. Hatch Nuclear Plant Unit 1 Full Core Drilled Conditions, s

NEDO-21580, February, 1977.

1 1

10.

Letter from R. H.'Buchholz (G. E.) to P. 5. Check (NRC), " Response to NRC request for information on ODYN computer model", September's, 1980.

l 11.

" Average Power Range Monitor, Rod Block Monitor and Technical Specification l

Improvement (ARTS) Program.for Edwin I. Hatch Nuclear Plant, Units 1 and 2,"

j NEDC-30474-P, December 1983.

l i

i J

I 1

Amendment No. 3), 33, M, pp,105 3.11-6 i

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FIGURE 3.11.4 MCPR LIMIT FOR 8X8R FUEL FOR RATED POWER AND RATED FLOW 6

Amendment No. SS, 31,-95,105 9

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1B FIGURE 3.11.5 IWCPR LIMIT FOR P8X8R FUEL FOR RATED POWER AND RATED FLOW l

t Amendment No. SS, 37, pp, - 105 8

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8 FIGURE 3.11.6 MCPR LIMIT FOR 7X7 FUEL FOR RATED POWER AND RATED FLOW Amendment No.105

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25 OPERAT198G LIMIT MCPR (P) = kp *OPERATifeG LIMOT MCPR lle01 -

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FOR P < 25%: 980 THERMAL LIMITS MONITORING REQUORED I

NO LIMITS SPECIFIED I

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Kp =.1.15 + e.seee71e0% - Pi-I

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