ML20247J716

Amend 42 to License NPF-43,revising Tech Specs Re Operation of Unit During Cycle 2 W/Reload of GE Mfg Fuel Assemblies & Analyses & Methodologies

ML20247J716
Person / Time
Site:	Fermi  (NPF-43-A-042, NPF-43-A-42)
Issue date:	09/13/1989
From:	Thoma J Office of Nuclear Reactor Regulation
To:
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ML20247J724	List: ML20247J716 ML20247J728
References
NUDOCS 8909200260
Download: ML20247J716 (45)

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DETROIT EDISON COMPANY WOLVERINE POWER SUPPLY COOPERATIVE, INCORPORATED DOCKET NO. 50-341 FERMI-2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 42 License No. NPF-43 1.

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

A.

The applications for amendments by the Detroit Edison Company (the licensee) dated April 3 and May 31, 1989, and as supplemented by letter dated August 23, 1989, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; C.

There is reasonable assurance (i) 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 in compliance with the Commission'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 E.

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

2.

Accordingly, paragraph 2.C.(1) of Facility Operating License No. NPF-43 is hereby amended to read as follows:

Maximum Power Level DECO is authorized to operate the facility at reactor core power levels not in excess of 3293 megawatts thermal (100% power) in accordance with the conditions specified herein and in Attachment I to this license. The items identified in Attachment I to this license shall be completed as specified. is hereby incorporated into this license.

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Also, 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. NPF-43 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 42, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

DECO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

4.

This license amendment is effective as of its date of its issuance and will be fully implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATO Y COMMISSION w

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John 0. Thoma, Acting Director

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Project Directorate III-1 A{L Division of Reactor Projects - JII, IV, V & Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: September 13., 1989 i

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[s ATTACHMENT TO LICENSE AMENDMENT NO. 42 FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

i REMOVE INSERT xxi xxi xxii xxii 1-2 1-2 1-5 1-5 2-1 2-1 2-4 2-4 B 2-2 B 2-2 B 2-3 8 2-3 8 2-4 B 2-4 3/4 1-16 3/4 1-16 3/4 2-1 3/4 2-1 3/4 2-4 3/4 2-4 3/4 2-4A 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-6A 3/4 2 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-8A 3/4 2-8B 3/4 2-9 3/4 2-9 3/4 3-44 3/4 3-44 3/4 2-10 3/4 2-10 B 3/4 1-3 8 3/4 1-3 8 3/4 1-4 8 3/4 1-4 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 B 3/4 2-4 8 3/4 2-4 B 3/4 2-4A B 3/4 2-5 8 3/4 2-5 5-5 5-5

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INDEX 1.

LIST OF FIGURES' FIGURE PAGE 3.1.5-1 S0DIUM PENTABORATE VOLUME / CONCENTRATION REQUIREMENTS...................................

3/4 1-21 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183..................

3/4 2-2 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL-TYPE 8CR233..................

3/4 2-3 3.2.1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, RELOAD FUEL TYPE BC318D........................

3/4 2-4 3.2.1-4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, RELOAD FUEL TYPE BC318E.........................

3/4 2-4A 3.2.3-1 BOC TO 12,700 MWD /ST, MINIMUM CRITICAL POWER RATIO (MCPR) VS. TAU AT RATED FLOW............

3/4 2-8

.3.2.3-1A

-12,700 MWD /ST TO 13,700 MWD /ST, MINIMUM CRITICAL POWER RATIO (MCPR) VS. TAU AT RATED FLOW......

3/4 2-8A 3.2.3-1B 13,700 MWD /ST TO EOC, MINIMUM CRITICAL POWER RATIO (MCPR) VS. TAU AT RATED FLOW............

3/4 2-8B 3.2.3-2 FLOW CORRECTION (K ) FACTOR....................

3/4 2-9 f

3.4.1.1-1 THERMAL POWER.VS. CORE FLOW....................

3/4 4-3 3.4.6.1-1 MINIMUM REACTOR PRESSURE VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE....................

3/4 4-21 4.7.5-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST.....

3/4 7-21 B 3/4 3-1 REACTOR VESSEL WATER LEVEL..................... B 3/4 37 B 3/4.4.6-1 FA'iT NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE........................ B 3/4 4-7 B 3/4.6.2-1 LOCAL POOL TEMPERATURE LIMIT.................... B 3/4 6-5 B 3/4.7.3-1 ARRANGEMENT OF SHORE BARRIER SURVEY POINTS...... B 3/4 7-6 5.1.1-1 EXCLUSION AREA.................................

5-2 5.1.2-1 LOW POPULATION ZONE............................

5-3 5.1.3-1 MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS......................................

5-4 FERMI - UNIT 2 xxi Amendment No. 30,42 I

INDEX LIST OF TABLES TABLE PAGE 1.1 SURVEILLANCE FREQUENCY NOTATION................

1-9 1.2 OPERATIONAL CONDITIONS.........................

1-10 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS......................................

2-4 1

3. 3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION......

3/4 3-2 3.3.1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES.......

3/4 3-6 4.3.1.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................

3/4 3-7 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION............

3/4 3-11 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS..

3/4 3-15 3.3.2-3 ISOLATION ACTUATION SYSTEM INSTRUMENTATION RESPONSE TIME..................................

3/4 3-18 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS..................................

3/4 3-20 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..........

3/4 3-24 3.3.3-2 EMERGENCY CORE COOLING SYSTEM ACTUA110N INSTRUMENTATION SETPOINTS......................

3/4 3-27 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES...

3/4 3-29 4.3.3.1.1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS......

3/4 3-30 3.3.4-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION................................

3/4 3-33 3.3.4-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS.....................

3/4 3-34 FERMI - UNIT 2 xxii Amendment No.' 42

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1 DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of

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. fuel, sources, incore instruments or reactivity controls within the-reactor pressure vessel with the vessel head removed and fuel in the vessel.

Normal movement of SRMs, IRMs, TIPS, or special movable detectors is not considered a CORE ALTERATION.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to'a safe conservative position.

CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of an NRC approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The' thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the-total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11-The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.

Times-shall include diesel generator starting and sequence loading delays where applicable.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

FRACTION OF LIMITING POWER DENSITY 1.12 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type.

FRACTION OF RATED THERMAL POWER 1.13 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

FERMI - UNIT 2 1-2 Amendment No. 42

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.The_ following terms are defined so that uniform interpretation of the specifications may be achieved.

The defined terms appear in capitalized eype and shall be applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

AVERAGE PLANAR EXPOSURE

1. 2 The AVERAGE PLANAR EXPOSURE shall be applicable.to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of. fuel rods in'the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATIN RATE

1. 3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1.4 A CfiXRIE WIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

Calibration of instrument channels with resistance temperature detectors (RTD) or thermocouple sensors shall consist of verification of operability of the sensing element and adjustment, as necessary, of the remaining adjustable devices in the channel.

CHANNEL CHECK

1. 5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation.by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument r.nannels measuring the same parameter.

CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b.

Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

FERMI - UNIT 2 1-1 Amendment No. D, 42

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DEFINITIONS 2.

Closed by at least one manual valve, blank flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

c.

Each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.

d.

The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e.

The suppression chamber is in compliance with the requirement of Specification 3.6.2.1.

f.

The sealing mechanism associated with each primary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.

g.

The suppression chamber to reactor building vacuum breakers are in compliance with Specification 3.6.4.2.

THE PROCESS CONTROL PROGRAM 1.30 The PROCESS CONTROL PROGRAM (PCP) shall contain the provisions to assure i

that the SOLIDIFICATION of wet radioactive wastes results in a waste form l

with properties that meet the requirements of 10 CFR Part 61 and of low-level radioactive waste disposal sites.

The PCP shall identify process parameters influencing SOLIDIFICATION, such as pH, oil content, H 0 content, solids content, ratio of solidification agent to waste 9

ahd/or necessary additives for each type of anticipated waste, and the acceptable boundary conditions for the process parameters shall be identified for each waste type, based on laboratory scale and full scale testing or experience.

The PCP shall also include an identification of conditions that must be satisfied, based on full scale testing, to assure that dewatering of bead resins, powered resins, and filter sludges will result in volumes of free water, at the time of disposal, within the limits of 10 CFR Part 61 and of low-level radioactive waste disposal sites.

PURGE - PURGING 1.31 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3293 MWT.

l FERMI - UNIT 2 1-5 Amendment No. 42

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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 POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 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 /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than the Safety Limit MCPR of 1.07 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS I and 2.

ACTION:

With MCPR less than the Safety Limit MCPR of 1.07 and the reactor vessel steso l

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 /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the rcactor 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 less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

l FERMI - UNIT 2 2-1 Amendment No. 42

j SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued)

REACTOR VESSEL WATER LEVEL 1

2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

APPLICABILITY:

OPERATIONAL C0h0ITIONS 3, 4 and 5 ACTION:

With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required.

Comply with the requirements of Specification 6.7.1.

FERMI - UNIT 2 2-2

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS g

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION' SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor' protection system instrumentation setpoints shall be set i

I consistent with the Trip Setpoint values shown in Table 2.2.1 o l

APPLICABILITY:

As shown in Table 3.5.1-1.

ACTION:

With a' reactor protection system instrumentation setpoint l'ess conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable 'and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its-setpoint adjusted consistent with.the Trip Setpoint value.

i FERMI - UNIT 2 2-3

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2.1 SAFETY LIMITS BASES

2.0 INTRODUCTION

~ 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 transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.06.

MCPR greater than 1.06 represents a conser-vative margin ielative 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, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a signi-ficant departure from the condition 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 103 lbs/hr, bundle pressure drop is tiearly 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 10s lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-cal power at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL PCHER.

Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

FERMI - UNIT 2 B 2-1

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. SAFETY LIMITS BASES

'2.1.2 THERMAL POWER, High Pressure and High Flow

.The fuel cladding integrity Safety Limit is set such that no mechanistic-fuel damage is calculated to occur if the limit is not violated.

Since the

_ parameters which result in fuel damage are not directly observable during reat-

. tor operation, the thermal and hydraulic conditions resulting in_a departure from nucleate boiling have been used to. mark the beginning of the region where fuel damage could occur. Although_it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.

However, the uncertainties in monitoring the core' operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to. avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in the operating parameters and in the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is determined using the approved critical power correlation.

Details of the fuel cladding integrity safety limit calculation are given in Reference 1.

Reference 1 includes a tabulation of the uncertainties used in the determination of the safety Limit MCPR and of the nominal values of parameters used in the Safety Limit MCPR statistical analysis.

Reference 1.

" General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (latest approved revision).

FERMI - UNIT 2 B 2-2 Amendment No. 42

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t FERMI - UNIT 2 B 2-3 Amendment No. 42

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FERMI - UNIT 2 B 2-4 Amendment No. 42

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~ REACTIVITY' CONTROL-SYSTEMS' s

' CONTROL' ROD DRIVE HOUSING ~ SUPPORT LIMITING CONDITION FOR OPERATION D

3.1.3.8-The control rod drive' housing support shall be in place.

p.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

h; With.the control rod' drive housing support not in place, be in at leest HOT

. SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ~and-in COLD SHUTDOWN _within the followingl24 hours.

SURVEILLANCE REQUIREMENTS 14.1.3.8 The control rod drive housing support shall be verified to be in place by a. visual inspection prior to startup any time it has been disassembled or when maintenance has been performed in the' control rod drive housing support area.

FERMI - UNIT 2 3/4 1-15

b REACTIVITY CONTROL SYSTEMS

- 3/4. I'. 4 CONTROL R0D PROGRAM CONTROLS ROD-WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1-The rod worth minimizer (RWM) shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 m.J 2*, when THERMAL POWER. is less than or equal to 20% of RATED THERMAL POWER, the minimum allowable preset power level.

ACTION:

a.

With the RWM inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or.other. technically qualified member of the unit technical staff who is present at the reactor control console.

Otherwise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the-Shutdown position.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.4.1

.The RWM shall be demonstrated OPERABLE.

a.

In OPERATIONAL. CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RWM automatic initia-

-tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.

b.

lInOPERATIONALCONDITION2within8hourspriortowithdrawalof control rods for the purpose of making the reactor critical, by veri-fying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod (after selection of first control rod).

c.

In OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RWM automatic initiation when reducing THERMAL POWER, by demonstrating the withdraw block and insert block functions.

d.

By demonstrating that the Banked Position Withdrawal sequence input to the RWM computer is correctly loaded following any loading of the program into the computer.

• Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

i FERMI - UNIT 2 3/4 1-16 Amendment No. 42 l

l L_____---_______________-_-_-_-________-_________________-_-___-______--_-_-__A

l4:

.a.

--3/4.2 LPOWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION-3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APiHGRs) shall not exceed:

a.

The MAPLHGR limit which has been approved for the respective fuel and lattice type as a function of the average planar exposure (as determined by the NRC approved methodology described in GESTAR-II),

-or b.

When hand calculations are required, the most limiting lattice type MAPLHGR limit as a function of the average planar exposure shown in the Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, and 3.2.1-4 for the applicable bundle type.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the above limits, initiate corrective action within l

15 minutes and restore APLHGR to within the required limits 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 i- -

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 limits required by Specification 3.2.1:

l 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.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, 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 R0D PATTERN for APLHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

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' POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS-

~ LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-high scram trip setpoint (S) and flow biased neutron flux-high control rod block trip setpoint (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE VALUE S < (0.58W + 59%)T S < (0.58W + 62%)T RB $ (0.58W f S W)T S

5 (0.58W + 53%)T S

RB

- where:

S and S are in percent of RATED THERMAL POWER, RB W = Loop recirculation, flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million 1bs/hr, at 100% of RATED THERMAL POWER T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

T is applied only if less than or equal to 1.0 APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of. RATED THERMAL POWER.

ACTION:

With the APRM flow biased neutron flux-high scram-trip setpoint and/or the

- flow biased neutron flux-high control rod block trip setpoint less conservative than the value shown in the Allowable Value column for 5 or SRB, as above determined, initiate corrective action within 15 minutes and adjust 5 and/or

.S to be consistent with the Trip Setpoint value* within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce RB 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 - 4.2.2 The'FRTP and the MFLPD for each class of fuel shall be determined, the value of T calculated, and the most recent actual APRM flow biased neutron flux-high scram and flow biased neutron flux-high control rod block trip setpoints verified to be within the above limits or adjusted, or the APRM gain readings shall be verified as indicated below,* 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.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, 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 MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

• With MFLPD greater than the FRTP during power ascension up to 90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

FERMI - UNIT 2 3/4 2-5 Amendrent No. 9, 42

r.

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL ~ POWER RATIO r

LIMITING CONDITION FOR OPERATION' 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figures 3.2.3-1 thru 3.2.3-1B times the K shown inLfigure 3.2.3-2, with:

f (Tave I )

B 7.=

T

'I A'

B where:

A = 1.096 seconds, control rod average scram insertion time T

limit to notch 36 per Specification 3.1.3.3,

]b.018, B = 0.813 + 1.65{

1 0

i 9

N-a 3

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nq N tj Iave =

i=1 j

n

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1=1 n = number of surveillance tests performed to date in cycle, th N. = number of active control rods measured in the i I

surveillance test, 9 = average scram time to notch 36 of all rods measured 1

th in the i surveillance test, and N

total number of active rods measured in Specification y = 4.1.3.2.a.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

' ACTION a.

Operating in the Control Cell Core (CCC) operating mode

• and MCPR less than I

the applicable MCPR limit shown in Figures 3.2.3-1 thru 3.2.3-1B

• The CCC operating mode includes operation with only A2 rods, Al shallow rods less than or equal to notch position 36, all peripheral rods inserted in the core, and rods inserted to position 46.

Normal control rod operability checks, coupling checks, scram time testing, and friction testing of non-CCC control rods does not require the utilization of the more restrictive non-CCC I

operational mode MCPR limits.

Any other operation is a non-CCC operating. mode.

FERMI - UNIT 2 3/4 2-6 Amendment No. 9, 39.42 i

___________.________m

.__._.____.__._-_______.__.___..__._a

._____._._,________._________._____m___.m_____.

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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO j

LIMITING CONDITION FOR OPERATION ACTION (Continued).

(Curve A) times the applicable K, curve shown in Figure 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required 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 />.

b.

Operating in the non-CCC operating mode

• and MCPR less than the applicable MCPR limit shown in Figures 3.2.3-1 thru 3.2.3-1B (Curve B)

Ltimes the applicable K, curve shown in Figure 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required 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 />.

Operating in either the CCC or non-CCC operating mode with either the c.

main turbine bypass system inoperable per Specification 3.7.9 or the moisture separator reheater inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within one hour, MCPR is determined to be equal to or greater than the MCPR limit as shown in Figures 3.2.3-1 thru 3.2.3-1B (Curve C) by the main turbine bypass or moisture separator reheater inoperable curve times the applicable K shown in Figure 3.2.3-2.

f d.

Operating in either the CCC or non-CCC operating mode with both the main turbine bypass system inoperable per Specification 3.7.9 and the meisture separator reheater inoperable, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within one hour, MCPR is determined to be equal to or greater than the MCPR limit as shown in Figures 3.2.3-1 thru 3.2.3-1B by the main turbine bypass and moisture separator reheater inoperabl-e curve times the applicable K shown in f

Figure 3.2.3-2.

• The CCC operating mode includes operation with only A2 rods, Al shallow rods less than or equal to notch position 36, all peripheral rods inserted in the core, and rods inserted to position 46.

Normal control rod operability checks, coupling checks, scram time testing, and friction testing of non-CCC control rods does not require the utili2ation of the more restrictive non-CCC operational mode MCPR limits.

Any other operation is a non-CCC operating mode.

i l

FERMI - UNIT 2 3/4 2-6a Amendment No. 42

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 MCPR, with:

l a.

T = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or b.

I as defined in Specification 3.2.3 used to determine the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equa! to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2:

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.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, 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 R0D PATTERN for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 Prior to the use of Curve A and whenever Surveillance Requirement 4.2.3.1 is performed while using Curve A of Figures 3.2.3-1 through 3.2.3-1B, verify that all non-CCC control rods are fully withdrawn from the core.

Non-CCC control rods are all control rods excluding A2 rods, Al shallow rods inserted less than or equal to notch position 36, all peripheral rods, and rods inserted to position 46.

Normal control rod operability checks, coupling checks, scram time testing, and friction testing of non-CCC control rods does not require the utilization of the more restrictive non-CCC operational mode MCPR limits.

FERMI - UNIT 2 3/4 2-7 Amendment No. 42

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CU RVE 3

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CURVE C - MCPR lin3t toe both CCC er mee*CCC operatiesel messa withest either turbine Dypeso or seletare se pareter reh eater.

CU RVE D - W CP R limit for both CCC er mon =CCC operellemal modes without both turbine bypass seg meistere separator reheater et rated feedweler temperature 12,700 MWD /ST TO 13,700 MWD /ST MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS TAU AT RATED FLOW F 8 0 U R E B.t.B-1 A FERMI - UNIT 2 3/4 2-8a Amendment No.42

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0.10.20.30.40.50.60.70.80.91.0 TAU CU R VE A = M CP R 81mit 'er CCf. ope rellemal m ode wit h bet R turbine bypaso ses selettre separator robeater in servlee.

CU RVE S - M CPR limit ter men CCC operetienet mese with bot h turbine typect and meistere separator rekester la service.

CURVE C + MCPR limit ter both CCC ees mem-CCC operettecal meses witbo.t either turbine bypass or seleture s eestator re keeter.

CURV E D

• M CP R limit ter both CCC or men.CCC operettemel m.se. witm e.1 t ot n f o r bi.e by p es o e..

m el.i ro e s p e r te r receeier et reis e to e s weie r te mp o ret.ro 13,70C MWD /ST TO EOC MINIMUM CRITICAL POWER RATIO VERSUS TAU AT RATED FLOW r io u R E s.a.s-it FERMI - UNIT 2 3/4 2-8b Amendment Nc. 42

1.4 1.3 AUTOMATIC FIAT CONTROL 1.2 K

f 1.1

~

MANUALFIAT CONTROL I IC00P TUBE SETPOINT CAURRATION POSITIONED SUCK THAT

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= tofs% -

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= 117A%

I I

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30 40 50 60 70 80 90 100 CORE FLOW (%)

FLOWCORRECT10N(y) FACTOR FIGURE 3.2.3-2 FERMI - Unit 2 3/4 2-9 Amendment Nc. 42

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kw/ft for bundle types BCR183 and BCR233 or 14.4 kw/ft for' bundle types 8D318D and BC318E.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

. equal to 25% of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action

, ithin 15 minutes and restore the LHGR to 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 w

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 4.2.4 LHGR's shall be determined to be equal to or less than the limit:

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.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, 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 on a LIMITING CONTROL ROD PATTERN FOR LHGR.

d.

The provisions of Specification 4.0.4 are not applicable.

FERNI - UNIf 2 3/4 2-10 Amendment No.42

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REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS'(Continued)

The surveillance requirement to measure and record the time that the accumu-lators maintain pressure above the alarm setpoint is intended to provide infor-mation rather than establish OPERABILITY of the accumulators.

No action is s

required if the accumulator pressure does not remain above the alarm setpoint for at least 10 minutes.

Control rod coupling integrity is required to ensure compliance with the analysis of the. rod drop accident in the FSAR.

The overtravel position feature provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after ccmpleting CORE ALTERATIONS that could have affected the control rod coupling integrity.

The subsequent check is performed as a backup to the initial demonstration.

In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod posi-tion indication system must be OPERABLE.

The control rod housing support restricts the outward movement of a control

-rod to less than 3 inches in the event of a housing failure.

The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.

The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.

The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.

3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.

The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.

When THERMAL POWER is greater than 20 percent of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20 percent of RATED THERMAL POWER provides adequate control.

The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

The analysis of the rod drop accident is presented in Section 15.4.9 of l

the UFSAR and the techniques of the analysis are presented in a topical report, Reference 1.

l The RBM is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation.

Two channels are provided.

Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage.

This system backs up the written sequence used by the operator for withdrawal of control rods.

FERMI - UNIT 2 B 3/4 1-3 Amendment No. 42

i l

I REACTIVITY CONTROL SYSTEMS I

BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The design objective of the Standby Liquid Control (SLC) System is two

fold, One objective is to provide backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.

The second objective of the SLC System is to meet the requiremsat of the ATWS Rule, specifically 10 CFR 50.62 paragraph (c)(4) which states that, in part:

"Each boiling water reactor must have standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution."

The SLC System uses enriched Boron-10 (contained in the sodium pentaborate solution) to comply with 10 CFR 50.62 paragraph (c)(4).

The methods used to determine compliance with ATWS Rule are in accordance with Reference 4.

To meet both objectives, it is necessary to inject a minimum quantity of 2350 net gallons of 65 atom percent Boron-10 enriched sodium pentaborate in a solution having a concentration of no less than 9.0 weight percent (see Figure 3.1.5-1 for equivalent volumes and concentration ranges).

The equivalent concentration of natural boron required to shutdown the reactor is 660 parts per million (ppm) in the 70 F moderator, including the Recirculation loops and with the RHR Shutdown Cooling Subsystems in operation.

In addition to this, a 25 percent margin is provided to allow for leakage and imperfect mixing (825 ppm).

The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible sodium pentaborate solution volume range, which adequately compensates for the positive reactivity effects due to moderator temperature reduction and xenon decay during shutdown.

The temperature requirement is necessary to ensure that the sodium pentaborate remains in solution.

With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperator or for longer periods of time with one of the redundant components inoperable.

The SLC tank heaters are only required when mixing sodium pentaborate and/or water to establish the required solution operating parameters during additions to the SLC tank.

Normal operation of the SLCS does not depend on these tank heaters to maintain the solution above its saturation temperature.

Technical require-ments have been placed on the tank heater circuit breakers to ensure that their failure will not degrade other SLC components (see Specification 3/4.8.4.5).

Surveillance requirements are established on a frequency that assures a high reliability of the system.

Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.

Analysis of Boron-10 enrichment each 18 months provides sufficient assurance that the minimum enrichment of Boron-10 will be maintained.

I FERMI - UNIT 2 B 3/4 1-4 Amendment No. N, 42

1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM (Continued)

Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.

i 1.

" General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A and U.S. (latest approved revision) 2.

General Electric Co., Licensing Topical Report " Anticipated Transient Without Scram; Response to the NRC ATWS Rule, 10 CFR 50.62" NEDE-31096-P-A dated February 1987.

FERMI - UNIT 2 B 3/4 1-5 Amendment No. 3$,42

.a

6 3

3/4.2 POWER DISTRIBUTION LIMITS BASES-The specifications of this section assure that the peak cladding tempera-ture following the postulated design basis loss of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46, 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 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 i

on the rod-to-rod power distribution within an assembly.

The peak clad tempera-ture is calculated assuming a 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

' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (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 Technical Specification MAPLHGR value is the most limiting composite of the fuel mechanical design analysis MAPLHGR and the ECCS MAPLHGR.

Fuel Mechanical Design Analysis:

NRC approved methods (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice, operating at the bounding power history, meet the fuel design limits specified in Reference 1.

This bounding power history is used as the basis for the fuel design analysis MAPLHGR value.

LOLA Analysis: A LOCA analysis is' performed in accordance with 10 CFR 50 Appendix K to demonstrate that the MAPLHGR values comply with the ECCS limits specified in 10 CFR 50.46.

The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

Only the most limiting MAPLHGR values are shown in the Technical Specification figures for multiple lattice fuel.

When hand calculations are required, these Technical Specification MAPLHGR figure values for that fuel type are used for all lattices in that bundle.

For Some fuel bundle designs MAPLHGR depends only on bundle type and burnup.

Other fuel bundles have MAPLHGRs that vary axially depending upon the specific combination of enriched uranium and gadolinia that comprises a fuel bundle cross section at a particular axial node.

Each particular combination of enriched uranium and gadolinia, for these fuel bundle types, is called a lattice type.

These particular fuel bundle types have MAPLHGRs that vary by lattice (axially) as well as with fuel burnup.

Reference 1.

" General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (latest approved revision).

FERMI - UNIT 2 B 3/4 2-1 Amendment No.42

u POWER DISTRIBUTION LIMITS BASES l

3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specifications 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRM instru-ments must be adjusted to ensure that the MCPR does not become less than the Safety Limit MCPR or that > 1% plastic strain does not occur in the degraded situation.

The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THER-MAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transieat would not be increased in the degraded condition.

FERMI - UNIT 2 B 3/4 2-2 Amendment No. 42

w-P r.

BASES TABLE B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS E-Plant Parameters:

Core THERMAL P0WER...............

3430 MWt* which corresponds to i

105% of rated steam flow n

' Vessel: Steam Output................~14.86 x 106 lbm/h'r which corresponds to 105% of rated steam flow Vessel' Steam Dome Pressure.......

1055 psia Design Basis Recirculation Line Break Area for:

a.

Large Breaks 4.1 ft2 b.

Small Breaks 0.1 ft I

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GE0 METRY (kW/ft)

FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18

-First Reload 8x8 14.4 1.4 1.18 A more detailed listing of input of each model and its source is presented in Section II of Reference I and subsection 6.3 of the FSAR.

• This power level meets the Appendix K requirement of 102%.

The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

1 FERMI - UNIT 2 B 3/4 2-3 Amendment No."42

a

• POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limiting MCPRs at steady-state operating conuitions as specified in Specification 3.2.3 are derived from the established fuel clad-ding integrity Safety Limit MCPR, and an analysis of abnormal operational tran-l-

sients.

For any abnormal operating transients 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 given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients 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 tem-perature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limiting MCPR of Specification 3.2.3 is obtained and presented in Figures 3.2.3-1, 3.2.3-1A, and 3.2.3-18.

The MCPR curves illustrated in Figures 3.2.3-1 thru 3.2.3-1B were derived as described above for the following assumed operating conditions:

Curve A - MCPR limit with turbine bypass system, moisture separator reheater systems in service and CCC (Control Cell Core) operational mode (A2 rods, Al shallows inserted less than or equal to notch posi-tion 36, all peripheral rods, and all rods inserted to position

46) inserted in the core.

The operating domain includes the 100%

po.ver/ flow region and extended load line region with 100%

power and reduced flow.

Curve B - MCPR limit with the turbine bypass system, moisture separator reheater systems in service and non-CCC operational mode (any control rod inserted in the core).

The operating domain includes the 100% power / flow region and the extended load line region with 100% power and reduced flow.

Curve C - MCPR limit for either CCC or non-CCC operational modes with either the main turbine bypass system inoperative and the mois-ture separator reheator system available or the main turbine bypass system available and the moisture separator reheater sys--

tem inoperable.

The operating domain includes the 100%

power / flow region and the extended load line region with 100%

j.

power with reduced flow.

Curve D - MCPR limit for either CCC or non-CCC operational modes with the main turbine bypass system inoperative and the moisture separator reheater system inoperable.

The operating domain includes the FERMI - UNIT 2 B 3/4 2-4 Amendment No. 39, 42

4 e

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO (Continued) 100% power / flow region and the extended load line region with 100% power and reduced flow.

Curve A provides the MCPR limit assuming operation above 25 percent RATED THERMAL POWER with the turbine bypass system and moisture separator reheater in service.

The curve was developed based upon the operating MCPR limits for a rod withdrawal error transient (UFSAR, Section 15.4.2) for operating within the CCC control rod patterns and a Main Turbine Trip with Turbine Bypass Failure transient (UFSAR, Section 15.2.3).

CCC control rods are A2 rods, Al shallow rods (inserted less than or equal to notch position 36), all peripheral rods, and all rods inserted to position 46.

The analysis of the Main Turbine Trip with Turbine Bypass Failure takes credit for the steam flow to the moisture separator reheater.

Curve B provides the MCPR limit assuming operation above the 25 percent RATED THERMAL POWER with the turbine bypass system and moisture separator reheater system in service and non-CCC control rods inserted in the core.

Non-CCC control rods are all rods excluding A2 rods, Al shallow rods (inserted less than or equal to notch position 36), all peripheral rods, and all rods inserted to position 46.

The curve was developed based upon the operating MCPR limits for a rod withdrawal error transient (UFSAR, Section 15.4.2) for any operating withdrawal sequence.

Curve C provides the MCPR limit assuming operation above the 25 percent RATED THERMAL POWER with the moisture separator reheater operable and turbine bypass system inoperable or the moisture separator reheater inoperable and the turbine bypass system operatie.

The curve was developed based upon the operat-ing MCPR limits for several combinations of Feedwater Controller Failure.

Operation with main turbine bypass inoperable or with a moisture separator reheater inoperable results in a total reactor steam flow bypass capability of approximately 10 percent and 26 percent, respectively.

The impact of operation with the moisture separator reheater inoperable but with bypass operable and utilization of Curve B is conservative because the 26 percent bypass capability is less limiting ith regard to the existing analysis used to establish Curve B which assumes only 10 percent bypass capability (with the main turbine bypass system inoperable).

Therefore, the operation above 25 percent RATED THERMAL POWER with either the moisture separator reheater inoperable or main turbine bypass system inoperable is bounded by the existing Curve B.

Curve D provides the MCPR limit assuming operation above the 25 percent RATED THERMAL POWER with both the moisture separator reheater inoperable and the turbine bypass system inoperable.

The curve was developed based upon the operating MCPR limits from the Feedwater Controller Failure.

There is no mode change restraint should the main turbine bypass or the moisture separator reheater be inoperable.

However, should the main turbine FERMI - UNIT 2 B 3/4 2-4a Amendment No. 39, 42

r ~

2 i.

e 1.....

POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO (Contir.ued)

I bypass system or the moisture separator rrtheater be inoperable as 25 percent RATED THERMAL POWER is exceeded, the MCPR check must be completed within one hour.

The evaluation of a given transient begins with the system initial parameters shown in UFSAR Table 15.0.1 that are input to a GE-core dynamic behavior transient computer program..The codes used to evaluate transients are described in GESTAR-II.

The principal result of this evaluation is the reduction in MCPR catsed by the transient.

The purpose of the K factor of Figure 3.2.3-2 is to define operating y

limits at other than rated core flow conditions.

At less then 100% of rated flow the required MCPR is the product of the MCPR and the K factor.

The 7

K factors assure that the Safety Limit MCPR will.not be violated during a flow ibcreasetransientresultingfromamotorgeneratorspeedcontrolfailure.

The K factors may be applied to both manual and automatic flow control modes.

f The K factor values shown in Figure 3.2.3-2 were developed generically f

and are applicable to all BWR/2, BWR/3, and BWR/4 reactors.

The K, factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow, although they are applicable for the extending operating region.

For the manual flow control mode, the K factors were calculated such that f

for the maximum flow rate, as limited by the pump scoop tube setpoint and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows.

The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the K.

f FERMI - UNIT 2 B 3/4 2-4b Amendment No. #, 42

)

1 POWER DISTRIBUTION LIMITS I

i BASES i

j 3/4.2.3 MINIMUM CRITICAL POWER RATIO (Continued)

J For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow.

The K factors shown in Figure 3.2.3-2 are conservative for the General f

Electric plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of K.

f At THERMAL POWER levels less than or equal to 25 percent of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod pat-terns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

During initial startup testing of the plant, a MCPR evaluation will be made at 25 percent of RATED THERMAL POWER level with minimum recirculation pump speed.

The MCPR margin will thus be demonstrated such that future MCPR evalua-tion below this power level will be shown to be unnecessary.

The daily require-ment for calculating MCPR when THERMAL POWER is greater than or equal to 25 per-cent of RATED THERMAL POWER is sufficient since power distribution shif ts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pattern is ap-proached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3.4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

1.

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

2.

" General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A, latest approved revision.

FERMI - UNIT 2 B 3/4 2-5 Amendment No. 42

r.

L:

l DESIGN FEATURES 5.3 REACTOR CORE l

FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 764 fuel assemblies which are limited to l

those fuel assemblies that have been analyzed with NRC approved codes and methods I

and have been shown to comply with all of the criteria in the latest approved version of GESTAR-II.

CONTROL R0D ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing 143 inches of bar:n carbide, 8 C, powder surrounded by a cruciform shaped stainless steel sheath.

4 5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

.1250 psig on the suction side of the recirculation pump.

2.

1500 psig from the recirculation pump disenarge to the outlet side of the discharge shutoff valve.

3.

1500 psig from the discharge shutoff valve to the jet pumps:

c.

For a temperature of 575*F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,034 cubic feet at a nominal steam dome saturation temperature of 540'F.

FERMI - UNIT 2 5-5 Amendment No.42

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DESIGN FEATURES 5.5' METEOROLOGICAL TOWER LOCATION 5.5.1 The 60 meter meteorological tower shall be located as shown on Figure 5.1.1-1.

5.6 FUEL STORAGE j

CRITICALITY 5.6.1 The spent fuel storage racks are designed and shall be maintained with:

A k,7f equivalent to less than or equal to 0.95 when flooded with a.

unborated water.which includes a conservative allowance of 1.9%

delta k/k for uncertainties as described in Section 9.1 of the FSAR.~

b.

A nominal 6.22 and 11.9 x 6.6 inch center-to-center distance between fuel assemblies placed in the high density and low density storage racks, respectively.

5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 660' 11\\".

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2414 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7.1-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7.1-1.

FERMI - UNIT 2 5-6

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