ML17278A846

Amend 28 to License NPF-21,revising Tech Specs to Support Operation of Facility at Full Rated Power During Cycle 2 & Establishing Operating Limits for All Fuel Types

ML17278A846
Person / Time
Site:	Columbia
Issue date:	05/23/1986
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML17278A847	List: ML17278A846 ML17278A848
References
TAC-60804, NUDOCS 8605300299
Download: ML17278A846 (77)

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET NO. 50-397 WPPSS NUCLEAR PROJECT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No.

NPF-21 1.

The'uclear Regulatory Commission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the Washington Public Power Supply System (the Supply System, also the licensee),

dated Februarv 26,

1986, as supplemented on April 7, 24, 25, 30 and May 22, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 regulations of the Commission; C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering tho health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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, the license is amended by changes to the Technical Specifica-tions as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.

NPF-21 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

28, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

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

Sb05300299'b0523 PDR ADOCK '05000397 P

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This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Enclosure:

Chanqes to the Technical Specifications Elinor G. Adensam, Director BWR Project Directorate No.

3 Division of BWR Licensing Date of Issuance:

May 23, 1986

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ENCLOSURE TO LICENSE AMENDMENT NO. 28 FACILITY OPERATIN6'ICENSE NO. NPF-21 DOCKET NO. 50-397 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 1llill xx XX1 XX1V 1-1 1-2 1-3 1-4 1-5 1-6 1-7 1-8

'8 2-1' 8 2-2 8 2-3 8 2-4 3/4 1-2 3/4 1-8 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4

'3/4 2-6 3/4 2-7 3/4 2-8 3/4 3-41 8 3/4 l-l 8 3/4 1-2 8 3/4 1-3 8 3/4 1-4 8 3/4 2-1 8 3/4 2-2 8 3/4 2-3 8 3/4 2-4 8 3/4 2-5 1ll 111 XX xx(a)

XX1 XX1V 1-1 1-2 1-3 1-4 1-5 1-6

'1-7 1-8 8 2-1 8 2-2 8 2-3 3/4 1-2 3/4 1-8 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-4A 3/4 2-48 3/4 2-4C 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 3-41 8 3/4 1-1 8 3/4 1-2 8 3/4 1-3 8 3/4 1-4 8 2/4 2-1 8 3/4 2-2 8 3/4 2-3 8 3/4 2-4

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INDEX DEFINITIONS SECTION

1. 0 DEFINITIONS 1. 1 ACTION...........

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1.2 AVERAGE BUNDLE EXPOSURE..

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l. 3 AVERAGE PLANAR EXPOSURE....................

1-1 1.4 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1: 5 CHANNEL CALIBRATION.

1. 6 CHANNEL CHECK............

1.7 CHANNEL FUNCTIONAL TEST....

1. 8 CORE ALTERATION....

1.9'RITICAL POWER RATIO.

1.10 DOSE E(UIVALENT I-131...................

1-2 1-2 1"2 1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY.........

1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME..

1-3 II 1-2 1;12 EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE TIME.........

1-2 1.14 FRACTION OF LIMITING POWER DENSITY.

1. 15 FRACTION OF RATED THERMAL POWER.

1. 16 FRE(UENCY NOTATION.

1. 17 GASEOUS RADWASTE TREATMENT SYSTEM.

l. 18 IDENTIFIED LEAKAGE.....

1.19 ISOLATION SYSTEM RESPONSE TIME 1.20 LIMITING CONTROL ROD PATTERN.............;

1-3 1-3 1-3 1" 3 1" 3 1-3 1-4 1.21 LINEAR HEAT GENERATION RATE......................

1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST 1.23 MAXIMUM FRACTION OF LIMITING POWER DENSITY.

1-4 1-4 WASHINGTON NUCLEAR - UNIT 2 Amendment No. 28

INDEX DEFINITIONS SECTION DEFINITIONS (Continued) 1.24 MAXIMUM TOTAL PEAKING FACTOR.

1. 25 MEMBER(S)

OF THE PUBLIC

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PAGE 1"4 1.26 MINIMUM CRITICAL POWER RATIO............

1.27 OFFSITE DOSE CALCULATION MANUAL........

1.28 OPERABLE - OPERABILITY.

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1-5 1.29 OPERATIONAL CONDITION - CONDITION.........,.............,..

1-5

l. 30 PHYSICS TESTS....

1.31 PRESSURE BOUNDARY LEAKAGE

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1. 32 PRIMARY CONTAINMENT INTEGRITY........'......................

1-5 1.33 PROCESS CONTROL PROGRAM.

. 1.34 PURGE " PURGING........

1.35 RATED THERMAL POWER.

1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME

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1"6 1"6 1-6 1.37 REPORTABLE EVENT 1.38 ROD DENSITY...

1.39 SECONDARY CONTAINMENT INTEGRITY.

1.40 SHUTDOWN MARGIN..

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1-6 1-6 1-6 1-7

1. 41 SITE BOUNDARY.........................

1.42 SOLIDIFICATION.

1.43 SOURCE. CHECK.

1.44 STAGGERED TEST BASIS

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1-7 1-7 1"7 1.45 THERMAL POWER...

1.46 TOTAL PEAKING FACTOR......

WASHINGTON NUCLEAR - UNIT 2

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Amendment No. 28

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INDEX DEFINITIONS SECTION DEFINITIONS (Conti nued) 1.47 TURBINE BYPASS SYSTEM RESPONSE TIME.

1.48 UNIDENTIFIED LEAKAGE....

1.49 UNRESTRICTED AREA...........

1.50 VENTILATION EXHAUST TREATMENT SYSTEM.

1. 51 VENTING.

PAGE 1-8

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1"8 1-8 1-8 WASHINGTON NUCLEAR - UNIT 2 Amendment No. 28

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LIST OF FIGURES INDEX FIGURE

3. 1. 5-1 PAGE SODIUM PENTABORATE SOLUTION SATURATION TEMPERATURE...

~34 1-21

3. 1. 5-2

3. 2. 1"1 3.2.1 2

3. 2. 1-3 SODIUM PENTABORATE TANK, VOLUME VERSUS CONCENTRATION REQUIREMENTS...........

MAXIMUM AVERAGE PLANAR LINEAR HfAT. GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIALCORE FUEL TYPE 8CR183........................

MAXIMUM AVERAGE PLANAR LINEAR HfAT GENERATION RATE (MAPLHGR) VERSUS AVfRAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR233.

MAXIMUMAVfRAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ENC XN-1.FUEL 3/4 1-22 3/4 2-2 3/4 2"3 3/4 2-4 3.2. 1-4

3. 2. 1-5

3. 2. 1-6 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183.

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

3/4 2-4B MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS BUNDLE AVERAGE EXPOSURE ENC XN-1 FUEL................,.

.'/4 2-4C

3. 2. 3"1

3. 2. 4-1

3. 3. 10-1 3.4. 1. 1-1

3. 4. 6. 1-1 REDUCED FLOW MCPR OPERATING LIMIT LINEAR HEAT GENERATION'ATE (LHGR) LIMIT VERSUS AVERAGf PLANAR EXPOSURE EXXON Bx8 FUEL THERMAL POWER LIMITS OF SPEC.

3. 3. 10-1 THERMAL POWER LIMITS OF SPEC. 3.4.1.1-1............-..

MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (INITIALVALUES)......

3/4 2-7 3/4 2-10 3/4 3-104 3/4 4-3a 3/4 4"20 3.4. 6. 1-2

4. 7"1

3. 9. 7-1 B 3/4 3-1 WEIGHT/HEIGHT LIMITATIONS FOR LOADS OVER THE SPENT FUEL STORAGE POOL REACTOR VESSEL WATER LEVEL 3/4 9"10 B 3/4 3-8 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR'VESSEL PRESSURf (OPERATIONAL VALUES)......... 3/4 4-21 SAMPLE PLAN 2)

FOR SNUBBER FUNCTIONAL TEST..........

3/4 7-15 WASHINGTON NUCLEAR - UNIT 2 XX Amendment No.

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LIST OF FIGURES INDEX PAG IGURE

~B3 4. 4. 6-1.

FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE..............

B 3/4 4-7 EXCLUSION AREA BOUNDARY....................,.........

5-2

5. 1-1

5. 1"2

5. 1"3 LOW POPULATION ZONE..

5-3 UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LI(UID EFFLUENTS.............

5"4

6. 2. 1-1

6. 2. 2-la

6. 2. 2-lb UNIT ORGANIZATION....................................

UNIT ORGANIZATION - OPERATIONS DEPARTMENT............

6-5 WASHINGTON NUCLEAR - UNIT 2 Amendment No. 29 xx(a)

OFFSITE ORGANIZATION 6-3

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

1-9 1.2 OPERATIONAL CONDITIONS 1-10 2.2. 1-1 B2. 1. 2-1

3. 2. 3-1

3. 3. 1-1

3. 3 ~ 1"2

4. 3. 1. 1-1

3. 3. 2-1

3. 3. 2-2

3. 3. 2-3

4. 3. 2. 1"1 3.3.3 1

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS..

2"4 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT B 2"3

'CPR OPERATING LIMITS FOR RATED CORE FLOW............

3/4 2-7 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

3/4 3"2 REACTOR PROTECTION SYSTEM RESPONSE TIMES.............

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

ISOLATION ACTUATION INSTRUMENTATION...

3/4 3-7 3/4 3-12 ISOLATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..

0 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.

3/4 3-22 3/4 3-26 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ~.......

3/4 3-16 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME.......

3/4 3-19

3. 3. 3-2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 3/4 3-30

3. 3. 3"3

4. 3. 3. 1-1 'MERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............

3/4 3-34 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES.........

3/4 3-33

3. 3. 4. 1-1

3. 3. 4. 1-2

4. 3.4. 1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION.

ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............

3/4 3"38 3/4 3-39 3/4 3-40 WASHINGTON NUCLEAR - UNIT 2 XX1 Amendment No.

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INDEX LIST OF TABLES (Continued)

TABLE 4.4. 6. l. 3-1

3. 6. 3-1

3. 6. 5. 2-1

3. 7. 6. 4-1

3. 7. 6. 5-1

3. 7. 8-1

4. 8. l.l. 2-1 4.8. 2.1-1 3: 8. 4. 2-1

3. 8.4. 3-1 4.11-1

4. 11-2

3. 12-1

3. 12-2

4. 12-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM--

WITHDRAWAL SCHEDULE..

PRIMARY CONTAINMENT ISOLATION VALVES SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION VALVES FIRE HOSE STATIONS YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES AREA TEMPERATURE MONITORING...

DIESEL GENERATOR TEST SCHEDULE BATTERY SURVEILLANCE REQUIREMENTS................

PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION RADIOACTIVE LI(UID WASTE SAMPLING AND ANALYSIS P ROGRAM

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RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS P ROGRAM

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RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM.;..

REPORTING LEVELS FOR RADIOACTIVITYCONCENTRATIONS IN ENVIRONMENTAL SAMPLES.....................,.

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DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS PAGE 3/4 4-22 3/4 6-21 3/4 6"39 3/4 7-25 3/4 7-27 3/4 7-31 3/4 8-9 3/4 8-14 3/4 8-23 3/4 8-26 3/4 11-2 3/4 11-9 3/4 12-3 3/4 12"9 3/4 12-10 WASHINGTON NUCLEAR - UNIT 2 XX1 V Amendment No.

28

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1. 0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifi-cations may be achieved.

The defined terms appear in capitalized type 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 BUNOLE EXPOSURE 1.2 The AVERAGE BUNOLE EXPOSURE is equal to the sum of the axially averaged exposure of all the fuel rods in the specified bundle divided by the number of fuel rods in the bundle.

AVERAGE PLANAR EXPOSURE 1.3 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 GENERATION RATE

.= 1.4 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APIHGR) 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 1.5 A CHANNEL CALIBRATION 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.

CHANNEL CHECK 1.6 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 channels measuring the same parameter.

WASHINGTON NUCLEAR - UNIT 2 Amendment No. 2B

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DEF INITIGNS CHANNEL FUNCTIONAL TEST 1.7 A CHANNEL FUNCTIONAL TEST shall be:

a.

Analog channels - the injection of a simulated signal into the channel

.as cl'ose 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.

CORE ALTERATION 1.8 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Sus-

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

CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be that power in the assembly which is calculated by application of the XN-3 correlation to cause some point in the assembly to experience boiling transition divided by the actual assembly operating power.

DOSE E UIVALENT I-131 1.10 DOSE.EQUIVALENT I-131 shall be that concentration of I-131, m'icrocuries 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."

K-AVERAGE DISINTEGRATION ENERGY

l. 11 E shall be the average, weighted in proportion to the concentration of each radionuclide 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 95K of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM ECCS)

RESPONSE

TIME

1. 12 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.

WASHINGTON NUCLEAR - UNIT 2 1-2 Amendment No.28

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DEFINITIONS END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM

RESPONSE

TIME shall be that time'nterval to energization of the recirculation pump circuit breaker trip coil from when the monitored parameter exceeds its trip setpoint at the channel sensor of the associated:

a.

Turbine throttle valves channel sensor contact opening, and b.

Turbine governor valves initiation'f valve fast closure.

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 1.14 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. 15 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

FEEIEOENOY NOTATEOII 1.16 The FRE(UENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM

1. 17 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivi ty prior to release to the environment.

IDENTIFIED LEAKAGE 1.18 IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the opera-tion of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

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.

WASHINGTON NUCLEAR - UNIT 2 1-3 Amendment No.

28

DEF INITIGNS LIMITING CONTROL ROD PATTERN 1.20 A LIMITING CONTROL R00 PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR,,LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod, It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.23 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be highest value of the FLPD which exists in the core.

MAXIMUM TOTAL PEAKING FACTOR 1.24 The MAXIMUM TOTAL PEAKING FACTOR (MTPF) shall be the largest TPF which exists in the core for a given class of fuel for a given operating condition.

MEMBER(S OF THE PUBLIC 1.25 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of

. the site for recreational, occupational or other purposes not.associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the current met'hodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints and in the conduct of the environmental radiological monitoring program.

WASHINGTON NUCLEAR - UNIT 2 Amendment No. 28

DEFINITIONS OPERABLE - OPERABILITY 1.28 A system, subsystem,

train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem,

train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL CONDITION - CONDITION 1.29 An OPERATIONAL CONDITION, i.e.,

CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 1.30 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation as (1) described in Chapter 14 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.31 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable,.fault in a reactor coolant system component body, pipe wall, or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.32 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a.

All primary containment penetrations required to be closed during acci dent conditions are either:

b.

C.

d.

e.

1.

Capable of being closed by an OPERABLE primary containment automatic isolation system, or 2.

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

All primary containment equipment hatches are closed and sealed.

Each primary containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

The primary containment leakage rates are within the limits of Specification 3.6. 1.2.

The suppression chamber is in compliance with the requirements of Specification 3.6.2. 1.

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

WASHINGTON NUCLEAR - UNIT 2 1-5 Amendment No. 2B

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DEF INITIONS PROCESS CONTROL PROGRAM l.33 The PROCESS CONTROL PROGRAM (PCP) shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

PURGE -

PURGING 1.34 PURGE or PURGING shall be 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 1.35 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3323 MWt.

REACTOR PROTECTION SYSTEM RESPONSE TIME

l. 36 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until deener gization of the scram pilot valve solenoids.

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

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Secti'on 50.73 to 10 CFR Part 50.

ROD DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a

fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 100K ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.39 SECONDARY CONTAINMENT INTEGRITY shall exist when:

a 0 b.

C.

All secondary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.

All secondary containment hatches and blowout panels are closed and sealed.

The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

WASHINGTON 'NUCLEAR " UNIT 2 Amendment No.28

l II l

DEFINITIONS SECONDARY CONTAINMENT INTEGRITY (Continued) d.

At least one door in each access to the secondary containment is closed.

e.

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

The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5. l.a.

SHUTDOWN MARGIN 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e., 68'F; and xenon free.

SITE BOUNDARY 1.41 The SITE BOUNDARY shall be that line beyond which the land is not owned,

leased, or otherwise.controlled by the licensee.

SOLIDIFICATION 1.42 SOLIDIFICATION shall be the conversion of, radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

SOURCE CHECK 1.43 A SOURCE CHECK shall be the qualitative assessment of channel response when the'channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.44 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.

b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

WASHINGTON NUCLEAR " UNIT 2 1"7 Amendment No.28

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DEFINITIONS THERMAL POWER 1.45 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TOTAL PEAKING FACTOR 1.46 The TOTAL PEAKING FACTOR (TPF) shall be the.ratio of local LHGR for any specific location on a fuel rod divided by the core average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.47 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

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

UNIDENTIFIED LEAKAGE 1.48 UNIDENTIFIED LEAKAGE shall be all.leakage which.is not IDENTIFIED LEAKAGE.

UNRESTRICTED AREA 1.49 An UNRESTRICTED AREA'hall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

0 VENTILATION EXHAUST TREATMENT SYSTEM 1.50 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Features (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.51 VENTING shall be 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 not provided or required during VENTING.

Vent, used in system

. names, does not imply a VENTING process.

WASHINGTON NUCLEAR - UNIT 2 1"8 Amendment No.

28

2.0 SAFETY LIMITS and LIMITING SAFETY SYSTEM SETTINGS BASES INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the envi rons.

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 for two recirculation loop operation and 1. 07 for single recircula-tion loop operation for both GE and ENC fuel.

MCPR greater than 1.06 for two recirculation loop operation and 1.07 for single recirculation loop operation represents a conservative margin relative to the conditions required to main-tain 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 sig-nal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

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

These conditions represent a significant departure from the condition intended by design for planned operation.

The MCPR fuel cladding integrity safety limit assures that during normal operation and during anticipated operational occurrences, at least 99.9 percent of the fuel rods in the core do not experience, transition boiling (Reference XN-NF-524 (A), Rev. 1).

2. 1 SAFETY LIMITS

2. l. 1.

THERMAL POWER Low Pressure or Low Flow For certain conditions of pressure and flow, the NN-3 correlation is not valid for all critical power calculations.

The XN-3 correlation is not valid for bundle mass velocities less than

. 25 x 10 lbs/hr-ft or pressures less than 585 psig. Therefore, the fuel cladding integrity Safety Limit is estab-lished 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 10'bs/h (approximately a mass velocity of.25 x 10'bs/hr-ft~),

bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10~ lbs/h.

Full scale ATLAS test data taken at pres-sures from 14.7 psia to 800 psia indicate that the fuel assembly critical power WASHINGTON NUCLEAR - UNIT 2 B 2-1 Amendment No. 28

SAFETY LIMITS BASES THERMAL POWER Low Pressure or Low Flow (Continued) at this flow is approximately 3.35 MWt.

With the design peaking factors, this corresponds to a THERMAL POWER of more than 50K of RATEO THERMAL POWER.

Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure below 585 psig is conservative.

2. l. 2 THERMAL POWER Hi h Pressure and Hi h Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage ar'e not directly observable during reactor opera-tion, 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 i'n 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'alue 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. 9X 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 the Exxon Nuclear Critical Power Methodology for boiling water. reactors which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.

The probability of the occur rence of boil-ing transition is determined using the Exxon nuclear critical heat flux-enthalpy XN-3 correlation.

The XN-3 correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.

The required input to the statistical model are the 'uncertainties listed in Bases Table B2. 1. 2-1 and the nominal values of the core parameters listed in Bases Table B2. 1. 2-2.

The bases for the uncertainties in the core parameters are given in XN-NF-524(A), Rev.

1 and the basis for the uncertainty in the XN-3 corr ela-tion is given in XN-FN-512(A), Rev.

1 The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen tb produce a skewed power distribution 'having the greatest number of assemblies at the highest power levels.

The worst distribution during any fuel cycle would not be as severe as the-distribution used in the analysis.

a.

Exxon Nuclear Critical Power Methodology for Boiling Water Reactors, XN-NF-524(A), Rev.

1.

b.

Exxon Nuclear Company XN-3 Critical Power Correlation, XN-NF-512(a),

Rev. l.

WASHINGTON NUCLEAR - UNIT 2 2"2 Amendment No. 28

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BASES TABLE B2.1.2"1 UNCERTAINTIES CONSIDERED IN THE MCPR SAFETY LIMIT Parameter Feedwater Flow Rate Feedwater Temperature Core Pressure Total Core Flow Rate Core Inlet Enthalpy XN-3 Critical Power Correlation Assembly Flow Rate Power Distribution:

Radial Peaking Factor Local Peaking Factor STANDARD DEVIATION"

. 0176

.0076

.0050

.0250

.0024

.0411

.0280

.0528

.0246

" Fraction of Nominal Value.

MASHINGTON NUCLEAR - UNIT 2 B 2-3 Amendment No. 28

l f

REACTIVITY CONTROL SYSTEMS 3/4. 1. 2 REACTIVITY ANOMALIES LIMITING CONDITION FOR OPERATION

3. 1. 2 The reactivity difference between the monitored core k ff and the predicted core k ff shall not exceed 1X delta k/k.

eff APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the reactivity equivalence differ ence exceeding 1X delta k/k:

a.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> perform an analysis to determine and explain the cause of the reactivity differ ence; operation may continue if the difference is explained and corrected.

b.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.1.2 The reactivity difference between the monitored core k ff and the pre-dicted core k ff shall be verified to be less than or equal to 1X delta k/k:

eff a.

During the first startup following CORE ALTERATIONS, and b.

At least once per 31 effective full power days.

WASHINGTON NUCLEAR - UNIT 2 3/4 1-2 Amendment No. 28

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REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, for the four control rods arranged in a two-by-two array, based on deenergiza-tion of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Full Withdrawn 45 39 25 5

Average Scram Inser-tion Time (Seconds) 0.455 0.920

2. 052

3. 706 APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

'h a ~

With the average scram insertion times of control rods exceeding the above limits:

1.

Declare the'ontrol rods with the slower than average scram insertion times inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and 2.

Perform the Surveillance Requirements of Specification 4. 1.3.2.c at least once per 60 days when operation is continued with an average scram insertion time(s) in excess of the average scram insertion time limit.

h Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

The provisions of Specification

3. 0. 4 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement

4. l. 3. 2.

WASHINGTON NUCLEAR - UNIT 2 3/4 1"8 Amendment No.

28

3/4.2 POWER 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 (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE for GE fuel and average bundle exposure for ENC fuel shall not exceed the limits shown in Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-3.

The limits for single loop operation are shown in Figures 3.2. 1-4, 3.2. 1-5, and 3.2. 1-6.

APPLICABILITY:

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

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, or 3.2.1-3, initiate corrective action within 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 25K 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 RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures

3. 2. 1-1,

3. 2. 1-2, and 3. 2. 1-3:

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'OWER increase of at least 15K 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 ROD PATTERN for APLHGR.

WASHINGTON NUCLEAR - UNIT 2 3/4 2-1 Amendment No.

28

12.5 12.7 12.8 12.9 12.7

~~12.0 11.5 wc 11.0 630 y 4J0 10.5 E e D (9 10.0 m~

9 5 C

9.0 8.0 12.1 12.1 12.7 12.8 12.9 12.7 11.7 10.8 10.0 9.4 1,102 5,512 11,023 16,535 22,046 27,558 33,069 38,581 44,093 Ordinate Abscissa 10.8 10.0 9.4 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 Average Planar Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Type 8CR183 Figure 3.2.1-1 860599.4A

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12.0 1,102 12.1 5,512 12.2 11,023 12.2 16,535 12.1 22,046 11.6 27,558 11.2 33,069 10.6 38,581 10.0 44,093 Ordinate Abscissa 12.2 12.2 11.6 11.2 10.6 10.0 8.0 5,000 10,000 15,000 20,000 25,00 0 30 000 35 000 40 000 Average Planar Exposure (MWD/MT) 0 CO Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Type SCR233 Figure 3.2.1-2 860599.5A

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5,000 10,000 15,000 20,000 25,000 30,000

'35,000 40,000 Bundle Average Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Ex'posure ENC XN-1 Fuel Figure 3.2.1-3 860599.2A

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'9.07 e.a 0 7.90 6.5 0

5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Average Planar Exposure (MWD/MT)

SINGLE LOOP Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Type 8CR183 Figure 3.2.1-4 O

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AI SQ tO 10.5 10.0-9.5 6.0 e

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10.08 10.16 10.08 10.16 10.25 10.25 10.16 9.74 9.41 8.90 8.40 1,102 5,512 11,023 16,535 22,046 27,558 33,069 38,581 44,093 Ordinate Abscissa 10.25-10.25 10.16 9.74 9.41 8.90

.40 0

5,000 SINGLE LOOP Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Type BCR233 Figure 3.2.1-5 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Average Planar Exposure (MWD/MT)

O

0 J

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8.0 Q o

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7.5 X O 7.0 C

6.5 10.01 10.01:

Bundle Average Exposure (MWD/MT) 0 5,000 10,000 15,000 10.01 MAPLHGR.

kw/ft 10.01 10.01 10.01 10.01 10.01 m

5.0 Qi 5,000 10,000 15,000 20,000 25,000 30,000 BUNDLE AVERAGE EXPOSURE (HWD/N)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus BUNDLE AVERAGE EXPOSURE ENC XN-1 FUEL Figure 3.2.1-6 35,000 40,000 45,000 O

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POWER DISTRIBUTION LIMITS 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be:

a.

Greater than or equal to the applicable MCPR limit determined from Table 3.2.3-1 during steady state operation at rated core flow, or b.

Greater than or equal to the greater of the two values determined from Table 3.2.3-1 and Figure 3.2.3-1 during steady state operation at other than rated core flow.

APPLICABILITY:

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

ACTION:

With MCPR less'han the applicable MCPR limit determined from Table 3.2.3-1 and Figure 3.2.3-1, initiate corrective action within 15 minutes and restore MCPR to within the requi red 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 percent 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 RE UIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or equal to the appli-cable MCPR limit determined from Table 3.2.3-1 and Figure 3.2.3-1.

a.

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

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 percent 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 ROD PATTERN for MCPR.

WASHINGTON NUCLEAR - UNIT 2 3/4 2-6 Amendment No. 28

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Table 3.2.3-1 MCPR OPERATING LIMITS FOR RATED CORE FLOW MCPR 0 eratin Limit E ui ment Status 1.

Normal*

2.

Control Rod Insertion Bounded by Tech.

Spec.

Limits (3. 1.3. 4 - p3/4 1-7) 3.

RTP Inoperable, Normal Scram 100 Core Flow 1.27 ENC Fuel 1.28 GE Fuel 1.32 Both Fuel Types 1.32 ENC Fuel 1.33 GE Fuel 106 Core Flow 1.27 ENC Fuel 1.28 GE Fuel 1.32 Both Fuel Types 1.33 ENC Fuel 1.34 GE Fuel This MCPR is based on the ENC reload safety analyses performed using the control rod insertion times shown.below (defined as normal scram).

In the event that surveillance

4. 1.3.2 shows these scram insertion times may be exceeded, the plant thermal limits of Step 1.

above are to default to the values in Step 2.

above and the scram insertion times must meet the requirements of Tech.

Spec.

3. 1.3.4.

Position Inserted From Full Withdrawn Notch 45 Notch 39 Notch 25 Notch 5

Slowest Measured Average Control Rod Insertion Time to Specified Notches for Each Group of 4 Control Rods Arranged in a Two-b -Two Arra Seconds)

.404

.660 1.504 624 WASHINGTON NUCLEAR " UNIT 2 3/4 2-7 Amendment No. 28

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1.6 1.5 1.0 20 30 40 50 60 70 80 Total Core Recirculating Flow (%Rated) 90 100 110 Reduced Flow MCPR Operating Limit Figure 3.2.3-1 860599.1A

POWER DISTRIBUTION LIMITS

\\

3/4. 2. 4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed 13.4 kW/ft.

The LHGR for ENC fuel shall not exceed the values shown in Figure 3.2.4-1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

~125 f RREER EEERRRE RRREER.

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 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 reduce THERMAL POWER to less than 25K 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 RE UIREMENTS 4.2.4 LHGRs shal'1 be determined to be equal to or less than the limit:

a.

At least once per 24.hours, 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 15K 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.

WASHINGTON NUCLEAR " UNIT 2 3/4 2"9 Amendment No.

l

Cn IPl XI EXP LHGR M

I C)

Permissible Region of Operation 10,000 20,000 30,000 40,000 Average Planar Exposure (MWD/MT) 50,000 0

510 2,580 5,230 7,940 10,470 13,220 15,990 18,708 21,590 24,420 27i280 30,150 33,050 35,960 38,900 41,830 44,760 15.62 15.621 15.10 14.71 14.19 14.13 14.06 14.06 14.00 13.93 13.93 13.08 12.24 11AO 10.47 9.55 8.65 7.77 O

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure Exxon 8x8 Fuel Figure 3.2A-1 860599.3A

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I INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of-Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3. 3. 4. 2-3.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to.

30 X of RATED THERMAL POWER.

ACTION:

a.

C.

d.

e'.

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

With the number o

OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel(s) in the tripped condition within one hour.

With the number of OPERABLE channels two or more less than required

.by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

l.

If the inoperable channels consist of one turbine governor valve channel and one turbine throttle valve channel, place both inoperable channels in the tripped condition within one hour.

2.

If the inoperable channels include two turbine governor valve channels or two turbine throttle valve channels, declare the trip system inoperable.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or evaluate MCPR to be equal to or greater than the applicable MCPR limit without EOC-RPT within one hour" or take the ACTION required. by Specification 3.2.3.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or evaluate MCPR to be equal to or greater than the applicable MCPR limit without EOC-RPT within one hour" or take the ACTION required by Specification 3.2.3.

"If MCPR is evaluated to be equal to or greater than the app1 icab 1 e MCPR 1 imit without EOC-RPT within one hour, operation may continue and the provisions of Specification 3.0.4 are not applicable.

WASHINGTON NUCLEAR - UNIT 2 3/4 3-41 Amendment No. 28

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3/4. 1 REACTIVITY CONTROL SYSTEMS BASES I'/4.

l. 1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that (1) the reactor can be made subcritical from all'operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3), the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold,. xenon-free condition and shall show the core to be sub-critical by at least R + 0.38K delta k/k or R + 0.28K delta k/k, as appropriate.

The value of R in units of I delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity.

The value of R must be positive or zero and must be determined for each fuel loading cycle.

Two different values are supplied in the Limiting Condition for Operation to provide for the different methods of demonstration of the SHUTDOWN MARGIN.

The highest worth rod may be determined analytically or by test.

The SHUTDOWN MARGIN is demonstrated by an insequence control rod withdrawal at the begin-ning of life fuel cycle conditions, and, if necessary',

at any future time in the cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure.

Observation of subcriticality in this condition assures subcriticality with the most reactive control rod fully withdrawn.

This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.

3/4. 1. 2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement is small, a careful check on actual reactor conditions compared to the predicted conditions is necessary.

Any changes in reactivity from that predicted (predicted core Keff) can be deter-mined from the core monitoring system (monitored core K ff).

In the absence eff

'f any deviation in plant operating conditions of reactivity anomaly, these values should be essentially equal since the calculational methodolo-gies are consistent.

The predicted core K

is calculated by a 30 core eff simulation code as a function of cycle exposure.

This calculation is performed for projected or anticipated reactor operating states/conditions throughout the cycle and is usually done prior to cycle operation.

The monitored core K ff is the K ff as calculated by the core monitoring system for actual plant eff conditions.

Since the comparisons are easily done, frequent checks are not an impo-sition on normal operation.

A 1 percent deviation in reactivity from that of the predicted is larger than expected for normal operation and, therefore, should be thoroughly evaluated.

A deviation as large as 1 percent would not exceed the design conditions of the reactor.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 1-1 Amendment No.

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REACTIYITY CONTROL SYSTEMS BASES 3/4. 1.3 CONTROL.RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the safety analyses, and (3) limit the potential effects of the rod drop accident.

The ACTION statements permit variations from the basic re-quirements but at the same time impose more restrictive criteria for continued operation.

A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum.

The re-quirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a

rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the core wide transient analyzed in XN-NF-85-143.

This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than the fuel cladding safety limit.

The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem 'with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reac-tor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods.with inoperable accumulators are declared inoperable and Specification

3. 1.3. 1 then applies.

This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure.

Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 1"2 Amendment No. 28

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REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)

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 completing 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 position 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 re-sult 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 20K 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'HERMAL POWER is less than or equal to 20K 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.

Parametric Control Rod Drop Accident analyses have shown that for a wide range of key reactor parameters (which envelope the operating ranges of these parameters) the fuel enthalpy rise during a postulated control rod drop acci-dent remains considerably lower than the 280 cal/gm limit.

For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective-delayed neutron fraction and maximum four-bundle local peaking factor are compared with the inputs to the parametric analyses to deter-mine the peak fuel rod enthalpy rise.

This value is then compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle.

If cycle-specific values of the above parameters are outside the range assumed in the parametric analysis, an extension of the analysis or a cycle-specific analy-sis may be required.

Conservatism present in the analysis, results of the para-metric stu'dies and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in XN-NF-80-19 Volume 1.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 1-3 Amendment No.28 II

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REACTIVITY CONTROL SYSTEMS BASES CONTROL ROD'S PROGRAM CONTROLS (Continued)

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.

3/4. 1.5 STANDBY LI UID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free

shutdown, assuming that none of the withdrawn control rods can be inserted.

To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.

A minimum quantity of 4587 gallons of solution containing a minimum of 5500 pounds of sodium pentaborate is required to meet this shutdown requirement.

There is an additional allowance of 150 ppm in the reactor core to account for imperfect mixing.

The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required minimum pumping rate is 41.2 gpm.

The minimum storage volume of the solution is estab-lished to allow for the portion below the pump suction that cannot be inserted and the filling of other piping systems connected to the reactor vessel.

The temperature requirement on the sodium pentaborate solution 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 inoperable or for longer periods of time with one of the redundant components inoperable.

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.

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.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 1-4 Amendment No. 28

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3/4. 2 POWER OISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 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 on the rod to rod power distribution within an assembly.

For GE fuel, the peak I

clad temperature 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 Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for GE fuel is this LHGR of the highest powered rod divided by its local peaking factor which results in a calculated LOCA PCT much less than 2200'F.

The Technical Specification APLHGR for ENC fuel is specified to assure the PCT following a postulated LOCA will not exceed the 2200'F limit.

The limiting value for APLHGR is'shown in Figures 3, 2. 1-1,

3. 2. 1-2, and 3. 2.1-3 for two recirculation loop operation.

These values shall be multiplied by a factor of 0.84 for single recirculation loop oper ation.

This multiplier is determined from comparison of the limiting analysis between two recirculation loop and single recirculation loop operation.

The calculational procedure used to establish the APLHGR shown on Figures

3. 2. 1-1,

3. 2. 1-2, and 3. 2. 1-3 is based on a loss-of-coolant accident analysis.

The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50.

These models are described in Reference 1 or XN-NF-80-19, Volumes 2, 2A, 2B and 2C, Rev.

1.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-1 Amendment No. 28

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POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The flow biased simulated thermal power-upscale scram setting and control rod block functions of the APRM instruments limit plant operations to the region covered by the transient and accident analysis.

In addition, the APRM setpoints must be adjusted for both two recirculation loop operation and single recirculation loop operation to ensure that the MCPR does not become less than the fuel cladding safety limit or that ) 1X pla'stic strain does not occur in the degraded situation.

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

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-2 Amendment No. 2B

POWER DISTRIBUTION LIMITS BASES 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO k

The required operating limit MCPRs at steady-state operating 'conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial condi-tion 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 Specifica-tion 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 temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification

3. 2.3 is obtained and presented in Table 3. 2.3-1.

The evaluation of a given transient begins with the system initial param-eters shown in XN-NF-85-143. that are'nput to a ENC-core dynamic behavior tran-sient computer program.

The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle.

The codes and methodology to evaluate pressurization and nonpressurization events are described in XN-NF-79-71.

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

The purpose of the MCPRf of Figure 3.2.3-1 is to define operating limits at other than rated core flow conditions.

At less than 100K of rated flow the requi red MCPR is the maximum of the rated flow MCPR determined from Table 3.2.3-1 and the reduced flow MCPR determined from Figure 3.2.3-1, MCPRf assures that the Safety Limit MCPR will not be violated.

MCPRf is only cal-culated for the manual flow control mode.

Automatic flow control operation is not permitted.

WASHINGTON NUCLEAR " UNIT 2 B 3/4 2-3 Amendment No. 28

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4'OWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

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

For all designated control rod patterns 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 start-up testing of the plant, a

MCPR evaluation will be made at 25K of RATED THERMAL POWER level with minimum recirculation pump speed.

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

The daily requirement for calculating MCPR when THERMAL POWER is greater'han or equal to 25K of RATED THERMAL POWER is sufficient since power distribution shifts 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 approached 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.

WASHINGTON NUCLEAR UNIT 2 B 3/4 2-4 Amendment No.28

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