ML17284A479

Amend 62 to License NPF-21,revising Tech Specs Re Limiting Conditions for Operation & Instrumentation Setpoints to Allow Operation Up to 75% Power W/One Recirculation Loop Operating to Design Burnup of Reload Fuel

ML17284A479
Person / Time
Site:	Columbia
Issue date:	08/05/1988
From:	Knighton G Office of Nuclear Reactor Regulation
To:
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ML17284A480	List: ML17284A479 ML17284A481
References
NUDOCS 8808190041
Download: ML17284A479 (56)

Text

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

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 62 License No.

NPF-21 1.

The Nuclear 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 March 7, 1988 as supplemented on May 13, 1988 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 the 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.

8809i9o pg000397 880905 gDOCK-o pNU

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

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.

62

, 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.

3.

This amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 5, 1988 George

. Knighto, Projec Directora e

Division of Reactor IV, V and Special Director V

Projects - III, projects

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ENCLOSURE TO LICENSE AMENDMENT NO.

62 FACILITY OPERATING LICENSE 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 vertical lines indicating the areas of change.

REMOVE V

V1 Xll X111 XX 2-4 3/4 2-4 3/4 2-4c 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 3-55 3/4 3-102 3/4 3-103 3/4 3-104 3/4 4-1 3/4 4-2 83/4 2-1 83/4 2-4 83/4 3-7 83/4 3-7a 83/4 4-1 INSERT V

V1 X11 X111 XX 2-4 3/4 2-4 3/4 2-4c 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-11 3/4 2-12 3/4 2-13 3/4 2-14 3/4 3-55 3/4 3-102 3/4 3-103 3/4 3-104 3/4 4-1 3/4 4-2 83/4 2-1 83/4 2-4 83/4 2-5 83/4 2-6 83/4 3-7 83/4 4-1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 0 APPLICABILITY....

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 SHUTDOWN MARGIN.

3/4. 1. 2 REACTIVITY ANOMALIES..........

3/4. 1. 3 CONTROL RODS Control Rod Operability..

Control Rod Maximum Scram Insertion Times.......

PAGE 3/4 O-1 3/4 1-1 3/4 1-2 3/4 1"3 3/4 1-6 Four Control Rod Group Scram Insertion Times...........

3/4 1-8 Control Rod Scram Accumulators Control Rod Drive Coupling......................

Control Rod Position Indication...............,........

Control Rod Drive Housing Support......................

3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer...........

Rod Sequence Control System.

Rod Block Monitor 3/4. 1.5 STANDBY LIQUID CONTROL SYSTEM..........................

3/4.2 POWER DISTRIBUTION LIMITS 3/4 1-9 3/4 1-11 3/4 1-13 3/4 1"15 3/4 1-16 3/4 1"17 3/4 1-18 3/4 1-19 3/4. 2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............

3/4 2-1 3/4 2.2 APRM SETPOINTS.................

3/4.2.3 MINIMUM CRITICAL POWER RATIO...

3/4.2.4 LINEAR HEAT GENERATION RATE................

3/4.2.5 (RESERVED FOR FFTR) 3/4. 2. 6 POWER/FLOW INSTABILITY....................

3/4.2.7 NEUTRON FLUX NOISE MONITORING...

3/4 2-5 3/4 2-6 3/4 2"9 3/4 2-11 3/4 2"13 WASHINGTON NUCLEAR " UNIT 2 Amendment No. 62

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 3 INSTRUMENTATION PAGE 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................

3/4 3-10 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.

3/4 3"25 3/4. 3. 5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION..................

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION.........

3/4. 3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.......

Seismic Monitoring Instrumentation.....

Meteorological Monitoring Instrumentation..

Remote Shutdown Monitoring Instrumentation.

Accident Monitoring Instrumentation.....

Source Range Monitors........

Traversing In-Core Probe System............

Fire Detection Instrumentation............

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Loose-Part Detection System.................

Radioactive Liquid Effluent Monitoring Instrumentation.

Radioactive Gaseous Effluent Monitoring Instrumentation.

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation..

End-of-Cycle Recirculation Pump Trip System Instrumentation..........

3/4 3-37 3/4 3"41 3/4 3-47 3/4 3-52 3/4 3-58 3/4 3-61 3/4 3"64 3/4 3-67 3/4 3"70 3/4 3-76 3/4 3"77 3/4 3"79 3/4 3"83 3/4 3-84 3/4 3-89 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM..................

3/4 3"96 3/4.3.9 FEEDMATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.

3/4 3-98 WASHINGTON NUCLEAR - UNIT 2 vi Amendment No.62

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INDEX BASES SECTION 3/4. 0 APPLICABILITY 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. l. 1 SHUTDOWN MARGIN...................

3/4. 1. 2 REACTIVITY ANOMALIES 3/4.1.3 CONTROL RODS..............

PAGE B 3/4 0-1 B 3/4 1-1 B 3/4 1-1 B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTROLS B 3/4 1-3 3/4. 1.5 STANDBY LIQUID CONTROL SYSTEM.....,..............

B 3/4 1"4 3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

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3/4. 2. 2 APRM SETPOINTS B 3/4 2-1 B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO.....................

B 3/4 2"3 3/4.2.4 LINEAR HEAT GENERATION RATE 3/4.2.5 (RESERVED FOR FFTR)

B 3/4 2-4 3/4.2.6 POWER/FLOW INSTABILITY...........................

B 3/4 2-4 3/4.2.7 NEUTRON FLUX NOISE MONITORING.............

B 3/4 2"5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION........ B 3/4 3-1 3/4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION.........~....

B 3/4 3"2 3/4.3.3 3/4. 3. 4 3/4.3.5 3/4.3.6 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.....................

RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.

REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.

CONTROL ROD BLOCK INSTRUMENTATION..

B 3/4 3"2 8 3/4 3"3 B 3/4 3"4 8 3/4 3-4 WASHINGTON NUCLEAR - UNIT 2 X11 Amendment No.

62

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INDEX BASES SECTION INSTRUMENTATION (Continued) 3/4.3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation............

Seismic Monitoring Instrumentation..............

Meteorological Monitoring Instrumentation.......

Remote Shutdown Monitoring Instrumentation......

Accident Monitoring Instrumentation.............

Source Range Monitors................

Traversing In-Core Probe System......

Fire Detection Instrumentation...........

Loose-Part Detection System.

Radioactive Liquid Effluent Monitoring Instrumentation..

Radioactive Gaseous Effluent Monitoring Instrumentation.

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PAGE B 3/4 3-4 B 3/4 3-4 B 3/4 3-5 B 3/4 3-5 8 3/4 3"5 B 3/4 3-5 B 3/4 3-5 8 3/4 3"6 B 3/4 3-6 B 3/4 3-6 B 3/4 3-7 3/4.3.8 3/4.3.9 FEEDMATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.......................

B 3/4 3-7 TURBINE OVERSPEED PROTECTION SYSTEM.............

B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM............................

3/4.4.2 SAFETY/RELIEF VALVES..........

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE B 3/4 4-1 B 3/4 4-1 3/4.4.4 Leakage Detection Systems.........

OPerational Leakage.

CHEMISTRY...............-.

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B 3/4 4-la B 3/4 4-2 B 3/4 4-2 3/4.4. 5 SPECIFIC ACTIVITY.............

B 3/4 4-3 3/4.4. 6 PRESSURE/TEMPERATURE LIMITS.....................

B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES B 3/4 4-5 WASHINGTON NUCLEAR - UNIT 2 X111 Amendment No 62

1

LIST OF FIGURES INDEX FEGURE

3. 1. 5-1 3.l. 5-2 SODIUM PENTABORATE TANK, VOLUME VERSUS CONCENTRATION RE(UIREMENTS...

3/4 1-22 PAGE SODIUM PENTABORATE SOLUTION SATURATION TEMPERATURE...

3/4 1"21 3.2. 1-1 MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIALCORE FUEL TYPE 8CR183..........

3/4 2-2

3. 2. 1"2 MAXIMUMAVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR233.................

3/4 2-3

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

3/4 2-4 3.2.1-4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183................

3/4 2-4A

3. 2. 1-5 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR233.

3/4 2-4B 3.2.3"1 3.2. 4"1 3.2. 6-1 3.2. 7-1 3.4. l. 1-1 3.4.6.1-1 3.4. 6. 1-2

4. 7"1

3. 9. 7-1 B 3/4 3-1 REDUCED FLOW MCPR OPERATING LIMIT................

3/4 2-8 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE EXXON 8x8 FUEL 3/4 2-10 THERMAL POWER LIMITS OF SPEC. 3.2.6"1........

3/4 2-12 OPERATING REGION LIMITS OF SPEC. 3.2.7-1.............

3/4 2-14 THERMAL POWER LIMITS OF SPEC. 3.4.1.1-1..............

3/4 4-3a MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (INITIALVALUES)......

3/4 4-20 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (OPERATIONAL VALUES).........

3/4 4-21 SAMPLE PLAN 2)

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

3/4 7-15 WEIGHT/HEIGHT LIMITATIONS FOR LOADS OVER THE SPENT FUEL STORAGE POOL..............................

3/4 9-10 REACTOR VESSEL WATER LEVEL.............

B 3/4 3-8 WASHINGTON NUCLEAR - UNIT 2 XX Amendment No. 62

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FUNCTIONAL UNIT TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS TRIP SETPOINT ALLOWABLE VALUES

1. Intermediate Range Monitor, Neutron Flux - High

2. Average Power Range Monitor:

< 120/125 divisions of full scale

< 122/125 divisions of full scale a.

Neutron Flux-High, Setdown

< 15K of RATED THERMAL POWER

< 20K of RATED THERMAL POWER b.

Flow Biased Simulated Thermal Power - High

1) Flow Biased

2) High Flow Clamped c.

Fixe'd Neutron Flux - High d.

Inoperative

3. Reactor Vessel Steam Dome Pressure - High

4. Reactor Vessel Water Level - Low,.Level 3

5. Main Steam Line Isolation Valve - Closure

6. Main Steam Line Radiation - High

< 0.66W + 51X, with a maximum of

< 113.5X of RATED THERMAL POWER

< 118K of RATED THERMAL POWER N.A.

< 1037 psig

> 13.0 inches above instrument zero"

< 10.0X closed

< 3.0 x full power background

< 0.66W + 54X, with a maximum of

< 115.5X of RATED THERMAL POWER

< 120K of RATED THERMAL POWER N.A.

< 1057 psig

> 11.0 inches above instrument zero

< 12.5X closed

< 3.6 x full powe~

background ee ases igure 3/4 3-1.

V7 CR O

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13.0 12.5 12.0 I

11.5

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11.0 m O 10.5 C

E E ~ 10.0 X

C5 +

9.5 9.0 8.5 0

5,000 10,000 15,000 20,000 25,000 30,000 35,000 13.0 13.0 13.0 13.0 13.0 11.3 9.4 7.9 Bundle Average Exposure MAPLHGR (MwD/MT) kw/ft Two Loop and Single Loop Operation O

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

ANF Sx8 Reload Fuel Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure Figure 3.2.1-3

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THIS PAGE INTENTIONALLY LEFT BLANK WASHINGTON NUCLEAR UNIT 2 3/4 2-4c Amendment No.

62

POWER DISTRIBUTION LIMITS 3/4.2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE VALUE S <

0.66W+ 5 T

S <

0.66W+ 54 T

SRB

< (0.66W + 42X)T SRB < (0.66W + 45K)T where:

S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million lbs/h.

T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY.

T is always less than or equal to 1, APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

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f RRRER RIIERRRE POllER.

ACTION:

with the APRN flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or SRB to be consistent with the Trip Setpoint value( ) 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 '

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 simulated ther-mal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

a.

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

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 with MFLPD greater than or equal to FRTP.

"With MFLPD greater than the FRTP during power ascension up to 90K 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 lOOX times MFLPD, provided that the adjusted APRM reading does not exceed 100X of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

WASHINGTON NUCLEAR UNIT 2 3/4 2-5 Amendment No. 62

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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 or above rated core flow in two loop operation, or when in single loop operation, 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 less than rated core flow'hen in two recirculation loop operation.

APPLICABILITY:

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

ACTION:

With MCPR less than 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 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 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.

62

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Table 3.2.3-1 MCPR OPERATING LIMITS Cycle

~Ex osure Equipment Status MCPR Operating Limit U

to 106K Core Flow GE Fuel ANF Fuel l.

0 MWD " 3750 MWD MTU MTU 2.

3750 MWD -

EOC MWD MTU MTU 3.

3750 MWD -

EOC MWD MTU MTU

1. 40 Normal scram times""

l. 40 Control rod insertion 1.50 bounded by Tech.

Spec.

limits (3.1.3.4-p 3/4 1-8)

1. 28

1. 31

1. 38 4.

3750 MWD -

EOC MWD

~MU MTU 5.

3750 MWD -

EOC MWD MTU MTU RPT inoperable Normal scram times RPT inoperable Control rod insertion bounded by Tech.

Spec.

limits (3. 1.3. 4-p 3/4 1-8)

1. 50

l. 55

l. 37

1. 43 6.

0 MWD -

EOC MWD MTU MTU Single loop operation 1.40 RPT operable Normal scram times""

l. 37 "In this portion of the fuel cycle, operation with the given MCPR operating limits is allowed for both normal and Tech.

Spec.

scram times and for both RPT operable and inoperable.

"*These MCPR values are based on the ANF Reload Safety Analysis 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 have been

exceeded, the plant thermal, limits associated with normal scram times default to the values associated with Tech.

Spec.

scram times (3. 1.3.4-p 3/4 1-8),

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

Spec.

3.1. 3.4.

Position Inserted From Full Withdrawn Slowest measured average control rod insertion times to specified notches for all operable control rods for each group of 4 control rods arranged in a a two-b -two arra seconds Notch 45 Notch 39 Notch 25 Notch 5

WASHINGTON NUCLEAR - UNIT 2 3/4 2-7

.404

.660

1. 504

2. 624 Amendment No. 62

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1.5 COc 1.4 CL0 1.3 9O 1.2 Two Loop Operatton 1.0 10 20 30 40 50 60 70 80 90 100 110 Total Core Recirculating Flow (% Rated)

O Reduced Flow MCPR Operating Limit Figure 3.2.3-1

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3/4. 2. 6 POWER/FLOW INSTABILITY LIMITING CONDITION FOR OPERATION 3.2.6 Operation with THERMAL POWER/core flow conditions which lay in the crosshatched region of Figure 3.2.6-1 is prohibited.

APPLICABILITY:

OPERATIONAL CONDITION 1 When THERMAL POWER is greater than 39K of RATED THERMAL POWER and core flow is less than or equal to 45K of rated core flow.

ACTION:

With THERMAL POWER/core flow conditions which lay in the crosshatched region of Figure 3.2.6-1, initiate corrective action within 15 minutes to establish a

THERMAL POWER/core flow condition which lays outside the crosshatched region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.6 The THERMAL POWER/core flow conditions shall be verified to lay outside the crosshatched region of Figure 3.2.6-1 once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

WASHINGTON NUCLEAR - UNIT 2 3/4 2-1>

Amendment No.62

COx RA OXx CA I

C CP I

BO CC g

50 oo40 0

E 30 o

20 OQ 10 0

20 30 40 50 Core Flow (% Rated) 60 O

Ch fO Operating Region Limits of Speclficatlon 3.2.6 Figure 3.2.6-1

POWER DISTRIBUTION LIMITS 3/4.2.7 NEUTRON FLUX NOISE MONITORING LIMITING CONDITION FOR OPERATION 3.2.7 The APRM and LPRM neutron flux noise levels shall not exceed three (3) times their established baseline values when operating in the region of AP P LICABILITY.

APPLICABILITY:

OPERATIONAL CONDITION 1 with THERMAL POWER/core flow in Region B of Figure 3. 2.7-1, with two reactor coolant system recirculation loops in operation and total core flow less than 45K of rated total core flow, or with one reactor coolant system recirculation loop not in operation.

ACTION:

a.

If baseline APRM and LPRM neutron flux noise levels have not been established for the appropriate reactor coolant system condition (one or two loop operation) since the most recent CORE ALTERATION, then:

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exit the region of APPLICABILITY. Establish baseline APRM and LPRM neutron flux noise.levels prior to re-entering Region B of Figure 3.2.7-1.

b.

If baseline APRM and LPRM neutron flux noise levels have been established for the appropriate reactor coolant system condition (one or two loop operation) since the most recent CORE ALTERATION, then:

With the APRM or LPRM neutron flux noise levels greater than three (3) times their established noise levels, initiate corrective action within 15 minutes to restore the noise levels 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 below the region of APPLICABILITYwithin the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.7.1 The provisions of Specification 4.0.4 are not applicable.

4.2.7.2 The APRM and LPRM neutron flux noise levels shall be determined to be less than or equal to three (3) times their established baseline values:

a.

At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and b.

Within 30 minutes after completion of a THERMAL POWER increase of greater than or equal to 5X of rated THERMAL POWER.

~Detector levels A and C of one LPRM string per core octant plus detector levels A and.C of one LPRM string in the center of the core should be monitored.

WASHINGTON NUCLEAR - UNIT 2 3/4 2"13 Amendment No.62

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50 40 30 I

oV Region Reglon B Line X 20 30 40 50 Core Floor (% Rated) 60 70 Operating Region Limits of Speclflcatlon 3.2.7 Figure 3.2.7-1

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TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPQINTS TRIP FUNCTION 1.

ROD BLOCK MONITOR b.

Inoperative

~m c.

Downscale 2.

APRM a.

Flow Biased Neutron Flux Upscale b.

Inoperative c.

Downscale d.

Neutron Flux - Upscale, Startup 3.

SOURCE RANGE MONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale 5.

SCRAM DISCHARGE VOLUME a.

Water Level-High b.

Scram Trip Bypass 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW TRIP SETPOINT

< 0.66 W + 40K N.A.

> 5X of RATED THERMAL POWER

< O. 66 W + 42K" N.A.

> 5X of RATED THERMAL POWER

< 12'f RATED THERMAL POWER N.A.

< 1 x 10 cps R.A.

> 0.7 cps N.A.

< 108/125 divisions of full scale R.A.

> 5/125 divisions of full scale

< 527 ft 2 in. elevation R.A.

ALLOWABLE VALUE

< 0.66 W+ 43K R.A.

> 3X of RATED THERMAL POWER

< 0.66 W + 45m<

~ l R.A.

> 3X of RATED THERMAL POWER

< 14X of RATED THERMAL POWER N.A.

< 1.6 x 10 cps

.R.A.

> 0 5 cps N.A.

< 110/125 divisions of full scale R.A.

> 3/125 divisions of full scale

( 527 ft 4 in. elevation R.A.

a.

Ups cal e

< 108/125 divisions of full scale

~ b.

Inoperative N.A.

c.

Comparator

< 10X flow deviation e Average ower ange Monitor rod block function is varied as a function of (W).

The trip setting of this function must be maintained in accordance with

< 111/125 divisions of full scale N.A.

< 13K flow deviation recirculation loop flow Specification 3.2.2.

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THIS PAGE INTENTIONALLYLEFT BLANK WASHINGTON NUCLEAR - UNIT 2 3/4 3"102 Amendment No.

THIS PAGE INTENTIONALLYLEFT BLANK WASHINGTON NUCLEAR - UNIT 2 3/4 3-103 Amendment No.

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3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2".

ACTION:

a 0 With one reactor coolant system recirculation loop not in operation:

1.

Within 15 minutes:

a ~

Verify that core flow is greater than or equal to 39K of rated core flow or that THERMAL POWER/core flow conditions lay below the line in Figure 3.4.1.1-1.

With core flow less than 39K of rated core flo> and THERMAL POWER/core flow conditions above the line in Figure 3.4.1.1-1, initiate action to reduce THERMAL POWER to below the line in Figure 3.4. 1. 1-1 or increase core flow to greater than or equal to 39K of rated core flow within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Verify that the requirements of LCO 3.2.7 are met, or comply with the associated ACTION statement within the specified time limits.

2.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a)

Place the recirculation flow control system in the Local Manual (Position Control) mode, and b)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2,

and, c)

Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for General Electric fuel limit to a value of Oe84 times the two recirculation loop operation limit per Specification 3.2. 1, and, d)

Reduce the volumetric flow rate of the operating recircula-tion loop to < 41,725"" gpm.

See Special Test Exception 3.10.4.

""This value represents the actual volumetric recirculation loop flow which produces 100K core flow at 100K THERMAL POWER.

This value was determined during the Startup Test Program.

WASHINGTON NUCLEAR " UNIT 2 Amendment No 62

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REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued ACTION:

(Continued) e)

Perform Surveillance Requirement 4.4.1.1.2 if THERMAL POWER is

< 25K*"" of RATED THERMAL POWER or the recirculation loop flow in the operating loop is < 10K""* of rated loop flow, f)

Reduce recirculation loop flow in the operating loop until the core plate hP noise does not deviate from the estab-lished core plate hP noise patterns by more than 100K.

3.

The provisions of Specification 3.0.4 are not applicable.

4.

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.

With no reactor coolant system recirculation loops in operation, immediately initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.1. 1. 1 With one reactor coolant system recirculation loop not in operation, at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> verify that:

a.

The recirculation flow control s'stem is in the Local Manual (Position Control) mode, and b.

The volumetric flow rate of the operating loop is

< 41,725 gpm.""

""This value represents the actual volumetric recirculation loop flow which produces 100K core flow at lOOX THERMAL POWER.

This value was determined during the Startup Test Program.

"""Final values were determined during Startup Testing based upon actual THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

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

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3/4. 2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.4S.

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 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 Speci-fication APLHGR for ANF fuel is specified to assure the PCT following a postu-lated LOCA will not exceed the 2200 F limit.

The limiting value for APLHGR is shown in Figures 3.2. 1-1 and 3.2. 1-2 for two recirculation loop operation and Figures 3.2. 1-4 and 3.2. 1-5 for single loop operation.

Figure 3.2.1-3 applies to both single and two loop operation.

The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5 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. l.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2"1 Amendment No. 62

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POWER 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 than 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.

3/4.2. 6 POWER/FLOW INSTABILITY At the high power/low flow corner of the operating

domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e. g., power shape, bundle power, and bundle flow).

In February,

1984, GE issued SIL 380 addressing boiling instability and supplying several recommendations.-

In this SIL, the power/flow map was divided into several regions of varying concern.

It also discussed the objectives and philosophy of "detect and suppress,"

coining the phrase.

The ANF topical report for COTRAN (XN-NF-691P) discusses boiling instability.

The SER written on this topical (dated May 10, 1984) interprets the topical to require that the detect-and-suppress surveillance be used in regions which have code calculated decay ratios

.75 or greater and that operation is forbidden in regions having calculated decay ratios of.9 and greater.

The NRC Generic Letter 86-02 addressed both GE and ANF (then EXXON) stability calculation methodology and stated that due to uncertainties, General Design Criterias 10 and 12 could not be met using analytic procedures on a BWR 5 design.

The letter espoused GE SIL 380 and stated that General Design Criterias 10 and 12 could be met by imposing the SIL 380 recommendations in operating regions of potential instability.

The NRC concluded that regions of potential instability constituted calculated decay ratios of.8 and greater by the GE methodology and.75 and greater by the EXXON methodology.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-4 Amendment No. 62

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POWER DISTRIBUTION LIMITS BASES POWER/FLOW INSTABILITY (Continued)

Predicated on the SIL 380 endorsement, WNP-2 has divided the power/flow map on the following boundary lines:

1.

2 3.

4.

5.

80K rod line 45K core flow line APRM rod block line minus 3X power Natural Circulation flow line Minimum Forced Circulation for normal recirculation lineup.

This division conforms to the SIL 380 recommendations with a 3X power penalty on the APRM rod block line.

For LCO 3.2.6, the region of concern is bounded by the APRM rod block line, minus 3X power, the natural circulation flow line, and the 45K core flow line.

Calculated decay ratios between the two flow lines and on the APRM rod block line minus 3X must be less than.9.

Operation in the region between the two flow lines and above the rod block line minus 3X is forbidden due to the potential for boiling instabilities.

For the ease of annual licensing submittals, a 3X margin from the rod block line is taken to avail the opportunity to submit with no Technical Specifica-tion changes under the provisions of 10 CFR 50.59.

This 3X provides margin to assure that vendor stability calculations can easily support the allowable operating region.

For calculational ease the power boundary is linearized between two points, (24K Flow, 39K Power) and (45K Flow, 62K Power).

3/4.2.7 NEUTRON FLUX NOISE MONITORING At the high power/low flow corner of the operating

domain, a small prob-ability of limit cycle neutron flux oscillations exists depending on combina-tions of operating conditions (e.g.,

rod patterns, power shape).

To provide assurance that neutron flux limit cycle oscillations are detected and sup-

pressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region.

Stability tests at operating BWRs were reviewed to determine a generic region of the power/flow map in which surveillance of neutron flux noise levels should be performed.

A conservative decay ratio of 0.75 was chosen as the basis for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs.

This generic region has been determined to correspond to a core flow of less than or equal to 45K of rated core flow and a thermal power greater than that specified in Figure 3.4. 1.1-1 (Reference).

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations.

BWR cores typically operate with neutron flux noise caused by random boiling and flow noise.

Typical neutron flux noise levels of 1-12K of rated power (peak-to-peak) have been reported for the range of low to high recirculation loop flow during both single and dual WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-5 Amendment No.62

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0 POWER DISTRIBUTION LIMITS BASES NEUTRON FLUX NOISE MONITORING (Continued) recirculation loop operation.

Stability tests at operating BWRs have demon-strated that when stability related neutron flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5-10 times the typical values.

Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern) with two reactor recirculation loops in operation.

Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows.

To maintain a reasonable variation between the low flow and high flow ends of the flow range, the range over which a specific baseline is applied should not exceed 20X of rated core flow with two recirculation loops in opera-tion.

Data from tests and operating plants indicate that a range of 20X of rated core flow will result in approximately a 50'ncrease in neutron flux noise level during operation with two recirculation loops; Baseline data should be taken near the maximum rod line at which the majority of operation will occur.

However, baseline data taken at lower rod lines (i.e., lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow.

In the case of single loop operation (SLO), the normal neutron flux noise may increase more rapidly when reverse flow occurs in the inactive jet pumps.

This justifies a smaller flow range under high flow SLO conditions.

Baseline data should be taken at flow intervals which correspond to less than a 50K in-crease in APRM neutron flux noise level.

If baseline data are not specifically available for SLO, then baseline data with two recirculation loops in operation can be conservatively applied to SLO since for the same core flow SLO will exhibit higher neutron flux noise levels than operation with two loops.

However, because of reverse flow characteristics of SLO, the core flow/drive flow re-lationship is different than the two loop relationship and therefore the base-line data for SLO should be based on the active loop recirculation drive flow, and not the core flow.

Because of the uncertainties involved in SLO at high reverse flows, baseline data should be taken at or below the power specified in Figure 3.4. 1. 1-1.

This will result in approximately a 25K conservative baseline value if compared to baseline data taken near the rated rod line and will therefore not result in an overly restrictive baseline value, while providing sufficient margin to cover uncertainties associated with SLO.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2"6 Amendment No. 62

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'NSTRUMENTATION BA'SES MONITORING INSTRUMENTATION (Conti nued) 3/4.3.7. 11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm/

trip setpoints for these instruments shall be calculated and adjusted in ac-cordance with the methodology and parameters in the ODCH to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM This specification is provided to ensure that the turbine overspeed protection system instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.

Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures.

3/4.3.9 FEEDWATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION The feedwater system/main turbine trip system actuation instrumentation is provided to initiate the feedwater system/main turbine trip system in the event of reactor vessel water level equal to or greater than the level 8 setpoint associated with a feedwater controller failure.

WASHINGTON NUCLEAR - UNIT 2 8 3/4 3"7 Amendment No. 62

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and been found to be acceptable provided the unit is operated in accordance with the single recirculation loop operation Technical Specifications herein.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criteria.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

Where the recircula-tion loop flow mismatch'imits cannot be maintained during two recirculation loop operation, continued operation is permitted in. the single recirculation loop operation mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop.

The loop temperature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.

Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145 F.

3/4.4.2 SAFETY/RELIEF VALVES The safety valve capacity is designed to limit the primary system pressure, including transients, in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section III, 1971, Nuclear Power Plant components (up to and including Summer 1971 Addenda).

The Code allows a peak pressure of 110K of design pressure (1250 (design)

X 1.10 = 1375 psig maximum) under upset conditions.

In addition, the Code specifications require that the lowest valve setpoint be at or below design pressure and the highest valve setpoint be set so that total accumulated pressure does not exceed llOX of the design pressure.

The safety valve sizing evaluation assumes credit for operation of the scram protective system which may be tripped by one of two sources; i.e.,

a direct position switch or neutron flux signal.

The direct scram signal is derived from position switches mounted on the main steamline isolation valves (MSIV's) or the turbine stop valve, or from pressure switches mounted on the dump valve of the turbine control valve hydraulic actuation system.

The posi-tion switches are actuated when the respective valves are closing, and follow-ing lOX travel of full stroke.

The pressure switches are actuated when a fast closure of the control valves is initiated.

Further, no.credit is taken for power operation of the pressure relieving devices.

Credit is only taken for WASHINGTON NUCLEAR - UNIT 2 B 3/4 4-1 Amendment No. 62

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