Text

Proposed Tech Specs Re single-loop Operation
	ML17279A919
Person / Time
Site:	Columbia
Issue date:	03/07/1988
From:	; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	;
Shared Package
ML17279A913	List: ML17279A912; ML17279A916; ML17279A917; ML17279A918; ML17279A919;
References
	NUDOCS 8803300298
	Download: ML17279A919 (69)
	v • d • e

ATTACHMENT TO MNP-2 SINGLE LOOP OPERATION

SUMMARY

REPORT TECHNICAL SPECIFICATION CHANGES (INCLUDING JUSTIFICATION)

The following Technical Specification pages are attached:

- Justification for Changes Revise Index to add new 3/4.2.6 and 3/4.2.7 V1 Revise Index to delete 3/4.3.10 X11 Revise (Bases) Index to add new 3/4.2.6 and 3/4.2.7 X111 Revise (Bases) Index to delete 3/4.3.10 XX Revise List of Figures 2-4 Revise Table 2.2.1-1 3/4 2-4 Revise Figure 3.2.1-3 3/4 2-4C Delete Figure 3.2.1-6 3/4 2-5 Revise as indicated 3/4 2-6 Revise as indicated 3/4 2-7 Revise MCPR Operating Limits 3/4 2-S Revise Figure 3.2.3-1 New Page (-1-) Add new Section 3/4.2.6 Power/Flow Instability 3/4 3-104 Adds new Figure 3.2.6-1 0 New Page 3/4 3/4 3-104 3-55

(-2-) Add new Section 3/4.2.7 Neutron Flux Noise Monitoring Adds new Figure 3.2.7-1 Revise Table 3.3.6-2 3/4 3-102 Delete Section 3/4.3.10 3/4 3-103 Delete Section 3/4.3.10 3/4 3-104 Delete Figure 3.3.10-1 3/4 4-1 Revise as indicated New Page (-4-) Add new Action Statement to LCO 3.4.1.1 3/4 4-2 Revise as indicated aS/4 2-1 Revise as indicated New Page (-5-) Add new Bases Section 3/4.2.6 New Page (-6-) Add new Bases Section 3/4.2.6 B3/4 3-7 Delete Bases Section 3/4.3.10 B3/4 3-7a Delete Bases Section 3/4.3.10 B3/4 3-7 Add new Bases Section 3/4.2.7 B3/4 3-7a Add new Bases Section 3/4.2.7 a > i(~ '3c i'. c B3/4 4-1 Revi se as i ndi cated 8803300298 880307 PDR ADOCK 05000397 p DCD

0 JUSTIFICATION FOR CHANGES TO TECHNICAL SPECIFICATION 3.3.10 i~ Move The this LCO to the intent of the POWER LCO DISTRIBUTION LIMITS section of Tech. Specs.

i s to moni tor neutron flux noi se l evel s to detect the approach of an unstable region of operation. The LCO has little or nothing to do with instrument calibration, and thus, will be more appropriately located in the POWER DISTRIBUTION LIMITS section of Tech. Specs, immediately adjacent to related LCO 3/4.2.6, Power/Flow Stability.

2) Modify wording in the LCO and APPLICABILITY sections to more clearly define the region of applicability where noise monitoring is required.

3) Clarify ACTION based upon whether baselining has been performed or not.

4) Incorporate Surveillance Requirements 4.3.10.2 and 4.3.10.3 (old) into the ACTION statement (where they belong, since they constitute action statements).

5) Modify Figure 3.3.10-1 (old) to mor e clearly identify the separate regions where 1) noise monitoring is required, and 2) operation is prohibited.

6) Remove ACTION statement b. from 3.3.10 and incorporate it in a separate LCO, 3/4.2.6, Power/Flow Stability. Currently, this ACTION statement exists with no stated LCO.

JUSTIFICATION TO CHANGES to 3/4.1.1 Gather 15 minute actions and remove them from within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months cri teri a.

2) Scram and rod block trip setpoints do not need to be changed as the analysis envelopes operation up to 75$ thermal power with single loop core flows and drive flows.

3) 34% core flow limit is in accordance with the SIL 380 recommendations of Minimum For ced Circulation. The true minimum forced circulation at WNP-2 is 24%, however, the normal flow line followed during reactor startup is the 2 pump, 15-Hz, FCY full open line, which is the 34K line. The previous value, 39K, was based on the flow that Duane Ar nold, a non-flow control valve plant, could achieve with the recirc pump motor generator scoop tubes set at minimum.

4) Move power/flow instability requirements under power distribution.

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INOEX LIMITING CONOITIONS FOR OPERATION ANO SURVEILLANCE REOUIREMENTS SECTSOR PAGE 3/4. 0 APPLICABILITY...'...... 3/4 0-1 3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4. l. 1 SHUTDOW MARGIN...... 3/4 1-1 3/4. 1. 2 REACTIVITY ANOMALIES....... 3/4 1-2 3/4. 1.3 CONTROL ROOS Control Rod Operability.............. '/4 1"3 Control Rod Maximum Scram Insert'io imes.............. 3/4 1"6 Control Rod Averaoe Scram Insert+ Times.............. 3/4 1-7 Four Control Rod Group Sera ertion Times.. 3/4 1-8 Control Rod Scram Accumul~~. 3/4 1-9 Control Rod Orive Coup 3/4 1-11 Control Rod Positio ication........... 3/4 1"13 Control Rod Oriv sing Support.. 3/4 1"15 3/4.1.4 CONTROL ROO PROGRAM CONTROLS Rod Worth Minimizer.........,....................... 3/4 1-16 Rod Sequence Control System....... 3/4 1-17 Rod Block Monitor...................... 3/4 1-"8 3/4.1.5 STANOBY LIQUIO CONTROL SYSTEM.......................... 3/4 1-19 3/<. 2 POMER OISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE............. 3/4 2-1 3/4 2.2 APRM SETPOINTS.. 3/4 2-5 3/4.2.3 MINIMUM CRITICAL POWER RATIO. 3/4 2-6 3/4 2 4 LINEAR HEAT GFNERATION RATE................... 3/4 2-8

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INOEX MITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4. 3 INSTRUMENTATION 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. 4 R ECIRCULATION PUMP QIP ACTUATION INSTRUMENTATION ATWS Recirculation P grip System Instrumentation.. 3/4 3"37 End-of-Cycle RecirculaM~Pump Trip System Instrumentation..........M~ ........................ 3/4 3-41

,3/4. 3. 5 REACTOR CORE ISOLATION COOL INSTRUMENTATION.............. ~ SYSTEM ACTUATION 3/4 3"47 3/4. 3. 6 CONTROL ROD BLOCK INSTRUMENTATION.................... 3/4 3-52

/4. 3. 7

~ ~ MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...A'............ 3/4 3-58 Seismic Monitoring Instrumentation........ 3/4 3"61 Meteorological Monitoring Instrumentation............ 3/4 3-64 Remote Shutdown Monitoring Instrumentation........... 3/4 3-67 Accident, Monitoring Instrumentation.................. 3/4 3"70 Source Range Monitors...........................'..... 3/4 3"76 Traversing In-Core Probe System. 3/4 3-77 Fire Detection Instrumentation....................... 3/4 3-79 Loose-Part Detection System.............. 3/4 3-83 Radioactive Liquid Effluent Monitoring Instrumentation......... 3/4 3-84 Radioactive Gaseous Effluent Monitoring Instrumentation........,.. 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. 36

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IHOEX BASES K

SBCTION PAGc 3/4e0 APPLICABILITYo ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ooo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o 8 3/4 Q-l 3/4.1 REACTIVITY CONTROL SY~i 3/4.1.1 SHUTDOWN MARGIN.................................. 8 3/4 REACTiVITY ANOMALIES............................. 8 3/4 l-l 1-1'/4.1.2 3/4.1. 3 CONTROL ROOS..................................... 8 3/4 1-2 3 /4. 1.4 3/4. 1.5 CONTROL ROO PROGRAM STANOBY LIOUID CONTROL CONTROLS.....

SYSiiH...........

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3/4 1-3 3/4 1-4 3/4. 2 POWER OISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR H - ERATION RATEo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o yo

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 3/4 2-1 3/4.2.2 APRH SETPOINTS...... ~RA............... 8 3/4 2-2 3/4.2. 3 MINIMUM CRITICAL RATIO................ ..... 8 3/4 2-4 3/4,2.4 LINEAR HAT G ION RATE...................... 8 3/4 2-5 3/4. 3 IHSTRUMEHTATIQN 3/4.3.1 REACTOR PROTECTION SYSTEM IHSTRUMEHTATIQN........ 8 3/4 3-1 3/4.3. 2 ISQLATION ACTUATIOH IHSTRUMEHTATION......... 8 3/4 3-2 3/4. 3. 3 EMERGENCY CORE COOLING SYSTEM ACTUATIOH IHSTRUMEHTATIOH................ 9 3/4 3-2 3/4. 3. 4 RECIRCULATION PUMP TRIP ACTUATIQH INSTRUMENTATION................. 8 3/4 3"3 3/4.3. S REACTOR CORE ISOLATION COOLIHG SY~ic.

ACTUATION INSTRUMM'ATIOH....................... 8 3/4 3-4 3/4. 3. o CONTROL RQO BLOCK IHSTRUMEHTATIQH............. 8 3/4 3-4 CQme ver @v ~t=TR)

POlgGR//FLOIV INSTAGIU~i/ lS 3/4 2.-

N~VTR0~ ~~ MolSG QOQiToaiH(> .~... 8 '5/g 2.-

WASHINGTON NUCLEAR " UNIT 2 X11

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CONTROLLED COPY INOEX SES SECTION PAGE INSTRUMENTATION (Continued) 3/4. 3. 7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...... 8 3/4 3"4 Seismic Monitoring Instrumentation........ 8 3/4 3-4 Meteorological Monitoring Instrumentation. 8 3/4 3"5 Remote Shutdown Monitoring Instrumentation 8 3/4 3-5 Accident Monitoring Instrumentation....... 8 3/4 3-5 Source Range Monitors.. 8 3/4 3-5 Traversing In-Core Probe System........... 8 3/4 3-5 Fire Detection Instrumentation. 8 3/4 3-6 Loose-Part Detection System............... 8 3/4 3-6 Radioactive Liquid Effluent Monitoring Instrumentation........................... 8 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instrumentation......... 8 3/4 3-7 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM....... 8 3/4 3-7 3/4.3.9 FEEDMATER SYSTEM/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION................. 8 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM.......................'..... 8 3/4 4-1 3/4.4. 2 SAFETY/RELIEF VALVES.... ~.............. 8 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....................... 8 3/4 4-la Operational Leakage...................... 8 3/4 4-2 3/4.4. 4 CHEMISTRY.................... 8 3/4 4-2 3/4.4. 5 SPECIFIC ACTIVITY.... 8 3/4 4"3 3/4. 4.6 PRESSURE/TEMPERATURE LIMITS..................... 8 3/4 4-4 3/4.4.7 MAIN,STEAM LINE ISOLATION VALVES;............... 8 3/4 4-5 WASHINGTON NUCLEAR - UNIT 2 X111 Amendment No: 38

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CONT HOLLEO COPY INDEX IST OF FIGURES

'URE PAG

. 1.5"1 SODIUM PENTABORATE SOLUTION SATURATION TEMPERATURE... 734 1-21

3. 1. 5-'2 SODIUM PENTABORATE TANK, VOLUME VERSUS CONCENTRATION REQUIREMENTS 3/4 1-22

3. 2. 1" 1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE'PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183. 3/4 2-2

3. 2. 1-2 MAXIMUM AVERAGE Pl.ANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PI.ANAR EXPOSURE, INITIAL CORE FUEL TYPE BCR233. 3/4 2-3

3. 2. 1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GfNERATION 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 (MAPl.HGR) 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) VfRSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE BCR233.....

aery 3/4 2-4B

3. 2. 3-1 REDUCED FLOW MCPR OPERATING LIMIT. 3/4 2-7

3. 2. 4-1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE EXXON 8x8 FUEL 3/4 2-10

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

3. 4. 6. 1-1 MINIMUM RfACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (INITIAL VALUES)...... 3/4 4-20

3. 4. 6. 1-2 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (OPERATIONAL VALUES)....,.... 3/4 4-21

4. 7-1 SAMPLE PI.AN 2) FOR SNUBBER FUNCTIONAL TEST .......... 3/4 7-15

3. 9. 7" 1 WEIGHT/HEIGHT LIMITATIONS FOR LOADS OVER THE SPENT FUEL STORAGE POOL. 3/4 e-ao B 3/4 3"1 RfACTOR VESSEL WATER LEVEL B 3/4 3-8 WASHINGTON NUCLEAR - UNIT 2 XX Amendment No. 28

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TAAI.E . l-l RFAClOR PROTECT tOII 5 ~ RE ." II.i'Hi~iTATln!I SEIPOltITS fn ALLOWABLE J

~ 1 FUIICT]DUAL U(IIT TRIP SETPOINT VALUES Cl

-I O 1. intermediate Range Hu>>itor, IIeutron F)ux - Iiigh < 120/125 divisions < l22/125 divisions of full scale of full scale P

Cl 2. Avorago Power Aange Ho>>itor:

I gal a. IIeutron Flux-IIIgh, Setdown < 15K of RATED TIIEAHAL POWEA < 20K of BATED TIIEAIQL POWEA

b. Flow Diased Slmulateil Thermal I'ower- Illgh peea44ou.

/pe Flow Biased < 0.66M + 51K, with < 0.66M + 54K, of with a oaxfooo of a maximiua z)~ Illgh Flow Clamped a 133.5% of RATED TIIERHAL PDMER

< 115.5X of RATED TIIERHAL POWER

~ ~0-.66M-+-SO-.~I-Hr-ox4mu~f

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~~exlmt~f ATKD- <-4845K-of Ref D WIER Bt ~IIEAHAL-POMER

c. Fixeil IIeutron Flux - Ilfgh < llQX of AA TIIEAIIAL POWER < 120X of AATED TIIERHAL POWER

d. Inoilerativo II.A.

3. Reactor Vessel Steam Dome Pressure - Iligh < 1037 psig < 1057 psig II. Reactor Vessel Wato> Level - Low, Level 3 > 13.0 inches above instrument > ll.O inches above zeros instrument zero

5. Hain Steam Line isolation Valve - Closure < )O.OX closed < 12.5X closed

6. Hain Steam Line Railiation - Iligli < 3.0 x full power background < 3.6 x full power background See Oases F~giire 8 3/4 3-1.

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SIAJS c.C 0 oooration p

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<<t e 10.5 Exposure (MWD/MT)

MAP LHGR kw/tt

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g$ 5,000 10,000 13.0 13.0 0

lQ 9.5 15,000 13.0 I e ITl C 20,000 13.0 25,000 11.3 U

9.0 30,000 9.4 A 35,000 7.9 0U 8.0 5,000 10,000 15,000 20,000 25,000 '0,000 '35,000 40,000 Bundle Average Exposure (MWD/MT)

ANF Gx8 Reload Fuel Maximum Average Planar Linear Heat Generation Rate (NIAPLHGR) Versus Bundle Average Exposure Figure 3.2.1-3

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10.0 10.01 10.Q1 10. 01: 0.01 L

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9.0 Sing I e Loo p Ope lo a 0 o cl Bundle cA 8.5- Average w 0 L

Exposure MAPLHGR Q

Q o 8.0 (MWDIMT) k /lt 4D C E a

~Q 0 10.01 I 7.5 S,QOQ 10.01 X y A gXle 10,000 10.01 C 15,000 10.01 8 7.0 C

6.0 5,000 ~ 10,000 . 15,0QO 20,QQO 25,000 30,0QO 35,000 4Q,QOO 45,MO llUHDLE AVERAGE EXPOSURE. (HMO/N)

Maxlmurn Average Planar Linear Heat Generation Rate (MAPLHGR) Versus BUHDLE AVERAGE EXPOSURE ANF Gx8 Reload Fuel Figure 3.2.1-6

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r CONTROLLED COPY POWER DISTRIBUTION LIMITS

/4 2 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (5) and flow biased neutron flux-upscale control rod block trip setpoint. (SRB) shall be established according to the following relationships:

&PG4+8A TRIP SETPOINT ALLOWABLE VALUE 5 < (0.66W + 54 )T SRB

< (0.66W + 42 )T SRB

< (0.66W + 46Ã)T where: 5 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.

PPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or qual to 25.o of RATED THERMAL POWER.

ACTION:

With the APRM 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 5 or SRB, as above determined, initiate corrective action within 15 minutes and adjust 5 and/or SRB to be consistent with the Trip Setpoint value(*) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS

4. 2.2 The FRTP and the MFLPO 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15 of RATED THERMAL POWER, and ,

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with MFLPO greater than or equal to FRTP.

>lith MFLPO greater than the FRTP during power ascension up to 90~ of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100'fo times MFLPO, provided that the adjusted APRM reading does not exceed 100~ of RATED

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THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

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WASHINGTON NUCLEAR - UNIT 2 3/4 2-5

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

Greater than or equal to the applicable MCPR 1jpt~dptermined from goo a.

P ag Table 3.2.3"1 during steady state operation at"rate core flow, PLRgTsoa Ot" 4VWZW Zu >ZrubC.& LOOP ogZrCA~rO~~ 0 R

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b. Greater Chan or equal to the greater of the two values determined from<Tabl,e 3.2.3-1 and Figure 3.2.3"1 during steady state operation at e440 than rated core flow~ su~a~ l~ rc o Rzcl~cu~a~iov LOOP Psm' A T'r O N.

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

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

II coNT4Q44".-A. co~~

MCPR OPERATING LIMITS MCPR Operating Limit U to 106K Core Flow Cycle Equi pment

~Ex asure Status GE Fuel ANF Fuel

1. 0 MWD - 4150 MWD 1. 29 1. 26 MMU ~MU

2. 4150 MWD

" EOC MWO Normal scram times*" 1. 32 1. 30 HTU HK

3. 4150 MWD

- EOC MWD Control rod insertion l. 39 1. 35

$ 1TU ~MU bounded by Tech. Spec.

limits (3.1.3.4-p 3/4 1"7) 4150 MWD

" EOC MWD RPT inoperable l. 37 1. 35 NMU NMU Normal scram times

5. 4150 MWD " EOC MWO RPT inoperable 1.43 l. 39 HHU 'MU Control rod insertion bounded by Tech. Spec.

limits (3.1.3.4-p 3/4 1-7)

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$ c'p,AH Y'sHES 95'o~Ac.

~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 1-7),

and the scram insertion time must. meet the requirements of Tech. Spec.

3.1.3.4.

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

Notch 45 .404 Notch 39 .660 Notch 25 1.504 Notch 5 2.624 WASHINGTON NUCLEAR - UNIT 2 3/4 2-7 Amendment No. 45

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1.0 20 30 40 50 60 70 BO 90 100 Total Core Recirculating Flow (% Bated) 11('educed Flow MCPB Operating Limit Figure 3.2.3-1

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POWER DISTRIBUTION LIMITS 3/4. 2. 6 POWER/FLOW INSTABILITY ITING 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 454 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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HINGTON NUCLEAR -. UNIT 2

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INSTRUMENTATION 4.2.7 NEUTRON FLUX NOISE MONITORING ITING CONDITION FOR OPERATION 3.2.7 The APRH and LPRH neutron flux noise levels shall not exceed three (3) times their established baseline values when operating in the region of APPLICABILITY.

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 455 of rated total core flow, or with one reactor coolant system recirculation loop not in operation.

ACTION:

a. I f base ine APRM and LPRM neutron f lux noi se 1 evel s have not been 1

established for the appropriate reactor coolant system condition (one or two loop operation) since the most recent CORE ALTERATION, then:

Within 2 hours exit the region of APPLICABILITY. Establish baseline APRH and LPRM neutron flux noise levels prior to re-entering Region 8 of Figure 3.2.7-1.

b If baseline APRH 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 APRH 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to below the region of APPLICABILITY within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

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

4.2.7.2 The APRH and LPRH 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 hours, and

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

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

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l. AOD BLOCK HOIIITOII

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Upscale

< 0.66 W + 40X < 0.66 W + 43X kh~SAX

b. I t nope ra i ve tl.A. tl.A.

c. Downsca1 e > 5X of RATED TIIERHAL POWER > 3X of RATED TIIEAHAL POWER

2. APAH

~ a. Flow Biased Heul.ron Flux Upscale b.

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Inoperative

< 0.66 H. A.

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0.66 W + 45X*

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"c.'ownscale > 5X of Rg~ EAHAL POWER > 3X of RATED TIIERHAL POWER

d. Neutron Flux - Upscale, Startup < 12X MPD TIIERHAL POWEA < 14X of RATED TIIEAHAL POWER

3. SOURCE AhtlGE HOtllTORS

a. Detector not full in H.A.

1.6 x 10 5 cps

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b. Upscale <

C. Inoperative l.A. tl.A.

d. Downscale > 0.5 cps

4. INTERMEDIATE AAtlGE HONITOAS

a. Detector not full in tl.A. tl A.

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b. Upscale < 100/125 divisions of full scale < 110/125 divisions of full scale

c. Inoperative tl.A. H.A.

d. Downscale > 5/125 divisions of full scale > 3/125 divisions of full scale

5. SCAAH DISCIIAAGE VOLUHE

a. Water Level-Illgh < 527 ft 2 in. elevation < 527 ft 4 in. elevation

b. Scram Trip Bypass H.A. N.A.

6. REACTOR COOLAtlT SYSTEtt RECIRCULATION FLOW

a. Upscale < iOB/125 divisions < ill/125 divisions of full scale of full scale

b. Inoperative tl.A. tl.A.

c. Compara tor < lOX flow deviation < llX flow deviation The Avera'ge Power Range Honitor rod block function is varied as a function of recirculation loop flow (W). Tl~e trip setting of this function must be maintained in accordance with Specification 3.2.2.

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CONTROLLED COPY INSTRUMENTATION

/4. 3. 10

~ NEUTRON FLUX MONITORING INSTRUMENTATION iQ

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L TING CONDITION FOR OPERATION 3 .3. 10 The APRM and LPRM" neutron flux noise levels shall not exceed t ree (3) times their established baseline values when operating in the allowab region of Figure 3.3.10-1.

APPLICABILI7h OPERATIONAL CONDITION 1 with two reactor coolant ystem recir-p hi p ti ith TEERRRL PtlflER p t th th in Figure 3.3.10-1 and total core flow less than 45K of rate total core flow ii it p iii d or with one reactqr coolant system recirculation loop not operation with THERMAL POWER greater than the limit specified in Figure .3.10-1.

a. With the APRM or LPRM" neutron flux noise level greater than three (3) times their esta lished baseline noiseglevels, initiate corrective action within 15 minu es to restore the praise levels to within the re-quired limits within 2 ours or reduce THERMAL POHER to 'less than or equal to the limit specs ied in Figur 3.3. 10-1 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

'b. ~ With reactor power/core flqw in t e crosshatched region of Figure 3.3.10-1, initiate corrective action withi

~ ~ ~ ~ ~

15 minutes to reduce power by control rod insertion to a reactor power/core flow below the crosshatched

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region within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

~

SURYEILLANCE RE UIREMENTS 4 .3.10.1 The provisio of Specification .0.4 are not applicable.

4.3. 10.2 With twoyeactor coolant system re irculation loops in operation, establish a baselPne APRM and LPRM" neutron flux noise level value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months upon en ering the APPLICABLE OPERATIONA CONDITION of Specifica-tion 3.3.10 pr vided that baselining has not bee performed since the most recent CORE TERATION.

4.3.10.3 With one reactor coolant system recirculation loop not in operation, establi h a baseline APRM and LPRM" neutron flux noise vel value with ess than or equal to the limit specified in Figure .3. 10-1 prior to THERMAL'OWER ent ing the APPLICABLE OPERATIONAL CONDITION of Specificat n 3.3.10 provided bgelining has not been performed with one reactor coolant sy tern recirculation loop not in operation since the most recent CORE ALTERATION.¹ WASHINGTON NUCLEAR - UNIT 2 3/4 3-102 Amendment No. 45

'I 4'NSTRUMENTATION CONTROLLED COPY NEUTRON FLUX, MONITORING INSTRUMENTATION JEILLANCE RE UIREMENTS Continued) 4.3.10.4 The APRM and LPRM" neutron flux noise levels shall be determined to be 1'ess'han or equal to the limit of Specification 3.3.10 and the rea/or power/core flow shall be verified to lie outside the crosshatched re on of Figure 3.3.10-1 when operating within the APPLICABLE OPERATIONAL C DITION of Specification 3.3.10:

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

b. Mithvq 30 minutes after completion of a THERMAL P ER increase of at least 5X of RATED THERMAL POMER.

"Detector levels A and C of one LPRM string per core octant plus detector levels A nd C of one LPRM string in the center of the core should be monitor d.

dThe aseline data obtained in Specification 4.3.30.3 is apPlicable to opera-ti n with one reactor coolant system recirculation loop not iq operation and T ERMAL POWER greater than the limits specified in Figure 3.3.X -1.

SHINGTON NUCLEAR - UNIT 2 3/4 3-103 Amendment No. 45

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3/4. 4. 1 RECIRCULATION SYSTEM CIRCULATION LOOPS MITING CONDITION FOR OPERATION

3. 4. 1..1 Two reactor coolant system recirculation l.oops shall be in operation.

APPLICABILITY: OPERATIONAL CONOITIONS 1~ and 2~.

ACTION:

a. Mith one reactor coolant system recirculation loop not in operation:

S~~+

1 Mithin 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a) Place the recirculation flow control system in the Local Manual (Position Control) mode, and b) The T AL POWER shal e less than or ual to t limit specific n Figure 3.4.1. -1 or the provi ns of Sp fi" cation 4.3. 3 are satisfie . Mith one reac coolant system recircu tion loop not i peration and w h THERMAL POWER greater tha the limit speci ed in Figure 3. .1.1-1, d the provisions o Specification 4. .10.3 having no bee satisfied, initiat action within 1 inutes to redu THERMA OMER to less than equal to the it specified in Figure 4. 1.1-1 within 4 rs. The provi ns of Specification 3. 10.3 must be sa 'ied prior to suming p er operation a e the limit specs ed in Figure . . l. 1-1.

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and, PoNR, 6 EKIEKAff E LEG T'SPIC F VC I Reduce the Max> um Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.84 times the two recirculation loop operation limit per Specification 3.2.1, and, duce the verage Power Rang Monitor (Applho Scram and Ro Block an od Bloc onitor .r'etpoin&~and Al lo-able lues to ose appl> ble for ngle recirculation loop op ation per ecificat' 2.2.1, 3.2.2, an 3.3.6.

Pf 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 100M core flow at 100 THERMAL POWER. This value was determined during the Startup Test Program.

WASHINGTON NUCLEAR - UNIT 2 3/4 4"1 Amendment No. 16

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INSERT

1. Within 15 minutes:

Verify that core flow is 39% 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 39% of rated core flow 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 39% of rated core flow within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

IU GGNTRGLm.EB GGPY REACTOR COOLANT SYSTEM ITING CONDITION FOR OPERATION Continued ACTION: (Continued)

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 < 10Ã""" of rated loop flow.

Reduce recirculation loop flow in the operating loop until

+ the core plate M noise does not deviate from the estab-lished core plate EP noise patterns by more than 100Ã.

i) With o reactor coola t system rec'ulation loo ot i operation d THERMAL PO greater th the limit s ci" fied in Figur .4.1.1-1 an ore flow les than 39~ o rated core flow, 'tiate aetio within 15 mz tes to redu ERMAL POWER to les han or equ to the limit ecified in . 3.4.1.1"1 or inc ase core f to greater n or equal 39K of rated core low within 4 urs.

3 The provisions of Specification 3.0.4 are not applicable.

A. ~ Otherwise, be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months verify that:

a. The recirculation flow control system 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 100K 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. 16

~ ~

CONTROLLED COPY 3/4. 2 POWER DISTRIBUTION LIMITS SES 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 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 uhich results in a calculated LOCA PCT much less than 2200'F. The Technical Specification APLHGR for ANF fuel is specified to assure the eadem-ee PCT following a will not exceed the 2200 F limit. The limiting value for APLHGRpostulated LOCA is shown OKER'~2 Figures 3. 2.1-1, 3 '. 1-2, and 3. 2. 1-3 for two recirculation loop operation.

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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. l.

WASHINGTON NUCLEAR " UNIT 2 B 3/4 2-1 Amendment No. 45

J POWER OISTRIBUTION LIMITS ASES

.2.6 POWER/FLOW INSTABILITY P

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., powershape, 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 Oesign 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 instabi lity constituted calculated decay ratios of .8 and greater by e GE methodology and .75 and greater by the EXXON methodology.

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

l. 805 rod line 2.'5K

~

core flow line

3. APRH rod block line minus 3g power

4. Natural Ci rcul ati on flow line

5. Hinimum Forced Circulation for normal recirculation lineup.

This division conforms to the SIL 380 recommendations with a 3X power penalty on the APRH rod block line. For LCO 3.2.6, the region of concern is bounded by the APRH rod block line, minus 3g power, the natural circulation flow line, and the 45K core flow line. Calculated decay ratios between the two flow lines and on the APRH rod block line minus 3g must be less than .9. Operation in the region between the two flow lines and above the rod block line minus 3/

is forbidden due to the potential for boiling instabilities. ~

1 For the ease of annual licensing submittals, a 3/ margin from the rod block line is taken to avail the opportunity to submit with no Technical Specifica-ion changes under the provisions of 10CFR50.59. This 3g provides margin to ure that vendor stability calculations can easily support the allowable rating region. For calculational ease the power boundary is linearized between two points, (241 Flow, 39%%d Power) and (45/ Flow,'25 Power).

L I ~

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IHSTRUMENTATIOH ASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING IHSTRUMENTATION 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 OOCM 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 HOLOUP-SYSTEM. The OPERABILITY and use of this instrumgntation is consistent with the requirements of General Oesign Criteria 60, 63~%ad 64 of Appendix A to 10 CFR Part 50.

3/4. 3. 8 TURBINE OVERSPEEO PROTECTION SYSTEM This specification is provided to re that the turbine overspeed protection system instrumentation and turbine speed control valves are OPERABLE and will protect the turbin om excessive overspeed. Protection from turbine excessive overspeed i uir ed since excessive overspeed of the turbine could generate potentiall maging missiles which could impact and amage safety-related component uipment or structures.

3/4. 3. 9 FEEOWATER SYSTEM/MA RBINE TRIP SYSTEM ACTUATION IHSTRUMENTATION 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.

X3/4 3. NEU N FL MONIT NG IN UMENTA H i At th high p er/lo flow c er of e opera ng doma' a sma prob-abill tion's of ls it cyc neut flux ci liat ns exis depend on corn na-o operat' con tions . g., r patter , power hape). o pr ovi a urance that ne tron f limi cycle o cillati s are d ected a sup-pressed, A and L M neut n flu oise els sho d be mo 'tored w 'le opera ing in is reg n.

Stab lity te s at o rating WRs we revie to det mine a eric regs n of t power+low ma in whi shoul be per rmed. > cons~. ative ay rat'f surveys lance o neutron or dete mining the gen ic .regs n for s veilla e to ac unt for

0. was chas lux nois levels as the e plant ses o

p nt var> bility f deca ratio th core nd fue designs. This ge ric reg' has b en-det mined corres nd to core f of le than or qual to 4 of rate core low and therma power eater tI n that ecified

.4.1. 1- (Refe ce). 'igure WASHINGTON NUCLEAR " UNIT 2 B 3/4 3"7 Amendment Ho. 36

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INSTRUMEHTATIOH A S ONITOfHHG IHSTRUMENTA'TION (C tinue 3/4. - 7. 11XRAD IOACTIVE GASEOUS E LUENT NITORI INSTRUAE TATIO Th radioaqtive gaseous ef fluent strume tation provid to mon'tor a control, as applicable, the +leases f radar ctive terials gaseo eff ents haring a'ctual or~potenti'a$ relea s of g eous e uents. he ala /

trip tpoints, for these instruments+hall b calcul~ ed and justed e rdanc with the methodology aqd param ters i the OD to ens e that e

'c-ala m/tri ,will occur prior to exceeding he lim s of 10 FR Par 0. Thi inst mentatqon al~ inclu'des provisions f monit ing and ontrol ' the ncen ationgof potgntialky, explosive gas 'xtures 'n the M TE GAS LOUP S TEM. he OPE)ABILIT+ and use of this, instr ntatio is cons'ent w h the requ*remen g of Gh eral 5 sign +iteria 5, 63~ 64 o ppendix to 10 R 3/ 3.8 RBIH OVERS ED PRO ECTION SYSTE T s spe ificat on is rovide to re that /he turbize over eed p otecti syst inst ument ion. an t bine speed contr 1 valve are OP ABLE a d wilKprote the rbin om ex essive ovqrspeed. Protec on from urbioe excesdiye ove speed 'qu'red s> ce excesb,'ve overs eed of e urbin could eneraWe pote iall magin miss> es which ould im ct and mage s ety-r ated mpone 'pment r str tures.

3/4. .9 FEE MATER YSTEM The feedwa r sys m/maie turbin trip stem tuati instrum tation provid to ini iate e fee ater s tern/m in tur 'ne tr'ystem -the eve t of re tor ve el wa r lev equa to or eater han t level 8 setpo' asso ated w th a edwate contro ler fa lure.

a.. 7 I 3/4.~

pressedd, NEUTRO FLUX 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 that neutron flux limit cycle oscillations are detected and sup- 'ssurance APRM and LPRM neutron flux noise levels shou1d be monitored while operating in this region.

6 i~

Stability tests at operating BMRs were reviewed o determine a generic.

region of the power/flow map in which surveillance o eutron flux noise levels should be performed. A cons~rvative decay ratio of was chosen as the bases 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 igure 3.4.1. l-l (Reference).

MASHINGTON NUCLEAR - UNIT 2 B 3/4 3-7 Amendment Ho. 36

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INSTRUMENTATION BASES k

MONITORING INSTRUMENTATION (Continued)

Pal S~

NEUTRON FLUX MONITORING (Continued)

Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux osci llations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron flux noise levels of 1-12 of rated power (peak-to-peak) have been reported for the range of low to high recirculation loop flow during both single and dual 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~los are sufficient to ensure early detection of limit cycle neutron flux os~ tions.

Typically, neutron flux n levels show a gradual increase in absolute magnitude as core flow is incrrea (constant control rod patt ~n) with two reactor recirculation loops in op ion. Therefore, the b: '.ne neutron flux noise level obtained at a specific flow can be. applied over a range of core flows. To maintain a reasonabl ~iation between the low flow and high flow ends of the flow range, the rangeMer which a specific baseline is applied

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should not exceed 20~ of rated core flow with two recirculation loops in opera" tion. Data from tests and operating plan dicate that a range of 20~ of rated

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core flow will result in approximately a 5

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crease in neutron flux noise

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level during operation with two recirculatio

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ps. Baseline data should be

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taken near the maximum rod line at which the gority of operation will occur.

However, baseline data taken at lower rod lines (i.esa 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 interva1s which correspond to less than a 50 in-crease in APRM neutron f'lux 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 flaw SLQ 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. l. 1-1. This will result in approximately a 25. 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 3-7a Amendment Ho. 16

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3/4.4 REACTOR COOLANT SYSTEM BASLS 3/4. 4. 1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and been found to be acceptabl '

' , provided the unit is operated in accordance with the single recirculation loop operation Technic I 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 coas;down from either recirculation loop following a LOCA. 'Ahere the recircula--

tion loop flow mismatch limits cannot be maintained during two recirculation loop operation, continued operation is permit ed in the single recirculation loop operation mode.

in order to prevent undue stress on the vessel nozzles and bottom he"d region, the recirculation loop temper'atures shall be witn n 50 F of each ot.:er prior to s artup of an idle loop. The loop temperature must also be within 50'F of the ",eactor pressure vessel coolant temperature to prevent thermal

'hock to ',he 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 s.ress on the vessel would result

~

temperature difference was greater than 145~F.

if the 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 lid of design pressure (1250 (design) X 1. 10 = '375 psig maximum) under upset condi tions. 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 110Ã of the design pressure.

The safety valve sizing evaluation assumes credit for operation of the scram protective system which may be trioped by one of two sources; i.e., a dir ect 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 clos;!>g, and follow-ing 10~ travel of full stroke. The pressure switches are actuated wnen a fast closure of the control valves is initiated. Further, no credit is taKen for power ooera ion of the pressure relieving devices. Credit is only'aken f ;,

WASHINGTON NUCLEAR - UNIT 2 B 3/4 4-1 Amendment No. 38

VANCEDNUCLEAR FUELS CORPORATION 00 106Lh AVENUE NE, PO BOX 90777, BELLEVUE. WA 9600MTTT 1206) 453.4300 February 22, 1988 g - elm/ge JBE:067:88 ANFWP-88-0020 Washington Public Power Supply System 3000 George Washington Way P. 0. Box 968 Richland, WA 98669-0968 Attn: Manager, Central Contracts Gentlemen:

In response to a telephone request by the Supply System on February 4, 1988, the single loop control rod withdrawal error results have been revised and are enclosed.

Very truly yours, J. B. Edgar an. ~

Contract Administrator tlm Enclosure

WNP-2 SINGLE LOOP CONTROL ROD WITHDRAWAL ERROR (REVISED)

The limiting Cycle 3 control rod withdrawal case has been rerun at the single loop conditions provided by the Supply System in Reference 1. The single loop power and flow conditions for the analysis were 2492 MWt (75%) and 57.8 Mlb/hr (53.3%). The initial control rod pattern for the analysis is shown on Figure

5. 1 of the Cycle 3 reload analysis report XN-NF-87-25. The calculated Cycle 3 hCPR and CPR values for single loop and'two loop are as follows:

Sin le Loo Result Rod Block Distance Sin le Loo ACPR Sin le Loo CPR Monitor Settin Withdrawn ft ANF GE ANF 0.0 1.544 1.819 106% 5.0 0.22 0.26 1.326 1.564 107% 5 ' 0.22 0.26 1.326 1.564 108% 6.0 0.24 0.29 1.302 1,526 Two Loo Result (from XN-NF-87-25)

Rod Block Distance Two Loo ACPR Two Loo CPR onitor Settin Withdrawn ft GE 0.0 1.369 1.605 106%* 4.5 0.20 0.23 1.173 1.376 107% 4.5 0.20 0.23 1.173 1.376 108% 5.0 0.22 0.25 1.154 1.352

The Cycle 3 setting is 106% and the CRWE based MCPR operating limit is 1.26 for the ANF fuel.

The above reported single loop control rod withdrawal error (CRWE) calculation was performed to demonstrate that the 1.35 CPR value used to initialize the single loop ECCS analysis could be used as the single loop CPR limit. The single loop CRWE results presented in this letter show that the single loop CPR values at the reduced power and flow conditions are significantly higher than the two loop CPR values (more margin to the CPR safety limit). The single loop hCPR values are slightly larger than the two loop hCPR values as expected for the high starting CPR values. Based on the ANF experience with the CRWE analysis, a lower starting CPR results in a smaller calculated ACPR value.

The original basis for the flow dependent

~

CPR limit for the two loop operation is the pump runup event. In the single loop configuration, however, the additional constraint of the reduced flow MCPR operating limits is no longer required (Reference 2). A two pump flow runup is not possible as the pump in

0 ~,iV *4 4. ~ 4I I ~ '.'t %8 'APL ~ * .'4J ~ 4 ~ >> h A ~ i" * % 'A 0 idle loop is not running. An inadvertent start of the idle pump cannot

~

he ~

ffect flow appreciably as the pump is interlocked to prevent starting unless

~ ~

its associated flow control valve is at the minimum position (Reference 3).

~ ~

~

For future operation, a constant single loop CPR limit of 1.35 is considered adequate for the single loop operational mode. This single loop limit is related to the two loop CRWE limit being equal to or less than 1.26. If the two loop CPR limit based on CRWE for the limiting fuel type is greater than 1.26, a cycle specific review of the single loop CPR limit may be required.

REFERENCES

1) Letter WPANF-28-87-0101 from RA Vopa]ensky to JB Edgar dated November 20, 1987.

2) JE Krajicek, "WNP-2 Single Loop Operation Analysis", ANF-87-119, Advanced Nuclear Fuels Corporation, Richland, WA 99352, September 1987.

3) WNP-2 FSAR, Chapter 4, Section 4.4.3.3.3, pages 4.4-5 and 4.4-6 (Design Features for Power Flow Control).

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ML17279A919

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SUMMARY

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