ML17285B402

Proposed Tech Specs,Revising Sections 1.0,3/4.2.1,3/4.2.3, 3/4.2.3,3/4.4.1,B3/4.2.1,B3/4.2.3 & 6.9 by Implementing Generic Ltr 88-16, Removal of Cycle-Specific Parameter Limits from Tech Specs

ML17285B402
Person / Time
Site:	Columbia
Issue date:	08/02/1990
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17285B401	List: GO2-90-132, Application for Amend to License NPF-21,implementing Generic Ltr 88-16, Removal of Cycle-Specific Parameter Limits from Tech Specs. Amend Revises Tech Specs 1.0,3/4.2.1,3/4.2.3, 3/4.2.4,3/4.4.1,B3/4.2.1,B3/4.2.3 & 6.9 ML17285B402
References
GL-88-16, NUDOCS 9008060171
Download: ML17285B402 (90)

Text

ATTACHMENT I PROPOSED TECH.

SPEC.

CHANGES CORE OPERATING LIMITS REPORT:

CYCLE 6 Qi.

(

9008060271 900802 PDR iADOCK 05000397 P

PDC

CONTROLLED COPY INDEX DEFINITIONS SECTION

1. 0 DEFINITIONS

1. 1 ACTION.

l. 2 AVERAGE 8UNDLE EXPOSURE PAGE 0

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~ l-l 1.3 AVERAGE PLANAR EXPOSURE 1-1 1.4 AVERAGE PLAHAR LINEAR HEAT GENERATION RATE.................

1-1

1. 5 CHANNEL CALIBRATIOH.

1. 6 CHANNEL CHECK.

1.7 CHANNEL FUNCTIONAL TEST 1 ~ 8 CORE ALTERATIOH.

t, $Q Copra gPg PCA 7IAJS l-IPfl7 S 1.9 CRITICAL POWER RATIO 1.10 DOSE E(UIVALEHT I-131.

1.11 E-AVERAGE DISINTEGRATION ENERGY

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1-2

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1-2 1-2

1. 12 EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE TIME.........

1-2

1. 12-A END-OF-CYCLE (EOC).

1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME..

1-3 1.13-A FINAL FEEDWATER TEMPERATURE REDUCTION (FFTR).............

1-3 1.14 FRACTION OF LIMITING POWER DENSITY

1. 15 FRACTION OF RATED THERMAL POWER.

1.16 FREQUENCY NOTATION.

1. 17 GASEOUS RADWASTE TREATMENT SYSTEM.

1. 18 IDENTIFIED LEAKAGE.

1.19 ISOLATION SYSTEM RESPONSE TIME..

1.20 LIMITING CONTROL ROD PATTERN.

1 21 LINEAR HEAT GENERATION RATE 1-3

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1 3 1-4 1-4 WASHINGTON NUCLEAR UNIT 2 Amendment No 77

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CONTROLLED COPY INDEX ADMIHISTRATIVE CONTROLS SECTIDN I

!. CORPORATE HUCL'E'AR'"'SAFETY REYIB( BOARD (Continued)..

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o ~ e ~ oe ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~

CO MEETING FRE)UENCY....................................

PAGE 6-11 6-11 UORUMe ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ o o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ o ~ o ~ o ~ ~ e 6

EYIEk o ~ o o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ e o ~ ~ ~ o ~ o e ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o o ~

R UDITSe ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ o o ~ ~ o o o ~ o e ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~

A CORDS ~ o ~ ~ ~ ~ ~ ~

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RE

~Q 6-12 6-13 6.6 REPORTABLE OCCURREHCE ACTION..............................

6-13 6.7 SAFETY LIMITVIOLATIOH......................

6-14 6.8 PROCEDURES AHD PROGRAMS.........................,.....,...

6-14

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6. 4 REPORTING RE UIREMENTS...........'.........................

6:9.1 ROUTINE REPORTS AND REPORTABl >> OCCURRENCES...........

STARTUP REPORT.........'...-"--"-.-..--..............-

ANNUAL.REPORTS............

MONTHLY OPERATING REPORTS............................

REPORTABLE OCCURENCES...............,................

6-16 6-16 6"16 6-17 6-17 PROMPT NOTIFICATION MITH '!WRITTEN FOLLMJP............

6-17 6.9.2

6. 10

6. 11 THIRTY DAY WRITTEN REPORTS...............,...........

ANNUAL RADIOLOGICAL EHYIROHMENTAL OPERATING REPORT...

SEMIAHHUAL RADIOACTIVE EFFLUENT RELEASE REPORT.......

SPECIAL REPORTS........-.............................

C >Ra oPsapwi~6

<itive RBPoa~

RECORD RETEHTIOHo ~ o ~ o o ~ o o o ~ ~ ~ o ~ ~ ~ ~ ~ o ~ e ~ ~ o ~ e ~ ~ ~ e ~ o ~ ~

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RADIATIOH PROTECTIOH PROGRAM.............................

6-19 6-20 6-21 6.12 HIGH RADIATIOH AREA.....................................,

6e13 PROCESS CONTROL PROGRAM. ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ oo ~ ~ ~ ~ ~ ~ ~e ~ ~ ~ ~ ~ ~ ~

6-24 6-25 6.14 OFFSITE DOSE CALCULATIOHMAHUAL....o.-..".-".."-...... 6-25 ke15 MAJOR CHANGES TO RADIOACTIYE LI UID GASEOUS AHD SOLID WASTE TREATMENT SYSTEMS.. -..... ~ - ~ ~ ~. ~. -. ~ ~. ~. ~.....6-26 WASHINGTON NUCLEAR - UNIT 2 xIx heendment Ho.

16

CONTROLLED COPY LIST OF FIGURES INDEX 3.1. 5-2 PAGE SODIUM PEHTABORATE SOL'UTIOH SATURATIOH TEMPERATURE...

3/4 1-21 y,l SODIUM PENT%BORATE TANK, VOLUME VERSUS COHCEHTRATIOH Rf(UIREMENTS.......................................;.

3/4 1-22 oWRoootAI ~ oiHI

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o MASHINGTON NUCLEAR - UNIT 2 Amendment No.

B4

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LIST OF FIGURES t":

F1GURE CONTROLLED COPY IHDEX 4EA~EAAxEHHb8&HHAT~BH St PAGE 3.2.6-1 3.2.7-1 3.2. 8-1 3.4.1. 1-1 3.4.6. 1"1 3.4.6.1-2

4. 7-1

3. 9. 7-1 6,

B 3/4 3-1 HEIGHT ABOVE SFP WATER LEVEL VS.

MAXIMUM LOAD TO BE CARRIED OVER SFP.....................................

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

B 3/4 3-8 OPERATING REGION LIMITS OF SPEC. 3.2.6................

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

3/4 2-14 OPERATING REGION LIMITS OF SPEC. 3.2.8...............

3/4 2-16 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 SAMPI E PLAN 2)

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

3/4 7-15 B 3/4.4.6"1 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE...

B 3/4 4-7

5. 1-1

5. 1"2

5. 1"3 EXCLUSION AREA BOUNDARY.............................

5-2 LOW POPULATION ZONE..................................

5-3 UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND, LIQUID EFFLUENTS............. '-4 (f

WASHINGTON NUCLEAR - UNIT 2 xx(a)

Amendment No.

84

.00NTRQLLEB GGPY INDEX LIST OF TABLES TABLE 1.2 SURVEILLANCE FREQUENCY NOTATION OPERATIONAL CONDITIONS PAGE 1-9 1-10

2. 2. 1-1

82. l. 2-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS..

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

3. 3. 1-1

3. 3. 1-2

4. 3. l. 1-1

3. 3. 2-1 3

3 2

2

3. 3. 2-3 REACTOR PROTECTION SYSTEM INSTRUMENTATION............

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

3/4 3"6 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ISOLATION ACTUATION INSTRUMENTATION.

3/4 3-7 3/4 3"12 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS....'....

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

3/4 3-19

4. 3. 2. 1-1

3. 3. 3"1 ISOLATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..........

3/4 3-22 3/4 3-26

3. 3. 3" 3

4. 3. 3. 1"1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES....

3/4 3-30 3/4 3-33 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............

3/4 3-34

3. 3. 4. 1-1

3. 3. 4. 1-2 ATMS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION...

ASS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS 3/4 3-38 3/4 3-39

4. 3. 4. 1" 1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............

3/4 3-40 MASHINGTON NUCLEAR - UNIT 2 XX1 Amendment No.

CGNTRGLLED COPY DEFINITIONS CHANNEL FUHCTIOHAL TEST 1.7 A CHANNEL FUNCTIONAL TEST shall:be:

a.

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

b.

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

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

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

Sus-ension of CORE ALTERATIOHS shall not preclude completion of the movement f a component to a safe conservative position.

M$HljL CRIT AL POWER RATIO The CRITICAL POWER RATIO (CPR) shall be that power in the assembly which is calculated by application of the appropriate critical power correla-tion to cause some point in the assembly to experience boiling transition divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.10 DOSE E(UIVALEHT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I"131, I-I32, I-I33, I-l34, and I-135 actually present.

"'he thyroid dose conversion factors used for this calculation shall be those listed in Table !II of TI0-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

f-AVERAGE DISINTEGRATION EHERGY 1.11 Z shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coo'lant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in HeV, for isotopes, with half-lives greater than 15 minutes, making up at least 95Ã of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM ECCS RESPOHSE TIME 1.'2 The-EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE

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

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

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

WASHINGTON NUCLEAR - UNIT 2 1-2 Amendment Ho. 84

INSERT A CORE OPERATING LIMITS REPORT 1.8A The CORE OPERATING LIMITS REPORT is the HNP-2 specific document that provides CORE OPERATING LIMITS for the current operating reload cycle.

These cycle-specific CORE OPERATING LIMITS shall be determined for, each reload cycle in accordance with Specification 6.9.3.

Plant operation within these Operating Limits is addressed in individual specifications.

CONTROLLED COPY l

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION gpgccgIwJ 3.2.1 All AVE GE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel fuel shall not exceed the limits

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APPLICABIL!TY:

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

ACTION:

~p4L~pi~

4imiss 8e Pn v<,

With an APLHGR exceeding the limits of

, initiate cor-

,rective action within 15 minutes and restore APLHGR to within the required

,limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits de-termined from

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> ~ ~ e lip f +aq ~ 1( ~ >>, na 1 s II V 13.0 12.5-12.7. 'I2.0 12.9 l2.7 12.0-l 11.5-a) c 11.0 0 QJ) EQ < o 105-E e DQ .E 10 0-X 9.5-c 9.0 0.0-12.1 12.1 12.7 12.8 12.9 . 12.7 11.7 10.0 10.0 9. I 1,102 5,512 11,023 16,535 22,046 27,55 3, 69 30,501 44,093 Ordinate Abscissa .7 10.0 10.0-9.4 5,000 1o,noo 15,ooo 2o,ooo 25,ooo 3o,ooo 35,ooo 4o,ooo Average Planar Exposure (MND/MT) Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Type OCR103 Figure 3.2.1-1 860599.4A \\ 9 lA 0D CO Il t4 M CD 13.0 ) Cl Ol C Q O ct h E e a (9 E X oX 6$4lC 11.5 11.0 10.5-10.0 9.5 9.0 8.0 f 12.0 Ordinate A lssa 12.0 '1,102 12.1 5,512 12.2 11,023 12.2 16,535 12.1 22,046 11.6 27,558 11.2 '3,069 10.6 38,581 10.0 44;093 12.2 .11.2 10.6 A0 10.0 Q IT1 U. O0 ..a 0 00 0 -5 0 'f0)000 15,000 20,000 '5,0 Average Planar Exposure (MWD/MT) Maximum Average Planar Linear Heat Generation Bate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Type BCR233 Figure 3.2.1-2 30,000 35)000 40,000 60599.5A Cl 0 z A 18.0 12.5 12.0 g o 11.5 g,~ 11.0 m o 10.5 C E~o 10.0 05 + 9.5 9.0 8.5 0 5,000 10,000 1 5,000'0,000 25,00 30 0 ,000 13.0 1 3.0 1 3.0 1 3.0 11.3 9.4 7.9 Bundle Average Exposure MAPLHGR (MwD/MT) kw/ft Two Loop a Single Loop eration A0 O O0 O C7l 0 5,000 10 000 15<000 20 000 25 000 30<000 35<0 Bundle Average Exposure (MWD/MT) ANF Sx8 Reload Fuel Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure 40,000 Figure 3.2.1-3 .=ig=l~ ~-'--"'gRY4'" ~lV~~%i:i:-=0: WWWNQiWQWWWMWWWWWgRgWWWR5; WMW~WCWN~WMMWWWAMW M WWWNMR~MWWWMWWWWN~M WMWWWa 'MWCWWNWNMWWWWM MNMWWWNMALWAgWNWAAWNWNAAMAMMA'%PiWCWWNWNNAWWMM MNMWWAWMWNA A WNWAWWW AWWMW >MWMWWWW'NWMWWMWMM ~ MNMW+Ag+WH%A ANAAggC WAAMg~kWi4NWMAWWWMMACAHg +RNMWWWWMCMWWNNWRWWMWWWQWQWM&OMAHA MA/MANA/M MWW WNWWRWWM'5&iHWWCCMWWCWMWWWCCWLa+WWWMM P I o o AAMWNWWNN>-QWAWWWNWWNWNWWWWNANA RWAMM M ANMWMWWMV~a'.OAWWWNNANWMOAXANNMWaWAMN WNMANWW~&iQWA~WM WNWgMWMWWWQMNMQWA~~MM gM WM MANWMNWMCNWWMWWMMNMWWMHMMAMWMAMA%% I, ~ MN~AWMWWMMWWWNQWMHAM1NNWWAWAMNWWAMN MFQNWMWWMMWAWMWWMMI+MWNWWWWCWMMWWWMM WM NANWNANNNWNANWWMN&NWWWWWMWWMNWWNMM Q .QOAAANAANNAAANAANNAKAAAAAARAQAAAHQN R I 'AAQANWRAAAAARAAHRARA>HAWAAAARAAAARA ~ M ~NMMWMWWMNWWWWMgMMWMhLWWWWMWWMWWWWMM MMNWMWMMPAMWWMMNMMMMMCNWMMMMWMWM~AMWMWMMMWWWMM QSSAAAAY>AMXOROAL'RNCCNAOASiMNAAAAA%AAAAMAAWQNM RNRAAAKAAQAAAHANAAAN~NAAANANNAWANARAWNANAAA KM I I I I I III I II I III III I I I I I I I I I I I ~ I I I ~ I ' 0 10.0 9.5 10.08 10.25-' 10.25 10.16 9.74-0$ ~ C C5 dlGl C Q 0 L) t5 + m E S D C5 OS XgK t5 C 9.0 8.5 8.0 7.5 7.0 6.5 10.08 '10.16 10.25 10.25 10.16 9.74 9.41 8.90 8.40 1,102 5,512 11,023 16,535 22,046 ~ 27,SSB 33,069 38,58 4, 93 Ordinate'bscissa 8.90 8.40 6.0 0 5,000 0 000 35,000 0,00 10,000.. 15,000 20,000 25,000 3, ~ Average Planar'Exposure (MWD/MT) 0 45,000 SINGLE LOOP Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure initial Core Fuel Type 8CR233 Figure 3.2.1-5 12.0 11.5 ~ ~ ~ ~ C 11.0 10.5 10.0 9.5 ClOl O Pu t4 9.0 8.5 E ca 8.0 L C 7.5 7.0 Bundlo Avorago Expo sumo LMMiHIl 0 5,000 10,000 15>000 20,000 25,000 30>000 35,000 MAPLHGR 11.2 11.2 11.2 11.2 11.2 8.1 6.8 6.0 0 5,000 10,000 15,000 20,000 25>000 30,000 35>000 Bundle Average Exposure (MWD/MT) ANF 9 X 9 - IX AND 9 X 9 - 9X Reload Fuel Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure Figure 3.2.1-6 0 p V 0 Ql I Al Qrc< ro~ 7 c ro ~ CL P ro o lg E ~~ 5 ~ 4 ~ ~ C ~ ~ ~ ~ ~ ~ ~ Bundle Avera Ex sure D/MTU) ,0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 MAPLHGR (Ilw/fl) 8.90 8.90 8.90 8.90 8.90 7.74 g!h5.44 5.4 l 0 10000 20000 30000 40000 Bundle Average Exposure (MWD/MT) hlaxknutn Average Planar Linear IIeal Generallon Rale (MAPLIIGR)Versus aundl ~ Average Expoaure SVEA 96 Lead Fuel Aaaembllea Figure 3.2.1-7 ~ ~ h( / f / 'k 1 't l ) f 8 Ol COK C0 OIC OI t3 CO Ol X: CO Ol C'O CQ o. OI OI CQ Ol E E )C 11 l4 4 s r I 4 Bundle Average Exposure (MWD/MTU) 0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 MAPLHGA 10.9 't0.9 10.9 10.9 10. .5 7.9 6.6 Bundle Average Exposure Two Loop Single Loo peratlon 35000 Maximum Average Planar Linear Heal Generallon Rate (MAPLHGR)Versus Bundle Average Exposure GE 11 Lead Fuel Assemblies Ftgure 3.'2.1-8 CONTROLLED COPY 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: 44 8 1 1 h 1i* 11 MCPR 1i R~~. n. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater equal to 25 percent of RATED THERMAL POWER. ACTION: .Hith HCPR less than the applicable HCPR limit determined from , initiate correcti've action within 15 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 POWER to less than 25 percent of RATED THERMAL POWER within the next 4 than or ....~o +~5 g minutes THERMAL hours. SURVEILLANCE RE UIREMENTS 4.2.3.1 MCPR shall be determined to be greater than or cable MCPR limit determined a. At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, equal to the appli-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 1 CONTROLLED COPY gs ~ ~ Table 3. 2. 3-1 HCPR OPERATING LIMITS MCPR Operating Limit U to 106K Core Flow cl e Ex sure 1. OMWD" 3 OMWD MTU HTU 2. 3750 MWD " EOC MTU M 3.. 3750 MWD " EOC MWD" "* HTU MTU 4. 3750 MWD " EOC MWD HTU HTU 5. 3750 MWD - EOC MWD HTU HTU 0 MWD - EOC MWD HTU HTU Equipment Status Normal scram times"" Control rod insertion bounded by Tech. Spec. limits (3.1.3.4-3/4 1-8) RP inoperable Horm 1 scram times*" RPT ino rable Control r d inse ion bounded by ech Spec. limits (3.1. p 3/4 1-8) Single loo oper ion RPT opera le Normal ram times" 8x8 AHF Fuel**"

1. 24

1. 31 1.

6

1. 36

1. 40

1. 35 SV

-96 FA FUEL

1. 37

1. 48

1. 55

1. 55

1. 61

1. 54 "In this portion of the fuel cyc, operation with th given HCPR operating limits is allowed for both nor al and Tech.

Spec. sera times and for both RPT operable and inoperable. ""These MCPR values are bas d on the ANF Reload Safety Analy s performed using the control rod inserti times shown below (defined as norm 1 scram). In the event that surveillanc 4.1.3.2 shows these scram insertion t> es have been

exceeded, the plant ermal limits associated with normal scram times default to the values assoc ted with Tech.

Spec. scram times (3.1.3.4-p /4 1-8), and the scram inse tion times must meet the requirements of Tech. ec. 3.1.3.4. ( Position nserted From Full Withdrawn otch 45 Hotch 39 Notch 25 Hotch 5 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 .404 .660

1. 504

2. 624 WA INGTOH NUCLEAR " UHIT 2 3/4 2"7 Amendment Ho.

84 ~ CONTROLLED COPY ~ Table 3.2.3-1 (Continued) MCPR OPERATIHG LIMITS t """The GE11 LFA fuel, the AHF fuel and the GE initial core el are also monitored to the AHF Bx8 fuel R Operating 'mits (Refe nc

Power Distribution Limits, Bases, 3/4. 2.

Minimum Critical wer Ratio, p. B 3/4 2-3). """"For Final Feedwater Temperature Reduction ra conditions beyond all rods out point, add .O2 to the MCPR for all fu in th 'P-2 core except for the SVEA-96 LFA fuel. For the SVEA"96 A fuel, add to the MCPR for Final Feedwater Temperature Reducti rated conditions be d the all rods out point. ASHIHGTOH HUCLEAR - UNIT 2 3/4 2-7a Amendment Ho. 1.8 Ywo Loop Operall n Cl n I foal 1.7 Total Core Flow Rate 0 . 90 80 70 60 50 40 PR peratlng Llmlt 1.07 'f.13 1.19 '1.26 ".;1.34 1.45 1;59 MI 00 1.0 10 20 30 40 50 60 70 80 90 100 110g O Total Core Flow (% Rated) Reduced Flow MCPR Operating Llmlt This Curve ls Applicable to AHF Reload Fuel, GE InlllalCore Fuel, ANF9 X 9 LFAFuel, GE 41 LFA Fuel, and SVEA-96 LFA Fuel This curve Is also applicable to FFTR operation Figure 3.2.3-1 ComaO<<~O CO< Y ~ POWER DISTRIBUTION LIMITS 3/4. 2. 4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION. 5pcCI5lc 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) %~X shall not exceed the values -shown in C @wc apc~aC ~ lr .C~ g 6 pong APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25~ of RATED THERMAL POWER. ACTION: With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. SURVEILLANCE RE UIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit: a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR. WASHINGTON NUCLEAR " UNIT 2 3/4 2"9 Amendment No. 84 EXP GA Permlsslble Region of Oporatlo 0 510 2,580 5,230 7,940 10,470 13I220 15,990 18,708 210590 24,420 27,280 30,150 33,050 35,960 381900 41,830 44,760 15.62 15.621 15.10 14.71

14. 'l9 14 13 EI, 14.06 "

14.06 14.00.= 13.93 13.93 13.0B 12.24 11.40 10.47 9.55 8.65 7.77 10,000 20,000 30,000 40,000 Average Planar Exposure (hllWD/MTj ANF 8x8 Reload Fuel Linear Heat Generation Rate (LHGR) Umlt Versus Average Planar Exposure 5, 0 Figure 3.2.4-1 0 C) C) Cl Cm 13 12 10 Oundlo Avccogo Exposuro fQWD/QQ 0 5,000 10,000 15,000 20,000 30,000 40 0 ,000 60,000 70,000 LIIGA gwlgl 13.7 13. .7 13.7 13.0 11.5 10.0 0.5 7.0 5.5 0 0 O CA lO 5,000 10,000 20,000. 30,000 40,000 50,000 60,000 '70,000 80,000 Average Planar Exposure (MWD/MT) ANF 9 X 9-IX Reload Fuel Linear Heat Generation Bate (LHGB) Limit Versus Average Planar Exposure C) CD I CD CD E CD Gl CC 0 CD CD (3 c0 CD 05 CDC 14 13 1'f 10 9 6 7 Bund[a Avarag8 Exposure NMMD 0 5,000 10,000 ~ 15,500 20,000 30,000 40,00 5 0 0,000 70,000 Ll{GR lmdli 13.1 13.1 13. 12.5 - 11.2 9.9 8.6 7.3 6.1 0R ZJ0 ro-a A0 ..0 0 5,000 10.000 20,000 30,000 40,000 50,000 60,000 70,000 00,000 Average Planar Exposure (MAD/MT) ANF 9 X 9 - 9X Reload Fuel Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure O CJl lO Figure 3.2A-3 Exposure (MWD/ 0} ~~ E I f Cg fC O P Gl e) 10 co 0) CJ o) 0 tokyo,aOO . 10000 20000 30000 400 Average Planar Exposure (MWDIMT) Linear Heal Generallon Rale (LRGR) LlmltVersus Average Planar Exposure SVEA-96 Lead Fuel Assemblies Figure 3.2AQ 1 'R 1J >h 0 5$ 0 2,580 5,230 7,940 $0,170 13,220 $5,990 $ 8,708 2$'$90 21,120 27gg-;~f'0,150 '3,050 35,960 38,900 1$,830 hh,760 Ll(GB $ 3.$ 13.$ 12.7 12.3 11.9 1$.4 11.8 1 $.4 1$.7 1$.7 11.7 11.0 10.3 0.8 4.9 4.0 7 3 'm 8.5 Q $%04 Average Planar Exposure (MWD/liflT) Uoear Heal Generation Bate (LHGR)Umll Vofsui Avdfagd Plbtlsf Exposure GE 11 Lead Fuel Asaemhllea Agua 32.1-5 O I 1 ONTROLLED COPY ~ s 3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4. l. 1 Two reactor coolant system recirculation loops shall be in operation. APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2*. ACTION: With one reactor coolant system recirculation loop not in operation: 1. Verify that the requirements of LCO 3.2.6 and LCO 3.2.8 are met, or comply with the associated ACTION statements (- 2. Verify that THERMAL POWER/core flow conditions lay outside Region B of Figure 3.4.1.1-1. With THERMAL POWER/core flow conditions which lay in Region B of Figure 3.4. 1. 1-1, as soon as practical, but in all cases within 15 minutes, initiate action to exit Region B by either decreasing THERMAL POWER with control rod insertion or increasing core flow with flow control valve manipulation. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> exit Region B. The starting or shifting of a recirculation pump for the purpose of exiting Region B is specifically prohibited. 3. 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) fop gregal Electric fuel limit to ~kue-s i 1 i 1 P , and, SPc,s:>f-F50 I& T'~Q came opreaWVIOC l-iMIT5 P &POR P d) Reduce the volumetric flow rate of the operating recircula-tion loop to < 41,725"" gpm. ~See peciai 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 3/4 4-1 Amendment No. 7l 'I ~e tp O'I 4 ll,~ ~' lg(>> g g lt ~ "+, g lt 1 )i 1 ~jl~ OONTROLLED COPY ~ 3/4.2 POWER DISTRIBUTION LIMITS 4 I BASES The specifications of this section assure that the peak cladding temperature followin~ 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 LMGR 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 7ECN. O'PECS. ~ ~ ~j'~c'>F>n'0 IN 7uz c a < E GI pc ggT'r A/6 L r hf fvD R EPEDfc7'pzcipraO s'pzciFrc ru 7.gc core, L z, ~ l', ply lptj al procedure used to establish the APLHGR Sheen-en:A'puces 1" 's based on a loss-of-coolant accident analysis'he analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. These models are ~~ed in E l ) 1 SBG rr om WASHINGTON NUCLEAR - UNIT 2 8 3/4 2-1 Amendment No. 71 1 fi I" II ifh M h h M 'M M %ONTROLI ED COPY BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit HCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR and an. analysis of abnormlal operational transients. For any abnormal operating transient analysis evaluation with the initial condi-tion of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specifica-tion 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRiTICAL POWER RATIO (CPR}. The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is +~~~PIg g g RRJ V'08 C C R4'P t'PrfP A<f46 41+f7 f8'R>- The evaluation of a given transient begins with the system i,nitial param-eters shown in the cycle specific transient analysis report that are input to an AHF core dynamic behavior transient computer program. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle. The codes and methodology to evaluate pressurization and nonpressurization events are 'n The principal result of this evaluation is the reduction/in'CPR. caused by the transient. I C+CV C~C+Cl SDC7lOH $. QM 3 Pt TPg 75CH,

5PQCSs, Fcew PsPsPdosur

$.Pgf-I P(60 iw ~88 come p .C "6 IfA7iktq fl p p I h II Rf'~lf dfl p l1'j,'~p limits at other than rated core flow conditions. At less than 10(C of rated fl h d I d MC II I h I f h d fl N PR~ h d fl MCPM ~NCPR-8oT~ S.<<I~~~ >M 7.-88 cRC Israel 7'ie 4 assures that the Safety Limit MCPR will not be violated. ~ MCPRf is only caI-culated for the manual flow control mode. Automatic flow control operation fg pod+ is not permitted. WASHIHGTOH HUCLEAR - UNIT 2 B 3/4 2"3 Amendment Ho. 45 <<<<<<y 74 ~ ~ y ( ~ " gl ) <<1 ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) I The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:, a. Container volume, b. Total curie quantity (specify whether de.ermined by measurement or ~ estimate), ~c.'rincipal radionuclides (specify whether determined by measurement .or estimate),'. Source of waste ard processing employed (e.g., dewatered spent resin, ~ compacted dry waste, evaporator bot oms),

e. 'ype of container (e.g.,

LSA, Type A, Type B, Large guantity}, and f. Solidification agent or absorbent (e.g.,'.cement, urea formaldehyde). 'I II The Radioactive Effluent Release Reports shall include.a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and 'liquid e ents made during the reporting period. The Radioactive Effluent Release Re shall include any change-made during the reporting period to the PROCESS L PROGRAM (PCP) and to she OFFSITE DOSE CALCULATION MANUAL (ODCM), as wel listing of new locat'-;ons for dose calculations and/or environmental m gring identified by t':e land use ~ census pursuant to Specification

3. 12.2.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the anal Administ..ator of the g ~/ Regi ona 1 0ffice of the NRC withi n the time peri od spe~i ed for each.report'. ~~ g 58'. 10 RECORD RETEHTIOH 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at..least the minimum period indicated.

6. 10. 2 b.

The following records shall be retained for at least 5 years: Records and logs of unit,operation coveri'ng time interval at each power'evel: 'I Records and logs of principal maintenance activities', inspections,

repair, and replacement of principal items of equipment related to nuclear safety.

C. All REPORTABLE OCCURRENCES submitted to the Commission d. Records of surveillance activities, inspections; and calibrations required by these Technical Specifications. 'WASHINGTON NUCLEAR - UNIT 2 6-22 INSERT B Core 0 eratin Limits Re ort 6.9.3.1 Core operating limits shall be established prior, to each reload cycle; or, prior, to any remaining portion of a reload cycle; for, the following: a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGR) for Specifications 3.2.1 and 3.4.1. b. The MINIMUM CRITICAL POWER RATIO (MCPR) for, Specification 3.2.3. c. The LINEAR HEAT GENERATION RATE (LHGR) for. Specification 3.2.4. and shall be documented in the CORE OPERATING LIMITS REPORT. 6.9.3.2 The analytical methods used to determine the core operating limits shall be those topical reports and those revisions and/or, supplements of the topical reports previously reviewed and approved by the NRC; which describe the methodology applicable to the current cycle.

For, WNP-2; the topical reports are:

1. XN-NF,-512(P)(A); "XN-3 Critical Power. Correlation" 2. ANF-1125(P)(A); "ANFB Critical Power, Correlation" 3. XN-NF,-524(P)(A); "Exxon Nuclear, Critical Power Methodology

for, Boiling Water, Reactors" 4.

XN-NF,-79-71(P)(A); "Exxon Nuclear, Plant Transient Methodology for, Boiling Water, Reactors" 5. ANF,-913(P)(A); "COTRANSA 2: A Computer. Program for, Boiling Water, Reactor, Transient Analysis" 6. XN-NF,-80-19(P)(A); "Exxon, Nuclear, Methodology for. Boiling Water, Reactors"'. XN-NF.-85-67(P)(A); "Generic Mechanical Design for,.Exxon

Nuclear, Jet Pump Boiling )later, Reactor:

Reload Fuel" 8. ANF.-89-014(P); "Generic Mechanical Design

for, ANF 9x9-IX and 9x9-9X BWR Reload Fuel" 9.

XN-NF,-81-22(P)(A); "Generic Statistical Uncertainty Analysis Methodology" 0 6.9.3.3 The core operating limits shall be determined such that all ~ ~ ~ ~ ~ ~ appl icabl e 1 imits (e.g.; fuel thermal -mechani cal 1 imits; core thermal -hydraul ic 1 imits, EGGS 1 imits, nucl ear 1 imits such as shutdown

margin, transi ent anal ysi s 1imits and acci dent anal ysi s limits) of the safety analysis are met.

6.9.3.4 The GORE OPERATING LIMITS REPORT; including any mid-cycle revisions or supplements, shall be provided upon issuance

for, each reload cycle; to the NRG Document Control Desk with copies to the Regional Administr ator, and Resident Inspector.

l t ~ ATTACNNENT II PROPOSED WNP-2 CYCLE 6 CORE OPERATING LIMITS REPORT Controlled Copy No. HNP-2 CYCLE 6 CORE OPERATING LIMITS REPORT August; 1990 HASHINGTON PUBLIC POHER SUPPLY SYSTEM PAGE 1 1 2 3 5 67, ,8 9 10ll 12 13 14 15 16 17 18 19 20 WNP"2 CYCLE 6 CORE OPERATING LIMITS REPORT LIST OF EFFECTIYE PAGES AMENDMENT 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 LEP-1 MNP-2 CYCLE 6 CORE OPERATING LIMITS REPORT TABLE OF CONTENTS

1.0 INTRODUCTION

AND

SUMMARY

2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LIMIT FOR USE IN TECHNICAL SPECIFICATION 3.2.1 3.0 MINIMUM CRITICAL PONER RATIO (MCPR)

LIMIT FOR USE IN TECHNICAL SPECIFICATION 3.2.3 4.0 LINEAR HEAT GENERATION RATE (LHGR)

LIMIT FOR USE IN TECHNICAL SPECIFICATION 3.2.4

5.0 REFERENCES

>II lit

1.0 INTRODUCTION

AND

SUMMARY

This report provides the AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) limits; the MINIMUM CRITICAL POWER RATIO (MCPR) limits; and the LINEAR HEAT GENERATION RATE (LHGR) limits for WNP-2; Cycle 6 as required by Techni cal Speci ficati on 6.9.3.1.

As required by Techni cal Specifications 6.9.3.2 and 6.9.3.3; these limits have been determined using NRC-approved methodology and are established such that all applicable limits of the plant safety analysis are met.

The reload design is discussed in detail in the Cycle 6

Reload Summary Report (Reference 1.0).

The ther mal limits given here are developed in the Cycle 6

Transient Analysis Report (Reference 2.0);

and the Cycle 6 Reload Analysis Report (Reference 3.0).

Included in the WNP-2 Cycle 6 reload are four, General Electric (GE) lead fuel assemblies (LFA's) and four; ABB Atom (ABB) LFA's.

These LFA's have been designed to be compatible with the reload fuel assemblies that will constitute the remainder.

of the reload batch for, Cycle 6.

The Supply System will load the LFA's in core locations which have been analyzed to have sufficient margin such that the LFA's are not expected to be the limiting assemblies in the core on either, a nodal or, an assembly power basis.

This approach is intended to prevent the possibility of the LFA's from ever, being the limiting fuel assemblies.

The GE11 LFA is described in the GE11 Lead Fuel Assembly Report for, Washington Public Power Supply System Nuclear.

Project No.

2 Reload 5',

Cycle 6

(Reference 4.0).

The SVEA-96 LFA is described in the Supplemental LFA Licensing Report-SVEA-96 LFA's

for, WNP-2 (Reference 5.0).

The process

for, developing thermal limits for, the SVEA-96 LFA fuel based upon the ANF Cycle 6 reload fuel thermal limits is described in this Reference.

Preparation; review and approval of this report were performed in accordance with applicable Supply System management; engineering and operating procedures.

References 6.0 through 13.0 identify the specific topical report revisions and supplements which describe the methodology utilized in this cycle specific analysis.

2.0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LIMIT FOR USE IN EH ICLS I

I All APLHGRs for, use in Technical Specification 3.2.1 for each fuel type as a function of AVERAGE PLANAR EXPOSURE for General Electric (GE) initial fuel and AVERAGE BUNDLE EXPOSURE for, Advanced Nuclear, Fuels (ANF) fuel; GE11 LFA fuel and SVEA-96 LFA fuel', shall not exceed the limits shown in Figures 1; 2; 3; 6; 7 and 8 when in two-loop operation and Figures 3; 4; 5;

6; 7

and 8

when in single loop operation.

For GE Fuel; with one recirculation loop not in operation; reduce the MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) to a

value of 0.84 times the two loop limits found in Figures 1 and 2.

13.0 12.5 12.0 f

11.5 CD hC G.

11.0

~~ c 0'

10.5 E

a)

CD x

10.0 9.5 9.0 12.1 12.7 12.8 12.9 12.7 11.7 10.8 10.0 9.4 1,102 5,512 11)023 16,535 22)046 27)558 33,069 381581 44,093 AVERAGE PLANAR KZQKB RKEQl~UE Two Loop Operation 8.0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Average Planar Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Average Planar Exposure Initial Core Fuel Type 8CR1 83 Figure 1 900541.1 (6I90)

125 9.0 g

13.0 ao

)

12.0

!O I

11.5 H~

~ 8 11.0 0

105 Em E'

100 x~

gg )

9.5 12.0 12.1 12.2 12.2 12.1 11.6 11.2 10.6 10.0 1,102 5)512 11,023 16,535 22)046 27,558 33,069 38,581 44,093 AVERAGE PLANAR J~L~

~EOQQQE Two Loop Operation 8.0 5)000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Average Planar Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Average Planar Exposure Initial Core Fuel Type 8CR233 Figure 2 900541.2 (6$0)

CD CQ O

RcO CD CD I

lO 13.0 12.5 f

12.0 5

m<

c o 115 e o 11.0 O C Q

CD E g 10.5 a (9 DC CD m f-10.0 c

9.5 9.0 13.0 13.0 13.0 13.0 13.0 11.3 9.4 7.9 0

5,000

-10,000 15,000.

20,000 25,000 30,000 35,000 AVERAGE BUNDLE

~E)!)LE RCEO!i!!BE Two Loop and Single Loop Operation 8.5 5,000 10,000 15)000 20,000 25,000 30,000 35,000 40,000 Average Bundle Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Bundle Average Exposure ANF 8x8 Reload Fuel Figure 3 9005412

($90)

10.5 10.0 C5

~ e 9.5 mn oc9.0 8.5 x -a) 8.0 cc7'5 7.0 APLHG 10.16 10.66 10.75 10.84 10.66 9.83 9.07 8AO 7.90 AVERAGE PLANAR

~EO~SU E

1,102 5,512 11,023 16,535 22,046 27,558 33)069 38,581 44,093 Single Loop Operation 6.5 5,000 10,000 15,000 20,000 25,000 30,000 Average Planar Exposure (MWD/MT) 35,000 40>000 45,000 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Average Planar Exposure Initial Core Fuel Type 8CR183 Figure 4 900541.4

($90)

10.5 10.0 g

9.5 E

CD cD 90 c$ 0 85 CD Q e E

cD 8.0 a@

cD 75 7.0 6.5 10.08 10.16 10.25 10.25 10.16 9.74 9.41 8.90 8.40 1,102

• 5,512 11,023 16,535 22,046 27,558 33,069 38,581 44,093 AVERAGE PLANAR

~MAPLH E

~EX OSU E

Single Loop Operation 6.0 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45,000 Average Planar Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Average Planar Exposure Initial Core Fuel Type 8CR233

.Figure 5 900541.5 (Gl90)

12.0 11.5 11.0 10.5 C5 a.

10.0 Ul C C$ 0 9.5 E o 9.0 a Q 8.5 8.0 7.0 MAPLHG 11.2 11.2 11.2 11.2 11.2 9.7 8.1 6.8 AVERAGE BUNDLE

~EXPO UR "0

5,000 10,000 15,000 20,000 25,000 30,000 35,000 Two Loop and Single Loop Operation 6.5 5,000 10>000 15,000 20,000 25,000 30,000 35,000 Bundle Average Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Bundle Average Exposure ANF 9x9 - IXAND 9x9 - 9X Reload Fuel Figure 6 900541.8 (680)

CD

D u

8 O,

C5 CDUl C CO 0 Q o 7

C DQ C$

DC CDX CD C

6 8.90 8.90 8.90 8.90 8.90 7.74 6.44 5.41 0

5,000 10,000 15,000 20,000 25,000 30,000 35,000 AVERAGE PLANAR M>!EL'XXOsURE Two Loop and Single Loop Operation 10,000 20,000 Average Planar Exposure (MWD/MT) 30>000 40>000 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Average Planar Exposure SVEA-96 Lead Fuel Assemblies Figure 7 900541.7 (680)

11.5 7.5 7.0 CX 0

z Co 11.0 5

toe m

m m 10.0 e a m 0 9.5 cf e E e 9.0 DQ m f-8.5 8.0 10.9 10.9 10.9 10.9 10.9 9.5 7.9 6.6 0

5,000 10,000 15,000 20,000 25,000 30,000 35,000 AVERAGE PLANAR

/HAPL~ Q~OJJJg Two Loop and Single Loop Operation 6.5 5)000 10,000 15,000 20,000 25,000 30,000 35,000 Bundle Average Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)Versus Bundle Average Exposure GE 11 Lead Fuel Assemblies Figure 8 900541.8 (GI90)

3. 0 MINIMUM CRITICAL POWER RATION (MCPR)

LIMIT FOR USE IN TECHNICAL

~

~

~

~

~

~

CI.ICA ION The MCPR limit for use in Technical Specification 3.2.3 shall be:

a.

Greater, than or. equal to the applicable MCPR limit determined from Table 1 during steady state operation at or above r ated core flow in two loop, or, when in single loop operation; or b.

Greater than or equal to the greater.

of the two limits determined from Table 1 and Figure 9 during steady state oper ation at less than rated core flow when in two recirculation loop operation.

10

TABLE 1 MCPR OPERATING LIMITS Cycle

~Ex osure Equipment Status NCPR OPERATING LIMIT UP TO 106%

CORE FLOW ANF SYEA"96 8x8 Fuel***

LFA Fuel 1.

0 NWD 3750 NWD

~NU

~MU 2.

3750 MWD - EOG MWD***kormal scram times**

~NU

~MU 1.24 1.31 1.37 1.48 3.

3750 MWD " EOG

~MU NWD***control Rod insertion

~N U

bounded by Tech.

Spec.

limits (3.1.3.4-p 3/4 1-8)'.36 1.55 5.

3750 NWD EOG

~MU NWD RPT inoperable

~M U

Control rod insertion b'ounded by Tech.'pec.

limits (3.1.3.4 "

p 3/4 1-8)

Single loop operation RPT inoperable Normal scram times**

6.

0 NWD " EOC NWD

~MU

~NU 4.

3750 NWD - EOG NWD RPT inoperable

~N U

~M U

Normal scram times**

1.36 1.40 1.35 1.55 1.61 1.54

• 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 MGPR 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.

11

~

~

TABLE 1 (Continued)

NCPR OPERATING LIMITS Position Inserted From Full Withdrawn Notch 45 Notch 39 Notch 25 Hotch 5

Slowest measured average control rod insertion times to specified notches for. all operable control rods for, each group of 4 contr ol rods arranged in a two-b -two arna (seconds)

.404

.660 1.504 2.624

• • • The GE11 LFA fuel; the ANF.

LFA fuel and the GE initial core fuel are also monitored to the ANF. 8x8 fuel NCPR Operating Limits.

• • • ~For, Final Feedwater Temperature Reduction rated conditions beyond all rods out point; add

.02 to the NCPR for, all fuel in the WNP-2 core except for, the SVEA-96 LFA fuel.

For, the SYEA-96 LFA fuel; add

.03 to the NCPR for, Final Feedwater.

Temperature Reduction rated conditions beyond the all rods out point.

12

2.3 2.2 Two Loop Operation 2.1 2.0 1.9 1.8 1.7 1.6 0

1.5 O

1.4 MCPR OPERATING LIMIT 1.07 1.13 1.19 1.26 1.34 1.45 1.59 1.77 2.27 TOTAL CORE FLOW~RQg 100 90 80 70 60 50 40 30 20 1.3 1.2 1.0 20 30 40 50 60 70 80 90 100 110 Total Core Flow (% Rated)

Reduced Flow MCPR Operating Limit Versus Total Core Flow AllFuel in WNP-2 Cycle 6 This curve is applicable to FFTR operation Figure 9 900541.S (Gl90)

4. 0 LINEAR HEAT GENERATION RATE (LHGR)

LIMIT FOR USE IN TECHNICAL

~

~

~

CI:ICA IO The LHGR limit for use in Technical Specification 3.2.4

for, GE initial core fuel,shall not exceed 13.4 kw/ft.

The LHGR limit for use in Technical Specification 3.2.4 for, reload fuel shall not exceed the values shown in Figures 10; ll; 12; 13 and 14.

14

16 14 12 10 8

itiGR 15.62 15.621 15.10 14.71 14.19 14.13 14.06 14AS 14AO 13.93 13.93 1308 1224 11AO 1a47 8.65 7.77 AVERAGE PLANAR EKBQKBE 0

510 2,580 5,230 7,940 10,470 13,220 15,990 18,708 21/90 24,420 27>280 30,1 50 33,050 35,960 38,900 41/30 44>760 10,000 20,000 30,000 Average Planar Exposure (MWD/MT) 40,000 50,000 Linear Heat Generation Rate (LHGR) Llmlt Versus Average Planar Exposure ANF Sx8 Reload Fuel Figure 10 900541.10 (6t90)

14 13 LHQB 13.7 13.7 13.7 13.7 13.0 11.5 10.0 8.5 7.0 5.5 BUNDLE AVERAGE EZQKBK 0

5,000 10,000 15,000 20,000 30,000 40,000 50)000 60)000 70,000 5,000 10,000 20)000 30,000 40,000 50,000 605000 70,000 80,000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure ANF9x9-IX Reload Fuel Figure 11 90054M1 (SI90)

14 13 f

5 E

11 I

Pg 10 C0 9

8 C

6 LAB 13.1 13.1 13.1 13.1 12.5 11.2 9.9 8.6 7.3 6.1 BUNDLE AVERAGE

~EX

'egg 0

5,000 10,000 15,000 20,000 30,000 40,000 50,000 60,000 70,000 5,000 10,000 20>000 30s000 40>000 50,000 60,000 70>000 801000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure ANF9x9 - 9X Reload Fuel Figure 12 900541.12

($90)

12 AVERAGE PLANAR JLLQ

~EXPOSU E

~ 11.6 0 to 40,000 10,000 20)000 30,000 40)000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure SVEA-96 Lead Fuel Assemblies Figure 13 900541.1 3

($90)

tl

Ol Cl O

z CO CD I

!O 14 13 11 Q

10 CO 9

Q 8

C 6

LUQB 13.1 13.1 12.7 153 11.9 118 118 118 11.7 11.7 11.7 113) 102 9.6 SAI LO 78 6$

AVERAGE PUNAR EKEQRllBK 0

510 2580 5,230 7,940 10,470 13,220 15,990 18,708 21/90 2~20 27/80 30,150 33,050 35,960 38,900 41JQO 44,760 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 45)000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure GE11 Lead Fuel Assemblies Figure 14 900541.14 (6I90)

5.0 REFERENCES

1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 14.0 WPPSS-EANF,-126; Rev.

1; "WNP-2 Cycle 6

Reload Summary Report";

Washington Public Power, Supply System; April 1990 ANF-90-01; "WNP-2 Cycle 6 Plant Transient Analysis Report"; Advanced Nuclear, Fuels Corporation',

January 1990 ANF,-90-02; "WNP-2 Cycle 6 Reload Analysis Report"',

Advanced

Nuclear, Fuels Corporation',

January 1990 GE11; "Lead Fuel Assembly Report forWashington Public

Power, Supply System Nuclear Project No.

2 Reload 5 Cycle 6";

General Electric Company; December.

1989 (Proprietary)

UK 90-126; "Supplemental Lead Fuel Assembly Licensing Report SVEA-96 LFA's for, WNP-2"; ABB Atom; January 1990 (Proprietary)

XN-NF-512(P)(A)',

"XN-3 Critical Power.

Correlation";

Revision 1;

Supplement 1; October, 1982 XN-NF-524(P)(A);

"Exxon Nuclear Critical Power Methodology for.

Boiling Water, Reactors";

Revision 1; November 1983 XN-NF-79-71(P)(A);

"Ex;xon Nuclear Plant Transient Methodology for Boiling Water, Reactors",

Revision 2;

November, 1981 and Revision 2;

Supplements 1;

2 and 3; March 1986 ANF-913(P) (A);

"A

Computer, Program

for, Boiling Water

Reactor, Transient Analysis"; Volume 1; Supplmenets 1;

2 and 3; May 1988 XN-NF-80-19(P)(A);

"Exxon

Nuclear, Methodology

for, Boiling

Water, Reactors";

Volume 1; May 1980, Volume 1; Supplements 1 and 2; March 1983; Volume 3', Revision 2; January 1987 XN-NF.-85-67(P)(A); "Genenc Mechanical Designs for, Exxon Nuclear. Jet Pump Boiling Water, Reactor, Reload Fuel"; Revision 1', September.

1986 ANF,-89-014(P);

"Generic Mechanical Design for, ANF 9x9-IX and 9x9-9X BWR Reload Fuel"; Revision 0; Supplement 1; June 1990 XN-NF-81-22(P) (A) ',

"Generic Statistical Uncertainty Analysi s Methodology";

November, 1983 YUF.:139: 89; ANFWP-89-0106; YU Fresk; Advanced Nuclear Fuels; to Manager.',

Central Contracts',

Supply System; "Justification for. Cycle 5 Reduced Flow MCPR Curve";

June 30; 1989 20

f 0