Text

Proposed Tech Specs Adding &/Or Deleting Portions of Paragraph 6.9.3.2 of Insert B
	ML17286A717
Person / Time
Site:	Columbia
Issue date:	04/19/1991
From:	; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	;
Shared Package
ML17286A716	List: ML17286A714; ML17286A717;
References
	NUDOCS 9104240166
	Download: ML17286A717 (49)
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INDEX k

DEFINITIONS SECTION

l. 0 OEF IH ITIOHS PAGe,

l. 1 ACTION. ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~

]-1 1.2 AVERAGE BUNDLE EXPOSURE ..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ l-l 1.3 AVERAGE PLANAR EXPOSURE

]. 4 AVERAGE PLANAR LINEAR HEAT GFHERATION RATE

1. 5 CHANNEL CALISRATIOH.

l. 6 CHANHEL CHECK....

1.7 CHAHHFL FUNCTIONAL TEST................... ]-~ I 1.8 1.9 CORE ALTERATION Q~gg CRITICAL POMER pQ gATI MG RATIO....

L I Rl TS ga PoHT'g 1-2 1.10 OOSe. E(UIVALEHT I-]31. 1-2

].ll E-AVERAGE DISINTEGRATIOH ENERGY ................... 1-2 1.12 Pi1ERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME......... 1-2

1. 12-A END-OF-CYCLE (EOC) 1-2 1 ]3 EHO-OF-CYCLE RECIRCULATION PUtlP TRIP SYSTEH RESPONSE TIME.. 1-3

" . ]3 A FINIAL FEEDMATER TEHPERATURE RFDUCTIOH (FFTR)............. 1 3

l. 14 FRACTIOH OF LIMITIHG POMER DENSITY..'.

1-3

]5 FRACTION OF RATED THERMAL, OWER. 1-3 1 16 FREOUENCY NOTATION..

l. 17 GASEOUS RADWASTF TREATHEHT SYSTEH...

l. 18 IDENTIFIED LEAKAGE 1-3

].]9 ISOLATION SYSTEM RESPONSE TIHE

l. 20 LIMITING CONTROL ROD PATTERN.

9104240166 910419 1 21 LINEAR HEAT GEHERATION RATF,, PDR Al30CK , 05000397...

P PDR MASHIiNGTON iHUCLEAR - UNIT 2 Amendment No 77

1 INDEX

'LIMIT'IHG CONDITIONS FOR OPERATION AHD SURVEILLANCE RE UIREMEHTS SECTION PAGE 3/4. 0 APPLICABILITY 3/4 O-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 SHUTDOWN MARGIN. 3/4 1-1 3/4:1. 2 REACTIVITY ANOMALIES 3/4 1-2 3/4. 1.3 CONTROL RODS Control Rod Operabi 1 ity. 3/4 1-3 Control Rod Maximum Scram Insertion Times... 3/4 1-6 Four Control Rod Group Scram Insertion Times.. 3/4 1-8 Control Rod Scram Accumulators................ 3/4 1-9 Control Rod Drive Coupling... 3/4 1-11 Control Rod Position Indication. 3/4 1-13 Control Rod Drive Housing Support. 3/4 1-15 3/4. 1.4 CONTROl ROD PROGRAM CONTROLS Rod Worth Minimizer 3/4 1-16 Rod Sequence Control System. 3/4 1-17 Rod Block Monitor 3/4 1-18 3/4. 1.5 STANDBY LI(UID CONTROL SYSTEM. ................... 3/4 1-19 3/4. 1.6 FEEDWATER TEMPERATURE . ......................... 3/4 1-23 3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE............. 3/4 2-1 3/4 2.2 APRM SETPOIHTS 3/4 2W W 3/4.2.3 MINIMUM CRITICAL POWER RATIO........................... 3/4 2-N +

3/4.2.4 LINEAR HEAT GENERATION RATE . 3/4 2-~ Q 3/4.2.5 (RESERVED FOR FFTR) 3/4.2.6 POWER/FLOW INSTABILITY. 3/4 2-m5 3/4.2.7 STABILITY MONITORING - TWO LOOP OPERATION.............. 3/4 2-X3 7 3/4.2.8 STABILITY MONITORING - SINGLE LOOP OPERATION........... 3/4 2-~ 9 WASHINGTON NUCLEAR - UNIT 2 v Amendment Ho. 77

1 r>> e Krerr ~

r I

~

1NDEX ADMINISTRATIVE CONTROLS SECTION PAGE

6. 1 RESPOHSI BI LITY. 6-1

6. 2 ORGANIZATION................ 6-1 6.2.1 OFFSITE AHD OHSITE ORGANIZATIONS...................... 6-1 6.2.2 U HIT STAFF ....................................... ~ ~ 6-1 6.2.3 NUCLEAR SAFETY ASSURANCE GROUP........................ 6-7 FUNCTION............ 6"7 RECORDS' COMPOSITION..... 6-7 RESPONSIBILITIES 6-7

....... 6-7 6.2.4 SHIFT TECHNICAL ADVISOR............................... 6-7

6. 3 UNIT STAFF UALIFICATIOHS.................................. 6-7

6. 4 TRAINING.... 6-.8

6. 5 REVIEW AHD AUDIT.......... 6-8 6.5.1 PLANT OPERATIONS COMMITTEE (POC)...... 6-8 FUNCTION........ 6-8 COMPOSITION. 6-8 ALTERNATES 6"8 MEETING FRE(UEHCY. 6-8 (UORUM............ 6-8 RESPOHSI BI LITI ES............... 6" 9 RECORDS ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6.5.2 CORPORATE NUCLEAR SAFETY REVIEW BOARD (CNSRB)......... 6-10 F UNCTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ 6" 10 COMPOSITION e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

' 6" 10

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

ALTERNATES.............. 6-10:

C oNSuL<H 875, g- )1 WASHINGTON NUCLEAR - UNIT 2 Xvl l 1 Amendment No. 63

TNDEX II I

(,

ADMINISTRATIVE CONTROLS SECTION PAGE HEETIHG UORUMI ~ ~ ~

~

CORPORATE HUCLEAR SAFETY REVIEW BOARD

-GQNSDtVANPS.

(Continued)

FRE(UEHCY....................................

o ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ o ~ o ~ ~ ~ ~ a ~ ~ o o ~ o ~ o ~ ~ o ~ ~ ~ e ~

6-11 R EVIBI ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ eo ~ ~ ooo ~ ~ o ~ o ~ eeo ~ o ~ ~ ~ ~ e ~ ~ ~ ~ eo ~ ~ o ~ o 6-11 AUDITSI I ~ a ~ aaeaa ~ aooao ~ ~ ~ oo ~ o ~ oo ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ e ~ ~ o ~ ~ ~ ~

R ECORDS ~ ~ ~ ~ ~ ~ ~ ~ ) ~ e ~ ~ ~ ~ ~ ~ ~ ~ o ~ o g ~ a ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ 6-13

6. 6 REPORTABLE OCCURRENCE ACTION.............................. 6" 13 6.7 SAFETY LIMIT VIOLATIOH.................................... 6-a4 6.8 PROCEDURES AND PROGRAMS................................... 6-14 6.9 REPORTIHG RE UIREMEHTS.............. 6-16 6.9.1 ROUTIHE REPORTS AHD REPORTABLE OCCURRENCES..........

STARTUP REPORT..................................... 6"16 ANNUAL REPORTS ~ ~ ~ ~ o ~ o o ~ ~ ~ o o ~ ~ o ~ e ~ ~ ~ o ~ ~ a ~ ~ ~ ~ o ~ ~ ~ o e o e o ~ 6-16 MONTHLY OPERATING REPORTS...................,........ 6-17 REPORTABLE OCCUREHCES...............'................. 6"17 PROMPT NOTIFICATION MITH WRITTEN FOLLOMUP............ 6"17 THIRTY DAY WRITTEN REPORTS........................... 6-aa. JS ANNUAL RADIOLOGICAL EHVIROHMEHTAL OPERATING REPORT... .6-ge i8 SEMIANNUAL RADIOACTIVE EFFLUEHT RELEASE REPORT....... 6-n i9 6.9.2 6 e'10 SPECIAL REPORTS CoQ&OPZPBrrHC RECORD RETEHTIOH a ivy QcPC¹"...,

~ o ~ ~ ~ o ~ ~ o ~ o o o ~ o oie o ~ o o ~ ~ ~ ~ ~ ~ ~ ~ o o o o o o o

~

6-22 DC c-6-22 QQ

6. 11 RADIATIOH PROTECTION PROGRAM......................... 6-24 6.12 HIGH RADIATIOH AREA.....-......-.-----..-.---.------.---. 6"24 o

6.13 PROCESS CONTROL PROGRAM...................... -- ~ 6-25

6. 14 OFFSITE DOSE CALCULATIOH HAHUAL.......................... 6-25 6.15 MAJOR CHANGES TO RADIOACTIVE LI UID GASEOUS ANO SOLID HASTE TREATMENT SYSTEMS.....-..............-...-.-. 6-26 WASHINGTON HUCLEAR - UNIT 2 xix Amendment No.. 16..

x INDEX LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTIOH SATURATION TEMPERATURE- - - 3/4 1-21 3.l. 5-2 SODIUM PEHTABORATE TANK, YOLUME VERSUS CONCEHTRATION REQUIREMENTS....................................;. 3/4 1" 22

3. 2. 1" 1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AYERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR183........... "~" " ~ ~ ~ 8F~""2 0<I<'lQdl

3. 2. 1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR233......... " ~ ~ ~ ~ M~~ Qel'EILCI

3. 2. 1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AYERAGE BUNDLE EXPOSURE ANF SxS RELOAO FUEL............................... A~~ QCjek'3

3. 2. 1" 4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AYERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE SCR183................................ 3~~ j3EI<h'CI

3. 2. 1-5 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE

3. 2. 1-6 (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR233.........,......................

MAXIMUM AVERAGE PLANAR LINEAR HEAT'EHERATION RATE SFA ~ ae(GI-Fd (MAPLHGR) VERSUS AYERAGE BUNDLE EXPOSURE AHF 9x9-IX AHD 9x9-9X FUEL........... . OE iC.keel

3. 2. 1-7 MAXIMUM AVERAGE PLAHAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS BUHDLE AVERAGE EXPOSURE SVEA-96 LEAD FUEL ASSEMBLIES.......

3. 2. 1" 8 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS BUNDLE AVERAGE EXPOSURE GEll LEAD FUfL ASSEMBLIES......;.................'......... 8/A~K

3. 2. 3-1 REDUCED FLOW MCPR OPERATING LIMIT.................... BL4 ~ QE.Je&

3. 2. 4-1 LIHEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGf PLANAR EXPOSURE ANF 8x8 RELOAD FUEL.......... W4-4 M Q<IG.kd 3.2. 4" 2 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VfRSUS AVERAGE PLANAR fXPOSURE AHF 9x9-IX FUEL.............. 3~~A.

3.2. 4-3 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF Sx9-9X FUEL.............. 'IFA-~SB Qf(SfC~i-

3. 2. 4" 4 LIHEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE SVEA-96 LEAD FUEL ASSEMBLIES........................................... 3~hlOC-WASHINGTON HUCLEAR - UNIT 2 Amendment Ho. 84

IROEX LIST OF FIGURES FIGURE PAGE

3. 2. 4-5 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE GE11 LEAD FUEL ASSEMBLIES.................

3. 2. 6-1 OPERATING REGION LIMITS OF SPEC. 3.2.6............... 3/4 2-~6

3. 2. 7-1 OPERATING REGION LIMITS OF SPEC. 3.2.7............... 3/4 2-X 8

3. 2. 8-1 OPERATING REGION LIMITS OF SPEC. 3.2.8............... 3/4 2-P /C 3.4. l. 1-1 THERMAL POWER LIMITS OF SPEC. 3.4.1.1-1.............. 3/4 4-3a 3.4.6.1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE....................... 3/4 4-20 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST .......... 3/4 7-15 3.9.7-1 HEIGHT ABOVE SFP MATER LEVEL VS. MAXIMUM LOAD TO BE CARRIED OVER SFP..................,........ 3/4 9"10 B 3/4 3-1 REACTOR VESSEL MATER LEVEL.......... .. ...... B 3/4 3-8 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 EXCLUSION AREA BOUNDARY ............................. 5-2

5. 1-2 LOW POPULATION ZONE . 5-3

5. 1-3 UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LI(UID EFFLUENTS............. 5-4 WASHINGTON NUCLEAR - UNIT 2 xx(a) Amendment No. .87

INOER LIST OF TABLES TAGLE PAGE SURVEILLANCE FREQUENCY NOTATION...................... 1-9 1.2 OPERATIONAL CONDITIONS............................... 1-10 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS.. 2-4 B2. 1. 2" 1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT........................... B 2-3

3. 2. 3" 1 NCPR OPERATING LIMITS FOR RATEO CORE FLON...........:Lf42.&O-P/<t7CA

3. 3. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............ 3/4 3-2

3. 3. 1-2 REACTOR PROTECTION SYSTEM RESPONSE TIMES............. 3/4 3-6

4. 3. 1. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3/4 3-7

3. 3. 2-1 ISOLATION ACTUATION IHSTRUMEHTATION. 3/4 3-12

3. 3. 2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS... 3/4 3-16

3. 3. 2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME....... 3/4 3-19

4. 3. 2. 1-1 ISOLATION SYSTEM INSTRUMENTATIOH SURVEILLANCE REQUIREMENTS 3/4 3"22 3.3.3"1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION. 3/4 3-26 3.3.3 2 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.................. 3/4 3-30 3.3.3"3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES......... 3/4 3-33 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............ 3/4 3-34

3. 3. 4. 1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION. 3/4 3-38 3.3.4.1-2 ATWS RECIRCUL'ATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS 3/4 3"39

4. 3.4. 1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS............ 3/4 3-40 WASHINGTON NUCLEAR " UNIT 2 XX1 Amendment No.

0

1. 0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifi-cations may be achieved. The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.

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

AVERAGE BUNDLE EXPOSURE 1.2 The AVERAGE BUNDLE EXPOSURE is equal to the sum of the axially averaged exposure of all the fuel rods in the specified bundle divided by the number of fuel rods in the bundle.

AVERAGE PLANAR EXPOSURE 1.3 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.4 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLMGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL 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 trip into the functions.

sensor to verify OPERABILITY including alarm and/or The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, entire channel is tested.

overlapping or total channel steps such that the MASHINGTON NUCLEAR - UN!T 2 1"1 Amendment No. 2B

DEFINITIONS CORE ALTERATION

1. 8 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Sus-pension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.8A The CORE OPERATIHG LIMITS REPORT is the MHP-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.

CRITICAL POMER RATIO 1.9 The CRITICAL POMER 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 E UIVALENT I"131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

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

E-AVERAGE DISINTEGRATION ENERGY

1. 11 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half-lives greater than 15 minutes making up at least 95K of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM ECCS RESPONSE TIME 1.12 The EMERGENCY CORE COOLING SYSTEM (ECCS} RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

MASHINGTON NUCLEAR - UNIT 2 Amendment No. 84

As MQDIF=iE.D PER.

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3.2.1 All AVE.. GE PLAHAR LIHEAR HEAT GEHERATION RATES (APLHGRs) for each type of fuel I ~ ~ ~ ~

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APPLICABILITY: OPERATIOHAL COHDITIOH 1, when THERMAL POMER is greater than or equal to 25 of RATED THERMAL POMER.

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rective a'ction wit.hin 15 minutes and res ore APLHGR to wi hin the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or r duce THECAL POMER to less than 2S of RATED THEiPCL POWER within he next, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

c. Ini ially 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 opera ing with a LIMITIXG COHTROL ROO PATTERN for APLHGR.

MASHIHGTOH NUCLEAR - UNIT 2 3/4 2-1 Amendment Ho. 84

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4I 0 CO Dun dlo Averago K

CO 4)

Exposure (MNDlfATLJ) MAPLllGA 0 C

0 10.9 CO 5,000 10.9 C7 CO 10,000 10.9 n

4J C1l

'15,000 20,000 0

CO I 25,000 4C 30,000 7.9 35,000 6.6 E

D C

X B

CO a-CD D 0 25000 35000 G.

B Dundlo Average Exposure CO ft O hlaxlmum Average Planar l.lnear l leal Genera(Ion llalo (MAPLllGA)Versus Dundlo Average Cxposuro-GL 11 Lead f-uel Assemblies

~ ~

Figure 3.2.1-0

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE VALUE S < 0 66W + 5 T S < 0.66W + 54 T SRB

< (0.66W + 42K)T < (0.66W + 45K)T SRB where: S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million lbs/h.

T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. T is always less than or equal to 1.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

~q1 1 RATER THERMAL TAHAR.

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 S or SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or SRB to be consistent with the Trip Setpoint value( ) within 2 hours2.314815e-5 days 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 MFLPD for each class of fuel shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated ther-mal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

a. At least once per 24 hours2.777778e-4 days 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 15K 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 MFLPD greater than or equal to FRTP.

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

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

e'4 l%'VIA'ATTa/AlmaeaIVV

~ ~

POMER DISTRIBUTION LIMITS 3/4.2.3 MIHII"UH CRITICAL POMER RA IO LIMITING CONDITION FOR OPERATION

'3.2.3 The MIHikUM CRITICAL POMER RATIO (MCPR) shall be:

Gre=

'.ar han or equal to tha applicable HCPR limit

~ V V V V C'l~ 4I V Vv ~ e I I I I I I II I T I ~ I~ ~ VV VV ~

Icae:= +

c:. e. e On.

APPLICABILITY: OPERATIONAL CONDITIOH 1, when THERMAL POMER is eater than or equal to 25 percent of RATED THERMAL POMFR.

e, ACTION: .Mith HCPR less than the applicable initiate MCPR limit correcti've action within 15 minutes Zg-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 percen of RATED THERMAL POWER within the next 4 hours.

r.>

SURVEILLANCE RE"'UIREMEHTS ~ I.r 4.2.3.1 be determined to greater than or equal to the appli-cab 1 MCPR sha e MCPR 1 im-i t P-be 7 ') 0 1 P'a

a. At least once per 24 o oiaacIfikd ir +C- .-V.

b. Mithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after completion of a THERMAL POMER increase of at lees- ~5 percent of RATED THERMAL POMER, and

c. Ini ";a11y 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 LIMITIHG CONTROL ROD PATTERN for MCPR.

3 WASHINGTON HUC~i=OR - UNIT 2 ve 2-p'mendment No. 62

WomRo<<ED copv (

Tabl e 3. 2. 3-1 MCP R OPERATING LIMITS MCPR Operating Limit

, Uo to 106Ã Core Flow SY -46 cl e Equipment BxB FA Ex sure Status AHF Fuel""* FUEL

1. 0 HMO "-3 0 M) 1. 24 1. 37 MTU MTU

2. 3750 MMD - EOC """" Hor, al scram times"" 1. 31 l. 48 MTU. M

3. 3750 $$ " EOC M)" "~ Control rod insertion l. 6 MTU MTU bounded by Tech. Spec.

limits (3.1.3.4-3/4 1-8)

4. 3750 HMO " EOC M) RP inoperable l. 36 1. 55 MTU MTU Horm 1 scram times""

5. 3750 %0 - EOC %$ RPT ino rable 1. 40 1. 61 MTU MTU Control r d inser ion bounded by ech Spec.

limits (3. 1.

p 3/4 1-8)

6. 0 WD - EOC MMD, Single loo oper ion 1. 35 1. 54 MTU MTU RPT opera le Normal ram times" "In this portion of the fuel cyc, operation with th given MCPR 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 Sa ety Analy s performed using the control rod inser i times shown below (defined as norm 1 scram). In the event that surveillanc 4.1.3.2 sho~s these scram insertion ts es have been exce ded, the plant ermal limits associated with normal scram times default to the values assoc t.ed 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.

4.

Slowest measured average control rod insertion times to specified notches for all operable control rods for each Position nserted From group of 4 control rods arranged in a Full Mithdrawn a two-b -two arrav seconds otch 45 .404 Hotch 39 .660 Notch 25 l. 504 Hotch 5 2. 624

'rlA IHGTOH NUCLEAR - UNIT 2 3/4 2-7 Amendment No. 84

+ CONTROLLED COPY (

Table 3.2.3-1 (Continued)

MCPR GPERATIHG LIMITS

"""The CEIL'LFA fuel, the AHF fuel and the CE ini>tial core el are also monitorekto '>"e AHF 8x8 fuel >, "R Qperatina Limi:s (Refe nca: Power Ois ribu.ion Limits, Bases, 3/4.2. Minimum Cr.t:cal wer Ratio,

p. B 3/4 2-3).

""""For Final Fee~a(atar THberature Reduction r a conditi>ons beyond all &ds cut point, add .02 to the MCPR for all fu in th 'P-2 core except >or t>".e SVEA-46 LFA fuel. For the SVEA-96 A fuel, add . ". to the MCPR for Final Feedwatar Temperat re Reduc i rated conditions be d the all rcds cut po >nz.

MASiiIHCTQH >'IUCLEAR " UHIT 2 3/4 2-7a Amendment >(o.

Vua Loop Operall n PB l.7 Total Core paratfng Flow Rate Llmlt 1.07 E . 90 'f.t3 l.s 80 1.19 CO '1.26 60 ":1 34 a) 14 50 1.45 Q. 1.59 0

fL 1.a 1.2 50 60 70 00 100 Total Coro Flow (% Bated)

Aeduced Flow MCPA Operallno Llmll Tl>ls Curve ls Appllcehla lo ANF Aeload Fuel, GE Inlllal Core Fuel, ANF 9 X 0 LFA Foal, GE 1l LFA Fuel, hand SYEA-96 LFA Fuel TI>ls curve ls also oppllcabla lo FFTA operation Floure 3.2.3-1

0 ~ ~" ~'qiFacD'~ (.l'"J P,c IViop PER

'",. CONTaOLLED COPY a POMER DISTRIBUTIOH LIMITS 3/4.2.4 LIHEAR HEAT GFHERATIOH RATE LIMITIHG CONDITION FOR OPERATION.

QP&CJ+le 3.2.4 The LINER HEAT GENERATIOH RATE (LHGR) e-'I- shall not exceed the values

~ho~ ln I ~I~W j ) ~

0 p~ a.<a~ g g. g C, I op~

APPLICABILITY: OPERATIOHAL COHDITIOH 1, when THERMAL POMER is greater than or equal to 25~ o, RATED THERMAL POMER.

ACTION:

With the LHGR of any fuel rod exec ding the limit, ini aa e corrective action

'irnH

.r.,;<~. C";.or~"~

mf <~~< a Rgov+

within 15 minu-es and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL PG'yiER to less than 2 ~ of RATED THERMAL POMER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURYEILLAHCE R"=.-.UT~rMgHTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:

a. At leas once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Llr ii,t Pep)*

b. Mith'.n "2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after completion cf a THERMAL POMER increase o at lees- "X of RATED THERMAL POMER, and C. Ini+'.ally 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 opera ing on a LIMITIHG COHTROL R00 PATTERN for LHGR.

WASHINGTON HUCLEAR UNIT 2 Amendment Ho. 84

EXP GA 0 5 5.62

. 540 45.625 2,500 15.10 5,230 14.7$

7,940 14'-1 9

$ 0,470 't 4. t 3.;( ~

co 12 "

K 43,220 14.05 0 15,990 14.06 10,700 0 4.00.

"'SQ 21,590 13.93 4l 24)420 13.93 U Permissible 27,200 13.00 cl Region nf

" 30,150 52.24 a)

Oporatlo 83,050 11.40 8 35,960 5 0.47 Cl

0) 30,900 9.55 C

CC,Oao "

6.65 44)760 7.77 0 i0,000 20,000 80,000 40,000 5, 0 Average Planar Exposure (MWD/MT)

ANF OxO A+toad Fuel Linear tleat Ge>>eratlon erato (LttGA) Limit Versus Average Planar Exposure Ftgure 3.2.4-$

~ i, g

's a ~

Dandlo Avcfooo-13 Exposnro LI lGA fQWA/M'Q owlf) a 13.7 5,0no 13.

10,000 .7

'l5,000 . 13.7 20,000 -13.0 3o,ono 11.5 1a.a

",000 a.5 1a . '.0 Gn,ono

'7o,non 5.5 0

7 5.Goo 1n.noo 2o.noo. 3o.ooo ao.non 5o.ooo Go,aoo 70,0oo no,ooa Average Planar Exposure (MWD/MT)

ANF 0 X 9- IX Helond Fuel Linear I-lent Generation Ante (LHGB) Limit Versus Average Planar Exposure Rrg))re 3.2A-2

Dundlo Avofogo Expoouco LllGA

~ r <<altoona hdll 0 13.1 5000 13.1 10>000

~ ~ 15,500 .1 E 20,000 12.5 30,000 ~ 11.2 a) 4n,oo 0.9 cd 5 0 f) 0,000 7.3 C: 70,000 6.1 0

cd n)

C a>

(3 cd rl) cd a)

C o 5.oon 1o.ooo 2o.ooo 3o,nno co,nno 5o,noo 50,000 . 7o,ooo Average Planar Exposure (MMlD/MT)

ANF 9 X 9- QX Beload F<lel I lnear I-leat Generation Bate (LHGB) Limit.

Versus Average Planar Exposllre

2 I/l C)

Exposur 'LfdGR Cl I

(Mwo/ 0) (kVl/tf) fn

ÃI E 0 fo 40,000

0) l1 rg K ~ ~ ~

l 0

a>

C 0) 10 lg a>

x V

C:

l0000 20000 30000 400 CD Average Plnnar Exposure (MND/MT)

D A.

9 40 D Linear l leal Generallon Bale (LINB) Llmll Vereua Averao ~ Planar Exposure SVEA 96 Lead Fuel Aaaen>bllea O

Floure 3.2.4-l

LflGB 0 13.1 F. 510 13.1

'2 2,500 12.7 5,230 12.3 7,040 11.Q 10,470 11.8 E

l1 13,220 1L8

4) 15,890 1 l.b 18,708 1 l.7

)0- 2 l 400 11.7 2i,420 11.7 0

27,28) ',.'. ll 11.0 T'1 30,150 10.3 c a 33,050 9.8 (3 35,06/ 8.0 38,000 $ .0 tO a ~

4) 41,830 7.3 L

44,7GD 8.5 Cl iD G

2 $ 000 000OI Avornoo Planar Exposura (MNDlMTj Linear l leal Generalfo11 Ante (LltGR) Llmll Vorsua Average Plalinr Exposure 06 1$ Lead Fuel hsaewbllea 0

POWER DISTRIBUTION LIMITS 3/4. 2. 6 POWER/FLOW INSTABILITY LIMITING CONDITION FOR OPERATION 3.2.6 Operation with THERMAL POWER/core flow conditions which lay in Region A of Figure 3.2.6-l is prohibited.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than 559A TIIERRAL PALTER d f1 I I tt I I IER f t d core flow.

ACTION:

With THERMAL POWER/core flow conditions which lay in Region A of Figure 3.2.6-1, then as soon as practical, but in all cases within 15 minutes, initiate a MANUAL SCRAM.

SURVEILLANCE RE UIREMENTS

4. 2. 6 The THERMAL POWER/core flow conditions shall be verified to lay outside Region A of Figure 3. 2. 6-1 once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when operating in the region of APP LICAB I LITY.

WASHINGTON NUCLEAR - UNIT 2 3/4 2% Amendment No. 71

70 egion A 60 K

~o 50 40 E

e 30 I-20 10 0

20 30 40 50 60 70 Core Flow (% Rated)

Operating Region Limits of Specification 3.2.6 Figure 3.2.6-1

3/4. 2 POWER DISTRIBUTION LIHITS 3/4. 2. 7 STABILITY HONITORING TMO LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.2.7 The stability monitoring system shall be operable~ and the decay ratio of the neutron signals shall be less than .75 when operating in the region of APPLICABILITY.

APPLICABILITY: OPERATIONAL CONDITION 1, with two recirculation loops in operation and THERMAL POMER/core flow conditions which lay in Region C of Figure 3.2.7-1.

ACTION:

a. Mith decay ratios of any two (2) neutron signals greater than .75 or with two (2) consecutive decay ratios on any single 'neutron signal greater" than .75:

As soon as practical, but in all cases within 15 minutes, initiate action to reduce the decay ratio by either decreasing THERHAL POWER with control rod insertion or increasing core flow with recirculation flow control valve manipulation. The starting or shifting of a recirculation pump for the purpose of decreasing decay ratio is specifically prohibited.

b. With the stability monitoring system inoperable and when operating in the region of APPLICABILITY:

As soon as practical, but in all cases within 15 minutes, initiate action to exit the region of APPLICABILITY by either decreasing THERMAL POMER with control rod insertion or increasing core flow with recirculation flow control valve manipulation. The starting or shifting of a recirculation pump for the purpose of exiting the region of APPLICABILITY when the stability monitoring system is inoperable is specifically prohibited. Exit the region of APPLICABILITY within one (1) hour.

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

4.2.7.2 The stability monitoring system shall be demonstrated operable~

within one (1) hour prior to entry into the region of APPLICABILITY.

4. 2. 7. 3 Decay ratio and peak-to-peak noise values calculated by the stability monitoring system shall be monitored when operating in the region of APPLICABILITY.

"Verify that the stability monitoring system data acquisition and calculational modules are functioning, and that displayed values of signal decay ratio and peak-to-peak noise are being updated. Detector levels A and C (or B and D) of one LPRH string in each of the nine core regions (a total of 18 LPRH detectors) shall be monitored. A minimum of four (4) APRHs shall also be monitored.

MASHINGTON NUCLEAR - UNIT 2 3/4 2~T 7 Amendment No. 71 I

70 Region A 60 K

o 50 Region C a 40 E

I 30 I-U 20 10 0

20 30 40 50 60 70 Core Flow (% Rated)

Operating Region Limits of Specification 3.2.7 Figure 3.2.7-1

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.8 STABILITY MONITORING - SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.2.8 The stability monitoring system shall be operable~ and the decay ratio of the neutron signals shall be less than .75 when operating in the region of APPLICABILITY.

APPL'ICABILITY: OPERATIONAL CONDITION 1, with one recirculation loop in p EEEERWALR WAR/ ff Eff Rfpfyf R f p f Figure 3.2.8-1.

ACTION:

a 0 With decay ratios of any two (2) neutron signals greater than .75 or with two (2) consecutive decay ratios on any single neutron signal greater than .75:

As soon as practical, but in all cases within 15 minutes, initiate action to reduce the decay ratio by either decreasing THERMAL POWER with control rod insertion or increasing core flow with recirculation flow control valve manipulation. The starting or shifting of a recirculation pump for the purpose of decreasing decay ratio is specifically prohibited.

b. With the stability monitoring system inoperable and when operating in the r egion of APPLICABILITY:

As soon as practical, but in all cases within 15 minutes, initiate action to exit the region of APPLICABILITY by decreasing THERMA POWER with control rod insertion. Exit the region of APPLICABILITY within one (1) hour.

SURVEILLANCE RE UIREMENTS 4.2.8.1 The provisions of Specification 4.0.4 are not applicable.

4.2.8.2 The stability monitoring system shall be demonstrated operable" within one (1) hour prior to entry into the region of APPLICABILITY.

4.2.8.3 Decay ratio and peak-to-peak noise values calculated by the stability monitoring system shall be monitored when operating in the region of APPLICABILITY.

~Verify that the stability monitoring system data acquisition and calculational modules are functioning, and that displayed values of signal decay ratio and peak-to-peak noise are being updated. Detector levels A and C (or B and D) of one LPRM string in each of the nine core regions (a total of 18 LPRM detectors) shall be monitored. A minimum of four (4) APRMs shall also be monitored.

WASHINGTON NUCLEAR " UNIT 2 3/4 2~ Amendment No. 71

70 Region A 60 K egion C o 50 Region B o 40

-E e 30 I-u 20 10 20 30 40 50 60 70 Core Flow (% Rated)

Operating Region Limits of Specification 3.2.8 Figure 3.2.8-1

3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTBI RECIRCULATION LOOPS LIMITING CONOITIOH FOR OPERATION

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

APPl ICABILITY: OPERATIOHAL CONDITIONS 1~ and 2~.

ACTION:

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

l. 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 8 of Figure 3.4.1.1-1.

Mith THERMAL POWER/core flow conditions which lay in Region 8 of Figure 3.4.1.1-1, as soon as practical, but in all cases within 15 minutes, initiate action to exit Region 8 by either decreasing THERHAL POMER with control rod insertion or increasing core flow with flow control valve manipulation.

Mithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months exit Region B. The starting or shifting of a recirculation pump for the purpose of exiting Region 8 is specifically prohibited.

3. 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 (Posit,ion 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)

Rate (tlAPLHGR) for gegegal Electric fuel limit to ~~

Reduce the l1aximum Average Planar Linear Heat Generation and, SPIC>f F50 I< ~H5 CoAG o pc nhVlg C tl WITS P &P<R 7 d) Reduce the volumetric flow rate of the operating recircula-tion loop to < 41,725"" gpm.

'Se Special Test Exception 3.10.4.

~~This value represents the actual volumetric recirculation loop flow which produces 100'. core flow at 100K THERMAL POWER. This value was determined durina the Startup Test Program.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-1 Amendment Ho. >l

I'

~ ~

vQNTROLLED COPY 3/4.2 POMER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following'he 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 PLAHAR 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 rite 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 PLAHAR LINEAR HEAT GENERATION RATE (APLHGR) for GE fuel is .

this LHGR of the highest powered rod divided by its local peaking factor which results in a calculated LOCA PCT much less than 2200'F. The Technical Speci-fication APLHGR for AHF 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 4, td d ~ 1 Cl

<<'l I I cl I

+y ~ glF/Po IA 7VC 4'0+~

'AP 1 c v pc $ 4TI'l(p 8, i& f7' E P4'R.V O

j'dld

- ~ .

I I I

-' d d

. ~

I Adlldll 5 p+clFIGO s based on a g< ~gg c- colt, c loss-or-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 @~~M in, l m ) 7 y g $ +~~?\4M 554Tiom g.g p oF i hE YscH. 5 pic MAS8INGTON NUCLEAR " UNIT 2 8 3/4 2-} Amendment Ho. 7l

POMMER OISTRIBUTIOH LIMITS BASES 3g4 2 3 MTHIMUM CRITICAL POMER RATIO The required operating limit HCPRs at steady-stat operati>ng conditions as specified in Soecificaticn 3. 2. 3 are derived from the established fuel cladding integrity Safety Limiit HCPR and an. analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial condi>-

tion o> the reactor beino at the steady-state operating limit, it is requir d that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transien assuming instr>Jm>, ent trip setting given in Specifica-tion 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnor>>>al operational transient, the most limiiting tran-sients have been analvzed to Ce.ermine which result in the largest reduction in CRITICAL POMER RATIO (CPR). The type of transients evaluated were less o>

flow, irscrease in pressure and power, positive reactivity inser ion, and coolant temperature decrease. The limiting,transient yields the laroes delta MCPR.

Shen added to the Safety Limit HCPR, the required minimum operating limit MCPR of Specifi>cat;on 3. 2. 3 is '-' " - ". " " ". ztzctl->> 0 '

eOnz OPCaa~>ua ~i~>7 S PZPde~.

The evalua ion of a given transient begins witn the system ini .ial param-eters shown in the cycle specific transient analysis report that are input to an AHF core dynamic behavior transient corn>puter prcgram. The outputs of this l program a',ong wi h he initial MCPR form the input >or fur.her analyses of the and nonpressuri-ation events are trans i ent. 1

~~

thermallv limiting bundle. The codes and methodology to evaluate pressurization Crs. V'v ms.wc>

in principal rasuit o-, this evaluaticnTis the reduc-'ice/in MCiR c=-used.by the

...'he SBCVIOhJ ~, g, p 9(- Spy s C u. 4 +p L C pd caw PEPENog>lr 5 Pg" F yN Tgg cd> gc o p~

NCPP,-:..

~

The purpose of the 1 is .o define operating'~i>~

T limits at other than rated core flow conditions. At less than 100~ of rated assures>>at tie flew the r equired MCPR is the ma" imum of the rated flow MCPR Safe y Limit MCPR

,'.Cl!

will 8>pru not. be

> ev (PICO

'!,-ju ..

We>"-

violated. MCPR.T is only cal-

!!P!!.

r-"uxC cp S'C CW 7 Sa g cula ad for the manual flow control mode. Automatic flow ccrtrol ooeration is not permitted.

MASHIHGTOH HUCL""AR " UHIT 2 B 314 2-3 Am> nor:.~~~. N~

ADMINISTRATIVE CONTROLS THIRTY DAY WRITTEN REPORTS

6. 9. 1. 9 DELETED ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.10 Routine Radiological Environmental Operating Repo'rts covering the operation of the unit during the pi'ev',ous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a compar-ison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by Specification 3. 12.2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations speci-fied in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiological envi ronmental monitor ing program; at least two legible maps" covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Inter laboratory Comparison Program, required by Specification 3.12.3; discussion of all deviations from the sampling schedule of Table 3.12-1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achi evabl e.

"One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

IZ Amendment No. 5 WASHINGTON NUCLEAR UNIT 2 6-W

ADHINISTRATIVE CONTROLS SEHIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.11 Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Heasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Haterials in Liquid and Gaseous Effluents from Light-Mater-Cooled Nuclear Power Plants,"

Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix 8 thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direc-tion, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability." This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from=the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to HEHBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5. 1-3) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall. be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION HANUAL (ODCH).

The Radioactive Effluent Release Report shall also include once a year an assessment of radiation doses to the likely most exposed HEHBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977,

In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

WASHINGTON NUCLEAR - UNIT 2 6W

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ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)

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 determined by measurement or estimate),

1 C. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large guantity), and Solidification agent or absorbent (e.g., cement, urea formaldehyde).

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 effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3. 12.2.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

p44 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.

The analytical methods used to determine the core operating limits shall be those topical reports and those revisions and/or supplements of the topical report previously reviewed and approved by the HRC, which describe the methodology applicable to the current cycle. For WNP-2 the topical reports are:

ANF-1125(P)(A), and Supplements 1 and 2, "ANFB Critical Power Correl ation", April 1990

2. Letter, R. C. Jones (HRC) to R. A. Copeland (AHF), "HRC Approval of ANFB Additive Constants for AHF 9x9-9X BWR Fuel", dated November 14, 1990 3~ XH-NF-524(P)(A), Revision 2 and Supplements 1 and 2, "Exxon Nuclear Critical Power Methodology for Boiling Water Reactors",

November 1990 AHF-913(P)(A), Volume 1, Revision 1 and Volume 1, Supplements 2, 3 and 4, '"COTRAHSA 2: A Computer Program for Boiling Water Reactor Transient Analysis", August 1990

5. AHF-CC-33(P)(A), Supplement 2, "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option", January 1991.

6. XH-NF-80-19(P)(A), Volume 1, Supplements 3 and 4, "Exxon Nuclear Methodology for Boiling Water Reactors", November 1990

7. XN-HF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads", June 1986

8. XH-HF-80-19(P)(A), Volume 3, . Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description", January 1987

9. XN-NF-85-67(P)(A), Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump Boiling Water Reactor Reload Fuel",

September 1986

10. ANF-89-014(P), "Generic Mechanical Design for'NF 9x9-IX and 9x9-9X BWR Reload Fuel", May 1989 AHF-89-014(P), Supplement 1, "Generic Mechanical Design of AHF.

9x9-IX and 9x9-9X BWR Reload Fuel", June 1990

12. (SER for 9x9 mechanical design)

13. XN-HF-81-22(P)(A), "Generic Statistical Uncertainty Analysis Methodology", November 1983

14. HEDE-24011-P-A-6, "General El ectric Standard Application for Reactor Fuel",'pril 1983

6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits; core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

6.9.3.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle, to the HRC Document Control Desk with copies to the Regional Administrator. and Resident Inspector.

6. 10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10 C ode of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6. 10.2 The following records shall be retained'for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level.

b. 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.

MASHINGTON NUCLEAR - UNIT 2 6-22

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