ML17285A360

Proposed Tech Specs 3/4.2.6, Power Flow Instability & 3/4.2.7, Neutron Flux Noise Monitoring.

ML17285A360
Person / Time
Site:	Columbia
Issue date:	03/31/1989
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17285A359	List: ML17285A358 ML17285A360
References
TAC-72924, NUDOCS 8904100263
Download: ML17285A360 (52)

Text

INDEX DEFINITIONS

.SECTION 1.0 DEFINITIONS PAGE 1.1 ACTION..~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~l-l 1.2 AVERAGE BUNDLE EXPOSURE.1.3 AVERAGE PLANAR EXPOSURE....................

1.4 AVERAGE PLANAR LINEAR HEAT GENERATION RATE l.5 CHANNEL CALIBRATION.

1.6 CHANNEL CHECK.~~~~~~~~~i~~~~~~~~~~~~~~~~~~~~~~~~~~~1 1 1.7 CHANNEL FUNCTIONAL TEST l.8 CORE ALTERATION.

1.9 CRITICAL POWER RATIO.1.10 DOSE E(UIVALENT I-131 1-2 1" 2 1-2 1-2 1.11 K-AVERAGE DISINTEGRATION ENERGY............................

1" 2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS)RESPONSE TIME.........

1-2 1.13 END"OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME..1-3 1.14 FRACTION OF LIMITING POWER DENSITY 1.15 FRACTION OF RATED THERMAL POWER..1.16 FRE(UENCY NOTATION.........

1-3~~~~~~~~4~1-3 1-3 1.17 GASEOUS RADWASTE TREATMENT SYSTEM..........................

1-3 1.18 IDENTIFIED LEAKAGE.)lg QNMBQXHT'O'J.p

..-~--~~1.M ISOLATION SYSTEM RESPONSE TIME 1.20 LIMITING CONTROL ROD PATTERN.~Ql 1.M LINEAR HEAT GENERATION RATE 1.g2 LOGIC SYSTEM FUNCTIONAL TEST cF3 1.M MAXIMUM FRACTION OF LIMITING POWER DENSITY..1-3 I-9 1-M WASHINGTON NUCLEAR-UNIT 2 p13R 85 o3p1 85'oooo Aaoe<osooosyz F'aC}Amendment No.28 1\'I I pt I I 4 INDEX DEFINITIONS SECTION DEFINITIONS (Continued) 1.W MAXIMUM TOTAL PEAKING FACTOR.1.29 MEMBER(S)OF THE PUBLIC.1.46 MINIMUM CRITICAL POWER RATIO.'a7 1.E7 OFFSITE DOSE CALCULATION MANUAL'a5 1.48 OPERABLE" OPERABILITY..............,'pv 1.29'PERATIONAL CONDITION-CONDITION.

1.30'HYSICS TESTS..........

9(1.Zl PRESSURE BOUNDARY LEAKAGE.1.~PRIMARY CONTAINMENT INTEGRITY............

89 l.~PROCESS CONTROL PROGRAM..........

1.34 PURGE-PURGING.'55 l.~RATED THERMAL POWER.1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME..97 1.%'EPORTABLE EVENT 98 1.38 ROD DENSITY...

99 1.~SECONDARY CONTAINMENT INTEGRITY..........

+0 1.40 SHUTDOWN MARGIN.fl 1.M SITE BOUNDARY.1.42 SOLIDIFICATION.

99 1.43 SOURCE CHECK...........

1A4" STAGGERED TEST BASIS 1AS THERMAL POWER..1.46 TOTAL PEAKING FACTOR.'f7 WASHINGTON NUCLEAR-UNIT 2 PAGE 1" 6 Amendment No.28 l I ,*"+q<II 4 INDEX DEFINITIONS SECTION DEFINITIONS (Continued) 1.47 TURBINE BYPASS SYSTEM RESPONSE TIME.....Y8 1.48 UNIDENTIFIED LEAKAGE 99 1A9 UNRESTRICTED AREA.................

90 1.5B VENTILATION EXHAUST TREATMENT SYSTEM.gl 1.&3: VENTING....

$2 PAGE MASHINGTON NUCLEAR" UNIT 2 Amendment No.28 l Ft 1~4 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.0 APPLICABILITY.................

3/4.1 REACTIVITY CONTROL SYSTEMS PAGE 3/4 0"1 3/4.1.1 SHUTDOWN MARGIN....................

........3/4 l-l 3/4.1.2 REACTIVITY ANOMALIES......................

3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability................................

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 LIQUID CONTROL SYSTEM...........

~..............

3/4 1-19 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 SETPOINTS.................................

3/4 2-5 3/4.2.3 MINIMUM CRITICAL POWER RATIO..........................

3/4 2-6 3/4.2.4 LINEAR HEAT GENERATION RATE...................

3/4 2-9 3/4.2.5 (RESERVED FOR FFTR)3/4.2.6 POWER/FLOW INSTABILITY.

5VFl8XJ.X'T Y 3/4.2.7 NBRRONWtU~OTSE MONITORING.

+~" 0~OP~+.g.g S~~eZ~X7-y rnO~r~oarN6" SXNQE COopaPHPHTS'oN WASHINGTON NUCLEAR" UNIT 2 v 3/4 2-11 3/4 2-13 3/y 8"/5 Amendment No.62 I pl 1 J e INDEX BASES SECTION" 3/4.0 APPLICABILITY..........

3/4.1 REACTIVITY CONTROL SYSTEMS PAGE 8 3/4 0"1, 3/4.1.1 SHUTDOWN MARGIN.......................

..........

B 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES.............................

8 3/4 1-1 3/4.1.3 CONTROL RODS..........................

B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTROLS B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.....,..............

B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION ATE~~~~~~~~~~~~~~~~~~~R~~~~~~~~o~~~~~~~~~B 3/4 2 1 3/4.2.2 APRM SETPOINTS................

~..............

B 3/4 2-2 3/4.2.3.MINIMUMCRITICAL POWER RATIO.....................

B 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE..~..................

B 3/4 2-4 3/4.2.5 (RESERVED fOR fFTR)3/4.2.6 POWER/FLOW INSTABILITY...........................

B 3/4 2"4 S<~a~zrY 3/4.2.7 NEUVRQ~M~KSB MONITORING".X~A

~K~.9PP~~.

+B 3/4 2"5 gq.>.8 BTA~LSTY R0NilotsNG-szMGLE'ooP oPARWQOg 8~/v Q-5 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION........

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

B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..............

..................

B 3/4 3"2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION......................

B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION........................

B 3/4 3-4 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION..................................

B 3/4 3"4 WASHINGTON NUCLEAR" UNIT 2 X11 Amendment No.62 "4~~0 P r J'N 1 tt JY LIST OF FIGURES INDEX FIGURE 3.1.5-1 3.1.5-2 SODIUM PENTABORATE TANK, VOLUME VERSUS CONCENTRATION, RE(UIREMENTS.........................................

3/4 1"22 PAGE'ODIUM PENTABORATE SOLUTION SATURATION TEMPERATURE...

3/4 1"21 3.2.1-1 3.2.1 2 3.2.1-3 3.2.1-4 3.2.1" 5 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183.......................

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

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)VERSUS AVERAGE BUNDLE EXPOSURE ANF Bx8 RELOAD FUEL..................................

MAXIMUM'AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183................................

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

3/4 2-2 3/4 2"3 3/4 2-4 3/4 2-4A 3/4 2-4B 3.2.3-1 3.2.4-1 3.2.6-1 REDUCED FLOW MCPR OPERATING LIMIT...................

3/4 2-8 LINEAR HEAT GENERATION RATE (LHGR)LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF Bx8 RELOAD FUEL..........

3/4 2-10 OPERATING REGION LIMITS OF SPEC.3.2.6...............

3/4 2"12 3.2.7-1 3.2.8-I 3.4.1.1"1 OPERATING REGION LIMITS OF SPEC.3.2.7.......

opzgb~g@Racmx xxmrrs oF sp~c....~THERMAL POWER LIMITS OF SPEC.3.4.1.1-1..............

3/4 2-14~h s-ic 3/4 4"3a 3.4.6.1-1 3.4.6.1-2 4.7" 1 3.9.7" 1 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 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (INITIAL VALUES)......

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

3/4 4-21 SAMPLE PLAN 2)FOR SNUBBER FUNCTIONAL TEST..........

3/4 7-15 WASHINGTON

'NUCLEAR" UNIT 2-XX Amendment No.63 D El-INITIONS END-OF"CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to energization of the recirculation

'pump circuit breaker trip coil from when the monitored parameter exceeds its trip setpoint at the channel sensor of the associated:

a.Turbine throttle valves channel sensor contact opening, and b.Turbine governor valves initiation of valve fast closure.The response time may be measured by any series of sequential, overlapping or total steps such that the'entire response time is measured.FRACTION OF LIMITING POWER DENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD)shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type.~FRACTION OF RATED THERMAL POWER 1.15 The FRACTION OF RATED THERMAL POWER (FRTP)shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.~RttRUE i t 1.16 The FRE(UENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table l.l.GASEOUS RADWASTE TREATMENT SYSTEM 1.17 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.18 IDENTIFIED LEAKAGE shall be: a.Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or b.Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the opera-tion of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.ISOLATION SYSTEM RESPONSE TIME 1.X9 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when gO the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions.

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

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.WASHINGTON NUCLEAR-UNIT 2 1-3 Amendment No.28 INSERT A IMMEDIATELY 1.19 ItNEDIATELY shall mean as soon as practical within the constraints of maintaining status and control of the reactor, core and plant systems, but in all cases within 15 minutes.

1 g~

DEFINITIONS LIMITING CONTROL ROD PATTERN 1.20'LIMITING CONTROL ROD PATTERN shall be a pattern which results in the gl core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.LINEAR HEAT GENERATION RATE 1.2X LINEAR HEAT GENERATION RATE (LHGR)shall be the heat generation per unit gg length of fuel rod.It is the integral of the heat flux over the heat transfer area associated with the unit length.LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components,i.e., all relays and contacts, all trip units, solid state logic elements, etc, of a logic circuit, from sensor through and including the actuated device, to veri.fy OPERABILITY.

The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.23'he MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD)shall be A highest value of the FLPD which exists in the core.MAXIMUM TOTAL PEAKING FACTOR 1.%The MAXIMUM TOTAL PEAKING FACTOR (MTPF)shall be the largest TPF which g5 exists in the core for a given class of fuel for a given operating condition.'

MEMBER(S OF THE PUBLIC 1.25 MEMBER(S)OF THE PUBLIC shall include all persons who are not occupationally

~associated with the plant.This category does not include employees of the utility, its contractors or vendors.Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recreational, occupational or other purposes not.associated with the plant.MINIMUM CRITICAL POWER RATIO 1.46 The MINIMUM CRITICAL POWER RATIO (MCPR)shall be the smallest CPR which , g7 exists in the core.OFFSITE DOSE CALCULATION MANUAL 1.gi'he OFFSITE DOSE CALCULATION MANUAL (ODCM)shall contain the currentmethodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints and in the conduct of the environmental radiological monitoring program.WASHINGTON NUCLEAR-UNIT 2 Amendment No.28 DEFINITIONS OPERABLE-OPERABII ITY 1.28 A system,'ubsystem; train, component or device shall be OPERABLE or haveOPERABILITY when it is capable of performing its specified function(s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL CONDITION" CONDITION 1.25 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive W combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.PHYSICS TESTS 1.30'PHYSICS TESTS shall be those tests performed to measure the fundamental 81'nuclear characteristics of.the reactor core and related instrumentation as (1)described in Chapter 14 of the FSAR, (2)authorized under the provisions of 10 CFR 50.59, or (3)otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1&I PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault 82 in a reactor coolant system component body, pipe wall, or vessel wall.PRIMARY CONTAINMENT INTEGRITY 1.32'RIMARY CONTAINMENT INTEGRITY shall exist when: ss a.All primary containment penetrations required to be closed during accident conditions are either: 2~Capable of being closed by an OPERABLE primary containment automatic isolation system, or Closed by at least one manual valve, blind flange, or deacti-vated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification 3.6.3.b.C.d.e.All primary containment equipment hatches are closed and sealed.Each primary containment air lock is in compliance with the requirements of Specification 3.6.1.3.The primary containment leakage rates are within the limits of Specification'3.6.1.

2.The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows, or O-rings, is OPERABLE.WASHINGTON NUCLEAR-UNIT 2 1-5 Amendment No.2q

~~t h DEFINITIONS PROCESS CONTROL PROGRAM 1AS The PROCESS CONTROL PROGRAM (PCP)shall contain the sampling, analysis,~5'nd formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.PURGE" PURGING 1.34 PURGE or PURGING shall be the controlled process of discharging air M or gas from a'confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.39'ATED THERMAL POWER shall be a total reactor core heat transfer rate to 9<the reactor coolant of 3323 MWt.REACTOR PROTECTION SYSTEM RESPONSE TIME 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from 87 when the monitored parameter exceeds its trip setpoint at the channel sensor until deenergization of the scram pilot valve solenoids.

The response time may be measured by any series of sequential, over lapping, or total steps such that the entire response time is measured.REPORTABLE EVENT 1.M A REPORTABLE EVENT shall be any of those conditions specified in 38 Secti'on 50.73 to 10 CFR Part 50.ROD DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a 9f fraction of the total number of control rod notches.All rods fully inserted is equivalent to lOOX ROD DENSITY.SECONDARY CONTAINMENT INTEGRITY 1.39 SECONDARY CONTAINMENT INTEGRITY shall exist when: 90 All secondary containment penetrations required to be closed during accident conditions are either:.1.Capable of being closed by an OPERABLE secondary containment automatic isolation system, or b.2.Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.All secondary containment hatches and blowout panels are closed and sealed.c.The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.WASHINGTON NUCLEAR" UNIT 2 1-6 Amendment No.28 4 Id."

DEFINITIONS SECONDARY CONTAINMENT INTEGRITY (Continued) d.At l~ast one door in each access to the secondary containment is'losed.e.The sealing mechanism associated with each secondary containment penetration, e.g., welds, bellows, or O-rings, is OPERABLE.f.The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.SHUTDOWN MARGIN 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is Wl subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e., 68'F;and xenon free.SITE BOUNDARY 1.4X'he SITE BOUNDARY shall be that line beyond which the land is not owned, F~leased, or otherwise controlled by the licensee.SOLIDIFICATION 1.42 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid gg systems to a<<homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

SOURCE CHECK 1.43 A SOURCE CHECK shall be the qualitative assessment of channel response'R when the channel sensor is exposed to a radioactive source.STAGGERED TEST BASIS lAC A STAGGERED TEST BASIS shall consist of: a e b.A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals.

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

WASHINGTON NUCLEAR-UNIT 2 1-7 Amendment No.28 DEFINITIONS THERMAL POWER 1.45 THERMAL POWER shall be the total reactor core heat transfer rate to the'/to reactor coolant.TOTAL PEAKING FACTOR 1.46 The TOTAL PEAKING FACTOR (TPF)shall be the ratio of local LHGR for any 97 specific location on a fuel rod divided by the core average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.TURBINE BYPASS SYSTEM RESPONSE TIME 1.47 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be that time interval from Q when the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

The response time.may be measur ed by any series of sequential, overlapping, or total steps such that the entire response time is measured.UNIDENTIFIED LEAKAGE 1.48'NIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.IV UNRESTRICTED'REA 1A9 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY 50 access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the site boundary used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.VENTILATION EXHAUST TREATMENT SYSTEM 1.50 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents)

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

VENTING 1.81 VENTING shall be the controlled process of discharging air or gas from$Q a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.Vent, used in system names, does not imply a VENTING process.WASHINGTON NUCLEAR-UNIT 2 1-8 Amendment No.28 I

t POWER DISTRIBUTION LIMITS'/4.2.6 POWER/FLOW INSTABILITY LIMITING CONDITION FOR OPERATION 3.2.6 Operation with THERMAL POWER/core flow conditions which lay in the IPegion 8 cemshatched-reg-ice of Figure 3.2.6-1 is prohibited.

APPLICABILITY:

OPERATIONAL CONDITION 1)g en THERMAL POWER is greater than 39K of RATED THERMAL POWER and core flow is less than or equal to 45K of rated.core flow.ACTION: wN Nzpzqgzzg Kaqion 5 With THERMAL POWER/core flow conditions which lay in th-vosshatehed-re~a of Rg 3.2.6-1, i i i'I'B'R~wsav~eii~~

~~I SURVEILLANCE RE UIREMENTS 4.2.6 The THERMAL POWER/core flow conditions shall be verified to lay outside Qg'eo/j'he-crosshatched-rein of Figure 3.2.6-1 once per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sg~hen cperg4i'n~

in%he, region cFr APPL.X'c R6sxgTQ.WASHINGTON NUCLEAR-UNIT 2 3/4 2-11 Amendment No.62 R Q O C R q 4l I 7 BO C g 50 o 40 m so P-I 20 0 0 10 0 20 30 40 50 Core Flow (%Rated)BQ?Q O Ch Operating Region.Llmlts of Speclficatlon 3.2.6 Figure 3.2.6-4 lJ ply~l+~~~t I IW+fgllg'I ll-f8\l~lll A IV r,

~~Q~(~g~Isw&IA'<lk PO I I+~r J NTROLLE0 COPY P WER ISTRIBUTION

-LIMITS 3/4.2;7 NEUTRON FLUX NOISE MONITORING LIMITING C DITION FOR OPERATION 3.2.7 The APR and LPRM neutron flux noise levels shall not exceed t ee (3)times their est lished baseline values when operating in the region f APPLICABILITY.

APPLICABILITY:

OPE TIONAL CONDITION 1 with.THERMAL POWER/core fl w in Region B of Figure 3..7-1, with two reactor coolant system recir ulation loops in operation and total core flow less than 45K of rated to al core flow, or with one reactor coo nt system recirculation loop not in op ration.ACTION: a.If baseline APRM and LPR neutron flux noise levels hy e not been established for the appro riate reactor coolant syst A,condition (one or two loop operation) since e most recent CORE ALTE TION, then: Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exit the regia of APPLICABILITY.

stablish baseline APRM and LPRM neutron flux noise le ls prior to re-e ering Region B of Figure 3.2.7-1.b.If baseline APRM and LPRM neutron lux noise evels have been established for the appropriate reactor coolant ystem c ndition (one or two loop operation) since the most recent COR ALTER TION, then: "With the APRM or LPRM neutron flux nois evels greater than three (3)times their established noise levels, i tiate corrective action within 15 minutes to restore the noise levels o 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 POWE to elow the region of APPLICABILITY within the next 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.SURVEILLANCE RE UIREMENTS 4.2.7.1 The provisions of Specifi tion 4.0.4 are not pplicable.

4.2.7.2 The APRM and LPRM neutr n flux noise levels shal be determined to be less than or equal to three (3 times their established ba eline values: a.At least once per-hours, and b.Within 30 minut s after completion of a THERMAL POWER increase of greater than equal to 5X of rated THERMAL POWER."Detector levels and C of one LPRM string per core octant plus dete tot levels A and C f one LPRM string in the center of the core should be ,monitored.

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

~'

?0 60 K g 50~e~o40 5 30 o 20 0 O 40'-'Pe-,io~

A n.0 C gJ, r, 0 I I U A 0 Q 0 20 30 40 50 Core Flovr (%Rated)60 Operating Region Umlts of Speclf ication 3.2.6 Figure 3.2.6-1 S l Y h R g 1 E 1 4/4.2 POWER DISTRIBUTION LIMITS 3/4.2.7 STABILITY MONITORING

-TWO 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 POWER/core flow conditions which lay in Region B of Figure 3.2.7-1 ACTION: a~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 IMMEDIATELY initiate action to reduce the decay ratio by either decreasing THERMAL POWER with control rod insertion orincreasing 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:

IMMEDIATELY initiate action to exit the region of APPLICABILITY 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 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 REQUIREMENTS:

4.2.7.1 The provisions of Specification 4.0.4 are not applicable.

4.2.7.2.4.2.7.3 4.2.7.4 The stability monitoring system shall be demonstrated operable*within one (1)hour, prior to entry into the region of APPLICABILITY.

The decay ratios from the stability monitoring system shall be monitored following reactivity manipulation when operating in the region of APPLICABILITY.

The decay ratios from the stability monitoring system shall be demonstrated to be less than 0.75 once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when operating in the region of APPLICABILITY.

• 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.

3/4 2-13 70 e0 C 50 Le o+40 N so I o 20 0 Region A~g>pw g n 0~c'3 0 t l ITl U A 0.0 0 20 30 40 50 Core Flow (%Bated)60 70 Operating Region Llmlts of Speclf Ication 3.2.7 Figure 3.2,7-5

.3/4.2 POWER DISTRIBUTION LIMIT 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.

APPLICABILITY:

OPERATIONAL CONDITION 1, with one recirculation loop in pPERE IRE PEIIERI 11 El I I I R gl E 1 Plg 3.2.8-1.ACTION a.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: IMMEDIATELY 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 region of APPLICABILITY:

IMMEDIATELY initiate action to exit the region of APPLICABILITY by decreasing THERMAL POWER via 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 The decay ratios from the stability monitoring system shall be monitored following reactivity manipulation when operating in the region of APPLICABILITY.

4.2.8'The decay ratios from the stability monitoring system shall be demonstrated to be less then 0.75 once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when operating in the region of APPLICABILITY.

• 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 should also be monitored.

3/4 2-15 b

x R Q O R R A I Ol C R q B C g 0 I o 4 I so o 20 Q eglon A Re glon B n 0't I rn U A'.'g C'0 20 30 40 50 Core Flow (%Rated)70 D O Ch Operating Region Llelts of Specification 3.2.7 Figure 3.2.7-1 0 C'

~'~5 II lI I'&A't'" 4h I%

x-R Q O R X C A l 70 Region C eglon O Ch 60 L" 50~40 E ao I o 20 O 10 20 Re r~B 40 50 Core Floe (A Rated)30 8'perating Region Llmlts of Speclflcatlon 3.22'igure 3.2p'-1 6'0 70 0 R 0 U 0 0.0

~-~~, 0~)4~, 1'4 W r A:<<J l 3/4~4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation., APPLICABILITY:

OPERATIONAL CONDITIONS 1" and 2".ACTION: a.With one reactor coolant system recirculation loop not in operation:

Within 15 minutes: a.'hat core flow i s greater than a to 39K of rated c flow or that THERM L-PO R/core flow conditions

" lay below the in re,3.4.1.1-1.

With core flow less than 39K o re flow and THERMAL POWER/core flow cond-ns above the n Figure 3.4.1.1-1, initiate'ct o reduce THERMAL POWER to the line in gure 3.4.1.1-1 or increase core flow to ter than or equal to 39K of rated core flow within the next ur~.L.cv 9 1.t and Verify that.the requirements of LCO 3.2.4 are met, or comply with the associated ACTION statement(s+4hm-Ne

~pea-i-f-4d-tAme

-l4a+ts-.8.%Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s: a)Place the recirculation flow control system in the Local Manual (Position Control)mode, and b)Increase the MINIMUM CRITICAL POWER RATIO (MCPR)Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and, c)Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)for General Electric fuel limit to a value of 0.84 times the two recirculation loop operation limit per Specification 3.2.1, and, d)Reduce the volumetric flow rate of the operating recircula-tion loop to<41,725"" gpm.3 113 1*pp 3.1.4.""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.62 0\~;1 t INSERT B 2.Verify that THERMAL POWER/core flow conditions lay outside Region B of Figur e 3.4.1.1-1.

With THERMAL POWER/core flow conditions which 1 ay in Region B of Figure 3.4.1.1-1, IMMEDIATELY ini ti ate acti on to exi t Regi on B by ei ther 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.

~'s.I'/1 I'l e~e (e REACTOR COOLANT SYSTEM CONTROLLED COPY LIMITING CONDITION FOR OPERATION (Continued ACTION (Continued) e)Perform Surveillance Requirement 4.4.1.1.2 if THERMAL POWER is<25K""" of RATED THERMAL POWER or the recirculation loop flow in the operating loop is<lOX*"~of rated loop flow.f)Reduce recirculation loop flow in the operating loop until the core plate hP noise does not deviate from the estab-lished cole plate EP noise patterns by more than 100K.The provisions of Specification 3.0.4 are not applicable.

+jV.Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.eIiange<<RA AP~b.With no reactor coolant system recirculation loops in operation, ediatel initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.SURVEILLANCE RE UIREMENTS 4.4.1.1.1 With one reactor coolant system recirculation loop not in operation, at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> verify that: a.The recirculation flow control system is in the Local Manual (Position Control)mode, and b.The volumetric flow rate of the operating loop is<41,725 gpm.""""This value represents the actual volumetric recirculation loop flow which produces 100K core flow at 10(C THERMAL POWER.This value was determined during the Startup Test Program."""Final values were determined during Startup Testing based upon actual THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

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

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4 s)lt t'h t gC t 70 80 C 50 g 40<ao o 20 0 20 eglon A R'c i~B 30 glon C 40 50 Core Floe ('k Rated)60 70 0 K 0-U O-0.0 0;Operating Region Llmlts of Speclflcatlon 3 bA,1.Figure 3,~~<-~

h C A'I I%4>>" 1>l, 4 g 1 I POWER DISTRIBUTION LIMITS~I BASES POWER/FLOW INSTABILITY (Continued)

Predicated on the SIL 380 endorsement, MNP-2 has divided the power/flow map on the following boundary lines: l.2 3.5.80K rod line 45K core flow line APRM-rod-NeeW+o~us-3X-pew~

Ioo5 rM lifJP Natural Circulation flow line Minimum Forced Circulation fo normal recirculation lineu.more coriseryohii<

i e joo redline.or a line cl@ini~o.calcoh decay Wbo cFr0.9 Co lCh This, dsv)sson conforms to the SIL 380 recommendatsons,~h-a-3X-power-pena+ted ea-4he-Af%M-rwd-lAodW-i-ne For LCO 3.2.6, the region of concern is bounded by c e-, Heu's, the natural circulation flow line, and the core ow one.a culate decay ratios between the two flow lines and on the APRM rod block line minus 3X must be less than.9.0 eratio in the region between the two flow lines and above is forbidden due to the potential for boiling instabilities.

7vr-th se of annual licensing submittals, a 3X margin f~arLblock line is taken il the opportunity to b's no Technical Specifica-, tion changes under the p 0 CFR 50.59.'his 3X provides margin to assure that vend&ity calcula n an easily support the al.lowable operati son.For'calculational ease the w oundary is linearized Qe een>two points, (24K Flow, 39K Power)and (45K Flow, Power).maerxzvY 3/4.2.7 NBRRONWLUX-NKSE-MONITORING

-P4'0~P OPBgAVZo N At the high power/low flow corner of the operating domain, a small prob-ability of limit cycle neutron flux oscillations exists depending on'combina-tions of,.operating condi,tions (e.g., rod.patterns, power shape).To provide assurance that neutron flux limit cycle oscillations are detected and sup-pressed, APRM and LPRM neutron flux n~e-'i~ebs should be monitored while operating in this region.ski,ea denny re6ios Stability tests at operating BWRs were reviewed to determine a generic region of the power/flow map in which surveillance of neutron flux noise levels should be performed.

A conservative decay ratio of 0.75 was chosen as the basis for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs.This generic region has been determined to correspond to a core flow of less than or equal to 45K of rated core flow and a thermal power greater than that specified in Figure 3.4.1.1-1 (Reference).

Ne flux noise limits are also established to ensure early detection of limit cycle ne o oscillations.

BWR cor es t ical h 5nSeW neutron flux noise cause n o n ow noise.Typical neutron C.flux noise o 1-12K of rated power~pe~-have been reported for ange of low to high recirculation loop flow during bo dual MASHINGTON NUCLEAR-UNIT 2 B 3/4 2-5 Amendment No.62 INSERT C Stability monitoring is performed utilizing the ANNA system.The ANNA system shall be used to monitor APRM and LPRM signal decay ratios when operat-ing in the reg'ion of concern.3/4.2.8 STABILITY MONITORING

-SINGLE LOOP OPERATION The basis for stability monitoring during single loop operation is con-sistent with that given above for two loop operation.

The defined region where surveillance is required is larger for single loop operation due to a potential reduction in the stabilizing effect of forced circulation.

g 4, Il I C

%1"<<v\aO QJ POWER DISTRIBUTION LIMITS~0 CONTROLLED COPY BASES NEUTRON FLUX NOISE MONITORING (Continued) irculation loop operation.

Stability tests at operating BWRs have demon str ed that when stability related neutron flux limit cycle oscillations ccur they suit in peak-to-peak neutron flux limit cycles of 5-10 times the ypical values.Therefore, actions taken to reduce neutron flux noise levels xceeding three (3)imes the typical value are sufficient to ensure early de ction of limit cycle eutron flux oscillations.

Typically, eutron flux noise levels show a gradual incr se in absolute magnitude as core low is increased (constant control rod p tern)with two reactor recirculatio loops in operation.

Therefore, the aseline neutron flux noise level obtained a a specific core flow can be app ed over a range of core flows.To maintain reasonable variation betwe the low flow and high flow ends of the flow rang the range over which a pecific baseline is applied should not exceed 20K of rat core flow with two ecirculation loops in opera-tion.Data from tests and ope ting plants ind ate that a range of 20K of rated core flow will result in approxi tely a 50K-crease in neutron flux noise level during operation with two re'ulatig loops.Baseline data should be taken near the maximum rod line at wh h lfe majority of operation will occur.However, baseline data taken at lower r lines (i.e., lower power)will result in a conservative value, since the neu on lux noise level is proportional to the power level at a given core flo In the case of'ingle loop peration (SLO), the normal neutron flux noise may increase more rapidly whe reverse flow occur in the inactive jet pumps.This justifies a smaller flo range under high flow LO conditions.

Baseline data should be taken at fl intervals which correspon to less than a 50K in-crease in APRM neutron f ux noise level.If baseline da are not specifically available for SLO, th baseline data with two recirculatl loops in operation can be conservativel applied to SLO since for the same core low SLO.pill exhibit higher ne ron flux noise levels than operation with t loops..However, because of rever e flow ch'aracteristics of SLO, the core flow/dr e flow re-lationship is fferent than the two loop relationship and therefo the base-line data fo SLO should be based on the active loop recirculation d ve flow, and not th core flow.Because of the uncertainties involved in SLO a high reverse ows, baseline data should be taken at or below the power speci ed in Fi e 3.4.1.1-1.This will result in approximately a 25K conservative base ne value if compared to baseline data taken near the rated rod line an wi therefore not result in an overly restrictive baseline.value, while oviding sufficient margin to cover uncertainties associated with SLO.WASHINGTON NUCLEAR-UNIT 2 8 3/4 2"6 Amendment No.62 v Oa h