Text

Amend 137 to License NPF-21,increasing Authorized Max Power Level of Reactor from Current Limit of 3,323 Mwt to 3,486 Mwt
	ML17291A776
Person / Time
Site:	Columbia
Issue date:	05/02/1995
From:	Russell W; Office of Nuclear Reactor Regulation
To:	;
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ML17291A775	List: ML17291A774; ML17291A776; ML17291A777;
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	NUDOCS 9505100210
	Download: ML17291A776 (82)
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t UNITED STATES t NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 WASHINGTON PUBLIC POWER SUPPLY SYSTEM OOCKE O. 50-397 NUCLEAR PROJECT NO AME DM NT TO FACI T

OP RATING LICE S Amendment No.

137 License No. NPF-21 l.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by the Washington Public Power Supply System (licensee) dated July 9,

1993, as supplemented by letters of October 8, and October 25, 1993, January 6, January 6, February 2,

May 3, May 13, September 26, and October 12,

1994, complies with the standards and requirements of the Atomic Energy Act of 1954,.as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraphs 2.C.(1) and 2.C.(2) of Facility Operating License No.

NPF-21 are hereby amended to read as follows:*

Page 3 is attached, for convenience, for the composite license to reflect this change.

Please remove page 3 of the existing license and replace with the attached page.

9505l002%0 950502 PDR ADOCK 05000397~

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A

Powe vel The licensee is authorized to operate the facility at reactor. core power levels not in excess of full power (3486 megawatts thermal).

Items in Attachment 1 shall be completed as specified.

Attachment 1 is hereby incorporated into this license.

(2) echn ca S eci ications and viro mental Protection Pla The Technical Specifications contained in Appendix A, as revised through Amendment No. i37 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This amendment is effective as of the date of issuance and is to be implemented prior to startup from the 1995 refueling outage.

FOR THE NUCLEAR REGULATORY CONHISSION Milliam T. Russell, Director Office of Nuclear Reactor Regulation Attachments:

1.

Page 3 of License 2.

Changes to the Technical Specifications Date of Issuance:

Nay 2, 1995

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(3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive,

possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive,

possess, and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not

separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(I) a imum Power Leve The licensee is authorized to operate the facility at reactor core power levels not in excess of full power (3486 megawatts thermal).

Items in Attachment I shall be completed as specified.

Attachment I is hereby incorporated into this license.

(2) Tec nical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No.

137

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TTACHM NT TO CENSE AM NDMENT NT NO.

137 TO FACILITY OPERATING LICENS NO.

NPF-21 DOCKET NO. 50-397 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

The corresponding overleaf pages are also provided to maintain document co'mpleteness.

REHKE ii xx(a) 1-5 1-6 2-4 3/4 2-2 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 3-4 3/4 3-5 3/4 3-15 3/4 3-16 3/4 3-24 3/4 3-43 3/4 3-55 3/4 4-3a 3/4 4-7 3/4 4-18 3/4 4-19 3/4 4-20 3/4 4-21 3/4 4-2la 3/4 4-23 3/4 5-3 3/4 6-1 3/4 6-2 3/4 6-3 3/4 6-4 3/4 6-5 3/4 6-6 3/4 6-33 3/4 6-44 3/4 7-8 ii xx(a) 1-5 1-6 2-4 3/4 2-2 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 3/4 2-10 3/4 3-4 3/4 3-5 3/4 3-15 3/4 3-16 3/4 3-24 3/4 3-43 3/4 3-55 3/4 4-3a 3/4 4-7 3/4 4-18 3/4 4-19 3/4 4-20 3/4 4-21 3/4 4-22 3/4 4-23 3/4 5-3 3/4 6-1 3/4 6-2 3/4 6-3 3/4 6-4 3/4 6-5 3/4. 6-6 3/4 6-33 3/4 6-44 3/4 7-8

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B 3/4 2-5 B 3/4 2-6 B 3/4 3-1 B 3/4 3-3 B 3/4 4-4 B 3/4 4-5 B 3/4 4-6 B 3/4 4-7 8 3/4 5-1 B 3/4 5-2 B 3/4 6-1 B 3/4 6-2 B 3/4 6-3 B 3/4 6-4 6-21 B 3/4 2-5 B 3/4 2-6 B 3/4 3-1 B 3/4 3-3 B 3/4 4-4 B 3/4 4-5 B 3/4 5-1 B 3/4 5-2 B 3/4 6-1 B 3/4 6-2 B 3/4 6-3 B 3/4 6-4 6-21

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INDEX DEFINITIONS SECTIDN

1. 0 DEFINITIONS PAGE 1.1 ACTION.....................................................

1 1 1.2 AVERAGE BUNDLE EXPOSURE....................... "---..."-. ~

1.3 AVERAGE PLANAR EXPOSURE....................................

1-1 1.4 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..-"-""".-." 1 1

1. 5 CHANNEL CALIBRATION........................................

1-1 1.6 CHANNEL CHECK..............................................

1-1 1.7 CHANNEL FUNCTIONAL TEST....................................

1-1 1.8 CORE ALTERATION............................................

1-2 1.8A CORE OPERATING LIMITS REPORT...............................

1-2 1.9 CRITICAL POWER RATIO.......................................

1-2 1.10 DOSE E(UIVALENT I-131......................................

1-2

1. 11 Z-AVERAGE DISINTEGRATION ENERGY............................

1-2

1. 12 EMERGENCY CORE COOLING SYSTEM (ECCS)

RESPONSE TIME.........

1-2 1.12-A END-OF-CYCLE (EOC).......................................

1-2 l.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME..

1-3

l. 13-A FINAL FEEDWATER TEMPERATURE REDUCTION (FFTR).............

1-3 1.14 FRACTION OF LIMITING POWER DENSITY...""...."............

1-3 1.15 FRACTION OF RATED THERMAL POWER............................

1-3

1. 16 FRE(UENCY NOTATION.........................................

1-3 1.~ 17 NOT USEDo ~ o ~ ~ s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~

1 3 1.18 IDENTIFIED LEAKAGE...............""".".................

1-3 1.19 ISOLATION SYSTEM RESPONSE TIME.............................

1-3 1.20 LIMITING CONTROL ROD PATTERN...............................

1-4 1.21 LINEAR HEAT GENERATION RATE................................

1-4 WASHINGTON NUCLEAR - UNIT 2 Amendment.No g. 98

INDEX DE F IN IT IONS

~SECT ON

~ddt d (t tt dt PAG 1.22 LOGIC SYSTEM FUNCTIONAL TEST...............................

1-4 1.23 MAXIMUM FRACTION OF LIMITING POWER DENSITY.................

1-4 1.24 MAXIMUMTOTAL PEAKING FACTOR...............................

1-4 1.25 MEMBER(S)

OF THE PUBLIC....................................

1-4 1.26 MINIMUM CRITICAL POWER RATIO...............................

1-4 1.27 OFFSITE DOSE CALCULATION MANUAL............................

1-4a 1.28 OPERABLE OPERABILITY.....................................

1-5

1. 29 OPERATIONAL CONDITION CONDITION..........................

1-5 1.30 PHYSICS TESTS..............................................

1-5 1.31 PRESSURE BOUNDARY LEAKAGE..................................

1-5 1.31a a

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P 1.32 PRIMARY CONTAINMENT INTEGRITY.....................

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5 1.33 PROCESS CONTROL PROGRAM....................................

1-6 1.34 PURGE PURGING............................................

1-6 1.35 RATED THERMAL POWER........................................

1-6 1.36 REACTOR PROTECTION SYSTEH RESPONSE TIME....................

1-6 1.37 REPORTABLE EVENT...........................................

1-6 1.38 OD DENSITYt ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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1.39 SECONDARY CONTAINMENT INTEGRITY............................

1-6 1.40 SHUTDOWN MARGIN............................................

1-7 1.41 SITE BOUNDARY..............................................

1-7 1.42 OT USED ~

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N 1.43 SOURCE CHECK...............................................

1-7 1.44 STAGGERED TEST BASIS.....................

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WASHINGTON NUCLEAR - UNIT 2 Amendment No. M;98, 137

LIST OF FIGURES

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3.2.4-5 3.2.6-1 3.2.7-1 3.2.8-1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE GEll LEAD FUEL ASSEMBLIES............................................

OPERATING REGION LIMITS OF SPEC. 3.2.6................

PAG Deleted 3/4 2-6 OPERATING REGION LIMITS OF SPEC. 3.2.7................

3/4 2-8 OPERATING REGION LIMITS OF SPEC. 3.2.8....'............

3/4 2-10 3.4.1.1-1 OPERATING REGION LIMITS OF SPEC. 3.4.1.1..............

3/4 4-3a 3.4.6.1A 3.4.6.1B MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE.........;..............

MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE........................

3/4 4-20 3/4 4-21 4.7-1 3.9.7-1 SAMPLE PLAN 2)

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

3/4 7-15 HEIGHT ABOVE SFP WATER LEVEL VS.

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

3/4 9-10 B 3/4 3-8 B 3/4 3-1 REACTOR VESSEL WATER LEVEL............................

B 3/4.4.6-1 (DELETED) 5.1-1 5.1-2 5.1-3 EXCLUSION AREA BOUNDARY...............................

LOW POPULATION ZONE

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UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS..............

5-2 5-3 5-4 WASHINGTON NUCLEAR UNIT 2 xx(a)

Amendment No. 94-, KM%&,137

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DEF INITIONS OPERABLE OPERAB TY 1.28 A system, subsystem,

train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s) and wlien 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 1.29 An OPERATIONAL CONDITION, i.e.,

CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

~PHV I I i ITS 1.30 PHYSICS TESTS shall be those tests performed to measure the fundamental 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.

PR SSUR OUNDA Y AKAG 1.31 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall, or vessel wall.

1.31a Pa (psig) is a the calculated peak containment internal pressure related to design basis accidents, and is equal to 38 psig.

PRIMARY CONTAINME T I TEGRITY 1.32 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a.

All primary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE primary containment automatic isolation system, or 2.

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.

All primary containment equipment hatches are closed and sealed.

c.

Each primary containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

d.

The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e.

The suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

f.

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. 28," 37

DEF IN ITIONS PROC SS CONT 0 P

OG 1.33 The PROCESS CONTROL PROGRAH (PCP) shall contain the current formulas,

sampling, analyses, test and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61 and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

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1.34 PURGE or PURGING shall be the controlled process of discharging air 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.

RAT D H RHAL POW 1.35 RATED THERHAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3486 HWt.

E CTOR PROT C

ON SYST S

ONS T

1.36 REACTOR PROTECTION SYSTEH

RESPONSE

TIHE shall be the time interval from 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, overlapping, or total steps such that the entire response time is measured.

~RNP R ARI I NT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

~000 RIN IT 1.38 ROD DENSITY shall be the number of control rod notches inserted as a

fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 100X ROD DENSITY.

SECONDARY CONTAINHENT TEGRITY 1.39 SECONDARY CONTAINHENT INTEGRITY shall exist when:

a.

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 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.

b.

All secondary containment hatches and blowout panels are closed and sealed.

Amendment No. iM-,98,"37 c.

The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.

WASHINGTON NUCLEAR UNIT 2 1-6

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.

APPLICABILITY:

As shown in Table 3.3.1-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

WASHINGTON NUCLEAR " UNIT 2 2-3

Vl CD C

OR PROTEC ON ST NS RU ONSTO 2.

Average Power Range Monitor:

TNILNEP0 T

g 120/125 divisions of full scale

~FNNUTIONNL U 1.

Intermediate Range Monitor, Neutron Flux - High ALLOWABLE MERE g 122/125 divisions of full scale a.

Neutron Flux-High, Setdown b.

Flow Biased Simulated Thermal Power - High g 15X of RATED THERMAL POWER

~ 20X of RATED THERMAL POWER

1) Flow Biased

2) High Flow Clamped c.

Fixed Neutron Flux - High d.

Inoperative 3.

Reactor Vessel Steam Dome Pressure High 4.

Reactor Vessel Water Level Low, Level 3

5.

Main Steam Line Isolation Valve - Closure 6.

DELETED

~ 0.58W + 59X, with a maximum of 6 113.5X of RATED THERMAL POWER g 118X of RATED THERMAL POWER N.A.

5 1060 psig h 13.0 inches above instrument zero 5 10.0X closed 5 0.58W + 62X, with a maximum of g 114.9X of RATED THERMAL POWER 6 120X of RATED THERMAL POWER N.A.

g 1074 psig I

2 11.0 inches above instrument zero 5 12.5X closed O

See Bases Figure B 3/4 3-1.

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel shall not exceed the limits specified in the Core Operating Limits Report.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than AREAR T RATER TRERRAE POPPER.

ACTION:

With an APLHGR exceeding the limits specified in the Core Operating Limits Report, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to 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

<.2. 1 411 AoL"'"-Rs shall be veri !e" '.o be equal o "." less than the limits specified in the Core Operating Limits Report.

a.

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

C.

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 Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

WASHINGTON NUCLEAR - UNIT 2 3/4 2-1 Amendment No. 94

PO R

D S RISUT ON IMITS 3 4, RM S TPO NTS LI TING CON ITION 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 (S,) shall be established according to the following relationships:

~TR1 EE Ol T S g (0.58W + 59X)T S

g (0.58W + 50X)T

~OBAB AL S g (0.58W + 62X)T SR~ 5 (0.58W + 53X)T where:

S and S~ are in percent of RATED THERMAL POWER, W - Loop re'circulation 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 l.

YLPPPI ABILITY:

OPEAATIOAAL OOEOITIOE I, I

TBEBIAL POIIEE I B

TA equal to 25X of RATED THERMAL POWER.

A~CT 0:

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 S

as above determined, initiate corrective action within 15 minutes and adjust

.S and/or S

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 TERNAL POWER to less than 25X 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 thermal 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 15X 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 90X 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 100X times MFLPD, provided that the adjusted APRM reading does not exceed 100X of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

WASHINGTON NUCLEAR UNIT 2 3/4 2-2 Amendment No. 94,i37

POWER D S RIBUTION LIM TS 3 4.2.6 POW 0

S ILITY LIMITING CONDITION FOR OPERATION 3.2.6 Operation with THERMAL POWER/core flow conditions which lay in Region A

of Figure 3.2.6-1 is prohibited.

EPPLICNNILITT:

OPENATIGNAL CIINOITIGN I, h

TIIEIINAL POIIEE I O

t tl 35.3X of RATED THERMAL POWER and core flow is less than or equal to 45X of rated core flow.

~C~O:

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.

SURVE L ANCE R UIR M NTS 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 APPLICABILITY.

WASHINGTON NUCLEAR UNIT 2 3/4 2-5 Amendment No. 94,137

C/l C)

Im m

80 60 cdKg y 40 0

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~2O (23.8F, 85.3P)

Region A (45F, 62.8P)

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0 40 70 O

CORE FLON (X RATED)

Operating Region Limits of Specification 3.2.6 Figure 3.2.6-1 p

3 4.2 POWER DISTRI T 0 LIMITS 3 4.2.7 STAB 0

TO ING TWO OOP OPERATION LIMITING CONDITION FO 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 regiori of APPLICABILITY.

A~PP I At I:

OPEONTIONAI OIINOITION I, Itt t I

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P operation and THERMAL POWER/core flow conditions which lay in Region C of Figure 3.2.7-1.

ACTION:

a.

With decay ratios of any two (2) neutron signals greater than or equal to 0.75 or with two (2) consecutive decay ratios on any single neutron signal greater than or equal to 0.75:

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 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 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 APPLICABILITYwithin one (1) hour.

SURVEILLANCE RE UIREM NTS 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 stabi1 ity moni toring system data aequi s ition 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 2-7 Amendment No. 94,"37

80 Region A 60 6'5 Kg y 40 0

CL aj (45F, 62.8P)

(45F, 50.8P)

(23.8F, 30.8P)

(28.8F, 35.8P) 0 CORE FLOW (X RATED)

Operating Region Limits of Specification 3.2.7 Figure 3.2.7-1 70

3 4.2 OW R DISTRIBUTION LIHITS 3 4.2.8 STABI IT HON TORING SINGLE LOOP OPERATION LIHITING 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.

N A~PPLICANI TP:

OPNPNTIONAL CIINOITION I, ~Itt I

I tt I

I I operation and THERHAL POWER/core flow conditions which lay in Region C of Figure 3.2.8-1.

])CT~0:

a.

With decay ratios of any two (2) neutron signals greater than or equal to 0.75 or with two (2) consecutive decay ratios on any single neutron signal greater than or equal to 0.75:

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 decreasing THERHAL POWER with control rod insertion.

Exit the region of APPLICABILITYwithin one (I) hour.

SURVEILLANCE RE UIREHENTS 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 (I) 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 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.

WASHINGTON NUCLEAR UNIT 2 3/4 2-9 Amendment No. 94, " 37

80 Region A (45F, 62.8P)

Region B (45F, 50.8P)

(28.8F, 30.8 (28.8F, 35.3 0

70 Core Flow (X RATED)

Operating Region Limits of Specification 3.2.8 Figure 3.2.8-1

TABLE 3. 3. 1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 6.

DELETED 7.

Primary Containment Pressure - High 8.

Scram Discharge Volume Water Level - High a.

Level Transmitter b.

Float Switch 9.

Turbine Throttle Valve - Closure APPLICABLE OPERATIONAL CONDITIONS

1) 2(f) 1, 2

5(h) 1 ) 2 5(h) 1(i)

MINIMUM OPERABLE CHANNELS PER TRIP SYSTEM a

2(g) 4(3)

ACT100 1

3 1

3 10.

Turbine Governor Valve Fast Closure, Valve Trip System Oil Pressure - Low 1(i) 2(3) 11.

Reactor Mode Switch Shutdown Position 12.

Manual Scram 1,

2 3, 4 5

1, 2

3, 4 5

1 7

3

\\

~

~TBLE..

(

ti d)

R ACTOR PROTECTION SYSTEM NSTRUMENTATION CTION ACTION 1 ACTION 2 ACTION 3 ACTION 4 ACTION 5 ACTION 6 ACTION 7 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Suspend all operations involving CORE ALTERATIONS* and insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

DELETED Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to a value such that THERMAL POWER is less than 30X of RATED THERMAL POWER, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Verify all insertable'ontrol rods to be inserted within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 8 ACTION 9 Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Suspend all operations involving CORE ALTERATIONS*, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Except movement of IRM, SRM or special movable detectors, or replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.

WASHINGTON NUCLEAR - UNIT 2 3/4 3-4 Amendment No. 137

AC OR PROTEC IO SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)

(b)

(c)

(d)

(e)

(g)

(h)

A channel may be placed in an inoperable status for up to six hours for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

The "shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn* and shutdown margin demonstrations are being performed per Specification 3. 10.3.

An APRM channel is inoperable if there are less than 2

LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

This function shall be automatically bypassed when the reactor mode switch is not in the Run position and reactor pressure

< 1060 psig.

This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10. 1.

This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

Also actuates the standby gas treatment system.

With any control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9. 10.2.

This function shall be automatically bypassed based on turbine first stage pressure when THERMAL POWER is less than 30X of RATEO THERMAL POWER.

Also actuates the EOC-RPT system.

Not required for control rods removed per Specification 3.9. 10. 1 or 3.9. 10.2.

WASHINGTON NUCLEAR UNIT 2 3/4 3-5 Amendment No. 99, 137

TABLE 3.3. 1"2 REACTOR PROTECTION SYSTEM RESPONSE TIMES FUNCTIONAL UNIT 1.

Intermediate Range Monitors:

a.

Neutron Flux - High b.

Inoperative 2.

Reactor Vessel Steam Dome Pressure - High Reactor Vessel Water Level - Low, Level 3 Main Steam Line Isolation Valve - Closure DELETED Primary Containment Pressure - High Scram Discharge Volume Mater Level - High a.

Level Transmitter b.

Float Switch 9.

Turbine Throttle Valve - Closure 10.

Turbine Governor Valve Fast Closure, Trip Oil Pressure - Low 11.

Reactor Mode Switch Shutdown Position 12.

Manual Scram 3.

4.

5.

6.

7.

8.

Average Power Range Monitor*:

a.

Neutron Flux - Upscale, Setdown b.

Flow Biased Simulated Thermal Power - Upscale c.

Fixed Neutron Flux - Upscale d.

Inoperative

RESPONSE

TIME Seconds N.A.

N.A.

6'**

< 0.09 H.A.

< 0.55

< 1.05

< 0.06 N.A.

N.A.

< 0.06

< 0.088 H.A.

N.A.

Neutron detectors are exempt from response time testing.

Response

time shall be measured from the detector output or from the input of the first electronic component in. the channel.

"Including simulated thermal power time constant.

Neasured from start of turbine control valve fast closure.

ACTION 20 ACTION 21 ACTION 22 ACTION 23 ACTION 24 ACTION 25 ACTION 26

,~TA I..

CC I

AT SO ATION ACTUATION I S

RUM NT TION A~CTI N

TATTN NT Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Be in at'east STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or close the affected system isolation valves within the next hour and declare the affected system inoperable or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ..

Lock close or close, as applicable, the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

JTTNT T IINN

Hay be bypassed with reactor steam pressure

~ 1060 psig and all turbine stop valves closed.

- When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

80uring CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

(b)

(c)

(d)

(e)

(f)

(g)

(h)

(a)

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

Also actuates the standby gas treatment system.

DELETED A channel is OPERABLE if 2 of 4 detectors in that channel are" OPERABLE.

Also actuates secondary containment ventilation isolation dampers per Table 3.6.5.2-1.

Closes only RWCU system outboard isolation valve RWCU-V-4.

Only valves RHR-V-123A and RHR-V-123B in Valve Group 5 are required for primary isolation.

Manual initiation isolates RCIC-V-8 only and only with a coincident reactor vessel level-low, level 3.

Not required for RHR-V-8 when control is transferred to the alternate remote shutdown panel during operational conditions 1,

2 8

3 and the isolation interlocks are bypassed.

When RHR-V-8 control is transferred to the remote shutdown panel under operational modes 1,

2, and 3 the associated key lock switch will be locked with the valve in the closed position.

Except RHR-V-8 can be returned to, and operated from, the control room, with the interlocks and automatic isolation capability reestablished in operational conditions 2 and 3

when reactor pressure is less than 135 psig.

WASHINGTON NUCLEAR UNIT 2 3/4 3-15 Amendment No. 58-,9M&8,137

R P

FUNC IO 1.

RIHARY CON TAB 3.3.

SOL ON CTUA ON NSTRU P

0 0

S POIN S

ALLOWABLE MILLIE a ~

b.

C.

d.

e.

Reactor Vessel Water Level 1)

Low, Level 3

2)

Low Low, Level 2

Drywell Pressure - High Hain Steam Line 1)

DELETED 2)

Pressure - Low 3)

Flow High Hain Steam Line Tunnel Temperature High Hain Steam Line Tunnel

Temperature - High Condenser Vacuum Low g.

Manual Initiation 2.

SECONOAR CO INN N ISOLA ION

~ 13.0 inches*

~ -50 inches*

g 1.68 psig h 831 psig g 115.6 psid 5 164'F S 80'F h 23 inches Hg absolute pressure N.A.

h 11.0 inches R -57 inches g 1.88 psig h 811 psig g 124.6 psid 5 170'F S 90'F

~ 24.5 inches Hg absolute pressure N.A.

a ~

b.

c ~

d.

Reactor Building Vent Exhaust Plenum Radiation High Drywell Pressure - High Reactor Vessel Water Level Low Low, Level 2

Manual Initiation

< 13.0 mR/h S 1.68 psig

~ -50 inches*

N.A.

5 16.0 mR/h g 1.88 psig

~ -57 inches N.A.

O

TABLE 4.3.2. 1-1 (Continued)

ISOLATION ACTUATION INSTRNENTATION SURVEILUWCE RE UIREMENTS CHANNEL TRIP FUNCTION CHECK 3.

REACTOR WATER CLEANUP SYSTEN ISOLATION a.

d Flow - High S

b.

Heat Exchanger Area Temperature - High c.

Heat Exchanger Area Ventilation 4 Teaperature-High N.A d.

Pump Area Temperature - High Pump Room A Pump Room B Pump Area Ventilation d Temp. - High Pump Room A Pump Room B f.

SLCS Initiation g.

Reactor Vessel Mater Level - Low Low, Level 2 h.

RMCU/RCIC Line Routing Area Temperature - High i

-RMCU Line Routing Area Temperature - High Manual Initiation N.A.

N.A N.A.

N.A.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION CHANNEL-FUNCTIONAL TEST SA SA SA SA SA R

1, 2, 3 I, 3, 3

- ~

10 20 3 10 20 3 R

R N.A.

10 20 3 2

3 1, 2, 3 1, 2, 3 1, 2, 3 OPERATIONAL CHANNEL CONDITIONS FOR MHICH CALIBRATION SURVEILLWCE RE VIREO a.

b.

Co N.A.

RCIC Steal Line Flow " High S

RCIC/RHR Stealw Line Flow - High S

RCIC Steam Supply Pressure-Low RCIC Turbine Exhaust Oiaphraya Pressure - High N.A.

RCIC Equipment Rooa Temperature - High RCIC Equipment Rooa d Teeperature - High N.A.

Q SA SA 10 20 1, 2, 3 1, 2, 3 1, 2, 3 1, 2, 3 20 3

~TAB 4.4..

fC ti 4)

SO T ON AC UA ION INSTRU ENT T 0 SUR IL NC R

U R

TRIP FUNC ON CHANNEL CIICEC CHANNEL FUNCTIONAL CHANNEL

~T 4~At INTia OPERATIONAL CONDITIONS FOR WHICH RV A

R 4.

R ACTOR CO SO NG SY TEN ISOLAT ON '(Continued) g.

RWCU/RCIC Steam Line Routing Area Temperature - High N.A.

h.

Drywell Pressure - High N.A.

i.

Hanual Initiation N.A.

5.

RHR S

ST N SHUT OWN COO G

E SOLA 0

SA R

R R

N.A.

1, 2, 3

a ~

b.

c ~

d.

e.

g.

N.A.

Reactor Vessel Water Level-Low, Level 3

S Reactor Vessel (RHR Cut-in Permissive)

Pressure - High N.A.

Equipment Area Temperature-High Equipment Area Ventilation 4 Temp.

High Shutdown Cooling Return Flow Rate High N.A.

RHR Heat Exchanger Area Temperature High Hanual Initiation SA SA SA R

R N.A.

1, 2, 3

TA E NOTAT ONS When reactor steam pressure

~ 1060 psig and/or any turbine stop valve is open.

When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

During CORE ALTERATION and operations with a potential for draining the reactor vessel.

TRIP UNC ION TABL 3.3 4.2-N -OF-C CLE RECI CULAT 0 U

P T S

ST STRU HINIHUH OPERABLE CHANNELS Plass I'

1.

Turbine Throttle Valve Closure 2.

Turbine Governor Valve - Fast Closure 2(b) 2(b)

A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance provided that the other trip system is OPERABLE.

This function shall be automatically bypassed when turbine first stage pressure is less than or equal to the pressure equivalent to THERNL POWER less than 30X of RATEO THERNL POWER.

TABLE 3.3.4.2"2 END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS TRIP FUNCTION 1.

Turbine Throttle Valve-Closure 2.

Turbine Governor Valve-Fast Closure TRIP SETPOINT

< 5X closed

> 1250 psig ALLIABLE VALUE

< 7X closed

> 1000 psig

TRIP FUNCTION l.

~ROO LKK ROIL OR a.

Upscale b.

Inoperative c.

Downscale CONTRO

~RRLR R.R.R-OD C

R A

ON S

PO S

TRIP S TPOI T g 0.58W + 48X N.A.

h 5X of RATED THERMAL POWER LO

~ 0.58W + 51X N.A.

~ 3X of RATED THERMAL POWER 2.

3.

4.

5.

6.

~PRM a.

Flow Biased Neutron Flux Upscale b.

Inoperative c.

Downscale d.

Neutron Flux - Upscale, Startup SOURCE NG 0

ORS a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale N

RMED G

ON ORS a.

Detector not full in b.

Upscal'e c.

Inoperative d.

Downscale SCRAM DISCHARG VO UM a.

Water Level-High b.

Scram Trip Bypass REACTOR COOLAN SYSTEM R

C RCULATION a.

Upscale b.

Inoperative c.

Comparator g 0.58W + 50X*

N.A.

h 5X of RATED THERMAL POWER g 12X of RATED THERMAL POWER N.A.

6 1x10 cps N.A.

~ 0.7 cps N.A.

g 108/125 divisions of full scale N.A.

h 5/125 divisions of full scale g 527 ft 3 in. elevation N.A.

FLO g 108/125 divisions of full scale N.A.

6 10X flow deviation g 0.58W + 53X*

N.A.

Z 3X of RATED THERMAL POWER g 14X of RATED THERMAL POWER.

N.A.

5 1.6 x 10 cps N.A.

~ 0.5 cps N.A.

6 110/125 divisions of full scale N.A.

h 3/125 divisions of full scale g 527 ft 5 in. elevation N.A.

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

g llX flow deviation

The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

TABLE 4.3.6-1 CONTROL ROD BLOCK IHSTRUHEHTATIOH SURVEILLANCE RE UIRENEHTS TRIP FUNCTION CHANNEL CHANNEL FUNCTIONAL CHECK TEST OPERATIONAL CHANNEL CONDITIONS FOR NICH ELLE)ILL)II))

SUWEIL~CE IEI)lit Elt I

I C

t'aitll EA 1.

2.

3.

4.

R00 BLOCK NNITOR a.

Upscale b.

Inoperative c.

Downscale APN a.

Flat Biased Neutron Flux Upscale b.

Inoperative c.

Downscale d.

Neutron Flux - upscale, 5tartup SOURCE RANGE NHITORS a.

Detector not full $n b.

Upscale c.

Inoperative d.

Downscale INTERHEDIATE RAHGE NNITORS H.A.

H.A.

N.A. ~

H.A.

N.A.

H.h.

H.A.

N.h.

5/U(b) ~ q 5/U(b)o q s/u(b), q 5/U(b), q H.h.

q 5] u(b), M(~)

5/u(b),

M S/u(b),

M 5/U(b), M N.A.

H.A.

5/U(b)(c), q(c) q 5/U(b)(c), q(c)

H.A.

5/U(b}(c), q(c) q 1A 1A 1A 1

1, 2, 5 12,5 2o 5 2,5 2,5 2,5 g.

o 5.

6.

a.

Detector not full in b.

Upscale c.

Inoperative d.

Downscale SCRAN DISCHARGE VOLNtE a.

Mater Level-High b.

Scram Trip Bypass REACTOR COOUNT SYSTN RECIRCULATION a.

Upscale b.

Inoperative c.

Coeparator H.A, H.A.

H.A.

H.h.

N.A.

H.h.

FLOM H.A.

H.A.

N.A.

s/u(b), M(~)

s/u(b),

M 5/u(b),

M 5/U(b), M s/u(b), q s/u(b), q s/u(b), q-N.A.

H.A.

R N.A.

H.A.

2,5 2,5 2,5 2,5 5%A 54 '

5

~

v t

~

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued) c.

Core flow is greater than or equal to 39'f rated core flow when core THERMAL POWER is greater than the limit specified in Figure 3.4.1.1-1.

4.4.1.1.2 With'one reactor coolant system recirculation loop not in operation, within no more than 15 minutes prior to either THERMAL POWER increase or recir-culation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 25K"*" of RATEO THERMAL POWER or the recirculation loop flow in the operating recirculation loop is

< 10K""* of rated loop flow:

a.

< 145'F between reactor vessel steam space coolant and bottom head Grain line coolant, b.

< 504F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure

vessel, and c.

< 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements of Specification 4.4.1.1.2b.

and c.

do not apply when the loop not in operation is isolated from the reactor pressure vessel.

4.4.1. 1.3 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:

a.

Verifying that the control valve fails "as is" on loss of hydraulic pressure (at the hydraulic control unit), and b.

Verifying that the average rate of control valve movement is:

1.

Less than or equal to illof stroke per second opening, and 2.

Less than or equal'o 11K of stroke per second closing.

"""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 4-3 Amendment No. 71

80 Region A 60 Region C (45F, 62.8P) 8 g~40 c8 20 (23.8F, 30.3P (23.8F, 35.3 Region 8 t 00% Rod Une 80% Rod Une (45F, 50.3P) 0 30 40 60 70 Core Flow (XRated)

OPERATING REGION LIHITS OF SPECIFICATION 3.4.1.1 FIGURE 3.4.1.1-1

R CTOR C

NT SYST M

3 4.4.

S F TY LI F VALVES IM T C NDI ON FOR OPERATION 3.4.2 a) The safety valve function of at least 12 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:*

2 safety/relief valves 9 1165 psig

%3X 4

safety/relief valves 9 1175 psig

%3X 4

safety/relief valves 9 1185 psig AX 4

safety/relief valves 9 1195 psig

%3X 4

safety/relief valves 9 1205 psig

%3X REPPICA TY:

OPERATIONAL CGNOITIONE I, d R, R

THEIIHAL POIIER I E

I than or equal to 25X of RATED THERMAL POWER.

b) The safety valve function of at least 4 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:*

2 safety/relief valves 9 1165 4

safety/relief valves 9 1175 4

safety/relief valves 9 1185 4

safety/relief valves 9 1195 4

safety/relief valves 9 1205 psig

%3X psig

%3X psig i3X psig i3X psig i3X CAB than 25X

~CION:

~TY:

OPENNTI NAL CIINOITTGNE I, I, d N.

A TEENAGE P IIER I of RATED THERMAL POWER.

a ~

b.

With the safety valve function of one or more of the above required safety/relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With one or more safety/relief valves stuck open, provided that suppression pool average water temperature is less than 90'F, close the stuck open safety/relief valve(s); if unable to close the open

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

WASHINGTON NUCLEAR UNIT 2 3/4 4-7 Amendment No. 88, 137

~

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY/RELIEF VALVES LIMITING CONDITiON FOR OPERATION ACTION: (Continued) valve(s) within 2 minutes or if suppression pool average water temperature is 110'F or greater, place the reactor mode switch in the Shutdown position.

SURVEILLANCE RE UIREMENTS MASHINGTON NUCLEAR UNIT 2 3/4 4-7a Amendment No.

~1 35

)

i r,"

TABLE 1.1.5-1 PRIMARY COOLNT SPECIFIC ACTIVITY SAMPLE ND NALYSIS PRONAN I

TYPE OF MEASUREMENT AND NALYSIS 1.

Gross Beta and Gaaaaa Activity Oeteraination 2.

Isotopic Analysis for OOSE EQUIVALENT l-131 Concentration 3.

Radiocheiaical for E Oeteraination h.

Isotopic Analysis for Iodine 5.

Isotopic Analysis of an Off-gas Sajaple Including guantitative Heasureaents for at least Xe-133'e-135 and Kr-Bs SAMPLE ND ANAl.YSIS FREIRIIENCY At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months At least once per 31 days At least once per 6 months~

a)

At least once per 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

~never the specific activity exceeds a liait, as required hy ACTION b..

b)

At least one seeple, bet~n 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following the change in THERMAL RNKR or off-gas level, as required by ACTION c.

At least once per 31 Ays OPERATIONAL CONDITIONS IN lSICH SAMPLE ND NALYSIS RE VIRED 1, 2, 3 lN,R,R,M 1,2 "Sample to be taken after a ainiae of 2 EFPO and 20 days of POMER OPERATION have elapsed since reactor was last subcritical for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months or longer.

NUntil the specific activity of the priaary coolant systea is restored to within its limits.

~

A

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I R ACTOR CO LAN SYS E

3

.4.6 PR SSUR T MPERATURE INITS R

CTOR CO A T S STE L HITING CONDITION FOR OPERATION 3.4.6. 1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4.6.1A, 3.4.6.1B, or 3.4.6.1.c*

(1) curve A or A'or hydrostatic or leak testing; (2) curve B or B'or heatup by non-nuclear

means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:

a.

A maximum heatup of 100 F in any 1-hour period, b.

A maximum cooldown of 100 F in any 1-hour period, c.

A maximum temperature change of less than or equal to 20 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.

The reactor vessel flange and head flange temperature greater than or equal to 80'F when reactor vessel head bolting studs are under tension.

~APLICAAILII: At 11 tt

]l~C~O:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UI EME TS 4.4.6. 1.1 During system heatup,

cooldown, and inservice leak and hydrostatic testing oper ations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figures 3.4.6.1A, 3.4.6.1.B, or 3.4.6.1.c curves A, A', B, B', or C, as applicable, at least once per 30 minutes.

Figure 3.4.6.1.c A'nd B'urves are effective for less than or equal to 8

EFPY of operation.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-18 Amendment No. 87-,HM,$ 37

RACO C

TS T

SURVE L ANC R

U R MENTS Continued 4.4.6.1.2 The reactor, coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1B curve C within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality and at least once per 30 minutes during system heatup.

4.4.6.1.3 The reactor vessel material surveillance specimens shall'e removed and examined, to determine changes in reactor pressure vessel material properties as required by 10 CFR Part 50, Appendix H in accordance with the NRC approved schedule.

The results of these examinations shall be used to update the curves of Figures 3.4.6.1A and Figure 3.4.6.1B.

4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80 F:

a.

In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:

1.

g 100'F, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.

g 90'F, at least once per 30 minutes.

b.

Mithin 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-19 Amendment No. %-,%7-,"37

WNP-2 PRESSlJRElTEMPERATURE UMITS EKC hN %OXENf0'KE'A'NP A

B Core beltliae limits.

Limits aft8r an asaaned 72 F core beltline temp shift from an ini)ial RTHDT of 28 F.

110 F 312 PSIG 140 F 312 PSIG Boltup limit 80'F 0

50 K

50 2X 250 3

350 4

450 5

FIGURE 3.4.6.1A HINIHUM REACTOR HETAL TEHPERATURE TEMPERATURE F

WASHINGTON NUCLEAR UNIT 2 3/4 4-20 Amendment No. 87.,137

Revis y

NRC letter dated:

Duly 14, 1994 WNP-2 PRESSURE/TEMPERATURE LIMITS FOR 8 EFPY TESTE'Nl IBtlC NKENBore beltlfne lpga)ts.

m

- S2 3u 6uu

~l 500 Lfalts after an assed 51.1 F core beltlkne temp shrift free an kn)Pal RTNDT of 28 F.

110 F

312 PSIG 140 F

312 PSIG Boltup l)mkt 80 F FIGURE 3.4.6.l.c MINIMUM REACTOR VESSEL METAL TEMPERATURE TEMPERATURE F WASHINGTON NUCLEAR UNIT 2 3/4 4-22 Amendment No. 488, 137

WNP-2 PRESSURE/TEMPERATURE LIMITS NUCLEAR HEATING CURVE

'C'ore beltline lieft.

Ueit after an assuee4 72"F core beltline teep shift froa an initial RT<>> of 28 F.

180 F 12 PSIG Boltup lisrlt 80 F I

I I

FIGURE 3.4.6.1B HINIHUH REACTOR METAL TEHPERATURE TEHPERATURE F

WASHINGTON NUCLEAR - UNIT 2 3/4 4-21 Amendment No. 42&,i37

REAC OR COOLANT SYSTEM REACTOR ST DON IHIT NG CON ION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1035 psig.

OPPIA I ':

OPENATIONAI IONOITIONN I 2 I hGBDk:

With the reactor steam dome pressure exceeding 1035 psig, reduce the pressure to less than 1035 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIRENENTS 4.4.6.2 The reactor steam dome pressure shall be verified to be less than 1035 psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Not applicable during anticipated transients.

WASHINGTON NUCLEAR UNIT 2 3/4 4-23 Amendment No. 137

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REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES h

LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall, be OPERABLE with closing times greater than or equal to 3 and less than or equal to 5 seconds.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a ~

With one or more MSIVs inoperable:

1.

Maintain at least one MSIV OPERABLE in each affected main steam line that is open and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , either:

a)

Restore the inoperable valve(s) to OPERABLE status, or b)

Isolate the affected main steam line by use of a deactivated MSIV in the closed position.

2.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The provisions of Specification

3. 0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4

~ 7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to Specification 4. 0. 5.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-24

' i

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EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION Continued ACTION:

(Continued) 2)

With the LPCS system inoperable and either LPCI subsystems "B"

or "C" inoperable, restore at least the inoperable LPCS system or the inoperable LPCI subsystem "B" or "C" to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

e.

3)

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and.in COLD SHUTDOWN within the following 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s*.

For ECCS divisions 1 and 2, provided that ECCS division 3 is OPERABLE and divisions 1 and 2 are otherwise OPERABLE:

1.

With two of the above required ADS valves inoperable, restore one inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to g 128 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2.

With three or more of the above required ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to ~ 128 psig within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In the event an ECCS system is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and submitted to the= Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the useage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.

WASHINGTON NUCLEAR UNIT 2 3/4 5-3 Amendment No. 137

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREHENTS 4.5.1 ECCS divisions 1, 2, and 3 shall be demonstrated OPERABLE by:

a.

At least once per 31 days for the

LPCS, LPCI, and HPCS systems:

l.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Verifing that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct" position.

b.

Verifing that, when tested pursuant to Specification 4.0.5, each:

1.

LPCS pump develops a flow of at least 6350 gpm at greater than or equal to 128 psid.¹ 2.

LPCI pump develops a flow of at least 7450 gpm at greater than or equal to 26 psid.¹ 3.

HPCS pump develops a flow of at least 6350 gpm at greater than or equal to 200 psid.¹¹ C.

For the

LPCS, LPCI, and HPCS systems, at least once per 18 months performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.

Actual injection of coolant into the reactor vessel may be excluded from this test.

d.

For the HPCS system, at least once per 18 months, verifying that the suction is automatically transfer red from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal.

"Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may.be in position for another mode of operation.

¹Pressure difference between the reactor and the suppression pool air volume.

¹¹Pressure differential above the suction source (Suppression pool or condensate storage tank).

WASHINGTON NUCLEAR - UNIT 2 3/4 5-4

w I

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3 4.6 CONTAINMENT SYSTEMS 3 4.6.

P IMARY CONTAI NT PRIMARY CONTAINME T INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

OOPPIA I ITT:

OPNNNTIONAI OONOITIONN I, I 2 N.

ACTION:

Without PRIMARY CONTAINHENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a ~

b.

c ~

d.

After each closing of each penetration subject to Type B testing, except the'primary containment air locks, if opened following Type A or 8 test, by leak rate testing the seals with gas at P,,

and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6. 1.2.d for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L,.

At least once per 31 days by verifying that all primary containment penetrations** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provided in Table 3.6.3-1 of Specification 3.6.3.

2 By verifying each primary containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

By verifying the suppression chamber is in compliance with the requirements of Specification 3.6.2.1.

See Special Test Exception 3. 10. 1.

- Except valves, blind flanges, and deactivated automatic valves which are within the primary containment, or other areas administratively controlled to prohibit access for reasons for personnel safety (i.e., radiation and temperature) and are locked,

sealed, or otherwise secured in'the closed position (l-l/2 inch and smaller valves connected to,vents, drains or test connections must be closed but need not be sealed).

Valves inside'ontainment shall be verified closed following primary containment de-inerting, but verification is not required more often than once per 92 days.

Valves in other administratively controlled areas shall be verified closed during each COLD SHUTDOWN, but verification is not required more often than once per 31 days.

WASHINGTON NUCLEAR UNIT 2 3/4 6-1 Amendment No. 88,137

CONTAIN NT SYSTEMS RIOR CONTAINHEN LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 a ~

b.

c ~

d.

C B

EZLOH:

With:

a ~

b.

c ~

d.

restore:

'a ~

Primary containment leakage rates shall be limited to:

An overall integrated leakage rate of less than or equal to L 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P,.

A combined leakage rate of less than or equal to 0.60 L, for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves* (and valves which are hydrostatically leak tested per Table 3.6.3-1), subject to Type B and C tests when pressurized to P,.

Less than or equal to 11.5 scf per hour for any one main steam line isolation valve when tested at P 25.0 psig.

A combined leakage rate of less than or equal to 1 gpm times the total number of ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment, when tested at 1. 10 P,.

ill YRIIIRRY TRNTRINIIENT INTEGRITYI, R,l, I R Specification 3.6.1.1.

The measured overall integrated primary containment leakage rate exceeding 0.75 L or The measured combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves* (and valves which are hydrostatically leak tested per Table 3.6.3-1),

subject to Type B and C tests exceeding 0.60 L or The measured leakage rate exceeding 11.5 scf per hour for any one main steam line isolation valve, or The measured combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment exceeding 1 gpm times the total number of such

valves, The overall integrated leakage rate(s) to less than or equal to 0.75 Land The combined leakage rate for all 'penetrations and all valves listed in Table 3.6.3-1, except for main steamline isolation valves* and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type B and C tests to less than or equal to 0.60 L and

Exemption to Appendix J of 10 CFR Part 50.

WASHINGTON NUCLEAR UNIT 2 3/4 6-2 Amendment No. i37

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CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION Continued

~ACTIO :

(Continued) c.

The leakage rate to less than or equal to 11.5 scf per hour for any one main steam line isolation valve, and d.

The combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the

'rimary containment to less than or equal to 1 gpm times the total number of such valves, prior to increasing reactor coolant system temperature above 200'F.

SURVEILLANCE RE UIREMENTS e

4.6. 1.2 The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:

a.

Three Type A Overall Integrated Containment Leakage Rate tests shall" be conducted at 40 i 10-month intervals during shutdown at P during each 10-year service period.

The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.

b.

If any periodic Type A test fails to meet 0.75 L, the test schedule for subsequent Type A tests shall be reviewed an) approved by the Commission.

If two consecutive Type A tests fail to meet 0.75 L a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L at which time the above test schedule may be resumed.

C ~

The accuracy of each Type A test shall be verified by a supplemental test which:

I.

Confirms the accuracy of the test by verifying that the supplemental test result, L minus the sum of the Type A and the superimposed leak, L are equal to or less than 0.25 L,.

2.

Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental, test.

3.

Requires the quantity of gas injected into the containment or bled from the containment during the supplemental test to be between 0.75 L, and 1.25 L,.

WASHINGTON NUCLEAR=- UNIT 2 3/4 6-3

'mendment No. 137

CONT INM SYST HS SURVEILLANCE RE UIREHENTS Continued d.

Type B and C tests shall be conducted with gas at P,*, at intervals no greater than 24 months except for tests involving:

1.

Air Locks 2.

Hain steam line isolation valves, 3.

Valves pressurized with fluid from a seal

system, 4.

ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment, and 5.

Purge supply and exhaust isolation valves'ith resilient seals.

e.

Air locks shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.3.

f.

Hain steam line isolation valves shall be leak tested at least once per 18 months.

g.

Leakage from isolation valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when determining the combined leakage rate provided the seal system and valves are pressurized to at least 1.10 P and the seal system capacity is adequate to maintain system pressure for at least 30 days.

h.

ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment shall be leak tested at least once-per 18 months.

i.

Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE per Surveillance Requirements 4.6.1.8. 1 and 4.6.1.8.2.

j.

The provisions of Specification 4.0.2 are not applicable to 24-month or 40 k 10-month surveillance intervals.

Unless a hydrosatic test is required per Table 3.6.3-1.

- - Forthose tests conducted during refueling outages, the 24-month interval may be exceeded by no more than 3 months.

WASHINGTON NUCLEAR UNIT 2 3/4 6-4 Amendment No. 44-,424,137

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CONTAINMENT SYSTEMS PRIMARY CON AINM NT IR LOCKS LIMITING CONDITION FOR OPERATION 3.6. 1.3 Each primary containment air lock shall be OPERABLE with:

a.

The interlock operable and engaged such that both doors cannot be opened simultaneously, and b.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, and c.

An overall air lock leakage rate of less than or equal to 0.05 L, at p,.

APPLICABILITY:

OPERATIONAL CONDITIONS I, 2" and 3.

~CTI ON:

a.

With the interlock mechanism inoperable:

Maintain at least one operable air lock door closed and either return the interlock to service within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock at least one operable air lock door closed.

2.

3.

Operation may then continue until the interlock is returned to service provided that one of the air lock doors is verified locked closed prior to each closing of the shield door and at least once per shift while the shield door is open.

h Personnel passage through the air lock is permitted provided an individual is dedicated to assure that one operable air lock door remains locked at all times so that both air lock doors cannot be opened simultaneously.

4.

The provisions of Specification 3.0.4 are not applicable.

b.

With one primary containment air lock door inoperable:

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2.

Operation may then continue until. performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked clos'ed immediately prior to each closing of the shield door and at least once per shift while the shield door is open.

See Special Test Exception 3. 10. 1.

WASHINGTON NUCLEAR UNIT 2 3/4 6-5 Amendment No. 9,137

CONT IN ENT SYSTEMS LIMITING CON ITION FOR OPERATION Continued KIL98:

(Continued) 3.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.

The provisions of Specification 3.0.4 are not applicable.

c ~

With the primary containment air lock inoperable, except as a result of an inoperable air lock door or an inoperable interlock mechanism, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.1.3 Each primary containment air lock shall be demonstrated OPERABLE:

a ~

b.

C ~

By verifying interlock operation (i.e., that only one door in each air lock can be opened at a time).

1.

Prior to using the air lock in Operating Conditions I, 2 and 3

but not required more than once per 6 months, 2.

Following maintenance that. could affect the interlock mechanism.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying seal leakage rate less than or equal to 0.025 L, when the gap between the door seals is pressurized to 10 psig.

By conducting an overall air lock leakage test at P,,

and by verifying that the overall air lock leakage rate is within its limit:

l.

At least once per 6 months¹¹¹,

and 2.

Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance had been performed on the air lock that could affect the air lock sealing capability*.

¹¹¹The provisions of Specification 4.0.3 are not applicable.

Exception to Appendix J of 10 CFR 50.

WASHINGTON NUCLEAR UNIT 2 3/4 6-6 Amendment No. 9,137

VA VE FUNCTIO N

TAB E 3.6.3-Contin ed PRIMARY CON NM ISO A ON ALVES ALVE GROU MAXIMUM ISOLATION TIHE c

d.

Other Containm t o

'on Valves (Continued)

Radiation Monitoring PI-V-X72f/1 PI-V-X73e/1 Tr ansversing Incore Probe System TIP-V-6 TIP-V-7,8,9,10,11(e)

TABLE NO ONS N.A.

N.A.

But greater than 3 seconds.

IProvisions of Technical Specification 3.0.4 are not applicable.

(a)

See Technical Specification 3.3.2 for the isolation signal(s) which operate each group.

(b)

Valve leakage not included in sum of Type B and C tests.

(c)

Hay be opened on an intermittent basis under administrative control.

(d)

Not closed by SLC actuation signal.

(e)

~ Not subject to Type C Leak Rate Test.

(f)

Hydraulic leak test at 1.10 P.

(g)

Not subject to Type C test.

fest per Technical Specification 4.4.3.2.2 (h)

Tested as part of Type A test.

(i)

Hay be tested as part of Type A test.

If so tested, Type C test results may be excluded from sum of other Type B and C tests.

(j)

Reflects closure times for containment isolation only.

(k)

During operational conditions 1,

2 E

3 the requirement for automatic isolation does not apply to RHR-V-8.

Except that RHR-V-8 may be opened in operational conditions 2

8 3 provided control is returned to the control room, with the interlocks reestablished, and reactor pressure is less than 135 psig.

(1)

The isolation logic associated with the reactor recirculation hydraulic control containment isolation valves need not meet single failure criteria for OPERABILITY for a period ending no later than Hay 15, 1995.

C'

CONTAINMENT SYSTEMS 3/4.6.4 VAcuuM RELIEF SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION.FOR OPERATION 3.6.4.1 Seven of the nine pairs of suppression chamber - drywell vacuum breakers shall be OP/RABLE and all nine pairs shall be closed.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

C.

With one or more vacuum breakers in up to two pairs of suppression chamber - drywell vacuum breakers inoperable for opening, verify both vacuum breakers of each pair to be closed within two (2) hours.

With one or more vacuum breakers in three or lore pairs of suppression chamber - drywell vacuum breakers inoperable for opening but known to be

closed, restore the inoperable pairs of vacuum breakers such that a

minimum of seven pairs are in an OPERABLE status ~ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLO SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With one suppression chamber " drywell vacuum breaker open, verify the t

other vacuum breaker in the pair to be closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; restore the open vacuum breaker to the closed position ~ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With one closed position indicator of any suppression chamber-drywe11 vacuum breaker inoperable:

1.

Verify the other vacuum breaker in the pair to be closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and at least once per 15 days thereafter, or 2.

Verify the vacuum breaker(s) with the inoperable position indicator to be closed by conducting a test which demonstrates that the hP is maintained at greater than or equal to 0.5 psi for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months without makeun within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 15 days thereafter.

3.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COED SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

WASHINGTON NUCLEAR - UNIT 2 3/4 6-34 Amendment No.

30

CONTAINMENT SYSTEMS 3 4.6.6 PR ARY CONTAIN T ATMOSPH E

CONTRO AN SUPP SS 0 C

B R HYDROG R COMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.6.1 Two independent drywell and suppression chamber hydrogen recombiner systems shall be OPERABLE.

~APP ICA ILITT:

PPENNTICINAL CIINPITIIINN I 2 2.

~AC ~0 With one drywell and suppression chamber hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.6. 1 Each drywell and suppression chamber hydrogen recombiner system shall be demonstrated OPERABLE:

a.

At least once per 6 months by verifying during a recombiner system warmup test that the minimum recombiner heater outlet temperature increases to greater than or equal to 500'F within 90 minutes.

b.

At least once per 18 months by:

1.

Performing a

CHANNEL CALIBRATION of all recombiner operating instrumentation and control circuits.

2.

Verifying the integrity of all heater electrical circuits by performing a resistance to ground test within 30 minutes following the above required functional test.

The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

3.

Verifying during a recombiner system functional test that, upon introduction of IX by volume hydrogen in a 140-180 scfm stream containing at least 1X by volume oxygen, that the catalyst bed temperature rises in excess of 120'F within 20 minutes.

4.

Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure; i.e.,

loose wiring or structural connections, deposits of foreign materials, etc.

c.

By measuring the system leakage rate:

l.

As a part of the overall integrated leakage rate test required by Specification 3.6. 1.2, or Amendment No. 26,137 3/4 6-44 2.

By measuring the leakage rate of the system outside of the containment isolation valves at P,,

on the schedule required by Specification 4.6. 1.2, and including the measured leakage as a

part of the leakage determined in accordance with Specification 4.6. 1.2.

WASHINGTON NUCLEAR UNIT 2

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PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 2.

Verifying that on each of the below pressurization mode actuation test signals, the train automatically switches to the pressurization mode of operation and the control room is gaintained at a positive pressure of 1/8 inch water 'gauge relative to the outside atmosphere during train operation at a flow rate less than or equal to 1000 cfm:

a)

Drywell pressure-high, b)

Reactor vessel water level-low, and c)

Reactor Building exhaust plenum-high radiation.

3.

Verifying that the heaters dissipate 5.0 + 0.5 kM when tested in accordance with ANSI N510-1980.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 while operating the train at a flow rate of 1000 cfm + lOX.

g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorber bank satisfies the inplace penetration and bypass leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-1980 for a halogenated, hydrocarbon refrigerant test gas while operating the train at a flow rate of 1000 cfm + 1(C.

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

36

J N

N

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N P

PAN S

S 3

.7.3 R

TO CO SO ATION COO ING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.**

OPIINTIONL ONOITIOOI I, I, d I Itt t

t d

pressure greater than 150 psig.

~AC ION:

With the RCIC system inoperable, operation may continue provided the HPCS system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2.

Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked,

sealed, or otherwise secured in position, is in its correct position.

3.

Verifying that the pump flow controller is in the correct position.

b.

When tested pursuant to Specification 4.0.5 by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1015 + 20, - 80 psig.*

The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

~*The ability of automatically taking RCIC suction from the suppression pool is not a requirement for RCIC OPERABILITY until May 17, 1993 or the beginning of the spring 1993 refueling outage when RCIC OPERABILITY is no longer required; whichever occurs first.

WASHINGTON NUCLEAR - UNIT 2 3/4 7-8 Amendment No. k%4,137

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POWER 0 STRIBUTION MITS BASES

v "k slightly increasing the time required for the normal scram to suppress the flux.

N T

G TIO RAT This specification assures that the Linear Heat'Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

At the high power/low flow corner of the operating

domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g.,

power shape, bundle power, and bundle flow).

In 1984, GE issued SIL 380 addressing boiling instability,and made several recommendations.

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

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

The SIL recommends that REGION A be bounded by the 100X rod line and REGION C be bounded by the BOX rod line.

The NRC Generic Letter 86-02 discussed both the GE and SIEMENS (then EXXON) stability methodology and stated that due to uncertainties, General Design Criteria 10 and 12 could not be met using available analytical procedures on a

BWR.

The letter discussed SIL 380 and stated that General Design Criteria 10 and 12 could be met by imposing SIL 380 recommendations in operating regions of potential instabilities.

The NRC concluded that regions of potential instability constituted decay ratios of 0.8 and greater by the GE methodology,.and 0.75 by the SIEMENS methodology which existed at that time.

SIEMENS Power Corporation has recently developed an improved stability computer code STAIF.

A topical report (EMF-CC-074P) which describes the STAIF stability code and provides benchmarking against reactor data was submitted to the NRC in 1993.

The NRC issued a

SER approving the STAIF stability code for establishing stability boundaries on April 14, 1994.

In the SER on STAIF the NRC stated the uncertainty in the STAIF code was 20X.

The STAIF stability code has been used to establish the stability region boundaries for WNP-2.

The lower boundary of REGION A was defined to assure it bounds a decay ratio of 0.9.

REGION C was conservatively defined to bound a

decay ratio of 0.75.

The stability REGIONS A and B are shown in Figure 3.2.6-1.

REGION A conforms to the recommendations of SIL 380 in that REGION A bounds a

calculated decay ratio of 0.9.

Operation in REGION A is proh'ibited.

REGION C

bounds a decay ratio of 0.75.

WASHINGTON NUCLEAR UNIT 2 B 3/4 2-5 Amendment No. 84, 137

POW D

R BU ION HITS BASES 3

G WO 00 OP RATIO At the high power/low flow corner of the operating

domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g.,

rod patterns, power shape).

To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRH and LPRH neutron flux signal decay ratios should be monitored while operating in this region (Region C).

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 45X of rated core flow and a thermal power at which the calculated decay ratio is less than 0.75.

Stability monitoring is performed utilizing the ANNA system.

The system shall be used to monitor APRH and LPRH signal decay ratio and peak-to-peak noise values when operating in the region of concern.

A minimum number of LPRH and APRH signals are required to be monitored in order to assure that both global (in-phase) and regional (out-of-phase) oscillations are detectable.

Decay ratios are calculated from 30 seconds worth of data at a

sample rate of 10 samples/second.

This sample interval results in some inaccuracy in the decay ratio calculation, but provides rapid update in decay ratio data.

A decay ratio of 0.,75 is selected as a decay ratio limit for operator response such that sufficient margin to an instability occurrence is maintained.

When operating in the i egion of applicability, decay ratio and peak-to-peak information shall be continuously calculated and displayed.

A surveillance requirement to continuously monitor decay ratio and peak-to-peak noise values ensures rapid response such that changes in core conditions do not result in approaching a point of instability.

3 4..8 TABI Y

0 0

G SING E

LOOP 0 TIO The basis for stability monitoring during single loop operation is consistent with that given above for two loop operation.

The smaller size of the region of allowable operation, Region C, is due to a limit on the allowed flow above the 80X rodline.

When operating above the 80X rodline in single loop operation, the core flow is required to be greater than 39X.

Continuous operation in Region B is not permitted.

Should Region B be entered the actions required by Technical Specification 3/4.4.1. 1 are to be complied with.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-6 Amendment No. 84,137

3 4.3 INSTRUMENTATION

'"""'ASES 3 4 3. 1 0

PROT C

ON SYSTEM INST UMENTATION The reactor protection system automatically initiates a reactor scram to:

a.

Preserve the integrity of the fuel cladding.

b.

Preserve the integrity of the reactor coolant system.

c.

Minimize the energy which must be adsorbed following a loss-of-coolant accident, and d.

Prevent inadvertent criticality.

This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance.

When necessary, one channel may. be made inoperable for brief intervals to conduct required surveillance.

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system.

The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.

The tripping of both trip systems will produce a reactor scram.

The system meets the intent of IEEE-279 for nuclear power plant protection systems.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC 30851 P, "Technical Specification Improvement Analyses for BWR Reactor Protection System,"

as approved by the NRC and documented in the SER (letter to T. A. Pickens from A. Thadani dated July 15, 1987).

The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2. 1.

The RPS instrumentation that provides

1) the Turbine Throttle Valve-Closure and 2) Turbine Governor Valve Fast Closure, Valve Trip System Oil Presure - Low trip signals measures first stage turbine pressure to initiate a trip signal.

The Load Rejection safety analysis (FSAR 15.2.2) bases initial conditions on rated power and specifies turbine bypass operability at greater than or equal to 30X of rated thermal power.

Because first stage pressure'can vary depending on operating conditions, the qualifying notes describing when the turbine bypass feature is to be disabled specify a turbine first stage pressure corresponding to less than 30X RTP (turbine first stage pressure is dependent on the operating parameters of the reactor, turbine, and condenser).

Therefore, because a value for turbine first stage pressure cannot be precisely fixed and because pressure measurement initiates the trip, the Technical Specification refers to a pressure associated with a specific Rated Thermal Power value rather than a value for pressure.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses.

No credit was taken for those channels with response times indicated as not applicable.

Response

time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.

WASHINGTON NUCLEAR UNIT 2 B 3/4 3-1 Amendment No. 90,137

INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems.

Shen neces-

sary, one channel may be inoperable for brief intervals to conduct required surveillance.

Some of the trip settings may have tolerances explicitly.stated where both the high and low values are critical and may have a substantial effect on safety.

The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.

Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected.

For D.C.-operated

valves, a 3-second delay is assumed before the valve starts to move.

For A.C.-operated valves, it is assumed that the A. C. power supply is lost and is restored by startup of the emergency diesel generators.

In this event, a time of 13 seconds is assumed before the valve starts to move.

In addition to the pipe break, the failure of the D.C.-operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 13-second diesel startup.

The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunc-tion with the 13-second delay. It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.

However, to enhance overall system reliability and to monitor instrument channel response time trends the isola-tion actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or

'ess than the drift allowance assumed for each trip in the safety analyses.

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.

This specification provides the OPERABILITY requirements, trip setpoints, and response times that will ensure effectiveness of the systems to provide the design protection.

Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 3-2

~

., INSTRUMENTATION BASES 3 4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.

The response of the plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349, dated March 1971, and NED0-24222, dated December 1979.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of the reactor protection system and is an essential safety supplement to the reactor trip.

The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle.

The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity.

Each EOC-RPT system trips both recircula-tion pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events.

The two events for which the EOC-RPT protective feature will function are closure of the turbine throttle valves and fast closure of the turbine governor valves.

A fast closure sensor from each of two turbine governor valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine governor valves provides input to the second EOC-RPT system.

Simil-

arly, a position switch for each of two turbine throttle valves provides input to one EOC-RPT system; a position switch from each of the other two throttle valves provides input to the other EOC-RPT system.

For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine governor valves and a 2-out-of-2 logic for the turbine throttle valves.

The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.

~

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled.

The manual bypasses and the automatic Operating Bypass at less than 30X of RATED THERMAL POWER are annunciated in the control room.

The EOC-RPT System instrumentation that provides a trip signal measures first stage turbine pressure to initiate a trip signal.

The safety analysis requiring an EOC-RPT bases initial conditions on rated power and specifies turbine bypass operability at greater than or equal to 30X of rated thermal power.

Because first stage pressure can vary depending on operating conditions, the qualifying notes describing when the turbine bypass feature is to disabled specify a turbine first stage pressure corresponding to less than 30X RTP (turbine first stage pressure is dependent on the operating parameters of the reactor, turbine, and condenser).

Therefore, because a

value for turbine first stage pressure cannot be precisely fixed and because pressure measurement initiates the trip the Technical Specification refers to a pressure associated with a specific Rated Thermal Power value rather than a

value for pressure.

The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of th'e electric arc, i.e.,

190ms, less the time allotted for sensor

response, i.e.,

10ms, and less the time allotted for breaker arc suppression determined by test, as correlated to manufacturer's test results, i.e.,

83ms, and plant preoperational test results.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 3-3 Amendment No. 137

INSTRUMENTATION BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.

Operation with a trip set less conservative than its Trip Setpoint but within'its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of Specifications 3/4.1.4, Control Rod Program Controls, 3/4.2, Power Distribution Limits and 3/4.3.1 Reactor Protection System Instrumentation.

The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

The test exception to the weekly Channel Functional Test of the SRM/IRM Detecto~ Not Full In instrumentation noted in Table 4.3.6-1, Control Rod Block Instrumentation Requirements, is intended to avoid cable damage and radiation exposure during operational condition 5 periods when outage work is being done in the under core region.

Upon completion of all the work in this area, when access for maintenance or construction efforts is no longer required, the test will be completed per the prescribed frequency within seven days.

3/4. 3. 7 MONITORING INSTRUMENTATION 3/4. 3. 7. 1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by'he individual channels; (2) the alarm is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63, and 64.

The criticality monitor alarm setpoints were calculated using the criteria from 10 CFR 70.24.a.l that requires detecting a dose rate of 20 Rads per minute of combined neutron and gamma radiation at 2 meters.

The alarm setpoint was determined by calculational methods using the gamma to gamma plus neutron ratios from ANSI/ANS 8.3-1979, Criticality Accident Alarm System, Appendix B and assuming a critical mass was formed from a seismic event, with a volume of 6' 6' 6't a distance of 27.7 feet from the two detectors.

The calculated dose rate using the methodology is 5.05 R/hr.

The allowable value for the alarm setpoint was, therefore, established at 5R/hr.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 3-4 Amendment No. 81

REACTOR COOLANT SYSTEM BASES 3/4.4. 5 SPECIFIC ACTIVITY The liaitations on the specific activity of the priaary coolant ensure that the 2-hour ttyroid and whole body doses resulting from a sain steam line failure outside the containment during steady. state operation vill not erceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on specific activity represent interim liaits based upon a parametric evaluation hy the NRC of typical site locations.

These values are conserva-tive in that specific site parameters, such as SITE SOUNDARY location and meteorological conditions,

>rare not considered in this evaluation.

The ACTION statement permitting POMER OPERATION to continue for liaited time periods ~5th the primary coolant's specific activity greater than 0.2 t.icrocurie per gram DOSE EQUIVALENT I-131, hut less than or equal to 4.0 @5cro-curies per gram OOSE EQUIVALENT J-131, acconeodates possible iodine spiking phenomenon which lay occur following changes in THERMAL POWER.

Closing the aain steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant vill he detected in sufficient tiae to take corrective action.

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

3g

REACTOR COOLANT SYSTE BASES P

U TS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 4.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady-state conditions, i.e.,

no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the

-controlling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup.and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The reactor vessel materials have been tested to determine their initial RT~,.

Reactor oper ation and resultant fast neutron irradiation, E greater than I HeV, will cause an increase in the RT~,.

Therefore, an adjusted reference temperature, based upon the fluence, nickel content, and copper content of the material in question, can be predicted using the fluence for 109.2X of original rated power and the recommendations of Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."

The pressure/temperature limit curves, Figures 3.4.6. 1A and 3.4.6. 1B include predicted adjustments for this shift in RT~, for the end of life fluence and are effective for 10 EFPY and 8 EFPY, respectively.

The actual shift in RT~, of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTH E185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figures 3.4.6.1A and 3.4.6. 1B shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.

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

, 488,137

5 ~

~

q I

e REACTOR COO NT S S

EH BASES SU P

TU (Continued)

The pressure-temperature limit lines shown in Figures 3.4.6.1A and 3.4.6.1B for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.

3 N S NE SO A ION Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break.

Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE.

The surveillance requirements are based on the operating history of this type valve.

The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks.

The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.

3 4.

S RUCTURA NT G

T The inspection programs for ASHE Code Class 1,

2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.

Access to permit inservice inspections of components of the reactor coolant system is in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code 1974 Edition and Addenda through Summer 1975.

The inservice inspection program for ASHE Code Class 1,

2 and 3

components will be performed in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code and applicable addenda.as required by 10 CFR 50.55a.

3 U

V A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.

WASHINGTON NUCLEAR UNIT 2 B 3/4 4-5 Amendment No. 87-,482-, H3, 137

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~

3 4.5 EM RGENCY CORE COOLING SYSTEM BASES 3 4.

nd 3

.5 CC - OP T NG nd S

UTDOWN ECCS division I consists of the low pressure core spray system and low pressure coolant injection subsystem "A" of the RHR system and the automatic depressurization system (ADS) as actuated by ADS trip system "A".

ECCS division 2 consists of low pressure coolant injection subsystems "B" and "C"

of the RHR system and the automatic depressurization system as actuated by ADS trip system "B".

The low pressure core spray (LPCS) system is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS.

The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that the LPCS

.system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident.

Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the. double-ended reactor recirculation line break, and for small breaks following depressurization by the ADS.

The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

ECCS division 3 consists of the high pressure core spray system.

The high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.

The HPCS system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.

The HPCS system operates over a range of 1210 psid, differential pressure between reactor vessel and HPCS suction source, to 0 psid.

WASHINGTON NUCLEAR UNIT 2 B 3/4 5-1 Amendment No.i37

~ ~

1 p

I

~

EHERGENCY CORE COO ING SYSTEH BASES E CS-TING d

SH TDOWN (Continued)

The capacity of the system is selected to provide the required core cooling.

The HPCS pump is designed to deliver greater than or equal to 516/1550/6350 gpm at differential pressures of 1160/1130/200 psig.

Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credi.t is taken in the safety analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition.

The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCS system to function properly after a small break loss-of-coolant

accident, the automatic depressurization system (ADS) automa-tically causes selected safety/relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200 F.

ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 psig.

This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.

ADS automatically controls seven selected safety/relief valves although the safety analysis only takes credit for five valves.

It is therefore appropriate to permit two valves to be out-of-service for up to 14 days without materially reducing system reliability.

3 4.5.3 SUPPR S

0 C A B The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the

HPCS, LPCS, and LPCI systems in the event of a LOCA.

This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core.

The OPERABILITY of the suppression chamber in OPERATIONAL CONDITION 1, 2, or 3 is required by Specification 3.6.2.1.

WASHINGTON NUCLEAR UNIT 2 B 3/4 5-2 Amendment No.137

3 4 6 ON NM NT SYST S

BASES 46 P

R 0

INE 6

I I

M INT GRIT PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted. to those leakage paths and associated leak rates assumed in the safety analyses.

This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

AR C

TAIN T L AKAG The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the calculated peak accident pressure.

As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50 with the exception of exemptions granted for main steam isolation valve leak testing and testing the airlocks after each opening.

3 4.6.

.3 P

IMA CONTA NM N AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the primary containment leakage rate given in Specifications 3.6. 1. 1 and 3.6.1.2.

The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during reactor operation.

Only one closed door in each air lock is required to maintain the integrity of the containment.

3 4.6

.4 V

AKAG CON ROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the main steamline isolation valves in the postulated LOCA situations would be a

small fraction of the 10 CFR Part 100 guidelines, provided the main steam line system from the isolation. valves up to and including the turbine condenser remains intact.

Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIVs such that the specified leakage requirements have not always been maintained continuously.

The requirement for the leakage control system will reduce the untreated leakage from,the MSIVs when isolation of the primary system and containment is required.

WASHINGTON NUCLEAR UNIT 2 B 3/4 6-1 Amendment No. 137

CONTAIN N

SYSTE S

BAS S

S L

G CONT 0 SYS E

(Continued)

Design specifications require the system to accommodate a leak rate of five times the Technical Specification leakage allowed for the MSIVs while maintaining a negative pressure downstream of the MSIVs.

The allowed leakage value per each valve is 11.5 scfm, or a total of 230 scfh.(3.8 scfm)."'hen corrected for worst case pressure, temperature and humidity expected to be seen during surveillance testing conditions, the flow would never exceed an indicated value (uncorrected reading from local flow instrumentation) of 5 cfm.

The 30 cfm acceptance criterion provides significant margin to tpis design basis requirement and provides a benchmark for evaluating long term blower performance.

The Technical Specification limit for pressure of -17" Hz0 W.C. was also established based on a benchmark of the installed system performance capability.

This -17" HzO W.C. provides assurance that the negative pressure criterion can be met.

6.

.5 C

NM S

UCTU G

This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the unit.

Structural integrity is required to ensure that the containment will withstand the maximum calculated pressure in the event of a LOCA.

A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

3 6.

DRYWE AND UPPR SSION CHAMB R TERNA PR SSUR The limitations on drywell and suppression chamber internal pressure ensure that the calculated containment peak pressure does not exceed the design pressure of 45 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 2 psid.

The limit of 1.75 psig for initial positive contain-ment pressure will limit the peak pressure to be less than the design pressure and is consistent with the safety analysis.

3

.6.

D W

G TE UR The limitation on drywell average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 340'F during LOCA conditions and is consistent with the safety analysis.

3 4.

YW SU P

SSION CHAM R

PURG SYST The 24-inch and 30-inch drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for inerting, de-inerting and pressure control.

Until all the drywell and suppression chamber valves have been qualified as capable of closing within the times assumed in the safety analysis, they shall not be open more than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months in any consecutive 365 days.

Valves not capable of closing from a full open position during a

LOCA or steam line break accident shall be blocked -so as not to open more than 70 (a)

Letter, G02-75-238, dated August 18,

1975, NO Strand (SS) to OD Parr (NRC),

"Response to Request for Information Main Steam Isolation Valve Leakage Control System" WASHINGTON NUCLEAR - UNIT 2 B 3/4 6-2 Amendment No. QS;-R4,i37

~

g I

CONTAIN NT SYST S

, <<"',$ j,;y BASES YW SUPP SS 0 C NBER PURG SYSTE (Continued)

'he time limit on use of the drywell and suppression chamber purge lines is not restricted when using the 2-inch purge supply and exhaust isolation valves since the 2-inch valves will close during a

LOCA or steam line break accident and therefore the SITE BOUNDARY dose guidelines of 10 CFR Part 100 would not be exceeded in the event of an accident during PURGING operati'ons.

The design of the 2-inch purge supply and exhaust isolation valves meets the requirements of Branch Technical Position CSB 6-4, "Containment Purging During Normal Plant Operations."

Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow the opportunity for repair before gross leakage failure develops.

Valves with metal to metal seals will be tested on a Type C schedule in accordance with Surveillance 4.6. 1.2.d to assure allowable leakage rates are not exceeded.

The 0.60 L

leakage limit shall not be exceeded when the leakage rates determined by tfie leakage integrity tests of those valves are added to the previously determined total

'or all valves and penetrations subject to Type B and C tests.

3 4.6 DEPRESSURI ATION SYST HS The specifications of this section ensure that the'primary containment pr'essure will not exceed the design pressure of 45 psig during primary system blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.

The suppression chamber'ater volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1040 psig.

Since all of the gases in the drywell are purged into the suppression chamber air space during a loss-of-coolant accident, the pressure of the liquid must not exceed 45 psig, the suppression chamber maximum pressure.

The design volume of the suppression

chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is below the design pressure of 45 psig.

Haximum water volume of 128,827 ft results i~ a downcomer submergence of 12 ft and the minimum volume of 127, 197 ft results in a submergence approximately 4 inches less.

The majority of the Bodega tests were run with a submerged length of 4 feet and with complete

'ondensation.

Thus, with respect to the downcomer submergence, this specification is adequate.

The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F and this is conservatively taken to be the limit for complete condensation of the reactor

coolant, although condensation would occur for temperatures above 170'F.

WASHINGTON NUCLEAR UNIT 2 B 3/4 6-3 Amendment No. 490-424, 137

CONTAINM NT SYSTEMS BASES D

SSU IO T

S (Continued)

Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.

Under full power operating conditions, blowdown from an initial suppression chamber water temperature of 90'F results in a water temperature of approximately 145'F immediately following blowdown which is below the 200'F used for complete condensation via quencher devices.

At this temperature and atmospheric

pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus, there is no dependency on containment overpressure during the accident injection phase.

If both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.

Experimental data indicate that excessive steam condensing loads can be avoided if the peak bulk temperature of the suppression pool is maintained below 200'F during any period of relief valve operation with sonic conditions at the discharge exit for quencher devices.

Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Because of the large volume and thermal capacity of the suppression

pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.

The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a

safety/relief valve inadvertently opens or sticks open.

As a minimum this action shall include:

(I) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety/relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety/relief valve to assure mixing and uniformity of energy insertion to the pool.

3

.6.3 P

IMARY CONTA NM SOLATION V VES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

Containment isolation within the time limits specified ensures for those isolation valves designed to close auto-matically that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 6-4 Amendment No. 80-,400,i 37

~

ADMINISTRATIVE CONTROLS t'

CORE OP RATING IM T R

PO (Continued) 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 report previously reviewed and approved by the NRC, which describe the methodology applicable to the current cycle.

For WNP-2 the topical reports are:

1.

ANF-1125(P)(A),

and Supplements 1 and 2, "ANFB Critical Power Correlation," April 1990 2.

Letter, R.

C. Jones (NRC) to R. A. Copeland (ANF),

"NRC Approval of ANFB Additive Constants for ANF 9x9-9X BWR Fuel," dated November 14, 1990 3.

ANF-NF-524(P)(A), Revision 2 and Supplements 1 and 2, "Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors,"

November 1990 4.

ANF-913(P)(A), Volume 1, Revision 1 and Volume 1, Supplements 2,

3 and 4, "COTRANSA':

A Computer Program for Boiling Water Reactor Transient Analysis," August 1990 5.

ANF-CC-33(P)(A), Supplement 2,

"HUXY:

A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option," January 1991 6.

XN-NF-80-19(P)(A), Volume 1, Supplements 3 and 4, "Advanced Nuclear Fuel Methodology for Boiling Water Reactors,"

November 1990 7.

XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,"

June 1986 8.

XN-NF-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 BWR Reload Fuel," September 1986 10.

ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, "Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9x9-IX and 9x9-9X BWR Reload Fuel," October 1991 11.

XN-NF-81-22(P)(A), "Generic Statistical Uncertainty Analysis Methodology," November 1983 12.

NEDE-24011-P-A-10-US, "General Electric Standard Application for Reactor Fuel," U.S. Supplement, March 1991 13.

NEDE-23785-1-PA, Revision 1,

"The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume III, SAFER/GESTR Application Methodology," October 1984 14.

NED0-20566A, "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10'CFR 50 Appendix K," September 1986 15.

EHF-CC-074(P)(A),

"Volume 1 STAIF A Computer Program for BWR Stability in the Frequency

Domain, Volume 2 STAIF A Computer Program for BWR Stability in the Frequency

Domain, Code gualification Report,"

July 1994.

WASHINGTON NUCLEAR UNIT 2 6-21 Amendment No. 94-,499,137

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) 6.9.3.3 The core operating limits shall be determined such that all appli-cable limits (c.g., fuel thermal-mechanical limits, core thermal-.

hydraulic lfmfts, ECCS lfafts, nuclear limits such as shutdown margin, transient analysis lfafts and accident analysis lfafts) of the safety analysis arc act.

6.9.3.4 The CORE OPERATING LIMITS REPORT, fncludfng any afd-cycle ravfsfons or supplements, shall be provided upon issuance for each reload cycle, to the NRC Document Control Oesk with copies to thc Regional Administrator and Resident Inspector.

6. 10 RECORD RETENTION 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 C.

d.

The following records shall be retained for at least 5 years:

Records and logs of unit operation covering time fnterval at each po~er level.

Records and logs of principal maintenance activities, inspections,

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

All REPORTABLE OCCURRENCES submitted to the Cownfssfon.

Records of surveillance activities, inspections, and calibrations required by these Tcchnical Specifications.

WASHINGTON NUCLEAR - UNIT 2 6-22 Amendment No.

94

ML17291A776

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