ML20082D944

Proposed Tech Specs Re Incorporation in Scope & Content Endorsed by NRC for License NPF-42

ML20082D944
Person / Time
Site:	Wolf Creek
Issue date:	04/06/1995
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20082D938	List: ML20082D937 ML20082D944
References
NUDOCS 9504110043
Download: ML20082D944 (45)

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TAttachment'II to CO 95-0032.

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REVISED TECHNICAL SPECIFICATION PAGES

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9504110043 950406 PDR ADOCK 05000482

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i DEFINITIONS g

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.-

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

i 1)

. Capable-of being closed by an OPERABLE containment automatic isolation valve system, or-2)

Closed by manual valves, blind. flanges, or'deactivat'ed automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment' hatches are closed and. sealed, c.

Each-air lock is in compliance with the requirements of Specification 3.6.1.3, k w n.n.a by seeiS d e d A l.t.4 E ;s+,A a+he Bases i

U e f.

The containment leakage rates are within the limits of Specification 3.C.I.2, and.

13/4. 6.t.t d p.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABLE,oM CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant -

pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION'shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT e

1.10 The CORE OPERATING LIMITS REPORT (COLR) _is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9.

Plant operation within these operating limits is addressed in individual Specifications.

WOLF CREEK - UNIT 1 1-2 Amendment No. 61 5

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t REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued)

-is yahr A or epcd b 1.3 % oK/g 3.

The rod is declaredninoperable and the SHUTDOWN MARGIN Me! : ::t d S pciceti:n 3.1.1.1 1: ::ti:ffed.

POWER OPERATION may then continue provided that:

a)

A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions;

-b rewir m at af ;,ecificatien 0.1.1.1

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4: h t: min:d :t 1:::t :::: per 12 h: r:;-

b jd A power distribution map is obtained from the movable inceredetectorsandF(Z)andFhareverifiedtobe 9

within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and c/

The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.

ACTION 4 - Restore the inoperable rods to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at

. least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

4.1.3.l.3 rh15EKT l - I ci WOLF CREEK - UNIT 1 3/4 1-15 Amendment No. 27 J

.i INSERT 1-15 4.1'3.1.3

. Prior to reactor criticality, the rod drop time of the individual full-length shutdown and control rods from the fully withdrawn position shall be ' demonstrated to be less than-or equal to 2.7 seconds from beginning 'of-

. decay of stationary gripper coil voltage to dashpot entry, with T.,, 2 551*F, i

and all. reactor coolant pumps operating:

a.

For all rods. following each removal of the reactor. vessel head, and l

b.

For specifically affected ' individual rods following any maintenance on or modification to the control rod' drive' system which could affect the drop time of those specific rods.

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l rr REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4. Both power-operated relief valves -(PORVs) and their associated block valves shall be OPERA 8LE.

j APPLICABILITY: MODES 1, 2, and 3.*

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a.

With one or both PORVs inoperable because of excessive seat leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

l b.

With one PORV inoperable due to causes other than excessive seat leakage, within I hou'r either restore the PORV to OPERABLE status or close its associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, j

i c.

With both PORVs inoperable due to caussi other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d.

With one or both block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the

. block valve (s) to OPERABLE status or place its associated PORV(s) in manual control. Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

e.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.4.2 Each block. valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless

)

the block valve is closed in order to meet the requirements of ACTION b, or c.

in Specification 3.4.4.

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• With all RCS cold leg temperatures cbove 368'F.

WOLF CREEK UNIT 1 3/4 4-10 Amendment No. 63

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IWSERT 4-10 7

I i-4.4.4.3 Both PORV pesition indica shall be demonstrated OPERABLE at least i

once per 31 days by performan f a CHANNEL CHECK unless the associated block

)

valve is in the closed ion.

4,4.4,4 Bot block valve position indicators ',- shall be demonstrated OPERABLE east once per 31 days by performance of a CHANNEL CHECK unless j

the ek valve is verified in the closed position and power is removed.

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3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within I hour or be in at least HOT STANDEY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS

4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least once per 31 days by verifying that all 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 their positions, except as provided in Table 3.6-1 of Specification 3.6.3; b.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; awus c.

After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure not less than P,, 48 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.5.1.2d. for all other Type B and C penetrations, the combined leakagedrate is less than 0.60 L,;

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• Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed, or otherwise secured I

in the closed position.

These penetrations shall be verified closed during each COLD SHUT 00WN except that such verification need not be performed more often than once per 92 days.

WOLF CREEK - UNIT 1 3/4 6-1 I

l INSERT 6-1 1

d.

By performing containment leakage rate testing, except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions; and e.

By verifying containment structural integrity in accordance with the Containment Tendon Surveillance Program of Specification 6.8.5.c.

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CONTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS 4.6.1.7.1 Each 36-inch containment shutdown purge supply and exhaust isolation valve (s)* shall be verified blank flar,ged and closed at least once per 31 days.

4.6.1.7.2 Each 36-inch containment shutdown purge supply and exhaust isolation valve and its associated blank flange shall be leak tested at least once per 24 months and following each reinstallation of the blank flange when pressurized to P, 48 psig, and verifying that when the measured leakage rate for these a

valves and flanges, includirg staa leakage, is added to the leakage rates deter-mined pursuant to Specification A 6.1.2d., for all other Type B and C penetra-tions, the combined leakage rete is less than 0.60 L,.

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-4.6.l.Id.

4,6.1.7.3 The cumulative time that all 18-inch containment mini purge supply and/or exhaust isolation valves have been open during a calendar year shall te determined at least once per 7 days.

4.6.1.7.4 At least once per 3 months each 18-inch containment mini-purge supply and exhaust isolation valve with resilient material seals shall be i

demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.05 L, when pressurized to P,.

"Except valves and flanges which are located inside containment.

These valves shall be verified to be closed with their biank flanges installed prior to entry into MODE 4 following each COLD SHUTDOWN.

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'a0LF CREE ( - UNIT 1 3/4 6-12

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ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 8)

Limitations on the annual and quarterly air doses resulting i

from noble gases released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, i

9)

Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 10)

Limitations on the annual dose or dose commitment to any MEMBER 0F THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

f.

Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) repre-sentative measurements of radioactivity in the highest potential exposurr, pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

e The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM.

2)

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and the modifications to the monitoring program are made if required by the results of this census, and 3)

Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sampli matrices are performed as part of the quality assurance program for environmental monitoring.

Tad 56RT 68. 5 -**

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.

WOLF CREEK UNIT 1 6-18 Amendment No. 42

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INSERT 6.8.5

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a.

Explosive Gas and Storage Tank Radioactivity Monitoring Program-This program provides controls for potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity r

contained in unprotected outdoor liquid storage tanks.

The program shall include:

1 1.

The limits for concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM and a surveillance program to ensure the limits are n.aintained.

2.

A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 20.5 rem to any individual i

in an UNRESTRICTED AREA in the event of an uncontrolled release of the tanks' contents, consistent with Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases due to Waste Gas System Leak or Failure."

3.

A surveillance program to ensure that the quantity of radioactivity contained in following outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area-drains connected to the liquid radwaste system, is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA, in the event of an uncontrolled release of the t

tanks' contents, a.

Reactor Makeup Water Storage Tank, b.

Refueling Water Storage Tank, c.

Condensate Storage Tank, and d.

Outside Temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste.

1 The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

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Reactor Coolant Pump Flywheel Inspection Program

-Each reactor coolant. pump' flywheel.shall-be ' inspected per.the recommendations of '. Regulatory Position' C.4.b of ' Regulatory Guide 1.14, ll

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Revision 1, dated August 1975.

c.

Containment Tendon Surveillance Program This program-provides controls for. monitoring tendon performance /.

including the offactiveness of the ' tendon corrosion ' protection medium, to ensure containment' structural integrity.

The program shall include -

baseline measurements prior to initial plant operation as well' as i

periodic testing thereaf.ter.

The containment. Tendon -Surveillance j

Program, and its inspection frequencies and acceptance criteria,.shall' j

be in accordance with Wolf Creek Generating Station position on draf t.

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Revision 3 of Regulatory Guide 1.35 dated April 1989.

I The provisions of Specifications 4.0.2 and 4.0.3 are applicable to'the containment Tendon Surveillance Program inspection frequencies.

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Attachment III to CO 95-0032

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t ATTACHMENT III j

REVISED PAGES FROM THE APPLICATION OF TME NRC FINAL POLICY STATEMENT l

ON TECHNICAL SPECIFICATION IMPROVEMENTS I

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TABLE I (Cont.)

Summary of Criteria Application Results Power Distribution Limits Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.2.1 3.2.1 Axial Flux Differ.

Retain Retain 3.2.2 3.2.2 Heat Flux Hot Channel Retain Retain Factor 3.2.3 3.2.3 Nuclear EnthalpyRise Retain Retain Hot Channel Factor 3.2.4 3.2.4 Quadrant Power Tilt Retain Retain Ratio 3.2.5 3.2.5 DNB Parameters Retain Retain TABLE I (Cont.)

Summary of Criteria Application Results Instrumentation Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.3.1 3.3.1 Reactor Trip System Retain Retain Instrumentation 3.3.2 3.3.2 Eng. Safety Feature Retain Retain Actuation System Instrumentation 3.3.3.1 3.3.3.1 Radiation Monitoring Retain Retain Instrumentation 3.3.3.2 3.3.3.2 Movable Incore Detectors Relocate Relocate 3.3.3.3 3.3.3.3 Seismic Instrumentation Relocate Relocate 3.3.3.4 3.3.3.4 Meteorological Relocate Relocate Instrumentation 3.3.3.5 3.3.3.5 Remote Shutdown Retain Retain Instrumentation 3.3.3.6 3.3.3.6 Accident Monitoring Retain Retain 5

Instrumentation 3.3.3.9 3.3.3.9 Loose Parts Detection Relocate Relocate System 3.3.3.11 Explosive Gas Monitoring Not Relocate 6

Instrumentation Reviewed 3.3.4 3.3.4 Turbine Overspeed Relocate Relocate

-Y-Protection 1

TABLE I (Cont.)

Summary of Criteria Application Results Reactor Coolant System Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.4.1.1 3.4.1.1 Reactor Coolant Loops Retain Retain and Coolant Circulation 3.4.1.2 3.4.1.2 RCS Hot Standby Retain Retain 3.4.1.3 3.4.1.3 RCS Hot Shutdown Retain Retain 3.4.1.4.1 3.4.1.4.1 Cold Shutdown Loops Retain Retain Filled 3.4.1.4.2 3.4.1.4.2 Cold Shutdown Loops Retain Retain j

Not Filled 3.4.2.1 3.4.2.1 Safety Valves -Shutdown Relocate Relocate 3.4.2.2 3.4.2.2 Safety Valves -Operatine Retain Retain 3.4.3 3.4.3 Pressurizer Retain Retain 3.4.4 3.4.4 ReliefValves Retain Retain 3.4.5 3.4.5 Steam Generators

!!doce^c "dca::,

3.4.6.1 3.4.6.1 Leakage Detection

" Retain

Retain Systems 4%

Re h 3.4.6.2 3.4.6.2 Operational Leakage Retain Retain 3.4.7 3.4.7 Chemistry Relocate Relocate 3.4.8 3.4.8 Specific Activity Retain Retain i

3.4.9.1 3.4.9.1 Pressure /femperature Retain Retain Limits 3.4.9.2 3.4.9.2 Pressurizer Relocate Relocate Pressure / Temperature 3.4.9.3 3.4.9.3 Overpressure Protection Retain Retain System 3.4.10 3.4.10 Structural Integrity Relocate Relocate 10 3.4. I 1 3.4.11 RCS Vents Relocate Relocate

TABLE 1 (Cont.)

Summary of Criteria Application Results Emergency Core Cooling Systems Tech STS Rev. 5 Technical Specification NRC WCGS Note i

Spec Number Title Results Results Number j

3.5.1 3.5.1 Accumulators Retain Retain 3.5.2 3.5.2 ECCS Subsystems Tavg Retain Retain 2350 F 3.5.3 3.5.3 ECCS Subsystems Tavg Retain Retain

< 350'F 3.5.4 ECCS Subsystems Tavg Not Retain 2,1 I s 200"F Reviewed 3.5.5 3.5.5 RWST Retain Retain 4

TABLE 1 (Cont.)

Summary of Criteria Application Results Containment Systems Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.6.1.1 3.6.1.I Containment Integrity Retain Retain 12, G 3.6.1.2 3.6.1.2 Containment Leakage See Note 12 See Note 12 12' 3.6.1.3 3.6.1.3 Containment Airlocks Retain Retain 3.6.1.4 3.6.1.5 Internal Pressure Retain Retain 3.6.1.5 3.6.1.6 Air Temperature Retain Retain 3.6.1.6 3.6.1.7 Contain. Vessel Structural Relocate Relocate 13 Integrity 3.6.1.7 3.6.1.8 Containment Ventilation Retain Retain 14 System 3.6.2.1 3.6.2.1 Containment Spray Retain Retain System 3.6.2.2 3.6.2.2 Spray Additive System Retain Retain 3.6.2.3 Containment Cooling Retain Retain System 3.6.3 3.6.3 Containment Isolation Retain Retain Valves 3.6.4.1 3.6.4.1 Hydrogen Analyzers Retain Delete 15 3.6.4.2 3.6.4.2 IIydrogen Control System Retain Retain

TABLE 1 (Cont.)

Summary of Criteria Application Results Plant Systems Tech STS Rev. 5 Technical Specification Title NRC WCGS Note Spec Number Results Results Number 3.7.1.1

.3.7.1.1 Safety Valves Retain Retain 3.7.1.2 3.7.1.2 Auxiliary Feedwater System Retain Retain 3.7.1.3 3.7.1.3 Condensate Storage Tank Retain Retain 3.7.1.4 3.7.1.4 Specific Activity Retain Retain 3.7.1.5 3.7.1.5 Main Steam Isolation Valves Retain Retain 3.7.1.6 Steam Generator Not Retain Atmospheric Relief Valves Reviewed 3.7.1.7 Main FeedwaterIsolation Not Add 16 Valves Reviewed 3.7.2 3.7.2 Steam Generator Relocate Relocate Pressurefremperature Limits i

3.7.3 3.7.3 Component Cooling Water Retain Retain 3.7.4 3.7.4 Essential Service Water Retain Retain System 3.7.5 3.7.5 Ultimate Heat Sink Retain Retain 3.7.6 Control Room Emerg.

Retain Retain Ventilation System 3.7.7 3.7.8 Emerg. Exhaust System -

Retain Retain Auxiliary Building 3.7.8 3.7.9 Snubbers Relocate Relocate

+7-3.7.9 3.7.10 Scaled Source Contamination Relocate Relocate

'.7.12 3.7.13 Area Temperature Relocate Relocate it a

Monitoring

TABLE I (Cont)

Summary of Criteria Application Results Special Test Exceptions Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results Number 3.10.1 3.10.1 Shutdown Margin Relocate Delete 20 3.10.2 3.10.2 Group Height, Insenion, Retain Retain and Power Distribution Limits 3.10.3 3.10.3 Physics Tests Retain Retain 3.10.4 3.10.4 Reactor Coolant Loops Retain Retain 3.10.5 3.10.5 Position Indication Relocate Relocate 20 System Shutdown TABLE I (Cont.)

Summary of Criteria Application Results Radioactive Effluents Tech STS Rev. 5 Technical Specification NRC WCGS Note Spec Number Title Results Results i

Number 3.11.1.4 3.11.1.4 Liquid Holdup Tanks Relocate Relocate

-M 6 3.I1.2.5 3.I1.2.5 Explosive Gas Mixture Relocate Relocate 6

3.11.2.6 3.11.2.6 Gas Storage Tanks Relocate Relocate 41-6 1

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i will' be' retained cas is.

- The surveillance associated. with LCO

i initial 1 condition and supports LCO 3.1;3.1..As'such,-LCO 3.1'.3.2 j

3.1.3.3 ~ is. not' required' for-any. retained LCO and,.therefore,.SR l

74'.1.3.3 will be. relocated.

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- The. NRC review: ofL this, LCO. concluded-that itJcould be?

relocated.

However, if an associated'SR?is'necessary to meet the

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operability requirements for a. retained LCO, the SR should be.

l relocated.to the retained:LCO.- SR 4.1.3.4. is required to ensure the L operabilityn of control rods 'under LCO '3;1.3.1f and willD.bei retained under:that LCO.;ith th n y ti ; lixit gi.x in'22:

""?? ":: f r 15.1_? 2. This is consistent with STS.

1 5.

The Regulatory Guide 1.97, Rev. 2, Type A variables' identified (

in USAR Appendix 7A c are retained.

' The - neutron flux - (Gamma-Metrics). and RVLIS instrumentation will be added. 'The non-Type A--

variables are identified and evaluated on the screening form. TheJ l

relocated instruraents ares

'i Containment Pressure'- Extended Range-PotW Mt'm rnaca+or PZR Safety Valve Position Indication.

POIN bc.k %)ve. bd h LMc. dot-

. Unit Vent-High Range Noble Gas Monitor-l i

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6.

This specification will be relocated fand an' Explosive Gas Monitoring Program statement will be~ incorporated,into new'Section 6.8.5.

end 5 h yr % k Ra&ou.Hd.+y..

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l i

l 10.

The LCO will be relocated and the associated SR regarding RCP

' j flywheel integrity will be retained in new -Section - 6.8.5 as a programmatic requirement.

i

11. This LCO is intended to prevent a loss of the decay. heat'-

removal function in Mode 5 and Mode 6 with the reactor vessel head

{

installed by allowing the safety injection pumps to be operable

)

when the water level is below the vessel flange.

The LCO will be

.c retained.

Consideration was given to incorporating the

N f

~

l, restrictions on pump operation,into LCO - 3. 4. 9. 3, Overpressure Protection, which would _ have been.in conformance with the - STS

' approach. :However, the Modes and RCS temperatures for which these specifications apply prevented-combining them into one specification.

3.6. l. 2 -

12. Containment testing is a requirement imposed by Appendix J of 10 CFR 50.

-h6e-LCO will' be relocated; however, the values of parameters defining leakage limits from 3.6.1.2 will be retained-under the Containment Integrity Bases.

SR 4.6.1.lc will be modified to/._ ell-i--te ref:rener t: : :p::ifi:: tion th:t a-

...a nn.,,

n.s-

$h$h 2 $ k En $7v$eN e do$e d *C'*ECA No d.Y.\\.\\.Y b bin E 55 locFR$o, Appen&NT.

13. This specification will be relocated and a Containment Tendon Surveillance Program -statemese will be incorporated into new.

Section 6.8.5. New SR 4.b. I.I.e. J D* cme 6 ** I'*P e** N S$* *

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14. SR 4.6.1.7.2 will be modified to eliminate reference to a specification that was relocated, and instead reference correrpending "E?_". Eertien 15.5.1.1. rua SR 4.fo.\\.\\.h.

15. LCO 3.6.4.1 is deleted since it is redundant. to LCO 3.3.3.6 and is obsolete per the STS.

16. A new Technical Specification. for operability of the Main Feedwater Isolation Valves (MFIVs) will be added to the Technical Specifications for Wolf Creek. The requirements will be identical to those in the Callaway Technical Specifications.

Inclusion of a specification for the MFIVs is consistent with NRC Policy Statement Criterion 3 regarding accident mitigating components.

1? Sie pecification rill be releerted :nd : Enubb:r Insp::ti:n Progr:r :::t:::nt will be included in !!:u E tien S.e.5.

IS.

Sir pecification Till be r:100 ted end en ?rt: Terperatur:

Menitoring Pr gree -tster:nt rill be included in nee Ee tien 5.0.5.

19. This specification places a lower limit on the amount of water above the top of the fuel assemblies in the reactor vessel during movement of control rods.

The Bases state that this ensures the water removes 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly in the event of a fuel handling accident during core alterations.

However, the movement of control rods is not associated with the -initial conditions of a fuel handling accident, and the Bases do not address any concerns regarding inadvertent criticality which could lead to a breach of the fuel rod cladding.

Inadvertent criticality during Mode 6 is prevented by maintaining proper boron concentration in the coolant in accordance with LCO 3.9.1.

Therefore, this LCO will be relocated.

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20. The ' NRC _ review concluded th,at (1). special test : exceptions 3.10.1 through 3.10.4 may be included with corresponding LCOs which are remaining in Technical Specifications, and (2) special p

L test exception _3.10.5 may be relocated along with LCO 3.1.3.3.

[

LCO.3~10.1'is only applicable in Mode 2.

As discussed in Note 1 above - the SHUTDOWN MARGIN requirements for ' Modes 1-and 2 are-retained in' other Reactivity Control System- -Technical i

specifications.

Retained Special Test Exceptions 3.10.2 and 3.10.3 address Special Test Exception 3.10.1'for LCOs 3.1.3.1 and 3.1.3.6.

Therefore, LCO 3.10.1 will be deleted.

Also, per the stated NRC conclusion, LCO 3.10.5 will be relocated.

LCOs 3.10.2 3

through 3.10.4 will be retained as they are.

21.

Sic specifiestir eili be relcestM and : Ster ;

Trnt

".2 dies:tifity " nitoring "regr : :tzt:- 7t rill be in:1Med in I

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TECHNICAL SPECIFICATION SCREENING FORM (1)

TECHNICAL SPECIFICATION 3.3.3S ACCIDENTMONITORINGfNSTRIAG'NTAllON i

Applicable Modes: 1,2, and 3 (2)

EVA1EATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO

,X, (1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant j

abnormal degradation of the reactor coolant pressure boundary.

,X, (2)

A process variabic, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 1

(3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or -

presents a challenge to the integrity of a fission product barrier.

1 (4)

A structure, system, or component whic.h operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Re instrumentation that satisfies criteria 3 or 4 are the Type A variables in USAR Appendix 7A, as well as the risk-significant variables listed in the discussion below. Some of the 13 3/4.3.3.6 matruments may be relocated to USAR Chapter 16. Others must be retained in the Tecimical Specifications. He Neutron Flux Monitors and the Reactor Vessel Water Level Indicating System (RVLIS) will be added to the Technical Specifications.

if the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Gpecifications.

If the ansur to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

j i

(3)

DISCUSSION Operability of the accident monitoring instrumentation ensures that suflicient information is available on selected plant parameters to monitor and assess these variables following an accident. He instrumentation allows the operator to verify the response of automatic safety systems and to take preplanned manual actions to accomplish a safe plant shutdown.

ne accident monitoring instrumentation is not intended to be a leading indicator of RCS leakage. Although accident monitoring instruments respond to the consequences of a LOCA, the instruments captured by criterion 1 are those that are latended to prevent a LOCA from occurring and to give some indication of RCS leakage prior to the LOCA. Derefore, accident monitoring instrumentation TS is e applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degr# za of the RCPD and does not satisfy criterion 1.

Accident monitoring li.hstation is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission pmduct barrier. Although some variables that are accident monitoring instruments may also establish initial conditions at the time of a DBA or transient (for example, pressurizer leve;), the post-event function is separate and distinct from the pre event function. Derefore, the accident monitoring instnunentation TS does not satisfy criterion 2.

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Anstliery Feedwater Flow Rate !

• Ihe auxiliary feeduster flow rate indicater luis not been shown to be sipiflamme to public h' anith and smisty by either -

i operational expwienes er PSA. R is not included in the WCOS IPE model. Houever, the AFW flow rate ind==e-should be retained for arveral reasons. First, it is included in NURBO 1431 Table 3.3.3-1. 6cand, SR 4.7.1.2.1 regunos AFW flow rate indic=ha= Third. AFW flow rate 3=dae=a==i is being renamed in 13 Table 3.3-9 for the j

. auxthery shutdown panel (ASP).' If an IED and SR for the ASP AFW Sow rate indication ese being :=e===d, than j

they should also be retained in 13 3.3.3.6.-

5

12..

FORY and PORY ins.,ar Velve Positten.aa u

PORY and POR.Y block valve positaan nuhanters(;\\\\ be rtlocA red.

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^

R is esther noted that these indicaters are not Type A variables at Wolf

's""

A4.0 Creek, nor are they RO 1.97 Categosy 1. " ^', "

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13.

Safety Valve Position Indicator i

This instrument b not a Type A or Category 1 indu=tian The safety valve position indicators have not been shoes l

to be signiikant to public health and safety by either operational experience or PSA, and they are not included in the WCOS IPE model. This instnanent is a Type D, Category 2 vanable and will be relocated.

l 14.

Unit Vent -IEgb Range Noble Gas Manitor l

1his instnanent is not a Type A or Category 1 indication. lhe unit vent. high range noble gas monitor has not been i

shown to be significant to public health and safety by either operational experience or PSA. It is not included in the WCOS IPE model. The WCOS Ensegancy Plan also has provissons to issue assite Protective Action l

Recomnwad=tions beasd on plant condition. ladar=taan fresa this anonitor'would not bc requued to make those detenamstions. This is a Type D, Category 2 variable and will be relocated.

e i

(4)

CONCLUSION This Technmal Specifiestaan is retained.

'As indicated above; Neutron Flux and RVLIS will be added.

The Techmcal Specification any be relocated to the following controlled document (s)-

foW foshWiw er,: pN e de 1*5NaaMed..

valve tion indicator;

• • USAR Chaptar 16 (Containment pressure-Extended-range; Sa Unit vent wide.ampe noble gas mani

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TECHNICAL SPECIFICATION SCREENING FORM (1)

TECHNICAL SPECIFICATION 3.3.3.11 EXPLOSIVE GAS MONITORINGINSTRUMENTA110N

(

Applicable Modes: Durmg Waste Gas Hold S

9 ystemoperation (2)

EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO 2.

(1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

J.

(2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X, (3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

2.

(4)

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is 'YES*, then the Technical Specification shall be retained in the Tecimical Specifications.

If the answer to all four of the above questions is 'NO', the Technical Specification may be relocated to a controlled document.

(3)

DISCUSSION The explosive gas monitoring instrumentation provides the capability to detect the concentration of oxygen and hydrogen in the waste gas holdup system (at the hydrogen recombiners) and provide an alarm if the concentrations exceed prescribed limits. According to LCO 3.3.3.11, this TS assures the operability of the instrumentation required for LCO 3.11.2.5, Explosive Gas Mixture of the Radioactive Efiluents TS. According to the Bases of LCO 3.11.2.5, the purpose of the limits on explosive gas concentrations and the monitoring instrumentation is to prevent an explosion in the waste gas holdup system.

(The Bases for 3.3.?.11 were deleted in Operating License Amendment No. 42.) An explosion could result in a release of radioactive materials contained in the gaseous waste holdup system. Although release of the contents of a waste gas decay tank is an analyzed DBA, the analysis assmnes that the tank ruptures non-mechanistically and not as the result of a hydrogen explosion. Herefore, the explosive gas limits are not an initial condition of a DBA.

De explosive gas monitoring instrumentation is not applicable to installed instrumentation used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB; therefore, this IS does not satisfy criterion 1.

He explosive gas monitoring instru;nentation is not applicable to a process variable, design feature, or operating restriction that is an initial condition of any DBA or transient analysis. Thus, this 13 does not satisfy criterion 2.

1 The explosive gas monitoring instrumentation is not assumed to function in the safety analysis. It is not a part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this 13 does not satisfy criterion 3.

From Reference 2, the Explosive Gas Monitoring Instrumentation has not been shown to be significant to public health and safety by either operational experience or PSA. He function of this instrumentation is to preclude inadvertent radioactivity releases fmm the Waste Gas lloldup System due to a tank failure from a waste gas explosion. Severe accidents dominate public risk, not inadvestent releases. His system is not modeled in the WCGS IPE. Thus, this TS does not satisfy criterion 4.

e '

Y

/

9 4

l (4)

CONCLUSION i

His Tarhaical Specirmabanis ed===d I

i

- 2.

The Technical Specificdian may be relocated to the following controlled @-*(s):

' USAR Chapter 16. (The LCO will be relocated but a programwwill be added to new

. TS Section 6.8.5).

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TECHNICAL SPECIFICATION SCREENING FORM

^

(I)

TECHNICAL SPECIFICATION 3.3.4 'IURBINE OVERSPEED PRO 1TC110N Applicable Modes: 1,2, and 3 (2)

EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO 1

(1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

1 (2)

A process variable, design feature, or operating restriction that is an initial condidon of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or.

presents a challenge to the integrity of a fission product barrier.

I 1

(4)

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is 'YES", then the Technical Specification shall be tr'ained in the Technical Specifications.

t

'If the answer to all four of the above questions is 'NO", the Technical Specification may be relocated to a controlled -

document.

I (3)

DISCUSSION f

he Turbine overspeed Protection System actuates to mitigate a potential turbine overspeed event. His prevents the generation of potentially damaging missiles from the turbine. De turbine overspeed event is not a DBA. His event is evaluated to detennine the probability of damage to equipment needed for safe shutdown. He turbine has a favorable orientation from the standpoint of low trajectory missiles; however, the combination of overspeed probability with high trajectory strike probability must meet the NRC's requirements for overall probability, i.e., less than 1 E-7 per year.

He Turbine Overspeed Protection System is not applicable to installed instrumentation used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

De hrbine Overspeed Protection System is not associated with a process variable, design feature, or operating restriction that is an initial condition of any DBA or transient analysis. Aus, this 15 does not satisfy criterion 2.

He Turbine Overspeed Protection System is not assumed to function in the safety analysis. It does not apply to any SSC that is a past of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Hus, this TS does not satisfy criterion 3.

From Reference 2, the turbine overspeed protection has not been shown to be significant to public health and safety by either operational experience or PSA. PRA studies discussed in Reference 2 indicate that the probability of turbine missile ejection and resultant damage to safety-related structures are so low that they have little or no impact on the quantification of core damage frequency. Due to the physical location and orientation of the turbine at WCGS in relation to the majority of safety-related equipment, these low probab!!ity values will be valid. Turbine overspeed protection is not included in the WCGS IPE model. nus, this TS does not satisfy criterion 4.

7.:

. I 1;

j 3

1 i

I J

.i(4)

CONCLUSION j

t

'Ihis Technical specirwahon is retamed.

]

l 2,.

• lhe Technemi Specification may be relocated to the following controlled document (s):

USAR Chapter 16. s"._ '_"..._, '_ _- '

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TECHNICAL SPECIFICATION SCREENING FORM l

I

- (1)

TECHNICAL SPECIFICATION 3.4.5 STEAM GENERATORS ApplicableModes 1,2,3,and4 I

(2)

EVALUATION OF POLICY STATEMENT CRITERIA j

Is the Technical Specification applicable to:

YES NO

.X, (1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_X, (2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

.X.,

(3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

i J

J.,.

(4)

A structure, system, or component which operating experience or probabilistic safety assessment las shown to be significant to public Iwalth and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be' reMcated to a controlled

.un,--,

e (3)

DISCUSSION His T3 establishes the inservice inspection requirements for the steam generator (SG) tubes which are part of the RCPD. It is intended to maintain the structural integrity of this portion of the RCPB. The LCO requires the SGs to be operable in Modes 1,2,3, and 4; operability in this case refers to the structural integrity of the SG tubes by means of an augmented inservice inspection (ISI) program that is performed periodically during plant outages.

This specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB; and, therefore, this 15 does not satisfy criterion 1.

His specification is not applicable to a process variable or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. He specification is applicable to the design feature of 50 tube strength which comes into play, for example, during a LOCA or MSLB to avoid a combined LOCNSGTR or MSLB/SGl'R event. However, tube integ Py is neither an active design feature nor monitored or controlled during plant operation, rather during shutdown conditions under the SG ISI program. Hus, the structural integrity and assumed passive post-accident perfonnance of the SG tubes is maintained by periodic inspcction.

Derefore, this 15 does not satisfy criterion 2.

The SG tubes are components of the RCS that are part of the prunary success path and stich function to mitigate a DBA or i

transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The post-accident or post-transient performance of the SGs, which is a passive function, is maintained by the periodic inspection and' repair of the SG tubes specified in this LCO. Ilowever, the operability of the SG tubes is not maintained during operation of the plant through any actions performed or parameters monitored by the operating staff. Also, the SG tubes do not perform any active function or actuation required for DBA or transient mitigation. Therefore, this TS does not satisfy criterion 3.

i I

e Reference 2 states that the steam generators have not been shown to be significant to public health and safety by either operational expenence or PSA. Also, for WCGS, the SGlR initiated event contributes less than 1.5% to the core damage frequency. WCOS does have three dominant (i.e. above 1.0E 07/ year) containment bypass sequences resulting from the SGTR initiated event. Hey are: 1. SGIR event, AFW and cooldown fail; 2. SGTR event, failure to stabilize RCS and ruptured SG pressure, secondary side relief valve (RV) closes; and 3. SGIR event, failure to stabilize RCS and ruptured SG j

pressure, accondary side RV sticks open, cooldown fails. However, steam generator bypass is controlled by Technical i

Specification 3.4.6, 'Imkage Detection Systems", which limits the leakage from all steam generators not isolated from the 5

RCS to I sallon per minute. His limitation assures that dosage contribution from the tube leakage will be limited to a small fraction of the 10CER Part 100 done guideline values in the event of a SGTR event. Herefore, this 15 does rot satisfy Criterion 4.

{

Ref. 4 conc, cud east this LCO could be relocated out of TS but that the SRs must be retamed.

l (4)

CONCLUSION X

His Technical Specification is retained. RgWer h re\\c>cde %h h ad @ n p% ram, Amriph % Sec.h to, % Ts ew de retahe6 % -r5,3/4,4,5.

j

-ff-he Technical Specification may be relocated to the following controlled document (s):

-De4EOsneyk;i

' : US.^.R '"_-.;-: M; h::::c, e4G!d: = :" r 7 ;__ _. _ __.. ap,.mww c_.:q e $

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TECHNICAL SPECIFICATION SCREENING FORM y

(I)

TECHNICAL SPECIFICATION 3 4.10 STRUCTURAL DTITGRTIY ApplicableModes: AllModes (2)

EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO 1

(1)

Installed instrumentation that is used to detect, acd indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary X

(2)

A process variable, design feature, or operating res riction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (3)

A structure, system, or component that is part of the primary success path and widch functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (4)

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Tecimical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3)

DISCUSSION His specification provides the inspection requirements for the ASME Code Class 1,2, and 3 components to ensure their stmeturalintegrity.

His specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPD. Derefore, the structural integrity requirements do not satisfy criterion 1.

His specification is not applicable to a process variable, design feature, or operating restriction that is an initial condition of DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

While the E imposes an operating restriction regarding pressure boundary integrity, it is not monitored or controlled during plant operation. He assumed integrity of Class 1,2, and 3 components is assured by means of periodic inspections.

Herefore, this E does not satisfy criterion 2.

ASME Code Class 1,2, and 3 components are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of or present a challenge to the integrity of a fission product barrier. Individual AShE Code Class I,2, and 3 components may satisfy criterion 3 and the requirements that ensure the integrity / operability of these components are included in the individual specifications that cover these components. Ilowever, as stated above, this I

specification addresses the passive, pressure boundary function of these components. Herefore, this E does not satisfy criterion 3.

Ref. 4 concluded that the LCO for this specification could be relocated out of E; however, the associated SR must be relocated to the E programmatic requirements.

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' From Reference 2, the structural integrity of ASME Code 1,2 and 3 -!

.--- " has not been shown to be significant to

.i public health and safety by either operational experience or PSA. Failme modes of these components would not be identified from the requiremads of this *=cl='eal specificahon, & structual integrity of ASME Code 1,2 and 3 components is not modeled in the WCOS IPE. Therefore, this 15 does not satisfy critarion 4.

l t

-(4)-

CONCLUSION.

t This Technical Specification is rd===8

.j

.X.

b Technical Specificahon may be relocated to the following controlled document (s):

}

USAR Chapter 16. (The LCO may be relocated but a progran!hwill be added to new 1E Section 6.8.5)..

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TECHNICAL SPECIFICATION SCREENING FORM (1)

TECHNICAL SPECIFICATION 3.6.1.2 CONTAINMENTLEAKAGE Applicable Modes: 1,2,3, and 4 (2)

EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO J.

(1) lastalled insaumentation that is used to detect, and indicate in the control room, a significant -

abnormal degradation of the reactor coolant pressure boundary.

J.

(2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product bamer.

2.

(3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

2.

(4)

A structure, system, or component which operating experience or probabilistic safety==ne== ment has shown to be significant to public health and safety.

If the answer to any one of the above questions is 'YES*, then the Technical Specification shall be retamed in the Technical Specifications.

If the answer to all four of the above questions is *NO", the Technical Specification may be relocated to a controlled document (3)

DISCUSSION This 13 identifies the allowable leakage rates for the containment structure which are established to meet 10 CFR 50, Appendix J. Rese requirements ensure that the leakage rates from containment will not exceed the value assumed in the safety analyses at the peak accident pressure.

His specification is not applicable to installed instrumentation that is used to deter *, nnd indicate in the control room, a significant abnormal degradation of a the RCPB; and, therefore, the 13 does not saGfy criterion 1.

His specification is applicable to parameters that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the process variables for which the requirements are applicable (contamment design pressure and allowable leakage rates) are not variables that are monitored and controlled during power operation such that process values remam within the analysis bounds. Containment integrity is assured by periodic inspection and testag. Herefore, this specification does net satisfy criterion 2.

He specification applies to containment leakage rate limits. Rus, it is applicable to a structure that is part of the primary success path and which function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the intent of criterion 3 is to capture only those SSC (and supporting systems) that are part of the primary success path of a safety sequence analysis. Operability of the containment is assured by a separate LCO (3.6.1.1), and the limits imposed by the leakage rate requurments are neither monitored or controlled during operation nor part of the primary success path of the containment function. Derefore, this 13 does not satisfy criterion 3.

lll From Reference 2, containment leakage las not been shown to be significant to public health and safety by either operational cwh or PSA.' PRAs indicate that risk is dominated by events in which the containment is bypassed, unisolated, or fails

. structurally. De technical specification value for overall containment leakage is included in the WCOS MA AP model, but m

contributes only a small fraction of the total release in the lael 2 IPE. Derefore, this 'IB does not satisfy criterion 4.

Ref. 4 concluded that this LCO could be retarmal out of"IE but that the limiting values of Paand Le must be retamedin75.

(4)

CONCLUSION i

his Technical Specificauon is ret===I He Technical Specification may be relocated to the following controlled document (s):

~* He LEO may be relocated to USAR Chapter 16, but the limiting values of Paand Lawillbe retamed in -

the Containment lategrity Bases.nRelocation of the LCO requires that revisions be made to SR4.6.1 lc and SR 4.6.1.7

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l TECHNICAL SPECIFICATION SCREENING FORM (1)

TECHNIC (L SPECIFICATION 3.61.6 CONTAINMENTVESSEL STRUCTURAL INTEGRITY Applicable Modes: 1,2,3, and 4 G)

EVALUATION OF POLICY STATEMENT CRITERIA 4

i Is the Technical Specification applicable to:

YES NO 1

1 (1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

1 (2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (3)

A stmeture, system, or component that is part of the primary success path and utich functions or actuates te mitigate a Design Basis Accident or Transient that ei'her assumes the failure of or l

1 presents a challenge to the integrity of a fission product barrier.

1 (4)

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

'If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled Am nment p)

DISCUSSION i

ne containm st serves as a barrict to prevent the release of fission products following a LOCA or MSLB inside containment.

1 To mitigate the potential consequences of a DBA, it is necessary that the containment structure meet its structural requrrements. This specification is intended to detect abnormal degradation of the contamment structural elements. This 13 outlines an appropriate mspection and testing pmgram to demontrate this capability. The program consists of the measurement of tendon liftofiforce, tensile tests of tendon wires, and visual examination of tendons, anchorages and exposed j

mienor and exterior surfaces of the containment.

l his specification is not applicable to installed ietrumentation that is used to detret, and indicate in the control room, a significant abnormal degradation of a the RCPB; and, therefore, this TS does not satisfy criterion 1.

This specification is applicable to a design feature (the containment) that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Containment structural integrity is assumed to be available for many DBAs. However, containment structuralintegrity is not monitored or controlled during plant operation but, rather, via periodic inspections and tests. Therefore, this is does not satisfy criterion 2.

The specification applies to the detection of abnormal degradation of containment stmetures and therefore to cor>+ainment structural integrity. Hus, it is applicable to a structure that is part of the primary success path which functions to mitigate a i

DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Ilowever, the fu tetional mode addressed by the is is mainhining the passive, pressure boundary integrity. This 13 does not eIdress the capability of the containment to function or actuate in order to mitigate the consequences of a DBA or transient.

Therefore, this TS is not required to ensure the operability of containment and, thus, does not satisfy criterion 3.

i i

~

Ref. 4 concluded that this LCO could be relocated out of TS but that the associated SRs should be retained to meet the operability requirements for a retamed LCO, in this case LCO 3.6.1.1. Ref. 2 incorporated the SRs regarding tendon surveillance into Section 6 of the 13.

From Reference 2, contamment leakage has not been shown to be significant to public health and safety by either operational experience or PSA. PRAs indicate that risk is dominated by events in which the containment is bypassed, unisolated, or fails structurally. None of the sequences addressed in the contamment and source tenn analysis could realistically threaten containment due to hydrogen combustion. No WCGS containment vulnerabilities were identified as a result of Supplement 3 to Genenc leter 88-20.

"Ihe WCGS IPE also showed that the overall release frequency per year is dominated by releases due to containment bypass sequences of which interfacing system LOCAs make up the vast majority.

The best estunate containment failure mode will occur at 2.13 times the design pressure due to membrane stresses in the contaimnent mid-height region which exceed the pre-stress and cause through-concrete cracking and yielding of the liner, reinforcing steel, and pre-stressing tendons 'Ihe material properties used in these calculations do not change rapidly, so testmg and inspection requirements of this technical specification are not critical. Therefore, this TS does not satisfy criterion 4.

(4)

CONCLUSION This Technical Specification is retamed.

jf.,

'Ihe Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (The LCO may be :rlocated; but a pro will be added to new 13 Section 6.8..

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'U CHNICAL SPECIFICATION SCREENING FORM (1)

TECHNICAL SPECIFICATION 3.7.8 SNUBBERS Applicable Modes: 1,2,3, and 4. Also Modes 5 and 6 for those systems required to be operable in these Modes.

(2)

EVALUATION OF POLICY STATEMENT CRITERIA is the Technical Specification applicable to:

YES NO 1

(1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

1 (2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (4)

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significaat to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retamed in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3)

DISCUSSION l

ne snubbers are required to be operable to ensure that the structural integrity of the RCS and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. He reoinig action of the snubbers ensures that the initiating event failure does not propagate to other parts of the failed system or to other safety systems.

Snubbers also allow normal thermal expansion of piping and nozzles to eliminate excessive thermal stresses during heatup or cooldown. Snubber surveillance is conducted under the requirements of the Wolf Creek Snubber Surveillance Program.

He E requirements for stubbers are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this B does not satisfy criterion 1.

He snubber TS is associated with a design feature or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Ilowever, the snubber requirements are not explicitly considered in the accident analysis. He availability of the snubbers is assumed based on the performance of a program of periodic augmented inspection and testing. Snubber operability is not required to be monitored and controlled during plant operation. Some snubbers (inaccessible) can only be inspected during plant outages.

i Rus, this U does not satisfy criterion 2.

Those snubbers that are required to function during DBAs or transients to prevent the initiating event from propagating to other systems or components that are part of the primary success path may be considered components that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Ilowever, snubbers are not explicitly considered in DBA or transient analyses but are a structural / design feature whose operability is assured by an inspection program. Herefore, tids B does not satisfy criterion 3.

i i

I From Reference 2, for non-RCS and other high energy systems such as the feodwater and main steam systems inside the l

containment building, the snubber technical specification has not been shown to be significant to public health and safety by either operational expenence or PSA. Reference 2 reviewed the Zion and Millstone PRAs and determmed that the snubbers, which ensure the operability of certam safety-related equipment during a seismic event, are not risk dominant for this scenano. While the seismic portion of the WCOS IPEEE is not complete at t% time, there is no reason to believe the results will be different. Thus, this TS does not satisfy critenon 4.

l For snubbers which are not past of the RCS or other high energy systems, then, this technical specification can be relocated.

(4)

CONCLUSION j

'Ihis Technical Specification is retamat-

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TECIENICAL SPECIFICATION 3.712 AREA TEMPERAlVRE MONfIDRING Applicable Modes: Whenever egmp-ar in the area is reqmrod to be OPERABLE.

Q) -

EVALUATION OF PO1JCY STATEMENT CRITERIA Is the Technical Specification applicab:e to:

YES NO j

1 (1)

Indaltad insta==aatd-that is used to detect, and indicate in the control room, a significant abnonnel degradstion of the reactor coolant pressure bor. Gary, i

1 (2)

A process variable, design featur. w operatag restriction that is an initial condition of a Design l

Basis Accident or Tranment analysis that either assumes the failure of or presents a challenge to the integrity of a tission product bemer.

{

3 1

(3)

A structure, system, or component that is part of the pnmary success path and which functions or actuates to mitigate a Demgn Basis Accident or Transient that either assumes the failure of or

~

presents a challenge to the intagnty of a fission product barner l

-1 (4)

A structure, system, or {

, - ' which operstmg experience or probabilistic safety===e== ment I

has shown to be significant to public health and safety.

l'i If the answer to any one of the above questions is "YES*, then the Technical Specification shall be retamed in the Technical Specifications.

If the answer to all fow of the above ge== is 'NO*, the Technical Specification may be relocated to a controlled '

decoment.

p)

DISCUSSION Dis specification places a limit on the '._,, __. of the areas of the plant which contain safety-related eqmpment. His is -

required to ensure that the temperature of the eqmpment does not exceed its environmental qualiGcation temperature dwing nonnal operation. Exposure to excessively high temperatures may degrade the eqmpment and cause a loss ofits operability.

De 13 regareinants for area -- ; =;, monitoring see not applicable to installed instrumentation used to detect a significant abnonnat degradataan of the RCPB; therefore, this TS does not satisfy criterion 1.

The area temperstwo monitoring TS is associated with the variable of room temperstwe which is not a process variable, design feature, or operstmg restriction that is an initial condition of a DBA or transient analysis that either assumes the failme of or presents a challenge to the integnty of a fission product barner. Rus, this 15 does not satisfy criterion 2.

De T5 for ares temperature momeonng does apply to the WiF*y of SSC: that are part of the primary success path which functions or actuates to mitigate a DBA or transiast that either assumes the failme of or presents a challenge to the integrity of a fission product barrier. However, the is is only induectly applicable to the operability of these systems and cP-Derefore, this TE does not satisfy criterion 3.

Froen Reference 2, the area temperstwe monitors have not been shown to be significant to public health and safety by either operational experience or PSA. The area temperstme monitors have not been included in the WCOS IPE. Herefore, this 15 does not satisfy criterion 4.

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- (4)

CONCLU580N

' Ibis Technical Specification is wtaiW I:

2.

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TECHNICAL SPECIFICATION SCREENING FORM (1)

TECHNICAL SPECIFICATION 3.1L1.4 LIOUIDliOLDUPTANKS ApplicableModes: Atalltimes (2)

EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO 1

(1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary 1

(2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

1 (4)

A structure, system, or component widch operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES*, then the Technical Specification shall be retained in the Technical -

Specifications.

If the answer to all four of the above questions is *NO", the Technical Specification may be relocated to a controlled document.

(3)

DISCUSSION He liquid holdup tank specifications impose limits on the quantity of radioactive material contained in specific outdoor tanks that may contain radwaste. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentration would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area. The tanks addressed by this specification are:

a.

Reactor Makeup Water Storage Tank b.

Refueling Water Storage Tank c.

Condensate Storage Tank d.

Outside temporary tanks, excluding demineralizer vessels and liners being used to solidify radioactive wastes.

Dese tanks are not addressed by the safety analysis of radioactive release from a subsystem or component.

He 'IS requirements for liquid holdup tanks are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

He liquid holdup tanks 13 is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission 4

product barrier. Aus, this TS does not antisfy criterion 2.

- _= _

The 15 for liquid holdup do not apply to an SSC that is part of the pnniary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failme of or pr= ente a challenge to the integrity of a fission product barner. Therefore, tids is does not satisfy criterion 3.

From Reference 2, the liquid holdup tanks, which hold radweste, have not been shown to be significant to public health and safety by either operational experience or PSA. Risk of r= L-- u uy release is domin= tad by severe accidents, not releases of radionuclides generated from normal operations For this reason, the liquid holdup tanks are not modeled in the WCOS IPE.

Therefore, this "IE do not satisfy criterion 4.

(4)

CONCLUSION 1his Technical Specification is retamed.

2.

'Ihe Technical Specification may be relocated to the following controlled document (s):

USAR Chapter 16. (The LCO may be relocated but a programmwill be added to new 15 Section 6.g.5).

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TECIDGCAL SPECD1 CATION SCR. EENING FORM

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TECHNICAL' SPECIPICATION 3.1155 EYPLOSIVE GAS MIXTURE l

Applicable Modes: At alltimes 1

.i

- (2).

EVALUATION OF POLICY STATEMENT CRITERIA i

i Is the Techmcal Specificatena applicable to:

-j

. YES NO jl 1

1 (1) 1=d=Il=d instr==w=* ion that is used to detect, and indicate in the control room, a significant

.i abnormal '. " ' - of the reactor coolant pressure boundary.

~

j 1

(2)

A process variable, design feature, or operstmg resenction that is an initial condition of a Design Basis Accidet or Tramment analysis that either assumes the failure of or presents a challenge to the intagnty of a fismen product barrier.

1 (3)

A structure, systeni, or counponent that is part of the pnmary success path and which functions or '

actuates to mitigate a Demgn Basis Accident or Transient that either assumes the far.:n of or presents a challenge to the insegnty of a fission product barrier.

l 1

(4)

A structure, system, ori, N which operating experience or probabilistic safety assessment l

has shown to be significant to public health and safety.

l

\\

If the answer to any one of the above questicas is 'YES", then the Technical Specification shall be included in the new 1

TechnicalSpecifications

),

If the answer to all four of the above questions is "NO", the Technm=1 Specification may be relocated to a controlled j

docenent.

(3)

DISCUSSION I

i Dis specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas I

holdi, system is mainemined below fie f1====hility limits of hydrogen and oxygen.- Main:ainmg these limits provides assurance that the releases of radioactive matenals will be controlled in conformance with the regarements of GDC 60 of Appandiv A to 10 CIR 50. The accident analysis concerning the gaseous redweste system assumes that a storage tank ruptures, from unspecified causes, and releases its contents without mitigation.

i j

De 13 requinsments for explosive gas mixture are not applicable to instattad nstrumentation used to detect a significant i

?

abnonnel degr=d=emn of the RCPB; therefore, this 13 does not satisfy criterion 1.

j De explosive gas mixture 13 is -= tad with a process variable, design feature, or operating restriction that is an initial condition of a DBA or treament analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barner. Dus, this 75 does not satisfy cntenon 2.

i q

De 13 for explosive gas mixture does not apply to an SSC that is part of the pnmary -*== path and which functions or actusses to mitigate a DBA or tranment that either assumes the failure of or presents a challenge to the integnty of a fission product barrier. Derefore, this 13 do not satisfy entenon 3.

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The explosive gas =% of the waste gas holdup tanks has not been shown to be significant to public health and safety by

[

either operational expenance or PEA. Ridt of radioactivity :elease is doeunated by severe accidents, not releases of l

~h r=dia==ctides generated from nonnel operations. In =M*aa fresa Reference 2 the quantity of radmactivity ca=t==ad in each l

pressarized gas storage tank in the waste gas holdup system is limited to assure a release would be substantially below the.

does guideline values of 10 CPR Part 100. The waste gas bokhy tanks are not modeled in the WCOS IPE. Therefore, this 15 -

- does not satisfy catenas 4.

i i

' CONCLUSION

-i (4) s Ibis Technical",

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TECHNICAL SPECIFICATION 3.11.2 6 GAS STORAGE TANKS Applicable Modes: At alltimes p)

EVALUATION OF POLICY STATEMENT CRITERIA la the Technical Specification applicable to:

YES NO

.X.

(1)

Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

.X.

(2)

A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

[

.X (3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or i

presents a challenge to the integrity of a fission product barrier.

.X.

(4)

A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

if the answer to any one of the above questions is "YES", then the Tecimical Specification shall be retained in the Technical l

Specificatens.

' If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled d-=~d p)

DISCUSSION The gas storage tank specifications impose limits on the quantity of radioactive material contained in those tanks for which the quantity of radioactivity contained is not limited dinctly or indirectly by another 'IS. Restricting the quantity of sudioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of the public at the nearest site boundary will not exceed 0.5 rem.

'Ihis is consistent with Branch Technical Position l'ISB 115, " Postulated Radioactive Releases Due to a Waste Oss System leak or Failure." "Ibe accident analysis concernmg the gaseous mdwaste system assumn a rupture of a storage tank without mitigation.

"Ihe 13 requirements for gas storage tanks are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The gas storage tank 'IS is associated with a process variable or operating restriction (quantity of contained radioactivity) that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. liomer, the barrier in this case is the tank itself which is not a barrier that is monitored and controlled during power operatic 4 the plant "Perefore, this TS does not satisfy criterion 2.

The TS for gas storage tanks does not apply to an SSC that is part rif the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product bamer Therefore, this TS does not satisfy criterion 3.

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Fman Reference 2, the waste gas holdup tanks, which hold radweste, haive not bosa shown to be significant to public health and safety by either operatonal experience or PSA. Risk of radioactivity release is dominated by severe accidents, not' selsenes of redenuclides generated from normal operations. In aMition, from Reference 2 the quantity of radioactivity contained in each pressurized gas storage tank in the waste gas holdup system is limited to assure.a release would be '

shially below the dose guideline values of 10 CFR Part 100. De waste gas holdup tanks are not modeled in the WCOS.

IPE. Derefore, this 13 does not satisfy criterion 4.

(4)

CONCLUSION i

This Technical Specification is retamed.

.X.

The Technical Specification may be relocated to the following controlled document (s):

t i USAR Chapter 16. (The LCO rnay be relocated but a is-- - -- ^

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