ML19301E618

Summary of January 28, 1986, Meeting W/Aif Subcommittee on Tech Spec Improvement - Wolf Creek TS Split

ML19301E618
Person / Time
Site:	Wolf Creek
Issue date:	02/06/1986
From:	Desai K Office of Nuclear Reactor Regulation
To:	Butcher E Office of Nuclear Reactor Regulation
References
NUDOCS 8602200650
Download: ML19301E618 (63)

Text

r.b aiA February 06, 1986 FEMORANDUM FOR:

Edward J. But-cher, Chief Technical Specifications Coordination Branch Division of Human Factors Technology, NRR FROM:

Kulin D. Desai, Reactor Engineer Technical Specifications Coordination Branch Division of Human Factors Technology, NRR

SUMMARY

OF MEETING WITH AIF SUBCOMMITTEE ON TECHNICAL SPECIFICATIONS IMPROVEMENT - WOLF CREEK TS SPLIT The NRC staff and the AIF TS subcommittee members met on January 28, 1986 in Bethesda, MD. te discuss Westinghouse - Wolf Creek Technical Specifications split using AIF/TSIP proposed criteria.

The purpose of this working session was to:

1.

determine validity and usefulness of the criteria; 2.

identify areas of agreement and disagreement of Wolf Creek TS split; 3.

discuss the differences and resolve these issues; and 4

identify any defects within the criteria for improvement or clarity.

Our meeting was constructive and helpful to all parties. The overall ccnclusion was that the proposed criteria work very well, however, criterion

• 2 and #3 need further clarity to be completely effective.

The Wolf Creek Technical Specifications have 133 Limiting Conditions for Operation (LCO). Our Wolf Creek TS split corrparison identified 34 LC0 as disegreements between the staff and the AIF. These 34 LC0 were discussed in detail for resolution. Out of these 34 LCO, we resolved the disagreement for 22 during this working session. The remaining 12 LC0 represent the differences due to criteria definition and interpretation, surveillance assurance requirements and surveillance of instrumentation related issues as listed below:

1.

Criteria Jefinional Differences LC0 3.1.1.3 Moderator Temperature coefficient (MTC)

' C0 3.2.2 Heat Flux Hot Channel Factor - FQ(Z)

LCO 3.2.3 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel Factor LCO 3.6.1.2 Containment Leakage 2.

Surveillance Assurence Requirements (LC0 that seem to be in existence only to assure appropriate surveillance)

B602200650 860206 CF ADOCK 05000482 CF

Edward J. Butcher February 06, 1986 LCO 3.4.5 Steam Generators (Inservice Inspection of Steam Generator Tubes)

LC0 3.4.10 Structural Integrity (Inservice Inspection of ASME Code Class 1, 2 and 3 components including each reactor coolant pump flywheel and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves)

LC0 3.7.8 Snubbers (Inservice Visual Inspections of Snubbers) 3.

Surveillance Instrumentation (These requirements can be combined with other LCO as surveillance requirements)

LCO 3.1.3.2 Control Rod Position Indication Systems - Operating (Digital and Demand)

LC0 3.1.3.3 Control Rod Position Indication Systems - Shutdown (Digital only - for Centrol Rods not fully inserted)

LCO 3.1.3.4 Pod Drop Time LCO 3.7.12 Area Temperature Monitoring 4

LCO 3.4.9.1 RCS - Pressure / Temperature Limits (Reactor Vessel material Surveillance Program required by 10 CFR 50, Appendix H and ASME Section 111 Appendix G)

This meeting was our first attempt for the criteria application to the Wolf Creek TS.

It is believed that the 12 remaining areas of disagreement can be resolved at future meetings between the NRC and the AIF.

The staff and the AIF TS subconmittee members are in the process of preparing a similar TS split for Linerick - a BWR plant. We plan to meet on February 26, 1986 todiscusstheresultofthiswork.

\\S) b 1 MW Kulin D. Desai, Reactor Engineer Technical Specifications Coordination Branch Division of Human Factors Technology, NRR Encloscres:

1.

List of Attendees 2.

Proposed Criteria 3.

Staff Wolf Creek TS Split 4.

AIF Wolf Creek TS Split cc w/encis:

Distribution H. Denton, NRR R. dernero, NRR D. Vassallo, NRR T5CB Rdg.

D. Eisenhut, NRR R. Iay, MITRE

5. Newberry, NRR W. Pusse11,itRR W. Cunningham, M:TRE R. Emch, NPR D. " ~ nann, NRR V. Benaroya, NRR TSCB Members H. Thompson, NRR W. Regan, NRR TSCB:DHFT:NRR KDDesai:dlm 02/6/86

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ATTACHMENT 1 Criterion 1: An installed system that is used to detect, by monitors in the control room a significant abnormal degradation of the reactor coolant pressure boundary, or:

DISCUSSI_0_N: A basic concept in the protection of the public health and safety is the prevention of accidents. Systems are installed to detect significant abncrmal degradation of the reactor coolant pressure buundary so as to allow operator actions to either correct the condition or to shutdown the plant safely, thus reducing the likelihood of a loss of coolant accioent.

This criterion is intended to ensure that Technical Specifications control those systems that detect excessive reactor coolant system leakage. Two specific examples of systems which are selected using Criterion 1 are:

Secondary System Radiation Monitors Reactor Building Sump Level Instrumentation Criterion 2: A process variable that is an initial condition of the Design Basis Accident Analysis, cr; DISCUSSION: Another basic concept in the protection of the public health and safety is that the plant shall be operated within the bounds of the initial conditions assumed in the existing Design Basis Accident (DBA) analysis. These analyses consist of postulated events, analyzed in the FinalSafetyAnalysisReport(F5AR)ionalgoals.

for which a structure, system, or com>onent must meet specified funct These analyses are contained in Chapters 6 and 15 of the F5AR (or equivalent chapters) and are identified as Condition II III or IV events (ANSI N 18.2) (or equivalent) that either assume the failure o,f or present a challenge to the ir.tr.grity of a fission product barrier.

Process variables are parameters for which specif(c values or ranges of values have been chosen as reference bounds in DBA analyses ard which are monitored and controlled in actual operation such that process values remain within the analysis bounds.

The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA analyses, which are monitored and controlled. So long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low.

2-Implicit in this criterion is the associated installed control room instrumentation that monitors and/or controls the selected process vairiable. Two specific examples of process variables selected using Criterion 2 are:

Koveable Group Assembly Rod Insertion Limits Deactor Coolant System Pressure Limits Criterion 3: A structure, system, or component that is part of the primary success path of a safety sequence analysis and functions or actuates to mitigate a Design Basis Accident.

DISCUSSION: A third concept in the protection of the public health and safcty is that in the event that a postulated DBA should occur, structures, systems, and components are available to function or to actuate in order to mitigate the consequence of the DBA. Safety sequence analyses or equivalent have been performed in recent years and provide a method of presenting the plant response to an accident.

A safety sequence analysis is a systematic examinaticn of the actions required to mitigate the consequences of events considered in the plant's DBA analysis, as presented in Chapters 6 and 15 of the plant's Final Safety Analysis Report. Such a safety sequcnce analysis considers all applicable events, whether explicitly or implicitly presented. The primary success path of a safety Jequence analysis consists of those actions assumed in the design basis accident analysis which limit the consequences of the events to within the appropriate acceptance criterja.

It is the intent of this criterion to capture into Technical Specifications only those structures, systems, components that are part of the primary succpss path of a safety sequence analysis.

Implicit in this criterion are those support systems that are necessary for items in the primary success path to successfully function. The primary success path is equivalent for each DBA to the combinations and sequences of equipment assumed to operate when responding to the event which results in acceptable plant accidant response (including consideration of the single failure criterion).

Two specific examples of structures, systems, and components which are selected using Criterion 3 are:

Reactor Trip System Instrumentation Primary System Safety Valves

ENctowns 3

DRAFT TEST APPLICATION OF TSIP TECHNICAL SPECIFICATION SELECTION CRITERIA "o

TEST APPLICATION OF TSIP TECHNICAL SPECIFICATION SELECTION CRITERIA TO VOLF CREEK TiiCHNICAL SPECIFICATIONS Backcround: On September 30, 1985, the Final Report of the Technical SpecificationImprovem:ntProjectwasforwardedtoHaroldDenton. This report discussed the problems assou.ted with Technical Specifications and included conclusions and recomendations of T5IP. One of the three root problems was lack of well defined criteria for what requirements should be included in Technical Specifications. T5IP recomended that a Comission Policy Statement be prepared to articulate the scope and putpose of Technical Specifications.

This Policy Statement would include specific criteria to identify Technical Specification content. After many discussions, including meetings with the AIF and the T5IP Advisory Group, criteria for selecting Technical Specifications were derived and recomended. These criteria would be used on a voluntary basis by licensees to determine which requirements would remain in Technical Specifications and which requirements would be placed in another controlled document. Detailed discussions on the criteria can be found in the T51P Final Report, Section 2.2.1, and in the AIF Technical Specification Ic:provement Report of October 1, 1985 (letter to H. Denton from M. Edelman dated October 8,1985). It was determined that one of the next necessary steps was to apply the criteria. This report describes this first appiitation of the recomended criteria.

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==

Description:==

The purpose of this trial application is to verify that the criteria work as described in the TSIP report and, if not, to make recommendations as to how the criteria should be altered or supplemented.

This application is also intended to provide the Technical Specification Coordination Branch (TSCB) with detailed results for their consideration and to serve as a basis for continuing dialogue with the industry and HRC staff. Verification of the success of the criteria is based on their practicality and clarity for application and a review of whether the final Technical Specifications capture those systems, components, and variables most important to safety. This determination is obviously based on judgement. In addition, special attention was paid to those LCOs that would be removed from Technical Specifications and have Action Statement that limit reactor power in some way, including shutdown. Each of these specifications.was specifically noted and recocnendations made on where they should gc.

This trial application is made on one PWR and one BWR set of Technical Specifications. The PVR used was Wolf Creek. The EVR used was Limerick.

The criteria are shown in Attachment I with supporting discussion (AIF report).

Included in this report are detailed results, conclusions and recommendations. These conclusions and recommendations, especially the reconoendations, should be considered preliminary. It is hoped that they will serve as a point of departure for the TSCB to continue dialogue with the industry and the NRC staff.

Results: Enclosures I and 2 provide the detailed results of the trial application. Enclosure 1 is the application of the criteria to the Wolf Creek Technical Specifications. Enclosure 2 covers the Limerick application.

Each Limiting Condition for Operation (LCO) in Section 3 of each plant's Technical Specifications is listed individually with its purpose (s). A column is then provided to indicate whether the specification remains in the tech-specs, or not, and if it remains, which criterion apply. The next column states d ether or rat the associated ACTION STATEMENT for the LCO requires a reactor shutdown after some time, or limits reactor power in some way. The last column provides coments and is intended to address:

1.

Interpretations of or difficulties with the criteria.

2.

Appropriate coments on the LCO importance (subjective criteria) 3.

Weaknesses in the BASES.

. provides a " count" of limiting conditions of operation (LCO) for ehchsetofTechnicalSpecificationstoprovideaperspectiveofhowmanystay and leave following the application of the criteria. Also indicated are how many LCOs have a reactor power limitation and how many do not. It is estimat(d that these criteria would allow placing about 40% of the current LCOs in other controlled documents. is a listing, or potpourri, of insights and comments thht should be used to supplement the details of enclosures 1 and 2, and this criteria application in general.

- Enclosure 5 is a discussion and list of the Limiting conditions of Operation (LCOs) that would leave the Technical Specifications, but have a power limitation of some sort.

==

Conclusions:==

1.

The proposed criteria, in general, provide on effective means to determine which Limiting Conditions for Operation (LCO)s) should remain

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in the Technical Specifications. With some exceptions, which are rather minor, the rer.aining LCOs appear to address those systems and components which are of immediate importance to the public health and safety. This conclusion is based on judgement regarding what systems are necessary to shutdown the reactor, cool the reactor, and provide containment. (To the extent possittle, risk assessment insights were considered - Core neit or core damage risk ar.d public health risk).

2.

Some rather unimportant (from a risk perspective) LCO's remain in the Technical Specifications. These include safety analysis initial conditions and non-reactor related requirements such as rad-vaste tank limits.

3.

Several minor problems with the criteria ant the supplementary information, such as definitions, were encountered:

6-a.

The term, " Design Basis Accident (DBA)" is not precise enough, causes confusion, and in some respects is wrong. The AIF report describes a DBA as a hypothetical event that is not expected to occur (pg.12). Yet, the definition of DBA on pg.14 includes Condition 11 events which are anticipated events analyzed in the FSAR.

b.

The criteria do not cover the normal decay heat removal function provided by the residual heat removal system.

c.

The criteria do not clearly cover the reactor vessel pressure-temperature limits during all modes of operation (see g below).

d.

Criterion 1 refers to systems to detect abnormal degradation of the reactor coolant pressure boundary in the control room. The "in the control room" limitatien appests to be an extraneous holdover from earlier proposed criteria.

e.

The term " process variable" in criterion 2 does not include, conditions, or assumptions which are important initial conditions of a Chapter 15 safety analysis. An example is the pressure interlocks on the RHR suction valves. These interlocks help limit LOCA; tc thos; that occur inside the containment. An interfacing systems LOCA is beyond the design basis.

f.

None of the criteria c::plicitly treat the refueling mode.

g.

Lov temperature overpressure protection transient analyses is not ecnsidered under Criterion 2 and Criterion 3.

This analysis is not found in Chapter 6 or 15.

Recomendations:

1.

This trial application, with the several problems and proposed solutions, should be considered by the TSCB and used as a basis for discussion with NRC and industry. The criteria worked reasonably well considering that this was the first real application.

2.

To support the subjective criteria, which pertains only to reactor operation, Criteria 2 and 3 should only apply to reactor transients and accidents analyzed in the F5AR. This clarification would remove all "non-reactor" LCO's (rad-waste tank limits) and possibly some refueling requirements. Previous studies such as VASH-1400 Jave estiented the rit L associated with all sources of radioactivity on a site. Studies conclude that a gross release of radioactivity can occur only if fuel melts and that, while releases involving waste storage tents would be " troublesome" particularly to in plant personnel, they could not result in public consequences nearly as serious as eccidents involving melting of fuel in the reactor core or spent fuel pool.

.g.

3.

The term Design Basis Accident should be replaced by safety analysis,"

" transients e.nd accidents," or a similar term, supplemented by the current AIF discussion of Condition II, III and IV events.

4.

The discussion on Criterion 3 should make it clear that decay heat removal is a necessary part of a primary success path for core cooling, such that the Residual Heat Removal System is included in the Technical Specifications.

5.

Criterion 2 discussion should be clarified such that conditions or assumptiens (like the RHR interlocks) which are feportant bounds of the safety analysis are covered.

6.

The phrase by monitt:rs in the control room" should be deleted from the first criterion.

7.

Lov temperature overpressure protection transient analyses should be considered part of the safety analysis (currently DEA) envelope, such that it is covered by Criterion 2 and Criterjon 3.)

8.

k'hile the criteria have been successful in focusing on the t.ey safety systems, the Technical Specifications, because the criteria tre based on the traditional licensing approach, will have requirements that are auch less icportant than others. It is recommended that ACTION

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STATEMENTallowedoutagetimeadjustmentsorflexibilitybeusedto reflect importance rather than removing iter.s from Techn'tal Specifications arbitrarily or thorugh additional revisi m of the criteria.

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ATTACHMENT 1 Criterion 1: An installed system that is used to detect, by monitors in the control room a significant abnormal degradation of the reactor coolant pressure boundary, or:

DISCUSSION: A basic concept in the protection of the public health and safety is the prevention of accidents. Systems are installed to detect significant abnormal degradation of the reactor coolant pressure boundary so as to allcw operator actions to either correct the condition or te shutdown the plant safely, thus reducing the likelihoed of a loss of coolant accident.

This crituse

's intended to ensure that Technical Specifications control those syr%i_, sat detect excessive reactor coolant system leakage. Two specific examples of systems which are selected using Criterion 1 are:

Secondary System Radiation Konitors Reactor B.rilding Sump Level Instrumentaticn Criterion 2: A process variable that is an initial condition of the Design Easis Accident Analysis, or; DISCUSSION: Another basic concept in the protection of the public health and saf ety is that the plant shall be operated within the bounds of the initial conditions assumed ir. the existing Design Basis Accident (DBA) analysis. These analyses consist of postulated events, analyzed in the FinalSafetyAnalysisReport(F5AR)Ionalgoals.

for which a structure, system, or component must e.eet specified funct These an:1yses are entained in Chapters 6 and 15 of the FSAR (or equivalent chapters) and are identified as Condition II, III, er IV events (ANSI N 18.2) (or equivalent) that either assume the failure of or present a challenge to the integrity of a fission product barrier.

Process veriables are parameters for which specific values or ranges of values have been chosen as reference bounds in DBA analyses and which are monitored and controlled in actual operation such that process values remain within the analysis bounds.

The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA analyses, which are monitored and controlled. So long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low.

2-Implicit in this criterion is the associated installed control room instrumentation that monitors and/or controls the selected process variable. Two specific examples of process variables selected using Criterion 2 are:

Moveable Group Assembly Rod Insertion Limits Reactor Coolant System Pressure Limits Criterion 3: A structure, system, or cocoonent that is part of the primary success path of a safety sequence analysis and functions or actuates to mitigate a Design Basis Accident.

DISCUSSION: A ihird concept in the protection of the public health and safety is that in the event that a postulated DBA should occur, structures, systems, and components are available to function or to actuate in order to mitigate the consequence of the DBA. Safety sequence analyses or equivalent have been performed in recent years and provide a tethod of presenting the plant response to an accident.

A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant's DBA analysis, as presented in Chapters 6 and 15 of the plantN Final Safety Analysis Report. Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented. The primary success path of a safety sequence analysis consists of those actions assumed in the design basis accident analysis which limit the consequences of the events to within the appropriate acceptance criteria.

It is the intent of this criterion to capture into Technical Specifications only those structures, systems, components that are part of the primary success path of a safety sequence analysis. Implicit in this criterion are those support systems that are necessary for items in the primary success path to successfully function. The primary success path is equivalent for each DBA to the combinations and sequences of equipment assumed to operate when responding to the event which results in acceptable plant accident response (including consideration of the single failure criterion).

Two specific examples of structures, systems, and components which are selected using Criterion 3 are:

Reactor Trip System Instrumentation Primary System Safety Valves

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LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMMuffS 3.1 HEALIIV11Y CONIROL SYSILHS To:

Boration control (system-3.1.1 BORAT10N CONTROL

1. Ensure ability to reach see 3.1. 0 to maintain 3.1.1.1 Shutdown Margin subcriticality from all Yes (#2)

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3. Preclude inadvertent under ECCS only. Shut-Criticality in shut-downmarginisa" process down. Most restrictive variable controlled by condition is steamline the operator. This LCO break at no load Tavg.

appears to be redundant to 3.1.3.6, Rod Insertion limits and less comprehensive than 3.1.3.6 (except for Modes 3 and 4).

3.1.1.2 Shutdown Margin Tavg Same as aaove.

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HAS RX. PnWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION C0tMENTS 3.1.2.6 Borated Water Source -

Provides two sources of No Yes Operating borated water for boron addition during operation.

3.1.3 MOVEABLE CONTROL ASSEMBLIES 3.1.3.1 Group Height Ensure:

Yes (#2)

Yes

1. Proper power distribution.

2. Sufficient shutdown margin.

3. Correct rod-alignment as assumed in accident analysis.

3.1.3.2 Position Indication To ensure control rod Yes - as a Yes systems - Operating alignment and insertion surveillance (Digital and Demand) limits.

for 3.1.3.1.*

3.1.3.3 Position Indication Systcss - To monitor rod position Yes - as a No Action requires opening Shutdown (Digital only -

during shutdown surveillar.te of reactor trip for rods not fully for 3.1.3.1.*

breaker. Bases do nct inserted) address shutdown case.

3.1.3.4 Rod Drop Time To ensure rod insertion Yes - as a Yes While not monitored rate is consistent with survelliance and controlled during accident analysis for 3.1.3.*

operation, surveillance assumption.

is necessary to ensure re stor trip system can perfom safety function.

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LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMKENTS 3.1.3.5 Shutdown Rod Insertion lo ensure that minimum Yes - But Yes (via Bases do not address Limit (modes 1 and 2) shutdown margin is redundant to 3.1. 3.1 this specification maintained.

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Action specifically. Maybe Statement),

should be part of group height or shutdown margin.

3.1.3.6 Control Red Insertion To ensure:

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on this specification,

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3.2 POWER DISTRIBUTION LIMITS In reality, small deviations from these conditions could result in localized overheating and fuel damage in the event of a transient or accident; this probably would not present an immediate threat to the public safety. Also, most of these parameters are set by the core and fuel design not by operatIonalmethods.

However, these

" Discussion under Criterion #2 says that installed control room instrumentation that monitors and/or controls the selected process variable is " implicit in this criterion."

LCO REMAINS ACTION STMT.

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HAS RX. POWER LCO NO.-

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMHENTS conditions are an integral part of our defense-in-depth philosophy, 3.2.1 Axial. Flux Difference Ensures that axial flux Yes (#2)

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analyzed bounds for DNB.

this T.S. from BASES.

It does appear to be a 'vunding condition to ensure clad temperature and DNB criteria are not violated in DBA.

3.2.2 Heat Flux Hot Channel Ensures:

Yes (#2)

Yes Factor - F (z)

1. local power density 9

ettd minimum ONBR ar* not exceeded.

2. LOCA peak clad temperature of 2200'F not exceeded.

3.2.3 RCS Flow Rate and Nucle'ar Same as 3.2.2 Yes (#2)

Yes Enthalpy Rise Hot Channel Factor.

3.2.4 Quadrant Power Tilt Ratio Ensures power tilt in Yes (#2)

Yes X-Y plane within bounds for DNB analysis.

3.2.5 DNS Parameters

. Ensures that RCS pressure Yes (#2)

Yes and temperature are within initial bounds for DNB analysis.

3.3 INSTRUMENTATIOil

-G-LCO REMAINS ACTION STMT.

TECH. SPEC.

IIAS RX. POWER LCD HO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMENTS 3.3.1 Reactor Irip System 10 provide reactor trip Yes (#3)

Yes Instrumentation initiation when specific parametcr limits are reached.

3.3.2 Engineered Safety Features To provide actuation of Yes (#3)

Yes Actuation System those engineered safety

(

Instrumentation features whose function is necessary to mitigate postulated LOCAs, transients and accidents.

3. 3. 3.1_

Radiation Monitoring for Plant Operations:

1. Containment

a. Atmospheric-Gaseous Provides automatic iso b Yes (#3)

No Action Statement requires Radioactivity.

tion of containment pun closing purge valves.

b. Gaseous Radiaattivity -

Monitor RCS Leakage Yes (#1)

Yes Provides surveillance

c. Particulate Radioactivity Monitor RCS Leakage Yes (#1)

Yes requirements and some-what redundant to

2. Fuel Building 3.4.6.1, Leakage

a. Exhaust-Gaseous Automatic switchover to Yes (#3)

No betection Systems.

Radioactivity Emergency Ventilation

b. Criticality To monitor and alarm No No fuel pool radiation level.

3. Control Room Air Automatic switchover to No No Although control room Intake Emergency Ventilation.

function is important, the criteria do not 3.3.3.2 Movable Incore Detectors

.To calibrate excore No No appear to cover this detectors and obtain specification as a flux maps.

primary success path.

3. 5 3.3 Seismic Instrumentation To determine the magnitude No No of a seismic event and evaluate equipment response.

LCO REMAINS ACTION STMT.

TECH. SPEC.

IIAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION C0fHENTS 3.3.3.4 Meteorological lo obtain data for No No Instrumentation estimation of dose to public for routine or accidental releases.

3.3.3.5 Remote Shutdown To ensure ability to No Yes This instrumentation Instrumentation achieve and maintain HOT is not part of a 4

SHUTDOWN from outside the primary success path control room and to ensure for a DBA (criteria 3).

that a fire will nnt preclude Almost " of this achieving safe shutdown.

Instrumentailon is on the auxiliary shutdown panel.

3.1.3.6 Accident Monitoring To provide sufficient No (see coment) Yes Instruments that " key" InstruwAation information following an manual actions which accident. (Consistant are on primary success with Reg. Gufde 1.97 and paths for a DBA would HUREG-0737).

remain.

to ensure capability to No No 3.3.3.7 Chlorine Detection Systems detect and initiate actions in th* event of an accidential chlorine release.

3.3.3.8 Fire Detection To provide detection of No No Instrumentation fires and actuation of suppression systems.

3.3.3.9 Loose-Fart Detection To provide capability to No No System detect loose metallic parts in the reactor coolant system.

3.3.3.10 Radioactive liquid To w nitor and control, as No Ho Ef fluent Monitorir.g applicable, the release of Instrumentation radioactive material in liquid cffluents during actual or potential releases of Ilquid effluents.

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LCO REMAINS ACTION STMT.

TECH. SPEC.

HA5 RX. POK R LCO H0.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMMENTS 3.4.3

a. Pressurizer Backup Heaten 150 Kw Enhances natural No Yes Pressurizer heaters not circulation control of credited in DBA analysis.

RCS pressure

b. Water Level < 92%

Ensures bubble as assumed Yes (#2)

Yes Redundant to High in accident analysis Pressurizer level Reactor Trip at 92%.

}

3.4.4 Relief Valves (All PORV's Minimize opening of safety 7es Yes Although not stated in and Block valves) valves. (ser: comment)

BASES, PORV is used to (modes 1, 2, and 3) reduce pressure in a SG tube ruptt re event.

Therefore, spec stays.

3.4.5 Steam Generators The purpose of this Yes (as Yes - does This is the specification (modes 1, 2, 3 and 4) specification is to ensure surveillance not allow with all the tube the structural integrity of primary heat-up above survelliance of this part of the RCS.

success path 200'F requirements. The BASES component the do not address the decay steam generators) heat removal function of the steam generator. The detailed survelliance could be removed with 151.

3.4.6 Reactor Coolant System Leakage 3.4.5.1 Leakage Detection Systems To detect leakage t' rom the Yes (#1)

Yes

• Particulate Radicactivity reactor coolant prt!ssure

• Containment Sump tevel boundary.

(consistent with

• Containment Air Cooler Reg. Guide 1.45.)

• Condensate Flow Rate

• Gaseous Radioactivity

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRiiERIA)

LIMITATION COMMENTS 3.4.6.2 (berational teakage 10:

Yes (#1, #2)

Yes No RC pressure boundary provide early detection of leakage impending failures igpm unidentified Icakage control SG 1eakage in Igpa total RCS-SG 1eakage accordance with accident 10gpm. identified leakage analysis assumotions 8gpa controlled leakage prevent identified leakage i

per RCP from interfering with Igpa leakage of RCS pressure leakage detection systems isolation valves ensure adequate perfonnance of RCP seals prevent over pressurization of low pressure systems outside of containment 3.4.7 Chemistry To ensure that corrosion of No Yes the RCS is mmimized and reuoces the potential for RCS leakage or failure due to stress corrosion.

3.4.8 Specific Activity To ensure that the Yes (#2)

Yes (reactor coolant) resulting 2 hr. doses at site boundary don't exceed Part 100 for steam generator tube rupture accident.

3.4.9.1 Pressure / Temperature Provide limits in Yes (#2)

Yes The BASES for this Limits -

accordance with App. G to specification addresses Reactor Coolant System ensure vessel integrity.

the reactor vessel, primarily. No specific discussion is provided on the p7ssurizer limits. While the vessel is likely the most limiting component, the pressarizer LCO does

LCO REMATHS ACTION STMT.

TECH. SPEC.

HAS RX. PtNER LCO NO.

LCO TITLE PURPOSE (CRITERIA) llHITATION C0fEENTS

,:.resure RC5 integrity but does not appear to be covered by the criteria.

The App. G pressure and temperature limits, while not exactly initial conditions of a DBA, I

they are reference bounds monitored and controlled within established limits.

3.4.9.2 Pressure / Temperature No No Limits -

Pressurizer 3.4.9.3 Overpressure Protection Provide protection such No No Postulated overpressure System that pressu e will remain events during shutdown (modes 3, 4, 5, 6 within App. G limits and are not DBAs as defined RCS <3684 )

vessel integrity wl11 be by the criteria.

~

ensured.

Therefore this spec leaves.

3.4.10 Structural Integrl'y 151 and IST pregrams for Yes (#3)

Yes Requirements are required ASME Code Class 1, 2, and 3 by 10 CFR 50.55a(g) components ensure that the except whera relief has structural integrity and been granted. Rceains operatit,nal readiness will be as surveillance of maintained.

primary success path systems.

3.4.11 RCS Vents To exhaust non condensible No Yes Vents are act relied gases from RCS that could upon in any DBA success inhibit natural circulation

path, core cooling.

LCO REMAIHS ACTION STMT.

TECil. SPEC.

IIAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMMENTS 3.5 LNERGENLY CORE COOLING SYSTEMS 3.5.1 Accumulators Ensures operability of Yes (#3)

Yes some " realistic" analyses system which provides indicate that acc mulator initial cooling for core not necessary. App. K during large LOCA.

analysis requires them..I 3.5.2 ECCS Subsystems Ensures operability of two Yes (#3)

Yes

-Tavg >350'F subsystems to provide coeling for core if LOCA initiates abcve '350F.

3.5.3 ECCS Subsystems Ensures operability of one Yes (#3)

Yes

-Tavg <350'F subsystem to provide cooling for core if LOCA initiates below 350'F.

3.5.4 ECCS Subsystems F Myents RCS No No See coment on 3.1.2.3

-Tavg >200'F ovc v essurization during regarding RCS shutote by making SI pumps overpressure during inoperable.

shutdown.

3.5.5 Refueling Water Storage Ensures adequate supply of Yes(#2)

Yes Tank water for ECCS within DBA analysis envelope.

3.6 CONTAINMENT SYSTEMS 3.6.1 PRIMARY CONTAINMENT

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITER;A)

LlHITATION COMMENTS

3. b.1.1 Containment Integrity Maintain primary containment Yes (#2, #3)

Yes Containment integrity is integrity; restrict nicases really a DBA bounding to rates and paths in DBA

" condition," not a

analyses, process variable as referred to by Criteria

1. 2. Also, containment integrity is not really a system as in Criteria

1. 3, but the containment is.

3.6.1.2 Containment Leakage Ensures that containment Yes Yes can A leak rate could be leakage is within bounds of not start-up. thought of as a process safety onalysis.

variable, but it is not monitored and controlled 01 ring nomal operation.

Appendix J 1eakage testing is very wportant to assurance of containment integrity.

Maybe this testing should be 1) similar to or part of ISI/i3T program or 2) included as surveillance requirements and acceptance e.riteria under T.S. 3.6.1.1.

3.6.1.3 Containment Air locks Ensures that containment Yes (#2, #3);

Yes leakage through air locks This LCO is is within bounds of safety really SR and analysis.

acceptance criteria for T.5. 3.6.1.1

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO HO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMMENTS 3.6.1.4 Internal Pressure Ensures containment Yes (#2)

Yes Analyses show that structure integrity by:

accident consequences

1) limiting negative are quite insensitive differential pressure to iinitial pressure

2) limiting initial internal assumption. Containments pressure to meet DBA initial are generally well conditions (limiting DBA is overdesigned for DBA

}

steamline break) pressure.

3.6.1.5 Air Temperature Ensures initial temperature Yes (#2)

Yes Coment on 3.6.1.4 meets DBA (steamline break) generally applies to bounding initial condition.

temperature as well.

3.G.1.6 Containment Vessel Ensure containment Yes Yes These tests are performed Structural Integrity integrity every few years; this

" process variable" is not monitored and controlled during operation. Similar coment to 3.6.1.2.

But this is a surveillance for a primary success path -

the containment.

3.6.1.7 Containment Ventilation Ensure conteinment Yes (#2, #3)

Yes System integrity by 1) purge valves being closed or

2) mini purge valves being operable.

3.6.2 DEPRESSURIZATION AND COOLING SYSTEMS

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCD NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMENTS 3.6.2.1 Containment Spray System Ensure ability to Yes (#3)

Yes depressurize and cool containntent during DBA via two operable sprays.

3.6.2.2 Spray Additive System Ensures lodine removal Yes (#3)

Yes Recent source term term efficiency for sprays work and analyses seem I

assumed in DBA.

to show that additives are not necessary.

3.6.2.3 Containment Cooling System Ensures ability to cool Yes (#3)

Yes containment in LOP along with sprays.

3.6.3 Containment Isolation Valves Ensures containment Yes (#3)

Yes isolation; basis for DBA analyses 2.5.4 COM00STIBLE GAS CONTROL Generation of significant amounts combustible gas is beyond DBA.

Therefore, none of these LCOs meet the criteria.

3.6.4.1 Hydrogen Analyzers Monitor build-up of No Yes A close call. Chapter 6 -

hydrogen inside of FSAR implies they containment post-LOCA are necessary in Containment Design Basis.

But they are not a primary success path.

3.6.4.2 Hydrogen Mixing Systems Prevent localized No Yes Same as 3.6.4.1.

accumulations of hydrogen 3.7 PLANT SYSTEMS

17 -

LCO REMAINS ACTION STMT.

TECif. SPEC.

IIAS RX. POWER LCO H9.

LCO TITLE PURPO t (CRITERIA)

LIMITATION COMMENTS 3.7.1 lVHBINE CYCLE 3.7.1.1 Safety Valves To ensure that secendary Yes (#3)

Yes Bases do not mention pressure does not exceed that these valves are 110% of design.

also necessary for removal of decay heat and overpressure I

protection of the RCS.

3.7.1.2 Auxiliary Feedwater System To ensure ability to cool-Yes (#3)

Yes The Bases for this down RCS following a loss specification are of offsite power (per Wolf incomplete. This system Creek BASES).

is necessary to mitigate a loss of normal feed-water, small loss of coolant accidents and maintain a safe hot shutdown.

3.7.1.3 Condensate Storage Tank To provide sufficient AFW Yes (#3)

Yes water supply to maintain hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and then cooldown to RHR cut-in.

3.7.1.4 Specific Activity To ensure steam line Yes (#2)

Yes Initial assumption in rupture event doses are steam line break within Part 100 analysis. Probably not critical a parameter.

3.7.1.5' Main Steam t.ine Isolation To ensure that no more than Yes (#3)

Yes Valves one steam generator will blowdown in the event of a steam line rupture 3.7.2 Steam Generator Pressure /

To ensure pressure reduced No Yes (prevents Does not clearly meet Temperature limitation stresses do not exceed the heat-up above any DBA condition or minimum fracture toughness 200'F.)

process variable (#2).

stress limits.

Only a factor at shutdown.

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATI0H COMMENTS 3./.3 Component Cooling Water To provide cooling to Yes (#3)

Yes System certain safety related equipment - consistent with accident analysis.

3.7.4 Essential Service Water To provide cooling to Yes (#3)

Yes g

System certain safety related equipment - consistent with accident analysis.

3.7.5 Ultimate Heat Sink To provide the heat sink Yes (#3)

Yes and temperature to ensure sufficient cooling capacity to mitigate the effects of accidents.

3.7.6 Control Room Emergency Ensures that 1) the ambient Yes (#3)

Yes Although not Ventilation System air temperature does not specifically stated in exceed the allowable BASES, it is assumed the temperature for equipment equipment and instruments i

and instrumentation cooled in control room support by the system, 2) the or are a primary success control room will be habitable path for DBAs.

3.7.7 Emergency Exhaust Systems Ensures that radioactive Yes (#3)

Yes materials leaking frow the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the environment. (The operation of this system was assumed in the safety analysis.)

. LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION C0tHENTS 3.7.8 Snubbers To ensure that the No Yes (as part While snubbers do indeed structural integrity of the of operability support RCS and other RCS and all other safety definition safety systems, the related systems is maintained of equipment snubbers are essentially during and following a seismic they support) part of the piping or other event initiating design itself. Thatis,l dynamic loads, the snubbers are assume to perfom in a certain way in the dynamic analysis. They are not explicitly considered in Chapter 6 or 15, but are a structural / design consideration.

Therefore, they do not meet criteria 3 and would not be in the tech-specs. Since the snubber are assumed to be present F an initial condition, Criterion 2 could apply.

Also, the " leak before break" analysis minimizes the need for snubbers. Snubber surveillance would be handled just like other IST on RCS and other comonents.

3.7.9 Sealed Source Leakage Ensures that leakage from No No byproduct, source, and Special Nuclear Material sources will not exceed allowable intake values.

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION C0tHENTS 3./.10 Fire 5uppression Systems Ensure adequate fire No No lhese systems are not a suppression capability is part of DBA mitigation available to confine and as defined by the extinguish fires in areas of criteria.

safety-related equipment, 3.7.11 Fire Barriers Confines fires and retards No No (same as 3.7.10.)

spread to adjacent areas.

3.7.12 Area Temperature Monitoring Ensures safety-related Yes (#2)

Yes (though Temperatures in the area equipment will not be specific of vital equipment are subjected to temperatures equipment LCO variables monitored by in excess of their action state-the operator during environmental qualification ment) normal operation. While temperatures.

they may not be "prucess variables," they represent conditions of equipment in the primary success path for mitigating DBAs, as well e

as an initial condition of the DBA. This meets criteria #2. Exceeding these may not be an immediate safety problem due to time available to correct and the equipment not truly being inoperable.

From a practical stand-point, its had to imagine exceeding some of these specs.

3.8 ELECTRICAL POWER SYSTEMS To provide sufficient power for safe shutdown and mitigation and control accidents

. LCO REMA1HS ACTIO*1 STMT.

TECH. SPEC.

HAS RX. POWER LCD NO.

LCO TITLE PURPOSE (CRITERIA) llMITATION COMMENTS 3.8.1.1 A.C. Sources Operating (modes 1, 2, 3, 4)

Yes (#3)

Yes 3.8.1.2 A.C. Sources No No BASES do not discuss the Shutdown (modes 5 and 6) need for A.C. power when shutdown (although needed to shutdown). Mode 5 f (

isients not DBAs.

3.8.2.1 0.C. Sources Yes (#3)

Yes This LCO remains Operating (modes 1, 2, 3, 4) because it contains the diesal generators, a primary success path.

This LCO contain offsite power limits which are not primary success path, Eut are included because of the regulation and its importance.

3.8.2.2 D. C. Sources No No BASES do not discuss the Shutdown (modes 5andb) need for D.C. power when shutdown (although needed to shutdown). Modes 5, 5 events not. DBAs.

The._

may very well be a tie between DC (and AC) power to the boren dilution event, a mode 5 and 6 event.

If so, there specs would stay.

This comment also appiles to 3.8.1.2 and 3.8.3.2.

3.8.3.1 Onsite Power Distribution Yes (#3)

Yes Operating (modes 1, 2, 3, 4)

. LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMMENTS 3.8.3.2 Onsite Power Distribution No Ho same as 3.8.1.2, 3.8.2.2 Shutdown (modes 5 and 6) 3.8.4 Electrical Equipment To protect containment No Yes These devices do : vel Protective Devices electrical penetrations serve a direct sue ess path function. Should a z

penetration fall, the plant is then in an action statement for containment integrity.

3.9 REFUELING OPERATIONS 3.9.1 Boten Concentration itaintain subcriticality Yes (#2)

No Initial condition of boron dilution accident.

3.9.2 Instrumentatfor. - Source To detect ratlicactivity Yes (#3)

No Success path for boren Range Monitors changes in the core dilution accident.

3.9.3 Decay Time Ensures sufficient tiix: has Yes (#2)

No elapsed to allow decay of i

shortlived fission pro (ucts.

It in consistent with tue assuretions in the accident analysis.

3.9.4 Containment Building Isolate containment to Yes (#2, #3)

No Penetrations

dtigate fuel handling accl6?nt.

3.9.5 Communications (bef. ween To provide prompt No No control room and refueling comunication with station station).

personnel.

3.9.6 Refueling Machine To ensure proper and safe No No machine operation.

LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER LCO NO.

LCO TITLE PURPOSE (CRITERIA)

LIMITATION COMMENTS 3.9./

Crane Travel Ensures that in the event Yes (#3)

No Based on need to mix a fuel assembly is dropped, boron in boron dilution the activity will be limited

accident, to that contained in a single assembly and to prevent fuel heat removal and boron mixing.

(

3.9,0 Residual Heat Removal and To ensure heat removal and Yes(#3)

No Based on need to mix Coolant circulation boron mixing.

boron in boron dilution accident.

3.9.9 Containment Ventilation Ensure automatic isolation Yes No Redundant to 3.9.4 and System of purge system.

should be combined.

3.9.10/11 Water Level - Reactor Vessel To filter radioactivity Yes (#2)

No Also important as and Storage Pool following ruptured fuel biological shield and asseebly event.

for fuel cooling. This is not addressed in the BASES.

3.9.12 SpentFuelAssembly5}orage To p nent inadvertent Yes (#2)

No Assumed to be a DBA criticality.

event.

3.9.13 Emergency Exhaust System To filter releases from Yes (#3)

No the fuel handling accident.

3.10 SPECIAL TEST EXCEPTIONS 3.10.1, Shutdewn Margin To allow radioactivity Yes (#2)

No (but requires measurements.

boration) 3.10.2 Group Ileight Insertion To allow physics testing.

Yes (#2)

Yes and Power Olstribution 3.10.3 Physics Tests To allow physics testing.

Yes (#2)

Yes 3.10.4 neactor Coolant loops To allow startup tests Yes (#2)

Yes LCO REMAlHS ACTION STMT.

TECll. SPEC.

IIAS RX. POWER LCO HO.

LCO TITLE PURPOSE (CRITERIA)

L1HITATION C0tHENTS 3.10.5 Position indication system To allow rod drop tests Yes (as Yes Note: All exceptions to surveillance Teclinical Specification see 3.1.3.3) for special testing should be in Technical Specifications specifically.

z 3.11 RADIDACTIVE EFFLUENTS 3.11.2.1, Liquid Effluents To control releases and Ho Ho 3.11.1l.2 associated doses 3.11.1.3 Liquid Rad waste Treatment To ensure treatent systen Ho No Systm availability.

3.11.1.4 Liquid floidup Tanks Limit release to amount Yes (#2)

Ho (quantity) assumed in accident analysis 3.11.2.1, Gaseous Effluents To Control releases and Ho No 3.11.2.2, associated doses 3.1?.2.3 i

3.11.2.4 Gascous Rad waste Treatment To ensure treatment system No No System availability 3.11.2.5 Explosive Gas Mixture To prevent an explosion Ho No 3.11.2.6 Cas Storage Tanks limit release to amount Yes (#2)

No (quantity) asstmied in accident analysis 3.11.3 Solid Rad Waste To provide good quality No No solid waste (50.36a) 3.11.4 Total Dose To limit total doses to No No pubile from fuel cycle

. LCO REMAINS ACTION STMT.

TECH. SPEC.

HAS RX. POWER it0 NO.

LCO TITtt PURPOSE

{t.RITERIA) llMITATION COMMENTS 3.12 RAD 10LC61 CAL ENVidUNHLHIAL HONITORING 3.12.1 Honitoring Progrem To monitor exposure No No pathways 3.12.2 d Use Census To rbtain Information No No t

nec~ssary to update ODCH for exp0sure pathways 3.12.3 Interlab Comparison To obtain independent No Ho checks on measurements I

ENCLOSURE 3 LCO " COUNT"*

WOLF CREEK Total Number 129 Power Limitation 73 No Power Limitation 55 Initial Split:

Total Jn:

79(51%)

Total Out 50 (39%)

Power Limitation 59 Power Limitation 14 No Power. imitation 20 No Power Limitation 35 LIMERICK TOTAL NUMBER 138 Power Limit; tion No Power Limitation Initial Split:

Total Jn:

Total Out Power Limitation Power Limitation No Power Limitation No Power Limitation

• Counts are approximate - LCO's can be grouped many ways-some included detailed tables and surveillance requirements.

ENCLOSURE 4 INSIGHT AND COMMENTS (Potpourri)

Hermal Boration comes out (secondary success path and path to cold shutdown).

-W Review made without benefit of primary success path analysis.

-G Criterion 2 includes surveillance systems of process variables.

-G Some LCO's stay because they are surveillance of primary success path.

-G Power distr!bution:

complex and intertwined -G not risk significant - but stays Some BASES very poor:

AFW, PORVs, Turbine Overspeed Protection -G Criteria don't address requirements redundant to regulations such as ISI/IST.

-G Definition for imediate threat, on. subjective criteria not yet docteented.

-G 0737 item, RCS vents, Pressurizer Heaters leave. - W, G Criteria (and discussion) don't reference staff SER.

Criteria do not discuss (or include) instrumentation which:

Triggers action in emergency procedures (non prie.ary path)

Confires operation of primary success path system:.

-G Criteria do no* discuss special exceptions to tech-specs (3.10)'.

-G W = Wolf Creek related corcent.

L = Limerick related coment.

G = General Comment.

ENCLOSURE 5 Limitino Cenditions for Operatien with Action Statements that Limit Reatter Power Based on the enclosed split, 14 LCOs are in this category for Wolf Creek.

LCOs are in this category for Limerick. This number is viewed to be quite small and their disposition or arguments concerning disposition, will hopefully not detract from the overall Technical Specifiestion improves.ent program objectives. Retomended dispomion is assigned in the Enclosure 5.

Some judgements are i sry close calls. Yvo points should be made. First, these items as a whole appear to have more safety significance than most (.COs removed.

Second, a reasonable argument to keep these requirements ir Technical Specifications, so that all such requirements are in one place, can be made.

This argument should not yet be discounted. Fe11owing is a list of the LCOs for each set of Technical Specifications.

Information is provided for each LCO that should be useful in making a final determination.

Wolf Creek 3.1.2.2/3.1.2.4/3.1.2.6 Reactivity Control /Boration Control 3.3.3.5 Ramcte Shutdown Instrumentatien L3. 3.6 Acciden. Monitoring Instrumentation 3.4.3 Pressurizer Heaters 3.4.7 Chemistry 3.4 4.2 Pressurizer Pressure-Temparature Limits 3.4.11 RCS Vents 3.6.4.1 Hydrogen Analyzer 3.6.4.2 Hydrogen Mixing 3.7.2 Steam Generator Pressure-Temperature Limits 3.7.8 Snubbers 3.8.4 Electrical Equipment Protective Devices

. 3.1. 2. 2/3.1. 2. 4 /3.1. 2. 6 - Boration Control - With RCS temperature equal to or greater than 350'F, a minimum of 2 boron injection flow paths are required to meet the single failure criterion. The technical basis for the requirement is to provide a shutdown cargin of 1.3% a t/k after xenon decay and cool down from operating conditions. Criterion 3, although it is not stated, only progresses to hot shutdown - since Chapter 15 si. ops at het shutdown also. Foration is necessary to proceed to cold shutdown. It is judged to be of low risk significance, but is clearly more important than many other LCOs be'ng removed.

3.3.3.5 Remote Shutdown Instrumentation - This requirement is to satisfy

~

GDC 19 on shutdown from outside the control room and Appendix R safe shutdown. Remote shutdown is not a primary success path. However, manual actions, based on plant conditions indicated to the operator, have Men found to be important in at least one PRA fire analysis.

3.3 3.6 Accident Monitorina Instrumentation - This requirement is consistent with revision 2 of Reg. Guide 1.97. Reactor trip system and Engineered Safety features Actuation System instrumer.tation appears to provide many of the same key parameters (although the range may be more limited). In ad"dition, RTS anc' ESTAS requirements apparently do not cover the control room indication which is the focus of this requirement.

3.4.3 Pressurizer Heaters - Pressurizer heaters allow pressure control to be maintained by the pressurizer and enhances the capability to esthblish natural circulation. They are not credited in any safety analysis. Iney are nomally not modeled in a PRA since core cooling can be maintained vithout them.

3.4.7 Chemistry - These requirements are intended to ensure that corrosion of the RCS is ne.3imized and to reduce the potential for leakage or failure due to stress co resion. Based on discussions with Chemical Engineering Branch personnel, the " leak before break" studies consider poor chemistry control.

- Further, industry programs have been initiated that control reactor water chemistry that are more conservative than the Technical Specifications.

Some utilities however, are apparently ng involved in these Owners Group programs.

3.4.9.2 Pressurizer Pressure - Temperature _ Limits - The Technical Specification BASES do not discuss these limits.

It is judged that these limits are not of imediate importance since the reactor vessel is the most limiting component and the vessel pressure temperature limits remain in Technical Specifications. The differential temoerature limit on the spray nozzle is noted to be very large, and practically speaking, very difficult to violate (maximum spray water temperature differential of 583*F).

3.4.H Reactor Coolant System Vents - The vents are provided to exhaust non-condensibles that could inhibit natural circulation core cooling. The vent requirement stems directly from Three Mile Island accident and ensures the capability to perform the venting funt' ion.

3.6.4 Combustible Gas Control (Hydrooen Analyzer and Hydrocen Mixing) - These s'vstems and the basis for the requirements are discussed in Chapter 6 of the FiAR. From the standpoint of the transient and accident analysis, and Criterion 3, tnese requirements would leave the Technical Specifications.

Reading of Chapter 6 sakes this a close call in that a small fraction (5%)

of the core is assumed to react to produce hydrogen and serves as the design bases for these requirements. This culd be interpreted lo mean including the requirements under Criterion L 3.7.2 Steam Generator Pressure - T tperature Limits - While steam generator i

press;;te and temperature cre process variables, the basis for these lititt is the brittle fracture concern of the steam generator (RT-h"J3T = 60*F). The limit is 70'F h2 200 psi which does not tppear to come into Disy at operating temperature and pressure or when using the steam generator to

-4 remove decay heat. The survellisnce only applies when water temperature is less than 70*F. This specification was not judged to meet any of the criteria.

3.7.8 Snubbers - While the snubbers are important to the dynamir response of the piping, components and systems they support, snubbers are judged as not seeting Criterion 3.

They are viewed as part of the piping design, which is not (and need not be) included in Techaical Specifications in its entirety.

In cperable snubbers would still nquire an engineering analysis and determination as to the " operability" of the system they support. This issue relates directly to any future work done on the operability definitiot:.

~

3.8.4 Electrical Equipment Protective Devices - These devices are required to protect containmer.t electrical penetrations and penetration conductors.

They do not meet any of the criteria. (The list of devices is the largest single LCO (by page volute) in this set of Technical Specifications). The primary success path of concern here appears to be contairrent integrity, hovercr. these requirements are not judged to be a direct support system for this success path.

EN CL o.TVR E 4

WOG TECH SPEC SUBCOMMITTEE MEETING JANUARY 21-23, 1986 WOLF CREEK TECH SPECS TEST AGAINST CRITERIA D

WDG YECN srtt sUBCornifftt MttflWG 1/21/85 1/23/06 08JECTIVE

• T0 SPLIT WOLF CWitt itCM srtCs SWtti 1 OF j WlmUts fiftt

• • 0st N fts? AGAlWsf CRiftRIA ACTION COMMENis routR RED.

Calf 1 Calf 2 CRIf3 3.1.1.1*

s WT00W4 MARGIW, f>200 DECF, M-1,2,3,4 EW5URE sWTV0W4 MARCIN >CR* 1.3%

WO fis NO WO sit CtFmENT #1 3.1.1.2*

sistJfDOWN MARCIN, it200 DECF, M3 tusuRE snufo0W4 MARcis >oR* 12 no Yts p0 no stt cursstri 82 3.1.1.3 MEDttATM TertRAftRE COEFFICitWT L.L. < MTC

• U.L.

NO NO NO Yts 3.1.1.

• MINIPRM ft?W'ttATURE CMFFICIENT T Avt > 331 DtcF 30 fts no Yts 3.1.2.1*

90RAfttet Ft0WrtfRS, M 4,5,6 FLOWPAT4 OrtRAtlLITY WO No Yts No stt CtF1 MENT F2 3.1.2.2*

90 RATIOS FLOWPAtws, M 1,2,3 FlesA-Ain OrtRAsttlff s0 No YEs Yts stt Ctmntuf #2 3.1.2.3*

CnAreclug FvMrs. sWic0Wu, n 4,5,6 tvMP ortRAstliff u0 no Yts w0 stt CtyntwTs #2,20 3.1.2.4*

CWARCiwe PtMrs M 1,2,3 FUMP OrtwAstLITY m0 no Yts Yts stt timMtufs #2, 3 3.1.2.3*

30RaftD vnftR sotmCt, M 3,6 po wo Yts s0 stt Coreituf F38 3.1.2.6*

90 RATED m itt, M 1,2,3,4 BORAftD WAftt AVAILAntLiff NO NO Yts YEs stt CtF1 MENT 83 3.1.3.1*

cP sf.

vtalFY DBA AsstF'rfl04 WO Yts Yts Yts stt COM'T4? Ei 3.1.3.2 RPI (PRFI vs DtMAuC)

Manif 0Rs FOR RPI LO wo No its s

(I) statement of purpose not included for all specs

S e

WG ltCW SPEC SURCDMITTEE M[tflWG 1/21/M - 1/23/M C8JttilYE TO SPLIT WLF Crttt itCN SPECS Sutt? 2 or

't NUMStt flTLE FUtPOSE TEST AGA!NST CalttalA (1) Actlas CUntuts PtM a atD.

CRIf1 CRif2 CRIT 3 3.1.3.3 Ret RPI MORift*5 no no no no 3.1.3.4 atn Detr flME vtalFT DsA ASSU1PTIONS NO 90 WO NO 3.1.3.3*

Stt;1toal RCD INStarian Limit vttIFY DSA ASSUMrflDt3 so TES No so 3.1.3.6*

CDITRot Rtc INSttTION LIMITS VttlFT 09A ASStmPfl0NS 50 ftS NO Wo 3.2.1*

ARIAL FLLT SIFF.

M RIFY DBA ASSUMPittWS WO TES WO TES 3.2.2 F-0 YttlFY 084 ASSUMPfl0NS 1p0 WO NO TES 3.2.3 F DELTA N AMD FLtd Mt]FT D9A ASSUMPil WS 50 NO IF)

TES SEE CUTIENT F$

3.2.4*

OLAD F0Wtt flLT VttIFY 08A ASSUMPfl0NS 90 TES

1. 0 ftS 3.2.3*

Ds3 VttlFT DBA ASStmPfl0NS NO TIS NO Yts 3.3.1*

af INSta, vtalFY OtA ASSUMPfl0NS 90 WO TES Yt3 Stt comtWT #6 3.3.2*

ES7 INSTS.

VttlFY DBA ASStFPflDt3 80 WO TES TIS Stt CU9' TWT #6 3.3.3.1*

RAD. MDif f03 INST'd.

I%ftilTIAL DBA NtQM50R YES 20 TES NG Stt COPrtWT U 9

e Woo TECW SPEC SUBCorfttfitt MttitwC 1/21/M 1/23/M DBJECTIVE - To SFtti b10tF CRtti itCM SPECS SNEff 10F M Ruistt 11TLE PURPost TEST ACAtwSt CatitRIA (1) ACTIC1 COMMtWTS POWER M P.

CRIf1 DtI2 Calf 3 3.3.3.2 MIDS 19R DIST SURY.

No No No no 3.3.3.3 SElmic twSTS.

SttSMIC Stev.

no no no no 3.3.3.4 ptitonotectCAL twSTS.

MtittactoctCAL SURY.

so no no no I

3.3.3.3 Rtnoit Snuteous Im3TS.

Alf. T-wRot etxm no yo No its 3.3.3.6*

ACCIPEWT MONITDtlWG twSTS.

PAM - TMt titlUIMENTS No 50 Ytt Y'S SC CtrestWYS M, 8 3.3.3.7 Cut 0Rtwt etittitos Ccuinot RoaPt Ainos. SURY.

so no no wo 3.3.3.8 Fits ett. twSTS.

.o no no wo 3.3.3.9 teost PARTS ctf. SYSt.

No no no no 3.3.3.10' #AD. Lle. EFFL. IwSTS.

084 ASSUPPTION Wo No Yt3 No Stt CDP 98twiS #7, to 3.3.3.11' RAD CA9tCUS EFFt.193fS.

DSA ASSUMPitCE Wo No YES No Sit Ctr1MENTS #7,10 3.3.4 fles. fTetSPtt0 TURStwt PtoTECTION Wo No No TES 3.4.1.1*

• aCS LOCPS, M 1,2 D8A ASSUMPTION No Yt3 No TES Stt Ct m EFT #11 e

• s WOG YttN srtC su COMMITitt MttTING 1/21/M ** 1/23/M 08JtcilVE

• TO srLif WOLF CRttE TECW srtCs SMtti i 0F f NUMBER flitt FtRf0?E fts? AGAltst CRlittlA (1) A.i A ComtWis E 1 RED.

CtIT1 Cttf2 CRIf3 3.4.1.2*

RCs LOCrs, M-]

DEA As30MPflows to Gs 40 No sit comtu' #12 3.4.1.3*

Rt3 Leurs, M 4 DsA Asst *1Pflots WO Yts

1. 0 WO stt C W NT f13 3.4.1.4.1* tts Locrs, M 3 esA AssuMPTIORs NO Yts NO no sit Co m TWT #13 3.4.1.4.2* RCs Ltres M 3, Locrs NOT FULL csA Ass 0MPflows 90 Yts so no stt CmMtWT #13 3.4.2.1 FRZt CEDE SAftff VLv., M 4,3 50 WO no no 3.4.2.2*

Ptzt a m sAFtTY vtv., M 1,2,3 esa Assunrflows m0 no YES Trs 3.4.3*

F1!ZR WO Yts Yts Yts 3.4.4*

PC2V's 50 WO Yts Yts stF C W WT #1&

3.4.3 sitAM CCuttAfots 90 WO No stt C N WT 513 3.4.6.1*

RCs LEAK ptitCTots Yts no no TES 3.4.6.2*

RCS LEAUGE LIMITS WO Yts NO Yts sit COMMtui #16 3.4.T RCS CptwisitY WO wo no YEs e

-s, s

WOG TECW SftC SusC09tlTTtt PittflWC 1/21/86 *a 1/23/86 08JtCTIVE TO SPLif WLF CRttt TECM $PtCS SMitY [ OF f IlUMUT1 TITLE PURf05E TEST AGAIWST CRITERIA {1) ACTION ComtuTS POWCR RtD.

CalT1 CRIT 2 CalT3 3.4.8*

RCS SPtc. ACTIVITY

1. 0 Tts NO Yts stt ComtwT FIT 3.4.9.1 P T LIMITS NO No no its 3.4.9.2 FRZR NEAftp AND COOLDGA NO No NO TES 3.4.9.3 COLD Ovtt. ?tts. F107.

WO NO NO 5tt ComtWY #18 3.4.10 RCs $TRUCT. Itits.

No 50 NO wo 3.4.11 ECS vtwit WO wo u0 sit Co mtut f19 3.3.1*

ECCS ACOMULATORS 50 ft1 WO TES 3.3.2' ECCS SUBSTSitW3 NO NO TES Yt3 3.3.3 '

ttCS SUBS'sitws, n 4 u0 no TES 3.3.,

Il PUNPS fuor, T @ etcr so w0 po stt ComEwY #20 3.3.3*

WWST I!O Yts Yts 3.6.1.1*

CatY. IWit9.

NO 90 YES TIS Stt ComEWT #21 e

a w

e P

WOG it*W SftC ELOCCrmtfftt ritilNS 1/21/86. 1/23/86 08JECTIVE TO SPLIT WOLF CRttK TECM SPECS SMtti A 0F

UUPetR TITLE Futrost TEST AGalNST CRITtt1A (1) ACTION Cu mENTS POWtt RED.

CRIT 1 CRIT 2 CRIT 3 3.6.1.2 CONTAIMMENT LEAKAGE

1. 0 WO m0 Sit CEPPtWT #22 3.6.1.3*

CCMTAINNENT Alt Letti WO NO TES TIS Stt CD etWT F23 3.6.1.4*

CD4TAIMMENT FWis!Utt Wo its

1. 0 fts 3.6.1.3*

CONTAtuntui itwttaitet NO Yts NO Yts 3.6.1.6*

CcsTAluntwT siitUCT. INTtc.

NO

1. 0 YES stt CIFTitNT #24 3.6.1.T*

Ftattt AnD EYW. Isot. YLYS.

80 uo YES 3.6.2.1*

CMTAlWNEWT $rRAY WO NO Yt3 3.6.2.2*

SrtAT ADO. SYST.

NO Yts Yts 3.6.2.3*

CDei. FAs toottas no no Yts 3.6.3*

CCNT. IscL. vtVI.

m0 no Yts stt COMMtut F23 3,6.4.1 N 2 ANALYZERS un no NO 3.6.4.2 N 2 C04tRUL SYTT.

so no u0 0

M TECW trtC SU8CO Mlfitt PEETIWC 1/21/M 1/23/M OBJECTIVE TO Stiff WDLF CRitt itCW SrtCS SPEET 2 0F j Nisutta fiftt FtRPOSE ftSt A mtes? CRITERIA (1) ACTION QPMENTS fTutt RID.

CRift CRif2 CRIT 3 3.7.1.1*

MRtlWE SartTT YtVS.

No NO TIS YES 3.7.9.t*

R7K. DEEDtmTER WO 18 0 YES TES 3.T.1.3 COMO. STOR. TANK r0

1. 0 WO Yts Stt CDettut F26 3.T.1.4*

SEC. COOLANT trtC. Act.

No its no Yts 3.T.1.5*

RISV OrttAtti 80 No Yts Tis 3.7.2 S.C. P f LIMlf3

1. 0 NO WO no 3.T 3*

CCW STsitM wo MO 7ts Yts 3.T.4*

tsu STsitM

1. 0 no Yts Yts 3.T.5*

ULT. ptAT SINE WO No Yt3 YES Stt CarmtNT #27 3.T.6*

CONT. RM. t* U S. W WT SYST.

No No YES Ytt

$tt CarettWTS F28, 27 3.T.T*

EMfG. EXE881ST SYST.

No B0 7t3 YES Stt CDettNis, #28, 29 3.T.B*

S3H.99tRS wo no Ytt WO 5tt CD*Tuf f30 e

900 ftCE SPfC IU8CDTttfitt #ttflNC 1/21/86 ** 1/23/86 OBJECTIVE TO SPLIT WOLF CRttt itCM SrtCs d

SWEET T OF -

RNBit flitt itST AGAINST CalittIA (1) ACfl04 COPPENTS POWtt StD.

CalT1 CRIf2 CRIT 3 3.T.9 SEALED 50LRCE trWTAftlNATIce NO WO No FItt RstttsSION ff3T.* f 1.1.9e.1'\\

3. /.10 1

NO WO NO Stt ComtNT K39 i

11 3.7.11 Fitt BAttita rtutts.

wo no me 9

j t

L 3.T.12 AttA ft?W'. McNITORS

1. 3 NO NO SE2 CoretNT #31 3.8.1.1*

A.C. SOURCES NO NO TES stt ComtNT #32 3.8.1.2*

A.C. sttmCts, M 3,6 20 No Yts 3.8.2.1*

D.C. 50Uttts 80 No Yts stt ComtWT K32 3.8.2.2*

D.C. SOURCES, M 3.6 mo No Yts 3.8.3.1*

Deslit rVa. DIST.

No No TES 3.8.3.2*

UtstTE PWR Ditt, M 3,6 No WO TEs 3.8.4.1 CONT. PtWET. OvTRCL*FtWT PtoitCTORS

fo No WO TIS Stt CarmtWT K33 3.9.1 RCS 90t0I CDRC., M 6 MO WO No

$tt ComtWT K34 e

s WDG TECW SMC 9ACEM9ffitt MttflNs 1/21/86 1/Z3/86 08JttflVE - TO 5511T WOLF CREtt itCN srtCs SWtti 9 0F H 81UNtt IIILE ftsi AGAltsi CR!itRI# (1) ACTION CaretWis rowtt Rto.

Ctifi CWlf2 CRif3 3.9.2 s.a. Iwsts, M 6 no no no 3.9.i' DECAT Titit, M 6 wo Yts no 3.9.45 COWT. PtWris., M 6 WO NO Yts 3.9.5 CDMUNICATIcts, M-6 N0 No No 3.9.6 attuttles MACNInt no ao so 3.9.P CRANE TRAVYt, M 6 Wo Yts no 3.7.8.1 est tocrs, M 6 AND L*Z3 FT 30 no y0 stt CuPrtWT F33 3.9.8.2 Rtut LCors, M-6 AMD L<23 FT so no 80 Stt COPritsf F33 3.9.9*

towf. vtwis, M 6 uo ao Yts 3.9.10.1*

R.V. Mitt LEYtt, M 6 20 Tis NO Stt ComtWT FM 3.9.10.Z* a.v. m itt Ltvit, M-6 30 ygg NO Stt COMMtWT FM 3.9.11*

st0RAct POOL w.L.

20 Yts NO Stt COMMtut f%

e

WOG ftCN SFTC SURCDTtlTitt tittTING 1/21/86 1/23/86 CeJECTIVE To SPtif WOLF Cattt itCN SrtCS SMtET d of i IlUM8tt flTLE FUtfest Ttti ACAlwST CRiftRIA (1) ACTION C N Nf3 F9WER RED.

CRIT 1 CRIf2 CRIT 3 3.*

trtWf FUEL STORAct 90 No No Sit C mMtsi F34 I'#*I *

• 3.9.13*

ENtts. Ext. SYST.

so Wo its Stt Corestui #29 eI

• • bO 3.10 SrtC. 7107 EXttPTS.<

• I

• 89 89

1. 0.

4t>

Stt Co mtNT F37 ey - En l

3.11.1.1 RAD. Llo. EFFLS.

L

.g

.,4 )

No Wo no 3.11.1.2 Dost no no go 3.11.1.3 Llo. Rac m tit TREATMtui so no no 3.11.1.4* RAD. LI0. HottMr iAmts so it5 wo 3.11.2.1 GASE013 EFTL. Dost R.

Wo No No 3.11.2.2 Westt tiAs Dost wo no no 3.11.2.3 1 131,133, M 3 Dost so no no 3.11.2.4 CAS RAD m tit TttAT18CNT uo No go 3.11.2.5 0 2 conc. In tatSit CAS TANK Wo No No e

WOG ftCW SrtC SU8 COMMITTEE MEfflWC 1/21/M - 1/23/M OBJECTIVT TO S?tif M.F CREtt itCW SrtCS SWtti _" Of _n RPett flitt l'tJRrost itST AGAIRST CRiftRIA (1) ACfl0W COMMENTS POWER RED.

CRift Calf 2 CRIT 3 3.11.2.6* Re. cAs st0RAct TAsts No Ytt to 3.11.3 SM9 RAO. WASTt NO WO N0 t

W3 WO NO 3.11.4 TUTAL DOSE,

3.12.1 R 2. tWV. 90tif0R1 WO N0 WO 3.12.2 LA e Ust ttW ws 80 80 No 9

1

COMMDTIS ON WDLF CREEX TECH SPEC SPLIT WOG WING JA!TJARY 21-23, 1986 A total of ';::I LCO's meet criteria A total of 9$ LCO's fail criteria E Total LCO's Evaluated m

LCO Hects Criter',a NO.

COMMO."I 1.

In todes 1 and 2, 31utdc a Margin (SDM) is not a "pmcess variable".

In modes 3 and 4 shutdown margin satisfies criteria #2 because SDM can be controlled via boron ooncentration.

2.

Criteria satisfied, based on info ution on Baron Dilution DBA htich requires boron injection for accident citigation - A plant specific DBA requirenent.

3 Redundant to ECCS Tech Spec.

4.

Criteria #3 satisfied because operable implies Trippable.

5 Ihe ROS now should be included with the Dl3 Tech Spec (3/4.2 5).

6.

The tech spec should only f nelude insts. asstned in the safety Analysis, for exa=ple, Int. Range I.evel Reactor Trip abould be deleted.

7 New Fuel Pool Radiation Honitcr not asstzned in any DBA analysis,

- therefore ocit fn:n Tech Spec.

8.

Instrtraentation needed to go from an accident condition to a Safe Shutdown condit. ion would satisfy criteria #3 and should be retained in the specs.

Any redundant insts. or insts required for-ER3 reasons should be renoved f rom the spe based on EGR's not being in the PSAR.

9 Criteria #3 satisfied, per FSAR Chapter 15 state =ent that S.G.

bMwdown automatically isolates on Hi Radiation Alaru. Other insts.

e assumed in Safety Analysis should be removed frcn the Spec.

10.

Insts. assu::ed in Safety balysis in TS 3 3 310 and 3 3 311 should be relocated to T.S. 3 3 31, then these two specs could be de.leted.

11.

Since the LCO surveillance is on flow and the D!G Tech Spee includes a flow LCO and Surveillance this Tech Spec could be deleted.

12.

Criteria #2 is satisfied due to the M-3 reactivity addition accident.

13 Per FSAR, loop operation is astned to assure complete riring in R.V.

14.

PORV in Safety Analysis is manually actuated, therefore Channel Calibration Surveillance should be deleted. Also, the PORV used to citigate a tube rupture accident is a plant specific analysis asseption.

15 The spee is used to include S.G. IS1 in the tech specs. If S.G.

operability is covered by another spec (3 4.1.1), for exa=ple, the su veillance should be removed to the ISI program and the LCO deleted.

16.

Criteria #2 is satisfied in ths.t leakage is Controllable by closing isolation valves, reducing pressure, etc.

17 Criteria #2 is satisfied by oparator via feed / bleed, Ph'R reduction, etc.

18.

Per FSAR, the DBA criteria is not met (i.e., COMS not described in Chapters 6 or 15).

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19 RCS vent not assumed in FSAR.

20.

Ibt required, based on C0KS not included in DBA. Also, restrictions on charging pu:ps based on COMS in other Tech Specs can be deleted.

-21.

LCO meets criteria 3, but the Surveillance references other LCO's and appears to be redaadant.

22.

LCO fails criteria #2 because leakage is not controllable by the operator.

23 Only Part a. of LCO satisfies criteria #3 Part b. - covering leakage fails criteria and should be deleted.

24.

Criteria #3 satisfied but the LCO only refers to surveillances. The Surveillances should be removed frcen the spec and a surveillance program referenced.

25 This LCO should be cambined with TS 3.6.1.1 and the valve list recnoved to another controlled doctraent.

26.

Wolf Creek PSAR asstnes ESW as source of Aux Feedwater.

27 A plant specific TS. Plant specific analysis may be able to show the UHS te=perature or level requirements are not necessary.

28.

The LCO satisfies criteria #3 for a DBA but the Chapter 15 Asseptions required by the SEP are excessively conservative. An analysis using consistent asseptions from other DBA calculations may show that the LCO is not needed for DBA citigation.

29 The surveillance requirements for these tech specs should be removed to another controlled docuenent.

30.

120 satisfies criteria f3 but surveillance should be re:noved to ISI prograu.

31.

LCO is Eq. Qual. Basis Only, not DBA, therefore no Tech Spec criteria tre satisfied.

32.

Surveillance should be reoved frota Tech Specs.

33 Cmponent operation not assaned by any DBA Asseptions, therefore LCO and Surveillance can be re:noved frm Tech Specs.

34.

No DBA Ass mption involved in LCO.

35 No Baron Dilution DBA in M:x$e 6, therefore no Tech Spee criteria apply.

36.

DBA Assmption in LCO.

37 Special Test Dcceptions LCO should be retained unless the LCO in the exception has been deleted.

38.

Criteria #3 only applies to L'ST portion of LCO and then only ten in mode 5 39 4 LCo's contained in Subsections of 3 7.10, all LCO's fail the criteria test.

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