ML20205P809

Excess Flow Check Valve Testing Relaxation

ML20205P809
Person / Time
Site:	Duane Arnold
Issue date:	11/30/1998
From:	Eftimie C, Fairfield R, Knecht P GENERAL ELECTRIC CO.
To:
Shared Package
ML20205P800	List: ML20205P803 ML20205P809 NG-99-0308, Application for Amend to License DPR-49,relaxing SR Frequency by Allowing Representative Sample of Excess Flow Check Valves to Be Tested Every 24 Months,Such That Each Will Be Tested at Least Once Every 10 Yrs
References
B21-00658-01, B21-658-1, NUDOCS 9904200380
Download: ML20205P809 (37)

Text

I GE Nuclear Diergy 1121-00658-01 Class 1 November 1998 Excess Flow Check Valve Testing Relaxation Prepared for:

Excess Flow Check Valve Committee llWR Owners' Group Prepared by: PD Knecht CC I!flimie RM Fairfield Project Manager:

oriuinal siened by Rick Ilill R.A. Ilill GI! Nuclear linergy San Jose. Califbrnia November 16,1998 9904200380 990412 PDR ADOCK 05000331 P

PDR

e (iE Xuelear Energy l.\\lPORTANT NOTICE REG ARDING CONTENTS OF Tills REPORT Please Read Carefully The only undertakings of General Electric Company (GE) respecting information in this document are contained in the contract between the Boiling Water Reactors owners' Group (BWROG) and GE. as identified in the respective utilities' BWROG Standing Purchase Order tiir the performance of the work described herein, and nothing in this document shall be construed as changing those individual contracts. The use of this inti)rmation, except as defined by said contracts, or ti>r any purpose other than that tiir w hich it is intended. is not authorized. and with respect to any other unauthorized use, neither GE. nor any of the contributors to this document makes any representation or warranty, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document.

PARTICIPATING liTILITIES:

lloston Edison Company Carolina Power & Light Commonwealth Edison (Comed)

Detroit Edison Company GPU Nuclear IES Utilities. Inc.

Illinois Power Company New York Power Authority Niagara NIohawk Power Corporation Northeast Utilities Northern States Power PECO Energy Pennsylvania Power & Light Public Service Electric & Gas Tennessee Valley Authority Washington Public Power Supply System This report is for 3 our use. Comments regarding the appropriateness and completeness of the schedule and the good practices provided in this report should be directed to the GE Project Nianager of the BWROG Excess Flow Check Valve Committee at N1 ail Code 182. GE Nuclear Energy,175 Curtner Avenue San Jose. CA 95125.

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j GE Nuclear Energy Excess Flow Check Valve Testing Relaxation Table of Contents

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1. Background

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2. Tech Specs and I.icensing Basis 3

2.1. Technical Specifications 3

2.2.1.icensing flases 4

2.2.1. General Design Criteria 4

2.2.2. Regulatory Guide 1.1 I (Reference 6) 5 2.2.3.10CFR50 Appendix J (Reference 7) 7 2.2.4. ASN1E ON1-10. Subsection ISTC (Reference 8) 7

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2.2.5. Other t.icensing Bases 8

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3. Consequence Evaluations 9

3.1. Itadiological Conseq uences 9

3.2. Operational Consequences 9

4. Risk Evaluation

!I 4.1. Failure Studes 11 4.2. Failure flates 1I 4.3. Instrument I.ine lireak 12 4.4. llelease Frequency 12

5. Discussion 16 i

6. Conclusions I9

7. References 20 aIttachtnent. l Survey Results Surninary 2l

,tttachinent B Instrutnent I.ine Break Radiological tnalysis 3l l

a

C (iE Nuclear IOrcrgy Sittrattrary of Results This report reviews the licensing requirements, operational experience and consequences associated with testing requirements for fixcess Flow Check Valves (liFCV)in j

instrument lines connected to the Reactor Coolant Pressure lloundary (RCPil). Testing ofI!FCVs in instrument lines connected to the containment atmosphere is not required.

The purpose of the report is to explore strategies for eliminating or extending the testing interval lbr !!FCVs. This would provide a benefit to all llWRs in the form of reduced outage time, occupational exposure and associated costs.

The review concludes that the safety significance ofI!FCVs is extremely small and that they need not be addressed by plant Technical Specitications. The review also concludes that demonstrated experience of valve reliability, coupled with low consequences of excess flow check valve failure, providejustitication tbr extending the test interval up to once in ten years. Ilowever, elimination of testing is not recommended since periodic testing provides assurance of reliable functioning of the salves to provide an effective means of protection of the working environment if an instrument line break were to occur.

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Q GE suclear Energy Excess Flow Check Valve Testing Relaxation

===1.

Background===

liFCVs are utilized in llWR containments to limit the release of fluid in the event of an instrument line break. Table 1-1 lists typical instrument functions for lines containing liFCVs.,\\ lark Ill plants in general do not contain as many I!FCVs because many of the instrument racks are located within the accessible containment building.

Table 1 Ty pical RCPil EFCV applications Instrument Typical te oIEFCVs Ensironment Sensed I

RPV Lesel Pressure 16 RPV RPV llead Pressure i

RPV Main Steam l.me Flow 16 RPV Core Plate dP 3

RPV l

Recirculation Pump Seal Pressure 4

RPV Rectreulation Pump Suction Pressure 2

RPV Recirculation Flow 24 RPV Recirculation Discharge 8

RPV Recirculation Pump DitTerential Pressure 4

RPV RCIC Steam Lme Flow 4

RPV llPCI Steam 1.ine Flow 4

RPV l'otal 86 EFCVs in instrument lines which connect to the RCPIl are normally tested during refueling outages to meet Technical Specification requirements. Instrument lines that connect to the containment atmosphere, such as those which measure drywell pressure, or monitor the containment atmosphere or suppression pool water level, are considered extensions of primary containment. A failure of one of these instrument lines during normal operation would not result in the closure of the associated liFCV, since normal operating containment pressure is not sufficient to operate the valve. Such I!FCVs will only close with a downstream line break concurrent with a I.OCA. Since these conditions are beyond the plant design basis, EFCV closure is not needed and containment atmospheric instrument line EFCVs need not be tested.

Testing can require several hundreds of manhours to complete and can be a critical path activity during some outages. Testing typically requires the reactor to be pressurized to normal operating pressure. EFCVs are generally tested by opening an instrument drain valve, and observing valve closure by either direct indication (valve position indicationL or by a combination ofindirect indications (audible sounds, pressure, temperature, level, or flowrate). Attachment A provides the results of a survey of flWR plant specilie design and operating practices associated with EFCVs.

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GE Nuclear Energv EFCVs are made by several manufacturers. Some are simple self-actuating ball check valves, with allowable leakage rates of up to 3 gpm when closed. Others are more sophisticated devices, with remote position indication and a solenoid-operated reset.

EFCVs are not required te be leak tight.

The purpose of this report is to address issues associated with extending the EFCV testing interval.

Current Impact of Testing The benefits of EFCV test relaxation lies in reduced cost oflabor during outages and reduction in outage lengths without significantly impacting the risk to the general public.

An estimate of the impacts of EFCV testing made at one BWR is shown in Table 1-2.

Note that at this BWR some of the EFCVs were bench tested during the outage rather than in place (in-situ), Bench testing increases the labor and decreases the exposures associated with EFCV testing.

Table 1 EFCV Impact Activity Labor Exposure Critical Path (Mau-hours)

(31an-RE31)

Time (hou rs)

Valving in/out 38

.75 In-situ Testing 92

.6 Bench Testing 224 0

Instrument line refill 38

.75 Total 392 2.1 16 The utility estimated that EFCV testing was costing them about $125,000! year (averaged over a 24 month cycle). In addition, a certain personnel risk occurs due to the need to test the valves.

Planned Approach The planned approach to reduce the exposures, costs and outage impact resulting from EFCV testing is tojustify a less frequent EFCV testim interval than is currently specified in Technical Specifications. The planned approach provides ajustitication (Section 2) for relocation of the EFCV testing requirements from Technical Specifications to administrative documents such as the Technical Requirements Manual (TRM). This conclusion is supported by offsite dose and reliability evaluations contained in Sections 3 and 4. Relaxation of the testing interval relies upon plant specific submittals based on Option B of 10CFR50, Appendix J.

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GE Nuclear Energe 2.

Tech Specs and Licensing Basis 2.1.

Technical Specifications Each BWR Technical Specification is unique. but has certain similarities to other Technical Specifications. BWR/4 Standard Specitications (Reference 1) Section 3.6.3 requires that EFCVs be operable during operating conditions with the reactor pressure above atmospheric. Surveillance requirement 4.6.3.4. originally based on the fuel cycle length, specifies that operability be demonstrated once per 18 months.

Improved BWR/4 Technical Specifications (Reference 2) provide relaxed restoration times (from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) if an EFCV is found to be inoperaba. but still require testing at an 18 month interval. SR 3.6.1.3.10 states:

'l'eri& each reactor instruntentation line EfTl' actuates {on a sinlulated instrument line break to restrict flow to < l gph). ' '

The specified interval is "[l8/ months ' The NRC provides the basis tbr the

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recommended interval in Reference 2 as follows:

"The [18] monthfiequency is haved on the need to perform this surveillance under the conditions that apply during a plant antage and the potentialJirr an unplanned transient ifthe surveillance were performed with the reactor at power. Operating eaperience has shown that these components usually pass this surveillance when performed at the [13l j

monthfrequency. Therefore theJiequency was concluded to he acceptable / rom a reliability standpoint. '

Reference 3 provides the criteria in Table 2-1 tbr establishing a Technical Specification 1 imiting Condition thr Operation (LCO). If an LCO is not needed any surveillance testing associated with that LCO similarly would not be needed.

' information included in brackets [ l represent plant specific information 1

l GE i 'uclear Energv Table 2-1 LCO Criteria Requirement Comment Criterion I installed mstrumentation that is used to detect I FCVs do not satisfy this cciterion and indicate in the control room. a sigm/icant abnormal degradation of the reactor coolant Criterion 2 ~

pressure boundary.

A process variah/c, designfeature, or operating liFCVs ua not satisly tlns criterion restriction that is an imtial condition ofa design basis accident or translent analysis that either anumes thejailure ofor presents a challenge to the integrity ofa lission product barrier.

Criterion 3 A structure, system. or component that is part of tiFCV closure is not needed for i

the primary success path and u hichfunctions or accident mitigation (see Section 3) actuates to mitigate a design basis accident or transient that either assumes thefailure of or presents a challenge to the integray of ajission product barrier.

Criterion 4 A structure. system, or component which liFCV closure is not significant to operating experience or probabilistte rid public health and safety tsee Section assessment has shoun to be sigmficant to public 4) health and safety.

From Table 2-1, and as discussed in later sections of this report, it is evident the EFCVs do not meet any of the criteria for having a LCO included in Technical Specifications.

Consequently, justification exists tbr relocating any testing requirements from the Technical Specifications. It can be noted that the improved BWR/6 Standard Technical Specifications (Reference 4) do not currently contain reference to EFCVs.

2.2.

Licensing Bases 2.2.1. General Design Criteria l

General Design Criteria (GDC) 55 and 56 contained in 10CFR50, Appendix A l

(Reference 5), provide design requirements for isolation oflines that penetrate the j

primary containment. Instrument lines which monitor the RPV or containment internal l

conditions are subject to isolation requirements, but as noted in the GDCs, "imless it can he demonstrated that the containment isolation provisionsfor a specific class oflines, i

such as insirtunent lines, are acceptable on some other defined basis ' An alternate licensing basis acceptable to the NRC for isolation ofinstrument lines connected to the RCPB is described in Regulatory Guide 1.11 (Reference 6). This is described in Section n..,...,..

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GE Nttelear Energy 2.2.2. Regulatory Guide 1.11 (Reference 6)

Instrument lines constitute closed, extended containment boundary system piping outside containment. Regulatory Guide 1.11," Instrument Lines Penetrating Primary Reactor Containment " accepts instrument lines as extensions of primary containment, and allows that their configuration satisfies the "other defined basis" requirements of GDC 55 and

56. Automatic isolation ofinstrument lines during a LOCA is not prudent, since these instrument lines provide safety functions for reactor protection and containment isolation which need to be operable during a I.OCA.

Regulatory Guide 1.11 design requirements are summarized in Table 2-2.

Table 2-2 Regulatory Guide 1.11 Requirements l

Line Ty pc Requiren'ent Comment with regards to EFCVs Lines that are a) Shindd sansly the requirements /or redundamy.

Instrument lines part af the independence, and testabihty of the protection system generally comply uith protection system this requirement. This requirement is not applicable to EFCV testing.

Lines that are b) Should be si:ed or ortficed to assure that in the event Ihis requirement is not part af the of a postulatedfailure of the piping or of any applicable to EYCV protection system component (including the posudated rupture of any testing.

valve body) in the line outside primary reactor and containment during normal reactor operanon.

Item 6)(3) may not be (lJ the leakage is reduced to the manmum extent satisfied if the UVCVs Lines that are not practical consi3 tent svuh other safety fail to function.(See part of the requirements section 3.2) protection system

(.') the rate and extent ofcoolant loss are svuhin the capabilay of the reactor coolant makeup Neetion 3 and system.

Attachment B (3) the integray andfunctmnal performance of demonstrate that the secondary emuainment. ilprovided. and designs generally meet associated sakty systems le g.Jihers.

the intent of this standby gas treatment system) sval be requirement, maintained, and H) the potential utfute exposure svill be substantially belosv the guidelines of 10CFR100.

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GE Nuclear Energy Lines that are c) Should be provsded wah an Isolatmn valve capable of LYCVs generally comply part af the automatic operation or remote operationfrom the uith this requirement.

protection system control room orfrom another appropriate location.

and located in the hne outs:Je the contamment as and close to the contamment as prac tocal. There should be a hich decree of avvurance that thiv valve:

Reliability oflifCVs is Lines that are not (I) uillnot close accidentally during normal discussed further in part of the reactor operation, Neetion 4.

protection system (2) uill close or be closed ) the instrument line integrity outside contamment is lost during normal reactor operation or under accident conditions, and (3) will reopen or can be reopened under the condaions that worddprevail when valve reopening is appropriate. power operated valves should remain as-is upon loss of power. The status (opened or closed) of all such isolati<m valves should be indicated in the control room. If a remotely operable valve is provided. sufficient information shotdJ be avadable in the control room or other appropriate location to assure timely andproper actions by the operator.

Lines that are th Should be conservatively designed up to and instrument hne part of the including the isolation valve and ofa quahty at least separation generally i

protection system eqmvalent to the containment, Theseportions of the addresses this lines should be located and protected so as to requirement.

and minimi:e hkelihood of their being damaged accidentally. They should be protected or separated This requirement is not Lines that are not to preventfailure ofone linefrom induemgJadure of apphcable to VVCV part of the any other hne. provisions should be inchided to testing.

protection system permit periodic visual mservice inspection.

partictdarly of those portions of the lines outside cemtainment up to and including the isolation valve.

Lines that are c) Should not be sa restricted by components m the lines.

Instrument line orincing part af the such as valves and orillees, that the response time of generally does not protection sy stem the connected instrumentation will be mcreased to an impact the performance unacceptable degree.

of the instrumentation.

and This requirement is not Lines that are not applicable to EFCV part of the testing.

protection system As shown in Table 2-2 the design criteria of RG l.1I are generally met in by the current BWR designs. It should be noted that with the exception of testing iniplied by RG 1.11 item e, no specific testing requirements are delined by the regulatory guide. The discussion portion of the regulatory guide states:

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i GE Nuclear Energv 1

" Sufficient experience with valves of a similar npc should he available to auure a high probability that the valve will not close when the instrmnent line is intact and its safetyJimetion is required, but that it will close ll the l

instrument line is ruptured downstream" The perli>rmance of EFCVs discussed in Section 4 provide this high degree of assurance.

l 2.2.3.10CFR50 Appendix J (Peference 7) i i

A Containment isolation Valve (CIV)is detined in Appendix J as "any valve which is relied upon to perform a containment isolation fimetion.

Appendix J prescribes air-testing rmluirements tiir containment isolation valves, and provides tbr exemptions li>r valves uhich are water sealed. Niost EFCVs are connected to water-filled systems, and j

are tested for operability with water.

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10CFR50 Appendix J testing is only applicable to EFCVs if they pertbrm a containment isolation function. EFCVs are not required to close in response to a containment isolation signal and are not required to operate under post-1.OCA conditions. As discussed in Section 3.1. the functioning of EFCVs is not necessary to remain within 10CFR100 limits.

Consequently tbr purpose of 10CFR50, Appendix J, CIV testing, EFCVs do not provide a containment isolation function and are exempt from consideration under Appendix J.

2.2.4. ASME OM-10, Subsection ISTC (Reference 8)

ASNIE Ohl-10. Subsection ISTC. "hnervice Testing of l'alves in I.ight-IVater Reactor Power Plants "(Reference 8) establishes the requirements tbr inservice testing of valves in light-water reactor nuclear power plants. Testing is required for valves " required ui perform a specilleJimetion in shutting down a reactor to the cold shutdown condition. in maintaining the cold shutdown condition or in mitigatmg the consequences of an accident' EFCVs are not needed to mitigate the consequences of an accident because an instrument line break coincident uith a design basis 1.OCA would be of sufficiently low probability to be outside of the design basis. Furthermore ibliowing a design basis 1.OCA. isolation is not necessary to achieve acceptable consequences of an accident (see Section 3.1)

Thercibre, because EFCVs do not peribrm the functions for uhich the ISTC applies (i.e.

they are not needed to mitigate an accident). EFCV testing is not required by the ASN1E i

code. Consequently, ONI-10 is not considered to be part of the licensing basis for EFCV testing.

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GE Nuclear Energy 2.2.5. Other Licensing Bases 10CFR50 Appendix A. General Design Criterion (GDC) 55 provides the isolation requirements for lines penetrating containment that are connected to the reactor coolant pressure boundary. The NRC has allowed exceptions to the GDC in previous evaluations in a similar manner to that provided through Regulatory Guide 1.11. Another similar evaluation has been established for the CRD withdrawal lines.

NUREG-0803," Generic Safety Evaluation Report Regarding Integrity of Hll'R Scrain System Piping' was issued in response to draft NUREG-0785, "Sa/ety Concerns ils30ciated It'ith Pipe Hrcaks in the Hil'R Scram System ' In response to NUREG-0785.

GE Nuclear Energy prepared NEDO-24342,"GE Evaluation in Re3ponse to XRC Request Regarding Hil'R Scram System Pipe Hrcaks ' In NEDO-24342, GE Nuclear Energy, contended that the CRD withdraw lines are small diameter (3/4") lines and pertbrm important safety functions, and, theretbre, automatic isolation valves should not be used.

fhe staff concluded that a departure from the explicit requirements of GDC 55, such as that represented by the CRD hydraulic design, isjustified without isolation valves. I his assessment was based on the fact that the CRD withdraw lines penetrating the containment and routed to the HCUs are small in diameter (3/4") and are conservatively designed and of high quality. Even when the staff postulated a break in one of these lines during reactor operation (including scram), they lbund that:

The restricted flow area of the CRD limits the reactor coolant leakage to a very small value (within the capabilities of the reactor coolant makeup capabilities), and The reactor can be shut down and cooled down in an orderly manner.

The similarity between the CRD line discussion and instrument lines provides a precedence for acceptability of the instrument lines without credit thr EFCVs. Both cases represent small lines which connect to the RPV pressure boundary and contain restricted flow areas. Based on the low consequence ofinstrument line breaks a case can be made that EFCVs are not required. Testing of the EFCVs is not justitiable if the valves themselves are not required.

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Q GE Nuclear Energv 3.

Consequence Evaluations 3.1.

Radiological Consequences Radiological consequences from RCPB instrument line breaks have been evaluated at most plants to show compliance with Regulatory Guide 1.11 and are documented in some UFSARs. A typic.il GE radiological evaluation of the mstrument line break with and without a %" oritice installed has been conducted ( Attachment B) using a GE methodology which has been accepted by the NRC in GE FSAR submittals. No credit is taken in these evaluations tbr the operation of the Standby Gas Treatment System (SGTS).

The results of the evaluation (see Attachment B) indicate that even without a %" oritice installed, the resulting thyroid dose at the site boundary is 16 Rem u hich is about 5"'o of the 10CFR100 limit and may be considered insignificant. Similarly, whole body exposures are shown to be about 1% of the 10CFR100 limits without credit Ihr an oritice.

With an orifice, the doses are reduced by a factor of more than 5, which provides added conservatism.

j The radiological consequence of EFCVs failing to function upon demand is sufticiently low to be considered insignificant. Specific analyses are needed to contirm this conclusion at each plant, but similar results would be expected. Because of the insignificant consequence with EFCV failure,it can be concluded that the EFCVs are not needed to assure a containment isolation function.

3. 2.

Operational Consequences The operational impact of an EFCV which is connected to the RPV pressure boundary failing to close is based on the environmental effects of a steam release in the vicinity of the instrument racks. With the exception of potentialjet impingement impacts, the environmental impact of the failure ofinstrument lines connected to the RPV pressure boundary is the average release of about 6,000 lb/hr of steam into the reactor building (assuming a %" restriction in the line). A release of this magnitude, with or without an orifice, is within the pressure control capacity of Reactor Building ventilation systems (typically greater than 100,000 cfm). Due to its large volume, the bulk reactor building j

temperature would not be expected to be signilicantly affected except in the vicinity of i

the break.

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this magnitude or greater may exceed the capability of the standby gas treatment system i

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GE Nuclear Energy (SGTS) to remove moisture and control humidity to the charcoal beds. I hmever, a detailed evaluation would be necessary to confirm SGTS performance. No credit tiir SGTS operation is taken in the otTsite dose calculations ( Appendix II) and the results are acceptable. Theretbre the intent to maintain the integrity c.nd functional performance of secondary containment ibliowing instrument line breaks is met independent of SUTS operability.

Separation of equipment in the reactor building minimizes the operational impact of an instrument line break on other equipment due to jet impingement. Nevertheless, the presence of an unisolated steam leak into the reactor building would require a reactor shutdow n and depressurization to allow access to manually isolate the line.

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1 4.

Risk Evaluation l

4.1.

Failure Modes The basis for the evaluation of public risk relies on an estimate of the failure of the EFCV to close when required, the instrument line break frequency, and the offsite consequence of the event. It can be speculated that most EFCVs fail to close due to sticking rather than some other mechanical problem Sticking is assumed to be a time dependent phenomenon which controls the EFCV unavailability as a function of the surveillance testing interval.

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4. 2.

Failure Rates The reliability of EFCVs was evaluated based on testing experience provided by different BWR utilities in response to the llWR Owners' Group survey ( Attachment A). The data selected from the survey tbr evaluation, as well as the result of the evaluation are l

presented in Table 4-1. The composite data show a very high reliability tbr the EFCVs with a total of 9 failures in over 10,000 valve years of operation.

The values shown in Table 4-1 for the " Upper Limit Failure Rate" were calculated using the following equation:

1 2r = H 2

,a W

Where:

2r is the upper limit failure rate per hour T

is the operating time in hours r

is the number of failures

y, is the value taken from the chi-square distribution tables which corresponds to 2r+2 degrees of freedom and a = 0.05 (l-u = 0.95 is the specified contidence level)

The last row of Table 4-1 shows the composite failure rate of EFCVs based on the data from the plants listed in the table. The best estimate and the confidence limit were calculated using the sum of the operating times and failures shown in the table. Table 4-2 shows failure rates calculated for each EFCV manuhteturer. based on the data presented in Table 4-1. The results of the failure rate analysis presented in Tables 4-1 and 4-2 show relatively consistent values if calculated for each plant or for each valve manufacturer.

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(iE Nuclear Energy Ilased on this data, the composite failure rate value of 1.67E-7/hr between testing associated with a 95% confidence level for EFCVs is considered a best estimate of the reliability of EFCVs based on the current testing experience.

4.3.

Instrument Line Break The instrument line break frequency was calculated based on the WASil-1400 small pipe break litilure rate of 6.1E-12 per hour per foot ofline, and a conservatively assumed average 100 feet ofline from the EFCV to the instrument. It has been assumed that this failure rate applies equally to all small pipe sizes (% ' to 2", per WASil-1400). It is also assumed that this value is independent of whether the line is pressurized or not.

l Therefore. for a single instrument line the resulting frequency is 5.34E-6 breaks per s ear l

(6.1E-12

• 8760 hrs / year

• 100 ft. = 5.34E-6).

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4.4.

Release Frequency The risk impact on the public health and safety from EFCVs can be evaluated as the product of a release frequency (due to a break in an instrument line concurrent with an EFCV failure to close) and the consequence of the release (see Section 3.1). The release frequency can be calculated based on the instrument line break frequency (Section 4.3) and EFCV litilure to close probability.

Table 4-3 shows the release frequency from a single instrument ;ine, assuming diff erent test intervals for the EFCV. The release frequency was calculated using the following equations:

RF = /* 2 (2) and

.1 = ). g (3)

Where:

RF is the release frequency per year

/

is the instrument line break frequency per year (Section 4.3) 2 is the EFCV unavailability (failure to close probabilim

).

is the EFCV failure rate per hour (Section 4.2) is the EFCV surveillance test interval in hours Based on the release frequency shown in Table 4-3 for one instrument line, and assuming 86 instrument lines with testable EFCVs in a plant. the release frequency from any broken instrument line is:

For 18 months sur eillance test interval 86

• 5.86E-9 = 5.04E-7 eventryear 12 4

4 (iE Nuclear Ener,<4y For 10 years surveillance test interval 86

• 3.91 E-8 = 3.361%

events / year These release l'requencies are sulliciently low that it can be concluded that a change in surveillance test l'requency has minimal impact on the valve reliability.

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1 (iE Nuclear Energy Table 4-1: EFCV Failure Rates l*lant Slaken)

Operating Operating Nurnber Best l'pper Notes l Note ll EFCF' Tinte Time of Estiinate f.imit l years]

l hours l Failure.s Failure Rate railure l Note 2l l Note 2l Virl Rate //h/

l Note 3l 13 row ns Ferry Marotta 100.5 8.80E4 05 3

3.411s-06 8.81 E-06

4. 6 Ilrunswick Valcor 267 2.34 E 4 06 0

0 1.28 E-06 5

Clinton Dragon 220 1.93 E + 06 0

0 1.5 5 E-06 DAl.C Marotta 1974 1.73l:407 0

0 1.73 E-07 Dresden Chemquip 922 8.071:+ 06 0

0 3.7 l E-07 fermi 2 Dragon 930 8.15 E + 06 0

0 3.68 E-07 Fitzpatrick Marotta 2019 1.77E+07 0

0 169E-07 Monticello Chemquip 2314 2.03 E+ 07 1

4.93 E-08 2 34F-07 Oyster Creek Chemquip 465 4.07E +06 0

0 7.36E-07 Susquehanna Marotta and 144 1.26 E + 06 4

3.17E-06 7.26E-06

6. 7 Valcor VY Chemquip 1725 1.51E+07 1

6.62E-08 3.14 E-07 WNP2 Dragon 1344 1.lHE+07 2

1 69E-07 5.341!-07 Composite 12424.5 1.09 E+ 08 il 1.0lE-07 1.670-07 Notes to Table 4-l:

1.

1.aSalle data was not included in the table due to inconclusis e reported data 2.

1)eternuned by multiplying the nurnber of tested l I CVs with the time period durmy which the number of occurring tailures was reported 3

'lhese tailure rates are obtained by dniding the number of f ailures by the operatmg time.

4 l'tihty reports 67 sahes tested durmg last outage, of which 3 tailed the test l'he I i CV operatmg time for ;l tailures is 67sabes

• 1.5)ean = 100 Sycan 5

Utihty reports 89 sahes per umt tested at an intenal ol' 18 months (15 ) cars) Ihere wer: no f ailures dunng the last tesi of each unit. Therefore, the 1.1 CV operatmg time for o tailures is 2umts

• M9 sakes

• I 5) cars = 267 ears 3

6 Consen atne results 'Ihe s abes reported " tailed" passed a bench test af ter bemg remos ed 7.

Unit 2 reports 96 f.ICVs tested per umt eser) outage. presentl) IM month egle and 4 ladures durmg the last outage of I mt 2

't herefore, the I i CV operating time is 96 $ahes* l 5 > cars = 144 yean Only l' nit 2 data shown

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(iE Xuelear Energr Table 4-2: EFCV Failure Itates by Manufacturer Make of Pla n t

• Operating Operating Number liest

!!pper EFCV Time Time of Estimate Limit l> carsj lhoursl Failu res Failure Failure Itate l/h)

Rate l/hl Chemquip Monticello, VY, 5426 4.75 E+07 2

4.21l!-08 1.331?-07 Oyster Creek, Dresden Dragon Clinton, Fermi 2494 2.181?+ 07 2

9.2[:-08 2.89E-07 2, WNP2 Marotta llrowns Ferry, 4093.5 3.59E+07 3

8.3 7E-08 2.161i-07

DAEC, Fitzpatrick, Susquehanna*

Valcor llrunswick, 67.5 5.91 E+ 05 0

0 5.07E-06 Susquehanna*

• Susquehanna has a combination o/ vahes; No breaLiown of numbers of failures he valve t)pe was available.

Table 4-3:

Itclease Frequency from a Single Instrument line (based on 5.34E-6 instrument line break frequency, and 1.67E-7 EFCV failure rate) i EFCV Test EFCV Test Interval EFCV l!navailability Release Frequency Interval l hours l

[

l/ carl 3

lyenrs)

(1 RF l.5 13140 1.10E-03 5.86E-09 2

17520 I.46E-03 7.8 I E-09 6

52560 4.39 E-03 2.34 E-08 10 87600 7.31 E-03 3.9 t E-08 15 y

(iE Nuclear Energy 5.

Discussion itisk of EFCV failure Failure of an EFCV to close will not involve a significant increase in the probability or consequences of an accident previously evaluated. If an EFCV fails to close, a low leak rate will exist due to the %" orifice, effective valve restriction, or instrument tubing size.

FSAR analyses for a ruptured instrument line have shown offsite doses well below 10CFR100 limits without EFCVs.

The ef fect of extending the EFCV testing intervals is a corresponding increase in the i

potential frequency li)r a release. However, esen if the test interval were increased to once in 10 years, the release frequency from an individual line remains very low at about 4E-8/ year. Therefore, considering the low consequence of release, the extension of the surveillance interval does not affect the risk to the public associated with a failure an i

instrument line and the failure of an EFCV to perform its intended function.

The risk to the public can be shown by combining the release frequencies in Section 4 with a consequence of release from Section 3.1. The corresponding public risk with the current testing basis can be shown to be ~3E-5 mrem /yr. (Whole Body)[5.04E-7 events!yr x.05 Rem / event = 2.5E-5 mRemlyrl. With an extended testing interval this value changes to ~2E-4 mRemlyr. (Whole Body). This is live to six orders of magnitude below 10CFR20.105 annual exposure limits to the general public of 500 mrem'yr (Whole Body).

Clearly the risk to public health and safety (based on plant experience) is extremely low and not impacted by the testing interval. This furtherjustities that the EFCVs do not meet the criteria ti)r being included in the Technical Specifications.

As discussed in Section 3.2, instrument lines in Mark I and Mark 11 plants are located outside of the Primary Containment in the Reactor Building,in an area served by the Standby Gas Treatment System following an accident. This provides additional mitigation of any postulated offsite release from a broken instrument line.

Personnel llazard Reduction Changing the testing interval, as with any reduction in required maintenance, inherently reduces the risk ofindustrial accidents including inadvertent exposure to radioactive liquid and occupational exposure. Furthermore there is a consequential reduction in the amount of liquid radwaste that requires processing. Both these issues provide lavorable benefits from an increased testing interval.

l6

GE Nuclear Energy l.icensing liasis Section 2 provided justification that Technical Specifications form the sole licensing basis fi>r testing of RCPil I!FCVs. The review also demonstrated that a basis for establishing an LCO in Technical Specifications for I!FCVs could not be justified. The bases li>r the surveillance interval is that testing during outages shows reliable perti>rmance. The evaluation in Section 4 shows that even with a 10-y ear testing interval, perti>rmance would not be significantly degraded.

No General Design Criteria, Regulatory Guide or ASNil! code requires testing ofI!FCVs unless they are considered to provide a containment isolation function. The results of Section 3.1 justify that the !!FCVs are not needed to provide a containment isolation function.

Furthermore, the RCPfl instrument lines, in w hich the liFCVs are installed, are similar to the Control Rod Drive System withdraw lines, in that they are normally pressurized to reactor operating pressure, are highly restricted and are unisolable from the reactor. 'I he CRD insert and withdraw lines have been accepted by the NRC without isolation provisions.

It can be concluded that I.COs for the I!FCVs are not required to be included in the Technical Specifications and that liFCV testing requirements can be deleted from plant Technical Specifications.

1 I

Reliable Design Requirements The design requirements contained in Regulatory Guide 1.1I provide a highly reliable design with insignificant consequences in the event of an instrument line break. These design requirements are not changed by an alternate testing interval.

Performance llased Testing Although I!FCVs are not required to be tested to meet the requirements of 10CFR50 Appendix J, the approach specified in 10CFR50, Appendi, ', option ll, can provide an approach familiar to the NRC for establishing an acceptable interval between tests of the I!FCVs.

10CFR50, Appendix J, Option 11 allows a perfbrmance-based approach tiir determining the leakage rate surveillance testing frequencies for Type A. Type it and Type C containment penetrations. lixtensions of ty pe 11 and Type C test intervals are allowed based upon completion of two consecutive periodic "as-found" tests where the results of t

17

t GE Nuclear Energy each test are within a licensee's allowable administrative limits. An "as-lbund" test is one perfbrmed prior to any periodic maintenance. Acceptable performance history of each component is used as the basis for extending the surveillance test interval. If performance experience satisfies the criteria test intervals may be increased up to a maximum of 120 months (10 years).

1 The Nuclear Energy Institute (NEI) prepared NEI 94-01 (Reference 9) to provide implementation guidelines ihr meeting 10CFR50. Appendix J, Option 11. NEl 94-01 (which is endorsed by the NRC in Regulatory Guide 1.163 (Reference 10)). states:

1

ldditional considerations used to determine appropriate fiequencies

\\

may include service life, environment. past performance. design. and safety impact. "

At most IlWRs historical EFCV pertbrmancejustifies an extension of the testing interval based on peribrmance based criteria, and the extension of the surveillance interval does i

not change the design, function, or operation of the EFCVs. When requesting test interval extension. EFCVs with similar operating and environmental conditions should be combined into groups.

I i

I L

18

4 (iE Nuclear Energy 6.

Conclusions Failure of RCPB EECVs does not have adverse consequences in terms of risk to the public. EFCVs are not necessary to provide a containment isolation function.

Consequently, the only licensing basis for testing requirements are contained in the Technical Specifications. The information in this report provides justification for relocating the RCPB EFCV testing requirements from Plant Technical Specifications to the Technical Requirements Manual.

Elimination of testing,is not recommended. Without periodic testing, a stuck EFCVs would not be detected. Automatic closure of EFCVs improves access to instrument lines in the event of a line break and improves maintainability and reduces contamination of the reactor building. Since EFCVs provide a simple method of avoiding these adverse operational considerations associated with recovery following instrument line breaks. the BWROG believes they should be retained in the plant design, and tested periodically in accordance with administrative controls. EFCV testing does not need to be retained in Plant Technical Specifications.

Reliability and consequence analyses provide a performance basis ihr extending the test interval up to 10 years without significantly affecting offsite risk. It is recommended that the testing interval be extended based on perfbrmance-based criteria included in 10CFR50, Appendix J Option B. An extended testing interval reduces plant costs and occupational exposures without significantly affecting offsite risk. When requesting test interval extension, EFCVs with similar operating and environmental conditions should be combined into groups. A test schedule based on actual valve peribrmance should be established and controlled in plant administrative procedures A staggered test schedule is suggested, in which a portion of the valves are tested each outage. Each EFCV should be tested at least once every ten years.

l9

)

I (iE Xuclear Energy I

l 7.

References 1.

" Standard l'ecimical Specifications. Lieneral Electric Plarus ' N\\ inE() 0123, Rey 3.

December 1980 2.

" Standard l'echnical Specifications. (ienera! Electric Plants tHIl'R/h '. N \\ !Rl%

1433, April 1995

3. " Technical Speci// cations",10CFR50.36, July 29,1996 4.

' Standard Technical Specill cations. General Electric Plants (Bil'IM ' NliRUli 1434, April 1995 5.

"(' ode of Federal Regulations. Part 10('FR50...Ippendix cl' 6.

" Instrument I.ines Penetrating Primary Reactor Containment ' Regulatory (iuide

1. i l (Safety Guide 11), 3/10!71 7

"( ' ode of Federal Regulations. Part IOCFR50 <lppendix J" 8.

"hiservice Testing of l'alves in Light-ll'ater Reactor Power Plants" ASNlE O.\\1-\\ 0.

Subsection ISTC,1995 9.

"hidustry Guidelines for implementing performance-based option vi 10( 'FR50

<lppendix J", NEI 94-01. July 26,1995 l0. " Performance Based ('ontainment Leak-Test Program", Regulatory (iuide l.I63, September 1995

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\\

GE Nuclear Energy Attachment B l

1 Instrument Line Break Radiological Analysis Analysis i

This analysis concerns the Instrument Line Ilreak Accident (ll. IIA). The calculation will i

be done using the standard Gli II. IIA based upon Reference [1-1.

GE Standard Analyses The Gli standard analysis is described in Reference 11-1. Chapter 8. which assumes a break in an unisolable small line (typically an instrument line) w hich is choked by a one-quarter inch orifice. The break continues unabated li>r ten minutes at which time temperature sensors in the reactor building (or whatever area outside containment) cause j

the operators to respond by scramming and shutting down the reactor. The reactor liillows a standard cool down of 100 F/hr for a period of 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at u hich time the accident is terminated.

The flow rate and depressurization curves ti>r the restricted case are given in Reference 11-

1. Figures 8-1 and 8-2 respectively. The total integrated release li)r the restricted case is I

approximately 33.000 lbs. For the unrestricted case, a total of 100.000 lbs of steam was l

assumed to be released to the reactor building.

1 The release of lission products is restricted to iodines uhich are radiologically the most significant isotopes ihr site dose considerations. The reactor water inventory is assumed to be a Technical Specification limits of 0.2 pCilgm Dose Equivalent I-131. Since the exact determination of separate isotopic species is dependent upon the dose data set used, the dose conversion factors consistent with TID-14844. Reference 11-3, are used in this analysis and are given in Table 11-1 below.

To account for iodine spiking source terms, the ibliowing algorithm is used:

At accident initiation.15% of the gap inventory ofiodines is released to the vessel water and is assumed homogeneously mixed in the water. The 95"o gap inventory is given in Ref.11-1 and is listed in Table Il-l.

As the reactor depressurizes, the remaining 85"6 is released proportional to the depressurization so that all the gap inventory is released to the vessel by the i

termination of the accident.

l l

1 29

GE Nuclear Energy As the vessel water is released to the reactor building, it is assumed that the iodine in the water is released to the air space directly proportional to the flash fraction of the released water. This is a significant conservatism in the calculation of the released iodine.

1 The tission products released to the air space are assumed released to the environment at i

an assumed rate of 4800% per day as a cold ground level release. Operation of the Standby Gas Treatment System is not assumed. Specilie computational detail is given in i

Table B-2.

Ilesults The results of an evaluation of postulated unrestricted and restricted instrument line breaks at Dresden are given in Figures B-1 and B-2. The results show the thyroid and whole body dose as a function of time for the site boundary (exclusion area boundary) and for the low population zone boundary (LPZ), and are summarized below.

{

I)ose Unrestricted Restricted

. Unrestricted Result Result fract' 'n of (Rem)

(Rem) 10CFR' ' Limit

\\

Thyroid LPZ

.5

.09 Site Boundary 16

.9 5.3 Whole Body LPZ

.003

.0006

.06 Site Boundary

.05

.008 1.0 30

s' (iE Xuclear Energy Table 11-1 Isotopic Input 1)ata DCl2 Cone (iapInven Isotope M/see)

(RI!N1/Ci)

(pCi/gm)

(Ci/flundle)

I-131 9.9771 li-7 I.4 Sli+ 6 0.047 2.I 1-132 8.42631i-5 5.35 f(+3 0.415 3.2 1-133 9.25681i-6 4.00li + 5 0.326 5.0 l-134 2.19631i-4 2.50l!+ 4 1.207 5.4 1-135 2.92391i-5 1.241!+5 0.755 4.8 Table 11-2 Other input Data Item Value liasis Nteteorology site boundary 2.615-4 sec/m3 Ref.11-2. Ul:S All I.PZ l.1 li-5 1

Number llundles 724 ling. Data Ilase Niass RPV Water 590.000 lb.

Ref.11-2. l!! S AR 31

GE Nuclear Energy Figure B-1 Thyroid Dose i

10E+2

-ii^.t _

?- -i f h--f^-JLI

. -.. _ _E ? i f.i __ x_ - -

g=nc ::_::-- 4:g==&msm-&; :.

~.~b.E:=?_~':.t r 5=,=5=M=_V: _

1CE*1

^

p_

_g g..

g_

l i

/ S;te Boundary Dose using Restncted Flow l

l

/'

ICE +0

-h.__ =g_

t=ge gg_

=. -

p

.=m.._m-,_._

- rp= - =

a.=

_..__y l

? Dosa usinq One inch t me l

~

i

[

1 CE-1

__ g 7 g 3._.== = = g g== ===

-==r-

=7

~*

. ]: r _{

g 1

_.l. _ ~~

~ l NLPZ Dose using Restricted FI A 1 CE 2 gg. -

_ _ _ = = =

- =====y=-3_

3g_; _.. __ __ -f-:3.g=q_ _ _

_=

+

j ---

1 CE-3

-=-

= _. _..

_ _ = _ _ = =

=-:==._._,g===_+..====y====,==-=-ggg_.3_.-

7

_r y.____._...-

T I

I i

l l

10E-4

~

I i

0 1

2 3

4 5

6 time in Hours Figure B 2 Whole Body Dose 10E+0

.___3g g_m=__ _

._t-

_a

..g._

._L.___

L.-

Site Boundary Dose using One Inch 10E d

_ { ;g =__.

7u2..g_g;m_==

_m___j

_ ;;_ _,_ __=;.

==r j

t 7

j

, Sde Boundary Dose using Restncted Flow {t 1 OE.2

_._gg __ _.

4- =_..

g =

r i

~

.g m

$LPZ Dose usingDne Inch Line 1 CE-3 l

7 g.:

-s==n =.

_ - - _ _, =

g gg=;ggge== = - t==_

-__.=j=ggg= g n.

-z- -.

p y.____._.

_ T-

'N

[

[

~ LPZ Dose using Restncted Flow 1 OE.4 y_

_ggg._.

._-7_

---.=m.-._

7

...._..__..,3 Rem

~

~T i

~ 1-1' _ _._ __

I i

--.- gg. p_.---.

10E-5 g.

= =--t gggg.

.ggg

=

==t_-_--

f_-.__

T 7 _ __. _ __

10E 6 0

1 2

3 4

5 6

Time in Hours t

32 1

l l

l

GE.\\'uclear En :rgy

References:

11-1. Careway, IIA, Nguyen, VD.. Stancavage, PP," Radiological Accident livaluation -

the CON AC Code", General filectric Report NI!DO-21143-1. December 1981.

Il-2. Dresden UFSAR, Subsection 15.6.2.

11-3.

DiNunno J.J., et al " Calculation of Distance Factors ihr Power and Test Reactor Sites', Technical Intbrmation Document 14844, NTIS, March 23.1962.

I l

1 f

.