ML20071B230

Confirms Results of 931204 Telcon Between Commonwealth Edison Co & Nrc,In Which Ceco Requested Notice of Enforcement Discretion from TS 1.0.3 for LaSalle County, Unit 1

ML20071B230
Person / Time
Site:	LaSalle
Issue date:	12/06/1993
From:	Piet P COMMONWEALTH EDISON CO.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
NUDOCS 9404250056
Download: ML20071B230 (9)

Text

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/ Commonw:altn Edison 1400 Opus Race

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/ G.'ovi ti"ros 60c15 December 6,1993 Mr. John Zwohnski i Assistant Director for Projects Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Cummission Washington, D. C. 20555

Subject:

LaSalle County Nuclear Power Station Unit 1 Reque.:t for NRR Enforcement Disciution Regarding Facility Operating License NPF-11, Appendix A, Technical Specification 4.0.3 NRC Docket No. 50423

Dear Mr. 2wolinski:

This letter documents the results of the teleconference held on December 4, 1993, between Commonwealth Edison (CECO) and tha NRC Staff, in which Commonwealth Edison requested a Notice of Enforcemen: Diceretion from Technical Spoeitication L0.3 for LaSalle County Unit 1.

At 1715 on December 4,1993, LaSalle County Unit 1 entered Technical Specification 3.4.2. Action a., due to the inoperability of two (2) Safoty/ Relief Valvce.

Theae two (2) Safety /Rolief valves woro conservatively determined to be inoperable por Technical Specification 4.0.3 due to the time period since verification of the liR settings for valves 1B21-F013B and IB21 F033J. Technical Specification 4.0.3 statea " Failure to perform a Surveillance Requirement within the apacified timo inteval shall constitute a failure to moet the OPERABILI'IY requirements for a T.imited Condition for Operation." Technical Specification 4.0.5 provides surveillance requirements for inservice inspectivu and testing of ASME Code Class 1,2, and 3 components per ASME Se:: tion XI of the Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g). The reactor coolant system Safety / Relief Valves are included in the ASME Code compononta to be tested. The survoillance intorrals for the Safety /Rclief Valves are based upuu ASME Section XI Table IWV 35101. Safety / Relief Valves ID21 F013B and IB21 F013d exceeded the specilled surveillance interval requirements (every 5 years) and were conservatively determined to be inoperable.

Because it is not possibic to perform the required testing with Unit 1 in operation and the cer.ctor coolant t,ystem pressurized, CECO requested that Unit 1 be allowed t-o continue to operate until the March 1994 Refueling Outage (L1R06) or until the next Cold Shutdown, whichaver occurs first, A Notice of Enforcement Discretion was verbally approved by NRR at 2025 CST on December 4,1993.

The basis of the request is provided in Attachment 1 and includes:

The Technical Specification that will be violated:

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301 504 3861:* 2/ 9 Mr. J. Zwolinski December 6,1993 The circumstances surrounding the condition, including the need for prompt action; The safety basis for the request that enforcement discretion be exercised, including an evaluation of the safety significance and potential consequences of the proposed course of action:

Any proposed componnatory measure (o);

The justification for the duration of the request; The basis for the conclusion that the request will not have a potential adverso impact on the public health and safety and that a significant safely hazard is not involved;

• The basis for the conclusion that the request will not invniva ndverse consequences to the environment.

If during duration of this Notice of Enforcement Discretion additional Safety / Relief valves are made or found to be inoperable for any reason, CECO will follow the Action requirements of the Technical Specifications. CECn rariuests that this Notice of Enforcement Discrotion be in effect until an Exigent Technical Specification Amendment is approved. A request for an Exigent Technical Specification Amendment will be submitted on December 10,1993, for NRC Staff review.

This request for Enforcement Discretion has been reviewed and approved by the LaSailo County On-Site Review Committee, in accordance with LaSalle County Station procedures.

CECO sincerely appreciates tha NRC atafia effort and participation in the review of this request. Please direct any questions or comments to Peter Pict, Nuclear Licensing Administrator, at (708) 003 7286.

W trulp urs, c

'd deter L. Piet Nuclear Licensing Administrator Attachment cc: J. B. Martin, Regional Administrator - RIII D. Hilla, Senior Resident Inapector - LaSalle County A. Gody, Proiect Manager - NRR NRC Document Control Desk k W14\lACalle\ncedl.wpfU 2 6

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301 504 3861:* 3/ 9 AITACHMENI

1. TECHNicALSPlCIElCATJONDR LICENSING CONDITION THAT WILLEE -

VIOLATED At 1715 on December 4,1993, LaSalle County Unit i entered Technical Specificadon 3.4.2, Action a., due to the inoperability of two (2) Safety / Relief Valves.

These two (2) Safety / Relief valves were conservadvely detennined to be Inoperable per Technical Specification 4.0.3 due to the t!me period since verification of the !!ft setdngs for valves 1821-F013B and iB21 F013J. Technical Specification 4.0.3 states "Fallure to perform a Surveillance Requirement within the specified dme inte val shall constitute a failure to meet the OPERABILITY requirements for a Limited Condition for Operation." Technical Specification 4.0.5 provides survelliance requirements for Inservice inspection and testing of ASME Code Class 1, 2, and 3 components per ASME Section XI of the Bol!er and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g). The reactor coolant system Safety /Reller Valves are included in the ASME Code components to be tested. The surveillance Intervals for the Safety / Relief Valves are based upon ASME Section XI Table IWV-35101.

Safety / Relief Valves IB21 F013B and IB21-F013) exceeded the specified surveillance Interval requirements (every 5 years) and were conservadvely determined to be Inoperable. It is not possible to perform the required tesdng with Unit I in operation and the reactor coolant system pressurized.

Therefore, Commonwealth Edison requests Enforcement Discretion from Technical Specification 4.0.3 for SRVs 1B21 F0138 and IB21-F013] be granted to allow continued unit operadon until approval of an exigent Technical Specification amendment. The amendment will request that valves IB21 F013B and IB21-F013] be exempted from the requirements of Tecirlical Specification 4.0.3 until the sixth refueling outage or the first Cold Shutdown prior to LI R06, whichever comes fhst. In addidon, the LCO for Technical Specification 3.4.2 will be modified to require all 18 SRVs to be OPERABLE, with the exemption to 4.0.3 for SRVs 1B21 F013B and iB21 F013).

LaSalle County Station (LaSalle) Unit 2 has been reviewed and each SRV has been tested within the last 5 years.

2. . CIRCUMSTANCES SURROUNDING THE SITUATION LaSalle's inservice Tesdng Program is based on ASME Secdon XI 1980 Edidon, Winter 1980 Addenda. Subsection IWV, Table IWV-35101 of Section XI requires

...that at each refueling all valves which have not been tested during the preceding 5 year period shall be tested." LaSalle Station's Engineering Department discovered that two (2) Unit 1 Main Steam Safety /Rellef Valves did not meet this requirement coming out of the unit's fifth refuel outage in January of 1993. The two SRVs,1B21 F013B l

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and IB21-F013), were last setpoint tested during the first Unit I refuel outage (LIR01), which ended in October,1986. Thus, the 2 valves have not been setpoint tested in approximately 7 years. SRVs IB21-F013B and IB21-F013) are presently scheduled to be tested during the sixth Unit I refuel (L1R06) In March 1994.

During previous cycles, it was LaSalle's Interpretation that each SRV required testing once within each fixed five (5) year period. Each SRV was seTolnt tested during the first fixed 5 year period, Januay 1984 to Januaw 1990. (This includes time from the end of the 5 year interval until the next scheduled Refue!.) The end of the third Unit 1 refuel (L1 R03) In January 1990 marted the start of tip second fixed 5 year period. LaSalle is presently on schedule to have each SRV setpoint ced again by the end of this period. The testing above meets LaS2!!a's odginal interpretation.

Recent review of Table IWV 3510-1 has rendered a different Interpretation.

This Interpretation concludes that when coming out of a refueling outage, each installed SRV must have been tested within the previous 5 Year span, Irregardless of the particular fixed 5 year period for the unit. Because the test sequencing was different within the two fixed 5 year periods, the most recent setpoint test of SRVs IB21 F013B and IB21 F013] do not fall wid11a the 5 year envelope preceding the 11rdi Unit I refuel (LI R05).

3. EVAlllATION OF SAFETY SIGNIFICANCE AND CONSEOUENCES The ponible effet.ta of the safety setpoint drift are:

1) Setpoint drift HIGH (delayed valve opening in the safety mode).

2) Sctpoint ddft LOW (carly safety mode opening).

EVALUAllON 01- CASE 1) SEIPOINI DRIFT HIGH:

The bounding transients for LaSalle Stadon were reviewed by General Electric.

The review showed that the limiting MCPR transient (Turbine Load Reject without Bypass, or LRNBP) is not affected because the minimum MCPR is reached prior to reactor pressure readling the lowest SRV Relief Setpoint (1076 psig). The affected SRVs are in the two highest actuation groups and therefore cannot Impact the results of th!s event. The other (non limiting) MCPR events did not require re-evaluation for the potential of becoming limiting for the following reasom

1. The non-Ilmiting pressurization events (Turbine Trip without Bypass, Feedwater Controlier Failure withour Rypass) add reactivity (and lost MCPR marBfn) by the same physical mechanism as the LRNBP.

Therefore, these events will abo experience the minimum MCPR pilor to the first SRV tellef setpoint, and the possible Safety Mode setpo!nt drift I

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cannot affect the outcome of the event analyses.

2. Non pressurization MCPR events (Loss of Feedwater Heaters, and Rod Withdrawal error) do not result in loss of the turbine pressure regulator, and do not involve SRV operation. Therefore SRV setpoint drift cannot affect the outcome of the event analyses.

EOC RPT widi one SRV Out of Service (005) is analyzed for both LaSalle Units. Thh analysis assumes the lowest pressure SRV does not actuate, and les response is only a function of relief mode operation. The effect of safety mode setpoint difft does not affect the relief mode of operation.

The limiting pressurintion transient for vessel over pressurization Is the MSIV closure w!th flux scram (which assumes MSIV limit switch scram falls). For this event, the peak reactor pressure of 1266 psig is only s!!ghdy impacted (by less than 10 psig),

because the remaining relief capacity is adequate. The ASME pressure limit of 1375 psig for the RPV hottom head (1325 usig steam dome pressure) is not exceeded.

EVALUATION OF CASE 2) SETPOINT DRIFT LOW:

For the second affect (setpoint drift low), the possible concents are:

1) Drift to below operating pressure and initiation of stuck open SRV au,ldent, and

2) Complication of an event which utillzes safety mode.

The historical data foi Safety Mode setpoint drift was reviewed. LaSalle has 18 Crosby Safety /Reliel Valves, Style HB-654P and Size 6xRx10, Installed on each unit.

In addition to these 36 valves, LaSa!!e also has 9 spare valves of the same style and size.

The valves have safety set pressures of 1150 psig,1175 psig,1185 psig,1195 psig, or 1205 psig. Of 76 test points from the Unit I and Unit 2 SRV populations, there have been a total of 19 failures to meet the setpoint uiteita of + 1% / 3% oflift pressure.

The data is evenly distributed between the units and evenly distributed between posidve drift and negative drift data. There also does not appear to be a correlation to age or life of plant for either the frequency or magnitude of setpoint drifts. Therefore, no indicadons of a systematic drift factor are present, and the performance of the two subject valves is expected to reflect the random variations experienced in the populadon sofar.

To check for the potential for an exaggerated drift due to the extended surveillance Interval, a possible drift of twice the worst observed setpoint drift was postulated. This would be approximately 8% of setpoint, and for the SRV 1B21-s anlausta!Iemoed3.wph3 3 l

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001 504 3861:# 6/ 8 F013B, a possible drlft down from 1195 (nominal setpoint) to 1099 psig might occur.

SRV 1B21 F0131 has a higher setpoint and greater margin for downward setpoint ddft.

This is above the unit operating pressure (nominally 1005 psig), and above the lowest pressure rellef setpoint (1076) psig. Therefore, a decreasing setpoint condition that induces the Stuck Open SRV Accident is not considered credible.

The relief mode was exercised during the Unit I scram dated September 14, 1993. During the scram, reactor pressure reached approximately 1075 psig. Neither SRV IB21-F013B or IB21 Fol3J expertenced safety mode actuations at this pressure.

This verifles dat their safety mode setpoint d!d not drlft below this pressure, and supports the expectation that a stuck open SRV accident is not credible (due to the concem of safety setpoint drift).

SRV 1B21-F0138 was manually opened twice during that event and SRV 1B21-F0131 was manually opened once. Bodi operations were satisfactory, Indicadng that operation and capability of SRV IB21 F013B and IB21-F013] are not impaired by any other factors.

Four SRVs exhibited indication of actuator air leaks during this event. None of these problems included safety mode anomalles or fa!!ures, and neithei SRV 1D21-F013B or IB21-F013] exhibited problems.

The potential for comp!!cadon or changes to the progreulon of events that would not reach pressures as high as the Safety Mode setpoints of SRV IB21-F013B or 1B21-F013J 15 minimal. The primary event ofInterest is the MSIV closure w!th flux scram, which is the only event analyzed each cycle that reaches pressures above the lowest Safety setpoint group (1146 psig). MSIV closure with flux scram opens all SRVs, so carfy actuadon of IB21 F013B or IB21 F013) would lessen the severity of the piessutizadon event. If (downward) setpoint drift occurred into the region of relief valve operadons (the first group to be Impacted would be the highest relief group at 1116 psig), the SRVs would have been expected to be open in the relief mode already (the rellef mode setpoint for SRV 1821-F013B is 1106 psig, and for SRV IB21 F013]

is 1116 pstg).

No other parameters (i.e., reactivity, cooldown rates, etc) are expected to be affected by setpoint drift in the downward direction.

4. COMPENSATORY ACT10HS The condition of concem affects only the safety function setpoints of SRVs 1B21 F013B and IB21 F013), and does not involve either their relief mode operation, Indicadons, or possible systern actuation effects.

Tlie following Compensatory Acdons will be placed in effect:

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1) Licensed Operators will be trained, prior to assumlng the shift beginning December 6,1993, diat for diese .2 5RVs, safety mode actuatforu are potentially affected by setpoint drift.

2) Caution cards will be placed on the control panel, by December 6,1993, to reinforce the Information that safety actuation pressures should be verified during appropriate conditions.

5. LUSTIFICATION FOR TFiEJURATION OF THE REOUEST:

The surveillance intervals are provided to periodically verify, to the extent poss!ble, that the survellied component will perform its desired safety funcdon when required. Typically these surveillance tests verify that the component is indeed performing or capable of performing its required safety function. The failure to perform a surveillance on a component does not, in itself, make the equipment unable to perform its function. In this specific case the survell!ance requirement is to verify the calibration (safety valve funcdon lift setting) prior to startup from a refueling outage where the last setpoint test exceeds 5 years. The startup from the last outage in which the two safety /rellef valves were tested occurred in October 1986, approximately 7 years ago. The next opportunity to perform this test, which requires the reactor to be in cold shutdown, would be the spring 1994 refueling outage, L1R06, presently scheduled for March 1994 or the first cold shutdown prior to LIR06, whichever comes first. The max! mum total fength of time since the last setpoint test, therefore, will be approximately 7.5 years.

6. EVALUAIlOhLOF SIGNIFICANIHAZARDS CONSIDERATION _

Commonwealth Edison has evaluated the pronmed Technical Specification Amendment and determined that It does not represent a signll! cant hazards consideradon. Based on the ultesla for defining a significant hazards consideration established in 10 CFR 50.92, operadon of LaSalle County Stadon Unit 1 in accordance with the proposed amendment will not:

1) lavolve a significant increase in the probability or consequences of an accident previously evaluated because:

2. There is no affect on accident inldators so there is no change in probability of an accident. The probability of a failed open Safety / Relief Valve (SRV) is not affected based on observed performance of setpoint drift.

b. There is no effect or minimal affect on the t.onsequences of analyzed accidents based on an evaluatlan that the highest reactor vessel pressure that will occur is 5dil less than the Safety Limit of kAn!a\lassbe\noW3 wpf\G

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.[SENT,BY: 4-18-01 : 9:21AM : -- 301 504 3861:# 8/ 9 1325 psig steam dorne pressure, for the bounding vessel pressurtzadon event. This evaluation assumed that both SRVs 1821 F013B and 1B21 F013] fall to open.

2) Create the possibility of a new or different kind of accfdent from any accident previously evaluated because:

The SRVs are not being used in any other mode than original design. The only affect is from the safety mode setpoint drift. This issue does not involve any plant modificadons or changes to operating procedures. Therefore, this issue does not create the possibility of a new or different kind of acc! dent from any previously evaluated accident.

3) Involve a significant reducdon In the margin of safety because:

The review of previous sensitivity analyses for peak accident pressure Indicates that extension of the surveillance intervals for SRV ID21-FOl3B and IB21-F013) can not result in exceeding the Safety Limit reactor pressure of 1325 psig steam dome pressure. Therefore, this issue does not involve a significant reduction in the margin of safety.

Guidance has been provided in " Final Procedures and Standards on No Significant Haurds Consideradons," Final Rule, 51 FR 7744, for the applicadon of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards consideradons. These proposed amendments most closely fit the example of a change which may either result In some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the applicable Standard Review Plan.

This proposed :;mendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a sign!ficant relaxation of the bases for the limiting conditions for operadons. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideradon.

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7. EliYlRONMENTAL ASSESSMENI LaSalle County Station has evaluated the proposed enforcement discretion agaInst the criteria for Identification of licensing and regulatory actlons requiring environmental assessment in accordance with 10 CFR 51.20. It has been determined that the proposed changes meet die citteda for a categorical exclusion as provided under 10 CFR 51.22(c)(9). This conclusion has been determined because the changes requested do not pose significant hazards consideradons or do not involve a signtt! cant increase in the amounts, and no signiffr2nt changes In the types, of any effluents that may be released off-site. Addidonally, this request does not involve a significant increase in Individual or cumulative occupational radladon exposure.

8. APEROVAL BY ON-SITE REVIEW The request has been approved by LaSa!!c County Senior Station Management and On-Site Review (OSR) In accordance with Station procedures.

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