ML20202C382

Requests NRC Review & Approval of Encl Engineering Evaluation of Elevated Tailpipe Temperatures Associated W/ SRV 20.-3B,IAW TS 3.6.D.4

ML20202C382
Person / Time
Site:	Pilgrim
Issue date:	01/30/1998
From:	Olivier L BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BECO-2.98.012, NUDOCS 9802120229
Download: ML20202C382 (8)

Text

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o so.amsa a i Pilgrim Nuclear Power Station g , Rody Hill Road Plymouth, Massachusetts 02360 l LJ. Olivier Vice Presided Nuclear and Citation Director January 30 ,1998 BECo Ltr. 2.98.012 l U.S. Nuclear Regulatory Commission l Attention: Document Control Desk Washington, DC 20555 l

Docket No. 50-293 License No. DPR-35 Request for NRC approval of an Engineering Evaluation:

Elevated Tailpipe Temperature on Safety Relief Valve 203 38 in accordance with Pilgrim Nuclear Power Station (PNPS) Technical Specification 3.6.D.4, Boston Edison requests NRC review and approval of the attached engineering evaluation of elevated tailpipe temperatures associated with safety-relief valve (SRV) 203-3B. This evaluation 1 was reviewed by the Operational Review Committee on January 15,1998.

The PNPS Technical Specification 3.6.D.4 states:

Any safety relief valve whose discharge pipe temperature exceeds 212*F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more shall be removed at the next shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in the as found condition and recalibrated as necessary prior to reinstallation. Power operation shall not continue beyond 90 days from the initial discovery of discharge pipe temperatures in excess of 212*F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC approval of the engineering evaluation delineated in 3.6.D.3.

Technical Specification 3.6.D.3 states, in part, that an engineering eva uation shall be performed justifying continued operation for the corresponding temperature increases.

An elevated tailpipe temperature associated with safety / relief valve 203-3B (i.e.,132*F) was first observed on December 11,1997. The temperature slowly increased, and on December 24, 1997, the temperature reached and remained in excess of 212 F for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The tailpipe temperature is currently approximately 217'F where it has remained relatively stable.

SRV 203-3B is operable in its present condition. The leakage is minor in nature and it has tentatively been attributed to pilot staDe leakage.

if SRV203-3B tailpipe temperature exceeds 235'F for a period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or exceeds 2ETF at any time, then an orderly shutdown of the reactor shall commence, as recommended in the attached engineering evaluation.

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.~vv"~ Page 1 of 2 9002120229 900130 PDR ADOCK 05000293 P PDR

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Pilgrim Nucle:r Power Station Your review cnd approv;l is requested pri:r 13 M:rch 24,1998 t3 preclud3 a shutdown cf Pilgrim Station. If addition:lInformation is required, pl2cs3 contact Mr. B:b C nnon tt (508) 830-8321.

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hev L. J. Olivier RLC/deg id: reqeat/radmisc

Attachment:

Engineering Evaluation i

cc: Mr. Alan B. Wang, Project Manager Project Directorate 1-3 Mall Stop: OWFN: 14B2 Office of Nuclear Reactor Regulatic7 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region 1-475 Allendale Road Kiag of Prussia, PA 19406 Senior Resident inspector Pilgrim Nuclear Power Station Page 2 of 2

. EE98-004

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• ATTACHMENT t .

BOSTON EDISON COMPANY ENGINEERING EVALUATION L Initiating Documenti PR97.3565

2. Affected (System. Subsystem. Train. ComDonent. or Device)

Target Rock Coriv .on Two-Stage Safety Relief Valve 203-3B

3. Specified Functions of the AtTected Item The safety relief valves are part of the reactor coolant pressure boundary and operate by power actuation (i.e., automatic depressurization system) or self-actuation by high process pressure. The safety relief valves limit peak vessel pressure during overpressure transients to satisfy ASME code requirements. The postulated transients for which safety / relief valve actuation is required are given in Chapter 14 and in Appendices R and Q of the FSAR.' The automatic depresrurization system provides a means to rapidly depressurize the primary system down to a pressure at which low pressure cooling systems can provide makeup. In the event of a small or medium break LOCA, this function would be required if high pressure ECCS was unable to maintain vessel water hvel.~

4. Enferences

1. Technical Specifications and associated bases 3.6.D.1,3.6.D.2,3.6.D.3,3.6.D.4 and 3.6.D.5.

2. General Electric Report NSE 13-0282, " Pilgrim Plant, SRV Tailpipe Steam Temperature Correlation for SRV Lee.kage Monitoring System," dated February 1982.

3. General Electric Report NEDE-30476, "Setpoint Drift Investigation of Target Rock Two-Stage Safety / Relief Valve (Final Report)," dated February 1984.

4. Operability Evaluation for Target Rock Corporation Two-Stage Safety Relief Valve 203-3D, dated 8/17/91.

5. NRC approval of operability evaluation for 203-3D, incoming NRC letter 1.91.288 dated 10/24/91.

6. TCH92-133," Root Cause/ Corrective Action Response for 203-3D" (PR92.0338/F&MR 91-373).

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7. Operability Evaluation for Target Rock Corporation Two-Stage Safety Relief

, , Valve 203-3 A dated 11/2/93.

8. Wyle Lab, Test Report No. 41211-0 dated,4/25/91,

9. Supplemental Reload I.icensing Report for PNPS Reload 11, Cycle 11, PDC 96-g 17," Reload 11/ Cycle 12 Core Design"

10. MR #19703052

11. Operability Evaluation for Target Rock Corporation Two-Stage Safety Relief Valve 203-3B, pilot serial n.imber 1025, dated 2/9/96.

E 12. Operability Evaluation for Target Rock Corporation Two Stage Safety Relief Valve 203-3D pilot serial number 1054, dated 11/4/97.

13. NEDO-22159 General Electric Boiling Water heactor Increased SRV S;mmer Margin Analysis for PNPS Unitl-June 1982.

5. Operability Concerl

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Safety Relief Valve (SRV) 203-3B is leaking. This condition was detected by tailpip-temperature monitoring instrumentation on 12/11/97 and documented in Problem Report 97-3565. The SRV 203-3B tailpipe temperature trended to 132 'F on 12/11/97,145 *F on 12/16/97,212 'F on 2/24/97 and stabilized at 217 *F on 1/2D8 This temperatuie profile supports a condition indicative of pilot stage leakage.

Technical Specification 3.6.D.3 requires an engineering evaluation to support continued operation if the temperature of any safety relief valve discharge pipe exceeds 212'F for a period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during normal reactor power operation (Ref.1). The Technical Specification Bases states that minimal les.kage exists when tht tailpipe temperature is 215 F, and therefore, a conservative temperature of 212 F was chosen.

6. Onera'oility Recommendation (Check one)

@ Operable -

O Inoperable Page 2

. EE98-004 7,. Basis for Recommendadon (Use additional sheets as necessary)

The most likely leakage paths through the Target Rock Corporation (TRC) two-stage safety relief valve are: (1) through the main stage, past the main disc and seat interface, or (2) through the pilot stage, past the disc and seat interface. General Electric (GE) and TRC (the valve's manufacturer) representatives have in the past indicated main stage leakage is typically substantial and increases faster than pilot stage leakage and pilot stage leakage is mon. on than main disc leakage.

SRV 203 3B pilot 1025 had high tailpipe temperatures between 1/24/96 and 2/6/96.

The operability evaluation determined the elevated temperature which eventually settled out at 217 'F was the result of minor pilot leakage. This was later confirmed at Wyle Labs where diagnostic "r .found" setpoint and leakage tests were performed.

SRV 203-3D pilot 1054 and main stage 008 were replaced on 12/3/97 due to high tailpipe temperature. Although diagnostic testing is incomplete, a leaking pilot valve is suspected.

Due to the similarity of the increase in tailpipe temperatures to SRV-3B pilot 1025 leakage, the most probable cause for the leakage presently experienced by SRV-3B is pilot leakage. This condition may clear with a lowering of reactor pressure; however, it will more likely remain at some low level (218-220 F ), for a period of months.

The consequences ofleakage across either the pilot or main stage boundary for SRV 203-3B must be addressed, since leakage increases may occur later and may occur at either location. Pilot stage leakage affects valve lift set point and response time while main stage leakage loes not.

Pilot Stage Leakage Pilot stage leakage can affect the performance of the two stage Target Rock SRV in the pressure actuated mode (i.e., safety mode). The effects of leakage on valve performance have been extensively studied and basically consist of the following: (1) setpoint drift, (2) response time changes (Ref. 3).

The leakage rates studied by GE and TRC range from 200 lbs/hr to 1000 lbs/hr. Test results indicate that set point prespire increased to approximately 1% at 225 lbs/hr and to 2% at a leakage rate of approximately 400 lbs/hr. The setpoint then decreased 2% per 100 lbs/hr of additio: alleakage. The eiTect ofleakage rate on setpoint is illustrated in Reference 3. Based on TRC test results, pilot stage leakage up to 1000 lbs/hr did not significantly affect the SRV setpoint (Ref. 3).

Response time is the interval from pilot actuation to main disc lift. The normal response time for a two stage TRC SRV is approximately 0.4 seconds. Response time varies with leakage rate A slower response time results in a higher peak reactor vessel pressure t . iring the safety mode, and a faster response time results in a lower peak reactor pressure. A slower response time also Page 3

a EE98-004 '

, results from a higher tailpipe temperature (i.e., higher steam leakage). The impact ofleakage on

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- response time is presented in the plant analysis section of this report.

Main Stage Leakage

' Main stage leakage is an uncommon problem in the industry according to TRC. This view is -

substantiated by the volumes ofinformation available on relief valve ler.kage, all of which is a result of pilot stage leakage, Leakage across the main stage boundary is an economic concern -

because of the rotential for seat and/or disc dam:_ge. TRC and GE adviw that leakage across the main disc will nat affect the ability of the SRV to operate in either the pressure actuated or power-actuated modea. . Leakage across the main stage should not cause the SRV to inadverte :tly open and cause a rapid depressurization or fail to reclose aRer operating.

Plant Analyss General Electric has performed sensitivity analyses on PNPS showing that even if driR results in an opening pressure 10% above the nominal setpoint for all Pilgrim's installed SRVs, the peak -

pressure for the MSIV closure-flux scram event is less than the upset limit of 1375 psig. Based . i on these results, GE concludes that BWRs with TRC two-stage SRVs can tolerate driRs significantly above the 1% technical specification setpoint tolerance.

'Also, the peak vessel pNssure would increase by 5 psig if one SRV experienced the leakage L induced maximum response time delay of 0.9 second (Ref. 3). This is much less than the 77 psig margin between reactor vessel pressure for cycle 12 and the upset limit of 1375 psig (Ref. 9).

The impact of either a delay in SRV response time or an increase in SRV opening pressure on critical power thermal margin is minimal. This is due to the rapid insertion oflarge negative-control reactivity during transients before the higher pressure can contribute to any significant _-

Ladditional core power production due to core void collapse. This was demonstrated in NEDO-22159 (Ref.13 ), where a 30 psig increase in SRV opening setpoint resulted in only 0.1% _ j increase in peak fuel rod heat flux following a limiting pressurization event. This was specifically

- evaluated for PNPS for cycle 6f However, it would also apply to cycles 7-12 due to the -

insignificant contribution of SRV pressure relief to the mitigation of the core power excursion -

associated with limiting pressurization events. Reactivity shutdown via reactor scram renders the core essentially subcritical before SRV pressure relief can be effective in moderating the void

- collapse due to the pressurization event.

SRV Leakane Versus Tailoipe Temocrature and SRV Setocint

, L The maximum allowable SRV 203-3B tailpipe temperature of approximately 255'F can be correlated to a ceam leakage flow rate of approximately 225 lbs/hr, while steam leakage of 1000 lbs/hr corresponds to a tail pipe temperature of approximately 275 F. It is acceptable to continue operation with a tailpipe temperature ofless than or equal to 255"F since test data has demonstrated that the possible relief valve setpoint drin at this temperature is equivalent to +1%

(Ref. 3).

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_ , f .; EE98 004 Riant Parameter Effects on Tailpine Temocrature

- Drywell temperature: Sensitivity analysis predicts that the tailpipe temperature is relatively in-sensitive to drywell temperature variations over the entire range of steam leakage (Ref. 2).-

Reactor pressure: The temperature of the steam at the exit of the relief valve decreases as reactor pressure increases. Any effect on downstream tailpipe temperature may be offset by increased leakage s ates at higher rmi tor pressure.l The temperature limit of 255'F was based on normal reactor operating pressure for the exit steam (Ref. 2).

- Containment pressure: The safety relief valve tail pipe is equipped with vacuum breakers that prevent drawing a column of torus water into the tailpipe. The tailpipe will be at atmospheric .

pressure prior to inetiing and slightly above atmospheric pressure after inerting the containment.

The effects of containment pressure on tailpipe temperature are negligible because the differeace in tailpipe pressure due to inerting is only a few psig. Also the maximum leakage flow rate of up to 1000 lbs/hr will na be sufficient to pressurize the tailpipe, thereby affectiag temperature (Ref. 2). Therefore, containment pressure effects are judged to be negligible.

Eauioment ouamation Each safety relief valve has one solenoid valve which is attached to a manifold mounted on the air

. operator for the valve. The leakage flow through the safety relief valve will raise the temperature -

of the main valve body, base, pilot assembly and associated tailpipe. The solenoid ' valve is environmentally qualified, considering in-part the normal ambient temperature to which it is exposed. The solenoid valve is not in direct contact with any part of the safety relief valve which will experience appreciable elevated temperature because of the leakage through the valve..

Therefore, the solenoid valve will not be exposed to any significant amount of conducted heat but could be exposed to a slightly higher ambient temperature.- The solenoid valve is mounted _as an appendage off the safety relief valve in a configuration that maximizes air circulation around it and -

minimizes the ambient temperature to which the solenoid ' valve is exposed. Therefore, the effects of minor leakage through the SRV 203-3B safety relief valve are judged to hr.ve no appreciable affect on the enviranmental qualification of the safety relief valve solenoid.

CQElusi9a SRV 203-3B is operable in its present condition. The leakage that has occurred is of a minor i nature attributed to pilot stage leakage and is acceptable as discussed previously. Either -

_ _ intermittent or continuous leakage within the limits described below is acceptable for continued operation.. Tests and analyses have shown that ieakage rates of approximately 225 lbs/hr (equivalent to 255'F) should not impact the SRV setpoint by more than +1%.

Based on past experience with leeking pilot valv's, a slightly lower action limit has been selected for this SRV in order to assure reliable SRV operation and reduce damage to the pilot seat and Page 5

,. EE98-004 disc. If the tailpipe temperature exceeds 235'F for a period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or exceeds 250*F at any time, an orderly shutdown of the reactor shall commence.

8.Comnensatory Measures / Conditions Required Justification for Continued Operation Technical specification surveillance 4.6.D.3 requires that SRV tailpipe temperature be logged

] daily. This surveillance shall be performed at an increased frequency of once per hour, to compensate for the reduced margin between the normal maximum tailpipe temperature of 212 F and 235'F.

Temporary Modification 97-66 has been installed to reconfigure the alarm circuitry on safety relief valve temperature recorder TR260-20. The alarm point of RV-203-3B which normally annunciates at 200*F has been reset to annunciate at 230 F.

If SRV 203-3B tailpipe temperature exceeds 235'F for a period greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or exceeds 250'F at any time, then an orderly shutdown of the reactor shall commence.

This relatively small leakage is not expected to cause torus water temperature or drywell temperature to change significantly; therefore, an increased surveillance interval for these parameters is not warranted.

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