ML042940375

Request for NRC Approval of Engineering Evaluation for Elevated Discharge Pipe Temperature of Safety Relief Valve 203-3C

ML042940375
Person / Time
Site:	Pilgrim
Issue date:	10/12/2004
From:	Balduzzi M Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.04.095
Download: ML042940375 (20)

Text

E ntergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Michael A. Balduzzi Site Vice President October 12, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No.: 50-293 License No.: DPR-35 Request for NRC Approval of Engineering Evaluation for Elevated Discharge Pipe Temperature of Safety Relief Valve 203-3C LETTER NUMBER: 2.04.095

Dear Sir or Madam:

In accordance with 10 CFR 50.90 and Pilgrim Station Technical Specification 3.6.D.4, Entergy Nuclear Operations, Inc. (Entergy) requests NRC approval of the engineering evaluation of elevated discharge pipe temperature of main steam safety relief valve RV-203-3C (SRV-3C).

NRC approval is requested by January 4, 2005 to avoid a shutdown of Pilgrim Station. provides the evaluation of this request. Enclosure 2 identifies the commitments contained in this letter.

Please feel free to contact Bryan Ford, (508) 830-8403, if you have any questions or require additional information.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the /-2 day of October 2004.

Sincerely, Michael A. Balduzzi DWE/dm : Evaluation of Request (16 pages) : Commitments (1 page)

Entergy Nuclear Operations, Iric. Letter Number: 2.04.095 Pilgrim Nuclear Power Station Page 2 cc: Mr. Lee Licata, Project Manager Office of Nuclear Reactor Regulation Mail Stop: 0-8B-1 U.S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19408 Senior NRC Resident Inspector Pilgrim Nuclear Power Station Mr. Robert Walker Radiation Control Program Commonwealth of Massachusetts Executive Offices of Health and Human Services 174 Portland Street Boston, MA 02114 Ms. Cristine McCoombs, Director Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702

ENCLOSURE 1 Evaluation of Request

1. DESCRIPTION

2. REQUESTED APPROVAL

3. TECHNICAL ANALYSIS

4. REGULATORY ANALYSIS 4.1 No Significant Hazards Consideration

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

Evaluation of Request

1. DESCRIPTION The temperature indicator for one main steam safety relief valve is indicating increased temperatures. The instrument is associated with the discharge pipe (tailpipe) for safety relief valve RV-203-3C (SRV-3C).

Pilgrim Nuclear Power Station Technical Specifications Section 3.6.D.3 states:

If the temperature of any safety relief valve discharge pipe exceeds 212 0 F during normal reactor power operation for a period of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an engineering evaluation shall be performed justifying continued operation for the corresponding temperature increases.

Pilgrim Nuclear Power Station Technical Specifications Section 3.6.D.4 states:

Power operation shall not continue beyond 90 days from the initial discovery of discharge pipe temperatures in excess of 212OF 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.

The discharge pipe temperature for SRV-3C exceeded 2120 F at 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on October 6, 2004 and by 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on October 7, 2004, the discharge pipe temperature had exceeded 212 0 F for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. REQUESTED APPROVAL Pilgrim requests NRC approval of this evaluation that it is safe to continue to operate with the leakage past SRV-3C as indicated by the discharge pipe temperatures greater than 212 0F. This approval is requested in accordance with 10 CFR 50.90 and Technical Specification 3.6.D.4, and is requested by January 4, 2005 to avoid a shutdown of Pilgrim Station.

Pilgrim commits to enforce the following limits on continued plant operation with higher discharge pipe temperatures for SRV-3C.

1. Ifthe discharge pipe temperature for the SRV exceeds 2350 F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> an orderly shutdown of the reactor shall commence and the reactor pressure shall be less than 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In addition, ifthe discharge pipe temperature for the SRV exceeds 2500 F an orderly shutdown of the reactor shall commence and the reactor pressure shall be less than 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Page I of 16

vv - -v td S s- 2*

2. Technical Specification surveillance 4.6.D.3 requires that SRV discharge pipe temperature be logged daily. This surveillance shall be performed at an increased frequency of obce per hour while the discharge pipe temperature is greater than 212 0F for SRV-3C.

3. TECHNICAL ANALYSIS

1. Problem Statement Safety Relief Valve SRV-3C is believed to be leaking. This condition was detected by discharge pipe temperature monitoring instrumentation on SRV-3C. The current temperature profile supports a condition indicative of pilot stage leakage.

This condition was detected by routine review of SRV discharge pipe temperature monitoring instrumentation on October 4, 2004. The SRV-3C discharge pipe temperature began to gradually increase from about 113OF beginning on October 1,2004, and eventually exceeded 212 0F at 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on October 6, 2004 and by 0309 hours0.00358 days <br />0.0858 hours <br />5.109127e-4 weeks <br />1.175745e-4 months <br /> on October 7, 2004 had exceeded 212 0 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 temperature was about 2130 F when this proposed change was prepared and based on past experience, is expected to stabilize between 215 0F to 220 0F. The current temperature 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 0F 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 (Reference A.1). The Technical Specification Bases states that minimal leakage exists when the discharge pipe temperature is 2150 F and therefore, a conservative temperature limit of 2120 F was chosen.

2. Assumptions Following an evaluation of the SRV-3C discharge pipe temperature profile it has been determined that the leakage is most likely due to pilot stage (pilot seat area) leakage, as explained in the following engineering evaluation.

3. Engineering Evaluation The safety relief valves (SRVs) are part of the reactor coolant pressure boundary and operate by power actuation (i.e., automatic depressurization system) or self-actuation by process high pressure. The SRVs limit peak vessel pressure during overpressure transients to satisfy ASME code Page 2 of 16

requirementsW The postulated transients for which safety/relief valve actuation is required are described in Chapters 4 and 14 and in Appendices Q and R of the Updated Final Safety Analysis Report (References B.1, B.2, B.3, and B.4). The automatic depressurization system (ADS) 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 Loss of Coolant Accident (LOCA), the ADS function would be required if the High Pressure Coolant Injection (HPCI) system is unable to maintain vessel water level.

The most likely leakage paths through Target Rock Corporation (TRC) two-stage SRVs 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.

Representatives of General Electric (GE) and TRC (the SRVs' manufacturer) have in the past indicated that main stage leakage is typically substantial and increases faster than pilot stage leakage, and that pilot stage leakage is more common than main disc leakage.

History Elevated SRV discharge pipe temperatures have been experienced previously at Pilgrim for which the required engineering evaluation(s) was completed and request(s) for approval was submitted to the NRC (References E.1, E.2, and E.3). After review and evaluation of each request, the request was approved by the NRC (References F.1, F.2, and F.3).

Inthe late 2003 - early 2004 timeframe, elevated temperatures were experienced on the discharge pipes for SRV-3A and SRV-3D for which a required engineering evaluation was completed and a request for approval was submitted to the NRC. Additional information related to the request was also submitted to the NRC. The request was later withdrawn before the requested approval because these SRVs were repaired during the March 22, 2004 outage. (Reference E.4). The NRC subsequently acknowledged the withdrawal of the request (Reference F.4).

The pilot stage (serial number 1048) currently installed in SRV-3C was installed during the 2003 refueling outage. The elevated SRV-3C discharge pipe temperature was evaluated for operability (Reference G.1) on October 8, 2004.

Page 3 of 16

. -I .I - . ,,4 - , , ; -

Correlation of Current Conditions to Historidal Experience Due to the similarity of the increase in SRV-3C discharge pipe temperatures to previous leaking pilot stages, the most probable cause for the leakage presently experienced by SRV-3C is pilot stage leakage. This condition may clear with a lowering of reactor pressure. However, it will most likely remain in the 215 to 230 OF ranges from now until the next refueling outage, in the spring of 2005.

The consequences of leakage across either the pilot stage or main stage boundary for SRV-3C must be addressed, since leakage increases may occur later. Pilot stage leakage affects valve lift setpoint and response time while main stage leakage does not. In either case neither the reactor dome safety limit (1325 psig), ASME code allowable for Upset conditions (1375 psig), or ASME code allowable for Emergency conditions (1500 psig) will be exceeded due to the SRV leakage bounded by this evaluation.

Pilot Stage Leakage Pilot stage leakage can affect the performance of a 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 consist of setpoint drift and response time changes (Reference C.4).

Leakage rates studied by GE and TRC range from 200 Ibm/hr to 1000 Ibm/hr. Test results indicate that setpoint pressure increased by approximately 1% at a leakage rate of approximately 225 Ibm/hr and by 2% at a leakage rate of approximately 400 Ibm/hr. The setpoint then decreased 2% per 100 Ibm/hr of additional leakage. The effect of leakage rate on setpoint is illustrated in Reference C.4. Based on TRC test results, pilot stage leakage up to 1000 Ibm/hr does not significantly affect the SRV setpoint (Reference C.4).

Response time is the interval from pilot stage actuation to main stage disc lift. The normal response time for a two-stage TRC SRV is approximately 0.4 seconds. The response time varies with leakage rate. A slower response time results in a higher peak reactor vessel pressure during the safety mode, and a faster response time results in a lower peak reactor vessel pressure. A slower response time results when discharge pipe temperature increases. The impact of leakage on response time is presented in the "Impact on Nuclear Safety" section of this report.

Page 4 of 16

Main Stage Leakage According to Target Rock, main stage leakage is an uncommon problem in the industry. This is substantiated by information available on relief valve leakage, most of which is a result of pilot stage leakage. Leakage across the main stage boundary is an economic concern because of the potential for seat and/or disc damage. TRC and GE advise that leakage across the main stage disc will not affect the ability of the SRV to operate in either the pressure actuated, or power actuated modes. Leakage across the main stage should not cause the SRV to inadvertently open and cause a rapid depressurization or fail to reclose after operating.

Impact on Nuclear Safety:

Overpressure Protection - MSIV Closure with Flux Scram Event The capacity of the relief and safety valves shall be sufficient to prevent the reactor pressure from exceeding the allowable overpressure of the ASME Code,Section III, during a main steam isolation valve closure with indirect scram on neutron flux. The ASME Code allowable pressure is the Upset limit of 1375 psig which is 10% above reactor vessel design pressure of 1250 psig.

Neither the reactor dome safety limit of 1325 psig nor the ASME code allowable of 1375 psig for peak vessel pressure are exceeded following a MSIV closure with flux scram. This conclusion takes into consideration the effects of the following variables that affect the overpressure protection analysis results performed for the current cycle.

• A 10% increase in the nominal setpoint of each of the four SRVs results in a peak pressure increase of 30 psig (Reference C.4).

This pressure difference is based on sensitivity analysis performed on PNPS using the standard reload licensing model and assumptions with varying SRV setpoints or response time (Reference C.4). Although the sensitivity analysis model is not entirely consistent with the current Pilgrim design, the analysis is believed to provide a valid prediction of the relative change in pressure from incremental setpoint drift.

Assuming a 1% drift of each SRV, the setpoint would equal the Technical Specification limit of 1126 psig (1115 psig plus 1%). At the Technical Specification limit, the sensitivity analysis resulted in a peak vessel pressure of 1330 psig. Assuming a 10% setpoint drift of each of the SRVs, the sensitivity analysis resulted in a peak vessel pressure of 1360 psig, a difference of 30 psi.

Page 5 of 16

-I The effect of a response time delay of 0.9 s6conds is 5 psig (Reference C.4).

Testing performed by Target Rock Corporation showed that gross leakage may increase the delay time for pilot stage actuation by a maximum of 0.5 seconds relative to standard reload licensing analysis which assumes a delay time of 0.4 seconds. Therefore, the delay time after gross leakage has occurred could be as high as 0.9 seconds (Reference C.4). The engineering evaluation considered SRV-3C is delayed by 0.9 seconds. The sensitivity analysis discussed in the preceding (first) bullet was used to estimate the effect of a 0.9 second delay time on each of the four SRVs. The analysis of four delayed SRVs resulted in a 21 psi increase in the peak vessel pressure (Reference C.4). The 21 psi increase was reduced to 5 psi to reflect a delay time of 0.9 seconds for one SRV. The 21 psi increase is conservative because it occurs at a leakage rate of 300 Ibm/hr. The steady state discharge pipe temperature limit of 2350 F corresponds to a pilot leakage limit of 75 Ibm/hr. At that steady state leakage rate, Target Rock Corporation test results indicate a potential delay time of 0.55 seconds, which is only 0.15 seconds higher than the standard value used in analysis of 0.4 seconds (Figure 32). Testing performed on PNPS pilot stages that had previously leaked and were operated within the proposed limits resulted in a statistical pilot stage delay time equal to 0.5 seconds, only slightly above the analysis value of 0.4 seconds. As-found stroke time testing performed on SRVs indicates that the main stage disc stroke time is typically 0.025 seconds as compared to the analysis input of 0.15 seconds which provides a margin of 0.125 seconds in overall valve response which is applicable to each of the four SRVs. Thus, the 0.1 second increase (difference of the statistical delay time and analysis value) in pilot stage operation is completely offset by the observed 0.125 second decrease in main stage operating time, and the total valve opening time is less than that assumed in the analysis. While test results indicate the potential effects of leakage on valve response time are relatively minor and could justifiably be neglected, the consideration of an increased delay time is conservative and provides a bounding estimate of the peak vessel pressure.

The calculated peak vessel pressure of 1305 psig for the Cycle 15 MSIV closure with flux scram event did not take credit for the currently installed SRVs that have a 5.125 inch diameter throat. Instead the SRV throat sizes used ranged between 4.905 to 5.030 inches (Reference D.1). The SRVs modeled in the Cycle 15 analysis (smaller throats) have a combined average rated steam flow of -800,000 Ibm/hr versus 870,000 Ibm/hr for the larger throat now installed in the SRVs.

Page 6 of 16

The smaller throat sizes used in the Cycle 15 reload licensing analysis represent conservative inputs that increase the predicted peak reactor pressure during pressurization transients. The installed larger throat size SRVs provide greater than a 7% increase in relief capacity and will lower the predicted peak reactor pressure during pressurization transients. The installed larger throat size SRVs will reduce the peak vessel pressure provided the valves are near full open prior to the point in time that the peak vessel pressure is reached.

Figure 32 (next page) from the MSIV Closure (Flux Scram) transient analysis results shows that all four SRVs are near full open at time (t) = -2 seconds. The peak vessel pressure is reached at t = -5 seconds (Reference C.6). Therefore, the larger throat SRVs currently installed, but not modeled or reflected in the Cycle 15 MSIV Closure (Flux Scram) results, will effectively dampen the vessel pressure response. This conservatism is estimated to reduce the maximum reactor vessel pressure by approximately 5 psig.

The peak vessel pressure result of 1305 psig is based on the following tabulation:

1085 psig Initial dome pressure (Reference C.6)

+ 18 Dsia Difference between peak vessel and dome pressure 1103 psig Initial maximum vessel pressure

+202 psia Vessel pressure rise during transient (Figure 32) 1305 psig Peak vessel pressure Cycle 15 MSIV Closure (Flux Scram)

(References C.6 & D.1)

The effects from SRV pilot leakage and throat size changes described above are combined as tabulated below to develop a conservative estimate of the peak vessel and dome pressures for the limiting MSIV Closure (Flux Scram) event:

1305 psig Peak vessel pressure for Cycle 15, MSIV Closure (Flux Scram) assuming a 1%increase in nominal setpoint on all four SRVs and small SRV throat sizes used in previous cycles (References C.6 & D.1)

+ 5 psig Opening delay of 0.9 seconds for 1 SRV (Reference C.4)

+ 30 psig 10% increase in nominal setpoint of 4 SRVs (Reference C.4)

(-) 5 psig Est. peak pressure reduction from large SRV throat size 1335 psig Peak vessel pressure < 1375 psig (1250 x 1.1)

(-)18 psia Difference between peak vessel and dome pressures 1317 psig Vessel dome pressure < Technical Specification Safety Limit of 1325 psig Page 7 of 16

The peak vessel pressure is estimated at 1335 psig which is significantly below the ASME code allowable of 1375 psig that is allowed to exceed the vessel design pressure of 1250 psig by 10% (i.e., 1.1 x 1250 psig = 1375 psig). The corresponding peak dome pressure of 1317 psig is less than the Technical Specification Safety Limit of 1325 psig.

Excerpt from SRLR for PNPS Cycle 15 Figure 32 Plant Response to MSIV Closure (Flux Scram) la an 4.0 Tirff (sac)

Notes:

1. The initial reactor vessel dome pressure used in the analysis shown in Figure 32 was 1085 psig that corresponds to the high-pressure scram analytical limit.

2. Inthis figure the Y-axis is a dual scale. The Y-axis is in the units of psi for the reactor vessel pressure rise data and % rated for valve flow. 0.0 on the Y-axis in this figure corresponds to the initial peak vessel pressure of 1103 psig as tabulated above.

Page 8 of 16

Overpressure Protection Analysis - ATWS The four installed two-stage SRVs and two spring operated safety valves protect the reactor coolant pressure boundary from exceeding the ASME Level C limit of 1500 psig during a full power ATWS. The limiting ATWS is a pressure regulator failure that causes the turbine control and bypass valves to fail open, leading to vessel depressurization and MSIV closure on low steamline pressure. This event was re-analyzed for the recent Thermal Power Optimization (TPO) power uprate from 1998 MWth to 2028 MWth as documented in Reference C.5. This reanalysis credited the installed throat sizes of 5.125 inches and assumed one SRV experienced a 21 psig setpoint drift (equivalent to 1.8%) that caused the valve to open at 1136 psig (SRV nameplate pressure of 1115 psig + 21 psi drift). The remaining three SRVs were assumed to open at 1126 psig (1115 psig + 1% (11 psi)). The calculated peak vessel pressure for the limiting ATWS is 1495 psig which meets the acceptance criteria of 1500 psig. The selected discharge pipe temperature limit is chosen to limit setpoint drift to less than 1%, and therefore, the ATWS analysis described above bounds the conditions permitted by the operability evaluation and ensures the ASME Level C limit is not exceeded during the worst-case ATWS.

The ATWS analysis performed for the TPO power uprate to determine peak vessel pressure assumed an SRV opening time delay of 0.4 seconds and a main disc stroke time of 0.15 seconds for a total delay between the start of pilot motion and completion of main disc motion of 0.55 seconds.

These values are the standard for reload licensing analysis and no additional delays were assumed. Other conservatisms in the analysis justify neglecting the potential pilot stage actuation time delay increase of 0.15 seconds described in the preceding section on the MSIV Closure with Flux Scram Event. Examples of these conservatisms are margin between the assumed setpoint for the SRVs and the analysis input value, margin between the typical SRV opening time and the analysis input value, and margin in the rate of pressure rise as compared to the rate used to measure the increase in delay on leaking SRV pilot stages. In the ATWS analysis, the setpoint for three SRVs is assumed to have drifted upward 1% (to 1126 psig) and for conservatism the setpoint of the fourth SRV was assumed to have drifted upward 1.8% (to 1136 psig). The steady state operational discharge pipe temperature limits are designed to limit the setpoint drift to less than 1% on the leaking SRV(s). Three SRVs remain leak tight and are unaffected by leakage. As discussed in the preceding section on the MSIV Closure with Flux Scram event, the observed 0.1 Page 9 of 16

second respons-e time increase in pilot stage bperation is completely offset by the 0.125 second decrease in main stage operating time and the total valve opening time is less than assumed in the analysis. The maximum response time delay of 0.9 seconds is associated with a leakage rate of 300 Ibm/hr and a 0.55 second delay is associated with a 75 Ibm/hr leakage rate. Each of these delay times is derived from testing that used a 60 psi/second rate of pressure rise or ramp rate (Reference C.4, Section 3.2). The ATWS analysis results indicate a higher ramp rate of approximately 90 psi/second (Reference C.5). This higher ramp rate provides more motive force to open the pilot in a shorter period of time, thus reducing the delay time. As stated in the preceding section on the MSIV Closure with Flux Scram event, for conservatism, a pressure increase for an increased response time delay was included in the assessment of leakage on the overpressure protection transient (MSIV closure flux scram event). Although the ATWS analysis does not include this potential response time delay, test results and conservatism in other inputs provides an adequate level of overall conservatism in this analysis.

On this basis, the ATWS analysis performed for the TPO power uprate includes adequate conservatism with regard to SRV performance characteristics and peak vessel pressure prediction.

Thermal Limits The impact on critical power thermal margin is minimal for either a delay in SRV response time or an increase in SRV opening pressure. This is due to rapid insertion of large negative control reactivity during transients before the higher pressure can contribute to any significant additional core power production from core void collapse. This was demonstrated in NEDO-22159 (Reference C.3) where a 30 psig increase in SRV opening setpoint resulted in only 0.1% increase in peak fuel rod heat flux following a limiting pressurization event. This was specifically evaluated for PNPS for fuel cycle 6. It also applies to the current fuel cycle due to the insignificant contribution of SRV pressure relief to the mitigation of the core power excursion associated with the limiting pressurization events on which the operating limit minimum core power ratio is based. 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.

Primary Containment Parameters The effect of the leakage on Torus water temperature is expected to be insignificant because the evaluated leakage is so small relative to the mass of the Torus water volume. Similarly, any effects on Drywell air temperature or containment pressure are also expected to be insignificant.

Page 10 of 16

No other systems are expected to be significantly impacted by this amount of leakage. The total leakage through the SRV is limited to 75 Ibm/hr.

This amount of leakage adds a minor amount 6f heat to the Torus water during power operations. The Torus water temperature is limited to a maximum of 800F and is routinely maintained below this limit using the residual heat removal system in the suppression pool cooling mode. The ATWS analysis contained in NEDE-24223 (Reference C.1) is the generic analysis performed for a representative BWR 3 with a Mark I containment.

The containment analysis contained in NEDE-24223 is considered the licensing analysis for PNPS with respect to the ATWS Rule, 1COR 50.62.

The initial suppression pool temperature used in NEDE-24223 is 900F.

Therefore, because PNPS maintains the suppression pool water temperature well below 900F, there is no impact on Torus water temperature from the standpoint of ATWS requirements.

Environmental Qualification Each SRV has one solenoid valve, which is attached to a manifold mounted on' the air operator for the SRV. The leakage flow through the SRV will raise the temperature of the main valve body, base, pilot stage and associated discharge pipe. The solenoid valve is environmentally qualified, which considers in part the normal ambient temperature to which it is exposed. The solenoid valve is not in direct contact with any part of the SRV, 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 SRV 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-3C is judged to have no appreciable affect on the environmental qualification of the SRV solenoid valve.

SRV Leakage Versus Discharge Pipe Temperature and SRV Setpoint An SRV discharge pipe temperature of approximately 2550 F can be correlated to a steam leakage flow rate of approximately 225 Ibm/hr, while steam leakage of 1000 Ibm/hr can be correlated to a discharge pipe temperature of approximately 275 0F. It is acceptable to continue operation with a discharge pipe temperature of less than or equal to 2550 F since test data has demonstrated that the possible relief valve setpoint drift at this temperature is equivalent to +1% (Reference C.4).

Page 11 of 16

Plant Parameter Effects on Discharge Pipe Temperature Drywell Temperature: Sensitivity analysis predicts that SRV discharge pipe temperature is relatively insensitive to Drywell temperature variations over the entire range of steam leakage (Reference C.2).

Reactor Pressure: The temperature of the steam at the exit of an SRV decreases as reactor pressure increases. Any effect on downstream discharge pipe temperature may be offset by increased leakage rates at higher reactor pressure. The temperature limit of 2551F in the preceding section was based on normal reactor operating pressure for the exit steam (Reference C.2).

Containment Pressure: The SRV discharge pipe is equipped with vacuum breakers that prevent drawing a column of Torus water into the discharge pipe. The discharge pipe is at atmospheric pressure prior to inerting and slightly above atmospheric pressure after inerting the containment. The effects of containment pressure on discharge pipe temperature are negligible because the difference in discharge pipe pressure due to inerting is only a few psig. A leakage flow rate of up to 1000 Ibm/hr will not be sufficient to pressurize the discharge pipe, thereby not affecting discharge pipe temperature (Reference C.2). Therefore, containment pressure effects are judged to be negligible.

Conclusion SRV-3C is operable in the present condition. The leakage that has occurred is minor in nature and it is attributed to the SRV pilot stage. The present leakage level is acceptable as discussed previously. Either intermittent or continuous leakage through and from the SRV within the limits is acceptable for continued operation. Tests and analyses have shown that leakage rates of approximately 225 Ibm/hr (equivalent to 2550F) should not impact the SRV setpoint by more than +1%.

Actions Based on past experience with leaking SRV pilot stages, a lower more conservative action limit has been selected for SRV-3C in order to assure reliable operation and reduce damage to the pilot stage seat and disc.

Therefore, if the SRV-3C discharge pipe temperature exceeds 2351F for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> an orderly shutdown of the reactor shall commence and reactor pressure shall be less than 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Inaddition, an increase in discharge pipe temperature to greater than 2500F may be an indication of a condition not previously observed which Page 12 of 16

may place valve performance outside the bounds of this evaluation. If the SRV-3C discharge pipe temperature exceeds 2500 F then an orderly shutdown shall'bommence and the reactor pressure shall be less than 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Technical Specification surveillance 4.6.D.3 requires that SRV discharge pipe 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 SRV discharge pipe temperature of 212 0F and 235 0F.

4. REGULATORY ANALYSIS 4.1 No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (Entergy) is requesting NRC approval of its evaluation of the acceptability of reactor operation at Pilgrim Nuclear Power Station for greater than 90 days with leakage past Safety Relief Valve SRV-3C as indicated by discharge pipe temperatures in excess of 212 0F. This approval is requested in accordance with Technical Specification 3.6.D.4. Entergy has evaluated whether or not the requested approval involves a significant hazards consideration by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

Indication of elevated SRV discharge pipe temperature is attributed to leakage past the SRV pilot valve. Excessive leakage, corresponding to temperatures greater than 2550 F, has the potential to affect SRV operability by affecting the SRV setpoint or response time. Continued operation with the discharge pipe of the SRV indicating temperatures less than 2550F ensures that the leakage past the SRV is maintained below the threshold for a leakage rate that would potentially have an effect on SRV setpoint or response time.

Administrative controls are in place to ensure that margin to the 2550F value is maintained to assure reliable operation and to reduce the potential for damage to the SRV pilot seat and disc.

The SRV continues to perform the intended design/safety function with no adverse effect because the leakage past the SRV is maintained below the threshold for a leakage rate that could potentially have an adverse impact on the ability of the SRV to perform the design function. The impact of the leakage on other Page 13 of 16

systems is small and all systems continue to be able to perform their intended design functions. Current accident analyses remain bounding and there is no significant increase in the consequences of any accident previously evaluated. In addition, as a result of the leakage, normal plant operating parameters are not affected and consequently there is no increased risk in a plant transient.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Continued plant operation with elevated SRV-3C discharge pipe temperature within the bounds of the established administrative controls ensures that the leakage past the SRV is maintained below the threshold for a leakage rate that would potentially have an effect on SRV setpoint or response time. This ensures that the SRV will perform the intended design/safety function. The leakage does not adversely impact the ability of any system to perform its design function. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

Continued operation with the SRV-3C discharge pipe indicating temperatures in excess of 21 20 F does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. The leakage does not result in excess SRV setpoint drift or response time changes. The imposed administrative controls on plant operation provide assurance that there will be no adverse effect on the ability of the SRV to perform the intended design/safety function. There are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 'no significant hazards consideration" is justified.

Page 14 of 16

5. ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6. REFERENCES A. Pilgrim Station Technical Specifications

1. Specification 3.6.D, Safety and Relief Valves.

B. Pilgrim Station Updated Final Safety Analysis Report

1. Section 4.4, Nuclear System Pressure Relief System.

2. Section 14, Station Safety Analysis.

3. Appendix Q, Supplemental Reload Licensing Reports.

4. Appendix R, Initial Core Station Safety Analysis.

C. General Electric Reports

1. NEDE-24223 Assessment of BWR/3 Mitigation of ATWS, dated December 1979.

2. NSE 13-0282 "Pilgrim Plant, SRV Tailpipe Steam Temperature Correlation for SRV Leakage Monitoring System," dated February 1982.

3. NEDO-22159, 'General Electric Boiling Water Reactor Increased SRV Simmer Margin Analysis for PNPS Unit 1," dated June 1982.

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

5. GENE-0000-0000-6653, 'Project Task Report T0902, Anticipated Transients Without Scram," Rev. 0, dated January 2002.

Page 15 of 16

4.

6. GNF Report 0000-0008-6613SRLR, "Supplemental Reload Licensing Report for PNPS Reload 14, Cycle 15," Rev. 1, dated March 2003.

D. Pilgrim Calculations

1. S&SA-174, uReload Licensing Analysis Inputs - OPL-3 for Reload 14 and Cycle 15."

E. Correspondence to NRC

1. BECo letter, Request for NRC Approval of Engineering Evaluation:

Elevated Tailpipe Temperature on Safety Relief Valve 203-3D, dated October 1,1991.

2. BECo letter, Request for NRC Approval of Engineering Evaluation:

Elevated Tailpipe Temperature of Safety Relief Valve 203-3B, dated January 30, 1998.

3. Entergy letter, Request for NRC Approval of Engineering Evaluation: Elevated Tailpipe Temperature of Safety Relief Valve RV-203-3B, dated December 27, 2000.

4. Entergy letter, Request for NRC Approval of Engineering Evaluation of Elevated Safety Relief Valves' Discharge Pipe Temperatures, dated January 16, 2004; Response to NRC Request for Additional Information, dated February 25, 2004; and Withdrawal of Request, dated March 26, 2004.

F. Correspondence from NRC

1. NRC letter (TAC No. 81678), Engineering Evaluation for Leaking Safety Relief Valve 203-3D, dated October 24,1991.

2. NRC letter (TAC No. MA0881), Engineering Evaluation for Leaking Safety Relief Valve 203-3B, dated March 19,1998.

3. NRC letter (TAC No. MB0874), Engineering Evaluation for Leaking Safety Relief Valve 203-3B, dated February 21, 2001.

4. NRC letter (TAC No. MC1799), Withdrawal of Request, dated April 26, 2004.

G. Pilgrim Engineering/Operability Evaluation

1. Operability Evaluation for Target Rock Two-Stage Safety Relief Valve 203-3C, OE number CR-PNP-2004-3013 and CR-PNP-2004-3047.

Page 16 of 16

ENCLOSURE 2 Commitments (1 page)

It is Entergy's commitment to enforce the following limits on continued operation with SRV discharge pipe temperatures:

NUMBER REGULATORY COMMITMENTS DUE DATE 1 Ifthe SRV-3C discharge pipe temperature exceeds 2350 F Whenever SRV-3C for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> an orderly shutdown of the reactor shall is above 2120 F commence and the reactor pressure shall be less than 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In addition, if the SRV-3C discharge pipe temperature exceeds 2500F, then an orderly shutdown shall commence and the reactor pressure shall be less than 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2 Technical Specification surveillance 4.6.D.3 requires that Whenever SRV-3C SRV discharge pipe temperature be logged daily. This is above 2120 F surveillance shall be performed at an increased I frequency of once per hour.