05000293/LER-2005-003

LER-2005-003,

Docket Number
Event date:	05-11-2005
Report date:	0-0-1007
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
Initial Reporting
2932005003R00 - NRC Website

� FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) �

BACKGROUND

The Pilgrim Station Pressure Relief System (PRS) is designed to prevent over-pressurization of the ASME Boiler and Pressure Vessel Code qualified nuclear steam supply system. The PRS consists of two safety valves and four two-stage relief valves. These valves are installed in the main steam system piping upstream of the main steam isolation valves and are located within the Drywell. The safety valves are self-actuating, provide over-pressure protection, and discharge directly into the Drywell atmosphere when actuated. The relief valves augment the safety valves and are sized to prevent unnecessary actuation of the safety valves. The relief valves are self-actuating and discharge into the suppression pool through discharge piping connected to the valves. Each two-stage relief valve consists of a pilot assembly and main stage. The pilot assembly provides the pressure sensing function and the main stage provides the pressure relieving function. The relief valves are also part of the Automatic Depressurization System (ADS). As part of the ADS, the relief valves are designed to automatically actuate as a result of a depressurization permissive signal, and can also be manually actuated from the control room for depressurization.

Technical Specification (TS) 3.6.D.1 specifies that the nominal setpoint of the relief valves shall be selected between 1095 and 1115 psig and that all relief valves shall be set at this nominal set point ± 11 psi. The valves' nameplate setpoint is 1115 psig. Therefore, based on the tolerance limit of 11 psi (±1%), a maximum pressure of 1126 psig and a minimum pressure of 1104 psig are allowed. The established TS limit is stricter than the standard allowable relief valve setpoint drift range of ± 3% given in Section XI of the ASME Boiler and Pressure Vessel Code.

The main steam relief valves were manufactured by Target Rock Corporation, model #7567F.

Since the early 1980s, increased initial lift pressure (or upward setpoint drift) has been an industry concern applicable to the two—stage relief valves found in BWRs. Industry investigation of relief valve reliability problems revealed the primary cause of upward setpoint drift in the two-stage relief valves was corrosion bonding of the pilot valve disk to its seat. Three different design modifications were found to reduce or counteract the corrosion bounding: 1) installation of ion beam implanted platinum pilot valve disks, 2) the installation of Stellite-21 pilot valve disks, and 3) the installation of additional pressure actuation switches. PNPS implemented changes to install the Stellite-21 pilot valve disk design in the mid 1980 timeframe.

NRC review of upward setpoint drift is documented in NRC Regulatory Issue Summary 2000-12, Resolution of Generic Safety Issue B-55, "Improved Reliability of Target Rock Safety Relief Valves," and Generic Issue 165, "Spring Actuated Safety and Relief Valve Reliability." In the review, the NRC staff found that the industry has significantly improved valve performance and is continuing efforts to evaluate and improve performance, as necessary. Therefore, the staff found no new requirements were necessary and that existing quality assurance, maintenance rule, and code testing requirements were adequate to ensure reliable valve performance in the future.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) All four pilot assemblies for main steam relief valves RV-203-3A, 3B, 3C and 3D were removed during the May 2005 Refueling Outage (RFO-15). RV-203-3C had experienced steam leakage problems while installed during the operating cycle (Cycle 15) and was the subject of License Amendment No.

208 which was issued on December 23, 2004. This amendment allowed for continued operation with the steam leakage provided that certain conditions were satisfied.

EVENT DESCRIPTION

On May 11, 2005 Pilgrim Station was notified that three of four pilot valve assemblies previously installed had as-found popping pressures that exceeded the maximum TS tolerance limit of 1126 psig.

The as-found popping pressures were 1161 psig (serial number 1046), 1139 psig (serial number was 1124 psig. These test results were received when the plant was shutdown, during the 2005 refueling outage.

CAUSE

The probable cause of the as-found popping pressures exceeding the TS tolerance limit was setpoint drift resulting from corrosion bonding of the pilot valve disk to its seat. The corrosion bonding most likely occurred while the pilots were installed and in-service.

CORRECTIVE ACTION

The following corrective actions were taken. Four spare pilot valve assemblies were installed to replace the pilot valve assemblies that were removed during RFO-15. The spare pilot valve assemblies that were installed were tested and certified to be within 1% of the nominal set point required by technical specifications. New insulation was installed on the SRVs which may inhibit corrosion bonding of the seat and disk assembly while the valve is in-service.

The following corrective actions are planned. Engineering is working with industry representatives to evaluate actions and potential modifications to reduce or mitigate set-point drift.

These actions are within the scope of the corrective action program.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) � PILGRIM NUCLEAR POWER STATION 05000-293 2005 003 00 Loss of Coolant Accident (LOCA) — Fuel Peak Clad Temperature:

Following a small break LOCA and vessel isolation, reactor pressure will remain high and is controlled by cycling (opening and closing) relief valves. The small break analysis for Pilgrim Station assumes both HPCI and RCIC are unavailable. Core cooling is provided by the Automatic Depressurization System (ADS) in combination with low-pressure Core Standby Cooling Systems (CSCS). Until ADS initiation, the loss of inventory from the vessel is a function of break area and reactor pressure as controlled by cycling relief valves. After the initial discharge of stored energy from the reactor vessel to the suppression pool by multiple relief valves, a single relief valve is capable of removing decay heat.

Since the lowest as-found popping pressure of the four relief valve pilot assemblies was 1124 psig (serial number 1208), the analysis of record is bounding with respect to reactor pressure and inventory loss from the vessel prior to depressurization by ADS. Therefore, the existing LOCA analysis provides a bounding prediction of core uncovery time, fuel clad heatup, and peak clad temperatures.

Anticipated Transient Without Scram (ATWS):

Beginning in Cycle 15, the licensed thermal power level for PNPS was increased from 1998 MWth to 2028 MWth. Each of the four main steam relief valves was modified to increase the throat diameter, which provides a 7.5% increase in steam flow capacity per valve. During the licensing phase of the power uprate, a revised ATWS analysis was performed. These analyses resulted in a peak vessel pressure of 1495 psig. An evaluation of the as-found condition indicates that in the worst case, the peak vessel pressure resulting from the as-found popping pressures is estimated to be slightly greater than 1500 psig for approximately 30 seconds, well below the stress analysis limit of 1875 psig for faulted conditions (i.e., conditions associated with extremely low probability postulated events which may impair the integrity and operability of the nuclear system to the point where public safety is involved). It also should be noted that the estimated peak vessel pressure does not exceed the hydrostatic test pressure (125% of design pressure or 1560 psig) required by the ASME code to verify Reactor Coolant System integrity prior to initial plant startup. Therefore, given an ATWS event considering the as-found relief valve opening pressures, the estimated peak vessel pressure did not increase significantly and based on engineering judgment, system integrity would not have been impaired.

FACILITY NAME (1)� DOCKET NUMBER (2) I � LER NUMBER (6)� PAGE (3)

REPORTABILITY

This report is submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) because it was conservatively assumed that the as-found popping pressures could have been the pressures at which the relief valves would have operated if a high reactor pressure condition had occurred while the pilot assemblies were installed. The condition is assumed to have existed for a period greater than the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limiting condition for operation specified in Technical Specification 3.6.D.2 for the relief valves.

SIMILARITY TO PREVIOUS EVENTS

A review was conducted of Pilgrim Station LERs. The review focused on LERs related to relief valve tests submitted since 2001. The review identified LER 2001-004-000, LER 2004-001-00 and LER 2004-03-00.

ENERGY INDUSTRY IDENTIFICATION SYSTEM (EIIS) CODES The EIIS codes for this report are as follows:

COMPONENTS� CODES Valve, Relief �

RV

SYSTEMS� CODES Main Steam� SB