NG-15-0003, Fifth Ten-Year Interval Inservice Testing Program Relief Requests
| ML15033A372 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/30/2015 |
| From: | Vehec T NextEra Energy Duane Arnold |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NG-15-0003 | |
| Download: ML15033A372 (28) | |
Text
NEXTera ENERG-15-00 1100 ARNOLD NG-1 5-0003 January 30, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Duane Arnold Energy Center Docket No. 50-331 Renewed Op. License No. DPR-49 Fifth Ten-Year Interval Inservice Testing Program Relief Requests The Duane Arnold Energy Center Inservice Testing Program fourth ten-year interval will end on January 31, 2016. Attachment 1 to the letter provides Pump Relief Requests for the fifth ten-year interval. Attachment 2 provides Valve Relief Requests for the Fifth ten-year interval.
In accordance with 10 CFR 50.55a, NextEra Energy Duane Arnold, LLC requests NRC approval of the proposed Relief Requests prior to the start of the fifth ten-year interval on February 1, 2016.
This letter makes no new commitments or changes to existing commitments.
If you have any questions, please contact J. Michael Davis at (319) 851-7032.
T. A. Vehec Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Attachments cc:
Regional Administrator, USNRC, Region III Project Manager, USNRC, Duane Arnold Energy Center Senior Resident Inspector, USNRC, Duane Arnold Energy Center Acý-?
Kwý NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324 to NG-15-0003 NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Inservice Testing Program Pump Relief Requests 6 pages follow
NextEra Energy Duane Arnold Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Pump Relief Request Relief Request Number:
PR-01 Pump
Description:
Positive Displacement Pump Plant System:
Standby Liquid Control (SBLC)
Applicability:
Proposed alternative for SBLC Pump Vibration Instruments Code Group:
B ASME Class:
2 Component No.:
1P230A, 1P230B P&ID:
126 Function:
These are two 100% capacity positive displacement pumps designed to inject sodium pentaborate solution into the reactor at a minimum rate of 26.2 gpm at a discharge pressure greater than or equal to 1150 psig as backup capability for reactivity control independent of normal reactivity control provided by the control rods.
Test Requirement (s):
ISTB-3510(e); General, Frequency Response Range; The frequency response range of the vibration measuring transducers and their readout system shall be from one-third minimum pump shaft rotational speed to at least 1000 Hz.
Basis for Relief:
The nominal shaft rotational speed of these pumps is 242 rpm, which is equivalent to approximately 4 Hz. Based on this frequency and ISTB-3510(e), the required frequency response range of instruments used for measuring pump vibration is 1.33 to 1000 Hz. Procurement and calibration of instruments to cover this range to the lower extreme (1.33 Hz) is impractical due to the limited number of vendors supplying such test equipment and the level of sophistication and cost of the equipment.
These pumps are of a simplified reciprocating (piston) positive displacement design with rolling element bearings, Model Number TD-60, manufactured by Union Pump Corporation. Union Pump Corporation has performed an evaluation of the pump design and has determined that there are no probable sub-synchronous failure modes associated with these pumps under normal operating conditions.
Furthermore, there are no known failure mechanisms that would be revealed by vibration at frequencies below that related to shaft speed (4 HZ); thus, no useful information is obtained below this frequency nor will indication of pump degradation be masked by instrumentation unable to collect data below this frequency to within tolerance prescribed by IST.
Page 1 of 6
NextEra Energy Duane Arnold Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Pump Relief Request Relief Request Number:
PR-01 Pump
Description:
Positive Displacement Pump Plant System:
Standby Liquid Control (SBLC)
Applicability:
Proposed alternative for SBLC Pump Vibration Instruments Code Group:
B ASME Class:
2 Sub-synchronous peaks are usually associated with sleeved bearing components. These frequencies detect shaft to sleeve rub and oil whirl. The IST requirement for detection to 1/3 running speed is to detect these failure mode types. However, this Union Pump Corporation design utilizes roller bearings, which do not have the same failure modes. For a roller bearing design, typical failure is ball or race related and occurs at frequencies greater than turning speed, classified as non-synchronous. As a roller bearing fails, a corresponding change in 1-times turning speed and harmonics indicating excessive looseness and random impacting, not sub-synchronous frequencies, will be seen.
Per the manufacturer, there is no internal gearing in this pump model; therefore, the input shaft rpm is also the crank rpm. The instrumentation for measuring vibration must be adequate for accurately assimilating information at this rpm. The significant modes of vibration with respect to equipment monitoring are as follows:
1-Times Crankshaft Speed - An increase in vibration at this frequency may be an indication of rubbing between a single crankshaft cheek and rod end, cavitation at a single valve, or coupling misalignment.
2-Times Crankshaft Speed - An increase in vibration at this frequency may be an indication of looseness at a single rod bearing or crosshead pin, a loose valve seat in the fluid cylinder, a loose plunger/crosshead stub connection, or coupling misalignment.
Other Multiples of Shaft Speed or Non-synchronous peaks - An increase in vibration at other frequencies may be an indication of cavitation at several valves, looseness at multiple locations, or bearing degradation.
Per the manufacturer, all failure modes that cause vibration in the pump will be at multiples greater than the crank rpm.
Based on the foregoing discussion, it is clear that monitoring pump vibration within the frequency range of 4 to 1000 Hz will provide adequate information for evaluating pump condition and ensuring continued reliability with respect to the pumps' function.
Relief is requested pursuant to 10 CFR 50.55a(a)(3)(i); the alternative testing will provide an acceptable level of quality and safety.
Page 2 of 6
NextEra Energy Duane Arnold Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Pump Relief Request Relief Request Number:
PR-01 Pump
Description:
Positive Displacement Pump Plant System:
Standby Liquid Control (SBLC)
Applicability:
Proposed alternative for SBLC Pump Vibration Instruments Code Group:
B ASME Class:
2 Proposed Test Alternative/Frequency:
Vibration levels of the Standby Liquid Control Pumps will be measured in accordance with the applicable portions of ISTB-3500 with the exception of the lower frequency response limit for the instrumentation (ISTB-3510(e)). In this case the lower response limit of the vibration measuring equipment will be 4.00 Hz.
In addition to the normal SBLC pump IST vibration peak overall result, DAEC engineering department personnel will routinely perform post spectral/waveform analysis of the vibration data to ensure no adverse trends toward mechanical degradation go undetected. This lower limit restriction will not affect the operational readiness of the Standby Liquid Control Pumps, and the OM Code maximum allowable vibration limits for the required action range are being maintained.
The proposed alternative will result in corrective action on a pump prior to the occurrence of significant degradation.
Reference:
Safety Evaluation for Relief Requests Related to the Fourth 10-year Interval Inservice Testing (IST)
Program (TAC NOS. MC8713, MC8784 and MC8785) (ML061870011)
Page 3 of 6
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Pump Relief Request Relief Request Number:
PR-02 Pump
Description:
Positive Displacement Pump Plant System:
Standby Liquid Control (SBLC)
Applicability:
Proposed alternative for SBLC Pump Flow Measurement Code Group:
B ASME Class:
2 Component No.:
1P230A, 1P230B P&ID:
126 Function:
These are two 100% capacity positive displacement pumps designed to inject sodium pentaborate solution into the reactor at a minimum rate of 26.2 gpm at a discharge pressure greater than or equal to 1150 psig as backup capability for reactivity control independent of normal reactivity control provided by the control rods.
Test Requirement (s):
ISTB-3550 Flow Rate: When measuring flow rate, a rate or quantity meter shall be installed in the pump circuit.
ISTB-5300 Positive Displacement Pumps, (a) Duration of Tests, (1) For the comprehensive test; after pump conditions are as stable as the system permits, each pump shall be run at least 2 min. At the end of this time at least one measurement or determination of each of the quantities required by Table ISTB-3000-1 shall be made and recorded.
Basis for Relief:
The positive displacement SBLC pumps are designed to pump a constant flow rate regardless of system resistance. The SBLC system was not designed with a flow meter in the flow loop. The system was designed with a test tank, where the change in level can be measured over time and a flow rate calculated. As part of the modifications made to the SBLC system for the ATWS Rule (10 CFR 50.62), DAEC installed instrumentation to measure the SBLC flow. The ultrasonic flow meter that was installed; however, was not intended to meet the accuracy requirements of the ASME OM Code, and has not proven to be capable of meeting Code accuracy requirements. The accuracy performance of the flow meter is attributed to the lack of adequate straight length of pipe to establish fully developed flow.
In March 2006, portable ultrasonic flow meters were installed on the common SBLC pump discharge piping to determine the practicality of using ultrasonic flow meters to measure flow per ASME OM Code requirements. The flow meter transducers were installed at three different locations on the discharge piping. A vendor representative was on-site to facilitate proper installation and setup of the transducers and flow meters.
Page 4 of 6
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Pump Relief Request Relief Request Number:
PR-02 Pump
Description:
Positive Displacement Pump Plant System:
Standby Liquid Control (SBLC)
Applicability:
Proposed alternative for SBLC Pump Flow Measurement Code Group:
B ASME Class:
2 Each location produced significantly different measured flow rates compared to the other locations and the test tank level method.
NUREG-1482, Revision 2 recognizes that plants may have difficulties with flow instrumentation. In Section 2.5.1, "Justifications for Relief or Alternatives," the NUREG states that compliance with the Code may be impractical because of design limitations. Section 2.5.1.(2) states, "Imposition of the Code requirements would require significant system redesign and modifications. For example, a flow meter does not meet the accuracy requirements of ISTB-3510 and Table ISTB-3510-1 because the present system configuration does not have a straight section of pipe of sufficient length in which to measure flow accurately (see Section 5.5)." In that respect, flow measurement cannot be achieved to the required accuracy using a flow meter. In addition, the SBLC test tank is not large enough to provide two minutes of flow prior to recording flow data. As discussed in NUREG-1482, Revision 2, Section 5.5.2, "Use of Tank Level to Calculate Flow Rate for Positive Displacement Pumps," states, "Requiring licensees to install a flow meter to measure the flow rate and to guarantee the test tank size, such that the pump flow rate will stabilize in 2 minutes before recording data would be a burden because of the design and installation changes to be made to the existing system.
Therefore, compliance with the Code requirements would be a hardship."
Relief is requested pursuant to 10 CFR 50.55a(a)(3)(i); the alternative testing will provide an acceptable level of quality and safety.
Proposed Test Alternative/Frequency:
As a proposed alternative, flow rate for the SBLC pumps will be determined by measuring the change in test tank level over time. The pump will be started with suction from the test tank and will discharge to storage barrels. The test tank level will be approximately the same at the beginning of each test to ensure repeatability. After at least two minutes of pump operation and a change of tank level of at least 20 inches, the time and level are recorded and the pump stopped.
The change in level over the measured time will be converted to flow rate by the following formula:
Q (GPM) = qt AL (inch) /At (Second)
Where: Q is flow rate EP is a constant which reflects tank dimensions and unit conversions AL is the measured change in level in the tank in time At.
Attachment I Page 5 of 6
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Pump Relief Request Relief Request Number:
PR-02 Pump
Description:
Positive Displacement Pump Plant System:
Standby Liquid Control (SBLC)
Applicability:
Proposed alternative for SBLC Pump Flow Measurement Code Group:
B ASME Class:
2 Pump discharge pressure will match system pressure up to the shutoff head of the positive displacement pump. Because of the characteristics of a positive displacement pump, there should be virtually no change in pump discharge flow rate as a result of the rising level in the temporary storage barrels. Therefore, increasing level will not have an impact on test results. By having approximately the same level in the tank at the beginning of each test, repeatable results can be achieved.
Per NUREG-1482, Revision 2, Section 5.5.2 states, "When flow meters are not installed in the flow loop of a system with a positive displacement pump, it is impractical to directly measure flow rate for the pump. The staff has determined that, if the licensee uses the tank level to calculate the flow rate as described in Subsection ISTB-3550, the implementing procedure must include the calculational method and any test conditions needed to achieve the required accuracy. Specifically, the licensee must verify that the reading scale for measuring the tank level and the calculational method yield an accuracy within +/-2 percent for Group A and B tests, and Preservice and Comprehensive Tests. If the meter does not directly indicate the flow rate, the record of the test shall identify the method used to reduce the flow data."
The test tank level will be measured in accordance with the accuracy requirements of OM Table ISTB-3510-1. The calculation method and test conditions required to achieve this accuracy are documented in the implementing procedures.
Reference:
Safety Evaluation of Relief Request for the Duane Arnold Energy Center Fourth 10-Year Pump and Valve Inservice Testing Program (TAC NO. MD1844) (ML070160009)
NUREG-1482, Rev. 2, "Guidelines for Inservice Testing at Nuclear Power Plants" Attachment I Page 6 of 6 to NG-1 5-0003 NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Inservice Testing Program Valve Relief Reguests 19 pages follow
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-01 Valve
Description:
Plant Systems:
Applicability:
Code Category:
Excess Flow Check Valves (EFCV)
Feedwater Control, Residual Heat Removal, Core Spray, Neutron Monitoring, Nuclear Steam Supply Shutoff, Reactor Vessel Recirculation, Reactor Non-Nuclear Instrumentation Proposed Alternative Test for Excess Flow Check Valves (EFCV) in accordance with 10 CFR 50.55a(a)(3)(i)
C ASME Class:
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NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-01 Valve
Description:
Plant Systems:
Applicability:
Code Category:
Excess Flow Check Valves (EFCV)
Feedwater Control, Residual Heat Removal, Core Spray, Neutron Monitoring, Nuclear Steam Supply Shutoff, Reactor Vessel Recirculation, Reactor Non-Nuclear Instrumentation Proposed Alternative Test for Excess Flow Check Valves (EFCV) in accordance with 10 CFR 50.55a(a)(3)(i)
C ASME Class:
1 Function:
Excess flow check valves (EFCVs) specifically designed by Marotta Scientific Controls Inc. for the DAEC are provided in each instrument process line that penetrates the drywell and connects to the reactor coolant pressure boundary. The EFCV is designed so that it will not close accidently during normal operation, will close if a rupture of the instrument line is indicated downstream of the valve, can be reopened when appropriate, and has its status indicated in the control room.
An orifice is installed just inside the drywell on each of these instrument lines. The orifice limits leakage to a level where the integrity and functional performance of secondary containment and associated safety systems are maintained, the coolant loss is within the capability of the reactor coolant makeup system, and the potential offsite exposure is substantially below the guidelines of 10 CFR 50.67.
Regulatory Guide 1.11 requested that an additional isolation valve capable of automatic operation be located outside containment on these instrument process lines. At the DAEC, these are the excess flow check valves.
Test Requirement (s):
ASME OM Code-2004 Edition through the 2006 Addenda.
ISTC-3510, Exercising Test Frequency. Active Category A, Category B, and Category C check valves shall be exercised nominally every 3 months, except as provided by ISTC-3520, ISTC-3540, ISTC-3550, ISTC-3570, ISTC-5221 and ISTC-5222. Power-operated relief valves shall be exercise tested once per fuel cycle.
Basis for Relief:
The excess flow check valve is a simple device: the major components are a poppet and spring. The spring holds the poppet open under static conditions. The valve will close upon sufficient differential pressure across the poppet. Functional testing of the valve is accomplished by venting the instrument side of the tube. The resultant increase in flow imposes a differential pressure across the poppet, which compresses the spring and decreases flow through the valve.
Page 2 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-01 Valve
Description:
Excess Flow Check Valves (EFCV)
Plant Systems:
Feedwater Control, Residual Heat Removal, Core Spray, Neutron Monitoring, Nuclear Steam Supply Shutoff, Reactor Vessel Recirculation, Reactor Non-Nuclear Instrumentation Applicability:
Proposed Alternative Test for Excess Flow Check Valves (EFCV) in accordance with 10 CFR 50.55a(a)(3)(i)
Code Category:
C ASME Class:
1 Excess flow check valves have been extremely reliable throughout the industry. In the first 40 years of operation at the DAEC, no excess flow check valve has failed to close due to actual valve failure (i.e.,
not related to test methodology). The DAEC Technical Specifications (TS) detail what frequency is required to maintain a high degree of reliability and availability, and provide an acceptable level of quality and safety. In the NRC's Safety Evaluation, which is associated with TS Amendment No. 230, the Staff concluded, "Based on the acceptability of the methods applied to estimate the release frequency, a relatively low release frequency estimate in conjunction with unlikely impact on core damage and negligible consequence of a release in the reactor building, we conclude that the increase in risk associated with the licensee's request for relaxation of EFCV surveillance testing to be sufficiently low and acceptable." DAEC requested this relief pursuant to 10 CFR 50.55a(a)(3)(i) to exercise excess flow check valves at the frequency specified in amended DAEC TS Surveillance Requirement (SR) 3.6.1.3.7.
The NRC's Safety Evaluation also states that the radiological consequences of an unisolable rupture of an instrument line were previously evaluated in response to Regulatory Guide 1.11, as documented in DAEC UFSAR Section 1.8.11. This evaluation assumed a continuous discharge of reactor water through an instrument line with a 1/4-inch orifice for the duration of the detection and cooldown sequence. The assumptions for the accident evaluation do not change as a result of the proposed change in test frequency, and the evaluation in DAEC UFSAR Section 1.8.11 remains acceptable.
General Electric NEDO-32977-A (Boiling Water Reactor Owner's Group (BWROG) Topical Report B21-00658-01), Excess Flow Check Valve Testing Relaxation, dated November 1998, (revised through June 2000) was approved by the staff on March 14, 2000. NEDO-32977-A provides additional bases for this relief request. The report concludes that the change in the test frequency had insignificant impact on valve reliability, and that the demonstrated reliability of EFCVs coupled with low consequences of EFCV failure provided adequate justification for extending the test interval up to once every 120 months.
Proposed Test Alternative/Frequency:
Excess flow check valves will be exercised at the frequency specified in the amended DAEC TS Surveillance Requirement (SR) 3.6.1.3.7. The surveillance requirement is to test a representative sample of Excess Flow Check Valves so that each Excess Flow Check Valve is tested at least once every 10 years.
Page 3 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-01 Valve
Description:
Plant Systems:
Applicability:
Code Category:
Excess Flow Check Valves (EFCV)
Feedwater Control, Residual Heat Removal, Core Spray, Neutron Monitoring, Nuclear Steam Supply Shutoff, Reactor Vessel Recirculation, Reactor Non-Nuclear Instrumentation Proposed Alternative Test for Excess Flow Check Valves (EFCV) in accordance with 10 CFR 50.55a(a)(3)(i)
C ASME Class:
1 The Excess Flow Check Valves have position indication in the control room. Check valve remote position indication is excluded from Regulatory Guide 1.97 as a required parameter for evaluating containment isolation. The remote position indication will be verified in the closed direction at the same frequency as the exercise test, which will be performed at the frequency prescribed in the amended DAEC TS Surveillance Requirement (SR) 3.6.1.3.7. After the close position test, the valves will be reset and the remote open position indication will be verified. Although inadvertent actuation of an EFCV during operation is highly unlikely due to the spring-poppet design, the DAEC will verify the EFCVs indicate open in the control room at a frequency greater than once every 2 years.
The failure of an EFCV to isolate would be evaluated per the DAEC corrective action program. The DAEC 10 CFR 50.65 Maintenance Rule Program specifies a performance criteria of less than or equal to 1 maintenance preventable failure to isolate per year on a 3-year rolling average for the combined total of identified automatic isolation Primary Containment valves.
Relief is requested pursuant to 10 CFR 50.55a(a)(3)(i); the alternative testing will provide an acceptable level of quality and safety.
Reference:
- Safety Evaluation related to DAEC Technical Specification Amendment 230 approved by the NRC on December 29, 1999 (ML003672720)
- Safety Evaluation for Relief Requests Related to the Fourth 10-year Interval Inservice Testing (IST)
Program (TAC NOS. MC8713, MC8784 and MC8785) (ML061870011).
Page 4 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 Component No.:
PSV4400, PSV4401, PSV4402, PSV4405, PSV4406, PSV4407 P&ID:
114 Function:
These Safety/Relief Valves (SRVs) provide automatic overpressure protection for the nuclear boiler system, thereby preventing failure of the nuclear process barrier (i.e., reactor coolant pressure boundary). (DAEC Updated Final Safety Analysis Report (UFSAR) Section 5.2.2) Each SRV is self-actuating at its prescribed set-point and resets at approximately 50 psi below its lift set-point.
The Automatic Depressurization System (ADS) utilizes four of the six SRVs provided in the nuclear boiler system to accomplish reactor vessel depressurization. The purpose of the ADS is to provide an automatic means of reducing reactor pressure for events such as pipe breaks or reactor loss of water level transients when the High Pressure Coolant Injection system is unable to maintain reactor water level. The pressure reduction enables low pressure make-up systems such as Low Pressure Coolant Injection (LPCI) and Core Spray to inject additional makeup water into the vessel to restore or maintain water level preventing overheating of the fuel cladding. (DAEC UFSAR Section 6.3.2.2.2)
The Low-Low Set (LLS) System utilizes the two remaining SRVs provided in the nuclear boiler system that are not used for the ADS function to mitigate the induced high frequency loads on the primary containment (torus) and the thrust loads on the SRV discharge lines and tailpipes. This reduces the possibility of a SRV tailpipe rupture occurring inside the torus above the suppression pool water level; thereby creating a bypass of the pressure suppression function. The LLS System automatically controls reactor pressure by opening and closing the LLS SRVs in the relief mode over a wider band of reactor pressure than the safety mode. The LLS valves are the two SRVs with the lowest safety mode pressure relief setpoints. This reduces the number and frequency of SRV actuations allowing the SRV discharge line vacuum relief valves time to clear the discharge lines of water, thus lowering the thrust loads.
In addition, each SRV may be operated in the relief mode from the Main Control Room with individual AUTO/OPEN switches and selected SRVs may be similarly operated from the Remote Shutdown Panel outside the Main Control Room. This capability allows the SRVs to be utilized for reactor vessel pressure control under emergency conditions.
Page 5 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 The six SRVs are Target Rock Three-Stage, Model 7467F design. The SRVs are dual-function valves capable of being independently opened in either the safety or relief mode of operation. The ADS SRVs are identical in construction to the LLS SRVs. Each valve is a pilot-controlled, pneumatically opened, spring shut, reverse-seated globe valve. It can be operated either as a self-actuated pressure relief function or a remote-actuated pressure control function.
The SRV consists of three main sections, the pilot valve, the remote actuator, and the main valve.
Reactor vessel pressure is felt on the top and bottom of the main valve piston, and on the pilot valve bellows via the pilot sensing port. Since reactor vessel pressure is equalized across the main valve piston due to the drilled orifice, it is the main valve spring and differential pressure across the main disc, which keeps the main disc seated. When reactor vessel pressure reaches the self-actuated pressure relief function opening set-point, the pilot valve is forced open against spring pressure, allowing steam to flow into the chamber above the second stage piston. Steam pressure exerted on the second stage piston causes the second stage disc to unseat, providing a relief path for steam above the main disc.
Since the steam can escape through the passage in the valve body faster than it can be admitted through the small main piston orifice, pressure above the main piston will decrease. Pressure will continue to decrease until reactor vessel pressure acting on the under-side of the main piston overcomes the spring pressure and forces the main disc off its seat. When reactor system pressure decreases to approximately 50 psi below the self-actuated pressure relief function opening set-point, the pilot valve will reseat and the main valve spring pressure will reseat the main disc.
Remote actuation of an SRV may be accomplished by its hand-switch located in the Main Control Room, selected switches on the Remote Shutdown Panel, or by automatic initiation signals (ADS or LLS). Upon receipt of a remote initiation signal from any of these sources, the SRV's solenoid operating valve (SOV) is energized and 90-psi nitrogen pressure is directed to a diaphragm near the top of the SRV. This causes the diaphragm and the attached rod to be forced downward. The rod then makes physical contact with the second stage piston causing the second stage disc to unseat. The remainder of the valve operation is the same as for self-actuation function previously described. Removal of the remote initiation signal allows nitrogen pressure to vent off of the diaphragm via the exhaust port of the SOV, thus permitting the spring to reseat the main disc.
Page 6 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Plant Systems:
Applicability:
Code Category:
Safety/Relief Valves (SRV)
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves B/C ASME Class:
1 The main steam relief valves are dual function safety/relief valves that operate as both a pilot operated relief valve (overpressure protection mode) and a power-operated relief valve (manual/ADS/LLS mode).
The SRVs are categorized as Category B and Category C valves in the Inservice Testing Program. The Category of C is consistent with the pilot operated relief valve function and is tested per Appendix I of the ASME OM Code. Category B is consistent with a power-operated relief valve and is tested per Section ISTC-5100 of the ASME OM Code.
Test Requirement (s):
Appendix I Section 1-1320, Test Frequencies, Class 1 Pressure Relief Valves (a) 5-Year Test Interval. Class 1 pressure relief valves shall be tested at least once every 5 years, starting with initial electric power generation. No maximum limit is specified for the number of valves to be tested within each interval; however, a minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 5 years.
(b)
Replacement with Pretested Valves. The Owner may satisfy testing requirements by installing pretested valves to replace valves that have been in service, provided that:
(1) For replacement of a partial complement of valves, the valves removed from service shall be tested prior to resumption of electric power generation; or (2) For replacement of a full complement of valves, the valves removed from service shall be tested within 12 months of removal from the system.
(c) Requirements for Testing Additional Valves. Additional valves shall be tested in accordance with the following requirements.
(1) For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the + tolerance limit of the Owner-established set-pressure acceptance criteria of 1-1310(e) or + 3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.
Page 7 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 (2) If the as-found set-pressure of any of the additional valves tested in accordance with 1-1320(c)(1) exceeds the criteria noted therein, then all remaining valves of that same valve group shall be tested.
(3) The Owner shall evaluate the cause and effect of valves that fail to comply with the set-pressure acceptance criteria established in 1-1320(c)(1) or the Owner-established acceptance criteria for other required tests, such as the acceptance of auxiliary actuating devices, compliance with Owner's seat tightness criteria, etc. Based upon this evaluation, the Owner shall determine the need for testing in addition to the minimum tests specified in 1-1320(c) to address any generic concerns that could apply to valves in the same or other valve groups.
Appendix I Section 1-3410, Class 1 Main Steam Pressure Relief Valves with Auxiliary Actuating Devices Each valve that has been maintained or refurbished in place, removed for maintenance and testing, or both, and reinstalled shall be remotely actuated at reduced or normal system pressure to verify open and close capability of the valve before resumption of electric power generation. Set-pressure verification is not required.
Section ISTC-5113, Valve Stroke Testinq (a) Active valves shall have their stroke times measured when exercised in accordance with ISTC-3500.
(b) The limiting value(s) of full-stroke time of each valve shall be specified by the Owner.
(c) The stroke time of all valves shall be measured to at least the nearest second.
(d) Any abnormality or erratic action shall be recorded (see ISTC-9120) and an evaluation shall be made regarding need for corrective action.
(e) Stroke testing shall be performed during normal operating conditions for temperature and pressure if practicable.
Section ISTC-5114 Stroke Test Acceptance Criteria Test results shall be compared to the reference values established in accordance with ISTC-3300, ISTC-3310, or ISTC-3320.
(a) Valves with reference stroke times of greater than 10 sec shall exhibit no more than 125% change in stroke time when compared to the reference value.
(b) Valves with reference stroke times of less than or equal to 10 sec shall exhibit no more than 150%
change in stroke time when compared to the reference value.
Page 8 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 (c) Valves that stroke in less than 2 sec may be exempted from ISTC-5115(b). In such cases the maximum limiting stroke time shall be 2 sec.
Basis for Relief from 1-1320(a) and 1-3410(d):
This Fifth 10-year IST Interval request for relief is based on Appendix I, ASME OM Code-2004 Edition through the 2006 Addenda. Exercising of the SRV after reinstallation could only be performed during reactor startup when there is sufficient steam pressure to actuate the main disk. Past history indicates that the main disks may not re-seat properly after being exercised during reactor startup resulting in steam leakage into the suppression pool. This leakage results in a decrease in plant performance and the potential for increased suppression pool temperatures and level, which could force a plant shutdown to repair a leaking SRV. Past operating history indicates that the exercising performed during reactor startup is of no significant benefit in ensuring the proper operation of the individual SRV subassemblies.
This request for relief also proposes to implement Code Case OMN-17 "Alternate Rules for Testing ASME Class 1 Pressure Relief/Safety Valves." OMN-17 states in Section (a) that safety valves shall be tested at least once every 72 months (6 years) with a minimum of 20% of the SRV group being tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve that is in service shall not exceed 72 months except that a 6-month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods.
Justification:
Leaking SRVs create operational problems associated with the suppression pool. SRV leakage increases both pool temperature and level, requiring more frequent use of the Residual Heat Removal (RHR) System to maintain the corresponding limits for the suppression pool in the plant's Technical Specifications (TS).
The SRV pilot assemblies removed during a refueling outage are tested at an offsite facility. The as-found testing is performed prior to the resumption of power operation from that refueling outage, meeting the OM code requirements. The valves are refurbished, as necessary, to meet the acceptance criteria of zero leakage, and are certified in writing as being leak free. The valves are then reinstalled in the plant in a subsequent refueling outage and proper pilot operation is confirmed through leak rate testing of the pilot air operators and associated accumulator piping followed by manual lift at reactor power.
Page 9 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 Several aspects of SRV design and operation can contribute to valve leakage. As mentioned earlier, these include test pressure, pilot valve disc and rod configuration, and overall system and valve cleanliness. Actuation of the SRVs after laboratory testing by any means allows these contributors to impact the ability of the valve to re-close completely. DAEC has made significant efforts to minimize the effects of these contributors. In 1999 the DAEC Technical Specifications were changed to permit an as-found tolerance of +/- 3% and +/- 1% as-left tolerance on the SRV opening setpoints. Since that time, DAEC has not had any SRV setpoint failures and had only one instance of seat leakage during testing at the offsite facility in 2009. There have been two instances of valve leakage during power operation; a pilot valve leak in 2004 and a second stage leak in 2010. This recent event occurred shortly after performance of the in-situ test at reduced system pressure and is believed to be a contributing cause of the valve failure.
In reference to the OM Code 2004 Edition through 2006 Addenda, Section 1-1320, "Test Frequencies, Class I Pressure Relief Valves," states that there is a five year frequency for SRV testing.
DAEC proposes to use Code Case OMN-17, "Alternate Rules for Testing ASME Class 1 Pressure Relief/Safety Valves," to change the frequency to six years, including a 6 month grace period, to coincide with the 24-month refueling cycle at DAEC.
Additionally, reducing challenges to the SRVs is a recommendation of NUREG-0737, "TMI Action Plan Requirements," Item I1.K.3.16. This recommendation is based on a stuck-open SRV being a possible Loss of Coolant Accident (LOCA). This relief request is consistent with that NRC recommendation.
Proposed Alternative Test:
As an alternative to the testing required by ASME OM Code-2004, Appendix I, paragraph 1-3410(d),
DAEC proposes to actuate the SRVs in the relief mode at the certified test facility. A test solenoid valve will be energized, the actuator will stroke, and the 2nd stage rod movement will be verified. This test will verify that, given a signal to energize the solenoid valve, the 2nd stage disc rod will travel to unseat the 2nd stage disc. The 2nd stage function will be recorded in the test documentation package for future reference, as needed. Alternate testing is justified since the remaining segments of the SRV relief mode of operation are verified by other tests. The ability of the pilot disc to open is demonstrated in the safety mode actuation bench test. The integrity of the pneumatic and solenoid system for the SRVs is verified by performance of post maintenance leak rate testing, continuity testing, and a functional testing of the solenoid valve while detached from the SRV.
Page 10 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 Automatic valve actuation is proven by Logic System Functional Tests which include verification that the SOV is energized by the automatic signal. The actuator to main body joint is inspected during ISI VT-2 exam performed prior to startup. The above proposed surveillance and testing of the SRVs and associated components provide reasonable assurance of adequate valve operation and readiness.
Following reinstallation, the electrical and pneumatic connections will be verified by energizing the SOVs using the respective control switches and inspecting the pneumatic actuator for movement and leakage (so-called dry lift test). While this test will actuate the SRV second stage, operating experience at other plants indicates that it does not initiate second stage leakage or otherwise damage the valve when performed with no steam pressure; thus, making it a better alternative test to an in-situ steam test during reactor startup.
DAEC proposes to implement Code Case OMN-17 that requires in section (a) a 72-month test interval for Class 1 pressure relief valves with a minimum of 20% of the SRV group being tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 72-month interval, if they exist. The test interval for any individual valve that is in service shall not exceed 72 months except that a six month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods. The removed main steam relief valves will be sent for as-found testing to the offsite test facility. Each main steam relief valve will then be disassembled and inspected for abnormal wear and the specific concerns documented in General Electric Company Service Information Letters (SIL) No. 196, Supplement 17 and No. 646, References 5 and 6 respectively. The post-maintenance tests required by Appendix I, Section 1-3310 will be conducted at the offsite testing facility. As part of implementation of this relief request, DAEC will institute measures to assure that each main steam relief valve will be disassembled and inspected prior to being placed on the new 72-month interval.
Testing will be performed as stated below:
Test Frequencies, Class I Pressure Relief Valves (a) 72-Month Test Interval. Class 1 pressure relief valves shall be tested at least once every 72 months (6-years), starting with initial electric power generation. A minimum of 20% of the valves from each valve group shall be tested within any 24-month interval. This 20% shall consist of valves that have not been tested during the current 5-year interval, if they exist. The test interval for any individual valve shall not exceed 72 months except that a 6 month grace period is allowed to coincide with refueling outages to accommodate extended shutdown periods.
Page 11 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 (b) Replacement With Pretested Valves. The Owner may satisfy testing requirements by installing pretested valves to replace valves that have been in service, provided that:
(1) For replacement of a partial complement of valves, the valves removed from service shall be tested prior to resumption of electric power generation and shall be subjected to the maintenance specified in (d); or (2) For replacement of a full complement of valves, the valves removed from service shall be tested within 24 months of removal from the system.
(c) Requirements for Testing Additional Valves. Additional valves shall be tested in accordance with the following requirements.
(1) For each valve tested for which the as-found set-pressure (first test actuation) exceeds the greater of either the +/- tolerance limit of the Owner-established set-pressure acceptance criteria or +/- 3% of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.
(2) If the as-found set-pressure of any of the additional valves tested in accordance with (c)(1) exceeds the criteria noted therein, then all remaining valves of that same valve group shall be tested.
(3) The Owner shall evaluate the cause and effect of valves that fail to comply with the set-pressure acceptance criteria established in (c)(1) or the Owner-established acceptance criteria for other required tests (e.g., acceptance of auxiliary actuating devices, compliance with Owner's seat tightness criteria). Based upon this evaluation, the Owner shall determine the need for testing in addition to the minimum tests specified.
(d) Maintenance. The Owner shall disassemble and inspect each valve after as-found set-pressure testing to verify that parts are free of defects resulting from time rated degradation or service induced wear. Based upon this inspection, the Owner shall determine the need for additional inspections or testing to address any generic concerns. As-left set-pressure testing shall be performed following maintenance and prior to retuning the valve to service.
(e) Each valve shall have been disassembled and inspected in accordance with (d) above prior to the start of the 72 month test interval. Disassembly and inspection performed prior to the implementation of this Code Case may be used.
Page 12 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Safety/Relief Valves (SRV)
Plant Systems:
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Applicability:
Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves Code Category:
B/C ASME Class:
1 Basis for Relief from ISTC-5113 and ISTC-5114:
The proposed alternatives provide adequate assurance that valve stroke time in the power-actuated mode will be acceptable. Stroke timing of the SRVs will be performed at the offsite test facility as described above. Currently, as-found stroke time testing is performed prior to and after performing maintenance at the offsite test facility. After completion of maintenance, plant surveillance tests with steam at reduced pressure are performed in order to detect gross failures of the SRVs to change position.
The tests performed at DAEC are not as refined as the valve response time test performed at the offsite test facility. The design requirement for the valve stroke time is 0.45 seconds from signal initiation to valve full open in the power-actuated mode (0.40 seconds from signal initiation to start of valve motion and 0.050 seconds (50 milliseconds) for valve stroke to full open). Measuring valve stroke times to this level of accuracy in-situ at the power plant is not practical and only possible under the controlled conditions of the offsite facility. Per ISTC-5114(c), the maximum permissible valve stroke time can be up to 2 seconds.
Consequently, the in-situ test acceptance criterion becomes essentially a "failure to open" criterion.
Therefore, the tests performed at DAEC can only detect gross failures to change position and cannot monitor for valve performance degradation between tests.
Justification:
In-situ stroke timing is not useful for identifying valve degradation over several operating cycles. Rather, an in-situ exercise test will be used to ensure that the valve will function in the power-actuated mode.
This test will be performed at the frequency prescribed in ISTC-351 0 for power-operated relief valves.
Stroke time at the offsite test facility will demonstrate that the valve performs acceptably compared to the stroke times of known good performing valves. Since the offsite test facility cannot duplicate the electrical control system at the plant, actuation of the valve at the test facility is accomplished through a simplified electrical actuation. Observation of the end of the operating stroke at the offsite test facility is indirect, based on evidence of steam flow and pressure, as it is at the nuclear facility, since the relief valves have no positive open indication. Although these differences may result in minor differences in measured stroke time compared to those measured when installed in the plant, the stroke times measured at the test facility will be comparable to each other and thus can be used to detect any abnormality in valve performance.
Proposed Alternative Test:
Stroke times will be measured at the offsite test facility. Stroke times will be measured following valve rebuild. The timing will begin with the actuating electrical signal and end with the indirect indication of the end of the operating stroke. Stroke time acceptance criteria will use a pre-established reference value that represents good performance for the valve type.
Page 13 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
VR-02 Valve
Description:
Plant Systems:
Applicability:
Code Category:
Safety/Relief Valves (SRV)
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves B/C ASME Class:
1 An in-situ exercise test of the valve in the power-actuated mode will be performed at the frequency prescribed in ISTC-3510. The in-situ exercise test will be performed prior to the resumption of electric power generation. Main disc movement and set-pressure verification are not required.
==
Conclusion:==
Based upon the above, the proposed alternatives provide an adequate assurance of quality and safety equal to that of the current Code of record. Consequently, the provisions of 10 CFR 50.55a(a)(3)(i) are judged to be met.
Duration:
The proposed alternatives identified in this relief request shall be utilized during the Fifth 10-year IST Interval that begins on February 1, 2016.
Precedents:
NUREG-1482, Rev. 2, Paragraph 4.3.2.1 states, "In recent years, the NRC staff has received numerous requests for relief or TS changes or both related to the stroke testing requirements for BWR dual-function main steam S/RVs. The 2003 Addendum and earlier editions and addenda to Mandatory Appendix I to the OM Code require the stroke testing of S/RVs after they are reinstalled following maintenance activities. A number of licensees have determined that in situ testing of the S/RVs can contribute to undesirable seat leakage of the valves during subsequent plant operation and have received approval to perform stroke testing at a laboratory facility coupled with in situ tests and other verifications of actuation systems as an alternative to the testing required by the OM Code."
Similar relief has been approved for DAEC (ML12170A421), Dresden and Quad Cities stations (ML081330557), and Peach Bottom Units 2 and 3 (ML081790539), which also use three-stage Target Rock SRVs. The alternative testing approved for these plants included an in-situ actuator test without live steam (dry lift test).
Page 14 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program Valve Relief Request Relief Request Number:
Valve
Description:
Plant Systems:
Applicability:
Code Category:
VR-02 Safety/Relief Valves (SRV)
Nuclear Boiler System Automatic Depressurization System (ADS)
Low-Low Set (LLS) System Proposed alternative test for ASME Class 1 Pressure Safety/Relief Valves B/C ASME Class:
1
References:
- 2. Code Case OMN-1 7, "Alternative Rules for Testing ASME Class 1 Pressure Relief / Safety Valves"
- 3. NUREG-1482, Rev. 2, "Guidelines for Inservice Testing at Nuclear Power Plants"
- 5. General Electric Co. Service Information Letter # 196, Supplement 17, "Target Rock SRV main disc spring relaxation and tip breakage," January 5, 1996.
- 6. General Electric Co. Service Information Letter # 646, "Target Rock safety relief valve failure to fully open," December 20, 2002.
- 8. DAEC UFSAR Section 6.3.2.2.2, Automatic Depressurization System Page 15 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program General Relief Request Relief Request Number:
VR-03
==
Description:==
Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Applicability:
All Pumps and Valves Contained within the Inservice Testing (IST)
Program Scope Applicable Code Edition and Addenda:
ASME OM Code 2001 Edition through 2003 Addenda
Applicable Code Requirement
This request applies to the frequency specifications of the ASME OM Code. The frequencies for tests given in the ASME OM Code do not include a tolerance band.
Code Paragraph Description ISTA-3120(a)
"The frequency for the inservice testing shall be in accordance with the requirements of Section IST."
ISTB-3400 Frequency of Inservice Tests ISTC-3510 Exercising Test Frequency ISTC-3540 Manual Valves ISTC-3630(a)
Frequency ISTC-3700 Position Verification Testing ISTC-5221(c)(3)
"At least one valve from each group shall be disassembled and examined at each refueling outage; all valves in a group shall be disassembled and examined at least once every 8 years."
Appendix 1, 1-1320 Test Frequencies - Class 1 Pressure Relief Valves Appendix 1, 1-1330 Test Frequencies - Class 1 Nonreclosing Pressure Relief Devices Appendix 1, 1-1340 Test Frequencies - Class 1 Pressure Relief Valves that are used for Thermal Relief Application Appendix 1, 1-1350 Test Frequencies - Class 2 and 3 Pressure Relief Valves Appendix 1, 1-1360 Test Frequencies - Class 2 and 3 Nonreclosing Pressure Relief Devices Appendix 1, 1-1370 Test Frequencies - Class 2 and 3 Primary Containment Vacuum Relief Valves Appendix 1, 1-1380 Test Frequencies - Class 2 and 3 Vacuum Relief Valves Except for Primary Containment Vacuum Relief Valves Appendix 1, 1-1390 Test Frequencies - Class 1 Pressure Relief Valves that are used for Thermal Relief Application Appendix II, 11-4000(a)(1)
Performance Improvement Activities Interval Appendix II, 11-4000(b)(1)(e)
Optimization of Condition Monitoring Activities Interval Page 16 of 19
NextEra Energy Duane Arnold, LLC Fifth Ten-Year Interval Duane Arnold Energy Center (DAEC)
Inservice Testing (IST) Program General Relief Request Relief Request Number:
VR-03
==
Description:==
Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Applicability:
All Pumps and Valves Contained within the Inservice Testing (IST)
Program Scope Applicable Code Edition and Addenda:
ASME OM Code 2001 Edition through 2003 Addenda
Reason for Request
Pursuant to 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(ii), relief is requested from the frequency specifications of the ASME OM Code. The basis of the relief request is that the Code requirement presents an undue hardship without a compensating increase in the level of quality or safety.
ASME OM Code Section Inservice Testing (IST) establishes the inservice test frequency for all components within the scope of the Code.
The frequencies (e.g., quarterly) have always been interpreted as "nominal" frequencies and are defined in plant Technical Specifications (TS) Section 5.5.6, "Administrative Controls, Programs and Manuals - Inservice Testing Program."
Licensees routinely applied the surveillance extension time period (i.e., grace period) contained in the plant TS Surveillance Requirements (SR) Applicability, specifically SR 3.0.2. This TS allows for a less than or equal to 25% extension of the surveillance test interval to accommodate plant conditions that may not be suitable for conducting the surveillance. However, regulatory issues have been raised concerning the applicability of the TS "grace period" to ASME OM Code required IST frequencies irrespective of allowances provided under TS SR 3.0.2.
The lack of a tolerance band on the ASME OM Code IST frequency restricts operational flexibility.
There may be a conflict where a surveillance test could be required (i.e., its frequency could expire,) but where it is not possible or not desired that it be performed until sometime after a plant condition or associated Limiting Condition for Operation (LCO) is within its applicability. Therefore, to avoid this conflict, the surveillance test should be performed when it can and should be performed.
The NRC recognized this potential issue in the TS by allowing a frequency tolerance as described in TS SR 3.0.2. The lack of a similar tolerance applied to OM Code testing places an unusual hardship on the plant to adequately schedule work tasks without operational flexibility.
Thus, just as with TS required surveillance testing, some tolerance is needed to allow adjusting OM Code testing intervals to suit the plant conditions and other maintenance and testing activities. This assures operational flexibility when scheduling surveillance tests that minimize the conflicts between the need to complete the surveillance and plant conditions.
Page 17 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program General Relief Request Relief Request Number:
VR-03
==
Description:==
Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Applicability:
All Pumps and Valves Contained within the Inservice Testing (IST)
Program Scope Applicable Code Edition and Addenda:
ASME OM Code 2001 Edition through 2003 Addenda Proposed Alternative and Basis for Use:
Code Case OMN-20 is included in the ASME OM Code, 2012 Edition and will be used as the alternative to the frequencies specified in ASME OM Code.
The requirements of Code Case OMN-20 are described below.
1 Section IST and earlier editions and addenda of ASME OM Code specify component test frequencies based either on elapsed time periods (e.g., quarterly, 2 years, etc.) or based on the occurrence of plant conditions or events (e.g., cold shutdown, refueling outage, upon detection of a sample failure, following maintenance, etc.).
a) Components whose test frequencies are based on elapsed time periods shall be tested at the frequencies specified in Section IST with a specified time period between tests as shown in the table below. The specified time period between tests may be reduced or extended as follows:
- 1) For periods specified as less than 2 years, the period may be extended by up to 25% for any given test.
- 2) For periods specified as greater than or equal to 2 years, the period may be extended by up to 6 months for any given test.
- 3)
All periods specified may be reduced at the discretion of the owner (i.e., there is no minimum period requirement).
Period extension is to facilitate test scheduling and considers plant operating conditions that may not be suitable for performance of the required testing (e.g., performance of the test would cause an unacceptable increase in the plant risk profile due to transient conditions or other ongoing surveillance, test or maintenance activities). Period extensions are not intended to be used repeatedly merely as an operational convenience to extend test intervals beyond those specified.
Period extensions may also be applied to accelerated test frequencies (e.g., pumps in Alert Range) and other less than two year test frequencies not specified in the table below.
Period extensions may not be applied to the test frequency requirements specified in Subsection ISTD, Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-water Reactor Nuclear Power Plants, as Subsection ISTD contains its own rules for period extensions.
Page 18 of 19
NextEra Energy Duane Arnold, LLC Duane Arnold Energy Center (DAEC)
Fifth Ten-Year Interval Inservice Testing (IST) Program General Relief Request Relief Request Number:
VR-03
==
Description:==
Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Applicability:
All Pumps and Valves Contained within the Inservice Testing (IST)
Program Scope Applicable Code Edition and Addenda:
ASME OM Code 2001 Edition through 2003 Addenda Frequency Specified Time Period Between Tests Quarterly (or every 3 months) 92 days Semiannually (or every 6 months) 184 days Annually (or every year) 366 days X Years X calendar years where X' is a whole number of years > 2 b)
Components whose test frequencies are based on the occurrence of plant conditions or events may not have their period between tests extended except as allowed by ASME OM Division: 1 Section IST 2009 Edition through OMa-2011 Addenda and earlier editions and addenda of ASME OM Code.
Duration of Proposed Alternative:
The proposed alternatives identified in this relief request shall be utilized during the Fifth 10-year IST Interval that begins on February 1, 2016.
Precedents:
Similar relief has been approved for DAEC (ML14144A002) and Quad Cities Nuclear Power Station, Units 1 and 2 (ML13042A348). The alternative testing approved for these plants included an in-situ actuator test without live steam (dry lift test).
References:
- 1. DAEC TS Section 1.4, "Frequency"
- 2. DAEC TS SR 3.0.2 [Specified Frequency (25% grace Period)]
- 4. DAEC TS Section 5.5.6, "Inservice Testing Program" Page 19 of 19