ML16224B014

Pilgrim Nuclear Power Station Response to Request for Additional Information - Relief Request Number PRR-52, Proposed Alternative to Pressure Testing of Mechanical Joints as a Result of Repair/Replacement Activity and the Use of Code Case N

ML16224B014
Person / Time
Site:	Pilgrim
Issue date:	08/02/2016
From:	Perkins E P Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.16.039, CAC MF7025
Download: ML16224B014 (13)

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• *===* Entergx August 2, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Nuclear Operations, Inc. Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth. MA 02360 SUBJECT: Response to Request for Additional Information -Relief Request Number PRR-52, "Proposed Alternative to Pressure Testing of Mechanical Joints as a Result of Repair/Replacement Activity and the Use of Code Case N-795" Pilgrim Nuclear Power Station Docket No. 50-293 Renewed License No. DPR-35 REFERENCE: NRC to Entergy Transmittal, "Request for Additional Information for Proposed Alternative Relief Request No. PRR-52 for Pilgrim (CAC NO. MF7025)" (1.16.023) LETTER NUMBER: 2.16.039 Dear Sir or Madam: Pursuant to the request by the U.S. Nuclear Regulatory-Commission Staff for additional information contained in the Reference, please find attached the Pilgrim Nuclear Power Station response. Please contact me at (508) 830-8323 or Murray Williams at (508) 830-8275 if you have any questions. There are no regulatory commitments included in this letter. Sincerely, Manager, Regulatory Assurance EPP/mw Attachment: Response to Request for Additional Information -Relief Request Number PRR-52, "Proposed Alternative to Pressure Testing of Mechanical Joints as a Result of Repair/Replacement Activity and the Use of Code Case N-795" Entergy Nuclear Operations, Inc. Pilgrim Nuclear Power Station cc: Mr. Daniel H. Dorman Regional Administrator, Region I U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd., Suite 100 King of Prussia, PA 19406-2713 Ms. Booma Venkataraman, Project Manager Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8C2A Washington, DC 20555 NRC Senior Resident Inspector Pilgrim Nuclear Power Station Letter No. 2.16.039 Page 2 of 2

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to Letter Number 2.16.039 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST NUMBER PRR-52, "PROPOSED ALTERNATIVE TO PRESSURE TESTING OF MECHANICAL JOINTS AS A RESULT OF REPAIR/REPLACEMENT ACTIVITY AND THE USE OF CODE CASE N-795" (10 Pages)

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to Letter No. 2.16.039 Page 1 of 10 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST NUMBER PRRa52 PROPOSED ALTERNATIVE TO PRESSURE TESTING OF MECHANICAL JOINTS AS A RESULT OF REPAIR/REPLACEMENT ACTIVITY AND THE USE OF CODE CASE N-795 PILGRIM NUCLEAR POWER STATION ENTERGY NUCLEAR OPERATIONS, INC. DOCKET NO. 50-293 (CAC NO. MF7025) Request for Additional Information (RAI) 1 Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(b)(2)(xxvi) requires repair/replacement activities use the provisions in IWA-4540("c) of the 1998 Edition of ASME Code,Section XI, for pressure testing Class 1, 2 and 3 mechanical joints when using the 2001 Edition through the latest edition and addenda incorporated by reference in paragraph 10 CFR 50.55a(a)(1)(ii). ASME Code edition 1998 with no addenda,Section XI, IWA-4540(c) references IWA-5211 (a) for system pressure tests. Please clarify which edition of the code IWA-5211 (a) will be drawn from. Response to RAI 1: Our understanding of the condition in 10 CFR 50.55a(b )(2)(xxvi) is to ensure that mechanical joints involved in a repair/replacement activity are pressure tested. Paragraph IWA-5211 (a) of the 2001 Edition with the 2003 Addenda will used in conjunction with IWA-4540(c) of the 1998 Edition for that pressure testing. Note that the Pressure Test Program for PNPS will continue use of the 2001 Edition with the 2003 Addenda of the Code per Request for Alternative (PRR-26). RAl2 Given a repair/replacement is made to a reactor coolant pressure boundary (RCPB) component which cannot be isolated during a maintenance or forced outage, indicate the expected dose associated with performing the testing and inspections in accordance with the requirements of:

• IWA-4540(c) of the 1998 edition of the ASME Code,Section XI, assuming the procedure used mirrors the procedure currently used during refueling outages.

• Code Case N-795. Discuss any As-Low-As-Reasonably-Achievable (ALARA) concerns and whether they could be mitigated in some other manner.

Response to RAI 2: Attachment to Letter No. 2.16.039 Page 2of10 Given a repair/replacement is made to a reactor coolant pressure boundary (RCPB) component which cannot be isolated during a maintenance or forced outage, utilizing historical data for Drywell entry from RF020 in 2015, ALARA personnel the expected dose associated with performing the testing and inspections in accordance with the requirements of IWA-4540(c) of the 1998 edition of the ASME Code,Section XI to be 919 mrem. The majority of the dose received is due to the closure of manual valves to establish the test line-up and reopening of the valves upon test completion. Performance of the testing and inspection in accordance with the requirements of Code Case N-795 is estimated to be 25 mrem. The licensee states that proposed alternative will be used on those con:iponents that cannot be isolated. Please clarify that for those components that can be isolated, they will be inspected following the guidance of ASME Code Edition 1998, no addenda,Section XI, IWA-4540(c). Response to RAI 3: For those mechanical joints that can be isolated, the requirements of the 2001 Edition with the 2003 Addenda for pressure testing will be.followed. Note that IWA-4540(c) of the 1998 Edition refers to IWA-5211 (a) for the requirements on how pressure testing is performed. The "Basis for Use" section of the proposed alternative states that the core decay heat during a maintenance outage is much higher than that after a refueling outage, and that the heat load is difficult to control once shutdown cooling is removed from service. I 1) What are the temperature and pressure limits for use of shutdown cooling? 2) Given these limits, please explain why pressurization to 90 percent normal operating pressure, with a hold for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, is possible but pressurizing to 100 percent normal operating pressure is unusually difficult. Response to RAI 4: RAI 4 has three specific questions: Why is there more decay heat in a maintenance outage than in a refueling outage (RFO)? The reason for increased decay heat load experienced for a maintenance outage versus a typical refueling outage is primarily due to the larger effective irradiated core size that would exist during the maintenance outage. During a refueling outage, on average, 25-30% of the fuel bundles are replaced with new fuel bundles, thereby removing 30% of the effective decay heat

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to Letter No. 2.16.039 Page 3of10 source. Additionally, the* new non-irradiated fuel bundles act as a heat sink for the other 70% of the irradiated core. In addition, the decay heat from the irradiated fuel bundles declines .over the duration of the refueling outage. In general terms, there would be approximately 2.5 to 3 times the decay heat to manage in a maintenance outage than following the core reload from an RFO. What are limits of the Shutdown Cooling (SOC) System? The Residual Heat Removal (RHR) suction piping rating of 80 PSI is the limiting system condition. As such, the normal system isolation setpoint is set at a conservative lower value. During conduct of the Class 1 leakage test, RHR SOC is secured from service and unavailable to aid in removal of decay heat for the full duration of the test which is typically up to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> for inspection, remediation and restoration from vessel solid conditions. Why is maintaining 90% reactor pressure for extended periods easier to control than 100% reactor pressure? A reactor pressure of 90% of the normal operating pressure is the nominal Mechanical Hydraulic Control (MHC) system setpoint of a General Electric (GE) Boiling Water Reactor (BWR) Mark 3/4. At 5% reactor power, there are two redundant pressure regulators controlling pressure with .the turbine bypass valves, bypassing the 5% steam, within its 25% steam capacity. This is a manageable position with minimal but steady steam demand above the decay heat contribution. This allows the reactor to maintain a balanced condition with water fed from a constant flow Control Rod Drive (CRD) system and rejected within the Reactor Water Cleanup (RWCU) System capacity. At this point during reactor startup, maintaining 90% reactor *pressure and no greater than 5% reaetor power, diverse and redundant protective controls are enforced by the Reactor Protection System (RPS) and the Primary Containment Isolation System (PCIS), providing the equipment safeguards for human access to the drywell while also providing the elevated temperature, pressure and process flow to enable the necessary conditions for component test and inspection. The GE BWR 3/4 uses the turbine inlet pressure as the control parameter to establish* turbine bypass and control valve positions. The control system uses a proportional band with no rate or reset feature. Due to line pressure losses with steam flow, reactor pressure rises as.the steam flow is increased. Pilgrim cannot use its designed pressure control systems to reach 100% reactor pressure without being near 100% reactor power, creating a drywell environment that would not allow human access for any inspection based on radiation and temperature safety concerns. The licensee states that in the unlikely event there is a leak in the repaired/replaced joint, their leakage detection system capability will catch it before the leak becomes a structural issue. Discuss whether the RCS leakage detection systems satisfy NRC Regulatory Guide 1.45, Revision 1, "Guidance on Monitoring and Respondfng to Reactor Coolant System Leakage" (ADAMS Accession No. ML073200271).

Response to RAI 5: Attachment to Letter No. 2.16.039 Page 4of10 Code Case N-795 is based on the assumption that a properly performed system leakage test (IWB-5220), conducted at a test pressure of at least 87% of the pressure required by IWB-5221 (a), will discover any existing source of leakage. There is no new or additional dependency ori the Reactor Primary Pressure Boundary Leakage Monitoring Systems as a consequence of using Code Case N-795. The principal characteristics of these comprehensive Leakage Monitoring Systems are nonetheless described below. The Pilgrim Relief Request (PRR)-52 states, as a matter of fact, that "In the unlikely event that leakage would occur subsequent to the VT-2 examination, at higher pressures associated with 100% rated reactor power, leakage would be detected by the drywell monitoring systems, which include drywell pressure monitoring, the Containment Atmosphere Monitoring system and the drywell floor drain sumps. Leakage monitoring is required by Technical Specifications." The PNPS Reactor Pressure Boundary Leak Detection Systems are fully described in the PNPS Final Safety Analysis Report (FSAR) Section 4.10 NUCLEAR SYSTEM LEAKAGE RATE LIMITS. The overall Leakage Detection System comprises the following individual systems: The Drywell Equipment and Drywell Floor Sump Collection System monitors the identified and unidentified condensate and leakage which flows 'to the Drywell Floor Drain and Equipment Drain Sumps. As the water which has been collected in the sumps is pumped out, the discharge flow from each sump is individually metered by flow integrators. Total Leakage Rate is routinely calculated from these flow integrators and a record is maintained and reviewed to detect increases in the leakage rate. Primary Containment Atmospheric Temperature, Humidity, and Pressure Instrumentation are also available as an aid to detecting leakage. Condensation from the drywell atmosphere occurs as the atmosphere is circulated through the Drywell Coolers. This condensation is collected and piped to the Drywell Floor Drain Sump. Fluid leakage from the Primary Pressure Boundary will result in increased cooling loads on the Drywell Air Coolers which will result in abnormal temperature measurements on the cooling units. The condensation on the coolers will increase and abnormally high condensate flows to the Floor Drain Sump will result. Condensation on the Drywell walls and structures within the primary containment will also collect in the Floor Drain Sump. The integrated Floor Drain Sump flow, the Drywell Atmosphere Pressure and Temperature, the Drywell Atmosphere Humidity, and the Drywell Air Cooler Temperatures are all employed as indicators of potential leakage from the Primary Pressure Boundary.

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to Letter No. 2.16.039 Page 5of10 FSAR Section 4.10.3.3 "Reactor Pressure Boundary Leak Detection System" further describes the leakage monitoring objective as follows: The monitoring of Airborne Radioactivity levels in the containment atmosphere permits operators to evaluate leakage relative to the probable source. For example, an abundance of iodine in the containment atmosphere would indicate water leakage, Wherea.s an abundance of gaseous activity would indicate steam leakage. If an unusual increase above background levels on any channel (as would be expected from a reactor pressure boundary leak) should occur, an alarm will sound in the control room. Daily equilibrium activity levels are taken which are compared to previous equilibrium levels to detect possible background level changes or instrument malfunction. Alarm settings are established at or below 1 Oe6 counts per minute (cpm). The unidentified leakage limit is 5 gallons per minute (gpm). Past experience has shown that the 1 Oe6 cpm alarm point has always been reached well below the 5 gpm limit (typical levels are between 2 and 3 gpm). A leakage rate of 150 gpm has been conservatively calculated to be the minimum liquid leakage from a crack large enough to propagate rapidly. An allowance for reasonable leakage which does not compromise barrier integrity and is not identifiable is made for normal operation. The unidentified leakage rate limit is established at 5 gpm which is far enough below the 150 gpm leakage rate to allow time for corrective action to be taken before the process barrier is significantly compromised. FSAR Section 4.10.4 Safety Evaluation provides the following summary: The unidentified leakage rate limit is based, with an adequate margin for contingencies, on th.e calculaJed leakage from a crack large enough to propagate rapidly. The established limit is sufficiently low so that even if the entire unidentified leakage rate were coming from a single crack in the nuclear system process barrier, corrective action could be taken before the )ntegrity of the barrier is threatened with significant compromise. The design of the Reactor Pressure Boundary Leak Detection Systems pre-dates Regulatory Guide 1.45 and the PNPS System has not been reconciled with the Guide's recommendations. The PNPS System is in general .conformance with the intent of the Guide; most differences are involved with the qualification requirements for the monitoring components. The NRC staff understands that the proposed alternative will use nuclear heat to generate the necessary pressure needed to perform the repair/replacement inspections. However, these pressure tests are necessary to ensure the integrity of the RCPB. Please discuss the operating experience associated with testing the integrity of repairs/replacements on the RCPB, considering the frequency and size of potential leaks associated with fleet-wide historical repairs which would fall in the scope of this request.

Response to RAI 6: Attachment to Letter No. 2.16.039 Page 6of10 An operating experience (OE) search for external leakage detected during pressure testing of code repaired/replaced items was performed utilizing the Institute of Nuclear Power Operations online search tool, the PNPS corrective action database and the PNPS historical code repair/replacement log dating back to 2011. Search criteria used was designed to capture BWR fleet-wide RCPB leakage events resulting from pressure testing of RCPB components that had been subjected to repair/replacement activities. The search yielded results (within the scope of PRR-52) that were limited to OE concerning leakage from undervessel CRD flanges that had been replaced during refuel outages. At PNPS, CRD flange leakage detected after replacement and during the Class 1 system pressure test following refuel outages is typically in the drops per minute range and is a result of the lower test temperatures seen during the Class 1 system pressure test. CRD flange leakage normally arrests itself with no corrective action as a result of thermal expansion of the mechanical joint due to higher moderator temperatures experienced during startup. With regard to potential RCPB external leakage during pressure testing of repair/replacement activities in BWRs, the following components are routine repair/replacement items, any of which have the potential to leak during pressure testing: o CRD and LPRM exchanges

• Safety relief valve replacements

• Piping and valve mechanical joints (flanged connections, packing leaks)

• Recirculation pump seals Leakage from the above listed components would be readily detectable by both visual examination and plant drywell leakage collection systems monitored in the control room. Leakage would be minimal from non-pressure boundary conditions such as inadequately tightened flange bolting or gasket conditions and would not challenge plant safety systems or Technical Specification leakage limits. Conversely, the worst case scenario where complete separation of a RCPB component causes unrestricted leakage is considered extremely unlikely due to approved plant maintenance and work process procedures combined with associated quality oversight inspection. In the unlikely event unrestricted leakage from an unsecured component did occur, it would be detected at much lower pressures long before reaching test pressure due to the excessive leakage which would be readily detected in the control room by plant leakage monitoring systems. In addition, in reviewing industry specific requests concerning the use of a lower test pressure, three approved requests were located. These requests were for the replacement of Main Steam relief valves that were found to be leaking internally past the valve seat causing high tail pipe temperatures during startup from a refueling outage.

The following is a list of those approved request for alternatives: Attachment to Letter No. 2: 16.039 Page 7of10

• Duane Arnold Energy Center-RE: Request for Authorization of Alternative Regarding Pressure Test Requirements (TAC NO. MC2328, ADAMS Accession Number ML040930429) dated April 13, 2004.

• Monticello Nuclear Generating Plant -One-time lnservice Inspection Program Plan Relief Request No. 8 for Leak Testing the "B" and "G" Main Steam Safety Relief Valves, (TAC No. MB9538, ADAMS Accession Number ML031640464), dated June 13, 2003.

• Duane Arnold Energy Center -Alternative Regarding Pressure Testing Requirements for Main Steam Safety Relief Valve PSV 4402 (TAC No. ME5143, ADAMS Accession Number ML 11237A105) dated September 6, 2011 The licensee states that there are risks associated with performing the repair/replacement required inspections and associated pressure testing at 100% reactor operating pressure. Please describe the procedure which would be used for this testing (similar to the procedure currently performed during refueling outages} and specify what the risks are. Include a discussion of:

• which equipment would be available during this testing if it were performed in

• accordance with IWA-4540(c} of the 1998 edition of the ASME Code,Section XI

• whether there would be changes to the shutdown interlocks

• the abnormal plant conditions/alignments necessary to complete this testing e the available sources for heat removal and pressure/inventory control and whether they would be sufficient for the purposes of this test following ma:intenance or forced outage. Response to RAI 7: RAI 7 requests a discussion of risk involved with testing to 100% pressure using procedure PNPS 2.1.8.5 "Reactor Vessel Pressurization and Temperature Control for Class 1 System Leakage Test." Which equipment would be available during testing in accordance with (IAW) PNPS 2.1.8.5? The equipment or systems that would be available and utilized for this test are Control Rod Drive System, Core Spray keepfill lines, RWCU, Reactor Recirculation System and the drain valves of the Main Steam Line Sys'tem. The Reactor Pressure Vessel and connected piping will be filled solid using the CRD system. A Recirculation System and CRD system pump, along with the reactors decay heat will be the energy source pressurizing the vessel and piping. The RWCU system and MSL drains will be the discharge flow path as necessary to assist in establishing and maintaining stable plant, test and PTLR conditions.

Whether there are changes to shutdown interlocks? Attachment to Letter No. 2.16.039 Page 8of10 The systems that will have interlocks/protective features defeated during the performance of PNPS 2.1.8.5 include the PCIS, RPS, and the Recirculation System speed control. RHR SOC is intentionally secured as a pre-requisite to establish test conditions, since it would exceed its PCIS isolation setpoint as pressure is increased beyond its suction piping limitation previously discussed. As setup for this test, the reactor is scrammed and the mode switch is placed in 'SHUTDOWN' position. For the proposed alternative test, RPS and mode switch shutdown interlocks would be fully functional. The abnormal plant conditions/alignment required to conduct PNPS 2.1.8.5. As previously stated, RHR SOC is intentionally secured in order .to establish test conditions. The is filled solid to the Main Steam Isolation Valves, removing the availability of both the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems. Once the vessel is filled and pressurized, the plant has only the RWCU and manual operation of Main Steam Line drains to control level, pressure, and temperature to assist in meeting test conditions, as well as ensuring that test conditions are maintained inside the Technical Specifications Pressure Temperature Limits Report Curve. This test condition results in a water solid and isolated reactor pressure vessel with the normal methods for decay heat removal secured. The available heat removal. and pressure/inventory system availability. It would be dependent on operator use of SRVs and certain system operation should component malfunction or miss-communication occur. PNPS 2.1.8.5 specifies the minimum Core Standby Cooling System (CSCS) availability that must be maintained. While SOC is secured for the test as previously stated, the RHR SOC mode also remains available. However, based on test setup and as previously stated, RWCU and MSL drains are the primary depressurization methods. There is very little margin for error should this letdown path be lost with a solid vessel to maintain the reactor pressure vessel within the Technical Specifications Pressure Temperature Limits Report Curve. The purpose of the test is to identify pressure boundary Wherever possible, Pilgrim would isolate and test those portions of Class 1 systems that were repaired/replaced during the outage to I code requirements as part of post work testing prior to any modified maintenance outage vessel leakage test as proposed. Using Code Case N-795, Pilgrim would be in a normal plant start-up alignment with all of the required safety and non-safety systems available (including the above systems) to heat up to test pressure*, maintain test conditions, and establish safe plant and human conditions for conduct of the inspection; all within the normal start-up procedure.

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to Letter No. 2.16.039 Page 9of10 Given that the proposed alternative will use nuclear heat to generate the necessary pressure needed to perform the repair/replacement inspections, the NRC staff needs additional information to confirm that the request for the proposed alternative pursuant to 10 CFR 50.55a(z)(2) on the basis that compliance with the specified requirements as described in this request would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Please provide the following information for cases where the post repair/replacement leak test is performed at the reduced pressure permitted by ASME Code Case N-795. Provide the information for two cases: 1) Using nuclear heat and 2) not using nuclear heat. For both cases, please provide:

• Available methods and systems for heat removal (including heat removal capacity for each available system)

• Available methods and systems for pressure control. Available methods and systems for inventory control.Available methods and systems for reactivity control.

• An indication of the mode and plant operating state.

• Any changes to normal interlocks.

• An indication of whether the heat removal, pressure, inventory, and reactivity control would be sufficient for the configuration of the plant. Providing this information in a tabular format is acceptable. Response to RAI 8: RAI 8 requests a comparison of both methodologies regarding hardship, quality, and safety. In summary, the Class 1 leakage test requires two full man days to identify the boundary for tagging, installation of jumpers if necessary, establishment of test conditions and then system restoration. This is in addition to actual inspection and repair time, which would be-the same for both test approaches. The quality of the inspection would be the same, with no major differences to drywell environment regarding personnel safety. However, the nuclear safety aspect of an inspection at approximately 5% power as part of a normal start-up evolution with all systems in normal alignment, performed with a full complement of CSCS available would be a more conservative approach when compared to the artificial test alignment which renders steam driven CSCS unavailable, secures decay heat removal systems, bypasses portions of the PCIS and RPS logic, and leaves minimal system options available for depressurization and heat rejection should the test need to be aborted:

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to Letter No. 2.16.039 Page 10of10 The following table is provided to contrast the differences in the test methods. Source Nuclear Non-Nuclear Heat Removal MHC System Manual drain to Condenser thru RWCU Pressure MHC System Manual control of CRD Pump Control Level Control Feed Water Level Control System Manual balance of drain & CRD with RWCU discharge path Reactivity RWM, RMCS, RPS,SBLC, Neutron All Rods In/ 2 SRMs Operable Control Monitoring Mode Startup Shutdown Normal All active Portions of PCIS and RPS Initiation Interlocks Logic defeated, Recirculation Pump Speed Control interlock defeated Heat Removal Within capacity of MHC system for Decay Heat would be within CSCS within capacity heat removal as operated during capacity. Limited Temp/Press Margin. normal start-up. HPCI and RCIC HPCI and RCIC inoperable. operable.