RS-15-003, Relief Request I3R-26 Associated with the Third 10-Year Lnservice Inspection Program Interval

Relief Request I3R-26 Associated with the Third 10-Year Lnservice Inspection Program Interval

ML15005A380
Person / Time
Site:	Byron
Issue date:	01/05/2015
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-15-003
Download: ML15005A380 (16)

Text

4300 Winfield Road IMO/ Warrenville, IL 60555 ANP- Exelon Generation 630 657 2000 Office RS-15-003 10 CFR 50.55a January 5, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Byron Station, Unit 1 Facility Operating License No. NPF-37 NRC Docket No. STN-50-454

Subject:

Relief Request I3R-26 Associated with the Third 10-Year lnservice Inspection Program Interval In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (a)(3)(ii), Exelon Generation Company, LLC, (EGC) hereby requests NRC approval of the attached relief request associated with the Third 10-year Inservice Inspection (ISI) Program Interval for Byron Station, Unit 1. The third interval of the Byron Station ISI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda.

NRC approval of the attached relief is requested on the basis that hardship and unusual difficulty exists in establishing a system configuration that will subject selected Class 1 components to Reactor Coolant System (RCS) pressure during the system pressure test as required by IWB-5220 without a compensating increase in the level of quality and safety. EGC is proposing an alternative to perform the examination of selected Class 1 piping and valves at plant conditions other than those required by IWB-5220.

The details of the 10 CFR 50.55a relief request are enclosed. EGC requests approval of this relief request by September 11, 2015 to support Byron Station, Unit 1 refueling outage (B1 R20) scheduled to commence on September 14, 2015.

January 5, 2015 U. S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Ms. Dwi Murray at (630) 657-3695.

Respectfully, David M. Gullott Manager Licensing Exelon Generation Company, LLC

Attachment:

10 CFR 50.55a Relief Request I3R-26 cc: NRC Regional Administrator Region III NRC Senior Resident Inspector Byron Station NRC Project Manager, NRR Byron Station Illinois Emergency Management Agency Division of Nuclear Safety

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 1 of 14 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1, IWB-5220, Code Case N-798, and Code Case N-800 Examination Category: B-P Item Number: B15.10

Description:

Code Class 1 pressure retaining components that are beyond the first normally closed valve to the second boundary component. Including RC, RY, SI, and RH piping, vents, drains, and valves as noted on Table 1 and Table 2.

Component Number: Piping components listed in Table 1 and Table 2 Drawing Numbers: Drawings listed in Table 1 and Table 2 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The Third Interval Inservice Inspection (ISI) program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda (Reference 1).

3.0 APPLICABLE CODE REQUIREMENT:

ASME Section XI, Table IWB-2500-1, "Examination Categories," Examination Category B-P, Item B15.10 requires that a system leakage test be conducted prior to startup following each refueling outage in accordance with the requirements of IWB-5220.

Paragraph IWB-5222(a) requires that the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup with the visual examination extended to include the second closed valve at the boundary extremity. Paragraph IWB-5222(b) requires the pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval shall extend to all Class 1 pressure retaining components within the system boundary.

ASME Code Case N-798, "Alternative Pressure Testing Requirements for Class 1 Piping Between the First and Second Vent, Drain, and Test Isolation Devices," (Reference 2) states that for portions of Class 1 vent, drain and test piping between the first and second isolation devices that normally remain closed during plant operation, only the boundaries of IWB-5222(a) shall apply.

ASME Code Case N-800, "Alternative Pressure Testing Requirements for Class 1 Piping Between the First and Second Injection Valves," (Reference 3) states that for portions of

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 2 of 14 the Class 1 boundary between the first and second isolation valves in the injection and return path of standby safety systems, the system leakage test may be conducted by pressurization of the Class 1 volume using the Class 2 safety system to pressurize the volume. It further states that such alternative tests shall be performed each inspection interval and the system leakage test shall be conducted using the pressure associated with the Class 2 system function that provides the highest pressure between the Class 1 isolation valves.

The Byron Station, Unit 1 Third 10-year ISI Interval ends on July 15, 2016.

4.0 REASON FOR REQUEST:

Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that hardship and unusual difficulty exists in establishing a system configuration that will subject selected Class 1 components to reactor coolant system (RCS) pressure during the system pressure test without a compensating increase in the level of quality and safety. EGC requests approval to perform the examination of selected Class 1 piping and valves at plant conditions other than those required by IWB-5222(b) by using the alternative boundaries permitted by Code Cases N-798 and N-800 at the end of the interval.

The design of some Byron Station, Unit 1 Class 1 process piping requires substantial effort to extend the boundary subject to RCS pressure where check valves or non-redundant components serve as the first system isolation from the RCS. Such configurations would require temporary piping installations, such as high-pressure hoses, and/or other unusual temporary system configurations in order to achieve test pressures at upstream piping and valves required by IWB-5222(b).

These components are located in areas involving occupational radiation exposure, and leakage testing of these lines would require significant dose. Establishing and restoring such temporary configurations would result in an unwarranted increase of worker radiation exposure. It is estimated that extending the boundary to the second isolation device for all Class 1 components would result in a total dose of 1.858 REM and require an estimated 644 person-hours. This estimate is based on the dose and personnel records of the activities performed during the previous Byron Station, Unit 1 10-year interval refueling outage (B1R13) in the Spring of 2005. The activities associated with this work include scaffold erection, insulation removal, valve manipulations, freeze seals, examinations, re-installation of insulation, and scaffold removal.

Based on the above, the extension of the boundary subjected to RCS pressure during system leakage tests to include all Class 1 pressure retaining components within the system boundary in accordance with IWB-5222(b) represents a hardship and unusual difficulty that does not provide a compensating increase in the level of quality and safety provided by the examination.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 3 of 14 The following tables contain specific information pertaining to the various pipe segments for which relief is being requested.

Table 1, "Class 1 Piping Between the First and Second Vent, Drain, and Test Isolation Devices (N-798)," identifies the Class 1 pressure retaining components associated with the requested relief that will remain in their normal operating configuration and will not be pressurized to RCS pressure during the system leakage test. The piping segments listed in the table contain 56 socket welds. None of the welds are selected under the criteria of the risk-informed ISI program. No leakage has been identified with these piping segments.

Table 2, "Class 1 Piping Between the First and Second Isolation Valves (N-800),"

identifies the Class 1 pressure retaining components associated with the requested relief that will remain in their normal operating configuration and will not be pressurized to RCS pressure during the system leakage test, however, will be examined at operating pressures associated with the outboard Class 2 system functional pressure. The piping segments listed in the table contain 301 socket, 284 butt, and 11 branch welds. Eight Safety Injection welds are selected under the criteria of the risk-informed ISI program.

No leakage has been identified with these piping segments.

Small Bore Class 1 RCS Manual Vent and Drain Lines Relief is requested from pressurizing piping between the first and second isolation device on small size vent and drain lines. The Class 1 vents or drain lines in the RCS identified in Table 1 range in size from 3/4 inch to two inches. The Class 1 vents and drains in the RCS are equipped with inboard isolation valves and outboard blind flanges or pipe caps that provide double isolation of the Reactor Coolant Pressure Boundary (RCPB). The valves are maintained in the closed position during normal plant operation and the downstream pipe and blind flange or pipe cap are not normally pressurized. To pressurize those piping segments as required by IWB-5222(b), it would be necessary to open the inboard valves manually to pressurize the downstream piping and connections.

Pressurization by this method defeats the double isolation and reduces the margin of personnel safety for those performing venting and draining operations on high-pressure components after the test. Furthermore, performing the test with the inboard isolation valves open requires several person-hours to position the valves for the test and restore the valves to their closed positions once the test is completed. These valves are located in close proximity to the RCS loop piping and would require personnel entry into high radiation areas within the containment and a consequential increase in radiation exposure. The required plant configuration for testing would defeat one of the two RCPB isolation devices, thereby increasing the likelihood of intersystem leakage and potentially allowing a class 2 system to be over pressurized or lifting a class 2 pressure relief valve. Thus, compliance with IWB-5222(b) requirement results in unnecessary hardship without a compensating increase in the level of quality and safety.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 4 of 14 These piping segments are included in the examination population for visual examination (VT-2) through the entire length as part of the Class 1 system leakage test at the conclusion of each refueling outage. The proposed leakage test will not specifically pressurize past the first isolation valve for this examination. No external or visible leakage will be allowed. Since this type of test will assure that the combined first and second isolation devices are effective in maintaining the RCPB at normal operating temperature and pressure, the increase in safety achieved from the IWB-5220 required leakage test is not commensurate with the hardship of performing such testing.

Reactor Head Vent The reactor head vent lines are configured into two groups; the main assembly consists of two lines for a parallel flow-path terminating in a non-class open-ended vent and an inboard single line vent terminating in a non-class capped piece. Byron Station Technical Requirements Manual (TRM) 3.4.e, "Reactor Vessel Head Vents," requires that two reactor vessel head vent paths shall be OPERABLE and closed in Modes 1, 2, 3, and 4; with no provision identified for testing these valves in elevated-pressure modes. Testing these lines may only be performed at lower pressure that would require access to the upper integrated head package to install manual pumps.

The proposed system leakage test will not specifically pressurize past the first isolation valve for this inspection. The examination of the vent lines will be performed using pressure associated with the normal startup valve configurations.

Large Bore Class 1 Safety Injection and Shutdown Cooling Piping Segments HPSI/Cold Leg Injection These piping segments, identified in Table 2, provide the flow path for High Pressure Safety Injection (SI) into the RCS. The primary isolation devices are the four 1.5-inch check valves at the cold leg oriented to flow into the RCS. The upstream isolation is at a single 3-inch check valve. The piping segments provide the required double isolation barrier for the RCPB. These lines are visually examined both during the RCS system leakage test as Class 1 boundary lines, and during the High Pressure SI System functional pressure test conducted at SI Pump discharge pressure in accordance with the requirement to examine systems at their highest operating pressure.

Leakage testing at RCS pressure requires a pressure source to be connected by way of temporary high pressure hose connections and is performed in conjunction with RCS Pressure Isolation Valve Seat Leakage Testing. For the proposed testing, the examination of these components will be performed during the associated isolation valve leakage surveillances scheduled at the end of the 10-year interval.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 5 of 14 Hot Leg/Cold Leg Safety Injection These large bore piping segments, identified in Table 2, provide the flow path for Safety Injection into the RCS. The primary isolation devices are the two 10-inch check valves oriented to flow into the RCS with 6-inch and 2-inch second isolation valves on branch lines. The piping segments provide the design required double isolation barrier for the RCPB. These lines are visually examined during the RCS system leakage test within the Class 1 boundary lines.

Leakage testing at RCS pressure would require unusual temporary system configurations, which would challenge the Class 2 piping and components should the Class 1 to Class 2 boundary valve leak by toward the Class 2 system(s). This configuration would violate the design requirement for two RCPB isolation devices during testing. For the proposed testing, the components will be subjected to the outboard Class 2 system functional pressure associated with the SI Accumulators.

Hot Leg Shutdown Cooling Suction There are two 12-inch Hot Leg Shutdown Cooling Suction lines, one each from RCS Loop A and C Hot Legs, identified in Table 2. These piping segments consist of piping between the two Shutdown Cooling Suction valves on each train of the system (valves 1RH8701B and 1RH8701A on Train A and valves 1RH8702B and 1RH8702A on Train B). These valves are open-interlocked at a required set point below 337 psig to avoid over-pressurization of the Shutdown Cooling System. The interlock prevents manual opening of the valves from the Control Room when RCS pressure is above the set point.

The piping segment is visually examined (VT-2) through the entire length as part of the Class 1 system inspection at the conclusion of each refueling outage, as well as when the Shutdown Cooling System is in service.

The examination of these components will be performed using the outboard Class 2 system functional pressure associated with the normal Shutdown Cooling system pressure and valve configurations during the performance of RCS pressure isolation valve test.

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ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 10 of 14 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

In accordance with 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that hardship and unusual difficulty exists in establishing a system configuration that will subject selected Class 1 components to RCS pressure during the system pressure test as required by IWB-5222(b) without a compensating increase in the level of quality and safety.

EGO proposes to perform the examination of selected Byron Station Unit 1 Class 1 piping and valves at the end of the third 10-Year ISI interval with plant conditions other than those required by IWB-5220 by using the alternative boundaries permitted by Code Cases N-798 and N-800.

The objective of the required extended pressure boundary conditions in accordance with IWB-5222(b) is to detect evidence of leakage, and thereby verify the integrity of the RCPB beyond the first isolation valve. As discussed above, in order to meet IWB-5222(b) requirement, the establishment of and the return from the required temporary configurations would involve considerable time to accomplish, result in an unwarranted increase of worker radiation exposure, exposure of personnel to industrial safety risks due to use of a single isolation valve, and potentially over pressurize downstream piping in the event of valve leakage. In addition, these temporary configurations could require bypassing Updated Final Safety Report (UFSAR) specified safety system protective features and interlocks. As a result, EGC has concluded that compliance with IWB-5222(b) requirement constitutes a hardship without a compensating increase in the level of quality and safety. The proposed alternative testing methods permitted in Code Cases N-798 and N-800 would provide a level of assurance that the RCS pressure boundary is maintaining structural integrity at elevated pressure.

In the unlikely event of a through wall leak in the piping segments identified in Tables 1 and 2 during normal operation, the leak would result in unidentified RCS leakage. RCS leakage detection instrumentation have been designed to aid operating personnel in differentiating between possible sources of detected leakage within the containment and identifying the physical location of the leak. The RCS leakage detection instrumentation consists of the containment sump monitor and the containment atmosphere radioactivity monitor (gaseous or particulate). Technical Specifications (TS) 3.4.15, "RCS Leakage Detection Instrumentation," requires the containment sump monitor, in combination with a gaseous or particulate radioactivity monitor to be operable.

The containment floor drain sump flow monitor (RF008) and the reactor cavity sump flow monitor (RF010) are normally utilized to fulfill the containment sump monitor requirement used to collect unidentified leakage. Alarms are provided to alert the operator of leakages of 1.0 gallon per minute (gpm). When the alarm function is not capable of detecting 1.0 gpm of unidentified leakage within one hour, the containment floor drain sump flow indication may be periodically monitored to ensure the capability of detecting 1.0 gpm of unidentified leakage within one hour. In lieu of the containment floor drain

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Revision 0 Page 11 of 14 sump flow monitor (RF008), either containment sump level monitor (PC002 or P0003) can be used by monitoring a change in sump level over a period of time in such a manner as to ensure the capability of detecting 1.0 gpm of unidentified leakage within one hour.

The reactor coolant contains radioactivity that, when released to the containment, can be detected by the gaseous (PRO11B) or particulate (PRO11A) containment atmosphere radioactivity monitor. Radioactivity detection systems are included for monitoring both particulate and gaseous activities because of their sensitivities and rapid responses to RCS leakage, but have recognized limitations. Reactor coolant radioactivity levels will be low during initial reactor startup and for a few weeks thereafter, until activated corrosion products have been formed and fission products appear from fuel element cladding contamination or cladding defects. If there are few fuel element cladding defects and low levels of activation products, it may not be possible for the gaseous or particulate containment atmosphere radioactivity monitor to detect 1.0 gpm increase within one hour during normal operation. However, the gaseous or particulate containment atmosphere radioactivity monitor is operable when it is capable of detecting 1.0 gpm increase in unidentified leakage within one hour given an RCS activity equivalent to that assumed in the design calculations for the monitors.

TS 3.4.13, "RCS Operational LEAKAGE," limit system operation in the presence of leakage from RCS components to amounts that do not compromise safety. Surveillance Requirement (SR) 3.4.13.1 requires the performance of RCS water inventory balance to verify RCS leakage is within limits to ensure that the integrity of the RCPB is maintained.

In the event that unidentified leakage increases greater than 0.10 gpm above the normal, steady state value for a given plant condition during the performance of the RCS water inventory balance, administrative procedures require that the controls and actions that establishes the controls and expectations for monitoring RCS leakage under the Boric Acid Corrosion Control (BACC) program be implemented.

Actions that implement the BACC program define the methodology used to establish acceptable baseline values, establish unidentified leakage action levels, the criteria used to ensure adequate monitoring of RCS leakage occurs, and the minimum actions that could be taken at each action level to ensure the safe operation of the plant. The program also addresses abnormal trends in RCS primary system leakage indicators, which may provide indication of leaks much smaller than TS and RCS leakage levels.

The majority of welds encompassed by the boundaries described in Table 2 were in the population of welds subject to examination under ASME Section XI (Examination Category B-J) in the first and second ISI intervals and in the population of the current third interval Risk Informed lnservice Inspection (RI-ISI) program. These programs are intended to verify the structural integrity of piping welds. The selection criteria for inspection under the current interval RI-ISI program are determined by the PRA risk rankings and degradation mechanism assessments. During the first, second, and third

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-26 Request for Relief for Alternative Requirements for Pressure Retaining Boundary during System Leakage Test In Accordance with 10 CFR 50.55a(a)(3)(ii)

Revision 0 Page 12 of 14 intervals at Byron Station, there have been no failures during examinations performed on selected welds.

Relief is requested from the extended pressure test boundaries of IWB-5222(b) required for the selected Class 1 components during the System Leakage Test conducted at or near the end of the inspection interval.

For those portions of Class 1 vent and drain lines between the first and second isolation devices that normally remain closed during plant operation, Byron Station will conduct the required end of interval system pressure tests as prescribed in IWB-5222(b) using the alternatives of Code Case N-798 with only the boundaries of IWB-5222(a) will apply as follows:

• The Class 1 vents and drain lines proposed system leakage test will not specifically pressurize past the first isolation valve for this inspection. No external or visible leakage will be allowed.

• The reactor head vent lines proposed system leakage test will not specifically pressurize past the first isolation valve for this inspection. The examination of the vent lines will be performed using pressure associated with the normal startup valve configurations.

For those portions of the Class 1 boundary between the first and second isolation valves, Byron Station will conduct the required end of interval system pressure tests as prescribed in IWB-5222(b) using the alternatives of Code Case N-800 with only the boundaries of IWB-5222(a) will apply as follows:

• The High Pressure Safety Injection (HPSI) examinations will be performed during the associated isolation valve leakage surveillances scheduled at the end of the 10-year interval.

• The Hot Leg/Cold Leg Safety Injection examinations will be performed using the outboard Class 2 system functional pressure associated with the SI Accumulators.

• The Hot Leg Shutdown Cooling Suction examination will be performed using the outboard Class 2 system functional pressure associated with the normal Shutdown Cooling system pressure and valve configurations during the performance of RCS pressure isolation valve test.

Basis for Use:

As described above, the activities associated with the requirements of IWB-5222(b) would incur increased radiation exposure, potential personnel hazards, and additional workforce resources. The associated activities would include scaffold erection,

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Revision 0 Page 13 of 14 insulation removal, valve manipulations, freeze seals, examinations, re-installation of insulation, and scaffold removal, which would not result in a commensurate increase in the quality and safety of the systems. Therefore, the extension of the boundary subjected to RCS pressure during system leakage tests to include all Class 1 pressure retaining components within the system boundary in accordance with IWB-5222(b) represents a hardship and unusual difficulty that does not provide a compensating increase in the level of quality and safety provided by the examination.

6.0 DURATION OF PROPOSED ALTERNATIVE:

The duration of the proposed alternative is requested for the remainder of the Byron Station Unit 1 Third lnservice Inspection Interval currently scheduled to end in July 15, 2016.

7.0 PRECEDENTS

The NRC has previously authorized similar relief requests for the use of ASME Code Cases N-798 and N-800. Authorization has been recently granted for relief requests from the following stations.

1. St. Lucie Plant, Unit 2, Third Inspection Interval Relief Request RFA 9, was authorized by NRC Safety Evaluation dated November 25, 2013, ADAMS Accession No. ML13308C426.

2. Sequoyah Nuclear Plant, Units 1 and 2, Third Inspection Interval Relief Request 11-SPT-1, was authorized by NRC Safety Evaluation dated May 20, 2014, ADAMS Accession No. ML14133A112.

3. Surry Power Station, Units 1 and 2, Fourth Inspection Interval lnservice (ISI)

Program, System Pressure Testing (SPT), SPT-003 and SPT-002, was authorized by NRC Safety Evaluation dated May 28, 2014, ADAMS Accession No. ML14142A089.

8.0 REFERENCES

1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components, 2001 Edition, through 2003 Addenda, the American Society of Mechanical Engineers, New York.

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2. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Case N-798, "Alternative Pressure Testing Requirements for Class 1 Piping Between the First and Second Vent, Drain, and Test Isolation DevicesSection XI, Division 1."

3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Case N-800, "Alternative Pressure Testing Requirements for Class 1 Piping Between the First and Second Injection ValvesSection XI, Division 1."

4. Byron Station, Units 1 and 2 Technical Requirements Manual 3.4.e, "Reactor Vessel Head Vents."

5. Byron Station, Units 1 and 2 Technical Specifications 3.4.15, "RCS Leakage Detection Instrumentation."

6. Byron Station, Units 1 and 2 Technical Specifications 3.4.13, "RCS Operational LEAKAGE."