LR-N09-0104, Relief Requests Associated with the Third Inservice Inspection Interval

Relief Requests Associated with the Third Inservice Inspection Interval

ML091410451
Person / Time
Site:	Salem
Issue date:	05/12/2009
From:	Keenan J Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N09-0104
Download: ML091410451 (9)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC 10 CFR 50.55a LR-N09-0104 MAY 12 2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Nuclear Generating Station, Units I and 2 Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

Relief Requests Associated with the Third Inservice Inspection (ISI)

Interval In accordancewith 10 CFR 50.55a(a)(3), "Codes and standards," PSEG Nuclear LLC (PSEG), hereby requests NRC approval of proposed Relief Requests SC-13R-91 and

$1 -13R-92 as alternatives to the requirements of the American- Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection and Testing of Components of Light-Water Cooled Plants," for system leakage test conducted at or near the end of each inspection interval.

PSEG requests approval of the proposed requests by 03/01/2010 to permit the proposed alternatives to be implemented during the refueling outages in Spring 2010 (Unit 1) and Fall 2012 (Unit 2). For Unit 1 the third interval will end on May 20, 2011 and for Unit 2 the third interval will end on November 27, 2013. The Code of Record for the third interval is ASME Code,Section XI, 1998 Edition through 2000 Addenda.

The proposed relief requests are provided in the attachments to this letter. Relief request S1-13R-92, Attachment 2, requests authorization to use ASME Code Case N-731 in lieu of ASME Code Section XI requirements. PSEG understands that Code Case N-731 may be approved for use in the next revision of Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1." However the request is being submitted because the applicable Regulatory Guide may not be issued prior to the next Unit 1 outage. The relief in attachment 2 is limited to the Spring 2010 outage.

There are no commitments contained in this letter.

Au4 -7 95-2168 REV. 7/99

LR-N09-0104 Page 2 If you have any questions or require additional information, please contact Mrs. Erin West of my staff at 856-339-5411.

Sincerely,

,Ianager - Licensing PSEG Nuclear LLC Attachments:

1. Relief Request SC-13R-91

2. Relief Request S1-13R-92 cc: S. Collins, Administrator, Region I, NRC R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector Salem P. Mulligan, Manager IV, NJBNE H. Berrick - Salem Commitment Tracking Coordinator L. Marabella - Corporate Commitment Tracking Coordinator

LR-N09-0104 ATTACHMENT 1 Salem Nuclear Generating Station, Unit Nos. I and 2 Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 Relief Request - SC-13R-91 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety

1. ASME Code Component(s) Affected Code Class: 1 Examination Category: B-P Item Number: B15.50 and B15.70

Description:

Vent and Drain Class 1 pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval.

Unit/Inspection: Salem Unit 1 & 2 / Third (3 rd) 10-Year Intervals Affected components consist of Class 1 Reactor Coolant System double isolation vent and drain valves and piping between valves. Valves are normally closed during plant operation and the outboard valves only see pressure if the inboard valve is open or leaks by the valve seat. The size of the affected piping and valves is 3/4" and 1" NPT.

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection and Testing of Components of Light-Water Cooled Plants," 1998 Edition through 2000 Addenda, for system leakage tests conducted at or near the end of each inspection interval. For Unit 1 the third interval began on May 19, 2001 and will end on May 20, 2011. For Unit 2 the third interval began on November 27, 2003 and will end on November 27, 2013.

3. Applicable Code Requirement

ASME Section XI IWB-5222(b) states, "The pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval shall extend to all Class I pressure retaining components within the system boundary."

4. Reason for Request

The many vent and drain connections within the RCPB have double manual isolation valves. The requirement to extend the system leakage test boundary to the outboard valve on these vent and drain connections results in a hardship without a

Salem Units 1 and 2 Inservice Inspection Program LR-N09-0104 Relief Request SC-13R-91 10 CFR 50.55a compensating increase in the level of quality and safety. Repositioning the inboard manual valves before and after the test will take considerable time and will result in an increase in radiological dose to plant personnel. These off-normal configurations may also contribute to the risk of delaying normal plant start-up because of the critical path time and effort required to ensure system configuration is restored.

PSEG Nuclear, LLC estimates that complying with the current IWB-5222(b) requirement would result in an accumulated dose of approximately 0.75 person-rem for Salem Unit 1 and an additional 0.75 person-rem at Salem Unit 2 during each respective outage when the end of interval pressure test would be completed.

The purpose of the required extended pressure boundary condition is to detect evidence of leakage resulting in a validation of the integrity of the RCS pressure boundary beyond the first isolation valve. Meeting those requirements involves considerable time to establish and return from the required temporary configuration resulting in both risk of delaying normal plant startup following a refueling outage and an increase in personnel radiation exposure.

5. Proposed Alternative and Basis for Use Pursuant to 10CFR50.55a(a)(3)(ii), PSEG Nuclear LLC requests authorization to utilize the following alternative requirements in lieu of the requirements of Article IWB-5222(b):

Perform visual leakage examination of Class 1 piping one inch NPS and less on vent and drain connections off the reactor coolant pressure boundary (RCPB) at or near the end of each inspection interval that have double manual isolation valves with the inboard isolation valve maintained in the normally closed position.

These double manual isolation valves perform no other safety function other than maintaining the Class 1 pressure boundary.

The vent and drain connections are normally closed during plant operation and the outboard valves only become pressurized if the inboard valve is open or leaks by the valve seat. Seat leakage, although undesirable, is not indicative of a flaw in the pressure boundary. The non-isolable portions of these vent and drain connections are normally pressurized during operation and are visually examined during the leak test conducted at each refueling outage. The portions of the vent and drain connections that are isolated are still within the scope for visual examination which ensures no adverse condition exist. Any through-seat leakage on the inboard vent or drain valve would result in pressurization of the outboard valve and piping segment between the valves during the visual examination.

Additionally, the plant technical specifications for RCPB leakage monitoring provide reasonable assurance that appropriate actions, including plant shutdown, would be taken if leakage exceeded specified limits.

Page 2 of 3

Salem Units 1 and 2 Inservice Inspection Program LR-N09-0104 Relief Request SC-13R-91 10 CFR 50.55a Based on the above justification, PSEG requests relief from the requirements of IWB-5222(b) pursuant to 10 CFR 50.55(a)(3) for the current third interval for the affected components and approval of the proposed alternative to complete visual examinations on the piping with vent and drain valves in their normal operation position. The proposed alternative provides reasonable assurance of structural integrity.

6. Duration of Proposed Alternative The duration of the request for proposed alternative for Salem Unit 1 is through the end of the 3 rd interval currently scheduled to end on 05/20/2011; for Salem Unit 2, the duration is through the end of the 3 rd interval currently scheduled to end on 11/27/2013.

7. Precedents In Reference 1, the NRC authorized Entergy's proposed alternative to complete visual examination of RCPB vent and drain connections for leakage with the inboard isolation valve in the normally closed position during the system leakage test conducted at or near the end of each inspection interval.

In Reference 2, the NRC authorized Calvert Cliff's proposed alternative to conduct the required system leakage examination of the Class 1 pressure retaining components with the first normal closed or "inboard" isolation valves in their normally closed positions for the remainder of their Third Ten- Year Inservice Inspection Interval.

In Reference 3, the NRC authorized the proposed alternative for Progress Energy Carolinas which was to pressurize only up to the inboard isolation valves of several systems that would exclude a small segment of the Class 1 pressure boundary from attaining test pressure. Visual examination during pressurization would include all components within the system boundary.

8. Reference

1. NRC Safety Evaluation dated Febuary 2, 2007 (TAC Nos. MD1399, MD1400, MD1401, MD1402, and MD1403), Arkansas Stations, Units 1 and 2, Grand Gulf Station, River Bend Station, Waterford Steam Electric Station, Unit 3. Docket NOS. 50-313, 50-368, 50-416, 50-458, 50-382

2. NRC Safety Evaluation dated February 12, 2009 (TAC Nos. ME0112, ME0113),

Calvert Cliffs Nuclear Power Plant, Unit NOS. 1 and 2. Docket NOS. 50-317, 50-318.

3. NRC Safety Evaluation dated January 21, 2009 (TAC Nos. MD8744), Shearon Harris Nuclear Plant. Docket NO. 50-400.

Page 3 of 3

LR-N09-0104 ATTACHMENT 2 Salem Nuclear Generating Station, Unit No. 1 Facility Operating License No. DPR-70 NRC Docket No. 50-272 Relief Request SI-13R-92 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

Acceptable level of quality and safety

1. ASME Code Component(s) Affected Code Class: 1 Examination Category: B-P Item Number: B15.50 and B15.70

Description:

Class 1 pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval.

Unit/Inspection: Salem Unit 1/Third (3 rd) 10-Year Intervals Affected components consist of portions of Class i Reactor Coolant Safety Injection lines from Accumulators to Reactor Coolant Cold Legs. Affected portions of Safety Injection ten inch piping are between valves (XX-SJ-54) to and including the first-off (XX-SJ-56) and second-off check valves (XX-SJ-55) from the Reactor Coolant System. Affected components also include portions of six inch Safety Injection piping downstream of check valves (XX-SJ43) to ten inch piping tie-in between check valves (XX-SJ-56) and (XX-SJ-55) and two inch Safety Injection piping downstream of check valves (XX-SJ-144) to piping tie-in between check valves (XX-SJ-56) and (XX-SJ-55). See Figure 1 for affected components in each of the four reactor coolant loops.

2. Applicable Code Edition and Addenda

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection and Testing of Components of Light-Water Cooled Plants," 1998 Edition through 2000 Addenda for system leakage tests conducted at or near the end of each inspection interval. For Unit 1 the third interval began on May 19, 2001 and will end on May 20, 2011.

Salem Unit 1 Inservice Inspection Program LR-N09-0104 Relief Request S1-13R-92 10 CFR 50.55a

3. Applicable Code Requirement

ASME Section Xl IWB-5221 (a) states, "The system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100%

rated reactor power." and ASME Section XI IWB-5222 (b) states, "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."

4. Reason for Request

In order to obtain the pressure corresponding to 100% rated reactor pressure a jumper (temporary connections) would have to be installed between the reactor coolant system (RCS) and the volume between the first-off check valves and accumulator isolation valves. This lineup is not allowed by Technical Specifications (all vents and drains are required to remain closed) in Mode 3, the mode the RCS would have to be in to be at the required pressure. The volume of pipe must be otherwise pressurized using hydrostatic testing pumps; this would result in excessive dose, unnecessary special test procedures and unnecessary expenditure of plant resources during the ascension to power phase followinga refueling outage. Testing in accordance with IWB-5221 (a) is not required for an adequate level of quality and safety because the associated components are designed to the full pressure rating of the RCPB. Additionally, these segments are isolated from the full RCS pressure during normal operations and are subject to ASME Code required VT-2 (Visual) inspections which are performed each refueling outage. These inspections would be sufficient to identify structural defects.

5. Proposed Alternative and Basis for Use Pursuant to IOCFR50.55a(a)(3)(i), PSEG requests authorization to utilize the alternative requirements in ASME Code Case N-731 in lieu of the requirements of IWB-5221(a). Code Case N-731 will be used for portions of Class 1 systems that are continuously pressurized during an operating cycle by statically pressurized safety injection systems. Code Case N-731 states that for portions of Class 1 safety injection systems that are continuously pressurized during an operating cycle, the pressure associated with the statically-pressurized safety injection systems may be used in lieu of the pressure corresponding to 100% rated reactor power. The test pressure will be 650 psig, corresponding to the minimum operating pressure for the affected components. Affected piping is bound by 10" valves XX-SJ-54, XX-SJ55, XX-SJ56, 6" valves XX-SJ43 and 2" valves XX-SJ-144. Reference Figure 1.

Some portions of the Class 1 pressure retaining boundary cannot be pressurized to the pressure corresponding to 100% rated reactor power Page 2 of 4

Salem Unit 1 Inservice Inspection Program LR-N09-0104 Relief Request S1-13R-92 10 CFR 50.55a without installation of jumpers or other extraordinary means which could result in entering system valve lineups not authorized by plant technical specifications for Mode 3. This Request for Alternative is for Class 1 portions of safety injection systems that are continuously pressurized during an operating cycle (i.e., the portion of the safety injection system between the first-off check valves and accumulator isolation valves from the RCS which are maintained pressurized during the operating cycle by the safety injection accumulators).

The proposed alternative provides acceptable level of quality and safety because these sections of piping are continuously pressurized. Therefore, adequate time exists for leakage to be identifiable at the lower pressure the safety injection accumulators provide. Additionally, the level and pressure of the safety injection accumulators are continuously monitored and any leakage identified from the safety injection accumulators would be investigated and identified in accordance with station operating procedures.

6. Duration of Proposed Alternative The duration of the request for proposed alternative is for the Salem Unit 1 refueling outage 20 scheduled for the spring of 2010.

7. Precedents In the Referenced safety evaluation, the NRC authorized Millstone, Units 2 and 3, alternative to use Code Case N-731 for testing of the piping segments.

8. Reference NRC Safety Evaluation dated September 27, 2007 (TAC Nos. MD2866, MD2867), Millstone Stations, Units 2 and 3, Docket Nos. 50-336, 50-342 Figure 1 Page 3 of 4

Salem Unit 1 Inservice Inspection Program LR-N09-0104 Relief Request S1-13R-92 10 CFR 50.55a Figure 1 X-WJ-144 Page 4 of 4