ML12125A279

Inservice Inspection Program Fourth Ten-Year Interval 10 CFR 50.55a Request No. RR-2-4

ML12125A279
Person / Time
Site:	Kewaunee
Issue date:	05/03/2012
From:	Jordan A Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
12-324
Download: ML12125A279 (131)

Text

Dominion Energy Kewaunee, Inc. 9@FrO.O N490 Hwy 42, Kewaunee, WI 54216-/ [*]uuu )Ju' Web Address: www.dom.com May 3,.2012 ATTN: Document Control Desk Serial No.12-324 U. S. Nuclear Regulatory Commission LIC/JG/RO 11555 Rockville Pike Docket No.: 50-305 Rockville, MD 20852-2738 License No.: DPR-43 DOMINION ENERGY KEWAUNEE. INC.

KEWAUNEE POWER STATION INSERVICE INSPECTION PROGRAM FOURTH TEN-YEAR INTERVAL 10 CFR 50.55a REQUEST NO. RR-2-4 Pursuant to the provisions of 10 CFR 50.55a(a)(3)(ii), Dominion Energy Kewaunee, Inc.

(DEK) hereby requests NRC approval of'the attached proposed 10 CFR 50.55a request (RR-2-4) for the Fourth Ten-year Interval of the Inservice Inspection Program for Kewaunee Power Station (KPS). This 10 CFR 50.55a request proposes a temporary deviation from the requirements of ASME Section XI, Appendix IX, Article IX-1000, Paragraph (c)(2), which prohibits the use of clamping devices on "... portions of a piping system that forms the containment boundary" and ASME Section XI, Appendix IX, Article IX-6000(a), which states that the area immediately adjacent to the clamping device shall be examined using a volumetric method.

KPS is currently in a refueling outage. The reactor has been refueled and the reactor vessel has been reassembled. The plant is currently in MODE 5 - Cold Shutdown with the residual heat removal system in operation. Per KPS Technical Specification (TS) 3.4.7, "RCS Loops - MODE 5, Loops Filled," one residual heat removal (RHR) loop is required to be operable and in operation; and either one additional RHR loop shall be operable, or the secondary side water level of at least one steam generator shall be greater than or equal to 5%. Recently, a leak was discovered at a socket weld in a 3/4-inch line that is common to both RHR loops rendering both loops of RHR inoperable.

During activities to install a leak-limiting device over this leak, a welder inadvertently created a very small through-wall perforation of the 3/4-inch line. Both leaks exist in a 3/4-inch line that is common to both RHR loops. Neither leak can be repaired without removing both RHR loops from service.

DEK is proposing to perform a temporary alternate repair of the RHR piping by maintaining the installed leak-limiting device over the socket weld leak (and its associated structural restraint, which serves as an added measure of safety to prevent a catastrophic separation of the 3/4-inch line above the leaking sockolet); and installing a second (similar) leak-limiting device on the newly created through-wall perforation.

These activities will ensure that containment integrity and structural integrity of the RHR system is maintained prior to proceeding from MODE 5 to MODE 4. This temporary alternate repair will remain in place until the section of pipe containing the leaks is

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Page 2 of 3 repaired. Repairs will be pursued expeditiously. Once in MODE 4, isolating and repairing the portion of the RHR system with the leaks will take approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, thereby eliminating the need for the leak-limiting devices.

To verify that the newly created through-wall perforation caused by the welder was not exacerbated by service related degradation (e.g., erosion or thinning), the accessible piping surrounding this through-wall perforation in the 3/4-inch line has been evaluated and found to be structurally sound. The 3/4-inch piping surrounding the through-wall perforation has been inspected using straight beam UT techniques and determined to be at or near nominal thickness (0.113 inches).

There is no acceptable alternative to performing the temporary repair since the affected RHR line cannot be isolated in the current plant condition. Permanent repair of the piping leaks would first require transitioning to a different mode. MODE 4 would be the optimal mode as it ensures two trains of decay heat removal and one train of ECCS injection while maximizing RCS water inventory. The alternative to going to MODE 4, is to return the plant to the refueling mode (MODE 6), remove the reactor-head, remove the upper core internals, and offload the core into the spent fuel pool. This option would result in an undue hardship and unusual difficulty without a compensating increase in the level of quality and safety, and is therefore justified under 10 CFR 50.55a(a)(3)(ii).

Since the piping remains seismically qualified, system leakage will be maintained within the current licensing basis; and since the clamp is structurally equivalent to the piping, there is no expected additional risk of pipe failure.

The details of 10 CFR 50.55a Request No. RR-2-4 are provided in Attachment 1 to this letter. Information on maneuvering the plant, isolating the leaks and performing the permanent repair (along with contingency measures) is provided in Attachment 2. The two temporary modification packages that will be used to perform the alternate repair are provided in Enclosures 1 and 2.

If you have questions or require additional information, please feel free to contact Mr.

Craig Sly at 804-273-2784.

Very truly yours, Site Sresident - Kewaunee Power Station

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Page 3 of 3

Attachment:

1. Kewaunee Power Station Fourth Ten-year Interval Inservice Inspection Program 10 CFR 50.55a Request No. RR-2-4

2. Repair Activities and Contingency Actions Description, RHR Piping Flaws

Enclosure:

1. Temporary Modification Package 2012-11 (Revision 4)

2. Temporary Modification Package 2012-12 (Revision 1)

Commitments made by this letter:

1. Following installation of the devices and until the system is isolated, a VT-2 examination will be performed and repeated a minimum of once every twelve hours, and leakage observed from the devices will be evaluated.

2. The sealant injection pressure and volume will be controlled by work instructions and procedures to ensure the sealant is not injected into the RHR system piping.

3. DEK will implement the contingencies described in Attachment 2 of this letter.(in the section titled "Contingency Actions During Repair Activities"), while in MODE 4, until ASME Code repairs to the affected piping are completed.

cc: Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Mr. Karl D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No.12-324 ATTACHMENT 1 FOURTH TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM 10 CFR 50.55a REQUEST NO. RR-2-4 KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 1 of 9 Kewaunee Power Station Fourth Ten-Year Interval Inservice Inspection Program 10 CFR 50.55a Request No. RR-2-4 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 COMPONENTS AFFECTED ASME Code,Section XI Code Class 2 Residual Heat Removal*(RHR) system 3/4-inch Sockolet to Valve RHR-600. Pipe is 3/4-inch schedule 40, ASTM A312, type 304 sample line. Fitting is 0.750-inch on 10-inch Sockolet, 3000 Ib, ASTM A182 F 304.

Code of record is USAS B31.1 - 1967.

RHR design temperature and pressure: 400 OF and 600 psig.

2. APPLICABLE CODE EDITION AND ADDENDA ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition, 2000 Addenda

3. APPLICABLE CODE REQUIREMENTS ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition, 2000 Addenda, IWA-4133 states that mechanical clamping devices used to replace piping pressure boundary shall meet the requirements of ASME Section XI, Appendix IX.

o ASME Section Xl, Appendix IX, Article IX-1000(c)(2) states that clamping devices shall not be used on portions of a piping system that forms the containment boundary.

o ASME Section Xl, Appendix IX, Article IX-6000(a) requires a plan for monitoring defect growth in the area immediately adjacent to the clamping device.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 2 of 9

4. REASON FOR REQUEST Currently, Kewaunee Power Station (KPS) is in MODE 5 - Cold Shutdown, and has declared both trains of residual heat removal (RHR) inoperable due to two through-wall leaks. One leak is on a.3/4-inch socket weld connection of ASME Code Class 2 piping in the common RHR pump A and B discharge piping (Note: Installation of a leak limiting device on this leak had successfully isolated leakage through this flaw). While completing installation of the leak limiting device on this leak, a newly created through-wall perforation on a section of this 3/4-inch RHR sample line was created when attempting to create a fillet weld as an added measure of safety to prevent a catastrophic separation of the 3/4-inch line above the leaking sockolet.

Although considered inoperable for purposes of Technical Specifications (TS),

compliance, the RHR system is currently providing decay heat removal for the reactor coolant system (RCS).

TS 3.4.7, "RCS Loops - MODE 5, Loops Filled," is applicable during MODE 5 with the RCS loops filled. TS LCO 3.4.7 requires one RHR loop to be Operable and in operation. In addition, TS LCO 3.4.7 requires one additional RHR loop be operable (but not necessarily in operation) or the secondary side water level of at least one steam generator shall be greater than or equal to 5%. Currently both loops of RHR are declared inoperable due to the through-wall leak at the 3/4-inch socket weld and due to the newly created through-wall perforation on a section of this 3/4-inch piping. Both loops of RHR are available to provide decay heat removal. Additionally, if both RHR loops become unavailable, two SGs are available to provide decay heat removal as well as feed and bleed with the Safety Injection system (SI).

In order to implement a permanent weld repair (ASME Code repair) for the 3/4-inch socket weld and for the newly created through-wall perforation on a section of this 3/4-inch piping, both RHR cooldown loops must be removed from service and isolated from the RCS. In order to remove both RHR cooldown loops from service, one of two options must be performed.

Option one is to return the plant to the refueling mode, remove the reactor head, remove the upper core internals, and offload the core into the spent fuel pool. Then, the RHR system could be isolated and drained to allow repair of the affected piping. This option would require maneuvering the plant from its current operating condition in MODE 5 with two steam generators (SG) available to provide an alternate method for decay heat removal to a "no-MODE" condition. This option would result in an undue

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 3 of 9ý hardship and unusual difficulty without a compensating increase in the level of quality and safety of the plant. This first option would require the following actions/conditions:

" RCS cooldown from current plant conditions, thus losing the SG decay heat removal capability, and entry to MODE 6 - Refueling.

• The RCS would have to be drained to 6-inches below the reactor vessel flange resulting in a reduction in RCS inventory,, and a shorter time-to-boil if decay heat removal were lost.*(ICCE)

" The reactor head would have to be disassembled and detensioned.

" The reactor head (heavy load) would have to be removed and the reactor cavity flooded to a level of 23 feet.*(ICCE)

• The reactor vessel upper internals (heavy load) would have to be removed..

" The reactor core would have to be offloaded to the spent fuel pool.*(ICCE)

" The estimated duration of this evolution from the start of cooldown to core offload is approximately 8 days.

" The estimated radiation dose for this overall reactor disassembly, core offload, and subsequent reload and reactor reassembly evolution is approximately 8 Rem based on actual exposure measured during the same activities conducted during the current ongoing refueling outage.

• The above noted items are Infrequently Conducted Complex Evolutions (ICCE)

Option two is to perform a temporary alternate repair of the RHR piping by installing a leak-limiting strong-back device on both the socket weld leak and on the newly created through-wall perforation (caused during welding of the original leak limiting strong-back device). These temporary alternate repairs will ensure structural integrity of the affected piping. This activity is needed to return the RHR system to operable. These temporary alternate repairs will remain in place until the unit achieves MODE 4. After the unit reaches MODE 4, core cooling is provided by the reactor coolant pumps circulating water through the core and to the steam generators. With core cooling being provided by the steam generators, both loops of RHR cooling can be secured, the affected piping isolated, and the piping repaired.

Once in MODE 4, TS 3.6.1, "Containment," LCO 3.6.1, requires that the containment is Operable. Containment integrity will be maintained in MODE 4 by the leak-limiting devices until such time that both loops of RHR can be secured and the affected piping isolated. The ASME Code repair will be pursued expeditiously. Once in MODE 4,

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 4 of 9 isolating and repairing the portion of the RHR system with the leak will take approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, eliminating the need for the leak-limiting devices.

A dose of 171 mRem has been accumulated to date during installation of the first leak limiting device and its associated structural restraint. A dose of 120 mRem is estimated for installation of the remaining temporary modifications and subsequent removal of all associated temporary modifications.

The mechanical clamping devices that will be used will comply with the applicable ASME Code requirements outlined in ASME Section XI, Appendix IX, with the exception that they will be located on piping that is considered a containment boundary. ASME Section Xl, Appendix IX, Article IX-1000, Paragraph (c)(2) prohibits the use of clamping devices on "... portions of a piping system that forms the containment boundary."

Therefore, in order to use the devices, DEK requires approval of an alternative to allow a temporary deviation from the requirements of Appendix IX, Article IX-1000, Paragraph (c)(2) in order to return the RHR system to an operable status for the purpose of performing the ASME Code repair.

Based on the discussion above, DEK requests NRC approval of an alternative to the repair requirements of ASME Boiler and Pressure Vessel Code, Section Xl, 1998 Edition, 2000 Addenda IWA-4133. Pursuant to 10 CFR 50.55a(a)(3)(ii), DEK is requesting approval to temporarily deviate from the requirements of Appendix IX, Article IX-1 000, Paragraph (c)(2) which prohibits the use of clamping devices on "... portions of a piping system that forms the containment boundary" and ASME Section Xl, Appendix IX, Article IX-6000(a), which states that the area immediately adjacent to the clamping device shall be examined using a volumetric method. This requested deviation is based on DEKs conclusion that compliance would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety.

5. PROPOSED ALTERNATIVE AND BASIS FOR USE The proposed alternative would allow use of leak-limiting strong-back devices on a piping system that forms part of the containment boundary. The devices accomplish two functions; limiting the leakage from the defects, and maintaining the structural integrity of the affected piping. The proposed alternative, for both the original socket weld defect and the newly created through-wall perforation, uses a leak-limiting device to ensure containment integrity, since this region of the RHR piping is a portion of the containment boundary. A sealant will be injected into leak-limiting enclosures (one each for the original socket weld defect and the newly created through-wall perforation) to

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 5 of 9 provide a temporary pressure boundary for the RHR system. The sealant, X-36, has a low concentration of halogens/chlorides, therefore it is safe for use on stainless steel.

The sealant injection pressure and volume will be controlled by Work instructions and procedures to ensure the sealant is not injected into the RHR system piping. The leak-limiting device on the original socket weld defect is mechanically fastened with clamps and bolts to the 10-inch diameter RHR pipe and the 3/4-inch diameter pipe is fastened to the leak sealant enclosure with set screws. As an added measure of safety, a structural restraint has been installed to physically restrain the 3/4-inch line. This restraint will prevent a catastrophic separation of the 3/4-inch line above the leaking sockolet. The leak-limiting device on the newly created through-wall perforation is mechanically fastened to the RHR-600 valve body and the 10-inch RHR pipe.

RHR-600 is an outside containment isolation valve for Containment Penetration 10, a Class 6 penetration. Penetration Class 6 is a system required to operate post-accident.

The design and operational criteria for penetration Class 6 isolation valves are governed by the functional requirements of the system. The isolation valves at Penetration 10 are not subject to the requirements of 10 CFR 50, Appendix J (reference 5).

The requirements for containment isolation (on the affected piping) will be satisfied by the leak sealant enclosures. Each enclosure is designed to RHR temperature and pressure requirements and ASME Section Xl, Appendix IX, Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary. The piping remains seismically qualified and the enclosures will prevent system leakage.

Application of the leak-limiting strong-back devices to maintain containment integrity and the structural integrity of the 3/4-inch line will ensure that the plant can transition from MODE 5 to MODE 4 and perform an ASME Code repair.

Each leak limiting strong-back device has been designed to accommodate thrust loads resulting from a complete failure of the welded connection of concern. A review of the piping stress analysis has been performed to ensure that the additional mass does not adversely affect the qualification of the existing system.

Following installation of the leak limiting strong-back devices, a VT-2 examination will be performed and repeated a minimum of once every twelve hours until a MODE change from MODE 5 to MODE 4 is satisfactorily completed and the resulting portion of residual heat removal piping needed to facilitate repair is isolated. Prior to making the change from MODE 5 to MODE 4, the situation will be re-evaluated if leakage is identified in the affected piping near RHR-600. No leakage is expected from the affected piping near

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 6 of 9 RHR-600 with the leak limiting devices installed. If leakage is detected, compliance with TS 3.4.13, "RCS Operational Leakage", shall be evaluated.

Original Socket Weld Defect ASME Section Xl, Appendix IX, Article IX-2000 states that, if the defect size cannot be directly determined, a conservative bound of the size shall be determined and documented. ASME Section XI, Article IX-6000 states that the area immediately adjacent to the clamping device shall be examined using a volumetric method. Visual examination identifies the existing leak size as a pinhole and has placed the defect at the toe of the socket weld to the 3/4-inch branch line to valve RHR-600. There is no other visual indication of degradation to the piping wall thickness. Therefore, DEK is requesting a deviation from Article IX-6000(a) in that no volumetric inspection in the area of the clamp will be performed on the 10-inch pipe or on the 3/4-inch pipe. The installation of the leak-limiting device also precludes volumetric inspection of the defect.

Therefore, the defect can be conservatively characterized as residing within the socket weld and any growth would be limited to the weld itself, effectively limiting the impact of the defect to that of a circumferential crack. The proposed alternative, as a conservative measure to account for nondestructive examination limitations of the sockolet, includes two aspects; installation of a leak-limiting strong-back device which includes a restraint on the 3/4-inch diameter pipe, and a VT-2 examination a minimum of once every twelve hours until the system is isolated for repair. The device will conservatively maintain structural integrity of the affected components during the duration of the proposed alternatives.

System operating loads and installation loads from sealing of the newly created through-wall performaton will have inconsequential impact on the original socket weld defect. The strong back device is designed per ASME Section Xl Appendix IX to accommodate complete ejection.

Newly Created Through-Wall Perforation During installation of the leak-limiting strong-back device on the original socket weld defect discussed above, a welder inadvertently created a very small through-wall perforation of the 3/4-inch line.

The accessible portions of piping in the area of the newly created through-wall perforation have been inspected using straight beam ultrasonic testing techniques.

These inspections confirmed that the 3/4-inch piping above the leak sealant enclosure to the bottom of valve RHR-600 is structurally sound and that the wall thickness is

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 7 of 9 consistent with the specified nominal piping thickness of 0.113 inches. No anomalies were detected during the ultrasonic inspection. Visual inspection has characterized the newly created through-wall perforation as having an indistinguishable axial length and a circumferential length of approximately 3/64-inch. In this situation, the perforation is known to have been created during maintenance and is not due to aging during operation of the plant. The location of the perforation is on 3/4-inch piping base metal.

The estimated length of the attempted weld is 0.25 inches.

The sealant box prohibits straight beam examination and visual examination of the 3/4-inch piping within the box which is on the bottom side of the perforation. An aspect ratio of 6:1 is used to conservatively bound the lower extent of the perforation contained in the sealant box, which is consistent with ASME Section Xl, Non Mandatory Appendix L.

Using an aspect ratio of 6:1 is equal to 0.678 inches in the axial direction. It is not necessary to assume additional degradation in the circumferential direction because restraining clamps have been manufactured and installed to mitigate the possibility of ejection.

Thus, the size of the perforation is conservatively bounded as follows:

" Depth = 0.113 inches (nominal specified piping thickness)

" Circumferential Length = 0.25 inches (the area of weld in contact with 3/4-inch pipe)

" Axial Length = 0.928 inches (based on the size of the weld in contact with the 3/4-inch pipe and sealant box (0.25 inches) and an aspect ratio in the axial direction of 6:1 (0.678 inches))

It has been determined that a 1.6 inch axial flaw is acceptable to maintain structural integrity.

There are two (2) possible degradation mechanisms that could cause the perforation to grow in size, both of which are related to service time. The first mechanism is fatigue.

The second mechanism is stress corrosion cracking. Perforation growth caused by erosion due to leaking water is considered negligible because the pressure is low and stainless steel is resistant to flow accelerated corrosion wastage mechanisms.

Once in MODE 4, isolating and repairing the portion of the RHR system with the leak will take approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Permanent repair of the affected piping will be performed in an expeditious manner. During this time, the size of the defects will have negligible fatigue crack growth because the number of fatigue cycles that could occur during this time period is low or nonexistent. Similarly, the short duration will be insufficient to initiate flaw growth due to stress corrosion cracking, as the environment

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 8 of 9 will not change significantly. Additionally, there have been no cases of stress corrosion cracking on stainless steel base metal at KPS.

The above described approaches will provide the safest and most expeditious method to complete an ASME Code repair of the affected piping given the current condition of the plant.

6. DURATION OF PROPOSED ALTERNATIVES This alternative would be applicable for the period of time it takes KPS to return the RHR system to Operable status until entry into MODE 4 and ASME Code repair of the affected piping is completed. Once in MODE 4, isolating and repairing the portion of the RHR system with the leak will take approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, eliminating the need for the leak-limiting devices.

7. PRECEDENTS Dominion is currently aware of three (3) situations (see references 1, 2, 3 and 4) where similar alternatives have been approved to facilitate repair of ASME Section Xl piping that forms the containment boundary.

8. REFERENCES

1. Letter from M. L Marchi (WPSC) to NRC, "Relief Request RR-2-1 to Allow Continued Plant Operation with Two Pin Hole Leaks in a 3/4 inch ASME Code Class 2 Chemical Injection Weldment," dated August 12, 1996. [ADAMS Accession No. ML1I11810480]

2. NRC SER, "Kewaunee Power Station - Approval of a Relief Request from the requirements of 10CFR50.55a for Repair of 3/4-inch ASME Code Class 2 Chemical Injection Weldment (TAC No. M96273)," dated September 13,1996.

3. Letter from M. L. Marshall (NRC) to J. A. Stall (Florida Power and Light Co.),

"Turkey Point Nuclear Plant, Unit 4 - Safety Evaluation for Relief Request Regarding Mechanical Clamping Device on Pressure Boundary Piping (TAC No.

MC7338)," dated August 15, 2005. [ADAMS Accession No. ML052090182]

4. Letter from R. J. Laufer (NRC) to M. Kansler (Entergy), James A. Fitzpatrick Nuclear Power Plant - Relief Request for Temporary Non-Code Repair of a

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 1, Page 9 of 9 Shutdown Cooling Pipe (TAC No. MC7544)," dated August 9, 2005. [ADAMS Accession No. ML052070047]

5. Letter from Darrell G. Eisenhut, NRC Director Division of Licensing, to C.W.

Giesler, Wisconsin Public Service Corporation, "Exemption to Certain 10 CFR 50 Appendix J Requirements," dated September 30, 1982.

Serial No.12-324 ATTACHMENT 2 REPAIR ACTIVITIES AND CONTINGENCY ACTIONS REPAIR OF RESIDUAL HEAT REMOVAL PIPING FLAWS Including three plant drawings:

RHR Cooldown Lineup RHR Injection Lineup RHR Split Train Lineup KEWAUNEE POWER STATION.

DOMINION ENERGY KEWAUNEE, INC.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 1 of 8 Kewaunee Power Station Repair Activities and Contingency Actions Repair of Residual Heat Removal Piping Flaws 2

BACKGROUND Current Plant Status The reactor is in MODE 5, with both residual heat removal (RHR) pumps operating in cooldown mode (see enclosed drawing). The reactor coolant pumps (RCPs) are both operating with steam generator (SG) levels in their normal operating band. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.7, "RCS Loops - Mode 5, Loops Filled", is not met. Current reactor coolant system (RCS) temperature is being held below 200'F and RCS pressure is being maintained at < 380 psig.

No leakage is expected from the affected section of piping by valve RHR-600.

System Description

The RHR system can be aligned in any of three configurations:

• Common cooldown lineup;

• Emergency core cooling system (ECCS) injection lineup; or

" Split train lineup (RHR Train A aligned for cooldown and RHR Train B aligned for ECCS injection).

In RHR cooldown lineup, both trains of RHR take suction from a common line connected to the RCS via a parallel setof valves connected to the two RCS loop hot legs. The common line supplies the Train B safety injection (SI) accumulator injection line via valve RHR-1 1 (see enclosed drawing).

When in the ECCS injection lineup, both trains are 100% independent, with each train taking suction from the refueling water storage tank (RWST) and injecting into the reactor vessel head (see enclosed drawing).

When in split train lineup, RHR Train A is aligned to take suction from the RCS hot legs and inject via the RHR-1 1 flow path. RHR Train B is aligned to take suction from the RWST and inject into the Reactor Vessel Head (see enclosed drawing). After

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 2 of 8 transitioning into MODE 4, the RHR system will be in the split train lineup prior to initiation of repairs.

Technical Specification Applicability NRC approval of Request RR-2-4 will allow both trains of RHR to be declared operable but nonconforming, upon acceptance of operability determination OD-481, for operation in MODE 5. At this point, at least one train of RHR will be in the cooldown lineup, with both SGs and RCS loops Operable. This meets the requirements of TS LCO 3.4.7,

"'RCS Loops - Mode 5, Loops Filled", which requires one RHR Loop Operable and in operation and.either one additional RHR loop Operable or secondary side water level of at least one SG greater than 5%.

To transition the unit above 200°F (into MODE 4), applicable Technical Specifications must be met (as directed by mode change checklists). Once in MODE 4, LCO 3.4.6, "RCS Loops - MODE 4", can be met by having two loops of RCS Operable and one in operation. This will be accomplished by having both SGs in their required range and both RCPs operable and at least one operating.

In addition, applicable TS LCOs 3.7.4, 3.7.5 and 3.7.6 will be met as they relate to SG heat sink.

In this configuration, RHR Train B will be aligned for ECCS injection to satisfy LCO 3.5.3, which requires one ECCS train to be operable. This alignment will leave RHR Train A in the cooldown alignment and maintain two RHR suction flow paths Operable for LTOP to satisfy LCO 3.4.12.

Containment integrity, as required by LCO 3.6.1 and 3.6.3, will continue to be met based on the leak sealant enclosures (coupled with NRC approval of Request RR- 2-4).

During maintenance activities in MODE 4 to implement the permanent repair, containment integrity will be ensured following the requirements of LCO 3.6.3.

Pressure and Temperature Range for Proposed Repair Upon NRC approval of Request RR-2-4 (with the leak limiting strong-back devices installed), the unit will be placed in MODE 4, the affected piping will be depressurized and isolated. RCS leakage will be monitored. No RCS leakage is expected into the affected section of piping and the ASME Code repair will commence.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 3 of 8 While in MODE 4, RHR Train A will be cooled down via recirculating the RHR loop after it is isolated per normal operating procedures. RCS temperature will be maintained less than 350°F and RCS pressure will be maintained less than 380 psig via the RCS Loops in operation.

PREVENTING PRESSURE BOUNDARY LEAKAGE Maintenance operating procedures will be followed to isolate RHR Train A for repair of the affected piping by valve RHR-600.

The initial system isolation will be made at the following valves:

" Discharge of the RHR Train A heat exchanger (RHR-9A, manual isolation);

• RHR train cross connect (RHR-1OA and RHR-10B, manual isolation);

" RHR recirculation line (RHR-500A, manual isolation valve);

" Letdown inlet (LD-60, motor operated valve located in containment);

" RCS injection (RHR-1 1, motor operated valve located in containment); and

" Reactor Vessel Injection (SI-302A, motor operated valve located in containment).

Additional isolation is as follows:

" Check valve SI-22B for the RCS injection line ; and

" Check valves SI-304A and SI-303A for the reactor vessel injection.

Of note is that the drain path for the repair activities by valve RHR-600 is at a high point in the system. This will require evacuating the water at the weld line when repairing the affected piping. This condition also allows a water column of approximately 12 feet between RHR-600 and inside containment valves RHR-1 1 and SI-302A, which provides additional protection of the containment boundary.

Valve LD-60, letdown line connection, will be closed. Thisline is within the test boundary for outboard containment isolation valve LD-6 on the letdown line. This line has been leak tested per the Local Leak Rate Testing (LLRT) Program and has successfully passed its test during this current outage.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 4 of 8 The affected portion of piping will be depressurized and water level will be monitored.

To monitor and minimize potential leakage outside of containment when the RHR system is opened, leakage past SI-22B and SI-304A and SI-303A will be monitored using open vent valves in containment, which are on the downstream side of RHR-1 1 and SI-302A respectively.

If leakage into the affected section of piping develops while the piping is breached for repair activities, TS LCO 3.4.13 would be evaluated. Although response contingencies are planned, leakage is not expected because of the passive nature of the isolation boundaries and the positive means in place to prevent inadvertent repositioning (tag out, breakers racked out, etc.).

Sequencing of the procedure will ensure that RHR-1 1 and SI-302A will remain closed.

RHR-600 BOUNDARY ISOLATION VALVES LD-60 RHR to CVCS Letdown Line LD-60 isolates the RHR system from CVCS Letdown system. LD-60 is a 2" motor operated globe valve inside containment. This valve is part of the test boundary for the Local Leak Rate Testing (LLRT) of valve LD-6 which has been successfully tested during both the current and previous refueling outages. Although the differential pressure that will exist across LD-60 during the RHR-600 repair efforts will be in the opposite direction of the differential pressure that was present during the LLRT, the LLRT result is still considered to have demonstrated the leak integrity of LD-60.

KPS has historically experienced. RWST Inleakage due to seat leakage past LD-60.

Following valve repairs on LD-60 during KR-30 and removal of RHR-44 and RHR-45 during KR-31, KPS did not experience RWST Inleakage prior to entering KR-32, indicating that LD-60 is no longer leaking.

Based on the review of operating performance and maintenance history for LD-60, boundary leakage across LD-60 is not expected to occur.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 5 of 8 RHR-11 I RHR Discharge to RCS SI-302A RHR Injection to Reactor Vessel RHR-1 I is a 10"X8"X1 0" double-disc motor operated gate valve inside containment. RHR-1 1 provides the flow path for Decay Heat Removal during refueling and provides isolation between' the RCS and RHR systems, along with check valve SI-22B, during operations. A seat leakage test is not performed for RHR-1 1. Check valve SI-22B has a history of seat leakage during both testing and operation. RCS leakage past SI-22B and RHR-1 1 into the RHR system would be evident by RWST Inleakage, similar to that experienced when LD-60 was leaking. As stated above, KPS did not experience RWST Inleakage during the cycle prior to KR-32, indicating that boundary leakage across RHR-1 1 is not expected.

SI-302A is a 6" flex wedge motor operated gate valve inside containment. SI-302A is open during normal operation and is closed when RHR Train A is aligned for Decay Heat Removal. There are two check valves (SI-303A and SI-304A) downstream of SI-302A, which are leak tested each refueling outage. Test performance for SI-303A and SI-304A has historically shown low leakage values. In addition, SI-302A has successfully performed as a boundary valve for maintenance performed during KR-31 to remove RHR-44 and RHR-45. Based on the combination of two check valves and a closed motor operated valve in series and prior boundary valve performance for SI-302A, leakage across SI-302A is not expected to occur.

Based on 1) review of operating performance and maintenance history for RHR-1 1 and SI-302A; 2) limiting the acceptable leakage across the upstream check valves to less than 5 gpm; and, 3) if needed, maintaining the vent valves open during the repair to limit the differential pressure across the boundary valves; boundary leakage 'across these valves is not expected to occur.

RHR-9A RHR Heat Exchanger A Outlet RHR-9A isolates the RHR system at the outlet of the A RHR Heat Exchanger. RHR-9A is a 8" manual split wedge gate valve outside containment. Past operating history for this valve relative to leakage is good as demonstrated by its isolation performance as the upstream isolation for the RHR system modification work performed in 2011

[removal of RHR-44 and RHR-45]. The modification work performed in 2011 involved the same RHR piping segment (from RHR-9A to RHR-1 1) as will the RHR-600 repair work. Based on prior boundary valve performance for RHR-9A, leakage across RHR-9A is not expected to occur.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 6 of 8 RHR-500A RHR Train A recirculation line RHR-500A is a 2" manual globe valve outside containment. RHR-500A isolates the RHR system from the recirculation line back to the A RHR pump suction. RHR-500A has successfully performed as a boundary valve for numerous maintenance activities on RHR Train A, most recently during KR-32 when SI-351A and SI-351 B were replaced.

Based on prior boundary valve performance for RHR-500A, leakage across RHR-500A is not expected to occur.

RHR-10A and RHR-10B - RHR train cross connect RHR-1 OA and RHR-1 OB are 8" manual split wedge gate valves outside containment that separate RHR Trains A and B downstream of the RHR Heat Exchangers and Flow Control Valves (RHR-8A and RHR-8B). RHR-1OA and RHR-IOB are closed during normal operation and are closed once RHR is aligned for split train mode. Similar to RHR-9A, these valves were boundary valves for the RHR system modification work performed in 2011 [removal of RHR-44 and RHR-45]. Based on prior boundary valve performance for RHR-1OA and RHR-1OB, leakage across RHR-1OA and RHR-1OB is not expected to occur.

SEQUENCE FOR PERMANENT REPAIR Sequence of Activities The following steps provide the general overall sequence for the ASME Code repair activities once authorization is obtained to start. Specific steps are included in work order instructions.

• Removal of the stainless steel tubing from the downstream side of RHR-600.

" Removal of the leak limiting strong-back device from the piping, including the two leak sealant enclosures.

• Removal of the sealant from the piping in the area of the sockolet.

" Cleaning the pipe in the area of the sockolet (weld preparation).

" Removal of the 3/4-inch pipe from the sockolet by grinding out the fillet weld.

" Preparation of the sockolet for welding and performing QC cleanliness inspection.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 7 of 8

• Performing cleanliness inspection and non-destructive examination of the prepared sockolet.

• Fitting up the new piping/valve assembly.

• Welding in the new piping/valve assembly.

- Performing visual and non-destructive examinations of the new fillet weld.

• Reinstalling the stainless steel tubing from the downstream side of valve RHR-600.

Contingency Actions During Repair Activities To ensure successful completion of the repair activities, various contingency measures will be established and implemented.

" Water level will be monitored during draining and repair activities to ensure weld capabilities.

" Although leakage is not expected, in the unlikely event leakage occurs from the RCS via RHR-1 1 and SI-302A, the vents in containment downstream of the isolation valve will be opened as described in Maintenance Operating procedures.

" A replacement valve and 3/4-inch piping will be utilized to complete the repairs. The replacement valve and piping will be prefabricated in the shop, including the 3/4-inch pipe to replacement valve weld. This will limit the number of field welds to a single weld (3/4-inch pipe to sockolet), minimizing the time the RHR piping system is open to the local environment.

A' maintenance mockup will be used to allow workers to demonstrate ability to perform the required actions and to validate estimated times for defect excavation, fit-up and welding of newsocket welds.

- Plant staff associated with repair activities during the period of time the RHR system is breached, including welders, nondestructive examination (NDE) staff, and quality control (QC) inspectors will be staged in the immediate vicinity to minimize delays.

Additionally, radiation protection (RP) personnel will provide continuous coverage during repair activities to minimize delays.

As an emergency measure, a temporary plugging device will be staged at the work location to mitigate the possibility of any leakage developing of sufficient magnitude to preclude completion of weld repair activities. This material would be of sufficient construction to restrict flow through the open section of affected piping. The temporary plugging device will be test fit utilizing a mockup to ensure proper fit and to orient the workers to its use.

Serial No.12-324 10 CFR 50.55a Request RR-2-4 Attachment 2, Page 8 of 8 FILLING AND VENTING RHR PIPING AFTER REPAIR Following completion of the ASME Code repair, filling and venting of the drained piping will be performed in accordance with Procedure MOP-RHR-010. This procedure will initially use the RWST as the fill source. Ultrasonic Testing (UT) will be performed to ensure the system is sufficiently full. An alternate fill and vent method is proceduralized using valve LD-60 and the letdown line.

CONTINGENCY ACTIONS FOR ABNORMAILITIES In the event abnormal conditions were to occur during the performance of these repair activities, operators would respond in accordance with the appropriate response procedures.

Operators are trained in shutdown loss of coolant events and shutdown loss of decay heat removal events. The system alignment during this repair allows operators to readily address both events. RCS makeup capability remains available via the charging system, the safety injection system, or via Train B of the RHR system. Core cooling remains available via both RCS loops and SGs if the RCS remains intact. Containment closure is covered in these abnormal operating procedures, including actuation of containment isolation. These actions are covered in Procedure AOP-RHR-001, Abnormal Residual Heat Removal System Operation, and Procedure AOP-RHR-002, Shutdown Loss of Coolant Accident.

Serial No.12-324 ENCLOSURE 1 FOURTH TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM 10 CFR 50.55a REQUEST NO. RR-2-4 TEMPORARY MODIFICATION PACKAGE 2012-11 (Revision 4)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Temporary Modification 0

2-IDDOMInela"..

Site9 Unit Year Temporary Modification Number Revision Number Work Order(s)

KPS 1 20-12 2012-11 4 KW1066940*6 KW1 00894787 KWI00895870 I High Risk? C3 YES Z NO Provide information as required and attach additional sheets as necessary for each item.

PartA. TM Description (To be completed by Requestor/Originator/Engineering)

1. Title RHR-600 Leak Repair

2. Expiration Date [Not to exceed one refuel cycle (unless approved by site VP)]

KR32 (Remove prior to entering Mode 3)

3. Affected Systems/Components/QA Class RHR (System 34)/RHR-600/SR

4. Reason (e.g., awaiting parts, testing, calibration, repairs, temporary power supply)

The 3/" socket weld upstream of RHR-600 has a pinhole leak. This TM performs a temporary repair of the leak by Injecting leak sealant into an enclosure around the leaking fitting to stop leakage. This TM will be installed in Mode 5 and removed prior to entering Mode 3. In addition, a NRC Relief Request is required before changing from Mode 5 to Mode 4 with this TM installed.

The additional flaw in the Y4" diameter pipe caused during TM installation has been evaluated and found nmt to strueturally impair the integrity of the /" pipe. The flaw was evaluated in both the axial and circumferential directions and concluded that the %"RHR pipe was stable and able to maintain Its structural integrity with the flaw. Therefore, this TM does not address the presence of this additional flaw.

REV 1 revised the section regarding Monitoring Requirements (IX-6000) to be consistent with the NRC relief request requirements. Also provided clarification to the system pressure requirements.

REV 2 will change the following:

1. Clarify that during sealant injection, the RCS temperature will not exceed 195 F. This was required to eliminate confusion regarding temperature'at the injection location and plant operating mode temperatures.

2. Add bolt torque values to the implementation sections. Values based on vendor installation prastlesa.

3. Clarify allowable sealant volume.

4. Enhanced discussion regarding thermal effects and constraints due te the Irnstailatieo ef the repair alump.

5. Revised Ref. 12 to Revision B for change to the 1ill Of Materials.

REV 5 will change the fellowing.,

The additional method 6f retaining the /" pipe will now be a structural restraining clamp rather than a 3/16" circumferential fillet weld between the leak sealant enclosure and the 3/"

pipe.

Form. No. YSWIAC4 lMw4 2011})

Temporary Modification DOM10111,60nm REV 4 changes the following:

1. Revises the acceptance criteria to 'no leakage'.

5. Description (e.g., specific details on the aspects of the modification): Work Order reference(s), and instrument index as applicable, and locations (e.g., racks, cubicles, building, area, elevation, and rooms to identify in detail the location of the modification).

Attach sketches as necessary.

See following pages

6. Documents/Drawings/Procedures to be Updated

a. Vendor procedure'for sealant iniection b.

C.

7. List any Mode Restrictions This TM can remain installed in Modes 4 and 5. It must be removed and the system restored prior to entering Mode 3.

8. Action Plan for Removal - Close-out Document (REA/DCP/DCR, Work Order, Procedure changes)

Repair leaking weld and remove TM per WO # KW1 00894787

9. Installation Instructions for TM Install leak sealant enclosure and inject leak sealant per approved procedure. Torque clamp bolting to 126 ft-lbs and strong back bolting to 75 - 90 ft-lbs. [16]

Isolate (tag shut) valve RHR-600.

Disconnect downstream tubing from RHR-600.

Install structural restraining clamp per Attachment 1 As required, reconnect downstream tubing from RHR-600 Tighten all clamp fasteners wrench tight When tagout is cleared, open RHR-600 as required

10. Required System Testing Following TM Installation Verify no leakage.

11. Removal Instructions for TM Per WO KW100894787

12. Required System Testing Following TM Removal Permanent repair of the leaking fitting will follow removal of the TM. Therefore, there are no Form No. 730749 (May 20U)

Temoorarv Modification

!"pDominion-removal test requirements for this TM.

13. Requested By (Name - Please Print) 14. leq eseBp (Signature) 15. Date Tim LaHann 05/02/2012 Part B. Design Engineering ReviewslScreening Evaluations Attached /

1. CM-AA-RSK-1 001, Engineering Risk Assessment, Attachments Z YES El[ NO 2 and 3

2. DNES-AA-GN-1002, Document Impact Summary, Attachment 1 Z] YES E. NO

3. DNES-AA-GN-1003, Design Effects and Considerations, 0 YES El NO Attachment 2

4. DNAP-3004, Dominion Program for 10 CFR 50.59 and 10 CFR 72.48 - Changes, Tests, and Experiments (KPS only - perform ONE IS REQUIRED GNP-04.04.01 for Applicability and Pre-Screening prior to DNAP-3004) 0 YES EI NO a 50.59/72.48 Screen, Attachment 4 rYES [] NO

• 50.59/72.48 Evaluation, Attachment 6

5. Additional Attachments a.

b.

C.

) d.

"6.Summary: Descrilptions & Conclusions Resulting from Reviews/Evaluationsý Performed in Step 3.1.8 and Above RSK-1001: The TMV is medium risk AA-GN-1002: A procedure for leak sealant injection is required AA-GN-1003: Components within the ISI boundary are affected.

60.69: The TMV screens out of 50.59

7. Additional Reviews and/or Comments - Provide Details Below None n

8. Prepared By (Print &Sign) -/9. Date

-rim LaHann I...*.,*.i?

10. Independent/Design Authority Review) ýaapial) 11. Date (Print & Sign)

Lord Christensen _ *. _ J-  ;/,

'A*dditional reviews for Implementing Organization, Training, etc. as necessary in accordance with Section 3.3

• 12. Name. (Print & Sign)/Department 13. Date

a. - -1, Form No.730749 (May 2021)1

Temnorarv Modification

P'Dommenieon"

b. 2.

C. 3.

d. 4.

14. Engineering Supervisor 15. Engineering Supervisor 16. Date (Name - Please Print) (Signature)

Part C. Operations Review (To be completred by Shift ManagerlDesignee)

1. Are controlled Station drawings affected by the TM and are [ YES D NO 0 NA they attached? I

2. Necessary Station personnel are informed of the TM? Z YES n NO F] NA

3. Temporary procedure changes and Temporary procedures Dl YES [1 NO Z'NA are implemented to support the TM?

4. Evaluated need for check valves and/or other anti-siphon [j YES LI NO Z NA protection, if TM utilizes piping or hoses?

5. Limiting conditions and special requirements indentified. Z YES NO [ NA Note: NRC Relief Request is required before mode change (5 to 4) with TM installed.

6. TM verified not to violate Technical Specifications, not to create a hazard to Station safety or personnel, or conflict with existing Station conditions? [R YES [I NO El NA Note: NRC Relief Request is required before mode change (5 to 4) with TM installed.

7. TM Tags generated in accordance with applicable Site Z YES El NO F] NA Tagging procedure and any special instructions included on the tags?

8. TM Log updated? I*[ YES El NO El NA

9. Engineering Post Installation Walkdown Requested? D] YES Z NO D1 NA

10. Quarterly walkdown required? DYES DNO I] NA

11. PRA risk impact has been assessed and, if determined to N YES L NO DNA be applicable, has been entered into the Risk Monitor and/or Shutdown Risk Assessment?

12. The affected Unit shall not exceed the operating mode of: 5 until NRC Relief Request is approved. Once the Relief Request is approved, the unit shall not exceed Mode 4.

13. Concurrent or Independent Verification: NA

14. FunctionalCheck or Visual Inspection: Per WO KW100894696

15. The opposite Unit shall not exceed the operating mode of: NA

16. STA Concurrence (Print & Sign) for Virginia Plants Only [Commitment 5.1.4] 17. Date NA Form No. 730749 (May 2011)

Temporary Modification I

P Dominiow AA -D 0 AT AH 3 P 5 o

18. Shift Manager/Designee (Prirjt & Sign) 19. Date Ethan Treptow 9-" -5/ 12 Part D. FSRC Review (as applicable, Refer to LI-AA-600)

NOTE: Ifthe TM is used to move radioactive fluids or gases, the Manager Radiological Protection or Radiological Protection alternate must be a member of FSRC NOTE: TMs which could affect Nuclear Safety must be reviewed by FSRC

1. Is FSRC Signature Required? Z YES E[ NO

2. FSRC Authorized Duration t ,4M,& '3 at

3. F8RC Chairman Approval4. ) / "}. Date a 7(P nro nt,:

Part E. TM Installation (To be completed by 'applicable personnel)

1. Shift Manager proval for TM Installation (Print & Sign) 2. Date

3. Shift Manager Comments (includes any additional requirements in accordance with Section 3.4, 'Implementation") - Provide Details Below

4. Installation Completed By (Name - 5. Installation Completed By 6. Date Please Print) (Signature)

7. Independently Verified By (Name - 8. Independently verified By 9. Date Please Print) (as applicable) (Signature) (as applicable)

Notify the Shift Manager that TM Is installed and the required post installation testing can be performed.

10. Instructions Used for Post Installation Testing: Per WO KW1 00894696

11. Testing Performed By 12. Testing Performed By (Signature) 13. Date (Name - Please Print)

14. Post Installation Testing Satisfactorily? [YES INO

15. Required Administrative Controls Established? [I YES EL NO

16. Is Engineering Post Installation Walkdown Required?

NOTE: Responsible Engineering signature is in accordance with IFYES [ NO Step 3.4.6

17. Responsible Engineer (Print & Sign) (Engineering walkdown completed) 18. Date NA FormN. 73W749 (May 20211

Temporarv Modification

- 2X)ýD o'mini 01r

19. Shift Manager (Name- Please Print) 20. Shift Manager (Signature) 21. Date Part F. TM Restoration (To be completed by applicable personnel)

1. Shift Manager Approval for TM Removal (Print & Sign) 2. Date ........

3. Shift Manger Comments (includes any additional requirements in accordance with Section 3.7, "TM Removal") - Provide Details Below

4. TM Restoration Completed By (Name - 5. TM Restoration Completed By 6. Date Please Print) (Signature)

7. Independently Verified By 8. Independently verified By 9. Date (Name - Please Print) (Signature)

Notify the Shift Manager that TM has been removed and the required restoration testing can be performed

10. Instructions Used for TM Restoration Testing

11. Testing Performed By 12. Testing Performed By 13. Date (Name - Please Print) (Signature)

14. Shift Manager (Print & Sign) 15. Date Part G. Post Restoration Review (Completed by Shift Manager/Designee following restoration)

1. Post restoration testing completed satisfactorily? [ YES [] NO

2. All Documentation satisfactorily completed? . YES LI NO

3. Temporary drawings removed from Control Room/any other El YES [ NO location?

4. Procedures changed to eliminate TM? (Review Part A) El YES [I NO

5. Necessary Station Personal notified? [D YES [L NO

6. TM Tags removed? Dl YES [1 NO

7. TM Log updated? D-] YES -INO

8. Shift Manager (Print & Sign) 9. Date Form No. 73D749 (May 2011)

Temporary Modification m mA_37A,1 ýg.

• Domm"Iffielollm Part H. Monthly Audit INITIAL DATE COMMENTS ON AUDIT 1."

2.

3.

4.

5:

6.

Part A, DESCRIPTION RHR-600 is a %"SR sample valve located off RHR line 1O-AC-601 R11. A leak repair enclosure, which incorporates a strongback design, will be installed to seal the leak. A structural clamp will also be installed to prevent the piping/valve from separating from the 10" RHR 1A HX discharge line in event of complete weld failure.

This is a short duration TM, which allows the unit to enter Mode 4 from Mode 5 for permanent repair of the fitting leak. Upon entry into Mode 4, the A train of RHR can be removed from service and the leaking weld repaired.

To install the structural restraint clamp, RHR-600 will be closed and the downstream tubing disconnected. During clamp installation, the tubing may be reconnected as required.

Quality Classification RHR-600 and associated pipe and tubing is SR [1]. All parts used for the leak sealant enclosure and structural restraint-clamp are SR, with the exception of the optional shim material, which may be non-safety related.

ASME B&PV Code RHR-600 and adjoining pipe is classified as ASME Section Xl Class 2 [2]. ASME Section Xl, Appendix IX provides direction on the use of mechanical dlamping devices on Class 2 and 3 piping pressure boundary. The requirements of this appendix will be used to demonstrate the acceptability of this repair on the leaking fitting.

Note that Article IX-1 000 prohibits the use of a clamping device on portions of a system that form the containment boundary. RHR-600 is credited with maintaining the containment boundary. and therefore the plant cannot change modes with this temporary modification installed without an approved NRC Relief Request.

Form No. 730749 (May 20111

Temporary Modification Auxiliary Cooling (AC) System Temperature and Pressure The design pressure and temperature of the piping, AC-601R-11, is 600 psig and 400 OF [5].

The leak sealant enclosure will be designed to these conditions.

The TM will be installed and the leak sealant injected in Mode 5. RCS pressure will be maintained greater than 340 psig. The operating pressure will be used to determine the injection pressure for the leak compound. This will assist in preventing injection of sealant into the system. Sealant injection will be performed In Mode 5. During leak sealant injection, the maximum RCS temperature will not exceed 1950 F [7]. The X-36 sealant is rated for injection from a temperature of 140OF up to 400 OF [15], and is therefore suitable for this application.

Overview The leak repair will be completed using a hub clamp assembly designed and fabricated by TEAM Industrial Services, Inc. This assembly incorporates a leak sealant enclosure Into a strongback clamp design. Two #10-24 set screws, tightened to 36 in-lbs; are included in the hub assembly. These set screws tighten onto the 3/4"pipe between the leaking sockolet fitting and RHR-600 and will prevent ejection of the 3/4"pipe and RHR-600 in the event of a full circumferential failure of the leaking W"sockolet weld. As an added measure of safety, a structural restraint clamp will. be installed above RHR-600 that will attach to the 10" diameter RHR piping to ensure structural Jntegrity (Attachment 1).

This structural clamp, in conjunction with the set screws, is more than adequate to restrain the 520 lbf separation thrust [12] and ensures that no catastrophic separation can occur due to a full circumferential failure of the W socket'weld.

A structural evaluation of the structural restraint clamp is attached to this temporary modification

[19].

The leak sealant enclosure is made of stainless steel and, along with the X-36 sealant, provides the pressure boundary up to the design temperature and pressure of the AC pipe, which is 600 psig and 400 0F. The studs, which secure the leak sealant enclosure, are tightened to 126 ft-lbs

[12].

General Design Requirements (IX-3100)

Defect Characterization:The defect is characterized as a pinhole.leak in the fillet weld of the W" sockolet upstream of RHR-600. The leak is documented in CRs 472654 and 472226. Since more accurate data on the defect size or potential for propagation is not available, this TM assumes catastrophic failure as characterized by full circumferential failure of the 3/4"sockolet weld. This failure will not result in ejection of any pipe or RHR-600 because of the restraint provided by the structural restraint clamp. The design of the structural restraint clamp is shown in Attachment 1. In the event of a full circumferential failure of the leaking W"sockolet weld, the rupture load would be transferred through the restraint clamp and into the 10" diameter RHR line.

Materialcompatibility [12]. The hub clamp is constructed of SA240 Gr 304/316 material. Studs and all-thread material are SA1 93 Gr B8, hex nuts are SA1 94 Gr 8, and set screws are ANSI B1 8.3 stainless steel. All sealant enclosure materials are acceptable for use with borated water Form No. 730749 (May 20111

Temporary Modification

~Dominion,*

in the AC system. The pipe coupling and arms of the structural restraint clamp are constructed of stainless steel, and are also acceptable for use Inthe AC system.

The Grinnell Figure 295 pipe clamps that attach the structural restraint clamp to the pipe are carbon steel. Stainless steel banding, as required, will be installed between the carbon steel clamps and the stainless steel 10" RHR pipe. While there may be some leakage of borated water from the %"pipe assembly until sealant is injected, this TM will be installed for only a short time. Therefore, potential for boric acid corrosion of the carbon steel is minimal and the potential clamp degradation from corrosion will be negligible.

TEAM Industrial Services compound X-36 will be injected into the enclosure to provide temporary repair of the system pressure boundary. Sealant X-36 has a low concentration of Halogens/Chlorides and is acceptable for use on stainless steel [8].

Defect size: The defect size assumed for the design of the sealant enclosure is a full circumferential failure of the leaking Y4"pipe socket weld. Per Reference 12, the leak sealant enclosure Is adequately designed for this application. These calculations establish the enclosure size and construction, as well as maximum sealant injection pressure fnd volume necessary to provide the system pressure boundary. Identifying maximum sealant Injection volume prevents over-injection and intrusion of the sealant into plant systems.

AddItional supports:The Installation of the enclosure adds a 71 lb point load [12] to line 10-AC-601 R1 I between hangers RHR-H9 and RHR-H55 [3]. The structural restraint clamp weighs approximately 57 lbs. Installation of the clamp will add two point loads, each approximately 28.5 Ibs, to this same line. Thus a total of approximately 128 lbs will be added to the pipe line. This load increase is less than 10% of the weight of the supported pipe. The structural qualification of the nearby supports is contained in calculation S-062-RHR-34-004 [18]. These supports (RHR-H9A, RHR-H9, RHR-H55, and RHR-H29) were qualified for loads that are 10% greater than the governing load case "to avoid future reanalysis for small load increases*. Therefore, no further evaluation of this load addition is required [4].

$y virtue of their relatively light weight and their being secured to the %"and 10" pipe, the leak sealant enclosure and the structural restraint clamp do not adversely affect the selsmic qualification of the pipe or adjacent equipment.

Clamplng Device (IX-3200)

As described above, the hub clamp Is constructed of SA240 Gr 304/316 material. Studs and all-thread material are SAI 93 Gr 1S, hex nuts are SA194 Gr 8, and set screws afe ANSI 151.3 stainless steel. The leak sealant enclosure is designed to meet Table IX-3200-1 Level A stress limits as well as meet the bounding pressure and temperature requirements of the Auxiliary Cooling System [(121.

Plonin System (LX-3300)

Leak Sealant

Enclosure:

The leak sealant enclosure will be secured rigidly to the 10-Inch pipe using a strongback design. The friction between the clamp and the pipe Is significant compared to the seismic forces of the relatively light weight clamp and enclosure. The clamp and enclosure are fabricated from SA-204 Gr 304/316 stainless steel and are well stiffened to resist the self weight seismic forces.

FormNa.730749 (Mv 2b1)

Temporary Modification Vibration: The pipe in question is at the' outlet to the RHR heat exchanger and is well supported near the repair area. Vibration is not considered to be the apparent eause of the defeat in the weld. Consequently, vibration was not considered in the design of the leak sealant enclosure.

PipingEvaluation/Stress:The proposed leak sealant enclosure and structural clamp Introduces a relatively small load to the pipe eonfiguratlen due to their total weight of approximately 128 lbls and the additional seismic forces associated with their weight multiplied by seismic accelerations. Seismic forces on this configuration will be resisted by the frictional force of the pipe clamps around the 10-inch pipe. The clamps' loads will be shared by two supports located 6 to 8 feet horizontally from the sample line's connection to this pipe. Refer to KPS Drawing M-962-2 for the piping and hanger configuration. The existing pipe stress analysis, KPS Calculation RHR-34-004, Indicates the stresses for the total span between the supports are relatively small on the order of less than half of the allowable stresses. Considering that the weight of the clamps is less than 10% of the weight of this span of 10 inch diameter pipe, and that the combined pipe stresses are low, It is reasonable to conclude that Installing this clamp would not have an adverse affect on the integrity of this piping system including design basis loading conditions.

Joint stiffness: There is no appreciable change in the stiffness of the RHR piping system due to the addition of the leak sealant enclosure and structural restraining clamp.

Constrainingeffects: The clamp holding the leak sealant enclosure is a single clamp band. The temperature Increase experienced by the clamp from the point of installation until Its removal will be no more than 95OF (From 1501F min to 2451F max [17]). This temperature increase will result in a radial expansion of the 10" diameter pipe by no more than 0.005". The effect of this radial expansion on the strongback bolts will be an Increase in tensile stress of approximately 20%. The initial torque on the strongback bolts is approximately 40% of the material yield.

Therefore, the additional tensile stress on these bolts will remain within the yield point of the material.

There will also be a constraining effect on the 3/4" pipe due to the addition of the structural clamp. If unrestrained, the thermal expansion from the center of the 10" pipe to top of valve RHR-600 where the structural clamp is located is approximately 0.011" [19]. This thermal movement was considered in the design of the structural restraint. A relatively small force of 130 lbs [19] will be imparted to the support from this thermal expansion. This load will also be subjected to the top of the 10" pipe as a compressive force. This compressive load is considered to have a negligible impact to the stresses In the 10" pipe, In summary, the constraining effect of the leak sealant enclosure, strongback, and structural clamp will not significantly impact the piping stresses, the strongback bolting, or the structural clamp stresses.

Defect growth: The friction resistance of the pipe clamp and the rigid connection of the leak sealant enclosure stabilizes the 3/ Inch pipe and valve, reducing stress. Therefore the defeet is not expected to grow during the very short timeframe this TM will be installed.

Monitoringi Reguirement (1X-6000)

Compliance with the monitoring requirements of IX-6000 shall be as described In the relief

request, No.7".0749 (May 20 2,*)

F~orm

Temporary Modification Sealant Injection The sealant will be injected at a pressure based on the system operating pressure. Thus, the system pressure will prevent the sealant from entering the system piping. The allowed quantity of sealant to be injected is calculated to be 8.5 In3 [12]. This volume will be procedurally controlled via GMP-206, Leak Sealant Injection Repair of Steel Components.

Sealant will be Injected to stop the leak up to maximum predetermined amount of sealant. The acceptance criteria for post-injection leakage from the joint is no leakage.

Containment Isolattion The leak enclosure will be located on piping outside of containment that is between the containment and the outside isolation valve. Therefore the leak enclosure will need to [91:

1. Meet Safety Class 2 design requirements.

2. withstand containment design temperatures

3. withstand Internal pressure from containment structural Integrity test

4. meet seismic Category I design requirements

5. be protected against a high energy line break outside of containment when need for eentainment isolation.

6. maintain leakage within the Current Licensing Basis RHR-600 is an outside Containment isolation valve for Containment Penetration 10, a Class 6 penetration. Penetration Class 6 Is a system required to operate post-accldent. The design and operational criteria for penetration Class 6 isolation valves are governed by the functional requirements of the system. The isolation valves at Penetration 10 are not relied upon to prevent the escape of Containment air to the atmosphere [14].

The requirements for Containment Isolation are satisfied since the enclosure will satisfy the design requifements of the RHR system. The leak sealant enelesule will be desigriel IR accordance with ASME Section XI, Appendix IX,Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary. The piping remains seismically qualified and the enclosure will remain intact during a seismic event. The leak sealant enclosure will prevent system leakage.

Operations The leak seal enclosure Is located near RHR-600. The installed TM will not prohibit operational use of RHR-800.

References

1. XK-100-18, Rev. SA

2. ISIXK-100-18, Rev. AC

3. M-962-2, Rev. A

4. ANSi/ASME Code Reconciliation For Replacement Material, Parts, And Components, Kewaunee Power Station, Revisien 3, July 8, 2010

5. XK-100-371, Rev. 5 IV~mi"o 750140 (May;0;1)

TemDorarv Modification io Domini@n=

6. ASME B&PV Code

7. OP-KW-GOP-102, Rev. 13

8. Consumable Material Evaluation (QME) 10000005744

9. ANSI N271-1976, Containment Isolation Provisions for Fluid Systems.

10. KPS Calculation RHR-34-004, Pipe Stress Report (Rhr-34-004) For Residual Heat Removal Piping System Analytical Part No. Rhr-34-004, Rev, 1, 19900228, CA706184

11. CEM-0049 and Addendum 00A, Rev 000, Evaluation Of Compensatory Measure Taken In Response To Identified Leakage At 3/41 Driain Line Valve 2-RH-33 Off Line 14"-RH-1 18-602

12. TEAM Industrial Services Engineering Order 91511 dated 04/29/2012, Rev B

13. System Integrity Program, Rev. 8

14. Darrell G. Eisenhut, NRC Director Division of Licensing, to C.W. Giesler, Wisconsin Public Service Corporation, Exemption to Certain 10 CFR 50 AppendIx J Requirements, dated September 30, 1982.

15. TEAM email, 04/29/2012

16. TEAM email, 4/29/2012, Strong Back Torque

17. OP-KW-NOP-RHR-001, Rev. 17, Sect. 5.11 RHR Alignment for Exceeding 245'F (Split Train Made)

18. KPS Calculation S-062-RHR-34-004 79-14 Hanger Design Verification

19. Evaluation of Pipe Clamp Assembly Components, 06/01/2012 : Structural Restraining Clamp Form No. 730749 (May 2011)

TURAT2012-I1 STRUCTURAL STRAINING CLAMP BILL OF MATERIAL ITEM OTY MATERIAL 1 2 CLAMP. PIPE 10. GRINNELL FIG. 295 2 2 Vz STAINLESS STEEL PLATE, ASTM A240-0BC TYPE 316 3 1 COUPLING, 114' SOCKOLET WELD PIPE, STAINLESS STEEL ASTM A-182 F304, 6000 LB

,(CENTER MACHINED OUT)

NI Z PIPE, I, STAINLESS STEEL, ASTM A3I2-8BA P304 SCH. 40 SEAMLESS 5 ETSS BANDING 10 GA. 6" WIDE. LENGTH AS NEEDED SECTION A-A IvbCIomp.dgn

Industrial Services Registration# F-003143 Engineerig Depaxiaent Tel: (281) 388-5695 Fax: (281) 388-5690 ROUTING SLIlP & COVER SHEET FOR NUCLEAR SAFETY RELATED JOBS Branch Work Order #: 203-04270 Status: Priority Caller: Chad Preston Customer: Domion Energy . Safety Review #: 91511 Engr Order #: 91511

____ Name: Signature: Date: Time:

Data Taken By: Heather Hodges, 4/28/2012 t 06:00 Designed By: Heather Hodges 4/28/2012 13:08 Verified By. Andrew Campbell 1 4/28/2012 14:15 Shop Received By: __.__ _ _ _

I QC Received By:-_, __ _

Specifications:

Design Pressure: 600 psi Design Temperature: 1400 OF Service: Reactor Coolant ITorque Value: 176 f b Total Weight: 70.85 . Void: 6.27 inj BC (A3)

,.ealant Type: X36 w/G-Fiber ** ,1

( bo Not Paint __ '_

Note- :ASME SECTION XI,

_______ _ APPENDIX IX ___

QC FINAL INSPECTION REQUIRED Nuclear - Safety Related MTRs and COCs Required PMI Required Bill of Materials: __

Description:

i Material: Qty:

Clamp Hub SA 240 GR 304 / 316 2 Strongback SA 240 GR 304 / 316 _ 1 J 5/8"-1lUNC STUDS SA 193 GRB8 /8M 4 5/8" HEX NUTS SA 194 GR 8/8M S _ _ _

1. 10-24UNC SET SCREWS !ANSI)B18.3 STAINLESS STEEL_2 _

,SEALANT i X-36 __1 TUBE

__________________________________ __________________________________________________________ ]

Rev. 3/3/2010 Rev. 3/3/2010

-n II~ J~ I~I1.1 rI~~f'q~ F~I I A I ~ A I ~AN * ~is1I

.rI T rl K31 - V %kLIIT ALVVA Tb 03/4"1 518" STUDS 03/4" (2) PLACES NOTES:

1. APPROVED TO MANUFACTURE

2. 1/4" X0.12" TUBING GROOVES INHUB FACE

3. 3/16" X0.09" TUBING GROOVES INBORE

4. INSTALL STAINLESS TUBING REGISTRATION # F-003143 5.

6.

DRILL &TAP (2)1/4-NPT INJECTION PORTS INCAVITY MAX INJECTION PRESSURE: 1000 PSI + STATIC Industrial Services, Inc. I PERIMETER VOL I 3

INA3 BC

7. (2)HALVES REQUIRED ENGINEERING ORDER# 91511 EM WT. 71 LBSIVOL 9ýý"7 1NA3 R(-.

8. ALL DIMENSIONS ARE TYPICAL UNLESS NOTED DRAWING # N/A MAc*~ADsUl~Acu DOMINION .4t% -,..

9. D/T (2) # 1O-24UNC CUP TIP SET SCREWS IN 1.07" BORE, TQ: 36 IN*LBS TECO PART # N/A IR~AXSHARP~tIAU .~S HUB CLAMP 1OUL~

10. NUCLEAR SAFTEY RELATED - DO NOT PAINT WPS: N/A

11. ASME SECTION XI, APPENDIX IX "AroZ l eF T- EG6VAMP- SIZE A REV 0 CHECKED BY: AC 14/28/2012 lP'A I C-I. -A, K£fZ-II:T I n"C 1)

R iS - Ba - i~ A V*P~P~~ ri 3%fl. n. . a * .n.~ ,At..,a~r a.,

t Z)Art I T rI K.'I - Wl-- qLlI T ALVVAY b 5(8" r r -I.I 5/811 IT 3" 1 3/4"

£ Q

v 03/4" (4) PLACES REGISTRATION # F-003143 NOTES:

1. (1) REQUIRED

...... Industrial Ser, PERIMETER VOL . INA3 BC ENGINEERING ORDER# 9151 1EM WT. Z7T LBSIVOL tI~1A3 BC

2. SEND (4)5/8-11 UNC X 24" LONG ALLTHREADS

3. SEND 5/8 HEAVY HEX NUTS DRAWING # N/A DOMINION In"'.

TECO-PART # N/A HUB CLAMP TOLERANCES WPS: N/A FRACTIONJAL I'll, ANOGWAR *fj2.

DRAWN BY: HH 4/28/2012 IV3OPALCEO(CMAL

-1O IlIREEPIACOEOMAAL iflros SIZE A REV 0 CHECKED BY: AC 4/28/2012 ALLOVAVISO1WHWINRJS SCALE: 1:4 SHFFT 9 nF 9

I TEAM INDUSTRIAL SERVICES, INC.

ENGINEERING DATA COVER SHEET I

. Rev. 12/16104 LRS:l B TS:] Other:[

PRESSURE TEMPERATURE Design: Design: Service. ~4- 0 t~-A Oprtn:Operating:.

Operting Line Material: 3/ go Ll 5

- 3a16 -5s Line Size: Flange Rating: Quantity. Sealant Selection: Material Requested:.

3/9- )c~r 1 ~ - 316 SS Package Requirem ents:

Seck all that are immediately needed) (Check one only) rawings For Irfimediate Manufacture k alc~latlons'. Wait for Approval of Package / Pdce" Pce Of so 0 ballpark or engineering) eby El Price Only / No Calculations/Dramwngs aeand "lme Required:

I' F __mal PrInts ard Calculations: john.p.schroeder@dbm.cor ustmer Name: T,'h ,'i FaxNo:MI&

I e

~,To CBtmr=

E Notify Branch Supervisor afterfaxing Name: _" Phone No:

El Immediately ENe't Business Day Special Requirements: Strongbacic S needed for separation I vibration El Stress Relief Required by Customer El If Lifting Lugs are required, where sho~tld they be located?

El PE Stamp El Are there specific requirements / codes for this customer?

El MTRs and COOs El Other U NDE DESCRJPTION OF ITEM TO BE BUILT (HTS ONLY) STYLE O5 FITTING (HTS ONLY)

Hot Tap Fitting: " [] Split Tee o Line Stop Fitting: El Full Encirclement Reinforcing Saddle (0] with Nozzle?)

(Saddle is non-pressure retaining)

El HI-Stop Fitting:

0 Other I00 Bolt-on Weld-o-let 1I I] Sadidle / Scarfed Nozzle Tap Size: Existing Line Sýhedulo; Requested Nozzle Schedule:

FORM 104.2 Quality System Supplement Rev: 5 Corporate Page I of 2 NUCLEAR AUTHORIZATION / CONTRACT REVIEW CHECKLIST (NACL)

I TMS Division e PART A: Used for material only orders and in conjunction with PART B for service related work.

Limit 3 jobs per Team Job # / CJ#. All threejobs must be within the same service line.

Submitted by: ,- Branch: n3 Date: /-g2f- I z Utility: ],,,, Plant ILocation: -,  !,

Contact:

/4,. i-e- Phone #: '?Z 0- 3 g'- ?3ýSb Customer ,Cont ork Order/PO#: [Team Job #:'&O 30 el7o Personnel: ~.el. L 2. ( tq4t-' e.-MK/?Ato,% .

Shilt: A 4*,14 Additional Personneh Pre-Job Check List:,A onnustbe eo edeachda each sh wad gdor to any rorkAef ggornmed

1. All Pre-JobSteps must be Initialedby the Team InduWrialServices, Inca technicianwho i leadingactiviie, at the work sitepriorto any work being performeL

2. The Owner Representatm must initial Step 8 priorto any work beingperformed All 7"!*I ,mnlnwp.j an the fob must slun* Sten 119 orior to any+work beinaneriormed Steps: Yes No Initial Date Comments:

1. Has JSA been coropleted? l IL.. _ _______

2. Has Tech. Support been contacted (Critical Job Review) El 91-/ . C.#: 11571

3. Have Team personnel READ and do they UNDERSTAND the Owner Engineering Control? 0ýj -. 0 z

4. Have Team personnel verified proedure(s) to be used? _ " " j' Procedurclk: 3 a0

( 5. Have Team personnel READ and UNDERSTOOD the Owner W.O. Scope I- " ...

//2% W.O.#: KLW,0069q-14 tP96

6. Have Team personnel been briefed on Owner W.O. Scope and received copy o '

7. Have Team personnel bad a Rad. Protection Brief! ' ... _ _,_ _/__

B. Has Owner Representative verified all Steps have been perInned? " *Owner Rep. to Initi. l.

9. Signature of all Team Personnel: ________

• PART B: Used in conjunction with PART A for all service related workL PART B is not required for material only orders.

For Technical Support Use Only Number of Jobs reviewed. I [] 2C] 31-] Comments:

Reviewed by:. ....

Signature: Date:

Job #1 Information ..... _ Reqfrements Service Line:/

• Leak Severity: / Notifieations: Safety Related: Yes No Line Size: /O fiS/g Service: r/'*'/' CO, i**

Quality Assurance: Yes Health Physics: Yes **' No No El El Revs.#:

5C9tRd Procedure ih#:

Unit:.Ep.Rate:

Op. Temp: /16O Op. Press: 3,50/ Materials & Fanipment DesignoTemp: &O0 DesignPress: 6j02* Sealant Type: ,-. - 44jC Lot#:

Equip. ID #: A'H.(- 600 Sealant Type: - - 'b r Ot IJ,.Lot#:

Comments/Description of Job: Fabrication Data jv!.C *.5U_.eL Mill Tests: Yes [] No El Sec. II: Yes [ No []

Has the Customer expressed mechanical itegr concerns for the component? Yes M No [I Plant Rep. (print):,* 4*+f,* ,,q' Initials:

• Design Calcs.: Yes'T Nor-l Engr. Dwgs.: Yes% No E]

TISI Tech. (print): Initials: t. P Positive Matr Identification: YesJI NoE

Quality System Supplement SFORM 5 Rev:1.04.2 Corporate Page 2 of 2 1 .f )RIZATION / CONTRACT REVIEW CHECKLIST (NACL)

TMS Division Job #2 Information _ Requirements Line: Leak Severity. N Ceations: Safety Relatri. Yes 0] No El Quality Assurance: Yes E] No 0l Line Si Service: ealth Physics: Yes [I No El Unit Exp. Rate: Procedure #: Rev. #:

Op. Tenp: Op. Press: MateRals & EQuipment Design Temp: Design Press: Sealant Type: Lot #:

Equip. ID # Sealant Type: Lot #:

Conmmenti/Description of Job: Eabr cat -III Mill Tests:Yes= NoDl se.iI-Y Nol Has the Customer expressed mechanical integi concems fr te component? Yes El No El Plant Rep. (print): Initials:, Design CaIcs.: Yes E] No E] Engr. Dwgs.: Yes [] aE TISI Tech. (print): - Initials: Positive Material Identification: Yes E] No 0l Job #3 Information Requirements ice Line: Leak Severity: tifications: Safety Related: Yes No Quality Assurance: Yes El No E-Line S* service:

• Health Physics: Yes El No [D Unit Exp. Rate: Procedure #-:"' Rev.-#-.

0.Te . Op. Press: rials & Eui meat Design Temp: Design Press: Sealant Type: Lot #.

Equip. ID Sealant Type: Lot#:

Comments,/Descipion i ofof Job:

J: *Fabrication Da, Mill Test: Yes Ml No [-] f Sec. Ill: NO El Has the Customer expressed mechanical integrity coýems for the component? IYes El No "l Plant Rep. (print): Initials:. Design Calcs.: Yes El No El Engr. Dwgs.: Yes NEo JZ1 TISI Tech. (print): Initials: pPositive Material Identification: Yes [E No El Post Job Steps:

1. All Post-job Steps must be initialedby the Team IndustrIalServices, Inc. technicianwho is leadingactivitesat the work site.

2. Post-Jobmust be signedby customer representative.

Steps: Yes No Initial Date Comments:

Complete sigcL-off ofjob package. EJC.Zgfiz 10 Complete sign-off of Team's procedure. jZEc 9 ~ '~/i-Complete Work Records and have Owner s0g. - I _/// 2 Plant Representative (print): 4 / Signature:

TISI Technician (print): - IjSignature: ,,,* " t Additional Comments:

L

• DS 136 TEE COUPLET ANDIOR WELD-O-LET

_[SCRE-WED/SOCIET E]SLS. BUTT. WELD FITTING GIVENBY: IC.tzI DATE: I  ?- ( IPLANT: I NT. : :6

. &0 CBD.Y:j~f~ 1 ~ DTE -2,&- rZ j.ISURFAcECONDITION:I (?&

LINE SIZE,: I0..'~ SEVER=T OF LEAK tON 10 SHIP TO:I \w IDIMENSIONAL DATA I LTR. DIM Al Bi OPen.ý Cl DI io, 7LS0 El 1,r-/7..r A2 B2 C2 D2

)E2 A3 B3 11.37S*"

C3 I, vs-*o I

D3 E3

,3.s~

W3___ (BCPT)-'---B1 -

• F 1.2510 CMTAO) d Al W4 ,___ [ OBSTRUCTIONS Gi  ?-O 02, . . ZS"-o H2 .z_-_

J2, x

/ 11 LOCATION OF BLOW-'D

,]SC;;;W AJWELED 1

COPYRIGHT 2000. Team Tndustri.1 Services. 1012/00 1

Designer HH Date 0412812012 TEAM Industrial Services Sheet 1 of 7 Checker AC Date 0412812012 9i511EM REGISTRATION # F-003143

.Split Circular Endplate Analysis

References:

A5 ME Boiler and Pressure Vessel Code,Section II, Part D, (Table for Maximum Allowable Stresses, 2010 Edition)

Fo rmulas for Stress and Strain by Roark and Young, Fifth Edition, Table 24, Case 31 5 i~pLy 5upportea Edge Fr* Ed-9o Data:

( ) Design Pressure P := 600.psi Modulus of Elasticity 8

E := 0.279.10 -psi Design Temperature T := 400-deg Poisson's Ratio v:= 0.30 Split Endplate OD OD: 4.0-in Maximum Allowable Deflection Ymax := 0.05-in Cover Wall Thickness tR=0.75-in Joint Efficiency JE:= 1" Split Endplate Thickness tý.dp1: 0.625*in External Corrosion Allowance ExtCA := 0-in Opening Hole Diameter HD: O-in Internal Corrosion Allowance IntCA:= 0.in (Conservative)

OD := OD - 2.ExtCA OD =4in Maximum Allowable Stress Sa 1:8300.psi tw.ed:= twau - ExtCA - IhtCA t,= 0.75-in t,,dpl := týýdpj- ExtCA - TntCA teadpl = 0.625-in Inside Radius [GD - 2(t,,)]

rR:=- -

.2

]R = 1.25-in Analysis:

Solving for Modulus of Rigidity E

G = 10730769.231-psi 2.(1 + v)

Solving for variables I. HD a -2 2 OD twau a:= a = 1.25-in C= c = 0.625-in 2 .2 b:= a- c b = 0.625-in

TEAM Industrial Services Sheet 2 of 7 REGISTRATION # F-003143 91511EM Solving for Constants K:= 0.42338. (b c) .58614.(b c) + c85046-lb+ c) ' b.+c)

+ 2.48483 K = 2.485 I2-b (4 .625-tendpl) rG( +L2 "f = 2.496

~y:= c j+4-1-I--I ~2-c}

"_ _ 1 4-b 1=+I--*24 2 e-"1 "/ = 2.463 4b24 _Y2 = 0.406 Y2 := 2 1--

X, = 4.231

_ - _Y2 + >xi}C*

_(b Y2)>t. b(^f + X = 0.292 2- + -Y 1

2) (_X - ).co s (. c, = 0.012

-1

.: I 1

C2 c2 = 0.409 12 -- lcosh 1.

C 1 2-Stress atA (maximum)

O~t6-P-c.b

.- ; (1 -1)ci(

+ CT{l crt = 7647.346-psi < 183 0 0

2. + - Sal,,, -psi bC) -'Y2 teadpl Deflection at B (maxdmum)

1. ! !

24-P-c2"b 2 b E rtWndpl Tc I e hL +c1cs-I b4) y = 0-in <Yzizax =0.05-in Minimum Cover Wall Thickness P-IR treqd = 0.04181-in tau = 0.75-in trq

=(JE-Saw 0 )- (0.6-P)

TEAM Industrial Services Sheet 3 of 7 REGISTRATION # F-003143 91511 EM Line Enclosure Analysis

Purpose:

This analysis will calculate the internal stresses and bolt load of a line enclosure.

References:

ASME Boiler and Pressure Vessel Code,Section II, Part D, (Table for Maximum Allowable Stresses), 2010 Edition Team Industrial Services, Teco Manufacturing, Engineering Department, ISO-9001 Quality Manual, EP8.7 R1R Dimensions Fr'ee Bodyj DlonrLdrn' Data:

Design Pressure P = 600-psi Length Between Centerline of Seals LS:= 3.0-in Design Temperature T = 400.deg Sidebar Length (at Centedine) LB := 3.0-in Inside Radius R:= IR - ItCA R = 1.25-in # of Studs per Half NS2 := 1 Cover Thickness t:= tan + IntCA t = 0.75-in Hole Size h:= 0.75-in Cavity to Stud CL A := 0.625-in Stud Tensile Area TA:= 0.226.in2 End of Sidebar to Stud CL B:= 0.875-in Stud Allowable Stress Ss:= 13800-psi Sidebar Thickness ts :=OD ts = 2-in Enclosure Allowable Stress Sai=ow 18300.psi 2

External Corrosion Allowance ExtCA= 0-in Internal Corrosion Allowance IntCA = 0-in R := R + IutCA R = 1.25-in B:= B - ExtCA B = 0.875-in t := t- ExtCA - IntCA t = 0.75-in LB := LB - 2.ExtCA LB = 3-in A:= A- IntCA A= 0.625-in ts:= ts - ExtCA ts = 2-in.

TEAM Industrial Services Sheet 4 of 7 REGISTRATION # F-003143 91511EM Analysis:

Solving for forces and moments F:= PR.LS F,:= F Fy:= F F = 2250-Ibf F, = 2250.1bf Fy = 2250.lbf Setting forces in x direction equal to 0 R 2 := F. R2 = 2250-lbf Setting moments around centerpoint of cavity equal to 0 t

2 BL:= F- BL = 2892.857-1bf BL := if(BL < F,F,BL) DL = 2892-857.lbf B

Allowable Bolt Load BLa:= TA.Ss-NS2 BLa = 3118.8.1bf Stresses in Shell (thin walled enclosure) o-  :=- = 1000-psi t

Sidebar Stress (at Bolt Centerline)

S R, := BL- F R 1,= 642.857.4bf RI.B.--sts 2

"-(LB - NS2-h)-ts3 &b2 = 375-psi 12 3 F

= 2s -

r = 750-psi 2 (LB - NS2.h).ts Results:

Less Than Bolt Load Allowable BL = 2892.86-Ibf BLa = 3118.8.1bf Stresses in Shell (thin walled enclosure) a = 1000-psi Saol,,= 18300-psi Sidebar Stresses (@ bolt centerline)

Orb2 = 375-psi Sallow = 18300-psi Shear Stresses in Sidebar

, = 750-psi 0.8."So,, = 14640-psi

TEAM Industrial Services Sheet 5 of 71 91511EM REGISTRATION # F-003143 Thrust Calculation Due to Unequal Bores Seal Length # I D1 := 3.5-in Seal Length # 2 D2:= 3.5.in Smaller Diameter d:= 1.05-in Injection Pressure IP:= 1000-psi Number of Studs N =4 Size of Studs 5/8-11 UNC Tensile Area of Studs At:= 0.226-m2 Stud Allowable Stress SS := 13800-psi H:= At.SS Allowable Load of Studs H = 3118.8.lbf Thrust Produced Thrust:= (D1.D2 - d2..H.IP Thrust = 11384.099-lbf Number of Studs Required Thrust ND -- . ND = 3.65 < I N=4 H

Force per Stud Thrust F = 2846.025-lbf N

Thrust Calculation Due to Separation Diameter d := 1.05-in Design Pressure p := 600-psi Number of Set screws Ns := 2 Size of Set Screws #10 x 24 UNC (Cup Tip)

Holding Power of Set Screws Hs:= 262.1bf Thrust due to separation Ts ( -,--.P

-= Ts = 519.541 -bf

,4 Number of Set Screws Required NDs:= NDs = 1.983 Hs Force per Set Screw P Ts Fs 259.77-lbf.

Ns

TEAM Industrial Services REGISTRATION # F-003143 Sheet 6 of 7 91511EM Thrust and Bending Calculation - Hub Thrust Produced Thrust= 11384.099*lbf Moment Arm X:= 3.375.in Total Width B:= OD B= 4-in Number of Studs per Half NS:= 2 Cavity b := (IR- IntCA).2 b = 2.5-in Allowable Stress Sa11ow = 18300-psi Thickness Provided (After Radius) tp = 2.3.75-in Joint Efficiency E:= 1.o Force per Half Thrust Fs- NS Fs = 5692.049.1bf Thickness Required tr= 6.Fs.X' tr (= tr = 2.049-in < . tp= 2.375-in C ~ Thrustand Bending Calculation - Flat Bar Thrust Produced Thrust = 11384.099-lbf Moment Arm X:= 3.9-in Total Width B := 3-in Number of Studs NS =.2 Cavity b:= 0-in Allowable Stress Sallow = 18300-psi Thickness Provided (After Radius) tp:-- 1.625-in Joint Efficiency E:= I Force per Half Thrust Fs:= Fs = 5692.049-lbf NS Thickness Required S B Fs .X (B- b-E.(sao,,) tr= 1.558-in < tp = 1.625-in I i

TEAM Industrial Services Sheet 7 of 7 REGISTRATION # F-003143 91511EM Weight and Void Void Injection Valves NI:= 2 TnjVlv :=.NIV-O.19in 3 InjVlv = 0.38-in' Cavity:= (2.5-in) 2.-(l.57.in(A1) Cavity = 7.707 -ji3 4

2 71 2 Ir Line := (1.5-in) .- .(.75-in) + (1.05-in) *- .(0.5650.in) 3 4 4 Lime= 1.815..

Void:= Cavity - Line + InjVlv Void = 6.272-in3 B.C.

Weight (Clamp & Stron back from SolidWorks Models)

Clamp := 17.96-lb-2 Clamp = 35.92-1b al1thread:= 4-0.087 lb in + 8-0.1 .lb allthread = 9.232.lb in Studs:= 4-0.087- in+ 8-0.11-lb Studs = 3.664-lb in lb Sealant := Void-1.35.0.043-- Sealant= 0.364-1b

.3 m

InjValves:= NIV.0.46-lb InjValves = 0.92-lb Stromgback:= 20.56-lb Strongback = 20.56-lb Weight:= Clamp + allthread + Sealant + InjValves + Strongback + Studs Weight = 70.66-lb

PevjeD b 0.E Pipe; OD VS T-D T~

PAe$SVjZZ. F6,ecz Page Iof 3 pq~ýAX> 111 *j,JVM01edMAAA

~eq~eW~ ~

Evaluation of Pipe Clamp Assembly Components:

-Prepared by:- Christopher M. Minton Date: q.S<) Z62I-Reviewed by: Vishwa M. Bhargava " Date:

Assumptions:

The pipe clamp assembly is sufficiently rigid, such that seismic loading is negligbl.e Therefore, the'only loads for consideration are those due to the design pressure (600psi) and thermal.

DO .= 1.05in Di 0.824in t:== 0.1 13in

-b 0 Pdes := 600 Apipe =Irr(D - .)2

.2 I. 2 In Determine force due to pressure:

Dete n or Apxpjdes = 520ib II Determine force due to thermal growth of vent line:

The temperature increase experienced by the clamp from the 0.0203- -0.009- point of installation until its removal will be no more than 95°F ft ft (From 150°F min to 245°F ma4). Thermal expansion value is that for 250°F minus 150°F.

Ad:= '1/2 po = 0.0113 in Eplat:=-29000000-b plate I (0-5".i) (1.75i)3 = 0.223-in4

.2 m

(30in)3 -130 "o Assume simply supported beam, maximum shear load at midspan is the following:

Fpr towa +th = 3 ttl 2 2 Pipe Clamp Evaluation:

Pipe Clamps are Figure 295. Per Anvil Catalogfor 10 inch pipe:

Rating is 3240 lbs @ 6507F. OKAY

I Page 2 of 3 Evaluation of Weld at Coupling:

Weld is parallel weld, length 1-75', spaced 1/2w Consider moment arm of 15.inches. I&

M= -15in-Fto0tal I = ,75in-lb F3 =Ftot=325Ib Assume an eccentric load acting on the structural steel, offset 1 inch:

M2 --- Fttrlin = 32S-in-lb Required weld size is determined to be o,,9'M7n (See Attached Spreadsheet)

Allowable weld size is based on the minimum of the following:

0.707 x 0.3 x 75ksi = r5,90S psi (Electrode .E -3o' ")

0.4xSy = 0.4 x 30 ksi= 12 ksi (Base Metal Shear, Temp considered as 1007F) v.51jt in <o.4,75' in, OK Evaluation of Stainless Steel Plate Steel: Consider 100 0F Temperature.

S -=30000-r Sy2 (ASME B&PVC, 2004, Section If.Part D)

. 2 Check Bending Stress:

23 4 1.75in Iplateo= 0 =- 2 0.875-in MI'¢=i, Z Ib " b M 1 .-- 2 Fb -o 6 6 Sy 1980oo Ob OKAY 2 Fb Check Shear Stress A 1.75iu-0-5i 0-875-rn

_FtOWs Aplate 0267 + 0.282 lb

.2 IA

=0-274 2

(Formula from Design of Welded rtorsion '---

a-(I.75in)(O5-m)2 m.2 Structures, Blodgett) lb T 0%27--

total Trtorsion + Tfor0e =

.2 M

F -=o04S= o12000

, 2 =.2516o OKAY FY

. ... .A.. .... A .N ...( Baaall1e Weld) Page 3 of 3

__Between SS plate and SS coupling 1 JKPJ~M~NF~WN ~

LOADS FROM: STRUDL =I CALC = OTER MEMBER. JOINT START COORDRNATE SYSTEK:- LOCAL.( ) GLOBAL MaTdalTbick-nofThiclrPatJined 0.500 in LOADING CAS: ERVNELOPE [ LOADING CONDMIION Uni? ,r A 1r AT C'RNTT 01 WFtfl-FP= 0 Ibf m, = 4 S-7 in - bf A

F,= 0 Ibf M2 = 325 in--T F3 = 32, Ibf M3= 0 in.Ibf Fc.m.*, =4 21000 psi WELD SIZE F"o~ý,4 = 30000 psi O'VIED: 0,-4.-15 in W .P ROPHTlM: CIEATED A ASAME)

A=2a = 3.50 in z= 1.75 in b= 0500 in 2?

Z, =(a23 = L02 i-2 Z.=a(b)= 0.88 b-2 L.

Imc = z, + z(ba) = 1.27 b3- cI3= z(ab) + z. = 4.45 in 2 F=[( F2 + Mi + f + C F1 +I M2 ) +

Z, 7-3 A A Ti3*c, A

( F3 + M2 A 1/C 3 F,[( o + 4-W + 0 + 0 + .25- )2 +

3.50 1.02 0.88 350 1L27 3.5 + 4325 )2.fn 3-50 4.45 FORCE ON WELD: FB= 4:19 - lbf/ia RFTTIRED W'T"1D .17E - ' (Base Meta Gov==n) FYT (BaseMetal) <- 33000 psi" W = 0 (sQF 0o.3q* ia <0o.4315 in 0.4 (FY (Base Met~al)) .4 ( 30000 )

CONCLUSION: USE OF INCHWELD IS ACCEPTABLE.

ProductRisk Assessment Worksheet 0906miniow Engineering Product: Temporary Modification (TMODI Date: 4128/2012 Engineering Product

Title:

RHR.600 Stronciback Prepared By. Lod Christensen Print Engineering Product No: TMOD-2012-11 Rev: 0 INSTRUCTIONS This worksheet has three parts:

" Engineering risk is the product of the probability of engineering error and the consequences of that error. The engineering risk can be managed by applying more stringent validation to those engineering products that exhibit the most risk. Part I of this worksheet which evaluates the probability of failure and Part 2 the consequences of that failure.

" Part 3 uses these results in a risk matrix to determine the risk rating for the engineering product.

"This risk rating determines which Actions for Engineering Risk Activities (Go to Attachment 5) apply to the engineering product PART 1: PROBABILITY OF ENGINEERING FAILURE Using the definitions listed in Section 3.1.2, the engineering product Is deemed complex by supervision. Consider the following:

"High Impact - The number of engineering disciplines, organizations, or processes Involved In the task or Its resolution is high (>two)

"The Growth Tendency is High due to:

>Outsourced engineering products and services

>New technologies, applications, or Infrequently performed tasks

>A work scope that Is not well defined

• The Urgency of the Project Is High, e.g., subject to significant time pressure, and the engineering product Is not simple.

The engineering product Is deemed moderately complex by supervision, if there is a growth tendency due to a degree of uncertainty regarding the task or there is an urgency with simple engineering product.

The engineering product Is deemed simple by supervision. These may include:

"Task performed by a single Individual

" Routine tasks

" Non-urgent tasks PART 2: POTENTIAL CONSEQUENCES OF AN ENGINEERING FAILURE The engineering product can impact: Significant impact 3

" nuclear safety e industrial safety Some impact

" environmental protection

" regulatory compliance Minimal impact L electric generatio,,n, reliability .......

PART 3: ENGINEERING RISK

" Enter Probability value from Part I 'RISK ASSESSMENT TABLE CONSEQUENCE "

SEnter Consequence value from Part 2_ 1 2 3

• Determine Risk rating (L/M/H) from Risk Table and enter it here. - I1 L L M (Use In Attachment 5.) " ro 2-L (M H md 3 M H H Note: Supervisor may upgrade this risk rting based on factors listed in Attachments 2 & 3. 1 0-Please provide a completed copy of this checklist to your E-HUDC.

FRun No. 731210 (Jun 2011)

EngineeringRisk Plan PDominloW it Engineering Product Number: TMOD-2012-11, Rev. 0 Product

Title:

RHR-600 Strongback Based on the Risk Rating derived in Attachment 4 the following validation processes/HU tools apply. The Manager may waive any required process as permitted by other governing documents.

Risk Based Validation Process/HU Tools Engineering Used' Us Waive Validation Used Risk Rating L M H 1(1) 2(2) X(3)

Risk Assessment - Validation Plan 0 R R R R X Initiate Form Project Team 0 0 R R R X Task Determine need for external expertise for the Project R R R R R X

_,K Yes - Needed No Project Scope of Work R R R R R X Plan Task Pre-Job Briefing Worksheet 0 R R R R X (PI-AA-HU-ENG-1 011)

Project Kickoff Meeting 0 R R R R X Ownedr ReviewNendor Oversight5 R R R R R X MCR (Modification Challenge Review) at 30% 0 0 R R R X MCR at 75% 00 R R R X MCRat 100% O R R R R X Engineer Task ECR Engineering Challenge Review O O R R R X (other than modifications)

FECR-Fleet ECR, or FMCR - Fleet MCR 0 0 0 R R X ITPR- Independent Third Party Review 0 0 0 0 R X QRT Review (CM-AA-QRT-1 001) S S S S S Form No. 731223 (Jun 2011)

EngineeringRisk Plan 0

Paominlow I.

NOTES:

1. Hi Risk with < 4 Probability of Failure Factors (Attachment 3)

2. Hi Risk with 5-8 Probability of Failure Factors (Attachment 3)

3. Hi Risk with > 9 Probability of Failure Factors (Attachment 3)

4. 0 = Optional, R = Required, S = Sampling

5. Only applies to outsourced engineering Work

6. Only required for a significant team effort High Risk Compensating Actions (Not Required for Low or Medium Risk)

Consequence Risk Mitigation: For each consequence risk factor identified. in Attachment 2, list the factor and the actions to be employed to mitigate that risk.

Risk Factor(s) Compensating Owner Due Date Tracking Mechanism

( Probability Risk Mitigation: For each Probability of Failure risk factor identified In Attachment 3, list the factor and the actions to be employed to mitigate that risk.

Risk Factor(s) Compensating Owner Due Date Tracking Mechanism Assigned Engineer Date 5eupervisor Date Engineering Manager (High-Risk Engineering Product Only) Date Provide a completed copy of this checklist to your E-HUDC.

Provide summary in ESOMS, FogmW. 731223(.hm'2011)

Design Effects Table

@Dmini@o INSTRUCTIONS:

This table will identify impacted programs. If a question is answered yes, the responsible engineer shall address and document in the discussion section of the Design Change. The responsible engineer shall check "No Impact" or "Impact" as appropriate. If necessary, consult the program owner to assist with the determination. If it is determined that there is an impact, identify program owner, check "Impact" in the applicable section, obtain the consulted individual(s) signature on the engineering product cover sheet and document the discussion in the change package. If an impact is determined for any program, then relevant portions of this attachment should be attached to the appropriate document. If it is determined that there is "no impact" in a section where a question is answered "Yes," document the basis for this determination in the change package. When all questions are answered "No" in a particular section, do not check "Impact" or "No Impact."

Station Unit Change Document Number Z] KPS [0 MPS 2 1 [E 2 TMOD-2012-11, Rev. 3

[1 NAPS n SPS [1 3 L ISFSI Change Document Title RHR-600 Strongback 1.0 Fire Protection / Appendix R 1.4 .Fire Protection Equipment or Features I Yes 0 No Program Owner Name:

El No Impact [I Impact The purpose of this section is to provide guidelines for the review of modifications/activities to ensure that the design conforms to the requirements set forth by the fire protection program documents, in each respective station's UFSAR, and to ensure that all modifications to the stations do not lessen the degree of fire protection at the stations.

Does the activity modify, add, or remove fire protection equipment including:

• fire barriers, doors, Or penetrations, and partial barriers credited in the fire hazards analysis

• fire dampers, doors, penetrations, hatches, cable tray fire stops; cable tray covers, or other elements installed In fire barriers

• non- Appendix R emergency lighting (e.g., exit lighting)

• fire resistant coatings including structural steel fireproofing and electrical raceway wrap

• fire suppression systems, including sprinklers, C02, and halon systems

• fire detection equipment, including smoke, heat, and flame detectors

• fire protection system interface devices such as HVAC shutdown trips, supervisory air supplies, and backup batteries

• fire fighting equipment or systems including hose stations and portable fire extinguishers

• supports and restraints for Fire Protection systems and equipment

• fire suppression water supply, including fire pumps & valves Does the activity modify any plant structure (including floors, doors, roofs, ceilings, drains, curbs, dampers, penetrations, hatches, equipment knockouts, stairwells, HVAC systems, pipe chases, elevator shafts, load bearing structural steel, etc.) such that it may block or otherwise interfere with the operation of any fire protection equipment?

Does the activity modify the occupancy or function of a room or structure such that it may affect any fire protection equipment?

She activity modify the nearby environmental conditions (room ambient temperature, nearby heat sources, etc.) that may affect any fire

'bation equipment?

" Signature required on Engineering Product Cover Sheet.

Design Effects Table 1.2 Combustible Loading El Yes 1Z No Program Owner Name:

El No Impact El Impact*

Does the activity add, modify, relocate, or remove combustible material within a fire zone, or relocate combustible material between fire zones?

Combustibles may include:

" Ordinary Combustibles (e.g., wood, paper, carpet)

• Combustible Liquids Grease, charcoal, and combustible insulation

• Plastic Materials (especially halogenated plastics such as PVC or Neoprene)

• Cables / Cable Tray Loading (Consult local procedures for tray fill criteria.)

• Cables or other materials which give off corrosive gasses when burned.

• Coatings (e.g., wall and floor coverings)

• Flammable gases 1.3 Hazards and Ignition Sources El Yes ED No Program Owner Name:

El No Impact El Impact*

Does the activity create, alter,. or remove any hazards or ignition source, such as:

Hydrogen or other explosive gasses

• Combustible metals (magnesium, zirconium, etc.)

() *New heat sources that could ignite combustible materials 1.4 Fire Safe-Shutdown Analysis El Yes !] No Program Owner Name:

.l No Impact Impact*

Il Does the activity involve any of the following relative to the Appendix R safe shutdown analysis (refer to local procedure/ site specific Appendix R information)?

• Credited Appendix R Equipment and flowpaths as identified in equipment databases (e.g., EDS, MEL, lAD) or as represented on P&IDs and in operating procedures.

• Cables, electrical schemes, and raceway associated with Appendix R credited equipment

" Radiant energy shields

• RCP lube oil collection systems Does the activity involve access to a fire zone/area, fire protection equipment or Appendix R credited safe shutdown equipment to perform manual operator actions or manual fire fighting activities?

Does the activity affect Time Critical operator or fire brigade response time?

Does the activity add, modify, relocate, remove, or obstruct any emergency righting on the manual action path required for compliance with Appendix R?

Does the activity add, remove, or affect the performance of any plant communications system relied upon for fire fighting or safe plant shutdown?

Does the activity impact any Appendix R exemption requests?

• Signature required on Engineering Product Cover Sheet.

Destqn Effects Table Domnio 2.0 Environmental Qualification (EQ)

E]Yes ED No Program Owner Name:

Ei No Impact I] impact The following questions provide guidance to ensure that the methodology used for compliance with 10 CFR 50.49, "Environmental Qualificatior of Electric Equipment Important to Safety for Nuclear Power Plants" will not be adversely impacted. The purpose of this review is to determine i the proposed design change could affect EQ Master List equipment, EQ files, EQ environmental conditions, or EQ/HELB boundaries.

Does the activity modify any EQ component as identified in (or to be added/removed to) the equipment database(s)?

Does the activity result in exceeding normal operating environmental conditions (i.e., temperature, pressure, humidity, radiation) as identified in Environmental Zone descriptions?

Does the activity affect or create new accident (HELB or other pipe breaks) environmental conditions (i.e., temperature, pressure, humidity, radiation, spray composition, or submergence) or the equipment operating time?

Does the activity add/remove any electrical system or portion of an electrical system credited for accident mitigation where the equipment is located in a harsh environment?

Does the activity alter the electrical portion or accuracy of any accident mitigating or monitoring system (including cable and interfaces) located in an area where the environment is affected by an accident?

D- he activity alter the physical arrangement or HELB boundary including doors, hatches, duct work, piping, or electrical penetrations, and s aral walls such that it could affect existing EQ equipment or add new equipment to the EQ Program?

Does the activity modify the operating conditions of EQ equipment (energized or de-energized, duration and/or component heat rise condition) during normal or accident operating times?

3.0 ASME Codes / ISI I IST ASME Section Xl and/or OM Code provides the requirements for preservice and inservice testing and examination of In scope piping, components, component supports, pumps, valves, pressure relief devices and dynamic restraints to assess their operational readiness.

3.1 Inservice Inspection 2 Yes [I No Program Owner Name: Phil Bukes 0 No Impact El Impact*

• Signature required on Engineering Product Cover Sheet.

C- - -.- I--- -

Design Effects Table S Domin)o NOTE: The ISI Classification Boundary Drawings identify the current ASME Code equipment.

1. Does the activity involve SSCs addressed by the ASME ISI Programs and require a change to one or more of these programs?

SSCs addressed by the ASME ISI program include any of the following:

- Class 1, 2, or 3 piping, components, or supports

- Class MC or CC - containment liner, concrete, arid post tensioning systems (no CC at Kewaunee)

• High Safety Significant Components (HSS - identified on Surry Unit 1 CBM drawings with an "H")

- ASME Code Inservice Inspection (ISI)

- Augmented ISI Program

- Constructed to a Nuclear Code, e.g., ASME Ill, ANSI B31.7, or ASME Draft Pump and Valve NOTE: A change to the ISI Program is defined as a revision to one of the following documents; ISI Plan or Schedule, System Pressure Test Program, Augmented Program, or ISI drawings as result of the activity. Examples of activities involving changes to SSCs such as the following may require a change in one of the above documents:

" Changing the diameter of pressure boundary bolting

" Changing the type of pipe support i.e., vertical/lateral to anchor, spring can to rigid strut, etc.

" Adding or removing pipe supports

• Adding, removing, or changing line numbers that are depicted on ISI documents

" Adding, removing, or changing the name of welds on lines that are depicted on ISI drawings

2. Will the activity result.in a permanent or temporary non-physical change to an ASME XI Class 1, 2, or3, MC, or CC pressure boundary item, core support, or component support item or piping system done by changing its design rating (e.g., intemal or external pressure, "mperature, or load)?

)

3. Will the activity result in a Repair/Replacement activity such as, removing, adding, and modifying items or systems (including welding, brazing, and defect removal activities)?

If Screening Questions 1, 2, and 3 above are all answered "No", then the program is checked "No" on the Design Effects and no further action is required.

If any Screening Question above is answered "Yes", then the program is checked "Yes" on the Design Effects and a Program Owner review is required.

If ISI program checklist is checked "Yes" on the Design Effects the following impacts shall be addressed with the program.owner.

1. A review of the proposed activity shall be performed to assure compliance with the applicable version of ASME Section XI (ISI) by contacting ISI personnel.

" Consider requirements for ASME XA Code acceptance inspection, base line (pre-service) inspection access, etc.

" A review of databases may be necessary to determine the correct design specification(s) related to the proposed change. Consider if an update to these documents is required as a result of the change package.

2. Consider access requirements for inspection including base line inspection requirements for repairs and replacements (i.e., design considerations) such as:

I insulation removal for later inspection necessary?

ls IIs pre-service (base line) inspection, such as UT necessary?

• Is the accessibility adequate for inspection? Equipment requiring ]SI that is being added or deleted or imposition requirements that are being changed for specific equipment shall be identified.

( I

5Design Effects Table

3. Consider the components being modified that are ASME Section III, Draft Pump and Valve, or ANSI B31.7 Code items. The Code Design Specification must be evaluated for technical accuracy prior to the change package being issued. Any change to an'ASME III, Draft Pump and Valve, or ANSI 831t.7 Code Design Specification shall be reviewed and certified by a Registered Professional Engineer competent in the area of design related to the change being proposed and the applicable nuclear code prior to issuance of the change package. A Registered Professional Engineer must sign the change package cover page and affix their seal for modifications per the previously mentioned codes when the proposed changes are being made prior to updating the specification. It is preferred that the specification be modified prior to implementation of the proposed change.

4. In all cases where the replacement part(s) do not meet the design bases for those installed, a.Report of Reconciliation shall be written in accordance with ASME Section XI. The change package itself may include the Report of Reconciliation.

5. Consider the documentation listed below. These documents must be maintained accurate and current per ASME Xl. The following design documents shall be addressed as applicable:

NOTE: Dominion design specifications may have procurement specification titles and vice versa.

• Design Specifications

- Design Reports (e.g., Overpressure Protection report, seismic analyses, etc.)

- Reports of Reconciliation (Used to document the engineering evaluation of acceptability) 3.2 Inservice Testing EI- Yes 10 No Program Owner Name:

Eli No Impact E] impact*

1. Does the activity permanently or temporarily add, remove, repair, replace, modify, affect, or involve a change to acceptance criteria or performance requirements for mechanical equipment or permanently installed plant instrumentation required for Inservice Testing (IST)

Programs?

NOTE: Examples of activities involving changes to SSCs such as the following may require a change to the IST Program.

• Adding or removing a valve or pump that is in the IST Program

• Changing the fail Safe direction of an AOV

• Permanently removing the internals of a check valve

• Changing the hydraulic characteristics of a pump in the IST Program MPS only-RCR-41747 (B16996-345): The DCM now requires an engineering program review by the IST program coordinator if there is an impact on ASME Code programs. This review during the initial stages of the design change process will help ensure that IST requirements are incorporated into the design change. (04093).

NOTE: The IST programs identify the current ASME Code OM equipment. Permanently installed plant instrumentation is often used when performing ASME Code tests. These instruments are subject to the accuracy and scaling requirements in the ASME Code and compliance with the requirements must be maintained. SSCs applicable to the IST program include pumps, valves, drives (motors, actuators) or instrumentation required for testing (pressure, flow, limit switches, etc.).

2. Does the activity involve SSCs Identified by the ASME IST Programs or require a change to these programs?

3. Review of DC by IST Program Coordinator if it impacts ASME programs. [CM 3.1.1 for MPS]

• Signature required on Engineering Product Cover Sheet.

Design Effects Table 4.0 Regulatory Guide 1.97 - Post Accident Monitoring El Yes E No Program Owner Name:

,I No Impact ii Impact*

In order to ensure compliance with Regulatory Guide 1.97, a review must be performed to determine the impact that the activity may have on the Regulatory Guide 1.97 program. Regulatory Guide 1.97 describes a method acceptable to the NRC for complying with the NRCs regulations to provide instrumentation to monitor plant variables and systems during and following an accident.

Does the activity alter any of the-following characteristics of Regulatory Guide 1.97 components identified in the electronic database, such as:

" Redundancy (Reduction of Category 1 variables only)

" Power Source (Category 1 and 2 variables only)

" Safety Classification

" Type of Display or Recording Instrumentation

" Control Room RG 1.97 Identification (Category 1 and 2, Type A, B, &C variables)

" Range or accuracy of Monitoring Instrumentation

" Electrical Interface between RG 1.97 loops and the transmission of this signal for other use prior to and including an isolation device (Category 1 and 2 variables only)

" Method of Measurement of a RG 1.97 parameter

• Location of Maintenance Isolation Device (Category 1 and 2 variables only)

" Accuracy requirements of any instrument credited during or following an accident

" Does the activity affect any Regulatory Guide 1.97 specification (MPS only)

D,,-q the activity add, delete, or physically modify the existing configuration of any analog inputs from RG 1.97 loops to a Plant Computer

,. )n (PCS)? This excludes any software changes to the PCS.

Does the activity appreciably increase or decrease a system parameter (i.e. temperature, flow, pressure, etc.) measured by Regulatory Guide 1.97 instrumentation such that the I&C calculations are affected?

5.0 Maintenance Rule EI Yes 0 No Program Owner Name:

No, Impact I Impact*

The guidance provided by the following questions assist the reviewer of the activity/modification in determining if any system/component functions are affected as delineated in the Maintenance Rule Scoping and Performance Criteria Matrix. This will ensure that compliance with 1 OCFR50.65 [Maintenance Rule] will be maintained.

Does the activity affect the Maintenance Rule function of a Structure, System, or Component (SSC) in any of the following ways?

" Adds a Maintenance Rule function

" Deletes a Maintenance Rule function

" Changes a Maintenance Rule function

" Changes safety classification

" Adds equipment utilized in Emergency Operating Procedures (EOPs)

Removes equipment utilized in Emergency Operating Procedures (EOPs)

• Changes the Maintenance Rule Performance Criterion:

• Significantly affects unavailability post modification (on an ongoing basis)

• Significantly affects reliability

• Signature required on Engineering Product Cover Sheet.

Design Effects, Table

I. fti no Does the activity affect component or system function, which could prevent SR SSCs from fulfilling the SR function?

Does the activity remove or install a system or train?

Does the activity affect component or system functions, which could cause a reactor trip, safety system actuation, or cause a power reduction' MPS only:

Does the activity affect component or system functions associated with the storage of nuclear fuel in Unit 1?

Does the activity transfer ownership of an existing system to another unit?

6.0 Radiological Protection Program (ALARA)

El Yes 0 No Program Owner Name:

El No Impact El Impact' The purpose of the following questions is to assist the reviewer in determining the impact of the modification/activity on the respective sites Radiation Protection Plan(s) which sets forth the requirements of the Radiation Protection Program. The ALARA Program(s) which ensures tha occupational radiation exposure, both individually and collectively, is maintained "As Low As Reasonably Achievable" (ALARA). And, the OffsitE Dose Calculation Manual(s) (ODCM) which establish requirements for the Radioactive Effluent and Radiological Environmental Monitoring P- -Tams.

Does the activity create a new radiation source or a new radiation area onsite or cause an increase in dose rates from an existing source?

(Consider types of welds that could become crud traps, addition of contaminated system piping, change in insulation on the component, etc.)

Does the activity create or increase routine maintenance, operation, service, or surveillance requirements in a radiation area?

Does the activity involve shielding changes, ventilation changes, or materials that contribute to radioactive crud, resin, or sludge treatment systems?

Does the activity involve the replacement of valves, valve internals, pump impellars, etc. that could impact source term (which contain cobalt alloys) on the primary system?

Does the activity potentially contaminate systems or components on the plant secondary side?

Does the activity cause an increase or potential increase in the amounts of radioactive airborne effluents or liquid fifluents or significantly alter the nuclide mix of such effluents?

Does the activity result in a new radioactive liquid or gaseous discharge point or decrease the ability to sample or monitor existing release paths?

Does the activity affect primary system chemistry controls?

Does the change increase the potential for the release of radioactivity to ground water?

Does the activity result in an increase of the potential for radioactive materials to be released into inaccessible locations, such as buried piping/cracks in concrete? Ifthe answer is "Yese then include site drawings and any temporary RCA's used to store licensed material.

• Signature required on Engineering Product Cover Sheet.

=- Mý -.- --- -.-

MINER Design Effects Table WDOMInIon 7.0 Environmental Impact (Non-Radiological) l[Yes Z No Program Owner Name:

El No Impact El Impact*

Does the activity alter water quality characteristics of any liquid discharge? (flow, chemical composition, temperature)

Does the activity create or increase a source of air pollution?

Does the activity alter any non-radiological environmental monitoring system? (meteorological or water quality monitoring system)

Does the activity:

" Take an oil pipeline out of service. Refer to station's SPCC Plan for guidance on 40 CFR 1.12(d)(2) requirements.

" Install or modify dams, wells or water treatment systems, sewage treatment plant, or sewage pump station

" Disturb more than 1/4 acre of land

" Add, modify, disable, remove, or relocate oil storage tanks or buried petroleum piping (Consider the potential for flooding)

• Change yard grading, storm drainage system, or yard surface

" Involve the demolition or construction of a building

" Involve the use of mobile or portable storage tanks or vessels

" Introduce a new source of hazardous waste

" Does the proposed change involve construction activity near catch basin/storm drain?

I~uclear Material Control EU Yes 0 'No Program Owner Name:

El No Impact E] Impact" The purpose of the following question is to assist the reviewer in determining the impact of the modification/activity on the Nuclear Material Control Program as it applies to each Dominion nuclear power station and to the Nuclear Analysis and Fuels (NAF) department. Nuclear material is a collective term for the NRC licensed materials: by-product, source, and special nuclear material. Nuclear material is contained in nuclear fuel assemblies, nuclear fuel rods, fuel assembly inserts, irradiated components, certain radiation detectors and sources, and various other nonfuel items.

Does the activity result in the procurement or shipment of items containing Uranium or Plutonium (special nuclear materials) or change the handling or storage of special nuclear materials?

• Signature required on Engineering Product Cover Sheet.

Form No. 731172 (Dec 20101

Design Effects Table Iq iýj~ p ,- 4 I 'l q- Sii 9.0 License Renewal Rule Program and Aging Management Activities Cl Yes Z No Program Owner Name:

ElNo Impact I[E rlpact*

In order to ensure compliance with the License Renewal Rule (10 CFR 54), a review must be performed to determine the impact that the activity may have on the License Renewal Rule (LRR) program. The LRR program credits aging management programs (AMAs and AMPs) for managing aging of plant structures, systems, and components (SSCs). AMAs and AMPs are intended to provide reasonable assurance of the ability of LRR required SSCs to perform their license renewal intended functions. [CM 3.1.7]

The purpose of this review is to determine if an activity requires a change to an AMA or AMP credited under the License Renewal Rule.

For example, changing the materials of construction of an SSC in the scope of LRR may create the potential for additional aging effects that were not considered or that the current aging management activity is not designed to detect. Removing or adding an SSC creates a potential for LRR documents to become inaccurate. Changes may also result in internal or external environments, chemistry parameters, or maintenance practices different than those previously assumed for plant SSCs as part of the LRR program evaluation.

" Does the activity affect a component identified in the Equipment Data System, the Master Equipment List, or License Renewal Drawings as being in scope for license renewal?

" Does the activity involve a component or structure associated with a fire protection system,, the SBO diesel generator, EQ cable, or anticipated transient without scram mitigation?

- Does the activity involve more than a minimal change in plant operating conditions of stress, temperature, radiation, or chemistry for more than 30 days and could affect the activities identified as in-scope for License Renewal in the first and second bullets?

" Does the activity involve the deletion or change in frequency of an inspection or method of discovering or evaluating the material condition of a plant SSC that is in-scope for licensing renewal aging management?

- Does the activity install a medium-voltage (2 kV-35 kV) power cable?

10.0 Generic Letter (GL) 89-13 Program El Yes 1] No Program Owner Name:

[I No Impact E I' mpact*

Does the activity impact any heat exchanger that uses Service Water for cooling?

Does the activity affect the chemical treatment of the SW system?

Does the activity affect the pressure boundary integrity of the Service Water System?

11.0 Station Blackout (SBO)

El Yes M] No Program Owner Name:

El No Impact El Impact*

Does the activity alter the design or operation of any SBO equipment? SBO equipment is identified in the MEL/BOM Database.

Does the activity potentially impact the operation of any SBO equipment (i.e., non-SBO equipment potentially affecting SBO equipment operation, such as an increase in the area temperature)?

• Signature required on Engineering Product Cover Sheet.

Form No. 731172 nlp3, 9inim

Design Effects Table FoominioW 12.0 Appendix J Program Z1 Yes DI No Program Owner Name: Paul Miller

• No Impact 1-I Impact Does the activity affect any contaihment boundaries? This includes changes which may affect any containment isolation valves (including closing force, stroke time, position indication, etc.) or the leak tightness of any containment isolation boundary, such as:

" Containment Isolation Trip Valves

" Containment Isolation Check Valves

• Manual Containment Isolation Valves

" Electrical Penetrations

" Personnel/Equipment Hatches

" Secondary Containment (MPS) 13.0 NERC - North American Electric Reliability Council

[*]Yes 1Z No Program Owner Name:

L- No Impact E] Impact*

Does the activity affect the net or gross electrical output, MW or MVAR, of the main generator?

Does the activity affect the allowable switchyard voltage range?

P ,.the activity affect any high transmission function or control; such as Electrical Bus Protection, Offsite Power Reliability; or I Ae/G enerator/Exciter Control/Model/Transformer?

Does the activity affect the NERC Compliance Commitments described Inthe Nuclear Switchyard Interface Agreements (NSIA's) located on the Nuclear Webpage - Nuclear Switchyard Reports?

Does the activity affect any Preventive Maintenance (PM) or Scheduled Maintenance Activity for Plant or Switchyard Equipment described in the Nuclear Switchyard Interface Agreements (NSIA's) located on the Nuclear Webpage - Nuclear Switchyard Reports?

Does the activity affect or potentially affect any Transmission Operating Guidelines or Work Procedures under Dominion agreements with our Transmission and Distribution organization (could be outside company or Dominion T&D depending on the station) as shown under website "Nuclear Switchyard Report"?

• Signature required on Engineering Product Cover Sheet.

FotrmNn- 7:1172 In-9fm1 tn

14.0 GSI-191 (Containment Recirculation Sump) Considerations Yes No Program Owner Name:

El No Impact LI Impact*

1. Does the change involve the addition, removal, or change of any fibrous material in containment (i.e., insulation, damming material, fire stops, cloth barriers, fire retardant, filters, or screens)?

2. Does the change involve the addition of any adhesive (labels, stickers, or signs) to containment?

3. Does the activity impact a containment SSC design basis (e.g., RWST level change)?

4. Does the change involve the addition of significant horizontal or vertical surface area (>100ft 2 ) inside containment that could collect dust oi dirt?

5. Does the change'affect the amount of qualified or unqualified coatings inside containment?

- Is a component with a qualified coating being added to containment within the GSI-1 91 limiting break areas?

- Does the change replace an unqualified coating with a qualified coating?

- Is a component with an unqualified coating being added anywhere inside containment?

- Are surfaces being left uncoated in containment?

6. Could the change affect the post-LOCA water level, sump temperature, or recirculation water flow paths in containment?

• Does the change add, move, or modify high-energy piping connected to the RCS?

• Will the activity involve moving, relocating, or repositioning any SSC, storage scaffolding, or temporary modification intended to be left inside containment during power operation?

7. Will the activity result in the conditions to produce any chemical effects or add to the generation of latent debris, concrete, or corrosion particulates as described in GSI-191?

" Does the change involve the installation or removal of reactive metals (aluminum), which will be subjected to containment spray or immersion during a LOCA?

" Is a change to an existing chemical quantity being made in containment (e.g., sodium hydroxide) or any new chemical being added to containment?

• Will the change involve exposing concrete by either the addition of new concrete, grout, or the removal of coating on existing concrete or grout surfaces, which will expose concrete surfaces to containment spray or immersion during a LOCA?

8. Will the activity result in the conditions to produce any downstream effects by adding to the generation of latent debris, concrete, or corrosion particulates as described in GS1-191?

• .Is any component or part being added which contains particulate (i.e. sand) or which could break down to particulate when exposed to high pressure steam/water jets or containment temperature and humidity during a LOCA (e.g., light bulbs)?

• Does the change affect the minimumflow clearance of any component in the ECCS system or make a change to the material of any component, which could experience abrasive or erosive wear from debris-laden water during post-LOCA recirculation? Does change

• affect Debris Plugging of ECCS Strainers? VA Plants [CM 3.1.41

9. Are there any components added, removed, or modified such that free volume or heat sink in containment is increased or decreased?

Ifyes, contact Nuclear Safety Analysis (NA&F) for input (may affect analysis).

Forth No. 731172 (Dec 201C

Document Impact Summary

PpDominioW Instructions: This checklist summarizes document and database impacts. The procedures, documents, software programs, and databases, which must be reviewed, are already listed on the Document Impact Summary form. However, the preparer may list (in the spaces provided) any other documents or other information, which require revising as a result of the activity.

Station Unit Change Document Number NKPS [i-MPS [INAPS LISPS 01 EI2 E]3 LI ISFSI TMOD-2012-11, Rev. 3

1. Procedures LIN/A Item Change Required?

a. Administrative Procedures EL Yes Z No

b. Operations Procedures El Yes Z No

c. Maintenance Procedures El Yes ONo

d. Engineering Procedures (including Rx Engineering) El Yes 0 No

e. Radiological Protection Procedures EYes ONo

f. Security Procedures El Yes ONo

g. Emergency Procedures El Yes ONo

h. B.5.b Procedures/Guidelines El Yes ONo

i. Severe Accident Management (SAM) Procedures/Guidelines EL Yes 0 No

j. Other Procedures (new Vendor procedure for leak seal injection) 0 Yes EL No

2. Drawings __ N/A Item Change Required?

i

a. Station Drawings []LYes ONo

3. Design Calculations Analysis E_ N/A Item Change Required?

a. Electrical Calculations LIYes El No

b. Mechanical Calculations (new vendor calc) 0 Yes EL No

c. I&C Scaling and Setpoint Documents / Calculations LI Yes 0 No

d. Computers / Software Design / Firmware LI Yes 0No

e. Class I Piping Stress Report EL Yes ONo

f. Non-Class 1 Piping Stress Analysis EL Yes ONo

g. Civil Engineering Building / Structural Analysis EL Yes 0 No

h. Pipe Support Calculations EL Yes 0 No

i. HVAC Calculations EL Yes ONo

j. Cable and Raceway EL Yes ONo

k. Nuclear Safety Analysis EL Yes 0 No I. Technical Reports (e.g., Westinghouse Scaling Document, etc.) EL Yes Z No

m. Other E] Yes ONo FormNo. 729529 (Dec 2010)

Document Impact Summary

Ppoominiow' Station Unit Change Document Number ZKPS LMPS E]NAPS rISPS N1 I-2 LI3 EL ISFSI TMOD-2012-11, Rev. 3

4. Other Documents [I N/A Item Change Required?

EL Yes . No

a. Updated Final Safety Analysis Report (UFSAR)/Update Safety Analysis Report (USAR)/DSARIISFSI-SAR

b. System and Plant Design Basis Documents (DBD) EL Yes Z No

c. Specifications El Yes [E No

d. Technical Requirements Manual(s) (TRM) EL Yes R No e.

1 Severe Accident Mitigation Guidelines (SAMG) Nuclear Safety Analysis [CM 4.1.1] for VA Plants LI Yes E_YesNo

[ No

f. Vendor Technical Manuals El Yes [No

g. Technical Specifications LI Yes Z No

h. Equipment Data System (EDS)/Q-List/EMPAC/MEL EL Yes [] No

i. Bill of Material(s) (BOM) / Spare Parts 0 Yes [No

j. Emergency Preparedness Program EL Yes [No

5. Key Regulatory Programs [NIA Item Change Required?

a. Fire Protection I Appendix "R" Program(s) o Fire Safe-Shutdown Analysis (e.g., Appendix "R" Report, or Branch Technical Position 9.5-1 shutdown analysis) o Fire Protection Program Document L] Yes El No

" Combustible Loading Analysis o Firefighting Strategies or Pre-fire Plans

" Other Program Documents, such as Exemption Requests or FP Engineering Evaluations

b. EQ Program L Yes El No

c. License Renewal (Aging Management) Program LI Yes EL No

d. Station Blackout Program El Yes El No

e. Regulatory Guide 1.97 EL Yes LINo

f. Generic Letter 89-13 LIYes LINo

g. Radiological Protection Program (ALARA) EL Yes EL No

h. Maintenance Rule Program EL Yes LINo

i. Appendix"J" Program El Yes L1 No

j. Inservice Inspection (ISI) El Yes EL No

k. Inservice Testing (IST) Program(s) El Yes El No

1. Environmental (Non-Radiological) El Yes LINo

m. GSI-191 Program El Yes LINo

n. Nuclear Material Control LIYes LINo

1. o. North American Electric Reliability Council (NERC) [ L Yes L] No Form No. 729529 {Dec 2010)

Document Impact Summary DominioW Station Unit Change Document Number MKPS LIMPS L-NAPS L']SPS 91 []2 I13 LI ISFSI TMOD-2012-11, Rev. 3

6. Other Process Documents and Databases [ N/A Item Change Required?

a. ERDS Datapoint Library EL Yes EL No

b. Preventive Maintenance Program (PM) EL Yes LINo

c. Coatings Program LI Yes LINo

d. Flow-Accelerated Corrosion (FAC) Program El Yes LI No'

e. Margin Management Program EL Yes EL No f, Equipment Reliability Program LI Yes LI No

g. In Containment Banned / Restricted Materials Program LI Yes LI No

h. Valve Packing Program El Yes EL No

i. Maintenance Check Valve Program LI Yes EL No

j. Maintenance Safety/Relief Valve Program EL Yes EL No

k. Snubber Program E]Yes LINo I. Human - System Interface Program LI Yes LINo

m. Seismic Qualification Program Lj Yes LINo

n. Post Maintenance Testing LI Yes EL No

o. Motor Operated Valve (MOV) Program LI Yes EL No

ý 11 p. Air Operated Valve (AOV) Program El Yes EL No

q. ReactiVity Management Program El Yes EL No

r. ODCM / REMP / REM El Yes EL No

s. Single Point Vulnerability Database EL Yes LINo

t. ASME Section VIII Vessel Program Database LI Yes El No

u. Containment Hydrogen Generation El Yes' EL No

v. Containment Debris Inventory El Yes LINo

w. Probabilistic Risk Assessment - PRA LI Yes EL No

x. ISFS1 - 10 CFR 72.212 Evaluation Report - for Spent Fuel EL Yes EL No

y. Critical Equipment List EL Yes [L No

z. Supply Chain Warehouse Stocking Database (SAP) LI Yes EL No aa. Underground Piping and Tank Integrity Program EL Yes [:.No

7. Training [NIA Item Change Required?

a. Nuclear Control Room Operator Development Program LI Yes YsN I] No (NCRODP) Training Modules

b. Simulator Changes El Yes [No

c. Other Training (e.g. Maintenance, NSS, ESP, etc.) I- Yes [ No

d. Training Dept. Training Impact Report (TIR) in accordance with TR-AA-530. [L Yese ]N No Form No. 729529 (Dec 2010)

Document Impact Summary

PODOMiniow Station Unit Change Document Number MKPS [-IMPS LINAPS E-SPS 01 [-2 1-13 El ISFSI TMOD-2012-11, Rev. 3

8. Kewaunee Only ZE N/A Item Change Required?

a. Concrete Rebar Cut Request E Yes 9 No b, Fuse Control Program El Yes 0 No

c. Motor Thermal Overload Heaters [I Yes 0 No

d. System Descriptions El Yes Z No

e. Improved Technical Specifications El Yes I] No

9. Millstone Only __ NIA Item Change Required?

a. Safety Functional Requirements Manual El Yes [] No

b. REMODCM E[ Yes El No

10. North Anna Only __ NIA Item Change Required?

a. North Anna Setpoint Document El Yes E]No

b. Precautions, Limitations, and Setpoint (PLS) Document El Yes El No

c. North Anna VPSSL El Yes ElNo

d. Lube Oil Manual ElYes ElNo

11. SurryOnly _ _ _NIA Item Change Required?

a. Lube Oil Manual El Yes ElNo

12. Remarks (Attach additional pages if needed):

Form No. 729529 (Dec 2010)

Form DDC-02, Revision 0 FDominion KPS Design Change Training Impact Evaluation E auto Process

- Refer to FORM INSTRUCTIONS for completing this form -

- For ResponsibleEngineer Use Design Change Number: Temporary Modification 2012-11 Rev.: 0

Title:

RHR-600 Strongback and Leak Sealant Enclosure A copy of the draft design description AND this form have been attached to the following CRS item:

CONDITION REPORT #: CR472226/472654 Responsible Engineer (RE): T LaHann / L Christensen 04/28/2012 Print Name Date For Training Use-_" ______ .. , ___ "_

Provide Training Provide Training Prior To Training Materials Training Required Turnover Plan Impacted Materials

?.  ? Tracking Impact Evaluation Tracking Discipline RFT No. Y N Y N Number(s) Performed By Y N Number(s)

Initial License CR472654 EI 9 El I] N/A Jon Robb E] 0 N/A Training Nuclear Auxdllary CR472654 El E9 [E M N/A Jon Robb El 0 N/A Operator License Operator CR472654 E] E El 0 N/A Jon Robb 11 0 N/A Requalification Shift Technical Advisor CR472654 F] 0 El 0 N/A Jon Robb D Z N/A Shift Manager CR472654 El 0 El 0 N/A Jon Robb El Z N/A Electrical CR472654 N/A Jon Robb El N N/A Maintenance l&C Maintenance CR472654E. Z El Z N/A Jon Robb El 0 N/A Mechanical CR472654 El NIA N

Jon Robb N/A Maintenance E2 Maintenance CR472654 El 0 El 0 N/A Jon Robb El 0 N/A N/A Supervisor Chemistry CR472654 El 0 El 0 N/A Jon Robb El 0 N/A Radiation C 2 0 N/A Jon Robb N/A Protection CR472654 El 0 El 0 Engineering CR472654 El 0 El Z N/A Jon Robb El 0 N/A Nuclear Employee CR472654 El 0 El 0z N/A Jon Robb El 0., N/A Training I I. .....

Other El El[ IEl _ ElEl Page 1 of I

Modification,Challenge Review SolDomilliow Attendance Sheet

, 1 , , ,

El Preliminary Design, 30%

Note: Not required for FMCR.

[] Intermediate Design, 75%

E] Final Design, 1009%

DC Number: TMod 2012-11 Rev 3 Date: 4/30/2012 DC

Title:

RHR 600 Leak Repair MODIFICATION DESIGN REVIEW BOARD Department PrintedName Signature

--b ~Al ~ j/ P-f~A4 14 CAA ;41-4 _____C-

= . .

Y2A*,

J4. co I keg

______ I_______ t________

ForniNo. 729556 (Jun 2011)

D DOIIO CM-AA-ECR-1001 REVISION 1 PAGE 23 OF 42

/

Modification Challenge Review

'PDom'iudow Attendance

,, . I, Sheet* -* * - - * . -,

E] Preliminary Design, 30%

Intermediate Design, 75% Note: Not required for FMCR. i LI Final Design, 100%

DC Number: I ,: QoI2 - 1/ Date: 4/29 /1 -

DCTitle: RL//g .ov L-- / ,-.'4..-

MODIFICATION DESIGN REVIEW BOARD Department PrintedName Signature

--4 J1 4

&.siVS c ' z4 T

,[21, -frk J ,.-..

FormNe. reSS58 (Jun 2M1)

Serial No.12-324 ENCLOSURE 2 FOURTH TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM 10 CFR 50.55a REQUEST NO. RR-2-4 TEMPORARY MODIFICATION PACKAGE 2012-12 (Revision 1)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Temporary Modification d

CM A A- 2 T.o- A Dominion Site Unit Year Temporary Modification Number Revision Number Work Order(s)

KPS 1 2012 2012-12 1 KW100896318*

KW1 00894787 I High Risk? QL YES 0 NO Provide information as required and attach additional sheets as necessary for each item.

PartA. TM Description(To be completed by Requestor/Originator/Engineering)

1. Title Second RHR-600 Leak Repair

2. Expiration Date [Not to exceed one refuel cycle (unless approved by site VP)J KR32 (Remove prior to entering Mode 3)

3. Affected Systems/Components/QA Class RHR (System 34)/RHR-600/SR

4. Reason (e.g., awaiting parts, testing, calibration, repairs, temporary power supply)

The WA" Pipe upstream of RHR-600 has a pinhole leak caused during welding associated with the installation of Temporary Modification'2012-11 in addition to the pinhole leak in the W" socket weld upstream of RHR-600. All aspects associated with the repair for thd pinhole leak in the 3/4" socket. weld upstream of RHR-600 are addressed under TM 2012-11.

This TM performs a repair of the leak in the W pipe caused during the installation of TM,.

2012-11.

5. Description (e.g., specific details on the aspects of the modification): Work Order (e.g., racks, oubicles, reference(s), and instrument Index as applicable, and locations building, area,,.elevation, and rooms to identifyin detail the location of the modification).

Attach sketches as necessary.

See following pages REVISION I makes the following changes:.

" ASTM Al 93/Al 94 Is an acceptable substitution for ASME $A1 93/194 material

" Specific reference to "RHR pressure" for determining maximum sealant Injection pressure is replaced with "operating pressure". -This is consistent with TM 2012,111 terminology.

" Editorial corrections

6. Documents/Drawings/Procedures to be Updated.

a. None

b. _"

C.

7. List any Mode Restrictions

• This TM may be installed in Mode 6 or Mode 5.

• The TM must be removed and the system restored prior to entering Mode 3.

• An NRC Relief Request must be approved before entering Mode 4 with the TM installed.

8. Action Plan for Removal - Close-out Document (RENDCP/DCR, Work Order, Procedure

Temporary Modification CMAA-D A H 3 Pag 2. o wr Dominion-changes)

Repair leaking weld and remove TM per WO # KW1 00894787. This WO also removes TM 2012-11.

9. Installation Instructions for TM The following steps for enclosure installation may be performed in any logical order

" Install leak sealant enclosure and torque the four hub clamp bolts to 33 ft-lbs

• Tighten the strongback halves together to 32 ft-lbs.

• Tighten bolting between the sealant enclosure and the strongback to seal the new hub clamp against the existing hub clamp (wrench tight). The vendor has no torque specification for this bolting [16]

" Tighten set screws to 165 in-lbs Inject sealant per GMP-206

10. Required System Testing Following TM Installation Verify no leakage

11. Removal Instructions for TM Per WO KW1 00894787.

12. Required System Testing Following TM Removal Permanent repair of the leak will follow removal of the TM. Therefore, there are no removal test requirements for this TM. ,R-

13. Requested By (Name - Please Print) 14/ ReIues d B (Signature) 15. Date Tim LaHann 05/02/2012 Part B. Design Engineering Reviews/Screening Evaluations Attached

1. CM-AA-RSK-1 001, Engineering Risk Assessment, Attachments , YES [ NO 2 and 3

2. DNES-AA-GN-1002, Document Impact Summary, Attachment 1 0 YES El NO

3. DNES-AA-GN-1003, Design Effects and Considerations, YES NO Attachment 2

4. DNAP-3004, Dominion Program for 10 CFR 50.59 and 10 CFR 72.48 - Changes, Tests, and Experiments (KPS only - perform ONE IS REQUIRED GNP-04.04.01 for Applicability and Pre-Screening prior to DNAP-3004) Z YES nI NO

• 50.59/72.48 Screen, Attachment 4 EYES I NO

• 50.59/72.48 Evaluation, Attachment 6

5. Additional Attachments

a. Attachment 1, Evaluation of 0.25" Circumferential & Axial Flaw on 3/4" RHR-600 Piping, KPS

b. DDC-02 20tl)

(May 730749 No.

No. 730749 (May 2011)Form Formo

Temporary A, 3 b Modification A *" o mw-Dominion-C.

d.

6. Summary: Descriptions & Conclusions Resulting from Reviews/Evaluations Performed in Step 3.1.8 and Above RSK-1001: The TM is medium risk AA-GN-1002: The procedure to be used for leak sealant injection for TM 2012-12 was previously approved for TM 2012-11.

AA-GN-1003: Components within the ISI boundary are affected.

50.59: The TM screens out of 50.59

7. Additional Reviews and/or Comments - Provide Details Below None

8. Prepared By (Print & *gn) 9. Date Tim LaHann 05/02/2012

10. Independent/DesignAuthori Revie 11. Date (Print & Sign) *L* *05/02/2012 Lori Christensen Additional reviews for Implementing Organization, Training, etc. as necessary in accordance with Section 3.3

12. Name (Print & Sign)/Department 13. Date

a. 1.

b. 2.

c. 3.

d. 4..

14. Engineering Supervisor 15. Engineering Supervisor 16. Date (Name - Please Print) (Signature)

Joe McNamara -?ne- 5/2 *./2.

Part C. Operations Review (To be completed by Shift Manager/Designee)

1. Are controlled Station drawings affected by the TM and are Ej YES Oj NO U NA they attached?

2. Necessary Station personnel are informed of the TM? [ YES LI NO LI NA

3. Temporary procedure changes and Temporary procedures L] YES rI NO I] NA are implemented to support the TM?

4. Evaluated need for check valves and/or other anti-siphon Li YES Ln NO R'NA protection, if TM utilizes piping or hoses?

5. Limiting conditions and special requirements indentified.

Note: This TM may be installed in Mode 6 or Mode 5. ' YES [: NO [L NA The TM must be removed and the system restored prior to entering Mode 3.

Form No. 730749 (May 2011)

Temporary Modification C -A - C2 AT 3 Pg 4 WDomin.ion-An NRC Relief Request must be approved before entering Mode 4 with the TM installed.

6. TM verified not to violate Technical Specifications, not to create a hazard to Station safety or personnel, or conflict with existing Station conditions?

Note: This TM may be installed in Mode 6 or Mode 5. .IYES r-] NO E] NA The TM must be removed and the system restored prior to entering Mode 3.

An NRC Relief Request must be approved before entering Mode 4 with the TM installed.

7. TM Tags generated in accordance with applicable Site X YES El NO DNA Tagging procedure and any special instructions included on the tags?

8. TM Log updated?. 4 YES El NO L] NA

9. Engineering Post Installation Walkdown Requested? LI YES M NO E] NA

10. Quarterly walkdown required? E] YES L1 NO [] N-A

11. PRA risk impact has been assessed and, if determined to YE be applicable, has been entered into the Risk Monitor and/or Y NO LI NA Shutdown Risk Assessment?

12. The affected Unit shall not exceed the operating mode of: 4 or 5

13. Concurrent or Independent Verification:

14. Functional Check or Visual Inspection: Per WO KW1 00896318

15. The opposite Unit shall not exceed thle operating mode of., NA

16. STA Concurrence (Print & Sign) for Virginia Plants Only [Commitment 5.1.4] 17. Date NA NA

18. Shift Manager/Desig pe Print & Sign) 19. Date Ethan Treptow -3 2.

Part D. FSRC Review (as applicable, Refer to LI-AA-600)

NOTE: If the TM is used to move radioactive fluids or gases, the Manager Radiological Protection or Radiological Protection alternate must be a member of FSRC NOTE: TMs which could affect Nuclear Safety must be reviewed by FSRC

1. Is FSRC Signature Required? I YES LINO

2. FSRC Authorized Duration S','"3

/ Form No. 730749 (May 2011)

Temporary Modification C AT 2 IA T A N. 3 ior.Domninion-S3. FSRC Chairman Approval (Print LI-/1

&. ign)

/ /'L

4. Date Part E. TM Installation (To be completed by applicable personnel)

1. Shift Manager Approval for TM Installation (Print & Sign) 2. Date

3. Shift Manager Comments (includes any additional requirements in accordance with Section 3.4, "Implementation")- Provide Details Below No0j

4. Installation Completed By (Name - 5. Installation Completed By 6. Date Please Print) (Signature)

7. Independently Verified By (Name - 8. Independently verified By 9. Date Please Print) (as applicable) (Signature) (as applicable)

Notity the Shift Manager that TM is installed and the required post installation testing can be performed.

10. Instructions Used for Post Installation Testing Verify no leakage

11. Testing Performed By 12. Testing Performed By (Signature) 13. D~ate (Name - Please Print)

14. Post Installation Testing Satisfactorily? LI YES ,I NO

15. Required Administrative Controls Established? EL YES [L NO

16. Is Engineering Post Installation Walkdown Required?

NOTE: Responsible Engineering signature is in accordance with Step EL YES [I NO 3.4.6 of CM-AA-TDC-204.

17. Responsible Engineer (Print & Sign) (Engineering walkdown completed) 18. Date

19. Shift Manager (Name - Please Print) 20. Shift Manager (Signature) 21. Date Part F. TM Restoration (To be completed by applicable personnel)

1. Shift Manager Approval for TM Removal (Print & Sign) 2. Date 3: Shift Manger Comments (includes any additional requirements in accordance with Section 3.7, "TM Removal") - Provide Details Below

4. TM Restoration Completed By (Name - 5. TM Restoration Completed By 6. Date Form No.730749 (May 2011)

Temporary Modification C A 3 Pg

• .,. 6 1 Please Print) (Signature)

7. Independently Verified By 8. Independently verified By 9. Date (Name - Please Print) (Signature)

Notify the Shift Manager that TM has been removed and the required restoration testing can be performed

10. Instructions Used for TM Restoration Testing

11. Testing Performed By 12. Testing Performed By 13. Date (Name - Please Print) (Signature)

14. Shift Manager (Print & Sign) 15. Date Part G. Post Restoration Review (Completed by Shift ManagerlDesignee following restoration)

1. Post restoration testing completed satisfactorily? El YES [j NO

2. All Documentation satisfactorily completed? El YES Li NO

3. Temporary drawings removed from Control Room/any other EZ YES EI NO location?

4. Procedures changed to eliminate TM? (Review Part A) [I YES EL NO

5. Necessary Station Personal notified? LI YES [1 NO

6. TM Tags removed? El YES [1 NO

7. TM Log updated? E YES E3 NO

8. Shift Manager (Print & Sign) 9. Date Part H. Monthly Audit INITIAL DATE COMMENTS ON AUDIT 1.

2.

3.

4.

5.

Form No. 730749 (May 2011)

Temporary Modification C T A H 3 P 7 of 1 PDominion-Form No. 730749 (May 2011)

Temporary Modification CWA-D-6 *TAHMN3 Pag 8 of 113

-W Dominion-Part A, DESCRIPTION RHR-600 is a 3/4" SR sample valve located off RHR line 10-AC-601R1 1. A leak repair enclosure, which incorporates a strongback design, was installed to seal the pinhole leak In the 3" socket weld under TM 2012-11. In addition, a structural restraining clamp was installed to ensure the %"pipe and/or RHR-600 could not catastrophically separate due to full circumferential failure of the leaking 3/4" sockolet weld.

The scope of this TM Is limited to installing an additional leak repair enclosure to address the leak caused by welding during TM 2012-11 installation. TM 2012-12 is a separate leak repatpr activity that does not postulate failure of TM 2012-11 as a design consideration.

This is a short duration TM, which allows the unit to enter Mode 4 from Mode 5 for permanent repair of the fitting leak. Upon entry Into Mode 4, the A train of RHR can be removed from service and the leaks repaired.

Quality Classification RHR-600 and associated pipe and tubing is SR [1]. All parts used for the leak repair enclosure are SR.

ASME 04PY Code RHR-600 and adjoining pipe.Is classified as ASME Section Xl Class 2 [2]. ASME Sectlon XI, Appendix IX provides direction on the use of mechanical clamping devices on Class 2 and 3 piping pressure boundary. The requirements of this appendix will be used to demonstrate the acceptability of this repair on the leaking pipe and fitting.

Note that Article IX-1 000 prohibits the use of a clamping device 6n portions of a. system that forms the oontainment boundary. RHR-600 Is credited with maintaining containment boundary and therefore restrictions on unit mode changes apply to this TM.

Auxilfary CoolIng (AC) Syste!m Tegmerature and Pressure The design pressure and temperature of the piping, AC-601 R-1 1, is 600 psig and 400 'F [5].

The leak repair enclosure will be designed to these eaordltions, The TM will be installed and the sealant compound Injected in Mode 6 or Mode 5. When the leak enclosure Is unvented, injection pressure will be limited to:

Operating pressure + Pressure required to initiate injection. (static) + 1 psi This injection pressure restriction, combined with the volume restriction on sealant injection, prevents intrusion of sealant into a plant system. During sealant compound injection, the maximum RCS temperature will not exceed 1950 F [7]. The X-36 sealant is rated for injection from a temperature of 00F up to 400"F [15], and is therefore suitable for this application.

Overview A hub clamp assembly (leak repair enclosure) designed and fabricated by TEAM Industrial ServIces, Inc will be used. The leak repair enclosure mounts directly on top of the TM 2012-11 enclosure [12].. 5/16-18-UNC cup tip set screws, tightened to 165 in-lbs, are included in the hub assembly. These set screws tighten against the body of valve RHR-600 near the pipe socket pisifft Mg,W'RIAP (MAY

ý011)

Temporary Modification I C-A-TC0 ATAHMN 3 Pag 9 of 13 OFDominion- I connection. The size, number, and installation torque on the sets screws are established to ensure RHR-600 does not separate from the 3/4" pipe during sealant injection [12].

The leak repair enclosure is made of stainless steel and, along with the X-36 sealant, provides the pressure boundary up to the design temperature and pressure of the AC pipe, which is 600 psig and 4000 F. The studs, which hold the two halves of the leak repair enclosure together, are tightened to 33 ft-lbs [12].

General Desicin Requirements (IX-3100)

Defect Characterization:

In addition to the original defect addressed by TM 2012-11, a second defect was created during welding associated with TM 2012-11 installation. This is a pinhole defect caused by melt through during welding. Ultrasonic examination has verified the structural integrity of the pipe around the circumference of the pipe to a close proximity of the defect. This defect was conservatively evaluated in Attachment 1 as .25" through wall hole. The evaluation concluded that the flaws in circumferential as well as axial direction are stable and the 314" RHR piping will maintain its structural integrity with the flaw.

Materialcompatibility [12]: The hub clamp is constructed of SA240/479 Gr 304 material. Bolting material is A/SAl 93 Gr B8 Class 1 or Class 2. Threaded rod for the strongback enclosure is A/SA193 Gr B8 Class 2 or A/SA193 Gr B7, and nuts are ANSA194 Gr 8 or 2H. Set screws are ANSI B18.3 stainless steel. All materials are acceptable for use in this application.

TEAM Industrial Services compound X-36 will be injected into the enclosure to provide temporary repair of the system pressure boundary. Sealant X-36 has a low concentration of Halogens/Chlorides and is acceptable for use on stainless steel 18].

SealantInjection: Per Reference 12, the leak repair enclosure is adequately designed for this application. These calculations establish the enclosure size and construction, as well as maximum sealant injection pressure and volume necessary to provide the system pressure boundary. Identifying maximum sealant injection volume prevents over-injection and intrusion of the sealant into plant systems.

Clamping Device (IX-3200)

As described above, the leak repair enclosure is constructed of SA2401479 Gr 304 material.

The sealant enclosure is a clamshell design with the halves joined together around the leak using four fasteners. The enclosure is then rigidly attached to clamps on the 10" RHR line.

The enclosure is designed to meet Table IX-3200-1 Level A stress limits as well as the bounding pressure and temperature requirements of the Auxiliary Cooling System [12].

Pipingi System (IX-3300)

Leak Repair

Enclosure:

The leak repair enclosure will be rigidly attached to the 10-inch pipe using a strongback design. The friction between the clamp and the pipe is significant compared to the seismic forces of the relatively light weight clamp and enclosure. This enclosure and the TM 2012-11 enclosure are rigid structures that are rigidly, mounted to the 10 inch pipe therefore, there will be no adverse interaction between the enclosures.

Form Noz730749 (May 2011)

Temporary Modification

• MA

• 0 A TA C.- I o WDominion-Vibration: The pipe in question is at the outlet to the RHR heat exchanger and is well supported near the repair area. *Consequently, vibration was not considered in the design of the leak sealant enclosure.

Pipingand PipingSupport Evaluation/Stress:

The following are the additional loads to the residual heat removal piping in analytical part number RHR-34-004 associated with TM 2012-11 and TM 2012-12:

Weight of the existing temporary modification is 71 lbs Weight of the structural clamp that was installed is 57 lbs (based on hand calculation)

Valve RHR-600 concentrated weight is 15 lbs Weight of the leak seal box is less than 55 lbs The total load addition on piping is 198 Ibs; this vertical load will be shared between supports with hanger mark number RHR-H9 (capacity is 3240 lbs, with a design margin of over 1200 lbs) and RHR-H29 (capacity is adequate per existing calculation, with a design margin of about 450 Ibs). A greater part of the load will be seen by RHR-H9 because it is closer to the point of application of the load (analytical node 2062). Conservatively assuming that RHR-H29 will see 25% of the applied loads, there will be approximately 400 lbs of margin remaining. Therefore, the RHR-H29 support can still accommodate an additional 330 lbs without failure after allowing for a safety factor of 1.2.

The maximum stress (combined stress) on the piping at node 2062 is approximately 30% of the allowable. Adding 198 lbs will not overstress the piping, so there is no concern from a membrane stress perspective. Fundamental frequency of vibration of the system is 22 Hz which is within the rigid region of the response spectra curve. This piping is therefore seismically stable and adding an extra 198 lbs will not result in a significant change in the dynamic response. Thermal forces and moments are low and within acceptable limits while displacements and rotations are negligible.

Based on the review of pipe stress analysis report, pipe support design calculation and design drawings there is adequate design margin for the piping and supports to take the additional loads to be imposed on the RHR piping system. [3, 18, 19, 20, 21]

Joint stiffness: There is no appreciable change in the stiffness of the RHR piping system due to the addition of the leak sealant enclosure.

Constrainingeffects: The clamp holding the TM 2012-12 leak repair enclosure is a single clamp band. The temperature increase experienced by the clamp from the point of installation until its removal will be approximately 150'F (from approximately I 00°F to 245°F [17]). This temperature increase will result in a radial expansion of the 10" diameter pipe by no more than 0.008". Assuming the entire stretch load of 0.008" is taken by the strongback bolts, the total stress within the bolts remains below 60% of material yield. Note that TM 2012-11 addresses the effects of the components that were added by that temporary modification.

Form No. 730749 (May 2011)

Temporary Modification MT TTA

• M Pae 1 WftDminion Defect growth: In addition to the original defect that is addressed by TM 2012-11, a second defect was created during welding associated with the TM 2012-11 installation. This is a pinhole defect caused by melt through during welding. Ultrasonic examination has verified the structural integrity of the pipe around the circumference of the pipe to a close proximity of the defect. This defect was conservatively evaluated in Attachment 1 as .25" through wall hole. The evaluation concluded that the defect in the circumferential as well as the axial direction is stable and the 3/4" RHR piping will maintain its structural integrity with the defect.

The defect is not expected to grow due to the shape of the defect, the integrity of the surrounding base metal, and the known cause of the defect being the melting of the pipe base metal. The friction resistance of the pipe clamp and the rigid connection of the leak repair enclosure stabilize the % inch pipe and valve, reducing stress. Therefore the defect is not expected to grow during the very short timeframe this TM will be installed.

Monitorng Requirement IX-6000)

Compliance with the monitoring requirements of IX-6000 shall be as described in the relief request.

Sealant Iniection The sealant will be injected at a pressure based on the system operating pressure. Thus, the system pressure will prevent the sealant from entering the system piping. The allowed quantity of sealant to be injected is calculated to be 5.7 in 3 [12]. This volume will be procedurally controlled via GMP-206, Leak Sealant Injection Repair of Steel Components.

Sealant will be injected to stop the leak up to maximum predetermined amount of sealant. The acceptance criteria for post modification testing is zero leakage.

Containment Isolation The leak repair enclosure will be located on piping outside of containment that is between the containment and the outside isolation valve. Therefore, the leak repair enclosure will need to [9]:

1. Meet Safety Class 2 design requirements.

2. withstand containment design temperatures

3. withstand internal pressure from containment structural integrity test

4. meet seismic Category I design requirements

5. be protected against a high energy line break outside of containment when need for containment isolation.

6. maintain leakage within the Current Licensing Basis RHR-600 is an outside Containment isolation valve for Containment Penetration 10, a Class 6 penetration. Penetration Class 6 is a system required to operate post-accident. The design and operational criteria for penetration Class 6 isolation valves are governed by the functional requirements of the system. The isolation valves at Penetration 10 are not relied upon to preventthe escape of Containment air to the atmosphere [14].

Form No. 730749 (May 2011)

Temporary Modification CMAATC24ATCMN A2 3 Pag of 13 Dominion-The requirements for Containment Isolation are satisfied since the enclosure will satisfy the design requirements of the RHR system. The leak repair enclosure will be designed in accordance with ASME Section Xt, Appendix IX,Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary. The piping remains seismically qualified and the enclosure will remain intact during a seismic event. The leak repair enclosure will maintain leakage within the Current Licensing Basis, which for this TM, is zero leakage. High Energy Lines Breaks are not postulated until the RCS temperature is greater than or equal to 540 F [22]. Since this TM will be removed before Mode 3, HELB does not apply.

Operations The leak seal enclosure is located near RHR-600. The installation of this TM does not restrict operation of the valve or require any modification to the valve itself other than the leak repair enclosure will be sealing against the valve body near the %"pipe to valve connection.

References

1. XK-100-18, Rev. BA

2. ISIXK-100-18, Rev. AC

3. M-962-2, Rev. A

4. ANSI/ASME Code Reconciliation For Replacement Material, Parts, And Components, Kewaunee Power Station, Revision 3, July 6, 2010

5. XK-100-371, Rev. 5

6. ASME B&PV Code, 1998

7. OP-KW-GOP-102, Rev. 13

8. Consumable Material Evaluation (CME) 10000005744

9. ANSI N271-1976, Containment Isolation Provisions for Fluid Systems.

10. Not used

11. CEM-0049 and Addendum O0A, Rev 000, Evaluation Of Compensatory Measure Taken In Response To Identified Leakage At 3/4" Drain Line Valve 2-RH-33 Off Line 14"-RH-1 18-602

.12. TEAM Industrial Services Engineering Order 91672

13. System Integrity Program, Rev. 6

14. Darrell G. Eisenhut, NRC Director Division of Licensing, to C.W. Giesler, Wisconsin Public Service Corporation, Exemption to Certain 10 CFR 50 Appendix J Requirements, dated September 30, 1982.

15. TEAM email, 05/01/2012

16. TEAM email, 05102/12

17. OP-KW-NOP-RHR-001, Rev. 17, Sect. 5.11 RHR Alignment for Exceeding 245'F (Split Train Mode)

Form No.730749 (May 2011)

Temporary Modification SCM A A3o M Pe f Dominion-

18. Calculation RHR-34-004 revision 1, Pipe Stress Analysis of Residual Heat Removal Piping, Analytical Part RHR-34-004

19. Calculation S-062-RHR-34-004 revision 5, 79-14 HANGER DESIGN VERIFICATION FOR HANGERS ASSOCIATED WITH ANALYTICAL PART RHR-34-004

20. Drawing MS-34-78 Rev. OOA

21. Drawing MS-34-79 Rev. 00B

22. KPS USAR IOA.5, Analysis of High Energy Line Breaks Form No. 730749 (May 2011)

hIndustrial Services Registration# F-003143 ,

Engineering Department Tel: (281) 388-5695 Fax: (281) 388-5690 ROUTING SLIP & COVER SHEET FOR NUCLEAR SAFETY RELATED JOBS Branch Work Order #: 203-05224 Status: Priority Caller: Josh Thompon Customer: Domion Energy Safety Review #: 91666 En Order #: 91672EM Rev. I Name: Signature: Date: Time:

Data Taken By: Andrew Campbell 4/30/2012 13:12 Designed By: Jack Quackenbush 04/30/12 20:00 Heather Hodges 05/01/2012 09:56 Mitchell Williamson 5/1/12 15:15 Verified By: Felipe Arreola 04/30/12 22:15 Mitchell Williamson 15/1/12 10:53 Andrew Campbell 5/1/2012 19:30 Shop Received By: _

QC Received QBBy: Cy ~~~~~ . . . . .. . . .Rci. e . . .. . ..... .~~ . . . . . ... ...........ii... . . . . . .. . . .. . . .. . . .... .......... i 1Specifications:

Design Pressure 600 psi Design Temperature: 1400 OF Service:

- ..- IBorated Water---

. .. .. .. . . ... -.. .... . ....... ....... r. Value:33 . . ...

...... .. ....  ! __ - FT*LB

_ - --- -. 3 4 .. .... .... .

Total Weight: 47.91 LB Void: - 4.24 in BC, Sealant Type: .--

---

....---.. ~ ~ wX36w/ G-Fiber Note: A ECO.#: 12505 Do Not Paint 1 DATE: 5/2/12 Note iNote:

... ............. .....- iASMEff

......................... SECTION XI, i*.... .. ................ . . .. ......... ....- ]

APPENDIX IX QC FINAL INSPECTION REQUIRED Nuclear - Safety Related MTRs and COCs Required Bll of Materials:

Description:

Material: j__y:

1 Hub Clamp 4SA-240/479 GR.-304 12 IStrongback (top.... SA-240/479 GR-304 .................... 2 Strongback (ottom) iSA-240/479 GR-304 12 5/8-11-UNC Studs.. SA-193 GR-B8 CL 1I6. . . .

.. Nuts . ......... ...... . .........

_5/8-ll-UNCNuts .. ASTMA-194GR-8 (Al) 18 5/16-18-UNC Set Screws j ANSI B18.3 Stainless Steel . 8 i Sealant i....

X-36 I ......

Tube Rev. 3/3/2010

NO-%

mg a I1w - I I 5IIvvy,-% a %

q ------

-EXISTING HUB CLAMP 91511EM NEW SHOWN

-NEW HUB CLAMP

1. ASSEMBLY VIEW 2, APPROVED TO MANUFACTURE

3. NUCLEAR - SAFETY RELATED

4. DO NOT PAINT 5.

6.

LIMIT INJECTION PRESSURE: 1000 PSI + STATIC ASME SECTION XI, APPENDIX IX T EA M Industrial

7. COMMERCIAL GRADE DEDICATION / PMI REQUIRED ENGINEERING ORDER# 91672EM

8. THE VENTED PRESSURE IS 1000 PSI + STATIC UUMOMMSroaME

9. THE NON-VENTED PRESSURE IS SYSTEM PRESSURE + I PSI + STATIC DRAWING # N/A MACHINEDSURFACES

.10. PLANT TO PROVIDE SA-193 GR B8 CL 2 ALLTHREADS CONECTING BRFKSWAP CORNERS

.0 THE 4 STRONG BACK STRAPS (4) TOTAL ALL THREAD TOLERANICES:

TEAM WILL PROVIDE THE CLAMP STUDS AND THE 2 STUDS CONECTING II.- MAC1MONAL 2l/32 THE CLAMP TO THE STRONG BACKING )RAWN BY: JMQ/HH/MCW 15/1/2012 ANGULAR TWOPALCE 11/2' DEC00AL10.01 SEND (18) 518-11-UNC HEAVY HEX NUTS INRAPLACUDECAWA20C05 SIZE SEND (4) EXTRA SET SCREWS PER PLANT REQUEST 'HECKED BY: FA/MCW/AC 15/1/2012 ALLDVAENSIONS WWHES ISCALE: 1:4

"SAFETY FIRST - QUALITY ALWAYS" 03/4" 5/8" STUDS (4) PLACES TQ: 33 FT*LB 5/8" 0 3/4" REGISTRATION # F-003143 1.

2.

2) HALVES REQUIRED, (1)SHOWN

.110 CRUNCH TEETH (2.00 BORE ONLY) -ITEA /,Industrial Services, Inc.NIT. PERIMETER V0l IA,&RC

3. 3/16 X 0.09 TUBING GROOVE (HUB FACE) ENGINEERING ORDER# 91672EM LBSIVOL INA3 BC LUSSnOIN00V10SPEWD

4. INSTALL STAINLESS TUBING DRAWING # N/A IACNE 3+/-0ACES Dominion Energy

5. (2)D/T 1/8-NPT INJECTION PORTS (CAVITY) TECO PART # N/A BREAKS/LRPCORNM.0M5

6. NO WELDING Hub Clamp WPS: N/A

7. DIMENSIONS TYPICAL UNLESS OTHERWISE NOTED F+/-ACSIONAL SI/O ANGUJLAR 11li?

DRAWN BY: JMQ/HH/MCW 5/1/2012 TWO0PALCIDEC/AL 10,01 TAMEEPLACE CEOMAL 10005 SIZE A REV. 1-CHECKED BY' FA/MCW/AC 5/1/2012 ISHEET SHEET 22 OF (k;3 ALLDMADMPIOSN ACHES 3CALE: 1:1.5 3 SCALE: 1:1.5

"SAFETY FIRST - QUALITY ALWAYS" M p 1 1/4" 5/8", TOP BOTTOM 5/8" 11/2" R5 3/8" R5 318" 13 1/41t DRILL &TAP 5/8"-11-UNC 1-1/4" DEEP R6 1/8" R6 1/8" 13 (1) PLACE 3/4" 0 3/4" 5/8 STUDS (2)PLACES TQ: 32 FT*LB 11/2" -2 1/41.' 1 REGISTRATION # F-003143

'JEAM Industrial Services, or-sr- I c: I v \r*

%/n1"* IKlA*

I A DAf" ENGINEERING ORDER# 91672EM NT. I RSIOL INA3 RC UNt05SOMERYMSPE0C+/-INE DRAWING# N/A MAC14NEDSLUFACES Dominion Energy

1. (2)STRONGBACK RING SETS REQUIRED TECO PART# N/A RE+/-AK SARPCORNERSXUS Hub Clamp.

2. (2)HALVES PER SET, (1)TOP HALF, (1)BOTTOM WPS: N/A IOLERAMM:5

3. DIMENSIONS TYPICAL UNLESS OTHERWISE NOTED FRACIINAL *132 A

4. SEND (2)5/8-11 UNC ALLTHREAD 6"LONG DRAWN BY: JMQ/HH/MCW 15/1/2012 ANGULAR IWOPALCEDECUAAL

• I/2' 20M0 LSEOMAL THREEPLACE *GM SIZE A REV CHECKED BY: FA/MCW/AC 5/1/2012 ININCHESU ALL MENSIONS SCALE: 1:3 ISET I

Industrial Services, Inc. Sheet 1 of 1 Registration # F-003143 MATERIAL SPECIFICATIONS AECO#: 12505 17 Non-Critical/Nuclear Z Critical/Nuclear Drawn By: JMQ/HH/MCW Date: 05/01112 Engineering Order No.: 91672EM Rev. 1 Checked By: FAIMCW/AC Date: 5/1/12 Enclosures Material Specification JMTR COO NR PIPE FITTING ROLLED PLATE BLOCK / PLATE / SIDEBARS SA-2401479 GR-304 X X ENDPLATES_

STRONGBACK BARS SA-240/479 GR-304 X X S.B. EARS/FINGERS Fasteners J...I.I..__

STUDS ENCLOSURE SA-193 GR B8 CL 1 X X STUDS STRONGBACK SA-193 GR B8 CL 1 X X NUTS ENCLOSURE SA ASTM A-194 GR 8 (Al) X X NUTS STRONGBACK SA ASTM A-194 GR 8 (Al) X X SET SCREWS Stainless Steel ANSI B18.3 X X HTS FLANGE, TEE ,,,,,

RUN (if fabricated)

BRANCH (if fabricated)

FITTING_

WELD-O-LET _

VALVE FES Miscellaneous 05/26/2009

.~ .~

• TEAM INDUSTRIAL SERVICES, INC.

ENGINEERING DATA COVER SHEET Rev. 9128106 0 Work Order#: Routne: Engr Order Customer.,LS T:FlOhr Safety Review #

Ctistomer# , Ship To:BP#

t. o.,I_____X__w Technician Name*s): Date: Operatio I HTS Tech Support Rep (If Applicable):

PRESSURE TEMPERATURE Design: Design: Service:

Operating: Operating: Une Material:

~so36q s.51 Ute Size Flange Ratn Quantitynt Selection: Material Requested:

Package Requirments:

Check all that are imnedatel needed) Check one onl)

Drawings R For Immediale Manufacture

! Calculaltions wait for Approval of Package I Price Prime (ifsoE]ballpark or EAby enginermng) II] Price Only / No Calcuatlons/Drawings Date and Time Required: ,44P FaxIEmaiI Prints and Calculations: Pf dX .b,'tes (9)&Atv, v To Customer Name: Fax No:.

To Branch E-mail:

OtifY Brach SUPervisor a&r Wng Name,: Ai Phone No: 6p 6&6N&-

f&Iiraimedistelv []Next Business M~y 4sref urj.:e/(-/5 '

Special Requirements: [] Strongback needed for separation vibration

[ Stress Relief Reluired by Customer -] IfLlflrg Lugs am required, where should they be locatel?

LI] PE Stamp [] Are there specifi requirements I codes for this customer?

MTRs and COCs [] Other WONDE DESCRIPTION OF ITEM TO BE BUILT (HTS ONLY) STYLE OF FITTING (HTS ONLY)

.

• Lne Sop* Fffin: *,' ' F-IFul Encircle tK~rdobdn9 Saddle (0[" with Nozzle?)

.E].... StopFilling (Saddle Is non-press ure.retalnng)

El-'HI-Stop Fiftfnt . .. Bolt-on, " ... ",

T iOther: Et

[13 SaddlaIaScarfeNo l-Tap Size: Efitng Line Schedule: Requested K'ozlrSchedula:

FORM 104.2 Quality System Supplement TEAM Corporate Rev: 5 Page I of 2 NUCLEAR AUTHORIZATION I CONTRACT REVIEW CHECKLIST (NACL)

I TMS Division

• PART A: Used for material only orders and in conjunction with PART B for service related work.

Submitted by: J". K.-U Branch: ,5 Date: d/-,5 Utility: Z *,, Plant I Location: ,

Contact:

/I4,ko.ben&.- Phone #: ' o &t - ,g ,,

Customer Contract/Work Order/PO#: Team Job #: &63*e-51 W Personnel: 1. 2dditiona3. 4.

Shift:)1~ ~Additional Personnel:

I. All Pre-JobSteps must be initialedby (leTeam IndustrialServices, Inc. technician tho is leading activities at the work sitepriorto any wtorbeingpelfobIe.L

2. The OwnerRepresentathemust initialStep 8priorto any work beingperformed.

S All *'fLq mn.nlnvw~,* nn 21w moh m~it't .'Mo' ,'%n *0 ,,rinr In) nm~ twtv5r Sine' nswfhrmnd 0;MS.-n R. "M

1. Has JSA been completed? ¶ ] " " q-30d1-DZ.

2. Has Tech. Support been contacted? (Critical Job Review) U l.5'L CJ '966(~

3. Have Tean personnel READ and do they UNDERSTAND the Owner Engineering Control? j.O -0 . FC, 17- - t I

4. HaveTeampcrsonneJverifiedprocedure(s)tobeused? Li Lj J. 'P-ot. Pocedurdl: 3,30 S. Have Team personnel READ ad UNDERSTOOD the Owner W.O. Scope? 0 1 T, 4q.30-f

• W.O- #:/ W (d %q~,Y"

6. Have Team personnel been briefd on Owner W.O. Scope anl received copy .i-r, q-o-ir-of W.O.?

7. Have Team pelonne had aRa d.Protection Brief?

• LI *YST S. HasOwnerRepresentativeverifiedallStepshavebeenperfonred? _ _ - S". 9"Sd *OwnerRep.to initial.#.

9. Signatures of all Tean1Perne>lI. C_

- PART B: Used in conjunction with PART A for all service related work. PART B Is not required for material only orders.

Lb..,..' .....

aAfl,=.-~fl,.ical Support LineSiz: /s sQuality Assurance: Yes No I]

3 Lieie:/, t 1q Service-&,.,,17 Healthhsc: Yes I-No D Unit: ;of 1 Exp. Rate: %5-0 p,L Procedure #:33,01, Rev. #:

OP. Te p: .. ,,.* Op. Press: 3 5S..6.,... Materials & Equipm..ent Design Temp: qN/6 Design Press: 6 6044 Sealant Type: )A-3 'Jt'.c Lot #:

Equip. ID#ff. 12 #JZ-6 6 SealantType: - 'e-,A't.4.c Lot#

Comments/Description of Job: Fabrication Data Kea e 0. 4- 6&m k4dv-- Mill Tests: YesI~ NoE] Sec. lllfý esNo)W Has the Customer expressed mechanical integrity concerns for the component? Yes CA No [1 Plant Rep. (print):/ (,SInitials: #& Design Cats.: Yeslt.y No s] Engr. Dwgs.: Yes . NooEl TISI Tech. (print): OQ Initials: T1 F j Po e Material Identification: Yes E l No

FORM 104.2 Quality System Supplement Rev: 5 Corporate Page 2 of 2 NUCLEAR AUTHORIZATION I CONTRACT REVIEW CHECKLIST. (NACL)

TMS&Division LeakSeverity: Notif-_, ons: Safety Related: Yes [] No I]

Service: Quality Assurance: Yes []

No eawl Physics: Yes El No El Unit: Exp. Rate: Procedure #: Rev. #:

Op. Tmp: Pess:Mat~e als & E xiiimelat Design Temp: Design Press: Sealant Type: Lot #*

Equip. ID #: Sealant Type: Lot #:

Commets/Docripton ofJob- abrication -Dalh Mill Tests: Yes [] Nor- Sec. IlJ: No-]

Has the Customer expressed mechanical integrit nces for the component? Yes [] No []

Plant Rep. (print): Initials: Design Calcs.: Yes E], No El Engr. Dwgs.: Yes o E]

TISI Tech. (print): Initials: Positive Material Identification Yes'[ No[]

TIMOM ize.'--* Leak Severity: Noti tions: Quality Safety atAssurance:

Related: Yes .Yes

[l No []

[]

Line S : Service: lth Physics: Yes [J .No E]

Exp. Rate: Procedure \I Rev. ff:

Op. Temp: Op. Press: Marials & Eauilmnent Design Temp: -pesign Press: Sealant Type: Lot #:

Equip. ID #f: Sealant TypeaLot Lot Croptimmnnoft sobe:M CoFaenisetionion sc ilTes ts:Yes0 No F1 sSee HN.e 0% No[No Has the Customer expressed mechanical integrity ol*,s for the component? !YosD[ Non[

• TISI Tech. (print): Initials: rl'Bmtiw Material Identification: "-es*lN L All FPstobStepy must be initialedby the Team Industrial Services,Inc. technician who is leadingactivities at the work site.

2. Post-Jobmust bSsigned by customer representativew

- . - ,- :*.W

,-,-= . ** * ,- .- -*4.= *%*,*

.* .-* 7:, -'  % ... . I.T . **.' Ik'j

• Z. w,-g,-..

Complete sign-offofjob package. . . ,

Complete sign-offof Team's procedum 0 1-Complete Work Recor&ids amve 0wner sign. . 0 Plant Representative (print): / . SSignature:

Additional Comments:

Designer: JMQIHHIMCW Date: 05101112 TEAM Industrial Sheet I of 10 Checker: FAIMCWJAC Date: 511112 ServicesOrder: .

Cover Thickness

Reference:

ASME Boiler and Pressure Vessel Code,Section II, Part D, 2010 Edition Data:

Design Pressure P := 600-psi Inside Radius R:= 1.125-in Allowable Stress S:= 18300-psi Joint Efficiency E:= 1.0 Supplied Thickness Ts:= 0.625-in Analysis:

P.R (E-S) - (0.6-P)

T = 0.03763-in Registration # F-003143

I Designer: JMQIHHIMCW Date: 05101112 TEAM Industrial Sheet 2 of 10 Checker: FAIMCW/AC Date: 511112 Order: 91672EM Rev. I Split Circular Endplate Analysis

References:

ASME Boiler and Pressure Vessel Code,Section II, Part D, (Table for Maximum Allowable Stresses, 2010 Edition Formulas for Stress and Strain by Roark and Young, Fifth Edition, Table 24,' Case 31 51mplq 5upported Edge F- Eg Data:

Design Pressure P:= 600-psi Maximum Allowable Stress Sallow:= 18300-psi Design Temperature 2 64 T:= 400.deg Modulus of Elasticity E:= 00000-psi Split Endplate OD OD:= 3.50-in Poisson's Ratio v:= 0.31 Cover Wall Thickness twall:= 0.625-in Maximum Allowable Deflection ymax:= 0.05-in Split Endplate Thickness tendpl:= 0.75-in External Corrosion Allowance ExtCA:= 0-in Opening Hole Diameter (consv) HD:= 0-in Internal Corrosion Allowance IntCA:= 0-in OD:= OD ExtCA OD = 3.5-in twall:= twall - ExtCA - IntCA wall = 0.625-in tendpl:= tendpl - ExtCA IntCA tendpl = 0.75.in Solving for Modulus of Rigidity E+

G:= 2-(1 + v) G= 10076335.878-psi Solving for variables HD a--

OD twall 2 a- a= 1.125-in C:= c = 0.563-in 2 2 b =a - c b = 0.563-in Registration # F-003143

Designer:. JMQIHHIMCW Date: 05101112 TEAM Industrial Sheet 3 of 10 Checker:. FAIMCWlAC Date: 5/1112 SeicesOrder: 91672EM Rev.

Solving for Constants b-)-1.58614. b-c + 2.85046- 3.1277.b+

0.238- c) b+ c) 2.88 (b__ c K= 2.485 l2-b( .625-tendpl>G

• G.+4.lI + = 2.358 4 2-b E 1+ 1 2 24 = 2.319 1 :=4- 1 "..(1 +t4 Xb = 3.562 1 - 2-c )L-E + C.6

-y,.b_ -*J2+ b 12-,)Jt. -, --I -

-I.

+ 2 X= 0.315 IT2"( - "12 + X " - "_(a11 2"W 1

c 1 := c 1 = 0.017 b ~1

- >x2- 1)-(cos l{'jJ7 1Ic 2 = 0.456 (C * " 1 2cs Stress at A (maximum)

-t._6"P--cb.(b- -1 ,Y12..S + C2T1 -,,22..S) + S

.b tendpl2 (

at = 4343.238-psi < Sallow= 18300-psi Deflection at B (maximum)

Y 24-P-c 2 -b2 l-esh + c 2 "cOsh(27 +

E'tendpl y = 0-in < Ymax = 0.05-in Registration # F-003143

Designer: JMQIHH/MCW Date: 05101112 TEAM Industrial Sheet 4 of 10 Checker: FAIMCWIAC Date: 511112Services Order: 91672EM Rev. I LineEosure Analsi

Purpose:

This analysis will calculate the internal stresses and bolt load of a line enclosure.

References:

ASME Boiler and Pressure Vessel Code,Section II, Part D, (Table for Maximum Allowable Stresses),

2010 Edition Team Industrial Services, Teco Manufacturing, Engineering Department, ISO-900o1 Quality Manual, EP8.7 R

I DImensIory Free Body Diagram Data:

Design Pressure P:= 600-psi Length Between Centerline of Seals LS:= 2.00-in Design Temperature T:= 400.deg Sidebar Length (at Centerline) LB:= 2.00-in Inside Radius R:= 1.125-in # of Studs per Half n:= 1 Cover Thickness t:= 0.625-in Hole Size h:= 0.75.in Cavity to Stud CL A:= 0.625-in Stud Tensile Area TA:= 0.226-in2 End of Sidebar to Stud CL B := 0.625-in Stud Allowable Stress Ss:= 13800-psi Sidebar Thickness ts:= 1.75-in Enclosure Allowable Stress Sallow:= 18300-psi External Corrosion Allowance ECA := 0-in Internal Corrosion Allowance ICA:= 0-in R:= R + ICA R = 1.125-in B:= B - ECA B = 0.625-in t t - ECA - ICA t = 0.625-in LB:= LB ECA LB = 2-in A:= A- ICA A = 0.625-in ts:= ts - ECA ts = 1.75-in Registration # F-003143

Designer: JMQIHH/MCW Date: 05101112 TEAM Industrial Sheet 5 of 10 Checker: FcMCWeAC Date: 5/1/12 Order: 91672EM Rev. I Analysis:

Solving for forces and moments F:= P-R-LS Fx:= F Fy:= F F = 1350-lbf Fx = 1350.1bf F = 1350.lbf Setting forces in x direction equal to 0 R2 := Fx R2 = 1350-lbf Setting moments around centerpoint of cavity equal to 0 t

2 BL:= F. BL = 2025-lbf BL := if(BL < F, F, BL) BL = 2025-lbf B

Allowable Bolt Load BLa:= TA-Ss-n BLa = 3118.8.1bf Stresses in Shell (thin walled enclosure) o-:= P--R o-= 1080-Dsi t

Sidebar Stress (at Bolt Centedine)

R 1 :=BL-F R 1 = 675-Ibf ts 2

'b2:= 1 3 ob2 = 661.224-psi

-- (LB - n-h)-ts 12 3 F TS 2 (-=

(LB - n.h).ts Ts = 925.714-psi Results:

Less Than Bolt Load Allowable BL - 2025-lbf BLa = 3118.8.lbf Stresses in Shell (thin walled enclosure) oa= 1080-psi Sallow = 18300-psi Sidebar Stresses (@ bolt centerline)

G-b2 = 661.22-psi Sallow= 18300-psi Shear Stresses in Sidebar "s = 925.71-psi 0.8-Sallow = 14640-psi Registration # F-003143

Designer: JMQIHHIMCW Date: 05101112 TEAM Industrial Sheet 6 of 10 Checker: FAIMCWIAC Date: 511112 Order: 91672EM Rev. 1 Services Thrust Calculation Due to Une~qual Bores & Separation Larger Diameter D:= 2.00-in Smaller Diameter d:= 1.05-in Desiqn Pressure P:= 600-psi Number of Set screws N:=4 Size of Set Screws 5/16-18-UNC (Cup Tip)

Holding Power of Set Screws H:= 728-1bf '

Limit Injection Pressure Pi := 1000.psi Thrust Produced T: (D 2 _ d 2).~.Epi+, (d 2).iE.p T=' 2795.232-1bf 4 4 Number of Set Screws Required T

ND: ND = 3.84 H

Thrust per Set Screws F

F= 698.81-1bf N

Thrust Calculation Due to Unequal Bores & Separation Larger Diameter D := 2.625-in Smaller Diameter d:= 1.05-in Limit Inlection Pressure Pi:= 1000-psi Deslan Pressure Pd:= 600-psi Number of Studs N:= 2 Size of Studs 5/8 x 11 UNC Stud Tensile Area TA:= 0.226-i'n2 Stud Allowable Stress SS := 13800-psi Thrust Produced T:= (D2 - d2).W-4.Pi + (d2).41.Pd T = 5065.524-lbf Number of Studs Required T

TA-SS ND= 1.624 Force per Stud T

F - F = 2532.762-lbf N

Registration # F-603143

Designer: JMQIHHIMCW Date: 05101/12 TEAM Industrial Sheet 7 of 10 Checker. FAIMCWIAC Date: 511112 Services Order: 91672EM Rev. I Strap Thickness:

REFERENCE:

Practical Stress Analysis in Engineering Design, Alexander Blake, 2nd Edition, pages 29-30 F

[B7 t.

DATA:

Number of Bolts N:= 1 Thrust per Bolt F:= 2532.762-lbf Ear Width B:= 1.25-in Maximum Allowable Stress S.:= 18300.psi Joint Efficiency E:= I Strap'lD ID:= 10.75-in Strap OD OD:= 12.25-in Bolt Circle BC:= 12.00-in Bar thickness(fumished) t " OD - ID t= 0.75-in 2

Moment Arm BC- OD+ ID X 0.25-in 2

Thickness Equation derivation:

Where, P:= N.F M:=P-X  ; C = t2  ; 1.= (1/12)B(t..qd )3 S =,(M*C)/I 9=(P*X*(t/2))I(1/12)*B*t 3 Bar thickness(required): Stress on Ear 6.M 6.M treqtd :- B-----E Stress 6-m t2.B.E treqd = 0.41-in Stress = 5403.23 .psi < S = 18300.psi Registration # F-003143

IDesigner: JMQIHHIMCW Date: 05101112 TEAM Industrial Sheet 8 of 10I Checker: FA/MCWIAC Date: 511112 ServicesOrder: 91672EM Rev.

Thrust and Bending Calculation (bending of strongback ear of hub clamp)

Thrust Produced T:= 5065.524-lbf Moment Arm X:= 5.125-in Total Width B:= 3.50-in Number of Studs N:= 2 Cavity b := 2.25-in Allowable Stress S:= 18300-psi Thickness Provided tp:= 2.00-in Joint Efficiency E:= 1.0 Force per Half T

Fs Fs = 2532.762.lbf N

Thickness Required S6-Fs-X tr = 1.845-in (B - b)-E.S Thrust and Bending Calculation (bending of ear on strongback ring)

Thrust Produced T:= 5065.524.lbf Moment Arm X:= 1.12-in Total Width B:= 1.25-in Number of Studs N:= 2 Cavity b:= 0-in Allowable Stress S:= 18300-psi Thickness Provided tp:= 1.50-in Joint Efficiency E:= 1.0 Force per Half T

Fs = 2532.762.lbf N

Thickness Required 6-Fs-X tr = 0.863-in

't:=4(B - b)-E-S Registration # F-003143

Designer: JMQIHHIMCW Date: 05101112 TEAM Industrial Sheet 9 of 10 Checker: FAIMCWIAC Date: 5/1112 T TheDescinAndBehvio Refeenc: A Inrodctin Services OfThe oltd Jintby ickord Order. 91672EM Rev. 1I Torque Analysis

Reference:

An Introduction To The Desigqn And Behavior Of The Bolted Joint by Bickford, Second Edition, Page 133.

Data:

Stud Tensile Area TA := 0.226.in2 Stud Allowable Stress SS := 13800.psi Allowable Strength of Stud Fp := TA-SS Fp 3118.8-1bf Torque Application Factor A:= 1.0 Pitch of Threads P := 0.0909.in Coefficient of Friction Nut/Stud pIt := .15 Effective Contact Radius of Threads rt:= 0.2822.in Half Angle of Threads 03:= 30.deg Coefficient of Friction Nut/Joint pIn:= .15 Effective Contact Radius Nut/Joint m:= 0.4219.in Analysis:

T:= Fp-A- + os(p)

(Ar t ,.M)P

+t' T = 32.91-ft-lbf OR T = 394.936-in-lbf Registration # F-003143

Designer. JMQHHi/MCW Date: 05101112 TEAM Industrial Sheet 10 of 10 Checker: FAIMCWIAC Date: 511/12 Te rial Order: 91672EM Rev. I Services Weight and Void Void RunCavity:= (2.25-in) 7r-(1.25-in) RunCavity = 4.97-rn3 A

2 w -3 BranchCavity:= (0-in) --- (0-in) BranchCavity = 0.-in 4

Cavity:= RunCavity + BranchCavity Cavity = 4.97-in3 27, Linel := (1.05-in). --. (1.25.in) Linel = 1.082-in3 A

2 71 3 Line2:= (0-in) .-. (0-in) Line2 = 0-in 4

3 Void: Cavity - Linel - Line2 Void= 3.888-in InjVlv:= 2-0.18-in3 Void:= Void + InjVlv Void = 4.248.in3 BC Weight Clamp:= 9.101b.2 Clamp = 18.2-1b SB1 := 5.031b.2 SBI = 10.06-lb SB2:= 4.561b-2 SB2 = 9.12-lb Plate:= Clamp + SB1 + SB2 Plate = 37.38-lb lb' StudsNutsl:= (0.11-lb-8) + (8-in-4-0.087- .) StudsNutsl = 3.664-1b StudsNuts2:= (0.1l-lb-8) + (10-in-4-0.087.1- StudsNuts2 = 4.36-lb StudsNuts3:= (0.11-lb-2) + (6-in.2.0.087- .lb StudsNuts3 = 1.264-lb k mn) lb Sealant:= Void. 1.35-0.043.-- Sealant = 0.247-lb

.3 In InjectionValves:= 0.501b-2 InjectionValves = 1-lb Weight := Plate + StudsNutsl + StudsNuts2 + StudsNuts3 + Sealant + InjectionValves Weight= 47.915-lb Registration # F-003143

TM 2012-12 Attachment 1, Page 1 of 4 Evaluation of 0.25" Circumferential & Axial Flaw on 314" RHR-600 Piping.'KPS:

Prepared By: Z Date Aifr: X'i2_

Reviewed Byl " Q W -A-~ Dt:' 4 5 -0 4

Area Reinforcement Evaluation for 3/4" R IR pipe at valve RHR-600 with 0.25" hole, KPS:

Pipe Minimum required wall thickness for design Pressure:

Design Pressure p:= 600 psi Outside Diameter D6:= 1.05 in' Material A 312 TP 304 Allowable Stress at 400 deg F S,:=14900 psi (for A312 TP304 per B31.7, 1969)

Coefficient y:= 0.4 p.Do

=2.(s + p.y) tm = o.02 in Area reinforcement evaluation:

Area required = Ar in2 Area Available = A. in2 diameter (hole)= d := 0.25 in ot:= 90-dog Minimum required thickness of run pipe t, := tm 0.021 Iin Thickness of the run pipe Tr:= 0.113 in iln2 Ar := [d-t1,(2 - sincx))]-1.07 A4= 5.566.107 ka:= d-(T - tr) Aa 4 ).023 jn2 Aa>Ar OK

Reference:

ANSI B31.7 2004 Section Xl. Appendix-C, Para C-5322:

Thickness of pipe t:= 0.113 in Depth of flaw a:= 0.113 in N513UMITO.75RHR_KPS.xmcd

TM 2012-12 Attachment 1, Page 2 of.4 314" Diameter pipe for RHR-600 valve:

Material: ASTM A 312 TP 316L, A312 TP 304, A182-F316L, A182-F304 Yield Stress Sy:= 25000 psi Ultimate Stress Su:= 65000 psi Lower bound properties are used for all four materials Flow Stress of := 0.5.(Sy + Su) Psi o-f = 4.5 x le psi Outside diameter of pipe = D := 1.05 in Pipe wall thickness = t:=0.113 in Mean radius of pipe= R:= - in 2

R = 0.469 in Half Crack length = 0.125 in Half Crack angle 0:= R -C 0 = 0.267 y:= asin (0.5).(.-}sin()] yp = 0.132 Safety factor for membrane stress for Service Level A SFm := 2.7 membrane stress at incipient plastic collapse = cre rm:= o*t- -(t.-~.)2--t] crc, =3.739x 164 psi The allowable membrane stress for a circumferentially flawed pipe =St

-m St= 1.385 x 101 psi S~m Outside diameter of pipe D:= 1.05 in Pipe wall thickness t := 0.113 In D 0 -t Mean radius of pipe = R := D in R = 0.469 in 2

Inside diameter of pipe = d:= Do - 2.t in d = 0.824 In Design pressure = p:= 600 psi p~d2 Primary membrane Stress = P.:= pd* si Do2 2 d2 Pm= 961. 895 psi < St OK N513L[MrT0.75RHRKPS.xmcd TM 2012-12 Attachment 1, Page 3 of 4 Available safety factor for membrane stress = = 38.872 > SF. = 2.7 OK 2004 Section Xl, Division I, Append ix-C,Through wall Axial Flaw, Para C-5410:

2.

Rj Inspected length of Flaw = 1 0.25 in Allowable length of axial flaw Ian in Do:= 1.05 Hoop Stress O -p.D R: t 2-t 2 22 1*1.58-12

=

1,U:= 1.584WR- al - = 5.857 > 0.25 OK 2004 Section XI, Division I, Appendix-C, Para C-6322:

This evaluation is performed assuming the flaw location being affected by weld Z factor for weld location for austenitic steel [Shield Metal Arc Welding (SMAW) is conservatively used]

Nominal Pipe Size (NPS) NPS:= 0.75 in Z:= 1.30.[l + .010-(NPS - 4)] Z 1.258 The allowable membrane stress for a circumferentially flawed pipe =St O'cm4 st := ' St =1.101 x 10 psi z SFý'

Primary membrane Stress Pm= 961.895 psi < St = 1.101 X 104 psi OK Available safety factor for membrane stress -!c-m- = 30.906 > SFm = 2.7 OK Z-Pm N5I3UMITO.75RHR_KPS.xmcd

TM 2012-12 Attachment 1, Page 4 of 4

==

Conclusion:==

The flaws in circumferential as well as axial direction are stable and the 3/4" RHR piping will maintain its structural integrity with the flaws.

N513UMITO.75RHRKPS.xmca 1

TEAM INDUSTRIAL SERVICES, INC.

TECO MANUFACTURING, INC.

Engiering Change Orde Item#/SWO# 191672EM Requested Change change sa 194 gr 8 to astm a-194 gr.8 ECO# 1250 NCR# 0 Letter Category iNudear-Safety Related Effect On Structural Integrity of Clamp Reason for Change IMaterial Availability N/A

[ QC Final Inspect]on Required Changed by IMCW Change date 15/2/2012 Changes Made Checked by ISLG Al. changed sa 194 gr8 to astm a-194 gr.8 on material sheet and on cover page.

Checked date 15/2/12

• Approved by IRD

'Approval date 52/2012 Manufacturing Received By Date W QC Received By Date

[] Stock Item

Document A Impact Summary 0,O.

Ofminion Instructions

This checklist summarizes document and database impacts. The procedures, documents, software programs, and databases, which must be reviewed, are already listed on the Document Impact Summary form. However, the preparer may list (In the spaces provided) any other documents or other information, which r'equire revising as a result of the activity.

Station Unit Change Dopument Number OKPS O]MPS []NAPS EJSPs 01 ['2 03 0 ISFSI TMOD-2012-12, Rev. 0

1. Procedures 7 N/A Item Chionas IRoquird?

a. Administrative Procedures l Yes [No

b. Operations Procedures "Yes ONo

., Maintenance Procedures OYes 0 No

d. Engineering Procedures (including Rx Engineering) El Yes 0 No

e. Radiologicsal Protection Procedures QYes No

f. $ecurity Procedures .Yes .No

.g. Emergency Procedures 0JYes ONo

h. 5.5.b ProcedEures/Guidelines E] Yes ONo

i. Severe Accident Management (SAM) Procedures/Guidelines . . . Yet .. No J. Other Procedures (new Vendor procedure for leak seal injection) 0 Yes E0 No

2. Drawings _ .__

KOem Change Required?

9. Station Drawings [] Yes 10 No

3. Design Calculations I Analysis _E N/A Item Change Required?

a. Electrical Calculations .Yes No,

b. Mechanical Calculations (new vendor calc) 19 Yes 0 No

c. I&C Scaling and $etpoint Documents / Calculations [ Yes No d, Computers I Software Design I Firmwre . Yes y N

e. Class I Piping Stress Report 0 Yes No

f. Non-Class I Piping Stress Analysis 0 Yes ONo g, Civil Engineering Building / Structural Analysis Qyes No

h. Pipe Suppor-t Calculations .Yes No.

I. HVAC Calculations yes No

j. Cable ond Raceway .. Yes ....... _No

k. Nuclear Safety Analysis El Yes ONo I. Technical Reports (eg., WestTnghouse Scaling Document, etc.) C[ Yes _ No

m. Other ._ ___Y_ ___N_

F-)N-3-'Mpo (POOK"O

Document Impact Summary 3 ,Ao . *. .- "J DominionW Station Unit Change Document Number RKPI5 []MPS []NAPS QISPS-M 0 [J3I2 [I ISPMI 2i TMVIOD-12-ig, Rtev. 0

4. Other IDocumenta 2 /A

a. Updated Finll Safety Anaysis Report (UFSAR)/Update Safety Analysis Report (U-SAR)/ISAR/ISSFi-SAR ___ Ye______No

b. System and Plant Design Basis Documents (DBD) _ _ _Yes _ __No C. Specifications [3Yes ONo

d. Technical Requirements Manual(s) (TRM) , Yee . No

e. Severe Accident Mitigation Guidelines (SAMG) Nuclear Safety Yes No Analysis [PM 4.1,11 for VA Plants ___ Yes____No f, Vendor Technical Manuals Yes N g., Technical Specifications El Yes No

h. Equipment Data System (F-DS)/Q-List/EMPACIMEL [2 Yos No.

Bill of Material(s) (BOM) / Spare Parts 13 Yes No.

j, Emergency Preparedness Program [: Yes No S. Key. Regulatory Programe NIA Ite- chan ROEred?

a. FiFlv *01etotlon IAppendix R"R Program~

SFire .afe-8hutdewnAnalysis (,., APPonolix "R" Repae, up Branch Toehnical Position 9.5-1.ahutd~wn analysli)

FFlie Proteetlen Program Document Soebuatible L ading AnalyNis Firefighting l otrataglea or Proe4ire Plans 9 Other Program Doeuments, such as Exemption Requests or PP Engineering Evvaluations

b. EQ Pgrorm, EDYes'N

c. License Renewal (Aging Management) Program . . Yes El No

d. Station Blackout Prograrm -. - Yes El No

e. Regulatory Guide 1.97 DYes .. No

f. Generic Letter 89-13 ElYou JN

. Roadiologioal Protection Program (ALARA) DYes iE No

h. Maintenance Rule Program [Yes .No

1. Appendix "JT Program Ql Yes GNo

j. Insetvice Inspection (1I1) []Yes -No,

k. Inservice Testing (1ST) Program(s) .EYes . Nei I. Environmental (Non-Radiological) .[Yes [No Im.GSI-101 Program Q Yes - No

n. Nuclear Material Control No

e. North American Electric: reliabiiity Council (NERO) I E- Ye0 M. tj' Pw9i

Document Impact Summary 5-P-Domi nion DNESAAkN-1 ATTCM. Page Station Unit Change Document Number NKPS [jMPS E]NAPS []SP$ 01 [12 []3 [] ISFSI TMOD-2012-12, Rev. 0

6. Other Process Documents and Databases /NA KOMI____________________________ Charnge Raquired?

a. ERDS Dtapoint Liobrary El Yes No

b. Preventive Maintenance Program (PM) L-Yes iNo' o, Oeatinge Pfrgram 1 Ye. . . No

d. iow-A.olermted Crfosion (FAC) Program- . YEJ . No

e. Margin.n aemrnt Program Vag_ _ No

f. Equipment Reliability Program .. Y. . No

g. InContainment Banned / Restricted Materials Program 0_Yes Q_ No

h. Valve Peckinig Progirai 12 Yes 1 No

i. Maintenance Check Valve Progrom 13 Yee No

j. :.Maintenance Safet*tRelief Valve Program 1. Yes [I No

k. Snubber Program L Yes 0 No

1. Human -System Interface Program 0 Yes EINo

.m. Seismic Qualification Program 0] yNO

n. Po5t Maintenance Testing . .. e.s NNO

o. Motor Operated Valve (MQV) Program 0 Ye No

p. Air Operated Valve (AOV) Program . Yes8 N .o

q. Reactivity Management Program El Yes [2 No

r. ODOM / REMP / REM 0_y_____ No

s. Single Point Vulnerability Database C] Yes No

t. ASME Section VIII Vessel Program Database 0 Yes 01No

u. Containment Hydrogen Generation LvsL~

v, Containment Debris Inventmr Noe

w. Probabillll Risk Af-esment - PRA

x. FSI - 10 CPR 72.212 Evaluatlon Repor-forSpeatFul Ye o

y. Critical Equipment List yov No

z. Sulpply Chain Warehouse Stocking Database (SAP) CYes N aa, Underground Piping and Tank Integrity Program E[ Yes i No Item Change Requlid?

a. Nuclear Control Room Operator Development Program (NCRODP) Training Mo0ukNes

b. Bimulatter Chvges QYes

c. Other Training (eg. Maintenance, NSS, ESP, eto.) . Yes . No

d. Training Dept. Training Impact Report (Irin accordance with N

___________________ Ye No~

Document Impact Summary I ,, , A: . ... A.,. -

PDominion Station Unit Change Document Number f@KP6 [OMPS E.JNAPS OSIPS 01 [1j2 Ej3 LII 18F8I TMD2612.12, Fkav. 0 B. Kewaunee Only I 2 NIA

a. Concrete Rebar Cut Request .1 EiYea ~ NO

15. Fuse Control Prooram EjYes rx@No c, Motor Thermal Overload Heaters I EiYes ________

d. Systemn Decriptions E21 Yea No____

e. In proved Technical Specifications :2 jNo miMiItalle Onl~y ______ NIA, Item ch~ange Rouired?

0. Safety Functionel Requiremonts Manuml - Yes JNo t.b REMODGM ElY@4. DNo Item equied~?

RIai

a. North Anna Setpoint Document Qyes a

b. Precautions, Limitations, and Setpolnt (PLS) Documnent QYes g o

a. Nofth Anna VP88L ciYet, No

d. Luba Oil M~anual YaN Item Change RrnnjiradI7

a. Luba Oil Manual ye N 12., Remar'ks (Attach addition~al pages if Meade'd):

FOFM W. 72MM (raemIt C10 1

Design Effects Table Dominion [DNEAL MGN S0 This table will identify Impacted programs. If a question is answered yes, the responsible engineer shall address and document in the discussion section of the Design Change. The respQnsible engineer shall check 'No Impact" or "Impact" as appropriate. If necessary, consult the program owner to assist with the determination. If it is determined that there is an impact, Identify program owner, check "Impact" in the applicable section, obtain the consulted Individual(s) signature on the engineering product cover sheet and document the dlscussion in the change package. If an impact Is determined for any program, then relevant portions of this attachment should be attached to the appropriate document. If it is determined that there Is "no impact" in a section where a question is answered "Yes," document the basis for this determination In the change package. When all questions are answered "No" in a particular section, do not check "Impact" or "No Impact!

Station Unit Change Document Number Vj KIPS MPS 1 2 MOD-0112.A2. Rev. 8 C3 WAPS C3 PS ~ ersi.

l1 89cod IlRH-MOO Leok Repair

,1.0 Filre Protection I Appendix R 1.1 Flre@ Protection P-qulpMent or Features tJ Yes In No Program Qwriar Name:

0 No Impact EJ mpace The purpose of this section Is to provide guidelines for the review of modifications/activities to ensure that the design conforms to the requirements set forth by the fire protection program documents, in each respective station's UFSAR, and to ensure that all modifications to the stations do not lebsen the degree of fire protection at the stations.

Poes the ctivity modify, acd, or remove fire protection equipment Including:

fire barriers, doors, or penetrations, end partial barriers credited Inthe fire hazards analysis

• fire dampers, doors, penetrations, hatches, cable tray fire stops, cable tray coverS, or other elem'ents Installed In fire boriere non- Appendix R emergency lighting (e.g., exit lighting)

• fire resistant coatings Includifig structural steel fireproofing 6nd electrical raceway Wrap

fire suppression systems, Including sprinklers, 002, and halen systems

. fire detectlph equipment, Including smoke, heat, and flame detectors

• fire protction system Interface devices such as HVAC shutdown trips, supervisory air supplies, and backup batteries fire fighting equipment or systems Including hQse stations and portable fire extinguishers

• supports and restraints for Fire Protection systems and equipment firo suppression water supply, Including fire pumps &valves Does the activity modify any plant structure (including floors, doors, roofs, ceilings, drains, .urbs, dampers, penetrations, hatches, equipment knockouts, stairwell, HVAC systems, pipe chases, elevator .hafts, load bearing structural steel, etc.) suoh that Itmay block or otherwise interfere with the operation of any fire protection equipment?

Does the activity modify the occupancy or function of a room or structure such that it may affect any fire protection equipment?

Does the activity modify the nearby environmental conditions (room amblent temperature, nearby heai sourees, etc.) that may affeat apy fire Signature required on Engineering Product Cover Sheet.

F'orm No, 73117R (Pop 21qi)

Design Effects Table ILA I MUM.

~jNo Inipaet t~impact, D~oes the activity odd, modify, relocate, or remove combustible material within a flre zone, or relocate combusibe131 Materia betweNH firO 201J881 Combustibles may ind-ude:

Ordinary Combustibles (e.g., wood, paper, carpet)

• Combusttble Liquido

" Grease, charcoal, and combustible insulation

" Plastic Materials (especially halogenated plastics sueh as PVC or Neoprene)

" Cables / Cable Tray Loading (Consult local procedures for tray fill criteria.)

" Cables or other materials which give off corrosive gasses when burned

" Coatlngs (e.g., wall and floor coverings)

" Flammable gases 1,3 Hazards and Ignition Sources Yes No Program Owner Name:

12No Impact 0 Impact*

Does the activity create, alter, or remove any hazards or ignition source, such as:

S.lydrogen or other explosive gasoes '

• • *owihustibie meItils {pia~neiu , *rnnium, eta,)

Nowheat seur-*eQ that eeuld Ignito qombustible materali*

-1.4. Flere Safe-Shutdown Analysis Yesp No Program Owner Name:

[3iomal)

No Impaet IH~t Doe.s the activity Involye any of the following relative to the Appendix R.safe shutdown analysis (refer.to local proeadure/ site speciflo AppendiX

.- Credited Appendix R Equipment and flowpath* as identified in equipment databases (e.g., EDM, MEL, lAD) or as represented on P&l1a and Inope'atirir prerdures.

• Cables; electrical schemes, and raceway associated with Appendix R credited equipment

• Radiant energy ohfelds

- RCP lube oil collection systems Does the activity involve access to a fire zone/area, fire protection equipment or Appendix R credited safe shutdown equipment to perform manual operator aq.tions er manual flt fighting aetivities?

Does the activity affect Time Critical operator or fire brigade response time?

Does the activity add, modify, relocate, remove, or obstruct any emergency lighting on the manual action ppth required for compliance with Appendix R?

Does the'activity add, remove, or affect the performance of any plant communications system relied upon for fire fighting or safe plant Dcss the aowyity inipeat any A4ppndix R ogemption requests.

slqnaturv required oil PEnineefing Produot Covet; she§t.

Design Effects Table

• C . . . .

D*omini*on 2.0 nvlral.*effntal Qualiieatiea (RO)

Yes No Pregiram Owner Name:

0 N9 Impact The following questions provide guidance to ensure that the methodology used for compliance with 10 CFR 50.49, 'Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants" will not be adversely impacted. The purpose of this review is to determine If the proposed design change could affect EQ Master List equipment, EQ files, EQ environmental conditions, or EQ/HELB boundaries.

Does the activity modify any EQ component as identified In (or to be added/removed to) the equipment database(s)?

Does the activity result in exceeding nonral operating environmental conditions (I.e., temperature, presure, humidity, radiation* as Identified In Does the activity affect or create new aeoldent (HELS or other pipe breaks) environmental conditions (1,e., temperature, prossure, humidity, radlaton, aprmy oompealtlon, or sum "menoe) of the equipment opevaotig Vme?

Does the activity add/remove any electrical system or portion of an electrical system credited for accident mitigation where the equipment Is located Ina hahrs anviranment?

Does the activity alter the electrical portion or accuracy of any accident mitigating or monitoring system (Including cable and interfaces) located inran.ara where the On Armei-ent !0 oftated by an accident?

Doesthe activity alter the physical arrangement or HELB boundary including doors, hatches, duct work, plplng, or electril penetrations, and structural walis aeidi that it could ýftct r.lstiong EQ equipment or add-new equipment to tha.l2QProgram?

Does the activity-modify the operating conditbons-of EQ equipment (energl;ýed or de-energized, duration and/or component heat rise conditIon) duNrng wemrsl or acidant covoting tOnes?

3.0 ASME Codes / ISI I IST ASME Section XA and/or OM Code provides the requirements for preservicaeand inservioe testihg end Oxamlnation of In scope piping, components, component supports, pumps, valves, pressure relief devices and dynamic restraints to assess their operatienal readiness; 3.1 Inservlce Inspection Em Yes Cl No Program Owiner Nano: Phil Oukes Signature requlred on Engineering lProduot Cover Sheet.

Fwm Mi. 731172 (DQo "10)

Design Effects Table E$ 2

'W fDminion AAGN03 ATTCMN Pg.4 of11 NOt'E: The 131 Olasihlea.tica 3ounda.y tDswlrige identify the euw-ent A8M* Cede equipment.

1. Does the activity involve $$CS oddressed by the AGME t13Prgrams ard require. change to oneer more of these eogFamsr SSCsoaddred by the A*MA 181 program ins luda any of the follwi[ng:

• Clesa , 2, or 3 piping, coponents. or supporns

• Class MC or CC - containment liner, concrete, and post tensioning systems (no CC ,t Kewaunee)

• High $afety Significant Components (H$S - Identified on Sully Unit I CBM drawings- with an HI')

• ASME Code Inservice Inspection (131)

• Augmented 181 Program

• Constructed to a Nuclear Code, eg., ASME III, ANSI 131.7, or ASME-Draft Pump and Valve NOTE., A change to the ISI Program Is defined as a revision to one of the following documents: 18! Plan or Schedule, System Pressure Test Program, Auigmented Prmgram, or ISl drawings as result of the activity, Examploo of ectivitles Involving changes to $SCssuch ao the following may require a change in one of the above documents-

.Changing the diameter of pressure boundary bolting

, Changing the type of pipe support I.e., vertical/lateral to anehor, spring can to dgrd strut, otp.

, Adding or removing pipe supports

• Adding, removing, or changing line numbers that are depicted on 181 documents

- Adding, removing, or changing the name of welda on lines that are depicted on 181 drawings

2. Will the activity result In a permanent or temporary non-physical change to an ASME XI Class 1,2, or 3, MO, or C0 pressure boundary item, cere Quppe)1, er coffplooent support item or piping system dens by changing it* dssloin rtinq ( i*fital to. or stomal pra6ure, temptrartue, or lood)?

S. Will the activty Fasuit In~Ra l ~~ysc alcmn s,raMogA9"ddIag, and W40viiM ftomr,4apw ytams (InasludingR§walf4pg1 braying, ard de r anet**vi1:ieIo Jmoval nnQu", then the program.is ohsl d 'on te 001gn Effets an no uther ae on" is relUired, .

If any Screening Question above Is answered Yt'esY , then the program Is checked "Yes" on the Design Effects and a Program Owner review Is If 181 program checklist Is checked 'Yes" on the Design Effects the following impacts shall be addressed with the program owner.

1. A review of the proposed activity shall be performed to.assure compliance with the applioable vermlon ofASME SectIon Xi (181) by

-'cetaeetng Il.l Ipar~ense.

• Consider requirements for A$ME Xl Code acceptance inspection, base line (pre-service) inspection access, etc, A review of detabases may be necessary to determine the correct design speciflctlon(s) related to the preppsod change. Conside? If an update to these documents Is required as a result of the change paekagq.

2. Consider access requlremento for inspection Including base ltin Inspeotio.n requlrememts far repairs and replaements (i.e.. des-ign egasid ioes)auh aa;

• Is Insulation removal for later ins sotion necessary?

.Is pIeseeioe (baao line) inapection, suoh qas UT M8080a~v?

Is the accessibility adequate for Inspection? Equipment requiring IS1 that Is being added or deleted or imposition Pequlemeanti 4tst are being changed for apeific equipmtilt shall be Iden~tifled.

- - -.....--- ~-. - - --

rwA Ný1.

144 Va (ft*-P)

Design Effects Table Dominion

.3. Consider the c-rnponent* being modified that are ASME Section III, Draft Pump 4nd Valve, or ANSI 831,7 Code items, The Code Design Specification must be evaluated for technical accuracy prior to the change paekage being Ilsued. Any chanrg to an AgME Ii, Draft Pump and Valve, or ANSI B31.7 Code Design Specification shall be reviewed and certified by a Registered Professional Engineer competent In the area of design related to the change being proposed and the applicable nuclear code prior to issuance of the change package. A Registered Profeslonal Engineer must sign the change package cover page and affix their seal for modifications por the previously mentioned codes when the proposed changes are being made prior to updating the spe~ifoatlon. It is preferred that the specificotion be miedified prior to impla mtation of the proposed change,

4. In all oases where the repiaemment part(s) do not meet the design bases for those Installed, a Repo of Reconcillation shall be Wryen In acordanee with ASME Section. X1. The change paokage itself may ilchide the Reoiet of Re"ooclliatin.

5. Conslder the documenttion listed below. These documente must be maintained ai curate and current per ASME ,i. The failowiRq desigin eoumeniapq ghrall 4ao edre d ps a ilcabi:

NOTrE: Domlnion design sptecificatlon may heve procurement speolflcation Wstle and viC- vYera.

, Realsn popef (eev.,

ov reeUssure Potedtlorureport, gWoeml evluatnsee esot.)

ieports of Reconalllatlori (Used to document the ongineering avaluation of soceptability)

R 3.2 lnservipp Testing E]e IN No Programn Owner NarnO.'

!JNo lmouet

1. Does the actiity permanently or temporarily add, remove, repair, replace, mo~dify, affect, or involve a change to acceptance crterld or perulormance requirements for mefhanIcal equipment or permanently Installed plant instrumentation required for In-ery iC Testing (IST)"

NOM- Uxqmple off aetivitles involving changes to SBoo sueh as the fellw*,In May coquiro s ehahnn to the laf ft-O0R..

, Addlng eor remaovig a valve eOPump that ia Inthe IS PCrOgm

,h* glig the fall eae dIrAtloi of sa AOV

• Permaaatiy ramovlmn the late ala of hookk valve PhavIging the hydre-lte C a hll rlatica otf a pump In the 1*T Pregoanm ReR4-i47 (13i09e8..45), The DIM new requlres aR enilneemlng pmomkvm review by the JS pre~olam eseerIinatea Ifthope Isan impact on ASME Code promrams, This review during the initial stages of the design change process will help ensure that IST requimments are incorporated Into the design change, (Q4093)7 NOTE: The IST programs ldentify the current ASME Qode OM equipment, Permanently Installed plant instrumentation is often used when performing AWME Pedo te ts. These instruments are subjeet to the ocourey and seailring requiremento In the ASME rode and compliancewith the requirements must be maintained. SSCs applic.able to the IlT program Include pumps, valves, drives (motorS, actuators) or Instrumentation requiried for teting (pressure, fley, limit ..wltoheo, eto.),

2. Does the activity involve $SCs identified by the ASME lST Programs or require a change to these programs?

3. Reviq.i of DC by IST Program Coordinatpr if it Impacts A5ME programs. [CM 3. 1.1 for-MP,-8

! Slne.upe Pequired en Preduat cover hieot.

Nlnenltg

Design Effects Table DNSAA-N10 ATTACMN 2* Pae6f1 S'DominioW 4.6 Regulatory Guide 1,0? - Pest Aceidevit Mnitoiag

! Yes NProgram N Owner Nmme:

t~No lmpaot In order to ensure compliance with Regulatory Guide 1.97, a review must be performed to determine the impaet that the activity may have on the Regulatory Guide 1.97 program. Regvlatory Guide 1.97 describes a method acceptable to the NRC for complying with the NRCS regulations to provide instrumentation to monitor plant variables and systems during and folWowing an aecident.

Does the ectivity alter any of the fkileWing charootedefles of Regulatory Guide I .97 compenento Identifled In the eleeFtnle database, oush as,

• Redundancy (Reduction of Categeoy I variables only)

• Pewor Summe (Cotlepo I arid 2 vw=ialao Poly)

SSafty tla"rIfeation

'F e f 1412014Y pf Rowgding laatpumej~t8Ol Tw

• eentol Roem 10 11.17 Idedtflf tion (qotegary I and 2, Type A, B, &0 vaftblas)

SRange r aeourey af Monito1rig Inetrimentatlon Electrical Interface between RG 1.97 loops and the transmission of this signal for other use prior to arid Including an Isolation device E

(Oetopply I and 2 vemobls only)

, Method of Measurement of ; RG 1.97 parameter Locaiton of Maintenonce lIolatton Device (OCteory 1 and 2 variables only) o Accuracy requirements of any Instrument cedited during or following bn accident

,Does the activity affect any Regulatory Guide 1.07 specification (MP$ only)

Does-the activity add, delete, or physleally modify the existing configuration of any analog Inputs from RQ 1i.97 loops to 0 Plant eomputer Sysitam (PTO)M Th[* Aoluds *ny a4we ahOages to the PC&.

Does the aeotvity apprealably inerease er deereasa a yatem parameote (I.e. temperature, Raw, presure, ote.).measurd.by Regulete ..

Guide 1.07 lgtNnekstotlon aveli that the I&C coloulatlono p affected?

5,0 MF1ntenenoe Rulo V04 IM No. pmrop'e ftwn! Now.,

C3 No Impait .. imp e ..

The guldance provided by the following questions alst the reviewer of the aotivlty/modifltatlon In deterwninig If any systefni/cmponenr functions are affected as dOllneated In the. Maintenance Rule 4coping and Performance Crlterla Matrix,. This will ensure that bompliaeno with IOQPFRV.0. [Maintenance Rule] will be maintained, Does tiie aMtIvity alfost the Maintenatiqe Ryis tuflOVIORiof 4 StPuotUr, System%, Or e M-POPIo~t (815) Ik $11Y Of thS fe~ollWia waV69 MAdds a Mlint"plnoe Ruleo *u*oi.o 108a0gea a MOe R"*. p ule fuN0Ia ahanges a mintalf4loAoull

, Adds equipmernt utillsed In 15meorency Opierthng Pro.'cures (EOPe)

• Removes equipment uIlI9ed In SmerONFey Operating Prcedures (10Pa)

• Ohangce the Maintenance Rule Performance Cterion,

• Siniofantly affects unavailabillty post modification (an an ongoing basls)

Significantly affetls rellbility

'$igriotur4 Faquired en Erniniereli iI*peouct e pove Short.

ftw 14a. Im

. (WApq)

Design Effects Table DNSIAGN100 ATAHE-2 ýq:7o Does the artivity affect c.omponont or ,iystem function, which Qould prevent SR SSCs from fulfiliing the SR function?

Does the activlty remvo or install a system or train?

Does the activity affect component or System functions, which could cause a reactor trIp, safety System actuation, or cause a pewer redut1teYn noas the aetivity affect eaparient OF system Aiunt'lans rau*tle*d w4i the atogigs af nualopi fuel in Usiit 4-Does tlie activity transfer ownership of en existing system to another unit?

8.0 Radlologleei Proteetion Program (ALARA) 0E Yes . No jPrograrm Owner Name:

El No Impact El Impaot, The purpose of the following questions is to assist the reviewer In determining the impact of the modification/activity on the respective sites Radiation Protection Plan(s) which sets forth the requirements of the Radiation Protection Prqgram, The ALARA Program(s) which ensures that occupational radiation exposure, both individually and collectively, Is maintained 'As Low As Reasonably Achievable" (ALARA). And, the Offelte DQse Calculation Manual(s) (ODCM) which establish requirements for the. Radioactive Effluent and Radiological Environmental Monitoring Programs,.

Pass the aetlvity pregte a new feodlafiea sepree oro new ,ifiaftlaOr ao@a si*te OF e-ausa( an il0oia 1"d0sa. rates fm aR, O sex1oft OSuMr'c.

(Consider types of welds that could become crud traps, addition of eontominated system piping, change InInsulation on the component, etc,)

Paes the AQtivity create or increase routine maintenance, operation, serviGe, or surveillance requirements in a radleton are@?

Does the activity involve shielding changes, ventilation changes, or materials that contribute to radioactive srud, resin, or sludge treatment Dpes:the activity involve the replacerment of valves, valve Internals, pump Impellers, etc. that could Impact source term (whloh co*tain cobalt alloys) en the pnmiay system?

Does the a.ctivty potentially contaminate systems or components on the plant secondary side?

Does the activity cause an Increase or potential Increase in the amounts of radioactive airborne effluents or liquid effluents or significantly alteo the nuallde mix of oueh effluento?

Does the activity result in a new radlQgctive liquid or gaseovs discharge point or decrease the ability to sample or monitor existing relea*e Paths?

Does the activity affect primary system chemistry controls?

Does the change increase the potontial for the release of radioactivIty to ground water?

Does the activity result in an Increase of the potential for radioactive materials to be released Into inaccessible locations, such as buried plpipg/elaake in conte'to? If tho anawor is "Yes"tharn include Mto drawings and any tempora-ry RCA's uiwd to Mtoem llornnd Mtef-ri.I,

'l910ntiuoa raquiFedon CR nflim2Qwl11 Praduet Ava 01

Design Effects Table

• ~fl *3 A A S ~** .0 7..0 Enlrnrfi~ermtal ivapst (Ne-ladlolgicai)

[* Yes

• No ;Program Owner Name:

[] Noelmpaot [J lmoset*

.Doesthe activty alter w~ater quality characteristios of any liquid discharge? (fiow, chemcical composition, temperature).

Doe the activity Oreete or I~or~oe a source, a? air pollution?

Does the activity alIter any non-red, oioglCaI environmental monitoring system? (meteorologlcal or water quallty m Itolt~ng system)

Does thle aetivity:

• Take an oil pipeline out of service. Refer to statlari's sF'ee Plan for guledaneB on 40 OFR ;I12(d)(*) requirerments.

- Install or modify dams, wells or water treatment systems, sewage treatment plant, or sewage pump station

• Disturb mere then 1/4 atcre of land

• Aidd,, modify, disable, remove, or relocate oil storage tanks or buried petroleum piping (Consider the potenltilI for flooding)

• Change yard g!radlma, sterm* dlrainage system, or yard surface

• Involve the demolition or construction of a building

• Involve the use of mobile or portable storage tanks or vessels

, I~ntroduce a new source of hazardous waste

• QOeS the proposed change involve construction activity near catch basin/storm drain?

0.0 ' Nuolsaw Mater!a Qontrel O] YeS

• Ne Program Owtisr Nsme:

I.O[ NolImpact C[ Impact 4 The purpose Control of the Program asfollowing It appliesquestion is to assistnuclear to each Dominion the reviewer power in determining station and to the the impact the modlflcatlonlaotivlty NuclearofAnalysis and Fuels (NAF) on the NuclearNuclear department. Matedal material Is a collective term for the Nl*C licensed materials: by-product, source, and speoial nuclear material. Nuclear materiel is contained in niuclear fuel assemblies, n~uclear fuel rods, fu,.el assembly ln~erts, irradiated components, certain radiation detectors.and sources, and various Does th, activity result in the procurement or shipmcent of items eontainlng Uranium or Plutonium (specIal nuc;lear matras Ii*h~e SSignature required on IEngineering Product Cover Sheet.

Design Effects Table

'-,2 Dominion . I

0. t.CARisa ReiieW0i RuIP Proafram ead A0149 Mnaogn"1t Activitlage 0l Yes 9 No Prourom ounr Name:

fl No impact tma t In order to ensure compliance with the License Renewal Rule (10 CFR 54), a review must be performed to determine the Impact that te activity may have on the License Renewal Rule (LRR) program, The LRR program credits aging management programs (AMAs and AMPs) for managing aging of plant structures, systems, and components (SSCs). AMAs and AMPs are intended to provide resseroable assurance of the ability of LRR required S$C9 to perform their license renewal Intended functions. [CM 3.1.71 The purpose of this review is to determine If an activity reqt.ires a change to an AMA or AMP credited under the License Renewal Rule.

For example, changing the materials of construction of an SSC In the scope of LRR may create the potential for additional aging effects that were not consider-d or ihat the cumrnt aging management activity is not designed to detect. Removing or adding an SMO creates a potential for LRR documents to become inaccurate. Changes may also result in Internal or external environments, chemistry parameters, or maintenance practices different than those previously assumed for plant $SCs as part of the LRR program evaluation.

- Does the activity affsct a component identified in thp Equipment Data System, the Master Equipment List, or License Renewal Dawingsa as loing In scope ftr liconse renewal?

- Does the activity Involve a component or structure associated with a fire protection system, the SBO diesel generator, EQ cable, or anti**patad traielont vAthout. 4rarn mitigation?

• Does the activity Involve more then a minimal change In plant operating condltions of stress, temperature, radiation, or chemistry for more than 30 days and culd affeat the activitles Identified a@ in-soope far License Renewal inthe firat and aconod bullets?

Does the activity involve the deletion or change in frequency of an inspection or method of discoverlng or evaluating the material conditilon of a plant M80 that t In-scope for licensing renewal eging management?- .

• Does the activity Install a medium-voltage (2 kV-35 kV) power cable?

10.0 Gaeie Laftte (0L) 80.13 P4ragm "__*

fl ou NOe Program OBwneF Name:,

w Impact No Ppe~

Does the aoctivity Impact any host exch.rnger that usos Service Water-for cooling?

Does the activity affect the oheriaml tMetment of the SW oystem?

Does the activity affect the pFessure boundary Integrity of the OeONiCe Water System?

11.0- Statlon Blackout (8W30)

L~Yes H No P1roprom ownw? Name,,

Does the activity alter the design or oppration of einy 81O .qulpment? 880 equipment Is Identified In the MEL/86M Databao,.

Does the activity potentially Impact the operation of any 80 equipment (i.e., non-8B0 equipment potentially affecting 8Be equipment potirntu, Sn r h i on inUeering n~etemperature)?

u SSignatuee required on Engineersing Produot eaver Sheet, Fqtrn No. 7s' ' ý (P90 2W 0)

Design Effe~cts Table t~yes L~No PFSQFS OWPU)P Nahl§' tPaUi M1l10?

Does the aetivity, affect any eentaIinmert baundwi"ea? This Itioludes ahangeii whieh m.-ay ai4fiat arity easntsimnmit Islatlar,Vvae (irwisudlni Ioloin force, -.troko ftie, poelticn indicton, etcj or the leak fightnoss of any containment isolationi boundary, such gig:

~onahinen lsiotonTrip Valvos

~orQ~mgsl 0l40614 Veolvea .eI SMantial Ontianffiaat Ir~afistof VIAIVc 8800oopgy eontalpment (MPS) 13.0 NERO North Arrierloart glettria Reliability Council OJ Yuri t No Program Owmer Nam.e, O No Imopact El IMPW~p Does the a1ctivity affact Vie riot or gross alectrical outut, MW or MVAR, of the main gonerato6?

Does the activity affect the allowable switchyard voltage range?

Does the actiit affect any high transmission fu~nction or control; such as Electrical Bu Protection, Ofibite Power Reliability, orp Tupblne/doenrstor/Exaiter Coreil/Modol/Tranofomar?

Does the activty affect the NERO tOemplianee Commitments desaribee lInthe Nuelear Swshyarcj Intodfase Apraemsnts (NSJAte) loemestdpn..

the Nualelar Wabpeov@ Nuelea? 8-favd ftpolda?

Do~es-ithe PactI'vity aiffe any' Pre'ventive, Mainteniance (PM) or Scheduled Mainltenance Activity for Plant or Switchyard Equipmmit desoribed hIn the Nuaear Swrltehyvd l~terfaae Agreemenits (NSIA's) located an die Nualoar W04498o - Nualoar $Vvitahyari ftprtlq?

Doeip the activity affoot or potentially affact any 'Vronamilisoln Operating Guldellnee op Wdrk.Preoedui,,Q@ unclopi Doolmlne" apiampits with eup TronerrIselen and DI ~builonorgramiptlen (could be outside company or Damlnlcii TAD dapoelndin on the Mahanr) aoshor Unden uow~bo§it

$ignature required on Engineering Product Cover Sheet.

Ffýn Nq, ?ý 1173 (12w 20, Q)

Design Effects Table Dofminione *NS-A N I00 'ATAHEN ,

`14,0 0*,Al1 (Centainment Recireulatioe Sump) eeneitiene MfsNo a11 PPOFrnr Owner Name-

1. Does the change Involve the addition, removal, or change of any fibrous material In eei~taiorvent (i..jinauiatlon, damming material, fife atops, cloth rifrre, fire retardant , flIerS, or soraens)?

2. Does the change involve the addition of any adhesive (labels, stickers, or signs) to eontainment'?

3. Does the activity Impact a containment $8C design besis (e.g., RWS level change)?

4, Does the change Involve the addition of sIgnflicant horizontal or vertical surface area (>100Q.t) Inside containment that could oileoot dust oP digt?

5. Does the Chanea affect the amount of qualified of unqualifod a*eotlag Insldt alnment' Is a eomponent with a qualified eoating being added to containment within the 081-101 fimiting break areas?

Dees the *chmg mpiac oan unqulIffled 4o0tlng with a qualsfed aoatlng?

• Is a component with an unqualified eoating being added anywhere inside containment?

• Are surfaoos beiNg left uncoated in agntoinment?

6. Could the change affect the p~st-LOCA water level, sumnp temperature, or recirculation water flow paths in containment?

• Does the change add, move, or modify high-energy piping gannected to the RCS?

• Will the activity inyolve moving, relocating, or repositioning any SSC, sterago scaffolding, oF temporary modification Intended to be left invde containrent during power operation?

7.A Will the actlyity resuit in the Conditions to produpe any chemibal effects or add to the generation of Iltent debris, cReta.or cor

.partiul~ito es da'sNbad In l091?W?

VDoes the change involve the Installation or removal of reoctlve metals (aluminum), wiilch Will be subjected to entalement qpray or Immarvaili dUgrig uLOQA?

• Is a change to an existing chemical quantity being made in'contaIrment (e.g,, sodium hydrodde) or any new ehemlcal being addeed to

• Will the change Involve exposing concrete by either the addltion of new conerete, groutr or the removal of seating on existing eqnamlo OF Smut autfae.s, whilh VAI 0 i anorote IU'dcos to oontaiFiifnt Upray of'if4rmii dugna i LOOA?

8. Will the activity result In the eonditions to produce-any downstream gffects by addlng.to the gereration of latent debris, eoHMte, GF corfeaenl partlulates sa 'derbed in (GaiIQ1e?

I any component or pert being added which contains par*lculiet (i.e. sand) or which eould break down to partleuiate when e@*Opsad to Is high praeqqra steam/water Jate qP erothkimment ternpr-04ur Aend humldity dudno a LOCA (@.g.: light buiblA)

• Does the change affect the minimum flow clearance of any qomponent In the ECOS system or make a change to the materiai of any component, which could experience abrasive or erosive wear from debris-laden water during post-LOCA rseiroulotion? Dose chaoinga affect. Debris Plugging of EC($ $trainers? VA Plants [CM:3.14]

9. Are there any components added, removed, or modlfled such that free volume or heat sink in 6entalnment is Inereaped or deoraased$

if yea, contact Nuleoar Safety Analysis (NA&U) for input (rogy affect analylil),

Form DDC-02, Revision 0 Dominion IIDesign KPS Change T~raining Impact Evaluation:

[1 Process _

- Refer to FORM INSTRUCTIONS for completing this form -

For Responsible Engineer Use Design Change Number: Temporary Modification 2012-12 Rev.: 0

Title:

Second RHR-600 Leak Repair A copy of the draft design description AND this form have been attached to the following CRS item:

CONDITION REPORT #: CR473227 Responsible Engineer (RE): Tim' LaHann/Jim Brandtjen 05/02/2010 Print Name Date

_-_______ ... '*. " : ,." "For Training. Use._ _ _ _ __'"

Provide, Training Provide Training Prior To Training Materials' Training Required Turnover Plan Impacted Materials

?  ? Tracking Impact Evaluation  ? Tracking Discipline RFT'No. Y N Y. N Number(s) Performed By Y N Number(s)

Initial Upense PM Morgan N/A Training CR473227 M E] NAMor Auwllsry O~sretor CR473227 12 ,

[ 1] [ N/A PM Morgan 0 N/A W

Ucense- -

Operator CR473227 E2 JD [] N NIA PM Morgan Ej 0 N/A Requalification

.Shlft Technilcal - -

Advisor '2 CR473227 [ N/A PM Morgan' E] N/A Shift Manager CR473227 12 [ 12 [ N/A PM Morgan N/A Electical Maintenance CR473227 12 , 0 1 , 0 ,*. N/A NIA.. .*. . PM Morgan 1] [ N/A Maintenance CR473227 N2 El1 0 N/A PM Morgan 12 0 NIA Matenanical Maintenance Maintenance CR473227 CR727 ] 0

] -2[ ] N/A NAPMogn PM Morgan ... El 0 N/A N

Cupervisor CR473227 12 0 12 [ N/A PM Morgan 12 [ N/A Chemistry CR473227 N/A PM Morgan ] N/A R adl --

tion Protection CR473227 9 0l El 0] N/A PM Morgan N0A Engineering CR473227 12 0 12 9 N/A PM Morgan* N/A Employee CR473227 El 0 Q2 0 N/A PM Morgan 12 j N/A Training-, N/ . . .

Other: N/A .N/A [ ] ]. N/Aw, NAI N/A.

Page 1 of I