HNP-18-035, Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval

Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval

ML18156A026
Person / Time
Site:	Harris
Issue date:	06/04/2018
From:	Hamilton T Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HNP-18-035
Download: ML18156A026 (20)

Text

( a, DUKE ENERGYGI JUN O 4 2018 Serial: HNP-18-035 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill NC 27562-9300 919-362-2502 10 CFR 50.55a

Subject:

Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, lnservice Inspection Program for Containment, Third Ten-Year Interval Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Progress, LLC (Duke Energy), hereby requests Nuclear Regulatory Commission (NRC) approval of the attached relief request for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) inservice inspection program for containment, third ten-year interval. The relief request proposes an alternative to the requirements in 10 CFR 50.55a(g)(4) with regard to the post repair pressure testing requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, 2007 Edition with 2008 Addenda, IWE-5000, as conditioned by 10 CFR 50.55a(b)(2)(ix)(J).

The provisions of this relief are applicable to the third ten-year containment inservice inspection interval for HNP, which commences on May 20, 2018, and is currently scheduled to end on September 8, 2027, as identified in the Third Interval Containment lnservice Inspection Plan, submitted to the NRC on October 23, 2017 (Agencywide Documents Access and Management System Accession No. ML17296A323). Enclosure 1 to this letter provides the relief request for the alternative repair and replacement testing requirements for the containment building equipment hatch sleeve weld. Enclosure 2 provides excerpts from the HNP design specification for the containment liner, air locks, and hatch, which is referenced in Enclosure 1.

Duke Energy requests approval of the proposed alternative by May 30, 2019, in order to support planning for the scheduled refueling outage in the fall of 2019.

This letter does not contain any regulatory commitments.

Please refer any questions regarding this submittal to Jeff Robertson, HNP Regulatory Affairs Manager, at (919) 362-3137.

Sincerely,

~-~

Tanya M. Hamilton

Tanya M. Hamilton Vice President Harris Nuclear Plant 5413 Shearon Harris Rd New Hill NC 27562-9300 919-362-2502 10 CFR 50.55a Serial: HNP-18-035 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63

Subject:

Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Progress, LLC (Duke Energy), hereby requests Nuclear Regulatory Commission (NRC) approval of the attached relief request for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP) inservice inspection program for containment, third ten-year interval. The relief request proposes an alternative to the requirements in 10 CFR 50.55a(g)(4) with regard to the post repair pressure testing requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section Xl, 2007 Edition with 2008 Addenda, IWE-5000, as conditioned by 10 CFR 50.55a(b)(2)(ix)(J).

The provisions of this relief are applicable to the third ten-year containment inservice inspection interval for HNP, which commences on May 20, 2018, and is currently scheduled to end on September 8, 2027, as identified in the Third Interval Containment Inservice Inspection Plan, submitted to the NRC on October 23, 2017 (Agencywide Documents Access and Management System Accession No. ML17296A323). Enclosure 1 to this letter provides the relief request for the alternative repair and replacement testing requirements for the containment building equipment hatch sleeve weld. Enclosure 2 provides excerpts from the HNP design specification for the containment liner, air locks, and hatch, which is referenced in Enclosure 1.

Duke Energy requests approval of the proposed alternative by May 30, 2019, in order to support planning for the scheduled refueling outage in the fall of 2019.

This letter does not contain any regulatory commitments.

Please refer any questions regarding this submittal to Jeff Robertson, HNP Regulatory Affairs Manager, at (919) 362-3137.

Sincerely, Tanya M. Hamilton

U.S. Nuclear Regulatory Commission Page 2 HNP-18-035

Enclosures:

1. Relief Request I3R-18

2. Applicable Sections of Design Specification for Containment Liner, Air Locks, and Hatch, "CAR-SH-AS-1" cc:

J. Zeiler, NRC Senior Resident Inspector, HNP M. Barillas, NRC Project Manager, HNP NRC Regional Administrator, Region II

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 HNP-18-035 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld Inservice Inspection Program, Third Ten-Year Interval for Containment Relief Request I3R-18 (12 pages plus cover)

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 1 of 12 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld Inservice Inspection Program, Third Ten-Year Interval for Containment Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Alternative Provides Acceptable Level of Quality and Safety

1.

ASME Code Components Affected Component:

Containment Building Equipment Hatch Body Ring and Sleeve Class Type:

Code Class MC, Seismic Category I

==

Description:==

Steel Class MC Equipment Hatch Body Ring & Sleeve, which penetrates a cylindrical concrete Containment Building Size:

24-0 Inside Diameter, 1-1/4 Thickness Sleeve Material Material:

ASME SA-516, Grade 70

2.

Applicable Code Edition and Addenda

Shearon Harris Nuclear Power Plant, Unit 1 (HNP), Inservice Inspection Program (ISI) - Third Interval for Containment American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code, Section Xl, 2007 Edition with 2008 Addenda HNP, Equipment Hatch Code of Construction ASME B&PV Code,Section III, 1974 Edition with Winter 1975 Addenda

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 2 of 12

3.

Applicable Code Requirements ASME Code, Section Xl, 2007 Edition with 2008 Addenda, IWE-5221(a) states:

Except as noted in IWE-5224, a pneumatic leakage test shall be performed in accordance with IWE-5223 following repair/replacement activities performed by welding or brazing, prior to returning the component to service.

ASME Code, Section Xl, 2007 Edition with 2008 Addenda, IWE-5223 states:

IWE-5223.1 Pressure. The pneumatic leakage test shall be conducted at a pressure between 0.96 Pa and 1.10 Pa, except when otherwise limited by plant technical specifications, where Pa is the design basis accident pressure.

IWE-5223.2 Boundaries. The test boundary may be limited to brazed joints and welds affected by the repair/replacement activity.

IWE-5223.4 Examination. During the pneumatic leakage test, the leak tightness of brazed joints and welds affected by the repair/replacement activity shall be verified by performing one of the following:

(a) A bubble test - direct pressure technique in accordance with Section V, Article 10, Appendix I, or any other Section V, Article 10 leak test that can be performed in conjunction with the pneumatic leakage test (b) A Type A, B, or C Test, as applicable, in accordance with 10 CFR 50, Appendix J.

10 CFR 50.55a(g)(4) states in part:

Components that are classified as Class MC pressure retaining components and their integral attachments, and components that are classified as Class CC pressure retaining components and their integral attachments, must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of the ASME BPV Code and addenda that are incorporated by reference in paragraph (a)(1)(ii) of this section, subject to the condition listed in paragraph (b)(2)(vi) of this section and the conditions listed in paragraphs (b)(2)(viii) and (ix) of this section, to the extent practical within limitation of design, geometry, and materials of construction of the components.

10 CFR 50.55a(b)(2)(ix)(J) states:

(J) Metal containment examinations: Tenth provision: In general, a repair/replacement activity such as replacing a large containment penetration, cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, pressurizers, or other major equipment; or other similar modification is considered a major containment modification. When applying IWE-5000 to Class MC pressure-retaining components, any major containment modification or repair/replacement must be followed by a Type A test to provide assurance of both containment structural integrity and leak-tight integrity prior to returning to service, in accordance with 10 CFR part 50, Appendix J, Option A or Option B on which the applicants or licensees Containment Leak-Rate Testing Program is based. When

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 3 of 12 applying IWE-5000, if a Type A, B, or C Test is performed, the test pressure and acceptance standard for the test must be in accordance with 10 CFR part 50, Appendix J.

10 CFR 50, Appendix J, Paragraph IV.A states in part:

IV. Special Testing Requirements Containment modification. Any major modification, replacement of a component which is part of the primary reactor containment boundary, or resealing a seal-welded door, performed after the preoperational leakage rate test shall be followed by either a Type A, Type B, or Type C test, as applicable for the area affected by the modification.

4.

Reason for Request

HNP's concrete containment structure is a steel-lined reinforced concrete structure in the form of a vertical right cylinder with a hemispherical dome and a flat base with a recess beneath the reactor vessel. Certain steel components in the containment system are designed and fabricated in accordance with the ASME B&PV Code Section III, Division I, Subsection NE, Class MC Components. One of these steel components is the equipment hatch. The equipment hatch is a welded steel assembly with an overall inside diameter of 24-0. A 15-0 inside diameter bolted cover is provided in the equipment hatch for passage of smaller equipment which is too large to pass through the breech-type personnel air lock. The HNP Final Safety Analysis Report (FSAR), Figure 3.8.1-14 shows the equipment hatch.

Duke Energy Progress, LLC (Duke Energy), plans to replace the reactor pressure vessel head at HNP in the fall of 2019. The 15-0 diameter bolted cover in the equipment hatch is not large enough to allow passage of the original reactor pressure vessel head or the replacement reactor pressure vessel head. Therefore, the equipment hatch will be severed at the 24-0 diameter portion of the equipment hatch called the penetration sleeve. The 24-0 diameter body ring welds to a 24-0 diameter steel sleeve that penetrates containment. The penetration sleeve is attached to the concrete wall using anchorages extended into the concrete. The penetration sleeve is welded to the containment liner plate on the inside of the containment concrete wall.

Removing the equipment hatch body ring and a portion of the penetration sleeve will create a 24-0 opening that will provide access for the removal of the original reactor pressure vessel head, and installation of the replacement reactor pressure vessel head. Following replacement of the reactor vessel head, the equipment hatch body ring will be re-welded to the sleeve with a full penetration weld, and the containment building equipment hatch will be restored to its original design requirements.

Once the equipment hatch body ring has been re-welded to the existing sleeve that penetrates the concrete containment wall, a leakage test in accordance with IWE-5223, as modified by 10 CFR 50.55a, Paragraph (b)(2)(ix)(J), would be required. IWE-5223.2 allows the licensee to limit the test boundary to the area impacted by the repair/replacement activity. However, 10 CFR 50.55a, Paragraph (b)(2)(ix)(J) requires an Integrated Leak Rate Test for any major containment modifications, and this paragraph defines major containment modifications as

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 4 of 12 cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, The repair restores the equipment hatch body ring weld to the containment penetration sleeve in accordance with ASME Code requirements. An effective post repair test of the equipment hatch weld's leak tight integrity can be performed by an alternate leakage test, which pressurizes only the area affected by the reinstallation weld. The proposed alternative is in lieu of the required Appendix J, Type A, integrated leak rate test (ILRT) following restoration of the equipment hatch body ring by welding to the penetration sleeve. Duke Energy proposes to conduct the local leak rate test at 1.15 x design pressure (Pd), where Pd = 45 psig, as required by the original design specification.

Duke Energy has determined that the localized leakage testing specified in this request will provide an acceptable level of quality and safety. Relief is requested in accordance with 10 CFR 50.55a(z)(1).

5.

Proposed Alternative and Basis for Use Proposed Alternative Duke Energy proposes to perform a localized leakage test on the equipment hatch body ring to the containment sleeve re-installation weld area in lieu of the Type A ILRT on the entire concrete containment structure as required by 10 CFR 50.55a(b)(2)(ix)(J). 100 percent radiographic and 100 percent magnetic particle testing of the reinstallation weld area shall first be performed in accordance with ASME Code,Section III, Subsection NE. Then, the equipment hatch reinstallation weld shall have a leak chase channel welded over it with a screwed half coupling to allow pressurization. Next, a channel strength test and simultaneous leakage (pressure decay) test shall be performed in accordance with the original design specification, CAR-SH-AS-1, Paragraph 26.2, which is provided in Enclosure 2 of this submittal. Additionally, leakage testing shall be performed as required by CAR-SH-AS-1, Paragraph 26.3, as shown in. Note that in lieu of the halogen sniffer test specified in CAR-SH-AS-1, Paragraph 26.3, a bubble test - direct pressure technique in accordance with ASME Code,Section V, Article 10, Appendix I shall be performed consistent with IWE-5223.4. The coupling will be capped after the test is complete. Illustrations of the cut location of the equipment hatch sleeve, the separation of the body ring from the equipment hatch sleeve after the cut, the leak chase test channel configuration, the reinstallation welds, and the leak chase test channel coupling assembly installation, are provided in Figures 1-5 of this Enclosure.

The channel strength test as specified in the original design specification requires that the leak chase channel be held at a pressure of 51.8 pounds per square inch gauge (psig) for two hours and then solution film tested. The basis for 51.8 psig is 1.15 x design pressure (Pd), where Pd =

45 psig. Testing at 51.8 psig (1.15 x Pd) requires relief from IWE-5223.1, which states that the pneumatic leakage test shall be conducted at a pressure between 0.96 Pa and 1.10 Pa. From HNP Technical Specifications, Section 6.8.4.k, Pa for HNP is 41.8 psig. The proposed test pressure is acceptable since this is consistent with the original design specification, and the pressure is only being applied to the local weld and not the entire containment. Compensation will be made for any change in ambient air temperature during the test, if necessary. In the

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 5 of 12 event that any leakage is identified, the cause of the leakage shall be determined and any required weld repairs shall be performed.

Following the channel strength test and simultaneous pressure decay test, the owner requirement to perform solution film testing will be met by performing a bubble test - direct pressure technique, as required by ASME Code,Section XI, IWE-5223.4 and ASME Code,Section V, Article 10, Appendix I. The pressure applied to the test channel shall be 51.8 psig.

Acceptance criteria shall be per ASME Code,Section V, Article 10, Appendix I. If leakage occurs, the weld shall be repaired and the test repeated. This leakage test shall be performed after reinstallation of the equipment hatch sleeve weld, prior to entry into Mode 4.

The localized leakage bubble test on the pressure boundary weld area of the equipment hatch sleeve will provide a more effective examination of the re-installation weld than the Type A ILRT as required by 10 CFR 50.55a(b)(2)(ix)(J). Therefore, an alternative to the 10 CFR 50.55a requirement is requested pursuant to 10 CFR 50.55a(z)(1) in that the proposed alternative provides an acceptable level of quality and safety.

Basis for Use The repair and replacement activities associated with temporary removal and reinstallation of the HNP equipment hatch body ring will be performed in accordance with the requirements of ASME Code,Section XI, 2007 Edition with 2008 Addenda. ASME Code,Section XI, Paragraph IWA-4411, identifies that welding and installation activities shall be performed in accordance with the Owners requirements and the original Construction Code. Fabrication and installation activities (i.e. cutting and welding) will be performed in accordance with the original construction code of ASME Code,Section III, Subsection NE, or as reconciled to a later edition. The reinstallation of the equipment hatch body ring and associated weld will return the equipment hatch body ring weld to its original design requirements.

The activities proposed and manner of testing proposed are in accordance with the original HNP FSAR design and licensing basis. FSAR, Section 3.8.1.1.3.3 states, The equipment hatch is a welded steel assembly having an inside diameter of 24 ft. 0 in. with a weld-on cover with sufficient material to initially allow for six removals and rewelding. As shown in Enclosure 2, the CAR-SH-AS-1,Section II, item 8, states in part, The equipment hatch which functions as a pressure retaining boundary and is designated a Part shall be a welded steel assembly having an inside diameter of 24 ft 0 in. with welded-on cover with sufficient material to allow 6 Removals and Re-Welding. HNP's concrete containment building has a steel penetration sleeve that is intended to be cut and re-welded to allow passage of large components during outages. In addition, ASME Code,Section XI, IWE-5000 does not require an ILRT after the type of activity that Duke Energy will be performing. IWE-5223.2 states, The test boundary may be limited to brazed joints and welds affected by the repair/replacement activity.

Prior to performing the reinstallation weld, the surfaces to be welded will be prepared in accordance with the approved ASME Code Repair and Replacement Plan. Welding will be performed by qualified personnel, and the weld shall comply with applicable ASME Code,Section III design and non-destructive examination requirements. Post weld examinations will be performed on the equipment hatch body ring weld, which will include both a 100 percent magnetic particle testing and 100 percent radiographic examination of the circumferential weld.

Therefore, the equipment hatch body ring weld will be restored to its original design

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 6 of 12 requirements. The examinations will ensure 100 percent volumetric weld integrity and will meet the acceptance criteria of ASME Code,Section III requirements.

The proposed local leakage test (bubble test) will provide further confirmation of leak tight integrity for the weld repair. The acceptance criteria for the local leakage test will ensure zero leakage at the weld area. A Type A ILRT would measure the total rate of containment leakage.

Therefore, the leakage acceptance criteria for the proposed local leakage testing is more stringent than that of a Type A ILRT. Pressurization to greater than or equal to design pressure will assure the structural integrity of the equipment hatch reinstallation weld. If there is any leakage of the equipment hatch at the repair weld, it would be identified by the local leakage test and corrected.

Prior to plant startup from the refueling outage, the equipment hatch will be tested to check for any leakage through the equipment hatch O-ring seals. This test is a Type B test, as described in 10 CFR 50, Appendix J. HNP currently completes this Type B test during each refueling outage to meet a 24 month frequency requirement, as required by HNP TS Surveillance Requirement 4.6.1.2.a.

The removal and restoration of the equipment hatch was performed at HNP in 2001 during the steam generator replacement (SGR) project, which included completion of a local leak rate test.

An ILRT was not performed in 2001 during the equipment hatch restoration since it was not required. The test configuration used in 2001 is similar to the test configuration proposed in this relief request. The local leak rate test performed in 2001 utilized a plant procedure, which required a minimum test pressure of 41.8 psig (Pa) and a maximum pressure of 45 psig (Pd). A test pressure of 51.8 psig will be used for the local leak rate test for this activity in the fall of 2019, which is greater than the test pressure used during the SGR replacement outage in 2001.

The 10 CFR 50, Appendix J program for HNP required a Type A ILRT of the containment building in 2012 (well after the SGR in 2001). There were no issues with leakage at the equipment hatch circumferential weld identified and the test results indicated that the Containment System is performing well within leakage limits as approximately 42 percent margin remained. This demonstrates that an effective local leak rate test was performed previously during the steam generator replacement in 2001.

CONCLUSIONS The combination of a 100 percent volumetric radiographic examination, 100 percent magnetic particle testing, and the localized leak rate test of the reinstallation weld at 1.15 x design pressure is sufficient to confirm the integrity of the equipment hatch replacement body ring weld.

In accordance with the requirements of 10 CFR 50.55a(z)(1), the localized leakage testing specified in this relief request provides an acceptable level of quality and safety in lieu of the required Type A test.

6.

Duration of Proposed Alternatives The performance of a localized leak rate test is a one-time alternative for the ASME Code repair and replacement activity associated with the containment building equipment hatch during the

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 7 of 12 reactor vessel head replacement scheduled for the fall of 2019 refueling outage. The repair and replacement activity using this proposed alternative shall be acceptable for the life of the plant, or until such time that a future repair and replacement activity of this nature is performed again, whichever comes first.

7.

Precedents

1. Entergy Operations, Inc. (Entergy) letter to Nuclear Regulatory Commission (NRC) dated July 27, 2011, Request for Alternative to ASME IWE-5221 Regarding Post Repair Testing of Waterford 3s Steel Containment Vessel Opening (Agencywide Documents Access and Management System (ADAMS) Accession Number ML112150195)

2. NRC letter to Entergy dated January 4, 2012, Waterford Steam Electric Station, Unit 3 -

Request for Alternative to ASME IWE-5221 Regarding Post-Repair Testing of Steel Containment Vessel Opening (ADAMS Accession No. ML113330137)

3. Tennessee Valley Authority (TVA) letter to NRC dated November 17, 2005, Watts Bar Nuclear Plant (WBN), Unit 1 - One Time Request for Relief from American Society of Mechanical Engineers (ASME),Section XI Code Requirements - Tests Following Repair, Modification, or Replacement (IWE-5221) (Adams Accession No. ML053260493)

4. NRC Letter to TVA dated August 30, 2006, Watts Bar Nuclear Plant, Unit 1 - One-Time Request for Relief from American Society of Mechanical EngineersSection XI Code Requirements - Tests Following Repair, Modification, or Replacement (IWE-5221) (ADAMS Accession No. ML061590111)

5. FirstEnergy Nuclear Operating Company (FENOC) letter to NRC dated March 8, 2013, 10 CFR 50.55a Request RR-E1, Proposed Alternative Regarding Post-Repair Pressure Testing Requirements (ADAMS Accession No. ML13070A226)

6. NRC Letter to FENOC dated May 8, 2013, Davis-Besse Nuclear Power Station, Unit No. 1

- Safety Evaluation In Support of Proposed Alternative Regarding Post-Repair Pressure Testing Requirements (ADAMS Accession No. ML13121A404)

8.

References

1. Design Specification CAR-SH-AS-1, Containment Liner, Air Locks, and Hatch, Revision 12

2. HNP FSAR, Amendment 61, Section 3.8.1 and Figure 3.8.1-14

3. ASME B&PV Code,Section XI, Article IWE-5000, System Pressure Tests, and Article IWA-4000, "Repair/Replacement Activities, 2007 Edition with 2008 Addenda

4. ASME B&PV Code,Section III, Article NE-5000, "Examination," and Article NE-6000, "Testing," 1974 Edition with 1975 Winter Addenda

5. ASME B&PV Code,Section V, Article 10, "Leak Testing," of Appendix I, 2017 Edition

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 8 of 12 Figure 1: Cut location of the Equipment Hatch Sleeve (Side View)

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 9 of 12 Figure 2: Separation of the Body Ring from the Equipment Hatch Sleeve (Side View)

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 10 of 12 Figure 3: Installation of the Leak Chase Test Channel (Side View)

Notes:

1. Leak chase channel (3 inches in width) still in place for steam generator replacement weld

2. Original hatch sleeve construction installation weld to liner sleeve (leak chase channel was removed)

3. The new cut location of the equipment hatch sleeve will be 6.5 inches toward the concrete containment from the centerline of the Item 1 steam generator replacement weld.

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 11 of 12 Figure 4: Reinstallation Weld (Side View)

The completed reinstallation weld Inside groove of reinstallation weld and non-destructive examinations performed first

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 1 Page 12 of 12 Figure 5: Installation of the Leak Chase Test Channel Coupling Assembly (Side View)

U.S. Nuclear Regulatory Commission Relief Request I3R-18 HNP-18-035 Enclosure 2 HNP-18-035 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400/Renewed License No. NPF-63 Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld Inservice Inspection Program, Third Ten-Year Interval for Containment Applicable Sections of Design Specification for Containment Liner, Air Locks, and Hatch, "CAR-SH-AS-1" (3 pages plus cover)

Purchaser's Identification No. CAR-SH-AS-1 REV. 12 EBASCO SPECIFICATION CONTAINMENT LINER, AIR LOCKS AND HATCH SECTION II - ARRANGEMENT

7. The steel Containment Liner which functions as a leak-GENERAL R9 tight membrane and is designated a Part shall be of the dimensions shown on the design drawings and shall have one-equipment access hatch with two covers, one personnel air lock, one personnel escape lock, a circular crane girder and rail, penetration sleeves and other items as shown on the design drawings. Also included are four (4) valve chambers and their appurtenances.

The valve R9 chambers and their appurtenances as shown on the design drawings are 9'*0 diameter by 19 1-0 long airtight enclosures which function as a secondary containment boundary to completely enclose the sump lines and isolation valves.

8.

The equipment hatch which functions as a pressure re-taining boundary and is designated a Part shall be a welded steel assembly having an inside diameter cf 24 ft O in. with welded-on cover with sufficient material to allow 6 Removals and Re-welding.

A 15 ft O in. inside diameter bolted cover shall be pro-vided in the equipment hatch cover for passage of equipment which is too large to pass through the personnel locks during plant shutdown.

Provision shall be made to pressurize the space between the gaskets of the bolted hatch cover to 51.8 psig.

.1 For the equipment hatch covers, lifting lugs shall be provided for moving each cover away from the opening when they are removed.

Mobile cranes or stationary hoists will be provided by Purchaser for liftings of hatch covers. Seller shall indicate the lif ting weight of each cover

• EQUIPMENT HATCH

• 2 Pads required for support of removable floor closure sections when the covers are in the closed position shall be an integral part of the cover and shall be designed by Purchaser and installed by Seller at locations shown on Purchaser's design drawings.

9.

One Breech-Type personnel air lock and one personnel escape lock which function as pressure retaining boundaries shall be furnished.

These shall be welded steel assemblies, and provision shall be made to pressurize the space between their door gaskets to 51.8 psig.

PERSONNEL LOCKS R9 R9 R9 R2 Rl R9 Rl R2

.l The Breech-Type personnel air lock which is designated R9 a Part and is used to permit personnel access and egress and passage of small equipment during plant operation shall have a 9'-0" inside R4 diameter and a clear internal length of 11 ft 9 in. with full diameter breech lock doors hinged to open outward from each end of the lock.

411 Minimum diameter viewports shall be provided on each door.

Doors R2 for the lock shall be hydraulically sealed and electrically interlocked.

Electrical limit switch contacts shall be provided to interlock two of R9 the Purchasers valves with the air lock locking pin movement. The wiring required to accomplish this shall be terminated in a junction box mounted outside the containment vessel.

9

EBASCO SPECIFICATION Purchaser's Identification CONTAINMENT LINER, AIR LOCKS AND HATCH SECTION VI* TESTING AND INSPECTION No.

CAR-SH-AS-1 REV. 12 2

REQUIREMENTS

5.

The Seller shall perform spot radiography and vacuum box FOR TESTING testing of the liner in accordance with the applicable requirements LINER of the ASME Code Section III, Division 2, CCSOOO to demonstrate the R6 integrity of the complete containment liner to the satisfaction of thePurchaser *

• 1 After completion of a successful vacuum box and radiography tests, and subsequent repair and retesting of any found defects the welds shall be covered by channels as indicated on the design drawings.

A channel strength and simultaneous leakage (pressure decay) test will then be performed by applyine ~

psig air pressure to the ch~nncls for Rt least two hours, after which all welds will be solution film tested. For those cases where a vacuum box test has been performed on the liner seam welds and/or penetration collar attachment welds, these welds need not be solution film tested. If there is any indicated loss (not to exceed pressure drop specified in Vendor's procedures) of channel test pressure within the two hour period, any such defects are to be repaired. Compensation for change in ambient air temperature will be made if necessary.

Leak testing shall be performed in ac-cordance witl1 the requirements of ASHE Section III Division 2 cc 5535.2 Rl R4 RIO R4 R2 R4 Rl, R2 R4

.2 All testing connections and accessories as applicable shall be permanently left in place for future testing withal! connections properly sealed.

3 For penetrations, leak testing procedures will be conducted in.accordance with ANSI N6.2.

Strength and leakage tests will be Rl performed in accordance with Paragraph 25. 1 and 25. 2.

Any leaks R2 found are to be repaired and retested.

This procedure to be repeated R6 until no leaks exist.

22

Purchaser's Identification No. CAR-S11-AS-l REV. 12 EBASCO SPECIFICATION CONTAI~'MENT LINER, AIR LOCKS AND HATCH SECTION VI TESTING AND INSPECTION

26.

For the Air Locks, Valve Chnmbers and Hatch, The Seller shall perform the following shop tests in accordance with applicable requirements of the ASME Code and with requirements hereafter specified to demonstrate the integrity of the equipment fabricated by him to the satisfaction of the Purchaser.

REQUIREMENTS FOR TESTING AIR LOCKS, HATCH AND VALVE CHAt-IBERS

.1 All butt welds shall be radiographed and magnetic particle inspected *

• 2 After repair of any defects found from radiography has been performed, the welds between the flanges and the barrel of the lock and hatch shall be covered by testing angles or pipes as shown on the design drawings.

A channel strength and simultaneous leakage (pressure decay) test will then be performed in accordance with paragraph 25.1 *

• 3 On successful conclusion of the strength test, halogen sniffer tests will be performed by introducing a mixture of air and Freon (0.3 ounces per cubic foot) into the test angles at a pressure of 45 psig.

The detection bf any amount of halogen shall indicate a leak requiring weld repairs and retesting.

An alternate halogen test and acceptance standard may be used upon approval by Purchaser.

The test angles shall be held at 45 psi(+ 0.5 psi) pressure for a period sufficiently long enough to complete testing.

Where leaks occur, welds will be repaired by arc gouging, grinding, chipping and/or machining and then rewelding.

The halogen leak test will also be applied to each door of the personnel locks and manway cover of valve chambers.

.4 A pressure test shall be conducted for valve chambers and air locks, in accordance with requirements of Section III of the ASME B & PV Code Article NE-6000.

27. After the containment structure is complete with liner, concrete structures, and all penetrations, the personnel locks and equipment hatch in place, a leakage and structural proof test will be made by the Purchaser and any leaks found in the liner, penetrations, air locks or hatch which are due to defective material or workmanship shall be repaired by the Seller.

P.6 Rl R4 R2 R4 Rl R4 R4 R7 FINAL LEAKAGE AND STRUCTIJRA L PROOF TESTS

28.

Leakage shall be determined using the "Absolute Method".

The "Absolute Method" is the measurement of the internal pressure, temperature and dew*point temperature over a predetermined period of time.

EQUIPMENT FOR TESTS Measurements for the structural proof test will consist of deflections, deformations and concrete crack sizes and spacing.

A detailed Specification will be issued for all the equipment and instruments to be used in the above tests.

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