Text

Units 1 and 2; Fourth 10-year Interval In-service Inspection Program ISI Program Update: Notification of Impractical ASME Code Requirements (NL-16-2707)
	ML16362A273
Person / Time
Site:	Hatch
Issue date:	12/27/2016
From:	Pierce C; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-16-2707
	Download: ML16362A273 (68)
	v • d • e

~ Southern Nuclear Charles R. Pierce 40 Inverness Center Parkway Regulatory Affairs Director Post Office Box 1295 Birmingham, AL 35242 205 992 7872 tel 205 992 7601 fax crpierce@southernco.com DEC 2 7 2016 Docket Nos.: 50-321 NL-16-2707 50-366 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Units 1 and 2 Fourth 10-Year lntervallnservice Inspection Program lSI Program Update: Notification of Impractical ASME Code Requirements Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(g)(5)(iii), Southern Nuclear Operating Company (8NC) hereby notifies the U.S. Nuclear Regulatory Commission (NRC) that 8NC has determined that conformance with certain ASME Section XI Code (Code) requirements is impractical for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (HNP). SNC submits the enclosed information to support the determinations of impracticality which are based on demonstrated limitations experienced when attempting to comply with the Code requirements during the fourth 10-year lSI program interval. Requests for relief are enclosed.

This letter contains no new NRC Commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.

Re*zc~ fj:;d*

C. R. Pierce Regulatory Affairs Director CRP/EFB/Irc

Enclosures:

1) 181-RR-13

2) ISI-RR-14

3) 181-RR-15

4) 181-RR-16

5) ISI-RR-17

6) 181-RR-18

7) 181-RR-19

8) 181-RR-21

9) 181-RR-22

10) 181-RR-23

11) 181-RR-24

U.S. Nuclear Regulatory Commission NL-16-2707 Page2 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M.D. Meier, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RType: CHA02.004 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch

Edwin I. Hatch Nuclear Plant- Units 1 and 2 Fourth 10-Year lntervallnservice Inspection Program lSI Program Update: Notification of Impractical ASME Code Reguirements Enclosures Relief Requests ISI-RR-13 ISI-RR-14 ISI-RR-15 ISI-RR-16 ISI-RR-17 ISI-RR-18 ISI-RR-19 ISI-RR-21 ISI-RR-22 ISI-RR-23 ISI-RR-24

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-13 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality-(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category 8-J, Item Number 89.11 :

1831-1RC-288 Austenitic steel circumferential butt weld (Pipe to Valve) which was overlaid during the 1990 refueling outage due to intergranular stress corrosion cracking (IGSCC) observed when this weld was examined. One circumferentially-oriented flaw was recorded on the pipe side of the weld with a length of 6.2-inches and a depth of approximately 47%. The original butt weld was a stainless steel weld joining wrought stainless steel piping to an austenitic stainless steel casting (CASS) valve and was overlaid with Type 308L stainless steel.

In the fourth lSI Interval, Hatch elected to implement 8WRVIP-75 for the examination of austenitic piping welds which were previously examined to the commitments of Generic Letter 88-01, "NRC Position on lntergranular Stress Corrosion Cracking in 8WR Austenitic Stainless Steel Piping ," dated January 25, 1988 and NUREG-0313, Revision 2, "Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping," dated January 1988. BWRVIP-75 provides revisions to the scope and frequencies of inspections for austenitic piping welds for Categories A through E.

Weld number 1831-1 RC-288-13 was examined for fourth lSI Interval credit during the 1R24 outage in Spring 2010 as a Category E weld.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code Requirements Piping weld overlays have been a successful mitigation approach to address 8WR IGSCC issues since the 1980s. The NRC and the industry have worked together on the technical issues related to the installation and examination of weld overlays.

Subsection IW8 of ASME Section XI of the 2001 Edition through the 2003 Addenda does not address the examination of weld overlays nor is this detail in later editions of ASME Section XI. Appendix Q provides the needed guidance but was not incorporated ISI-RR-13 Version 1.0 Page 1 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-13 into ASME Section XI until the 2005 Addenda to the 2004 Edition. The examination of weld overlays is performed using qualified procedures, equipment, and personnel meeting Supplement 11 of Appendix VIII for ASME Section XI. The nondestructive (NDE) procedures provide the detailed examination volume. The typical weld overlay configuration is shown below with the required examination volume being shown in Figure RR-13-1. It should be noted that the typical piping weld overlay is installed such that there is an equal amount of weld overlay material on each side of the original butt weld. The design of the weld overlay on piping weld 1831-1RC-28B-13 was a non-typical weld overlay as shown on Figure RR-13-2 to address materials issues as discussed below.

1/21n.---..-~ ~-112in.

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\

\ I I

I *c t

\ I

\~

"- J J

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As-found flaw Examination Volume A-8-C-D Figure RR-13-1 Typical Weld Overlay

4. Impracticality of Compliance Overlay Material and Design:

This weld overlay was installed during the Spring 1990 refueling outage with final documentation of the IGSCC examinations and evaluations plus weld overlay designs submitted to the NRC on June 29, 1990 per Hatch Letter HL-1158. The report number is SIR-90-039 and was developed by Structural Integrity Associates (SIA) for Hatch.

Figure RR-13-2 shows the final configuration of the weld overlay including the Pipe to Valve weld and the installed weld overlay. This weld overlay was designed to the requirements of USNRC NUREG-0313, Revision 2. The analytical bases for the design of the repairs are in accordance with the requirements of ASME Section XI, IWB-3641 as specified in NUREG-0313. It should be noted that this design was issued several years prior to ASME Code Case N-504 being included in Regulatory Guide 1.147, Revision 11 in October 1994.

ISI-RR-13 Version 1.0 Page 2 of5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-13 The downstream edge of the weld overlay (valve side) stops on the circumferential butt weld to avoid the possibility of opening casting discontinuities during the welding process. Also since the cast austenitic stainless steel casting valve and the weld material are known to be IGSCC-resistant, the overlay did not need to cover these components from a material viewpoint. SIA has recently confirmed that no incidences of IGSCC in weld metal or castings in BWRs have been reported. Therefore, the operating experience of these materials supports their IGSCC resistance.

The length of the weld overlay was sufficient to transmit the load from one end of the cracked pipe through the overlay to the other side. At the time that this weld overlay was designed, no specific requirements for the length existed and an evaluation was performed using finite element techniques to determine the effect of these short overlays on the stress limits in the ASME Code, Section Ill design rules. The acceptance criterion used was subsequently adopted in ASME Code Case N-504. The results of the evaluation indicated that the calculated stresses met the applicable ASME Code allowable design stresses.

Overlay lnspectabilitv:

As noted previously, this is a non-typical weld overlay which results in examination limitations not seen on the typical weld overlay. In addition, the examination volume has had to be modified to fit the specific configuration of this weld overlay. Figure RR-13-3 shows the required examination volume associated with this weld overlay and the limitations imposed by the configuration. The composite ultrasonic examination coverage was calculated as 60% by examination personnel. The most susceptible portion of the examination volume was examined with automated phased-array ultrasonic techniques. The remainder of the examination volume is located above cast material which as discussed earlier, has no history of IGSCC.

Limitations and coverage for specific types of examinations is discussed below.

Circumferential Flaws Examinations from the upstream and downstream sides included 25°, 45°, 60°, 70° refracted longitudinal waves plus a oo longitudinal wave. Only non-relevant indications and acoustic interface were recorded.

Axial Flaws Examinations looking for axial flaws included 45°, 60°, and 70° refracted longitudinal waves. Only non-relevant indications and acoustic interface were recorded.

5. Burden Caused by Compliance To appreciably increase the examination volume coverage of this weld overlay would require a modification of the original weld overlay. In addition to significant radiation ISI-RR-13 Version 1.0 Page 3 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-13 dose associated with this work, SNC believes the cost would nominally cost 2.0 million dollars based on Hatch experience.

The other option would be to replace this section of piping with IGSCC-resistant material which would result in more radiation dose and higher costs. SNC considers these options as an undue burden considering the strong operating experience of weld overlays at Hatch and in the industry. In addition, the examination of the most susceptible volume has been performed on this weld. Therefore, it is concluded that the ASME Code requirement is impractical.

6. Proposed Alternative and Basis for Use A significant volume of the examination coverage along with VT-2 examinations associated with the Class 1 leakage test performed each refueling outage provide reasonable assurance that unacceptable flaws have not developed in the subject weld or that they will be detected and repaired prior to the return of service . During operation, leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Section 4.10 for Unit-1). Thus an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examinations in lieu of the Code requirement. Relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents Previous examinations of this weld were performed to address the Hatch commitments of Generic Letter 88-01, "NRC Position on lntergranular Stress Corrosion Cracking in BWR Austenitic Stainless Steel Piping," dated January 25, 1988 and NUREG-0313, Revision 2, "Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping," dated January 1988.

9. References None ISI-RR-13 Version 1.0 Page 4 of5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-13

/

Pipe Side Valve Side Flow------

Avg Pipe Thickness -1.36" Avg Overlay Thickness- .60" Weld Overlay Width - 4.350" Weld Overlay for 1831-1 RC-288-13 Figure RR-13-2 Legend:

Pipe Original Wold Overlay Exam Area Susceptible Area

/

Pipe Side Valve Side Susceptible Area ~ /

Flow-------

60% composite coverage Weld Overlay for 1831-1Rc-2BB-13 Figure RR-13-3 ISI-RR-13 Version 1.0 Page 5 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-14 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category 8-F, Item 85.20, 2.5-inch Nozzle-to-Safe End butt welds for nozzles N16A and N168.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code Requirements Table IW8-2500-1, Examination Category 8-F, Item 85.20 requires a surface examination of the nozzle-to-safe-end welds on pressure retaining dissimilar metal welds as defined by Figure IW8-2500-8(b). Relief is requested from performing the Code-required surface examination of the above identified RPV nozzle-to-safe-end piping welds.

4. Impracticality of Compliance The nozzle welds are inaccessible due to the design of the RPV insulation, the concrete shield wall and the shield blocks that surround the piping. The shield blocks are welded together and cannot be removed without a significant amount of work. The piping exits the shield wall through an access port of approximately five inches square which severely limits access. Pictures RR-14-01 and -02 show this configuration.

5. Burden Caused by Compliance To allow access to perform these examinations on Hatch-1, a redesign of the 1N 16 nozzles would be required. Therefore, it is concluded that the ASME Code requirement is impractical.

6. Proposed Alternative and Basis for Use VT-2 examination in conjunction with the Class 1 system leakage/hydrostatic test each refueling outage will provide adequate assurance that any flaw(s) that might have propagated through the subject welds are identified and repaired prior to returning the plant to power operation. During operation, leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Section 4.10 for Unit-1).

ISI-RR-14 Version 1.0 Page 1 of 3

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-14 Thus an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examination in lieu of the Code requirement. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth ISIInservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents The physical limitations described in this relief request are identical to Hatch for the third lSI Interval with Relief Request RR-6 which was approved by the NRC on June 16, 1997.

Alternate approaches for the remaining nozzles discussed in RR-6 have been utilized for the fourth lSI Interval. In addition, Hatch submitted Revision 2 of RR-6 on June 8, 1998 and the NRC approved it on June 3, 1999. This revision did not address the Hatch-1 N16 nozzles.

9. References The TAC Numbers for the original NRC SER dated June 16, 1997 for the third lSI Interval relief request RR-6 are M93918 and M93919.

ISI-RR-14 Version 1.0 Page 2 of 3

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-14 Picture RR-14-01 Picture RR-14-02 ISI-RR-14 Version 1.0 Page 3 of3

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6}(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category 8-0, Item 814.10, Pressure Retaining Welds in Control Rod Housings. The extent of examination is 10% of the peripheral Control Rod Housing welds.

The general configuration of the Control Rod Drive (CRD) Housing welds is similar for both Hatch units. The bottom head of the Reactor Pressure Vessel (RPV) has a penetration for each CRD location. Outside of the RPV, there is a housing to flange weld (located in the Subpile Room under the RPV) and an intermediate housing to housing weld (located inside the RPV Support Skirt). Each Hatch unit has 36 peripheral CRD housings and therefore,Section XI requires four CRD housings to be examined for a total of eight CRD housing welds per unit. Figure RR-15-1 shows the CRD housing welds for Unit-1 while Figure RR-15-2 shows the Unit-2 configuration . Figure RR-15-3 gives the Top View of the CRD housings for both Hatch units.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code Requirements Table IW8-2500-1, Examination Category 8-0, Item 814.10 requires either a surface or volumetric examination of the Control Rod Housing welds on pressure retaining butt welds as defined by Figure IW8-2500-18. Southern Nuclear (SNC) evaluated the two examination techniques and determined that surface examinations (liquid penetrant) should be performed to maximize the coverage.

4. Impracticality of Compliance Due to accessibility issues for Unit-1, no CRD housing welds were examined as discussed in the description below. On Unit-2, eight CRD housing welds were examined; however, all eight were the intermediate housing to housing welds below the RPV bottom head with the detailed discussion described below.

Physical limitations exist on both Hatch units which limit the examination of the lower CRD housing weld (housing to flange weld). In particular, accessibility does not exist for the 36 peripheral CRD housing welds on both Hatch units due to the close proximity of adjacent ISI-RR-15 Version 1.0 Page 1 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 CRD housing flanges, neutron monitoring instrumentation and associated cabling plus the horizontal beams, etc. supporting the CRD piping in the Subpile room. Picture RR-15-1 shows the configuration from the Subpile room floor while Pictures RR-15-2 and RR-15-3 show two views of a specific CRD flange highlighting the physical limitations involved.

The second CRD housing weld is located near the RPV bottom head inside the RPV support skirt. The Unit-1 vessel was designed prior to the development of ASME Section XI requirements and access through the support skirt was not included. The Unit-2 design addressed ASME Section XI requirements and an access port was included providing access into the vessel support skirt. Therefore, Hatch-1 has no access while Hatch-2 has access for the housing to housing welds. Since Unit-2 had available coverage, four CRD housing to housing welds were examined to meet the original requirement plus an additional four CRD housing to housing welds were selected for examination so that a total of eight CRD housing to housing welds were examined at Hatch-2 (see Table RR-15-1 for a listing of these eight Unit-2 welds). No limitations were observed nor were any surface indications recorded .

5. Burden Caused by Compliance To appreciably increase the examinations of the CRD housing welds would require a redesign of the RPV bottom head, which would be an undue burden. Therefore, it is concluded that the ASME Code requirement is impractical.

6. Proposed Alternative and Basis for Use For Hatch-1, the VT-2 examination in conjunction with the Class 1 system leakage test each refueling outage will provide adequate assurance that any flaw(s) that might have propagated through the subject welds are identified and repaired prior to returning the plant to power operation. During operation, leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Section 4.10 for Unit-1).

The surface examination of the eight Hatch-2 CRD housing to housing welds plus the VT-2 examination in conjunction with the Class 1 system leakage test each refueling outage will provide adequate assurance that any flaw(s) that might have propagated through the subject welds are identified and repaired prior to returning the plant to power operation.

During operation, leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Section 5.2.7 for Unit-2).

Thus an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examination in lieu of the Code requirement. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

ISI-RR-15 Version 1.0 Page 2 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15

8. Precedents In the third lSI Interval (and prior to that interval}, the CRD Housing welds were not required to be examined due to the IWB-1220(a) exemption and therefore, there is no Hatch precedent. The exemption criterion was changed during the fourth lSI Interval which resulted in the CRD Housing welds being examined .

Other BWRs have submitted similar relief requests. The relief request (RR-46) from Progress Energy Carolinas, Inc. Brunswick Steam Electric Plant dated May 8, 2009 and December 2, 2009 (ADAMS Accession Numbers ML09134011 and ML093440850, respectively) aided SNC in the development of this relief request.

9. References The ADAMS Accession Number for the NRC SER, dated April 7, 2010, for the Brunswick Steam Electric Plant is ML100491269.

ISI-RR-15 Version 1.0 Page 3 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 TABLE RR-15-1 Hatch-2 CRD Housing to Housing Weld Examinations Weld Number Percent Coverage Limitations 2811\2-02-19-1 100% None 2811\2-06-11-1 100% None 2811\2-10-07-1 100% None 2B 11\2-18-03-1 100% None 2B 11\2-26-03-1 100% None 2B 11\2-34-03-1 100% None 2811\2-50-19-1 100% None 2B 11\2-50-23-1 100% None ISI-RR-15 Version 1.0 Page 4 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 UPPER HOUSING IJ£LD

Hl 1811\1-02-19-2 PIPE TO FLANGE STUB TUBE TO JOTTOM HEAD IJELD CRD HOUSI NG TO STUB TUBE IJELD RPV BOTTOM HEAD SA- 533 GR B CLASS 1 TYPICAL CRD HOUSING ASSEMBLY CRD H D U S I N O / IJELDS (304SS)

SECTION A-A B CAP SCREIJS 8 IJASf£RS TITLE CRD HOUSING DETAIL UrEI£-IIIIMIIJCS SOUTHERN NUCLEAR COMPANY G£ -919D274 REV 9 C[-C!~-474 R(V 1 FrGURE RR-15-1 181-RR-15 Version 1.0 Page 5 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 UPPER HOUSI Nu 'WELD CHEAR RPV BOTTOM HEAD>

H2 2Bll\2- 02-19-l PIPE TD PIPE LO'WER HOUSI NG 'WELD

([N SUBPILE Rf][]K>

H2 2Bll\2-02-19-2 PI P E TO fLA NGE STUB TUBE TO BDTTilM HEAD \/ELD CRD HOUSI NG TD STUB TUBE 'W[LD RPV BOTTOM HEAD SA-533 GR B CLASS 1 T YPICAL CRD HOUSING ASSEMBLY CRD HOUSING \/ans C304 SS>

SECTION A-A 8 CAP SCRE\/S 8 WtSHERS nn.E CRD Hll.ISING DETAIL If: 919De74 R['l CJ 851-003 REV I SOUTHERN NUCLEAR CIJMPI\NY GE 9191!260 REV 16 841-003 REV 3 FIGURE RR 2 CE 115701 ISI-RR-15 Version 1.0 Page 6 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number 181-RR-15

. - - - CRD PENET RATIO CTYPICAL 137>

00 02 Of> 10 14 lB 22 26 30 34 Ja *2 *6 50 INCORE FLUX MO NITOR 5t PENETRATIONS

<T YP I CAL 4 3) 47 43 BOTTOM HEAD 39 DRAIN PENETRATIO 3S 31 270° 27 23 19 l:i 11 07 03 INCORE FLUX MONITOR PENETRATION TOP VIE'vl TYPICAL FO R BOTH HATCH UNITS TlTLE CRD 1\Nll IM-<I:RE PEJ£TV,1JDN PLAN 1----------1 11DUDCE ..,..,billS CE 11570:

SOUTHERN NUCLEAR COMPANY 11Sl-G03 REV 0 GE 197R629 REV 9 FIGURE RR-15-3 II I 181-RR-15 Version 1.0 Page 7 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 Picture RR-15-1 ISI-RR-15 Version 1.0 Page 8 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 Picture RR-15-2 ISI-RR-15 Version 1.0 Page 9 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-15 Picture RR-15-3 ISI-RR-15 Version 1.0 Page 10 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii) .

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category B-0, Item 83.90, nozzle to vessel welds. Unit 1 welds are shown in Table RR-16-1 and Unit 2 welds are shown in Table RR-16-2.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code Requirements Table IWB-2500-1, Examination Category B-0, Item 83.90 requires that the examination volume shown in Figures IWB-2500-7(a) through (d) be met. Per Code Case N-460, coverage greater than 90% is acceptable. ASME Code Case N-613-1 was used on barrel-type nozzles to increase the ASME Code percentage.

4. Impracticality of Compliance Coverage was limited due to the geometry of the nozzles and in some cases the proximity of other nozzles or components. When automated scanning was limited, qualified supplemental manual examinations were used to increase the coverage where possible; therefore, coverage was maximized to the extent practical and it would be impractical to obtain any more appreciable coverage.

5. Burden Caused by Compliance Increasing the coverage would require replacing the RPV nozzles with a new design.

6. Proposed Alternative and Basis for Use Coverage was limited due to the geometry of the nozzles and in some cases the proximity of other nozzles or components, as defined in the attached tables. In general, the barrel-type nozzle configuration [Section XI Figure IWB-2500-7(a)] had less coverage than the flange-type nozzle configuration [Section XI Figure IWB-2500-7(b)]. Figure RR-16-1 shows the 1N 1 (flange-type nozzle) while Figure RR-16-2 shows the 1N2 (barrel-type nozzle) for comparison. Figure RR-16-3 has also been included to show the Unit-1 limitations related to the insulation support ring for the Reactor Recirculation nozzles.

ISI-RR-16 Version 1.0 Page 1 of 8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 In most cases, examination for axially-oriented flaws could not be performed from the nozzle side of the weld due to the configuration of the nozzle; however, the presence of an axial flaw does not have a significant impact on the structural integrity of a nozzle weld.

The examination of circumferentially-oriented flaws was typically obtained for these welds.

Selected limitations existed due to the insulation support ring on Unit-1 and adjacent nozzles limited certain examinations (the tables discuss the specific limitations). A significant volume of these welds along with the visual (VT-2 examinations) associated with the Class 1 leakage test performed each refueling outage provide adequate assurance that any flaw(s) that might have propagated through the subject welds are identified and repaired prior to returning the plant to power operation. During operation, leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Sections 4.10 for Unit-1 and 5.2.7 for Unit-2). Thus an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examinations in lieu of the Code requirement. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents The physical limitations described in this relief request are similar to Hatch for the third lSI Interval (RR-44) which was submitted for NRC approval per NL-06-1159 on July 10, 2006.

The ultrasonic examinations performed for the 4th lSI Interval under this relief request contained similar or higher composite coverages with all examinations performed to the Performance Demonstration Initiative techniques.

9. References The NRC SER for the third lSI Interval relief request RR-44 has ADAMS Accession Number ML071360297, dated June 5, 2007.

ISI-RR-16 Version 1.0 Page 2 of 8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 TABLE RR-16-1 Weld Number Description Coverage Basis for Limited Coverage 1B11\1N1A Recirculation Outlet Nozzle to Shell Weld 81.1% Automated and Manual UT scanning was performed from the (Flange-Type Nozzle) vessel OD surface. Automated and Manual UT scanning was restricted due to the nozzle configuration and the proximity of the insulation support ring.

1B11\1 N2A Recirculation Inlet Nozzle to Shell Weld 51% Automated and Manual UT scanning was performed from the (Barrel-Type Nozzle) vessel OD surface. Automated UT scanning was restricted due to the nozzle configuration and the proximity of the insulation support ring. The Manual UT was restricted due to the proximity of the insulation sup~ort bracket.

1B11\1 N2B Recirculation Inlet Nozzle to Shell Weld 52 .2% Automated UT scanning was performed from the vessel OD 1B11\1N2D (Barrel-Type Nozzle) surface and was restricted due to the nozzle configuration and 1B11\1N2E insulation support ring . Supplemental manual examinations 1B11\1 N2G were performed in the restricted areas.

1B11\1 N2H 1B11\1N2K 1B11\1N2C Recirculation Inlet Nozzle to Shell Weld 45.3% Automated scanning was performed from the vessel OD surface 1B11\1N2F (Barrel-Type Nozzle) and was limited due to the nozzle configuration and the proximity 1B11\1 N2J of an insulation support bracket and an insulation support ring.

Manual examinations were performed in areas limited to the automated system.

1B11\1 N3A Main Steam Nozzle to Shell Weld 84.4% Automated UT scanning was performed from the vessel OD (Flange-Type Nozzle) surface. Automated scanning was restricted due to the nozzle configuration . Automated and Manual UT composite coverage =

84.4%.

1B11\1 N3B Main Steam Nozzle to Shell Weld 84.2% Automated UT scanning was performed from the vessel OD (Flange-Type Nozzle) surface. Automated scanning was restricted due to the nozzle configuration .

1B11\1N3C Main Steam Nozzle to Shell Weld 40.9% Automated UT scanning was performed from the vessel OD (Barrel-Type Nozzle) surface and was restricted due to the nozzle configuration .

ISI-RR-16 Version 1.0 Page 3 of8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 TABLE RR-16-1 Weld Number Description Coverage Basis for Limited Coverage 1B11\1 N4B Feedwater Nozzle to Shell Weld 39.2% Automated UT scanning was performed from the vessel OD (Barrel-Type Nozzle) surface. Automated scanning was restricted due to the nozzle configuration and N9 Nozzle. The Manual UT was not restricted.

1B11\1N4C Feedwater Nozzle to Shell Weld 31 .9% Automated UT scanning was performed from the vessel 00 (Barrel-Type Nozzle) surface. Automated and Manual scanning was restricted due to the nozzle configuration and N 11 B Nozzle. The Manual UT was restricted due to the proximity of the N11 B Nozzle.

1B11\1 N4D Feedwater Nozzle to Shell Weld 35.7% Automated UT scanning was performed from the vessel 00 (Barrel-Type Nozzle) surface. Automated scanning was restricted due to the nozzle configuration.

1B11\1 N5A Core Spray Nozzle to Shell Weld 33.3% Automated scanning was performed from the vessel OD surface (Barrel-Type Nozzle) and was limited due to the nozzle configuration and the proximity of an insulation support ring .

1B11\1N5B Core Spray Nozzle to Shell Weld 37.3% Automated scanning was performed from the vessel 00 surface.

(Barrel-Type Nozzle) Automated and Manual scanning was restricted due to the nozzle configuration and insulation support ring.

1B11\1N9 Control Rod Drive Nozzle to Shell Weld 57.5% Manual UT scanning was performed from the vessel 00 surface.

(Barrel-Type Nozzle) Scanning was restricted due to the nozzle configuration and the proximity of the insulation support ring.

ISI-RR-16 Version 1.0 Page 4 ofB

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 TABLE RR-16-2 Weld Number Description Coverage Basis for Limited Coverage 2811\2N1A Recirculation Outlet Nozzle to Shell Weld 87% Automated UT scanning was performed from the vessel OD (Flange-Type Nozzle) surface and was restricted due to the nozzle configuration.

2811\2N2C Recirculation Inlet Nozzle to Shell Weld 85.6% Automated UT scanning was performed from the vessel OD 2811\2N2E (Flange-Type Nozzle) surface and was restricted due to the nozzle configuration.

2811\2N2H 2811\2N4A Feedwater Nozzle to Shell Weld 89% Automated UT scanning was performed from the vessel OD 2811\2N4C (Flange-Type Nozzle) surface and was limited due to the nozzle configuration and the proximity of the 2N 11 A/2N 11 B instrumentation nozzles.

ISI-RR-16 Version 1.0 Page 5 of 8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 Hatch 1 N1 Nozzle- Recirculation Outlet 70°T 145?60"T Typical Flange-Type Nozzle for Hatch-1 Figure RR-16-1 ISI-RR-16 Version 1.0 Page 6 of8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 Hatch 1 N2 Nozzle - Recirculation Inlet

/

I 45"T 60"T I

70"T Typical Barrel-Type Nozzle for Hatch-1 Figure RR-16-2 ISI-RR-16 Version 1.0 Page 7 of 8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1&2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-16 TYPICAL COVERAGE RESTRICTION FOR HATCI+1 RECIRCULATION NOZZLES (N1 & N2)

DUE TO THE INSULATION SUPPORT RING

' Reactor Pressure Vessel Restricted Area ~Insulation support ring N1 OR N2 NOZZLE Figure RR-16-3 ISI-RR-16 Version 1.0 Page 8 of 8

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-17 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category 8-A, Item Number 81.30 reactor vessel shell-to-flange weld.

2811/2C-1 -Low Alloy Steel-Inspected during the 2R23 outage in Spring 2015.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code Requirements Examination Category 8-A, Table IW8-2500-1 of the 2001 Edition with 2003 Addenda of the ASME Section XI Code requires a volumetric examination be performed on this weld. The examination volume is shown in ASME Section XI Figure IW8-2500-4 and includes essentially 100% of the weld length. The examinations were performed from the outside of the reactor vessel using both automated and manual examination procedures, personnel, and equipment qualified in accordance with Appendix VIII, Supplements 4 and 6, as amended by the conditions set forth in 10 CFR 50.55a. The use of Appendix VIII was allowed by alternative ISI-ALT-01, which was approved by NRC safety evaluation dated January 3, 2006 (NRC ADAMS Accession Number ML053470091).

4. Impracticality of Compliance The ultrasonic examinations could only be performed from the shell side of the weld because of the weld and flange taper as shown in the attached figures. Ultrasonic examinations from tapered surfaces have not been qualified by the Performance Demonstration Initiative (PDf).

The weld was examined to the maximum extent possible using a combination of automated and manual scanning. The manual scanning was limited to those areas where the automated scanning had limitations (e.g ., at the thermocouple pads) .

The composite ultrasonic examination coverage was calculated as 50.6% by examination personnel. In addition to the flange taper, the examinations were limited due to three thermocouple pads on the shell side of the weld. Figures RR-17 -1 and -2 are attached to provide details on the configuration and examination results.

ISI-RR-17 Version 1.0 Page 1 of4

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-17

5. Burden Caused by Compliance To appreciably increase the examination volume coverage of weld 2C-1 would require a redesign of the RPV flange, which would be an undue burden. Therefore, it is concluded that the ASME Code requirement is impractical.

6. Proposed Alternative and Basis for Use A significant volume of the weld was examined and no unacceptable indications were found .

Coverage for circumferential flaws originating at the inside surface or middle of the examination volume was in excess of 90%. Additionally, the reactor vessel vertical welds and the reactor vessel accessible bottom head welds were examined using Appendix VIII techniques without any unacceptable indications; therefore, it is unlikely that any pattern of degradation exists in the reactor vessel that has gone undetected. This examination coverage along with VT-2 examinations associated with the Class 1 leakage test performed each refueling outage provide reasonable assurance that unacceptable flaws have not developed in the subject weld or that they will be detected and repaired prior to the return of service. During operation, Hatch-2 leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Section 5.2.7). Thus, an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examinations in lieu of the Code requirement. Therefore, relief should be granted per 10 CFR 50.55a(g}(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents This is the first examination of this weld performed per Appendix VIII on Hatch-2.

This same weld was examined on Unit-1 with similar-type limitations. Relief Request ISI-RR-02 per NL-10-0989, dated July 8, 2010 was submitted to the NRC.

9. References ML 1164A 13 is the NRC ADAMS Accession Number for the NRC Safety Evaluation, dated July 15, 2011 for ISI-RR-02.

ISI-RR-17 Version 1.0 Page 2 of4

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-17 Figure RR-17 -1 ISI-RR-17 Version 1.0 Page 3 of4

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-17 "lbmnloc.alpllt Figure RR-17-2 ISI-RR-17 Version 1.0 Page 4 of4

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-18 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(ili)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category 8-J, Item 89.11, Dissimilar Metal Pressure Retaining Welds in Piping NPS 4 or Larger Circumferential Welds, 2E11-1RHRM-24A-10 Stainless Steel Pipe to Carbon Steel Elbow Weld -Inspected during the 2R23 Outage in Spring 2015 and 2E11-1RHRM-24B-10 Stainless Steel Pipe to Carbon Steel Elbow Weld -Inspected during the 2R23 Outage in Spring 2015.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code RequirementsSection XI, Table IWB-2500-1, Examination Category 8-J, Item 89.11 requires that essentially 100% of the weld length be examined by the volumetric and surface methods.

ASME Code Case N-460, as an alternative for use by the NRC RG 1.147, Revision 17, states that a reduction in examination coverage due to part geometry or interference for the ASME Class 1 or 2 weld is acceptable provided that the reduction is less than 10%, i.e.,

greater than 90% examination coverage is obtained.

ASME Code Case N-663, as an alternative for use by the NRC RG 1.147, Revision 17, states that in lieu of the surface examination requirements for the piping welds of Examination Category 8-F (NPS 4 and larger), 8-J (NPS 4 and larger), C-F-1, and C-F-2, surface examinations may be limited to areas identified by the Owner as susceptible to outside surface attack.

The requirement for a surface examination was addressed by an N-663 evaluation which determined that both Hatch units had no locations that were susceptible to this cracking.

Therefore, surface examinations of the welds in this relief request were not performed.

4. Impracticality of Compliance The examination coverage was limited due to taper created by the difference in the outside diameter of the elbow versus the outside diameter of the pipe (Figure 1). Based on the ISI-RR-18 Version 1.0 Page 1 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-18 configuration of the weld, it is impractical to appreciably increase the coverage. Composite Code coverage was 87% . Examinations were performed as follows:

Circumferentially-oriented flaw coverage was obtained from both the upstream and downstream sides of the weld using a 45°/60° Refracted Longitudinal Wave (RL) with 100% Code coverage (Figure 1).

The base material was examined using a 45° shear wave with 100% coverage (Figure 2).

Axially-oriented flaw coverage was obtained from both the downstream side and the upstream side of the weld using a 45° RL with 61% Code (Figure 3).

5. Burden Caused by Compliance Compliance would require the replacement of the elbows with new components fabricated with a special design to allow examination.

6. Proposed Alternative and Basis for Use Although the ultrasonic examination was limited for axially-oriented flaws, the circumferential flaw coverage of the weld joint was scanned from both sides with both the 45° RL transducer and the 60° RL transducer for 100% coverage. This coverage provides assurance that unacceptable flaws have not developed in the subject welds or that they will be detected and repaired prior to the return of service. Therefore, based on the UT examination of the subject areas to the maximum extent practical, there is reasonable assurance of the structural integrity and safety of the welds because the information and data obtained from the volume examined provided sufficient information to judge the overall integrity of the weld.

Furthermore, a VT -2 visual examination of these welds is performed each refueling outage as part of the Class 1 leakage test. During operation, Hatch-2 leakage can be determined by the leakage detection system (LOS) located in the drywell (as described in the Hatch FSAR Section 5.2.7). Thus, an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examinations in lieu of the Code requirement. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents A similar relief request (RR-55) was submitted for the third lSI Interval related to the examination of these two 24-inch RHR welds per NL-06-1159 on July 10, 2006. In addition, a relief request for the Unit-2 20-inch RHR weld with an identical configuration (elbow to pipe weld) was submitted per NL-10-0989 on July 8, 2010.

ISI-RR-18 Version 1.0 Page 2 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-18

9. References The ADAMS Accession Number for the NRC SER, dated July 20, 2007 for the third lSI Interval for these two 24-inch RHR welds is ML071830010. The NRC SER, dated for the fourth lSI Interval for the 20-inch RHR weld has an ADAMS Accession Number of ML11164A133.

ISI-RR-18 Version 1.0 Page 3 of 6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-18 PHASED AIIRAY UI.'DIASOMC EXAMINATION CPVERAGE ASP'SSMPfT J.DO:I(, ol required cowar;ee ob[:ll nad In me :.xl:ll sun d iNctiGn of weld :l1:1.1-1RH~M -:Z4A-30.

Weld :ZE11- lllHRM*24A-lD 1.()---

\-

,... \ITcoo U..lt o...tq '

............ ,1-0oo...-t Elb<aw !lldto CZJ\II!f'aiiiE! bile.

Figure RR-18-1 ISI-RR-18 Version 1.0 Page 4 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-18 Wllld 2EU-JRHIIM-24A-10 z.tco:U...,.td ' "' c.lc.-fl.2M F.- IMw u... ~ l 1-:>a.-...,. l l.oo...- 1 1-o-.....,.t~-- a r-- -L.ooot~ l-rn.te-.--. 1 Weld 2EU-1RHRM*2AA*10 z.oaa..._*

r.lo VIew ""'" a....,

c.&c.-u..

?

H)u-....,ji~Oc>W- 1 l-o-..., j 2-o-._ j r-- -~3oolrrr~ -..r~a j --.1 Pipe Side~ lladl Figure RR-18-2 ISI-RR-18 Version 1.0 Page 5 of 6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-18 2E11-1RHRM-Z4A-10 0..15 11\

1- l.OO in ... ,

I ., 0.90 sq in tequited ASM£ covetage I ~ )_ I 0.59 "~ '" achieved As depicted in the illustration above, 65.5% of the Code-required examination coverage for axial flaws was achieved due to the inability to scan on the weld. Examination personnel also documented that 81.0% of the wetted surface was examined for the presence of axial flaws.

Coverage obtained for circumferential flaws was 100%.

Combined Code coverage was 82.5%.

Figure RR-18-3 ISI-RR-18 Version 1.0 Page 6 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 1, ASME Section XI Category 8-J, Item 89.11, Pressure Retaining Welds in Piping NPS 4 or Larger Circumferential Welds.

A total of nine welds are addressed in this relief request. Of these nine welds, all are austenitic piping welds except for Hatch-1 piping weld 1E21-1 CS-1 0-7 which is ferritic material. The nine piping welds with limited examination apply to both Hatch units and are shown on Table-RR-19-1 for Unit-1 and Table-RR-19-2 for Unit-2.

In the fourth lSI Interval, Hatch elected to implement BWRVIP-75 for the examination of austenitic piping welds which were previously examined to the commitments of Generic Letter 88-01, "NRC Position on lntergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," dated January 25, 1988 and NUREG-0313, Revision 2, "Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping," dated January 1988. BWRVIP-75 provides revisions to the scope and frequencies of inspections for austenitic piping welds for Categories A through E.

The eight austenitic piping welds shown in Tables RR-19-1 and RR-19-2 are shown as either BWRVIP Category A or C.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code RequirementsSection XI, Table IWB-2500-1, Examination Category 8-J, Item 89.11 requires that essentially 100% of the weld length be examined by the volumetric and surface methods.

ASME Code Case N-460, as an alternative for use by the NRC RG 1.147, Revision 17, states that a reduction in examination coverage due to part geometry or interference for the ASME Class 1 or 2 weld is acceptable provided that the reduction is less than 10%, i.e.,

greater than 90% examination coverage is obtained.

ASME Code Case N-663, as an alternative for use by the NRC RG 1.147, Revision 17, states that in lieu of the surface examination requirements for the piping welds of Examination Category 8-F (NPS 4 and larger), 8-J (NPS 4 and larger), C-F-1, and C-F-2, ISI-RR-19 Version 1.0 Page 1 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 surface examinations may be limited to areas identified by the Owner as susceptible to outside surface attack.

The requirement for a surface examination was addressed by an N-663 evaluation which determined that both Hatch units had no locations that were susceptible to this cracking.

Therefore, surface examinations of the welds in this relief request were not performed.

4. Impracticality of Compliance The examination limitations for the above welds are due to the design of components (e.g.,

penetration, valve, sweepolet, etc.) which restricts the access for the ultrasonic (UT) examinations shown in Tables RR-19-1 and RR-19-2. With the exception of the Unit-1 Core Spray piping weld 1E21-1 CS-1 OB-20A, the examinations are primarily a one-side examination from either the pipe, elbow, manifold, or cap side of the weld due to configuration. For Core Spray weld 1E21-1CS-108-20A, the weld description is Safe-End Extension to Safe-End with limitations on the upstream side due to the taper of the Safe-End Extension. In these nine cases, it would be impractical to increase the coverage.

5. Burden Caused by Compliance Compliance would require the replacement of the existing components with new components fabricated with a special design to allow examination.

6. Proposed Alternative and Basis for Use Per the NRC staff position found in Generic Letter 88-01, Category A welds are those with no known cracks and have a low probability of experiencing IGSCC. The one ferritic weld also has a low probability of cracking.

Category C welds are considered susceptible to IGSCC but have been mitigated by stress improvement after more than two cycles of operation. Each weld was stress improved using either the induction heating stress improvement (IHSI) or the mechanical stress improvement (MSIP). Of the eight austenitic welds, six are protected by effective hydrogen water chemistry. The UT examination performed from at least one side of the weld in conjunction with the stress improvement and the hydrogen protections should provide reasonable assurance that unacceptable flaws have not developed in the subject weld or that they will be detected and repaired prior to the return of service. The two welds without effective hydrogen water chemistry (1 831-1 RC-4A-10A and 1E21-1CS-108-20A) are being examined per the requirements of BWRVIP-75. All nine piping welds were examined to the maximum extent practical with no unacceptable indications recorded.

Therefore, based on the UT examination of the subject areas to the maximum extent practical, there is reasonable assurance of the structural integrity and safety of the welds because the information and data obtained from the volume examined provided sufficient information to judge the overall integrity of the welds.

Furthermore, a VT-2 visual examination on the subject welds are performed each refueling outage as part of the leakage test. During operation leakage can be determined by the ISI-RR-19 Version 1.0 Page 2 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 leakage detection system (LOS) located in the Drywell. The LOS is described in HNP-1 FSAR Section 4.10 and HNP-2 FSAR Section 4.1 0. Based on the above information, relief should be granted per 10CFR50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the 4th lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents Similar relief requests (RR-52; RR-53; and RR-56) were submitted for the third lnservice Inspection Interval and accepted by the NRC.

9. References The ADAMS Accession Number for the NRC Safety Evaluation, dated July 20, 2007 was ML071830010.

ISI-RR-19 Version 1.0 Page 3 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Table RR-19-1 Hatch-1 Weld Number I BWRVIP Description I Exam Outage Coverage Basis for Limited Coverage Category No.

Coverage was limited to a one-sided examination due to the proximity of the sweepolet taper to the weld (Figure 1). Circumferential flaw Branch Connection to Cap I coverage was obtained using 60° shear waves and 60° Refracted 1B31-1 RC-4A-10A I C 50%

1R26- 2014 Longitudinal Waves. Axial flaw coverage was limited to the pipe side.

Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved in 1985 using the lHSl process.

Coverage was limited to a one-sided examination due to the proximity of the sweepolet taper to the weld (Figure 2). Circumferential flaw Branch Connection to Pipe I coverage was obtained using 60° shear waves and 60° Refracted 1B31-1RC-12BR-E-11 C 50%

1R26- 2014 Longitudinal Waves. Axial flaw coverage was limited to the pipe side.

Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved in 1985 using the lHSl process.

Coverage was limited to a one-sided examination due to the proximity of the valve taper to the weld (Figure 3). Coverage was also reduced in 1E21-1 CS-1 OA-7 Valve to Elbow /1R26- 2014 65% the area on the inside radius (intradose) of the elbow. Circumferential (Ferritic Weld) flaw coverage was obtained using 60° shear waves. Axial flaw coverage was limited to the elbow side.

Scanning for circumferential flaws on the upstream side (Safe-End Extension) was not possible due to the OD taper. No limitations were Safe-End Extension to Safe-End 1E21-1CS-10B-20A I C 75% noted for axial flaws. Examinations were performed to the maximum I 1R22A - 2006 extent possible (Figure 4). This weld was stress improved in 1993 using the MSlP process.

Coverage was limited to a one-sided examination due to the proximity of the valve taper to the weld (Figure 5). Circumferential flaw coverage 1G31-1RWCUM-6-D-51 was obtained using 60° shear waves and 60° Refracted Longitudinal Valve to Elbow /1R26- 2014 50%

A Waves. Axial flaw coverage was limited to the elbow side. Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved in 1993 usingthe MSlP process.

lSl-RR-19 Version 1.0 Page 4 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Table RR-19-1 Hatch-1 Weld Number I BWRVIP Description I Exam Outage Coverage Basis for Limited Coverage Cate_g_o_ry No.

Coverage was limited to a one-sided examination due to the proximity of the penetration taper to the weld (no Figure). Circumferential flaw 1G31-1RWCUM-6-D-16/ Penetration to Pipe I 1R22A - coverage was obtained using 60° shear waves and 60° Refracted 50%

Table 6 2006 Longitudinal Waves. Axial flaw coverage was limited to the elbow side.

Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved in 1993 using the MSIP process.

ISI-RR-19 Version 1.0 Page 5 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Table RR-19-2 Hatch-2 Weld Number I BWRVIP Description I Exam Outage Coverage Basis for Limited Coverage Category No.

Coverage was limited to a one-sided examination due to the proximity of the penetration taper to the weld (Figure 6). Circumferential flaw Cross to Manifold I 2R20 - coverage was obtained using 45° and 70° shear waves. Axial flaw 2831-1RCM-22A-11 A 65%

2009 coverage was limited to the pipe side. Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved during the 1984 outa_g_e using the IHSI process.

Coverage was limited to a one sided examination due to the proximity of the valve taper to the weld (Figure 7). Circumferential flaw coverage was obtained by the use of 45° and 70° shear waves. Axial flaw 2831-1 RCM-28AD-5/ A Pipe to Cross 12R23 - 2015 50%

coverage was limited to the pipe side. Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved during the 1984 outage using the IHSI process.

Coverage was limited to a one sided examination due to the proximity of the valve taper to the weld (Figure 8). Circumferential flaw coverage was obtained by the use of 45° and 67° shear waves for coverage of 2831-1RCM-28BD-5/ A Pipe to Cross I 2R23 - 2015 50% 50%. Axial flaw coverage was limited to the penetration side. Per the requirements of 10 CFR 50.55a(b)(2)(xv)(A)(2) coverage is 50%. This weld was stress improved during the 1984 outage using the IHSI process.

ISI-RR-19 Version 1.0 Page 6 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Figure 1 - Illustration of the Weld Geometry for 1831*1 RC-4A*1 OA CAP Figure 2 -Illustration of the Code Coverage Obtained for 1831-1 RC-128R-E-1 FLC)W .._

t ISI-RR-19 Version 1.0 Page 7 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Figure 3 - Illustration of the Code Coverage Obtained for 1E21-1 CS-1 OA-7 flOW'!;:,.

I

t. ~ I, v~l~e

-~--~zu~~~-------

\ : l"v I //

~----,n -- -, ~

........... . ...L------ :*- ..... ---** *--- - -

Figure 4 -Illustration of the Code Coverage Obtained for 1E21-1CS-108-20A SJ£AR VAV£ COVERAGE RL COVERAGE:

ISI-RR-19 Version 1.0 Page 8 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Figure 5- Illustration of the Code Coverage Obtained for 1831-1 RWCUM-6-D-5 Vf\Lvf_ fLOW

_j**

Figure 6 - Illustration of the Code Coverage Obtained for 2831-1 RCM-22A-1 FLOW

\ r-~--

-*~*--MAN__IFOLD

, I CROSS REQUIRED CODE COVERAGE CODE COVERAGE OBTAINED = 65%

ISI-RR-19 Version 1.0 Page 9 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 and 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-19 Figure 7- Illustration of the Code Coverage Obtained for 2831-1 RCM-28AD-5 PrP£.. TEE.

l o" Figure 8 -Illustration of the Code Coverage Obtained for 2831-1 RCM-2880-5 F/ov)

/

ISI-RR-19 Version 1.0 Page 10 of 10

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-21 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 2, ASME Section XI, Code Category C-A, Item Numbers C1.10, shell circumferential welds and C1.20, head circumferential welds for Residual Heat Removal (RHR) Heat Exchanger 2E11-2HX-A.

The specific welds are:

Item Number C 1.1 0:

2E11-2HX-A-2 Upper Shell Ring to Lower Shell Ring Examined during 2R23; Spring 2015 2E11-2HX-A-3 Lower Shell Ring to Flange Examined during 2R 19; Spring 2009 Item Number C1.20:

2E11-2HX-A-1 Shell Head to Upper Shell Ring Examined during 2R20; Spring 2011

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code Requirements Table IWB-2500-1, Examination Category C-A, Item Numbers C1.10 and C1.20 require examination per Figure IWC-2500-1.

4. Impracticality of Compliance The limitations are described in Table RR-21-1. Each weld had specific configurations which limited the examination. It would be impractical to appreciably increase the coverage obtained.

5. Burden Caused by Compliance Compliance would require replacement of the existing heat exchanger with a new heat exchanger fabricated with a special design to allow examination.

ISI-RR-21 Version 1.0 Page 1 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-21

6. Proposed Alternative and Basis for Use A significant percentage of these three heat exchanger welds was examined and no unacceptable indications were found. These examinations and the ongoing leakage tests once every lSI Period provide reasonable assurance that unacceptable flaws have not developed in these welds or that they will be detected and repaired prior to the return of service. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i) .

7. Duration of Proposed Alternative The proposed relief request is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31,2015.

8. Precedents The three RHR heat exchanger welds were examined in the third lSI Interval with limitations. Weld A-1 limitations were documented by RR-5 in a submittal to the NRC on October 17, 1995. The limitations for welds A-2 (RR-62) and A-3 (RR-60) were documented in a submittal on July 10, 2006 (ADAMS Accession Number ML061910425).

9. References The NRC SER for RR-5 was issued on June 16, 1997 (TAC Numbers M93918 and M93919) while RR-60 and RR-62 were approved by the NRC on June 5, 2007 (ADAMS Accession Number ML071360297).

ISI-RR-21 Version 1.0 Page 2 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-21 Table ISI-RR-21-1 Weld No. Description Coverage Basis for Limited Coverage 2HX-A-1 Shell Head to Upper Shell Ring 40% Obstruction due to lugs.

2HX-A-2 Upper Shell Ring to Lower Shell Ring 73% Limitation due to support brackets.

2HX-A-3 Lower Shell Ring to Flange 85% Limitation due to component configuration.

ISI-RR-21 Version 1.0 Page 3 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-21

't~

lSo.. Of IHUD t6J"""" ~IL UQUIIt.D ellAM 101... o&~cr.f!.b t>uE..- t"t> LAJ~ *

11. I 1!,c. 4-o-1..

"! CA"ti.AIA! D*B f>>u..,~b Weld No. 2E11-2HX-A-1 FIGURE RR-21-1 ISI-RR-21 Version 1.0 Page 4 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-21

~<---- s".1s "cl.-o..--->~*

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~~--2. s- ~----91 Weld No. 2E11-2HX-A-2 Figure RR-21-2 ISI-RR-21 Version 1.0 Page 5 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-21 FJo.M~~

~~_,..,;~~tDt-4 C::::O._,C.,..Cl.:J~ ~~~ c...-t~" C.<!)~

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.3=t>~-* .. ~.:J, /a& 3-7-o=t Weld No. 2E11-2HX-A-3 Figure RR-21-3 ISI-RR-21 Version 1.0 Page 6 of6

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-22 Relief Request In Accordance with 10 CFR 50.55a(g)(5) (iii)

--lnservice Inspection Impracticality--

(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i) .)

1. ASME Code Component(s) Affected Class 2, ASME Section XI Code Category C-8, Item C2.21, nozzle to shell examinations for Residual Heat Removal (RHR) Heat Exchanger welds listed in Tables RR-22-1 and RR-22-2, plus Figures RR-22-1 through 3.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code RequirementsSection XI, Table IWC-2500-1, Examination Category C-8, Item C2.21 requires that essentially 100%

of the weld length be examined by the volumetric and surface methods per Figure IWC-2500-4.

ASME Code Case N-460, as an alternative for use by the NRC RG 1.147, Revision 17, states that a reduction in examination coverage due to part geometry or interference for the ASME Class 1 or 2 weld is acceptable provided that the reduction is less than 10%, i.e., greater than 90% examination coverage is obtained .

4. Impracticality of Compliance The ultrasonic (UT) examination technique was performed to satisfy ASME Section XI requirements.

In addition, a magnetic particle (MT) examination was performed to address the surface examination requirements.

In the examination of the Hatch-1 weld 1E 11-2HX-A-l during the 1R23 Spring 2008 outage, two UT indications were recorded. The indications were evaluated to ASME Section XI and were found to be unacceptable. Hatch had Structural Integrity Associates perform a fracture mechanics and fatigue crack growth evaluation. The results of that analysis demonstrate that crack growth for the remaining life of the plant is minimal and that the flaw is acceptable since the calculated maximum allowable flaw size is much larger than the existing flaw size including growth until the end of plant life. This weld was examined two more times during the fourth lSI Interval and no changes were noted. In addition, the nozzle weld on the Hatch-1 "8" Loop heat exchanger was examined during the Spring 2008 outage with no indications noted. The scope expansion of this weld was performed to address IW8-2430 of the 2001 Edition through the 2003 Addenda.

ISI-RR-22 Version 1.0 Page 1 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-22 Tables RR-22-1 and -2 show the results of the UT examinations performed on these three welds.

Figures RR-22-2 and RR-22-3 provide additional details for Hatch-1 and -2 RHR heat exchanger nozzles, respectively. Coverage varied from 58% to 85% of these welds. Due to the configuration, there was no scanning from the nozzle side and scans for axially-oriented flaws were limited due to the weld configuration. Greater than 90 percent coverage was obtained for circumferentially-oriented cracking from the shell side. These welds were examined to the maximum extent possible and it would be impractical to appreciably increase the coverage.

5. Burden Caused by Compliance Compliance would require replacement of the existing heat exchanger with a new heat exchanger fabricated with a special design to allow examination.

6. Proposed Alternative and Basis for Use A significant percentage of the RHR heat exchanger nozzle to shell welds were examined during the fourth lSI Interval. These examinations and the ongoing leakage tests once every lSI Period provide reasonable assurance that structural integrity is being maintained; therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed relief request is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents Two similar relief requests were submitted for these welds during the third lSI Interval for Hatch.

Relief Requests RR-58 and RR-59 were submitted to the NRC by NL-06-1159, dated July 10, 2006.

9. References The ADAMS Accession Number for the NRC Safety Evaluation, dated June 5, 2007, is M L071360297.

ISI-RR-22 Version 1.0 Page 2 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-22 Table ISI-RR-22-1 Weld No. Description I Exam Outage Coverage Basis for Limited Coverage Inlet Nozzle to RHR Hx Shell/

1E11-2HX-A-I 58% Limitation due to nozzle configuration as shown on Figure 2.

1R23- 2008 RHR Hx Shell to Outlet Nozzle I 1E11-2HX-B-O 68% Limitation due to nozzle configuration as shown on Figure 2.

1R25- 2010 - '--- - - - - - --- - I Table ISI-RR-22-2 Weld No. Description I Exam Outage Coverage Basis for Limited Coverage I RHR Hx Shell to Outlet Nozzle I 2E11-2HX-A-O 85% Limitation due to nozzle configuration as shown on Figure 3.

2R20- 2009 i ISI-RR-22 Version 1.0 Page 3 of5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-22 118' FLOOR. GRATlt.'G

~SSEL TrE DOWN i!RMJ<Ei-(4) AT oa,

'QIO ,. lSO(' 2,01 I

NOTE: 0 2 REFEREf~CE lS OVTI.ET t.IOZZLE CEHTER UNE AZIMUTH REFER£t.ICE 15 CLOCKWISE LOOKING DOWN RT 1E11-1HX-A(B)-O

' '-..JL---------.J I ~/

Figure 1 - Hatch-1 RHR Heat Exchanger Layout (Typical for Both Hatch Units)

ISI-RR-22 Version 1.0 Page 4 of5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-22 These detail drawings show the layout of these welds. Differences exist between the units and a drawing is included for each unit. The inlet nozzle is located on the heat exchanger head while the outlet nozzle is located on the lower shell. In general, the scans looking for circumferential flaws from either the heat exchanger head or the shell can be performed while the scans looking for axial flaws on the weld are limited because of the weld contours .

Weld detail taken from detail drawing- illustrative only Nozzle 45 degree

/

Shell

~~ .49 square inch examination volume

$~ .49 square inches examined (1000.4) in the .axial scan.~tion

.. [D .08 sq\W"C incbes examined (16%) in the c:tre. scan directaon

Examined 58% of required volume Scale: 50%

Figure 2 - Hatch-1 RHR Heat Exchanger Nozzle Weld Coverage Figure 3 - Hatch-2 RHR Heat Exchanger Nozzle Weld Coverage ISI-RR-22 Version 1.0 Page 5 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 and Unit-2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-23 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(ili)

--lnservice Inspection Impracticality-(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 2, ASME Section XI Category C-F-2, Item C5.51, Circumferential Pressure Retaining Welds in Carbon or Low Alloy Steel Piping:

Hatch-1 welds are shown in Table RR-23-1 and Hatch-2 welds are shown in Table RR-23-2.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code RequirementsSection XI, Table IWC-2500-1, Examination Category C-F-2, Item C5.51 requires that 100%

of each weld requiring examination receive a surface and volumetric exam.

ASME Code Case N-460, as an alternative for use by the NRC RG 1.147, Revision 17, states that a reduction in examination coverage due to part geometry or interference for the ASME Class 1 or 2 weld is acceptable provided that the reduction is less than 10%, i.e.,

greater than 90% examination coverage is obtained.

ASME Code Case N-663, as an alternative for use by the NRC RG 1.147, Revision 17, states that in lieu of the surface examination requirements for the piping welds of Examination Category 8-F (NPS 4 and larger), 8-J (NPS 4 and larger), C-F-1, and C-F-2, surface examinations may be limited to areas identified by the Owner as susceptible to outside surface attack.

The requirement for a surface examination was addressed by an N-663 evaluation which determined that both Hatch units had no locations that were susceptible to this cracking.

Therefore, surface examinations of the welds in this relief request were not performed.

4. Impracticality of Compliance Coverage was limited on these piping welds based on the information provided in the two tables and Figures 1 through 4. Scans for axial flaws were not required for these ferritic welds. Scans for circumferential flaws were performed to the maximum extent practical.

ISI-RR-23 Version 1.0 Page 1 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 and Unit-2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-23

5. Burden Caused by Compliance Increasing the coverage would require the re-design of these welds and components to allow access.

6. Proposed Alternative and Basis for Use A volumetric examination was performed on these welds with results as noted on the attached tables. In addition, VT-2 visual examinations associated with the Class 2 leakage test are performed each inspection period for these welds. These examinations and the ongoing leakage tests once every lSI Period provide reasonable assurance that unacceptable flaws have not developed in these welds or that they will be detected and repaired prior to the return of service. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31,2015.

8. Precedents None; this is the first examination of these welds performed per Appendix VIII of ASME Section XI.

9. References None ISI-RR-23 Version 1.0 Page 2 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 and Unit-2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-23 Table RR-23-1 Hatch-1 Weld Number Description Coverage Basis for Limited Coverage 1E11-2RHR-10B-SWDS-4 Elbow to Pipe /1 R25- 2013 85.7% A welded attachment limited the examination of this weld (Figure 2).

No examination was performed from the torus pipe side due to the 1E21-2CS-16A-TS-2 Pipe to Elbow /1 R27- 2016 70%

torus configuration (Figure 3j_. - -

Table RR-23-2 Hatch-2 Weld Number I Description I Coverage I Basis for Limited Coverage 2E11-2RHR-6B-TSP-5 I 45° Elbow to Pipe /2R21 - 2011 I 78.9% I A restraint limited the examination of this weld (Figure 4). - - -

ISI-RR-23 Version 1.0 Page 3 of5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 1 and Unit-2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-23 2E41-2HPCI-16-TS-18 FIDW ELBOW CODE REQUIRED VOLUME= 0.2 SQ. INCHES LIMITED DUE TO WELD CROWN 0.04 SQ. INCHES 0.04 DIVIDED BY 0.2= 0.2 OR 20%

TOTAL VOLUME EXAMINED 0.16 SQ. INCHES OR 80%

Figure 1 - Illustration of the Welded Crown, Tapered Tee, and Examination Angles

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1-Ill )

~

0"

1;114, ,;

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-=--"-~~-------=- FLOw I"'

Figure 2 -1E11-2RHR-108-SWDS-4 ISI-RR-23 Version 1.0 Page4 of5

Southern Nuclear Operating Comp~ny Plant Edwin I. Hatch, Unit 1 and Umt-2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-23

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Figure 3 -1E21-2CS-16A-TS-2 PIPE S,,, ~ 1"'otAL. ~4t>tll.IED ~'tAM vowM£.

.1~) o~ !.tAM VOWME Nof" E*1!1\Mitoi.Eb bu£.. '1b RU.11ZN...rr' fa.oM 15" 1b ,q ~

Figure 4- 2E11-2RHR-6B-TSP-5 ISI-RR-23 Version 1.0 Page 5 of 5

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-24 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality-(Note: Licensees request under 10 CFR 50.55a(g)(5)(iii).

The NRC grants under 10 CFR 50.55a(g)(6)(i).)

1. ASME Code Component(s) Affected Class 2, ASME Section XI Category C-G, Item C6.1 0, Pressure Retaining Welds in Pump Casings:

2E21-2CS-POP-A -2 Core Spray (CS) Circumferential Pump A Outlet Nozzle Elbow to Flange Weld- Carbon Steel- Inspected during the 2R21 Outage in Spring 2011.

2. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

3. Applicable Code RequirementsSection XI, Table IWC-2500-1, Examination Category C-G, Item C6.10 requires that 100%

welds in all components in each piping run examined under Examination Category C-F receive a Surface Examination. In case of multiple pumps of similar design, size, function, and service in a system, required weld examinations may be limited to all the welds in one pump in the same group or distributed among any of the pumps of the same group.

ASME Code Case N-460, as an alternative for use by the NRC RG 1.147, Revision 17, states that a reduction in examination coverage due to part geometry or interference for the ASME Class 1 or 2 weld is acceptable provided that the reduction is less that 10%, i.e.,

greater than 90% examination coverage is obtained.

4. Impracticality of Compliance Coverage was limited for the Surface Examinations due to the proximity of the support structure for the pump motor to the weld and the flange bolting on the pump casing (Figure 1), 2E21-2CS-POP-A-2. Surface Examination coverage was calculated to be 50%. The yoke was limited to one direction due to configuration. The magnetization direction that was utilized was parallel to the weld and would detect axial surface flaws.

5. Burden Caused by Compliance Increasing the surface examination coverage would require a redesign of the pump flange and pump motor support and room to allow for additional space around the weld.

ISI-RR-24 Version 1.0 Page 1 of 3

Southern Nuclear Operating Company Plant Edwin f. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-24

6. Proposed Alternative and Basis for Use SNC performs a leakage test each lSI Period. Therefore, based on the surface examination performed to the maximum extent practical and the leakage test of the subject areas, there is reasonable assurance of the structural integrity and safety of the welds because the information and data obtained from the area examined provided sufficient information to judge the overall integrity of the welds. Therefore, relief should be granted per 10 CFR 50.55a(g)(6)(i).

7. Duration of Proposed Alternative The proposed alternative is applicable for the fourth lnservice Inspection Interval, extending from January 1, 2006 through December 31, 2015.

8. Precedents Similar relief request RR-06 was submitted for the fourth lnservice Inspection Interval and accepted by the NRC.

9. References The ADAMS Accession Number for the NRC Safety Evaluation, dated July 15, 2011 was ML11164A133.

ISI-RR-24 Version 1.0 Page 2 of 3

Southern Nuclear Operating Company Plant Edwin I. Hatch, Unit 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number ISI-RR-24 Figure 1 -Photograph of the proximity of the support structure to the examination area.

MOTOR FLANGE SUPPORT BOLTING STRUCTURE ISI-RR-24 Version 1.0 Page 3 of3

v t e Letter Sequence Request
CAC:MF9027, (Open)
Initiation Request, Request Acceptance... Administration Questions 26 April 2017 Answers 31 May 2017 Results Approval... Other: ML17268A044
MONTHYEAR2017-10-20 [Table View]

NL-16-2707, Units 1 and 2; Fourth 10-year Interval In-service Inspection Program ISI Program Update: Notification of Impractical ASME Code Requirements (NL-16-2707)

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NL-16-2707, Units 1 and 2; Fourth 10-year Interval In-service Inspection Program ISI Program Update: Notification of Impractical ASME Code Requirements (NL-16-2707)

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Enclosures:

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