DCL-15-021, Response to NRC Request for Additional Information Regarding Relief Request RES-SI

Response to NRC Request for Additional Information Regarding Relief Request RES-SI

ML15036A606
Person / Time
Site:	Diablo Canyon
Issue date:	02/05/2015
From:	Welsch J Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PG&E Letter DCL-15-021
Download: ML15036A606 (24)

Text

Pacific Gas and Electric Company James M. Welsch Diablo Canyon Power Plant Site Vice President Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 February 5, 2015 805.545.3242 Internal: 691.3242 PG&E Letter DCL-15-021 Fax: 805.545.4884 U.S. Nuclear Regulatory Commission 10 CFR 50.55a ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant Unit 1 and Unit 2 Response to NRC Request for Additional Information Regarding Relief Request REP-SI

References:

1. PG&E Letter DCL-14-060, "ASME Section Xllnservice Inspection Program Request for Alternative REP-S I: Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded attachments," dated July 21, 2014 (ML14202A613)

2. NRC Letter, "Request for Additional Information -lnservice Inspection Request for Alternative REP-SI," dated December 23, 2014 In Reference 1, Pacific Gas and Electric Company (PG&E) submitted Relief Request for Alternative REP-S I for Diablo Canyon Power Plant Units 1 and 2 for NRC approval.

The NRC Staff provided a request for additional information (RAI) via letter dated December 23, 2014 (Reference 2). The Enclosure 1 to this letter provides PG&E's response to seven out of the ten RAI questions. PG&E will coordinate submission of a supplement to the NRC with the NRC project manager to address the remaining RAI questions.

A revised version of the Enclosure to Reference 1 is included in Enclosure 2 to this letter. The following three reports that were included in Reference 1 are current, still valid, and applicable to the revised relief request that is included in Enclosure 2:

• Weld Procedure Specification No. 149

• PG&E ATS Report 420DC-14.20: Welding Procedure Qualification Record (PQR) 771 and Associated Documents A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

Document Control Desk PG&E Letter DCL-15-021 February 5, 2015 Page 2

• SIA Report No. 1301620.402: Stress and Fracture Mechanics Evaluation of Type 410 Stainless Steel Weldments in Safety Injection Pumps at Diablo Canyon Power Plant (Revision 2)

This communication does not contain regulatory commitments (as defined by NEI 99-04).

If you have any questions or require additional information, please contact Mr. Philippe Soenen at (805) 545-6984.

Sincerely, Site Vice President RNTT/4231/50600119 Enclosures cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator Thomas R. Hipschman, NRC Senior Resident Inspector Siva P. Lingam, NRC Project Manager Gonzalo L. Perez, Branch Chief, California Department of Public Health State of California, Pressure Vessel Unit A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek

Enclosure 1 PG&E Letter DCL-15-021 PG&E Response to NRC Request for Additional Information Regarding Relief Request REP-51 NRC Question RAI-EVIB 1.0:

The relief request is based *on the premise that existing Sl pump vent and drain socket welds may be determined acceptable as-is for continued service. However, the possibility exists that all vent and drain socket welds could fail on an Sl pump . To properly evaluate the relief request, the NRC staff requests that the licensee describe:

(a) The safety significance if all vent and drain weld connections are severed on a single Sl pump.

(b) In case of failure of all vent and drain weld connections on a single Sl pump, please explain if the safety function of nearby equipment and personnel will be affected.

PG&E Response:

Failure of the subject socket welds on safety injection (SI) pump vent and drain connections are not a part of the Diablo Canyon Power Plant (DCPP) design bases calculations. The impact of postulated failures of vent and drain connections are being evaluated. PG&E will coordinate submission of a supplement to the NRC with the NRC project manager to address this request for additional information (RAI).

NRC Question RAI-EVIB 2.0:

The relief request partly bases the acceptability of existing Sl pump vent and drain socket welds on the performance of a welding procedure qualification test with representative type 41 0 stainless steel (P-6) and type 304 stainless steel (P-8) base materials using type 309 filler metal per the production welding procedure*

parameters without post-weld heat treatment, and that the qualification test assembly passed the required ASME Code destructive tests. However, for this material combination, the possibility exists that untempered martensite may be formed in the heat affected zone of the qualification weldment. The NRC staff requests that the licensee discuss the following:

(a) Whether hardness testing was performed in the P-6 heat affected zone of the qualification weldment to objectively measure the possibility and extent of 1

Enclosure 1 PG&E Letter DCL-15-021 untempered martensite formation.

(b) Since the destructive test specimens were machined down from 0.375-inch in thickness to 0.300-inch in thickness, please justify that the machined destructive test specimens are representative of the qualification weldment in terms of microstructure. Discuss whether the possibility exists for detrimental untempered martensite to be formed during welding but removed during the preparation of the destructive test specimens.

PG&E Response:

(a) Hardness testing was not performed in the heat affected zone of the qualification weldment to measure the possibility and extent of untempered martensite.

Fracture mechanics evaluations described in SIA Report No. 1301620.402, Revision 2, conservatively assume a worst-case K1c fracture toughness of 25 ksiv'in based on estimated toughness properties for untempered type 410, as reviewed in Section 3.2.5.

The 41 0 base metal of the qualification coupon was chosen to have a high carbon content, and it is probable the microstructure in the heat affected zone contains untempered martensite with properties consistent with those assumed in the analysis contained in the SIA report.

(b) The destructive test specimens were machined from the qualification test coupon as required by ASME Section IX. "Qualification of the procedure specification demonstrates the mechanical properties of the joint made using a joining process," (Reference ASME IX, Section UG-1 02). It follows that specimens machined as part of the ASME qualification process are representative of the qualification weldment.

. Two types of destructive test specimens were machined from the qualification weldment:

• Reduced section tension specimens

• Transverse face and root bend specimens The reduced section tension specimens were machined as required by Figure QW-462.1 (a) of ASME IX. Machining was performed to "obtain approximately parallel surfaces." The maximum reduction in thickness was 0.074 inch. The weld layers varied from approximately 0.125 to 0:150 inch in thickness. Therefore no weld layer was completely 2

Enclosure 1 PG&E Letter DCL-15-021 removed by machining and any untempered martensite would also not be completely removed.

The transverse face and root bend specimens were machined as required by Figure QW-462.3(a) of ASME IX. All machining was performed on what became the interior surface (i.e., intrados surface) of the bent specimens. Therefore, face bend specimens were machined on the root side of the plate, and root bend specimens were machined from the face side of the plate. As both surfaces of the welded plate are represented on the outer surface of either the face and root bends, the entire weld and heat affected zone cross-section and any untempered martensite are contained in the bend specimens.

NRC Question RAI-EPNB 1.0:

The licensee submitted the relief request under paragraph 50.55a(a)(3)(i )

of Title 10 ofthe Code of Federal Regulations (1 0 CFR). The NRC staff believes that the use of the proposed alternative (i.e., continued operation with the existing non-conforming welds) provides a lower level of quality and safety than with ASME Code compliant welds. The NRC staff suggests that it is more appropriate to request the relief pursuant to 10 CFR 50.55a(a)(3)(ii).

Please propose the alternative to the ASME Code under 10 CFR 50.55a(a)(3)(ii) or justify why it is appropriate to submit this relief request under 10 CFR 50.55a(a)(3)(i). The licensee should also provide hardship justification for not complying with the ASME Code requirements.

PG&E Response:

As suggested by the NRC staff, the enclosure to Reference 1 is revised to request the relief pursuant to 10 CFR 50.55a(z)(2), formerly 10 CFR 50.55a(a)(3)(ii). The revised relief request is included in Enclosure 2.

Hardship justification for not complying with the ASME Code requirements is also provided, in the revised relief request. Five photographs of welded attachments are included in Attachment 1 of Enclosure 2 for reference.

The following three reports that were included in Reference 1 are current, still valid, and applicable to the revised relief request that is included in Enclosure 2:

• Weld Procedure Specification No. 149

• PG&E ATS Report 420DC-14.20: Welding Procedure Qualification Record (PQR) 771 and Associated Documents 3

Enclosure 1 PG&E Letter DCL-15-021

• SIA Report No. 1301620.402: Stress and Fracture Mechanics Evaluation of Type 410 Stainless Steel Weldments in Safety Injection Pumps at Diablo Canyon Power Plant (Revision 2)

Reference:

1. PG&E Letter DCL-14-060, "ASME Section Xllnservice Inspection Program Request for Alternative REP-S I: Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded attachments," dated July 21, 2014

.(ML14202A613)

NRC Question RAI-EPNB 2.0:

The industry operating experience has shown that socket welds are susceptible to through-wall cracking.

(a) . Please discuss whether any leakage detection systems are available to detect any potential leak from the subject socket welds and whether the operators in the control room would be notified of the leakage.

(b) Please discuss the consequence of a through-wall

• leak and a complete severance at any of the subject socket welds.

• PG&E Response:

Leakage detection systems are not credited to detect any potential leak from the subject socket welds in the Sl pump rooms. The leakage potentially could be detected either during operator walk-downs or through several indirect indications in the control room, which are still being assessed. Failure of the subject socket welds on Sl pump vent and drain connections are not a part of the DCPP design bases calculations. The impact of postulated failures of vent and drain connections are being evaluated. PG&E will coordinate submission of a supplement to the NRC with the NRC project manager to address this RAI.

NRC Question RAI-EPNB 3.0:

When the safety injection pumps are running; Please discuss whether vibration exists on the drain lines and vent lines. If yes, discuss the potential for the vibration that may cause cracking at the socket welds.

4

Enclosure 1 PG&E Letter DCL-15-021 PG&E Response:

PG&E has collected vibration data on the vent and drain connections, and the data is being analyzed by an external vendor. PG&E will submit the response to this RAI question in a supplement along with responses to RAI questions RAI-EVIB 1.0 and RAI-EPNB 2.0. PG&E will coordinate submission of a supplement to the NRC with the NRC project manager to address this RAI.

NRC Question RAI-EPNB 4.0:

Section 2.4, page 2-3, of the stress calculation states that unit axial load of 1000 pounds (lbs). was used as an input.

Please discuss where and how 1000 lbs. was obtained and derived.

PG&E Response:

The unit axial load of 1000 lbs was used in the evaluation to derive the associated stresses for convenience. This is a standard practice for establishing stresses for input to linear elastic fracture mechanics evaluations. As explained in Sections 3.2.3.2 and 3.2.4 of SIA Report 1301620.402, Revision 2, the stresses from the actual loads (shown in Table 3-2 of SIA Report 1301620.402, Revision 2) are determined by linearly scaling these loads to the unit load stresses, based on the ratio of the actual applied load to the 1000 lb unit axial load.

NRC Question RAI-EPNB 5.0:

Section 3 of the stress calculation states that when analyzing the outside diameter flaw, the methods of the ASME Code,Section XI, Appendix C, C-7300 and American Petroleum Institute, API-579, were used to obtain stress intensity factors. However, when analyzing the inside diameter flaw, it does not appear that the API-579 method was used.

Please clarify whether API-579 method was used to evaluate the inside diameter flaw. If not, discuss the reference of the flaw evaluation method.

5

Enclosure 1 PG&E Letter DCL-15-021 PG&E Response:

API-579 method was used to evaluate the stress intensity factor for the inside diameter flaw. This is identified in Section 3.3.1 (pages 3-8) and Section 3.3.4 (pages 3-9) of SIA Report No. 1301620.402, Revision 2.

NRC Question RAI-EPNB 6.0:

Section 3.3.2, page 3-8, of the stress calculation states that" ... residual stresses would not contribute to fatigue crack growth ... " The NRC staff believes that although residual stresses are steady state in nature (i.e.,

constant), they affect the maximum tensile stress and R [stress ratio] and may, therefore, affect fatigue crack growth.

Please clarify the above quoted statement. Discuss whether the above statement is specifically applied to the inside diameter flaw in the subject component.

PG&E Response:

Although residual stresses are steady-state in nature and would influence the R-ratio and therefore the fatigue crack growth rate, (which depends on both R and ~K, according to Eq. 8 on p. 4-3 of SIA Report No. 1301620.402, Revision 2), for an inside diameter flaw, the ~K values are found to be smaller than the threshold stress intensity factor, ~Kth, except for deep flaws at least 50 percent through-wall, (as seen in Figure 4-2 in SIA Report No. 1301620.402, Revision 2). Because the applied ~K is below the threshold ~Kth except for deep flaws, no fatigue crack growth is predicted.

NRC Question RAI-EPNB 7.0:

Section 4 of the stress calculation discusses fatigue crack growth. Section 4 also discussed a postulated inside diameter initiated flaw.

If an inside diameter flaw is postulated, please discuss the likelihood of stress-corrosion cracking in the subject welds.

PG&E Response:

The likelihood of stress corrosion cracking (SCC) is very low. This is because the Sl system operates with controlled purity water and it operates at low temperatures (less than 200°F). It is difficult to initiate SCC cracking at temperatures below 200°F. In general, lower 6

Enclosure 1 PG&E Letter OCL-15-021 temperatures require higher strain and higher water impurity levels to initiate SCC. For these reasons, SCC is not a concern. In addition, service experience to date at OCPP has not shown any evidence of SCC for the Sl system.

NRC Question RAI-EPNB 8.0:

Page 5-1 of the stress calculation states that the allowable flaw depth for an inside diameter flaw exceeds 80 percent of the wall thickness. IWB-3643 of the ASME Code,Section XI, 2003 addenda (Code of record) limits the maximum allowable flaw depth to 75 percent through-wall.

Please explain why the allowable flaw depth for an inside diameter flaw exceeds (i.e., non-conservative) the ASME Code,Section XI limitation.

PG&E Response:

The allowable flaw sizes for an outside diameter (00) flaw and an inside diameter (10) flaw are determined in Sections 3.4.1 and 3.4.2 of SIA Report No. 1301620.402, Revision 2, respectively.

Per section 3.4.1, the allowable flaw depth-to-thickness ratio for an 00 flaw was determined as 0.716 (71.6 percent).

For an 10 flaw, the allowable flaw depth to thickness ratio could be as high as 80 percent, which is the limit of validity of the K solution. As explained in Section 3.4.2, for the 10 flaw, the stress intensity factor K does not exceed the fracture toughness K1c for all flaw depths over which the stress intensity factor solution is valid (80 percent of wall thickness), as defined by API-579 . The 80 percent of wall thickness refers to the limit of validity of the K solution rather than the allowable flaw depth. Therefore, the allowable flaw depth for the 10 flaw can be conservatively set to the ASME Section XI maximum allowable flaw depth of 75 percent through wall.

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Enclosure 2 PG&E Letter DCL-15-021 10 CFR 50.55a Request Number REP-51, Revision 1 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2) 1

Enclosure 2 PG&E Letter DCL-15-021 10 CFR 50.55a Request Number REP-51, Revision 1 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

Compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety Table of Contents

1. ASME Code Component Affected

2. Applicable Code Edition and Addenda

3. Applicable Code Requirement

4. Reason for Request

5. Proposed Alternative and Basis for Use 5.1 Welding Procedure Qualification Tests 5.2 Stress and Fracture Mechanics Evaluation 5.3 Nondestructive Examinations 5.4 Review of Safety Injection Pumps Operating History 5.5 Hardship 5.6 Conclusion

6. Duration of Proposed Alternative

7. References : Photographs of vent and drain connections on Safety Injection Pump 1-1 (typical) and suction drain connection on Safety Injection Pump 1-2 (showing coupling attachment) 2

Enclosure 2 PG&E Letter DCL-15-021 10 CFR 50.55a Request Number REP-51 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(2)

-Compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety-

1. ASME Code Components Affected Diablo Canyon Power Plant (DCPP), Unit 1, ASME Code Class 2, Safety Injection (SI) Pumps 1-1 and 1-2 nominal pipe size (NPS)% inch vent and drain connection socket weld attachments (four attachment welds per pump); and DCPP, Unit 2, ASME Code Class 2, Sl Pump 2-1 NPS% inch vent and drain connection socket weld attachments (four attachment

• welds). (Note: DCPP, Unit 2, Sl Pump 2-2 vent and drain connections were manufactured differently and are not affected).

2. Applicable Code Edition and Addenda ASME Section XI, 2001 Edition through 2003 Addenda.

3. Applicable Code Requirement IWA-4000, "Repair/Replacement Activities," including IWA-4130, "Alternative Requirements," and IWA-4131, "Small Items," as corrective action for the four affected Code Class 2, NPS % inch socket welds on each pump.

4. Reason for Request Relief is requested from implementing the Section XI repair/replacement rules for nonconforming % inch nominal diameter vent valve and drain pipe fitting attachment socket welds. These welds connect to four integrally attached stub piping nipples on each of the three subject Sl Pumps. (Note: larger diameter pipe connections to these pumps were supplied with integral flanged connections and are not affected).

The Unit 1 Sl Pumps 1-1 and 1-2 and Unit 2 Sl Pump 2-1 are size 2 ~,

Model Number JTCH, manufactured by Pacific Pumps. The pump casings are fabricated from martensitic stainless steel and were each supplied with four integrally attached % inch nominal diameter Type 410 3

Enclosure 2 PG&E Letter DCL-15-021 martensitic stainless steel (ASME material Type P-6) pipe nipple stubs.

One integral vent stub nipple and three integral drain stub nipples were supplied with each pump. The pump casings including the pipe nipples and their attachment welds to the pump casings were heat treated during pump manufacture and supplied as an integral pump assembly.

The Unit 1 Sl pumps and connected piping were installed in 1974 and the Unit 2 Sl pump 2-1 and connected piping was installed in 1975 by the original plant construction piping and equipment installation contractor.

During original installation of the pump assemblies in the plant, Type 316 austenitic stainless steel (ASME material Type P-8) isolation valves were welded to the integral vent stub nipple connections, and Type 304 austenitic stainless steel (ASME material Type P-8) pipe fittings (elbows or tees, or for the 1-2 pump suction drain a coupling) were welded to each of the integral drain stub nipple connections supplied with each pump.

The valve or'fitting-to-stub nipple attachment welds were made using the pipe and equipment installation contractor's welding procedure Specification Number 149 (see Attachment 1 of Reference 1) using Type 309 stainless steel filler metal. Procedure 149 was qualified for welding carbon steel (ASME material Type P-1) to austenitic stainless steel (ASME material Type P-8). Procedure 149 was not qualified for welding martensitic stainless steel (ASME material Type P-6) to austenitic stainless steel (ASME material Type P-8); and therefore, does not contain provision for post-weld heat treatment that would potentially be required by a P-6 to P-8 Procedure. The discrepancy in welding procedure qualification was discovered in December 2013 during material verification as part of the planning process for anticipated replacement of the Pump 1-1 vent valve due to boric acid leakage from the valve packing.

ASME Section XI would require use of IWA-4000 repair/replacement rules for correction of the four nonconforming % inch nominal diameter socket welds on each subject pump.

5. Proposed Alternative and Basis for Use PG&E proposes to accept the existing Sl Pumps 1-1, 1-2, and 2-1 vent and drain attachment socket welds as-is.

To confirm acceptability of the existing Sl pumps vent and drain socket welds, PG&E has:

• conducted welding procedure qualification tests with representative 41 0 stainless steel and 304 stainless steel base materials using Type 309 filler metal as per the original Welding Procedure Specification 149 parameters without post-weld heat treatment (see Attachment2 of Reference 1);

• 4

Enclosure 2 PG&E Letter DCL-15-021

• performed a Stress and Fracture Mechanics Evaluation of Type 41 0 Stainless Steel Weldments in Sl Pumps at DCPP (see Attachment 3 of Reference 1);

• performed nondestructive examinations (NOEs) of the subject welds to determine and verify current conditions; and

• performed a review of the Sl pumps operating histories including pressure test records.

Each of these actions are discussed below and detailed in the attachments of Reference 1.

5.1 Welding Procedure Qualification Tests Welding Procedure Qualification Test Report is presented in Attachment 2 of Reference 1. For the weld qualification tests, Arc-Met testing to determine carbon content of the existing Sl pumps, 41 0 stainless steel pipe nipples were attempted but proved unsuccessful due to the small pipe size, short lengths of the drain nipples and adverse component configurations. As a result, Type 41 0 stainless steel material with the highest carbon content readily available (0.13 percent) was used for the qualification testing. To qualify the procedure, 3/8 inch thick Type 410 stainless steel plate was welded to 3/8 inch thick Type 304 stainless steel plate using a combination of gas tungsten arc welding (GTAW) at the root with shielded metal arc welding (SMAW) for the cover passes. Ambient condition preheat of 66.5°F was used with maximum interpass temperature of 297°F recorded. No post weld heat treatment was used.

The final weld was sectioned to provide two tensile and four bend test specimens which were tested by an independent laboratory. Two of the bend specimens were subjected to root bending, 180 degrees, and two were subjected to face bending, 180 degrees, over rollers with diameter of 4 times the bend specimen thickness, with the weld and heat-affected zones centered within the convex length of bent samples per ASME Section IX, Table QW-451.1 and QW-160, 2013 Edition. The samples were subsequently examined for cracks and other defects and all were found acceptable.

The two tensile test specimens were tested in accordance with ASME

• Section IX, Table QW-451.1 and QW-150, 2013 Edition, with required ultimate tensile strength of 65 Kips (1 000 pounds) per square inch (ksi).

Actual ultimate tensile strengths of 75.5 ksi and 76.0 ksi respectively were recorded, with the breaks occurring in the 410 stainless steel parent metal in both instances.

5

Enclosure .2 PG&E Letter DCL-15-021 5.2 Stress and Fracture Mechanics Evaluation Stress and Fracture Mechanics Evaluation Report prepared by Structural Integrity Associates (SIA) is presented in Attachment 3 of Reference 1.

SIA's evaluation of the % inch Type 410 stainless steel nipples welded to Type 316 valves or Type 304 fittings without post weld heat treatment on the DCPP Sl Pump vent and drain lines consisted of stress analysis, evaluation of allowable flaw size under maximum loading, and evaluation of crack propagation of postulated flaws under cyclic fatigue loading. A fracture mechanics approach analogous to the methods of ASME Code Section XI, supplemented with procedures from American Petroleum Institute Standard API-579, was used because the ASME Section XI methods do not address Type 41 0 martensitic stainless steels, evaluation of (postulated) flaws on piping outside diameter (OD) surfaces, or evaluation of flaws in piping of diameter 4 inches or less.

The postulated flaw extends from the socket weld toe on the Type 41 0 stainless steel nipple, which is the region where cyclic stresses are the largest, and grows from the OD toward the inside diameter (I D).

Additionally, a postulated flaw originating at the ID was evaluated due to the presence of residual tensile stresses as a result of welding.

The depths of OD and ID flaws located along the largest cyclic stress path that would cause crack instability under maximum operating loads and pressure, including seismic/abnormal loads and applicable structural factors, were evaluated. The allowable flaw depth for an OD flaw was determined to be 0.110 inch, approximately 71.6 percent of the wall thickness of 0.154 inch. The allowable flaw depth for an ID flaw was found to exceed 80 percent of the wall thickness.

For cyclic loading, postulated ID flaws are not predicted to grow as all cyclic stress intensity factors are below the fatigue threshold.

For postulated OD crack analysis, 7000 thermal transient cycles, 400 design earthquake cycles, and 20 Hosgri earthquake cycles were assumed. For the postulated OD crack to grow by fatigue under cyclic operating loads, and pressure to the allowable flaw size in the evaluated number of cycles, an initial crack of at least 0.104 inch depth is required.

This depth corresponds to a surface length of 0.832 inch for a crack aspect ratio of 4.

For nondestructive test minimum length detection limits of 1/16 inch (such as for liquid penetrant examinations), fatigue crack growth will not occur 6

Enclosure 2 PG&E Letter DCL-15-021 for a postulated OD flaw where surface length is equal to the detection limit, even for load cycles associated with the Hosgri earthquake.

For a postulated 10 percent through-wall OD flaw, no growth is predicted except for the 20 cycles assumed for the Hosgri event. For that case, the associated crack extension is 8.3 x 1o- inch.

6 For a postulated OD crack 0.026 inch deep Oust exceeding the fatigue crack growth threshold), the amount of crack extension under the evaluated cyclic loading is 0.0015 inch.

The evaluations of the postulated OD and ID flaws show that crack growth under anticipated cyclic loading is minimal.

5.3 Nondestructive Examinations During the operating history of the plant, the subject welds have been examined by qualified VT -2 visual examiners every 40 months during scheduled ASME Section XI system pressure tests. No leakage from any of the welds has ever been identified.

Liquid penetrant examinations of all subject welds were performed between December 18 and 20, 2013, with specific attention focused for crack-like indications. No linear or crack-like indications were detected.

5.4 Review of Safety Injection Pumps Operating History The cumulative number of starts is a measure of the cyclic loading experienced by the pumps, as analyzed in the stress and fracture mechanics evaluation. The Sl pumps were each started several times during testing prior to plant operation. During plant operation, the pumps normally function in a stand-by capacity and are periodically started for pump readiness testing and system pressurizations for leak testing, as well as a small number of starts in support of the Slfunction.

Preoperational starts are an estimate of the number of Sl pump starts during preoperational startup testing activities and during three Plant Hot Functional Testing programs. Each pump is estimated to have had 25 preoperational starts.

The total number of operational starts for Sl Pumps 1-1, 1-2, and 2-1 through the end of 2013 was estimated using the operating data of each of these pumps to establish an annual average. This average, 11 starts per year for each pump, was extrapolated back to the commencement of plant operation.

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Enclosure 2 PG&E Letter DCL-15~021 Total preoperational and operational start estimates were then added together. The resulting estimated number of starts for each Sl pump during the life of the plant was multiplied by 2 as a conservative measure allowing for a higher number of starts per year at beginning of plant life plus any pressurizations of the Sl piping by means other than a pump start, such as hydro testing.

The calculation of total starts for each pump is as follows: [Number of preoperational starts plus (Average number of starts per year multiplied by number of years of plant operation)] multiplied by 2.

Total starts for Sl Pumps 1-1 and 1-2: [25 starts + (11 starts/year X 29 years)] X 2 = 688 starts Total starts for Sl Pump 2-1: [25 starts+ (11 starts/year X 28 years)] X 2

= 666 starts.

The total number of starts to date (approximately half of plant life assuming a 20 year license renewal extension) for each of the subject Sl pumps is conservatively estimated to be less than 700 starts.

Conservatively assuming an additional 700 starts during the second half of plant life (including the assumed 20 year license extension period), the total number of Sl pump starts during all of plant lifetime is estimated to

• be less than 1400 starts. This is well under the 7000 thermal transient cycles assumed in the fatigue crack growth analysis.

5.5 Hardship The existing vent stub nipple to vent valve and drain stub nipples to drain pipe fittings socket welds were made in accordance with the original construction contractor's Welding Procedure Specification 149 which did not provide for post-weld heat treatment. There may have been some self-tempering from the weld process itself. The stub nipples are short, typically less than 4 inches in length, with the casing drain nipples being less than 1 inch long. The suction drain nipple on Pump 1-2 is also approximately 1" long. Photographs of each attachment to Pump 1-1 which are typical of all the pumps are included in Attachment 1 of this Enclosure for reference. A photograph of the non-typical exception, the much shorter suction drain connection for pump 1-2 which has a coupling welded to the shorter martensitic nipple is also included.

Attempting to perform localized heat treatment would jeopardize the factory heat treatment of the entire pump casing. Heat treatment of the entire pump casing to include these small welds cannot be performed in 8

Enclosure 2 PG&E Letter DCL-15-021 situ under the same controlled conditions available during factory fabrication. Removing the pump for heat treatment would entail disassembly, cutting pipe connections, removal of each pump, transport and heat treatment of a large machine under conditions of radiological contamination, followed by a reverse process for reinstallation. Either means of heat treatment could result in warping or damage to the pump casings. Removal of the pumps would involve additional handling of

• contaminated equipment and entails risk of damage. These factors constitute a hardship for compliance with the specified requirements.

An alternative design using threaded fittings was also considered. The existing welded fittings could be removed and the remaining martensitic nipples could be threaded to accept threaded austenitic fittings, although the case drains and the suction drain on Pump 1-2 are very short and threading may not be practicable. For threaded fittings, since seal welding would introduce the same heat treatment hardship as socket welds, the pipe thread would be the only seal for the *boric acid solution process fluid at system operating pressure of up to or exceeding 1520 psig. The construction code limit for threaded connections in% inch pipe is 1500 psig. Experience with the boric acid corrosion control program has shown that threaded connections such as pipe caps often leak boric acid even at comparatively low pressures. These factors would also constitute a hardship.

Contact radiation levels at the Sl pumps are relatively low, approximately 0.5 mR per hour. However, a number of craft personnel would be required for disassembly, rigging, removal and transport of the pumps, as well as handling during the heat treatment process. Given involvement of 6 personnel for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per pump for disassembly and another 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> for reassembly, incremental radiation exposure of 720 mR would be expected for the 3 pumps in addition to the hardship with heat treating large contaminated machines.

5.6 Conclusion As discussed above and demonstrated and documented in Attachment 2 and Attachment 3 of Reference 1, attempting to heat treat the nipple to valve or fitting socket welds would require either localized heating or heating the entire pump cas.ing. Either evolution could not be performed in situ under the same controls as available during original factory fabrication of the pumps and could thus result in warping or other damage to the pump casings. Removing the pumps for heat treatment would require disassembly, cutting pipe connections, removing the pump and transporting the contaminated casing for heat treatment, then reversing 9

Enclosure 2 PG&E Letter DCL-15-021 the process to reinstall. The risk of damage from the pump removal and reinstallation process can not be eliminated.

The existing Sl pumps vent and drain socket welds were fabricated using a procedure satisfactory in all respects other than post-weld heat treatment. This procedure has been demonstrated to produce the required mechanical properties of a post-weld heat treated procedure.

Some degree of self-tempering may have been provided by the weld process itself. Welding process tests, nondestructive examination of the existing welds, and analysis of the existing welds have been performed to demonstrate acceptability of the existing configuration. Accordingly, compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety in accordance with 10 CFR 50 .55a(z)(2), thus the existing weldments may be determined acceptable as-is for continued service.

6. Duration of Proposed Alternative The proposed alternative will apply for the remaining service life of Sl Pumps 1-1, 1-2, and 2-1, including the duration of the current operating licenses plus a contemplated license extension period of 20 years.

7. References

1. PG&E Letter DCL-14-060, "ASME Section XI lnservice Inspection Program Request for Alternative REP-S I: Proposed Alternative to Requirements for Repair/Replacement Activities for Certain Safety Injection Pump Welded attachments," dated July 21, 2014 (ML14202A613) 10

Attachment 1 Enclosure 2 PG&E Letter DCL-15-021 Photograph 1: Sl Pump 1-1: Pump Case Drain

Attachment 1 Enclosure 2 PG&E Letter DCL-15-021 Photograph 2: Sl Pump 1-1: Pump Discharge Drain 2

Attachment 1 Enclosure 2 PG&E Letter DCL-15-021 Photograph 3: Sl Pump 1-1: Pump Suction Drain 3

Attachment 1 Enclosure 2*

PG&E Letter DCL-15-021 Photograph 4: Sl Pump 1-1: Pump Vent 4

Attachment 1 Enclosure 2 PG&E Letter DCL-15-021 Photograph 5: Sl Pump 1-2: Pump Suction Drain 5