LR-N20-0058, Response to Request for Additional Information License Amendment Request Re Application of Leak-Before-Break for Accumulator, Residual Heat Removal, Safety Injection and Pressurizer Surge Lines

Response to Request for Additional Information License Amendment Request Re Application of Leak-Before-Break for Accumulator, Residual Heat Removal, Safety Injection and Pressurizer Surge Lines

ML20260H194
Person / Time
Site:	Salem
Issue date:	09/16/2020
From:	Mcfeaters C Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2020-LLA-0088, LAR S19-08, LR-N20-0058
Download: ML20260H194 (8)

Text

September 16, 2020 10 CFR 50.90 LR-N20-0058 LAR S19-08 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

Subject:

Response to Request for Additional Information License Amendment Request Regarding Application of Leak-Before-Break for Accumulator, Residual Heat Removal, Safety Injection and Pressurizer Surge Lines References 1. PSEG letter to NRC, "License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology, dated April 24, 2020 (ADAMS Accession No. ML20115E374)

2. NRC e-mail to PSEG, "Salem 1 and 2 - Final RAI RE: Leak-Before-Break Methodology (L-2020-LLA-0088), dated August 6, 2020 (ADAMS Accession No. ML20219A651)

In the Reference 1 letter, PSEG Nuclear LLC (PSEG) submitted a license amendment request for Salem Generating Station (Salem) Units 1 and 2. The proposed amendment requested to exclude the dynamic effects of specific postulated pipe ruptures from the design and licensing basis based on leak-before-break methodology.

In Reference 2, the U.S. Nuclear Regulatory Commission staff provided PSEG a Request for Additional Information (RAI) to support the NRC staffs technical review of Reference 1. to this submittal provides the response to the RAI.

PSEG has determined that the information provided in this submittal does not alter the conclusions reached in the 10 CFR 50.92 no significant hazards determination previously submitted. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

No new regulatory commitments are established by this submittal. If you have any questions or require additional information, please do not hesitate to contact Mr. Brian Thomas at (856) 339-2022.

LR-N20-0058 10 CFR 50.90 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

9/16/2020 Executed on _______________________

(Date)

Respectfully, Charles V. McFeaters Site Vice President Salem Generating Station Response to Request for Additional Information License Amendment Request Regarding Application of Leak-Before-Break for Accumulator, Residual Heat Removal, Safety Injection and Pressurizer Surge Lines cc: Mr. D. Lew, Administrator, Region I, NRC Mr. J. Kim, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Site Regulatory Commitment Tracking Coordinator

LR-N20-0058 LAR S19-08 Response to Request for Additional Information License Amendment Request Regarding Application of Leak-Before-Break for Accumulator, Residual Heat Removal, Safety Injection and Pressurizer Surge Lines 1

LR-N20-0058 LAR S19-08 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST REGARDING APPLICATION OF LEAK-BEFORE-BREAK FOR ACCUMULATOR, RESIDUAL HEAT REMOVAL, SAFETY INJECTION AND PRESSURIZER SURGE LINES PSEG NUCLEAR LLC SALEM GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-272 AND 50-311 EPID: L-2020-LLA-0088 By letter dated April 24, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20115E374), PSEG Nuclear LLC (PSEG or the licensee) submitted a license amendment request (LAR) for the U. S. Nuclear Regulatory Commission (NRC) review and approval. In the LAR, the licensee proposed to apply leak-before-break (LBB) methodology to specific portions of accumulator, safety injection (SI) and residual heat removal (RHR) lines and the entire pressurizer surge lines at Salem Generating Station (Salem) Units 1 and 2. To complete its review, the NRC staff requests the following additional information.

1. Section 2.1.3 in Enclosure 1 of the licensees letter describes the RHR system piping and associated LBB scope. That section does not clearly describe whether the scope of the LAR includes the RHR return line piping to the reactor coolant system (RCS) primary loop. Clarify whether the RHR return line piping is included in the scope of the LAR.

PSEG Response:

At Salem the RHR return line piping does not directly connect to the RCS primary loops.

The RHR return to the RCS cold legs is through the 6-inch Safety Injection lines which are attached to the 10-inch accumulator lines (see segment SI-CL-I in Figure 3-1 of Reference 6). As discussed in Section 2.1 of Enclosure 1 of Reference 1, the scope of the piping segments included in the LAR stop at the first pressure isolation valve off of the RCS. For this return path, the scope of the LAR is from the RCS cold leg to the SJ56 check valve (see segment ACC-I in Figure 3-1 of Reference 4).

The RHR return to the RCS hot legs is through the 6-inch Safety Injection lines which connect to the #3 and #4 hot legs through the SJ156 check valves (see segment S1-HL-II and S1-HL-III on Figure 3-4 in Reference 6). For this return path, the scope of the LAR is from the RCS hot leg to the SJ156 check valve which is the first check valve off of the RCS.

2. Section 4.1 in each of the WCAP-18248, WCAP-18249, WCAP-18253 and WCAP-18261 reports indicates that the nozzles of the subject piping at the RCS primary loop is the A182 F316 material that has material properties identical to the A376 TP316 material. However, the basis of the determination on the identical properties is not clear. Describe how the licensee confirmed that the mechanical properties of the A182 F316 and A376 TP316 materials used in the subject piping are identical (e.g., by comparing the data in certified materials test reports (CMTRs)).

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LR-N20-0058 LAR S19-08 PSEG Response:

The statement indicating identical properties is intended to refer to the ASME Code Section II properties for Modulus of Elasticity (E), Yield Strength (Sy), and Ultimate Strength (Su) used for the Leak-Before-Break evaluations. SRP 3.6.3 (References 8 and 9) requires the evaluation of the least favorable combination of stress and material properties. LBB stresses in the nozzle are significantly lower than the stresses in the piping due to a larger geometry for the distribution of stress. Differences in the specific Certified Material Test Report (CMTR) properties of the nozzles are negligible since the analyses are bounded by the piping locations. As such, CMTRs for the nozzle were not reviewed or compared against the average and minimum CMTR properties of the piping.

3. Section 4.1 of the WCAP-18261 report states that the material of fabrication for the pressurizer surge lines is A376 TP316 stainless steel. ASTM A376 TP316 provides a material specification for seamless pipes. Section 4.1 of the WCAP report does not identify a material of fabrication for pipe fittings (e.g., elbows). If pipe fittings are used in the pressurizer surge lines, describe the material specification for the pipe fittings and discuss how the mechanical properties of the pipe fittings are determined in the LBB analysis.

PSEG Response:

The surge lines of Salem Units 1 and 2 are fabricated entirely of straight pipe and pipe bends; both of which are the A376 TP316 stainless steel material identified in the report.

There are no elbows or other pipe fittings with alternate material types.

4. Section 3.6 of the WCAP-18261 report describes the equation that the licensee used to calculate the piping minimum thickness at the weld counter-bore in the pressurizer surge lines (i.e., (0.041 + 1.75 x pipe nominal thickness)/2). The section does not clearly describe a technical basis or relevant reference for the equation. Describe the technical basis and relevant reference for the equation.

PSEG Response:

The pipe weld minimum thickness is determined based on the American National Standard, ANSI B16.25-1979, Buttwelding Ends (Reference 10). Specifically, subsection 4.1.5 equation:

C = A - 0.031 - 1.75t - 0.010 Where C is the inside diameter of the weld preparation counter-bore, A is the outside diameter of the pipe, and t is the nominal wall thickness of the pipe.

Minimum wall thickness is then back-calculated to be; tmin = [A - C] / 2 = [A - (A - 0.031 - 1.75tnom - 0.010)] / 2 tmin = [0.041+1.75tnom] / 2 3

LR-N20-0058 LAR S19-08

5. Table 5-1 of the WCAP-18248 report describes the critical locations for the SI-HL-I segment of the SI lines at Salem Units 1 and 2. Table 5-1 of the WCAP report indicates that location 2-RH-113-A is identified as the Unit 2 critical location for the SI-HL-I segment. However, Table 3-35 of WCAP-18248 indicates that location FW-RC-231-20 has a higher total faulted stress than location 2-RH-113-A for the segment of Salem Unit

2. Explain why location 2-RH-113-A is identified as the Unit 2 critical location for the segment even though the total faulted stress of location 2-RH-113-A is less than that of location FW-RC-231-20.

PSEG Response:

Weld 2-RH-113-A is the limiting location in Segment SI-HL-I for the SAW (submerged arc weld) welding process. Weld FW-RC-231-20 used a SMAW (shielded metal arc weld) welding process and is categorized separately from weld 2-RH-113-A. The limiting SMAW location for segment SI-HL-I is weld FW-1-RH-72-43A. The faulted stress for weld FW RH-72-43A (Table 3-32 of Reference 6) is more limiting that the faulted stress for weld FW-RC-231-20 (Table 3-35 of Reference 6).

Analysis Segment Weld Location Weld Process Faulted Stress (ksi) 2-RH-113-A SAW 19834 SI-HL-I FW-RC-231-20 SMAW 20601 FW-1-RH-72-43A SMAW 23246

6. Section 2.0 of the WCAP-18516 report describes the fatigue crack growth analysis with a postulated initial surface crack in the accumulator line. For the transients without accumulator/SI/RHR actuation, the generic transients, which were used in the fatigue analysis, do not totally match the Salem plant-specific transients. For example, the generic transients do not include heatup, cooldown or hot standby operations transients that are included in the plant-specific transients (as shown in Tables 2-1 and 2-2 of WCAP-18516 for generic and plant-specific transients, respectively).

Given the differences between the generic transients and plant-specific transients, describe how the licensee determined that, for the transients without accumulator/SI/RHR actuation, the total number of generic transient cycles is slightly less than the total number of Salem 60-year projection cycles. As part of the response, clarify whether the cycles of certain generic transients were added up to represent the cycles of similar plant-specific transients (e.g., unit loading transient cycles in the generic transients are added up to represent the heatup transient cycles in the plant-specific transients).

PSEG Response:

As indicated in WCAP-18516-P (Reference 7), the fatigue crack growth for the Accumulator Line would be dominated by the transients which experience actuation of the Accumulator, Safety Injection, or Residual Heat Removal systems. These transients generate flow in the normally-stagnant Accumulator line piping and produce a thermal shock with the potential to produce much more significant flaw growth. Alternately, for the transients without actuation of the Accumulator, Safety Injection, or Residual Heat Removal systems, the Accumulator piping remains stagnant. Any thermal transient effects transfer very slowly 4

LR-N20-0058 LAR S19-08 from the primary loop (cold leg) piping and produce very small changes in the stress intensity factors, most often leading to negligible flaw growth. As such, the comparison of transients in WCAP-18516-P was predominantly focused on the transients with actuation of the Accumulator, Safety Injection, or Residual Heat Removal systems and showing that the cycles for the generic FCG evaluation clearly bound the Salem design transient set.

For the non-actuation transients past experience with fatigue and flaw growth evaluations provide the bases for the following justification of the flaw growth effects:

Heatup, Cooldown, Hot Standby Operations, Primary Leak Test, Primary Hydrostatic Test, and Secondary Hydrostatic Test (approximately 1570 combined cycles) are very slow transients which produce negligible thermal stresses and flaw growth, regardless of the number of cycles.

Reactor Trip, Loss of Flow, Loss of Load, Loss of Power, and Inadvertent Auxiliary Spray (approximately 360 combined cycles) may be more severe transients for the primary loop piping, with faster temperature changes and greater thermal range, but these effects are greatly attenuated within the Accumulator Line piping. As such, it is judged that the flaw growth of these 360 cycles would be bounded by the additional cycles of Unit Loading/Unloading (approximately 10,000+ additional cycles) and Step Load/Unload (approximately 1,900+ additional cycles).

Since the non-actuation transients are either negligible or much less significant than the transients with actuation of the Accumulator, Safety Injection, or Residual Heat Removal systems the conclusions of WCAP-18516-P remain applicable. Growth of small surface flaws is very small, very slow, and the flaws are not capable of becoming through-wall flaws.

7. Subparagraph 11.(C).(i) of Standard Review Plan 3.6.3, Subsection III indicates that the accuracy of the leak rate computational methods should be demonstrated by comparison with other acceptable computational procedures or with experimental data. Describe how the licensees leak rate computational methods have been compared with other acceptable computational procedures or with experimental data.

PSEG Response:

The basis for the Westinghouse leak rate calculation has been long-established and used consistently on all Leak-Before-Break applications for more than three decades. Leak rate calculations using this evaluation model have been compared against alternate calculation models and benchmarked against leak rate data that was available from laboratory testing.

Furthermore, an excerpt from NUREG-1512 (Reference 11) regarding the Westinghouse leak-rate computer code reads Westinghouse software had been adequately benchmarked and was acceptable. While NUREG-1512 had been intended to provide the safety evaluation of the AP600 standard plant design, the conclusions related to the Westinghouse leak rate calculation methodology are relevant in general applications and demonstrates a history of review and approval by the NRC.

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LR-N20-0058 LAR S19-08

References:

1. PSEG Letter, License Amendment Request to Exclude the Dynamic Effects of Specific Postulated Pipe Ruptures from the Design and Licensing Basis Based on Leak-Before-Break Methodology, April 24, 2020 (ADAMS Accession No. ML20115E374).

2. NRC e-mail to PSEG, "Salem 1 and 2 - Final RAI RE: Leak-Before-Break Methodology (L-2020-LLA-0088), dated August 6, 2020 (ADAMS Accession No. ML20219A651)

3. Westinghouse Reports, WCAP-18261-P (Proprietary) and WCAP-18261-NP (Non-proprietary), Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology, March 2018.

4. Westinghouse Reports, WCAP-18249-P (Proprietary) and WCAP-18249-NP (Non-proprietary), Technical Justification for Eliminating Accumulator Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology, March 2018.

5. Westinghouse Reports, WCAP-18253-P (Proprietary) and WCAP-18253-NP (Non-proprietary), Technical Justification for Eliminating Residual Heat Removal Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology, March 2018.

6. Westinghouse Reports, WCAP-18248-P (Proprietary) and WCAP-18248-NP (Non-proprietary), Technical Justification for Eliminating Safety Injection Line Rupture as the Structural Design Basis for Salem Units 1 and 2, Using Leak-Before-Break Methodology, March 2018.

7. Westinghouse Reports, WCAP-18516-P (Proprietary) and WCAP-18516-NP (Non-proprietary), Fatigue Crack Growth Evaluations of Salem Units 1 and 2 Accumulator, RHR, Pressurizer Surge and Safety Injection Lines Supporting Expanded Scope Leak-Before-Break, February 2020.

8. Standard Review Plan: Public Comments Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register/Vol. 52, No. 167/Friday August 28, 1987/Notices, pp.

32626-32633.

9. NUREG-0800, Revision 1, Standard Review Plan: 3.6.3 Leak-Before-Break Evaluation Procedures, March 2007.

10. American National Standards, Buttwelding Ends, ANSI B16.25-1979.

11. NUREG-1512, Volume 1, Final Safety Evaluation Report Related to Certification of the AP600 Standard Design, U.S. Nuclear Regulatory Commission, September 1998 (ADAMS Accession No. ML081080310).

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