ML15281A164
| ML15281A164 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 12/03/2015 |
| From: | Siva Lingam Plant Licensing Branch IV |
| To: | Halpin E Pacific Gas & Electric Co |
| Singal B, NRR/DORL/LPLIV-1 | |
| References | |
| CAC MF4988, CAC MF4989 | |
| Download: ML15281A164 (21) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 3, 2015 Mr. Edward D. Halpin Senior Vice President and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424
SUBJECT:
DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: APPLICATION OF LEAK-BEFORE-BREAK IN THE FUEL STRUCTURAL ANALYSIS RESULTS TO SATISFY 10 CFR 50.46(b)(4) (CAC NOS. MF4988 AND MF4989)
Dear Mr. Halpin:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 221 to Facility Operating License No. DPR-80 and Amendment No. 223 to Facility Operating License No. DPR-82 for the Diablo Canyon Power Plant (DCPP), Units No. 1 and 2, respectively. The amendments consist of changes to the licensing basis as described in the DCPP Updated Final Safety Analysis Report (UFSAR) Sections 3.6.2.1.1.1, "Reactor Coolant System Main Loop Piping (Leak-Before-Break)," and 4.2.1.1.2, "Fuel Assembly Structure," in response to your application dated October 2, 2014.
The amendments approve performance of the fuel assembly structural analysis based on a pipe break that considers the application of leak-before-break (LBB), which allows the exclusion of the dynamic effects of certain pipe breaks (primary loop piping) and use the fuel assembly structural analysis results to satisfy, in part, the emergency core cooling system performance acceptance criterion, paragraph 50.46(b)(4) of Title 10 of the Code of Federal Regulations (1 O CFR). The amendments revise the DCPP UFSAR to document that the fuel assembly structural analyses are based on a pipe break location that considers the application of LBB and the results of the fuel assembly structural analyses are used as an input to demonstrate compliance with 10 CFR 50.46(b)(4).
E. Halpin A copy of the related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.
Sincerely,
~<j'*~
Siva P. Lingam, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50~275 and 50-323
Enclosures:
- 1. Amendment No. 221 to DPR-80
- 2. Amendment No. 223 to DPR-82
- 3. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 221 License No. DPR-80
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Pacific Gas and Electric Company (the licensee), dated October 2, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
- 2. Accordingly, the license is amended as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 221 are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance. In addition, the licensee shall include the revised information in the next Updated Final Safety Analysis Report update submitted to the NRC in accordance with 10 CFR 50.71(e), as described in the licensee's application dated October 2, 2014, and evaluated in the staff's safety evaluation enclosed with this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. DPR-80 Date of Issuance: December 3, 2015
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 223 License No. DPR-82
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A The application for amendment by Pacific Gas and Electric Company (the licensee), dated October 2, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 2
- 2. Accordingly, the license is amended as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 223, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance. In addition, the licensee shall include the revised information in the next Updated Final Safety Analysis Report update submitted to the NRC in accordance with 10 CFR 50.71(e), as described in the licensee's application dated October 2, 2014, and evaluated in the staff's safety evaluation enclosed with this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. DPR-82 Date of Issuance: December 3, 2015
ATTACHMENT TO LICENSE AMENDMENT NO. 221 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 223 TO FACILITY OPERATING LICENSE NO. DPR-82 DOCKET NOS. 50-275 AND 50-323 Replace the following pages of the Facility Operating License Nos. DPR-80 and DPR-82 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License No. DPR-80 REMOVE INSERT Facility Operating License No. DPR-82 REMOVE INSERT (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 221 are hereby incorporated in the license.
Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3) Initial Test Program The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Company's Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:
- a. Elimination of any test identified in Section 14 of PG&E's Final Safety Analysis Report as amended as being essential; Amendment No. 221
(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This License shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.
(2) Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 223, are hereby incorporated in the license.
Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3) Initial Test Program (SSER 31, Section 4.4.1)
Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.
- The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Amendment No. 223
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 221 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 223 TO FACILITY OPERATING LICENSE NO. DPR-82 PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323
1.0 INTRODUCTION
By application dated October 2, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14276A195), Pacific Gas and Electric Company (the licensee) requested changes to Facility Operating License Nos. DPR-80 and DPR-82 for the Diablo Canyon Power Plant (DCPP), Units 1 and 2, respectively. The licensee proposed changes to the licensing basis as described in the DCPP Updated Final Safety Analysis Report (UFSAR) Sections 3.6.2.1.1.1, "Reactor Coolant System Main Loop Piping (Leak-Before-Break),
and 4.2.1.1.2, "Fuel Assembly Structure." Portions of the letter dated October 2, 2014, contain sensitive unclassified non-safeguards information (proprietary) and accordingly, have been withheld from public disclosure pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 2.390.
In its letter dated October 2, 2014, the licensee requested U.S. Nuclear Regulatory Commission (NRC) staff approval to perform the fuel assembly structural analysis based on a pipe break that considers the application of leak-before-break (LBB), which allows the exclusion of the dynamic effects of certain pipe breaks and use the fuel assembly structural analysis results to satisfy, in part, the emergency core cooling system (ECCS) performance acceptance criterion in 10 CFR 50.46(b)(4). The licensee has proposed to revise the DCPP UFSAR Sections 3.6.2.1.1.1 and 4.2.1.1.2 to document that the fuel assembly structural analyses are based on a pipe break location that considers the application of LBB and the results of the fuel assembly structural analyses are used as an input to demonstrate compliance with 10 CFR 50.46(b)(4).
2.0 REGULATORY EVALUATION
The fuel assembly structural analyses are performed in accordance with NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:
Enclosure 3
LWR [Light-Water Reactor] Edition" (SRP), Section 4.2, "Fuel System Design," Revision 3 (ADAMS Accession No. ML070740002). SRP Section 4.2 states, in part, that:
The fuel system safety review provides assurance that (1) the fuel system is not damaged as a result of normal operation and anticipated normal occurrences (AOOs), (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained.
General Design Criterion (GDC) 10 ... establishes specified acceptable fuel design limits (SAFDLs) that should not be exceeded during any condition of normal operation, including the effects of AOOs. Therefore, the SAFDLs are established to ensure that the fuel is not damaged. Within this context, "not damaged means that the fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those assumed in the safety analysis. The design limit of GDC 10 (i.e., the SAFDLs) accomplish these objectives. In a "fuel rod failure," the fuel rod leaks and the first fission product barrier (the cladding) is breached. The dose analysis required by 10 CFR Part 100 for postulated accidents must account for fuel rod failures. "Coolability," in general, means that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual heat even after a severe accident. The general requirements to maintain control rod insertability and core coolability appear repeatedly in the GDC found in Appendix A to 10 CFR Part 50 (e.g., GDC 27 and 35). In particular, 10 CFR 50.46 provides the specific coolability requirements of the loss-of-coolant accident (LOCA).
Appendix A, "Evaluation of Fuel Assembly Structural Response to Externally Applied Forces," to SRP Section 4.2 describes the applicable analytical procedures for fuel assembly structural response to seismic and LOCA loads.
DCPP, Units 1 and 2, were licensed for construction prior to the issuance of 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," and so the DCPP operating license was issued based on compliance with the draft General Design Criteria (Draft GDC)
(Federal Register notice dated July 11, 1967 (32 FR 10213)). Therefore, the licensing basis for DCPP remains the Draft GDC, with some exceptions noted in the DCPP UFSAR, Section 3.1 and Appendix 3.1A1 . For the purposes of the NRC staff review of this LAR, the following GDCs apply: GDC 2, "Design bases for protection against natural phenomena" (DCPP licensing basis is the Draft GDC, "Criterion 2 - Performance Standards (Category A")) and GDC 4, "Environmental and dynamic effects design bases" (DCPP licensing basis is the revised GDC 4 issued in 1987).
Draft GDC 2 states:
Those systems and components of reactor facilities that are essential to the prevention of accidents which could affect the public health and safety, or to 1 The licensee incorrectly quoted 10 CFR 50, Appendix A, Criterion 2, in its application dated October 2, 2014.
mitigation of their consequences, shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area, and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.
GDC 4 (revised in 1987 as discussed above) states:
Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
The regulations in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," establish standards for the calculation of emergency core cooling system (ECCS) performance and acceptance criteria for that calculated performance. Paragraph 10 CFR 50.46(b)(4) provides the following ECCS performance criteria for core geometry:
Coo/able geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
GDC 4 provides a regulation permitting the use of LBB. The LBB concept is based on calculations and experimental data demonstrating that certain pipe material has sufficient fracture toughness (ductility) to prevent a small through-wall flaw from propagating rapidly and uncontrollably to catastrophic pipe rupture and to ensure that the probability of a pipe rupture is extremely low. The small leaking flaw is demonstrated to grow slowly and the limited leakage would be detected by the reactor coolant system (RCS) leakage detection systems early on such that licensees can shut down the plant to repair the degraded pipe long before the potential catastrophic pipe rupture.
SRP Section 3.6.3, Revision 1, "Leak-Before-Break Evaluation Procedures" (ADAMS Accession No. ML063600396), provides guidance on screening criteria, safety margins, and analytical methods for the piping systems to be qualified for LBB. NUREG-1061, Volume 3, "Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe
Breaks," November 1984 (ADAMS Accession No. ML093170485), provides the technical basis for the LBB analyses.
Licensees need to submit, for NRC review and approval, a fracture mechanics evaluation of specific piping configurations to meet the requirements of GDC 4. A candidate pipe should satisfy the screening criteria of SRP Section 3.6.3 by demonstrating that it experiences no active degradation mechanisms. The pipe also needs to satisfy the safety margins described in SRP Section 3.6.3 via fracture mechanics analyses. Finally, the LBB application is predicated on the ability of the RCS leakage detection systems to detect a certain leak rate, with certain margin, that corresponds to the leakage flaw size of the candidate pipe. NRC Regulatory Guide (RG) 1.45, Revision 1, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage," May 2008 (ADAMS Accession No. ML073200271), provides acceptance criteria for the RCS leakage detection systems.
Hence, this amendment request was reviewed against the guidance in NUREG-0800, Sections 3.6.3 and 4.2, and the regulatory requirements of Draft GDC 2, GDC 4, and 10 CFR 50.46(b)(4).
3.0 TECHNICAL EVALUATION
3.1 Background By letter dated March 16, 1992 (Legacy Accession No. 9203230334), the licensee requested to eliminate from the design basis the dynamic effects of postulated pipe ruptures in the reactor coolant loop piping for DCPP, Unit Nos. 1 and 2. The request was based on a LBB analysis performed by the licensee in accordance with GDC 4. By letter dated March 2, 1993 (Legacy Accession No. 9302250062), the NRC concluded that the licensee's LBB analysis complies with GDC 4 and the probability of large breaks occurring in the RCS line is sufficiently low that the dynamic effects associated with the postulated pipe breaks need not be a design basis. Hence, the NRC staff approved the licensee's request.
The original DCPP, Units 1 and 2, fuel structural analysis for the 17x17 Optimized Fuel Assembly was based on the NRG-approved methodology in Westinghouse Electric Company LLC's proprietary topical report WCAP-9401-P-A, "Verification and Testing Analysis of the 17x17 Optimized Fuel Assembly," dated August 1981. In 2010, the licensee performed a new fuel assembly structural analysis, which incorporated the application of LBB based loads as an input to the analysis. The use of this new LBS-based input in the revised analysis modified the location of the break from the reactor vessel inlet to the limiting RCS branch line piping location (e.g., accumulator line, pressurized surge line, and residual heat removal system line).
However, the change in break location in the 2010 analysis was not addressed by the 10 CFR 50.59 evaluation performed in support of activities associated with installation of an Integrated Head Assembly. The licensee determined that the revised licensing basis regarding the fuel assembly structural analysis with LBB based inputs could not be incorporated into DCPP, Units 1 and 2 UFSAR without prior approval of the NRC staff, and, therefore, submitted this LAR for approval.
3.2 Licensee's Evaluation In its application dated October 2, 2014, the licensee stated, in part, that:
The fuel assembly structural analyses were performed to satisfy SRP Section 4.2, Appendix A [Reference 2 of the application dated October 2, 2014].
One of the objectives of the fuel assembly structural analyses is to confirm that a fuel coolable geometry is maintained.
The fuel assembly structural analyses were performed using the methods described in UFSAR Section 3.7.3.15.2, "Fuel Assembly Evaluation," including running a detailed fuel assembly model using the Westinghouse WEGAP code.
This detailed model (depicted in UFSAR Figure 3.7-27F) conservatively represents an entire row of full-length fuel assemblies (15 total).
The DCPP LBB analyses were performed in accordance with GDC 4. GDC 4 states: "However dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping." The exclusion of the dynamic effects associated with postulated pipe rupture of the main coolant loop piping was allowed generically for nuclear power units with the NRG approval of LBB in [10 CFR 50, Appendix A, GDC 4] and specifically for DCPP with the NRG approval of LBB in
[the NRG letter dated March 2, 1993 (Legacy Accession No. 9302250062)].
Section C.2 of the NRG Leak-Before-Break Knowledge Management document
[dated May 29, 2007 (ADAMS Accession No. ML092430585)] discusses that the decompression waves within the intact portion of the piping system are a local dynamic effect. The fuel is inside the reactor vessel and therefore it is within an intact portion of the piping systems. This allows the consideration of the dynamic effects associated with the postulated pipe rupture of the RCS branch line piping (e.g., accumulator line, pressurizer surge line, [residual heat removal] line), for the fuel assembly structural analyses.
LBB was applied when determining the local dynamic effects associated with postulated pipe ruptures in RCS branch line piping that resulted in the most limiting LOCA loads, as an input to the fuel assembly structural analyses. The accumulator line break in the cold leg was determined by analysis to be the RCS branch line piping rupture that resulted in the highest impact force of the fuel grids. Therefore, the LOCA loads for an accumulator line break were used as input to the fuel assembly structural analyses.
The [best-estimate large break loss-of-coolant accident (BELOCA)] analyses were performed in accordance with the NRG approved methodology contained in WCAP-12945-P-A, and WCAP-12945-P-A, Addendum 1-A, Revision 0 (for Unit 1) [ADAMS Accession No. ML043550246] and WCAP-16009-P-A, Revision 0 (for Unit 2) [ADAMS Accession No. ML080640898]. The BELOCA analyses were approved by the NRG for Diablo Canyon Units 1 and 2 in License
Amendments 191 and 192, respectively [ADAMS Accession Nos. ML063040036 and ML063380018].
The BELOCA methodologies confirmed that the criteria of 10 CFR 50.46(b)(1) through (b)(4) are met, and considered a break equivalent in size to double ended rupture of the main coolant loop piping, consistent with 10 CFR 50.46(c).
The NRC approved BELOCA methodology also addresses grid deformation in fuel assemblies that are located on the periphery of the core, if it is determined to occur, based on the fuel structural integrity analyses [WCAP-12945-P-A, "Code Qualification Document for Best Estimate LOCA Analysis," Appendix C:
"Requests for Additional Information" [non-publicly available]. However no grid deformation was calculated to occur in the fuel structural integrity analyses that were performed for DCPP.
The fuel structural integrity analyses considered the LOCA loads associated with an RCS branch line break, consistent with the approval of LBB for DCPP that excluded the dynamic effects associated with a postulated rupture in the main coolant loop piping.
In its letter dated October 2, 2014, the licensee further stated, in part, that:
The LOCA load for DCPP was based on the loads associated with a postulated pipe rupture in the accumulator line, the RCS branch line that produced the most limiting loads. The earthquake load for DCPP was based on the loads associated with the Hosgri earthquake. The Hosgri earthquake grid loads were determined by analysis to be more limiting than the Double Design Earthquake and Design Earthquake loads. The limiting LOCA load, combined with the Hosgri earthquake load using the Square Root Sum of Squares (SRSS) method did not result in any grid deformation for DCPP. Additional details on the grid impact forces are contained in Attachment 1 to [the licensee's letter dated October 2, 2014). Therefore, the coolable geometry criterion of SRP 4.2 was satisfied, as well as the coolable geometry criterion of 10 CFR 50.46(b)(4).
The BELOCA analyses do not credit control rod insertion to shutdown the reactor. A reactor shutdown is caused by the voiding associated with a large break LOCA.
However, one of the other objectives of the fuel structural integrity analyses performed to satisfy SRP 4.2 is to confirm control rod insertability. Since the fuel structural integrity analyses that were performed for DCPP did not calculate any grid deformation, control rod insertability was confirmed.
Based on the above discussion, in its letter dated October 2, 2014, the licensee concluded:
The BELOCA analyses that were performed for DCPP confirmed that the criteria of 10 CFR 50.46(b)(1) through (b)(4) are met, and considered a break equivalent in size to double ended rupture of the main coolant loop piping, consistent with 10 CFR 50.46(c).
LBB was applied to the local dynamic effects associated with postulated pipe ruptures in the RCS branch line piping that resulted in the most limiting LOCA loads, as an input to the fuel assembly structural analyses, based on the elimination of the dynamic effects associated with postulated pipe ruptures in the main loop piping, as approved in [the NRC's letter dated March 2, 1993].
The LOCA load for DCPP was based on the loads associated with a postulated pipe rupture of the RCS branch line that produced the most limiting loads. The limiting LOCA load, combined with the Hosgri earthquake load using the SRSS method did not result in any grid deformation for DCPP. Therefore, the coolable geometry criterion of SRP 4.2 was satisfied, as well as the coolable geometry criterion of 10 CFR 50.46(b)(4).
Attachment 4 [of the Enclosure to the licensee's letter dated October 2, 2014],
Summary of the Diablo Canyon Units 1 and 2 Fuel Structural Analysis Results for the Hosgri Earthquake and the Accumulator Line Break, provides a summary of the results of the DCPP Fuel Structural Analysis.
In its LAR dated October 2, 2014, the licensee also requested to utilize the LBB based fuel assembly structural analysis results to satisfy, in part, the ECCS performance criterion in 10 CFR 50.46(b)(4), "Coolable geometry." The results of the fuel assembly structural analysis are used as an input to the ECCS analyses to determine compliance with 10 CFR 50.46(b)(4) coolable geometry requirements in accordance with the NRG-approved Westinghouse BELOCA methodology (WCAP-12945-P-A, Appendix C).
The licensee performed its original fuel structural analyses for the original Vantage 5 fuel assembly based on the break locations on a reactor pressure vessel inlet and a reactor coolant pump outlet pipe break (i.e., primary loop pipe breaks). However, for the proposed license amendment request, the licensee postulated an accumulator line break (the limiting RCS branch line break), in lieu of a primary loop pipe break, as an input to the fuel structural analyses for the new Westinghouse 17 x 17 Optimized Fuel Assembly.
In its LAR dated October 2, 2014, the licensee stated that it combined the LOCA loads from a postulated rupture of the accumulator branch line piping with the Hosgri earthquake load using the square root sum of the squares (SRSS) method per SRP Section 4.2. The licensee's analysis did not result in any grid deformation, and, therefore, concluded that the coolable geometry ECCS performance criterion of 10 CFR 50.46(b)(4) was satisfied.
3.3 NRC Staff Evaluation By letter dated March 2, 1993 (see Section 3.1 of this Safety Evaluation), the NRC approved the application of LBB for the RCS (primary loop) piping at DCPP, Units 1 and 2. The NRC staff
notes that the ECCS at DCPP, Units 1 and 2, was originally designed and its performance analyzed using break locations in the primary loop piping, not the RCS branch lines. Based on the above, the ECCS at DCPP, Units 1 and 2, retains its designed heat removal and mass replacement capacity based on postulated pipe ruptures of primary loop piping, and that ECCS is designed without considering the benefit of LBB even though the primary loop piping is approved for LBB. In the original ECCS analysis, the licensee combined the seismic loads with the LOCA loads as inputs to the fuel structural analysis. The results of the original fuel structural analysis showed that the original fuel assembly (the Vantage 5 fuel assembly) satisfied the regulations regarding core geometry and ECCS performance following a LOCA.
To analyze the current fuel assembly (17 x 17 Optimized Fuel Assembly), the licensee performed two sets of calculations postulating a LOCA occurring in the primary loop piping and in the accumulator line.
The licensee calculated the maximum impact forces on the fuel assembly (or fuel grids) in the current fuel assembly based on pipe breaks occurring in the primary loop piping (cold leg). For this set of calculations, the licensee did not take credit for LBB. As shown in Tables 3-1 to 3-4 of Attachment 4 to the licensee's letter dated October 2, 2014 (contains proprietary information, and is therefore withheld from public disclosure), the resulting impact loads (LOCA loads) on the fuel grids based on primary loop pipe ruptures are bounded by the allowable loads (i.e., the crush strength limit of the fuel grid). However, in this set of calculations, the licensee did not combine the resulting impact loads (LOCA loads) with the seismic loads to be applied to the fuel grids as detailed in SRP Section 4.2. The NRC staff noted that if the resulting impact loads (LOCA loads) are combined with the seismic loads as detailed in SRP Section 4.2, the total loads on the fuel grids would exceed the allowable loads. This means that the post-LOCA fuel structure would not likely satisfy the coolable geometry requirements of 10 CFR 50.46(b)(4).
As an alternative, and the subject of the licensee's LAR, the licensee took credit for LBB by excluding the dynamic effects of a primary loop piping rupture in performing its fuel structural analysis for the current fuel assembly. As noted in Section 3.1 of this SE, the NRC approved the application of LBB for the RCS primary loop piping at DCPP, Units 1 and 2, in March 1993.
Accordingly, the licensee excluded the dynamic effects of a primary loop rupture and assumed an RCS piping branch line rupture in order to calculate LOCA loads. The licensee determined that the branch line piping with the most limiting LOCA loads would be the RCS accumulator line location because it will provide the maximum impact loads on the fuel assembly structure.
The licensee combined the seismic loads with the LOCA loads based on SRP Section 4.2 and was able to demonstrate that for an RCS accumulator line break, the fuel assembly structural analysis satisfies 10 CFR 50.46(b)(4).
While reviewing both sets of calculations, the NRC staff evaluated if the use of an accumulator line break in lieu of a primary loop pipe break for the fuel assembly structural analysis is acceptable under the LBB regulation. The NRC staff's primary consideration was whether a primary loop pipe break could be considered as local dynamic effects as opposed to global dynamic effects. On April 6, 1998, the NRC published a Federal Register notice (53 FR 11311),
"Leak-Before-Break Technology; Solicitation of Public Comment on Additional Applications,"
which permitted consideration of LBB in the design basis of relevant safety-related structures, systems, and components (SSCs) if the dynamic effects of a pipe break would result in local, not global, phenomena. That is, if the dynamic effects of a primary loop pipe break are
considered "local" and not "global" phenomena, the licensee may take credit for LBB and therefore, may not need to postulate a primary loop pipe break to analyze the fuel assembly.
In 1975, the NRC staff identified an unresolved safety issue with respect to the newly defined asymmetric blow down loads that are generated by postulating double-ended ruptures of pressurized-water reactor (PWR) primary loop piping. The NRC noted that in the event of a postulated instantaneous, double-ended, offset rupture at a reactor vessel nozzle, asymmetric loading could result from forces induced on the reactor internals by transient differential pressures across the core barrel and by forces on the vessel caused by transient differential pressures in the reactor cavity. Such differential pressures, although of short duration, could place a significant load on the supports of the reactor vessel and on other components such as the fuel assembly, possibly affecting their integrity. The NRC designated the problem as Unresolved Safety Issue A-2 and documented the issue in NUREG-0609, "Asymmetric Slowdown Loads on PWR Primary Systems: Resolution of Generic Task Action Plan A-2,"
dated January 1981 (ADAMS Accession No. ML13255A427). NUREG-0609 provides guidance and acceptance criteria for the structural analysis, which includes the specification of mechanical response (loads, deflections, and stresses) of the primary coolant piping, major components and their supports, and reactor internals and fuel assembly.
To address Unresolved Safety Issue A-2, the industry proposed a concept of LBB, which states that the piping has sufficient fracture toughness to resist uncontrollable crack propagation such that the pipe will leak first for an extended period of time during which the operators can safely shutdown the plant before the pipe ruptures, and submitted topical reports for NRC review and approval, as detailed below.
On February 1, 1984, the NRC published the SEs for two Westinghouse Topical Reports (TRs)
(WCAP-9558, Revision 2, "Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack," May 1981, and WCAP-9787, "Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation," May 1981 (Legacy Accession No. 8402130257). These two TRs provided a technical basis for LBB and dealt with the elimination of postulated pipe breaks in PWR primary loop piping as part of NRC Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in the Primary Main Loops," dated February 1, 1984 (ADAMS Accession No. ML031150562). The NRC SEs closed Unresolved Safety Issue A-2 and permitted the asymmetric blowdown loads resulting from double-ended pipe breaks in primary coolant loop piping to be excluded from the design basis in the application of LBB.
Subsequently, the proposed rule to modify GDC 4, dated July 1, 1985 (50 FR 27006), stated that the direct dynamic effects of pipe rupture are missile generation, pipe whipping, pipe-break reaction forces, and discharging fluids. The influence of discharging fluids includes impingement forces, decompression waves within the intact portion of the piping system, and pressurization in cavities, subcompartments, and compartments.
Since the inclusion of LBB in the GDC, licensees and industry groups have requested clarification regarding whether the NRG-approved LBB application is applicable to relax the design requirements of the containment, ECCS, and equipment qualification. In a Federal Register notice dated April 8, 1988 (53 FR 11311 ), the NRC stated that LBB is not applicable to
the design of containment, ECCS, equipment qualification of safety-related electrical and mechanical equipment because the dynamic effects of a pipe break on the design of containment, ECCS, and equipment qualification are deemed to be global, not local. The NRC clarified that when LBB is approved, ECCS must retain its heat removal and mass replacement capacity needed to mitigate the effects of postulated primary loop piping ruptures. This means that even if LBB is approved for a plant's primary loop piping, the ECCS must be designed to perform based on primary loop pipe ruptures. The NRC staff's determination is based on the position that LBB cannot be used to relax the design requirements of safety-related SSCs such as containment, ECCS, and equipment qualification because a pipe break would result in global dynamic effects, affecting the overall of the safety function of the containment, ECCS, and SSCs.
However, as discussed in Federal Register notice 53 FR 11311, the NRC staff clarified that local dynamic effects uniquely associated with pipe ruptures may be deleted from the design basis of containment systems, structures, and boundaries, from the design basis of ECCS hardware (such as pumps, valves, accumulators, and instrumentation), and from the design basis of safety-related electrical and mechanical equipment when LBB is accepted. "Local dynamic effects" are defined as missile generation, pipe whipping, pipe-break reaction forces, and discharging fluids.
The NRC staff further clarified the definition of local dynamic effects in Sections C.2 and C.3 of an NRC interoffice memorandum dated May 29, 2007 (ADAMS Accession No. ML092430585).
In the memorandum, the NRC staff explained that LBB may be used, upon NRC staff prior approval, so that dynamic effects from the pipe breaks such as decompression waves within the intact portion of the piping system and pressurization in cavities may be excluded from the design basis.
The NRC staff concludes that a postulated primary loop break affecting the fuel assembly is considered a local, not global, dynamic effect. As stated above, the direct dynamic effects of primary loop pipe rupture are missile generation, pipe whipping, pipe-break reaction forces, and discharging fluids. The influence of discharging fluids includes impingement forces, decompression waves within the intact portion of the piping system, and pressurization in cavities. The fuel assembly is located inside the reactor vessel. The reactor vessel and the primary loop piping are an enclosed system. If the rupture occurs on a primary loop pipe, the loading on the fuel assembly would come from the hydrodynamic forces generated by the decompression waves and pressurization in cavities fluids caused by the discharging fluid from the break location. The impact of the hydrodynamic forces on the fuel assembly would be localized within the confines of the reactor vessel. As such, the dynamic effects from a primary loop pipe break could be considered as a local phenomenon.
The NRC staff concludes that the impact of a primary loop pipe break on the fuel structure is considered local dynamic effects under the requirements of GDC 4 and that the licensee can take credit for LBB in the licensing basis. The licensee does not need to postulate any breaks in the primary loop piping to analyze the fuel assembly structure. Thus, the NRC staff concludes that it is acceptable that the licensee selected the accumulator line break to analyze the structural integrity of the fuel assembly. The NRC staff notes that in the LAR dated October 2, 2014, the licensee did not request to reduce or relax performance of ECCS such as pump capacity, mass flow rate, or any other system changes.
By letter dated March 2, 1993, the NRC staff approved LBB application for the RCS (primary loop) at DCPP, Units 1 and 2, and no further evaluation regarding the use of LBB at DCPP is needed. Based on the above discussion and the information provided by the licensee in its letter dated October 2, 2014, the NRC staff concludes that the fuel assembly structural analyses is consistent with the guidance of SRP Section 4.2, Appendix A and complies with the requirements of Draft GDC 2, GDC 4, and 10 CFR 50.46(b) and the proposed changes to the DCPP UFSAR are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official, Mr. S. Hsu, was notified on October 8, 2015, of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on March 3, 2015 (80 FR 11496). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Balwant K. Singal, NRR/DORL/LPL4-1 Andrea George, NRR/DORL/LPL4-1 John Tsao, NRR/DE/EPNB Siva Lingam, NRR/DORL/LPL4-1 Date: :=-December 3 , 2015
ML15281A164 *Previously concurred OFFICE NRR/DORL/LPL4-1 /PM NRR/DORL/LPL4-1 /PM NRR/DORL/PM/LPL4-1 NRR/DORL/LPL4-1 /LA NRR/DE/EPNB/BC NAME BSingal* AGeorge* Slingam* JBurkhardt* DAiiey*
DATE 10/26/15 11/9/15 11/6/15 10/9/15 10/26/15 OFFICE NRR/DSS/SRXB/BC OGC NLO w/comments NRR/DORL/LPL4-1 /BC NRR/DORL/LPL4-1 /PM NAME CJackson* CKanatas* RPascarelli Sling am DATE 11/6/15 11 /19/15 12/03/15 12/03/15