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| number = ML13261A264 | | number = ML13261A264 | ||
| issue date = 09/25/2013 | | issue date = 09/25/2013 | ||
| title = | | title = Redacted Version, Issuance of Amendment No. 195, Revise Criticality Safety Analysis and Technical Specification 4.3.1, Criticality, and Delete Spent Fuel Pool Loading Criteria License Condition | ||
| author name = Wang A | | author name = Wang A | ||
| author affiliation = NRC/NRR/DORL/LPLIV | | author affiliation = NRC/NRR/DORL/LPLIV | ||
| addressee name = | | addressee name = | ||
| Line 9: | Line 9: | ||
| docket = 05000416 | | docket = 05000416 | ||
| license number = NPF-029 | | license number = NPF-029 | ||
| contact person = Wang A | | contact person = Wang A | ||
| case reference number = TAC ME7111 | | case reference number = TAC ME7111 | ||
| document type = Letter, License-Operating (New/Renewal/Amendments) DKT 50, Safety Evaluation | | document type = Letter, License-Operating (New/Renewal/Amendments) DKT 50, Safety Evaluation | ||
| Line 18: | Line 18: | ||
=Text= | =Text= | ||
{{#Wiki_filter:OFFICIAL USE ONLY PROPRI&TARY UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 September 25,2013 Vice President, Operations Entergy Operations, Inc. Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 GRAND GULF NUCLEAR STATION, UNIT 1-ISSUANCE OF AMENDMENT RE: CHANGES TO THE NUCLEAR CRITICALITY SAFETY ANALYSIS (TAC NO. ME7111) | {{#Wiki_filter:OFFICIAL USE ONLY PROPRI&TARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 25,2013 Vice President, Operations Entergy Operations, Inc. | ||
Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 | |||
==SUBJECT:== | |||
GRAND GULF NUCLEAR STATION, UNIT 1-ISSUANCE OF AMENDMENT RE: CHANGES TO THE NUCLEAR CRITICALITY SAFETY ANALYSIS (TAC NO. ME7111) | |||
==Dear Sir or Madam:== | ==Dear Sir or Madam:== | ||
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 195 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS). This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 9, 2011, as supplemented by letters dated September 8 and November 23,2010; March 9, April 21 , May 3, and November 21,2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16, 2013. The letters dated September 8 and November 23, 2010, and March 9 and May 3, 2011, are incorporated by reference in the September 9, 2011, license amendment request (LAR) as allowed by Section 50.32, "Elimination of replication," | |||
of Title 10 of the Code of Federal Regulations (10 CFR). The amendment approves | The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 195 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS). | ||
: 1) additional requirements for the spent fuel and new fuel storage racks in TS 4.3.1, "Criticality," | This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 9, 2011, as supplemented by letters dated September 8 and November 23,2010; March 9, April 21 , May 3, and November 21,2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16, 2013. The letters dated September 8 and November 23, 2010, and March 9 and May 3, 2011, are incorporated by reference in the September 9, 2011, license amendment request (LAR) as allowed by Section 50.32, "Elimination of replication," of Title 10 of the Code of Federal Regulations (10 CFR). | ||
The amendment approves: 1) additional requirements for the spent fuel and new fuel storage racks in TS 4.3.1, "Criticality," 2) a revision to the current Nuclear Criticality Safety Analysis, which is described in GGNS Updated Final Safety Analysis Report Sections 9.1.1, "New Fuel Storage," and 9.1.2, "Spent Fuel Storage," to reflect changes resulting from the extended power uprate, and 3) deletion of the spent fuel pool loading criteria Operating License condition. | |||
and 9.1.2, "Spent Fuel Storage," | The NRC has determined that the related safety evaluation, provided in Enclosure 2, contains proprietary information pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, the NRC staff has also prepared a non-proprietary version of the NOTICE: Enclosure 2 to this letter contains Proprietary Information. Upon separation from | ||
to reflect changes resulting from the extended power uprate, and 3) deletion of the spent fuel pool loading criteria Operating License condition. | : Enclosure 2, this letter is DECONTROLLED. | ||
The NRC has determined that the related safety evaluation, provided in Enclosure 2, contains proprietary information pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
Accordingly, the NRC staff has also prepared a non-proprietary version of the NOTICE: Enclosure 2 to this letter contains Proprietary Information. | |||
Upon separation from : Enclosure 2, this letter is DECONTROLLED. | OFFICIAL Y81i ONbV PROPRIIiTARY INFO~MATlON | ||
OFFICIAL USE ONLY PROPRIETARY OFFICIAL Y81i ONbV PROPRIIiTARY -2 safety evaluation, which is provided in Enclosure | -2 safety evaluation, which is provided in Enclosure 3. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. | ||
Sincerely, Alana::pr~:t.r Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416 | |||
Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416 | |||
==Enclosures:== | ==Enclosures:== | ||
: 1. Amendment No. 195 to NPF-29 2. Proprietary Safety Evaluation | : 1. Amendment No. 195 to NPF-29 | ||
: 3. Non-proprietary Safety Evaluation cc w/Enclosures 1 and 3: Distribution via Listserv OFFICIAL Y8E ONLY PROPRIETARY ENCLOSURE AMENDMENT NO. 195 TO FACILITY LICENSE NO. ENTERGY OPERATIONS, INC" ET GRAND GULF NUCLEAR STATION, UNIT DOCKET NO. | : 2. Proprietary Safety Evaluation | ||
UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 ENTERGY OPERATIONS, INC. SYSTEM ENERGY RESOURCES, INC. SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPI, INC. DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 195 License No. NPF-29 The Nuclear Regulatory Commission (NRC, the Commission) has found that: The application for amendment by Entergy Operations, Inc. (the licensee), | : 3. Non-proprietary Safety Evaluation cc w/Enclosures 1 and 3: Distribution via Listserv OFFICIAL Y8E ONLY PROPRIETARY INFORMATION | ||
dated September 9, 2011, as supplemented by letters dated September 8 and November 23, 2010; March 9, April 21, May 3, and November 21, 2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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. | |||
-2 Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 195 are hereby incorporated in the license. | ENCLOSURE 1 AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC" ET AL. | ||
Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. In addition, Paragraph 2.C.(45) of Facility Operating License No. NPF-29 is hereby amended to read as follows: Deleted. This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance. | GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 | ||
In addition, the licensee will maintain a minimum distance of 12 inches between any fuel stored in the Control Blade/Defective Fuel Storage Rack (Module H1) and in the surrounding high-density spent fuel pool storage racks as described in the licensee's letter dated September 5, 2013, and the NRC staff's safety evaluation for this amendment. | |||
In addition, the licensee shall include the revised information in the Grand Gulf Nuclear Station Updated Final Safety Analysis Report in the next periodic update in accordance with 10 CFR 50.71(e), | UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 ENTERGY OPERATIONS, INC. | ||
as described in the licensee's application dated September 9, 2011, as supplemented by letters dated September 8 and November 23,2010; March 9, April 21, May 3, and November 21, 2011; April 18, October 1, and October 22, 2012; and July 2, September 5, and September 16, 2013, and the NRC staff's safety evaluation for this amendment. | SYSTEM ENERGY RESOURCES, INC. | ||
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor licensing Office of Nuclear Reactor Regulation | SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPI, INC. | ||
DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 195 License No. NPF-29 | |||
: 1. The Nuclear Regulatory Commission (NRC, the Commission) has found that: | |||
A. The application for amendment by Entergy Operations, Inc. (the licensee), dated September 9, 2011, as supplemented by letters dated September 8 and November 23, 2010; March 9, April 21, May 3, and November 21, 2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and 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. | |||
-2 | |||
: 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-29 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. 195 are hereby incorporated in the license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. | |||
In addition, Paragraph 2.C.(45) of Facility Operating License No. NPF-29 is hereby amended to read as follows: | |||
(45) Deleted. | |||
: 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 will maintain a minimum distance of 12 inches between any fuel stored in the Control Blade/Defective Fuel Storage Rack (Module H1) and in the surrounding high-density spent fuel pool storage racks as described in the licensee's letter dated September 5, 2013, and the NRC staff's safety evaluation for this amendment. In addition, the licensee shall include the revised information in the Grand Gulf Nuclear Station Updated Final Safety Analysis Report in the next periodic update in accordance with 10 CFR 50.71(e), as described in the licensee's application dated September 9, 2011, as supplemented by letters dated September 8 and November 23,2010; March 9, April 21, May 3, and November 21, 2011; April 18, October 1, and October 22, 2012; and July 2, September 5, and September 16, 2013, and the NRC staff's safety evaluation for this amendment. | |||
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor licensing Office of Nuclear Reactor Regulation | |||
==Attachment:== | ==Attachment:== | ||
Changes to the Facility Operating License No. NPF-29 and the Technical Specifications Date of Issuance: | Changes to the Facility Operating License No. NPF-29 and the Technical Specifications Date of Issuance: September 25, 2013 | ||
September 25, 2013 ATTACHMENT TO LICENSE AMENDMENT NO, FACILITY OPERATING LICENSE NO, DOCKET NO, Replace the following pages of the Facility Operating License No. NPF-29 and the Appendix A, Technical Specifications 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 Remove 4 4 16a 16a 16b 16b Technical Specifications Remove 4.0-1 4.0-1 4.0-2 4,0-2 4.0-2a SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54. In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10CFR 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: Maximum Power Level Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 4408 megawatts thermal (100 percent power) in accordance with the conditions specified herein. Technical The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 195 are hereby incorporated into this license. | |||
Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. During Cycle 19, GGNS will conduct monitoring of the Oscillation Power Range Monitor (OPRM). During this the OPRM Upscale function (Function 2.f of Technical Specification Table 3.3.1.1 1) will be disabled and operated in an "indicate only" mode and technical specification requirements will not apply to this function. | ATTACHMENT TO LICENSE AMENDMENT NO, 195 FACILITY OPERATING LICENSE NO, NPF-29 DOCKET NO, 50-416 Replace the following pages of the Facility Operating License No. NPF-29 and the Appendix A, Technical Specifications 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 Remove 4 4 16a 16a 16b 16b Technical Specifications Remove 4.0-1 4.0-1 4.0-2 4,0-2 4.0-2a | ||
During such time, Backup Stability Protection measures will be implemented via GGNS procedures to provide an alternate method to detect and suppress reactor core thermal hydraulic instability oscillations. | |||
Once monitoring has been successfully completed, the OPRM Upscale function will be enabled and technical specification requirements will be applied to the function; no further operating with this function in an "indicate only" mode will be conducted. Amendment No. 195 The first performance of the periodic assessment of eRE habitability, Specification 5.5.13.c. | (b) SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54. | ||
(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from March 2005, the date of the most recent successful tracer gas test, as stated in the June 30, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years. The first performance of the periodic assessment of the CRE boundary, Specification 5.5.13.d, shall be within the next 18 months, plus the 136 days allowed by SR 3.0.2, as measured from the date of issuance of this amendment. Leak rate tests associated with Surveillance Requirements (SR) 3.6.1.1.1, 3.6.1.3.5, and 3.6.1.3.9, as required by TS 5.5.12 and in accordance with 10 CFR 50, Appendix J, Option B, and SRs 3.6.5.1.1 and 3.6.5.1.2 are not required to be performed until their next scheduled performance dates. The tests will be performed at the EPU calculated peak containment pressure or within EPU drywell bypass leakage limits, as appropriate. Deleted. | In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility. | ||
16a Amendment No. +/-9+/-, 195 This license condition provides for monitoring, evaluating, and taking prompt action in response to potential adverse flow effects as a result of power uprate operation on plant structures, | C. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10CFR 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 Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 4408 megawatts thermal (100 percent power) in accordance with the conditions specified herein. | |||
This plan shall include: Criteria for comparison and evaluation of projected strain and acceleration with on-dryer instrument data. Acceptance limits developed for each on-dryer strain gauge and accelerometer. Tables of predicted dryer stresses at CLTP, strain amplitudes and PSDs at strain gauge locations, acceleration amplitudes and PSDs at accelerometer locations, and maximum stresses and locations. PAT plan shall provide correlations between measured accelerations and strains and the corresponding maximum stresses. | (2) Technical ~pecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 195 are hereby incorporated into this license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. | ||
The PAT plan shall be submitted to the NRC Project Manager no later than 10 before start-up. GGNS shall monitor the main steam line (MSL) strain gages and on-dryer instrumentation at a minimum of three power levels up to 3898 MWt. Based on a comparison of projected and measured strains and accelerations, GGNS will assess whether the dryer acoustic and structural models have adequately captured the response significant to peak stress projections. Amendment No. 195 Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The for Grand Gulf Nuclear Station is located in Claiborne County, Mississippi on the east bank of the Mississippi River, approximately 25 miles south of Vicksburg and 37 miles north-northeast of Natchez. | During Cycle 19, GGNS will conduct monitoring of the Oscillation Power Range Monitor (OPRM). During this time, the OPRM Upscale function (Function 2.f of Technical Specification Table 3.3.1.1 1) will be disabled and operated in an "indicate only" mode and technical specification requirements will not apply to this function. During such time, Backup Stability Protection measures will be implemented via GGNS procedures to provide an alternate method to detect and suppress reactor core thermal hydraulic instability oscillations. Once monitoring has been successfully completed, the OPRM Upscale function will be enabled and technical specification requirements will be applied to the function; no further operating with this function in an "indicate only" mode will be conducted. | ||
The exclusion area boundary shall have a radius of 696 meters from the centerline of the reactor. | 4 Amendment No. 195 | ||
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 800 fuel assemblies. | |||
Each assembly shall consist of a matrix of Zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide ( | (b) The first performance of the periodic assessment of eRE habitability, Specification 5.5.13.c. (ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from March 2005, the date of the most recent successful tracer gas test, as stated in the June 30, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years. | ||
4.2.2 Control Rod Assemblies The reactor core shall contain 193 cruciform shaped control rod assemblies. | (c) The first performance of the periodic assessment of the CRE boundary, Specification 5.5.13.d, shall be within the next 18 months, plus the 136 days allowed by SR 3.0.2, as measured from the date of issuance of this amendment. | ||
The control material shall be boron carbide or hafnium metal, or both. 4.3 Fuel Storage 4.3.1 Criticality | (44) Leak rate tests associated with Surveillance Requirements (SR) 3.6.1.1.1, 3.6.1.3.5, and 3.6.1.3.9, as required by TS 5.5.12 and in accordance with 10 CFR 50, Appendix J, Option B, and SRs 3.6.5.1.1 and 3.6.5.1.2 are not required to be performed until their next scheduled performance dates. The tests will be performed at the EPU calculated peak containment pressure or within EPU drywell bypass leakage limits, as appropriate. | ||
GRAND Amendment No. | (45) Deleted. | ||
Design Features 4.0 DESIGN FEATURES 4.3.1.1 (continued) | 16a Amendment No. ~, ~, +/-9+/-, 195 | ||
: d. Fuel assemblies having | |||
: e. Region II racks are controlled as follows: | (46) This license condition provides for monitoring, evaluating, and taking prompt action in response to potential adverse flow effects as a result of power uprate operation on plant structures, systems, and components (including verifying the continued structural integrity of the steam dryer) for power ascension from the CLTP (3898 MWt) to the EPU level of 4408 MWt (or 113 percent of CLTP or 115 percent of OLTP). | ||
: 1. cells with any Boraflex which has received a gamma dose in excess of 2.3E10 rads or which has a Boron-l0 areal less than 0.0165, which are the Spent Fuel Pool Rack Boraflex are treated as II Storage cells face-acent either restricted from fuel the isolated cells as a minimum (i.e., | (a) The following requirements are placed on operation of the facility before and during the power ascension to 3898 MWt: | ||
GRAND 4.0 2 Amendment No. 195 Design Features 4.0 4.0 DESIGN FEATURES (continued) | : 1. GGNS shall provide a Power Ascension Test (PAT) Plan for the Steam Dryer testing. This plan shall include: | ||
The new fuel storage racks are designed and shall be maintained with: keff S 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.1 of the UFSARj A nominal fuel assembly center to center storage spacing of 6.535 inches within rows and 11.875 inches between rows in the new fuel storage racks. Fuel assemblies having a maximum k-infinity of 1.26 in the normal reactor core configuration at cold conditions; | * Criteria for comparison and evaluation of projected strain and acceleration with on-dryer instrument data. | ||
: d. assemblies having a maximum nominal U-235 enrichment of weight percent. | * Acceptance limits developed for each on-dryer strain gauge and accelerometer. | ||
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 202 ft 5.25 inches. 4.3.3 Capacity The spent fuel storage pool shall be maintained with a storage capacity limited to no more than 4348 fuel assemblies. | * Tables of predicted dryer stresses at CLTP, strain amplitudes and PSDs at strain gauge locations, acceleration amplitudes and PSDs at accelerometer locations, and maximum stresses and locations. | ||
No more than 800 fuel assemblies may be stored in the upper containment pool. GRAND 4.0 2a Amendment No. 195 ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR RELATED TO AMENDMENT NO. 195 FACILITY OPERATING LICENSE NO. ENTERGY OPERATIONS, INC., ET GRAND GULF NUCLEAR STATION, UNIT DOCKET NO. (NON-PROPRI OfflGIAl us. ONlY PROPRI.TARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL. GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 | The PAT plan shall provide correlations between measured accelerations and strains and the corresponding maximum stresses. The PAT plan shall be submitted to the NRC Project Manager no later than 10 days before start-up. | ||
: 2. GGNS shall monitor the main steam line (MSL) strain gages and on-dryer instrumentation at a minimum of three power levels up to 3898 MWt. Based on a comparison of projected and measured strains and accelerations, GGNS will assess whether the dryer acoustic and structural models have adequately captured the response significant to peak stress projections. | |||
16b Amendment No. ~, 195 | |||
Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for Grand Gulf Nuclear Station is located in Claiborne County, Mississippi on the east bank of the Mississippi River, approximately 25 miles south of Vicksburg and 37 miles north-northeast of Natchez. The exclusion area boundary shall have a radius of 696 meters from the centerline of the reactor. | |||
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 800 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2 ) as fuel material, and water rods. Limited sUbstitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. | |||
4.2.2 Control Rod Assemblies The reactor core shall contain 193 cruciform shaped control rod assemblies. The control material shall be boron carbide or hafnium metal, or both. | |||
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: | |||
: a. keff S 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the UFSAR; | |||
: b. A nominal fuel assembly center to center storage spacing of 6.26 inches in the storage racks. | |||
: c. Fuel assemblies having a maximum K-infinity of 1.26 in the normal reactor core configuration at cold conditions; (continued) | |||
GRAND GULF Amendment No. ~,+/-+4,195 | |||
Design Features 4.0 4.0 DESIGN FEATURES 4.3.1.1 (continued) | |||
: d. Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 percent; | |||
: e. Region II racks are controlled as follows: | |||
: 1. cells with any Boraflex which has received a gamma dose in excess of 2.3E10 rads or which has a Boron-l0 areal less than 0.0165, which are the Spent Fuel Pool Rack Boraflex are treated as II | |||
: 2. Storage cells face- acent either restricted from fuel the isolated cells or are conf as a minimum (i.e., additional cells may the Region II fuel storage configuration requirements in Figure 4.3 1. | |||
: 3. When a 4x4 array of cells is classified as Region II and face-adjacent to another II 4x4 storage array, the new II 4x4 array is to be blocked in the same 8-of-16 and at the same orientation as the acent Region II 4x4 storage Location Blocked to Prevent (continued) | |||
GRAND GULF 4.0 2 Amendment No. ~, 195 | |||
Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with: | |||
: a. keff S 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.1 of the UFSARj | |||
: b. A nominal fuel assembly center to center storage spacing of 6.535 inches within rows and 11.875 inches between rows in the new fuel storage racks. | |||
: c. Fuel assemblies having a maximum k-infinity of 1.26 in the normal reactor core configuration at cold conditions; | |||
: d. Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent. | |||
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 202 ft 5.25 inches. | |||
4.3.3 Capacity 4.3.3.1 The spent fuel storage pool shall be maintained with a storage capacity limited to no more than 4348 fuel assemblies. | |||
4.3.3.2 No more than 800 fuel assemblies may be stored in the upper containment pool. | |||
GRAND GULF 4.0 2a Amendment No. 195 | |||
ENCLOSURE 3 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL. | |||
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 (NON-PROPRI ETARY) | |||
OfflGIAl us. ONlY PROPRI.TARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL. | |||
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 | |||
==1.0 INTRODUCTION== | |||
By letter dated September 9, 2011 (Reference 1), as supplemented by letters dated November 21,2011 (Reference 2), April 18, 2012 (Reference 3), October 1,2012 (Reference 4), October 22,2012 (Reference 5), July 2,2013 (Reference 6), September 5,2013 (Reference 7), and September 16, 2013 (Reference 8), Entergy Operations. Inc. (Entergy, the licensee), submitted a request to amend Facility Operating License No. NPF-29 and revise the Grand Gulf Nuclear Station, Unit 1 (GGNS) Technical Specifications (TS). The letters dated September 8,2010 (Reference 9), and November 23,2010 (Reference 10); and March 9,2011 (Reference 11), April 21, 2011 (Reference 12), and May 3, 2011 (Reference 13), were incorporated by reference by the licensee in the September 9, 2011, license amendment request (LAR) as allowed by Section 50.32, "Elimination of replication," of Title 10 of the Code of Federal Regulations (10 CFR). Specifically, the proposed amendment requested to: | |||
: 1) Revise the current Nuclear Criticality Safety Analysis (NCSA), which is described in GGNS Updated Final Safety Analysis Report (UFSAR) Sections 9.1.1, "New Fuel Storage," and 9.1.2, "Spent Fuel Storage," to reflect changes resulting from the extended power uprate (EPU), | |||
: 2) Revise TS 4.3.1.1, "Criticality," to add requirements for two design parameters to spent and new fuel storage racks resulting from the EPU and add requirements to specify a spent fuel storage configuration for Region II cells (Region I and Region /I cells are described in Section 3.2 of this safety evaluation) to account for potential degradation of Boraflex, and | |||
: 3) Delete the spent fuel pool (SFP) loading criteria Operating License condition, which was approved for Cycle 19 or until a new NCSA is approved. | |||
Portions of the letters dated September 8 and November 23, 2010, April 21, 2011, October 1, 2012, and July 2,2013, contain proprietary information and, accordingly, have been withheld from public disclosure. Accordingly, the supplemental letters dated September 8 and November 23,2010; March 9, April 21, May 3, and November 21,2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16. 2013, provided additional OFFICIAL USE ONLY - PROPRIETARY INFORMATION | |||
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- 2 information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 24, 2012 (77 FR 3511). | |||
The following is the NRC staff's evaluation of the NCSA and associated TS changes. | |||
==2.0 REGULATORY EVALUATION== | |||
The following regulatory requirements and guidance documents were applicable to the NRC staff's review of the licensee's amendment request: | |||
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 62 states: | |||
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. | |||
-information that clarified the application, did not expand the scope of the application as originally | |||
2.0 REGULATORY EVALUATION The following regulatory requirements and guidance documents were applicable to the NRC staff's review of the licensee's amendment request: | |||
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 62 states: Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. | |||
Paragraph 50.68(b)(1) of 10 CFR requires: | Paragraph 50.68(b)(1) of 10 CFR requires: | ||
Plant procedures shall prohibit the handling and storage at anyone time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water. | |||
Paragraph 50.68(b)(2) of 10 CFR requires: | |||
The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used. | |||
Paragraph 50.68(b)(3) of 10 CFR requires: | |||
If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks | |||
Bounding Lattice Identified with 95 percent Probability and 95 percent Confidence The proposed technical specifications will require that the SCCG peak k-infinity must be no greater than 1.26. For a specific SCCG k-infinity, the in-rack keff value varies with the lattice. | Bounding Lattice Identified with 95 percent Probability and 95 percent Confidence The proposed technical specifications will require that the SCCG peak k-infinity must be no greater than 1.26. For a specific SCCG k-infinity, the in-rack keff value varies with the lattice. | ||
The bounding fuel assembly used in the analysis has a SCCG k-infinity that is slightly above the 1.26 k-infinity limit and its in-rack keff value must be bounding at a 95/95 probability and confidence. | The bounding fuel assembly used in the analysis has a SCCG k-infinity that is slightly above the 1.26 k-infinity limit and its in-rack keff value must be bounding at a 95/95 probability and confidence. It was not clear from the analysis description provided in NEDC-33621 P that the identified bounding lattice indeed bounded the other lattices at a 95/95 probability/confidence level. | ||
It was not clear from the analysis description provided in NEDC-33621 P that the identified bounding lattice indeed bounded the other lattices at a 95/95 probability/confidence level. Sufficient additional discussion was provided in the response to RAI-2 in Attachment 1 to GNRO-2013/0050 (Reference | Sufficient additional discussion was provided in the response to RAI-2 in Attachment 1 to GNRO-2013/0050 (Reference 6) to adequately demonstrate that the bounding lattice is indeed bounding at a 95/95 level. | ||
3.4.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties Fuel assembly manufacturing tolerances and uncertainties were evaluated using standard techniques based on keff sensitivity studies of parameter variation around the nominal model. | |||
The response to RAI-16 provided in Attachment 1 to GNRO-2012/00120 (Reference | During the review, it was noted that some potentially significant tolerances and uncertainties were not evaluated. | ||
The response to RAI-16 provided in Attachment 1 to GNRO-2012/00120 (Reference 4) included expanded analysis of tolerances and uncertainties. The response adequately addressed the NRC staff's concerns. | |||
The response adequately addressed the NRC staff's concerns. | 3.4.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on uranium-235 enrichment, fuel rod gadolinia content and distribution, and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also carry through to the spent nuclear fuel. Common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is more problematic. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup. | ||
3.4.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on uranium-235 enrichment, fuel rod gadolinia content and distribution, and various manufacturing tolerances. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. | |||
These tolerances and bounding values would also carry through to the spent nuclear fuel. Common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
The characterization of spent nuclear fuel is more problematic. | - 18 3.4.3.4 Burnup Uncertainty | ||
Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. | (( | ||
That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup. OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION | 11 3.4.3.5 Axial Apportionment of the Burnup or Axial Burnup Profile The standard BWR SCCG peak k-infinity analysis technique uses either two-dimensional models or a three-dimensional model with uniform axial burnup distributions. Generally, this is appropriate because the peak in limiting assembly reactivity occurs at lower burnups where the uniform axial burnup distribution is conservative. If one were to credit assembly burnup beyond the limiting peak reactivity burnup, at some assembly burnup value, the use of the uniform axial burnup would become non-conservative. | ||
-18 3.4.3.4 Burnup Uncertainty | |||
Generally, this is appropriate because the peak in limiting assembly reactivity occurs at lower burnups where the uniform axial burnup distribution is conservative. | |||
If one were to credit assembly burnup beyond the limiting peak reactivity burnup, at some assembly burnup value, the use of the uniform axial burnup would become non-conservative. | |||
The GGNS Region I analysis includes a departure from the standard method in that the explicit modeling of Boraflex gaps introduces additional axially dependent features (Le., Boraflex panel gaps) to fuel storage rack models. The concern is that these axially dependent features may impact the determination that it is conservative to use a uniform axial burnup distribution rather than a bounding axially-varying burnup distribution. | The GGNS Region I analysis includes a departure from the standard method in that the explicit modeling of Boraflex gaps introduces additional axially dependent features (Le., Boraflex panel gaps) to fuel storage rack models. The concern is that these axially dependent features may impact the determination that it is conservative to use a uniform axial burnup distribution rather than a bounding axially-varying burnup distribution. | ||
In response to this concern, additional analysis was performed and is documented in the response to RAI-1 provided in Attachment 1 to GNRO-2013/00050 (Reference 6). The discussion provided demonstrates that use of the uniform axial burnup distribution is still conservative for normal and credible abnormal conditions in this BWR SCCG peak k-infinity analysis. | In response to this concern, additional analysis was performed and is documented in the response to RAI-1 provided in Attachment 1 to GNRO-2013/00050 (Reference 6). The discussion provided demonstrates that use of the uniform axial burnup distribution is still conservative for normal and credible abnormal conditions in this BWR SCCG peak k-infinity analysis. | ||
3.4.3.6 Burnup History/Core Operating Parameters The reactivity of light-water reactor fuel varies with the conditions the fuel experiences in the reactor. | 3.4.3.6 Burnup History/Core Operating Parameters The reactivity of light-water reactor fuel varies with the conditions the fuel experiences in the reactor. This is particularly true for BWR fuel NCSA using the SCCG peak k-infinity analysis method. As a result of the usage of Gd 20 3 in fuel rods, fuel assembly reactivity initially increases as depletion of the gadolinium isotopes dominates the change in reactivity. At some point, the gadolinium isotopes are sufficiently depleted that the 235U depletion dominates the change in reactivity. The peak reactivity occurs at the transition point. The value of the in-rack keff at peak reactivity is affected by the reactor depletion parameters in several ways. | ||
This is particularly true for BWR fuel NCSA using the SCCG peak k-infinity analysis method. As a result of the usage of | Factors that lead to a more thermal neutron energy distribution cause the 155&157Gd and fission products to deplete more quickly and reduce plutonium generation. This causes the peak reactivity condition to be reached earlier, achieving a higher in-rack keff value. Increased water OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
At some point, the gadolinium isotopes are sufficiently depleted that the 235U depletion dominates the change in reactivity. | |||
The peak reactivity occurs at the transition point. The value of the in-rack keff at peak reactivity is affected by the reactor depletion parameters in several ways. Factors that lead to a more thermal neutron energy distribution cause the 155&157Gd and fission products to deplete more quickly and reduce plutonium generation. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
This causes the peak reactivity condition to be reached earlier, achieving a higher in-rack keff value. Increased water OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION | - 19 density and decreased void fraction lead to a more thermal neutron energy distribution and to lower fuel rod temperatures due to improved fuel rod cooling. | ||
-19 density and decreased void fraction lead to a more thermal neutron energy distribution and to lower fuel rod temperatures due to improved fuel rod cooling. | Factors that lead to a less thermal neutron energy distribution cause the 155&157Gd and fission products to be depleted more slowly and result in increased plutonium generation. Decreased water density, increased void fraction, and control rod usage all result in neutron energy spectrum hardening. | ||
Factors that lead to a less thermal neutron energy distribution cause the 155&157Gd and fission products to be depleted more slowly and result in increased plutonium generation. | The GGI\IS analysis included consideration of reactor operating parameter variation by (1) evaluating the bounding lattices at (( )) void, and (2) using sensitivity study results of the impact of (( | ||
Decreased water density, increased void fraction, and control rod usage all result in neutron energy spectrum hardening. | )) to generate bias terms that are applied to the maximum in-rack keff value. | ||
The GGI\IS analysis included consideration of reactor operating parameter variation by (1) evaluating the bounding lattices at | This approach is conservative because it effectively applies multiple conflicting conditions at the same time. For example, it is not realistic to include (( | ||
The uncertainty associated with calculating the residual gadolinium isotopes is included in the "burnup uncertainty" discussed above in Section 3.3.3.4. | )) All sensitivities that result in increased in-rack keff values are incorporated in the analysis as bias terms added to the in-rack maximum kelt value. | ||
3.4.4 Analysis of Abnormal Conditions | 3.4.3.7 Integral Burnable Absorbers GGNS uses fuel assemblies in which some of the fuel rods contain Gd 2 0 3 . Pin-by-pin fuel compositions are calculated using the TGBLA code and are used in MCNP models of the fuel storage racks. At peak reactivity, the residual gadolinium isotopes are credited. The uncertainty associated with calculating the residual gadolinium isotopes is included in the "burnup uncertainty" discussed above in Section 3.3.3.4. | ||
3.4.4 Analysis of Abnormal Conditions | |||
- | (( | ||
3.4.5 Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated non-conservatively. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
3.4.5.1 Potential Nonconservatisms Correlations between gap locations in co-located panels ignored As was noted in Section 3.3.2.2, there appears to be some correlation in gap locations in poison panels around each cell. Axially co-located gaps around one storage cell would increase neutronic interaction and increase the keff of the system. Ideally, the Monte Carlo style sampling and modeling of Boraflex degradation should have included correlated sampling of gap locations. | |||
This effect is ameliorated by the artificial adjustment of the range of gap locations and the number of gaps simulated per panel. The gap location distribution limits gap to the central 6 feet of the fuel assembly. | OFFIGIAL USE ONLY PROPRIETARY INFORMATION | ||
The measured gap distributions show a significantly wider distribution in the gap locations. | - 20 | ||
This restriction increases the probability that some axially co-located gaps are modeled. | )) The NRC staff concludes that this TS change is acceptable. | ||
The distribution of the number of gaps is that all panels have at least one gap and 50 percent have one gap and the other 50 percent have two gaps. Measurements reported in Figure 2 of Attachment 1 of GNRO-201 0100073 (Reference | 3.4.5 Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated non-conservatively. | ||
3.4.5.1 Potential Nonconservatisms Correlations between gap locations in co-located panels ignored As was noted in Section 3.3.2.2, there appears to be some correlation in gap locations in poison panels around each cell. Axially co-located gaps around one storage cell would increase neutronic interaction and increase the keff of the system. Ideally, the Monte Carlo style sampling and modeling of Boraflex degradation should have included correlated sampling of gap locations. This effect is ameliorated by the artificial adjustment of the range of gap locations and the number of gaps simulated per panel. The gap location distribution limits gap to the central 6 feet of the fuel assembly. The measured gap distributions show a significantly wider distribution in the gap locations. This restriction increases the probability that some axially co-located gaps are modeled. The distribution of the number of gaps is that all panels have at least one gap and 50 percent have one gap and the other 50 percent have two gaps. | |||
Measurements reported in Figure 2 of Attachment 1 of GNRO-201 0100073 (Reference 10) indicate that there may be no gaps in 10 percent of the panels and more than two gaps in 20 percent of the panels. The artificially restricted distribution requires one or two gaps in every panel and ensures that the size of these one or two gaps will be large enough to conservatively bound the existence of more, but smaller, gaps. | |||
-21 | (( | ||
The uncredited nuclides could be worth up to about 1 to 2 percent Llk. Thus, these uncredited nuclides represent a significant unquantified conservatism. | )) As discussed in Section 3.3.1.1, the NRC staff noted other conservatisms that would bound this effect. | ||
Modeling of Boraflex degradation The analysis credits residual Boraflex with a 10B loading of 0.0133 g 10B/ | 3.4.5.2 Potential Analysis Conservatisms The analysis includes aspects that add margin to the analysis. | ||
The unquantified margin associated with this 20 percent 10B areal density reduction is significant conservatism. | (( | ||
3.4.5.3 Conclusion on Analysis of Margins The potential non-conservatisms associated with | OFFIGIAL USE ONLY* PROPRIETARY INFORMATION | ||
The !\IRC staff has concluded that not assuming that correlations between gap locations in co-located panels exist is ameliorated by the licensee's modeling artificiality there remains an uncertainty as to whether or not that artificiality is wholly sufficient. | |||
Therefore, the NRC staff has allocated the remaining analytical margin and unallocated conservatisms from the analysis to reach a reasonable assurance decision that the licensee's SFP TS will ensure compliance with 10 CFR 50.68. 3.5 The NFV NCSA NEDC-33621 P, Revision 0 (Reference 14), includes an analysis of the new fuel storage racks (NFSR). This analysis was needed to support inclusion of maximum nominal enrichment and maximum k-infinity limits in the revised TS. Specifically, the proposed TS limit the fuel assemblies stored to a maximum nominal uranium-235 enrichment of | OFFICIAL USE ONLY* PROPRIETARY INFORMATION | ||
This section documents the review of the NFV NCSA. 3.5.1 New Fuel Vault (NFV) NCSAs Method The analysis is to demonstrate compliance with the following requirements from 10 CFR 50.68: OFFICIAL USE ONLY* PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION | - 21 | ||
-Paragraph 50.68(b)(2) of 10 CFR requires, The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used. Paragraph 50.68(b )(3) of 10 CFR requires, If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used. The GGNS NFV NCSA demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95 percent probability/95 percent confidence level. The licensee's analYSis documented in NEDC-33621 P, Revision 0, demonstrates that the keff value, includil1g bias and uncertainties, for full density water conditions is more than | )) | ||
The new fuel storage rack kef! values are all based on unirradiated fuel bundles. | Uncredited actinides and fission products As was noted in Section 3.3.1, there are a large number of minor actinides and fission products that are not credited in this analysis. The uncredited nuclides could be worth up to about 1 to 2 percent Llk. Thus, these uncredited nuclides represent a significant unquantified conservatism. | ||
The standard cold core geometry maximum k-infinity values for each array were determined using TGBLA06A and its ENDF/B-V cross-section data. Consequently, the uncertainties include a contribution for the use of TGBLA06A to determine the in-core cold geometry k-infinity values. As noted earlier MCNP-05P is a GEH proprietary version of MCNP5, which has been validated for use at the GGNS and ENDF/B-VII.O nuclear data were used for the new fuel storage rack analysis. | Modeling of Boraflex degradation The analysis credits residual Boraflex with a 10B loading of 0.0133 g 10B/cm 2 | ||
The supporting validation is documented in Appendix A of NEDC-33621 P, Revision O. Section 2 of HI-2094416 states the bias and 95/95 uncertainty from the validation are | * The TSs require that cells with Boraflex determined to have a 1°B loading of less than 0.0165 g 10B/cm 2 be removed from service or moved to Region II, which does not credit any residual Boraflex. The unquantified margin associated with this 20 percent 10B areal density reduction is significant conservatism. | ||
OFFICIAL USE ONLY PROPRIETAR¥ 9FFICIAL USE ONLY PROPRIETARY | 3.4.5.3 Conclusion on Analysis of Margins The potential non-conservatisms associated with (( | ||
-The NRC staff reviewed the computational method and supporting validation and concludes it is acceptable. | )) can be quantified to some degree and have been offset as previously described. The !\IRC staff has concluded that not assuming that correlations between gap locations in co-located panels exist is ameliorated by the licensee's modeling artificiality there remains an uncertainty as to whether or not that artificiality is wholly sufficient. Therefore, the NRC staff has allocated the remaining analytical margin and unallocated conservatisms from the analysis to reach a reasonable assurance decision that the licensee's SFP TS will ensure compliance with 10 CFR 50.68. | ||
3.5.2 NFV Fuel Storage Racks The steel structures that comprise the NFV fuel storage racks were conservatively not modeled. | 3.5 The NFV NCSA NEDC-33621 P, Revision 0 (Reference 14), includes an analysis of the new fuel storage racks (NFSR). This analysis was needed to support inclusion of maximum nominal enrichment and maximum k-infinity limits in the revised TS. Specifically, the proposed TS limit the fuel assemblies stored to a maximum nominal uranium-235 enrichment of (( )) weight percent and limit the assembly k-infinity to no more than 1.26 in normal reactor core configuration and cold conditions. | ||
Without the steel, the rack model simplifies down to constraints on the spacing and location of the fuel assemblies. | This section documents the review of the NFV NCSA. | ||
All rack structures were modeled as full-density water. The analysis was performed using radially and axially infinite arrays of fuel assemblies. | 3.5.1 New Fuel Vault (NFV) NCSAs Method The analysis is to demonstrate compliance with the following requirements from 10 CFR 50.68: | ||
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- 22 Paragraph 50.68(b)(2) of 10 CFR requires, The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used. | |||
Paragraph 50.68(b )(3) of 10 CFR requires, If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used. | |||
The GGNS NFV NCSA demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95 percent probability/95 percent confidence level. The licensee's analYSis documented in NEDC-33621 P, Revision 0, demonstrates that the keff value, includil1g bias and uncertainties, for full density water conditions is more than (( )) .6.k below the applicable limit from 10 CFR 50.68. | |||
Typically, an optimum moderation study is also performed that identifies the maximum keff at reduced water densities for fuel stored in normally dry NFV storage racks. Section 5.5.1 of Reference 14 indicates the provision of 10 CFR 50.68(b)(3) that stipulates an optimum moderation analysis need not be performed is met. As the current GGNS TS do not have an optimum moderation keff limit, the NRC staff did not revisit that item in its review of this license amendment request. | |||
The new fuel storage rack kef! values are all based on unirradiated fuel bundles. The standard cold core geometry maximum k-infinity values for each array were determined using TGBLA06A and its ENDF/B-V cross-section data. Consequently, the uncertainties include a contribution for the use of TGBLA06A to determine the in-core cold geometry k-infinity values. | |||
As noted earlier MCNP-05P is a GEH proprietary version of MCNP5, which has been validated for use at the GGNS and ENDF/B-VII.O nuclear data were used for the new fuel storage rack analysis. The supporting validation is documented in Appendix A of NEDC-33621 P, Revision O. | |||
Section 2 of HI-2094416 states the bias and 95/95 uncertainty from the validation are (( | |||
)) This bias and bias uncertainty are similar to values reported for other analyses using ENDF/B-VII.O nuclear data. Considering that the values are consistent with similar analyses and that there is more than (( )) .6.k margin to the limits, further review of the validation was deemed unnecessary. | |||
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- 23 The NRC staff reviewed the computational method and supporting validation and concludes it is acceptable. | |||
3.5.2 NFV Fuel Storage Racks The steel structures that comprise the NFV fuel storage racks were conservatively not modeled. | |||
Without the steel, the rack model simplifies down to constraints on the spacing and location of the fuel assemblies. All rack structures were modeled as full-density water. | |||
The analysis was performed using radially and axially infinite arrays of fuel assemblies. | |||
Replacing new fuel storage vault walls, floor, and overhead detail with additional fuel is a conservative modeling approximation. | Replacing new fuel storage vault walls, floor, and overhead detail with additional fuel is a conservative modeling approximation. | ||
The licensee is introducing a standard cold core geometry k-infinity limit into the relevant TS. Consequently, the peak reactivity criticality analysis method used in the spent fuel storage racks is also applied to the new fuel storage racks. 3.5.3 Fuel Assemblies The new fuel storage rack analysis presented in Section 5.0 of NEDC-33621, Revision 0, evaluates only GE14 and GNF2 fuel assembly designs. | The licensee is introducing a standard cold core geometry k-infinity limit into the relevant TS. | ||
While the analysis looked at a very limited number of fuel bundles variations, this is acceptable because of the conservative modeling approach and large margin, | Consequently, the peak reactivity criticality analysis method used in the spent fuel storage racks is also applied to the new fuel storage racks. | ||
The fuel assembly uncertainties considered in the analysis did not include contributions related to | 3.5.3 Fuel Assemblies The new fuel storage rack analysis presented in Section 5.0 of NEDC-33621, Revision 0, evaluates only GE14 and GNF2 fuel assembly designs. While the analysis looked at a very limited number of fuel bundles variations, this is acceptable because of the conservative modeling approach and large margin, (( )) ilk to the 0.95 keff limit. | ||
3.5.4 MCNP Calculation Convergence Issue During the review, the NRC staff requested (see response to RAI-25 in Reference | The fuel assembly design is conservative in that the fuel is effectively modeled as infinitely long. | ||
Pin-by-pin enrichments, gadolinium fuel rods were explicitly modeled. | |||
The fuel assembly uncertainties considered in the analysis did not include contributions related to (( )) In response to RAI-16 documented in Attachment 1 to GNRO-2012/00120 (Reference 4), these uncertainties were estimated for the spentJuel storage racks. The combined uncertainty for these two parameters was (( )) for the spent fuel racks. It is expected that the contribution from these uncertainties would be similar for the new fuel storage racks. Combination of these uncertainties with the other analyzed uncertainties would raise keff by about (( )) | |||
- | Based on the NRC staff's review of the above, the fuel assembly model used in the NFV NCSA is adequately conservative. | ||
3.5.4 MCNP Calculation Convergence Issue During the review, the NRC staff requested (see response to RAI-25 in Reference 4) that fission source convergence be checked for all MCNP cases. According to the RAI response, (( | |||
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)) | |||
3.5.5 Analysis of Margins The licensee determined that the maximum kelf value, including biases and uncertainties, for the full density flooding case was (( )) The applicable limit is 0.95. The margin to this limit is (( )) The NRC staff concludes that this is acceptable. | |||
A few small potential nonconservatisms were identified in the review. The margins to the limits are large enough to cover potential nonconservatisms associated with deficiencies in the analysis of tolerances and failure to consider a broader range of GE14 and GNF2 fuel assembly variations. | A few small potential nonconservatisms were identified in the review. The margins to the limits are large enough to cover potential nonconservatisms associated with deficiencies in the analysis of tolerances and failure to consider a broader range of GE14 and GNF2 fuel assembly variations. | ||
3.4.6 Boraflex Monitoring Program By letter dated November 9, 2010 (Reference 15), the NRC staff issued an RAI. RAI-4 of that letter requested that the licensee provide more specific information on the nuclear criticality analysis that demonstrates compliance with GDC 62. In the licensee's response dated November 23,2010 (Reference 10), the licensee provided supplemental information on Boraflex performance and its associated monitoring program. | 3.4.6 Boraflex Monitoring Program By letter dated November 9, 2010 (Reference 15), the NRC staff issued an RAI. RAI-4 of that letter requested that the licensee provide more specific information on the nuclear criticality analysis that demonstrates compliance with GDC 62. In the licensee's response dated November 23,2010 (Reference 10), the licensee provided supplemental information on Boraflex performance and its associated monitoring program. | ||
The licensee utilizes Boraflex as the neutron absorber material to maintain the required subcriticality margin in the SFP storage racks. The Boraflex material is sandwiched between sheets of stainless steel with the edge strips welded in place to frame the Boraflex. | The licensee utilizes Boraflex as the neutron absorber material to maintain the required subcriticality margin in the SFP storage racks. The Boraflex material is sandwiched between sheets of stainless steel with the edge strips welded in place to frame the Boraflex. The licensee stated that a Boraflex monitoring program has been established at GGNS to monitor Boraflex performance. The Boraflex monitoring program includes gap measurements, and SFP silica evaluations based on the EPRI RACKLIFE system, which is a computer program used to assess in-service performance of Boraflex. In addition, the licensee indicated that Blackness testing (i.e., an in-situ measurement technique) was performed on Boraflex panels placed in high gamma dose locations to evaluate the size of panel gaps. These panels received dose from freshly discharged fuel following a refueling outage for approximately 1 year before Blackness testing commenced, constituting one campaign. This testing was performed after each refueling outage, totaling seven campaigns. | ||
The licensee stated that a Boraflex monitoring program has been established at GGNS to monitor Boraflex performance. | The licensee stated that the results of the Blackness testing indicated that the total panel gap, as a percent of the initial panel length versus dose, follows the EPRI Boraflex shrinkage model until the dose exceeds 2.3x1010 rads. Furthermore, the licensee stated that the loss accelerates as the panels approach 3.0x10 10 rads. | ||
The Boraflex monitoring program includes gap measurements, and SFP silica evaluations based on the EPRI RACKLIFE system, which is a computer program used to assess in-service performance of Boraflex. | The licensee reported that a Boron-10 (B-10) Areal Density Gauge for Evaluating Racks (BADGER) test campaign was conducted in 2007 to measure the B-10 areal density and panel loss from the gaps. BADGER testing is an in-situ technique used to measure B-10 areal density. The licensee stated that the BADGER test results confirmed that gap measurements for panels with doses below 2.3x1010 rads were consistent with the maximum shrinkage predicted by the EPRI Boraflex shrinkage model. The licensee also stated that the results of the BADGER test were consistent with the results of the previous seven Blackness tests. | ||
In addition, the licensee indicated that Blackness testing (i.e., an in-situ measurement technique) was performed on Boraflex panels placed in high gamma dose locations to evaluate the size of panel gaps. These panels received dose from freshly discharged fuel following a refueling outage for approximately 1 year before Blackness testing commenced, constituting one campaign. | OFFICIAL USE ONLY PROPRIEfARY INFORMAflON | ||
This testing was performed after each refueling outage, totaling seven campaigns. | |||
The licensee stated that the results of the Blackness testing indicated that the total panel gap, as a percent of the initial panel length versus dose, follows the EPRI Boraflex shrinkage model until the dose exceeds 2.3x1010 rads. Furthermore, the licensee stated that the loss accelerates as the panels approach 3. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
The licensee stated that the BADGER test results confirmed that gap measurements for panels with doses below 2.3x1010 rads were consistent with the maximum shrinkage predicted by the EPRI Boraflex shrinkage model. The licensee also stated that the results of the BADGER test were consistent with the results of the previous seven Blackness tests. OFFICIAL USE ONLY PROPRIEfARY INFORMAflON OFFICIAL USE ONLY PROPRIETARY INFORMATION | - 25 Furthermore, the licensee determined that the Boraflex material experienced additional degradation since the last blackness tests conducted in 1999. | ||
-Furthermore, the licensee determined that the Boraflex material experienced additional degradation since the last blackness tests conducted in 1999. The licensee stated that the 2007 BADGER test was conducted on 32 panels in Region I and Region II cell locations of the SFP. The licensee stated that the GGNS SFP is divided into cells that are labeled either Region lor Region" locations; the labels are used to identify the level of dose accumulated. | The licensee stated that the 2007 BADGER test was conducted on 32 panels in Region I and Region II cell locations of the SFP. The licensee stated that the GGNS SFP is divided into cells that are labeled either Region lor Region" locations; the labels are used to identify the level of 10 dose accumulated. The Region I panels had accumulated doses up to 1.77x1 0 rads; 10 whereas, the Region" panels had accumulated doses as high as 3.83x10 rads. A Region I and Region II cell was reported to have a minimum B-10 areal density of 0.0182 and 0.0166 gram/square centimeter (g/cm2), respectively. The licensee stated that the Region I minimum areal density is above the criticality safety analysis assumption of 0.0133 gm/cm 2 and the Region II analysis does not credit any Boraflex. The licensee reported that the difference between the Region I BADGER test results and RACKLIFE results are bounded by a 95/95 2 | ||
The Region I panels had accumulated doses up to 1.77x1 | uncertainty of 0.0022 g/cm . The licensee also stated that the 0.0022 g/cm 2 uncertainty is applied in the Borallex monitoring program to determine if a panel falls below the criticality analysis minimum areal density assumption of 0.0133 g/cm 2. Any cell that meets or exceeds the dose acceptance criterion of 2.3x1 010 rads or falls below the areal density acceptance criterion of 0.0182 g/cm 2 is configured as a Region II cell. | ||
respectively. | After reviewing the information the licensee provided regarding Boraflex performance at GGNS, the NRC staff determined that more information was needed to complete its review. The NRC staff issued RAls 2, 4, 5, 6, and 7 to the licensee on February 8, 2011 (Reference 25). The staff requested the licensee to discuss in detail the surveillance approach that will be used in the Boraflex monitoring program, specifically the methods of neutron attenuation testing (i.e., in-situ testing), frequency of inspection, sample size, data collection, and acceptance criteria. The staff requested the licensee to describe how the program's acceptance criteria account for potential degradation between surveillance periods. In addition, the staff requested the licensee to discuss the EPRI shrinkage model dose as it relates to the acceptance criteria analysis for continued operation. The staff also requested the licensee to discuss how the Region II locations are determined and added after each Blackness and BADGER test campaign; and the calibration technique and reference panel used for these types of tests. | ||
The licensee stated that the Region I minimum areal density is above the criticality safety analysis assumption of 0.0133 gm/ | In the licensee's response dated March 9, 2011 (Reference 11), the licensee stated that RACKLIFE will remain a significant component of the GGNS Boraflex monitoring program. The RACKLIFE calculations will continue to be performed each cycle and include projections of rack performance to the next RACKLIFE calculation. The licensee stated that the RACKLIFE analysis will compare the predicted silica to the plant measured silica. The RACKLIFE parameters will be adjusted as needed based on the comparison. The licensee stated that the analysis will include projections to the next planned RACKLIFE analysis date to ensure current Region I storage locations will not need to be reclassified as Region II storage locations. The licensee further stated that the EPRI shrinkage model has no direct impact on the criticality safety analysis; however, GGNS observed that a change in the Boraflex gap performance can occur at doses above that value. Boraflex panels which have received a gamma dose in excess of 2.3x10 10 rads or which have an areal density of less than 0.0165 g/cm2 are treated as Region II panels and are no longer credited in the criticality safety analysis. The licensee reported that the dose limit of 2.3x1 010 rads ensures the Boraflex gap configurations meet the NCSA assumptions. The licensee stated that the B-10 areal density acceptance criterion has been established from the summation of the NCSA assumed areal density, the | ||
The licensee reported that the difference between the Region I BADGER test results and RACKLIFE results are bounded by a 95/95 uncertainty of 0.0022 g/ | :OFFICIAL USE ONLY* PROPRIETARY INFORMATION | ||
frequency of inspection, sample size, data collection, and acceptance criteria. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
The staff requested the licensee to describe how the program's acceptance criteria account for potential degradation between surveillance periods. | - 26 BADGERlRACKLIFE uncertainty, and the design areal density tolerance (Le., 0.0133 + | ||
In addition, the staff requested the licensee to discuss the EPRI shrinkage model dose as it relates to the acceptance criteria analysis for continued operation. | 0.0022 + 0.001 = 0.0165 g/cm2). | ||
The staff also requested the licensee to discuss how the Region II locations are determined and added after each Blackness and BADGER test campaign; and the calibration technique and reference panel used for these types of tests. In the licensee's response dated March 9, 2011 (Reference 11), the licensee stated that RACKLIFE will remain a significant component of the GGNS Boraflex monitoring program. | The licensee stated that an additional BADGER measurement will be performed prior to the end of 2013 to ensure that the BADGERlRACKLIFE uncertainty remains valid. The 2013 BADGER test campaign results and rack performance will be evaluated to determine the need for additional tests. It was reported that the 2013 test campaign will consist of at least 30 Boraflex panels. The licensee indicated that the BADGER to RACKLIFE uncertainty will be developed based on the test results of the planned 2013 test. The licensee stated that if the uncertainty value is less than the existing BADGERlRACKLIFE uncertainty of 0.0022 g/cm 2, the more conservative value will be considered acceptable. In addition, the current uncertainty value of 0.0022 g/cm2 will be adjusted if evaluation of the test results indicates that the uncertainty is less than the existing value. The minimum areal density results will be confirmed to be greater than the NCSA assumption of 0.0133 g/cm 2. The licensee stated that the acceptability of the minimum areal density and uncertainty will be based on verifying that all the NCSA distributions bound the corresponding BADGER measured distributions. | ||
The RACKLIFE calculations will continue to be performed each cycle and include projections of rack performance to the next RACKLIFE calculation. | The NRC staff reviewed the information the licensee provided regarding Boraflex performance at GGNS and determined that more information was needed to complete its review. By electronic mail dated April 4, 2011 (Reference 26), the NRC staff issued additional RAls. | ||
The licensee stated that the RACKLIFE analysis will compare the predicted silica to the plant measured silica. The RACKLIFE parameters will be adjusted as needed based on the comparison. | The NRC staff requested the licensee to discuss in detail the future (i.e., after 2012) surveillance approach and BADGER testing for the Boraflex material. | ||
The licensee stated that the analysis will include projections to the next planned RACKLIFE analysis date to ensure current Region I storage locations will not need to be reclassified as Region II storage locations. | In the licensee's response dated May 3,2011 (Reference 13), the licensee stated GGNS will perform periodic testing of the Boraflex neutron absorbing material on a frequency of 5 years using BADGER testing. During the period between BADGER testing, the licensee intends to perform analyses to confirm subcriticality is maintained. The NRC staff notes that the licensee committed to performing BADGER and RACKLI FE testing at the intervals mentioned above (i.e., 5 years and each cycle, respectively). In a letter dated July 2, 2013 (Reference 6), the licensee further stated that the options for long-term strategy for maintaining subcriticality in its SFP per its licensing basis and the regulations include the following: 1) making use of additional dry cask storage to store spent fuel, 2) perform rack replacement, and/or 3) install neutron absorber inserts. | ||
The licensee further stated that the EPRI shrinkage model has no direct impact on the criticality safety analysis; | After reviewing the information the licensee provided on the Boraflex neutron absorber material and the Boraflex monitoring program, the NRC staff concludes that the program provides reasonable assurance that it will be able to detect, monitor, and mitigate Boraflex degradation. | ||
This is accomplished by detection and monitoring of degradation in the use of the RACKLIFE computer code and BADGER testing, and reclassifying panels that do not meet the acceptance criteria from Region I to Region II locations. The NRC staff concludes that comparing predicted data from RACKLIFE to plant SFP measured silica each cycle will also provide information on the accuracy of the model to actual conditions. The NRC staff concludes that the 5-year frequency for BADGER testing and the conducting of further evaluation and analysis of material degradation between testing provides reasonable assurance that in between surveillances the Boraflex performance will be monitored and mitigated. In addition, performing in-situ BADGER testing will provide information on the amount of B-10 areal density present and material OFFICIAL USE ONLY PROPRIETARY INFORMATION | |||
The licensee reported that the dose limit of 2.3x1 010 rads ensures the Boraflex gap configurations meet the NCSA assumptions. | |||
The licensee stated that the B-10 areal density acceptance criterion has been established from the summation of the NCSA assumed areal density, the :OFFICIAL USE ONLY* PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION | OFFICIAL' USE ONLY PROPRIETARY INFORMATION | ||
-BADGERlRACKLIFE uncertainty, and the design areal density tolerance (Le., 0.0133 + 0.0022 + 0.001 = 0.0165 g/cm2). The licensee stated that an additional BADGER measurement will be performed prior to the end of 2013 to ensure that the BADGERlRACKLIFE uncertainty remains valid. The 2013 BADGER test campaign results and rack performance will be evaluated to determine the need for additional tests. It was reported that the 2013 test campaign will consist of at least 30 Boraflex panels. The licensee indicated that the BADGER to RACKLIFE uncertainty will be developed based on the test results of the planned 2013 test. The licensee stated that if the uncertainty value is less than the existing BADGERlRACKLIFE uncertainty of 0.0022 g/ | - 27 degradation (i.e., reduction of neutron-absorbing capacity and loss of material) of the Boraflex SFP racks. | ||
In addition, the current uncertainty value of 0.0022 g/cm2 will be adjusted if evaluation of the test results indicates that the uncertainty is less than the existing value. The minimum areal density results will be confirmed to be greater than the NCSA assumption of 0.0133 g/ | Based on a review of the dose limit and areal density acceptance criteria, the NRC staff has concluded that they provide reasonable assurance that the program will be able to mitigate degradation of Boraflex before it will challenge the criticality analysis as evaluated above. | ||
The NRC staff reviewed the information the licensee provided regarding Boraflex performance at GGNS and determined that more information was needed to complete its review. By electronic mail dated April 4, 2011 (Reference 26), the NRC staff issued additional RAls. The NRC staff requested the licensee to discuss in detail the future (i.e., after 2012) surveillance approach and BADGER testing for the Boraflex material. | 3.6 Summary The NRC staff review of the GGNS new and spent fuel storage racks NCSA, documented in NEDC-33621 P, Revision 0, identified some non-conservative items. These items were evaluated against the margin to the regulatory limit and what the NRC considers an appropriate amount of margin attributable to conservatisms documented in the analyses. | ||
In the licensee's response dated May 3,2011 (Reference 13), the licensee stated GGNS will perform periodic testing of the Boraflex neutron absorbing material on a frequency of 5 years using BADGER testing. | The spent fuel storage analysis for Region I included a novel Monte Carlo sampling technique for crediting residual Boraflex. The NRC staff identified two issues with the method. The first issue is that statistical analysis of measured data was used to conclude that the distributions used to characterize the residual Boraflex were not correlated. From a physical perspective this does not seem realistic. All four panels around anyone storage location receive a major part of their radiation dose from the fuel assemblies stored in that cell. Clearly, the source of the damage is correlated. The NRC staff concluded this is acceptable as the licensee has offset this potential non-conservatism as discussed in Section 3.4.2.2, "Correlations between distributions." Secondly, the spent fuel manufacturing process very likely introduced correlated Boraflex panel restriction points and water flow paths. This would occur due to how the components were assembled and welded together. From the review of a sample of measured data for one randomly selected storage cell indicates that there is some correlation. The axial co-location of gaps around a storage cell would increase keff. This effect is offset by artificial adjustments made to the distributions used to simulate the Boraflex damage. In particular, the simulated damage is restricted to the central 6 feet of the panel and all panels are modeled as having either one or two Boraflex gaps. In reality, the gaps are spread over a much larger range, 10 percent of the examined panels had no gaps and 20 percent of the panels had more than two gaps. More small gaps should yield a lower keff compared to the same total loss modeled in one or two gaps. The NRC staff concluded this is acceptable as additional margins are available to offset this potential non-conservatism discussed in Section 3.4.5, "Margin Analysis and Comparison with Remaining Uncertainties." | ||
During the period between BADGER testing, the licensee intends to perform analyses to confirm subcriticality is maintained. | The distributions used to simulate the Boraflex degradation can be compared with new data as data become available. This will ensure that the characteristics of the Boraflex degradation have not changed enough to invalidate the distributions used in criticality analysis. Further, following some initial radiation dose, panel dissolution continues even in the absence of significant continuing dose. As new data becomes available regarding the continuing dissolution and evolution of Boraflex gap formation. the bases for the GGNS SFP criticality analysis can be confirmed. | ||
The NRC staff notes that the licensee committed to performing BADGER and RACKLI FE testing at the intervals mentioned above (i.e., 5 years and each cycle, respectively). | The NRC staff has also reviewed the licensee's evaluation of the effects of the proposed changes on the Boraflex neutron absorber material used at GGNS and concludes that the licensee has addressed the impact of the changes on the Boraflex SFP racks. Further, the licensee has provided the NRC staff with reasonable assurance that the Boraflex monitoring OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
In a letter dated July 2, 2013 (Reference 6), the licensee further stated that the options for long-term strategy for maintaining subcriticality in its SFP per its licensing basis and the regulations include the following | |||
: 1) making use of additional dry cask storage to store spent fuel, 2) perform rack replacement, and/or 3) install neutron absorber inserts. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
After reviewing the information the licensee provided on the Boraflex neutron absorber material and the Boraflex monitoring | - 28 program will continue to detect degradation of Boraflex material and meet the requirements of 10 CFR 50.68(b)(4), SRP Section 9.1.2, and GDC 62. Therefore, the NRC staff concludes the proposed changes are acceptable with respect to the Boraflex material and monitoring program. | ||
Based on the an acceptable Boraflex monitoring program, the review of the supporting analysis reports and based on the margins to regulatory limits, crediting some analysis conservatism, and including consideration of the identified potential non-conservatisms, the NRC staff concludes that there is a reasonable assurance that the GGNS SFP and NFV fuel storage racks meet the applicable NCSA regulatory requirements. Therefore, the NRC staff concludes that the TS changes are acceptable. | |||
This is accomplished by detection and monitoring of degradation in the use of the RACKLIFE computer code and BADGER testing, and reclassifying panels that do not meet the acceptance criteria from Region I to Region II locations. | In addition as stated in Section 3.2, the licensee has proposed to remove License Condition 45. | ||
The NRC staff concludes that comparing predicted data from RACKLIFE to plant SFP measured silica each cycle will also provide information on the accuracy of the model to actual conditions. | The license condition was approved for one cycle. Therefore, the NRC staff concludes that the deletion of this license condition is acceptable with the approval of this amendment. | ||
The NRC staff concludes that the 5-year frequency for BADGER testing and the conducting of further evaluation and analysis of material degradation between testing provides reasonable assurance that in between surveillances the Boraflex performance will be monitored and mitigated. | 4.0 REGULATORY COMMITMENTS By letters dated October 1, 2012, and July 2 and September 5, 2013, Entergy made the following regulatory commitments: | ||
In addition, performing in-situ BADGER testing will provide information on the amount of B-10 areal density present and material OFFICIAL USE ONLY PROPRIETARY OFFICIAL' USE ONLY PROPRIETARY INFORMATION | : 1) The [NCSA] LAR submitted via Entergy letter GNRO-2011/00076 (ADAMS Accession #ML1125321287) proposed a change to TS 4.3.1 that specified six of 16 cells in a Region 114x4 cell array will be blocked. | ||
-degradation (i.e., reduction of neutron-absorbing capacity and loss of material) of the Boraflex SFP racks. Based on a review of the dose limit and areal density acceptance | In order to reflect that eight rather than six cells will be blocked as determined in the criticality safety analysis, Entergy is processing a change to proposed TS 4.3.1. This revision will be submitted under separate letter on or before October 22, 2012. | ||
: 2) While the contribution to the calculated panel dose from diagonally adjacent fuel assemblies is relatively small, an adjustment to the dose calculations will be augmented to include this effect for empty cells. | |||
The spent fuel storage analysis for Region I included a novel Monte Carlo sampling technique for crediting residual Boraflex. | : 3) The zero-dose panels in cells Z014, Z016, and Zr15 will be included in the upcoming BADGER test campaign. | ||
The NRC staff identified two issues with the method. The first issue is that statistical analysis of measured data was used to conclude that the distributions used to characterize the residual Boraflex were not correlated. | : 4) In order to ensure the criticality analysis remains applicable, the Boraflex monitoring program will be modified to incorporate updating the gap growth due to dissolution following each BADGER campaign. The updated dissolution rate will be applied to the most recent BADGER results through the end of the next BADGER test interval plus one year to confirm the continued applicability of the criticality analysis. | ||
From a physical perspective this does not seem realistic. | : 5) The Boraflex monitoring program will be modified to confirm that the analyzed dissolution features remain bounding. | ||
All four panels around anyone storage location receive a major part of their radiation dose from the fuel assemblies stored in that cell. Clearly, the source of the damage is correlated. | : 6) Entergy will maintain a minimum distance of 12 inches between any fuel stored in the Control Blade/Defective Fuel Storage Rack (Module H1) and in the surrounding high density spent fuel pool storage racks. | ||
The NRC staff concluded this is acceptable as the licensee has offset this potential non-conservatism as discussed in Section 3.4.2.2, "Correlations between distributions." | OFFICIAL USE ONLY - PROPRIETARY INFORMATION | ||
From the review of a sample of measured data for one randomly selected storage cell indicates that there is some correlation. | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
The axial co-location of gaps around a storage cell would increase keff. This effect is offset by artificial adjustments made to the distributions used to simulate the Boraflex damage. In particular, the simulated damage is restricted to the central 6 feet of the panel and all panels are modeled as having either one or two Boraflex gaps. In reality, the gaps are spread over a much larger range, 10 percent of the examined panels had no gaps and 20 percent of the panels had more than two gaps. More small gaps should yield a lower keff compared to the same total loss modeled in one or two gaps. The NRC staff concluded this is acceptable as additional margins are available to offset this potential non-conservatism discussed in Section 3.4.5, "Margin Analysis and Comparison with Remaining Uncertainties." | - 29 The licensee stated that Regulatory Commitment 1 was closed in the letter dated October 22, 2012. The NRC staff concludes that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment are best provided by the licensee's administrative processes, including its commitment management program. | ||
The distributions used to simulate the Boraflex degradation can be compared with new data as data become available. | |||
This will ensure that the characteristics of the Boraflex degradation have not changed enough to invalidate the distributions used in criticality analysis. | ==5.0 CONCLUSION== | ||
the bases for the GGNS SFP criticality analysis can be confirmed. | 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. | ||
The NRC staff has also reviewed the licensee's evaluation of the effects of the proposed changes on the Boraflex neutron absorber material used at GGNS and concludes that the licensee has addressed the impact of the changes on the Boraflex SFP racks. Further, the licensee has provided the NRC staff with reasonable assurance that the Boraflex monitoring OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION | |||
-program will continue to detect degradation of Boraflex material and meet the requirements of 10 CFR 50.68(b)(4), | ==6.0 REFERENCES== | ||
SRP Section 9.1.2, and GDC 62. Therefore, the NRC staff concludes the proposed changes are acceptable with respect to the Boraflex material and monitoring program. | : 1. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request, Criticality Safety Analysis and Technical Specification 4.3.1, Criticality, Grand Gulf Nuclear Station, Unit 1," dated September 9, 2011 (GNRO-2011/00076) (ADAMS Accession No. ML112521287). | ||
Based on the an acceptable Boraflex monitoring | : 2. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Criticality Safety Analysis, Grand Gulf Nuclear Station, Unit 1," dated November 21,2011 (GNRO-2011/00104) (ADAMS Accession No. ML113320260). | ||
: 3. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to NRC Requests for Additional Information Pertaining to License Amendment Request for Criticality Safety Analysis, Grand Gulf Nuclear Station, Unit 1," | |||
Therefore, the NRC staff concludes that the TS changes are acceptable. | dated April 18, 2012 (GNRO-2012/00027) (ADAMS Accession No. ML12109A281). | ||
In addition as stated in Section 3.2, the licensee has proposed to remove License Condition | : 4. Ford, B.S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to NRC Requests for Additional Information - GGNS Criticality, Safety Analysis License Amendment Request, Grand Gulf Nuclear Station, Unit 1," dated October 1, 2012 (GNRO-2012/00120) (ADAMS Accession No. ML12276A152). | ||
: 5. Richey, M. L., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to NRC Request for Additional Information - GGNS Criticality Safety Analysis License Amendment Request, Grand Gulf Nuclear Station, Unit 1," dated October 22,2012 (GNRO-2012/00124) (ADAMS Accession No. ML12296A417). | |||
4.0 REGULATORY COMMITMENTS By letters dated October 1, 2012, and July 2 and September 5, 2013, Entergy made the following regulatory commitments: | : 6. Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request - Responses to NRC Requests for Additional Information, Grand Gulf Nuclear Station, Unit 1," dated July 2, 2013 (GNRO-2013/00050)(ADAMS Accession No. ML13190A043). | ||
: 1) The [NCSA] LAR submitted via Entergy letter GNRO-2011/00076 (ADAMS Accession | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
# | |||
In order to reflect that eight rather than six cells will be blocked as determined in the criticality safety analysis, Entergy is processing a change to proposed TS 4.3.1. This revision will be submitted under separate letter on or before October 22, 2012. 2) While the contribution to the calculated panel dose from adjacent fuel assemblies is relatively small, an adjustment to the dose calculations will be augmented to include this effect for empty cells. 3) The zero-dose panels in cells Z014, Z016, and Zr15 will be included in the upcoming BADGER test campaign. | OFFIGIAL USE ONLY PROPRIETARY INFORMATION | ||
: 4) In order to ensure the criticality analysis remains applicable, the Boraflex monitoring program will be modified to incorporate updating the gap growth due to dissolution following each BADGER campaign. | - 30 | ||
The updated dissolution rate will be applied to the most recent BADGER results through the end of the next BADGER test interval plus one year to confirm the continued applicability of the criticality analysis. | : 7. Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request - Supplemental Information, Grand Gulf Nuclear Station, Unit 1," dated September 5, 2013 (GNRO-2013/00066) | ||
: 5) The Boraflex monitoring program will be modified to confirm that the analyzed dissolution features remain bounding. | (ADAMS Accession No. ML13249A235). | ||
: 6) Entergy will maintain a minimum distance of 12 inches between any fuel stored in the Control Blade/Defective Fuel Storage Rack (Module H1) and in the surrounding high density spent fuel pool storage racks. OFFICIAL USE ONLY -PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION | : 8. Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request - Supplemental Information, Grand Gulf Nuclear Station, Unit 1," dated September 16, 2013 (GNRO-2013/00073) | ||
-The licensee stated that Regulatory Commitment 1 was closed in the letter dated October 22, 2012. The NRC staff concludes that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment are best provided by the licensee's administrative processes, including its commitment management program. 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. | (ADAMS Accession No. ML13260A084). | ||
: 9. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request, Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated September 8, 2010 (GNRO-201 0/00056) (ADAMS Accession No. ML102660409). | |||
(ADAMS Accession No. | : 10. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Supplemental License Amendment Request, Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated November 23, 2010 (GNRO-2010100073) (ADAMS Accession No. ML103330093). | ||
(ADAMS Accession No. | : 11. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated March 9,2011 (GNRO-2011/00017) (ADAMS Accession No. ML110680507). | ||
-GGNS Criticality, Safety Analysis License Amendment | : 12. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated April 21, 2011 (GNRO-2011/00025) (ADAMS Accession No. ML11112A098). | ||
: 13. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1,I! dated May 3, 2011 (GNRO-2011/00034) (ADAMS Accession No. ML111240288). | |||
(ADAMS Accession No. | : 14. GE-Hitachi Nuclear Energy Americas, LLC, "Grand Gulf Nuclear Station, Fuel Storage Criticality Safety Analysis of Spent and New Fuel Storage Racks," NEDC-33621 P, Revision 0, November 2010 (not publicly available - proprietary); public version designated as NEDO-33621, Revision 0, November 2010 (ADAMS Accession No. ML103330092). | ||
-GGNS Criticality Safety Analysis License Amendment | : 15. Wang, A S., U.S. Nuclear Regulatory Commission, letter to Entergy Operations, Inc., | ||
"Grand Gulf Nuclear Station. Unit 1 - Supplemental Information Needed for Acceptance of License Amendment Request for an Extended Power Uprate (TAC No. ME4679)," | |||
(ADAMS Accession No. | |||
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFIGIAL USE ONLY PROPRIETARY | |||
-Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request -Supplemental Information, Grand Gulf Nuclear Station, Unit 1," dated September 5, 2013 (GNRO-2013/00066) | |||
(ADAMS Accession No. | |||
(ADAMS Accession No. | |||
(ADAMS Accession No. | |||
(ADAMS Accession No. | |||
(ADAMS Accession No. | |||
(ADAMS Accession No. | |||
(ADAMS Accession No. ML111240288). GE-Hitachi Nuclear Energy Americas, LLC, "Grand Gulf Nuclear Station, Fuel Storage Criticality Safety Analysis of Spent and New Fuel Storage Racks," NEDC-33621 P, Revision 0, November 2010 (not publicly available | |||
-proprietary); | |||
public version designated as NEDO-33621, Revision 0, November 2010 (ADAMS Accession No. | |||
Unit 1 -Supplemental Information Needed for Acceptance of License Amendment Request for an Extended Power Uprate (TAC No. ME4679)," | |||
dated November 9,2010 (ADAMS Accession No. ML103010200). | dated November 9,2010 (ADAMS Accession No. ML103010200). | ||
OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY* PROPRIETARY INFORMATION | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
-31 Electric Power Research Institute Report NP-6159, "An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks," prepared by Northeast Technology Corp., December 1988 (ADAMS Accession No. ML003736666). U.S. Nuclear Regulatory Commission, Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks," dated June 26, 1996 (ADAMS Accession No. ML031110008). | |||
prepared for Entergy Operations, Inc., July 1, 1999. Excerpts provided as enclosure 2 of Entergy Operations Inc. letter dated April 21, 2011 (ADAMS Accession No. | OFFICIAL USE ONLY* PROPRIETARY INFORMATION | ||
prepared for Entergy Nuclear Operations, Inc., October 1, 2010. Provided as enclosure 1 of Entergy Operations Inc. letter dated April 21, 2011 (ADAMS Accession No. | - 31 | ||
dated August 19, 1998 (ADAMS Accession No. ML003728001). U.S. Nuclear Regulatory Commission, "Final Division of Safety Systems Interim Staff Guidance DSS-ISG-2010-01, Revision 0, 'Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools,'" | : 16. Electric Power Research Institute Report NP-6159, "An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks," prepared by Northeast Technology Corp., December 1988 (ADAMS Accession No. ML003736666). | ||
dated October 13, 2011 (ADAMS Accession No. ML110620086). J.M. Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses Criticality (keff) Predictions," | : 17. U.S. Nuclear Regulatory Commission, Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks," dated June 26, 1996 (ADAMS Accession No. ML031110008). | ||
NUREG/CR-7109 (ORNLITM-2011/514), | : 18. HOLTEC Report HI-992255, "Blackness Testing of Boraflex in Selected Spent Fuel Storage Rack Cells of the Grand Gulf Nuclear Station," prepared for Entergy Operations, Inc., July 1, 1999. Excerpts provided as enclosure 2 of Entergy Operations Inc. letter dated April 21, 2011 (ADAMS Accession No. ML111120329). | ||
U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012 (ADAMS Accession No. ML12116A128). J.C. Wagner and C.v. Parks, "Recommendations on the Credit for Cooling Time in PWR Burnup Credit Analyses," | : 19. Northeast Technology Corp. Report NET-287-01, Revision 1, "BADGER Test Campaign at Grand Gulf Nuclear Station," prepared for Entergy Nuclear Operations, Inc., | ||
NUREG/CR-6781 (ORNLlTM-2001/272), | October 1, 2010. Provided as enclosure 1 of Entergy Operations Inc. letter dated April 21, 2011 (ADAMS Accession No. ML11112A099). | ||
U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2003 (ADAMS Accession No. ML030290585). D.E. Mueller, K.R. Elam, and P.B. Fox, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," NUREG/CR-6979, ORNLlTM-20071083, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2008. Wang, A. B., U.S. Nuclear Regulatory Commission, e-mail to Entergy Operations, Inc., "Grand Gulf, Unit 1, Request for Additional Information, NRRlDCIICSGB Review, Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt (TAC No. ME4679)," | : 20. Kopp, Sr., L., U.S. Nuclear Regulatory Commission, memorandum to T. Collins, U.S. Nuclear Regulatory Commission, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," dated August 19, 1998 (ADAMS Accession No. ML003728001). | ||
dated February 8, 2011 (ADAMS Accession No. | : 21. U.S. Nuclear Regulatory Commission, "Final Division of Safety Systems Interim Staff Guidance DSS-ISG-2010-01, Revision 0, 'Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools,'" dated October 13, 2011 (ADAMS Accession No. ML110620086). | ||
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY* PROPRIETARY INFORMATION | : 22. J.M. Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses Criticality (keff) Predictions," NUREG/CR-7109 (ORNLITM-2011/514), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012 (ADAMS Accession No. ML12116A128). | ||
-Wang, A. B., U.S. Nuclear Regulatory Commission, e-mail to Entergy Operations, Inc., "Grand Gulf, Unit 1, Request for Additional Information, Round 2, NRRlDCI/CSGB Review, Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt (TAC No. ME4679)," | : 23. J.C. Wagner and C.v. Parks, "Recommendations on the Credit for Cooling Time in PWR Burnup Credit Analyses," NUREG/CR-6781 (ORNLlTM-2001/272), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2003 (ADAMS Accession No. ML030290585). | ||
dated April 4, 2011 (ADAMS Accession No. ML110940136). | : 24. D.E. Mueller, K.R. Elam, and P.B. Fox, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," NUREG/CR-6979, ORNLlTM-20071083, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2008. | ||
Principal Contributors: | : 25. Wang, A. B., U.S. Nuclear Regulatory Commission, e-mail to Entergy Operations, Inc., | ||
A. Obodoako and K. Wood Date: September 25, 2013 OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY -PROPRIETARY INFORMATION | "Grand Gulf, Unit 1, Request for Additional Information, NRRlDCIICSGB Review, Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt (TAC No. ME4679)," dated February 8, 2011 (ADAMS Accession No. ML110390173). | ||
-safety evaluation, which is provided in Enclosure | OFFICIAL USE ONLY PROPRIETARY INFORMATION | ||
OFFICIAL USE ONLY* PROPRIETARY INFORMATION | |||
- 32 | |||
: 26. Wang, A. B., U.S. Nuclear Regulatory Commission, e-mail to Entergy Operations, Inc., | |||
"Grand Gulf, Unit 1, Request for Additional Information, Round 2, NRRlDCI/CSGB Review, Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt (TAC No. ME4679)," dated April 4, 2011 (ADAMS Accession No. ML110940136). | |||
Principal Contributors: A. Obodoako and K. Wood Date: September 25, 2013 OFFICIAL USE ONLY PROPRIETARY INFORMATION | |||
OFFICIAL USE ONLY - PROPRIETARY INFORMATION | |||
- 2 safety evaluation, which is provided in Enclosure 3. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. | |||
Sincerely, IRA! | |||
Alan Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416 | |||
==Enclosures:== | ==Enclosures:== | ||
: 1. Amendment No. 195 to NPF-29 2. Proprietary Safety Evaluation | : 1. Amendment No. 195 to NPF-29 | ||
: 2. Proprietary Safety Evaluation | |||
: 3. Non-proprietary Safety Evaluation cc w/Enclosures 1 and 3: Distribution via Listserv DISTRIBUTION: | : 3. Non-proprietary Safety Evaluation cc w/Enclosures 1 and 3: Distribution via Listserv DISTRIBUTION: | ||
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* OFFICE NRR/DORLlLPL4/PM NRRlDORLlLPL4/LA NRRlDSS/STSB/BC NRRlDE/ESGB/BC'" | |||
*SE in ut via email *OFFICE NRR/DORLlLPL4/PM NRRlDORLlLPL4/LA NRRlDSS/STSB/BC NRRlDE/ESGB/BC'" | ----~--------------r-------------~. | ||
JBurkhardt GKulesa 9/23/13 9/12/13 DSS/SRXB/BC* | JBurkhardt RElliott GKulesa | ||
OGC -DORLlLPL4/BC NRRlDORLlLPL4/P MMarkley (CFLyon for) AWang 9/25/13 *9/25/13 OFFICIAL USE ONLY PROPRIETARY INFORMATION}} | --------~---------- | ||
9/18/13 9/23/13 9/12/13 DSS/SRXB/BC* OGC - NLO DORLlLPL4/BC NRRlDORLlLPL4/P | |||
-----------+--------------~----------- | |||
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-=-- | |||
9/25/13 9/25/13 *9/25/13 | |||
~--~-======-~--~~O~F~F~IC~IA~L~R~E7C~O~RDCOPY OFFICIAL USE ONLY PROPRIETARY INFORMATION}} | |||
Latest revision as of 03:39, 20 March 2020
| ML13261A264 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/25/2013 |
| From: | Wang A Plant Licensing Branch IV |
| To: | Entergy Operations |
| Wang A | |
| References | |
| TAC ME7111 | |
| Download: ML13261A264 (46) | |
Text
OFFICIAL USE ONLY PROPRI&TARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 25,2013 Vice President, Operations Entergy Operations, Inc.
Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1-ISSUANCE OF AMENDMENT RE: CHANGES TO THE NUCLEAR CRITICALITY SAFETY ANALYSIS (TAC NO. ME7111)
Dear Sir or Madam:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 195 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS).
This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated September 9, 2011, as supplemented by letters dated September 8 and November 23,2010; March 9, April 21 , May 3, and November 21,2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16, 2013. The letters dated September 8 and November 23, 2010, and March 9 and May 3, 2011, are incorporated by reference in the September 9, 2011, license amendment request (LAR) as allowed by Section 50.32, "Elimination of replication," of Title 10 of the Code of Federal Regulations (10 CFR).
The amendment approves: 1) additional requirements for the spent fuel and new fuel storage racks in TS 4.3.1, "Criticality," 2) a revision to the current Nuclear Criticality Safety Analysis, which is described in GGNS Updated Final Safety Analysis Report Sections 9.1.1, "New Fuel Storage," and 9.1.2, "Spent Fuel Storage," to reflect changes resulting from the extended power uprate, and 3) deletion of the spent fuel pool loading criteria Operating License condition.
The NRC has determined that the related safety evaluation, provided in Enclosure 2, contains proprietary information pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, the NRC staff has also prepared a non-proprietary version of the NOTICE: Enclosure 2 to this letter contains Proprietary Information. Upon separation from
- Enclosure 2, this letter is DECONTROLLED.
OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL Y81i ONbV PROPRIIiTARY INFO~MATlON
-2 safety evaluation, which is provided in Enclosure 3. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, Alana::pr~:t.r Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosures:
- 1. Amendment No. 195 to NPF-29
- 2. Proprietary Safety Evaluation
- 3. Non-proprietary Safety Evaluation cc w/Enclosures 1 and 3: Distribution via Listserv OFFICIAL Y8E ONLY PROPRIETARY INFORMATION
ENCLOSURE 1 AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC" ET AL.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 ENTERGY OPERATIONS, INC.
SYSTEM ENERGY RESOURCES, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPI, INC.
DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 195 License No. NPF-29
- 1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:
A. The application for amendment by Entergy Operations, Inc. (the licensee), dated September 9, 2011, as supplemented by letters dated September 8 and November 23, 2010; March 9, April 21, May 3, and November 21, 2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and 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.
-2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-29 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. 195 are hereby incorporated in the license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
In addition, Paragraph 2.C.(45) of Facility Operating License No. NPF-29 is hereby amended to read as follows:
(45) Deleted.
- 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 will maintain a minimum distance of 12 inches between any fuel stored in the Control Blade/Defective Fuel Storage Rack (Module H1) and in the surrounding high-density spent fuel pool storage racks as described in the licensee's letter dated September 5, 2013, and the NRC staff's safety evaluation for this amendment. In addition, the licensee shall include the revised information in the Grand Gulf Nuclear Station Updated Final Safety Analysis Report in the next periodic update in accordance with 10 CFR 50.71(e), as described in the licensee's application dated September 9, 2011, as supplemented by letters dated September 8 and November 23,2010; March 9, April 21, May 3, and November 21, 2011; April 18, October 1, and October 22, 2012; and July 2, September 5, and September 16, 2013, and the NRC staff's safety evaluation for this amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License No. NPF-29 and the Technical Specifications Date of Issuance: September 25, 2013
ATTACHMENT TO LICENSE AMENDMENT NO, 195 FACILITY OPERATING LICENSE NO, NPF-29 DOCKET NO, 50-416 Replace the following pages of the Facility Operating License No. NPF-29 and the Appendix A, Technical Specifications 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 Remove 4 4 16a 16a 16b 16b Technical Specifications Remove 4.0-1 4.0-1 4.0-2 4,0-2 4.0-2a
(b) SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54.
In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility.
C. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10CFR 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 Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 4408 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2) Technical ~pecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 195 are hereby incorporated into this license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
During Cycle 19, GGNS will conduct monitoring of the Oscillation Power Range Monitor (OPRM). During this time, the OPRM Upscale function (Function 2.f of Technical Specification Table 3.3.1.1 1) will be disabled and operated in an "indicate only" mode and technical specification requirements will not apply to this function. During such time, Backup Stability Protection measures will be implemented via GGNS procedures to provide an alternate method to detect and suppress reactor core thermal hydraulic instability oscillations. Once monitoring has been successfully completed, the OPRM Upscale function will be enabled and technical specification requirements will be applied to the function; no further operating with this function in an "indicate only" mode will be conducted.
4 Amendment No. 195
(b) The first performance of the periodic assessment of eRE habitability, Specification 5.5.13.c. (ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from March 2005, the date of the most recent successful tracer gas test, as stated in the June 30, 2005 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.
(c) The first performance of the periodic assessment of the CRE boundary, Specification 5.5.13.d, shall be within the next 18 months, plus the 136 days allowed by SR 3.0.2, as measured from the date of issuance of this amendment.
(44) Leak rate tests associated with Surveillance Requirements (SR) 3.6.1.1.1, 3.6.1.3.5, and 3.6.1.3.9, as required by TS 5.5.12 and in accordance with 10 CFR 50, Appendix J, Option B, and SRs 3.6.5.1.1 and 3.6.5.1.2 are not required to be performed until their next scheduled performance dates. The tests will be performed at the EPU calculated peak containment pressure or within EPU drywell bypass leakage limits, as appropriate.
(45) Deleted.
16a Amendment No. ~, ~, +/-9+/-, 195
(46) This license condition provides for monitoring, evaluating, and taking prompt action in response to potential adverse flow effects as a result of power uprate operation on plant structures, systems, and components (including verifying the continued structural integrity of the steam dryer) for power ascension from the CLTP (3898 MWt) to the EPU level of 4408 MWt (or 113 percent of CLTP or 115 percent of OLTP).
(a) The following requirements are placed on operation of the facility before and during the power ascension to 3898 MWt:
- 1. GGNS shall provide a Power Ascension Test (PAT) Plan for the Steam Dryer testing. This plan shall include:
- Criteria for comparison and evaluation of projected strain and acceleration with on-dryer instrument data.
- Acceptance limits developed for each on-dryer strain gauge and accelerometer.
- Tables of predicted dryer stresses at CLTP, strain amplitudes and PSDs at strain gauge locations, acceleration amplitudes and PSDs at accelerometer locations, and maximum stresses and locations.
The PAT plan shall provide correlations between measured accelerations and strains and the corresponding maximum stresses. The PAT plan shall be submitted to the NRC Project Manager no later than 10 days before start-up.
- 2. GGNS shall monitor the main steam line (MSL) strain gages and on-dryer instrumentation at a minimum of three power levels up to 3898 MWt. Based on a comparison of projected and measured strains and accelerations, GGNS will assess whether the dryer acoustic and structural models have adequately captured the response significant to peak stress projections.
16b Amendment No. ~, 195
Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for Grand Gulf Nuclear Station is located in Claiborne County, Mississippi on the east bank of the Mississippi River, approximately 25 miles south of Vicksburg and 37 miles north-northeast of Natchez. The exclusion area boundary shall have a radius of 696 meters from the centerline of the reactor.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 800 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 2 ) as fuel material, and water rods. Limited sUbstitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 193 cruciform shaped control rod assemblies. The control material shall be boron carbide or hafnium metal, or both.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. keff S 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the UFSAR;
- b. A nominal fuel assembly center to center storage spacing of 6.26 inches in the storage racks.
- c. Fuel assemblies having a maximum K-infinity of 1.26 in the normal reactor core configuration at cold conditions; (continued)
GRAND GULF Amendment No. ~,+/-+4,195
Design Features 4.0 4.0 DESIGN FEATURES 4.3.1.1 (continued)
- d. Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 percent;
- e. Region II racks are controlled as follows:
- 1. cells with any Boraflex which has received a gamma dose in excess of 2.3E10 rads or which has a Boron-l0 areal less than 0.0165, which are the Spent Fuel Pool Rack Boraflex are treated as II
- 2. Storage cells face- acent either restricted from fuel the isolated cells or are conf as a minimum (i.e., additional cells may the Region II fuel storage configuration requirements in Figure 4.3 1.
- 3. When a 4x4 array of cells is classified as Region II and face-adjacent to another II 4x4 storage array, the new II 4x4 array is to be blocked in the same 8-of-16 and at the same orientation as the acent Region II 4x4 storage Location Blocked to Prevent (continued)
GRAND GULF 4.0 2 Amendment No. ~, 195
Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
- a. keff S 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.1 of the UFSARj
- b. A nominal fuel assembly center to center storage spacing of 6.535 inches within rows and 11.875 inches between rows in the new fuel storage racks.
- c. Fuel assemblies having a maximum k-infinity of 1.26 in the normal reactor core configuration at cold conditions;
- d. Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent.
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 202 ft 5.25 inches.
4.3.3 Capacity 4.3.3.1 The spent fuel storage pool shall be maintained with a storage capacity limited to no more than 4348 fuel assemblies.
4.3.3.2 No more than 800 fuel assemblies may be stored in the upper containment pool.
GRAND GULF 4.0 2a Amendment No. 195
ENCLOSURE 3 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 (NON-PROPRI ETARY)
OfflGIAl us. ONlY PROPRI.TARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 195 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
By letter dated September 9, 2011 (Reference 1), as supplemented by letters dated November 21,2011 (Reference 2), April 18, 2012 (Reference 3), October 1,2012 (Reference 4), October 22,2012 (Reference 5), July 2,2013 (Reference 6), September 5,2013 (Reference 7), and September 16, 2013 (Reference 8), Entergy Operations. Inc. (Entergy, the licensee), submitted a request to amend Facility Operating License No. NPF-29 and revise the Grand Gulf Nuclear Station, Unit 1 (GGNS) Technical Specifications (TS). The letters dated September 8,2010 (Reference 9), and November 23,2010 (Reference 10); and March 9,2011 (Reference 11), April 21, 2011 (Reference 12), and May 3, 2011 (Reference 13), were incorporated by reference by the licensee in the September 9, 2011, license amendment request (LAR) as allowed by Section 50.32, "Elimination of replication," of Title 10 of the Code of Federal Regulations (10 CFR). Specifically, the proposed amendment requested to:
- 1) Revise the current Nuclear Criticality Safety Analysis (NCSA), which is described in GGNS Updated Final Safety Analysis Report (UFSAR) Sections 9.1.1, "New Fuel Storage," and 9.1.2, "Spent Fuel Storage," to reflect changes resulting from the extended power uprate (EPU),
- 2) Revise TS 4.3.1.1, "Criticality," to add requirements for two design parameters to spent and new fuel storage racks resulting from the EPU and add requirements to specify a spent fuel storage configuration for Region II cells (Region I and Region /I cells are described in Section 3.2 of this safety evaluation) to account for potential degradation of Boraflex, and
- 3) Delete the spent fuel pool (SFP) loading criteria Operating License condition, which was approved for Cycle 19 or until a new NCSA is approved.
Portions of the letters dated September 8 and November 23, 2010, April 21, 2011, October 1, 2012, and July 2,2013, contain proprietary information and, accordingly, have been withheld from public disclosure. Accordingly, the supplemental letters dated September 8 and November 23,2010; March 9, April 21, May 3, and November 21,2011; April 18, October 1, and October 22,2012; and July 2, September 5, and September 16. 2013, provided additional OFFICIAL USE ONLY - PROPRIETARY INFORMATION
OFFICIAL USE ONLY PROPRIETARY INFORMATION
- 2 information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 24, 2012 (77 FR 3511).
The following is the NRC staff's evaluation of the NCSA and associated TS changes.
2.0 REGULATORY EVALUATION
The following regulatory requirements and guidance documents were applicable to the NRC staff's review of the licensee's amendment request:
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 62 states:
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
Paragraph 50.68(b)(1) of 10 CFR requires:
Plant procedures shall prohibit the handling and storage at anyone time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.
Paragraph 50.68(b)(2) of 10 CFR requires:
The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.
Paragraph 50.68(b)(3) of 10 CFR requires:
If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with lOW-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98,at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.
Paragraph 50.68(b)(4) of 10 CFR requires:
If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with 6FFICIAL USE ONLY* PROPRIETARY INFORMATION
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- 3 unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.
Neutron absorber materials utilized in spent fuel racks exposed to treated water or treated borated water may be susceptible to reduction of neutron-absorbing capacity, changes in dimension that increase Keff and loss of material. The NRC staff reviewed the effects of the proposed changes on the Boraflex neutron absorber material utilized at GGNS and the adequacy of the licensee's Boraflex monitoring program to assure that degradation of SFP neutron absorbing material that could affect the criticality analysis will be detected, monitored and mitigated. The NRC's acceptance criteria for the Boraflex monitoring program are based on the 10 CFR Part 50.68(b){4) reqUirements as stated above. Specific review criteria are contained in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (SRP) Section 9.1.2, "New and Spent Fuel Storage," Revision 4, and 10 CFR Part 50, Appendix A, GDC 62, "Prevention of criticality in fuel storage and handling."
Paragraph 50.36{c){4) of 10 CFR states that the TSs will include items in the category of:
Design features. Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.
The GGNS SFP and containment pool (CP) NCSA does not take credit for soluble boron for normal operating conditions. Therefore, the regulatory requirement is for the GGNS keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.
The GGNS New Fuel Vault (NFV) NCSA demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95 percent probability/95 percent confidence level. Typically, an optimum moderation study is performed that identifies the maximum keff at reduced water densities for fuel stored in normally dry NFV storage racks. Section 5.5.1 of General Electric Hitachi Nuclear Energy Americas LLC (GEH) report NEDC-33621 P, Revision 0, "Grand Gulf Nuclear Station, Fuel Storage Criticality Safety Analysis of Spent and New Fuel Storage Racks," November 2010 (Reference 14), indicates the provision of 10 CFR 50.68(b)(3) that states an optimum moderation analysis need not be performed is met. As the current GGNS TS do not have an optimum moderation kett limit, the NRC staff did not review this item in its review of this LAR.
3.0 TECHNICAL EVALUATION
3.1 Background By letter dated September 8,2010 (Reference 9), and November 23,2010 (Reference 10); and March 9, 2011 (Reference 11). April 11, 2011 (Reference 12). and May 3, 2011 (Reference 13),
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-4 Entergy requested to amend Facility Operating License No. NPF-29 and to revise the GGNS TSs to support an EPU at a licensed core thermal power level increased from 3898 megawatts thermal (MWt) to 4408 MWt including revising the NCSA. The NCSA needed to be revised to reflect changes to include consideration of reactor operating condition changes related to the EPU and to include consideration of Boraflex degradation in the spent fuel storage racks. By letter dated November 9, 2010 (Reference 15), the NRC staff requested, among other things, that the licensee provide an NCSA that demonstrates compliance with General Design Criterion (GDC) 62, "Prevention of criticality in fuel storage and handling," of Appendix A to 10 CFR Part 50. A nuclear criticality safety analysis was subsequently documented in GEH report NEDC-33621P, Revision 0 (Reference 14), which was transmitted to the NRC as Attachment 2 to a letter from Entergy dated November 23,2010 (Reference 10).
The GEH report NEDC-33621 P presents the NCSA for the GGNS new and spent fuel storage racks. The report describes the methodology and analytical models used in the NCSA to show that the new and spent fuel storage racks maximum k-effective (keff) will be no greater than 0.95 when flooded with unborated water. An optimum moderation analysis of fresh fuel storage racks, consistent with 10 CFR 50.68(b)(3) requirements was not performed for the GGNS. The GEH report NEDC-33621 P provided a basis for omitting the optimum moderation analysis for the new fuel storage racks. GGNS has two pools for spent fuel storage: the SFP and the containment pool (CP) as described below:
The GGNS SFP contains 16 rack modules of varying size. Fifteen of the SFP rack modules are high-density storage modules in which fuel is stored on a 6.26-inch center-to-center spacing.
Originally, these 15 rack modules utilized Boraflex as a neutron absorber. Boraflex degradation, described generically in the Electric Power Research Institute (EPRI) Report NP-6159, "An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks," prepared by Northeast Technology Corp., December 1988 (Reference 16), and NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks," dated June 26, 1996 (Reference 17),
has been observed (See HOLTEC Report HI-992255 in Reference 18 and Northeast Technology Corp. Report NET-287-01, Revision 1, in Reference 19) in the GGNS high-density storage rack modules.
Eight high-density storage modules are also installed in the GGNS CPo These rack modules are the same design as the high-density storage modules installed in the SFP and use Boraflex neutron absorber panels. The total CP fuel assembly storage capacity is limited to 800 fuel assemblies. The GEH criticality analysis report (Reference 14) also covers fuel storage in the CP fuel assembly storage racks.
As a result of the Boraflex degradation, the new criticality analysis includes provisions for crediting residual Boraflex in Region I storage locations and elimination of credit for Boraflex in Region II storage locations. Entergy has proposed a two-region fuel storage approach in the spent fuel storage racks. Region I credits residual Boraflex and Entergy has requested that all Region I storage locations may be used. Region" storage cells are defined in TS 4.3.1.1.e as all storage locations face adjacent to a Boraflex panel that has either received a gamma dose in excess of 2.3x1010 rads or has been determined to have a boron-10 areal density less than 0.0165 grams boron-10 per square centimeter. Fuel stored in Region II must be stored in an 8-out-of-16 checkerboard arrangement in minimum increments of 4x4 arrays as smaller patterns were not analyzed. In addition, to address misloading events, Entergy has physically blocked OFFICIAL USE ONLY PROPRIETARY INFORMATION
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- 5 certain locations in Region II to prevent storage of spent fuel in those locations as described in TS 4.3.1.1.e. It should be noted that Region I and Region II can exist in both the SFP and the CP. The 16th rack is a legacy low-density storage rack module designated as module H 1.
No criticality analysis was provided for the single low-density storage rack module (H 1), which was designed to store fuel with 12-inch center-to-center spacing. The licensee has proposed a limitation to maintain a 12 inch buffer between any spent fuel stored in H1 and the spent fuel racks. The total SFP fuel assembly storage capacity is limited to 4348 fuel assemblies.
New fuel assemblies may be stored in what are normally dry conditions in the GGNS new fuel racks in the New Fuel Vault (NFV). The NFV NCSA is updated to support modification of the NFV technical specifications to include a maximum nominal enrichment limit of 4.9 weight percent uranium-235 (new TS 4.3.1.2.d) and a maximum normal reactor core configuration at cold conditions k-infinity limit of 1.26 (new TS 4.3.1.2.c). No credit is taken in fresh fuel storage rack analysis for neutron absorption in storage rack structural materials. The NFV storage capacity is limited to 300 fuel assemblies.
3.2 Proposed Changes to Facility Operating License and TSs Current paragraph 2.C.(45) of Facility Operating License No. NPF-29 states:
(45) Through Cycle 19 or until the revised criticality safety analysis has been approved, whichever comes first, the storage cells in the GGNS SFP racks shall be categorized as either Unrestricted or Restricted.
(a) Unrestricted cells (Region I) are cells with a minimum panel B10 areal density greater than 0.0179 gm/cm 2 and that have received an exposure less than 2.3E10 rads. Unrestricted cells may contain fuel assemblies up to the maximum k-infinity of 1.26 (cold core configuration).
(b) Restricted cells (Region II) are cells with either a minimum panel B10 areal density less than 0.0179 gm/cm 2 or that have received an exposure in excess of 2.3E1 0 rads. Storage in Restricted cells shall not credit any Boraflex. Storage shall be controlled in a 10-of-16 configuration (see below). In addition, only fuel assemblies with a k-infinity of less than 1.21 (cold core configuration) may be stored in a Region II cell.
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-6 Region II 4X4 Storage Configuration B B B
B B B
D Fuel Assembly Storage Location
[!] Location Physically Blocked to Prevent Storage Revised Paragraph 2.C.(45) of Facility Operating License No. NPF-29 would state:
(45) Deleted Current TS 4.3.1.1 has been revised to add the following new requirements, which would state:
- c. Fuel assemblies having a maximum k-infinity of 1.26 in the normal reactor core configuration at cold conditions;
- d. Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent;
- e. Region II racks are controlled as follows:
- 1. Storage cells with any Boraflex panel which has received a gamma dose in excess of 2.3E10 rads or which has a Boron-10 areal density less than 0.0165, which are designated within the Spent Fuel Pool Rack Boraflex Monitoring Program, are treated as Region II panels.
- 2. Storage cells face-adjacent to Region II panels are either restricted from fuel storage by physically blocking the isolated cells or are configured to meet, as a minimum (i.e., additional cells may be blocked), the Region II fuel storage configuration requirements in Figure 4.3-1 .
- 3. When a 4x4 array of cells is classified as Region II and face adjacent to another Region II 4x4 storage array, the new Region II 4x4 array is required to be blocked in the same 8-of-16 pattern and at the same orientation as the adjacent Region II 4x4 storage configuration.
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-7 Figure 4.3-1 Region II 4X4 Storage Configuration D Fuel Assembly Storage Location D Location Physically Blocked to Prevent Storage Current TS 4.3 .1.2 has been revised to add the following new requirements, which would state:
- c. Fuel assemblies having a maximum k-infinity of 1.26 in the normal reactor core configuration at cold conditions;
- d. Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent.
These changes are related to include consideration of reactor operating condition changes related to the EPU and to include consideration of Boraflex degradation in the spent fuel storage racks .
The reactor operating condition changes related to the EPU affect fuel depletion, which is considered in the new and spent fuel NCSA. EPU-related changes impacting NCSA may include higher power density, and fuel and moderator temperature changes. Fuel depletion EPU-related changes affect the NFV analysis because the TS are being revised to include a limit on new fuel storage based on the maximum k-infinity in normal reactor core configuration at cold conditions. The k-infinity limit includes consideration of reactivity peaking as the gadolinia within the fuel is depleted with fuel burnup.
By letter dated September 5, 2013 (Reference 7), the licensee proposed a limitation requiring that no fuel in low-density storage rack module H 1 will be stored within 12 inches of the surrounding high-density SFP storage racks. This limitation is necessary since no analysis was presented to address the effect on kelf of the interface between any fuel that is stored in the low density storage rack module H1 and the surrounding high-density SFP storage racks.
Otherwise the storage of fuel in low-density storage rack module H 1 is governed by the legacy analysis for that module.
In Reference 7, the licensee has proposed to remove License Condition 45. The license condition was approved for one cycle . Therefore, the NRC staff concludes that the deletion of this license condition is acceptable with the approval of this amendment.
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- 8 3.3 Method of Review This safety evaluation involves a review of the NEDC-33621 P (Reference 14) which provides the NCSA for the new and spent fuel storage racks at GGNS provided as Attachment 2 to the licensee's letter dated April 21, 2011 (Reference 12). The NCSA includes the effects of the changes in reactor operating conditions associated with the EPU and supports use of spent fuel storage rack requirements that include use of two storage regions with different storage configurations. The review was performed consistent with SRP Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling."
On August 19, 1998, the NRC staff issued guidance for performing the review of SFP/CP NCSA. This memorandum is known colloquially as the 'Kopp Memo' (Reference 20), after the author. While the Kopp memorandum does not specify a methodology, it does provide some guidance on the more salient aspects of an NCSA, including computer code validation. The guidance is germane to boiling-water reactors (BWRs) and pressurized-water reactors (PWRs),
borated and unborated. The Kopp memorandum has been used for virtually every light-water reactor SFP NCSA since, including this GGNS analysis.
This safety evaluation also considered the guidance provided in NRC Interim Staff Guidance DSS-ISG-2010-01, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," dated October 13, 2011 (Reference 21).
3.4 SFP/CP NCSA Review 3.4.1 SFP/CP NCSA Method There is no generic or standard methodology for performing NCSA for fuel storage and handling. The methods used for the NCSA for fuel in the GGNS SFP and CP are described in NEDC-33621P, Revision 0 (Reference 14). Additional information describing the methods used is provided in the RAI responses attached to References 5, 10, 11, and 12). Some SFP/CP analysis deficiencies were identified during the review, but as will be discussed below, sufficient margin is built into the analysis methodology to offset the deficiencies. The methodology is specific to this analysis and is not appropriate for other applications.
3.4.1.1 Computational Methods The criticality analysis considers the increase in fuel reactivity seen in BWR fuel as the gadolinia (Gd 20 3 ) fabricated into some of the fuel rods is depleted during reactor operations. This method is frequently used in BWR criticality analyses and is sometimes referred to as the standard cold core geometry (SCCG) peak k-infinity method or, more simply, the in-core kM method. In this method, the lattices used in the fuel assemblies are characterized by their SCCG peak k-infinity.
The analyst demonstrates that lattices with an SCCG peak k-infinity meeting the TS k-infinity limit will have an in-rack keff value no greater than 0.95 with a 95 percent probability and 95 percent confidence. The SCCG peak k-infinity values and associated burned fuel compositions were calculated using the TGBLA06 computer code with its ENDF/B-V nuclear data.
In Reference 14, the licensee stated that the fuel storage criticality (keff) value calculations for the SFP and CP were performed using the GEH/GNF proprietary version of the Monte Carlo OFFICIAL USE ONLY PROPRIETARY INFORMATION
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- 9 N-Particle (MCNP)-05P, used with ENDF/B-VII.O continuous energy cross-section data and the burned fuel compositions calculated by TGBLA06. MCNP-05P is a Monte Carlo program for solving the linear neutron transport equation for a 'fixed source or an eigenvalue problem. The code implements the Monte Carlo process for neutron, photon, or electron transport or coupled transport involving all these particles, and can compute the eigenvalue for neutron-multiplying systems. For the present application, only neutron transport was considered. Reference 14 provides a plant specific validation for the use of MCNP-05P at the GGNS. The NRC staff has concluded that with this validation the in-core k-infinity method and these computer codes and their associated nuclear data sets are acceptable for use at GGNS. In addition, the in-core k-infinity method and these computer codes and their associated nuclear data sets have been used in many NCSAs, and are industry standards The spent fuel analysis (( ))
(Reference 12) to cover lack of validation of spent fuel compositions, including fission products, and (( )) to cover the lack of validation for keft calculations of burned fuel systems containing minor actinides and fission products. These methods of determining the uncertainties are consistent with guidance in References 20 and 21. Recent work published in NUREG/CR-71 09, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Criticality (keff) Predictions," April 2012 (Reference 22), indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with calculating keff for systems with minor actinides and fission products. From the response provided for RAI-5.b in Reference 6, the burned fuel compositions included some additional actinides and fission products beyond those considered in NUREG/CR-71 09. Some of the additional nuclides are radioactive and have half-lives that are too short to be credited. ((
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- 10
))
During the review, NRC staff requested (see response to RAI-25 in Reference 4) that fission source convergence be checked for all MCNP-05P cases. According to the RAI response, all SFP/CP MCNP-05P cases were verified to be converged.
3.4.1.2 Computer Code Validation Since the NCSA credits fuel burnup, it is necessary to consider validation of the computer codes and data used to calculate burned fuel compositions and the computer code and data that utilize the burned fuel compositions to calculate keff for systems with burned fuel.
((
)) This uncertainty was calculated by the licensee and applied correctly.
The study used to support validation of keff calculations using MCNp*5 was documented in Appendix A to GEH report NEDC-33621 P. The validation study was revised as discussed in RAI responses provided in References 4 and 6. The validation set included ((
)) During the review, it was noted that no analysis of trends in the calculated bias and bias uncertainty was performed. In response to various RAls, the licensee revised the validation (( )) and to perform trending analysis. The revised bias uncertainty was used in the calculation of the maximum keff.
Appropriate critical experiment data were not available to validate keff calculations crediting minor actinides or fission products. To address this validation deficiency, ((
))
Recent work published in NUREG/CR-7109 indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with minor actinides and fission products. ((
))
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- 11 3.4.2 SFP and Fuel Storage Racks 3.4.2.1 SFP Water Temperature NRC guidance provided in the Kopp memorandum states the NCSA should be done at the temperature corresponding to the highest reactivity. Analysis was performed with water temperatures of (( )) The water densities used reflect the water density at each water temperature. The temperatures examined reflect ((
)) The (( )) cross-section data set was used for all MCNP calculations.
The NRC staff identified two issues with the SFP water temperature bias determination. The first is that the analysis did not examine enough water temperatures between ((
)) to ensure that kelf did not peak at some value between the evaluated water temperatures. The response to RAI-4 presented in Reference 6 includes results from additional calculations that demonstrate that the Region I and II kelf values do not peak between the already evaluated temperatures.
The second issue is that (( )) cross-sections were used for all temperature/density variations evaluated. The critical experiments used in the validation study have temperatures around (( )) There are limited data available to validate the impact on kelf of the variation of cross-sections with temperature over the 300 to 400 oK range. As can be seen from Figure 4-1 and Figure 4-2 provided in the response to RAI-4 in GNRO-2013/00050 (Reference 6), Region I is limiting at low temperature and Region II is limiting at high temperature. The Region II kmax(95195), which includes a large adder for the misplaced assembly accident, is only (( )) (see response to RAI-12 in Reference 4).
Consequently, there is adequate margin to cover poor validation of the variation of kelf with temperature at (( )) The Region I kmax(95195) is limiting at low temperature and the impact of poor validation of the variation of kelf with temperature at (( )) as compared to the validation study temperature of (( )) is expected to be negligible.
3.4.2.2 SFP/CP Storage Rack Models The GGNS SFP contains 16 rack modules of varying size for a total capacity of 4393 storage cells. Fifteen of the SFP rack modules are high-density storage modules in which fuel is stored on a 6.26-inch center-to-center spacing. Originally, these 15 rack modules utilized Boraflex as a neutron absorber. Boraflex degradation, described generically in References 16 and 17) has been observed (References 18 and 19) in the GGNS high-density storage rack modules. As a result of the Boraflex degradation, the new criticality analysis includes provisions for crediting residual Boraflex in Region I storage locations and elimination of credit for Boraflex in Region II storage locations.
No criticality analysis was provided for the single unpoisoned low-density storage rack module (H1), which was designed to store fuel with 12-inch center-to-center spacing. Consequently, the OFFICIAL USE ONLY* PROPRIETARY INFORMATION
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- 12 total SFP fuel assembly storage capacity is limited to 4348 fuel assemblies. Since the current operating license includes a 4348 fuel assembly SFP fuel assembly storage capacity, prohibiting storage in low-density storage rack module H-1 is consistent with the current operating license.
The high-density spent fuel (HDSF) racks used in the SFP and CP rack modules are constructed from stainless steel cruciform, 'T" and "L" shaped components welded together with a 6.26-inch center-to-center spacing. Each wing of each of these components is constructed of a Boraflex sheet that originally had a nominal thickness of 0.070 inches and a 95/95 minimum lOB loading of about 0.0019 g 1°B/cm2 between two 0.063-inch thick stainless steel sheets.
Blackness testing reported in Reference 18 and Boron-10 (B-10) Areal Density Gauge for Evaluating Racks (BADGER) testing reported in Reference 19 has revealed significant Boraflex degradation. The nature of the degradation is described in these references and includes shrinkage, general dissolution, and gap and scallop formation. While the degradation was initially caused by radiation damage to the polymers in the Boraflex, dissolution of the damaged Boraflex material continues even where the radiation field is reduced. Consequently, the licensee needs to include consideration of both the current condition of the Boraflex and its continuing degradation.
No credit was taken for residual Boraflex in highly damaged panels. The TS proposed in Reference 5 require that any cells face-adjacent to a panel that has received a gamma dose greater than 2.3 x 1010 rads or which has a lOB areal density less than 0.0165 g 10B/cm 2 be either blocked off or configured to be in Region II storage. Fuel will be stored In Region II in a 2-out-of-4 checkerboard arrangement with a minimum size of 4x4 storage locations. No credit is taken for Boraflex in Region II.
Many of the RAls associated with this LAR review are related to the methods used to credit residual Boraflex in the Region I storage locations. The panel width is reduced ((
)) to account for radiation induced shrinkage and the lOB areal loading is reduced
(( )) to account for dissolution. Due to the random-looking damage patterns measured in some Boraflex panels, a Monte Carlo style approach has been taken to quantify the impact of gaps in the Boraflex panels. In this approach, probability distributions were determined based on evaluation of measured data. Distributions were developed for the number of gaps per panel, total panel loss per panel, gap size per panel, and gap axial location per panel. These distributions were then sampled to create (( )) storage cell models that (( )) The resulting keft values were then used to determine an ((
))
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- 13 NRC staff had several questions and concerns about this technique. The following issues are the most significant items and are described and discussed below:
- Number of calculations
- Derivation of gap distributions
- Continuing time-dependent evolution of distributions
- Correlations between distributions
- Artificial adjustments to distributions
(( ))
Number of calculations The original analysis included only (( )) This did not appear to be enough samples to fully populate the high-keff tail on the results distribution. The response to RAI3.c.2) provided in Reference 6, notes that the analysis was expanded to include (( ))
configurations. This expansion of the analysis was appropriate and is sufficient.
Derivation of gap distributions The NRC staff requested additional information concerning the development of the Boraflex gap distributions in the Monte Carlo sampling method to define Region II degraded Boraflex models.
Additional information was provided in References 4, 6, 10, and 11. The additional information provided an improved understanding concerning the development of the probability distributions used in the development of the degraded Boraflex models.
Continuing time-dependent evolution of distributions After a panel receives some minimal radiation dose, the Boraflex material starts to break down and dissolve even if no additional radiation dose is received. An issue raised by the NRC staff is that from discussion in various documents, it was not clear that the continuing dissolution of Boraflex was adequately incorporated into the criticality analysis. Discussion was provided in the response to RAI-3 in Reference 6. The information provided a reasonable assurance that consideration of the continuing dissolution is included in the analysis.
Correlations between distributions The degraded Boraflex is characterized by distributions describing (1) total panel loss due to gaps, (2) number of gaps, (3) size of gaps, and (4) locations of gaps. The NRC staff had questions concerning the potential for correlations between these distributions in anyone panel and between gap characterization in the four panels around each cell.
The following may result in correlations:
- Storage rack design - each rack module is built from cruciform, 'T', and "L" shaped components. It is likely that water flow paths into and through each shape are different and that gap formation and dissolution patterns and rates for each shape may be correlated with the component shape.
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- 14
- Fabrication - Boraflex panel gap formation is affected by how the steel, holding the Boraflex in position, is welded together in each component and how the components are welded together. Assembly and welding processes common to all panels may increase the likelihood of damage occurring at correlated locations. For example, spot welds may be at similar axial locations for all components.
- Most of the radiation dose received by anyone panel is from irradiated fuel stored in the two storage cell locations face adjacent to the panel. All four panels around anyone cell receive about half of their dose from the same spent fuel assemblies. This common radiation source term could result in related/similar damage patterns in the Boraflex panels surrounding each cell.
- Within a panel, the number of gaps is likely anti-correlated with gap sizes.
Multiple axially co-located gaps around an assembly would yield a higher keff than randomly separated gaps. Due to the potential for correlations between distributions and between panels around each storage cell, NRC staff requested additional information concerning the potential impact of correlations on the Monte Carlo sampling method used to generate the degraded Boraflex models for Region I. In its RAI responses provided for RAI-4 in Reference 4 and for RAI-3 in Reference 6, the licensee maintained that there are no statistically significant correlations observed in measured data. It appears that this conclusion was based on a "null hypothesis" that no correlations existed unless proven otherwise. The NRC staff' did not reach the same conclusion from its interpretation of the measured data. Since the maximum keff will be determined by the minimum critical volume and not the entire SFP volume the measured data must be interpreted with that in mind. The NRC staff's observations are that when the measured data of multiple panels surrounding the same storage cell are considered, a distinct correlation is readily apparent. Therefore, the NRC staff did not conclude that there is no correlation. The NRC staff believes it would have been more appropriate to assume there are correlations unless proven otherwise, thus overestimating the impact of Boraflex gaps on SFP/CP rack keff.
However, even with the licensee's conclusion concerning the handling of correlations in the data, the NRC staff has concluded that the licensee's practice of artificially limiting the distributions for gap axial location to the central 6 feet of the fuel assembly and for restricting the number of gaps to one or two gaps is acceptable as it ameliorates the lack of simulating correlations between distributions and in panels around each storage cell because it forces the licensee's degradation modeling to be more densely packed than a purely random distribution would achieve. As more data becomes available from future surveillance activities at GGNS, this issue could be looked at again. As the Boraflex continues to degrade, additional cells might need to be removed from service or the criticality analyses might need to be updated.
Artificial adjustments to distributions During its review of the probability distribution functions used to simulate Boraflex degradation, questions were raised by NRC staff that some of the distributions did not compare well with the measured data and were in some cases artificially adjusted. Examples include limitation of gap OFFICIAL USE ONLY PROPRIETARY INFORMATION
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- 15 locations to the central 6 feet of the fuel assembly, differences between the measured and assumed gap sizes, renormalization of individual gap sizes to match total panel loss, and limitation of the number of gaps per panel to one or two.
Additional information and discussion was provided in the response provided for RAI-4 in Reference 4 and in the RAI responses provided in Reference 6. Additional calculations were performed by the licensee and documented in Reference 6 to demonstrate that renormalization of the gap sizes to be consistent with the sampled total gap size had no impact on the analysis.
(( ))
The Monte Carlo Boraflex degradation modeling technique was used to simulate (( ))
degraded Boraflex configurations. The licensee used the (( )) keff results to ((
))
The total storage capacity is 4,348 assemblies in the SFP and 800 assemblies in the CPo ((
))
In discussions with the Entergy, it stated that it had revised the analysis to use the ((
)) but had neglected to inform the NRC staff in its RAt response.
This was subsequently documented in a letter to the NRC dated September 16, 2013 (Reference 8).
3.4.2.3 SFP/CP Storage Rack Models Manufacturing Tolerances and Uncertainties Region I and II racks are the same design, with no credit taken for Boraflex in the Region II racks. Nominal values were used for rack wall thicknesses and center-to-center spacing.
Sensitivity studies were performed to quantify the impact of manufacturing tolerances and uncertainties on the SFP keff.
((
)) The NRC staff has concluded this method adequately quantifies the uncertainty in keff due to variations in Boraflex degradation.
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- 16 3.4.3 Fuel Assembly 3.4.3.1 Bounding Fuel Assembly Design The fuel assemblies currently used at GGNS are GE14 and GNF2 fuel bundles. These BWR assemblies are 1Ox1 0 lattices (( ))
Assemblies may include fuel rods, containing gadolinia (Gd 20 3 ), in selected locations in the lattice. Assemblies may contain ((
)) GGNS has also used several other fuel assembly designs during its operating history. In addition to GE 14 and GNF2 bundles, previously used assembly designs include GE 6/7, ANF 8x8, ANF 9x9, ATRIUM 10, and GE11.
In the peak reactivity method, the bounding fuel assembly design is determined by depleting various lattices at a variety of reactor conditions and using the resulting burned fuel compositions to calculate the in-rack keff value for each reactor depletion/lattice variation combination. The identified bounding fuel bundle, which is conservatively above the SCCG k-infinity limit value of 1.26, has an in-rack keff less than 0.95 that bounds 95 percent of the resulting in-rack keff values for all lattices with a 95 percent confidence, and, therefore, the NRC staff concludes the identified bounding fuel bundle to be acceptable.
Three concerns were identified by NRC staff during review of the criticality analysis provided in support of the LAR:
- 1. It appears that the analysis did not include consideration of all fuel assembly designs that have been used at GGNS,
- 2. The determination of bounding lattice may have been based on examination of too few lattices, and
- 3. Since the bounding lattice analysis was based on a relatively small number of lattices, the most reactive or 95/95 bounding lattice may not have been identified.
These issues were resolved satisfactorily, as described below.
Legacy Fuel The criticality analysis documented in NEDC-33621 P, Revision 0 (Reference 14), used only the GE14 and GNF2 fuel designs to identify the bounding fuel assembly design. Additional information concerning other "legacy fuel" was provided in Attachment 1 to Reference 6.
Justification was provided for all legacy fuel designs except the GE 6/7 assemblies. The response to RAI-2 provided in Attachment 1 to GNRO-2013/0050 (Reference 6) addressed the GE 6/7 assemblies. As such the licensee has demonstrated that other legacy fuels used at GGNS are bounded by the GE 14 and GNF2 fuel designs and therefore, the NCSA is acceptable for legacy fuel designs.
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- 17 Number of Lattices Considered Table 17 of NEDC-3362'I P, Revision 0, identified only (( )) lattices that were examined to determine the bounding lattice. Each of the lattices was depleted at varying reactor depletion conditions. The NRC staff's concern was that the number of lattices examined may have been too small to adequately characterize the inventory of fuel used at the GGNS. The response to RAI-2 provided in Attachment 1 to GNRO-2013/0050 (Reference 6) presents an analysis expanded to consider (( )) additional lattices. The expanded analysis adds reasonable confidence that an appropriate bounding lattice has been identified.
Bounding Lattice Identified with 95 percent Probability and 95 percent Confidence The proposed technical specifications will require that the SCCG peak k-infinity must be no greater than 1.26. For a specific SCCG k-infinity, the in-rack keff value varies with the lattice.
The bounding fuel assembly used in the analysis has a SCCG k-infinity that is slightly above the 1.26 k-infinity limit and its in-rack keff value must be bounding at a 95/95 probability and confidence. It was not clear from the analysis description provided in NEDC-33621 P that the identified bounding lattice indeed bounded the other lattices at a 95/95 probability/confidence level.
Sufficient additional discussion was provided in the response to RAI-2 in Attachment 1 to GNRO-2013/0050 (Reference 6) to adequately demonstrate that the bounding lattice is indeed bounding at a 95/95 level.
3.4.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties Fuel assembly manufacturing tolerances and uncertainties were evaluated using standard techniques based on keff sensitivity studies of parameter variation around the nominal model.
During the review, it was noted that some potentially significant tolerances and uncertainties were not evaluated.
The response to RAI-16 provided in Attachment 1 to GNRO-2012/00120 (Reference 4) included expanded analysis of tolerances and uncertainties. The response adequately addressed the NRC staff's concerns.
3.4.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on uranium-235 enrichment, fuel rod gadolinia content and distribution, and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also carry through to the spent nuclear fuel. Common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is more problematic. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup.
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- 18 3.4.3.4 Burnup Uncertainty
((
11 3.4.3.5 Axial Apportionment of the Burnup or Axial Burnup Profile The standard BWR SCCG peak k-infinity analysis technique uses either two-dimensional models or a three-dimensional model with uniform axial burnup distributions. Generally, this is appropriate because the peak in limiting assembly reactivity occurs at lower burnups where the uniform axial burnup distribution is conservative. If one were to credit assembly burnup beyond the limiting peak reactivity burnup, at some assembly burnup value, the use of the uniform axial burnup would become non-conservative.
The GGNS Region I analysis includes a departure from the standard method in that the explicit modeling of Boraflex gaps introduces additional axially dependent features (Le., Boraflex panel gaps) to fuel storage rack models. The concern is that these axially dependent features may impact the determination that it is conservative to use a uniform axial burnup distribution rather than a bounding axially-varying burnup distribution.
In response to this concern, additional analysis was performed and is documented in the response to RAI-1 provided in Attachment 1 to GNRO-2013/00050 (Reference 6). The discussion provided demonstrates that use of the uniform axial burnup distribution is still conservative for normal and credible abnormal conditions in this BWR SCCG peak k-infinity analysis.
3.4.3.6 Burnup History/Core Operating Parameters The reactivity of light-water reactor fuel varies with the conditions the fuel experiences in the reactor. This is particularly true for BWR fuel NCSA using the SCCG peak k-infinity analysis method. As a result of the usage of Gd 20 3 in fuel rods, fuel assembly reactivity initially increases as depletion of the gadolinium isotopes dominates the change in reactivity. At some point, the gadolinium isotopes are sufficiently depleted that the 235U depletion dominates the change in reactivity. The peak reactivity occurs at the transition point. The value of the in-rack keff at peak reactivity is affected by the reactor depletion parameters in several ways.
Factors that lead to a more thermal neutron energy distribution cause the 155&157Gd and fission products to deplete more quickly and reduce plutonium generation. This causes the peak reactivity condition to be reached earlier, achieving a higher in-rack keff value. Increased water OFFICIAL USE ONLY PROPRIETARY INFORMATION
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- 19 density and decreased void fraction lead to a more thermal neutron energy distribution and to lower fuel rod temperatures due to improved fuel rod cooling.
Factors that lead to a less thermal neutron energy distribution cause the 155&157Gd and fission products to be depleted more slowly and result in increased plutonium generation. Decreased water density, increased void fraction, and control rod usage all result in neutron energy spectrum hardening.
The GGI\IS analysis included consideration of reactor operating parameter variation by (1) evaluating the bounding lattices at (( )) void, and (2) using sensitivity study results of the impact of ((
)) to generate bias terms that are applied to the maximum in-rack keff value.
This approach is conservative because it effectively applies multiple conflicting conditions at the same time. For example, it is not realistic to include ((
)) All sensitivities that result in increased in-rack keff values are incorporated in the analysis as bias terms added to the in-rack maximum kelt value.
3.4.3.7 Integral Burnable Absorbers GGNS uses fuel assemblies in which some of the fuel rods contain Gd 2 0 3 . Pin-by-pin fuel compositions are calculated using the TGBLA code and are used in MCNP models of the fuel storage racks. At peak reactivity, the residual gadolinium isotopes are credited. The uncertainty associated with calculating the residual gadolinium isotopes is included in the "burnup uncertainty" discussed above in Section 3.3.3.4.
3.4.4 Analysis of Abnormal Conditions
((
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- 20
)) The NRC staff concludes that this TS change is acceptable.
3.4.5 Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated non-conservatively.
3.4.5.1 Potential Nonconservatisms Correlations between gap locations in co-located panels ignored As was noted in Section 3.3.2.2, there appears to be some correlation in gap locations in poison panels around each cell. Axially co-located gaps around one storage cell would increase neutronic interaction and increase the keff of the system. Ideally, the Monte Carlo style sampling and modeling of Boraflex degradation should have included correlated sampling of gap locations. This effect is ameliorated by the artificial adjustment of the range of gap locations and the number of gaps simulated per panel. The gap location distribution limits gap to the central 6 feet of the fuel assembly. The measured gap distributions show a significantly wider distribution in the gap locations. This restriction increases the probability that some axially co-located gaps are modeled. The distribution of the number of gaps is that all panels have at least one gap and 50 percent have one gap and the other 50 percent have two gaps.
Measurements reported in Figure 2 of Attachment 1 of GNRO-201 0100073 (Reference 10) indicate that there may be no gaps in 10 percent of the panels and more than two gaps in 20 percent of the panels. The artificially restricted distribution requires one or two gaps in every panel and ensures that the size of these one or two gaps will be large enough to conservatively bound the existence of more, but smaller, gaps.
((
)) As discussed in Section 3.3.1.1, the NRC staff noted other conservatisms that would bound this effect.
3.4.5.2 Potential Analysis Conservatisms The analysis includes aspects that add margin to the analysis.
((
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- 21
))
Uncredited actinides and fission products As was noted in Section 3.3.1, there are a large number of minor actinides and fission products that are not credited in this analysis. The uncredited nuclides could be worth up to about 1 to 2 percent Llk. Thus, these uncredited nuclides represent a significant unquantified conservatism.
Modeling of Boraflex degradation The analysis credits residual Boraflex with a 10B loading of 0.0133 g 10B/cm 2
- The TSs require that cells with Boraflex determined to have a 1°B loading of less than 0.0165 g 10B/cm 2 be removed from service or moved to Region II, which does not credit any residual Boraflex. The unquantified margin associated with this 20 percent 10B areal density reduction is significant conservatism.
3.4.5.3 Conclusion on Analysis of Margins The potential non-conservatisms associated with ((
)) can be quantified to some degree and have been offset as previously described. The !\IRC staff has concluded that not assuming that correlations between gap locations in co-located panels exist is ameliorated by the licensee's modeling artificiality there remains an uncertainty as to whether or not that artificiality is wholly sufficient. Therefore, the NRC staff has allocated the remaining analytical margin and unallocated conservatisms from the analysis to reach a reasonable assurance decision that the licensee's SFP TS will ensure compliance with 10 CFR 50.68.
3.5 The NFV NCSA NEDC-33621 P, Revision 0 (Reference 14), includes an analysis of the new fuel storage racks (NFSR). This analysis was needed to support inclusion of maximum nominal enrichment and maximum k-infinity limits in the revised TS. Specifically, the proposed TS limit the fuel assemblies stored to a maximum nominal uranium-235 enrichment of (( )) weight percent and limit the assembly k-infinity to no more than 1.26 in normal reactor core configuration and cold conditions.
This section documents the review of the NFV NCSA.
3.5.1 New Fuel Vault (NFV) NCSAs Method The analysis is to demonstrate compliance with the following requirements from 10 CFR 50.68:
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- 22 Paragraph 50.68(b)(2) of 10 CFR requires, The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.
Paragraph 50.68(b )(3) of 10 CFR requires, If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.
The GGNS NFV NCSA demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95 percent probability/95 percent confidence level. The licensee's analYSis documented in NEDC-33621 P, Revision 0, demonstrates that the keff value, includil1g bias and uncertainties, for full density water conditions is more than (( )) .6.k below the applicable limit from 10 CFR 50.68.
Typically, an optimum moderation study is also performed that identifies the maximum keff at reduced water densities for fuel stored in normally dry NFV storage racks. Section 5.5.1 of Reference 14 indicates the provision of 10 CFR 50.68(b)(3) that stipulates an optimum moderation analysis need not be performed is met. As the current GGNS TS do not have an optimum moderation keff limit, the NRC staff did not revisit that item in its review of this license amendment request.
The new fuel storage rack kef! values are all based on unirradiated fuel bundles. The standard cold core geometry maximum k-infinity values for each array were determined using TGBLA06A and its ENDF/B-V cross-section data. Consequently, the uncertainties include a contribution for the use of TGBLA06A to determine the in-core cold geometry k-infinity values.
As noted earlier MCNP-05P is a GEH proprietary version of MCNP5, which has been validated for use at the GGNS and ENDF/B-VII.O nuclear data were used for the new fuel storage rack analysis. The supporting validation is documented in Appendix A of NEDC-33621 P, Revision O.
Section 2 of HI-2094416 states the bias and 95/95 uncertainty from the validation are ((
)) This bias and bias uncertainty are similar to values reported for other analyses using ENDF/B-VII.O nuclear data. Considering that the values are consistent with similar analyses and that there is more than (( )) .6.k margin to the limits, further review of the validation was deemed unnecessary.
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- 23 The NRC staff reviewed the computational method and supporting validation and concludes it is acceptable.
3.5.2 NFV Fuel Storage Racks The steel structures that comprise the NFV fuel storage racks were conservatively not modeled.
Without the steel, the rack model simplifies down to constraints on the spacing and location of the fuel assemblies. All rack structures were modeled as full-density water.
The analysis was performed using radially and axially infinite arrays of fuel assemblies.
Replacing new fuel storage vault walls, floor, and overhead detail with additional fuel is a conservative modeling approximation.
The licensee is introducing a standard cold core geometry k-infinity limit into the relevant TS.
Consequently, the peak reactivity criticality analysis method used in the spent fuel storage racks is also applied to the new fuel storage racks.
3.5.3 Fuel Assemblies The new fuel storage rack analysis presented in Section 5.0 of NEDC-33621, Revision 0, evaluates only GE14 and GNF2 fuel assembly designs. While the analysis looked at a very limited number of fuel bundles variations, this is acceptable because of the conservative modeling approach and large margin, (( )) ilk to the 0.95 keff limit.
The fuel assembly design is conservative in that the fuel is effectively modeled as infinitely long.
Pin-by-pin enrichments, gadolinium fuel rods were explicitly modeled.
The fuel assembly uncertainties considered in the analysis did not include contributions related to (( )) In response to RAI-16 documented in Attachment 1 to GNRO-2012/00120 (Reference 4), these uncertainties were estimated for the spentJuel storage racks. The combined uncertainty for these two parameters was (( )) for the spent fuel racks. It is expected that the contribution from these uncertainties would be similar for the new fuel storage racks. Combination of these uncertainties with the other analyzed uncertainties would raise keff by about (( ))
Based on the NRC staff's review of the above, the fuel assembly model used in the NFV NCSA is adequately conservative.
3.5.4 MCNP Calculation Convergence Issue During the review, the NRC staff requested (see response to RAI-25 in Reference 4) that fission source convergence be checked for all MCNP cases. According to the RAI response, ((
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- 24
))
3.5.5 Analysis of Margins The licensee determined that the maximum kelf value, including biases and uncertainties, for the full density flooding case was (( )) The applicable limit is 0.95. The margin to this limit is (( )) The NRC staff concludes that this is acceptable.
A few small potential nonconservatisms were identified in the review. The margins to the limits are large enough to cover potential nonconservatisms associated with deficiencies in the analysis of tolerances and failure to consider a broader range of GE14 and GNF2 fuel assembly variations.
3.4.6 Boraflex Monitoring Program By letter dated November 9, 2010 (Reference 15), the NRC staff issued an RAI. RAI-4 of that letter requested that the licensee provide more specific information on the nuclear criticality analysis that demonstrates compliance with GDC 62. In the licensee's response dated November 23,2010 (Reference 10), the licensee provided supplemental information on Boraflex performance and its associated monitoring program.
The licensee utilizes Boraflex as the neutron absorber material to maintain the required subcriticality margin in the SFP storage racks. The Boraflex material is sandwiched between sheets of stainless steel with the edge strips welded in place to frame the Boraflex. The licensee stated that a Boraflex monitoring program has been established at GGNS to monitor Boraflex performance. The Boraflex monitoring program includes gap measurements, and SFP silica evaluations based on the EPRI RACKLIFE system, which is a computer program used to assess in-service performance of Boraflex. In addition, the licensee indicated that Blackness testing (i.e., an in-situ measurement technique) was performed on Boraflex panels placed in high gamma dose locations to evaluate the size of panel gaps. These panels received dose from freshly discharged fuel following a refueling outage for approximately 1 year before Blackness testing commenced, constituting one campaign. This testing was performed after each refueling outage, totaling seven campaigns.
The licensee stated that the results of the Blackness testing indicated that the total panel gap, as a percent of the initial panel length versus dose, follows the EPRI Boraflex shrinkage model until the dose exceeds 2.3x1010 rads. Furthermore, the licensee stated that the loss accelerates as the panels approach 3.0x10 10 rads.
The licensee reported that a Boron-10 (B-10) Areal Density Gauge for Evaluating Racks (BADGER) test campaign was conducted in 2007 to measure the B-10 areal density and panel loss from the gaps. BADGER testing is an in-situ technique used to measure B-10 areal density. The licensee stated that the BADGER test results confirmed that gap measurements for panels with doses below 2.3x1010 rads were consistent with the maximum shrinkage predicted by the EPRI Boraflex shrinkage model. The licensee also stated that the results of the BADGER test were consistent with the results of the previous seven Blackness tests.
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- 25 Furthermore, the licensee determined that the Boraflex material experienced additional degradation since the last blackness tests conducted in 1999.
The licensee stated that the 2007 BADGER test was conducted on 32 panels in Region I and Region II cell locations of the SFP. The licensee stated that the GGNS SFP is divided into cells that are labeled either Region lor Region" locations; the labels are used to identify the level of 10 dose accumulated. The Region I panels had accumulated doses up to 1.77x1 0 rads; 10 whereas, the Region" panels had accumulated doses as high as 3.83x10 rads. A Region I and Region II cell was reported to have a minimum B-10 areal density of 0.0182 and 0.0166 gram/square centimeter (g/cm2), respectively. The licensee stated that the Region I minimum areal density is above the criticality safety analysis assumption of 0.0133 gm/cm 2 and the Region II analysis does not credit any Boraflex. The licensee reported that the difference between the Region I BADGER test results and RACKLIFE results are bounded by a 95/95 2
uncertainty of 0.0022 g/cm . The licensee also stated that the 0.0022 g/cm 2 uncertainty is applied in the Borallex monitoring program to determine if a panel falls below the criticality analysis minimum areal density assumption of 0.0133 g/cm 2. Any cell that meets or exceeds the dose acceptance criterion of 2.3x1 010 rads or falls below the areal density acceptance criterion of 0.0182 g/cm 2 is configured as a Region II cell.
After reviewing the information the licensee provided regarding Boraflex performance at GGNS, the NRC staff determined that more information was needed to complete its review. The NRC staff issued RAls 2, 4, 5, 6, and 7 to the licensee on February 8, 2011 (Reference 25). The staff requested the licensee to discuss in detail the surveillance approach that will be used in the Boraflex monitoring program, specifically the methods of neutron attenuation testing (i.e., in-situ testing), frequency of inspection, sample size, data collection, and acceptance criteria. The staff requested the licensee to describe how the program's acceptance criteria account for potential degradation between surveillance periods. In addition, the staff requested the licensee to discuss the EPRI shrinkage model dose as it relates to the acceptance criteria analysis for continued operation. The staff also requested the licensee to discuss how the Region II locations are determined and added after each Blackness and BADGER test campaign; and the calibration technique and reference panel used for these types of tests.
In the licensee's response dated March 9, 2011 (Reference 11), the licensee stated that RACKLIFE will remain a significant component of the GGNS Boraflex monitoring program. The RACKLIFE calculations will continue to be performed each cycle and include projections of rack performance to the next RACKLIFE calculation. The licensee stated that the RACKLIFE analysis will compare the predicted silica to the plant measured silica. The RACKLIFE parameters will be adjusted as needed based on the comparison. The licensee stated that the analysis will include projections to the next planned RACKLIFE analysis date to ensure current Region I storage locations will not need to be reclassified as Region II storage locations. The licensee further stated that the EPRI shrinkage model has no direct impact on the criticality safety analysis; however, GGNS observed that a change in the Boraflex gap performance can occur at doses above that value. Boraflex panels which have received a gamma dose in excess of 2.3x10 10 rads or which have an areal density of less than 0.0165 g/cm2 are treated as Region II panels and are no longer credited in the criticality safety analysis. The licensee reported that the dose limit of 2.3x1 010 rads ensures the Boraflex gap configurations meet the NCSA assumptions. The licensee stated that the B-10 areal density acceptance criterion has been established from the summation of the NCSA assumed areal density, the
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- 26 BADGERlRACKLIFE uncertainty, and the design areal density tolerance (Le., 0.0133 +
0.0022 + 0.001 = 0.0165 g/cm2).
The licensee stated that an additional BADGER measurement will be performed prior to the end of 2013 to ensure that the BADGERlRACKLIFE uncertainty remains valid. The 2013 BADGER test campaign results and rack performance will be evaluated to determine the need for additional tests. It was reported that the 2013 test campaign will consist of at least 30 Boraflex panels. The licensee indicated that the BADGER to RACKLIFE uncertainty will be developed based on the test results of the planned 2013 test. The licensee stated that if the uncertainty value is less than the existing BADGERlRACKLIFE uncertainty of 0.0022 g/cm 2, the more conservative value will be considered acceptable. In addition, the current uncertainty value of 0.0022 g/cm2 will be adjusted if evaluation of the test results indicates that the uncertainty is less than the existing value. The minimum areal density results will be confirmed to be greater than the NCSA assumption of 0.0133 g/cm 2. The licensee stated that the acceptability of the minimum areal density and uncertainty will be based on verifying that all the NCSA distributions bound the corresponding BADGER measured distributions.
The NRC staff reviewed the information the licensee provided regarding Boraflex performance at GGNS and determined that more information was needed to complete its review. By electronic mail dated April 4, 2011 (Reference 26), the NRC staff issued additional RAls.
The NRC staff requested the licensee to discuss in detail the future (i.e., after 2012) surveillance approach and BADGER testing for the Boraflex material.
In the licensee's response dated May 3,2011 (Reference 13), the licensee stated GGNS will perform periodic testing of the Boraflex neutron absorbing material on a frequency of 5 years using BADGER testing. During the period between BADGER testing, the licensee intends to perform analyses to confirm subcriticality is maintained. The NRC staff notes that the licensee committed to performing BADGER and RACKLI FE testing at the intervals mentioned above (i.e., 5 years and each cycle, respectively). In a letter dated July 2, 2013 (Reference 6), the licensee further stated that the options for long-term strategy for maintaining subcriticality in its SFP per its licensing basis and the regulations include the following: 1) making use of additional dry cask storage to store spent fuel, 2) perform rack replacement, and/or 3) install neutron absorber inserts.
After reviewing the information the licensee provided on the Boraflex neutron absorber material and the Boraflex monitoring program, the NRC staff concludes that the program provides reasonable assurance that it will be able to detect, monitor, and mitigate Boraflex degradation.
This is accomplished by detection and monitoring of degradation in the use of the RACKLIFE computer code and BADGER testing, and reclassifying panels that do not meet the acceptance criteria from Region I to Region II locations. The NRC staff concludes that comparing predicted data from RACKLIFE to plant SFP measured silica each cycle will also provide information on the accuracy of the model to actual conditions. The NRC staff concludes that the 5-year frequency for BADGER testing and the conducting of further evaluation and analysis of material degradation between testing provides reasonable assurance that in between surveillances the Boraflex performance will be monitored and mitigated. In addition, performing in-situ BADGER testing will provide information on the amount of B-10 areal density present and material OFFICIAL USE ONLY PROPRIETARY INFORMATION
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- 27 degradation (i.e., reduction of neutron-absorbing capacity and loss of material) of the Boraflex SFP racks.
Based on a review of the dose limit and areal density acceptance criteria, the NRC staff has concluded that they provide reasonable assurance that the program will be able to mitigate degradation of Boraflex before it will challenge the criticality analysis as evaluated above.
3.6 Summary The NRC staff review of the GGNS new and spent fuel storage racks NCSA, documented in NEDC-33621 P, Revision 0, identified some non-conservative items. These items were evaluated against the margin to the regulatory limit and what the NRC considers an appropriate amount of margin attributable to conservatisms documented in the analyses.
The spent fuel storage analysis for Region I included a novel Monte Carlo sampling technique for crediting residual Boraflex. The NRC staff identified two issues with the method. The first issue is that statistical analysis of measured data was used to conclude that the distributions used to characterize the residual Boraflex were not correlated. From a physical perspective this does not seem realistic. All four panels around anyone storage location receive a major part of their radiation dose from the fuel assemblies stored in that cell. Clearly, the source of the damage is correlated. The NRC staff concluded this is acceptable as the licensee has offset this potential non-conservatism as discussed in Section 3.4.2.2, "Correlations between distributions." Secondly, the spent fuel manufacturing process very likely introduced correlated Boraflex panel restriction points and water flow paths. This would occur due to how the components were assembled and welded together. From the review of a sample of measured data for one randomly selected storage cell indicates that there is some correlation. The axial co-location of gaps around a storage cell would increase keff. This effect is offset by artificial adjustments made to the distributions used to simulate the Boraflex damage. In particular, the simulated damage is restricted to the central 6 feet of the panel and all panels are modeled as having either one or two Boraflex gaps. In reality, the gaps are spread over a much larger range, 10 percent of the examined panels had no gaps and 20 percent of the panels had more than two gaps. More small gaps should yield a lower keff compared to the same total loss modeled in one or two gaps. The NRC staff concluded this is acceptable as additional margins are available to offset this potential non-conservatism discussed in Section 3.4.5, "Margin Analysis and Comparison with Remaining Uncertainties."
The distributions used to simulate the Boraflex degradation can be compared with new data as data become available. This will ensure that the characteristics of the Boraflex degradation have not changed enough to invalidate the distributions used in criticality analysis. Further, following some initial radiation dose, panel dissolution continues even in the absence of significant continuing dose. As new data becomes available regarding the continuing dissolution and evolution of Boraflex gap formation. the bases for the GGNS SFP criticality analysis can be confirmed.
The NRC staff has also reviewed the licensee's evaluation of the effects of the proposed changes on the Boraflex neutron absorber material used at GGNS and concludes that the licensee has addressed the impact of the changes on the Boraflex SFP racks. Further, the licensee has provided the NRC staff with reasonable assurance that the Boraflex monitoring OFFICIAL USE ONLY PROPRIETARY INFORMATION
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- 28 program will continue to detect degradation of Boraflex material and meet the requirements of 10 CFR 50.68(b)(4), SRP Section 9.1.2, and GDC 62. Therefore, the NRC staff concludes the proposed changes are acceptable with respect to the Boraflex material and monitoring program.
Based on the an acceptable Boraflex monitoring program, the review of the supporting analysis reports and based on the margins to regulatory limits, crediting some analysis conservatism, and including consideration of the identified potential non-conservatisms, the NRC staff concludes that there is a reasonable assurance that the GGNS SFP and NFV fuel storage racks meet the applicable NCSA regulatory requirements. Therefore, the NRC staff concludes that the TS changes are acceptable.
In addition as stated in Section 3.2, the licensee has proposed to remove License Condition 45.
The license condition was approved for one cycle. Therefore, the NRC staff concludes that the deletion of this license condition is acceptable with the approval of this amendment.
4.0 REGULATORY COMMITMENTS By letters dated October 1, 2012, and July 2 and September 5, 2013, Entergy made the following regulatory commitments:
- 1) The [NCSA] LAR submitted via Entergy letter GNRO-2011/00076 (ADAMS Accession #ML1125321287) proposed a change to TS 4.3.1 that specified six of 16 cells in a Region 114x4 cell array will be blocked.
In order to reflect that eight rather than six cells will be blocked as determined in the criticality safety analysis, Entergy is processing a change to proposed TS 4.3.1. This revision will be submitted under separate letter on or before October 22, 2012.
- 2) While the contribution to the calculated panel dose from diagonally adjacent fuel assemblies is relatively small, an adjustment to the dose calculations will be augmented to include this effect for empty cells.
- 3) The zero-dose panels in cells Z014, Z016, and Zr15 will be included in the upcoming BADGER test campaign.
- 4) In order to ensure the criticality analysis remains applicable, the Boraflex monitoring program will be modified to incorporate updating the gap growth due to dissolution following each BADGER campaign. The updated dissolution rate will be applied to the most recent BADGER results through the end of the next BADGER test interval plus one year to confirm the continued applicability of the criticality analysis.
- 5) The Boraflex monitoring program will be modified to confirm that the analyzed dissolution features remain bounding.
- 6) Entergy will maintain a minimum distance of 12 inches between any fuel stored in the Control Blade/Defective Fuel Storage Rack (Module H1) and in the surrounding high density spent fuel pool storage racks.
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- 29 The licensee stated that Regulatory Commitment 1 was closed in the letter dated October 22, 2012. The NRC staff concludes that reasonable controls for the implementation and for subsequent evaluation of proposed changes pertaining to the above regulatory commitment are best provided by the licensee's administrative processes, including its commitment management program.
5.0 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.
6.0 REFERENCES
- 1. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request, Criticality Safety Analysis and Technical Specification 4.3.1, Criticality, Grand Gulf Nuclear Station, Unit 1," dated September 9, 2011 (GNRO-2011/00076) (ADAMS Accession No. ML112521287).
- 2. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Criticality Safety Analysis, Grand Gulf Nuclear Station, Unit 1," dated November 21,2011 (GNRO-2011/00104) (ADAMS Accession No. ML113320260).
- 3. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to NRC Requests for Additional Information Pertaining to License Amendment Request for Criticality Safety Analysis, Grand Gulf Nuclear Station, Unit 1,"
dated April 18, 2012 (GNRO-2012/00027) (ADAMS Accession No. ML12109A281).
- 4. Ford, B.S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Responses to NRC Requests for Additional Information - GGNS Criticality, Safety Analysis License Amendment Request, Grand Gulf Nuclear Station, Unit 1," dated October 1, 2012 (GNRO-2012/00120) (ADAMS Accession No. ML12276A152).
- 5. Richey, M. L., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to NRC Request for Additional Information - GGNS Criticality Safety Analysis License Amendment Request, Grand Gulf Nuclear Station, Unit 1," dated October 22,2012 (GNRO-2012/00124) (ADAMS Accession No. ML12296A417).
- 6. Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request - Responses to NRC Requests for Additional Information, Grand Gulf Nuclear Station, Unit 1," dated July 2, 2013 (GNRO-2013/00050)(ADAMS Accession No. ML13190A043).
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- 30
- 7. Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request - Supplemental Information, Grand Gulf Nuclear Station, Unit 1," dated September 5, 2013 (GNRO-2013/00066)
(ADAMS Accession No. ML13249A235).
- 8. Ford, B. S., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Criticality Safety Analysis License Amendment Request - Supplemental Information, Grand Gulf Nuclear Station, Unit 1," dated September 16, 2013 (GNRO-2013/00073)
(ADAMS Accession No. ML13260A084).
- 9. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request, Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated September 8, 2010 (GNRO-201 0/00056) (ADAMS Accession No. ML102660409).
- 10. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Supplemental License Amendment Request, Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated November 23, 2010 (GNRO-2010100073) (ADAMS Accession No. ML103330093).
- 11. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated March 9,2011 (GNRO-2011/00017) (ADAMS Accession No. ML110680507).
- 12. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1," dated April 21, 2011 (GNRO-2011/00025) (ADAMS Accession No. ML11112A098).
- 13. Krupa, M. A, Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Request for Additional Information Regarding Extended Power Uprate, Grand Gulf Nuclear Station, Unit 1,I! dated May 3, 2011 (GNRO-2011/00034) (ADAMS Accession No. ML111240288).
- 14. GE-Hitachi Nuclear Energy Americas, LLC, "Grand Gulf Nuclear Station, Fuel Storage Criticality Safety Analysis of Spent and New Fuel Storage Racks," NEDC-33621 P, Revision 0, November 2010 (not publicly available - proprietary); public version designated as NEDO-33621, Revision 0, November 2010 (ADAMS Accession No. ML103330092).
- 15. Wang, A S., U.S. Nuclear Regulatory Commission, letter to Entergy Operations, Inc.,
"Grand Gulf Nuclear Station. Unit 1 - Supplemental Information Needed for Acceptance of License Amendment Request for an Extended Power Uprate (TAC No. ME4679),"
dated November 9,2010 (ADAMS Accession No. ML103010200).
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- 16. Electric Power Research Institute Report NP-6159, "An Assessment of Boraflex Performance in Spent-Nuclear-Fuel Storage Racks," prepared by Northeast Technology Corp., December 1988 (ADAMS Accession No. ML003736666).
- 17. U.S. Nuclear Regulatory Commission, Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks," dated June 26, 1996 (ADAMS Accession No. ML031110008).
- 18. HOLTEC Report HI-992255, "Blackness Testing of Boraflex in Selected Spent Fuel Storage Rack Cells of the Grand Gulf Nuclear Station," prepared for Entergy Operations, Inc., July 1, 1999. Excerpts provided as enclosure 2 of Entergy Operations Inc. letter dated April 21, 2011 (ADAMS Accession No. ML111120329).
- 19. Northeast Technology Corp. Report NET-287-01, Revision 1, "BADGER Test Campaign at Grand Gulf Nuclear Station," prepared for Entergy Nuclear Operations, Inc.,
October 1, 2010. Provided as enclosure 1 of Entergy Operations Inc. letter dated April 21, 2011 (ADAMS Accession No. ML11112A099).
- 20. Kopp, Sr., L., U.S. Nuclear Regulatory Commission, memorandum to T. Collins, U.S. Nuclear Regulatory Commission, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," dated August 19, 1998 (ADAMS Accession No. ML003728001).
- 21. U.S. Nuclear Regulatory Commission, "Final Division of Safety Systems Interim Staff Guidance DSS-ISG-2010-01, Revision 0, 'Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools,'" dated October 13, 2011 (ADAMS Accession No. ML110620086).
- 22. J.M. Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses Criticality (keff) Predictions," NUREG/CR-7109 (ORNLITM-2011/514), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012 (ADAMS Accession No. ML12116A128).
- 23. J.C. Wagner and C.v. Parks, "Recommendations on the Credit for Cooling Time in PWR Burnup Credit Analyses," NUREG/CR-6781 (ORNLlTM-2001/272), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2003 (ADAMS Accession No. ML030290585).
- 24. D.E. Mueller, K.R. Elam, and P.B. Fox, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," NUREG/CR-6979, ORNLlTM-20071083, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2008.
- 25. Wang, A. B., U.S. Nuclear Regulatory Commission, e-mail to Entergy Operations, Inc.,
"Grand Gulf, Unit 1, Request for Additional Information, NRRlDCIICSGB Review, Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt (TAC No. ME4679)," dated February 8, 2011 (ADAMS Accession No. ML110390173).
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- 26. Wang, A. B., U.S. Nuclear Regulatory Commission, e-mail to Entergy Operations, Inc.,
"Grand Gulf, Unit 1, Request for Additional Information, Round 2, NRRlDCI/CSGB Review, Amendment Request for Extended Power Uprate to Increase the Maximum Reactor Core Power Operating Limit from 3898 to 4408 Mwt (TAC No. ME4679)," dated April 4, 2011 (ADAMS Accession No. ML110940136).
Principal Contributors: A. Obodoako and K. Wood Date: September 25, 2013 OFFICIAL USE ONLY PROPRIETARY INFORMATION
OFFICIAL USE ONLY - PROPRIETARY INFORMATION
- 2 safety evaluation, which is provided in Enclosure 3. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, IRA!
Alan Wang, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosures:
- 1. Amendment No. 195 to NPF-29
- 2. Proprietary Safety Evaluation
- 3. Non-proprietary Safety Evaluation cc w/Enclosures 1 and 3: Distribution via Listserv DISTRIBUTION:
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