BVY 17-031, Response to Request for Additional Information Re Exemption Request from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign
| ML17255A236 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee, Holtec  |
| Issue date: | 09/07/2017 |
| From: | Chappell C Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| References | |
| BVY 17-031, CAC L25219 | |
| Download: ML17255A236 (11) | |
Text
- ~Entergy Entergy Nuclear Operations, Inc.
Vermont Yankee 320 Governor Hunt Rd.
Vernon, VT 05354 802-257-7711 Coley C. Chappell Manager, Design and Programs 10 CFR 72.7 BVY 17-031 September 7, 2017 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Response to Request for Additional Information Related to Exemption Request from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign (CAC No. L25219)
Vermont Yankee Nuclear Power Station License No. DPR-28 Docket Nos. 50-271, 72-59 and 72-1014
REFERENCES:
- 1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Exemption Request from Certain Requirements of 10 CFR 7?.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign," BVY 17-006, dated May 16, 2017 (ML17142A358)
- 2. Letter, USNRC to Holtec International, "Certificate of Compliance No. 1014, Amendment No. 1O for the HI-STORM 100 Cask System (CAC No. L24979)," dated May 25, 2016 (ML16144A177)
- 3. Letter, Holtec International to USNRC, "Holtec International HI-STORM 100 Multipurpose Canister Storage System Amendment Request 1014-11," dated January 29, 2016 (ML16029A528 - package)
- 4. Letter, Holtec International to USNRC, "Holtec International HI-STORM 100 Multipurpose Canister Storage System Amendment Request 1014-12," dated June 14, 2016 (ML16169A363- package)
- 5. Letter, USNRC to Holtec International, "Amendment No. 12 to Certificate of Compliance No. 1014 for the HI-STORM 100 Cask System - Request for_Additional Information, dated July 21, 2017 (ML17181A015)
- 6. E-Mail to T.B. Silko/Entergy from Yen-Ju Chen/DSFM re: "Response:
Additional Information Requested," dated August 3, 2017 (ML17219A047)
Dear Sir or Madam:
By letter dated May 16, 2017 (Reference 1), Entergy Nuclear Operations, Inc. (ENO) submitted ) [) O(
a request for exemption from certain requirements of 10 CFR 72.212 and 72.214 for Vermont
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BVY 17-031 I Page 2 of 3 Yankee Nuclear Power Station (VY). These regulations require, in part, compliance with the terms and conditions of the Holtec International (Holtec) Cask System Certificate of Compliance (CoC) for spent fuel storage at the VY's independent spent fuel storage installation.
Specifically, the requested exemption would allow the loading of selected fuel assemblies with shorter cooling times and higher heat loads than those specified in Reference 2 into MPC-68M multi-purpose canisters (MPCs), as well as an optional loading pattern for the MPC-68M. VY's requested exemption would permit specific changes as contained within proposed Amendments Nos. 11and12 of the HI-STORM 100 CoC No. 72-1014 (References 3 and 4), as incorporated by reference in VY's exemption request.
Subsequent to VY's exemption request, the U.S. Nuclear Regulatory Commission (NRC) submitted to Holtec a Request for Additional Information (RAI) (Reference 5) in response to Holtec's Amendment 12 request. As documented in Reference 6, the NRC requested VY provide a response to RAI questions 8-5, 8-7, 8-1 O and 8-11 as provided in the letter to Holtec dated July 21, 2017 (Reference 5), tailoring these questions as applicable to VY's exemption request. The response to the RAI is provided in Attachment 1.
In addition, Attachment 2 of the VY exemption request provided supporting markups of the Final Safety Analysis Report pages from the HI-STORM 100 Amendment request No.12, and it has been identified that the proposed markup to Figure 4.111.4 should have been included.
Accordingly, this is provided in Attachment 2 to this submittal.
This letter contains no new regulatory commitments.
Should you have any questions concerning this letter, please contact me at (802) 451-3374.
Sincerely, CCC/tbs Attachments: 1. Response to Request for Additional Information Related to Exemption Request from Certain Requirements of 10 CFR 72.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign
- 2. Markup of Figure 4.111.4 from the HI-STORM 100 Amendment Request No.12 cc: Mr. Daniel H. Dorman Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713
BVY 17-031 I Page 3 of 3 Mr. Jack D. Parrott, Sr. Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-8F5 Washington, DC 20555 Ms. Yen-Ju Chen, Sr. Project Manager Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Mail Stop T-48-34 Washington, DC 20555 Ms. June Tierney, Commissioner Vermont Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05602-2601
BVY 17-031 Docket Nos. 50-271, 72-59 and 72-1014 Attachment 1 Vermont Yankee Nuclear Power Station Response to Request for Supplemental Information Related to Exemption Request from Certain Requirements of 10 CFR 72.212 and 1o CFR 72.214 to Support the Dry Fuel Loading Campaign
BVY 17-031 I Attachment 1 I Page 1 of 5 REQUEST FOR ADDITIONAL INFORMATION RELATED TO EXEMPTION REQUEST FROM CERTAIN REQUIREMENTS OF 10 CFR 72.212 AND 10 CFR 72.214 TO SUPPORT THE DRY FUEL LOADING CAMPAIGN FOR VERMONT YANKEE NUCLEAR POWER STATION
Background
By letter dated May 16, 2017 (Reference 1), Entergy Nuclear Operations, Inc. (ENO) submitted a request for exemption from certain requirements of 10 CFR 72.212 and 72.214 for Vermont Yankee Nuclear Power Station (VY). These regulations require, in part, compliance with the terms and conditions of the Holtec International (Holtec) Cask System Certificate of Compliance (CoC) for spent fuel storage at the VY's independent spent fuel storage installation.
Specifically, the requested exemption would allow the loading of selected fuel assemblies with shorter cooling times and higher heat loads than those specified in Reference 2 into MPC-68M multi-purpose canisters (MPCs), as well as an optional loading pattern for the MPC-68M. VY's requested exemption would permit specific changes as contained within proposed Amendments Nos. 11 and 12 of the HI-STORM 100 Coe No. 72-1014 (References 3 and 4). A portion of VY's exemption request incorporated by reference Holtec's above discussed amendment requests.
Subsequent to VY's exemption request submittal, the U.S. Nuclear Regulatory Commission (NRG) submitted to Holtec a Request for Additional Information (RAI) (Reference 5) in response to Holtec's Amendment 12 request. As documented in Reference 6, the NRG requested VY provide a response to RAI questions 8-5, 8-7, 8-1 O and 8-11 as provided in the letter to Holtec dated July 21, 2017 (Reference 5), tailoring these questions as applicable to VY's exemption request.
In support of VY's exemption request and in support of our response to the subject RAI, ENO hereby incorporates by reference the documents listed in the References section of this attachment.
Request for Additional Information:
Chapter 8 - Materials Evaluation 8-5 Provide a justification for the overpack inner shell maximum temperature during normal storage with a MPC-68M (proposed FSAR Table 4.lll.3b) which exceeds the long-term normal design temperature limits in FSAR, Rev. 13, Table 2.2.3. This information is necessary to assure compliance with 10 CFR 72.236(b).
Response
The discussion provided in the response to 8-7 below with describing the substantial margin from the maximum aggregate heat load of any cask in the VY specific loading plan to the cask heat load limit in Amendment 1O is also applicable to this question related to overpack inner shell temperatures.
BVY 17-031IAttachment1IPage2 of 5 8-7 Provide a justification or analysis to support the off-normal and accident temperatures for concrete in the proposed FSAR Tables 4.111.7, .4.111.15, and 4.111.17 that exceed the maximum accident temperature limits for concrete listed in ACl-349-85. In addition, clarify the required recovery plan for the overpack following the off-normal and accident conditions where the concrete temperatures exceed the maximum accident temperature limits in ACl-349-85.
NUREG/CR-6900 (NRG, 2006) includes a summary of the potential concrete degradation mechanisms that occur at elevated temperatures. The analysis should include an assessment of these degradation mechanisms and their effects on the safety function of the concrete in the HI-STORM overpack.
This information is necessary to assure compliance with 10 CFR 72.236(b).
Response
Proposed Amendment 12 (Reference 3) introduces a new loading pattern for the MPC-68M, identified as new Appendix B Figure 2.4-1. Under this loading pattern, an MPC-68M could be loaded to 42.8 kW (maximum) as compared to the approved Amendment 1O Design Basis heat load of 36.9 kW (maximum heat load for regionalized loading), as described in HI-STORM 100 FSAR, Revision 14, Section 2.1.9.1.2. The additional heat load allowed by the proposed Amendment 12 loading pattern is the underlying basis of question 8-7. While the technical question can be satisfactory addressed through the review of proposed Amendment 12, in order to resolve this question for the review of the VY exemption request, ENO is proposing to limit the actual total (aggregate) cask heat load to less than or equal to the 36.9 kW as allowed in Amendment 10, thus remaining within the approved Amendment 1O analysis applicable to this question related to overpack concrete integrity.
The VY loading plan for all casks has been developed and provides considerable conservatism with respect to the Amendment 1O limit of 36.9 kW.
For example, the aggregate heat load of any cask scheduled to be loaded utilizing the subject exemption request is not greater than 24.5 kW, ranging from 17.0 kW to 24.5 kW. Therefore, for the VY exemption request, there is a conservatism of at least approximately 33% from the cask heat load limit in Amendment 10. ,
In addition, on a per cell basis, there is also considerable conservatism between VY's actual cask loading plan and the per cell heat load limits allowed by the Quarter Symmetric Heat Load (QSHL) pattern described in Appendix B Figure 2.4-1 of Holtec's proposed Amendment 12. For example, the QSHL pattern of Figure 2.4-1 provides sixteen (16) cell locations (per cask) with a decay heat load limit of 1.2 kW for loading fuel assemblies. The VY site specific loading plan includes 132 fuel assemblies with decay heat loads >0.71 kW1, ranging up to 0.91 kW for the highest. Therefore, for the VY site specific exemption request, there is a conservatism of at least 24% for the VY fuel assemblies loading into these locations with respect to the maximum allowable per cell heat load limit in Figure 2.4-1 of the proposed Amendment 12.
710 watts (.71 kW) is the maximum per cell heat load limit under Amendment 10 utilizing regionalized loading.
BVY 17-031 I Attachment 1 I Page 3 of 5 Accordingly, the substantial margin from the cask aggregate decay heat loads in the VY specific loading plan to the cask heat load limit in the existing Amendment 1O provides reasonable assurance that the analyses supporting Amendment 10 with respect to the integrity of the overpack concrete adequately address the changes proposed for the VY exemption request.
8-10 Revise Attachment 2 for Holtec Letter 5014812, "Structural Calculation Package for MPC, (Holtec Report Hl-2012787) and provide the following:
a) Fracture toughness estimations for Metamic-HT as a function of temperature using the mechanical properties reported in "Metamic-HT Qualification Sourcebook, (Holtec Report Hl-2084122 Revision 10) to support the calculation of a minimum flaw size for crack propagation.
b) Additional information with respect to the composition of Metamic-HT including the minimum and maximum boron carbide and aluminum oxide loading to allow a comparison of the estimated fracture toughness values for Metamic-HT to measured fracture toughness values of particle reinforced aluminum metal matrix composites.
In applicant's structural calculation analysis, the Metamic-HT fracture toughness value provided was stated to be based on an estimate by NRC staff using Charpy impact data and a correlation between Charpy data and fracture toughness based on pressure vessel steels. The applicant stated that:"[ ... ] Based on CVE correlations for steels, the critical stress intensity factor of Metamic-HT basket was estimated by the NRG reviewer [1] to 112 be Ktc= 30 ksi in {. ** }." The applicant further stated that the estimated value of fracture was comparable to the range of fracture toughness for aluminum alloys, which tend to be 112 in the range of 18.2 to 45.5 ksi in
- Based on the estimated fracture toughness value, the applicant calculated a minimum flaw size for crack propagation, a, using the equation for the stress intensity for a plate with an edge crack (Holman and Langland, 1981) to be 1.275 inches or more than 20x greater than the allowable or detection flaw size stated by the applicant to be 1/16" (0.0625").
The staff reviewed the applicant's analysis and determined that the statement claiming that NRC staff estimated a value for the Metamic-HT critical stress intensity factor is not accurate. A review of the conversation record (ML092330054 and ML092440495) show that the NRC staff at the time were questioning the fracture toughness of Metamic-HT.
The NRC staff estimation of the fracture toughness of an aluminum metal matrix composite (MMC) using a correlation developed for pressure vessel steels was understood to be an estimate. In addition, the staff note that estimating the fracture toughness for aluminum MMCs using data for aluminum alloys is not an established practice. Fracture toughness of aluminum alloys is dependent on a number of factors, including composition and condition for aluminum alloys that can be age hardened (ASM, 1998).
The Metamic-HT Qualification Sourcebook (Holtec Report Hl-2084122 Revision 10) includes mechanical properties of Metamic-HT over a wide range of temperatures.
Although fracture toughness is not directly measured, the Metamic-HT Qualification Sourcebook Attachment E includes: (1) calculation of the required fracture toughness as
BVY 17-031 /Attachment 1 I Page 4 of 5 a function of peak stress and crack size and (2) a correlation equation developed for structural steels that can be used to calculate fracture toughness using the Charpy Notch (Cv) data and Young's Modulus (E) data.
The staff notes that the estimation of fracture toughness for Metamic-HT using a correlation equation developed for structural steel is not an established practice.
Estimated values of fracture toughness for Metamic-HT can be compared to measured values of fracture toughness for particle reinforced aluminum metal matrix composites if the composition of Metamic-HT including the minimum and maximum boron carbide and aluminum oxide loading are provided. This is necessary to show that the use of the correlation equation for pressure vessel steels provides similar values of fracture toughness compared to reported fracture toughness values for analogous particle reinforced aluminum metal matric composites. The staff reviewed the available literature on fracture toughness measurements for aluminum metal matrix composites (Flom et al.,
1989; Flom and Arsenault, 1989; Lewandowski, 2000; Miserez, 2003; Rabiei et al.,
2008). Numerous aluminum MMCs exist which utilize a variety of aluminum alloys and ceramic particle compositions. Commonly used particle compositions include SIC, Al 20 3 ,
and 84C. Based on the available information on fracture toughness of aluminum MMCs, the range of fracture toughness values spans from 8 to 30 ksi*in 112
- Several factors can influence the fracture toughness of aluminum metal matrix composites including:
(1) particle composition, (2) particle size, (3) particle loading, (4) particle distribution or clustering, (5) alloy composition and, (6) alloy condition for aluminum alloys that can be age hardened. The applicant should consider whether technical information in addition to the Metamic-HT composition should be provided to support a comparison of the fracture toughness values for Metamic-HT.
This information is necessary to assure compliance with 10 CFR 72.236{b).
Response
The response to this question provided by Holtec in Reference 6 is applicable to the VY exemption request with no changes or additions, and is incorporated by reference.
8-11 Provide information on potential strength degradation of aluminum basket shims by thermal over-aging of precipitation-hardened microstructure.
The application addressed the use of aluminum alloy basket shims primarily in thermal performance. The applicant assumes aluminum alloy to be effective for the short duration dynamic loading from the tip-over accident. Aluminum alloy, such as Alloy 2219, used by Holtec is precipitation-hardened alloy. The application shows the shims temperature could be as high as 295°C {563°F) under normal conditions (FSAR Table 3.111.3 and Table 4.lll.3b). Literature data shows that over-aging and accompanying strength degradation could occur at 21 O - 240°C in a few hours (for Alloy 2219 in Rafi Raza et al., 2011 ).
It is unclear to the staff whether the structural analysis adequately accounts for potential degradation of strength of aluminum alloy for prolonged conditions including normal conditions as discussed in HI-STAR SAR Section 2.2 (Holtec International, 2017). The
BVY 17-031 I Attachment 1 I Page 5 of 5 staff requests that the applicant (i) provide justification that the current tip-over analysis in the design basis is valid, (ii) revise the analysis to adequately account for the degradation of aluminum alloy strength, or (iii) state that the type of Alloy 2219 (e.g., 2219-0) is in the annealed conditions which would not be subje;:ct to degradation of strength due to over-aging.
This information is needed to determine compliance with 10 CFR 72.236{b).
Response
The response to this question provided by Holtec in Reference 6 is applicable to the VY exemption request with no changes or additions, and is incorporated by reference.
References
- 1. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Exemption Request from certain requirements of 10 CFR 72.212 and 10 CFR 72.214 to Support the Dry Fuel Loading Campaign," BVY 17-006, dated May 16, 2017 (ML17142A358)
- 2. Letter, USNRC to Holtec International, "Certificate of Compliance No. 1014, Amendment No. 1O for the HI-STORM 100 Cask System (CAC No. L24979)," dated May 25, 2016 (ML16144A177)
- 3. Letter, Holtec International to USN RC, "Holtec International HI-STORM 100 Multipurpose Canister Storage System Amendment Request 1014-12," dated June 14, 2016 (ML16169A363 - package)
- 4. Letter, USNRC to Holtec International, "Amendment No. 12 to Certificate of Compliance No. 1014 for the HI-STORM 100 Cask System - Request for Additional Information."
Dated July 21, 2017(ML17181A015)
- 5. E-Mail to T.B. Silko/Entergy from Yen-Ju Chen/DSFM re: "Response: Additional Information Requested," dated August 3, 2017 (ML17219A047)
- 6. Letter, Holtec International to USNRC "Submittal of Responses to NRC's RAls for HI-STORM 100 LAR 1014-12," dated August 25, 2017
BVY 17-031 Docket Nos. 50-271, 72-59 and 72-1014 Attachment 2 Vermont Yankee Nuclear Power Station Markup of Figure 4.111.4 from the HI-STORM 100 Amendment Request No.12 (2 pages total including this cover sheet)
BVY 17-031 I Attachment 2 I Page 1*of 1 Table 4.:rn.4: MaximumPresS1rres Under Normal Long Tenn Storage Condition Pressul'e (psig)
Initial maxinnun backfill* (at 468-.5 70°F)
Nonnal:
intact rods 98.7:5.5 1% rods ruprure** 9.9..U Off-Nonnal (10% rods rupture) 104.'Q.(b3.
Accident (100% rods rupture) 152.04~
- Comervativdy asswned at the Tech. Spec. maxinu.un value (see Table 4.4.12).
- Per NUREG-1536!1 pressure analysis with :ruptured fuel rods (meluding BPRt\: feels fur PJJJR.foel) is perforated with release of 100% of the ruptured fuel rod fill gas and 30% of the significant :radioactive gaseous fission products.