L-19-008, Partial Response to NRC Letter Dated April 30, 2019, Request for Additional Information for the Technical Review of the Application for Renewal of the Humboldt Bay Independent Spent Fuel Storage Installation License No. SNM-2514 (CAC No. 00: Difference between revisions

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{{Adams
#REDIRECT [[WBL-19-008, American Society of Mechanical Engineers, Section Xl, Third 10-Year Inspection Interval, Inservice Inspection Owners Activity Report for Cycle 15 Operation]]
| number = ML19197A026
| issue date = 07/01/2019
| title = Partial Response to NRC Letter Dated April 30, 2019, Request for Additional Information for the Technical Review of the Application for Renewal of the Humboldt Bay Independent Spent Fuel Storage Installation License No. SNM-2514 (CAC No. 00
| author name = Welsch J
| author affiliation = Pacific Gas & Electric Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NMSS
| docket = 07200027
| license number = SNM-2514
| contact person =
| case reference number = CAC 001028, HIL-19-008
| package number = ML19197A028
| document type = Legal-Affidavit, Letter type:L, Response to Request for Additional Information (RAI)
| page count = 35
| project = CAC:001028
| stage = Request
}}
 
=Text=
{{#Wiki_filter:Enclosure 3 to this letter contains Confidential Information - Withhold Under 10 CFR 2.390 Ill Pacific Gas and
* r[tf&~.* Electric Company" James M. Welsch
* Senior Vice President Generation and Chief Nuclear Officer Diablo Canyon Power Plant P.O. Box 56 Avila Beai:h, CA 93424 805.545.3242 E-Mail: James.Welsch@pge.com
~uly 1, 201'9 PG&E Letter HIL-19"'.008 ATTN:
* Document Control Desk Director, Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission
. Washington, DC 20555-0001 Docket No. 72-27, Materials License No. SNM-2514 Humboldt Bay Independent Spent Fuel Storage Installation 10 CFR 72.42 Partial Response to NRC Letter dated April 30, 2019, "Request for Additional Information for the Technical Review* of the Application for Renewal of the Humboldt Bay Independent Spent Fuel Storage Installation License No. SNM-2514 (CAC No. 001028)"
 
==References:==
: 1. PG&E Letter HIL-:18-006, License Renewal Application for the Humboldt B:ay Independent Spent FueJ Storage Installation, dated July 10, 2018 (ML 1_8215A180 and ML18215A213)
: 2. PG&E Letter HIL-18-008, Response to NRC Letter dated September 20, 2018, "Request for Supplemer:,tal Information for the Technical Review of the Application for Renewal of the Humboldt Bay lndependentSpent Fuel Storage Installation License No. SNM-2514 (CAC No. 001028)", dated October 22, 2018 (ML18330A050)
: 3. NRC Letter, "Request for Additional Information for the Technicc;:11 Review of the Application for Renewal of the Humboldt Bay Independent Spent Fuel Storage Installation.License No. SNM.:.2514 (CAC No. 001028)", dated April 30, 2019 (ML19122A231 and ML 19_122A230)
 
==Dear.Commissioners and Staff:==
By Pacific Gas and Electric Company (PG&E) letters HIL-18-006, dated July 10, 2018 (Reference 1) and HIL-18-008, dated October 22, 2018 (Reference 2), PG&E*
submi~ed a License Renewal Application (LRA) to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Materials License SNM-2514, for the Humboldt Bay (HB) Independent Spent Fuel Storage Installation (ISFSI).
* *.. 5 S ZD
/Jl-1
. Enclosure 3 contains Confidential information - Withhold Under 10 CFR 2.390
*1v~S SZ/p When separated from Enclosi:fre 3, this document i_s decontrqlled to this letter contains Confidential ln{orniatiori - Withhold Under,11 O CFR 2.390 Document Control Desk July 1, 2019 Page 2*
PG&E Letter HIL-19-00.8 By NRC {{letter dated|date=April 30, 2019|text=letter dated April 30, 2019}}, (CAC/EPID No. 001028/L-2018-RNW-0016, Reference 3) the NRC Staff requested additional information to support their review of the HB ISFSI LRA. As discussed with the NRC Project Manager, PG&E will provide responses to request for additional information (RAI) 2-1 Part 9, RAI 3-4 Part 1, RAI A-6, and RAI A-7 by July 31, 2019.
* Enclosure 1 contains PG&E's responses to the RAls except for those noted above. contains Revision 2 of the LRA, resulting *from the RAI responses with the changes designated by change bars in the left margin. The LRA is being provided on one disk labeled, "Humboldt Bay Independent Spent Fuel Storage Installation Site Specific License Renewal Application, Revision 2, June 2019."
*Enclosure 3 provides the HI-STAR overpack closure.plate seals certificate of conformance, as referenced in response to RAI 3-3. Enclosure 3 contains *
* confidential.information that should be withheld from public disclosure in accordance with 10 CFR 2.390. contains an Affidavit pursuant to 10 CFR 2.390. The Affidavit sets forth the basis for which specific information includ~d in Enclosure 3 may be withheld from public disclosure by the Commission and addresses the considerations listed in 10 CFR 2.390(b)(4). All documents within the scope of this affidavit are marked as "Confidential Information -Withhoi'd Under 10 CFR 2.390."
New regulatory commitments (as defined by NEI 99-04) are provided in.
If you have any questions regarding this response, please contact.
Mr. Philippe Soenen at (805) 459-3701.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on July 1, 2019..
* Sincerely,
'~~
James M. Welsch Senior Vice President Generation and Chief Nuclear Officer j contains Confidential information -Withhold Under10 CFR 2.390 When separated from Enclosure 3, this document is decontrolled to this letter contains*Confidential Information -Wfthhold Under 10 CFR 2.390 Document Control Desk July 1, 2019 Page 3*
Enclosures cc:
Hum.boldt Distribution cc/enc:
William C. Allen, NMSS Project Manager Christopher T. Markley, NMSS Project Manager Scott A. Morris, Region IV Admiriistrc!tor
* PG&E Letter HIL-19:-008 Gonzalo*L. Perez, California Department of Public Health (without Enclosures 3 and 4) contains Confidential information - Withhold Under 10 CFR 2.390 When sep~rated frorn Enclosure. 3, this document Is decontrolled
 
Holtec Confidential Information Affidavit PG&E Letter HIL-19-008 I
 
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 3015002-AFF Non-Proprietary Attachment 1 AFFIDAVIT PURSUANT TO 10 CFR).390 I, Kimberly Manzione, being duly sworn, depose and state as follows:
(1)
I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withpolding.
(2)
The information sought to be withheld is information provided in reference documents listed below as noted in the responses to the NRC's Request for Additional Information for the Technical Review of the Humboldt Bay Independent Spent Fuel Storage Installation License No. SNM-2514. These following references contain Holtec Proprietary information:
Holtite A: Development History and Thermal Performance Data, Holtec Report HI-2002396, Rev. 5.
American Seal Certificate of Conformance# 34048-CE for, HI-STAR Overpack Closure Plate Seals.
(3)
In making this application for withholding of proprietary information of which it is the owner, *Holtec International relies upon the exemption from disclosure set forth in the Freedom oflnformation Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4)~ and 2.390(b)(l) for "trade secrets and commercial or
* financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought.
is all "confidential commercial information", and some portions als~ qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes ofFdIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992),
and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir.
1983).
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U.S. Nuclear Regulatory Commission
. ATTN: Document Control Desk Document ID 3015002-AFF Non-Proprietary Attachment 1.
AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4)
Some examples of categories of information which fit into the definition of
. proprietary information are:
: a.
Information that discloses a process, method~ or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors
* without license from Holtec International constitutes a competitive economic advantage over other companies;
: b.
Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive _-position in the design, manufacture, shipment, installation, assurance.of quality, or licensing of a similar product. 1
: c.
Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies ofHoltec International, its customers, or its suppl_iers;
: d.
Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holteq Intem3:tional;
: e.
Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.
The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a and 4.b above.
(5)
The information sought to be withheld is being; submitted to the NRC in confidence. The information (including that compile~ from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so
. held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have 2 of 5
 
.\\.
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 3015002-AFF Non-Proprietmy Attachment l AFFIDAVIT PURSUANT TO 10 CFR,2.390
* been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the.subsequent steps taken to prevent its unauthorized disclosure, are as set fort"f:i in paragraphs ( 6) and (7) following:
( 6)
* Initial approval of proprietary treatment of a document is made by th~ manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.
Access to such documents within Holtec International is limited on a "need to lmow" basis.
(7). The procedure for approval of external release of '.such a docum~nt typically requires review by the staff manager, project manager,_ principal scientist or other equivalent authority, by the manager of the cognizant marketing function
( or his de~ignee ), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.
Disclosures. outside Holtec International are
* limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information,
* and then only in accordance with appropriate regulatory provisions =or proprietary agreements.
I (8)
The information class.ified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information,is classified as proprietary because it contains detailed descriptions of analytical
* approaches and methodologies not available elsewhere. Thi.~ information would provide\\yther parties, including competitors, with. information from Holtec Intemational's technical database and the results of.evaluations performed by Holtec **International.
* A substantial effort has been
* expended by Holtec International to develop this information. Release, of this information would
* improve a competitor's position because it would enable Holtec' s competitor to copy our technology and offer it for sale in competition* with our company, causing us financial injury..
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. \\
 
U.S. Nuclear *Regulatory Commission ATTN:\\Document Control Desk Document ID 3015002-AFF
* Non-Proprietary Attachment 1 AFFIDAVIT PURSUANT TO 10 CFR*~.390 (9)
Public disclosure of the information sought to be *.withheld is likely to cause substantial harm to Holtec Intemational's competitiy:e position and foreclose or reduce the availability of profit-making opportunitiep._ The information is part of Holtec Intemational's comprehensive spent fuel sto~e technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physica~ database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process.
1 The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holte*clntemational.
The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.
Holtec Intemational's competitive advantage will be.lost if its competitors are able to use the results of the Holtec International experience to normalize or
* verify their own process or if they are able to claim an 1 equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.
The value of this information to Holtec International would be lost if the information were disclosed to the public. Making suqh information available to competitors without there having been required to undertake a similar expenditure* of resources would unfairly provide comp.etitors with a windfall and deprive Holtec International of the opport:uriity to** exercise its competitive advantage to seek an adequate return on its large investrnentin developing these very valuable analytical tools.
4of5
 
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Document ID 3015002-AFF Non-Proprietary Attachment 1.
AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY
)
)
ss:
COUNTY OF BURLINGTON )
Kimberly Manzione, being duly sworn, deposes and says:.
That she has read the foregoing affidavit and the matters stated therein are true and correct to the.best of her knowledge, information, and belief.
Executed at Camden, New Jersey, this 22 day o:fMay 2019.
Kimberly Manzione Holtec International Erika Grandrimo NOTARY PUB UC STATE OF NEW JERSEY MY COMMISSION EXPIRES January 17, 2022 5 of5 PG&E Letter HIL-19-008 1 PG&E Response to NRC Letter dated April 30, 2019, "Request for Additional Information for the Technical Review of the Application for Renewal of the Humboldt Bay Independent Spent Fuel Storage lnstaUation License No. SNM-2514 (CAC No. 001028)"
PG&E Letter HIL-19-008 Page 1 of 19 PG&E Response to NRC Letter dated April 30, 2019, "Request for Additional Information for the Technical Review o*f the Application for Renewal of the.
Humboldt Bay Independent Spent Fuel Storage Installation License No. SNM-2514 (CAC No. 001028)"
RA/ 1-1 Provide the current estimated operating and.maintenance costs for the Humboldt Bay (HB) Independent Spent Fuel Storage -Installation (ISFSI}, as well as sources of funds to cover those costs, over the planned life of the /SFSI during the proposed license renewal period (years 2025 to 2065r
* Additionally, provide the rationale for these cost projections.
By {{letter dated|date=July 10, 2018|text=letter dated July 10, 2018}} (Agencywide Documents Access and Management System Accession No. ML18215A202), Pacific Gas & Electric Company (Pµ&E).
requested renewal of the Humboldt Bay /SFSI, (SNM 2514, Docket No. 72-27), foran additional 40-year period beyond the end of the current license term. The original 20-year /SFSI license expires on November 17, 2025.
In its submittal, PG&E stated, in part, that the Humboldt Bay ISFSI will remain financially qualified to carry out the operation and decommissioning of the /SFSI during the period of the renewed material license as required by 10 CFR 72.22(e).
The regulation at 10 CFR 72.22(e) "Contents of application: General and Financial Information," states:
Except for DOE, information sufficient to demonstrate4o the Commission the financial qualifications of the applicant to carry out, in accordance with the regulations in this chapter, the activities for which the license is sought. The information must state the place at which the activity is to be performed, the general plan for carrying out the
* activity, and the periocj of time for which the license is requested. The information must show that the applicant either possesses the necessary funds, or that the applicant has reasonable assurance of obtaining the necessary; funds or that by a combination of the two, the applicant will have the necessary funds available to* cover the following:
: 1. Estimated construction costs;
: 2. Estimated operating costs over the planned life of the /SFSI; and
: 3. Estimated decommissioning costs, and the necessary financial arrangements to provide reasonable assurance before licensing, that decommissioning will be carried out affer the removal of spent fuel, high./eve/ radioactive waste,
* and/or reactor related greater than class C (GTCC) waste from storage.
'. *..:~*
PG&E Letter HIL-19-008 Page 2 of 19 After reviewing PG&E's submittal, it appears that the estimated operating and maintenance costs, as well as sources of funds to operate the Humboldt Bay /SFSI
* were not specifically provided in the application for license renewal, nor could this information be easily obtained from staff's review of the PG&E annual report.
This information is needed to confirm compliance with 10 CFR 72.22(e).
I PG&E Response to RAI 1-1 The estimated operating and maintenance costs, as well as sources of funds to operate the HB ISFSI are submitted to the Nuclear Regulatory Commission (NRC) in the Humboldt Bay Power Plant Unit 3 Decommissioning Funding Report on an annual basis in accordance with 10 CFR 50.82(a)(8)(vii). PG&E Letter HBL-19-007, dated March 28, 2019 (ML19087A094), provides the most recent Decommissioning Funding Report, including the sources of funds to operate the HB ISFSI (see cover letter) and estimated
~
operating and maintenance costs (see Enclosure 5). As described in Enclosure 5, the estimated operating and maintenance (O&M) costs for the HB ISFSI from 2019 through 2033 is approximately $8.9 million ($2018). In addition to O&M costs, Enclosure 5 provides other costs such as security, staffing and infrastructure expenses associated with spent fuel management Under approved California Public Utilities Commission ratemaking, these costs will be funded by customer rates.
The Decommissioning Funding Report assumes the Department of Energy (DOE) will complete transfer of spent fuel and GTCC waste from the HB ISFs*1 in 2032. PG&E will continue to request additional funding through the California Public Utilities Commission Nuclear Decommissioning Cost Triennial Proceeding process if the DOE transfer date is delayed. provides revised License Renewal Application (LRA), Section 1.3.6. In addition, Enclosure 2 provides revised LRA Section 1.4 and Appendix G to reference the most recent decommissioning funding plan for the HB ISFSI.
RAJ 2-1 Clarify the Scoping Evaluation with regard to the following items. and their safety functions, modifyingthat evaluation and the aging management review as necessary. *
: 1. The soil around the vault. The renewal application should address the. soil around the vault, considering it in the scoping evaluation and aging management review or justifying why that is not necessary, since the soil is in the shielding analysis models (see Final Safety Analysis Report (FSAR) Figure.7.3-4) or
* influences how the analysis was done, such as locations where dose rates are calculated (e.g.,.seeFSAR Sections 7.3.1, 7.3.2, andl.3.2.2). The soil being in
. the shielding models or influencing how the shielding analysis was done means the soil has a safety function.
PG&E Letter HIL-19-008.
* Page 3 of 19
: 2. The reference drawing for the damaged fuel container (DFC). The FSAR contains Figure 4.2-3, which describes the DFC and should be referenced in the renewal scoping evaluation.
: 3. Inclusion otboth a shielding and a criticality function in the safety functions for
. the DFC subcomponents thf1t confine fuel.assembly material to the known volume of the DFC. Both the criticality analysis and the shielding analysis rely on the DFC to confine fuel material to a specified vglume (the DFC's cavity).
Otherwise, these analyses would need to consider the effects of fuel material from damaged fuel relocating to other areas within the Multi-Purpose Canister (MPC)-HB. DFC subcomponents having lhis function include the container wall (or tube) and top and bottom subcomponents, including the mesh, that enclose the DFC cavity. Thus, the safety functions of these subcomponents should include criticality and shielding.
: 4. Inclusion of a criticality and a shielding function in the safety functions for the MPC-HB fuel spacers and upper fuel spacers. These spacers keep the fuel
* assemblies axially positioned so that the active fuel region remains within the axial zone covered by the neutron absorber panels. This positioning is credited in the criticality analysis even though the spacers themselves are not included in the models. These spacers also have a shielding function in terms of maintaining the spent fuel assemblies' position in the MPC-HB relative to other components that are credited for shielding the radiation source from the spent fuel assemblies (such as the MPC-HB lid and the basket). Thus, the safety functions of these spacers should include criticality and shielding.
: 5. Inclusion of a shielding function in the safety functions for the fuel basket cell spacer plates. From the drawings, at least some of these spacer plates form basket cell walls, which are credited in the shielding analysis. Thus, the safety functions of these plates should include shielding.
: 6. Inclusion of both a shielding and a criticality function in the safety functions for the sheathing in the MPC-HB. The absorber sheathing is included in the analyses for both shielding and criticality. Thus, the safety functions for the sheathing should include both shielding and criticality.
. 7. Inclusion of a shielding function for the trunnions. The trunnions are insertedinto.
the top flange of the overpack, both for the HB overpack and the GTCC waste overpack. While the portion of the trunnions that extends beyond the outer surface of the. top flange is not credited in the shielding analysis, the analysis credits material in the area where. the trunnions are within the top flange. Thus,..
the safety functions of the trunnions should include shielding.
* Enclosure 1 PG&E Letter HIL-19-008 Page 4 of.19
: 8. Inclusion of a shielding function in the safety functions for the HB overpack's neutron cover plate. This cover plate is included in the shielding analysis model;.
. thus, its safety functions should include shielding. This also applies to steel subcqmponents above and below the neutron shielding material.
\\
: 9. Inclusion of a shielding function for porlplugs, base plugs, and similar
* subcomponents of the overpacks. These items are relied on to prevent radiation streaming from the openings in the overpacks and minimize occupational exposures from these streaming paths. Thus, these subcomponents should
* have a shielding safety function.
. 10. Inclusion of a criticality function in the safety functions for the steel shells, lid, and base of the HB overpack arid the lid and base*ofthe MPC-HB. The HB overpack's steel shells are included in the criticality model (as are the overpack's and MPC's lids and bases) and help to absorb thermal neutrons in the model.
The overpack's neutron shielding is given a criticality safety function. Thus, these steel shells should also have a criticality safety functio11.
: 11. lnc/usion of a shielding function in the safety functions for the process waste container (PWC). The shielding model includes the materials of the PWC. Thus, the relevant subcomponents should be credited with a shielding function.
: 12. Listing a shielding safety function for the outer container in the GTCC waste container (GWC). This component is included in the shielding analysis model.
Thus, the outer container, including its lid, should scope in and have a shielding.
safety function.
: 13. Confirmation that there is no lid for the GWC's inner shell. The referenced GWC drawings do not include an inner shell lid; however, the shielding analysis for the GTCC waste is based on that waste remaining within the GWC's inner shell.
Thus, a lid may be needed for that inner shell to ensure the waste remains within it, which also means that this lid would. scope in and have a shielding safety function
* 14. Inclusion of a shielding function in the safety functions for the vault shell and the vault lid top plate, base plate, and outer she/I;
* These subcomponents are included in the shielding analysis model. Thus, these subcomponents should
* have a shielding safety function. This may also apply to the vault shell lid ring.
This information is needep to confirm compliance with 10 CFR 72.42(b), 72.24(d) and (e), 72.104, 72.106, 72.124, and 72.126.
* PG&E Response to RAI 2-1
* The Scoping Evaluations are clarified as follows:
PG&E Letter HIL-19:.008 Page 5 of 19
: 1. As discussed in thE3 HB ISFSI FSAR, Section 7.3.2.2, soil was ir:icluded in the analytical model in the HB ISFSI dose assessment calculation. However, considering the thickness of the shielding materials surrounding the spent fuel and GTCC waste (including steel and Holtite in the HI-STAR 100's body and vault's concrete), the soil does not provide any significant additional shielding.
An evaluation was completed to determine if there is any need to validate the soil shielding during the period of extended operation.
The evaluation reperformed two new scenarios of the HB ISFSI dose assessment using the sanie analytical model, assumptions; and methodologies, except (1) replacing all soil' with air (i.e., vault is.sitting on the ground), and (2) replacing portions of soil with air (i.e.,.pockets of soil are gone, but the vault is below grade) - see Table 1 below.
The site boundary is located slightly greater than 50 feet from the ISFSI: Doses are calculated at approximately 50 feet from the infinite line of casks from the
* ground elevation to approximately 15 feet high in segments of approximat~ly four~
feet. The exact elevations of the dose locations (in centimeters [cm]) and the results are provided Table 1 below. *
* Assuming 2,080 hours per year of occupancy, these values meet the 25 millirem (mrem) per year requirement of 10 CFR 72.104. Thus, there are no soil validations (aging.management) that are necessary during the period of extended operation.
Table 1: HB ISFSI Doses Elevation Scenario 1 Maximum Dose Scenario 2 Maximum Dose (cm)
(mrem/year)*
(mrem/year)*
0 to 120 9.5 10.4 120to240 10.2 10.7 240 to 360 10.7 11.0 36.0 to 480 11.2 11.7
* *Based on 2,080 hours/year occupancy. provides revised LRA Table 2-9 and Appendix D, Section 0~2 which includes the soil surrounding the ISFSI "vault and the basis for not listing a shielding function for the period of extended operation.*
: 2. HB ISFSI LRA Tc1ble 2-3 has been updated to include FSAR Update Figure 4.2-3 as a referenced drawing for the DFC. Enclosure 2 provides the revised LRA table.
PG&E Letter HIL-19-008.
page 6 of 19
: 3. HB *1sFSI LRA Tables 2-3 and 3.3-1_ have been updated to include criticality and shieldi_ng safety functions for those subcomponents that confine fuel assembly material to the known volume of the DFC. DFC subcomponents having these functions include the tube; pan base, side, and top; mesh, mesh ring, and mesh plate; lock bolt and plate; baseplate; and base feet. Enclosure 2 provides the revised LRA tables.
: 4. HB ISFSI LRA Tables 2-4 and 3.4-1 have been updated to include criticality and shielding safety functions for the MPC-HB fuel spacers and upper fuel spacers. provides the revised LRA tables.
: 5. HB ISFSI LRA Tables 2-4 and 3.4-1 have b.een updated to include a shielding safety function for the fuel basket cell spacer plates. Enclosure 2 provides the revised LRA tables.
: 6. HB ISFSI LRA Tables 2-4 and 3.4-1 have been updated to include criticality and shielding safety functions for the sheathing in the MPC-HB. Enclosure 2 provides the revised.LRA tables.
: 7.
* HB ISFSI LRA Tables 2-5 and 3.5-1 have been updated to include shielding and structural integrity safety functions for the HI-STAR HB Overpack lifting trunnions. HB ISFSI LRA Tables 2-8 and 3.8-1 have been updated to include shielding and structural integrity safety functions for the HI-STAR HB GTCC.
Overpack lifting trunnions. _ Enclosure 2 provides the revised LRA tables, I
: 8. HB ISFSI LRA Tables 2-5 and 3.5-1 have been updated to include a shielding safety function for the HI-STAR HB Overpack neutron cover plate, top ring plate, and toe ring plate. Enclosure 2 provides the revised LRA tables.
: 9. As discussed with the NRC, PG&E will respond to this RAI in a separate letter prior to July 31, 2019. See PG&E commitment in Enclosure 5.
v,
: 10. HB ISFSI LRA Tables 2-4 and 3.4-1 have been updated to include criticality and shielding safety functions for the MPC-HB base plate, lid, closure ring, and port cover plate. HB ISFSI LRATables 2-5 and 3.5-1 have been updated to include a criticality safety function for the HI-STAR HB Overpack bottom plate,- inner shell, intermediate.shells, and closure plate. Enclosure 2 provides the revised LRA tables.
1
: 11. HB ISFSI LRA Tables 2-6 and 3.6-1 have been updated to include a shielding -
safety function for the PWC container shel_l and top plate. Enclosure 2 provides the revised LRA tables.
PG&E Letter HIL-19~008 Page 7 of 19
: 12. The following subcomponents make up the "GWC outer container" as listed in HB ISFSI LRA Tables 2-7 and 3.7-1:
su.bcomponerit Part.*
* Reference..
Number Drawing MPC Base Plate 1
6023993-26 MPC Shell 2, 3 6023.993-26 and FSAR Update Figure 3.3-4 MPC Lid Top 18 6023993-26 MPC Lid Bottom 19 6023993-26 Closure Ring 21 6023993-26 Drain Shield Block 22 6023993-26 Port Cover Plate 25 6023993-26 HB ISFSI LRA Tables'2-7 and 3.7-1 correctly list the shielding function for the MPC base plate, MPC shell, MPC lid top and bottom, and drain shield block. HB ISFSI LRA Tables 2-7 and 3.7-1 have been updated to include a shielding safety.
function for the closure ring and port cover plate. Enclosure 2 provides the revised LRA tables.
: 13. There is no lid for the GWC's inner shell. There is a space between the top of the inner shell and the bottom of the GWC lid bottom which stores GTCC waste components. The sizes and shapes of the GTCC waste pieces preclude movement out of the inner shell (see drawings referenced in revised LRA Table 2-12). However, because the GTCC waste pieces protrude above the top of the inner shell inconsistent with the GTCC waste shielding analysis, the shielding analysis is being updated to reflect the actual configuration. This finding has been entered into the PG&E and Holtec Corrective Action Programs for evaluation. As discussed with the NRC, PG&E will notify the NRC of the completed shielding analysis update and evaluation findings in a separate letter by July 31, 2019. See PG&E commitment in Enclosure. 5.
To ensure that there is no potential for aging of the GTCC waste during the period of extended operation that could lead to portions coming out of the inner shell, the HB ISFSI LRA was revised. Enclosure 2 provides new or revised LRA.
Sections 2.3, 3.12, and D.3, and LRA Tables 2-1, 2-12, and 3.12-1 to (1) scope in the GTCC waste for Criterion 2, and (2) perform an aging management evaluation. The aging management evaluati6n demonstrates there are no potential aging effects/mechanisms that would lead to GTCC waste degradation such that portions would come out of the inner shell.
J
: 14. HB ISFSI. LRA Tables 2-9 and 3.9-1 have been updated to indude a shielding safety function for the vault shell base plate, vault shell, vault shell lid ring, vault lid base and top plates, and vault lid outer shell. Enclosure 2 provides the revised LRA tables.
 
RAJ 3-1 PG&E Letter HIL-19-008 Page 8 of 19 Clarify the environments for the components below and revise the aging management review tables, as appropriate.
: 1. LRA Table 3.5-1, "Aging Management Review of HI-STAR HB Overpack,"
* contains line items for the neutron cover plate exposed to a sheltered environment and the neutron rib exposed to an embedded environment.. The staff notes that these two components may be expected to be embedded in Holtite-A.
I
: 2. LRA Table 3.8-1, ''Aging Management Review of HJ-STAR GTCC Overpack,"
contains a line item for the nickel alloy lifting trunnion exposed to enclosed air (internal) and sheltered (external) environments. It is unclear to the staff how the trunnion is exposed to the enclosed air environment, rather than being embedded in steel.
: 3. LRA Table 3. 8-1, ''Aging Management Review of HI-STAR G TCC Overpack,"
contains a line item for the intermediate shells exposed to enclosed air (internal) and sheltered (externalj environments. It is unclear to the staff how an intermediate shell is exposed internally to the enclosed air environment, rather than being embedded in steel.
: 4. LRA Table 3.8-1, ''Aging Management Review of HI-STAR GTCC Overpack,"
contains a line item for the "shell" exposed to enclosed air (internal) and sheltered (external) environments. It is unclear to the staff if this shell is referring to the inner shell, and if so, how this shell is exposed to-*a sheltered (external) environment and managed by the HB ISFSI External Surfaces Monitoring AMP.
The staff requires clarification of the exposure environments to ensure that the aging effects are appropriately evaluated.
This information is required to demonstrate compliance with 10 CFR 72.42(a).
PG&E Response to RAI 3-1 The exposure environments are clarified as follows:
: 1. The neutron cover plate is the outer-most shell on the HI-STAR HB Overpack.
* The inner diameter surfac~ is embedded in Holtite-A, while the outer diameter surface is~exposed te> the external sheltered environm~nt. HB ISFSI LRA Table 3.5-1 has been updated to designate the sheltered environment as "external" and add a new internal environment of ''Embedded (Holtite-A)". The neutron rib is fully embedded in Holtite-A. HB ISFSI LRA Table 3.5-1 has been
)
PG&E Letter HIL-19-008 Page 9 of 19 updated to clarify that the embedded medium is Holtite-A. Enclosure 2 provides the revised LRA table.
: 2. HB ISFSI LRA Tables 3.5-1 and 3.8-1 andAppendix D, Table 9.4-1 have been updated to show the lifting trunnion internal environment as "Embedded (Metal)"
and revised associated aging effects and mechanisms. Enclosure 2 provides the*
revised LRA tables.
: 3. HB ISFSI LRA Table 3.8-1 has been updated to show the intermediate shell internal environment as "Embedded (Metal)" and revised associated aging effects and mechanisms. Enclosure 2 provides the revised LRA table.
: 4. The "shell" listed in HB ISFSI LRA Table 3.8-1 refers to the inner-most shell on the HI-STAR HB GTCC Overpack. The inner diameter surface is exposed to an enclosed air environment, while the outer diameter surface is embedded in the intermediate shells. HB ISFSI LRA Table 3.8-1 and Appendix D, Table 9.4-1 have been updated to show the shell external environment as "Embedded (Metal)" and.revised associated aging effects and mechanisms. Enclosure 2 provides the revised LRA tables.
RA/ 3-2 Provide justification for not identifying cracking due to stress corrosion cracking as a.
credible aging mechanism and effect for welded stainless steel components exposed to a sheltered environment for the external surfaces of the H /-STAR HB Overpack.
LRA Table 3.5-1, ''Aging Management Review of HI-STAR HB Overpack," contains line items for stainless steel port plugs, closure plate overlay, and flange overlay exposed to a sheltered environment. Pitting and crevice corrosion are identified as credible aging, mechanisms.
The Staff notes that one of (he reports used in the LRA to evaluate aging mechanisms.
(Draft NUREG-2214, "Managing Aging Processes in Storage (MAPS) Report) identifies cracking due. to stress corrosion cracking as a credible aging mechanism for welded stainless steel components in a sheltered environment. Stress corrosion cracking is identified in that report as being credible due to the potential exposure* to moisture and chloride-containing contaminants.
To ensure that potential degradation of the HI-STAR HB overpack is appropriately managed, the Staff requires the technical basis for excluding cracking due to stress corrosion cracking as an aging effect.
This information is required to demonstrate compliance with 10 CFR 72.42(a).
 
PG&E Response to RAI 3-2 PG&E Letter HIL-19-008 Page 10 of 19 Stainless steel port plugs and rupture disk components in HB ISFSI LRA Table 3.5-1 exposed to a sheltered environment are consistent with the Draft MAPS.Report, Table 3-2, for not considering stress corrosion cracking as an applicable aging mechanism because these components have no welds or heat-affected zones.
. Therefore, sufficient stress does not exist in the components to support stress corrosion cracking. HB ISFSI LRA Table 3.5-1 was updated to irclude this basis as a table note
* for these components.
The closure plate. overlay and flange overlay are exposed to the external overpack sheltered environment. HB ISFSI LRA Table 3.5-1, Section 3.5.4 1 Table A-1., and Appendix D, Table 9.4-1 were updated to reflect the potential for stress corrosion cracking and its aging management. provides the revised LRA tables.
RAJ 3-3 Provide details of self-energizing seals in Section 3. 5. 1: The staff requests the potential changes of mechanical properties (e.g., yield stress, or creep if applied) with time of the self-energizing seals, Alloy X750. The seal manufacture's data or open literature data could be provided.
The HI-STAR 100 HB overpack is a heavy-walled steel cylindrical vessel that provides the helium retention boundary during storage operations. The helium retention boundary is comprised of the overpack inner shell welded to a cylindrical forging at its bottom and a heavy flange with a bolted closure plate at its top. The closure plate is equipped with two concentric grooves for self-energizing seals. The staff requests the function, properties and materials of the self-enf]rgizing seals.
This information is needed for evaluating Hf-STAR Humboldt Bay (HB) /SFSI Renewal, in compliance with 10 CFR 72.-122(b),(c), 10 CFR 72.42(a)(1).
PG&E Response to RAI 3-3
* The requested manufacturer's data for the HI-STAR 10.0 HB overpack closure plate self-energizing seals is provided in Enclosure 3.
The self-energized seals are allmade,of lnconel Alloy X750 material and governed by SAE AMS 5598 and SAE AMS 5699. As described in HB ISFSI LRA Section 3.5.4.1 and the HI-STAR 100 FSAR Update Section 9.2.2Ahe Closure seals ensure a pressure boundary is maintained and that the MPG is in an inert environment. There we no PG&E Letter HIL-19-008 Page 11 of 19 credible normal, off-normal, or accident events.which can cause the failure of the overpack helium retention boundary seals.
I Nickel alloys, such as X750, have good oxidation resistance up to 1300°F and therefore are not susceptible to corrosion under the environmental conditions at Humboldt Bay.
Likewise, creep is not credible due to the comparable creep-rupture values presented in ASME Section ll for SB-637 N07718 at temperatures up to 1300°F. The normal '.
condition temperature of the HI-STAR.HS overpack lid is calculated at 225°F and has been measured at a max.imum temperature of 113.3°F. Per the Draft MAPS Report, creep is not credible in steel components of Dry Storage Systems and aging
* management is not required during the 60-year timeframe. Considering the seals are fabricated*from Alloy X750, which israted to handle higher temperatures than standard stainless and carbon steels used to fabricate the overpacks, the.conclusion that creep is not credible for the overpack seals is valid.
The most important characteristic of the seals' ability to perform is the useful spring back. The spring back is related to the material's modulus of elasticity. The
. temperature effect on the modulus of elasticity due to the *operating temperatures is*
negligible.
Table* 4-9 and Section 3.2.4 of the Draft MAPS Report concludes that no aging management is necessary on the HI-STAR 100 closure seals.
RA/.3-4 Clarify the following items, modifying the renewal application and analyses as
* necessary.
: 1. The fraction of boron-10 in the Holtite-A shielding material that is estimated to be depleted over the 60 years of storage (20-year initial license period plus the 40-year period of extended operations). The renewal application indicates this*
fraction will be Jess than 5x10-10; however, the original evaluation on which this is based, the 10 CFR Part 71 safety analysis for the HI-STAR 100 transportation
*
* package, indicates the fraction for 50 years is 4. Ox10-8. Thus, it is not clear how the 5x10-1° fraction was derived.
: 2. The location in the Humboldt Bay JSFSI FSAR that establishes the design basis limits for surface dose rates.. The fourth paragraph of Element 5 of the HB ISFSJ
* Reinforced Concrete Structures AMP (Table A-2 of the renewal application) indicates that these limits are in Chapter 5 of the /SFSI FSAR. However, the.
staff did not find where the dose rate limits were established in that chapter of the FSAR.
.PG&E Letter HIL-19-008 Page 12 of 19 This information is needed to confirm compliance with 10 CFR 72.42(a).
PG&E Response to RAI 3-4
: 1. As discussed with the NRC, PG&E will respond to this RAI in a separate letter prior to July 31, 2019. See PG&E commitment in Enclosure 5.
: 2. Design basis limits for surface dose rates at the HB ISFSI are identified in HB ISFSI FSAR Chapter 7, Radiation Protection. Element 5 of the HB ISFSI Reinforced Concrete Structures AMP (HB ISFSI LRA Table A-2) has been updated to reflect the correct FSAR Chapter reference. Enclosure 2 provides the revised Element 5.
RAJ A-1 State how the visual inspection parameters will be controlled to ensure* that there is sufficient resolution and lighting for the inspections of the Cask Transporlation System AMP.
I LRA Appendix A-3, "Cask Transporlation System AMP," states that visual inspections of the transporler structure, cask restraint system, and wedge Jock assembly are performed with sufficient resolution and lighting to identify the degradation.
It is unclear to the staff how the Humboldt Bay processes and procedures are controlled to ensure that inspectors will use sufficient resolution and lighting to identify the parameters monitored in the Cask Transportation System AMP (e.g., discontinuities indicative of pitting, crevice, general, and galvanic corrosion). Describe either site operation practices or AMP-specific requirements that will be used to establish resolution and lighting requirements for the transportation system insp$ctions.
This information is required to demonstrate compliance with 10 CFR 72. 42( a).
PG&E Response to RAI A-1 The HB ISFSI LRA has been updated to reflect that VT-3 inspections will be performed for the transporter structure, cask restraint system, and wedge lock assembly.
VT-3 requirements specify resolution and lighting requirements to identify the degradation. Enclosure 2 provides the revised Table A-3, Table A-4, and Appendix D,
.Section 9.4.3.3.3.
RAJ A-2 In FSAR Section 9.4.3.3.3, "Cask Transporlation System AMP," clarify the acceptance criteria for the tactile inspections of polymers that are subject to hardening.
' \\
PG&E Letter HIL-19-008 Page 13 of 19 LRA Appendix A-3, "Cask Transportation System AMP," and FSAR Section 9.4.3.3.3 state that tactile inspections are used to evaluate hardening of polymers. However, the acceptance criteria for polymers appear to be relevant only to visual inspections (e.g.,
erosion, cracking, crazing, checking, and chalks)..
Describe the tactile inspection acceptance criteria that are capable of evaluating
* polymer hardening.
. This information is required to*demonstrate compliance with 10 CFR 72.42(a).
PG&E Respons~ to RAI A-2 The HB I Sf SI LRA has been updated to reflect that polymer Cask Transporter Adjustable Cask Bumper replacement will occur prior to the start of each cask transfer campaign if the Cask Transporter has been in service greater than 20 years. These bumper replacements preclude the need for specific bumper inspections and associated acceptance criteria. Enclosure 2 provides the revised Table A-3, Table A-4, and Appendix D, SectiC?n 9.4.3.3.3.
RA/ A-3 State the frequency of the Cask Transportation System AMP inspections following the initial inspections that are to occur prior to first use.
LRA Appendix A-3, "Cask Transportation System AMP," and FSAR Section 9.4.3.3.3 state that the AMP inspections occur prior to first use of the system after components reach 20 years of service. However, there is no description of subsequent inspections.
It is unclear to the staff whether the initial inspection described in the AMP is the only inspection that will be performed in the 40-year period of extended operation. If so, the staff requires technical justification that the initial inspection is sufficient to ensure that
* the transportation system will perform its safety functions for the entire license term. If subsequent inspections are intended to be performed, the AMP should describe the required frequency (for example, Draft NUREG-2214 recommends a 5-year inspection interval fortransport cask inspections while transport casks are in use).
This information is required to demonstrate compliance with 10 CFR 72.42(a).
. PG&E Response to RAI A-3 HB ISFSI LRA Appendix A, Table A-3 and Table A-4, and Appendix D,Section 9.4.3.3.3 have been updated to clarify that the Cask Transporter visual inspections are performed prior to use (e.g., prior to the first use for on-site ca.sk handling operations or prior to the first offsite transport) if the Cask Transporter has been in-service for greater than 20 PG&E Letter HIL-19-008 Page 14 of 19
* years and every 5 years thereafter. Enclosure.2 provides the revised Table A-3, Table A-4, and Appendix D, Section 9.4.3.3.3.
RAJ A-4 Clarify the conditions under which below-grade concrete will be inspected and provide justification if such inspections are not conducted at every opportunity.
. LRA Table A-2, "HB ISFSI Reinforced Concrete Structures AMP," includes conflicting information on when below-grade concrete will be inspected.
AMP Element 3, "Parameters Monitored or Inspected," states that "[i]nspections of exposed portions of the below grade c9ncrete are conducted when excavated for any reason."
Conversely, AMP Element 4, "Detection of Aging Effects," states that
'1e]xaminations of representative samples -of the exposed portions of the below grade concrete are conducted when excavated for any reason if conditions exist in accessible areas that could indicate the presence of or result in degradation to inaccessible below-grade concrete structural elements." [emphasis added]
The Staff notes that AC/ 349.3R-18, "Report on. Evaluation and Repair of Existing Nuclear Safety-Related Concrete Structures," Chapter 6, "Evaluation Frequency,"
recommends, that, for structures with non-aggressive exposures, representative samples *of below-grade concrete be examined when excavated for any reason.
Sectio!J 3.4 of AC/ 349.3R states that the combination of soillgroundwater_chemistry monitoring and opportunistic inspections of below-grade concrete can verify that periodic inspections of accessible above-grade structures can serve as a leading indicator of degradation.
Without opportunistic inspections of below-grade concrete, it is unclear to the staff that
* the periodic AMP in,spections of accessible concrete will evaluate worst-case* conditions.
This information is required to demonstrate compliance with 10 CFR 72.42(a).
- PG&E Response to RAI A-4 Consistentwith ACl349.3R-18 recommendations, HB ISFSI LRA Table A-2, E_lem~nt4 was updated to require examinations of representative samples of the exposed portions of the below grade concrete are conducted when excavated for any reason_ (i.e.,
opportunistic inspections). Enclosure 2 provides the revised Element 4.
 
RAJ A-5 PG&E Letter HIL-19-008 Page 15 of 19 Revise the HB ISFSI Reinforced Concrete Structures AMP (Table A-2 of the renewal application) to clearly identify and describe the management of the Holtite-A aging and its criticality safety function.
In accordance with Table 3.5-1 of the renewal application, cracking and radiation embiittlement aging effect and mechanism are to be managed as part of the HB JSFSI Reinforced Concrete Structures AMP (Table A-2 of the renewal application).
Table 3.5-1 also assigns the Holtite-A a criticality safety function. However, the AMP does not call out.this intended function. Also, the descriptions of the AM P's elements do not clearly include or address the Holtite-A
* material.
This information is needed to confirm compliance with 10 CFR 72.42(a) and 72.124.
PG&E Response to RAI A-5 HB ISFSI LRA Tables 2-5 and 3.5-1 incorrectly listed a criticality safety function for Holtite-A. The Holtite-A neutron shield is for shielding purposes only. HB ISFSI LRA Tables 2-5 and 3.5-1.have been updated to remove the criticality safety function for the HI-STAR HB Overpack neutron shield. This change makes HB ISFSI LRA Chapter 2 and 3 consistent with the existing HB ISFSI Reinforced Concrete Structures AMP, which does not include management of Holtite-A for criticality. Enclosur.e 2 provides the revised LRA tables.
RAJ A-6 Provide an evaluation of the public and occupational doses for operations for overpacks and the /SFSI vault that:
: 1. accounts for the combined effects of potential degradation of the carbon steel, Holtite-A neutron shielding, and the concrete sub-components*
: 2. demonstrates that the proposed aging management programs ensure the shielding function will be maintained when considering the combined degradation effects
: 3. demonstrates that the.doses will remain within the design basis limits described in Chapter 7 of the FSAR and the regulatory limits in 10 CFR Part 72 and 10 CFR Part 20 when considering the combined degradation effects, and
* 4. addresses all relevant operations configurations within the design basis..
The r~newalapplication includes discussion of ~ging effects andmechanis.rns for the carbon sffiel subcomponents of the overpacks and the JSFSI vault as. well as the vault PG&E Letter HIL-19-008
. Page 16 of 19 concrete and the Holtite-A neutron shielding for the overpacks containing spent fuel.
. These components *are included in the shielding analysis fordetermining overpack dose rates, demonstrating compliance with regulatory dose limits(e.g., 10 CFR 72.104(a) and 10 CFR 72.106(b)), and determining occupational dose estimates. In the renewal application, the licensee addresses each the effects of aging for each subcomponent separately and only for the configuration of the overpack in its /SFSI.vault cell. Since these subcomponents all contribute to the shielding function, the licensee should evaluate the combined effect of their degradation, as evaluated in the proposed analyses and allowed in the acceptance criteria of the proposed aging management
.* programs. Additionally, the license design basis includes operations with configurations in addition to the configuration of the overpacks being in their respective vault cells with the cell lid in place. At least some of these operations may be encountered during operation of the:!SFSI (e.g.; operations with the vault cell /id removed, operations with the overpack out ofthe vault cell for preparation for transport). Thus,/he licensee's.
evaluation should address configurations of the subcomponents for the relevant operations allowed by the license. The following discussion provides additional detail regarding items the requested evaluation should address.
* The proposed aging management of the concrete subcomponents uses the AC/ 349.3R evaluation criteria. These criteria are intended forensuring structural performance of the concrete, not ensuring the shielding function. So, the evaluation should address the degradation that use of these criteria would allow before the degradation would be entered into the licensee's corrective action program. The licensee has performed some analysis for loss of material for carbon steel subcomponents; however, the staff cannot determine that the analysis is adequate to account for the amount of corrosion of carbon steel subcomponents that is discussed in the renewal application (e.g., the estimated annual corrosion rates discussed in the application). The scope of this analysis is limited to the overpack being in its vault cell with no consideration for the impacts on shielding from degradation of the Ho/fife-A and the concrete subcomponents.
~
. The design bases in Chapter 7 of the FSAR include evaluations of the dose rates and
* doses and evaluation of compliance with regulatory limits that address the configurations and operations included in the design basis and described in the FSAR.
The evaluation in the renewal application shouid demonstrate that the actions and evaluation criteria in the proposed aging_management programs are.sufficient to ensure the shielding function is maintained for these configurations and operations, not just the configuration with the overpacks in their vault cells with the vault cell /ids in place. The evaluation should consider relevant transfer operations, periodic maintenance activities and activities required by technical specifications,* if any, and should consider that operations may be for multiple overpapks within a given year period. In instances where the evaluation may indicate that design bases or compliance with regulatory
.limits may be challenged (e-.g., 10 CFR Part 20 occupational dose limits), the evaluation should describe the actions that would be taken, 'controls that would be imposed, or conditions that would assure compliance is maintained. Guidance such as is provided PG&E Letter HIL-19-008 Page 17 of 19 in Section 11.4.3.1 of NUREG-1567, particularly the bulleted list at the end of the section, should be considered, as needed.
I This information is needed to determine compliance with 10 CFR 72.24(e),
72.122(h)(5), 72.104, 72.106, 72.124, 72.126, and 72.42(a).
PG&E Response to RAI A-6 As discussed with the NRG, PG&E will respond to this RAI in a separate letter prior to July 31, 2019. See PG&E commitment in Enclosure 5.
RAJ A-7 I
. Provide justification that the proposed aging management program for managing degradation of the Holtite-A shielding material is adequate to ensure the shielding function of this material is maintained for the period of extended operation for each spent fuel overpack.
The proposed aging management of the Holtite-A shielding material includes quarterly radiation surveys of the ISFSl's vault cells' lids and general area, quarterly evaluation of TLD dose data, and dose rate measurements on the vault cells' lids every 5 years along with dose rate measurements on the closure plates (the lid area) of the overpack in the vault cell which. is opened for more detailed inspections of the vault cell inter/or. While surveys of the /SF Si's general area or the dose data from the TLDs will provide an indication of the overall lSFSI dose rates and doses, the licensee should justify how these data will enable identification of degradation of an individual overpack's Holtite-A material that requires further action to ensure the shielding function is maintained. All overpacks will contribute to the measurements (survey and TLD) and, depending on the area of the Holtite-A that is degraded, the surveys and TLDs may be at locations that will not detect the effects of the degraded Holtite-A. Additionally, the measurements on the vault cell lid and the overpack closure plates are in areas where there is no Holtite-A or in locations where the Holtite-A does not have any expected impact on dose rates.
The Holtite-A is on the radial side of the overpacks, whereas all of the dose rate measurements are on the top of the overpack or directly above the overpack (on the vault cell lid).
Thus, the staff currently finds that the proposed measurements are not sufficient to detect degradation of Holtite-A on an individual overpack that would require corrective action. The justification should include discussion of sensitivity of the measurement techniques and how that is sufficient to identify an issue with the Holtite-A., The justification should also include discussion of the locations at which dose rates will be measured on the vault lids and overpack lid, including whether the measurements will be taken at multiple lid locations, and the basis for the measurement location selection, including the number of locations and the appropriateness and adequacy to detect*
Holtite-Adegradation on an individual overpack that requires corrective*action. The
 
I PG&E Letter HIL-19-008 Page 18 of 19.
discussion should also explain the adequacy of the proposed acceptance criteria to ensure the measurements are sufficient to detect Holtite-A degradation on an individual overpack that requires corrective action. The justification should demonstrate that the actions and evaluation crite,:ia
* are sufficient to ensure the Holtite-A shielding function is maintained for the configurations and operations which are part of the license design
. basis.
This information is needed to determine compliance with 10 CFR 72.24(e),
72.122(h)(5), 72.104, 72.106, 72.124, 72.126, and _72.42(a).
PG&E Response to RAI A-7
. As discussed with.the NRC, PG&E will respond to this_RAI in a separate letter prior to July 31, 2019. See PG&E commitment in Enclosure 5.
RAJ D-1 Clarify if Bora/ is used as a neutron poison in the MPC-HB.
In LRA Appendix D, "Final Safety Analysis Report Update Supplement and Chfmges,"
"Bora/" was deleted from the text in Sections 4.2.3.3. 7 and 4.6.4. However, the staff notes that other sections of the FSAR include the use of this neutron poison material,
_ but they were not revised in the update. For example, Bora/ is included in FSAR Table 4.6-1 and FSAR Sections 4.4.3.6, 4.6.1.2, and 4.6.3.
The staff requires clarification of the use of Bora/ to ensure that the neutron poison in the MPC-HB is appropriately evaluated for aging.
.This information is required to demonstrat~ compliance with 10 CFR 72.24(c) and 72.42(a).
PG&E* Response to RAI D-1 Baral is not used as a neutron poison in the MPC-HB, but was evaluated and approved for use in the original HB ISFSI design and licensing. The HB ISFSI LRA, Section D.2 HB ISFSI FSAR mark ups for Sections 4;2.3.3.7 and 4.6.~ have been updated to remove the deletions associated with Bor:al because it is still approved for use at the HB ISFSI. Enclosure 2 provides the revised LRA Appendix D.
* RAID-2 I
Provide the rationales for the use of previous versions of Interim Staff Guidance (ISG)
The applicant does not use updated versions of NRG JSG in Table 4.2-12 of -
HUMBOLDT BAY /SFSI FSAR UPDATE:
 
/SG 2. Fuel Retrievability
. Enclosure 1 PG&E Letter HIL-19-008 Page 19 of 19 As the functional definitions may have been modified/added in the newer version, any potential aging effects.may need to be assessed/addressed accordingly with the revision. For example, undamaged fuel defined in the later version is not included.
Clarify that only intact fuel and damaged fuel are considered. There is no aging issue ass.ociated with undefined fuel if any. The clarification-of /SG 1 would be applied to ISG 2 in terms of retrievability requirements.
This information is needed for evaluating HI-STAR Humbo1dt Bay (HB) ISFSI Renewal, in compliance with 10 CFR 72.42(a)(1), 10 CFR 72.122(b) (0.
PG&E Response to RAL D-2 The HB ISFSI LRA, Section 2.2 was updated to clarify that the scoping evaluation was
. completed using the retrievability definition as defined in ISG-2, Revision 2. Thus, the HB ISFSI LRA is consistent with the current NRC definition for retrievability. Enclosure 2 provides the revised LRA Section 2.2.
PG&E Letter HIL-19-008 License Renewal Application (LRA) Revision 2 Affected LRA Sections and Tables PG&E Letter HI L-19-008 Page 1 of 3 License Renewal Application (LRA) Revision 2 Affected LRA Sections and Tables
* LRA Section/Table Reason for Change Section 1.3.6 RAI 1-1 Section 1.4 RAI 1-1 Section 2.2 RAI D-2 Section 2.3 RAI 2-1, Part 13 Table 2-1 RAI 2-1, Part 13 Table 2-3 RAI 2-1, Parts 2, 3 Table 2-4 RAI 2-1, Parts 4, 5, 6, 10 Table 2-5 RAI 2-1, Parts 7, 8, 10 RAI A-5 Table 2-6 RAI 2-1, Part 11 Table 2-7 RAI 2-1, Part 12 Table 2-8 RAI 2-1, Part 7 Table 2-9 RAI 2-1, Parts 1, 14 Table 2-12 RAI 2-1, Part 13 Section 3.5.4 RAI 3-2 Section 3.12 RAI 2-1, Part 13 Table 3.3-1 RAI 2-1, Part 3 Table 3.4-1 RAI 2-1, Parts 4, 5, 6, 10 Table 3.5-1 RAI 2-1, Parts 7, 8, 10 RAI 3-1, Parts 1 and 2 RAI 3-2 RAI A-5 Table 3.6-1 RAI 2-1, Part 11 Table 3.7-1 RAI 2-1, Part 12 Table 3.8-1 RAI 2-1, Parts 7, 8 RAI 3-1, Parts 2, 3, 4 Table 3.9-1 RAI 2-1, Part 14 Table 3.12-1 RAI 2-1, Part 13 Appendix A, Table A-1 RAI 3-2 Appendix A, Table A-2 RAI 3-4, Part 2 I
RAI A-4 Appendix A, Table A-3 RAI A-1 RAI A-2 RAI A-3 Appendix A, Table A-4 RAI A-1 RAI A-2 RAI A-3
 
'\\
Appendix 0, Section 0.2 Appendix 0, Section 0.3 Appendix G, Section G.1 u
PG&E Letter HIL-19-008 Page 2 of 3 RAI 2-1, Part 1 RAI 0-1 RAI 2-1, Part 13 RAI 3-1, Parts 2 and 4 RAI 3-2 RAI A-1 RAI A-2 RAI A-3 RAI 1-1 PG&E Letter HIL-19-008 Page 3 of 3 Humboldt Bay Independent Spent Fuel Storage Installation Site Specific License Renewal Application, Revision 2,June 2019 w'
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* Regulatory Corrimitment PG&E Letter HIL-19~008
 
\\.
Regulatory Commitment PG&E Lefter HIL-19-008 Pacific Gas and Electric Company (PG&E) is making the following new regulatory
* commltments (as defined by NEI 99-04) in this submittal:
Commitment Due Date PG&E will provide responses to. request for additional information (RAI) 2-1 Pc3rt 9, RAI 3-4 Part 1, RAI A-6, and RAI July 31, 2019 A-.7 by July 31, 2019.
. RAI 2-1, Part 13: PG&E will notify the NRG of the completed shielding analysis update and evaluation findings by July 31, July 31, 2019 2019.}}

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