ML19122A230

Enclosurai (Letter to J. Welsch 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))

ML19122A230
Person / Time
Site:	Humboldt Bay
Issue date:	04/30/2019
From:	Christopher Markley Renewals and Materials Branch
To:	Welsch J Pacific Gas & Electric Co
Markley C
Shared Package
ML19122A229	List: ML19122A230 ML19122A231
References
CAC 001028, EPID L-2018-RNW-0016
Download: ML19122A230 (10)

Text

REQUEST FOR ADDITIONAL INFORMATION Pacific Gas & Electric Company (PG&E)

License Renewal Application Docket No. 72-27 License No. SNM-2514 This request for additional information (RAI) identifies information needed by the U.S. Nuclear Regulatory Commission (NRC) staff in connection with its review of the license renewal application (LRA). NUREG-1927, Revision 1, Standard Review Plan for Renewal of Specific Licenses was used by the staff in its review of the application. Each individual RAI describes information needed by the staff for it to complete its review of the application and to determine whether the applicant has demonstrated compliance with the regulatory requirements.

In responding to the following RAIs, the staff notes that activities a licensee is performing during the current licensing period may be credited towards aging management in the renewed period, provided that the applicant demonstrates that the activities can effectively manage the effects of aging.

CHAPTER 1: GENERAL INFORMATION RAI 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 ISFSI during the proposed license renewal period (years 2025 to 2065). Additionally, provide the rationale for these cost projections.

By letter dated July 10, 2018 (Agencywide Documents Access and Management System Accession No. ML18215A202), Pacific Gas & Electric Company (PG&E) requested renewal of the Humboldt Bay ISFSI, (SNM 2514, Docket No. 72-27), for an additional 40 year period beyond the end of the current license term. The original 20 year ISFSI 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 ISFSI 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 demonstrate to 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 period 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:

Enclosure

1. Estimated construction costs;

2. Estimated operating costs over the planned life of the ISFSI; and

3. Estimated decommissioning costs, and the necessary financial arrangements to provide reasonable assurance before licensing, that decommissioning will be carried out after the removal of spent fuel, high-level radioactive waste, and/or reactor related greater than class C (GTCC) waste from storage.

After reviewing PG&Es submittal, it appears that the estimated operating and maintenance costs, as well as sources of funds to operate the Humboldt Bay ISFSI were not specifically provided in the application for license renewal, nor could this information be easily obtained from staffs review of the PG&E annual report.

This information is needed to confirm compliance with 10 CFR 72.22(e).

CHAPTER 2: SCOPING EVALUATION RAI 2-1: Clarify the Scoping Evaluation with regard to the following items and their safety functions, modifying that 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., see FSAR Sections 7.3.1, 7.3.2, and 7.3.2.2). The soil being in the shielding models or influencing how the shielding analysis was done means the soil has a safety function.

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 of both a shielding and a criticality function in the safety functions for the DFC subcomponents that 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 volume (the DFCs 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 this 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 2

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 inserted into 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.

8. Inclusion of a shielding function in the safety functions for the HB overpacks 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 subcomponents above and below the neutron shielding material.

9. Inclusion of a shielding function for port plugs, 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 and the lid and base of the MPC-HB. The HB overpacks steel shells are included in the criticality model (as are the overpacks and MPCs lids and bases) and help to absorb thermal neutrons in the model. The overpacks neutron shielding is given a criticality safety function. Thus, these steel shells should also have a criticality safety function.

11. Inclusion 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 GWCs 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 GWCs 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 3

top plate, base plate, and outer shell. 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 needed to confirm compliance with 10 CFR 72.42(b), 72.24(d) and (e),

72.104, 72.106, 72.124, and 72.126.

CHAPTER 3: AGING MANAGEMENT REVIEW RAI 3-1: 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.

2. LRA Table 3.8-1, Aging Management Review of HI-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 GTCC Overpack, contains a line item for the intermediate shells exposed to enclosed air (internal) and sheltered (external) 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).

RAI-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 HI-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 the 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 4

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).

RAI 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-energizing seals.

This information is needed for evaluating HI-STAR Humboldt Bay (HB) ISFSI Renewal, in compliance with 10 CFR 72.122(b),(c), 10 CFR 72.42(a)(1).

RAI 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 less 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 4.0x10-8. Thus, it is not clear how the 5x10-10 fraction was derived.

2. The location in the Humboldt Bay ISFSI FSAR that establishes the design basis limits for surface dose rates. The fourth paragraph of Element 5 of the HB ISFSI Reinforced Concrete Structures AMP (Table A-2 of the renewal application) indicates that these limits are in Chapter 5 of the ISFSI FSAR. However, the staff did not find where the dose rate limits were established in that chapter of the FSAR.

This information is needed to confirm compliance with 10 CFR 72.42(a).

APPENDIX A: AGING MANAGEMENT PROGRAM (AMP)

RAI 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 Transportation System AMP.

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LRA Appendix A-3, Cask Transportation System AMP, states that visual inspections of the transporter structure, cask restraint system, and wedge lock 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 inspections.

This information is required to demonstrate compliance with 10 CFR 72.42(a).

RAI A-2: In FSAR Section 9.4.3.3.3, Cask Transportation System AMP, clarify the acceptance criteria for the tactile inspections of polymers that are subject to hardening.

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).

RAI 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 for transport cask inspections while transport casks are in use).

This information is required to demonstrate compliance with 10 CFR 72.42(a).

RAI 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.

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• AMP Element 3, Parameters Monitored or Inspected, states that [i]nspections of exposed portions of the below grade concrete are conducted when excavated for any reason.

• Conversely, AMP Element 4, Detection of Aging Effects, states that [e]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 ACI 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. Section 3.4 of ACI 349.3R states that the combination of soil/groundwater 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 inspections of accessible concrete will evaluate worst-case conditions.

This information is required to demonstrate compliance with 10 CFR 72.42(a).

RAI A-5: 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 embrittlement aging effect and mechanism are to be managed as part of the HB ISFSI 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 AMPs 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.

RAI A-6: Provide an evaluation of the public and occupational doses for operations for overpacks and the ISFSI 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 7

4. addresses all relevant operations configurations within the design basis.

The renewal application includes discussion of aging effects and mechanisms for the carbon steel subcomponents of the overpacks and the ISFSI vault as well as the vault concrete and the Holtite-A neutron shielding for the overpacks containing spent fuel. These components are included in the shielding analysis for determining 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 ISFSI 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 ISFSI (e.g., operations with the vault cell lid removed, operations with the overpack out of the vault cell for preparation for transport). Thus, the licensees 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 ACI 349.3R evaluation criteria. These criteria are intended for ensuring 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 Holtite-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 should 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 lids 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 overpacks 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 in Section 11.4.3.1 of NUREG-1567, particularly the bulleted list at the end of the section, should be considered, as needed.

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).

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RAI A-7: 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 ISFSIs 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 interior. While surveys of the ISFSIs general area or the dose data from the TLDs will provide an indication of the overall ISFSI dose rates and doses, the licensee should justify how these data will enable identification of degradation of an individual overpacks 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-A degradation on an individual overpack that requires corrective action. The 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 criteria 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).

APPENDIX D: FINAL SAFETY ANALYSIS REPORT UPDATE SUPPLEMENT AND CHANGES RAI D-1: Clarify if Boral is used as a neutron poison in the MPC-HB.

In LRA Appendix D, Final Safety Analysis Report Update Supplement and Changes, Boral 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, Boral is included in FSAR Table 4.6-1 and FSAR Sections 4.4.3.6, 4.6.1.2, and 4.6.3.

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The staff requires clarification of the use of Boral to ensure that the neutron poison in the MPC-HB is appropriately evaluated for aging.

This information is required to demonstrate compliance with 10 CFR 72.24(c) and 72.42(a).

RAI D-2: Provide the rationales for the use of previous versions of Interim Staff Guidance (ISG)

The applicant does not use updated versions of NRC ISG in Table 4.2-12 of HUMBOLDT BAY ISFSI FSAR UPDATE:

ISG 2. Fuel Retrievability 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 associated with undefined fuel if any. The clarification of ISG 1 would be applied to ISG 2 in terms of retrievability requirements.

This information is needed for evaluating HI-STAR Humboldt Bay (HB) ISFSI Renewal, in compliance with 10 CFR 72.42(a)(1), 10 CFR 72.122(b) (l).

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