ML23059A513

Table 1 Responses to NEI Comments on Predecisional Draft Guide ML22276A149

ML23059A513
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Issue date:	03/01/2023
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Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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1 General The basic idea is good-some safety-related SSCs None Thanks could be designed to a lesser seismic design criterion, which would make the plant less expensive and, if applied properly, just as safe.

2 Introduction In the advanced reactor public meeting that took Consider This is now place on 10-12-2022, NRC explained that the guides addressing how to outside the scope B. Proposed Options Options 2 and 3 can also be used under part 50 as apply Options 2 and of the two Draft long as: 3 to Part 50. RGs.

D. Implementation

• A singular SSE applies and its value is not NRC staff to below per Part 50 Appendix S decide whether a separate RG

• Safety criteria similar to those of Part 53 could be Framework A are clearly defined if the SDC is developed to below a Category 5. realize these options.

3 B. Proposed Options The DG provides an example for Option 2, but not for Consider providing Develop one or Option 3. Option 3 may be the most desirable, but it a similar example more example is also the one that will be hardest to reach for Option 3. applications of consensus on without further guidance. Option 3 (with specifications for Framework A and Framework B) as Appendix B to Draft RG 1410.

4 B. Proposed Options The problem with Option 2 is the requirement to See Comment The information and perform a generic SPRA. The DG is light on Basis analyses necessary content, so one needs to refer to the RIL since the to address these DG points to it. Here are some thoughts: comments are important, but too

a. The plants are not designed to a uniform hazard detailed and response spectrum, but to a CSDRS that the vendor extensive to include in the Draft RG

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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selects to hopefully bound the sites where they 1410. We propose might build (or at least yield a design that can be to expand on all qualified for those sites even if the curve is exceeded these points as we to some extent at certain frequencies). revise the RIL and turn it into a

b. The RIL suggests that one approach could be to NUREG/CR that select a bounding hazard. There are two parts: provides many of b.i. Bound the spectral shape. This means to take the the technical bases actual spectral shape for each SDC at all (or a subset) for the two draft for the existing plant sites in the US (there bounds the RGs.

set at all frequencies. This would not be a UHRS, so would not actually provide an accurate SPRA for any site (a point that the RIL acknowledges but does not make a judgement on).

b.ii. Bound the PSHA. This appears to mean to take the hazard curve for the same sites and draw a shape that bounds the AEF at each pga level. This further compounds the issue that it would not apply to any site, and doesnt correspond to the spectral shape being used.

b.iii. By definition, this should be adequately conservative to assure that the categorization would work at any site that was included in the set of sites used for the bounding process.

b.iv. This double bounding approach will almost assuredly result in over-categorization of SSCs for most designs. The issue is that you are matching a PSHA (frequency) curve with a spectrum that is likely quite irrelevant and very conservative (e.g., the bounding PSHA yields a high pga that is mostly

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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associated with sites with low spectral acceleration, but the bounding spectrum has high spectral accelerations associated with sites with low pga).

b.v. It may also miss some things, since site-specific issues such as liquefaction, subsidence, SSI, etc.

could be missed. This could be an issue later, since it is also required to perform the final SPRA and confirm the categorization. At this point it is too late -

the plant has been designed (so it is not too late to change something, but that defeats the purpose of standardized design).

c. The RIL suggests a second approach, which would be to perform what is effectively a series of PRAs by using the UHRS, PSHA, and site-specific conditions for multiple sites. The implication is you then categorize the SSCs based on the most restrictive (which will likely come from different sites).

This would likely give a better answer, but also be quite a bit more expensive to implement. It may result in less over-categorization and be less likely to be challenged by a site-specific PRA.

d. Either approach requires selection of an adequate range of sites to envelope where a plant might be built.

Otherwise, it could result in a failure to confirm the categorization at a particular site.

5 B. Proposed Options The key problem with Option 3 is that the DG is very See Comment Some minor light on content and the RIL doesnt discuss it at all. Basis clarification has This is highly problematic. There needs to be some been added in this actual guidance on what would be acceptable, draft. More

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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otherwise it may take so long to reach agreement enhanced that it may not be timely enough to be useful. discussion will be included in

a. Option 3 does not require a SPRA, either for the Appendix B and categorization or for the confirmation. It allows other revised RIL, which risk-informed approaches but does not provide any we will publish as indication of what these might be. a NUREG.

b. Categorization could be based in large measure on the consequences of the failure of the SSC. For some designs, this approach would be very desirable.

b.i. Again, there needs to be some indication regarding the selection of the scenarios.

b.ii. There also needs to be some guidance on what would be acceptable for determining that the resulting consequences would not exceed the F-C curve if the design criteria were relaxed. There would need to be some consequence margin based on some judgment of the potential accident frequency but taken too far it approaches Option 2.

6 B. Proposed Options Both Option 2 and Option 3 require that the design See Comment Discussions in the decisions be confirmed. For option 2, this is done Basis planned with SPRA. For Option 3, either by SPRA or other NUREG/CR will risk-informed approaches. The DG does not actually be expanded to discuss what this means, but the implication is that clarify that the use this is the site-specific analysis associated with a of a F-C curve is COL or construction permit application. The RIL not mandatory in doesnt mention this at all. Again, the lack of any Option 3.

guidance brings up concerns because once the However, both design is done you would not want to have to change Framework A and an SSC from, say SDC-4, to SDC-5. There needs to Framework B be some finality to the design requirement that would ultimately require

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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not result in a categorization change based on the a SPRA type specific site where a plant would be built. analysis.

a. Would it be adequate to show that the overall seismic risk is low, or would you also have to show that the F-C for each seismic scenario that includes an SDC-4 SSC is under the curve?

b. Would you have to demonstrate that there are no SDC-4 SSCs that are risk-significant regardless of the overall or individual scenario risk?

7 Introduction Based on the advanced reactor public meeting that Consider The revised draft took place on 10-12-2022, NRC also discussed the addressing how to RGs now address B. Proposed Options possibility to use the options under Framework B as apply Option 2 and both Framework A long as one adopts safety criteria from Framework A 3 to Framework B. and Framework B.

instead of Framework Bs principal design criteria. However, the draft RGs do not include These clarifications allow more flexibility to the industry any 10 CFR Part and should be covered in the DGs otherwise the 50 discussions. A industry may assume that if they use Framework B or reference to AERI Part 50, they are not allowed to use the methodology will be added to the from the DG. Draft RGs in the May revision.

8 Introduction It may be useful to address somewhere, not Consider the best NRC staff will necessarily in these guides, what NRC will require to way to provide this decide whether a see if exemptions are considered for either Part 53 or insight to the separate RG Part 50/Part 100 seismic requirements. NRC has industry. could be proposed in the past that some advanced reactors developed to may need to take exemptions from Parts 50/100. realize these Designers in the industry are considering such options under 10 exemptions for their design, beyond the alternatives CFR Part 50 or offered in the DG. Part 52.

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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Therefore, it may be useful to address this pathway as well as to what NRC needs to see to approve potential exemptions. Such exemptions may be related to the SSE/DBGM, and site specific seismic/geotechnical investigations/analyses which are not required for some nuclear facilities in the U.S (see comment under B Proposed Option below for context).

9 Related Guidance RG 1.232. 1.143, 1.166 and NUREG-0800 are Add as needed. These references mentioned in the corresponding section of the are now included seismic isolator DG but not in this DG. in both revised draft RGs.

10 B. Discussion ASCE 43-19 will be endorsed based on this DG while Consider different NRC plans to 43-05 is endorsed based on RG 1.208. It is not clear revisions of ASCE revise RG 1.208 if NRC plans to update RG 1.208 to change the 43 relied upon by because the ASCE revision or if the industry can continue to rely the NRC and clarify current version of on the 43-05 version when operating under Part 50. if the latest version RG 1.208 only only should be refers to SDC-5.

followed. The two draft RGs reference ASCE 43-19 to cover SDC-2 to SDC-5.

If NRC staff choose to provide separate guidance on 10 CFR Part 50, this topic will be included there.

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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11 B. Proposed Options It appears that there should be an option 4 that Consider an option The scope of the follows NRCs process for review and approval of for the use of two draft RGs is facilities such as medical isotope facilities but applied IBC/ASCE 7 based based on ASCE to advanced reactors that have a similar risk level on medical isotope 43-19.

due to their low inventory and dose consequence. and DOE nuclear facility precedents. The IBC/ASCE 7 Shine medical and Northwest medical clearly detail in approach to their PSAR in Chapter 2 that they rely solely on the characterizing the seismic methodology of IBC/ASCE 7. Despite these hazard and evaluating SSCs facilities processing radioisotopes under Part 50 differs from that in licensing, their low-risk thresholds allow the NRC to ASCE 43-19 and approve their application despite it utilizing industrial current NRC non-nuclear codes. IBC/ASCE 7 are also the codes guidance. An used for DOE for its nuclear facilities. evaluation of ASCE 7 for use It is unclear why the same cannot be done for micro- with commercial reactors that present a very low risk and can prove it nuclear power through a source term analysis. plants has not been performed by It is understood that IBC/ASCE 7 could be used as part the NRC staff or of Option 3, however it is not clear if sufficiently low industry for its source term results based on an extreme accident applicability to would be sufficient. lower-risk facilities.

We propose a technical meeting on this topic to address this comment.

12 B. Proposed Options The following statement is somewhat vague: NRC should Some minor elaborate on this clarifications have C.2.4 and provide been added in this

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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risk insights; and a combination of approximate, examples. The draft. More bounding, and conservative analyses and Appendix may be a enhanced quantitative risk information candidate for this. discussion, including some other risk-informed approaches examples, will be included in Several reactor designers will be interested in how to Appendix B and comply with this. the revised RIL/NUREG.

13 Proposed Options Echoing previous comments, Option 1s methodology Preferably, the Some minor is laid out in RG 1.208, while Option 2 is backed up requested clarifications have by NEI 18-04s documentation. Something similar is clarifications for been added in this needed to support Option 3, so it is clear what is Option 3 should be draft. More acceptable. This would help avoid back and forth part of this DG but enhanced with the NRC which could be due to the industrys could be discussion, interpretation of what is reasonable for Option 3. incorporated in an including some external document. examples, will be included in Appendix B and the revised RIL/NUREG 14 C.2 and C.3 Option 3 is still tied to NEI-18-04 because of the Option 3 should be Some minor emphasis on IDP. The connection to IDP is revised to better clarifications are confusing. clarify that using planned for the Option 3 is not the next version.

same as applying More enhanced NEI 18-04 (and its discussion, corresponding IDP). including some examples, will be also included in Appendix B and

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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the revised RIL/NUREG 15 C. Staff Pre-decisional The document references RIL 2021-04 which states: Considering these The scope of the assertions from the two draft RGs is Appendix A The approach in this report can also accommodate NRC and previous based on ASCE codes other than ASCE 43, such as ASCE 7 comments, NRC 43-19.

(ASCE/SEI, 2010), for the design of low-risk facilities. should generate (Microreactors are special systems of relatively low guidance that is The IBC/ASCE 7 risk, and the regulatory framework for them is specific to ASCE 7 approach to evolving (see, e.g., BNL, 2020). and its use, taking characterizing The process uses ASCE 43, for consistency with the into account how it hazard and evaluating rest of this report; however, compliance with risk is already structures and criteria may be shown using other design codes, implemented for systems is entirely such as ASCE 7. medical isotope different than facilities. ASCE 43-19 and It is envisioned that some of the very small advanced current NRC reactors and microreactors (with negligible offsite guidance. It would consequences) can be seismically designed in require an in-depth accordance with ASCE 7. The LMP would impose an evaluation to additional and undue burden in the reactor design assess its process. applicability to lower-risk facilities.

Based on NRC direction, ASCE 7 may be referenced in future revisions to the two draft RGs.

16 C.3.1 Some SDCs below 5 are discussed as examples; It would be helpful SDC-1 is outside however, alternative SDCs were also discussed in to explain that SDC the scope of the advanced reactor public meeting that took place 2 and 1 can be ASCE 43.

on 10-12-2022 as a possibility. Technically both SDC attained based on Therefore, SDC-1

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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2 and 1 are possibilities depending on the dose the overall risk of is not considered consequence of the reactor. the facility. in this guide.

Although, one can imagine other SDCs, ASCE 43 provides specific guidance for SDCs -2, -3, -4, and -5, assuring the target performance in a risk-graded approach is achieved.

17 C.3.3 The section states Seismic loads are prescribed as Consider adding OBE or one-half SSE. some clarification Discussion will be considering the 1/3 expanded in the However, OBEs greater than 1/3 SSE per 10 CFR 50 SSE stipulated in revised proposed Appendix S require additional work and it is not clear Appendix S. NUREG/CR.

if it is also the case for this DG. Appendix S states: However, we conclude that an A value greater than one-third of the Safe Shutdown OBE specific Earthquake Ground Motion design response spectra. design is Analysis and design must be performed to demonstrate impractical when that the requirements associated with this Operating multiple design Basis basis motions are used. For Earthquake Ground Motion in Paragraph proposed Part 53, (a)(2)(i)(B)(I) are satisfied. The design must take into we do not include account soil-structure interaction effects and the an OBE design duration of vibratory ground motion. option.

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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18 C. Staff Pre-decisional There is a lot of discussion of peer reviews related to Consider what This topic will be this section and it is not clear how relevant/required clarification might considered for that is when codes other than ASCE 43 and ASCE 4 be needed in light latest versions as are relied upon, or when a non-PRA method is used of the comment that per NRC direction.

per Option 3. supports the However, both introduction to other draft RGs codes and describe initiatives standards. that are being proposed for the first time in the nuclear industry.

Many situations will likely arise that will require judgement and review by experts, for example non-linear analysis.

19 Appendix A ASCE 349 and N690 are mentioned in Appendix A. The guidance The NRC will There may be cases where these codes are not used should be updated consider the B. Proposed Options as part of Option 3. In those cases, their to clarify that the applicability of commercial/industrial counterparts ASCE 318 and use of alternative ASCE 7 after we ASCE 360 would be used. This would be the case codes is not only have an when the IBC/ASCE 7 is used. applicable to the opportunity to substitution of assess its ASCE 43 and 4 as applicability. In the called out in Figure proposed 1 and extends to approach it is other nuclear crucial that an codes. SSC achieves a design target that allows it to

Part 1: Comments on Pre-decisional DG ML22276A149, Technology-Inclusive Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants Section Comment/Basis Recommendation SwRI Team

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perform its needed safety function and meet safety criteria. The current version of the two draft RGs do not include application of IBC/ASCE 7.

Part 2: Comments on Pre-decisional DG ML22276A154, Seismically Isolated Nuclear Power Plants Section Comment/Basis Recommendation NRC Response 1 General All comments from the above table that are also Make consistent Understood applicable to this DG should also be considered updates as and are not repeated here. needed 2 Related Guidance It may be worthwhile to call out the NUREG Add as needed. We agree and we series on seismic isolation NUREG/CR-7253 to made this revision.

7255 in this section.

3 Background It may be good to also mention the ITER fusion Add as needed. We agree and we reactor under construction in addition to the made this revision.

Horowitz reactor in France as it also uses seismic isolation based on NUREG/CR-7255.

This may become relevant when NRC generates regulation for fusion reactors.

4 Background This section states ASCE/SEI 43-19 is the only Consider The scope of the Draft consensus standard to provide criteria for the modifying RG 1307 is based on seismic design criteria for applications of SI background to ASCE 43-19.

systems in nuclear facilities. account for the broader reference The IBC/ASCE 7 However, as noted in the previous table some on seismic approach to nuclear facilities rely on ASCE 7. ASCE 7 2021 isolation in other characterizing hazard has seismic isolation provisions in Ch 17 would standards. and evaluating be suitable for nuclear structures where the dose structures and systems consequence justifies the use of that code under is entirely different than Option 3. ASCE 43-19 and current NRC guidance.

It would require an in-depth evaluation to assess its applicability to lower-risk facilities.

Based on NRC direction, ASCE 7 may be referenced in future revisions to the two draft RGs.

Part 2: Comments on Pre-decisional DG ML22276A154, Seismically Isolated Nuclear Power Plants Section Comment/Basis Recommendation NRC Response 5 Figure 2 The figure mentions Advanced LWR under The language We agree and we Option 1, however not all advanced reactors should be made modified the text are LWRs. technology accordingly.

inclusive.

6 Figure 2 Given that at least ASCE 7 also provides Propose making In the current version criteria for seismic isolators and testing, the note conditional of the draft RG, only requirement to follow the testing requirement and less restrictive ASCE 43-19 is from 43-19 and 4-16 in cases where ASCE 7 or remove note as considered. An would be acceptable is not fully justified and it is implied that applicant can always may be an undue burden. the testing should use a different code if be in accordance adequate justification with the codes demonstrates that the used. intent of the technical positions is achieved.

We will add this language back into the draft RGs in the May revision.

7 Figure 2 The graded SPRA and RG 1.233 being the Consider adding primary requirements for Option 3 is a the flexibility of The reference to RG departure from the DG documented in the and Other Risk- 1.233 was removed previous table. Informed and some clarifying Approaches for language was added to However later discussions related to Option 3 Option 3. the draft RG text.

state: Option 3 based on a broad spectrum of approaches that include deterministic inputs, risk This would be insights, and a combination of \approximate, consistent with bounding or conservative analyses, and C.3.4.

quantitative risk information.

8 Figure 2 Based on earlier comments in this table and the Consider adding In the current version previous one, it is not reasonable to require all the flexibility of of the draft RG, only seismically isolated advance reactors to use and other codes ASCE 43-19 and 4-17 ASCE 43-19 and 4-16. and standards with are considered. An proper applicant can always justification. use a different code if

Part 2: Comments on Pre-decisional DG ML22276A154, Seismically Isolated Nuclear Power Plants Section Comment/Basis Recommendation NRC Response adequate justification This would be demonstrates that the consistent with intent of the technical C.3.3. positions is achieved.

9 Figure 3 It is also important to note that if other codes Consider similar We will consider this in are allowed, earthquake recurrence and recommended possible future performance targets need to be commensurate relaxation as revisions to the draft with those codes and the selected SDC. above. RG. However, to move in this direction, NRC This impacts the performance targets and will need additional

-5 criteria discussed for DBE at 10 and BDBE at technical information.

-6 10 , probability of unacceptable performance is We propose a to be less than 1% and 10% under DBE DBGM technical meeting to discuss how to align and BDBE DBGM well as the 167% of the other codes and DBGM noted in the figure.

standards and to define performance Note also that requirements to also consider targets.

performance of the isolators during BDBEs could lead to over-designing the isolators.

10 General NUREG/CR-7253 offers: Consider including We are taking this into this fact. consideration.

The ground motion response spectrumare NUREG/CR-7253 calculated for design of nuclear power plants ASCE 43-19 will be by multiplying the ordinates of a uniform hazard reviewed to determine response spectrum at the specified hazard whether and how we exceedance frequency by a design factor that is can modify the draft greater than or equal to 1.0. The factor can be RG as requested, for seismically isolated nuclear set equal to 1.0 for the planned May design of a power plant if the earthquake risk revision.

is dominated by horizontal ground shaking and a stop is provided...

This is understood to mean that the DF utilized in RG 1.208 to convert the UHRS into the GMRS is effectively 1 when using isolators where the

Part 2: Comments on Pre-decisional DG ML22276A154, Seismically Isolated Nuclear Power Plants Section Comment/Basis Recommendation NRC Response horizontal seismic force control and adequate CS is provided. This is an important fact that is not discussed in the DG.

11 Figure 2 Option 2 later discusses 1.5 x DBE but that is Consider clarifying We are considering the not mentioned in the figure. this similar to what clarification, however, was done for the discussion of the PGA is discussed for this DG and the previous, Option 3 in the 1.5 factor in Option 2 is however when other codes are used per Option figure. partly in reference to 3, PGA may not be the key seismic input. When when contact loads ASCE 7 is used, PGA informs the foundation Consider if the should be considered design, however the design of SSCs and the focus on PGA in a design.

DRS is based on bounding spectral might be acceleration rather than PGA. misleading. As stated earlier, the discussion in this guide is related to ASCE 43-19. Both the regulations and ASCE 43 characterize design basis motion by a DRS. The use of PGA to describe the DRS is a historical practice. PGA can be related to another descriptor of a DRS.

12 C.3.4 This statement appears to provide some Consider making We agree and we flexibility but the RG 1.233 occurring after even restrictive. have modified the text other risk-informed approaches still makes it less in the current version restrictive: of the draft RG.

The applicant should demonstrate that the final design satisfies Part 53 safety criteria which could be accomplished through a graded SPRA (or other risk-informed approaches) in accordance with the guidance provided in RG 1.233.

Part 2: Comments on Pre-decisional DG ML22276A154, Seismically Isolated Nuclear Power Plants Section Comment/Basis Recommendation NRC Response 13 C.4.5 The PRISM design referenced in the DG called Consider clarifying This topic will be for a qualification program including the whether any of this evaluated and C.4.6 following to determine horizontal static and still applies in case expanded on the dynamic stiffness, vertical stiffness, damping it is not captured revised RIL/NUREG.

and margin to failure/failure modes: by latest to be endorsed versions

• The testing of high damping bearings of ASCE 43/4 (when they are

• The qualification of expansion joints for used).

the secondary heat transfer system piping

• Large building tests with prototype isolators

• Scale model tests of reactor structure with isolators on a shake table

• The development of analytical models

• Bearing material optimization and qualification

• The development of seismic isolation guidelines

• Seismic margin assessment 14 General Overall Option 3 which should provide the most Consider previous We agree with flexibility is still tied to the used of LMP recommendations decoupling it from the guidance which is not the case for Option 3 on this to give LMP, NEI 18-04/IDP, discussed in the previous table. more flexibility and RG 1.233. We under Option 3 modified the text in and optionally the current version of decouple it from the draft RG to reflect LMP, NEI 18- this. However, ASCE

Part 2: Comments on Pre-decisional DG ML22276A154, Seismically Isolated Nuclear Power Plants Section Comment/Basis Recommendation NRC Response 04/IDP, RG 1.233 43-19 is still the and ASCE 43/4. primary code we rely on for the reasons discussed in the guide.