ML22102A014
| ML22102A014 | |
| Person / Time | |
|---|---|
| Issue date: | 05/13/2022 |
| From: | Carolyn Lauron NRC/NRR/DNRL/NRLB |
| To: | |
| Lauron C, 301-415-2736 | |
| Shared Package | |
| ML22062A007 | List: |
| References | |
| DPO-2020-004, OEDO-22-00085, OEDO-22-00086, OEDO-22-00087 | |
| Download: ML22102A014 (7) | |
Text
Responses to Submitters Comments in February 14, 2022, Letter to the Executive Director for Operations (EDO)1 In the table below, the Office of Nuclear Reactor Regulation (NRR) staff provides its responses to comments in the submitters letter to the EDO in response to the EDOs decision on Differing Professional Opinion (DPO) case number DPO-2020-004.2 Comment Reference
- to Submitters Comment NRR Staffs Response
[Staff Submitters Assigned] Letter Comments Regarding DPO Appeal Issue 1 DPO Appeal Issue 1, as stated in the EDO decision: The NuScale reactor building design is incomplete, inadequate, and unsafe for the design basis earthquake [DBE] (safe shutdown earthquake [SSE] / Certified Seismic Design Response Spectra [CSDRS]).
This is primarily because no design modifications were made when demand forces exceeded capacity.
1A February 14 The more ductile of a structural element, the The NRR staff disagrees with the comment.
Letter at 2 more stress excessive over the (elastic) capacity Resolution of localized demand-to-capacity available at can be redistributed to its neighboring structural exceedances from linear elastic analysis is not DPO-2020- elements. A structural analysis is required generally a safety-significant structural analysis 004 Case under such a condition and the analysis result issue. Resolving these exceedances by averaging File at 105 will show whether the structural element can with adjacent elements after assessing the specific redistribute the excessive stress over its (elastic) area or element(s) with exceedance is a normal, capacity to its neighboring structural elements or accepted approximation in professional not without failure. No one should use his/her engineering practice and applicable to the NuScale judgement to determine whether the stress design. This approach has satisfactorily achieved redistribution is possible or not and how much results consistent with the expected performance and to how many structural elements because goal (limit state) of essentially elastic structural that subjective approach has no basis just like behavior (i.e., allowing for only limited localized the stress averaging issue in issue #1. This is damage and/or inelastic behavior).
a structural analysis issue not a judgment issue. For the NuScale design, the overwhelming majority of elements meet code acceptance criteria 1 Letter dated February 14, 2022, from John Ma to EDO titled Respond and Request to EDO (February 14 Letter), at pp. 104 - 111 of the DPO-2020-004 Case File, redacted, public version, Agencywide Documents Access and Management System (ADAMS) Accession No. ML22056A017.
2 Memorandum from D. Dorman to J. Ma, Differing Professional Opinion Appeal Concerning DPO-2020-004, dated February 8, 2022, ADAMS Accession No. ML22021B617.
Enclosure 2
on an element-basis (without further evaluation or averaging) based on demands from linear elastic analyses. This fact provides sufficient assurance of the general essentially elastic behavior of the relevant NuScale structures. Since the structural analysis and design process involves making appropriate modeling idealizations, assumptions, and approximations, the structural analyst and engineer(s) of record are expected to interpret analysis results and exercise professional engineering judgements, consistent with expected structural behavior and analysis objectives.
1B February 14 I want to point out that (1) the AISC [N690-18] The NRR staff disagrees with the comment in part.
Letter at 2 standard is only applicable to steel structures The NRR staff agrees that concrete material is available at and steel material is inherently ductile, and does inherently brittle under certain load conditions.
DPO-2020- not apply to concrete structures, such as the Physical characteristic can be improved through 004 Case NuScale reactor building, because concrete well-designed structures using steel reinforcement File at 105 material is brittle, and (2) if the stress or steel composite (SC) construction. The stress averaging is limited to no larger than twice the averaging provision mentioned in the comment is section thickness for ductile steel material, how only applicable to SC construction addressed by could anyone justify the use of stress the new Appendix N9, Steel-plate Composite (SC) averaging with four times the section thickness Walls of ANSI/AISC N690-18, Specification for for brittle concrete material for the NuScale Safety-Related Steel Structures for Nuclear reactor building as stated in the NuScale DC Facilities.3 It is not applicable to steel or
[design certification] application? reinforced concrete structures. The ANSI/AISC N690-18 specification has no stress averaging requirements for steel structures, its primary scope, and the ACI 349-064 code has no stress averaging requirements for reinforced concrete structures.
3 American Institute of Steel Construction Standard ANSI/AISC N690-18, Specification for Safety-Related Steel Structures for Nuclear Facilities, Appendix N9, Steel-plate Composite (SC) Walls, June 28, 2018.
4 American Concrete Institute Code, ACI 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary. ACI 349-06 is the code of record for the NuScale reinforced concrete design.
Enclosure 2
SC construction is a relatively new composite concept consisting of plain concrete sandwiched between two steel faceplates. Since reinforced concrete has superior ductility characteristics compared to SC construction, and better ability to redistribute forces and moments through cracking, even prior to reinforcing steel (rebar) yielding, as well as by rebar yielding, reinforced concrete structures are able to redistribute forces and moments over lengths used in the NuScale analysis as reviewed by the NRR staff.
1C February 14 The above excerpt from the EDOs letter does The NRR staff disagrees with this comment.
Letter at 3 not include the example that I provided in my Although the EDOs letter did not explicitly mention available at DPO report and in Appeal to EDO report. That element 4942, in Item 6, Enclosure 1, the NRR DPO-2020- example has a structural element with D/C > 3.0 staff has provided a detailed assessment of in-004 Case [i.e., D/C = 3791/1184 = 3.2], much greater than plane shear, including exceedances for element File at 106 the D/C > 0.8, and used ten structural elements, numbers 4942, which is discussed by the DPO many more than the three elements as stated submitter, and 4951. These elements are above, for stress averaging. That example in reentrant corner elements at the north and south my Appeal to EDO report is copied below: ends of the top of the short partition weir wall along reactor building (RXB) grid line 3, as shown in The in-plane shear force (the demand) acting DCA Part 2, Tier 2, Figures 3B-10 and 3B-11.
on Element number 4942 is 3791 kips (1 kip =
1000 pounds) but the structural element only In Item 6, the NRR staff discussed how NuScale has a shear capacity (or strength) of 1184 addressed in-plane shear in two ways. Briefly, kips. The force (the demand) acting on the NuScale addressed the issue by incorporating it in element is more than three times greater than its the main reinforcing steel design based on capacity. No design modification was done, and element demands and by performing an additional no post-yield structural element properties were gross structural wall check, consistent with created and used to capture the condition or provisions in Section 21.7.4 and the related behavior of these overstressed structural Section 11.10 of the ACI 349-06 code. As stated elements when the reactor building is only in the Item 6, the NRR staffs review concluded subjected to the design-basis (CSDRS) that the applicant addressed in-plane shear, earthquake. The applicant arbitrarily brought including the example cited by the submitter, in an down the high shear stress by averaging the acceptable and appropriate manner.
shear stress of ten structural elements (see Enclosure 2
page15 in my DPO report). (emphasis in original)
Comments Regarding DPO Appeal Issue 2:
DPO Appeal Issue 2, as stated in the EDO decision: Structural collapse due to shaking from the review level earthquake (RLE) was not evaluated for the NuScale reactor building, so there is no seismic margin incorporated into the structural design. This is, in part, because the NRC has not provided a definition or interpretation of the NRC policy in SECY-93-087 with respect to seismic margin. Using a probabilistic risk assessment (PRA) method alone for evaluation of building safety at the RLE is incorrect.
2 February 14 I am not disputing the adequacy of the EDOs The NRR staff disagrees with this comment. The Letter at 3-8 answer, but the above answer does not address Executive Director for Operations (EDO) response available at or apply to my DPO issue. My DPO issue is not to the Differing Professional Opinion Appeal DPO-2020- about the PRA-based seismic margin analysis Concerning DPO-2020-004 addressed the concern 004 Case for the entire plant. My DPO issue is about that raised in the DPO submitters appeal. Also, the File at 106- the required seismic margin for the reactor technical issues discussed in the DPO submitters 111 building has not been designed into the February 14, 2022 letter are not materially different building. The PRA-based seismic margin from the technical issues that the DPO submitter analysis for the entire plant and the required previously provided. As a result, the NRR staff seismic margin for the single reactor building response focuses on the DPO submitters concern are two different subjects that require two that the seismic margin for the reactor building different approaches. The former belongs to the (RXB) could be incorrectly represented in the discipline (or field) of probability while the latter staffs analysis.
belongs to the discipline of structural engineering. The reason that no seismic margin In response to the DPO submitters concerns that had been designed into the reactor building was one could draw an incorrect conclusion from the because the lack of recognition of this distinction NRR staffs analysis about the RXB seismic between the two subjects. The lack of this robustness during a review level earthquake distinction was caused by that the previous NRO (RLE), the NRR staff provides additional (now NRR) management had prohibited the use discussion of its analysis as previously of the structural engineering approach (method documented in the Final Safety Evaluation Report and process) for seismic margin design for the (FSER). In Section 19.1.4.8.1, Seismic Risk reactor building and replaced it by the PRA Evaluation, Chapter 19 Probabilistic Risk approach and moved the review responsibility Assessment and Severe Accident Evaluation, the from structural engineers (Structural peak ground acceleration (pga) corresponding to Engineering Branch) to probabilistic risk the NuScale design basis safe shutdown Enclosure 2
analysts (PRA Branch). This management earthquake (CSDRS or SSE) is 0.5g. Using the action resulted not only in no seismic margin 1.67 seismic margin figure cited by the DPO being designed into the reactor building but also submitter, the pga acceleration corresponding to in a false claim or implication that the building that would be 0.84g.
possessed a seismic margin of 1.67 and would not collapse during the review level earthquake In Section 19.1.4.8.1.2, Seismic Fragility (RLE) without being noticed to causal readers. Evaluation, the NRC structural engineering staff The no seismic margin analysis/design and the reviewed the seismic fragility evaluation of false claim or implication are presented and structures and structural components (SSC), as discussed below. documented in FSER. The staff documented that a separate fragility analysis was performed for each structure listed in DCA Part 2, Tier 2, Chapter 2.1 No seismic margin was designed into the 19, Section 19.1.5.1 Seismic Risk Evaluation, reactor building while other important buildings Table 19.1-35 Structural Fragility Parameters and have including the AP1000 shield building Results. The RXB structural components evaluated included the RXB crane, RXB exterior walls, module supports, bioshield, pool walls, crane support walls, bay walls, roof, and basemat.
2.2 The subtly false claim or implication that the The fragility analyses were performed using NRC-reactor building possessed a seismic margin of endorsed methods in DC/COL-ISG-20 1.67 and would not collapse during RLE should (conservative deterministic failure margin method be corrected in the FSER for the NuScale or separation of variables method). The staff also design certification application documented that it audited a summary of the fragility calculations of several PRA-critical structures, including the reactor building structures listed above, and the staff verified the 2.3 The lack of distinction between the PRA- assumptions, controlling failure modes, and the based SMA for the entire plant safety and the results of the seismic evaluation (performed by the seismic margin for the single reactor building applicants structural engineers) presented in DCA safety has caused the unsafe design for the Part 2, Tier 2, Table 19.1-35, Table 19.1-38 reactor building Seismic Correlation Class Information, and Table 19.1-40 Key Assumptions for the Seismic Margin Assessment. The results of these structural fragility evaluations included the median seismic 2.4 The unsafe design of the certified reactor capacity and uncertainty parameters (randomness building and the subtle claim or implication that and modeling uncertainties). Using the fragility Enclosure 2
the reactor building possessed a seismic margin parameters, a ground motion representing high of 1.67 and it would not collapse are wrong and confidence (95 percent) of low probability (5 need to be corrected percent) of failure (HCLPF) was calculated for each SSC. The fragility parameters were then used as inputs to the PRA model for the seismic margins analysis used to determine the plant-level 2.5 Moving the structural engineers review HCLPF.
responsibility to the probabilistic risk analysts is improper (this is the first time occurred in my As shown in DCA Part 2, Tier 2, Tables 19.1-35 more than 47-year service in the NRC) and that and 19.1-38, the lowest design-specific (DS) improper action resulted in unsafe design and HCLPF seismic capacity values calculated for that action should be corrected structural failure events of the RXB structural components were 0.88g for the RXB crane and 0.92g for the RXB exterior walls. Thus, these SSCs have demonstrated adequate seismic 2.6 The two major problems for the certified robustness when exposed to pga of 0.84g or 1.67 reactor building and their proper Resolution times the CSDRS acceleration. As documented in FSER Section 19.1.4.8.1.2, the staff verified that The problems as stated above include (1) no no SSCs with HCLPF capacities less than 0.84g, seismic margin was explicitly designed into the as indicated in DCA Part 2, Tier 2, Table 19.1-38, reactor building while other important buildings contribute to the seismic margin.
have, including the AP1000 shield building, and (2) the PRA staff had concurred with the In FSER Section 19.1.4.8.1.2, the staff applicants false claim or implication that the documented its review of NuScales assumption5 reactor building possessed a seismic margin of that seismic Category I structures meet the 1.67 and would not collapse during the RLE, seismic margin criteria of 1.67 times the CSDRS which is obviously wrong. for site-specific seismic hazards (e.g., sliding, overturning). The staffs review concluded that it is Request #2: a reasonable assumption for the purpose of the DCA. Consistent with Tier 2 COL information 2.1 The EDO needs to obtain an answer from items, the combined license (COL) applicant will the NRR on whether the certified reactor need to confirm the validity of this assumption with building will collapse during the RLE or not, and other Tier 2 information items as part of its COL the basis for that answer, and the value application.
5 NuScale DCA Part 2, Tier 2, Table 19.1-40, Key Assumptions for the Seismic Margin Assessment, ADAMS Accession No. ML20224A508.
Enclosure 2
(numerical number, such as 1.5 or 1.67 or any other numbers) of seismic margin that the The EDO response acknowledges the important reactor building possessed so that the public role that structural engineers have in the can see the adequacy of the reactor building development of seismic margins analysis.
design and its actual seismic margin value. Structural engineers, similar to the role of mechanical engineers in reviewing fragility 2.2 The EDO needs to obtain an answer from analysis of mechanical and electrical equipment, the NRR explaining its logic and reason for review the fragility evaluation of structures and prohibiting the use of structural engineering structural components including any related approach and replacing it by the PRA approach structural analyses. The NRC structural for assessing the seismic margin and safety of engineering staff actively participated and the reactor building so that the public can see reviewed the applicants seismic fragility evaluation and judge whether such an action is proper, or it of the structures and structural components had resulted in unsafe design for the reactor (including RXB structural components) and its building. results (median capacity, uncertainty parameters, and HCLPF) that were used to develop inputs to (emphasis in original) the plant-level PRA model, consistent with the NRC guidance in DC/COL-ISG-20. The results of the fragility evaluation indicate that, in addition to the plant level HCLPF, the HCLPF values of the RXB structural components examined were also above 0.84g. For the reasons given above, the NRR staff concludes that the FSER accurately documents the staffs review and conclusions.
Enclosure 2