ML24135A194
| ML24135A194 | |
| Person / Time | |
|---|---|
| Site: | 99902056 |
| Issue date: | 05/14/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL |
| References | |
| Download: ML24135A194 (21) | |
Text
From:
Sean Gallagher Sent:
Tuesday, May 14, 2024 10:43 AM To:
Schiele, Raymond Joseph Cc:
Allen Fetter; Greg Cranston; ClinchRiver-CPSafDocsPEm Resource
Subject:
TVA CRNS Readiness Assessment Phase 5 Initial PSAR Observations Attachments:
Readiness Assessment Phase 5.pdf Good morning, Attached are initial NRC staff observations on draft PSAR Sections 9.3, 9.4, 9.5, 9A and all of Chapter 15 that the staff viewed in TVAs electronic reading room as part of Phase 5 of the Readiness Assessment of the draft Clinch River CP application.
After all six phases of the Readiness Assessment are completed, NRC will transmit, via letter, a compilation of all final Readiness Assessment Observations on the Clinch River CP PSAR chapters and sections. The nomenclature initial is being used to account for potential TVA updates to the PSAR before the end of the Readiness Assessment.
If TVA makes any future updates to the draft PSAR chapters and sections for follow up observations by NRC staff, please contact myself, Greg Cranston or Allen Fetter.
- Regards, Sean Gallagher, Project Manager NRR/DNRL/NLIB
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TVA CRNS Readiness Assessment Phase 5 Initial PSAR Observations Sent Date:
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Readiness Assessment - Phase 5 1
Clinch River Nuclear Site Unit 1 BWRX-300 Preliminary Safety Analysis Report March 20, 2024 Chapter/Section 9.3, 9.4, 9.5, 9A Section Basis for Observation/Comment Readiness Assessment Observations/Comments 9.3 - Process Auxiliaries 10 CFR 50.62(c)(4)
Section 9.3.5 identifies NEDC-33912-P-A, "BWRX-300 Reactivity Control" as approving the removal of the Standby Liquid Control System, however this approved LTR did not approve an exemption. The SER noted, "The staff concludes that an analysis that demonstrates P(ATWS) is less than 1x10-5 per reactor year
[without the use of an SLCS] could support an exemption from 10 CFR 50.62(c)(4). When the NRC receives an application for a BWRX-300, the staff will conduct an evaluation of reliability or probabilistic analysis that demonstrates the P(ATWS) criterion is met [without the use of an SLCS], conforms with Limitation and Condition 5.1 of this SE, and confirms that special circumstances justify an exemption from 10 CFR 50.62(c)(4)." No exemption or mention of an exemption are listed in Section 9.3.5, nor an evaluation pursuant to Limitation and Condition 5.1 of NEDC-33912-P-A, "BWRX-300 Reactivity Control."
9.4 - Air Conditioning, Heating, Cooling and Ventilation Systems N/A None
Readiness Assessment - Phase 5 2
9.5 - Other Auxiliary Systems SRP 9.5.8 Section 9.5.8 does not discuss the location of air intake in relation to SDG exhaust in order to prevent dilution or contamination.
9.5 - Other Auxiliary Systems GDC 2 Section 9.5.4 references GDC 2 but does not provide detail on how the system meets the GDC, such as designation as seismic category I.
P&IDs have historically been used as an aid to show where seismic/quality classification changes occur within a system, if the entire system is not a single classification 9.5 - Other Auxiliary Systems Clarity The text in Section 9.5.1.2.6, subsection Fire Pumps, identifies where two of the three fire pumps are located. It is unclear where the second electric fire pump is located.
9.5 - Other Auxiliary Systems Clarity In Section 9.5.1.2.6, it is unclear whether the Fire Water Enclosure identified for an electric fire pump or the unnamed fire-rated enclosure for the diesel-driven fire pump are the same structure as the Fire Pump House identified on Figure 9.5-1.
9.5 - Other Auxiliary Systems Information Section 9.5.1.6 identifies RG 1.189, Revision 4 as applicable guidance. The staff notes that Revision 5 to Regulatory Guide 1.189 has been issued.
9.5 - Other Auxiliary Systems RG 1.189 Section 3.2.1 While the level of detail provided for the exception to RG 1.189 position 3.2(j) [seismic water supply for manual firefighting] in Section 9.5.1.9 is likely sufficient for the PSAR (construction permit) review, additional detail and justification will likely be required for the FSAR (operating license) review.
9.5 - Other Auxiliary Systems Clarity/RG 1.189 NFPA standards 16, 92A, and 1081 are mentioned in in the text of Section 9.5.1, but do not appear in Table 9.5-1, List of Applicable Codes, Standards and Regulatory Guidance for Fire Protection.
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9A - Fire Hazards Analysis RG 1.189 Section 8.2 The strategy described in section 9A.3.2 seems to imply that SSCs within the fire area where the fire occurs may be relied on to achieve safe shutdown. This would be contrary to the guidance in RG 1.189 section 8.2, but no deviation from the guidance has been identified.
Readiness Assessment - Phase 5 1
Clinch River Nuclear Site Unit 1 Preliminary Safety Analysis Report - Chapter 15 April 16, 2024 Chapter/Section 15 - General Observations Section Basis for Observation/Comment Readiness Assessment Observations/Comments 15 - General Observations Information/Clarity Section 15.2.2 of the PSAR discusses categorization of events according to their frequency of occurrence. The section mentions that selection of event frequency is described in subsection 15.6.3.1. However, justification or documentation for selection each of the events analyzed in Section 15.5 to their assigned classification is not provided.
For example, Section 15.5.3.1.1 analyzes Loss of Feedwater Heating event as a BL-AOO while 15.5.4.1.1 has Loss of All Feedwater heating as a CN-DBA. Similarly, Section 15.5.3.2.1 analyzes Generator Load Rejection or Turbine Trip as BL-AOO and Section 15.5.4.2.1 has the same event, with passive common cause failure of the DL2 functions, analyzed as CN-DBA.
The applicant should provide a detailed evaluation for selection of each of the events analyzed in Section 15.5 to their respective assigned event categories. For example, in some cases these event classifications do not coincide with the ESBWR event classifications.
15 - General Observations Regulatory Guide (RG) 1.70, Standard Format and Content of Safety Analysis Report for Nuclear Power Plants Chapter 15 of the PSAR lists various LTRs including BWRX-300 Safety Strategy and other LTRs for various codes (e.g. TRACG, PANAC11, PAVAN etc.) and methodologies.
Regulatory Guide (RG) 1.70, Standard Format and Content of Safety Analysis Report for Nuclear Power Plants, provides guidance regarding which material must be incorporated by reference. The PSAR Chapter 15 should also disposition all the limitation and conditions associated with the LTRs incorporated in the Chapter 15 safety analyses.
Readiness Assessment - Phase 5 2
Chapter/Section 15.5 - Deterministic Safety Analysis Section Basis for Observation/Comment Readiness Assessment Observations/Comments 15.5 -
Deterministic Safety Analyses 10 CFR 50.2/ 10 CFR 50, Appendix A Section 15.5.3 of the PSAR presents analysis of the Anticipated Operation Occurrences (AOOs). The analysis of events listed within the sub-sections of Section 15.5.3 shows DL2 functions being credited to mitigate the AOOs. The BWRX-300 Safety Strategy LTR states that that SC3 and SCN, Non-Safety-Related SSCs perform the DL2 primary and support functions while SC1, important to safety or safety related, are required to perform DL3 primary and integral support functions. Based on the description provided in the BWRX-300 Safety Strategy LTR, DL2 SSCs are the primary success path for protecting multiple fission product barriers (e.g., SAFDLs, RCPB) during design-basis transients (i.e., AOOs).
The NRC staff notes that AOOs are considered design-basis events, and consistent with the definition of safety-related in 10 CFR 50.2, any SSCs that perform the functions prescribed in the definition are required to be classified as safety-related." The 10 CFR 50.2 states that safety related SSCs are those that are relied on during or following a design basis event to assure, in part, (1) The integrity of the reactor coolant pressure boundary, and (2)
The capability to shut down the reactor and maintain it in a safe shutdown condition. AOOs are defined in Appendix A to 10 CFR Part 50, as those conditions of normal operation that are expected to occur one or more times during the life of the nuclear power unit.
The general design criteria (GDC) in 10 CFR 50, Appendix A provides the minimum requirements and criteria for maintaining the integrity of the reactor coolant pressure boundary (RCPB), and for shutting down the reactor and maintaining it in a safe condition for AOOs and postulated accidents such that there is reasonable assurance that the
Readiness Assessment - Phase 5 3
facility can be operated without undue risk to the health and safety of the public. For AOOs, the GDCs prescribe a safe shutdown condition to be one where decay heat is being sufficiently removed and the fuel integrity barrier is maintained by demonstration of appropriate margin to the specified acceptable fuel design limits (SAFDLs) (e.g., GDC 34). The NRC staff notes that classification of the entire DL2 as non-safety related places significantly more reliance on non-safety related equipment to mitigate AOOs and will require appropriate justification.
15.5 -
Deterministic Safety Analyses 10 CFR 50.12/ 10 CFR 50, Appendix A The deterministic safety analysis performed in Section 15.5 of the PSAR shows that the Single Failure criterion is only applied to DL3 Safety Class 1 (SC1) SSCs. This is consistent with the BWRX-300 Safety Strategy LTR which states that meeting single failure criterion is not a design rule applied to Safety Category 2 or 3 functions.
The LTR states that redundancy may be required for other reasons and points out that sufficient mechanical, electrical and instrumentation component should be provided.
The staff notes that 10 CFR Part 50, Appendix A, General Design Criteria require that the design include the capability to withstand single failures. These requirements are not exclusive to design-basis accidents.
Several GDCs have requirements for SSC reliability for mitigation of AOOs (e.g., GDC 17, 21, 22, 25, and 34). Alternate approaches to application of Single Failure criterion need to provide appropriate justification and may require exemptions to applicable regulatory requirements in accordance with 10 CFR 50.12.
15.5 -
Deterministic Safety Analyses Information/Clarity The input parameters used in the analysis for the AOOs, and the DBAs are presented in Table 15.5-3. Based on the information provided in the Table, it appears to the NRC staff that nominal values were used as an
Readiness Assessment - Phase 5 4
input for the various plant parameters for the analyses performed and no information is provided on the biasing of the values. The CP application should include justification on use of nominal values for the subject analyses or provide details on the conservative biasing used for each of the input parameters.
For example, Section 15.5.3.2.3 of PSAR provides AOO analysis of loss of condenser vacuum. However, the Table 15.5-3 does not contain the bounding conservative value for initial rated core power considering measurement uncertainties, feedwater flow rate and feedwater runback coast down time and initial MCPR of the hot channel.
15.5 -
Deterministic Safety Analyses 10 CFR 50.46/10 CFR 50, Appendix A/10 CFR 50.12 The deterministic safety analysis discussed for the LOCA events in Chapter 15.5 of the PSAR does not provide any evaluation for losses of coolant at the reactor isolation valve to the reactor vessel connections. The NRC staff notes that there is no analysis performed elsewhere in the PSAR for this specific location or a similar location yielding representative comparable results for the hypothetical loss of coolant at the reactor vessel to isolation valve flange. LOCAs are defined in 10 CFR 50.46(c) and 10 CFR Part 50, Appendix A, and analyzed in SAR Chapter 15 as non-mechanistic hypothetical breaks to establish the design-basis of the ECCS. 10 CFR 50.46(a)(1)(i) states, ECCS cooling performance must be calculatedfor a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.
The PSAR must include an analysis, and the ECCS sized, for all losses of coolant from the reactor coolant pressure boundary or else provide an appropriate justification for any exemption(s).
Readiness Assessment - Phase 5 5
15.5 -
Deterministic Safety Analyses NUREG-0800 Chapter 15/10 CFR 50, Appendix A
Section 15.5.3.2.2 of the PSAR lists closure of One Main Steam Reactor Isolation Valve (MSRIV) as an BL-AOO event. Section 15.5.5.2.1 of the PSAR later lists the closure of one MSRIV event as an EX-DEC event.
The postulated initiating event for the EX-DEC one MSRIV closure is same as the BL-AOO event but assumes a common cause failure hydraulic scram failure. The NRC staff notes that one main steam line isolation valve closure event is listed in the SRP (NUREG-0800) Chapter 15 as events that occur with moderate frequency and have been analyzed for other BWRs as DBEs, including the ESBWR DCD which lists it as an AOO. Please provide justification on classification of event of moderate frequency like closure of one MSRIV as an EX-DEC event.
15.5 -
Deterministic Safety Analyses 10 CFR 50, Appendix A/10 CFR Part 100 The safety analysis performed in Chapter 15.5 of the PSAR does not include the inadvertent loading and operation of fuel assembly in an improper position. Analysis of this event is required to meet the acceptance criteria for GDC 13, pertaining to providing instrumentation to monitor variables over anticipated ranges for normal operations, for anticipated operational occurrences, and for accident conditions, as well as compliance to 10 CFR Part 100, as it relates to offsite consequences resulting from reactor operations with an undetected misloaded fuel assembly. The event is listed in SRP (NUREG-0800) Chapter 15.4.7. As an example, the NRC staff notes that ESBWR DCD lists Fuel Assembly Loading Error, Mislocated Bundle as Infrequent Event.
Infrequent Event is defined as the DBE with probability of occurrence < 1/100 and radiological consequence less than a design basis accident ( 2.5 rem). The ESBWR DCD states that the GESTAR analysis (GESTAR II, Amendment 28) for Fuel Assembly Loading Error, Mislocated Bundle bounds the ESBWR design. For the event, the DCD also lists that Should a bundle
Readiness Assessment - Phase 5 6
mislocation, misorientation, and improper seating occur and go undetected, the plant specific acceptance of the generic GESTAR analysis is revoked, and the classification of this event is changed from infrequent incident (infrequent event) classification to an incident of moderate frequency (AOO) classification immediately for that plant. The CP application needs to provide justification for not analyzing the inadvertent loading and operation of fuel assembly in an improper position event in the deterministic safety analyses.
15.5 -
Deterministic Safety Analyses Information/Clarity Section 15.5.5.1 of the PSAR states that the NRC approved Control Rod Drive Accident (CRDA) methodology (LTR NEDE-33885-P-A, Revision 1) will be applied to the BWRX-300 to demonstrate that the cladding failures do not occur for postulated CRDA. It further states that results of rod drop calculations will be discussed in PRA which will be summarized in future licensing activity. The PSAR cites L&C 5.2 of the SER to the CRDA methodology and states that An exemption is provided in Section 3.1 to justify the CRDA as practically eliminated event with accompanying probabilistic risk assessment.
The NRC staff notes that Section 3.1 of the PSAR is not available at this time and NRC staff cannot provide any feedback on Technical Basis and Probability Analysis provided in subsections of Section 15.5.5.1 in absence of the exemption request along with the PRA.
15.5 -
Deterministic Safety Analyses Information/Clarity Section 15.5.2 of the PSAR provides the stability analysis results regarding the core wide mode of density wave oscillations. It is concluded that the system decay ratio is less than the acceptable limit for normal power operation. Although BWRX-300 operating domain is different from operating BWRs due to the in-vessel natural circulation, the stability analysis should reflect the possible range of RPV pressure, feedwater temperature and potentially the downcomer water level permitted for operation.
Readiness Assessment - Phase 5 7
15.5 -
Deterministic Safety Analyses Information/Clarity Section 15.5.3.2.2 provides the analysis results of an AOO of the Closure of One Main Steam Reactor Isolation Valve. Section 15.5.4.2 provides the analysis results of a DBA of the closure of all MSRIVs and FWRIVs. The listing of DBAs appears to be missing an event for the simultaneous closure of all MSIVs (without FWRIV closure), which could be more severe if the closure of the FWRIVs is not assumed.
Provide justifications that the closure of all MSIVs is not a DBA event or the relevant analysis results of all MSIVs closure as a DBA.
15.5 -
Deterministic Safety Analyses Information/Clarity Section 15.5.4.5 states that acceptance criteria for fuel integrity for LOCA DBA is demonstrated by showing that the water level does not fall below the Top of Active Fuel (TAF), or the fuel cladding temperature does not exceed the cladding temperatures during normal operation. However, the results presented for LOCA in Table 15.7-3 only list peak cladding temperatures and do not provide information on water level. The water level results for each of cases listed should to be included.
15.5 -
Deterministic Safety Analyses 50.34(a)(1)(ii)(D) and RG 1.183.
There appears to be an inconsistency in the cited references to the RADTRAD versions used in the calculation of radiological consequences. Section 15.5.1.2.6 RADTRAD, states that The dose consequences of postulated design basis accidents are calculated using the RADTRAD Version 3.10 computer code (Reference 15.5-11). Reference 15.5-11 is listed as NUREG/CR-7220, SNAP/RADTRAD 3.10: Description of Models and Methods, June 2016. However, the title of NUREG/CR-7220 is SNAP/RADTRAD 4.0: Description of Models and Methods. Note: RADTRAD 3.10 was translated into JAVA from Fortran to provide the base program that became SNAP/RADTRAD. Section 15.5.9.1 Analysis of LOCA Outside Containment states that dose consequence analyses are performed
Readiness Assessment - Phase 5 8
using NUREG/CR-6604 RADTRAD Version 3.10. NUREG/CR-6604 describes the original RADTRAD code version 2.20 1997.
Clarification as to the version of RADTRAD used to compute design basis accident dose consequences should be provided.
Chapter/Section 15.6 - Probabilistic Safety Analysis Section Basis for Observation/Comment Readiness Assessment Observations/Comments Table 15.6 Probabilistic Safety Assessment Objectives Table 15.7 Core Damage Frequency Results Table 15.7 Large Release Frequency Results Section 15.6.9 -
Results of the Level 1 Probabilistic Safety Assessment, Section 15.6.9.1.4
- Large Release Frequencies, Section 15.6.10.1
- Summary 10 CFR 50.34, 50.2, 50.36, Appendix A, Severe Accident Policy Statement, Safety Goal Policy Statement, SRP Chapters 15 and 19.3 Overall Observations - The NRC staff identified major gaps in the content of information in the areas of key risk insights and quantitative results for the design, dominant accident sequences, design features that are major risk contributors, and quantified risk metrics.
If not addressed, these gaps have the potential to challenge an effective and efficient review of the CP application, including, but not limited to, reaching positive regulatory findings on (1) implementation of the Safety Strategy, particularly the uses of the PRA in the methodology; (2) achievement of the stated objectives of the PRA in Table 15.6-1 of the PSAR; (3) meeting the Commissions Safety Goals and, consequently, the assurance of no undue risk to public health and safety (objective 4 of the PRA in Table 15.6-1); (4) meeting the Commissions containment performance goals (probabilistic and deterministic) as described in SECY 90-016 and SECY 93-087 and associated SRMs; (5) conformance with the Commissions Severe Accident Policy Statement; (6) conformance with the Commissions direction on the Regulatory Treatment of Non-Safety Systems as described in SECY 94-084 and SECY 95-132 and associated SRMs; and (7) support findings against applicable regulations such as 10 CFR 50.34 (a)(1)(ii) (extremely low probability for accidents that could result in
Readiness Assessment - Phase 5 9
the release of significant quantities of radioactive fission products).
Specific examples of these gaps are identified below. If the preliminary nature of this information is the reason for the exclusion, appropriate qualifiers can be used in the PSAR by the applicant and in the SER by the NRC staff (e.g., preliminary).
Table 15.6 Probabilistic Safety Assessment Objectives As above A clarification on why Table 15.6-1 does not include using the PSA as described in the LTR NEDC-33934P, BWRX-300 Safety Strategy needs to be provided.
Table 15.6 Probabilistic Safety Assessment Objectives As above Objective 6 of Table 15.6-1 states that the PSA is used to identify facility vulnerabilities and systems for which design improvements or modifications to operational procedures could reduce the probabilities of severe accidents or mitigate their consequences.
With the understanding that procedures may not be developed at this stage, the draft PSAR lacks examples of design changes made based on risk information and insights to achieve the stated objective. Appendix 15B briefly discusses risk reduction included as Defense Line 4 functions for mitigating design extension conditions. The CP application needs to include clarification on whether this section is intended to represent risk reduction measures based on risk insights from the PSA and provide specific examples of the risk reduction achieved or whether these measures were based on deterministic analysis/defense in depth.
15.6.3.5 - Event Sequence Frequency Quantification 15.6.10.1 -
Summary As above The draft PSAR lacks a list, with summary description, of dominant sequences for CDF and LRF for the full-power internal events PRA model.
Readiness Assessment - Phase 5 10 15.6.7 -
Uncertainty and Sensitivity Analysis 15.6.7.2.7 - Task Outputs and Preliminary Results As above The draft PSAR section is lacking the insights from the full-power internal events PRA such as a list of key assumptions and sources of uncertainty, including design features and design assumptions, impacting the application and a list of sensitivity analyses performed to address the assumptions.
Table 15.7 Core Damage Frequency Results Table 15.7 Large Release Frequency Results As above There are no CDF and LRF results.
Furthermore, Table 15.7-9, CDF Results and Table 15.7-10, LERF Results, are intentionally left blank and state that the tables will be populated in the FSAR. The staff notes that Section 3.2.7.3.4 of the GEH Safety Strategy LTR states that a demonstration that BWRX-300 risk metricsis provided.
15.6.3.5 - Event Sequence Frequency Quantification 15.6.10.1 -
Summary As above The draft PSAR is lacking information of accident sequence analyses including (a) summaries of event trees for each initiating event identified in the initiating event analysis, including a discussion of the sequences for each event tree; (b) a description of necessary and sufficient equipment (safety and non-safety-related) reasonably expected to be used to mitigate initiators; and (c) a description of individual function mission times for each safety function and time windows for each operator action included in the PRA.
15.6.3.4.2 - Fault Trees, Data Analysis As above The draft PSAR is lacking information on data analysis including design-specific justification for the failure rates used for first-of-a-kind components.
15.6.3.4 - System Analysis (no dedicated section re: passive safety systems)
As above The draft PSAR is lacking information on passive safety system reliability for the full-power internal events PRA including:
(a) identification of all key thermal hydraulics parameters that could affect the reliability of a passive system and introduce uncertainty into the determination of success criteria and (b) accounted for the uncertainty in the analyses that establish the success criteria.
Readiness Assessment - Phase 5 11 15.6.3.4.3 -
Human Reliability Analysis As above Section 15.6.3.4.3 states, Due to the incompleteness of the design and the lack of procedures, this [PRA] analysis is not expected to meet all the requirements of the standard at this time. With the understanding that this is preliminary information and for a CP application all the requirements of the PRA standard are not expected to be met, the draft PSAR is lacking (a) the identification and description of HFEs that result in initiating events, (b) identification and description of pre-and post-accident HFEs that impact the mitigation of initiating events, (c) identification of any dependent HFEs, and (d) any recovery action credit taken, with justification.
15.6.3.6 - Internal Fire Hazard As above For a CP application, either an internal fire PRA or a non-PRA evaluation is recommended. The staff notes that the draft PSAR describes performing an internal fire PSA. Given that an internal fire PSA was performed, the draft PSAR lacks (a) a summary of changes made to the internal events PSA to develop the internal fire PSA addressing each of the topics identified previously for internal events PSA and (b) a description of the risk insights.
15.6.3.7 - Internal Flooding Hazard As above For a CP application, either an internal flood PRA or a non-PRA evaluation of the risk is recommended. Given that an internal flood PSA was performed, the draft PSAR lacks (a) a summary of changes made to the internal events PSA to develop the internal flood PSA addressing each of the topics identified previously for internal events PSA and (b) a description of the risk insights.
15.6.3.9 - Low Power and Shutdown Probabilistic Safety Assessment As above For a CP application, its recommended low power and shutdown events be evaluated using a PRA or non-PRA method. As described in PSAR Section 15.6.3.9, the applicant developed a Low-Power and Shutdown (LPSD) probabilistic safety assessment (PSA) using ANS/ASME-58.22-2014, the trial use Low Power and Shutdown PRA standard. The LPSD PSA also includes
Readiness Assessment - Phase 5 12 heavy load drops that can cause fuel damage, core damage, and large release. To develop a shutdown model, plant operating states (POSs) were defined in relation to decay heat and the availability of systems.
Given that a low power and shutdown PSA was performed, the draft PSAR lacks (a) a summary of the POS analysis, (b) a summary of the systematic identification of potential LPSD initiating events, (c) shutdown event trees, (d) identification of key assumptions used in the evaluation, and (e) the quantitative results and dominant quantified accident sequences.
15.6.4 - Level 2 Probabilistic Safety Assessment As above The Draft PSAR lacks the following information regarding Level 2 analysis including (a) event trees and key phenomena for Level 2 PRA and (b) demonstration that the design at the CP-stage meets the Commissions expectations for containment performance for new reactors.
15.6.1.1 -
Probabilistic Safety Assessment Scope 15.6.2 -
Probabilistic Safey Assessment Overview 15.6.2.3 - Internal and External Events and Level 1 Probabilistic Safety Assessment As above Section 15.6.2.3, Internal and External Events and Level 1 Probabilistic Safety Assessment, states Internal and external events and Level 1 PSA also include: Self-assessment results using NEI 17-01 guidance... The draft PSAR does not include (a) a description of the self-assessment or peer review, and a summary of any limitations identified by the self-assessment arising from the level of maturity of design and operational details and (b) a description of the applicants plan for meeting each identified PRA element in the OL PRA. This information provides confidence to the staff that the applicant has the ability to develop and maintain the PRA post-CP approval to reflect the as-built plant.
15.6.1.1 -
Probabilistic Safety Assessment Scope As above The draft PSAR lacks a description of a PRA configuration control plan including (a) a description of the process to track assumptions and monitor inputs for PRA and non-PRA evaluations supporting the CP application; (b) a description of how new information will be collected and included in
Readiness Assessment - Phase 5 13 15.6.2 -
Probabilistic Safety Assessment Overview the PRA to maintain the PRA consistent with the as-built, as-to-be-operated plant design; and (c) a description of how reviews of the PRA will be conducted (i.e., self-assessment, peer review, etc.), including the frequency of such reviews.
15.6.3.8.1 - High Wind Hazard 15.6.3.8.3 -
Seismic Hazard 15.6.3.8 -
Probabilistic Safety Assessment External Hazards As above For a CP application, its recommended external events (e.g., seismic, high wind, and external flood) be evaluated using a PRA or non-PRA method. The staff notes that the draft PSAR describes performing a seismic PRA and high wind PRA. Given that such PRAs were performed, the draft PSAR lacks (a) a summary of changes made to the internal events PRA to develop the seismic PRA and high wind PRA addressing each of the topics identified previously for internal events PSA and (b) a description of the risk insights.
15.6.3.7 - Internal Flooding Hazard Table 15.6 External Hazards Screening As above The draft PSAR describes that external flooding was excluded from evaluation.
Section 15.6.3.7 states that, external flooding events are reasonably precluded from the BWRX-300 probabilistic flood analysis based on adherence to the design conditions set forth in the envelope of BWRX-300 standard plant site parameters.
Table 15.6-3, External Hazards Screening, includes an entry Extreme Rain that addresses external flooding; however, the screening description for this entry is incomplete. The site-specific basis for screening of the external flooding should be provided in the draft PSAR.
15.6.3.8.1 - High Wind As above Section 15.6.3.8.1 discusses high wind risk evaluation and states that, BWRX-300-specific high wind event frequency and plant effects are applied... to obtain risk results.
In the same section, the applicant also states that, site-specific data are inputs for the external hazards PSA analyses. It is not clear to the staff whether the high wind PRA performed at the CP stage will be based on the BWRX-300 generic high wind data or the site-specific data at Clinch River.
Readiness Assessment - Phase 5 14 15.6.3.8.3 -
Seismic Hazard As above In parallel to the above question on the high wind PRA, it is not clear to the staff whether the seismic PRA at the CP stage will be based on the BWRX-300 generic seismic data or the site-specific data at Clinch River.
The staff recommends that the external hazards PRA at the CP stage be performed using site-specific hazard information when available, which will provide more realistic risk insights to the staff at an early stage of the licensing process.
15.7 - Safety Analyses Appendix 15A.4 -
Reference Source Term for Conditions that are Practically Eliminated Events Table 15A Practically Eliminated Conditions As above Table 15A-1, Practically Eliminated Conditions describes many conditions that are practically eliminated. For each practically eliminated condition, the table includes a description of risk reduction design features and design and supporting operating provisions that make each condition extremely unlikely. This section does not include any quantitative risk-information to support the assertion that the conditions are extremely unlikely. In addition, since during the design stage some of the risk-reduction features may change, the section does not describe any quality control measure to ensure changes to the design/operation would not alter the likelihood of these conditions such that their elimination status would change.
Readiness Assessment - Phase 5 15 Chapter/Section 15.7 - Results of Deterministic Safety Analyses and Probabilistic Safety Assessment Section Basis for Observation/Comment Readiness Assessment Observations/Comments 15.7 - Results of Deterministic Safety Analyses and Probabilistic Safety Assessment 50.34(a)(1)(ii)(D) and RG 1.183.
There appears to be a discrepancy between the atmospheric dispersion coefficients for the off-site locations in Section 15.5.8 on page 15-109 and the main control room envelope (MCRE) for a release from the ICS pools and for a Diffuse Source Release from the RB in Section 15.5.9 on pages15-117 and 15-119 respectively. For these release points the 0-2 hour MCRE atmospheric dispersion factor is lower than the 0-2 hour factor for the EAB. The MCR has no credited filtration. The MCR unfiltered in-leakage is artificially assumed to be 38,000 cfm so the control room airborne concentrations would be essentially the same as the outdoor concentrations. However, Table 15.7-4 indicates that the MCR FHA dose is lower than the EAB dose and roughly proportional to the difference in the X/Q values. Tables 15.7-5, 15.7-6, 15.7-7, and 15.7-8 indicate similar discrepancies The NRC staff does not see these discrepancies as being credible.
These discrepancies need to be reconciled.
15.7 - Results of Deterministic Safety Analyses and Probabilistic Safety Assessment 50.34(a)(1)(ii)(D) and RG 1.183.
The section of the PSAR that describes the deterministic evaluation of the design basis substantial core melt accident needed to satisfy the requirements of 50.34(a)(1)(ii)(D) needs to be identified.
TVA Readiness Assessment Ch 15 Comments
- 1. Chapter 15 of the PSAR states that "BWRX-300 Safety Analyses incorporates selected guidance from the International Atomic Energy Agency (IAEA) Safety Standards Specific Safety Requirements No. SSR-2/1, Revision 1, Safety of Nuclear Power Plants:
Design, (Reference 15.1-2) and formed from the safety strategy discussed in Licensing Topical Report (LTR) NEDC-33934P, Revision), BWRX-300 Safety Strategy, (Reference 15.1-3)." The NRC staff notes that the review of the subject PSAR will be performed under 10 CFR Part 50 framework and that the NRC does not endorse IAEA standard listed by the applicant for the PSAR review.
- 2. Chapter 15 of the PSAR states that The BWRX-300 Safety Analyses is formed from the safety strategy discussed in Licensing Topical Report (LTR) NEDC-33934P, Revision 0, BWRX-300 Safety Strategy. The LTR is currently under NRC review. The NRC staff has provided GEH with observations on the LTR as part of the pre-application readiness assessment. Since the safety analyses performed in Chapter 15 of the PSAR are based on the Safety Strategy LTR framework, the NRC staff notes that the observations provided on the LTR are applicable here as well. Some of the observations are reiterated here for the specific sections.
- 3. All the LTRs incorporated by reference as part of the application should be tabulated in the PSAR