ML24107B098
| ML24107B098 | |
| Person / Time | |
|---|---|
| Site: | 99902056 |
| Issue date: | 04/16/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL |
| References | |
| Download: ML24107B098 (8) | |
Text
From:
Greg Cranston Sent:
Tuesday, April 16, 2024 1:35 PM To:
Schiele, Raymond Joseph; Klein, Spencer Paul; Montague, Kelvin Jevon Cc:
Shanlai Lu; Zhian Li; Rosie Sugrue; Dan Widrevitz; Allen Fetter; Michelle Hayes; Sean Gallagher; ClinchRiver-CPSafDocsPEm Resource
Subject:
Readiness Assessment Observations on draft PSAR Chapter/Sections 4.0, 4.2, 4.3, 4.4, 4.5,mand 4.6 (Phase 2) - draft Clinch River CP application Attachments:
Initial Observations PSAR 4 Readiness Assessment.pdf Good afternoon, Attached are initial NRC staff observations on draft PSAR Chapter/Sections 4.0, Reactor; 4.2, Fuel System Design; 4.3, Design Basis; 4.4 Thermal and Hydraulic Design; 4.5 Reactor Materials; and 4.6, Functional Design of Reactivity Control Systems, that staff viewed in TVAs electronic reading room as part of Phase 2 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 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 me, Allen Fetter or Sean Gallagher.
- Thanks, Greg
Hearing Identifier:
ClinchRiver_CPSafDocs_Public Email Number:
18 Mail Envelope Properties (DM6PR09MB5271EEE39456220D5CFA9AEC90082)
Subject:
Readiness Assessment Observations on draft PSAR ChapterSections 4.0, 4.2, 4.3, 4.4, 4.5,mand 4.6 (Phase 2) - draft Clinch River CP application Sent Date:
4/16/2024 1:35:21 PM Received Date:
4/16/2024 1:35:25 PM From:
Greg Cranston Created By:
Gregory.Cranston@nrc.gov Recipients:
"Shanlai Lu" <Shanlai.Lu@nrc.gov>
Tracking Status: None "Zhian Li" <Zhian.Li@nrc.gov>
Tracking Status: None "Rosie Sugrue" <Rosemary.Sugrue@nrc.gov>
Tracking Status: None "Dan Widrevitz" <Dan.Widrevitz@nrc.gov>
Tracking Status: None "Allen Fetter" <Allen.Fetter@nrc.gov>
Tracking Status: None "Michelle Hayes" <Michelle.Hayes@nrc.gov>
Tracking Status: None "Sean Gallagher" <Sean.Gallagher@nrc.gov>
Tracking Status: None "ClinchRiver-CPSafDocsPEm Resource" <ClinchRiver-CPSafDocsPEm.Resource@nrc.gov>
Tracking Status: None "Schiele, Raymond Joseph" <rjschiele@tva.gov>
Tracking Status: None "Klein, Spencer Paul" <spklein@tva.gov>
Tracking Status: None "Montague, Kelvin Jevon" <kjmontague@tva.gov>
Tracking Status: None Post Office:
DM6PR09MB5271.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 949 4/16/2024 1:35:25 PM Initial Observations PSAR 4 Readiness Assessment.pdf 201712 Options Priority:
Normal Return Notification:
No Reply Requested:
No Sensitivity:
Normal Expiration Date:
Readiness Assessment - Phase 2 Chapter 4, Reactori Clinch River CP application Revision 3/20/24 Chapter/Section 4 - Reactor Design Section Basis for Observation/Comment Readiness Assessment Observations/Comments 4.0 - Reactor 10 CFR 50.46; 10 CFR Part 50 Appendix A, GDC 10 & GDC 13; 10 CFR 50.61; 10 CFR 50.61a; 10 CFR 50 Appendix H The following information should be provided:
The nuclear instrumentation description needs to include the detector types at the different power ranges.
4.2 - Fuel System Design Chapter/
Section Basis for Observation/Comment Readiness Assessment Observations/Comments 4.2 - Fuel
System Design
10 CR 50.46; 10 CFR Part 50 Appendix A, GDC 10 & GDC 27 The following information will need to be provided. This information may be contained in references 4.2-9 and 4.2-10, which were not available for staff review as part of the readiness assessment. The following information needs to be provided as part of reports incorporated by reference into the CPA, or directly in Chapter 4 of the PSAR:
Justifications regarding the applicability of the fuel performance evaluation codes cited.
A detailed fuel performance evaluation.
A demonstration of compliance with SAFDLs, such as rod internal pressure at the end of fuel assembly life.
A cladding fatigue analysis for the BWRX-300 core.
Applicability of the referenced control rod design evaluation to the BWRX-300 design.
Readiness Assessment - Phase 2 Chapter/Section - 4.3 - Design Basis Chapter/
Section Basis for Observation/Comment Readiness Assessment Observations/Comments 4.3 - Design Basis 10 CFR 50.34; 10 CFR 50.46; 10 CFR Part 50 Appendix A,
GDC 10 Nuclear computer code benchmarking information for the BWRX-300 design should be provided. There is a reference 4.3-2 for TGBLA06/PANAC11 applicability to BWRX-300, however this was not made available for the readiness assessment. The staff expects that this reference may contain the missing information. The benchmarking analyses, or applicability of benchmarking to the BWRX-300 design, are needed for the NRC-approved nuclear analysis computer codes to determine the biases and uncertainties of the codes when applied to analyzing the BWRX-300 reactor because of the differences in designs and working principle. This information may be in reports incorporated by reference into the CPA or be included as a part of Chapter 4 of the PSAR.
4.3 - Design Basis 10 CFR 50.34; 10 CFR 50.46; 10 CFR Part 50 Appendix A,
GDC 10 The application is missing information regarding the core design. There is a reference 4.3-1 to NEDC-34044P, BWRX-300 GNF2 Equilibrium 12-month cycle nuclear design report, Rev. 0, November 2023, which the staff expects may include the missing information, but was not included as part of the readiness assessment. The following information needs to be provided (either as part of a document incorporated by reference into the CPA, or as part of Chapter 4 of the PSAR):
Fuel bundle and core enrichments and burnable poison loadings for the initial cycle or the equilibrium cycle, at a minimum.
Integrated and differential control rod worth information.
Key nuclear design parameters such as the effective delay neutron fraction eff, average prompt neutron lifetime, Doppler reactivity coefficients (max and min).
Key core performance parameters at BOC, middle of cycle (MOC), and end of cycle (EOC) as a function of power level based on a nominal cycle depletion.
Readiness Assessment - Phase 2 Chapter/Section - 4.4 - Thermal and Hydraulic Design Chapter/Section Basis for Observation/Comment Readiness Assessment Observations/Comments 4.4 - Thermal and Hydraulic Design 10 CFR 50.34 The PSAR discusses the methodologies the applicant intends to use for thermal and hydraulic design of the BWRX-300 reactor. However, there is no information on the calculations for the thermal and hydraulic design. Missing information for thermal analyses should be provided and include descriptions of the models, the input parameters, the values used in the calculations and the results of the calculations.
Some of this information may be included in references 4B-1 and 4.4-3, however these were not available for staff review for the readiness assessment.
Readiness Assessment - Phase 2 Chapter/Section - 4.5 - Reactor Materials Chapter/Section Basis for Observation/Comment Readiness Assessment Observations/Comments 4.5 - Reactor Materials NUREG-0800, SRP 4.5.1 & 4.5.2.
ASME Code Sections II
& III Recent international experience corroborates that modern low sulfur 3XX steels may be more susceptible to stress corrosion cracking than generally acknowledged. As noted in SRP 4.5.2 Section 1, under Materials, the adequacy and suitability of the materials specified are reviewed for, among several aspects, stress corrosion resistance. Discussion of how the potential for modern low sulfur 3XX steels relatively larger susceptibility to stress corrosion cracking (relative to historical 3XX steel supplies) will assist in addressing this review area.
NOTE: A big part of review in 4.5.1 and 4.5.2 is material suitability for the operating environment.
The recent stress corrosion cracking in low sulfur steels in France is not an isolated event and mentioning it as an observation/comment flows from the review guidance in SRPs 4.5. The ASME Code was developed when steel impurity control was less stringent, so sulfur content is often not mentioned in Section 2, especially for older specs.
3XX steels are in both the CRD and internals sections.
4.5 - Reactor Materials NUREG-0800, SRP 4.5.1 & 4.5.2.
Sections 4.5.1 and 4.5.3 reviewed do not include diagrams of the reactor vessel internal components and core supports (diagrams may exist elsewhere in the application). These diagrams will be important in enabling the staff to make findings for Sections 4.5.1 and 4.5.3. and should be included in the CPA so staff can better understand the general shape, design, and operating environment of all the subject components, especially those listed in Tables 4.5-1 and -2.
NOTE: Such diagrams need not be dimensionally precise or fully accurate but are necessary for staff to confirm their understanding of the subject components, interfaces between these components, and operating environment of subject components in order to make findings under the general design criteria and as described in SRP 4.5.1 and SRP 4.5.2. Sole reliance on textual descriptions has not proven sufficient in
Readiness Assessment - Phase 2 previous reviews for these sections due to the complexity of the subject components.
4.5 - Reactor Materials NUREG-0800, SRP 4.5.1, 4.5.2 Sections 4.5.1 and 4.5.3 reviewed do not include diagrams of the reactor vessel internal components and core supports (diagrams may exist elsewhere in the application). These diagrams will be important in enabling the staff to make findings for Sections 4.5.1 and 4.5.3. and should be included in the CPA so staff can better understand the general shape, design, and operating environment of all the subject components, especially those listed in Tables 4.5-1 and -2.
NOTE: Such diagrams need not be dimensionally precise or fully accurate but are necessary for staff to confirm their understanding of the subject components, interfaces between these components, and operating environment of subject components in order to make findings under the general design criteria and as described in SRP 4.5.1 and SRP 4.5.2. Sole reliance on textual descriptions has not proven sufficient in previous reviews for these sections due to the complexity of the subject components.
4.5 - Reactor Materials 10 CFR Part 50.55a NUREG-0800, SRP 4.5.1 & 4.5.2.
ASME Code Sections II
& III Section 4.5.3.1 does not explicitly cite ASME Code Section II but indicates that materials listed in Table 4.5-1 are merely representative. This appears inconsistent with SRP 4.5.2 which relies on the use of ASME Code citations to underpin NRC review (and for incorporation by reference of ASME Code requirements in 10 CFR 50.55a). There description given is not sufficient for NRC staff to draw conclusions regarding this in Section 4.5.3.1.
4.5 - Reactor Materials RG 1.171 ASME Code Sections IX Section 4.5.3.2 indicates that welder and welding procedures are qualified according to ASME Code,Section IX, in conformance with RG 1.71. It is unclear whether welding procedures are qualified according to ASME Code,Section IX, when used for areas without limited accessibility (the topic of RG 1.71). Clarify extend of conformance with ASME Code,Section IX.
4.5 - Reactor Materials 10 CFR Part 50 Appendix A, GDC 1 10 CFR Part 50 Appendix A Section 4.5.3.2 is generally lacking in detail and specificity. High-level statements of conformance with 10 CFR 50 Appendix A and 10 CFR 50 Appendix B leave little basis for staff to make findings.
4.5 - Reactor Materials ASME Code Sections III, sub-articles NB-2500 and NB-5000 Section 4.5.3.3 cites ASME Code,Section III, sub-articles NB-2500 and NB-5000. It is unclear which components falling under Section 4.5.3 would be subject to the requirements of NB sub-articles.
Readiness Assessment - Phase 2 4.5 - Reactor Materials Reg Guide 1.44 Section 4.5.3.4 makes no mention of any testing for sensitization (e.g. use of RG 1.44). It is unclear whether this is due to an intent to select materials for which testing is not necessary. Clarifying specifically how material selection and control precludes the need for testing; amending the section to include testing for sensitization; or otherwise providing a complete description of how sensitization will be avoided and the lack-thereof verified is necessary to support the staff making a finding.
4.5 - Reactor Materials NUREG-0800, SRP 4.5.1 & 4.5.2.
Table 4.5-1 only provides SFA numbers for weld materials, and Table 4.5-2 has no weld materials listed. This level of detail does not support review under SRP 4.5.1 or SRP 4.5.2. Application should include a description of which welding specifications will be used, and whether there are or are not welds in the control rod drive components and materials.
Chapter/Section - 4.6 - Functional Design of Reactivity Control Systems Chapter/Section Basis for Observation/Comment Readiness Assessment Observations/Comments 4.6 - Functional Design of Reactivity Control Systems 10 CFR Part 50 Appendix A, GDC 26 10 CFR 50.34 Exemption to GDC-26 is intended but not submitted yet and will be reviewed when received.
i Input provided by NRR/DEX/EICB; /DSS/SNRB; /DNRL/NVIB