ML25275A438
| ML25275A438 | |
| Person / Time | |
|---|---|
| Site: | 05000615 |
| Issue date: | 10/01/2025 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML25275A435 | List: |
| References | |
| NNP-25-009 | |
| Download: ML25275A438 (1) | |
Text
U.S. Nuclear Regulatory Commission NNP-25-009 October 1, 2025 1
ENCLOSURE 2 CRN Site CPA, NRC Audit Questions and TVA Responses (Public Version)
U.S. Nuclear Regulatory Commission NNP-25-009 October 1, 2025 1
The following NRC Audit Questions Responses and PSAR markups are provided in this Enclosure. The corresponding CPA revisions are included in Enclosure 3 (Public Version) and Enclosure 4 (Non-Public Version).
Revision Type Identifier Description PSAR Section(s)/
Revised Pages(1)
NRC Audit Question A-3.9-1 PSAR Section 3.9.2.1 should reference ASME Operation and Maintenance (OM)
Code Part 3 and Part 7.
3.9, pp. 3.9-11 NRC Audit Question A-4.3-2 Provide a justification for the approach stipulated in PSAR Section 4.3.1.2; Control Requirements (Shutdown Margins) is correct, which addresses, at minimum, control rod drive system design and the relevant safety analysis.
3.1, pp. 3.1-19, 3.1-20 4.3, pp. 4.3-1, 4.3-4, 4.3-6 15.3, pp. 15.3-6 NRC Audit Question A-4.5-1 Provide how the delta ferrite is controlled and measured for austenitic stainless steel materials for the CRD system.
4.5, pp. 4.5-5 NRC Audit Question A-4.5-2 Provide the maximum carbon content and the heat treatment (i.e., solution annealing) for the austenitic stainless steel material taking into account the influence carbon content and heat treatment has on the susceptibility to stress corrosion cracking in stainless steel for the CRD system.
4.5, pp. 4.5-2, 4.5-4, 4.5-9, 4.5-11 5.2, pp. 5.2-38, 5.2-39 NRC Audit Question A-4A-1 Consider revising the PSAR to appropriately reference and adequately describe, outline, or summarize the NEDC-34270 methodology in Section 4A.6 as a preliminary methodology for the PSAR.
4A, pp. 4A-2, 4A-3, 4A-4 NRC Audit Question A-5.2.3-1 Provide the filler metal classification for each of the ASME specifications listed in Table 5.2-2 to be used (e.g.,
Nickel-based alloy ASME SFA-5.14, ERNiCr-3).
5.2, pp. 5.2-36 through 5.2-39 5.3, pp. 5.3-2, 5.3-9 NRC Audit Question A-5.2.3-2 Provide the maximum carbon content for the stainless steel weld filler metal taking into account the influence carbon content has on the susceptibility to stress corrosion cracking in stainless steel.
5.2, pp. 5.2-36 through 5.2-39 NRC Audit Question A-5.2.3-3 Provide the material types that could be used for the isolation condenser system condensate return valves.
5.2, pp. 5.2-33, 5.2-34, 5.2-37
U.S. Nuclear Regulatory Commission NNP-25-009 October 1, 2025 2
Revision Type Identifier Description PSAR Section(s)/
Revised Pages(1)
NRC Audit Question A-5.2-4 Confirm that RCPB components will meet accessibility requirements and Preservice Inspection requirements as required by 10 CFR 50.55a(g)(2),
Accessibility requirements, and 10 CFR 50.55a(g)(3), Preservice examination requirements.
5.2, pp. 5.2-18, 5.2-19 NRC Audit Question A-8.1-1 PSAR sections 8.1.3 and 8.2.1 do not discuss how the offsite power system conforms with NUREG-0800, Revision 5, section 8.2, criteria 1, 2, and 5 and GDC 2, 4 and 18.
1.9, pp. 1.9-69, 1.9-70 NRC Audit Question A-8.1-2 Provide the safety classification of the offsite power system consistent with the descriptions in Section 3.2.2.
3A, pp. 3A-10 8.1, pp. 8.1-3 NRC Audit Question A-12.2-1 Update the PSAR to provide the maximum radionuclide inventory of the reactor core, considering long term operation of the facility. Also update the PSAR to discuss the methods, models, and assumptions for developing the source term.
12.2, pp. 12.2-1 NRC Audit Question A-13.7-1 PSAR Table 13.4-1 should use the correct milestones descriptions in NEI 06-06 as endorsed by the NRC in RG 5.84.
13.4, pp. 13.4-5, 13.4-6, 13.4-7 NRC Audit Question A-13.7-2 PSAR Section 13.7 will be modified to make explicit references to 10 CFR 26.4(e) and (g) individuals subject to FFD program.
13.7, pp. 13.7-1 NRC Audit Question A-17.1-1 Explain why Section 17.1 of the PSAR incorporates by reference the QA Program used for BWRX-300 GEH design activities documented in topical report NEDO-11209-A.
17.1, pp. 17.1-1 through 17.1-4 17.4, pp. 17.4-5 NRC Audit Question A-17.5-1 Please submit a revised version of the NNP-TR-001-NP-A which includes the staff's safety evaluation.
1.6, pp 1.6-1 17.5, pp. 17.5-3 NRC Audit Question A-17.5-2 Discuss why the description of Senior Vice Present, Engineering and Operations Support as provided in the TVA fleet QAPD does not fully align with the description provided in the PSAR or the New Nuclear QAPD.
17.5, pp. 17.5-1
U.S. Nuclear Regulatory Commission NNP-25-009 October 1, 2025 3
Revision Type Identifier Description PSAR Section(s)/
Revised Pages(1)
NRC Audit Question A-17.5-3 Explain why RG 1.33 is referenced in the CPA.
1.9, pp. 1.9-152 TVA-identified Change LCR 25-044 Update of PSAR Chapter 2 to correct inconsistencies with the Courtesy Copy of Chapter 2 posted in eRR.
2.1, pp. multiple 2.2, pp. multiple 2.3, pp. multiple 2.4, pp. multiple 2.5, pp. multiple TVA-identified Change LCR 25-048 Update of PSAR Section 13.3 to correct inconsistencies with the Courtesy Copy of Chapter 13 posted in eRR.
13.3, pp. 13.3-1, 13.3-2, 13.3-4.
13.3-5, 13.3-7, 13.3-9 TVA-identified Change LCR 25-051 Materials are updated in Tables 4.5-1, 4.5-2, 5.2-2a, and 5.2-3. Some of these new material specifications have been transmitted to the NRC by Audit Questions 5.2.3-1 and 5.2.3-3. This change updates the remaining tables entries to the current design.
4.5, pp. 4.5-4, 4.5-8 through 4.5-11 5.2, pp. 5.2-29 through 5.2-37, 5.2-40 TVA-identified Change LCR 25-052 Editorial Corrections to PSAR 3.7, pp. 3.7-29, 3.7-35 3.9, pp. 3.9-14 4.3, pp. 4.3-6 15.5, pp. 15.5-65 Note:
(1) The page numbers reflect the page numbers in the revised PSAR included in (Public Version) and Enclosure 4 (Non-Public Version).
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-3.9 Receipt Date:
8/25/2025 Question: The NRC staff provided readiness assessment observation that PSAR Section 3.9.2.1 should reference ASME Operation and Maintenance (OM) Code Part 3, Vibration Testing of Piping Systems and Part 7, Thermal Expansion Testing of Nuclear Power Plant Piping Systems; however, this feedback was not addressed in the PSAR. SRP Section 3.9.2.1 is regarding piping vibration and thermal expansion testing, and lists OM Code Part 3 and Part 7 as acceptance methods. PSAR, Table 1.9-3, Conformance with NUREG-0800 (Chapter 3 Design of Structures, Components, Equipment, and Systems) lists that PSAR Section 3.9.2 conforms to SRP Section 3.9.2. PSAR Section 3.9.2.1, Flow Induced Vibration, Thermal Expansion, and Dynamic Effects states: Piping vibration, thermal expansion, and dynamic effects testing is performed in accordance with the startup administrative manual as described in Chapter 14 and follows guidance from ASME OM Code, Nonmandatory Appendix M. The NRC staff notes that ASME OM Code, Nonmandatory Appendix M, Design Guidance for Nuclear Power Plant Systems and Component Testing is related to design of pumps, valves, and dynamic restraints to support preservice and inservice testing.
Response
PSAR Section 3.9.2.1 will be revised to conform with SRP 3.9.2,Section II, SRP Acceptance Criteria, Criterion 1. References to ASME OM Code Nonmandatory Appendix M will be removed. New references for OM Code Parts 3 and 7 will be added.
CPA Update on Docket: A markup of PSAR Section 3.9.2.1 is provided. This markup will be used as the basis for a future docketed PSAR update.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3.9-11 Revision 0 Additional Information on GDC compliance is found in Section 3.1.
10 CFR 50, Appendix B, as it relates to design quality control, is discussed in Section 17.1.
10 CFR 50, Appendix S, as it relates to the suitability of the plant design bases for mechanical components established in consideration of site seismic characteristics, is discussed in Section 3.7.
3.9.2.1 Flow Induced Vibration, Thermal Expansion, and Dynamic Effects Piping vibration, thermal expansion, and dynamic effects testing is performed in accordance with the startup administrative manual as described in Chapter 14 and follows guidance from ASME OM Code, Nonmandatory Appendix M (Reference 3.9-11).
3.9.2.2 Seismic Qualification of SC1 Mechanical Equipment (Including Other Reactor Building Vibration Induced Loads)
This subsection discusses the testing and analytical qualification of the SC1 mechanical equipment, and other ASME BPVC,Section III equipment including equipment supports.
3.9.2.2.1 Tests and Analysis Criteria and Methods The testing and analytical criteria for the seismic qualification of SC1 mechanical equipment are discussed in Section 3.10.
3.9.2.3 Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady-State Conditions The reactor internal components within the vessel are subjected to extensive dynamic system analysis and testing to evaluate the dynamic response of the components during normal reactor operation (i.e. steady state conditions) and operational flow transients. The main concern for steady state dynamic response is to ensure that flow-induced vibration (FIV) experienced during normal operation will not cause structural failure or degradation. FIV analysis is performed as explained in the following paragraph. Component loads due to operational flow transients and non-normal operational transient conditions are performed by analysis, as described in Section 3.9.2.5.
As a FIV scoping activity, reactor internal components are evaluated by analytical calculations to identify any components that may require design changes. Vibration levels of each component are scaled based on legacy BWR plant measurements as detailed below. Analytical calculations are performed as stated below and follow the four main FIV mechanisms that occur in BWRs:
Turbulence forcing function scaling was performed on existing (legacy) BWR component measured strains, flows and frequencies. The BWRX-300 component strain (and stress) is calculated from scaling the legacy plant conditions and component natural frequencies to the BWRX-300 conditions and component natural frequencies. Turbulence forcing function scaling uses the fundamental flow physics laws:
Division 2, Part 3, "Vibration Testing of Piping Systems" and Part 7, "Thermal Expansion of Nuclear Power Plant Piping Systems"
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-4.3-2 Receipt Date:
8/11/2025 Question:
Several statements in PSAR 4.3 concerning shutdown margin stuck rod assumptions appear to conflict. PSAR Section 4.3.1.2, Control Requirements (Shutdown Margins) states that The core must be capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn or rod pair associated with the common Hydraulic Control Unit (HCU) postulated stuck in the full out position, with all other rods fully inserted. It goes on to say The Shutdown Margin (SDM) is determined by using the 3D BWR core simulator code PANAC11 to calculate the core multiplication at selected exposure points with the strongest control rod fully withdrawn. PSAR Section 4.3.4.3.1, Core Reactivity Effects, states The core is capable of being made subcritical with the highest worth control rod fully withdrawn. PSAR Section 4.3.4.4, Shutdown Margins, states The core must be capable of being made subcritical with margin in the most reactive condition throughout an operating cycle with the most reactive control rod pair in their full out position and all other control rods fully inserted.
Please clarify the requirements placed on the control rod reactivity in order to ensure that the BWRX-300 design is consistent with GDC 26 and GDC 27, as well as the evaluations performed to ensure that these requirements are met. If some or all shutdown margin evaluations do not assume that the most limiting condition of either (1) the most reactive rod or (2) the most reactive rod pair associated with a common HCU is stuck at the fully withdrawn position, provide a justification for this approach which addresses, at minimum, control rod drive system design and the relevant safety analysis. As necessary, please provide PSAR markup to clarify BWRX-300 design bases.
Response
The requirement stipulated in PSAR Section 4.3.1.2; Control Requirements (Shutdown Margins) is correct: The core is capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn or rod pair associated with the coupled Hydraulic Control Unit (HCU) postulated stuck in the full out position, with all other rods fully inserted. As background, for GE BWRs equipped with Fine Motion Control Rod Drives (FMCRDs), most Hydraulic Control Units (HCUs) that actuate during a scram serve two (2) control rods. As such, the cold shutdown margins scenario postulates that both control rods assigned to a coupled HCU may be inoperable and fully withdrawn. The assignment of control rod pairs to HCU is made such that there is significant physical separation between control rod locations assigned to a coupled HCU resulting in little, or no change in reactivity should the highest worth rod be withdrawn or, in addition, its HCU partner (i.e., the cold shutdown margins is essentially the same whether the highest worth control rod or the control rod pair is fully withdrawn). The cold shutdown margins values are consistent with the highest worth HCU coupled control rod pair fully withdrawn. The language in the PSAR that references the cold shutdown margins scenario will be clarified for consistency.
CPA Update on Docket: Markups of CRN-1 PSAR Chapters 3, 4, and 15 are provided to add clarity to the language that references the cold shutdown margins scenario and to ensure consistency with how the PSAR addresses this topic.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3.1-19 Revision 0 Reactivity control methods utilized during normal operation and AOOs employ control rod assemblies. The CRD system is designed such that no single active failure of a CRD component or supporting system function will prevent a reactor scram. Positive insertion means of controlling reactivity include: 1) the hydraulic scram control rod insertion function using the Hydraulic Control Units (HCUs) and control rods, and 2) the motor-driven control rod positioning function for normal power maneuvering which includes a separate motor-driven control rod run-in insertion function using the Fine Motion Control Rod Drives (FMCRDs) and control rods.
The CRD system is capable of maintaining the reactor core subcritical under cold conditions, even if the highest worth pair of control rods associated with an HCU is assumed to be in the fully withdrawn position. This shutdown capability of the CRD system is made possible by designing the core configuration to ensure the reactor core is held subcritical under cold conditions.
The means of normal reactivity control use a common set of control rods; however, the hydraulic scram function is not precluded by the operation or malfunction of the motor-driven FMCRD control rod positioning or run-in design features. The FMCRDs motor-driven function is independent from the hydraulic-driven function such that the use of a common set of control rods to control normal power operations does not introduce any additional challenges or failure mechanisms to the scram function. The FMCRDs and associated control rods are located separately and physically independent in the reactor core (i.e., each control rod is physically separated from each other by fuel assemblies). There are no identified mechanical interactions between control rods that would affect more than one control rod pair at a time.
The Boron Injection System (BIS) provided in the BWRX-300 is used as a Defense-in-Depth measure to reduce reactor power to achieve and then maintain the reactor core subcritical under cold conditions, even if all control rods remain fully withdrawn. The BIS assures reactor shutdown by mixing a neutron absorber with the reactor coolant.
Accordingly, the requirements of GDC 26 are met. Additional information discussing the compliance approach to GDC 26 is described in the applicable PSAR sections listed in Table 3.1-1.
3.1.3.8 GDC 27 - Combined Reactivity Control Systems Capability GDC 27 Statement The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.
Evaluation Against GDC 27 The control rods are capable of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods, the capability to shut down the reactor (and cool the core) is maintained. Positive automatic insertion of the control rods (i.e., scram) is provided by the HCU accumulators in response to design basis accidents
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3.1-20 Revision 0 (DBAs) that require emergency core cooling system operation (i.e., initiation of reactor isolation and the isolation condenser system (ICS)). Successful scram and initiation of reactor isolation and the ICS is shown to both achieve and maintain safe shutdown of the reactor without the possibility of a return to criticality, even with the unlikely occurrence of two stuck rods or a single failure of an HCU accumulator during scram. GDC 27 is satisfied for the BWRX-300 without the injection of soluble boron.
Accordingly, the requirements of 10 CFR 50, Appendix A, GDC 27 are met.
Additional information discussing the compliance approach to GDC 27 is described in the applicable PSAR sections listed in Table 3.1-1.
3.1.3.9 GDC 28 - Reactivity Limits GDC 28 Statement The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold-water addition.
Evaluation Against GDC 28 The combined features of the CRD system and the rod control system incorporate appropriate limits on the potential amount and rate of reactivity increase. The fine motion capability of the FMCRD allows reactivity additions from rod withdrawal to be finely controlled by the rod control system. Normal rod movement and the rod withdrawal rate are limited by the rod control system and follow sequences assigned by the core designers to operate the core within thermal limits.
The rod control system controls rod patterns and provides control rod blocks to limit the rate and amount of reactivity addition for control rod movement. The BWRX-300 FMCRD and control rod guide tube designs prevent rod drop and rod ejection events through positive design means.
Control rod drop is prevented by use of a bayonet style coupling between the FMCRD and the control rod, CRD mechanism latches that maintain the control rod inserted after a scram, and FMCRD separation switches that detect a control rod that has become separated from the FMCRD ball nut. Upon detection of separation from the ball nut, the rod control system prevents further outward rod motion of the ball nut by enforcing a withdrawal block. Control rod ejection is prevented by physical constraints including a flange at the top of the control rod guide tube that is above the core plate to prevent guide tube ejection and the FMCRD connection to the control rod guide tube via a bayonet coupling. The FMCRD includes an electro-mechanical brake that further prevents inadvertent rod withdrawal by engaging to prevent rod motion if the control system is not commanding rod motion. For the case of a scram line break, the FMCRD includes an internal ball check valve which reduces the chances of rapid rod withdrawal. The ball check valve functions as SC1 component because it serves as a RCPB function and prevents reverse
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.3-1 Revision 0 4.3 Nuclear Design This section describes the design bases and functional requirements used in the nuclear design of the fuel, core, and reactivity control system and relates these design bases.
4.3.1 Nuclear Design Bases This section describes the design bases and functional requirements used in the nuclear design of the fuel, core, and reactivity control system.
4.3.1.1 Reactivity Feedback Reactivity coefficients representing the differential changes in reactivity produced by differential changes in core conditions are useful for calculating stability parameters and evaluating the response of the core to external disturbances. The base initial condition of the reactivity control system and the Postulated Initiating Event (PIE) determine which of the several defined coefficients are significant in evaluating the response of the reactor.
There are two primary reactivity coefficients that characterize the dynamic behavior of BWRs; these are the Doppler reactivity coefficient and the moderator void reactivity coefficient. Also associated with the BWR are a power reactivity coefficient and a temperature coefficient. The power coefficient is a combination of the Doppler and moderator void reactivity coefficients in the power operating range, and the temperature coefficient is merely a combination of the Doppler and moderator temperature coefficients. Power and temperature coefficients are not specifically calculated for reload cores.
The fuel reactivity acceptance criteria are established in NEDE-24011-P-A-31-US, General Electric Standard Application for Reactor Fuel (GESTAR II) (Supplement for United States),
(Reference 4.3-1).
4.3.1.2 Control Requirements (Shutdown Margins)
The core must be capable of being made subcritical, with margin, in the most reactive condition throughout the operating cycle with the most reactive control rod fully withdrawn or rod pair associated with the common Hydraulic Control Unit (HCU) postulated stuck in the full out position, and all other rods fully inserted. The Shutdown Margin (SDM) is determined by using the 3D BWR core simulator code PANAC11 to calculate the core multiplication at selected exposure points with the strongest control rod fully withdrawn. The SDM is calculated based on the carryover of the minimum expected exposure at the end of the previous cycle. The core is assumed to be in the cold, xenon-free condition to produce conservative calculated values.
4.3.1.3 Control Requirements (Overpower Basis)
Technical Specification limits on Minimum Critical Power Ratio (MCPR) and the LHGR are determined such that the fuel will not exceed required licensing limits during abnormal operational occurrences or accidents. Explicit Maximum LHGR Limit and MCPR parameter definitions are provided below.
or HCU coupled control rod pair
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.3-4 Revision 0 4.3.4.2 Reactivity Coefficients The reactivity coefficients evaluation is performed as part of new fuel design development ensuring compliance with USNRC fuel licensing acceptance criteria. The GNF2 results are documented in NEDC-33270P, GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II), (Reference 4.3-5), which concludes that the criteria defined in GESTAR II have been met for the GNF2 fuel design.
4.3.4.3 Reactivity Variation The excess reactivity designed into the core is controlled by the CRD system supplemented by gadolinia fuel rods. Control rods are used during the cycle partly to compensate for burnup and partly to control the power distribution. The hot excess reactivity and associated control rod patterns for the reference BWRX-300 equilibrium cycle are documented in NEDC-34044P (Reference 4.3-2).
4.3.4.3.1 Core Reactivity Effects This section discusses the different effects associated with reactivity.
Control Reactivity The core is capable of being made subcritical at any time or at any core condition with the highest worth control rod fully withdrawn.
The CRD system is designed to provide adequate control of the maximum excess reactivity anticipated during the plant operation. The shutdown capability is evaluated assuming a cold, xenon-free core.
The excess reactivity designed into the core is controlled by the CRD system supplemented by gadolinia fuel rods. Control rods are used during the cycle partly to compensate for burnup and partly to control the power distribution.
Control rods are the primary means of achieving shutdown in normal operations, AOOs, PIEs, and beyond design basis events and severe accident scenarios. During power operation, changes in core reactivity are controlled by movement and positioning of the control rods within the core, in fine increments, using FMCRD electric motors. Reactivity feedback mechanism is key for some PIE groups (see Table 15.2-1) in which rods are withdrawn in error.
Doppler Reactivity Coefficient The Doppler reactivity coefficient is of high importance in reactor safety. The Doppler reactivity coefficient is a measure of the reactivity change associated with a change in the temperature of the fuel material. An increase in fuel temperature causes an increase in the absorption of resonance energy neutrons and a decrease in reactivity. The Doppler reactivity coefficient provides instantaneous negative reactivity feedback to any rise in fuel temperature, on either a gross or local basis. Although the reactivity change caused by the Doppler effect is small compared to other power-related reactivity changes during normal operation, it becomes very
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.3-5 Revision 0 important during postulated rapid power excursions in which large fuel temperature changes occur. The GNF2 Doppler reactivity coefficient is negative for any operating conditions.
Moderator Void Coefficient The moderator void coefficient of reactivity is associated with the change in moderating capability of the in-channel water. The moderator void coefficient should be large enough to prevent power oscillation due to spatial xenon changes yet small enough that pressurization transients do not unduly limit plant operation. In addition, the moderator void coefficient has the ability to flatten the radial power distribution and to provide ease of reactor control due to the void feedback mechanism. The overall void coefficient is always negative over the complete operating range because the design is under moderated. The analysis performed to calculate the moderator void coefficient used the lattice physics code TGBLA06 and the 3D BWR core simulator code PANAC11.
Table 15.2-1 focuses on events initiated from conditions off normal power operation. In the Deterministic Safety Analysis (DSA) TRACG is used for event analysis discussed in Subsection 15.5.1.
Moderator Temperature Coefficient The moderator temperature coefficient is associated with a change in the moderating capability of the water. Once the reactor reaches the power producing range, nucleate boiling begins and the moderator temperature remains essentially constant. The moderator temperature coefficient is negative during power operation.
Boron Reactivity For the BWRX-300, the Boron Injection System (BIS) acts as an emergency backup to the insertion of control rods to provide a diverse means of making the reactor subcritical and is discussed in Subsection 9.3.10.
Xenon Reactivity Xenon reactivity feedback is not typically accounted for during events in DSA because BWRs do not have instability problems due to xenon.
Subsection 15.2.4.3 discusses bounding event selection for AOO and Design Basis Accident (DBA) reactivity and power distribution anomalies.
4.3.4.4 Shutdown Margins The minimum cold SDM for the reference BWRX-300 equilibrium core is presented in Figure 4.3-8 and documented in NEDC-34044P (Reference 4.3-2).
Details on how the minimum cold SDM has been quantified can be found in NEDC-34039P (Reference 4.3-4).
for the most reactive HCU coupled control rod pair
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.3-6 Revision 0 The core must be capable of being made subcritical with margin in the most reactive condition throughout an operating cycle with the most reactive coupled control rod pair in their full out position and all other control rods fully inserted. This calculation is performed at cold temperatures, which are between 20°C and 286°C. The amount of reactivity in the core that is below the target eigenvalue is the cold SDM. The typical value of cold SDM required by plant Technical Specifications is 0.38% k/k. SDM is dependent upon the core loading. It is calculated for each plant cycle prior to the operation of that cycle.
The calculations that demonstrate compliance with this requirement are performed for every fuel reload. The results of the cycle-specific calculations are documented in the reload license report for that cycle.
4.3.4.5 Criticality of Reactor During Refueling The minimum allowable SDM will be provided in the Technical Specifications.
4.3.4.6 Thermal Limit The margins to thermal limits for the reference BWRX-300 equilibrium core are presented in Figure 4.3-7 and Figure 4.3-8 and documented in NEDC-34044P (Reference 4.3-2).
4.3.4.7 Xenon Stability BWRs are not susceptible to xenon instabilities. The xenon stability evaluation has been demonstrated by:
No observed xenon instabilities in operating BWRs
Special tests conducted on operating BWRs in an attempt to force the reactor into xenon instability
Calculations 4.3.4.8 Thermal-Hydraulic Stability The BWRX-300 core remains stable throughout the entire operating domain. Refer to Appendix 4A for a discussion of thermal-hydraulic stability.
4.3.4.9 Load Following Operation Core power maneuvering is a normal occurrence for BWRs in response to several motivations (e.g., periodic surveillance testing, managing equipment performance degradation, control rod exchanges to reduce fuel duty, electricity demand). The load follow operation involves a rapid power increase or reduction. The thermal limits established for the fuel are developed to assure conformance to regulatory requirements during all modes of operation, including load following.
Additionally, BWR Operating Guidelines have been developed to assist in mitigating low frequency Stress Corrosion Cracking (SCC) pellet clad interaction.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 15.3-6 Revision 0 Primary Containment Containment pressures and temperatures are maintained below the design values.
The calculated containment pressure does not exceed the design pressure 60 psig (0.414 MPaG).
The calculated containment shell temperature does not exceed the design temperature 330°F (165.6°C).
The local combustible gas concentrations in the containment are within the range where deflagration or detonation cannot occur.
Containment atmosphere remains sufficiently mixed such that deflagration or detonation thresholds are not exceeded.
Containment capability will be retained to reduce the containment pressure and temperature following a DBA to minimize the release of fission products to the environment and to preserve containment integrity and leak tightness.
The calculated containment pressure reduces to less than 50%
of the calculated peak pressure for the most limiting LOCA within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Reactivity Control Reactivity control required to bring the reactor to cold shutdown is maintained.
Shutdown margin is established to assure that the reactor can be brought subcritical with the highest-worth control rod pair withdrawn when the core is in its most reactive condition. The subcriticality value is 0.38% k/k with the highest-worth control rod pair analytically determined.
Table 15.3-2 Design Basis Accident Deterministic Safety Analysis Acceptance Criteria (Sheet 2 of 3)
Fission Product Barrier or Fundamental Safety Function Qualitative Acceptance Criteria Quantitative Acceptance Criteria
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-4.5.1-1 Receipt Date:
9/3/2025 Question:
In order to make a reasonable assurance finding of the suitability of the material for its application, as they relate to quality standards for design, fabrication, erection, and inspection in accordance with meeting the requirements of GDC 1 and 10 CFR 50.55a, the control of delta ferrite in austenitic stainless steel materials should be specified to control cracking. The NRC staff notes that PSAR Section 4.5.3 for Reactor Internals and Section 5.2.3.4.2 for Reactor Coolant Pressure Boundary Material (along with PSAR Table 1.9-20 specifying conforms with RG 1.31) provide the necessary information for controlling delta ferrite. Therefore, provide how the delta ferrite is controlled and measured for austenitic stainless steel materials for the CRD in order to make a reasonable assurance finding that the material is compatible with the reactor coolant environment in accordance with GDC1 and 10 CFR 50.55a.
Response
Austenitic stainless steel materials for the CRD are required by specification to have an average minimum delta ferrite content of 8 FN (ferrite number) and a maximum of 20 FN for 308L and 16 FN for 316L as determined on undiluted weld pads by magnetic measuring instruments calibrated in accordance with American National Standards Institute (ANSI)/American Welding Society (AWS) A4.2M/A4.2, Standard Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite Content of Austenitic and Duplex Ferritic-Austenitic Stainless Steel Weld Metal.
CPA Update on Docket: Markups to CRN-1 PSAR Chapter 4 Subsection 4.5.3.2 are made to specifically note delta ferrite and point to Subsection 5.2.3.4.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-4 Revision 0 (Reference 4.5-1). Table 4.5-1 lists the applicable material product forms as well as the ASME BPVC,Section II (Reference 4.5-2) specifications.
Cast Austenitic Stainless Steel These cast versions of the wrought materials rely on carbon control and the presence of ferrite to assure SCC resistance which requires the ferrite level of 8% minimum for SCC. Additionally, the ferrite levels are also restricted to less than 20% to assure adequate toughness after long-term exposure to higher temperatures. The castings are given solution annealing to maintain SCC resistance.
Nickel Base Alloys Similar composition controls and product form processing to stainless steels are used to assure that wrought Ni-Cr-Fe Alloy 600M and the associated Weld Metal Alloy 82 are free of any sensitization risks. Nickel base wrought materials are Alloy 600M with niobium modified chemistry as well as Alloy 625.
RG 1.84 lists acceptable Code Cases. The primary one of concern is the use of Code Case N-580-2 which supports the use of Alloy 600M to address SCC concerns. Alloy 600 is not used in the BWRX-300.
High Strength Alloys In special locations, the higher strength alloy, Alloy X-750, is used for bolting or fastener type of applications. The main approach used to prevent SCC susceptibility are accomplished by limiting the applied stresses in the design process. Additionally, Alloy 718 may be used for high strength applications.
Alloy 17-4 PH is a martensitic precipitation hardening stainless steel of high strength, and therefore, also potentially susceptible to SCC if the hardness is not limited.
4.5.3.2 Controls on Welding All welding, welding qualification, preheating, post weld heat treatment, and related nondestructive examinations are performed in accordance with ASME BPVC Section IX, Qualification Standard for Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Welding, Brazing, and Fusing Operators, (Reference 4.5-4), and when applicable, ASME BPVC,Section III (Reference 4.5-1). For non-ASME BPVC components or systems, alternate standards such as American Welding Society (AWS) may be applied if specified in the applicable design documents. For ASME BPVC components, welder and welding procedures are qualified in accordance with the ASME BPVC,Section IX (Reference 4.5-4). General welding controls are described in Subsection 5.2.3.
Welding of austenitic stainless steel is performed in compliance with RG 1.31, Control of Ferrite Content in Stainless Steel Weld Metal, and RG 1.44, Control of the Processing and Use of Stainless Steel.
, including control and measure of delta ferrite for austenitic stainless steel materials, Subsection 5.2.3.4
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-4.5.1-2 Receipt Date:
9/2/2025 Question:
PSAR Section 4.5.1.2 states that the primary alloys used for the reactor internal components are austenitic structural alloys. In order to make a reasonable assurance finding of the suitability of the material for its application, as they relate to quality standards for design, fabrication, erection, and inspection in accordance with meeting the requirements of GDC 1 and 10 CFR 50.55a, the following should be specified to minimize stress corrosion susceptibility materials:
- a. maximum carbon content for the austenitic stainless steel materials in Section 4.5.1.2 and Table 4.5-2.
- b. The heat treatment (i.e., solution annealing) of the austenitic stainless steel materials in Section 4.5.1.2 and Table 4.5-2 and whether guidelines in RG 1.44 are used.
Note that PSAR Section 4.5.3.4 and 4.5.4 for Reactor Internals provides the necessary information for heat treatment and use of RG 1.44 guidelines. Therefore, provide the maximum carbon content and the heat treatment (i.e., solution annealing) for the austenitic stainless steel material taking into account the influence carbon content and heat treatment has on the susceptibility to stress corrosion cracking in stainless steel for the CRD in order to make a reasonable assurance finding that the material is compatible with the reactor coolant environment in accordance with GDC1 and 10 CFR 50.55a.
Response
The maximum carbon content and the heat treatment for the austenitic stainless steel materials is discussed in Subsection 5.2.3.4.1 and Table 5.2-2a. PSAR Table 5.2-2 was divided into Table 5.2-2a and Table 5.2-2b by NRC Audit Question Response A-5.2.3-1 and A-5.2.3-2. The BWRX-300 design conforms with RG 1.44 and with the guidelines of Generic Letter 88-01 and NUREG-0313, Revision 2, to avoid sensitization through the use of reduced carbon content and process controls.
CPA Update on Docket: Markup of CRN-1 PSAR Chapter 4 Subsection 4.5.1.2 is provided to add appropriate pointers to Chapter 5; and markup of Table 4.5-2 is provided to add clarity of the maximum carbon content for austenitic stainless steel material. Weld material details are moved from Tables 4.5-1 & 4.5-2 and placed in Table 5.2-2b, with appropriate pointers to Table 5.2-2b added in Section 4.5.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-1 Revision 0 4.5 Reactor Material This section describes the CRD System Structural Materials (Subsection 4.5.1) and the Reactor Internals Materials (Subsection 4.5.3).
4.5.1 Control Rod Drive System Structural Materials This section discusses materials used in the Control Rod Drive Mechanism (CRDM) and extends to the coupling interface with control rod assemblies in the reactor vessel. The materials comprising the fuel and control rod assemblies are described in Section 4.2.
The RCPB components of the CRD, consist of the FMCRD middle flange including the ball check valve, and the lower component housing, which encloses the lower part of the drive. These components are made with 300 series stainless steel materials compatible with the reactor coolant in accordance with ASME BPVC,Section III (Reference 4.5-1). Compliance with RG 1.84, which addresses ASME BPVC,Section III Code Cases, is discussed in Subsection 5.2.1.2, and Code Cases, if used, are listed in Table 5.2-1. The installation bolts used to attach the middle flange and the lower component housing to the CRD housing are low alloy steel material as shown in Table 5.2-2.
The metallic structural components of the CRDM are made from four types of materials: 300 series stainless steel, Nickel-Chrome-Iron Alloy X-750, XM-19 stainless steel and 17-4 PH stainless steel materials. The properties of the non-primary pressure retaining materials selected for the CRDM are equivalent to those given in Parts A, B and D of ASME BPVC,Section II, Materials, (Reference 4.5-2) and Section III (Reference 4.5-1), or are included in RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III. The requirements are intended to address the cold work concerns - both cold work from forming operations and abrasive grinding. Surface cold work can be detrimental to SCC resistance of austenitic stainless steels. In conformance with RG 1.44, Control of the Processing and Use of Stainless Steel, cold worked 300 series austenitic stainless steels are not used except for minor forming and straightening. Cold work concerns are controlled by limiting the material hardness, bend radius, or the amount of strain induced by a process. In these uses, the maximum yield strength used is 90 ksi. There are special materials also used for such components as the bayonet coupling (or coupling spud), the latch and latch spring which are made from XM-19 stainless steel, Alloy X-750 and 17-4 PH stainless steel. These materials have been used successfully in operating BWRs.
4.5.1.1 Materials Specifications The CRD systems material specification defines the material requirements for FMCRD components and is intended to support design requirements. The FMCRD material specification is used with the design specification required by ASME BPVC,Section III, NCA-2021.
When applicable, such as for ASME Class 1 components, ASME BPVC,Section III, NB-2300 applies.
A list of materials and specifications is provided in Table 4.5-2.
Table 5.2-2b lists material specifications for welds.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-2 Revision 0 4.5.1.2 Austenitic Stainless Steel Components The primary alloys used for the reactor internal components are austenitic structural alloys.
These materials have a long history of use in all BWRs. Consequently, there is substantial experience with understanding and mitigating failures in these alloys. Specifically, the wrought austenitic stainless steel components are Type 304, 304L, 316, 316L. XM-19 stainless steel with its higher strength, developed for CRD components, is used in special locations due to its excellent corrosion resistance.
4.5.1.3 Other Materials The structural components of the FMCRD are made from four types of materials: 300 series stainless steel, Nickel-Chrome-Iron Alloy X-750, XM-19 stainless steel, and 17-4 PH stainless steel materials. The RCPB components of the CRD, consist of FMCRD middle flange including the ball check valve, and the lower component housing, which enclose the lower part of the drive.
The flange and lower housing are made with 300 series stainless steel materials in accordance with ASME BPVC,Section III (Reference 4.5-1).
Alloy 17-4 PH is a martensitic precipitation hardening stainless steel of high strength, and therefore, also potentially susceptible to SCC if the hardness is not limited.
4.5.1.4 Cleaning and Cleanliness Control The degree of surface cleanliness meets the requirements of RG 1.28, Quality Assurance Program Criteria (Design and Construction). The Quality Assurance (QA) Program complies with 10 Code of Federal Regulations (CFR) 50, Appendix B and ASME NQA-1-2015, Quality Assurance Requirements, for Nuclear Facility Applications, (Reference 4.5-3). The QA Program also complies with RG 1.28 and is organized to show its relationship to NQA-1-2015 (Reference 4.5-3) and the exceptions and clarifications noted in RG 1.28. The GEH QA Program is discussed in Chapter 17.
4.5.2 Regulatory Requirements Compliance with 10 CFR 50, Appendix A The CRD system structural materials comply with the following GDC, which has plant-wide applicability, as discussed in Section 3.1:
GDC 1 - Quality standards and records.
The following GDC have aspects specific to the CRD system structural materials and compliance is addressed as follows:
GDC 14 - Reactor coolant pressure boundary. The ICS forms part of the RCPB and, as such, compliance with GDC 14 with regard to RCPB materials is addressed in Chapter 5.
GDC 26 - Reactivity control system redundancy and capability. Compliance to GDC 26 is discussed in Section 3.1 and Subsection 7.1.3.
Maximum carbon content and heat treatment (i.e., solution annealing) of austenitic stainless steel materials are discussed in Subsection 5.2.3.4.1 and Table 5.2-2a.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-4 Revision 0 Table 4.5-1 lists the materials used for the internal components including the shroud and core support ring, chimney, steam dryer assembly, top guide, core support structures, CRD housings, and incore housings (Figure 4.1-1). All of these materials are procured to specifications from ASME BPVC,Section II (Reference 4.5-2) as required by ASME BPVC,Section III (Reference 4.5-1). Table 4.5-1 lists the applicable material product forms as well as the ASME BPVC,Section II (Reference 4.5-2) specifications.
Cast Austenitic Stainless Steel These cast versions of the wrought materials rely on carbon control and the presence of ferrite to assure SCC resistance which requires the ferrite level of 8% minimum for SCC. Additionally, the ferrite levels are also restricted to less than 20% to assure adequate toughness after long-term exposure to higher temperatures. The castings are given solution annealing to maintain SCC resistance.
Nickel Base Alloys Similar composition controls and product form processing to stainless steels are used to assure that wrought Ni-Cr-Fe Alloy 600M and the associated Weld Metal Alloy 82 are free of any sensitization risks. Nickel base wrought materials are Alloy 600M with niobium modified chemistry as well as Alloy 625.
RG 1.84 lists acceptable Code Cases. The primary one of concern is the use of Code Case N-580-2 which supports the use of Alloy 600M to address SCC concerns. Alloy 600 is not used in the BWRX-300.
High Strength Alloys In special locations, the higher strength alloy, Alloy X-750, is used for bolting or fastener type of applications. The main approach used to prevent SCC susceptibility are accomplished by limiting the applied stresses in the design process. Additionally, Alloy 718 may be used for high strength applications.
Alloy 17-4 PH is a martensitic precipitation hardening stainless steel of high strength, and therefore, also potentially susceptible to SCC if the hardness is not limited.
4.5.3.2 Controls on Welding All welding, welding qualification, preheating, post weld heat treatment, and related nondestructive examinations are performed in accordance with ASME BPVC Section IX, Qualification Standard for Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Welding, Brazing, and Fusing Operators, (Reference 4.5-4), and when applicable, ASME BPVC,Section III (Reference 4.5-1). For non-ASME BPVC components or systems, alternate standards such as American Welding Society (AWS) may be applied if specified in the applicable design documents. For ASME BPVC components, welder and welding procedures are qualified in accordance with the ASME BPVC,Section IX (Reference 4.5-4). General welding controls are described in Subsection 5.2.3.
Table 5.2-2b lists material specifications for welds.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-8 Revision 0 Table 4.5-1 Reactor Internal Components and Materials Component(s)
Form Material Type ASME Specification Steam Dryer Assembly Plate, Bar or Forgings Stainless Steel SA-240/SA-240M, Type 304/304L, 316/316L SA-182/SA-182M, Grades F304/F304L, F316/F316L SA-479/SA-479M, Type 304/304L, 316/316L Shroud and Core Support Ring Plate or Forgings Stainless Steel SA-240/SA-240M, Type 316L SA-182/SA-182M, Grade F316L Core Support Structure Plate or Forgings Stainless Steel SA-240/SA-240M, Type 316L SA-182/SA-182M, Grade F316L Core Support Legs Plate, Bar or Forgings Ni-Cr-Fe SB-168, SB-166, or SB-564 Nickel Alloy modified per ASME Code Case N-580-2 Top Guide Plate or Forgings Stainless Steel SA-240/SA-240M, Type 304/304L SA-182/SA-182M, Grade F316/F316L Chimney Plate or Forgings Stainless Steel SA-240/SA-240M, Type 304/304L SA-182/SA-182M, Grade F316/F316L Fuel Supports Castings Stainless Steel SA-351/SA-351M, Grade CF3 Core Plate Bolts Bar Stainless Steel SA-479/SA-479M, Type XM-19 Stainless Steel Welds Covered Electrode or Filler Metal Stainless Steel SFA-5.4 SFA-5.9 Nickel Alloy Welds Filler Wire Nickel alloy SFA-5.14 (Note 1)
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-9 Revision 0 Table 4.5-2 Control Rod Drive Components and Materials Structural Component(s)
Form Material Type ASME Specification Coupling Spud Bar and Forgings Nickel Alloy SB-637 UNS N07750 Type 3, or UNS N07718 Type 2 (per ASME Code Case N-60-6)
Outer Tube Seamless Pipe Stainless Steel SA-312/SA-312M, Grade TP XM-19 Outer Tube Head Bar and Shapes Stainless Steel SA-479/SA-479M, Type XM-19 Ball Screw Bar and Shapes Stainless Steel SA-276, Type 440C Ball Nut Bar and Shapes Stainless Steel SA-564/SA-564M, Type 630(1)
Lower Component Drive Shaft Bar and Shapes Stainless Steel SA-479/SA-479M, Type 316 Stop Piston Seamless Pipe, Bar and Shapes Stainless Steel SA-312/SA-312M, Type 304/316 SA-376/SA-376M, Type304/316 SA-479/SA-479M, Type 304/316 Guide Tube Seamless Pipe Stainless Steel SA-312/SA-312M, Type 304/316 SA-376/SA-376M, Type 304/316 Stainless Steel Welds Covered Electrode or Filler Metal Stainless Steel SFA-5.4, Type E308L, E316L SFA-5.9, Type ER308L, ER316L Stainless Steel Welds Consumable Inserts Stainless Steel SFA-5.30, Type IN308L Note:
- 1. To reduce risk of cracking, only the H1075, H1100, and H1150 conditions are allowed in 17-4 PH /Type 630 stainless steel.
(Note 2)
Table 5.2-2b Reactor Coolant Pressure Boundary Welding Filler Metals Base Material Filler Metal Type ASME Classification AWS Classification Carbon Steel Welds Covered Electrode SFA-5.1 E7018 Filler Metal SFA-5.18 ER70S-2, ER70S-3, ER70S-6 Low Alloy Steel Welds (Note 1)
Covered Electrode SFA-5.5 E9018-B3 Filler Metal SFA-5.23 EF2, EF3 or other mutually agreed filler metal with appropriate chemistry selected for base metal composition and mechanical property requirements including toughness SFA-5.28 ER90S-XX with appropriate chemistry selected for base metal composition, shielding gas, and mechanical property requirements including toughness Low Alloy Steel Piping Welds (21/4 Cr-Mo)
Covered Electrode SFA-5.5 E9018-B3 Filler Metal SFA-5.28 ER90S-XX with appropriate chemistry selected for base metal composition, shielding gas, and mechanical property requirements including toughness Stainless Steel Welds (Note 2)
Covered Electrode or Filler Metal SFA-5.4 E308L-XX, E309L-XX, E316L-XX Filler Metal SFA-5.9 ER308L, ER308LSi, ER309L, ER316L Nickel Alloy Welds Filler Wire SFA-5.14 ERNiCr-3 Notes:
- 1. Dissimilar metal welds between carbon/low-alloy steel and stainless steel are made with a layer (or layers) of 309L, followed by completion of the weld with either 308L, 316L, or 309L. Alternatively, the welds may be completed with Nickel Alloy 82 for the entire weld.
Dissimilar metal welds between nickel-based alloy and carbon/low-alloy steel or stainless steel are performed using Nickel Alloy 82. Welds between carbon steel and low-alloy steel may be made with the filler metals listed for either base material, except that partial penetration welds may also be made with Nickel Alloy 82.
- 2. The maximum carbon content for stainless steel weld filler metal will be less than or equal to 0.035% in accordance with GL 88-01.
Consumable Inserts Stainless Steel Welds SFA-5.30 IN308L, IN316L
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-4.A-1 Receipt Date:
8/11/2025 Question:
Section 4A.1 notes several different methodologies as relevant for the stability analyses performed in the PSAR to demonstrate compliance with GDCs 10 and 12. This section also notes that NEDC-34270, BWRX-300 Stability Analysis Licensing Topical Report" is specific to the BWRX-300 design. However, the staff notes that this topical report is not clearly included within the Clinch River CP licensing basis. Further, Section 4A.6 "Stability Analysis Methods" does not include reference to NEDC-34270, and the section instead notes that the detailed discussion of the methods used to analyze BWRX-300 thermal-hydraulic stability is presented in NEDC-34043P, "BWRX-300 TRACG Application". However, the scope of this report covers the justification for the use of the TRACG code for AOO, Design Basis Accident and Design Extension Conditions. It does not justify the use of the TRACG code to determine stability margin and provide the stability analysis results. Without the inclusion of NEDC-34270 topical report in Section 4A.6, the staff believes that the documentation of the applicant's stability analysis method is incomplete.
Consider revising the PSAR to appropriately reference and adequately describe, outline, or summarize the NEDC-34270 methodology in Section 4A.6 as a preliminary methodology for the PSAR. The staff notes that an eventual -A version could be referenced and adequately described, outlined, or summarized at the OL stage.
Response
It is agreed that Licensing Topical Report NEDC-34270P, BWRX-300 Stability Analysis, Revision 0, which is currently under review by the NRC, presents the methods used to analyze BWRX-300 thermal-hydraulic stability. TVA has provided markups to update PSAR Section 4A.6 to properly summarize and reference NEDC-34270P. As per the clarification call with the NRC held on 8/14/25, the pointers to Subsection 15.5.2 will remain in the PSAR text at this time.
CPA Update on Docket: Markup of CRN-1 PSAR Section 4A.6 provided to appropriately summarize and reference NEDC-34270P.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4A-2 Revision 0 In the time domain analysis, decay ratio is defined as the ratio of the amplitude of two successive peaks. For the BWRX-300, the decay ratio is averaged over several cycles to reduce analytical uncertainty.
Conservative design criteria are imposed on the core-wide, regional, and single-channel decay ratios under all conditions of normal operation and anticipated transients. The limiting mode (i.e.,
highest decay ratio) for thermal-hydraulic oscillations is the core-wide mode.
Analyses are performed to confirm that the acceptance criterion for the core decay ratio of 0.80 is met, and to show that the BWRX-300 core will be stable through normal operations and AOOs considering uncertainties. Along with confirming stable operation, this also ensures that the Safety Limit MCPR is protected.
4A.5 Stability Solution Design The BWRX-300 design has features that result in stable behavior in normal operation and minimize the effects of potential oscillations in off-normal conditions include:
Small Core:
The small core size of the BWRX-300 reduces the likelihood of regional mode oscillations.
Core-wide oscillations are the dominant mode. Tighter neutronic coupling precludes regional mode oscillations. Mitigation of core-wide oscillations is described in Subsection 15.5.2.
Natural circulation:
As there are no recirculation pumps, recirculation pump trip transients that can induce oscillations have been eliminated.
Loss of Feedwater Heating (LFWH) AOO effect on stability is mitigated by SCRRI operation that reduces the core thermal power. The SCRRI function is described in Subsection 7.3.3. The LFWH AOO analysis is described in Subsection 15.5.3.
Tall chimney:
Increases volume of water Increases driving head and natural circulation flow Dampens oscillations
Optimized inlet core orifice design 4A.6 Stability Analysis Methods A detailed discussion of the methods used to analyze BWRX-300 thermal-hydraulic stability is presented in NEDC-34043P, BWRX-300 TRACG Application, (Reference 4A-3).
TRACG is a GEH proprietary version of the Transient Reactor Analysis Code. TRACG uses a multi-dimensional, two-fluid model for reactor thermal-hydraulics and a 3D reactor kinetics model. The models can be used to accurately simulate a large variety of test and reactor configurations. These features allow for realistic simulation of a wide range of BWR phenomena and are described in detail in NEDE-32176P, TRACG Model Description, (Reference 4A-4).
NEDC-34270P, "BWRX-300 Stability Analysis," (Reference 4A-7). NEDC-34270P provides a methodology utilizing implicit numerical integration for channel components and a nominal core wide decay ratio acceptance criterion.
The methodology described in NEDC-34270P leverages
, documenting its applicability to stability analysis
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4A-3 Revision 0 TRACG has been extensively qualified against separate effects tests, component performance data, integral system effects tests, and operating BWR plant data. The details are presented in NEDE-32177P, TRACG Qualification, (Reference 4A-5). TRACG has been used in a variety of applications for the operating BWRs as well as design and analysis for the ESBWR. Application for the ESBWR stability analysis is documented in NEDC-33083P-A, TRACG Application for ESBWR, (Reference 4A-6).
The latest TRACG base deck and PANAC core simulator wrapups have been used to evaluate the BWRX-300 stability performance. This evaluation supports the BWRX-300 design to evaluate decay ratio performance for the 100% nominal power condition where the BWRX-300 would have the least margin for stability in normal operation.
This stability analysis performed in accordance with the approved TRACG methodology demonstrates that the BWRX-300 maintains a coupled power-flow response such that any operational perturbation, maneuver, or AOO that does not cause an immediate scram is naturally damped, and decays quickly to steady state for all modes of operation and prevents the SAFDLs from being exceeded. The BWRX-300 is not susceptible to regional mode oscillations. Therefore, a special stability detection and trip system is not required as documented in NEDC-33912P-A (Reference 4A-1).
This analysis, including the decay ratio, is discussed in Subsection 15.5.2, and the acceptance criteria for the SAFDLs is provided in Section 15.3.
4A.7 References 4A-1 NEDC-33912P-A, BWRX-300 Reactivity Control, GE-Hitachi Nuclear Energy Americas, LLC, Revision 2, February 2023.
4A-2 NEDC-33922P-A, BWRX-300 Containment Evaluation Method, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, June 2022.
4A-3 NEDC-34043P, BWRX-300 TRACG Application, GE-Hitachi Nuclear Energy Americas, LLC, Revision 0, May 2024.
4A-4 NEDE-32176P, TRACG Model Description, GE-Hitachi Nuclear Energy Americas, LLC, Revision 4, January 2008.
4A-5 NEDE-32177P, TRACG Qualification, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, August 2007.
4A-6 NEDC-33083P-A, TRACG Application for ESBWR, GE Nuclear Energy, Revision 1, March 2005.
NEDC-34270P using 4A-7 NEDC-34270, Revision 0, "BWRX-300 Stability Analysis," GE-Hitachi Nuclear Energy Americas, LLC, March 2025
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-5.2.3-1 Receipt Date:
8/25/2025 Question:
In order to make a reasonable assurance finding of the suitability of the material for its application, as they relate to quality standards for design, fabrication, erection, and inspection in accordance with meeting the requirements of GDC 1, 4, 14 and 30 and 10 CFR 50.55a, the materials type/classification for pressure-retaining metals, including weld filler metals that are used for each component in the RCPB should be specified. Table 5.2-2, Reactor Coolant Pressure Boundary Materials, provides the material, ASME Specification and grade/classification for the reactor coolant pressure boundary components. However, the table only provides the ASME specification (e.g., Nickel-based alloy ASME SFA-5.14) for the weld filler metals to be used. This filler metal specification is broad and includes many types of filler. Therefore, provide the filler metal classification for each of the ASME specifications listed in Table 5.2-2 to be used (e.g., Nickel-based alloy ASME SFA-5.14, ERNiCr-3) in order to make a reasonable assurance finding that the material is compatible with the reactor coolant environment in accordance with GDC1, 4, 14, 30 and 10 CFR 50.55a.
Response
Table 5.2-2 will be revised to provide the AWS filler metal classification for each of the ASME specifications listed in the table (e.g., Nickel Alloy Welds, ASME SFA-5.14, ERNiCr-3). The table is being split into Table 5.2-2a covering materials and Table 5.2-2b covering weld metal. The TVA Response for Question Number A-5.2.3-3 will be transmitted separately.
CPA Update on Docket: See attached markup of PSAR Table 5.2-2a and Table 5.2-2b. This markup also includes PSAR changes in response to NRC Audit Question A-5.2.3-2. The CPA markup for Question Number A-5.2.3-3 will be transmitted separately.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-1 Revision 0 4.5 Reactor Material This section describes the CRD System Structural Materials (Subsection 4.5.1) and the Reactor Internals Materials (Subsection 4.5.3).
4.5.1 Control Rod Drive System Structural Materials This section discusses materials used in the Control Rod Drive Mechanism (CRDM) and extends to the coupling interface with control rod assemblies in the reactor vessel. The materials comprising the fuel and control rod assemblies are described in Section 4.2.
The RCPB components of the CRD, consist of the FMCRD middle flange including the ball check valve, and the lower component housing, which encloses the lower part of the drive. These components are made with 300 series stainless steel materials compatible with the reactor coolant in accordance with ASME BPVC,Section III (Reference 4.5-1). Compliance with RG 1.84, which addresses ASME BPVC,Section III Code Cases, is discussed in Subsection 5.2.1.2, and Code Cases, if used, are listed in Table 5.2-1. The installation bolts used to attach the middle flange and the lower component housing to the CRD housing are low alloy steel material as shown in Table 5.2-2.
The metallic structural components of the CRDM are made from four types of materials: 300 series stainless steel, Nickel-Chrome-Iron Alloy X-750, XM-19 stainless steel and 17-4 PH stainless steel materials. The properties of the non-primary pressure retaining materials selected for the CRDM are equivalent to those given in Parts A, B and D of ASME BPVC,Section II, Materials, (Reference 4.5-2) and Section III (Reference 4.5-1), or are included in RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III. The requirements are intended to address the cold work concerns - both cold work from forming operations and abrasive grinding. Surface cold work can be detrimental to SCC resistance of austenitic stainless steels. In conformance with RG 1.44, Control of the Processing and Use of Stainless Steel, cold worked 300 series austenitic stainless steels are not used except for minor forming and straightening. Cold work concerns are controlled by limiting the material hardness, bend radius, or the amount of strain induced by a process. In these uses, the maximum yield strength used is 90 ksi. There are special materials also used for such components as the bayonet coupling (or coupling spud), the latch and latch spring which are made from XM-19 stainless steel, Alloy X-750 and 17-4 PH stainless steel. These materials have been used successfully in operating BWRs.
4.5.1.1 Materials Specifications The CRD systems material specification defines the material requirements for FMCRD components and is intended to support design requirements. The FMCRD material specification is used with the design specification required by ASME BPVC,Section III, NCA-2021.
When applicable, such as for ASME Class 1 components, ASME BPVC,Section III, NB-2300 applies.
A list of materials and specifications is provided in Table 4.5-2.
Table 5.2-2a.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-5 Revision 0 10 CFR 50, Appendix B, Criterion XIII, as it relates to onsite material cleaning control, as discussed in Subsection 5.3.3.4.
10 CFR 50, Appendix G, as it relates to materials testing and acceptance criteria for fracture toughness of the RCPB as discussed in Subsection 5.2.3.3.1 10 CFR 50.55a, as it relates to quality standards applicable to the RCPB, by compliance with ASME BPVC,Section III and conformance with RG 1.31, RG 1.50, RG 1.71, Generic Letter 88-01, and NUREG-0313, Revision 2. RG 1.43 does not apply as discussed in Subsection 5.3.1.4.
Section 3.1 summarizes compliance with GDC.
5.2.3.1 Material Specifications This subsection discusses the specifications for pressure-retaining ferritic materials, austenitic stainless steels, and nonferrous metals, including bolting and weld materials, that are used for each RCPB component, e.g., vessels, piping, and valves. The adequacy and suitability of the ferritic materials, austenitic stainless steels, and nonferrous metals specified for the above application are discussed in Subsection 5.2.3.1 through Subsection 5.2.3.4.
Table 5.2-2 lists the principal pressure-retaining materials and the appropriate material specifications for the RCPB components and welds. RCPB materials comply with ASME BPVC,Section III, NB-2000 for ASME Class 1 components and NCD-2000 for ASME Class 2 components and ASME BPVC,Section II, Materials, (Reference 5.2-4). RCPB ferritic materials comply with 10 CFR 50, Appendix G. Compliance with RG 1.84, which addresses ASME BPVC,Section III Code Cases, is discussed in Subsection 5.2.1.2, and Code Cases, if used, are listed Table 5.2-1. No exceptions are taken to the material specifications of the ASME BPVC.
BWR piping materials and materials processing conforms to Attachment A to Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, and considers the recommendations in NUREG-0313, Revision 2, Technical Report on Material Selection and Processing Guidelines for Boiling Water Reactor (BWR) Coolant Pressure Boundary Piping, related to minimizing Stress Corrosion Cracking (SCC) in susceptible piping. Compliance with Attachment A to Generic Letter 88-01 includes carbon content and ferrite limits for stainless steel base and weld metals, and nickel-based alloy material selection. Further discussion of the materials associated with the RPV pressure-retaining components is provided in Subsection 5.3.1.
Cast austenitic stainless steels used in the RCPB have material specifications such as maximum ferrite levels that ensure adequate fracture toughness over the design life to support use of the material under the expected environmental conditions, e.g., exposure to the reactor coolant operating temperatures.
Nickel-chromium-iron alloys such as Alloy 600M consistent with Code Case N-580-2 along with Alloy 82 weld metal are used in the RCPB. Alloy 600 and associated weld materials are not used in the RCPB, due to their susceptibility to SCC.
Table 5.2-2a and Table 5.2-2b list
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-6 Revision 0 RPV and RPV Appurtenances Table 5.2-2 lists materials and specifications for the RPV and RPV appurtenances which form part of the RCPB. This includes RPV nozzles, stub tubes, CRD flanges/housings, and incore instrumentation housings and their associated RCPB components.
RPV Attachments Table 5.2-3 lists materials for selected RPV attachments. RPV attachments are those components welded directly to the RPV but do not form part of the RCPB. Many of these attachments are internal to the RPV and in contact with reactor coolant, while some are external to the RPV and not in contact with reactor coolant.
RPV attachments internal to the RPV and in contact with reactor coolant include the core support legs.
RPV attachments external to the RPV and not in contact with reactor coolant include the vessel stabilizer brackets and the vessel support skirt.
RPV Internals Materials for RPV internals, such as steam dryer assembly, chimney, and core support structures, are listed in Table 4.5-1. Discussion of RPV internals in this chapter is limited to their compatibility with reactor coolant and certain corrosion aspects or effects on water chemistry.
5.2.3.2 Compatibility with Reactor Coolant General corrosion and SCC induced by impurities in the reactor coolant can cause failures of the RCPB. The chemistry of the reactor coolant and any additives whose function is to control corrosion are discussed in Subsection 5.2.3.2.1 and Subsections 9.3.7, 9.3.8, and 9.3.9. The compatibility of the materials of construction employed in the RCPB with the reactor coolant, contaminants, or radiolytic products to which the system is exposed was considered by employing proven materials and processes. The extent of time-dependent general corrosion and other degradation mechanisms of materials of construction in contact with the reactor coolant was considered. Similarly, consideration was given to the control of the use of austenitic stainless steels in the sensitized condition.
5.2.3.2.1 Chemistry of Reactor Coolant Proper material selection, fabrication and process controls, and control of water chemistry ensures the integrity of the RCPB. This subsection provides a brief review of the relationships between water chemistry variables and RCPB materials performance, fuel performance, and plant radiation fields. Electric Power Research Institute (EPRI) guidelines in EPRI BWRVIP-190, BWR Vessel and Internals Project, Volume 1: BWR Water Chemistry Guidelines - Mandatory, Needed and Good Practice Guidance and Volume 2: BWR Water Chemistry Guidelines - Technical Basis, (Reference 5.2-1) are used as inputs to develop the basis for the BWRX-300 water chemistry program. The BWRX-300 water chemistry program may continue to be revised based on new EPRI guidelines, and as operational experience is gained.
Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-11 Revision 0 For the reactor internals, austenitic stainless steel is primarily used to minimize levels of radiation from corrosion products generated in the core. Where higher strength requirements than those for austenitic stainless steels are imposed, nickel-based alloys, nitrogen strengthened alloys, or precipitation hardening stainless steels are used.
The stainless steel structural materials used for the BWRX-300 reactor core internals are primarily wrought materials along with their weld metals, although cast austenitic stainless steels are used as well.
In summary, construction materials of the RCPB and reactor internals exposed to the reactor coolant consist of the following:
Solution-annealed austenitic stainless steels (both wrought and cast), Types 304, 304L, 316, 316L, CF3M, CF8M, and XM-19
Solution-annealed tempered martensitic stainless steels, Type 410
Nickel-based alloys (including niobium-modified Alloy 600M with filler metal Alloy 82, Alloy 625 with filler metal ERNICMO-3, Alloy 718, and X-750)
Precipitation hardening stainless steels such as Type 630 and Type XM-13
Carbon steels and low-alloy steels Table 5.2-2 lists materials and material specifications for RCPB components. Table 4.5-1 lists material and material specifications for reactor internals.
Suitability of Materials of Construction Exposed to Reactor Coolant Materials selection and processing controls address materials degradation issues in the RCPB such as SCC, general corrosion, and FAC. The specific categories of materials that are used are wrought austenitic stainless steels, cast austenitic stainless steels, martensitic stainless steels, precipitation hardening stainless steels, and nickel-based alloys along with their weld metals used as part of construction. General corrosion for the plant components is managed in the design process based on the ASME BPVC, which sets minimum corrosion allowances.
Enhanced IGSCC resistance has been achieved through the use of IGSCC resistant materials such as Type 316L with controlled composition stainless steel, XM-19, and stabilized nickel-based niobium-modified Alloy 600M and Alloy 82.
BWRX-300 materials provide adequate strength, fracture toughness, fabricability, and compatibility with the BWR environment. Their suitability has been demonstrated by long-term successful OPEX in reactor service.
Material Contamination and Transgranular Stress Corrosion Cracking The requirements of GDC 4 relative to compatibility of components with environmental conditions are met by compliance with the applicable provisions of the ASME BPVC such as corrosion allowances and by conformance with the recommendations of RG 1.44, Control of the Processing and Use of Stainless Steel. Material contamination that can lead to transgranular Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-29 Revision 0 Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 1 of 7)
Component Form Material (Note 1)
ASME Specification Reactor Pressure Vessel RPV Shells and Heads Forging Plate Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1, SA-533/SA-533M Type B Class 1 RPV Shell and Head Flange Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 RPV Main Closure Bolting Bolting Low-Alloy Steel SA-540/SA-540M Gr B23 or B24 Class 3 RPV Standard Flange Bolting Bolting Low-Alloy Steel SA-540/SA-540M Gr B23 or B24 Class 3 RPV Nozzles Forging Low-Alloy steel SA-508/SA-508M Gr 3 Class 1 RPV Level Instrumentation Nozzles Forging Stainless steel or Ni-Cr-Fe SA-182/SA-182M, SA-965/SA-965M Gr F316/F316L, or SB-168, SB-166, or SB-564 (Modified per ASME Code Case N-580-2)
Main Steam Reactor Isolation Valves (MSRIV)
MSRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSRIV Cover Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSRIV Poppet Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
MSRIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 MSRIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Main Steam Containment Isolation Valves (MSCIV)
MSCIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSCIV Cover Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSCIV Poppet Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSCIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
MSCIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-30 Revision 0 MSCIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Main Steam Line (RCPB Portion)
MS Piping Seamless Forged & Bored Carbon steel Low-Alloy steel SA-333/SA-333M Gr 6, SA-335/SA-335M Gr P22, SA-369/SA-369M Gr FP2 MS Contour Nozzle Forging Low-Alloy steel SA-508/SA-508M Gr 3 Class 1 MS 200 mm 1500 lb. Large Groove Flange Forging Carbon steel SA-350/SA-350M Gr LF2 Class 1 MS 50 mm Special Nozzle Forging Carbon steel SA-350/SA-350M Gr LF2 Class 1 MS Elbow Seamless Fitting Carbon steel SA-420/SA-420M Gr WPL6, SA-508/SA-508M Gr 1 MS Head Fitting/
Penetration Piping Forging Carbon steel SA-350/SA-350M Gr LF2 Class 1 MS Other Fittings Forging Low-Alloy steel SA-234/SA-234M Gr WP22, SA-336/SA-336M Gr F22 Reactor Pressure Vessel Head Vent Head Vent Piping Piping Stainless Steel SA-312/SA-312M Gr TP316L Head Vent to MSL Valves Forging Carbon Steel SA-350/SA-350M Gr LF2 Class 1 Head Vent to Quench Tank Boundary Valves Forging Stainless Steel SA-182/SA-182M Gr F316/F316L RPV Appurtenances Control Rod Drive Flanges/Housings CRD Housings Forging Stainless steel SA-182/SA-182M, SA-965/SA-965M Gr F304/F304L/F316/F316L Lower Component Housings Forging Stainless Steel SA-965/SA-965M Gr F304 Middle Flange Forging Stainless Steel SA-965/SA-965M Gr F304 Upper Flange Forging Stainless Steel SA-965/SA-965M Gr F316 Ball Check Valve Retainer Bar and Shapes Stainless Steel SA-479/SA-479M Type 316 CRD Spool Piece (if used)
Forging Stainless steel SA-965/SA-965M Gr F304/F304L Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 2 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-31 Revision 0 CRD Mounting Bolts Bolting Alloy steel SA-193/SA-193M Gr B7 CRD Penetration Stub Tubes Bar, Seamless Pipe, Forging Ni-Cr-Fe SB-166, SB-167, SB-564 (Modified per ASME Code Case N-580-2)
RPV Appurtenances Incore Instrumentation Housings Incore Instrumentation Nozzles Forging Bar, Seamless Pipe, Forging Stainless steel or Ni-Cr-Fe SA-182/SA-182M, SA-965/SA-965M Gr F304/F304L/F316/F316L or SB-166, SB-168, or SB-564 (Modified per ASME Code Case N-580-2)
Incore Instrumentation Stub Tubes Bar, Seamless Pipe, Forging Ni-Cr-Fe SB-166, SB-168, or SB-564 (Modified per ASME Code Case N-580-2)
Incore Housings Seamless Pipe, Forging Stainless Steel SA-312/SA-312M Gr 304/304L or, SA-182/SA-182M Gr F316/F316L Isolation Condenser System Piping and Components IC Steam Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
IC Standby Gas Purge Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
IC Cover Bolting (Note 3)
Forging Stainless steel SA-705/SA-705M Type 630 IC Drum and Drum Covers Forging Ni-Cr-Mo-Nb SB-564 Gr 1 UNS N06625 (Note 8)
IC Condensate Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
ICS Recombiner Vessel Forging Stainless steel SA-182/SA-182M, SA-965/SA-965M Gr F316 ICS Recombiner Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 3 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-32 Revision 0 Isolation Condenser System Isolation Condenser Supply Reactor Isolation Valve (ICSRIV)
ICSRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICSRIV Disc Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICSRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Isolation Condenser System Isolation Condenser Return Reactor Isolation Valve (ICRRIV)
ICRRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICRRIV Disc Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICRRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Isolation Condenser System ICS to SDC Containment Isolation Valves (CIV)
ICS CIV Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L SDC Interface Redundant Isolation Valve (ICS Outboard CIV)
Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L Isolation Condenser System Other ICS Purge Valves Forging or Casting Stainless Steel SA-182/SA-182M Gr F316/F316L, SA-351/SA-351M Gr CF3M ICS Purge Piping Seamless Pipe Stainless Steel SA-312/SA-312M Gr TP316L ICS Purge Fittings Forging Stainless Steel SA-182/SA-182M Gr F316/F316L SA-403/SA-403M Gr WP 316/
316L ICS Condensate Return Valves (Note 7)
(Note 7)
(Note 7)
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 4 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-33 Revision 0 Condensate and Feedwater Heating System Feedwater Piping (RCPB Portion)
FW Piping Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr TP316L FW Fittings Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr TP316L Condensate and Feedwater Heating System Feedwater Reactor Isolation Valves (FWRIV)
FWRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 FWRIV Cover Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 FWRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
FWRIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 FWRIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Condensate and Feedwater Heating System Feedwater Containment Isolation Valves (FWCIV)
FWCIV Body Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L FWCIV Cover Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L FWCIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
FWCIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 FWCIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Reactor Water Cleanup System (RCPB Portion)
CUW Piping Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr TP316L CUW Fittings Forging Stainless Steel SA-182/SA-182M, SA-965/SA-965M Gr F316/F316L, SA-403/SA-403M Gr WP 316/
316L Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 5 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-34 Revision 0 CUW RIV Body (Note 4)
Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 CUW RIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
CUW CIV Body Casting Stainless Steel SA-351/SA-351M Gr CF8M/CF3M CUW CIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Boron Injection System (RCPB Portion)
BIS CIV Body Forging Stainless Steel SA-182/SA-182M Gr F316/F316L BIS CIV Disc Forging Stainless Steel SA-182/SA-182M Gr F316/F316L BIS CIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
BIS Check Valve Body Forging Stainless Steel SA-182/SA-182M Gr F316/F316L BIS Piping Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr Type 316/316L BIS Fittings Seamless Forged & Bored Stainless Steel SA-182/SA-182M Gr F316/F316L Additional RCPB Bolting Material Flanges, Covers, and Bonnets Bolting Stud or Bolting Alloy & Carbon Steels SA-354, SA-449 Welding Filler Metals Base Material Filler Metal Type ASME Classification (Note 6)
Carbon Steel Welds Covered Electrode or Filler Metal SFA-5.1, SFA-5.18 Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 6 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-35 Revision 0 Low-Alloy Steel Welds Covered Electrode or Filler Metal SFA-5.5, SFA-5.23, SFA-5.28 Low-Alloy Steel Piping Welds (21/4 Cr-1 Mo)
Covered Electrode or Filler Metal SFA-5.5, SFA-5.28 Stainless Steel Welds Covered Electrode or Filler Metal SFA-5.4, SFA-5.9 Nickel-Based Alloy Welds Filler Wire SFA-5.14 Notes:
1.
Carbon content of RCPB wrought austenitic stainless steel (304/304L/316/316L) is 0.02%
maximum.
2.
To reduce risk of cracking in valve stems, the minimum nominal aging temperature in precipitation hardened stainless steels shall be 580°C (H1075) or higher.
3.
SA-705/SA-705M Type 630 bolting is limited to the H1100 condition per Table 4 of ASME BPVC,Section II, Part D.
4.
Any wetted surfaces in carbon or low-alloy steel bodied valves in the CUW are clad with stainless steel.
5.
Reference Table 5.3-1 and Subsection 5.3.1.5 for core beltline material composition limits.
6.
Dissimilar metal welds between carbon/low-alloy steel and stainless steel are made with a layer (or layers) of 309L, followed by completion of the weld with either 308L, 316L, or 309L.
Alternatively, the welds may be completed with Nickel Alloy 82 for the entire weld. Dissimilar metal welds between nickel-based alloy and carbon/low-alloy steel or stainless steel are performed using Nickel Alloy 82. Welds between carbon steel and low-alloy steel may be made with the filler metals listed for either base material, except that partial penetration welds may also be made with Nickel Alloy 82.
7.
Information currently not available from component supplier.
8.
Code Case N-943 for use of Alloy 625 (UNS N06625) is provided in Table 5.2-1.
9.
The minimum tempering temperature shall be 595°C (~1100°F) for a minimum of two hours.
The maximum hardness following final heat treatment shall be Rockwell C 28 for Type 410.
- 10. Items fabricated from these materials are wetted by reactor coolant but may not necessarily be part of the design ASME BPVC pressure boundary. Therefore, some of the specific material specifications or material specification grades identified under this category are not required to be listed in the ASME BPVC,Section II, Part D, Table 2A.
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 7 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2b [See INSERT next page]
Table 5.2-2a Not Used.
Table 5.2-2b Reactor Coolant Pressure Boundary Welding Filler Metals Base Material Filler Metal Type ASME Classification AWS Classification Carbon Steel Welds Covered Electrode SFA-5.1 E7018 Filler Metal SFA-5.18 ER70S-2, ER70S-3, ER70S-6 Low Alloy Steel Welds (Note 1)
Covered Electrode SFA-5.5 E9018-B3 Filler Metal SFA-5.23 EF2, EF3 or other mutually agreed filler metal with appropriate chemistry selected for base metal composition and mechanical property requirements including toughness SFA-5.28 ER90S-XX with appropriate chemistry selected for base metal composition, shielding gas, and mechanical property requirements including toughness Low Alloy Steel Piping Welds (21/4 Cr-Mo)
Covered Electrode SFA-5.5 E9018-B3 Filler Metal SFA-5.28 ER90S-XX with appropriate chemistry selected for base metal composition, shielding gas, and mechanical property requirements including toughness Stainless Steel Welds (Note 2)
Covered Electrode or Filler Metal SFA-5.4 E308L-XX, E309L-XX, E316L-XX Filler Metal SFA-5.9 ER308L, ER308LSi, ER309L, ER316L Nickel Alloy Welds Filler Wire SFA-5.14 ERNiCr-3 Notes:
- 1. Dissimilar metal welds between carbon/low-alloy steel and stainless steel are made with a layer (or layers) of 309L, followed by completion of the weld with either 308L, 316L, or 309L. Alternatively, the welds may be completed with Nickel Alloy 82 for the entire weld.
Dissimilar metal welds between nickel-based alloy and carbon/low-alloy steel or stainless steel are performed using Nickel Alloy 82. Welds between carbon steel and low-alloy steel may be made with the filler metals listed for either base material, except that partial penetration welds may also be made with Nickel Alloy 82.
- 2. The maximum carbon content for stainless steel weld filler metal will be less than or equal to 0.035% in accordance with GL 88-01.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.3-2 Revision 0 Section IX, Qualification Standard for Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Welding, Brazing, and Fusing Operators, (Reference 5.3-3), and Section XI, Rules for Inspection and Testing of Components of Light-Water-Cooled Planted, (Reference 5.3-4) and by conformance with RG 1.31, RG 1.50, RG 1.65, Generic Letter 88-01 and NUREG-0313, Revision 2.
10 CFR 50, Appendix A, GDC 4, as it relates to compatibility of components with environmental conditions, by conformance with RG 1.44 and Generic Letter 88-01 and NUREG-0313, Revision 2.
10 CFR 50, Appendix A, GDC 14, as it relates to prevention of rapidly propagating fractures of the RCPB, by compliance with 10 CFR 50, Appendix G and conformance with RG 1.31 and Generic Letter 88-01 and NUREG-0313, Revision 2.
10 CFR 50, Appendix A, GDC 31, as it relates to material fracture toughness, by compliance with 10 CFR 50, Appendix G, and conformance with RG 1.65.
10 CFR 50, Appendix A, GDC 32, as it relates to the requirements for a reactor vessel material surveillance program, by compliance with 10 CFR 50, Appendix H and ASTM International (ASTM) E185, Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels, (Reference 5.3-5). See Subsection 5.3.1.6.1 for discussion on applicable Edition of ASTM E185.
10 CFR 50, Appendix B, Criterion XIII, as it relates to onsite material cleaning control, as discussed in Subsection 5.3.3.4.
10 CFR 50, Appendix G, as it relates to materials testing and acceptance criteria for fracture toughness, by compliance with 10 CFR 50, Appendix H, and conformance with RG 1.65.
10 CFR 50, Appendix H, as it relates to the determination and monitoring of fracture toughness, by compliance with 10 CFR 50, Appendix H, and ASTM E185.
Section 3.1 summarizes compliance with GDC.
5.3.1.1 Material Specifications The materials used in the RCPB, RPV, appurtenances and selected attachments are listed in Table 5.2-2 and Table 5.2-3, with the applicable specifications.
The RPV materials comply with ASME BPVC,Section III; ASME BPVC,Section II, Materials, (Reference 5.3-1); and 10 CFR 50, Appendix G. The RPV materials also meet the additional requirements discussed in the following subsections.
These materials provide adequate strength, fracture toughness, fabricability, and compatibility with the BWR environment. Their suitability has been demonstrated by long-term successful OPEX in reactor service.
Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.3-9 Revision 0 The design and analysis of RPV bolting materials complies with ASME BPVC,Section III, Subsection NB requirements. The RPV main closure bolting materials are specified in Table 5.2-2. The bolting undergoes NDE per ASME BPVC,Section III, NB-2580 before being placed into service. The NDE includes visual examination, magnetic particle or liquid penetrant examination, and ultrasonic examination. The maximum yield strength is specified in Table 5.3-1.
The lateral expansion and the Charpy V values are specified in Table 5.3-1.
The RPV main closure bolting is not metal-plated. Bolting material may have a manganese phosphate (or other acceptable) surface treatment.
Lubricants are permissible provided that they are stable at operating temperatures and are compatible with the bolting and RPV materials and with the surrounding environment. Lubricants are to contain low levels of halogens, sulfur, lead, or other low melting point metals consistent with proven lubricants used in similar nuclear applications. Lubricants containing molybdenum sulfide (disulfide or polysulfide) should not be used for any Safety Class applications. Lubricants shall be approved by GEH prior to use.
During the venting and filling of the RPV and while the head is removed, the studs and stud holes in the reactor vessel shell flange are protected from corrosion and contamination.
5.3.2 Pressure-Temperature Limits The regulations requiring the imposition of pressure-temperature limits on the RCPB are the following:
10 CFR 50.55(i) requires that SSCs subject to the codes and standards in 10 CFR 50.55a must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed. In addition, GDC 1, requires that the codes and standards used to ensure quality products in keeping with the safety function be identified and evaluated to determine their adequacy.
GDC 14, requires that the RCPB be designed, fabricated, erected, and tested in order to have an extremely low probability of abnormal leakage, of rapid failure, and of gross rupture. Likewise, GDC 31, requires, in part, that the RCPB be designed with sufficient margin to ensure that when stressed under operating, maintenance and testing, and postulated accident conditions, the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. In order to assess the structural integrity of the reactor vessel, GDC 32 requires, in part, an appropriate material surveillance program for the reactor vessel beltline region.
The acceptability of the BWRX-300 RCPB pressure-temperature limits is demonstrated by meeting the relevant requirements of the following regulations:
10 CFR 50.55a, as it relates to quality standards for design, and determination and monitoring of material fracture toughness, by compliance with ASME BPVC,Section XI, Nonmandatory Appendix G.
Table 5.2-2a.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.3-16 Revision 0 Table 5.3-1 Reactor Pressure Vessel Material Controls Component Control(s)
Specified limits for RPV materials used in the core beltline limiting forging.
(See Table 5.2-2 Notes, Table 5.3-2, Subsection 5.2.3.3.1, Subsection 5.3.1.2, Subsection 5.3.1.5, and Subsection 5.3.2.1)
Base Material Residual Elements: 0.05% maximum copper, 0.006% maximum phosphorous, and 0.010% maximum sulfur Weld Material Residual Elements: 0.05% maximum copper, 1.0% maximum nickel, 0.008% maximum phosphorous, 0.010% maximum sulfur, and 0.05% maximum vanadium Studs, nuts, and washers for the main closure flange (See Subsection 5.3.1.2 and Subsection 5.3.1.7)
Material specification per Table 5.2-2 having maximum yield strength level of 1034 MPa (150 ksi)
RPV post-weld heat treatment of low-alloy steel welds (See Subsection 5.2.3.3.2, Subsection 5.3.1.2, and Subsection 5.3.3.2) 593°C (1100°F) minimum and not exceeding 635°C (1175°F) is applied to low-alloy steel welds in accordance with NB-4622.
Specific post-weld time and temperature requirements are provided in NB-4622 Toughness of bolting material exceeding 25 mm (1 in.)
diameter (See Subsection 5.2.3.3.1, Subsection 5.3.1.5, and Subsection 5.3.1.7)
Minimum of 61 J (45 ft-lbf) Charpy upper-shelf energy and 0.64 mm (0.025 in.) lateral expansion at the minimum bolt preload temperature per NB-2333 for ASME Class 1 bolting and at or below lowest service temperature per NCD-2332.3 for ASME Class 2 bolting The 61 J (45 ft-lbf) requirement of the ASME BPVC that applies to bolts over 101.6 mm (4 in.) in diameter, is conservatively applied to nominal bolt diameters exceeding 25 mm (1 in.)
RPV Design Data (See Subsection 5.3.3)
The RPV design pressure is 10.342 MPa gauge (1500 psig) and the design temperature is 314.4°C (598°F). The preservice hydrostatic test pressure is 1.25 times the design pressure per NB-6221 Average rate of change of reactor coolant temperature during heatup and cooldown (See Subsection 5.3.2.1)
Not to exceed 111.1°C (200°F) during any one-hour period Reference nil-ductility temperature (See Subsection 5.3.2.1)
-20°C for core beltline shell course
-25°C for nozzle forgings, except nozzles integral with shell forgings
-12°C for upper shell courses and head Table 5.2-2a and Table 5.2-2b Table 5.2-2a
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-5.2.3-2 Receipt Date:
8/25/2025 Question:
In order to make a reasonable assurance finding of the suitability of the material for its application, as they relate to quality standards for design, fabrication, erection, and inspection in accordance with meeting the requirements of GDC 1, 4, 14 and 30 and 10 CFR 50.55a, the materials type/classification for pressure-retaining metals, including weld filler metals that are used for each component in the RCPB should be specified. Table 5.2-2, Reactor Coolant Pressure Boundary Materials, provides the maximum carbon content of wrought austenitic stainless steel materials to minimize stress corrosion susceptibility, but not for the weld filler metal.
Therefore, provide the maximum carbon content for the stainless steel weld filler metal taking into account the influence carbon content has on the susceptibility to stress corrosion cracking in stainless steel in order to make a reasonable assurance finding that the material is compatible with the reactor coolant environment in accordance with GDC1, 4, 14, 30 and 10 CFR 50.55a.
Response
As stated in PSAR Subsection 5.2.3.1, Material Specifications: BWR piping materials and materials processing conforms to Attachment A to Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, and considers the recommendations in NUREG-0313, Revision 2, Technical Report on Material Selection and Processing Guidelines for Boiling Water Reactor (BWR) Coolant Pressure Boundary Piping, related to minimizing Stress Corrosion Cracking (SCC) in susceptible piping. Compliance with Attachment A to Generic Letter 88-01 includes carbon content and ferrite limits for stainless steel base and weld metals, and nickel-based alloy material selection.
A note will be added to Table 5.2-2b to state that the maximum carbon content for stainless steel weld filler metal will be less than or equal to 0.035% in accordance with Generic Letter 88-01.
The table is being split into Table 5.2-2a covering materials and Table 5.2-2b covering weld metal.
The TVA Response for Question Number A-5.2.3-3 will be transmitted separately.
CPA Update on Docket: See attached markup of PSAR Table 5.2-2a and Table 5.2-2b. This markup also includes PSAR changes in response to NRC Audit Question A-5.2.3-1. The CPA markup for Question Number A-5.2.3-3 will be transmitted separately.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 4.5-1 Revision 0 4.5 Reactor Material This section describes the CRD System Structural Materials (Subsection 4.5.1) and the Reactor Internals Materials (Subsection 4.5.3).
4.5.1 Control Rod Drive System Structural Materials This section discusses materials used in the Control Rod Drive Mechanism (CRDM) and extends to the coupling interface with control rod assemblies in the reactor vessel. The materials comprising the fuel and control rod assemblies are described in Section 4.2.
The RCPB components of the CRD, consist of the FMCRD middle flange including the ball check valve, and the lower component housing, which encloses the lower part of the drive. These components are made with 300 series stainless steel materials compatible with the reactor coolant in accordance with ASME BPVC,Section III (Reference 4.5-1). Compliance with RG 1.84, which addresses ASME BPVC,Section III Code Cases, is discussed in Subsection 5.2.1.2, and Code Cases, if used, are listed in Table 5.2-1. The installation bolts used to attach the middle flange and the lower component housing to the CRD housing are low alloy steel material as shown in Table 5.2-2.
The metallic structural components of the CRDM are made from four types of materials: 300 series stainless steel, Nickel-Chrome-Iron Alloy X-750, XM-19 stainless steel and 17-4 PH stainless steel materials. The properties of the non-primary pressure retaining materials selected for the CRDM are equivalent to those given in Parts A, B and D of ASME BPVC,Section II, Materials, (Reference 4.5-2) and Section III (Reference 4.5-1), or are included in RG 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III. The requirements are intended to address the cold work concerns - both cold work from forming operations and abrasive grinding. Surface cold work can be detrimental to SCC resistance of austenitic stainless steels. In conformance with RG 1.44, Control of the Processing and Use of Stainless Steel, cold worked 300 series austenitic stainless steels are not used except for minor forming and straightening. Cold work concerns are controlled by limiting the material hardness, bend radius, or the amount of strain induced by a process. In these uses, the maximum yield strength used is 90 ksi. There are special materials also used for such components as the bayonet coupling (or coupling spud), the latch and latch spring which are made from XM-19 stainless steel, Alloy X-750 and 17-4 PH stainless steel. These materials have been used successfully in operating BWRs.
4.5.1.1 Materials Specifications The CRD systems material specification defines the material requirements for FMCRD components and is intended to support design requirements. The FMCRD material specification is used with the design specification required by ASME BPVC,Section III, NCA-2021.
When applicable, such as for ASME Class 1 components, ASME BPVC,Section III, NB-2300 applies.
A list of materials and specifications is provided in Table 4.5-2.
Table 5.2-2a.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-5 Revision 0 10 CFR 50, Appendix B, Criterion XIII, as it relates to onsite material cleaning control, as discussed in Subsection 5.3.3.4.
10 CFR 50, Appendix G, as it relates to materials testing and acceptance criteria for fracture toughness of the RCPB as discussed in Subsection 5.2.3.3.1 10 CFR 50.55a, as it relates to quality standards applicable to the RCPB, by compliance with ASME BPVC,Section III and conformance with RG 1.31, RG 1.50, RG 1.71, Generic Letter 88-01, and NUREG-0313, Revision 2. RG 1.43 does not apply as discussed in Subsection 5.3.1.4.
Section 3.1 summarizes compliance with GDC.
5.2.3.1 Material Specifications This subsection discusses the specifications for pressure-retaining ferritic materials, austenitic stainless steels, and nonferrous metals, including bolting and weld materials, that are used for each RCPB component, e.g., vessels, piping, and valves. The adequacy and suitability of the ferritic materials, austenitic stainless steels, and nonferrous metals specified for the above application are discussed in Subsection 5.2.3.1 through Subsection 5.2.3.4.
Table 5.2-2 lists the principal pressure-retaining materials and the appropriate material specifications for the RCPB components and welds. RCPB materials comply with ASME BPVC,Section III, NB-2000 for ASME Class 1 components and NCD-2000 for ASME Class 2 components and ASME BPVC,Section II, Materials, (Reference 5.2-4). RCPB ferritic materials comply with 10 CFR 50, Appendix G. Compliance with RG 1.84, which addresses ASME BPVC,Section III Code Cases, is discussed in Subsection 5.2.1.2, and Code Cases, if used, are listed Table 5.2-1. No exceptions are taken to the material specifications of the ASME BPVC.
BWR piping materials and materials processing conforms to Attachment A to Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, and considers the recommendations in NUREG-0313, Revision 2, Technical Report on Material Selection and Processing Guidelines for Boiling Water Reactor (BWR) Coolant Pressure Boundary Piping, related to minimizing Stress Corrosion Cracking (SCC) in susceptible piping. Compliance with Attachment A to Generic Letter 88-01 includes carbon content and ferrite limits for stainless steel base and weld metals, and nickel-based alloy material selection. Further discussion of the materials associated with the RPV pressure-retaining components is provided in Subsection 5.3.1.
Cast austenitic stainless steels used in the RCPB have material specifications such as maximum ferrite levels that ensure adequate fracture toughness over the design life to support use of the material under the expected environmental conditions, e.g., exposure to the reactor coolant operating temperatures.
Nickel-chromium-iron alloys such as Alloy 600M consistent with Code Case N-580-2 along with Alloy 82 weld metal are used in the RCPB. Alloy 600 and associated weld materials are not used in the RCPB, due to their susceptibility to SCC.
Table 5.2-2a and Table 5.2-2b list
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-6 Revision 0 RPV and RPV Appurtenances Table 5.2-2 lists materials and specifications for the RPV and RPV appurtenances which form part of the RCPB. This includes RPV nozzles, stub tubes, CRD flanges/housings, and incore instrumentation housings and their associated RCPB components.
RPV Attachments Table 5.2-3 lists materials for selected RPV attachments. RPV attachments are those components welded directly to the RPV but do not form part of the RCPB. Many of these attachments are internal to the RPV and in contact with reactor coolant, while some are external to the RPV and not in contact with reactor coolant.
RPV attachments internal to the RPV and in contact with reactor coolant include the core support legs.
RPV attachments external to the RPV and not in contact with reactor coolant include the vessel stabilizer brackets and the vessel support skirt.
RPV Internals Materials for RPV internals, such as steam dryer assembly, chimney, and core support structures, are listed in Table 4.5-1. Discussion of RPV internals in this chapter is limited to their compatibility with reactor coolant and certain corrosion aspects or effects on water chemistry.
5.2.3.2 Compatibility with Reactor Coolant General corrosion and SCC induced by impurities in the reactor coolant can cause failures of the RCPB. The chemistry of the reactor coolant and any additives whose function is to control corrosion are discussed in Subsection 5.2.3.2.1 and Subsections 9.3.7, 9.3.8, and 9.3.9. The compatibility of the materials of construction employed in the RCPB with the reactor coolant, contaminants, or radiolytic products to which the system is exposed was considered by employing proven materials and processes. The extent of time-dependent general corrosion and other degradation mechanisms of materials of construction in contact with the reactor coolant was considered. Similarly, consideration was given to the control of the use of austenitic stainless steels in the sensitized condition.
5.2.3.2.1 Chemistry of Reactor Coolant Proper material selection, fabrication and process controls, and control of water chemistry ensures the integrity of the RCPB. This subsection provides a brief review of the relationships between water chemistry variables and RCPB materials performance, fuel performance, and plant radiation fields. Electric Power Research Institute (EPRI) guidelines in EPRI BWRVIP-190, BWR Vessel and Internals Project, Volume 1: BWR Water Chemistry Guidelines - Mandatory, Needed and Good Practice Guidance and Volume 2: BWR Water Chemistry Guidelines - Technical Basis, (Reference 5.2-1) are used as inputs to develop the basis for the BWRX-300 water chemistry program. The BWRX-300 water chemistry program may continue to be revised based on new EPRI guidelines, and as operational experience is gained.
Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-11 Revision 0 For the reactor internals, austenitic stainless steel is primarily used to minimize levels of radiation from corrosion products generated in the core. Where higher strength requirements than those for austenitic stainless steels are imposed, nickel-based alloys, nitrogen strengthened alloys, or precipitation hardening stainless steels are used.
The stainless steel structural materials used for the BWRX-300 reactor core internals are primarily wrought materials along with their weld metals, although cast austenitic stainless steels are used as well.
In summary, construction materials of the RCPB and reactor internals exposed to the reactor coolant consist of the following:
Solution-annealed austenitic stainless steels (both wrought and cast), Types 304, 304L, 316, 316L, CF3M, CF8M, and XM-19
Solution-annealed tempered martensitic stainless steels, Type 410
Nickel-based alloys (including niobium-modified Alloy 600M with filler metal Alloy 82, Alloy 625 with filler metal ERNICMO-3, Alloy 718, and X-750)
Precipitation hardening stainless steels such as Type 630 and Type XM-13
Carbon steels and low-alloy steels Table 5.2-2 lists materials and material specifications for RCPB components. Table 4.5-1 lists material and material specifications for reactor internals.
Suitability of Materials of Construction Exposed to Reactor Coolant Materials selection and processing controls address materials degradation issues in the RCPB such as SCC, general corrosion, and FAC. The specific categories of materials that are used are wrought austenitic stainless steels, cast austenitic stainless steels, martensitic stainless steels, precipitation hardening stainless steels, and nickel-based alloys along with their weld metals used as part of construction. General corrosion for the plant components is managed in the design process based on the ASME BPVC, which sets minimum corrosion allowances.
Enhanced IGSCC resistance has been achieved through the use of IGSCC resistant materials such as Type 316L with controlled composition stainless steel, XM-19, and stabilized nickel-based niobium-modified Alloy 600M and Alloy 82.
BWRX-300 materials provide adequate strength, fracture toughness, fabricability, and compatibility with the BWR environment. Their suitability has been demonstrated by long-term successful OPEX in reactor service.
Material Contamination and Transgranular Stress Corrosion Cracking The requirements of GDC 4 relative to compatibility of components with environmental conditions are met by compliance with the applicable provisions of the ASME BPVC such as corrosion allowances and by conformance with the recommendations of RG 1.44, Control of the Processing and Use of Stainless Steel. Material contamination that can lead to transgranular Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-29 Revision 0 Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 1 of 7)
Component Form Material (Note 1)
ASME Specification Reactor Pressure Vessel RPV Shells and Heads Forging Plate Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1, SA-533/SA-533M Type B Class 1 RPV Shell and Head Flange Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 RPV Main Closure Bolting Bolting Low-Alloy Steel SA-540/SA-540M Gr B23 or B24 Class 3 RPV Standard Flange Bolting Bolting Low-Alloy Steel SA-540/SA-540M Gr B23 or B24 Class 3 RPV Nozzles Forging Low-Alloy steel SA-508/SA-508M Gr 3 Class 1 RPV Level Instrumentation Nozzles Forging Stainless steel or Ni-Cr-Fe SA-182/SA-182M, SA-965/SA-965M Gr F316/F316L, or SB-168, SB-166, or SB-564 (Modified per ASME Code Case N-580-2)
Main Steam Reactor Isolation Valves (MSRIV)
MSRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSRIV Cover Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSRIV Poppet Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
MSRIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 MSRIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Main Steam Containment Isolation Valves (MSCIV)
MSCIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSCIV Cover Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSCIV Poppet Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 MSCIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
MSCIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-30 Revision 0 MSCIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Main Steam Line (RCPB Portion)
MS Piping Seamless Forged & Bored Carbon steel Low-Alloy steel SA-333/SA-333M Gr 6, SA-335/SA-335M Gr P22, SA-369/SA-369M Gr FP2 MS Contour Nozzle Forging Low-Alloy steel SA-508/SA-508M Gr 3 Class 1 MS 200 mm 1500 lb. Large Groove Flange Forging Carbon steel SA-350/SA-350M Gr LF2 Class 1 MS 50 mm Special Nozzle Forging Carbon steel SA-350/SA-350M Gr LF2 Class 1 MS Elbow Seamless Fitting Carbon steel SA-420/SA-420M Gr WPL6, SA-508/SA-508M Gr 1 MS Head Fitting/
Penetration Piping Forging Carbon steel SA-350/SA-350M Gr LF2 Class 1 MS Other Fittings Forging Low-Alloy steel SA-234/SA-234M Gr WP22, SA-336/SA-336M Gr F22 Reactor Pressure Vessel Head Vent Head Vent Piping Piping Stainless Steel SA-312/SA-312M Gr TP316L Head Vent to MSL Valves Forging Carbon Steel SA-350/SA-350M Gr LF2 Class 1 Head Vent to Quench Tank Boundary Valves Forging Stainless Steel SA-182/SA-182M Gr F316/F316L RPV Appurtenances Control Rod Drive Flanges/Housings CRD Housings Forging Stainless steel SA-182/SA-182M, SA-965/SA-965M Gr F304/F304L/F316/F316L Lower Component Housings Forging Stainless Steel SA-965/SA-965M Gr F304 Middle Flange Forging Stainless Steel SA-965/SA-965M Gr F304 Upper Flange Forging Stainless Steel SA-965/SA-965M Gr F316 Ball Check Valve Retainer Bar and Shapes Stainless Steel SA-479/SA-479M Type 316 CRD Spool Piece (if used)
Forging Stainless steel SA-965/SA-965M Gr F304/F304L Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 2 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-31 Revision 0 CRD Mounting Bolts Bolting Alloy steel SA-193/SA-193M Gr B7 CRD Penetration Stub Tubes Bar, Seamless Pipe, Forging Ni-Cr-Fe SB-166, SB-167, SB-564 (Modified per ASME Code Case N-580-2)
RPV Appurtenances Incore Instrumentation Housings Incore Instrumentation Nozzles Forging Bar, Seamless Pipe, Forging Stainless steel or Ni-Cr-Fe SA-182/SA-182M, SA-965/SA-965M Gr F304/F304L/F316/F316L or SB-166, SB-168, or SB-564 (Modified per ASME Code Case N-580-2)
Incore Instrumentation Stub Tubes Bar, Seamless Pipe, Forging Ni-Cr-Fe SB-166, SB-168, or SB-564 (Modified per ASME Code Case N-580-2)
Incore Housings Seamless Pipe, Forging Stainless Steel SA-312/SA-312M Gr 304/304L or, SA-182/SA-182M Gr F316/F316L Isolation Condenser System Piping and Components IC Steam Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
IC Standby Gas Purge Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
IC Cover Bolting (Note 3)
Forging Stainless steel SA-705/SA-705M Type 630 IC Drum and Drum Covers Forging Ni-Cr-Mo-Nb SB-564 Gr 1 UNS N06625 (Note 8)
IC Condensate Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
ICS Recombiner Vessel Forging Stainless steel SA-182/SA-182M, SA-965/SA-965M Gr F316 ICS Recombiner Piping Pipe or tubing Ni-Cr-Mo-Cb SB-444 UNS N06625 (Note 8)
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 3 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-32 Revision 0 Isolation Condenser System Isolation Condenser Supply Reactor Isolation Valve (ICSRIV)
ICSRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICSRIV Disc Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICSRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Isolation Condenser System Isolation Condenser Return Reactor Isolation Valve (ICRRIV)
ICRRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICRRIV Disc Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICRRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Isolation Condenser System ICS to SDC Containment Isolation Valves (CIV)
ICS CIV Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L SDC Interface Redundant Isolation Valve (ICS Outboard CIV)
Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L Isolation Condenser System Other ICS Purge Valves Forging or Casting Stainless Steel SA-182/SA-182M Gr F316/F316L, SA-351/SA-351M Gr CF3M ICS Purge Piping Seamless Pipe Stainless Steel SA-312/SA-312M Gr TP316L ICS Purge Fittings Forging Stainless Steel SA-182/SA-182M Gr F316/F316L SA-403/SA-403M Gr WP 316/
316L ICS Condensate Return Valves (Note 7)
(Note 7)
(Note 7)
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 4 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-33 Revision 0 Condensate and Feedwater Heating System Feedwater Piping (RCPB Portion)
FW Piping Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr TP316L FW Fittings Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr TP316L Condensate and Feedwater Heating System Feedwater Reactor Isolation Valves (FWRIV)
FWRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 FWRIV Cover Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 FWRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
FWRIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 FWRIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Condensate and Feedwater Heating System Feedwater Containment Isolation Valves (FWCIV)
FWCIV Body Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L FWCIV Cover Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L FWCIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
FWCIV Body Bolting Bolting Alloy steel SA-540/SA-540M Gr B23 Class 5 FWCIV Hex Nuts Bolting Nuts Alloy steel SA-194/SA-194M Gr 7 Reactor Water Cleanup System (RCPB Portion)
CUW Piping Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr TP316L CUW Fittings Forging Stainless Steel SA-182/SA-182M, SA-965/SA-965M Gr F316/F316L, SA-403/SA-403M Gr WP 316/
316L Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 5 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-34 Revision 0 CUW RIV Body (Note 4)
Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 CUW RIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
CUW CIV Body Casting Stainless Steel SA-351/SA-351M Gr CF8M/CF3M CUW CIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Boron Injection System (RCPB Portion)
BIS CIV Body Forging Stainless Steel SA-182/SA-182M Gr F316/F316L BIS CIV Disc Forging Stainless Steel SA-182/SA-182M Gr F316/F316L BIS CIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
BIS Check Valve Body Forging Stainless Steel SA-182/SA-182M Gr F316/F316L BIS Piping Seamless Forged & Bored Stainless Steel SA-312/SA-312M Gr Type 316/316L BIS Fittings Seamless Forged & Bored Stainless Steel SA-182/SA-182M Gr F316/F316L Additional RCPB Bolting Material Flanges, Covers, and Bonnets Bolting Stud or Bolting Alloy & Carbon Steels SA-354, SA-449 Welding Filler Metals Base Material Filler Metal Type ASME Classification (Note 6)
Carbon Steel Welds Covered Electrode or Filler Metal SFA-5.1, SFA-5.18 Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 6 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-35 Revision 0 Low-Alloy Steel Welds Covered Electrode or Filler Metal SFA-5.5, SFA-5.23, SFA-5.28 Low-Alloy Steel Piping Welds (21/4 Cr-1 Mo)
Covered Electrode or Filler Metal SFA-5.5, SFA-5.28 Stainless Steel Welds Covered Electrode or Filler Metal SFA-5.4, SFA-5.9 Nickel-Based Alloy Welds Filler Wire SFA-5.14 Notes:
1.
Carbon content of RCPB wrought austenitic stainless steel (304/304L/316/316L) is 0.02%
maximum.
2.
To reduce risk of cracking in valve stems, the minimum nominal aging temperature in precipitation hardened stainless steels shall be 580°C (H1075) or higher.
3.
SA-705/SA-705M Type 630 bolting is limited to the H1100 condition per Table 4 of ASME BPVC,Section II, Part D.
4.
Any wetted surfaces in carbon or low-alloy steel bodied valves in the CUW are clad with stainless steel.
5.
Reference Table 5.3-1 and Subsection 5.3.1.5 for core beltline material composition limits.
6.
Dissimilar metal welds between carbon/low-alloy steel and stainless steel are made with a layer (or layers) of 309L, followed by completion of the weld with either 308L, 316L, or 309L.
Alternatively, the welds may be completed with Nickel Alloy 82 for the entire weld. Dissimilar metal welds between nickel-based alloy and carbon/low-alloy steel or stainless steel are performed using Nickel Alloy 82. Welds between carbon steel and low-alloy steel may be made with the filler metals listed for either base material, except that partial penetration welds may also be made with Nickel Alloy 82.
7.
Information currently not available from component supplier.
8.
Code Case N-943 for use of Alloy 625 (UNS N06625) is provided in Table 5.2-1.
9.
The minimum tempering temperature shall be 595°C (~1100°F) for a minimum of two hours.
The maximum hardness following final heat treatment shall be Rockwell C 28 for Type 410.
- 10. Items fabricated from these materials are wetted by reactor coolant but may not necessarily be part of the design ASME BPVC pressure boundary. Therefore, some of the specific material specifications or material specification grades identified under this category are not required to be listed in the ASME BPVC,Section II, Part D, Table 2A.
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 7 of 7)
Component Form Material (Note 1)
ASME Specification Table 5.2-2b [See INSERT next page]
Table 5.2-2a Not Used.
Table 5.2-2b Reactor Coolant Pressure Boundary Welding Filler Metals Base Material Filler Metal Type ASME Classification AWS Classification Carbon Steel Welds Covered Electrode SFA-5.1 E7018 Filler Metal SFA-5.18 ER70S-2, ER70S-3, ER70S-6 Low Alloy Steel Welds (Note 1)
Covered Electrode SFA-5.5 E9018-B3 Filler Metal SFA-5.23 EF2, EF3 or other mutually agreed filler metal with appropriate chemistry selected for base metal composition and mechanical property requirements including toughness SFA-5.28 ER90S-XX with appropriate chemistry selected for base metal composition, shielding gas, and mechanical property requirements including toughness Low Alloy Steel Piping Welds (21/4 Cr-Mo)
Covered Electrode SFA-5.5 E9018-B3 Filler Metal SFA-5.28 ER90S-XX with appropriate chemistry selected for base metal composition, shielding gas, and mechanical property requirements including toughness Stainless Steel Welds (Note 2)
Covered Electrode or Filler Metal SFA-5.4 E308L-XX, E309L-XX, E316L-XX Filler Metal SFA-5.9 ER308L, ER308LSi, ER309L, ER316L Nickel Alloy Welds Filler Wire SFA-5.14 ERNiCr-3 Notes:
- 1. Dissimilar metal welds between carbon/low-alloy steel and stainless steel are made with a layer (or layers) of 309L, followed by completion of the weld with either 308L, 316L, or 309L. Alternatively, the welds may be completed with Nickel Alloy 82 for the entire weld.
Dissimilar metal welds between nickel-based alloy and carbon/low-alloy steel or stainless steel are performed using Nickel Alloy 82. Welds between carbon steel and low-alloy steel may be made with the filler metals listed for either base material, except that partial penetration welds may also be made with Nickel Alloy 82.
- 2. The maximum carbon content for stainless steel weld filler metal will be less than or equal to 0.035% in accordance with GL 88-01.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.3-2 Revision 0 Section IX, Qualification Standard for Welding, Brazing, and Fusing Procedures; Welders; Brazers; and Welding, Brazing, and Fusing Operators, (Reference 5.3-3), and Section XI, Rules for Inspection and Testing of Components of Light-Water-Cooled Planted, (Reference 5.3-4) and by conformance with RG 1.31, RG 1.50, RG 1.65, Generic Letter 88-01 and NUREG-0313, Revision 2.
10 CFR 50, Appendix A, GDC 4, as it relates to compatibility of components with environmental conditions, by conformance with RG 1.44 and Generic Letter 88-01 and NUREG-0313, Revision 2.
10 CFR 50, Appendix A, GDC 14, as it relates to prevention of rapidly propagating fractures of the RCPB, by compliance with 10 CFR 50, Appendix G and conformance with RG 1.31 and Generic Letter 88-01 and NUREG-0313, Revision 2.
10 CFR 50, Appendix A, GDC 31, as it relates to material fracture toughness, by compliance with 10 CFR 50, Appendix G, and conformance with RG 1.65.
10 CFR 50, Appendix A, GDC 32, as it relates to the requirements for a reactor vessel material surveillance program, by compliance with 10 CFR 50, Appendix H and ASTM International (ASTM) E185, Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels, (Reference 5.3-5). See Subsection 5.3.1.6.1 for discussion on applicable Edition of ASTM E185.
10 CFR 50, Appendix B, Criterion XIII, as it relates to onsite material cleaning control, as discussed in Subsection 5.3.3.4.
10 CFR 50, Appendix G, as it relates to materials testing and acceptance criteria for fracture toughness, by compliance with 10 CFR 50, Appendix H, and conformance with RG 1.65.
10 CFR 50, Appendix H, as it relates to the determination and monitoring of fracture toughness, by compliance with 10 CFR 50, Appendix H, and ASTM E185.
Section 3.1 summarizes compliance with GDC.
5.3.1.1 Material Specifications The materials used in the RCPB, RPV, appurtenances and selected attachments are listed in Table 5.2-2 and Table 5.2-3, with the applicable specifications.
The RPV materials comply with ASME BPVC,Section III; ASME BPVC,Section II, Materials, (Reference 5.3-1); and 10 CFR 50, Appendix G. The RPV materials also meet the additional requirements discussed in the following subsections.
These materials provide adequate strength, fracture toughness, fabricability, and compatibility with the BWR environment. Their suitability has been demonstrated by long-term successful OPEX in reactor service.
Table 5.2-2a
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.3-9 Revision 0 The design and analysis of RPV bolting materials complies with ASME BPVC,Section III, Subsection NB requirements. The RPV main closure bolting materials are specified in Table 5.2-2. The bolting undergoes NDE per ASME BPVC,Section III, NB-2580 before being placed into service. The NDE includes visual examination, magnetic particle or liquid penetrant examination, and ultrasonic examination. The maximum yield strength is specified in Table 5.3-1.
The lateral expansion and the Charpy V values are specified in Table 5.3-1.
The RPV main closure bolting is not metal-plated. Bolting material may have a manganese phosphate (or other acceptable) surface treatment.
Lubricants are permissible provided that they are stable at operating temperatures and are compatible with the bolting and RPV materials and with the surrounding environment. Lubricants are to contain low levels of halogens, sulfur, lead, or other low melting point metals consistent with proven lubricants used in similar nuclear applications. Lubricants containing molybdenum sulfide (disulfide or polysulfide) should not be used for any Safety Class applications. Lubricants shall be approved by GEH prior to use.
During the venting and filling of the RPV and while the head is removed, the studs and stud holes in the reactor vessel shell flange are protected from corrosion and contamination.
5.3.2 Pressure-Temperature Limits The regulations requiring the imposition of pressure-temperature limits on the RCPB are the following:
10 CFR 50.55(i) requires that SSCs subject to the codes and standards in 10 CFR 50.55a must be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed. In addition, GDC 1, requires that the codes and standards used to ensure quality products in keeping with the safety function be identified and evaluated to determine their adequacy.
GDC 14, requires that the RCPB be designed, fabricated, erected, and tested in order to have an extremely low probability of abnormal leakage, of rapid failure, and of gross rupture. Likewise, GDC 31, requires, in part, that the RCPB be designed with sufficient margin to ensure that when stressed under operating, maintenance and testing, and postulated accident conditions, the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized. In order to assess the structural integrity of the reactor vessel, GDC 32 requires, in part, an appropriate material surveillance program for the reactor vessel beltline region.
The acceptability of the BWRX-300 RCPB pressure-temperature limits is demonstrated by meeting the relevant requirements of the following regulations:
10 CFR 50.55a, as it relates to quality standards for design, and determination and monitoring of material fracture toughness, by compliance with ASME BPVC,Section XI, Nonmandatory Appendix G.
Table 5.2-2a.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.3-16 Revision 0 Table 5.3-1 Reactor Pressure Vessel Material Controls Component Control(s)
Specified limits for RPV materials used in the core beltline limiting forging.
(See Table 5.2-2 Notes, Table 5.3-2, Subsection 5.2.3.3.1, Subsection 5.3.1.2, Subsection 5.3.1.5, and Subsection 5.3.2.1)
Base Material Residual Elements: 0.05% maximum copper, 0.006% maximum phosphorous, and 0.010% maximum sulfur Weld Material Residual Elements: 0.05% maximum copper, 1.0% maximum nickel, 0.008% maximum phosphorous, 0.010% maximum sulfur, and 0.05% maximum vanadium Studs, nuts, and washers for the main closure flange (See Subsection 5.3.1.2 and Subsection 5.3.1.7)
Material specification per Table 5.2-2 having maximum yield strength level of 1034 MPa (150 ksi)
RPV post-weld heat treatment of low-alloy steel welds (See Subsection 5.2.3.3.2, Subsection 5.3.1.2, and Subsection 5.3.3.2) 593°C (1100°F) minimum and not exceeding 635°C (1175°F) is applied to low-alloy steel welds in accordance with NB-4622.
Specific post-weld time and temperature requirements are provided in NB-4622 Toughness of bolting material exceeding 25 mm (1 in.)
diameter (See Subsection 5.2.3.3.1, Subsection 5.3.1.5, and Subsection 5.3.1.7)
Minimum of 61 J (45 ft-lbf) Charpy upper-shelf energy and 0.64 mm (0.025 in.) lateral expansion at the minimum bolt preload temperature per NB-2333 for ASME Class 1 bolting and at or below lowest service temperature per NCD-2332.3 for ASME Class 2 bolting The 61 J (45 ft-lbf) requirement of the ASME BPVC that applies to bolts over 101.6 mm (4 in.) in diameter, is conservatively applied to nominal bolt diameters exceeding 25 mm (1 in.)
RPV Design Data (See Subsection 5.3.3)
The RPV design pressure is 10.342 MPa gauge (1500 psig) and the design temperature is 314.4°C (598°F). The preservice hydrostatic test pressure is 1.25 times the design pressure per NB-6221 Average rate of change of reactor coolant temperature during heatup and cooldown (See Subsection 5.3.2.1)
Not to exceed 111.1°C (200°F) during any one-hour period Reference nil-ductility temperature (See Subsection 5.3.2.1)
-20°C for core beltline shell course
-25°C for nozzle forgings, except nozzles integral with shell forgings
-12°C for upper shell courses and head Table 5.2-2a and Table 5.2-2b Table 5.2-2a
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-5.2.3-3 Receipt Date:
8/25/2025 Question:
In order to make a reasonable assurance finding of the suitability of the material for its application, as they relate to quality standards for design, fabrication, erection, and inspection in accordance with meeting the requirements of GDC 1, 4, 14 and 30 and 10 CFR 50.55a, the materials type/classification for pressure-retaining metals that are used for each component in the RCPB should be specified. Note 7 to Table 5.2-2, Reactor Coolant Pressure Boundary Materials, states that the material information is currently not available from the component supplier for isolation condenser system condensate return valves. To make a reasonable assurance finding that the material is compatible with the reactor coolant environment in accordance with GDC1, 4, 14, 30 and 10 CFR 50.55a, a range of material types that could be used should be specified for the isolation condenser system condensate return valves.
Response
PSAR Table 5.2-2 will be revised to provide the material types for the isolation condenser system condensate return valves.
CPA Update on Docket: See attached markup of PSAR Table 5.2-2 that is revised based on NRC Feedback.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-32 Revision 0 Isolation Condenser System Isolation Condenser Supply Reactor Isolation Valve (ICSRIV)
ICSRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICSRIV Disc Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICSRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Isolation Condenser System Isolation Condenser Return Reactor Isolation Valve (ICRRIV)
ICRRIV Body Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICRRIV Disc Forging Low-Alloy Steel SA-508/SA-508M Gr 3 Class 1 ICRRIV Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10),
SA-479/SA-479M Type XM-19 or 410 Condition 2 (Note 9)
Isolation Condenser System ICS to SDC Containment Isolation Valves (CIV)
ICS CIV Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L SDC Interface Redundant Isolation Valve (ICS Outboard CIV)
Forging Stainless Steel SA-182/SA-182M Gr F304/F304L/F316/F316L Isolation Condenser System Other ICS Purge Valves Forging or Casting Stainless Steel SA-182/SA-182M Gr F316/F316L, SA-351/SA-351M Gr CF3M ICS Purge Piping Seamless Pipe Stainless Steel SA-312/SA-312M Gr TP316L ICS Purge Fittings Forging Stainless Steel SA-182/SA-182M Gr F316/F316L SA-403/SA-403M Gr WP 316/
316L ICS Condensate Return Valves (Note 7)
(Note 7)
(Note 7)
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 4 of 7)
Component Form Material (Note 1)
ASME Specification see INSERT on next page
A-5.2.3-3 INSERT for PSAR Table 5.2-2:
Isolation Condenser System Condensate Return Valves (CRV)
ICS CRV Body Forging Stainless Steel SA-182/SA-182M Gr F304/F304L, Gr F316/F316L (Dual Grade)
ICS CRV Disc Forging Stainless Steel SA-182/SA-182M Gr F304/F304L, Gr F316/F316L (Dual Grade)
ICS CRV Valve Stem Rod or Bar Stainless Steel SA-564/SA-564M Type 630 or XM-13 (Notes 2,10) or SA-479/SA-479M Type XM-19
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-35 Revision 0 Low-Alloy Steel Welds Covered Electrode or Filler Metal SFA-5.5, SFA-5.23, SFA-5.28 Low-Alloy Steel Piping Welds (21/4 Cr-1 Mo)
Covered Electrode or Filler Metal SFA-5.5, SFA-5.28 Stainless Steel Welds Covered Electrode or Filler Metal SFA-5.4, SFA-5.9 Nickel-Based Alloy Welds Filler Wire SFA-5.14 Notes:
1.
Carbon content of RCPB wrought austenitic stainless steel (304/304L/316/316L) is 0.02%
maximum.
2.
To reduce risk of cracking in valve stems, the minimum nominal aging temperature in precipitation hardened stainless steels shall be 580°C (H1075) or higher.
3.
SA-705/SA-705M Type 630 bolting is limited to the H1100 condition per Table 4 of ASME BPVC,Section II, Part D.
4.
Any wetted surfaces in carbon or low-alloy steel bodied valves in the CUW are clad with stainless steel.
5.
Reference Table 5.3-1 and Subsection 5.3.1.5 for core beltline material composition limits.
6.
Dissimilar metal welds between carbon/low-alloy steel and stainless steel are made with a layer (or layers) of 309L, followed by completion of the weld with either 308L, 316L, or 309L.
Alternatively, the welds may be completed with Nickel Alloy 82 for the entire weld. Dissimilar metal welds between nickel-based alloy and carbon/low-alloy steel or stainless steel are performed using Nickel Alloy 82. Welds between carbon steel and low-alloy steel may be made with the filler metals listed for either base material, except that partial penetration welds may also be made with Nickel Alloy 82.
7.
Information currently not available from component supplier.
8.
Code Case N-943 for use of Alloy 625 (UNS N06625) is provided in Table 5.2-1.
9.
The minimum tempering temperature shall be 595°C (~1100°F) for a minimum of two hours.
The maximum hardness following final heat treatment shall be Rockwell C 28 for Type 410.
- 10. Items fabricated from these materials are wetted by reactor coolant but may not necessarily be part of the design ASME BPVC pressure boundary. Therefore, some of the specific material specifications or material specification grades identified under this category are not required to be listed in the ASME BPVC,Section II, Part D, Table 2A.
Table 5.2-2 Reactor Coolant Pressure Boundary Materials (Sheet 7 of 7)
Component Form Material (Note 1)
ASME Specification
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-5.2-4 Receipt Date:
9/15/2025 Question:
PSAR Section 5.2.4, Reactor Coolant Pressure Boundary Inservice Inspection and Testing, provides how the BWRX-300 design meets the regulations in 10 CFR 50.55a for preservice inspection (PSI) and inservice inspection (ISI). In order to make a reasonable assurance finding for the CP review regarding the design and construction of the facility to facilitate PSI and ISI of the reactor coolant pressure boundary (RCPB) in accordance with meeting the requirements of 10 CFR 50.55a and GDC 1 and 32, the following information is needed to verify the specific requirements that will be used to develop the PSI and ISI program as it relates to the BWRX-300 design:
- a. That the RCPB components are designed and provided with the access necessary to perform the required preservice and inservice inspections set forth in Section III and XI of the ASME Code, including the examinations required by ASME BPVC,Section XI. As part of this confirmation of the inspectability of the design, the following should be considered:
- i.
PSI for Class 1 RCPB will be specified in the design specification and be in accordance with ASME Section III, NB-5281 and NB-5282 and Section XI, Table IWB-2500-1, and will be implemented during construction.
ii.
Additional requirements such as NUREG-0313, Revision 2, and Generic Letter 88-01 that may include additional inspections.
iii. Any augmented PSI and ISI for Class 1 RCPB welds.
iv. The ability to perform system hydrostatic tests and system leakage tests.
- b. In addition, provide the requirements for accessibility that the design will use in order to meet the requirements of 10 CFR 50.55a(g)(2), Accessibility requirements, and 10 CFR 50.55a(g)(3), Preservice examination requirements. (e.g., access to both sides of the weld, removable insulation, etc.).
Response: RCPB components will meet accessibility requirements and Preservice Inspection requirements as required by 10 CFR 50.55a(g)(2), Accessibility requirements, and 10 CFR 50.55a(g)(3), Preservice examination requirements. Accessibility requirements include access to both sides of the weld, removable insulation, etc. Subsection 5.2.4 will be revised to further define inspection and accessibility requirements.
CPA Update on Docket:
See attached markup of PSAR Subsection 5.2.4.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 5.2-18 Revision 0 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing The BWRX-300 meets the relevant requirements of the following regulations:
10 CFR 50.55a, as it relates to the requirements for inspecting and testing ASME Class 1 and ASME Class 2 components of the RCPB as specified in ASME BPVC,Section XI. The BWRX-300 ISI program complies with the rules of 10 CFR 50.55a as indicated in Subsection 5.2.1.1 and Subsection 5.2.1.2.
10 CFR 50, Appendix A, GDC 32, as it relates to periodic inspection and testing of the RCPB.
BWRX-300 is designed for the performance of preservice inspection and ISI of ASME Class 1 and ASME Class 2 components within the RCPB in accordance with ASME BPVC,Section XI pursuant to 10 CFR 50.55a(g). The edition of ASME BPVC,Section XI is discussed in Subsection 5.2.1.1.
The RCPB consists of ASME Class 1 and ASME Class 2 components as discussed in Subsection 5.2.1.1 and illustrated on Figure 5.1-2. RCPB components which require inspections and testing to satisfy ASME BPVC,Section XI are examined by appropriate inspection and testing methods provided in ASME BPVC,Section XI. Examination categories, methods, and procedures for RCPB components are in accordance with ASME BPVC,Section XI, Subsection IWA, Subsection IWB, and Subsection IWC. Examination results are evaluated and accepted in accordance with Subsection IWA, Subsection IWB, and Subsection IWC.
5.2.5 Reactor Coolant Pressure Boundary Leakage Detection RCPB leakage detection systems are designed to provide a means of detecting and, to the extent practical, identifying the source of the reactor coolant leakage. Although NEDC-33910P-A, BWRX-300 Reactor Pressure Vessel Isolation and Overpressure Protection, (Reference 5.2-3) discusses conformance to RG 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, the BWRX-300 design includes only unidentified leakage that is potentially from the RCPB. As identified leakage is not defined, there is no leakage separation.
These are alternatives to RG 1.45 General Positions 1.1 and 1.2.
The RCPB leakage detection system detects leakage after an earthquake for an early indication of degradation so that corrective action can be taken before such degradation becomes severe enough to result in a leakage rate greater than the capability of the makeup system to replenish the coolant loss.
Historically, separation of identified leakage was accomplished by routing leakage from known sources, such as the recirculation pump seal leakage, to an equipment drain system instrumented to quantify the total identified leakage rate.
The BWRX-300 is a natural circulation design and has no pumps within the RCPB. Eliminating pump seal leakage significantly reduces anticipated leakage which could be identified.
Additionally, the BWRX-300 does not make use of traditional safety or relief valves within the RCPB.
RCPB components are designed and provided with the access necessary to perform the required preservice and inservice inspections set forth in ASME BPVC,Section III and Section XI, including the examinations required by ASME BPVC,Section XI. The following are considered in the inspectability of the design:
PSI for Class 1 RCPB are specified in the design specification and are in accordance with ASME BPVC,Section III, NB-5281 and NB-5282 and ASME BPVC,Section XI, Table IWB-2500-1, and will be implemented during construction.
Additional requirements such as NUREG-0313, Revision 2, and Generic Letter 88-01 that may include additional inspections.
Any augmented PSI and ISI for Class 1 RCPB welds, e.g., augmented inspections in Subsection 3.6.1.1.3.
The ability to perform system hydrostatic tests and system leakage tests.
RCPB components and their supports are designed and provided with the access necessary to perform the required preservice and inservice examinations set forth in ASME BPVC,Section III and Section XI, e.g., IWA-2200(b) for weld surface preparation for NDE, or the optional ASME Code Cases listed in Regulatory Guide 1.147. Adequate clearance is provided in accordance with ASME BPVC,Section XI, IWA-1500 by the design and arrangement of system components.
RCPB components and their supports meet the preservice examination requirements set forth in ASME BPVC,Section III and Section XI or the optional ASME Code Cases listed in Regulatory Guide 1.147, applied to the construction of the particular component.
The implementation schedules in Table 13.4-1 are applicable to the Preservice Inspection Program and Inservice Inspection Program.
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-8.1-1 Receipt Date:
8/11/2025 Question:
In PSAR section 8.1.3, the applicant stated that the offsite power system is not required to meet GDC 2, 4, and 18. PSAR Table 1.9-8 indicates that the offsite power system conforms with NUREG-0800, Revision 5, section 8.2, criteria 1, 2, and 5, that provide the acceptance criteria for meeting the requirements of GDC 2, 4, and 18. PSAR Table 1.9-8 also indicates that this conformance is discussed in PSAR sections 8.1.3 and 8.2.1. It appears that PSAR sections 8.1.3 and 8.2.1 do not discuss how the offsite power system conforms with NUREG-0800, Revision 5, section 8.2, criteria 1, 2, and 5 and GDC 2, 4 and 18. Provide this discussion.
Response
The sections in Chapter 8 are correct; therefore, editorial updates have been made in Table 1.9-8 to revise the conformance fields for NUREG-0800, Revision 5, Section 8.2, Criteria 1, 2, and 5 from conforms to N/A because the offsite power system is not required to meet GDC 2, 4, and 18.
CPA Update on Docket: Sheet 1 of 7 of Table 1.9-8 on page 1.9-69 of the CRN-1 PSAR has been revised for editorial corrections.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 1.9-69 Revision 0 Table 1.9-8 Conformance with NUREG-0800 (Chapter 8 Electric Power)
(Sheet 1 of 7)
SRP Section Rev Title Criterion Topic Conformance PSAR Section 8.1 4
Electric Power-Introduction II Reference to other sections for specific acceptance criteria Conforms 8.1 8.2 5
Offsite Power System 1
Withstanding effects of natural phenomena per GDC 2 requirements Conforms 8.1.3 8.2.1 8.2 5
Offsite Power System 2
Protection against dynamic effects, including missiles generated from equipment failure per GDC 4 requirements Conforms 8.1.3 8.2.1 8.2 5
Offsite Power System 3
Sharing of SSCs at multi-unit stations per GDC 5 requirements N/A - CRN-1 application is for a single unit.
N/A 8.2 5
Offsite Power System 4
Capability, capacity, availability of preferred power source per GDC 17 requirements N/A - Offsite power is not required for accident mitigation.
N/A 8.2 5
Offsite Power System 5
Inspection and testing per GDC 18 requirements Conforms 8.1.3 8.2 5
Offsite Power System 6
Operation per GDCs 34, 38, 41, and 44 requirements N/A - Offsite power system not required for accident mitigation.
N/A 8.2 5
Offsite Power System 7
Alternate AC power source per 10 CFR 50.63 requirements N/A - BWRX-300 design is a passive design with a 72-hour coping capability.
N/A 8.2 5
Offsite Power System 8
Maintenance per 10 CFR 50.65 requirements FSAR FSAR
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-8.1-2 Receipt Date:
8/18/2025 Question:
PSAR section 3.1.4.4 states - GDC 33 applies to small leaks in the RCPB that are smaller than the double-ended rupture of an instrument line (e.g., leakage from flanges or cracks in piping or other components), and which do not exceed the capability of the SC3 high-pressure CRD system used as normal reactor coolant makeup during power operations. The high-pressure CRD system pumps can operate using either onsite AC power (assuming offsite power is not available) from the SC3 SDGs or offsite AC power (assuming onsite power is not available).
The staff notes that although the offsite power is not required for a DBA, it will be used to supply the high-pressure CRD system to provide reactor coolant makeup for small breaks, which do not exceed the capability of the high-pressure CRD system. PSAR Section 8.1.1. does not identify the safety class for the offsite power system, including interfaces with other systems that are classified as SC 3 (including the CRD system). Provide the safety classification of the offsite power system consistent with the descriptions in Section 3.2.2.
Response
SECY-94-084 states that passive designs do not require any active power; therefore, an offsite power system is not required. The offsite power system has no Safety Category function due to the passive design of the BWRX-300. The BWRX-300 offsite power system is classified as SCN.
CPA Update on Docket: Markups to the CRN-1 PSAR Table 3A-1 and Subsection 8.1.1.1 are provided to clarify the safety classification of the offsite power system.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 3A-10 Revision 0 Battery Monitors SC3 TB N/A Non-Seismic Battery Banks SC3 TB N/A Non-Seismic Cable and Raceway System Reactor Building Raceway for DL3 SC1 RB N/A I
Non-Reactor Building Raceway for DL3 SC1 CB, TB, RWB N/A Non-Seismic Reactor Building Raceway for DL4a SC2 RB N/A Non-Seismic Non-Reactor Building Raceway for DL4a SC2 CB, TB, RWB N/A Non-Seismic Reactor Building Raceway for DL4b SC3 RB N/A Non-Seismic Non-Reactor Building Raceway for DL4b SC3 CB, TB, RWB N/A Non-Seismic Reactor Building Raceway for DL2 SC3 RB N/A Non-Seismic Non-Reactor Building Raceway for DL2 SC3 CB, TB, RWB N/A Non-Seismic Other Reactor and Non-Reactor Building Raceway SCN RB, CB, TB, RWB N/A Non-Seismic Grounding and Lightning Protection System All components SCN ALL N/A Non-Seismic AUXILIARY SYSTEMS FUEL STORAGE AND HANDLING New and Spent Fuel Storage Fuel Rack SC1 RB N/A I
Table 3A-1 Preliminary BWRX-300 Component Classification List (Sheet 9 of 22)
Principal Component Safety Class (Note 1)
Location (Note 2)
Quality Group (Notes 3,5)
Seismic Category (Notes 4,5,7,9,12)
Notes Offsite Power System SCN N/A N/A Non-Seismic
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 8.1-3 Revision 0 Figure 8.1-1 depicts the main BWRX-300 electrical one-line diagram for the Generator and Exciter, Standby Power System, and Preferred Power System.
For further explanation on the assignment of safety classes, refer to Subsection 3.2.2.
8.1.1 Offsite Power System The BWRX-300 has two connections to the switchyard. The Preferred Power System provides the interconnecting EDS elements between the plant main generator, onsite power system, and the offsite power source. The preferred power sources for the plant are the two switchyard connections. Note that for the BWRX-300 design, a LOPP or a Loss-of-Offsite Power (LOOP) are the same and can be used interchangeably to represent a loss of the switchyard as a power source to the plant.
The BWRX-300 switchyard design allows significant flexibility. Because of the passive safety design of the BWRX-300 plant, only a single Preferred Power System connection between the plant and switchyard is required. However, two sources of offsite power are provided to improve reliability and operational flexibility.
8.1.1.1 Offsite Power System Design Basis Safety Category Function Offsite power has no Safety Category 1 function due to the passive design of the BWRX-300.
Therefore, offsite power supplies are not required based on Principal Design Criteria (PDC) 17, Electric Power Systems. Further discussion is contained in Section 3.1.
8.1.2 Onsite Power Systems The onsite AC power system starts at the Generator Step Up Transformer (GSU) and Reserve Auxiliary Transformer (RAT), and transitions from there to include the main generator, the main generator circuit breaker, the Unit Auxiliary Transformer (UAT), and the rest of the plant AC distribution as depicted on Figure 8.1-1. The typical BWRX-300 switchyard interface is depicted on Figure 8.1-2.
The Preferred Power System source is from the switchyard breakers through the High Voltage (HV) side of the GSU, through the UAT, and to the Medium Voltage buses. A Preferred Power System source from the switchyard breakers to the RAT primary side is provided for maintenance and resiliency purposes. The UAT and RAT secondaries are connected to the plant MV buses.
The MV bus configuration and load allocation are such that buses can be shut down for maintenance during outages and can be shut down one at a time in normal operation while minimizing the effect to plant power generation.
The MV buses are normally aligned to the UAT. The MV buses can be aligned to the RAT while offline or during power generation, either manually or via automatic transfer. The RAT is normally energized and ready to accept the plant load. The MV buses have protective relays and are
, and the offsite power system is classified as SCN.
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-12-2-1 Receipt Date:
7/28/2025 Question:
PSAR Section 12.2.1, indicates that the reactor vessel core source is described in the manner needed as input to the shield design calculation and that the doses from the reactor core neutron and gamma sources, including the effects of (neutron and gamma) interactions, are considered. However, while the Clinch River PSAR provides numerous source terms and information for numerous radiation sources in the facility, including the core source term 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown in Table 15.5-35 for assessing fuel handling accidents and some neutron fluence information in PSAR Table 5.3-4, the application does not provide the maximum radionuclide inventory for the reactor core during operation (i.e. maximum core source term at shutdown time 0). The maximum core inventory is used and/or related to various source terms and analysis in the PSAR, including the accident analysis. The core source term is also the origin of numerous other source terms already provided in PSAR Chapters 11, 12, and 15. In addition, PSAR Section 12.3.2.2 provides some discussion of RPV, containment, and reactor building shielding, but it is unclear if the concrete and steel of these structures are sufficient to shield neutron radiation or if any specific neutron radiation shielding is needed or has been considered (such as polyethylene, borated material).
Information Needed:
Update the PSAR to provide the maximum radionuclide inventory of the reactor core, considering long term operation of the facility. Also update the PSAR to discuss the methods, models, and assumptions for developing the source term. The core source term should include the principal radionuclides in the reactor core, such as the radionuclides that are important for accident analysis, effluent releases, and radiation shielding design consideration.
Response
The reactor core source is represented by the neutron Watt Fission spectrum in shielding analysis to support the BWRX-300 (X300) conceptual design. It is assumed that the delayed fission gamma source contribution outside of the primary shielding is not statistically significant in comparison to the prompt neutron and neutron induced gamma sources.
Core radioisotope inventories are not provided in the PSAR for shielding, effluent, or accident analysis description purposes as they are not required to support analyses presented as follows:
- The shielding of primary components (i.e., major plant structural components such as the Reactor Pressure Vessel Pedestal) are initially established using the neutron Watt Fission spectrum.
Clarification changes to PSAR Section 12.2.1 are proposed, as shown below.
- Routine radiological effluent releases are calculated using NUREG-0016 and ANSI/ANS-18.1
methodology, which does not require plant-specific reactor core isotopic inventories as an input. See PSAR Sections 11.2.4 and 11.3.4 for calculation descriptions of liquid and gaseous effluent releases.
- Accident analyses presented in PSAR Chapter 15 do not include events that result in core damage.
With the exception of the Fuel Handling Accident where isotopic compositions are provided in PSAR Table 15.5-35, events involving radiological releases are derived from design basis reactor coolant source terms presented in PSAR Section 11.1.
CPA Update on Docket:
PSAR Section 12.2.1 text under the Reactor Vessel Core Sources subheading will be revised as follows:
The reactor vessel core source is described in the manner needed as input to the shield design calculations. To determine Ddose rates external to shielding and within occupied areas are determined based on reactor core neutron and gamma sources including, a reactor core source model equivalent to the neutron Watt Fission spectrum scaled to the BWRX-300 power density at full power is applied. This spectrum is representative of the energy dependent probabilities of fission-borne neutrons and neutron leakage expected in a core of this size. Tthe effect of (n, gamma) interactions on the reactor vessel core source is considered.
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests This package contains: A-13.7-1 and A13.7-2 Responses Question Number:
A-13.7-1 Receipt Date:
8/4/2025 Question:
The PSAR should refer to the requirements in 10 CFR 26.4(a) as summarized in NEI 06-06, Table 3-1, "FFD Program Applicability and MilestonesFor Construction and Transition to Operations," or NUREG-0800, Chapter 13.7.2, Table 1 - "FFD Program Applicability and Milestones," to ascertain when an FFD program for operation must be implemented. For example, FFD Program for Security personnel is required prior to the establishment of the protected area vice "Protected area containing new fuel" as stated in the Table 13.4-1, "Initial Operations" of the PSAR.
Additionally, the FFD Program for Operation must be established to the earlier of the following, "Licensees receipt of fuel assemblies onsite or establishment of a protected area or the 10 CFR 52.103(g) finding" vice "Fuel receipt in a protected area" as stated in the PSAR. The PSAR should use the correct milestones descriptions in NEI 06-06 as endorsed by the NRC in RG 5.84.
Response
Table 13.4-1, Operational Programs and Implementation Milestones will be modified to update implementation milestones to be consistent with NEI 06-06, Table 3-1, FFD Program Applicability and Milestones For Construction and Transition to Operations and NUREG-0800, Chapter 13.7.2, Table 1, FFD Program Applicability and Milestones. Refer to Attachment A: NRC Audit Response A-13.7-1 Markup, which updates the Implementation Milestones column of the table.
CPA Update on Docket: Attachment A will be used as the basis for a future docketed PSAR update to Table 13.4-1.
Question Number:
A-13.7-2 Receipt Date:
8/11/2025 Question:
NRC-endorsed NEI 06-06, Revision 6, only provides one acceptable method for FFD program guidance for individuals categorized in 10 CFR 26.4(f) and not for individuals categorized in 10 CFR 26.4(e), (g), etc. While table 13.4-1 lists the regulation in which individuals subjected to an FFD program are described, the NRC requests TVA clarify which portions of Chapter 13.7 describes or refers to individuals who are subject to an FFD program that are covered by 10 CFR 26.4(e) and (g).
Response
PSAR Section 13.7 will be modified to make explicit references to 10 CFR 26.4(e) and (g) individuals. Refer to Attachment B: NRC Audit Response A-13.7-2 Markup.
CPA Update on Docket: Attachment B will be used as the basis for a future docketed PSAR update to PSAR Section 13.7.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 13.4-5 Revision 0 Security Program for aggregated category 1 or category 2 quantity of radioactive material (applicable to protection of aggregated quantity of radioactive material prior to fuel on site) 10 CFR 37.41(a)(1)
Prior to receipt of fuel onsite (protected area) 13.6 Fitness for Duty (FFD) Program for Construction (workers and first-line supervisors) 10 CFR 26.3(c) 10 CFR 26.4(f)
Prior to initiating 10 CFR Part 26 construction activities 13.7 Fitness for Duty (FFD) Program for Construction (management and oversight personnel) 10 CFR 26.3(c) 10 CFR 26.4(e)
Prior to initiating 10 CFR Part 26 construction activities 13.7 FFD Program for Security Personnel 10 CFR 26.4(e)(1)
Prior to initiating 10 CFR Part 26 construction activities 13.7 10 CFR 26.4(a)(5)
Protected Area containing new fuel FFD Program for FFD Program personnel 10 CFR 26.3(c);
Prior to initiating 10 CFR Part 26 construction activities 13.7 FFD Program for persons required to physically report to the Technical Support Center (TSC) or Emergency Operations Facility (EOF) 10 CFR 26.4(c)
Prior to the conduct of the first full-participation emergency preparedness exercise under 10 CFR Part 50, Appendix E, Section F.2.a 13.7 FFD Program for Operation 10 CFR 26.4(a) and (b)
Fuel Receipt in a Protected Area 13.7 Table 13.4-1 Operational Programs and Implementation Milestones (Sheet 4 of 5)
Program Title Source Requirement Implementation Milestone FSAR Section Delete text and Insert:
" Prior to the earlier of:
A. Receipt of fuel assemblies on site or B. Establishment of a protected area" Delete text and Insert:
" Prior to the earlier of:
A. Receipt of fuel assemblies on site or B. Establishment of a protected area" Attachment A - Markup for A-13.7-1 and A-13.7-2 Responses
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 13.7-1 Revision 0 13.7 Fitness for Duty The Fitness for Duty (FFD) Program is implemented and maintained in multiple and progressive phases dependent on the activities, duties, or access afforded to certain individuals at the construction site. In general, two different phases of the FFD program will be implemented: a construction phase FFD program and an operating phase FFD program. TVA will maintain responsibility for the FFD program and has oversight of any duties that are delegated to selected contractors. The construction and operating phase programs are implemented as identified in Table 13.4-1 requirements.
The construction phase FFD program is consistent with NEI 06-06 (Reference 13.7-1), as endorsed in Regulatory Guide 5.84, and meets the applicable requirements of 10 CFR Part 26.
The construction FFD program will meet or exceed the requirements in NEI 06-06, and any exceptions will be documented in Chapter 1. NEI 06-06 applies to persons constructing or directing the construction of SC1 and security-related SSCs performed onsite where the new reactor will be installed and operated.
The FFD program during the construction phase implements different requirements for different individuals by specific milestones, as required in regulation. This includes:
Construction personnel (workers and first line supervisors as described in NEI 06-06) will be subject to 10 CFR Part 26, Subpart K.
Construction management and oversight personnel, as further described in NEI 06-06, and security personnel will be subject to 10 CFR Part 26, Subparts A through H, N, and O.
FFD program personnel will be subject to 10 CFR Part 26, Subparts A, B, D-H, N & O.
Upon the receipt of special nuclear material onsite in the form of fuel assemblies, security personnel will also be subject to 10 CFR Part 26, Subpart I.
Upon the establishment of a protected area in accordance with 10 CFR 73.55 and receipt of fuel assemblies, the site will transition to the operating phase of the FFD program in accordance with the applicable requirements of 10 CFR Part 26.
The operating phase of the FFD program will be consistent with TVAs existing FFD program for operating nuclear power plants, which conforms with the applicable requirements of 10 CFR Part 26.
13.7.1 References 13.7-1 NEI 06-06, Fitness for Duty Program Guidance for New Nuclear Power Plant Construction Sites, Nuclear Energy Institute, Revision 6, May 2012.
Delete text and insert:
"Personnel performing duties described in 10 CFR 26.4(e)"
Insert: "performing duties described in 10 CFR 26.4(g) "
Attachment A - Markup for 13.7-1 and 13.7-2 Responses
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-17-1-1 Receipt Date:
8/4/2025 Question:
Section 17.1 of the PSAR states that the QA Program for site design, construction, and operation activities controlled by TVA is described in Section 17.5. TVA considers GEH a vendor under the oversight of the TVA Quality Assurance Program Description in Section 17.5. The QA Program used for BWRX-300 GEH design activities is based on the standard QA Program documented in topical report NEDO-11209-A, "GE hitachi Nuclear Energy Quality Assurance Program Description," (Reference 17.1-1) and the additional information in the chapter, which describes and clarifies GEH interfaces and responsibilities with BWRX-300 project participants." Given that GEH is identified as a vendor under the oversight of the TVA QAPD, please explain why Section 17.1 of the PSAR incorporates by reference the QA Program used for BWRX-300 GEH design activities documented in topical report NEDO-11209-A, and the additional information in the chapter, which describes and clarifies GEH interfaces and responsibilities with BWRX-300 project participants.
Response
Sections 17.1 and 17.4 have been revised to remove reference to GEH topical report NEDO-11209-A. See attached mark-up of Sections 17.1 and 17.4.
CPA Update on Docket: See attached mark-up of PSAR Sections 17.1 and 17.4.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 17.1-1 Revision 0 CHAPTER 17 QUALITY ASSURANCE The Quality Assurance (QA) Program described in Section 17.1 is applicable to the BWRX-300 standard design activities used by GEH. The QA Program for site design, construction, and operation activities controlled by TVA is described in Section 17.5.
17.1 Quality Assurance During the Design and Construction Phases The QA Program for site design, construction, and operation activities controlled by TVA is described in Section 17.5. TVA considers GEH a vendor under the oversight of the TVA Quality Assurance Program Description in Section 17.5.
The QA Program used for BWRX-300 GEH design activities is based on the standard QA Program, documented in topical report NEDO-11209-A, GE Hitachi Nuclear Energy Quality Assurance Program Description, (Reference 17.1-1) and the additional information in this chapter, which describes and clarifies GEH interfaces and responsibilities with BWRX-300 project participants.
The standard QA Program is already in use on nuclear power plant work and is U.S. Nuclear Regulatory Commission (USNRC) accepted. The QA Program complies with 10 CFR 50, Appendix B and American Society of Mechanical Engineers (ASME) NQA-1-2015, Quality Assurance Requirements for Nuclear Facility Applications, (Reference 17.1-2). The QA Program also complies with RG 1.28, Quality Assurance Program Criteria (Design and Construction) and is organized to show its relationship to ASME NQA-1-2015 and the exceptions and clarifications noted in RG 1.28, Revision 5.
Conformance with regulatory guides is described in Section 1.9.
17.1.1 Organization NEDO-11209-A, Section 1, describes requirements for the organization and the organizational structure used during design of the BWRX-300.
17.1.2 Quality Assurance Program NEDO-11209-A, Section 2, describes the requirements for the QA Program used during design of the BWRX-300.
The classification of Structures, Systems, and Components (SSCs) for the BWRX-300 is described in Section 3.2. The GEH QA Program is applied to Safety Class 1 (SC1) SSCs relative to performance of their Safety Category 1 functions. Quality controls for SSCs within the scope of Design Reliability Assurance Program (D-RAP) are described in Subsection 17.4.7.
NEDO-11209-A, Section 2, addresses personnel qualification requirements for the QA Program consistent with ASME NQA-1 2015.
TVA owns responsibility for design activities related to the CRN site and has delegated the responsibility for BWRX-300 design activities to GEH. GEH performs BWRX-300 design activities in accordance with the GEH Quality Assurance Program Description (QAPD).
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 17.1-2 Revision 0 17.1.3 Design Control NEDO-11209-A, Section 3, establishes requirements for design control used during design of BWRX-300, including software design.
17.1.4 Procurement Document Control NEDO-11209-A, Section 4, establishes requirements for procurement document control used during design of the BWRX-300.
17.1.5 Instructions, Procedures, and Drawings NEDO-11209-A, Section 5, establishes requirements for instructions, procedures, and drawings used during design of the BWRX-300.
17.1.6 Document Control NEDO-11209-A, Section 6, establishes requirements for document control used during design of the BWRX-300.
17.1.7 Control of Purchased Material, Equipment, and Services NEDO-11209-A, Section 7, establishes requirements for control of purchased material, equipment, and services used during design of the BWRX-300.
17.1.8 Identification and Control of Materials, Parts, and Components NEDO-11209-A, Section 8, establishes requirements for identification and control of materials, parts, and components during design of the BWRX-300.
17.1.9 Control of Special Processes NEDO-11209-A, Section 9, establishes requirements for control of special processes used during design of the BWRX-300.
17.1.10 Inspection NEDO-11209-A, Section 10, establishes requirements for inspection used during design of the BWRX-300.
17.1.11 Test Control NEDO-11209-A, Section 11, establishes requirements for test control used during design of the BWRX-300.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 17.1-3 Revision 0 17.1.12 Control of Measuring and Test Equipment NEDO-11209-A, Section 12, establishes requirements for control of measuring and test equipment during design of the BWRX-300.
17.1.13 Handling, Storage, and Shipping NEDO-11209-A, Section 13, establishes requirements for handling, storage, and shipping used during design of the BWRX-300.
17.1.14 Inspection, Test, and Operating Status NEDO-11209-A, Section 14, establishes the control of inspection, test, and operating status used during design of the BWRX-300.
17.1.15 Nonconforming Materials, Parts, or Components NEDO-11209-A, Section 15, establishes requirements for the control of nonconforming materials, parts, or components used during design of BWRX-300.
17.1.16 Corrective Action NEDO-11209-A, Section 16, establishes requirements for the corrective action program used during design of the BWRX-300.
17.1.17 Quality Assurance Records NEDO-11209-A, Section 17, establishes requirements for control of QA records used during design of the BWRX-300.
17.1.18 Audits NEDO-11209-A, Section 18, establishes requirements for a system of QA audits used during design of the BWRX-300, which is a support activity.
17.1.19 References 17.1-1 NEDO 11209-A, GE Hitachi Nuclear Energy Quality Assurance Program Description, GE-Hitachi Nuclear Energy Americas, LLC., Revision 17, December 2022.
17.1-2 ASME NQA-1-2015, Quality Assurance Requirements for Nuclear Facility Applications, American Society of Mechanical Engineers, 2015.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 17.4-5 Revision 0 lower-level ones, mapping is performed to understand the risk significance of components, systems, and functions. The risk significance criteria are applied at whatever level of entity being examined (i.e., basic event, component, system, or function).
17.4.6 Design Considerations The predicted reliability of SSCs in the scope of D-RAP are evaluated at the detailed design stage by appropriate design reviews, PSA, and reliability analyses. If a proposed design change affects the PSA model, then the PSA is revised in accordance with the PSA update process, and the new results will be implemented into the D-RAP process.
17.4.7 Quality Controls Applicable to Design Reliability Assurance Program The quality controls applicable to the D-RAP process during the design phase are described in Section 17.1. Specifically, the design control process is subject to the requirements in Subsection 17.1.3. Procedures used by D-RAP process are subject to requirements in Subsection 17.1.5. The corrective action program described in Subsection 17.1.16 is applied to identify and resolve issues related to D-RAP. Records generated by D-RAP activities are handled in accordance with the requirements described in Subsection 17.1.17, and the D-RAP process is subject to audit requirements described in Subsection 17.1.18.
Quality controls for non-SC1 SSCs that are risk-significant are consistent with those for Non-safety related SSCs that are significant contributors to plant safety described in NUREG-0800, Standard Review Plan, Section 17.5, during the design phase of D-RAP. Quality controls for non-SC1 SSCs that are risk-significant are described in Section 17.5 during the construction phase.
17.4.8 References 17.4-1 NEDC-33934P, BWRX-300 Safety Strategy, GE-Hitachi Nuclear Energy Americas, LLC., Revision 1, June 2024.
This includes the design control process, procedures used by D-RAP process, the corrective action program, records and audits
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-17-5-1 Receipt Date:
8/11/2025 Question:
Section 17.5 of the PSAR identifies that NNP-TR-001-NP-A, Revision 2, dated June 2023, is incorporated by reference. However, NNP-TR-001-NP-A, Revision 2, dated June 2023 does not include the NRC staffs safety evaluation, which define the staffs conditions and limitations for approving NNP-TR-001-NP-A, Revision 2. The transmittal letter from TVA, Approved -
A Version of the Quality Assurance Program Description for Tennessee Valley Authority (TVA) New Nuclear, Revision 2, dated February 5, 2024 (ADAMS Accession No. ML24037A144) included NNP-TR-001-NP-A, Revision 2, dated June 2023, the NRC staffs final safety evaluation for the TVA topical report NNP-TR-001-NP, Revision 2, and responses to the NRCs request for additional information (eRAI-386) as enclosures to the letter. As such, it is not clear whether the safety evaluation for this topical report is included by reference into the PSAR. Please submit a revised version of the NNP-TR-001-NP-A which includes the staff's safety evaluation.
Response
TVA will resubmit NNP-TR-001-NP-A, which includes the safety evaluation.
See attached Compiled letter and Enclosure.
CPA Update on Docket: See attached Compiled letter and Enclosure.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 1.6-1 Revision 0 1.6 Material Referenced Documents that are incorporated by reference into the PSAR are identified using the bases provided in Nuclear Energy Institute (NEI) 98-03, Guidelines for Updating Final Safety Analysis Reports, (Reference 1.6-1), which is endorsed by RG 1.181. Subsection 1.6.1 lists Topical Reports which provide BWRX-300 design information that supports the CPA.
Incorporation of the SSAR is addressed in Section 1.8.
1.6.1 Topical Reports The following Topical Reports are Incorporated By
Reference:
NEDC-33922P-A, BWRX-300 Containment Evaluation Method, (Reference 1.6-2),
discusses the analysis method used for the thermal hydraulics performance to demonstrate that the containment design satisfies the acceptance criteria listed in NEDC-33911P-A, BWRX-300 Containment Performance, (Reference 1.6-4).
NNP-TR-001-NP-A, Quality Assurance Program Description for TVA New Nuclear, (Reference 1.6-3), describes the quality program for the site-selection, design, construction, and operation of CRN-1.
Other topical or technical reports referenced in the PSAR provide design and analysis information that supports the CPA.
1.6.2 References 1.6-1 NEI 98-03, Guidelines for Updating Final Safety Analysis Reports, Nuclear Energy Institute, Revision 1, June 1999.
1.6-2 NEDC-33922P-A, BWRX-300 Containment Evaluation Method, GE Hitachi Nuclear Energy Americas, LLC, Revision 3, June 2022.
1.6-3 NNP-TR-001-NP-A, Quality Assurance Program Description for TVA New Nuclear, Tennessee Valley Authority, Revision 2, June 2023.
1.6-4 NEDC-33911P-A, BWRX-300 Containment Performance, GE-Hitachi Nuclear Energy Americas, LLC, Revision 3, January 2022.
September 15, 2025 (ML25258A031)
Supplement 1, Audit Question A-17.5-1
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 17.5-3 Revision 0 Supplement 1 PSAI 4.7 An application for an OL that references the TVA New Nuclear QAPD must specify that requisite design verification activities will be completed prior to relying on the design outputs in fulfilling the required information needed for OL issuance.
Response: PSAI 4.7 is not applicable to a CP application.
PSAI 4.8 An application for COL, CP, or OL that references the TVA New Nuclear QAPD must describe the qualification requirements for outside fire protection consultants who will be relied upon to conduct fire protection equipment and program inspections.
Response: No action required. Qualification requirements for fire protection staff are described in Section 1.6.1.a of RG 1.189, Fire Protection for Nuclear Power Plants, Revision 4. TVA New Nuclear will conform with these requirements for the selection of outside fire protection consultants. Section 1.9 addresses conformance to RG 1.189, Revision 4.
PSAI 4.9 An application for SDA, DC, ESP, COL, CP, or OL that references the TVA New Nuclear QAPD must provide information on conformances and exceptions to RG 1.8, Revision 4, RG 1.26, Revision 6, RG 1.28, Revision 5, RG 1.54, Revision 3, RG 1.164 Revision 0, RG 1.231, Revision 0, RG 1.234, Revision 0.
Response: Information on conformances and exceptions to Regulatory Guides is summarized in Section 1.9.
17.5.1 References 17.5-1 NNP-TR-001-NP-A, "Quality Assurance Program Description for TVA New Nuclear,"
Tennessee Valley Authority, Revision 2, June 2023.
September 15, 2025 (ML25258A031)
Supplement 1, Audit Question A-17.5-1
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-17.5-2 Receipt Date:
8/11/2025 Question:
"The staff would like to have a clarification call with TVA to get a better understanding of the organizational structure as outlined in the PSAR, TVA's New Nuclear QAPD, and TVA's Fleet QAPD, as there are discrepancies in how certain positions, roles, and responsibilities are described in the two QAPDs and the PSAR. For example, Section 17.5 of the PSAR references NNP-TR-001-NP, Quality Assurance Program Description for TVA New Nuclear, but indicates that information in the QAPD needs to be revised to reflect the current responsibilities for the Senior Vice Present, Engineering and Operations Support. Specifically, PSAR Section 17.5 states that the last sentence in the New Nuclear QAPD, Part II, Section 1.2.1 should be read as follows: The Senior Vice President, Engineering and Operations Support, is also responsible for governance, oversight, and support of existing nuclear procedures, processes, and personnel, which will be leveraged for the TVA New Nuclear Program. The staff currently understands that the position identified as the Senior Vice President, Engineering and Operations Support, is the same position as the Senior Vice President, Engineering and Operations Support described in the TVA Fleet QAPD (e.g., Revision 44). However, the description of this position as provided in the TVA fleet QAPD does not fully align with the description provided in the PSAR or the New Nuclear QAPD (e.g., the Fleet QAPD notes the responsibilities for this position include governance, oversight, and support of TVA New Nuclear Program and Projects activities).
(1) How does TVA intend to align the PSAR and the New Nuclear QAPD to ensure they reflect recent changes in the organizational structure?
(2) What are TVAs longer term plans to align and/or merge the New Nuclear and Fleet QAPDs to ensure consistency and alignment in organizational responsibilities, including organizational changes, to the extent applicable?"
Response
To address the concern that the description provided in the PSAR for the Senior Vice President, Engineering and Operations Support, TVA will add and Projects after Program in Section 17.5.
The responses to the two additional questions follow:
- 1) How does TVA intend to align the PSAR and the New Nuclear QAPD to ensure they reflect recent changes in the organizational structure?
The second paragraph of PSAR Chapter 17.5 states that Revisions to NNP-TR-001-NP will be performed periodically in accordance with a defined process and transmitted to the NRC.
(2) What are TVAs longer term plans to align and/or merge the New Nuclear and Fleet QAPDs to ensure consistency and alignment in organizational responsibilities?
TVA intends to revise the New Nuclear QAPD to Revision 3 in accordance with TVA procedures to
reflect changes in the organizational structure. Revision 3 will be transmitted to the NRC prior to implementation.
Background - TVA Executive Team decided to have two plans, one for the TVA Fleet and one for New Nuclear Program/Project (notably for CRN-1 construction), to avoid distracting the operating fleet with CRN project demands during construction. New Nuclear will continue to use fleet procedures and processes where appropriate. New Nuclear specific procedures and processes developed and implemented as required.
TVA does not plan to merge the New Nuclear and Fleet QAPDs. However, the QAPDs share a similar structure and commitment to NQA-1-2015 to enable merging the QAPDs in the future if it is beneficial to do so.
TVA will utilize internal reviews and revisions of the QAPDs to maintain alignment of the QAPD organization section content that is common to the QAPDs.
CPA Update on Docket: See attached markup of PSAR Section 17.5.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 17.5-1 Revision 0 17.5 Quality Assurance Program Description - Design Certification, Early Site Permit, and New License Applicants The Quality Assurance Program Description (QAPD) for TVA New Nuclear is provided in the topical report NNP-TR-001-NP-A, Quality Assurance Program Description for TVA New Nuclear, (Reference 17.5-1). The classification of Structures, Systems, and Components (SSCs) for the BWRX-300 is described in Section 3.2. The TVA New Nuclear QAPD is applied to Safety Class 1 (SC1) SSCs relative to performance of their Safety Category 1 functions.
Revisions to NNP-TR-001-NP will be performed periodically in accordance with a defined process and transmitted to the NRC. The topical report is Incorporated by Reference with the following departures and/or supplements.
1.
The last sentence in Part II, Section 1.2.1 should be read as follows:
The Senior Vice President, Engineering and Operations Support, is also responsible for governance, oversight, and support of existing nuclear procedures, processes, and personnel, which will be leveraged for the TVA New Nuclear Program.
2.
The final safety evaluation of NNP-TR-001-NP contains the following limitations and conditions.
FINAL SAFETY EVALUTION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGARDING THE TOPICAL REPORT OF THE QUALITY ASSURANCE PROGRAM DESCRIPTION FOR THE TENNESSEE VALLEY AUTHORITY NEW NUCLEAR PROGRAM EPID NO. L-2022-TOP-0055 LIMITATIONS AND CONDITIONS An application for SDA, DC, ESP, COL, CP, OL, or LWA (as applicable) may reference the approved TVA New Nuclear QAPD provided that the application satisfies the following conditions and limitations, as applicable. The conditions and limitations are intended to address plant-specific aspects of TVA New Nuclear QA program that have not been described in the TVA New Nuclear QAPD. The TVA New Nuclear QAPD and the NRC staff's safety evaluation provides context and basis for the required additional information.
The following plant-specific action items (PSAIs) must be satisfied by applications referencing the TVA New Nuclear QAPD:
PSAI 4.1 An application for SDA, DC, ESP, COL, CP, OL, or LWA (as applicable) that intends to credit the approved version of the TVA New Nuclear QAPD for activities where the regulations other than 10 CFR Part 50 and Part 52 established QA requirements must include an analysis that demonstrates the QA requirements set forth in these regulations are met by the TVA New Nuclear QAPD.
Response: The CP application is submitted under 10 CFR Part 50.
and Projects
TVA Response to CP Application Preliminary Safety Analysis Report Audit Information Requests Question Number:
A-17-5-3 Receipt Date:
8/4/2025 Question:
Table 1.9-20, Conformance with Regulatory Guides (Sheet 5 of 38) identifies that the PSAR conforms to RG 1.33, Revision 3 and RG 1.28, Revision 5. However, RG 1.33 is applicable only to quality assurance program requirements during the operation phase as required by 10 CFR 50.34(b)(6)(ii). It is not clear why this table identifies PSAR sections for demonstrating conformance to RG 1.33, Revision 3. Please explain why RG 1.33 is referenced in the CPA.
Response
The TVA New Nuclear Quality Assurance Program Description Topical Report (NNQAP) covers the design, construction, and operations phases of CRN. The NNQAP documents a commitment to RG 1.33 for the operations phase.
During the design and construction phase, TVA will be developing procedures and processes for use in the operation phase of Clinch River. The content of those procedures and processes will need to comply with RG 1.33 prior to entering the operational phase, and therefore the commitment to this RG is included in the CPA.
RG 1.70 Section 13.1.2, Operating Organization, states that the SAR should describe the structure, functions, and responsibilities of the onsite organization established to operate and maintain the plant. Regulatory Guide 1.33 provides guidance for the establishment of these organizations.
In addition, RG 1.70, Section 13.5.1.1, Conformance with Regulatory Guide 1.33 states:
Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)," contains guidance on facility administrative policies and procedures. The SAR should specifically indicate whether the applicable portions of Regulatory Guide 1.33 concerning plant procedures will be followed. If such guidance will not be followed, the SAR should describe specific alternative methods that will be used and the manner of implementing them.
Full compliance with RG 1.33 and ANS-3.2-2012 will be documented in the FSAR for the Operations License Application. TVA does not expect finality regarding conformance to RG 1.33 in accordance with 10 CFR 50.34(b)(6)(ii). Table 1.9-20, RG 1.33, will be revised to add the following note: Conformance of Section 17.5 will apply to the FSAR.
CPA Update on Docket: See attached markup of PSAR Table 1.9-20.
Clinch River Nuclear Site Construction Permit Application, Preliminary Safety Analysis Report 1.9-152 Revision 0 Supplement 1 1.33 Quality Assurance Program Requirements (Operation) 3 Conforms 5.4.7 13.1 13.5 17.5 1.34 Control of Electroslag Weld Properties 1
N/A The design does not use electroslag welding except for RPV cladding.
5.2.3 5.3.1 1.35.1 Determining Prestressing Forces for Inspection of Prestressed Concrete Containments 0
N/A BWRX-300 design does not include prestressed concrete.
N/A 1.36 Nonmetallic Thermal Insulation for Austenitic Stainless Steel 1
Conforms 5.2.3 6.1.4 1.40 Qualification of Continuous Duty Safety-Related Motors for Nuclear Power Plants 1
Conforms 3.11.2 1.41 Preoperational Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments 0
FSAR FSAR 1.43 Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components 1
N/A The requirements of RG 1.43 are not applicable to the BWRX-300 vessel because the RPV is constructed from SA-508/SA-508M, Grade 3, Class 1 or SA-533/SA-533M, Type B, Class 1 (low-alloy steel forgings or plates).
5.2.3 5.3.1 Table 1.9-20 Conformance with Regulatory Guides (Sheet 5 of 38)
Reg Guide Title Rev Conformance Alternatives PSAR Section Conformance of Section 17.5 will apply to the FSAR.