ML23117A327
| ML23117A327 | |
| Person / Time | |
|---|---|
| Issue date: | 05/17/2023 |
| From: | Joseph Donoghue NRC/NRR/DSS |
| To: | |
| Shared Package | |
| ML23117A323 | List: |
| References | |
| RG 1.7 Rev 3 | |
| Download: ML23117A327 (5) | |
Text
Regulatory Guide Periodic Review Regulatory Guide Number:
1.7, Revision 3
Title:
Control of Combustible Gas Concentration in Containment Office/Division/Branch:
NRR/DSS/SCPB Technical Lead:
Hanry Wagage Staff Action Decision:
Revise Date:
05/17/2023
- 1.
What are the known technical or regulatory issues with the current version of the Regulatory Guide (RG)?
The staff has identified the following issues, including those identified previously during the RG 1.7 Periodic Review July 2013 (Ref. 6.1).
1.1 Improvements Needed Based on Fukushima Lessons Learned One of the primary lessons learned from the accident at the Fukushima Dai-ichi nuclear power plant was the significance of the challenge presented by a loss of multiple safety-related systems following the occurrence of a beyond-design-basis external event. In the case of the Fukushima Dai-ichi accident, the loss of all alternating current power led to loss of core cooling, and ultimately to core damage and a loss of containment integrity. The design basis for U.S. nuclear plants includes bounding analyses with margin for external events expected at each site. Extreme external events (e.g., seismic events, external flooding, etc.) beyond those accounted for in the design basis, while unlikely, could present challenges to nuclear power plants.
In response to lessons learned from the Fukushima Dai-ichi accident, the U.S. Nuclear Regulatory Commission (NRC) promulgated Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Section 50.155, Mitigation of beyond design-basis events, to improve the capability of nuclear power plants to address beyond-design-basis external events (Ref. 6.2).
The staff issued RG 1.226 (Ref. 6.3), providing guidance for meeting requirements of 10 CFR 50.155. This guidance references capabilities to support combustible gas control that are applicable to (a) boiling water reactors (BWRs) with Mark III type containments and (b) pressurized water reactors with ice condenser containments. This change was part of the resolution of Generic Safety Issue 189 (Ref. 6.4), and therefore, Footnote 1 of Regulatory Position C.1 should be deleted. The staff recommends referencing 10 CFR 50.155 and associated RG 1.226 in the revision to RG 1.7.
1.2 Miscellaneous Issues (a) Regulatory positions in RG 1.7, Revision 3, are not aligned with the subsections of 10 CFR 50.44(c) (i.e., 1. Mixed atmosphere, 2. Combustible gas control,
- 3. Equipment survivability, 4. Monitoring, and 5. Analyses):
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The staff recommends aligning the regulatory positions as proposed in Table 1 (below) to enhance clarity.
RG 1.7, Revision 3, Regulatory Position C.4, Hydrogen Gas Production, is not supported by the regulation and suggest deleting it.
The following text from RG 1.7, Revision 3, Regulatory Position C.1 needs to be moved to RG 1.7, Revision 4, Regulatory Position C.3 Equipment Survivability:
Table 1. RG 1.7 Regulatory Position Mapping for being Consistent with 10 CFR 50.441 Rev. 4 (proposed update)
Rev. 3 (current version)
C.1 Mixed atmosphere C.3 Atmosphere Mixing Systems C.2 Combustible gas control C.1 Combustible Gas Control Systems, excluding Equipment survivability expectations under severe accident conditions should consider the circumstances of applicable initiating events... This guidance was used to review the design of evolutionary and passive plant designs, as documented in NUREG-1462 [Ref. 6.5], NUREG-1503
[Ref. 6.6], and NUREG-1512 [Ref. 6.7].
C.3 Equipment Survivability The part excluded above after revising as proposed in 2(d) below C.4 Monitoring C.2 Hydrogen and Oxygen Monitors, excluding C.2.1(1)/C.2.2(1) which provided guidance on equipment survivability C.5 Analyses C.5 Containment Structural Integrity 1 RG 1.7, Revision 3, Regulatory Position C.4, Hydrogen Gas Production, is not in the regulations and the staff suggests deleting it.
(b) RG 1.57 and RG 1.136 for steel and concrete containments provide guidance on load combinations, analysis methods, and acceptance criteria necessary for structural analysis (Refs. 6.8 and 6.9). Recommend adding these references to Regulatory Position C.5 (of proposed RG 1.7 Revision 4 (see Table 1)).
(c) RG 1.7 Revision 3 Regulatory Positions 2.1 and 2.2 providing guidance for hydrogen and oxygen monitors, respectively:
Identical guidance is provided for both hydrogen and oxygen monitors on the following: Equipment Survivability, Power Source, Display and Recording, Range, Human Factors, and Direct Measurement.
Guidance is provided for hydrogen monitors differ from that for oxygen monitors for the following:
o Guidance is provided for Channel Availability and Interfaces of oxygen monitors and not for hydrogen monitors.
o Additional guidance is provided on Servicing, Testing, and Calibration for oxygen monitors but not for hydrogen monitors: The location of the isolation
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device should be such that it would be accessible for maintenance during accident conditions.
o Guidance provided on Quality Assurance is more detailed for oxygen monitors than that for hydrogen monitors.
Recommend providing guidance common to both hydrogen and oxygen monitors and oxygen monitor specific guidance separately to avoid repetition.
(d) The RG directs to Chapter 19 of the ABWR FSER (Ref. 6.6) and the AP1000 FSER (Ref. 6.10) for acceptable approaches for demonstrating equipment survivability.
Instead, the staff recommends including relevant guidance from these two FSERs as well as FSERs from later design reviews (i.e., ESBWR, APR1400, and NuScale Design Certification Application (Refs. 6.11 through 6.1)).
(e) Page 10, Regulatory Analysis/Backfit Analysis: Recommend replacing this with Section D, Implementation, as given in recently revised RGs, e.g., RG 1.82, Revision 5 (Ref. Error! Reference source not found.).
- 2.
What is the impact on internal and external stakeholders of not updating the RG for the known issues, in terms of anticipated numbers of licensing and inspection activities over the next several years?
The staff is engaged in the review of NuScale Power, LLC application for NuScale US460 Standard Design Approval and the following pre-application activities for small modular reactors of light-water reactor (LWR) designs. These pre-applications will benefit from updating RG 1.7 guidance because it will help them meet the requirements of 10 CFR 50.155 and offer other guidance improvements discussed under item 1.2 above:
(a) US460 at Idaho National Laboratory Site - Utah Associated Municipal Power Systems and Carbon Free Power Project Pre-Application for a Combined License (b) SMR-160 - SMR, LLC, a subsidiary of Holtec International Pre-Application (c) BWRX-300 - GE-Hitachi Nuclear Energy Pre-Application (d) BWXT mPower' - BWXT mPower, Inc. Pre-Application There will be no impact on the existing U.S. fleet of operating LWRs because the RG provides methods for meeting 10 CFR 50.44(b) and the plants have already shown compliance to those requirements.
There will be no impact on Vogtle Electric Generating Plant Units 3 and 4, currently undergoing construction, fuel load, and startup testing activities per the regulations in 10 CFR Part 52. RG 1.7 Revision 3 provides methods for meeting 10 CFR 50.44(c),
which the plants have already shown compliance during the reviews of design certification and combined license applications.
There will be no impact on the NuScale Design Certification Application (DCA), which has already been approved by staff. The DCA, which includes approving NuScales proposed exemption from the requirements of 10 CFR 50.44(c)(2). The staff has determined that 10 CFR 50.44(c), as applicable to the design, was met (Ref. 6.1).
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- 3.
What is an estimate of the level of effort needed to address identified issues in terms of full-time equivalent (FTE) and contractor resources?
0.2 FTE for the revision
- 4.
Based on the answers to the questions above, what is the staff action for this guide (Reviewed with no issues identified, Reviewed with issues identified for future consideration, Revise, or Withdraw)?
Revise
- 5.
Provide a conceptual plan and timeframe to address the issues identified during the review.
Revise RG 1.7 Revision 3 to address issues as identified above in the answer to Question 1. The target date for DSS issuance of RG 1.7 Revision 4, which includes addressing the public comments, is anticipated by the end of NRC fiscal year 2025. This timeline will help to support the staff review of upcoming small modular reactor design certification applications.
- 6.
References 6.1 NRC, RG 1.7 Periodic Review July 2013, Washington, DC, March 23, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17047A181).
6.2 U.S. Code of Federal Regulations (CFR), Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy.
6.3 NRC, RG 1.226, Revision 0, Flexible Mitigation Strategies for Beyond-Design-Basis Events, Washington, DC, June 2019, ML19058A012.
6.4 NRC, Technical Report Supporting Closure of Generic Issue 189: Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen Combustion During a Severe Accident, Washington, DC, February 2013, ML13008A361.
6.5 NRC, NUREG-1462, Volumes 1 and 2, Final Safety Evaluation Report Related to the Certification of the System 80+ Design, Washington DC, August 1994, ML100780157 and ML100430017.
6.6 NRC, NUREG-1503, Final Safety Evaluation Report Related to the Certification of the Advanced Boiling-Water Reactor Design, Docket No.52-001, Washington, DC, July 1994.
6.7 NRC, NUREG-1512, Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003, U.S. Nuclear Regulatory Commission, Washington, DC, September 1998, ML081080310 and ML081080320.
6.8 NRC, Regulatory Guide 1.57, Revision 2, Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components, May 2013, ML12325A043.
6.9 NRC, RG 1.136, Revision 4, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containment, Washington, DC, February 2021, ML20301A167.
6.10 NRC, NUREG-1793, Final Safety Evaluation Report Related to the Certification of the AP1000 Standard Design, (NUREG-1793, Initial Report), Washington, DC, September 2004.
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6.11 NRC, NUREG-1966, Final Safety Evaluation Report Related to Certification of the Economic Simplified Boiling-Water Reactor Standard Design, Washington, DC, April 2014, ML14100A304.
6.12 NRC, Advance Power Reactor 1400 (APR1400) Final Safety Evaluation Report, Washington, DC, September 2018, ML18087A364.
6.13 NRC, Final Safety Evaluation Report for the NuScale Standard Plant Design, Washington, DC, August 2020, ML20231A804.
6.14 NRC, RG 1.82, Revision 5, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Washington, DC, August 2022, ML22152A114.