ML20296A425
| ML20296A425 | |
| Person / Time | |
|---|---|
| Issue date: | 11/19/2020 |
| From: | Mark Blumberg, Matthew Smith NRC/NRR/DRA/ARCB |
| To: | |
| Smith M | |
| References | |
| RG-1.183 | |
| Download: ML20296A425 (28) | |
Revision of Regulatory Guide 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors Mark Blumberg, Senior Reactor Engineer (Technical Lead)
NRR/DRA/ARCB
mark.blumberg@nrc.gov Micheal Smith, Health Physicist (Project Lead)
NRR/DRA/ARCB
micheal.smith@nrc.gov November 19, 2020
Agenda
- Key Messages
- Background
- Regulatory Guide (RG) Update Process
- RG 1.183 Guidance Updates Under Consideration
- Looking Forward
- Feedback/Discussion
- Comments and input from the public
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Key Messages
- The NRC staff has restarted efforts to revise RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.
- The objectives of the revision are to:
- incorporate lessons learned from recent NRC staff reviews of Alternative Source Term (AST) and Main Steam Line Isolation Valve (MSIV) leakage LARs;
- incorporate relevant operating experience as well as recent post-Fukushima seismic risk insights and walkdowns;
- respond to change of regulatory environment (e.g., backfit guidance SRM-
SECY-18-0049 & NuScale SRM-SECY-19-0036);
- make the guidance more useful by considering feedback and comments from licensees;
- ensure sufficient guidance is in place for licensing advanced light-water reactors (LWRs), accident tolerant fuel (ATF), high-burnup, and increased enrichment fuel; and,
- incorporate insights from new research activities.
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Key Messages (Contd)
- RG 1.183 Rev. 0 and Rev. 1 will co-exist as a result of SRM-
SECY-18-0049, Management Directive and Handbook 8.4, Management of Backfitting, Issue Finality, and Information Collection.
- NRC staff will hold public meetings for external stakeholder engagement on the revision of RG 1.183.
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Background
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Background
- Origin: Footnote to 10 CFR 100.11(a) is a performance-based rule to evaluate the defense-in-depth provided by the containment.
- TID-14844 Source term provided guidance which assumed the source term is instantaneously available in the containment.
- Radionuclide behavior observed during the TMI accident did not appear at all similar to the TID-14844 source term.
- NRC initiated research effects in the area of severe accidents which culminate in publication of NUREG-1150.
- NUREG-1465 source term was derived from the sequences in NUREG-1150.
- RG 1.183 Rev. 0 adopted the NUREG-1465 early in-vessel fuel melt source term.
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Background (contd)
- NRC staff developed RG 1.183 Rev. 0 (July 2000) to support implementation of 10 CFR 50.67, Accident source term
- RG 1.183 Rev. 0 is applicable to nuclear power reactor applicants and licensees adopting 10 CFR 50.67
- Limited range of applicability on Non-LOCA release fractions
- RG 1.183 Rev. 0 identified the significant attributes of an acceptable accident AST based on NUREG-1465, Accident Sources Terms for Light-Water Nuclear Power Plants (1995)
- RG 1.183 Rev. 0 provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST
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DG-1199
- In October 2009, the NRC issued for public comment DG-1199 as a proposed Rev. 1 of RG 1.183.
- Staff received 150 public comments
- The reasons for revision of RG 1.183 in DG-1199 were:
- Providing additional guidance for modeling BWR MSIV leakage,
- Expand applicability of Non-LOCA release fractions to support modern fuel utilization,
- Extending the applicability of the proposed RG for use in satisfying the radiological dose analysis requirements contained in 10 CFR Part 52 for advanced LWR design and siting,
- Providing additional meteorological assumption guidance.
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Modern Fuel Utilization
- Since DG-1199 was issued for public comment, NRC issued several license amendments to support modern fuel utilization.
- Oconee Units 1, 2, and 3 (2019)
- Shearon Harris (2018)
- H.B. Robinson (2017)
- Catawba Units 1 and 2, McGuire Units 1 and 2, Oconee Units 1, 2, and 3 (2016)
- Diablo Canyon Units 1 and 2 (2015)
- Reinforced need for expanded Non-LOCA release fractions
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2019 License Amendment Requests
- As a result, work on DG-1199 was postponed to allow NRC staff to incorporate lessons learned, from evaluation of the LARs, into the revised RG 1.183:
- James A. FitzPatrick Amendment No. 338 for AST, July 21, 2020
- Quad Cities Nuclear Power Station, Units 1 & 2 - Amendment Nos. 281 and 277 to increase allowable MSIV leakage, June 26, 2020
- Nine Mile Point Nuclear Station, Unit 2 - Amendment No. 182 to change allowable MSIV leak rates, October 20, 2020 (ML20241A190)
- Dresden Nuclear Power Station, Units 2 & 3 - Amendments Nos. 272 and 265 to increase allowable MSIV leakage, October 23, 2020
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Regulatory Guide Update Process
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Regulatory Guide Update Process
- Identify which RGs need to be revised based on:
- Rulemakings
- Lessons learned
- Stakeholder feedback
- Periodic reviews
- Develop draft RG through internal collaboration
- Internal staff comment resolution
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RG 1.183 Guidance Updates Under Consideration
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Expected General Updates
- The intent of the NRC staff is for RG 1.183 Rev. 0 and Rev. 1 to co- exist
- With the exception of items discussed later, NRC will consider changes proposed in DG-1199 as modified by public comments.
- Incorporate updates, new or withdrawn regulatory guidance (i.e., RG 1.194 (meteorology)).
- Guidance for modern fuel utilization (non-LOCA gap fractions).
- Changes due to Regulatory Information Summaries (i.e., 06-04,
01-19).
- Lessons learned from license reviews (i.e., clarify DFs and containment isolation as used in the FHA).
- Clarify TEDE calculation terminology (i.e., EDEX vs. EDE).
- Remove environmental qualification guidance from RG and refer to RG 1.89.
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ATF, High-Burnup, Extended Enrichment
- Applicability of Rev.1 expanded to encompass fuel burnup extension to 68 GWd/MTU (rod average) and 235U enrichments up to 8.0wt%.
- Applicability of Rev.1 to near-term ATF design concepts being considered.
- Non-LOCA release fractions sensitive to fuel design
- Utilize accident source terms from Sandia National Laboratories report SAND2011-0128, Accident Source Terms for Light Water Nuclear Power Plants Using High-Burnup of MOX Fuel, and non-loss-of-coolant accident (non-LOCA) source terms based on FAST calculations (similar to those calculated in the proposed update to RG 1.183, Draft Guide 1199).
NRC Memorandum, Applicability of Source Term for Accident Tolerant Fuel, High Burn Up and Extended Enrichment, dated May 13, 2020, ADAMS Accession Number ML20126G376
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Draft Guide DG-1199 Non-LOCA Release Fractions DG-1199 (2011) included the following components:
1. Revised Table 3 Non-LOCA release fractions based on expanded power profile
2. New Table 4 RIA transient fission gas release fractions
3. New analytical procedure for revising release fractions
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Planned Updates for Non-LOCA Release Fractions
1. Maintain Table 3 release fractions up to 62 GWd/MTU rod average burnup
2. New table for release fractions with expanded applicability up to 68 GWd/MTU rod average burnup
3. Update Table 4 RIA transient fission gas release to include burnup-dependent correlations
4. Update example calculation based on FAST
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DG-1327 CRE/CRD Public Comments (2019)
- Many of the planned changes to RG 1.183 were included in draft regulatory guide DG-1327
- Revised Table 3 Non-LOCA gap fractions using new version of FAST fuel rod thermal-mechanical code
- Revised Table 4 RIA transient fission gas release with BU-dependent correlations
- Acceptable analytical procedure for revising Non- LOCA gap releases
- Public comments received on these topics will be reflected in RG 1.183
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FAST Calculations (1)
- Extended rod average power profiles out to 68 GWd/MTU
- Preliminary calculations show no increase in release fractions
- Axial Power Distribution:
- Sweeping (3 cycles) AXPDs with the following peak Fz peaking factors
- PWR Peak Fz = 1.144
- BWR Peak Fz = 1.228
- Is this sufficient to support future reloads?
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FAST Calculations (2)
tables?
- How to address BWR
part-length fuel rods?
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Revised Fuel Handling Accident
- Revisited the original studies forming the technical basis for the FHA and incorporate updated information.
- Model improvements established from the current understanding of reactor fuel pin physics and iodine chemistry under the environmental conditions in which fuel handling operations are taking place.
- Concluded that considerable margin exists regarding the scrubbing effects of iodine in the spent fuel or reactor pool and that the current staff DBA FHA fission product transport model can be refined while still maintaining conservatism.
- Reference: Memo from RES to NRR, Closeout to Research Assistance Request for Independent Review of Regulatory and Technical Basis for Revising the Design-basis Accident Fuel Handling Accident, November 23, 2019 (ML19270E335)
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Additional Method for Aerosol Deposition Models
- Staff is considering an additional method for aerosol deposition models
- Staff is addressing issues in RIS 2006-04, Experience with Implementation of Alternative Source Terms (considering reconstitution of AEB-98-03 and reviewing the multigroup method).
- Regulatory position in Rev. 0 continues to be acceptable. As a result, RG
1.183 Rev. 0 and Rev. 1 will co-exist.
- Over the last 10 years no applicant or licensee has adopted the methodology from SAND2008-6601, Analysis of Main Steam Isolation Valve Leakage in Design Basis Accident Using MELCOR 1.8.6 and RADTRAD.
- There have been no communications that applicants or licensees intend to adopt the SAND2008-6601 methodology.
- NRC staff plans to consider stakeholder input/feedback to inform the NRCs decision on what methodology to include in RG 1.183 Rev. 1.
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Lessons Learned from Licensing Reviews
- Staff are considering whether to clarify:
- the expectations for containment spray in BWR
drywells/containments (i.e., Rev. 0 Appendix A Assumption 3.3)
- the expectations for performing and using sensitivity analysis (i.e., Rev. 0, RPs 1.3.3 and 5.1.3)
- if crediting pathways should be consistent with design requirements for safety (i.e. technical specifications, safety related, Rev. 0, RP 5.1.2)
- RG wording to assume that a LOOP is coincident with a turbine trip (not with initiation of the accident)(i.e., Rev. 0, App. F & G
Assumption 5.4)
- the expectations for BWR MSIV Leakage LOCA analysis assumptions with respect to pipe breaks
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Use of Risk and Engineering Insights
- Update the expectations for use of risk insights as directed in SRM-
- NRC staff has developed a technical assessment on this topic considering 20+ years of operational and seismic risk insights.
- Assessment will be publicly available via the NRCs Interim Staff Guidance process.
- Four issued safety evaluations are supported by risk and engineering insights.
- Staff is exploring streamlined approach for quantitative credit for hold-up and retention of MSIV leakage within the power conversion system for BWRs.
- Is there interest in a streamlined approach?
- What portion(s) of the alternative pathway justification in Rev. 0 are resource intensive (availability of pathway, seismic robustness steps, both?)
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Additional Considerations
- Consider revising footnote 7 which provides an incorrect method to convert thyroid dose to TEDE
- Implies a back-of-the-envelope calculation appropriately converts between ICRP 2 and ICRP 26/30 dosimetry methodologies.
- There is no simple methodology to convert between these two systems of dosimetry.
- To correctly calculate the radiological dose consequences for design basis accidents the appropriate dose methodology (and DCFs) must be applied.
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Looking Forward
- Consider feedback from stakeholders
- Develop updated draft RG 1.183 Rev. 1
- Hold additional public meeting 1st Quarter CY 2021
- Staff review and disposition of public comments
- Update of draft RG 1.183 Rev. 1 as necessary
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Discussion/Feedback
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Questions/Comments?
Mark Blumberg, Senior Reactor Engineer (Technical Lead)
NRR/DRA/ARCB
mark.blumberg@nrc.gov Micheal Smith, Health Physicist (Project Lead)
NRR/DRA/ARCB
micheal.smith@nrc.gov
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