ML25262A075
| ML25262A075 | |
| Person / Time | |
|---|---|
| Issue date: | 09/19/2025 |
| From: | Elijah Dickson NRC/NRR/DRA |
| To: | |
| References | |
| Download: ML25262A075 (1) | |
Text
Margin Recovery Assessment for Environmental Qualification Accident-specific Integrated Dose Rates Elijah Dickson, Ph.D.
Sr. Reliability and Risk Analyst Division of Risk Assessment Office of Nuclear Reactor Regulation September 24, 2025
Assessment Approach
- Review
- The Issue - Impact of Updated Accident Source Terms on Environmental Qualification of Equipment
- Applicable Regulations - 10 CFR 50.49(e)(4), 100.11, and 50.34
- Current Agency Position - Generic Issue 187
- Agencys Policy on Backfitting and Forward - FRN 64 71990, NUREG/BR-0058, NUREG/BR-0184
- Traditional Method to Develop EQ Integrated Dose Curves - How are these calcs done?
- Leverage Modern Reactor Accident Analysis Methods
- Investigate current reactor accident analysis and review methods and to propose ways, if possible, to modernize the approach while ensuring appropriate conservatisms in the regulatory accident analysis.
- Focus resources on limiting Equipment Dose Calculations (BWR suppression pool / piping)
- Calculations consisted of two steps:
Reproduce NEI results using their assumptions (where available)
Repeat the EQ dose calculations with each source term vintage, using DG-1425 reg. positions and modern analytical tools with realistic core design parameters.
At Issue
- NRC regulatory requirements include the environmental qualification of equipment for the duration that it is needed to perform its safety function.
This includes qualification for radiation.
Source terms are used to define the radiation environment for design purposes.
- Results of regulatory source term updates:
TID-14844 Reg Guide 1.183, Rev. 0, Rev.1 and SAND 2023-01313 amount or kinds and amounts of radionuclides.
specified radiation hazard environment
- Regulatory uncertainty regarding the use of updated accident source terms for equipment qualification.
- Impact: Need to backfit or forward fit?
0%
20%
40%
60%
80%
100%
120%
Noble Gases Halogens Alkali Metals Tellurium Molybdenum Accident Source Term: BWR Total Release Fractions TID-14844,Section V.A RG 1.183 Rev. 0, Table 1 RG 1.183 Rev. 1, Table 1 SAND2023-01313, Table 5-3 Question: Are EQ issues due to historical modeling and calculational assumptions or some underlying safety issue uncovered by updates to the regulatory accident source term?
Source: NEI White Paper - Impact of Higher Source Term fractions on EQ Doses (ML24165A085)
RG 1.183 Rev.0 > TID-14844 100hrs.
RG 1.183 Rev.1 > TID-14844 10.5hrs.
SAND2023-01313 (RG 1.183 Rev. 2) > TID-14844 9.5hrs.
At Issue (Cont.)
Applicable Regulations The regulation containing the requirements for environmental qualification is 10 CFR 50.49. Regarding the radiation environment, 10 CFR 50.49(e)(4) requires:
The radiation environment must be based on the type of radiation, the total dose expected during normal operation over the installed life of the equipment, and the radiation environment associated with the most severe design basis accident during or following which the equipment is required to remain functional, including the radiation resulting from recirculating fluids for equipment located near the recirculating lines and including dose-rate effects.
This requirement has generally been met using the TID-14844 source term (a fuel melt source term that far exceeds design basis accident source terms), originally developed to meet the requirements in 10 CFR 100.11. As noted in the regulation, this source term is:
based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.
Bubble Chart of Normalized DBA Source Terms using RG 1.183 Rev. 0 Core Inventory (100%)
MHA-LOCA Source Term
(~25%)
Control Rod Ejection Source Term
(~1.4%)
Fuel Handling Source Term
(~0.002%)
Reactor Coolant Source Term
(~0.0003%)
10 CFR 50.46 LOCA Source Term
(~0.0003%)
Each fuel melt source term vintage bounds all DBA source terms specifically required by 10 CFR 50.49(e)(4).
Applicable Regulations (cont.)
Source terms required to be assessed by 10 CFR 50.49(e)(4).
Current Agency Position Federal Register Notice 64 71990, Use of Alternative Source Terms at Operating Reactors, December 23, 1999.
- NUBARG Comment Response (Backfitting).
When radiological consequence analyses are involved, the NRC expects to use a technically appropriate AST in evaluating generic and plant-specific backfitting analyses, including those proposed for facilities that have not implemented an AST. The NRC agrees with the NUBARG position that the NRC has discretion to take all new information on accident source terms into account.
The NRC's guidance for evaluating proposed NRC regulatory actions (including backfitting) are contained in NUREG/BR-0058, "Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission, and NUREG/BR-0184, "Regulatory Analysis Technical Evaluation Handbook." These documents state that value and impact (including adverse effects on health and safety) parameters are to be best estimates, preferably mean or expected values. These documents also provide that analyses are to be based largely on risk considerations.
FOOTNOTE 2: As provided in § 50.109, Backfitting is defined as the modification of or addition to systems, structures, components, or the design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct or operate a facility; any of which may result from a new or amended provision in the Commission rules or the imposition of a regulatory staff position interpreting the Commission rules that is either new or different from a previously applicable staff position.
Current Agency Position (cont.)
Federal Register Notice 64 71990, Use of Alternative Source Terms at Operating Reactors, December 23, 1999.
NEI Comment Response (Equipment Qualification).
The NRC has determined that it is necessary to consider the potential impact of the postulated cesium concentration in the containment sump water as it applies to all operating power reactors, not just to those licensees amending their design basis to use an AST.
Since the postulated increase in the integrated dose occurs only following an accident, there is no adverse effect on equipment relied upon to perform safety functions immediately following an accident.
Rather, this issue affects equipment that is required to be operable longer than about 30 days to 4 months after an accident.
As such, the NRC determined that continued plant operation does not pose an immediate threat to public health and safety.
Also, should such long-term equipment fail there will not be an undue threat to public health and safety as protective actions for the public would have already been implemented by the time the postulated failure could occur.
In addition, the time period between the onset of the event and the projected failure allows compensatory measures to be taken to prevent the equipment failure or to restore the degraded safety function.
The NRC will evaluate this issue as a generic safety issue to determine whether further regulatory actions are justified. The final regulatory guide, or subsequent revisions thereto, is expected to reflect the resolution of this generic safety issue.
Current Agency Position (cont.)
Generic Issue 187: The Potential Impact of Postulated Cesium Concentrations of Equipment Qualification CONCLUSION The staff concluded that there was no clear basis for backfitting the requirement to modify the design basis for equipment qualification to adopt the AST. There would be no discernible risk reduction associated with such a requirement. Licensees should be aware, however, that a more realistic source term would potentially involve a larger dose for equipment exposed to sump water for long periods of time. Longer term equipment operability issues associated with severe fuel damage accidents, (with which the AST is associated) could also be addressed under accident management or plant recovery actions as necessary. Thus, the issue was DROPPED from further pursuit.1776
Traditional method to develop EQ integrated dose curves.
- Traces back to the 1960s and 1970s.
- Lump parameter modeling - coarse assumptions and worst-case scenario inputs.
- Advantages.
- Evaluation model - simple to execute and review.
- Intention - Ensures robust safety margin.
- Disadvantages.
- Large analytical margin which can limit operational margin - rarely, if ever, quantified.
- Misrepresentative - without understanding EQ analysis purpose and underlying assumptions can lead to risk-communication misunderstanding.
- Deterministic without risk-insights - no consideration of compensatory measures.
(1) Reactor Core Inventory (2) Accident Source Terms (3) Transport and Mitigation of Accident Source Term (4) Radiation (,)
Spectra at plant-specific locations (5) Shielding and Dose Rate calculations (6) Integrated Dose Rates
(2) Depletion Analysis Oak Ridge ORIGEN code methodology Max U235 weight-%.
Max Burnup.
No radial or axial.
dependencies.
O-D modeling.
Evaluation Model Example (1) Simplify Operational Parameters Single assembly to represent core.
simulated gamma spectra (3) System Transport SNAP/RadTrad code methodology Simple compartment modeling from containment to sump, suppression pool, or recirc pipe.
Reactor core nuclide inventory as a function of enrichment and burnup Containment sump (4) Simulate Gamma Emission Spectra Data Analysis and prep for shielding calculations (5) Shielding and Dose Analysis MCNP code methodology Simple geometries like a component exposed to a 100ft pipe length RCS pipe Pipe Pipe Component
Source: NEI White Paper - Impact of Higher Source Term fractions on EQ Doses (ML24165A085)
RG 1.183 Rev.0 > TID-14844 100hrs.
RG 1.183 Rev.1 > TID-14844 10.5hrs.
SAND2023-01313 (RG 1.183 Rev. 2) > TID-14844 9.5hrs.
Evaluation Model Example (cont.)
- Areas of Modernization or Margin Recovery:
(1) Reactor Core Inventory
- Lots of Margin.
- Model, conservatively, realistic core operating parameters (2) Accident Source Terms
- Not much Margin.
- Would need to adjusting fundamentals (3) Transport and Mitigation
- Lots of Margin.
- Improvements in source term transport models and mitigation (4) Radiation
(,) Spectra
- None.
- Physics is Physics (5) Shielding and Dose Rate
- Lots of Margin.
- Simplistic to Realistic Geometries Margin Recovery Assessment Approach
- Focus - Margin Recovery: Reactor Core Inventory (1) Reactor Core Inventory
- Lots of Margin.
- Model, conservatively, realistic core operating parameters
- Approach
- Replace lump-parameter assumptions with realistic, yet defensible, ones.
Lump Realism Enrichment:
Max U235 Enrichment:
Weighted Ave.
Burnup:
Full Core Max Burnup:
Batch Dependent 0-D:
Power shape, Moderator den.,
Fuel comp.,
1-D:
Power shape, p()
Moderator den., m Fuel comp.,
Provides info on:
Realistic core designs 5 to 10% U235 enrichment 60-80 GWd/MTU batch-cycle burnups 2-3 year cycle lengths Margin Recovery Assessment Approach (Cont.)
New SCALE tool, Fuel Cycle Estimator, quickly generates realistic inventories based on specific LWR core loading characteristics, including HBU/IE
- Focus - Margin Recovery: Reactor Core Inventory: Enrichment
- BWR fuel have varying enrichment along its axis and per-pin N-T - 15.24 cm, Natural Enrichment VAN - 229.60 cm, Max = 6%, Ave. = 5.21 U235 DOM - 91.44 cm, Max = 6%, Ave. = 5.17 U235 NAT - 30.48 cm, Natural Enrichment Lump Realism Enrichment:
Max U235 Enrichment:
Weighted Ave.
Margin Recovery Assessment Approach (Cont.)
- Focus - Margin Recovery: Reactor Core Inventory: Burnup
- BWR fuel have varying burnup loading patters Lump Realism Burnup:
Full Core Max Burnup:
Batch Dependent Once burned Twice burned Core Burnup (GWd/MTU)
Ave Max Ave 6%
24 52.96 59.41 39.39 Discharg Burnup (GWd/MTU)
Max Enrichment Cycle Length (m)
Margin Recovery Assessment Approach (Cont.)
Margin Recovery Assessment Approach (Cont.)
- Focus - Margin Recovery: Reactor Core Inventory: Spatial Variation
- BWR fuel have varying moderator density and power shaping along its axis radially with core Lump Realism 0-D:
Power shape, Moderator den.,
Fuel comp.,
1-D:
Power shape, p()
Moderator den., m Fuel comp.,
Margin Recovery Assessment Approach (Cont.)
RG 1.183 Rev.0 RG 1.183 Rev.1 RG 1.183 Rev.2 TID-14844
- TID-14844 as an accident source term could still be acceptable when acknowledging the degree of conservatisms built into the original licensing basis.
- Realistic (but still conservative) operational core parameters to define the reactor core inventory demonstrates sufficient analytical margin and reinforces original licensing basis safety case.
- Regulatory design criteria and methods are intended for defense-in-depth and safety margin rather than from actual events.
- None of the accident source terms were intended to protect against a particular event. These source terms have always served as useful design inputs for the purposes of defense-in-depth and safety margin.
- GI-187 conclusion is still applicable to update accident source terms
- Longer term equipment operability issues associated with severe fuel damage accidents, (with which the AST is associated) are addressed under accident management or plant recovery actions, as necessary.
Key Takeaways
References Federal Register Notice 64 71990, Use of Alternative Source Terms at Operating Reactors, December 23, 1999 NRC, Regulatory Guide 1.183 Rev 0), Alternative Radiological Source Terms for Evaluating DBA at Nuclear Power Reactors. dated July 2000.
NRC, Regulatory Guide 1.183 Rev 1), Alternative Radiological Source Terms for Evaluating DBA at Nuclear Power Reactors. dated October 2023.
NRC, DG-1425 (RG 1.183 Rev 2), Alternative Radiological Source Terms for Evaluating DBA at Nuclear Power Reactors. ACRS Version, dated 2024.
(ML24304A864).
NEI, White Paper: Impacts of Higher Source Term Release Fractions on Environmental Qualifications, Nuclear Energy Institute, dated June 2024.
Sandia National Laboratories, SAND2023-01313, High Burnup Fuel Source Term Accident Sequence Analysis, Albuquerque, New Mexico, April 2023 (ML23097A087).
Sandia National Laboratories, SAND2024-10674, Multi-region Tabular Source Terms for BWR Containment Design Leakage Assessments, Albuquerque, New Mexico, June 2024 (ML24229A044).
NRC, Management Directive 8.4, Management of Backfitting, Forward fitting, Issue Finality, and Information Requests, September 20, 2019.
Title 10 of the Code of Federal Regulations (10 CFR), Section 50.109, Backfitting NUREG/BR-0058, Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission, Revision 4 Resolution of Generic Safety Issues: Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification ( NUREG-0933, Main Report with Supplements 1-35 )
Memorandum for A. Thadani from B. Sheron, "Proposed Generic Safety Issue The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump," December 16, 1999. (ML993610109)