Text

Comment (2) of Kati R. Austgen on Draft Regulatory Guide: General Site Suitability Criteria for Nuclear Power Stations
	ML23326A031
Person / Time
Site:	Nuclear Energy Institute
Issue date:	11/17/2023
From:	Austgen K; Nuclear Energy Institute
To:	; Office of Administration
References
	88FR7177 00002, NRC-2023-0153, DG-4034
	Download: ML23326A031 (1)
	v • d • e

11/22/23, 8:22 AM blob:https://www.fdms.gov/e207b396-e5f4-4d3f-ab34-4c68b674c77b SUNSI Review Complete Template=ADM-013 As of: 11/22/23, 8:21 AM E-RIDS=ADM-03 Received: November 17, 2023 PUBLIC SUBMISSION ADD: Bridget Curran; Status: Pending_Post Edward O'Donnell, Tracking No. lp3-a6s8-gi7h Mary Neely Comments Due: November 17, 2023 Comment (2)

Publication Date:

Submission Type: Web 10/18/2023 Citation 88 FR 7177 Docket: NRC-2023-0153 Draft Regulatory Guide: General Site Suitability Criteria for Nuclear Power Stations Comment On: NRC-2023-0153-0001 Draft Regulatory Guide: General Site Suitability Criteria for Nuclear Power Stations Document: NRC-2023-0153-DRAFT-0003 Comment on FR Doc # 2023-22980 Submitter Information Email: atb@nei.org Organization: Nuclear Energy Institute General Comment See attached file(s)

Attachments 11-17-23_NRC_NEI Comments on DG-4034 NEI White Paper - Advanced Reactor Population-Related Siting Considerations blob:https://www.fdms.gov/e207b396-e5f4-4d3f-ab34-4c68b674c77b 1/1

Kati R. Austgen Phone: 202.340.1224 Sr. Project Manager, New Nuclear Email: kra@nei.org November 17, 2023 Office of Administration Mail Stop: TWFN-7-A60M U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Program Management Announcements and Editing Staff

Subject:

Comments on Draft Regulatory Guide (DG), DG-4034, General Site Suitability Criteria for Nuclear Power Stations. (Docket ID: NRC-2023-0153) (Federal Register Notice 88 FR 71777)

Project Number: 689 Submitted via Regulations.gov

Dear Program Management,

Announcements, and Editing Staff:

The Nuclear Energy Institute (NEI)1, on behalf of its members, submits the following comments on Draft Regulatory Guide (DG), DG-4034, General Site Suitability Criteria for Nuclear Power Stations, published in the Federal Register on October 18, 2023, for public comment by November 17, 2023.

The most significant comment on the NRCs DG-4034 is regarding the dose criterion for advanced reactors, which sets the population density distance (PDD) at twice the distance to 1 rem in 30 days. The original basis for the PDD was to minimize the potential for permanent relocation due to land contamination, and to meet the qualitative health objectives (QHOs) for latent cancer fatalities. The Commission has determined that the current 20 mile PDD for large light water reactors (LLWRs) provides adequate protection of the public health and safety, however, to our knowledge, the NRC has not analyzed the anticipated doses for the current 20 mile PDD for LLWRs to determine whether the proposed dose criterion of twice the distance to 1 rem in 30 days for advanced reactors provides a similar level of protection.

In the attached paper, NEI performed these evaluations and concluded that the NRCs proposed criterion associated with very low frequency events, twice the distance of 1 rem in 30 days is more than a factor of five times more conservative compared to previous LLWR licensing when calculated using best estimate consequence methods. That is, 1 rem is approximately five times lower than the level of protection currently accepted for LLWRs using a population density distance of 20 miles. Doubling the distance to 1 rem in 30 days for advanced reactors will further lower the dose, resulting in more than a factor of five conservatism given the decrease in dose with distance, e.g., 4.7 rem/0.36 rem = a factor of 13. Under more conservative modeling assumptions, the NRCs proposed dose for advanced reactors could be more than 70 times more conservative compared to previous LLWR licensing. For example, use of a 95% weather in lieu of best estimate mean weather yields a factor of 5.5 increase in conservatism; 13 x 5.5 = a factor of 71.5.

The NRCs Policy Statement on the Regulation of Advanced Reactors states, Regarding advanced reactors, the Commission expects, as a minimum, at least the same degree of protection of the environment and public 1

The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.

Office of Administration Nuclear Energy Institute November 17, 2023 Page 2 health and safety and the common defense and security that is required for current generation light-water reactors (LWRs). Furthermore, the Commission expects that advanced reactors will provide enhanced margins of safety and/or use simplified, inherent, passive, or other innovative means to accomplish their safety and security functions. Although the NRC policy statement might permit a performance-based dose criterion for advanced reactors that is somewhat more restrictive than that of LLWRs, the staffs proposal would impose an unnecessary level of regulatory burden.

In order to meet the expectations of the NRC Advanced Reactor Policy Statement, the NRC should revise the dose criterion for the PDD of advanced reactors, even if this requires the Commissions approval. We recommend that the dose criterion for the PDD of advanced reactors be established as 5 rem in 30 days (rather than the currently proposed PDD of 2x the distance of 1 rem in 30 days). This would align with the current NRC accepted level of protection for LLWRs calculated using best estimate methods at 20 miles. This alternative should be pursued to acknowledge the features of advanced reactors and to avoid undue burden on the owner of an advanced reactor such that population related siting considerations do not unnecessarily limit allowable sites when compared to the accepted level of protection for existing plants by aligning requirements with advanced reactor safety profiles. As an example of the impacts, the paper calculates that the PDD for a representative advanced reactor would be 300 meters (using the NRC accepted current level of protection), but would be 1,690 meters if the NRC imposes the proposed excessively conservative dose criterion. Furthermore, excessively conservative dose criterion for the PDD will have negative societal impacts in that it may potentially exclude the use of nuclear energy in some locations, where it would otherwise be needed to meet energy, environmental, and national security goals.

In the meantime, we encourage the NRC staff to use DG-4034 and the NEI paper to accept applications with proposals to meet a dose criterion of twice the distance of 1 rem in 30 days using best estimate modeling, since this is already approved by the Commission, to establish the PDD for advanced reactors.

While reviewing the guidance for an alternate PDD dose criterion, it was identified that other, related siting criteria such as the population center distance in 10 CFR 100.3 may benefit from the establishment of alternative dose-based criterion. We recommend the NRC engage with stakeholders in discussions to determine which other siting criteria should be addressed by alternative criteria for advanced reactors, commensurate with their safety profiles.

The full set of NEI comments are in the attached table. We appreciate the NRCs effort in developing this draft guidance and encourage your consideration of all stakeholder comments. We would be pleased to answer any comments or questions you might have on the contents of this letter and its attachments.

Sincerely, Kati R. Austgen, Sr. Project Manager, New Nuclear Attachments c Belkys Sosa, NRR, NRC Edward ODonnell, RES, NRC Steven Lynch, NRR, NRC Mohamad Shams, NRR, NRC Robert Taylor, NRR, NRC

Nuclear Energy Institute 1 NEI Comments on Draft Regulatory Guide (DG), DG-4034, General Site Suitability Criteria for Nuclear Power Stations Affected Section Comment/Basis Recommendation

1. GENERAL The draft continually refers to assessments Recommend using area of potential effect instead or adding it in regions or discusses surveys without into areas where direction is given to perform an activity.

specifics.

2. GENERAL Impacts appear to be consistently Modify language where appropriate to acknowledge that there discussed in a negative manner. are beneficial/positive aspects to nuclear power plants. This is especially true in the area of socioeconomics where there are a number of benefits to a community from the presence of a plant.

3. B. Discussion, Reference to Electric Power Research Update refence on this page and in the REFERENCES list to Scope of Institute document No. 3002005435, Site EPRI document No. 3002023910, Advanced Nuclear Regulatory Guide Selection and Evaluation Criteria for New Technology: Site Selection and Evaluation Criteria for New 4.7, Page 4, 1st Nuclear Power Generation Facilities (Siting Nuclear Energy Generation Facilities (Siting Guide)2022 paragraph Guide), issued June 2015 is out of date. Revision, which can be found at EPRI issued an update in November 2022. https://www.epri.com/research/products/000000003002023910

4. B. Discussion, It is not necessary to state, Selecting a Delete this sentence.

Site Selection, suitable site for a commercial nuclear Page 5, 1st power station may require a significant paragraph, 1st commitment of time and resources. Many sentence small modular or advanced reactors may be selected for use in specific locations for very specific purposes. This statement could discourage new potential applicants.

5. B. Discussion, The draft guide currently states, It also Proposed, It also reduces potential doses and property Population reduces potential doses and property damage from radiological contamination in the event of a Considerations, damage in the event of a severe accident. severe accident that results in a significant release of Page 6, 1st radiological material to the environment.

paragraph, 2nd Given that the property damage in view is sentence assumed to be associated with radiological contamination rather than direct physical damage that could occur from other types of events, and the population considerations are also in view of a

Nuclear Energy Institute 2 Affected Section Comment/Basis Recommendation significant radiological release, recommend adding clarification as shown with redline insertion.

6. B. Discussion, The draft guide currently states, The Proposed, For large LWRs, tThe numerical values in this guide Population numerical values in this guide are generally are generally consistent with past NRC practice and reflect Considerations, consistent with past NRC practice and consideration of severe accidents for large LWRs, as well as Page 6, 1st reflect consideration of severe accidents for the demographic and geographic conditions of the United paragraph, last large LWRs. States. For advanced reactors, the numerical values are sentence conservative with respect to past NRC practice and reflect Per dispersion calculations developed and consideration of EPA protective action guidelines.

discussed in the NEI white paper (Advanced Reactor Population-Related Siting Considerations), the proposed 500 ppsm dose criterion of 2x the distance to 1 rem in 1 month for advanced reactors is quite conservative as compared to the current level of protection accepted for large LWR siting. Per NRC discussion during the October 27, 2023, public meeting, the advanced reactor dose criterion is significantly based on the desire to have the dose criterion below that associated with the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) of 1 rem in 4 days and 2 rem in 1 year.

As described in our comments on Appendix A and in the NEI White Paper Advanced Reactor Population-Related Siting Considerations, the basis for the proposed dose criterion should be reconsidered. At the very least, clarification on the basis of the currently proposed values for advanced

Nuclear Energy Institute 3 Affected Section Comment/Basis Recommendation reactors should be added as shown with redline/strikeout.

7. B. Discussion, Consider adding clarifying language, as Based on plant design (e.g., large light water reactor (LLWR),

Hydrology, Page shown with redline/strikeout. advanced reactor), cConsumption of water may necessitate an 7, Water evaluation of existing and future water uses in the area to Availability ensure adequate water supply during droughts for both station operation and other water users (i.e., commercial nuclear power station requirements versus public water supply).

8. B. Discussion, Consider adding clarifying language, as When the ecological sensitivity of a site within the area of Ecological shown with redline/strikeout. potential effect under consideration cannot be established from Systems and existing information that would be considered representative of Biota, Page 9, 2nd the current ecological system, more detailed studies, as paragraph discussed in RG 4.2, Rev. 3, might be necessary.

9. B. Discussion, Consider adding clarifying language, as Based on location, nNoise levels at nuclear stations during both Noise, Page 9 shown with redline/strikeout. construction and operation could have undesirable impacts. For example, cooling towers (if present), turbines, and transformers contribute to the noise during station operation, and such noise could have varying levels of environmental impact, depending on the site.

10. C. Staff It may be useful for an evaluation of the An evaluation of the water requirements for the ultimate heat Regulatory water requirements for the ultimate heat sink should consider, if available, a minimum 30-year weather Guidance, Page sink to consider a minimum 30-year record and should follow the guidance provided in RG 1.27.

13, Section weather record, but this should not be 1.2.1.3, 2nd strictly necessary. In any case, the NRC paragraph staff expect the applicability of data to be substantiated. Consider clarification as shown with redline insertion.

11. C. Staff Suggest an insertion after the 1st paragraph Some reactor plant designs with relatively small cores, passive Regulatory to provide distinction regarding unique safety features, or other design features, which are anticipated Guidance, Page features of advanced reactors. to result in smaller postulated accident releases and associated 16, Section 1.3 radiological doses, may identify the potential for the exclusion area boundary (EAB) to be smaller than historically determined for LLWRs.

Nuclear Energy Institute 4 Affected Section Comment/Basis Recommendation

12. C. Staff Suggest an insertion after the 1st paragraph Some reactor plant designs with relatively small cores, passive Regulatory to provide distinction regarding unique safety features, or other design features, which are anticipated Guidance, Page features of advanced reactors. to result in smaller postulated accident releases and associated 17, Section 1.3.3 radiological doses, may identify the potential for the exclusion area boundary (EAB) to be smaller than historically determined for LLWRs.

13. C. Staff The draft guide misstates the content of 10 Correct the statement to reflect the rule text, As stated in 10 Regulatory CFR 100.21(h) as follows (emphasis CFR 100.21(h), reactors should be located away Guidance, Page added), As stated in 10 CFR 100.21(h),

18, Section 1.4.3, reactors are to be located away from very 1st paragraph densely populated centers; areas of low population density are generally preferred.

The rule text states, reactor sites should.

14. C. Staff Insert clarifying text as shown with redline ...for a site located away from a very densely populated center Regulatory insertion. but not in an area of low density, acceptability will be Guidance, Page determined after consideration of safety, environmental, 18, Section 1.4.3, economic, and other factors. Other factors may include reactor 1st paragraph plant designs with relatively small cores, passive safety features, or other design features, which are anticipated to result in smaller postulated accident releases and associated radiological doses.

15. C. Staff The draft guide currently states, The Proposed, For large LWRs, tThe numerical values in this guide Regulatory numerical values in this guide are generally are generally consistent with past NRC practice and reflect Guidance, Page consistent with past NRC practice and consideration of severe accidents for large LWRs, as well as 18, Section 1.4.3, reflect consideration of severe accidents for the demographic and geographic conditions of the United 2nd paragraph large LWRs. States. For advanced reactors, the numerical values are conservative with respect to past NRC practice and reflect Per dispersion calculations developed and consideration of EPA protective action guidelines.

discussed in the NEI white paper (Advanced Reactor Population-Related Siting Considerations), the proposed 500 ppsm dose criterion of 2x the distance to 1 rem in 1 month for advanced reactors is quite conservative as compared to the

Nuclear Energy Institute 5 Affected Section Comment/Basis Recommendation current level of protection accepted for large LWR siting. Per NRC discussion during the October 27, 2023, public meeting, the advanced reactor dose criterion is significantly based on the desire to have the dose criterion below that associated with the Environmental Protection Agency (EPA) Protective Action Guides (PAGs) of 1 rem in 4 days and 2 rem in 1 year.

As described in our comments on Appendix A and in the NEI White Paper Advanced Reactor Population-Related Siting Considerations, the basis for the proposed dose criterion should be reconsidered. At the very least, clarification on the basis of the currently proposed values for advanced reactors should be added as shown with redline/strikeout.

16. C. Staff Insert clarifying text as shown with redline Other factors, such as safety, environmental, or economic Regulatory insertion. characteristics, may render the site with higher population Guidance, Page density acceptable including reactor plant designs with 19, Section 1.4.3, relatively small cores, passive safety features, or other design 1st full paragraph features, which are anticipated to result in smaller postulated accident releases and associated radiological doses.

Nuclear Energy Institute 6 Affected Section Comment/Basis Recommendation

17. C. Staff Add clarifying language to reflect the Insert after 3rd paragraph:

Regulatory forthcoming final rule, Emergency Guidance, Page Preparedness for Small Modular Reactors As a note, some reactor plant designs with relatively small 20, Section 1.5.3 and Other New Technologies, as shown cores, passive safety features, or other design features, which with redline insertion. are anticipated to result in smaller postulated accident releases, may follow alternative emergency preparedness requirements for small modular reactors and other new technologies per 10 CFR 50.160 which may reduce the size of their emergency planning zone.

18. C. Staff Consider adding clarifying language, as The Clinch River FSER highlights that emergency plans for one Regulatory shown with redline insertion. or more reactors should be considered for green-field, or retired Guidance, Page fossil fuel plant, siting of nuclear power plants.

21, Section 1.5.3, final paragraph, final sentence

19. C. Staff Add clarifying text after the last paragraph As a note, some reactor plant designs with relatively small Regulatory in Section 1.5.3, as shown with redline cores, passive safety features, or other design features, which Guidance, Page insertion. are anticipated to result in smaller postulated accident releases, 21, last may follow alternative emergency preparedness requirements paragraph before for small modular reactors and other new technologies per 10 Security section. CFR 50.160 which may reduce the size of their emergency planning zone.

20. C. Staff Consider adding clarifying language, as Both the frequency and duration of periods of low flow or low Regulatory shown with redline insertion. water level should be determined from a reasonable and Guidance, Page representative period of time based on the historical record 25, Section and, if the cooling water is to be drawn from impoundments, for 1.7.2.3, 1st projected operating practices.

paragraph, 2nd sentence

Nuclear Energy Institute 7 Affected Section Comment/Basis Recommendation

21. C. Staff Consider adding clarifying language, as The acceptability of a site depends on establishing that (1) an Regulatory shown with redline/strikeout. postulated accident at a nearby industrial, military, or Guidance, Page transportation facility will not result in radiological 26, Section 1.8.3, consequences that exceed the dose specified in 10 CFR 50.34, bottom or (2) such an postulated accident poses no undue risk paragraph, 1st because the annual frequency of its occurring is sufficiently low sentence (less than about 1x10-7 per year).

22. C. Staff Numbering is off in Section C.2, e.g., Generic editorial correction to check section numbers and Regulatory subsection numbering goes from 2 to 2.1.1. sequencing.

Guidance, Later subsections often skip 2.X.1 or Section C.2, 2.X.X.1, e.g. 2.1.3 is followed by 2.1.3.2 on Page 27-48 Page 29.

23. C. Staff Add clarifying language, as shown with Depending on the plants cooling system design, wWater and Regulatory redline/strikeout. water vapor released to the atmosphere Guidance, Page 29, Section 2.1.3.4, 1st paragraph

24. C. Staff Add clarifying language, as shown with Depending on the plants cooling system design, Regulatory redline/strikeout. cConcentrations of chemicals, dissolved solids, and suspended Guidance, Page solids in cooling tower drift could affect terrestrial biota within 30, Section the area of potential effect and result in unacceptable damage 2.1.4.4, 2nd to vegetation and other resources.

paragraph

25. C. Staff Add clarifying language, as shown with The potential impacts of commercial nuclear power stations on Regulatory redline/strikeout. water quality would are likely to be acceptable if they satisfy Guidance, Page effluent limitations, water quality criteria for receiving waters, 32, Section and other requirements under the CWA. The applicant should 2.2.2.4, 1st also identify any other relevant state, local, or federal paragraph regulations current at the time when it is considering sites.

26. C. Staff Modify text for clarity, as shown with Although effluent discharges are governed under the authority Regulatory redline/strikeout. of the Federal Water Pollution Control Act (FWPCA) to ensure Guidance, Page that established existing water quality standards are maintained 32, Section to be protective of the public and environment, wWhere station

Nuclear Energy Institute 8 Affected Section Comment/Basis Recommendation 2.2.2.4, 5th construction or operation could degrade water quality to the paragraph detriment of other users as a result of waters being listed as impaired under Section 303(d) of the CWA, more detailed analyses and evaluation of water quality may be necessary

27. C. Staff Modify text for clarity, as shown with If applicable, potential sources of cooling water should also be Regulatory redline/strikeout. screened to determine if the proposed plant cooling water Guidance, Page system meets all three criteria of the CWA 316(b) intake flow 34, Section limitations specified in 40 CFR 125.81: (1) is a point source that 2.2.3.4, 1st uses or proposes to use a cooling water intake structure; (2) paragraph has at least one cooling water intake structure that uses at least 25 percent of the water it withdraws for cooling purposes based on an average monthly measurement; and (3) has a design intake flow greater than two (2) million gallons per day. by their capacity to meet intake flow limitations specified in CWA section 316(b), as implemented by the EPA in 40 CFR Parts 9, 122, 123, 124, and 125. CWA section 316(b) identifies criteria based on type of water body in order to reduce environmental impact. It is typically administered by State programs.

28. C. Staff Consider adding clarifying language, as The following are examples of potential environmental effects Regulatory shown with redline insertion. of station construction and operation that should be assessed Guidance, Page and which the applicant can reference per NUREG-1437, 34, Section Revision 1, when addressing operational impacts, as 2.2.3.4, 2nd applicable:

paragraph

29. C. Staff Consider adding clarifying language, as The ecological systems and biota at potential sites and their Regulatory shown with redline insertion. environs within the area of potential effect should be sufficiently Guidance, Page well known 36, Section 2.3, 1st paragraph, 2nd sentence

30. C. Staff Consider adding clarifying language, as To limit the potential for entrapment of aquatic organisms by Regulatory shown with redline insertion. intake or discharge structures, evaluations of potential sites Guidance, Page should consider of the entrainment, impingement and heat 41, Section 2.3.5

Nuclear Energy Institute 9 Affected Section Comment/Basis Recommendation shock requirements of applicable Federal, State and local regulations.

31. C. Staff Consider adding clarifying language, as Locating a commercial nuclear power station adjacent to lands Regulatory shown with redline insertion. devoted to public use when present might be unacceptable to Guidance, top of Federal, State, or local jurisdictions.

Page 43, Section 2.4.2

32. C. Staff Consider adding clarifying language, as For example, nuclear power station siting in areas uniquely Regulatory shown with redline insertion. suited for growing specialty crops, if present, may be Guidance, Page considered a type of land conversion involving unacceptable 43, Section economic dislocation.

2.4.2.4, 3rd paragraph

33. C. Staff Consider adding clarifying language, as Where applicable, sSome areas might be unsuitable for siting a Regulatory shown with redline/strikeout. nuclear power station because of public interest in reserving Guidance, Page land for future to public scenic, recreational, or cultural use.

44, Section 2.4.3

34. C. Staff Consider adding clarifying language, as Unless adequate mitigation measures can be implemented to Regulatory shown with redline/strikeout. offset impacts, tThe use of a proposed site that could Guidance, Page disproportionately affect minority or low-income communities 48, Section 2.6 should be avoided as sites for nuclear power stations.

35. C. Staff The draft guidance only references use of a Please clarify discussion to include CP and ESP, in addition to Regulatory Limited Work Authorization in support of a COL.

Guidance, combined license (COL). Per 10 CFR Section C.3, 50.10(d), limited work authorization may be Page 49 requested in support of a combined license application or construction permit (CP) application or by an applicant or holder of an early site permit (ESP).

36. C. Staff Consider adding clarifying language, as

Identify the geographic area of potential effect and time period Regulatory shown in redline/strikeout. to be considered in evaluating the cumulative impact.

Guidance, Page

Collect information on the relevant impacts of the proposed 50, Section 3.3, action within the identified geographic area of potential effect.

bulleted list

Nuclear Energy Institute 10 Affected Section Comment/Basis Recommendation

Identify other past, present, or reasonably foreseeable actions that would contribute to the cumulative impact within the area of potential effect when added to the proposed action.

Determine the cumulative impact on the resource area within the area of potential effect.

Identify plans or actions (if any) to avoid, minimize, or mitigate adverse cumulative impacts within the area of potential effect.

37. Appendix A The NRCs proposed criterion associated The NRC should revise the dose criterion for the population with very low frequency events, twice the density distance (PDD) of advanced reactors, even if this distance of 1 rem in 30 days is more than a requires the Commissions approval. We recommend that the factor of five times more conservative dose criterion for the PDD of advanced reactors be established compared to previous large LWR licensing as 5 rem in 30 days (rather than the currently proposed PDD of when calculated using best estimate 2x the distance of 1 rem in 30 days). To be consistent with the consequence methods. That is, 1 rem is current level of protection accepted by the NRC, this dose limit approximately five times lower than the should be based upon implementation of a performance-based approach to advanced reactor siting with required consequence level of protection currently accepted for analyses performed using best estimate approaches rather large LWRs using a population density than conservative approaches.

distance of 20 miles. Doubling the distance to 1 rem in 30 days for advanced reactors will further lower the dose, resulting in more than a factor of five conservatism given the decrease in dose with distance, e.g., 4.7 rem/0.36 rem = a factor of 13.

The NRC proposed dose criterion results in undue burden on the owner of an advanced reactor with population related siting considerations unnecessarily limiting allowable sites when compared to the accepted level of protection for existing plants. It is also not consistent with the NRCs Advanced Reactor Policy Statement that expects at least the same level of protection.

Nuclear Energy Institute 11 Affected Section Comment/Basis Recommendation

38. Appendix A Related to the consideration of uncertainty, Provide for simplified assessment of uncertainty as described in per dispersion calculations developed and NEI white paper, Advanced Reactor Population-Related Siting discussed in the NEI white paper, Considerations.

Advanced Reactor Population-Related Siting Considerations, the proposed 500 ppsm dose criterion of 2x the distance to 1 rem in 30 days for advanced reactors is quite conservative as compared to previous large LWR licensing when calculated using best estimate consequence methods. The conservatism based on best-estimate calculations would be more than a factor of five. It is also noted that the LWR base case would have uncertainty associated with it, such that the margin could be even more.

While consideration of uncertainty is important, the margin in the proposed population density criterion for advanced reactors as compared to historical large LWR siting should limit the need for extensive uncertainty assessment.

39. Appendix A The draft guidance would benefit from NRC should consider inclusion of examples and graphs in the inclusion of examples and graphs of various guidance.

siting scenarios (such as those provided in the SECY 20-0045 Enclosure and the referenced ORNL report) but with real numbers rather than variables to provide and promote greater clarity, understanding, and transparency such that potential siting scenarios for advanced reactors are more readily apparent to stakeholders, including members of the public.

Nuclear Energy Institute 12 Affected Section Comment/Basis Recommendation

40. Appendix A The following discussion provided by the NRC should ensure guidance is more widely applicable for NRC in the Enclosure to SECY 20-0045, future advanced reactor applications, and not just consider Furthermore, while NRC regulations do feedback from stakeholder engagement on near-term needs for restrict the siting of plants relative to a this guidance.

population center containing more than about 25,000 residents, the staffs interactions with stakeholders have not identified any near-term proposals to site reactors within a population center exceeding 25,000 residents, and no changes to the relevant regulatory criteria and associated guidance are being proposed at this time, appears to be a little short-sighted and not consistent with the clean sheet of paper approach advocated for in developing a risk-informed, performance-based regulatory framework for advanced reactors. One would expect a more far-reaching vision to be embraced as part of this regulatory framework development.

41. Appendix A The draft guidance should make it clear that Specify in Appendix A that best estimate assumptions for best estimate assumptions for weather, weather, wind direction and shielding may be used in offsite wind direction and shielding may be used in consequence analyses to inform the alternative population-offsite consequence analyses to inform the related siting considerations for advanced reactors.

alternative population-related siting considerations for advanced reactors.

42. Appendix A Information that is equally applicable across Consider organizing Appendix A to minimize repetition of the approaches described in Appendix A, information that is equally applicable/available to each e.g., the limiting population-related siting approach.

consideration may be population center distance instead of population density

Nuclear Energy Institute 13 Affected Section Comment/Basis Recommendation distance, could be consolidated into one section to minimize repetition and clarify the common performance-based foundation for all approaches.

43. Section A-1, The language in 10 CFR 100.21(h) sets a Ensure the language of 10 CFR 100.21(h) is captured Background, flexible regulatory requirement (i.e., accurately and continue the use of flexible language, e.g.,

Section A-3.1, Reactor sites should be located away should, when providing guidance throughout the RG, Section A-3.2 from very densely populated centers. Areas including Appendix A, if 500 ppsm is indeed not meant to be a of low population density are, generally, regulatory limit.

preferred.) rather than a strict regulatory limit. (emphasis added)

The guidance in Appendix A seems to impose the 500 ppsm guideline as a limit rather than as guidance for meeting a flexible regulatory requirement.

B. Discussion, Population Considerations, on Page 6 does state, The numerical values in this guide are generally consistent with past NRC practice and reflect consideration of severe accidents for large LWRs, as well as the demographic and geographic conditions of the United States.

Additionally, Sec. 1.4.3 continues the flexible language of preferably and should when introducing the 500 persons per square mile.

However, the approach for advanced reactors provided in the draft guidance of Appendix A does not consistently use flexible language. See for example, A-3.2,

Nuclear Energy Institute 14 Affected Section Comment/Basis Recommendation 1st paragraph, for which population density is not to exceed 500 ppsm.

See NRC precedent in Florida Power &

Light Co. (Turkey Point Units 6 and 7),

Commission Memorandum and Order CLI-18-01, Docket Nos. 52-040-COL & 52-041-COL (Apr. 5, 2018), at 29-32 (ML18095A117), including The Staff noted that the highest population density at any radial distance out to 20 miles for the Turkey Point site is comparable to that of previously licensed sites. Id. For example, the Limerick site had a density of 789 persons per square mile at 5 miles; the Zion site had a density of 668 persons per square mile at 10 miles; and the Connecticut Yankee site had a density of 716 persons per square mile at 20 miles. Id.

(citing Metropolitan Siting - A Historical Perspective, NUREG-0478 (Oct. 1978),

tbl.I (ML12187A192)).

44. Section A-3, The introduction to the bulleted list states Proposed new bullet:

page A-3, the NRC staff provides the following list of Perform an atmospheric dispersion analysis (i.e., dose bulleted list key actions for siting analyses that should assessment), based on mean dose.

be considered regardless of the approach taken:

In the NRC public meeting October 27, 2023, slide 7 for Non-LWR under LMP indicated that risk metrics under LMP are based on mean frequency and mean consequence, with decision making informed by other factors such as

Nuclear Energy Institute 15 Affected Section Comment/Basis Recommendation uncertainty. In public discussion, the NRC confirmed that while EAB and LPZ dose calculations are performed with conservative inputs for atmospheric dispersion analysis, the population density (i.e., 500 ppsm) atmospheric dispersion calculations may be performed in a best estimate manner consistent with LMP. The public noted that the same best estimate dispersion analysis should seem to be applicable to the other two approaches of LWR under traditional approach and the non-LWR under traditional approach.

One clear way to indicate this is to include an additional bullet in this list regarding the atmospheric dispersion analysis.

45. Section A-3.1, Section A-3.1 Regulatory Guide 1.233 Proposed additional sentence:

page A-4, 2nd Approach (non-LWRs) second paragraph The atmospheric dispersion analysis (i.e., dose assessment) paragraph includes the following: may be based on best estimate inputs and assumptions, mean The calculation of offsite doses should be dose and consideration of uncertainties.

Section A-3.2, page in accordance with NRC-accepted A-5, 2nd paragraph methodologies, including associated Include a footnote with some example best estimate inputs /

computer models for the plant response to assumptions, such as:

Section A-3.3, page an accident, the performance of various Examples of best estimate inputs and assumptions associated A-6, 2nd paragraph barriers to the release of radioactive with the atmospheric dispersion modeling include average materials, and the atmospheric dispersion weather conditions, average dose of varied wind direction, of any released materials to the area credit for normal shielding (e.g., public in their homes at night.)

surrounding the plant.

In the NRC public meeting October 27, 2023, slide 7 for Non-LWR under LMP indicated that risk metrics under LMP are based on mean frequency and mean

Nuclear Energy Institute 16 Affected Section Comment/Basis Recommendation consequence, with decision making informed by other factors such as uncertainty. In public discussion, the NRC confirmed that while EAB and LPZ dose calculations are performed with conservative inputs for atmospheric dispersion analysis, the population density (i.e., 500 ppsm) atmospheric dispersion calculations may be performed in a best estimate manner consistent with LMP. This should be clarified.

Further, the public noted that the same best estimate dispersion analysis should be applicable to the other two approaches of LWR under traditional approach and the non-LWR under traditional approach. Use of best estimate / mean dispersion analysis for all three approaches (i.e., A-3.1, A-3.2, A-3.3) makes sense in view of an overall performance-based approach and, as demonstrated in the NEI white paper on Advance Reactor Population-Related Siting Considerations, reduces undue conservatism as compared to previous large LWR siting criterion.

One clear way to indicate this is to include additional text in the corresponding paragraph for each section regarding the acceptability of mean/best estimate atmospheric dispersion analysis.

46. A-3.1, Regulatory The last sentence of this section states, Although they are their own license class, the precedent set by Guide 1.233 "However, an advanced reactor with research and test reactor siting should be considered for

Nuclear Energy Institute 17 Affected Section Comment/Basis Recommendation Approach (non- estimated doses below 1 rem at the site consistency, i.e., the safety profile of an advanced reactor may LWRs), Page A-4 boundary over the month following the support siting within towns with a population greater than to A-5 assumed postulated accident could be sited approximately 25,000 residents.

within towns with populations of no more than approximately 25,000 residents." This is potentially a very conservative interpretation of the regulations that may counteract efforts to provide technology-inclusive, risk-informed, and performance-based guidance. The safety profiles of some advanced reactors, e.g., micro-reactors, will be analogous to research and test reactors. Further, there may be defense-in-defense features associated with advanced reactors that outweigh the benefits of establishing additional defense-in-depth via locating reactors away from population centers and/or there may be performance characteristics that could demonstrate safety in even closer proximity to population centers.

47. A-3.2, Regulatory The last sentence of this section states, Although they are their own license class, the precedent set by Guide 1.183 Plus "However, an advanced reactor with research and test reactor siting should be considered for Severe Accidents estimated doses below 1 rem at the site consistency, i.e., the safety profile of an advanced reactor may Approach boundary over the month following the support siting within towns with a population greater than (LWRs), Page A- assumed postulated accident could be sited approximately 25,000 residents.

5 within towns with populations of no more than approximately 25,000 residents." This is potentially a very conservative interpretation of the regulations that may counteract efforts to provide technology-inclusive, risk-informed, and performance-based guidance. The safety profiles of some advanced reactors, e.g., micro-

Nuclear Energy Institute 18 Affected Section Comment/Basis Recommendation reactors, will be analogous to research and test reactors. Further, there may be defense-in-defense features associated with advanced reactors that outweigh the benefits of establishing additional defense-in-depth via locating reactors away from population centers and/or there may be performance characteristics that could demonstrate safety in even closer proximity to population centers.

48. A-3.3, For Non- The last sentence of this section states, Although they are their own license class, the precedent set by LWRs Not Using "However, an advanced reactor with research and test reactor siting should be considered for RG 1.233 and estimated doses below 1 rem at the site consistency, i.e., the safety profile of an advanced reactor may Using Traditional boundary over the month following the support siting within towns with a population greater than Analysis of a assumed postulated accident could be sited approximately 25,000 residents.

Containment- within towns with populations of no more Type Barrier, than approximately 25,000 residents." This Page A-6, 2nd is potentially a very conservative paragraph interpretation of the regulations that may counteract efforts to provide technology-inclusive, risk-informed, and performance-based guidance. The safety profiles of some advanced reactors, e.g., micro-reactors, will be analogous to research and test reactors. Further, there may be defense-in-defense features associated with advanced reactors that outweigh the benefits of establishing additional defense-in-depth via locating reactors away from population centers and/or there may be performance characteristics that could demonstrate safety in even closer proximity to population centers.

Nuclear Energy Institute 19 Affected Section Comment/Basis Recommendation

49. A-3.3, (c) The last sentence of this section states, Provide clarity on any expectation to follow specific Consideration of However, use of a hazard assessment methodologies for hazard assessment.

Uncertainty, coupled with provisions of defense in depth Page A-8, last (multiple means of satisfying a safety paragraph function to reduce or eliminate the likelihood of phenomena) can satisfactorily address phenomenological uncertainties.

Is the term hazard assessment used generally or is a specific methodology expected to be followed, e.g., seismic hazard assessment has a specific meaning.

WHITE PAPER NEI White Paper: Advanced Reactor Population-Related Siting Considerations Prepared by the Nuclear Energy Institute November 2023

November 2023 Acknowledgements This document was prepared for the Nuclear Energy Institute by JENSEN HUGHES. NEI acknowledges and appreciates the contributions of NEI members providing input, reviewing and commenting on the document, particularly the NEI Advanced Reactor Regulatory Task Force and NEI Siting Task Force.

NEI Project Lead: Kati Austgen Notice Neither NEI, nor any of its employees, members, supporting organizations, contractors, or consultants make any warranty, expressed or implied, or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately owned rights.

November 2023 Executive Summary The approach approved in SRM-SECY-20-0045, Population Related Siting Considerations for Advanced Reactors, was created to support a more performance-based approach to the policy on siting away from population centers and increase the number of allowable sites for advanced reactors in comparison to current guidance. To understand the consistency of the levels of public protection afforded by the siting of the current fleet of light-water reactors and the NRC proposed performance-based approach for advanced reactors, the industry has conducted scoping analyses to assess the guidance under development in NRCs pre-decisional white paper, Alternative Approaches to Address Population-Related Siting Considerations, made public in April 2023, and subsequently incorporated in draft Regulatory Guide DG-4034 (Proposed Revision 4 to Regulatory Guide 4.7, General Site Suitability Criteria for Nuclear Power Stations), then discussed in a public meeting October 27, 2023.

The conclusion of these analyses is that the NRC population density distance (PDD) dose criterion for advanced reactors of twice the distance to 1 rem in 30 days is excessively conservative, resulting in undue burden on the owner of an advanced reactor with population related siting considerations unnecessarily limiting allowable sites when compared to the accepted level of protection for existing plants. Therefore, it is recommended that a PDD dose criterion for advanced reactors of 5 rem in 30 days be used to achieve a similar level of protection to that which is already accepted by the NRC.

Furthermore, this dose criterion is based upon implementation of a performance-based approach to advanced reactor siting where required consequence analyses can be performed using best estimate approaches rather than conservative approaches. This white paper provides a description of the scoping analyses performed along with related observations, conclusions, and recommendations.

The NRCs current guidance for the population siting criterion with respect to population density (the focus of this white paper) prescribes that the average density does not exceed 500 persons per square mile for any radial distance out to 20 miles. The NRC has long held that this criterion provides an acceptable level of protection for the public.

The NRCs Policy Statement on the Regulation of Advanced Reactors [23] states that:

Regarding advanced reactors, the Commission expects, as a minimum, at least the same degree of protection of the environment and public health and safety and the common defense and security that is required for current generation light-water reactors (LWRs). Furthermore, the Commission expects that advanced reactors will provide enhanced margins of safety and/or use simplified, inherent, passive, or other innovative means to accomplish their safety and security functions.

The policy statement also notes attributes that should be considered in advanced designs include, designs that incorporate the defense-in-depth philosophy by maintaining multiple barriers against radiation release, and by reducing the potential for, and consequences of, severe accidents. It is important to note that the policy statement did not require advanced reactors to meet higher safety standards, and the NRC has been implementing the policy statement in ways that align requirements with advanced reactor safety profiles. Given these expectations and the complementary functional nature of NRCs various siting criteria, it is unclear whether the NRC alternative approaches align fully with industry and NRC efforts to implement performance-based criteria for siting that better account for the safety profiles of advanced reactors. Indeed, the NRC staff recently noted in SECY 20-0045, Population Related Siting Considerations for Advanced Reactors [2] that its efforts to address

November 2023 population-related siting considerations are an important part of the integrated approach to help inform the design and siting processes and the related content of applications for licenses, certifications, and approvals for advanced reactors. It further acknowledged that the Nuclear Energy Innovation and Modernization Act of 2019 (NEIMA) requires the NRC to develop and implement, where appropriate, strategies for the increased use of risk-informed, performance-based techniques to resolve policy issues such as siting considerations that may unnecessarily restrict the development of commercial advanced nuclear reactors.

The purpose of this paper is to 1) put the NRCs population-density siting considerations in context with other siting elements and defense-in-depth considerations, 2) compare the level of protection afforded by the NRC proposed siting criteria for advanced reactors to that provided by the guidance currently applied to existing large LWRs, and 3) evaluate the effects of the NRC proposed approach on the ability to flexibly and maximally site advanced reactors with similar levels of public protection as todays reactors. In doing so, we present information on the historical origins and evolution of the NRCs current reactor siting and population density criteria. We conclude that the level of protection established by the NRCs pre-decisional white paper approaches would impose undue burden and excessive restrictions compared to what the NRC currently finds acceptable for large LWRs. Therefore, we propose an alternative dose criterion for defining the population density distance that appropriately credits features of advanced reactors (e.g., smaller source terms, passive safety features, enhanced release barriers).

November 2023 Table of Contents Introduction and Background .......................................................................................................... 1 Purpose ............................................................................................................................................ 2 Overview of NRCs Current Population Siting Criteria and Guidance .............................................. 3 Reactor Siting Distances................................................................................................................... 5 Historical Context - Origins and Evolution of the NRCs Population Siting Criteria ........................ 7 Population Siting Purposes ............................................................................................................ 11 Population Density Distance Dose Historical Effectiveness........................................................... 14 7.1 Large LWR Model Inputs ................................................................................................... 14 7.2 Large LWR Model Results ................................................................................................. 15 Advanced Reactors Population Density Distance Dose Estimates ................................................ 19 8.1 HTGR Model and Results .................................................................................................. 19 8.2 Molten Salt Reactor Model and Results ........................................................................... 21 8.3 WinMACCS Model Summary Insights ............................................................................... 23 Defense-In-Depth Considerations.................................................................................................. 23 Alternate PDD Dose Criterion for Discussion................................................................................. 25 Conclusions .................................................................................................................................... 25 References ..................................................................................................................................... 26 Table of Figures Figure 1: Reactor Siting Population Limitations (SECY-20-0045 Enclosure) ................................................. 7 Figure 2 - Large LWR Base Case 30-Day Dose vs. Distance ......................................................................... 17 Table of Tables Table 1a - Large LWR Siting Criteria Protective Function ........................................................................... 11 Table 1b - EPZ Relationship to Siting Criteria Protective Function ............................................................. 13 Table 2 - Large LWR Release 30-Day Dose .................................................................................................. 17 Table 3 - HTGR Release 30-Day Dose .......................................................................................................... 20 Table 4: MSR Release 30-Day Dose............................................................................................................. 22 Table 5 - Defense-In-Depth Considerations for Siting Advanced Reactors ................................................ 24

November 2023 INTRODUCTION AND BACKGROUND In Title 10 of the Code of Federal Regulations (CFR) Part 100, Section 100.21, the NRC requires that:

(h) Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable3.

3 Examples of these factors include, but are not limited to, such factors as the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better rail and highway access, shorter transmission line requirements, or less environmental impact on undeveloped areas, wetlands or endangered species, etc. Some of these factors are included in, or impact, the other criteria included in this section.

For implementation of 10 CFR 100.21(h), Regulatory Guide (RG) 4.7, General Site Suitability Criteria for Nuclear Power Stations, [1] contains current NRC guidance for assessing the population around potential reactor sites using the criterion of the population density not exceeding 500 persons per square mile (ppsm) out to 20 miles. The NRCs guidance and experience for siting nuclear power plants relate to large LWRs. The population-related siting considerations for large LWRs are based on a hypothetical accident and fission product release with the following typical attributes:

A reactor core inventory associated with a large reactor power level (e.g., 3,000 MWth)

A substantial meltdown of the core (e.g., 100% noble gases, 50% iodine)

Subsequent release into the containment of appreciable quantities of fission products, and

Consideration, via defense-in-depth, of the eventual release of appreciable quantities of fission products into the environment (e.g., sufficient to impact populations twice the distance of the customary 10-mile plume exposure pathway for the 500 ppsm to 20 miles criterion)

Compared to previous generations of reactor designs, advanced reactor designs are expected to have reduced likelihoods of accidents, reduced fission product releases, and potentially slower releases of radioactive material in the unlikely event of an accident. This is discussed in detail in the Enclosure to SECY-20-0045, Population Related Siting Considerations for Advanced Reactors [2]. In addition to describing the relevant NRC regulatory requirements, guidance, and history, the Enclosure sets forth four options for the Commissions consideration to provide an alternative to the current guidance in RG 4.7 related to 10 CFR 100.21(h).

NEI provided input to the NRC staff as they considered the four options in SECY-20-0045, and supported the staffs recommended Option 3, under which the NRC would revise the guidance in RG 4.7 relating to 10 CFR 100.21(h) to include provisions for advanced reactor designs. According to the NRC staff, Option 3 allows consideration of the design and site-specific accident consequences and specific features of an advanced reactor design (e.g., fuel designs, inherent safety features, and other contributors to the retention of radionuclides) that may limit the release of radionuclides. This approach is intended to

November 2023 support a more performance-based approach to the policy on siting away from population centers to increase the number of allowable sites for advanced reactors in comparison to current guidance while still controlling societal risks. Under the new Option 3 graded approach, the population density would be assessed against a criterion in the revised guidance that the population density is expected to be no greater than 500 ppsm out to twice the distance at which the 1 rem dose was calculated based on design-specific events (i.e., design basis or beyond-design-basis events).

In SRM-SECY-20-0045 [3], the Commission approved the staffs recommended Option 3 to provide technology-inclusive, risk-informed, and performance-based criteria to assess certain population-related issues in siting advanced reactors. Regarding the traditional dose assessment approach described by the NRC staff in the Enclosure to SECY-20-0045, 1 the Commission instructed the staff to provide appropriate guidance on assessing defense-in-depth adequacy and establishing hypothetical major accidents to evaluate. Notably, in his Response Sheet, Commissioner Wright emphasized that the development of population-related siting criteria for advanced reactors is critical to establishing a timely, efficient, and predictable regulatory framework for such technologies. He further underscored the flexibilities inherent in the Option 3 approach:

This approach is consistent with the Commissions long-standing recognition that improvements in reactor design may potentially affect siting decisions. This technology-inclusive, risk-informed approach moves away from deterministic siting criteria, consistent with the NRCs approach to establish emergency preparedness requirements for small modular reactors. This approach would also leverage existing requirements, operating experience, analyses, and research and allow for flexibilities based on design-specific characteristics of advanced reactor technology, including inherent safety features and lower power levels that may limit the release of radionuclides. The staffs recommended approach should effectively meet the NRCs important safety and security mission in a manner that does not inhibit the development and licensing of new technologies. [emphasis added]

Chairman Hanson similarly noted that for advanced reactors demonstrating enhanced safety attributes, it is reasonable to have a regulatory pathway that gives applicants the flexibility to justify sites closer to population centers compared to historical siting of large light water reactors. He also emphasized the need for regulatory clarity on this issue to ensure effective and predictable licensing of advanced reactors.

PURPOSE The purpose of this paper is to assess the level of public protection provided by the NRC proposed dose criterion of twice the distance of 1 rem for 30 days, and to evaluate various modeling parameters in order to make recommendations to the NRCs draft guide DG-4034 that will result in NRC dose criterion and guidance that provide a similar level of public protection as the current 20 miles for large LWRs.

In consideration of SRM-SECY-20-0045 [3], NEI performed scoping consequence analyses to compare the level of protection afforded by the NRC proposed siting criteria for advanced reactors to that provided by the guidance currently applied to existing large LWRs and evaluated the effects of the NRC proposed 1 The staff suggested that the size of the area within which the population density would be assessed could be determined using the risk-informed methodology described in DG-1353 (subsequently issued as Regulatory Guide 1.233), or using traditional, deterministic practices with a hypothetical major accident and guidance such as Regulatory Guide 1.183, but that [t]hese methods will necessitate a thorough, upfront analysis of design and site characteristics.

November 2023 approach on the ability to flexibly and maximally site advanced reactors with similar levels of public protection as todays reactors. NEI presented an overview of these analyses at a public meeting [16] on July 20, 2023. NEI found that the proposed dose criterion is orders of magnitude less than what has been accepted for existing large LWRs assuming conservative modeling assumptions as employed for exclusion area boundary (EAB) and low population zone (LPZ) siting consequence analyses. As a result, the proposed criterion would be excessively restrictive for siting for advanced reactors.

After issuing Draft Regulatory Guide DG-4034 [11] for public comment, the NRC conducted a public meeting and presented slides [12] clarifying that for those reactor designs that pursue the Licensing Modernization Project (LMP) approach (e.g., NEI 18-04) mean consequence results would be acceptable provided that uncertainties and cliff-edge effects were assessed. In public discussion, the NRC specifically stated that the conservative consequence analysis assumptions used for the EAB and LPZ dose calculations were not required to be applied to the population density distance calculation.

The public noted that similar language was not presented in the slides for light-water SMRs using a traditional approach (e.g., RG 1.183), non-LWRs using a traditional approach, or in DG-4034. The NRC appeared to indicate that mean consequence dose assessment could be acceptable for light-water SMRs or non-LWRs using a traditional approach with justification, e.g., describing how uncertainties were assessed.

Based on NRC public interactions to date, this NEI white paper presents population-related considerations that would apply to advanced reactor siting with a focus on the 500 ppsm dose criterion, but within the broader context of the various siting criteria. It also presents historical perspectives, recent WinMACCS scoping modeling results, and defense-in-depth elements to inform the siting considerations.

OVERVIEW OF NRCS CURRENT POPULATION SITING CRITERIA AND GUIDANCE The NRCs siting regulations are codified in 10 CFR Parts 50, 52, and 100. Section 100.21, Non-Seismic Siting Criteria, requires that the site have an exclusion area surrounding the reactor in which there are no permanent residents and the reactor licensee has the authority to determine all activities, including exclusion or removal of personnel and property from that area. Section 100.21 also requires an LPZ immediately surrounding the exclusion area. The number and density of residents within the LPZ must be such that there is a reasonable probability that protective measures could be taken on their behalf in the event of a serious accident. Specific distances to the outer boundary of the exclusion area (i.e., the EAB), and to the outer boundary of the LPZ are not prescribed by regulation. Rather, the boundaries of the exclusion area and LPZ are set by dose limits of 25 rem total effective dose equivalent (TEDE) over the most limiting 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the exclusion area and over the entire passage of the radioactive cloud for the LPZ, respectively. Consequently, the distances to the EAB and LPZ may vary from reactor site to reactor site.

10 CFR 100.21 also requires that the consequences of postulated accidents meet the radiological dose criteria in 10 CFR 50.34(a)(1), considering the sites atmospheric dispersion characteristics. Specifically, 10 CFR 50.34(a)(1) requires a description and safety assessment of the site and a safety analysis of the facility, including analysis of the engineered safety features and barriers that must be breached because of an accident that can then result in a release of radioactive material to the environment, with special

November 2023 attention to plant design features that are intended to mitigate the radiological consequences of accidents. 10 CFR 50.34(a)(1)(ii)(D) states that the safety assessment by the applicant:

shall assume a fission product release6 from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable site characteristics, including site meteorology, to evaluate the offsite radiological consequences. Site characteristics must comply with part 100 of this chapter.

Footnote 6 of this regulation describes the assumed release to containment for this analysis as follows:

6 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.

Safety analysis requirements in 10 CFR 52.17(a)(1)(ix), 10 CFR 52.47(a)(2), and 10 CFR 52.79(a)(1)(iv), for new reactor applications for an early site permit, standard design certification, and combined operating license, respectively, describe the same type of analysis and use the same evaluation criteria of 25 rem TEDE at the EAB and LPZ as set forth in 10 CFR 50.34(a)(1). 2 As a siting criterion, 10 CFR 100.21(b) requires that the distance to the nearest population center of more than about 25,000 people - the population center distance (PCD) - is at least one and one-third times the distance to the outer boundary of the LPZ. For operating reactors, EABs and LPZ boundary distances may vary from site to site and reactor to reactor, because the distance determination depends in part on the need to meet the siting dose acceptance criteria at the chosen distance. Generally, large LWR EABs are about 0.5 miles, and LPZs are about 3 to 5 miles. The 10 CFR 100.21(b) PCD requirements result in the closest population center to the reactors lying outside of the LPZ; however, the population center could be inside the 10-mile plume exposure emergency planning zone (EPZ) used for large LWRs.

Reflecting longstanding NRC policy for large LWRs, 10 CFR 100.21(h) provides that:

Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining the acceptability of a particular site located away from a very densely populated center but not in an area of low 2 For design basis accidents (DBAs) other than loss of coolant accident (LOCA) (or bounding scenario as applicable to small modular reactor (SMR) and non-LWR designs), NRC regulatory guidance provides that the dose acceptance criteria are fractions of the 25 rem TEDE criterion, either 25 percent (6.5 rem TEDE) or 10 percent (2.5 rem TEDE). These dose acceptance criteria are to be generally commensurate with the higher likelihood of the accident scenario, such that the more likely scenarios (e.g., rupture of one steam generator tube with the coolant at equilibrium activity concentration) are compared to a lower dose criterion (2.5 rem TEDE). For the set of DBAs evaluated for offsite consequences, doses are also calculated within the control room to show compliance with the 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19 control room radiological habitability dose criterion of 5 rem TEDE.

November 2023 density, consideration will be given to safety, environmental, economic, or other factors, which may result in the site being found acceptable. 3 RG 4.7 [1] provides guidance for assessing the population around potential reactor sites in accordance with 10 CFR 100.21(h). RG 4.7 adds a population density criterion to the foregoing regulatory requirements, as follows:

A reactor should be located so that, at the time of initial plant approval within about 5 years thereafter, the population density, including weighted transient population, averaged over any radial distance out to 20 mi (cumulative population at a distance divided by the circular area at that distance), does not exceed 500 persons per square mile. A reactor should not be located at a site where the population density is well in excess of this value. [emphasis added]

RG 4.7 explains that [l]ocating reactors away from densely populated centers is part of the NRCs defense-in-depth philosophy and facilitates emergency planning and preparedness, as well as reduces potential doses and property damage in the event of a severe accident. It further notes that the numerical values in the guide are generally consistent with past NRC practice and reflect consideration of severe accidents, as well as the demographic and geographic conditions characteristic of the United States.

REACTOR SITING DISTANCES As the foregoing reflects, power reactor siting has typically involved assessment of certain key distances which are depicted in Figure 1 (taken from the Enclosure to SECY-20-0045 [2]), including:

Exclusion Area Boundary (EAB) - The area surrounding the reactor in which the licensee has the authority to determine all activities including exclusion or removal of personnel and property.

The EAB distance is determined such that an individual located at any point on its boundary for any 2-hour period following the onset of the postulated release would not receive a radiation dose (TEDE) in excess of 25 rem. (10 CFR 100.21(a) / 100.3 / 50.34 / 52.17 / 52.79). The EAB is sometimes referred to as the Owner Controlled Area.

Low Population Zone (LPZ) - The area immediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken on their behalf in the event of a serious accident. The LPZ distance is determined such that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose (TEDE) in excess of 25 rem. (10 CFR 100.21(a) / 100.3 / 50.34 / 52.17 / 52.79)

Population Center Distance (PCD) - The distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents. The boundary of the population center shall be located at least one and one third times the distance from the reactor to the outer boundary of the LPZ. (10 CFR 100.21(b) / 100.3) 3 Footnote 3 to 10 CFR 100.21(h) lists examples of such factors. They include, but are not limited to, the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better rail and highway access, shorter transmission line requirements, or less environmental impact on undeveloped areas, wetlands or endangered species, etc.

November 2023

Population Density Distance (PDD) - The PDD is determined at the time of initial plant approval so that, at that time and for about five years thereafter, the population density, including weighted transient population, averaged over any radial distance out to 20 miles (cumulative population at a distance divided by the circular area at that distance), does not exceed 500 persons per square mile. (10 CFR 100.21(h), RG 4.7 [1])

Emergency Planning Zone (EPZ) - While not a siting decision element for a reactor plant (and therefore not depicted on Figure 1), the EPZ relates with various siting distances and is therefore included for context and discussion. The EPZ is determined in relation to local emergency response needs and capabilities; per NUREG-0396 [13], the EPZ size for the current fleet of LWRs was determined using dose-distance curves to arrive at a radial distance of about 10 miles for the plume exposure pathway. As discussed in NUREG-0396, an EPZ boundary distance of about 10 miles is adequate to provide for a substantial reduction in early health effects (injuries or deaths) in the event of a severe reactor accident. A dose of 200 rem whole body was used as the dose at which significant early health effects begin to occur. The report goes on to say that planners should use judgement to determine the precise size and shape of an EPZ by considering local conditions such as demography, topography and land use characteristics, access routes, jurisdictional boundaries, and arrangements with the nuclear facility operator for notification and response assistance. Generally, the plume exposure pathway EPZ for the current fleet of LWRs with an authorized power level greater than 250 MWth consists of an area about 10 miles (16 km) in radius and the ingestion pathway EPZ consists of an area about 50 miles (80 km) in radius. The size of the EPZs may be determined on a case-by-case basis for reactors with an authorized power level less than 250 MWth. (10 CFR 50, Appendix E)

In the forthcoming final rule, Emergency Preparedness for Small Modular Reactors and Other New Technologies, the NRCs final alternative framework for SMRs and ONTs consists of two major elements - an EPZ size determination process and a set of performance-based requirements. The size of an EPZ determined by this process is scalable based on factors such as accident source term, fission product release, and associated dose characteristics, and the same process can be applied to all SMR and ONT designs. Once published, the final rule 10 CFR 50.33, Contents of applications; general information, is expected to state, The plume exposure pathway EPZ is the area within which: (A) Public dose, as defined in § 20.1003 of this chapter, is projected to exceed 10 mSv (1 rem) total effective dose equivalent over 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from the release of radioactive materials from the facility considering accident likelihood and source term, timing of the accident sequence, and meteorology. Further, the performance-based requirements in § 50.160, Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities, do not contain any technology specific language.

November 2023 Figure 1: Reactor Siting Population Limitations (SECY-20-0045 Enclosure)

HISTORICAL CONTEXT - ORIGINS AND EVOLUTION OF THE NRCS POPULATION SITING CRITERIA NUREG-0625 (1979) [4] discusses the historical development and implementation of 10 CFR 100 and RG 4.7 and provides some valuable technical insights that are relevant to advanced reactor siting. Prior to Part 100, the Atomic Energy Commissions (AEC) general policy regarding power plant siting was to provide both site isolation and plant design (primarily containment) as elements of defense-in-depth to assure no undue hazard to the health and safety of the public. The concept of a maximum credible accident (MCA) was developed during this time to test whether the degree of site isolation and plant design was sufficient. As NUREG-0625 further explains:

The maximum credible accident concept was carried into Part 100 In Part 100, the maximum credible accident is defined as 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. [10 CFR 100.11(a)

November 2023 footnote 1]. Although more severe accidents (now generally referred to as Class 9 accidents 4) are conceivable, the consequences of such accidents were normally not analyzed for assessing the suitability of a proposed site and plant design.

When Part 100 was first developed, the MCA was assumed to be a loss-of-coolant accident (LOCA) that would result in a substantial core meltdown with subsequent release of appreciable quantities of fission products into containment. Given the relatively small size of reactors at the time, it was believed that this postulated core meltdown could be accommodated without a loss of containment integrity, thus providing an effective upper bound on offsite radiological consequences. As power reactors increased in size by roughly an order of magnitude, they were required to have emergency core cooling systems (ECCS) in accordance with Appendix K to 10 CFR Part 50. While effective ECCS operation reduced the potential for a substantial meltdown, concerns about containment failure due to steam or hydrogen explosions and containment overpressure failures remained. As increasing credit was given to engineered safety features, plants began to be located closer to population centers. This trend prompted a concern that [a]n unbounded reduction of the distance factor as a tradeoff for added safety features can lead to an erosion of the protection provided by distance that was originally contemplated in Part 100.

At the time NUREG-0625 was authored (1979), the primary objectives of the various siting-related distances discussed herein as perceived by the staff were noted as follows:

EAB - Control of land use close to the plant. Protection of the public in the event of an accident, and protection of the plant from man-made sources.

LPZ - A zone where evacuation was feasible. A buffer zone between the EAB and large population concentrations to control or minimize societal consequences given an accident.

PCD & PDD - No objectives attributed to the NRC staff were identified. The Task Force members (authors of NUREG-0625) noted that their rationale for recommending population density and distribution limits both within and beyond the EPZ is to provide some additional assurance that the societal risk from Class 9 [beyond design basis accidents (BDBAs)] for populations within about 20 miles of a nuclear plant is kept at reasonable levels.

To summarize, the discussion in NUREG-0625 provides the following key technical insights:

The concept of MCA was intertwined with the development of the original 10 CFR 100 siting criteria, and the threshold for credible was design basis accidents (DBAs). The MCA was a release to containment, with containment function credited for establishing the EAB.

Containment integrity and siting distance were defense-in-depth elements for BDBAs.

Containment served as defense-in-depth protection for failure of engineered safety systems to prevent the MCA. Siting distance served as defense-in-depth protection for failure of containment.

The potential for BDBAs to impair containment function resulting in a release to the environment was a growing concern as reactor size increased. The independence of the 4 Class 9 accidents are referenced in NUREG-0625 as accidents beyond the design basis.

November 2023 containment function as a defense-in-depth element for significant accidents was considered diminished for large LWRs.

The credit for additional ECCS functions resulting in decreased EAB and LPZ distances was a growing concern. The independence of siting distance as a defense-in-depth element for significant accidents was considered diminished for large LWRs.

Given the foregoing historical developments, and as discussed in Oak Ridge National Laboratorys 2019 Advanced Reactor Siting Policy Considerations report [17], the distance and population density values in RG 4.7 thus also evolved over time. (But, as noted in the 2019 ORNL report, the original genesis of the values is based on containment as the only engineered safety feature and very conservative estimates of radioactivity release following a LWR LOCA.) In 1974, the AEC drafted guidance for nuclear power plant site suitability that included population guidance. The initial RG 4.7 guidance summarized previous thoughts on siting but lacked definitive population density numbers (ppsm) as suggested by WASH-1308 [18]. However, the 1974 version of RG 4.7 (Rev. 0) draft provided a minimum exclusion area distance and a maximum required distance to the LPZ boundary:

Areas of low population density are preferred for nuclear power station sites. High population densities projected for any time during the lifetime of a station have been a source of contention during both the Regulatory staff review and the public hearing phases of the Licensing process. If the population density at a proposed site is not acceptably low, then the applicant will be required to give special attention to alternative sites with lower population densities.

Based on past experience, the Regulatory staff has found that a minimum exclusion distance of 0.4 mile, even with unfavorable design basis atmospheric dispersion characteristics, usually provides assurance that engineered safety features can be designed to bring the calculated dose from a postulated accident within the guidelines of 10 CFR Part 100. If the minimum exclusion distance is less than 0.4 mile, it may be necessary to place special conditions on station design (e.g., added engineered safety features) before the site can be considered acceptable. Also, based on past experience, the Regulatory staff has found that a distance of 3 miles to the outer boundary of the LPZ is usually adequate.

The AEC staff revised RG 4.7 a year later to include a specific evaluation of population density out to 30 miles while retaining the previous guidance. That revision (Rev. 1) allowed consideration of a population density up to 1,000 ppsm over the facility lifetime:

If the population density, including weighted transient population, projected at the time of initial operation of a nuclear power station exceeds 500 persons per square mile averaged over any radial distance out to 30 miles (cumulative population at a distance divided by the area at that distance), or the projected population density over the lifetime of the facility exceeds 1,000 persons per square mile averaged over any radial distance out to 30 miles, special attention should be given to the consideration of alternative sites with lower population densities.

In 1998, NRC revised RG 4.7 again (Rev. 2) to reflect the population density guidance that still appears in the current version of RG 4.7 (Rev. 3 issued in 2014). The guidance no longer includes a firm number for

November 2023 the minimum exclusion area distance and a minimum acceptable LPZ distance. Instead, RG 4.7 Rev. 2 states:

Preferably a reactor would be located so that, at the time of initial site approval and within about 5 years thereafter, the population density, including weighted transient population, averaged over any radial distance out to 20 miles (cumulative population at a distance divided by the circular area at that distance), does not exceed 500 persons per square mile. A reactor should not be located at a site whose population density is well in excess of the above value.

The regulatory basis for the 1998 revisions can be found principally in two documents: (1) the Statement of Considerations (SOC) for the NRCs 1996 revisions to its reactor siting criteria regulations in Part 100

[19], and a 1997 NRC-commissioned report prepared by Brookhaven National Laboratory (BNL)

(NUREG/CR-6295 [20]). The 1996 SOC states, in relevant part:

Regulatory Guide 4.7 does provide effective separation from population centers of various sizes.

The Commission has examined these guidelines with regard to the Safety Goal. The Safety Goal quantitative health objective in regard to latent cancer fatality states that, within a distance of ten miles (16 km) from the reactor, the risk to the population of latent cancer fatality from nuclear power plant operation, including accidents, should not exceed one-tenth of one percent of the likelihood of latent cancer fatalities from all other causes. In addition to the risks of latent cancer fatalities, the Commission has also investigated the likelihood and extent of land contamination arising from the release of long-lived radioactive species, such as cesium-137, in the event of a severe reactor accident.

The results of these analyses indicate that the latent cancer fatality quantitative health objective noted is met for current plant designs. From analysis done in support of this proposed change in regulation, the likelihood of permanent relocation of people located more than about 20 miles (32 km) from the reactor as a result of land contamination from a severe accident is very low. A revision of Regulatory Guide 4.7 which incorporated this finding that population density guidance beyond 20 miles was not needed in the evaluation of potential reactor sites was issued for comment at the time of the proposed rule. No comments were received on this aspect of the guide. [emphasis added]

The unidentified analysis mentioned in the foregoing excerpt from the 1996 SOC appears to be NUREG/CR-6295 [20]. That report documents BNLs performance of a series of probabilistic consequence assessment calculations for nuclear reactor siting, in which BNL considered insights into severe accident source terms from NUREG-1150 [21] and examined consequences in a risk-based format consistent with the quantitative health objectives (QHOs) of the NRC's Safety Goal Policy. BNL found that based on the new source terms, the prompt and latent fatality risks at all five generic sites evaluated in NUREG-1150 meet the QHOs of the NRC's Safety Goal Policy - for all combinations of site parameters and source terms - by margins ranging from one to more than three orders of magnitude.

In addition to the evaluation of site acceptability criteria, BNL conducted sensitivity studies of selected site parameters (reactor power level, emergency response, and long-term dose criteria) on appropriate consequence measures including the individual risks of prompt fatalities, permanent relocation (i.e., condemnation of non-farm property), and latent cancers for the most severe source terms. Those calculations showed that the mean value of the individual risks of prompt fatalities drops below 10-8 per year at about 3 miles; the mean value of the individual risk of latent cancer fatalities drops below 10-8

November 2023 per year (when long-term protective actions are applied) at about 15 miles; and the mean value of the individual risk of permanent relocation drops below 10-8 per year at about 20 miles from the plant.

Significantly, in the 1996 SOC, the Commission explicitly distinguished between current large LWRs and advanced or next-generation reactors with regard to siting distances and the QHOs:

In summary, next-generation reactors are expected to have risk characteristics sufficiently low that the safety of the public is reasonably assured by the reactor and plant design and operation itself, resulting in a very low likelihood of occurrence of a severe accident. Such a plant can satisfy the QHOs of the Safety Goal with a very small exclusion area distance (as low as 0.1 miles). The consequences of design basis accidents, analyzed using revised source terms and with a realistic evaluation of engineered safety features, are likely to be found acceptable at distances of 0.25 miles or less. With regard to population density beyond the exclusion area, siting a reactor closer to a densely populated city than is current NRC practice would pose a very low risk to the populace. [emphasis added]

The Commission nevertheless concluded that defense-in-depth considerations and the additional enhancement in safety to be gained by siting reactors away from densely populated centers should be maintained.

POPULATION SITING PURPOSES In view of the historical context of population siting criteria, dose criteria attributes, and past analyses (e.g., MACCS analyses for atmospheric dispersion), Table 1a presents a proposed summary of the functional purpose (i.e., protective attribute for the public) and defense-in-depth provision of each of the siting criteria currently implemented for the commercial fleet of large LWRs, along with Table 1b providing the EPZ for further context. While various siting criteria may provide overlapping protective attributes, the focus is on their generally intended protective feature(s).

Table 1a - Large LWR Siting Criteria Protective Function Siting Dose Criteria Functional Purpose Containment Defense-In-Depth Element Status Provision EAB < 25 rem for any

Protection of the plant Intact (within Minimizes early 2-hr period for from man-made sources technical radiological fatalities an individual at that might initiate a plant specification and injuries from dose the EAB leakage limits) 5 streaming associated accident with fission products

Protection from retained in containment prolonged exposure of (i.e., containment radiation from normal shine) and leakage to operation the environment if

Protection from acute and plant safety systems fail to prevent a significant potentially fatal exposure core melt.

following an accident 5 Technical specification leakage refers to the maximum allowed containment leakage associated with the primary containment. A postulated severe accident may result in a very small fractional release of radionuclides to the environment via such leakage.

November 2023 Siting Dose Criteria Functional Purpose Containment Defense-In-Depth Element Status Provision involving significant quantities of fission products associated with DBAs released into containment LPZ < 25 rem for the Protection from acute and Intact (within Minimizes early cloud passage potentially fatal exposure technical radiological fatalities if period following an accident specification containment and (implemented involving significant leakage limits), evacuation fail.

as 30 days) for quantities of fission which should an individual products associated with result in a very located at any DBAs released into limited cloud point on the containment. If released to outer boundary the environment, previous MACCS analyses have shown that individuals within 2 to 3 miles are typically most at risk for early fatalities.

PCD Extension of LPZ Protection from acute Intact (within

Minimizes early criterion (1.333 radiological injuries technical radiological injuries times the LPZ (fatalities are addressed by specification and the potential for distance) the LPZ) and societal leakage limits) evacuation to be impacts (e.g., land within LPZ, but condemnation) following an impaired with ineffective for the accident involving respect to most at-risk residents significant quantities of societal impacts (closest to the plant) fission products released to beyond LPZ. by minimizing the the environment. Land population nearest within 5 miles is most at the plant within the risk for condemnation given EPZ.

a significant release to the environment.

Minimizes the number of people whose land may be condemned due to radiological contamination.

PDD Dose criterion is Protection from societal Impaired Minimizes societal not explicit. The impacts (e.g., land impacts if containment established interdiction, fails.

distance is 20 decontamination) following miles. an accident involving significant quantities of

November 2023 Siting Dose Criteria Functional Purpose Containment Defense-In-Depth Element Status Provision fission products released to the environment. Land within 20 miles is typically most likely to require decontamination, but decontamination may be required even further (e.g.,

40 miles downwind).

Table 1b - EPZ Relationship to Siting Criteria Protective Function Siting Dose Criteria Functional Purpose Containment Defense-In-Depth Element Status Provision EPZ 6 Current large Protection from acute Impaired. EPA Minimizes early LWRs: NUREG- radiological fatalities and Protective radiological fatalities 0396 considered radiological injuries Action Guides close to the plant and EPA PAGs (< 25 following an accident (PAGs) are not radiological injuries rem over 96 involving significant expected to be further from the plant if hours) and quantities of fission triggered by containment fails early threshold doses products released to the technical in the accident.

for early severe environment. specification health effects. leakage.

The established distance is about 10 miles. 7 EP for SMR &

ONT Alternative:

< 1 rem over 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from the release of radioactive materials. 8 6 As noted previously, the EPZ was not a siting decision element for large LWRs but is included for context and discussion purposes.

7 The EPA PAGs [22] do contain dose criteria for implementation of protective actions within the EPZ, but the size of the EPZ is not specifically determined by dose criteria for each individual reactor for large LWRs. Instead, per 10 CFR 50.47(c)(2), the plume exposure pathway EPZ is specified to consist of an area about 10 miles in radius. This radius was selected as discussed in NUREG-0396 [13], Appendix I, The study concluded that the higher PAG plume exposures of 25 rem (thyroid) and 5 rem (whole body) would not be exceeded beyond 10 miles for any site analyzed. Even under the most restrictive PAG plume exposure values of 5 rem to the thyroid and 1 rem whole body, over 70 percent of the plants would not require any consideration of emergency responses beyond 10 miles.

8 In the forthcoming final rule, Emergency Preparedness for Small Modular Reactors and Other New Technologies, the NRCs final alternative framework for SMRs and ONTs consists of two major elements - an EPZ size determination process and a set of performance-based requirements. The size of an EPZ determined by this process is scalable based on factors such as accident source term, fission product release, and associated dose characteristics, and the same process can be applied to all SMR and ONT designs. Once published, the final rule 10 CFR 50.33, Contents of applications; general information, is expected to state, The plume exposure pathway EPZ is the area within which: (A)

Public dose, as defined in § 20.1003 of this chapter, is projected to exceed 10 mSv (1 rem) total effective dose equivalent over 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months from the release of radioactive materials from the facility considering accident likelihood and source term, timing of the accident sequence, and meteorology. Further, the performance-based requirements in § 50.160, Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities, do not contain any technology specific language.

November 2023 POPULATION DENSITY DISTANCE DOSE HISTORICAL EFFECTIVENESS The NRC staffs proposed Option 3 (approved by the Commission in SRM-SECY-20-0045) for PDD for advanced reactors is 500 ppsm out to twice the distance at which the 1 rem dose in 30 days is calculated from postulated accidents. This differs from the previous more generically prescriptive requirement for the existing fleet of large LWRs of 500 ppsm out to 20 miles. It should be noted that the distance x 2 approach proposed for advanced reactors will lead to a received dose less than 1 rem for any individual beyond the 500 ppsm boundary due to the decrease of dose with distance. As scoping analyses for advanced reactors later show, the dose received at the 500 ppsm boundary could be 32% to 40% of the 1 rem criterion (e.g., 360 mrem in 30 days), such that the proposed dose criterion associated with an extremely low frequency event roughly equates to the average annual exposure received by individuals in U.S. from natural and man-made sources (i.e., 620 mrem/yr) 9.

Given the generically prescriptive nature of the previous large LWR-based PDD criterion, a scoping atmospheric dispersion consequence analysis was conducted using the WinMACCS code, the standard code used for probabilistic risk assessment (PRA)-related Level 3 consequence analyses. The purpose of this scoping model was to estimate dose received using the existing (large LWR based) PDD siting criterion (i.e., the level of protection actually provided by the 20-mile limit for the current fleet) so that it can be compared with the NRC proposal of twice the distance to 1 rem dose from an advanced reactor.

7.1 Large LWR Model Inputs The scoping WinMACCS model was based on the following inputs and assumptions:

The NRC Linear No-Threshold (LNT) Point Estimate Sample Problem distributed with WinMACCS version 3.10 was the starting point for the model.

Core inventory was taken from the NRCs State-of-the-Art Reactor Consequence Analysis (SOARCA) for the Surry reactor (2546 MWth) [5] and was ratioed up by 50% (effectively representing 3819 MWth) to reflect a larger reactor power level. This reactor power level is still less than other licensed LWRs but is reasonably considered to be representative of a typical large LWR (neither the lowest nor highest power level in the existing fleet).

Radionuclide release magnitudes (to the environment) were based on calculating an average of the total release fraction of the highest releases from thirteen recent Severe Accident Management Alternatives (SAMA) analyses (both PWR and BWR) using a frequency screening.

The highest release fraction for each plant was based on the Cesium (Cs) group (which historically correlates with societal impacts) for those release categories that exceeded a total release frequency (including internal events, internal flooding, fire and seismic where available or estimated from SAMA external event multipliers) of 5E-7/yr. SAMA release categories with frequencies below 5E-7/yr are estimated to be low in the frequency range of beyond design basis events (e.g., NEI 18-04 [9] defines BDBE as those in a frequency range of 1E-4/yr to 5E-7/yr) and were therefore excluded so as not to over-estimate the level of protection provided to the population by the 20-mile PDD used for the existing fleet. The calculated average total release fraction used for the model was about 32% for iodine and 34% for cesium and 9 https://www.nrc.gov/about-nrc/radiation/around-us/doses-daily-lives.html

November 2023 represents a typical large LWR release. The total release was allocated into three plumes assuming 65% of the release occurred initially (e.g., after containment failure), 20% occurred in the second plume, and 15% in the final and longest plume. Modeling multiple plumes is typical in WinMACCS analyses where an annual hourly weather data file is employed, as was used for this model. Such modeling supports meteorological changes during the postulated release (e.g., wind direction at the time of the release) and generally results in lower peak doses in any one direction since the released material is modeled to be dispersed more widely if the wind direction changes with each plume. The release fraction for each plume is shown below.

Xe/Kr I Cs Te Ba Ru / Mo La Ce Plume 1 5.71E-01 2.08E-01 2.21E-01 2.01E-01 2.23E-02 2.85E-02 1.63E-03 8.64E-03 Plume 2 1.76E-01 6.42E-02 6.79E-02 6.19E-02 6.87E-03 8.78E-03 5.03E-04 2.66E-03 Plume 3 1.32E-01 4.81E-02 5.10E-02 4.64E-02 5.15E-03 6.59E-03 3.77E-04 1.99E-03 Total 8.79E-01 3.21E-01 3.40E-01 3.10E-01 3.43E-02 4.39E-02 2.51E-03 1.33E-02

Other plume modeling aspects are shown below. Plume heat was based on the average of heat values calculated using the MAAP code as available in the SAMA documents (i.e., generic heat values assumed in many SAMA analyses were not included in the average). The release height was taken as approximately one half of the typical containment height of 60 meters (m), a recommended modeling approach as identified in NEI 05-01 [6] for SAMA analyses. The plume delay (from accident initiation to release) and duration were postulated based on past SAMA analyses, but it is noted that the delay time is irrelevant given that protective actions such as evacuation are not modeled.

Height (m) Heat Delay (hr) Duration (w) (hr)

Plume 1 30 2E+6 2 3 Plume 2 30 2E+6 5 5 Plume 3 30 2E+6 10 10

Surry meteorology data (which are distributed as example data with the WinMACCS code) were used. The meteorology data are hourly for 16 compass directions for one year. All 8,760 weather sequences were used in the model quantification to develop a statistical distribution of results based on weather variation (e.g., wind direction, wind speed, precipitation).

Dose was calculated for 30 days, with no credit for protective actions (e.g., no evacuation, no dose-based hotspot relocation).

7.2 Large LWR Model Results A number of large LWR WinMACCS scoping cases were performed to examine the potential impact of different modeling assumptions upon the results, including:

Average weather conditions (best estimate) vs 95% weather conditions (conservative)

Average dose of 16 wind directions (best estimate) vs maximum dose of 16 wind directions (conservative)

November 2023

Partial credit for structure shielding over the 30 days (e.g., sleeping indoors at night considered best estimate) versus no credit for structure shielding (e.g., open field conditions 24/7 considered conservative)

Based on the recent NRC public meeting slides [12] and public meeting discussion associated with the Draft Regulatory Guide (DG-4034) for proposed RG 4.7 Revision 4 [11], the base case is considered to be those assumptions aligned with a best-estimate dose analysis for PDD rather than the typical conservative assumptions that are employed for the EAB and LPZ dose assessments (e.g., conservative weather, maximum wind direction, open field conditions) 10. Sensitivity cases individually examine the potential impact of more conservative assumptions such as:

95% weather for dose 11

Maximum dose of the 16 wind directions

No credit for shielding over the 30 days (i.e., open field conditions)

Specifically, the large LWR WinMACCS base case dose results are based on:

Mean (average) weather

Average dose of the 16 wind directions

Normal activity shielding values (e.g., inside structures at night)

No protective actions such as evacuation or temporary relocation12 The calculated base case dose at 20 miles for a large LWR would be approximately 5 rem in 30 days as shown in Table 2. The dose result as a function of distance for the base case is presented graphically in Figure 2. A 1 rem dose would be achieved at approximately 45 miles.

10 NRC slide #7 for non-LWR reactor designs using the LMP approach specifies that risk metrics under LMP are based on mean frequency and mean consequence of event sequences, with decision-making informed by other factors such as uncertainties and assessment of cliff-edge effects. Meeting discussion clarified that PDD dose under LMP may be modeled using best-estimate type approaches with uncertainties considered rather than the conservative dose modeling approaches employed for EAB and LPZ. It is noted that the NRC slides #8 (LWR under traditional approach) and #9 (Non-LWR under traditional approach) do not specifically mention use of mean consequence results, however both mention analyses ranging from bounding assumptions and simple modeling to a high level of realism and detail. Use of mean dose results would reflect a higher level of realism for these two approaches. Evaluation of uncertainties could provide justification.

11 WinMACCS output includes representation of the dose statistical distribution based on calculations for each of the 8760 weather sequences.

The 95th percentile result represents that 95% of the weather sequences would be expected to have a dose less than the reported value.

12 The omission of protective actions such as shelter-in-place, evacuation or temporary relocation over the 30-day period is a conservatism that is retained in the base case modeling and could be further evaluated via sensitivity cases by reactor vendors. These protective actions represent broader societal impacts (e.g., displacing individuals) and are omitted to support consideration of dose absent these other societal impacts.

November 2023 Table 2 - Large LWR Release 30-Day Dose Dose @ 20 Ratio of

Case miles Sensitivity to (rem) Base Case 1 Base Case 4.7 (Mean weather, WD average, normal activity shielding) 2 Sens Case: 95% weather 26 5.5 3 Sens Case: WD maximum 7.0 1.5 4 Sens Case: No shielding 12 2.5 5 Sens Case: 1 year exposure 10 2.1 Figure 2 - Large LWR Base Case 30-Day Dose vs. Distance As can be seen from the base case results, the proposed PDD approach for advanced reactors is estimated to be greater than a factor of five times more stringent than that of existing large LWRs given that the proposed criterion is 2x the distance to 1 rem. If the criterion was the distance to 1 rem (rather than 2x the distance), the proposed criterion would be approximately a factor of five more stringent (4.7 rem/1 rem = 4.7). Given that the dose decreases with distance, the distance multiple of two will result in additional margin, e.g., if the 30-day dose at 2x the distance to 1 rem is approximately 360 mrem (due to the decrease of dose with distance), then the comparison is 4.7 rem/0.36 rem = a factor of 13. This provides substantial margin for uncertainties, but also creates undue burden and excessive restrictions on siting advanced reactors compared to the accepted level of protection for existing plants.

November 2023 The sensitivity case results demonstrate the following:

Case 2 demonstrates that if the advanced reactor PDD dose calculation required the use of statistically conservative weather results (i.e., 95th percentile weather), the inherent conservatism would be more than 26 times that of existing large LWR siting (given the doubling of the distance to 1 rem). For example, if the 30-day dose at 2x the distance to 1 rem is approximately 360 mrem (due to the decrease of dose with distance), then the comparison is 26 rem/0.36 rem = a factor of 72. Considering best estimate analyses, this sensitivity case layers an additional factor of 5.5 on the base dose calculation.

Case 3 demonstrates that if the advanced reactor PDD dose calculation required use of the peak wind direction results (in lieu of averaging the wind direction results), the inherent conservatism would be more than 7 times that of existing large LWR siting (given the doubling of the distance to 1 rem). For example, if the 30-day dose at 2x the distance to 1 rem is approximately 360 mrem (due to the decrease of dose with distance), then the comparison is 7 rem/0.36 rem = a factor of 19. Considering best estimate analyses, this sensitivity case layers an additional factor of 1.5 on the base dose calculation.

Case 4 demonstrates that if the advanced reactor PDD dose calculation required open field conditions for the public (e.g., not in structures at night), the inherent conservatism would be more than 12 times that of existing large LWR siting (given the doubling of the distance to 1 rem). For example, if the 30-day dose at 2x the distance to 1 rem is approximately 360 mrem (due to the decrease of dose with distance), then the comparison is 12 rem/0.36 rem = a factor of 33. Considering best estimate analyses, this sensitivity case layers an additional factor of 2.5 on the base dose calculation.

Case 5 demonstrates that if the advanced reactor PDD dose calculation was performed for a 1-year exposure period instead of a 30-day exposure period the inherent conservatism would be more than 10 times that of existing large LWR siting (given the doubling of the distance to 1 rem). Considering best estimate analyses, this sensitivity case layers an additional factor of 2.1 on the base dose calculation, thus case 5 also demonstrates that a 1-year exposure period does not result in doses that are 12 times that of the 30-day exposure, but rather doses that are only about double. This is attributed to natural decay and weathering (e.g., migration deeper into the soil) of the fission products through time. As such, for advanced reactors, if the 30-day dose at 2x the distance to 1 rem is approximately 360 mrem (due to the decrease of dose with distance), then a 1-year dose (without any credit for decontamination activities) could be approximately 720 mrem, not appreciably higher than the average annual exposure of 620 mrem for an individual in the U.S. from natural and man-made sources. This dose is well below the EPA PAGs [22] dose for intermediate phase relocation of the public of 2 rem in the first year.

In summary, the large LWR base case and sensitivity cases demonstrate that the proposed PDD dose criterion for advanced reactors is significantly more stringent than that of the existing fleet when assessed with best estimate consequence modeling, and would be excessively more stringent if conservative modeling were required. Case 5 (i.e., 1-year exposure) also indicates that the proposed criterion is only slightly above that of the average annual exposure for an individual in the U.S.

The Commission has previously stated that the current generation of plants is adequately safe, and so we find that the requirement for a PDD limit out to 2x the distance to 1 rem is exceedingly conservative

November 2023 and would place an undue burden on the owner of an advanced reactor with population related siting considerations unnecessarily limiting allowable sites when compared to the accepted level of protection for existing plants.

ADVANCED REACTORS POPULATION DENSITY DISTANCE DOSE ESTIMATES In order to consider the implications of the above on siting of advanced reactors, scoping dose estimates were calculated for two advanced reactor designs, a High Temperature Gas Reactor (HTGR) using TRISO (TRi-structural ISOtropic) fuel and a Molten Salt Reactor (MSR) based on the generic large LWR WinMACCS model. The purpose of these calculations was to estimate the PDD for these two generic designs for comparison to the NRC proposed dose criterion and the accepted level of protection for existing plants.

The release-related inputs for these scoping HTGR and MSR models were based on data developed by Sandia in SAND2020-0402 [8] for performing simplified scoping assessments of non-LWRs. The advanced reactor scoping models are based on a postulated MCA based upon review of the Sandia study and other publicly available materials. As noted for the large LWR model, an accident release frequency less than 5E-7/yr (i.e., once in 2 million years) is one potential benchmark screening value to justify development of the MCA for those reactor designs that have a PRA. Absent specific design information and accident analyses, the model inputs and results developed for these models are considered scoping estimates. In general, the MCA for each considers:

Small core inventory (e.g., 250 MWth) as compared to the large LWRs (e.g., 3000 MWth).

Consideration of percentage of the fuel subject to fuel damage.

Consideration of fission product migration from the fuel to the environment, including containment/confinement degradation. In the absence of severe accident induced containment failure (as is commonly postulated for large LWRs), containment degradation is viewed as leakage well beyond technical specification.

8.1 HTGR 13 Model and Results The HTGR model postulates a release with the following characteristics:

250 MWth power level based on fission product inventory scaling of approximately 10% of the Peach Bottom SOARCA LWR inventory (per Sandia). This is modeled as 15% of the Surry SOARCA inventory.

Normal activity shielding values are used (e.g., individuals inside structures at night).

1E-4 failure for TRISO fuel particles based on Sandias Table 7-2 [8] baseline probability and EPRIs report [14] on safety test data for TRISO fuel associated with the Advanced Gas Reactor (AGR) program. Failure of the particles allows the release to move from the fuel kernel into the pebble.

13 It is recognized that the data developed by Sandia and leveraged in this scoping model may not accurately reflect the latest design features planned for reactor technologies using TRISO fuel. Nevertheless, publicly available information from the Sandia report was used for inputs.

November 2023

Varied release fractions from the fuel pebble to the reactor coolant system based on the chemical properties of the various radionuclide groups. Sandia Table 4-2 references 95% for noble gases (Kr, Xe), 35% for halogens (I, Br), 25% for alkali metals (Cs, Rb), down to 2E-4 for lanthanides.

No credit is taken for deposition or holdup in the reactor coolant system (consistent with Sandia).

Degraded containment/confinement is assumed (i.e., leakage of the reactor enclosure assumed above technical specification for MCA purposes) of 10% for noble gases and 1% for other fission product groups. The Sandia report considered an intact reactor enclosure would provide 0.1%

leakage and a significantly impaired enclosure could provide 40% leakage.

A release to the environment with three plumes, with lower assumed heat content than the large LWR case. Given that the TRISO fuel failures are driven by a slow heat-up of the fuel to levels that may result in temperature induced damage, the total release is allocated to plumes 1, 2, and 3 as 10%, 50%, and 40% respectively, as follows:

NG I Cs Te Ba Ru / Mo La Ce Plume 1 9.50E-07 3.50E-08 2.50E-08 5.00E-09 2.00E-09 2.00E-09 2.50E-10 2.00E-11 Plume 2 4.75E-06 1.75E-07 1.25E-07 2.50E-08 1.00E-08 1.00E-08 1.25E-09 1.00E-10 Plume 3 3.80E-06 1.40E-07 1.00E-07 2.00E-08 8.00E-09 8.00E-09 1.00E-09 8.00E-11 Total 9.50E-06 3.50E-07 2.50E-07 5.00E-08 2.00E-08 2.00E-08 2.50E-09 2.00E-10 Height (m) Heat Delay (hr) Duration (w) (hr)

Plume 1 10 1E+5 2 3 Plume 2 10 1E+5 5 10 Plume 3 10 1E+5 15 10 The scoping WinMACCS model results for the representative HTGR release assuming best estimate consequence modeling (similar to the large LWR base case) are shown in Table 3.

Table 3 - HTGR Release 30-Day Dose Distance Mean Weather WD Average Shielding (rem) 100m 2.1E-3 200m 6.8E-4 300m 3.9E-4 600m 1.5E-4 1200m 6.0E-5 (0.75 miles)

November 2023 Based upon the best estimate inputs for scoping purposes, the distance for a dose of 1 rem in 30 days would be well under 100m. Considering the NRC proposed PDD criterion of 2x the distance to 1 rem, the 500 ppsm restriction would be expected to apply at a distance <100m, such that the PDD for a 250 MWth HTGR could be well within a 100m site boundary. It is noted however, that analysis variations (e.g., larger reactor enclosure leakage, ground level release, less plume heat reducing buoyancy of the plume, alternate near field dispersion modeling) could increase the calculated dose. Table 3 also shows that for the doubling of a distance, the dose decreases to 32% to 38% of the original value.

8.2 Molten Salt Reactor 14 Model and Results The MSR model postulates a release with the following characteristics:

250 MWth power level (assumed; it is not specifically presented in the Sandia report [8] but it is estimated that Sandia proposed a 250 MWth power level for each advanced reactor design assessed) based on a fission product inventory estimate scaling of approximately 1% of the Peach Bottom SOARCA inventory (per Sandia). This is modeled as 1.5% of the Surry SOARCA inventory.

100% of the molten salt fuel is assumed to be discharged into the reactor cell or drain tank from a large LOCA (per Sandia)

Fission product releases from the salt fuel into containment via vaporization are estimated based on ORNL-TM-732 [15] from the Molten Salt Reactor Experiment (MSRE) safety analysis.

Sandia presented postulated bounding cases. One case assumed high vaporization essentially equivalent to large LWR releases (low fission product solubility assumption). A second case assumed no vaporization (high fission product solubility assumption) except for noble gases. The ORNL-TM-732 safety analysis provides data for a maximum credible accident estimate of fission product release from the salt of 100% for noble gases, 10% for iodine, and 10% for release of solids.

ORNL-TM-732 [15] estimates a 50% plate-out of the iodine release on the secondary containment surfaces.

Degraded containment (leakage assumed above technical specification for MCA purposes) of 10% for noble gases and 1% for other fission product groups.

A release to the environment with three plumes, with negligible assumed heat content given the expected rapid cooling of the dispersed salt. The total release is allocated to plumes 1, 2, and 3 as 65%, 20%, and 15% respectively, consistent with the large LWR case, as follows:

NG I Cs Te Ba Ru / Mo La Ce Plume 1 6.50E-02 1.63E-04 6.50E-04 6.50E-04 6.50E-04 6.50E-04 6.50E-04 6.50E-04 Plume 2 2.00E-02 5.00E-05 2.00E-04 2.00E-04 2.00E-04 2.00E-04 2.00E-04 2.00E-04 Plume 3 1.50E-02 3.75E-05 1.50E-04 1.50E-04 1.50E-04 1.50E-04 1.50E-04 1.50E-04 14 It is difficult to use a single calculation to capture a bounding release for an MSR considering that most (if not all) MSR design concepts have a dedicated system to handle the volatile radionuclides that are released from the salt during normal operations. While the inclusion of such a system may reduce the volatility (i.e., the release fraction) of mobile radionuclides from a spill of fuel salt, some of the risk is then transferred to the off-gas handling system, where volatile (i.e., mobile) radionuclides may exist in nontrivial amounts. For the purposes of this scoping model, the Sandia report was used for representative MSR inputs.

November 2023 Total 1.00E-01 2.50E-04 1.00E-03 1.00E-03 1.00E-03 1.00E-03 1.00E-03 1.00E-03 Height (m) Heat Delay (hr) Duration (w) (hr)

Plume 1 10 0 2 3 Plume 2 10 0 5 5 Plume 3 10 0 10 10 The scoping WinMACCS model results for the representative MSR release assuming best estimate consequence modeling (similar to the large LWR base case) are shown in Table 4.

Table 4: MSR Release 30-Day Dose Distance Mean Weather WD Average Shielding (rem) 100m 23 200m 7.9 300m 4.5 600m 1.7 845m(1) 1.0 1200m 0.61 1690m 0.36(1)

(2x distance to 1 rem)

Note (1): Estimated based on trend line.

Based upon the best estimate inputs for scoping purposes, the distance for a dose of 1 rem in 30 days would be approximately 845m. Considering the NRC proposed PDD criterion of 2x the distance to 1 rem, the 500 ppsm restriction would be expected to apply within approximately 1690m (about 1.05 miles) of the facility. By comparison, if the accepted level of protection for the existing large LWR fleet is simply held constant, it is evident that the PDD for a 250 MWth MSR could be approximately 300m. It is noted however, that analysis variations (e.g., fission product solubility, reactor enclosure leakage, alternate near field dispersion modeling) would impact the calculated dose.

Table 4 also shows that for the doubling of a distance, the dose decreases to 34% to 38% of the original value, such that the estimated dose at 2x the distance is only about 360 mrem over thirty days. Based on the large LWR sensitivity case for a 1-year exposure time, the 1-year dose at the PDD of 1690m would be estimated to be about 720 mrem, not appreciably higher than the average annual exposure of 620 mrem for an individual in the U.S. from natural and man-made sources. Furthermore, this dose is well below the EPA PAGs [22] dose for intermediate phase relocation of the public of 2 rem in the first year.

November 2023 8.3 WinMACCS Model Summary Insights Based on the WinMACCS models for a generic large LWR representing the existing fleet of licensed LWRs and two notional advanced reactor designs, the following modeling-related insights are presented:

The proposed PDD criterion of 500 ppsm out to twice the distance at which the 1 rem dose was calculated for 30-days following any design-basis or beyond-design-basis event for advanced reactors is more than five times more stringent than that for existing large LWRs using best-estimate consequence assumptions, given that the dose decreases as a function of distance.

Dispersion modeling assumptions and variability (e.g., mean weather vs. 95% weather, shielding credit vs. no shielding credit over the 30 days) lead to significant differences in results. Requiring conservative analyses assumptions similar to those used for EAB or LPZ would be excessive compared to the accepted level of protection provided by the PDD for the existing large LWR fleet. Development of the PDD dose criterion for advanced reactors needs to clearly consider these pertinent modeling aspects.

The PDD estimates for the notional HTGR and MSR were calculated to range from <100m (for the HTGR) to approximately 1 mile (for the MSR). The 30-day dose at the PDD for the MSR would be about 360 mrem. It is noted that these results are based on generic scoping estimates, which are subject to some uncertainty and applicants would provide justification as otherwise required by NRC regulations regarding the MCA and containment response (e.g., the assumption of excessive containment leakage vs. containment failure) in performing their own analyses.

However, there is not so much uncertainty as to warrant anything more conservative than best estimate analysis.

The following additional items are noted with regard to the WinMACCS modeling results:

WinMACCS is an atmospheric dispersion code that only calculates the dose from the passing plume and deposited materials. For distances close to the reactor (e.g., within several hundred meters), the dose from fission products remaining in containment (i.e., containment shine) may be appreciable. Such dose does not represent societal impacts.

Modeling for the HTGR and MSR were generally consistent with the generic large LWR model to support comparison purposes. For distances close to the reactor (e.g., <500m), alternate near field dispersion model options available within WinMACCS may be more appropriate. Given the large uncertainties associated with other inputs, however, these modeling refinements were not pursued for purposes of this scoping evaluation.

WinMACCS has the ability to determine distances associated with land remediation following an atmospheric release to the environment (e.g., contamination, condemnation). The models developed for this white paper were simplified and did not employ those elements of WinMACCS. Such results could provide additional perspective on societal impacts.

DEFENSE-IN-DEPTH CONSIDERATIONS As presented previously in the last column of Table 1a, each siting distance provides defense-in-depth for protection of the public from postulated accidents due to potential failures of systems and fission

November 2023 product barriers. In regard to small modular reactors and micro-reactors, dose calculations (as illustrated in this white paper) may show very close distances for siting elements like the EAB, LPZ, PCD, and PDD. These close distances may appear to reduce the defense-in-depth protection afforded by siting distances; however, it is noted that these close distances are due to smaller potential fission product inventories and the inherent safety features of advanced reactor designs which deserve credit.

Given the varied designs of advanced reactors, an in-depth assessment of defense-in-depth is beyond the scope of this white paper. However, defense-in-depth considerations are presented in Table 5 for each siting distance to facilitate consideration by reactor designers. Specifically, an evaluation could identify how the reactor design protects the public given assumed prior failures (e.g., failure of active safety systems). With regards to the societal benefit of the PDD, it is noted that the protection afforded by PCD may largely address protection afforded by PDD as the distances decrease.

Table 5 - Defense-In-Depth Considerations for Siting Advanced Reactors Defense-In-Depth EAB LPZ PCD PDD Layer Passive Features If PFs inadequate If PFs inadequate If PFs inadequate If PFs inadequate (PF)

Active Systems & If AS/OAs If AS/OAs If AS/OAs If AS/OAs Operator Actions inadequate inadequate inadequate inadequate (AS/OA)

In Plant Fission If FPBs If FPBs If FPBs If FPBs Product Barriers inadequate inadequate inadequate inadequate (FPB)

Environmental EB success. EB success. If EB inadequate / If EB inadequate /

Barrier (EB) 15 Generally relied Generally relied degraded degraded upon for EAB upon for LPZ considerations. considerations.

EPZ Protective PAs not expected If PAs required, If PAs required, If PAs required, Actions (PA) to be needed the LPZ promotes the PCD promotes the PDD portion their success. their success. within the EPZ may promote their success.

Distance (to Societal impacts Broader societal PCD should PDD should minimize broader not expected impacts not minimize broader minimize broader societal impacts 16) expected societal impacts societal impacts It should also be noted that the design of the advanced reactors is such that the likelihood of a severe accident scenario is much lower than it is for the current fleet (e.g., more passive safety features). This indicates that it is not necessary to apply a stricter PDD criterion to these reactors as their safety is already significantly enhanced.

15 The Environmental Barrier can be viewed as the last substantial fission product barrier that prevents release to the environment. For many reactor designs this is the containment / confinement barrier.

16 Societal impacts can be viewed to include evacuation or temporary relocation of individuals, while broader societal impacts can be viewed as those with longer term and /or larger regional disruptions which could include impacts from decontamination activities or condemned land.

November 2023 ALTERNATE PDD DOSE CRITERION FOR DISCUSSION Based on the analysis presented above, it is evident that the NRC staffs proposed Option 3 (approved by the Commission in SRM-SECY-20-0045) for PDD for advanced reactors of 500 ppsm out to twice the distance to 1 rem in 30 days is very conservative as compared to the NRC accepted level of protection for currently licensed large LWRs. Further, it is overly restrictive with respect to societal risk given that the decrease of dose with distance (based on the large LWR ratio of 1 year exposure sensitivity case to base case, i.e., factor of approximately 2) is expected to result in an annual dose within the EPA PAGs

[22] dose for intermediate phase relocation of the public of 2 rem in the first year for advanced reactors without multiplication of the distance to 1 rem. Therefore, an alternate criterion is proposed and discussed.

For perspective, the PDD for large LWRs has typically been the longest siting distance as compared to the EAB, LPZ, and PCD. The criteria for these distances are summarized as follows:

EAB - 25 rem in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (using conservative modeling assumptions)

LPZ - 25 rem in 30 days (using conservative modeling assumptions)

PCD - 1.33 times the LPZ distance, which equates to approximately 17 rem in 30 days for the select advanced reactor scoping cases evaluated.

Alternate PDD - 5 rem in 30 days using best estimate dose calculations This alternative proposes that the PDD dose criterion be aligned to the current NRC accepted level of protection for large LWRs calculated using best estimate methods (e.g., average weather conditions, average of wind directions, normal activity shielding credit for individuals being in structures at night).

The sensitivity cases described in this paper serve as justification for not using conservative methods (e.g., 95% weather, peak wind direction, open field conditions 24/7) applied to other siting dose calculations such as the EAB and LPZ. A dose value of 5 rem in 30 days for advanced reactors still provides added margin given the NRCs Policy Statement on the Regulation of Advanced Reactors [23]

expectations that advanced reactors will provide enhanced margins of safety and/or use simplified, inherent, passive, or other innovative means to accomplish their safety and security functions. Further, this alternative would continue the NRC practice of implementing the policy statement in ways that align requirements with advanced reactor safety profiles.

CONCLUSIONS Based upon the considerations outlined above and scoping modeling performed, the following conclusions are drawn with regards to the NRC proposed 500 ppsm population density distance dose criterion of 2x the distance of 1 rem in 30 days:

The original basis for the NRCs establishment of a 20-mile PDD was to 1) minimize land contamination that would require permanent relocation, and 2) achieve dose rates that would meet the QHOs for latent cancer fatalities.

A PDD of 5 rem in 30 days (rather than 2x the distance of 1 rem in 30 days) would align with the current NRC accepted level of protection for large LWRs calculated using best estimate methods

November 2023 at 20 miles. This alternative should be pursued to acknowledge the features of advanced reactors and avoid undue burden for their siting by aligning requirements with advanced reactor safety profiles.

The NRCs proposed criterion associated with very low frequency events, 2x the distance of 1 rem in 30 days, is more than a factor of five times more conservative compared to previous large LWR licensing when calculated using best estimate consequence methods. If NRC proceeds with a PDD of twice the distance to 1 rem in 30 days, the associated guidance should accept best estimate methods from all applicants. The sensitivity analyses described in this white paper may be referenced as justification.

If the NRC imposes a dose criterion on advanced reactors that is less than one-fifth the acceptable dose for large LWRs, then this excessive baseline margin to previous large LWR licensing experience should limit the need for extensive uncertainty analysis given that similar uncertainty as determined in the large LWR comparison sensitivity cases would be expected for advanced reactors. This supports the application of best estimate methods, without further justification, to a variety of advanced reactor licensing regimes (i.e., LMP, LWR and non-LWR using traditional approach).

If the NRC imposes more conservative assumptions and modeling approaches, then, based on sensitivity cases alone, the level of public protection will additionally layer anywhere from 19 to 72 times more conservatism than what the NRC currently accepts for large LWRs.

The dose criterion calculations are highly dependent upon modeling assumptions (e.g., assumed weather conditions, shielding) that need to be clearly considered with respect to the realistic exposure risk to the public.

Advanced reactor safety profiles already offer a greater level of protection against large release accidents as these new designs have advanced safety features and levels of defense in depth that make the frequency of such accidents much lower than for the current fleet. This further supports the conclusion that best estimate analysis approaches are more than adequate for defining the PDD.

REFERENCES

[1] Regulatory Guide 4.7 General Site Suitability Criteria for Nuclear Power Stations, Revision 3, March 2014.

[2] SECY-20-0045, Population Related Siting Considerations for Advanced Reactors, May 2020.

[3] SRM-SECY-20-0045, Staff Requirements - SECY-20-0045 - Population-Related Siting Considerations for Advanced Reactors, July 2022.

[4] NUREG-0625, Report of the Siting Policy Task Force, August 1979.

[5] NUREG/CR-7110, State-of-the-Art Reactor Consequence Analyses Project, Surry Integrated Analysis, Vol. 2, Revision 1, August 2013.

November 2023

[6] NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis Guidance Document, November 2005.

[7] NUREG/CR-7270, Technical Bases for Consequence Analyses Using MACCS (MELCOR Accident Consequence Code System), October 2022.

[8] SAND2020-0402, Simplified Approach for Scoping Assessment of Non-LWR Source Terms, January 2020.

[9] NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, Revision 1, August 2019.

[10] NRC Draft White Paper, Alternative Approaches to Address Population-Related Siting Considerations, April 2023.

[11] Draft Regulatory Guide DG-4034, Proposed Revision 4 to Regulatory Guide 4.7: General Site Suitability Criteria for Nuclear Power Stations, ML23123A090, October 2023.

[12] NRC Public Meeting Slides, DG-4034 (RG 4.7 Revision 4), Appendix A, Alternate Approaches to Address Population-Related Siting Considerations, ML23298A134, October 27, 2023.

[13] NUREG-0396, Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants, December 1978.

[14] EPRI-AR-1(NP)-A, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO)-Coated Particle Fuel Performance, November 2020.

[15] ORNL-TM-732 MSRE Design and Operations Report, Part V, Reactor Safety Analysis Report, August 1964.

[16] NRC Advanced Reactor Stakeholder Public Meeting Slides, ML23200A301, July 20, 2023.

[17] ORNL/TM-2019/1197, Advanced Reactor Siting Policy Considerations, June 2019.

[18] WASH-1308, Population Distribution around Nuclear Power Plant Sites, 1973.

[19] NRC Final Rule, Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants, 61 Fed. Reg. 65157, Dec. 11, 1996.

[20] NUREG/CR-6295, Reassessment of Selected Factors Affecting Siting of Nuclear Power Plants, February 1997.

[21] NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, 1990.

[22] EPA-400/R-17/001, PAG manual: Protective Action Guides and Planning Guidance for Radiological Incidents, January 2017.

[23] NRC-2008-0237, Policy Statement on the Regulation of Advanced Reactors, 73 Fed. Reg. 60612 Oct. 14, 2008

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