_____________ 

POLICY ISSUE 

(Information) 

February 7, 2016 SECY-16-0012 

FOR: The Commissioners 

FROM: Victor M. McCree 

Executive Director for Operations 

SUBJECT: ACCIDENT SOURCE TERMS AND SITING FOR SMALL 

MODULAR REACTORS AND NON-LIGHT WATER REACTORS 

PURPOSE: 

To inform the Commission of the status of staff activities related to accident source terms for 

small modular reactors (SMRs) and non-light water reactors (non-LWRs), and the staff’s current 

assessment of potential policy issues associated with use of mechanistic source terms (MSTs) 

in design-basis accident (DBA) dose analyses and siting. 

This paper responds to Commission direction in Staff Requirements Memorandum 

(SRM)-M110329, "Staff Requirements – Briefing On Small Modular Reactors, 9:00 A.M., 

Tuesday, March 29, 2011, Commissioners’ Conference Room, One White Flint North, Rockville, 

Maryland (Open to Public Attendance),” dated April 14, 2011 (Agencywide Documents Access 

and Management System (ADAMS) Accession No. ML110880535). This paper does not 

address any new commitments or resource implications. 

SUMMARY: 

As a result of pre-application activities over recent years and earlier work by the U.S. Nuclear 

Regulatory Commission (NRC) staff and the Commission, the NRC staff previously identified a 

number of potential policy and licensing issues for SMRs and non-LWRs. One issue is 

associated with the determination of source terms and the resulting dose calculations and siting 

evaluations. Current regulations and guidance, while maintaining the requirement to assume 

CONTACTS: Arlon Costa, NRO/DEIA 

 (301) 415-6402 

 Michelle Hart, NRO/DSEA 

 (301) 415-1265 
The Commissioners - 2 - 

substantial meltdown of the core and substantial release into containment, permit the use of 

MSTs to account for the design-specific accident scenarios and accident progression in 

developing DBA radiological source terms. These same source terms are then used as a basis 

for evaluating equipment qualification, vital area access, some shielding calculations, and as 

part of the site suitability determinations. Near-term SMR applicants will likely propose to follow 

the existing regulations and guidance to develop design-specific DBA scenarios, coupled with 

more realistic mechanistic methods to model the accident progression and develop the source 

terms for these scenarios. Evaluation of the mechanistic methods will be an important part of 

the staff’s review, but the approach proposed by near-term SMR design certification applicants 

does not currently appear to raise any policy issues. 

Use of MSTs for DBAs for SMRs, given the expected smaller amount of fuel and unique and 

passive nature of these designs, is expected to result in reduced source terms when compared 

to large LWRs. These reduced source terms could form the basis for an applicant request to 

establish emergency planning zones that are smaller than what is currently required by Title 10 

of the Code of Federal Regulations (10 CFR) 50.47(c)(2). In addition, the reduced source terms 

could result in smaller exclusion areas and low population zones (LPZs) as defined in 10 CFR 

100.3, as determined in accordance with the safety assessment and dose criteria in 10 CFR 

50.34(a)(1). Any NRC-approved reduction in the size of the LPZ could, in turn, allow such a 

reactor to be sited in closer proximity to a large population center as compared to large LWRs, 

as provided under 10 CFR 100.21(b). Any proposed site would also need to be consistent with 

other NRC requirements including 10 CFR 100.21(h), which limits, in qualitative terms, how 

close to the large population center a site can be. 

The staff is informing the Commission of this potential policy issue in accordance with 

Commission direction to the staff to think expansively about upcoming SMR issues and engage 

the Commission early on matters of policy.1 Some industry organizations have expressed a 

desire to deploy SMRs to replace existing coal plants, and acknowledge that to do so would 

require siting SMRs closer to population centers than large LWRs typically have been. For 

example, the Electric Power Research Institute’s Utility Requirements Document states that 

“[T]he deployment of [SMRs] to replace retired coal-fired power stations in critical locations for 

electric grid stability requires acceptance of small nuclear plants closer to population centers.” 

Staff intends to further engage stakeholders to understand whether proposed industry plans 

include siting of SMR facilities closer than has typically been allowed for large LWRs, and 

whether a clarification of the NRC’s position on siting is warranted. 

DISCUSSION: 

Calculations of Source Terms 

Quantitative determination of radioactive material that could potentially escape from the fuel and 

subsequently from the reactor core during normal operation or as a result of an accident, and 

ultimately be released outside the containment structure, plays a critical role in the facility’s 

design and the agency’s requirements to protect public health against radiation hazards. The 

analysis of the details of the timing and type of accident that could occur and the related amount 

of radioactive material that could be released in the event of that accident involve some 

 

1 SRM, “Briefing on Small Modular Reactors,” dated April 14, 2011, (ADAMS Accession No. ML110880535).
The Commissioners - 3 - 

uncertainty; therefore, DBA dose assessments typically are performed using methods that yield 

conservative dose estimates that are then compared to the regulatory requirements. The 

quantification of the accidental release of fission products into the containment atmosphere, or 

accident radiological “source term,” affects the design of plant systems and is one element used 

to determine site suitability by meeting the dose criteria established for power reactor siting and 

safety system design in 10 CFR 50.34(a)(1)(ii)(D). A method to develop an MST is applicable to 

both SMRs and non-LWRs; however, there will be differences in the details of the analysis 

resulting from differences in the designs. 

Significant progress has been made through the years in understanding reactor accident 

behavior for LWRs, including fission product release and transport. This increased technical 

understanding results in more detailed mechanistically based assessments of source terms to 

estimate the release and behavior of these fission products, which can substantially attenuate 

the magnitude of the release as compared to a more deterministic approach. The use of a 

mechanistic analysis includes accounting for fission product retention and removal processes. 

A DBA source term is intended to be representative of a major accident scenario involving 

significant core damage, with radiation released to the environment through expected leakage 

from an intact containment. 10 CFR 50.34(a)(1)(ii)(D), which is referenced by 10 CFR 100.21, 

“Non-Seismic Siting Criteria,” requires that the total effective dose equivalent (TEDE) from this 

source term to an individual at the exclusion area 2 boundary (EAB) and at the outer boundary of 

the LPZ not exceed 25 rem TEDE within any 2-hour period of a release and for the duration of 

the release, respectively. Among other siting criteria, one criterion contained in 

10 CFR 100.21(b) requires that the distance to the nearest population center of at least 25,000 

people is at least one and one-third times the distance from the reactor to the outer boundary of 

the LPZ. Other NRC requirements (10 CFR 100.21(h)) address the Commission’s stated policy 

to site reactors “away from very densely populated centers” with a preference for “areas of low 

population density.” More information on the specific NRC reactor siting and safety analysis 

requirements can be found in the enclosure. 

A staff memorandum 3 issued on June 20, 2014, documented the status of this issue through 

the previous years. That memorandum included information on staff interactions with 

stakeholders and vendors such as the Nuclear Energy Institute, Generation mPower LLC, 

NuScale Power, and Department of Energy/Idaho National Laboratory. Communications and 

meetings with potential applicants were focused on design-specific activities associated with the 

DBA scenario. Since the 2014 memorandum, the staff has continued interactions with NuScale 

and, to a lesser degree, with the other stakeholders mentioned. In December 2015, NuScale 

submitted a licensing topical report4 for staff review and approval of the report’s accident source 

term methodology. The enclosure provides more background information on the evaluation of 

MSTs. 

The staff is well prepared to review DBA source terms for SMRs, including MST analyses. The 

staff has the tools and technical expertise to evaluate the appropriate topics, such as scenario 

 

2 See 10 CFR 100.3, “Definitions.” 

3 Commission Memorandum, “Status of Mechanistic Source Term Policy Issue for Small Modular Reactors” 

June 20, 2014 (ADAMS Accession No. ML14135A482). 

4 NuScale Power, Topical Report TR-0915-17565, "Accident Source Term Methodology," Revision 0, dated 

December 2015 (ADAMS Accession No. ML16004A217). 
The Commissioners - 4 - 

selection, design-basis accident progression, fission product transport and removal processes, 

and the treatment of uncertainty. Although the staff has not yet developed source term tools 

and technical expertise for non-LWRs to the same level as that for SMRs, the staff believes a 

mechanistic approach could also be applied to non-LWR designs, subject to the availability of 

adequate tools and analysis approaches. In pre-application discussions with SMR designers, 

an overview and outline of proposed methods have been discussed. The methods proposed by 

the potential applicants in those interactions appear to generally build on currently approved 

methods. The staff did not note any specific area of concern during those interactions, nor any 

potential policy issues regarding implementation of mechanistic modeling of DBA source terms 

for SMRs. 

Potential Effects of Source Terms on Siting 

As discussed above, SMRs and non-LWRs may be designed with smaller cores and passive 

safety features, which may result in the calculation of smaller releases to the environment for a 

DBA and the resulting hypothetical dose calculation. This in turn would provide additional 

margin with respect to the dose related portions of the current NRC siting requirements focused 

on radiological safety. For example, future SMR applicants may be able to show that an 

individual at the EAB would not receive a dose that exceeds 1 rem TEDE within any 2-hour 

period of a release. This calculated dose is much lower than the current regulatory requirement 

that the dose not exceed 25 rem TEDE.

This results in the potential for the EAB and LPZ to be at the same distance around a very small 

site, potentially at a few hundred meters from the center of the reactor location or facility. The 

dose criteria which would allow for a smaller LPZ would also potentially allow the reactor to be 

considered for a location at a distance that is relatively close to a population center with at least 

25,000 people. This reduced distance would be considered in combination with other siting 

requirements, including 10 CFR 100.21(h), which states 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 density, consideration will be given to 

safety, environmental, economic, or other factors, which may result in the site 

being found acceptable. 

Although 10 CFR 100.21(h) could allow siting a reactor closer to a “densely populated center,” it 

also requires a plant to be sited away from a “very densely populated area.” 

It is important to note that dose to individuals is only one element in the considerations of siting 

and proximity to densely populated centers. Siting requires considerations of other factors as 

cited in 10 CFR 100.20, “Factors to be considered when evaluating sites.” These include 

population density and use characteristics of the site environs, nature and proximity of manrelated hazards, and physical characteristics of the site, which includes seismology, 

meteorology, geology, and hydrology. 

There are currently no combined license (COL) applications scheduled that employ an SMR or 

non-LWR design, nor have any pre-application discussions to date specifically indicated that an 

applicant plans to site its facility closer to densely populated centers. However, the staff 
The Commissioners - 5 - 

understands that future applicants may request to site SMRs at retired fossil fuel plants or other 

industrial sites. For example, one SMR vendor has stated that its design “can be sited next to 

population centers without any threat to the local environment or populace.” Therefore, the staff 

will proactively engage with stakeholders to determine whether clarification on siting is 

needed. If so, the staff will inform the Commission and request approval to develop a draft 

regulatory basis for Commission consideration that can be subsequently included in regulatory 

guidance. The staff anticipates that development of the draft regulatory basis will take 

approximately 18 months and, if approved by the Commission, another year will be needed to 

finalize the regulatory guidance. In light of resource constraints and other higher priorities, the 

staff believes this approach to develop guidance is the most prudent means of handling this 

issue. Of course, the staff will ensure that it remains abreast of industry plans so that the staff 

can complete the regulatory guide ahead of the time it will be needed. 

CONCLUSIONS: 

The Commission has previously provided guidance to the staff on the use of MST analysis and 

the staff has implemented it in appropriate stages, such as the development of guidance for 

justifiable mechanistic analysis for an alternative accident source term. SMR and non-LWR 

applicants can employ modern analysis tools to demonstrate quantitatively the safety features of 

those designs. MST analysis methods can also be used by applicants to demonstrate the ability 

of the enhanced safety features of plant designs to mitigate accident releases. This would also 

allow future COL applicants to consider reduced distances to EABs and LPZs, and potentially 

increased proximity to population centers. 

All currently known near-term applicants are utilizing sites that appear to comply with the 

directive in 10 CFR 100.21(h) to locate away from very densely populated areas. However, the 

staff may need additional direction from the Commission if an applicant were to propose a site 

that is significantly closer to a very densely populated center than previously approved. 

Staff will explore these issues with interested stakeholders over the next 12 to 18 months. 

Engaging stakeholders on this issue relatively early in the licensing process for these designs 

will provide time for the staff to propose any changes to the Commission. Any policy issues and 

the staff’s proposed path forward will be provided to the Commission in a future paper. 
The Commissioners - 6 - 

COORDINATION: 

The Office of the General Counsel has reviewed this paper and has no legal objection. 

 /RA/ 

Victor M. McCree 

Executive Director 

 for Operations 

Enclosure: 

Background Information Regarding 

 Accident Source Terms and Siting 
The Commissioners -6- 

 

COORDINATION: 

The Office of the General Counsel has reviewed this paper and has no legal objection. 

Victor M. McCree 

Executive Director 

 for Operations 

Enclosure: 

Background Information Regarding 

 Accident Source Terms and Siting 

WITS 201100172 

 ADAMS Accession No.: ML15309A319 via e-mail* SECY-012 

OFFICE NRO/DARR:PM NRO/DSEA:TR NRO/DARR:BC NRO/DARR:D 

NAME ACosta MHart ABradford MMayfield 

DATE 11/05/2015 11/06/2015 11/27/2015 12/01/2015 

OFFICE NRO/DSEA:D RES:D NRR:D NSIR:D 

NAME SFlanders MWeber BDean (J.Giitter 

for) 

B. Holian (K. Coyne 

for) 

DATE 12/01/2015 12/08/2015 12/14/2015 12/11/2015 

OFFICE Tech Editor OGC NRO:D EDO 

NAME CHsu PMoulding 

(S. Kirkwood for) 

JUhle VMcCree 

DATE 12/07/2015 01/15/2016 01/15/2016 02_/07/2016 

OFFICIAL RECORD COPY 
Enclosure 

Background Information Regarding Accident Source Terms and Siting 

Relevant Documents Summary 

The following is a short summary of relevant Nuclear Regulatory Commission (NRC) documents 

related to accident source term analyses, which are discussed in more detail later in this 

enclosure: 

• TID-14844 (1962):1 Used a conservative and deterministic approach to estimate 

light-water reactor (LWR) off-site doses from a postulated core melt accident to account 

for site suitability in compliance with Title 10 of the Code of Federal Regulations

(10 CFR) Part 100, “Reactor Site Criteria.” 

• SECY-90-016 (1990):2 Provided recommendations concerning departures from 

regulations for evolutionary LWRs while recognizing that deviations from practices at the 

time could be warranted in future designs. This led to the development of a more 

physics-based design-basis accident (DBA) source term than the TID-14844 

source term. 

• SECY-93-092 (1993):3 Defined mechanistic source terms (MSTs) and recommended 

that scenario-specific (mechanistic) source terms be allowed. This included a caveat 

such that there should be sufficient understanding of fuel performance, fission product 

behavior, and accident selection to bound uncertainties. The Commission approved 

this staff recommendation in a July 30, 1993, staff requirements memorandum (SRM) 

(Agencywide Documents Access and Management System (ADAMS) Accession 

No. ML003760774). 

• NUREG-1465 (1995):4 Provided insights from severe accident research on fission 

product release and transport that were used in developing a revised source term for 

large LWRs, which was expressed in terms of times and rates of appearance of 

radioactive fission products into the containment, the types and quantities of the species 

released, and other important attributes such as the chemical forms of iodine, given a 

severe core-melt accident. 

• RG 1.183 (2000):5 Provided elements of a mechanistic approach with the deterministic 

analysis and, for regulatory purposes, a more realistic estimate of the amount of fission 

 

1 Technical Information Document TID-14844, “Calculation of Distance Factors for Power and Test 

Reactor Sites,” Atomic Energy Commission, March 23, 1962 (ADAMS Accession No. ML021750625). 2 SECY-90-016, “Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship 

to Current Regulatory Requirements,” January 12, 1990 (ADAMS Accession No. ML003707849). 3 SECY-93-092, “Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, PIUS) and CANDU 3 Designs 

and Their Relationship to Current Regulatory Requirements,” dated April 8, 1993 (ADAMS Accession 

No. ML040210725). 4 NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants,” February 1995 

(ADAMS Accession No. ML041040063). 5 Regulatory Guide RG 1.183, “Alternative Radiological Source Terms for Evaluating Design-Basis Accidents 

at Nuclear Power Reactors,” July 2000 (ADAMS Accession No. ML003716792). 
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products present in the containment from a postulated severe accident for large LWRs 

than was included in TID-14844. 

• SECY-03-0047 (2003):6 Retained previous Commission's guidance in an SRM 7

(consistent with previous Commission and Advisory Committee on Reactor Safeguards 

(ACRS) views contained in SRM-SECY-93-092) that allowed the use of scenario-specific 

source terms for non-LWRs, provided that there were sufficient understanding and 

assurance of plant and fuel performance and deterministic engineering judgement used 

to bound uncertainties. It also provided support for moving this issue toward 

performance-based regulations and away from prescriptive regulations. 

• SECY-05-0006 (2005):8 Proposed a framework for the “Use of Scenario-Specific 

Source Terms for Licensing Decisions” and licensing approach for new plant licensing, 

in support of SRM-SECY-03-0047. 

• NRC Policy Statement (2008):9 Defined along with NUREG-1226 10 , advanced reactors 

as those with innovative designs for which licensing requirements will be significantly 

different from the existing LWR requirements. These documents also provide guidance 

for developing new regulatory requirements supporting licensing advanced reactor 

designs. 

• Commission Memorandum (2011):11 Provided the status on planned activities to 

address methods for determining a MST and to describe the circumstances in which a 

source term determined by such methods could be appropriate. 

• Commission Memorandum (2013):12 Provided the status of activities for the siting 

source term and emergency plan (EP) issues for small modular reactors (SMRs). 

• Commission Memorandum (2014):13 Documented the status of this issue since 

issuance of SECY-10-0034 and also described staff interactions with stakeholders, such 

as public meetings during the 2010–2014 period. 

 

6 SECY-03-0047, “Policy Issues Related to Licensing Non-Light-Water Reactor Designs,” March 28, 2003 

(ADAMS Accession No. ML030160002). 7 SRM-SECY-03-0047, “Policy Issues Related to Licensing Non-Light Water Reactor Designs,” dated 

June 26, 2003 (ADAMS Accession No. ML031770124). 8 SECY-05-0006, “Second Status Paper on the Staff’s Proposed Regulatory Structure for New Plant Licensing 

and Update on Policy Issues Related to New Plant Licensing,” dated January 7, 2005 (ADAMS Accession 

No. ML043560093). 9 73 FR 60612, “Policy Statement on the Regulation of Advanced Reactors,” October 14, 2008. 10 NUREG-1226, “Development and Utilization of the NRC Policy Statement on the Regulation of 

Advanced Nuclear Power Plants,” June 1988. 11 Commission Memorandum, “Status of Staff Activities to Address Mechanistic Source Term Methodology and 

Its Application to Small Modular Reactors,” December 29, 2011 (ADAMS Accession No. ML113410366). 12 Commission Memorandum, “Current Status of the Source Term and Emergency Preparedness Policy 

Issues for Small Modular Reactors,” May 30, 2013 (ADAMS Accession No. ML13107A052). 13 Commission Memorandum, “Status of Mechanistic Source Term Policy Issue for Small Modular 

Reactors” June 20, 2014 (ADAMS Accession No. ML14135A482). 
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DBA Source Term and MST Requirements 

The regulations in 10 CFR 50.2, “Definitions,” define source term as “the magnitude and mix of 

the radionuclides released from the fuel, expressed as fractions of the fission product inventory 

in the fuel, as well as their physical and chemical form, and the timing of their release.” The 

safety analysis described in 10 CFR 50.34, “Contents of Applications; Technical Information,” 

requires consideration of an accident that results in core melting and release of fission products 

to the containment. For large LWRs, the loss-of-coolant accident (LOCA) is the DBA that most 

fits the required analysis as described in 10 CFR 50.34(a)(1), and it models an arrested core 

melt with releases from the reactor coolant system (RCS) to the intact containment, fission 

product transport, and removal in the containment and other engineered safety features (ESFs), 

and release to the environment through design containment leakage at the leak rate 

demonstrated through technical specification surveillance. DBA dose analyses for accidents 

other than the LOCA are also evaluated in the safety analysis report. Non-LOCA DBA dose 

analyses also have radiological source terms, but in general terms they are based on calculated 

bounding fuel failure rates (if applicable to the accident), or the reactor coolant may be the only 

source of radioactivity for release through design-specific calculated coolant steaming and leak 

rates. 

The deterministic LOCA source term in TID-14844 was used by most of the operating reactors 

in initial licensing, and is the technical basis to calculate distance requirements originally used 

for 10 CFR Part 100 since 1962. The TID-14844 LOCA assumed an instantaneous and 

homogenous release of 100 percent of the noble gases and 50 percent of the iodines from the 

core to the containment, 50 percent of which was assumed to deposit on surfaces. These 

assumptions were applied to all large LWR designs. 

In 2000, guidance in Reg Guide 1.183 for an alternative source term (AST) originated from 

severe accident research 14 and physically based (mechanistic) models. It contains one LOCA 

source term for all pressurized-water reactors (PWRs) and a second one for all boiling-water 

reactors (BWRs). The AST LOCA accident scenario similarly deterministically assumed initial 

failure of core cooling, but the progression and release to containment was mechanistically 

modeled to be applicable to large LWRs (either PWRs or BWRs). The AST release to 

containment occurs over time instead of instantaneously, as in TID-14844. 

Earlier discussions regarding SMRs and non-LWR designs focused on a full mechanistic 

approach, meaning that the “mechanistic source term,” or MST, uses scenario-specific 

mechanistic or physically based modeling of accident progression to estimate a design-specific 

DBA radiological source term release from the RCS to the containment to demonstrate 

compliance with reactor siting and safety assessment regulations. It is a further evolution of 

the DBA radiological source term. 

The “Policy Statement on the Regulation of Advanced Reactors” states that the Commission 

expects that “advanced reactors will provide enhanced margins of safety and/or utilize 

simplified, inherent, passive, or other innovative means to accomplish their safety functions.” 

The staff understands that SMR and non-LWR designers are including safety margins or using 

simplified inherent, passive, or other innovative means to accomplish safety functions in their 

 

14 Because of SECY-90-016 and the related SRM, the revised source term for LWRs in NUREG-1465 was 

developed to provide a more physically based DBA source term to replace the TID-14844 source term. 
-4- 

specific reactor design features. One such feature is the passive emergency core cooling 

function. In general, the designs rely on limited amounts of equipment, rely on natural 

circulation, and do not call for operator action to initiate the function. Under current NRC 

regulations, SMR and non-LWR designers are required to provide in their applications the 

appropriate technical bases for the assumptions and methods they use to model the DBA 

scenarios and system response, including the bases for any passive safety functions that are 

credited in fission product release mitigation. This is in line with the evaluations of large LWR 

passive design features as has already been done for the certified reactor designs of 

AP600 and AP1000, as well as the Economic Simplified Boiling-Water Reactor. 

SECY-02-0139 15 framed the proposal to use MSTs that are plant-specific, by stating: 

Current LWRs use site-specific parameters (e.g., exclusion area boundary) and a 

predetermined source term into containment to analyze the effectiveness of the 

containment and site suitability for licensing purposes. These source terms are 

described in documents TID-14844 and NUREG-1465 and are based upon 

enveloping the fission product releases that would be predicted to occur given 

a core melt accident. On the other hand, future plants, particularly non-LWRs, 

propose not to use a predetermined source term for assessing the effectiveness 

of plant mitigation features or site suitability, but rather to use plant-specific 

accident source terms corresponding to each of the AOOs [anticipated 

operational occurrences] and DBEs [design basis events] defined for the plant. 

Such an approach puts a burden on the applicant and staff to understand the 

fission product release characteristics and uncertainties associated with a variety 

of accident scenarios. 

SECY-93-092 and its associated SRM also provided the following definition and staff 

recommendation on MST for advanced non-LWR designs: 

[Definition:] 

A mechanistic source term is the result of an analysis of fission product release 

based on the amount of cladding damage, fuel damage, and core damage 

resulting from the specific accident sequences being evaluated. It is developed 

using best-estimate phenomenological models of the transport of the fission 

products from the fuel through the reactor coolant system, through all holdup 

volumes and barriers, taking into account mitigation features, and finally, into the 

environs. 

[Recommendations:] 

Advanced reactor and CANDU 3 source terms should be based upon a 

mechanistic analysis and will be based on the staff's assurance that the 

provisions of the following three items are met: 

• The performance of the reactor and fuel under normal and off normal 

conditions is sufficiently well understood to permit a mechanistic analysis. 

Sufficient data should exist on the reactor and fuel performance through 

 

15 SECY-02-0139, “Plan for Resolving Policy Issues Related to Licensing Non-Light Water Reactor Designs,” 

dated July 22, 2002 (ADAMS Accession No. ML021790610). 
-5- 

the research, development, and testing programs to provide adequate 

confidence in the mechanistic approach. 

• The transport of fission products can be adequately modeled for all 

barriers and pathways to the environs, including specific consideration of 

containment design. The calculations should be as realistic as possible 

so that the values and limitations of any mechanism or barrier are not 

obscured. 

• The events considered in the analyses to develop the set of source 

terms for each design are selected to bound severe accidents and 

design-dependent uncertainties. 

The design-specific source terms for each accident category would constitute one 

component for evaluating the acceptability of the design. 

The staff indicated that this approach provides, for regulatory purposes, a more realistic 

estimate of the amount of fission products present in the containment from a postulated severe 

accident than was included in TID-14844. It was recognized that this approach to source term 

modelling can have implications on issues related to 10 CFR Part 100, equipment qualification, 

control room habitability, and assessments of severe accident risks in plant environmental 

impact statements. The SRM further stated “that a mechanistic ‘scenario specific’ source term 

for each reactor concept warrants further consideration before evaluating the acceptability of the 

design.” 

In SECY-03-0047, the staff reaffirmed the earlier approach in SECY-93-092, recommending the 

following, which was approved by the Commission in its associated SRM dated June 26, 2003: 

• Retain the Commission’s guidance contained in the July 30, 1993, SRM 

that allows the use of scenario-specific [mechanistic] source terms, 

provided there is sufficient understanding and assurance of plant and fuel 

performance and deterministic engineering judgement is used to bound 

uncertainties2. 

Footnote 2 above states: “This represents a fundamental change in practice from that used in 

LWRs, in that the source term used for siting considerations may not be that associated with a 

core melt accident.” 

• This recommendation...would allow credit to be given for the unique 

aspects of plant design and builds[sic] upon the recommendation under 

Issue 4[16]. Furthermore, this approach is consistent with prior 

Commission and ACRS views. However, this approach is also 

dependent upon understanding fuel and fission product behavior under 

a wide range of scenarios and on ensuring fuel and plant performance is 

maintained over the life of the plant. This approach is also very 

dependent on the event selection process. 

 

16 Probabilistic Event Selection, Safety Classification and Reliability Criteria. 
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The staff updated the Commission on the use of scenario-specific source terms in 

SECY-05-0006.17 Among other topics within SECY-05-0006, it specifically addressed how the 

staff proposed to integrate scenario-specific source terms into the proposed regulatory structure 

for new plant licensing. The staff proposed using a flexible, performance-based approach to 

establish scenario specific licensing source terms. The key features of the staff’s approach 

were as follows: 

• Scenarios are to be selected from a design-specific probabilistic risk assessment (PRA). 

• Source term calculations are based on verified analytical tools. 

• Source terms for compliance should be 95 percent confidence level values based on 

best estimate calculations. 

• Source terms for emergency preparedness should be mean values based on best 

estimate calculations. 

• Source terms for licensing decisions should reflect scenario-specific timing, form, and 

magnitude of the release. This approach puts the burden on the applicant to develop 

the technical basis. An applicant could, however, propose to use a conservative 

source term. 

Based on the documents listed at the beginning of this enclosure, the approach used by 

applicants for MST methodologies for SMRs and non-LWRs, should be, at a minimum, as 

consolidated in the following list: 

1. The reactor must be considered to be an advanced reactor in accordance with the 

description in the NRC’s Policy on Advanced Reactors. 

2. An MST used in siting and DBA dose analyses is defined as the result of an analysis of 

fission product release based on the amount of cladding damage, fuel damage, and core 

damage resulting from the specific accident sequences being evaluated. It is developed 

using best-estimate phenomenological models of the transport of the fission products 

from the fuel through the reactor coolant system, through all holdup volumes and 

barriers, taking into account mitigation features, resulting in release into the 

environment. The design-specific source terms for each accident category would 

constitute one component for evaluating the acceptability of the design. In developing 

an MST, the following conditions apply: 

a. The performance of the reactor and fuel under normal and off-normal conditions 

must be sufficiently well understood to permit a mechanistic analysis. Sufficient 

data should exist on the reactor and fuel performance through research, 

development, and testing programs to provide adequate confidence in the 

mechanistic approach. Deterministic engineering judgment should be used 

to bound uncertainties. 

 

17 SECY-05-0006, “Second Status Paper on the Staff’s Proposed Regulatory Structure for New Plant Licensing 

and Update on Policy Issues Related to New Plant Licensing,” dated January 7, 2005 (ADAMS Accession 

No. ML043560093). 
-7- 

b. The transport of fission products can be adequately modeled for all barriers and 

pathways to the environment, including specific consideration of containment 

design. The calculations should be as realistic as possible so that the values and 

limitations of any mechanism or barrier are not obscured. 

c. Credit may be given for the unique aspects of plant design 

(i.e., performance-based), which builds upon probabilistic event section, safety 

classification, and reliability criteria to determine the appropriate accident 

scenario and progression. 

d. The events considered in the analyses to develop the set of source terms for 

each design are selected to bound severe accidents and design-dependent 

uncertainties. Scenarios are to be selected using information from a 

design-specific PRA. 

e. Source term calculations are based on verified analytical tools. 

3. Source terms for licensing decisions should reflect scenario-specific timing, form, and 

magnitude of the release. This approach puts the burden on the applicant to develop 

the technical basis. An applicant could, however, propose to use a conservative 

source term. 

a. Source terms for compliance related to safety assessment and siting should be 

95 percent confidence level values based on best estimate calculations. 

b. By comparison, source terms for emergency preparedness should be mean 

values based on best estimate calculations. 

In discussions with SMR and non-LWR (e.g., Next Generation Nuclear Plant) designers, the 

designers indicated that they believed that the current siting regulatory framework is applicable 

to their designs. As described previously, the current regulatory approach requires the analysis 

of a large release to containment. To fulfil this requirement, the SMR or non-LWR designer 

would need to develop accident scenarios, specific to each design, using a mechanistic 

approach to developing the DBA large radiological source term to containment. 

The selection of appropriate scenarios for the purposes of developing design-specific DBA 

source terms is crucial and will require close review and coordination with the reviews of event 

classification and PRA. Furthermore, mechanistic modeling of fission product transport and 

mitigation in the core, RCS, containment, and ESFs will require adequate technical basis. The 

staff already has been evaluating similar aspects within reviews of new large LWR designs. As 

an example, the staff has previously evaluated and given approval of design-specific 

mechanistic modeling of aerosol deposition in containment for the AP600 and AP1000.18

 

18 NUREG-1512, “Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design 

Certification Docket No.52-003,” September 1998 (ADAMS Accession No. ML081020331) and NUREG1793, “Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design,” September 

2004 (ADAMS Accession Nos. ML043450344, ML043450354, ML043450284, ML043450290, and 

ML043450274). 
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Reactor Siting and Safety Analysis Regulatory Requirements 

Using MST analysis affects the application of the siting criteria contained in 10 CFR Part 100 

and the calculation of the DBA radiological consequences when comparing the results to the 

dose criteria in 10 CFR 50.34(a)(1). The regulations in 10 CFR 100.21, “Non-Seismic Siting 

Criteria,” require 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. An low-population 

zone (LPZ) immediately surrounding the exclusion area is also required.19 The number and 

density of residents within the LPZ are to 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, called the exclusion area boundary 

(EAB), and to the outer boundary of the LPZ are not prescribed by regulation, so they may vary 

from reactor site to reactor site. As a siting criterion, 10 CFR 100.21(b) requires that the 

distance to the nearest population center of at least 25,000 people is at least one and one-third 

times the distance to the outer boundary of the LPZ. 

The siting regulation in 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 site’s 

atmospheric dispersion characteristics. Specifically, the analysis described by 

10 CFR 50.34(a)(1) is a description and safety assessment of the site and a safety analysis of 

the facility, including analysis of the ESFs and those barriers that must be breached because of 

an accident that can result in a release of radioactive material to the environment, with special 

attention to plant design features that are intended to mitigate the radiological consequences of 

accidents. The regulation in 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. The evaluation 

must determine that: 

(1) An individual located at any point on the boundary of the exclusion area for 

any 2 hour period following the onset of the postulated fission product release, 

would not receive a radiation dose in excess of 25 rem total effective dose 

equivalent (TEDE). 

(2) An individual located at any point on the outer boundary of the low population 

zone, who is exposed to the radioactive cloud resulting from the postulated 

 

19 The regulations in 10 CFR 100.3, “Definitions,” provide the regulatory definitions of the exclusion area and 

low population zone, including more specific detail on the residents and population surrounding the site. 
-9- 

fission product release (during the entire period of its passage) would not receive 

a radiation dose in excess of 25 rem TEDE. 

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

as follows: 

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 given in 

10 CFR 50.34(a)(1). For DBAs other than LOCA (or bounding scenario as applicable to SMR 

and non-LWR designs), the regulatory guidance 20 is 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 radiological habitability dose criterion of 5 rem 

TEDE. 

The siting dose criteria are expected to be evaluated through DBA dose analyses on a 

per-reactor basis, even for multimodule plants. This is because of the design protection 

against external events that may affect more than one module concurrently, separation and 

independence of the modules’ systems, structures and components and safety functions, and 

design against common cause failures among modules, in accordance with GDCs 2, 4, and 5. 

This means that the siting of a multimodule plant, including the determination of the EAB, LPZ 

and population center distances, is currently expected to be based upon the evaluation of a 

single reactor. 

Implications for SMRs and Non-LWRs 

The siting regulations require that residents in a densely populated center of 25,000 people or 

more be located at minimum specified distances from the reactor relative to the LPZ outer 

boundary. For currently operating reactors and recently licensed new reactors, EABs and LPZ 

boundary distances may vary from site to site, and reactor to reactor, because the distance 

determination is partly a function of the capability to meet the siting dose acceptance criteria at 

the chosen distance. In general, large LWR EABs are around 0.8 km (0.5 miles), and the LPZs 

may be around 4.8 to 8 km (3 to 5 miles). The 10 CFR 100.21(b) requirements on reactor 

location relative to population centers result in the closest population center to the reactors lying 

 

20 NUREG-0800 (SRP) Section 15.0.3, “Design Basis Accident Radiological Consequences of Analyses for 

Advanced Light Water Reactors” (ADAMS Accession No. ML070230012). 
-10- 

outside the LPZ, but the population center could be inside the 10-mile plume exposure EPZ 

used for large LWRs. 

In addition as stated in 10 CFR 100.21(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 acceptable. 

Regulatory Guide 4.7 21 provides guidance related to location of LWR sites relative to major 

population centers. Additionally, regarding considerations of a low likelihood of severe 

accidents, the NRC has previously provided the following discussion in the Federal Register

regarding “Siting Away From Densely Populated Centers”:22

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 [quantitative 

health objectives] 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. 

Nevertheless, the Commission concludes 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. 

As is reflected in the above, the Commission has previously considered whether ESFs could 

allow for siting of reactors near densely populated cities. Further discussion of past 

metropolitan siting issues can be found in “Containing the Atom, Nuclear Regulation in a 

Changing Environment,” 1963-1971, J. Samuel Walker, Chapter 4, “Reactors Downtown? The 

Debate over Metropolitan Siting,” as well as in NUREG-0478.23 The staff anticipates examining 

this history more closely as it engages with stakeholders over the next several months. 

 

21 Regulatory Guide 4.7, Rev. 3, “General Site Suitability Criteria for Nuclear Power Stations,” March 21, 2014 

(ADAMS Accession No. ML12188A053). 22 Federal Register, 61 FR 65157, RIN 3150-AD93, “Reactor Site Criteria Including Seismic and Earthquake 

Engineering Criteria for Nuclear Power Plants,” December 11, 1996. 23 NUREG-0478, “Metropolitan Siting - A Historical Perspective,” D.F. Bunch, October 31, 1978 (ADAMS 

Accession No. ML12187A192). 
-11- 

MST Relationship to the Basis for EPZ Size Evaluations 

In SECY-11-0152,24 the staff discussed possible consideration of potential changes to the 

emergency planning and preparedness framework for SMRs. Among the considerations for 

establishing the size of EPZs for SMRs, the staff stated its expectation that the dose 

assessments that provide the basis for the EPZ distances would evaluate a spectrum of 

accidents, using the plant design PRA, as well as including current insights on severe 

accident progression. 

Subsequently in SECY-15-0077,25 the staff proposed a consequence-based approach to 

establishing requirements for offsite emergency planning, as determined to be necessary 

for SMRs and other new technologies. In the related SRM dated August 4, 2015 

(ADAMS Accession No. ML15216A492), the Commission approved the staff’s proposal to 

revise NRC regulations and guidance through rulemaking to “demonstrate how their proposed 

facilities achieve U.S. Environmental Protection Agency (EPA) Protective Action Guide (PAG)26

dose limits at specified EPZ distances, which may include the site boundary.” During the 

upcoming rulemaking process, the technical basis for the emergency planning framework for 

SMRs and non-LWRs will be developed, including guidance and criteria for radiological 

consequence analyses performed by the applicant to justify an EPZ size commensurate with the 

potential offsite radiological risk posed by the facility. 

The DBA dose analyses performed by an applicant to show compliance with the siting safety 

analysis regulations and their related radiological releases to the environment are expected to 

be included in the spectrum of analyses that form the technical basis for the EPZ distance. 

However, the EPZ size basis dose analysis would be done for the appropriate exposure period 

for comparison to the EPA PAG dose criterion of 1 rem TEDE, which is comparably less than 

the siting dose criterion of 25 rem TEDE. In this case, EABs and LPZ boundaries, which are 

located at shorter distances, may coincide with the plume exposure EPZ. Satisfying a 1 rem 

TEDE dose criterion for an EPZ of the same size as the EAB and LPZ would likely also 

demonstrate compliance by a large margin if compared with the 25 rem TEDE siting 

requirement in 10 CFR 50.34(a)(1)(ii)(D). Additionally, the technical basis for both emergency 

plans and EPZs needs to include consideration of beyond design basis accidents.

 

24 SECY-11-0152, “Development of an Emergency Planning and Preparedness Framework for Small Modular 

Reactors,” October 28, 2011 (ADAMS Accession No. ML112570439). 25 SECY-15-0077, “Options for Emergency Preparedness for Small Modular Reactors and Other New 

Technologies,” May 29, 2015 (ADAMS Accession No. ML15037A176). 26 EPA PAGs, are given in EPA-400-R-92-001, “Manual of Protective Action Guides and Protective Actions for 

Nuclear Incidents.” The PAG is defined as the dose at which public protective actions should be considered 

and undertaken.