POLICY ISSUE 

(Information) 

June 9, 2015 SECY-15-0081

FOR: The Commissioners 

FROM: Mark A. Satorius 

Executive Director for Operations 

SUBJECT: STAFF EVALUATION OF APPLICABILITY OF LESSONS LEARNED FROM 

THE FUKUSHIMA DAI-ICHI ACCIDENT TO FACILITIES OTHER THAN 

OPERATING POWER REACTORS 

PURPOSE: 

The purpose of this paper is to inform the Commission of the results of the staff’s evaluation of 

the applicability of lessons learned from the accident at the Fukushima Dai-ichi nuclear power 

plant in Japan to facilities other than operating power reactors. This paper does not address 

any new commitments or resource implications. 

SUMMARY: 

Shortly after the March 11, 2011, earthquake and tsunami that led to the accident at the 

Fukushima Dai-ichi nuclear power plant in Japan, the U.S. Nuclear Regulatory Commission 

(NRC) staff was directed to undertake near-term and longer-term actions to assess the accident 

and identify lessons learned for U.S. nuclear facilities (“Tasking Memorandum – COMGBJ-11-

0002 – NRC Actions Following the Events in Japan,” dated March 23, 2011 (Agencywide 

Documents Access and Management System (ADAMS) Accession No. ML110820875)). The 

Near-Term Task Force (NTTF) was created to review the events and the possible implications 

for the safety of U.S. nuclear power plants. On July 12, 2011, the NTTF issued its report, titled 

“Near-Term Report and Recommendations for Agency Actions Following the Events in Japan” 

(ADAMS Accession No. ML11186A950), which includes 12 recommendations. NRC staff 

prioritized the NTTF recommendations and has been actively implementing these 

recommendations through orders, requests for information, rulemaking, and other regulatory 

initiatives. 

CONTACTS: Margie A. Kotzalas, NMSS/FCSE John T. Adams, NRR/DPR 

 301-415-7298 301-415-1901
The Commissioners - 2 - 

As one of the longer-term activities, COMGBJ-11-0002 also directed the staff to assess the 

applicability of the lessons learned from the accident to non-operating reactors and non-reactor 

facilities. Very shortly after the accident, NRC staff performed limited assessments to ensure 

that no immediate safety concerns existed at these facilities and took safety measures where 

necessary, such as an independent verification of fuel facility licensees’ abilities to prevent or 

mitigate the consequences of events that could challenge the safety or licensing bases of those 

facilities. With insights gained from NRC activities related to power reactors and from the 

results of inspections at fuel cycle facilities, NRC staff has more fully evaluated issues and 

possible actions related to other NRC-licensed materials, devices, and facilities. NRC staff’s 

detailed evaluation can be found in Enclosure 1 of this paper. Based on the evaluation of 

lessons learned from the Fukushima accident, NRC staff concludes that with the exception of 

some limited scope additional analyses, there is no need for regulatory action for facilities other 

than operating power reactors in light of the accident. 

NRC staff is continuing additional assessment related to Fukushima lessons learned for 

NRC-regulated fuel facilities and a small subset of research and test reactors (RTRs). For all 

other facilities and licensees reviewed, NRC staff has determined that further assessments are 

not needed based on Fukushima lessons learned and that the existing regulatory requirements 

and processes ensure adequate protection of public health and safety. 

BACKGROUND: 

The series of events that resulted in core damage at the Fukushima Dai-ichi nuclear power plant 

was initiated by an earthquake and the resulting tsunami. The tsunami exceeded the design 

bases for the facility and inundated emergency power systems, leading to an extended loss of 

electrical power and, eventually, core damage and release of large amounts of radioactive 

material. 

In response to the Fukushima Dai-ichi accident, the Commission established the NTTF to 

conduct a systematic and methodical review of NRC processes and regulations and determine if 

the agency should make improvements to its regulatory program. COMGBJ-11-0002 stated 

that “...applicability of the lessons learned to non-operating reactor and non-reactor facilities 

should also be explored.” In SECY-11-0117, “Proposed Charter for the Longer-Term Review of 

Lessons Learned from the March 11, 2011, Japanese Earthquake and Tsunami,” dated 

August 26, 2011 (ADAMS Accession No. ML11231A886), NRC staff defined the scope as 

follows: 

The scope of the NRC’s longer-term review will include those items identified in 

the Chairman’s tasking memorandum for longer-term review, recommendations 

for evaluation that were provided by the Near-Term Task Force and are 

approved by the Commission, and any other review topics the Commission 

directs. The scope of the steering committee’s review will include power and 

non-power reactors, non-operating reactors, and non-reactor NRC licensees, and 

will be informed by interactions with external stakeholders. 

In SECY-11-0137, “Prioritization of Recommended Actions to Be Taken in Response to 

Fukushima Lessons Learned,” dated October 3, 2011 (ADAMS Accession No. ML11272A111), 

NRC staff developed longer-term plans to address the recommendations from the NTTF and 

other stakeholders. In SECY-11-0137, the staff indicated that it would continue to evaluate the 

applicability of lessons learned to licensed facilities other than power reactors (e.g., RTRs, 

independent spent fuel storage installations, and reactors that have permanently ceased 

operations but still maintain fuel in a spent fuel pool), and take appropriate actions. 
The Commissioners - 3 - 

For fuel cycle facilities, the staff issued Temporary Instruction (TI) 2600/15, “Evaluation of 

Licensee Strategies for the Prevention and/or Mitigation of Emergencies at Fuel Facilities,” 

dated September 30, 2011 (ADAMS Accession No. ML111030453). The inspection activities of 

TI 2600/15 included those licensees required by Title 10 of the Code of Federal Regulations

(10 CFR) Part 70, “Domestic Licensing of Special Nuclear Material,” to conduct an integrated 

safety analysis (ISA); those licensees under 10 CFR Part 40, “Domestic Licensing of Source 

Material,” required by their license to conduct an ISA; and currently operating certificate holders 

under 10 CFR Part 76, “Certification of Gaseous Diffusion Plants.” As described in TI 2600/15, 

the types of events NRC assessed included seismic hazards, external flooding hazards, internal 

flooding hazards, wind and tornado loading, extended loss of alternating current or emergency 

power, and fires. 

NRC staff evaluated the findings from TI 2600/15 using its normal inspection processes. As a 

result of the inspection, NRC staff identified a concern at the Honeywell Metropolis Works 

Facility. This facility converts uranium concentrate received from uranium mines to uranium 

hexafluoride. Specifically, as a result of its inspection activities, the staff found that the facility 

may not have been adequately protected from an unlikely but credible event, such as an 

earthquake or tornado. After the inspection, the licensee agreed to fortify its building and 

processing equipment. The licensee shut down its facility until NRC staff confirmed that 

appropriate upgrades had been made. In August 2013, NRC authorized the licensee to restart 

the facility following the completion of the necessary upgrades. 

With insights gained from NRC activities related to power reactors, such as seismic and flooding 

walkdowns, and the results of inspections at fuel cycle facilities, NRC staff has fully evaluated 

the issues and possible actions related to NRC licensees other than operating power reactors. 

Similar to TI 2600/15, the types of events that NRC staff assessed for these facilities included 

postulated external events, seismic hazards, external flooding hazards, internal flooding 

hazards, wind and tornado loading, extended loss of alternating current or emergency power, 

and fires, to determine if existing regulatory requirements appropriately address such hazards. 

In addition to the evaluation of initiating events and external hazards, NRC staff assessed these 

licensees qualitatively in terms of (1) policy issues related to Fukushima, (2) the NTTF’s findings 

and recommendations, and (3) other domestic and international studies and evaluations. NRC 

staff’s review was broad in scope and was not limited to specific recommendations and 

considerations provided by the NTTF, which tend to be discussed in the context of nuclear 

power reactors. 

DISCUSSION: 

NRC staff’s assessment includes a screening and evaluation of specific types of NRC licensees 

other than operating power reactors. These evaluations were completed through document 

reviews of license requirements and the regulatory framework, and internal meetings with 

subject matter experts. The extent of these evaluations depended on the nature of the licensed 

facility activities, devices, or material (e.g., amount of radioactive material, complexity in a given 

facility’s design, inherent safety, etc.). 

NRC staff factored several other evaluations and actions into its reviews. For example, as 

discussed above, fuel cycle facilities were inspected shortly after the Fukushima Dai-ichi 

accident and appropriate actions were taken to ensure adequate protection of public health and 

safety (including the multi-month shutdown of the Honeywell Metropolis Works Facility to allow 

for upgrades to the facility). Another example of such an evaluation is the NRC staff’s 

assessment of the safety of RTRs in Pacific coastal States in light of the tsunami forecasted to 

impact the Pacific coast of the United States as a result of the March 11, 2011, earthquake off 

the coast of Japan. As part of that evaluation, NRC staff concluded that none of the five RTRs 

in the Pacific States were vulnerable to a tsunami reaching the locations of the facilities. In 
The Commissioners - 4 - 

cases where NRC staff performed previous evaluations, they are discussed in the 

facility-specific assessments supplied in Enclosure 1 of this paper. 

NRC staff’s assessments address the following general areas: 

• spent fuel storage and transportation 

• fuel facilities 

• radioactive material users 

• irradiators licensed in accordance with 10 CFR Part 36, “Licenses and radiation safety 

requirements for irradiators” 

• low-level waste disposal facilities 

• uranium recovery facilities and uranium mill tailings 

• decommissioning reactors and complex materials facilities 

• non-power reactors (e.g., RTRs) 

For each of these areas NRC staff assessment provides the following: 

• the current regulatory framework for the area 

• relevant pre-Fukushima assessments 

• insights from experience gained from international activities, if applicable 

• an assessment of the facilities’ capabilities to address or mitigate the following events: 

flood, seismic, high wind and missiles, lightening, snow and ice loads, temperature 

extremes, fire, and loss of power 

• an assessment of each of the 12 NTTF recommendations for each of the areas 

Results of Stakeholder Interactions: 

NRC staff provided a draft white paper to stakeholders for their review and comment, which 

contained much of the staff’s assessment found in Enclosure 1 of this paper. As appropriate, 

changes have been made to the NRC staff’s assessment found in Enclosure 1 as a result of 

comments received from these stakeholders. The stakeholder comments were gathered in two 

ways: 

• For classes of licensees that are regulated by both Agreement States and the NRC, 

NRC staff solicited comments on the NRC staff’s draft white paper through a letter to 

Agreement States dated February 20, 2015 (ADAMS Accession No. ML15050A430 

(nonpublic)). The February 20, 2015, letter included the sections of the NRC staff’s draft 

white paper relevant to the Agreement States. This included sections of the NRC staff’s 

draft white paper associated with: (1) radioactive material users, (2) irradiators licensed 

in accordance with 10 CFR Part 36, (3) low-level waste disposal facilities, (4) uranium 

recovery and uranium mill tailings, and (5) decommissioned reactors and complex 

materials facilities. 

• For licensees and other external stakeholders, the staff solicited comments through a 

March 13, 2015, Category 3 public meeting1

 on the topic. The NRC staff’s draft white 

paper was supplied as part of the notice for the public meeting and stakeholders had the 

opportunity to offer written comments through an email address associated with the 

meeting or by verbal comments at the meeting. The NRC staff’s draft white paper 

 

1

 Category 3 public meetings are fully engaged discussions between the NRC and the public (as well as 

stakeholders that might include other government agencies, the industry and others). Public participation 

is actively sought at this type of meeting, which has the widest participation opportunities and is 

specifically tailored for the public to comment or ask questions. 
The Commissioners - 5 - 

provided to support the March 13, 2015, public meeting included the same material 

provided in the February 20, 2015, letter to the Agreement States, plus the following 

three additional draft assessments: (1) spent fuel storage and transportation, (2) fuel 

facilities, and (3) NRC-licensed non-power reactors. 

Agreement States’ Comments

The Commonwealth of Virginia and the Organization of Agreement States (OAS) provided 

comments on the NRC staff’s draft white paper in letters dated March 19, 2015 (ADAMS 

Accession No. ML15083A303), and March 23, 2015 (ADAMS Accession No. ML15083A281), 

respectively. The two letters generally provided the same comments. The major comment was 

that the Commonwealth of Virginia and the OAS Board agreed with NRC staff’s conclusion that 

no further study or regulatory action is warranted beyond those already being implemented as 

part of the regulatory oversight of the licensees discussed in the February 20, 2015, draft white 

paper. Other comments provided in the two letters included proposed technical clarifications 

and requests for more explicit statements recognizing the role that the Agreement States have 

in regulating some of the facilities that are within the scope of the NRC staff’s assessment. 

NRC staff generally incorporated the comments provided in the two letters and adjusted the 

assessment found in Enclosure 1 of this document. A more detailed discussion of the NRC 

staff’s resolution of the Agreement State comments can be found in an NRC letter to OAS dated 

April 3, 2015 (ADAMS Accession No. ML15091A169). 

Comments Associated with March 13, 2015, Public Meeting

Enclosures 2 and 3 of this document provide a listing of the verbal and written comments 

received, respectively, as a result of the March 13, 2015, public meeting. The meeting 

summary was issued on April 13, 2015 (ADAMS Accession No. ML15077A410). 

Enclosure 2 documents NRC staff’s response to the 41 verbal comments provided during the 

public meeting. The comments were generally focused on the spent fuel storage and 

transportation and decommissioning portions of the NRC staff’s draft white paper. Many of the 

comments involved the independent spent fuel storage installations found at those power 

reactors that recently permanently ceased operation and the decommissioning of the same 

reactors. As a result of these comments, NRC staff added discussions to the spent fuel storage 

and decommissioning portions of the Enclosure 1 assessment to provide additional basis for its 

conclusions in these areas. 

NRC staff also addressed 14 sets of email comments provided as a result of the public meeting. 

Enclosure 3 provides a summary of these comments and the NRC staff’s response to each. As 

documented in Enclosure 3, NRC staff considered comments on the NRC staff’s draft white 

paper from the licensees for the Massachusetts Institute of Technology research reactor 

(MITR), the University of Missouri at Columbia research reactor (MURR), and the National 

Bureau of Standards and Technology test reactor (NBSR). The comments from these licensees 

mainly involved clarifications of the various RTR attributes and capabilities to respond to 

accidents. Other email comments included concerns about the NRC staff’s assessment of 

spent fuel storage at power reactors that have ceased operations and the NRC staff’s 

assessment of other decommissioned facilities. NRC staff made changes to the Enclosure 1 

assessment to address these comments, as appropriate. 

Summary of Results: 

For the eight classes of licensees discussed in the enclosure, the NRC staff’s overall conclusion 

is that existing regulatory processes provide for adequate protection of public health and safety. 
The Commissioners - 6 - 

NRC staff’s evaluation found two areas where work will continue to ensure that Fukushima 

lessons learned are appropriately considered: 

• NRC staff will continue its efforts to resolve concerns with fuel facilities’ safety 

assessments and the supporting documentation with respect to the treatment of natural 

phenomena hazards. NRC staff expects to issue a generic letter on this subject in 2015. 

• For the three non-power reactors that have thermal power ratings in excess of 

2 Megawatt thermal (MWt), NRC staff is performing further assessments to determine if 

regulatory actions are necessary to mitigate certain beyond-design-basis external 

events. 

Fuel Cycle Facility Assessment

As discussed in SECY-11-0137 and SECY-13-0095, “Fourth 6-Month Status Update on 

Response to Lessons Learned from Japan’s March 11, 2011, Great Tohoku Earthquake and 

Subsequent Tsunami” (ADAMS Accession No, ML13213A304), NRC staff completed 

inspections at selected fuel facilities and the results were used to perform a systematic 

evaluation of the processes and regulations applicable to fuel facilities. As a result of this 

evaluation, NRC staff concludes that the current regulatory approach and requirements of these 

licenses continue to serve as a basis for reasonable assurance of adequate protection of public 

health and safety. However, as described in greater detail in Enclosure 1 of this document, for 

the Honeywell Metropolis Works Facility, the inspections found potentially significant safety 

issues and NRC staff took immediate steps to ensure corrective actions were taken. In addition, 

NRC staff found generic issues regarding compliance with the current regulatory framework with 

regard to the treatment of certain natural phenomena events in the facilities’ (uranium 

conversion, enrichment, and fuel fabrication) safety assessments. As discussed in 

SECY-15-0045, “Issuance of Generic Letter 2015-01, ‘Treatment of Natural Phenomena 

Hazards in Fuel Cycle Facilities,’” dated March 27, 2015 (ADAMS Accession No. 

ML15036A141), NRC staff is developing a generic letter to request information from licensees 

that was not readily available during the inspections, to verify that compliance is being 

maintained with regulatory requirements and license conditions regarding the treatment of 

natural phenomena hazards. 

Research and Test Reactor Assessment

NRC staff’s review associated with the three highest-powered RTRs (MITR, MURR, and NBSR) 

identified the need for additional limited-scope assessments. For these three reactors, NRC 

staff will perform additional assessments regarding the reactors’ capabilities to prevent or 

mitigate loss of coolant accidents as a result of beyond-design-basis natural phenomena (e.g., 

seismic events). This is because for these RTRs, the early loss of reactor coolant can result in 

failure of the fuel cladding and subsequent radiological release unless reactor coolant makeup 

can be provided from installed facility equipment or from portable external sources. 

Both the MITR and MURR reactors are tank type reactors. Because of their low power, natural 

convection flow of reactor coolant is sufficient to remove decay heat from these reactors and 

prevent bulk boiling, even in the event of a loss of all electrical power and active decay heat 

removal systems. Therefore, there is not a near-term need to replenish the water around the 

reactor fuel lost by evaporation. However, if the initiating external event also causes (or occurs 

concurrently with) the failure of the core tank and the loss of the reactor pool integrity, then the 

resulting loss of coolant inventory could result in inadequate decay heat removal and fuel 

damage. 
The Commissioners - 7 - 

NBSR is the highest-powered RTR licensed by the NRC at 20 MWt. In the case of NBSR, if 

there is an extended loss of electrical power to operate the active decay heat removal systems 

or damage to the active decay heat removal system that prevents its use, fuel damage could 

occur unless the decay heat removal systems are restored or reactor coolant makeup using 

portable external sources can be initiated to make up for evaporation and boil-off. As such, in 

addition to performing additional assessments regarding NBSR’s capabilities to prevent or 

mitigate loss of coolant accidents as a result of beyond-design-basis natural phenomenon (e.g., 

seismic events), NRC staff will also assess the NBSR’s capabilities to prevent or mitigate loss of 

decay heat removal capabilities due to an extended loss of alternating current power (e.g., that 

could result from an extreme flood or other extreme natural phenomenon). 

The additional assessments for MURR, MITR, and NBSR will be performed by NRC staff with 

expertise in the areas of seismology, hydrology, structural analysis, and RTR design and 

operation. These assessments are expected to be limited in nature, will use existing budgeted 

resources, and will be prioritized considering other high-priority work that is being performed by 

these staff. Upon completion, NRC staff will provide a Commissioners’ Assistants Note 

presenting the results of the assessments and, if appropriate, a discussion of the work 

necessary to resolve any remaining concerns at these facilities. 

CONCLUSION: 

For the majority of regulated facilities discussed in this paper, NRC staff has determined that no 

other analysis or regulatory action is needed based on its review of the lessons learned from the 

Fukushima Dai-ichi accident. If responses to the fuel facility generic letter or the further 

assessments of the three highest-powered RTRs cause NRC staff to determine that additional 

regulatory actions are needed to address Fukushima lessons learned, then staff will interact 

with stakeholders and, as appropriate, engage the Commission. 

 /RA/ 

Mark A. Satorius 

Executive Director 

 for Operations 

Enclosures: 

1. Applicability of Fukushima Lessons Learned 

to Facilities Other Than Operating Power Reactors 

2. Discussion of Public Comments Received During 

March 13, 2015, Meeting 

3. Discussion of Email Comments Associated with 

Staff’s Draft Assessment 
Enclosure 1 

i 

APPLICABILITY OF FUKUSHIMA LESSONS LEARNED TO FACILITIES OTHER THAN 

OPERATING POWER REACTORS 

Table of Contents

Acronyms ...................................................................................................................................... ii

1. Spent Fuel Storage and Transportation Systems ............................................................. 1

2. Fuel Facilities .................................................................................................................... 13

3. Radioactive Material Users ............................................................................................... 22

4. Part 36 Irradiators ............................................................................................................. 32

5. Low-Level Waste Disposal Facilities ................................................................................ 42

6. Uranium Recovery Facilities and Uranium Mill Tailings .................................................... 52

7. Decommissioning Reactors and Complex Materials Facilities ......................................... 61

8. NRC-Licensed Non-Power Reactors ................................................................................ 81
ii 

Acronyms 

ac alternating current 

ACI American Concrete Institute 

ADAMS Agencywide Documents Access and Management System 

ALARA as low as is reasonably achievable 

AMP aging management program 

AV Apparent Violation 

BWR boiling water reactor 

CFR Code of Federal Regulations 

DBA design basis accident 

DOE Department of Energy 

DOT Department of Transportation 

DP decommissioning plan 

EP emergency plan 

EPA Environmental Protection Agency 

EPZ Emergency Planning Zone 

FBI Federal Bureau of Investigation 

FEMA Federal Emergency Management Agency 

FEP facility effective power 

GTCM greater than critical mass 

IAEA International Atomic Energy Agency 

IMC Inspection Manual Chapter 

IN Information Notice 

ISFSI Independent Spent Fuel Storage Installation 

ISR in situ recovery 

LEU low-enriched uranium 

LIP local intense precipitation 

LLW low-level waste 

LOCA loss of coolant accident 

LTP license termination plan 

LRWPA Low-level Radioactive Waste Policy Amendments Act of 1985 

MHA maximum hypothetical accident 

MIT Massachussetts Institute of Technology 

MOX mixed oxide 

MURR University of Missouri at Columbia research reactor 

MWt megawatts-thermal 

NBSR National Institute of Standards and Technology test reactor 

NEA Nuclear Energy Agency 

NPH natural phenomena hazard 

NPR nonpower reactors 

NTTF Near Term Task Force 

NRC Nuclear Regulatory Commission 

NSTS National Source Tracking System 

NYSERDA New York State Energy Research and Development Authority 

PMF probable maximum flood 

PSDAR post-shutdown decommissioning activities report 

RASSC Radiation Safety Standard Committee
iii 

Acronyms (continued) 

RWMC Radiation Waste Management Committee 

RTR research and test reactors 

SAR Safety Analysis Report 

SBO station blackout 

SSCs structures, systems, and components 

TI temporary instruction 

TLAA time-limited aging analysis 

UF6 uranium hexaflouride 

UFSAR Updated Final Safety Analysis Report 

USEC United States Enrichment Corporation 

USGS United States Geological Survey 

URI Unresolved Item 

WASSC Waste Safety Standards Committee 

WVDP West Valley Demonstration Project 

WVDPA West Valley Demonstration Project Act of 1980 
1 

1. Spent Fuel Storage and Transportation Systems 

I. Current Regulatory Framework 

Radiation exposure is the main hazard associated with spent fuel storage facilities, systems, 

and transportation packages. These facilities, systems, and packages have multiple 

components and use specific materials of construction to protect workers, the public, and the 

environment. The U.S. Nuclear Regulatory Commission’s (NRC’s) responsibility is to review, 

approve, and inspect the design of storage facilities, systems, and transportation packages to 

ensure the safe and secure storage and transport of spent nuclear fuel. The U.S. Department 

of Transportation (DOT) is the United States’ competent authority for defining requirements for 

radioactive materials in transit. 

The following sections summarize the existing regulatory framework to ensure that radioactive 

materials under NRC’s authority are stored and transported safely and securely. 

Licensing

The regulatory infrastructure for licensing spent fuel storage (Title 10 of the Code of Federal 

Regulations (10 CFR) Part 72)1

 and transportation package designs (10 CFR Part 71)2 allows 

for issuing certificates of compliance, general licenses, or specific licenses. Certificates of 

compliance are granted to vendors for a particular storage cask or transportation system design. 

The following sections describe the NRC’s licensing requirements for the safe and secure 

storage and transportation of spent nuclear fuel. 

a. Storage of Spent Nuclear Fuel

The NRC’s responsibility is to review and approve the design of storage systems and 

facilities for spent nuclear fuel and radioactive material in accordance with applicable 

regulations and guidance. Dry storage systems perform the safety functions of ensuring 

subcriticality, and confining and shielding the radioactive contents during the period of 

storage. For the design of the dry cask storage systems, vendors are required to 

consider natural phenomena hazards (NPHs) based on the 10 CFR Part 72 regulations. 

Vendors perform bounding analyses to verify the adequacy of the dry cask storage 

systems to comply with the pertinent regulations. Upon review of the applicant’s safety 

analysis report, if appropriate safety and security margins are maintained, the results of 

the review are documented in a safety evaluation report. Then, the NRC issues a 

certificate of compliance accompanied with technical specifications to the cask vendor. 

A utility that has an existing 10 CFR Part 50 or Part 52 license can have a general or 

specific license to store spent nuclear fuel (i.e., an independent spent fuel storage 

installation). For a general license, the utility is responsible for verifying that the design 

parameters of the selected cask system bound the NPHs that may occur at their site. A 

 

1

 10 CFR Part 72, “Licensing requirements for the independent storage of spent nuclear fuel and highlevel radioactive waste, and reactor- related greater than Class C waste.” 2

 10 CFR Part 71, “Packaging and transportation of radioactive material.”
2 

specific license is required for any entity that does not already hold a 10 CFR Part 50 or 

10 CFR Part 52 license or for any utility with site-specific parameter(s) that are not 

bounded by the certified vendor’s designs. The applicants that apply for a specific 

license must verify the adequacy of the cask design to withstand the specified NPHs for 

the given site. 

The United States has 55 general licensees and 15 specific licensees of independent 

spent fuel storage installations. The two types of independent spent fuel storage 

installations are dry storage or wet storage (i.e., spent nuclear fuel pool). There is only 

one wet storage facility licensed under 10 CFR Part 72 in the United States, G.E. Morris, 

and its fuel is more than 20 years old.3

 

Initial cask certifications and specific licenses were issued for a 20-year term. In 2011, 

NRC revised the requirements in 10 CFR Part 72 to allow for initial license and 

certification periods of up to 40 years. The regulations also allow for 40-year renewals of 

both the specific license and cask certifications. The requirement for renewal of specific 

licenses is contained in 10 CFR 72.42, “Duration of license; renewal.” This regulation 

specifies that the renewal application include: 

(1) Time-limited aging analyses (TLAAs) that demonstrate that structures, 

systems, and components important to safety will continue to perform their 

intended function for the requested period of extended operation; and 

(2) A description of the aging management program (AMP) for management of 

issues associated with aging that could adversely affect structures, systems, 

and components important to safety. 

The requirements for storage cask certification renewals are contained in 10 CFR 

72.240, “Conditions for spent fuel storage cask renewal.” Analyses of aging effects and 

an associated aging management program are also required for cask certification 

renewal. 

NRC staff reviews renewal applications to ensure the independent spent fuel storage 

installation will continue to provide safe and secure storage of spent nuclear fuel 

throughout the renewal period. The performance of the storage system in both the initial 

and renewal terms must meet all regulatory requirements. This evaluation considered 

storage casks whether in their initial term or a renewed license or certification term. 

b. Transportation of Spent Nuclear Fuel

The NRC’s regulatory authority for the transport of radioactive material is delineated in 

the memorandum of understanding with DOT4

 and 10 CFR 71.5, “Transportation of 

licensed material.” DOT is the competent authority for all transportation matters in the 

 

3

 NRC Inspection Report 072-00001/11-01, dated March 13, 2011. 

4

 Memorandum of Understanding between the U.S. Department of Transportation and the U.S. Nuclear 

Regulatory Commission published on July 2, 1979; 44 FR 38690. 
3 

United States. The NRC’s responsibility is to ensure that packages for fissile materials 

and Type B quantities5

 are designed, manufactured, and handled safely, as well as 

transported securely. Packages approved by the NRC shall comply with 10 CFR Part 71 

and can be used at any place in the United States. 

Domestically, NRC supports DOT on the review of packages that would be used in the 

United States that have been approved in other countries for transporting radioactive 

material. This review process is called revalidation. During the revalidation process, the 

NRC determines if the package complies with the International Atomic Energy Agency 

(IAEA) regulatory requirements for transportation of radioactive materials.6

 

Internationally, NRC participates in forums related to the transportation and storage of 

radioactive materials. The NRC also assesses if changes should be made to 10 CFR 

Part 71, when IAEA regulations are revised. 

Inspection and Oversight

NRC staff inspects certificate holders and independent spent fuel storage installations. NRC 

staff inspects certificate holders to ensure compliance with quality assurance requirements for 

manufacturing and handling storage system or package designs. The staff also inspects the 

operational aspects of the independent spent fuel storage installations and spent fuel cask 

loading campaigns. 

Emergency Preparedness

a. Storage of Spent Nuclear Fuel

Contents of an emergency plan (EP) for spent fuel storage facilities are delineated in 

10 CFR 72.32. Per 10 CFR 72.32, an application for a specific-license for an 

independent spent fuel storage installation must include an EP if the facility is: 

1. not located: 

i. on the site of a nuclear power reactor, or 

ii. within the exclusion area as defined in 10 CFR Part 100, “Reactor site 

criteria,” of a nuclear power reactor, or 

2. located on the site of a nuclear power reactor: 

i. which does not have an operating license, or 

ii. that is not authorized to operate. 

 

5

 A Type B quantity is defined in 10 CFR 71.4, “Definitions.” 

6

 TS-R-1, “Regulations for the Safe Transport of Radioactive Material, 2009 Edition Safety 

Requirements.” 
4 

For independent spent fuel storage installations located on the site or within the 

exclusion area of a nuclear power reactor licensed for operation, 10 CFR 72.32(c) 

specifies that “the EP required by 10 CFR 50.47 shall be deemed to satisfy” the 10 CFR 

72.32 requirements for an EP. There are no current regulatory requirements for 

establishing emergency planning zones in the EP for independent spent fuel storage 

installations. 

b. Transportation of Spent Nuclear Fuel

As previously mentioned, the NRC and the DOT share responsibility for regulating the 

transportation of radioactive materials. The NRC is responsible for ensuring that escorts 

are properly trained to respond to events as specified in 10 CFR Part 73, Appendix D,7

including severe weather conditions and security events. The DOT defines 

requirements for the shippers of radioactive materials. Regulatory requirements for 

shippers related to emergency response are specified in 49 CFR Part 172. 

Security

Existing NRC regulations include physical security requirements for facilities with a general or 

specific license for an independent spent fuel storage installation. The NRC and the DOT share 

regulatory responsibility for the safe transport of shipments. 

a. Storage of Spent Nuclear Fuel

Licensees with either a specific or general license for an independent spent fuel storage 

installation must maintain their security systems in an operable condition. This includes 

measures to maintain security systems available during power losses per 10 CFR 73.51 

and 10 CFR 73.55. Generally-licensed independent spent fuel storage installation 

security requirements are found in 10 CFR 72.212 and 10 CFR 73.55. The security 

requirements for a specific license of independent spent fuel storage installations are 

found in 10 CFR 72.180–194 and in 10 CFR 73.51. 

After the terrorist attacks of September 11, 2001, the NRC issued security orders to all 

spent nuclear fuel storage facilities to impose greater security measures beyond the 

requirements in the NRC regulations at the time. The NRC continues to issue security 

orders to independent spent fuel storage installations and is revising 10 CFR Part 73 to 

codify the requirements of the security orders. 

b. Transportation of Spent Nuclear Fuel

NRC’s security requirements for spent nuclear fuel in transit include: 

1. training requirements for escorts 

 

7

 10 CFR Part 73, “Physical protection of plants and materials,” Appendix D, “Physical protection of 

irradiated reactor fuel in transit, training program subject schedule.” 
5 

2. review and approval of shipment routes of spent nuclear fuel 8 

3. requirements for personnel accessing the shipment9

 

II. Post-Fukushima Event Evaluations/Assessment 

NRC staff evaluated both transportation systems and storage designs to determine the effect of 

NPHs and the applicability of the Near Term Task Force (NTTF) recommendations. For the 

transportation assessment, the staff used NUREG-2125, “Spent Fuel Transportation Risk 

Assessment,”10 published in January 2014, as the basis for the evaluation. 

Also, NRC staff participated in a series of table top exercises to evaluate the effect of selected 

NPHs to existing storage designs. NRC staff did not evaluate site-specific designs, because the 

main design characteristics11 of storage systems were the same for general and specific 

licensees. NRC staff’s expertise included knowledge of 10 CFR Part 72 requirements to 

evaluate the adequacy of the storage design in the following areas: 

• confinement evaluation 

• criticality safety evaluation 

• materials evaluation 

• shielding evaluation 

• structural evaluation 

• thermal evaluation 

The staff divided its qualitative assessment into the following phases: 

Phase 1. Determine initial applicability of NTTF recommendations to spent fuel 

storage and transportation certificate holders and licensees. 

Phase 2. Determine commonalities among the design characteristics of the 

storage systems. (Note: This phase did not apply to the NRC staff’s 

assessment of transportation of spent fuel). 

Phase 3. Determine qualitative magnitude of consequences of NPHs and 

external events. (Rely on NUREG-2125 as basis for the transportation 

qualitative assessment.) 

Phase 4. Document findings, determine path forward, and develop input. 

NRC staff’s qualitative assessment is based on a single canister and assumes that all spent fuel 

systems and operations are in compliance with 10 CFR Part 72 regulations. Loading activities 

in the pool are performed in accordance with the licensee’s 10 CFR Part 50 (or 10 CFR Part 52) 

safety programs. Final closure and on-site transfer are performed in accordance with technical 

specifications and site procedures. For welded canister-based systems, the transfer cask 

 

8

 10 CFR 73.37, “Requirements for physical protection of irradiated reactor fuel in transit” 

9

 10 CFR Part 73.38, “Personnel access authorization requirements for irradiated reactor fuel in transit” 10 ADAMS Accession No. ML14031A323. 

11I.e., wet storage or dry storage, horizontal or vertical system orientation, and welded or bolted closure. 
6 

provides additional protection to the storage canister during on-site transfer, while for directly 

loaded systems, the cask is fully prepared for storage before on-site transfer. 

a. Phase 1. Determine initial applicability of NTTF recommendations to spent 

fuel storage and transportation certificate holders and/or licensees.

The NTTF report, “Recommendations for Enhancing Reactor Safety in the 21st Century,” has 12 

recommendations. NRC staff considered these recommendations for applicability to 

independent spent fuel storage installations and the transport of spent nuclear fuel and 

evaluated the recommendations to determine if any future action is warranted. The following 

table summarizes the result of the review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed.12 No regulatory gaps were 

identified as a result of the staff’s initial review. 

The decision was confirmed after additional 

assessment (discussed below) was completed. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of structures, systems and 

components (SSCs). 

No Action Needed. Based on initial review and 

confirmed by bounding analysis discussed 

below, the staff did not identify a need for 

additional analysis in this area. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. As discussed below, 

ISFSIs are made of material that is not 

flammable, and bounding flooding analysis 

shows no need for additional analysis. 

Moreover, combustible loads are currently 

minimized at ISFSIs. For the one wet storage 

facility in the U.S., the stored fuel is more than 

20 years old. This wet storage facility is a 

below-ground reinforced concrete pool with a 

welded stainless steel liner. Based on a review 

of multiple external scenarios (as discussed 

below), NRC staff determined that fuel melt is 

not credible as a result of these scenarios. 

4 The task force recommends that the 

NRC strengthen station blackout 

(SBO) mitigation capability at all 

operating and new reactors for 

design-basis and beyond-designNo Action Needed. ISFSIs remove decay heat 

by passive mechanisms. The use of power is 

limited to monitoring systems. Loss of power 

will not lead to fuel damage. The fuel at the 

existing wet storage facility is more than 20 

 

12 No Action. No further study or regulatory action by NRC staff is warranted.
7 

Recommendations Review Result 

basis external events. years old and does not have a near term need 

for active decay heat removal systems. 

5 The task force recommends requiring 

reliable hardened vent designs in 

boiling-water reactor (BWR) facilities 

with Mark I and Mark II containments. 

Not Applicable.13 There are no credible 

scenarios that create hydrogen in quantities of 

concern for spent fuel storage and 

transportation systems. 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identify insights about hydrogen 

control and mitigation inside 

containment or in other buildings as 

more information is revealed through 

further study of the Fukushima Daiichi accident. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for spent fuel storage and 

transportation systems. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

No Action Needed. The fuel at the U.S. wet 

storage facility is more than 20 years old and 

does not have a near term need for active 

decay heat removal systems or near term 

makeup. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs [emergency operating 

procedures], SAMGs [severe accident 

management guidelines], and EDMGs 

[extensive damage mitigation 

guidelines]. 

No Action Needed. Fukushima lessons learned 

for power reactors showed prolonged loss of 

AC power and multiunit events could have 

severe consequences leading to a need to 

update EP response. Prolonged loss of AC 

power or multiunit event is not a concern for 

passively cooled ISFSIs or the existing wet 

storage facility with fuel more than 20 years old. 

No need to upgrade existing EP requirements 

for spent storage and transportation systems 

was identified. 

9 The task force recommends that the 

NRC require that facility EPs address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 8 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

No Action Needed. See item 8 above. 

 

13 Not applicable. The NTTF recommendation for commercial nuclear power reactors does not apply to 

this “other regulated facility.”
8 

Recommendations Review Result 

to decision-making, radiation 

monitoring, and public education. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s initial review. 

The decision was confirmed after additional 

assessment (discussed below) was completed. 

b. Phase 2. Determine commonalities among the design characteristics of the 

storage systems (Note: this phase did not apply to NRC staff’s assessment of 

transportation of spent fuel) 

Based on the current regulatory framework for licensing, independent spent fuel storage 

installations were categorized into two major groups, general licenses and specific 

licenses.14 NRC staff further grouped the casks systems into different design 

characteristics to simplify the qualitative assessment. Design characteristics included: 

1. orientation on the independent spent fuel storage installation pads (i.e., vertical, 

horizontal, or underground) 

2. closure type (i.e., bolted or welded)15

3. vents on overpack (i.e., vented or not vented) 

After grouping the storage systems, NRC staff qualitatively evaluated the 

effects of selected natural phenomena and external events. 

c. Phase 3. Determine qualitative magnitude of consequences of NPHs and 

external events.

NRC staff qualitatively evaluated the radiological consequences of the natural 

phenomena and external events. The magnitudes of the consequences were 

defined as follows: 

1. Low – No radiation-related deaths or injuries expected; no offsite contamination. 

2. Medium – Few radiation-related deaths or injuries; little to no offsite 

contamination. 

 

14 Each type of licensee uses a dry cask storage system to place the fuel removed from the spent fuel 

pool. Then, these casks systems are placed on the independent spent fuel storage installation concrete 

pad. 

15 Vertical systems only. 
9 

3. High – Significant radiation-related deaths or injuries and significant offsite 

contamination or property damage. 

NRC staff’s initial screening found that natural phenomena such as lightning, snow 

and ice loads, external fire, extreme temperature, and drought were either 

determined to be not applicable or low to no consequence events. Dry cask 

storage systems are passive and designed for the heat loads associated with 

spent nuclear fuel. Natural phenomena of extreme cold or heat (low or high 

ambient temperatures) will not impact the safety performance of the storage 

system such that confinement of the system would fail. Similarly, lightning and fire 

would not negatively impact the safety performance of the storage system. The 

storage system is comprised of materials that are not flammable (concrete and 

steel). The presence of flammable materials, like diesel fuel, is limited by technical 

specification around the dry storage system, however, the temperatures 

associated with fire would not impact the containment function of the storage 

system. Drought is not applicable because there is no use of water in dry storage. 

NRC staff performed more in-depth assessment for the following natural 

phenomena and external events for dry storage cask systems: 

1. seismic 

2. flooding 

3. high winds (e.g., hurricanes), 

4. tornado winds and tornado missiles, and 

5. loss of off-site power 

The regulations in 10 CFR Part 72 require the applicants to assess natural phenomena 

events as part of the safety basis for a storage facility or container design. For this 

assessment, NRC staff considered the potential impacts of natural phenomena that 

would be beyond the design basis evaluation that would be required to meet 10 CFR 

Part 72. Based on the qualitative evaluation, NRC staff determined that these events 

also would result in low radiological consequences. The following summarizes the NRC 

staff’s qualitative evaluation: 

1. Seismic – Per 10 CFR Part 72.10216 and 10 CFR Part 72.10317 the design-basis 

earthquake for dry cask storage systems varies depending on the user’s location 

of the storage cask design. Nevertheless, for a storage system, the 

consequences of an earthquake are bounded by a hypothetical event called 

“non-mechanistic tip-over.” Dry casks storage systems are designed to 

demonstrate structural adequacy under this hypothetical scenario. This event is 

independent of any natural phenomena events, and the structural integrity of the 

 

16 72.102, “Geological and seismological characteristics for applications before October 16, 2003, and 

applications for other than dry cask modes of storage.” 

17 72.103, “Geological and seismological characteristics for applications for dry cask modes of storage on 

or after October 16, 2003.” 
10 

fuel and the cask body is verified for their adequacy to withstand the potential of 

a cask tipping over and hitting the concrete supporting pad.18 

2. Flood – NRC staff considered scenarios such as a (1) storage system fully 

submerged in water, (2) partial flooding around the storage system, and (3) 

blockage of the overpack vents for those canister-based systems with natural 

convection cooling vented overpacks. The worst case scenario for loss of heat 

transfer is a partial flood around the storage system. This scenario would not 

cause a radioactive release to the environment because the temperature would 

not exceed allowable temperatures for the materials of construction for system 

confinement. 

For the third scenario, NRC staff assumed that all vents of an overpack system 

were blocked with debris, the debris was not removed, and the temperature of 

the fuel increased (i.e., cooling capability of the system decreased). The results 

of the steady-state thermal analysis for a vertical cask system and a horizontal 

cask system indicate that a release from the confinement is not expected to 

occur. The heat load in the cask is at its highest when initially placed into 

storage and the system cools overtime. NRC staff considered the highest 

allowable heat loads in the cask system for this evaluation. 

3. High Winds – For this scenario, NRC staff considered hurricane force winds. 

Regulatory Guide 1.221, “Design-Basis Hurricane and Hurricane Missiles for 

Nuclear Power Plants,” and NUREG/CR-7005, “Technical Basis for Regulatory 

Guidance on Design-Basis Hurricane Wind Speeds for Nuclear Power Plants,” 

are NRC’s guidance used for addressing hurricane winds in the design of a 

nuclear power plant. Most of the independent spent fuel storage installations are 

collocated with a nuclear power plant; therefore, the same guidance applies. The 

hurricane wind speeds in NUREG/CR-7005 were based on a model developed 

for the American Society of Civil Engineers.19

Based on NUREG/CR-7005, the resulting wind speeds are nominal 3-second 

peak-gust values at a height of 33 feet in flat open terrain. Using Regulatory 

Guide 1.221, the wind speed may vary from 130 mph (Texas) to 290 mph 

(Southeastern Atlantic Coastline) depending on the independent spent fuel 

storage installation’s location. Missiles generated by hurricanes were not 

considered in this portion of the assessment, because the tornado assessment 

would bound high winds generated-missile events. Given the size and weight of 

spent fuel storage systems, it is not anticipated that high winds will displace or 

overturn the cask; however, the worst case in a beyond design basis event due 

to high winds would be bounded by the “non-mechanistic tip-over” which is 

described above for the seismic evaluation. 

 

18 Independent spent fuel storage installation pad. 19 ASCE/SEI 7-05, “Minimum Design Loads for Buildings and Other Structures.” 
11 

4. Tornado Winds and Tornado Missiles – Dry cask storage systems’ structures, 

systems, and components that are important to safety must be designed to 

withstand the effects of natural phenomena hazards without losing the capability 

to perform their safety function. NRC Regulatory Guide 1.76, “Design-Basis 

Tornado and Tornado Missiles for Nuclear Power Plants,” Revision 1, March 

2007, offers guidance that NRC staff considers acceptable in selecting the 

design-basis tornado and tornado-generated missiles for a nuclear power plant. 

Vendors of dry cask storage systems use these criteria to comply with the 

10 CFR Part 72 requirements. 

During the qualitative assessment, NRC staff considered the potential for 

damage to the cask confinement boundary for the bolted system due to a beyond 

design basis tornado missile event. Given the robustness of the spent fuel cask 

design, this would not result in a significant release or increase in the radiation 

levels in the vicinity of the cask. The consequence of this event is categorized as 

low since there would be no off-site contamination. 

5. Loss of Off-Site Power – The surveillance systems at the independent spent 

fuel storage installations that would need power to operate are pressure sensors 

and security systems. During the qualitative assessment, NRC staff found that 

the loss of power of pressure sensors would not cause a release of radioactive 

material. Requirements of security surveillance systems are included in 10 CFR 

Part 73 as discussed above. 

There is only one wet storage independent spent fuel storage installation in the United 

States (i.e., G.E. Morris) and the stored fuel is more than 20 years old.20 The G.E. Morris 

facility is a below-ground reinforced-concrete pool with a welded stainless steel liner. This 

facility was inspected in March 2011,21 after the event at Fukushima. NRC inspectors 

reviewed scenarios, such as station blackout, seismic, tornado, flood, and fire. NRC staff 

found that the fuel would not melt as a result of these events because of its limited heat 

load. 

NRC staff considered the findings and conclusions documented in NUREG-2125, 

“Spent Fuel Transportation Risk Assessment” (ADAMS Accession No. 

ML12125A218), in evaluating the consequence of the NPHs with respect to the 

transport of spent nuclear fuel. NUREG-2125, published in January 2014, 

estimated the risk associated with the transportation of spent nuclear fuel by 

examining the behavior of three NRC-certified casks during routine transportation 

and in transportation accidents. NRC staff’s qualitative evaluation focused on 

comparison of the impact and fire accident scenarios in NUREG-2125 to the NPH. 

Flooding is not an accident scenario considered in NUREG-2125 because of the 

regulatory requirement for spent fuel transportation packages to be designed and 

constructed and the contents limited such that it would be subcritical if water were 

to leak into the containment system such that maximum reactivity of the fissile 

 

20 NRC Inspection Report 072-00001/11-01, dated March 13, 2011. 21 NRC Inspection Report 072-00001/11-01, dated March 13, 2011. 
12 

material would be attained (10 CFR 71.55, “General requirements for fissile 

material packages”). Therefore, flood is already an analyzed condition for spent 

fuel transportation packaging. The staff concluded that the NPHs are bounded by 

the accident scenarios evaluated in NUREG-2125 and would have low 

consequence. 

III. International Experience 

Since the March 11, 2011, seismic and tsunami events at the Fukushima Dai-ichi nuclear power 

plants in Japan, NRC has been actively involved in information exchanges with the international 

community at conferences and forums such as the Structural Mechanics in Reactor Technology 

and IAEA. NRC staff has coordinated and chaired technical sessions to address issues related 

to the safe storage and transportation of spent nuclear fuel as well as presented papers at these 

forums. The insights gained during these technical exchanges have been used, as applicable, 

for facilities within the United States. NRC staff has followed the progress by IAEA on the 

Action Plan on Nuclear Safety (GOV/2011/59-GC(55)/14) and the results of the IAEA gap 

analysis, as it relates to spent fuel storage and transportation. NRC staff will continue to 

participate on working groups such as the IAEA safety guide SSG-15: Storage of Spent Nuclear 

Fuel and SSR-6: Regulations for the Safe Transport of Radioactive Material, are evaluated for 

revision. Any enhancements that are identified in these IAEA documents will be considered for 

applicability to the NRC’s regulations for spent fuel storage and transportation. 

IV. Conclusions and Recommendations 

NRC staff found that the NRC’s existing regulatory framework ensures the safe and secure 

storage and transportation designs for radioactive material licensed by the NRC. Based on the 

qualitative assessment, NRC staff did not find safety concerns associated with the designs of 

spent fuel storage and transportation systems. 

The defense-in-depth philosophy is embedded in 10 CFR Part 71 and 10 CFR Part 72 

regulations. NRC staff ensures that designs proposed by applicants, licensees, and certificate 

holders consider industry standards and include layers of protection to maintain doses within the 

regulatory limits. The systems are built and handled under an NRC approved quality assurance 

program. In term of oversight of NRC approved systems, NRC staff has guidance to inspect 

and schedule inspections to ensure that storage facilities, storage systems, and transportation 

systems are built, and operate, as specified in the design approved by the NRC.22 Therefore, 

NRC staff concludes that no further regulatory action or study is necessary. 

 

 

22Inspection Manual Chapter 2690, “Inspection Program for Dry Storage of Spent Reactor Fuel at 

Independent Spent Fuel Storage Installations and for 10 CFR Part 71 Transportation Packagings.” 
13 

2. Fuel Facilities 

I. Overview of Fuel Cycle Facilities, Academic and Other Institutions 

Fuel cycle facilities involved in conversion, enrichment, and fuel fabrication are regulated 

through a combination of regulatory requirements; licensing; safety oversight (including 

inspection, assessment of performance, and enforcement); evaluation of operational 

experience; and regulatory support activities. These facilities turn the uranium that has been 

removed from ore (as yellowcake) into fuel for nuclear reactors. In this process, the conversion 

facility converts yellowcake into uranium hexafluoride (UF6). Next, an enrichment facility heats 

the solid UF6 enough to turn it into a gas, which is “enriched,” or processed to increase the 

concentration of the uranium-235 (U235). Then the enriched uranium is manufactured into 

pellets. These pellets are placed into fuel assemblies and ultimately into nuclear reactors. 

Fuel Cycle Process, Facilities, and Associated Hazards

The regulations governing the licensing and operation of fuel cycle facilities are found below. 

Each type of facility presents different levels of risk to both the worker and offsite from events 

involving chemical and radiological material. 

• 10 CFR Part 20, “Standards for protection against radiation” 

• 10 CFR Part 21, “Reporting of defects and noncompliance” 

• 10 CFR Part 40, “Domestic licensing of source material” 

• 10 CFR Part 70, “Domestic licensing of special nuclear material” 

• 10 CFR Part 73, “Physical protection of plants and materials” 

• 10 CFR Part 74, “Material control and accounting of special nuclear material” 

• 10 CFR Part 76, “Certification of gaseous diffusion plants” 

A. Uranium Conversion 

Process: After the yellowcake is produced at the mill, the next step is conversion into pure UF6

gas suitable for use in enrichment operations. During this conversion, impurities are removed 

and the uranium is combined with fluorine to create the UF6 gas. The UF6 is then pressurized 

and cooled to a liquid. In its liquid state it is drained into 14-ton cylinders where it solidifies after 

cooling for approximately five days. The UF6 cylinder, in the solid form, is then shipped to an 

enrichment plant. 

Hazards: As with mining and milling, the primary risks associated with conversion are potential 

chemical and radiological events. Strong acids and alkalis are used in the conversion process, 

which involves converting the yellowcake (uranium oxide) powder to very soluble forms, leading 

to possible inhalation of uranium. In addition, conversion produces extremely corrosive 

chemicals that could cause fire and explosion hazards. Fire and explosion hazards are also a 

concern in areas where liquid UF6 is stored and processed. When liquid UF6 is released to the 

atmosphere, it reacts with the moisture in the air to form a dense vapor cloud that contains 

hydrogen fluoride gas (chemical hazard), a nonradioactive, extremely toxic substance. 

Plants: Honeywell International Inc. Metropolis, Illinois. 
14 

B. Uranium Enrichment

Process: Enriched uranium is required in commercial light-water reactors to produce a 

controlled nuclear reaction. Enriching uranium increases the proportion of uranium atoms that 

can be “split” by fission to release energy (usually in the form of heat) that can be used to 

produce electricity. Not all uranium atoms are the same. When uranium is mined, it consists of 

about 99.3 percent uranium-238 (U238), 0.7 percent uranium-235 (U235), and less than 0.01 

percent uranium-234 (U234). The fuel for nuclear reactors has to have a higher concentration of 

U235 than exists in natural uranium ore. Normally, the amount of the U235 isotope is enriched 

from 0.7 percent of the uranium mass to about 5 percent. 

Hazards: The principal hazards at an enrichment plant are the chemical hazards in handling 

UF6. When UF6 contacts moisture in air, it reacts to form hydrogen fluoride and uranyl fluoride. 

The chemical hazards of compounds of uranium in soluble form such as UF6 and uranyl fluoride 

are much greater than the radiological hazards of those same compounds. In addition, 

hydrogen fluoride can be very dangerous if inhaled; inhalation is the principal hazard at an 

enrichment plant. These hazards are controlled by plant design and administrative controls to 

confine soluble uranium compounds. The radiological hazards are relatively low and containers 

of natural, enriched, and depleted uranium can be handled without extra shielding. Another 

hazard for this type of facility is the potential for mishandling the enriched uranium, which could 

create a criticality accident (inadvertent nuclear chain reaction). 

Plants: 

• Gaseous Diffusion Uranium Enrichment Facility 

− United States Enrichment Corporation (USEC) Inc. in Paducah, Kentucky (No 

longer a NRC-licensed facility) 

• Gas Centrifuge Uranium Enrichment Facilities 

− American Centrifuge Plant, LLC (USEC) in Piketon, OH (License issued, 

construction halted) 

− Louisiana Energy Services in Eunice, NM 

− AREVA Enrichment Services Eagle Rock, LLC , Idaho Falls, ID (License issued, 

construction not started) 

• Laser Separation Enrichment Facility 

− GE-Hitachi in Wilmington, NC (License issued, construction not started) 

C. Uranium Fuel Fabrication

Process: Fuel fabrication facilities convert enriched UF6 into fuel for nuclear reactors. 

Fabrication also can involve mixed oxide (MOX) fuel, which is a combination of uranium and 

plutonium components. The NRC regulates several different types of nuclear fuel fabrication 

operations, such as light water reactor low-enriched uranium fuel and light water reactor mixed 

oxide fuel. 
15 

Fuel fabrication for light water power reactors typically begins with receipt of low-enriched 

uranium (LEU) hexafluoride from an enrichment plant. The UF6, in solid form in containers, is 

heated to gaseous form, and the UF6 gas is chemically processed to form LEU dioxide (UO2) 

powder. This powder is then pressed into pellets, sintered into ceramic form, loaded into 

zircaloy tubes, and constructed into fuel assemblies. Depending on the type of light water 

reactor, a fuel assembly may contain up to 264 fuel rods and have dimensions of 5 to 9 inches 

square by about 12 feet long. 

MOX fuel differs from LEU fuel in that the dioxide powder from which the fuel pellets are 

pressed is a combination of UO2 and plutonium oxide (PuO2). Congress directed NRC to 

regulate the Department of Energy's (DOE's) fabrication of MOX fuel, which uses repurposed 

plutonium from international nuclear disarmament agreements. 

Hazards: Chemical, radiological, and criticality hazards at fuel fabrication facilities are similar to 

hazards at enrichment plants. 

Plants: 

• Uranium Fuel Fabrication 

− Global Nuclear Fuel-Americas, LLC in Wilmington, NC 

− Westinghouse Electric Company, LLC in Columbia, SC 

− Nuclear Fuel Services, Inc. in Erwin, TN 

− AREVA NP, Inc. in Richland, WA 

− Babcock & Wilcox Nuclear Operations in Lynchburg, VA 

• Mixed Oxide Fuel Fabrication 

− Shaw AREVA MOX Services, LLC in Aiken, SC (Construction Authorization 

issued, construction ongoing) 

D. Uranium Hexafluoride Deconversion 

Process: As U235 is extracted, converted, and enriched in the uranium recovery, conversion, 

and enrichment processes for use in fabricating fuel for nuclear reactors, large quantities of 

depleted uranium hexafluoride (DUF6), or “tailings,” are produced. These tailings are 

transferred into 14-ton cylinders which are stored in large yards near the enrichment facilities. A 

process called "deconversion" is then used to chemically extract the fluoride from the DUF6 

stored in the cylinders. This deconversion process produces stable compounds, known as 

uranium oxides, which are generally suitable for disposal as low-level radioactive waste. 

Hazards: Chemical exposure is the dominant hazard at deconversion facilities because 

uranium chemical compounds and other chemical compounds (such as hydrogen fluoride) are 

hazardous at low levels of exposure. 

Plant: International Isotopes in Hobbs, NM (license issued, construction not started) 
16 

E. Academic and Other Institutions

Academic and other institutions use radioactive material in classroom demonstrations, 

laboratory experiments, and research, and to offer health physics support to other institutional 

radioactive materials users. These facilities are licensed in accordance with 10 CFR Parts 30, 

40, or 70 depending on the type of materials possessed. These programs may vary in size from 

large, broad-scope programs involving chemical, physical, biological engineering, and 

biomedical research, to small programs using only gas chromatographs or self-shielded 

irradiators. 

Hazards: These licensees have limited amounts of materials and have demonstrated by 

NRC-approved evaluation that no member of the public will exceed the thresholds of the 

regulations in 10 CFR Part 70. That is not to say that accidents cannot occur with these 

licensees, but because of the limited amount of materials possessed and the primarily sealed 

nature of the material, the effect to the public and the environment is limited. 

II. Post-Fukushima Event Evaluations/Assessment 

NRC staff evaluated and inspected selected fuel cycle facilities to confirm that licensees 

complied with regulatory requirements and license conditions; and to evaluate their readiness 

under NPH events and other licensing bases events related to NPHs. NRC staff’s assessment 

considered the NTTF recommendations to determine whether other regulatory actions by the 

NRC are warranted. This assessment included consideration of new seismic hazards 

information from the U.S. Geological Survey (USGS) for the central and eastern United States 

which was the subject of an NRC generic communication to fuel facilities in NRC Information 

Notice 2010-19, “Updated Probabilistic Seismic Hazard Estimates in Central Eastern United 

States” (ADAMS Accession No. ML102160735). 

As discussed in SECY-11-0137 and SECY-13-0095, NRC staff completed inspections at 

selected fuel facilities and the results were used to perform a systematic evaluation of the 

processes and regulations applicable to fuel facilities. Because of the evaluation, NRC staff 

concludes that the current regulatory approach and requirements of these licensees continues 

to serve as a basis for reasonable assurance of adequate protection of public health and safety. 

However, as described in greater detail below, for the Honeywell Metropolis Works Facility, the 

inspections found potentially significant safety issues and the NRC staff took immediate steps to 

ensure corrective actions were taken. In addition, NRC staff found generic issues regarding 

compliance with the current regulatory framework with regards to the treatment of certain 

natural phenomena events in the facilities’ (uranium conversion, enrichment and fuel fabrication) 

safety assessments. As discussed in SECY-15-0045, “Issuance of Generic Letter 2015-01, 

‘Treatment of Natural Phenomena Hazards in Fuel Cycle Facilities,’” dated March 27, 2015 

(ADAMS Accession No. ML15036A141), NRC staff is developing a generic letter to request 

information from licensees that was not readily available during the inspections, to verify that 

compliance is being maintained with regulatory requirements and license conditions regarding 

the treatment of NPHs. 
17 

A. Uranium Conversion, Enrichment, Fuel Fabrication, and Deconversion Licensees

On March 31, 2011, NRC staff issued Information Notice (IN) 2011-08, “Tohoku–Taiheiyou–Oki 

Earthquake Effects on Japanese Nuclear Power Plants—for Fuel Cycle Facilities,” (ADAMS 

Accession No. ML110830824) to inform addressees of the potential challenges associated with 

preventing or mitigating the effects of natural phenomena events. IN 2011-08 recommended 

that addressees review the information for applicability to their facilities and consider actions, as 

appropriate, to ensure that features and preparations necessary to withstand or respond to 

severe external events from natural phenomena (e.g., earthquakes, tsunamis, floods, 

tornadoes, and hurricanes) are reasonable and consistent with regulatory requirements. 

From December 2011 through May 2012, NRC staff conducted inspection activities in 

accordance with Temporary Instruction (TI) 2600/015, “Evaluation of Licensee Strategies for the 

Prevention and/or Mitigation of Emergencies at Fuel Facilities” (ADAMS Accession No. 

ML12286A284). The NRC completed the TI in three phases. In the initial phase, NRC staff 

reviewed licensing documents, including the safety assessments and EPs. The second phase 

consisted of NRC inspectors evaluating accident prevention measures and emergency actions 

through onsite evaluations that focused on credible natural phenomena and loss of utilities that 

support onsite systems (e.g. electricity and water). The third phase involved assessing whether 

the strategies and equipment were effective to prevent or mitigate emergencies during selected 

beyond licensing basis natural events and the extended loss of utilities that support onsite 

systems. In the review of licensing basis events, the NRC considered the following NPHs: 

seismic, flooding, and high winds (caused by hurricanes or tornadoes). The NRC also 

evaluated onsite fires that may result from seismic related equipment failures. Particular 

attention was given to earthquakes and flooding because of recent events and significant 

advancements in the state of knowledge of these hazards. 

In addition, during the TI 2600/015 inspections, the staff also considered operating experience 

on the implementation of mitigation strategies and emergency procedures used by licensees to 

cope with natural phenomena events. The staff performed interviews of licensee personnel and 

walk-downs of the facility to assess how the facility coped with the occurrence of previous 

natural phenomena events (such as floods or hurricanes). 

After the implementation of TI 2600/015, the NRC determined that the evaluated facilities had 

established programs, procedures, and equipment to respond to licensing basis events 

involving fire, flooding, and loss of utilities. However, based on information obtained from the 

inspection activities, NRC staff found that the assumptions used by licensees in developing the 

independent safety analysis and other safety assessments are not clearly described and 

documented. The NRC primarily attributed this to the lack of available facility design information 

and significant variations in the level of detail and rigor of implementation in the facility safety 

assessments with regards to the treatment of natural phenomena events. Therefore, the NRC 

inspectors were unable to verify that these facilities complied with their licensing basis and 

regulatory requirements. The NRC inspectors opened unresolved items (URIs) to further 

assess whether the evaluated licensees complied with license conditions, the requirements of 

10 CFR 70.61 and 10 CFR 70.62(c), regarding NPH event accident sequences. A summary of 

the results of the inspections can be found in the table below. NRC staff has determined that for 

all the facilities inspected (except the Honeywell Metropolis Works Facility) in consideration of 
18 

inherent seismic capacity in facility structures, existing safety programs in place, and 

radiological/chemical source terms, continued operation does not pose an imminent risk to 

public health and safety. 

For the Honeywell Metropolis Works Facility, the NRC determined that the site Emergency 

Response Plan underestimated the amount of UF6 and hydrogen fluoride that could potentially 

be released during credible seismic events or tornadoes. Specifically, the inspectors found that 

the process equipment in the licensee’s Feed Materials Building lacked seismic restraints, 

supports, and bracing that would ensure process equipment integrity during certain credible 

seismic events or tornadoes. The NRC issued a confirmatory order that required the licensee to 

demonstrate its SSCs relied on for safety were adequate for seismic and tornado events. The 

facility structure and internal components were significantly retrofitted to improve the 

performance under seismic and tornado events. Additional information is available in NUREG0090, Volume 36, “Report to Congress on Abnormal Occurrences: Fiscal Year 2013” (ADAMS 

Accession No. ML14150A073). 

Summary of TI 2600/015 results 

Facility Summary of Issues Identified

Part 76 

Paducah 

(ADAMS Accession 

No. ML12131A437) 

• Tornadoes were not considered a credible event because of 

the return period chosen for the evaluation basis event. The 

team determined that a tornado could be considered a 

credible event for the site if newer data is used to evaluate 

the probability of occurrence. However, the consequences of 

a tornado event are bounded by other safety basis events. 

Part 70 

AREVA 

(ADAMS Accession 

No. ML12122A094) 

• Unresolved Item (URI) 70-1257/2012-006-01 was opened to 

further evaluate whether the licensee complied with the 

requirements of 10 CFR 70.62(c) and 70.61 performance 

requirements regarding natural phenomena events accident 

sequences. 

Babcock & Wilcox 

Nuclear Operations 

Group 

(ADAMS Accession 

No. ML12121A574)

• URI 70-27/2012-006-01 was opened to further evaluate 

whether the licensee complied with the requirements of 10 

CFR 70.62(c) and 70.61 performance requirements 

regarding natural phenomena events accident sequences. 

Global Nuclear Fuel −

Americas 

(ADAMS Accession 

No. ML12209A276)

• URI 70-113/2012-006-01 was opened to further evaluate 

whether the licensee complied with the requirements of 10 

CFR 70.62(c) and 70.61 performance requirements 

regarding natural phenomena events accident sequences. 

Nuclear Fuel Services 

(ADAMS Accession 

No. ML12122A186)

• URI 70-143/2012-06-01 was opened to further evaluate 

whether the licensee complied with Table 2.2 of the license 

application regarding management measures for items relied 

on for safety PREP-A and PREP-B. 

• URI 70-143/2012-006-03 was opened to further evaluate 

whether the licensee complied with the requirements of 10 

CFR 70.62(c) and 70.61 performance requirements 

regarding natural phenomena events accident sequences. 
19 

Facility Summary of Issues Identified

Westinghouse −

Columbia Fuels 

(ADAMS Accession 

No. ML12122A083)

• URI 70-1151/2011-07-01 was opened to review 

Westinghouse’s response to the failure to ensure that the risk 

of an earthquake was limited by applying sufficient 

engineered controls, administrative controls, or both, to the 

extent needed so that, upon implementation of such controls, 

the event was highly unlikely. 

• URI 70-1151/2011-07-02 was opened to review 

Westinghouse’s evaluation regarding whether all nuclear 

process under an earthquake were subcritical. 

Part 40 

Honeywell 

(ADAMS Accession 

No. ML12222A163)

• URI 40-3392/2012-006-01 was opened to evaluate whether 

the Metropolis Works Facility integrated safety analysis 

appropriately considered credible high consequence seismic 

and tornado events and subsequently designated plant 

features and procedures and management measures to 

ensure that the accident sequences (public and workers 

health and safety) remained highly unlikely or the 

consequences were mitigated to acceptable levels. 

• Apparent Violation (AV) 40-3392/2012-006-02 was identified 

for the failure to identify all relevant accident sequences 

related to credible seismic events and tornadoes that could 

result in large uranium hexafluoride (UF6) releases for which 

protective actions may be needed. 

• AV 40-3392/2012-006-003 was identified for the failure to 

supply complete and accurate information related to MTW’s 

emergency response plan. 

Note: Louisiana Energy Services was not inspected because it is a new facility designed and constructed 

with more demanding criteria as required by 10 CFR 70.64, “Baseline Design Criteria”. 

Review of Near-Term Task Force recommendations

NRC staff considered the 12 NTTF recommendations for applicability to fuel cycle facilities and 

evaluated to determine if any future action is warranted. The following table summarizes the 

result of the review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establish a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 
20 

Recommendations Review Result 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Immediate Action. Based on lessons 

learned from TI 2600/15 inspections staff 

identified a need to verify compliance regarding 

treatment of natural phenomena events (see 

Generic Letter discussion below). 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. Staff assessment noted 

inherent safety margin in fuel cycle facilities and 

NRC staff assessment considered new seismic 

hazards information from USGS which was 

provided to fuel facilities by NRC Information 

Notice 2010-19. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

No Action Needed. Staff assessment 

considered the inherent safety margin in fuel 

cycle facilities, including but not limited to, 

existing safety programs in place and 

radiological and chemical source terms to 

determine that SBO mitigation capabilities do 

not need to be strengthened beyond those that 

already exist. 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

Not Applicable. There are no credible scenarios 

that create hydrogen in quantities of concern for 

these licensees. 

6 The task force recommends, as part 

of the longer term review, that the 

NRC points out insights about 

hydrogen control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

through further study of the 

Fukushima Dai-ichi accident. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

Not Applicable to fuel cycle facilities. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. Fukushima lessons learned 

for power reactors showed prolonged loss of 

AC power and multiunit events could have 

severe consequences leading to a need to 

update EP response. No need to upgrade 

existing EP requirements based on inherent 

safety margin in fuel cycle facilities, including 

but not limited to, existing safety programs in 

place, and considering radiological and 

chemical source terms associated with fuel 
21 

Recommendations Review Result 

cycle facilities. 

9 The task force recommends that the 

NRC require that facility EPs address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 8 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 8 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

Generic Letter: Treatment of Natural Phenomena Hazards in Fuel Cycle Facilities

Because of inspections, and as discussed in SECY-15-0045, NRC staff is proposing issuing a 

generic letter due to the generic applicability of the URIs across the nuclear fuel facility industry. 

Current NRC regulations require the evaluation of site hazards including natural phenomena 

events. The purpose of the generic letter is to request information from licensees to verify that 

compliance is being maintained with regulatory requirements and license conditions regarding 

the treatment of natural phenomena events. 

NRC staff maintains engagement with industry and other stakeholders through public meetings 

and presentations during the annual Fuel Cycle Information Exchange. NRC staff plans to issue 

the generic letter in 2015 and expects to close out the review of the responses in 2016. 

B. Academic and Other Institutions (Greater Than Critical Mass)

One of the considerations coming out of the fuel cycle review described above was to evaluate 

the readiness of greater than critical mass (GTCM) licensees as it applies to the lessons from 

the Fukushima accident. The readiness of GTCM facilities was evaluated by performing 

desktop reviews (ADAMS Accession No. ML14111A087). The recommendations of the NTTF 

were reviewed for applicability, as well as considerations found during the review of fuel cycle 

facilities. On the basis of this review, the continued operation of GTCM licensees do not pose 

an imminent risk to the public health and safety. The current regulatory approach and 
22 

requirements of these licensees continues to serve as a basis for reasonable assurance of 

adequate protection of public health and safety. For these facilities, compliance with the current 

regulatory framework ensures that the consequences of any potential natural phenomena event 

are low to offsite receptors and workers. 

III. International Experience 

NRC staff has been actively involved in International forums with regards to discussions and 

lessons from the accident as it relates to fuel facilities. For example, NRC staff participated in 

an IAEA consultancy and technical meeting to develop a standard on the assessment of fuel 

cycle facilities in light of the Fukushima Dai-ichi accident. The standard aims to provide 

practical information and experiences on the performance of assessments of fuel cycle facilities. 

In addition, NRC staff, in coordination with the Office of International Programs, met with the 

Director and Principal Inspector of the Decommissioning, Fuel, and Waste Program (Office of 

Nuclear Regulation) from the United Kingdom to broaden cooperation with the United States in 

areas related to the fuel cycle, storage, disposal, decommissioning, and activities associated 

with the post-Fukushima event for fuel cycle facilities. The insights and lessons learned 

gathered through interactions with the international community were used during the NRC staff 

evaluation of NRC regulated fuel cycle facilities. 

IV. Conclusions and Recommendations 

Because of the systematic and methodical evaluation of fuel facilities and GTCM licensees in 

light of the lessons learned from Fukushima, NRC staff concludes that the current regulatory 

approach and requirements for fuel cycle licensees offers reasonable assurance of adequate 

protection of public health and safety. For these facilities, compliance with the current 

regulatory framework ensures that the consequences of any potential natural phenomena event 

to offsite receptors and workers are within the dose limits stipulated by the regulatory 

requirements. NRC staff will continue its efforts to resolve concerns with fuel facilities’ safety 

assessments and the supporting documentation with respect to the treatment of NPHs. NRC 

staff expects to issue a generic letter on this subject in 2015. 

3. Radioactive Material Users 

I. Current Regulatory Framework 

The NRC regulates about 2,900 research, medical, industrial, government, and academic 

materials licensees. In addition, the NRC has agreements with 37 States, under which the 

States assume regulatory responsibility for the use of certain radioactive materials. These 

Agreement States oversee about 18,000 licensees. The quantities that medical and academic 

licensees possess can range from millicurie quantities of radionuclides to thousands of curies 

contained in self-shielded irradiators and gamma knife devices. Industrial uses of sealed source 

devices include a variety of applications and devices that include density gauges, thickness 

gauges, prompt gamma neutron activation analysis gauges, well logging gauges, moisture 

density gauges, industrial radiography sources, irradiators, as well as others. 
23 

Safety evaluations for sources and devices used in industrial, academic and medical settings 

are mostly evaluated as part of the sealed source and device registration process. Licensees 

are authorized to possess and use only those sealed sources and devices specifically approved 

or registered by NRC or an Agreement State. The NRC or Agreement State evaluate the safety 

of gauges, radiography source assemblies, exposure devices, source changers, and welllogging sources before authorizing a manufacturer or distributor to distribute the gauges to 

specific licensees. The safety evaluation is documented in a sealed source and device 

registration certificate. 

Sealed sources are required to satisfy rigorous design and performance criteria. To prevent 

accidental dispersion from the device into the environment, the licensed material should be a 

chemical and physical form that is as insoluble and non-dispersible as practical. The material is 

doubly encapsulated; the capsule should be resistant to extreme changes in temperature, 

pressure, and vibration; and it is resistant to impact and puncture. The evaluation of the 

sources and devices include a review of the design, manufacturing, prototype testing, and 

proposed uses. The sources and devices are designed to survive normal conditions of use and 

likely accidental conditions. 

Self-shielded irradiators (e.g., blood irradiators) incorporate many engineering features to 

protect individuals from unnecessary radiation exposure. Many other devices (e.g., gamma 

knife, radiography cameras, gauges, etc.) have some engineering features (shielding, 

connectors, switches, etc.) that prevent unnecessary exposure. These devices are usually 

designed for use in an industrial or laboratory environment, i.e., inside a building, protected from 

the weather, and in many cases without wide variations in temperature and humidity. 

Facilities and equipment must be adequate to protect health, minimize danger to life or property, 

minimize the possibility of contamination, and keep exposure to occupationally exposed workers 

and the public as low as is reasonably achievable (ALARA). Licensed materials located in an 

unrestricted area and not in storage must be under the constant surveillance and immediate 

control of the licensee. Areas where material is used or stored, including below ground bunker 

storage areas, should (1) be accessible only by authorized persons, and (2) secured or locked 

when an authorized person is not physically present. 

NRC regulations applicable to devices containing byproduct material and to NRC radiation 

safety evaluations are found in the following: 

• 10 CFR Part 20, “Standards for protection against radiation” 

• 10 CFR Part 21, “Reporting of defects and noncompliance” 

• 10 CFR Part 30, “Rules of general applicability to domestic licensing of byproduct material” 

• 10 CFR Part 31, “General domestic licenses for byproduct material” 
24 

• 10 CFR Part 32, “Specific domestic licenses to manufacture or transfer certain items 

containing byproduct material” 

• 10 CFR Part 33, “Specific domestic licenses of broad scope for byproduct material” 

• 10 CFR Part 34, “Licenses for industrial radiography and radiation safety requirements for 

industrial radiographic operations” 

• 10 CFR Part 35, “Medical use of byproduct material” 

• 10 CFR Part 37, “Physical protection of category 1 and category 2 quantities of radioactive 

material”23

• 10 CFR Part 40, “Domestic licensing of source material” 

• 10 CFR Part 70, “Domestic licensing of special nuclear material” (for pacemaker devices) 

• 10 CFR Part 71, “Packaging and transportation of radioactive material” 

• 10 CFR Part 150, “Exemptions and continued regulatory authority in Agreement States and 

in offshore waters under section 274” 

Licensees are responsible for the security and safe use of all licensed material (sealed and 

unsealed) from the time it arrives at their facility, during its use and storage, and until it is 

transferred or disposed. The licensee should be able to account for the location of all materials 

possessed, whether the material is located in a secured laboratory cabinet, a locked refrigerator 

or freezer, or in an appropriate waste container awaiting disposal. Security for category 1 and 2 

sources (e.g., gamma knife, blood irradiators, etc.) is now described in 10 CFR Part 37, 

“Physical protection of category 1 and category 2 quantities of radioactive material.” This 

regulation establishes security requirements for the use and transport of the most risk-significant 

quantities of radioactive materials. 

 

23 Category 1 sources and practices - personally extremely dangerous: This amount of radioactive 

material, if not safely managed or securely protected, would be likely to cause permanent injury to a 

person who handled it, or were otherwise in contact with it, for more than a few minutes. It would 

probably be fatal to be close to this amount of unshielded material for a period of a few minutes to an 

hour (see IAEA-TECDOC-1344, Categorization of radioactive sources, 2003 for more details). 

Category 2 sources and practices - personally very dangerous: This amount of radioactive material, if not 

safely managed or securely protected, could cause permanent injury to a person who handled it, or were 

otherwise in contact with it, for a short time (minutes to hours). It could possibly be fatal to be close to this 

amount of unshielded radioactive material for a period of hours to days. 
25 

Licensing

The NRC’s NUREG-1556 technical report series, “Consolidated Guidance about Materials 

Licenses,” offers a comprehensive source of reference information about various aspects of 

materials licensing and materials program implementation. These reports, where applicable, 

describe a risk-informed, performance-based approach to licensing consistent with the current 

regulations. Specific guidance is found in the following: 

Volume 1 - Program-Specific Guidance About Portable Gauge Licenses 

Volume 2 - Program-Specific Guidance About Industrial Radiography Licenses 

Volume 3 - Applications for Sealed Source and Device Evaluation and Registration 

Volume 4 - Program-Specific Guidance About Fixed Gauge Licenses 

Volume 5 - Program-Specific Guidance About Self-Shielded Irradiator Licenses 

Volume 7 - Program-Specific Guidance about Academic, Research and Development, and 

Other Licenses of Limited Scope Including Gas Chromatographs and X-Ray Fluorescence 

Analyzers 

Volume 11 - Program-Specific Guidance about Licenses of Broad Scope 

Volume 14 - Program-Specific Guidance About Well Logging, Tracer, and Field Flood Study 

Licenses 

Most of these volumes have been updated and published in the Federal Register for public 

comment. The rest of the volumes are being updated and will be published in the Federal 

Register for public comment by the end of fiscal year 2016. As applicable, additional guidance 

on source security and safety culture have been included. 

Possession limits must cover the total anticipated inventory, including licensed material in 

storage and waste. If the type, form, and amounts of any of the materials possessed exceed 

those for Category 1 and Category 2 sources (e.g., cesium-137 or cobalt-60), they must be 

reported to, and tracked in, the National Source Tracking System (NSTS) in accordance with 

10 CFR 20.2207, “Reports of transactions involving nationally tracked sources.” Such sources 

also may have other requirements for security of these materials under 10 CFR Part 37, 

“Physical protection of category 1 and category 2 quantities of radioactive material.” Refer to 

Appendix E, “Nationally Tracked Source Thresholds,” to 10 CFR Part 20, “Standards for 

protection against radiation,” for a list of radionuclides of interest and Category 1 and 2 

quantities. This regulation establishes security requirements for the use and transport of the 

most risk-significant quantities of radioactive materials. 

Gamma well-logging sources often contain Category 2 amounts of cesium-137 while neutron oil 

well-logging sources contain Category 2 amounts of americium and beryllium. Radiography 

exposure devices also contain Category 2 amounts of cobalt-60 or iridium-192. Academic and 

most medical licensees possess small quantities of unsealed source material, far less than 
26 

Category 2 quantities. However, hospitals may possess blood irradiators, teletherapy devices, 

gamma knives, or brachytherapy sources that contain Category 1 or 2 quantities. 

Inspection and Oversight

For the fabrication and use of sealed sources and devices used by radioactive material users, 

the NRC ensures compliance with the regulatory requirements through both periodically 

scheduled and special inspections. Manual Chapter 2800, “Materials Inspection Program,” 

establishes the inspection program for licensees authorized to possess, use, transfer, and 

dispose of radioactive material. Inspections are conducted in accordance with formalized 

procedures, specifically, Inspection Procedure No. 87125, “Materials Processor/Manufacturer 

Programs.” The objective of the procedure is to assure that (1) licensed activities are being 

conducted in a manner that will protect the health and safety of workers and the general public, 

(2) the licensed programs are being conducted in accordance with NRC requirements, and (3) 

the licensee is manufacturing sources or devices in accordance with commitments made to 

NRC. 

Emergency Preparedness

Facility design and available equipment must supply sufficient engineering controls and barriers 

to protect the health and safety of the public and licensee employees, keep exposures to 

radiation and radioactive materials ALARA, and minimize the danger to life and property from 

the uses of the types and quantities of radioactive materials possessed by the licensee. The 

licensee should have and follow emergency and operational event procedures, appropriate for 

the sealed or unsealed material possessed, and for natural phenomena (if required), including 

an earthquake, a tornado, hurricane, flooding, or other phenomena as appropriate for the 

geographical location of the facility. 

Loss or theft of licensed material, sabotage, fires, floods, etc., can adversely affect the safety of 

workers and members of the public. Therefore, written procedures must be developed to 

minimize, as much as possible, the effect of these incidents on workers, members of the public, 

and the environment. Licensees who possess radioactive materials in unsealed form, on foils or 

plated sources, or sealed in glass in excess of the quantities in 10 CFR Part 30.72, 

“Schedule C—Quantities of Radioactive Materials Requiring Consideration of the Need for an 

Emergency Plan for Responding to a Release,” must either (1) demonstrate that the maximum 

dose to a person offsite due to a release of radioactive materials will not exceed 0.01 Sv (1 rem) 

effective dose equivalent or 0.05 Gy (5 rem) to the thyroid; or (2) develop an EP for responding 

to a release of radioactive material. That plans should include a facility description, means and 

equipment for mitigating the consequences of each type of accident, methods and equipment to 

assess releases of radioactive material responsibilities of licensee personnel, a description of 

the means to promptly notify offsite response organizations, a description of the means of 

restoring the facility to a safe condition, and provisions for biennial onsite exercise. The 

emergency preparedness plan (or contingency plan) at nuclear pharmacy manufacturing 

facilities such as Mallinckrodt address how the facility staff will respond to major spills, fire, 

tornado, and earthquake. Worst case scenario assessments assume complete release of the 

entire radioactive inventory of affected buildings and assess the offsite total effective dose 

equivalent and thyroid organ exposure to members of the public. At present, most medical and 

academic licensees do not possess quantities of radioactive materials that exceed the values 
27 

described in 10 CFR 30.72, Schedule C, and are not required to develop an EP. 

 

Licensees are required to develop and follow operating and emergency procedures. Each year, 

some gauges experience equipment failures because of corrosion caused by the harsh 

environmental conditions. After discovery of an equipment failure, it is important to determine if 

the shielding and source are intact. Procedures for routine inspection, maintenance, and 

operability of exposure devices, survey instruments, transport containers, and storage 

containers are license requirements. Additional instructions are required to minimize exposure 

of persons in the event of an accident as well as source retrieval instructions. 

The NRC and Agreement States consider that theft or the loss of sources of larger activities 

may cause a substantial hazard. These licensees are required by regulation to report the theft 

or loss of these sources to their regulatory authority to aid in the prevention of personnel 

exposures. Immediate NRC notification is required if radioactive material or equipment may 

have caused, or threaten to cause, an exposure in excess of 5 rem TEDE or 15 rem lens dose 

equivalent. Lost sources have, in the past, resulted in exposures to members of the public and 

contamination of unrestricted areas. 

II. Evaluations and Assessments Prior to Fukushima 

Although most equipment that uses sealed sources is very rugged, accidents occur that can 

temporarily or permanently damage the equipment. If this occurs while the source is outside its 

shielded container, there is potential for worker and public exposure. On numerous occasions 

ice, blowing snow or freezing rain have prevented the retraction of a source into a radiography 

exposure device. Structural and vehicle fires can damage the outer casing or over pack of a 

radiography exposure device, but the source generally remains intact and in the shielded 

position. Several instances have occurred where devices have been washed overboard while 

on route to a job site or off a deep sea oil drilling platform; divers recovered most, but not all of 

these devices. In other instances, trucks or tractor trailers have run over moisture density 

gauges with minor damage to the storage container. In those few instances where the source 

was dislodged from the source holder, the source was quickly found and re-inserted into the 

source holder and protective shielding. 

In 2006, a nuclear pharmacy reported flood damage from a storm. All radioactive material 

(sealed sources and radiopharmaceutical material) was present and no removable 

contamination was present at the pharmacy. A September 2001 tornado damaged multiple 

buildings at the U.S. Department of Agriculture facility. The buildings contained various forms of 

carbon-14, tritium, iodine-125, phosphorous-32, and sulfur-35. All fume hoods and storage 

refrigerators, where the radioactive material was stored, were intact and the tornado did not 

affect the material. No other instances involving Category 1 or 2 medical or academic sources 

to external events are documented in the nuclear material event database. No instances in the 

nuclear material event database document the loss of licensee control of Category 1 industrial, 

medical or academic sources due to severe weather, earthquake, or flooding. 

For more than 30 years, the NRC has put improvements in place in its emergency 

preparedness and incident response programs especially upon reviewing lessons learned after 

several severe natural disasters. For example, in 2005, Hurricane Katrina struck the Gulf 

Coast. The Federal Emergency Management Agency (FEMA) described Hurricane Katrina as, 
28 

“the single most catastrophic natural disaster in U.S. history,” with estimated damage exceeding 

$100 billion. In preparation for and after landfall, the NRC contacted its Category 1 and 

Category 2 licensees to obtain information on the physical status and the security of facilities 

and materials in those states potentially affected by the hurricane. Coordination with the 

Agreement States proved successful in obtaining current information regarding the status of 

radioactive materials located in those states. The NRC emergency response was coordinated 

with other Federal agencies, such as the Centers for Disease Control and Prevention, the U.S. 

DOE, the Environmental Protection Agency (EPA), FEMA, and the U.S. Army Corps of 

Engineers. If a licensee had lost control of a radiation source, DOE’s aerial monitoring system 

was available to search for and find any missing or misplaced radiation source. The 2005 

Hurricane Season Lessons Learned Task Force proposed a series of recommendations to 

include a recommendation involving reporting the status of risk-significant materials. These 

improvements were applied to NRC responses later in the 2005 hurricane season. 

Extreme external events such as earthquakes, tornadoes, hurricanes, flooding, or wildfires 

occur every year in the United States. Many of these events have the potential to cause the 

loss of licensee control of radioactive material. One tool that is available to assist the NRC and 

Agreement States monitor the status of radioactive material is the NSTS. The NSTS is a secure 

web-based database designed to document the location and status of Category 1 and 

Category 2 radioactive sources regulated by the NRC and the Agreement States. About 1,300 

licensees began reporting their Category 1 and Category 2 source information for inclusion into 

the NSTS in January 2009. The tracking spans the life cycle of the source from manufacture 

through shipment receipt, decay, and disposal. NSTS enhances the ability of the NRC and 

Agreement States to inspect and investigate, communicate information to other government 

agencies, and verify legitimate ownership and use of nationally tracked sources. 

III. Post-Fukushima Event Evaluations/Assessments 

Although the source integrity of a sealed source after exposure to a natural event is very likely 

to be retained, loss of control of a source by the licensee is a possibility. The tsunami that 

struck the northeast coast of Japan destroyed thousands of buildings and created approximately 

20 million tons of debris. Five million tons of this debris were swept out to sea (3.5 million tons 

of debris was deposited along the coast of Japan and another 1.5 million tons became floating 

debris in the Pacific Ocean). The Government of Japan does not believe radioactive material 

floats among the debris. 

Since the Fukushima accident, the NRC also has developed a mapping tool to give situational 

awareness of the NSTS’s Category 1 and 2 licensees and sources as distributed across the 

United States. The mapping tool was developed with help from the Federal Bureau of 

Investigation (FBI) to support the NRC’s Operations Center for incident response. The 

information for the mapping tool is updated quarterly and is being shared with other agencies 

(FBI, DOE, National Nuclear Security Administration, and Department of Homeland Security) to 

support their situational awareness. Tools like this, along with NSTS, are used regularly to 

monitor and verify source security after natural events like earthquakes, wildfires, tornadoes, 

hurricanes, and flooding. 

NRC staff evaluated radioactive material licensees through document review of license 

requirements and regulatory framework, incident reports contained in the nuclear material 
29 

events database, and internal meetings with license review experts. The current assessment 

also reviewed various external events to determine if a failure of a sealed or unsealed source 

could reasonably be expected to result from the event that would be more severe than 

previously evaluated. The following table gives an overview of the potential outcome and 

overall assessment of several types of events on industrial, academic, and medical uses of 

radioactive material. The natural events assessed and an initial evaluation is summarized 

below: 

Effect of External Events on Radioactive Material Users 

External Event Outcome Assessment 

Flood Challenge to structures and 

vehicles in which unsealed 

and sealed sources are used, 

stored, or transported; 

potential loss of control 

• Licensees are required to meet any 

city, county or state requirements/ 

regulations regarding building 

construction. 

• The greatest concern is the loss of 

licensee control of radioactive 

material. During the license 

application process, the location of 

all unsealed and sealed source 

materials must be described. For 

those radioactive materials with 

activities that exceed those for 

Category 1 (e.g., irradiators) and 

Category 2 (e.g., well logging) 

sources, they must be reported to 

and tracked in the NSTS in 

accordance with 10 CFR 20.2207. 

 

• Emergency plans that address 

natural phenomenon, including an 

earthquake, a tornado, flooding, or 

other phenomena, generally are not 

required by industrial licensees, 

unless the activity limits exceed 

those in 10 CFR 30.72, 

Schedule C. However, the licensee 

must develop operating and 

emergency procedures and the 

operator must demonstrate an 

understanding of these procedures. 

Seismic Challenge to structures in 

which unsealed and sealed 

sources are used or stored. 

Manufacturer must check 

Gamma knife equipment 

before patient use. 

High Wind and 

Missiles 

Challenge to structures and 

vehicles in which unsealed 

and sealed sources are used, 

stored, or transported; 

potential loss of control 

Lightning Challenge to structures in 

which unsealed and sealed 

sources are used, stored, or 

transported 

Snow and Ice Loads Challenge to structures and 

vehicles in which unsealed 

and sealed sources are used, 

stored, or transported 

Drought None 

Temperature 

Extremes 

Failure to retract radiography 

source because of ice 

formation. 

External Fire Challenge to structures and 

vehicles in which unsealed 

and sealed sources are used, 

stored, or transported. 

Potential damage to shielding 

material (e.g., lead) upon 

exposure to extreme heat. 

Loss of Power Some gauges and devices 
30 

External Event Outcome Assessment 

would not be affected. 

Challenge to transport of 

sealed and unsealed source 

material. Facility control, 

monitoring and security 

systems would be inoperable 

once battery backup power is 

exhausted 

NRC staff considered the 12 NTTF recommendations for applicability to radioactive material 

users and evaluated to determine if any future action is warranted. The following table 

summarizes the result of the review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Action Needed. Greatest concern is loss of 

licensee control. NSTS requires tracking of 

sources. NRC has upgraded its capability to 

provide the status of risk-significant materials in 

response to extreme events. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. See item 2 above. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

No Action Needed. See item 2 above. Some 

licensees require power to ensure facility 

controls for monitoring and security. For these 

facilities, processes and procedures are in 

place to ensure material is secured in the event 

of a loss of power. 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 
31 

Recommendations Review Result 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identify insights about hydrogen 

control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

through further study of the 

Fukushima Dai-ichi accident. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

Not Applicable to these licensees. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. NRC upgraded its 

capabilities to provide situational awareness of 

the NSTS’s Category 1 and 2 sources. As 

discussed above extreme natural phenomena 

events have occurred that have affected this 

class of licensees. NRC staff continually 

assesses the need to improve emergency 

response capabilities as a result of lessons 

learned from these events. 

9 The task force recommends that the 

NRC require that facility EPs address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional 

emergency plan (EP) topics related to 

multiunit events and prolonged SBO. 

No Action Needed. See item 8 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 8 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 
32 

IV. International Experience 

NRC staff, in concert with the IAEA secretariat and other Member States, support an IAEA 

Action Plan on Nuclear Safety that includes an action entitled, “Review and Strengthen IAEA 

Safety Standards and Improve Their Implementation.” This action calls on the Commission on 

Safety Standards and its’ Safety Standards Committees to review and revise, as necessary, 

relevant IAEA safety standards. The Radiation Safety Standards Committee (RASSC) has 

participated in the extensive program of work undertaken by the IAEA to review all of its safety 

standards following the Fukushima accident. This program focused initially on the review and 

revision of safety requirements, followed by the supporting safety guides. This topic was first 

discussed by RASSC in June 2011 and is a standard agenda item at all subsequent RASSC 

meetings. 

NRC staff has participated in a series of IAEA-sponsored international conferences. For 

example, the 2nd International Conference on Occupation Radiation Protection was convened 

to provide an opportunity for experts to discuss the various radiation protection issues, that have 

been highlighted by the Fukushima accident and to consider how these should be addressed at 

both the national and international levels. The conference agenda covered the wide range of 

occupational exposure situations associated with fuel cycle, reactor, industrial, and medical, 

activities, and from occupational exposure situations associated with naturally occurring 

radioactive materials. Recommendations from these conferences will be included in a broader 

IAEA post-Fukushima accident review that should be finalized in 2015. 

V. Conclusion and Recommendations 

NRC staff concludes that unsealed radioactive materials and sealed sources and devices used 

in industry, academia, and medicine are appropriately licensed and have sufficient engineering 

controls to protect the health and safety of workers and members of the public. Worker 

exposures are kept as low as is reasonably achievable and minimize the danger to life and 

property. The safety evaluation as documented in a sealed source and device registration 

certificate conducted before authorizing a manufacturer or distributor to distribute the radioactive 

sources to specific licensee assures the integrity of the device. Thousands of industrial sources 

have been exposed to harsh environmental stressors and licensees have developed operational 

and emergency procedures to deal with unplanned accidents and emergencies. Databases and 

mapping tools designed to track Category 1 and Category 2 radioactive sources are continually 

being improved and refined. Finally, specific guidance for licensing radioactive material has 

been updated and published for public comment. Therefore, no further study or regulatory 

action is warranted. 

4. Part 36 Irradiators 

I. Current Regulatory Framework 

Gamma radiation is routinely used in industry to eliminate disease-causing insects and microorganisms such as E. coli and salmonella. Food products, food containers, spices, fruits, plants 

and medical supplies are the products most commonly irradiated. To be effective, radiation 

exposure upward of hundreds to thousands of gray is required to sterilize these products. The 
33 

process does not leave a radioactive residue or cause the treated products to become 

radioactive. 

A Part 36 irradiator is defined as a facility that uses radioactive sealed sources for the irradiation 

of objects or materials and in which radiation dose rates exceeding 5 gray (500 rads) per hour 

exist at 1 meter from the sealed radioactive sources, but does not include irradiators in which 

both the sealed source and the area subject to irradiation are contained within a device (selfshielded) and are not accessible to personnel such as a blood/tissue irradiator. Part 36 

irradiators generally contain Category 1 sources of cobalt-60 with activities that range from 9 

PBq (250 KCi) to 1 EBq (30 MCi). 

The NRC and Agreement States license several types of irradiators. Underwater irradiators are 

devices in which sealed sources always remain underwater and workers do not have direct 

access to the sources or the area subject to irradiation without entering the pool. Panoramic 

irradiators include those facilities where the irradiations occur in air and workers potentially 

could have access to the source while it is in operation. If the source is stored in a pool of water 

when not in use, the facility is called a panoramic wet-source-storage irradiator. The United 

States has over 50 wet-source-storage irradiators in use, which are licensed to use between 37 

to 520 PBq (1 to 14 MCi) of cobalt-60. If a Part 36 irradiator’s sources are not stored in a pool 

of water when not in use, it is referred to as a dry-source-storage irradiator. In this case, the 

sources are stored in a large shielded container. Beam type dry-source storage irradiators emit 

a narrow beam of radiation which irradiates the target material in air in areas potentially 

accessible to workers. 

In a typical irradiation system, products are loaded into irradiation containers that are conveyed 

through a maze into a concrete or metal shielded room. Inside the shielded room, the products 

are exposed to gamma rays coming from a rack of cobalt-60 sources. The concrete or metal 

shield is designed to limit worker and public exposure to no more than 2 mrem (20 μSv) per 

hour at 1 meter from the shield. 

Radiation safety requirements for irradiators are described in 10 CFR Part 36, “Licenses and 

radiation safety requirements for irradiators.” This part describes the requirements for the 

design, construction and safe operation of irradiators under normal and emergency operating 

conditions. More on these issues will be described below. Security requirements for irradiators 

are described in 10 CFR Part 37, “Physical protection of category 1 and 2 quantities of 

radioactive material.” This regulation establishes security requirements for the safe use and 

transport of the most risk-significant quantities of radioactive materials. 

NRC regulations applicable to devices containing byproduct material and to NRC radiation 

safety evaluations are found in the following: 

• 10 CFR Part 20, “Standards for protection against radiation” 

• 10 CFR Part 21, “Reporting of defects and noncompliance” 
34 

• 10 CFR Part 30, “Rules of general applicability to domestic licensing of byproduct 

material” 

• 10 CFR Part 32, “Specific domestic licenses to manufacture or transfer certain items 

containing byproduct material” 

• 10 CFR Part 36, “Licenses and radiation safety requirements for irradiators” 

• 10 CFR Part 37, “Physical protection of category 1 and category 2 quantities of 

radioactive materials” 

• 10 CFR Part 71, “Packaging and transportation of radioactive material” 

Licensing

The NRC’s NUREG-1556 technical report series, “Consolidated Guidance about Materials 

Licenses,” offers a comprehensive source of reference information about various aspects of 

materials licensing and materials program implementation. These reports, where applicable, 

describe a risk-informed, performance-based approach to licensing consistent with the current 

regulations. Specific guidance applicable to industrial irradiators is described in NUREG-1556 

Volume 6, “Program-Specific Guidance about 10 CFR Part 36 Irradiator Licenses.” This 

NUREG volume is being updated from the previous version to include safety culture, security of 

radioactive materials, protection of sensitive information, and clarification of regulatory policies 

and guidance. The draft revision should be finalized in 2015. 

Facility design and available equipment must have sufficient engineering controls and barriers 

to protect the health and safety of workers and members of the public, keep exposures to 

radiation and radioactive materials as low as is reasonably achievable, and minimize the danger 

to life and property from the uses of the types and quantities of radioactive materials possessed 

by the licensee. The licensee must have and follow emergency or abnormal event procedures 

including external events such as an earthquake, a tornado, flooding, or other phenomena as 

appropriate for the facility’s geographical location. The design of the facility should consider 

abnormal events such as water loss or leakage from the source storage pool, a prolonged loss 

of electrical power, or a fire alarm or explosion in the radiation room. 

For example, during the 1980s, there was concern that source material used in irradiators might 

be accidentally released into the environment. With these concerns in mind, the radioactive 

material used in an irradiator must be as nondispersible and as insoluble as practical if the 

source is used in a wet-source-storage or wet-source pool irradiator. As a result, colbalt-60, a 

metal, is used as a source instead of cesium-137 (a soluble chloride salt). Furthermore, the 

source material must be doubly encapsulated in a material that is corrosion resistant. 

Licensees, during the application process, must supply the manufacturer’s (or distributor’s) 

name and model number for each requested sealed source and device. Licensees will be 

authorized to possess and use only those sealed sources and devices specifically approved or 

registered by the NRC or an Agreement State. 
35 

In most instances, steel reinforced concrete walls and ceilings are built to enclose these 

irradiators to ensure that public exposure is no greater than 2 mrem per hour when measured 

1 meter from the shield surface. For panoramic irradiators built in seismic areas such as 

California, the licensee must design the reinforced concrete radiation shields to retain their 

integrity in the event of an earthquake by designing to the seismic requirements of an 

appropriate building code such as American Concrete Institute (ACI) Standard ACI 318-89, 

“Building Code Requirements for Reinforced Concrete,” Chapter 21, “Special Provisions for 

Seismic Design,” or local building codes, if current. The foundation of the facility should be 

designed with consideration given to soil characteristics to ensure that it is adequate to support 

the weight of the facility shield walls. Soil failure could result in shifting of, and possibly damage 

to, the irradiator shield. For pool irradiators, the pool should be designed to assure that it is leak 

resistant and the pool walls are strong enough to bear the weight of the pool water and shipping 

casks. The pool is also designed to ensure that a dropped cask would not fall onto the irradiator 

sources, and the metal components are corrosion resistant. 

Inspection and Oversight

The NRC assures compliance with the regulatory requirements through both periodically 

scheduled and special inspections. Inspections of irradiators licensed under 10 CFR Part 36 

are conducted in accordance with formalized procedures, specifically Inspection Procedure 

No. 81722, “Irradiator Programs.” The objective of this procedure is to determine if licensed 

activities are being conducted in a manner that will protect the health and safety of workers and 

the general public. The procedures are also used to determine whether the licensed programs 

are being conducted in accordance with NRC requirements. The NRC also uses Inspection 

Manual Chapter 2815, “Construction and Preoperational Inspection of Panoramic, Wet-SourceStorage Gamma Irradiators,” to determine if panoramic wet-source-storage gamma irradiators 

are constructed, equipped and operated in accordance with the license application. 

Emergency Preparedness

The licensee is required to develop and implement emergency and abnormal event procedures 

for its facility. These event procedures generally address: (1) low or high water level indicator 

readings, an abnormal water loss, or leakage from the source storage pool; (2) a prolonged loss 

of electrical power; (3) a fire alarm or explosion in the radiation room; (4) interlock failure; and 

(5) natural phenomena, including an earthquake, tornado, flooding, or other phenomenon as 

appropriate for the facility’s geographical location. 

In addition to emergency procedures, the irradiator facility is designed and equipped to respond 

to numerous emergencies. For example, if electrical power at any panoramic irradiator is lost 

for longer than 10 seconds, the sources must automatically return to their shielded position. 

One facility is equipped with backup electrical generators with 5 to 7 days fuel supply because 

of hurricane-related frequent extended power outages. In the event of a fire, the radiation room 

at a panoramic irradiator must be equipped with a fire extinguishing system capable of 

extinguishing a fire without the entry of personnel into the room. 

As part of the licensee’s emergency or abnormal operating procedures, NRC notification is 

required when sources are stuck in an unshielded position; any fire or explosion occurs in the 

radiation room; there is damage to the source racks; there is structural damage to the pool liner 
36 

or wall; or abnormal water loss or leakage occurs from the source storage pool. Reports to the 

NRC must include a telephone report within 24 hours and a written report within 30 days. 

There have been several instances where source racks did not return to a shielded position. In 

one instance loss of electric power damaged a programmable logic control to an irradiator 

causing the failure of the source racks to return to a shielded position. The source racks were 

safely lowered mechanically. Two serious radiation-related injuries occurred at irradiator 

facilities during the 1970s. In both cases, the overexposures occurred when operators walked 

into the exposure room when the source was unshielded. The overexposures occurred 

because safety systems were intentionally bypassed or operating procedures were not followed. 

Two deaths unrelated to radiation exposure occurred at U.S irradiators. These workers were 

crushed while moving materials to be irradiated on a conveyor. 

II. Evaluations and Assessments Prior to Fukushima 

Irradiators have been safely used to sterilize food, spices, and medical supplies since the 

mid-1960s. First published in 1993, 10 CFR Part 36 was based, in part, on recommendations 

contained in American National Standards Institute standard N43.10, “Safe Design and Use of 

Panoramic, Wet Source Storage Gamma Irradiators.” Each license application review includes 

a review of the design, manufacturing, prototype testing, and proposed uses of the sources and 

the irradiator. Irradiators contain certain safety features. Source racks are designed to return to 

a safe or shielded position in the event of a loss of power, air pressure, or other external factors. 

Irradiators must have access controls to prevent inadvertent entry of personnel if the sources 

are not in the shielded position. The sources and devices are designed to survive normal 

conditions of use and likely accidental conditions. Most irradiators licensed for use in the United 

States have seismic detectors. When this detector senses excessive vibration, the source rack 

is automatically lowered into the pool. 

III. Post-Fukushima Event Evaluations/Assessment 

Since 2013, security requirements for irradiators are described in 10 CFR Part 37. This 

regulation describes a performance based program of security measures required for licensees 

that possess risk significant quantities of materials. If there is an actual, attempted, or 

suspicious activity related to the possible theft and diversion or radiological sabotage of risksignificant material, the licensee is required to notify both local law enforcement and the NRC 

immediately. 

The current assessment reviewed the generic facility design and available equipment to 

determine if engineering controls and barriers are sufficient to protect the health and safety of 

the public and irradiator workers in the event of severe natural phenomena such as an 

earthquake or tsunami. 

In the Unities States, 55 panoramic, wet-source-storage irradiators and one underwater pool 

irradiator, each containing a minimum of 37 PBq (1 MCi) of cobalt-60, are licensed for use. 

Nine of these irradiators are in seismic areas. The NRC defines a seismic area as any location 

where the probability of a horizontal acceleration in rock of more than 0.3 times the acceleration 

of gravity in 250 years is greater than 10 percent. Although not required, each of these nine 

irradiators is equipped with a seismic detector that causes the radiation source holder to 
37 

become fully shielded automatically in the source storage pool should the seismic detector be 

actuated. When returned to the storage pool, the radiation reading at the pool surface should 

be less than 2 mrem per hour. 

The average source storage pool holds between 16,700 and 18,100 gallons of water. The 

water level is typically maintained 3 meters (10 feet) above the top of the source material when 

the source rack is in the fully shielded position. The pool liner is fabricated using 3/16-inch 

stainless steel, encased in high density, steel reinforced concrete, generally 40 to 50 cm (15 to 

20 inches) thick. The pool is specifically designed without a tie-in to the main shield so that it is 

free to move independently of the shield in the event of an earthquake. Water chilling systems 

maintain pool water temperatures between 18 degrees to 27 degrees Celsius (65 degrees to 

80 degrees Fahrenheit). The water chilling system is necessary because a single 370 TBq 

(10,000 Ci) cobalt-60 sealed source has an estimated surface temperature of 130 degrees 

Celsius (265 degrees Fahrenheit). During normal operations several hundred individual 

sources are loaded into the source rack. Without chilling capacity, the pool water temperature 

will begin to elevate and level off after 5 to 7 days at approximately 65 degrees Celsius (150 

degree Fahrenheit). The pool water evaporation rate will be slow, such that it would take weeks 

before the source material is exposed to the air. Water can be manually added to the pool. For 

example, two wet storage irradiators can be quickly refilled by the fire department through a 

designed connection point located outside each shield. It is important to note that there is no 

adverse effect to the source material resulting from the pool being empty. Each source is safety 

tested to withstand temperatures in excess of 800 degrees Celsius (1,470 degrees Fahrenheit) 

for 1 hour, which simulates the temperature of many hydrocarbon fires. 

In the event the source rack does not return to the shielded position, the radiation shield 

surrounding the source is extremely robust. The walls of the shield are designed such that the 

radiation dose rate in areas that are normally occupied during operation of a panoramic 

irradiation may not exceed 2 millirem per hour at any location 30 cm or more from the wall or 

ceiling of the room. To achieve this level of protection, the concrete wall and ceiling thickness 

generally exceeds 165 to180 cm (65 to 70 inches). The design of the walls and roof is required 

to comply with the structural requirements of the local building code or the American Concrete 

Institute Standard 318-89; hence, the concrete density, rebar size, the rebar cross spacing and 

the number of rows is appropriate for use in a seismic area. Assuming a minimum concrete 

density of 147 pounds/cubic foot, the approximate weight of each radiation shield structure is at 

least 2,000 tons. An NRC staff review demonstrates that the structural components of Part 36 

irradiators have considerable seismic capacity and will maintain their structural integrity under 

quite severe ground motion conditions (greater than several g’s). However, during the review of 

the regulatory framework, the staff identified several equivalent methods for radiation shield 

construction as described in ANSI N43.10, “Safe Design and Use of Panoramic, Wet Source 

Storage Gamma Irradiators (Category IV) and Dry Source Storage Gamma Irradiators 

(Category II).” The staff will consider future rulemaking activities to include these equivalent 

methods. 

In the event of a significant earthquake, the warehouse roofing and walls may collapse 

restricting access to the radiation control room and primary building power may be lost for an 

extended period of time. However, the shield is designed to retain its integrity and to ensure 

that no radioactive material will be released and that there will not be any off site radiation 

exposure. 
38 

The NRC previously evaluated the potential impact of an earthquake generated tsunami to a 

pool irradiator built in Honolulu, Hawaii. NRC staff considers this evaluation to be generally 

bounding for other irradiators. This is because, other than the irradiator located in Hawaii, the 

licensed irradiator facilities are either located inland or in locations that would not generate 

tsunamis to the extent possible in Hawaii. For the evaluation of the Hawaii irradiator, at shore, 

tsunami waves up to 10 m (33 ft) can reach velocities of up to 13 m/s (29 mph). Given the 

weight of a single sealed source and the shear velocity needed to lift a source out of the storage 

pool, the wave velocities associated with the largest historical tsunami would be insufficient to 

remove a source from the bottom of the irradiator pool even if the facility had sustained enough 

damage that the source rack had been destroyed. In addition, the wave velocity of a windgenerated (i.e., hurricane) storm surge is less than that associated with a tsunami. Therefore, 

NRC staff concludes that it is unlikely that a tsunami or hurricane will result in a loss of control of 

radioactive material from an industrial irradiator that would have an adverse effect on public 

health and safety or the environment. 

During the license application process, the location of all Category 1 and Category 2 sources 

must be reported to and tracked in the NSTS, which the NRC maintains. The NSTS is a secure 

web-based database designed to document the location and status of Category 1 and 

Category 2 sources. This information can be used to give situational awareness in the event of 

a severe weather event. 

NRC staff evaluated panoramic irradiator licensees through document review of license 

requirements and regulatory framework, incident reports contained in the nuclear material 

events database and the daily event notification report, and internal meetings with license 

review experts. The current assessment also reviewed various external events to determine if a 

failure of a sealed or unsealed source reasonably could be expected to result from the event 

that would be more severe than previously evaluated. The following table gives an overview of 

the potential outcome and overall assessment of several types of events on academic and 

medical uses of radioactive material. The natural events assessed and an initial evaluation is 

summarized below: 

Effect of External Events on Part 36 Irradiators. 

External Event Outcome Assessment 

Flood Challenge to external 

structures in which irradiator 

sources are used. No loss of 

source control. 

• Warehouse roofing and walls may 

be damaged thus restricting access 

to the irradiator control room and 

access door until debris is 

removed. 

• The concrete/steel reinforced 

radiation shield, 70 to 74 inches 

thick, is designed to retain its 

integrity in the event of an 

earthquake by designing to the 

Seismic Challenge to external 

structures in which irradiator 

sources are used. Radiation 

shield has considerable 

seismic capacity and will 

maintain structural integrity 

under quite severe (several 

g’s) ground acceleration. 
39 

External Event Outcome Assessment 

High Wind and 

Missiles 

Challenge to external 

structures in which irradiator 

sources are used. No damage 

to shield integrity. 

seismic requirements of an 

appropriate source such as ACI 

Standard ACI 318-89. 

• A radiation shield constructed with 

steel, 12 to 14 inches thick, is 

designed to retain its integrity in the 

event of a severe earthquake. 

• Emergency or abnormal event 

procedures must address natural 

phenomenon, including an 

earthquake, a tornado, flooding, or 

other phenomena, as appropriate 

for the facility’s geographical 

location. 

Lightning Challenge to external 

structures in which irradiator 

sources are used. 

Snow and Ice Loads Challenge to external 

structures in which irradiator 

sources are used. No damage 

to shield integrity. 

Drought none 

Temperature 

Extremes 

none 

External Fire Challenge to external 

structures in which irradiator 

sources are used. No damage 

to sealed sources or shield 

integrity.. 

Loss of Power Source automatically returns 

to shielded position. 

Malfunction of source rack 

possible, thus requiring 

mechanical intervention to 

return sources to shielded 

position. Control system door 

interlocks failsafe closed and 

locked in the event of a power 

loss. All facility control, 

monitoring, and security 

systems would be inoperable 

once battery backup power is 

exhausted, estimated to be 

within 24 hours. 

NRC staff considered the 12 NTTF recommendations for applicability to Part 36 irradiators to 

determine if any future action is warranted. The following table summarizes the result of the 

review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

However, during the review of the regulatory 
40 

Recommendations Review Result 

balances defense-in-depth and risk 

considerations. 

framework, the staff identified several 

equivalent methods for radiation shield 

construction as described in ANSI N43.10, 

“Safe Design and Use of Panoramic, Wet 

Source Storage Gamma Irradiators (Category 

IV) and Dry Source Storage Gamma Irradiators 

(Category II).” The staff will consider future 

rulemaking activities to include these equivalent 

methods. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Action Needed. The concrete/steel 

reinforced radiation shield, 70 to 74 inches 

thick, is designed to retain its integrity in the 

event of an earthquake by designing to the 

seismic requirements of an appropriate source, 

such as ACI Standard ACI 318-89. A radiation 

shield constructed with steel, 12 to 14 inches 

thick, is designed to retain its integrity in the 

event of a severe earthquake. 

Staff assessment determined it is unlikely that a 

tsunami or hurricane will result in a loss of 

control of radioactive material from an industrial 

irradiator that would have an adverse effect on 

public health and safety or the environment. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. See item 2 above. For wet 

sources, sources are stored in pools of water. 

In addition each sealed source is safety tested 

to withstand temperatures in excess of 800 

degrees Celsius for 1 hour, which simulates the 

temperature of many hydrocarbon fires. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

No Action Needed. See items 2 and 3 above. 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identify insights about hydrogen 

control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 
41 

Recommendations Review Result 

through further study of the 

Fukushima Dai-ichi accident. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

No Action Needed. Pool water evaporation rate 

will be slow, and take weeks before the source 

is exposed to air. Water can be manually 

added to the pool. It is important to note that 

there is no adverse effect to the source material 

resulting from the irradiator pool being empty. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. Emergency or abnormal 

event procedures must address natural 

phenomenon, including an earthquake, a 

tornado, flooding or other phenomena, as 

appropriate for the facility’s geographical 

location. Because of the robust nature of the 

source material offsite consequences are 

unlikely due to extreme events. 

9 The task force recommends that the 

NRC require that facility EP address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 8 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 8 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

IV. Conclusion and Recommendations 

NRC staff concludes that Part 36 irradiators are licensed appropriately and have sufficient 

engineering controls to protect the health and safety of workers and members of the public. 

Worker exposures are kept as low as is reasonably achievable and minimize the danger to life 

and property. Based on the available information, it is unlikely that natural phenomena (e.g., 

tsunamis, earthquakes, hurricanes, or fire) will result in a loss of control of radioactive material 
42 

from an industrial irradiator that would have an adverse effect on public health and safety or the 

environment. Program specific guidance for irradiator licensing is being revised to include 

additional guidance on seismic design requirements for steel radiation shields, the security of 

radioactive materials, and clarification of regulatory policies and practices. As such, NRC staff 

concludes that irradiators are licensed as appropriate to the scope and potential hazard created. 

No further study or regulatory action for these facilities is warranted. 

5. Low-Level Waste Disposal Facilities 

I. Current Regulatory Framework 

Low-level waste (LLW) includes items that have become contaminated with radioactive material 

or have become radioactive through exposure to neutron radiation. LLW is not high-level 

radioactive waste, transuranic waste, spent nuclear fuel, or byproduct material as defined in 

paragraphs (2), (3), and (4) of the definition of byproduct material set forth in 10 CFR 20.1003, 

“Definitions.” LLW typically consists of contaminated protective shoe-covers and clothing, 

wiping rags, mops, filters, reactor water treatment residues, equipment and tools, luminous 

dials, medical tubes, swabs, injection needles, syringes, and laboratory animal carcasses and 

tissues. The radioactivity can range from just above background levels found in nature to levels 

found in parts that are inside the reactor vessel in a nuclear power plant. Another type of LLW, 

with large volumes waiting for disposal, is depleted uranium. Low-level waste is typically stored 

on-site by licensees, either until the radioactivity has decayed away and the affected waste can 

be disposed of as ordinary trash, or until amounts are large enough for shipment to a LLW 

disposal site in containers approved by the U.S. Department of Transportation. 

LLW disposal occurs at commercially operated LLW disposal facilities that must be licensed by 

either the NRC or an Agreement State. The Low-level Radioactive Waste Policy Amendments 

Act of 1985 (the LRWPA) gave States responsibility for the disposal of their low-level radioactive 

waste. The Act encouraged States to enter into compacts that would allow them to dispose of 

radioactive waste at a common disposal facility. Currently, four commercial disposal facilities 

accept low-level radioactive waste: (1) EnergySolutions in Barnwell, SC; (2) U.S. Ecology in 

Richland, WA; (3) EnergySolutions in Clive, UT; and (4) Waste Control Specialists, LLC, in 

Andrews, TX. The EnergySolutions site in Barnwell is licensed by South Carolina to receive 

wastes in Classes A, B, and C. The facility accepts waste from Connecticut, New Jersey, and 

South Carolina. The State of Washington licensed the U.S. Ecology site to receive wastes in 

Classes A, B, and C. It accepts waste from States that belong to the Northwest Compact 

(Washington, Alaska, Hawaii, Idaho, Montana, Oregon, and Wyoming) and the Rocky Mountain 

Compact (Colorado, Nevada, and New Mexico). The State of Utah licenses the 

EnergySolutions site to accept Class A waste only. The facility accepts waste from all regions 

of the United States. Finally, the Texas Commission of Environmental Quality licenses the 

Waste Control Specialists site to receive waste in Classes A, B, and C. This site is the most 

recent of the four, opening in 2012, and accepts LLW from Texas and Vermont, 34 States that 

do not have operating facilities, and the Federal government. 

Storage of LLW requires an NRC or Agreement State license. NRC or Agreement State 

regulations require the waste to be stored in a manner that keeps radiation doses to workers 

and members of the public below NRC-specified dose levels. Licensees must further reduce 
43 

these doses to levels that are ALARA. Actual doses, in most cases, are a small fraction of the 

NRC limits. 

Licensing 

The NRC and the Agreement States share responsibilities to protect workers, the public, and 

the environment from LLW. Federal regulations in 10 CFR Part 61 and the NRC’s LLW 

regulatory program, as well as compatible Agreement State regulations and oversight, play a 

key role in protecting the public and the environment. However, the licensees and operators of 

disposal facilities ultimately bear the primary responsibility for safely handling and using LLW 

materials. 

Regulations and Guidance Documents

The NRC regulations for shallow land disposal, which are undergoing revision, are in 10 CFR 

Part 61. NRC staff has issued several guidance documents on 10 CFR Part 61 requirements, 

including the evaluation of disruptive phenomenon (e.g., NUREG-1623). Also, in certain cases, 

guidance for other NRC regulatory programs (e.g., NUREG-1757) may be adapted to the 

10 CFR Part 61 analyses. Early guidance for the implementation of 10 CFR Part 61 (e.g., 

NUREG-1199 and NUREG-1200) was generally prescriptive. More recently, the review of sitespecific performance assessments has become more performance-based, driven in part by the 

Commission’s 1995 probabilistic risk assessment policy statement (60 FR 42622, August 16, 

1995). NRC staff has developed more recent guidance (e.g., NUREG-1573 and NUREG-1854) 

for detailed LLW performance assessment and probabilistic risk-informed analyses using an 

enhanced performance-based approach. NRC staff developed the “Branch Technical Position 

(BTP) on Concentration Averaging and Encapsulation,” on January 17, 1995, which covered, in 

part, the waste classification technical position. The BTP also defined a subset of concentration 

averaging and encapsulation practices that NRC staff finds acceptable in determining the 

concentrations of the radionuclides tabulated in 10 CFR 61.55, “Waste classification.” The 

latest version of the BTP was issued on February 15, 2015 (80 FR 10165). 

Emergency Preparedness

The LLW disposal facilities have emergency procedures they adhere to in their license. Each of 

the four sites has security and emergency preparedness conditions in their license. The 

regulations in 10 CFR Part 62, “Criteria and procedures for emergency access to non-Federal 

and regional low-level waste disposal facilities,” authorize the NRC to grant emergency access 

to any non-Federal or regional LLW disposal facility or to any non-Federal disposal facility within 

a State that is not a member of a Compact. The terms and conditions upon which the 

Commission will grant this emergency access are found in 10 CFR Part 62 Subparts B and C. 

The regulations in this part apply to all persons who have been denied access to existing 

regional or non-Federal low-level radioactive waste disposal facilities and who submit a request 

to the Commission for a determination pursuant to this part. 10 CFR Part 62 applies only to the 

LLW that the States have the responsibility to dispose of pursuant to section 3(1)(a) of the 

LLRWPA. 

The appropriate States and their licensees consider emergency preparedness for LLW facilities. 

As indicated above, risk or hazard at most LLW disposal facilities are minor. However, for those 
44 

facilities that may experience severe climate conditions such as earthquakes, flooding, 

tornadoes, and high speed winds, emergency preparedness should be addressed and potential 

risks and hazards need to be evaluated. The main concern is severe land erosion or landslides 

that may cause radioactive waste either in storage or in waste disposal cells to be transported 

offsite to cause harm or damage to humans or the environment. Under such conditions, sitespecific performance assessments, mitigation, waste containment, and emergency 

preparedness would be further developed. Under the proposed rule for the revision of 10 CFR 

Part 61, a safety case would be developed for LLW disposal facilities to ensure that they are 

meeting the performance objectives. The safety case would be updated periodically and use 

monitoring data to reflect actual site performance while taking extra protective measures as 

necessary. 

 

External Events

External events are events that originate either off the site or within the boundaries of the site 

but from sources that are not directly involved in the operational states of the LLW disposal site 

or facility. The regulations in 10 CFR Part 61.50, “Disposal site suitability requirements for land 

disposal,” requires site suitability for near-surface disposal. In this context, the primary 

emphasis in disposal site suitability is given to the isolation of wastes, and minimizing the 

potential for the occurrence of external events through site location. The disposal site’s features 

must ensure that the long-term performance objectives of 10 CFR Part 61, Subpart C, are met. 

In addition, the disposal site should be capable of being characterized, modeled, analyzed and 

monitored. Within the facility’s region or State, a disposal site should be selected so that 

projected population growth and future developments are not likely to affect the ability of the 

disposal facility to meet the performance objectives of 10 CFR Part 61, Subpart C. Areas 

having known natural resources which, if exploited, would result in failure to meet the 

performance objectives of 10 CFR Part 61, Subpart C, must be avoided. 

The disposal site must be well drained and free of areas of flooding or frequent ponding. Waste 

disposal should not take place in a 100-year flood plain, coastal high-hazard area or wetland, as 

defined in Executive Order 11988, “Floodplain Management Guidelines.” Upstream drainage 

areas must be minimized to decrease the amount of runoff which could erode or inundate waste 

disposal units. The disposal site must offer sufficient depth to the water table that ground water 

intrusion into the waste, perennial or otherwise, will not occur. The NRC will consider an 

exception to this requirement to allow disposal below the water table if it can be conclusively 

shown that disposal site characteristics will result in molecular diffusion being the predominant 

means of radionuclide movement and the rate of movement will result in the performance 

objectives of 10 CFR Part 61, Subpart C, being met. In no case will waste disposal be 

permitted in the zone of fluctuation of the water table. As stated in 10 CFR 61.50(a), the hydrogeologic unit used for disposal must not discharge ground water to the land surface within the 

disposal site. Areas must be avoided where tectonic processes such as faulting, folding, 

seismic activity, or volcanism may occur with such frequency and extent to significantly affect 

the ability of the disposal site to meet the performance objectives of 10 CFR Part 61, Subpart C, 

or may preclude defensible modeling and the prediction of long-term effects. Areas must be 

avoided where surface geologic processes such as mass wasting, erosion, slumping, land 

sliding, or weathering occur with such frequency and extent to significantly affect the ability of 

the disposal site to meet the performance objectives of Subpart C of this part, or may preclude 

defensible modeling and prediction of long-term impacts. The disposal site must not be located 
45 

where nearby facilities or activities could adversely affect the ability of the site to meet the 

performance objectives of 10 CFR Part 61, Subpart C, or significantly mask the environmental 

monitoring program. 

Before granting a license, all of the above site suitability features must be addressed in terms of 

potential hazards that may affect compliance with the LLW disposal facility performance 

objectives. In addition, the licensee shall follow emergency and operational event procedures 

which are appropriate for the LLW facilities and onsite activities (e.g., waste storage or 

treatment). Except for certain radionuclide releases within regulatory acceptable limits, no 

instances involving external events at LLW facilities have been reported. It is unlikely that 

events such as severe weather, earthquake, or seismic events would have significant safety 

effects during a compliance period of 1,000 to 2,000 years. 

II. Evaluations and Assessments Prior to Fukushima 

The NRC relies on the four performance objectives for 10 CFR Part 61 (i.e., protection of the 

public, protection of intruders, protection of workers, and site stability) to mitigate disruptive 

processes as well as other associated requirements that were developed, in part, based on the 

technical analyses of the consequences of disruptive events. 

Disruptive Events and Processes

The potential for disruptive processes was recognized when the NRC developed 10 CFR 

Part 61 in 1982. The performance objectives and other requirements were designed to limit the 

effects of disruptive events on the performance of a low-level waste disposal facility. The 

requirements applied a defense-in-depth approach, which include: 

• waste characteristics – provides limits to radiological concentrations and hazardous 

characteristics (e.g., flammability, pyrophoricity) 

• siting characteristics – supplies required or exclusionary characteristics that limit the 

potential for disruptive processes and events 

• waste classification system – requires higher concentration waste to be disposed of 

more robustly (e.g., deeper) 

• system design – provides a description of the active systems that cannot be used for 

long-term safety 

• performance objectives – provides that a system must be capable of being modeled and 

analyzed to demonstrate that the performance objectives will be met (performance 

objectives include radiological protection as well as site stability) 

After waste is placed in disposal cells and those cells are closed, the waste is covered by about 

3 to 10 meters or more of soil, rock, and other materials. Placement of the waste underground 

makes rapid release of the waste to the environment extremely unlikely. Rapid release of 

radioactivity to the environment from a closed LLW disposal facility requires a high energy event 

such as from a volcano. The limited radiological content of the waste combined with the high 

dispersion associated with the event results in a risk-limiting state. Disruptive events could 
46 

affect the passive performance of a LLW disposal facility, if passive performance is relied upon. 

Because the waste is buried at moderate to significant depths below the land surface, there are 

substantial delays between the release time from the facility and the release time to the 

accessible environment. The delay times are in the range of tens to thousands of years 

depending on the facility, location, and relevant pathways. Even at the minimal end, adequate 

time is available for remedial response. 

III. Post-Fukushima Event Evaluations/Assessments 

In response to the Fukushima Lessons Learned activities, the NRC staff has assessed the 

mitigation of effects of disruptive events and processes on the disposal of LLW. The 

assessment involved a review of the pertinent regulatory requirements, guidance documents, 

and technical analyses that have been performed. The combination of historical regulatory 

requirements and guidance documents, technical analysis, newly proposed regulatory 

requirements, and revised guidance documents are sufficient to mitigate the risk from disruptive 

events on LLW disposal facilities. In addition, because of the inherent characteristics of the 

waste and disposal facility designs, LLW disposal facilities are inherently resistive to the more 

likely disruptive processes. 

Disruptive Events Guidance Documents

NRC staff has developed new draft guidance to support the revisions to 10 CFR Part 61 and to 

supplement existing guidance. Chapter 2 of the draft guidance document and its associated 

appendices offer detailed guidance for determining the proper scope of the analysis (i.e., the 

Features Events and Process) to incorporate the site-specific analysis of disruptive 

phenomenon. NRC staff developed hazard maps of many disruptive processes to facilitate this 

screening and analysis process. After a licensee identifies and characterizes potential 

disruptive processes, they may use different technical assessment methods to understand the 

effect of the disruptive processes on the ability of a disposal site to meet the 10 CFR Part 61 

performance objectives. 

In addition to the technical analyses, licensees may use engineered barriers to mitigate the 

effect of disruptive processes. Based on the characteristics of low-level waste disposal 

facilities, NRC staff expects erosive processes to be the most likely of all of the disruptive 

processes to affect the long-term stability of most disposal facilities. Therefore, licensees 

develop robust erosion control designs using durable materials which are usually developed 

based on the consideration of low-probability events, such as the “Probable Maximum 

Precipitation (PMP) and corresponding Probable Maximum Flood (PMF),” NUREG-1623. The 

PMF is defined as the hypothetical flood that is considered to be the most severe flood 

reasonably possible. It is derived based on (1) comprehensive hydro-meteorological application 

of the PMP, and (2) other hydrologic factors favorable for maximum flood runoff, such as 

sequential storms and snowmelt. Geomorphic hazard assessment may be used to develop the 

erosion protection designs. 

The inadvertent intruder analysis for 10 CFR 61.42 has a key role with respect to mitigating the 

effect of disruptive events on a low-level waste disposal facility. In 10 CFR Part 61, NRC 

envisioned that the inadvertent use of the disposal site may occur after closure. To ensure that 
47 

a member of the public who inadvertently intrudes into the disposal site was protected, NRC 

developed intruder scenarios and calculated the doses that could result from exposure to a unit 

concentration of the isotopes anticipated to be present in future LLW disposals. A concentration 

that was equivalent to 500 mrem was then calculated and formed the basis for the waste 

classification tables. The intruder scenarios considered intrusions such as, drilling, home 

construction, discovery, and agriculture development. These calculations and results are 

described in the draft and final Environmental Impact Statements for 10 CFR Part 61, as well as 

an update to the analysis published in 1986 (NRC, 1981; NRC, 1982; NRC,1986). NRC staff 

has determined that the intruder scenarios bound the risk from other potential disruptive events. 

The unlikely probability of the scenarios occurring and the limited dispersion involved in the 

intruder scenarios combine to make the intruder scenarios more limiting. In addition to the 

intruder scenarios, exposed waste scenarios and operational scenarios were also assessed. As 

discussed above, robust erosion control covers are expected. However, the risk associated 

with total or partial removal of the cover was also assessed. The covered waste was assumed 

to be exposed and dispersed by wind or water. The transport mechanism used was the 

transport of radionuclides via surface water runoff leading to the exposure of individuals from a 

contaminated water body or wind transport. The entire disposal facility was assumed to be 

exposed. The surface water erosion rate was about 0.82 tons/acre-yr. Effects were calculated 

for time periods of up to 1,000 or 2,000 years after closure. The maximum estimated dose was 

about 1 mrem per year. 

Operational scenarios assessed included a container accident and an accident involving fire. In 

the container-accident scenario, an accident is assumed to occur in which a container drops 

from a significant height, breaks open, and releases a portion of its contents into the air. 

Atmospheric dispersion was included for offsite impact calculations. The fire scenario involves 

a hypothetical fire in a disposal cell that lasts for two hours. The fire causes radioactivity to be 

released into the air. Exposures from a fire scenario are a strong function of the waste form, 

facility design, and operation. The maximum volume-weighted effect from a fire scenario was 

about 32 mrem per year. 

NRC staff evaluated seismic activities as potential natural disruptive events that could affect 

current commercial LLW disposal facilities. In this regard, NRC staff evaluated the most recent 

USGS seismic activity maps, particularly recent maps showing 2 percent probability 

exceedance in 50 years of different peak ground acceleration distributions. Based on this 

evaluation, NRC staff concluded that LLW commercial disposal facilities are located in areas 

with peak ground acceleration in the range of 0.02 – 0.2g, where g is the standard gravity 

9.8 m/s2

. The Fukushima earthquake was estimated at 2.99 g which is about 15 times, or more, 

stronger than any anticipated earthquake at United States commercial LLW disposal facilities. 

Therefore, the potential damage for typical buildings above the surface is anticipated to be light 

to moderate. For shallow land disposal facilities, waste is placed below the surface and covered 

with a thick earthen cover of soil and rocks. The potential damage to the LLW structure and the 

potential release of radioactive materials in this type of a facility are negligible. Thus, NRC 

staff’s predicted seismic hazards events would cause no significant radiological effects to the 

public. Based on the above discussion and analysis, the doses to members of the public under 

extreme natural events are not expected to exceed the EPA protective action guideline levels 

under emergency situations. The following table gives an overview of the potential outcome 

and overall assessment of several types of events on LLW facilities. The natural events 

assessed and an initial evaluation is summarized below: 
48 

Effect of External Events on LLW Facilities 

External Event Outcome Assessment 

Flood • Challenge to LLW storage 

and treatment facilities, as 

well as disposal cells 

before emplacement of 

cover. 

• Challenge to onsite waste 

shipments waiting for 

disposal. 

• Challenge to waste 

acceptance criteria for 

contents of liquids. 

• Challenge to integrity of 

the disposal cell after 

placement of cover. 

The three LLW disposal sites in Utah, 

Richland, and Texas are located in a 

dry environment. Therefore, severe 

flooding is unlikely to occur. For South 

Carolina, LLW disposal site flooding 

may cause release of LLW materials 

particularly from uncovered cells; 

however; potential hazards/risk would 

be anticipated to be low because of 

dilution and mixing with surface 

materials. For all LLW facilities, 

disposal cells are designed to allow 

rain drainage, thus avoiding release of 

materials in disposal cells. It is noted 

that all four LLW disposal sites are in 

low seismic hazards areas. In addition, 

because of design, and the engineered 

and natural barrier systems at these 

facilities, there should be no significant 

safety effects. Furthermore, waste 

packages and disposal operation at the 

facility should minimize release of 

radioactive materials, and thereby 

minimize intrusions and provide 

additional defense-in-depth for safety. 

More importantly, the current waste 

acceptance criteria of waste classes 

and disposal at the sites should offer 

additional assurance of low hazard or 

risk from such external event. 

External events have been addressed 

in accordance with 10 CFR Part 61 

requirements; therefore, EPs that 

address natural phenomena, including 

earthquake, a tornado, flooding, or 

other phenomena, generally should not 

be required for LLW disposal facilities. 

For certain specific waste streams with 

large volumes and long-lived 

radionuclides (e.g., DU) more guidance 

and site-specific risk/hazard 

Seismic • Challenge to structures 

and integrity of containers. 

• Challenges to vaults and 

covers to contain LLW and 

minimize intrusion. 

• Challenges to integrity of 

disposal cells because of 

mechanical stresses or 

because of water 

infiltration through faults or 

cracks. 

High Wind and 

Missiles 

Challenge to structures in 

which LLW radioactive 

material is contained or stored 

in a transition before disposal. 

Lightning Challenge to structures that 

could catch fire during 

lightning which may contain 

stored pyrophoric radioactive 

materials in a transition before 

disposal. 

Snow and Ice Loads Challenge to structures in 

which radioactive material is 

stored or being treated for 

disposition. 

Drought None 

Temperature 

Extremes 

None except for pyrophoric 

materials stored before 

disposition. 
49 

External Event Outcome Assessment 

External Fire Challenge to structures in 

radioactive materials stored in 

a transition for disposition 

particularly pyrophoric 

materials. 

assessments are being developed in 

the context of a limited rulemaking for 

10 CFR Part 61. Such site-specific 

analysis should offer extra protective 

assurance particularly if climate 

conditions change in the long-term 

future. 

Loss of Power None 

NRC staff considered the 12 NTTF recommendations for applicability to LLW facilities and 

evaluated to determine if any future action is warranted. The following table summarizes the 

result of the review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Action Needed. Because of the inherent 

characteristics of the waste and disposal 

facilities’ designs, LLW disposal facilities are 

inherently resistive to disruptive events. Waste 

acceptance criteria provide additional 

assurance of a low hazard or low risk from 

external events. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. See item 2 above. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

Not Applicable. See item 2 above. 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 
50 

Recommendations Review Result 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identifies insights about 

hydrogen control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

through further study of the 

Fukushima Dai-ichi accident. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

Not Applicable. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. Licensees consider 

emergency preparedness of LLW facilities. 

Main concern is severe land erosion or 

landslides that may cause radioactive waste in 

storage or waste disposal cells to be 

transported offsite. However, risks or hazards 

are considered low based on the classification 

of the waste disposed in these facilities and 

design of the facilities being resistive to 

disruptive events. 

9 The task force recommends that the 

NRC require that facility EPs address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 8 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 8 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

For certain specific waste streams with large 

volumes and long-lived radionuclides (e.g., DU) 

more guidance and site-specific risk/hazard 

assessments are being developed in the 

context of 10 CFR Part 61 limited rulemaking. 
51 

IV. International Experience

NRC staff are actively engaged in international activities pertaining to post-Fukushima lessons 

learned applicable to NRC activities in LLW management, uranium recovery (discussed below 

in Section 6), and decommissioning (discussed below in Section 7). The NMSS/DUWP Director 

is the United States representative on the IAEA Waste Safety Standards Committee (WASSC). 

The NMSS/DUWP Director participated in the WASSC meetings after the Fukushima accident, 

as well as in the “IAEA Expert Meeting on Decommissioning and Remediation after Accident,” 

which was held at IAEA Headquarters in January 2013. NRC staff also participated in the 

review of IAEA standards (e.g., safety requirements), especially those modified in response to 

the IAEA “ACTION Plan” after the Fukushima accident. NRC staff also participated in Nuclear 

Energy Agency (NEA)/OECD activities particularly the “Radioactive Waste Management 

Committee” (RWMC). The NMSS Director is the United States representative on RWMC. The 

NRC staff participated in RWMC workgroup or task groups such as the Working Party on 

Dismantling and Decommissioning and its subsidiary task groups. 

It is noted that no IAEA waste safety standards requirement were identified to be modified 

based on Fukushima lessons learned. In addition, throughout NRC staff and management 

international participation, there were no significant safety issues or actions identified that need 

to be undertaken by the NRC to modify its regulations for activities involving LLW management, 

decommissioning, and uranium recovery. In fact, in several cases, NRC regulations are more 

restrictive to ensure safety than IAEA safety standards, particularly those pertaining to safety 

under emergency situation and post remedial actions after a severe accident. 

V. Conclusion and Recommendations 

The combination of current regulatory requirements and guidance documents, technical 

analysis, and the newly proposed regulatory requirements and revised guidance documents, 

are sufficient to mitigate the risk from disruptive events on LLW disposal facilities. The current 

locations of commercial LLW disposal facilities are in low to moderate potential seismic damage 

regions; thus anticipated damage to disposal facilities with significant releases of radioactive 

materials is unlikely. No other technical analyses are necessary at this time. However, for 

licensing of new facilities or renewal of existing facilities, the addition of the features, events, 

and process analyses requirements combined with the site-specific intruder assessment and 

site stability analyses will ensure that disruptive events are appropriately accounted for in 

licensing of LLW disposal facilities that may dispose of new waste streams, such as depleted 

uranium. The LLW commercial disposal facilities are sited to avoid disruptive events or 

processes as practicable. There are no specific protective action guideline limits required for 

LLW similar to reactors to design barriers to mitigate potential radionuclide releases from 

seismic events. Nevertheless; regulatory requirements for stability of LLW disposal facilities, as 

evident from current locations of commercially licensed LLW disposal facilities, ensure that 

safety of the public and the environment are protected from such disruptive events. 
52 

6. Uranium Recovery Facilities and Uranium Mill Tailings 

I. Current Regulatory Framework 

Uranium recovery involves extracting natural uranium ore from the earth and concentrating (or 

milling) that ore. These recovery operations produce a product, called “yellowcake,” which is 

then taken to a fuel cycle facility for further processing and enrichment. The yellowcake is 

transformed into fuel for nuclear power reactors. In addition to yellowcake, uranium recovery 

operations generate waste products, called byproduct material, which contain low levels of 

radioactivity. The NRC does not regulate conventional uranium mining (where ore is removed 

from open pits or underground shafts). The NRC regulates in situ recovery (ISR) (formerly 

known as in situ leach recovery) where the uranium ore is chemically altered underground 

before being pumped to the surface for further processing. The NRC also regulates 

conventional uranium milling where uranium ore is processed at the surface. 

In the United States, the ISR process is the most widely used technique to extract uranium from 

below the ground surface. In the ISR process, injection wells pump a chemical solution into the 

layer of earth containing uranium ore. The solution dissolves the uranium from the deposit in 

the ground, and is then pumped back to the surface through recovery wells and sent to the 

processing plant to be converted into uranium yellowcake which is a solid form of mixed 

uranium oxide. Yellowcake is commonly referred to compounds such as U3O8, and UO4H2O, 

because these chemical compounds comprise more than 85 percent of the yellowcake 

produced by UR facilities. This product (i.e., yellowcake) is then transported to a uranium 

conversion facility, where it is transformed into uranium hexafluoride (UF6), in preparation for 

fabricating fuel for nuclear reactors. Monitoring wells are installed at ISRs to ensure that 

uranium and chemicals are not escaping from the drilling area. 

At a conventional uranium mill, uranium ore from a mine is first crushed into sand sized 

particles. Then a leaching agent is used to remove the uranium from the ore. In addition to 

extracting 90 to 95 percent of the uranium from the ore, the leaching agent also extracts several 

other "heavy metal" constituents, including molybdenum, vanadium, selenium, iron, lead, and 

arsenic. A series of other chemical processes are used to further refine, remove chemical 

impurities, and concentrate the uranium into yellowcake. Similar to what occurs at an ISR 

facility, the finished yellowcake is packed in 55-gallon drums and transported to a uranium 

conversion facility. The mill tailings generated during the conventional milling process are 

considered byproduct material. Mill tailings typically are disposed of in an impoundment and are 

covered with a radon barrier to minimize the amount of radon leaving the impoundment. In 

addition to radon emanation, mill tailings also contain other heavy metals that are typically found 

with uranium. 

In general, the primary industrial hazards associated with uranium milling are the occupational 

hazards found in any metal milling operation that uses chemical extraction, as well as the 

chemical toxicity of the uranium itself. The main risk from uranium is intake through ingestion or 

inhalation. In addition, hazards of radon gas and decay progenies need to be addressed for 

protection of workers and the public. Because the uranium produced at these facilities is not 

enriched, there is no criticality hazard and little concern of fire or explosive hazards. 

Radiological hazards are also of negligible concern because uranium has little penetrating 
53 

radiation and only moderate non-penetrating radiation. Therefore, the primary radiological 

hazard is attributable to the presence of radium in the byproduct material. 

Licensing and Regulations 

The NRC currently regulates uranium recovery operations in Wyoming, New Mexico and 

Nebraska. By issuing or amending a current license, the NRC authorizes the licensee to 

construct and operate (with specified conditions) a uranium recovery facility, expand an existing 

facility, or restart an existing facility at a specific site, in accordance with established laws and 

regulations. 

Uranium mill tailings are primarily the sandy processed waste material from a conventional 

uranium mill. This ore residue contains the radioactive decay products from the uranium chains 

(mainly the U-238 chain) and heavy metals. As defined in 10 CFR Part 40, the tailings or 

wastes produced by the extraction or concentration of uranium or thorium from any ore 

processed primarily for its source material content is byproduct material. This includes discrete 

surface waste resulting from uranium solution extraction processes, such as in situ recovery, 

heap leach, and ion-exchange. Byproduct material does not include underground ore bodies 

depleted by solution extraction. The wastes from these solution extraction facilities are 

transported to a mill tailings impoundment for disposal. 

The requirements for this type of byproduct material are described in 10 CFR Part 40, 

Appendix A, “Criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings 

or Wastes Produced by the Extraction or Concentration of Source Material from Ores 

Processed Primarily for Their Source Material Content.” In particular, the criteria in Appendix A 

cover the siting and design of tailings impoundments, disposal of tailings or wastes, 

decommissioning of land and structures, ground water protection standards, testing of the radon 

emission rate from the impoundment cover, monitoring programs, airborne effluent and offsite 

exposure limits, inspection of retention systems, financial surety requirements for 

decommissioning and long-term surveillance and control of the tailings impoundment, 

and eventual government ownership of the tailings site under an NRC general license. 

Licensees and applicants may propose alternatives to the specific requirements of Appendix A 

for approval by the Commission. 

Mill tailings could pose a hazard to public health and safety. As the radium decays over 

thousands of years, tailings produce a radioactive gas called radon. To keep them isolated, the 

tailings are placed in “tailings piles” for long-term storage or disposal. A tailings pile could be a 

large trench or a former mine pit and must meet NRC regulations. These piles generally are 

lined, covered, and monitored to detect leaks. The NRC also requires that adequate funds will 

be available to decommission the site, properly close the tailings pile, and maintain and monitor 

the site over the long term. 

The U.S. Congress passed the Uranium Mill Tailings Radiation Control Act in 1978. This law 

created two programs to protect the public and the environment from uranium mill tailings. 

Among other things, this law made DOE responsible to assume control of uranium processing 

sites after closure for long term monitoring, although a State may voluntarily assume such 

control. 
54 

Title I – Legacy Sites 

Many sites that produced uranium for the early nuclear power and weapons programs closed 

before requirements had been established for cleanup and long-term maintenance. Under the 

law, as of May 18, 2015, the DOE was responsible for cleanup at 22 legacy sites. All of these 

sites are in western states, except for two sites in Pennsylvania. 

The NRC reviewed DOE’s plans for cleanup and long-term maintenance of the Title I sites. The 

DOE has completed surface clean-up and is continuing to clean up ground water at several 

sites. The DOE maintains these sites and supplies long-term surveillance under a general 

license from the NRC. 

Title II – Sites Licensed in 1978 or Later 

The NRC’s primary implementing regulations for these sites are found in 10 CFR Part 40. 

These sites must be licensed by the NRC or an Agreement State. For all uranium sites, the 

following apply: 

• Holders of licenses to recover uranium are responsible for compliance with and the cost 

of site decontamination, decommissioning, and reclamation requirements issued by the 

NRC. 

• Facility design must be submitted for review with the application. 

• Plans to clean up any surface or ground water contamination must be approved by the 

NRC or the Agreement State in which the site is located. 

For conventional mill facilities, the following apply: 

• When a site has ceased operations and is being prepared for closure, DOE develops a 

long-term surveillance plan for the site that must be reviewed by the NRC. 

 

• When the NRC finds a site meets the cleanup standards or concurs that a state-licensed 

site meets the standards, and the tailings pile meets the approved design criteria, the 

NRC or state can terminate the license. 

• After the NRC has accepted DOE’s long-term surveillance plan and terminated the 

license, the site transfers to DOE for long-term surveillance and maintenance under a 

general license. 

 

• A State can assume responsibility for long-term care and maintenance of the site, but to 

date, this has not occurred. 

In 1983, the EPA issued uranium mill tailings standards for both Title I and Title II sites. In 1985 

and 1987, the NRC updated its regulations for Title II sites to be consistent with EPA's 

standards. In 1995, the EPA issued final standards for cleaning up ground water at Title I sites. 
55 

Currently there are 28 uranium mill sites that are in varying licensing status. Eleven NRClicensed uranium mill sites closed and are being decommissioned. Nine sites licensed by 

Agreement States are being decommissioned. Six sites are fully decommissioned (four NRC 

sites and two in Agreement States) and were transferred to DOE for long-term monitoring. The 

NRC licensed a facility in Clive, Utah, in 1993, to take mill tailings for disposal. In 2004, Utah 

assumed regulatory authority for the Clive disposal site as an Agreement State. The Texas 

Commission for Environmental Quality (TCEQ) licensed a facility to take mill tailings for disposal 

near Andrews, Texas, in 2008. 

A uranium recovery license is typically valid for 10 years from the date of the Commission 

finding under 10 CFR 40.32 based on current NRC licensing practice. A uranium recovery 

license can be renewed in 10-year increments. As of April 2014, the NRC is reviewing eight 

such actions for a new facility, expansion of an existing facility, or renewal of an existing facility. 

Additional applicable regulations include: 

• 10 CFR Part 20, “Standards for protection against radiation” 

• Appendix A to 10 CFR Part 40, “Criteria Relating to the Operation of Uranium Mills and 

the Disposition of Tailings or Wastes Produced by the Extraction or Concentration of 

Source Material from Ores Processed Primarily for Their Source Material Content” 

Appendix A to 10 CFR Part 40 sets forth the criteria relating to the operation of uranium mills 

and the disposition of tailings or wastes produced by the extraction or concentration of source 

material from ores processed primarily for their source material content. In general, these 

criteria require uranium recovery facilities to control industrial hazards and address waste and 

decommissioning concerns. 

Inspection and Oversight

NRC inspects licensed facilities and covers areas such as training of personnel who operate a 

uranium recovery facility, radiation protection programs, dose records, and security of nuclear 

materials onsite and during shipment. 

The goals of the inspection program at uranium recovery facilities are: 

• Ensure operations are conducted safely and in accordance with license conditions and 

NRC's regulations and requirements. 

• Ensure, through direct observation and verification, that uranium recovery facilities are 

taking appropriate corrective actions to address violations of operational safety 

procedures. 

 

• Ensure that a uranium recovery site is minimizing contamination and exposure to the 

workers and to the public using ALARA concept throughout its operation including 

packaging of yellowcake into safe containers as required by NRC and DOT regulations. 

The NRC does not monitor radioactive releases independently on a continuous basis, but 

performs planned and unplanned inspections in coordination with EPA and Agreement States. 
56 

Independent measurement of the releases to verify the licensee's samples may be conducted in 

coordination with the concerned State. 

Emergency Preparedness

Previous NRC staff assessments have evaluated the risks at both conventional uranium mills 

and ISR facilities. Specifically, NUREG-0706, “Final Generic Environmental Impact Statement 

on Uranium Milling,” considered risks at conventional uranium mills and evaluated several 

scenarios resulting in a release of radioactive materials. NUREG-0706 found situations where 

workers could receive a dose from these scenarios (but still within the Part 20 limits); the report 

did not find significant doses to members of the public resulting from the evaluated scenarios. 

NUREG/CR-6733, “A Baseline Risk-Informed, Performance-Based Approach for In Situ Leach 

Uranium Extraction Licensees,” presents NRC staff analyses of in situ leach uranium extraction 

facility operations and accidents that consider both likelihood of occurrence and the 

consequence. The analyses in NUREG/CR-6733 are conservative and demonstrate that in situ 

leach uranium extraction facilities operated with properly trained workers and effective 

emergency response procedures generally pose low levels of radiologic risk. NRC staff 

considers analyses similar to, or based on, those in NUREG/CR-6733 to be an appropriate 

basis for licensee safety analyses. 

The guidance in NUREG-1569, “Standard Review Plan for In Situ Leach Uranium Extraction 

License Applications,” does not state that applicants should have complex accident analyses, 

consequence evaluations, and probability determinations. However, site-specific conditions and 

circumstances must be addressed in any uranium recovery application. The NRC has 

evaluated the effects of accidents at conventional uranium mills (NUREG-0706). These 

analyses demonstrate that, for most credible potential accidents, consequences are minor so 

long as effective emergency procedures and properly trained personnel are used. The previous 

studies found situations where workers may be subjected to higher than anticipated normal 

doses, but the doses would remain within the 10 CFR Part 20 limits. Doses to members of the 

public were expected to be lower than the worker doses. 

 

External Events

External events are events that originate either off-site or within the boundaries of the site but 

from sources that are not directly involved in the operation of uranium recovery facilities. 

External events hazards and potential effects are appropriately evaluated during licensing or 

license amendment reviews. External events are important aspects of uranium recovery 

licensing and inspection process. 

The regional location and site layout for a proposed uranium recovery facility is extensively 

reviewed to show the relationship of the site to local water bodies (lakes and streams); 

geographic features (highlands, forests); geologic features (faults, folds, outcrops); 

transportation links (roads, rails, airports, waterways); political subdivisions (counties, 

townships); population centers (cities, towns); historical and archeological features; key species 

habitat; and non-applicant property (farms, settlements). In this regards, maps are reviewed to 

examine geologic features, structures such as surface impoundments, diversion channels, 

monitoring wells, and recovery plant buildings. 
57 

NRC staff also considers meteorology through a description of the general climate of the region 

and local meteorological conditions based on appropriate data from the National Weather 

Service, military, or other stations recognized as standard installations. NRC staff considers 

potential events based on review of data on precipitation, evaporation, and joint-frequency 

distribution data by wind direction, wind speed, stability class, period of record, and height of 

data measurement. The on-site program should be designed in accordance with Regulatory 

Guide 3.63, “Onsite Meteorological Measurement Program for Uranium Recovery Facilities—

Data Acquisition and Reporting” (NRC, 1988). The meteorological data used for assessing 

effects are substantiated as being representative of expected long-term conditions at and near 

the site. 

NRC staff also considered the seismicity and the seismic history of the region. In an 

application, historical seismicity data should be summarized on a regional earthquake epicenter 

map, including magnitude, location, and date of all known seismic events. Where possible, 

seismic events associated with the tectonic features described in the geologic structures should 

be included. Flooding, faulting, folding, seismic activities, volcanism, meteorology, 

climate/climatology, and surface water, as well as the hydrological system at the facility must be 

addressed in terms of potential hazards that could affect compliance with the uranium recovery 

facility safety objectives. 

II. Post-Fukushima Event Evaluations/Assessments 

NRC staff evaluated various external events to determine if a failure of safety systems or 

barrier(s) at uranium recovery or mill tailing facilities could cause significant release of 

radioactive materials to harm workers or the public or cause any significant damage to the 

environment. In summary, external events such as flooding, seismic, or external fire hazards 

are not expected to result it doses in excess of NRC dose limits. For ISR facilities, the process 

of uranium extraction is conducted below ground. Therefore, potential effects from seismic or 

flood hazards would be negligible. For mill tailing facilities, the tailings are placed in piles at or 

below surface and stabilized with thick clay covers and heavy rip-rap boulders to minimize 

erosion and to stabilize materials under severe conditions. Therefore, effects from external 

events on mill tailing piles should not exceed public dose limits. Under severe external events 

conditions, potential dose effects to members of the public should not exceed EPA protective 

action guidelines under emergency situation based on an NRC staff assessment. Therefore, 

the potential risk for all external events is low. In summary, for licensed uranium recovery and 

mill tailings facilities, NRC staff has found that under extreme external events or conditions, 

dose impacts would not exceed regulatory criteria for members of the public under an 

emergency situation. The following table gives an overview of the potential outcome and overall 

assessment of several types of events on uranium recovery and mill tailing facilities. The 

natural events assessed and an initial evaluation is summarized below: 

Effect of External Events on Uranium Recovery and Mill Tailing Facilities 

External Event Outcome Assessment 

Flood • Challenge to uranium 

recovery structures, 

impoundments, storage, 

External events are addressed in the 

license application. Therefore, severe 

flooding is unlikely to occur or cause 
58 

External Event Outcome Assessment 

and treatment facilities 

(e.g., ion exchange, 

calcination, and drying), as 

well as byproduct waste 

before disposition or 

permanent disposal. 

• Challenge to onsite 

yellowcake packaging and 

shipments waiting for 

transport to fuel cycle 

treatment and enrichment 

facilities. 

• Challenge to acceptance 

criteria for environmental 

releases and contents of 

liquids. 

• May challenge integrity of 

the impoundment and 

buried piping between the 

well fields and the 

processing facilities. 

• Uncontrolled chemical 

releases. 

significant harm to uranium recovery 

facilities. Although severe flooding 

may cause low-activity releases from 

uranium recovery impoundments, the 

consequences of such a release would 

be low. The combination of siting and 

design requirements, existing drainage 

systems, and engineered and natural 

barrier systems at uranium recovery 

facilities are expected to prevent 

significant safety effects from external 

events. 

Consequences of extreme conditions 

(e.g., from (i) radon releases from 

process streams; (ii) yellowcake dryer 

hazards; (iii) lixiviate leaks in buried 

piping between the well fields and the 

processing facility; and (iv) chemical 

accidents) are expected to be less than 

EPA protective action guidelines. 

Seismic • Challenge to uranium 

recovery structures and 

integrity of containers. 

• Challenges to vaults and 

covers to contain uranium 

recovery impoundments 

and byproduct materials 

waste. 

• Challenges to integrity of 

process streams and 

potential radon release. 

• Challenges to the buried 

piping and infrastructure 

because of lixiviant leaks. 

• Challenges to integrity of 

chemical treatment system 

and potential 

environmental damage. 

• Challenges to integrity of 

yellowcake. 
59 

External Event Outcome Assessment 

High Wind and 

Missiles 

Challenge to structures in 

which uranium recovery 

equipment and engineered 

systems are housed. Radon 

releases as well as low-activity 

uranium and chemicals may 

be released. 

Lightning Challenge to structures that 

may catch fire during lightning 

strikes which may contain 

stored yellowcake. 

Snow and Ice Loads Challenge to structures in 

which ore is being processed 

for concentration, calcination, 

and drying; and to byproduct 

waste in a transition before 

disposal. 

Drought None 

Temperature 

Extremes 

None in the cold regions. In 

the hot and dry regions 

concerns of fire from oxidizing 

material stored before 

shipment. 

External Fire Challenge to structures in 

which uranium recovery plant 

and equipment are housed. 

Effects on uranium oxides, 

yellow cake, stored before 

shipment could be significant. 

Oxidizing nature of yellowcake 

could cause a fire hazard. 

Assignment of risk from 

external hazard due to the 

oxidizing nature of yellowcake 

is low. 

Loss of Power Possible release of radon. 

NRC staff considered the 12 NTTF recommendations for applicability to uranium recovery and 

mill tailing facilities to determine if any future action is warranted. The following table 

summarizes the result of the review. 
60 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Action Needed. For ISR facilities, uranium 

extraction is conducted in the subsurface. For 

mill tailings facilities, the tailings are placed in 

piles at or below the surface and stabilized 

(e.g., use of clay and rip-rap boulders). Under 

severe external event conditions dose effects to 

members of the public are not expected to 

exceed EPA protective action guidelines. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. See item 2 above. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

Not Applicable. See item 2 above. 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identify insights about hydrogen 

control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

through further study of the 

Fukushima Dai-ichi accident. 

Not Applicable. There are no credible 

scenarios that create hydrogen in quantities of 

concern for these licensees. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

Not Applicable. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. Under severe external 

event conditions, dose effects to members of 

the public are not expected to exceed EPA 

protective action guidelines. 
61 

Recommendations Review Result 

9 The task force recommends that the 

NRC require that facility EPs address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 8 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 8 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

III. Conclusion and Recommendations 

NRC staff concludes that the uranium recovery and mill tailing facilities licensed by the NRC or 

Agreement States are appropriate to the scope and potential hazard. The current assessment 

reviewed various external events to determine if a failure of safety system(s) or barrier(s) at 

uranium recovery facilities could cause significant release of radioactive materials that would 

harm workers or the public or cause significant damage to the environment. Potential hazards 

from external events are believed to be insignificant or low for the above facilities. Therefore, 

NRC staff concludes that no further study or regulatory action is necessary for uranium recovery 

facilities. 

7. Decommissioning Reactors and Complex Materials Facilities 

I. Current Regulatory Framework 

The NRC regulates the decommissioning of power reactors, research and test reactors, 

materials and fuel cycle facilities, and uranium recovery facilities. The decommissioning 

program ensures that NRC-licensed sites are decommissioned in a safe, timely, and effective 

manner so that they can be returned to beneficial uses. Each year, the NRC terminates about 

150 materials licenses. Most of these license terminations are routine, and the sites require 

little, if any, remediation to meet the NRC’s unrestricted release criteria. 
62 

In general, most decommissioning research and test reactors, materials and fuel cycle facilities, 

and uranium recovery facilities, or sites, present low hazard or risk because radioactive sources, 

including spent fuel, are typically removed after operations cease when entering into the 

decommissioning mode. Decommissioning involves subsequent decontamination and remedial 

actions to reduce residual radioactivity to a level corresponding to the dose criteria of 10 CFR 

Part 20, Subpart E. 

The potential radiological hazard at decommissioning facilities is normally low, as it is typically 

associated with residual radioactivity on equipment, surfaces, and surface or subsurface soil. 

Low hazard may also be associated with radioactive waste materials resulting from cleanup, 

decontamination, and dismantlement that may be transition activities before permanent disposal 

or disposition. Decontamination and treatment processes during decommissioning present 

minor hazards to workers. 

Radionuclides remaining after cessation of operation are typically contained on surfaces of 

equipment or buildings, or in the soil. The potential for a release of the radioactive materials 

requires severe environmental or climate conditions such as flooding, tornado, or earthquake. 

For power reactor decommissioning sites, the principal radiological risks are associated with the 

storage of spent fuel onsite. Most decommissioning sites have spent fuel stored only in 

independent spent fuel storage installations, which are addressed in Section 1 of this enclosure. 

Five power reactors recently permanently shut down and defueled and entered into a 

decommissioning status (Kewanee Power Station; San Onofre Generating Station Units 1 and 

2; Crystal River Unit 3 Nuclear Generating Station; and Vermont Yankee Nuclear Power 

Station). These recently permanently shut down power reactors still have spent fuel stored in 

the spent fuel pool (SFP), as well as in independent spent fuel storage installations (except for 

Crystal River, which does not yet have an independent spent fuel storage installation). In 

addition, one other older permanently shutdown power reactor (Millstone Unit 1) still has fuel 

stored in the SFP. Note that several decommissioning power reactors no longer have any fuel 

stored onsite and the risks are the same as for the material decommissioning sites discussed 

below. The applicability of Fukushima Dai-ichi lessons learned to the decommissioning power 

reactor with fuel still stored in the SFP are addressed separately from the assessment of other 

decommissioning material sites. 

Licensing 

The NRC, its licensees, and the Agreement States share responsibility to protect public health 

and safety and the environment during decommissioning of reactors and material facilities. The 

decommissioning sites cover a variety of licensing activities including power reactors, research 

and test reactors, complex materials, uranium recovery, and fuel cycle facilities. There are 

about 49 sites in Agreement States undergoing decommissioning (36 materials and 13 uranium 

recovery facilities). NRC decommissioning sites include: (1) 18 power reactors, (2) 8 research 

and test reactors, (3) 16 complex non-reactor facilities, (4) 39 uranium recovery facilities, and 

(5) 2 fuel cycle facilities. 
63 

Regulations and Guidance

The following regulations specifically apply to facilities undergoing decommissioning: 

• 10 CFR Part 20, Subpart E, “Standards for protection against radiation, radiological 

criteria for license termination” 

• 10 CFR 30.36, byproduct materials: “Expiration and termination of licenses and 

decommissioning of sites and separate buildings, and outdoor areas” 

• 10 CFR 40.42, source material: “Expiration and termination of licenses and 

decommissioning of sites and separate buildings, and outdoor areas” 

• 10 CFR 40, Appendix A, “Criteria Relating to the Operation of Uranium Mills and the 

Disposition of Tailings or Wastes Produced by the Extraction or Concentration of Source 

Material from Ores Processed Primarily for Their Source Material Content” 

 

• 10 CFR 50.82, production and utilization facilities: “Termination of license” 

• 10 CFR 50.83, “Release of part of a power reactor facility or site for unrestricted use” 

• 10 CFR 70.38, special nuclear material: “Expiration and termination of licenses and 

decommissioning of sites and separate buildings, and outdoor areas” 

• 10 CFR 72.54, “Expiration and termination of licenses and decommissioning of sites and 

separate buildings, and outdoor areas” 

• 10 CFR 72.218, “Termination of licenses” 

The primary decommissioning guidance documents are NUREG-1757, “Consolidated 

Decommissioning Guidance, Volumes I-IV;” NUREG-1700, Rev. 1, “Standard Review Plan for 

Evaluating Nuclear Power Reactor License Termination Plans;” NUREG-1575, “Multi-Agency 

Radiation Survey and Site Investigation Manual (MARSSIM);” and NUREG-1537, “Guidelines 

for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors.” These 

NUREGs describe: (1) methods acceptable to NRC staff in carrying out specific parts of the 

Commission's regulations; (2) techniques and criteria used by NRC staff in evaluating 

decommissioning actions; and (3) guidance to licensees responsible for decommissioning NRClicensed sites. 

The decommissioning process for materials licensees takes a risk-informed, performance-based 

approach and ensures compliance with the radiological criteria for license termination in 10 CFR 

Part 20, Subpart E. The approaches to license termination described in the regulatory guidance 

provide the information (subject matter and level of detail) needed to terminate a license by 

considering the specific circumstances of the range of radioactive materials users licensed by 

the NRC. Volume 1 is applicable to the decommissioning of materials facilities licensed under 

10 CFR Parts 30, 40, 70, and 72 and to the ancillary surface facilities that support radioactive 
64 

waste disposal activities licensed under 10 CFR Parts 60, 61, and 63. For complex material 

sites, a decommissioning plan is required to be developed within one year of site shutdown. 

When a power reactor permanently shuts down, a post-shutdown decommissioning activities 

report (PSDAR) is provided by the licensee, as required by 10 CFR 50.82. Guidance on the 

information and content contained in the PSDAR is specified in Regulatory Guide 1.185, 

“Standard Format and Content for Post-Shutdown Decommissioning Activities Report.” The 

purpose of the PSDAR is to provide the NRC and the public with a general overview of the 

licensee’s proposed decommissioning activities and to inform NRC staff of the licensee’s 

expected activities and schedule so that NRC staff can plan for inspections and make decisions 

about its oversight activities. The PSDAR is also a mechanism that informs the public of the 

proposed decommissioning activities before the conduct of those activities. Additional guidance 

on the power reactor decommissioning process is provided in NRC Regulatory Guide 1.184, 

“Decommissioning of Nuclear Power Reactors.” 

Most of the regulations applicable to operating power reactor licensees in 10 CFR Part 50 

continue to apply to power reactors that have permanently shut down. Because of the reduced 

risk posed by decommissioning power reactor sites when compared to operating reactors, many 

licensing actions – primarily involving exemptions and amendments – have been submitted by 

the licensees and processed by NRC staff in order to establish a long-term licensing framework 

that is representative of the reactor’s permanently shut down and defueled status. These 

license actions involve emergency planning, security, onsite and offsite insurance, operator 

staffing and training, and technical specifications. Decommissioning power reactor licensees 

maintain the licensing basis of the facility and reflect the status of the facility as 

decommissioning progresses in the updated final safety analysis report (UFSAR). 

In addition, a license termination plan (LTP), must be submitted two years or more before 

license termination. The level of detail submitted in the LTP will vary depending on when the 

licensee submits the LTP. The information submitted in the LTP should reflect the current 

status of the decommissioning at the facility. The regulations in 10 CFR 50.82(a)(9)(ii) require 

that the LTP include the following information: site characterization, identification of the 

remaining dismantlement activities, plans for site remediation, detailed plans for the final 

radiation survey, description of the end use of the site, an updated site-specific estimate of 

remaining decommissioning costs, and a supplement to the environmental report describing any 

new information or significant environmental change associated with the licensee’s proposed 

termination activities. The LTP must be submitted as a supplement to the licensee’s final safety 

analysis report or as an equivalent document. 

 

Inspection, Licensing, and Oversight

The goals of the inspection and licensing program at sites undergoing decommissioning are to: 

• Ensure, through the review of the decommissioning submittals and surveys, that the 

decommissioning can be conducted safely and in accordance with NRC’s regulations 

and requirements. 

 

• Ensure, through direct observation and verification, that decommissioning is being 

conducted in accordance with the decommissioning activities as described in the 
65 

respective decommissioning submittal and NRC's regulations. 

 

• Ensure that corrective actions are taken by facilities undergoing decommissioning if they 

are not conforming to the decommissioning activities as described in the respective 

decommissioning submittal or NRC’s regulations. 

 

• Ensure that the site, at the completion of decommissioning, complies with NRC's criteria 

for license termination. 

The NRC does not independently monitor radioactive releases on a continuous basis, but 

performs an independent measurement of the releases to verify the licensee’s samples. The 

NRC also reviews licensee’s documents and procedures for monitoring releases and reviews 

the results periodically. The inspection program, which contains the objectives and procedures 

for each type of inspection, is described in Inspection Manual Chapter (IMC) 2602, 

“Decommissioning Oversight and Inspection Program for Fuel Cycle Facilities and Material 

Licensees;” IMC 2545, “Research and Test Reactor Inspection Program;” and IMC 2561, 

“Decommissioning Power Reactors Inspection Program.” 

Emergency Preparedness

Emergency preparedness considerations for dismantlement and decontamination of facilities 

take into account the reduced risk of accidents, particularly for power reactors and fuel cycle 

facilities. As indicated above, risk or hazard would be greatly reduced for facilities undergoing 

decommissioning shortly after cessation of operations and upon entering into the 

decommissioning mode (e.g., after removal of spent fuel and/or radioactive sources). 

For facilities transitioning into defueling, dismantling, decontamination, and decommissioning; 

emergency preparedness would overlap with decommissioning planning to assess and evaluate 

other safety issues and hazards. Under such conditions, a site-specific emergency 

preparedness plan would be developed. In this context, EPA’s protective action guidelines 

would be used in the early and intermediate phases of an emergency. 

For power reactors undergoing decommissioning, exemptions from certain emergency 

preparedness requirements for normal operation (e.g., 10 CFR 50.47(b) and Appendix E to 10 

CFR Part 50) may be granted. The NRC has recently approved emergency preparedness 

exemptions based on when the decay heat level of the spent fuel in the SFP has lowered to the 

point that the risk of a zirconium fire is negligible. Licensees are still required to maintain an 

onsite emergency plan addressing the classification of an emergency, and notification and 

coordination with designated offsite government officials following an event declaration. If 

needed, offsite authorities may implement protective actions using a comprehensive emergency 

management (all-hazard) approach to protect public health and safety. These licensees are no 

longer required to maintain formal offsite radiological emergency preparedness, including the 

10-mile emergency planning zone. 

As part of the process, the licensee submits an amendment to their emergency plan that 

conforms to the exemptions granted by the NRC. The licensee will continue to notify the NRC 

and the State following the declaration of an emergency classification, and maintain 

communications and interface responsibilities with Federal, State, and local organizations that 
66 

may provide assistance in the event of an emergency declaration. In addition, provisions for 

fire, law enforcement, ambulance and medical services will continue to be provided via letters of 

agreement with local entities. 

External Events

External events, human-induced or natural events, are typically evaluated in the preliminary 

phases of the site evaluation process. 

The decommissioning plan (DP) typically addresses meteorology, climate, and surface water, 

as well as the hydrological system at the facility. The decommissioning process and plan also 

includes aspects of survey and monitoring, storage and control of licensee’s material as well as 

decommissioning material not in storage (see NUREG-1757, Appendix H). Further, the DP 

includes auditing and surveillance. Notification and inspection, including criticality safety, are 

significant aspects of the decommissioning reviews. External events such as flooding, 

earthquakes, tornadoes, volcanic eruptions, or severe climate conditions are usually addressed 

in the design of the facility24 and in the decommissioning plan (or UFSAR for decommissioning 

power reactors). In the unlikely event that such an external event were to occur, the status of 

the site is reported and a site-specific assessment of significant safety impacts is conducted to 

identify immediate actions that need to be taken to mitigate significant radiological releases. 

Under 10 CFR 20.1406, licensees must describe how the facility design and procedures for 

operation will minimize, to the extent practicable, contamination of the facility and the 

environment, to facilitate eventual decommissioning, and minimize, to the extent practicable, the 

generation of radioactive waste. In addition, licensees should, to the extent practical, conduct 

operations to minimize the introduction of residual radioactivity into the site, including the 

subsurface, in accordance with the existing radiation protection requirements in Subpart B and 

the radiological criteria for license termination in 10 CFR Part 20 Subpart E. 

For materials decommissioning sites the risk from external events is low, because of the low 

activity of the sources and the low potential of movement by external events for such sources. If 

such movement of sources at decommissioning materials facilities occurred, there would be 

dilution of such sources (e.g., from surface water or ground water, or mixing with clean soil) and 

potential dose effects would be less than public dose limits under 10 CFR Part 20. Under 

extreme external event conditions, NRC staff determined that the potential dose effects to 

members of the public would not exceed EPA protective action guidelines. 

 

24 For power reactors, 10 CFR Part 50.2 provides a definition for “design bases.” For these facilities, 

“design basis events” consist of those events and accidents that the NRC has – as a matter of historical 

practice – required licensees to consider when identifying safety-related SSCs needed to provide key 

safety functions (see 10 CFR 50.49). For facilities other than power reactors regulated by the NRC, 

“design basis” may not be defined. Regardless, for the purposes of this assessment the staff focused its 

review on extreme external events that go beyond those that were considered in the licensing of the 

facility. 
67 

West Valley Demonstration Project 

NRC staff received several stakeholder comments on a draft of its assessment. The draft 

assessment was issued to support a Category 3 public meeting on the topic. One of the 

stakeholder comments provided in response to the Category 3 public meeting (which was 

submitted jointly by the Alliance for a Green Economy, Citizens’ Environmental Coalition, and 

the Nuclear Information and Resource Service), expressed concern regarding the assessment’s 

applicability to the West Valley Demonstration Project (WVDP) in Western New York. 

The WVDP is a unique case and different from other NRC-licensed complex sites. The WVDP 

Act of 1980 (WVDPA) gives the NRC an advisory and a monitoring role with respect to the 

WVDP activities. Under the WVDPA, the Department of Energy (DOE) has responsibility to 

decontaminate and decommission: (a) the tanks and other facilities of the Western New York 

Service Center in West Valley in which high-level waste was solidified or stored, (b) the facilities 

used in the solidification of waste, and (c) any material and hardware in connection with the 

project. As a result of the Act, the WVDP NRC license was put into abeyance. When the 

DOE’s work under the Act is completed, the NRC license will be reactivated and the NRC 

licensee, New York State Energy Research and Development Authority (NYSERDA), will be 

responsible for the license. 

Nevertheless, until the WVDP NRC license is reactivated, the NRC responsibilities under the 

WVDPA are limited. NRC responsibilities under the WVDPA are to provide decommissioning 

criteria, informal review and comment, and monitoring. The objective of the monitoring activities 

is to help provide the bases for the Office of Nuclear Material Safety and Safeguards to advise 

the DOE of whether activities conducted at the WVDP facilities and their operation are adequate 

to protect the public health and safety and the environment from radiological hazards. However, 

protection of public health and safety and the environment is the sole responsibility of the DOE 

and its contractors. 

The WVDP decommissioning will be conducted in two phases. NRC staff provided a technical 

evaluation report to the DOE on February 25, 2010, documenting the staff’s review and 

comment on Phase I of the WVDP decommissioning plan (ADAMS Accession 

No. ML100360030). The key objective of the staff’s review of this plan was a determination on 

whether the proposed action described in the Phase 1 decommissioning plan satisfies the 

decommissioning criteria for unrestricted use specified in 10 CFR 20.1402. Based on its review, 

NRC staff determined that the plan provides reasonable assurance that the proposed action will 

meet the decommissioning criteria. 

Natural phenomena and external events were addressed in DOE’s Final Environmental Impact 

Statement (EIS) which was issued in January 2010 (ADAMS Accession No. ML100750122). In 

addition, an earlier Safety Analysis Report (SAR) addressed external events in more detail (see 

for example SAR document ID Number WVNS-SAR-023 (ADAMS Accession No. 

ML032940574). The NRC may informally review documents associated with WVDP activities as 

a result of its monitoring activities or it may do so as a result of commitments that DOE made in 

its WVDP Phase I Decommissioning Plan, or as a result of learning about development of DOE 

documents in NRC’s informal consultancy role. Phase II of the WVDP Decommissioning Plan is 

not yet developed. DOE and NYSERDA plan to develop the supplemental EIS by 2020. A 

Decommissioning Plan for Phase II will also be developed. 
68 

The primary responsibility for assessing Fukushima lessons learned with respect to the WVDP 

lies with DOE. The NRC recently became aware of a DOE report published in December 2014 

(http://energy.gov/sites/prod/files/2014/12/f19/2014_WVDP_Emergency_Mgmt_Tech._Basis_an

d_Emerg._Preparedness-December%20_2014.pdf) that assessed the DOE-WVDP’s progress 

on implementing the Fukushima lessons learned into its Emergency Management Program. 

Through the DOE Office of Enterprise Assessments (EA), a review was performed in part to 

determine WVDP’s progress in implementing enhancements recommended in DOE policy that 

stemmed from the DOE’s evaluation of Fukushima Lessons Learned for its facilities (OE-1, 

“Improving DOE Capabilities for Mitigating Beyond Design Basis Events (BDBEs).” 

The DOE review found, among other things, that the method DOE used to quantify the material 

at risk in one of its storage areas was not adequate, and that the WVDP plans and procedures 

may not support appropriate and conservative event categorization and classification, protective 

actions, and offsite notification. Due to the nature of some of the material that must be stored, 

handled, and transported during decommissioning, the area of emergency response may need 

to be enhanced. NRC staff will continue to monitor this situation. 

II. Evaluations and Assessments Prior to Fukushima 

Before Fukushima, NRC staff evaluated the potential effects of certain external events on 

nuclear power reactor facilities that had permanently ceased operation and defueled. The 

concern was for facilities where irradiated fuel remains in the SFP. In this context, two analyses 

and studies were conducted for SFPs subjected to postulated beyond-design-basis external 

events. NUREG-1738, “Technical Study of Spent Fuel Pool Accident Risk at Decommissioning 

Nuclear Power Plants,” contains an evaluation of the potential accident risk in a spent fuel pool 

at decommissioning plants in the United States. This study was prepared to supply a 

conservative technical basis for decommissioning rulemaking for permanently shut-down 

nuclear power plants. It describes: 

• the modeling approach of a typical decommissioning plant with design assumptions and 

industry commitments, 

• the thermal-hydraulic analyses performed to evaluate the behavior of spent fuel stored in 

the SFP at decommissioning plants, 

• the risk assessment of SFP accidents, 

• the consequence calculations, and 

• the sensitivity study and implications for decommissioning regulatory requirements. 

NUREG-1738 concluded that: “Given the robust structural design of SFP, it is expected that a 

seismic event with peak spectral acceleration several times larger than the SSE [safe shutdown 

earthquake] would be required to produce catastrophic failure of the structure. The estimated 

frequency of events of this magnitude differs greatly among experts and is driven by modeling 

uncertainties.” In summary, the results of the study indicate that the risk at SFPs is low and well 
69 

within the Quantitative Health Objectives. The risk is low because of the very low likelihood of a 

zirconium fire even though the consequences from a zirconium fire could be serious. The study 

includes use of a pool performance guideline as an indicator of low risk at decommissioning 

facilities. 

In 2005,25 NRC staff studied SFPs at Zion 1 and 2, Millstone Power Station 1, La Crosse BWR, 

Humboldt Bay 3, Indian Point 1, and GE Morris. The study concluded that 

it is not necessary to develop additional mitigation strategies for drain-down events for 

the five decommissioning power reactor SFPs. The heat-up is slow, because of the long 

decay time of the most recently discharged fuel in these SFPs. The minimum decay 

time of the fuel in these SFPs is 8.5 years. The decay power for 8.5 year old fuel is a 

factor of six lower than 1 year old fuel. Because of the slow heat-up time associated 

with such long decay times, current mitigation strategies augmented by ad hoc mitigation 

strategies are adequate. Also there appears to be no way to gravity drain the GE Morris 

SFP, even by introducing a hole in its side or floor. 

More recently, NUREG-2161, “Consequence Study of a Beyond-Design-Basis Earthquake 

Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor,” dated September 2014 

(ADAMS Accession No. ML14255A365), was issued. The study provides consequence 

estimates of a hypothetical spent fuel pool accident initiated by a low likelihood seismic event at 

a specific reference plant. The study compares high-density and low-density loading conditions 

and assesses the benefits of post-9/11 mitigation measures. Past risk studies have shown that 

storage of spent fuel in a high-density configuration is safe and risk of a large release due to an 

accident is very low. This study’s results are consistent with earlier research conclusions that 

spent fuel pools are robust structures that are likely to withstand severe earthquakes without 

leaking. 

III. Post Fukushima Event Evaluations/Assessments 

Assessment for Facilities Undergoing Decommissioning Other than Power Reactors

The current assessment reviewed various external events to determine if a failure of safety 

barrier(s) at decommissioning facilities could cause significant release of radioactive materials 

to harm workers or the public or cause any significant damage to the environment. The nature 

of events assessed and initial evaluation are summarized in the table below. The potential dose 

impact to members of the public under extreme external events do not exceed EPA protective 

action guideline levels under emergency situation. 

 

 

25 The 2005 study assessed the risk at decommissioning these power plants and GE Morris due to 

catastrophic failure of the spent fuel pool. While the study contains sensitive information the conclusion is 

not. 
70 

Effect of External Events on Decommissioning Facilities 

External Event Outcome Assessment 

Flood Challenge to structures at 

decommissioning facilities 

which may contain residual 

radioactive materials higher 

than release limits on 

surfaces. 

Challenge to areas with 

residual radioactivity above 

release limits in soil. 

Challenge to areas storing 

decommissioning waste in a 

transition for onsite disposal or 

ultimate off site disposition. 

During the decommissioning process 

licensees are required to submit either 

a “Decommissioning Plan” or a 

“License Termination Plan” describing 

all safety aspects to protect workers, 

the public, and the environment from 

potential release of radioactivity. In 

addition, as indicated above, licensees 

are required to assess site location, 

climate conditions, and potential safety 

issues related to certain events such 

as extreme rain or flooding conditions, 

tornadoes, and high wind events. 

During decommissioning, radioactive 

materials are either on building 

surfaces, in soil, or could be stored in 

piles or in containers for ultimate 

packaging and disposition. 

NRC staff does not expect radioactive 

materials to be released from 

decommissioning facilities that exceed 

Category 1 or Category 2 sources 

(e.g., sources that must be reported to 

and tracked in accordance with 10 

CFR 20.2207). For decommissioning 

facilities other than decommissioning 

power reactors emergency plans that 

address natural phenomena, including 

an earthquake, a tornado, flooding, or 

other phenomena, generally are not 

required, though external events are 

addressed in safety guides. 

Under extreme event conditions, NRC 

staff determined that the potential dose 

effects to members of the public would 

not exceed EPA protective action 

guidelines. 

Seismic Challenge to structures in 

which radioactive materials are 

stored or present as residual 

radioactivity above release 

limits. 

High Wind and 

Missiles 

Challenge to structures in 

which radioactive material is 

present above release limits or 

stored in a transition for 

disposal. 

Lightning Challenge to structures that 

may catch fire during lightning 

strikes which may contain 

stored radioactive materials in 

transition before disposition. 

Snow and Ice Loads Challenge to structures in 

which radioactive material is 

stored or being treated for 

disposition. 

Drought None 

Temperature 

Extremes 

None except for pyrophoric 

materials stored before 

disposition. 

External Fire Challenge to structures in 

radioactive materials stored in 

a transition for disposition. 

Loss of Power None 
71 

NRC staff considered the 12 NTTF recommendations for applicability to decommissioning 

facilities other than power reactors to determine whether future action is warranted. The 

following table summarizes the result of the review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Action Needed. For research and test 

reactors, complex non-reactor facilities, 

uranium recovery facilities, and two fuel cycle 

facilities undergoing decommissioning, the risk 

from external events is low because of the low 

activity of the sources and the low potential of 

movement by external events for such sources. 

Under extreme event conditions, NRC staff 

determined that the potential dose effects to 

members of the public would not exceed EPA 

protective action guidelines. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. See item 2 above. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

No Action Needed. 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

No Action Needed. For research and test 

reactors, complex non-reactor facilities, 

uranium recovery facilities, and two fuel cycle 

facilities undergoing decommissioning 

scenarios that create hydrogen in quantities of 

concern are not credible. 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identifies insights about 

hydrogen control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

through further study of the 

No Action Needed. For research and test 

reactors, complex non-reactor facilities, 

uranium recovery facilities, and two fuel cycle 

facilities undergoing decommissioning 

scenarios that create hydrogen in quantities of 

concern are not credible. 
72 

Recommendations Review Result 

Fukushima Dai-ichi accident. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

No Action Needed. Research and test reactor 

facilities undergoing decommissioning have 

had their spent fuel removed from the facility. 

For complex non-reactor facilities, uranium 

recovery facilities, and two fuel cycle facilities 

undergoing decommissioning this lessons 

learned item is not applicable. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. See item 2 above. 

9 The task force recommends that the 

NRC require that facility EPs address 

prolonged SBO and multiunit events. 

No Action Needed. See item 2 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 2 above. 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 2 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

Assessment for Power Reactors Undergoing Decommissioning

As discussed above, there are 18 power reactors undergoing decommissioning. Five of these 

reactors have recently shutdown permanently and removed fuel from their reactor vessels: 

Crystal River Unit 3, Kewaunee, San Onofre 2 and 3, and Vermont Yankee. These reactors still 

have spent fuel stored in SFPs, as well as in independent spent fuel storage installations 

(ISFSIs) (except for Crystal River, which does not yet have an ISFSI). Of the remaining 13 

power reactors undergoing decommissioning, only one of the reactors (Millstone Unit 1) stores 

spent fuel in a SFP. The remaining power reactors undergoing decommissioning either have 

their spent fuel stored in an ISFSI, in an operating unit’s SFP, or the fuel has been shipped 

offsite. 
73 

Power reactors undergoing decommissioning have two risks associated with them: (1) control 

of radioactive material other than spent fuel (similar to the decommissioning of other material 

sites) and (2) proper storage, protection, and handling of spent fuel (for those facilities that still 

have fuel onsite in either an independent spent fuel storage installation or SFP). Regarding the 

risks associated with the control of radioactive material other than spent fuel, the assessment 

found above for decommissioning material facilities is considered applicable to power reactors 

undergoing decommissioning. This assessment concludes that additional regulatory action is 

not necessary based on Fukushima lessons learned for these facilities. 

For power reactors undergoing decommissioning that store their spent fuel in ISFSIs, the 

assessment found in Section 1 of this enclosure is applicable. This assessment concludes that 

additional regulatory action is not necessary based on Fukushima lessons learned for these 

facilities. If the spent fuel is stored in an operating power reactor SFP, post-Fukushima actions 

taken for operating power reactors are applicable and safety is assured through those actions. 

Assessment for Decommissioned Power Reactors with Irradiated Fuel Stored in Spent Fuel 

Pools

As described above the risk associated with these facilities is associated with loss of cooling to 

the spent fuel leading to cladding failure and offsite release. For the five power reactors that 

have recently permanently shutdown and defueled their reactors, the orders associated with 

Fukushima lessons learned were rescinded. Specifically, the following orders were rescinded 

for these five power reactors: 

• Order EA-12-051, which requires reliable SFP instrumentation be installed and 

maintained. This order was issued to ensure that licensees were able to monitor SFP 

level to appropriately prioritize mitigation and recovery actions during an event. 

• Order EA-12-049, directing licensees to develop and implement strategies to maintain or 

restore core cooling, containment cooling, and spent fuel cooling capabilities following a 

beyond-design-basis external event. Compliance with this order includes an initial 

phase, a transition phase, and a final phase. 

In addition, Order EA-13-109 was rescinded for the lone boiling water reactor (Vermont Yankee) 

that is included in the group of five power reactors that have permanently shutdown recently. 

Order EA-13-109 requires that all licensees that operate boiling water reactors with Mark I and 

Mark II containment designs, including the Vermont Yankee Nuclear Power Station, implement 

requirements for reliable hardened containment vents capable of operation under severe 

accident conditions at their facilities. 

The basis for rescinding Order EA-12-051 is that because the licensees certified the permanent 

removal of fuel from the reactor vessel, the SFP becomes the primary safety concern for site 

personnel. In the event of a challenge to the safety of fuel stored in the SFP, decision-makers 

would not have to prioritize actions and the focus of response personnel would be on the SFP. 

Thus, the basis for this order no longer applies. 
74 

The basis for rescinding Order EA-12-049 for these units includes: 

• Guidance and strategies to maintain or restore core cooling and containment are no 

longer necessary because the nuclear fuel has been permanently removed from the 

reactor vessels and containments. 

• The decay heat levels in the SFPs at the time the unit would have been required to be in 

compliance with the order are low. At these low decay heat levels, reliance on the SFP 

coolant inventory for passive cooling will provide an equivalent level of protection to that 

required by the initial phase of Order EA-12-049. 

• At the time compliance with the order would have been required, the low decay heat 

levels will provide a long time to boil off the SFP inventory, and consequently a long time 

to reach a point at which makeup would be necessary for radiation shielding purposes. 

This eliminates the need for a transition phase of Order EA-12-049, where makeup 

water would be added using portable onsite equipment. 

• The low decay heat and long boil-off period at the time compliance with the order would 

have been required provides sufficient time for the licensee to obtain off-site resources 

on an ad hoc basis to sustain the SFP cooling function indefinitely, eliminating the need 

for the final phase required by Order EA-12-049. 

The basis for rescinding Order EA-13-109 for Vermont Yankee is found in an April 15, 2014, 

letter (ADAMS Accession No. ML14055A323). The basis states that following submittal of the 

10 CFR 50.82(a)(1) certifications for permanent cessation of operations and the permanent 

removal of fuel from the reactor vessel, the primary containment will no longer provide a safety 

function and the underlying purpose of Order EA-13-109 will no longer be served. On 

January 12, 2015, the Vermont Yankee licensee provided the certifications for permanent 

cessation of operations and the permanent removal of fuel from the reactor vessel (ADAMS 

Accession No. ML15013A426). 

In addition to the rescission of the orders, NRC staff has determined that the March 12, 2012, 

10 CFR 50.54(f) request for information for flooding and seismic reevaluations and for 

emergency planning applicable to operating reactors is generally not applicable to the five 

power reactors that have recently permanently shutdown and defueled their reactors, for the 

following reasons:26

 

26 The reasons given hear are generally applicable to the five reactors that have recently permanently 

ceased operations. The basis for the March 12, 2012, request for information responses not being 

necessary for San Onofre Units 2 and 3 and Crystal River Unit 3 are found in letters dated January 22, 

2014 (ADAMS Accession Nos. ML13329A826, and ML13325A847, respectively). The staff documented 

that Kewaunee is no longer obligated to respond to the March 12, 2012, request for information in letters 

dated October 9, 2013, and January 22, 2014 (ADAMS Accession Nos. ML13261A128, and 

ML13322B255, respectively). The staff is currently considering a March 12, 2015, Entergy letter 

providing a position that the March 12, 2012, request for information is no longer applicable to Vermont 

Yankee. 
75 

• No further responses or actions associated with the 10 CFR 50.54(f) letter are 

necessary, because the licensee is no longer authorized to load fuel into the reactor 

vessel and potential fuel-related accident scenarios are limited to the SFP. Unlike the 

reactor, the safety of fuel located in the SFP is assured for an extended period through 

maintenance of pool structural integrity, which preserves coolant inventory and 

maintains margin to prevent criticality. Small changes in the flooding hazard elevation 

would not threaten the structural integrity of a flooded pool. Further, previous 

evaluations of SFP structures have determined that seismic margins are very large. As 

seismic and flooding studies continue for the remainder of the operating fleet, new 

information concerning the adequacy of design bases of SFPs will be evaluated for 

applicability to decommissioned sites using existing NRC processes. 

• Based on the discussion above, the safety of the fuel stored in SFPs would not be 

substantially affected by potential changes in the flooding or seismic hazard levels. 

Furthermore, for beyond design basis external events challenging the safety of the spent 

fuel, recovery and mitigation actions could be completed over a long period of time due 

to the slow progression of any accident as a result of the very low decay heat levels 

present in the pool within a few months following permanent shutdown of the reactor. 

Thus, SFP beyond design basis accident scenarios at decommissioning reactor sites do 

not require the enhanced communication and staffing that may be necessary for the 

reactor-centered events the 10 CFR 50.54(f) letter addresses. 

NRC staff also assessed the Millstone Unit 1 SFP. The logic for rescission of the Orders 

EA-12-051 and EA-12-049, and the 10 CFR 50.54(f) request for information, to the five recently 

permanently shutdown and defueled power reactors is also the basis for why these orders and 

request for information did not apply to Millstone Unit 1. Millstone Unit 1 permanently shutdown 

on July 21, 1998, so the decay heat load associated with fuel in its spent fuel pool is a fraction 

of that found in the reactors that recently entered the decommissioning mode. In addition, 

although a boiling water reactor, Order EA-13-109 associated with Mark I and Mark II 

containments is not applicable to Millstone 1, because the fuel is in the SFP and the 

containment no longer provides a safety function. 

Additional Capabilities at Recently Permanently Shutdown and Defueled Power Reactors 

As previously noted, NUREG-1738 and NUREG-2161 suggest that the radiological 

consequences of a postulated beyond-design-basis accident that results in a zirconium fire of 

the fuel stored in the spent fuel pool can be extensive; however, the likelihood of such an 

accident is very low, even considering the most severe natural phenomena. In addition, for 

SFPs with sufficient decay time, there is little potential to cause offsite early fatalities regardless 

of the type of offsite response. Furthermore, even though the possibility of a beyond-designbasis SFP accident that results in a significant loss of SFP cooling water inventory is very 

unlikely, the recently decommissioned reactors have multiple strategies for providing makeup to 

the SFP which include, for example: using existing plant systems for inventory makeup, 

supplying water through hoses to a spool piece connection to the existing SFP piping, or using a 

diesel-driven portable pump to take suction from other sources of water to provide makeup or 

spray to the SFP. NRC staff has assurance that these strategies will continue to be maintained 

at the five recently permanently shutdown power plants based on a license condition for 
76 

mitigation strategies that was established in response to the terrorist attacks of September 11, 

2001, and was maintained even after these facilities permanently ceased operation. 

For this and other reasons, the five recently permanently shutdown power plants have proposed 

exemptions that would result in elimination of the requirements for formal offsite radiological 

emergency plans at these sites (but would require the maintenance of certain onsite capabilities 

to communicate and coordinate with offsite response authorities). Before granting the 

requested emergency planning exemptions, NRC staff has sought Commission approval. 

In general, exemptions to certain emergency planning requirements have been granted when 

the site-specific analyses show that at least 10 hours is available for mitigative actions from a 

postulated beyond-design-basis SFP drain-down event where cooling of the spent fuel is not 

effective. The assumptions include beyond-design-basis accidents that result in either a partial 

drain-down or the complete drainage of the SFP with rearrangement of spent fuel rack geometry 

and/or the addition of rubble to the SFP that would effectively impede any decay heat removal 

through all possible modes of cooling. The analysis postulates that decay heat transfer from the 

spent fuel via conduction, convection, or radiation would be impeded. This analysis is often 

referred to as an adiabatic heatup. The analysis demonstrates that, assuming a loss of all 

modes of cooling for the spent fuel stored in the SFP, there would still be a minimum of 10 

hours for the hottest fuel assembly to reach 900°C, which is the critical temperature threshold 

for self-sustained oxidation of zirconium cladding in air. This will ensure that sufficient time 

exists to initiate appropriate mitigating actions. 

The Commission has determined that the risk of offsite radiological release from these facilities 

is sufficiently low to exempt the licensee from certain requirements related to the need to 

maintain offsite emergency planning. The Commission has approved exemptions from certain 

emergency planning requirements through Staff Requirements Memorandum for San Onofre 

Units 2 and 3, Vermont Yankee, Crystal River Unit 3, and Kewaunee (ADAMS Accession Nos. 

ML15061A521, ML15061A516, ML14364A111, and ML14219A366, respectively). 

Longer-Term Actions

On December 30, 2014, the Commission directed NRC staff to proceed with rulemaking on 

decommissioning (SRM-SECY-14-0118 at ADAMS Accession No. ML14364A111). This 

rulemaking will address issues discussed in SECY-00-0145, such as: (1) the graded approach 

to emergency preparedness, (2) lessons learned from the plants that have already (or are 

currently) going through the decommissioning process, (3) the advisability of requiring a 

licensee’s PSDAR to be approved by the NRC, (4) the appropriateness of maintaining the three 

existing options for decommissioning and the timeframe associated with those options, (5) the 

appropriate role of State and local governments and non-governmental stakeholders in the 

decommissioning process, and (6) any other issues deemed relevant by NRC staff. Therefore, 

based on Commission direction, issues pertaining to emergency preparedness safety, defensein-depth, and potential risk from external events would be anticipated to be addressed in more 

detail in the context of the anticipated rulemaking for decommissioning. 

The table below summarizes the external events reviewed by NRC staff, the impact on the 

facility and a summary of the basis for no additional action being needed for this class of 

decommissioned power reactors. 
77 

Effect of External Events on Decommissioned Power Reactors 

External Event Outcome Assessment 

Flood Challenge to spent fuel pool 

cooling and to spent fuel pool 

integrity. 

• No assessment of the reactor or 

containment systems are 

necessary because the licensee is 

no longer authorized to load fuel 

into the vessel and potential fuelrelated accident scenarios are 

limited to the spent fuel pool. 

• The safety of fuel located in the 

spent fuel pool is assured for an 

extended period through 

maintenance of pool structural 

integrity, which preserves coolant 

inventory and maintains margin to 

prevent criticality. 

• Exemptions from certain 

emergency planning requirements 

have been granted based on the 

low likelihood that spent fuel would 

reach the zirconium ignition 

temperature in less than 10 hours, 

which was determined using 

conservative assumptions. These 

assumptions include beyonddesign-basis accidents that result 

in either a partial drain-down or the 

complete drainage of the SFP with 

rearrangement of spent fuel rack 

geometry and the addition of 

rubble to the SFP that would 

effectively impede any decay heat 

removal through all possible 

modes of cooling. 

• Recent and past studies have 

concluded that spent fuel pools are 

robust structures that are likely to 

withstand severe earthquakes 

without leaking 

• Decay heat levels have decreased 

with time since these reactors 

permanently shut down such that 

Seismic Challenge to spent fuel pool 

cooling, spent fuel pool 

integrity, and spent fuel pool 

geometry. 

High Wind and 

Missiles 

Challenge to spent fuel pool 

cooling, spent fuel pool 

integrity, and spent fuel pool 

geometry. 

Lightning Challenge to spent fuel pool 

cooling. 

Snow and Ice 

Loads 

Challenge to spent fuel pool 

cooling and to spent fuel pool 

integrity. 

Drought Challenge to spent fuel pool 

cooling. 

Temperature 

Extremes 

Challenge to spent fuel pool 

cooling. 

External Fire Challenge to spent fuel pool 

cooling. 

Loss of Power Challenge to spent fuel pool 

cooling. 
78 

External Event Outcome Assessment 

the boil off of the SFP inventory 

would take days, and 

consequently, there is sufficient 

time to restore SFP cooling or 

implement other mitigating actions.

• Recently decommissioned 

reactors have additional 

capabilities to provide makeup 

water and sprays to the SFPs that 

are required by a license condition.

NRC staff considered the 12 NTTF recommendations for applicability to decommissioning 

power reactors to determine whether future action is warranted. The following table 

summarizes the result of the review. 

Near-Term Task Force Recommendations and Future Actions 

Recommendations Review Result 

1 Establishing a logical, systematic, and 

coherent regulatory framework for 

adequate protection that appropriately 

balances defense-in-depth and risk 

considerations. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

Based on Commission direction, issues 

pertaining to emergency preparedness safety, 

defense-in-depth, and potential risk from 

external events are anticipated to be addressed 

in more detail in the context of the anticipated 

rulemaking for decommissioning. 

2 The task force recommends that the 

NRC require licensees to reevaluate 

and upgrade as necessary the designbasis seismic and flooding protection 

of SSCs. 

No Action Needed. The 10 CFR 50.54(f) 

request for information is no longer applicable 

in light of the movement of the fuel to SFPs, the 

robust nature of the SFPs, and the low decay 

heat levels. Issues could be revisited based on 

lessons learned from the analysis of response 

from operating reactors. 

3 The task force recommends, as part 

of the longer term review, that the 

NRC evaluate potential 

enhancements to the capability to 

prevent or mitigate seismically 

induced fires and floods. 

No Action Needed. 

4 The task force recommends that the 

NRC strengthen SBO mitigation 

capability at all operating and new 

reactors for design-basis and beyonddesign-basis external events. 

No Action Needed. Order-12-049 was 

rescinded for five reactors that have recently 

permanently ceased operation. Basis for order 

rescission is provided above, including fuel 

being moved to SFPs and low decay heat 

levels. 
79 

Recommendations Review Result 

5 The task force recommends requiring 

reliable hardened vent designs in 

BWR facilities with Mark I and Mark II 

containments. 

No Action Needed. Nuclear fuel has been 

permanently removed from the reactor vessels 

and containments. 

6 The task force recommends, as part 

of the longer term review, that the 

NRC identifies insights about 

hydrogen control and mitigation inside 

containment or in other buildings as 

additional information is revealed 

through further study of the 

Fukushima Dai-ichi accident. 

No Action Needed. Spent fuel is stored in 

SFPs that are robust in nature. Low decay heat 

levels provide for time to restore cooling and 

prevent hydrogen generation. Issues could be 

revisited based on lessons learned from the 

analysis for this item for operating reactors. 

7 The task force recommends 

enhancing spent fuel pool makeup 

capability and instrumentation for the 

spent fuel pool. 

No Action Needed. Once fuel is in the SFP, the 

SFP becomes the primary safety concern for 

site personnel. In the event of a challenge to 

the safety of fuel stored in the SFP, decisionmakers would not have to prioritize actions and 

the focus of response personnel would be on 

the SFP condition. Thus, the basis for this 

recommendation and Order EA-12-051 no 

longer applies. 

Recently decommissioned reactors have 

additional capabilities to provide makeup water 

and sprays to the SFPs that are required by a 

license condition. 

8 The task force recommends 

strengthening and integrating onsite 

emergency response capabilities such 

as EOPs, SAMGs, and EDMGs. 

No Action Needed. Based on SFPs being the 

focus of licensee’s attention, the robust nature 

of designs, and low decay heat loads, beyonddesign-basis accident scenarios at 

decommissioning reactor sites do not require 

the enhanced communication and staffing that 

may be necessary for the reactor-centered 

events. 

9 The task force recommends that the 

NRC require that facility EP address 

prolonged SBO and multiunit events. 

No Action Needed. See item 8 above. 

10 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue additional EP 

topics related to multiunit events and 

prolonged SBO. 

No Action Needed. See item 8 above. 
80 

Recommendations Review Result 

11 The task force recommends, as part 

of the longer term review, that the 

NRC should pursue EP topics related 

to decision-making, radiation 

monitoring, and public education. 

No Action Needed. See item 8 above. 

12 The task force recommends that the 

NRC strengthen regulatory oversight 

of licensee safety performance (i.e., 

the reactor oversight process) by 

focusing more attention on defensein-depth requirements consistent with 

the recommended defense-in-depth 

framework. 

No Action Needed. No regulatory gaps were 

identified as a result of the staff’s assessment. 

Based on Commission direction, issues 

pertaining to emergency preparedness safety, 

defense-in-depth, and potential risk from 

external events are anticipated to be addressed 

in more detail in the context of the anticipated 

rulemaking for decommissioning. 

IV. Conclusion and Recommendations 

For research and test reactors, complex non-reactor facilities, uranium recovery facilities, and 

the two fuel cycle facilities undergoing decommissioning, the current assessment reviewed 

various external events to determine if a failure of safety barrier(s) at these facilities could cause 

a significant release of radioactive materials that would harm workers or the public or cause 

significant damage to the environment. NRC staff concluded that these facilities undergoing 

decommissioning are appropriately licensed, given the radiological risk and potential hazards. 

Potential hazards for these facilities from external events are believed to be insignificant or low 

because the potential dose impact to the members of the public under extreme external events 

do not exceed EPA protective action guideline levels under an emergency situation. Therefore, 

no further study or regulatory action is recommended for such facilities. 

The conclusion found above for decommissioning material facilities is considered applicable to 

power reactors undergoing decommissioning regarding the risks associated with the control of 

radioactive material other than spent fuel. The staff also assessed the risk of external events for 

the decommissioned power reactors that have fuel stored in their SFPs (including those that 

have recently permanently ceased operation). Previous studies and analyses (e.g., NUREG1738 and NUREG-2161) have shown that the SFP structure is extremely robust and capable of 

withstanding the external events addressed by this paper. In addition, based on the decay heat 

levels of the five recently permanently shutdown reactors and the time available to take 

mitigating actions, there are no identified safety concerns that need further analysis. Further, as 

directed by the Commission in the staff requirements memorandum for SECY-14-0118, NRC 

staff is developing a comprehensive power reactor decommissioning rulemaking that will 

address the technical basis for an appropriate long-term regulatory framework for permanently 

shutdown reactors with fuel stored in the SFPs. Among other things, this rulemaking will likely 

examine the appropriate timing of certain regulatory transition points where current operating 

reactor requirements can be reduced based on the risks posed by external events in 

conjunction with thermal-hydraulic criteria that would be a function of the decay time of the 

spent fuel. 
81 

8. NRC-Licensed Non-Power Reactors 

I. Introduction

NRC staff evaluated the 31 NRC-licensed research and test reactors (RTRs) to assess 

applicability of Fukushima Dai-ichi lessons learned. The RTRs were categorized into two 

categories based on the RTR’s licensed thermal power. Twenty six RTRs licensed for less than 

2 MWt comprised Category 1, and the five RTRs greater than 2 MWt comprised Category 2. 

The assessment concluded that all of Category 1 and two 2 MWt research reactors from 

Category 2 were highly resilient to the loss of electrical power, active decay heat removal 

systems, and heat sink. The robust nature of these reactors is the result of their minimal decay 

heat generation – air cooling is sufficient to remove decay heat to prevent fuel failure. The three 

largest of the Category 2 reactors, two research and one test reactor, do not share the same 

level of resilience as the Category 1 reactors because of their reliance on water for adequate 

decay heat removal. NRC staff has concluded that additional assessment is needed to 

determine the resilience of the primary coolant system integrity to a beyond-design-basis 

seismic event for the research reactors and resilience of emergency power, active decay heat 

removal, and coolant make up systems to flooding scenarios and to a beyond-design-basis 

seismic event for the test reactor. However, the assessment of resilience for the three largest 

facilities is not urgent, because they do not have significant off-site consequences even under 

the most severe assumptions of damage. Such damage was assumed in the post-9/11 security 

assessments of multiple sabotage scenarios that considered the potential radiological 

consequences resulting from the severe damage to reactor fuel and facility containment or 

confinement systems. 

II. Background 

Non-power reactors (NPRs) are designed for many purposes, including commercial, research, 

testing, and education. These reactors are not used to produce electricity, process steam, or for 

desalination. All 31 operating NRC-licensed NPRs are used for research, testing, or education 

in nuclear engineering, physics, chemistry, biology, anthropology, medicine, materials sciences, 

and related fields and are all classified as RTRs. Among the RTRs that the NRC has licensed, 

the range of licensed thermal power varies from five watts to 20 megawatts-thermal (MWt). 

There are also eight RTRs shut down and in various stages of decommissioning. This 

assessment considered only the 31 RTRs with current operating licenses. The eight RTR 

facilities in the decommissioning phase are included in the assessment described in Section 7.

The predominant radiological hazard associated with RTRs is radiological exposure from the 

mishandling of radioactive materials or experiments by operators or researchers. These 

materials or experiments may or may not contain nuclear fuel and can be located inside or 

outside the core in experimental facilities. The neutron and gamma ray beams, as well as 

byproduct materials produced by the activation of non-radioactive materials, may present 

radiological hazards. These materials typically do not present a hazard to the public due to the 

licensing limitations for experiments. The greatest hazard by far to the public results from the 

accumulation of fission products resulting from the fission of the uranium fuel. These 

radioactive materials are normally contained within the fuel cladding. In most cases, the reactor 

fuel is inside the core within pools of water to shield personnel from radiation and to remove 

heat produced from nuclear reactions. The reactors are housed within confinement or 
82 

containment buildings. Therefore, an uncontrolled release of radioactive material could happen 

only in a case where both a loss of the fuel cladding integrity and the failure of the confinement 

controls or the containment barriers occur simultaneously. 

The nature of the hazard and the mechanisms necessary for the accidental release of 

radioactive material from an RTR is similar to that of a power reactor. The difference between 

the current NRC-licensed RTRs and power reactors is the magnitude of their potential accident 

consequences. The difference in consequences results from the RTR’s significantly different 

operating characteristics including maximum operating power level, temperatures, pressures, 

and the duration and frequency of operation. These operating characteristics result in 

significantly smaller fission product inventories for RTRs because of the significantly smaller 

quantity of fuel at the facility. To put these differences into perspective, consider the following: 

(1) the maximum licensed thermal power level of the smallest power reactor is about two orders 

of magnitude greater than that of the largest RTR; (2) the duration of operations of a power 

reactor is nearly continuous, while most RTRs are operated only periodically, typically a fraction 

of the normal work week; and (3) the onsite accumulation of significant quantities of spent fuel 

(20 to 40 years’ worth) in fuel pools at power reactors as compared to little or no spent fuel at 

RTRs. These factors contribute to a significantly smaller nuclear material inventory, accident 

source term, and demand for active decay heat removal at RTRs. 

III. Licensing 

The NRC’s authority to license and regulate NPRs is provided in Sections 103 and 104 of the 

Atomic Energy Act (Act). Section 103 of the Act pertains to the licensing of industrial or 

commercial reactors that can consist of both power and NPRs. Section 104 of the Act relates to 

the licensing of NPRs for the purpose of medical therapy and research and development. All 

RTRs currently licensed by the NRC are licensed under Section 104 of the Act. Unique to this 

authority are the provisions contained in Paragraph 104c of the Act that directs the “Commission 

to impose the minimum amount of such regulation and terms of license that will permit the 

Commission to fulfill its obligation under this Act to promote the common defense and security 

and to protect the health and safety of the public with the intent to permit the conduct of 

widespread and diverse research and development.” 

RTRs have been licensed under 10 CFR Part 50, “Domestic Licensing of Production and 

Utilization Facilities,” using the concept of defense-in-depth (DID). The concept of DID was 

applied at initial licensing to compensate for recognized uncertainties at the time (1950s and 

1960s) related to nuclear reactor design, operation, and consequences associated with potential 

accidents. As such, a comprehensive DID approach forms the foundation for the design and 

licensing of all RTRs. Even with the accumulation of many reactor-years of operating 

experience and the development of more advanced analytical capabilities for the assessment of 

safe reactor operation and reactor accident consequences, the concept of DID remains as a 

relevant and effective means to address uncertainties. 

Also, the implementation of 10 CFR Part 50, as it applies to RTRs, has been achieved using 

only deterministic methods and acceptance criteria. Inherent in these methods and criteria are 

the inclusion of highly conservative safety margins. 
83 

Similar to power reactors, a set of RTR licensing-basis events were established that are 

intended to ensure conservatism in design and protection from a wide spectrum of postulated 

accidents. Those accidents are highly stylized and do not consider multiple failures of safety 

systems. Qualitative approaches for ensuring reliable safety systems, such as the single failure 

criterion, were put in place. Testing plans and operational limits are established in technical 

specifications to ensure a high degree of confidence that safety systems would accomplish their 

designed safety functions if called upon during an accident. 

It is common that the analysis of a set of postulated accidents for RTRs do not result in a 

radiological release. In order to assess the dose impact to the public, an incredible but 

hypothetical event that results in a radiological release is assumed to occur. This event must 

bound all the credible hazards resulting from the postulated accidents and is referred to in the 

siting and licensing of RTRs as the maximum hypothetical accident (MHA). The MHA assumes 

a failure of the fuel or a fueled experiment that results in radiological consequences (a release 

of radioactive material) that exceed those of credible accidents. The MHA is not expected to 

occur; therefore, only the potential consequences are analyzed and not the initiating event or 

scenario details. Guidance for the licensing of RTRs is provided in NUREG-1537, “Guidelines 

for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors,” (Volume 1, 

“Format and Content,” and Volume 2, “Standard Review Plan and Acceptance Criteria”). 

Licensing acceptance criteria allowing construction and permitting operation of a research 

reactor (not a test reactor) are based on meeting conservative assurance that the public and 

occupational dose resulting from the radiological release assumed in the MHA remain bounded 

by the normal (non-accident) occupational and public dose limits contained in 10 CFR Part 20, 

“Standards for protection against radiation.” When the occupational or public dose resulting 

from the release from a research reactor exceeds the occupational and public dose limits of 

Part 20, the staff conducts additional review to ensure that the supporting MHA analysis is 

based on conservative assumptions. 

The siting and licensing of a test reactor also requires the development of an MHA from which a 

postulated radiological release is calculated. For the test reactor only, the public and 

occupational dose resulting from the release must meet 10 CFR Part 100, “Reactor site criteria.” 

The effects of external events are also considered specifically in the RTR licensing process. 

NRC-licensed RTRs are required to demonstrate, in their design, reasonable assurance that 

external events would not prevent safe operation and shutdown of the reactor. They must also 

demonstrate that provisions are included to mitigate or prevent an uncontrolled release of 

radioactive material. The consequences of external events are considered and bounded by 

analyzed accidents, particularly the MHA. The traditional licensing approach for the 

consideration of external events for RTRs relies on documented historical averages and 

extremes, credible event frequencies, and predictive potential for the specific external events. 

At a minimum, each research reactor facility is required to meet the local building codes for the 

specific type of event (e.g., seismic, flooding) and test reactors are required to meet the 

requirements of 10 CFR Part 100. 
84 

IV. Inspection and Oversight 

NRC staff inspects each facility periodically to ensure that licensees safely conduct regulated 

activities and maintain their facility in compliance with regulatory requirements. The NRC 

employs a graded inspection program for operating RTRs with two separate levels of inspection 

based on the maximum licensed power level of the facility. The scope of the inspection 

program for RTRs licensed to operate at power levels of 2 MWt or greater, is more 

comprehensive and completed annually, whereas the inspection program for RTRs licensed to 

operate at power levels below 2 MWt is completed every two years. Inspection programs for 

RTRs include: organizational structure, qualifications and responsibilities, operational activities, 

design and design control, review and audit functions, radiation and environmental protection, 

operator requalification, maintenance and surveillance activities, fuel handling, experiments, 

procedures, emergency preparedness, and safeguards and security. Other inspections, beyond 

those required by the routine inspection program, are completed as needed (e.g., in response to 

an event). 

Those reactors that are shut down but are not being decommissioned have an abbreviated 

inspection program completed triennially. NRC inspects decommissioning reactors to verify 

their safe condition and the safe conduct of dismantlement and decontamination. 

V. Emergency Preparedness 

Emergency planning considerations for RTRs and power reactors are similar. The substantially 

smaller accident source term of the currently-licensed RTRs results in a significant reduction of 

potential radiological consequences of an accident. RTR licensees are required to identify the 

area for which emergency planning is performed. This area is defined as an Emergency 

Planning Zone (EPZ). The EPZ27 comprises the area defined by emergencies that present 

potential radiological consequences that can result in off-site plume exposures that exceed 10 

milliSieverts (mSv) [1 rem] deep dose or 50 mSv [5 rem] to the thyroid. For all currently 

licensed RTRs, the EPZ boundaries are established well within an area under the control of the 

licensee. For example, for 26 of the RTRs (those less than 2 MWt), the EPZ is the operational 

boundary28 and the EPZ of the highest thermal power NRC-licensed RTR (20 MWt) is a 400 

meter radius surrounding the facility. 

RTRs also carry out an emergency classification scheme equivalent to that used at power 

reactors. The four emergency classes (notification of unusual event, alert, site area emergency, 

and general emergency) defined in 10 CFR Part 50, Appendix E, “Emergency planning and 

preparedness for production and utilization facilities,” are used to classify RTR emergencies. 

However, the classification of general emergency is not used in emergency plans for any of the 

RTRs licensed because the postulated radiological consequences of the MHA at each RTR 

 

27 ANSI/ANS-15.16-2008, “Emergency Planning for Research Reactors” 28 The operational boundary is the area within the site boundary, such as the reactor building, where the 

chief administrator has direct authority over all activities. 
85 

does not meet the criteria requiring a general emergency29 declaration at the site 

boundary30 (offsite). 

VI. International Assessment of RTRs of the Fukushima Dai-ichi Accident

The NRC RTR staff has participated with our international partners in developing an 

understanding of the lessons learned from the Fukushima accident. This included work with 

both the Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA) on 

the development of international assessment guidance which included the application of a 

graded approach where appropriate. This work was deemed necessary because the stress test 

method initially used in Western European post-Fukushima assessments was designed 

specifically for power reactors and was applied to some of the European research reactors. 

Based on presentations at technical meetings from countries that applied the method to their 

research reactors, the stress test method was found to be much more beneficial in providing 

safety insight for the higher powered research reactors than for those of lower power. 

Application of the stress test method did not include grading of the assessment of the lower 

powered research reactors which resulted in the application of significant assessment resources 

that returned a diminished safety benefit.31 

The NRC did not conduct stress tests for either power or non-power reactors. Instead, for 

power reactors, the staff followed the recommendations of the Near-Term Task Force, which 

are substantially addressing the same issues. For RTRs, the staff developed and followed the 

NRC assessment guidance for facilities other than power reactors. This guidance was 

compared to IAEA’s Safety Report Series No. 80, “Safety Reassessment for Research Reactors 

in Light of the Accident at the Fukushima Dai-ichi Nuclear Power Plant,” following issuance and 

found to be generally consistent. 

International meetings, both at the NEA and IAEA, have been held to present the results of 

Member State reassessment of research reactor safety. Most countries have not fully 

implemented the draft guidance provided in IAEA Safety Report Series No. 80, opting only to 

revisit the seismic analyses of record for their research reactors. A few countries (mostly 

Western European) have met the intent of the IAEA guidance. NRC staff’s assessment to date 

is that U.S. RTRs, while not fully compliant with IAEA Safety Report Series No. 80, do meet the 

intent of the document. 

VII. Pre-Existing RTR Assessment Information Useful to the Assessment of the 

Fukushima Accident

Because of the terrorist attacks on September 11, 2001, the security at NRC-licensed facilities 

received increased attention. On September 28, 2001, the NRC Chairman directed NRC staff 

to undertake a thorough review of NRC’s safeguards and security programs. The focus of this 

review was to examine basic assumptions underlying the regulatory framework and to find any 

 

29 10 CFR Part 50 Appendix E 

30 The site boundary is that boundary, not necessarily having restrictive barriers, surrounding the 

operations boundary wherein the reactor administrator may directly initiate emergency activities. 31 European Research Reactor Conference 2012. 
86 

necessary changes to optimize NRC, licensee, local, State, and Federal response capabilities. 

In carrying out this directive, NRC staff completed an integrated and comprehensive security 

assessment to determine if physical security improvements were warranted at the various types 

of NRC-licensed facilities, including RTRs. 

The 2006 RTR security assessments focused on theft and sabotage scenarios at each RTR. 

These scenarios were chosen as they were viewed as having the potential to adversely affect 

public health and safety. The postulated consequences of these scenarios were compared to 

NRC-established radiological consequence screening criterion to determine if adequate 

protection of public health and safety is maintained. NRC staff employed a three-phase, 

assessment process that the Commission had approved.32 The 2006 security assessments are 

not publically available because they contain sensitive security-related information. 

The post-9/11 RTR security assessments concluded that the security posture at NRC-licensed 

RTRs remains adequate to promote the common defense and security and protect public health 

and safety. The extension of security assessment conclusions related to the analysis of 

sabotage scenarios can, at a minimum, offer useful insights to assist in the assessment of the 

consequences associated with a beyond-design-basis external event because both may result 

in severe damage states to the facility that may exceed those assumed in RTR’s design and 

accident analysis. The post-9/11 security assessments considered the potential radiological 

consequences for the various sabotage scenarios. The supporting analyses of the radiological 

consequences demonstrated that RTR sabotage scenarios would result in doses to the public 

that are fractions of the 10 CFR Part 100 reactor siting dose criteria.33 NRC staff considered the 

sabotage event assessment and its conclusions as part of the beyond-design-basis external 

event assessment and determined that the sabotage assessment conclusions could not be 

extended to the beyond-design-basis external event case without additional site specific 

assessment. 

VIII. Prompt Post-Fukushima Assessments 

In the days immediately following the Fukushima accident, NRC staff collected available 

information related to accident initiation, progression, and consequences. The NRC RTR staff 

used this information to inform a prompt assessment of the safety of NRC-licensed RTRs. The 

goal of the assessment was to determine if insights gained from the accident revealed any 

conditions or lessons learned that would call into question the safety of the NRC-licensed RTRs 

or reveal the need for immediate regulatory action. NRC staff conclusions were based on the 

available information and engineering judgment. The areas considered by NRC staff during the 

prompt assessment included: 

• natural events 

 

32 SRM SECY-04-0222, dated January 19, 2005 33 Per cited regulations, an individual located at any point on the exclusion area boundary for two hours 

immediately following onset of the postulated fission product release would not receive a total radiation 

dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid 

from iodine exposure. 
87 

• electrical power 

• decay heat removal 

• spent fuel 

• combustible gas control 

• reactor containment/confinement. 

Natural Events

The licensing process for RTRs requires that the applicant describe and discuss the 

geographical, geological, seismological, hydrological, and meteorological characteristics of the 

site. This information must be presented in sufficient detail to allow NRC staff to compare the 

site’s potential natural hazards to the applicant’s proposed facility design. This allows NRC staff 

to decide on the acceptability of the proposed design at that site. To support such a conclusion, 

the design must provide reasonable assurance that structures, systems and components will 

remain capable of performing safety functions during and after postulated natural events. 

After the Fukushima Dai-ichi accident, the staff reviewed the natural event descriptions and 

discussions each RTR licensee supplied in licensing documentation. NRC staff performed the 

first of these assessments the day of the accident. The staff specifically assessed the designs 

of those RTRs in Pacific coastal states in light of the predicted tsunami forecasted to impact the 

Pacific coast of the United States. The staff concluded that none of the five RTRs in Pacific 

States were vulnerable to the tsunami reaching the locations of the facilities. 

In the weeks after the accident, NRC staff started a broader assessment of natural events 

originally considered in the design bases for the NRC-licensed RTRs. A specific focus was 

placed on the seismic and flooding events. NRC staff reached the following general conclusion 

related to the current design and siting of the NRC-licensed RTRs: 

• The treatment of natural events for each NRC-licensed RTR is reviewed at the time of 

licensing and considers the average and extreme (historical worst case) values as well 

as predictive potentials (where appropriate and available) in the assessment of the 

potential hazards presented by a specific natural event. 

• Because of their low power, NRC-licensed RTRs do not rely on large sources of water 

for makeup and heat sinks which allows siting of RTR facilities a significant distance 

from potential sources of flooding. 

• Each research reactor licensee has demonstrated through various analyses that the 

radiological consequences associated with the maximum predicted seismic event at the 

facility site, using information available at the time of licensing, meets accident analysis 

acceptance criteria which have been based on highly conservative public and 

occupational dose limits. (RTRs do not update their seismic analyses with the latest 

seismic hazards information unless undergoing relicensing) 

• Similar to power reactors, the one test reactor currently licensed by the NRC must meet 

10 CFR Part 100 seismic requirements and radiological limits. 
88 

Electrical Power

Given the Fukushima scenario, NRC staff included a review of an extended loss of electrical 

power on the safety of NRC-licensed RTRs as part of their immediate assessment. NRC staff 

reached the following conclusions related to the NRC-licensed RTRs’ reliance on electrical 

power and sensitivity to its loss: 

• Most RTRs have some level of emergency power, typically to power area radiation 

monitors, evacuation alarms and lighting, and security systems. None require electrical 

power (normal or emergency) to safely shut down the reactor. 

• 28-out-of-31 RTRs are air-coolable and thus do not require electric power for decay heat 

removal. 

• 2-out-of-31 RTRs may require ac power to replenish the reactor coolant inventory lost 

because of seismically-induced failure of the reactor coolant system boundary (pool or 

tank) or boil-off. 

Decay Heat Removal

At Fukushima it was ultimately the inability to remove the decay heat34 from the reactor cores 

that resulted in catastrophic failures. The generated decay heat is not unique to power reactors 

and is also generated in the cores of RTRs. The major differences between power reactors and 

RTRs are the magnitude of their maximum operating power, the duration of operation, and the 

smaller quantities of spent or irradiated fuel onsite. These qualities equate to a significantly 

smaller fission product inventory and a significantly reduced decay heat generation rate at 

RTRs. These along with other factors make RTRs much less susceptible to core damage from 

overheating when compared to power reactors and generally eliminate the need for highly 

complex, diverse, and redundant active decay heat removal systems typical of those found at 

power reactors. For example, for the NRC-licensed RTRs that are less than 2 MWt, decay heat 

can be adequately removed through air cooling of the core.35 

Given the Fukushima scenario, NRC staff included a review of the decay heat removal 

capability as part of the immediate assessment. NRC staff reached the following conclusions 

related to the decay heat removal at NRC-licensed RTRs: 

 

34 The majority of the decay heat generated is attributable to the decay of the radioactive fission products. 

After reactor shutdown, decay heat is about 6.5 percent of the previous operating power and decreases 

over time; 1.5 percent after 1 hour, 0.4 percent after 24 hours, and 0.2 percent after one week. Ref. DOEHDBK-1019, “DOE Fundamentals Handbook – Nuclear Physics and Reactor Theory,” January 1993. 35 S. Hawley and R. Kathren, Pacific Northwest Laboratory, 1982. “NUREG/CR-2387, Credible Accident 

Analyses for TRIGA and TRIGA Fueled Reactors.” 
89 

• Natural convection of primary coolant provides adequate decay heat removal for all RTR 

designs in the short-term (0.5 to 2.5 hours). 

• The one test reactor licensed by the NRC requires active systems for adequate longterm decay heat removal. 

• In cases of loss of coolant scenarios, air cooling is sufficient to remove decay heat for all 

RTRs with maximum licensed power levels of less than or equal to 2 MWt (28 of the 31 

RTRs). 

• Loss of coolant scenarios in the greater than 2 megawatt reactors (3 of the 31 RTRs) 

rely on designs that: 

− Maintain floodable volume that maintains the core covered with water at those 

facilities. 

− Provide the capability to spray the core with diverse sources of water via active 

(pumped) and/or passive (gravity drain) means. 

− Are equipped with emergency power systems or batteries sufficient to power 

cooling and makeup systems for the time required following reactor shutdown. 

Spent Fuel Pool Cooling

Given the public interest in spent fuel pools following the Fukushima accident, NRC staff 

included a review of the storage of irradiated and spent fuel at RTRs. NRC staff reached the 

following conclusion: 

• All RTRs have very small inventories of spent or irradiated fuel. 

• At all but the three largest RTRs, spent fuel is not routinely discharged due to low power 

and limited operational duration and frequency (practically speaking, the majority of the 

less than 2 MWt RTR reactor cores can be considered lifetime cores). 

• The Department of Energy (DOE) recovers spent and/or unwanted irradiated fuel from 

28 of 31 RTRs owned by government or academic institutions which includes the largest 

generators of spent fuel preventing the accumulation of large on-site spent fuel 

inventories. 

• Typically, spent, irradiated and excess fuel is stored dry, in the reactor pools, or in 

dedicated spent fuel pools when provided by design. 

• In no case does the storage of spent or irradiated RTR fuel require active cooling to 

remove adequate decay heat. 
90 

Combustible Gas Control

Significant damage resulted to the primary and secondary containment structures at Fukushima 

Dai-ichi due to ineffective control of hydrogen. Given the Fukushima scenario, NRC staff 

included a review of hydrogen generation and control at RTRs. NRC staff reached the following 

conclusions: 

• Low-power RTRs (those that are licensed to operate at less than or equal to 2 MWt) do 

not generate sufficient quantities of hydrogen from the radiolytic decomposition of water 

to reach combustible or explosive concentrations of hydrogen given the volume and 

mixing of the hydrogen with the reactor building atmosphere. 

• RTRs do not inject hydrogen into the primary coolant to scavenge dissolved oxygen. 

• The higher-powered RTRs either prevent the formation of combustible or explosive 

concentrations of hydrogen through the use of a dedicated hydrogen control system or 

have demonstrated by analysis that the formation of combustible or explosive 

concentrations are not possible under normal operating and accident conditions. 

• Concerning the formation of hydrogen from metal water reactions during accident 

conditions: 

− The amount of reactive metal present in a RTR core is several thousand times 

less than the reactive metal contained in a power reactor core. 

− Only one RTR uses zirconium cladding and this reactor does not generate 

sufficient decay heat to reach the elevated cladding temperature necessary to 

initiate the zirconium-water reaction. 

− RTRs with stainless steel fuel cladding also lack sufficient decay heat to reach 

the elevated cladding necessary to initiate a metal-water reaction. 

− An aluminum-water reaction at an RTR with aluminum fuel cladding is also 

extremely unlikely due to the necessity to establish precise physical conditions 

related to the form of the aluminum (i.e., small droplets of molten aluminum) to 

initiate the reaction. 

Reactor Containment/Confinement

As discussed above, the Fukushima Dai-ichi accident resulted in catastrophic failure of both the 

primary and secondary containments at some of the affected units. These failures did not occur 

directly from either the earthquake or the tsunami but only because of the inability to remove 

decay heat at a sufficient rate. This led to the eventual failure of all barriers intended to prevent 

the release of radioactive materials. 

All but three of the RTRs are designed with confinement structures that are not pressure tight 

structures. They typically employ a ventilation system to maintain the reactor building at a slight 
91 

negative pressure and discharge through a controlled pathway with a filtered and elevated 

release point. 

Three research reactors use containment buildings that provide a controlled leakage boundary. 

The containment structures are equipped with both under and over pressure protection thereby 

preventing the failure of the containment building from excessive differential pressures. 

Given the insights gained from the Fukushima accident, NRC staff reviewed the containment 

and confinement structures used at RTRs to prevent or control the release of radioactive 

materials to the environment. NRC staff reached the following conclusions: 

• NRC-licensed RTRs operate at very low power, low temperature (well below the boiling 

point of water at atmospheric pressure), and low pressure (typically at or near 

atmospheric pressure) and as such the energy that must be controlled or dissipated 

during an accident is very low and orders of magnitude less than at a power reactor. 

• Energetic releases from the reactor or reactor coolant systems to the containment or 

confinement are not expected to challenge the design limits during normal operating, 

transient, or accident conditions due to low operating temperatures and pressures, large 

containment/confinement volumes, and relatively low decay heat generation rates. 

Prompt Assessment Conclusions

The RTR prompt assessment was completed within weeks of the Fukushima Dai-ichi accident. 

As discussed above, NRC staff concluded that there were no safety concerns revealed by the 

accident for which immediate actions were necessary, nor was any new information revealed 

that would contradict or invalidate assumptions used in the safety basis of any of the RTRs 

licensed by the NRC. 

IX. Assessment of Near-Term Task Force Report Recommendations to RTRs 

In days after the Fukushima Dai-ichi accident, the Commission directed NRC staff to review 

methodically the NRC’s processes and regulations to determine whether the NRC should make 

improvements to its regulatory system and to make recommendations to the Commission 

regarding policy direction. In response to the Commission’s direction, the Executive Director for 

Operations established the Near-Term Task Force (NTTF) to conduct a near-term evaluation of 

the need for agency action. 

On July 12, 2011, the NTTF issued their report, “Recommendations for Enhancing Reactor 

Safety in the 21st Century” (ADAMS Accession No. ML111861807). The report has 12 

recommendations. The focus of the NTTF’s review was specific to power reactors, but their 

recommendations can be generally considered for any type of licensed facility. NRC staff 

developed a review process that provided general guidance for the review of the 12 NTTF 

recommendations, external events, and other Fukushima lessons learned for licensed facilities 

other than power reactors. 
92 

In assessing the applicability of lessons learned from the Fukushima Dai-ichi accident to RTRs, 

NRC staff chose to first group the RTRs based on their licensed thermal power level, because a 

given facility’s susceptibility to core damage under conditions similar to those encountered 

during the Fukushima Dai-ichi accident correlate directly to thermal power and the decay heat 

generation rate. This resulted in two review categories. Category 1 included research reactors 

with a licensed thermal power level of less than 2 MWt, which included 26 of 31 NRC-licensed 

research reactors. Category 2 included the remaining four research reactors, and the one test 

reactor, all of which were licensed for thermal power levels of 2 MWt or greater. 

Using a 2 MWt threshold for categorizing RTRs was not arbitrary. Some analyses36 exist that 

demonstrate that air cooling is sufficient to remove decay heat from research reactors licensed 

for operation at less than 2 MWt after a complete loss of coolant (dry core). One analysis of 

research reactor plate-type fuel37 (commonly referred to as MTR-type fuel) concludes that 

sufficient decay heat removal by air cooling alone exist for reactors up to 3 MWt. More recent 

facility-specific accident and thermal hydraulic analyses conducted in support of highly-enriched 

uranium (HEU) to low enriched uranium (LEU) conversions, or renewed operating licenses at 

research reactors of less than 2 MWt, have confirmed those conclusions. 

Recent security-related studies and assessments also contribute to the justification of the 2 MWt

categorization criteria for RTRs. These studies and assessments did not focus on fuel failure 

due to inadequate decay heat removal but rather on intentional damage or destruction of the 

fuel. Of primary concern was the magnitude of the potential radiological consequences 

resulting from a sabotage event. The sabotage studies conducted as part of the regulatory 

basis for the rulemaking that added 10 CFR 73.60(f), “Additional requirements for physical 

protection at non-power reactors,” provided the initial justification for greater security 

requirements to protect against sabotage at RTRs rated at thermal power greater than or equal 

to 2 MWt. The sabotage study results were later expanded upon by the 2006 security 

assessments conducted in response to the terrorist attacks of September 11, 2001. These 

subsequent RTR security assessments reviewed the adequacy of the existing security 

framework for the protection of nuclear materials from theft and sabotage in light of the new 

threat. The review concluded that the existing protections against sabotage remained 

adequate. 

Assessment of External Events on RTRs 

NRC staff assessed the various external events listed in the staff guidance to determine if an 

assumed external event exceeding the magnitude of the external events analyzed during the 

licensing process would result in more severe radiological consequences than those postulated 

in accident analyses. The external events analyzed during the licensing of each RTR were 

based on the average and extreme historical values and predictive potentials for the specific 

type of event and location of the facility. 

 

36 S. Hawley and R. Kathren, Pacific Northwest Laboratory, 1982. “NUREG/CR-2387, Credible Accident 

Analyses for TRIGA and TRIGA Fueled Reactors.” 

37 C. Webster. “Water-Loss Tests in Water-Cooled and Moderated Research Reactors.” Nuclear Safety, 

Volume 8, Number 6, November-December 1967. 
93 

RTRs less than 2 MWt (Category 1 Research Reactors) 

NRC staff found that the risk of a significant release of radioactive material because of external 

events exceeding the severity of the external events considered during the licensing is very low 

for the research reactors licensed for operation at less than 2 MWt. As discussed above, the 

low thermal power rating of these research reactors results in a low decay heat generation rate 

such that air cooling of the fuel would prevent overheating of the fuel cladding even if the 

external event in question causes or happens concurrently with the complete loss of coolant, the 

loss of electrical power, and the loss of all active decay heat removal systems. 

The table below summarizes the external events reviewed by NRC staff, the impact on the 

facility and a summary of the basis for no additional action being needed for this class of RTRs. 

Effect of External Events on Category 1 Research Reactors 

External Event Potential Effect on the 

Facility 

Assessment 

Flood • Loss of all electrical power 

and any active heat removal 

systems. 

• Potential damage to facility. 

• Active decay heat removal 

systems and electrical power are 

not needed. Decay heat is 

removed by natural convection of 

pool water or by air cooling if pool 

inventory is lost. 

• The operation of these facilities is 

not reliant on large bodies of water 

and therefore not typically sited 

directly adjacent to them (with one 

exception)38 thereby reducing the 

susceptibility to flooding. 

Seismic • Loss of all electrical power 

and any active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Physical damage to reactor 

components (e.g., pool 

integrity, passive cooling 

features or structural 

components). 

• Active decay heat removal 

systems and electrical power are 

not needed. Decay heat removed 

by natural convection of pool 

water or by air cooling if pool 

inventory is lost. 

• Facilities in this category are 

insensitive to changes in the 

seismic hazard since they are air 

coolable in the event of a seismic 

induced loss-of-coolant accident. 

• Building code bases is to prevent 

building collapse. 

 

38 A 100 watt critical assembly is located directly adjacent to a river and has historically experienced 

flooding that resulted in minimal impact to the facility. While water would damage the facility, it will not 

have a detrimental effect on decay heat removal which can be achieved by air cooling alone. 
94 

External Event Potential Effect on the 

Facility 

Assessment 

• Loss of air cooling from debris 

obstruction highly unlikely. 

• Fuel cladding remains intact. 

• No radiological release expected. 

High Wind, Tornado, 

and Missiles 

• Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Below grade design or above 

grade reinforced concrete 

biological shield prevents impact 

damage to fuel. 

• Bounded by seismic assessment 

above. 

Lightning • Loss of all electrical power 

and active heat removal 

systems. 

• Bounded by flood and seismic 

assessment above. 

Snow and ice loads • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements/containments. 

• Physical damage to reactor 

components (e.g., pool 

integrity, passive cooling 

features or structural 

components). 

• Bounded by flood and seismic 

assessment above. 

Drought and 

Temperature 

Extremes 

None • Bounded by flood and seismic 

assessment above. 

Fire • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Physical damage to reactor 

components (e.g., pool 

integrity, passive cooling 

features or structural 

components). 

• Low combustible loading and 

compliance with fire codes at the 

reactor facility. 

• Below grade design or above 

grade reinforced concrete 

biological shield prevents direct 

heating of pool liner or the fuel. 

• Bounded by flood and seismic 

assessment above. 

Loss of Power • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

• Bounded by flood and seismic 

assessment above. 
95 

External Event Potential Effect on the 

Facility 

Assessment 

confinement due to loss of 

ventilation systems. 

Loss of Heat Sink • Require a reactor shutdown 

at most RTRs to prevent 

exceeding temperature 

limits. 

• Bounded by flood and seismic 

assessment above. 

RTRs of Greater Than or Equal to 2 MWt (Category 2 Research and Test Reactors)

Air cooling for the Category 2 RTRs can no longer be assumed to provide adequate decay heat 

removal based solely on the RTRs maximum licensed thermal power. The analysis referenced 

in footnote 37 concluded through loss of coolant testing that a research reactor operating at 

3 MWt for 1 week could lose coolant via a tank rupture without experiencing fuel clad failure. As 

a matter of conservatism, the NRC has limited consideration of air cooling to less than 2 MWt. 

More recent testing referenced in footnote 36 also confirmed that a TRIGA research reactor 

operating at 2 MWt when subjected to an instantaneous loss of coolant via a beam tube, piping, 

or tank rupture would not experience fuel cladding failure. Given the potential for increased 

decay heat generation at reactors licensed for a thermal power of 2 MWt or greater requires 

reliance on specific operating conditions such as the availability of reactor coolant, a heat sink, 

and electrical power become more important. 

The decay heat studies referenced previously in this paper assumed the reactor that operated 

continuously at its maximum licensed power of 2 MWt. If one considers a more realistic and 

representative power history for a specific reactor, some of the Category 2 research reactors 

would respond similarly to that of the research reactors in Category 1 to severe external events. 

Adopting the concept used in the 2006 RTR security assessments of calculating a facility 

effective power (FEP) will represent a more accurate approximation of the actual decay heat 

that must be removed to prevent fuel failure. The FEP for the two 2 MWt facilities produce less 

decay heat than for a 1 MWt research reactor (0.12 MWt and 0.81 MWt). Given the fractional 

use of their maximum licensed power over the life of these facilities, it can be demonstrated that 

decay heat can be removed adequately by natural convection of pool water or by air cooling in 

the event that coolant is lost without need for active decay heat removal systems and electrical 

power. Because of the similarities between the 2 MWt Category 2 reactors and the Category 1 

reactors, the radiological consequences postulated in their MHAs would bound external events. 

Therefore, NRC staff considers the analysis summarized in the above table for RTR less that 

2MWt to be applicable to the 2 MWt reactors. 

For the three remaining reactors in Category 2, FEPs exceeded 2 MWt indicating that air cooling 

would not be adequate to prevent fuel cladding failure. NRC staff’s focus on external event 

assessment of Category 2 RTRs will be on the 6 and 10 MWt, research reactors and the 20 MWt

test reactor only. 
96 

 Research Reactors Licensed for Greater Than 2 MWt

There are two high powered research reactors in Category 2. They include the University of 

Missouri at Columbia (MURR) and Massachusetts Institute of Technology (MITR) reactors. 

Both of these reactors are tank type reactors capable of removing adequate decay heat by the 

natural convection flow of the reactor coolant following a severe external event even if that 

event results in the loss of all electrical power and active decay heat removal systems. In this 

case, decay heat is not sufficient (given the passive heat sink) to raise the temperature of the 

water above bulk boiling. Therefore, for this scenario, there is not a near-term need to replenish 

the water around the reactor fuel lost by evaporation. It is only when the initiating external event 

also causes (or occurs concurrently with) a loss of primary coolant, a condition which would 

require the failure of the core tank and reactor pool integrity, do the conditions exist that result in 

inadequate decay heat removal. 

A type of severe external event that could result in conditions where there is a loss of coolant 

accident (LOCA) and concurrent loss of ac power is a seismic event. A seismic event would be 

required to exceed significantly the predictive potential for seismic activity in the areas where 

these reactors are located to cause the failure of the core tanks and the reactor pools. For 

example, the MITR is in an area where the USGS’s 2014 seismic hazard data for the reactor’s 

location predicts a maximum peak ground acceleration of 0.16 g. A seismic analysis referenced 

in the MITR Safety Analysis Report for the core tank demonstrates that it is capable of 

withstanding, at yield stresses, static forces corresponding to 5.1 g horizontal simultaneously 

with 3.4 g vertically. Based on this analysis, the licensee concluded that the seismically-induced 

loss-of-coolant accident is not credible. Considering the similarity between the designs of the 

MIT and MURR, it would be reasonable to assume that both will have seismic margins; 

however, the magnitude of that margin will vary depending on local seismic hazards. The staff 

plans to confirm that the seismic margins for these research reactors are such that a LOCA 

induced from a seismic event is not credible. If necessary, the staff may look at the time 

needed before core damage and inform the size and location of the break through structural 

evaluations. If the seismic margins are such that additional regulatory steps are necessary, the 

staff will follow the appropriate regulatory process to mitigate the potential for a seismicallyinduced LOCA concurrent with loss of ac power. 

NRC staff is also assessing the capability of other external events (e.g., missiles from high 

winds) resulting in a LOCA and concurrent loss of electrical power. Both of the research 

reactors in this category have containments and the designs are such that the reactor core is 

protected by above grade reinforced concrete biological shield that provide an additional barrier 

(in addition to the containment) that minimize the potential for impact damage to the pool liner, 

core tanks and fuel. As part of the seismic evaluation above, NRC staff will confirm that 

external events other than seismic events are not credible causes of a LOCA concurrent with a 

loss of ac power. 

For both the seismic evaluation assessment and the assessment of the capability of other 

external events to cause a LOCA and concurrent loss of ac power NRC staff will consider the 

results of the previous security assessments before other regulatory actions, if any, are 

considered. As part of their assessment of these events, NRC staff plans to gather the 
97 

appropriate structural expertise, supplemented by NRC staff members who have performed 

assessments of power reactor responses to NRC’s Order EA-12-049, to inform its assessment 

on whether additional regulatory action is needed. The table below summarizes the external 

events reviewed by NRC staff, the impact on the facility, and the staff’s assessment. 

Effect of External Events on MURR and the MITR Research Reactors 

External Event Potential Effect on the 

Facility 

Assessment 

Flood • Loss of all electrical power 

and any active heat 

removal systems. 

• Challenge to facility 

equipment and reactor 

structures including 

confinements or 

containments. 

• Decay heat is adequately 

removed by natural convection of 

pool water. Inundation by flood 

waters will undoubtedly damage 

facility equipment but it is not 

expected to interrupt decay heat 

removal through natural 

convection. 

Seismic • Loss of all electrical power 

and any active heat 

removal systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Physical damage to 

reactor components (e.g., 

pool integrity, passive 

cooling features or 

structural components). 

• The building codes are 

established to prevent building 

collapse and would limit the 

likelihood of catastrophic failure of 

structures. 

• The reactors in this class have 

better seismic capability than the 

Category 1 research reactors. 

• The research reactor decay heat 

is adequately removed by natural 

convection of the pool water. 

Active decay heat removal 

systems and electrical power are 

not needed. 

• If the seismic event either causes 

or occurs concurrently with the 

loss of coolant, these research 

reactors may require support from 

portable equipment if installed 

equipment cannot be promptly 

restored. NRC staff to continue 

assessment. 

High Wind, Tornado, 

and Missiles 

• Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Above grade reinforced concrete 

biological shield and containment 

wall may prevent impact damage 

to the reactor pool and fuel. 

• NRC staff to continue assessment 

to ensure that there is not an 

event-specific vulnerability. 
98 

External Event Potential Effect on the 

Facility 

Assessment 

Lightning • Loss of all electrical power 

and active heat removal 

systems. 

• Decay heat is adequately 

removed by natural convection of 

pool water. Active decay heat 

removal systems and electrical 

power are not needed. 

Snow and ice loads • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Bounded by seismic and high 

winds assessment. 

• NRC staff will continue 

assessment to ensure there is not 

an event-specific vulnerability. 

Drought and 

Temperature 

Extremes 

• It may be necessary to 

shut down the reactors 

based on an inability to 

continue to meet Technical 

Specifications. 

• Active decay heat removal 

systems and electrical power are 

not needed. 

Fire • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Low combustible loading and 

compliance with fire codes at the 

reactor facility. 

• Fire-induced LOCA highly unlikely. 

In the event of a fire the biological 

shield and water in reactor core 

tank and reactor pool will cool 

these components such that fireinduced LOCA is not credible. 

Loss of Power • Loss of all electrical power 

and active heat removal 

systems. 

• Active decay heat removal 

systems and electrical power are 

not need. 

Loss of Heat Sink • Require a reactor 

shutdown at most RTRs to 

prevent exceeding 

temperature limits. 

• Active decay heat removal 

systems and electrical power are 

not needed. 

20 MWt Test Reactor 

The National Institute of Standards and Technology (NBSR) test reactor is protected initially 

from fuel cladding failure via a passive coolant makeup system combined with natural 

convection cooling following the loss of active decay heat removal capability. Specifically, the 

inner reserve tank will make up the inventory boiled off during the first half hour. The 

emergency tank will continue to replenish the inventory for an additional 2 hours. As the 

inventory of coolant contained in the passive makeup system becomes depleted, the pool 

coolant level boils down until the fuel becomes uncovered and fuel temperatures start to 

increase to a point where fuel cladding can fail if the facility’s light water makeup water source to 

the core or electrical power and a means of decay heat removal are not restored to operation or 
99 

otherwise provided via portable equipment in a timely manner. This portable equipment can 

allow domestic light water to be supplied to the reactor following installation of a spool-piece. 

Emergency power to the active decay heat removal system can be supplied by the 125 volt 

direct current station batteries for a limited time or from one of the two on-site emergency diesel 

generators for a longer period of time. In the event of a LOCA, the heavy water can be 

collected in sumps and then pumped back into the reactor for decay heat removal; however, this 

requires ac power. The test reactor fuel cladding is vulnerable to failure to beyond-design-basis 

seismic events that potentially result in an extended loss of electrical power, active decay heat 

removal systems, or coolant inventory makeup capability. 

The radiological consequences resulting from a severe external event may exceed those 

assumed in the MHA but would not exceed 10 CFR Part 100 siting criteria used in licensing the 

facility. This has been confirmed by the post-9/11 security assessment of sabotage scenarios 

which assumed massive damage states to the facility. Because of the malicious intent and the 

extreme assumptions of facility damage used in the sabotage assessment, the postulated 

radiological consequences from the worst case sabotage event are expected to bound the 

postulated radiological consequences of all external events. The radiological consequences 

predicted by the worst case sabotage event analysis are a fraction of the 10 CFR Part 100 

reactor siting dose criteria. 

Nevertheless, NRC staff plans to review the seismic capability for this reactor to determine 

whether an extended loss of ac power (both offsite and loss of on-site emergency diesel 

generators) from a seismic event is credible. If the seismic assessment’s results are such that 

additional mitigative capabilities are necessary, NRC staff will follow the appropriate regulatory 

process to mitigate the potential for a seismically-induced extended loss of ac power. 

Flooding at the test reactor is not a likely external event. No major bodies of water are near to 

the facility and the facility is above the 500 year flood plain. Because no major sources of water 

are near the facility, there are no water control structures, such as dams, dikes or levees subject 

to seismically induced failure which could result in the flooding of the test reactor. Records 

show flooding from local intense precipitation (LIP) events is extremely rare. The flooding of the 

facility from a LIP event which causes the extended loss of all electrical power and active decay 

heat removal is also not of concern because such an event would not adversely impact the 

availability of heavy and light water makeup sources to the core. In the long-term, makeup 

water can be replenished though connections to the city water system, which do not require 

electrical power. NRC staff plans to continue its assessment of flooding events to determine if 

additional regulatory actions are needed to address this event. 

NRC staff is also assessing the capability of other external events (e.g., missiles from high 

winds) causing an extended loss of all (offsite, onsite, and emergency sources) electrical power. 

As part of its assessment of these events NRC staff intends to gather the appropriate structural 

and flooding expertise, supplemented by NRC staff members who have assessed power reactor 

responses to NRC’s Order EA-12-049, to inform its assessment on whether additional 

regulatory action is needed. The table below summarizes the external events reviewed by NRC 

staff, the effect on the facility and the NRC staff’s assessment. 
100 

Effect of External Events on NBSR 

External Event Effect on the Facility Assessment 

Flood • Loss of all electrical power 

and any active heat 

removal systems. 

• Challenge to facility 

equipment and reactor 

structures including 

confinements or 

containments. 

• Passive coolant makeup system 

removes decay heat for 2.5 hours. 

• City-water backup can supplement 

makeup coolant inventory after 2.5 

hours, if available. 

• Bounded by seismic results 

Seismic • Loss of all electrical power 

and any active heat 

removal systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Physical damage to 

reactor components (e.g., 

pool integrity, passive 

cooling features or 

structural components). 

• Test reactor meets 10 CFR Part 

100 siting criteria. 

• The test reactor decay heat is 

adequately removed by natural 

convection of reactor vessel water 

for a minimum of 0.5 hours. An 

additional 2 hours of makeup water 

is provided via the emergency tank 

if it survives the seismic event. 

Continued adequate decay heat 

removal will require either a source 

of makeup water or AC power and 

active decay heat removal. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability 

High Wind, 

Tornado and 

Missiles 

• Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• The confinement building and the 

biological shield provide barriers to 

wind driven missiles. 

• The confinement building and 

reinforced concrete biological 

shield may prevent impact damage 

to reactor vessel and fuel. Subject 

of additional assessment. 

• Test reactor decay heat is 

adequately removed by natural 

convection of pool water initially. 

Continued adequate decay heat 

removal will require either a source 

of makeup water or active decay 

heat removal systems for the longterm. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability. 

Lightning • Loss of all electrical power 

and active heat removal 

• Test reactor decay heat is 

adequately removed by natural 
101 

External Event Effect on the Facility Assessment 

systems. convection of reactor vessel water 

initially. Continued adequate decay 

heat removal will require active 

decay heat removal systems in the 

long-term. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability. 

Snow and ice loads • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Test reactor decay heat is 

adequately removed by natural 

convection of reactor vessel water 

initially. Continued adequate decay 

heat removal will require active 

decay heat removal systems in the 

long-term. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability. 

Drought/ 

Temperature 

Extremes 

It may be necessary to shut 

down the reactors based on an 

inability to continue to meet 

Technical Specification. 

• Test reactor decay heat is 

adequately removed by natural 

convection of pool water initially. 

Continued adequate decay heat 

removal will require active decay 

heat removal systems in the longterm. 

• Long-term coolant makeup 

requirements very minimal for this 

reactor because of low decay heat. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability. 

Fire • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

structures including 

confinements or 

containments. 

• Low combustible loading and 

compliance with fire codes at the 

reactor facility. 

• Above grade reinforced concrete 

biological shield prevents direct 

heating of the reactor vessel and 

fuel. 

• Test reactor decay heat is 

adequately removed by natural 

convection of reactor vessel water 

initially. Continued adequate decay 

heat removal will require active 

decay heat removal systems in the 
102 

External Event Effect on the Facility Assessment 

long-term. 

• Staff will continue assessment to 

ensure there is not an eventspecific vulnerability. 

Loss of Power • Loss of all electrical power 

and active heat removal 

systems. 

• Challenge to reactor 

confinement. 

• Test reactor decay heat is 

adequately removed by natural 

convection of reactor vessel water 

initially. Continued adequate decay 

heat removal will require active 

decay heat removal systems in the 

long-term. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability 

Loss of Heat Sink • Require a reactor 

shutdown at most RTRs to 

prevent exceeding 

temperature limits. 

• Test reactor decay heat is 

adequately removed by natural 

convection of reactor vessel water 

initially. Continued adequate decay 

heat removal will require active 

decay heat removal systems in the 

long-term. 

• NRC staff will continue assessment 

to ensure there is not an eventspecific vulnerability. 

Assessment of NTTF Recommendations

Assessment Conclusions of NTTF Recommendations for RTRs <2 MWt

Regarding the NTTF recommendations associated with ensuring protection and enhancing 

mitigation, the research reactors in Category 1 are resilient to cladding failure due to decay heat 

induced overheating. They do not require electric power, active decay heat removal systems, or 

the presence of reactor coolant to adequately remove decay heat. They are small facilities that 

operate infrequently and at low power and as such present minimal radiological hazards to 

public health and safety. Confinement or containment structures are not challenged by 

energetic releases from primary or secondary cooling systems or by hydrogen generation from 

metal-water reactions during accidents. The research reactors in this category do not generate 

significant quantities of spent fuel and the minimal quantities that exist can be adequately 

cooled by air. 

Regarding the NTTF recommendations associated with emergency preparedness, NRClicensed research reactors are not collocated with other reactors. Emergency preparedness 

response would be anticipated to be uncomplicated due to their simplistic design, low decay 

heat and small source terms. EPZs are small. Facility recovery and reentry following an 

accident would likely be successful through the implementation of uncomplicated manual 

actions and prudent health physics precautions and controls. 
103 

Assessment Conclusions of NTTF Recommendations for RTRs 

Greater Than or Equal to 2 MWt

Regarding the NTTF recommendations associated with ensuring protection and enhancing 

mitigation, the three highest power Category 2 reactors (MURR, MITR, and NBSR) are 

particularly sensitive to the availability of reactor coolant. The early loss of reactor coolant can 

result in failure of the fuel cladding and subsequent radiological release unless reactor coolant 

makeup can be provided from installed facility equipment or from portable external sources. If 

coolant remains available, all three of these reactors will initially be provided with adequate 

decay heat removal via natural circulation cooling. However, if there is an extended loss of 

electrical power to operate the active decay heat removal systems or damage to the active 

decay heat removal system that prevents its use, the one test reactor could require the recovery 

of those systems or be provided with coolant make up from portable external sources to prevent 

fuel cladding failure. Therefore, the NTTF recommendations related to these scenarios could 

prove beneficial, under rare circumstances, in preventing fuel cladding failure at the test reactor. 

As discussed in the section above regarding external events, NRC staff is continuing its 

assessment of the need for protective or mitigative strategies for these reactors. If NRC staff 

concludes that additional protective or mitigative strategies are appropriate, NRC staff will use 

existing regulatory processes to carry them out. 

Regarding the NTTF recommendations associated with enhancing mitigation for containments, 

the Category 2 reactors present an increased radiological hazard due to decay heat generation 

rates as well as a proportional increase in fission product inventory. Category 2 reactor 

confinement or containment structures are not challenged due to the absence of any significant 

energetic releases from primary or secondary cooling systems or by the generation of 

combustible or explosive concentrations of hydrogen from metal-water reactions during 

accidents. 

Regarding the NTTF recommendation associated with enhancing mitigation for spent fuel pools, 

the RTRs in this category generate small quantities of spent fuel that can be adequately cooled 

by air shortly after discharge from the core. Additionally, the DOE routinely recovers spent fuel 

from these facilities preventing the long-term accumulation of spent fuel on site. 

Regarding the NTTF recommendations associated with strengthening emergency 

preparedness, NRC-licensed research reactors are not collocated with other reactors. 

Emergency preparedness response would be anticipated to be uncomplicated due to their 

relative simplistic design, low decay heat and small source terms. EPZs are small and do not 

extend beyond the owner controlled area. Facility recovery and reentry following an accident 

would likely be successful through the implementation of uncomplicated manual actions and 

prudent health physics precautions and controls. 

X. Conclusions and Recommendations

Research Reactors of Less than 2 MWt (Category 1)

The reactors in this category represent most of the NRC-licensed research reactors (26 of 31) 

with a maximum licensed thermal power of 1.1 MWt. They do not rely on the availability of 
104 

electrical power, active decay removal systems, or coolant inventory to adequately remove 

decay heat and prevent fuel clad failure. Even if a loss of coolant results from, or occurs 

concurrently with, an external event, sufficient decay heat can be removed via air cooling of the 

core. This resiliency is attributable to short duration and low power operation typical of this 

category of reactor and results in a minimal fission product inventory and decay heat 

generation. 

RTRs of Greater Than or Equal to 2 MWt (Category 2)

The reactors in this category represent the highest powered of the NRC-licensed RTRs (5 of 31) 

with licensed thermal power levels from 2 to 20 MWt. NRC staff has not found other actions are 

needed for the two 2 MWt research reactors in Category 2. 

The three highest powered RTRs in this category do rely on the presence of a coolant inventory 

to adequately remove decay heat. The research reactors in this category are not reliant on 

active decay heat removal systems and electrical power to prevent fuel cladding failure; 

however, the test reactor is. After the depletion of the coolant from the test reactor’s passive 

coolant makeup system and heavy water storage tank, the test reactor would require electrical 

power and the active decay heat removal system or additional sources of coolant. 

As discussed above, NRC staff is performing additional assessments related to the peak ground 

acceleration required to result in the failure of the primary coolant system integrity for the two 

high power research reactors. As part of this assessment NRC staff will also assess the 

capability of other external events (e.g., missiles from high winds) causing an issue with primary 

coolant integrity. The goal of this assessment is to determine if it is necessary to include 

additional features that would prevent or mitigate the loss of primary coolant. If it can be shown 

that pool failure is not credible under these circumstances, then NRC staff is unlikely to pursue 

additional actions for these research reactors. 

It is important to note that the test reactor was sited in accordance with 10 CFR Part 100. As 

part of the licensing process, the test reactor applicant provided evidence with the analysis of 

postulated accidents, including the MHA, that the 10 CFR Part 100 reactor siting dose criteria 

were met. The 2006 security assessments performed in response to the terrorist attacks of 

September 11, 2001, used the 10 CFR Part 100 reactor siting dose criteria as criteria to justify 

that no additional security-related regulatory actions were necessary to fulfill the Commission’s 

obligations under the Atomic Energy Act39 for the RTR sabotage event. The Commission found 

that the 10 CFR Part 100 reactor siting dose criteria were appropriate as an upper dose limit for 

the potential radiological consequences resulting from sabotage events. The worst case 

sabotage event for the test reactor resulted in doses that were a fraction of the 10 CFR Part 100 

reactor siting dose criteria. The extent of facility damage assumed in the worst case sabotage 

scenario may bound the facility damage of beyond-design-basis external events. Considering 

 

39 The Commission is directed by Section 104c of the Atomic Energy Act to impose only such minimum 

amount of regulation of the licensee as the Commission finds will permit the Commission to fulfill its 

obligations under the Act to promote the common defense and security and to protect the health and 

safety of the public and will permit the conduct of widespread and diverse research and development. 
105 

the bounding nature of the worst case sabotage scenario, would require additional assessment. 

Given that assessment, it may be possible to conclude that the radiological consequences 

resulting from any severe external event will likely not exceed the test reactor licensing criteria. 

As discussed above, for the test reactor, NRC staff is performing additional assessments of the 

resiliency of the facility’s emergency power and active decay heat removal systems to a severe 

seismic and or other external event. If the additional assessment demonstrates that the 

emergency power and active decay heat removal systems remain capable to perform their 

functions following such events, NRC staff is unlikely to pursue additional actions. If the 

additional assessment concludes otherwise, that is, emergency power and active decay heat 

removal would not be available following a severe seismic or external event, then NRC staff will 

use the appropriate regulatory process to determine whether to implement additional features to 

prevent or mitigate these severe seismic or external events at the test reactor. 

The additional assessments for MURR, MITR, and NBSR will be performed by NRC staff with 

expertise in the areas of seismology, hydrology, structural analysis, and RTR design and 

operation. These assessments are expected to be limited in nature, will use existing budgeted 

resources, and will be prioritized considering other high-priority work that is being performed by 

these staff. Upon completion, NRC staff will provide a Commissioners’ Assistants Note 

presenting the results of the assessments and, if appropriate, a discussion of the work 

necessary to resolve any remaining concerns at these facilities.
Public Comments Received During March 13, 2015, Public Meeting to Discuss Staff’s Prelimi

Assessment of Applicability of Fukushima Lessons Learned to Facilities Other than Operating Powe

This enclosure provides a list of public comments received during the March 13, 2015, public meeting. The pubbroken into six major sections. In the first five sessions, the staff described its preliminary assessment of the Fulearned for the following facilities: • Session 1 consisted of the NRC staff’s overview of its preliminary assessment for spent fuel storage inst

transportation packages, and decommissioned reactors and complex materials facilities. 

• Session 2 consisted of the NRC staff’s overview of its preliminary assessment for fuel cycle facilities. • Session 3 consisted of the NRC staff’s overview of its preliminary assessment for radioactive materials u

• Session 4 consisted of the NRC staff’s overview of its preliminary assessment for low-level waste dispos

recovery facilities, and uranium mill tailings. 

• Session 5 consisted of the NRC staff’s overview of its preliminary assessment for research and test reacAfter each of the first five sessions the staff provided stakeholders the opportunity to ask questions and provide 

staff’s preliminary assessment. For session six the staff did not make any presentations and stakeholders were opportunity to ask questions and provide comments on the topics discussed in the staff’s preliminary assessmeThe following table provides a description of the comments received, the staff’s response to the comment, and, 

changes made to the staff’s assessment based on the comment. Each comment was provided a unique identifsession in which it was received. For example the first comment received in session 1 is provided the identificaand the sixth comment in session 5 is provided the identification number S5-6. The unique identifier can also be 13, 2015, meeting summary (Agencywide Documents Access and Management System (ADAMS) at Accession 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

S1-1 Donna Gilmore - Recommends the staff 

qualify the conclusion for the independent spent fuel storage installation (ISFSI) such that it is clear that it does not appl

y to the ISFSI 

The staff has processes in place to address aging maperiod and is actively involved in research activities reissues associated with spent fuel storage systems. Thassessment is based on casks meeting regulatory reqcasks meet the re

gulator

y requirements the staff has c
- 2 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

renewal period. Commenter referenced an August 5, 2014, meeting summary associated with potential cracking in dry shielded canisters (DSCs) over time and a concrete workshop that suggests degradation of concrete over time is possible. 

conclusion that no radiation-related death or injuries ocontamination is expected. As a result of the comment NRC staff added a discussstorage and transportation section that addresses the fuel storage system license and certificate renewal anapplicabilit

y to the renewal period.

S1-2 Donna Gilmore 

– Consequences of a 

DSC crack are not addressed. Inspections have not been developed to find such cracking. Paper should address both issues. 

The staff is continuing its research to ensure aging meaddressed. In the initial license term, there are fabricainspections and processes in place to ensure integrity the time it is put into service. In recognition that storabe in service into their renewal period, the NRC and inresearch to ensure potential aging mechanisms, inclucracking, are identified and that appropriate detection Staff anticipates that non-destructive inspection technwithin approximately five years. Staff finds these activensure that the ISFSIs continue to meet NRC regulatiperiod. The staff is also developing a report to evaluafuel aging which will be issued in the near term. NRC staff did not identify changes to the assessment comment. 

S1-3 Marvin Lewis 

– Paper does not 

recognize that Fukushima happened and that transportation accidents happen. 

The purpose of the staff’s assessment is to address leFukushima accident. The staff’s transportation assesthe draft white paper (see ADAMS Accession No. ML1NUREG-2125, “Spent Fuel Transportation Risk Assesrecognizes transportation accidents happen. NRC stchan

ges to the assessment as a result of this commen

S1-4 Marvin Lewis 

– NUREG 2125 uses 

optimistic assumptions. 

The staff believes that the assumptions in NUREG 21NRC staff did not identify changes to the assessment comment.

S1-5 Marvin Lewis 

– Meeting does not have 

adequate public participation.

The staff issued a press release on March 2, 2015, to stakeholders about the meetin

g and provided a draft o
- 3 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

23, 2015. 35 non-NRC personnel either registered for participated via phone in the meeting. NRC staff did nthe assessment as a result of this comment.

S2-1 Bob Link – Paper states that generic 

letter (GL) for fuel facilities will be issued in March 2015, and during the presentation the staff stated that GL will be issued in April. 

Paper changed to reflect expectations regarding issua

S3-1 George Rudy – Paper does not mention 

initiators associated with terrorist attacks or aircraft impacts. 

The Paper’s focus is for Fukushima lessons learned. aircraft impacts are addressed in separate analysis peresponse to the September 11, 2001, terrorist attack. has done extensive analysis for theft and diversion anadditional security requirements. The staff is continuosecurity threat and the current regulatory for security tadditional regulatory actions are needed. NRC staff did not identify changes to the assessment comment.

S4-1 Donna Gilmore – Does the earthquake 

map provided on slide 56 of the staff’s presentation (ADAMS Accession No. ML15070A530) take into account the new long-term forecast for California from the US Geological Survey issued on March 10, 2015? If the paper does not, should the assessment be changed to reflect the latest seismic information? 

No, the paper does not take into account the March 10earthquake forecast for California. The staff’s assessqualitative analysis of its regulations and the staff’s cothe licensing processes that ensure that facilities othereactors are robust and that systems structures and cnecessary to protect the public will function properly in earthquake. The staff continuously assesses new infooperating assessment programs to ensure that new inearthquake forecast for California are evaluated to detthe NRC has to take action to ensure public safety. NRC staff did not identify changes to the assessment comment. 

S4-2 Donna Gilmore – Paper does not 

address aging management and 

The staff’s assessment is based on the staff’s practicprocesses are periodically updated to address new inf
- 4 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

recommendation made to add assumptions used in the paper. 

renewed licenses, the staff looks at the effects of aginthat they are properly addressed before a license is reassumption is therefore based on licensees’ complianregulations and that this compliance will be maintainelevel waste (LLW) facilities, the NRC regulations addrof the public by ensuring barriers are in place to ensurAs a result of this and other comments NRC staff addSECY paper to discuss aging management of dry cas

S4-3 George Rudy – Should the Department 

of Energy’s Waste Isolation Pilot Project be addressed in the paper?

The Waste Isolation Pilot Project is not regulated by thoutside the scope of the paper. NRC staff did not idenassessment as a result of this comment.

S5-1 George Rudy – Did the research and 

test reactor assessment assume postaccident monitoring systems? 

No, the assessment did not consider the effects of Fuon the research and test reactors (RTRs) post-accidenRTRs are not required to have a post-accident monitooperating power reactors. They are required to have, equipment and instrumentation necessary for the safe and needed to maintain the facility safely shutdown. Wand instrumentation is required at a RTR is determineprocess for each facility. This includes nuclear instrumlevel and temperature indication to assess the conditioradiation monitoring systems to monitor radiation levereactor. The higher powered reactors have more facilinstrumentation including backup and emergency powheat removal systems. The staff did assess Fukushima lessons learned for Rpaper, concluded that additional assessment is neede3 RTRs that are greater than 2 megawatts in power. NRC staff did not identify changes to the assessment comment.

S5-2 Marvin Lewis – Did the scope of the The scope of the paper is limited to NRC-licensed RT
- 5 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

 paper include military and Department 

of Energy reactors? 

Act of 1954, as amended, does not give the licensing 

the NRC for the DOE and DOD reactor facilities. Thernormally licensed by the NRC. In special cases, the Nprovide licensing and oversight of a specific DOD faciestablishment of an appropriate legal agreement. For oversight of the Armed Forces Radiobiological Researesponsibility of the NRC and it was included in the asNRC staff did not identify changes to the assessment comment.

S5-3 Ruth Thomas – Does the staff consider 

Defense Nuclear Safety Board regulated facilities and DOE facilities?

The scope of the paper is limited to NRC – licensed R

to Comment S5-2 

S5-4 Ruth Thomas – The paper and 

presentation uses the term “very unlikely.” Is the term very unlikely defined in probabilistic terms? 

The use of the terms in the paper such as “highly unlikand is not based on a quantified probabilistic risk asseThe staff used highly conservative assumptions as doFor example, for RTRs less than 2 megawatts the stafloss of coolant and determined that air cooling alone wdecay heat and prevent fuel damage such that fission NRC staff did not identify changes to the assessment comment.

S5-5 George Rudy – Do RTRs have a no fly 

zone around them? 

No, RTRs do not have a no-fly zone around them. Altof the meeting the staff has done extensive review of based on lessons learned from the September 11, 200on the location of the RTR the staff performs additionaone of the RTRs is located next to a runway. The liceimpact event as one of the design base external evenadditional information regarding hazards associated wNRC staff did not identify changes to the assessment comment.

S5-6 Ray Lutz – Requested that audio file be Audio files from the March 13, 2025, meeting were ma
- 6 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

 made available. the NRC website. The files can be found at the follow

http://www.nrc.gov/reactors/operating/ops-experiencebriefing.html

S5-7 Ray Lutz – RTRs should be reviewed 

for effects to the public from terrorist attacks. 

The scope of the staff’s assessment is limited to FukuEnclosure 1 to this document, Section 8, Subsection Vdiscussion of the 2006 RTR security assessments and with the Fukushima lessons learned assessment. In asecurity assessment, the impact to the public has also sabotage studies and aircraft impact assessments. NRC staff did not identify changes to the assessment comment.

S6-1 George Rudy – Does the NRC 

coordinate Fukushima lessons learned with other Federal agencies so that other Federal agencies can take advantage of NRC’s work for reviewing their regulated facilities? For example, is the NRC coordinating with DOE regulated facilities? 

NRC shares Fukushima lessons learned and other opwith other Federal agencies and with the international agencies can take advantage of NRC’s work and the Nof their work. NRC staff did not identify changes to the assessment comment. 

S6-2 Ace Hoffman – NRC staff is exhibiting 

gratuitous ignorance as it applies to aircraft impacts and dry cask storage breaches and is not properly assessing these events. 

The staff has performed detailed security assessmentspent fuel storage systems. The staff has discussed tassessments with the appropriate licensee or vendor changes to the ISFSI design as a result of these evaluNRC staff did not identify changes to the assessment comment.

S6-3 Ace Hoffman - There is an absence of 

discussion of the need for prompt evacuation or for the impacts to the Price Anderson insurance protection in the event of an accident. 

The staff’s assessment does include discussions of wemergency plans are needed as a result of applying Flearned to facilities other than power reactors. The stachanges needing to be made in this area as a result oThe staff did not identify changes needing to be made 
- 7 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

protections based on the result of its assessment.

S6-4 Donna Gilmore – The Fukushima dry 

cask storage system is different than U.S. dry cask storage systems. Did the NRC consider what would happen to U.S. cask designs had they been at Fukushima? 

The staff evaluated the existing US NRC approved dry designs, grouped by common design characteristics, a

of applying Fukushima lessons learned to these facilitfuel storage cask in use at Fukushima would be most evaluated with common characteristics of vertical orieclosure. All spent fuel storage cask designs must meerequirements of 10 CFR Part 72. NRC staff did not identify changes to the assessment comment.

S6-5 Donna Gilmore – Humboldt Bay does 

not have air vents to remove heat because of low heat load. The San Onofre and Callaway underground design have air vents which would let water into the system. The staff should consider this. 

As documented on page 11 of the staff’s assessment, 

spent fuel storage systems in the following conditionswater, 2) partial flooding, and 3) blockage of the vents The results of the steady-state thermal analysis for a va horizontal cask system indicate that a release from texpected. NRC staff did not identify changes to the assessment comment.

S6-6 Donna Gilmore – What is the 

remediation plan if a DSC fails? 

NRC regulations are developed to ensure that contain

spent fuel storage system. The DSC is one componespent fuel storage system. The staff has performed sdetermined that the probability of a DSC failing and thconsequences are very low. NRC staff did not identify changes to the assessment comment.

S6-7 Donna Gilmore – The security 

assessments referenced in response to S6-6 should be provided to the public 

NRC staff continually balances the public’s need to knsame security assessment information available to postaff has provided classified briefings on the security aindividuals with a need to know. There is no plan to massessments publicly available at this time.
- 8 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

NRC staff did not identify changes to the assessment comment

S6-8 Ray Lutz – Presentation does not 

discuss Fukushima lessons learned. 

The staff is seeking comments not on the presentation paper that was provided as part of the meeting noticemeant to be a brief summary of the staff’s assessmenanalysis found in the draft white paper.

S6-9 Ray Lutz – Slide 12 from the meeting 

handouts states that Near Term Task Force Fukushima lessons learned number 2 regarding the need to reevaluate and upgrade design basis flood and seismic design basis events is not applicable. This does not appear to be accurate. 

The staff did evaluate flooding and seismic scenarios as documented in the draft white paper and found the Based on this comment the staff changed the recommapplicability for Fukushima item 2 from “Not applicable

S6-10 Ace Hoffman – Aircraft hazards and 

terrorist events should be considered by the staff. 

The scope of the paper is limited to Fukushima lessonnoted in response to comments S3-1, S5-7 and S6-2, security assessments for some of the facilities discuss

S6-11 George Rudy – A naturally generated 

missile or seismic event could cause a degraded DSC to fail, therefore, the staff needs a preplanned mitigation plan. 

The scope of this assessment was to evaluate the applearned from the event at Fukushima to NRC-regulatestakeholders, other than commercial power reactors. tornado missile events were assessed. Based on the 

staff did not identify needed changes to its current regISFSIs and therefore did not identify a need for a mitig

S6-12 Marvin Lewis – Do the spent fuel 

storage cask designs consider the enrichment of Uranium 235? 

Yes. The staff reviews and captures in technical speccritical fuel characteristics which include the enrichmespent in the reactor, and the condition of the fuel itselfNRC staff did not identify changes to the assessment comment.

S6-13 Marvin Lewis – 5/8 inch thick DSC are Spent fuel storage cask systems must meet the requir
- 9 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

 deficient. 72 to be authorized for use by the NRC. The DSC is o

the spent fuel storage system. The NRC has approveDSC designs that are approximately 1⁄2-inch thick. Thand stored in overpacks that protect the DSC from envas high winds and tornado borne missiles. In addition evaluated against a non-mechanistic tip over. The cademonstrate structural adequacy for this event. The sover assessment bounds the consequences of an earconcluded that the current regulatory requirements enstorage and transportation of spent nuclear fuel.

S6-14 Donna Gilmore – Commenter 

expressed concern that San Onofre is in the process of changing the spent fuel pool cooling system to a spent fuel pool island such that ocean cooling will not be needed. 

Specific changes to the spent fuel pool cooling system the draft white paper. The draft white paper documenof decommissioned reactors’ spent fuel pools to deterneeded based on Fukushima lessons learned. The drconcludes changes to the NRC’s oversight of these fa

at this time. Note: Subsequent to the meeting the Office of NucleaSan Onofre project manager followed up with the comadditional information in response to the commenter’s to the San Onofre spent fuel pool cooling system

S6-15 Donna Gilmore – In a previous 

discussion NRC indicated that reactor vessels that experienced inner wall cracking did not leak. Commenter is not concerned about water leaking from the DSC but rather gas leaking through DSC cracks. 

See response to comment S1-2

S6-16 Donna Gilmore – Concerned that 

environmental monitors are being removed from around ISFSIs. Decommissioned reactors and ISFSI should have real time environmental 

Based on the staff’s review of Fukushima lessons learidentify a need for real time environmental monitors fodecommissioned reactors. Limits for the release of ra

ISFSIs can be found in 10 CFR 72.104, “Criteria for raeffluents and direct radiation from an ISFSI or MRS.” 
- 10 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

monitors. requirements for ISFSIs can be found in 10 CFR 72.32

S6-17 Donna Gilmore – San Onofre ISFSIs 

are only 100 ft above the sea level. With global warming this is not enough elevation. 

The NRC continuously reviews operating experience iif changes to its oversight of licensees need to be madpage 26 of the draft white paper, for more than 30 yeaimplemented improvements in its emergency preparedresponse programs especially upon reviewing lessons natural disasters. The draft paper goes onto describe in the aftermath of Hurricane Katrina. The staff also notes that for the San Onofre ISFSIs to level rise of 100 ft, the effects of global warming would magnitude higher than currently predicted. 

S6-18 Donna Gilmore – Commenter 

concerned that NRC is not reviewing the San Onofre ISFSI before it is installed and will only perform inspections prior to its operation. 

Implementation of the ISFSI at San Onofre has gone athorough NRC review. NRC authorizes storage of speISFSI under two licensing options: site-specific license San Onofre, the licensee plans to use the general licethe general license provision, the licensee intends to ucask. An NRC approved cask is one that has undergoits safety aspects and has been found to be adequate the bounding site characteristics evaluated in the safelicensees using such approved casks must show that 

technical specifications can be met and document the by 10 CFR 72.212, “Conditions of general license issulicensee’s evaluation is subject to NRC inspection. 

S6-19 Donna Gilmore – NRC processes 

should be changed to require NRC approval prior to installing an ISFSI or making major changes to the spent fuel pool. 

The requirements for installing an ISFSI can be found evaluation requirements before a licensee can make a pool can be found at 10 CFR 50.59. These requiremeappropriate NRC review is performed before an ISFSI change to a spent fuel pool reaches a threshold for whis required.

S6-20 George Rudy – There is inadequate See response to comment S5-1.
- 11 - 

Comment Number

Person Providing Comment and Description of Comment

Response to Comment 

 post-accident consideration for research 

and test reactors. 

S6-21 Ray Lutz 

– The conclusion for the spent 

fuel storage section of the paper states that the no further regulatory action or study is necessary. This conclusion does not recognize that the staff needs to continuously assess its regulations in light of Fukushima lessons learned.

The conclusion in the spent fuel storage section of the 

reflects that no further regulatory action or study is neassessment of the lessons learned from the Fukushimcontinuously evaluates if its regulations and guidance in light of new pertinent information and will continue t

S6-22 Ray Lutz 

– ISFSIs should have a 

regulatory requirement to have Emergency Planning Zones. 

ISFSIs are required to have in effect an emergency plrequirements of 10 CFR 72.32 (for a specific licensed (general licensed ISFSI). The NRC believes that emerequirements in 10 CFR 72.32 are adequate for ISFSInotice for 10 CFR Part 72 (60 FR 32430, 32435) descfor not including Emergency Planning Zones for ISFSidentify chan

ges needin

g to be made in this area as a 

S6-23 Ray Lutz 

–

Assessment is deficient and 

needs to be redone because it does not provide enough of a discussion describing the Fukushima cask designs and compare them against the US desi

gns to identify lessons learned.

See response to comment S6-4.

S6-24 Donna Gilmore 

– NRC should require a 

license amendment for ISFSI changes or for chan

ges to the spent fuel pool.

See response to comment S6-19
Discussion of Email Comments Received Associated with the Staff’s 

Preliminary Assessment on the Applicability of Fukushima Lessons 

Learned to Facilities Other than Operating Power Reactors 

This enclosure provides a list of public comments received via email associated with the NRC staff’s preliminary applicability of Fukushima lessons learned to facilities other than operating power reactors. The 14 sets of ema

as a result of the March 13, 2015, meeting can be found in an enclosure to the meeting summary (Agencywide and Management System (ADAMS) Accession No. ML15077A410). 

The following table provides a synopsis of the email comment documented in the March 13, 2015, meeting sumresponse to the comment, and any changes made to the NRC staff’s assessment based on the comment. Each provided a unique identifier based on the origin of the comment. For example, the Missouri at Columbia Reseacomments are listed as the first set of email comments documented in the March 13, 2015, meeting summary. comment is provided comment number E1-1. The “E” designation is used to indicate that the basis for the commThis designation was used as an aid to distinguish the comments received verbally during the meeting and thos

Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

E1-1 Missouri at Columbia Research Reactor 

(MURR) - On page 69, the 3rd bullet under Electrical Power, “3-out-of-31 RTRs may require ac power to replenish inventory because of seismic activity or boil-off.” Assume by replenishing inventory you mean water inventory for the reactor and/or spent fuel pools? As it currently stands, MURR does not need electrical power to fill the pool in an emergency situation on a lowering of pool level. MURR has a dedicated water line with a manual isolation valve that will provide virtually an unlimited quantity of water to the 

Given that this comment is related to the immediate aby NRC staff following the Fukushima accident and wdetermination adequacy of a facility’s design basis asscommenter’s statement is correct under the assumed As a result of the comment NRC staff changed the refout-of-31 RTRs may require ac power to replenish the inventory lost as a result of seismically induced failure system or boil-off.” 
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

pool. 

E1-2 MURR - On page 71, the 5th bullet 

under Spent Fuel Pool Cooling, “Typically, spent irradiated and excess fuel is stored dry in the reactor pools, or in dedicated spent fuel pools provided by design.” Not sure what is meant by “Stored dry in the reactor pools”? How can you store something dry in a water pool? 

Comment is correct. Comma added to correct sentenc

E1-3 MURR - Page 71, 3rd bullet under 

Combustible Gas Control, “The higher powered RTRs have dedicated hydrogen control system that are operated during reactor operation; however, upon reactor shutdown, those systems are no longer required to prevent the formation of combustible or explosive concentrations of hydrogen.” MURR does not require a hydrogen control system. MURR attached response to a relicensing request for additional information regarding this subject and it has been accepted by the NRC. 

Comment is correct and the report has been modified 

E2-1 National Institute of Standards and 

Technology – NIST provided several pages of clarifying comments regarding the capabilities of the test reactor. 

The staff incorporated the comments as appropriate wprovided clarification or corrected language regarding test reactor. 

E3-1 Massachusetts Institute of Technology The staff recognizes the concern presented by the com
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

 (MIT) – The paper discusses 10 CFR 

Part 20 limits in the context of a maximum hypothetical accident (MHA). The part 20 limits were not specifically written to be applied to accident scenarios. 

separate activity associated with updating licensing reresearch reactors under the rulemaking process, NRC 

developing accident dose criteria for research reactors

E3-2 MIT – The paper places radiological 

consequences in context of a facility’s effective power. The concern is that a facility’s effective power could be seen as precedence for using values other than license limits for other regulatory activities. 

The NRC staff recognizes the comment’s concern regvalues other than license limits for regulatory determintopic of this paper is lessons learned from the accidenDaiichi, which evaluates accident conditions caused bgreater magnitude than those considered sufficiently clicensing. Therefore, the NRC staff’s assessment consuch as effective power to ensure the assessment refeffects. No changes were made as result of this comm

E4-1 PBorchmann - Comment discusses 

Superflex and industry preferences for cheaper and less extensive site upgrades. The comment also references two non-concurrences associated with the draft assessment. 

The staff believes the majority of the comment refers tpaper that was issued in the process of developing CO“Integration of Mitigating Strategies for Beyond-Design Events and the Reevaluation of Flooding Hazards.” Tavailable in the Agencywide Documents Access and M(ADAMS) at Accession No. ML14238A616. Two non-associated with COMSECY can be found in the encloCOMSECY. The staff’s draft assessment on the applicability of Fuklearned to facilities other than power reactors did not hconcurrences. 

E4-2 PBorchmann – Comment mentions that 

stakeholders in southern California are concerned about Section 7 of the white paper regarding decommissioned 

Comment does not provide specific concerns regardindecommissioning portion of the white paper, thereforedocument was identified by the staff. 
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

reactors. 

E5-1 Alliance for a Green Economy, Citizens’ 

Environmental Coalition, Nuclear Information and Resource Service – Citing the West Valley Nuclear site in Western NY, challenges statements regarding timeliness in decommissioning activities as related to the West Valley Demonstration Project (WDVP). 

NRC staff recognizes that the WVDP is a unique case other NRC licensed complex sites. In fact, the WVDP Act) gives the NRC an advisory and a monitoring role 

WVDP activities. Under the Act, the Department of Endecontaminate and decommission: (a) facilities of the Service Center in West Valley in which high-level wasstored, (b) the facilities used in the solidification of wamaterial and hardware in connection with project. As the NRC’s WVDP license was put into abeyance. Whthe Act is completed, the NRC license will be reactivatState Energy Research and Development Authority wthe license. Therefore, due to the complex nature of tinterdependence of multiple Federal and State authorunique nature and process of decommissioning activitcosts, more time is needed to complete decommissionfacilities and land. As a result of comments received regarding the WVDPdiscussion of this project to the assessment found in E

E5-2 Commenters found in comment 5-1, 

stated that the NRC should list the decommissioned facilities that the white paper is evaluating and mention West Valley, Savannah River, Los Alamos, and Hanford Sites. The commenters request whether or not NRC staff consider these sites low risk. 

The decommissioned facilities evaluated in this draft wall decommissioned facilities with active NRC or AgreeThese facilities can be found in the 2014 Status of the Annual Report, an NRC document, which can be acceAccession No. ML1429A239 (http://pbadupws.nrc.gov/docs/ML1429/ML14294A239DOE has the lead (direct responsibility) for West Vallelicense for that site is reactivated—the NRC license is abeyance. The other sites referenced are also DOE’s 
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

The NRC’s role at the Savannah River, Hanford and Winvolves overview and consultation on implementation are therefore not appropriate for consideration in an Nwhether further regulatory action is necessary to addrfrom the Fukushima Daiichi accident at NRC licensed 

E5-3 Commenters found in comment 5-1, 

stated that a vulnerability assessment for earthquakes and climate change for complex decommissioning sites should be performed. The commenters referred 

to WVDP as indicated in the statements: “The reality is that at West Valley and many other complex material facilities, there are interred materials that are anything but securely contained;” and ... “Radioactive releases from West Valley nuclear waste site could also occur with extreme rainfall events, and conditions which result from climate change will inevitably increase the risk and consequences at the site.” 

NRC staff assessed potential risk at complex material NRC or Agreement States. As stated in the draft whitrisk from natural phenomena is low under severe scenusing Fukushima lessons learned potential dose imparisk is not expected to exceed Environmental ProtectioProtective Action Guideline levels for members of the emergency situation. Regarding, the West Valley Dem(WVDP), the natural phenomena, including containmeevents (e.g.; severe rainfall and flooding), were adequDOE’s Final Environmental Impact Statement (EIS) (wJanuary 2010 and can be found in ADAMS at AccessiML100750122) and earlier Safety Analysis Report (SAfound in ADAMS at Accession No. ML032940574). FChapter 4 of the EIS and Chapters 2 and 9 of the SARpotential public hazards and events related to containconfinement, fire hazard, criticality, seismology, meteosurface/subsurface hydrology. 

E5-4 Commenters found in comment 5-1, 

citing West Valley, challenge statements in the white paper that a decommissioning plan is required to be developed. 

Due to the unique and complex nature of the WVDP dfacility, the decommissioning plan is being submitted iNRC staff reviewed the Phase I Decommissioning PlaFebruary 25, 2010, ADAMS Accession No. ML100360implemented by DOE. The Phase II decommissioning by 2020. 

E5-5 Commenters found in comment 5-1, 

indicate that the white paper contains 

The EPA Protective Action Guideline levels under em

were used as guidelines to evaluate necessary regula
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

no data, facts, or science supporting the discussion that external events pose low risk to decommissioned facilities. 

addition, NRC staff evaluated “Decommissioning PlanTermination Plans” describing all safety aspects to propublic, and the environment from potential release of rNRC licensees assess site location, climate conditionsissues related to events such as extreme rain or floodtornadoes, and high wind events. Since the location omaterials during decommissioning are either on buildicould be stored in piles or in containers for ultimate padisposition, the risk from potential radioactivity releasefacilities was found to be low. Section 5 of the draft wreferred to studies already carried out and reported as technical reports. 

E5-6 Commenters found in comment 5-1, 

request the West Valley site-specific emergency preparedness plan. 

The DOE is responsible for operating and implementindecommissioning activities. The DOE is self-regulatinemergency plan in place. 

E5-7 Commenter found in comment 5-1, 

question whether a site evaluation was ever performed for the West Valley Nuclear facility. 

In the Final WVDP EIS, DOE addresses the site evalu

E5-8 Commenters found in comment 5-1, 

challenge whether the West Valley Nuclear facility considered external events and that NRC would be able to mitigate the release of radioactive material from the site. 

DOE is responsible for mitigation and cleanup. The NPhase I of the decommissioning activities associated wadvising DOE and monitoring their decommissioning aactivities. The Phase II decommissioning plan is anticdeveloped around 2020. Therefore, defining the specauthority/agency in Phase II is premature. 

E6-1 Donna Gilmore – Near term task force 

(NTTF) recommendation 1 not adequately addressed because the draft white paper does not address aging management of dry cask storage. 

See response to comments S1-1 and S1-2. 
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

E6-2 Donna Gilmore – The dry cask storage 

at Fukushima are superior to those licensed in the US. 

See response to comment S6-4. 

E6-3 Donna Gilmore – NTTF 

recommendation 2 is not adequately addressed. Aging management, new seismic science, and climate change not adequately assessed for ISFSIs. 

NRC staff disagrees with this comment. See responsfor additional detail. 

E6-4 Donna Gilmore – NTTF 

recommendation 8 regarding emergency planning not adequately addressed. 

NRC staff disagrees with this comment. See responsfor additional detail. 

E6-5 Donna Gilmore – NTTF 

recommendation 11 regarding emergency planning not adequately addressed. Continuous remote radiation monitoring systems should be required for ISFSIs. 

The paper’s focus is for Fukushima lessons learned, nof individual licensees. The staff does routinely assesperformance and did not identify changes needing to bas a result of its review. 

E6-6 Donna Gilmore – Provided a comment 

related to NTTF recommendation 12 regarding strengthening the oversight process. There is a lack of defense in depth for ISFSIs and NRC oversight for utilities that have shown a pattern of mismanagement should be increased. 

See responses to E6-1 through E6-5 above. 

E7-1 Joe Holtzman – Supports the comments 

from Donna Gilmore and indicates that he totally disagrees with the staff conclusions in the draft white paper. 

NRC staff disagrees with the comment. For responsecomments see responses to comment sets E6 and S6

E8-1 Marv Lewis – General comments that NRC staff disagrees with the comment. 
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

 NRC has not adequately addressed 

Fukushima lessons learned and provides general concerns including transportation of other hazardous materials on roads and railways along with radioactive shipments. 

As the comments relate to the transport of other hazaroad and railways, the Department of Transportation hrelated to shipment routing and mixing of different hazthe same conveyance in the Hazardous Materials RegParts 100 to 177. The responsibility of the safe transpradioactive materials is shared between NRC and the Transportation through a memorandum of understand

E9-1 Ruth Thomas – general comments such 

as draft white paper: 1) does not define very low risk, 2) does not address the consequences of an airplane hitting a reactor, and 3) does not address the vulnerability of non-power reactors. 

Regarding the definition of risk – the assessment foungenerally qualitative in nature and therefore, risk is noin terms of an overall number. NRC staff’s assessmeNUREG 2125, “Spent Fuel Transportation Risk Assesprovide a quantitative analysis. Where qualitative termstaff attempts to provide perspective. For example, Nmagnitude of the consequences of a natural phenomeexternal events as “low” for the NRC’s spent fuel and assessment if no radiation deaths or injuries are expecontamination is expected. Regarding airplane hitting a non-power reactor, and vpower reactors, several other comments were receiveresponse to questions S3-1, S5-7, and S6-2. 

E10-1 Paul Frey – Provided power point 

presentation on economic damage to California due to meltdown and fuel pool fires at Diablo Canyon. 

The paper is limited to Fukushima lessons learned for power reactors. NRC staff notes that regarding spent assessments have been made including those referenEnclosure 1. These assessments include NUREG-17of Spent Fuel Pool Accident Risk at Decommissioning Plants,” and more recently, SECY-13-0112, Enclosure Study of a Beyond-Design-Basis Earthquake Affecting for a U.S. Mark I Boiling Water Reactor” (ADAMS Acc
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

No. ML13256A342). 

E11-1 Bruce Campbell – Disagrees with 

overall conclusions of the draft white paper. 

NRC staff acknowledges the comment. 

E11-2 Bruce Campbell – Does not believe that 

ISFSIs are properly addressed, and asserts that NRC staff needs to ensure best quality casks like those in Germany. NRC staff also needs to assure reputable transport company and a place to store the waste. 

Regarding the ISFSI design portion of the question secomment S6-13. Regarding the portion of the comment associated with waste, the long-term plan for the disposal of spent fueof the President and Congress; the current plan estabcodified in the Nuclear Waste Policy Act (NWPA). Whrecognizes that the implementation of the current plan societal challenges, it is not the NRC’s role to set policNRC has developed regulations for the independent snuclear fuel in accordance with the NWPA. Staff belieevaluation adequately addresses the recommendationdid not identify changes needing to be made in this arreview. Regarding the portion of the comment regarding transresponsibility of the safe transportation of radioactive between NRC and the Department of Transportation tmemorandum of understanding (44 FR 38690). Thessafe transportation of radioactive materal. 

E11-3 Bruce Campbell – Low level waste site 

leak tritium and other radioactive material and are not properly regulated. 

Tritium release may occur at a LLW Disposal facility brelease would be insignificant and may represent a smpermissible public dose limit. Nevertheless, if such reconcerned State and operator should take appropriate deploying geo-membranes) to limit surface-water infilt
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

The current assessment pertains to natural phenomenby assuming that radioactive materials may be releaseevents as such: flooding, surface erosion, or earthquaconclusion derived in this analysis is that, under these the risk is low and the potential dose impact to membewould not exceed EPA Protective Action Guidelines limunder emergency situation. Considering the low dose from a tritium leak would be far below the 10 CFR Parrequirements for public dose limit. 

E11-4 Bruce Campbell – Mill tailings are not 

adequately regulated. 

The regulations that the NRC has established for mill 10 CFR Part 40, Appendix A, “Criteria Relating to the Mills and the Disposition of Tailings or Wastes Producor Concentration of Source Material from Ores ProcesTheir Source Material Content.” In addition, 10 CFR Pfor protection of the public apply to mill tailings. 

E11-5 Bruce Campbell – Nonpower reactors 

not adequately regulated. Lawrence Livermore Lab plutonium experiment released radioactivity to the environment. 

The scope of the paper is limited to NRC-licensed RTEnergy Act of 1954, as amended, does not give the licauthority to the NRC for the DOE and DOD reactor facthey are not normally licensed by the NRC. In special agreed to provide oversight of a specific DOD facility festablishment of an appropriate legal agreement. Furhas no knowledge of any reactors located at the LLNL

E11-6 Bruce Campbell – Fuel cycle generic 

letter is deficient. There should be specific site analysis to ensure safety. 

The generic letter requests fuel cycle facility licensees information on site specific analyses of natural phenomdemonstrate how they meet the NRC regulations and license. Compliance with the regulations provides reaof public health and safety. 

E11-7 Bruce Campell – NTTF 

recommendation associated with upgrading design basis components 

See response to comment S6-9. 
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

related to seismicity and flooding should not be considered “not applicable.” 

E12-1 Marni Magda – It is inappropriate to 

allow San Onofre to have buried canisters left on a bluff for 20 years 

See response to comment S6-18. 

E13-1 Michel Lee – Document uses overly 

optimistic assumptions. Climate change will exert a multiplier effect on aging mechanisms such that risks will increase. 

For comments related to aging management of spent see response to comments S1-1 and S1-2. For LLW disposal NRC staff performs assessments of performing the assessments, NRC staff assumes that packages/containers would degrade over time and, as materials may reach the receptor through multiple patconsiders natural phenomena events such as flooding change for LLW facilities. Aging mechanisms that maradioactive material from LLW packages/containers arNRC staff’s assessment. 

E13-2 Michel Lee – Economic risk placed on 

taxpayers is not addressed. 

The purpose of the draft white paper is to address potnatural phenomena on regulations of other than operafacilities. In other words, it is intended to address necactions to ensure protection of the health and safety o

E13-3 Michel Lee – Paper does not address 

the radioactive release of entire fuel cycle. 

The purpose of the draft white paper is to address potnatural phenomena on regulations of other than operafacilities. Therefore, NRC staff did not address the radthe entire fuel cycle. 

E13-4 Michel Lee – Paper is based on flawed 

assumptions including assuming nothing will go wrong, compliance with regulations ensures safety, and corporations will address radioactive releases from their facility. 

The paper conservatively assumes that extreme naturhazards happen and based on this performs an assesnot additional regulatory action is needed. As noted in identified two areas where additional assessment is nletter for fuel cycle facilities, and 2) additional assessmreactors. NRC staff does assume that licensees will m
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Comment Number 

Person Providing Comment and Description of Comment 

Response to Comment 

responsibilities; otherwise, appropriate regulatory actio

E13-5 Michel Lee – Paper does not address 

cyber security 

Cyber security is outside the scope of the Fukushima assessment. NRC staff is taking several actions in thioperating power reactors and other facilities the NRC 

E14-1 PBorchmann supplemental comment – 

Draft white paper does not recognize climate change impacts. 

Discussion of climate change found in response to coE5-5, E6-3 and E13-1. The analysis performed by the severe flooding (independent of the cause of the floodfollowing: 

• fully and partially submerged ISFSI cask (see Enc

• lessons learned based on Hurricane Katrina (see E