Regulatory Approaches for 

Addressing Reprocessing 

Facility Risks: An Assessment

Office of Nuclear Regulatory Research

NUREG/CR-7168
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Regulatory Approaches for 

Addressing Reprocessing 

Facility Risks: An Assessment

Manuscript Completed: December 2011

Date Published: February 2015

Prepared by: 

G. Martinez-Guridi, V. Mubayi, R. A. Bari, and *F. Gonzalez

Brookhaven National Laboratory

Nuclear Science and Technology Department

P.O. Box 5000

Upton, NY 11973

*U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001

 

NRC Job Code V6091

NUREG/CR-7168

 

iii

ABSTRACT

This report addresses methods for assessing the risks posed by a reprocessing facility, 

which have not previously been quantified relative to other fuel-cycle facilities. 

Reprocessing facilities can have higher potential source terms than other fuel-cycle 

facilities, which heighten the risk relative to the other facilities. This report explores the 

potential hazards that these facilities pose to the public, workers, and the environment 

by discussing literature on the regulation of these facilities and reviewing the experience 

of current operating facilities worldwide. It offers an overview of actual events and their 

consequences at these facilities. It also contains supporting information for assessing

the feasibility, advantages, and disadvantages of undertaking detailed versus simplified

quantitative risk assessments, for the range of events associated with large reprocessing 

facilities. The report gleans insights on regulating reprocessing hazards and risks from 

reports such as NUREG-1909 [Croff et al. 2008], and a white paper from the Nuclear 

Energy Institute [NEI 2008].

v

CONTENTS

ABSTRAC ..............................................................................................................................iii

LIST OF FIGURES ...................................................................................................................vii

LIST OF TABLES......................................................................................................................vii

EXECUTIVE SUMMARY ...........................................................................................................ix

ACKNOWLEDGMENTS.............................................................................................................xi

ACRONYMS AND ABBREVIATIONS ...................................................................................... xiii

1. INTRODUCTION................................................................................................................... 1

1.1 Objectives.............................................................................................................. 1

1.2 Background ........................................................................................................... 1

1.3 International Regulatory Contexts.......................................................................... 2

1.4 Organization of the Report..................................................................................... 3

2. REVIEW OF RECENT WORK .............................................................................................. 5

2.1 ACNW&M Evaluation of Potential Regulatory Changes......................................... 5

2.2 NEI Report............................................................................................................. 8

3. ACCIDENTS AND RISKS ASSOCIATED WITH REPROCESSING FACILITIES .................11

3.1 Summary of Accidents at Reprocessing Facilities.................................................11

3.1.1 Overview of Reprocessing Facilities Worldwide....................................................11

3.1.2 Accidents at Reprocessing Facilities Worldwide ...................................................14

3.1.3 International Literature on Safety and Regulation of Reprocessing Facilities ........18

3.2 Results of Risk Assessments................................................................................21

3.2.1 Rokkasho Reprocessing Plant ..............................................................................23

3.2.2 Tokai Reprocessing Plant .....................................................................................26

3.2.3 Thermal Oxide Reprocessing Plant.......................................................................27

3.2.4 La Hague..............................................................................................................28

3.2.5 Electrochemical Processing (Pyroprocessing) ......................................................28

3.3 Insights Gained from the Survey of Accidents and Risk Analyses.........................30

4. QUALITATIVE VS. QUANTITATIVE RISK METHODS.........................................................33

4.1 Introduction...........................................................................................................33

4.2 Approach ..............................................................................................................34

4.3 Insights from the Technology Neutral Framework of NUREG-1860 ......................34

4.4 PRA for Reprocessing Facilities............................................................................35

5. SUMMARY AND OBSERVATIONS .....................................................................................39

5.1 Summary ..............................................................................................................39

5.2 Observations ........................................................................................................39
vi

5.3 Suggested Considerations....................................................................................40

6. REFERENCES ....................................................................................................................41

APPENDIX A: Brief Description of Events at Reprocessing Facilities.....................................A-1

APPENDIX B: Brief Description of Electrochemical Processing (Pyroprocessing)..................B-1
vii

LIST OF FIGURES

Page

Figure 1: Trends in spent-fuel management (from [IAEA 2008a]) ........................................13

Figure 2: Variations in the frequency of occupational exposure with dose 

(from [ICRP 1993])................................................................................................19

Figure 3: Approach for evaluating frequencies of accidents (from [Shoji et al. 2005]) ..........24

Figure 4: Conceptual comparison of risk (from Kohata et al. [2004])....................................25

Figure B-1: Diagram of Electrochemical Processing Operations ...........................................B-2

LIST OF TABLES

Page

Table 1: Past, Current, and Planned Reprocessing Capacity in the World ...........................12

Table 2: Summary of Accidents in Reprocessing Facilities Worldwide .................................16

Table 3: Facility and Analysis Ranking (from IAEA [2002])...................................................22

Table 4: Significant Accidents in Reprocessing Facilities Worldwide....................................30

Table A-1: Events at Reprocessing Facilities in France..........................................................A-2

Table A-2: Events at Reprocessing Facilities in Germany ......................................................A-4

Table A-3: Events at Reprocessing Facilities in India.............................................................A-5

Table A-4: Events at Reprocessing Facilities in Japan...........................................................A-5

Table A-5: Events at Reprocessing Facilities in the Russian Federation ................................A-5

Table A-6: Events at Reprocessing Facilities in United Kingdom .........................................A-10

Table A-7: Events at Reprocessing Facilities in United States of America............................A-11

ix

EXECUTIVE SUMMARY

In 2008 AREVA NC Inc. [AREVA 2008] and other industry entities indicated to the Nuclear 

Regulatory Commission (NRC) their interest in commercial reprocessing in the U.S. In 

December 2008, the Nuclear Energy Institute submitted its white paper on reprocessing 

regulations [NEI 2008]. As a result, the NRC began assessing the need to expand its current 

regulatory regimes under Title 10 of the Code of Federal Regulations (CFR) Part 70 to 

encompass large spent fuel reprocessing facilities. In a series of SECY documents and their 

associated Staff Requirements Memoranda (SRMs), the NRC updated its planning for revising 

the regulatory framework for regulating reprocessing facilities. In SECY-09-0082 [NRC 2009]

the NRC staff presents a gap analysis, as required by the Commissioners, associated with the 

development of the new framework. The staff presented twenty-three gaps in the current 

regulatory structure in SECY-09-0082 along with their descriptions, and priority.

This report relates to Gap #5, described in the enclosure to SECY-09-0082, that is concerned 

with risk considerations for a production facility licensed under 10 CFR Part 70. The risk 

assessment required by Part 70 involves an Integrated Safety Analysis (ISA) with a 

characterization of the likelihood and consequences of credible accident sequences. SECY-09-

0082 notes that the existing requirements do not adequately address the increased risk posed 

by a reprocessing facility relative to that of other fuel-cycle facilities. Furthermore, it points out 

that reprocessing facilities can have higher potential source terms than other fuel-cycle facilities, 

which may heighten the risk of the former facilities. 

A suitably performed risk assessment of a reprocessing facility potentially can characterize the 

associated risks of concern adequately. This is the general sentiment of the NRC Advisory 

Committee on Nuclear Waste and Materials (ACNW&M) (see NUREG-1909 [Croff et al. 2008]).

SECY-09-0082 comments on the need to revise 10 CFR 70 to adequately address the unique 

hazards and risks related to these facilities.

Fuel-cycle facilities are distinguished from power reactors mainly by the diversity of their

strongly interrelated inherent hazards, and by the large distribution and mobility of the hazards 

throughout the plant. A comprehensive identification and quantification of initiating events and 

scenarios is a challenge when performing a fully integrated Probabilistic Risk Assessment

(PRA) for these facilities. A very consistent effort is deemed necessary to provide a realistic, 

accurate quantification of the risk; significant uncertainties generally are expected in the results. 

Beginning in mid-2009, under contract with the Office of Nuclear Regulatory Research (RES), 

Brookhaven National Laboratory (BNL) prepared this report which provides information to 

support NRC’s assessment of the feasibility, advantages and disadvantages of conducting 

detailed quantitative vs. simplified qualitative risk assessments for the range of accidents 

associated with reprocessing. For this report, BNL 1) explored the potential hazards that large 

reprocessing facilities pose to the public, workers, and the environment, 2) searched the 

literature for reports and other documents related, in particular, to risk assessments conducted 

for such facilities, 3) reviewed the experience of current operating facilities for example, in 

Japan, France, U.K., and elsewhere, and, 4) gleaned insights on regulating reprocessing 

hazards and risks from reports such as NUREG-1909 [Croff et al. 2008], and a white paper from 

the Nuclear Energy Institute (NEI) [2008]. 
x

This report is the product of a short-term, limited-scope study and did not fully survey the range 

of material existing in this subject area. Moreover, because some information was proprietary, 

some of the body of literature was unavailable. Nevertheless, enough material was available 

from which to make sound observations and offer supportive insights.
xi

ACKNOWLEDGMENTS

The authors are indebted to colleagues in France, Japan, and the United Kingdom for providing 

valuable information for this study. They are F. Bertrand, G-L. Fiorini, R. Nakai, G. Vaughan, 

and D. Watson. Comments and insights from the following U.S. NRC staff who reviewed earlier 

drafts of this report were invaluable: D. Damon, Y. Faraz, A. Murray, W. Reed, P. Reed and T. 

Sippel.

xiii

ACRONYMS AND ABBREVIATIONS

ACNW&M U.S. NRC Advisory Committee on Nuclear Waste and Materials

ACRS Advisory Committee of Reactor Safeguards

AIChE American Institute of Chemical Engineers

ANL Argonne National Laboratory

BDC Baseline Design Criteria

BNL Brookhaven National Laboratory

BSL Basic Safety Limits

BSO Basic Safety Objectives

CFR Code of Federal Regulations

DOE U.S. Department of Energy

ER Electrorefiner

EPA U.S. Environmental Protection Agency

FCF Fuel Cycle Facilities

FINAS Fuel Incident Notification and Analysis System

FMEA Failure Mode and Effects Analysis

GDC General Design Criteria

HAZOP Hazard and Operability Analysis

HSE U.K. Health and Safety Executive

IAEA International Atomic Energy Agency

ICRP International Commission on Radiological Protection

INES International Nuclear and Radiological Event Scale

INPRO International Project on Innovative Nuclear Reactors and Fuel Cycles

IROFS Items Relied on for Safety

ISA Integrated Safety Analysis

JAEA Japan Atomic Energy Agency

LOPA Layers of Protection Analysis (LOPA)

LWR Light Water Reactor

MFFF Mixed Oxide Fuel Fabrication Facility

NEA Nuclear Energy Agency

NEI Nuclear Energy Institute

NFCIS Nuclear Fuel Cycle Information System

NMED Nuclear Materials Events Database

NRC U.S. Nuclear Regulatory Commission

NRNF Non-Reactor Nuclear Facility

NSC Japanese Nuclear Safety Commission

OECD Organisation for Economic Co-operation and Development

PHA Process Hazard Analysis

PRA Probabilistic Risk Assessment (or Probabilistic Risk Analysis)

PSA Probabilistic Safety Assessment

PUREX Plutonium-Uranium Recovery by Extraction

RRP Rokkasho Reprocessing Plant

SAP Safety Assessment Principles

SNF Spent Nuclear Fuel

TAG Technical Assessment Guide

THORP Thermal Oxide Reprocessing Plant

TRP Tokai Reprocessing Plant

TRU Transuranic

UREX Uranium Recovery by Extraction

1

1. INTRODUCTION

1.1 Objectives

In 2008, the Nuclear Regulatory Commission (NRC) began assessing the need to expand 

current regulatory regimes under Title 10 of the Code of Federal Regulations (CFR) Part 70 to 

encompass large spent fuel reprocessing facilities. Beginning in mid-2009, under contract with 

the Office of Nuclear Regulatory Research (RES), Brookhaven National Laboratory (BNL) 

prepared this report which provides information to support NRC’s assessment of the feasibility, 

advantages, and disadvantages of conducting detailed quantitative vs. simplified qualitative risk 

assessments for the range of accidents associated with reprocessing. For this report, BNL 1) 

explored the potential hazards that large reprocessing facilities pose to the public, workers, and 

the environment, 2) searched the literature for reports and other documents related, in 

particular, to risk assessments conducted for such facilities, 3) reviewed the experience of 

current operating facilities for example, in Japan, France, U.K., and elsewhere, and, 4) gleaned

insights on regulating reprocessing hazards and risks from reports such as NUREG-1909 [Croff

et al. 2008], and a white paper from the Nuclear Energy Institute (NEI) [2008]. This report 

documents the observations of this study and supporting information.

This report is the product of a short-term, limited-scope study. It did not fully survey the range of 

material existing in this subject area. Moreover, because some information was proprietary, 

some of the body of literature was unavailable. Nevertheless, enough material was available 

from which to make sound observations and offer supportive insights.

The main observations from the international activities are limited and somewhat general. 

Accordingly, Section 1.3 gives a concise pertinent overview. 

1.2 Background

In response to government- and industry-initiatives over the past few years, the NRC has been 

considering revising its regulatory structure for spent-fuel reprocessing. In a series of SECY 

documents and their associated Staff Requirements Memoranda (SRMs), the NRC updated its 

planning for revising this regulatory framework. In SECY-09-0082 [NRC 2009], along with their 

priority, the NRC staff presents a revised gap analysis, as required by the Commissioners, 

associated with the development of the new framework. The staff presented twenty-three gaps 

in the current regulatory structure in SECY-09-0082.

This report relates to Gap #5, described in the enclosure to SECY-09-0082, that is concerned 

with risk considerations for a production facility licensed under 10 CFR Part 70. The risk 

assessment required by Part 70 involves an Integrated Safety Analysis (ISA) with a 

characterization of the likelihood and consequences of credible accident sequences. SECY-09-

0082 notes that the existing requirements do not adequately address the increased risk posed 

by a reprocessing facility relative to that of other fuel-cycle facilities. Furthermore, it points out 

that reprocessing facilities can have higher potential source terms than other fuel-cycle facilities, 

which may heighten the risk of the former facilities. 

A suitably performed risk assessment of a fuel reprocessing facility potentially can characterize 

the associated risks of concern adequately. This is the general sentiment of the NRC Advisory 

Committee on Nuclear Waste and Materials (ACNW&M) (see NUREG-1909 [Croff et al. 2008]).
2

SECY-09-0082 comments on the need to revise 10 CFR 70 to adequately address the unique 

hazards and risks related to these facilities.

Fuel-cycle facilities mainly differ from power reactors by the diversity of their strongly 

interrelated inherent hazards, and by the large distribution and mobility of the hazards 

throughout the plant. A comprehensive identification and quantification of initiating events and 

scenarios is a challenge when performing a fully integrated Probabilistic Risk Assessment 

(PRA) for these facilities. A very consistent effort is deemed necessary to provide a realistic, 

accurate quantification of the risk; significant uncertainties generally are expected in the results. 

1.3 International Regulatory Contexts

A brief synopsis of regulatory approaches to reprocessing facilities in other countries is provided

here. It is based, in part, on private communications with officials and researchers in those 

countries and may not necessarily represent the stated positions of their respective 

governments.

In the United Kingdom (U.K.), the Health and Safety Executive (HSE) applies PRA to all 

facilities (nuclear, chemical, heavy industry, etc.). These PRAs must be suitable to the type of 

facility and sufficient to show that the numerical targets in the regulations are met. The targets 

are based on doses or risk to workers and the public; the same targets that apply to nuclear 

power plants and other facilities. In revising its Safety Assessment Principles (SAPs) for fuel 

facilities, in 2006 the U.K. removed the two targets on the release of radioactivity and plant 

damage, as they were considered as reactor-orientated and unsuitable for fuel facilities. The 

new SAPs are published at www.hse.gov.uk/nuclear/saps/index.htm. This page also explains

the numerical targets. The U.K.’s SAPs are expressed in terms of Basic Safety Objectives 

(BSO) and Basic Safety Limits (BSL). The former is a level of risk considered as negligible (i.e., 

risks below the BSO do not require further regulatory attention), while the latter corresponds to a 

risk level analogous to a regulatory limit, viz. risks above the BSL would be considered 

unacceptable. The region lying between the BSO and the BSL is regarded as a “tolerable” risk 

region. In the latest revision of the U.K.’s SAPs, the BSO for the individual risk of an offsite 

fatality due to internally and externally initiated accidents that occur at a nuclear facility is set at 

1 X 10-6 per year, while the corresponding BSL is 1 X 10-4 per year. (The U.K. guidelines do not 

distinguish between early- or prompt-fatalities and latent cancer fatalities). In addition, the 

U.K.’s HSE recently published a Technical Assessment Guide (TAG) on Probabilistic Safety 

Analysis (PSA)1 [HSE 2009].

After the fire and explosion incident in Japan in 1997 ([IAEA 1999a] and [IAEA 2007]), safety 

reassessments of the Tokai Reprocessing Plant (TRP) were undertaken from 1998 to 1999;

also various activities were carried out to enhance plant safety. From 2002 to 2003, the relative 

importance of safety functions at the TRP was evaluated by applying a PRA. Thereafter, from 

2004 to 2005, a PRA also was applied to four representative accident scenarios to assess 

quantitatively the effectiveness of the hardware modification and operating procedure 

improvement which were implemented based on the PRA results. Since then, the Japan Atomic 

 1

 Some publications use the term Probabilistic Risk Assessment (or Analysis) (PRA), while others use 

Probabilistic Safety Analysis (PSA). These terms have the same meaning and are used interchangeably 

on this report. 
3

Energy Agency (JAEA) has conducted a study estimating the component failure rates for a 

reprocessing plant based on the maintenance records stored in the TRP.

In France, there has been an effort to develop the necessary knowledge of existing riskmanagement tools, thereby allowing their use to assess the risk of nuclear facilities. In this 

context, the PRAs are of particular interest because of their significant role in the safety culture

of the nuclear industry. There appear to be no distinctions in the approaches of the French 

nuclear authorities to various types of fuel-cycle facilities.

1.4 Organization of the Report

The report has seven chapters and two appendices. Chapter 2 briefly reviews previous work. 

Chapter 3 summarizes accidents at, and risk assessments for reprocessing facilities. Chapter 4

discusses applying qualitative versus quantitative risk methods. Chapter 5 summarizes the

insights gained about the current regulatory situation and offers some observations. Finally, 

Chapters 6 and 7, respectively, contain the references, and acknowledgments. Appendix A 

concisely presents the accidents that have happened in reprocessing facilities worldwide. 

Appendix B gives a short description of electrochemical processing (pyroprocessing) and its 

literature. 

As a guide to the reader, it is noted that there is no one-to-one correspondence between the 

sections of the report and the four tasks stated in Section 1.1. The aim was to provide an 

integrated, holistic view of the subject area, and thus, the report sections do not draw on the 

elements of inquiry given by the four tasks in isolation from one another.

5

2. REVIEW OF RECENT WORK

2.1 ACNW&M Evaluation of Potential Regulatory Changes

In 2008, the Advisory Committee on Nuclear Waste and Materials (ACNW&M) of the NRC 

published a report on the Background, Status and Issues Related to the Regulation of 

Advanced Spent Nuclear Fuel Recycle Facilities (NUREG-1909 [Croff et al. 2008]). In the 

context of this present paper, the ACNW&M define recycle as involving (a) reprocessing the 

spent nuclear fuel (SNF) to separate it into its constituent components, (b) refabricating fresh 

fuels containing plutonium and minor actinides, and, (c) managing and storing the gaseous-, 

liquid-, and solid-wastes generated along with spent fuel. The ACNW&M report describes the 

historical approach to the SNF recycle, reviews recent advances in technology, and evaluates 

the technical- and regulatory-issues that will need to be addressed to assure the viability of 

commercially reprocessing spent fuel. 

The following are of particular interest to the work documented in this report: (1) The 

ACNW&M’s evaluation of potential modifications to Part 70, perhaps including creating a new 

rule that would have to be considered for an effective, efficient regulatory process, (2) the 

impact of reprocessing on other NRC regulations dealing with nuclear wastes, and, (3) the 

complexity of some advanced approaches to reprocessing, such as the UREX process1 or 

electrochemical processing that need more technical development and evaluation before an 

adequate regulatory framework including regulatory guidance can be established. 

The ACNW&M report points out that fuel-fabrication facilities, which now are licensed under 10 

CFR Part 70, utilize an ISA to assess the safety of the design and to identify equipment relied 

upon for safety. While using the ISA is an important step towards risk quantification and the 

expanded use of risk-informed regulations, the ACNW&M report indicates that the “Joint 

Subcommittee of ACRS and ACNW&M noted shortcomings in ISAs that would likely need to be 

addressed to expand its role in regulatory decisions involving reprocessing facilities.” 

In discussing ISA versus PRA for analyzing risk in reprocessing facilities, the ACNW&M report 

comments:

“The primary reason for using ISA rather than full scope PRA is that the consequences of likely 

accidents in or routine releases from fuel cycle facilities are believed to be small compared to 

the consequences of accidents at reactors, and does not justify the effort of doing probabilistic 

analyses. However, the effort required to prepare an ISA for complex SNF recycle handling 

liquids containing substantial quantities of concentrated cesium, strontium, and TRU2 elements 

is likely to approach the effort that would be required to evaluate risks using a PRA. The 

Committee and the ACRS have previously advised [ACNW&M, 2002, 2006] that a regulation 

that utilizes PRA insights is preferable to one based on ISA because the latter has significant 

limitations in its treatment of dependent failures, human reliability, treatment of uncertainties, 

and aggregation of event sequences.”

 1 There is a suite of UREX processes, each of which consists of a series of steps designed to remove 

specific groups of radionuclides to tailor products and compositions of the desired product and waste 

streams [Laidler, 2006].

2 TRU stands for transuranic.
6

A related methodological issue discussed in the ACNW&M report is a best estimate versus a 

conservative approach:

“A companion issue to that of probabilistic versus deterministic approaches is whether analyses 

should be based on data and models that represent the best estimate of what might really occur 

with an associated uncertainty analysis to explore the effects of incorrect data or models, or 

should be based on demonstrably conservative data and models. Most regulations and license 

applications for fuel cycle facilities have used a conservative, deterministic approach. The 

Committee has letters on record pointing out problems with using this approach (see Appendix 

C [of NUREG-1909]). Some of the most important problems are that using very conservative 

assumptions can mask risk-significant items and most conservative analyses are not 

accompanied by a robust uncertainty analysis.”

In overall terms, the ACNW&M report determined that the experience and lessons learned from 

licensing fuel fabrication facilities under 10 CFR Part 70 to some extent are applicable to 

reprocessing facilities; however, several features of reprocessing facilities may require 

additional regulation. In particular, ACNW&M indicates that “a new rule” could be formulated 

specifically for licensing reprocessing facilities, and they point to the development of a riskinformed performance-based framework for licensing new reactor designs (published as 

NUREG-1860 [NRC 2007]) as an example of an approach that “...may be advantageous 

because of its flexibility.” In ACNW&M’s view, an advantage of formulating a new rule is that it 

would “...avoid the need to write exemptions for rules already in place and would place all the 

regulations relevant to the recycle facilities under one part of the regulations, effectively leaving 

other parts of the regulations unchanged.” The drawback to this approach would be the 

additional time and resources required to develop such a rule, although ACNW&M also stated 

that “....it is unclear whether the requirements for developing a new rule are significantly greater 

than those of other approaches.” 

The approach that ACNW&M implicitly recommends in considering a new regulation is the one 

set out in NUREG-1860 as a technology-neutral framework for a risk-informed and 

performance-based approach to licensing a future generation of nuclear power reactors. The 

ACNW&M report emphasizes the following aspects of this technology-neutral framework: 

[The framework put forth in NUREG-1860] “...integrates safety, security, and emergency 

preparedness to establish a comprehensive set of requirements as a license condition. The 

approach focuses on the most risk-significant aspects of plant operations and uses the 

Commission’s safety goals (separate goals would need to be developed for recycle facilities) as 

top-level regulatory criteria that designers must meet to ensure adequate safety. The approach 

eliminates the need for exemptions by implementing guidance to accommodate technological 

differences between designs.” 
7

The ACNW&M report details the following new activities and facilities that will require decisions 

about the appropriate licensing regulations. 

• reprocessing fuels from light water reactors (LWR), and later, from other advanced 

reactors 

• fabricating fuels to recycle TRU- or fission product-elements or fuels for some new 

reactor designs (e.g., graphite-moderated reactors) 

• disposing of new types of wastes, such as cladding and TRU (GTCC) waste3

• extending the interim storage of intermediate-lived radionuclides (cesium and 

strontium), followed by in situ disposal.

Since there are uncertainties about future reprocessing technologies that may be developed 

and implemented, ranging from aqueous technologies, like PUREX (plutonium-uranium 

extraction) and UREX, to dry technologies like electrochemical processing, a technology neutral 

set of regulations supplemented by technology specific regulatory guidance seemingly offers a 

balanced, flexible approach to creating the regulatory framework for reprocessing.

Another feature of any new or modified rule for reprocessing identified in the ACNW&M report is 

that it should “...be consistent with Commission policies including the Commission’s PRA policy 

statement [NRC 1995]. The latter states, in part, “The use of PRA technology should be 

increased in all regulatory matters to the extent supported by the state of the art in PRA 

methods and data, and in a matter that complements the NRC’s deterministic approach and 

supports the NRC’s traditional defense-in-depth philosophy.” The ACNW&M report reiterates 

“The Committee has gone on record repeatedly in letters to the Commission about the use of 

risk-informed decision making, starting in October 1997 and most recently in the letter of May 2, 

2006.” Appendix C of the ACNW&M report lists the committee letters related to risk-informed 

activities and PRA.

A new rule that licenses reprocessing also will need to specify limits on gaseous- and liquideffluents generated during operation. The ACNW&M report comments on the need to formulate 

“ALARA requirements” for reprocessing facilities that establish design objectives and limiting 

conditions for radioactive material effluents. These requirements will be analogous to the 

current Part 50 Appendix I, “Numerical Guides for Design Objectives and Limiting Conditions for 

Operation to Meet the Criterion ‘As Low as Is Reasonably Achievable’ for Radioactive Material 

in Light-Water-Cooled Nuclear Power Reactor Effluents.

In summary, the ACNW&M report has two main conclusions: (1) It determined that no existing 

regulation in the United States is “entirely suitable” for licensing reprocessing facilities. Existing 

fuel-cycle plants, such as fuel-fabrication facilities, handle relatively small amounts of 

radioactive materials but future reprocessing plants likely will process much larger quantities of 

radioactive materials in solids, liquids, and gases distributed in various locations in the facility. 

(2) It concludes that compared to other fuel-cycle facilities regulated under 10 CFR Part 70, the 

possibility of larger source terms, the presence of both radioactive and chemical hazards, as 

well as the greater complexity of equipment and operations suggests employing 

 3 GTCC waste stands for “greater than Class C” waste. 10 CFR Part 61.55 categorizes low-level waste 

from Class A (least hazardous) to GTCC (most hazardous).
8

correspondingly more sophisticated methodologies, such as PRA, to analyze the risk of 

reprocessing facilities. Following the conclusions of the ACNW&M report, should PRA be 

adopted as the method for carrying out risk assessments of reprocessing plants, several 

improvements and enhancements to the traditional PRA methods used in reactors are 

warranted for the latter facilities since they pose both radiological- and chemical-risks. 

Examples of these potential enhancements are briefly mentioned in Chapter 4.

2.2 NEI Report

The NEI [2008] proposed a licensing framework for a reprocessing facility that “...is modeled 

under the risk-informed and performance-based approach of Part 70 supplemented with 

provisions from Part 50.” The framework is designed to implement a new part, labeled part 7x, 

under Title 10 of the CFR. The NEI proposes that the framework is technology neutral, and 

sufficient to encompass licensing of the different reprocessing technologies that industry is 

studying. 

From a substantive and technical standpoint, the framework basically adopts the approach and 

requirements that mirror those in Part 70. It requires performing an ISA to identify facility 

accidents and items relied on for safety (IROFS), management measures to assure the 

availability and reliability of the IROFS, and other associated administrative requirements. It 

requires quantitative assessments of risk to a member of the public located outside the 

controlled area from high consequence accidents involving fission products to the extent 

practicable based on the availability of data to support quantitative analysis, and establishment 

of Technical Specifications for IROFS identified for such accidents. 

In offering a rationale of the need for a new part beyond Part 70 to license a reprocessing 

facility, NEI identifies one regulatory argument, and one substantive technical one. The 

regulatory argument is derived from the fact that a reprocessing facility is considered a 

“production facility” under the Atomic Energy Act, and hence, is subject to 10 CFR 50,

“Domestic Licensing of Production and Utilization Facilities.” NEI cites the portions of the 

existing Part 50 that refer to reprocessing facilities but argues, citing several NRC SECY 

documents such as SECY-08-0134 [NRC 2008b], that Part 50 mainly was used to license 

operating light-water reactor (LWR) power plants and it “....would not be effective or efficient to 

revise Part 50 to license reprocessing facilities.” The substantive technical argument rests on 

the claim that reprocessing facilities would have a “...greater source term than other fuel cycle 

facilities.” Additionally, the NEI report indicates that as a production facility, regulatory 

requirements such as Technical Specifications, which are lacking in Part 70, would have to be 

identified.

The NEI proposes an additional technical feature beyond Part 70, viz., for “...accident scenarios 

that could result in a high consequence event involving fission product releases to an individual 

located outside the controlled area, the ISA is to be supported by a quantitative assessment of 

the risk to the extent practicable based on the availability of data to support quantitative analysis 

including determination of the margins of safety during normal operations and transient 

conditions anticipated during the life of the facility, and the adequacy of structures, systems, and 

components.” Since a quantitative risk assessment generally signifies a PRA, the NEI proposal 

essentially is to perform an ISA (a qualitative analysis but might be supported by semiquantitative assessments) and then a limited PRA that analyzes only a subset of potential 

accidents at the facility. The NEI mentions that the likelihood determination made in the ISA will 

be “...supported by quantitative analysis” for a “...relatively limited subset of IROFS.” The NEI 
9

report also states that the Technical Specifications will be developed only for this limited set of 

IROFS that protect against accidents that might entail high consequences from releases of 

fission-products to an off-site individual.

In addition, the NEI proposal for Part 7x incorporates a set of baseline design criteria (BDC) for 

the reprocessing facility that are modeled on corresponding design criteria in 10 CFR Part 50 

and Part 70. Appendix A of Part 50 specifies General Design Criteria (GDC) for nuclear-power 

plants, and Part 70.64 specifies BDC for fuel-cycle facilities licensed under Part 70. The BDC 

proposed in Part 7x are a combination of the Part 50 GDC and the Part 70 BDC.

11

3. ACCIDENTS AND RISKS ASSOCIATED WITH REPROCESSING 

FACILITIES

3.1 Summary of Accidents at Reprocessing Facilities

This chapter briefly summarizes accidents at nuclear-fuel reprocessing facilities throughout the 

world. It is organized into four main sections. Section 3.1.1 gives an overview of reprocessing 

facilities, and Section 3.1.2 summarizes accidents therein. Some reports published by 

international agencies related to the safety and regulation of such facilities, were identified while 

searching the literature for the accidents; Section 3.1.3 briefly describes them. The focus of this 

literature search was on the accidents, and not on the reports; hence, those included in Section 

3 are publications that appear relevant, but were not obtained through an exhaustive search. 

3.1.1 Overview of Reprocessing Facilities Worldwide

Table 1, shows past, current, and planned reprocessing facilities worldwide. Both current and 

projected reprocessing capacities shown in this table are a relatively small fraction of the total 

spent fuel generated by nuclear power plants worldwide, as expressed by the International 

Atomic Energy Agency (IAEA) TECDOC-1587 [IAEA 2008a]:

“Currently about 10,500 tHM1 spent fuel are unloaded every year from nuclear power 

reactors worldwide (Figure 1). This is the most important continuous growing source of 

civil radioactive materials generated, and thus need to be managed appropriately. Also, 

this annual discharge amount is estimated to increase to some 11,500 tHM by 2010. 

The total amount of spent fuel cumulatively generated worldwide by the beginning of 

2004 was close to 268,000 tHM of which 90,000 tHM has been reprocessed. The world 

commercial reprocessing capacity is around 5,550 tonnes per year.2

 Projections 

indicate that the cumulative amount generated by the year 2010 may be close to 

340,000 tHM with a corresponding increase in reprocessed fuel. By the year 2020, the 

time when most of the presently operated nuclear power reactors will approach the end 

of their licensed operation life time, the total quantity of spent fuel generated will be 

approximately 445,000 tHM.”

Discussing spent fuel produced in the United States, the NRC [2011] indicates 

“As of January 2011, the amount of commercial spent fuel in safe storage at commercial 

nuclear power plants was an estimated 63,000 metric tons. The amount of spent fuel in 

storage at individual commercial nuclear power plants is expected to increase at a rate 

of approximately 2,000 metric tons per year.”

 1 tHM means metric tonnes of heavy metal (MtHM). 

2 Table 1 indicates that the world’s current reprocessing capacity is 5,950 MtHM/year. The cause of the 

difference from 5,550 MtHM/year was not found in the IAEA-TECDOC-1587; however, this difference is 

not significant. 
12

Table 1: Past, Current, and Planned Reprocessing Capacity in the World

(in tHM/year) (from [IAEA 2008a])

Country Site Plant Type of Fuel 

Processed

Operation Capacity

Start Shutdown Present Future

Belgium MOL Eurochemic LWR 1966 1975

China Jiuquan RPP LWR ? 25

Lanzhou LWR 2020 800

France

Marcoule APM FBR 1988 1996

Marcoule UP1 GCR 1958 1997

La Hague UP2 LWR 1967 1000 10003

La Hague UP3 LWR 1990 1000 10007

Germany Karlsruhe WAK LWR 1971 1990

India

Trombay PP Research 1964 60 60

Tarapur PREFRE 1 PHWR 1974 100 100

Kalpakkam PREFRE 2 PHWR 1998 100 100

Kalpakkam PREFRE 3A PHWR 2010 150

Tarapur PREFRE 3B PHWR 2012 150

Japan

Tokai-mura JAEA TRP LWR 1977 90 90

Rokkashomura

JNFL RRP LWR 2007 800

Russian

Fed.

Chelyabinsk RT1

WWER-440,

BN-350,

BN-600 RR

1977 400 400

Krasnoyarsk

RT2 WWER1000 2025 1500

Demonstrative 

facilities

VVER-1000

RBMK 2013 50-100

U.K.

Sellafield B205 GCR 1967 1500

Sellafield Thorp LWR/AGR 1994 900 1000

Dounreay UKAEA RP FBR 1980 2001

USA

West Valley NFS LWR 1966 1972

Hanford Rockwell U metal 1956 1989

Savannah 

River SR U metal 1954 1989

Idaho Falls R U-Al alloy 1959 1992

Total Capacity 5950 6525

 3 1000 MtHM for each plant, with a cumulated maximum of 1700 MtHM for the La Hague site. 
13

Figure 1: Trends in spent-fuel management (from [IAEA 2008a])4

Routine Releases of Radioactivity

Reprocessing facilities apparently used to release significantly more radioactivity during normal 

operation than other nuclear facilities. In particular, Schneider et al. [2001] stated,

“Reprocessing operations release considerably larger volumes of radioactive discharges than 

other nuclear activities, typically by factors of several 1,000 compared with nuclear reactor 

discharges. In the U.K., Fairlie [1997] has estimated that about 90% of nuclide emissions and 

discharges from the U.K. nuclear programme result from reprocessing activities.” On the other 

hand, a draft report by the IAEA [2007] pointed out:

“With the state-of-art technology, it is now possible to design and operate fuel 

reprocessing plants with as low an environmental release as any other conventional 

chemical industry. For instance, it has now been demonstrated that the volume of waste 

generated in the reprocessing operations have been considerably reduced in recent 

years. The dose to the public due to waste discharged from reprocessing has also 

shown a steady decrease [IAEA 2005]. A combination of learning from experience and 

continuous improvements, modifying both plant and practice, such as introduction of 

automated operations has reduced the average employee radiation exposures at 

reprocessing facilities from over 10 mSv to 1.5 mSv per person pa over the past two 

decades. (By comparison, the average annual exposure for airline crew is about 2 

mSv). Radiation exposure of the public has also been reduced, largely in line with the 

reductions in radioactive discharges. As the quantity of radioactivity being discharged 

has declined each year, the proportion of radiation exposure that is attributable to 

 4

 “SF Discharged” means SF generated.
14

current discharges has declined. In the U.K. today, the average annual exposure of 

individuals due to radioactive discharges is less than 0.1 mSv. By comparison, the 

average annual exposure to individuals in the U.K. from natural background radiation is 

about 2.2 mSv.”

3.1.2 Accidents at Reprocessing Facilities Worldwide

BNL conducted a survey of accidents at reprocessing facilities, and this subsection summarizes 

them. McLaughlin, Frolov, et al. [2000] reviewed in detail criticality accidents in nuclear 

facilities, including several at reprocessing facilities around the world. Discussing such 

accidents, the IAEA [2007] states

...Of the nearly 60 criticality accidents, which have occurred since 1945, about a third 

occurred at nuclear fuel cycle facilities...Twenty of these accidents involved processing 

liquid solutions of fissile materials, while none involved failure of safety equipment or 

faulty (design) calculations. The main cause of criticality accidents appears to be the 

failure to identify the range of possible accident scenarios, particularly those involving 

potential human errors.

Some accidents occurred due to a chemical (“red oil”) reaction. Hyder [1994a and 1994b], and 

Paddleford and Fauske [1995] describe the reactions of tri-n-butyl phosphate (TBP) with nitric 

acid and nitrates.

In addition, an annex to the report by Schneider et al. [2001] contains some events that 

occurred at the La Hague reprocessing facility; COGEMA5 reported them to the French nuclearsafety authorities between 1989 and the end of the first half of 2001, and the safety authorities 

published them. 

Table 2 lists the major accidents that were categorized using the International Nuclear and 

Radiological Event Scale (INES), as presented in the literature, as well as other events whose 

consequences were of different degrees of severity and that have not been classified using this 

scale yet. Events having relatively small consequences were not included in this table, but 

nevertheless occur in reprocessing facilities. 

Cadwallader et al. [2005] discuss this situation when they state the following:

Searching the DOE Occurrence Reporting and Processing System for [Idaho Chemical 

Processing Plant] ICPP events returned ~ 600 events in the past 14 years. Many of 

these events were personnel anti-contamination clothing or skin contamination events, 

and a number of false fire alarms (as well as false criticality, security, and evacuation 

alarms) were included in the 600 events. There were several power outage and voltage 

dip events, and a number of personnel safety issues, including lockout-tagout 

deficiencies, chemical overexposure, radiation overexposure, procedure violations, and 

industrial injuries. There was a lightning strike at the facility in June 1998, but there was 

little damage after alarm systems were returned to normal. Several small fires were 

 5 COGEMA (Compagnie générale des matières nucléaires) is a French company, created in 1976 from 

the production division of the Commissariat à l'Energie Atomique (CEA), which is the French Atomic 

Energy Commission. In 2001, COGEMA became part of the larger group Areva; the subsidiary's name 

was changed to Areva NC in March 2006.
15

reported, in electrical distribution and other equipment, and two small fires with kerosene 

in the calciner apparatus. Other events included wrong casks and wrong fuel moved or 

stored, environmental contamination by chemicals (motor oil, diesel fuel, etc.), 

equipment failures, a few dropped items, and a bomb threat hoax. An ICPP worker was 

killed in an industrial accident when he was struck by a forklift truck in May 1991. These 

events are tragic and nonetheless endemic to many types of industrial facilities.
Table 2: Summary of Accidents in Reprocessing Facilities Worldwide

(The number in parenthesis after an accident is the number of resulting early fatalities, if available.)

INES Kind of Hazard

Level Type of Event Criticality Radiation Fire an

7 Major accident

6 Serious accident Mayak PA6, 9/29

5 Accident with wider 

consequences

4 Accident with local 

consequences

Windscale, 1973

3 Serious incident

THORP, 8/2004 La Hague, 1/6/1

Siberian Chemic

Tokai-mura, 3/11

2 Incident

1 Anomaly

Classification not available

Mayak PA, 3/15/1953

Mayak PA, 4/21/1957 (1)

Mayak PA, 1/2/1958 (3)

Oak Ridge Y-12 Plant, 6/16/1958

Los Alamos Laboratory, 12/30/1958 (1)

ICPT,7 10/16/1959

Savannah River, 9/13/1960

Mayak PA, 12/5/1960

ICPT, 1/25/1961

Hanford Works, 4/7/1962

Mayak PA, 9/7/1962

Siberian Chemical Combine, 1/30/1963

Siberian Chemical Combine, 12/2/1963

Mayak PA, 12/16/1965

Mayak PA, 12/10/1968 (1)

Windscale, 8/24/1970

ICPT, 10/17/1978

Siberian Chemical Combine, 12/13/1978

Mayak PA, 8/19/1950

Mayak PA, 9/13/1950

Mayak PA, 9/20/1950

Mayak PA, 10/1/1951 (1)

Mayak PA, 9/20/1952

Mayak PA, 11/6/1954

Mayak PA, 12/22/1955

Hanford Works, 6/18/1956

Hanford Works, 9/3/1963

La Hague, 10/2/1968

La Hague, 10/1/1976

La Hague, 1/2/1980

La Hague, 4/1980

La Hague, 1983

Sellafield, 1983

La Hague, 2/13/1990

La Hague, 3/11/1997

Karlsruhe, 1999

KARP, 1/21/2003

Dounreay, 9/2005

Hanford Works, 

Savannah River

Hanford Works, 

Hanford Works, 

Hanford Works, 

ORNL, 11/20/19

Fontenay-Aux-R

Hanford Works, 

La Hague, 1/14/

Savannah River

Hanford Works, 

UTP, Ontario, 19

Karlsruhe, 1985

Hanford reproce

 6 Mayak PA stands for Production Association

7 ICPT stands for Idaho Chemical Processing Plant.

16
17

The scope of BNL’s literature search is restricted to those events with serious consequences, 

that is, ones with any of the following characteristics:

1. Events involving criticality, fire, explosion, or substantial leak of radioactive material.

2. Events causing fatalities or injuries to people in the reprocessing facility’s site. 

3. Events involving off-site releases of radioactivity. 

All the events identified in this study were grouped according to the following three main kinds of 

hazard: 

1. Criticality. This group includes events with one or more criticality excursions that 

exposed workers directly to radiation. 

2. Radiation. This group includes events in which criticality did not occur, but one or more 

persons were exposed directly to radiation from an accidental release of radioactivity 

onsite and/or offsite from the reprocessing facility. 

3. Fire and/or explosion. This group includes events involving chemical exothermic 

reactions. Consequently, in addition to the fire and/or explosion hazard itself, 

radioactivity and/or hazardous chemicals may have been released onsite and/or offsite 

from the reprocessing facility.

The public is notified of the safety significance of events associated with sources of radiation, 

via the INES. International experts convened in 1990 by the IAEA and the Nuclear Energy 

Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD)

developed this scale. According to the INES User’s Manual [IAEA 2009], events are classified 

on the scale at seven levels: Levels 4–7 are termed “accidents” and Levels 1–3 “incidents;” 

events without safety significance are classified as “Below Scale/Level 0.” Events that have no 

safety relevance for radiation or nuclear safety are not classified on the scale. 

Table 2 summarizes all the events identified in this study as a function of the type of hazard 

involved and the INES classification. It shows that most of them have not been classified 

according to this scale, or their classification was not found in the public literature. If early 

fatalities resulted from an accident, the number of deaths is included in parenthesis after the 

accident. Here, “early fatality” is defined loosely as a death happening within a few months after 

the accident. In addition, other consequences such as “latent fatalities” (those that may occur 

with some time lag following exposure, such as latent cancers due to radiation exposure) may 

have occurred in some events, but this information was unavailable to this study. 

All early fatalities were workers at the reprocessing facilities. Accounting only for these

fatalities, apparently the event at Mayak Production Association (PA) on 1/2/1958 (3 early 

fatalities) would be categorized as Level 5, “Accident with wider consequences,” and the events 

involving one early fatality (Mayak PA, 10/1/1951; Mayak PA, 4/21/1957; Los Alamos Scientific 

Laboratory, 12/30/1958; Mayak PA, 12/10/1968) would be classified as Level 4, “Accident with 

local consequences,” in the international scale. 

Appendix A gives additional information about each event in Table 2. 
18

3.1.3 International Literature on Safety and Regulation of Reprocessing Facilities

Some reports published by international agencies about the safety and regulation of this type of 

facility were identified while searching the literature for the accidents presented in Section 3.1.2. 

Most of them were published by the International Atomic Energy Agency (IAEA), and the 

Nuclear Energy Agency (NEA). The focus of the literature search documented in this report 

was on the accidents and not on the reports, so the reports included are publications that 

appear relevant, but were not retrieved via an exhaustive search. This section briefly addresses 

them.

The report IAEA-TECDOC-1221 [IAEA 2001] contains the results of a meeting of the IAEA 

Technical Committee in 2000 whose main objective was to compile information on the nature of 

the safety concerns and status of the regulations on nuclear-fuel-cycle facilities in IAEA’s 

Member States. It states, “...Although some similar safety hazards may be posed at reactor 

and non-reactor fuel cycle facilities, the differences between them give rise to specific safety 

concerns at the non-reactor fuel cycle facilities that must be especially taken into consideration 

in the design and operation of these facilities...” It further points out, “The IAEA maintains a 

database of nuclear fuel cycle facilities in Member States in the Nuclear Fuel Cycle Information 

System [IAEA NFCIS]. This system reasonably represents the approximate number and 

worldwide distribution of nuclear fuel cycle facilities. The system may not necessarily be up to 

date in all respects as the IAEA relies on Member States to refresh the information periodically; 

however, the information is sufficiently representative to be used in assessing the relative 

magnitude and diversity of existing and projected fuel cycle facilities in Member States.”

The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was 

launched in the year 2000, based on resolutions of the IAEA General Conference, and 

developed a set of basic principles, user requirements and criteria together with an assessment 

method, which taken together, comprise the INPRO methodology, for the evaluation of 

innovative nuclear energy systems. The results of this work were initially documented in IAEATECDOC-1362, “Guidance for the evaluation for innovative nuclear reactors and fuel cycles” 

[IAEA 2003], and in IAEA-TECDOC-1434, “Methodology for the assessment of innovative 

nuclear reactors and fuel cycles” [IAEA 2004]. INPRO prepared additional guidance in using its 

methodology to assess the sustainability of an innovative nuclear energy system (INS) in the 

form of an INPRO assessment manual. The resulting INPRO manual is comprised of an 

overview volume (IAEA-TECDOC-1575, Rev. 1) [IAEA 2008c], and eight additional volumes 

covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management 

(Volume 4), proliferation resistance (Volume 5), physical protection (Volume 6), environment 

(Volume 7), safety of reactors (Volume 8), and safety of nuclear fuel cycle facilities (Volume 9). 

Volume 9 [IAEA 2008d] presents the safety issues related to design and operation of mining, 

milling, refining, conversion, enrichment, fuel fabrication, fuel storage and fuel reprocessing 

facilities. It further discusses adapting the INPRO methodology in terms of identifying indicators 

and acceptance limits of various criteria for these facilities. For example, with respect to 

occupational exposure criteria, it states:

It is realized that the experience with respect to nuclear fuel cycle facilities in various 

countries has not been collated and harmonized to the extent that has been done for the 

reactor systems. For example, the limits for exposures vary from country to country (see 

Figure 2 [ICRP 1993]). Arriving at such limits, falls strictly under the purview of the 

regulatory body in the respective country, even though it is presumed that the ICRP 

limits in general form the guidelines for the regulatory process.
19

Figure 2: Variations in the frequency of occupational exposure with dose

(from [ICRP 1993])

8

The goal of Volume 9 of IAEA-TECDOC-1575, Rev. 1, is to provide guidance to an assessor of 

the safety of nuclear-fuel- cycle facilities in a country or region (or even on a global scale) that is 

planning to install a nuclear-power program (or maintaining or enlarging an existing one) on how 

to apply the INPRO methodology in this specific area. As part of Volume 9, Section 5.6, “Fuel 

reprocessing facilities” specifically addresses these facilities, and Section 6.2, “Safety-related 

RD&D9 areas,” states10:

More research will be needed to bring the knowledge of plant characteristics and the 

capability of computer codes to model phenomena and system behavior for innovative 

fuel cycle installations to at least the same confidence level as for existing nuclear power 

plants. In addition, a method should be developed for quantifying the safety of such 

facilities. 

 8 1 Sv = 100 rem. 

9 RD&D stands for research, development, and demonstration. 

10 The spelling in this quote was changed to conform to U.S. spelling.
20

Further development of Probabilistic Safety Analyses (PSA) methods, including best 

estimate plus uncertainty analysis, and their supporting data bases are required and 

need to be capable of:

• Assessing innovative nuclear designs implemented with lines of defense composed 

of inherent safety characteristics and passive, as well as active systems;

• Assessing total risk from various states and considering both internal and most 

external initiating events;

• Accounting for safety culture and human factors;

• Accurately accounting for ageing effects; and

• Quantifying the effects of data and modeling uncertainties.

• Identify all important phenomena and try to computer simulate them.

• Validate computer codes in all regimes of fluid and solid material behavior. 

Simulation may compensate for lack of operating experience if limited experimental 

results are available and they have been used to validate the computer code 

employed.

• Justify scaling to commercial size installations, and

• Obtain reliability data.

Further, the IAEA points out in its document IAEA-TECDOC-1587 [IAEA 2008a] that

Safety requirements for reprocessing plants are reflected at national level in regulations 

and standards. However, there is a trend toward internationalization of safety standards for 

the nuclear fuel cycle in general and spent fuel management facilities in particular. This 

issue has been examined at the IAEA and a system of international safety standards for 

fuel cycle facilities is in development. A safety guide on spent fuel reprocessing facilities is 

also in preparation.

Meanwhile, a Joint Convention on the Safety of Spent Fuel Management and the Safety of 

Radioactive Waste Management has been agreed to, and entered into effect on 18 June 

2001. The Joint Convention is the first international treaty relating to these areas of safety 

which is legally binding. It represents a commitment by States to achieve and maintain a 

high level of safety in the management of spent fuel and radioactive waste. The first 

Review Meeting of the Joint Convention was held in November 2003 [IAEA 1997] and the 

second in May 2006.

With regard to reprocessing, the scope of application of this Convention is “...Spent fuel held at 

reprocessing facilities as part of a reprocessing activity is not covered in the scope of this 

Convention unless the Contracting Party declares reprocessing to be part of spent fuel 

management.”

In addition, the IAEA authored a report, “Safety of Nuclear Fuel Cycle Facilities: Safety 

Requirements,” [IAEA 2008b] whose objective “...is to establish requirements that, in the light of 

experience and the present state of technology, must be satisfied to ensure safety, for all stages 

in the lifetime of a nuclear fuel cycle facility, i.e. its siting, design, construction, commissioning, 

operation and decommissioning. This publication is intended to be used by designers, 

operating organizations and regulators for ensuring the safety of fuel cycle facilities.” The scope 

of this publication includes reprocessing facilities. Ranguelova, Niehaus, and Delattre11 also 

discuss the safety of these facilities. 

 11 The authors of this report were affiliated with the IAEA. This report is undated.
21

The NEA also carried out technical activities in this area. In particular, the book “The Safety of 

the Nuclear Fuel Cycle” [NEA 2005] discusses the safety of these facilities; however, due to the 

limited scope of this work, this publication was not included in this study. Furthermore, the 

NEA’s Working Group on Fuel Cycle Safety (WGFCS) of the Committee on the Safety of 

Nuclear Installations (CSNI) organized the recent workshop “Fuel Cycle Safety – Past, Present 

and Future” [NEA 2007]. Representatives from Australia, Canada, France, the IAEA, Japan, the 

Russian Federation, U.K., and U.S. made presentations in the following sessions: “Legacy 

Waste Concerns,” “Fuel Cycle Safety Communication Issues,” “Ensuring the Safety of the 

Facility for the Future” (which included operational experience and research issues), and, 

“Regulation and Licensing of Nuclear Fuel Cycle Facilities.” 

Ueda [2005] describes using risk information for Japanese fuel-cycle facilities, indicating that 

studies on risk assessment methodology and related research conducted by Japanese 

organizations were based on methodologies, such as the probabilistic safety assessment (PSA) 

used for nuclear reactors, and methods of hazard analysis and risk evaluation for chemical 

plants. 

3.2 Results of Risk Assessments

This section summarizes the results of some Probabilistic Risks Assessments (PRAs) of 

reprocessing facilities worldwide. The inclusion of the approaches and results herein does not 

necessarily endorse them. 

As an introduction to this subject, a brief discussion by the IAEA in IAEA-TECDOC-1267 [2002] 

on the level of detail to use when conducting a PRA of a non-reactor nuclear facility (NRNF) is 

first presented:12 

Facility hazard can influence the depth of analysis, since it may be appropriate to 

analyze lower hazard facilities to less depth than higher hazard facilities (i.e., the depth 

of analysis is commensurate to hazard). Similarly, facility complexity can influence the 

depth of analysis, since it may be appropriate to analyze simple facilities to less depth 

than more complex facilities.

Thus, the concept of hazard-graded depth of analysis is appropriate for NRNF PSAs. This 

report seeks to provide a comprehensive guidance for assessing the risk of a high hazard 

NRNF for regulatory purposes. Table 3 illustrates the concept of a graded approach and 

provides some guidance as to how to reasonably apply reduced depth of analysis for facilities of 

lower hazard.

The following main observations were obtained from the information in Table 3: 1) A large 

reprocessing facility has a large radioactive inventory; 2) a detailed quantitative PRA is 

applicable for this kind of facility, and to some extent, to a medium-sized reprocessing facility, 

and, 3) this kind of PRA would include a Human Performance Analysis. 

 12 The spelling in this quote was changed to conform to U.S. spelling. 
22

Table 3: Facility and Analysis Ranking (from IAEA [2002])

Hazard Rank

Low (low activity 

inventory)

Medium (medium 

activity inventory)

High (large activity 

inventory)

Examples of Facilities13

Radioisotope Lab Fuel Fabrication 

Facility Research Reactor

Small Calibration 

Facility

Waste Treatment 

Facility

Large Reprocessing 

Plant

Hot Cell Facility Low-Level Waste 

Storage Facility

High-Level Waste 

Storage Facility

Depth of Analysis

Simple Intermediate Detailed

PSA Tasks (qualitative to semiquantitative) (semi-quantitative) (quantitative)

Familiarization Simple, minor effort ↔ Detailed diverse review

Hazard Identification, 

Initiating Events 

Selection

Simple systematic or 

engineering 

evaluation

↔

Detailed systematic 

review (FMEA, 

HAZOP, etc.)

Undesirable End States ↔ Detailed development

Safety Measures 

Identification Simple, minor effort ↔ Detailed identification

Safety Measures 

Information ↔ Detailed information

Event Grouping Simple grouping ↔ Detailed development

Event Sequence 

Modelling

Simple modeling or 

engineering 

evaluation

↔

Complex modelling 

(Failure Tree Analysis, 

Event Tree Analysis, 

etc.)

Human Performance 

Analysis Simple (judgment) ↔

Detailed analysis 

(Human Reliability 

Analysis, Task 

Analysis, etc.)

Consequence Analysis Simple analysis ↔ Detailed analysis

Parameter Estimating

Few parameters, 

bounding case, 

qualitative 

frequencies

↔ Many parameters, best 

estimates

Sequence 

Quantification

Simple (dose, 

qualitative 

frequency)

↔

Complex (uncertainty 

analysis, sensitivity 

analysis, distributions)

Documentation Basic ↔ Detailed

 13 Examples of facilities that may rank differently depending upon the facility’s size (for instance, a 

smaller reprocessing facility or plant involving a lower activity inventory may be ranked as a medium 

hazard).
23

3.2.1 Rokkasho Reprocessing Plant

It appears the safety analysts in Japan regard PRA as an important aspect of risk evaluation of 

their reprocessing facilities, even before risk-informed requirements were promulgated by the 

Japanese Nuclear and Industrial Safety Agency. Takebe et al. [2007] noted that “...Performing 

PSA and utilizing the risk information for non-reactor nuclear facilities also has the same role in 

securing safe operation effectively and rationally as in the nuclear power plants. Taking into 

account the safety characteristics of reprocessing plants in which radioactive materials exist 

scattered in several chemical processes and storage facilities, we should evaluate risks of many 

events efficiently and systematically with various types, scenarios, frequencies and 

consequences in order to assess whole risk and its profile of the plant...” A simplified PRA 

method for the Rokkasho Reprocessing Plant (RRP) was developed [Shoji et al. 2005], 

incorporating previous detailed PRA results of some representative events. The intent of this 

work was to use the risk information in operating and managing the facility (e.g., classifying the 

components and systems in classes of importance to determine the terms of periodical 

inspections). In their preliminary results for an evaluation of hydrogen explosions in a Pu 

concentrate vessel, they found that the risk contribution is higher for a high-consequence, lowfrequency explosion rather than a lower-consequence explosion at a higher frequency of 

occurrence. 

Shoji et al. [2005] compared the so-called “risk index method” for the ISA, with their proposed 

“improved risk index method” (also called a simplified PSA) and a detailed PSA (Figure 3). 

Takebe et al. [2007] point out that a detailed PRA for 15 selected events was first carried out, 

and the results “...showed us applicability of detailed PSA to RRP and quantified unavailability 

of utilities, electricity, cooling water and compressed air, those are commonly used in every 

processes. Some of the results were reflected upon the detail design works...In order to assess 

risk of many events efficiently and effectively seeking risk profile of the whole plant, a simplified 

PSA tool was developed because most part of events would have relatively simple sequence 

and low consequence and should not need detailed PSA. As the following step, using the 

simplified PSA, assessment works on 655 events are on the way.”

Kohata et al. [2004] introduce the overall approach and results of the detailed PSA of RRP that 

included “...frequencies of occurrence of events, release pathways and the amount of activities 

released, damage probabilities of high efficiency particulate air (HEPA) filters, radiation 

exposure dose (mSv), risks coming from accidents (both social and individual risks).” They offer 

a conceptual comparison of the risk from reactor plants and reprocessing facilities (Figure 4), 

and observe that for practical purposes, the risk of the former can be discussed in terms of only 

core damage frequency without regard for the consequence; however, discussing the risk of the 

latter requires accounting for both frequency of occurrence and consequence.
Figure 3: Approach for evaluating frequencies of accidents (from [Shoji et al. 2005])

24
25

Figure 4: Conceptual comparison of risk (from Kohata et al. [2004])

Kohata et al. [2004] indicate that the usage of risk information from the detailed PSA will be as 

follows:

• “PSA can give useful information on how to ensure and enhance the safety rationally 

including operational procedures and maintenance scheme

• To perform PSA extensively and to utilize the knowledge obtained to successfully cope with 

risk-informed regulation that will be implemented in the not too distant future

• Knowledge obtained from PSA will be applied to:

o Maintenance scheme including Allowed Outage Time (AOT) and Limiting Condition for 

Operation (LCO).

o Operational procedures against events beyond design base etc.

o Education and training of operators and staffs

o Rationalization of total inspection scheme including mandatory periodic inspection, 

voluntary periodic inspection, in service inspection, etc. considering the importance

• Risk-Informed Regulation

o A type of performance-based regulation

o What are the suitable performance indicators suitable for reprocessing plants to diagnose 

plant operational status?

o How to relate performance indicators to safety goal?

o Development of in-house oversight process to provide clear and appropriate information 

to each level of personnel concerned”

Shoji et al. [2005] state that in the original "risk index method" the items relied on for safety 

(IROFSs) are assumed to be independent, but that the simplified PRA approach is an 

improvement compared to the original method due to, among other factors, the inclusion of 

“Evaluation for reliability of IROFS based on failure frequency or probability of individual 

systems, components and human actions, in consideration of dependency between each 

IROFS.”
26

The simplified PRA applied to the RRP relies on information obtained from the detailed PRA 

that was performed for a limited set of accident scenarios. For example, Takebe et al. [2007]

state:

First of all, accident sequences and safety functions should be identified. Afterward, 

systems, components, and human action related to safety functions are identified from 

the design and operation information. As thus far explain, there is no difference from 

detailed PSA procedure...Analyst inputs prescribed failure rates for system, component 

and human action related to initiating events and safety function ... selecting from a list 

“Database of system, component and human action” that have been set conservatively 

based on the published documents (IEEE-std 500, NUREG-1363, etc.). Human error 

rate was determined based on detailed PSA... Unavailability of such support systems as 

utilities, which related to many events, has been set ... with simplified fault tree equation 

based on detailed PSA results. 

3.2.2 Tokai Reprocessing Plant

Ishida et al. [2003] indicate that 

With the fire and explosion accident at the Tokai Bituminization Demonstration Facility in 

March 1997, JNC [Japan Nuclear Cycle Development Institute] had carried out the 

safety reassessment of the TRP [Tokai Reprocessing Plant] in 1999... The PSR 

[Periodic Safety Review] of the TRP has been carrying out to obtain an overall view of 

actual plant safety. As a part of the PSR, based on the results of the safety 

reassessments of the TRP, PSA methodology has been applied to evaluate the relative 

importance of safety functions that prevent the progress of events causing to postulated 

accidents...As evaluation methods, event tree and fault tree methodologies were 

selected by taking into account of the power plant PSA [NRC 1982], PSA specialist 

opinions and document [IAEA 2002]...PSA methodologies have been applied on all 

postulated accidents.

Concerning dependencies and common-cause failures, Ishida et al. [2003] state, “Dependent 

failures could be dominant contributors to the frequency of the postulated accidents and they 

should be taken into account in the analysis regardless of the selected modeling approach...” 

They also point out “...human reliability analysis was carried out based on the operation manual 

by using the Technique for Human Error Rate Prediction (THERP)...” 

Ueda [2005] indicated that the Japanese Nuclear Safety Commission (NSC) and the Nuclear 

and Industrial Safety Agency (NISA) are planning to use risk information for nuclear-safety

regulation, and outlined studies in Japan on using risk information for reprocessing facilities. He 

pointed out “...Studies on risk assessment methodology and related researches have been 

conducted by [Japanese organizations] based on methodologies such as the probabilistic safety 

assessment (PSA) used for nuclear reactors and the methods of hazard analysis and risk 

evaluation for chemical plants...” He presented the PSA results for some postulated events at 

the TRP including risk importance factors and frequency vs. consequence plots. He noted that 

a greater number and variety of events must be evaluated as possible major contributors to risk 

in a PSA for TRP than those for nuclear reactors. He considered this is because 1) radioactiveand nuclear-materials are processed throughout a facility with a variety of chemical and physical 

forms; 2) there is a wide distribution of potential hazardous sources, e.g., radioactivity, heating 
27

sources, and flammable and explosive materials, in many parts of a facility; and, 3) there is a 

wide variety of postulated events in many parts of a facility. For one of the postulated events,

hydrogen explosions in a Pu purification process, Ueda gave a plot of “relative frequency” 

versus “relative consequence” of the related PSA accident sequences. 

3.2.3 Thermal Oxide Reprocessing Plant

PRA was used during the design and the start of operation of the U.K.’s Thermal Oxide 

Reprocessing Plant (THORP), and was used subsequently in its Periodic Safety Reviews, and 

for the Magnox reprocessing plant [Vaughan 2009]. In particular, James and Sheppard [1991] 

discuss the risk of thermal runaway in a nuclear-fuel-reprocessing plant due to red-oil reactions 

using, for illustration, the uranyl-nitrate evaporator in the THORP plant. Their paper sets down 

the lessons-learned from previous incidents and discusses the research and development work 

undertaken to enhance understanding of the nature and kinetics of these reactions. They use 

their findings to analyze evaporator behavior and to identify scenarios that entail thermal 

runaway. Then, they outline their PRA approach for defining the frequency of red-oil hazards.

No other PRA studies of THORP were found in the public literature, partly due to commercial 

confidentiality and partly to security concerns. One issue recognized by the U.K. authorities is 

that PRAs cannot be split so easily into Levels1, 2, and 3 as the plant does not have a simple 

set of barriers. Two other problems are the changing nature of the hazard as the material 

moves through the plant and changes its physical state, and the need to account for much more 

human interaction as the process develops. In the U.K. analyses, this makes the form of the 

PRA somewhat different from that for reactors as the balance tends more towards fault trees, 

rather than event trees. 

The U.K.’s HSE issued its Safety Assessment Principles (SAPs) for Nuclear Facilities [2006], 

and more recently published a Technical Assessment Guide (TAG) on Probabilistic Safety 

Analysis (PSA) [HSE 2009]. The purpose of the TAG is to “...provide an interpretation of those 

Safety Assessment Principles ... related to PSA and to provide specific guidance to inspectors 

engaged in the assessment of PSAs and PSA related submissions ... from Licensees, License

Applicants or Generic Design Assessment (GDA) Requesting Parties...” This guide does not 

apply different approaches or methods for evaluating reactor and non-reactor nuclear facilities. 

However, it recognizes that a comprehensive PRA, including quantitative evaluations, must be 

carried out for reprocessing facilities, as reflected by the following quote from its Section 3.2,

“Fault analysis: PSA – Need for a PSA – FA.10”

14:

The depth of the PSA for a given facility may vary depending on the magnitude of the radiological 

hazard and risks and the complexity of the facility. For example, for some facilities simplified 

analyses, or even qualitative arguments, application of good practice and DBA [Design Basis 

Analysis] may be sufficient to demonstrate that the risk is ALARP [As Low As Reasonably 

Practicable]. However, for complex facilities such as power reactors or reprocessing facilities, 

comprehensive PSAs that meet modern standards should be developed for all types of initiating 

faults and all operational modes.

 14 Fault analysis principle “FA.10,” which is one SAP, was titled “Need for PSA.” 
28

3.2.4 La Hague

Simonnet [2004] reports that PRA was used in parts of the design of the La Hague reprocessing 

facility in France. No other PRA study of this facility was found in the public literature. 

On the other hand, the French Atomic Energy Commission (Commissariat à l'Energie Atomique, 

or CEA) published a study on applying PRA to non-reactor nuclear facilities (NRNFs) [Bassi

2005]. It points out that “The bibliography analysis shows that the PSA [Probabilistic Safety 

Assessment] approach for NRNF is close to that currently adopted for the NPPs, but it has to be 

adapted due to the specificities of these plants...” As part of adapting the PRA method of power 

reactors to assessing the risk of an NRNF, Bassi proposes quantifying the risk using the 

“ARAMIS” approach [Delvosalle et al. 2006] that involves developing and quantifying a 

somewhat different version of fault trees and event trees. Bassi considers that “...ARAMIS ...

provides a semi-quantitative approach of the risk, potentially interesting for fuel cycle facilities, 

oriented towards supplying of whole risk management processes and regulatory demonstration, 

rather than towards an accurate quantification of the risk.” This approach is semi-quantitative

because it includes some qualitative aspects. 

The French study recognizes that reprocessing plants are amongst the three most hazardous

NRNFs, as Subsection 5.1.3, “Safety philosophy,” stated, 

“Because chemical processes form an integral part of the nuclear fuel cycle facilities, 

insurance of safety requires the control of both the chemical and nuclear hazards. It’s 

important to notice that the hazards vary from one facility to another, depending on the 

processes employed, the age, the output, the physical and/or chemical properties of the 

substances, and possibly the specific conditions [TECDOC1221]...Therefore, reprocessing 

plants, high activity liquid waste tanks, and plutonium handling plants are the most 

dangerous facilities, even if the nature of the dangers is globally the same in the whole fuel 

cycle.”

3.2.5 Electrochemical Processing (Pyroprocessing)

SECY-08-0134 [NRC 2008b] notes that the two main methods of reprocessing of spent nuclear 

fuel (SNF) used to date are aqueous separations and electrochemical processing. The former 

employs solvent-extraction techniques for purification. The latter uses an electrochemical 

technique to purify spent fuel. Electrochemical techniques generate a fuel that is not as pure as 

aqueous reprocessed fuel, and consequently, this fuel currently is only suitable for the recycling 

of fuel in advanced burner reactors (ABRs) (fast-neutron reactors), where these “impurities” can 

be burned. 

This subsection summarizes some observations related to the safety of electrochemical 

processing of SNF. In particular, NUREG-1909 points out the following:

• Electrochemical processing inherently is a batch process so that materials must be 

moved as solid physical objects in most of the various steps involved. The size of the 

batches is limited by criticality considerations. The large number of movements of highly 

radioactive objects containing fissile materials is likely to necessitate high equipmentreliability, low accident-likelihood, and a great need for nuclear-material accountability.
29

• There is no estimate of the amount and characteristics of failed or used equipment, such 

as electrodes and crucibles.

• Electrochemical processing per se does not use organic chemicals. This avoids the 

potential for accident scenarios involving organic chemical reactions (e.g., fire, red oil, 

resin explosions) and wastes from the cleanup of organic solvents and extractants.

Mariani et al. [1995] studied the criticality safety of the electrorefiner (ER) of the Fuel Cycle 

Facility (FCF) at Argonne National Laboratory West. They state, “Since the FCF ER is a 

complicated assembly of hardware, and the ER processes themselves are complex, the number 

of issues relevant to evaluation of the FCF ER process for criticality safety is substantial. The 

strategy to maintain criticality safety in the FCF ER process is summarized here, giving a few 

detailed examples of how the strategy is applied to static inventories, process items, and 

operations. A full account of the applied strategy has already been given [Mariani et al. 1993].” 

They also indicate, “In the absence of extensive statistical data on the operation of similar 

facilities, no formal PRA was performed. Consequently, the definition and classification of 

abnormal events required careful examination of the design and operation of the ER and 

required application of sound technical judgment. The classification of individual events as 

unlikely or extremely unlikely was based on the following technical issues: 

a) Equipment and container designs, 

b) Physical limits and controls, 

c) Administrative controls, 

d) Criticality control limits, 

e) Process variations, 

f) Sampling and analysis uncertainties, 

g) Distinguishability of different material forms and containers, and 

h) Number of steps or length of time required for the event to develop without notice.

According to the report by the Committee on Electrometallurgical Techniques for Department of 

Energy (DOE) Spent Fuel Treatment [NAS 2000], the Argonne National Laboratory’s (ANL)

Experimental Breeder Reactor-II (EBR-II) Spent Nuclear Fuel Treatment Demonstration Project 

began in June 1996 and ended in June 1999. Four criteria evaluated its success, addressing

the process, the waste streams, and the safety of the electrometallurgical demonstration project. 

Criterion 4 required demonstrating that safety risks, environmental impacts, and nuclearmaterials accountancy are quantified and acceptable within regulatory limits. One goal for 

meeting this criterion was to estimate the safety risks, environmental impacts, and material 

accountancy for the inventory operations. The Committee believed that this goal was met, 

based on ANL’s safety analysis [Garcia et al. 1999].

15

In previous reports, this Committee noted their concerns about the scale-up of the HIP16

process. In particular, the National Research Council [1999] pointed out that ANL-West “...is 

working with an outside vendor to produce larger beryllia17 crucibles needed to increase the 

 15 This document was not found in the public literature.

16 Salt-loaded zeolite is mixed with a borosilicate glass and consolidated at high temperature (850 to 

900°C) and pressure (14,500- to 25,000-psi) in a hot isostatic press (HIP) to make the final waste form.

17 Beryllia is another name for beryllium oxide (BeO), a white crystalline oxide. It is notable as it is an 

electrical insulator with a thermal conductivity higher than any other non-metal except diamond, and 

actually exceeds that of some metals. Its high melting point leads to its use as a refractory. 
30

throughput of the cathode processor. The scale-up of beryllia crucibles continues to be a 

problem at the outside vendor. The larger beryllia crucibles are failing mechanically, apparently 

due to thermal stresses.”

Appendix B briefly describes electrochemical processing and the associated literature. 

3.3 Insights Gained from the Survey of Accidents and Risk Analyses

The following insights were obtained from the survey of accidents:

1. The most important conclusion is that accidents at reprocessing facilities can result in 

very severe consequences, up to, and including early fatalities and injuries of personnel, 

and substantial releases of radioactivity to the environment. The definition of highconsequence events in 10CFR70.61, “Performance requirements,” includes, but is not 

limited to, those internally or externally initiated events that result in an acute worker 

dose of 1 Sv (100 rem) or greater total effective dose equivalent, or an acute dose of 

0.25 Sv (25 rem) or greater total effective dose equivalent to any individual located 

outside the controlled area. The information available about these accidents was not 

enough to assess which ones met these criteria, if any. On the other hand, those 

involving early fatalities clearly exceeded them. In addition, those having a classification 

of 4, “Accident with local consequences” or higher in the INES scale also are significant 

accidents. Table 4 presents accidents with documented early fatalities and accidents 

satisfying this classification. 

Table 4: Significant Accidents in Reprocessing Facilities Worldwide

(The number in parenthesis after an accident is the number of resulting early fatalities, if 

available.)

INES Kind of Hazard

Level Type of Event Criticality Radiation Fire and/or explosion

6 Serious 

accident

Mayak PA, 9/29/1957

4

Accident with 

local 

consequences

Windscale, 1973

Classification not 

available

Mayak PA, 4/21/1957 (1)

Mayak PA, 1/2/1958 (3)

Los Alamos Laboratory, 

12/30/1958 (1)

Mayak PA, 12/10/1968 (1)

Mayak PA, 10/1/1951 (1) Tomsk, Russia, 

4/6/1993

The accident at the Mayak PA facility on September 29, 1957 due to a chemical 

explosion may be considered a very high-consequence accident as it released a total 

activity of approximately 74 petabecquerels (PBq18) which were dispersed offsite over

the territory of the Chelyabinsk, Sverdlovak and Tyumen regions.

 18 A PBq is equal to 1015 disintegrations per second (dis/s), and 1 Ci is equal to 3.7 x 1010 dis/s. Hence, 

74 PBq is approximately equal to 2 x 106 Ci. 
31

2. Reprocessing facilities entail the risks associated with the traditional process industries, 

such as chemical toxicity and chemical reactions leading to fire and explosion, as well as 

the risks related to nuclear materials (such as radioactive contamination and criticality 

excursions). An accident may involve a combination of nuclear- and non-nuclearhazards. 

3. Many of the identified accidents occurred during the 1950s and 1960s, presumably 

because at that time, in comparison to current regulations and standards, there were 

minimal safety standards and/or regulatory oversight was relaxed or lacking. While 

accidents during the last two decades were fewer, serious events still occurred, such as 

those at Siberian Chemical Enterprises (Russian Federation, 1993), THORP (U.K., 

August 2004), and Tokai-mura (Japan, March 11, 1997). Hence, apparently there is a 

trend toward decreasing number and severity of consequences of accidents. This is 

likely due to a combination of implementing safer work practices, increased operating 

experience, better safety standards, and/or stricter regulatory oversight; it may also 

reflect the fact that several countries have shut down their main reprocessing facilities 

(Table 1). 

4. Accidents of differing severities (from relatively mild to severe) have occurred in

practically all countries having reprocessing facilities (Table 1). No known accidents at 

reprocessing facilities have occurred in Belgium and China.

5. Two of the identified events (Sellafield, 1983; Karlsruhe, 1999) happened during 

shutdown of a facility, indicating that there also is risk associated with the shutdown of a 

reprocessing facility. 

The following conclusions were reached after studying the methods and results of risk 

assessments of reprocessing facilities:

1. A thorough approach, based on a comprehensive PRA that takes into account both 

radiological and chemical hazards, needs to be considered to assess the risk to 

receptors from reprocessing operations. 

2. One Japanese PRA study used a simplified PRA for a reprocessing facility. A detailed 

PRA performed for a limited set of accident scenarios associated with this facility first 

was undertaken, and its models, data, and results were used to develop the simplified 

PRA. The Japanese researchers that developed and applied the simplified PRA 

approach to this facility consider it an improvement over the “risk index method” that is 

typically used for the ISA. 

3. The comprehensive risk assessment includes a quantitative evaluation of the risk to 

receptors. Though failure data may be scarce for this kind of facility, the available data 

is supplemented with engineering judgment, as was done in the early days of applying 

the PRA method for nuclear power reactors, and still is used for some technical issues. 

4. Since nuclear- and chemical-hazards are distributed throughout reprocessing facilities, 

approaches that are typically not used for the PRA of nuclear-power reactors, such as 

methods of hazard analysis and risk evaluation for chemical plants, are employed to 

identify, model, and quantify the risk of individual processes or the entire facility. 

33

4. QUALITATIVE VS. QUANTITATIVE RISK METHODS

4.1 Introduction

SECY-09-0082 [NRC 2009] updated the progress towards developing a regulatory framework 

for licensing reprocessing facilities, including a regulatory gap analysis. In terms of risk 

considerations, Gap #5 indicated that (1) reprocessing facilities would have a higher source 

term, and thus, present a greater relative risk compared to fuel-cycle facilities licensed under 10 

CFR Part 70, and, (2) the ISA methodology for risk assessment in Part 70 does not adequately 

address the risks posed by facilities with higher risk. The gap analysis recognized the 

ACNW&M’s recommendation for a quantitative approach to risk, such as PRA, rather than ISA 

because of the limitation in the ISA’s treatment of dependent failures, human reliability, 

uncertainties, and its aggregation of event sequences. However, the gap analysis also 

expressed some reservations about PRA use in reprocessing facilities, based on lack of 

relevant, reliable data. 

Part 70 requires an ISA, and the risk index method used by most applicants and licensees 

involves an order-of-magnitude analysis. However, for a reprocessing facility, risk-informing the 

facility’s design and operation through a quantitative risk assessment, in conjunction with the 

quantitative risk guidelines proposed by the staff in SECY-04-0182 [NRC 2004] and later 

updated in [NRC 2008a], could have major benefits, such as:

1. Significantly enhance the ability of the NRC’s staff to better understand and categorize 

risk-significant issues. This clearly would be useful in the license review process, in 

ensuring that the risks posed by a new facility or by additions to an existing one are well

characterized and understood. 

2. Play an important role in the inspection process in evaluating the risk significance of the 

inspection findings, in a manner similar to the risk-oversight program for power reactors.

3. Identify risk-important sequences and components to support the allocation of resources 

to decrease the risk.

Some experience was gained in applying PRA techniques to facilities carrying out operations 

somewhat similar to large-scale reprocessing facilities. BNL conducted a limited PRA to 

analyze the risks of “red oil” explosions in the mixed oxide fuel-fabrication facility (MFFF) 

currently under construction at the DOE’s Savannah River site in Aiken, SC. The MFFF uses 

aqueous processing based on the PUREX process to separate impurities in plutonium 

feedstock in manufacturing mixed oxide (MOX) fuel. The MFFF is being licensed under Part 70 

using ISA methodology, but a limited PRA was carried out because it was felt that PRA methods 

could offer useful risk insights to NRC’s staff reviewers in their analysis of some highconsequence events, such as red-oil explosions. The experience gained through this exercise 

indicated that while ISA is a useful starting point, it has several limitations, particularly in 

analyzing common-cause failures and human reliability. PRA has advantages in this regard.

The experience of applying PRA techniques to MFFF showed that employing surrogate data 

from related facilities with similar equipment affords usable results for analyzing the safety of the 

facility. Over time, the issue of data should become less important for facilities, such as fuel 

reprocessing plants, as demonstrated by experience with data needs for reactor PRA over the 

last three decades. However, PRA techniques, developed mainly for operating light-water 

reactors, must be enhanced in several ways to facilitate their application to the safety of 
34

nuclear- and chemical-processes that are characteristic of reprocessing plants. Section 4.4 

offers suggestions for these enhancements.

4.2 Approach

As discussed in Chapter 2, based on work carried out by the ACNW&M [Croff et al. 2008] as 

well as the NEI report [2008], the approach to conducting risk assessment of a reprocessing 

facility should be a technology neutral one since the overall regulation may need to 

accommodate different reprocessing technologies currently in various stages of development. 

While aqueous technologies such as the PUREX process are mature, other technologies are 

currently under development. Technology-specific regulatory guides could then be used to 

supplement the regulation. The ACNW&M mentioned the technology neutral approach 

identified in NUREG-1860 [NRC 2007] for risk-informed performance-based licensing of future 

power reactors (referred to in Chapter 2) as offering a valuable basis for developing a similar 

approach to reprocessing facilities.

At a high level, the following features of the technology neutral approach identified in NUREG1860 are relevant:

1. The approach should be risk-informed and performance-based. Risk-informing is a 

philosophy that considers risk insights together with other factors to establish 

requirements that better focus the attention of the licensee and regulatory body on 

issues of design and operation commensurate with their importance to public health and 

safety. Risk insights are derived from a risk assessment of the facility; ACNW&M stated 

its preference for PRA over ISA, as pointed out in Chapter 2, above. 

2. The facility design must encompass defense-in-depth and safety margins. 

3. Technology-neutral risk-acceptance criteria need to be formulated to help in developing 

the licensing basis for the facility. Here, the frequency-consequence curve, discussed 

below, based on the one offered in NUREG-1860 is a possible candidate. 

It should be noted that analogs of the reactor safety goals, known as quantitative risk guidelines, 

were developed for non-reactor nuclear facilities, such as fuel-cycle facilities, in SECY-04-0182

[NRC 2004], and later updated in [NRC 2008a]; they could also apply to reprocessing facilities, 

regardless of technology.

4.3 Insights from the Technology Neutral Framework of NUREG-1860

The provision of a technology-neutral framework that can be created for diverse technologies,

using important probabilistic- and deterministic-criteria governing risk and performance, will 

facilitate developing a consistent, stable, and predictable set of requirements that are both riskinformed and performance-based. One important feature of the framework is developing a risk 

acceptance criterion that relates both elements of the risk, frequency, and consequence, posed 

by the facility or process regardless of technology. NUREG-1860 achieved this via constructing

a frequency-consequence curve.

A criterion that specifies limiting frequencies for a spectrum of consequences, from very small to 

very large, can be denoted via a frequency consequence (F-C) curve. On the F-C plane, this
35

curve provides an acceptable limit in terms of the frequency of potential accidents and their 

associated consequences. The objective of such a curve would be to establish the licensing 

basis, i.e., to identify the event sequences that must be mitigated via the design and operation 

of the plant. This objective involves first establishing criteria for ensuring that the frequency of 

occurrences of event sequences is inversely related to the consequences, e.g., high-frequency 

events such as minor transients should have low consequences, and high-consequence events 

like an explosion followed by a large release of radionuclides should have low frequency. 

Second, the objective involves establishing criteria that define the acceptable frequencies for 

different levels of consequences.

4.4 PRA for Reprocessing Facilities

PRA is a fairly mature technology for understanding the vulnerabilities, and predicting the risks 

posed by commercial nuclear-power reactors. However, fuel-cycle facilities also are chemical 

processing plants, which present a different set of challenges than power reactors. These 

differences include the nature and type of hazards they pose, the kinds of accidents that can 

occur, and the recipients of the risk. This is recognized in the 10 CFR Part 70 regulations 

governing the licensing of these facilities; in particular, the performance criteria in Part 70.61 

specify both the radiological- and chemical-hazards posed by fuel-cycle facilities and establish 

limits on their consequences for the public, workers, and the environment as a function of the 

likelihood of events.

Furthermore, the risk assessments for facilities with nuclear- and chemical-processes like 

reprocessing facilities must account for several features that distinguish such facilities from 

reactors. These elements, analyzed below for enhancing traditional PRAs, are based on the 

following: (1) The insights gained from previous work on the risk assessment of possible red-oil 

excursion events in the proposed MFFF currently under construction at the DOE’s Savannah 

River site, (2) a review of both domestic- and international-safety activities and databases (e.g., 

FINAS and NMED, mentioned by the end of this section) of non-reactor nuclear facilities, and 

(3) a review of the challenges faced by the ISA methods that are mandated by 10 CFR Part 70 

regulations. 

The risk assessment methods able to respond to the needs of the reprocessing facilities should

account for the unique features of these facilities that make their risk profile different from those 

of commercial power reactors. Some of the major differences are noted below:

1. The hazards posed by the facility include toxic chemical- and explosion-hazards in 

addition to radiological hazards.

2. There is no analog of the reactor core as the main source of hazard in the facility; the 

source term for both chemical- and nuclear-hazards might be distributed throughout the 

plant, with the amount in each location varying depending on processing operations.

3. As noted in the report IAEA-TECDOC-1267, “Procedures for Conducting Probabilistic 

Safety Assessment for Non-Reactor Nuclear Facilities” [IAEA 2002], many current nonreactor nuclear facilities rely heavily on manual control during normal operation, as well 

as manual actuation to respond to faults and potential accident conditions. Future 

facilities, as reflected in the designs of proposed fuel-fabrication or reprocessing 

facilities, are transitioning to fully automated actuation, control, and monitoring.
36

4. There is a major reliance on operating systems for prevention and control, with less 

reliance on standby systems.

Due to these differences, the risk and reliability methods, data, and software tools developed for 

commercial nuclear power plants could be modified to become more suitable for use in the risk 

assessment of non-reactor nuclear facilities, such as reprocessing plants. In some cases, new 

tools or new databases may need to be developed.

The following two examples illustrate some of the major differences in the risk- assessment 

methods needed for these facilities versus those employed in the current PRAs of power 

reactors. 

Example 1: An accident scenario starts with the introduction of some undesired processing 

chemicals from one vessel to another. Once the materials are introduced into the recipient 

vessel, a vulnerable condition can be created that will last for a period governed by the 

response time of the process-control mechanism. For an accident to occur, however, another 

condition (say, high temperature) may need to also occur during the time that the undesired 

processing chemical remains and is not cleared.

This example identifies several PRA modeling issues as follows:

1. The initiating event frequency is the frequency of occurrence of the vulnerable condition 

and its intensity (i.e., the frequency and amount of undesired chemical transfer to the 

improper location).

2. The period of vulnerability is a random variable governed by the process unit, and 

response time of the associated process control. 

3. Since the systems are running continuously, the failure frequency (not the unavailability

as usually is estimated in PRA codes) of the systems responsible for controlling the 

mixture temperature within the period of vulnerability must be estimated. 

This example highlights three important issues relating to PRA models:

1. The chronological/temporal sequence of events is important. Hence, event A followed 

shortly by event B will not have the same impact as event B followed shortly by event A. 

This is equivalent to saying the failure of A will challenge B but not the reverse.

2. The system’s survival probability within the period of vulnerability must be assessed, 

rather than the average system unavailability. 

3. Both the duration and the chronological sequence of events need to be explicitly 

considered as a part of the model.

Example 2: A low set point on a relief valve could cause it to open and subsequently close 

under normal pressure variation during operation. The low set point caused by drift is such that 

it opens the relief valve on average once every 10 hours. Each time the relief valve opens, the 

unwanted materials that are transferred to the vessel in which operations are taking place 

accidentally can be transferred to another vessel. The probability must be calculated of a 

scenario wherein sufficient material is transported and accumulates in the second vessel 
37

leading to an undesired consequence. Enhancements to fault trees or other methods should be 

defined to evaluate such cumulative probabilities. 

Several areas require further development to meet the needs for undertaking a risk assessment 

for fuel-cycle facilities. Enhancing the PRA tools to address more systematically the heavy 

reliance on process control and the use of operating systems is a high priority. The reliability 

and risk assessment of these areas require a dynamic approach, namely, treating time as an 

explicit independent variable. The objective is to develop tools and techniques for performing a 

reliability evaluation of the time-evolutionary path of a dynamic system.

This objective could be accomplished by methods to enhance the existing fault tree and event 

tree routines and their evaluation algorithms to address the risk and reliability issues associated 

with the fuel cycle facilities. The evaluation algorithms could include stochastic point process 

models including Markov and Semi-Markov models. These are considered as the preferred 

methods for quantification rather than the simple reliability equations currently used.

The approach needs to review and integrate the existing risk and reliability methods used in 

several industries, such as the nuclear-, chemical-, and aircraft-industries. The following are 

examples of such methods: PRA, ISA, Layers of Protection Analysis (LOPA), Process Hazards 

Analysis (PHA), Hazard and Operability Analysis (HAZOP), Event Sequence Diagram (ESD), 

and Petri nets. In particular, attention must be paid to the methods recommended in IAEATECDOC-1267, “Procedures for Conducting Probabilistic Safety Assessment for Non-Reactor 

Nuclear Facilities” [IAEA 2002].

The approach needs to determine the best, most suitable methods, and then identify the 

additional algorithms necessary for enhancing the existing methods to facilitate a more 

systematic evaluation of the issues emerging from risk evaluations of fuel-cycle facilities. They 

will include addressing the methodological needs of, and developing the tools for enhancing 

risk-assessment techniques. Also, identify those algorithms useful for integrating and 

transferring the modified fault trees and event trees to an equivalent Markov/Semi-Markov 

model; this will support the evaluation, and identify the needed parameters, along with the 

potential performance indicators that could aid in monitoring vulnerable states (also of 

importance for use in oversight program). Lastly, the types of models and potential sources of 

data needed for such analyses must be addressed. 

Substantive reviews are needed of several important and available data sources, including the 

ones identified below:

1. Fuel Incident Notification and Analysis System (FINAS) database established by the 

Committee on Safety of Nuclear Installations of the Nuclear Energy Agency since 1992, 

now containing 90 events [NEA 1996]. 

2. Nuclear Material Events Database (NMED) maintained at Idaho National Laboratory 

(INL) for the NRC.

3. Russian statistical analyses on fuel-cycle facilities [Kolesnikov 2000].

4. Non-reduced (i.e., raw) data from Savannah River, Hanford, and other domestic DOE 

facilities.
38

5. Data from various experiments and accidents that will help in developing the knowledge 

base database for constructing fault trees. 

6. Data from other large international fuel-cycle facilities.

These databases will help to identify the types of data needs and modeling requirements for 

assessing accident progression and conducting consequence analyses in fuel-cycle facilities. 

This includes identification of the end-states of accident sequences that involve failure of 

systems designed to control the releases of radiological materials or toxic chemicals that can 

lead to exposure of the public, workers, or the environment; the estimation of source terms

resulting from the accident; and the calculation of consequences. The chemical source terms 

and their consequences demand particular attention, including mechanisms and amounts of 

releases, the transport of releases, and the resulting health effects models and standards for 

different chemicals. Detailed consideration is also needed of the radiological consequence 

models recommended in the report IAEA-TECDOC-1267 [IAEA 2002] (the chemical 

consequences were out-of-scope of this IAEA report, but are vital in NRC-regulated fuel-cycle 

facilities, as recognized in 10 CFR Part 70.61 criteria). Various governmental and industrial 

organizations that developed or adopted models for chemical release, transport, and 

consequence include the U.S. DOE (e.g., the models studied in the DOE’s Accident 

Progression and Consequences, APAC, program [Chung 2002]), the U.S. Environmental 

Protection Agency (e.g., ALOHA [EPA 2007]), the chemical industry (e.g., the models 

recommended by the American Institute for Chemical Engineers, i.e., [AIChE 2000] and [AIChE 

2001]), and those mentioned in previous NRC work (NUREG/CR-6410 [NRC 1998]). These 

models and approaches need to be reviewed along with chemical-exposure standards and 

limits developed by various industry bodies for their applicability and usefulness to fuel-cycle 

facilities. The potential interactions between nuclear- and chemical-hazards should be 

addressed as well.
39

5. SUMMARY AND OBSERVATIONS

5.1 Summary

This report discussed the current regulatory context for reprocessing facilities both nationally 

and internationally. As noted in SECY-09-0082 [NRC 2009], the existing requirements in 10

CFR Part 70 do not adequately address the increased risk posed by a reprocessing facility 

compared to that of other fuel-cycle facilities. Reprocessing facilities can have higher potential 

source-terms than other fuel cycle facilities, and accordingly, this may increase their overall risk. 

The present report concurs with this observation as a result of examining events that have 

happened in such facilities, and other risk information related to their hazards. In particular, as 

demonstrated in our review of accidents in reprocessing facilities, it is evident that the potential 

for health risk to the public is significantly greater than for other types of fuel-cycle facilities. 

Moreover, accidents at reprocessing facilities can result in very severe consequences, 

especially onsite including early fatalities and injuries of personnel. Accidents with documented 

early fatalities and accidents having a classification of 4, “Accident with local consequences” or 

higher in the INES scale are significant accidents; Table 4 presents them. The accident at the 

Mayak PA facility in September 1957 due to a chemical explosion released a total activity of 

approximately 74 PBq (approximately 2 x 106 Ci) which were spread offsite over the 

Chelyabinsk, Sverdlovak and Tyumen regions of the then Russian Federation.

5.2 Observations

NRC licensees and applicants currently implement varying degrees of ISAs, that is, from 

qualitative to semi-quantitative. Most of them use the order-of-magnitude approach described in

NUREG-1520; Revision 1 of this NUREG was published in 2010 [NRC]. For relatively simple 

nuclear fuel-cycle systems, the ISA approaches may identify potential weaknesses in a facility’s 

design or operation, and enable the identification of IROFS. However, since the approaches do 

not incorporate inter-system dependencies, nor provide an integrated assessment of risk, they 

could miss some essential risk outliers in more complex facilities.

The Commission’s PRA policy statement issued in 1995 states in part, “The use of PRA 

technology should be increased in all regulatory matters to the extent supported by the state of 

the art in PRA methods and data, and in a manner that complements NRC’s deterministic 

approach and supports the NRC’s traditional defense-in-depth philosophy.” The authors of the 

present report observe that reprocessing facilities are distinct from other fuel-cycle facilities,

which are less complex and do not represent a comparable level of risk.

Safety analyses of a reprocessing facility can benefit greatly from the systematic, disciplined 

procedures embodied in risk assessment methodology. As discussed in Sections 3.2 and 3.3, 

varying degrees of PRAs for reprocessing facilities already have been carried out in several 

countries. Notwithstanding the limited data available for PRAs of this type of facilities compared 

to that for power reactors, the safety analyses of these facilities can benefit from the potential 

understanding gained by uncovering potential weaknesses in design and identifying dominant 

contributors to the risk of a plant or facility, such as human errors and dependencies. Moreover, 

PRAs can indicate those areas of a facility that are not significant from a risk viewpoint. 

Consequently, in a reprocessing facility, regulatory attention and resources can be allocated 

according to the risk importance of operator actions, and of its structures, systems, and 
40

components. In addition, the approaches to uncertainty analysis commonly employed in reactor 

PRA studies can be very useful in expressing risks for reprocessing facilities. 

5.3 Suggested Considerations

This report documented several accidents at reprocessing facilities, some of which entailed 

high-consequence events as previously discussed. Studies of these accidents might well offer 

more details to gain a better understanding of their causes and mechanisms of occurrence. The 

potential benefits from such studies could lie in identifying weaknesses in the design and/or 

operation of these facilities, and in establishing areas requiring additional regulatory attention. 

In addition, as Volume 9 of the IAEA-TECDOC-1575, Rev. 1 [IAEA 2008d] points out, further 

development of PRA methods and their supporting data bases are needed to assess the risk 

associated with reprocessing facilities. 

Some PRA studies of reprocessing facilities were carried out in Japan, and the U.K. The review 

of these studies documented in this report was brief for two main reasons: 1) The studies are 

not available publicly, so only brief articles were consulted, and, 2) the limited scope of this 

work. However, if obtained, a review of the original studies could be conducted to assess 

valuable information, such as the techniques, scope, data, and computer codes employed, and 

differences between them and a typical reactor PRA. In addition, insights might be obtained 

about these facilities’ safety. 
41

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 This document contains a note stating that it “is a working Document for the 'STOA Panel'. It is not an 

official publication of STOA. This document does not necessarily represent the views of the European 

Parliament.” 
A-1

APPENDIX A: Brief Description of Events at Reprocessing Facilities

Tables A-1, A-2, A-3, A-4, A-5, A-6, and A-7, respectively, summarize accidents at reprocessing 

facilities in France, Germany, India, Japan, Russian Federation, U.K., and the USA. In some 

events included in the tables for the Russian Federation and the USA, the information about an 

accident did not specify the type of nuclear facility (i.e., whether it was a reprocessing facility or 

not); these events were included for the sake of completeness. 

The International Nuclear and Radiological Event Scale (INES) is used for communicating to the 

public the safety significance of events associated with sources of radiation. The scale was 

developed in 1990 by international experts convened by the IAEA and the Nuclear Energy 

Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD). 

According to the INES User’s Manual [IAEA 2009], events are classified on the scale at seven 

levels: Levels 4–7 are termed “accidents” and Levels 1–3 “incidents” events without safety 

significance are classified as “Below Scale/Level 0.” Events that have no safety relevance with 

respect to radiation or nuclear safety are not classified on the scale. 

The tables in this appendix include the INES classification of each event. However, most of 

these events have not been classified according to this scale or their classification was not 

found in the public literature. In this case, the column with the heading “INES Level” contains 

“NA” (not available). Some of the unclassified events have relatively minor consequences, such 

as an inadvertent spill of radioactivity within a room of a facility with no injuries to workers, and 

no release of radioactivity from the room. However, it is clear that other unclassified events 

would have severe consequences and be categorized as accidents in this scale. For example, 

early fatalities have occurred as a result of an accident; here, “early fatality” is defined loosely as 

a death that occurred within a few months after the accident occurred. If an accident description 

does not have an indication of early fatalities, it does not necessarily mean that there were 

none; it may be due to lack of information about the consequences of the accident. In addition, 

other “latent fatalities” (those that may happen with a much longer time lag following exposure, 

such as latent cancers due to radiation exposure) may have occurred as a result of some of the 

accidents, but this information was unavailable to this study. 

All early fatalities were workers of the reprocessing facilities. Taking into account these fatalities 

only, apparently the event at Mayak PA in the Russian Federation on 1/2/1958 (3 early fatalities) 

would be categorized as Level 5, “Accident with wider consequences,” and the events involving 

one early fatality (Mayak PA, 10/1/1951; Mayak PA, 4/21/1957; Los Alamos Scientific 

Laboratory, 12/30/1958; Mayak PA, 12/10/1968) would be classified as Level 4, “Accident with 

local consequences,” in the international scale. 

Some tables contain more accidents than others; however, this does not necessarily reflect the 

safety record of a country or facility. This difference simply is considered to be mainly due to 

the availability of information from different countries about this type of accident. 
A-2

Table A-1: Events at Reprocessing Facilities in France

Facility and 

Date of 

Accident

Summary INES 

Level

Type of 

Hazard 

Facility at 

FontenayAux-Roses, 

June 26, 1962

Explosion (no detailed information was found about this accident). NA Fire / 

explosio

La Hague, 

October 2, 

1968

An abnormal rate release of gaseous iodine-131 (131I) was detected at the 

UP2-400 factory. This high-level release (37,000 Bq/s) continued for eight 

hours, and then decreased for 15 hours before activity was under the 

authorized limit (370 Bq/s). The total amount of iodine-131 released was 

evaluated at 185 GBq (the current authorization for iodine is 110 GBq per 

year). This accident was caused by treating insufficiently cooled gasgraphite fuels. Also, a release of iodine-129 (129I) was reported (7.4 GBq). 

There was no evaluation of the impact on public health. 

NA Radiatio

La Hague, 

January 14, 

1970

During the chemical dissolution of the fuel, the temperature of the reaction 

increased sharply and an explosion occurred due to the emanation of 

hydrogen gas (H2). The filters of the chimney recorded an activity of 5,900 

GBq, mainly due to antimony-125 (125Sb), 95%, and iodine-131 (131I), 5%. 

The iodine-129 activity released was 2.7 times higher than the annual 

authorized limit (110 GBq). No measurements in the environment were 

undertaken after this accident, nor was there an evaluation of the impact on 

public health.

NA Fire / 

explosio

La Hague, 

October 1, 

1976

A badly designed packaging of tritium (3H)-rich wastes allowed the release 

of large amounts of tritium in the “Sainte-Hélène” stream: the activity 

reached 7,400 Bq/l in October 1976. Streams outside nuclear areas 

usually exhibit a level below 1 Bq/l. From 1977 to 1983, the mean annual 

tritium activity released was about 5,000 Bq/l. These unexpected high 

levels of contamination were due to this leakage. Re-packaging operations 

entailed the further release of 52,000 GBq of tritium. Many other 

radionuclides were detected over time in this stream, like strontium-90 

(

90Sr), caesium-137 (137Cs), and cobalt-60 (60Co).

NA Radiatio
A-3

Table A-1: Events at Reprocessing Facilities in France

Facility and 

Date of 

Accident

Summary INES 

Level

Type of 

Hazard 

La Hague, 

January 2, 

1980

A one-meter long through-wall crack was discovered on the La Hague 

discharge pipe at about 200 meters from the shore. Fishermen received 

3.5 times the annual authorized dose, namely 3.486 mSv, compared with 

the current European limit of 1 mSv/year.

NA Radiatio

La Hague, 

April 1980

There was a loss of coolant resulting in gaseous releases (no detailed 

information about this accident was found1). 

NA Radiatio

La Hague, 

January 6, 

1981

Graphite elements burned for 24 hours in a waste silo. The uranium metal 

caught fire following a mechanical shock during operations. The maximum 

measured level of air contamination, 700 Bq/m3

, was reached 10 hours 

after the fire began. The released activity was mainly due to caesium-137 

and -134 (137Cs and 134Cs), and ranged between 740 GBq and 1,850 GBq. 

Strontium-90 (90Sr) was detected in rainwater. A worker received in one 

day the annual admissible dose, 50 mSv. There was no assessment of the 

off-site health impact.

3 Fire / 

explosio

La Hague, 

1983

COGEMA said that strontium-90 contamination was caused by metallic 

waste stored in concrete pools that released radionuclides into the 

groundwater and nearby streams. At the end of 1999, the contamination 

still was detected: 20 Bq/l in 1991 and between 5 and 10 Bq/l later 

(maximum admissible concentration in drinking water: 36 Bq/l). 

NA Radiatio

La Hague, 

February 13, 

1990

A routine replacement of a chimney filter led to the release of non-filtered, 

contaminated air for 10 minutes. An estimated 3.7 MBq of caesium-137 

were released in the atmosphere. The radiological consequences were not 

evaluated.

NA Radiatio

 1

 The internet site http://prop1.org/2000/accident/facts5.htm briefly describes this alleged event. However, the site’s reputa
A-4

Table A-1: Events at Reprocessing Facilities in France

Facility and 

Date of 

Accident

Summary INES 

Level

Type of 

Hazard 

La Hague, 

March 11, 

1997

An independent laboratory measured a high level of dose- equivalent rate 

close to the discharge pipe at low tide. COGEMA confirmed the value of 

0.3 mSv/hour, giving an individual the annual dose in three and a half 

hours. With this exposure and the strontium-90 (90Sr) concentration 

reached inside the pipe, the vicinity of the pipe would have required 

classification as a “nuclear facility” in accordance with French legal 

framework on radioprotection.

NA Radiatio

Table A-2: Events at Reprocessing Facilities in Germany

Facility and 

Date of 

Accident

Summary INES 

Level

Type of 

Hazard 

Karlsruhe, 

1985

Fire with resulting gaseous releases (no detailed description of this event 

was found). 

NA Fire / 

explosio

Karlsruhe, 

between 

March and 

April, 1999

Radioactivity escapes from the shutdown reprocessing plant through a 

defective ventilation system, briefly exposing 31 employees to 

contaminated air in three cases.

NA Radiatio

 2 PortalU is a cooperation of the German "Länder" and the German Federal Government. Its website is http://www.portalu.
A-5

Table A-3: Events at Reprocessing Facilities in India

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Kalpakkam

Reprocessing 

Plant (KARP), 

January 21, 

2003

A leak in an isolation valve separating a high-level liquid waste tank from 

a low-level liquid waste tank resulted in an increase in the radioactivity 

level in the latter. Consequently, the six personnel who were involved in a 

transfer operation were exposed to a higher dose than the annual dose 

limit. There was no release of radioactivity to the environment. 

NA Radiatio

Table A-4: Events at Reprocessing Facilities in Japan

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Tokai-mura,

March 11, 

1997

A fire and explosion contaminated 37 workers with radioactive material. 

The activity released to the environment was in the order of 10-3 GBq for 

alpha nuclides and several GBq for beta nuclides. The committed dose 

equivalent was of a magnitude of 10-3 to 10-2 mSv.

3 Fire / 

explosio

Table A-5: Events at Reprocessing Facilities in the Russian Federation

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Mayak PA,3

August 19, 

1950

Azizova et al. [2005] characterized this accident as “...technology violation 

at the radiochemical plant.” One person suffered acute radiation 

syndrome (ARS) as a result of the accident. 

NA Radiatio

 3 Azizova et al. [2005] describe the Mayak Production Association (Mayak PA) as “...Mayak PA is located in the Southern 

region, which is approximately 1000 miles east of Moscow. Mayak PA includes nuclear reactors, radiochemical and pluton

waste storage areas.” González [1998] points out “Operation of uranium-graphite reactors for plutonium production a

began in 1948.”
A-6

Table A-5: Events at Reprocessing Facilities in the Russian Federation

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Mayak PA, 

September 

13, 1950

Azizova et al. [2005] characterized this accident as “...technology violation 

at the radiochemical plant.” One person suffered ARS as a result of the 

accident.

NA Radiatio

Mayak PA, 

September 

20, 1950

Azizova et al. [2005] characterize this accident as “...technology violation 

at the radiochemical plant.” One person suffered ARS as a result of the 

accident.

NA Radiatio

Mayak PA, 

October 1, 

1951

Azizova et al. [2005] characterize this accident as “...technology violation 

at the radiochemical plant.” Three people suffered ARS, and one died as 

a result of the accident. 

NA Radiatio

Mayak PA, 

September 

20, 1952

Azizova et al. [2005] characterize this accident as “...technology violation 

at the radiochemical plant.” One person suffered ARS as a result of the 

accident.

NA Radiatio

Mayak PA, 

March 15, 

1953

The accident occurred in a plutonium-processing building. The plutonium 

had been recovered from irradiated uranium rods. The plutonium solution 

was transferred between two vessels, assuming that one vessel was 

empty that actually was not. There was a single criticality excursion in a 

plutonium nitrate solution in an interim storage vessel. As a result, one 

operator received an estimated dose of 100 rad, and another operator 

received an estimated dose of 1,000 rad. The accident caused no 

physical damage to any equipment. 

NA Criticality 

Mayak PA, 

November 6, 

1954

Azizova et al. [2005] characterize this accident as “...technology violation 

at the radiochemical plant.” One person suffered ARS as a result of the 

accident.

NA Radiatio

Mayak PA, 

December 22, 

1955

Azizova et al. [2005] characterize this accident as “...technology violation 

at the radiochemical plant.” One person suffered ARS as a result of the 

accident.

NA Radiatio

Mayak PA, 

April 21, 1957

The accident occurred in a glove box in which an excess of uranium 

accumulated during the filtration of uranyl oxalate precipitate. The history 

of the criticality excursion is unknown. One operator died, and five others 

received doses estimated to be upwards of 300 rad.

NA Criticality 
A-7

Table A-5: Events at Reprocessing Facilities in the Russian Federation

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Mayak PA, 

September 

29, 1957

One of the cooling systems of the concrete high-level liquid waste storage 

tanks broke down permitting the tank to dry out and overheat. The 

chemical reaction of dry nitrate and acetate salts in the waste tank 

containing highly active waste caused a chemical explosion and a large 

release of radionuclides. The total activity dispersed off-site over the 

territory of the Chelyabinsk, Sverdlovak, and Tyumen regions was 

approximately 74 PBq. This area was later called the “Kyshtym footprint.” 

6 Fire / 

explosio

Mayak PA, 

January 2, 

1958

The accident involved an enriched uranyl nitrate solution that was being 

transferred from an experimental vessel to bottles. Rather than use a 

drain line as prescribed, three workers lifted the experimental vessel and 

began to move it (to pour the contents manually into bottles) when the 

criticality excursion occurred. A fourth worker, who was 2.5 m away, was 

also exposed to irradiation. Total neutron- and gamma-absorbed doses 

were estimated at 6,000 ± 2,000 rad for the three who lifted the vessel 

(who died in five to six days), and 600 rad for the coworker at 2.5 m. 

NA Criticality 

Mayak PA, 

December 5, 

1960

This accident occurred in a building where waste solutions were 

processed to recover plutonium. Recovery consisted of three successive 

stages of purification, and took place in two glove boxes. There were 

multiple criticality excursions associated with a plutonium-carbonate 

solution in a holding vessel. During the accident and the subsequent 

cleanup phase, five individuals received doses in the range 0.24 rem to 

about 2.0 rem. There was no contamination or damage to any of the 

equipment.

NA Criticality 

Mayak PA, 

September 7, 

1962

The accident occurred in a building that housed operations associated 

with converting plutonium feed-material to metal. The metal subsequently 

was purified in several processes and then cast into ingots. In each of 

these steps, dry residues were generated that contained recoverable 

quantities of plutonium. The accident occurred during the chemical 

dissolution of some of these residues in a vessel of unfavorable geometry. 

Three criticality excursions associated with a plutonium-nitrate solution in 

a dissolution vessel occurred, but there were insignificant exposures. 

NA Criticality 
A-8

Table A-5: Events at Reprocessing Facilities in the Russian Federation

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Siberian 

Chemical 

Combine,4

January 30, 

1963

This accident occurred in a waste recovery line of a uranium-metal 

production building. The waste feed was a dry precipitate. The first step 

in recovering the uranium was a concentrated nitric acid dissolution 

process. Multiple criticality excursions associated with a uranyl nitrate 

solution happened in a collection vessel. Four people standing 10 m 

away from the collection vessel received radiation doses of 6 to 17 rad. 

No damage occurred to the vessel, nor was there any contamination of 

the surroundings.

NA Criticality 

Siberian 

Chemical 

Combine, 

December 2, 

1963

This accident occurred in an enriched uranium reprocessing and 

purification facility. The combination of the unfavorable geometry of the 

holding vessel and the unplanned accumulation of much larger than 

expected quantities of organic solutions led to multiple criticality 

excursions associated with an organic solution of uranium in a vacuumsystem holding vessel. The largest individual dose received was less 

than five rem. There was no damage to the equipment or radioactive 

contamination.

NA Criticality 

Mayak PA, 

December 16, 

1965

This accident occurred in a residue-recovery area of a metal- and fissilesolution processing building. Multiple criticality excursions happened, 

associated with an uranyl nitrate solution in a dissolution vessel. Of the 

personnel in the area at the time of the accident, 17 received doses of 0.1 

rem or less, 7 between 0.1 and 0.2 rem, and 3 between 0.2 and 0.27 rem. 

The process equipment was not damaged, and no contamination 

occurred.

NA Criticality 

 4 According to the NTI website (http://www.nti.org/db/nisprofs/russia/fissmat/putomsk/tomsk7.htm#SKHK), the closed city o

Tomsk-7, is the location of the Siberian Chemical Combine (SKhK). The SKhK comprises several large facilities: The Rea

plutonium-production reactors); the Isotope Separation Plant; the Radiochemical Plant; the Conversion Plant; the Chemica

Scientific Research and Design Institute; fissile-material storage facilities, radioactive-waste-management facilities, and sev
A-9

Table A-5: Events at Reprocessing Facilities in the Russian Federation

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Mayak PA, 

December 10, 

1968

The accident occurred in a building where various chemical and 

metallurgical operations with plutonium were performed. A vessel with 

unfavorable geometry was being used temporarily in an improvised and 

unapproved operation for storing a plutonium organic solution. Two 

independent handling operations with this same vessel and same 

contents less than one hour apart resulted in two prompt critical 

excursions. A weak excursion occurred between the two energetic ones, 

when no personnel were present. One person received an estimated 

absorbed dose of about 700 rem; another received about 2,450 rem and 

died about one month afterwards. 

NA Criticality 

Siberian 

Chemical 

Combine, 

December 13, 

1978

There was a single criticality excursion associated with plutonium metal 

ingots in a storage container. An operator received an estimated total 

body dose of 250 rad, and more than 2,000 rad to his arms and hands. 

Seven other people received doses between 5 and 60 rad. The 

equipment was not damaged, and no contamination resulted.

NA Criticality 

Siberian 

Chemical 

Enterprises,5

April 6, 1993

The accident occurred during reprocessing of irradiated reactor fuel, and 

damaged both the reprocessing line and the building, resulting in the 

release of about 30 TBq of beta and gamma-emitting radionuclides and 

about 6 GBq of 239Pu. No personnel were overexposed. The accident 

happened due to a chemical (red oil) reaction. 

3 Fire / 

explosio

 5 The Siberian Chemical Enterprises (SCE) site is located in the Russian Federation. The countryside around SCE is relat

apart from the regional capital Tomsk and Tomsk-7 (now known as Seversk), which lie south of the SCE site.
A-10

Table A-6: Events at Reprocessing Facilities in United Kingdom

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Windscale, 

August 24, 

1970

The plant was being used to recover plutonium from miscellaneous scrap. 

A criticality accident happened with an organic solution of plutonium in a 

transfer vessel. A 150 mm diameter hole was cut through the concrete 

roof, and a line to the vessel was opened. Two people were in the plant 

at the time of the accident; one received an estimated dose of 2 rad, the 

other less than 1 rad. 

NA Criticality 

Windscale, 

1973

Radioactive material was released accidentally into an operating area in 

the plant after an exothermal chemical reaction in a reprocessing tank. 

4 Fire / 

explosio

Sellafield, 

during 

November 

1983

Highly radioactive waste liquor was discharged accidentally into the sea. 

The accident was due to a failure of communication between shifts, so 

that a tank that was assumed to contain liquid suitable for marine 

disposal, in fact contained highly radioactive material that was discharged 

into the sea creating an environmental hazard. This incident occurred 

during plant shutdown for routine annual maintenance.

NA Radiatio

THORP. The 

leak began 

before August 

28, 2004, and 

remained 

undiscovered 

until April 20, 

2005. Likely 

the leak was 

relatively 

small until 

January 2005. 

On 20 April 2005, British Nuclear Group Sellafield Limited (BNGSL) 

discovered a leak from a pipe that supplied highly radioactive liquor to an 

accountancy tank in a part of the THORP at Sellafield, known as the ‘feed 

clarification cell’. Approximately 83 000 liters of dissolver product liquor, 

containing about 22,000 kilograms of nuclear fuel (mostly uranium 

incorporating around 160 kilograms of plutonium), had leaked onto the 

floor of the cell. The leak came from a pipe that was severed completely 

at a point just above where it enters a tank. The leak remained 

undetected for about eight months. THORP was shut down after the 

incident; consent to begin reprocessing operations again was granted on 

January 9, 2007. 

3 Radiatio

 6 These slides have no date.
A-11

Table A-6: Events at Reprocessing Facilities in United Kingdom

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Dounreay, 

September 

2005

The accident occurred during decommissioning of the facility. Some 

radioactive reprocessing residues in the form of dissolved spent fuel, kept 

in underground tanks, are pumped to an area where they are mixed with 

cement, then stored in 500-litre drums. A robot carries out these 

operations, but on the day of the accident, 266 liters of radioactive 

material and 300 kgs of cement were spilled and solidified onto the floor of 

a treatment cell. There were no injuries to workers and no radiological 

release occurred from the cell itself.

NA Radiatio

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Hanford 

Works, 

December 4, 

1951

A fire occurred in the waste storage room of a small laboratory building. 

Apparently, nitric acid on rags used for cleaning and decontaminating 

parts that had been boxed for disposal spontaneously combusted. 

Firefighters used water to extinguish the fire. The fire burned for about 

four hours. Firefighting was hampered by the need for protection against 

inhaling plutonium. The fires in the building’s air filters were difficult to 

extinguish; they had to be removed to prevent further release of Pu-239 to 

the atmosphere. The fire damage was not severe, but the contamination 

spread throughout the building was quite severe, so that it had to be 

abandoned.

NA Fire / 

explosio

 7 McGuire [1988] states that “The Nuclear Fuel Services reprocessing plant was plagued by many small releases into grou

and air as well as unusually high occupational radiation exposures. However, the plant never had an accident of significan

emergency preparedness.” Since he does not provide information about these events, they are not included in this table. 
A-12

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Savannah 

River (facility 

code named 

TNX), January 

12, 1953

This was a test bed facility built in 1951 to perform experiments using 

mockups of the prototype of the equipment for the tributyl-phosphate 

(TBP) process. One part tested was a nitric-acid evaporator to remove 

excess nitric acid from the plutonium stream. The TNX evaporator 

exploded due to a “red oil” reaction. Investigation revealed that TBP and 

a kerosene-like diluent, which were immiscible in the aqueous acid, 

nonetheless had transferred into the evaporator and then chemically 

reacted with the materials in solution, causing the explosion. This was a 

matter of concern since it might occur in production units.

NA Fire / 

explosio

Hanford 

Works, 19538

During a process evolution, an acid solution and a caustic solution were 

mixed improperly. Metal waste supernatant was being pumped from a 

holding tank to a blend tank to prepare process feed for the TBP. Sixty 

percent nitric acid first was added to the blending tank, then the metal 

waste, and the agitator for the tank was started. A geyser of liquid quickly 

rose about 10 meters, and the prevailing wind carried the liquid onto 

several workers, contaminating them with about 4,000 counts per minute 

(cpm); they experienced stinging sensations from the liquid. Investigation 

revealed that the operator had activated the agitator switch; however, 

whilst it started to operate, it failed to continue. Areas near the blend tank 

contained a yellowish liquid that read 35 rem/hour at 6 inches. There 

were a significant number of reports of personnel contamination from 

maintenance and operations tasks. 

NA Fire / 

explosio

 8 Cadwallader et al. do not give the exact date of this event. 
A-13

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Hanford 

Works 

(REDOX 

plant), June 

18, 1956

An operator was in the control room of the plutonium-concentration facility 

when a continuous survey meter began to “break down’, that is, saturate 

counts at its given setting. The operator investigated and found the meter 

to be working correctly; the surfaces of the control room had become 

contaminated to greater than 7 million disintegrations per minute (dpm), 

off the scale of the instrument. Two spills of process solution (highly 

concentrated acidic plutonium product) were found on the floor behind the 

control panel. The solution had entered the control room by an instrument 

air-line. The skin of the operator who noted the survey meter’s 

“breakdown” was contaminated to over 40,000 dpm. Nasal smears read 

30,000 dpm, and all of his clothing was contaminated. A rough estimate 

was that only about 350 cm3 was spilled in the control room. 

NA Radiatio

Hanford 

Works (silver 

reactors), 

February 

1958

Silver reactors were used to react iodine from the Hanford stack effluents. 

A silver reactor had been flushed with ammonium hydroxide and then with 

water to remove a plug of silver compounds that had accumulated after 

repeated regenerations. The ammonia compound allowed a rapid 

exothermic reaction. 

NA Fire / 

explosio

Hanford 

Works (silver 

reactors), 

19589

Operators noted that after processing ammonia-laden off-gas, the silver 

reactors would experience exothermic reactions that would increase outlet 

temperatures. 

NA Fire / 

explosio

Oak Ridge Y12 Plant, June 

16, 1958

This accident occurred in a building in a process designed to recover 

enriched uranium from various solid wastes. Multiple criticality excursions 

occurred from an uranyl-nitrate solution in a water-collection drum. Eight 

people received significant radiation doses (461, 428, 413, 341, 298, 86.5, 

86.5, and 28.8 rem). At least one person owes his life to the prompt and 

orderly evacuation plans that were followed.

NA Criticality 

 9 Cadwallader et al. do not give the exact date of this event. 
A-14

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Los Alamos 

Scientific 

Laboratory, 

December 30, 

1958

The operations performed at the facility where the accident occurred were 

those chemical steps used to purify and concentrate plutonium from slag, 

crucible, and other lean residues remaining after recovery processes. A 

single criticality excursion was associated with a plutonium organic 

solution in an organic treatment tank. The accident resulted in the death, 

36 hours later, of an operator; the estimated dose to his upper torso was 

12,000 ± 50% rem. Two other persons received radiation doses of 134 

and 53 rem. No equipment was contaminated or damaged even though 

the shock from the accident displaced the tank about 10 mm at its 

supports.

NA Criticality 

Idaho 

Chemical 

Processing 

Plant, October 

16, 1959

This accident occurred in a chemical processing plant that accepted 

spent-fuel elements from various reactors, among other items. The fissile 

material involved in the accident (34 kg of enriched uranium, in the form of 

uranyl nitrate) was stored in a bank of cylindrical vessels with favorable 

geometry. The initiation of a siphoning action, inadvertently caused by an 

air-sparging operation, transferred about 200 liters of the solution to a 

15,400 liter tank containing about 600 liters of water. Multiple criticality 

excursions were associated with the uranyl-nitrate solution in a wastereceiving tank. Because of thick shielding, nobody received significant 

prompt gamma- or neutron-doses. During evacuation of the building, 

airborne fission products (within the building) resulted in combined beta 

and gamma doses of 50 rem (one person), 32 rem (one person), and 

smaller amounts to 17 people. 

NA Criticality 
A-15

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Oak Ridge 

National 

Laboratory 

(ORNL) 

(Radiochemic

al Processing 

Pilot Plant), 

Nov. 20, 1959

A nonnuclear explosion involving an evaporator occurred in a shielded 

cell. Plutonium was released from the processing cell, probably as an 

aerosol of fine particles of plutonium oxide. No radioactive material was 

released from the ventilation stacks; no contamination of grounds and 

facilities occurred outside of a relatively small area of ORNL immediately 

adjacent to the site of explosion. No one was injured by the explosion, 

and no one received more than 2% of a Iifetime body burden of plutonium 

or an overexposure to sources of ionizing radiation, either during the 

incident or the subsequent cleanup operations. The explosion is 

considered to have resulted from the rapid reaction of nitrated organic 

compounds formed by inadvertent nitration of about 14 liters of a 

proprietary decontaminating reagent.

NA Fire / 

explosio

Savannah 

River, 

September 

13, 1960

A valve corridor was highly contaminated by leaking coolant water that 

had been exposed to high activity waste from the PUREX process, and 

then had leaked through a defective waste-evaporator reboiler. 

Approximately 5,000 Curies was released with the water. Most of the 

radioactivity was contained in the building and was flushed to the wastehandling facilities. Minor amounts of liquid entered the seepage basins 

outside the building. A series of operating errors and miscommunications 

allowed the initial leakage water to flow back through open valves on 

steam traps and then out into the corridor. The measured radiation 

readings after the incident were as high as 400 rad/hour at 30 cm. 

Cleanup was costly and time consuming. 

NA Criticality 

Idaho 

Chemical 

Processing 

Plant, January 

25, 1961

This accident occurred in the main process building where fission 

products were separated chemically from dissolved spent fuel. The 

uranium then was concentrated via evaporation. Multiple criticality 

excursions occurred from an uranyl-nitrate solution in a vapordisengagement vessel. All employees evacuated promptly, and were 

exposed only to minimal doses (<60 mrem) caused by airborne fission 

products after leaving the building.

NA Criticality 
A-16

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Hanford 

Works, April 

7, 1962

This accident occurred in a plutonium-waste chemical recovery facility, 

and involved: 1) Cleaning up the floor of a solvent extraction hood, 2) a 

product-receiver tank that could overflow into the hood, 3) a temporary 

line running from the hood floor to a transfer tank, and, 4) the apparent 

improper operation of valves. There were multiple criticality excursions 

associated with a plutonium solution in the transfer vessel. Three people 

received significant doses of radiation (110, 43, and 19 rem). The 

accident itself caused no damage or contamination, but precipitated the 

final shutdown of the plant.

NA Criticality 

Hanford 

Works 

(PUREX 

plant), 

September 3, 

1963

The plant stacked 62 curies of I-131 over a three-day period, beginning on 

September 3. Several “short cooled” fuel elements were placed in the 

dissolver, releasing significant amounts of I-131. 

NA Radiatio

Hanford 

Works 

(PUREX 

plant), 

November 6, 

1963

Resin in a plutonium-extraction column caught fire. The fire spread 

plutonium contamination throughout the building and some outside of the 

building.

NA Fire / 

explosio
A-17

Table A-7: Events at Reprocessing Facilities in United States of America7

Facility and 

Date of 

Accident

Summary INES 

Level

Type o

Hazard 

Savannah 

River, 

February 12, 

1975

There was a ‘red oil’ explosion in the Thermal Denitrator at the 

reprocessing plant. The cause was attributed to the exothermic reaction 

between TBP and uranyl nitrate. 

NA Fire / 

explosio

Hanford 

Works, 197610

An ion-exchange column underwent a chemical reaction and exploded. A 

worker, who was in front of the glovebox that housed the column at that 

time was injured by glass from the glovebox’s shattered window, and 

contaminated with americium-241. The americium that had collected in 

the ion-exchange resin before the explosion was dispersed throughout the 

room during the explosion.

NA Fire / 

explosio

Idaho 

Chemical 

Processing 

Plant, October 

17, 1978

The accident occurred in a shielded operation of a fuel-reprocessing plant 

in which solutions from the dissolution of irradiated reactor fuel were 

processed by solvent extraction to remove fission products and recover 

the enriched uranium. The history is unknown of criticality excursions 

associated with an uranyl-nitrate solution in a lower disengagement 

section of a scrubbing column. There were no significant exposures to 

personnel and no damage to process equipment. As a direct result of this 

event, the plant underwent an extended, expensive shutdown.

NA Criticality 

UTP, Ontario, 

1980

An explosion occurred with gaseous releases.11 NA Fire / 

explosio

Hanford 

reprocessing 

plant, May 14, 

1997

Explosion (no additional information about this event was found). NA Fire / 

explosio

 10 Cadwallader et al. do not give the exact date of this event. 

11 Schneider et al. [2001] do not provide more information about this event, and no additional information was found. Since 

this accident, it is included in this table, even though it happened in Canada. 

B-1

APPENDIX B: Brief Description of Electrochemical Processing 

(Pyroprocessing) 

Section 6.2, “Pyroprocessing,” of NUREG-1909 [Croff et al. 2008] gives an overview of the 

technology of electrochemical processing, and points out that there are many manifestations of 

electrochemical processing in the nuclear industry, several of which are directed at spent fuel 

recycle. As applied to reprocessing of SNF, electrochemical processing involves the use of 

molten salts and metals in an electrochemical cell to separate the SNF constituents. 

Electrochemical processing involves anodization (oxidation) of a metal feed material into a 

molten salt electrolyte and then reduction at a cathode. 

The feed to electrochemical processing was originally intended to be metallic spent fuel, and the 

process lends itself best to reprocessing this type of fuel. As a consequence, the current 

Department of Energy (DOE) plans call for electrochemical processing to be used to reprocess 

metallic or possibly nitride SNF containing the transuranic (TRU) actinide elements after 

irradiation in a fast-spectrum transmutation reactor.1 However, oxide fuels such as those from 

light-water reactors (LWRs) can be electrochemically processed by first converting them to 

metal through a head-end step that reduces the oxide to metal. This reduction is best 

accomplished using finely divided oxide, which can be prepared using voloxidation2 to pulverize 

the oxide fuel. Process modifications are possible that separate uranium, plutonium, and other 

actinides from the remainder of the radionuclides. Figure A-1 [ANL 2002] presents the 

electrochemical processing flowsheet for oxide SNF under development by Argonne National 

Laboratory and other organizations such as Korean Atomic Energy Research Institute (KAERI). 

The following are the major steps in this flowsheet:

• Oxide SNF is chopped into segments and voloxidized (not shown).

• Most of the oxides in the SNF are reduced to the metal. 

 1 Long-lived radioactive isotopes, especially actinides such as plutonium and neptunium but also selected 

fission products such as 99Tc and 129I, are converted to shorter-lived fission products or stable isotopes by 

fission and/or neutron capture from neutrons generated in a transmutation reactor. In this reactor, the 

TRU elements would be fissioned to produce energy and what is primarily a fission product waste, thus 

removing by transmutation the principal long-term heat-producing actinides from the wastes.

2 The spent fuel is chopped (sheared) into segments using the voloxidation (volume oxidation) process. 

This process depends on the oxidation of the UO2 spent fuel matrix to lower density U3O8 to break down 

the fuel matrix and release trapped gases from it.
B-2

Figure B-1: Diagram of Electrochemical Processing Operations

• The metal from oxide reduction or metallic SNF, including the cladding in either case, 

becomes the anode in an electrorefiner (ER). The ER is essentially a crucible containing 

a molten electrolyte salt (a mixture of LiCl and KCl) atop a layer of liquid cadmium metal.

The anode and two cathodes operating at different voltages are inserted into the molten 

salt. After operating for about 12 hours, the ER contains the following:

– The anode contains elements that are stable as metals under the conditions in 

the ER (e.g., zirconium, technetium, iron, molybdenum).

– One cathode contains most of the uranium as metal.

– The other cathode (liquid cadmium) contains some of the uranium and rare earth 

fission products plus essentially all of the TRU elements as metal.

– The molten salt contains most of the fission products that are stable as chlorides 

under the conditions in the ER (e.g., cesium, strontium, barium).

• The metallic products associated with all three electrodes also contain entrained 

electrolyte salt and cadmium.

• The cathodes are separately inserted into a cathode processor in which the entrained 

electrolyte salt and cadmium are recovered for recycle by vacuum distillation.

• The uranium metal is converted to an appropriate form, either hexafluoride for reenrichment or oxide for direct reuse or disposal. The extent to which additional cleanup 

of the uranium might be necessary before conversion is not known.
B-3

• The TRU metal goes to an injection casting furnace (not shown) where it is refabricated 

into new fuel for a fast transmutation reactor.

• The metal left at the anode, including the cladding, is heated in a metal waste furnace to

produce a solid metallic waste form having zirconium as the major constituent for LWR 

fuels and iron as the major constituent for stainless-steel clad fuels.

• The fission-product-laden salt is circulated through a zeolite ion exchange bed which 

incorporates the fission products into the zeolite matrix. The loaded zeolite is 

consolidated into a monolithic form by combining it with borosilicate glass frit and 

sintering it, which converts the zeolite to the mineral sodalite in a waste form called 

glass-bonded zeolite [NAS, 2000] [Kim, 2006]. Processes to improve the removal of

fission products from the salt and recycle the salt are under development [Simpson, 

2007].

A report by the Committee on Electrometallurgical Techniques for DOE Spent Fuel Treatment

[NAS 2000] and a paper by Laidler et al. [1997] describe electrochemical processing technology 

for SNF in more detail.

Several reports by the International Atomic Energy Agency (IAEA) address electrochemical 

processing; in particular, the report “Guidance for the Application of an Assessment 

Methodology for Innovative Nuclear Energy Systems, International Project on Innovative 

Nuclear Reactors and Fuel Cycles manual — Safety of Nuclear Fuel Cycle Facilities,” [IAEA 

2007] describes facilities of the nuclear fuel cycle, including electrochemical processing. An 

IAEA report [IAEA 2008] describes the technologies used for spent-fuel reprocessing, including 

electrochemical processing, and their implementation in several countries.

A report by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation 

and Development [NEA 2004] gives a detailed description of the implementation of 

electrochemical processing in countries worldwide, including the USA. 

Willit et al. [1992] present a comprehensive review of the literature on uranium and plutonium 

electrorefining in molten salts. It covers work published from 1943 to November 1991. 

Electrodeposition and electrodissolution at solid and liquid metal electrodes are discussed as 

well as mass transfer in liquid metal and molten salt phases. The journal “Nuclear Technology” 

dedicated the issue of May 2008 (Vol. 162) to electrochemical processing. Recently, a paper by 

Goff et al. [2011] also briefly describes electrochemical processing.





NUREG/CR-7168 Regulatory Approaches for Addressing Reprocessing Facility Risks: 

An Assessment