Review of Spent Fuel 

Reprocessing and 

Associated Accident 

Phenomena

Office of Nuclear Regulatory Research

NUREG/CR-7232
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Review of Spent Fuel 

Reprocessing and 

Associated Accident 

Phenomena

Manuscript Completed: October 2016 

Date Published: February 2017

Prepared by: N. E. Bixler, F. Gelbard, D. L. Y. Louis and J. Phillips*

*Sandia National Laboratories

 Albuquerque, New Mexico 87185

 Operated for the U.S. Department of Energy

M. Fuhrmann, NRC Project Manager

NRC Job Code Number V6050

Office of Nuclear Regulatory Research

NUREG/CR-7232
_____________________________

NUREG/CR-7232 has been 

reproduced from the best 

available copy.

_____________________________
iii

ABSTRACT

The work presented in this report supports the advancement of the safety assessment 

capabilities of the NRC staff in terms of evaluating accidents that could release radioactivity into 

the environment and estimating the associated risk during the reprocessing of nuclear fuel. 

This report lays out much of the background and the concepts needed to construct a 

computational tool for reprocessing facility source terms. It reviews the following topics: (1) the 

status of past and current reprocessing plants throughout the world; (2) an overview of a typical 

plant design, its functions, and its layout; (3) historical accidents and the phenomena that are 

relevant to these accidents; and (4) the models that are needed to describe these accidents. 

This document does not propose or describe a specific computational tool that can be used to 

estimate source terms for a reprocessing facility. Rather, it supports the NRC’s capability to 

evaluate potential accident source terms that would be used in a license application for a 

reprocessing facility.

v 

TABLE OF CONTENTS

ABSTRACT...................................................................................................................................iii

TABLE OF CONTENTS................................................................................................................ v

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

LIST OF TABLES......................................................................................................................... ix

EXECUTIVE SUMMARY ............................................................................................................. xi

ACKNOWLEDGMENTS.. ...........................................................................................................xiii

ABBREVIATIONS AND ACRONYMS......................................................................................... xv

1 INTRODUCTION............................................................................................................ 1-1

2 STATUS OF FUEL REPROCESSING PLANTS............................................................ 2-1

3 OPERATIONS AND SYSTEMS..................................................................................... 3-1

3.1 Operations.......................................................................................................... 3-1

3.1.1 Shipping and Receiving Operations .......................................................3-2

3.1.2 Storage ...................................................................................................3-3

3.1.3 Crushing, Shearing, and Chopping.........................................................3-3

3.1.4 Extraction and Separation ......................................................................3-3

3.1.5 Waste Management................................................................................ 3-4

3.2 Systems............................................................................................................ 3-10

3.2.1 Confinement ......................................................................................... 3-13

3.2.2 Ventilation............................................................................................. 3-14

3.2.3 Fire ....................................................................................................... 3-16

3.2.4 Miscellaneous Support Systems........................................................... 3-17

4 ACCIDENTS/INCIDENTS – HISTORY AND DISCUSSION .......................................... 4-1

4.1 Chemical Explosions ..........................................................................................4-8

4.1.1 History.....................................................................................................4-8

4.1.2 General Discussion............................................................................... 4-12

4.2 Inadvertent Criticalities ..................................................................................... 4-14

4.2.1 History................................................................................................... 4-14

4.2.2 General Discussion............................................................................... 4-20

4.3 Fire ................................................................................................................... 4-22

4.3.1 History................................................................................................... 4-22

4.3.2 General Discussion............................................................................... 4-22

4.4 Spills and Leaks ............................................................................................... 4-23

4.4.1 History................................................................................................... 4-23

4.4.2 General Discussion............................................................................... 4-26
vi

5 ACCIDENT AND AEROSOL GENERATION MODELING............................................. 5-1

5.1 Inventory in Operations ...................................................................................... 5-1

5.2 Source Term Potentials and Release Mechanisms............................................ 5-7

5.2.1 Explosions .............................................................................................. 5-8

5.2.2 Release Due to Nuclear Criticality........................................................ 5-21

5.2.3 Fires...................................................................................................... 5-21

5.2.4 Spills ..................................................................................................... 5-23

5.3 Basic Principles Approach to Determining Releases ....................................... 5-31

6 REFERENCES ...............................................................................................................6-1

Appendix A: BARNWELL NUCLEAR FUEL PLANT DESCRIPTION .................................... A-1

Appendix B: SELECTED PAST CHEMICAL EXPLOSION ACCIDENT CONDITIONS 

AND POSSIBLE EXPLOSION CALCULATION INPUT MODELS .................... B-1

Appendix C: MISCELLANEOUS DATA ................................................................................. C-1
vii

LIST OF FIGURES

Figure 3-1 Typical operations at spent fuel reprocessing facilities that uses a 

co-extraction process with countercurrent concept ..............................................3-1

Figure 3-2 Joule-heated melter concept [DOE 1999]............................................................. 3-7

Figure 3-3 Induction melting used at the French AVM process [Marsden 2010] ................... 3-7

Figure 3-4 A typical component list of a PACT melter [DOE 1999]........................................ 3-8

Figure 3-5 PHP prototype design [DOE 1999] ....................................................................... 3-8

Figure 3-6 Feed preparation and melter schematic for the arc melter furnace 

design [Jain 1998b]............................................................................................... 3-9

Figure 3-7 Three-tiered confinement zones used in radiological plant ................................ 3-14

Figure 4-1 Chronology of process criticality accidents. Plot shows number and 

location of accidents by year from 1945-2000. [McLaughlin 2000]..................... 4-15

Figure 5-1 Important fission products, which contributed to decay heat for typical 

light water reactor fuel for cooling times from 1 to 1,000 years 

[OECD 2011] ........................................................................................................ 5-3

Figure 5-2 Operation inventory for an UREX type extraction process (with no Pu 

separated out) of a 5-year cooled and 45 GWd/t burnup [NEA 2012-15]............. 5-5

Figure 5-3 Operation inventory for a UREX Type extraction process (without Pu 

separated) of 30-year cooled fuel with 45 GWd/t burnup [NEA 2012-15]............. 5-6

Figure 5-4 Median particle diameter as a function of mass ratio (rm) 

[Steindler 1980]................................................................................................... 5-11

Figure 5-5 Initial AR of airborne versus mass ratio [Steindler 1980] .................................... 5-12

Figure 5-6 Explosive-induced liquid aerosol experiment [Ishmatov 2010] ........................... 5-14

Figure 5-7 Aerosol size distribution for (a) final aerosol distribution in 

measurement (b) initial aerosol distribution calculated [Ishatov 2010] ............... 5-15

Figure 5-8 Pressurized radioactive aerosol tank configuration for measuring the 

pressurized release of solution from PARE [NUREG/CR-3093]......................... 5-27

Figure 5-9 PARE modification for pressurized liquid release experiment 

[NUREG/CR-3093] ............................................................................................. 5-28

Figure 5-10 Measured and predicted temperatures for flashing-spray releases 

[NUREG/CR-4779] ............................................................................................. 5-29

ix

LIST OF TABLES

Table 2-1 World fuel reprocessing plants [GFMR 2010]....................................................... 2-1

Table 2-2 U.S. enrichment and mixed oxide plants [GFMR 2010]........................................ 2-6

Table 2-3 Status of high level waste vitrification facilities [Jain 1998, DOE 1999]................ 2-7

Table 3-1 Solidification technology [DOE 1999]. .................................................................. 3-6

Table 3-2 Selected nuclear regulatory commission guidance documents for fuel 

reprocessing. ...................................................................................................... 3-11

Table 3-3 Confinement needs for handling radioactive materials 

[Cadwallader 2005]............................................................................................. 3-13

Table 3-4 Zoning of facilities based on radioactivity of materials handled 

[DOE 2003]. ........................................................................................................ 3-15

Table 3-5 Zoning of facilities based on chemical contamination levels 

[DOE 2003]. ........................................................................................................ 3-15

Table 4-1 Potential incidents at a spent fuel reprocessing facility [DP-1558]. ...................... 4-1

Table 4-2 Summary of possible major accidents at a spent nuclear fuel 

reprocessing facility. ............................................................................................. 4-5

Table 4-3 Potential accident initiators from external events and natural 

phenomena [IAEA DS360].................................................................................... 4-9

Table 4-4 Explosion accidents in fuel reprocessing plants. ................................................ 4-10

Table 4-5 Blast overpressure on structural damage [Merrifield 1990]. ............................... 4-14

Table 4-6 A summary of solution criticality incidents [McLaughlin 2000]............................ 4-16

Table 4-7 Single-parameter limits for uniform aqueous solutions of fissile 

nuclides (U-235 and Pu-239) [ANS-8.1]. ............................................................ 4-20

Table 4-8 Activity of important gaseous and volatile radionuclides released 

during a 1019 fission excursion in spent fuel solution*......................................... 4-21

Table 4-9 A summary of fire incidents at the fuel reprocessing facilities*........................... 4-22

Table 4-10 Spill accidents in fuel reprocessing plants. ......................................................... 4-24

Table 5-1 Typical low-enriched light water reactor spent fuel inventory (per 

metric ton of fuel) for two cooling periods and burnup ranges 

[NEA 2012-15]. ..................................................................................................... 5-2

Table 5-2 Process vessel characteristics and its material properties 

[NUREG/CR-4593]. ............................................................................................ 5-10

Table 5-3 TNT equivalence for various gas, combustible gas and combustible 

liquids released during vessel rupture events [NUREG/CR-4593]. .................... 5-10

Table 5-4 Time of complete evaporation of water from aerosol drops of a 

20% salt solution [Ishmatov 2010]. ..................................................................... 5-14

Table 5-5 Scaled overpressure and impulse correction factors for vessels situated 

near ground [Ferradas 2006]. ............................................................................. 5-18

Table 5-6 Fitted Ps and i equations* for spherical vessels [Ferradas 2006]. ..................... 5-18

Table 5-7 Summary of the airborne release in combustible liquid 

[NUREG/CR-4736]. ............................................................................................ 5-22

Table 5-8 Source term for burning of combustible liquid [NUREG-1320]. .......................... 5-23

Table 5-9 Comparison of properties of water and experimental liquids*. ........................... 5-24

Table 5-10 Experimental results on liquid spills with correction 

[NUREG/CR-4997]*. ........................................................................................... 5-25

Table 5-11 Estimates of mole fraction of the pressurized gases dissolved in 

uranine solution* [NUREG/CR-4779].................................................................. 5-28
x 

Table 5-12 Airborne release for pressurized release of uranine solutions 

[NUREG/CR-4779]. ............................................................................................ 5-28

Table 5-13 Airborne release for flashing-spray releases of uranine solutions 

[NUREG/CR-4779]. ............................................................................................ 5-29

Table 5-14 Airborne release for the pressurized unh solutions in air 

[NUREG/CR-3093]. ............................................................................................ 5-30

Table 5-15 Comparison of the measure and correction data of aerosol size 

distribution for liquid pressurized release [NUREG-1320]. ................................. 5-31
xi

EXECUTIVE SUMMARY

In response to some commercial interests in future reprocessing plant construction and 

operation, the U.S. Nuclear Regulatory Commission (NRC) is investing in developing expertise 

to support advanced nuclear fuel reprocessing. The work presented in this report supports the 

advancement of the safety assessment capabilities of the NRC staff in terms of evaluating

accidents that could release radioactivity into the environment and estimating the associated 

risk. 

The history of fuel reprocessing operations in the United States has been limited to the nuclear 

weapon complex for material production. Licensing of a civilian aqueous reprocessing facility 

was last sought for the Barnwell Nuclear Fuel Plant about the year 1970. Along with historical 

reprocessing efforts in the United States, foreign fuel facilities, supporting both weapon 

development and civilian spent fuel processing, have contributed to the development of various 

chemical techniques for reprocessing fuel. The co-extraction process is considered to be a 

candidate for any future domestic civilian aqueous reprocessing without plutonium separation. 

However, no request for a license has been submitted to date nor has a specific plant design 

been proposed to be built. Therefore, much of this document focuses on historical plant 

designs (e.g., Barnwell) to provide reasonable expectations for a future aqueous reprocessing 

plant. 

This report lays out much of the background and the concepts needed to construct a 

computational tool for aqueous reprocessing facility source terms. It reviews the following 

topics: (1) the status of past and current reprocessing plants throughout the world; (2) the 

design, functions, and layout of a typical plant; (3) historical accidents and the phenomena that 

are relevant to these accidents; and (4) the models that are needed to describe these accidents. 

This document does not propose or describe a specific computational tool that can be used to 

estimate source terms for a reprocessing facility. 

The various accidents considered are fires, spills, inadvertent criticality events, and explosions. 

Modeling the relevant phenomena for each accident type is a prerequisite for creating a sourceterm tool to provide the release rate and aerosol characteristics of a release to the environment. 

A purpose of this document is to evaluate and describe the relevant phenomena that could 

contribute to a source term. Later work will focus on developing the source-term tool. 

xiii

ACKNOWLEDGMENTS

Guidance, information, and comments were provided by the original Nuclear Regulatory 

Commission Contracting Officer’s Representative (NRC COR), Phil Reed. Near the end of this 

project, Mark Fuhrmann took the role of COR and provided the final guidance needed to 

complete this project. Funding for this project was provided through the NRC. 

Scott Weber, Wendy Reed, Christina Leggett, and Yawar Faraz provided helpful comments on 

this document. Finally, Alex Brown and Chengcheng Feng provided significant guidance and 

support for carrying out the smoothed particle hydrodynamics modeling described in Chapter 5.

xv

ABBREVIATIONS AND ACRONYMS

AHA Acetohydroxamic Acid

ALI Annual Limits on Intake

AMMD Aerodynamic Mass Mean Diameter

APM Atelier Pilote Marcoule

ARF Airborne Release Fraction

ASME American Society of Mechanical Engineers

AVM Marcoule Vitrification Facility

BNFP Barnwell Nuclear Fuel Plant

BRF Base Recycling Facility

BWR Boiling Water Reactor

CCD-PEG Chlorinated-Cobalt Dicarbollide-Polyethylene Glycol

CEMG Crane and Equipment Maintenance Gallery

CERS Contact Equipment Removal Station

CFAST Consolidated model of Fire and Smoke Transport

CFD Computational Fluid Dynamics

CFM Cubic Feet per Minute

CRA Control Room Area

CLS Cask Loading Station

DAC Derived Airborne Concentration

DBE Design Basis Earthquake

DOE Department of Energy

DOG Dissolver Off-Gas

DR Damage Ratio

DTPA Diethylenetriaminepentaacetic acid

EMS Equipment Maintenance Station

FDS Fire Dynamics Simulator

FP Fission Product

FPIG Filter Piping and Instrument Gallery

FRSS Fuel Receiving and Storage Station

GNEP Global Nuclear Energy Partnership

GP General Purpose

GSD Geometric Standard Deviation

HA High Activity

HAN Hydroxylamine Nitrate

HAS Hydroxylamine Sulfate

HEME High Efficiency Mist Eliminators

HF Hydrofluoric acid

HILC High Intermediate Level Cell

HLC High Level Cell

HLLW High-Level Liquid Waste

ICPP Idaho Chemical Processing Plant

ILC Intermediate Level Cell

ILLW Intermediate-Level Liquid Waste

INRA International Nuclear Recycling Alliance

LPF Leak Path Factor

LWR Light Water Reactor

MAR Material At Risk

MOX Mixed oxide fuel
xvi

NPH Normal Paraffin Hydrocarbon

NRC Nuclear Regulatory Commission

NUEX Modified PUREX process

OBE Operating Basis Earthquake

OML Oxalate Mother Liquor

PACT Plasma Arc Centrifuge Treatment

PARE Pressurized Airborne Release Equipment

PHP Plasma Hearth Process 

PPC Plutonium Product Cell

PNC Plutonium Nitrate Cell

PNSA Plutonium Nitrate Storage Area

PNSL Plutonium Nitrate Storage and Load-out

PVB Pressure Vessel Breach

PWR Pressurized Water Reactor

RF Respirable Fraction

RMSC Remote Maintenance and Scrap Cell

RPC Remote Process Cell

SAC Sample and Analytical Cell

SNL Sandia National Laboratories

ST Source Term

TALSPEAK Trivalent Actinide-Lanthanide Separation by Phosphorus reagent Extraction from

Aqueous Komplexes

TBP Tributyl Phosphate

THM Ton of Heavy Metal

THORP Thermal Oxide Reprocessing Plant

TRU Transuranic

TRUEX Transuranic Extraction

TVF Tokai Vitrification Facility

UDS Undissolved Solids

UN Uranyl Nitrate

UNH Uranyl Nitrate Hexahydrate

UPC Uranium Product Cell

USBM United States Bureau of Mines

VOG Vessel Off-Gas

WIP Waste Immobilization Plant

WTC Waste Tank Cell

WTEG Waste Tank Equipment Gallery

WVDP West Valley Demonstration Project

WVP Waste Vitrification Plant
1-1 

1. INTRODUCTION

In recent years, there has been some domestic interest in reprocessing commercial spent 

nuclear fuel. The Department of Energy (DOE) is investing resources to study advanced 

reprocessing technologies and methods. Before a domestic commercial reprocessing facility 

can be built and operated, it must be licensed by the U.S. Nuclear Regulatory Commission 

(NRC). 

The purpose of this report is to provide the background research and analyses to develop a 

computational tool to estimate accidental releases of hazardous substances to the environment, 

or source terms, for fuel reprocessing facilities. Best estimate analyses incorporate state-of-theart methods, models, and computational codes intended to adequately model and predict the 

progression of an accident from the onset of the initiating event up to and including 

characterizing the release of any hazardous material to the environment. 

The risk of an accident is defined as the product of the probability and the consequence of the 

accident. The development of a source term tool should provide a methodology to estimate the 

consequences of an accident. The probability of an accident is determined separately from the 

source term and consequence analysis, and may be determined through hazard analysis 

methods such as failure modes and effects analysis or event tree analysis. Once a 

determination of the probability of a given event is made, the source term tool estimates the 

timing and magnitude of potential release to the environment. The resulting source term 

coupled with a consequence analysis tool, like MACCS, can be used to estimate the distribution 

of radioactive materials to the surrounding environment and to assess the consequences of the 

accident. 

The various accidents considered are fires, leaks and spills, inadvertent criticality events, and 

explosions; each of which falls under the purview of the source term tool. Modeling the relevant 

phenomena for each accident type is a prerequisite to creating a source term tool to provide the 

release rate and aerosol characteristics of a release to the environment. The purpose of this 

document is to evaluate and describe the relevant phenomena that could contribute to a source 

term. Later work will focus on developing the source term tool. 

The history of fuel reprocessing operations in the United States has been limited mostly to the 

nuclear weapon complex for material production. Licensing of a civilian facility was last sought 

for the Barnwell Nuclear Fuel Plant near the Savannah River Site in South Carolina around the 

year 1970, while the West Valley reprocessing facility, located near Buffalo New York, operated 

from 1968 to 1972. Along with historical efforts in the United States, foreign fuel reprocessing 

facilities, supporting both weapons development and civilian fuel reprocessing, have contributed 

to the development of various chemical techniques for reprocessing fuel. A brief status of the 

fuel reprocessing facilities operated to date is presented in Section 2. The co-extraction process 

is considered to be the most likely candidate for future domestic civilian reprocessing without 

plutonium separation. 

A review of the operations conducted and systems employed at reprocessing facilities is 

presented in Section 3. The reprocessing stages presented include receiving, storage, 

separation, and waste management. Brief discussions on the various engineered systems such 

as ventilation, confinement, and fire sprinkler systems are provided as well. 
1-2 

From the historical operations of fuel reprocessing plants, descriptions of accidents, with some 

information on offsite consequences, are available. Section 4 presents a history of the 

accidents categorized by accident type. A general discussion of the relevant phenomena is also 

presented. A list of these accidents is presented to build familiarization with these events as 

well as their potential consequences. Of the reported accidents, nitration-oxidation-reaction 

events comprise the largest category. 

Given the occurrence of an accident, an estimate of transport of released radioactive material 

and ultimate release into the environment is the purpose of a source term tool. Each type of 

accident and related transport of nuclear materials are controlled by a set of phenomena. The 

phenomena pertinent to accident progression and transport for the set of accident types are 

discussed in Section 5. 
2-1 

2. STATUS OF FUEL REPROCESSING PLANTS

Most of the early fuel reprocessing plants in the world were built for military applications. In the 

U.S., defense fuel reprocessing was performed at several Department of Energy sites, such as 

Hanford, Idaho, Oak Ridge, and Savannah River. Commercial fuel reprocessing briefly existed 

in the U.S. until President Carter stopped civilian fuel reprocessing to reduce the possibility of 

nuclear proliferation. 

The following tables give a brief description of defense and civilian fuel reprocessing facilities

around the world. Table 2-1 and Table 2-2 summarize the past, current, and planned fuel 

reprocessing facilities in the world and the current and planned U.S. enrichment and mixed 

oxide production plants, respectively. Since many modern spent fuel reprocessing facilities 

include actual fuel production operations and solidification of liquid wastes (vitrification), a list of 

these facilities is also included in this section. A list of vitrification facilities is presented in

Table 2-3. 

Table 2-1 World fuel reprocessing plants [GFMR 2010]

Facility Type** Operational 

Status

Capacity

(THM/year*) Notes

Belgium

Dessel Civilian Decommissioned This reprocessing plant (called “Eurochemic”) was 

built as a joint facility with 12 other OECD states, 

and was operated between 1957 and 1974. It has 

been dismantled [IAEA-T1529].

China

Pilot Plant Civilian Starting up 50-100 A long-delayed pilot reprocessing plant at 

Yumenzhen site in Gansu Province. Another 

source rated this facility at 0.1 THM/year [NUREG1909].

Lanzhou [IAEAT1467]Civilian 800 A future plan for the high capacity fuel 

reprocessing plant that is planned to start in 2020.

France

APM [OCED 

2005]

Shut-down in 

1997

5 APM (Atelier Pilote Marcoule) is a pilot 

reprocessing plant built at Marcoule. This plant 

was commissioned in November 1962 to validate 

the processes at pilot scale. The actual fuel 

reprocessing, which began in 1987, was done for 

several fuel types, including natural uranium-gasgraphite fuel, mixed fuel and fast breeder oxide 

fuels.

UP1 Dual Shut-down in 

1992

600 [NUREG1909]UP1 is located at Marcoule. Large-scale 

separation of plutonium was carried out for both 

military and civilian purposes between 1958 and 

1997. Reprocessing for military purposes was 

ceased in 1992.

UP2A,1 Operating 1000 The UP2 facility reprocessed gas-graphite reactor 

fuel from 1966 to 1987.The UP2 plant started 

reprocessing light water reactor oxide fuel in 1976. 

In 1994, the UP2 facility was expanded, and 

renamed UP2-800. One source rated this facility 

at 850 THM/year [NUREG-1909].
2-2 

Table 2-1 World fuel reprocessing plants [GFMR 2010]

Facility Type** Operational 

Status

Capacity

(THM/year*) Notes

UP3A,1 Operating 1000 UP3 facility is a first major extension at the La 

Hague site, and started operating in 1989. This 

facility has been rated to produce 850 THM/year 

[NUREG-1909].

Germany

Karlsruhe A

[NUREG-1909]

Civilian Decommissioned 25 This facility was operated between 1970 and 

1991. [IAEA-T1529]. 

India

Trombay NoncivilianOperating 50 This facility started operating in 1964 to produce 

plutonium using the PUREX process. The plant 

was augmented and restarted operations in 1983

[Dey 2005].

Tarapur Dual Operating 100 The original facility was commissioned in 1975 

and was intended to reprocess Zircaloy clad oxide 

spent fuel. This plant uses a chop-leach 

technique for the head-end and uranous nitrate 

stabilized by hydrazine as the reductant for 

partitioning [Dey 2005]. New plant commissioned 

in 2011.

Kalpakkam Dual Operating 100 This facility is intended to reprocess Zircaloy clad 

natural uranium oxide spent fuel [Dey 2005].

Israel

Dimona NoncivilianOperating 40-100 The fuel reprocessing capability includes a 

plutonium separation plant (Machon 2).

Italy

Saluggia Civilian Decommissioned 0.1 [NUREG1909]This pilot plant was named “EUREX”, and 

operated from 1970 to 1983 [ODEC Italy].

Trisaia Civilian Decommissioned This pilot plant was operated during 1970s to 

process the uranium-thorium cycle fuels from the 

U.S. Elk River reactor. It also included a 

solidification of liquid reprocessed waste 

(cementation) [OECD Italy].

Japan

Rokkasho Civilian Delayed 800 This facility, which is operated by Japan Nuclear 

Fuel Limited [Ueda 2005], has design capacity of 8 

tons of plutonium or 800 tons of uranium annually. 

Delayed until at least 2018

A MOX fuel fabrication plant was originally

planned to be completed by 2015. This facility 

remains non-operational [GFMR 2010]. 

Tokai Civilian Temporarily shut 

down

200 This facility, which is operated by Japan Nuclear 

Cycle Development Institute, started in 1977 

[Ueda 2005].

North Korea

Yongbyon NoncivilianOn standby 100-150 

Pakistan

Nilore NoncivilianOperating 20-40

Chashma NoncivilianUnder 

construction

50-100

Russia
2-3 

Table 2-1 World fuel reprocessing plants [GFMR 2010]

Facility Type** Operational 

Status

Capacity

(THM/year*) Notes

RT-1 Dual Operating 200-400 RT-1 is located at the Mayak site. It reprocesses

naval and research-reactor fuel. This plant also 

separates reactor-grade plutonium and uranium.

Seversk Dual To be shut down 

after cleanup

6000 It contains an isotope separation plant, which 

became operational in 1953.

Zheleznogorsk Dual To be shut down 

after cleanup

3500

United Kingdom

AEA 

Reprocessing 

plants [NUREG1909]C

NoncivilianDecommissioned 0.02 There were two plants, one for material test 

reactors, and one for mixed oxide reactors.

B204C NoncivilianShut down B204 was the first reprocessing plant to extract 

plutonium for military applications in the UK.

B207 NoncivilianDecommissioned This was a uranium purification plant [NUREG1909].

B205 Civilian Operating 1500 The B205 Magnox reprocessing plant processes 

foreign light-water reactor fuel. The original 

schedule was to shut down this facility in 2012. 

However due to backlog, shutdown is now

planned for 2017.

THORPA Civilian Temporarily shut 

down

1200 Thermal Oxide Reprocessing Plant (THORP) at 

Sellafield reprocesses oxide fuel from its AGR 

reactors and foreign LWR reactors. Due to many 

incidents at the plant, THORP throughput varies 

greatly. Once the current obligations are met, 

regulators plan to shutdown THORP in 2020 [MSF 

2011]. According to [NUREG-1909], there was a 

miniature pilot plan that was shutdown at THORP.

United States

Oak RidgeA

[Manhattan 

1999]

NoncivilianDecommissioned X-10 was constructed in 1940s. It contained a 

pilot plant for a chemical separation process to 

recover plutonium from the experimental pile 

experiment within X-10.

Hanford 

A,B[Bastin 2008], 

[Manhattan 

1999]

NoncivilianDecommissioned Total: 96,900 

T and B 

plants: 8100 

REDOX 

plant: 22,400 

PUREX plant: 

66,400 

[Gephart 

2003]

Hanford Engineering Works became the 

production location for producing weapons grade 

materials during and following World War II. 

There were three plants – “T”, “U” and “B” plants. 

Because of the large size (60 feet high by 700 feet 

to 1,100 feet with thick-walled heavily reinforced 

concrete structures) of the processing equipment, 

these plants were sometimes referred as “Canyon” 

[Bastin 2008]. Remotely operated and 

maintained equipment was installed at the bottom 

of the structure to carry out the chemical 

processing. The “T” plant separated plutonium out 

of irradiated fuel from Hanford production reactors 

[Manhattan 1999]. The “U” plant was to recover 

uranium not recovered earlier using a solvent 

extraction process, and the “B” plant was to 

recover isotopes from nuclear waste [Bastin 2008]
2-4 

Table 2-1 World fuel reprocessing plants [GFMR 2010]

Facility Type** Operational 

Status

Capacity

(THM/year*) Notes

Idaho [Pace 

2006], [Bastin 

2008]

Dual Shutdown Over 10 THM 

during the 

operating life 

of the plant 

[Pace 2006]

Idaho Chemical Processing Plant (ICPP) was 

configured like the Oak Ridge reprocessing plant. 

ICPP was built to reprocess all highly enriched 

uranium irradiated in U.S. production reactors in 

the 1950s [Bastin 2008]. ICPP was intended to 

recycle uranium from the plutonium production 

reactors [Pace 2006]. In 1957, ICPP was an 

important player for processing non-defense 

nuclear fuel. All highly enriched fuels including 

those from the Shipping-port Core were processed 

at ICPP. Thus ICPP has been reprocessing fuels 

with cladding in aluminum, zirconium, stainless 

steel, graphite, and many other unique materials 

[Pace 2006]. It processed the fuel for space 

applications and from research reactors as well 

[Pace 2006]. ICPP was ordered to shut down in 

1992.

Savannah River A [Savannah

50_CTheseh11], 

[Bastin 2008], 

[GFMR 2010]

Dual H-Canyon is still 

operating, and the 

F-Canyon is 

being 

decommissioning

15

[GFMR, 

2010]

Savannah River Site in 1950s contained two 

primary fuel reprocessing plants for defense use – 

200-F [uses PUREX process] and 200-H (today, 

they are often referred to as F-Canyon and HCanyon), which were based on wet processes 

used to dissolve fuel assemblies and on chemical 

separation to extract both uranium and plutonium 

[Savannah50_Ch11]. H-Canyon’s current mission 

is to blend down the highly enriched uranium to 

low enriched uranium for power reactor fuel 

applications [GFMR 2010]. Some impure 

plutonium may be reprocessed in this facility. The 

extracted plutonium would either be used as feed 

for the mixed-oxide plant being built at Savannah 

River Site or vitrified to be disposed as waste.

Morris [West 

Valley 

1996],[NUREG1909]

Civilian Never operated 300 This was a General Electric Midwest Fuel 

Recovery Plant at Morris, Illinois. It was declared 

inoperable in 1974.

Barnwell [West 

Valley 

1996],[NUREG1909]

Civilian Never operated 1500 This facility was constructed in 1970 by Allied 

General Nuclear Services (AGNS) at Barnwell, 

South Carolina. This plant was planned to operate 

in 1974. The plant never operated due to delay in 

construction and licensing, and the 1977 decision 

by President Carter to defer indefinitely all 

reprocessing of commercial nuclear fuel.

West Valley 

[West Valley 

1996]

Civilian Decommissioned 300 This facility is the first commercial spent fuel 

reprocessing plant in the US and was operated by 

Nuclear Fuel Services (NFS) between 1966 and 

1972. This facility used the PUREX process. 

Other operations included in this facility were 

chopping the assembly rods, dissolving the spent 

fuel in acid, separating and storing radioactive 

wastes.
2-5 

Table 2-1 World fuel reprocessing plants [GFMR 2010]

Facility Type** Operational 

Status

Capacity

(THM/year*) Notes

Energy Solutions 

[Energy 

Solutions 2008]

Civilian Proposed A Base Recycling Facility (BRF) was proposed for 

GNEP. BRF is drawn heavily on the THORP 

Facility at Sellafield in the UK. The facility 

includes a head-end process facility, where the 

spent fuel is removed from its transport casks 

under water, and is placed into temporary storage 

containers. This facility uses the UREX process 

method to extract fissile materials.

International 

Nuclear 

Recycling 

Alliance (INRA) 

[INRA 2008]

Civilian Proposed 800 INRA proposed a fuel reprocessing plant that is 

adapted from the technology at the La Hague and 

Rokkasho reprocessing plants. The plant contains 

a number of operations: used fuel receipt and 

storage; shearing, dissolution, and compaction; 

separations, purification, and concentrations; fuel 

fabrication; fission products vitrification; high active 

solid waste interim storage; and waste processing, 

such as vitrification.

*THM stands for ton of heavy metal.

**Dual means both civilian and non-civilian applications. APUREX Operation BREDOX/PUREX Operations

CBUTEX Operation

1The La Hague Reprocessing Site includes two current operated reprocessing plants – UP2-800 and UP3.
2-6 

Table 2-2 U.S. enrichment and mixed oxide plants [GFMR 2010]

Facility Type Operational 

Status

Capacity *

(THM/year) Comments

Oak Ridge 

[Manhattan 

1999]

NoncivilianK-25 was 

decommissioned. 

Y-12 is 

operational.

In 1942, a reprocessing pilot plant to be built at Oak 

Ridge, Tennessee was initiated. K-25 at Oak 

Ridge was the nation’s first diffusion plant for 

enriched uranium for weapons applications in 1943. 

This plant used the gaseous diffusion method to 

enrich uranium. Additional enrichment facilities 

(such as K-27) were also built near K-25 and the 

entire complex was called the “Oak Ridge Gaseous 

Diffusion Plant”. The K-25 building was demolished 

in 2013.

Y-12 was built in the1940s and operated in early 

1943 for enriching uranium. Y-12 produces 

uranium with higher enrichment than had been 

produced at K-25.

Paducah, 

Kentucky

Civilian Shut down 11300 Paducah Gaseous Diffusion Plant started the first 

production cells for enriched uranium in September 

1952 to fuel production reactors. The intent of this 

plant was to produce low-enriched uranium. The 

low-enriched uranium was then further enriched at 

K-25. In the 1960s, this plan was converted for 

producing fuels for commercial nuclear power 

plants [Paducah 2011].

Eunice, NM Civilian Operating 5900 In 2006, Louisiana Energy Service was issued a 

construction license to build Urenco, a European 

consortium of enrichment facilities. This facility is 

based on a gas centrifuge technology to enrich 

uranium fuel for power reactor use. This facility is 

the first enrichment plant since 1956.

Savannah 

River, S.C. 

[NUREG1767]

Civilian Under 

construction 

3.5 (Pu) Department of Energy’s Mixed (MOX) Fuel 

Fabrication Facility at Savannah River Site has 

been under construction since 2007. This facility 

was scheduled to be completed in 2019, but may 

be canceled. 

Piketon, Ohio Civilian Under 

construction 

3800 USEC American Centrifuge Plant was first built as 

a pilot plant. Once the Department of Energy 

grants the loan guarantee, then the full-scale plant 

will be built.

Areva Eagle 

Rock, Idaho

Civilian Planned 3300-6600 Areva planned to begin construction in 2011 but the 

project is on hold.

GLE, 

Wilmington, 

NC

Civilian Planned 3500-6000 Global Laser Enrichment (GLE) delayed again a 

decision on building a commercial-scale facility. 

*THM stands for ton of heavy metal.
2-7 

Table 2-3 Status of high-level waste vitrification facilities [Jain 1998, DOE 1999]

Site Location Technology* Started Status 

Marcoule 

Vitrification Facility 

(AVM)

Marcoule, 

France

Induction 

melting

1978 Facility shut down for 

decommissioning (1997)

La Hague R7 and 

T7 Vitrification 

Facilities

La Hague, 

France

Induction 

melting

1989 (R7 facility)

1992 (T7 facility)

The La Hague facilities are 

operational and are used to 

support spent nuclear fuel 

reprocessing.

Waste Vitrification 

Plant (WVP)

Sellafield, 

U.K.

Induction 

melting

1991 WVP is operational and is 

used to support spent 

nuclear fuel reprocessing. 

Pamela 

Vitrification Facility

Mol, Belgium Joule heating 1985 Facility shut down for 

decommissioning (1991)

West Valley 

Demonstration 

Project (WVDP)

West Valley, 

New York, 

U.S.A.

Joule heating 1996 Facility shut down for 

decommissioning (2002) 

[Meess 2011]

Defense Waste 

Processing 

Facility (DWPF)

Savannah 

River Site, 

Aiken, South 

Carolina, 

U.S.A.

Joule heating 1996 Operating

Tokai Vitrification 

Facility (TVF)

Tokai, Japan Joule heating 1995 Shut down

Mayak Vitrification 

Plant

Mayak, S. 

Ural, Russia

Joule heating 1987 (I melter)

1991 (II melter)

Operating 

Waste 

Immobilization 

Plant (WIP)

Trombay, 

India

Pot glass 

induction 

melting

1986 Operating

*See Section 3.1.5.2 for the vitrification technology description.

3-1 

3. OPERATIONS AND SYSTEMS

In the U.S., the majority of spent fuel is a product of operating light-water reactor plants. 

Presently, the enrichment of fresh fuel is up to approximately 5%, and the typical burnup of 

spent fuel is about 60 gigawatt-days per metric ton of uranium. For fuel reprocessing, spent fuel 

is cooled for at least 5 years, although the majority of fuel in the waste stream is significantly 

older. Accounting for the range of fuel types and ages in the design and operating procedures 

for a fuel reprocessing facility is essential. 

There are four main separation processes that are either in use or proposed for future use. 

Briefly, these are the following:

1. UREX Process – This process is similar to the PUREX process except that the

chemistry is adjusted so that uranium and technetium are extracted in two product

streams while product waste stream contains transuranics and fission products.

2. PUREX Process –This process uses nitric acid for dissolution and tributyl phosphate

(TBP) as an organic solvent [Pace 2006]. This process is commonly used to extract

plutonium for use in weapons or mixed oxide fuel for reactors and forms the base

separation process for fuel reprocessing in the world today.

3. Co-Extraction Process – This process was developed by CEA (Commission of Atomic

Energy in France) and AREVA, and is a modified version of the PUREX process. It is

divided into two main separation phases: a) extraction cycles for separating and

purifying U-Pu mixtures and b) forming a U-Pu nitrate solution to convert U and Pu into

(U, Pu)O2 [NUREG-1909]. A detailed AREVA process description is provided in [Bader

2011].

4. NUEX Process – This is also a modified PUREX separation process producing four

outputs: (a) recycled uranium, (b) mixed plutonium, neptunium, and uranium to enhance

proliferation resistance, (c) mixed americium and curium, and (d) a waste stream for

vitrification containing the remainder of the spent fuel isotopes [Energy Solutions 2008].

Currently, as indicated in the previous section, there is no operating spent fuel reprocessing 

facility in the U.S. Section 3.1 describes a reprocessing facility based on the co-extraction 

process as an example of a facility that might be built in the U.S. Section 3.1 also describes the 

typical operations of a co-extraction facility. Section 3.2 describes the typical physical systems 

that can be found in such a facility.

3.1 Operations

In most spent fuel reprocessing facilities, the following operations are typical (illustrated in

Figure 3-1): 

1. Receiving and shipping

2. Storage (spent fuel pool and fuel container processing)

3. Crushing/shearing/chopping (processing of spent fuel)

4. Chemical processing

a. Extraction

b. Separation
3-2 

c. Purification

d. Solvent recovery

e. Enrichment (optional)

5. Waste management, which includes waste storage and vitrification

Chopping, shearing and 

crushing

Dissolution in nitric acid 

(Leaching)

Fission products

uranium and transuranics

separation

Spent Nuclear Fuel

Off-gases

Off-gases and cladding

Nitric acid

Solvent tributyl 

phosphate and kerosene

Uranium and Plutonium 

partition

Nitric acid recovery

High-level waste 

(fission products)

Uranyl nitrate

Purification and 

conversion to oxide

U/Pu nitrate

UO3 product (U,Pu)O2 product

Liquid Waste

Solidification

(Vitrification)

Fuel Processing Chart_3.vsd

Store in spent fuel pool

Purification and 

conversion to oxide

Off-gases

Liquid waste 

storage

Solid waste 

storage

Co-extraction 

U and PU

Solvent

FP scrubbing

FP 

Raffinates 

Most U 

scrubbing Pu stripping U stripping

Solvent

U solution Diluted nitric acid

Reducing solution

Loaded solvent

Pu/U solution Scrub

U and Pu partition

FP, U, transuranics separations

Countercurrent 

concept

Figure 3-1 Typical operations at spent fuel reprocessing facilities that uses a coextraction process with countercurrent concept1

The following subsections provide a brief discussion of the processing facilities in which the 

operations described above are performed. 

3.1.1 Shipping and Receiving Operations

Shipping and receiving operations use certified shipping packages. Once spent fuel arrives at 

the facility, it may be moved to a spent fuel pool where the spent fuel is stored until reprocessing 

begins. 

1 Note that the countercurrent flow diagram of the extraction phase was adapted from [Todd 2008], but it has been 

enhanced for the co-extraction process. A detailed description of each operation is presented in subsequent sections.
3-3 

3.1.2 Storage

Fuel reprocessing facilities have an area designed to store incoming spent fuel ready for 

reprocessing. This can be in dry storage or in a spent fuel pool. Since the spent fuel is allowed 

to cool for at least five years before being accepted for reprocessing, the decay heat of the 

spent fuel stored in the spent fuel pool at a fuel reprocessing facility is typically lower than that 

stored in the spent fuel pool at a nuclear power plant. Potential events that can occur in a 

shipping and receiving facility are mostly from dropping of shipping packages, fuel assemblies, 

or fuel rods. 

3.1.3 Crushing, Shearing, and Chopping

Spent fuel assemblies are chopped, crushed, or sheared to increase the surface area for 

interaction between solvent and fuel. During this operation, fission product gases may be 

released and thus adequate ventilation and filtration systems must exist to capture these gases. 

Accidents related to oxidation of cladding and fuel can occur, especially when the interactions 

with air occur at elevated temperatures. Because of the high radiation level associated with 

spent fuel, this operation is often performed behind a shielded wall or in a hot cell where the 

ventilation and atmosphere are controlled. This portion of the facility has a potential for fires 

involving pyrophoric metals. A loss of ventilation/atmosphere control could also lead to a 

radioactive release. 

3.1.4 Extraction and Separation

There are several separation processes employed in spent fuel reprocessing plants. All but the 

electrochemical process involves dissolution of spent fuel and cladding materials as well as 

extraction of certain constituents for further processing. Below is a list of chemical processes 

that have been developed for use since World War II:

1. Bismuth Phosphate Process – This process was developed during World War II with the

specific purpose of recovering plutonium (uranium is not recovered). Six steps are used

in this process: (1) fuel dissolution in nitric acid, (2) adjustment of plutonium to the

tetravalent state Pu(IV) with ferrous iron or nitrite, (3) addition of bismuth and phosphoric

acid to separate Pu(IV) and BiPO4 from uranium, aluminum, and most fission products,

(4) re-dissolution of the Pu in highly concentrated nitric acid, (5) incorporation of

dichromate to oxidize plutonium to the hexavalent state, and (6) precipitation of BiPO4 to

obtain Pu(VI) [Campbell 1990]. Steps (2) to (6) can be repeated until the desired purity

is achieved. This process was used at Hanford starting in late 1944 [Gephart 2003].

2. Ion Exchange Process – This process was developed about the same time as the

Bismuth Phosphate Process. This process uses sulfonic acid and organic cation

exchange resins. A distinct disadvantage of this process is that it is difficult to obtain a

highly pure Pu product [Campbell 1990].

3. REDOX Process – This process uses nitric acid and aluminum nitrate for dissolution and

hexone (methyl isobutyl ketone) as an organic solvent [Pace 2006]. It is a continuous

process to recover both U and Pu with high yield and high purity. The process was

developed at Argonne National Laboratories, and tested in a pilot plant at Oak Ridge in

1948-49. A REDOX plant was built at Hanford in 1951. Hexone is highly flammable and

volatile [Todd 2008]. Hexone is also unstable in high concentrations of nitric acid and a
3-4 

large amount of neutral salt, such as aluminum nitrate, is required to maintain chemical 

stability. Because of the waste management issue related to large volumes of waste, a 

process with emphasis on recovered and recycled waste is preferred [Campbell 1990].

4. BUTEX Process – This process uses dibutyl carbitol as a solvent and nitric acid as a

salting agent. This method eliminates the need for aluminum nitrate used in the REDOX

process.

5. PUREX Process –This process uses nitric acid for dissolution and tributyl phosphate

(TBP) as an organic solvent [Pace 2006]. This forms the base separation process for

fuel reprocessing in the world today. This process was used at Hanford in the PUREX

plant starting in the early 1950s [Gephart 2003].

6. Co-Extraction Process – This process was developed by CEA and AREVA , and is

divided into three main phases: a) extraction cycles for separating and purifying U-Pu

mixtures, b) forming a U-Pu nitrate solution to convert U and Pu into (U, Pu)O2, and c) a

powder metallurgy process to produce the fresh mixed oxide fuel [NUREG-1909]. A

detailed AREVA process description is provided in [Bader 2011].

7. NUEX Process – This is a separation process producing four outputs: (a) recycled

uranium, (b) mixed plutonium, neptunium, and uranium to enhance proliferation

resistance, (c) mixed americium and curium, and (d) a waste stream for vitrification

containing the remainder of the spent fuel isotopes [Energy Solutions 2008]. The NUEX

process contains three primary separation processes: (a) Primary separation using TBP

dissolved in odorless kerosene, similar to the primary process used in THORP; (b)

TRUEX (Transuranic Extraction); and (c) TALSPEAK (Trivalent Actinide-Lanthanide

Separation by Phosphorus reagent Extraction from Aqueous Komplexes). Similar

descriptions are given elsewhere [NUREG-1909].

8. FLUOREX [Ogata 2003] – A hybrid system of using the fluoride volatility process to

recover most of uranium prior to the conventional PUREX process, where the residual

uranium and plutonium are purified (see PUREX process above). This process is being

employed at the Rokkasho Reprocessing Plant.

9. Electrochemical [NUREG-1909] – This process requires that any oxides in the spent fuel

be reduced to metallic form – either chemically by molten lithium or by electrolytic

reduction in molten lithium chloride. The metal from oxide reduction or metallic spent

fuel including cladding materials becomes the anode in an electro-refiner. This refiner

contains a molten electrolyte salt (a mixture of LiCl and KCl) atop of cadmium metal. For

an operating condition of 12 hours, the products from the electro-refiner contain: an

anode with elements that are stable as metals (e.g., zirconium, technetium, iron, and

molybdenum), a cathode containing most of the uranium as metal, and a cathode

containing TRU elements as metal, and rare earth fission products. The molten salt

contains most of the fission products as chlorides.

3.1.5 Waste Management

There are a number of waste streams produced by fuel reprocessing. High-level waste may be 

stored onsite or placed in a certified shipping container to be shipped to other sites to be 
3-5 

processed. If the facility is capable of enrichment, the processed uranium and plutonium may 

be further enriched. If not, these materials are placed in certified Special Nuclear Material 

(SNM) containers to be shipped elsewhere for further enrichment. Other residual chemicals 

may also be shipped to other facilities. 

After the spent fuel has been processed, the product, waste, and recycle streams may require 

storage prior to being shipped, subject to further reprocessing, or re-use in operation processes. 

Recycled acid must be contained in a robust and noncorrosive container. The end products, 

uranium oxide and plutonium oxide/mixed oxide, can be stored in a certified shipping container 

to be shipped out to fuel fabrication or enrichment facilities. The high-level waste can be stored 

on-site in solidified form or for further reprocessing of the waste. Storage-related accidents are 

possible, especially for high-level liquid waste, in which radiolysis of aqueous solutions may 

result in high concentrations of hydrogen. Inadequate ventilation can lead to explosions. 

As a byproduct, the waste stream from spent fuel reprocessing contains most of the non-usable 

fission products and impurities that may not be suitable for further reprocessing. Some streams 

may be separated into low-level, intermediate, and highly radioactive liquid waste. Liquid waste 

is stored in tanks, while any processed waste in solid form can be vitrified. Similarly, some 

types of liquid waste may be vitrified once dried. The vitrification process combines solid waste 

with glass frit, which is placed into a melter for heating until a glass is formed. The vitrification 

process has been used extensively for immobilizing waste for long-term disposal options.

The potential accidents associated with the waste operation include:

1. Storing liquid waste – potential H2 explosion (due to radiolysis of water) and leaks in

tanks, valves, and pipes

2. Drying and solidification of the liquid waste – external heating that could cause runaway

chemical reactions and possibly gas explosions

3. Loading of the waste into containers – spills and potential fires

4. Loss of cooling of high-level waste (HLW) storage tank

Note that human errors/machine malfunctions typically cause the accident conditions during 

loading of waste into the containers. 

3.1.5.1 Liquid Waste Storage

Onsite liquid waste storage is often needed as part of the fuel reprocessing operation. Several 

types of liquid waste can be handled that have low, intermediate, and high levels of radioactivity. 

High-level liquid waste requires shielding and additional equipment and monitoring systems to 

ensure that hydrogen buildup due to radiolysis is controlled. Portions of these liquid wastes may 

be further processed into solid forms, as discussed in the next section. 

3.1.5.2 Drying and Solidification (Vitrification)

In order to be solidified, liquid waste must undergo drying to remove solvent. This drying 

generally requires heating. First, a facility may use a concentrator to reduce the amount of 

solvent and to increase the concentration of the waste. Then, an evaporator may be used to 

further reduce the solvent content. Finally, a denitrator (or calciner) is used to completely 
3-6 

remove solvent. Once dried, the waste is subjected to a solidification process. The 

solidification process is often referred to as vitrification, which immobilizes the waste material in 

glass form. There are other non-vitrification solidification technologies that use sintering or a

molten metal process [Jain 1998b]. Other solidification processes, such as for uranium and 

plutonium, are discussed in the description of the extraction cycle and in the descriptions of the 

Barnwell spent fuel reprocessing plant in Appendix A, and a recently proposed plant from 

AREVA [Bader 2011]. This section only describes the vitrification technology for creating solid 

waste forms. 

Vitrification processes demonstrated or used by the Department of Energy (DOE) include: the

joule-heated melter process, plasma arc centrifuge treatment (PACT) system, plasma hearth 

process (PHP), American Society of Mechanical Engineers (ASME)/United Status Bureau of 

Mines (USBM) arc melter furnace, and in-situ vitrification (see Table 3-1 for a description of 

each of these processes). Among these vitrification processes, only joule-heated melters have 

been used on a production scale at DOE.

Table 3-1 Solidification Technology [DOE 1999]

Technology* Description Comments

Joule-heated 

melter

This technology utilizes the melter furnace that incorporates a 

process of heating the glass by conducting electric current 

through the glass matrix. Both the waste concentrate and the 

glass frit (glass former) are directly introduced into the melting 

chamber. (See Figure 3-2 for the conceptual design of this 

technology.)

This technology is well 

established and mature 

for the INCONEL melter 

used in the U.S., 

Japanese and Belgian

plants over 20 years 

[Jain 1998b].

Induction melting 

[Jain 1998a] 

The French AVM process (Atelier de Vitrification Marcoule)

technology utilizes a 2-stage process that consists of a 

combination of rotary kiln (a calciner) and an induction-heated 

metal glass-melting crucible [Baehr 1989]. (See Figure 3-3).

This technology is used 

at both United Kingdom 

and French vitrification 

facilities.

Plasma arc 

centrifuge 

treatment (PACT)

This technology utilizes the heat transferred from a plasma arc 

torch to create a molten bath. In the process, the solids are 

melted into the bath (~1760 °C) while organics are evaporated 

and/or pyrolized and partially oxidized. The off-gas leaving the 

primary chamber enters the secondary combustion chamber 

(SCC), where a complete oxidation takes place. (See Figure 3-4

for the schematic).

This technology has 

been demonstrated 

under the Environmental 

Protection Agency 

superfund innovative 

technology evaluation.

Plasma hearth 

process (PHP)

This technology is intended to stabilize mixed wastes. It is a hightemperature thermal treatment process using a plasma directcurrent arc torch in a stationary, refractory-lined melter. The 

plasma arc destroys organics and stabilizes residuals in a vitrified 

slag. The drummed waste is fed to the plasma chamber. Similar 

to PACT, the off-gas is transported to the cyclonic combustion 

chamber, where a natural gas burner is used to completely burn 

organics to yield CO2 and water. (See Figure 3-5 for the prototype 

design.)

This technology was 

developed for the DOE 

mixed waste treatment.

Arc melter 

furnace

This technology utilizes a plasma arc furnace to pyrolyze 

combustible organic material and melt residual inorganic materials 

with an electric arc. This furnace contains three carbon 

electrodes, a continuous feed system, and an off-gas treatment 

system with slag- and metal-trapping capabilities. (See Figure 3-6

for the design).

This technology was

used for the Phase 1 

testing of Hanford lowlevel waste [Jain 1998b].
3-7 

Technology* Description Comments

In-situ vitrification 

ISV) 

This technology uses electrical power to melt in-situ contaminated 

earthen media such as soil, sediment, and mine tailings. The ISV 

process is initiated by formation of a pool of molten soil at the 

surface of a treatment zone between four electrodes.

This technology has 

been evaluated and 

implemented within the 

DOE and other federal 

agencies, and aboard 

[Jain 1998b]. It has 

largely been rejected as 

a viable approach. 

* There are other technologies that are not included in this table. The reader is encouraged to consult [Jain 1998a]

and [DOE 1999].

Figure 3-2 Joule-heated melter concept [DOE 1999]

Figure 3-3 Induction melting used at the French AVM process [Marsden 

2010]
3-8 

Figure 3-4 A typical component list of a PACT melter [DOE 1999]

Figure 3-5 PHP prototype design [DOE 1999]
3-9 

Figure 3-6 Feed preparation and melter schematic for the arc melter 

furnace design [Jain 1998b]

The following processes are included within the vitrification operation:

1. Waste pretreatment and feed preparation

2. Vitrification

3. Off-gas treatment system

Many common problems facing the development of a reliable and safe melter technology for the 

vitrification of waste include [DOE 1999]:

• Inadequate design considerations for maintenance and surveillance

• Lack of a comprehensive failure modes and effect analysis

• Lack of a systems engineering analysis on design and construction

• Leakage through penetrations (for instrumentation, drains, etc.) of the melter shell(s) and

liner, which are typical weak spots

• Clogging of the melt discharge orifice

• Low reliability of critical support systems, including cranes, water cooling systems,

remote mechanical equipment, and maintenance support systems.
3-10

• Corrosiveness of the waste materials, especially where electrodes are concerned in

joule-heated melters

• Incompatibility of the refractory lining materials for the melter with the waste/glassforming materials

3.2 Systems

The following physical systems are included in a typical spent fuel reprocessing facility: 

1. Confinement – This system provides the mitigation and/or prevention of an

environmental release of radioactive elements and chemical products. This system also

has a significant impact on the safety of the plant, particularly the dose to onsite

employees as well as potential dose offsite.

2. Ventilation and filtration – This system is designed to reduce radioactive and chemical

contamination of various atmospheres within the facility through the use of filters. A

negative pressure differential maintained by the ventilation system reduces offsite

leakage through environmental pathways. Ultimately, the ventilation system processes

the air released to the environment to maintain acceptable radiation and chemical

releases. This system is also important to worker safety.

3. Fire suppression and detection – This system is used to mitigate the spread and

duration of a fire and to reduce the suspension of radioactive isotopes and harmful

chemical species.

4. Miscellaneous support systems – These systems can include electrical power, including

backup power for the plant, chemical systems supporting various plant operations, and

water resources that are used in the extraction and cleaning processes.

All of these systems are engineered safety systems and provide potential mitigation of an

accident inside the reprocessing facility. Each is described in more detail in the subsequent 

sections. Power backup is important since many of the ventilation/filtration systems and fire 

sprinklers may rely on this system if offsite power is lost. 

The NRC has published a number of guidance documents for the systems in fuel reprocessing 

plants (see Table 3-2).
3-11

Table 3-2 Selected Nuclear Regulatory Commission guidance 

documents for fuel reprocessing2

Regulatory 

Guide No. Title [Reference] Safety Functions or Pertinent Information

3.3 Quality Assurance 

Program Requirements 

for Fuel Reprocessing 

Plants and for 

Plutonium Processing 

and Fuel Fabrication 

Plants

• 10 CFR Part 50 Appendix B establishes quality assurance requirements

for nuclear power plants and fuel reprocessing plants.

3.6 Content of Technical 

Specifications for Fuel 

Reprocessing Plants

• Technical specifications should include both technical and administrative

matters.

• The use of safety limits and limiting control settings for establishing

technical specifications.

• In terms of administrative matters, four principal functions should be

performed:

o Operation of plant equipment

o Maintenance of equipment

o Record keeping

o Audits, reviews, and evaluation of operations

3.17 Earthquake 

Instrumentation for 

Fuel Reprocessing 

Plants

• Identification of the locations and requirement of earthquake detection

instruments and recorders for the structures, systems and components

important to safety.

3.18 Confinement Barriers 

and Systems for Fuel 

Reprocessing Plants

• Confinement systems minimize the radioactive materials to restricted

areas according to the limits specified in §20.103 of 10 CFR Part 20.

• These systems should minimize the radioactive materials to unrestricted

areas according to the annual limit prescribed by § 20.106(a) of 10 CFR

Part 20.

3.19 Reporting of Operating 

Information for Fuel 

Reprocessing plants

• Establishment of the reporting requirement, including types and time

intervals of the documents submittal.

3.20 Process Off -Gas 

Systems for Fuel 

Reprocessing Plants

• All parts of the process off-gas system should be designed to limit the

radioactive and noxious materials release

• The design must withstand postulated accident conditions

• Backup systems should exist to prevent any single failure

3.21 Quality Assurance 

Requirements for 

Protective Coatings 

Applied to Fuel 

Reprocessing and to 

Plutonium Processing 

and Fuel fabrication 

Plants

• This guide provides quality assurance requirements for protective

coatings over ferritic steels, aluminum, stainless steel, zinc-coated

(galvanized) steel, concrete or masonry surfaces of the plants.

3.22 Periodic Testing of 

Fuel Reprocessing 

Plant Protection 

System Actuation 

Functions

• The protection system that is important to safety should be designed to

permit periodic testing to extend to and include the actuation devices and

actuated equipment.

2 NRC guidance documents may need to be updated to support licensing of a modern reprocessing facility. 
3-12

Table 3-2 Selected Nuclear Regulatory Commission guidance 

documents for fuel reprocessing2

Regulatory 

Guide No. Title [Reference] Safety Functions or Pertinent Information

3.26 Standard Format and 

Content of Safety 

Analysis Reports for 

Fuel Reprocessing 

Plants

• This guidance provides the format and content to document a safety

analysis report for fuel reprocessing plant.

3.27 Nondestructive 

Examination of Welds 

in the Liners of 

Concrete Barriers in 

Fuel Reprocessing 

Plants

• This guide provides the requirement to examine (nondestructively) the

welds for the liners of concrete structures. The liners are important to

mitigate any spills inside the facility.

3.28 Welder Qualification for 

Welding in Areas of 

Limited Accessibility in 

Fuel Reprocessing 

Plants and in 

Plutonium Processing 

and Fuel Fabrication 

Plants

• This document describes the welder qualifications for weld fabrication

and repair of the safety related structures within the fuel reprocessing

facility.

3.29 Preheat and Interpass 

Temperature Control 

for the Welding of LowAlloy steel for Use in 

Fuel Reprocessing 

Plants and in 

Plutonium Processing 

and Fuel Fabrication 

Plants

• This document describes a method that meets the temperature-control

requirements for the welding of low-alloy steel components for fuel

reprocessing, plutonium processing and fuel fabrication plants.

3.30 Selection, Application, 

and Inspection of 

Protective Coatings 

(Paints) for Fuel 

Reprocessing Plants

• This guide provides a standard for evaluating and selecting protective

coatings (paints) for nuclear plants.

• This guide also provides guidance for preparing the surfaces to be coated

and for applying and inspecting the coatings under both shop and field

conditions.

3.31 Emergency Water 

Supply Systems for 

Fuel Reprocessing 

Plants [NRC 1975b]

• This guide discusses provision of water to meet the heat rejection

requirements for the fuel storage pool, waste storage tanks, and certain

process vessels in abnormal conditions.

• It also discusses provision of water for the fire protection system in case

the normal supply is disrupted, and

• It requires that water must be provided for any safety related equipment

or systems in accident conditions.

3.32 General Design Guide 

for Ventilation Systems 

for Fuel Reprocessing 

Plants

• This guide provides the requirements for a well-designed ventilation

system for fuel reprocessing plants.

3.37 Guidance for Avoiding 

Intergranular Corrosion 

and Stress Corrosion 

in Austenitic Stainless 

Steel Components of 

Fuel Reprocessing 

Plants

• This guide documents the guidance and steps that can avoid intergranular and stress corrosion in austenitic stainless steel components

used in fuel reprocessing.
3-13

Table 3-2 Selected Nuclear Regulatory Commission guidance 

documents for fuel reprocessing2

Regulatory 

Guide No. Title [Reference] Safety Functions or Pertinent Information

3.38 General Fire Protection 

Guide for Fuel 

Reprocessing Plants

• This guide documents a requirement to establish a fire protection

program within the facility to minimize and mitigate the fire accidents in

the facility. It also provides steps to mitigate fire situation with fissile

materials presented and critical control equipment presented.

3.40 Design Basis Floods 

for Fuel Reprocessing 

Plants and for 

Plutonium Processing 

and Fuel Fabrication 

Plants

• This guide describes methods of determining the design basis floods for

nuclear facilities (fuel reprocessing, plutonium processing and fuel

fabrication plants).

3.2.1 Confinement

One of the most important safety features in a spent fuel reprocessing plant is the confinement 

system. The building enclosing the extraction or storage operations is designed to minimize

radiation exposure to the workers and potential airborne release to the public and environment. 

Typically, the confinement requirements are based on radiation levels for standard operations. 

Table 3-3 shows the working environment enclosure requirements as a function of the radiation 

level, mainly for inhalation and direct exposure.

Table 3-3 Confinement needs for handling radioactive materials [Cadwallader 2005]

Work Environment Radioactivity Level (ALI*) Radioactive Quantity 

(Curies)

Laboratory bench top < 0.1 < 10-9

Laboratory fume hood 0.1 to 10 10-9 to 10-6

Unshielded glove box enclosure 10 to 1,000** 10-6 to 10-3

Shielded glove box enclosure 10 to 1×106 10-3 to 100

Hot cell ˃ 1×106 100 to 106

* ALI stands for annual limits on intake.

** The glove box may protect radiation level > 1,000 ALI if the emitted radiation is α or β particles rather than ɣ or 

neutron. If the radiation is penetrating, then a shielded glove box may need to be used at a small ALI value.

For direct radiation exposure, the use of physical barriers such as concrete walls, are used to 

reduce worker exposure from extraction operations and high-level waste storage cells.

To mitigate the release of airborne material during an accident, multiple-tiered confinement 

buildings are generally used. Figure 3-7 shows a 3-tiered confinement system. As shown in 

this figure, the primary confinement/barrier (Zone I) has either a storage container, process 

vessel, glove-box, or processing equipment that contains the radioactive and/or toxic materials. 

The primary confinement may be contaminated and is treated as such. To contain the 

contamination if Zone I is breached, the primary confinement/barrier is contained within a 

secondary confinement/barrier (Zone II), which is physically an operation room or cell. 

Depending on the contamination/acidity level expected in this secondary confinement/barrier, 

the floor, walls, and ceiling may be designed to protect building structures from chemical 
3-14

interactions caused by leaks or spills. This is usually done by using a nonreactive liner such as 

stainless steel. In most extraction operations, the processes are performed inside hot cells that 

have walls thick enough to shield workers from direct exposure to radiation from the spent fuel. 

Tertiary

Confinement 

(Barrier)

Building 

Structure, 

including 

corridors**

Normally Clean

Primary 

Confinement 

(Barrier)

Container, Process 

Vessel (equipment), 

Glove boxes, or 

Fume Hood*

Radiotoxic 

(contaminated)

Secondary

Confinement 

(Barrier)

Operation/Maintenance 

Cells or Rooms*

Potentially Contaminated

Air Flow

Air Flow

Air Flow

*This barrier/confinement may contain its own ventilation system that includes off-gas treatment

**This barrier/confinement may require airlock for personnel to enter the secondary confinement

Confinement_diagram(v0).vsd

Zone I

Zone II

Zone III

Figure 3-7 Three-tiered confinement zones used in radiological plant

In most instances, the secondary confinements/barriers reside within a third building structure

that acts as a tertiary confinement/barrier (Zone III). For example, the Barnwell Nuclear Fuel 

Plant was designed with 2 to 4 barriers. In this plant, all chemical extraction operations are 

done inside hot cells for which the floors and walls are lined with stainless steel and contain a 

sump for spill collection. The hot cells are contained in the main process building. All hot cell 

walls are thick enough to minimize worker exposure outside the cells, such as in operating 

corridors. 

Unlike nuclear power plants for which the containment is designed to withstand internal or 

external challenges such as high internal pressure and temperature and airplane crashes, many 

spent fuel reprocessing plants contain only confinement systems that may not be capable of 

withstanding any high energy accidents, such as an explosion inside or outside the facility. 

3.2.2 Ventilation

The ventilation systems provide pretreated airflow to the processing areas, and maintains 

pressure differentials between zones as required to confine radioactivity to designated locations

in the plant (see Figure 3-7). The ventilation system is designed to confine and channel any 

airborne release to treatment systems such as filters and off-gas treatment systems to capture 

volatile fission products or other toxic gases before they can exit via the exhaust stack. For 

example, the Engineering Alternative for Extraction study [Bader 2011] shows that the process 

*This barrier/confinement may contain its own ventilation system than includes off-gas treatment.

**This barrier/confinement may require airlock for personnel to enter secondary confinement. 
3-15

vessel ventilation system is used to treat any off-gas from the operations to ensure that no 

significant amount of either gaseous (or volatile) fission products or toxic gases are released.

Cooling and recirculation for the process cells are also provided by the ventilation systems.

Maintaining proper pressure differentials between the processing cells and the corridors or 

between the glove box and the operating cell is essential to keeping contamination within its 

intended confinement. Table 3-4 and Table 3-5 show the permissible quantity of materials 

within each confinement zone. Similar permissible levels of radioactivity and toxicity of 

materials are expected for a commercial reprocessing facility per NRC regulations [NRC 1979].

Table 3-4 Zoning of facilities based on radioactivity of materials handled [DOE 2003]

Quantity of Material Permitted in Zone at Any One Time*

Radiotoxicity of

Isotope

Primary Confinement Secondary 

Confinement

Tertiary 

Confinement

Very High ˃ 10 mCi 0.1 μCi – 10 mCi 0-0.1 μCi

High ˃ 100 mCi 1.0 μCi – 100 mCi 0-1.0 μCi

Moderate ˃ 1 Ci 10 μCi – 1 Ci 0-10 μCi

Negligible ˃ 10 Ci 100 μCi – 10 Ci 0-100 μCi

*There are practical upper limits to the material quantity in any particular zone, based on the material type and design of the

confinement system. For example, criticality safety concerns may restrict the amount of fissile material that can be handled at

one time, fire protection concerns may limit the amount of pyrophoric materials, and shielding considerations may limit the

amount of materials when penetrating radiation is emitted. The above criteria are based on the potential for the activity to

generate airborne radioactive materials.

Table 3-5 Zoning of facilities based on chemical contamination levels [DOE 2003]

Anticipated Contamination Levels

Type of Contamination Primary 

Confinement

Secondary Confinement Tertiary 

Confinement

Airborne a ˃ 100 × DAC 1 × DAC – 100 × DAC < 1 × DAC

Removable surface b ˃> RSCV c ˃ RSCV c < RSCV a The DAC is the derived airborne concentration value listed in 10 CFR 835, Appendix A for the type and chemical form of the 

material being handled. 

b

 The RSCV is the removable surface contamination value listed in 10 CFR 835, Appendix D, for the type of the material being 

handled. 

c

 RSCV level do not always directly lead to an increasing level of airborne contamination. The level of airborne contamination 

strongly depends on the potential for the particular activity to re-suspend the deposited particles into the atmosphere.

To capture any airborne release in the ventilation system, the ventilation system must contain 

the proper equipment such as high efficiency particulate air (HEPA) filters, charcoal filters, and 

sand filters). There are generally several ventilation systems supporting each confinement zone 

in the plant. 

Off-Gas System - This system (Zone I) is used in conjunction with process operations in which 

significant amounts of airborne and gaseous radioactive and toxic materials may be present 

(see Figure 3-7 for Zone I definition). In the proposed AREVA plant [Bader 2011], the off-gas 

system is used to remove noble and volatile fission products such as xenon, krypton, iodine, 

ruthenium, tritium, and carbon-14. Additionally, NOx must also be removed in the off-gas 

system before the effluent is released through the exhaust stack. Removal and trapping 

processes used in off- gas systems are also found in the proposed AREVA plant [Bader 2011]. 
3-16

Depending on the removal system used, the decontamination factor may be reduced when 

there are impurities in the gas stream. For example, an off-gas system used in the extraction 

operation may contain silver mordenite, an iodine sorbent to capture airborne iodine. However, 

experiments have shown that the decontamination factor of the iodine by this sorbent may be 

reduced in the presence of the airborne TBP vapor [Parker 1980]. 

Process Room System – This system is used to provide ventilation in the process cells or rooms 

where the spent fuel reprocessing operation resides. In most cases, the operations take place 

inside a hot cell, where the operator behind external walls uses manipulators to conduct 

operations. Within the process room, the atmospheric pressure is higher than that of the off-gas 

system for the processing equipment or process vessel, so that contamination in the process 

room is controlled. In the event of a chemical fire, the process room atmosphere may be 

inerted. 

Filter System – This system is a safety system that is used to capture airborne materials 

including an abnormal release of airborne radionuclides. The airborne materials may be in the 

form of vapors or aerosol particles. The filter systems can be located at the inlet and/or exhaust 

portions of the ventilation systems. Inlet filters are used to trap aerosol particles that enter the 

facility. Exhaust filters also trap aerosol particles that are generated during normal operations 

such as where crushing or chopping of the spent fuel assemblies takes place. During abnormal 

situations, filters are designed to be able to decontaminate 99.97% of 0.3 μm airborne materials 

and are designed to survive the accident. 

Zone II System – This system may include the Process Room System previously described, and 

is used to provide a secondary confinement zone so that airborne release from Zone I to Zone II 

is minimized (see Figure 3-7). 

Zone III System – This system is the outermost ventilation system that keeps the recirculation of 

the air clean for areas occupied by workers. This area includes maintenance rooms and 

corridors. 

Construction materials for the ventilation ducting are also important. These ducts may need to 

be able to withstand acidic environments, especially the ducts connected to the rooms or 

chambers with extraction equipment. 

3.2.3 Fire

Unlike a nuclear power plant, where spray systems in the containment are used to reduce the 

pressure from an energetic accident and wash down fission products, fire suppression systems 

are primarily designed to mitigate fires. However, these sprays can also reduce airborne vapors 

and aerosol particles. Fire detection systems do not perform the same function directly, but can 

facilitate human interdiction were a fire to occur. 

Fire sprinkler systems are active engineered safety features in a spent fuel reprocessing facility. 

They require a water source, pumps, and electrical power in order to function. Because the 

amount of fissile materials processed in the fuel reprocessing facility is often large, this system

must be designed to fight fire without inducing inadvertent criticality. The use of Halon fire 

suppression systems may be considered in areas where water could induce additional 

consequences.
3-17

Other than the process operations, sprinkler systems may be used in a storage building, pump 

houses, and in ventilation systems where filters are located.

3.2.4 Miscellaneous Support Systems

There are a number of miscellaneous support systems in a fuel reprocessing facility. These 

systems include:

• Electrical backup system – the function is to provide backup power in the case of loss of

offsite power. It must be able to provide adequate power to activate engineered safety

systems and cooling to processes that contain materials with high decay heat. Typical

backup systems include standby diesel generators that can provide power following a

loss of offsite power.

• Chemical supply system – the function is to provide the fresh chemicals and recycled

chemicals to support extraction and solidification operations in the facility.

• Water system – the function is to provide water to support various operations in the

facility, which can include:

o Spent fuel pools

o Steam to support concentrators, evaporators, denitrators, and calciners

o Extraction processes

o Fire-fighting

• Process control instrumentation – the function is to provide information to the plant

operators to evaluate the conditions in various processes in the facility, which may

include:

o Pressure control

o Temperature control

o Concentration control

o Volume control

4-1 

4. ACCIDENTS/INCIDENTS – HISTORY AND DISCUSSION

This section describes potential accidents or incidents that could occur at spent fuel 

reprocessing facilities (see Table 4-1). Reprocessing facility accidents involve spills, fires, 

explosions, and inadvertent criticality events; incidents are more minor occurrences. As shown 

in Table 4-1, the potential incidents are identified for each specific operation at the facility. The 

consequence of each incident is categorized in this table. Because of the types of chemicals 

used in the separation processes, chemical reactions that may lead to fire and explosions are 

included as possible accidents. In addition, because of radiation emitted from the spent fuel, 

any radiolysis process such as those involving water, can create flammable gases such as 

hydrogen that can lead to a deflagration or detonation. Inadvertent criticality events are 

possible since fissile material from the spent fuel is extracted and concentrated during the 

recovery process. A failure of the physical boundary of the processing equipment can permit a 

release. If the contents are liquid, a spill accident is also possible. These major anticipated 

accidents are identified and described in more detail in Table 4-2. 

Table 4-1 Potential incidents at a spent fuel reprocessing facility [DP-1558]

Process Low Consequence Intermediate 

Consequence High Consequence

Fuel Receiving and 

Storage Operation

• Fuel damage in transit

or upon arrival

• Fuel cooling time too

short

• Loss of cooling

• Surface contamination

not removed from cask

• Low water level

• Rupture of fuel during

storage

• Cask inadvertently vented

• Cask dropped

• Damage to fuel assembly

outside cask

• Damage to stored fuel

• Loss of cooling capability

• Hydrogen explosion in fuel

cleaning vessel

• Criticality

Shearing Operation • Fuel jammed or 

stranded in shear

• Irregular length of hulls

• Shear jammed

• Malfunction of inert gas

system

• Chopped fuel overheating

• Pressurization of shear

• Fuel element overheating

• Pyrophoric fire1

• Release of volatile and

particulate activity into

room

Dissolving 

Operation

• Un-complexed fluoride

in solvent

• Pu-rich residue settling

in dissolver, lines, and

other process vessels

• Dissolver seal failure

• Inadequate cladding

rinse

• Transfer error of

dissolver solution

• Suck back (due to

eructation and

pressurization of

dissolver)

• Siphoning

• Contact of sheared fuel at a

temperature about 300  ̊C

above boiling point of

dissolver solution

• Pressurization of the

dissolver

• Dissolver leakage (due to

corrosion gasket/valve

failure)

• Charge of inadequately

cooled fuel

• Malfunction of dissolver offgas (VOG) iodine adsorbers

• Excessive interaction of fuel

external to dissolver (slightly

reduction in criticality safety

margin)

• Excessive pressure in water

wash compartment except

during basket exchange.

• Precipitation of Pu polymer

(potential criticality)

• Zirconium fire1

• Zirconium explosion (due to

significant fines in the

dissolver)1

• Explosion in the iodine

adsorber

• Overconcentration in

dissolver (potential

criticality)

• Self-concentration of

dissolver solution (potential

criticality)

• Low H+/Pu ratio in dissolver

solution (potential criticality)

• Inadequate poison in

dissolver or leach solution

(potential criticality)
4-2 

Table 4-1 Potential incidents at a spent fuel reprocessing facility [DP-1558]

Process Low Consequence Intermediate 

Consequence High Consequence

• Dissolver pot coils not

submerged during shutdown

(high release of volatiles to

room atmosphere)

• High fuel loss in leached

fuel fragments

• Failure of cladding monitor

(potential criticality)

• Absence of basket in

dissolver

• Explosion in dissolver

vessel or hold tanks (H2

explosion)

Solvent Extraction 

Operation

• Plutonium concentration

in a solvent extraction

contractor increases

beyond normal values

• Potential backup of

radioactivity by air lines

• Flow reduction or

interruption of scrub

streams to contactors

(uranium)

• Loss of organic flow to a

contactor (uranium)

• Low temperature in

scrub stream to

contactor (uranium)

• Low temperature in

contactor

• Emulsion in solvent

washer

• High plutonium losses

in waste streams from

contactors (in PUREX

process)

• High plutonium

concentration in

contactor (in PUREX

process)

• Excess gamma activity

in plutonium product

• Potential for abnormal

accumulation of fissile

material by plutonium reflux

in contractor, particularly

during high activity waste

rework (decreased criticality

safety margin)

• Potential for excessive

product loss to spent organic

effluent (reduced criticality

safety margin)

• Accumulated solvent in

rooms

• Solvent extraction system

contents removed by

overflow (in sump)

• Solvent fire2

• Potential for buildup of

unsafe amounts of fissile

material in organic storage

tank (potential criticality)

• Potentially unsafe feed

concentration (potential

criticality)

• Potential for excessive

plutonium loss to aqueous

waste (potential criticality)

• Incorrect material transfer

of feed chemicals, product

materials or incorrect

materials (potential

criticality, excessive

radioactive material

release)

Product Evaporation • Leaks in tank containing 

concentrated uraniumplutonium solutions

• Transfer error in a product

evaporator system

• Coil or tube-bundle (in

reboilers and condensers)

failure

• Overflow of an evaporator

• Potential reaction between

TBP and UNH (red oil

explosion) in a product

concentrator

• Overconcentration of U/Pu

product (potential criticality)

• Boil-over (release of

radioactive materials from

primary containment)

Waste Evaporation • Overconcentration of 

waste

• Ruthenium escapes to stack

• Leaks (to sumps)

• Transfer error in the

evaporator system

• Coil or tube-bundle (in

reboilers and condensers)

failure

• Boil-over (release of

radioactive materials from

primary containment)

• Hydrogen explosion in

high-activity waste

evaporator
4-3 

Table 4-1 Potential incidents at a spent fuel reprocessing facility [DP-1558]

Process Low Consequence Intermediate 

Consequence High Consequence

• Overflow of a feed or bottom

tank

• High radioactivity in highactivity waste evaporator

condenser

• Inadequate vessel cooling

capacity that leads to release

of aerosols to room

atmosphere

• Leak of high-level waste

concentrate

Acid Recovery 

Operation

• Pressurization of acid

absorber/fractionator

• Leaks due to corrosion

• Eructation due to foreign

materials in feed

• Solvent addition to feed tank

• Red oil explosion due to

solvent in feed or

temperature > 140  ̊C

• High radioactivity in

recycled acid and/or water

Recovery Ion 

Exchanger 

Operation

• Overflow

• Leakage due to corrosion

• Uncontrolled reaction

between nitric acid and

anion exchange resin in the

primary recovery column

• Ion exchange resin fire

(leads to radioactivity

release to environment)

Off-Gas Treatment 

Operation

• Loss of off-gas header

volume

• Iodine removal

inadequate (increased

iodine accumulation in

krypton solvent

recovery)

• Excess nitrogen oxides

in VOG iodine adsorber

stream

• High radioactive particulate

releases to building

ventilation filters (increase

release to sand filter)

• Off-gas heater inadequately

heats streams to iodine

adsorbers

• High ruthenium adsorber bed

temperature

• High krypton-85 releases

• Filter failure

• Process vent system

pressurized (leads to

release to sand filter and

smaller amount to

atmosphere

Uranyl Nitrate 

Operation

• Overflow from storage

tank

• High uranium

concentration in

recovered acid

• Plugging of instrument lines

and sensors during

evaporation

• Steam coil leak in evaporator

reboiler

• Cooling coil leak due to

corrosion

• Overflow in concentration

system

Uranium Denitration 

and Reduction 

Operation

• Pressurization of

denitrator

• High uranium

concentration in

recovered acid

• Over-pressurization of

reductor

• Fire in ammonia dissociator

cubicle

• Reductor malfunction which

leads to release of UO2 and

hydrogen

• Denitrator eructation (leads

to fire and explosions).

Red oil explosion can occur

in a denitrator if TBP is

present.

• Hydrogen explosion

(hydrogen leaked from

reductor)

Uranium 

Hexafluoride 

Operation

• Uranyl nitrate solution

leaks

• Uranium hexafluoride

release

• Airborne uranium oxide

• Criticality due to excessive

uranium accumulation (high

enrichment fuel)

• Fire (due to solvent in feed,

hydrogen leak and
4-4 

Table 4-1 Potential incidents at a spent fuel reprocessing facility [DP-1558]

Process Low Consequence Intermediate 

Consequence High Consequence

presence of pyrophoric 

oxide powders)

Co-conversion 

Process Operation 

(UO2-PuO2

production)

• Transfer errors

• Process solution leak

• Overflows

• Chemical addition error

• Pressurization of calciner

• Calciner breached from

internal corrosion

• Calciner breached from

impact (crane drop)

• Excessive penetration of

calcine through the calciner

primary filter

• Filter system breached

• Fire suppression system

failure

• Uranium in steam

condensate or cooling water

returns the uranium oxide

dissolver

• Loss of power

• Ammonium nitrate

explosion in calciner

• Hydrogen explosion in

product storage due to leak

in hydrogen delivery

system

• Hydrogen explosion in Pu

nitrate storage tank due to

radiolysis

• Uncontrolled reactions

• Fire in process room

• Hydrogen explosion in cold

chemical area

Waste Calcination 

Operation

• Excessive penetration of

calcine through sintered

metal filters

• High temperature breach of

the calciner

• Calciner breached from

internal corrosion, thermal

shock, pressurization and

impact

• Filter systems breached

Solid Waste 

Processing 

Operation

• Airborne cement dust in

grout mixer area

• Radioactive

contamination in

cement preparation

area

• Failure of contaminated

process components

• Drums improperly filled

(no cement added)

• Violation of stack release

guide

• Container failure

• Power failure

• Fire in fuel hardware

fixation area

• Drum overflow in fuel

hardware fixation area

• Excessive fissile material in

hulls

• Waste container failure

after filling

• Fire in β-ɣ waste facility

• Criticality potential in α

waste

Solidification of 

Intermediate-Level 

Liquid Waste

• Transfer error that leads

to high activity waste

• Overexposure of

personnel to radiation

• Mixer plugging, which

leads to worker

exposure

• Airborne activity due to leaks,

overflows, foam-out from

mixer, suck back, and loss of

vessel ventilation (power

failure)

• Waste container failure due

to weld failure, fault in drum,

corrosion, or impact

• Fissile material in feed

• Uncontrolled reaction in

mixer or product container

Vitrified High-Level 

Waste Storage 

Operation

• Contaminated canisters

• Water loss from storage

pool

• Loss of cooling water and

shielding

• Canister stress corrosion

• Canisters raised above

adequate shield level

• Canisters dropped during

handling

• Waste transfer line rupture

due to the high viscosity

sludge causing plugging,

• Fire
4-5 

Table 4-1 Potential incidents at a spent fuel reprocessing facility [DP-1558]

Process Low Consequence Intermediate 

Consequence High Consequence

leading to potential 

accumulation of flammable 

gases and heating of the 

decay heat in high-level 

waste [Jain 1998] 

• Electrically heated melter

short due to accumulation of

noble metal on melter floor

[Jain 1998]

• Waste pretreatment issues

that lead to release of Pu with

the effluent in the evaporator

[Jain 1998]

• Improper use of chemicals in

feed preparation that leads to

erosion and corrosion of online instruments, causing

malfunctions of on-line

pressure, temperature and

level instruments [Jain 1998] 1

During the head-end process when the spent fuel rod is being chopper or crushed, magnesium-clad fuel and uranium/plutonium 

metal fuel are exposed to air, meaning that combustion may occur at normal temperatures [OECD 2005]. Zircaloy debris or dust 

could pose a potential fire or explosion if this material is used for the cladding [IAEA DS360]. 

2

N-dodecane in TBP in the PUREX process poses potential combustion conditions [Ikeda 2003]. 

Table 4-2 Summary of possible major accidents at a spent nuclear 

fuel reprocessing facility

Accident Cause Consequence 

Spill

Rupture of a 

high-level liquid 

waste storage 

tank

• Corrosion and erosion

• Mechanical stress from overpressure

and loss of cooling

• Simultaneous failure of the vessel off

gas (VOG) system (e.g., by a H2

explosion from loss of air sparge)

• Release of tank contents to room floor

• Airborne release of tank contents in cell

• Potential for large release to process stack if

safety feature fails

• Decay heat in the released liquid may cause

any liquid to boil away and melts the remaining

solids that can cause additional gases and

airborne aerosols

Fire

Solvent fires • High temperature in organic streams

• Solvent leak and vaporization

• Solvent sprays on hot equipment (e.g.,

steam line, evaporator reboiler)

• Solvent spill from skimmer overflow,

flooded decanter, and makeup tank

overflow

• Ignition source present

• Room atmosphere oxygen concentration

above limit

• Airborne activity in process room

• Possible overheating and plugging room

ventilation filters

• Air reversal

• Loss of process control

• Equipment damage

• Energy release by solvent burning somewhat

limited by amount in the process room
4-6 

Table 4-2 Summary of possible major accidents at a spent nuclear 

fuel reprocessing facility

Accident Cause Consequence 

Ion exchange 

resin fire or 

explosion

• Self-heating of resin due to abnormal

conditions in column: high nitric acid

concentration (9 molar [NUREG-1909]),

column overloading, dry resin in column,

and high column temperature

• Self-heating of spent resin in waste

• Spontaneous combustion of spilled

resins

• Exposure of resin to strong oxidants

other than nitric acid, such as

permanganate or chromate ions, to high

radiation doses, to strong reducing

agents such as hydrazine and to

catalytic metals such as iron, copper or

chromium [NUREG-1909]

• Allowing resin to remain in a stagnant,

non-flow condition while loaded with

exchanged metal and/or in contact with

process concentrations of nitric acid

[NUREG-1909]

• Column pressurization

• Column rupture

• Eruption

• Airborne activity in room, release of fission

products to ventilation system

• Possible plugging or overheating of VOG filters

• Waste fire

• Room pressurization

Hydrogen 

Explosion

Feed tank or 

vessel vent 

system

• Concentration of feed solution

• Hydrogen produced by radiolysis of feed

solution

• Loss of purge to feed tank

• Oxygen source

• Ignition source

• Severe damage to equipment

• Potential damage to off-gas system

• Release of radioactive material to process

room

• Potential initiation of fire

High activity 

feed, plutonium 

evaporator 

bottom tank, 

product storage 

tank, mixedoxide calciner

• Hydrogen produced by radiolysis of

product solutions

• Ignition source

• Loss of vessel purge

• Damage to off-gas system and release of off

gas

• Equipment damage

• Release of fissile to room floor

• Room pressurization

VOG header • High H2 or HT content in VOG stream,

air in stream, and ignition source

• Propagation from explosions in other

equipment in room

• Aerosol release by filters

• Damage to VOG system

• Pressurization of process room via room

exhaust header

Process room • Ignition presence

• Excessive oxygen presence in the room

atmosphere

• Damage to shielding room, ventilation and

VOG systems

Other 

Explosion1

Red oil explosion 

in concentrator, 

uranium 

evaporator 

(denitrator)

• Solvent in feed

• Temperature > 130  ̊C

• Heavy metal (U, Pu) present

• Rupture of concentrator may cause violent

ejection of liquid with formation of aerosol

• Damage to VOG treatment system, leading to

the release through stack.

• Induced fire

• Projectile generation in room

• Room pressurization
4-7 

Table 4-2 Summary of possible major accidents at a spent nuclear 

fuel reprocessing facility

Accident Cause Consequence 

Hydrazoic acid 

explosion

• Uncontrolled reaction of hydrazine

products used in solvent recovery.

Hydrazine is used in the extraction step

during the Pu stripping operations

• Damage to VOG system

• Fire in solvent recovery

• Room pressurization

• Airborne activity

Criticality

Fuel storage • Distortion of fuel storage array

• Fuel improperly stored

• Fuel assembly dropped into fuel storage

array

• Fissile material on pool water filter

• Damaged elements could release short-lived

noble gases and iodine

• Radiation and neutron locally high but largely

shielded by pool water

Inadequate 

poison in solvent

• Poison concentration too weak

• Wrong chemical added

• Failure to add poison at correct volume

• Dissolver (in conjunction with undetected

plugging of liquid or solids) could release

volatile fission products

Digester (in conjunction with undetected 

accumulations of solids or with overconcentrated solution) could release volatile 

fission products

In mechanical 

processing and 

feed preparation 

operations

• Inadequate poison in solvent

• Overconcentration of solution in digester

or feed adjustment, followed by

precipitation

• Accumulation of fissile residue in

digester or in solids recycle tank

• Voloxidizer flooded with water

• Caustic added to feed

• Dissolver blockage in addition to loss of

poison

• Accumulation of chopped fuel on

undetected stuck-shut voloxidizer and

dissolver

• Criticality potential in feed tanks

• Probable release of airborne activity, including

noble gases and iodine to room atmosphere

and off-site system

• Possible severe equipment damage

• High radiation levels in rooms.

• Mass and energy probably contained in room

• Energy release on the order of 1018 fissions

Solvent 

extraction 

operations

• Excessive fissile material in high activity

centrifuge bowl

• Fissile material in unintended location

• Damage to equipment

• Pu reflux in first, second, or plutonium

purification cycle

• Pu precipitation

• Pu over-evaporation

• Fissile uranium reflux

• Fissile material in solvent tank

• High local radiation

• Release of gaseous fission products to room

or to VOG

• Damage to equipment

In miscellaneous 

systems 

• General purpose concentrator (GPC)

bottoms-caustic routed from rework

decanter to digester in mechanical

process; plutonium precipitation

• GPC bottoms routed from rework

decanter to high-level aqueous waste

concentrator

• Fissile material in high activity waste;

plutonium precipitation

• Excessive fissile material accumulation

in organic phase in rework decanter

• High local radiation levels

• Release of gaseous fission products to room

and VOG systems

• Damage to process equipment; possible loss

of process control

Loss of Coolant
4-8 

Table 4-2 Summary of possible major accidents at a spent nuclear 

fuel reprocessing facility

Accident Cause Consequence 

• Loss of power or other equipment failure

can cause a loss of coolant to maintain

temperature in a high-level waste (HLW)

tank.

• Heating of the HLW tank leads to

evaporation of solvents and radioactive

material. It can also lead to chemical

instabilities or fires.

• Release of fission products and other isotopes

from the HLW tank in the form of vapors and

aerosols

• High local radiation levels

• Potential release through off-gas systems into

the atmosphere

• Possible damage to process equipment from

fire

*This summary is obtained from [NUREG-1320] but was further condensed for this report. The accident and consequence

information originated from [DP-1558]. Additional information is also provided in the indicated references. 1

An explosion due to the decomposition products of hydrazine, hydrazoic acid (HN3) and nitrates is also possible. The formation 

of these chemicals depends on the concentration of technetium and the flow of plutonium in the PUREX process [OECD 2005]. 

Equipment that can produce red oil include evaporators (these evaporate the diluted aqueous process streams, mainly metal 

nitrates, and are used to boiling away volatile water and nitric acid components), acid concentrators (these distill the acid 

solutions), and denitrators (these are known as calciners, which are heating devices, that heat concentrated solutions of metal 

nitrates to the point of decomposition) [DNFSB_T33].

According to a recent study [BNL 2011], analytical approaches for hazard analysis have been 

developed for reprocessing facilities that include methods of hazard analysis for chemical 

plants. These methods are employed to identify, model, and quantify failure frequencies of 

individual processes or the entire facility. Information about these quantifications and models is 

described in [BNL 2011]. Table 4-3 lists potential accident initiators. 

Additional sections describe each type of accident in more detail. Since chemical explosions are 

highly energetic, these are discussed first, followed by inadvertent criticality, then fire, and finally 

spill accidents. In each of these sections, brief descriptions are provided for previous accidents 

and incidents. Finally, a discussion is provided regarding the phenomena related to the 

generation of airborne material. 
4-9 

Table 4-3 Potential accident initiators from external events and natural 

phenomena [IAEA DS360]

Accident Initiator Description 

External Event

Plane crash If the facility is located near an airport, there is a potential for an airplane to crash into 

the facility. The probability of a plane crash is greatly dependent on the footprint of the 

facility and the distance from the airport. The over-flight impact probability is much

smaller than impacts when the plane is landing or taking off.

Vehicle accidents If there is inadequate distance from the facility to roads or no means of protecting the 

facility from vehicle accidents, there would be a potential of vehicle accidents that may 

impact the facility.

Explosion at nearby 

facilities

If a nearby facility can produce an explosion that can impact the facility of interest, 

then there is an explosion concern for the facility of interest. These concerns include 

explosion fragments flying into the facility and/or significant overpressure that could 

impact the operations of the facility. For example, a blast may rupture a solution tank 

which leads to a spill of radioactive materials, fire or explosion.

Natural Phenomena

Earthquakes Without seismic qualified equipment and containers used in the facility, seismicinduced fire and explosions are possible.

Severe weather 

conditions

Wind, rain, lightning, snow, and extreme temperatures are identified as severe 

weather conditions. If the roof was not designed to withstand the design loads of 

winds, rain, and snow, disruption of the operations within the facility may be possible. 

Lightning protection design and adequate ventilation can deal with lightning strikes and 

extreme temperature changes, respectively. Temperature is very important because 

many of the chemical processes used in the fuel reprocessing facility require a certain 

range of operating temperature. Without this control, fire and/or explosions due to 

elevated temperature may be possible.

4.1 Chemical Explosions

This section presents an overview of the historical explosion accidents at reprocessing facilities 

and discusses the pertinent phenomena associated with chemical explosions in Section 4.1.1

and Section 4.1.2, respectively. 

4.1.1 History

As described in the introduction, both fires and explosions are potential accidents because of 

the corrosive and reactive chemicals used in the extraction and purification phases in fuel 

reprocessing. Both the temperature and molarity of the chemicals in the process phases must 

be controlled to prevent conditions conducive for fires and/or explosions. Table 4-4 lists

explosion accidents that have occurred throughout the history of reprocessing facility operations 

along with a description of the events that led to the explosions. As shown in Table 4-4, the 

dominant historical explosive event has been attributed to red oil. Chemical explosions can also 

result from hydroxylamine nitrate (HAN) mixtures with nitric acid. Hydroxylamine sulfate (HAS) 

has similar properties similar to HAN. Both have been used in the PUREX process in extraction 

and purification operations. An explosion of approximately 75 tons of TNT equivalent is the 

largest explosion in a reprocessing facility that has been reported (see Table 4-4). 
4-10

Table 4-4 Explosion accidents in fuel reprocessing plants

Facility Year Accident

Type Description

Hanford, WA, 

U.S. 

[DNFSB_T33]

1953 Nitration 

oxidation 

reaction – 

equipment 

failure

The accident occurred in Building 321 where a feed pump failed during 

the initial operation of a new evaporator using fresh uranium. TBP was 

inadvertently present in the feed. Despite the pump failure, the 

evaporation process continued causing normal concentrations and 

temperatures to be exceeded. A great deal of red fumes escaped 

leading to an explosion.

Savannah River 

Site, SC, U.S. 

[DNFSB_T33]

1953 Nitration 

oxidation 

reaction – 

equipment 

failure/ 

operator error

A UN* solution was being batch-concentrated in the TNX Facility to 

remove excess nitric acid from solution. The UN was in contact with 

30% TBP solvent in kerosene. Approximately 80 lbs. of TBP was

inadvertently present in the feed, causing the temperature to rise, which 

ultimately led to an explosion. 

Mayak 

Production 

Association, 

Russia [Todd 

2008]

1957 Nitration 

oxidation 

reaction

Liquid high-level waste containing sodium nitrate and acetate salts from 

the acetate precipitation process was stored in underground tanks. The 

cooling system in one of the tanks failed, and the temperature in the 

tank rose to 350 °C. The tank contents evaporated to dryness, causing 

a massive explosion equivalent to 75 tons of TNT. Over 20 million Ci of 

radioactivity was released to the environment.

La Hague, 

France 

[NUREG-1909], 

[Scheider 2001]

1970 Explosion The temperature of the dissolution reaction of the graphite fuel 

increased sharply, resulting in hydrogen generation, which led to an 

explosion and release of fission products. 5900 GBq was recorded at 

the filter of the stack, mainly due to Sb-125 (95%) and I-131 (5%) 

[Scheider 2001]

Windscale, U.K. 

[NUREG-1909]

1973 Fire/explosion An exothermic chemical reaction in a reprocessing tank.

Savannah River 

Site, SC, U.S.

[DNFSB_T33]

1975 Nitration 

oxidation 

reaction – 

equipment 

failure/ 

human error 

At the H-Canyon Outside Facilities, UN was being calcined to UO3

along with an unknown quantity of organic material. Prior to the 

explosion, dense red fumes were emitted into the denitrator room.

Hanford 

[Cadwallader 

2005]

1976 Explosion An ion exchange column suffered a chemical reaction and explosion. A 

worker was in front of the glovebox that housed the column at the time 

of the explosion. The worker was injured by glass from the glovebox’s 

shattered window, and the worker received Americium-241 

contamination as well. The americium had been collected on the ion 

exchange resin before the explosion and was dispersed throughout the 

room in the course of the explosion.

Idaho 

[Cadwallader 

2005]

1991 Explosion On February 9, 1991, a small explosion occurred at the Idaho Chemical 

Processing Plant that ruptured a 6-inch diameter borosilicate glass 

vessel during dissolving operations to retrieve 1.9-kg of highly enriched, 

unirradiated uranium fuel from a scrap fuel can of predominantly 

uranium-zirconium. The uranium was contaminated with thorium, 

plutonium, and other nuclides. The dissolver vessel was set up within a 

walk-in confinement hood but in the course of the explosion, some of 

the hot nitric acid and uranium was splattered onto three operations 

personnel in the area. They were not injured by the acid or by glass 

shards but the three operators and one emergency response worker 

were alpha contaminated; the contamination was not significant. The 

second dissolver, next to the failed unit, was intact and there was no 

damage to other equipment. Evidence suggested that a reaction of 

epsilon phase U-Zr3 alloy and nitric acid occurred.
4-11

Table 4-4 Explosion accidents in fuel reprocessing plants

Facility Year Accident

Type Description

Mayak 

Production 

Association, 

Russia 

[NUREG-1909] 

1993 Fire/explosion A tank containing a solution of paraffin hydrocarbon and TBP exploded. 

The explosion was strong enough to knock down walls on two floors of 

the facility and caused a fire. According to [IAEA 1998], the accident 

caused damage to both the reprocessing line and the building. This

resulted in the release of about 30 TBq (810 Ci) of beta and gammaemitting radionuclides and about 6 GBq (0.162 Ci) of 239Pu. 

Tokai, Japan 

[Todd 2008]

1997 Fire 

/explosion – 

operator error

A fire occurred in the bitumen waste facility of the reprocessing plant. 

Bitumen is used to solidify intermediate-level liquid radioactive waste. 

The fire apparently occurred after errors were made in the monitoring of 

a chemical reaction. The fire was not completely extinguished and 

about ten hours after the chemicals had accumulated, an explosion 

occurred which failed the confinement of the facility.

Hanford, WA, 

U.S. [Todd 

2008] 

[DOE/EH-0555]

1997 Explosion –

operator error

A storage tank containing HAN and nitric acid was allowed to evaporate 

to dryness. The resulting explosion destroyed the tank and blew a hole 

in the roof of the building. 

*UN is uranyl nitrate

The formation of a nitration oxidation reaction can be explained as follows [IRSN 2008]. The 

organic solution of tributyl phosphate (TBP), diluents (such as kerosene, n-dodecane, 

hydrogenated tetra-propylene), and the products of their degradation by hydrolysis and 

radiolysis can react violently with oxidizing products. At high temperatures, these solvents in 

the presence of nitric acid or heavy metal nitrates (such as plutonium nitrate and uranyl nitrate) 

can lead to the formation of what has been traditionally called red oil. Red oil is a substance of 

non-specific composition formed when an organic such as TBP and kerosene comes in contact 

with concentrated nitric acid and is heated above 120 °C under reflux. The red oil is the result 

of formation of nitrated organics. In the PUREX process, this red oil can be formed in 

evaporators that concentrate aqueous flows containing nitrates and traces of organic solvents. 

These flow streams include (a) “inter-cycle” concentrating of uranyl nitrate solutions; (b) 

concentrating fission products; (c) acid recuperation, where the raffinates from various 

extraction cycles are concentrated; and (d) effluent stream flow after the oxalate precipitation of 

plutonium. Red oil formation requires prolonged heating and involves complex processes such 

as radiolysis, hydrolysis, oxidation, and nitration reactions. In addition, red oil formation 

depends not only on the level of nitrates and temperature of the medium, but also on the level of 

radiolytic degradation of the solvent.

In French reprocessing facilities, the risk of a nitration oxidation reaction is mainly associated 

with the following evaporator operations [IRSN 2008]:

• Inter-cycle concentration of re-extraction solutions of uranium from the first extraction

cycle. The first extraction cycle is composed of three stages: co-extraction of uranium

and plutonium, followed by selective partitioning of plutonium, and then uranium

partitioning before spent solvent treatment.

• Concentration of fission product solutions from the first extraction cycle.

• Recovery of tritiated and non-tritiated acids where aqueous acid solutions of low and

medium activity are concentrated.
4-12

• Treatment of the OML before recycling the upstream phase of oxalic plutonium

precipitation.

• Treatment of organic effluents by distillation.

Based on the lessons-learned from nitration oxidation reactions at DOE facilities, the following 

controls have been identified to mitigate or prevent a nitration oxidation reaction [DNFSB_T33]:

1. Maintaining a temperature of less than 130 ̊C to prevent explosion conditions. This

means adequate cooling is required.

2. Process equipment/vessel design that includes sufficient venting to prevent an

overpressure event from destroying the equipment/vessel.

3. Decanters or other liquid-liquid separation equipment that may be used to remove

excess TBP from feed streams entering the heated process equipment (such as

evaporators and denitrators [NUREG-1909]).

4. Limiting the process equipment/vessel size and organic concentration to to restrict

explosive energy to be low enough to mitigate this type of accident.

5. Limiting the nitric acid concentration to less than 10 molar.

4.1.2 General Discussion

This section provides a general discussion of the phenomenological effects associated with an 

explosion. An explosion event in a fuel reprocessing facility begins with an increase in mass or 

energy that produces a stress load upon the equipment barrier. If the load is sufficient, a failure 

of the process equipment barrier can occur. The pressurization event could result from

chemical reactions due to equipment or control system failures, temperature regulation failure, 

relief valve failures, etc. 

Either as a consequence of the vessel depressurizing or the kinetics of the energetic event, a 

pressure pulse is generated and imparts a force on the constituents of the reprocessing 

component such as the liquid or solid contents. The force imparted by the pressure wave can 

overcome liquid adhesion and gravitational forces, effectively separating the fluid body into 

suspended droplets. Coupled with a failure of a vessel, a portion of the newly suspended 

material may partition to the vapor space of the surrounding atmosphere. 

The initially suspended mass is a distribution of various droplet sizes. The size distribution is 

dependent on the energy of the event, the density of the susceptible material, evaporation from 

the material, etc. Proper characterization of the droplet size distribution permits meaningful 

determination of effective removal processes that would reduce the ultimate airborne material

released to the environment. 

A threshold aerosol diameter may be applied to simplify the calculation space since gravitational

settling is the overwhelming deposition mechanism for larger particulates. Particles with an

aerodynamic diameter of 50 micrometers or less can be categorized as an aerosol. Aerosol 

deposition mechanisms such as diffusiophoresis, Brownian motion, and thermophoresis 

become relevant deposition processes much below this 50 micrometer threshold. Effects on 

larger droplets such as evaporation, impacting and splattering are phenomena that may further 

increase the fraction of airborne material considered to be aerosolized. 
4-13

In addition to the deposition mechanics, aerosols interact directly with one another. Again, 

based on the size of the aerosol, various agglomeration mechanisms occur at different rates. 

Agglomeration due to the following mechanisms must be considered: Brownian motion, 

gravitational settling, and shear and inertial turbulence. 

Similar to the release of liquids, vapors within the process equipment may be available for 

transport as a consequence of the equipment failure. In addition to any direct release of vapor, 

evaporation from the surface of the remaining fluid within the vessel, content spilled from the 

vessel, and airborne products will each contribute to the generation of vapor species resulting

from an explosion. Condensation of vapor species can reduce overall availability to contribute 

to the environmental source term. Vapor condensation onto structures, pool surfaces, and 

aerosols present may occur, each of which affects the total airborne material.

An explosion may damage more than the equipment or space where the event originated. Any 

remaining energy from the explosion (in the form of a pressure pulse) is imparted to the 

surrounding walls and ceiling. Once the explosion occurs in a vessel or container, the 

overpressure may further damage the surrounding walls and ceiling in a room. The propagation 

of the explosion greatly depends on the geometry of the room and the connecting rooms, and 

the initial explosion location. A determination of damage to the subsequent connecting rooms 

and near field equipment could indicate possible cascading failures resulting from the initiating 

explosion.

Typical blast overpressures for a given explosion are given in Table 4-5. This table provides the 

differential pressure (or overpressure) that may be required to cause damage to common 

structures. Typically, the retaining walls used in the fuel reprocessing facilities are thick to 

shield intensive radiation emanating from the radioactive spent fuel materials. The shielding 

thickness can be as high as several feet of concrete with rebar reinforcement and may include a 

steel liner. The ceiling of the fuel reprocessing facilities, particularly in the area of the chemical 

processing areas, may be constructed to accommodate the ventilation and other piping systems 

above it. Thus, the ceiling could possibly withstand overpressure but may not be as strong as 

surrounding walls made from reinforced concrete.

A more detailed discussion on the blast overpressure and the failure models is presented in 

Section 5.2.1.4. 
4-14

Table 4-5 Blast overpressure on structural damage [Merrifield 1990]

Pressure 

Differential Damage

(psig) (kPa)

0.15 102.359 Typical pressure for glass failure

0.4 104.083 Limited minor structural damage

0.5-1 104.772-

108.220

Large and small windows usually shattered, occasional damage to window 

frames

0.9 107.530 Roof damage to oil storage tanks

1.3 110.288 Steel frame of clad buildings slightly distorted

1.5 111.667 Slight damage to window frames and doors

2-2.5 115.114-

118.562 Some frame distortion of steel framed buildings

2-3 115.114-

122.009

Concrete or cinder brick walls 8” (20.32 cm) to 12” (30.48 cm) thick, not 

reinforced shattered

3-4 122.009-

128.904 Rupture of oil storage tanks

4 128.904 Cladding of light industrial buildings ruptured

4.5 132.351 Severe distortion to frames of steel girder framed buildings

7-9 149.588-

163.378 Collapse of steel girder famed buildings

8-10 156.483-

170.272 Brick walls completely demolished

>10 >170.272 Complete destruction of all unreinforced buildings

13 190.957 18” (45.72 cm) brick walls completely destroyed

280 2031.857 Lip of crater

4.2 Inadvertent Criticalities

This section presents an overview of the historical criticality accidents at reprocessing facilities

and discusses the generalities of criticality accidents in Section 4.2.1 and Section 4.2.2, 

respectively.

Much of the information presented in this section is adapted from LA-13638 [McLaughlin 2000]. 

Similar criticality accidents are discussed in [Todd 2008]. 

4.2.1 History

As of the year 2000, there have been a total of 22 major process-related criticality incidents in 

the world [McLaughlin 2000]. Figure 4-1 shows the chronology of these 22 accidents These 

accidents can be summarized as follows:

• 21 of 22 excursions occurred in solutions or slurries.

• 18 accidents occurred in manned and unshielded facilities.

• Nine fatalities resulted from these accidents.
4-15

• No equipment was damaged in these accidents.

• One accident resulted in measurable fission product contamination beyond the plant

boundary and in measurable exposures to the public at Tokai Fuel Fabrication Plant.

Figure 4-1 Chronology of process criticality accidents. Plot shows number and 

location of accidents by year from 1945-2000. [McLaughlin 2000]

Among these incidents, one is related to metal ingots, which is not included in further

discussions. Additionally, not included in further discussions are process accidents related to 

enrichment and uranium hexafluoride such as: 

1. Siberian Chemical Combine, July 14, 1961 – Accident occurred in a gaseous diffusion

uranium enrichment facility.

2. Electrostal Machine Building Plant, November 3, 1965 – Accident related to conversion

of the uranium hexafluoride to uranium oxide.

3. Siberian Chemical Combine, December 13, 1978 – Accident related to plutonium metal

ingots in a storage container.

4. Novosibirsk Chemical Concentration Plant, May 15, 1997 – Accident related to chemical

etching process in the fuel rod production.

Table 4-6 provides the remaining criticality accidents that can be related to fuel reprocessing 

facilities. As shown in this table, the location of the accident, cause, accident description, and 

total fissions of the accidents are identified.
4-16

Table 4-6 A summary of solution criticality incidents [McLaughlin 2000]

(#)Accident Operation 

[Cause] Description Total Fissions

(1)

3/15/1953, Mayak 

Production

Association, Russia

Plutonium 

purification 

[operator 

error/procedural 

error]

There were 18 vessels containing plutonium 

nitrate solution in a staging area where 

operations of mixing, dilution, volume 

measurement, sampling, and interim storage 

were done. Each vessel had a 500 g plutonium 

mass limit. An excursion occurred in Vessel 18 

(originally contained 26 liters), which contained 

848±45 g of plutonium in 31 liters at the time of 

the accident. The cause of this excursion was 

due to unrecorded transfer of 5 liters of solution 

from Vessel 1 to Vessel 18.

2×1017 fissions. This 

was based on a 

temperature rise of 

60  ̊C in 31 liters of 

the solution.

(2)

4/21/1957, Mayak 

Production 

Association, Russia

Uranium 

purification

[equipment 

failure – defect 

in the filter 

fabric]

Impure uranyl nitrate (90% enriched uranium) 

and oxalic acid solution were mixed in a vessel 

to precipitate the uranyl oxalate trihydrate, which 

was done within a glovebox. This precipitate 

slurry was then vacuumed to a holding tank from 

which it was drained into a filter vessel. The 

precipitate containing uranium was collected on 

the filter fabric, and the filtrate was pulled 

through by vacuum and collected in a filtrate 

receiving vessel. An excursion occurred in this 

vessel that caused the filter vessel to bulge 

upward, followed by a violent release of gas and 

ejection of some of the precipitate out of the filter 

vessel and onto the glovebox floor. 

A fission yield of 

1×1017 fissions. This 

yield was based on 

the measurement in 

the room and the 

concentration of 24Na 

in the operator’s 

blood (17 hours after 

the accident, the 

dose was 245 

Bq/cc).

(3) 

1/2/1958, Mayak, 

Production 

Association, Russia

Solution critical 

experiment 

[operator error]

Based on the 4/21/1957 accident, a solution 

criticality experiment was conducted using highly 

enriched uranyl nitrate solution (90% enriched). 

The experiment was conducted in a cylindrical 

vessel that could hold an excess of 400 liters of 

solution to measure a wide range of critical 

states. Typically after each experiment, the 

solution must be drained to a smaller vessel 

which was criticality-safe. Experimenters 

decided one day to circumvent the routine by 

manually pouring the solution out of the 

experiment vessel. This routine created an 

excursion because human bodies provided 

additional reflection of the neutrons, which 

created a criticality-unsafe configuration.

Based on the fission 

product activity in the 

solution, a single 

pulse event was 

about 2×1017

fissions.

(4) 

6/16/1958, Y-12 Plant, 

Oak Ridge, TN, U.S.

Uranium 

recovery 

[equipment 

failure]

Solid wastes with highly enriched uranium (93% 

enriched) were dissolved in nitric acid, purified, 

concentrated, and then converted to uranium 

tetrafluoride. A 55-gallon drum served as a 

residue collection tank. Because there were

apparent leaks in the system, liquid accumulated 

in the drum to about ~56 liters with 2.1 kg of235U 

mass. This condition yielded excursions that 

lasted about 20 minutes.

Based on the 

instrument at the site 

and detectors 

nearby, the first 

spike was about 

6×1016 fissions, with 

a total of 1.3×1018

fissions.

(5) 

12/30/1958, Los 

Alamos, NM, U.S.

Plutonium 

recovery 

[equipment 

design]

An organic treatment tank, which typically 

contained solutions < 0.1 g Pu/liter and traces of 

americium. This tank was made of stainless 

steel and had a volume of 1000 liters. Before 

the accident, the tank contained an aqueous 

layer of 330 liters with 60 g of Pu, and an organic 

layer of 160 liters with 3.1 kg of Pu. The 

Based on the 

reactivity of the 

system, the 

excursion yielded 

1.5×1017 fissions.
4-17

Table 4-6 A summary of solution criticality incidents [McLaughlin 2000]

(#)Accident Operation 

[Cause] Description Total Fissions

aqueous layer was located at the bottom of the 

tank. This configuration was subcritical. When 

the stirrer was started, the motion altered the 

reactivity of the tank, and an excursion occurred 

in about 3 seconds. The excursion was 

terminated by the stirring action. After this 

accident, a criticality-safe tank was used.

(6) 

10/16/1959, Idaho 

Chemical Processing 

Plant, ID, U.S.

Uranium 

extraction 

[operator error, 

equipment 

failure]

The impurity extraction process occurs in three 

pulse columns. Intermediate between the first 

and second cycle extraction, the uranyl nitrate 

solution (91% enriched uranium) was stored in 

two banks of long pipes (referred to as pencil 

tanks). There was a line leading from the 

interconnected pipes to the 18,900 liter waste 

receiving tank. As required, operators needed to 

conduct analysis by collecting samples using a

sparging operation. Because the air sparging 

was so forceful, and the pressure gauge of the 

other pipe was functioning improperly, the liquid 

in the pipes was forced to the top of the loop that 

drained into the receiving tank. Excursions 

boiled away half of the 800 liter solution volume 

that terminated the criticality.

Based on the strip 

chart recordings 

from the continuous 

air monitors at 

various locations, the 

initial spike of at 

least 1017 fissions, 

followed by boiling of 

15 to 20 minutes to 

yield a total of 

4×1019 fissions.

(7) 

12/5/1960, Mayak 

Production 

Association, Russia

Plutonium 

recovery 

[operator error]

The accident occurred in the waste solution 

process to recover plutonium. A total of 4 

purification vessels were used in the process. 

All vessels were contained in glove boxes. A 

holding tank was used to keep solutions for 

various stages of the purification process. 

Because of overloading, the mass limit in the 

initial purification process was passed, and 

subsequent purification process resulted in a 

criticality accident. Excursions occurred in the 

holding tank, which contained 894 g of 

plutonium. 180 g of Pu was flushed out of the 

holding tank during the accident.

The excursions 

yielded a total of 

2.5×1017 fissions.

(8) 

1/25/1961, Idaho 

Chemical Processing 

Plant, ID, U.S.

Uranium 

purification

[operator error, 

equipment 

design]

The accident occurred in the upper 

disengagement head of the H-110 product 

evaporator. Because the disengagement head 

was not a favorable geometry, any solution that 

reached this section may yield criticality. Despite 

an overflow line to prevent a significant amount 

of solution reaching this section, 200 g of uranyl 

nitrate solution with 90% enrichment per liter was 

apparently rapidly ejected up into this section 

causing excursions. This action was due to 

operators attempting to clear the plugged line 

with air. The cause of the ejection of the solution 

upward was due to the high air pressure which 

forced the solutions in the favorable geometry 

section to move upward into the disengagement 

head.

Because of no 

instrument readout 

of the event, it was 

calculated to be 

6×1017 fissions.

(9) 7/14/1962, Siberian 

Chemical Combine, 

Russia

Uranium 

enrichment plant 

[equipment 

The accident was caused by the accumulation of 

high concentration of UF6 in the vacuum pump 

oil reservoir. Four days before the accident, 

there was a sudden flow rate increase of UF6

The total number of 

fissions for both 

excursions was 

estimated by fission 
4-18

Table 4-6 A summary of solution criticality incidents [McLaughlin 2000]

(#)Accident Operation 

[Cause] Description Total Fissions

failure and 

operator error]

into the reservoir The operator turned on the 

vacuum pump and approached the equipment 

(~0.5 m) to open a valve between the pump and 

holding vessels. This resulted in a criticality

event. 

product analysis to 

be about 1.2 × 1015.

(9) 

4/7/1962, Hanford 

Works, U.S.

Plutonium 

recovery 

[operator error]

The transfer tank with a capacity of 69 liters 

received solution containing 45 g Pu per liter and 

later additions of other diluted solutions which 

caused an excursion that lasted for 37.5 hours. 

The shutdown mechanism for the accident was

due to a central pipe that entered the bottom of 

the vessel. Fission reactions may have driven 

off CCl4 and converted most of the remaining 

organic to dibutyl phosphate. This phosphate, 

with extracted plutonium, may have gone to the 

bottom of the vessel and into the pipe to stop the 

excursion.

An initial spike 

estimated to be no 

more than 1016

fissions, and lasted 

for 37.5 hours to 

total fissions of 

8×1017.

(10) 

9/7/1962, Mayak 

Production 

Association, Russia

Plutonium 

recovery

[operator error 

and equipment 

design]

Dissolution vessels were used to dissolve solid 

Pu residue to further purify plutonium. Due to 

the unfavorable geometry of the vessels and 

undetermined plutonium concentration, 

excursions occurred in the vessels.

A total of three 

excursions were 

detected and 

estimated to have 

total of 2×1017

fissions.

(11) 

1/30/1963, Siberian 

chemical Combine 

(Tomsk), Russia

Uranium 

recovery

[operator error]

The recovery process required that dry 

precipitate of uranium (90% enriched) be placed 

into concentrated nitric acid for the dissolution 

process. The criticality control was based on the 

total fissile mass and fissile concentration. 

Because the operator recorded the concentration 

of the uranium precipitate incorrectly, this action 

led to the first excursion in a large diameter 

collection vessel. Then for the next 10 hours, 

eight additional excursions occurred.

The total fissions 

was estimated to be 

7.9×1017.

(12) 

12/2/1963, Siberian 

Chemical Combine 

(Tomsk), Russia

Uranium 

purification

[equipment 

design]

A number of process vessels containing both 

aqueous and organic solutions of uranium (90% 

enriched) were used for the purification process. 

In each vessel, a level indicator showed the level 

of the solution. However, this indicator worked 

well with aqueous solutions, but not with organic 

solutions. Since the organic solution is less 

dense than an aqueous solution, the organic 

solution was transferred to other vessel via the 

vacuum system. Over time, a final vessel 

accumulated excessive organic solution which 

led to excursions, which lasted for hours. 

Based on the 140La 

fission product 

analysis, the total 

fissions was about 

6×1016.

(13) 

7/24/1964, United 

Nuclear Fuel Recovery 

Plant (Wood River), 

U.S.

Uranium 

recovery

[operator error]

An operator mistook a highly concentrated uranyl 

(93% enriched uranium) nitrate solution as 

trichloroethane as the washing fluid for the 

solvent extraction columns. The operator pulled 

this solution into a vessel containing a large 

quantity (41 liters) of sodium carbonate solution. 

A series of excursions occurred after the 

combined solution (about 51 liters) was stirred 

automatically.

The first excursion 

was estimated as 

1.0×1017 to 1.1×1017, 

with 20% of the 

combined solution 

ejected. The total 

fissions for the entire 

event were about 

1.30×1017 to 

1.55×1017.
4-19

Table 4-6 A summary of solution criticality incidents [McLaughlin 2000]

(#)Accident Operation 

[Cause] Description Total Fissions

(14) 

12/16/1965, Mayak 

Production 

Association, Russia

Uranium 

recovery 

[equipment 

design]

The residues (90% enriched uranium) being 

recovered were produced from dissolution, 

precipitation, and reduction processes. The 

accident occurred in a dissolution vessel or 

holding vessel. A total of 11 excursions 

occurred. The cause of the accident was 

probably due to unfavorable critical 

configurations of the vessels.

The total fissions for 

the accident were 

about 5.5×1017. 

(15) 

12/10/1968, Mayak 

Production 

Association, Russia

Plutonium 

processing 

[equipment 

design and 

operator error]

An unfavorable vessel geometry was used to 

store a plutonium organic solution. The operator 

was instructed to pour 35 liters of plutonium 

organic and aqueous solutions into a 60-liter 

vessel. A first excursion occurred with 16 liters 

containing 880 grams of plutonium ejected. The 

remaining plutonium in the vessel was about 709 

grams.

The total fissions 

was about 1.3×1017

with the first 

excursion was about 

3×1016. 

(16)

8/24/1970, Windscale 

Works, United 

Kingdom

Plutonium 

recovery

[equipment 

design]

The plutonium recovery was done using a 

dissolver, conditioner, and transfer vessel. The 

criticality accident occurred in the transfer 

vessel, where excess plutonium solution 

accumulated, and was well above the criticality 

limit. A large mass of plutonium accumulated in 

the vessel over a period of 2 years. 

The total fissions 

were about 1015. It 

lasted to less than 

10 seconds.

(17)

10/17/1978, Idaho 

Chemical Processing 

Plant, U.S.

Uranium 

extraction

[equipment 

failure and 

operator error]

A pulsed column with larger diameter regions at 

the top and bottom of the column is used in the 

uranium solvent extraction phase. Two streams 

of organics (with a density lighter than aqueous 

solutions), and the aqueous solutions were 

flowing in the column. The organic solution 

containing uranium was extracted in the column. 

The accident was due to a number of causes. 

The first cause was the leaking water valve on 

the make-up tank. The chart recorder had run 

out of paper and procedures were not followed. 

The event was caused by the accumulation of 

the solution over time (~1 month).

Total fissions were 

about 2.7×1018. 

(18)

9/30/1999, JCO Fuel 

Fabrication Plant, 

Japan

Uranium 

processing 

[operator error]

A special request to process uranyl nitrate 

solution at 18.8 % enrichment of uranium was 

done for this operation. The operation had been 

done routinely for uranium enrichment of up to 

6%. Manual pour of the solution from one vessel 

to another vessel was performed in the 

operation. To save time for dissolution, the 

operators did not follow the procedure, and 

poured the bucket of solutions to a precipitation 

vessel with an unfavorable geometry. Eventually 

excursions occurred.

Total fissions were 

about 2.5×1018. 

Based on these process accidents, the following lessons have been identified:

1. Vessels with geometries that favor criticality should be avoided, particularly when high

concentration solutions might be present.

2. No verbal procedures should be used. Instead, documented procedures should be

used.
4-20

3. Operators should be aware of any abnormal conditions of the processes.

4. Fissile material accountability must consider criticality control when tracking material

accumulations in the equipment.

5. Operators should be trained to respond to potential (based on past experience or PRA

analysis) equipment failure or human errors.

6. Operators should be trained to deal with possible off-normal conditions for the processes

involving both organic and aqueous solutions.

7. Operators should be trained to follow procedures.

8. Operators should be trained to deal with failure of hardware that is important to criticality

control.

Table 4-7 provides the single-parameter criticality safety limits that are often used to prevent 

inadvertent criticality events from happening. An inadvertent criticality event is unlikely to occur 

when one of these limits is used to control the operation. However, this is not always practical. 

For example, a spent fuel reprocessing facility that processes a large amount of uranium cannot 

be limited to the amount of fissile and fissionable materials given in Table 4-7. Therefore, 

criticality safety evaluations are often performed to determine the limits to prevent inadvertent 

criticality. 

Table 4-7 Single-parameter limits for uniform aqueous solutions of 

fissile nuclides (U-235 and Pu-239) [ANS-8.1]

Parameter Subcritical Upper Limit for Fissile Solute

235UO2F2 235UO2(NO3)2 239Pu(NO3)4

Mass of fissile nuclide, kg 0.76 0.78 0.48

Diameter of cylinder of solution, cm 13.7 14.4 15.4

Thickness of slab of solution, cm 4.4 4.9 5.5

Volume of solution, liter 5.5 6.2 7.3

Concentration of fissile nuclide, g/liter 11.6 11.6 7.3

Atomic ratio of hydrogen to fissile nuclide* 2250 2250 3630

Areal density of fissile nuclide, g/cm2 0.40 0.40 0.25

*Lower limit

4.2.2 General Discussion

Nuclear criticality depends on the volume, mass, and concentration of the fissile materials. 

Criticality is a great concern when aqueous solutions are involved because water is a very good 

neutron moderator. Potential conditions of inadvertent criticality may occur during a fire fighting 

situation where water is introduced. In all cases, the chemical process associated with 

separation and purification operations require that containers used in the process be in criticality 

safe configurations. However, accidents initiated by corrosion of equipment, which can lead to 

leakage, may pose a criticality concern. 
4-21

Many short-lived fission products are generated when a nuclear excursion occurs. The fissile 

materials commonly found in commercial spent nuclear fuel include 233U, 235U and 239Pu. 

Inadvertent criticality may occur with sufficient mass, high fissile-to-moderator ratio, and certain 

geometrical configuration. Good moderators contain hydrogen, such as water and 

hydrocarbons. Once criticality occurs, the measure of the fission energy is often referenced by

the number of fissions. Typically, a fission produces about 200 MeV (3.204×10-11 joules). An 

excursion of 1018 fissions in 1 second would produce 32 MW of thermal power. In addition, the 

fission products produced by this number of fissions may be significant if the duration of the 

excursion lasts for a significant amount of time (see Table 4-6 for the duration of the excursions 

in past criticality accidents). A representative fission value and time can be input into a code

such as ORIGEN, to obtain the fission product inventory. For a spent fuel solution criticality 

accident with a total of 1019 fissions, the anticipated quantities and radionuclides generated are 

shown in Table 4-8. As shown in this table, the quantities of the noble gases and iodine fission 

products in terms of activity are given from 0 to 8 hours following the excursion.

Table 4-8 Activity of important gaseous and volatile radionuclides

released during a 1019 fission excursion in spent fuel 

solution* 

Nuclide Half-Life Radioactivity, Ci1

0 to 0.5 hr 0.5 to 8 hr Total

Kr-83m 1.8 hours 3.7E+00 3.3E+01 3.7E+01

Kr-85m 4.5 years 1.6E+01 1.5E+02 1.7E+02

Kr085 1.7 years 1.5E-04 1.4E-03 1.6E-03

Kr-87 76.3 minutes 1.0E+02 9.0E+02 1.0E+03

Kr-88 2.8 hours 6.5E+01 5.9E+02 6.6E+02

Kr-89 3.2 minutes 4.1E+03 3.7E+04 4.1E+04

Xe-131m 11.9 days 3.8E-04 3.5E-03 3.9E-03

Xe-133m 2.0 days 5.5E-02 1.9E+00 2.2E+00

Xe-133 5.2 days 1.3E+00 2.3E+01 2.7E+01

Xe-135m 15.6 minutes 1.1E+01 2.8E+03 3.3E+03

Xe-135 9.1 hours 1.6E+01 3.5E+02 4.1E+04

Xe-137 3.8 minutes 3.8E+03 4.2E+04 4.9E+04

Xe-138 14.2 minutes 1.2E+03 9.5E+03 1.1E+04

I-129 2x107 years 3.9E-10 4.3E-10 4.3E-10

I-131 8.1 days 1.8E-01 1.6E+00 1.8E+00

I-132 2.3 hours 6.7E-01 6.1E+00 6.7+00

I-133 0.8 hour 3.5E+00 3.1E+01 3.5E+01

I-134 52.6 minutes 4.8E+01 4.3E+02 4.8E+02

I-135 6.6 hours 1.2E+01 1.0E+02 1.2E+02

*Spent fuel contains < 5% U-235 enrichment [DOE 1994]

1

Total Ci. The assumption of cumulative yield is very conservative (i.e., it does not consider appropriate decay schemes). 
4-22

4.3 Fire

This section presents an overview of the historical fire accidents at reprocessing facilities and 

discusses the pertinent phenomena associated with aerosol generation and transport due to 

fires in Section 4.3.1 and Section 4.3.2, respectively.

4.3.1 History

A fire in a fuel reprocessing facility is possible due to physical heating and/or heating from 

chemical reactions. Table 4-9 lists fire accidents that have occurred in reprocessing facilities. 

In most cases, a fire also occurs in conjunction with an explosion. 

Table 4-9 A summary of fire incidents at the fuel reprocessing facilities

Facility Year

Accident

Type Description

Hanford

[Cadwallader 

2005]

1951 Fire –

human 

error

A laboratory room suffered a fire. Apparently, nitric acid on rags used for 

cleaning and decontaminating parts had been boxed for disposal. The 

rags spontaneously combusted in the waste storage room of a small 

laboratory building. The security patrol noted smoke from the ventilation 

stack after the normal work day and heard crackling noises over the 

night alarm public address system. Firefighters arrived and used water to 

extinguish the fire. The fire burned for about four hours.

Hanford

[Cadwallader 

2005]

1963 Fire –

equipment 

failure

In the PUREX plant, resin in a plutonium extraction column caught fire. 

The fire spread plutonium contamination throughout the building and 

some outside of the building.

La Hague 

[Scheider 

2001]

1981 Fire –

equipment 

failure

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 beginning of the fire. Graphite 

elements had been burning for 24 hours in a waste silo.

4.3.2 General Discussion

A fire starts with ignition of a combustible material (solid or liquid) in a room. The combustible 

material reacts with the available oxygen and releases heat and combustion products (including 

soot). The heated combustion products rise as a buoyant plume above the fire, entraining air as 

the plume rises to form a hot gas layer of the combustion products and fresh air under the 

ceiling. As the hot air rises, cooler air fills in to provide oxygen to sustain the fire. Eventually, 

this layer grows by pushing air out of any opening that connects with the room. Any gases and 

combustion products that leave the room containing the fire forms a hot gas layer near the 

ceiling of a connecting room. Although this room may not have a fire source, a layer grows

downward from the ceiling as more gas flows into the room from the fire.

The basic concept and model of a fire in a room and fluid exchange between connecting rooms 

have been documented [Tanaka 1983, Ho 1988, and Beall 1997].

In terms of modeling an accident involving a fire, the important phenomena include:

1. Heat released from the fire is directly related to the amount of combustibles available to

burn and available oxygen in the room. Both energy and duration are important

parameters to determine the extent of the accident.
4-23

2. Thermal mass in the room and surrounding rooms are the major heat sinks for the fire.

The heat fluxes can be calculated from the fire to the walls, ceiling and floor and other

objects in the room. (No credit is usually given for sprinklers.)

3. Hot gas layer build-up near the ceiling is an important phenomenon in a fire. The

stratification behavior captures the thermal plume from the fire. The layer is dynamic in

terms of the interaction with the plume, incoming cooler flow from doors and ventilation

system (mixing). As the hot gas layer grows, the temperature of the room fire increases,

which may initiate or induce a secondary accident such as another fire and/or explosion.

4. Combustion products are important because they influence the transport of

radionuclides from processing solution to other locations within the facility.

4.4 Spills and Leaks

This section presents an overview of historical spill accidents at reprocessing facilities and 

discusses the pertinent phenomena associated with aerosol generation and transport due to 

spills in Section 4.4.1 and Section 4.4.2, respectively. This section, similar to the overall report, 

specifically focuses on potential events that lead to creation of vapors and aerosols that could 

be released into the atmosphere. Spills and leaks can lead to groundwater contamination as 

well, but that topic is not discussed here. 

4.4.1 History

Spill accidents in a fuel reprocessing facility are mainly caused by equipment, piping, or 

container failure. The failures are often caused by a combination of corrosion and pressure 

transients. Corrosion results from the reaction of strong acid solutions with vessel and piping 

components. The primary cause of pressurization is chemical reactions inside the process 

vessel or holding tank. As given in Table 4-10, eleven spill accidents have been documented.
4-24

Table 4-10 Spill accidents in fuel reprocessing plants

Facility Year

Accident

Type Description

Hanford 

[Cadwallader 

2005]

1953 Pressurized 

spill – 

operator 

error

During a process evolution, an acid solution and a caustic solution were 

improperly mixed. Metal waste supernatant was being pumped from a 

holding tank to a blend tank to prepare process feed for the TBP. 

Approximately 2,800 gallons of 60% nitric acid was first added to the 

blending tank. The metal waste addition began, and the agitator for the 

tank was started. After about 100 minutes, a supervisor noted yellowish 

ooze coming from the tank cover blocks. As he rushed to the controls to 

stop the supernatant flow, a geyser of liquid quickly rose about 10 meters 

in the air for about 30 seconds, and the prevailing wind carried the geyser 

liquid onto the supervisor and other nearby workers. They were 

contaminated at ~4,000 counts per minute (cpm) and they had stinging 

sensations from the liquid. The supervisor called to shut the roadway that 

had been contaminated by the geyser. Investigation revealed that the 

operator had activated the agitator switch but the agitator had “bumped;” 

it started to operate but did not continued to run. The operator had not 

paused to verify that the agitator was continuing to run. Trough areas 

near the blend tank contained up to about 6-inch deep yellowish liquid 

that read a 35 rem/hour at a distance of 6-inches.

Hanford 

[Cadwallader 

2005]

1956 Spill –

equipment 

failure

At the REDOX plant, two process solution (highly concentrated acidic 

plutonium product) spills were found on the floor behind the control panel. 

The solution entered the control room by an instrument air line. The 

instrument airline operated a flow control valve in the process transfer line 

from the final product concentrator tank to a receiver vessel. When air 

pressure was relieved from the flow control valve, the air was intended to 

bleed harmlessly into the control room. The flow control valve stem had a 

bellows to protect it, but the bellows had failed and process solution had, 

over time, collected in the valve superstructure. The bleed hole in the 

valve superstructure still had the manufacturer’s dust plug in place. Over 

time, with no other pathway, the process solution backed up into the air 

line and out of the vent in the control room. The operator who noted the 

survey meter “break down” was contaminated on all skin surfaces to over 

667 Bq. Nasal smears read 500 Bq and all of his clothing was 

contaminated. A rough estimate was that only about 350 cm3 was spilled 

in the control room, and the clean-up cost was $2,000 - $2,500.

Savannah River 

Site 

[Cadwallader 

2005] 

1960 Spill –

equipment 

failure

A valve corridor was highly contaminated by leaking coolant water. The 

coolant water had become contaminated by high activity waste from the 

PUREX process, and then the water had leaked through a defective 

waste evaporator reboiler. Approximately 5,000 Curies were released 

with the water. Most of the released radioactivity was contained in the 

building and was flushed to the waste handling facilities. Minor amounts 

of liquid were released to 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 corridor was flooded with water, and the 

contaminated water flowed down a stairwell and spread widely 

throughout normally nonradioactive sections of the building. Radiation 

readings as high as 400 rad/hour at 30 cm were measured after the 

incident.
4-25

Table 4-10 Spill accidents in fuel reprocessing plants

Facility Year

Accident

Type Description

Hanford 

[Cadwallader 

2005]

1961 Spill –

equipment 

failure

At Hanford Works in the PUREX plant, the acid waste concentrator used 

a heat exchanger to help condense acid. The heat exchanger had to be 

taken out of service after only 18 months of operation. The failure was 

caused by improperly placed weld bead at several tube to tube sheet 

welds. The welder had laid bead by hand and apparently had deviated by 

as much as 50 mils (1.27 mm). The tube wall thickness was affected by 

the weld placement and this led to premature failure of the tube bundle. 

The investigators also noted that some tubes had been plugged with solid 

material and other tubes had solid films, which would have reduced the 

effective heat transfer.

Savannah River 

Site [DOE/EH0555]

1972 Pressurized 

spill – 

chemical 

reaction

A pressure surge in a waste evaporator being brought up to operating 

temperature caused the eructation of approximately 6,000 pounds of 

process solutions from the vessel. The solution was a dilute mixer-settler 

flush of 1 M nitric acid and 0.1 M HAN at the time of accident.

Savannah River 

Site [DOE/EH0555]

1978 Pressurized 

spill – 

chemical 

reaction

A sudden eructation occurred from a tank thought to be empty after 

addition of 25% (~2,000 pounds) nitric acid. However, the tank had 5% 

HAS in the heel of the tank. The eructation sprayed 3 individuals.

Savannah River 

Site [DOE/EH0555]

1980 Pressurized 

spill – 

chemical 

reaction

Dilute HAS/nitric acid solution in a tank was inadvertently heated over a 

period of a few days due to a leaky steam coil. The heating allowed the 

autocatalytic condition to be reached, causing the release of acid solution 

to the surrounding of the tank. A ruptured elbow, bent piping and ejection 

of the anchor bolts are the result of this accident.

Hanford 

[DOE/EH-0555]

1989 Pressurized 

spill – 

chemical 

reaction

An exothermic reaction involving a solution of HAN, nitric acid and 

hydrazine occurred in a one-inch chemical makeup line in the PUREX 

plant aqueous makeup area. The chemical reaction pressurized the 

isolated piping and blew out a gasket at the flange connection.

Savannah River 

Site [DOE/EH0555]

1996 Pressurized 

spill – 

chemical 

reaction

The temperature of a tank containing HAN in strong nitric acid was not 

monitored and rose to 40  ̊C where an autocatalytic event resulted in an 

eructation of 250 gallons.

THORP 

[DOE2005-11]

1998 Spill –

equipment 

failure

Erosion of an outlet pipe in the dissolver cell resulted in the leak of highly 

radioactive dissolver solution into the secondary containment. Despite 

sump-level indication, sampling and contaminated radiological probes, 

this leak went unnoticed for years.

THORP 

[DOE2005-11], 

[Todd 2008]

2005 Spill –

equipment 

failure

A pipe failure resulted in about 83,000 liters of highly radioactive dissolver 

solution leaking into the stainless-steel lined feed clarification of the 

THORP facility in 2005. This solution contained about 19 metric ton of 

uranium and plutonium. The leak went undetected for months (possibly 

since July 2004), before being discovered. 

Since complex chemical processes are used in highly radioactive spent fuel reprocessing, the 

chemical and radiation conditions are such that corrosion of equipment is a major concern in a 

spill accident. The corrosion problems could occur in (1) dissolvers where hot or boiling nitric 

acid is used to dissolve the spent fuel; (2) solvent extraction vessels for the separation 

processes of uranium, plutonium and fission products; (3) evaporators to concentrate these 

materials in hot nitric acid; and (4) storage tanks for radioactive waste of warm concentrated 

nitric acid [Mudali 1993]. To prevent and mitigate these types of accidents, frequent visual 

inspection of the process equipment is necessary.
4-26

4.4.2 General Discussion

Spill accidents are the least energetic event compared with inadvertent criticality events, fires, 

and explosions. Often a spill may not be detected immediately. Impaction of liquid against a 

solid surface may create aerosols and air flow may entrain some of the material in a pool 

created by the spill. Many aerosol models for spill accidents focus on air flow parallel to the 

solution pool to determine the amount entrained from the pool surface. A more energetic spill is 

based on a free fall concept from a few feet to 10 feet. Break-up phenomena account for the 

aerosol generation/suspension when the liquid hits the ground. 
5-1 

5. ACCIDENT AND AEROSOL GENERATION MODELING

In previous sections of this report, chemical separation processes and anticipated accidents 

(based on historical experience) are described to provide the context for identifying models and 

methodologies needed to characterize a source term. Characterization of a source term 

requires understanding the evolution of hazardous materials from the initial accidental release, 

through transport within the facility, and finally release to the environment. 

To properly characterize a source term, a sequential process is presented in this chapter. First, 

a determination of the available inventory and form of the material associated with a process or 

system within a reprocessing facility is discussed in Section 5.1. Then the more likely accidents 

are discussed in Section 5.2. For each accident the current techniques for estimating the 

released aerosol are given. Current techniques are primarily empirical and therefore limited, and 

may not be appropriate for reprocessing facilities. Therefore, Section 5.3 describes a basic 

principles approach that is proposed to replace the empirical approach for some phenomena. 

The transport of material through the facility is not discussed. 

5.1 Inventory in Operations

This section describes the spent fuel inventory during each step of the operations in a typical 

fuel reprocessing facility. This inventory greatly depends on the initial spent fuel received at the 

facility. From the operational history and cool down period of a given fuel, a determination of 

the inventory can be made to characterize the composition of the fuel, i.e., fission products and 

transuranics present. Fuel characterization is readily performed using ORIGEN or a similar 

code. As an example, the spent fuel presented has a burnup of 60 GW-day per metric ton of 

fuel with a minimum cooling period of 5 years. Table 5-1 shows the spent fuel inventory in 

terms of elements per metric ton for a typical low-enriched, light-water-reactor spent fuel for two 

burnups, 45 and 60 GWd/t, with 5- and 30-year cooling periods. The burnup in commercial 

plants could be lower than those used in this report. Both mass and decay heat power are 

presented in this table. The decay heat power comes from the isotopes of the elements 

reported. 
5-2 

Table 5-1 Typical low-enriched light water reactor spent fuel inventory 

(per metric ton of fuel) for two cooling periods and burnup 

ranges [NEA 2012-15]
5-3 

A representative list of the fission products that contribute to the decay heat as a function of the 

cooling time in terms of years is shown in Figure 5-13. As shown in this figure, the principal 

fission products at times < 10 years are 134Cs, 137Cs + 137mBa progeny, 90Sr + 90Y progeny, 106Rh, 

154Eu, 144Ce + 144Pr progeny and 147Pm. Based on a DOE report [DOE/SNF078], elemental 

decay heat as shown in Table 5-1 is primarily from Kr as 85Kr, Ba as 137mBa, Sr as 90Sr, Y as 90Y, 

Sb as 125Sb, Ru as 106Ru, Rh as 106Rh, Eu as 154Eu, Pm as 147Pm, Ce as 144Ce, Pr as 144Pr, U as 

234U and 238U, Pu as 238Pu, 239Pu and 240Pu, Np as 137Np, Am as 241Am, and Cm as 242Cm.

Figure 5-1 Important fission products, which contributed to decay heat 

for typical light water reactor fuel for cooling times from 1 to 

1,000 years [OECD 2011]

The accident type and process stage determines the material suspended as vapors and 

aerosols. For example, spent fuel before the dissolution process is in solid form, and may be 

contained in a fuel rod. Spent fuel in the extraction and dissolution processes is in solution, and 

therefore behaves differently. This, understandably, requires resolution of materials into the 

isotopic amounts and physical forms throughout the various stages of the reprocessing facility.

An example of the inventory during the operations of a UREX type of extraction is shown in 

Figure 5-2 and Figure 5-3 for 5-year and 30-year cooled fuel, respectively. As shown in these 

figures, most of the volatile fission products are released during the voloxiditation process. In 

this process, UO2 solution is heated to convert UO2 to U3O8, and volatiles escape as in the 

proposed AREVA plant [Bader 2011]). During the dissolution operation, the off-gas contains 

iodine vapors that are captured by a silver impregnated zeolite bed (it can be found in the 

proposed AREVA plant [Bader 2011]). In this dissolution operation hulls and other assembly 

hardware waste are produced (it can be found in the proposed AREVA plant [Bader 2011]). 

3 This figure is based on calculations performed for an enrichment of 4.5 wt% and 50 GWD/T burnup.
5-4 

The remaining solution operations, after the dissolution operation, separate out U, U-Pu-Np, and 

various fission product streams (it can be found in the proposed AREVA plant [Bader 2011]). 

Similar determination of fission product and transuranic inventories need to be established to 

quantify the material at risk in the event of an accident. 
5-5 

Figure 5-2 Operation inventory for an UREX type extraction process (with no Pu 

separated out) of a 5-year cooled and 45 GWd/t burnup [NEA 2012-15]. 
5-6 

Figure 5-3 Operation inventory for a UREX Type extraction process (without Pu 

separated) of 30-year cooled fuel with 45 GWd/t burnup [NEA 2012-15]. 
5-7 

5.2 Source Term Potentials and Release Mechanisms

Currently, many non-nuclear-reactor facilities, such as those at many Department of Energy 

sites, use a five-factor formula to identify their source term releases [DOE 1994]. These factors 

are all multiplied to arrive at a source term. Each of the factors is defined below:

1. MAR – for our purposes, material-at-risk is the radioactive inventory of the spent fuel in

an operation. For a criticality accident, additional radioactive inventory as a result of

fissions are included in the computation of MAR. For a tank containing spent fuel

solution materials, the radioactive inventory in the tank is the MAR.

2. DR – damage ratio is the fraction of MAR involved in the accident. For example, if a

tank bursts, the fraction of the initial MAR that is spilled is the DR.

3. ARF – airborne release fraction is the fraction of MAR that is airborne.

4. RF – respirable fraction is the fraction of MAR that is respirable. For example, the

fraction of the airborne spill content that is in the respirable size range. Particles smaller

than about 20 μm in aerodynamic diameter are considered respirable.

5. LPF – leak path factor is the fraction of MAR released out of the facility. For example,

the fraction of respirable aerosol from the airborne spill content that gets out of the

facility. This factor considers the room from which the accident occurred to the various

open pathways to release out of the facility.

Based on the discussions in Section 4, four major accident categories can be identified: 

criticality, explosion, fire, and spill. Among these accident categories, explosions contain the 

most energy and may fail the confinement boundary of the facility especially in locations where 

direct radiation levels are low and robust shielded structures are not required. However, a 

smaller explosion may only damage the primary container of the MAR, and possibly the room in 

which the container resides. In this case, a LPF calculation is necessary to determine source 

term from the facility. Therefore, the following sections provide more detail about the 

phenomena associated with criticality, explosion, fire and spill accidents, including any past 

accidents in fuel reprocessing facilities.

Several computer codes are widely available to determine the source term from a facility. Many 

of these codes came from the severe accident analysis of nuclear power reactors. Below is a 

list of the system codes which can be used for source term analysis:

1. MELCOR – This code was developed for the Nuclear Regulatory Commission (NRC) by

Sandia National Laboratories. Since its inception in 1980s, MELCOR has been used in

analyzing nuclear power reactor accidents, including spent fuel pool accidents. In

addition, MELCOR is used for non-reactor applications to estimate the source term from

nuclear and non-nuclear facilities at Department of Energy (DOE) sites. These facilities

include:

a. Plutonium processing facilities

b. Weapon assembly facilities

c. Weapon material processing facilities

d. Nuclear material processing facilities

e. Nuclear material testing facilities
5-8 

2. MAAP – This code was developed by Fauske and Associates for the U.S. nuclear

utilities. The code is used for supporting licensing of commercial nuclear power plants.

3. FATETM – This computer code was developed by Fauske and Associates and is

intended for fuel cycle and non-reactor facilities. It has been used for Hanford waste

facilities [NS-ES-0194].

4. ASTEC – This code was developed by IRSN (Institut de Radioprotection et de Surete

Nucleaire) and GRS (Gesellschaft fur Anlagen und ReaktorSicherheit), and serves a

similar function as MELCOR [Lajitha 2008].

The above system computer codes can predict the room-to-room transport based on the 

pathways in the facility during an accident. Almost all of these system codes have been 

developed for nuclear reactor applications. Only the FATETM code includes specific models for 

fuel cycle facilities and other facilities [NS-ES-0194]. MELCOR has been used for estimating 

the LPF values for DOE nuclear and non-nuclear facilities as indicated in Item 1 above. 

MELCOR can also be used to model spent fuel pools. To model a spent nuclear fuel 

reprocessing facility, additional models need to be added to the existing system codes to 

simulate accidents in a reprocessing facility. The next subsection discusses the phenomena 

threat need to be modeled in the system codes, and the data or correlations that are available.

5.2.1 Explosions

Explosions can occur in the fuel reprocessing facilities when process vessels over pressurize as 

a result of chemical reactions or physical processes. Physical explosions are a result of 

process vessels not being able to relieve pressure because of a cooling system or pressure 

relief valve failure. Chemical explosions include “red oil” reactions, HAN reactions, hydrogen 

combustion, and combustion of other gases or solids (such as zirconium fines) inside process 

vessels. Of course, the impacts of these explosions are not restricted to vessels and containers 

containing the unstable reactants, but may also affect other process equipment, confinement 

barriers, or ventilation system. The potential for a steam explosion is possible in the waste 

vitrification facility. However, this section primarily discusses explosions associated with 

process vessels as a result of chemical reactions.

An explosion can be delineated into various phases, each requiring predictive modeling to 

determine the potential aerosol and vapor generation that is available for release. The following 

subsections include methods for determining vessel failure pressure, airborne release, 

evaporation, and overpressure blast effects.

5.2.1.1 Vessel Failure Pressure

When an explosion occurs inside process equipment, it must first fail the vessel to be of 

significance. An explosion that leads to vessel failure is called a pressure vessel burst (PVB) 

[AIChE 2010]. Examples of PVBs include (a) chemical reaction or combustion inside the 

vessel, (b) overpressure due to external sources, such as high pressure supply from 

pressurized equipment (pumps), and (c) failure of the pressure relief valves if overpressure of 

the vessel is an anticipated situation. The equations presented below can be used to estimate 

the bursting pressure of a process vessel [NUREG/CR-4593].
5-9 

The hoop-stress equation is commonly used to determine the limiting stress on thin-walled 

process vessels 

σσtt = SSpp = PP ddoo

22 tt

Equation 5.1

where σσtt= ultimate tensile strength (psi), SSpp= hoop stress (psi), PP= burst pressure (psi), ddoo= 

outside diameter of vessel (inches), and tt= thickness of the vessel (inches). To solve for PP, 

Equation 5.1 becomes

PP = 22 tt σσtt

ddoo

Equation 5.2 

An empirical correlation for PP that accounts for the fatigue of the process vessel due to repeated 

cycling of pressure or by the notching of the vessel walls is given by the Bach equation

[NUREG/CR-4593]: 

PP = σσtt�ddoo

22 − ddii

22�

11. 33 ddoo

22 + 00. 44 ddii

22 Equation 5.3

where ddii = inside diameter of vessel (inches).

A modified Svensson equation for PP is given by: 

PP = σσttFFcccccc �

22 tt

ddoo

� �11 − tt

ddii

� Equation 5.4

where FFcccccc= cohesion factor for cylinders and is given as:

FFcccccc = � 00. 2222

εεuu + 00. 222222� �

ee

εεuu

�

εεuu

Equation 5.5

where ee= base of natural log, and εεuu= true strain rate at maximum log.

To demonstrate the differences in computing the burst-pressure from Equation 5.2 to 

Equation 5.4, consider a 304 stainless steel process vessel with a 48-inch outer diameter and 

0.25 inch thickness, and the material properties given in Table 5-2. The burst pressures 

calculated with Equation 5.2, Equation 5.3, and Equation 5.4 are 854, 1005, and 643 psi, 

respectively. 

In addition to empirical correlations, as discussed above, computational tools are available to 

simulate in detail the rupturing of a vessel of essentially arbitrary geometry, materials of 

construction, explosive energy, and explosive location within the vessel. The simulation 

provides not only the rupture size, geometry, and location, but also can estimate the mass and 

particle size distribution of liquid ejected. The simulation provides the liquid mass that (1) 

remains in the vessel, (2) settles to the ground or adheres to surfaces, and (3) becomes 

aerosol. These results are obtained by using a structural mechanics code to solve the solid 

mechanics problem of material deformation on short time scales, as discussed in Section 5.2. 
5-10

Table 5-2 Process vessel characteristics and its material properties [NUREG/CR-4593]

Facility Type Process Vessel 

Materials

Vessel 

Thickness 

(inches)

Outside 

Diameter 

(inches)

Length (inches)

MOX fuel fabrication Stainless steel Up to 0.25 3.6 – 72 13.8 - 384

Fuel reprocessing Stainless steel

Titanium

Up to 0.25

--

3.6 – 132

72

216

120

Solidification* Stainless steel

Hastelloy X

Up to 0.25

--

--

36

--

120

Material Ultimate Tensile Strength (psi) True Stain Rate, εεuu

Carbon steel AISI-SAE 1020 65,000 -- 

304 stainless steel 82,000 0.585

Titanium alloy 144,000 -- 

Hastelloy X 110,000 -- 

*Solidification includes a number of methods, including using inorganic additives, organic additives and vitrification

5.2.1.2 Airborne Release Estimate

Extrapolation of experiments used in studies of explosive-induced dispersal of both solids and 

liquids in a closed chamber of sufficient volume to preclude build-up of substantial pressure 

were suggested by Steindler and Seefeldt [Steindler 1980]. The experiments are for spherical 

explosives or cylindrical explosives with a diameter to height ratio of one. Based on the 

extrapolation, an empirical procedure has been created for producing an airborne release 

estimate for use in facilities [Steindler 1980]. The steps in this approach are as follows.

1. Estimate the mass ratio (rrmm) as the ratio of the weight of the inert material to the weight

of explosive. The explosive used is the TNT-equivalent weight of the explosive material

(such as the unstable solvent in the case of the red oil). Table 5-3 shows the TNT

equivalence for various systems of combustible/pressurized gases and liquids.

Table 5-3 TNT equivalence for various gas, combustible gas and combustible liquids 

released during vessel rupture events [NUREG/CR-4593]

Material and Conditions

Pound of TNT Equivalence* 

Per ft3 of Material Released 

under Stated Condition

Gas at 100 psi and 20  ̊C 0.02

Gas at 1000 psi and 20  ̊C 0.42

Gas at 10,000 psi and 20  ̊C 6.53

Argon at 20,000 psi and 20  ̊C ~1.0

Hydrogen gas at 20,000 psi and 20  ̊C ~1.7

Hydrogen explosion at 50% efficiency, including hydrogen gas 

released to atmosphere at 20,000 psi and 20  ̊C ~170

Mineral oil at 20,000 psi and 20  ̊C ~0.06

Red oil explosion 0.23

*1 pound of TNT is equivalent to the release of 1830 BTU (~5x105 cal)
5-11

2. Estimate the mass median diameter (ddgg) from Figure 5-4. Note the experimental data

only contains data up to rrmm of about 15 μm. Anything above this rrmm is considered an

extrapolation.

Figure 5-4 Median particle diameter as a function of mass ratio (rrmm) 

[Steindler 1980]

3. Estimate the ratio of mass airborne per mass of explosive, AAAA (g aerosol/g explosive)

using the expression [NUREG/CR-4593]:

AAAA = 22. 77 rrmm00.3333 Equation 5.6

or using Figure 5-5. 

Note that the airborne release is assumed to be instantly distributed over the room 

volume, following which settling immediately begins according to Stokes Law, as well as 

other forms of deposition. Also, an inherent assumption in the transformation from the 

experimental results to fuel cycle facility applications is the assumption of lognormal 

distributions with a geometric standard deviation (GG GGgg) of 2. 
5-12

Figure 5-5 Initial AAAA of airborne versus mass ratio [Steindler 1980]

Additional steps for the procedure have been proposed in [NUREG/CR-4593]. These steps 

allow for the creation of the particle size distribution. For each size range selected starting from 

the smallest diameter, it is necessary to calculate the fraction of aerosol in the size range by 

subtracting the two endpoint values of the cumulative normal frequency distribution, P(Z) (see 

Figure 3-4 [from NUREG/CR-3593]), where Z is computed as �ln dd − ln ddgg��ln GG GGgg. Finally, 

the fraction of the aerosol in each range is multiplied by the initial AAAA computed in Equation 5.6

to obtain the initial aerosol mass for each size range.

Instead of the additional steps described in [NUREG/CR-4593], [Bloom 1994] introduced the 

fraction of airborne, AAAAdd1−dd2 within the aerosol size range, dd1 to dd2 : 

AAAAdd11−dd22 = 00. AAAA ⎝

⎛ee ee � llll �

dd11

mm�

√22 llll(GG GG)

� − ee ee � llll �

dd22

mm�

√22 llll(GG GG)

�

⎠

⎞ Equation 5.7

Where AAAA = Equation 5.6, ddii= the aerosol diameter (dd2 > dd1) (μm), mm = mass median diameter 

(μm), which is defined as 38.962 rrmm0.3617, GG GG = geometric standard deviation, and ee eeee = the 

complementary error function. 

Note that [Bloom 1994] had pointed out that the method provided by Steindler and Seefeldt 

[Steindler 1980] was a “straw man” to encourage further work in this field, but apparently, no 

further work has been done. Cautions are that the use of the method from Steindler and 

Seefeldt [Steindler 1980] may require benchmarks from high-explosive experimental data. Red 
5-13

oil explosions in a pressure vessel may not be similar in terms of energy density, time-scale, 

etc., to high-explosive experiments used in the model. When a process vessel is ruptured due to 

an internal explosion and the rupture pressure is less than 500 psi, the airborne release 

described in Section 5.2.4 may be used. 

In recent years, researchers have begun to study the physics of generating liquid droplets by 

explosions below the liquid surface [Kudryashova 2011, Davydov 2003, Kerdrinskii 1993]. 

[Kerdrinskii 1993] describes the transformation of a cavitating liquid into a gas-droplet system. 

The fragmentation of liquids under explosive loading consists of the following four steps: 

1. formation and dynamics of bubble clusters in the liquid,

2. unbounded multiplication of cavitation nuclei leading to a foamy structure,

3. breakup of the foamy structure into cavitating fragments, and

4. transition into the droplet state and its evolution.

5.2.1.3 Vaporization

The droplet size distribution from an explosive event changes due to liquid evaporation and the 

vaporization rate is affected by soluble impurities in the solution. A recent paper [Ishmatov 

2010] describes this phenomenon as discussed in the following paragraphs. 

In the absence of foreign impurities, liquid evaporation is limited by diffusive transport of vapor 

away from the drop. For diffusion-limited evaporation the time rate of change of the droplet 

radius rrdd, is given as:

ddrrdd

dd = rrdd ρρdd

ρρvv − ρρaa

MMww

Equation 5.8

where = diffusivity of the vapor, ρρdd= liquid density, ρρvv= vapor density just outside the liquid 

drop, ρρaa aa= vapor density in air far from the drop and MMww= molecule mass of the liquid. 

The effect of soluble impurities is to reduce the vapor density above the liquid. Equation 5.8

becomes

ddrrdd

dd = rrdd ρρdd

xxdd(tt) ∙ ρρvv − ρρaa

MMww

Equation 5.9

where, xxdd(tt) = mole fraction of liquid as a function of time, which is given by:

xxdd(tt) =

11

MMHH22OO�11 − CCii(tt)�

CCii(tt)

MMii

+ 11 − CCii(tt)

MMHH22OO

Equation 5.10

where MMHH2OO= water molar mass, CCii(tt)= impurity concentration in solution as a function of time 

and MMii= impurity molar mass. 

In Table 5-4 the above equation is used to calculate the evaporation time of water from 20% salt 

solution drops. As shown in the table, small drops evaporate faster for the same reduction in 

size. 
5-14

[Ishmatov 2010] applies the above analysis to estimate the initial aerosol distribution when the 

aerosol is a result of pulsed explosive spraying of the liquid as shown in Figure 5-6. As shown 

in this figure, the explosive is located at the bottom of the apparatus and the liquid splashes on 

the reflector to form the aerosol. Figure 5-7 shows the final aerosol size distribution measured 

in the experiment (a), and the extrapolated initial size distribution using the above equations (b). 

The authors of this paper indicate that this pulsed method would yield high velocity droplets up 

to 300 m/s resulting in high flow turbulence. The evaporation of liquid aerosol is a complex 

process. Unfortunately, not enough detail is given in the paper to determine the conditions of 

the experiments or the measurement technique.

Table 5-4 Time of complete evaporation of water from aerosol drops of 

a 20% salt solution [Ishmatov 2010]

Initial drop size, μm 100 50 40 30 20 15 10 5 

Final size, μm 45 22 18 13 9 7 4 2 

Evaporation time, s 17 3.85 2.44 1.32 0.68 0.42 0.20 0.04

Figure 5-6 Explosive-induced liquid aerosol experiment [Ishmatov 2010]
5-15

Figure 5-7 Aerosol size distribution for (a) final aerosol distribution in 

measurement (b) initial aerosol distribution calculated 

[Ishatov 2010]

5.2.1.4 Blast Models

There are several simplified empirical models available to predict the destruction of walls and 

ceiling within a room where an explosion has occurred. In this section, two models are 

discussed:

• Venting model

• Overpressure model

Venting Model

Using the methodology discussed in [Rodriguez 1995], the compartment pressure can be 

estimated. Furthermore, one can assume that the same thermodynamic relation (Equation 

5.11) can be extended to estimate the pressures of the secondary compartments which are 

adjacent to the explosion compartment. This method may break down if the explosion is a 

detonation instead of deflagration. Also, the thermodynamic relation may not be correct if the 

explosion is directional and the expanded gas jet released from the vessel breach impacts the 

walls directly. Because a detonation produces a shockwave (with a time scale on the order of 

milliseconds), any reflected pressure waves from surfaces complicates calculating the failure of 

the surrounding structures or connecting structures. However, if structural data are available, 

the effects of deflagration and non-reflected pressure pulses can be used for modeling a 

reprocessing facility accident [Rodriguez 95].

Determining the opening size in a wall or ceiling also requires knowledge of the structure 

geometry and material properties. Gases from an explosion travel through penetrations such as 

doors, windows, ventilation ducting or piping holes before or at the same time as the walls or 

ceiling fail. In hazard calculations, there is a method called “explosion venting,” which provides 

a pathway for discharging combustion gases during a deflagration to maintain pressures below 

the room damage threshold. This venting method is determined by the size of the vent in order 

to keep the equipment or other structures intact without damage. The fire standard NFPA 68, 

entitled “Standard on Explosion Protection by Deflagration Venting” describes this venting 

method. A discussion of explosion venting can be found in the fire handbook [NFPA 2008]. 

Similarly, a safe venting estimate for a red oil explosion (runaway reaction) has been discussed 
5-16

elsewhere [Paddleford 1994]. In this reference, the venting is intended to prevent the red oil 

explosion because the accelerating temperature increase is a result of pressure build-up and 

run-away reaction rates at high-pressure conditions. 

For a low-strength enclosure, the basic relation for a deflagration vent size is given as [NFPA 

2008]:

AAvv = CC AASS PPrr

11/22 ⁄ Equation 5.11

where AAvv= required minimum vent area (m2), C = venting equation constant, which is a function 

of the flame burning velocity (see Equation 5.12 below), AASS= enclosure surface area including 

ceiling, floor and walls (m2), and PP rr = maximum pressure developed in vented deflagration 

(bar1/2). Note that this maximum pressure for this equation must be < 0.1 bar-g (gauge 

pressure). C for the low-strength enclosure is given as a function of burning velocity, vvbb (cm/s): 

CC = 11. × 11 − vvbb

22 + 11. × 11 −44vvbb − 00. 00 0000 Equation 5.12

Equation 5.11 is restricted in application to cases in which the longest dimension of the

enclosure (LLmm mm), measured from the vent is [NFPA 2008]:

LLmm ≤ ccLL

AACC

PP

Equation 5.13

where LL= 12 for relatively unobstructed enclosures with quiescent gas mixtures or 8 for 

enclosures that have many internal obstructions or possibly highly turbulent gas mixtures, AACC = 

the cross-sectional area normal to LLmm mm, and PP = perimeter of the cross section.

For the high-strength enclosures (PP rr > 0.1 bar-g) [NFPA 2008], AAvv is given by

AAvv = �(00. 11 ll11 KKGG − 00. 00 )PPrr

−00. + 00. 11 PPrr−00. (PPss − 00. 11)�VV22/33 Equation 5.14

where PPss = vent deployment pressure (bar-g), and KKGG = 100 SSuu⁄46, which is a function of the

combustible gas, and SSuu= burning velocity (cm/s). Note that Equation 5.14 is for L/D ≤ 2, where 

L = longest dimension of an enclosure, and D is the equivalent diameter of the cross sectional 

area.

The challenge here is how to apply the above equations for the red oil explosions, and what 

burning velocity is appropriate for the red oil explosions.

Overpressure Model

Blast overpressure models discussed herein are based on shock wave propagation into the 

surroundings once the PVB has occurred [AIChE 2010]. Many of the analytical explosion 

methods, especially in the area of gas explosions in terms of detonation/deflagration (in relation 

to the flame speed) and confined explosions have been examined in [Louie 2004]. While direct 

simulation of structural damage is possible with CFD analyses, it is unclear whether a CFD code 

will be available or chosen for future applications given the limited availability of data to 

represent a facility. The analytical expressions presented below provide a means to estimate 

whether structural damage occurs. However, the methods are somewhat simplified and often 

may not be applicable for certain geometries. 
5-17

In the last quarter century, a number of experiments and studies correlated PVB with solid 

explosive blast effects [Baker 1977, Coleman 1988 and Cain 1996]. Many experiments were 

performed in open air to observe PVBs due to pressurized gases or liquids [Cain 1996]. In the 

1970s, the Baker model [Baker 1977] was developed for estimating the pressure wave effects in 

exploding propellant tanks and gas storage vessels. Experiments have confirmed Baker’s 

model for determining overpressure and impulse as a function of distance from the tank rupture 

using both pressurized vessels and solid explosives [Baker 1977, Coleman 1988 and Cain 

1996]. In the Baker-Strehlow model, Sach’s scaling (or Energy scaling), is used to estimate the 

overpressure and impulse as a function of the scaled distance ( �) [AIChE 2010] as,

� = � PP00

EEee

�

11/33

Equation 5.15

where PP0= ambient pressure, EE = explosion energy, and = standoff distance. The scaled 

overpressure (PP�), is defined as, 

PP� = PPss

PP00

Equation 5.16

where PPss= side-on pressure.

A shock wave consists of positive and negative phases. Most of the time, only the positive 

phase is considered because it usually represents the largest change in pressure. 

The scaled impulse (ii) is defined as:

ii = ii aa00

PP00

22/33

EEee

11/33 Equation 5.17

where ii= impulse, and aa0= sound speed at the ambient condition.

To obtain the reflected overpressure (PP ) experienced by structures (e.g., walls), the following 

equations are needed [TIC-11268]:

PPrr = PPrr

PP00 � Equation 5.18

And

PPrr = 22 PP + (γγ + 11) ∙ PP�22

(γγ − 11) ∙ PP� + 22

Equation 5.19

As indicated in Equation 5.19, the scaled reflected overpressure (PP ) is more than twice the 

scaled side-on pressure (PP). For air, γγ= 1.4.

Based on the above scaled quantities, and solid explosive blast data, one could obtain Ps and i

graphically. In this approach [Baker 1977], explosive blast curves for Pentolite for PVBs have 

been matched closely to gas burst data. However, this model is based on a spherical Pentolite 
5-18

explosion in free air (which is away from the ground). To extrapolate these curves for cylindrical 

shaped explosives, correction factors must be applied [Ferradas 2006 and Geng 2011]. A 

recent paper has suggested that these correction factors should also take into account the 

length to diameter ratio of the cylindrical vessel, and the ratio of burst pressure to ambient 

pressure [Geng 2011]. For this study the use of the spherical burst overpressure and the 

inclusion of a reflecting surface (e.g., the ground) may be adequate to estimate the 

overpressure and impulse. In this approach, the burst pressure through the opening is

assumed to be isotropic. Table 5-5 shows the correction factors to account for the situation 

near the ground. Graph lookup may not be an efficient method, so a recent paper has 

converted PP� and ii estimates from this model into a set of equations that are functions of � in 

Table 5-6 for the spherical vessels. 

Table 5-5 Scaled overpressure and impulse correction factors for 

vessels situated near ground [Ferradas 2006]

� Correction Factor

PP� ii

< 1 2 1.6

≥ 1 1.1 1

Table 5-6 Fitted PPss and ii equations* for spherical vessels [Ferradas 2006]

Interval Scaled Overpressure 

Equation Scaled Impulse Equation

0.1 ≤ ≤ 0.2 PP = 1.25 −1.92

ii = 0.0164 −1.97

0.2 < ≤ 1.5 PP = 0.58 −2.39

ii = 0.0622 −1.14

1.5 < ≤ 1000 PP = 0.26 −1.11

ii = 0.0550 −1.04

*Equations have included ground effects, see Table 5-6. The variables for these equations are defined in

Equation 5.15, Equation 5.16 and Equation 5.17, and the units are in the MKS (Meter-Kilogram-Second) 

system. For example, the pressure is in Pascals, distance is in meters, and energy is in joules. Note is 

measured from the center of the explosion, which means that this value cannot be zero because of the 

presence of the vessel.

For a typical steel vessel rupture [AIChE 2010], the rupture energy is on the order of 100 to 

103 J. Energies as high as 108 to 1010 J were predicted for the TOMSK explosion accident (see 

Table 4-4). About 20% to 50% of the explosion energy is used in propelling fragments. These 

fragments include the pieces from the rupture vessel and liquid and solid materials from inside 

the vessel. 

About 50% to 80% of the remaining energy may damage surrounding structures. Once the 

vessel fails, the expanded gas fills the accident room first and some energy is dissipated by this 

process. Any weak point in the accident room, such as unbolted equipment, ventilation ducting, 

windows, or doors can be propelled by the expanding gas. A portion of the remaining energy 

may fail walls or ceiling, or drive the gas to adjoining rooms. Eventually, the expanding gas 
5-19

dissipates its remaining kinetic energy and attains equilibrium with the surrounding air. 

Therefore, both the shock overpressure and expanding gases from the PVB event should be 

accounted for when assessing damage to the facility. 

One may assume that the pressure due to the explosion is on the order of the failure pressure 

of the vessel. This assumption is consistent with documented events. [Rodriguez 1995]

discusses a forensic trace of the red oil explosion event at the TOMSK-7 radiochemical facility 

in 1993. In this event, the propelled ceiling plug of the room landed some distance from the site. 

From the calculations, the compartment with the vessel failure had a pressure of about 4-5 ×105

Pa. Vessel failure pressure was determined to occur at about 30×105 Pa. This failure 

pressure corresponds to the failure property of the steel vessel. Note that the compartment 

pressure after PVB is based on the thermodynamic relation

PP11VV11 = PP22VV22 Equation 5.20

where subscript 1 is the vessel and 2 is the compartment, and PP and VV are the pressure and 

gas volume, respectively.

Another approach for estimating the overpressure in an enclosure is to assume an ideal gas 

expansion at a constant volume with the final temperature as the adiabatic flame temperature

from NUREG-1805 [Iqbal 04]. 

The overpressure may also be determined from high-explosive experimental data. In these 

experiments, the PVB is a result of a burst [Cain 1996]. Both compressed liquid and gas inside 

the vessel are used. Based on the burst, the explosion energy is calculated. 

The explosion energy (EE ) in Equation 5.15 needs to be determined to find the overpressure 

or specific impulse. A simple approach is to assume that this energy is equal to or greater than 

the design pressure if a material-defect or missile impact is the cause. These failure modes are 

not considered in this report.

The equations for the explosion energy are given below [Coleman 88]. Considering an ideal gas 

and isentropic condition, the explosion energy (EE ,ii ii) is given by:

EEee ,ii ii = PPvvVVvv

γγ − 11 ∙ �11 − PP00PPvv

�

γγ−11

γγ Equation 5.21

where γγ=heat capacity ratio of the gas, VVvv= volume of the vessel, PPvv= failure pressure of the 

vessel, and PP0= ambient pressure. 

For adiabatic ideal gas conditions, the explosion energy (EE ,aa aa) is given by:

EEee ,aa aa = PPvvVVvv

γγ − 11 ∙ �11 − PP00ρρvvPPvvρρ00

� Equation 5.22

where ρρvv= gas density inside the vessel before failure, and ρρ0= ambient gas density. During 

PVB, gas from the vessel does not have time to transfer significant heat to the surrounding gas. 

This equation does not assume the expansion to be reversible, and it is not widely used 

because the value of ρρvv is needed.
5-20

For the isothermal ideal gas condition, the explosion energy (EE ,ii ) is given by:

EEee ,ii = PPvvVVvv ∙ �

PPvv

PP00

� Equation 5.23

Equation 5.23 assumes that the temperature does not change through the PVB. Also note that 

Equation 5.21 to Equation 5.23 can be applied only for the ideal gas situation. These equations 

need to be modified for non-ideal gas behavior.

Ideal gas behavior is adequate for most low pressure situations, less than about 1.03x107 Pa 

(1500 psi). To correct for real gas behavior, a compressibility factor, Z is defined as [Coleman 

88]:

ZZ = VVmm aa

VVmm (ii )

= PP

cc TT ∙ VVmm Equation 5.24

where VVmm = molar volume (mole/unit volume) for the real gas, and cc = gas constant. The 

term RRccccccccccTT

PP is the molar volume for ideal gas, where TT = temperature, and PP = pressure. For 

ideal gases, ZZ = unity. For real gases, Z can be greater than or less than unity. Therefore, for 

real gases, Equation 5.21 to Equation 5.23 become, respectively:

EEee ,ii ii = PPvvVVvv

ZZ(γγ − 11) ∙ �11 − PP00PPvv

�

γγ−11

γγ Equation 5.25

EEee ,aa aa = PPvvVVvv

ZZ(γγ − 11) ∙ �11 − PP00ρρvvPPvvρρ00

� Equation 5.26

EEee ,ii = PPvvVVvv

ZZ ∙ �

PPvv

PP00

� Equation 5.27

Both the isothermal and isentropic expansion of a gas for predicting blast overpressure can be 

either real or ideal. However, both models tend to over predict the overpressure. 

To utilize the extensive high explosive data for determining the blast overpressure [Cain 1996, 

Baker 1977, Geng 2011], this explosion energy is often converted to the equivalent mass of 

high explosive (such as TNT or Pentolite). TNT equivalents for various gases are given in 

Table 5-3. In addition to this table, an empirical correlation to convert the explosion energy from 

PVB or gas explosions (i.e., hydrogen or natural gas explosions in the room) to the TNT 

explosive equivalent mass, MMTTTTTT is given by [Louie 04] as

MMTTTTTT = ββ

EEee

HHTTTTTT

Equation 5.28

where ββ = a TNT equivalence factor, yield factor, or efficiency factor, and HHTTTTTT = TNT blast 

energy. A value of HHTTTTTT is 4.68 MJ/kg [AIChE 10]. Note that ββ in Equation 5.28 is for 

combustion gases and varies from 0.03 to 1.0 [Louie 04]. 
5-21

Unlike the vapor cloud explosion condition, the assumption of a PVB is that the explosion 

occurs inside a vessel, so that ββ is unity. Once MMTTTTTT is estimated, then it can be used as input 

to the code BLASTX. 

5.2.2 Release Due to Nuclear Criticality

A release can occur due to an inadvertent criticality event (often referred to as a nuclear 

excursion). For a closed vessel to be ruptured, the fission energy of the criticality event must be 

sufficiently large to overcome the design pressure of the vessel. If this energy is sufficient to

rupture the vessel, the remaining energy may expel solution from the breached vessel, and 

some of this expelled solution can form aerosols and vapors. This accident has a similar 

aerosol formation process as discussed previously for explosions. However, fission energy is 

usually distributed uniformly over the volume, unlike a chemical explosion, which may be 

concentrated within a smaller region of a vessel. Moreover, criticality events do not create 

gaseous products from a chemical reaction, but only from change of phase due to an increase 

in temperature. In view of these arguments, a criticality event can be modeled similarly as a 

chemical explosion in terms of the formation of vapors and aerosols, with the exception that 

gas-phase products are not directly formed from the fission energy.

For an inadvertent criticality condition, the explosion energy should be calculated based on the 

total number of fissions. If the fission energy is large enough to exceed the failure pressure of 

the vessel, then Equation 5.21 to Equation 5.27 are not needed. 

5.2.3 Fires

A limited number of experiments have been performed for liquid combustion in the context of 

reprocessing solvents, which involve both nitric acid and organic solutions [NUREG/CR-4736]. 

In one set of experiments, small volumes of 30% tributyl phosphate (TBP) in normal paraffin 

hydrocarbon (NPH, a kerosene-like diluent used in fuel reprocessing) were burned. Ignition 

was provided by a methane flame located above a beaker containing the liquids. The NPH was 

the only combustible material in the experiment. The acids were included to see if fire and 

heating would cause migration of the fission products or transuranic elements across the phase 

boundaries.

An empirical correlation was developed using the combustible liquid of 30% TBP/kerosene for 

the rate of release of airborne uranium release rate with an aerodynamic mass mean diameter

(AMMD) of 0.6 μm, and geometric standard deviation (GSD) of 3.1 [NUREG-1320]. This 

correlation is given as

mṁ rr = 11. 33 SSrrwwrr Equation 5.29

where mṁ rr = mass release rate of radioactive particles (g/s), ww = mass of the radioactive 

material in the solution (g), and SS = smoke release dimensionless rate (s-1). SS is given as 

[NUREG-1320]:

SSrr = γγss �qq̇

rrrr

" + qq̇

" + qq̇" � AAbb

hhcc

mm Equation 5.30

where γγss= fractional yield of smoke (for TBP/kerosene: 0.002 to 0.087), qq̇" = surface radiation 

heat loss (for kerosene: 8 kW/m2), qq̇ff

" = flame convective heat flux (for kerosene: 11 kW/m2), 
5-22

qq̇ff

" = flame radiative heat flux (for kerosene: 14 kW/m2), AAbb= burning surface area of the 

combustible liquid (m2), hcc= heat required to generate a unit mass of fuel vapors (for kerosene: 

1.5 kJ/g), and mmff= mass of the combustible liquid. 

In addition to this empirical correlation associated with combustible liquids, airborne release and 

aerosol size are provided in Table 5-7 and Table 5-8 for combustible liquids. 

Table 5-7 Summary of the airborne release in combustible liquid [NUREG/CR-4736]

Experiment Airborne Release 

Fraction Notes** 

Pure acid/organic with uranium 0.004-0.0056 100 mL organic liquid was placed on top of 100 mL of 

acid in a metal beaker. External heat was supplied. 

Substitution of a glass beaker would only delay the 

acid boil over. Approximately 40-60% of organic layer 

remained unburned. The amount of smoke generated 

is 0.0076 to 0.0189, and the burn time ranges from 

27.5 to 53.3 minutes. 

Acid with fission products/organic 

with uranium (30% TBP in NPH*)

0.025-0.027 uranium The presence of fission products in the acid phase 

seems to enhance uranium release from the organic 

phase. The amount of smoke generated is from 

0.0288 to 0.0318 and the burn time ranges from 24.8 

to 34 minutes.

Acid with fission products and 

uranium/pure organic (TBP in 

NPH)

0.06-0.071 uranium These experiments did burn to solid residue, with the 

initial solution volume of 100 mL in a metal beaker.

Acid with fission products and 

uranium/pure organic (TBP in 

NPH)

0.002 uranium This experiment did not burn to solid residue. Its initial 

solution volume was 150 mL of acid with 50 mL of 

organic in a glass beaker with a heat pump to provide 

external heat flux. The smoke release is from 0.0644 

to 0.0891, and burn time ranges from 40 to 65 

minutes.

Acid with fission products and 

uranium/organic with uranium

0.008-0.016 uranium Experiments involved solutes in both the organic and 

acid layers. Experiments did not burn to a solid 

residue. Note that one experiment has provided 

aerosol size: AMMD at 0.6 μm, GSD of 3.1, and 99% < 

10 μm. The smoke release is from 0.0303 to 0.0651 

and burn time ranges from 51 to 57.3 minutes.

*NPH has the following properties: specific gravity at 15.6  ̊C is 0.76, viscosity at 25  ̊C is 1.8 centipoise, and flash

point at 70  ̊C.

** The lower burn time value should be used to estimate the release rate.
5-23

Table 5-8 Source term for burning of combustible liquid [NUREG-1320]

Release Mechanism Airborne 

Release Fraction

Airborne Particle 

Size

AMMD 

(μm) GSD

Burning of contaminated combustible liquids

U or Pu liquid (Combustible liquid is spilled over

large amount of radioactive material, then

 ignited)

U or Pu nitrate

Non-volatiles other than U or Pu

Semi-volatiles

Volatiles

0.114

0.003

0.0077

0.01

0.843

2.4

2.4

2.4

2.4

2.4

3.8

3.8

3.8

3.8

3.8

Heating of unpressurized radioactive liquids

Pre-boiling

Boiling

1.06x10-10/s

4.76x10-7/s

20.0

20.0

1.5

1.5

Fires also induce a buoyant plume which affects the transport of airborne particles. Fire codes 

are available and the ones listed below can estimate the transport of energy and mass, and 

characterize plume rise. However, aerosol generation and dynamics are not determined with 

these codes. The following codes could be used to perform estimates of burn rate, energy 

generation, fire duration, and plume rise as inputs to another code. The fire codes include: 

CFAST (Consolidated model of Fire and Smoke Transport) and FDS (Fire Dynamics Simulator). 

The descriptions of these codes are briefly provided below:

CFAST simulates impact of fires and smoke in a specific building environment. A twozone fire model is used to calculate evolving distribution of smoke, fire gases, and 

temperature throughout rooms of a building during a fire.

FDS models the fire-driven fluid flow. This code solves numerically a form of the NavierStokes equations appropriately for low-speed, thermally-driven flow, with an emphasis 

on smoke and heat transport from fires.

5.2.4 Spills

Spill accidents in a fuel reprocessing facility are mainly due to leaks from valves and lines, and 

cracks in tanks. The leaks are due to corrosion, overpressure, and mechanical impact. A spill 

due to overpressure is called a “pressurized spill” and is discussed later in this section.

In terms of the radionuclide behavior, particularly in creating an airborne release, a free-fall spill 

usually exhibits break-up of the liquid into drops due to the instability and shear stress at the 

surface of the liquid. Once the liquid/drops hit a hard, unyielding surface to create splashing, 

drops are broken into smaller ones and some of these drops are resuspended. The amount of

aerosol created by spills is less than that from more energetic events such as fire, criticality, and 

explosion accidents. For a non-pressurized spill, the energy source that creates suspended 

drops is gravitational potential. 

Non-Pressurized Spill

NUREG/CR-4997 documents the performance of a number of free-fall spill experiments 

involving powders and solutions. The properties of the experimental solutions compared with
5-24

water are shown in Table 5-9. Because it was impractical to sample the entire air volume 

immediately after the spill, a correction was made to account for evaporation effects in the data. 

For example, uranine solution particles were most affected by evaporation because uranine 

solution contains only 1% non-volatile solids; whereas, the UNH solutions contained 28% nonvolatile solids and a large percent of lower volatility HNO3. 

Table 5-9 Comparison of properties of water and experimental liquids*

Material Density (g/cc) Viscosity 

(centipoise)

Surface Tension 

(dyne/cm)

Vapor Pressure 

(mm Hg)

Water 0.997 0.89 72.0 23.8

UNH a 1.54 1.70 66.4 13.6b

Uranine c 0.993 0.79 52.1 23.7

Sucrose d 1.0-1.3 1.3-47.1 67.0-83.0

Slurry 1.1-1.4 1.3-3.1 58.2-68.4

*at 25  ̊C, and adapted from Table 3.2 of [NUREG/CR-4997]. a Uranyl nitrate hexahydrate solution in nitric acid with 208.7 g uranium/liter (0.877 Molar), 8.4 Molar HNO3. b Calculated

c 10 g/liter uranine in water

d Varied from 0 to 60% sucrose

Table 5-10 shows the airborne release data for the free-fall spill experiments. The initial data 

are generally higher than the impactor data, since the initial data account for evaporation and 

settling (see [NUREG/CR-4997] for more details about the data corrections). 
5-25

Table 5-10 Experimental results on liquid spills with correction [NUREG/CR-4997]*

Applied Condition Initial a Measured b Initial/Measured

All Spills1

Average AMMD, μm

Average GSDc

Average Airborne Fraction 

21.5

7.3

28.5 x 10-6

8.9

5.4

25.5 x 10-6

2.4

1.4

1.1

UNH Spills2

Average AMMD, μm

Average GSDc

Average Airborne Fraction 

27.2

5.9

9.3 x 10-6

20.1

4.5

7.9 x 10-6

1.4

1.3

1.2

Uranine Spills3

Average AMMD, μm

Average GSDc

Average Airborne Fraction 

27.0

3.0

56.2 x 10-6

7.3

2.9

50.7 x 10-6

3.7

1.0

1.1

Sucrose Spills4

Average AMMD, μm

Average GSDc

Average Airborne Fraction 

11.0

14.2

10.6 x 10-6

2.7

9.4

9.7 x 10-6

4.1

1.5

1.1

Slurry Spills4

Average AMMD, μm

Average GSDc

Average Airborne Fraction 

15.8

10.1

23.3 x 10-6

3.1

6.7

20.7 x 10-6

5.1

1.5

1.1

Low Surface Tension Spills5

Average AMMD, μm

Average GSDc

Average Airborne Fraction 

20.2

3.7

45.9 x 10-6

6.5

4.2

41.5 x 10-6

3.1

0.9

1.1

*The experiments included a spill height from 1 to 3 m onto a dry stainless steel floor, liquid volume of 125 to 1000

ml. Liquids spilled included slurries and solutions of varying viscosities. The spill experiment is for a single spill

with a spill diameter (of spill container) from 6.2 to 12.4 cm.

a Recalculated to account for any evaporation and settling effect.

b Experimental impactor data

c Geometric standard deviation, lognormal distribution

1 If the liquid cannot be specified, use the data in this set

2 It represents heavy metal, aqueous solutions with densities > 1.2 g/cc

3 It contains less than 1% non-volatiles 4 See Table 5-9 for properties identified 5 Surface tension is 33.3 dyne/cm, viscosity is 1.14-1.24 centipoise, and density of 1.004 g/cc.

NUREG/CR-4997 includes empirical correlations for the airborne release fraction (ARF) for 

liquid spill experiments. Equation 5.31 includes an evaporation and settling correction:

AAAA = . × 11 11 AA 00. Equation 5.31

where Arch is the Archimedes number, which is defined as

AA = ρρaa 22HH33

μμ

Equation 5.32

where ρρaa aa= air density (g/cc), HH= spill height (cm, extrapolating above 3 m may not be 

reasonable), gg = gravity (981 cm/s2) and μ= solution viscosity (poise). For a very low-density 

aqueous solution, the right hand side of Equation 5.31 should be multiplied by 3.
5-26

An empirical correlation similar to Equation 5.31, but which includes the Froude number and 

dimensionless density, can be used to estimate ARF 

AAAA = . 33 × 11 − FF 00.33 AA 00.44 �

ρρaa

ρρss

�

22.22

 Equation 5.33

where FF = Froude number and ρρss = density of the solution. FF in Equation 5.33 is given by

FF = vv22

Equation 5.34

where vv = impact velocity (vv = �2 gg HH, cm/s) and = radius of equivalent sphere of liquid 

spilled (cm).

Note that Equation 5.34 includes a dependence on impact velocity and drop size, which 

Equation 5.31 does not. Also, note that the data in Table 5-10 and the above correlations are 

only applicable for a single spill event. For spills where the drop diameter is < 8 mm, and the 

spill height > 4.26 m, the results of the correlations are likely to be conservatively high because 

of the way velocity is computed. In the case of continuous spills, where a falling stream 

encounters less air drag than individual droplets, the correlations may be non-conservative.

As described in [NUREG/CR-4997], attempts have been made to correlate the lognormal 

aerosol size parameters, AMMD and GSD. Only the slurry spill data yields a correlation with a 

least square greater than 60%, and it is given as [NUREG-1320]:

AA (μμ ) = 22. 33 × 11 − WW 11.

00.11 Equation 5.35

Where WW = Weber number and = Reynolds number and these are given as:

WW = ρρss dd vv22

εεtt

Equation 5.36

where dd= diameter (cm), εεtt = liquid surface tension (dyne/cm), and vv = velocity (cm/s), which is 

�2 gg HH, and

Equation 5.37

A sample calculation in NUREG/CR-4997 evaluates a hypothetical leak in a solvent line in a 

reprocessing plant. The solvent is 30% tributyl phosphate (TBP with kerosene as a diluent). 

This solvent has a viscosity of 0.0207 poise. It contains 0.1 Molar UNH and 1 Molar HNO3. It is 

assumed that the leak is at 25  ̊C and falls 366 cm (12 ft) to the floor. 

From Equation 5.32 with air density of 0.00155 g/cc, 

AA = (00. 0000 /cccc)22(333333 cc )33� cc /ss22�

�00. 00 00

cc ∙ ss

�

22 = 22. × 11 Equation 5.37

μ

ρ vd s Re =
5-27

Using Equation 5.32, the AAAA is 3.86 x 10-5 and from Table 5-10, AMMD = 21.5 μm and GSD = 

7.3.

Pressurized Spill

A pressurized spill is a release caused by a pressure that provides an initial velocity to the 

ejected liquid. It is expected that the ARF is higher than for the free-fall spill case.

A series of experiments were performed on pressurized releases of solutions [NUREG/CR3093, NUREG/CR-4779]. These experiments were conducted using so-called PressurizedAirborne Release Equipment (PARE) inside a cylindrical enclosure called the Radioactive 

Aerosol Release Tank. As shown in Figure 5-8, the chamber is about the size of a small room. 

Also indicated in this figure, the release is upward toward the ceiling of the chamber. This 

configuration should provide an upper bound on the ARF. The size of the opening of the 

rupture disk limits the amount of the liquid expelled. Aerosol is collected by impactors and filters 

that are located near the curved walls of the cylinder. Since PARE is designed for pressurized 

releases of powders, PARE was modified to measure liquid releases as shown in Figure 5-9. 

The initial set of the experiments were conducted using air as the pressurized gas at pressures 

ranging from 50 to 500 psig. The initial solution of uranine and uranyl nitrate had volumes 

between 100 cm3 and 350 cm3 [NUREG/CR-3093]. Because of concern that the ARF may be a 

function of the amount of gas dissolved in the liquid, additional tests were performed with CO2

[NUREG/CR-4779]. As shown in Table 5-11, the solubility of CO2 in uranine solution is 50 times 

that of air. Table 5-12 provides the airborne release data for the comparison of air and CO2 for 

pressurized uranine solution releases. As shown in this table, using air as the pressurized gas 

tended to yield a lower airborne ARF value in comparison with CO2. 

Figure 5-8 Pressurized radioactive aerosol tank configuration for 

measuring the pressurized release of solution from PARE 

[NUREG/CR-3093]
5-28

Figure 5-9 PARE modification for pressurized liquid release experiment 

[NUREG/CR-3093]

Table 5-11 Estimates of mole fraction of the pressurized gases dissolved 

in uranine solution* [NUREG/CR-4779]

Pressure (psig) Pressurized Gas Mole Fraction Dissolved

500 Air

CO2

4.2x10-4**

2.0x10-2

250 Air

CO2

2.3x10-4

1.3x10-2

50 Air

CO2

4.8x10-5

2.5x10-3

*Assume solubility of gases the same as in water. This solution contains a uranium concentration of 10 g/liter.

** Air value calculated assuming 78% N2, 21% O2; 1% inert (not included)

Table 5-12 Airborne release for pressurized release of uranine solutions 

[NUREG/CR-4779]

Chamber 

Pressure 

(psig) 

350 cc Volume 100 cc Volume

Air CO2 Air CO2

ARF AMMD

(μm) GSD ARF AMMD(μm) GSD ARF AMMD(μm) GSD ARF AMMD(μm) GSD

500 0.05 4 2.7 0.13 3.3 11 0.15 4 3.5 0.22 2.7 5.5

250 0.01 4 2.3 0.039 2.3 5.2 0.06 3 3.8 0.50 2.1 2.6

50 0.0008 2 3.5 0.014 1.5 3.8 0.005 2 4 0.006 1.9 3.3
5-29

In addition to a pressurized release, a flashing-spray release experiment was also conducted for 

an uranine solution at the same concentration. In this experiment, the rupture pressure was the 

vapor pressure of superheated liquid [NUREG/CR-4779]. Figure 5-10 shows the measured and 

predicted temperatures of the flashing-spray release experiment. The ARF data for this 

flashing-spray experiment are shown in Table 5-13. 

Figure 5-10 Measured and predicted temperatures for flashing-spray 

releases [NUREG/CR-4779] 

Table 5-13 Airborne release for flashing-spray releases of uranine 

solutions [NUREG/CR-4779]

Chamber 

Pressure 

(psig) 

700 cc Volume 350 cc Volume 100 cc Volume

ARF AMMD

(μm) GSD ARF AMMD(μm) GSD ARF AMMD(μm) GSD

240 0.0496 6.25 2.7

124 0.0151 5.9 2.6 0.0222 7.7 2.7 0.0851 6.4 3.3

57 0.0089 8.4 5.6
5-30

Unlike the uranine solution, UNH solution contains more non-volatiles in solution, and the

density is about 1.54 g/cc. The ARF results for pressurized release of UNH solution using air 

are given in Table 5-14. 

Table 5-14 Airborne release for the pressurized UNH solutions in air 

[NUREG/CR-3093]

Chamber 

Pressure 

(psig) 

350 cc Volume 100 cc Volume

ARF AMMD

(μm) GSD ARF AMMD(μm) GSD

500 0.009 16 3.4 0.01 14 2.7

250 0.005 29 4.4 0.0009 13 3.6

50 0.0002 4 5.0 0.0003 7 6.1

The empirical correlation for ARF during a pressurized release reported in NUREG-1320 is

AAAA = 00. 33 xx 00. Equation 5.38

where xxgg = mole fraction of pressurizing gas or vapor. The AMMD associated with 

Equation 5.31 is 6.8 μm and the GSD is 3.2.

Similar to the spill cases in the previous section, a correction was made to the measured data 

set for evaporation and settling [NUREG-1320]. With this correction, Equation 5.39 becomes

AAAA = 00. 3333 xx 00. Equation 5.39

A comparison of the aerosol size distribution data for liquid pressurized releases with and 

without the evaporation and settling correction is given in Table 5-15. As concluded in 

[NUREG/CR-3093], measured ARF values were a function of pressure, but were relatively 

independent of the volume of the experimental configuration in Figure 5-8. 
5-31

Table 5-15 Comparison of the measure and correction data of aerosol 

size distribution for liquid pressurized release [NUREG-1320]

Experiment 

Measured Data Evaporation/Settling Corrected 

Data

AMMD (μm) GSD AMMD (μm) GSD

All Data 7.3 3.9 12.0 6.6

Uranine/Air 3.3 3.2 12.4 4.5

UNH/Air 13.6 4.2 8.4 11.1

Uranine/CO2 2.3 5.2 9.3 5.6

Flashing Sprays 6.8 3.2 21.0 3.0

5.3 Basic Principles Approach to Determining Releases

As discussed in the previous section, progress has been reported for determining the factors in

the five-factor formula used to determine the source term. For convenience, the formula is 

MAR × DR × ARF × RF × LPF, 

which in expanded form is

Material At Risk × Damage Ratio × Airborne Release Fraction × Respirable Fraction ×

Leak Path Factor.

For this application, the MAR is the total mass of solution inside a unit being considered. 

Deposition and filtration effects may significantly reduce the LPF below unity, and can be 

determined with transport codes. However, estimates of DR, ARF, and RF are very difficult to 

obtain. In particular, scaling from 350 cm3 beaker-size experiments to cubic meter reprocessing 

units is questionable. Furthermore, the particle size distribution is not determined by the fivefactor formula approach, neither is the variation in radiological content with particle size. The 

particle size distribution and radiological content greatly influences transport within facilities, 

dispersion in the atmosphere, and health consequences. For the five-factor formula approach, 

the respirable fraction (RF) is used as a rough measure to encompass these effects. 

Furthermore, there are no experiments that we found on the scale used in reprocessing 

facilities. Instead, suggestions for ARF and RF given in the previous section are for isolated 

effects, and these are generally not on the scale of interest.

As part of this program, a basic principles approach is considered that not only provides the DR, 

ARF, and RF, but also provides the particle size distribution and radiological content of the 

aerosol with particle size, while simultaneously determining the failure of the containing vessel.

The approach is to model from basic principles the coupled problems of explosion dynamics, 

liquid solution release and break up, and structural vessel rupture and failure. These 

phenomena are simulated by coupled structural and fluid mechanics codes, including a drop 

break-up model that is stable and accurate. All these codes have been developed under a 

common architecture called Sierra to facilitate the multi-physics coupling [Edwards 2002]. 
5-32

Sandia’s Presto structural mechanics code, which is a Lagrangian, three-dimensional, explicit, 

transient code is used to solve the structural problem with large deformations and short time 

scales. Radionuclide liquid particle dispersion is modeled with Smoothed Particle 

Hydrodynamic (SPH) elements, which are coupled to the explosive and the processing vessel 

walls [Monaghan 2005]. SPH allows for radionuclide solutions to both impart momentum to 

solid structures and for the solutions to be dispersed upon ejection from the vessel. The solid 

walls and equipment in the vessel are modeled with structural finite elements because these 

elements may deform and separate, but they do not get atomized by the explosive. After the 

fluid has been ejected and drops on the order of centimeters to millimeters have separated, the 

system is then modeled with Sandia’s Fuego fluid mechanics code. Fuego is a low Mach 

number control-volume finite element code that solves the Navier-Stokes equations for the flow 

induced by the explosion and the ejected drops. Drop break-up is modeled using the Taylor 

Analogy Break-up (TAB) model [O’Rourke 1987; Desjardin 2002] employing a Lagrangian 

transport framework coupled to the Eulerian gas phase solvers. The Presto/Fuego coupling is 

one-directional in that the structural code with SPH is run first, and then the fluids code with the 

TAB model is used to determine the aerosol particle size distribution and concentration. 

Typically, the fluid is initially modeled with several hundred thousand SPH particles, and then 

the final aerosol has orders of magnitude more particles that result from the break-up of SPH 

particles. These are demanding calculations that require a parallel machine running for several 

days, but for the first time reprocessing accidents can be simulated in detail from first principles 

instead of relying on limited and conservative correlations for the explosion, structural 

deformation of the reprocessing unit, and liquid aerosolization that are often limited to only one 

or a few of the phenomena.
6-1 

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A-1

APPENDIX A

BARNWELL NUCLEAR FUEL PLANT DESCRIPTION

This appendix describes the Barnwell Nuclear Fuel Plant (BNFP) in terms of the major 

operations, designs, and processed data [NUREG-0278, AGNS 1975]. Major portions of BNFP 

have been constructed, which includes [NUREG-0278]: 

1. Main process building (26 m high, 36 m wide and 68 m long)

2. Fuel Receiving and Storage Station (FRSS) (28 m high, 45 m wide and 45 m long)

3. Control Room Area (CRA)/Plutonium Nitrate Storage and Load-out (PNSL) (18 m high,

20 m wide and 25 m long)

4. Waste Tank Equipment Gallery (WTEG) (12 m high, 31 m wide and 34 m long) Waste

Tank Cells (WTCs)

The plot plan for the BNFP is shown in Figure A-1. As shown in this plot plan, the overall 

process building including the FRSS and CRA and PNSL is shown in the center near the road 

entry at the south. Both WTSG and WTC are shown just west of the overall process building. 

The overall process building layouts are shown in Figure A-2 through Figure A-7. Shown in 

these figures are the wall thickness and dimensions of the building. In principle, thick concrete 

walls (up to 6 feet) are used in the area where high radiation is expected. For fuel storage 

pools, depth of water is used to guard against radiation. Windows for viewing at high radiation 

areas are constructed with high-density lead glass oil-filled materials to minimize any exposure. 

The lower right corner of the layout of the top view in Figure A-2 to Figure A-5 shows the FRSS. 

The upper right corner of the layout of the top view in Figure A-2 to Figure A-5 shows the 

CRA/PNSL and offices and cold/hot laboratories. 

The throughput capacity of the BNFP was designed for reprocessing 5 metric tons per day

using a PUREX separation process, which amounts to 1500 metric tons of spent fuel per year 

[AGNS 1975]. This plant can handle fuel with initial fissile content of up to 5% enrichment, or 

the equivalent for plutonium recycle fuels.

The overall detailed process is presented in Figure A-8. In Figure A-9, a simplified process 

block diagram is presented. As shown in this figure, the spent fuel is received in heavily 

shielded casks and is unloaded and stored underwater in a pool in the FRSS. During 

processing, each fuel assembly is transferred under water to the main process building, where 

fuel assemblies are partly disassembled, chopped into pieces up to 5” long and dropped into a 

dissolver vessel. The dissolver vessel contains sufficient nitric acid to dissolve the fuel. Volatile 

fission products (such as noble gases, tritium halogens and carbon) are released to the effluent 

cleanup system during this step. The fuel cladding hulls and assembly parts are packaged and 

taken to a bunker-type interim storage area outside. 

The dissolved fuel material (in nitric acid aqueous solution) is centrifuged and sent to the first 

stage of solvent extraction using organic solution of 30% tributyl phosphate (TBP), where the 

uranium and plutonium are extracted into the organic phase in a centrifugal contactor while 

most of the other transuranic elements and fission products remain in the aqueous nitrate 

solution. This nitrate solution also contains about 0.5% of the unrecovered uranium and 
A-2

plutonium, which goes into the high-level liquid waste (HLLW). After concentration, the HLLW is 

transferred to underground tanks for interim storage.

The organic stream (plutonium and uranium) from the contactor is scrubbed in a pulse column 

contactor to remove additional fission products and other transuranic elements. Then this 

solution is processed through an electro-pulse column where the chemical valence of the 

plutonium is reduced, so that the uranium can be separated. The plutonium aqueous stream is 

processed through two additional stages of solvent extraction to remove residual fission 

products and other impurities. Uranium in the organic solvent solution is processed through a 

solvent extraction to remove any impurities. Silica gel filtration adsorption is used in the final 

uranium solutions to remove any residual zirconium. The uranium nitrate conversion to the 

uranium hexafluoride facility is included in the BNFP. However, the plutonium nitrate is stored 

in tanks and sent to other facilities for further processing. The waste streams at the BNFP 

include intermediate and high level wastes, which are stored in tanks, waiting for further 

solidification and disposal. The nitric acid used in the various processes is recovered for re-use. 

All aqueous raffinates containing small quantities of fissile material, except solvent treatment 

wastes, are passed through a recovery extraction system prior to concentration and storage. 

Iodine scrubbers and inorganic iodine adsorption beds are used to minimize any iodine effluent 

through the stack. The iodine remaining in the dissolver solution may be removed by mercury 

addition to the general purpose concentrator and acid fractionator overhead vaporizer before 

discharge to the stack. The high level wastes are chemically denitrated to a nitric acid 

concentration of 1-5 Molar before being sent to interim liquid waste storage. Details of the 

processes at BNFP will be described in the following sections.

In this appendix, the descriptions of each of the above 5 areas of the BNFP is discussed along 

with the process operations. Support systems and operations are also described, especially 

safety related. 

Chemical inventory at the BNFP includes nitric acid, gadolinium nitrate, tri-butyl phosphate, 

hydrazine, hydrogylamine and mercury [NUREG-0278, vol.1 5-27].
A-3

Figure A-1 Plot Plan for BNFP [NUREG-0278]
A-4

Figure A-2 Top view of overall process buildings at elevation 270’ 6” [AGNS 1975]
A-5

Figure A-3 Top view of overall process buildings at elevations 270’-6” to 279’-0” 

[AGNS 1975]
A-6

Figure A-4 Top view of overall process buildings at elevations 279’-0” to 297’-0” 

[AGNS 1975]
A-7

Figure A-5 Top view of overall process buildings at elevation above 297’-0” [AGNS 

1975]
A-8

Figure A-6 Side view of overall process buildings at A-A and C-C planes [AGNS 1975]
A-9

Figure A-7 Side view overall process buildings at D-D and B-B Planes [AGNS 1975]
A-10

Figure A-8 Overall Process Diagram for the BNFP [AGNS 1975]
A-11

Figure A-9 Simplified block process flow diagram
A-12

A.1 Main Process Building

The main process building contains the majority of the processes at the BNFP, which include:

1. Shearing of fuel assemblies

2. Dissolution of fuel

3. Solvent extraction, partitioning and stripping

4. Fissile extraction cycle (uranium and plutonium)

Structures included in this building are the following (see Figure A-2 to Figure A-7): 

1. Remote Process Cell (RPC)

2. Remote Maintenance and Scrap Cell (RMSC)

3. High Level Cell (HLC)

4. High-Intermediate Level Cell (HILC)

5. Intermediate Level Cell (ILC)

6. Uranium Product Cell (UPC)

7. Plutonium Product Cell (PPC)

The following sections describe each of the structures in more details, including the operations 

within each structure.

A.1.1 Remote Process Cell (RPC)

This cell is a high-level radiation area used primarily for the head-end mechanical process 

operations and initial chemical processing. As its name implied, all operations conducted in this 

cell are performed remotely. The fuel assemblies are transferred from the fuel transfer pool of 

FRSS to the feed mechanism of the mechanical shear in the cell via the use of overhead crane 

through a transfer tunnel between this cell and FRSS (see Figure A-10). The mechanical shear 

chops the fuel assemblies into small segments about 2 to 5 inches long. These segments are 

contained in the dissolver basket inside the dissolver that contains hot nitric acid and gadolinium 

nitrate (which acts a soluble neutron absorber). The un-dissolved cladding (hulls of stainless 

steel and Zircaloy) remaining in the basket is lifted from the dissolver and is transferred to 

RMSC. 
A-13

Figure A-10 A-A View of the equipment arrangement in remote process cell [NUREG0278]
A-14

During the chop and shear process, the sweep-air system provides a positive flow of air from 

the cell into the shear feed magazine, so that the air sweeps any fission product gas and 

particulate down the diverter chute into the dissolver system to prevent any release into the cell 

during shearing. 

The cell floor and lower walls are lined with 304L stainless steel to contain spills of the process 

solutions. The stainless steel liner is welded to anchors in the underlying concrete. The spilled 

solutions may be transferred from the cell sumps to the intermediate level liquid waste system 

or returned to the process. Any exposed concrete surfaces in the cell are coated with an acidresistant phenolic paint system such as Amercoat 90 or Phenoline 305. Most of the external 

walls contain 5’ thick concrete where the high radiation is expected. The 5-foot thick concrete 

structure provides shielding to personnel in normal operating areas surrounding the RPC to a 

radiation exposure no more than 1 mrem per hour. Interior walls and lower radiation area 

(above the process equipment) are generally thinner. There is a removable (wheel-mounted) 

shield door allowing the crane and power manipulator access to this cell from the Crane and 

Equipment Maintenance Gallery (CEMG) where the equipment is normally stored and 

maintained (see Figure A-4). See Section A.6 for the detailed description of CEMG function. 

There are five RPC shielding windows and frames are designed to withstand an operating basis 

earthquake (OBE) and design basis earthquake (DBE) or an in-cell blast and still maintain 

confinement. Table A-1 lists the major equipment in RPC.

Table A-1 Major equipment description in RPC [AGNS 1975]

Equipment Operations Descriptions 

Sheared fuel 

diverter

Once the fuel assemblies are chopped, the fuel segments 

are diverted to the dissolver

This diverter has a dimension of 8’9” 

diameter × 11’1” high.

Fuel 

dissolver (15-

C-101, 102, 

103)

The purpose of the dissolver is to perform the following 

dissolution sequence (see Figure A-11 for the dissolver 

flow diagram):

1. Chopped fuel is placed into approximately 3-8 Molar

HNO3 and 3 g/liter gadolinium recirculating dissolver

solution in the dissolver basket.

2. Dissolver product is withdrawn continuously by air

lifting the feed surge tank when solution is within

proper density of ~ 1.49 g/cc.

3. After the dissolve cycle, the dissolved fuel is exposed

to simmer solution (8 Molar HNO3 and 3 g/liter Gd),

which is air lifted to the dissolver section for

dissolving the incoming chopped fuel.

4. Leached hulls are rinsed with hull rinse solution (0.2

Molar HNO3 and 7.5 g/liter Gd).

5. Leached hulls contained in the removable dissolver

basket are dumped in the hull loadout station.

The dissolver includes three identical 

dissolver barrels, each capable of 

operating in dissolve, simmer, or rinse 

modes as previously described. Each 

right cylindrical barrel, which is made 

of 304L stainless steel, measures in 

3’2” in diameter at the top section, 

2’7.5” in diameter at the bottom, and 

15’7” high. Each barrel is also 

accompanied by a heat exchanger 

(15-E-132, 133, 134) to maintain 

proper temperature. This exchanger 

contains a heating capability to 240  ̊F 

during simmer cycle and to cool to 190 

 ̊F during dissolve cycle. The process 

fluid is outside the shell. Heating or 

cooling uses neutron poisoned water 

on the tube side.

High activity 

waste (HAW) 

concentrator

This equipment concentrates the high-activity waste 

steam (which contains about 99% of the nonvolatile 

fission products from spent fuel) from the extraction 

operations (see Figure A-1). The concentrate is sampled 

and then sent to the high level liquid waste storage 

facilities.

This concentrator (41-E-460) is 

constructed of titanium (vessel and 

jacket). It uses a titanium 

thermosyphon reboiler that is supplied 

with 25 psig steam at 260  ̊F. It 

measures 5’3’ in diameter × 18’0.25” 

high. 
A-15

Figure A-11 Typical dissolver flow diagram [AGNS 1975]
A-16

A.1.2 Remote Maintenance and Scrap Cell (RMSC)

This cell is used to manage the scrap from the cladding hulls and other radioactive scrap. In 

addition, this cell is used to decontaminate and repair process equipment. As its name implied, 

all operations are done remotely. All walls and floor are lined with 5-mm thick 304L stainless 

steel. Exposed surfaces other than the stainless steel are coated with phenolic paint. 

Figure A-12 shows the layout of the cell and its equipment. As shown in this figure, both A-A 

and B-B views show the locations of the dissolver basket dumper which is located at the roof 

line of the cell, the cladding hull disposal containers, and the 5-ton overhead crane used in the 

container transfer and for other purposes. When the filled dissolver basket is transferred from 

the remote process cell to this cell where the dumper is located, the dumper rotates the basket 

to allow all cladding scrap to fall into the chute which forms the lower portion of the dumper, 

where it is dropped into an open concrete hull disposal container (measuring 3’ inside diameter 

and 7’4” high) with a total capacity of 52.5 ft3. Sand can be added to the container after 

dumping, if the spent hull fines should become ignited (Zircaloy fines can be pyrophic under 

certain conditions). A total of three dissolver baskets full of the spent hull can be contained 

inside a disposal container. Once it is capped, it is moved into a storage location by a hull 

transfer car. The dumping station is located at one end of the cell and the loadout station at the 

other, where the Cask Loading Station is located. The loadout station is used to load the cask 

with the decontaminated filled disposal containers at the cask loading station above the cell. A 

removable hatch separates this cell and the cask loading station. In addition, a direct straightline 8-inch chute from the Sample and Analytical Cell (SAC) is provided to transfer solid waste 

to the SAC waste retainer in the RMSC. Other activities that can be performed in the cell 

include: examination of partially sheared fuel assembly via the fuel transfer hatch, and conduct 

large equipment maintenance, examination and decontamination from the remote process cell 

to this cell via a hatch located between the floor of the CEMG and the ceiling of this cell.
A-17

Figure A-12 Equipment arrangement in RMSC [NUREG-0278]

A.1.3 High Level Cell (HLC)

This cell is used to provide chemical adjustment and centrifuging of the main process stream 

from the dissolver solution of the remote process cell. It also handles the high and intermediate 

level waste solutions. The cell is designed and constructed (with reinforced concrete) to provide 

confinement during normal, abnormal operations, seismic or meteorological conditions. This 

cell is equipped with a controlled ventilation supply/exhaust system to confine gaseous 

discharges and to transfer such discharges to the process ventilation system for treatment, 

monitoring, and discharge. The doors, hatches and penetrations which are an integral part of 

the structure do not compromise its confinement and shielding functions. This cell is designed 
A-18

for remote operation and contact maintenance. This cell contains a stainless steel floor pan and 

sump to catch any process spills. Exposed concrete surfaces are coated with phenolic paint. 

There are two ways to access this cell. The first one is through a shield door from the Filter 

Piping and Instrument Gallery (FPIG), and the second one is through a removable concrete 

ceiling hatch to the Contact Equipment Removal Station (CERS). This cell contains a fire 

detection and suppression system. In addition, the sumps located in this cell are configured to 

prevent an advertent criticality. This cell has a dimension of 26ft × 28ft × 51.5ft [AGNS 1975]. 

The major equipment in the cell is listed in Table A-2. 

Table A-2 Major equipment descriptions at HLC [AGNS 1975]

Equipment Operations Descriptions 

Hull rinse 

surge tank

It is used in the hull rinse process. It has a dimension of 5’ diameter × 

8’1” high

Dissolver 

acid surge 

tank

It is used to contain the feed from the hull rinse surge tank 

and the recovered nitric acid feed before sending the solution 

back to the dissolver.

It has a dimension of 5’10” 

diameter × 9’10” high [NUREG0278]. 

Dissolver 

head pot tank

It is a disengaging vessel for air lefts and steam jets. A total of four such tanks exist in 

the cell. Each measures 1’1” 

diameter × 3’8” high [NUREG0278] 

Accountability 

tank

It provides input accountability point. Dissolver solution is 

mixed and then accurately measured for volume and density, 

and sampled for chemical analyses.

It measures 6’2” diameter and 

14’9” high [NUREG-0278]

Feed 

adjustment 

tank

It provides surge and batch adjustment of acidity and 

concentration of feed to recovery section.

A total of 2 tanks are located in 

the cell. Each measures 6’ 

diameter × 15’8” high [NUREG0278] .

General 

purpose (GP) 

concentrator 

feed tank

It receives process wastes continuously and others on a batch 

basis. It also receives feed of oxalate solutions from the silica 

gel bed regeneration and low activity general process wastes. 

The oxalate is destroyed by periodically raising the solution 

temperature in the tank to boiling, and adding hydrogen 

peroxide.

It measures 8’ diameter × 15’1” 

high [NUREG-0278].

A.1.4 High Intermediate Level Cell (HILC)

This cell is to provide the solvent extraction of uranium and plutonium from fission products, 

separation of uranium from plutonium, concentrating intermediate-level liquid wastes and 

solvent cleanup and storage. The high activity (HA) centrifugal contactor (see Figure A-13) and 

IB electro-pulse column are installed in removable shield plugs in the ceiling. Most of the 

process equipment in the cell is constructed of stainless steel. The IB electro-pulse column has 

titanium and platinum internals. Table A-3 lists the major equipment in the cell.

Under normal process conditions, uranium concentrations are maintained well below 75% of the 

minimum critical concentration corresponding to the specified maximum enrichment. In the 

event of DBE, major components including the connecting piping must withstand the accident 

condition. In addition, a sump located on the cell floor contains instrumentation for detection of 

the process losses. Two jet sump transfer systems are used to remove liquids from the sump: 

one system is designed to operate under all abnormal conditions to transfer the liquid to the 

HLLW tanks, and the other is designed for normal operating transfers to the sump collection 

tank. It has a dimension of 40ft × 27ft × 52ft [AGNS 1975].
A-19

Figure A-13 HA centrifugal contactor in HILC [AGNS 1975]
A-20

Table A-3 Major equipment descriptions in HILC [AGNS 1975]

Equipment Operations Descriptions 

Centrifugal 

contactor (21-C225)

After acid adjustment, the feed (about 70% 

is the adjusted dissolver product from the 

HAF feed tank and about 30% is the HS 

column raffinate) is clarified by this 

equipment, and the solution is contacted 

counter-currently in 10-stage with an 

organic solution of 30% tributyl phosphate 

(TBP) in normal paraffin hydrocarbon 

diluent (dodecane). The organic solution 

preferentially extracts the tetravalent 

plutonium and hexavalent uranium, leaving 

about 95% of fission products in the 

aqueous solution. The aqueous solution is 

feed to the HAW surge tank in RPC. The 

organic solution is feed to HS column.

It is about 3’1” diameter × 4’9” high for the 

section below the shield plug (see Figure A-12). 

The contact time between phases is 10 minutes, 

compared to two hours in typical pulse columns. 

The reduction of contact time radically reduces

solvent degradation due to radiation. The 10-

stage centrifugal contactor used was developed 

and is used in France at the La Hague Plant. 

HS column The organic solvent stream from the 

centrifugal contactor is heated to 60  ̊C to 

promote back-extraction of additional fission 

products during the scrubbing action of a 

relatively high acid stream (3 Molar HNO3).

It measures 1’10” diameter × 46’9” high 

[NUREG-0278]. This equipment has been met 

the criticality safety requirement, which is based 

on the fissile material concentration expected. 

Electro-pulse 

column

This column is used to partition uranium 

and plutonium without the use of chemical 

reducing agents such as ferrous sulfamate. 

The concept is based on electrolytic 

reduction of U+6 to U+4. The reduced U 

then reduces Pu+4 to Pu+3 so that the 

plutonium is extracted into the aqueous 

phase, effecting a separation of Pu and U. 

The Pu stream is transferred to the 1BP 

surge tank before sending to the PPC.

The plutonium concentrations in the electropulse column are maintained at approximately 

75% of the minimum critical concentration. For 

the abnormal condition where feed to the 

column continues, but no plutonium is removed, 

the time required to achieve a critical 

concentration is conservatively estimated to be 

> 6 minutes. Alarm system is in place to ensure 

continuously flow of plutonium out of the 

column. In addition, hydrogen can be generated 

in this column under certain conditions. The 

column vent is continuously monitored to detect 

any potential explosive conditions.

General purpose 

(GP) concentrator

It receives feed from the GP concentrator 

feed tank from HLC by a continuous air lift. 

The concentrator liquid is processed to a 

specific gravity of 1.35, at which time the 

concentrate is batch transferred to GPW 

check tank. 

It measures 7’8” diameter × 15’ high [NUREG0278]. Note that steam to the concentrator is 

supplied from the 25-lb closed loop system. 

This ensures that processing temperatures 

below 135  ̊C (minimum required for “red oil” 

explosion) are maintained. This concentrator is 

equipped with interlock systems that shut off the 

steam to the concentrator reboiler automatically 

in the event that pressure or temperature 

parameters are exceeded. In addition, sugar 

solution is added to the concentrator 

continuously to maintain a residual nitrite 

concentration and suppress ruthenium 

volatilization. Iodine in the feed may be 

removed by mercury addition before discharge 

to the stack.

GP waste check 

tank

It receives waste from the GP concentrator. 

It also receives mercury-bearing solutions 

from iodine scrubbers and purged solutions 

of the acid fractionator from the ILC. When 

it is full, the concentrate is sampled for 

It measures 2’10” diameter and 6’8” high 

[NUREG-0278].

. 
A-21

Table A-3 Major equipment descriptions in HILC [AGNS 1975]

Equipment Operations Descriptions 

accountability, H+ concentration and 

activity. Then it is transferred to the 

intermediate level liquid waste tanks. 

1SF tank It collects all aqueous process wastes, 

except those in the HAW stream. It feeds to 

the 1S column.

It measures 3’ diameter × 31’7” high [NUREG0278]. This tank is designed to safely contain 

higher plutonium concentrations than the 1S 

column, since this tank feeds to the 1S column. 

Measurement is done to monitor the contents. 

Fixed poison of Borosilicate glass raschig rings 

is added to the tank to minimize criticality 

potential.

1S column It functions to recover fissile materials from 

wastes generated by 2nd uranium cycle, and 

2nd and 3rd plutonium cycle. The fissile 

material is extracted as the organic phase 

via air lift from its top to the 1BU recycle 

tank. The remaining aqueous waste is air 

lifted to the LAW concentrator feed tank in 

ILC.

The main section is about 1’ diameter × 25’ high 

with a bottom disengaging section ~ 3’ diameter 

× 3’ high.

1BP Surge tank It receives the Pu stream from the electropulse column. It measures 4’ diameter × 14’8” high. 

Borosilicate—glass Raschig rings are added to 

the vessel for criticality safety.

A.1.5 Intermediate Level Cell (ILC)

This cell houses the off-gas treatment, spent solvent treatment and combustion, acid recovery, 

and low-level liquid waste concentration. All operating activities are performed remotely. Most 

equipment and piping are constructed of 304L stainless steel; otherwise it is noted. This cell 

has a dimension of 27ft × 38ft × 52ft. The major equipment in the cell is listed in Table A-4. 

Table A-4 Major equipment descriptions in ILC [AGNS 1975]

Equipment Operations Descriptions 

NO2 absorber It is a dissolver off gas absorber to scrub any 

NO2, primarily from the dissolvers.

It measures 5’ diameter × 31’7” high.

LAW 

concentrator, feed 

tank, reboiler

It concentrates feed from the NO2 absorber, 

and the acid stream from the HILC. 

Approximately the concentrate is about 20 fold 

than the aqueous waste from the feed tank. 

Once it is distilled, the overhead vapor 

containing about 3% nitric acid in water vapor 

is fed to the acid fractionator. The concentrate 

is sent to the LAWB check tank.

The concentrator measures 7’10” diameter × 

20’8” high. The reboiler measures 4’ 

diameter × 13’8” high. The feed tank 

measures 7’4” diameter × 16’6” high 

[NUREG-0278]. To ensure that the 

plutonium concentration is critical safe. 

Soluble poison may be added to the LAW 

concentrator feed tank based on the 

plutonium monitor upstream to the feed tank.

LAWB check tank This tank samples the feed, then the feed is 

transferred to the HAW surge tank in HLC.

It is about 3’ diameter ×6’ high. This tank 

equipped with a sampler, a cooling water 

jacket and air sparger.

Acid fractionator, 

reboiler, and 

condenser

The fractionator is used to concentrate and 

recover nitric acid. The reboiler supplies heat 

to distill and fractionate dilute nitric acid to form 

The fractionator measures 5’6” diameter and 

38’8” high [NUREG-0278]. It is constructed 

of titanium. Mercury addition may remove 
A-22

Table A-4 Major equipment descriptions in ILC [AGNS 1975]

Equipment Operations Descriptions 

12 Molar acid and water. In addition, the 

fractionator has draw offs to remove chlorides 

(nitrosyl chloride), thus preventing 

accumulation of corrosion contributing 

chlorides. Chloride concentrations above 0.05 

wt% are corrosive. Therefore, purging is 

performed once a month to reduce this 

concentration below this limit. The purged 

solutions are sent to the GPW check tank in 

HILC. The condenser is used to condense 

water vapor from the fractionator which is??

collected in the accumulator, and routed to the 

recycle water head pot. Then any water vapor 

is discharged through the stack.

iodine remaining in the feed in the acid 

fractionator overhead vaporizer before 

discharge to the atmosphere through the 

stack. The addition of sugar for the vaporizer 

also suppresses ruthenium volatility if any. 

The control of minimum acidity in the reagent 

streams is to prevent plutonium 

polymerization using the redundancy in the 

acid strength control devices.

Recovered acid 

storage tank

It stores recovered acid from the bottom 

products of the acid fractionator.

It measures about 7’4” diameter × 18’2” high 

[NUREG-0278].

Solvent batch 

stripping tank

It collects all organic wastes and acts as a 

central point. Types of chemicals used to strip 

the fissile materials and to decontaminate the 

organic waste vary. Repeated separate 

contacting with nitric acid and sodium 

carbonate is sufficient to decontaminate the 

solvent; however, oxalic acid and caustic can 

generally be used if additional and more radical 

treatment is required. The aqueous waste 

solutions from the solvent treatment are 

transferred to the GP concentrator feed tank in 

HILC. 

It measures 3’9” diameter ×7’10” [NUREG0278].

Solvent burner 

feed tank

It receives treated waste solvent from the 

stripping tank. It has a capacity of ~4300 

gallons. Once it is full, it feeds to the solvent 

burner.

It measures 8’2” diameter × 16’6” high 

[NUREG-0278].

Solvent burner 

and quench pot

The burner is operated at 1550  ̊F, using 

propane gas. Once the combustion chamber 

is maintained at this temperature for 30 

minutes, the waste solvent flows into the 

burner. Any P2O5 produced in the burner is 

converted to phosphoric acid in the quench 

pot, and sent as ILLW.

Both solvent burner quench pot and waste 

solvent burner have a combined dimension 

of 5’11” ×19’10” high. Both temperature and 

pressures in the burner, its blower and 

quench pot and air flow are included in the 

monitoring parameters in the safety interlock 

system. If any of these parameters are out 

of the tolerance condition, the burner will 

shut down. The alarm system is used to 

provide alerts to the operators.

A.1.6 Uranium Product Cell (UPC)

This cell provides a purification of the uranium product stream. Basically the aqueous strip 

solution containing the uranium is concentrated from 0.3-Molar to 1.5-Molar uranium and 

adjusted with nitric acid to ~2.05 Molar, and is again preferentially extracted by another 30% 

TBP organic solution in a pulsed column. Before leaving the column, the organic solution is 

scrubbed with 0.01 Molar and 2.3 Molar nitric acid solution to remove extracted ruthenium and 

zirconium-niobium. Hydroxylamine nitrate and hydrazine are also added to the scrub solution to 

remove any residual plutonium. Uranium is stripped from the organic solution in another pulsed 

column, using acidified water (0.01 Molar HNO3). The solution is concentrated by evaporation 

from 0.4 Molar to 1.5 Molar uranium. Finally, the concentrated uranium solution is passed 
A-23

through silica gel beds to remove residual traces of zirconium-niobium. Uranyl nitrate product is 

analyzed and transferred to the UF6 facility for shipment or conversion. This cell has a 

dimension of 31ft × 27 ft × 52 ft. The major equipment for this cell is provided in Table A-5. 

Table A-5 Major equipment* descriptions in UPC [AGNS 1975]

Equipment Operations Descriptions 

Decanter This removes organic from the 

feed to the concentrator and thus 

optimizes operability of the 

concentrator. 

Two decanters (1CU and 2EU) exist in this cell. Each 

measures at 1’2” diameter × 4’6” high.

Concentrator 

with 

condenser 

and reboiler

This system provides an effective 

control of the uranium 

concentration in the feed, and 

maximizes the decontamination 

factor in the extraction cycle. Each 

concentrator contains three 

stripping bubble caps, trays and 

mist eliminator York mesh-421 or 

equivalent, vortex breaker in lower 

section.

The two reboilers are vertical thermosyphon vessels that heat 

the feed to the concentrators to 116 F for 1CU system and to 

126 F for the 2EU system. The 1CU concentrator is about 17’ 

overall, top section contains a 5’6” diameter × 5’3” high, and 

bottom section contains a 2’6” diameter × 16’6” high. There 

are two systems (1CU and 2EU). Both condensers are to cool 

1CU concentrator overheads at 5 MTU/day to temperature of 

110 F. Cooling water at 85 F is used. 1CU condenser 

measures at 19” diameter × 168” tube length. Note that 

concentration control is used to ensure no criticality event. 

The parameters monitored include solution density, feed flow, 

process liquid temperature, and steam pressure. The 

monitoring is done in the redundant safety instrumentation and 

control. The system is shutdown when 87% of critical 

concentration is reached. In addition, if a high-pressure or 

high-temperature condition is detected in the concentrator, the 

reboiler stream is cutoff automatically.

U product 

cooler

This cooler provides cooling to the 

end product prior to transfer to the

catch tanks.

It measures at 1’ × 7’6” high.

U product 

catch tank

This tank is used to collet uranium 

product.

It has a volume capacity of 1210 gallons, which measures at 

5’6” diameter ×8’ long.

U product 

sample tank

It holds collected U product 

solution for sampling.

It has a volume capacity of 1210 gallons, which measures at 

5’6” diameter × 8’ long. Eight-hour production run batches 

may be collected and sampled.

Silica Gel 

beds

These beds are the final treatment 

for separating Zr and Nb from the 

U solution

Each bed measures about 1’8” × 13’ high. These beds are 

provided and are used alternately to permit continuous 

operation. If the processing rate of 5 MTU/day is used, the 

spent silica gel adsorbent unit is regenerated about once a 

week by flushing the bed with a 0.4 Molar oxalic acid. The 

regeneration solution is directed to GP concentrator in the 

HILC for oxalic acid destruction and concentration of the 

solution.

*All equipment is constructed using the 304 stainless steel.

A.1.7 Plutonium Product Cell (PPC)

This cell provides the purification of the plutonium process stream. Plutonium aqueous stream 

leaving the partitioning column (electro-pulse column in HILC) is re-oxidized to the extractable 

tetravalent state with nitrate and is preferentially extracted into the TBP organic solution in the 

pulsed extraction column off the second plutonium cycle. This resulting organic solution is 

further scrubbed with 1.0 Molar nitric acid solution to remove extracted ruthenium and 

zirconium-niobium in the top portion of the column. Then this solution passes through a strip 

column where tetravalent Pu is transferred to an aqueous stream of dilute (0.3 Molar) nitric acid. 

A 3rd plutonium cycle is repeated for extraction and scrubbed as in the second cycle to remove 
A-24

any fission product. A 3rd strip column is used to achieve higher Pu concentration using 

hydroxylamine nitrate to convert Pu to the strippable trivalent state. A TBP organic scrub 

solution also is used to remove any uranium from the plutonium aqueous stream as it leaves the 

3rd strip column. After TBP is removed, the final plutonium concentration from 60 g/liter to 350 

g/liter is achieved in a concentrator. After analysis, the final plutonium solution is stored in the 

product slab tanks before sending to the PNSL. Because of the concentration level (much 

above the ANS-8.1 subcritical limit), tanks must be criticality safe. This cell has a dimension of 

20ft × 27ft × 51.5ft. Table A-6 lists the major equipment in this cell.

Table A-6 Major equipment used in PPC [AGNS 1975]

Equipment Operations Descriptions 

2A/2B columns These columns provide 2nd extraction cycle. It 

receives feed from the 1BP surge tank in HILC. The 

resulting Pu concentration is 10-20 g/liter.

2A measures 3’5” diameter × 43’7” 

high, and 2B measures 3’5” diameter 

× 30’3” high. Note that the 

concentration of Pu in the organic 

solvent exiting the 2B column exceeds 

0.01 g/liter, an alarm is initiated to alert 

the operator of upset conditions. 

3A/3B columns These columns provide 3rd extraction cycle. They 

receive feed from the 2nd extraction cycle Pu stream. 

The resulting Pu concentration is about 40-50 g/liter.

3A measures 3’5” diameter × 43’7” 

high, and 3B measures 4’ diameter 

×46’3” high. Same alarm is issued for 

the Pu concentration in the organic 

solvent exceeding 0.01 g/liter.

3P concentrator, 

reboiler

This concentrator provides Pu concentration up to 

350 g/liter.

The combined dimension is 4’6”×13’9” 

high.

Pu product 

storage tanks

These tanks provide an interim storage before 

transferring the Pu product solution to the PNSL.

Each measures 2.25”×7’6”×11’4”.

A.2 Fuel Receiving and Storage Station (FRSS)

This station is designed to receive, handle and temporarily store spent fuel assemblies from 

nuclear light-water power reactors (see Figure A-2 to Figure A-7). A detailed top view and side 

view of the FRSS are shown in Figure A-13 and Figure A-14, respectively. Spent fuel 

assemblies are received in shielded casks by either truck or rail and delivered to FRSS in one of 

two bays as shown in this figure. 
A-25

Figure A-14 Detailed top view of the FRSS [NUREG-0278]
A-26

Figure A-15 Side view of FRSS [NUREG-0278]
A-27

The fuel storage pool consists of 2 cask unloading pools where casks are opened, 1 fuel 

storage pool, 1 failed fuel pool where leaking fuel assemblies can be canned or uncanned, and 

1 fuel transfer pool where fuel assembly is transferred to RPC. A 135-ton capacity cask 

handling crane is used for moving casks in and out of the cask unloading pool. Analysis of the 

dropped cask accident in the Unloading Bay the Test and Decontamination Pit or 

Decontamination Pit indicated that the integrity of the pool recirculation system can be 

breached. The water could be drained by gravity flow from the rupture lines; however, the 

placement of these lines prevents the pools from draining below10 feet of water shielding over 

the storage fuel. An aluminum storage canister is used to accommodate 8 pressurized water 

reactor (PWR) or 18 boiling water reactor (BWR) fuel assemblies. This canister is designed to 

provide geometric spacing for criticality and to withstand highest handling height drop to another

canister at the highest rated speed of the canister handling crane without canister distortion. 

Spacing between loaded canister compartments is limited to the multiplication factor ≤ 0.95. In 

addition, administrative procedures limit the movement of more than one fuel assembly in either 

cask unloading pool or transfer pool at a time.

Underwater hoods are provided as a means of containing the gaseous fission products during 

storage in the fuel storage pool. Any gases collected by the hood will be transported to the offgas treatment system. Fuel assemblies are unloaded and stored under water to provide cooling 

and shielding. The 12’ water shielding is provided in the pools which limit the surface dose to a 

calculated 0.08 mrem per hour. Handling systems are designed with special limit switches and 

mechanical stops to prevent fuel from raising fuel higher than the design depth of the shielding 

water. Storage capacity is 360 metric ton uranium of fuel. 

The pool water is channeled and treated to promote maximum clarity, to control temperature, 

and to minimize corrosion and radioactivity by using continuous filtration through 95% efficient 5 

μm pore size filter elements, cooling in heat exchanges to hold water temperature below 105  ̊F 

and demineralization. All expected wet surfaces in the FRSS is covered by a stainless steel 

liner up to the expected highest water mark.

Note that the FRSS structure is safety related up to 25 feet above the berm level and non-safety 

related above this level. The conditioned ventilation air to the FRSS is provided from an 

independent unit and a major portion of the air is channeled to the analytical filter station, is 

gathered and released to the stack. Dropped fuel element accident has been included in the 

safety analysis report, which shows the immediate release is negligible. A list of the major 

equipment in the FRSS is provided in Table A-7.
A-28

Table A-7 Major equipment description in FRSS [AGNS 1975]

Equipment Operations Descriptions

Cask 

handling/unloading 

cranes

These cranes are used for cask operations including 

unloading cask contents.

It includes the supports and 

restrainers to be used in cask 

operations. One crane has a capacity 

of 135 tons.

Fuel storage 

canister crane

This crane is used in the canister operations. It includes the supports and 

restrainers to be used in canister 

operations (15-ton capacity)

Fuel storage 

canisters

They are used to store fuel assemblies inside the 

spent fuel pool.

Each canister can accommodate 8 

PWR or 18 BWR fuel assemblies. 

The design of the canister and the 

spacing in the pool accounts for 

criticality concerns.

Fuel transfer 

conveyer

It is used to transfer fuel assembly from FRSS to 

RPC for processing

It is a cart-track arrangement which 

transports fuel assemblies from a 

vertical position in the fuel transfer 

pool to a horizontal loading position in 

the RPC.

Spent fuel pools These pools provide radiological shielding and 

cooling for the stored spent fuel. The failed pool is 

intended to handle the leaking fuel pins.

Spent fuel poolis divided into five 

pools consisting of two cask 

unloading pools (CUP), one fuel 

storage pool, one failed pool, and one 

fuel transfer pool. Normally, 12-ft of 

water shielding is provided for the 

pools.

A.3 Control Room Area/ Plutonium Nitrate Storage and Load-out

Both Control Room Area (CRA) and Plutonium Nitrate Storage and Load-out (PNSL) are 

located outside high radiation process area, where thick concrete structures are used to provide 

adequate shielding for radiation (see Figure A-2 to Figure A-7). The description of CRA is 

discussed first, followed by the description of the PNSL.

A.3.1 Control Room Area (CRA)

CRA is an important area that houses the process control instrumentation, display and 

monitoring panels for all safety-related instruments within the process facility. This area also 

serves as the communication center for which operators can be directed to perform manual 

functions. An electrical cable distribution room is below this area providing space for wire and 

cable routing (see Figure A-7). 

A.3.2 Plutonium Nitrate Storage and Load-out (PNSL)

This operation is to receive plutonium nitrate (250g/liter Pu at 3-Molar HNO3) from the interim 

storage tanks in the PPC and store in geometrically favorable storage tanks within two storage 

cells. A total of 24 such slab tanks (each has a capacity of 232 gallons) distributed into four 

modules in a cell. All safety related storage tanks are constructed with titanium to minimize 

corrosion and contamination. The stored nitrate solution is cooled by a filtered, chilled 

recirculating air system to maintain temperature at about 140  ̊F. The nitrate solution is loaded 

from the storage tanks in the Plutonium Nitrate Cell (PNC) into a 50-liter sample tank, to the 10-

liter load-out tank, and finally to 10-liter capacity shipping containers. Optionally, the nitrate 

solution from PPC can be transferred directly to the 50-liter sample tank. The shipping 
A-29

containers are decontaminated and stored (up to five containers) in the Plutonium Nitrate 

Storage Area (PNSA). Loading onto the truck for shipment is also done in this facility. Similarly, 

neptunium nitrate solution can be stored and processed in this facility.

A.4 Waste Tank Equipment Gallery (WTEG)

This gallery serves as the radioactive liquid storage processing facility, which houses 

processing equipment and utilities to store radioactive liquid waste. This building has an area 

about 30 m2, divided into three main sections: (a) a cold area which is non-radioactive, (b) a hot 

area which is radioactive, and (c) an intermediate area [NUREG-0278].

The cold area houses the waste tank operation instrumentation, which is used to monitor liquid 

level, specific gravity, temperature and pressure of the tank content.

A.5 Waste Tank Cells (WTCs)

This cell contains two high-level liquid waste (HLLW) storage tanks and two intermediate-level 

liquid waste (ILLW) tank. Only one tank is used at the time. The other tank is a spare. These 

tanks form the waste tank cells. Radioactive liquid is transferred from the main process building 

to WTEG through underground pipe vaults and distributed to the tanks through underground 

lines via the diverter cell in the WTEG.

A.5.1 High-Level Liquid Waste

Each stainless HLLW tank is 16.5 m in diameter by 6.1 m high with a capacity of 300,000 

gallons. Each tank contains 48 5-cm diameter cooling coils, 18 air-operated ballast tanks round 

the perimeter of the tank, 9 air-operated ballast tanks in the main part of the tank, 22 air lift 

circulators, 5 steam-operated ejector pumps, water-seal type pressure/vacuum relief system, 

external temperature sensors, and 10 instrument dip tubes to measure liquid level and specific 

gravity. This tank is contained within an underground, cylindrical, stainless steel lined (up to 5.5 

m on the vault side), concrete vault, which measures 18.3 m diameter and 7.6 m high. The 

thickness of concrete for the vault is 1.2 m, 0.9 m and 1.7 m for the floor, side and top, 

respectively. The tank is sloped inside the vault to facilitate drainage in the unlikely event of a 

leak. A sump around the perimeter of the vault is provided for containing any leak. Figure A-16

provides a cut-away view of the HLLW tank section.
A-30

Figure A-16 Cut-away view of the HLLW tank section [NUREG-0278]

The stored HLLW will be maintained below 140  ̊F and will not be allowed to reach boiling 

temperature under any credible set of circumstances. The cooling in the tank is provided by the 

48 cooling coils. Settling of the solids in the tank is prevented by using the ballast tanks, and air 

lift circulators. Any airflow through the waste solution tends to concentrate the solution by 

evaporation, which can amount to as much as 500 gallons/day/tank. If over-pressurization of 

the dissolve-off gas/vessel-off gas system occurs, which is an unlikely event, an automatic 

diversion of the HLLW off-gas to the UPC occurs, which will provide for de-entrainment of 

aerosols. Once the pressure surge has been relieved, normal venting of the HLLW-off-gas 

would be re-established.

A.5.2 Intermediate-Level Liquid Waste

The construction of the ILLW tank is very similar to the construction of the HLLW tank, except 

that ILLW tank does not have cooling coils, air lift circulators or ballast tanks, because the 

expected dose in the ILLW tank is much lower than that of HLLW. To minimize any solid 

settling, an air sparge system is used to agitate the waste. The feed for ILLW include the 

solvent treatment, laboratory, cell/equipment decontamination, blowdown from the acid 

fractionator overhead vaporizer and from solvent burner quench pot, off-gas scrubber solutions, 

HLLW off-gas condense, and silica gel regeneration from UPC. All of the ILLW solution is 

transferred to the tank from the GPW check tank in the main process building. All 
A-31

interconnecting lines between the process building and the diverter, and between the diverter 

and the ILLW vault are enclosed within stainless steel lined concrete pip vaults. As in the 

HLLW, ILLW is equipped with instrument dip tubes and thermocouples.

The tank is designed to provide a narrow operating range with respect to the vapor phase 

pressure, which the vapor space operating pressure is not expected to be positive to the 

surroundings. Thus pressure operation is never allowed. The designed operating pressure is 

minus 3 inches of water to plus 10 inches of water.

A.6 Galleries

The galleries provide enclosure and protection for piping, process support equipment, and 

instrumentation components. These galleries are used by operating personnel and are used for 

many operations [NUREG-0278].

• Head Tank Galley/Pulser Equipment Gallery – They are used to house air pulsers for the

solvent extraction columns, recovered acid tanks, recycled water tanks, and the cold offgas blower. The galleries are located above, and provide access to, the process cells

(see Figure A-5). They are constructed of reinforce concrete with 0.5 m thick walls.

• Piping and Instrument Gallery – This gallery contains piping, valves and process control

instrument to serve the adjacent cells. It is located the north side of the process cells

(see Figure A-3 and Figure A-4). As shown in Figure A-3, the lower piping and

instrument gallery is identified.

• Filter Piping and Instrument Gallery (FPIG) – This gallery contains piping, valves and

process control equipment necessary for the operation of the adjacent cells. It also

contains contact cell ventilation air filter housing, Sample and Analytical Cell filter

housings, Plutonium Product Cell filter housing, the 2.7 atmosphere steam system, and

miscellaneous pumps (see Figure A-2)

• Crane and Equipment Maintenance (CEMG) Gallery – This gallery serves the crane

maintenance for the RPC. A removable shielding door divides between this gallery and

RPC (see Figure A-4). This is also used for equipment items entering and leaving RPC.

This gallery is also interfaced with RMSC below through floor shielding plugs that occupy

roughly about half of its floor space. A shielding personnel access door, a workbench,

lifting device for the floor plugs, crane maintenance access platform, and an ultrasonic

cleaning tank (0.5 m3 capacity) can be found in this gallery.

• Service Concentrator Gallery – This gallery houses service concentrator equipment,

which is isolated by concrete walls and an airlock and is located on the southwest corner

of the lower viewing and operating station (see Figure A-2). Both floor and walls are

covered with acid-resistant paint up to a height of 0.9 m. Floor drains are provide to the

feed tank to route any leakage back to the service concentrator or for transfer to the

general purpose concentrator.

• Hull Monitor Maintenance Gallery – This gallery is used to maintain hull monitor and

drive assembly, and is located between the PRC and RMSC, and is accessed from the

lower viewing and operating station (see Figure A-2).
A-32

A.7 Stations

Stations are locations around the process cells that provide various functions for the facility. 

The major stations are described below [NUREG-0278]:

• Cask Loading Station (CLS) – This station is located east of the RMSC (see Figure A-4,

Figure A-6 and Figure A-7), which is used to remove loaded solid scrap disposal

containers and to secure them in the shipping cask for transport to the onsite

underground solid waste storage area. Empty disposal containers also enter the RMSC

from this station. Decontamination of the shipping cask is also done in this station.

• Viewing and Operating Stations – These stations are located on the south side of the

RPC and RMSC (see Figure A-2 to Figure A-5). These stations are:

o Lower (LVOS) – It has an area of ~4.9 m wide which bounds the RMSC on the

south side and west end, and the lower portion of the RPC on the south. Its

function is to load leached hulls in interim burial containers and remote

maintenance of equipment. It also contains 4 view windows that are 0.6 m by 0.7

m lead glass, oil-filled shielding windows installed in the concrete wall for

observing activities within the RMSC. At each view window there are two

master-slave manipulator wall tubes and one periscope wall tube. Additional

periscope wall tube is provided for reviewing the hull dumping operation. A port

for passing items into or out of RMSC is located in the east wall of this station.

o Grade (GVOS) – It has an area of ~4.9 m by 40 m, separated from the south side

of the RPC and CEMG by a concrete wall (1.5 m thick). This station is used to

conduct remote maintenance for the CEMG and operation of fuel shear

mechanism and subsequent transport of leached hulls to the hull dumper. Four

view shield windows are located in this station to observe operations in PRC and

CEMG.

o Analytical (AVOS) – It has an area of ~5.1 m by 55 m, located north of the SAC.

The operations in the SAC are conducted from this station through shielded view

windows and manipulators. This station also contains glove boxes for taking

samples from the UPC, PPC, and ILC.

o Top (TVOS) – It is a 4-m diameter circular area in the cold chemical station. It

provides top viewing of the RPC operations. A shielding window is provided on

the station’s floor. Four penetrations to the floor for supplying auxiliary services

to the RPC.

• Contact Equipment Removal Station (CERS) – This station provides access into the

process cells below (see Figure A-5). Materials may be installed in and removed from

these cells through shielding hatches in the cell ceilings. This station is served by a 55

metric ton crane.
A-33

• Equipment Maintenance Station (EMS) - This station performs final decontamination and

repairs where items from the cells are repaired, prepared for transferring out of the

building, or packaged for transfer to the underground waste storage. Occasion

maintenance activities and repair of equipment can be done in this station. Concrete

floor plugs provides access to the CEMG. This station contains a 55 metric ton crane for

any major lifts as needed. This station is located above the CEMG and is located the

top floor of the process building (see Figure A-5).

A.8 Sample and Analysis Cell (SAC)

This cell is used to provide a shielded area for remote sampling and analysis of highly 

radioactive materials from the process cells. It contains eight compartments, each having 

master-slave manipulators and viewing windows (see Figure A-5). Each compartment has a 

floor drain leading to the general concentrator for liquid waste, and a stainless steel liner to the 

floor and part of the walls. A power hoist is used to transport large items of equipment from 

compartment to compartment. The following functions are identified for the compartments:

• Two compartments contain samplers and equipment for remote sampling of the process

streams and tanks.

• One compartment contains sample storage capability and equipment for disposal of

surplus samples to a recovery tank.

• Two compartments contain equipment for remote radiochemical analyses.

• One compartment is used to provide decontaminated samples and to prepare weighed

dilutions of samples to be analyzed in the HCLA.

• One compartment provides equipment for processing chemistry studies.

• One compartment has facilities to transfer equipment to and from the cell via a transfer

drawer, and a scrap chute to the RMSC for disposal of canned solid waste.

A.9 Support Systems

This section describes the support systems used at the facility. These systems include the 

ventilation system, electrical power system, fire protection system, water reservoir, and 

laboratory area (hot and cold).

A.9.1 Ventilation System

This system provides pretreated airflow to the processing areas, maintains pressure differentials 

between zones as required to confine radioactivity to designated locations. The ventilation 

systems are designed to confine and channel any airborne release to treatment systems such 

as filters. Cooling is provided by the ventilation systems for the process cells. Figure A-17

shows the air flow diagram of the ventilation system. As shown in this figure, the air circulation 

for the FRSS is about 84,000 cubic feet per minute (CFM) and the process building air 

circulation is about 92,000 CFM. Table A-8 provides some operating conditions of the 

ventilation systems.
A-34

Figure A-17 Air flow diagram in the ventilation system [AGNS 1975]
A-35

Table A-8 Operating conditions of the ventilation systems at selected 

locations (see Figure A-17)* 

Location Temperature 

( ̊F)

∆Pressure

(inches of water 

gauge)

Zone Volume

(ft3) 

Remote Process Cell 130 -2 68,000

Remote Maintenance and Scrap Cell 100 -2 36,000

High Level Cell 130 -2 37,500

High Intermediate Level Cell 140 -2 56,200

Intermediate Level Cell 135 -2 53,300

Uranium Product Cell 143 -2 43,500

Plutonium Product Cell 140 -2 28,200

Waste Tank Equipment Gallery 93 -0.25 -- 

Waste Tank Cell 140 -2 110,000

Crane and Equipment Maintenance Gallery 76 -0.5 20,200

CERS/EMS 73 +0.25 33,100

*Both temperature and pressure shown are design values. ∆Pressure is based relative to atmosphere and volume

of zone shown is respective location.

According to [AGNS 1975], the supply system fan is sized to produce a total differential head of 

6 inches of water gauge and deliver air at its discharge at +2.7 inches of water gauge. In the 

event the supply unit is off and the process building exhaust blowers continue to operate, the 

areas of the process building that are normally slightly positive will see negative pressure 

somewhere between atmospheric and -6 inches of water gauge. There are a total of three 

exhaust blowers. One is in stand-by mode. It would start up automatically on low negative 

pressure at the outlet of the primary HEPA filter bank in the process building. Figure A-18

shows the blower station building. As shown in this figure, this building contains the final filter 

banks, which are enclosed inside a concrete room. It also houses the process exhaust blowers, 

analytical exhaust blowers. The process building ventilation supply is also shown.
A-36

Figure A-18 Blower station building [NUREG-0278]

A.9.2 Off-Gas System

This system provides treatment of off-gases to remove contained particulates, iodine and oxides 

of nitrogen. There are two major off-gas systems: a dissolver off-gas (DOG) and vessel off-gas 

(VOG). The scrubbed and filtered gaseous effluents from these systems are released to the 

atmosphere via the stack (see Figure A-19). As shown in Figure A-19 [NUREG-0278], there are 

two types of off-gas systems (DOG and VOG). Note that this figure is slightly different from 

what was reported in [NUREG-0278], because solidification processes were not included in the 

original BNFP. Each of the two systems is described below:
A-37

Figure A-19 Major off-gas systems
A-38

The DOG system provides treatment of off-gases from the dissolvers, high-activity waste 

concentrator condenser, plutonium vent knockout pot, acid fractionator condenser, and general 

purpose concentrator condenser. It contains an iodine scrubber, which uses the circulating 

mercuric nitrate-nitric acid solution to remove radioactive iodine. Since of the nitrogen oxides 

are primarily in the dissolvers, the NO2 absorber is included in the DOG.

The VOG system provides treatment of off-gases from the high and intermediate-level liquid 

waste storage tanks, instrument purges, samplers, air lifts, sparges, steam from blowing or 

venting jets, and the waste solvent burner.

A.9.3 Electrical Power System

Normal electrical power is fed to the plant from the commercial substation. In the absence of 

the normal power, the emergency power is supplied by two independent diesel engine-driven 

generators. Within 10 seconds of normal power loss, both generators will start automatically. 

DC power is provided for instrumentation in the main control room and the waste tank 

equipment gallery area.

A.9.4 Fire Protection System

The source of the water for the fire protection system is based on the 60.5x106-gallon Beacon 

Pond in the area of the plant [AGNS 1975]. 

The process cells are protected by individual safety automatic fire detection and suppression 

system in each cell: Uranium Product Cell, Intermediate Level Cell, High Level Cell, High 

Intermediate Level Cell, and Plutonium Product Cell. Each cell is provided with a dual fire 

detection system consisting of one smoke detector and one combination rate-of-rise and fixedtemperature detector. The suppression system uses Halon 1301 as the fire suppressant agent.

Other areas in the plant are protected by the fixed wet-pipe sprinkler systems, which include the 

storage building, galleries, pump house, analytical filter station and ventilation filter station.

A.9.5 Hot and Cold Laboratory Area

This area contains 13 individual laboratories (two-story complex) that are equipped to provide 

specific types of analyses or services. One laboratory is equipped for receiving samples from 

the Sample and Analysis Cell. Both fume hoods and glove boxes are located in the 

laboratories. Waste drains from the sinks, fume hoods and glove boxes are routed to the 

laboratory waste catch tank or fume hood drain tank. A special α laboratory includes two 

special glove boxes for confinement of concentrated α-emitted radionuclides.
B-1 

APPENDIX B

SELECTED PAST CHEMICAL EXPLOSION ACCIDENT

CONDITIONS AND POSSIBLE EXPLOSION

CALCULATION INPUT MODELS

This appendix provides insights gained from chemical explosion accidents that occurred in

spent fuel reprocessing facilities in the past. Of particular importance is understanding of the 

chemical explosions in process vessels used in the operations of the reprocessing facilities. 

Chemical explosions at a spent fuel reprocessing facility are most likely to occur in the 

operations involved with evaporation and concentration processes. The reason is that external 

heat is applied to remove unwanted liquid. The combination of heat and operation errors during 

transfer of TBP into the process vessel, where nitric acid and metal nitrates are present resulted 

in the “red oil” explosion condition. Lux and Vail [Lux 1995] classified the process vessels into 4 

types: feed tank, sump receipt tank, mixer settler, and evaporator. Among these types, feed 

tank, sump receipt tank and mixer settler are categorized as unheated vessels. They are 

unlikely to yield a red oil explosion condition, because no heat is applied. Thus their accident 

frequency is < 1x10-6/yr. Unlike the unheated vessels, the evaporator, concentrator or denitrator 

vessel is subjected to external heat. If an operation errors results in unwanted TBP in the 

heated vessel, the red oil explosion condition is possible. Therefore, the accident frequency of 

this heated vessel is likely higher than that of the unheated vessel.

A selected set of the past explosion accidents are described in this appendix, along with details 

on the accident conditions, particularly for the solution states before explosions. An estimate of 

the total energy of the explosion is calculated. The description should be sufficient to allow the 

simulation of the accident through the use of explosion computer codes. 

B.1 TNX Evaporator Nitration-Oxidation-Reaction Event (1953)

On January 12, 1953, a TNX evaporator exploded during the concentration of a uranyl nitratenitric acid solution. The TNX Semiworks program involved testing of equipment using the 

PUREX extraction operations. The Extraction solution included uranyl nitrate, nitric acid and 

30%/70% TBP/Amsco [Colven 1953, Durant 1988]. Amsco is a kerosene-like diluent. As a part 

of the testing, all aqueous uranyl nitrate fed to the evaporator had at one time been in contract 

with this organic solvent. At the time of the explosion, the operator heard a rumbling noise and 

then observed orange-colored fumes evolving from the top of the column with a high-pitched 

roar. The explosion destroyed the evaporator by ripping it into 6 major pieces. Most of the 

damage was located in the lower half of the evaporator, including the pot. In the column, the 

bottom tray was bowed downward more than 1 inch, while other trays were bowed to a lesser 

extent - some upward and some downward. This explosion resulted in extensive damages to 

the building, including the sheet metal roofing and siding that were extensively dislodged 

throughout the building. It also resulted in two minor injuries of personnel.

The evaporator cross section is shown in Figure B-1, and its components and dimensions are 

shown in Table B-1. 
B-2 

Figure B-1 Cross section view of TNX evaporator [Colven 1953]
B-3

Table B-1 Description of the TNX evaporator [Colven 1953]

Parameter Dimension

Evaporator Shell

Diameter

Height

Dished Ends

Wall Thickness

Volume*

6 feet (1.83 m)

5 feet (1.52 m)

1/4 –inch (0.64 cm) stainless steel

3/16 –inch (0.48 cm) stainless steel sheet (welded)

141.37 ft3 (1057.52 gallons, 4 m3)

Evaporator Pot

Diameter

Height

Wall Thickness

Heating Coils

Volume*

3 feet (0.91 m)

18 inches (45.72 cm)

3/16 –inch (0.48 cm) stainless steel

6 concentric banks of 1-inch schedule 40 stainless steel pipe

10.60 ft3 (79.52 gallons, 0.3 m3)

Column

Diameter

Height

Number of Plates

Humber Bubble Caps per Plate

Wall Thickness

Volume*

2.5 feet (0.76 m)

16.5 feet (5.03 m)

8

30

3/16 –inch (0.48 cm) stainless steel

80.99 ft3 (605.85 gallons, 2.29 m3)

Total capacity

Shell and Pot

Column

1825 gallons (6.91 m3)

1300 gallons (4.92 m3)

525 gallons (1.99 m3)

*Calculated based on the diameter and height of the component.

The cause of the accident could have been due to the fact that the liquid temperature recorder 

was broken and since the required degree of concentration was beyond the range of the 

specific gravity recorder, the evaporation was carried out for a specific length of time based on 

judgment. Also none of the organic solvent was thought to have been present in the feed, but 

after the explosion, it was discovered that Amsco diluent was present in the distillate catch tank. 

The constituents thought to be present at the time of the explosion are shown in Table B-2. 

This solution is based on the fourth and final charge consisting of the 70 gallon heel of the 

original solution and 160 gallons of previously evaporated material which had been diluted with

water. After the explosion, it was noted that the column bubble-trays revealed partially plugging 

with solid materials. This partial plugging of the column could well have facilitated a pressure 

buildup in the evaporator prior to its rupture.

Table B-2 TNX explosion solution contents and conditions [Colven 1953]

Parameter Values

Volume of Solution (liquid level recorder) 230 gallons (0.87 m3)

Composition (analysis of feed) UN-44 wt%, HNO3 -2.4 wt%

Specific Gravity (specific gravity recorder) 1.56*

TBP Content (from phosphorous analysis of residue 

after accident) 80 lbs (36.29 kg)

Amsco Content (found in distillate) 10-15 gallons (0.04 – 0.06 m3)

Pot Maximum Temperature (calculated) 164  ̊C

*At the time of the explosion, the specific gravity recorder had gone off scale (> 2.0)
B-4

To model this explosion, one may want to ignore the volume in the column to simulate the 

plugging of the bubble-trays in the observation. Therefore, the model volume should be 1300 

gallons, and the solution occupied about 230 gallons, which results in 1070 gallons for the gas 

headspace (see Table B-3).

Table B-3 Model input for TNX evaporator explosion

Parameter values

Evaporator volume and dimension

Volume (excluding Colum)

Solution

Gas headspace

1300 gallons

230 gallons

1070 gallons

Solution (1)

UO2(NO3)2 + HNO3

Volume

Mass*

UO2(NO3)2

Mass**

Molar

HNO3

Mass**

Molar

Solution

Density*

Viscosity$

Surface Tension$

230 gallons (0.87 m3)

1.26x106 g

5.97x105 g

1.74

3.26x104 g

0.6

1.56 g/cc 

3.05 mPa-s

77 mN/m

Solution (2)

TBP/Amsco

TBP

Mass

Mole1

Energy2

Amsco

Volume

Solution

Volume

Density3

Viscosity3

Surface Tension3

30 vol%/70 vol%

80 pounds (36.29 kg)

130.4

5.7581x107 Joules

10-15 gallons (medium: 12.5 gallons or 4.73x10-2 m3)

17.9 gallons (6.78x10-2 m3)

0.77 g/cc

1.2 mPa-s

22 mN/m

*Calculated using the given volume and the specific gravity of 1.56 provided in Table C.1-1.

**Calculated using the wt% given in Table C.1-1.

$Adapted from Table II and Table III of Ref.[Ondrejcin 1961] for 2 M UN solution.

1Molecular weight of TBP assumes to be 278.28 g/mole

2Assumed an energy of 4.4158x105 J/mole at 120  ̊C [Hyder 1994] 3These properties are approximated values from Appendix B of Ref. [Klophenstein 1958].

B.2 A-Line Denitration Explosion (1975)

The A-Line facility in F Area at the Savannah River Plant began operation in 1954 [McKibben 

1976]. The facility’s function is to convert uranium nitrate (UN) solution to UO3 powder. The 

process flow diagram is shown in Figure B-2. As shown in this figure, the overall processes can 

be summarized by the following:

1. Solvent decantation – to move organics (such as 30% TBP and 70% n-paraffin diluent) 

from the UN solution before it is sent to the evaporation operation.
B-5 

2. Evaporation – to concentrate the UN solution to increase the uranium concentration in

the evaporators.

3. Purification with silica gel – to reduce the 95Zr and 95Nb content in the UN solution from 2

above before sending to the hydrate process in the silica gel columns.

4. Evaporation to hydrate – to concentrate UN and water to corresponding uranyl nitrate

hexahydrate (UNH) or UO2(NO3)2•6H2O by evaporation in the hydrate evaporators

before sending to the denitrators.

5. Denitration – to converts UNH to UO3 by thermal decomposition in the denitrators.
B-6

Figure B-2 A-Line facility process flow diagram [McKibben 1976]
B-7 

The layouts of the equipment and facility are shown in Figure B-3 to Figure B-5. As shown in 

these figures, there are a total of 6 denitrators located on the first floor in the facility. The 

material inputs to these denitrators are provided by the 3 hydrate evaporators. Ultimately the 

silica gel columns provided the feed for the hydrate evaporators. 
B-8

Figure B-3 Vertical sectional elevation view of the A-Line facility [McKibben 1976]
B-9

Figure B-4 First floor view of the A-Line facility [McKibben 1976]
B-10

Figure B-5 A-Line denitrator assembly [McKibben 1976]
B-11

The explosion and fire accident on February 12, 1975 in the facility was due to a number of 

mishaps in the operations prior to the accident [McKibben 1976]. When the facility was shut 

down a month before the accident, one of the silica gel columns was regenerated by flushing 

with 14,000 liters of dilute nitric acid before sending to the concentrate hold tanks (S-1-8 and S1-9), which resulted a nominal uranium concentration of 250 g per liter and a specific gravity 

(SG) of 1.3, compared to the normal concentration of 400 g per liter and SG of 1.53. Normally, 

the TBP may have been present in an organic layer atop the silica gel column (at U 

concentration of 250 g per liter). With this flushing the resulting organic adduct became the UO2 

(NO3)2 • (TBP) 2 or TBP-UN. At the concentrate hold tanks, the aqueous solution (UN) is 

normally heavier than the organic TBP, when the UN is sent to the hydrate evaporator(s) then to 

the denitrator(s). However, because the resulting adduct at U concentration of 400 g per liter is 

heavier than the aqueous solution (UN) compared to a U concentration of 250 g per liter as 

shown in Figure B-6, the adduct was sent to the hydrate evaporators then to the denitrators. 

As shown in this figure, UN is heavier than adduct at the U concentration of 250 g per liter. 

Figure B-6 Specific Gravity (SG) for the A-Line Solution at 25  ̊C [McKibben 1976]

On February 11, 1975, these adducts made their ways to the denitrators. Noticeable foaming 

was observed at 200-300  ̊C instead of at 500  ̊C [McKibben 1976]. On February 12, 1975, 

operators added a commercial silicone antifoam agent to denitrator pot C-3-4 and started its pot 

agitator. The temperature of the pot was increased to 400 C. Approximately 2 minutes later 

two operators in the denitrator room observed a puff of red-brown smoke from this pot. They 

immediately turned off the heat to C-3-4 and turned on an emergency off-gas system to provide 

extra exhaust capacity for the pot. Explosion and fire occurred afterward. Both of these 

operators received minor injury from the explosion. The destruction caused by the accident is 

shown in Figure B-7 to Figure B-11. As shown in these figures, there was considerable damage 

to the “Transite” walls in the building, but no damage was done to the building’s structural 

supports. As shown in Figure B-7 to Figure B-10, the ejected materials from the exploded 
B-12

denitrator pot have accumulated on the equipment, walls and pipes. Figure B-11 shows the 

external walls of the facility at the east side which have been blown out.

Figure B-7 North view of the damaged denitrator room after explosion [McKibben 

1976] 

Figure B-8 South view of the damaged denitrator room after explosion [McKibben 

1976] 
B-13

Figure B-9 Deposits on walls in denitrator room after explosion [McKibben 1976]

Figure B-10 Close-up of deposits on walls in denitrator room after explosion [McKibben 

1976]
B-14

Figure B-11 View of the facility outside of east wall after explosion [McKibben 1976]

To construct the solution content at the time of the explosion, the following data is assumed and 

considered (see Table B-4): 

1. Approximately 14,000 liters of dilute HNO3 added to S-1-8 and S-1-9 tanks, which yields

250 g U/liter at Specific Gravity (SG) of 1.3.

2. Using Figure B-6 and 250 g U/liter, the following SGs for TBP and aqueous solution

indicate that the TBP adduct becomes heavier than the aqueous solution, which is then

sent to the hydrate evaporators and subsequently to the denitrators.

a. SG = 1.43 for 100% TBP

b. SG = 1.375 for 90% TBP

c. SG = 1.3 for aqueous solution

3. Using Figure B-6 and 400 g U/liter, the following SGs for TBP and aqueous solution

indicate that normally the aqueous solution is heavier than the TBP solution, so that only

the aqueous solution is sent to the hydrate evaporators and subsequently to the

denitrators.

a. SG = 1.45 for 100% TBP

b. SG = 1.38 for 90% TBP

c. SG = 1.535 for aqueous solution
B-15

4. Approximately 120 liters of TBP-UN adduct were added to the C-3-4 denitrator pot prior 

to the explosion [McKibben 1976].

5. Because of the temperature at 121 ̊C, the TBP-UN adduct sent to the hydrate 

evaporators did not yield accident conditions.

6. According to Figure B-5, the typical A-Line pot (denitrator) has a diameter of 6 ft. There 

were no additional dimensions given. This pot contains a motor drive for the agitator, 

which includes a plow. The bottom of the pot is round, but not hemispherical. Assuming 

one can scale the information from this figure, the deepest part of the pot is ~2.05 ft, and 

shadow part is ~1.42 ft. Then the average height is 1.74 ft. It is assumed that the 

volume of the pot can be thought to be a right cylinder with the given diameter and 

average height calculated: 49.2 ft3 or 1.3932 m3 (1.3932x103 liters).

7. The total energy of the explosion is about 7x105 kcal or 2.93x109 J, and the shock 

pressure is ~30 to 40 atmospheres and temperature is ~4650 K [McKibben 1976].

8. Using the data from 1,2 and 4 above, the mass of TBP is 1.43 x 120 liters = 171.60 kg, 

assuming 100% TBP. Using a molecular weight of 278.28 g/mole, it yields 616.65 

moles.

9. Using the specific reaction energy for TBP (4.4158x105 J/mole) given in Ref.[Hyder 

1994] , the moles given in 8 above yields an energy of 2.72x108 J which is an order of 

magnitude smaller than that given in 7 above.

10. Using the upper bound specific energy range of (1610-1830 kcal/mole) for TBP as given 

in Ref. [Hyder 1994] and averaging this range, then the total energy is 4.44x109 J, which 

is slightly higher than the value given in 7 above.

11. Based on the calculations in 9 and 10 above, the possible specific energy for the TBP in 

this accident is given by 2.93x109 J/616.65 moles = 4.75x106 J/mole.

12. Based on the specification for each denitrator, a single batch contains about 2000 

pounds of uranium or 907.18 kg [Harmon 1976].

13. According to the functionality of the hydrate evaporator, the goal is to produce UNH at 

1.2 kg U per liter before sending to the denitrators [McKibben 1976]. Based on this 

density, and 12 above, the volume for the UNH in the denitrator should be 756 liters.

14. Total solution volume occupied by the TBP-UN and UNH is 756+120 = 876 liters.

15. Assuming the calculated available volume from 6 above occupied by machinery and 

piping of 10%, then the gas space in the denitrator is 1260-876 = 384 liters.

16. Since there was no specification of the material make-up of the denitrator given, it is 

assumed to be stainless steel with an average thickness of 0.25 inches.

17. Since there was no specification about the pot agitator, it is assumed to be stainless 

steel. See Figure B-5 for the location of the agitator and plow component.
B-16

18. The properties of UNH are assumed as [Lang 1968]:

a. Density, according to 12 above is 1.2 g U/cc. Using the value of 1.2 g U/cc, the

UNH density is 2.465 g/cc. However, at the accident condition, the aqueous

solution is heavier than the organic, so it should be at 1.535 g/cc according to 3

above.

b. Viscosity of 19 centipoises at 85 °C.

c. Surface tension of 45 dynes per cm at 90-100 °C.

Table B-4 Input model for A-Line explosion

Parameter Value

Tank Geometry and Information

Diameter

Average Height

Thickness

Volume

Material

Internal Component: pot agitator

6 ft (1.82 m)

1.74 ft (0.53 m)

0.25 inch (0.635 cm)

~1400 liters

Stainless steel

Stainless steel

Solution (TBP-UN):

Volume

Mass 

Energy1

Density2

Viscosity2

Surface Tension2

TBP Concentration

U Concentration

HNO3 Concentration

120 liters

171.60 kg

2.93x109 J 

1.45 g/cc

3.885 mPa-s 

27.79 mN/m

80 mg/liter4

250g/liter5

3 Molar

Solution (UNH):

Volume

Mass of U

Density3

Viscosity3

Surface Tension3

HNO3 Concentration

756 liters

907.18 kg

1.535 g/cc

19 mPa-s 

45 mN/m

3 Molar6

Headspace Gas:

Volume 384 liters 1See text on assumptions

2Assumed 100% TBP, and use values reported at 20 °C [Brennan 1951]. 3See Assumption 18 above.

4Based on the solubility of the TBP in aqueous UN solution at 25 °C [McKibben 1976].

5Accident condition. 

6Assumed to be same as TBP-UN.

B.3 TOMSK-7 Explosion (1993)

A Red oil explosion at the Siberian Chemical Combine in TOMSK-7, Russia on April 6, 1993 

caused extensive damage to the nuclear fuel reprocessing facility and a large release of 

radioactive materials. This explosion was due to a runaway exothermic chemical reaction of a 

process vessel containing a concentrated solution of uranyl nitrate, plutonium nitrate, nitric acid, 

organics (Tributyl phosphate, TBP), and fission products [Rodriguez 1995]. It has been 

reported that there were two explosions, the first explosion was caused by the decomposition of 
B-17

TBP in the nitric acid, and the second explosion was due to the flammable gases released from 

the first explosion, which caused extensive damages to the concrete structures which housed 

the exploded process vessel [Usachev 2003]. In order to model this accident, the information 

about the constituents in the process vessel, as well as the vessel dimensions and properties 

must be given.

The exploded process vessel is detailed in a Los Alamos National Laboratory (LANL) report 

[Rodriguez 1995]. The dimensions and its contents before the explosion, and mechanical 

properties of the vessel are given in Figure B-12. As indicated in this figure, the vessel is a 

cylinder with a cone-type bottom. To simplify the modeling of this accident, all inter-connection 

piping and internal can be ignored. Furthermore, the assumption of a right cylinder for modeling 

this vessel may be used. Table B-5 shows the content and thermal conditions of the vessel, 

and the calculated dimension for the vessel. The average thickness is calculated in this table. 

Since the vessel is made of stainless steel 304, and the exploded conditions calculated about 

the vessel breach [Rodriguez 1995], Table B-6 provides the properties to be used for modeling

this vessel.

Figure B-12 Pressure vessel and surrounding structures, dimensions, and 

contents [Rodriguez 95]
B-18

Table B-5 Vessel contents and calculated dimensions from Figure B-12

Contents Volume (liter) Density (g/cc) Temperature (C)

headspace 10,900 -

Organic layer (TBP) 500 0.9 85

HNO3 (12 M) 1,500 1.4 35

UNH (0.66 M) 23,500 1.5 85

Calculated Vessel Dimension

Available volume 36.4A m3 Inside radius 1.384C m

Average wall thickness 1.6B cm Inside height 4.238D m A The sum of the content volumes

B The average of the reported thickness between 14 to 18 mm [Rodriguez 1995]

C Calculated by subtracting the average wall thickness to the diameter of 2.8 m from Figure C.3-1. D Calculated equivalent height, assuming the vessel is a right cylinder. 

Table B-6 Stainless steel 304 properties [ASM 2012]

Property Metric English

Density 8 g/cc 0.289 lb/in2

Ultimate tensile strength 505 MPa 73.2 ksi*

Yield tensile strength 215 MPa 31.2 ksi**

Young’s modulus (modulus of elasticity) 193-200 GPa 28000-29000 ksi***

Poisson’s ratio 0.29 0.29

*In comparison to reported value of 70 ksi at ~40  ̊C, and 62 ksi at 135  ̊C for Gr.70 and 80 ksi at ~40  ̊C and 77 ksi at

135  ̊C of 304 SS [Rodriguez 1995]

**In comparison to reported value of 25 ksi at ~40  ̊C, and 20 ksi at 135  ̊C for Gr.70 and 35 ksi at ~40  ̊C and 26 ksi 

at 135  ̊C of 304 SS [Rodriguez 1995]

***In comparison to calculated value using Equation 12 in [Rodriguez 1995], the value is ranging from 27920-28170 

ksi @ 100  ̊F, and 26670 ksi @316  ̊F.

Table B-5 lists the densities and volumes of the constituents in the vessel at the time of the 

explosion, which was based on the material flows into the vessel. The total content mass 

calculated from this table is 37,800 kg, and the bulk density is about 1.4824 kg/liter or 1.4824 

g/cc. According to another LANL report [Eisenhawer 1995], there should be a total of five layers 

in the vessel before the explosion (see Figure B-13). As shown in Figure B-13, the reaction 

layer is not identified in Figure B-12, because information about this layer is not well known. 

However, brown NO2 fumes were seen escaping from the plant stack [Eisenhawer 1995], which 

provided clues about this reaction layer. The following section discusses the possible reactions 

and release energy of the accident.
B-19

Figure B-13 Layers of contents in TOMSK-7 process vessel at accident time 

[Eisenhawer 1995]

At high temperatures, TBP tends to decompose, especially with nitric acid and other 

constituents. The reactions are complicated, and there are still many unknowns about the end 

products, which depends on the concentrations of the reactants involved, impurities, and 

temperatures. The complete oxidation of TBP by the nitric acid is illustrated below [Hyder 

1994]:

(CC4HH9)3PPPP4 + 14.4 HHNN 3 → 12 CC 2 + 7.2 NN2 + 19.2 HH2 + HH3PPPP4 Equation B-1 

The total heat generation from this reaction is about 1610 – 1830 kcal/mole (6.7407x106 – 

7.6618 x106 J/mole) TBP, assuming the reaction takes place at 25  ̊C. The molecular weight of 

TBP is 278.28 g/mole. In the same report [Hyder 1994], the decomposition of TBP yields a heat 

generation rate of about 105.47 kcal/mole (4.4158x105 J/mole) at 120  ̊C based on a series of 

chemical reactions, considering the decomposition of TBP to n-butanol, and then the reaction of 

n-butanol with HNO3, etc. Because there was evidence of NO2 release from the plant, the 

reaction in Equation (B-1) may not adequate to explain the NO2 release.

Experiments on the reaction between TBP and nitric acid solutions have been performed by 

Smith [Smith 94]. In the experiments, Smith examined both single phase (organic) and two 

phase (organic and aqueous) TBP/HNO3 reacting solutions at temperature above 100  ̊C. For 

closed systems, such as in TOMSK-7, the oxidation rate constant was about 1.33x10-3 per 

minute. In the experiments, Smith measured the product gases from the TBP/HNO3 reactions, 

which include CO2, CO, NO, N2O, N2 and NO2. In a two phase experiment (#13) [Smith 1994], 

the following chemical reaction for the n-butanol (CC4HH9 ) and nitric acid, after the TBP 

saturated with 3.6 M HNO3 and 10 M HNO3 at 110  ̊C were mixed:
B-20

CC4HH9 + 1.81 HH 3

→ 0.42 CC4HH8 2 + 0.32 CC3HH6 2 + 0.26 CC2HH4 2 + 0.75CC 2

+ 0.09CC + 0.31NN2 + 0.14NN2 + 1.4NN + 0.24NN 2

+ 0.26HH 2 + 0.17HH2

Equation B-2 

The butanol can be formed by the hydrolysis of TBP in the following reactions [DNFSB T33]:

(CC4HH9)3PPPP4 + HH2 → HH(CC4HH9)2PPPP4 + CC4HH9 Equation B-3 

HH(CC4HH9)2PPPP4 + HH2 → HH2CC4HH9PPPP4 + CC4HH9 Equation B-4 

HH2CC4HH9PPPP4 + +HH2 → HH3PPPP4 + CC4HH9 Equation B-5 

Using Eq. B-3, each mole of butanol would yield about 3 moles of gases. Based on the 

hydrolysis of TBP from Eq. B-4 to Eq. B-5, each mole of TBP would decompose to 1 mole of 

butanol. Knowing the number of moles of TBP, one could determine how much gases would be 

generated. Smith has correlated the gas generation with the organic to aqueous thickness ratio. 

Using Table B-5 for the organic and aqueous volumes measured for the TOMSK-7 accident, 

assuming the cross section area of both volumes are the same, then the organic to aqueous is 

1/3. Reading from Figure B-14, the gas generation rate is about 0.01 liter per minute per M 

(M=moles of HNO3/liter). As shown in Table B-5, the nitric acid layer is at 12 M. Therefore, the 

gas generation rate is about 0.12 liter per minute. It has been estimated that the total gas 

generated from the accident is about 250 m3 [Usachev 2003], which could be relatively high. 

Using the experimental results from Run#13 of Smith [Smith 1994], the initial organic density is 

~1.065 g/cc, which utilizes about 3 moles of TBP per liter. Using this number of moles of TBP 

per liter, then the organic layer as described in Table B-5 would yield about 1500 moles of TBP. 

For each mole of TBP decomposed, there would be about 3 moles of gases generated. 

Therefore, about 4500 moles of gas generated. For an ideal gas, each mole of gas has a 

volume of 22.4 liters at the standard condition, which implies that the total volume generated by 

the organic layer is about 100 m3. In terms of the energy generation, the energy generated is 

about 1.0111x1010 J to 1.1493x1010 J, using the energy released from Eq. B-1, and the 

previously calculated number of moles of TBP. If using the energy generated suggested by 

[Hyder 1994], the total energy generated for the same number of TBP moles is reduced to about 

6.6237x108 J. 

In terms of the fluid properties, [Bajoria 2012] published the TBP with nitric acid property plots 

(Figure B-15 to Figure B-18). 

The first simulation for modeling this explosion should consider the input parameters as 

provided in Table B-7. 
B-21

Figure B-14 Gas generation rate as a function of the organic to aqueous 

thickness ratio [Smith 1994]

Figure B-15 Effect of TBP on density of the aqueous solutions [Bajoria 

2012]
B-22

Figure B-16 Effect of TBP on viscosity of the aqueous solutions [Bajoria 

2012]

Figure B-17 Effect of TBP on interfacial tension between dodecane and 

aqueous solutions [Bajoria 2012]
B-23

Figure B-18 Effect of nitric acid concentration on the solubility of TBP in 

the aqueous phase [Bajoria 2012]
B-24

Table B-7 Summary of inputs for modeling TOMSK-7 accident

Parameter Value Comments

Vessel 

Dimension

Right cylinder geometry is assumed for simplifying the modeling effort, using 

data from Table B-5. The properties for this vessel are given in Table B-2.

Inside radius 1.384 m

Inside height 4.238 m

Wall thickness 1.6 cm

Contents

Liquid density 1.4824 g/cc It is based on the total contents divided by the total liquid volume as shown in 

Table B-5.

Liquid volume 25,500 liter

Gas density 7.29x10-4 g/cc It is assumed that the gas composition followed Eq. B-2, then normalized all 

gases produced, and evaluated at the average design pressure (12-16 atm) and 

design temperature of 135  ̊C [Rodriguez 1995]. It is assumed that ideal gas 

applied. The average MW of the gas is about 36.14 g/mole.

Gas volume 10,900 liter See Table B-5.

Fluid 

Properties

Viscosity 1.2 mPa-s Using the top curve of Figure D-5

Surface 

Tension

45 mN/m Using the lowest data point in Figure D-6

Explosion It is assumed that the explosion is due to the decomposition of TBP and the 

subsequent reactions of the butanol

Energy 

Generated

1.0x1010 J Using the lower end reported value from Eq. B-1, accounting for the uncertainty 

of the reaction zone before the explosion.

Gas generated 90 m3 It is assumed that the calculated gas volume of 100 m3 minus the initial gas of > 

10 m3 as shown in Table B-5. The gas release rate should be corresponding to 

the explosion rate modeled in the simulation.

B.4 Tank A-109 HAN-Nitric Acid Explosion (1997)

A hydroxylamine nitrate (HAN) explosion had occurred on May 14, 1997 in Tank A-109 in Room 

400 of the Plutonium Reclamation Facility (PRF), which is a part of the Plutonium Finishing 

Plant (PFP) located in the 200 West Area of the Hanford Site [DOE 1997]. Figure B-19 shows 

the schematic layout of the fourth floor for the PRF, where Tank A-109 is located. As shown in 

this figure, this floor houses Operating Control Room 44, Glovebox Room 43, Miscellaneous 

Treatment Room 41, and Column Room 42 in which vertical sections of two liquid-extraction 

columns (housed in a glovebox) penetrate the room from above and below. A-109 is one of the 

chemical makeup vessels that are located in Chemical Preparation Room 40 (see Figure B-19). 

This room does not contain radioactive materials. The facility ventilation system is designed for 

once-through flow to allow outside air to flow from uncontaminated areas to contaminated 

areas, then through a high-efficiency filter system before being exhausted by fans to the 

atmosphere via the PFP’s main stack. An alarmed monitor that was capable of detecting NO 

and NO2 gases in Room 40 had been installed. Figure B-20 shows the Schematic of the Tank 

A-109 and its components. As shown in this figure, Tank A-109 was equipped with a motorized 

agitator that was attached to the outside of the tank’s lid. Both the agitator shaft and mixing 

propellers were suspended from the tank lid to mix the solution inside the tank. The lid was 

attached with a large number of steel bolts. This tank has a 2” sloped bottom to ensure that the 

tank contents drain completely. Temperature indication capability is not equipped for this tank. 

This tank is linked to the vent header above the tank lid. The header is connected to all tanks in 

Room 40 through an exhaust duct, which vents to the facility exhaust system and then through 

the stack. No valve exists on either the overflow line, the vent line or vent header. As shown in 
B-25

this figure, there is a connection to a hydrazine Tank A-109A. Since hydrazine was no longer 

used, the connecting tank is emptied. The tank has a capacity of 400 gallons and is used to 

prepare a HAN and HNO3 solution.

Figure B-19 Schematic of the fourth floor of the PRF and location of Tank A-109 [DOE 

1997] 

Figure B-20 Schematic of Tank A-109 [DOE 1997]
B-26

The sequence of the events that developed the conditions prior to the explosion on May 14, 

1997 is given below [DOE 1997]:

• The HAN and NNO3 solution in the tank had evaporated for nearly four years. It was

reported a final sample contained 0.2534 M HNO3 and 0.3542 HAN solution at 370

gallons (3400 lb) before the facility was shut down on June 18, 1993. The inventory in

the tank was at 2440 lb on December 29, 1993. The last recording was done in October

28, 1996, which showed a weight of 295 lb (< 30 gallons) remained in the tank.

• The evaporation process, which concentrated the HAN and HNO3 and the potential

effect of catalysts (such as iron from erosion of the tank wall) created the conditions that

led to an autocatalytic chemical reaction.

• The explosion inside the tank tore the lid from the tank and propelled it and its

attachments upward with enough force to sever a 1.5” fire-suppression water line and to

severely damage the ceiling and roof. At some point, the agitator motor broke loose

from the tank lid and landed on the floor, while the agitator shaft became embedded in

the ceiling. The tank was displaced from the scale on which it was set and came to rest

at an angle against Tank A-102, which has been displaced 6” without damaging A-102.

The impact of the objects hitting the ceiling, creating a 2-foot bulge in the roof above

Room 40, 6-foot-long separation at the roof-wall interface, and a single tear near the

bulge. A small crack was observed in the insulation material of a ventilation supply duct

on the facility roof. (See Figure B-21 to Figure B-24 for the post-explosion conditions of

the facility.)

• Estimating the explosion due to the gas pressure inside of A-109 that blew the tank’s lid

and displaced the tank [DOE 1997]:

o Fauske and Associates estimated 200 to 300 psi of pressure with concentrated

HAN > 8 M.

o Babcock and Wilcox Alliance Research Center estimated a pressure buildup of

150 to 250 psi would be sufficient to cause this type of failure to the tank.
B-27

Figure B-21 A post-explosion photo of the tanks inside Room 40 and the location of 

agitator motor [Harlow 1998]

Figure B-22 A post-explosion photo showing the displacement of A-109 Tank [Halow 

1998]
B-28

Figure B-23 A post-explosion photo showing the agitator lodging into the roofing area 

[DOE 1997]

Figure B-24 A post-explosion photo showing the roof bulging [DOE 1997]
B-29

Using the pressure data given, the estimation of the explosion energy is calculated using the 

volume of the tank (400 gallons or 1.51 m3): 

• Using 200 psi (1.38x106 Pa) and the tank volume, the explosion energy is 2.08x106

Joules.

• Using 250 psi (1.72x106 Pa) and the tank volume, the explosion energy is 2.60x106

Joules.

If using the last recorded and sampled inventory before shutdown on June 18, 1993, the HAN 

loss was assumed due to the transfer of 1000 lb out before December 29, 1993. The following 

calculation of the HAN decomposition energy can be calculated:

1. Last recorded composition of 0.2534 M HNO3 and 0.3542 M HAN solution of 3400 lb

(1542 kg) and a volume of 370 gallons.

2. 370 gallons equal to 1400.6 liters in volume.

3. HAN quantities are:

a. Molecular weight of HAN is ~96 g/mole, assuming that chemical formula of

NH2OH•HNO3.

b. Using the volume from 2 and 0.3542 M HAN from 1 above, 496.09 moles of HAN

are calculated. Adjusting the 1000 lb loss after the recorded sample, then this

quantity is reduced to 350 moles.

c. Using the molecular weight given, there is 47644.7 g of HAN. Adjusting the 1000

lb, this quantity is reduced to 33632 g.

4. HNO3 quantities re:

a. Molecular weight of HNO3 is ~63 g/mole

b. Using the volume from 2 and 0.2534 M HNO3 from 1 above, 354.91 moles of

HNO3 is calculated. After adjusting a 1000 lb loss, this quantity is reduced to

251 moles.

c. Using the molecular weight given, there is 22359.46 g of HNO3. After adjusting a

1000 lb loss, this quantity is reduced to 15788 g.

5. According to DOE/EH-0555 [Harlow 1998], it gives a number of decomposition energy

values, depending on the reactions and concentration of HAN in the HNO3 solution:

a. For 5NH2OH•HNO3 → 3N2 + 8H2O + 4HNO3, decomposition energy is -68.3x103

cal/mole.

b. For 4NH2OH•HNO3 → 2N2O + 7H2O + 2HNO3, decomposition energy is -

49.1x103 cal/mole.

c. For a 0.5 M HAN solution, decomposition energy of 34.1 cal/g of solution.
B-30

6. Using the adjusted mole value for the HAN in 3 above, the decomposition energy

according to the first two values in 5 above are:

a. 21 MJ for 5NH2OH•HNO3 → 3N2 + 8H2O + 4HNO3 reaction

b. 18 MJ for 4NH2OH•HNO3 → 2N2O + 7H2O + 2HNO3 reaction

Using the adjusted mass value for the HNO3 in 4 above, the decomposition energy according to 

the last value in 5 above is given as 3.2 MJ for 0.5 M HAN solution with HNO3. 

The properties of the HAN solution are estimated as:

1. Density of the final solution prior to explosion is calculated based on the remaining volume

of <30 gallons (114 liters) and 295 lb (133.81 kg):

a. Specific gravity (SG) of 1.08518 for HAN and 1.3795 for HNO3 at 25 C assumed

[9]

b. Assuming HAN mass of 33.6 kg remained, the total volume occupied by HAN is

31 liters, using the SG given in a above.

c. Assuming HNO3 mass of 15.8 kg remained, the total volume occupied by HNO3

is 11.45 liters using the SG given in a above.

d. The remaining mass is assumed occupied by water which is 133.81 kg – 33.6 kg

– 15.8 kg is 84.41 kg. Assuming SG of 1 for water, this mass occupies a volume

of 84.41 liters.

e. Then, the total volume of HAN, HNO3 and water is 126.86 liters, which exceeds

the value given above. Therefore, the water mass is adjusted to be 114 – 31 –

11.45 liters = 71.55 liters. Thus the water mass is 71.55 kg.

f. Therefore, the density of the solution is 1.06 g/cc.

2. Using the final masses of HNO3 and HAN computed from 1 above, the solution is 3.42 M

HAN and 15.3 M HNO3. In terms of mass ratio, this solution is 25.5wt% HAN and 74.5wt%

HNO3. According to Ref. [10], the reactions at higher concentrations (> 6 M HAN) result in

the complete vaporization of the solution.

3. It is assumed that all HAN is consumed during the explosion, therefore, the remained

solution is HNO3 and water. The properties for HNO3 are used for aerosol calculations.

The input model for this accident is given in Table B-8. 
B-31

Table B-8 Model input for Tank A-109 explosion

Parameter Values

Tank volume and dimension

Volume

Height x Diameter

Wall thickness

Material

Solution (mainly HNO3 after explosion)

Gas headspace

400 gallons (1.51 m3) 

5 ft (1.52 m) x 4 ft (1.22 m)

3/16-inch (4.76x10-3 m)

347 stainless steel*

30 gallons (0.11 m3) 

370 gallons (1.4 m3)

Solution

Density

Viscosity$

Surface Tension$

1.06 g/cc

0.749 mPa-s 

41.2 mN/m

Explosion Energy

Pressure-Volume Relation

Decomposition Energy

Placement of Explosive

2.08-2.6 MJ

3.2-21 MJ

Above the solution

*Using properties as given before for 304 stainless steel

$These values are for HNO3. HAN values are 1 mPa-s and 40 mN/m, respectively [Vosen 1990]

B.5 Additional Possible Vessel Geometries for Considering in 

Explosions

The process vessels used in spent fuel reprocessing can be classified into two types: unheated 

and heated [Lux 1995]. The unheated process vessels are the mixer-settlers, holding tanks and 

storage tanks that do not be externally heated. The heated process vessels, which need 

external heating are the evaporators, denitrators, and concentrators. External heat used include 

propane, natural gas, and steam. In both types, organics (TBP or HAN) mat be present in the 

vessels. Although the accident frequency for the unheated vessel is generally lower than the 

heated vessels, the possible chemical explosion does exist. The heated process vessels 

requires accurate monitoring and controlling of the solution temperature and pressure within the 

vessel to reduce any potential explosions due to chemical reactions and overpressure.

This section describes the possible vessel geometries in the operations that could potentially 

have an explosion in a spent fuel reprocessing facility. 

Unheated process vessels used in the spent fuel reprocessing facilities include mixer-settlers, 

holding tanks, and storage tanks. Mixer-settlers (see Figure B-25) are used in the extraction 

operations of the reprocessing, which separates out the aqueous and organic streams. There 

are also two other process vessels that are used similarly as mixer-settlers: pulse columns (see 

Figure B-26) and centrifugal contactors (see Figure B-27). 
B-32

Figure B-25 Mixer-settler schematic [Simpson 2010]

Figure B-26 Schematic of Pulse Colum with Perforated Concept [Simpson 2010]
B-33

Figure B-27 Schematic of a Centrifugal Contactor [Simpson 2010]

Mixer-Settlers: This equipment provides a single stage of extraction. As shown in Figure B-25, 

both mixtures enter the mixing section of the equipment. After mixing, the solution flows into the 

settling section where the separation of the phases takes place. The principal behind the 

separation is based on the density difference of the phases, using gravity effect. Usually the 

lighter material (organic) flows over a weir; whereas, the heavier material (aqueous solution)

flows out near the bottom of the settling section. The residence time takes from several to tens 

of minutes, and requires a large facility footprint, but not much headspace and only limited 

remote maintenance capability because of the nature of the process.

Centrifugal contactors: This equipment is similar to the mixer-settlers, with a single stage of 

extraction. As shown in Figure B-27, the mixing of the less and more dense phase of fluid is 

located in the annular region of the contactor. Once mixed, the solution enters at the bottom of 

the rotor located at the center, where the centrifugal force separates the denser and lighter 

solutions. The use of rotor to separate the phases is the main difference between this and the 

mixer-settler. This equipment uses a residence time and requires a small facility footprint and 

minimal headspace. However, the use of the rotor requires remote maintenance capability for 

the motor and/or rotor. A cut-away view of the contactor is shown in Figure B-28. This 

contactor is being used at La Hague plant. Similarly, this type of contactor was to be used at 

the Barnwell Nuclear Fuel Plant (see Figure B-29). This contactor was used in the first 

extraction cycle for separating fission products and uranium-plutonium for the head activity 

separation. This contactor is a stacked 10-stage contactor, which takes in the aqueous feed 

consisting of ~70% from the adjusted dissolver product from the HA feed tank, and ~30% from 

the HS column raffinate. The output from this contactor consists of two streams: uranium and 

plutonium and 30% TBP solution, and fission product aqueous waste stream. 
B-34

Figure B-28 Cut away view of a multi-stage centrifugal contactor [Drain 2003]
B-35

Figure B-29 Cross Sectional View of High Activity Centrifugal Contactor at Barnwell 

Nuclear Fuel Plant (see Appendix A) 

In terms of heated process vessels, concentrators, evaporators and denitrators are used to 

concentrate the product stream and waste stream to reduce the water or liquid contents. 

Solidification processes may be followed once these contents are reduced. Examples of the 

evaporator and denitrators can be found in Appendix C. 

Both Appendix A and Bader [2011] provide a number of unheated and heated process vessels 

that are used in the design of the spent fuel reprocessing facilities. Readers are encouraged to 

review these appendices for the examples of the process vessels:

Barnwell Nuclear Fuel Plant (Appendix A): 
B-36

• General Purpose Concentrator

• Acid Fractionater, reboiler and condenser (steam is heated fluid)

• Solvent burner and quench pot

• UPC Concentrator with condenser and reboiler

• PPC concentrator and reboiler

Engineered Alternative Studies [Bader 2011]: 

• UREX Process Train – Off-specification evaporator (heated by steam)

• U/Tc Separation Train

o Pyrolysis oven for removing nitrates from the spent resin, and

o Carbon reforming oven to remove carbon from the resin.

• CCD-PEG Train – Raffinate evaporators – 40% reduction in volume

• U Solidification Train

o U hydrate evaporators to concentrate uranium solution

o U denitrators/reactors (electrical heated) to form UO3

• U/TRU Solidification Train – U/TRU calciner

• Fission Product Solidification Train – calciners (electrical heated)

• Cs/Sr Solidification – steam reformers
B-37

B.6 References

[ASM 2012] Data sheet from http://asm.matweb.com for AISI Type 304 Stainless 

Steel, July 3, 2012.

[Bajoria 2012] S.L. Bajoria et al., “Effect of tri-n-butyl phosphate on physical properties of 

dodecane-nitric acid system,” J Radioanal Nucl Chem, DOI 

10.1007/s10967-012-1832-9, May 25, 2012.

[Brennan 1951] KAPL-M-MEB-1, M.E. Brennan, “Physical and Chemical Properties of 

TBP,” Knolls Atomic Power Laboratory, March 9, 1951.

[Colven 1953] T.J. Colven, Jr. et al., Interim Technical Report: TNX Evaporator Incident 

January 12, 1953, E.I. du Pont de Nemours & Co., May 15, 1953.

[Decker 1987] BRL-TR-2864, M.M. Decker et al., “HAN-Based Liquid Gun Propellants: 

Physical Properties,” Technical Report, AD-A195 246, US Army Ballistic 

Research Laboratory, November 1987.

[DNFSB-T33] DNFSB/TECH-33, “Control of Red Oil Explosions in Defense Nuclear 

Facilities 2003,” Defense Nuclear Facilities Safety Board Technical 

Report, November 2003.

[DOE 1997] DOE/RL-97-59, Accident Investigation Board Report on the May 14, 1997 

Chemical Explosion at the Plutonium Reclamation Facility, Hanford Site, 

Richland, Washington, Revision 0, Department of Energy, July 26, 1997.

[Drain 2003] F. Drain et al., “40 years of Experience with Liquid-Liquid Extraction 

Equipment in the Nuclear Industry,” Waste Management 2003 

Conference, February 23-27, 2003, Tucson Arizona, U.S.A.

[Durant 1988] Savannah River Laboratory, W.S. Durant et al., “Explosions and Other 

Uncontrolled Chemical Reactions at Non-Reactor Nuclear Facilities of the 

Savannah River Plant,” DP-MS-88-15, E.I. du Pont de Nemours & Co., 

December 1988.

[Eisenhawer 1995] LA-UR-95-1712, S. W. Eisenhawer, “Accident Reconstruction Using 

Process Trees,” Los Alamos National Laboratory, 1995.

[Hang 1998] WSRC-TR-98-00292, T. Hang, “The FB-Line and F-Canyon HAN/Nitric 

Acid Decomposition Study,” Westinghouse Savannah River Company, 

August 19, 1998.

[Harlow 1998] DOE/EH-0555, D.G. Harlow et al., “Technical Report on Hydroxylamine 

Nitrate,” Department of Energy, February 1998.

[Harmon 1976] H.D. Harmon et al., “Behavior of Tributyl Phosphate in A-Line Processes,” 

DP-1418, E.I. du Pont de Nemours & Company, Savannah River 

Laboratory, August 1976.
B-38

[Hyder 1994] WSRC-TR-94-059, M.L. Hyder, “Safe Conditions for Contacting Nitric 

Acid or Nitrates with Tri-n-butyl Phosphate (TBP),” Westinghouse 

Savannah River Company, January 1994.

[Klophenstein 1958] R.K. Klophenstein, Studies of the Behavior of the TBP-Kerosene Solvent 

in Uranium Refining: Degradation of the Kerosene Diluent, NLCO-769 

(TID-4500, 14th Ed.), National Lead Company of Ohio, August 12, 1958.

[Lang 1968] G.P. Lang et al., “Some Physical Properties of Uranyl Nitrate Solutions – 

Topical Report,” Report No. MCW-1424 (TID-4500, 14th Ed.), 

Mallinckrodt Chemical Works, St. Louis, December 15, 1968.

[Lux 1995] WSRC-RP-95-910, C.R. Lux and R.E. Vail, “Overview of “Red Oil” 

Frequency Analyses for F-Canyon (U),” Rev. 0, Westinghouse Savannah 

River Company, September 21, 1995.

[Mckibben 1976] J.M. McKibben, “Explosion and Fire in the Uranium Trioxide Production 

Facilities at the Savannah River Plant on February 12, 1975,” DPSPU 76-

11-1, E.I. du Pont de Nemours & Company, Savannah River Plant, 

October 1976.

[Ondrejcin 1961] R.S. Ondrejcin, “Physical Properties of Uranium Process Solutions,” DP653, E.I. du Pont de Nemours & Co., Savannah River Laboratory, 

October 1961.

[Paddleford 1994] WSRC-MS-94-0649, D.F. Paddleford and H.K. Fauske, “Safe Venting of 

‘Red Oil’ Runaway Reactions,” Westinghouse Savannah River Company, 

December 21, 1994.

[Rodriguez 1995] LA-UR-95-1756, E.A. Rodriguez and R.F. Davidson, “TOMSK-7 

Radiochemical Facility Explosion: Chemical Vessel Burst Failure 

Analysis,” Los Alamos National Laboratory, 1995.

[Simpson 2010] INL/EXT-10-11753, M.F. Simpson and J.D. Law, “Nuclear Fuel 

Reprocessing,” Idaho National Laboratory, February 2010.

[Smith 1994] WSRC-TR-94-0540, J.R. Smith and W.S. Cavin, “Isothermal Heat 

Measurement of TBP-Nitric Acid Solutions,” Westinghouse Savannah 

River Company, December 16, 1994.

[Usachev 2003] V.N. Usachev and G.S. Markov, “Incidents Caused by Red Oil 

Phenomena at Semi-Scale and Industrial Radiochemical Units,” 

Radiochemistry, Volume 45, No. 1, 2003.

[Vosen 1990] S.R. Vosen, “Hydroxylammonium Nitrate – Based Liquid Propellant 

Combustion – Interpretation of Strand Burner Data and the Laminar 

Burning Velocity,” Combustion and Flame, 82 (1990), pg 376-388.
C-1 

APPENDIX C

MISCELLANEOUS DATA

C.1 Program Review Comment Responses Given February 7, 2013

Comments on Sandia National Laboratory Research Program Review February 7, 2013

Source Term Releases from Spent Nuclear Fuel Reprocessing Facilities, JCN V6050

1. Ensure the aerosol modeling incorporates current aerosol physics and chemistry to

describe aerosol transport and behavior during all releases.

We plan to use best-estimate aerosol and phase modeling to describe transport and

releases into the environment.

2. The radionuclide source term releases should be expressed in radioactivity units of

Ci/day, Ci/month, etc. and concentrations in releases to the atmosphere and aqueous

water bodies, including groundwater, where appropriate. Exposure of radioactive

releases needs to consider both workers and members of the public (e.g., reasonable

maximally exposed individual).

The modeling approach that we envision will treat the mass of radionuclides in various

locations, including suspended particles (aerosols). The radionuclides will be assumed

to be distributed uniformly in the masses that are tracked, so the mass and activity

fractions are identical. Thus, at the end of the analysis, releases in terms of isotopic

activities are readily calculated by the tool and provided to the user.

3. Clarify how fires will be treated empirically and how fires will be modeled.

Codes such as MELCOR are not designed to model fire physics explicitly. The fire can

be approximately modeled in MELCOR and other similar codes by adding the fire mass

(such as soot/by-product) and fire energy into the room(s) or control volume(s). An

empirical correlation is mainly involved with the development of the mass and energy

source profile(s) that can be input to codes, such as MELCOR. These empirical

correlations (temperature profile in the accident room) may be developed using codes

specifically designed for fire analysis, such as FDS or CFAST. The correlation is

developed using this approach for a range of the fire size (in energy) and room size, etc.

Ultimately the developed source term tool should be able to be benchmark against

“defined” fire experiments that were used to validate the fire analysis codes. Depending

on the resources available, it may be worthwhile to modify codes, such as MELCOR, to

be able to model fire scenarios, without relying on the empirical correlation approach. In

fire analysis codes, the atmosphere in the room can be modeled with two gas layers (hot

layer and cooler layer). The height of the hot layer is adjustable depending on the fire

energy added to the room. For a code such as MELCOR, hot gas layer modeling cannot

be performed. Rather, the modeling approach would rely on the user to divide the room

into appropriate, smaller control volumes in order to capture the stratification effect. We

currently intend to use the correlational approach when developing the characterization

of fires within the source term tool.
C-2 

4. For the major process, include the off-gas treatment processes, red-oil mechanisms,

calcination and oxalate processes during the uranium/plutonium purification processes.

It is our intention to cover these processes, including the vitrification process, and for the 

range of accident conditions. Below is a table (Table C-1) from the DP-1558 report. We 

intend to add additional scenarios as needed to cover all known processes and 

accidents in the literature we identified. 

Table C-1 Potential incidents possible at a spent fuel reprocessing facility [DP-1558]

Low Consequence Intermediate Consequence High Consequence

Fuel Receiving and Storage Operation

• Fuel damage in transit or upon

arrival

• Fuel cooling time too short

• Loss of cooling

• Surface contamination not

removed from cask

• Low water level

• Rupture of fuel during storage

• Cask inadvertently vented

• Cask dropped

• Damage to fuel assembly outside

cask

• Damage to stored fuel

• Loss of cooling capability

• Hydrogen explosion in fuel

cleaning vessel

• Criticality

Shearing Operation

• Fuel jammed or stranded in shear

• Irregular length of hulls

• Shear jammed

• Malfunction of inert gas system

• Chopped fuel overheating

• Pressurization of shear

• Fuel element overheating

• Pyrophoric fire1

• Release of volatile and particulate

activity into room

Dissolving Operation

• Un-complex fluoride in dissolvent

• Pu-rich reside settling in dissolver,

lines, and other process vessels

• Dissolver seal failure

• Inadequate cladding rinse

• Transfer error of dissolver solution

• Suckback (due to eructation and

pressurization of dissolver)

• siphoning

• Contact of sheared fuel at a

temperature about 300  ̊C above

boiling point of dissolver solution

• Pressurization of the dissolver

• Dissolver leakage (due to

corrosion gasket/valve failure)

• Charge of inadequate cooled fuel

• Malfunction of dissolver off-gas

(VOG) iodine adsorbers

• Excessive interaction of fuel

external to dissolver (slightly

reduction in criticality safety

margin)

• Excessive pressure in water wash

compartment except during

basket exchange.

• Dissolver pot coils not submerged

during shutdown (high release of

volatiles to room atmosphere)

•

• Precipitation of Pu polymer

(potential criticality)

• Zirconium fire1

• Zirconium explosion (due to

significant fines in the dissolver)1

• Explosion in the iodine adsorber

• Overconcentration in dissolver

(potential criticality)

• Self-concentration of dissolver

solution (potential criticality)

• Low H+/Pu ratio in dissolver

solution (potential criticality)

• Inadequate poison in dissolver or

leach solution (potential criticality)

• High fuel loss in leached fuel

fragments

• Failure of cladding monitor

(potential criticality)

• Absence of basket in dissolver

• Explosion in dissolver vessel or

hold tanks (H2 explosion)

Solvent Extraction Operation

• Plutonium concentration in a

solvent extraction contractor

increases beyond normal values

• Potential backup of radioactivity

by air lines

• Potential for abnormal

accumulation of fissile material by

plutonium reflux in contractor,

particularly during high activity

waste rework (decreased

criticality safety margin)

• Solvent fire2

• Potential for buildup of unsafe

amounts of fissile material in

organic storage tank (potential

criticality)

• Potentially unsafe feed

concentration (potential criticality)
C-3 

Table C-1 Potential incidents possible at a spent fuel reprocessing facility [DP-1558]

Low Consequence Intermediate Consequence High Consequence

• Flow reduction or interruption of

scrub streams to contactors

(uranium)

• Loss of organic flow to a contactor

(uranium)

• Low temperature in scrub stream

to contactor (uranium)

• Low temperature in contactor

• Emulsion in solvent washer

• High plutonium losses in waste

streams from contactors (in

PUREX process)

• High plutonium concentration in

contactor (in PUREX process)

• Excess gamma activity in

plutonium product

• Potential for excessive product

loss to spent organic effluent

(reduced criticality safety margin)

• Accumulated solvent in rooms

• Solvent extraction system

contents removed by overflow (in

sump)

• Potential for excessive plutonium

loss to aqueous waste (potential

criticality)

• Incorrect material transfer of feed

chemicals, product materials or

incorrect materials (potential

criticality, excessive radioactive

material release)

Product Evaporation

• Leaks in tank containing

concentrated uranium-plutonium

solutions

• Transfer error in a product

evaporator system

• Coil or tube-bundle (in reboilers

and condensers) failure

• Overflow of an evaporator

• Potential reaction between TBP

and UNH (red oil explosion)

• Overconcentration of U/Pu

product (potential criticality)

• Boil-over (release of radioactive

materials from primary

containment)

Waste Evaporation

• Overconcentration of waste • Ruthenium escapes to stack

• Leaks (to sumps)

• Transfer error in the evaporator

system

• Coil or tube-bundle (in reboilers

and condensers) failure

• Overflow of a feed or bottom tank

• High radioactivity in high-activity

waste r evaporator condenser

• Inadequate vessel cooling

capacity (that leads to release

aerosols to room atmosphere.

• Leak of high-level waste

concentrate

• Boil-over (release of radioactive

materials from primary

containment)

• Explosion in high-activity waste

evaporator

Acid Recovery Operation

• Pressurization of acid

absorber/fractionator

• Leaks due to corrosion

• Eructation due to foreign materials

in feed

• Solvent addition to feed tank

• Explosion due to solvent in feed or

temperature > 140  ̊C

• High radioactivity in recycled acid

and/or water

Recovery Ion Exchanger Operation

• Overflow

• Leakage due to corrosion

• Uncontrolled reaction between

nitric acid and anion exchange

resin in the primary recovery

column

• Ion exchange resin fire (lead to

radioactivity release to

environment)

Off-Gas Treatment Operation
C-4 

Table C-1 Potential incidents possible at a spent fuel reprocessing facility [DP-1558]

Low Consequence Intermediate Consequence High Consequence

• Loss of off-gas header volume

• Iodine removal inadequate

(increase iodine accumulation in

krypton solvent recovery)

• Excess nitrogen oxides in VOG

iodine adsorber stream

• High radioactive particulate

releases to building ventilation

filters (increase release to sand

filter)

• Off-gas heater inadequately heats

streams to iodine adsorbers

• High ruthenium adsorber bed

temperature

• High krypton-85 releases

• Filter failure

• Process vent system pressurized

(leads to release to sand filter and

smaller amount to atmosphere

Uranyl Nitrate Operation

• Overflow from storage tank

• High uranium concentration in

recovered acid

• Plugging of instrument lines and

sensors during evaporation

• Steam coil leak in evaporator

reboiler

• Cooling coil leak due to corrosion

• Overflow in concentration system

Uranium Denitration and Reduction Operation

• Pressurization of denitrator

• High uranium concentration in

recovered acid

• Over-pressurization of redactor

• Fire in Ammonia dissociator

cubicle

• Reductor malfunction which leads

to release of uo2 and hydrogen

• Denitrator eructation (leads to fire

and explosions)

• Hydrogen explosion (hydrogen

leaked from reductor)

Uranium Hexafluoride Operation

• Uranyl nitrate solution leaks

• Uranium hexafluoride release

• Airborne uranium oxide

• Criticality due to excessive

uranium accumulation (high

enrichment fuel)

• Fire (due to solvent in feed,

hydrogen leak and presence of

pyrophoric oxide powders)

Co-conversion Process Operation (UO2-PuO2 production)

• Transfer errors

• Process solution leak

• Overflows

• Chemical addition error

• Pressurization of calciner

• Calciner breached from internal

corrosion

• Calciner breached from impact

(crane drop)

• Excessive penetration of calcine

through the calciner primary filter

• Filter system breached

• Fire suppression system failure

• Uranium in steam condensate or

cooling water returns the uranium

oxide dissolver

• Loss of power

• Ammonium nitrate explosion in

calciner

• Hydrogen explosion in product

storage due to leak in hydrogen

delivery system

• Hydrogen explosion in Pu Nitrate

storage tank due to radiolysis

• Uncontrolled reactions

• Fire in process room

• Hydrogen explosion in cold

chemical area

Waste Calcination Operation

• Excessive penetration of calcine

through sintered metal filters

• High temperature breach of the

calciner

• Calciner breached from internal

corrosion, thermal shock,

pressurization and impact

• Filter systems breached

•

Solid Waste Processing Operation
C-5 

Table C-1 Potential incidents possible at a spent fuel reprocessing facility [DP-1558]

Low Consequence Intermediate Consequence High Consequence

• Airborne cement dust in grout

mixer area

• Radioactive contamination in

cement preparation area

• Failure of contaminated process

components

• Drums improperly filled (no

cement added)

• Violation of stack release guide

• Container failure

• Power failure

• Fire in fuel hardware fixation area

• Drum overflow in fuel hardware

fixation area

• Excessive fissile material in hulls

• Waste container failure after filling

• Fire in β-ɣ waste facility

• Criticality potential in α waste

Solidification of Intermediate level liquid waste

• Transfer error that leads to high

activity waste

• Overexposure of personnel to

radiation

• Mixer plugging, which leads to

worker exposure

• Airborne activity due to leaks,

overflows, foam-out from mixer,

suckback, and loss of vessel

ventilation (power failure)

• Waste container failure due to

weld failure, fault in drum,

corrosion, or impact

• Fissile material in feed

• Uncontrolled reaction in mixer or

product container

Vitrified High-Level Waste Storage Operation

• Contaminated canisters

• Water loss from storage pool

• Loss of cooling water and

shielding

• Canister stress corrosion

• Canisters raised above adequate

shield level

• Canisters dropped during

handling

• fire

1During the head-end process when the spent fuel rod is being chopper or crushed, magnesium cladded fuel and 

uranium/plutonium metal fuel are exposed to air, combustion may occur at normal temperatures [OECD 2005]. 

Zircaloy debris or dust could pose a potential fire or explosion if this material is used for the cladding [IAEA DS360]. 2N-dodecane in TBP in the PUREX process poses potential combustion conditions [Ikeda 2003]. 

5. Much of the description on aerosol behavior seems to be derived from spent fuel type

releases where solid materials become vaporized as aerosols. However, for

reprocessing spent fuel, the aqueous separation process involves aqueous solutions of

nitric acid and organic solvents. Fine powders following heating and evaporation are

other major sources of aerosols. For these situations, where aqueous and organic

liquids and fine powders may be converted to aerosols, the processes appear to be less

understood including the aerosol sizes that are likely to be formed. It is not clear where

vapors are distinguished from aerosols during the source term release processes.

The commenter correctly identifies two general processes for creating aerosols: (1) 

thermal processes that vaporize materials and subsequent condensation of the vapors 

to generate the aerosol particles (e.g., a tea kettle that creates steam that subsequently 

condenses to form fog) and (2) mechanical formation of aerosols by explosions or other 

highly energetic events that generate large pressure waves and induce stresses and 

accelerations. Transport codes, such as MELCOR, can handle the thermal processes 

that form aerosols but lack a model for aerosol generation by mechanical processes. 

The focus of Gelbard’s presentation was on mechanical generation of aerosols, 

specifically creation of liquid aerosols (containing solvent and dissolved or suspended 

radionuclides). The mechanical processes are simulated using a suite of codes known 

as the Sierra Codes.
C-6 

The Sierra codes model the explosion and approximately up to about 10 seconds after 

the explosion. During this period both gases and particles can escape from the 

processing equipment. The processes modeled with the Sierra codes are particle 

formation by the forced rapid ejection and breakup of the liquid by mechanical 

processes. We simulate this release period by modeling both the gases and particles, 

with the assumption that all the gases spread throughout the room containing the 

equipment that has failed. The drops may turn into solid particles by vaporization of 

volatile liquids. The Sierra models also include liquid drop particles vaporizing. 

Subsequent to the release period, MELCOR models any further particle vaporization, 

condensation, deposition, and transport to other locations. Thus, the output from the 

Sierra codes is used as input for MELCOR. (MELCOR does not model the initial release 

period when material is ejected from processing equipment.) The particles, whether 

they are liquid or solid are called aerosols when suspended, but are no longer 

considered to be aerosols if deposited on fixed surfaces such as a wall, floor, or even 

furniture. 

6. Fission products, transuranic nuclides and activated materials should be considered in

development of the source term release models.

We agree with this statement and intend that the tool will include each type of

radionuclide at the discretion of the user.

7. For chemical releases, it is our understanding that chemical reactions, or reactive

transport models, will not be considered in the source term models.

We agree with this statement, with a few caveats. First we do intend to include phase

change. This would be done by introducing an expression for vapor pressure as a

function of temperature. Second, empirical fire models convert organic solvents to

combustion products and energy. We do not intend to include a general treatment of

either thermodynamic equilibrium or reaction kinetics in the model.

8. The major assumptions listed in the overview presentation appear reasonable, although

additions are expected to be needed as we begin to fully understand the processes

involved and when further information becomes available.

We agree.

9. The equation for the source term, ST = MR x DR x ARF x RF x LPF, is based largely on

empirical correlations and may not be applicable to source term releases from

reprocessing facilities using aqueous separations. Limitations of this equation need to be

recognized for reprocessing. The parameters, data for the parameters, and all

assumptions need to be thoroughly examined for application to reprocessing facilities.

The proposed mechanistic modeling would seem to be the appropriate way to address

reprocessing facility source terms. A deterministic approach may be reasonable, but a

probability approach using Monte Carlo approach or Latin Hypercube Sampling would

provide a more realistic means of assessing source term releases.

We agree with the assessment of the five-factor equation often used by the DOE, and it

is our intention to improve on this traditional approach to the extent that we are able

within the scope of the project.
C-7 

The initial source-term model is being developed as a deterministic tool, but there are a 

number of parameters needed to run the model. It is for these parameters that a 

probability approach could be used in the future, but is not planned as part of this work, 

which is focused on development of the model. Initially, the user of the source-term tool 

may need to perform sensitivity analyses to determine ranges of results based on 

reasonable ranges of input parameters. A probabilistic approach based on the 

deterministic model could be developed as a follow-on effort if this would have value to 

the NRC.

10. What is the approach used to model red-oil explosions, oxalate explosions, and

potentially hydrogen (from radiolysis effects) explosions?

All explosions are modeled with the Sierra codes. What differentiates the explosions for 

red-oil, oxalate, or hydrogen are (1) the amount of energy released, (2) the size of the 

explosive region within the system, (3) the shape of the explosive region (such as 

confined to a thin layer, or a small puck-shaped region), (4) the amount of other fluids in 

the system, and (5) the location of the explosive (i.e. where in the processing 

equipment). The Sierra models can simulate the explosive release of liquids to form 

particles once these parameters are specified. Parameters associated with the five 

features listed above will be different for red-oil, oxalate, and hydrogen.

11. The slide “COEX-like Process – Daily Throughput Diagram” in the presentation-Fuel

Reprocessing Facility and Accidents – is a model of an extraction process separating

and combining plutonium with uranium to produce oxide powders for storage and further

mixed-oxide reactor fuel. Let’s call this Co-Extraction Process I. Another figure should

be added that more closely follows the COEXTM – AREVA separation process. We will

call this figure Co-Extraction Process II. Both concepts should be discussed in terms of

the co-extraction process for the aqueous separation of uranium and plutonium.

Because we have no license at this time for the construction and operation of a spent

fuel reprocessing facility, it is unknown which of the two separation processes may

ultimately be submitted. But it is always easy to scale back from the more sophisticated

proposed process, Process I, to the lesser complicated process, Process II, than it is to

scale up an analysis from Process II to Process I. This needs further discussion.

We agree that both concepts should be included, since they both have a U product 

stream and U-Pu oxide product stream. The only difference between them is in the 1st

and intermediate extraction cycles where U and U/Pu streams are produced. The other 

emphasis is the fission product separations in terms of the end FP stream(s) desired. 

Currently, it is our intention to include Process I (as a detail description, including flow 

diagram, conceptual plant layout, etc., since we have the information) and the 

description of the conceptual plant in our Task 2 report. If the description of Process II is 

available, similar to the level of Process I, we may want to include Process II in the writeup; otherwise, we may include it as a process description. The Barnwell plant has been 

included in the write-up at present.

12. The presentation Fuel Reprocessing Facility and Accidents appears to discuss the

important aspects of an aqueous separation process for the co-extraction of uranium

and plutonium, and provide background on the history of chemical explosions, major

accidents at reprocessing facilities, and the role of engineered safety features. Some

editorial comments regarding the presentation are:
C-8 

a. The color coding of Slide 2 is not clearly stated on the slide. First-time readers

may have some difficulty in understanding the color scheme. Also some boxes

in the separation diagram are not colored.

b. Slide 5 should have columns for “Ci/MTHM” for the major radionuclides.

c. Slide 5: Indicate initial enrichment percent of U-235 (e.g., 5%).

d. Indicate color scheme for Slide 6.

e. In upper right hand corner indicate U-235 enrichment percent.

f. Be aware that this slide provides a separation for all of the TRUs that will be

combined with uranium to form the oxide fuel, while Process II (Comment 11

above) only provides for uranium and plutonium to be separated together. The

remaining TRUs will be combined with the fission products and vitrified in glass

for ultimate disposal as HLW.

g. In Slides 8-11 somewhere, indicated the plutonium oxalate problems that

occurred at the Russian facilities. May want to reference some of the papers

published by AREVA regarding the plutonium and uranium oxalate separation

issues.

h. Spell out names for acronyms when the first appear (e.g., HAN, JCO)

i. Provide list of chemical and expected concentrations that may be expected to be

used in the separation processes.

j. Slide 21: Why is nuclear Excursion Accident type listed as a minor public severity

for source term releases from reprocessing facilities? The analyses have not yet

been performed for reprocessing facilities so one cannot predetermine a severity

index for both workers and the public

k. Back-up slides are instructive.

These corrections are noted and will be incorporated into the final report for Task 2.

13. We may wish to select particular processes and/or vessel configurations that should be

included in future computational aerosol dynamics simulations of airborne releases from

detonations (F. Gelbard’s presentation). There are a number of different types of vessel

configurations that are incorporated in the reprocessing separation processes. Among

them are vessels for centrifugal separation, liquid and solid storage, centrifuge

operations, evaporation processes, calcination heaters, etc. These need to be

considered.

The plan is to compute releases for different vessels using the detailed code (in Sandia’s

Sierra code suite) that produced the movie shown at the meeting. Only for

demonstration purposes did we use a simple cylinder. The Sierra codes will be used for

developing particle release correlations for the different vessels, and these correlations

will be incorporated into the tool provided to the NRC. As discussed below for comment

#21, the Sierra code suite will not be part of the tool.

14. Similarly, we need further discussion of the direction on the “prototype” plant design that

will be used to establish default parameters (Slide 5 of Synopsis presentation).
C-9 

We agree. It would seem desirable for Sandia to work more closely with those who are 

developing the prototype plant design so we can better understand evolutions in the 

prototype design.

15. We need to ensure that SNL has captured all of the possible accidents that could occur

at a reprocessing facility since the modeling is going to be responsive rather than

predictive. With that, there is a need to address the off-gas system and consider

possible accidents. If one takes a look at the diagram on page 6 of David Louie’s

handout (Fuel Reprocessing Facility and Accidents), in particular the off gases coming

from the fuel shearing/voloxidation box, Ru gas is a major component. Ru is not a gas

that we have previously considered. Will this be stored temporarily on site until it has

decayed? It is a beta emitter with a half-life of approximately 374 days. It decays to Rh106, which in turn immediately decays (~ 30 s) to Pd-106 and emits intense gamma rays

at 512 keV and 622 keV, which could cause a problem with shielding. Also, they need to

consider the cryogenic storage of Kr-85 and its release if the cooling system fails.

It is our intention to cover all anticipated scenarios for the spent fuel reprocessing plants

in the report (see comment 4 response). Note that the source term tool we are

developing is general and relies on the user to describe the accident simulated. To

evaluate if the shielding is adequate or not for Ru decay and subsequent decay

products, the user may need to treat this issue outside the source term tool, which is not

intended to evaluate radiation shielding.

16. Following on from a comment in the meeting about considering iodine in the waste

solvent. Iodine is likely to be in the solvent at very low concentrations; in the order of

ppm. Therefore, Nate was right and we don’t need to consider it in terms of release

during an accident. Isn’t 129I a major radionuclide release that is needed to assess both

the dose and activity compliance levels with EPA’s 40 CFR 190 regulations?

Iodine-129 may need to be included as a fission product in a source-term analysis;

however, it is not one of the more important isotopes for accident analyses because of

its long half-life and low specific activity. On the other hand, iodine-129 may be important

for normal plant operations if it is routinely emitted from the reprocessing facility. The

difference is that iodine-129 could build up in the environment over an extended period

of time due to years of plant operation, depending on the fraction of iodine that is

released routinely as opposed to being captured in the off-gas processing.

17. We like that SNL is looking at different types of fuel, namely different cooling times and

burn ups.

We will continue to take account of the fact that a range of fuel types and fuel ages are

likely to be reprocessed.

18. Need to check flow diagram on slide 6 of David Louie’s presentation. Where was the

data obtained and how was it developed?

To avoid making the presentation OUO, this information was not provided. It can be

provided separately if needed, but the information would need to be treated as OUO.

19. MELCOR would seem to be the best code for the task. However, we need to consider its

limitations. For example, there is no plume rise model available. Is this a concern?
C-10

We agree that limitations of MELCOR and other codes need to be considered in the final 

evaluation and selection process, but that is part of our Task 3 that will be performed in 

the future. 

Plume rise, in the context of an atmospheric release, is properly treated as part of a 

level-3 consequence analysis, so this is not an issue for a source-term tool. Plume rise, 

in the context of a fire model, is discussed under Comment 3.

20. A definition of source term releases is provided on Slide 2 of Fred Gelbard’s 

presentation. Should not this definition be expanded to include radioactivity levels (e.g., 

Curies) and gaseous releases (e.g., 85Kr, 14C). 

We agree that gaseous releases should have been specifically called out. The source 

term tool needs to ultimately produce release information in terms of activities, and this 

should have been discussed as well.

21. It is not clear if the aerosol model Fred Gelbard presented, actually shown as a nice 

movie, will be incorporated as part of the source term code. Will there be size limitations 

that need to be considered in running a model of this type from our desktops?

From the Sierra calculations, correlations for releases will be developed and either 

incorporated into the tool or a methodology for their inclusion in the analysis process will 

be provided to the NRC. The Sierra calculations typically require days of computation 

on a very large parallel computer, and hence are not appropriate for desktop computers.




NUREG/CR-7232 Review of Spent Fuel Reprocessing and Associated Accident Phenomena