i

Three Mile Island

Accident of 1979

Knowledge

Management Digest

The Cleanup Experience

A Literature Review

NUREG/KM-0001, Supplement 2

Office of Nuclear Regulatory Research
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iii

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TABLE OF CONTENTS

ACKNOWLEDGEMENTS.............................................................v

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

1.1 Purpose...................................................................................... 1

1.2 How To Use This Report .......................................................... 1

1.3 Report Organization and Readability........................................ 2

2 MANAGEMENT AND PLANNING.....................................5

2.1 Organizations............................................................................ 5

2.2 Program Planning.................................................................... 12

2.3 Documentation ........................................................................ 20

3 REGULATORY OVERSIGHT............................................23

3.1 Programmatic Environmental Impact Statement .................... 23

3.2 Technical Specifications ......................................................... 27

3.3 Conduct of Operations ............................................................ 28

3.4 Advisory Panel for the Decontamination of TMI-2 ................ 33

4 DEFUELING ..........................................................................37

4.1 Defueling Safety...................................................................... 37

4.2 Planning .................................................................................. 39

4.3 Quick Look ............................................................................. 49

4.4 Crane Operations..................................................................... 53

4.5 Defueling Work Platform........................................................ 54

4.6 Work Inside Containment ....................................................... 57

4.7 Defueling Tools....................................................................... 60

4.8 Defueling Canisters................................................................. 66

4.9 Defueling Operations.............................................................. 72

5 FUEL DEBRIS TRANSPORTATION.................................85

5.1 Planning .................................................................................. 85
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5.2 Shipping Cask ......................................................................... 88

6 DECONTAMINATION AND DOSE REDUCTION..........95

6.1 Management and Planning...................................................... 95

6.2 Dose Reduction ..................................................................... 103

6.3 Decontamination Methods and Agents................................. 107

6.4 Radioactive Contamination Penetration................................ 112

6.5 Recontamination ................................................................... 115

7 PERSONNEL PROTECTION............................................121

7.1 Dose Exposure Control ......................................................... 121

7.2 Worker Wellness Considerations.......................................... 124

8 WASTE MANAGEMENT ..................................................131

8.1 Liquid Waste ......................................................................... 131

8.2 Solid Wastes.......................................................................... 138

8.3 Onsite Interim Waste Storage and Staging ........................... 140

9 BIBLIOGRAPHY ................................................................143

10 REFERENCES .....................................................................149

Figure on Front Cover: The view of lower core support structure inside the reactor 

vessel with fuel debris removed from the core region. Still filled with reactor 

coolant, the next defueling activities were the lower reactor vessel head region 

(debris can be seen under the grid structure) and behind the core baffle plates of 

the core former.

Figure on Back of Front Cover: Mosaic panorama of the reactor core cavity from 

comprehensive video mapping in April 1984. Shown are hanging control rod 

assemblies, and broken fuel rods and control rod upper end fittings on top of the 

rubble bed.

Figure on Page vi: TMI-2 ventilation stack and the top rim of the reactor 

containment building. Ventilation exhausts from the containment building and 

auxiliary and fuel handling building were processed by filtration systems before 

released to the stack (also see the following figure).
v

ACKNOWLEDGEMENTS

The editors of this knowledge management supplement greatly appreciate the 

contributions from the following individuals and organizations in their pursuit to 

preserve knowledge and experiences of the cleanup of Three Mile Island Nuclear 

Station, Unit 2 (TMI-2). The knowledge contributors listed below contributed 

historically important reports, papers, video tapes, presentations, and supplemental 

information. These individuals were actively involved with the cleanup effort 

while employed by the licensee (employee, contractor), U.S. Nuclear Regulatory 

Commission, and U.S. Department of Energy. The management contributors

assisted in the drafting, reviews, and fact checking of this supplement. Many

legacy documents and videos related to the TMI-2 cleanup are accessible from the 

document collections list below.

Knowledge

Lake Barrett

Willis Bixby

James Byrne

Don Marksberry

Michael Masnik

Charles Negin

Gordon Skillman

James Tarpinian

Management

Don Marksberry and Michel Call (Co-Editors)

QTE (Technical Editors)

Theodore Smith

Richard Lee

Damian Peko

Joy Rempe

Jake Fay

Document Collections

NRC Agencywide Documents Access and Management System

Publications Branch of the NRC Office of Administration

NRC Office of the Secretary

U.S. Department of Energy

Idaho National Laboratory

American Nuclear Society

Electric Power Research Institute

Dickinson College Community Studies Center

Pennsylvania State University Libraries

National Archives and Records Administration
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1

1 INTRODUCTION

1.1 Purpose

The safe, expeditious stabilization, defueling, and cleanup (a) of Three Mile Island 

Nuclear Station, Unit 2 (TMI-2), including removal of the fuel from the 

accident-damaged reactor, were necessary for the long-term protection of public 

health and safety and the environment. The goal of the cleanup campaign was to 

ensure that the TMI-2 site would not become a long-term or permanent waste 

repository. The recovery activities that unfolded at TMI-2 in the weeks and 

months (and then years) after the March 28, 1979, accident were the result of a 

multiorganizational effort that included hundreds of dedicated and highly skilled 

individuals. About $2.4 billion (in 2019 dollars) (1) were spent by the time the 

defueling was nearing completion in 1989.

The purpose of this supplement is to catalog many of the experiences and insights 

documented in numerous reports and papers spanning the 1980 to 1993 cleanup 

period at TMI-2. The experiences in this report focus on those aspects of TMI-2 

relating to long-term plant stabilization, cleanup, and defueling. The description of 

these experiences are based on an extensive review of a wide range of reports, 

papers, presentations, and interviews with personnel formerly from the key 

organizations involved in the cleanup. Many insights were taken from discussions 

during a workshop that was organized by the U.S. Department of Energy (DOE) 

and the Idaho National Laboratory (INL) in 2016. This workshop brought together

many alumni involved with the TMI-2 cleanup and research, and 31 invited guests 

from 7 Japanese organizations involved with the cleanup of the Fukushima 

Dai-ichi Nuclear Power Station. 

To keep the number of cleanup experiences and insights to a manageable size, the 

list was limited to those previously documented in reports, papers, and 

presentations. Text searches using key words, such as “lesson,” “problem,”

“insight,” and “concern,” helped to sift through over 120,000 pages of documents 

to identify notable insights. Discussions with those involved in the cleanup 

significantly helped to identify documented experiences, lessons, and insights. The 

transportation of radioactive wastes and the long-term post-defueling monitored 

storage of the plant are not within the scope of this report.

1.2 How To Use This Report

Experiences and insights from TMI-2, Chernobyl, and Fukushima Dai-ichi

revealed that each severe accident behaves differently, and the subsequent cleanup 

will also be different. Nevertheless, experiences and lessons from these accidents 

a The term “recovery” is used in this NUREG/KM to mean actions taken to keep the plant 

in a stable condition and to prevent the inadvertent release of radioactivity. The term 

“cleanup” is used to mean action taken to decontaminate and defuel the plant and dispose 

of radioactive waste. These two terms are often used interchangeably for certain actions.
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and other large-scale cleanup programs can be adapted to most such programs, 

especially programmatic experiences.

This supplement is a continuation of Supplement 1 to NUREG/KM-0001, “Three 

Mile Island Accident of 1979 Knowledge Management Digest: Recovery and 

Cleanup.” (2) Supplement 1 includes summary descriptions of structures, systems, 

components, and activities associated with the recovery and cleanup at TMI-2.

Most importantly, most open-source references cited in this supplement can be 

found in the DVDs associated with Supplement 1. The document collections from 

these DVDs can be accessed at this time through the Idaho National Laboratory 

Research Library Digital Repository (https://tmi2kml.inl.gov).

Thorough overviews of the TMI-2 recovery and cleanup, including many 

experiences and lessons, can be found in the series of publications by the 

American Nuclear Society, Electric Power Research Institute (EPRI), International 

Atomic Energy Agency (IAEA), DOE, and NRC. Section 10 of this report lists

these publications. The reader is strongly encouraged to refer to these excellent 

sources for details associated with the experiences summarized in this report, as 

well as additional experiences and insights. Each experience or insight has a 

numbered source, which is listed in the reference section of this report.

The reader should be cautioned that, while the experiences documented in this 

report came from largely previously published sources, they may have been 

superseded by subsequent experiences at TMI-2, lessons from other cleanup 

projects, and research results, as well as changes in regulatory requirements. 

Contributions to this report by knowledge providers and their organizations should 

not be considered as endorsements of the contents of this report.

1.3 Report Organization and Readability

Experiences are organized into major subjects with a brief background section to 

orient the reader to that subject. The subjects are divided into subtopics, each with 

a brief discussion and a series of experiences. The experiences are brief, and each 

is preceded with a keyword phrase to highlight its specific topic. References are 

given so that the details of the experience can be further investigated.

To improve readability, abbreviations were minimized (b) and others that were 

frequently repeated were spelled out in the section that contained them. The text 

was structured using past tense, except for text describing this supplement. 

b Abbreviations used throughout this supplement: as low as reasonably achievable

(ALARA); Code of Federal Regulations (CFR); U.S. Department of Energy (DOE); 

Electric Power Research Institute (EPRI); General Public Utilities Nuclear Corporation 

(GPU); Idaho National Engineering Laboratory (INEL, now called INL); International 

Atomic Energy Agency (IAEA); U.S. Nuclear Regulatory Commission (NRC); 

programmatic environmental impact statement (PEIS); and Three Mile Island Unit 2 

(TMI-2).
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Footnotes provide editorials and additional clarifications not specific to the 

historical record. Some terms, which varied in use across documents and time,

were standardized. The text uses the unit of measure (in English units or the 

International System of Units) in the original source document (in some cases, the 

source provided both). A conversion chart is provided on the back cover. End 

notes refer to the source reference. The end note number at the beginning of the 

subsection (or bullet) refers to the entire subsection; the number at the end of a 

paragraph refers to that paragraph; the number in a sentence only applies to that 

sentence. Due to limitations in word processing, end notes embedded in figure 

captions are listed in the last endnote (see Section 10).

Section 2 provides experiences of the high-level management and planning of the 

overall cleanup program, with further descriptions included in the first subsection 

of subsequent sections. Section 3 describes unique experiences of the NRC’s 

regulatory oversight of the TMI-2 recovery and cleanup activities. Experiences 

from regulatory oversight of the TMI-2 cleanup had not been widely documented 

previously. Sections 4 through 8 describe experiences from an assortment of 

topics, including long-term stabilization; defueling, packaging, and onsite storage

of radioactive waste; the front end of the fuel debris transportation campaign; 

decontamination of the auxiliary and fuel handling building and containment

building; personnel protection; and onsite waste management. Section 9 is a 

bibliography of important resources that preserved the decisions made and 

associated consequences during the TMI-2 cleanup. The final section, Section 10,

lists the references cited in this report. The format of most references provides the 

file name of those documents included on the DVDs for NUREG/KM-0001, 

Supplement 1.

Figure on Next Page: Temporary filtration system for the auxiliary and fuel 

handling building installed on the auxiliary building roof. TMI-2 containment

building in the background; fuel handling building at center right.
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2 MANAGEMENT AND PLANNING

Many technical decisions considered issues on funding, public perception, and the 

regulatory environment. Each decision also involved a choice among several 

strategies, such as manual or robotic techniques to defuel the reactor; 

demineralization, evaporation, or solidification to process radioactive water; and 

gross decontamination or dose reduction to support the other cleanup work. The 

most important technical influence on decisionmaking was the relevant data 

available when a decision was made. In many cases, limited data or inaccurate 

assumptions about conditions were serious handicaps to both planning and 

operations. A central lesson of the TMI-2 cleanup was the importance of 

proceeding methodically to understand conditions, to develop a simple 

engineering approach to handle known conditions, and then to repeat this sequence 

until recovery operations were complete. (3)

The section covers experiences on organizations, project and work planning, and 

the retention of knowledge.

2.1 Organizations

Organizations that supported the licensee included the original architect engineers 

of Three Mile Island Nuclear Station (TMI), Units 1 and 2 (Gilbert Associates and 

Burns and Roe, respectively); the TMI-2 nuclear steam supply system vendor 

(Babcock & Wilcox); many volunteers from other nuclear power plants; the 

U.S. nuclear industry; universities; research laboratories; and several international 

organizations. (4) The licensee’s two principal support contractors were Bechtel 

Northern Company (architect engineer and construction) and Bechtel National, 

Inc. (decontamination and technical support). (5)

The licensee instituted several review and advisory groups that focused on cleanup 

activities. In 1983, the NRC approved a change in the licensee’s organization that 

created an onsite safety review group for independent review and audit. Although 

not required by regulatory requirements, other corporate review groups were 

established by the licensee to provide independent technical oversight on safety 

and cleanup technology issues, such as the Technical Assessment and Advisory 

Group and the Safety Advisory Board. (6, 7)

The DOE provided the technical assistance needed to enable removal and 

evaluation of the damaged reactor core and to perform other safety and severe 

accident research for the benefit of nuclear power technology. GPU, EPRI, the 

NRC, and the DOE formed a collective group called GEND to jointly sponsor and 

participate in the DOE’s TMI Information and Examination Program. (8)

• Program Uniqueness. All organizations involved in the cleanup of TMI-2 

realized early in the cleanup effort that the stabilization, defueling, 

decontamination, and decommissioning of an already damaged reactor were

unique and unlike anything previously encountered in U.S. commercial nuclear 
6

power reactors. The cleanup required special efforts to bring together all the

resources needed to ensure worker and public safety. Past regulatory practices 

alone were neither sufficient nor appropriate. All organizations required a 

documented, unified common goal of a safe and prompt cleanup of TMI-2. The 

licensee, support organizations, and the NRC had clearly defined and documented 

roles, responsibilities, and authorities. (9)

Expertise from external organizations was needed because many aspects of the 

cleanup were beyond the expertise of the licensee. The cleanup required skills and 

special facilities from the DOE and its national laboratories that did not exist 

elsewhere. Expertise from other utilities, service companies, and universities 

provided valuable resources. Combining these outside experts with the onsite work 

force was difficult; however, the combination brought much needed technical 

support, new ideas, and a channel to the worldwide technical community. (10)

• Safety Review Group. (11, 12) In 1983, the NRC approved a change in the 

licensee’s organization that created an onsite safety review group (SRG) for 

independent review and audit. The SRG was a permanent, full-time group of 

qualified individuals designated to perform this function, replacing pre-accident 

review committees which convened periodically and sometimes with rotating 

personnel. The staff were assigned onsite and reported to the TMI-2 licensing and 

nuclear safety director. The SRG conducted an ongoing program to evaluate the 

technical adequacy of those procedures and design changes important to safe 

operation of the plant as required by regulatory requirements and implementing 

procedures. Additionally, the SRG reviewed results of audits conducted by the 

quality assurance department and made recommendations, as appropriate. The 

SRG operated independently from both plant operations and engineering and had 

the charter to advise the director of TMI-2 on all safety matters. The SRG manager 

had the authority and the responsibility to bring to the attention of the licensee 

president any issues that were not being addressed with adequate consideration of 

nuclear or radiological safety. The SRG consisted of a manager and at least 

5 engineers with a bachelor’s degree in engineering or physical sciences and 

5 years of experience in the nuclear power field, or 9 years of appropriate 

experience. In addition, several technical analysts were included in this group to 

conduct operational trending analyses.

The safety review topics included: (●) written safety evaluations of changes to the 

facility, procedures, or tests and experiments, as described in the final safety 

analysis report, technical evaluation reports, or select system descriptions; 

(●) proposed changes in the facility, procedures, or tests and experiments, of 

which involved a change in the technical specifications or an unreviewed safety 

question; (●) proposed changes to technical specifications; (●) violations, 

deviations, or select reportable events to the NRC; (●) investigation of all 

violations of the technical specifications, as well as recommendations to prevent 

recurrence; (●) special reviews, investigations or analyses and reports, as 

requested; (●) summaries of audit reports of activities listed in the TMI-2 recovery 

quality assurance plan; (●) recognized indications of an unanticipated deficiency 
7

in some aspect of design or operation of structures, systems, or components, that 

could affect nuclear safety or radioactive waste safety; and (●) any other matters 

involving safety operations which the SRG deemed appropriate for consideration, 

or which was referred to the group.

• Independent Advisory Committees. The licensee voluntarily established two 

subsequent advisory groups: Technical Assessment and Advisory Group (TAAG) 

and Safety Advisory Board. The TAAG provided technical critiques of plans and 

operations, and the Safety Review Board reviewed safety practices. These two 

functions were kept separate to ensure clear definition of purpose; however, both 

groups were aware of each other’s activities. Thus, the technical group considered 

safety aspects in its recommendations, and the safety group was aware of the 

technical practicality of its recommendations. Both groups reported to the 

licensee’s corporation president. The NRC and the DOE observed meetings of 

both groups.

o Technical Assessment and Advisory Group. The TAAG evaluated the cleanup 

in terms of experience and techniques that proved successful in other 

industries as well as the nuclear industry. This group ensured that approaches 

to the various cleanup and defueling operations were technically adequate and 

that consideration was given to keeping radiation exposures at “as low as 

reasonably achievable” (ALARA) levels. The TAAG consisted of about

10 permanent members with a broad range of experience and additional ad 

hoc members when their special expertise was needed. The group responded 

to specific requests from the licensee’s recovery organization, the NRC, and 

the DOE. The DOE funded the group’s work through INEL. (13)

o Safety Advisory Board. The safety advisory board, which was different from 

the safety review group, evaluated the cleanup with a focus on public and 

worker health and safety. The board membership was composed of nationally 

and internationally recognized specialists in the fields of nuclear science, 

engineering, physics, economics, government and medicine. Members were 

drawn principally from university faculties and government research 

laboratories. The first chairman, Dr. James Fletcher, was former 

Administrator of NASA and returned to that position in 1986 following the 

Space Shuttle Challenger accident. Dr. Robert. Marston, who succeeded him, 

was former Director of the National Institutes of Health and former President 

of the University of Florida. The board met every 3 months and reviewed 

many aspects of the recovery activities, including regulations; nuclear 

criticality safety; worker and public safety; risk assessment; project 

organization; project financing; project procedures; technical planning; public 

communications; and conflict resolution. The board’s final report (14) 

summarized its activities from its establishment in 1981 through its final 

meeting in December 1989. (15)

• Integration of Licensee Organizations. In September 1982, the licensee

integrated its two principal support contractors, Bechtel Northern Company 
8

(architect engineer and construction) and Bechtel National, Inc. (decontamination 

and technical support), directly into its organizational structure. This 

reorganization reduced redundancy and concentrated the licensee’s resources on 

completing the cleanup and supporting the restart of Unit 1. Integration occurred at 

all levels, with licensee workers reporting to Bechtel managers and vice-versa. The 

organizational hierarchy remained flexible. For example, replacement of a licensee

manager who left the organization by a Bechtel employee was possible, depending 

on who was best qualified to fill the opening. Many organization charts did not 

even refer to the parent company of the employee. This lack of emphasis on 

corporate identity helped to create a TMI-2 team feeling. The team approach 

included contractors that provided much of the union manpower for the cleanup 

activities. However, experience at TMI-2 showed that the integration was not 

achieved to the same degree within all parts of the organization. In addition, 

jurisdictional and organizational disputes were not always eliminated. 

Nonetheless, for the companies involved, there was a general lack of corporate 

posturing and competition. (16, 17, 18)

• Separation of Operations and Research. (19, 20) A clear demarcation of 

responsibilities and the creation of onsite organizations for coordination and 

cooperation allowed plant cleanup and research to proceed efficiently and to 

complement each other. Throughout the TMI-2 cleanup, the DOE was primarily 

The licensee’s organizational structure at the beginning of defueling 

operations as described in the TMI-2 organization plan (supplement to the 

recovery technical specifications). The manager of the safety review group 

(different from the safety advisory board not shown on the chart) reported to 

the director of TMI-2 with the responsibility and authority to notify the 

licensee’s president of inadequate consideration of nuclear and radiological 

safety issues. (481.1)
9

responsible for conducting research, with EPRI providing research direction in a 

few specific areas. The DOE assembled, over time, a large team at TMI and at the 

national laboratories to support various research programs. While the licensee’s 

staff was essential to assisting the DOE and performing many of the research 

tasks, the success of the research program was due, in large measure, to the 

division of research and operations responsibilities. 

The licensee’s management was focused on the accident cleanup and sometimes 

development of the tools, systems, and procedures needed for cleanup and sample 

acquisitions. The DOE’s management was focused on the research. Each could 

devote most of its resources to its area of responsibility, with the necessary 

integration occurring primarily at working levels within the two organizations at 

the TMI-2 site. In addition, the licensee allowed DOE contractor personnel access 

to the TMI-2 facilities and permitted them, in special instances, to perform 

research and data acquisition tasks.

Additionally, in a cooperation agreement (21, 22) between the DOE and 17 Japan 

nuclear power organizations, the Japanese research staff at TMI-2 and DOE 

national laboratories worked in almost every area of the TMI-2 cleanup project 

during a 5-year period in the latter 1980s.

• The U.S. Department of Energy. (23) The uniqueness of the accident recovery 

and cleanup provided significant opportunities for reactor safety research and the 

DOE and its contractors were best suited fully to exploit them. In recognition of 

these opportunities, it was decided to establish a four-party coordination 

agreement among the licensee (GPU), EPRI, the NRC, and the DOE, collectively 

referred to as GEND. Accordingly, in March 1980, 1 year after the accident, the 

four parties signed a basic agreement in which each agreed to cooperate in areas of 

common research and to disseminate fully the results of the TMI-2 recovery 

operations to the world.

o The DOE’s Contributions. As its contribution to the effort, the DOE agreed to 

help develop the unique technology needed to enable removal and evaluation 

of the damaged reactor core and to perform other safety and severe accident 

research. In 1982, the DOE role in waste immobilization and reactor 

evaluation was substantially augmented in recognition of the unique 

capabilities of the U.S. Government for ensuring safe isolation and disposal of 

radioactive waste materials, as well as to conduct associated research and 

development that would be of general benefit. The DOE also agreed to accept 

the damaged core and transport it to its Idaho site for temporary storage and 

research. The DOE selected the INEL to manage its research and development 

program and to conduct a substantial part of the work at its laboratory in 

Idaho. In addition, many other DOE laboratories and contractors participated 

in the program, including Oak Ridge National Laboratory, Pacific Northwest 

National Laboratory, Hanford Operations, Sandia National Laboratories, Los 

Alamos National Laboratory, and Brookhaven National Laboratory.
10

o Technical Integration Office. In 1980, the DOE established a Technical 

Integration Office at TMI-2 to provide an onsite presence to carry out the 

DOE’s work more efficiently. Over the cleanup period, the office served as 

the DOE’s primary data gathering and distribution group. Its primary tasks 

included: (●) providing technology support to the licensee for recovery 

operations; (●) supporting the core debris shipping program through onsite 

preparations and monitoring; (●) providing samples and other data in support 

of the accident evaluation effort; and (●) disseminating technical information 

to the public, industry, and scientific community.

For most of the cleanup, the DOE maintained a relatively small department

staff and a somewhat larger contractor staff at the TMI-2 site. These onsite 

technical personnel had routine access to the plant and were permitted to work 

directly with the organizations conducting the cleanup. The DOE manager at 

TMI-2 reported to the DOE’s Idaho Operations Office but had considerable 

personal budget authority and the latitude to discuss issues directly with DOE 

headquarters officials in Washington, DC. The DOE developed most of the 

specialized equipment for data acquisition and recovery off site and 

thoroughly tested it in mockup facilities before using it at TMI-2.

o TMI-2 Research Budget. The total cost of the DOE TMI-2 research and 

development program was about $188 million (or $382 million in 

2019 dollars). Direct funding of research and development at TMI-2 was 

approximately $78 million. Offsite technology support cost an additional 

$29 million, with the remaining $81 million devoted to research on the 

accident and its consequences.

o Key Experiences. The 10-year involvement of the DOE and its national 

laboratories in the TMI-2 cleanup and research programs yielded many 

lessons of value for nuclear power programs around the world. Some of the 

key experiences included the following: (●) Success at TMI-2 was generally a 

result of innovative engineering applied to existing technology, rather than 

due to the development of entirely new approaches. (●) Data acquisition 

sometimes conflicted with production line work; however, data acquisition 

findings contributed to key program successes. (●) New and improved 

technologies for collecting, concentrating, transporting, and disposing of 

accident-generated radioactive wastes were developed and applied at TMI-2. 

(●) Concern for worker protection during decontamination resulted in a 

variety of innovations, including new surface cleaning techniques; new 

radiation survey equipment to quantify contamination levels; improvements in 

protective clothing; techniques for reducing worker heat stress; and 

improvements in beta dosimetry. (●) Documentation processes captured 

unique technologies developed during the cleanup. (c)

c Editor’s Note: Documentation is also important for long-term cleanup programs that may 

span one or more generations of technical support.
11

Issues and problems breakdown structure in the areas of plant accident response and recovery developed by the DOE’s Technical Information and Examination Program. Data potentially available at TMI-2 would contribute to progress in the general areas of future licensing criteria; analytical model development and verification; accident consequence mitigation; safety system design; system operation; system reliability; and requalification of key light-water reactor components and equipment. (481.2)
12

2.2 Program Planning

The first plan for the decontamination and defueling of TMI-2 (24) was developed 

at the request of Congress and issued in December 1979. However, this early plan 

did not have the benefit of information from the first manned containment building

entry in the summer of 1980 and the first “Quick Look” camera inspection of the 

damaged reactor core in the summer of 1982. The licensee’s “TMI-2 Program 

Strategy Technical Plan” (25)

, issued in June 1984, defined the recovery program in 

three phases: stabilization, fuel removal, and cleanup. The details of the plan 

evolved in view of new data, available technology, regulatory guidelines, and 

financial constraints. The “Strategy for Recovery Program Completion and 

Post-Recovery Configuration” plan (26)

, issued 2 years later, added a fourth phase 

for interim monitored storage. Many task forces and planning studies evaluated

options and provided recommendations in every aspect of the cleanup 

campaign. (27)

• Time at Risk. The cleanup operation involved many risks; however, the time 

that radioactive materials remained outside the fuel cladding was an important 

safety consideration. The risks associated with the cleanup activities to place the 

once-melted fuel radionuclides into engineered containers needed to be balanced 

against the risks associated with fuel debris not contained in an engineered 

containment system for an extended period of time. Both the licensee and the NRC 

balanced these competing risks in a proper manner to assure public health and 

safety. The longer that the once-melted fuel remained in an unengineered

condition, the greater the general overall risk. (28)

• Conservatisms. Experience at TMI-2 showed that excessive conservatism 

applied to reducing defueling risks could lengthen the time that the damaged fuel 

remained in an undesigned condition. This delay could increase overall risks when 

compared with proceeding with timelier defueling using more moderately

conservative criteria. A proper judgmental balance had to be achieved for overall 

Four phases of the TMI-2 postaccident plan: stabilization, fuel removal, cleanup, and 

interim monitored storage. (481.3)
13

societal benefit. (29)

• Innovative Applications of Existing Technology. Success in the TMI-2 

cleanup generally resulted from innovative engineering applied to existing 

technology. Building on existing technology allowed the cleanup to proceed in 

small, steady, incremental steps. In most instances, this proved to be faster and to 

cost less than engineering entirely new approaches to problems. The approach at 

TMI-2 was to try to use existing technology in a creative manner. This produced 

some simple and clever solutions to intimidating problems. The small step at a 

time or “learn as you go” approach was successful because there were so many 

surprises throughout the cleanup program. (30)

• Accommodation of Uncertainties. No one knew all the engineering challenges 

of a cleanup following a severe accident—TMI-2 was not a traditionally plannable 

decommissioning project. TMI-2 showed that uncertainties were large and there 

were many “unknown unknowns” in making “a priori” decisions. Licensee and 

NRC decision processes required a flexible, adaptive, and iterative (step-by-step)

approach that included constant organizational self-reflection to gather lessons

learned from previously performed steps. Classical conservative deterministic 

criteria alone were not completely sufficient. (31)

Although formal risk-informed processes did not exist at the time of the TMI-2 

cleanup, risk-informed aspects that focused on contamination and radioactive 

materials controls were the predominate safety concerns at TMI-2. Traditional 

reactor regulation focused on core cooling and protection aspects. To 

accommodate uncertainties, practical and basic safety precautions, such as 

monitoring and hold points, were established to minimize the time duration of the 

at-risk situation of the damaged core being in an unengineered configuration. 

Parallel engineering on difficult cleanup issues and evolutions provided readily 

available alternatives for the “unknown unknown” situations. (32, 33)

• Importance of Data Acquisition. (34, 35, 36) Balancing data-gathering tasks 

against production tasks was one of the most difficult challenges during the 

cleanup. Because of the DOE’s involvement in the TMI-2 cleanup, there was 

considerable emphasis on research that would benefit the nuclear industry. Early 

in the TMI-2 cleanup, the plant operations staff objected to the delays resulting 

from research tasks. However, it was recognized that advance survey data about 

conditions inside the reactor vessel, containment building, and many locations of 

the auxiliary and fuel handling building were important for cleanup planning.

Data acquisition became such an important part of the overall cleanup program 

that a group was created within the TMI-2 organization to support the DOE’s data 

acquisition needs. This group acquired and interpreted its own data that 

specifically benefited the cleanup program. Competition for resources never

allowed for the complete elimination of disagreements over the value of data 

acquisition tasks, obtained at extra effort, time, and exposure. Nonetheless, there 

was a general recognition that the cleanup, and the condition of the damaged core, 
14

held so many potential surprises that data acquisition was an important part of 

each step in the program. 

• Focus on Water Processing. The potential for TMI-2’s accident-generated 

radioactive water to leak into the Susquehanna River, contaminating both the river 

in Pennsylvania and the downstream Chesapeake Bay, was a major concern for the 

licensee leadership, the leaders and residents of communities downstream, and the 

State and local governments in both Pennsylvania and Maryland. The importance 

and urgency of removing the contaminated water from the entire TMI-2 site, as 

well as the TMI-2 containment building basement, became very significant. There 

was enormous focus on water processing throughout the entire TMI-2 cleanup. 

Decontaminated (processed) accident-generated water was temporarily stored on

site and reused as radiation shielding in the spent fuel pools inside the fuel 

handling building and the deep end of the refueling canal inside the containment 

building. This water was reprocessed and evaporated over a 2.5-year period at the 

end of the cleanup campaign. (37)

• Focus on Defueling. Early in the planning stages of recovering from the 

accident, the licensee envisioned returning TMI-2 to operation. As the extent of 

the damage to the reactor core and the expense of refurbishment became evident, a 

decision was made to work without regard to the final disposition of the plant. 

This decision focused available resources on immediate tasks. (38) In March 1984, 

the licensee changed its main focus at TMI-2 by advancing a “fast track” defueling 

concept to begin removing fuel within a year, with significant fuel removal 

planned to begin in April 1985 (although actual early defueling did not begin until 

later that year). In addition, the licensee focused efforts on modifying regulatory 

requirements to TMI-2 conditions, which involved significantly different safety 

controls and a different public risk profile. The concept focused available 

resources on near-term issues. (39)

• Conflicting Requirements. Potential sources of conflict developed early in the 

TMI-2 program because of differences in regulations and quality assurance (QA) 

practices followed by the licensee, the DOE, and the NRC. To promote smooth 

operations, it became necessary for each party to acknowledge the proper role of 

the other organizations and to sometimes accept the preeminence of another’s 

practices. For example, though many projects were research related, NRC 

regulations (which govern nuclear utilities) generally superseded those of the DOE 

when working with the licensee. When fabricating research equipment, the DOE’s 

QA requirements were generally followed. However, when the equipment was 

brought to TMI-2, the equipment was generally operated under the licensee’s QA 

requirements. (40)

Another example of a conflict that occurred early in the cleanup involved the 

shipment of abnormal waste from TMI-2 to INEL for research. The DOE required 

a task-oriented process for QA. TMI and other NRC licensees used a 

programmatic approach to QA. The DOE required the licensee to develop and use 

special procedures and checklists that were specific to the abnormal waste item 
15

instead of using more generic documentation and procedures, which were used for 

NRC-regulated activities. Further, DOE requirements were different from the 

regulations of the U.S. Department of Transportation and the NRC. For example, 

DOE definitions of fissile material and accountable materials included additional 

radioisotopes. (41)

• Plan with Available Resources. Considerations of bankruptcy and an order

from the Pennsylvania State Public Utility Commission impacted the schedules 

and planning of cleanup activities. The licensee had to consider whether to remove 

the reactor vessel head before all resources (e.g., financial, engineering, personnel, 

training) were available to complete the effort or to reinstall the reactor head, if 

necessary. In late 1980, the public utility commission would not allow the use of 

any operating revenues for cleanup and restoration costs at TMI-2 that were not 

covered by insurance. However, the NRC Commissioners emphasized, in their

policy statement that was attached in a subsequent letter (42) from the NRC 

Chairman to the licensee, that the licensee had to fully comply with all NRC 

health, safety, and environmental requirements applicable to TMI-2, regardless of 

whether these requirements appeared to conflict with the utility commission’s 

order. Further, the Chairman’s letter listed the activities required to be performed 

during the period of ongoing discussions between the licensee and the State 

agency. The list provided the minimum activities required to maintain the TMI-2 

reactor in a safe-shutdown condition and to ensure public and worker health and 

safety and environmental protection in the near term. In addition, the list provided 

some activities required for reducing the intermediate and long-term threats to 

public and worker health and safety and the environment. Although the licensee 

was already performing the listed activities, this list (Table 1) was an early 

“roadmap” for protecting public and worker safety. (43, 44, 45)

• Project End State. Toward the end of the cleanup, some questions arose about 

the end state of the TMI-2 site. The NRC’s independent Advisory Panel for the 

Decontamination of TMI-2 urged the NRC staff to produce a third supplement to 

the programmatic environmental impact statement to address the topic of the final 

end state. The NRC Commissioners directed the staff to produce the 

supplement. (46)

 

The project end state, known as “post-defueling monitored storage,” differed from 

any of the then-NRC-approved decommissioning modes, reflecting the fact that 

none of the established decommissioning end states would be achievable and 

appropriate for TMI-2. Post-defueling monitored storage was determined to be 

beneficial for the following reasons: (●) Occupational dose in the plant would be 

reduced during monitored storage because of natural decay of radioactive 

contamination (the remaining amount of cesium-137 would be 29–50 percent and 

cobalt-60 would be 1.9–13 percent). (●) A monitored storage period would allow 

time for continued development of decontamination technology. (●) Further 

reduction of occupational exposure would be achieved using advanced robotic 

technology, automatic cleaning and chemical cleaning techniques, and advanced 

waste treatment methods. (●) Developing technology for radioactive waste 
16

packaging and volume reduction could result in a reduction in the total volume of 

radioactive waste generated following post-defueling monitored storage. In 

addition, the licensee had stated that placing the TMI-2 facility in storage would 

eliminate any possible impact of TMI-2 decontamination and decommissioning 

efforts on the operating TMI Unit 1 facility. (47, 48)

• End State Specification Criteria. An IAEA working group developed an 

example of end-state specification criteria for the postaccident cleanup end state,

with examples taken from TMI-2 (see Table 2). (49)

Top: The decision-making ranking 

process used by the TMI-2 Strategic 

Planning Task Force to rank or 

prioritize various cleanup tasks 

necessary to achieve and maintain a 

post-recovery configuration. 

Bottom: The ranking criteria included 

acceptance criteria that represented 

essential attributes of potential final 

choices (called “Musts”) and desirable 

features (called “Wants’”). The latter 

were weighed based on importance. 

(481.3)
17

Table 1. NRC staff list of activities required in 1981 to be performed by the 

TMI-2 licensee. (50)

1. Maintain adequate control and confinement of radioactive materials. (Note 1)

2. Minimize the volume of water used in cleanup activities, maximize reuse of 

processed water, and reduce cross-contamination of processed water to the 

maximum practicable extent. (Note 1)

3. Maintain the facility and perform surveillance activities required by the 

technical specifications. Included in this activity was the performance of a test 

to verify stable core cooling by losses to ambient. (Note 1)

4. Perform necessary maintenance, including minor modifications, of equipment 

and facilities (i.e., winterization of the containment building cooling system). 

Included in this activity was the associated training of operators, and minor 

modifications identified as being desirable based upon design reviews and

initial operating experience. (Note 1)

5. Decontaminate the auxiliary building as necessary to facilitate maintenance

and operation of safety-related equipment. (Note 1)

6. Process radioactive waste generated by activities in this list, including handling 

and packaging for offsite disposal. Continue shipments for disposal of 

radioactive material. (Note 1)

7. Perform radiological controls necessary to support activities in this list,

including: surveillance of work in radiation areas; in-plant surveys and 

monitoring; personnel exposure, measurement, and documentation;

maintenance and calibration of equipment; emergency response capability; and 

training. Included in this activity was a continuation of the upgrading of the 

radiological controls program to meet revised performance standards which 

had resulted from the TMI-2 accident. (Note 1)

8. Perform measurement, analysis and documentation of the environmental 

impact of the facility. (Note 1)

9. Perform engineering functions (e.g., review of plant procedures) incident to 

conduct of required operation and surveillance of conditions of plant

equipment, systems, and facilities. (Note 1)

10. Administer the technical and administrative interface with the regulatory

agencies of the federal and state governments, including maintaining 

knowledge of current and proposed regulatory requirements. (Note 1)

11. Provide technically oriented services to implement quality assurance, training, 

emergency preparedness, and independent safety assessment programs. (Note 1)

12. Provide minimum required services in various support functions

(e.g., security). (Note 1)
18

Table 1. NRC staff list of activities required in 1981 to be performed by the 

TMI-2 licensee. (Continued)

13. Provide repair or replacement of the failed source range neutron monitor.

Continue efforts to achieve improved monitoring of core neutronics, utilizing

other instruments. (Note 1)

14. Perform decontamination efforts inside the containment building necessary to 

support required maintenance activities. (Note 1)

15. Complete, subject to NRC approval, an interim solid waste storage

facility. (Note 1)

16. Support finalization of the Draft Programmatic Environmental Impact

Statement. (Note 2)

17. Develop the capability to decontaminate the radioactive water within the 

containment building sump and the reactor coolant system. (Note 2)

18. Continue the technical planning effort, including the gathering of data on 

conditions inside the containment building and the development of criteria to 

be applied to cleanup activities. (Note 2)

19. Provide engineering to support licensing and completion of base line

engineering documents. (Note 2)

20. Complete development, engineering, and design efforts necessary to obtain

NRC approval for construction and operation of a system for solidification of 

EPICOR II liners or propose alternative methods for the stabilization of these 

wastes. (Note 2)

21. Continue decontamination of the auxiliary building. (Note 2)

22. Continue to improve the company’s management control programs. (Note 2)

23. Continue operation of the EPICOR II system on an as needed basis. (Note 2)

Notes:

(1) Minimum activities required to maintain the TMI-2 reactor in a safe shutdown 

condition and to insure public and worker health and safety and environmental 

protection in the near term.

(2) Activities required for reducing the intermediate and long-term threats to public and 

worker health and safety.
19

Table 2. Example of end state specification criteria for the postaccident cleanup 

end state. (51)

Criteria subjects Criteria statements 

(examples based from TMI-2)

Structural and boundary 

integrity

Structural and boundary integrity will be such that: 

(a) inspection personnel are safe, 

(b) contamination or hazardous materials remaining in the 

facility are contained, and 

(c) intrusion by unauthorized personnel, as well as animals and 

plants, are prevented.

Nuclear materials and 

criticality

Nuclear fuel and debris will be removed to the extent practical. 

Residual fissile material must be reduced to a level such that 

criticality cannot occur.

Hazardous materials Hazardous materials and chemicals will be removed in 

accordance with environmental regulations. Fixed in place 

hazardous materials remaining in the facility will be contained 

in limited areas or stabilized to prevent release. The amount and 

location of remaining hazardous materials will be documented.

Process systems and 

equipment

Process systems and equipment have been abandoned in place, 

isolated or sealed off for the safety of future personnel, or 

removed where there is a compelling reason to do so.

Service and utility 

systems and equipment

Only systems required to support the SAFSTOR state and 

maintain the stable condition are operational. Other utility 

systems will be abandoned in place, isolated or sealed off for 

the safety of personnel, or removed where there is a compelling 

reason to do so.

Personnel safety Inspection personnel are safeguarded by stable conditions, 

postings and written procedures established in accordance with 

standard procedures for radiological protection and industrial 

safety practice.

Waste and liquid 

effluents

Waste will have been removed to the extent practical. Waste 

may remain if removal is extremely difficult. The only liquids 

remaining are minor quantities that cannot be readily removed 

with installed equipment.

Radiation protection Established in accordance with standard procedures. In 

particular, the periodic inspection path will be subjected to 

ALARA review. Contamination remaining in the facility will

be contained in limited areas or stabilized to prevent release.

Housekeeping and 

miscellaneous materials

Valuable materials will be removed. Rubbish and noncontaminated furniture, loose equipment, etc. will be removed.
20

2.3 Documentation

• Target the Audience. Documenting the DOE research at TMI-2 was largely 

done by and written for the national laboratory audience. However, the DOE

research had many practical applications and benefits for nuclear utilities, and the 

utility audience required a completely different documentation approach. 

Technical detail and background were less important than specific direction on the 

practical aspects of implementing the results derived from the DOE research. 

Generally, reports for the utility audience were required to be concise and stress 

the cost, performance, or productivity improvement that could be expected. (52)

• Consolidate Reporting. The DOE published many research reports as 

“GEND” documents, even though the work was done by a variety of organizations 

that had their own technical documentation systems. GEND was the acronym for 

the principal participants in the TMI-2 research program: the licensee (GPU), 

EPRI, NRC, and DOE. The GEND reports came to be recognized as an important 

definitive source of TMI-2 data and results. This system allowed multiple 

organizations to publish TMI-2 information in a standard format. GEND reports 

are publicly available with no copyright restrictions. (53,54)

• Archived Data. In 1980, the DOE established a microcomputer database that 

eventually indexed about 20,000 documents on the TMI-2 research and cleanup 

programs. The database was designed to be easily accessed by researchers 

throughout the country. Periodic reviews of the database contents, particularly in 

the early stages of development, were important for ensuring that the most useful 

documents were being included. The reviewers included both technical and 

documentation support staff. (d, 55, 56)

• Repositories of TMI-2 Knowledge. With the advances of information 

technologies since the TMI-2 days, document collections can be easily captured 

and stored for future use by all interested stakeholders. The challenge today is

organizing the vast amounts of electronic media for easy retrieval. The technical 

aspects of the TMI-2 cleanup were well documented. Key sources of TMI-2 

cleanup experiences include the following: (57)

o American Nuclear Society (ANS) (www.ans.org) maintains an extensive 

collection of its journal articles and proceedings on every aspect of the TMI-2 

accident. ANS published the proceedings of the topical meeting, “The TMI-2 

Accident: Materials Behavior and Plant Recovery Technology,” held in 

d Editor’s Note. The TMI-2 database is no longer available for the obvious technical reason 

that the database platform is obsolete. Future endeavors to capture knowledge for 

long-term and future decommissioning projects like Fukushima Dai-ichi and Chernobyl

should be mindful of changes in information technologies as well as changes in 

organizational ownerships. Further, existing nondigital information (reports, photographs, 

and videos) from TMI-2 and other past cleanup programs should be converted to digital

format before being lost or degraded beyond use.
21

Washington, DC, in 1988 in Volume 87 (58) of the Nuclear Technology journal. 

The papers in the proceedings and other papers may be purchased from the 

ANS Web site (free for ANS members).

o Electric Power Research Institute (www.epri.com) (EPRI) made available on 

its Web site many of its research reports that supported the cleanup and 

understanding of the accident. This collection is copyright protected. Six

comprehensive reports on the accident, recovery, and cleanup include,

“Analysis of Three Mile Island—Unit 2 Accident” (59)

; “The Cleanup of Three 

Mile Island Unit 2, A Technical History: 1979 to 1990” (EPRI-NP-6931);

“TMI-2 Waste Management Experience” (EPRI-TR-100640); “TMI-2 PostAccident Data Acquisition and Analysis Experience” (EPRI-NP-7156);

“Radiation Protection Management Programs at TMI-2: Noteworthy Practices 

and Accomplishments” (EPRI NP-5338); and “Decontamination Experience 

During the Cleanup of Three Mile Island Unit 2” (EPRI-NP-7157).

o Idaho National Laboratory (INL) (https://inldigitallibrary.inl.gov) maintains a 

collection of research reports in its INL Knowledge eRepository, including the 

GEND report collection. Most of these documents are on the DVDs for

NUREG/KM-0001. The contents of the DVDs are posted on INL’s Web site

(https://tmi2kml.inl.gov).

o Pennsylvania State University libraries (www.libraries.psu.edu) maintain the 

TMI-2 Recovery and Decontamination Collection that contains several 

thousand videotapes of the recovery and cleanup during the 1979–1990 time 

period, as well as a few reports and photographs. GEND cosponsored a project 

with the university to catalog and maintain the extensive videotape library for 

future researchers. However, very few of the 3,000 video tapes are digitized,

and these tapes are deteriorating at an alarming rate.

o The DOE Office of Scientific and Technical Information Web site

(www.osti.gov) contains over 1,000 electronic full-text reports and papers 

relating to DOE-funded research in support of the recovery and cleanup efforts 

at TMI-2. Most of these documents are on the DVDs for NUREG/KM-0001.

o The NRC (www.nrc.gov) made available all its correspondence and safety 

evaluations with the licensee in microfiche format and an electronic cataloging 

system (the first electronic system for the NRC). Over 25,000 documents were 

preserved on microfiche. The electronic catalog platform has been kept up to

date throughout the years and is available online for public access; however, 

only a very few documents captured on microfiche have been digitized. About 

4,000 of the key documents from the NRC collection are on the DVDs for

NUREG/KM-0001.

Figure on Next Page: The INEL report “TMI-2 Accident Scenario Update” 

(EGG-TMI-7489, December 1986) identified four regions within the original core 

volume: upper cavity void region, debris bed region, previously molten region, and 

partially standing fuel assemblies (or “stubs”) region.
22
23

3 REGULATORY OVERSIGHT

The NRC realized that the recovery and cleanup of an already damaged reactor 

were unique and unlike anything encountered before. Special efforts were required 

to bring together all resources within the NRC and NRC technical support 

contractors. Past regulatory practices alone were neither sufficient nor 

appropriate. (60) Notable successes in the NRC’s oversight of the cleanup program 

can be attributed to the programmatic environmental impact statement (PEIS); the 

onsite and offsite NRC TMI project offices; the NRC concept of operations for the 

TMI-2 cleanup; the recovery technical specifications; and the Advisory Panel for 

the Decontamination of TMI-2. Experiences from these topics are presented 

below.

3.1 Programmatic Environmental Impact Statement

In their November 21, 1979, policy statement (61)

, the NRC Commissioners 

directed the NRC staff to prepare a PEIS on the decontamination and disposal of 

radioactive waste resulting from the accident. The Commissioners’ decision was 

influenced by a lawsuit about 6 months earlier from the City of Lancaster and 

others concerning the proposed operation of the EPICOR II system and by NRC 

discussions with the Council on Environmental Quality, within the Executive 

Office of the President. This holistic review (a thousand-page final PEIS 

document) was needed because the National Environmental Policy Act (NEPA) 

required a complete review of the environmental impact of an action—the TMI-2 

cleanup in this case—and not a piecemeal approach. A piecemeal approach would 

have only looked at incremental impacts, which could be small individually but in 

total could be large. External stakeholders were concerned about these incremental 

impacts, as well as the lack of documented consideration of alternative approaches 

for the cleanup. The PEIS provided the bases for the environmental impact 

assessment for all cleanup activities. (62, 63, 64)

• Scope of the PEIS. (65) The PEIS was intended to provide an overall evaluation 

of the environmental impacts that could result from the various cleanup activities. 

These activities began when the plant conditions were stabilized after the accident 

and continued through the completion of the cleanup. Impacts included the 

transportation of radioactive wastes and fuel debris to offsite storage locations. 

The PEIS included an overall description of the activities and a schedule for their 

completion, along with a discussion of alternatives considered and the rationale for 

choices made. 

o Public Meetings. The NRC discussed the proposed scope of the original PEIS

with representatives of the President’s Council on Environmental Quality, the 

licensee, and several State agencies. Early in the process of developing the 

draft PEIS, scoping sessions with the public took place in Harrisburg and

Middletown, PA, and in Baltimore, MD. After publication of the draft PEIS

for comment, the staff held 31 meetings with the public, local officials, and 
24

interested organizations to obtain, first hand, the comments and concerns of 

meeting participants and to foster an interchange of ideas.

o Uncertainties. Because they were programmatic in nature, the reports were

not intended to provide a step-by-step work plan. However, the most probable 

sequences and methods for cleanup had been assumed to predict the resulting 

environmental impacts. The best available information had been used and 

documented in these impact analyses. Where uncertainties existed, 

conservative assumptions had been made and documented in the main text 

and appendices as appropriate. If, when more information became available

(for example, conditions inside the reactor building and reactor vessel), 

proposed activities were found to be significantly beyond the scope of these 

assessments, the NRC would issue appropriate supplements to the PEIS.

o Scope Expectations. The ultimate disposition of the facility, whether to 

decommission or restore it to a condition acceptable for licensed operation, 

was not within the scope of the original PEIS. In addition, in their policy 

statement of April 27, 1981, which approved the NRC staff’s use of the PEIS, 

the Commissioners wanted to decide on the disposition of accident-generated 

water at a later date. Later supplements to the PEIS further addressed these 

two exceptions.

• Supplements to the PEIS. The PEIS (NUREG-0683, “Final Programmatic 

Environmental Impact Statement Related to the Decontamination and Disposal of 

Radioactive Wastes Resulting from March 28, 1979, accident, Three Mile Island 

Nuclear Station, Unit 2,” issued March 1981) had three supplements that were 

considered part of the original PEIS:

o Supplement 1, “Final Supplement Dealing with Occupational Radiation 

Dose.” The original PEIS stated that the most significant environmental 

impact of cleanup activities at TMI-2 would result from the radiation dose to 

the cleanup work force. This supplement, issued in October 1984, reevaluated

the occupational radiation dose and resulting health effects from cleanup and 

addressed additional alternative cleanup approaches using information 

gathered since the PEIS was prepared in 1980. Higher estimates resulted from 

a more accurate characterization of radiation fields in the reactor building 

based on numerous worker entries. (66) However, by the end of 1989, with the 

fuel debris about 99 percent removed, the collective dose to all workers fell 

within the range estimated in the original PEIS. (67)

o Supplement 2, “Final Supplement Dealing with Disposal of 

Accident-Generated Water.” This supplement, issued in June 1987, updated

the environmental evaluation of accident-generated water disposal alternatives 

published in the original PEIS, using more complete information. The 

supplement also included a specific environmental evaluation of the licensee’s 

proposal for water disposition. (68)
25

o Supplement 3, “Final Supplement Dealing with Post-Defueling Monitored 

Storage and Subsequent Cleanup.” This supplement, issued in August 1989,

evaluated the licensee’s proposal to complete the cleanup effort and place the 

facility into monitored storage for an unspecified period of time. The 

supplement provided an environmental evaluation of the licensee’s proposal, 

as well as several alternative courses of action, from the end of defueling 

efforts to the beginning of decommissioning. However, it did not evaluate the 

environmental impacts associated with decommissioning. (69)

• Approval of Cleanup Proposals: NRC Staff. The NRC Commissioners’ policy 

statement that endorsed the PEIS provided the staff with the authority to approve 

most cleanup activities. The Commissioners stated that, as the licensee proposed 

specific major decontamination activities, the NRC staff would determine whether

these proposals, and the associated impacts that were predicted to occur, were

within the scope of those already assessed in the PEIS. Except for the disposition 

of processed accident-generated water (which the Commissioners wanted to 

decide on later), the staff was allowed to act on each major cleanup activity 

without the Commissioners’ approval if the activity and the associated impacts 

were within the scope of those assessed in the PEIS. (70)

The Programatic Environmental Immpact Statement (PEIS) had three supplements 

that were considered part of the original PEIS.
26

• Approval of Cleanup Proposal: NRC Commissioners. In the year following 

the accident, the NRC Commissioners approved radiological effluent criteria for 

the interim period before the issuance of the PEIS for radiological releases from

data-gathering and maintenance operations. Following the issuance of the PEIS, if 

a cleanup task was evaluated to be outside the scope of the PEIS, then the NRC 

TMI-2 project office (TMIPO) would recommend to the NRC Commissioners 

either their approval of the task or the development of a supplement to the PEIS. 

This action was never necessary. The licensee never submitted a proposal for a 

cleanup activity that TMIPO determined would result in an environmental impact 

outside the scope of the PEIS and its supplements. (71)

• Success of the PEIS. The TMI-2 PEIS became a bounding safety case analysis 

document and served as an envelope within which cleanup operations could be 

efficiently approved by the onsite NRC staff. The NRC TMI project office was 

able to approve proposals, sometimes within days instead of the months that would 

have been needed had the reviews required NRC Commissioner approvals. The 

determination to develop a PEIS that covered the entire cleanup process, rather 

than separate environmental assessments for each major activity, helped to 

expedite environmental reviews. The PEIS’ success was based on the following 

additional contributions: (●) The requirements and intent of the National 

Environmental Protection Act were compiled using a programmatic approach to 

assessing impact. (●) The PEIS summarized the various components of the TMI-2 

cleanup that had appeared in many documents. (●) The document was written to 

be understood by members of the public. (●) The PEIS evaluated alternatives to 

the various evolutions. (●) The PEIS included technical information on the 

cleanup, including estimates of the impact on the environment, workers, and the 

public. (●) Development of the PEIS encouraged and factored in public 

involvement, including 31 meetings with the public, local officials, and interested 

organizations. (●) The PEIS was a living document that ultimately had three 

supplements. (●) Both the industry and the public considered the PEIS to be a 

comprehensive assessment. (72, 73)

• PEIS-Like Document for Other Cleanup Programs. The PEIS satisfied the 

requirements of the National Environmental Protection Act and then became the

basis for the NRC staff’s approval of cleanup proposals. The latter use of the PEIS 

can be applied to other severe accident cleanup programs. A PEIS-like document 

could be useful in the following ways: (●) Represent a national-level 

comprehensive document that addresses impacts from all segments of the cleanup. 

(●) Consider and incorporate input from all stakeholders, including the public, the 

licensee, and government agencies. (●) Evaluate alternatives. (●) Improve public 

understanding of the complexity and difficulty of the cleanup. (●) Be a living 

document, with the publication of supplements, as needed, with public input. (74)

• PEIS Delayed at TMI-2. The NRC did not decide to prepare a PEIS until 

nearly 8 months after the accident. An earlier decision to prepare this document 

would have allowed the PEIS to be published earlier than March 1981, which 

would have facilitated some of the early cleanup activities, including an earlier 
27

startup of the submerged demineralizer system to remove contaminated water on 

the containment building floor. (75) The potential for this radioactive water to leak 

into the Susquehanna River, contaminating both the river and the downstream 

Chesapeake Bay, as well as downstream drinking water supplies, was a major 

concern for the licensee’s leadership, the leaders and communities downstream, 

and the State and local governments in both Pennsylvania and Maryland. (76)

3.2 Technical Specifications

Within a year following the accident, the NRC issued an order that established the 

new “recovery technical specifications” (Appendix A to the facility operating 

license) that considered the condition of plant systems at that time. The purpose of 

these new technical specifications was to ensure that the damaged plant would 

remain in a safe and stable condition during the recovery mode. The then-existing 

preaccident technical specifications imposed for the protection of the environment 

(Appendix B to the facility operating license), including the established limitations 

on effluent releases and discharges, were unchanged and were to remain in effect 

except as provided in the order. (77)

• Recovery Technical Specifications. One important lesson from the cleanup 

was that many of the administrative inefficiencies were related to the constraints 

of the preaccident TMI-2 technical specifications. (78) As a result of the core and 

equipment damage, various requirements set forth in the preaccident technical 

specifications governing operations were no longer appropriate. For example, 

certain equipment that was required to be operable was no longer operable as 

defined in the technical specifications. Other systems not generally relied on for 

safe shutdown of the reactor maintained the facility in a stable mode of heat 

removal. High radiation levels in the containment building, reactor coolant, and 

certain areas throughout the plant had limited personnel access to certain 

components or had limited the ability to operate certain systems or components in 

their original design mode. Several systems and components had been modified to

respond to the initial emergency condition. The postaccident recovery technical 

specifications saved time and resources that would have been needed to modify 

the facility license for each change to the plant’s technical specifications. (79)

• Recovery Operations Plan. The “recovery operations plan” defined the 

surveillance requirements to be performed to ensure equipment operability as 

required by the recovery technical specifications. This plan was included as 

Section 4 of the recovery technical specifications. However, the plan was not 

considered a part of the technical specifications. As such, the NRC staff approved 

changes made to surveillance requirements without the need to modify the facility 

license, which changes to the technical specifications would have required. (80)

• TMI-2 Organization Plan. The licensee’s organization plan provided the 

organizational structure (e.g., charts) for managing the TMI-2 recovery operations, 

including the support functions of engineering and administration. The plan was 

cited in the organization section of the proposed technical specifications. The NRC 
28

approved the licensee’s concept of providing charts of the TMI-2 recovery 

management in the organization plan, instead of in the recovery technical 

specifications, so that future changes could be made effective in a timely manner. 

Changes made to the organization plan required NRC approval, but most changes 

did not require a modification of the recovery technical specifications. (81)

• Recovery Procedures. The recovery technical specifications also imposed a 

requirement that the NRC would approve recovery mode implementation 

procedures. The specific procedures affected by this requirement were those that: 

(●) specifically related to core cooling; (●) could cause the magnitude of releases 

to exceed limits established by the NRC; (●) could increase the likelihood of 

failures in systems important to safety and radioactive waste processing or storage; 

or (●) could alter the distribution or processing of significant quantities of 

contaminated water stored or being released through known flowpaths. The 

implementation of the procedure approval requirement necessitated a high degree 

of involvement by the NRC’s onsite staff in planning operations. During the first

5 years of recovery operations, the NRC approved over 1,000 procedures,

including revisions. (82)

3.3 Conduct of Operations

Soon after the creation of the TMI-2 project office, the office staff issued 

NUREG-0698, “NRC Plan for Cleanup Operations at Three Mile Island Unit 2,”

in July 1980. This plan defined the functional role of the NRC in cleanup 

operations at TMI-2 to ensure that agency regulatory responsibilities and 

objectives would be fulfilled. The plan outlined NRC functions in TMI-2 cleanup

operations in the following areas: (●) functional relationship of the NRC to other 

government agencies, the public, and the licensee to coordinate activities;

(●) functional roles of these organizations in cleanup operations; (●) NRC review 

and decisionmaking procedures for the licensee’s proposed cleanup operation;

(●) NRC/licensee estimated schedule for major actions; and (●) the NRC’s 

functional role in overseeing implementation of approved licensee activities. The 

NRC revised the plan in 1982 and 1984. (83, 84)

• Adjustments to Meet Challenges. As the cleanup progressed, it became 

apparent that TMI-2 should be treated more like a waste management facility than 

a reactor facility. This required different skills and different mind sets. Sometimes 

traditional reactor safety perspectives were counterproductive. The NRC’s 

oversight responsibilities at TMI-2 required special delegated authorities and 

accountability requirements. The NRC required technical staff who possessed 

excellent communication skills. Decisions were generally more effective when 

made at the local site versus at NRC Headquarters or NRC Region I, although 

NRC Headquarters expertise was often used. The issues under consideration were 

very technical and site driven. (85)
29

Key: Advisory Committee for Reactor Safeguards (ACRS), President’s Council on 

Environmental Quality (CEQ), Nuclear Reactor Regulation (NRR), Technical 

Advisory and Assistance Group (TAAG)

Major NRC functional roles in TMI-2 cleanup operations in 1984. (481.4)
30

• Regulatory Goals. The NRC’s regulatory goals for TMI-2 cleanup operations 

were well understood and documented in NUREG-0698. The NRC kept the plan

current with subsequent revisions and included the following objectives: 

(●) ensure reactor safety and control of radioactivity (e)

; (●) ensure minimal 

environmental impacts; (●) maintain the radiation exposures of workers, the 

public, and the environment within regulatory limits and at ALARA levels; and 

(●) achieve interim safe storage and disposal of radioactive wastes from the 

cleanup operation. (86)

• Onsite Presence. The NRC had a daily onsite presence with prompt, open 

access to all information. Information exchanges between the licensee (and its

contractors) and the NRC included preliminary planning and scheduling. This 

information involved more than just being aware of daily plant status; it also 

included being part of the intellectual planning aspects. The licensee decided what 

activity was to be proposed and determined when the proposed activity was safe 

enough to move forward. Additionally, the NRC staff monitored licensee planning 

meetings to keep the agency’s managers informed of upcoming activities, 

including those with possible safety or regulatory issues. This information fostered 

more complete and faster safety evaluations by the licensee and the NRC. In all 

cases, for both organizations, the dominant criterion was to do what was best for 

public and worker safety. (87)

• NRC TMI Project Office (TMIPO). (88, 89, 90, 91) At TMI-2, the normally 

separate NRC functions of regulatory approval of proposed changes and the 

inspection of operations to ensure compliance with approved license requirements 

were combined in a single group. Most of NRC staff who conducted the regulatory 

reviews and inspections were physically located onsite in mobile trailers. Some 

NRC reviewers and inspectors of unique activities were occasionally loaned to the 

site from NRC Region I or NRC Headquarters. In this manner, the staff could 

accomplish the review and approval process of proposed activities in a matter of 

days instead of months, and it could inspect all the activities immediately. 

o Authority and Capabilities. The NRC Commissioners gave TMIPO 

unprecedented authority and capabilities to meet regulatory responsibilities.

These included ample staffing and sufficient funding for technical support to 

use outside experts (national laboratories, consultants) and access to NRC 

technical experts at NRC Headquarters and regional offices. This organization 

was also unique because it combined in one unit both the management and 

professional staff necessary to carry out safety and environmental reviews and

the direct inspection of nearly all aspects of the cleanup.

o TMIPO Locations. TMIPO had a strong onsite presence, with two offices at 

TMI and a staff of about 20 individuals. The offices included a public office in 

e Editor’s Note: “Radioactivity” was a legacy term used during the TMI-2 period to mean 

most things radioactive, such as radioactive materials, contaminated waste water, surface 

contamination, and area sources of radiation. 
31

Middletown and an onsite working office. The onsite office was staffed by 

Headquarters and regional inspectors and technical staff, clerical staff, foreign 

assignees, part-time cooperative students, and summer students. This office 

was responsible for the day-to-day review of all licensee activities that 

pertained to the cleanup and information flow to other NRC offices, interested 

government agencies, and the public. The town office provided opportunities 

for the public to stop by and speak to public affairs and technical staff about 

the status of the plant and cleanup, including voicing their concerns about 

public safety. (The office included a rug for use by the children of visiting 

mothers.) An additional TMIPO office at NRC Headquarters, located in 

Bethesda, MD, functioned as the liaison with the Commissioners, executive 

managers, and NRC Headquarters offices.

o Staffing and Support. TMIPO had an integrated onsite staff for analysis, 

licensing, operations, inspections, and public communications. TMIPO also 

had strong support at NRC Headquarters for Federal interfaces, legal expertise, 

and special issues (e.g., transportation, health physics, environmental). 

National laboratories and contractors contributed special expertise to the NRC 

site office. Within the first 3 years following the accident, the NRC assigned 

30 to 40 staff members to the TMI-2 cleanup, initially divided equally between 

the site and NRC Headquarters offices. This level of staffing allowed for a 

high level of NRC scrutiny of the unique cleanup activities, including the 

review and approval of the detailed procedures for implementing the cleanup. 

As the project progressed, the NRC gradually reduced its staff and the level of 

staff review of detailed cleanup activities.

o Public Outreach. Public communication was most visible through weekly 

plant status reports on the cleanup; notification reports of unusual occurrences;

participation in several public and civic group meetings every week; and 

interviews with the local news media. The reports provided to the public and 

media were the same ones distributed within NRC Headquarters.

• Roles, Responsibility, and Authorities. The NRC staff involved in the 

regulatory oversight at TMI-2 required clearly defined and documented roles, 

responsibilities, and authorities on the cleanup, including regulatory interactions 

with all involved organizations, while maintaining regulatory independence.

NUREG-0698 documented the NRC’s role in cleanup operations at TMI-2 and its

regulatory responsibilities in fulfilling this role. (92)

The purpose of this NRC plan was to (●) define the functional role of the NRC in 

cleanup operations to ensure that agency regulatory responsibilities and objectives 

were fulfilled and (●) provide a general schedule of major cleanup actions and the 

NRC’s role in meeting these milestones. 

The plan outlined NRC functions in the following areas: (●) relationship of the 

NRC to other government agencies, the public, and the licensee for coordinating

activities, (●) review and decisionmaking process for the licensee’s proposed 
32

cleanup activities, and (●) roles in overseeing implementation of approved 

licensee activities. The NRC issued two revisions of its plan. (93)

• Sharing Information. All organizations at TMI-2 proactively shared

information while maintaining their respective responsibilities and duties. Open 

discussions of mutual concerns and options were encouraged so that all 

organizations could understand the others’ views. Differing professional opinion 

discussions were welcomed and resolved promptly with an established 

management process. Constant self-assessment was necessary; operational 

feedback reflection was an important part of the never-ending improvement 

learning process. (94)

• Regulatory Independence. The NRC TMI project office (TMIPO) used DOE 

national laboratories extensively to conduct independent confirmatory safety and 

environmental reviews relative to the cleanup. For example, Argonne National 

Laboratory supported the development of the PEIS, and Pacific Northwest 

Laboratory measured the nuclear fuel remaining in the reactor coolant system and 

the reactor vessel after defueling. In such cases, the NRC was careful to avoid the 

appearance of a conflict of interest, ensuring that when a specific national 

laboratory did work for the NRC, it did not work on that same subject for the 

licensee. This practice helped maintain independence. (95, 96)

On only one occasion was the use of separate laboratories not practical. That 

occasion required the use of two experts, who were the only ones highly 

specialized in fuel measurement, and both came from the same national laboratory. 

The licensee contracted one expert to assess the remaining fuel in the reactor 

vessel following defueling. TMIPO contracted the other specialist to review this 

assessment. Luckily, both experts came from different organizations within the 

laboratory, and the laboratory established a special “firewall” to separate the two 

to prevent collaboration and organizational influences in their work. The NRC 

found this arrangement acceptable. (97)

• Independent Research. The NRC signed a coordination agreement with the 

DOE, the licensee, and EPRI in 1980 to jointly sponsor and participate in the 

DOE’s TMI Information and Examination Program. This program (known as 

“GEND”) was established to acquire data to improve the understanding of nuclear 

plant accidents and the phenomena that contributed to those accidents. Like other 

research agreements that the NRC had with industry organizations on 

non-TMI-related research collaboration (e.g., EPRI, owners’ groups), all parties 

agreed upon the data collection methods and procedures before work began, and 

participating organizations used the impartial results independently to meet their

individual needs. Since NRC TMI project office (TMIPO) observed daily 

meetings of the various cleanup groups, the NRC staff was responsible for 

reviewing the data acquisition tasks to ensure that they were implemented in 

coordination with the ongoing cleanup schedule. In addition, TMIPO ensured that 

the acquired data were used for the benefit of the cleanup to the maximum 

possible extent. (98)

 
33

3.4 Advisory Panel for the Decontamination of TMI-2

In October 1980, the NRC established a 12-member TMI-2 advisory panel to 

consult with and advise the Commissioners and staff on major activities related to 

the decontamination and cleanup of TMI-2. The panel consisted of State 

employees of the Commonwealth of Pennsylvania, local government, the scientific 

community, and residents near TMI. The TMI-2 project office acted as a liaison 

between the NRC and the panel and provided information to the panel on the 

status of the cleanup. Panel meetings were open to the general public, and 

transcriptions were produced for the public record. Panel members traveled to 

Washington, DC, at least once a year to meet with the Commissioners and report 

on panel activities. The panel held its last meeting, the 78th overall, in 

September 1993. (99)

• Lessons Learned from the Advisory Panel. (100) NUREG/CR-6252, “Lessons 

Learned from the Three Mile Island Unit 2 Advisory Panel,” issued August 1994, 

included observations and lessons based on interviews of former members and 

reviews of transcriptions of public meetings. All 36 individuals interviewed for 

NUREG/CR-6252 considered the advisory panel a success, particularly its 

effectiveness in accomplishing its mission. The following summarizes some key 

observations from that document on the achievements of the panel:

o The panel established a communication channel between the public and the 

NRC Commissioners and helped to defuse hostility among the stakeholders. 

o The panel provided a way for the NRC and the utility to report on the progress 

of the cleanup and to gauge the public’s reaction to various alternative actions.

o The panel kept the importance of the cleanup before the NRC Commissioners 

through periodic public meetings.

o The consensus was that the panel had an influence on cleanup activities,

although not in the form of technical advice or guidance. 

o The most crucial panel influence on cleanup activities was the increased public 

scrutiny of both NRC and licensee decisions and activities. The panel 

facilitated communication with the public for both the NRC and the licensee. 

This communication helped sensitize the NRC and the licensee to public 

concerns. 

o The panel encouraged the licensee and the NRC to fully consider alternatives 

and to carefully think through cleanup activities and how these activities would

be presented to the community. 

• Success of the Advisory Panel. (101) The panel held 78 meetings over 13 years. 

Panel members were dedicated to their appointments (and not paid, except for 
34

nominal travel reimbursements). The success of the panel could be attributed to 

the following: 

o The NRC had a genuine interest in opening two-way communications with the 

public. The licensee recognized the value of openness in dealing with the 

regulatory agencies, employees of the Commonwealth of Pennsylvania, and 

the public. The NRC consistently supported the panel.

o Although funded by the NRC, the panel enjoyed a great deal of independence. 

Periodic meetings with the NRC Commissioners lent legitimacy to the panel.

o The panel was led by two chairpersons who were outstanding in their ability to 

deal with complex technical issues in a highly charged public environment. 

These chairpersons were fair, effective, and concerned when leading the 

meetings. A good chairperson was an absolute necessity for success.

o The makeup of the panel was balanced and represented all stakeholders. Panel 

members ranged in expertise and perspectives. This makeup contributed to the 

perception that the panel was a credible and legitimate forum for discussion of 

the cleanup activities. Longevity of panel membership allowed the 

development of the technical understanding required to ask the right questions 

and understand the answers.

o The structure of meetings was effective. The agenda for each meeting was 

predictable, time constraints were followed, and time was scheduled for all 

parties to speak.

o Panel members were committed and willing to spend the time between 

meetings to prepare. Having the meetings in the early evening made the 

meetings more accessible to the public; it also had the benefit of allowing the 

panel members to dine together, which markedly improved their understanding 

of each other’s concerns.

o Transcribing the meetings was important in that people were careful in what 

they said. Additionally, the transcription provided a historical record of the 

cleanup as told by all the stakeholders.

o Frequent news media coverage of the panel meetings disseminated cleanup 

information to a wider audience than was reached through the panel meetings

themselves. Media coverage encouraged high-quality presentations about the

cleanup.

o Members from the licensee’s Safety Advisory Board frequently attended panel 

meetings. This also helped inform cleanup planning.
35

Key to Figure on Next Page:

Workers on the defueling work platform.(a) camera operator; (b) tool operator; 

(c) crane operator; (d) probably the camera controller acting as an assistant to the 

tool operator; (e) radiation control technician; (f) step stool for disconnecting tools 

from the jib crane when tied off in the slot; (g) hydraulic control skid used to power 

long-handled tools; (h) long-handled tool tied off in the slot; (i) stepoff pad to the 

defueling water cleanup equipment; (j) small long-handled tools tied off in the slot; 

(k) long-handled tools tied off along the fuel canal wall; (l) monitor console; 

(m) rack containing disposable booties and gloves; (n) possibly a container of 

discarded booties and gloves awaiting transfer; and (o) camera control console and 

radio repeater equipment. (481.5)
36
37

4 DEFUELING

The 10-year overall defueling effort at TMI-2 expended a collective manpower 

effort of over 3.6 million person-hours. The reactor vessel defueling operations 

alone spanned a 5-year period, from October 1985 through January 1990, and 

involved over 2 million person-hours. The 5-year effort removed a total of about 

133,000 kilograms of fuel, cladding, and structural and control materials from the 

reactor vessel. In July and August 1991, the reactor vessel was drained to make 

final measurements of the residual fuel remaining in it. The estimated residual fuel 

quantity that remained in the reactor vessel following defueling was approximately 

1 percent of the original 94,000 kilograms of uranium oxide fuel inventory. (102)

The total quantity of residual fuel (uranium dioxide) was estimated to be less than

1,125 kilograms, distributed in four major plant locations as follows: (●) auxiliary 

and fuel handling buildings (less than 17 kilograms); (●) reactor building, 

excluding the reactor coolant system (less than 75 kilograms); (●) reactor coolant 

system, such as steam generators, pressurizer and surge line, and other low points, 

excluding the reactor vessel (less than 113 kilograms); (●) reactor vessel (less than 

900 kilograms). (103)

The following defueling topics include safety; planning; predefueling activities; 

the defueling work platform; work inside the containment building; defueling 

tools; defueling canisters; and defueling operations.

4.1 Defueling Safety

The near-term cooling of the reactor core was stable within the first week after the 

accident. However, many technical issues involving safety and control were 

considered in the following weeks and months. Other unique safety concerns and 

solutions were identified during the defueling years. The list below presents a brief 

overview of these concerns related to defueling activities. More thorough 

descriptions appear in EPRI-NP-6931 and NUREG/KM-0001, Supplement 1.

• Defueling Safety Concerns. When preparations for defueling began, each 

proposed activity required NRC approval, in accordance with requirements in the 

recovery technical specifications. The licensee’s safety analysis report of a 

proposed activity and the NRC’s safety evaluation report considered a list of 

safety concerns, along with any mitigation measures related to the proposed 

activity. The typical list of safety concerns addressed in these reports included: 

(●) criticality (in-vessel, ex-vessel, and containment building sump); (●) decay 

heat removal; (●) hydrogen evolution; (●) pyrophoricity; (●) heavy load handling 

and load drops; (●) fire protection; (●) impacts on TMI-1 and TMI-2 plant 

operations; (●) submerged combustion; (●) instrument interference; (●) reactor 

pressure vessel integrity; (●) electric shock; (●) worker exposure; and (●) release 

of radioactivity, including offsite exposure. (104)

• Criticality Concerns. The licensee instituted the necessary controls to prevent 

the damaged reactor core from achieving a critical condition during defueling 
38

operations. The reactor vessel at TMI-2 remained intact, allowing the use of boric 

acid in the reactor coolant for criticality control. Analysts predicted how much fuel 

would be at each location (in-vessel and ex-vessel), and they conservatively 

estimated the amount of boron required to preclude inadvertent criticality at each 

location. Safety evaluations included every possible scenario that could result in 

an inadvertent boron dilution. Numerous criticality analyses included conservative 

assumptions to account for the unknowns about the condition of the core. Further, 

there was reasonable assurance that the necessary controls were in place to prevent

an inadvertent criticality. The NRC’s independent analyses supported its approvals 

of all cleanup activities that could impact subcriticality. (105, 106)

• Boric Acid for Criticality Control. Six elements (boron, cadmium, gadolinium, 

lithium, samarium, and europium) were studied for potential addition into the 

coolant system to maintain the neutron multiplication factor (keff) below 0.95. 

Boron (as boric acid) was found to have a variety of advantages, including 

minimum impact on water cleanup systems, lack of serious materials compatibility 

problems, and lower costs. In addition, boron could be added to the coolant system 

using existing chemical addition equipment. The dissolved boron minimum level 

of 4,350 parts per million was found to be adequate to maintain the TMI-2 core 

debris subcritical under all feasible configurations. Boron additions had to be 

made before lowering water in the vessel because, once the water level was 

lowered, gas pockets (at the tops of the steam generators) would prevent mixing of 

the boron throughout the reactor coolant system. (107, 108)

• Boration Dilution Concerns. (109) Licensee evaluations considered that 

unborated or underborated water could be unintentionally injected into the reactor 

vessel, resulting in boron dilution and possible inadvertent criticality. The 

underborated water could come from several sources, such as demineralized water 

used in the plant or “slugs” of water trapped in pipes since the accident. All 

proposed cleanup activities at TMI-2 were analyzed for boron dilution and 

mitigation.

o Analysis. The boron dilution analysis typically included the following steps: 

(1) Identify the potential points of water injection into the reactor coolant 

system (RCS) (e.g., core flood tanks, pressurizer, reactor coolant pump seals, 

steam generator secondary side, reactor vessel nozzles, top of the open reactor 

vessel). (2) Track each potential RCS injection point connected from potential 

dilution sources (e.g., tanks, coolers, demineralizers, evaporators, heaters, 

closed cooling water systems, spent fuel pool). (3) Identify isolation barriers 

for each dilution source (e.g., removed spool pieces, closed valves, heat 

exchanger or pumps with elevation or head differences). (4) Determine the 

probability of failure of the isolation barrier configuration due to hardware 

faults and human error. (5) Estimate the total plant boron dilution potential by 

considering the number of injection paths, the reliability of each isolation 

barrier, and the potential for operator error, or failure, in identifying and 

terminating a boron dilution event. 
39

o Mitigation. Credit for mitigation was heavily dependent on the detection 

capability. At TMI-2, the means of detecting a boron dilution event included

monitoring: (●) reactor coolant level; (●) levels of dilution sources 

(e.g., tanks); (●) mass balance of the RCS; (●) status of the positions of valves, 

pumps, and breakers by using equipment checklists; (●) neutron detection with

operable source range neutron detectors; and (●) RCS boron concentration by 

routine sampling.

• Pyrophoric Reaction Concerns. Finely divided metallic Zircaloy is 

pyrophoric. Even though evidence showed that most of the Zircaloy fuel rod 

cladding had been oxidized by the accident, the fact that it had fragmented caused 

concern over a pyrophoric reaction (and a consequent large metal fire) if core 

debris were exposed to air. Various studies were performed in which samples of 

fine core debris were subjected to ignition tests. The studies indicated no potential 

for pyrophoric reactions. When the water level was subsequently lowered to the 

top of the plenum assembly before head lift (thereby uncovering fine debris resting 

on the top of the plenum) and air was allowed into the system, no pyrophoric 

reactions occurred. Nor were any such reactions observed during the entire reactor 

defueling operations, including sawing and plasma arc cutting of core 

materials. (110)

4.2 Planning

The TMI-2 experience demonstrated that damaged fuel could be safely handled 

and stored in a practical and effective manner. (111) However, the uncertainty about 

the scope of the defueling tasks and the unfounded hope that damage was minimal 

were the major shortcomings of early defueling planning. (112)

• Data. One lesson to come from the cleanup was the importance of accurate 

data about conditions inside the containment building, reactor vessel, and reactor 

coolant system. The licensee recognized this need and created an engineering 

organization responsible for data acquisition and analysis. The licensee’s Safety 

Advisory Board believed that data acquisition and analysis played an important 

role in the final success of the cleanup. However, the Board concluded that the 

licensee did not always assign adequate time or resources to the acquisition of 

accurate data and that this hindered many cleanup plans and subsequent 

operations. One impact of a lack of adequate knowledge about core conditions was 

the difficulty in planning for adequate tooling. Because data were often 

insufficient or not timely, tooling requirements were poorly defined. Defueling 

safety was not compromised by tooling difficulties, but the project schedule was 

probably delayed to some extent. (113)

• Examinations. Defueling could not be completely engineered at the start. 

Instead, a novel approach was required: engineering to gather data first and to take 

initial steps; gathering more data; engineering to defuel; and then repeating the 

steps as new areas and information were encountered. (114) Data on core conditions 

were obtained by the following in-containment activities: (●) camera examination
40

Flow diagram showing the examinations performed on core debris grab samples. (481.6)
41

of the upper core region (“Quick Look”); (●) uncoupling of all control rod drive 

mechanism leadscrews; (●) removal of three control rod drive mechanism 

leadscrews and analysis at offsite laboratories; (●) a probe insertion into the core 

debris bed; (●) core topography examination and explorations using ultrasound 

equipment; (●) underhead characterization program (visual inspections, radiation 

measurements, pyrophoricity tests); (●) axial power shaping rod assembly 

insertion test; (●) core debris grab samples; (●) solid-state track recorder 

measurements of the lower head region; (●) lower head visual examination; 

(●) debris samples from the lower head; (●) operating experience from defueling; 

(●) core void video mapping; (●) in-core instrument probing; (●) in-core 

thermocouple and self-powered neutron detector data; (●) core bore samples and 

video inspections; and (●) core bore machine data (i.e., depth, rotation speed, 

torque) to interpret elevation and thickness information on penetrated material. (115)

• Planning Study. (116) Within a few months prior to the start of preliminary 

defueling, the licensee planning study, “Core Conditions Design Basis,” issued in 

March 1985 (117) and revised in July 1986 (118)

, identified physical conditions that 

had a reasonable probability of being encountered during fuel removal operations. 

This study provided the design bases for the development of techniques and tools 

to remove core material. The report described core conditions that were based on 

fuel condition data obtained from various inspections and defueling operations 

performed in the reactor vessel. 

o Study Inputs. The combination of investigations and methods included:

(●) initial defueling experience; (●) visual and ultrasonic examination of the 

reactor vessel internals; (●) physical and chemical examinations of materials 

removed from the reactor vessel; (●) examination of materials transported 

throughout the RCS and containment system components; (●) interpretation of 

the response of online instrumentation during the accident; (●) calculations of 

accident damage and fission-product behavior using severe-accident analysis 

codes; and (●) first-principle engineering calculations of specific phenomena.

o Actual Conditions. Actual core conditions experienced throughout the 

defueling operations included: (●) loose debris; (●) fused debris; (●) intact 

assemblies; (●) a partially embrittled zone; (●) guide tubes and instrument 

tubes; (●) fused adjacent fuel assemblies; (●) fuel assembly/core former 

interface; (●) a lower core monolith; (●) lower reactor vessel head debris; and 

(●) end fittings stuck in the lower grid.

• Research Topics. Planning required testing, evaluation, and resolution of 

safety concerns. The DOE’s Hanford Operations evaluated safety concerns in 

GEND-051, “Evaluation of Special Safety Issues Associated with Handling the 

Three Mile Island Unit 2 Core Debris,” issued June 1985 (119)

, such as 

pyrophoricity, radiolytically generated hydrogen and oxygen, and the potential for 

steam generation in core debris canisters from an accidental fire during a 

transportation accident. Various other organizations conducted criticality studies. 

Recommendations drawn from these results included the following: 
42

(●) Hydrogen-oxygen recombiners should be installed in each core debris canister. 

(●) Water could be removed from each canister by drip drying (no vacuum 

pumping was required). (●) The maximum weight of the loaded, dewatered 

canisters and the minimum volume of gas/vapor in each canister should be 

controlled and measured by weighing before and after dewatering. (●) A cover gas 

of approximately 2 atmospheres of argon should be added to each canister. 

(●) Each canister should be weighed and pressure checked before shipping. 

(●) The shipping cask (f)should be designed to limit the temperature of the canister 

contents after the standard hypothetical accident (fire), such that the design 

pressure of the canister or cask would not be exceeded. (●) Provisions should be 

made for canister venting during long-term storage and for cask venting in the 

event of an overpressure condition resulting from an “extended” fire. (●) Some 

pyrophoricity testing of samples should be conducted during defueling to assure 

adequate safety-related information during canister opening. (120)

• Evaluation of Uncertainties. (121, 122) The NRC PEIS for the cleanup of TMI-2 

and design of the defueling systems were developed before the extent of core 

damage and radiological conditions were fully understood. The PEIS was based on 

NSAC-1, “Analysis of Three Mile Island—Unit 2 Accident,” (123) issued in 

March 1980. The early defueling designs were based on INEL’s GEND-007, 

“Basis for Tool Development for Reactor Disassembly and Defueling,” (124) issued 

May 1981, which compiled the core damage estimates previously performed by 

five independent groups. Neither study had the benefit of visual inspections inside 

the reactor vessel nor grab samples from the “Quick Look” program. 

o Worst Case Estimate. To compensate for uncertain core conditions, the PEIS 

postulated best and worst case core conditions. GEND-007 went further to 

identify the minimum and maximum bounds of damage and established a 

“reference” description for the status of the damaged core. The different 

degrees of damage present in the reference core were considered during 

planning of contingency tooling and procedures for reactor disassembly, fuel 

removal, core inspections, and fuel sample acquisitions.

o Observed Damage. The true extent of damage was not understood until after 

video inspections and sonar mapping of the rubble bed in 1983, video 

inspections of the lower head region in 1985, the core sample drilling program 

in 1986, and video inspections behind the core former walls in 1987. During 

f Editor’s Note: While referred to in the documents as a shipping cask, the terminology in 

the regulations for both the NRC (Title 10 of the Code of Federal Regulations

(10 CFR) Part 71, “Packaging and Transportation of Radioactive Material”) and the 

U.S. Department of Transportation (49 CFR 173, “Shippers—General Requirements for 

Shipments and Packagings,” Subpart I, “Class 7 (Radioactive Materials”) is package, or 

transportation package (or packaging when referring to only the shipping cask and not to 

the container and its contents). See also the definitions in the IAEA’s transportation 

regulations, Specific Safety Requirements No. 6, “Regulations for the Safe Transport of 

Radioactive Material,” 2012 Edition.
43

early planning through 1982, GEND-007 was adequate to bound the core 

damage for tool development, at least to the extent necessary to start defueling.

• Unsuccessful Data Acquisitions. Some data collection activities at TMI-2 

were not conclusive. These activities included the following: (●) Insertion test of 

axial power shaping rods to the fully inserted position to ascertain core conditions; 

control rods were not connected to the control rod drive mechanism leadscrew. 

(●) Vertical gamma profiles of the reactor vessel’s lower head region via the 

in-core tubes; tubes were blocked. (●) Use of solid-state track recorder neutron 

dosimetry in the reactor cavity between the vessel and biological shield for 

nondestructive assessment of fuel distribution; results were questionable. (125, 126)

• Systems Integration. The integrated design of the fuel removal, transfer, 

storage, and cask loading system was vital to defueling success. Defueling 

handling and processing changed little throughout the cleanup. The notable 

exception was the use of new and modified fuel debris removal equipment in the 

reactor vessel (e.g., variations of handheld and pneumatic tools, core bore 

machine, plasma arc cutting torch). These changes were made as defueling 

progressed and unexpected conditions were encountered, such as a solidified lower 

Maximum damage configuration of the TMI-2 core as estimated in 

1981. The graph shows regional average fuel cladding 

oxidation. (481.7)
44

core region, solidified core melt in the lower core support structure, and significant 

debris behind core baffle plates. (127, 128)

• Mockups. (129) The typical progression of activities at TMI-2 involved 

procurement, design, testing, development, shipment, more testing and 

development, training, assembly, and finally operation. The latter stages of 

development and training used mockups extensively. Used before every major 

cleanup activity, the mockups proved effective in improving work efficiency and 

minimizing radiation exposure. Mockups ranged from partial workspace 

arrangements to a more expensive detailed re-creation of an entire work area and 

activity. Detailed mockups were developed for activities in radiation areas that 

involved limited stay times, high collective doses, critical paths, routine tasks by 

many workers, complex tasks, or use of special equipment. Structures, systems, 

and components were also staged in a mockup setting in the turbine building 

before being reassembled in the containment building. The advantages of the use 

of mockups included the following: 

o Effective Planning Tool. Planners used mockups to identify necessary tools, 

materials, and services; to provide walk-throughs by planners, developers, and 

workers; to identify real and potential problems; and to plan and train for the 

use of robots. Tools that were tested in noncontamination conditions and did 

not work as expected could be disposed of as nonradiological waste.

o Effective Development Tool. Developers used mockups for fabrication, 

operating, and contingency procedures; to test equipment and tools; to check 

intersystem interfaces; and to check for tooling clearances and interferences.

For example (130)

, the INEL staff based in Idaho performed the bulk of the 

development work for the core bore drilling machinery. The staff designed, 

procured, assembled, and tested the equipment in an old unoccupied 

containment building at INEL. The facility provided full-scale testing of the 

equipment and the writing and validation of (verbatim compliant) procedures. 

The final core bore operating procedure contained 300 pages.

o Effective ALARA Tool. Radiation protection specialists used mockups to 

evaluate dose reduction measures and to reduce the time required in a radiation 

area, thus reducing worker exposure. 

o Effective Operations Tool. The operations staff used mockups to facilitate 

maintenance; increase the efficiency of operations; assess impacts on work 

schedules; simulate upset conditions; and solve operational problems.

For example (131)

, after initial operations, trouble shooting, and repairs, the core 

bore system full-scale mockup at INEL was used for data acquisition. Because 

the TMI core was an unknown environment, a data base was developed that 

consisted of drilling through a variety of materials (end fittings, standing glass 

rods, vertical metallic tube arrays, ceramic plates, carbon steel plates, stainless 
45

steel plates, sand, and even voids) to characterize the core locations drilled. 

The resultant data base allowed researchers to categorize materials drilled from 

the core, by comparison of the known data with parameters coming from actual 

core drilling.

• Contingency Planning. Contingency plans were formulated to address 

credible events that could impede the timely completion of the plenum assembly 

lift and transfer operations. Many subsequent contingency plans were a direct 

result of lessons learned during reactor vessel head lift operations. (132)

• System Checkouts. A checkout program for the plenum lift equipment 

identified several needed modifications. Special equipment was operationally and 

functionally checked out as part of an integrated system before use in the 

containment building. The onsite checkout was performed under conditions that 

closely simulated the actual interfacing conditions the equipment would be 

Safety Advisory Board inspecting the mockup in the TMI-2 turbine 

building of the rotating defueling work platform used to train 

defueling operators and test new equipment.
46

subjected to during use in the containment building. (133)

• As-Built Dimensions. Inspections revealed that the as-built dimensions of the 

reactor vessel internals were not consistent with the designed dimensions. These 

lessons showed that inspections for as-built dimensions should be made at critical

interface locations between reactor intervals (and perhaps all nuclear steam supply 

system equipment) and specially designed recovery equipment well in advance of 

final manufacture to avoid reworks or unusable equipment. (134)

• Use of Video to Survey the Damaged Core. The use of video equipment for 

reactor vessel characterization was extremely effective for seeing the conditions in 

the vessel. This information was eventually used to develop methods for fuel 

removal. The use of video cameras had some problems, such as the effects of the 

intense gamma radiation fields in the reactor vessel on the video camera and the 

challenge of relating the information obtained from viewing a small area to a 

larger area. To overcome the challenge of dimensional interpretation, objects of 

known dimensions in the reactor vessel were used as benchmarks to scale up the 

video image to the larger area. The effects of water clarity and shadows on the 

black-and-white video images (high-resolution color video technology was limited 

at the time) also presented problems. The use of auxiliary lights greatly improved 

the quality of the video. A little extra effort in properly positioning these lights 

helped eliminate the distracting shadows and provided a clearer image. (135)

• Single Defueling Contractor. As early as 1981, the DOE urged the TMI-2 

project management to select a single contractor to design and supply defueling 

tools and to pursue the post-head-removal defueling operation. Advantages of a 

single contractor included the following: (●) Cradle-to-grave responsibility would 

help ensure a thorough and competent job. (●) Relevant defueling experience 

could be applied to reduce the time and resources needed to develop a satisfactory 

level of expertise. (●) Because no one right approach existed, the process of 

selecting a single contractor would focus attention on the earliest possible 

selection of a defueling approach, which would then have a champion in the 

contractor proposing it. (●) One contractor would better ensure that tradeoffs 

between one defueling subsystem and another (e.g., one to control contamination 

in defueling water) were coordinated. (●) A single contractor could ensure the

integration of equipment, procurement, installation, and operations. In the spring 

of 1983, the licensee decided to retain Westinghouse Electric Corporation as the 

defueling contractor because of its experience in the areas of tool design, robotics, 

and nuclear fuel handling. The licensee retained control over the tool development 

program and produced performance specifications. (136)

• Remote Defueling. (137) One of the key questions facing planners was the use 

of remote equipment for defueling. An alternative defueling method that was 

considered was to convert the entire core into vacuumable rubble and transfer it 

directly into canisters in the fuel handling building, instead of transferring the fuel 

debris into shipping containers in the vessel or the fuel transfer canal. This 

approach proposed a remotely operated service arm (ROSA), a large shredder, and 
47

a debris vacuum/transfer system. ROSA was a computer-controlled and 

programmable electro-servo-powered arm. 

o Advantages. The advantages of a remotely operated service arm included the 

following: (●) Avoided the time-consuming process of placing fuel into 

shipping canisters within the confines of the reactor vessel and then 

transferring the canisters into the fuel handling building. (●) Minimized

in-containment work associated with preparing for or conducting defueling

activities. (●) Reduced the number of in-containment work hours and could 

reduce the need for decontamination. (●) Reduced water clarity problems

because ROSA could be programmed to work blind, thus reducing the volume 

of water to be processed. (●) Eliminated problems associated with tools long 

enough to be operated from an elevation above the reactor vessel. (●) Avoided 

material handling problems associated with moving shipping canisters inside 

the containment building.

o Challenges. One of the most technically challenging elements of the approach 

was the development of the shredder. Technical concerns with the 

shredder/slurry system centered on the development and licensing of an

unproven technology. In addition, the uncertainties surrounding the release rate 

of radionuclides during the shredding process were a concern. The pumping of 

debris outside of the containment building without packaging it first was a 

further concern. Developing this technology was determined to require an 

estimated 3-year engineering development program. Other concerns expressed 

during the review of the proposed system (138) included delays caused by 

maintenance issues and the uncertainty as to whether the system would work. 

The remote reconnaissance vehicle called Rover operated in the containment building 

basement throughout the cleanup campaign.
48

o Manual Defueling. Manual defueling would be more flexible. The ability of 

the program to reduce doses would be effective; therefore, the need to use 

robotic equipment would not be necessary

o Decision. Rather than attempting to minimize the theoretical overall time 

necessary to defuel (which was the promise of this alternate approach), the 

licensee wanted a method to start defueling as soon as possible. Otherwise, 

the success or failure of a complex system using unproven technology would

not be known for 2 to 3 years.

• Use of a Robotic Arm. An automated cutting equipment system was originally 

conceived to work with a robotic service arm, called MANFRED, which would 

have deployed tools and handled pieces of the lower core support assembly. This 

equipment was on site; however, it was never used because of concerns about the 

The MANFRED robotic service arm was on site; however, it was 

never used because of concerns about the complexity of operation.
49

complexity of operation and the vulnerability of cables and hoses to accidental 

severance. MANFRED would also had been extremely difficult to decontaminate 

when it had to be removed from the vessel. (139)

• Integration of Core Examinations with Defueling Plans. The DOE became 

concerned about the effects of one of the potential defueling methods under 

consideration at the time on its research and development objectives focused on 

the effects of the fuel melting. This method was completely robotic and involved 

conversion of the entire core into vacuumable rubble that would be transferred 

directly into canisters in the fuel handling building. If the core was ground up and

sluiced into canisters, all the spatial information on fission products and control 

material would be lost. As a result, the DOE developed a core sampling program 

that used a core boring machine to retrieve 8-foot-long bore samples from the 

damaged reactor core. (140)

• Use of Water for Shielding. (141) Reprocessed accident-generated water was 

used to fill spent fuel pool “A” to shield the highly radioactive submerged 

demineralizer system vessels, pool “B” to shield the defueling canisters, and the 

deep end of the fuel transfer canal inside the containment building to shield 

defueling canisters being staged for transfer to the fuel handling building.

The volume of water was minimized to reduce decontamination. Lessening the 

volume of water was crucial to accelerating the start of defueling by minimizing 

equipment development and logistics. The fuel transfer canal was only flooded in 

the deep end where the fuel transfer mechanisms were located. A dam was 

installed in the deep end of the fuel transfer canal to provide the water to shield the 

plenum and the transfer of fuel canisters, which were to be loaded inside the 

vessel. This arrangement allowed the fission products in the reactor coolant to be

retained in the minimum volume of water. Further, to simplify the water cleanup, 

the water in the fuel transfer canal and in the reactor vessel was separated to

minimize cross contamination.

4.3 Quick Look

The first video inspection of the upper reactor core region took place on 

July 21, 1982. This activity was the first major planned activity inside the 

containment building. A camera 1.5 inches in diameter and 12 inches long was 

inserted through an empty leadscrew support tube (inside a control rod drive 

mechanism), and then into a central control rod guide tube (inside the upper 

plenum). As the camera was lowered into the upper core region, it revealed a bed 

of rubble approximately 5 feet below the normal location of the top of the fuel 

assemblies.

• Quick Look—The Pathfinding Examination. Quick Look provided solid 

evidence that the core had been severely damaged. Until the video inspection of 

the core cavity in 1982, the licensee hoped that the TMI-2 plant could be 

refurbished and restarted as early as 1985. By one account, it was Quick Look that 
50

galvanized the attention of all organizations and provided the common focus that 

was needed to move forward with core debris removal. The licensee could not

hope for a reactor restart. (142)

Another account called the Quick Look inspection a “pathfinding” examination. 

Such an examination could not answer all the basic questions; however, once 

information and experience from the examination became available, it was

possible to develop more sophisticated examination equipment with a more 

realistic set of functional and operational requirements. (143) Follow-on 

examinations included the reactor vessel underhead video and radiological survey

in the summer of 1983; reactor core topography program in August 1983; retrieval

of “grab” samples from the debris bed in October 1983; and reactor core video 

mapping in April 1984. These and other examinations paved the way for the 

removal of the reactor vessel head on July 25, 1984, and the removal of the upper 

plenum assembly on May 15, 1985. (144)

The experience gained from this inspection supported the consideration of a 

manual defueling concept for removing the core and its debris. Further, the 

observation that the reactor internals were essentially intact, and that opening and 

working in and around the reactor vessel head area was reasonable, provided the 

basic insight that supported a manual “dry defueling” concept without having to 

flood the refueling canal. (145) The manual dry defueling concept was decided 

2 years later. (146)

• Quick Look Safety Evaluations. Quick Look was the first complex activity 

inside the reactor vessel that required formal safety review by the licensee and 

approval from the NRC. The licensee’s 169-page safety evaluation report included

a comprehensive criticality analysis for the proposed activities that could cause 

fuel rearrangements. The safety evaluations associated with the effects of Quick 

Look on safety functions included: (●) radiological environmental releases due to 

the release of krypton-85 from the reactor coolant system (RCS); (●) reactivity 

changes as a result of postulated fuel disturbances; (●) effects of the drain down of 

the RCS on decay heat removal capabilities; (●) boron dilution events from 

potential dilution paths; (●) releases of trapped hydrogen and krypton gases into 

the containment building atmosphere during the venting of the RCS; (●) potential 

for a combustible gas burn inside the containment building; (●) occupational 

exposures associated with decontamination or contamination control, rigging, RCS 

venting, installation of temporary power and lighting, removal of a control rod 

drive mechanism leadscrew, and inspection inside the reactor vessel; (●) RCS 

chemistry changes due to exposure of the RCS to the containment building

atmosphere; and (●) fire protection. (147)

The onsite NRC TMI-2 project office reviewed the licensee’s proposal, safety 

evaluation report, and associated operating procedures. The NRC concluded that 

Quick Look could be conducted in a safe manner with the measures and 

precautions provided for each of the major activities (i.e., venting,
51

Observation summary from central core bore location (K9). Video data from the 

interior of the core bore holes revealed a region of previously molten material directly 

below the hard crust and a region of intact standing fuel rods extending from the 

bottom of the previously molten region to the bottom of the core. (481.8)
52

depressurization, water-level lowering, lead screw removal, and camera insertion) 

associated with Quick Look. In addition, the NRC determined that conducting this 

program would pose little risk to the offsite public and the occupational workforce. 

Further, all the Quick Look activities were within the scope of those analyzed in 

the TMI-2 PEIS. (148)

• Other Quick Look Experiences. The Quick Look inspection also provided the 

following experiences and insights that helped with planning future cleanup 

activities: (149)

o Preparation of In-Vessel Activities. Experiences from the preparation work 

included the following: (●) Depressurization of the RCS and lowering the 

coolant level to prepare for the inspection were completed as planned. (●) The 

control rod drive mechanism drive shaft assembly was removed from the 

control rod without difficulty. (●) Cost in resources and time was high; 

27 containment building entries with 243 person-hours spent in the 

containment building to prepare for and execute the operation. 

(●) Challenging technical work could be conducted in the reactor vessel.

o Radioactivity Control. Observations from the radioactive contamination and 

exposure encountered during the preparation and conduct of the inspection 

The first closed circuit video inspection of the upper reactor core region took place on 

July 21, 1982. A technician sitting on a platform positioned on top of the reactor vessel 

service structure is inserting a camera, 1.5 inches in diameter and 12 inches long, 

through an empty leadscrew guide tube.
53

included the following: (●) Airborne activity and surface contamination were

not a problem. (●) Inspection equipment could be installed in the reactor core 

region and be satisfactorily decontaminated by simply wiping the equipment 

off, very much as was done in servicing operating nuclear plants. (●) Alpha 

contamination was not a problem. (●) A unique form of cesium contamination 

was discovered on the surfaces of the control rod drive mechanism leadscrew 

(examined at INEL) that was exposed to the accident environment inside the 

reactor vessel; this unique cesium surface contamination was very adherent 

and could not be removed easily. (●) Work could proceed in some areas of the 

containment building without excessive personnel exposure. (●) Work could 

be performed within the reactor vessel from a work platform under acceptable 

radiation levels with the potential to further reduce the radiation levels. 

(●) Actual exposure to complete the first Quick Look was 23 person-rem; the 

licensee’s initial estimate of 1,600 person-rem had cancelled or delayed the 

inspection, but experienced members of the Technical Assistance Advisory 

Group (an external review group) estimated 45 person-rem (NRC estimate 

was 60 person-rem).

o In-Vessel Visual Observations. The following visual observations contributed 

to the consideration of other defueling concepts: (●) The observable extent of 

damage to the upper core was enveloped within one of the worst case 

predictions of core damage as described in GEND-007 (150)

, which showed 

that computer models predicting reactor core damage were at least partially 

correct. (●) Upper reactor vessel internals were essentially intact (more 

damage had been expected) and should be structurally sound for removal 

purposes. (●) Core debris was not found on the top cover of the upper plenum 

assembly. (●) Core debris was not plastered up against the underside of the 

reactor vessel head.

4.4 Crane Operations

Several cranes were located inside the containment building: 

(●) 456,000-kilogram polar crane that was used to remove the reactor vessel head 

and plenum; (●) 23,000-kilogram auxiliary hoist that provided defueling support; 

(●) 4,500-kilogram service crane that was installed to support handling of the 

long-handled tools and other equipment; (●) wall-mounted jib crane above the 

reactor vessel head storage stand that was used to place the sand columns around 

the head storage stand; and (●) two jib cranes that were mounted on the rotating 

defueling work platform to aid the operators in manipulating the long-handled 

tools through a slot in the platform. Initially, the auxiliary hoist was not 

refurbished to avoid detracting from the critical path; however, the decision was 

later changed when defueling experience indicated the need to securely grapple 

core debris and to apply lift forces greater than those achievable with the service 

crane. (151)

• Crane Height. Experiences from crane operations inside the containment 

building during TMI-2 defueling included the following: (●) Cranes and hoists 
54

should have had sufficient lift to handle the longest tool with allowance for the 

rigging, a load scale, and over-travel for the upper limit switch. (●) The service 

crane height should be high enough to handle tools that could reach to the bottom 

of the reactor vessel or the lowest possible point. (●) The jib cranes mounted on 

the rotating work platform should have had telescoping posts to provide more 

usable hook height. (152)

• Crane Load Tests. The use of in-containment components for load testing 

cranes eliminated the need to bring other weights into the contaminated 

environment. The load test of the containment building polar crane consisted of 

removing the four 39,000-kilogram missile shields over the reactor vessel 

refueling canal and the 30,000-kilogram pressurizer missile shield. A stand was 

constructed to hold the missile shields after removal. (153)

• Crane Load Cells. Based on experience, one account suggested that cranes 

should include load cells to detect tools that were hung up on debris or work

platform equipment. (154)

• Crane Preventive Maintenance. Two failures occurred during the head lift;

failure of the polar crane pendant switch followed by failure of the relay in the 

polar crane hoist circuit. (155) As a result of the lessons learned from the head lift, 

the polar crane’s yearly preventive maintenance was scheduled to be completed 

just before the beginning of the plenum lift evolution. (156)

4.5 Defueling Work Platform

The defueling work platform (also known as the shielded work platform), with a 

Containment building polar crane load test using 40-ton missile shield blocks that are 

normally positioned over the reactor vessel service structure.
55

17-foot diameter rotatable surface and a 6-inch-thick steel plate shield, was 

installed 9 feet above the reactor vessel flange. By placing the work platform in 

the fuel transfer canal that had no water, the canal walls provided a shield from 

significant radiation in the lower levels of the reactor building. The platform 

provided a shielded work area for defueling operations; a support for manual, 

hydraulic, and mechanical defueling tools; and a method for removing defueling 

canisters. An adjustable slot and hand rail spanning the diameter provided access 

to the reactor core. The entire reactor vessel defueling effort was performed from 

this platform.

• Rotating vs. Stationary Platform. A stationary work platform would had been 

simpler to build and install and would not have required any components with long

lead times. However, a rotatable platform offered the advantage of a more precise 

alignment of tools and equipment over the desired section of the core. Loads could 

also be picked up in one radial core location and moved to another by rotating the 

platform. (157)

• Worker Safety. Two safety issues were revealed from minor incidents that 

could had resulted in significant injuries to workers on the work platform:

o In one case, the hoist on a jib crane came off the end of the jib arm I-beam on 

which it was being moved and slightly injured an operator. The jib crane

hook, which was mounted on the defueling work platform, was moved along 

an I-beam by a hand-operated chain-driven gear. A mechanical stop was 

mounted at the end of the I-beam to prevent the jib crane from running off the 

end. The stop was mounted by a bolt in a set screw arrangement. This set 

screw came loose with excessive use. (158)

Diagram of the rotating defueling work platform. (481.9)
56

o In another case, a worker fell into the reactor vessel through an open access 

port on the platform. The port provided access to one of the five defueling 

canister sleeves on the canister carousel (canister positioning system) in the 

reactor vessel. The worker backed into the open port and fell into the reactor 

vessel while catching the platform by the upper arms and elbows. The 

worker’s legs were submerged in the reactor coolant. Preliminary dose 

estimates indicated that the worker was expected to receive about a 

200-millirad skin dose to the legs. The worker received only minor bruises 

and did not require any medical treatment. (159)

• Tool Wash. A clean water wash system was installed on the work platform to 

provide clean borated water at a slow flow rate to wash off tools being lifted out of 

the coolant in the reactor vessel. (g) This helped to control contamination and hot 

particles. (160)

• Separation of Ground Fault Protection. When more than one light was served 

by a ground fault interrupter (GFI), any lamp failure due to water in-leakage 

would trip the GFI, resulting in the loss of all lights. The effort to identify the 

defective light was substantial. Each permanently installed underwater light that 

was required to operate the equipment should have its own GFI. (161)

• Underwater Electrical Components. Experience revealed that critical 

electrical components, such as limit or interlock switches, under water should be 

avoided. (162)

• Simplify Control Systems. Complexity in control systems should be 

minimized, where possible. This would include minimizing the number of 

interlocks, protective circuits, and monitoring devices to those essential for

ensuring safe operation. (163)

• Hydraulic Fluid. The low-viscosity hydraulic fluid contributed to reduced 

reliability of the aluminum sandwich valves used for the low-flow circuits. Valve 

malfunctions would lock pressure in the hydraulic cylinders and cause the quick 

disconnect fittings to leak. (164)

g Editor’s Note: Current practice includes wetting or washing tools being lowered into 

(potentially) contaminated water to prevent the adherence of hot particles, radioactive 

particles, or contamination in pores and cracks in the tools. This practice facilitates ease of 

decontamination of tools along with the washing as the tools are removed from the 

(potentially) contaminated water. This is a common practice for moving spent fuel into 

storage casks and transportation packages at U.S. plants. This could apply to the tools and 

any other items (e.g., canisters, packages, casks) inserted into the reactor pressure vessel, 

spent fuel pool, or other containers of contaminated water.
57

4.6 Work Inside Containment

Work inside the containment building at TMI-2 was more challenging than 

activities during common outages at nuclear power plants. The basement floor and 

walls were covered with highly radioactive contamination. These contaminants 

produced increased background radiation levels throughout the containment 

building. Structures, systems, and components inside the containment building 

were contaminated during the accident with high levels of beta radiation. Various 

activities inside the containment building also generated airborne radioactivity. 

Heat stress was a problem during the early cleanup campaign before an air chiller 

was installed inside the containment building. All of these personnel safety issues 

presented unique challenges for work planning and implementation. Even the 

simple tasks common to routine outage work presented challenges at TMI-2. The

licensee and others involved in the cleanup described the following experiences

and insights:

• Equipment Supplies. (165) The need for an adequate supply of reliable 

equipment, especially equipment that was essential to the critical path, was a 

primary lesson derived from the reactor vessel head lift operation. Some 

experiences from the reactor vessel head removal operations reported by the 

licensee (see GEND-044, “TMI-2 Reactor Vessel Head Removal,” issued in 

September 1985) included the following:

o Inventory. When making unscheduled containment building entries, 

improvements were needed to make sure that items were logged in and out so 

that the next crew was aware of what was already in the containment building

and what needed to be taken into it. Based on this experience, an inventory 

accounting system and an adequate backup supply of equipment and tools 

should be maintained, based on conservative estimates of potential needs. 

Personnel protective equipment, such as vests, oversized hoods, and 

respirators, should be substantially stocked, and planning should provide for 

sufficient personnel to process the respirators at peak periods.

o Testing and Evaluation. Measures should be taken to ensure that off-the-shelf 

equipment would perform satisfactorily. Procurement papers should be verified 

to include proper “Important to Safety” designations with quality 

assurance/quality control involvement. Testing should be performed, as 

applicable, before acceptance.

o Repairs. Potential repair and maintenance tasks should be thoroughly 

evaluated before an activity to ensure that they could be conducted with a

minimum impact upon the schedule if they needed to be done during the 

activity. All equipment should be supplied with scheduled (periodic) 

maintenance requirements and instructions.

o Backup and Spares. The camera, which provided the polar crane operator with 

the information on the preplaced targets for crane location, failed before the 
58

start of the reactor vessel head lift. There was no installed backup for this 

camera, although a spare camera and spare parts were available on site. 

Contingency plans, procedures, backups, and spares should be readily 

available for all sorts of schedule-important support equipment. A repair team 

with spare parts should be readily available during critical operations, while 

assuring proper radiological controls. Preparations could reduce delays from 

failures to last only minutes instead of hours.

• Work Documentation. (166) The documentation required for major operations 

was extensive, requiring multiple levels of review and approval. Planning was 

required to ensure that the operation followed procedures and that any changes 

could be expedited by available personnel. Some additional experiences from the 

reactor vessel head removal operations included the following:

o Approvals. This system facilitated rapid review and approval of necessary 

changes from planned operations and to provide technical assistance. During 

the TMI-2 reactor vessel head lift, a task force of planners and staff members 

was available in the coordination center to expedite changes to procedures or 

work instructions. The individuals had signature authority for reviews and 

approvals. Because they were aware of actual operations, they were able to 

support alternative courses of action quickly. Based on experience, one 

account suggested that personnel should be on 12-hour shifts during 

operations to facilitate communication and maintain continuity.

o Procedures. The procedures and documents for operations should be 

evaluated to ensure there are no duplicate steps. One action should not be 

controlled by more than one document, as this introduces the likelihood of 

overlooking details and increasing both the potential for conflicts and the 

effort required to make changes. This was a problem with the sequence 

document and the head lift procedure for TMI-2. Similar steps were in both 

documents. 

• Worker Effectiveness. Measures should be taken to keep morale high and to 

encourage teamwork. Some experiences from the reactor vessel head removal 

operations included the following: (167)

o Worker Fatigue. Workers should be rested, cool, and calm before a 

containment building entry. Supervisors should be sensitive to the stress 

experienced by making entries. 

o Shift Turnovers. A single shift turnover meeting involving all participants 

should be held, and all personnel should work the same shift schedule to 

maintain a smooth flow of work.

o Coordination Center. Too many people were in the Coordination Center 

during head lift; however, the question “Who is excess?” was the real issue. 

One possible arrangement to prevent or solve such an issue could involve the 
59

issuance of a limited number of passes or tickets per department. When that 

number of passes was in use, no other personnel could enter until someone 

from that department left. There should be no exceptions and no access lists 

beyond those authorized to hold passes. Individuals responsible for operations, 

by procedure, should make the determination.

• Training. Workers should receive as much preliminary training as possible to 

familiarize them with the working conditions and required operations. Training on 

accurate mockups using the actual procedures and job walkdowns inside the 

containment building represented the most significant contribution to successful 

operations. During the walkdown, personnel were able to identify locations of 

potentially useful equipment for contingencies. They were also able to identify the

location of equipment that could cause interference and thereby result in more 

time and cost due to the more severe radiological conditions. (168, 169)

• Electrical Maintenance. Electrical equipment installed in the containment 

building or radiation areas should have easily accessible diagnostic points and 

simple component replacement. Decontamination activities should be conducted to

avoid contaminating electrical equipment, which could complicate the 

maintenance of such equipment. (170)

Floor plan of the containment building air control envelope located outside the 

equipment hatch. (481.10)
60

• Moving Equipment into Containment. Components that were too large to fit 

through a personnel airlock were designed in subassemblies. These subassemblies 

were assembled and tested, sometimes using a mockup, before being disassembled 

and transported into the containment building. Having assembled and tested the 

components outside the containment building, the reassembly required less time, 

minimized radiation exposure, and increased reliability and functionality. (171)

• Contaminent Building Air Control Envelope. This was a controlled area 

outside the equipment hatch or emergency airlock where cleanup equipment and 

materials were assembled and staged before transfer into the containment building. 

This structure reduced worker stay-times in radiation areas, resulting in 

occupational exposure savings. The envelope also functioned as a staging area for 

contaminated material removed from the containment building, but it was not 

designed to be a storage area for radioactive wastes. (172)

• Equipment Hatch Removal. The licensee asked the NRC for approval to

remove the equipment hatch to the containment building on a contingency basis. 

The licensee considered that removal and reinstallation of the hatch could facilitate 

the movement of several pieces of large defueling equipment without the need for 

disassembly and reassembly inside the containment building, thus reducing time 

and radiation exposure. The NRC reviewed the licensee’s safety analysis report; 

however, the licensee withdrew its request because improved access would not 

offset the resources, time, and worker exposure involved in removing, handling, 

decontaminating, and storing the hatch. (173, 174)

• Water vs. Sand Shielding. Shielding around the reactor vessel head storage 

stand was used to attenuate radiation from the underside of the head. A shield wall 

consisted of fiberglass cylinders, each 0.6 meter in diameter and stacked 

3.6-meters high. Each cylinder had a concave interlocking pattern for maximum 

shielding effect. Initially, these cylinders were filled with water; however, leaks 

occurred, and the cylinders were subsequently filled with sand. (175)

4.7 Defueling Tools

Unique systems and equipment were designed and used to remove damaged fuel 

and structural debris from the reactor vessel. In the early defueling phase, tools 

were designed for “pick-and-place,” in which debris was picked up manually and 

placed into fuel containers (baskets) or specially engineered defueling canisters for 

shipping. Some long-handled tools had various hydraulically actuated fittings to 

tackle the larger pieces and smaller bits of debris. The core bore machine, 

previously used to obtain core samples, was used to bore holes in the resolidified 

mass of the previously molten material in the reactor vessel. This allowed the hard 

mass to be broken up. Combinations of tools to assist defueling the lower reactor 

vessel region included the core bore machine and plasma arc torch. (176)

• Designing Tools On Site. Over 100 long-handled tools were used to remove 

most of the damaged reactor core at TMI-2. Tool design requirements changed as 
61

new challenges arose. The first set of tools was manufactured off site. Many tools 

needed modifying as core conditions became better understood. After the start of 

defueling, an onsite machine shop was established for quick response to the need 

for tools. Most long-handled tools were fabricated on site, increasing efficiency 

and applicability (because of the interaction of the machine shop staff with 

operators and engineers). Simple tools could be built within 2 days. Based on this 

experience, the tool design and fabrication processes were able to modify or 

redesign tools quickly. (177, 178, 179, 180)

• Tool Inventory and Control. Experience revealed that tool inventory and 

control were essential to efficient operations. Spare parts should be provided for 

all tools that risked breakdown during normal defueling activities and that could 

cause a significant delay in defueling activities. Tools requiring maintenance 

should be segregated to prevent inadvertent use. (181, 182)

• Idle Tools. The time required to move tools between the reactor and the tool 

storage racks was significant. Idle tools that were not removed from the reactor 

tended to interfere with ongoing operations. (183)

• Peters Tool. A popular tool named after its vendor consisted of a very light 

aluminum rod and a vice-grip-like gripper on the end. The tool was operated by a 

mechanical lever at the top end. The tool was light and could be handled by hand 

without the use of a crane. These tools of various lengths were tied in the platform 

work slot for quick use. (184) 

• Hydraulic Clam-Shell Tool. The first defueling tool used to scoop up fuel 

debris on top of the core debris bed was a large hydraulic clam shell. This tool was 

ineffective because the debris bed was not very thick and contained partial fuel 

rods scattered across it. Additionally, the tool stirred up the fuel fines and 

eliminated all visibility. (185)

• Heavy-Duty Tools. The center of gravity of the tool should be determined for 

use in rigging and positioning the tool for proper operation. Experience revealed 

that the heavy-duty tools were harder to handle. (186)

• Hydraulic Fluids. Various fluids were tested for use in the defueling hydraulic 

system. After initiating defueling, a problem was encountered with the 

Peters vice-grip tool. (481.11)
62

Core bore machine, initially used to retrieve bore 

samples, was used to break apart once molten core 

and lower core support structure.

precipitation of boric acid crystals. To prevent further deterioration of the 

hydraulic fluid in the system and potential adverse effects on defueling tools, the 

hydraulic system was drained and flushed. Then the fluid was replaced with a 

tested substitution. When selecting a hydraulic fluid, several factors were 

considered: (●) the boron remaining mixed with the hydraulic fluid and 

(●) compatibility with RCS chemistry and processing, the canister recombiner 

catalyst, and the hydraulic system. (187, 188)

Another problem involving hydraulic fluid was encountered shortly after initiating 

defueling. The original hydraulic fluids were carbon based and contributed to 

microbial growth in the reactor vessel when hydraulic fluid leaked into the reactor 

coolant containing dormant microorganisms. The carbon-based borated hydraulic 

fluid used in the tools served as an unintended nutrient. To limit the growth of 

microorganisms, hydrogen peroxide was periodically added to the reactor coolant,

and a water-based hydraulic fluid with a 5-percent organic addition was used in 

the long-handled defueling tools. (189) (The subsection on Defueling Operations, 

below, provides further details.)

• Tool Connecting Joints. 

Experience revealed that the 

use of flanges to join handling 

poles should be avoided to 

prevent hangups on hidden 

obstacles when raising or 

lowering the tool into position.

(190)

• Useful Heavy-Duty Tools 

at TMI-2. Two important 

defueling machines used at 

TMI-2 were an off-the-shelf 

oil-rig-type vertical drilling 

machine (called core bore) and 

the underwater plasma arc 

cutting device. These machines 

provided the primary means of 

delivering cutting energy to the 

resolidified fuel mass and 

reactor internals to break the 

material up into sizes small 

enough for manual and 

hydraulic (vacuum, water lift) 

removal. (191)
63

• Core Bore Design Tests. Numerous bit styles and designs were tested using 

core-region mockups containing both actual materials and suitable substitutes in 

place of radiologically or economically prohibitive materials. The drilling mockup, 

located at INEL, included composite test samples consisting of Zircaloy-clad 

quartz rods, Inconel spacer grids, 300-series stainless steel end fittings, concrete 

blocks, ungraded 3/8-inch gravel, and hard-fired alumina plates. Information 

obtained from these tests was evaluated to select the drill bit cutter most 

appropriate for the work at TMI-2. Subsequent proof testing showed the bit to be 

successful in cutting all the anticipated material configurations. (192, 193, 194)

• Plasma Arc Cutting. The underwater plasma arc cutting device was chosen to 

cut apart and remove the massive stainless steel lower core support assembly 

(LCSA) to access the debris on the reactor bottom. The arc torch was used 

previously to cut upper fuel assembly end fittings. (195) A broad range of 

techniques were evaluated to cut the LCSA. Some techniques were ruled out, but 

others required experiments under simulated TMI-2 conditions. Plasma arc cutting 

was chosen over spark erosion machining; thermic rod; cutting water jet; arc saw; 

mechanical shear; explosive cutting; oxygen burning; sawing; drilling/milling; 

ultrasonic disintegration; and laser cutting. An automated cutting equipment 

system was designed and installed in the reactor vessel to perform the task. The 

LCSA was cut into about 50 pieces; disassembly required about 1 year. Though 

actual cutting was quick, productivity was hampered by equipment failures, torch 

failures, and control system redesign. (196, 197)

• Plasma Arc Difficulties. Torch burnouts and other problems (e.g., electronic 

failures, a seal failure on a servo-motor housing, a transformer failure, feedback 

circuitry difficulties, and position limit switch failures) initially affected system 

performance. Improvements and experience with the system reduced maintenance 

dependency. (198) Difficulty with the plasma arc cutting equipment slowed the 

cutting process for several reasons.

o Torch Burnouts. Frequent torch replacements were required due to burnout 

(average torch life was about 10 cuts), arc starting problems, and mechanical 

damage in low visibility and restricted access areas. A common failure was 

loss of electrical conductivity caused by heavily borated vessel water leaking 

into the torch head. When failures occurred, the entire torch assembly had to 

be raised from the vessel and reconstructed on the defueling work 

platform. (199)

o Cutting Problem. Fuel debris adhering to the lower core support assembly 

(LCSA) plates made cutting difficult. (200, 201)

o Turbulence. Plasma cutting and cover gases acted as an airlift in the LCSA 

forging, which caused loose debris to move from the forging flow holes and to 

be deposited in the X-Y bridge drive system of the automated cutting 

equipment system. This required the cutting operation to be shut down; the 

bridge to be removed from the vessel; the bridge’s various motors and 
64

positioning components to be disassembled, cleaned, lubricated, and 

reassembled; and the bridge to be reinstalled in the vessel. (202) Material 

suspended as a result of the airlift effect also interfered directly with the 

plasma arc cutting process, requiring recuts to be made. (203)

o Contamination. Bridge contamination impeded work progress. The bridge 

was extensively decontaminated before removal from the vessel. However, 

loose contamination levels increased as the liquid contamination dried on the 

bridge. This required extensive decontamination of the work areas where the 

bridge was handled during repairs. (204)

o Krypton-85 Releases. While performing plasma arc cutting, small amounts of 

krypton-85 gas were released when the ceramic fuel fragments were heated. 

The quantities released to the containment building and subsequently to the 

atmosphere through the monitored vent path were small and well within 

regulatory limits. (205)

• Plasma Arc Cutting Hazards. One concern from the use of the plasma arc 

torch was toxic gas generation. By-product gases included fissile material 

vaporization; krypton-85; carbon monoxide; small quantities of hydrogen (less 

than 1/10 standard cubic feet per minute); nickel carbonyl vapor; and oxides of 

nitrogen. The principal oxides of concern were nitric oxides and nitrogen dioxide. 

Schematic of the automated plasma arc cutting equipment 

system used to cut apart the massive stainless-steel lower core 

support assembly for removal from the reactor vessel. (481.12)
65

Safety evaluations concluded that gases that escaped from the water surface would

be captured by the defueling work platform off-gas system. The gas was expected 

to be diluted to insignificant concentrations in the containment building near the 

plant purge system exhaust suction point, even during periods of no purge 

operation. Condensable gas products were retained in the water. Administrative 

controls were required to ensure that personnel access to the discharge areas inside 

the containment building was prohibited during plasma arc cutting. The safety 

department monitored the work area for by-product gas to ensure it did not exceed 

occupational exposure limits. (206)

• Plasma Arc Cutting Performance. Several actions were taken to address 

concerns about automated (plasma arc) cutting equipment system performance: 

(●) improvements to increase the reliability of this system, (●) increase in the size 

of the pieces cut from the lower core support assembly to decrease the number of 

cuts, and (●) development of the core boring machine as a backup. (207)

• Vacuum and Air Lift Tools. Inlet velocities of 1.5 to 3.0 meters per second 

were required to entrain and carry the fuel debris into the vacuum and air lift 

systems. The effective capture distance for entraining debris was less than the inlet 

diameter of the suction nozzle. (208) The major concern during airlifting was the 

continual loss of visibility and redistribution of small and medium particles over 

Schematic of the air-lift system. A simple air-lift 

system injects air near the foot of the pipe: bubbles 

rise through the pipe; the liquid/air mixture within the 

pipe rises to the surface of the tank and draws water 

and solid debris through the opening at the foot of the 

pipe; and the debris eventually settles into a debris 

bucket. (481.12)
66

every horizontal surface in the vessel. This problem was addressed to some degree 

by modifications to the defueling water cleanup system’s in-vessel filtration

system, improving recovery time from 4 hours to 2 hours. (209) Based on a review 

of this experience, one account suggested that a series of filter devices could be 

used to capture all particulate material during similar defueling operations. Such a 

setup would require some additional filter changes but would also possibly 

improve defueling efficiency. (210)

• Air-Lift System. The air-lift system (ALS) successfully loaded many thousands 

of kilograms of debris from the core region rubble bed and from the lower head 

region of the reactor vessel. Debris, water, and air separation techniques 

effectively packaged the debris into a debris bucket, which was then placed into a 

defueling canister. The ALS used compressed air to produce the motive force for 

effectively transporting fuel debris (ceramic-like rubble) from beneath the lower 

core support assembly (LCSA) into a standard debris bucket. Gravity caused the 

debris to separate from the transport stream. The entire method did not rely on 

moving parts. INEL performed a full-scale test of the ALS using a mockup section 

of the LCSA placed in the bottom of a tank containing lead shavings; cubes of lead 

measuring 2.5 centimeters; lead shot of various sizes; and sections of 

1-centimeter-diameter stainless steel tubing that was 2–5 centimeters long. Tests 

demonstrated that the ALS could transport fuel debris from beneath the LCSA into 

a standard debris bucket at a minimum rate of 230 kilograms per minute. (211, 212)

4.8 Defueling Canisters

The defueling canisters were designed to accept and confine core debris ranging in 

size from particles (known as fines) of about 0.5 micron in diameter up to 

partial-length fuel assemblies of full cross section. The canisters were intended to 

provide confinement for offsite transport using a shipping cask and for long-term 

storage of core debris. Three types of defueling canisters were designed and 

fabricated: (●) The fuel canister was designed as a receptacle for large pieces of 

core material, which were picked up and placed either directly into the canisters or 

into another container that would then be inserted into the canister. (●) The 

knockout canister was designed for use in the fuel debris vacuum system to 

separate debris particles ranging from about 140 microns up to full pellet size or 

larger. (●) The filter canister was designed for use in the fuel debris vacuum 

system (particles that passed through the knockout canister), the defueling water 

cleanup system, and the canister dewatering system. (213)

• Design Factors. Designers began work on the defueling canister in 1980 

without knowing the physical details of the fuel debris. The final canister design 

influenced the designs of interfacing systems and the shipping cask, as well as the 

defueling process. Factors that influenced the design of the canisters included: 

(●) criticality safety; (●) structural integrity; (●) radiolysis of water; (●) ease in 

handling in the reactor vessel; (●) ease in loading; (●) limiting dimensions and 

weight restrictions of the fuel transfer system and shipping cask; (●) the weight 

restriction in the INEL storage pool; (●) the DOE requirement for canister vents; 
67

and (●) considerations of whether to use multiple single-purpose canisters or a 

single, multipurpose canister (the single-purpose canister was selected). Finally, 

since the canisters were part of the NRC’s certified transport package, they had to 

comply with NRC transportation regulations. (214, 215)

• Canister Diameter. The canister design process determined that an increase in 

the outer diameter to 14 inches from an originally planned diameter of 

13.25 inches was desirable. The increase in canister diameter considered the 

following factors: (●) The Boral plate shroud assembly for the fuel canister design 

would be an off-the-shelf-design item for a 14-inch diameter canister but would 

need to be redesigned for a smaller diameter canister. (●) A shroud for a 14-inch

diameter canister would have had a relatively larger cross-sectional area than a 

shroud for the smaller diameter canister and would make loading of damaged fuel 

assemblies an easier task. (●) A larger diameter canister would have a larger 

volume per canister for loading fuel, and fewer canisters would be required to load 

the entire core. (●) A larger outer diameter would be needed for the hydraulic 

performance of the knockout canisters since smaller diameter canisters would have

increased flow velocities and less settling of small particles. (216) It had been 

reported that the inside dimensions of the fuel canister should have been larger to 

better accept distorted debris and end fittings. Some debris had to be resized and 

some end fittings had to be placed in storage drums to avoid difficult 

resizing. (217, 218)

• Canister Length. One important experience from TMI-2 was the potential for 

conflict between maintaining flexibility to redesign as conditions became known

Drawings of three types of defueling canisters. Canister design criteria included 

criticality control, hydrogen generation control, and survivability during a 

transportation accident. (481.13)
68

and the lead time required for design and safety assessments. (219) Some options to 

modify the canister design based on core conditions and operating experience 

would have been useful. (220)

An early (September 1981) study of fuel debris canister designs proposed one that 

could fit a full-length fuel assembly. (221) The sonar mapping of the core’s 

topography in August 1983 indicated that few full-length fuel assemblies were left 

standing in the reactor vessel. This opened the possibility that the damaged fuel 

could be shipped in shorter length canisters, rather than full-length 

431.8-centimeter (170-inch)-long canisters. The possible use of canisters that were

only 330.2 centimeters (130 inches) long provided an opportunity to evaluate 

improvements in shipping economics and logistics using Government-owned 

M-130 rail casks, which had an inside cavity length of 330.2 centimeters 

(130 inches). After receipt of the cask proposals in June 1984, canister length was 

increased to 150 inches, which was the maximum length specified by potential 

suppliers. The slightly shorter length improved the ease of handling in the reactor 

vessel. (222) An even shorter canister design option (75 inches long) was considered 

to restrict the weight of the contents while maintaining a stack height of 

150 inches, but other options were not discussed further in the literature. (223)

• Weighing of Canisters. One major problem area that was encountered during 

the design stage was a simple and reliable way to measure the weight of the three 

different types of canisters while they were being loaded with debris in the reactor 

vessel. The total loaded dry shipping weight restriction was 2,800 pounds, which 

was a limitation of the canister storage pool at INEL (5 percent of the canisters 

could be up to 5-percent overloaded or 2,940 pounds). Several factors that 

contributed to the difficulty in developing a practical system for weighing fuel 

canisters in the reactor vessel included the following: (●) the expected wide 

variation in the debris density; (●) the mechanical and operational conditions 

during which the weighing system needed to be able to function; (●) the 

multielevation positions of canisters on the carousel (canister position system);

and (●) the rotational motion of the carousel. (224) Solutions included the 

following: (225)

o Knockout Canister. The knockout canister was weighed during all vacuuming 

operations using the knockout canister connect assembly module, which was 

mounted to the underside of the defueling work platform. 

o Filter Canister. The filter canister was weighed continuously during all 

vacuuming operations by the filter canister weighing system, which was

attached to the filter canister and suspended from the defueling work platform

deck shielding plate. 

o Fuel Canister. The fuel canister was weighed on an as-needed basis using the 

canister grapple tool (an open fuel can lifting tool) and a typical weight scale in 

the rigging of the containment building service crane. Since fuel canisters were 

open topped during loading, visibility during loading permitted the operators to 
69

accurately judge the fullness of the canister, and thus the canister was only 

weighed once or twice for verification.

o Dewatered Canisters. All three canister designs were weighed following 

dewatering in either the containment building or the fuel handling building. 

Each canister was picked up and raised out of the water by the canister 

handling trolley. This allowed the load cell in the canister handling trolley to 

sense the canister out-of-water weight to within 35 pounds. Each canister was 

weighed before and after dewatering to determine the amount of water 

removed during dewatering and to demonstrate that the dewatered canister 

complied with the maximum shipping weight restrictions.

• Design Controls. A detailed canister design interface control program 

document coordinated the numerous organizations involved in canister and 

supporting equipment designs. Key areas of design interfaces were established, 

and the lead design organization controlled all exchanges of design interface 

information. (226)

• Decontamination. (227) The external radiation dose rates on most canisters were 

very high, which made hands-on smear wipes not possible at TMI-2 because of the 

licensee’s efforts to keep worker doses ALARA. The focus at TMI-2 was 

decontamination. At INEL, the canisters were measured for contamination.

o Measures at TMI-2. In the fuel handling building at TMI-2, a decontamination 

spray ring with borated hot water was used to clean the loaded canister as it 

was being lifted from the spent fuel pool to be transferred to the shipping cask.

The licensee conducted several experiments with an empty canister to improve 

the decontamination procedures. These included: (●) a high-pressure water 

spray ring system, which failed by a factor of 50 to meet the INEL criteria; 

(●) multiple soakings of a canister in hydrogen peroxide solutions followed by 

hydrogen peroxide solution spraying and hand wiping, which indicated that 

another factor of two for decontamination would be required; (●) hand wiping 

and cleaning with a bristle brush, which showed that hand wiping was the most 

effective but still did not meet the requirements; and (●) using a 

decontamination spray ring with cold water and heated water, which showed 

that heated water was best and came closest to meeting the requirements.

o Measures at INEL. INEL was able to use the hot shop overhead manipulator to 

take smears of the external surface of a canister remotely as each was removed 

from a cask. INEL provided feedback on surface contamination levels for the 

licensee to improve its decontamination process. The problem was never 

completely eliminated but was significantly improved by the licensee’s efforts.

• Contamination Traps. An evaluation of the design of the canister closure 

cover considered that the cover could trap contamination and carry it from the 

reactor to the spent fuel pool. (228) Each canister had process connections located 
70

on the lid of the canister for filling, closing, dewatering, inerting, and 

monitoring. (229)

• Lid Gasket Seal. The metal gasket seal between the removable upper head (lid) 

and the bulkhead on the canister body was found to leak too easily during remote 

installation of the heads to the bodies using manual long-handled tools. Leaks 

occurred during defueling operations, even though the metal seals were able to 

pass the pneumatic pressure test of 150 pounds per square inch as required for an 

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel 

Code-stamped pressure vessel. The seal material was changed to an elastomer with 

DOE approval. (230)

• Serial Numbering. Recordkeeping of loaded canisters was a challenge for 

canister components with more than one serial number (e.g., lid and body). Since 

the fuel canisters were ASME-Code-stamped pressure vessels that had been 

pressure tested with a “matching” head and body (i.e., both head and body had the 

same number), INEL was concerned that interchanging heads would possibly 

negate the ASME Code stamp and complicate recordkeeping of the fabrication, 

documentation, and identification of canisters. The licensee determined that 

interchanging the heads did not negate the ASME Code stamp but committed to 

limiting the interchanging of heads and bodies to special situations only. (231)

• Canister Positioning. The canister positioning system (CPS) supported the 

fuel and knockout canisters in the reactor vessel. The CPS had five canister 

holding positions (sleeves), a means for lowering canister positions into three 

preset vertical positions, and a manual rotational drive unit. Various handheld 

tools were used to facilitate operation. The CPS was mounted to the defueling 

work platform. The elevation of each canister in the CPS could be changed to 

minimize the canister height above the core debris bed as the bed level was 

lowered. The canister elevation was adjusted by using a long-handled tool to 

change the height of the support sleeves. (232)

Suggested improvements in the design of the CPS based on actual defueling

included the following: (●) More adjustability for vertically positioning the 

canisters in the reactor vessel would have improved the working height above the 

rubble bed. (●) The canister and sleeve could have been designed to use a single 

handling tool to minimize tool changeouts. (233)

• Canister Filler Material. A lightweight concrete was used to fill the void 

between the square inner shroud and circular outer shell. This filling provided 

continuous lateral support to both the outer shell and the shroud. This resulted in a 

distributed loading function for horizontal drops and only insignificant 

deformations in the shroud shape, thus preserving a safe physical dimension for 

criticality safety. The fill also provided weight to ensure that an empty canister 

would be nonbuoyant, a general canister design criterion. The low-density 

concrete mixture contained cement, glass bubbles, and demineralized water. (234)

The water present in the concrete mixture was extremely difficult to remove 
71

during the drying process at INEL. Canister drying is needed for long-term storage 

of fuel canisters; therefore, in hindsight, the material was not a wise choice. Many 

debris components oxidized, leaving a net surplus of hydrogen that required

venting. Other options for filler material could have been explored, such as 

aluminum or glass beads. (235)

• Materials Compatibility. Water and materials that could come into contact 

with the canister’s internal components were evaluated for potential impacts. 

Modifications to water cleanup systems and water inhibiters were also evaluated 

for impacts. For example, the TMI-2 project experienced problems with the 

premature plugging of filter canisters by microorganisms and fine core materials. 

A biocide (hydrogen peroxide solution), a diatomaceous earth body feed, and a 

coagulant were used to enhance performance of the filter coating. The DOE 

reviewed the use of these materials for both transportation and long-term storage 

at INEL. The NRC’s transportation certification staff also reviewed and approved

the potential impact of these materials on the safety of the shipments. Reviewers

evaluated the effect of adding the body feed, coagulant, and biocide on the 

canister’s catalyst hydrogen recombiners. (236)

• Quality Assurance. A valuable lesson from TMI-2 was that production support 

functions, such as material traceability, documentation, quality control, and quality 

assurance (QA), should be sufficiently staffed to support the production 

schedule. (237)Based on this experience, QA involvement of the manufacturer of 

defueling canisters should have started upon receipt of the material. The NRC and 

the licensee identified substantial problems relating to the documentation of 

material used in the fabrication of defueling canisters. This led to serious delays in 

fabrication and narrowly avoided an adverse effect on the defueling schedule. It 

also highlighted the dangers of relying on one low-bid vendor for such difficult 

and important work. After a period of delays and uncertainty, the contract was 

withdrawn from the initial vendor and given to two new vendors that produced 

satisfactory canisters. (238)

The licensee’s contractor QA department was responsible for manufacturing 

inspections during canister fabrication, and it was decided to establish a full-time 

inspector in the shop. The licensee’s QA department provided periodic oversight. 

INEL (receiver of the loaded canisters) and the NRC conducted spot checks as 

well. (239) As a result of a DOE request, the NRC inspected defueling canisters 

during the fabrication process, which included welding observation,

nondestructive examinations, and fitting of components. (240)

• Vendor Inspections. (241) At the DOE’s request, which was the receiver of 

loaded defueling canisters, the NRC expended about 782 inspection hours related 

to the TMI-2 defueling canisters and transportation casks. The focus during these 

inspections shifted from a previous “paper” review to a safety review that involved

a more thorough examination of equipment and the implementation of the quality 

assurance (QA) program during the fabrication of components. The NRC 

interaction with vendor personnel was largely positive. The vendors responded to 
72

the inspection findings by promptly initiating corrective actions that addressed 

root-cause issues and gained a better understanding of, and insight into, the 

regulations. Because of the interactions during the inspection process, 

manufacturing and design controls were improved, thus resulting in an improved 

product. The NRC staff believed that this was a direct result of the shift in 

inspection focus to hardware and implementation of quality activities.

Based on reviews of the canisters and cask designs, the NRC’s vendor inspectors

identified fabrication processes to be inspected. Inspection plans identified the 

specific areas to be examined. Inspection areas included: (●) fabrication and 

inspection processes, such as welding and nondestructive examination;

(●) material identification and nonconforming item control throughout the 

facilities to ensure that nonconforming items were not used; (●) fabrication and 

acceptance tests and test results to ensure that acceptability criteria had been met; 

(●) training and qualification records of welders, nondestructive examination

inspectors, and QA inspectors; (●) material storage to ensure that the material was

not exposed to adverse conditions; (●) a calibration program for measuring and 

test equipment to ensure that calibrations were performed as required, by qualified

personnel, with standards traceable to the National Bureau of Standards (currently 

the National Institute of Standards and Technology), and with calibrated 

instruments; (●) a procurement program, including purchase orders, to ensure that

materials were bought from qualified subvendors, that receipts were inspected, and 

that the vendor had some control over the QA program in place at the subvendor’s 

facility; and (●) design and drawing control to ensure that changes were made in a 

controlled manner and all documents were the latest revision.

4.9 Defueling Operations

Removal of damaged fuel and structural debris from the reactor vessel started on 

November 12, 1985, 6.5 years after the accident. Numerous manual and 

hydraulic-powered long-handled tools were used to perform a variety of functions, 

such as pulling, grappling, cutting, scooping, and breaking up the core debris. 

Eventually, more powerful tools, such as the core bore machine, plasma arc torch, 

and water jet cutting system, were used to disassemble or cut apart reactor vessel 

components and break up resolidified core material. After breaking up and 

sectioning oversized debris, long-handled tools manually loaded debris into 

defueling canisters positioned underwater in the reactor vessel. The larger pieces 

of vessel internal components, such as lower core support assembly sections, were 

lifted out of the vessel by crane and stored in the modified core flood tank “A.”

Smaller pieces or “fines” were vacuumed into knockout canisters and filter 

canisters. Residual fuel remaining after defueling operations is listed in Table 3.

Other defueling activities included transferring the loaded defueling canisters from 

the containment building to the fuel handling building, dewatering the filled 

canisters, and placing canisters into the canister storage racks located in spent fuel 

pool ”A.” Removal of fuel and structural material from the reactor vessel was 

completed in stages. The plan for each stage incorporated experience gained from 
73

previous stages and activities, including visual in-vessel inspections and sample 

examinations. (242)

• Water Safety. The Occupational Safety and Health Administration rules on 

longshoring and marine terminal operations were useful guidelines for worker 

safety over water. (243) Only one incident report involved a worker on the defueling 

work platform who slipped through an open service hole and fell partway into the 

reactor vessel. The worker had temporarily and inappropriately unhooked himself 

from a sling to accomplish a work task. (244)

• Personnel Contamination. Workers experienced many instances of personnel 

contamination, which was most likely the result of reactor coolant spray or splash 

when tools were raised from and inserted into the reactor vessel. (245)

• Defueling Staffing. Initially, the work was conducted by teams of five. These 

teams were later reduced to as little as two, due to improved skills and experience 

of the workers. The original team included a tool handler, camera handler, crane 

operator, radiation control technician, and camera controller. Eventually, the 

camera controller function was performed by the tool handler when the camera 

controls were moved closer to the work slot in the defueling work platform. The 

radiation control technician, with a high-range radiation detector, was only 

involved when tools were taken out of the work slot. Supplementary personnel 

included the polar crane operator and the construction work force to assist in the 

tool movements outside the canal area and to close the debris canister. The 

NRC-licensed defueling senior reactor operator, stationed in the command and 

control center located in the turbine building, was technically in charge of the 

defueling operations requiring reactivity manipulations and was in constant radio 

contact with the defueling team members. (246)

• Opening the Reactor Coolant System to Atmosphere. Although the reactor 

head lift was found to be straightforward and was performed without technical 

difficulties, the operation required a great deal of planning and management 

resources because of concerns about opening the reactor coolant system to the 

atmosphere. The lift also drew considerable attention from the news media and 

public. (247)

• Hose Leaks. Hose failures occurred several times during the defueling process. 

Failures were typically caused by cracked hoses and snagging while raising tools 

out of the vessel. Even rubber hoses of the highest quality and pressure rating were 

known to fail. Based on this experience, it was reported that the use of rubber 

hoses should be minimized. Applications with rubble hoses should consider the 

consequences of failure in the design and response procedure. (248)

• Microbial Growth. Only a few months after defueling started, water clarity 

began to deteriorate. Video camera visibility was rapidly lost, and defueling was 

hampered. The problem turned out to be microbial growth (and, to a lesser extent,

fine particulate debris) that first plugged the sintered metal filters in the defueling 
74

water cleanup system and then began accumulating in algae-like masses within the 

reactor vessel. It took more than a year to analyze the source, decide and gain 

approval for a method to prevent future occurrence, and then to completely restore 

clarity and visibility. Studies revealed that small amounts of hydraulic fluid from 

the defueling tools had leaked into the reactor coolant and provided the organic 

food source for the microorganisms. This was aided by the appropriate levels of 

water temperature and light intensity from the underwater video camera lights to 

promote growth. Ultimately, hydrogen peroxide was used to kill the growth and 

prevent recurrence. (249, 250)

• Biocide. Considerable time was spent attempting to improve visibility inside 

the reactor vessel. Numerous techniques, with varying degrees of success, were 

evaluated serially to bring microorganism growth under control and to improve 

clarity by the end of 1986. One account suggested that, had potential solutions

been investigated in parallel (instead of serially), the microbial growth issue could 

have been resolved earlier. (251) Insights from biocide evaluations and 

implementation included the following:

o Biocide Concentrations. As the result of extensive research and testing, the 

NRC approved 200 parts per million hydrogen peroxide as the biocide. The 

NRC’s evaluation concluded that the biocide concentration was compatible 

with the following: (●) the existing water chemistry in the reactor coolant 

system; (●) water processing systems (i.e., EPICOR II, submerged 

demineralizing system, reactor vessel filtration system); (●) the defueling 

canister’s catalytic hydrogen recombiners; and (●) waste disposal 

Microorganism growth in the reactor vessel slowed defueling to a halt within 

a month after defueling started. Shown is the vessel annulus region between 

the lower core support assembly and the vessel wall.
75

requirements. In addition, the biocide did not negatively affect criticality

safety, given that hydrogen peroxide was borated to levels that complied with

recovery technical specifications. The evaluation also concluded that the 

increase in reactor coolant activity levels due to the increase in the rate of 

cesium leaching from the debris bed caused by the hydrogen peroxide would 

not have a significant impact on worker safety with the use of normal 

radiological control practices. (252, 253, 254)

o Initial Shock Treatment. Industrial strength hydrogen peroxide (15 percent) 

was added to the reactor coolant in the reactor vessel and spent fuel pool where 

the defueling canisters were stored before shipment. About 750 gallons 

(2,800 liters) were added initially to provide a shock treatment, with smaller 

concentrations added thereafter. (255)

o Gas Generation. Large amounts of hydrogen peroxide produced large 

quantities of gas that had to be vented. This gas produced significant liquid 

carryover onto the defueling platform surfaces and fuel handling building 

floors that required aggressive decontamination. (256)

o No Chloride-Induced Corrosion. Hydrogen peroxide contained no chlorides 

that would corrode stainless steel and left no residue in the reactor coolant after 

oxidation. (257)

o No Microbial-Induced Corrosion. Inspections, corrosion studies, and 

laboratory tests found no microbial-induced corrosion of components and 

defueling equipment in the reactor vessel. (258)

o Other Disinfection Approaches. Disinfection, a process in which organisms are 

destroyed or inactivated, could be accomplished by several different 

physiochemical treatments, including thermal energy; ultraviolet, gamma, 

X-rays, and microwave irradiation; ultrasonic disruption; and chemical agents,

with the latter the most common method of disinfecting water and waste 

waters. (259)

• Water Clarity. One account indicated that the reactor vessel defueling water 

cleanup system could have been sized to restore water clarity more rapidly. 

Operation of underwater equipment with limited visibility greatly reduced 

productivity while increasing the risk of equipment damage and lost time. The 

design basis for water clarity should have been 0.1 nephelometric turbidity unit 

(ntu) (about 0.1 parts per million suspended solids) instead of 1.0 ntu for manual 

tool operation at depths of 6 to 10 meters. (260) Working inside the reactor vessel 

often resulted in the resuspension of very fine particulates, the mass of which

precluded settling. This resulted in a smoky effect that reduced visibility from

12 meters of water shielding to less than 1 meter, and often much poorer. (261)
76

• Molten Material Formation. Analysts had not predicted the extent to which 

the molten material had formed into a large monolith within the reactor vessel or

understood the material’s resistance to fracture. (In fact, fuel melt was first 

reported in 1985 after careful examination of the first “grab” samples from the 

damaged core, 6 years following the accident.) The molten material that formed in 

the TMI-2 core was a complex mixture of the major core constituents (uranium, 

zirconium, and iron) with lesser amounts of control rod material (silver, indium, 

and cadmium) and other alloy constituents (chromium and nickel). The molten 

material also contained a significant quantity of oxygen, making the solidified 

melt a highly refractory ceramic. A portion of the molten material solidified into a 

heterogeneous, funnel-shaped disk over a meter thick. The mass was located in the 

center of the reactor vessel below the plenum. Elsewhere in the reactor vessel,

molten material flowed down between the baffle plates and the core barrel and 

solidified in the reactor vessel’s lower head region. Both solidified masses 

weighed many tons. Heavy-duty defueling tools (impact chisels and wedges) were 

unable to break apart the monolith. A specialized drilling apparatus (core bore) 

was finally used to pulverize the material. (262, 263)

Technician at the controls of the core bore machine located on an 

elevated platform mounted to the defueling work platform. The core 

bore was used to drill large holes in the solidified molten core 

region. The drill string, which was mounted through the center of 

the motor, can be seen at the right. A drill bit was mounted at the 

end of the string; additional strings were fastened together as 

needed. 
77

• Remote Coordination Center. The remote coordination of defueling operations 

increased productivity and minimized radiation exposure. The defueling activities 

on the defueling work platform were monitored from a coordination center 

established in the nearby turbine building. Support staff observed the work teams 

using closed circuit video cameras mounted in the containment building. The 

camera pan, tilt, and zoom controls were remotely operated from the coordination 

center. Video monitors in the coordination center also displayed images from the 

in-vessel video cameras used by the defueling teams. In addition, the coordination 

center was in radio communication with each member of the defueling team. The 

center recorded the video and audio of the operations for archival purposes, 

analysis of the footage, and worker training. Support for the defueling operators 

(e.g., special data; engineering expertise; procedural changes; suggestions for 

solving problems; ALARA advice) came by communications from the 

coordination center rather than by sending additional workers into the containment 

building. (264, 265)

• Number of Canisters. Estimating the number of canisters to be used in 

defueling evolved into a high art form and one with significant potential cost. The 

original estimate was 243. The first half of the INEL pool could store 

288 canisters. The project’s total estimate grew as defueling progressed because 

the weight loaded into each canister was usually less than anticipated. As 

important, the relative numbers of fuel, filter, and knockout canisters changed in 

response to the techniques used and unexpected conditions in the reactor vessel. 

The final estimate was between 349 and 360 canisters. (266) At the conclusion of 

the defueling effort, a total of 342 canisters of core debris, in 22 rail shipments,

were transported to INEL. The total number of canisters included 286 fuel 

canisters, which contained partially intact fuel assemblies and large debris picked 

up from the reactor vessel; 12 knockout canisters, which contained core debris 

vacuumed from the reactor vessel and reactor coolant system; and 62 filter 

canisters, which contained fine debris that had passed through the knockout 

canisters. (267)

• Extraction of Standing Fuel Assemblies. The outer ring of the core contained 

standing fuel assemblies of various lengths. Once the first assembly was removed 

by the lasso tool, the others came out easily. The tools that were developed for 

removing standing assemblies included hydraulic side grippers, spears, and long 

L-shaped assembly lifters. (268)

• Cutting Incore Guide Tubes. The core boring machine proved very adept at 

cutting out in-core guide tubes and support posts, which was a weakness of the 

automated cutting equipment system. However, cutting was only effective when 

enough space existed for the chips to fall away. (269)

• Baffle Plate Handling. (270, 271) Vertical baffle plates formed the peripheral 

boundary of the core. Roughly 4,000 kilograms of fuel debris relocated behind the 

baffle through one large hole in the baffle plates caused by the flow of molten core 

material from the core region. Several options were considered for removing the 
78

highly irradiated baffle plates (with an estimated peak contact gamma radiation 

field of approximately 3,000 roentgen per hour in air) to gain access to this debris. 

o Options. The removal options included the following: (●) Cut the plates into 

small pieces and load them into fuel canisters or other specially designed 

containers. (●) Remove the small pieces of plates from the vessel and store 

them either in a remote location (e.g., the containment building basement) or in 

a modified core flood tank (sections of the lower core support assembly were 

stored in these tanks). (●) Cut the plates into large sections, rotate each section, 

and remove debris one plate at a time. The latter option, which was selected,

required no plant modifications; resulted in lower radiation exposures; required 

minimum tool development; and did not introduce any new safety concerns 

(e.g., lifting and handling out of the vessel) or difficult failure scenarios.

o Selection. The plates were cut into eight sections using the plasma arc torch. A 

total of 864 bolts that held the baffle plates to the core barrel were removed 

using an untorquing tool and a drill tool. The 4-meter-long pieces of highly 

irradiated stainless steel were highly radioactive with an estimated peak of 

3,000 roentgen per hour gamma. Baffle plate handling exposed the upper core 

support assembly for defueling of the core former area. Two of the eight baffle 

plate sections were removed and hung from vent valve seats. The exposed area 

was defueled before removal of the next plate section. A computer model of 

the vessel was used to help select how to shuffle the baffle plates from hanger 

to hanger in a manner that least affected defueling and required the least 

Piece of the lower grid forging being removed from the reactor vessel. Pieces 

of the lower core support structure were moved in a modified core flood tank 

for storage.
79

handling. Each plate was essentially rotated 90 degrees from its original 

location to its final location.

• Defueling Incidents. Very few incidents, two injuries, and a few other 

relatively minor incidents were reported during defueling operations:

o Dropped Canister. On December 14, 1985, a load drop event occurred early in 

the defueling operations when a defueling canister and support sleeve fell into 

the reactor vessel. The licensee had been loading fuel assembly end fittings 

into a defueling canister when an end fitting became stuck in the canister. 

During attempts to reposition the stuck end fitting with the 1-ton jib crane, the 

defueling canister and support sleeve were dislodged from the canister 

positioning system and dropped 1.5 feet onto the top of the debris bed in the 

reactor vessel. The dropped load weight was 2,200 pounds; the jib crane was 

rated at 2,000 pounds. No increases in radiation levels and airborne activity 

were detected during the event and subsequent retrieval and recovery. (272)

o Dislodged Defueling Tool. On May 22, 1986, the trolley on the number one jib 

crane on the rotating work platform disengaged and fell into the reactor vessel. 

An operator was struck by a portion of the falling apparatus and received a 

scalp laceration. Approximately one-half of the trolley was recovered from the 

vessel and examined to determine the failure mechanism. The trolley on the 

number one jib crane was replaced with another of a different design. (273)

o Dislodged Defueling Tool. On August 19, 1986, a worker on the defueling 

platform was injured when a long-handled tool in the reactor vessel was 

dislodged from its temporary storage location and fell onto the worker’s right

hand, pinning it between the tool and the tool rack. The worker sustained an 

apparently severe laceration and possible fracture of one finger. (274)

o Worker Slipped into Reactor Vessel. On May 22, 1988, a worker on the 

defueling platform slipped and fell partway into the reactor vessel through an 

open access hatch. The individual worker was installing equipment in the 

reactor vessel in preparation for the reinstallation of the plasma arc cutting 

assembly. The worker had temporarily unhooked himself from a sling to 

accomplish the work. The fall was broken when the worker grabbed the 

defueling platform. The worker’s legs, up to his knees, were wet and 

contaminated. Protective clothing included a double protective clothing wetsuit 

and respiratory protection. Preliminary licensee dose estimates indicated that 

the worker was expected to receive about a 200-millirad skin dose to the legs. 

The worker received only minor bruises and did not require any medical 

treatment. (275)

o Workers in Contact with Fuel Debris-1. A small section of a fuel rod was 

unknowingly withdrawn from the reactor vessel when the defueling tool was 

removed from the reactor vessel, and the section of fuel rod fell on the work 
80

platform. The piece measured 10 roentgen per hour on contact and was picked 

up with a Peters tool (vice grip) and dropped back into the vessel. (276)

o Workers in Contact with Fuel Debris-2. In another incident, in 1989, one

worker received an overexposure of about 55 rem to the hand (which was in 

excess of regulatory limits) while unknowingly handling a piece of fuel debris 

in the TMI-2 containment building decontamination facility. A coworker 

received an unplanned exposure of about 13 rem to the skin on a hand. This 

facility was employed to repair, disassemble, and decontaminate equipment 

used in the defueling operation. During this cleaning evolution, one worker 

(without wearing a beta glove) picked up what he believed to be a piece of

debris from the floor and threw it in the area of a trash container. This same 

individual then picked up the item a second time because he wanted the

radiological controls technician to survey the item at the first opportunity. 

Shortly thereafter, the second worker picked up what was apparently the same 

piece of debris and briefly held it in his hand. Subsequently, in response to a 

concern raised by the first individual, the radiological control technician

responsible for monitoring the cleanup activities in the facility surveyed the 

debris and determined that it was a highly radioactive fuel fragment. The 

radiation survey (277) of the material indicated contact does rates of 1,320 rem 

per hour of gamma radiation and 11,580 rem per hour of beta radiation. The 

technician assigned to monitor the work was unaware that either individual had 

picked up the fuel fragment and failed to investigate. An earlier radiation 

survey of the work area did not identify the fuel fragment because other debris 

on top may had shielded the fragment. (278)

The NRC concluded that this incident demonstrated the need for (●) improved 

planning and preparation of work activities in contaminated or potentially 

contaminated areas, (●) improved radiological oversight and control of such 

work activities to ensure that such activities are conducted safely and in 

accordance with regulatory requirements, and (●) improved communications 

between the radiation protection personnel and the work crews they are 

monitoring. (279)

o Inadvertent Core Alteration-1. In March 1986, the defueling canister 

positioning system mounted in the reactor vessel was rotated with canisters 

containing core material without direct supervision of a licensed fuel handling 

senior reactor operator. This activity was classified as a core alteration in 

accordance with the recovery technical specifications. The lead engineer 

should have requested permission from the command center before initiating 

the core alteration activity. (280)
81

o Inadvertent Core Alteration-2. The air-lift vacuum system (ALS) used a debris 

bucket to collect debris. The debris bucket had a trap door that unloaded the 

bucket’s contents into the defueling canister. In 1989, unlicensed defueling 

operators concluded that the ALS did not contain a significant quantity of fuel 

debris when the system was removed from the reactor vessel for eventual 

placement in temporary storage. Visual inspection of the debris bucket before

its removal was hampered due to a lack of water clarity in the reactor vessel. 

To facilitate flushing of the bucket, the bottom door of the debris bucket was 

opened using an extension pole. Personnel stationed near the reactor vessel 

observed an unknown quantity of core debris fall from the debris bucket into 

the reactor vessel. This incident constituted an inadvertent and unplanned core 

alteration since a licensed fuel handling senior reactor operator did not 

authorize this action as required by the recovery technical specifications. The 

absence of a detectable increase in the weight of the debris bucket did not 

provide positive verification of the absence of fuel debris. The licensed 

operator and task supervisor should have assumed the presence of core debris 

in the debris bucket and classified this event as a core alteration before removal 

of the ALS. (281)

Defueling canister transfer cask positioned over a port hole in the defueling work 

platform. The canister transfer bridge raised the defueling canister inside the 

transfer cask, which was then moved to the deep end of the refueling pool.
82

Table 3. Residual uranium dioxide fuel (UO2) remaining after defueling 

operations. (282)

Reactor Vessel (kg UO2)

Defueling Work Platform Region

In-Vessel Filtration System 16.3

Canister Positioning System 13.7

Downcomer Region

Cold Leg Flow Deflectors 11.9

Hot Leg Bosses in Core Support Structure (CSS) 26.6

Surveillance Specimen Capsule Holders 3.5

Thermal Shield Support Blocks (Top Surface) 15.2

Thermal Shield Inner Surface and Annular Gap 118.6

Core Catchers/Seismic Restraint Blocks 2.9

Internals Indexing Fixture (IIF) Region

Reactor Vessel Flange, IIF Flange, CSS Flange 4.9

CSS Region

Vent Valve Seats (Inner Surfaces) 8.7

Top of Lower CSS Flange 1

Upper Core Support Assembly Region

Baffle Plate Inside Surface 17

Baffle Plate Outside Surface 17

Baffle Plate Flow Holes and Bolt Holes 10.5

Former Plates Top and Bottom Surfaces 39.9

Lower Core Support Assembly Region

Lower Grid Rib Section (LGRS) (a) 41.3

Between LGRS and Lower Grid Distributor Plate (LGDP) 12.8 (b)

Between LGDP and Forging 48.2 (b)

Forging Peripheral Flow Holes 110.1

Inside Support Post Stubs 1.4

Between Forging and Incore Guide Support Plate (IGSP)(c) 174.3 (b)

Between IGSP and Flow Distributor 39.7 (b)

Bottom Head Region

Head Surface 104.6

Incore Instrument Nozzles 29.4

Standing Incore Guide Tubes 17.6

Surface Film Deposits 2.1
83

Table 3. Residual UO2 remaining after defueling operations. 

(Continued)

Reactor Coolant System (Excluding Reactor Vessel) (kg UO2)

Decay Heat Drop Line 1.5

"A" Side Steam Generator

Upper Tubesheet, Tube Bundle, Lower Head, and J-Legs 4.1

Cold Legs 34.1

"B" Side Steam Generator

Upper Tubesheet, Tube Bundle, Lower Head, and J-Legs 51.7

Hot Leg 1.7

Cold Legs 21.3

Core Flood Line (Before Check Valve) 1.2

Reactor Coolant Pump 14.7

Reactor Building (Excluding Vessel and RCS) (kg UO2)

Fuel Transfer Canal 12.7

"A" D-ring 24.3

Storage Area, Upper Endfitting (55-Gallon Drums) 7.7

Letdown Coolers 3.7 (d)

Containment Basement and Sump 1.3

Defueling Water Cleanup System 2.3

Defueling Tool Rack 4.8

Temporary Reactor Vessel Filtration System 4.4

Containment Building Drains 5.1

Auxiliary and Fuel Handling Building (kg UO2)

Spent Fuel Pool "A" 4.9

Reactor Coolant Bleed Tanks 4.31

Subtotal: All Other Cubicles (Each < 1 kg) 7.79

Grand Total (Including Areas < 1 kg Not Shown Above) 1120 kg

Notes

a. Top surface and peripheral flow holes

b. Includes fuel pellets and fuel rod pieces

c. Includes IGSP flow holes

d. Below minimum detection level

Key to Figure on Next Page:

Defueling canister transfer process (right-to-left): (1) Workers on the defueling 

platform loaded fuel canisters. (2) A fuel canister bridge transferred a canister 

from the reactor vessel to the fuel transfer tube and mechanism inside the 

containment building. (3) The second fuel canister bridge in the fuel handling 

building loaded the canister into the fuel storage rack in the spent fuel pool. 

(4) The fuel transfer cask and building crane transferred the processed canister 

from the pool to the shipping cask loading station. (5) The Model 125-B shipping 

cask was lowered from its upright position onto the rail car and prepared for 

shipment.
84
85

5 FUEL DEBRIS TRANSPORTATION

Transport of the damaged core materials from TMI-2 to INEL for examination and 

storage presented many technical and institutional challenges. These challenges

included: (●) assessing the ability to transport the damaged core; (●) removing and 

packaging core debris in ways suitable for transport; (●) developing a transport 

package that could both meet Federal regulations and interface with the facilities 

at TMI-2 and INEL; and (●) developing a transport plan, support logistics, and 

public communications channels suited to the task. The DOE’s “Historical 

Summary of the TMI-2 Core Debris Transportation Campaign,” issued 1993, 

provides a thorough historical summary of how the DOE addressed these technical 

and institutional challenges and transported, received, and stored the TMI-2 core 

debris at INEL. The scope of the experiences presented below is limited to the 

transportation preparations at TMI-2. (283)

5.1 Planning

• Coordination. (284, 285) The close coordination of all organizations involved 

with the packaging and shipment of fuel debris from TMI-2 was required because 

of the broad scope, number of organizations involved, and required interfaces. In 

September 1983, INEL asked the licensee to consider a cooperative coordination 

effort for the TMI-2 core debris transport program. The TMI-2 Core Shipping 

Technical Working Team was formed to coordinate information among member 

organizations involved with the shipment of the TMI-2 core. This group solved 

major technical issues, such as hydrogen gas generation, canister criticality, 

canister design and testing requirements, and cask handling design requirements at 

both INEL and TMI-2. 

o Membership. All organizations involved with the transportation of fuel debris 

were members of the shipping working group. These organizations included 

those responsible for program management (DOE/INEL TMI-2 site office); 

core defueling coordination (licensee, Westinghouse); canister design 

coordination (licensee, Babcock & Wilcox); TMI-2 facility preparation

(licensee, Bechtel); cask supplier (Nuclear Packaging, Inc.); hazards 

evaluations (Hanford Operations in Washington State); transportation 

technology (Sandia National Laboratory Transportation Technology Center); 

transportation support (INEL, TMI-2 site office); and INEL receiving facility 

preparation (Idaho). Other approving Federal agencies, such as the DOE, the 

NRC, and the Federal Railroad Administration, also attended most meetings

and provided valuable input. 

o Meetings. The team provided a focal point for each program task, where 

activity status could be exchanged and potential problems could be identified 

for resolution in a timely manner. These principal organizations, and 

occasionally other special support organizations, attended the regularly held 

team meetings, during which many attendees would hear of progress on 

program tasks. The team approach was essential to the success of the program.
86

• Truck vs. Rail. Truck versus rail shipments and wet versus dry loading were 

the principal alternatives evaluated before procuring the casks. For each 

alternative, the team determined the associated costs and schedules for the number 

of shipments necessary to move the complete core to INEL. Studies showed that 

the licensee’s costs and INEL’s unloading costs would be substantially less for rail 

casks compared to truck casks. Further, rail shipments could transport heavier and 

multiple casks. A rail cask could carry multiple defueling canisters, thus limiting 

the total number of shipments and reducing the chance of an accident. (286)

• Wet vs. Dry Loading. The fuel debris handling studies performed by the 

licensee on truck casks showed that the use of a dry-loading method rather than a 

wet-loading method would reduce cask loading turnaround times at TMI-2. The 

wet-loading method was previously used for loading submerged demineralizer

system liners into a shipping cask under water in the spent fuel pool. The 

dry-loading approach reopened the possibility that loading rail casks in the fuel 

handling building truck bay would be a viable alternative to the procurement of 

truck casks. (287, 288)

• Loading Approach. (289) The licensee opted for dry loading of the cask instead 

of wet loading in the spent fuel pool. That decision was based on operational 

efficiencies, including the fact that much of the equipment dedicated to support of 

the cleanup effort occupied space in the pool. Safety considerations for dry loading 

of the cask included the handling of heavy loads and related accidents.

o Heavy Loads. The major concern with cask loading activities was the handling 

of the heavy loads in the fuel handling building. TMI-2 had an NRC-approved 

heavy-load handling program that was expanded to allow for dry loading of the 

shipping cask. The program complied with the requirements of NUREG-0612, 

“Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic 

Technical Activity A-36,” (290) issued July 1980. Heavy-load drop concerns 

caused a major change in the dry-loading approach during development of the 

cask handling equipment. The hydraulic cylinders used to upright and lower 

the Model 125-B cask for loading at TMI-2 were a practical and safe method 

for allowing use of a cask that was too heavy for the existing overhead crane.

o Accident Analyses. To be loaded, the shipping cask had to be rotated 

90 degrees from a horizontal to vertical orientation. Accident analyses found

that failure of the fuel handling building crane during cask rotation could 

result in dropping the cask onto the railcar, which had the potential to cause 

damage to the truck bay floor and fuel handling building systems in the 

basement (Unit 1 had redundant safety cables in the basement). The cask 

hydraulic lift assembly was built to upright the cask for loading and therefore 

eliminated the potential for failure of the crane during this operation. 

o Dose Reduction. The use of dry-loading equipment improved dose reduction 

and operational efficiency. In comparison to wet loading a cask, there was no 

need for hands-on decontamination of the exterior cask surfaces after removal 
87

from the water. With dose rates allowed by shipping regulations to reach 

200 millirem per hour at a cask’s surface, the potential savings in dose would 

be significant. 

o Cost Savings. In addition to savings in dose, the dry-loading equipment would 

be cost effective for campaigns requiring many shipments. Many working 

hours would be saved per cask loading by eliminating underwater handling of 

a cask and its lid, draining, and external surface decontamination. Further, as 

previously mentioned, the TMI-2 fuel handling building crane would have 

required extensive modifications to permit the lifting of a large and heavy 

cask from the spent fuel pool to the loading bay.

• Loading Equipment Requirements. The dry-loading equipment included some 

general requirements that were applicable to all components. These requirements

included the following: (●) All lifting and handling equipment was designed to 

meet the requirements in NUREG-0612 and American National Standards Institute

N14.6, “Radioactive Materials—Special Lifting Devices for Shipping Containers 

Weighing 10,000 Pounds (4500 kg) or More.” (●) System components were 

designed with redundant safety and operating features to accommodate off-normal

operating conditions (the cited reference did not provide examples). (●) System 

equipment was designed to fail in a safe manner assuming a failure would occur 

(fail safe). (●) Equipment included lead shielding to reduce personnel radiation 

doses. (●) System components were designed as modules to facilitate installation 

Left: Fuel canister lowered into the Model 125-B shipping cask from a fuel transfer 

cask. Right: Shipping cask loading tower lowered the shipping cask to a horizontal 

position onto a skid that would be attached to the rail car.
88

and maintenance. (●) Equipment materials and coatings were selected to facilitate 

decontamination of radioactive materials. (●) Equipment to support and stabilize 

the cask was designed to withstand the seismic requirements for a safe-shutdown 

earthquake. (291)

• Keep Equipment Simple. Most difficulties encountered during loading of the 

cask involved automated interlock systems (e.g., motor driven), particularly those 

in the fuel transfer cask. Many of these interlocks probably could have been 

designed for manual rather than automatic operation (the cited reference did not 

provide examples). Based on this experience, one account reported that, to avoid 

difficulties, equipment should be simple and, where practical, capable of manual 

operation. (292)

• Impact of Unrelated Events. Events unrelated to the shipping activity had an

unintended influence on the TMI-2 shipping campaign. Examples included the 

hazardous material train derailment in Pittsburgh (which led to the evacuation of 

approximately 15,000 people and occurred on the track used for TMI-2 shipments) 

and the bridge derailment near St. Louis (also on the track used for TMI-2 

shipments). Even the Space Shuttle Challenger disaster in 1986 had an effect, 

raising concerns about the behavior of O-ring seals at low temperatures. O-rings; 

although different, were used to seal the lid of the shipping cask. (293)

• Neutron Sources. Special attention was needed to identify and track neutron 

startup sources from the reactor core. Monitoring of a cask at INEL discovered 

one of the sources from TMI-2. Its presence was not documented in the canister 

shipping data and resulted in higher than expected radiation dose to personnel. 

Fortunately, the dose was well below acceptable levels. It was difficult to 

distinguish these sources from other debris while loading canisters and almost 

impossible to track them without neutron monitoring in all subsequent canister 

handling steps. (294)

5.2 Shipping Cask

• Cask Selection. The DOE had primary responsibility for managing the design 

and certification of the shipping casks and their procurement. The following 

factors influenced the selection of a cask for transporting the core to INEL: 

(●) Core debris contained enough plutonium that, in accordance with

10 CFR 71.63, “Special Requirement for Plutonium Shipments,” the material had 

to be transported in a separate inner container placed within an outer packaging 

(i.e., the package had to provide double containment). (h) That decision was 

reached in consultation with the NRC’s transportation group. (●) Breached fuel 

h Editor’s Note: In a revision to the rule, published in the Federal Register on 

January 26, 2004 (69 FR 3698), the NRC modified 10 CFR 71.63 to remove the 

requirement for double containment of plutonium (see the discussion for Issue 17 in the 

cited Federal Register notice). Thus, since October 1, 2004, the effective date of the 

revised rule, NRC regulations no longer require double containment of plutonium.
89

rods could not be considered a level of containment. (●) Participants in the TMI-2 

program decided that the defueling canisters would not be designed to provide a 

level of containment. The canisters needed removable lids and other loading 

features that made them difficult to qualify as a level of containment. Therefore, 

the decision was made that the shipping cask would provide both levels of 

containment. The canister design served other functions related to the integrity of 

the shipping cask as required by the regulations, such as criticality control and 

hydrogen and gas control. (295)

• Regulatory Coordination. Each shipment was thoroughly inspected before 

leaving TMI. The DOE, the NRC, and the U.S. Department of Transportation 

ensured that cask, cargo, and railcar met all necessary Federal requirements for 

safe shipment. A thorough inspection of the railcar, including cask tie-down, was 

conducted to verify that it met the safety requirements of the American 

Association of Railroads. In addition, before transport, radiation surveys were 

conducted by the licensee and NRC inspectors from the NRC TMI-2 project 

office, as well as State inspectors at selected interchange stops enroute to INEL. 

Before the start of shipments, the Office of the Federal Railroad Administration

inspected the entire rail route. (296, 297)

• Cask Fabrication. The Model 125-B rail casks were fabricated in parallel with 

cask certification activities. The fabricator accepted the risk that the cask might not 

be approved as built. This risk was minimized by frequent meetings with the NRC 

to present the cask design as it evolved and by the successful completion of the 

quarter-scale cask model drop test program. Only long-lead-time materials (shells 

and forgings) were ordered before completion of the drop tests. Following the

successful tests, fabrication of the cask components proceeded with some certainty 

that the cask design would not change. (298)

• Cask Certification. Certification of the TMI-2 rail cask was completed in 

record time, 23 months from the first meeting with the NRC to receipt of the 

certificate of compliance. The TMI-2 Core Shipping Technical Working Team that 

represented all the involved organizations met monthly. (299) The NRC certified 

this cask only after complete review of the application for certification. This 

review included the safety analysis report for the cask; data from drop tests using a 

1/4-scale cask model; data from drop tests of full-sized fuel and knockout 

Left: Model 125-B rail shipping cask under fabrication showing internal structure 

without inner containment shell. Right: Outer containment shell. (481.14)
90

canisters; and resolution of many design-related and test-related questions from 

the NRC. Before the application was submitted to the NRC for review and 

approval, it was subjected to one of the most intense reviews in the history of 

transporting radioactive materials at that time. Reviewers included personnel from 

DOE national laboratories, the licensee, and several subcontractors. The scrutiny 

and analysis expended on the application by the NRC were as thorough as the 

agency has given to any application for any rail, or truck, spent fuel cask. (300)

• Cask Tests Beyond Requirements. Computer modeling and analysis of certain 

structural design features of the TMI-2 cask system would have been adequate

without drop tests. However, the DOE conducted drop tests of a 1/4-scale model 

of the cask and a full-scale model of the knockout canister to verify performance 

during simulated accidents. Although component fabrication was expensive and 

Knockout canister inside a simulated shipping cask following a 

drop test. (481.14)
91

the testing costly, some estimates suggested that as much as a year was saved in 

acquiring certification. The drop test data and photographs were also helpful in 

reassuring the public that the TMI-2 transport system was safe. Conducting tests 

beyond the strict requirements of the regulating agency proved beneficial in 

licensing the equipment. (301)

• Integrated Cask and Equipment Tests. The DOE procured two new rail 

shipping casks, railcars, and some of the cask support equipment. The licensee 

procured cask loading and facility interface equipment. The integrated test of the 

cask loading equipment at the Hanford Engineering Development Laboratory was 

very successful in identifying interface problems and training operators. Insights 

from these tests included the following: (●) The integrated test was conducted at 

Hanford away from TMI-2, which allowed small anomalies in installation and 

checkout of the equipment to be corrected more easily and at less cost than 

3,000 miles away at TMI-2. (●) Several necessary mechanical and electrical 

modifications and equipment improvements were uncovered by the integrated test. 

Changes were engineered and implemented within hours and days rather than days 

or weeks, which would have been needed had the equipment been set up at TMI-2 

initially. (●) The test enabled many TMI-2 operators to gain first-hand knowledge 

of the equipment’s design and operation, including an understanding of the 

functional requirements, by direct discussions with shipping cask design 

engineers. This transfer of information was very valuable to the straightforward 

installation and use of the equipment at TMI-2. (●) The integrated test lasted a 

month, including initial system assembly, testing, disassembly, and packing for the 

shipment to TMI-2. The cost effectiveness of the integrated test was proven by the 

fact that equipment went from receipt at TMI-2 to NRC approval for use in less 

Model 125-B rail shipping cask 1/4-scale model following a drop 

test. (481.14)
92

than 2 months. (302, 303, 304)

• Number of Shipping Casks. In addition to the two shipping casks purchased 

by the DOE, the vendor built a third cask and leased it to the licensee to expedite

the shipments to INEL. However, delays encountered during defueling operations 

slowed the rate at which canisters were loaded. The casks were often on site 

awaiting the generation of enough canisters for a shipment. (305)

• Cask Lid Seal. The only cask-related delay in shipment occurred in 1988, 

when concerns over the O-ring material in the cask resulted in the material being 

changed to a different grade of elastomer. (306) The vendor identified a design test 

problem with the two O-rings used to seal the fuel shipment casks. The problem 

was that a design leak test of the cask with the existing O-ring material was not 

conducted at design cold temperatures (-40 degrees Fahrenheit). Design testing 

showed a potential for excess leakage using the existing Neoprene material at cold 

temperatures. (307, 308)

• Cask Improvements. As a result of the comprehensive inspection and 

preventive maintenance program initiated on the Model 125-B casks and railcars, 

numerous improvements were made to the operations, thus reducing the 

maintenance efforts. Table 4 identifies items that required more maintenance on 

the casks and railcars than anticipated and the corrective actions that were taken to 

remedy the situations. (309)

• Single Transportation Incident. The only rail shipment incident occurred in 

1987, when the train traveling at slow speed struck a car. The driver of the 

automobile sustained minor injuries and received a traffic citation from local 

authorities. The train’s engine received minor damage, but the shipping casks were 

not damaged. The carrier (Union Pacific Railroad) inspected the train’s engine and 

determined that it could remain in service. (310)

Rail shipment of three Model 125 B shipping casks 

containing defueling canisters. (481.14)
93

Table 4. Items of 125-B rail cask system requiring more maintenance than 

anticipated. (311)

Item Initial situation Improvement

Cask

Internal impact 

limiters

Thin stainless-steel sheet 

around honeycomb energy 

absorption media failed at the 

adhesive joint and resulted in 

constant cleanup and repair.

Replaced the thin sheet with 

thicker sheet and welded the 

sheet in place.

Internal impact 

limiters

Removing water from the 

cavity of the cask’s inner 

containment vessel required 

removal of the lower impact 

limiter.

A small-diameter tube was 

installed through the center of 

the lower impact limiter,

allowing removal of water by 

pumping.

Lanyards on pins

attached to the 

skid

Vinyl coating broke at crimp 

tie, allowing tie and coating to 

slide over cable, causing 

lanyard loop to open.

Replaced with uncoated 

stainless-steel cables.

Overpacks Difficult to install bolts because

they were heavy, long, and 

hard to maneuver into blind 

holes.

Added tapered lead-in collars 

around each bolt hole inside 

overpacks.

Railcar (a)

Excessive brake

shoe replacement

Pads on shoes cracked before

wearing out. Brake shoes were 

faulty.

Ordered new brake shoes from 

another manufacturer. Improved 

controls for releasing brakes.

Span bolster 

center bowl wear 

ring cracking

Wear ring and attachment weld

cracked.

Repairing and building up 

welds. Forged ring with 

machined press fit into center 

bowls was an alternative.

Tilt of railcar bed Lube disks were too hard and 

failed to compress. Motion 

from railcar movement caused 

disk to tear at center pinhole 

and ball up, causing bed to tilt.

Replaced by lube material 

melted into bowl.

Wheels Grade U wheels had excessive

tread wear.

Replaced with harder Grade C 

wheels.

a. Due to space limitations in the fuel handling building, the 125-B rail casks were not 

centrally located on the rail car; they were placed closer to one end of the car. This could

have contributed to the increased maintenance of the rail car. (312)

Figure on Next Page: Rover or remote reconnaissance vehicle (RRV) was used in 

the containment building’s basement to perform video and radiation surveys, 

collect sludge samples from the floor, collect core samples from the wall surface, 

flush walls with high-pressure water, remove the surface of the walls using an 

ultrahigh-pressure scarification system, and remove sludge.
94
95

6 DECONTAMINATION AND DOSE REDUCTION

The loss-of-coolant accident exposed about 7,200 square meters of concrete 

surfaces within the containment building to liquid- and vapor-phase contaminants. 

Most of those surfaces were protected by coatings of epoxy-based, nuclear-grade 

paints. Cesium and strontium were transported to containment building surfaces in

water droplets, and most of their measured surface activities were deposited after 

the nuclides had reached chemical equilibrium in the water in the basement. About 

2.4 million liters of reactor coolant were deposited in the containment building

basement during the accident. The water level eventually reached a depth of 

2.56 meters, and quantities of radionuclides were absorbed into the concrete walls 

of the basement. The principal radionuclides in the basement were 310,000 curies 

of cesium-137 and 11,000 curies of strontium-90. (313)

The understanding of decontamination techniques and methods has improved 

significantly since the TMI-2 accident. Nevertheless, the TMI-2 cleanup provided 

a unique experience in the context of a severe accident.

The following experiences from decontamination and dose reduction activities

include topics covering management planning; dose reduction; decontamination;

radioactivity penetration in concrete and paint; and recontamination problems.

6.1 Management and Planning

The cleanup strategy was to first decontaminate as much as possible, referred to as 

“gross” decontamination. Presumably, the thought process was that first reducing 

dose rates throughout would lead to conducting work more easily later. With this 

strategy, decontamination became the major priority during the initial stage of the 

cleanup. After poor results from the gross contamination experiment inside the 

containment building, the priority was changed to decontamination on a 

case-by-case basis to support priority tasks for moving forward with defueling 

activities.

• Gross Decontamination Experiment. The DOE funded the gross 

decontamination experiment in 1982 as part of its research and development 

charter to provide the industry with access to decontamination engineering and 

operational experience. In addition, the experiment documented effectiveness;

criteria; techniques; and radiation monitoring activities. Key conclusions included 

the following: (●) Results were often inconclusive because not all the data for preand post-activity characterization were collected; some collected data were

insufficient or not timely enough to accurately record the effects. (●) Radiation 

from other sources often masked the contamination reductions achieved during 

decontamination. (●) Knowledge about decontamination effectiveness had to be 

extrapolated or estimated with less precision than desirable. (●) Tested techniques 

were effective to varying degrees in removing surface contamination. (●) Some 

areas were recontaminated because of technique or procedure (see Section 6.5 of 

this document). (●) Organizational inefficiencies were revealed, including a 
96

complex review and approval process, as well as redundancy, duplication, and 

overlapping responsibilities in site procedures. (●) Knowledge about large-scale 

operations in the containment building was obtained, such as the movement of 

large numbers of workers in and out of the containment building, radioactive 

waste disposal, and communications and procedures for conducting 

in-containment work. (●) Decontamination training required rigorous 

standards. (314, 315, 316)

• Decontamination Sequence. The proposed decontamination sequence, as 

described in the gross decontamination experiment report, included: (1) remove all 

the storage, trash, and waste; (2) decontaminate vertical and overhead surfaces, 

especially those surfaces that had not been decontaminated previously;

(3) decontaminate floor surfaces; and (4) locate hot spots and shield, clean, or 

remove the source. (317)

• Plans and Priorities. Following the completion of the gross decontamination 

experiment in the containment building (October 1981 to March 1982), the 

licensee proposed several decontamination and dose reduction plans to the NRC 

for approval. The key plans included the following:

o Reduce Respiratory Protection. The decontamination efforts in 

September 1982 were directed toward the relaxation of respiratory protection 

and protective clothing requirements in the containment building, as 

appropriate. The objective was to increase worker efficiency by economizing 

person-rem resources, minimizing heat stress, and reducing waste volumes. In 

addition, expectations were that some reduction in area dose rates would be 

realized to the extent that superficial contaminants contributed to general area 

dose rates. The planned decontamination efforts included all levels of the 

containment building except the basement level. (318)

o Dose Reduction Task Force. In early 1983, the recovery team realized that the

effectiveness of dose reduction by means of gross decontamination techniques 

proved disappointing. The licensee formed the Dose Reduction Task Force to 

perform extensive radiological characterization to (●) identify the radiation 

sources that produced general area dose rates and (●) recommend actions to 

reduce dose rates from those sources in areas of intensive worker activities. 

(319)

o Strategic Decontamination and Dose Reduction. (320, 321, 322) In mid-1984, the 

licensee’s program for cleanup of the TMI-2 containment building, as 

presented in its TMI-2 Program Strategy plan (323) and the first supplement to 

NUREG-0683, the NRC’s PEIS, changed the priority from a focus on gross 

decontamination in the containment building to strategic decontamination and 

dose reduction. The strategic objectives were twofold: (●) reduce the dose to 

workers in transit to and from their workstations on the 347-foot elevation 

(operating level) and defueling work platform and (●) reduce or eliminate the 

need for respiratory protection for routine reactor building entries. This new 
97

focus resulted in the greatest reduction in personnel exposures during defueling 

preparations and operations. The revised cleanup program entailed the same 

milestones as the initial schedule, but the sequence of tasks was altered as 

follows: (1) dose reduction to continue during reactor disassembly; (2) reactor 

disassembly and defueling; (3) defueling; (4) primary system decontamination; 

and (5) cleanup of the containment building, including equipment, and 

auxiliary building to proceed as resources allowed.

Reducing general area dose rates was achieved by installing shielding for 

specific sources of radiation including the enclosed stairwell and the floor 

hatch on the 305-foot elevation (entry level), and the aircoolers, floor drains, 

the open stairwell, and the seal table on the 347-foot elevation. Removal of 

fixed contamination on floors also contributed to reducing general area dose 

rates. 

Initial efforts at dose reduction proved successful. The goals were achieved 

and, in some cases exceeded, at reducing: (●) the transit dose for personnel 

from air lock to work areas on the 347-foot elevation and returning to air lock 

from 0.4 to 0.18 millisievert (0.04 to 0.018 rem) and (●) the general area 

radiation levels on the 305-foot elevation from 3.5 to 2 millisievert/hour.

By attaining these early goals, the collective radiation does to workers was 

significantly reduced in the labor-intensive construction efforts in preparing for 

reactor vessel head lift and defueling. Later improvements in airborne 

contamination by decontamination of air-cooling units permitted manned 

access without respirators, allowing more effective use of limited personnel. 

Dose reduction efforts continued throughout the defueling period.

o Support Reactor Disassembly and Defueling. The TMI-2 Program Strategy 

plan (324) recognized that some decontamination activities would be required in 

the containment building during reactor disassembly and defueling to maintain 

surface and airborne radioactivity at acceptable levels. Additionally, the 

strategy recognized that some decontamination activities in the auxiliary 

building would be necessary to permit access for required safety surveillance 

operations and maintenance. Beyond these, incidental decontamination 

activities, such as small-scale demonstrations to support post-defueling 

activities, were to be conducted on a low-priority, noninterference basis. (325)

o Support Long-Term Storage. (326) The decontamination task force report (327)

, 

issued in December 1985, was prepared by a joint group of TMI-2 organization 

representatives to review the effort required to decontaminate TMI-2 and to 

evaluate the reduction in occupational exposure during post-defueling 

monitored storage. The report was prepared by a task force formed in 1985 to 

evaluate the problems and activities associated with achieving the final 

radiological completion criteria by mid-1985. The objective of the task force 

was to arrive at a consensus in the technical approach to each of the major 

areas of decontamination work in the containment. The major areas evaluated 
98

by the task force included: (●) remote equipment development; (●) sludge 

transfer and disposal; (●) steam generator D-ring shielding dose reduction and 

decontamination; (●) containment building basement recovery; (●) auxiliary 

and fuel handling building surface decontamination; (●) non-reactor coolant 

system decontamination; (●) containment building heating, ventilation, and air 

conditioning system modifications; (●) containment building Phase III (final) 

surface decontamination; (●) reactor coolant system decontamination; and 

(●) containment building Phase III decontamination waste management. The 

task force based its evaluations, technical approaches, and schedules on 

available technical plans.

The task force concluded that deferring decontamination for a period of 

30 years would result in a potential occupational exposure savings in the range 

of 4,500 to 9,800 person-rem. This savings was based, in part, on reduction in 

radiation dose rates due to the natural decay of radioactive materials and to 

advances in both remote cleanup technology and chemical decontamination 

methods.

• Experienced Workers, Simple Tools. Decontamination was labor intensive

and required a range of skills. A mixture of skills and hands-on supervision by the 

licensee contributed to an effective range of capabilities. Simple, traditional tools 

used by an experienced work force contributed to the effectiveness of the 

decontamination work. The workforce provided feedback based on experience to 

improve existing tools, including the potential use of more sophisticated 

techniques when traditional tools were not effective in unique situations. Often, an 

improvement was a traditional tool mounted on a sophisticated delivery system. 

Feedback was important for every aspect of the recovery effort. (328, 329)

• Flexible Plans. Planning the general approach was important; however, the 

plan needed to be flexible. Problems were approached in manageable segments. 

The specific task planning horizon at TMI-2 was 1 to 2 weeks, with changes 

occurring almost daily, based on the previous day’s progress and events and new 

information. Too much planning for hypothetical situations consumed already 

limited resources. Accurate, timely data about conditions were vital for realistic 

planning. Analysis needed to be based on experience and hard data, not estimates; 

therefore, it was often necessary to start work and learn. (330)

• Plant Knowledge. Knowledgeable personnel and accurate, as-built plant 

drawings were required for planning, efficient operations, and the effective use of 

resources. Knowledge of the TMI-2 plant was important to examine all 

interrelated aspects of decontamination operations, such as systems, cubicles, 

logistics of movement to avoid recontamination, and control of ventilation 

flowpaths to prevent recontamination. (331)

• Research vs. Operations. Research and development needs often conflicted 

with decontamination activities due to time and resource constraints. On occasion, 

adequate baseline data were not obtained before completing a decontamination 
99

activity. Consequently, research and development could only produce marginally 

useful data to measure the effectiveness of a decontamination project. (332)

• Support Facilities. New support facilities were required early in the 

decontamination program. Important facilities at TMI-2 included: (●) a changeout

facility, called the coordinated personal access facility, for reducing skin 

contaminations while undressing as well as reducing time spent in entering and 

exiting radiation areas; (●) an onsite laundry and respirator cleaning facility;

(●) small tool decontamination facility for reducing the number of tools needed,

which decreased expenses, waste volume, and storage inside the containment 

building; and (●) the remote coordination center (located in the turbine building) 

for directing decontamination and defueling activities inside the containment 

building. (333, 334)

• Training and Staffing. The gross decontamination experiment report (335)

concluded that training must be completed to rigorous standards. Some training 

deficiencies occurred due to equipment shortages for training and poor instruction. 

During a few of the later entries, training occurred just before the scheduled entry. 

This resulted in no time for a review of training effectiveness. This schedule left 

the discovery of deficiencies to the task supervisor during the actual entry. An 

insufficient number of backup personnel were trained; as a result, considerable 

scheduling flexibility was lost. The number of crew changes for sickness, 

absenteeism, accidents, and other incidents for each entry was underestimated. 

One backup for every four crew members (laborers and crafts) performing 

decontamination work was shown to be optimal. (336)

The remote coordination center located in the TMI-2 turbine 

building contained the control functions for entry into the 

containment building with positions for safety; radiological 

engineering; operations; entry coordinator; and command center 

management. (481.15)
100

• Technique Selection. The selection of decontamination techniques was based 

on the following attributes: (●) effectiveness; (●) resources; (●) time; (●) material

compatibility with cleanup systems; (●) radioactive waste forms and constituents 

generated and associated disposability and disposal costs; and (●) personnel 

radiation exposure. (337, 338)

• Importance of Frequent Surveys. Frequent surveys of the decontaminated 

areas were essential to tracking the progress made to reduce contamination and to 

identify recontamination sources. Though difficult, monitoring of surface 

contamination levels during the decontamination efforts by surveying surfaces 

during brief interruptions of the operation improved the effectiveness of 

decontamination. (339)

• Sample Analyses vs. Contact Surveys. Cores from concrete surfaces at the 

containment building upper levels were extracted to determine the activity 

concentration on those surfaces. Results suggested that measurements of the 

contact beta and gamma exposure rates of a surface could be used instead of 

collecting samples from the surface. Because of the requirement that the survey 

instruments used for this purpose be highly collimated, the use of beta exposure 

rates was likely to be a more reliable method than one that used gamma exposure 

rates. (340)

• Radiological Survey Data. (341, 342) Rigorous evaluation of the effectiveness of 

decontamination techniques required radiological survey data and a large database 

for source term characterization. Constraints on data acquisition included the 

following: 

o Factors that influenced the collection of radiological survey data included: 

(●) self-shielding effects of the survey technicians with different physical 

stature; (●) instrument variations causing different readings; (●) ongoing 

operations causing dynamic radiation fields in a specific area; (●) surveys not 

organized to systematically and precisely define the radiation field; 

(●) inconsistency of the surveyors; (●) and other varying factors with 

individual instruments creating uncertainties.

o Problems in establishing a rigorous database included: (●) heterogeneity of 

containment building surfaces and coatings; (●) irregular geometry of 

equipment surfaces; (●) heterogeneity of the contamination pathway and 

resulting deposition patterns; (●) environment of multiple sources;

(●) changing patterns of resuspension and deposition; and (●) high relative 

humidity. 

o The costs in resources and person-rem to collect and analyze samples and to 

interpret the data added constraints on the ability to statistically compare 

decontamination techniques. 
101

• Radiological Survey Using Thermoluminescent Dosimeters. To resolve the 

problems associated with radiological survey data collection, strings of 

thermoluminescent dosimeters (TLDs) were used to characterize radiation profiles 

inside the containment building. The advantages included precise spacing and 

orientation of data collection points, reproducible measurements, and ability to use

TLDs in high-radiation areas and through small openings (e.g., the containment 

building basement). The subsequent source term modeling of the containment 

building proved generally accurate and permitted later dose reduction efforts to be 

effective. (343)

• Communications. Operational problems with radios (e.g., too few radios, 

inoperative equipment, poor transmission quality) caused a significant increase in 

the number of person-hours spent in the containment building and decreased the 

effectiveness of the decontamination tasks and data acquisition. (344)

Locations of thermoluminescent dosimeter (TLD) exposure rate readings at the 

282-foot elevation (basement level) inside the containment building. Dosimeter data 

were obtained by lowering TLD strings at various points to define principal radiation 

sources such as the bathtub ring and the elevator shaft.
102

• Onsite Radiochemical Laboratory. (345, 346, 347) The inability of the onsite 

laboratory facilities to accommodate high-activity samples proved to be a 

hindrance to efficient operation in terms of turnaround time and ALARA 

considerations. The DOE placed two mobile laboratories on site to reduce the time 

and capital expenditures required to analyze samples at TMI-2. The combined 

capabilities of the two laboratories greatly enhanced onsite analytical capabilities 

to analyze samples of fission products, fuel, transuranic, and elemental core 

debris:

o Radiochemistry Laboratory. This laboratory, located in the TMI-2 fuel 

handling building, received and handled solid and liquid samples with 

activities up to 5 roentgen per hour. This laboratory, which had a sample 

hood, glove box, and analytical chemistry facilities, performed radiochemical 

separations and preparation of alpha, beta, gamma, and x-ray emitting 

radionuclides for quantification by instrumental analysis. In addition, the 

laboratory performed traditional wet chemical elemental analysis.

o Counting Laboratory. This laboratory, located in the Unit 2 turbine building, 

received alpha, beta, gamma, x-ray, and elemental samples produced by the 

chemical separations in the radiochemistry laboratory. The counting 

laboratory handled very small quantities of material with activities of less than 

1 milliroentgen per hour. The samples were analyzed instrumentally by 

spectroscopic techniques to produce quantitative analysis data.

Pre-accident construction-related equipment and materials stored inside the auxiliary 

building (top) containment building (bottom) became sources of radiation hot spots

that complicated decontamination.
103

6.2 Dose Reduction

• Dose Reduction Program. The effectiveness of dose reduction techniques in 

the containment building during the first 2 years of decontamination activities was 

reported to be disappointing. Dose reduction in the containment building proved to 

be a much more formidable task than in the auxiliary building. Source 

identification and characterization were difficult. The relatively congested physical 

arrangement of machinery, components, pipes, cable trays, and structural surfaces, 

all with potentially unique distributions of fission products, made the task of 

quantifying the discrete contributors to the overall high ambient radiation levels a 

complex puzzle. Gamma dose rates in the containment building remained at 

elevated levels despite substantial progress in gross decontamination and 

processing of contaminated water from the basement. Recognizing that the 

decontamination efforts would likely not result in significant reductions in general 

area radiation levels in the building, the licensee launched a dose reduction 

program at the end of 1982. (348, 349)

• Dose Reduction vs. Decontamination. The dose reduction program 

emphasized dose reduction rather than decontamination in the containment 

building. The program identified radiological sources that were classified as 

principal or discrete; modeled the source and estimated its contribution to the

general area exposure rates; and recommended actions to achieve dose reductions 

consistent with general exposure rate goals. Principal sources were physically 

large and had a major general area dose effect, such as the containment building

basement, the enclosed stairwell, and the air coolers. Discrete sources were 

movable or had localized effects, such as floor drains, abandoned equipment, and 

trash. A list of source terms arranged in order of their contribution to the ambient 

radiation levels was essential in developing an efficient dose reduction 

strategy. (350)

• Radiation Sources. (351) The dose reduction study classified radiological 

sources in the containment building as principal sources and discrete sources. The 

study also listed the major contributions of these sources. Sources and contributors 

are discussed below:

o Principal Sources. Major principal contributors to gamma exposure in the 

containment building included the following: (●) The outer containment 

building walls in the basement consisted of a ring of contaminants, (i) located 

8 feet from the floor. The ring on the poured concrete walls was caused by 

absorption into a 30-inch band of uncoated concrete 5.5 feet above the floor. 

Walls that were unpainted showed a marked increase in source strength over 

i The ring was residue as water from the basement was processed and resulted in lowering 

the level in the containment basement. This ring was the major contributor to dose rates 

within the upper containment. This was one reason for having the defueling platform 

inside the defueling water canal (without water) where the walls of the canal provided 

shielding for defueling workers.
104

painted surfaces. (●) The basement floor area consisted of sludge and water, 

both of which were assumed to be uniformly distributed over the floor. This 

slab source resulted in radiation fields that were as high as 1,000 roentgen per 

hour and made this area inaccessible to workers. (●) Reactor coolant system 

water was continually dissolving cesium. The reactor coolant system required 

periodic processing to remove radionuclides leaching from the fuel at a slow 

rate of 0.5 curie per day. (●) Areas of surface contamination included the 

containment building dome; polar crane; upper level (refueling deck) floor 

and walls; and the entry level floor, overhead, and walls. 

o Discrete Sources. Major discrete contributors to gamma exposure in the 

containment building included the following: (●) About 30 floor drains

contained fixed contamination and presented high local area gamma exposure 

rates. (●) Trash consisted of materials that were in the containment building at 

the time of the accident. Trash was usually low-level waste; however, it 

masked gamma surveys and could mask other sources.

(●) loss-of-coolant-accident ducts (both run up the wall of the containment 

building) were sources of exposure to personnel climbing up to the polar 

crane. (●) A resin column used to test the submerged demineralizer system 

ion exchange mixture had a contact dose rate of 12 roentgen (R) per hour.

(●) Two welding machines that were in the containment building at the time 

of the accident had contact does rates of 2 R per hour. (●) Both discharge 

lines from the core flood tanks had dose rates of 2 R per hour at elbows and 

joints. (●) The in-core instrument service area (seal table) on the defueling 

deck had boron accumulation resulting in dose rates of 10 R per hour.

(●) Polar crane components were not decontaminated due to their sensitivity 

to decontamination agents or techniques and so had high contact dose rates.

• Dose Reduction Techniques. The decontamination and dose reduction efforts

focused on tasks to minimize cumulative worker exposures through the end of the 

cleanup. Areas most frequently accessed by many workers were given the highest

priority. Dose reduction solutions included: (●) keeping inner and outer personnel 

airlock doors open (with NRC approval) to reduce time to stage equipment and 

materials; (●) changing entry and transit routes to new ones through lower dose 

fields; (●) restaging the tool storage cabinet to a lower dose area; (●) removing

trash and welding machines; (●) eluting and removing the submerged 

demineralizer system resin test column; (●) flushing areas and equipment, such as 

air coolers, loss-of-coolant-accident (LOCA) ducts, the seal table, D-ring interiors 

and equipment surfaces, the enclosed stairwell, and the elevator shaft; 

(●) decontaminating the seal table, reactor vessel head service structure, and walls 

of the refueling canal; (●) applying shielding, as necessary, to air coolers, the seal 

table, core flood tank discharge lines, hatch covers, LOCA ducts, the head service 

structure, the open stairwell, polar crane components, and cable trays; and 

(●) conducting extensive scabbling operations on concrete surfaces. (352, 353)

• Use of Video Technology. The TMI-2 cleanup effort pioneered the use of 

video camera technology for surveillance and inspection in nuclear power plants. 
105

The remote coordination center, located in the TMI-2 turbine building, used many 

cameras to survey 75 percent of the containment building with remote pan, tilt,

and zoom capabilities. The cameras proved extremely valuable to task 

management and personnel safety by allowing supervisory work guidance without 

the supervisor being in the area. This resulted in significant personnel dose savings 

in the early years when radiation levels were considerably higher before

decontamination and dose reduction planning took effect. (354)

• Use of Computer-Aided Design Models. An effort was undertaken to create a 

three-dimensional (3-D) computer-aided design model of in-containment systems 

and structures from the design drawings. This was the early days for such 

modeling. The value was that viewing the 3-D displays served to better understand 

the positions of structures, systems, and components within the containment 

building. For example, the 3-D model showed the existence in the containment 

building basement of two previously unknown small lubricating oil tanks for the 

main coolant pumps.

• Equipment and Materials Control. The amount of materials and equipment 

brought into contaminated areas was controlled to prevent interference with work 

and unnecessary contamination. Every effort should be made to use the shielding 

already available in the containment building. Although shielding lowered the 

initial person-rem-per-dollar cost, the overall cost increased when workers 

eventually returned to clean up a shielded area. (355, 356)

• Piping Access. The limited number of piping system access points, such as 

high-point vent or low-point drain valves, resulted in additional time spent in 

radiation areas to remove the valve internals or “rig” an opening in a pipe. (357)

Lead shielding around the enclosed stairwell at the 305-foot elevation (entry 

level) inside the containment building.
106

Need established 

for surface 

decontamination

Are surfaces 

equipment 

surfaces?

Are surfaces 

vertical?

Is 

surface diamond 

plate?

Is 

surface stainless 

steel?

Is surface 

covered with 

loose debris?

HP spin jet

6000 psi

15 gpm

ambient temp.

Mechanical 

scrubbing 

chemicals

Establish need 

for contamination 

control measures

HP spray

3 – 4000 psi

7 – 12 gpm

ambient temp.

HP spin jet

6000 psi

15 gpm

ambient temp.

LP flush

1 – 2000 psi

20 – 25 gpm

ambient temp.

Is 

surface 

stainless 

steel?

LP flush

1 – 2000 psi

20 – 25 gpm

ambient temp.

HP spray

3 – 4000 psi

7 – 12 gpm

ambient temp.

LP flush

1 – 2000 psi

20 – 25 gpm

ambient temp.

Abrasive 

technique

Yes

Yes

Yes

Yes Yes

No

No

No

No

No

No

Yes

Decontamination logic diagram based on results of the gross decontamination 

experiment in March 1982. (481.15)

(Abbreviations: gallons per minute---gpm; high pressure---HP; low pressure---LP;

pounds per inch---psi; temperature---temp)
107

6.3 Decontamination Methods and Agents

The decontamination logic diagram summarizes the conclusions on

decontamination effectiveness and efficiency from the gross decontamination 

experiment. Table 5 is the TMI-2 decontamination method reference table. Other 

experiences are summarized below.

• Water Flushing Techniques. High-pressure flushing was generally as 

effective as both low- and high-pressure flushing used in sequence. However, if 

the surface debris was loose, the low-pressure flush was used first because it was 

more controllable and less likely to disperse particulates. Experience from the 

gross decontamination experiment revealed the following insights: (●) Hydrolance 

operations were limited to 3,000 pounds per square inch (psi) during the gross 

decontamination experiment because of the inability of workers to maintain their 

footing while spraying. However, large workmen (in the 250- to 270-pound range) 

could satisfactorily spray with 6,000 to 7,000 psi. (●) High flow rates with high 

pressures caused problems on horizontal surfaces because of the inability of 

operators to properly direct the water to a drain and the attendant mist that 

recontaminated cleaned surfaces. (●) The lances, as shipped from the supplier, 

were deemed too short by the site safety office because of the potential for the 

lance operators to shoot their foot during high-pressure flushing. The short-tip 

lances were replaced with longer tip lances. (●) High-pressure spray activity with 

the lance and water floor scrubber caused contamination of other surfaces. Thus, 

the experiment report concluded that procedures should provide for an adequate 

flush after high-pressure activity and that the pressures should be sufficiently 

constrained to reduce contamination overspray and debris splatter. (358, 359, 360)

• No Water Temperature Effect. Surface decontamination effectiveness was not 

significantly altered by the flushing water temperature, despite other studies 

Decontamination using a high-pressure hydrolance.
108

indicating hot water to be superior. (361, 362)

• Scabbling. Scabbling was an aggressive decontamination technique that

roughened concrete surfaces and removed surface coatings with toothed pistons or 

a rotating drum. Testing conducted in 1984 evaluated the effectiveness of 

removing the surface coating from areas on the upper 347-foot elevation 

(operating level) in the containment building using a scabbler machine. The 

scabbling operation was performed in two passes, each pass removing 

approximately 1/16-inch from the surface. Radiation measurements showed 

approximate dose rate reductions of 50–60 percent gamma and 80–98 percent

beta. The resultant surface was suitable for application of epoxy, polymer, or 

similar finishes. The scabbler was modified to include a vacuum shroud connected 

to a high-efficiency particulate air-filtered vacuum system for collecting 

contaminated dust. Approximately 930 cubic meters of the containment floor 

coatings were removed, and the cleaned floors were recoated. (363)

• Remote Scabbler. A remotely operated scabbler was designed for use at 

TMI-2, although it was developed too late to be used. The device was successfully 

tested in other decontamination applications. The scabbler could be operated from 

50 feet away. It could cut a path 18 inches wide and could decontaminate concrete 

surfaces at a rate of 400 square feet per hour, about three times faster than manual 

scabbling. A self-contained vacuum system equipped with a high-efficiency 

particulate air filter eliminated dusting. (364, 365)

• Vacuum System. A high-efficiency vacuum collection system was designed to 

attach to a variety of devices, including manual scabblers. The system removed

residual loose debris on the cleaned surface. Operators did not contact the 

contaminated material because the design of the system permitted the exchange of 

full waste drums with empty ones while the vacuum system was operating. (366)

Underside of a typical hand-pushed scabbler machine with vacuum used to roughen 

concrete surfaces and remove contamination in the surface coatings or embedded deep 

into the concrete surface.
109

• Steam/Vacuum System Performance. A modified commercial steam/vacuum 

system was effective in decontaminating overhead areas and cable trays, which 

contained too many complex surfaces to clean effectively using flushes or hand 

wiping. The steam supply and vacuum removal system was an integrated unit that 

collected water and contaminants simultaneously. This type of system required 

fewer operators than separate water flush and vacuum systems and reduced the 

potential for the spread of contamination. Recontamination was reduced by 

collecting water and filtering it to remove radionuclides before it was reused. The 

powerful vacuum capability of the combined unit allowed the decontamination 

tool to be located as far as 175 feet from the steam supply and collection systems. 

Thus, only the decontamination tool and hoses were in the contamination zone. 

One problem encountered with this system was blockage by sheet materials that

were drawn into the vacuum inlet. (367, 368, 369)

• Chemical Additives for Decontamination. (370, 371, 372, 373) Chemical 

decontamination agents, such as phosphoric acid foam, sulfamic acid gel, and 

citric acid complexes, were evaluated at TMI-2 to improve the performance of 

decontamination systems. Small-scale use of specific chemicals was subject to 

evaluation, and chemicals were controlled, used, and disposed of in accordance 

with appropriate procedures and applicable Federal, State, and local regulations.

o Advantage. The biggest advantage of the use of chemical treatments was that 

they could be used in those areas where flushing or abrasive techniques could 

not be used, such as external surfaces of pipes and tanks and system internals. 

Plot of decontamination factor (based on contact measurements) 

verses water rate. (481.15) 
110

o Disadvantages. The disadvantages of chemical treatments included the 

following: (●) Techniques and reagents needed to be compatible with reactor 

coolant system component materials to avoid excessive degradation of such 

components. (●) A relatively large volume of chemical had to be applied to an 

expansive surface area or intricate piping system to provide coverage at the 

optimum concentration. (●) Chemical waste had to be treated and processed 

for disposal in a manner that followed regulations.

• Recontamination Protection. The strippable coating that was used after initial 

decontamination by water spraying adsorbed residual loose debris. The coating 

also protected against recontamination of the underlying surface. (374, 375)

• Strippable Coatings. (376, 377, 378) The strippable coating was a synthetic 

polymer compound that was used to remove oxides and radioactive contamination 

from oxidized ferrous and nonferrous surfaces and from nonporous surfaces. 

Experiences from the use of strippable coatings included the following:

o Coating Application. The coating was applied by an airless spray gun, allowed

to dry, and then removed manually by cutting, peeling, and rolling the material 

up to contain the contamination. A 55-gallon waste disposal drum 

accommodated the refuse from 2,000 to 2,500 square feet of coated surface. 

Successful application and removal of strippable coatings were dependent on 

operator experience. Thus, dedicating a trained group of workers would be cost 

effective.

o Coating Removal. Workers at TMI-2 had some difficulty in removing coatings 

from highly porous surfaces. A cheesecloth base was tested to determine the 

tensile strength and body of the coating to make removal easier. The base did 

make the coating come off more completely, but it was unnecessary for most 

applications. The self-stripping coating could not be removed easily; too much 

effort was spent removing the coating manually. However, self-stripping 

coatings were useful in high radiation areas because the coating could be 

applied and removed by remote vehicles.

o Coating Effectiveness. The copolymer coating produced decontamination 

factors of up to 245. Based on experimental data on carbon steel, the surface 

condition after using the copolymer coating for oxide removal was comparable 

to that achieved by commercial blast cleaning. The rate of application was 

comparable to that of high-pressure flushing with vacuuming to pick up water. 

The application and removal times for the strippable coating were also 

comparable to the high-pressure flush with vacuuming. Low-pressure flushing 

plus strippable coating produced substantial reductions and decontamination 

factors in area readings, contact readings, and smearable contamination results.

o Recontamination Protection. The strippable coating that was used after initial 

decontamination by water spraying adsorbed residual loose debris. The coating 

also protected against recontamination of the underlying surface.
111

• Vertical vs. Horizontal Surfaces. The decontamination of vertical surfaces 

proceeded somewhat differently from that of horizontal surfaces. Vertical surfaces

generally did not have debris embedded in their surfaces as floors did. Beyond 

that, decontamination factors observed on vertical surfaces were also higher than 

those observed on some horizontal surfaces. Experience showed that water volume 

per unit area required for run off was considerably less than that for horizontal 

surfaces. Additionally, as a vertical surface was flushed from top to bottom, the 

buildup of water flowing down the surface increased the volume per surface area 

and was effective in the elution of solubles and the transport of particulates. Such 

elution and transport were not so readily achieved on horizontal surfaces. (379)

• Use of Containment Building Spray. The use of the containment emergency

spray system to spray down large volumes of water, detergents, chemicals, or

steam on the structures, systems, and components inside the containment building

was not pursued. This approach was not necessary, mainly because lower than 

predicted radiation fields were found in the containment building as the result of 

2 years of dripping condensation (rain) caused by high humidity. The effectiveness 

was uncertain, and the spray would have generated a large volume of waste 

(1,500 gallons per minute pump flow rate). In addition, the spray down could 

cause damage to important equipment and instrumentation. (380)

• Use of Remote Control Equipment (Robotics). An important experience in the 

decontamination area was the use of robotics. Starting with simple commercial 

devices and borrowed DOE robots, TMI-2 developed unique robots designed for 

specialized accident recovery tasks. These tasks included radiation measurements;

video camera inspections; data acquisition; sediment sample acquisition;

Strippable coating used to decontaminate floor surfaces. (481.15)
112

high-pressure water flushing; acquisition of concrete core samples; concrete 

scabbling and scarification; sludge vacuuming; and debris pickup and removal. 

Remote-controlled robotic vehicles and supporting control equipment were used 

extensively to perform work in extremely high radiation areas, including the 

containment building basement, the makeup demineralizer room in the auxiliary 

building, and the reactor coolant pump seal injection valve room in the fuel 

handling building. Robotics research at TMI-2 produced innovations in robot 

design, deployment, and operator training. (381, 382)

The use of robots at TMI-2 did not require any special NRC licensing reviews; 

however, activities in which robots were used, like most recovery and cleanup 

activities, required safety evaluations for NRC review and approval. (383)

• Robot Maintenance. Based on experience, it was reported that robotic designs 

should be considered for reducing the contamination of internal parts. The design 

should also provide for easy maintenance by workers wearing anticontamination 

clothing with multiple sets of rubber gloves and respiratory protection. (384)

6.4 Radioactive Contamination Penetration

Radioactive contaminants that penetrated surfaces inside the containment building

included concrete floor coatings (paint), basement walls, and cinder blocks

(elevator shaft, stairwell). Contamination in the cork seam of expansion joints was 

discovered in the control and service building clean areas (adjacent to the fuel 

handling building).

• Concrete Coatings (Upper Levels). Protective coatings over concrete surfaces 

provided significant protection against radionuclide penetration. Analysis of 

concrete core samples from the entry and operating levels (305-foot and 347-foot

elevations, respectively) inside the containment building in 1984 indicated that

most of the radionuclides released from the reactor coolant system into the 

containment building environment were trapped in the concrete’s surface coating. 

Cesium generally was confined to within a few millimeters of the top surface of 

the coating layer. However, at floor locations that had coatings damaged before

the accident, cesium penetrated the subsurface concrete to a depth of several 

centimeters. A breach or partial breach in the coating layer combined with 

long-term pooling of contaminated water at the location of the damage was 

thought to be the cause of this phenomenon. The outward spread of subsurface 

contamination at such locations appeared to have been confined to the near 

vicinity of the site of the damage. The removal of the coating from the samples 

revealed that up to 99 percent of the total measured activity was removed with the 

coating. (385)

• Concrete Coatings (Basement). The quantity of radionuclide deposition in the 

containment building concrete was affected by the type and density of the 

concrete. Analysis of concrete core samples from containment building basement 

surfaces in 1987 indicated that much of the radionuclide content of the unpainted 
113

and painted concrete cores was deposited near the sample surface exposed to the 

accident water. Paint appeared to substantially restrict the uptake of fission 

products by the concrete. For the high-strength concretes (3,000 and 5,000 pounds 

per square inch), radionuclide penetration was limited to about 1.0 centimeters, 

with about 90 percent of the activity located in the first 0.5 centimeter. The data 

suggested that the average penetration into the unpainted concrete was slow 

(approximately 0.1 centimeter per month) during the approximately 9-month 

exposure to accident water and that much of the activity was probably deposited in 

the first day after the material was submerged in accident water. (386)

• Cinder Block Absorption. Cinder blocks were used in the containment 

building as walls for the elevator shaft and one of two stairwells; these blocks 

were only coated to 1.68 meters (5.5 feet) above the basement floor. The concrete 

block was considerably more porous that the other concrete structures, and 

significant amounts of cesium-137 were transported through the block. Radiation 

levels in the porous concrete block were the highest in the basement area,

exceeding 1,000 roentgen per hour, and were a major obstacle to basement 

cleanup. The blocks were only submerged in coolant for a period of about 

10 months; however, capillary action in the concrete tended to carry water into the 

concrete some distance above the surface of the free-standing water level. (387)

• Leaching Tests. Tests determined the leachability of the radionuclides 

absorbed into the concrete. The tests included three core concrete samples from 

the containment building basement and an in situ leaching test of concrete blocks 

surrounding the elevator shaft. Although the tests proved the potential of leaching

as a decontamination method, other programmatic considerations ruled against

reflooding the basement, such as schedule, water processing concerns, and an 

emphasis on defueling rather than decontamination. (388)

o Concrete Core Samples. The results of 4-month leach tests of concrete core 

samples from the containment building basement showed that the concrete 

could be leached of cesium and strontium. These results suggested that 

leaching could be an alternative to the mechanical removal of 

surface-deposited or absorbed radionuclides, or both, from concrete surfaces 

that had been submerged in reactor coolant for a period of time. Leaching 

cesium and strontium from containment building surfaces was significant, as 

up to 93 percent of the strontium and up to 78 percent of the cesium could be 

removed by leaching over the 4-month period. Also, the data suggested that 

longer leach periods with fresh coolant would desorb, or leach, additional 

quantities of fission products from the concrete. The leach rate for the 

basement block wall core sample was lower than that for the 3,000-pounds per 

square inch (psi) concrete wall because of the lower diffusion length for 

isotopes through the 3,000-psi concrete. (389)

o Concrete Block Test. The licensee conducted an in situ leaching test to reduce 

the radiation levels from the concrete block walls surrounding the elevator in 

the containment building. The walls were porous, hollow concrete block. This 
114

test consisted of injecting a high volume of water under low pressure into 

penetrations in the block wall to maintain a water level as high as possible 

within the block wall. The fill-and-leach operation reduced radionuclide 

content by 33 percent in the treated areas and removed 1,200 curies of 

cesium-137, or approximately 7.1 percent of the total block wall 

radioactivity. (390)

• Floor Expansion Joint Contamination. In late 1980, contamination in the 

cork seam of expansion joints was first discovered during a routine radiation 

survey in the control and service building area (adjacent to the auxiliary and fuel 

handling building). The seam was a cork-filled construction joint located between 

major structures to accommodate differential expansion between building 

structures and to attenuate vibration and wave motions during a seismic event. 

During the period following the accident, the cork seam located in the auxiliary 

and fuel handling building seal injection valve room was saturated with reactor 

coolant water due to leaking valves. Initial decontamination attempts were not 

successful. Over the years, the radioactive material had spread along the joint into 

noncontaminated areas inside the plant. However, the radioactive contamination 

was prevented from entering the ground water table by a water stop barrier

imbedded in the floor. Modifications were made to the cork seam to allow periodic 

monitoring of the water levels in the joint, to permit periodic water removal, and 

to prevent water and contamination migration within the cork-filled joint. (391)

Contamination in the cork seam of expansion joints. Modifications were 

made to the cork seam to allow periodic monitoring of the water levels in the 

joint, to permit periodic water removal, and to prevent water and 

contamination migration within the cork-filled joint. (481.16)
115

6.5 Recontamination

Recontamination of previously decontaminated areas was caused by splashing, 

overspray, and drainage from decontamination efforts in adjacent areas. Airborne 

transport of contamination was also a significant problem.

• Recontamination Concerns. Recontamination was a concern that kept dose 

rates from being reduced to anticipated levels. The contaminated waste water from 

the low- and high-pressure flushing operations was the major source of 

recontamination. The large volume of water produced was difficult to control

splashing and overspray onto adjacent surfaces contributed to the problem. Once 

surfaces were decontaminated and nearby unconfined sources of water were 

eliminated, stepoff pads and barriers were quickly implemented to reduce the 

opportunity for recontamination. In late 1988, most of the remaining plant 

decontamination was postponed until the completion of defueling due to the 

recontamination of some previously cleaned areas. This also focused project 

resources on the highest priority work; only decontamination that directly 

supported defueling was performed. (392, 393, 394)

• Contamination Control. The results of the gross decontamination experiment 

and actual experience provided the following insights: (●) Sequencing of 

decontamination activities to take into consideration gravity flow and location did 

not prevent recontamination. (●) Personnel traffic from the air lock to various 

areas of the containment building caused recontamination of surfaces where there 

was a significant difference in contamination levels between areas traveled.

(●) Adequate controls to isolate clean areas from contaminated areas reduced

recontamination. (●) High-pressure spray activity with the lance and water floor 

scrubber caused contamination of other surfaces. (●) A sufficient amount of time 

should be spent with the decontamination crews in training on the recontamination 

problem (i.e., how to recognize it and the corrective actions to be taken during 

decontamination activities). (●) Work procedures should consider recontamination 

of areas being decontaminated. (●) Moving equipment and large tools through 

contaminated areas to support decontamination efforts at higher elevations 

required protection, or the equipment had to be cleaned before it was placed on 

clean surfaces. This was especially true if the hoist area was where a great quantity 

of debris had been dumped or flushed and not cleaned.

• Contaminant Migration in Coatings. Conventional decontamination methods 

were only partially successful because of recontamination caused by contaminants 

embedded in the epoxy system’s primer coat layers migrating to the surface. For 

some areas, the recontaminations were so frequent that floors were scrubbed 

mechanically and wet-vacuumed as often as three times per week. An alternate 

solution was to remove the coatings. (395)

• Airborne Recontamination. Significant recontamination problems in the 

containment building were largely due to the airborne transport of contamination 

from highly contaminated surfaces. One major source of recontamination of floor 
116

surfaces was airborne contaminants primarily originating in the building air 

handling systems. This source was identified by tests that measured the 

contamination levels above and under a plastic plate placed over a cleaned test 

area for 24 hours. Another source of contaminants was friable, powdery film or 

boric acid crystals caused by the use of borated water for flushing building 

surfaces. This source was associated with the airborne activity concentrations 

measured on personnel samplers. This activity was caused by resuspension due to 

personnel movement. To reduce airborne transport, the containment building air 

cooler fans were reduced to two fans running at slow speed, unneeded ducts were 

closed, and some top-to-bottom ducts and air coolers were flushed. (396, 397) 

• Misting System. Removal of the reactor vessel head required that the water 

level be lowered to just below the top of the internal plenum assembly. The finely 

divided debris that had settled onto the top of the plenum assembly was thought to 

have the potential to become airborne, particularly when the water film left on the 

top surface evaporated. Accordingly, a misting system was installed to wet the 

exposed plenum surface to reduce airborne contamination. (398, 399)

• Minimized Welding. Welding inside the containment building was avoided to

minimize airborne contamination. (400)

A drill-vacuum system was developed by the INEL to collect the 

surface deposition samples. The sample sequence consisted of 

(a) setting up the drill-vacuum system at the sample location; 

(b) vacuuming an small area about 40 cm2 into a filter;

(c) changing the filter and boring a 1/2-inch-diameter hole with a 

flat nose bit while continually vacuuming all debris from the bored 

hole; and (d) moving the drill-vacuum equipment a short distance 

away and repeating the sequence for the deeper hole. Several 

samples of varying depth were collected at each location to aid in 

determining the extent of contamination. An ion chamber 

instrument provided beta and gamma readings at same locations as 

surface samples. (481.15)
117

Table 5. TMI-2 decontamination method reference table. (401)

Method Description Benefits Disadvantages

LowPressure 

Flushing

Flush wand using 

<1,000 pounds per 

square inch (psi), high 

volume of water

Approx. 1,000 square 

foot per hour (sqft/hr); 

large areas quickly 

flushed; easily 

controlled

Large volume of 

waste generated; 

unable to remove 

more tightly adherent 

contamination

HighPressure 

Flushing

Flush wand using 

10,000 psi, low 

volume of water

200 sqft/hr; removed 

oxide layers; most 

effective flushing 

treatment

Not easily controlled; 

personnel safety 

concern; can rapidly 

spread contamination

Multiplaner 

Flushing

Two- or threedimensional multijet

sprays using low 

pressure and high 

volume of water for 

gross flushing

360 degrees, four-pie 

coverage of 

target/cubicle; effective 

volume dose reduction

Large volume of 

waste generated; 

usually unable to 

remove more tightly 

adherent 

contamination

Reflooding Leaching of 

contaminants from 

concrete through water 

resaturation

Total leaching possible; 

however, several years 

of processing would be 

required

Very expensive to 

reflood expansive 

areas; high volume of 

waste water 

generated

Fill and 

Leach

Block wall leaching 

from inside out via 

low pressure and high 

volume of injected 

water

Reduced radionuclide 

content 33 percent in 

tested areas; method 

performed remotely

Cost/benefit 

marginal; high 

volume of waste 

water generated

Simple 

Tools

Mop and bucket, towel 

wipes, scrub brushes

Inexpensive tools; little 

training needed; 

smearable 

contamination directly 

reduced

Sometimes personrem intensive; extra 

worker care must be 

taken

Dry/Wet 

Vacuuming

Basic industrial wet 

and dry vacuum with 

high-efficiency 

particulate air filters

Effective in collecting 

loose debris on a small 

scale; easily operated 

equipment

Dry-vacuum could

increase airborne

levels; vacuuming is 

not for large-scale 

decontamination

Mechanical 

Scrubbing

Basic industrial floor 

scrubbers with 

abrasive pads

100-150 sqft/hr; 

excellent results on 

painted concrete, more 

tightly adherent 

contamination

Splatter from rotary 

pads; operator fatigue
118

Table 5. TMI-2 decontamination method reference table. (Continued)

Method Description Benefits Disadvantages

HydroScabblingConcrete surface 

removal using ultrahigh-pressure water, 

10,000 to 40,000 psi; 

manual or remote 

operations

Could easily remove 

surface concrete from 

walls and floors

Manual operations 

are dangerous 

w/ultrahigh

pressures; need 

subsequent processes 

to pick up and 

dispose of waste and 

to smooth surfaces 

prior to refinishing

HydroScarifying20–40 sqft/hr; 

demonstrated excellent 

vertical surface 

capability

Scabbling Concrete surface 

removal by 

pneumatically driven, 

toothed pistons

400 sqft/hr; remotely

operated; 1/16-inch 

depth of surface 

removed per pass; little 

or no airborne

Manual operation 

quickly fatigues 

operators; poor 

vertical surface 

capability

Scarifying Concrete surface 

removal using rotary, 

drum-housed blades

100 sqft/hr achievable; 

radiation levels 

effectively reduced in 

conjunction with scrub 

and vacuum

Increases airborne 

activity; many passes 

to remove; leaves 

unfinished surface

Steam 

Vacuuming

Enhanced steam 

cleaner with highefficiency vacuum

Decontamination factor 

>10; easily operated up 

to 175 feet from the 

base unit; little system 

residue

Some surfaces could

clog the vacuum inlet

Dry 

Abrasive 

Blasting

Particles of steel, sand, 

glass, etc. entrained in 

air or water, driven at 

high velocities to blast 

surfaces clean; internal 

vacuum reclaims 

particles

Slow but effective; 

leaves smooth surface 

ready for protective 

coatings

Difficult to reclaim 

blasting media

Liquid 

Abrasive 

Blasting

Known to have good 

decontamination factors 

on relatively smallscale projects

Relatively expensive 

process with many 

operational 

constraints

Flex 

Hone/Mole 

Nozzle

Pipe snakes with 

rotary brush and highpressure heads enter 

and scrub pipes of 

various sizes

Inner-pipe oxide layers 

were removed; a 

polishing effect was 

observed; easily 

operated

Relatively slow 

production rates
119

Table 5. TMI-2 decontamination method reference table. (Continued)

Method Description Benefits Disadvantages

Strippable 

Coating

Organic coatings 

containing chemicals 

that bond to and 

remove surface 

contaminants; 

strippable requires

manual removal; selfstripping blisters and

peels by itself

1,000 sqft/hr applied, 

775 sqft/hr removed; 

reduced loose 

contamination level; 

prevented 

recontamination

Organic-based 

materials pose waste 

disposal concerns

SelfStripping 

Coating

Saved worker resources 

by self-strip; effectively 

reduced surface 

contamination levels

Sometimes “self” 

strip was not 

complete with certain 

coatings

Chemical 

Foams

Expanding foam 

mixed with cleaning 

solution increasing 

coverage and contact 

time

Increased residence 

time of chemical 

decontaminants on 

target surfaces, 

especially verticals

Waste disposal 

concerns; 

interference with 

water processing 

systems

Reagents/ 

Detergents

Surfactants and 

chelating surface 

cleaning agents 

usually applied with 

water spray

Proved to be more 

effective than plain 

water; increased 

scrubbing effectiveness

High pressure spin jet floor cleaner with pressures up to 10,000 

psig at a cost of $54,000 (1983 dollars). The water seals on the 

rotating nozzles developed leaks probably caused by operation 

of the nozzle rotational motor without water flow provided to 

the seal. (481.15)

Figure on Next Page. An attempt was made on May 20, 1980 to enter the 

containment building prior to the purging of radioactive krypton gas. However, the 

initial entry was delayed until after the purging because of a malfunction of the 

inner door to the building’s personnel airlock. (This photograph maybe the actual 

attempt or a training exercise.)
120
121

7 PERSONNEL PROTECTION

The extensive contamination and the concern for worker protection during 

decontamination activities prompted a variety of innovations, including new 

surface cleaning techniques and new radiation survey equipment to quantify 

contamination levels; improvements in protective clothing; techniques for 

reducing worker heat stress; and improvements in beta dosimetry. In prescribing 

protective clothing in the hostile environment, the focus was on the overall risk to 

the worker, (402) including heat stress, visual and hearing acuity, and 

cardiopulmonary stress. One measure of TMI-2 success in personnel protection 

was the radiation exposure record. Even though decontamination had been an 

ongoing effort throughout virtually the entire decade-long cleanup, total worker

exposure was less than half of the revised estimates of Supplement 1 to the TMI-2 

PEIS. (403)

7.1 Dose Exposure Control

• Cumulative Worker Exposure Over 10 Years. Although worker activities at 

TMI-2 had been quite different from those at operating power plants, the 

cumulative doses at TMI-2 since the accident had been lower than the average 

doses experienced at operating reactors. By the end of 1989, the collective dose to 

all workers was 6,180 person-rem. This was comparable to the collective 

occupational radiation exposure of 2,000 to 8,000 person-rem (404) estimated in the 

original TMI-2 PEIS. The revised estimates from Supplement 1 of the PEIS (405)

were between 13,000 and 46,000 person-rem. The cumulative occupational dose 

for defueling and defueling support activities was much less, below 2,000 personrem. The exposure rate to defueling workers averaged less than 10 millirem per 

hour. (406)

.

Annual personnel cumulative dose during 1979 to August 31, 1989. Cleanup was 

considered completed in January 1990. During the last 5 months defueling of the core 

former region behind the baffle plates was started and completed, lower reactor vessel 

head airlift/vacuum was completed, and ex-vessel defueling was completed. (481.17)
122

Worker standing next to radiation 

shielding of interlocking fiberglass 

columns filled with sand that was used to 

shield the enclosed stairwell and the 

reactor vessel head.

• ALARA Review Criteria. The criteria used at TMI-2 to determine the need for

an ALARA review for a task included: (●) any task anticipated to accumulate 

5 person-rem or more of total exposure; (●) any task for which the dose or dose 

rate to the skin or the extremities, or both, might be limiting without special 

radiological controls; (●) any task in which the airborne concentration was 

expected to exceed the limits specified in 10 CFR Part 20, “Standards for 

Protection Against Radiation,” Appendix B, “Annual Limits on Intake (ALIs) and 

Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure: 

Effluent Concentrations; Concentrations for Release to Sewerage,” Table 1, 

“Occupational Values,” by a factor of 1,000 times (i.e., respiratory protection of 

1,000 times was inadequate); (●) any task that could release radioactive material 

directly to the environment; (●) work with highly contaminated systems or 

components, as identified by radiological engineering; and (●) reactor disassembly 

and defueling operations involving core alterations. (407)

Additional considerations for performing an ALARA review included evaluating 

the relative risks of radiation exposure verses physical safety hazards. For 

example, ALARA reviews were conducted where fall hazards were present and 

the restricted visibility and physiological stress when wearing respirators could 

increase the risk of falling. (408)

• ALARA Evaluations. A typical 

ALARA evaluation at TMI-2 included 

the following steps: (1) Evaluate 

radiological conditions in each location 

of work to determine the sources and 

their relative percentages of contribution 

to the total area dose rate. (2) Evaluate 

the area work occupancy in terms of the 

total job hours and the schedule of 

work. (3) Determine applicable dose 

reduction methods (i.e., shielding or 

source removal). (4) Identify options 

and estimate the degree of reduction 

from each. (5) Calculate the net positive 

benefit derived from each combination 

of options. (6) Select options with the 

highest net positive benefit for 

prioritization and implementation. (409)

• External Exposure Control. For 

jobs involving major, or significant, 

exposures, the following controls were 

required to reduce external exposure: 

(●) radiological controls incorporated in 

the design; (●) written instructions; 

(●) getting workers involved in the 
123

A radiation work permit was 

required to access contaminated 

ceiling areas above the clean lower 

areas of the auxiliary building 

main corridors.

development of tools and procedures; (●) use of mock ups to practice and to test 

tools and work plans; (●) prejob briefings before beginning work; and (●) postjob 

debriefings for lessons learned. (410)

• Skin Dose from Hot Particles. The release of reactor coolant containing finely 

divided fuel debris that generated discrete radioactive particles (or “hot particles”)

was a problem at some locations within the plant. Hot particles were small, 

sometimes microscopic, particles of contamination with relatively high specific 

activity. Hot particles at TMI-2 were primarily fuel fines from damaged fuel and 

activation particles originating mainly from the high-cobalt alloy used in valve 

seats. A computer code and other published methods of assessing skin dose were 

used to estimate doses from hot particle contamination. (411, 412) 

• Control of Hot Particles. To control hot particles during defueling operations, 

the containment building was arranged in a set of successive rings of increasing 

contamination with the defueling platform at the center. Disposable gloves and 

booties were staged at sticky stepoff pads. A layer of gloves and booties were 

removed when moving into a cleaner area. Sometimes a clean set of booties could 

be used to cover the dirty pair. (413) Additional precautions were taken to ensure 

that hot particles remained in the containment building, such as (●) frequent 

wipedowns of workers; (●) ventilation controls in the desuiting area; (●) increased 

frequency of personnel monitoring; and (●) automated personnel contamination 

measuring devices. These control mechanisms were very effective at limiting 

personnel exposure to hot particles. At TMI-2, no worker exceeded the licensee’s 

administrative dose limit for the skin (50 millisievert/calendar quarter) as a result 

of hot particle exposure. (414)

• Skin Protection. During early containment 

building entries, workers entering the 

containment building wore two sets of cloth 

coveralls. Some workers who were entering 

areas of very high surface contamination also 

had an outer plastic suit over the two sets of 

cotton coveralls. The coveralls were 

commercially laundered and reused as was 

typical of most nuclear power plants. 

However, the radioactive contamination on the 

coveralls from TMI-2 proved more difficult to 

remove by laundering. Further, significant 

amounts and types of contamination on the 

coveralls from TMI-2 were not typical of other 

nuclear facilities. Despite efforts to try 

different methods to launder the clothing, some 

contamination remained “fixed” in the cloth 

material. The residual contamination did not 

present a significant skin dose to the worker so 

the continued used of the contaminated 
124

garments was considered acceptable. However, the number of skin contamination 

incidents continued to increase with the workload and contamination levels 

workers encountered. It was eventually determined that workers’ physical exertion 

and perspiration contributed to contamination leeching out of the material of the 

inner set of coveralls and onto the skin. This leaching resulting in minor but 

numerous detectable skin contamination incidents. Consequently, it was decided 

that the inner set of laundered cloth coveralls should be replaced with a new, clean 

set of disposable, breathable paper coveralls. The clean layer of paper helped to 

prevent skin contaminations caused by cross contamination from the laundered 

cloth coveralls. (415)

TMI-2 workers who wore paper garments beneath their cloth coveralls reduced 

instances of skin contamination by 64 percent while working in the containment 

building. Skin contaminations were limited to those areas on the body not 

protected by the paper garments. In addition, paper garments were more 

comfortable than two sets of cloth coveralls. (416)

• Containment Atmosphere Dose Control. As a result of the accident significant 

quantities of fission gases and volatile radionuclides, primarily radioiodine, were 

released into the enclosed containment building atmosphere from the damaged 

reactor core. Approximately 1 year after the accident, air samples of the 

containment building atmosphere showed that krypton-85 was the principal 

remaining radionuclide. Krypton-85 concentration was estimated to be 

3.77 x 10+10 becquerel/cubic meter. (417) To permit less restricted access to the 

containment building and to proceed toward decontamination of the TMI-2 

facility, it was necessary to remove the krypton gas and to provide a suitable 

environment for workers. The purging began on June 28, 1980 and continued until 

the morning of July 11, 1980. (418) Removal of krypton-85 from the containment 

building atmosphere was estimated to reduce the radiation dose rate for workers 

by a factor of about 4. (419) The removal of 44,000 curies of radioactive krypton-85 

gas from the containment’s atmosphere allowed workers to begin to clean up the 

containment building, to maintain instruments and equipment, and to remove the 

damaged fuel from the reactor vessel. (420)

7.2 Worker Wellness Considerations

Considerations for worker safety and convenience included heat stress controls;

psychological concerns; respiratory protection; whole body counting; and a 

dedicated dressing facility. 

• Heat Stress Control. Heat stress concerns are common at all nuclear power 

plants, but at TMI-2, workers wore more lavers of protective clothing, including 

respirators and plastic suits. The potential for heat stress became the limiting

stay-time factor, especially during the summer months. Experience had shown that 

summer temperatures could reach between 79 and 89 degrees Fahrenheit in the 

containment building, thus limiting working time to 1.5 hours per crew. The 

licensee’s industrial safety office, EPRI, and Pennsylvania State University 
125

developed a comprehensive heat stress control program consisting of employee 

training, administrative controls, and personal cooling devices. The most 

significant means for reducing heat stress was the installation of an air chiller unit 

inside the containment building. 

o Initial Program. The licensee’s initial program for reducing heat stress 

focused on three actions: (●) medical screening, (●) education of workers to 

recognize heat stress, and (●) controlling work times. Protective clothing 

requirements were sometimes reduced, when possible; however, the types of 

tasks required for cleanup did not allow much flexibility. Attempts to screen 

and partially acclimate workers using mockup simulations, or a standard 

exercise, were very time consuming and did little to improve the productivity 

of the cleanup effort. Administrative controls to control work times resulted in 

short time limits that hindered productivity and caused an accumulation of 

dose as more workers had to travel through areas of high radiation levels. (421)

A computer program estimated a safe stay-time for a specific task in a specific 

work area. Estimates were based on work rate, air temperature, and insulating 

effects of protective clothing types. (422)

o Ice Garments. The frozen water garment, or ice vest, was a tight-fitting vest 

holding 8 pounds of ice packets worn under protective clothing. Although 

melt time was limited, workers in the high heat environment wearing the 

garments doubled their stay times. Field tests conducted at TMI-2 showed that 

the garments lowered body core temperature and heart rate. Prepared 

garments and additional ice packets were kept in a freezer in the worker 

dressout area. Experiences reported from the use of ice garments included the 

following: (●) About 5 to 10 percent of ice packets broke or leaked during 

each use and had to be replaced. (It was suggested that the use of a heavier 

Lead blankets used to shield worker pathways. (Upper left: core flood tank; upper 

middle: containment building air chillers; right: plenum of air-cooling assembly.)
126

plastic could have prevented breakage.) (●) A more durable cotton twill 

material had a very long service life and cost substantially less than the 

original ripstop nylon garment. (●) Garments were difficult to decontaminate; 

those heavily contaminated had to be disposed. (●) To avoid cross 

contamination, garments were laundered separately from contaminated 

protective clothing. (●) Substantial ice melting occurred if there were delays 

between the dressup period and entry into the work area. (●) Melted ice added 

weight to the worker’s heat stress burden. (●) Liquid nitrogen was used to 

quick-freeze the water packets to save time, but this practice was discontinued 

because the ice packets were so cold that frost bite was a safety concern. (423)

o Air-Flow Cooling Garments. Other commercially available body-cooling 

garment designs were considered, including those using circulating water 

tubes in the garment to remove body heat. These garments were found to be 

difficult to decontaminate and expensive, and their pumps required regular 

maintenance. The vortex cooling suit, which was tested at TMI-2, produces a 

flow of cool air to the skin that removes body heat through convection and 

increases sweat evaporation. The cooling effect produced by the vortex tube 

(a passive tube that separates a compressed gas into hot and cold streams) 

successfully protected workers from heat stress. However, problems limited 

the use of vortex suits at TMI-2 for the following reasons: (●) The part of the 

vortex tube outside the protective clothing could not be easily 

decontaminated. (●) A large volume of service air was required to 

simultaneously operate vortex suits for several workers. (●) Worker mobility 

was restricted by the umbilical hoses supplying air to the suits. (424)

o New Air Chiller. The installation of an additional air chiller to the 

containment building air handling system proved to be the most successful 

strategy for preventing heat stress. (425) The new chiller system reduced the 

temperature to 64 degrees F and permitted worker stay-times of 3 to 4 hours. 

This also limited the use of short frozen water garments to an exception-only

basis. (426) However, air cooling resulted in some condensation when the 

humidity was high, thus creating additional, although not significant,

contaminated water. (427)

• Psychological Concerns. One reported issue that would have benefited from 

additional attention at TMI-2 was the emotional and psychological concerns of the 

cleanup workers. Management decisions often did not factor in workers’ real

concerns associated with the safety of their families, or “taking doses home.” An 

example that was cited where management tried to overcome these concerns 

involved the licensee’s requirement to reduce the use of respirators in areas of the 

containment building or during activities where airborne contamination was not a 

problem. From an ALARA perspective, working without respirators improves 

work efficiency, thus reducing time and dose. The licensee assembled survey data 

and technical expert assessments to show the workers that the air in the 

containment building was safe to breathe and within regulatory limits.

Management and health physics technicians met with the workers to discuss their 
127

concerns and to try to put at rest their fears of incurring additional internal doses

and taking contamination home. Building and maintaining trust among coworkers 

and confidence in the health physics department were the keys to easing the 

workers concerns. The cleanup required special attention to worker morale. (428)

• Respiratory Protection. Of the numerous types of respirators used at TMI-2, 

the two most useful devices from the standpoint of worker comfort, productivity,

and dose reduction were powered air purifying respirators (PAPRs) and supplied 

air hoods. The PAPRs were used almost exclusively for cleanup and recovery 

operations at TMI-2. Advantages of the PAPRs were that they limited dead 

airspace and lens fogging while providing a cooling effect on the face. Advantages

of the supplied air hoods were that they were much more comfortable to wear and 

very useful in mitigating heat stress because the exhausted air was directed down 

the wearer’s torso. (429)

For the bulk of the decontamination and defueling efforts PAPRs were used where 

worker mobility was required, such as maintenance and decontamination 

activities. Supplied air hoods were used where worker mobility was limited and 

where worker stay times were extended over several hours. (430)

Left: Typical defueling entry radiological control suit during dressing. The powered air 

purifier battery and pump pack as well as the dosimetry were attached at the waist. The 

plastic coat would be pulled up over the shoulders and the lower hood and taped in 

place. Right: Containment building entry and exit facility called the contamination 

control corridor, also known as “C Cube”. (481.18)
128

• Dedicated Dressing Facility. A dedicated dressing facility, known as the 

personnel access facility (PAF), prepared workers for highly contaminated 

environments and reduced incidents of skin contamination. The PAF staff helped 

workers put on their protective clothing correctly. In addition, they verified that 

the provisions of the radiation work permit were met, assured proper respirator fit, 

provided special dosimetry, and assembled any necessary support equipment 

(e.g., tools, radio equipment). The PAF also served as a staging area for personnel 

awaiting authorization from the remote coordination center to enter the 

containment building, thereby reducing queuing at the access control point. The 

PAF was an effective tool in managing the containment entry process, reducing 

worker skin contaminations, and improving productivity. (431, 432, 433, 434)

In order to minimize cross contamination from workers and equipment exiting the 

containment building, a contamination control corridor (CCC) was established 

within the PAF. In the CCC health physics technicians assisted in the removal of 

equipment and in undressing the workers. The CCC was very effective in 

reducing the risk of contamination spread beyond the radiologically controlled 

areas. (435)

• Whole-Body Counting. The whole-body counter for the measurement of 

radioactivity within the human body was open for use whenever it was needed, 

which was approximately 20 hours per day, during operations to remove and store 

the reactor vessel head. For any large-scale operations after the reactor vessel head 

removal, it was recommended that the counter continue to be open for use, as 

needed, to support the remaining work. (436)

• TMI-2 Worker Registry. In response to an April 1, 1979, memorandum from 

the Secretary of Health, Education, and Welfare, a cooperative effort between the 

NRC and the U.S. National Institute for Occupational Safety and Health 

established a registry of occupational exposure at TMI-2. This registry would 

facilitate future radiation epidemiological studies. (437)

In 1980, the NRC concluded that the preexisting dosimetry program supported by 

personnel and medical records was fully satisfactory for carrying out future 

studies. The licensee retained those essential elements of a work registry 

(i.e., personnel records, medical records, exposure history) of every radiation 

worker. The NRC recognized that work registry information would not be useful 

in distinguishing a causal-effect relationship between TMI-2 worker radiation 

exposures and future health effects. Individual worker doses at TMI-2 were 

subject to the same NRC limits as at any other licensed facility. Further, almost all 

worker exposures had been maintained administratively well below NRC limits. 

Within these limits, the statistical likelihood of any resulting health effects from 

occupational radiation exposure (i.e., cancer, generic defects in future generations) 

would be very small and would not be discernible from health effects from any 

other cause. (438)
129

Considerations applied at TMI-2 for maintaining personnel radiation exposure 

ALARA during plant maintenance, repair and recovery. (481.19)

Figure on Next Page: The submerged demineralizer system in the spent fuel pool. 
130
131

8 WASTE MANAGEMENT

The TMI-2 accident and subsequent cleanup presented challenges in terms of the 

management of various forms and concentrations of radioactive waste. The 

management of highly contaminated water, fuel debris, and related solid waste 

byproducts included various tasks, such as handling, processing, temporary onsite 

storage, transportation, and final disposal. Decontamination activities resulted in 

substantial quantities of contaminated water and organic resins and inorganic 

zeolites produced from water processing systems. Fuel debris that spread 

throughout the plant created unique radiological waste characteristics. Also, some 

waste did not fit into established regulatory waste classification categories for 

transportation and disposal, and the possible generation of flammable gases inside 

sealed radioactive waste containers was a potential hazard. (439)

Most radioactive trash and solid decontamination wastes were handled as at other 

nuclear power plants but on a much larger scale. TMI-2 generated approximately 

6,000 cubic meters of radioactive waste (excluding fuel debris). About 98 percent

was classified as low-level radioactive waste that could be commercially buried; 

the remaining 2 percent was disposed of by special arrangements with the DOE.

Since TMI-2 was a newer plant, several unusual factors made waste management 

there much less difficult, such as limited quantities of fission and activation 

products (especially cobalt-60), two empty spent fuel pools, and an 

uncontaminated steam generator chemical cleaning building set up for water 

processing. (440)

A few of the many experiences from the management of liquid and solid wastes 

during the TMI-2 cleanup are presented below. Thorough discussions of these 

topics appear in EPRI-TR-100640 and the DOE report, “Historical Summary of 

the Fuel and Waste Handling and Disposition Activities of the TMI-2 Information 

and Examination Program (1980–1988),” issued October 1988 (EGG-2529).

8.1 Liquid Waste

Over one million gallons of contaminated water existed in the plant shortly after 

the accident. The water ranged in concentration from less than 1 microcurie per 

milliliter to hundreds of microcuries per milliliter and existed in auxiliary building 

tanks and systems, the containment building basement, and the reactor coolant 

system. New water processing systems that were designed and installed to 

decontaminate the accident-generated water included the EPICOR II system, 

submerged demineralizer system, and defueling water cleanup system. The 

processed water disposal system removed residual contaminants from processed 

accident-generated water through a closed-closed evaporator. (441)

• Water Management. (442, 443) Over the first few years, a major portion of the 

licensee’s resources were spent on water management. Before the accident, 

radioactive waste tanks were already 60 percent full of waste water from the TMI 

Unit 1 shutdown outage just before the accident. In addition, leakage from various 
132

operating systems in the auxiliary and fuel handling building added about 800 to 

1,000 gallons per day. The water was initially distributed in the containment 

building basement, reactor coolant system, and auxiliary building sumps and tanks 

and over the lower elevation floor of the auxiliary building. The water, with the 

associated high radiation fields, prevented system maintenance and hindered 

cleanup work. 

During the early weeks and months following the accident, plant operators 

transferred contaminated water between existing tanks. The existing plant systems 

were unable to process any of this water, which contained cesium-137 

concentrations that were initially several terabecquerels per cubic meter. Railroad 

tank cars were considered for onsite storage of low-level radioactive waste water; 

however, they were never used to store contaminated or processed water. In 

July 1979, the fuel pool waste storage system (known as the “tank farm”) was 

available with 110,000 gallons of storage capacity. In July 1981, two new 

processed water storage tanks were available (each with a capacity of 

500,000 gallons). 

• EPICOR I. The EPICOR I system was installed at TMI before the accident to 

process low-activity, nonaccident-generated liquid waste water, mainly from the 

preaccident TMI Unit 1 outage. (444) In their policy statement of May 25, 1979, the 

NRC Commissioners permitted discharge of preaccident waste water 

decontaminated by the existing EPICOR I decontamination system and discharge 

Simplified schematic of the EPICOR II process flowpath.
133

of industrial waste water (water slightly contaminated because of leakage from 

secondary plant service support systems) into the Susquehanna River, as consistent 

with the facility operating license and NRC regulations. However, restrictions 

were imposed on the allowed discharges of EPICOR I and industrial waste water. 

The statement required the NRC staff to prepare an environmental impact 

statement for public comment and NRC Commissioner approval before the 

discharge of other (accident) waste water and the operation of the EPICOR II 

system. (445)

• EPICOR II—Initial Cleanup Mode. The existing plant water processing 

system lacked the capability to process accident-generated water, which contained 

cesium-137 concentrations initially ranging from 1 to over 100 microcuries per 

milliliter. The auxiliary building emergency liquid cleanup system (known as 

“EPICOR II”) was designed and installed after the accident to clean up about 

450,000 gallons of intermediate-level waste water from October 1979 to 

December 1980. This water was held in various storage tanks and sumps inside the 

auxiliary and fuel handling building. Consistent with the design objectives of 

simplicity and use of proven technology, the EPICOR II system employed a series 

of disposable ion exchangers, preloaded with organic and inorganic resin media 

selected specifically for the physical and 

radiochemical characteristics of the 

contaminated water. EPICOR II reduced 

the volume of radioactive waste by a 

factor of 10 over conventional waste 

processing systems. (446) The system was 

in the preexisting chemical cleaning 

building. (447) Problems were encountered 

during the initial mode of EPICOR II 

processing.

• EPICOR II: Hydrogen 

Generation. (448) An unexpected problem 

with the storage of spent EPICOR II liners 

that became programmatic at TMI and the 

U.S. nuclear industry was the discovery of 

gas generation inside heavily loaded 

(radioactive) demineralizer liners that used 

resin-based media for ion exchange. 

During preparations to ship a heavily 

loaded prefilter liner to the DOE’s Battelle 

Columbus Laboratories for 

characterization, (449) a flash ignition of 

hydrogen occurred at the EPICOR II cask 

loading station in the TMI-2 waste 

packaging and handling facility. Hydrogen 

and oxygen gas generated by radiolytic 

decomposition of residual water in the 

Prototype gas sampler installed on an 

EPICOR II liner. INEL designed and 

built the device to sample and vent the 

liners remotely and to add recombiner 

catalyst. (481.20)
134

liners became a safety concern during handling, transportation, and reception

of the liners.

INEL designed and built a device to sample and vent the liners remotely and 

add recombiner catalyst. A remotely operated vent tool was devised to remove 

the vessel vent plug while maintaining a sealed environment around the storage 

module cell. The liners were purged of hydrogen gas, and the hydrogen gas 

generation rate was quantified before shipment to comply with 

U.S. Department of Transportation shipping regulations.

The DOE had studied the generation of hydrogen gas as a result of radiolysis of 

water before the TMI-2 accident; however, these studies were limited to

high-level and transuranic wastes. The NRC sponsored later studies of 

hydrogen generation in EPICOR II vessels, which resulted in changes to the 

certification of shipping casks. In Information Notice 84-72, “Clarification of 

Conditions for Waste Shipments Subject to Hydrogen Gas Generation,”(450) 

issued September 1984, the NRC required plants to demonstrate, by tests or 

measurements, that combustible mixtures of gases were not present in 

radioactive waste shipments; otherwise, the waste was to be vented within 

10 days of shipping. A task force, formed by the Edison Electric Institute to 

evaluate these NRC requirements, developed a calculational method to 

quantify hydrogen gas generation in sealed containers. EPRI then demonstrated 

this calculational method using a desktop computer at TMI-2, a method the 

NRC accepted.

• EPICOR II: Processing Problems. Other problems encountered during 

EPICOR II processing included the following: (●) Repetition of problems 

encountered during EPICOR I operation. (Constant changing of the proprietary 

resin media allowed for very little preplanning of liner loading. Slow 

turnaround of chemistry and radiochemistry on samples did not allow the 

monitoring of liner corrosion breakthrough to be done in a timely fashion,

requiring significant quantities of water to be reprocessed.) (●) Two loop seals

in the line between the auxiliary building and the chemical cleaning building, 

where the EPICOR II system was housed, presented difficulties in initiating

and maintaining flow to the EPICOR II system from the auxiliary building. 

(●) The installed automatic sampler never worked as designed. (●) The 

relatively high curie loadings of liners required special handling for 

transportation and disposal, including venting of combustible gases, 

overpacking with concrete high integrity containers, or providing additional 

intrusion protection in the burial trench. (451)

• Submerged Demineralizer System. The submerged demineralizer system

(SDS) was located in the TMI-2 spent fuel pool (filled with processed water for 

shielding) to clean up high-level radioactive accident-generated waste water from 

the containment’s basement, reactor coolant system, and reactor coolant bleed 

tanks. The very successful operation of EPICOR II established the confidence and 

experience that was to prove invaluable in dealing with the more highly 
135

contaminated water in the containment building basement. The SDS design started

in the summer of 1979 and was designed and installed on an accelerated basis over 

the next 2 years. (452) The SDS design had a 2-year life expectancy, but it was 

operated for 7 years. (453)

Problems encountered during the operation of the SDS included the following: 

(●) Maintenance of valves and equipment in shielded manifolds and gloveboxes 

was difficult. (●) Maintenance of diaphragm valves in the leakage containment 

system required movements of highly radioactive vessels to permit underwater 

access by divers for repairs. (●) Postfilter cartridges were initially being plugged 

constantly by zeolite fines in the system effluent, which was most probably caused 

by insufficient flushing of the liners before use. (●) Early SDS vessels were 

designed with flow restricters, which were removed through modification. 

(●) Installed flow meters/totalizers were very erratic. (●) The lack of a sample 

point between the prefilter and the final filter required curie loading calculations 

on these liners, based on underwater dose rate surveys and computer models.

(●) Tank farm eductors more than doubled the volume of water that they 

transferred. (●) The long-handled tools used to couple and decouple liners lacked 

the sensitivity required to ensure positive seating of quick disconnect fittings,

which resulted in several leaks to the spent fuel pool. (454)

• EPICOR II—SDS Polishing and High-Integrity Container Polishing Modes. 

EPICOR II removed residual radioactivity from submerged demineralizer system 

(SDS) effluents and processed miscellaneous wastes during a 6-year period. After 

the SDS was removed from operation in 1988, EPICOR II was the primary system 

to clean up the waste water mainly generated from building decontamination 

Flowsheet for water processing through the submerged demineralizer system. (481.21)
136

activities. The system configuration was the same as before; however, a highintegrity container (HIC) was placed in the first position to act as a roughing filter. 

EPICOR II processed water more than 14 years, beginning with the initial cleanup

in October 1979. (455)

Problems encountered during these modes of operation of EPICOR II included: 

(●) Decreased cartridge postfilter life was caused by the lack of adequate 

prefiltration, allowing particulates to pass through ion exchangers. (●) Floating or 

swelling organic resins were responsible for some, but not all, of the erratic level 

indication and control problems. (●) The 108 rad integrated dose limit on the HIC

lowered the allowable cesium-137 loading from 1,000 curies to 348 curies. (456)

• Defueling Water Cleanup System. The defueling water cleanup system

(DWCS) removed organic carbon, soluble fission products, and particulate matter 

from the fuel transfer canal (FTC), spent fuel pool “A” (SFP-A), and the reactor 

vessel. The DWCS was two independent systems, one for cleaning up the 

FTC/SFP-A and the other dedicated to the reactor vessel. A temporary reactor 

vessel water filtration system was used until solutions could be incorporated in the 

DWCS to stop the microorganism growth and to remove the colloidal material. 

The processed water disposal system consisted of the following: (1) a vapor 

recompression distillation unit (main evaporator) that distilled the processed water in a 

closed cycle and collects the purified distillate for subsequent release by vaporization; 

(2) an auxiliary evaporator that further concentrated the bottoms from the main 

evaporator; (3) a flash vaporizer unit that heated and vaporized the purified distillate 

from the main evaporator and released the vapor to the atmosphere in a controlled and 

monitored manner; (4) a waste dryer that further evaporated water from the 

concentrated waste and produced a dry solid; and (5) a packaging system that prepared 

the dry solid waste in containers acceptable !or shipment and for burial in a 

commercial low-level radioactive waste disposal site. (481.22)
137

The temporary system operated from February 1986 to May 1987, after which the 

improved DWCS took over the water filtration function. (457)

The problems experienced with the DWCS included the following: (●) Initial filter 

canister throughputs were less than 10 percent of expected levels due to fouling 

caused by organic colloids or biological growth in both the reactor vessel and the 

FTC/SFP-A. This problem was compounded by the presence of organically 

stabilized colloids and was solved in the reactor coolant system through 

modifications to the DWCS, providing coagulant and filter-aid (body feed) 

injection systems. However, success with the same modifications to the 

FTC/SFP-A portion of the DWCS was marginal. (●) The DWCS was plagued with 

hose failures and coupling leaks, many of which were caused through external 

damage resulting from improper handling of tools and equipment. Many coupling 

problems were caused by interference with inlet and outlet lines inside the reactor 

vessel and the rotation of the defueling work platform. (●) The inline 

nephelometers (turbidimeters) experienced ranging and fouling problems. (458)

• Improved Defueling Water Cleanup System (DWCS). The very fine 

particulates and suspended solids in the reactor coolant quickly plugged the 

DWCS filters. Diatomaceous earth (DE) was injected into the process stream to 

create a fine cake on the filter element surface in a filter canister. The DE was 

about 90 percent silicon dioxide. This cake collected the solids, thus protecting the 

metal filter material. Once the filter was no longer effective (indicated by high 

differential pressure), the process was secured, and the DE was knocked off the 

filter elements and collected on the bottom of the canister. A fresh coat of DE was 

applied, and the process was repeated until the canister was eventually filled with 

spent DE. Adding DE increased the service life from as short as 1 day to a month 

or longer. (459)

Processed water disposal system closed-cycle evaporator/vaporizer used to 

dispose of processed accident-generated water.
138

• Use of Accident-Generated Water. The accident-generated water (AGW) was 

processed and recycled for decontamination to minimize the amount of fresh water 

likely to be contaminated. The NRC issued Supplement 2 of the TMI-2 PEIS after 

public comment in 1987. The processed AGW was stored on site in two 

500,000 gallon tanks. A recycling system transferred the processed water into the 

containment building for decontamination activities. The processed water was also 

used as radiation shielding for both spent fuel pools in the fuel handling building 

and for the refueling canal in the reactor building. The 2.3 million gallons of 

processed AGW remained on site for about 10 years before it was evaporated and 

vaporized into the atmosphere over a 30-month period, starting in 1991, after the 

NRC granted its approval. This water contained 1,020 curies of tritium and 

2.3 curies of all other contaminants. Other nonradioactive contaminants included 

150 tons of boric acid and 11 tons of sodium peroxide. About 99.9 percent of the 

dissolved radioactive contaminants (other than tritium) contained in the evaporator 

influent were collected as dry solid waste. (460)

• Core Material Leaching. In mid-1980, a reactor coolant sample was collected 

and analyzed to assess the extent of degradation of the reactor core and to provide 

long-term information concerning the continued leaching of radionuclides from 

the core material. At this time, the archived sample that was taken the day of the 

accident was divided into smaller portions, which were also analyzed. The results 

of these analyses indicated that radionuclide leaching from the core material was 

insignificant. The data were used to plan for reactor coolant system water 

processing by the SDS. (461)

8.2 Solid Wastes

Over 180,000 cubic feet of solid radioactive waste were generated and shipped 

during the cleanup period. Solid waste included spent EPICOR I and EPICOR II 

resin liners; spent submerged demineralizer system vessels; contaminated clothing,

tools, and equipment; and decontamination materials. The DOE accepted about 

3,000 cubic feet of certain wastes that exceeded commercial burial limits (called 

abnormal waste) for research and development of radioactive waste disposal

technology, such as vitrification, hydrogen generation control, and the design and 

testing of high-integrity containers. (462) Fuel debris, including filter canisters from 

the defueling water cleanup system, was shipped to INEL for storage under unique 

agreements among the DOE, the NRC, and the licensee. (463)

• High-Integrity Containers (HICs) to Dispose of EPICOR II Prefilter Liners. 

The DOE designed and fabricated a special HIC to dispose of the original 

45 highly loaded EPICOR II prefilter liners at the commercial low-level 

radioactive waste burial facility near Richland, WA. The HIC was an overpack 

(EPICOR II liner sealed inside the HIC) that would remain stable below ground 

for a minimum of 300 years (about 10 half-lives of predominant isotopes). The 

HIC consisted of a cylinder made of reinforced concrete and a permanently sealed 

lid. A vent system cast in the lid provided passive venting of the container. This 

HIC was restricted for use with only EPICOR II prefilter liners generated at 
139

TMI-2. The State of Washington issued Certificate of Compliance 

No. WN-HIC-01 for this HIC on March 23, 1984. (464, 465)

• High-Integrity Containers (HICs) to Dispose of SDS Liners. An attempt to 

qualify the design of the submerged demineralizer system (SDS) vessel as a burial 

container was unsuccessful because analysis indicated that pin-hole leaks in the 

stainless steel shell could not be ruled out within 300 years (10 half-lives of 

cesium-137) of burial. Therefore, the SDS vessels were buried inside polyethylene 

HICs. These poly HICs were permitted exclusively at the commercial low-level 

radioactive waste burial facility in Barnwell, SC. However, one problem existed 

with the use of the polyethylene HICs—some of the SDS liners exceeded 108 rads 

absorbed dose, which was the maximum authorized limit for these HICs. This 

problem was solved using a concrete liner inside the polyethylene overpack. With 

the concrete liner installed, the HIC was shielded from the SDS liner so that this

absorbed dose limit was not exceeded over the required 300-year period. (466)

• Radiolytic Gas Generation. Radiolytic gas generation in highly loaded

submerged demineralizer system (SDS) vessels was calculated to be significant. 

Measurements of gas buildup confirmed that this could be a problem during 

storage and transportation, leading a DOE research program to develop several

techniques to deal with it. To reduce buildup of flammable gas during 

transportation and storage, the vessels were drained and vacuum-pumped to 

remove free water. A catalytic recombiner system was developed, tested, and 

installed in SDS vessels that recombined the hydrogen and oxygen gases generated 

by radiolysis back into water, thereby eliminating gas generation. (467)

A pressure relief system, consisting of a burst diaphragm and micropore graphite 

filter, was also added to each SDS vessel to prevent the uncontrolled, long-term 

buildup of noncombustible gas mixtures. This addressed the possible net buildup 

of hydrogen due to oxygen scavenging by various chemical reactions, such as the 

formation of carbon monoxide and carbon dioxide from the oxidation of organic 

materials trapped within the zeolites. (468)

Design configuration of the HIC without an enclosed EPICOR II 

prefilter. (481.23)
140

• Solidification of EPICOR II resins. The licensee pursued the solidification of 

EPICOR II resins in response to an NRC order; however, the value of such a 

complex, expensive, and non-ALARA operation was debated. As the regulatory 

and burial criteria were clarified (partly in response to this issue), alternative 

disposal methods became more appealing. The eventual agreement by the DOE to 

accept commercially nondisposable wastes from TMI-2 to research and develop 

high integrity containers satisfactorily resolved the issue without the need to 

solidify the resins at TMI-2. (469)

• Dry Waste Characterization. Dry active waste at TMI-2 consisted of 

radiologically contaminated paper, plastic, rags, wood, and metal resulting from 

personnel access and work within contaminated areas of the plant. The dry active 

waste had high concentrations of fission products, with strontium-90 being the 

dominant isotope. The isotopic mix of the dry active waste from various points of

generation throughout the plant was highly variable. To account for this 

variability, the dry active waste was categorized into distinct waste streams based 

on the ratio of cesium-137 to strontium-90, measured on contamination smears 

from various locations of the plant. Isotopic distributions, determined by gamma 

scans of the representative smears from these locations, were formulated for each 

waste stream. The transuranic and difficult-to-measure nuclides reportable under 

10 CFR Part 61, “Licensing Requirements for Land Disposal of Radioactive 

Waste” (i.e., technetium-99, carbon-14, nickel-63, iodine-129), were inferred from 

scaling factors(j) developed for the waste streams at TMI-2. A computer code 

developed a correlation between gamma radiation intensity and the isotopic 

distributions of each waste stream. Using this correlation, individual conversion 

factors for each waste stream were formulated based on various package sizes and 

waste densities. These dose-to-curie conversion factors determined the isotopic 

concentration of packaged dry active waste based on measured dose rates. (470)

• Incineration. As an alternative to compacting lower activity solid waste, the 

project team investigated the use of an incinerator to reduce the volume. In 1981, 

the idea was studied (see GEND-021, “Controlled Air Incinerator Conceptual 

Design Study,” (471) issued January 1982) and finally discarded because the 

volume of waste that could be incinerated was too small to justify the cost and 

technical difficulties. Regulatory uncertainties for air discharge permits were

another factor in the decision to cease consideration. (472, 473)

8.3 Onsite Interim Waste Storage and Staging

Several onsite facilities were constructed for temporary storage of solid 

radioactive waste products from cleanup activities that were being readied for 

transportation.

j Scaling factors were decay-corrected ratio correlations used to infer quantities of 

difficult-to-measure radionuclides from concentrations of tracer nuclides that are easily 

measured by gamma spectroscopy methods and routinely reported in sample results.
141

• Interim Waste Storage. Waste disposal regulations established following the 

accident required considerable analysis and conditioning before waste shipment. 

This created a need to establish facilities or locations for the interim storage of 

waste to avoid impending delays in the decontamination of auxiliary and 

containment buildings. Interim waste storage was needed for the temporary 

storage of decontamination wastes. (474) Several temporary storage and staging 

facilities were constructed at TMI-2: (●) The interim storage facility was built in 

late 1979 to temporarily store EPICOR resin liners and filters until the 

construction of the solid waste staging facility was completed. (●) The interim 

solid waste staging facility (also known as the “car port”) was opened in late 1982 

and used to collect and temporarily store (stage) low-level solid waste packages 

from both Units 1 and 2. (●) The solid waste staging facility was used to collect 

and temporarily stage radioactive waste, such as dewatered resins, filters, and 

sludge, from both Units 1 and 2 before shipment. The first pad of 60 storage cells 

opened for use in early 1980. (●) The waste handling and packaging facility was 

used to process and package solid radioactive waste (see below). (475)

• Waste Handling and Packaging. Completed in early 1987, the waste handling 

and packaging facility was a major support for the decontamination efforts. The 

facility processed and packaged solid radioactive waste, such as contaminated 

clothing, tools, and equipment. Processing of contaminated material consisted of 

compaction, size reduction, and decontamination for reuse at TMI-2. No 

radioactive waste was stored in this facility. (476) Sixty-nine percent of items were 

releasable after decontamination. (477)

Solid waste staging facility under construction.

Figure on Next Page: Model 125-B shipping cask loading tower. A fuel canister 

was lowered into the shipping cask (lower center) from the fuel transfer cask (not 

shown). The loading tower lowered the shipping cask to a horizontal position onto 

a skid that would be attached to the rail car.
142
143

9 BIBLIOGRAPHY

Thorough overviews of the TMI-2 recovery and cleanup experiences, including 

lessons learned, can be found in the series of publications by the American 

Nuclear Society, EPRI, IAEA, DOE, and NRC. The following documents were 

reviewed for TMI-2 insights and lessons:

• American Nuclear Society (ANS). A special volume of the ANS Nuclear 

Technology journal includes 138 papers presented at the meeting held in 

Washington, DC, from October 30 to November 4, 1988. Volume 87 consists of 

four issues, each devoted to one of the specialized areas of technology covered 

during the meeting: Materials Behavior, Health Physics and Environmental 

Releases, Remote Technology and Engineering, and Decontamination and Waste 

Management. The papers in the proceedings may be purchased from the ANS Web

site (http://www.ans.org/pubs/journals/nt/v_87).

• Electric Power Research Institute. EPRI staffed its TMI-2 site office from 

1980 to 1992 with the purpose of transferring TMI-2 accident cleanup data and 

experience to member utilities and to provide technical support to the cleanup 

program. This technology transfer program involved EPRI, along with the 

licensee, the DOE, and the NRC. In the early stages, EPRI focused on research

and development in support of decontamination and dose reduction. Towards the 

end of the cleanup, EPRI focused on all aspects of the cleanup program but placed 

the greatest emphasis on identifying developments and technologies that were

applicable to operating plants.

EPRI sponsored the following series of reports to document every aspect of the 

decade-long cleanup at TMI-2. These reports are available to the public on its Web

site (www.EPRI.com), except where noted:

o Contamination and Decontamination Experience with Protective Coatings at 

TMI-2, EPRI Report NP-5206, 1987

This report (478) documents the review of the TMI-2 experience with protective 

coatings, including the performance of protective coatings at TMI-2 before and 

after the accident. The report also documents the significant technical findings 

on coating behavior, including pertinent findings on the generic problems of 

coatings in contamination and decontamination procedures.

o Radiation Protection Management Programs at TMI-2: Noteworthy 

Practices and Accomplishments, EPRI NP-5338, 1989

This report presents an overview of the radiation protection tools and 

techniques that proved useful in the TMI-2 accident cleanup. It discusses

protection training; exposure measurement; radiological engineering; and 

protection program management. Each discussion describes the challenge that 

confronted the cleanup effort, examines the tool or procedure that was designed 
144

to meet that challenge, and evaluates its effectiveness in contributing to dose 

reduction. This report is not available from EPRI’s Web site.

o The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 

1990, EPRI-NP-6931, 1990

This report identifies the major questions and challenges that faced

management, describes the influencing factors and the available options, and 

presents the final decisions and their consequences. This history of TMI-2 

focuses on decisions related to seven major aspects of the cleanup: cleanup 

management; postaccident stabilization; personnel protection; data acquisition;

radioactive waste management; decontamination; and defueling. A detailed

chronology and extensive bibliography accompany the text.

o Decontamination Experience During the Cleanup of Three Mile Island 

Unit 2, EPRI-NP-7157, 1990

This report documents those lessons learned at TMI-2 in selecting and using 

various decontamination tools, products, and techniques to accomplish plant 

cleanup.

o TMI-2 Post-Accident Data Acquisition and Analysis Experience,

EPRI-NP-7156, 1992

This report documents the data acquisition techniques used at TMI-2, describes

the extent of damage and hazard conditions within the damaged plant, and 

provides a base of ideas and approaches for utilities that face an accident or 

other circumstances requiring special sampling and data acquisition techniques.

o TMI-2 Waste Management Experience, EPRI-TR-100640, 1992

This report provides comprehensive documentation on all the important aspects 

of the management of radioactive waste that resulted from the recovery and 

cleanup at TMI-2. In addition, this report provides a historical perspective to 

the recovery period by documenting the actual volumes of radioactive waste 

that were generated.

o Final TMI-2 Technology Transfer Progress Report, EPRI-TR-100643, 1992

This report discusses the technical accomplishments during the 1987–1992 

period in the areas of decontamination; defueling; dose management;

operations; radioactive waste management; reactor system characterization;

robotics; and reactor damage assessment.

• International Atomic Energy Agency. IAEA publications provide technical 

guidance and recommendations for postaccident planning. The reports listed 

below, and others, are available to the public on its Web site (www.iaea.org):
145

o Recovery Operations in the Event of a Nuclear Accident or Radiological 

Emergency, Proceedings of a Symposium, IAEA-STI/PUB/826, 1990

This report (479) reviews the actual experience gained and lessons learned from 

recovery techniques and operations in response to serious accidents at nuclear 

facilities and accidents associated with radioactive materials, and to consider 

the development of emergency planning and preparedness resources.

o Catalogue of Methods, Tools and Techniques for Recovery from Fuel 

Damage Events, IAEA-TECDOC-627, 1991

This report provides information about possible methods and equipment to 

manage a nuclear fuel damage accident. The report serves as a reference 

catalogue of existing techniques. It draws primarily on work done at Chernobyl 

and TMI, but it also includes experience from minor fuel accidents and 

describes commercially available equipment that might be applicable.

o Management of Severely Damaged Nuclear Fuel and Related Waste, 

IAEA-TRS-321, 1991

This report (480) addresses onsite, postaccident management leading to a stable 

condition with respect to gaining control over damaged fuel, related waste, and 

radiological releases to the environment.

o Cleanup and Decommissioning of a Nuclear Reactor after a Severe Accident, 

IAEA-TRS-346, 1992

This report provides an overview of factors that are relevant to the 

identification of cleanup requirements and to the choice of a decommissioning 

option for a severely damaged nuclear power plant. A methodology is proposed 

to evaluate various options and to select an appropriate action in a particular 

accident situation.

o Experiences and Lessons Learned Worldwide in the Cleanup and 

Decommissioning of Nuclear Facilities in the Aftermath of Accidents, 

IAEA-NW-T-2.7, 2014

This report documents the review of experiences from IAEA Member States in 

the cleanup and decommissioning of nuclear facilities in the aftermath of an 

accident.

• U.S. Department of Energy. The DOE documented lessons learned during its 

decade-long participation in the research and accident cleanup project at TMI-2. 

Reports are based on a review of a wide range of project documents and 

interviews with personnel from the many organizations involved:
146

o TMI-2: Lessons Learned by the U.S. Department of Energy—

A Programmatic Perspective, DOE-ID-10276, 1990

This report summarizes the lessons, both technical and administrative, learned 

by the DOE during its decade of involvement at TMI-2. It addresses a broad 

range of topics, including decontamination; robotics; radioactive waste 

management; reactor defueling; accident analysis; and project management and 

administration.

o Historical Summary of the TMI-2 Core Debris Transportation Campaign, 

DOE-ID-10400, 1993

This report describes the TMI-2 core debris transportation campaign. Subjects 

include preparations for transport; loading at TMI-2; institutional issues;

transport operations; receipt and storage at INEL; governmental inquiries and 

investigations; the exchange of information between the program and the 

public; and lessons learned.

o Historical Summary of Fuel and Waste Handling and Disposition Activities 

of the TMI-2 Information and Examination Program (1980–1988), 

EGG-2529, 1988

This report provides an historical summary of the major activities conducted by 

the TMI-2 Information and Examination Program in managing fuel and special 

radioactive wastes resulting from the accident at TMI-2. The activities included 

the development and use of advanced handling, processing, and disposal 

technologies for those wastes.

• U.S. Nuclear Regulatory Commission. 

o Three Mile Island Accident of 1979 Knowledge Management Digest: 

Recovery and Cleanup, NUREG/KM-0001, Supplement1

The main objective of this supplement is to provide, in electronic format, the 

key historical documents that were issued during the recovery and cleanup 

efforts. The seven major aspects of the recovery and cleanup, as presented in 

EPRI-NP-6931, were used to organize the contents in this supplement into the 

following sections: management and oversight; plant stabilization; worker 

protection; data acquisition and analysis; radioactive waste management; 

decontamination; and defueling. It also includes an additional section on 

post-defueling activities. This supplement chronicles those activities that began 

a week following the accident and ended with the completion of disposal of 

accident-generated water and the entry into post-defueling monitored storage 

in 1993.
147

NUREG/KM-0001 and over 4,000 documents, photographs, and videos 

provided on the companion DVD set are available to the public from INL’s 

Web site (https://tmi2kml.inl.gov). (481)

Model 125-B rail shipping cask. (481.24)

Figure on Next Page: Reactor vessel head (not shown) and service structure 

resting on its stand surrounded by shielding. Shown are white interlocking tubes 

filled with sand and yellow lead blankets. (Sand was found easier than water.)
148
149

10 REFERENCES

Reference citations are file names of documents on the DVDs for

NUREG/KM-0001, Supplement 1, “Three Mile Island Accident of 1979 

Knowledge Management Digest: Recovery and Cleanup,” except for reports from 

EPRI and IAEA, journal papers, and presentations. File names start either with a 

date or report number. Other sources that are not on the DVDs are provided in a 

traditional citation format.

End notes embedded in figure captions are listed in the last endnote.

Many of the documents on the DVD are historical in nature and might contain 

information that is obsolete or superseded by current regulations and research 

results. The historical documents on the DVDs are for historical reference only 

and are not official NRC records. Please refer to the NRC’s public Web site 

(http://www.nrc.gov) for current information on regulations, policy statements, 

regulatory guidelines, regulatory processes, and research results.

1 USNRC, “Status of and Remaining Actions for Cleanup of Three Mile Island Nuclear 

Station, Unit 2,” Commission Paper SECY-90-350, October 9, 1990. 2 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest: Recovery 

and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 3 Holton, W.C., C.A. Negin, R.W. Lambert, “The Technical History of the Three Mile 

Island Unit 2 Cleanup: Factors, Options and Decisions.” International Symposium on 

Recovery Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria, IAEA-SM-316, 1990. 4 Skillman, Gordon R. (2014, March). “The 35th Anniversary of the Three Mile Island 

Nuclear Power Plant Accident of 1979: Working at TMI During and Following the 

Accident.” Video Seminar presented at the U.S. Nuclear Regulatory Commission, 

Rockville, MD. (Retrieved from NUREG/KM-0001 at 

https://tmi2kml.inl.gov/HTML/Search.html?folder=0d-Video-35th) 5 Clark, Philip R., “Establishing the TMI-2 Cleanup Organization,” American Nuclear 

Society Executive Conference on TMI-2: A Learning Experience, Hershey PA, 

October 13–16, 1985. 6 (1983-09-19) TMI-2 Organization Plan Revision 6 and 7 7 (1984-06-15) TMI-2 Organization Plan Revision 8 8 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest: Recovery 

and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 9 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 10 Holton, W.C., C.A. Negin, R.W. Lambert, “The Technical History of the Three Mile 

Island Unit 2 Cleanup: Factors, Options and Decisions.” International Symposium on 

Recovery Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 11 (1983-09-19) TMI-2 Organization Plan Revision 6 and 7 12 (1984-06-15) TMI-2 Organization Plan Revision 8 13 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016.
150

14 (1990-03-01) Final Report, Three Mile Island-Unit 2 Safety Advisory Board 15 (1990-03-01) Final Report, Three Mile Island-Unit 2 Safety Advisory Board 16 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 17 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979–1990 (1990-03) 18 Clark, Philip R., “Establishing the TMI-2 Cleanup Organization,” American Nuclear 

Society Executive Conference on TMI-2: A Learning Experience, Hershey PA, 

October 13–16, 1985. 19 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979–1990 (1990-03) 20 Clark, Philip R., “Establishing the TMI-2 Cleanup Organization,” American Nuclear 

Society Executive Conference on TMI-2: A Learning Experience, Hershey PA, 

October 13–16, 1985. 21 (1984-04-16) DOE, Agreement with DOE for Japanese Participation in the TMI-2 

Research and Development 22 Tabata, N., F. Masuda, “Utilization of Three Mile Island Unit 2 Research and 

Development Information in Japan,” Nuclear Technology, Vol. 87, No. 4, 

pp. 814–823, December 1989. 23 McGoff, D.J., et al., “The Role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International Symposium on Recovery 

Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 24 (1979-12-12) GPU, Summary Technical Plan for TMI-2 Decontamination and Defueling 25 GPU Nuclear, “Technical Plan, TMI-2 Program Strategy,” GPU-TPO-TMI-115, 

Revision 0, June 1986. (From the personal file of a knowledge contributor.) 26 GPU Nuclear, “A Strategy for Recovery Program Completion and Post-Recovery 

Configuration, Volume 1, Executive Summary” GPU-TPO-186, July 1986. Retrieved 

from https://libraries.psu.edu/about/collections/three-mile-island-2-recovery-anddecontamination/three-mile-island-reports 27 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 28 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 29 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 30 McGoff, D.J., et al., “The Role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International Symposium on Recovery 

Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 31 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 32 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 33 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 34 F.R. Standerfer, “TMI-2: Plant Recovery.'' Transactions of the American Nuclear Society 

Winter Meeting Washington DC, Vol. 57, November 1988, p. 397. 35 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 36 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992.
151

37 Skillman, Gordon R. (2014, March). “The 35th Anniversary of the Three Mile Island 

Nuclear Power Plant Accident of 1979: Working at TMI During and Following the 

Accident.” Video Seminar presented at the U.S. Nuclear Regulatory Commission, 

Rockville, MD. (Retrieved from NUREG/KM-0001 at 

https://tmi2kml.inl.gov/HTML/Search.html?folder=0d-Video-35th) 38 Holton, W.C., C.A. Negin, R.W. Lambert, “The Technical History of the Three Mile 

Island Unit 2 Cleanup: Factors, Options and Decisions.” International Symposium on 

Recovery Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 39 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 40 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979–1990 (1990-03) 41 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 42 (1981-01-12) NRC Chairman to GPU, NRC Activities Required to be Performed by the 

TMI Licensee (re 09-12-1980) 43 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 44 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 45 (1981-01-12) NRC Chairman to GPU, NRC Activities Required to be Performed by the 

TMI Licensee (re 09-12-1980). 46 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 47 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 48 IAEA, “Experiences and Lessons Learned Worldwide in the Cleanup and 

Decommissioning of Nuclear Facilities in the Aftermath of Accidents,” 

IAEA NW-T-2.7, September 2014. 49 IAEA, “Experiences and Lessons Learned Worldwide in the Cleanup and 

Decommissioning of Nuclear Facilities in the Aftermath of Accidents,” 

IAEA-NW-T-2.7, September 2014. 50 (1981-01-12) NRC Chairman to GPU, NRC Activities Required to be Performed by the 

TMI Licensee (re 09-12-1980) 51 IAEA, “Experiences and Lessons Learned Worldwide in the Cleanup and 

Decommissioning of Nuclear Facilities in the Aftermath of Accidents,” 

IAEA-NW-T-2.7, September 2014. 52 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979–1990 (1990-03) 53 GEND-022, TMI-2 Technical Information and Examination Program Annual Report 

1981 (1982-04) 54 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979–1990 (1990-03) 55 GEND-003, TMI-2 Technical Information and Examination Program Annual Report 

1980 (1981-02) 56 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979–1990 (1990-03) 57 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 58 American Nuclear Society, Nuclear Technology, Volume 87, Numbers 1-4, 1989. 59 NSAC-80-1, NSAC-1 Revised, Analysis of Three Mile Island Unit 2 Accident (1980-03) 60 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 61 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS (TMI-2)
152

62 (1980-06-23) U.S. Senate, Nuclear Accident and Recovery at Three Mile Island, A 

Report 63 Byrne, J.J., Rogan, R.E., “Three Mile Island Unit 2 Licensing History,” Nuclear 

Technology, Vol. 87, No. 3, pp. 683–689, November 1989. 64 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 65 NUREG-0683, Vol. 1, “PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)” 66 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10) 67 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 68 NUREG-0683, SUPP. 2, PEIS, Disposal of Accident-Generated Water (1987-06) 69 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08) 70 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 71 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 72 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 73 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 74 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 75 Byrne, J.J., Rogan, R.E., “Three Mile Island Unit 2 Licensing History,” Nuclear 

Technology, Vol. 87, No. 3, pp. 683–689, November 1989. 76 Skillman, Gordon R. (2014, March). “The 35th Anniversary of the Three Mile Island 

Nuclear Power Plant Accident of 1979: Working at TMI During and Following the 

Accident.” Video Seminar presented at the U.S. Nuclear Regulatory Commission, 

Rockville, MD. (Retrieved from NUREG/KM-0001 at 

https://tmi2kml.inl.gov/HTML/Search.html?folder=0d-Video-35th) 77 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 78 (1990-03-01) Final Report, Three Mile Island-Unit 2 Safety Advisory Board 79 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 80 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 81 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 82 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID 83 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 84 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980-07) 85 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 86 NUREG-0698, Rev. 2, NRC Plan for Cleanup Operations at TMI-2 (1984-03) 87 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 88 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID.
153

89 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 90 Masnik, Michael, (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 91 Travers, W.D., “United States Nuclear Regulatory Commission’s Regulatory Oversight 

of the Cleanup Operations at the Three Mile Island Unit 2 Station (1979-1989),”

International Symposium on Recovery Operations in the Event of a Nuclear Accident 

or Radiological Emergency. International Atomic Energy Agency (IAEA), Vienna. 

6–10 November 1989. IAEA: Vienna, Austria. IAEA-SM-316. 1990. 92 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 93 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980-07) 94 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 95 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 96 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 97 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 98 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 99 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 100 NUREG/CR-6252, Lessons Learned from the Three Mile Island-Unit 2 Advisory Panel 

(1994-08) 101 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 102 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 103 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 104 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 105 Barrett, Lake (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 106 Byrne, James (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 107 GEND-026, Addition of Soluble and Insoluble Neutron Absorbers to the RCS of TMI-2 

(1984-07) 108 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 109 (1985-09-27) GPU Hazards Analysis, Potential for Boron Dilution of Reactor Coolant 

System, Rev. 2 110 GEND-051, Evaluation of Special Safety Issues Associated with Handling the TMI-2 

Core Debris (1985-06) 111 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 112 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 113 (1990-03-01) Final Report, Three Mile Island-Unit 2 Safety Advisory Board 114 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990.
154

115 GPU Nuclear, “Planning Study - Core Conditions Design Basis,” GPU-TPO-TMI-151, 

Rev. 1, December 1986. Retrieved from https://libraries.psu.edu/about/collections/

three-mile-island-2-recovery-and-decontamination/three-mile-island-reports 116 GPU Nuclear, “Planning Study - Core Conditions Design Basis,” GPU-TPO-TMI-151, 

Rev. 1, December 1986. 117 GPU Nuclear, “Planning Study - Core Conditions Design Basis,” GPU-TPO-TMI-151, 

Rev. 1, December 1986. 118 GPU Nuclear, “Planning Study - Core Conditions Design Basis,” GPU-TPO-TMI-151, 

Rev. 1, December 1986. 119 GEND-051, Evaluation of Special Safety Issues Associated with Handling the TMI-2 

Core Debris (1985-06) 120 GEND-051, Evaluation of Special Safety Issues Associated with Handling the TMI-2 

Core Debris (1985-06) 121 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 122 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 123 NSAC-80-1, NSAC-1 Revised, Analysis of Three Mile Island Unit 2 Accident (1980-03) 124 GEND-007, Basis for Tool Development for Reactor Disassembly and Defueling 

(1981-05) 125 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 126 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 127 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 128 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 129 IAEA, “Catalogue of Methods, Tools and Techniques for Recovery from Fuel Damage 

Events,” IAEA-TECDOC-627, October 1991. 130 Croft, Kevin (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 131 Croft, Kevin (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 132 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools 

(1988-10) 133 GEND-054, TMI-2 Reactor Vessel Plenum Final Lift (1986-01) 134 GEND-054, TMI-2 Reactor Vessel Plenum Final Lift (1986-01) 135 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 136 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 137 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 138 Byrne, James (personal communications, May 2019) provided clarifications based on his 

experiences at TMI-2. 139 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 140 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990.
155

141 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 142 Skillman, Gordon R. (2014, March). “The 35th Anniversary of the Three Mile Island 

Nuclear Power Plant Accident of 1979: Working at TMI During and Following the 

Accident.” Video Seminar presented at the U.S. Nuclear Regulatory Commission, 

Rockville, MD. (Retrieved from NUREG/KM-0001 at 

https://tmi2kml.inl.gov/HTML/Search.html?folder=0d-Video-35th) 143 Cole, Norman M., “Preparation for the Cleanup: Assessing the Damage,” ANS Executive 

Conference on TMI-2: A Learning Experience, Hershey Lodge, 

October 14–16, 1985. 144 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016 145 Cole, Norman M., “Preparation for the Cleanup: Assessing the Damage,” ANS Executive 

Conference on TMI-2: A Learning Experience, Hershey Lodge, 

October 14–16, 1985. 146 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 147 (1982-07-06) GPU Safety Evaluation, Insertion Camera Through Reactor Vessel 

Leadscrew Opening, Rev. 2 148 (1982-07-13) NRC Review, Control Rod Drive Mechanism Quick Look Camera 

Inspection (re 07-06-1982) 149 Cole, Norman M., “Preparation for the Cleanup: Assessing the Damage,” ANS Executive 

Conference on TMI-2: A Learning Experience, Hershey Lodge, 

October 14–16, 1985. 150 GEND-007, Basis for Tool Development for Reactor Disassembly and Defueling 

(1981-05) 151 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 152 Standerfer, F.R., “Three Mile Island Unit 2: Plant Recovery,” Nuclear Technology, 

Vol. 87, No. 1, pp. 54-56, August 1989. 153 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 154 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 155 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 156 GEND-054, TMI-2 Reactor Vessel Plenum Final Lift (1986-01) 157 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 158 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 159 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 160 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 161 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989.
156

162 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 163 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 164 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 165 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 166 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 167 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 168 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 169 GEND-054, TMI-2 Reactor Vessel Plenum Final Lift (1986-01) 170 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 171 (1984-10-30) GPU Technical Evaluation, Containment Air Control Envelope Design 

Criteria (Rev. 4), TER (Rev. 2) 172 (1984-12-04) NRC Safety Evaluation, Containment Air Control Envelope Design 

(re 07-22-1983, 10-30-1984) 173 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 174 (1985-01-18) GPU Safety Evaluation, Removal of the Equipment Hatch, Rev. 3 175 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 176 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 177 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 178 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 179 IAEA, “Catalogue of Methods, Tools and Techniques for Recovery from Fuel Damage 

Events,” IAEA-TECDOC-627, October 1991. 180 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 181 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 182 GEND-INF-073, TMI-2 DEFUELING TOOLS ENGINEERING REPORT (1986-02) 183 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 184 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 185 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 186 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 187 (1985-10-29) GPU, Defueling Hydraulic System Fluid Testing (re 10-24-1985) 188 (1985-11-25) GPU, Defueling Hydraulic System Fluid, Precipitation of Boric Acid 

Crystals (re 10-29-1985)” 189 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990.
157

190 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 191 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 192 EGG-M-08986 (CONF), TMI-2 Core boring machine (1986) 193 EGG-TMI-7385-R1, TMI-2 Core Bore Acquisition Summary Report (1987-02) 194 EGG-TMI-7385-R1, TMI-2 Core Bore Acquisition Summary Report (1987-02) 195 (1987-08-20) NRC Review, Use of Plasma Arc Torch to Cut Upper End Fittings 

(re various letters) 196 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 197 McGough, Michael S., et al., “Performance of the Automated Cutting Equipment 

System During the Plasma Cutting of the TMI-2 Lower Core Support Assembly,”

Transactions of the American Nuclear Society Winter Meeting in Washington, DC, 

Vol. 57. November 1988, p.504. 198 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools 

(1988-10) 199 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools 

(1988-10) 200 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 201 McGough, Michael S., et al., “Performance of the Automated Cutting Equipment 

System During the Plasma Cutting of the TMI-2 Lower Core Support Assembly,”

Transactions of the American Nuclear Society Winter Meeting in Washington, DC, 

Vol. 57, November 1988, p.504. 202 (1988-09-14) TMIPO Monthly (see TMI Project Office Status Reports folder in 

NUREG/KM-0001) 203 (1988-09-14) TMIPO Monthly (see TMI Project Office Status Reports folder in 

NUREG/KM-0001) 204 (1988-09-14) TMIPO Monthly (see TMI Project Office Status Reports folder in 

NUREG/KM-0001) 205 (1988-09-14) TMIPO Monthly (see TMI Project Office Status Reports folder in 

NUREG/KM-0001) 206 (1987-06-25) GPU, Off-Gas Releases Generated During Plasma Arc Cutting 

(re 05-07-1987) 207 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 208 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 209 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 210 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 211 IAEA, “Catalogue of Methods, Tools and Techniques for Recovery from Fuel Damage 

Events,” IAEA-TECDOC-627, October 1991. 212 EGG-M-92038 (CONF), TMI defueling project fuel debris removal system (1992) 213 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 214 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 215 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03)
158

216 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 217 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 218 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 219 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 220 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 221 GEND-011, Canister-Design Considerations for Packaging of TMI-2 Damaged Fuel and 

Debris (1981-09) 222 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 223 (1985-03-06) TAAG, Tenth Report of the Technical Assistance and Advisory Group 

(TAAG) 224 (1985-03-06) TAAG, Tenth Report of the Technical Assistance and Advisory Group 

(TAAG) 225 (1985-10-03) GPU Response to NRC, Early Defueling SER (07-29, 08-13-1985) 226 Hayes, G.R, Marsden, J.F., “Quality Assurance in the Removal and Transport of the 

Three Mile Island Unit 2 Core,” Nuclear Technology, Vol. 87, No. 4, pp. 721–728, 

December 1989. 227 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 228 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 229 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 230 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 231 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 232 GEND-INF-073, TMI-2 Defueling Tools Engineering Report (1986-02) 233 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 234 (1985-05-24) B&W, TMI-2 Defueling Canisters Final Design Technical Report 235 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition 

that Apply to Fukushima (2012) 236 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 237 Hayes, G.R., Marsden, J.F., “Quality Assurance in the Removal and Transport of the 

Three Mile Island Unit 2 Core,” Nuclear Technology, Vol. 87, No. 4, pp. 721–728, 

December 1989. 238 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 239 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 240 Abbate, C.M., and J.W. Craig, “U.S. Nuclear Regulatory Commission Inspection of 

Transportation Casks,” Nuclear Technology, Vol. 87, No. 4, pp. 755-758, 

December 1989.
159

241 Abbate, C.M., and J.W. Craig, “U.S. Nuclear Regulatory Commission Inspection of 

Transportation Casks,” Nuclear Technology, Vol. 87, No. 4, pp. 755-758, 

December 1989. 242 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 243 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 244 PNO-I-88-055 (see Preliminary Notification folder in NUREG/KM-0001) 245 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 246 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 247 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 248 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 249 Hofstetter, K.J., "Identification and Control of Microorganisms at Three Mile Island 

Unit 2," Transactions of the American Nuclear Society Winter Meeting in 

Washington, DC, Vol. 57, November 1988, p.530. 250 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 251 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 252 Masnik, Michael (2016, October). Presentation at the Meeting on US-Japan TMI-2 

Knowledge Transfer and Relevance to Fukushima, Idaho Falls, ID. 253 (1986-05-05) GPU Safety Evaluation, Addition of a Biocide to the TMI-2 Reactor 

Coolant System, Rev. 0 254 (1986-05-06) NRC Review, Addition of a Biocide to the TMI-2 Reactor Coolant System 

(re 05-05-1986) 255 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 256 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 257 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 258 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest 

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 259 (1986-05-05) GPU Safety Evaluation, Addition of a Biocide to the TMI-2 Reactor 

Coolant System, Rev. 0 260 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island Unit 2,” Nuclear 

Technology, Vol. 87, No. 3, pp. 587-594, November 1989. 261 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 262 Broughton, J.M., E.L. Tolman, P. Kuan, D.A. Petti, "A Scenario of the TMI-2 

Accident," Transactions of the American Nuclear Society Winter Meeting in 

Washington, DC, Vol. 57, November 1988, p.395.
160

263 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 264 EPRI, “Radiation Protection Management Program at TMI- 2: Noteworthy Practices and 

Accomplishments, EPRI-NP-5338, August 1987. 265 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 266 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 267 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition 

that Apply to Fukushima (2012) 268 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 269 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools 

(1988-10) 270 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 271 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 272 PNO-TMI-85-011 (see Preliminary Notification folder in NUREG/KM-0001) 273 TMIPO 5-19-86 (see TMI Project Office Status Report folder in NUREG/KM-0001) 274 PNO-TMI-86-004 (see Preliminary Notification folder in NUREG/KM-0001) 275 PNO-I-88-055 (see Preliminary Notification folder in NUREG/KM-0001) 276 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 277 Licensee Event Report LER-89-006; NRC Preliminary Notification PNO-I-89-81 278 (1990-01-11) NRC, Notice of Violation, Over and Unplanned Exposures 279 (1990-01-11) NRC, Notice of Violation, Over and Unplanned Exposures 280 LER-89-005 (see Licensee Event Report folder in NUREG/KM-0001) 281 LER-89-001 (see Licensee Event Report folder in NUREG/KM-0001) 282 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 283 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 284 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 285 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 286 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 287 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 288 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 289 EGG-2529, Historical Summary of Fuel and Waste Handling and Disposition at TMI-2, 

1980-1988 (1988-10) 290 USNRC, “Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic 

Technical Activity A 36,” NUREG 0612, July 1980. 291 EGG-2529, Historical Summary of Fuel and Waste Handling and Disposition at TMI-2, 

1980-1988 (1988-10) 292 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 293 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 294 INL/LTD-11-22779, Rev. 1, Three Mile Island Core Debris Packaging, Transportation, 

and Storage Summary (2012-02)
161

295 EGG-2529, Historical Summary of Fuel and Waste Handling and Disposition at TMI-2, 

1980-1988 (1988-10) 296 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 297 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 298 INL/LTD-11-22779, Rev. 1, Three Mile Island Core Debris Packaging, Transportation, 

and Storage Summary (2012-02) 299 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 300 McGoff, D.J., et al., “The role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International Symposium on Recovery 

Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 301 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 302 G.R. Hayes, "QA in the Design and Fabrication of the TMI-2 Rail Cask," Fifteenth 

Annual American Society for Quality Control National Energy Division Conference, 

San Antonio, TX, 24-26 October 1988, 1988. 303 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 304 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition 

that Apply to Fukushima (2012) 305 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 306 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 307 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016 308 (1988-12-15) TMIPO Monthly (see TMI Project Office Status Report folder in 

NUREG/KM-0001) 309 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 310 PNO-TMI-87-02 (see Preliminary Notification folder in NUREG/KM-0001) 311 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 312 INL/LTD-11-22779, Rev. 1, Three Mile Island Core Debris Packaging, Transportation, 

and Storage Summary (2012-02) 313 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 314 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 315 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 316 GEND-034, Gross Decontamination Experiment Report (1983-07) 317 GEND-034, Gross Decontamination Experiment Report (1983-07) 318 (1982-09-23) GPU Safety Evaluation, Ongoing Containment Building Decontamination 

Activities, Rev. 0 319 GPU Nuclear, “Planning Study, Dose Reduction,” GPU-TPO-TMI-039, Rev. 0, 

March 1983. (From the personal file of a knowledge contributor.) 320 GPU Nuclear, “Technical Plan, TMI-2 Program Strategy,” GPU-TPO-TMI-115, 

Revision 0, June 1986. (From the personal file of a knowledge contributor.) 321 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10)
162

322 Daniels, R. S, “Three Mile Island Unit 2 Reactor Building Dose Reduction Task Force,” 

Transactions of the American Nuclear Society Winter Meeting in Washington, DC, 

Vol. 87, October 1989, p. 555. 323 GPU Nuclear, “Technical Plan, TMI-2 Program Strategy,” GPU-TPO-TMI-115, 

Revision 0, June 1986. (From the personal file of a knowledge contributor.) 324 GPU Nuclear, “Technical Plan, TMI-2 Program Strategy,” GPU-TPO-TMI-115, 

Revision 0, June 1986. (From the personal file of a knowledge contributor.) 325 GPU Nuclear, “Technical Plan, TMI-2 Program Strategy,” GPU-TPO-TMI-115, 

Revision 0, June 1986. (From the personal file of a knowledge contributor.) 326 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report 327 Bengel, P. R., et al., "Decontamination Task Force Report," GPU Nuclear, 

December 18, 1985. 328 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 329 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 330 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 331 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 332 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 333 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 334 GEND-034, Gross Decontamination Experiment Report (1983-07) 335 GEND-034, Gross Decontamination Experiment Report (1983-07) 336 GEND-034, Gross Decontamination Experiment Report (1983-07) 337 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 338 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 339 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 340 GEND-INF-054, ANALYSES ON CONCRETE CORES FROM FLOORS AND 

D-RING WALLS OF RB (1984-06) 341 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 342 GEND-034, Gross Decontamination Experiment Report (1983-07) 343 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 344 GEND-034, Gross Decontamination Experiment Report (1983-07) 345 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 346 GEND-032, TMI-2 Technical Information and Examination Program Annual Report 

1982 (1983-04) 347 GEND-032, TMI-2 Technical Information and Examination Program Annual Report 

1982 (1983-04) 348 (1984) Process in the Recovery Operations at TMI-2 (Nuclear Safety) 349 (1986) Process in the Recovery Operations at TMI-2, An Update (Nuclear Safety) 350 (1983-01-21) GPU, TMI-2 Recovery Program Estimate, Rev. 2 (re 12-30-1982)
163

351 GPU Nuclear, “Planning Study, Dose Reduction,” GPU-TPO-TMI-039, Rev. 0, 

March 1983. (From the personal file of a knowledge contributor.) 352 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 353 GPU Nuclear, “Planning Study, Dose Reduction,” GPU-TPO-TMI-039, Rev. 0, 

March 1983. (From the personal file of a knowledge contributor.) 354 Hildebrand, J.E., “The Three Mile Island Unit 2 Recovery: A Decade of Challenge.”

International Symposium on Recovery Operations in the Event of a Nuclear Accident 

or Radiological Emergency. International Atomic Energy Agency (IAEA), Vienna. 

6–10 November 1989. IAEA: Vienna, Austria. IAEA-SM-316. 1990. 355 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 356 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools 

(1988-10) 357 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 358 GEND-034, Gross Decontamination Experiment Report (1983-07) 359 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 360 GEND-034, Gross Decontamination Experiment Report (1983-07) 361 GEND-034, Gross Decontamination Experiment Report (1983-07) 362 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 363 (1986-03-28) GPU Safety Evaluation, Reactor Building Decontamination and Dose 

Reduction Activities for 1986, Rev. 0 364 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, 1992. 365 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 366 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 367 EPRI, "Decontamination with a Steam-Water Supply and Vacuum Removal System," 

Technical Brief from the Electric Power Research Institute, Nuclear Power Division, 

TB NPD 18.2.86, 1986. 368 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 369 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 370 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 371 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 372 (1986-03-28) GPU Safety Evaluation, Reactor Building Decontamination and Dose 

Reduction Activities for 1986, Rev. 0 373 GEND-034, Gross Decontamination Experiment Report (1983-07) 374 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, 1992. 375 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 376 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 377 GEND-034, Gross Decontamination Experiment Report (1983-07) 378 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, 1992. 379 GEND-034, Gross Decontamination Experiment Report (1983-07) 380 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 381 McGoff, D.J., et al., “The role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International Symposium on Recovery 

Operations in the Event of a Nuclear Accident or Radiological Emergency. 
164

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 382 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 383 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 384 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 385 GEND-INF-054, Analyses on Concrete Cores from Floors and D-Ring Walls of RB 

(1984-06) 386 GEND-INF-081, Examination of Concrete Samples From the TMI-2 Reactor Building 

Basement (1987-02) 387 GEND-INF-081, Examination of Concrete Samples From the TMI-2 Reactor Building 

Basement (1987-02) 388 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 389 GEND-INF-081, Examination of Concrete Samples From the TMI-2 Reactor Building 

Basement (1987-02) 390 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990 391 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 392 GEND-034, Gross Decontamination Experiment Report (1983-07) 393 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 394 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 395 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 396 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 397 (1983-09-29) GPU Safety Evaluation, Ongoing Reactor Building Decontamination-Dose 

Reduction, Rev. 0 398 C.W. Hultman and R.W. Jackson, "TMI-2 Reactor Vessel Head and Plenum Removal,” 

Transactions of the American Nuclear Society Winter Meeting in Washington, DC, 

Vol. 57, November 1988, p.491. 399 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 400 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 401 EPRI, “Decontamination Experience During the Cleanup of Three Mile Island Unit 2,” 

EPRI-NP-7157, 1990. 402 Hildebrand, J.E., “The Three Mile Island Unit 2 Recovery: A Decade of Challenge.”

International Symposium on Recovery Operations in the Event of a Nuclear Accident 

or Radiological Emergency. International Atomic Energy Agency (IAEA), Vienna. 

6–10 November 1989. IAEA: Vienna, Austria. IAEA-SM-316. 1990. 403 McGoff, D.J., et al., “The role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International symposium on recovery 

operations in the event of a nuclear accident or radiological emergency. International 

Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: Vienna, 

Austria. IAEA-SM-316. 1990. 404 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)
165

405 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10) 406 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 407 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 408 Tarpinian, James (personal communications, April 2020) provided clarifications based 

on his experiences at TMI-2. 409 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 410 (1985-12-27) GPU, Corporate Radiation Protection Plan, Rev. 1 411 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 412 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, 1992. 413 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011.

414 Hilderbrand, James E., “Personnel Contamination Protection Techniques Applied 

During the Three Mile Island Unit 2 Cleanup,” Transactions of the American Nuclear 

Society Winter Meeting in Washington, DC, Vol. 87, October 1989, p. 531. 415 Tarpinian, James (personal communications, April 2020) provided clarifications based 

on his experiences at TMI-2. 416 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, 1992. 417 Lodde, G.M. et al, “Assessment and Control of the Three Mile Island Unit 2 Reactor 

Building Atmosphere,” Transactions of the American Nuclear Society Winter 

Meeting in Washington, DC, Vol. 87, October 1989, p. 535. 418 GEND-013, TMI-2 Reactor Building Purge-Kr-85Venting, (March 1981 419 IAEA, “Cleanup and Decommissioning of a Nuclear Reactor after a Severe Accident, 

IAEA-TRS-346, 1992. 420 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 421 Schork, J. S., B. A. Parfitt, "Heat Stress Control in the TMI-2 Defueling and 

Decontamination Activities," Transactions of the American Nuclear Society Winter 

Meeting in Washington DC, Vol. 57, November 1988, p. 467. 422 Hildebrand, J.E., “The Three Mile Island Unit 2 Recovery: A Decade of Challenge.”

International Symposium on Recovery Operations in the Event of a Nuclear Accident 

or Radiological Emergency. International Atomic Energy Agency (IAEA), Vienna. 

6–10, November 1989. IAEA: Vienna, Austria. IAEA-SM-316. 1990. 423 Schork, J. S., B. A. Parfitt, "Heat Stress Control in the TMI-2 Defueling and 

Decontamination Activities,” Transactions of the American Nuclear Society Winter 

Meeting in Washington, DC, Vol. 57, November 1988, p. 467. 424 Schork, J. S., B. A. Parfitt, “Heat Stress Control in the TMI-2 Defueling and 

Decontamination Activities,” Transactions of the American Nuclear Society Winter 

Meeting in Washington, DC, Vol. 57, November 1988, p. 467. 425 Schork, J. S., B. A. Parfitt, “Heat Stress Control in the TMI-2 Defueling and 

Decontamination Activities,” Transactions of the American Nuclear Society Winter 

Meeting in Washington, DC, Vol. 57, November 1988, p. 467. 426 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 427 Negin, Charles (personal communications, January 2019) provided clarifications based 

on his experiences at TMI-2. 428 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990.
166

429 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 430 Tarpinian, James (personal communications, April 2020) provided clarifications based 

on his experiences at TMI-2. 431 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 432 EPRI, “TMI-2 Technology Transfer Progress Report,” EPRI-NP-5821, May 1988. 433 Hildebrand, James E., "Personnel Contamination Protection Techniques Applied During 

the TMI-2 Cleanup," Transactions of the American Nuclear Society Winter Meeting 

in Washington DC, Vol. 57, November 1988, p. 531. 434 Hildebrand, J.E., “The Three Mile Island Unit 2 Recovery: A Decade of Challenge.”

International Symposium on Recovery Operations in the Event of a Nuclear Accident 

or Radiological Emergency. International Atomic Energy Agency (IAEA), Vienna. 

6–10 November 1989. IAEA: Vienna, Austria. IAEA-SM-316. 1990. 435 Tarpinian, James (personal communications, April 2020) provided clarifications based 

on his experiences at TMI-2. 436 GEND-044, TMI-2 Reactor Vessel Head Removal (1985-09) 437 1980 NRC Annual Report (NUREG-0774) 438 USNRC, Letter from NRC Executive Director for Operations to Senator Specter

Regarding the Philadelphia Inquirer Article on the TMI-2 Cleanup, Letter Dated 

February 26, 1985. (ADAMS Accession No. 8503070122) 439 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 440 Holton, W.C., C.A. Negin, R.W. Lambert, “The Technical History of the Three Mile 

Island Unit 2 Cleanup: Factors, Options and Decisions.” International Symposium on 

Recovery Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 441 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 442 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 443 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 444 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 445 (1979-05-26) NRC Policy, NRC to Advise Use of EPICOR II After Public Comment 446 McGoff, D.J., et al., “The role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International Symposium on Recovery 

Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 447 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 448 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 449 Skillman, Gordon R. (2014, March). “The 35th Anniversary of the Three Mile Island

Nuclear Power Plant Accident of 1979: Working at TMI During and Following the 

Accident.” Video Seminar presented at the U.S. Nuclear Regulatory Commission, 

Rockville, MD. (Retrieved from NUREG/KM-0001 at 

https://tmi2kml.inl.gov/HTML/Search.html?folder=0d-Video-35th) 450 USNRC, “Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas 

Generation,” Information Notice 84-72, September 1984. 451 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992.
167

452 McGoff, D.J., et al., “The role of the United States Department of Energy in Recovering 

from the Three Mile Island Accident.” International Symposium on Recovery 

Operations in the Event of a Nuclear Accident or Radiological Emergency. 

International Atomic Energy Agency (IAEA), Vienna. 6–10 November 1989. IAEA: 

Vienna, Austria. IAEA-SM-316. 1990. 453 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 454 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 455 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 456 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 457 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 458 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 459 Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 460 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 461 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 462 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 463 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 464 EGG-TMI-6304, Design Analysis, High-Integrity Container for Disposal of EPICOR-II 

Prefilter Liners (1983) 465 GEND-045, Disposal Demo of High Integrity Container with an EPICOR II Prefilter 

from TMI (1985) 466 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992. 467 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 468 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 469 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 470 EPRI, “TMI-2 Post-Accident Data Acquisition and Analysis Experience,” 

EPRI-NP-7156, 1992. 471 GEND-021, Controlled Air Incinerator Conceptual Design Study (1982-01) 472 GEND-021, Controlled Air Incinerator Conceptual Design Study (1982-01) 473 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 474 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” 

IAEA-TRS-321, 1991. 475 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 476 USNRC, “Three Mile Island Accident of 1979 Knowledge Management Digest:

Recovery and Cleanup,” NUREG/KM-0001, Supplement 1, June 2016. 477 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990. 478 EPRI, “Contamination and Decontamination Experience with Protective Coatings at 

TMI-2,” EPRI-NP-5206, 1987. 479 IAEA, “Recovery Operations in the Event of a Nuclear Accident or Radiological 

Emergency,” Proceedings of a Symposium, IAEA Report STI/PUB/826, 1990.
168

480 IAEA, “Management of Severely Damaged Nuclear Fuel and Related Waste,” IAEA 

Report TRS 321, 1991. 481 Endnotes of references to selected figures: 1. (1982-11-29) GPU, Proposed Organization Plan Revision No. 6 2. GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982-10) 3. Tarpinian, J.E., “Three Mile Island Unit-2 D&D” [PowerPoint Presentation], IAEA 

Workshop on Approaches to Decommissioning Works in Highly Contaminated or 

High Dose Areas, Vienna, Austria, 5-8 October 2015. Provided by the Author. 4. NUREG-0698, Rev. 2, NRC Plan for Cleanup Operations at TMI-2 (1984-03) 5. Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 6. GEND-INF-075-PT-1, TMI-2 Core Debris Grab Sample Examination and Results 

(1986-09) 7. GEND-007, Basis for Tool Development for Reactor Disassembly and Defueling 

(1981-05) 8. EGG-TMI-7385-R1, TMI-2 Core Bore Acquisition Summary Report (1987-02) 9. GEND-INF-073, TMI-2 Defueling Tools Engineering Report (1986-02) 10. (1983-07-22) GPU Technical Evaluation, Containment Air Control Envelope (with 

Design Criteria) 11. GEND-INF-065, TMI-2 Defueling System Design Description (1985-03) 12. GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools 

(1988-10) 13. INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and 

Disposition that Apply to Fukushima (2012) 14. DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03) 15. GEND-034, Gross Decontamination Experiment Report (1983-07) 16. (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage 

Safety Analysis Report 17. Hildebrand, J.E., “The Three Mile Island Unit 2 Recovery: A Decade of Challenge.” 

International Symposium on Recovery Operations in the Event of a Nuclear 

Accident or Radiological Emergency. International Atomic Energy Agency (IAEA), 

Vienna. 6–10 November 1989. IAEA: Vienna, Austria. IAEA-SM-316. 1990. 18. Green, Russ, et al., “TMI Unit 2 Post-accident Defueling: Experience and Lessons 

Learned,” Proceedings of GLOBAL 2011: 10th International Conference, Toward 

and Over the Fukushima Daiichi Accident, Chiba (Japan), 11–16 December 2011. 

(Photographs courtesy of the lead author who is shown being dressed in the left 

photograph.) 19. GPU Nuclear, “Three Mile Island Occupational Radiation Exposure Report: 

1979-1988,” Undated. (From the personal files of a knowledge contributor.) 20. GEND-INF-025, Development of a Prototype Gas Sampler for EPICOR II Prefilter 

Liners (1982-09) 21. GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water 

(1983-02) 22. (1992-05-12) GPU Technical Evaluation, Processed Water Disposal System, Rev. 4 23. GEND-045, Disposal Demo of High Integrity Container with an EPICOR II 

Prefilter from TMI (1985-02) 24. DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation 

Campaign (1993-03)
169


CONVERSIONS

Radiation Dose

1 mrem (1 millirem, 10-3) = *10 microsieverts (10 μSv, 10-5) 

100 mrem = *1 millisievert (1 MSv)

1 rem = *10 mSv

100 rem = *1 Sv

Radioactive Concentration 

27 picocuries (27 pCi, 2.7×10-11) = *1 becquerel (1 Bq)

1 millicurie (1 mCi, 0.001) = *37 megabecquerels (37 MBq, 3.7×107) 

1 curie (1 Ci) = *37 gigabecquerels (37 GBq, 3.7×1010) 

Radiation Absorbed Energy

1 roentgen = *0.877 rad = *0.00877 gray (Gy) 

100 rad = *1 Gy

Length 

1 inch (in.) = *2.54 centimeters (cm)

1 foot (ft) = 0.3048 meter (m)

Volume and Weight

1 gallon (gal) = 3.7854 liters (L) 

1 pound (lb) = 0.4536 kilograms (kg)

1 ton (U.S.) = *2000 lb = 907.1847 kg

Pressure

1 pound per square inch (psi) = 6.8948 kilopascals (kPa)

1 atmosphere (atm) = *101.325 kPa

Temperature 

Degrees Celsius (°C) = 5/9 × (°F - 32) 

Degrees Fahrenheit (°F) = (9/5 × °C) + 32

* Exact conversion factor

NUREG/KM-0001, Supplement 2 

December 2020