Three Mile Island

Accident of 1979

Knowledge

Management Digest

Recovery and Cleanup

NUREG/KM-0001, Supplement 1

Office of Nuclear Regulatory Research
TMI-2 radioactive material location map (see GEND-057).
iii

x 

Contents

Digital Versatile Disc (DVD) Contents ............................................. v

1 Introduction ............................................................................... 1

2 Management and Oversight ...................................................... 7

3 Plant Stabilization.................................................................... 41

4 Worker Protection.................................................................... 69

5 Data Acquisition and Analysis ................................................. 81

6 Waste Management .............................................................. 101

7 Decontamination ................................................................... 131

8 Defueling............................................................................... 147

9 After Defueling ...................................................................... 179

10 Postscript .............................................................................. 195

11 DVD Navigation and Interpretation........................................ 199

12 Contributions to Supplement 1 .............................................. 203

13 Endnotes............................................................................... 205

On April 15, 1990, the last rail shipment of fuel debris packaged in 

three Model 125-B shipping casks departs Three Mile Island for the 

Department of Energy’s Idaho National Engineering Laboratory.
iv

Above: First entry into the reactor building (shown) occurred on 

July 23, 1980. Heavy duty outerwear provided high-energy beta ray protection. 

Below: Water level in the basement of the reactor buidling reached the first 

stairwell landing indicating about 8 feet of water accumulation. One-third of 

this highly radioactive water came from leakage of non-radioactive river water 

coolant from a leaking reactor building air-handling cooler.
v 

Digital Versatile Disc (DVD) Contents

Supplement 1, Disc 1 (NUREG/KM-0001, DVD-2) Folder

Status and Summary Reports

TMI Program Office Weekly Status Reports 4a-TMIPO

NRC Preliminary Notifications 4b-PN

Licensee Event Reports 4c-LER

NRC Status Reports 4d-NRC

NRC Annual Reports 4e-Annual 

DOE/National Laboratory Status Reports 4f-DOE

DOE/National Laboratory Conference Papers 4g-DOE-CONF

GPU Status Reports 4h-GPU

Other Status Reports 4i-Other 

Licensing Actions

Operating License Amendments 5a-OL Amend

Orders 5b-Order

Order Amendments 5c-Order-Amend

Regulation Exemptions 5d-Exemption 

Recovery Operations Plan Changes 5e-ROP-Change 

Organization Plan Changes 5f-OP-Rev

Management and Oversight

General Management and Oversight 6a-Plans

Commission/Licensee Correspondence 6b-Commission

Commission SECY Papers/Staff Requirements 6c-SECY

Advisory Groups 6d-Advisory

Guidance-PEIS 6e-Guide-PEIS

General Project Design Criteria 6f-Design

One of many attachments to long-handled defueling tools used to move core 

debris in the reactor vessel.
vi

Supplement 1, Disc 2 (NUREG/KM-0001, DVD-3) Folder

Plant Stabilization 

General Plant Stabilization 7a-Stabilization

Planning Meetings 7b-Meeting

Recovery Procedures 7c-Procedure 

Industry Advisory Group Reports 7d-IAG-Report 

Reactor Building Entries 7e-RB-Entries

Worker Protection 8a-Protection 

Data Acquisition and Analyses

Recovery Data Analyses 9a-Data-Recovery

Accident Data Analyses 9b-Data-Accident

Instrumentation and Electrical Evaluations 9c-Data-Equipmt 

Criticality Analyses 9d-Criticality

Decontamination 11a Decon 

Supplement 1, Disc 3 (NUREG/KM-0001, DVD-4) Folder

Waste Management 

General Waste Management 10a-Waste

Water Processing: EPICOR 10b-EPICOR

Water Processing: SDS 10c-SDS

Solid Waste 10d-Solid

Defueling 12a-Defuel

After Defueling 13a-Post-Defuel

Pre-Accident: Plant References

Final Safety Analysis Report 14a-FSAR

System Descriptions 14b-Sys-Desc

Plant Procedures 14c-Procedure 

Other Plant References 14d-Other
vii

Above: Close up view of the upper grid rib section where fuel bundles were 

once held in place. Indication of melting of the stainless steel rids was 

observed during an inspection of the underside of the upper plenum as it 

was lifted from the reactor vessel. Below: Broken fuel rods in the void area 

at the top of the damaged reactor core.
viii

Oyster Creek Unit 1 (left) and the proposed Unit 2. In 1968, Jersey Central Power and Light Company applied for a reactor construction permit for the proposed unit. Instead, the plant was built on Three Mile Island adjacent to Unit 1. The new application included Metropolitan Edison and Pennsylvania Electric as additional owners.
1 

1 Introduction

The safe, expeditious recovery and cleanup of Three Mile Island 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 cleanup campaign ensured that the TMI site did not 

become a long-term or permanent waste repository.

1 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 multi-organizational effort that 

included hundreds of dedicated and highly-skilled individuals. 

Implementation of recovery and cleanup activities was the responsibility of 

the licensee with support from their many contractors. Organizations that 

supported the licensee included the original architect engineers of TMI 

Units 1 and 2 (Gilbert Associates and Burns and Roe, respectively2

); the 

TMI-2 nuclear steam supply system vendor (Babcock & Wilcox); many 

volunteers from other nuclear power plants; the U.S. nuclear industry; and 

several international organizations.3

The U.S. Nuclear Regulatory Commission (NRC) was responsible for the 

regulation of TMI-2 cleanup operations to ensure the health and safety of 

the public, and the TMI-2 occupational workforce, as well as the protection 

of the environment.4

 The U.S. Department of Energy (DOE) was 

responsible for supporting an extensive research program, as directed by the 

U.S. Congress. In addition, DOE agreed to the removal and disposition of 

the entire reactor core for research, as well as certain solid nuclear wastes 

generated during the cleanup of TMI-2.5, 6, 7, 8

 The DOE and its national 

laboratories provided much-needed technical support to the licensee and the

NRC in almost every aspect of the many TMI-2 research and recovery 

programs.9 The environmental agencies and nearby communities from the 

states of Pennsylvania and Maryland were responsible for ensuring that 

water qualities of the Susquehanna River and the downstream Chesapeake 

Bay were not adversely impacted by the damaged plant, or by the cleanup 

activities.10

Need for Cleanup. The NRC’s Programmatic Environmental Impact 

Statement (PEIS) related to decontamination and disposal of radioactive 

waste resulting from the TMI accident (NUREG-0683) concluded that the 

decontamination of the TMI-2 facility, including the removal of the nuclear

fuel and radioactive waste from the TMI site, was necessary for the 

long-term protection of public health and safety. The PEIS also concluded 

that methods existed, and could be suitably adapted to perform the cleanup 

operations with minimal releases of radioactivity to the environment.11
2 

The cleanup operations removed sources of potential radiation exposure that 

posed risks to the health and safety of the workers and the public. 

Accident-generated radiation sources were present in the form of airborne 

contamination; contaminated waste water; absorption of radioactive material 

on building and equipment surfaces; contaminated sludge in tanks, sumps, 

and building basement floors; contaminated filter cartridges and 

demineralizer resins; and damaged fuel and reactor components. As long as 

radioactive water occupied sumps and tanks, there existed a possibility of 

leakage into the groundwater, and subsequently, into the Susquehanna 

River. The contaminated water was also a source of direct radiation to 

workers requiring access to buildings in order to perform critical 

maintenance and repairs needed to keep the reactor in a safe-shutdown 

condition.12

The PEIS categorized cleanup into four fundamental activities: building and 

equipment decontamination; fuel removal and decontamination of the 

TMI-2 reactor building: a comparison (approximately to scale). The TMI-2 

reactor building is 130 feet in diameter and rises 191 feet from the floor to the 

dome, with a volume of about 2.4 million cubic feet. The U.S. Capitol Rotunda 

is 96 feet in diameter and rises 180 feet from the floor to the canopy, with a 

volume of about 1.3 million cubic feet.
3 

primary coolant system; treatment of radioactive liquids; and packaging, 

handling, storage, and transportation of radioactive solid wastes.

13

About This Supplement. The main objective of this supplement is to 

provide key historical documents in electronic format that were issued

during the recovery and cleanup efforts. Brief overviews of various 

structures, systems, equipment, and activities that were associated with the 

recovery and cleanup of TMI-2 are provided in the written portion of this 

NUREG/KM to describe the contents of the many document collections in 

the accompanying DVDs. 

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

learned, can be found in the Electric Power Research Institute (EPRI) report,

“The Cleanup of Three Mile Island Unit 2, A Technical History: 

1979 to 1990” and the special volume of the Nuclear Technology journal of 

the American Nuclear Society documenting 138 papers presented at the 

TMI-2 topical meeting in 1988.a

The seven major aspects of the recovery and cleanup, as presented in the

EPRI report, 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; an additional section on after-defueling 

activities follows. This supplement chronicles those activities, which began 

a week following the accident, and ended with the completion of disposal of 

accident-generated water, and entry into post-defueling monitored storage 

in 1993.

The document collections in this supplement were mainly derived from 

publicly available correspondence, including attached reports, between the 

licensee and the NRC, and the results of research activities sponsored by the 

NRC and DOE. In all, the accompanying DVDs contain about 

4,000 documents. Although an attempt was made to find and include a wide 

range of key documents, the collections provided on the DVDs are not 

complete. As such, a document collection might not provide a complete 

chronology of recovery, cleanup, and regulatory actions. A listing of 

documents in each document-collection folder is provided in spreadsheet 

format on each DVD (see the DVD folder, Common). Also, a list of more 

 

a EPRI-NP-6931 and many of EPRI’s historical reports on the TMI-2 accident and cleanup 

are currently (at the time of this publication) available from the EPRI’s website. Individual 

papers from Nuclear Technology, Vol. 87, Nos. 1 through 4, are currently available from the 

American Nuclear Society’s website. 
4 

than 25,000 TMI-2 records indexed in the Public Legacy Library, dating 

from 1979 to 1999, is included in the spreadsheet.b

How to Use this Supplement. A few suggestions for navigating through 

this supplement, and the many documents on the enclosed DVDs, are 

provided at the end of this NUREG/KM (see the section on DVD

Navigation and Interpretation). Documents in the DVD folders, Status and 

Summary Reports, Licensing Actions, and Management and Oversight, 

might be applicable to all sections in this supplement. The historical 

documents provided on the DVDs are for reference only, and are not official 

NRC records. End notes to this supplement provide filenames of the cited 

documents on the DVDs. The units of measure used in this NUREG/KM 

reflect those used in the original source reports. In some cases, conversions 

were provided in the original reports. Refer to the back cover for conversion 

factors and formulas.

 

b Public Legacy Library of the NRC Agencywide Documents Access and Management 

System (ADAMS) provides bibliographic citations for earlier documents during the period 

from 1979 through 1999, which are available in microfiche formats. This library can be 

viewed from the NRC public website.

TMI-2 reactor building under construction. Shown is the top dome of the 

reactor building. The Susquehanna River can be seen in the background. 
5 

Installation of a reactor vessel and steam generators inside containment 

similar to the TMI-2 design. Concrete shielding structures will be build 

around these components. Notice the ladder next to the reactor vessel. (From 

a training visual aid from the Babcock & Wilcox Company, the fabricator of 

the primary reactor components at TMI-2.)
6 

Above: NRC’s Harold Denton (left) and Victor Stello briefing President Jimmy Carter, 

Governor Richard Thronburgh, and others at the Air National Guard Facility in 

Middletown, Pennsylvania. Below: Senior managers from the Office of Nuclear 

Reactor Regulation in the NRC office trailer at Three Mile Island. Left-to-right: Roger 

Mattson, Harold Denton, Denwood Ross (on telephone), and Victor Stello.
7 

2 Management and Oversight

The NRC was responsible for the regulation and oversight of TMI-2 cleanup 

operations to ensure the health and safety of the public, and the TMI-2 

occupational workforce, as well as the protection of the environment. 

Implementation of recovery and cleanup activities was the responsibility of 

the licensee with support from their many contractors.14 NRC’s involvement 

covered two major areas: approving the recovery methods employed by the 

licensee and responding to public concerns over radiation exposure resulting 

from the accident and cleanup activities.15 In addition, the TMI-2 license 

remained under the regulatory requirements of an operating reactor (Title 

10, “Energy,” of the Code of Federal Regulations (10 CFR) Part 50, 

“Domestic Licensing of Production and Utilization Facilities”); many 

formal licensing actions were required to modify the pre-accident operating 

license to reflect the safety requirements of the plant in its unique shutdown 

condition.16 

Initially, the NRC did not have any specific guidelines or criteria pertaining 

to the recovery effort but followed its traditional regulatory practices of 

reacting to licensee proposals. This process sometimes required a time 

consuming and iterative exchange of written correspondence for questions 

and answers. Some actions required the majority approval of the NRC 

Commissioners, even for minor technical and radiological problems.17 At 

the end of the first year after the accident, it became apparent that a new 

approach was needed for maintaining the TMI-2 license during the cleanup 

and NRC oversight of the licensee. This approach consisted of an increased

onsite presence, as well as a new decision-making process for NRC 

approval of cleanup activities.18 The PEIS provided the environmental 

impact bases for all cleanup activities; the Commission policy statement that 

endorsed the PEIS provided staff with the authority to approve most cleanup 

activities.19 

The NRC accomplished its regulatory responsibilities for all post-accident 

operations at TMI 2 through licensing actions; safety evaluations of 

recovery and cleanup activities; inspections; daily interactions with the 

licensee and their contractors; communications with State and local 

governments and the public; coordination with other Federal agencies 

involved in the cleanup; and sometimes direction from the NRC 

Commissioners by majority vote. Formal licensing actions were required for 

proposed changes to NRC orders, the facility license, technical 

specifications, the recovery operations plan, the organization plan, and 

exemptions to regulations.20
8 

The licensee’s organizational structure and functions changed as the cleanup 

progressed through various stages of the recovery effort. Plans, schedules, 

resources, and cost estimates changed as more information became known 

about the condition of the damaged reactor core; reliability of plant systems 

and structures; and radiological characterization of accident-generated 

water, building structures, reactor components, and systems. Various 

organizations formed working groups to provide guidance on tackling 

specific issues, problems, and research activities. Independent oversight 

groups reviewed, monitored, and advised on the overall direction of 

recovery and cleanup plans and activities.21 NRC staff attended meetings 

conducted by the licensee’s working groups and independent oversight 

groups.22

The licensee and its contractors made changes to the facility in order to 

stabilize the reactor and clean up the damaged plant. New facilities and 

systems were built, and existing ones were used in different capacities than 

Side view of reactor coolant system components. Shown are the reactor vessel 

(center), both once-through steam generators (left and right), one of four 

reactor coolant pumps (upper left), both hot-legs (or “candy canes”) from the 

reactor vessel to top of the steam generators, and one of four cold-legs (or “Jlegs”) from the bottom of the steam generators to the reactor coolant pumps.
9 

those for which they were originally designed and approved. The changing 

configuration and operation of the damaged plant during the cleanup stages 

required changes to licensing documents. These changes imposed new 

requirements to ensure safety and eliminated pre-accident requirements that 

no longer applied to the shut-down plant.23 To implement these changes, the 

licensee developed formal documents such as plans, system description 

reports, safety analysis reports, technical evaluation reports, environmental 

assessment reports, experiments, and procedures. Formal documentation 

was required for most, if not all, data-collection and cleanup activities. 

Report of the Special Task Force on TMI Cleanup. Eight weeks before 

the first anniversary of the accident, a special NRC task force reported a 

variety of resource, policy, legal, and technical factors that could adversely 

influence, if left unchanged, the forward motion of the cleanup process and 

schedule. The task force observed frustration with the pace of the cleanup; 

the lack of interim criteria for the conduct of daily activities pending 

completion of the Programmatic Environmental Impact Statement (PEIS); 

the tedious NRC decision-making process; erosion of what was once a 

high-priority program; and strong public opinion on cleanup activities. In 

their report to the NRC Executive Director of Operations (EDO) on 

February 28, 1980, the task force made 15 recommendations stemming from 

their quick (9-day) evaluation of cleanup operations. 

The next day, the report was endorsed by the EDO and forwarded to the 

Commissioners with commitments to: (1) establish conservative interim 

criteria and a proper level of authority for staff decisions to permit small 

radiological releases associated with plant maintenance and data-gathering 

activities pending PEIS completion; (2) expedite the completion of 

assessments supporting the purging of radioactive gases from the reactor 

building and the completion of the PEIS; (3) increase permanent staffing of 

the NRC’s Middletown office and onsite support group, including a fulltime spokesperson to keep State and local government officials and the 

public informed on a continuing basis about the progress and the status of 

cleanup operations and about future plans; (4) transfer assignments to 

prepare environmental assessments to headquarters; and (5) prepare a plan 

and a schedule of activities for proceeding with the cleanup.24

The NRC Commissioners took actions that addressed other concerns 

expressed by the Special Task Force. The Commissioners approved by 

majority vote the interim criteria for radiological releases from data 

gathering and maintenance activities (SECY-80-175 and NUREG-0681), a 

contingency plan for cleanup in case of the financial failure of the licensee 

(SECY-80-416A and NUREG-0689), and the formation of the NRC 
10

Advisory Panel for the Decontamination of TMI-2. In their April 27, 1981 

policy statement that endorsed the PEIS, the Commissioners stated how and 

by whom major cleanup decisions would be made after the PEIS was 

complete, as well as the expected role of the PEIS in making those 

decisions. The Commissioners also stated that the licensee should accelerate 

the pace of the cleanup and that they expected to receive continuous advice 

from the TMI-2 Advisory Panel on major activities required to accomplish 

expeditious and safe cleanup of the TMI-2 facility.25

Document collections provided under the Management and Oversight 

category include Commission Actions, Licensing Actions, Planning and 

Guidance, and collections from the Advisory Groups and from the NRC’s 

TMI Project Office. Key documents in these document collections are 

summarized below. 

Isometric view of the reactor coolant system components.
11

Commission Actions. The five NRC Commissioners (also known 

collectively as the “Commission”) sometimes took actions to direct NRC 

staff to take certain licensing actions relating to policy issues. Issues before 

the Commission are decided by majority vote. The Commission’s decisionmaking tools used during the TMI-2 recovery and cleanup included policy 

statements, Commission orders, SECY papers, staff requirement 

memoranda, and general correspondence. 

• Policy statements. A “Statement of Policy” is not a regulation and does 

not impose specific regulatory requirements, but rather, provides the 

Commission’s rationale and motivation for future regulatory positions. 

Several policy statements that the Commission issued were directly 

applied to TMI accident investigation and recovery activities. These 

Commission policy statements (provided in the DVD folder, Policy 

Statements) are summarized below: 

o On May 25, 1979, the Commission directed staff to prepare an 

environmental assessment, with opportunity for public comment, 

regarding proposals to decontaminate and dispose of radioactive

contaminated waste water. The Commission’s statement required 

assessments on decontamination of intermediate-level waste water 

using the EPICOR II system; alternatives to discharge of waste 

water into the Susquehanna River; and decontamination and 

disposal of high-level waste water. The statement permitted 

discharge of pre-accident waste water decontaminated by the 

The five NRC Commissioners participated in an all-day public hearing in 

Harrisburg, PA on November 9, 1982. Residents and community groups in the 

Three Mile Island area were invited to express their views and concerns 

regarding the future of TMI-2. Shown (left to right), Commissioners Roberts 

and Ahearne, Chairman Palladino, and Commissioners Gilinsky and 

Asselstine.
12

existing EPICOR I decontamination system and discharge of 

industrial waste water (water slightly contaminated because of

leakage from secondary plant service support systems), 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 discharge of other 

(accident) waste water, and the operation of EPICOR II.26

o On November 21, 1979, the Commission directed staff to prepare a 

programmatic environmental impact statement (PEIS) on the 

decontamination and disposal of radioactive waste resulting from 

the accident.27 There were no identifiable legal requirements of the 

Atomic Energy Act of 1954, as amended, that required different 

radiological release criteria be applied to the cleanup of TMI-2 than 

were applied to an operating plant. In keeping with the purposes of 

the National Environmental Policy Act (NEPA), the Commission 

decided to prepare a PEIS on the decontamination and disposal of 

TMI-2 radioactive wastes.28 Under the terms of the policy 

statement, the Commission stated that development of the 

A computer generated cut-away drawing of the reactor building showing 

reactor coolant system components and “D-Ring” concrete radiation shields.
13

programmatic statement would not preclude prompt Commission 

action when needed. Such prompt actions would require 

consideration of the advice of the Council on Environmental 

Quality about the Commission’s NEPA responsibilities and would

require an environmental review with opportunity for public 

comment (in accordance with the May 25, 1979 policy statement). 

The policy statement also allowed rapid action for an emergency 

situation. In this situation, the Commission would consult the 

Council to the extent practical.29

o On September 26, 1980, the Commission issued a policy statement 

concerning the Pennsylvania Public Utility Commission’s order to 

the licensee to cease and desist from using any operating revenues 

for cleanup and restoration costs at TMI-2 which were not covered 

by insurance. The NRC Commissioners emphasized that all NRC 

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

must be fully complied with by the licensee, regardless of whether

or not these requirements appeared to conflict with the Utility 

Commission’s order.30, 31

o On April 27, 1981, the Commission issued a policy statement 

endorsing the final Programmatic Environmental Impact Statement 

(NUREG-0683) related to the decontamination and disposal of 

radioactive wastes resulting from the accident. The Commission 

stated that the licensee should accelerate the pace of the cleanup to 

complete expeditiously all decontamination activities, consistent 

with ensuring protection of public health and safety, and the 

environment. The policy statement also indicated that, as the 

licensee proposed specific major decontamination activities, the 

NRC staff would determine if these proposals, and associated 

impacts that were predicted to occur, fell within the scope of those 

already assessed in the PEIS. With the exception of 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 Commission’s approval if the 

activity and associated impacts fell within the scope of those 

assessed in the PEIS. The policy statement required the staff to keep 

the Commission informed of staff actions before staff approval of 

major activities. Further, the statement indicated that the cleanup 

should be carried out in accordance with the criteria in Appendix R 

of the PEIS, “Proposed Additions to Technical Specifications for 

TMI-2 Cleanup Program,” as well as in conformance to the existing 

operating license, and previously-imposed orders. The Commission 
14

expected to receive continuous advice from the Advisory Panel on 

the Decontamination of TMI-2 regarding major activities required 

to accomplish expeditious and safe cleanup of the TMI-2 facility.32

• Commission orders. Commission orders address appeals or motions 

before the Commission in such matters as amendments to nuclear 

facility licenses, license transfers, license renewals, and enforcement 

Components inside the reactor building.
15

matters.33 Two Commission orders resulted in the modification of the 

operating license at TMI-2 to require prompt operation of the 

EPICOR II system to decontaminate intermediate-level radioactive 

waste water in the auxiliary building, and to release krypton-85 from the 

reactor building’s atmosphere by controlled purging (see discussions 

below). 34, 35 These orders also resulted in amendments to the technical 

specifications. These orders are summarized later in this section and 

provided on the DVD (see the DVD folder, Orders).

• SECY papers and staff requirements memoranda. The primary 

decision-making tool for the Commission is a written issue paper 

submitted by NRC staff to the Commission. Policy, security, 

rulemaking, adjudicatory matters, and general information are provided 

in a stylized document referred to as a “SECY Paper.” After the 

Commissioners vote on a SECY Paper, the Office of the Secretary 

(SECY) records the decision in a memorandum to the staff called a 

“Staff Requirements Memorandum” (SRM), and also issues a 

“Commission Voting Record” which includes the record of votes and 

individual views of the Commissioners.36 SRMs may be issued 

following Commission meetings to document any discussion or requests 

made at the meeting. Many of the SECY papers and SRMs related to the 

TMI-2 recovery and cleanup efforts and found in the public record are 

provided on the DVD (see the DVD folder, Commission SECY 

Papers/Staff Requirements). 

• General correspondence. The Chairman is the official spokesperson for 

the agency. On a few occasions, the Chairman exchanged 

correspondence with licensee corporate executives. The examples of 

such correspondence that were found in the public record are provided 

on the DVD (see the DVD folder, Commission/Licensee 

Correspondence).

One notable letter from the NRC Chairman to the President of General 

Public Utilities (GPU) Corporation, dated January 12, 1981, responded 

to an earlier letter from GPU concerning near-term planning of the 

cleanup of TMI-2 in light of an order issued by the Pennsylvania Public 

Utility Commission. The State agency ordered the licensee to cease and 

desist from using any operating revenues for cleanup and restoration 

costs which were not covered by insurance. The Commission responded 

by issuing a Statement of Policy on the matter (see above) which was 

discussed in the Chairman’s response. In addition, the response 

provided a list of activities required to be performed during the period 

of ongoing discussions between the licensee and the State agency. The 
16

list provided a “road map” of minimum activities required in the nearterm to keep the TMI-2 reactor in a safe condition, and activities 

required to reduce potential threats over a longer term.37

Licensing Actions. Changes in the facility’s post-accident mode of 

operations required unique regulatory and licensing actions. In order to 

properly reflect evolving plant status, the NRC issued orders, modified 

those orders, approved license amendment requests, and granted relief from 

certain regulatory requirements. Most of the NRC correspondence 

approving each licensing action (including the correspondence supporting 

Internal view of the reactor pressure vessel at TMI-2. (From a training visual 

aid from the Babcock & Wilcox Company, the fabricator of the TMI-2 reactor 

vessel. This diagram was used to brief President Carter during his visit at TMI 

on April 1, 1979, see photo at the beginning of this section).
17

NRC safety evaluations and environmental reviews, where applicable) is 

provided on the DVD as indicated below: 

• Orders. The Commissioners (by majority vote), or designated 

senior-level officials, may issue an order that directs the licensee to take 

a prescribed action. Key orders related to the recovery effort are 

summarized below. These orders are provided on the DVD (see the 

DVD folder, Orders).

o On July 20, 1979, an Order for Modification of License was issued 

to suspend the power operation of TMI-2 and require that the 

facility be kept in a shutdown condition, in accordance with 

NRC-approved operating and contingency procedures.38

o On October 18, 1979, an Order for Modification of License was 

issued to require the licensee to promptly operate the EPICOR II 

filtration and ion-exchange decontamination system to 

decontaminate intermediate-level radioactive waste water held in 

tanks in the TMI-2 auxiliary building. The order further required the 

licensee to monitor EPICOR II discharge paths, maintain suitable 

tankage in TMI-1 as a contingency, and solidify spent resin before 

its shipment offsite.39, 40 The requirement to solidify resin was later 

removed in a modification to that order.41

o On February 11, 1980, an Order for Modification of License 

established new technical specifications governing operation of the 

facility in what was called a “recovery mode.” The proposed 

technical specifications (also known as “recovery technical 

specifications”) took into account the present condition of plant 

systems to ensure that the unit would remain in a safe and stable 

posture during the recovery mode.42 The order explicitly prohibited 

venting or purging, or other treatment of the reactor building’s

atmosphere; the discharge of water decontaminated by the 

EPICOR II system; and the treatment and disposal of high-level, 

radioactively-contaminated water in the reactor building, until each 

of these activities had been approved by the NRC, in accordance 

with the Commission’s November 21, 1979 policy statement. The 

attached safety evaluation and environmental assessment explained 

that these activities could have been allowed under the same 

effluent limitations as would apply in the case of a normallyoperating facility, had the Commission not determined that public 

interest warranted prohibiting these undertakings pending 

completion of an environmental review. The order provided that the 
18

existing pre-accident 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 would remain in effect except 

as provided in the order. These prohibitions effectively precluded 

the planned release of any radioactive liquid and gaseous material 

from TMI-2 without prior Commission approval (by majority vote). 

Low-level solid waste, such as rags and clothing, generated during 

cleanup operations in the auxiliary building, was permitted to be 

transported offsite to a commercial, low-level radioactive waste 

burial facility.43

The order did not change existing pre-accident Appendix B 

technical specifications, except that the licensee’s pre-accident 

management organization for activities addressed by those

specifications was replaced with sections in the proposed 

Appendix A technical specifications that addressed current 

requirements imposed on the licensee’s management organization 

for all post-accident licensed activities.44 The order proposed that 

The reactor pressure vessel on its way for installation at TMI-2. (This photo 

was taken either during the loading operation at the Babcock and Wilcox 

fabrication facility in Mount Vernon, Indiana, or during the unloading near 

Havre de Grace, Maryland.)
19

the facility operating license would be formally amended to include 

the proposed technical specifications. As discussed later, this did 

not happen until many years later. The technical specifications 

originally proposed were documented in NUREG-0432.

45 The 

related safety evaluation report and environment assessment were 

documented in NUREG-0647.

46

o On June 12, 1980, an Order for Temporary Modification of License 

was issued to require the licensee to release krypton-85 from the 

reactor building’s atmosphere by controlled purge. The order set 

offsite dose limits for the purge. The related final environmental 

assessment for the reactor building purge was documented in 

NUREG-0662.

47

o On June 18, 1981, an order was issued to require the licensee to 

promptly operate the submerged demineralizer system with effluent 

polishing by the EPICOR II system to process the 

intermediate-level contaminated water in the auxiliary building’s

tanks, and the highly-contaminated water in the reactor building’s

sump, and in the reactor coolant system. The related safety 

evaluation report was documented in NUREG-0796.

48

• Modification of orders. From time to time, some orders required 

modification to clarify language, or reflect changes in plant status, as a 

result of ongoing decontamination and maintenance efforts. The most 

overarching order that required numerous amendments was related to 

the proposed technical specifications. Although the NRC aimed to

incorporate the proposed recovery technical specifications in facility 

operating license (Appendix A technical specifications to the facility 

operating license), several requests for a hearing were filed in 

connection with the order. After that time, discussions with the 

recognized parties led to mutual agreement in the areas of concern, and 

resulted in the resolution of all outstanding issues and the withdrawal of 

all petitions. On November 8, 1985, the NRC Atomic Safety Licensing 

Board issued an order terminating the proceeding. During the pendency 

of this matter, a number of changes in the proposed technical 

specifications were necessary, each requiring a Modification of Order.49

Most modifications and related safety evaluations are provided on the 

DVD (see the DVD folder, Order Amendments).

• Changes to the recovery operations plan. The recovery operations plan

defined the surveillance requirements to be performed to ensure 

equipment operability as required by the plant’s technical specifications. 
20

The recovery operations plan was included as Section 4 of the proposed 

technical specifications. However, the plan was not considered a part of 

the proposed technical specifications. As such, changes made to 

surveillance requirements were approved by NRC staff outside of a 

Modification of Order.50 Some changes were issued concurrently with a 

Modification of Order, instituting corresponding changes to the 

proposed technical specifications. Approvals of most changes to the 

recovery operations plan are provided on the DVD (see the DVD folder,

Recovery Operations Plan Changes).

• Changes to the organization plan. The licensee’s organization plan 

provided the organizational structure (e.g., charts) for management of 

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 approved the 

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

management in the organization plan, instead of the proposed technical 

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

timely manner. Changes made to the organization plan required NRC 

approval, but certain changes did not require a Modification of Order.51

The approvals of most changes to the organization plan are provided on 

the DVD (see the DVD folder, Organization Plan Changes).

View of the cold-water basin inside a natural draft cooling tower at Three 

Mile Island. The cross columns support the curved concrete wall of the tower. 
21

• Amendments of license. The TMI-2 facility operating license, which 

included the technical specifications, was amended, modified, extended, 

and transferred during the three decades following the accident.

A chronology of key amendments is summarized below. Most 

amendment actions are provided on the DVD (see the DVD folder,

Operating License Amendments).

o On March 12, 1980, Amendment No. 10 was issued to revise the 

Appendix B technical specifications relating to the operation of the 

EPICOR II filtration and ion-exchange decontamination system to 

decontaminate intermediate-level waste water held in tanks in the 

TMI-2 auxiliary building. This amendment was approved under the 

NRC Commissioners’ Order of October 18, 1979.52

o On June 24, 1980, Amendment No. 11 was issued to approve a 

temporary change to Appendix B technical specifications relating to 

the bypassing of interlocks between the reactor building exhaust 

monitors and dampers during purging of the reactor building’s

atmosphere. This amendment was approved under the NRC 

Commissioners’ Order of June 12, 1980.53 

o On December 30, 1981, Amendment No. 18 was issued to reflect 

that GPU Nuclear Corporation was added as the licensee for TMI-2 

and replaced Metropolitan Edison Company as the licensee 

authorized to operate (maintain) TMI-2.54

o On January 27, 1986, Amendment No. 26 was issued to formally 

incorporate the proposed “recovery mode” technical specifications 

that were established by the February 11, 1980, Order and 

subsequently modified numerous times. The amendment was 

allowed by the NRC Atomic Safety Licensing Board following the 

resolution of all outstanding issues and the withdrawal of all 

petitions related to the order. Changes to the recovery technical 

specifications now required a license amendment (except for 

changes to the recovery operations plan and organization plan, 

which continued to be approved through a separate process).55

o On May 27, 1988, Amendment No. 30 was issued to extensively 

revise the technical specifications by aligning licensing 

requirements with appropriate plant conditions, through the 

remainder of the cleanup operations. Three facility modes were 

defined in the revision, which allowed transition from the defueling 

phase by incorporating technical specifications that were applicable 
22

during specific phases or modes of the cleanup. These modes were

as follows:

56

Mode 1: The reactor coolant shall be subcritical, with an average 

reactor coolant temperature of less than 200 degrees Fahrenheit.

Mode 2: This mode shall exist when the following conditions are 

met: (a) the reactor vessel and reactor coolant system are defueled 

to the extent reasonably achievable; (b) the possibility of criticality 

in the reactor building is precluded; and (c) there are no defueling 

canisters containing core material in the reactor building.

Mode 3: This mode shall exist when the conditions for Mode 2 are 

met, and no defueling canisters containing core material are stored 

on the TMI-2 site.

o On September 11, 1989, Amendment No. 35 was issued to modify 

the technical specifications, by deleting the prohibition for 

disposing of accident-generated water. The amendment retained the 

requirement for prior NRC approval of procedures associated with 

the disposal of the accident-generated water. The associated NRC 

safety evaluation approved disposal of the accident-generated water 

by evaporation, subject to restrictions.57

East view of TMI-2 buildings (left of TMI-1).
23

o On September 14, 1993, Amendment No. 45 was issued to modify 

the facility operating license to be a possession-only license. The 

NRC planned to issue the post-defueling monitored storage (PDMS) 

technical specifications after the licensee substantially satisfied the 

PDMS commitments and requirements.58

o On December 28, 1993, Amendment No. 48 was issued to 

extensively modify the technical specifications in ways consistent 

with the licensee’s plans for post-defueling monitoring storage of 

the facility.59

o On June 21, 1995, Amendment No. 49 was issued to extend the 

license until April 19, 2014, to allow both units at TMI to be 

decommissioned at the same time.60

• Exemptions. The NRC granted exemptions from certain requirements 

of the regulations for nuclear power plants, but only under special 

circumstances as permitted in the regulation (see Title 10, “Energy,” of 

the Code of Federal Regulations (10 CFR) 50.12, “Specific 

Exemptions”). Exemptions were necessary at TMI-2 because of the 

plant’s damaged configuration and changing status during cleanup. The 

approvals of most exemptions are provided on the DVD (see the DVD 

folder, Regulation Exemptions).

Planning and Guidance. Planning and guidance documents that were 

essential to the formulation of recovery and cleanup plans and activities 

included the NRC’s Programmatic Environmental Impact Statement (PEIS); 

the NRC policy statement endorsing the PEIS; the licensee’s planning 

studies; the recovery quality assurance plan; and the general project design

criteria document. These and other planning and guidance documents are 

summarized below. These documents and others, except as noted in the 

sections that follow, are provided in the DVD folder, General Management 

and Oversight. 

• Planning Study for Containment Entry and Decontamination. The 

licensee contracted Bechtel Power Corporation to develop a conceptual 

plan for reentry and decontamination of the reactor building. The 

primary objective of the July 2, 1979 report was to develop a plan for 

placing the reactor building in a configuration for removal of the reactor 

vessel head. The report provided an assessment of the reactor building’s

radiation environment and the physical condition of (and degree of 

damage to) it; evaluated alternatives for reactor building 
24

decontamination and reentry; and provided conceptual designs for new 

systems that may be needed to support reentry and decontamination.61, 62

• Summary Technical Plan for TMI-2 Decontamination and Defueling. 

The licensee’s plan for decontamination and defueling at TMI-2 was 

contained in their December 12, 1979 report, “Summary Technical Plan 

for TMI-2 Decontamination and Defueling.” The technical plan 

identified the major steps to clean up the plant, which are as follows: 

(1) decontamination of the auxiliary and fuel handling building, 

including removal of contaminated water held in storage tanks and 

sumps; (2) decontamination of the reactor building, including removal 

of the radioactive, gaseous atmosphere and contaminated water in the 

sump; (3) reactor examination and defueling; (4) decontamination of the 

reactor coolant system; (5) radioactive waste processing, including 

construction of the EPICOR II system, submerged demineralizer 

system, and evaporator and solidification system; and (6) solid 

radioactive waste management.63

• Interim criteria for radiological effluents from TMI-2 data gathering 

and maintenance operations (SECY-80-175). On April 14, 1980, the 

Commission approved radiological effluent criteria for the interim 

period before the issuance of the programmatic environmental impact 

statement for the purpose of data-gathering and maintenance operations. 

Releases which were specifically not covered by these criteria were 

purging of the reactor building’s atmosphere, disposal of EPICOR II 

decontaminated water, and treatment and disposal of high-level, 

radioactively-contaminated water in the reactor building. The interim 

criteria provided a mechanism by which the licensee may request to 

make small radioactive releases resulting in data collection and 

maintenance operations. These criteria described the information that 

the licensee must submit to the NRC for approval before performing 

these operations, and the type of review that the staff will perform to 

approve each request.64, 65

A lack of definitive release criteria hampered planning and engineering 

activities, resulted in a dilution of personnel resources to obtain specific 

Commission approval for activities that have insignificant impacts, and 

caused public concern over minor unplanned releases that would be 

inconsequential in a normal operating plant. An example of the former 

was the perceived need to obtain Commission approval to initially open 

the outer personnel air lock door which would release about 0.05 curie

of radioactive krypton gas (see SECY-80-10566). An example of the 

latter was the occurrence on February 11, 1980, in which the incidental 
25

off-gassing of about 0.3 curie of radioactive krypton gas from a leaked 

primary water system caused considerable public concern (see NRC 

Preliminary Notification 80-0367). To put these releases in perspective, 

TMI-2 had been releasing between 65 to 80 curies of krypton gas per 

month; at that time, a normal operating facility of this type may release 

over a 1,000 curies of radioactive gasses per month. In 1978, TMI-1 

released an average of 1,300 curies of radioactive gasses per month.68

• Report of the Governor's Commission on Three Mile Island. On 

May 14, 1979, Governor Richard Thornburgh of Pennsylvania 

established a special commission under the chairmanship of Lieutenant 

Governor William Scranton III to study and evaluate the consequences 

of the accident. The results of that commission's work were released on 

February 26, 1980, in a report entitled, "Report of the Governor's 

Commission on Three Mile Island." It contained a number of 

recommendations and findings aimed at protecting public health and 

safety in the wake of the TMI-2 accident. The seven-month 

investigation assessed the environmental, economic, health, legal, and 

social effects of the accident, and made recommendations for action or 

further study. One of the recommendations submitted to Governor 

Thornburgh was that Unit 2 be promptly decontaminated, under proper 

safety controls, in order to avoid possibly serious and uncontrolled 

releases of radiation.69, 70

• Programmatic Environmental Impact Statement (PEIS). The NRC’s 

PEIS related to the decontamination and disposal of radioactive wastes 

resulting from the accident (NUREG-0683), and three supplements to 

the PEIS were an important set of guidance documents for the NRC and 

licensee. The PEIS discussed the options and associated environmental 

impacts of four fundamental activities necessary to the cleanup: 

treatment of radioactive liquids; decontamination of the building and 

equipment; removal of fuel and decontamination of the coolant system; 

and packaging, handling, storing and transporting nuclear waste. The 

draft PEIS and supplements underwent comment periods by the 

licensee, Federal, State and local government agencies, and the public. 

The final PEIS was issued in March 1981.71

In terms of radiological health and safety, there was no known technical 

reason for the radiological release criteria to be more restrictive than 

had been acceptable at normal operating facilities. However, because of 

the unique characteristics of the cleanup operation that were not 

considered and evaluated in the pre-accident safety review of the plant, 

there was a need to define what keeping radiation exposure “as low as 
26

reasonably achievable” or “ALARA” meant with respect to offsite 

releases and occupational exposures. The PEIS provided the basis for 

making that determination.72 

The PEIS had three supplements that were considered part of the 

original PEIS. (See the DVD folder, Guidance-PEIS.) 

o Supplement 1, “Final Supplement Dealing with Occupational 

Radiation Dose.” The earlier 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 

was issued in October 1984 to reevaluate the occupational radiation 

dose and resulting health effects from cleanup and to address 

additional alternative cleanup approaches using information 

gathered since the PEIS was prepared in 1980.73 

o Supplement 2, “Final Supplement Dealing with Disposal of 

Accident-Generated Water.” This supplement was issued in 

June 1987 to update the environmental evaluation of 

accident-generated water disposal alternatives published in the 
27

original PEIS, using more complete and current information. Also, 

the supplement included a specific environmental evaluation of the 

licensee's proposal for water disposition.74

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

Monitored Storage and Subsequent Cleanup.” This supplement was 

issued in August 1989 to evaluate the licensee's proposal to 

complete the current 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, and a number of alternative courses of action from the end 

of current defueling efforts, to the beginning of decommissioning. 

However, it did not provide an evaluation of the environmental 

impacts associated with decommission.75 

Because these reports 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 in order 

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 appendixes as appropriate.76

• NRC policy statement endorsing PEIS. On April 27, 1981, the NRC 

Commissioners issued by majority vote a policy statement endorsing the 

final PEIS. The policy statement concluded that the PEIS satisfied the 

NRC’s obligations under the National Environmental Policy Act 

(NEPA). The policy statement also indicated that, as the licensee 

proposed specific major decontamination activities, the NRC staff 

would determine whether these proposals, and associated impacts that 

were predicted to occur, fell within the scope of those already assessed 

in the PEIS. With the exception of the disposition of processed 

accident-generated water (the Commissioners wanted to decide this 

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

without Commissioners’ approval if the activity and associated impacts 

fell within the scope of those assessed in the PEIS. The policy statement 

required the staff to keep the Commissioners informed of staff actions 

before staff approval of major activities.77

• TMI-2 Program Strategy. In June 1984, the licensee issued an internal 

technical plan, “TMI-2 Program Strategy,” that provided an overview of 

the recovery program, established program priorities, established policy 

and technical guidance, and provided the means for communicating the 
28

program to external people and organizations. The report provided 

program policies that addressed the following issues: (1) generic issues,

such as recovery program objectives; definition of program phases, 

priorities, and end points; use of the ALARA concept in decisions about 

radiation exposure; use of remote technology; application of regulations 

and regulatory guides; preservation of plant equipment and structures; 

permanence of recovery facilities; sharing of facilities and systems with 

TMI-1; and opening containment; (2) characterization of plant 

conditions, including data-gathering; (3) fuel control, such as criticality 

and reactivity control, accountability of fuel, definition of core waste, 

and fuel storage and disposal; (4) methods and end points for defueling;

(5) disassembly and storage of large radioactive components, including 

primary system integrity; (6) decontamination and dose reduction, such 

as decontamination during the fuel-removal stage, criticality prevention 

during decontamination, worker efficiency, and re-flooding of the 

reactor building’s basement; and (7) waste management, such as storage 

of waste and closure of commercial disposal sites, abnormal waste, 

reuse of processed water, and segregation of water at TMI-2.78 (Note: 

The program strategy is not available on the DVDs.) 

• Post-defueling monitored storage (PDMS) plan. This licensee report 

was submitted to the NRC on December 2, 1986, to provide a plan for 

plant conditions following completion of the cleanup program.79 This 

initial plan was used to develop the proposed license amendment, 

proposed technical specifications, and safety analysis report for 

implementation of the proposed PDMS plant configuration. The request 

was submitted to NRC on August 16, 1988.80 In response to the 

submittal, the NRC developed and issued Supplement 3 to the 

Programmatic Environmental Impact Statement (NUREG-0683), which 

dealt with PDMS and subsequent cleanup.81 The NRC issued the 

possession-only license on September 14, 1993, and issued the PDMS 

technical specifications on December 28, 1993.82 See further 

discussions on PDMS in the section on After Defueling. 

• Recovery quality assurance plan. This licensee plan ensured regulatory 

compliance for recovery and cleanup activities such as decontamination; 

assessment of damage; design; procurement; fabrication; handling; 

shipping; storage; cleaning; construction; installation; inspection; test; 

operation; maintenance; repair; and modification. This plan was 

periodically revised.83 (See the DVD folder, Guidance-PEIS.) 

• TMI-2 radiation protection plan. This plan provided the philosophies, 

basic policies, and objectives of the licensee’s program for radiological 
29

controls in accordance with NRC regulations and guidance. The 

objectives of the radiological controls program were to control radiation 

hazards in order to avoid accidental radiation exposures, to keep 

exposures within their regulatory required limits, and to keep exposures 

of workers and the general population at ALARA levels. The plan was 

implemented within radiological controls procedures. The TMI-2 

radiological controls program was fully-integrated into each phase of 

the recovery effort at TMI Unit 2. This plan was periodically revised.84

(See the DVD folder, Worker Protection.)

• General project design criteria. The general project design criteria 

document was developed by the licensee to provide an adequate basis 

for the design of recovery facilities and systems. The criteria included a 

general section that provided information common to all engineering 

disciplines. The general section defined the regulatory requirements; 

operating conditions; dose-reduction considerations; environmental 

considerations; sharing of existing facilities and services; and protection 

from severe natural phenomena and human-caused events. The 

discipline-specific sections provided detailed, generic criteria for 

architectural design; civil-structural design; control-system design; 

non-safety-related electrical design; mechanical design; and shielding 

design and access control. To ease the review of future submittals, the 

licensee requested that the NRC review the general section and 

subsequent revisions.85 (See the DVD folder, Guidance-PEIS. Early 

revisions of the discipline-specific sections were provided to the NRC 

for information only in the licensee’s submittal dated 

December 22, 1981.) 

• Special nuclear material (SNM) accountability plan. This licensee 

plan identified the methods and sequence of SNM accountability; the 

quality assurance program that was built into SNM measurement 

activities; and the areas, systems, and components that had undergone 

formal SNM measurement and those that did not require SNM 

assessment. SNM accountability is required of all licensee holders of 

reactor fuel and other SNM. As the result of the accident, the damaged 

fuel debris was dispersed throughout the plant, and the origin of the 

debris could not be traced to specific fuel assemblies. The NRC and 

DOE (receiver of the fuel debris) allowed the final SNM accountability 

for TMI-2 to be performed after defueling was completed.86 Further, the 

NRC granted the licensee exemptions regarding regulatory requirements 

for record keeping, inventorying, and reporting of special nuclear, 

source, and byproduct materials.87 Accountability was based on a 

thorough post-defueling survey of areas, systems, and components. The 
30

results of materials accountability surveys and analyses were important 

in evaluating the potential for recriticality of the remaining fuel debris 

under postulated conditions during the plant’s long-term, post-defueling 

monitored storage.88 (See the DVD folder, Guidance-PEIS; see also the 

DVD folder, After Defueling for post-defueling survey reports and 

post-defueling completion reports.)

• Licensed operator qualification and training procedures. These 

licensee procedures provided the processes and requirements for the 

training and certification of licensed operator candidates, and for the 

renewals of their licenses. Licensed operators included reactor 

operators, senior reactor operators, and fuel-handling-only (defueling) 

senior reactor operators. These NRC-approved training and qualification 

procedures reflected the unique and rapidly-changing plant conditions 

as recovery and cleanup progressed. As such, licensed operators were 

trained on new systems and defueling operations as part of their 

ongoing requalification training.89, 90, 91 Training of fuel-handling-only 

senior reactor operators began in December 1984. NRC granted fuelhandling-only licenses to the first five senior reactor operators in 

October 1985.92 (See the DVD folder, Guidance-PEIS.) 

Advisory Groups. Advisory and working groups were formed by GPU, the 

NRC, and the DOE to provide advice, and sometimes direction, on 

important recovery activities. Documents included in this collection (see the 

DVD folder, Advisory Groups) provide insights into the technical issues and 

problems encountered during plant recovery and cleanup. Some important 

advisory and working groups are summarized below: 

• Industry Advisory Group (March 30 to May 6, 1979). The Industry 

Advisory Group of outside organizations and individuals was formed by 

the licensee three days after the accident to help determine plant 

conditions, and evaluate approaches to achieving a stable condition. The 

scope of the group was expanded to evaluate operations and 

modifications proposed by the recovery organization; to independently 

assimilate, integrate, and interpret plant status information and data; and 

to review detailed procedures for plant recovery operations. Once the 

formal TMI-2 recovery organization was established on April 4, 1979, 

the information flow between the Industry Advisory Group and 

elements of that organization was accommodated through the Technical 

Working Group (described below). Activities for the Industry Advisory 

Group were assigned by the Technical Working Group.93 Some reports 

by the Industry Advisory Group are provided in the DVD folder,

Industry Advisory Group Reports. 
31

• Technical Working Group (1979). The Technical Working Group was 

an executive-level committee established by the licensee a week after 

the accident to propose and discuss actions associated with daily 

operations. The managers of the newly-established TMI-2 recovery 

organization attended the initial twice-daily meetings to coordinate 

activities, assign task responsibilities, and obtain general agreement on 

activities for the 24-hour period following each meeting. 

Representatives from outside organizations, such as the TMI-2 nuclear 

steam supply system vendor (Babcock & Wilcox); the original TMI-2 

architect engineer (Burns and Roe); the NRC; and the Industry Advisory 

Group, also attended these meetings. Licensee personnel chaired the 

meetings and were responsible for making the decisions, and 

a consensus was achieved in most cases. The NRC, by virtue of its legal 

authority to issue orders, did have the authority and therefore, the 

implicit responsibility, to override a decision whenever that seemed 

necessary.94 The Technical Working Group’s daily meeting minutes for 

the first four months after the accident are provided on the DVD (see 

the DVD folder, Planning Meetings).

• Advisory Panel for the Decontamination of TMI-2 (1980 to 1993).

In October 1980, the Commission established a 12-member TMI-2 

advisory panel to consult with and provide advice to the NRC 

Commissioners and staff on major activities related to the 

decontamination and cleanup of TMI-2. The panel consisted of 

members from the Commonwealth of Pennsylvania, local government, 

and the scientific community, as well as residents in the vicinity of TMI. 

The NRC TMI Program Office (discussed below) acted as a liaison 

between the NRC and the TMI-2 advisory panel, and also provided 

information to the panel on the status of the cleanup. Panel meetings 

were open for the general public to attend and transcriptions were 

produced for the public record. The panel provided an outlet for the 

public, 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. 

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 agency and the 

licensee to public concerns. It also kept the cleanup before the NRC 

Commissioners through periodic meetings. Panel members traveled to 

Washington, DC at least once each year to meet with the 

Commissioners and provide a report on current panel activities. The last 

panel meeting, the 78th overall, was held in September 1993.95 The 

“Lessons Learned from the Three Mile Island Unit 2 Advisory Panel” 
32

report (NUREG/CR-6252) provided lessons learned based on interviews 

of former members and reviews of transcriptions of public meetings 

(see the DVD folder, Advisory Groups). Appendix A, “Panel Meeting 

Dates and Transcript Microfiche Addresses,” to the report listed dates of 

panel meetings. Panel members were listed in an appendix of the NRC 

annual reports (see DVD folder, NRC Annual Reports).

The Advisory Panel for the Decontamination of Three Mile Island Unit 2 held 

its last meeting on September 23, 1993, in Harrisburg, Pennsylvania (Pa). 

Panel members attending the final meeting are pictured. They are, left-toright, front row: Ann Trunk, Resident of Middletown, Pa.; Arthur E. Morris 

(Panel Chairman), Resident and former Mayor of Lancaster, Pa.; Joel Roth 

(Panel Vice Chairman), Resident of Harrisburg, Pa.; Elizabeth Marshall, 

Resident of York, Pa. In the back row, left-to-right, are: Kenneth L. Miller, 

Director of the Division of Health Physics and Professor of Radiology, Milton 

S. Hershey Medical Center, Hershey, Pa.; Thomas Smithgall, Resident of 

Lancaster, Pa.; Lee H. Thonus, Alternate Designated Federal Official, NRC 

Office of Nuclear Reactor Regulation (Region I); John Leutzelschwab, 

Professor of Physics, Dickinson College, Carlisle, Pa.; Niel Wald, Professor, 

Department of Environmental and Occupational Health, University of 

Pittsburgh, Pittsburgh, Pa.; Michael T. Masnik, Designated Federal Official, 

NRC Office of Nuclear Reactor Regulation; Frederick S. Rice, Resident of 

Harrisburg, Pa.; and Gordon Robinson, Associate Professor of Nuclear 

Engineering, Pennsylvania State University, University Park.
33

• TMI-2 Safety Advisory Board (1981 to 1989). The TMI-2 Safety 

Advisory Board was established by the licensee on March 16, 1981, to 

provide the licensee with an independent appraisal of the recovery 

program that gave particular emphasis to the assurance of public and 

worker health and safety. The advisory board met every three months, 

and reviewed many aspects of recovery activities, including regulations; 

nuclear criticality; safety; risk assessment; project organization; project 

financing; project procedures; technical planning; and public 

communications. Additionally, the advisory board regularly expressed 

its views to the NRC, and to the Board of Directors of GPU. 

Periodically, the advisory board participated in public hearings, 

especially those of the NRC’s Advisory Panel for the Decontamination 

of TMI-2. The first advisory board chairman was Dr. James C. Fletcher, 

former administrator of the National Aeronautics and Space 

Administration (NASA). He later resigned as chair of the advisory 

board at the time of his second appointment as NASA Administrator, 

following the Challenger space shuttle accident in 1986.96 The final 

report of the Safety Advisory Board summarized its activities during its 

8-year period of existence (see the DVD folder, Advisory Groups). 

Appendix A, “Possible Research Opportunities as a Result of the TMI-2 

Accident and Cleanup,” to that report provided a set of 

recommendations on possible research opportunities.

TMI-2 Safety Advisory Board. Front row, left-to-right: Bruce Lundin, 

Robert Marston (Chairman 1986-1989), Jacob Fabrikant, Howard Raiffa.

Back row, left-to-right: Lombard Squires, Ronald Fillnow, Norman 

Rasmussen, Merril Eisenbud, Robert Friedman, William Stratton, John 

Auxier. Not shown: James Fletcher (Chairman 1981-1986).
34

• Technical Assistance and Advisory Group (1981 to 1989). The TMI-2 

Technical Advisory and Assistance Group (TAAG) was established by 

the licensee, with the cooperation of the DOE and NRC, to provide 

independent technical assessment and advice on the decontamination 

and defueling of TMI-2. 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 ALARA 

levels. The TAAG consisted of approximately 10 permanent members 

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 DOE. The group’s work was funded 

through the Idaho National Engineering Laboratory (INEL) by the 

DOE. Representatives from INEL, DOE, and the NRC attended TAAG 

meetings as observers. The TAAG reported directly to the president of 

GPU and documented its results to INEL to assist in execution of the 

DOE program.97, 98 Many of the TAAG reports are provided on the 

DVD (see the DVD folder, Advisory Groups).

• Technical Working Group for the TMI Information and Examination 

Program. The TMI Information and Examination Program was 

established to acquire data to improve current understanding of nuclear 

plant accident environments and of the phenomena which contributed to 

Two key planning documents used by NRC and DOE at TMI-2. 

NUREG-0698 report defined NRC’s roles in the cleanup operations and 

associated regulatory responsibilities. GEND-INF-036 report identified 

tasks to be planned and administered by the DOE to retrieve useful 

information from the cleanup effort to address key problems and issues 

in the areas of plant accident response and recovery. 

TASK PLAN FOR THE U.S. DEPARTMENT OF ENERGY

TMI-2 PROGRAMS

Technical Integration Office

Prepared for the

U.S. DEPARTMENT OF ENERGY

Three Mile Island Operations Office

Under DOE Contract No. DE-AC07-76ID01570
35

those environments. The licensee (GPU), EPRI, NRC, and DOE (known

as “GEND”) signed a coordination agreement on March 26, 1980, to 

jointly sponsor and participate in this program. The program was 

administered through DOE’s prime contractor at TMI-2 and was staffed 

by INEL personnel. In addition to the participation of NRC in the 

technical working group for this program, the NRC (1) reviewed the 

data-acquisition tasks to ensure that they were implemented in 

coordination with ongoing cleanup activities and (2) used the data 

acquired for the benefit of the cleanup to the maximum possible 

extent.99

Results of this program were summarized in periodic newsletters and 

annual reports (see the DVD folder, DOE/National Laboratory Status 

Reports). Technical details of the results were documented in 

GEND-series reports, reports from DOE’s national laboratories, the 

NRC’s NUREG-series reports, and EPRI reports. Most DOE-sponsored 

reports were digitized by the Idaho National Laboratory (formerly 

INEL) and the Oak Ridge National Laboratory and are currently 

available from their public websites. Most of these reports and 

NUREG-series reports are provided on the DVDs. 

NRC’s TMI Program Office. The NRC was responsible for the regulation 

of TMI-2 cleanup operations. Regulatory objectives were to (1) maintain 

reactor safety and control of radioactivity; (2) ensure that environmental 

impacts were minimized and that the radiation exposure of workers, the 

NRC office trailers were relocated in late April 1979 from the TMI Visitors’ 

Center to this location at the south end on Three Mile Island. Another office 

in nearby Middletown was opened and staffed on a regular basis, including 

evenings, to provide the public an opportunity to talk with NRC.
36

public, and the environment was within regulatory limits and were at 

ALARA levels; and (3) ensure interim safe storage and disposal of 

radioactive wastes from cleanup operations.100

On April 7, 1979, the director of the Office of Nuclear Reactor Regulation 

(NRR) formalized the NRC operations at TMI with three principal 

organizations: NRR Operations, NRR Technical Review, and the Office of 

Inspection and Enforcement Operations. The NRR operations function 

interacted with the NRR staff in the TMI-2 control room and with the GPU 

Task and Schedule Managerial Team. The GPU managerial team prioritized 

work and NRR provided liaison. The NRR technical review function 

provided safety reviews of plant modifications proposed by GPU’s Plant 

Modification Team. Early modifications included the new decay heat 

removal system, electrical power distribution, primary system 

instrumentation alternatives, radioactive waste systems, secondary cooling 

systems, and other plant modifications. The NRR technical review group 

also provided liaisons to GPU’s Industry Advisory Group and the GPU 

Technical Working Group. The inspection and enforcement function 

provided surveillance of TMI-2 operations and in-plant health physics and 

continued monitoring and analysis of the environment.101

The TMI support organization was renamed the TMI Program Office 

(TMIPO) within NRR on April 1, 1980; it provided the NRC’s overall 

direction of TMI-2 recovery and cleanup operations.102, 103 The TMIPO 

established a staff with management and technical expertise in key TMI-2 

cleanup activities such as radiation protection, radiological assessment, 

NRC staff suited up for an inspection

inside the reactor building. Left: 

Lake Barrett (onsite TMIPO 

manager) in 1984. Above: Michael 

Masnik and Lee Thomus (TMIPO 

staff and later licensing project 

managers of TMI-2) in 1993. 
37

radiological waste treatment, and nuclear safety. Support by experts in other 

disciplines was available from other NRC staff and from contractors, under 

arrangement with the DOE. The TMIPO coordinated its activities with the 

licensee, the DOE, other Federal agencies, State and local government 

officials, and the public.104 

Information flow was a major responsibility of the site office. A weekly 

status report containing pertinent reactor, radiological, and environmental 

information was prepared and distributed to all NRC offices. This report 

was also distributed to the public, with copies available at the Middletown 

office. The Middletown office was open and staffed on a regular basis, 

including evening hours, to provide the public an opportunity to remain 

informed of the cleanup progress. Information was also supplied to the 

public through press releases, television and radio interviews, and direct 

responses to both written and oral public concerns. Information-exchange 

meetings were also held periodically with officials of the DOE and the 

U.S. Environmental Protection Agency (EPA).

105

The TMIPO had the following regulatory responsibilities: (1) planning and 

managing all NRC involvement in TMI-2 cleanup activities; (2) obtaining 

information about and evaluating the current facility status; (3) analyzing 

and reviewing the licensee's proposed actions and procedures; (4) preparing 

technical review documents on the safety and environmental impacts of 

licensee-proposed cleanup actions; (5) approving or disapproving the 

licensee's proposed actions and procedures; (6) advising the NRC 

Commissioners on major cleanup actions; (7) coordinating the NRC's 

TMI-2 cleanup activities with other governmental agencies, as necessary, 

such as the DOE and EPA; (8) informing State and local governments and 

the public on the status and plans for cleanup activities; (9) overseeing 

day-to-day licensee activities to ensure that operations were implemented in 

accordance with NRC regulations, the facility's operating license, technical 

specifications, NRC orders, recovery plans, and approved procedures;

(10) ensuring that activities are carried out in compliance with approved 

NRC limits and procedures; and (11) coordinating with the NRC Office of 

Inspection and Enforcement on its TMI-2 inspection activities.106

The TMIPO report, “NRC Plan for Cleanup Operations at Three Mile Island 

Unit 2” (NUREG-0698, as revised) defined the NRC’s role in cleanup 

operations at TMI-2, outlined the NRC’s regulatory responsibilities in 

fulfilling this role, and provided the NRC’s review and decision-making 

procedures. The TMI-2 Cleanup Project Directorate (formerly TMIPO) was 

dissolved on February 1, 1988. The NRC’s TMI resident inspector office 

took over the inspection program for TMI-2 and a headquarters project 
38

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 

(see NUREG-0698, Revision 2).
39

directorate assumed responsibility for technical review and project

management functions.107

The TMIPO weekly status reports provided a detailed chronology of plant 

status, environmental monitoring results, the licensee’s recovery activities, 

NRC actions, and public meetings. Reports during the period from 1980 

to 1990 are provided on the DVD (see the DVD folder, TMI Program Office 

Weekly Status Reports). NUREG reports and other TMIPO-related 

correspondence are provided on the DVD (see the DVD folder, General 

Management and Oversight).

Other Documents. Other documents relating to the management of the 

recovery and cleanup efforts provided on the DVD (see the DVD folder, 

General Management and Oversight) include the following: 

• licensee recovery organizations (see also the DVD folder, Organization 

Plan Changes) 

• defueling and decontamination plans, schedules, and cost estimates 

• NRC and DOE work plans and agreements

Two diesel generators (center) were temporary installed at TMI-2 to provide 

diverse backup power to the balance-of-plant (BOP) electrical power buses in 

the event of a failure of normal offsite power sources. Backup protection was 

required for new recovery systems and existing BOP systems that were used to 

keep the plant stable. Many of the safety systems used during normal 

operations were rendered inoperable or inaccessible due to the accident. 
40

TMI-2 control room after the accident. The plant computer alarm station is shown in the foreground.Core exit thermocouple temperature indication panel which was installed shortly after the accident is shown in the backgound (center left) behind the hanging sign.
41

3 Plant Stabilization 

The near-term cooling of the reactor core was considered to be stable within 

the first week after the accident. However, many technical issues involving 

safety and control were identified in the following weeks and months. In the 

licensee’s report to the NRC, “Interim Report on the Three Mile Island 

Nuclear Station Unit 2 Accident,” dated May 5, 1979,108 the major 

near-term objectives of the recovery organization included the following: 

(1) keep the plant in a stable shutdown condition; (2) control and manage 

the volumes of existing radioactivity; (3) develop an overall 

waste-management plan for liquid, gas, and solids; (4) develop a strategy to 

reach cold shutdown safely and expeditiously; (5) modify the procedures, 

facilities, and equipment necessary to accomplish the above; and 

(6) institute the plan for accomplishing a transition into the organization 

necessary to proceed with the more long-term recovery efforts.

As a result of the degradation of the reactor core and plant equipment, 

certain equipment, required to be operable, was no longer available. Other 

non-safety-related systems, not generally relied on for safe shutdown, were 

used to keep the plant in a stable condition. High radioactivity in 

containment, reactor coolant, and certain auxiliary building areas limited 

access to maintain important components needed for safe shutdown. Harsh 

environmental conditions and high-radiation levels limited the ability of 

certain components and instrumentation to survive for long periods.

109

Early Plant Status. Maintaining the plant in a stable shutdown condition 

required: (1) reliable means to remove decay heat from the reactor core to 

the ultimate heat sink; (2) maintenance of reactor core sub-criticality to limit 

heat production; and (3) confinement of radioactivity within the reactor 

building.110 Decay heat removal required the maintenance of critical 

functions; namely, the reactor coolant system (RCS) flow control; RCS 

pressure control; RCS water inventory control; and RCS heat removal

control. The control of RCS flow initially required the removal of noncompressible gases, such as accident-generated hydrogen, which could 

potentially block flow at the high points of both hot-legs and damage an 

operating reactor coolant pump. The removal of gases was successful and 

one reactor coolant pump provided forced circulation for almost one month, 

followed by natural circulation.111 

The control of RCS pressure was important to maintain an adequate net 

positive suction head to prevent reactor coolant pump damage and to 

maintain the hot coolant at a subcooled condition to prevent boiling. 

Pressure was controlled initially by the pressurizer heaters, spray, and 
42

instrumentation. Contingency procedures were developed in the anticipation 

of failures of heater cable installation due to high radiation levels in the 

reactor building or failures of pressurizer instrument transmitters located in 

the flooded basement of the reactor building. Upon the loss of the last 

pressurizer level instrument channel on April 27, 1979, pressure was 

controlled with the plant solid for almost a year by manually balancing 

makeup injection and letdown flows. Letdown flow was fixed at a reduced 

one-third capacity due to blockages in the system. The desirable pressure 

was achieved by throttling the manual valves on the reactor coolant pump 

Systems and components inside the reactor building (RB). Key: core flood tank 

(CFT), decay heat (DH) system, once-through steam generator (OTSG), 

pressurizer (PZR), reactor coolant drain tank (RCDT), and reactor vessel (RV).
43

seal injection lines with the makeup injection valve completely closed. The 

lack of routine maintenance due to adverse radiological conditions inside 

the auxiliary and fuel handling building contributed to makeup pump 

failures and valve leaks, which further exuberated radiological 

contamination and releases inside the building. In March 1980, the makeup 

system was replaced with the newly-installed reactor coolant pressure 

control system; this new system was in operation until preparations to 

remove the reactor vessel head in 1984.112

Systems and components inside the auxiliary and fuel handling building. 

Key: waste disposal liquid (WDL) system, makeup and purification (MU&P) 

system, makeup filter (MU-F), and makeup pump (MU-P).
44

The control of RCS heat removal was through the “A” steam generator 

during the periods of forced and natural circulation cooling. The secondary 

or steam side of the steam generator was maintained in a partial vacuum by 

the main condenser mechanical vacuum pumps. A partial vacuum allows 

boiling and steaming at temperatures below 212 degrees Fahrenheit (°F). 

Steam flowed from the steam generator to the main condenser through a 

turbine bypass valve. The condensate system was used to feed water from 

the main condenser hotwell to the steam generator. The circulating water 

system removed heat from the condenser to the natural draft cooling 

towers.113 The use of secondary systems to achieve and maintain the plant in 

cold shutdown was never performed in commercial pressurized water 

reactors, especially for a long period of time. Cold shutdown condition was 

initially achieved on April 27th when reactor coolant temperature decreased 

to 188°F.114 Plant modifications improved the reliability of electrical power 

supplies to non-safety grade secondary system components that provided 

long-term cooling. A backup mini-condensate system was installed to 

replace the existing condensate pumps, if needed. Other diverse backup 

cooling systems were installed, such as the long-term “B” backup cooling 

system and mini decay heat removal system (discussed later).

115 

The control of the reactor core sub-criticality was accomplished by 

increasing the concentration of neutron-absorbing boron solution in the 

reactor coolant. However, the lack of information on the state of the reactor 

core and the condition of control rods, caused difficulty in accurately 

calculating the boron concentration required to maintain subcritical 

conditions.116 The required minimum boron concentration in the reactor 

coolant system was increased to account for the potential (actual) melting of 

control rod materials during the accident and to support defueling 

operations. Sources of makeup to the reactor vessel were limited and 

controlled to prevent boron dilution. Nuclear source range instrumentation 

used to detect re-criticality was down to one operable channel prior to its 

refurbishment during early reactor building entries.117 The potential for recriticality of the damaged reactor core, as well as fuel “fines” (very small 

fuel debris) consolidated in sludge on the reactor building basement floor, 

and in filters of cleanup systems, remained a concern during the first half of 

the new decade.118

Large quantities of radioactive krypton-85 gas (about 44,000 curies)119 and 

highly-contaminated water in the reactor building’s basement 

(approximately 630,000 gallons of water, which contained about one-half of 

the core inventory of the radionuclide cesium-137)120 prevented personnel 

access into the reactor building to assess damage and maintain important 

equipment. Radioactive gas in waste gas decay tanks leaked into the 
45

auxiliary building, causing airborne contamination problems. About 66 to 

80 curies of krypton gas was leaking out of the reactor building every 

month. The partial vacuum on the secondary side of the steam generator 

caused a pressure difference between the reactor building and turbine 

building, which enhanced leakage through the packing of various steam 

valves. These gases were subsequently discharged from the secondary 

system through the auxiliary building ventilation system to the 

environment.121 Waste-water storage at TMI was already at 60 percent (full) 

capacity before the accident, and that water was subsequently contaminated 

during the accident.122 Operational leakage from support systems located in 

the auxiliary and fuel handling building added to the mix about 800 to 

1,000 gallons a day of mostly uncontaminated water.123 

Organizations involved in the early recovery effort tackled many unique 

technical issues aimed at ensuring the safe shutdown of the reactor core, and 

ensuring sufficient removal of decay heat from the reactor core to the 

ultimate heat sink. A number of systems and components were modified in 

the near-term to maintain a stable plant. Some of these systems were placed 

in standby readiness and never had to be used. Many new systems and 

facilities were also built to improve the diversity and reliability of existing 

systems. These new systems provided capabilities to immediately clean up 

radioactivity hazards that would help future plant recovery operations. New 

procedures were developed, (many early ones handwritten), to provide 

guidance for operating the plant in its degraded condition. First-of-theirkind procedures were written, reviewed, and approved to handle many 

imaginable contingencies (see the DVD folder, Recovery Procedures). 

The Industry Advisory Group provided the utility with expert advice on 

many technical issues during the first 5 weeks following the accident (see 

the DVD folder, Industry Advisory Group Reports). The NRC extensively 

studied methods and associated plant modifications for achieving and 

maintaining cold shutdown. Other early issues and concerns that were 

studied included recriticality of the reactor core; impact of core damage on 

decay heat removal; reliability and diversity of decay heat removal, 

including electrical power; loss of instrumentation; post-accident hydrogen 

production; mitigation of radiological releases; radioactive waste 

characterization; contaminated water storage; potential for groundwater 

contamination; fire protection; quality assurance of plant modifications; 

radiological protection of plant workers; and safety analysis of the damaged 

plant (see the DVD folder, General Plant Stabilization).

Near-Term Recovery Actions. Key systems that were used for near-term 

plant stabilization are listed below. Some systems were placed in standby 
46

and never had to be used. Summary descriptions of these systems were 

provided in a series of monthly and quarterly licensee status reports that 

were submitted to the NRC over the first 18 months following the accident 

(see the DVD folder, GPU Status Reports). 

• Sub-criticality control. The reactor core was maintained in a shutdown 

condition by boron solution that was injected in the reactor coolant 

system. Since the integrity of the control rods and fuel rods was 

unknown, the reactor coolant boron concentration was maintained 

between 3,000 and 4,500 parts per million (ppm). The maximum boron 

concentration and a minimum reactor coolant temperature of 50 °F were 

specified in the proposed technical specifications to ensure that boron 

precipitation blockage would not occur.124 Criticality analyses of the 

reactor core to ensure adequate shutdown margin under worst case 

conditions had been made by the licensee, NRC, Babcock and Wilcox, 

Brookhaven National Laboratory, and Oak Ridge National Laboratory. 

A further evaluation of the risk of recriticality by NRC in April 1980 

concluded that the most probable mechanism for criticality was boron 

dilution and that the process was slow enough to detect and correct the 

approach to criticality given appropriate instrumentation and 

procedures.125 

In April 1984, the maximum allowable boron concentration in the 

proposed technical specifications was increased to 6,000 ppm to support 

defueling options that could potentially rearrange the core, or portions 

of the core, into a more reactive configuration. The increase in boron 

concentration, if needed, would be buffered by sodium hydroxide 

addition to maintain the pH in the reactor coolant system between 7.6 

and 7.8 for corrosion control. The modification to the technical 

specifications included an upper pH limit of 8.4 to insure no boron 

precipitation in the reactor coolant system (RCS) and associated sample 

lines (solubility of boric acid decreases with increasing pH).126 In July 

1984, the minimum boron concentration was increased to 3,500 ppm to 

ensure at least 1 percent shutdown margin, based on all creditable core 

configurations, including those resulting from a reactor vessel head drop 

accident. The licensee’s criticality analyses conservatively assumed a 

hypothetical 100 percent fuel failure; no neutron leakage or absorption 

by structural or poison material; no fuel burnup; maximum fuel 

enrichment; and optimum fuel-moderator ratio. For the out-of-core 

criticality model, the analysis assumed that 50 percent of the core 

formed a hemisphere in the bottom of the reactor vessel.127, 128, 129 In 

April 1985, the technical specifications were modified to increase the 

minimum boron concentration to 4,350 ppm to ensure that the fuel in 
47

the reactor coolant system would remain subcritical throughout all 

reactor disassembly and defueling operations. Boron concentration 

requirements for the spent fuel storage pool “A” and the fuel transfer 

canal were added in the technical specifications. The values were the 

same as those required for the reactor coolant system.130, 131, 132

Other criticality analyses were performed to support safety evaluations 

of defueling systems and equipment, defueling and cleanup operations, 

and the fuel debris remaining inside and outside the reactor vessel.

Other criticality control methods were evaluated in GEND-026, 

“Addition of Soluble and insoluble Neutron Absorbers to the Reactor 

Coolant System of TMI-2.” Most criticality safety analysis reports are 

provided in the DVD folder, Criticality Analysis. Results of criticality 

analyses that were performed in support of the post-defueling monitored 

storage licensing application are provided in the DVD folder, After 

Defueling. 

• Transition to natural circulation core cooling. Since the start of the 

accident until 8:00 p.m. on the same day when the “1A” reactor coolant 

pump was restarted, decay heat was primarily removed by the release of 

reactor coolant to the reactor building through the stuck-open 

pressurizer pilot-operated relief valve, as the operator opened its block 

valve to maintain pressure. For about a month, a reactor coolant pump 

provided forced circulation from the reactor core through the “A” steam 

generator. The reactor decay heat was transferred from the reactor 

coolant side of the “A” steam generator tubes, and out to the main 

turbine condenser in the form of steam, generated in the secondary side 

of the steam generator tubes.133

On April 27, the failure of the last remaining pressurizer level indication 

prompted the operators and engineers to enact the emergency procedure 

to stop the fourth and last remaining reactor coolant pump. That 

afternoon, the reactor coolant pump was intentionally shut down by a 

control room operator. The “forward flow” in both of the reactor coolant 

system loops decreased exponentially, as indicated on the then-operable 

flow instrumentation. Within a minute, TMI-2 entered smooth, natural 

circulation without incident. (The driving head for natural circulation 

was the difference in density between the hot 12-foot core water 

elevation, against a comparable 12-foot cooler elevation of water in the 

tubes in either or both of the adjacent steam generators.) The feasibility 

of long-term, natural circulation as a viable means for placing the 

damaged reactor core into a long-term, stable condition was extensively 

studied and evaluated by the licensee, the TMI-2 reactor vendor 
48

(Babcock and Wilcox), and the NRC (see reports in DVD folder, 

General Plant Stabilization).134

With the passage of time, and the associated reduction of decay heat 

generation rate (about 75 kilowatts 20 months following the accident), 

the natural circulation flow slowed and then changed from continuous to 

cyclic with increasing intervals between the cyclic flow “burps.” When 

the steam generator and cold-legs gradually cooled until the density was 

high enough to initiate natural circulation flow, flow then diminished as 

the warmer coolant from the reactor vessel displaced the colder water in 

the steam generator and cold-legs.135

• Alternate instrumentation. Alternate plant and reactor core instruments 

were used and new ones were installed to replace those that were 

damaged during the accident, or that failed because of harsh 

environmental or high-radiation conditions. Instrumentation and 

electrical equipment were exposed to moderately severe accident 

conditions, including steam, reactor building spray, high radiation, 

burning hydrogen, and the resultant overpressure. The equipment that 

survived the accident was then exposed to long-term moisture 

Core flood tank 1A level monitor CF-2-LT2 showing the condition of the 

enclosed electronic module. There was severe corrosion and degradation of 

internal components rendering the unit inoperable.
49

conditions, including high humidity, elevated temperature, and 

condensation. The TMI-2 Instrumentation and Electrical Program 

funded by DOE identified and analyzed a number of pre-accident 

installation problems and instrument response characteristics that led to 

misleading information and equipment failures. These problems 

included faulty component seals and inadequate drains and vents to 

protect enclosed equipment against moisture, anomalous response of 

radiation monitors, and substantial corrosion of electrical contacts over 

a period of a few years.136 (See the DVD folder, Instrumentation and 

Equipment Evaluations.)

Emergency procedures were written by the licensee and approved by the 

NRC to respond to losses of important instrumentation (see the DVD 

folder, Recovery Procedures). Indications from other systems were 

sometimes used to replace similar information from failed instruments. 

Installation of recovery systems provided alternative means for 

monitoring conditions of the reactor. Probes were used to measure

conditions inside the reactor building through its penetrations before the 

first reactor building entry in July 1980. Once entrance began, important 

electronic components were replaced or repaired.

• Modified decay heat removal system. Installed to improve the reliability 

of the existing decay heat removal system in order to reduce its 

Long-term “B” backup cooling system (horizontal heat exchanger shown)

provided a high pressure, closed cooling loop for the water filled “B” steam 

generator. This backup system was never used.
50

radiological impact in the auxiliary building and on the environment in 

the event that the system was needed to cool the core. This system was 

placed in standby and never had to be used.137

• Alternate decay heat removal system. Intended as a backup to existing 

and planned decay heat removal capabilities. The new system was to be 

built outside the fuel handling building and was to include a complete, 

integral closed-loop system to circulate reactor coolant through

connections from the existing decay heat removal system. Heat would 

be removed through its own intermediate component cooling water 

system through connections to the existing nuclear services river water 

system; this ambitious project was never completed.138

• Long-term “B” backup cooling system. Installed to provide a 

high-pressure, closed cooling loop for the water-filled “B” steam 

generator. This new system was a planned backup to the preferred 

steaming of the “A” steam generator to the main condenser. The system 

was installed in the turbine building and included a new heat exchanger, 

pump, surge tank, piping, and valves. Heat was transferred from the 

secondary side of the “B” steam generator to a new heat exchanger. The 

Mini decay heat removal system being fabricated at the Babcock and Wilcox 

facility. Two heat exchangers on stand in back; two pumps on floor. 
51

Mini decay heat removal system installed at TMI-2. Both heat 

exchangers are shown. The system became operational in 

October 1980.

heat exchanger was cooled by the existing secondary services closed 

cooling water system, which was then cooled by water from the existing 

nuclear services river water system. The system was installed, but never 

required use. An identical system for steam generator “A” was 

designed, but never built.139

• Mini decay heat removal system. Installed to replace the decay heat 

removal system and served as a technical specifications required means 

for boron injection and makeup to the reactor vessel. It was a 

skid-mounted system with two pumps and two heat exchangers that was 

fabricated at the Babcock & Wilcox facility in Lynchburg, Virginia.140

The system was installed in the fuel handling building with connections 

to the existing decay heat removal piping. The two heat exchangers 

were cooled by water from the existing nuclear services closed cooling 

system. The system was installed but was not used because loss of heat 

from the reactor coolant system to the reactor building’s ambient 

environment was shown to be adequate.

141 The system became 
52

operational on October 29, 1980,142 and could also be used during 

defueling operations when the reactor vessel was not configured to 

maintain forced cooling. However, loss-to-ambient cooling mode, as 

described below, was the preferred method to remove the very low 

decay heat levels.143

• Reactor coolant pressure control system. Installed to provide long-term 

reactor coolant system (RCS) pressure control and inventory control. 

This system kept the RCS in a water-solid condition for natural

circulation core cooling; maintained sufficient available net positive 

suction head for reactor coolant pump operation, if needed; controlled 

the quality of the makeup water; and maintained pressure while 

accommodating thermal and volumetric changes in the RCS inventory. 

The system included a makeup subsystem to maintain the reactor 

coolant system in a water-solid condition for natural circulation cooling; 

the makeup subsystem also injected chemically-treated water. 

A letdown subsystem provided RCS overpressurization protection by 

increasing letdown. Three surge tanks arranged in series provided RCS 

pressure control.144 The system was installed in the fuel handling 

building and placed in service in March 1980. The existing makeup 

system and letdown through the reactor coolant pump seal return line 

were secured to reduce leakage and contamination in the auxiliary and 

fuel handling building. The new system was no longer needed when the 

Reactor coolant pressure control system installed in fuel handling building. 

Three 900-gallon surge tanks shown.
53

RCS was depressurized in the summer of 1984 in preparation for 

removal of the reactor vessel head.145

• Alternate condensate pumps system. Installed to provide backup to the 

existing condensate pumps in order to supply feedwater to the steam 

generators for decay heat removal. In addition, the system could provide 

feedwater to the new temporary auxiliary boiler. The system included 

two 50-gallon-per-minute pumps that took suction from the condenser 

hotwell and discharged to the steam generator through either of the two 

newly-installed demineralizers.146 

• Reliability improvements to balance-of-plant (BOP) electrical 

distribution. Modifications to the existing BOP electrical power system 

provided backup protection in the event of a failure of normal offsite

power sources to BOP buses. Backup protection was required for new 

recovery systems and existing BOP systems that were used to keep the 

plant stable, such as decay heat removal, pressure control, plant 

instrumentation, and ventilation. Early modifications included two 

Reactor coolant pressure control system provided long-term reactor coolant 

system pressure and inventory control. The system became operational in 

March 1980.
54

independent power buses, each supplied by a new 2,500-kilowatt-rated 

diesel generator, and two existing circulating water pump buses 

supplied by one new 13.2-kilovolt line from a nearby offsite electrical 

substation.147 The new diesel generators became available on May 16, 

1979.148 The 13.2-kilovolt line was powered by an offsite 115-kilovolt 

network which was backed by combustion turbine generators capable of 

being energized independently of the normal offsite power source from 

the 230-kilovolt network.149 Later in 1980, the NRC granted approval 

for removing the BOP diesel generators given that the improved 

reliability of the 115-kilovolt network with the combustion turbines 

allowed the network to be classified as an acceptable backup source of 

power for decay heat removal.150

• Additional hydrogen recombiner. Installed to provide additional 

capacity to process hydrogen inside the reactor building following the 

accident, due to uncertainty regarding the quantity of hydrogen being 

generated. A skid-mounted, thermal-type hydrogen recombiner was 

installed next to the existing thermal-type hydrogen recombiner located 

in the fuel handling building.151 

• Portable disposable demineralizer system. Installed to remove 

radioactive fission products in the “B” steam generator in order to 

minimize personnel exposure and potential for contamination of the 

turbine building before the new long-term “B” closed-loop cooling 

system was placed into service. The system included a disposable 

demineralizer (18 inches in diameter and 30 inches in height) that was 

attached to the discharge and suction of the closed-loop pump. The 

portable disposable demineralizer system was used for wet layup of the 

long-term “B” cooling system during standby.152 

• Groundwater monitoring system. Installed to detect radioactivity 

leakage in the ground around the reactor building and auxiliary 

building. The Reactor Building Integrity Assessment Program was 

established by the licensee, at the direction of the NRC, to monitor 

potential leakage paths from the TMI-2 reactor building sump. The 

leakage monitoring points, which were based on engineering 

evaluations,153 included groundwater monitoring wells; storm drainage 

areas; cork seals (concrete joint seals) in structures surrounding the 

reactor building; and the tendon access gallery (a passageway 

surrounding the reactor building below the basement, approximately 

20 feet below the water surface in the reactor building.154) The system 

initially included eight monitoring wells located around the TMI-2 

reactor building and other nearby locations. Seven observation wells 
55

were installed early in 1980 to determine the source of tritium found in 

the wells. Monitoring wells had pumps; observation wells had grabsampling capability. Additional wells were installed in later years.155

The Programmatic Environmental Impact Statement (NUREG-0683) 

concluded that an accidental release of contaminated water in the 

basement of the reactor building was highly unlikely. Had this water 

leaked through the steel-lined concrete base of the reactor building, the 

wave fronts of tritium, strontium-90, and cesium-137 in the ground 

would reach the Susquehanna River after a minimum of 350 days, 23 

years, and 284 years, respectively; and would continue to enter the river 

over periods of about 130 days, 8.5 years, and 140 years, respectively. 

Accounting for radionuclide decay and total mixing of effluents in the 

river, the peak radionuclide concentrations would be orders of 

magnitude below concentration limits from 10 CFR Part 20, “Standards 

for Protection Against Radiation,” for liquid effluents released to the 

general environment. Furthermore, the peaks of these three 

radionuclides would occur at different times.156

Groundwater monitoring wells (MW) and observation wells (OW) were 

installed in early 1980 to detect potential radioactivity leakage in the 

ground around the reactor building and the auxiliary and fuel 

handling building.
56

• Nuclear sampling system. Installed to be used as a temporary 

liquid-waste sampling facility in TMI-2 without interfering with the 

existing sampling system that was shared with TMI-1. TMI-2 sample 

lines were rerouted to a new sample sink located in the fuel handling 

building. Sampling capabilities were provided for various existing and 

new storage tanks, the reactor coolant system, and continuous 

monitoring of boron concentration inside the reactor vessel.157

Temporary filtration system for the auxiliary and fuel handling 

building installed on the auxiliary building roof. TMI-2 reactor 

building in the background; fuel handling building at center 

right.
57

• Auxiliary and fuel handling building supplementary air filtration 

system. Installed to temporarily replace the existing charcoal filter trains 

in order to reduce offsite releases from the auxiliary and fuel handling 

building. Replacements of the existing filters were not practical because 

of the high radiation and contamination levels in the filter areas. The 

new system interfaced with existing ventilation systems and consisted 

of four parallel filter units. Each unit had an exhaust fan and a filter unit. 

Each filter unit consisted of a prefilter, a high-efficiency particulate air 

(HEPA) filter, a charcoal filter, and a second HEPA filter. The 

ventilation stack was capped to direct ventilation from the buildings into 

the new filtration units. The filtration units were installed on the roof of 

the auxiliary building.158 Both trains of the existing auxiliary building 

exhaust system were placed back into service by the end of April 1980. 

The supplementary units were secured and the ventilation stack was 

uncapped for the upcoming reactor building purge of krypton gas.159, 160

• Main condenser air extractor filtration system. Installed within two 

weeks of the accident to remove contamination extracted from the 

condenser condensate by the mechanical vacuum pumps before 

discharging to the unfiltered segment of the auxiliary building 

ventilation stack. The system included a pre-filter, a HEPA filter, a 

charcoal filter, and a second HEPA filter. The system remained in 

operation until steam generator cooling was secured in January 1981.161

• Fuel pool waste storage system (also known as the “tank farm”). 

Installed to provide temporary storage of 110,000 gallons of radioactive 

water from the reactor building sump and miscellaneous waste holdup 

tanks. The system consisted of 4 upper tanks (capable of holding 

15,000 gallons each) and 2 lower tanks (25,000 gallons each) forming 

two separate storage areas located in TMI-2 spent fuel pool “A.” 162 In 

order to preserve the integrity of the structure, tanks were not attached 

to the spent fuel pool structure. Therefore, the pool was not filled with 

water for shielding to prevent the tanks from floating. Installation began 

on April 6, 1979, and was completed in July 1979. On May 17, 1984, 

the removal of these tanks started following the cleanup of the highlycontaminated water in the reactor building to make room for the transfer 

and storage of reactor fuel debris.163

• Temporary auxiliary boiler system. Installed to supply steam to the 

TMI-2 main turbine gland seals while the existing auxiliary boilers 

shared with TMI-1 were serviced. Turbine gland seals were needed to 

maintain condenser vacuum for steam generator heat removal.164
58

• Waste water cleanup systems. EPICOR I, EPICOR II, and the 

processed water storage and recycling system are discussed in the 

section on Waste Management. 

• Solid waste storage facilities. The interim solid waste staging facility 

and the solid waste storage facility are discussed in the section on Waste 

Management. 

Longer-Term Recovery Actions. Key activities that were taken to keep the 

plant stable in the longer term are summarized below: 

• Transition to loss-to-ambient core cooling On January 5, 1981, the 

licensee stopped cooling (also known as steaming) the "A" steam 

generator by shutting a turbine bypass valve to the main condenser. This 

action put the reactor coolant system (RCS) in a "loss-to-ambient" mode 

of cooling. Decay heat was then removed by heat losses from the 

system to the air inside the reactor building without the need of any 

RCS circulation. A test of the loss-to-ambient cooling mode 

Fuel pool waste storage system or “tank farm” provided temporary storage of 

110,000 gallons of radioactive water. Lower tank shown next to inactive fuel 

transfer upenders (to be used later to move fuel canisters from the reactor 

building into the spent fuel pool).
59

commenced on November 6, 1980, for 33 days. Technical specifications 

were then modified to recognize loss-to-ambient mode as an acceptable 

means for long-term cooling of the reactor core. Operating procedures 

were prepared, and loss-to-ambient cooling was subsequently adopted 

as the primary means for cooling the reactor core. By February 1, 1981, 

decay heat diminished to about 43 kilowatts of thermal energy.165 The 

long-term “B” backup cooling system and the mini decay heat removal 

system were available as alternative cooling modes. The discontinued

use of the “A” steam generator to remove decay heat allowed its 

removal from service and technical specifications for several major 

balance-of-plant systems and equipment, such as the circulating water 

system; main steam system; “A” steam generator; condensate pumps; 

condensate and feedwater systems; main condenser; and associated 

support systems.166

• Waste water storage management. The storage of pre-accident and 

accident-generated radioactive water challenged the licensee for many 

months and years. Before the accident, radioactive waste tanks were 

already 60 percent full with waste water from the recent TMI-1 

shutdown outage.167 In addition, leakage from various operating systems 

in the auxiliary and fuel handling building added about 800 to 

1,000 gallons per day.168 On May 25, 1979, the Commission issued a 

policy statement prohibiting the discharge of accident-generated water 

into the river. The installation of additional water storage tanks and the 

start of the EPICOR I and II water cleanup systems provided 

much-needed relief in storage capacity for varying radioactive 

concentrations. During the early weeks and months following the 

accident, plant operators transferred contaminated water between 

existing tanks (see correspondence in DVD folder, General Plant 

Stabilization). The licensee considered using railroad tank cars for 

onsite storage of low-level radioactive waste water. The licensee asked 

the NRC to help them find 100 tank cars.169 Two tank cars were 

purchased by the licensee; however, they were never used to store 

contaminated or processed water. In July 1979, the fuel pool waste 

storage system (“tank farm”) was available with 110,000 gallons of 

storage capacity. In July 1981, two new processed-water storage tanks 

were available (each with 500,000 gallons of capacity).170

• Purge of the reactor building’s atmosphere. The controlled purging of 

radioactive krypton gas inside the reactor building was carried out under 

detailed procedures approved by the NRC staff. The purging began on 

June 28, 1980, and continued until the morning of July 11. The removal 

of radioactive krypton-85 gas from the reactor building’s atmosphere 
60

The TMI-2 ventilation stack being uncapped for the 

planned purge of accident-generated radioactive krypton 

gas from the reactor building. The ventilation stack was 

previously capped to redirect ventilation from the 

auxiliary and fuel handling building to the temporary 

supplementary air filtration system.

allowed workers to begin work to clean up the reactor building, to 

maintain instruments and equipment, and to remove the damaged fuel 

from the reactor vessel.

171

The approval process for an environmental assessment of a number of 

alternatives for the decontamination of the reactor building’s 

atmosphere, began with the issuance of a draft for public comment. 

Approximately 800 responses were received from various Federal, 

State, and local agencies and officials, as well as from nongovernmental 

organizations and 

private citizens. 

Based on the 

comments received, 

the “Final 

Environmental 

Assessment for 

Decontamination of 

the Three Mile 

Island Unit 2 

Reactor Building 

Atmosphere” 

(NUREG-0662) was 

issued in May 1980. 

Having reviewed the 

staff assessment and 

recommendations, 

along with the 

comments from the 

public, the Governor 

of Pennsylvania, and 

many others, the 

Commission issued 

a Memorandum and 

Order on 

June 12, 1980, 

which authorized the 

licensee to clean the 

reactor building’s 

atmosphere by 

means of a 

controlled purge. On 

the same day, the 

Commission issued 
61

a temporary modification of the TMI operating license setting offsite

dose limits for the purge.172 The order included a one-time waiver of the 

environmental technical specifications (Appendix B) requirements for 

the instantaneous and quarterly average limits for the release of noble 

gases.

173 On July 24, 1980, Amendment No. 11 to the facility’s

operating license was issued to allow the bypassing of the interlocks 

from the reactor building’s exhaust radiation monitors to the reactor 

building’s exhaust purge dampers for the duration of the purge.174

The vented activity was estimated to range from 38,302 to 50,254 curies 

of krypton-85 with a median value of 44,132 curies during the release 

period. Environmental monitoring was performed with substantial 

instrumentation, with both fixed monitors in the ventilation system, and 

mobile sampling around the plant.175 Offsite radiation monitoring 

programs were conducted by the licensee, the NRC, the United States 

Environmental Protection Agency (EPA), the Department of 

Environmental Resources of the Commonwealth of Pennsylvania, and 

also by local citizens working through the Community Radiation 

Monitoring Program set up by the DOE, and the Commonwealth of 

Pennsylvania. The maximum cumulative radiation dose and the 

maximum dose rate measured at offsite locations were a fraction of the 

limits allowed under NRC regulations.176 There were subsequent purges 

to release the krypton gas that was slowly desorbing from the water and 

walls. The monthly releases that occurred between September 1980 and 

December 1980 decreased in an approximately exponential manner 

(emitting 27, 15, 12, and 7.5 curies respectively).177

• Reactor building entries. Entries into the reactor building were initially 

desired for damage assessment, data collection, and equipment 

maintenance; in support of these entries, a set of experiments was 

performed to identify any safety issues. Experiments included weekly 

analyses of airborne samples from the reactor building; a gamma scan 
62

of the equipment hatch; a gamma scan of the basement water through 

penetration R605; an analysis of samples of the basement water through 

penetration R401 with support from the Oak Ridge National Laboratory 

(ORNL); radiation mapping of the no. 2 personnel air lock; plate out 

analysis of the spool piece from the hydrogen recombiner with ORNL 

support; analysis of hazardous gas in the reactor building’s atmosphere; 

operating level (347-foot elevation) gamma scan and video survey 

through penetration R626; and testing of protective clothing, 

telemetered dosimetry, and dose-rate instrumentation.178

Entries began on July 23, 1980,179 after the reactor building’s 

atmosphere was purged of krypton, though initially, the plan was to 

enter the reactor building prior to purging.180 An attempt was made on 

May 20, 1979, but was delayed until after the purging because of a 

malfunction of the inner door to the reactor building’s personnel 

airlock.181 The initial entries used the following equipment: (1) portable 

lighting; (2) radios (for communication); (3) respiratory protection, such 

as self-contained closed-circuit breathing apparatus for the first entry 

and a positive-pressure filtered breathing air mask for subsequent 

entries; (4) protective clothing, such as paper overalls, cotton overalls, 

and a firefighter’s suit; (5) protective shoes, such as cloth shoe covers, 

three pairs of plastic disposable shoe covers, and firefighter’s boots;

View from inside the reactor building of the inner flange to reactor building 

penetration R626 with video camera and telemeter dosimetry antennas 

(center). Initial experiments through this penetration supported the initial 

reactor building entry program.
63

(6) protective gloves, such as one pair of latex gloves, two pairs of 

neoprene gloves, and one pair of electric power lineman’s gloves;

(7) head protection, such as a cotton hood and a rain suit hood; 

(8) personnel dosimetry, such as telemetered dosimetry, self-reading 

digital dosimeters, thermoluminescent dosimeters, and film badges; and 

(9) handheld radiation instrumentation, such as a telescoping gamma 

dose-rate meter for taking measures at a distance and a prototype survey 

meter for taking high-range beta-ray measurements. The heavy-duty 

outerwear provided high-energy beta-ray protection. During the first 

20-minute stay inside the reactor building, each engineer received a 

whole-body gamma exposure of about 200 millirem with no beta skin 

exposure.182 On May 24, 1989, the 2000th reactor building entry was 

logged by the licensee and support personnel.183 Photographs, 

communication transcripts, and reports of the first six entries are 

provided in the DVD folder, Reactor Building Entries. 

• Reactor building sump water removal. By the time cleanup systems 

were ready to remove contaminated water from the reactor building, 

about 624,000 gallons of contaminated reactor coolant was standing on 

the reactor building basement floor. About 264,000 gallons was spilled 

Engineers being suited up for the initial entry attempt into the reactor 

building. The orange underwater diver’s dry suit was used to prevent inleakage of radioactive material and to provide sufficient density to attenuate 

the anticipated high-energy beta radiation. The yellow rain suit was used to 

prevent direct contamination of the dry suit. After the purge of krypton-85 

from the reactor building, the dry suit was no longer required.
64

through the stuck-open pressurizer pilot-operated relief valve during the 

accident. The pressurizer continued to leak during cooldown operations 

and contributed another 178,000 gallons. The automatic actuation of the 

reactor building spray pumps during the hydrogen burn contributed an 

additional 1,700 gallons of borated water directly into the sump. An 

additional 180,000 gallons of non-contaminated river water leaked from 

the river water cooling system through a leaking relief valve on a

reactor building cooling unit.184 

Although the contaminated water was safely contained in the reactor 

building sump, its presence there constituted a continuing risk of 

leakage to the environment, and prevented or hindered the performance 

of the major decontamination activities.185 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 GPU 

leadership, the leaders and residents of communities downstream, and 

the State and local governments of both Pennsylvania and Maryland. 186

Further, an early NRC task force concluded that the greatest risk of loss 

of containment was a confined leak of the reactor building water to the 

lower parts of the auxiliary building, where water in the reactor building 

stood almost 30 feet above the decay heat removal pumps and the 

reactor building spray pumps. These pumps may have been needed later 

during the recovery.187

First entry into the reactor building (shown) occurred on July 23, 1980. Heavy 

duty outerwear provided beta ray protection.
65

The reactor building was constructed to prevent water in the sump from 

leaking into the environment; however, early concerns existed over the 

long-term storage of the highly-contaminated water in the basement, 

and the integrity of the containment following the hydrogen burn inside 

the reactor building during the first day of the accident.188 The reactor 

building was designed and fabricated to withstand an internal pressure 

of 60 psig as well as major earthquakes. A structural integrity test was 

conducted prior to unit startup to pressurize the containment to 69 

pounds per square inch, 

115 percent of the design 

pressure.189 The 

foundation mat for the 

reactor building is 11.5-

feet-thick and rests on 

bedrock. The reactor 

building has a carbon 

steel liner, 3/8-inch-thick 

on the sides, 1/2-inchthick dome, and 1/4-inchthick base. A concrete 

slab 2-feet-thick was 

poured above the liner 

base plate. The exposed 

face of the liner was 

coated with a prime and 

finish coat of an epoxy 

lining material. An 

additional leakage barrier 

was provided by a 40-milthick polyvinyl chloride 

polymer waterproofing 

membrane which was 

installed over the outer 

surfaces of the foundation 

mat.190

The submerged demineralizer system (discussed later) and the surface 

suction system191 were designed and installed to clean up the highlycontaminated water inside the reactor building. A homemade floating 

pump, carefully crafted using polystyrene foam, glue, and epoxy from a 

Corvette repair kit, was designed and tested to initially float at an angle 

and then lie flat on its side as it approaches the bottom. In 60 roentgen 

per hour radiation field, a team of four wearing over 100 pounds of antiCross-section view of the reactor building and 

auxiliary building with water level inside the 

reactor building as of April 1980. Gound water 

elevation shown on the right represented the water 

table outside the buildings.
66

contamination clothing and equipment, dropped the suction pump with 

rubber hose over the open stairwell into the reactor building basement in 

less than 60 seconds.192 Details of this experience are provided in the 

seminar video, “The 35th Anniversary of the Three Mile Island Nuclear 

Power Plant Accident of 1979: Working at TMI During and Following 

the Accident,” (see the first volume of this NUREG/KM).

• Floor expansion joint contamination. On November 26, 1980, 

contamination in the cork seam of expansion joints was first discovered 

during a routine radiation survey in the control and service building 

area. The seam is 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 time period following the accident, the cork seam 

located in the auxiliary 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 non-contaminated areas inside the plant. 

However, the radioactive contamination is prevented from entering the 

groundwater table by a water-stop barrier. Modifications have been 

made to the cork seam to allow periodic monitoring of the water levels 

The surface suction system and floating pump. Picture shows a floating pump

during a test in the outlet flume between the two natural draft cooling towers.
67

in the joint, to permit periodic water removal, and to prevent water and 

contamination migration within the cork-filled joint. 193, 194, 195 

• Auxiliary and fuel handling building decontamination (discussed in 

the section on Decontamination) 

Document Collections. Documents relating to plant stabilization efforts are 

grouped into five document collections on the DVDs. 

• Correspondence between the NRC and the licensee (i.e., notifications, 

requests, reviews, and approvals) relating to the installation and 

operation of plant systems and documents associated with other plant

stabilization operations (see the DVD folder, General Plant 

Stabilization). Enclosures (such as reports of system descriptions, 

technical evaluations, and safety evaluations) are included with 

correspondence. NUREG reports that supported stabilization activities 

are also included.

• Agenda and minutes of daily meetings of the Technical Working 

Group during the first four months after the accident (see the DVD 

folder, Planning Meetings). As previously noted, this group was an 

executive committee established by the licensee within a week after the 

accident to propose and discuss actions associated with daily operations. 

Many of the meeting handouts contained notes by NRC staff attending 

these meetings.

• Emergency and special operating procedures that were initially 

developed to cope with the plant’s damaged state during the first 

months after the accident (see the DVD folder, Recovery Procedures). 

These procedures were reviewed and approved by NRC staff located at 

TMI. Many early procedures that were approved for use were 

handwritten. 

• Industry Advisory Group task close-out reports (see the DVD folder, 

Industry Advisory Group Reports). As discussed previously, this group 

of industry organizations and individuals assisted the licensee to 

determine plant conditions and evaluate approaches to stabilize the 

plant. Tasks assigned to the group were documented in these reports.

• Photographs, communication transcripts, and reports of the first six 

reactor building entries (see DVD folder, Reactor Building Entries).
68

Technicians inside the personnel airlock to the reactor building, as seen 

through the round window of the outer personnel airlock door.
69

4 Worker Protection 

Post-accident radiological conditions at TMI-2 were substantially different 

from those normally encountered at commercial operating nuclear plants 

because of the magnitude and specific mix of the radionuclide 

contamination. Radiation surveys made shortly after the accident showed 

that general area radiation readings ranged from 150 to 500 millirem per 

hour in the fuel handling building, and 50 to 5,000 millirem per hour in the 

auxiliary building. Hot spots were measured in the auxiliary building 

reaching up to 125 rem per hour, and exceeding 1,000 rem per hour in some 

cubicles. The high-energy beta component was up to one hundred times the 

gamma component. 196

During the first entry into the reactor building in July 1980, dose rates at the 

305-foot entry-level elevation ranged from 400 to 600 millirem per hour. 

Localized areas of high radiation were measured at 18 rem per hour over the 

open stairwell and 2 to 5 rem per hour at floor drains. The general-area floor 

and wall beta-radiation readings ranged from 1 to 2 rad per hour.197 Surveys 

performed during the second entry at the next-higher level, the 347-foot 

operating-floor elevation, showed general radiation readings of 100 to 

400 millirem per hour. 

Below the 305-foot entry level elevation was the 282-foot basement-level 

elevation, which was flooded with highly-contaminated water and sludge. 

A telescoping radiation detector was inserted down through one of the 

reactor building stairwells and measured 40 to 45 rem per hour at 5 to 7 feet 

from the surface of the basement water.198 Once the water was drained and 

processed through the submerged demineralizer system, dose rates from the 

remaining sludge ranged from 1 to 1,000 rem per hour, depending on 

location and distance from the floor.199

A unique concern at TMI-2 was high-energy beta contamination from

fission products in the reactor coolant. The generation of activation products 

from corrosion such as cobolt-60 was minimal because the new plant had 

less than a year of full-power operation. Areas in the auxiliary building that 

had experienced coolant leakage were measured in the 10- to 

100-rem-per-hour gamma range with associated beta dose rates in the 

1,000- to 10,000-rad-per-hour range. Similar gamma-to-beta ratios were 

measured on surfaces in the reactor building. The cesium isotopes have beta

energies in the 0.5-million-electron-volt (0.5-MeV) range; strontium-89, 

which has the highest concentration of beta emissions, has a maximum 

energy of about 1.5 MeV; yttrium-90, the decay daughter of strontium-90, 
70

has a 2.3-MeV beta energy.

200 Typically, beta radiation from beta-emitting 

radionuclides at operating plants is low-energy, so protective clothing 

provides sufficient shielding. The high-energy beta emitters present at 

TMI-2 required special radiological protection practices, such as monitoring 

equipment, personnel dosimetry, heavy protective clothing, procedures, and 

training.201 Access into many areas in the plant with high levels of 

high-energy beta-emitting yttrium-90 required a combination of 

double-respirator face pieces or face shields and safety glasses.202 The need 

for these special worker protection practices became clear after several 

incidents of skin and extremity exposures exceeding regulatory limits.

Radiation Exposure Events. The licensee reported several accident-related 

exposures to the whole body and extremities during the response to the 

accident on March 28th and 29th in excess of NRC regulatory limits. During 

the very early phases of post-accident recovery, about five months after the 

accident, six workers received overexposures to the skin and extremities. 

The last extremity overexposure of two workers occurred in 1989. These 

events are summarized below: 

• Whole-body exposure. On the evening of March 28, 1979, an auxiliary 

operator received a collective whole-body exposure in excess of the 

NRC quarterly limit, while on two tours in the auxiliary building. The 

auxiliary operator made two entries into the auxiliary building without 

radiation work permits, a high-range self-reading dosimeter, and proper 

planning with health physics support. There was no one to challenge the 

operator’s entry and no locked access to restrict entry into the building 

at that time. Upon exiting the building both times, the operator’s 

low-range, (0 to 200 millirem), self-reading pocket dosimeter read

off-scale high. In addition, the operator failed to perform surveys for 

personal contamination. Upon entering the TMI-2 control room after the 

first job and then again after the second, a count rate meter on the table 

alarmed both times. After being told by a radiation protection 

technician, who happened to see the alarm the first time, to 

decontaminate, the operator decided to enter the auxiliary building again 

for a second 10-minute job (the operator reasoned that he was already 

contaminated). On the operator’s return to the control room the second 

time, the operator told the shift supervisor that the self-reading

dosimeter had gone off scale. The shift supervisor told the operator to 

decontaminate. After decontamination, the operator’s personal 

dosimeter was read, indicating a whole-body gamma exposure of 

3.170 rem. An NRC investigation concluded that this entry exemplified 

the lack of good health physics practices.203 (See NUREG-0600, 

Section 3.2.4.8.) 
71

(The NRC radiation protection regulations that were applicable during 

the first decade of the recovery effort allowed up to 3 rem per calendar 

quarter for a total of 12 rem in one year with proper documentation of 

the worker’s exposure history. This higher limit is no longer stated in 

current regulations.204) 

• Extremity exposures. During the afternoon of March 29, 1979, two 

workers received overexposures to their hands while taking reactor 

coolant samples to measure boron concentration, in order to ascertain 

whether the reactor core was still critical. The team discussed the 

assignment for a period of less than one hour. No one wore extremity 

monitoring on their hands. Self-reading pocket dosimeters of one 

worker were worn inside the protective clothing, making them 

inaccessible for periodic checks. The other worker wore a high-range 

self-reading pocket dosimeter taped to the forearm, but it was knocked 

off during collection of the sample. Before taking the sample, the room 

was surveyed and found to have an average exposure rate of 8 roentgen 

per hour. About 300 milliliters (ml) was drawn in a sample bottle, of 

which 100 ml was poured into a graduated cylinder for gross radiation 

measurement. Using a telescoping radiation-detection instrument, the

cylinder measured off-scale high on contact (greater than 1,000 rem per 

hour), 400 rem per hour at 1 foot, and 10 to 15 rem per hour at 3 feet. 

A second 40-ml sample was drawn and carried into an adjacent

laboratory for boron concentration measurement. The area radiation 

monitor in the primary sample laboratory increased from 20 to 

800 millirem per hour, as the worker entered the room. Both workers 

handled the sample containers unshielded at contact with gloved hands, 

while hand-carrying the samples for disposal.

205

Both workers and a third worker involved with the boron measurement 

received contamination on a finger, wrist, and other parts of their 

bodies. The workers attempted decontamination for several hours. All

went home with some areas of their bodies having fixed contamination, 

such as 50 millirem per hour on a finger and 30 millirem per hour on a 

wrist. Residual contamination remained on small areas of one worker’s 

skin for over 30 days, which added to the total dose to the skin. Each 

worker supplied urine samples the next day, but these were not 

promptly analyzed by the licensee. The doses that the two workers 

received were later calculated by an NRC radiation specialist. One 

worker received an estimated exposure of 147 rem to the fingers, a skin 

dose to the top of the head in the range of 10 to 32 rem, and a 

whole-body exposure of 4.25 rem. The other worker received an 

estimated extremity dose in the range of 44 to 54 rem to the forearm 
72

skin and a whole-body exposure of less than 1 rem.206 (See 

NUREG-0600, Section 3.2.4.10.) 

• Whole-body exposures. On the evening of March 29, 1979, one of two 

engineers received whole-body exposure in excess of the quarterly limit 

while he toured the auxiliary building to check for leaks. Before the 

tour, the engineers were briefed by a radiation protection foreman, 

dressed appropriately for the degree of hazard, and were provided with 

two handheld radiation-detection instruments, one of which was a 

NRC inspectors monitoring radiological controls 

activities shortly after the accident. On March 30, the 

NRC became more actively involved in the radiation 

protection program by assigning NRC radiation 

specialists to each of the three shifts, providing around 

the clock coverage.
73

high-range gamma instrument. Shortly after entering the auxiliary

building, the high-range gamma instrument failed. They continued on 

together noting that the other instrument frequently pegged off-scale 

high (2 rem per hour). Upon exiting the building, the high-range

self-reading pocket dosimeter read about 3 rem for one individual and 

less than 1 rem for the other. Their personal dosimeters were read and 

indicated 3.14 rem and 0.170 rem. An NRC investigation concluded that 

this entry demonstrated a lack of implementation of basic radiation

protection training that could have resulted in serious consequences.207

(See NUREG-0600, Section 3.2.4.11.) 

• Whole-body exposures. On August 28, 1979, six workers entered a 

room in the TMI-2 fuel handling building to inspect and tighten leaking 

valves in preparation for decontamination of the area. Reactor coolant 

water, extremely contaminated from the March 28 accident, was leaking 

from the valves. The radiation survey instrument used by the workers 

showed a gamma dose rate in the room of 10 to 15 rem per hour and, in 

one small zone, 25 rem per hour. It was decided that the time limit on 

the presence of each worker in the radiation area was four minutes. 

What the survey instrument failed to disclose was that the beta radiation 

rates in the room were running as high as 2,500 rem per hour. It was 

later estimated that the workers had received doses from the beta 

radiation in excess of regulatory limits. The doses were as high as 

166 rem to the whole body in one instance, and 161 rem in another. No 

indication of medically significant effects on the personnel was 

identified by medical examination. The causes of the overexposures

were determined to be inadequate instrumentation for radiation 

detection and a failure to require adequately-protective clothing for the 

workers. Corrective action was taken under NRC direction.208

• Extremity exposures. On September 25, 1989, two workers received 

overexposures to their extremities while unknowingly handling a piece 

of fuel debris in the TMI-2 reactor building decontamination facility

inside the reactor building. Both workers were performing flushing 

decontamination operations when one worker picked up what was 

thought to be a nut lying on top of the drain grating in the flush facility. 

The worker picked up the material and tossed it toward a trash bag, but 

missed. The other worker picked up the material, presumably to place it 

in the trash bag, and was told by the first worker to leave it alone. The 

first worker picked up the debris again and set it down 8 feet away.

Management was not informed of this event until the following day 

when the first worker asked a radiation control technician (RCT) about 

the implications of handling fuel debris. Fortunately, the RCT realized 
74

from his questions and further discussions that he had handled a piece of 

corium. The licensee conducted numerous reviews and assessments of 

the incident, as well as medical examinations of and consultations with 

the two workers. 

A subsequent radiation survey performed on the material indicated 

contact dose rates of 1,320 rem per hour of gamma radiation, and 

11,580 rem per hour of beta radiation. The highly-radioactive material 

was placed in the reactor vessel using long-handled tools. The dose 

assessment, including re-enactments using a full-scale mockup of the 

decontamination facility, estimated that one worker might have 

experienced an extremity exposure to the hands in the range of 75 to 

375 rem, and the second worker might have experienced an exposure to 

the hands in the range of 18.75 to 75 rem. The wide range of these 

estimates was attributed to the uncertainty in the estimates of the time 

that the workers were handling and were near the radioactive source.

The licensee instituted a site-wide training program on the appropriate 

handling of fuel debris.209 The RCT demonstrated an attitude that 

needed to be fostered for conducting good health physics practices.

Radiological Protection Programs and Activities. Many programs and 

activities helped to improve radiation protection practices at TMI-2. New 

approaches were needed in a number of basic worker-protection and 

dose-reduction areas, including protective clothing, respiratory protection, 

Technicians taking radiation readings inside the reactor building. Notice the 

personnel dosimetry attached to the back of the technician.
75

dosimetry, radiation field and contamination characterization, 

exposure-tracking systems, dose-reduction planning, procedures, training, 

and robotics. In the summer, when the temperature in the reactor building 

approached 33°C (90°F), an ice vest was commonly used by workers to 

control heat stress and extend work periods.210, 211 Several respiratory 

protection breathing apparatuses were developed or adapted to extend staytimes in the reactor building, including a power air-purifying respirator and 

a power air-purifying hood.212 Some of the radiological protection programs 

and activities are summarized below: 

• Increased radiation protection oversight. On March 30, 1979, more 

control was established by the licensee over in-plant radiation hazards. 

Radiation protection technicians were being used onsite to establish 

control points. Temporary dose-accountability forms were implemented 

that evening at the site access point. The whole-body counter was in 

operation and individuals were being counted. On the afternoon of 

March 30, the NRC became more actively involved in the radiation

protection program. Five NRC radiation specialists were assigned to 

each of the three shifts, providing around-the-clock coverage. In 

addition, two NRC health physicists provided technical guidance.213

A summary of radiation protection findings that were identified by NRC 

inspectors during the early months of the recovery, were documented in 

the NRC investigative report, “Investigation into the March 28, 1979 

Three Mile Accident by Office of Inspection and Enforcement” 

(NUREG-0600).214

• NRC Special Panel on TMI-2 Radiation Protection Program. On 

September 26, 1979, the director of NRC’s Office of Nuclear Reactor 

Regulation, created a special panel to provide an independent review of 

the licensee’s existing and planned radiation protection program that 

was intended to keep personnel’s exposure to radiation at ALARA 

levels during the recovery and cleanup of Unit 2. The panel of external 

experts recommended the following in its report (NUREG-0640): an 

upgraded radiation safety program for major radiological recovery 

efforts; independent assessment of the proposed upgraded program; and 

a management plan and firm schedule for resolution of existing 

technical and management deficiencies in the radiation safety 

program.215

• Dose-reduction program. Because of the increasing amount of work 

inside the reactor building, a major effort to reduce dose rates was 

initiated by the NRC and the licensee in December 1982. The objective 

of this effort was to identify and eliminate or shield as many sources of 
76

radiation exposure as possible in occupied portions of the reactor 

building before and during reactor disassembly and defueling. The 

dose-reduction program initially focused on both the 305-foot 

entry-level elevation and the 347-foot operating-floor elevation, where 

most of the defueling work took place.216 Dose-reduction techniques 

included (1) shortening the transit time of workers in the reactor 

building by opening both personnel airlocks and modifying the 

ingress/egress paths; (2) decontaminating by water flushing discrete 

radiation sources, such as the air coolers, elevator shaft, and enclosed 

stairwell; (3) eliminating other discrete radiation sources by removal of 

trash and contaminated equipment; and (4) placing shielding at the 

305-foot elevation, such as lead curtains around the core flood tank, 

lead sheets on the covered floor hatch, and water columns and bladder 

shields around the open stairwell, elevator, and enclosed stairwell.217

Some of the more complex dose-reduction activities included extensive 

pre-task planning and mockup training for each task, decontamination 

of selected surfaces with chemicals, removal of paint, and scabbling (the 

mechanical removal of a thin layer) of concrete floors. These efforts 

resulted in significant reductions in the dose rate in the reactor 

building.218 In July 1984, workers entered the reactor building without 

respiratory protection for the first time since the accident, and, in 

accordance with ALARA principles, subsequent entries were made 

without respirators.

219 Dose-reduction activities to support defueling 

operations were completed in March 1985.220

Protective clothing worn during early entries into the reactor building. 
77

• PEIS supplement on occupational radiation dose. In October 1984, the 

NRC TMI Program Office issued Supplement 1 to the Programmatic 

Environmental Impact Statement or PEIS (NUREG-0683) after a public 

comment period. The purpose of this supplement was to re-evaluate the 

occupational radiation dose, and consequent health effects from 

cleanup, and address additional alternative cleanup approaches using 

information gathered since the original PEIS was prepared in 1980. The 

total radiation dose to cleanup workers was estimated to range between 

13,000 and 46,000 person-rem as opposed to earlier estimates of 

2,000 to 8,000 person-rem. The higher estimates resulted from a more 

accurate characterization of radiation fields in the reactor building based 

on numerous worker entries.221

Radiation Detection Instruments and Systems. Special instrumentation, 

systems, and techniques were developed or modified to measure and 

characterize the unique radiation situation at TMI-2 for ensuring worker 

safety and determining the effectiveness of decontamination processes. Key 

instruments are summarized below:222

• Thermoluminescent dosimeter (TLD) pseudo cores to take beta 

radiation measurements of the building floors. This simple device 

included a beta measuring TLD mounted on top of a plastic ring to 

shield the TLD from beta radiations coming from the floor area outside 

A technician lowering a telescoping radiation detector down into the open 

stairwell to the reactor building basement. Measurements revealed radiation 

levels of 40 to 45 rem per hour at 5 to 7 feet from the surface of water. 
78

the ring. TLD pseudo cores replaced the need to manually drill and 

process concrete core samples.

• Wall and floor sampler to mill the concrete surface and collect the 

sample in a filter for offsite analyses. 

• Modified handheld ion chamber detector to provide omnidirectional 

detection for gamma measurements and over 180 degrees for beta 

measurements. This modification provided a wider field of view for 

rapid directional exposure measurement of beta and gamma 

contamination. 

• Modified handheld tungsten-shielded, Geiger-Mueller detector with a 

conical lead collimator on the face of the probe to reduce the field of 

view from 140 degrees to 80-90 degrees. This modification provided 

rapid and accurate directional exposure measurements.

• Mobile radiochemistry laboratory to perform transuranic and 

radionuclide analyses of high-activity (less than 5 R per hour) liquid and 

solid samples. Higher activity samples had to be packaged, 

characterized, and shipped to offsite laboratories for analyses. This 

mobile laboratory was provided by the DOE. 

Water level in the basement reached the first stairwell landing indicating 

about 8 feet of water accumulation.
79

• Radiation mapping and ALARA planning system to provide 3-

dimensional maps of radiation exposures in plant areas and components, 

and to train and plan for missions in contaminated areas, and track their 

results.

• Improved personnel dosimetry system to approximate in-containment 

beta source conditions. The system included a modified 4-element 

dosimeter and an automated system used at TMI-2 each month to 

process that data from about 6,000 dosimeters.

Cumulative Worker Exposure Over 10 Years. Although worker activities 

at TMI-2 have been quite different than 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, with 

the cleanup about 99 percent completed, the collective dose to all workers 

was 6,180 person-rem. This was comparable to the collective occupational 

radiation exposure that was estimated in the original PEIS (2,000 to 

8,000 person-rem).223, 224

Document Collections. Documents relating to worker-protection and 

dose-reduction activities provided on the DVDs include the following: 

• an assortment of documents relating to radiation protection, such as 

annual reports of personnel exposures; radiation protection plans; 

dose-reduction plans; radiological training programs; and 

correspondence (see the DVD folder, Worker Protection) 

• an annual summary of dose-reduction activities provided in the NRC’s

annual reports (see the DVD folder, NRC Annual Reports) 

• the report, “TMI-2: Lessons Learned by the U.S. Department of Energy, 

A Programmatic Perspective,” which summarizes applications of 

research and development in the areas of worker protection (see the 

DVD folder, DOE/National Laboratory Status Reports) 

Defueling tool: spade bucket hydraulic attachment.
80
81

5 Data Acquisition and Analysis

The collection and analysis of data contributed to the safe stabilization of 

the damaged plant and supported planning of decontamination and defueling

activities. The TMI-2 accident also provided research opportunities that 

added to the knowledge about light-water-reactor behavior following a 

severe accident. Information that was required to ensure stable plant 

operations included measurement of reactor core and plant conditions; 

characterization of radioactivity inside the plant (for the design of recovery 

systems and planning of activities); and verification of recovery system and 

activity performance. The characterization of damaged fuel and internal 

reactor vessel components supported defueling planning, preparations, and 

operations, as well as interim storage, transportation, final storage, and 

disposal of fuel debris. 

The severe reactor core damage and behavior of fission products within the 

plant systems at TMI-2 gave government and industry researchers 

opportunities to: (1) measure the performance of instrumentation and of 

electrical and mechanical equipment within the reactor building during and 

after the accident; (2) determine physical damage to surfaces, components, 

and equipment caused by radiation exposure; (3) assess metallurgical and 

physical behavior of fuel, clad, and core components during and after the 

accident; and (4) assess new technologies for decontamination, and the 

disposal of radioactive waste.225

The DOE supported an extensive research program as directed by the 

U.S. Congress. Four organizations with a common interest in obtaining 

valuable information from the TMI-2 accident jointly established the TMI-2 

Information and Examination Program. The GPU, EPRI, NRC and DOE 

(hence the acronym GEND) signed the Joint Coordination Agreement226 on 

March 26, 1980, which identified the objectives and defined methods to 

implement the TMI-2 Information and Examination Program. The Technical 

Integration Office was established at TMI according to the terms of the Joint 

Coordination Agreement, with responsibility for implementation and daily 

management of the DOE programs. The office was staffed by INEL

personnel and reported to DOE’s site office. A technical working group (see 

above section on Advisory Groups) was formed to coordinate research

opportunities and activities.

227 In April 1984, the DOE signed a cooperation 

agreement with 17 Japanese nuclear power organizations to participate in 

research and development for recovery activities. As part of the agreement, 

DOE permitted Japanese organizations to participate in DOE research and 

development activities. In return, Japanese participants provided funding to 
82

DOE and research staff (up to 22 engineers at a time) at DOE laboratories 

and the TMI site.228, 229 

The NRC was responsible for evaluating and approving proposed in-plant 

research activities and implementing procedures to ensure that they were

conducted in a safe manner.230

Reactor Core Inspections. Of particular interest to all organizations 

involved in the cleanup and research, was the extent of damage to the 

reactor core and internal reactor vessel components. Data acquisition and 

analysis that were conducted inside the reactor vessel included visual “quick 

look” inspections using video cameras; ultrasonic measures; radiation dose 

measurement; grab sampling of loose core debris; and core samples using a 

core bore drilling machine.231 The most revealing examination, with the 

most far-reaching implications, was the first “quick look” visual inspection 

in July 1982, which showed extensive reactor core damage (fuel melting 

would not be identified and announced until four years later). Up to the time 

just before the quick look, the licensee had held out hope that the plant 

would one day be restarted. After quick look, the licensee’s primary focus 

The first closed-circuit video inspection of the upper reactor core region was 

performed 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 support tube.
83

was cleanup.232 A chronology of the reactor core inspection activities is

summarized below: 

• Insertion test of axial power shaping rods. In June 1982,233 an axial 

power shaping rod (APSR) insertion test was performed in an attempt to 

move each APSR leadscrew inside the control rod drive mechanism to 

its fully inserted position. During normal power operations, eight 

APSRs helped shape the power generation (neutron flux) uniformly 

across the reactor core to ensure even fuel burnup during the core’s

lifetime. These APSRs do not perform safety functions and do not drop 

into the reactor core during an automatic reactor shutdown. During the 

accident, the eight APSRs rods remained at a 25 percent withdrawn 

position while 75 percent of their length remained inserted in the fuel 

assemblies. One purpose of the test was thought to provide insight into 

the extent of damage to the reactor core and upper plenum. The analysis 

of the data concluded that the test provided little definitive information 

about the physical condition of the reactor core.234 (See

GEND-INF-038.) 

• “Quick look” camera inspection. The first closed-circuit video

inspection of the upper reactor core region was performed on 

July 21, 1982. A camera 1.5 inches in diameter and 12 inches long was 

inserted through the 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 five feet below the 

normal location of the top of the fuel assembly. It was believed that the 

rubble bed contained oxidized cladding, fuel fragments and/or pellets, 

poison material, and core structural components. No evidence of melted 

fuel pellets was found. Another inspection, on August 4, midway 

between the periphery and the center of the reactor core, also revealed a 

rubble bed approximately five feet below the top of the core region. 

Intact pellets, which might have been fuel or poison material, were 

visible on the top of the rubble. During a third inspection, which took 

place on August 12, a probe was poked through the rubble and it 

penetrated approximately one foot below the surface, indicating that the 

rubble in this region was composed of loose material.

235 The visual 

inspections did not find any apparent distortion of the upper plenum.236

(See GEND-030.) 
84

• Underhead characterization study. In late summer of 1983, a series of 

activities conducted for the underhead characterization study provided 

radiological data to support the radiological protection measures that 

would be needed to conceptualize and plan procedures for removal of 

the reactor vessel head. Previously, in December 1982, a simple vertical 

“quick scan” of the underhead region (using an ionization chamber 

lowered into an empty control rod drive mechanism opening) found 

higher-than-expected radiation levels, which ranged from 40 to 

600 roentgen per hour with an unknown beta-radiation contribution.

237

The results from the quick scan prompted a more detailed study.238

The initial underhead characterization study obtained dose rates around 

the reactor vessel head and service structure; re-measured dose rates 

under the head using a beta-shielded ionization chamber (“quick 

On July 21, 1982, the first video camera was inserted through a control rod 

drive mechanism motor tube for a “quick look” of the upper reactor core 

region. Shown is the top of the rubble bed of the reactor core cavity.
85

scan-2”), and a string of thermoluminescent dosimeters; performed 

remote visual inspections of the top of the upper plenum and underside 

of the reactor vessel head; obtained debris samples from the top of the 

plenum for pyrophoricity tests; examined the lower end of a leadscrew 

to characterize the stainless steel in the underhead environment; and 

evaluated the dose rate increase associated with moving several 

radioactive leadscrews from inside the reactor vessel to a position in the 

control rod drive mechanism’s motor tubes in the reactor vessel head’s

service structure. The study required the depressurization of the reactor 

coolant system to atmospheric pressure, a slight draindown of the 

reactor vessel’s water level to uncover the top one foot of the upper 

plenum (10 feet above the reactor core region), and the removal of a 

control rod drive mechanism.239 The NRC established a contract with 

DOE’s Pacific Northwest Laboratory to review the radiation 

measurements, fission products plate-out (absorption) on the upper 

plenum, and other chemical phenomena.240

The results of the data analysis revealed the following: no visual 

structural damage on the upper plenum; areas inspected on top of the 

plenum appeared relatively free of debris (though some sedimentation 

appeared to be present on some horizontal surfaces);241 pyrophoricity 

tests on two samples of material from the plenum surface proved 

negative;

242 gamma radiation fields in the range of 300 to 700 roentgen 

per hour were measured in the space formed by the underside of the 

reactor vessel head and the top of the plenum;243 and no significant 

dose-rate increase was observed on the service structure platform 

following the withdrawal of four leadscrews.244 The analysis of the data 

from the underhead characterization study supported plans to remove 

the reactor vessel head without flooding the refueling canal (also called 

“dry lift”).245

• Reactor core debris sampling program. The safety evaluation report for 

the underhead characterization study was amended to include two new 

activities that were sponsored by DOE’s TMI Reactor Evaluation 

Program: a reactor core debris sampling program and reactor core 

topography program. The reactor core debris sampling program 

provided data that was essential to prepare for future reactor vessel 

defueling activities and for the design of water cleanup systems and 

defueling canisters. In September and October 1983, the program 

obtained six specimens of reactor core debris by lowering specially 

designed tools into the reactor. The tool scooped up small samples of 

loose debris and transferred them into small shielded casks for offsite

shipment and analysis. The analysis of the samples included 
86

determining their particulate composition, particle size, fission-product 

content, and drying properties, as well as the fission-product 

leachability from the debris and the pyrophoricity of zirconium hydride 

or partially unoxidized zirconium fines.

246

The first three samples were taken at various depths in the center of the 

reactor core. The radiation field at one foot from one sampler (which 

had a capacity of one cubic inch) was approximately 3 roentgen per 

hour.247 The other three samples were taken midway between the center 

and the periphery of the reactor core at various depths in the debris 

bed.248 The six samples obtained during the grab-sample work were 

analyzed at the INEL and the Babcock & Wilcox (one sample) research 

facilities.249 Gamma-radiation levels of five samples, using a telescoping 

radiation instrument from 2.5 cm away, ranged from 3 to 36 roentgen 

per hour. Particle sizes ranged from about 0.6 cm to a fine debris.250

Five more samples were obtained in March 1984.251 Results from the 

sampling program were documented in the INEL report “TMI-2 Core 

Debris Grab Samples—Examination and Analysis” (GEND-INF-075).

• Reactor core topography program. In August and September of 1983, 

the core topography system, designed and built by DOE at INEL, was 

used to conduct an ultrasonic profile of the void area in the upper 

Core grab sample tool with fuel debris from the rubble bed obtained for 

the reactor core debris sampling program in September 1983.
87

reactor core region. A total of about 500,000 data points was obtained 

during system operation.252 The data and information obtained included 

the radial and axial extent of the reactor core cavity, the location of 

supported and unsupported fuel assembly end-fittings, and the location 

of the core cavity boundary with respect to structurally intact fuel 

assemblies. The analysis of this data supported upper plenum lift and 

defueling operations. 253

A clear plastic scale model of the damaged upper reactor core region 

was constructed at INEL in late 1983 based on ultrasonic 

measurements. This topographic model provided the most accurate 

indication of the extent of reactor core damage at that time. The volume 

of the cavity in the damaged area of the reactor core was measured at 

330 cubic feet or 26 percent of the original core volume. The bottom of 

the cavity ranged from 5 to 6 feet below the top of the core. Of the 

original 177 fuel assemblies, 42 appeared to contain some full-length 

fuel rods, but 23 of those 42 had less than 50 percent of the rods intact. 

The sonic mapping also revealed several partial fuel assemblies hanging 

from the underside of the upper plenum and indicated some distortion of 

the core former wall.254, 255 Each half of the plastic model was 

transferred to the NRC by DOE and the Smithsonian Institution. The 

complete model now resides at the NRC at Rockville, Maryland.

Topographic model of one-half of the cavity in the upper core region based on 

an ultrasonic profile from the reactor core topography program during 

August and September of 1983. The individual layers of the model represent 

2-inch contours within the cavity. The rods represent the locations and 

approximate lenghts of the axial power shaping rods within the core at the 

time of measurement. This model is currently located at NRC Headquarters.
88

• Reactor core video mapping. In April 1984, a comprehensive video 

mapping of the upper reactor core region between the plenum and 

rubble bed was completed. Video snapshots were assembled into a 

mosaic panoramic view of the rubble bed, core periphery, and the 

underside of the upper plenum grid section. Videos of the rubble bed 

showed broken fuel rods scattered around, fuel rod internal springs, 

intact fuel pellets, control rod assembly end couplings, and partially 

intact fuel assemblies around the periphery of the reactor core region. 

The video also showed unsupported partial fuel assemblies hanging 

from the underside of the upper plenum grid section, which had to be 

removed before plenum lifting. This map supported defueling 

operations planning, and eventually the removal of fuel debris.256, 257, 258

A few photographs and partial mosaics from the video mapping activity 

are provided in the DVD folder, Photo Gallery. 

• Video inspection of the lower head of the reactor vessel. In 

February 1985, the first video inspection of the reactor vessel’s lower 

head region was performed by guiding a small video camera and light 

through a gap between the upper plenum and core support flange during 

plenum jacking. The video inspection revealed re-solidified mass in the 

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.
89

reactor vessel lower head on February 20, 1985.259The camera followed

a path down to the lower head region between the reactor vessel wall 

and the internal thermal shield. The video revealed the accumulation of 

a substantial quantity (estimated at 10 to 20 tons) of accident-generated

debris with the appearance of a gravel pile. Over the next two years, 

four additional camera inspections of the lower head region were 

performed at different quadrants in the lower core support assembly

(July 1985, December 1985, July 1986, and February 1987). The 

analysis of the video data documented in the INEL report “TMI-2 

Lower Plenum Video Data Summary” (EGG-TMI-7429), revealed that 

large inhomogeneity existed in the physical appearance of the debris 

bed, ranging from a very fine dust-like and smooth surface, to a 

relatively flat, but coarse surface with large chunks, to a solid wall of 

lava-like debris.260 A few photographs from the video inspections are 

provided in the DVD folder, Photo Gallery.

• Vertical gamma profiles of the reactor vessel’s lower head region. In 

March 1985, an attempt was made to insert a miniature

gamma-radiation sensor into several incore instrument tubes to obtain 

vertical gamma profiles for characterizing fuel deposits that had settled 

on the bottom of the reactor vessel.261 The incore instrument tubes 

In February 1985, the first video inspection of the reactor vessel lower head

region revealed a gravel-like pile of core debris. Shown left is an instrument 

tube penetration.
90

penetrate the bottom of the reactor vessel, and extend upward into the 

reactor core region. A dummy detector wire of the same size and 

stiffness as the actual probes was inserted into 26 of the 52 incore 

instrument tubes; only one of the 26 wire probes reached the reactor 

vessel. The remaining 25 wires were blocked at various locations along 

the 120 feet of pipe between the bottom of the vessel and the pipe 

access at the incore seal table located in the reactor building. The single 

wire probe was inserted 22 inches into the lower reactor core region. 

A gamma sensor with a slightly larger diameter was inserted into the 

same instrument tube and reached the vicinity of the 

5-and-3/8-inch-thick lower head of the reactor vessel.

262

• Core stratification sample acquisition program. In July 1986, core 

boring operations were performed using a special computer-controlled 

drilling machine. The core stratification sample acquisition program 

was conducted as part of DOE’s TMI-2 Accident Evaluation Program to 

provide data on the material properties of the core debris. The core bore 

samples provided insights into fission-product release from the fuel, 

fission-product retention in the core, maximum temperature during the 

accident, and reactor core material interactions.263,264 A special

commercially available drilling rig was assembled on top of the 

defueling work platform to bore into the hard crust. Ten full-length core 

A damaged fuel assembly being examined at the INEL.
91

bore samples were obtained from all regions of the lower reactor core; 

these samples (approximately 2.5 inches in diameter and eight feet long) 

were analyzed at INEL, along with earlier samples of debris collected 

from the reactor vessel’s lower head. Video inspections of the reactor 

core below the debris bed were performed through several of the bore 

holes created by the drilling operations. Initial inspections indicated that 

peripheral fuel assemblies were essentially intact below the hard crust 

layer, but that the central reactor core region consisted largely of a fused 

mass of material.

265 The INEL report “TMI-2 Core Bore Examinations” 

(GEND-INF-092) documented results of the physical, metallurgical, 

and radiochemical examinations of the core bore samples and evaluated 

these results as they relate to the progression of core damage and 

fission-product behavior in the lower region of the reactor core. 266, 267

• TMI-2 vessel investigation project. In July 1989, a video inspection of 

the reactor vessel’s lower head revealed several cracks that appeared to 

be associated with incore instrument penetration nozzles. Higher-quality 

color videos and a mechanical probe were used in August to obtain 

Tears or cracks were found in the cladding of the lower reactor vessel head

around nozzle E-7 (instrumentation penetration stub upper right). These 

cracks were analyzed by Argonne National Laboratory and were attributed to 

stresses associated with the thermal gradient in the thick-walled carbon steel 

vessel during the heating and cooling phases of the accident.
92

better information on the cracks. The cracks appeared to be up to 

approximately 6 inches in length, 0.25 inches wide, and about 

0.19 inches deep.

268 Penetration into the thick reactor vessel steel was 

later determined to be superficial.

269

In February 1990, an international research effort obtained metallurgical 

samples from the reactor vessel’s lower head after defueling was 

completed. The program, which was sponsored by the NRC’s Office of 

Nuclear Regulatory Research and the Organisation for Economic 

Co-operation and Development, evaluated the potential modes of failure 

and the reactor vessel’s margin of failure during the TMI-2 accident. 

The condition and properties of material extracted from the reactor 

vessel’s lower head were investigated to determine the extent of damage 

to the lower head by chemical and thermal attack, the thermal input to 

the reactor vessel, and the margin of structural integrity that remained 

during the accident. A total of 15 “boat” samples were obtained from 

the reactor vessel’s lower head. In addition, 14 incore instrument 

penetration nozzles were cut off 1 to 2 inches above the reactor vessel’s

lower head, and obtained as samples. Two incore instrument guide tubes 

were cut free from the flow distributor head as samples.270 Results from 

Investigations from the TMI-2 Vessel Investigation Project identified the 

location of a hot spot in the lower reactor vessel head, but concluded that the 

hot spot would unlikely have caused vessel failure due to creep rupture for the 

temperatures and pressures that occurred during the accident.
93

the reactor vessel investigation project were documented in a series of 

NRC reports by the INEL and Argonne National Laboratory. (See the 

DVD folder, Accident Data Analysis.) 

Fuel Detection Techniques. Various fuel detection techniques were used to 

measure fuel debris and deposits internal and external to the reactor vessel. 

The licensee was required to document the measurements and calculations 

that were performed to ensure that the plant had been defueled to the extent 

reasonably achievable, and that the potential for a nuclear criticality had 

been precluded during normal and accident conditions.271 Measurements of 

residual fuel were required before the plant could begin post-defueling 

monitored storage (see section on After Defueling). Five general methods 

were used for fuel detection (detection of gamma rays, neutrons, alpha 

particles; sample and analysis; visual evidence). Key detection methods are 

summarized below:272, 273

• Sodium iodide (NaI) gamma spectrometer was used to survey much of 

the fuel deposits outside of the reactor building. This detector measured 

Ce-144 from fission products. 

• Other gamma spectrometers were used to measure fuel deposits in the 

auxiliary building. A high-purity germanium (HPGe) detector measured 

gamma radiation emitted from both Cerium-144 (Ce-144) and 

Europium-154 with improved energy resolution capability. A silicon 

lithium Si(Li) Compton recoil gamma ray spectrometer measured Ce144 gamma radiation in the A and B makeup and purification system 

demineralizer cubicles. 

• Solid-state track recorders were used to measure residual fuel in 

demineralizer cubicles and to profile fuel distribution within the reactor 

vessel from the annular gap outside the vessel. This passive detector 

used 93 percent enriched Uranium-235 fissile isotope foils to emit 

induced fast fission neutrons that created visible tracks in adjacent 

acrylic sheets.

• Copper activation coupons that become irradiated in the presence of a 

neutron flux. This method is insensitive to high gamma radiation fields. 

• Boron trifluoride (BF3) neutron detection system that thermalized fast 

neutrons from fissions for efficient counting in high gamma fields.

• Alpha fuel detectors were used to measure thin films of fuel debris on 

steam generator tube surfaces and on reactor coolant system surfaces.
94

• Active neutron detection technique was used to quantify smaller 

quantities of residual fuel. This photo neutron interrogation method used 

an antimony-beryllium neutron source to induce fission neutrons in the 

fuel that were measured by a helium-4 fast neutron recoil proportional 

counter. This method was more sensitive than passive counting for 

quantifying small deposits of fuel. 

Document Collections. Research tasks funded by NRC were documented 

in NUREG reports. Research results from the TMI-2 Information and 

Examination Program were documented in the GEND-series reports and 

DOE laboratory reports. GPU and EPRI documented the data acquisition 

and analysis tasks they performed in their own reports. Most reports from 

Government-sponsored work are provided on the DVDs. The EPRI report

“TMI-2 Post-accident Data Acquisition and Analysis Experience” (NP7156s) provides a concise overview of data activities during the TMI-2 

cleanup (not provided on the DVDs). 

The first sample of sludge from the reactor building basement was taken using 

a small scoop in June 1982. Later in 1982, three additional samples were 

taken from different locations using solenoid-operated samplers (top right).
95

Document collections relating to data acquisition and analysis tasks are 

listed below: 

• Tasks that generally supported plant recovery and cleanup activities are 

provided in the DVD folder, Recovery Data Analysis. Research topics 

included the following:

o pyrophoricity studies of potentially combustible core debris 

materials

o examination of debris on the reactor vessel’s lower head

o examination of fuel assembly components 

o criticality studies of the reactor core, reactor vessel, reactor coolant 

system, reactor building sludge, and cleanup system filters

o the reactor building entry program

o the dose-rate reduction program (see the section on Worker 

Protection above) 

A highly-radioactive EPICOR II prefilter liner being examined in a hot 

cell at the Battelle Columbus Laboratory. This examination and research 

at other national laboratories provided information on the processing of 

contaminated ion exchange media and on the degradation of these media 

and liners.
96

o polar crane refurbishment

o characterization of the sludge in the reactor building’s basement

o radiation and contamination surveys of building surfaces and 

internal to system components 

o characterization of the reactor building’s atmosphere 

o the reactor building gross decontamination experiment 

o the fission-product inventory program, which included radionuclide 

mass balance in the plant

o characterization of accident-generated water samples in various 

systems and components 

o sample acquisition from reactor building concrete surfaces

o characterization of and accountability for fuel deposits outside the 

reactor vessel

o the citizen radiation-monitoring program 

A telephone inside the reactor building during the accident was determined 

by INEL to sustain high temperatures from the hydrogen burn of 200 to 

220 degrees Fahrenheit. The telephone case, numeral ring, and dial were 

distorted and the cord coils relaxed.
97

o characterization of radioactive waste forms

o hydrogen production in spent filter canisters 

o effects of radiation on organic ion-exchange resins

o characterization and testing of EPICOR II waste form, resin, and 

liner integrity

• Tasks that improved the understanding of severe core damage accident 

phenomenology and its effects are provided in the DVD folder,

Accident Data Analysis. Research topics included the following: 

o TMI-2 accident forensic investigations

o benchmark thermal-hydraulic and severe-accident computer 

analysis codes with TMI-2 data

o external influences affecting the accident

o core relocation and debris-bed coolability

55-gallon drums were partially collapsed by the external pressure caused by 

the hydrogen burn inside the reactor building. This drum damage and lack of 

air duct damage (upper right) was consistent with a pressure pulse that 

developed over seconds by a deflagradion and not by the passage of a 

detonation wave. (The undamaged drum on the left was open to the 

atmosphere, thus did not experience a crushing differential pressure.)
98

o radionuclide (source term) behavior inside the plant

o in-vessel fuel distribution characterization using gamma-ray and 

neutron dosimetry 

o response of nuclear and non-nuclear instrumentation during the 

accident

o hydrogen generation and burning inside the reactor building

A sampler, called a water and sludge sampling device, was designed to 

simultaneously draw eight samples from the reactor building basement 

water.
99

• Tasks that were associated with the DOE-sponsored TMI-2 

Instrumentation and Electrical Program are provided in the DVD folder,

Instrumentation and Electrical Evaluations. The program evaluated 

instrumentation and electrical equipment for the effects of exposure to 

steam, spray, radiation, hydrogen burning, and resultant overpressure, as 

well as long-term exposure to moisture. Components inside the reactor 

building that were studied included radiation monitors, pressure 

transmitters, loose-parts monitors, various switches and contacts, 

solenoid operators for valves, and various other devices that suffered 

moisture intrusion. Summary reports are also provided in the DVD 

folder. 

Several electrical components from the reactor building were retrieved and 

examined as part of the DOE-sponsored Instrumentation and Electrical 

Program. Shown is the removal of an area radiation monitor (HP-R-211) 

from the northeast elevator wall.

Thermowell of a “worst case” platinum resistance thermometer removed from 

the reactor coolant system hot-leg showed radioactive surface deposits (right). 

This device was exposed to superheated steam during the accident; however, 

its calibration or its time response was not adversely affected by the accident.
100

Submerged demineralizer system inside the spent fuel pool.
101

6 Waste Management

The TMI-2 accident and subsequent cleanup challenged the management of 

various forms and concentrations of radioactive waste. The management of 

highly contaminated water, fuel debris, and related solid-waste byproducts 

included 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, while the possible generation of flammable 

gases inside sealed radioactive waste containers was a potential hazard. 

As part of the TMI-2 Information and Examination Program, DOE and the 

NRC sponsored research and development in the areas of volume-reduction 

techniques, performance of ion-exchange media, control of combustible 

gases, waste-disposal techniques, and radioactive material shipping cask

designs.274 The NRC and DOE signed a memorandum of understanding 

(MOU) that specified interagency procedures for the removal and 

disposition of nuclear waste resulting from the cleanup, including fuel 

debris. The memorandum and addendum ensured that the TMI site did not 

become a long-term waste disposal facility.275

Key Actions. A chronology of key NRC and DOE actions related to waste 

management is provided below. Most actions required research, 

development, safety evaluations, reviews, and approvals. Thorough 

overviews of waste management and transportation at TMI-2 are provided 

in the EPRI report, “TMI-2 Waste Management Experience” 

(EPRI TR-100640) and the DOE-sponsored reports, “Historical Summary of 

the Fuel and Waste Handling and Disposition Activities of the TMI-2 

Information and Examination Program: 1980–1988” (EGG-2529), and 

“Historical Summary of the Three Mile Island Unit 2 Core Debris 

Transportation Campaign” (DOE-ID-10400). Commission policy statements 

and NRC orders relating to waste management are further discussed in the 

prior section on Management and Oversight. 

• Policy statement on environmental assessments. On May 25, 1979, the 

Commission issued a policy statement directing NRC staff to prepare an 

environmental assessment regarding proposals to decontaminate and 

dispose of radioactively contaminated waste water. The policy required 

assessments on the decontamination of intermediate-level waste water 
102

using the EPICOR II system, alternatives to discharge of waste water 

into the Susquehanna River, and decontamination and disposal of 

high-level waste water.276

• EPICOR II environmental assessment. On August 14, 1979, in 

response to the Commission policy statement of May 25, 1979, NRC 

staff prepared and sent out for public comment an environmental 

assessment for the use of the EPICOR II filtration and ion-exchange 

decontamination system to remove radionuclides from 

intermediate-level radioactive waste water held in storage tanks in the 

TMI-2 auxiliary building (NUREG-0591). The proposed action was 

limited to cleanup and storage of waste water. The action also included 

the impacts of wet solid waste generated from EPICOR II operation, 

such as temporary storage, packaging, handling, transportation, and 

burial.277

• Order to operate EPICOR II. On October 16, 1979, the Commission 

issued a “Memorandum and Order” directing the licensee to operate the 

EPICOR II to decontaminate intermediate-level radioactive waste water 

held in storage tanks in the TMI-2 auxiliary building. In response to that 

The basement in the reactor building contained about 8 feet of highly 

contaminated water. The submerged demineralizer system started processing 

basement water on September 9, 1981. 
103

order, NRC staff issued an “Order for Modification of License” two 

days later to permit EPICOR II system operations. EPICOR II began

operation on October 22.278

• City of Lancaster agreement. On February 27, 1980, the NRC and the 

City of Lancaster, Pennsylvania, signed a litigation settlement 

agreement that prohibited the discharge of accident-generated water in 

the Susquehanna River until December 31, 1981, or until NRC 

completed necessary environmental reviews.279 

• GEND research coordination agreement. On March 26, 1980, the 

coordination agreement for the TMI-2 Information and Examination 

Program was signed by GPU, EPRI, NRC, and DOE (also known as 

GEND) to provide research and development coordination on efforts to 

manage wastes at TMI-2.280

• Final PEIS. On March 9, 1981, the NRC staff issued the Final 

Programmatic Environmental Impact Statement (PEIS) relating to the 

decontamination and disposal of radioactive wastes resulting from the 

TMI-2 accident (NUREG-0683). The PEIS was an overall study of the 

activities necessary for decontamination of the facility, defueling, and 

disposition of the radioactive wastes.281

• Policy statement endorsing PEIS. On April 27, 1981, the Commission 

issued a policy statement endorsing the final PEIS. With the exception 

of the disposition of processed accident-generated water (the 

Commission wanted to decide this issue later), the Commission directed 

the staff to act on each major cleanup activity (without further 

Commission direction) if the activity and associated impacts fell within 

the scope of those assessed in the PEIS.282

• Order to operate the submerged demineralizer system (SDS). On 

June 18, 1981, the NRC staff issued an order to direct the licensee to 

promptly commence the operation of SDS with effluent polishing by the 

EPICOR II system. The order directed the complete processing of the 

remaining intermediate-level contaminated water in the auxiliary 

building tanks (100,000 gallons), and the highly contaminated water in 

the reactor building sump (700,000 gallons) and the reactor coolant 

system (95,000 gallons). The approval to operate SDS did not include 

water disposal. All processed water had to be stored in existing onsite 

tanks; however, portions were allowed to be cycled for reuse within the 

plant. The order required that decisions related to the disposition of 

processed water be made by the Commission at a future date.283 The 
104

NRC review of the SDS formally started when the licensee submitted 

their technical evaluation report on April 10, 1980. After a number of 

design changes by the licensee and technical questions from the NRC, 

the NRC safety evaluation report (NUREG-0796) was issued in 

June 1981.284 In its review, the NRC determined that the potential 

environmental impacts from the proposed operation of the SDS were

within the scope of the PEIS. On July 10, 1981, the SDS began 

processing water from the reactor coolant bleed tanks. On 

September 22, 1981, the water from the reactor building sump and 

basement was pumped to the SDS feed tanks located outside the reactor 

building. The following day, the SDS began processing this water. The 

processing of water from the reactor building’s basement was 

completed in May 1982. In June 1981, the NRC issued the certificate of 

compliance for the shipping cask that was designed and fabricated to 

transport spent SDS vessels at first to a DOE research facility and 

subsequently to a commercial low-level radioactive waste burial 

facility.285

• NRC-DOE MOU. On July 15, 1981, the NRC and DOE signed MOU to 

formalize the working relationship between the two agencies with 

respect to the removal and disposition of solid nuclear wastes generated 

during the cleanup of TMI-2. The DOE agreed to carry out research and 

Submerged demineralizer system started processing highly contaminated 

water in June 1981 from the reactor building basement, reactor coolant 

system, and auxiliary building tanks.
105

conduct tests on solid wastes shipped from the plant to DOE facilities 

that had generic information value. DOE also agreed to accept other 

waste that were too highly radioactive for disposal in commercial 

facilities and provided no research value. The licensee provided 

reimbursement to the DOE for the shipment, storage, and disposal of 

such waste. Most low-level waste associated with decontamination, 

such as some ion-exchange media, boots, gloves, and trash, was

disposed of by the utility in licensed commercial low-level radioactive

waste burial facilities. The original MOU provided plans for DOE to 

accept some fuel assemblies and samples for analysis characterization 

and archiving. Moreover, the original plan was to place the remaining 

balance of the reactor core in fuel storage containers and store the fuel 

in the TMI-2 spent fuel storage pool to await resolution of the 

commercial nuclear power spent fuel storage issue.286 However, this 

MOU was formulated before the first “quick look” inside the reactor 

vessel that revealed a severely damaged reactor core. 

• NRC-DOE MOU addendum. On March 15, 1982, the NRC and the 

DOE signed a revision to the MOU. The DOE agreed to accept the 

entire reactor core for research and development and for temporary 

storage at a DOE facility.287 During the first decade after the accident, 

the DOE accepted 50 EPICOR II pre-filters during the initial processing 

of radioactive waste water from the auxiliary building, 19 highly loaded 

SDS vessels, 3 SDS vessels with highly loaded cartridge filters, cut 

control rod leadscrew segments, and original cartridge filters from the 

makeup and purification system that filtered reactor coolant system 

letdown during the accident. The disposal by DOE of waste products 

that were not useful for research was funded by the licensee.288

• DOE-GPU agreement. On March 19, 1982, the DOE and GPU signed 

an agreement for the acquisition of the damaged TMI-2 reactor core by 

DOE. In the agreement, DOE acquired ownership of the damage core 

from GPU and arranged for the shipment of the entire core to a DOE 

facility for research. GPU reimbursed DOE for shipping, storage after 

the research period, and ultimate disposal.

289

• Environmental assessment for processed water disposal. In June 1987, 

the NRC issued Final Supplement No. 2 to the Programmatic

Environmental Impact Statement (NUREG-0683) dealing with disposal 

of accident-generated water. The supplement presented NRC’s 

evaluation of the licensee’s July 1986 proposal for disposing of 

approximately 2.3 million gallons of slightly radioactive water. This 

water was contaminated either during the accident or during subsequent 
106

cleanup operations. The proposed method involved the forced 

evaporation of the water at the TMI site over a period of two and 

one-half years. The residue from this operation contained small amounts 

of the radioactive isotopes tritium (hydrogen-3), cesium-137, and 

strontium-90 and larger amounts of nonradioactive boric acid and 

sodium hydroxide. This residue required solidification and disposal as 

EPICOR II system originally decontaminated 380,000 gallons of 

intermediate-level radioactive water held in the auxiliary building 

tanks. Each vessel contained ion-exchange resin. The vessel at the top 

of the photo was the prefilter demineralizer, the center vessel was a 

cation ion-exchanger, and the third vessel was a mixed-bed polishing 

ion-exchanger. Each was fitted with three quick-disconnect hoses: a 

liquid waste influent line, a processed waste effluent line, and a vent 

line with attached overflow hose.
107

low-level waste. The NRC evaluated the licensee’s proposal together 

with eight alternative approaches, giving consideration to the risk of 

radiation exposure to workers and to the general public; the probability 

and consequences of potential accidents; the necessary commitment of 

resources, including costs; and regulatory constraints. The NRC

concluded in the supplement that the licensee’s proposal to dispose of 

the water by forced evaporation to the atmosphere, followed by onsite 

solidification of the remaining solids and their disposal at a commercial 

low-level radioactive waste burial facility, was an acceptable plan. The 

supplement also established that no alternative method of disposing of 

the contaminated water was without question, clearly preferable to the 

licensee’s proposal.290, 291

Public hearings on the licensee’s proposal to evaporate 2.3 million

gallons of accident-generated water were held by an NRC Atomic 

Safety and Licensing Board, and concluded on November 15, 1988. On 

February 3, 1989, the Board issued a decision finding in favor of the 

licensee on all relevant issues. On April 13, 1989, the Commission 

upheld the Board’s decision. Construction of the evaporator system 

began in August 1989.292

• License amendment permitting processed water disposal. On 

September 11, 1989, the NRC issued Amendment No. 35 to the 

facility’s operating license, modifying the plant’s technical 

specifications by deleting the prohibition for disposing of 

accident-generated water. Disposal was allowed in accordance with 

NRC-approved procedures.293 The evaporator system began vaporizing 

accident-generated water on January 24, 1991, after a prolonged period 

of system testing, modification, and repair.294 On August 12, 1993, the 

decontamination and evaporation of 2.23 million gallons of 

accident-generated water was completed.295 The system evaporated

about 99 percent of the initial pre-processing volume of 2.3 million

gallons. The residual volume that remained in various tanks and 

building sumps was estimated to be about 18,500 gallons.296 The 

evaporator system was disassembled and shipped offsite by the owner 

and operator of the system.297

Water Processing and Storage Systems. Before defueling operations 

could begin, pre-accident and accident-generated waste water required 

decontamination of low, intermediate, and high levels of radioactivity. 

Additional holding tanks were constructed and existing tanks were managed 

to maximize storage capacity. After decontamination, the slightly 

radioactive water (still contaminated with tritium, a radioisotope of 
108

hydrogen, and traces of fission products) had to be stored onsite until the 

NRC Commissioners decided the ultimate disposal requirements. Until then, 

the NRC permitted the licensee to reuse this water for surface and 

equipment decontamination activities inside the auxiliary and fuel handling 

building and reactor building. This water was reprocessed through water 

cleanup systems and stored onsite in large carbon-steel plate-welded tanks.

During defueling operations, several water cleanup systems were designed 

and installed to maintain water clarity for the manual removal of fuel debris 

inside the reactor vessel and remove fuel debris contamination resulting 

from the movement of defueling canisters from the reactor vessel to the 

spent fuel storage pool. Keeping the water inside the reactor vessel free of 

microorganisms was a complex technical problem that hampered visibility 

and eventually disrupted defueling operations for a period of time. 

More than 2.3 million gallons of processed accident-generated waste water 

that were stored in various tanks at TMI-2 was disposed of by evaporation 

over a 31-month period. The systems and facilities that were reviewed and 

approved by the NRC to store and process waste water are described below: 

• The fuel pool waste storage system (also known as the “tank farm”)

was installed to provide feed staging for the processing of the water 

from the reactor building’s basement by the EPICOR II system and 

submerged demineralizer system. (See Plant Stabilization section.) 

 Typical SDS vessel. Typical EPIROR II liner.
109

• The EPICOR I system was installed within a week after the accident to 

process low-activity, non-accident-generated liquid waste water, mainly 

generated from the TMI-1 outage. The system included one pre-filter 

liner, one demineralizer liner, and two receiving tanks. Each 

6-foot-by-6-foot liner was made of carbon steel. Over its 19-month 

lifetime, EPICOR I processed over 1.3 million gallons at an average 

flow rate of 10 gallons per minute.

298

• The auxiliary building emergency liquid cleanup system (also known 

as “EPICOR II”) was designed and installed after the accident to clean

up about 450,000 gallons of intermediate-level waste water. This water 

was held in various storage tanks and sumps inside the auxiliary and 

fuel handling building. The EPICOR II system was a liquid waste 

processing system to decontaminate water by filtration and

ion-exchange demineralization. This process for treatment of 

radioactive water was (and currently is) standard practice in nuclear 

power plants. The EPICOR II system was housed in an existing onsite 

building that was originally intended for chemical cleaning of the steam 

generators for TMI Units 1 and 2.299 

EPICOR II system originally decontaminated 380,000 gallons of intermediatelevel radioactive water held in auxiliary building tanks. Shown is the 

schematic of the flow path in 1980.
110

The EPICOR II system operated in three main configurations during its 

service life at TMI-2. 

o The original configuration included one pre-filter liner followed by 

two demineralizer liners and cartridge filters to retain any resin 

fines and particulates. The 4-foot-by-4-foot carbon-steel pre-filter 

liner contained a pre-coat material to remove particulate radioactive 

waste, such as activated corrosion products, and other suspended 

solids. The pre-filter also contained cation resin which was highly 

efficient for the removal of cesium and other cationic radionuclides 

from the waste stream. The first demineralizer (a 4-foot-by-4-foot 

carbon-steel liner) contained cation resins to remove cesium. The 

second demineralizer (a 6-foot-by-6-foot carbon-steel liner) 

contained mixed cation and anion resins to remove cesium and 

iodine, respectively. After processing, the water was collected in a 

clean water receiving tank for sample measurements of radionuclide 

concentrations. The water could be reprocessed as necessary to 

achieve the desired concentration of radionuclides.300, 301 The 

processing of intermediate-level waste water inside the auxiliary 

building began on October 22, 1979, and was completed on 

December 2, 1980. During this period, a total of 1,051,047 gallons 

of water was processed, with about half of this amount being 

recycled processing.302 

o EPICOR II was used to remove residual radioactivity from 

submerged demineralizer system (SDS) effluents and to process 

miscellaneous wastes from September 11, 1981, to April 1987.303

This configuration included two 6-foot-by-6-foot liners in series 

followed by one 4-foot-by-4-foot liner. The first-stage 

demineralizer generally removed sodium and other non-radioactive 

chemicals, although low concentrations of radionuclides were also 

removed. (SDS removed most of the gross cesium and strontium.) 

The second demineralizer reduced radioactive concentrations 

through ion exchange and filtering. The last liner refined the 

effluent water and caught resins that broke through the second 

liner’s retention screens.

304

o After the SDS was removed from operation in 1988, EPICOR II 

was the primary system to clean up waste water that was mainly 

generated from building decontamination activities. The system 

configuration was the same as before; however, a high-integrity 

container (HIC) loaded with zeolite resins was placed in the first 

position to act as a roughing filter to remove gross cesium and 
111

strontium radionuclides. The HIC was followed by the two standard 

EPICOR II carbon-steel liners. This configuration allowed 

EPICOR II to replace the aging SDS. The HIC was similar in size to 

the carbon-steel 4-foot-by-4-foot liner, but the HIC was constructed 

from very high-grade stainless steel.The HIC allowed the loading 

of higher concentrations of radionuclides for burial at a commercial 

low-level radioactive waste burial facility.

305

Upon completion of the TMI-2 cleanup activities, the EPICOR II system 

was released to general site use and included under the TMI-1 facility 

operating license.306

• The submerged demineralizer system (SDS) was designed and installed 

in the TMI-2 spent fuel pool to clean up high-level radioactive 

accident-generated waste water from the reactor building’s basement, 

reactor coolant system, and reactor coolant bleed tanks. Processed water 

from the processed water storage tanks was used to fill the spent fuel 

pool for shielding of the highly radioactive spent filters. The SDS 

consisted of a liquid waste treatment subsystem, a gaseous waste 

treatment subsystem, and a solid waste handling subsystem. The liquid 

waste treatment subsystem was designed to remove cesium and 

Submerged demineralizer system started processing highly contaminated water 

in June 1981 from the reactor building basement, reactor coolant system, and 

auxiliary building tanks.
112

strontium from the high-activity waste water by filtration and ion 

exchange. The liquid waste treatment subsystem included the following 

primary components (in the order of the flow path):307

o Pre-filter and final filter consisting of sand to remove particulates. 

o Feed tank system (known as the “tank farm”) of four 15,000-gallon 

storage tanks to hold contaminated water before processing it in the

ion-exchange vessels. 

A typical submerged demineralizer system ion exchange vessel. About 

4.5 feet in height and 2 feet in diameter, each vessel included five 

nozzles on the top: inlet, outlet, vent, and general vessel access (two 

nozzles) for loading and unloading the ion exchange media.
113

o A 30-gallon-per-minute pump. 

o Two parallel trains of identical inorganic zeolite-filled ion-exchange 

vessels. The first train included two vessels to process water from 

the reactor coolant system and reactor coolant bleed tanks. The 

second train included four vessels to process the highly 

contaminated water from the reactor building’s basement. Each of 

these vessels was a stainless steel pressure vessel two feet in 

diameter and four feet tall. Each vessel also contained a catalyst bed 

to recombine the available hydrogen and oxygen (generated by 

radiolysis of water) back into water.

o Post-filter to retain zeolite fines from new zeolite beds.

o Two 12,000-gallon monitoring tanks to hold processed water for 

measurement. SDS effluents in the tanks could be reprocessed or 

sent to the EPICOR II system to remove residual contamination (a 

process known as “polishing”). The final processed water was 

transferred from the monitoring tanks to the processed water storage 

tanks for reuse or deposition. 

o Sample connections, sample glove boxes, and a continuous in-line 

radiation monitor to estimate curie loadings and monitor vessel 

efficiency for process control.

The first vessel in a train upstream was the first to be removed from 

service. A spent vessel was moved to the dewatering station and then to 

the storage rack in the spent fuel pool using the fuel handling building’s

crane. Downstream vessels were disconnected and moved upstream one 

position. A new vessel was installed at the last open position. Spent 

ion-exchange resins and filters were loaded into a shielded 

transportation cask underwater. The cask with the vessel inside was

moved from the spent fuel pool, sealed, decontaminated, and loaded 

onto a truck trailer for shipment.308

DOE national laboratories provided technical assistance to the licensee 

and their contractors to develop the SDS.

309 The SDS started testing on 

July 12, 1981, started processing reactor building basement water on 

September 9, 1981, and started processing reactor coolant system water 

on May 21, 1982.310 During its 7-year service life, SDS processed 

4,566,000 gallons of water.

311 Details of the SDS design, installation, 

and testing were documented in the INEL report “Submerged 

Demineralizer System Processing of TMI-2 Waste Water” (GEND-31).
114

Following the initial SDS pumping of 50,000 gallons of water from the 

reactor building sump on September 27, 1981, the incore thermocouple 

temperatures increased. On September 28, 1981, the temperatures 

stabilized and started a slight downward trend. The licensee reported 

that the temperature change appeared to have occurred as a result of 

lowering the water level in the reactor building’s basement below the 

point at which the water was in contact with the lower dome of the 

reactor vessel. The heat-transfer characteristics had changed because 

heat might no longer be conducted to the water in contact with the 

vessel and the temperatures would change until equilibrium conditions 

were established. The highest temperature of an incore thermocouple 

was approximately 147°F (a 12°F increase) and the calculated average 

temperature was 117°F (a 2°F increase).312

• The processed-water storage and recycling system was installed to hold 

processed water from the EPICOR II and SDS cleanup systems. The 

processed water was stored onsite in two 500,000-gallon tanks. The 

tanks became operational on July 8, 1981.313 A recycling system was 

later installed to transfer the processed water into the reactor building 

for decontamination activities. This water was subsequently reprocessed 

by the SDS and returned back to the processed water storage tanks for 

further use or to await ultimate disposal. The storage system included

two steel plate-welded tanks that were epoxy-lined and insulated. 

Processed water storage tanks, each with 500,000-gallon capacity, were 

installed to hold processed water from EPICOR-II and SDS cleanup systems. 

Insulation and heat tracing were installed for freeze protection (not shown).
115

A heat-traced subsystem prevented freezing in winter months. Two 

transfer pumps were used to transfer water from the processed water 

tanks.314 A decision was made not to build a dike around the tanks for 

water retention because it was determined that any liquid spillage from 

a tank rupture would not represent a significant radiological health, 

safety, or environmental hazard.315, 316 On completion of the TMI-2 

cleanup activities, the processed water storage system was released to 

general site use and included under the TMI-1 facility operating 

license. 317

• The internals indexing fixture (IIF) processing system provided 

interim reactor coolant system processing capability following reactor

vessel head removal to minimize radiation dose rates around the IIF. 

The IIF cylinder was mounted on the reactor vessel’s flange and the 

reactor coolant level was raised to partially fill the IIF. The IIF 

processing system consisted of a submersible pump located inside the 

IIF which transferred water from the IIF through the submerged 

demineralizer system to a reactor coolant bleed holdup tank. Reactor 

coolant-grade water was concurrently returned to the reactor vessel 

from a second bleed tank by a waste transfer pump to maintain the IIF 

level.318 The system was placed in operation in August 1984 and was 

replaced by the defueling water cleanup system in November 1985.319

• The defueling water cleanup system (DWCS) was used to remove 

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 system was composed of two independent systems: 

the reactor vessel cleanup system and the FTC/SFP-A cleanup system.

Each system included four modified defueling canisters (filter canisters) 

with sintered metal filter media to remove debris, mainly fuel fines 

(uranium oxide) and core debris (zirconium oxide), down to a 

0.5-micron rating. Each system had either one (FTC/SFP-A cleanup) or 

two (reactor vessel cleanup) 4-foot-by-4-foot demineralizer liners filled 

with zeolite resins to remove cesium. The liners were high-integrity 

containers, similar to those later used in the EPICOR II system. Each 

system had two submersible pumps that fed two filter canisters (two 

trains of two filters each), followed by one or two demineralizers.320 The 

FTC/SFP-A cleanup system was ready when the upper plenum was 

moved from the reactor vessel to the deep end of the refueling canal in 

May 1985. The startup of the reactor vessel cleanup system followed in 

November 1985.321
116

• Treatment of microorganisms. Problems plagued the cleanup systems 

early because of microorganism growth in the reactor coolant. In 

January 1986, a filamentous growth was observed on surfaces inside the 

reactor vessel. A rapid growth occurred when hydraulic fluid from the 

defueling tools was inadvertently spilled into the reactor vessel, along 

with the use of lighting in the reactor vessel, use of biologically 

contaminated water for makeup, and aeration of the coolant by the 

movement of tools in and out of the water and by startup of water 

cleanup systems. The resulting loss of visibility halted all defueling 

operations. Principal microorganisms found in the reactor coolant were 

bacteria and fungi, which were common in river water sediment. During 

the months following the accident, a large volume of river water entered 

into the reactor building through a leaking air cooler. River water and 

accident-generated water on the reactor building’s basement floor were 

processed by the submerged demineralizer system and stored in the 

processed water storage tanks. The biologically contaminated, 

processed water was reused for reactor coolant makeup. The hydraulic 

fluid used in defueling tools was also found to be contaminated. These 

microorganisms were resistant to high radiation fields, borated reactor 

coolant chemistry, and biocides.322 

Considerable time was spent attempting to control 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 1986.323 As the result of extensive 

research and testing, the NRC approved 200 parts per million hydrogen 

peroxide as the biocide. NRC’s evaluation concluded that the biocide 

concentration was compatible with existing water chemistry and water 

processing systems, as well as the effects on criticality, defueling 

canister catalytic hydrogen recombiners, and waste disposal. In 

addition, the increase in reactor coolant activity levels due to the 

increase in rate of cesium leaching from the debris bed would not 

impose a significant impact on worker safety with the implementation 

of normal radiological control practices.324 Inspections, corrosion

studies, and laboratory tests found no microbial-induced corrosion of 

components and defueling equipment in the reactor vessel.

325

• The modified DWCS. In addition to the microorganisms being filtered 

by the DWCS, colloidal suspensions of fine particles of silicon and iron 

oxides caused premature plugging of the sintered metal filter media in 

the filter canisters. A temporary reactor vessel water filtration system 

(see below) was installed until solutions to stop the microorganism 

growth and remove the colloidal material could be incorporated in the 
117

DWCS design. Under normal conditions, the filter efficiency increases 

as a cake is built up on the surface of the media. However, the 

suspended colloids were plugging the pores of the sintered stainless 

steel filter media before the filter cake could form. Modifications to the 

DWCS included a filter-aid feed system that injected a coagulant with 

body-feed material into the filter canisters to agglomerate the colloids to 

filterable sizes and promote cake buildup on filter media.326, 327 The 

modified reactor vessel cleanup system was restarted on 

January 8, 1987.328 Later in 1987, a crosstie was installed between the 

FTC/SFP-A cleanup system and the “B” train of the reactor vessel 

cleanup system that was modified with the coagulant and filter-aid feed 

system. This crosstie permitted the processing of FTC and SFP-A water 

without the need to make a similar modification in the FTC/SFP 

cleanup system located in the fuel handling building.329 Other 

operational improvements were made to the DWCS during its service 

lifetime.330

• The temporary reactor vessel water filtration system was used to clean

up reactor vessel water above the rubble bed and to maintain an 

adequate visibility for defueling. The system took suction from the 

internals indexing fixture and pumped the water back to the fixture

through a filter assembly. The temporary system was installed after 

discovering difficulties that the defueling water cleanup system was 

having with cleaning up reactor vessel water. The system included a 

filter, a pump, and a “knockout” canister. (The knockout canister, filter

canister, and fuel canister were specifically designed and fabricated to 

contain fuel debris.) The filter was a diatomaceous-earth pre-coat filter. 

A loaded filter was backwashed of diatomaceous earth and debris into a 

knockout canister or filter canister for disposal. The system was restored 

by establishing flow and injecting six pounds of clean diatomaceous 

earth into the pump suction to coat the filter’s internal leaves.

331 The 

temporary system operated from February 1986 to May 1987, after 

which the improved defueling water cleanup system took over the water

filtration function.

332 

• The in-vessel filtration system was a modified defueling water cleanup 

system that supported vacuuming operations for final cleanup of the 

defueled reactor vessel. The in-vessel vacuum system used a knockout 

canister and filter canister in series.333
118

• The processed water disposal system (evaporator) was used to dispose

of 2.3 million gallons (8.7 million liters) of processed 

accident-generated water. The system processed the water through a 

closed-cycle evaporator. The purified distillate was reheated and 

discharged into the atmosphere as a vapor, using the 100-foot-high 

TMI-2 exhaust stack. The vapor contained essentially all of the water’s 

tritium and a small fraction of its particulate contamination. The vapor 

was released to the atmosphere in a controlled and monitored manner. 

The remaining particulate contamination was concentrated in the 

evaporator bottoms, collected, and further concentrated to a dry solid. 

A packaging system prepared the dry solid waste in containers 

acceptable for shipment and burial in a commercial low-level

radioactive waste burial facility.

334 The 2.3 million gallons of processed 

water (influent) contained about 1,020 curies of tritium and about 

2.3 curies of all other radionuclides, such as 42 percent strontium-90, 

38 percent carbon-14, and 14 percent cesium-137. The water also 

contained nonradioactive contaminants, such as 150 tons 

(136,000 kilograms) of boric acid and 11 tons (10,000 kilograms) of

sodium hydroxide.335 About 99.9 percent of the dissolved radioactive 

contaminants (other than tritium) contained in the evaporator influent 

were collected as dry solid waste.336 

Temporary Solid Waste Storage. The NRC stated in its Programmatic 

Environmental Impact Statement (PEIS) that the TMI site should not 

Processed water disposal system disposed of 2.3 million gallons of processed 

accident-generated water by evaporation and disposal of solid residue.
119

become a permanent radioactive waste disposal site. This statement was 

based on the conclusion that TMI did not meet the criteria for a safe 

long-term waste disposal facility for storing damaged fuel and radioactive 

waste.337 (The Commission endorsed the PEIS in their April 27, 1981, 

policy statement.

338) The objective of the memorandum of understanding 

between the NRC and DOE, as amended, for the removal and disposition of 

radiological waste resulting from the cleanup of TMI-2 was to ensure that 

the TMI site did not become a long-term waste disposal facility.339, 340

Several onsite facilities were constructed for temporary storage of solid 

radioactive waste products from cleanup activities that were being readied 

for transportation. Solid waste included spent EPICOR I and EPICOR II 

resin liners; contaminated clothing, tools, and equipment; and 

decontamination materials.

• The interim storage facility was built to temporarily store EPICOR 

resin liners and filters until the construction of the solid waste staging 

facility was completed. The earthen facility was constructed 

underground from corrugated galvanized cylinders with a welded

bottom. A 3-foot-thick concrete cover provided shielding at the top. The 

ground provided radial shielding. The loaded liners were placed on a 

galvanized drip pan. The design requirements for the facility were 

provided to the licensee by the NRC341. The facility was ready for use 

on November 5, 1979.342 The facility was abandoned in place after the 

liners were relocated to the new solid waste staging facility.343

• The interim solid waste staging facility (also known as the “car 

port”344) was used to collect and temporarily store (stage) low-level

solid waste packages from both Units 1 and 2. All waste packages 

placed in the facility, such as liners, drums, and metal boxes, were 

already prepared for shipment. The facility performed passive functions 

of protecting waste packages from precipitation and provided the means 

to load packages onto trucks. The facility included a truck bay for 

loading and unloading. The facility was originally sized to 

accommodate the waste generated by both units over a 6-month period. 

Waste packages could be stored for up to 5 years.345 The facility was 

ready for use on December 16, 1982.346

• 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. Two of six planned concrete 

storage modules were built, each consisting of 60 cells. Each cell was 

constructed of galvanized corrugated steel cylinders with a welded steel 

base plate, and was sized to accommodate any combination of waste 
120

containers, such as resin liners, metal boxes, and drums. A concrete 

cover 3-feet-thick and weighing about 14 tons covered each cell. Both 

modules shared a sump compartment to collect drainage. The facility 

performed no active function other than storage.

347, 348 Storage 

Module “A” went into service in January 1980.349

• The waste handling and packing facility was used to process and 

package solid radioactive waste, such as contaminated clothing, tools, 

and equipment. Processing of contaminated material consisted of 

compaction, size reduction, and sometimes decontamination for reuse. 

No radioactive waste was stored in this facility.350 The facility was 

ready for use in January 1987.351

• The TMI-2 spent fuel pools “A” and “B” were used to stage spent 

resin vessels from the submerged demineralizer system and loaded 

defueling canisters. The spent fuel pools were filled with water for 

shielding. The submerged demineralizer system’s liners were prepared 

for shipment and lowered into the shipping cask underwater. The loaded 

shipping cask was lifted out of the “B” spent fuel pool, moved by crane 

over to the truck bay in the fuel handling building, and lowered onto the 

truck bed. Defueling canisters had to be loaded individually into the 

shipping cask located on a special stand in the truck bay. The weight of 

Solid waste staging facility was constructed to temporarily store radioactive 

wastes, such as dewatered resins, prior to shipment. Two module structures 

were built (center). The crane is loading a spent liner in a shipping cask. The 

temporary interim liner staging modules shown along the bottom.
121

the shipping cask prohibited loading it inside the “A” spent fuel pool 

where the defueling canisters were stored. Each defueling canister was 

lifted out of the pool, raised into the fuel transfer cask, moved to the 

truck bay, and lowered into one of seven canister slots in the shipping 

cask. The loaded shipping cask was then lowered horizontally and 

mounted onto a railroad car. Each fuel shipment took about 

1,000 person-hours to prepare.

352 

Packaging and Transportation. The NRC regulates packaging for the 

transport, storage, and disposal of nuclear materials and waste. In addition, 

the NRC regulates the design, fabrication, use, and maintenance of 

containers for high-level radioactive shipments, including spent nuclear 

fuel. The U.S. Department of Transportation (DOT) regulates packaging for 

the transport of lower-level radioactive materials and waste. In addition, 

DOT regulates shippers of all types of radioactive material and oversees 

vehicle safety, routing, shipping papers, emergency response, and shipper 

training requirements. The NRC requires its licensees to comply with 

DOT’s safety regulations in addition to the NRC’s own requirements. NRC 

regulations for the safety of transport and certification of packages for large 

quantities of radioactive materials, including spent nuclear fuel, can be 

found in 10 CFR Part 71, “Packaging and Transport of Radioactive 

Material.”353, 354 Also, NRC regulations under 10 CFR Part 61, “Licensing 

Requirements for Land Disposal of Radioactive Waste,” provide 

requirements for shipments of radioactive waste to commercial land 

disposal facilities. These regulations provide limits on radionuclide 

concentrations and requirements for the long-term structural stability of the 

disposal container.355 State regulatory agencies, responsible for land 

disposal facilities, and the DOE, responsible for accepting certain wastes 

from TMI-2, also provided additional requirements.

The shipment of radioactive waste from TMI-2 by the licensee was required 

to follow general regulatory requirements that applied to all licensees of 

nuclear power plants. The NRC certified new shipping casks that were 

designed and fabricated for the transportation of unique high-level waste 

that came out of TMI-2 during cleanup. The NRC issued a “certificate of 

compliance” for each new shipping cask or reissued an existing certificate 

of compliance for a modified cask. The certificate of compliance provides 

the terms and conditions under which the cask can be used, such as the type, 

form, and quantity of material authorized for shipment in the cask, as well 

as hardware specifications and internal packaging requirements. A user of a 

cask must be certified by the NRC as a registered user. To become a 

registered user, an applicant must have a quality assurance program in 

accordance with 10 CFR 71 and a copy of the certificate of compliance 
122

Model CNS 1-13C-II shipping cask on the 

truck transporter.

(from the holder/owner of the cask), and must maintain maintenance records 

on its casks. The NRC also reviewed and approved special waste containers 

that were required to be shipped inside a shipping cask that provided special 

safety features during transportation. In addition, the NRC reviewed

containers that were required to meet regulatory requirements (e.g., those of 

10 CFR 61 and the NRC Technical Position Paper on Waste Forms356) for 

disposal in commercial low-level radioactive waste burial facilities. The 

types of shipping casts used at TMI-2 are summarized below: 

• NRC Certificate of Compliance No. 9152/B, Revision 0, 

Model CNS 1-13C II Type B Shipping Cask was issued on 

April 4, 1982, and subsequently 

revised. This new cask was 

developed for the transport of 

spent SDS vessels with 

dewatered resins. The cask 

consisted of a cylinder made of 

lead and steel lining, a bolted 

lid, and two impact limiters 

(upper and lower) made of 

steel-lined rigid polyurethane 

foam.357 This cask was later 

modified and recertified for the 

transport of core debris samples 

to a DOE laboratory

(Revision 8, dated 

August 14, 1985). The 

modification included a 

secondary package to fill the 

void of the cask, a DOT 

Specification 2R container to 

provide secondary containment 

for the samples, a limit of total 

fissile material to a subcritical 

quantity, and a restriction to the 

transport of only TMI-2 fuel 

debris.358, 359

• NRC Certificate of Compliance No. 5026, Revision 10, 

Model CNS 14-190 Shipping Cask was reissued in January 1984. This 

existing cask was recertified to transport a high-integrity container that 

was used as an overpack (or outer container) for an EPICOR II pre-filter 

liner. The cask was used by the DOE to ship EPICOR II liners from 
123

INEL to the commercial low-level radioactive waste burial facility near 

Richland, Washington. The shipments began in May 1984 and were 

completed in February 1985. 360, 361, 362

• NRC Certificate of Compliance No. 9200, Revision 0, 

Model No. 125-B Shipping Cask was issued on April 11, 1986, and 

subsequently revised. The new cask was designed specifically to 

transport the loaded defueling canisters. Given that DOE took 

possession (ownership) of the fuel debris in its agreement with GPU, 

DOE had authority under DOT regulations to self-certify its own 

radioactive material shipping packages. Further, DOT regulations 

required that DOE packages meet the requirements and standards of 

10 CFR 71. DOE chose the NRC to be the regulator for approval of the 

Model 125-B shipping cask. Its design provided two testable levels of 

leak-tight containment in accordance with NRC regulations. NRC 

regulations also required the cask to provide protection against any 

radiological release during normal conditions of transport and during 

hypothetical accident conditions. The cask safety analysis report was 

independently evaluated by the NRC in accordance with regulatory 

requirements of 10 CFR 71. The NRC issued the certificate of 

compliance for the rail cask to DOE on April 21, 1986. Three casks 

were fabricated for the exclusive shipment of TMI-2 fuel debris. Two 

casks were purchased and owned by DOE. The third cask was owned by 

the fabricator and leased to GPU. 363

Model 125-B rail shipping cask used to transport seven defueling 

canisters loaded with core debris. Impact limiters shown with flags.
124

The Model 125-B fuel cask consisted of five major components: the 

outer containment vessel, inner containment vessel, upper and lower 

canister impact limiters, canister shield plugs, and cask impact limiters. 

Each cask had its own transportation system of a skid and rail car. Gross 

shipping weight of the shipping cask was about 183,000 pounds. The 

primary and secondary containment lids weighed 3,000 and 

5,200 pounds respectively. The cask impact limiters weighed about 

11,700 pounds each and the sunshield weighed about 500 pounds.364 Up 

Exploded view of the Model No. 125-B Shipping Cask. Fuel debris was 

contained in the seven defueling canisters.
125

to 21 defueling canisters could be sent in a single rail shipment (seven 

canisters per cask, three casks per shipment). The fuel shipments began 

on July 20, 1986, and were completed on May 9, 1990. There were 

22 rail shipments for a total of 342 canisters of core debris transported 

to the INEL. The total number of canisters shipped included 286 fuel 

canisters (which contained partially intact fuel assemblies and large 

debris picked up from the reactor vessel), 12 knockout canisters 

(containing core debris vacuumed from the reactor vessel and reactor 

coolant system), and 62 filter canisters (containing fine debris that 

passed through the knockout canisters).365

• HIC overpacks to dispose of EPICOR II pre-filter liners. A special 

high-integrity container (HIC) was designed and fabricated for DOE to 

dispose of the original 45 highly loaded EPICOR II pre-filter liners at 

the commercial low-level radioactive waste burial facility near 

Richland, Washington. 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 the 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 pre-filter liners generated at TMI-2. Certificate of 

Compliance No. WN-HIC-01 was issued for this HIC by the State of 

Washington on March 23, 1984.366, 367

• HIC overpacks to dispose of SDS vessels. After an unsuccessful 

attempt to qualify the design of the SDS pressure vessel as a HIC, SDS 

vessels were shipped and buried inside a polyethylene HIC. These poly 

HICs were permitted exclusively at the commercial low-level

radioactive waste burial facility in Barnwell, South Carolina. A concrete 

liner was inserted between the SDS vessel and the poly HIC to provide 

the shielding necessary to ensure the authorized service life of the poly 

material.368

• Commercial metal HIC. A commercially available HIC was used at 

TMI-2 to dispose of radioactive waste that contained high 

concentrations of strontium-90 and transuranic material. (At high 

concentrations, the waste classification system of 10 CFR 61 required 

stabilization of the waste either by solidification or HIC.) The metallic 

HICs were made of a corrosion-resistant alloy material. In 1984, the 

State of Washington granted interim authorization for the first shipment 

to the commercial low-level radioactive waste burial facility near 

Richland before the NRC completed its review of the technical 
126

evaluation report for these HICs. In April 1988, NRC completed its 

review and concluded that the HIC met or exceeded all of the 

requirements of 10 CFR Part 61 and the recommendations of the NRC 

Technical Position Paper on Waste Forms. The metallic HICs were 

modified and qualified to function as demineralizer vessels in the 

defueling water cleanup system and later in the last EPICOR II 

configuration.369 

Hydrogen Generation in Waste Containers. Generation of flammable 

gases inside sealed radioactive waste containers was a potential hazard. Two 

reactions contribute to gas generation: (1) the reaction between metals and 

water, which oxidizes the metal and releases hydrogen gas; and (2) longterm exposure of water and organic materials to ionizing radiation 

(radiolysis). Because many TMI-2 wastes—particularly SDS zeolites and 

canisters of fuel debris—were loaded with high concentrations of 

radioactivity, the latter reaction was a potential source of flammable gaseous 

mixtures (oxygen plus hydrogen).370 

• EPICOR II liners. A unique gas sampler and vent tool was designed by 

DOE to remotely remove the vent plug from the EPICOR II ion 

exchange resin liner, sample the gas content, vent the container, purge 

Demonstration high integrity container (HIC) being covered in the trench at 

the commercial radioactive waste disposal site near Richland, Washington. 

The HIC contained the highly loaded EPICOR II prefilter No. 18. The 

demonstration set the precedent for disposing of the remaining EPICOR II 

prefilters that were temporarily stored at INEL.
127

the container with inert gas, and reinstall the vent plug. The venting of 

the liners was performed at the Solid Waste Storage Faculty in the 

individual cells where EPICOR liners were stored. All operations were 

done remotely under a portable concrete shielding blockhouse over the 

storage cell. A portable remote support trailer that was located on top of 

the storage modules contained the gas circulator for purging the 

atmosphere of a storage cell with inert gas, a gas chromatograph for online sampling, controls for the gas sampler including TV monitors.371, 372

Laboratory analysis of several highly loaded EPICOR II pre-filter liners 

identified the average hydrogen production rate of a 2000 Curie liner 

would generate a 4 percent mixture of hydrogen in a nitrogen and 

carbon dioxide atmosphere in about 100 days. The vent tool lowered the 

content that would be generated during shipment to less than the 

flammable limit of 4 percent.

373 

• SDS vessels. Radiolytic gas generation in highly loaded SDS vessels 

was calculated to be significant. Measurements of gas buildup 

confirmed that this could be a problem during storage and 

transportation. A DOE research program developed a number of 

techniques to deal with the problem. To reduce buildup of flammable 

gas during transport and storage, the vessels were drained and vacuum 

pumped to remove free water. A catalyst was added to each vessel to 

recombine the hydrogen and oxygen as it was generated. A pressure 

relief system, consisting of a burst diaphragm and micropore graphite 

filter, was also added to each vessel to prevent uncontrolled, long-term 

buildup of non-recombinable gas mixtures. (A net hydrogen buildup can 

occur due to oxygen scavenging by various chemical reactions, such as 

the formation of carbon monoxide and carbon dioxide from oxidation of 

organic materials trapped within the zeolites.)374, 375

• Defueling canisters. Catalyst beds were used in the TMI-2 core debris 

canisters to recombine radiolytic hydrogen and oxygen and prevent 

buildup of combustible mixtures of gases. Each canister contained a 

recombiner catalyst package incorporated into the upper and lower 

heads. The catalyst bed was designed with porous metal filters at each 

end to contain the panicles while allowing gases to flow to and through 

the catalyst.

376, 377, 378, 379

Document Collections. Documents relating to waste management activities 

are grouped into the following document collections on the DVDs: 

• EPICOR II system (see the DVD folder, Water Processing: EPICOR) 
128

• submerged demineralizer system (see the DVD folder, Water 

Processing: SDS) 

• solid waste processing and storage (see the DVD folder, Solid Waste) 

• other waste-related processing systems and activities (see the DVD 

folder, General Waste Management) 

• status and lessons learned reports (see the DVD folders under the Status 

and Summary Reports collection) 

• incidents and deficiencies (see the DVD folder, NRC Preliminary 

Notifications) 

The following types of documents are included in the above document 

collections: 

• system and facility description reports 

• technical evaluation reports and safety evaluation reports

• certificates of compliance for shipping casks (current at the time of first 

shipments) 

• research reports, NRC technical NUREG reports, GEND reports, and 

technical reports from DOE national laboratories

• other correspondence between the NRC and the licensee relating to the 

waste management systems and activities, such as notifications, 

requests, reviews, and approvals

Defueling tool: three point gripper hydraulic attachment.
129

Spent EPICOR II resin liner inside the shielded transfer cask (yellow) being 

lowered into a shipping cask.

A highly radioactive spent SDS resin liner (center) being transferred into a 

shipping cask (lower center) while underwater in the spent fuel pool. 
130

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 (see GEND-034).

(Abbreviations: Gallons per minute---gpm, high pressure---HP, low pressure---LP,

pounds per inch---psi, temperature---temp)
131

7 Decontamination

The 1979 accident involved a loss of reactor coolant and resulted in severe

damage to the reactor fuel. When the reactor coolant pump flow was 

restored, radioactive contamination in the form of fuel debris and fission 

products was distributed by the coolant throughout the reactor coolant 

system. Reactor coolant, carrying fuel debris and fission products as 

dissolved and particulate material, flowed out of the reactor coolant system 

through the stuck-open pressurizer pilot-operated relief valve and into the 

reactor building’s basement and sump. During the first hour of the accident, 

spilled reactor coolant in the reactor building sump was automatically 

pumped into the auxiliary building holding tanks, which then overflowed 

along with the sumps. Although this radioactive water did not initially 

contain damaged fuel, later, equipment leakage from the makeup and 

purification system did contain fuel debris and fission products which mixed 

with the water in tanks, sumps, and floor drains. Exposed surfaces in the 

reactor building and the auxiliary and fuel handling building were 

contaminated with radionuclides in the reactor coolant, as well as 

radionuclides that became airborne, as hot reactor coolant flashed into 

steam. Airborne contamination entered the ventilation systems and spread 

throughout the auxiliary and fuel handling building.380 For some unknown 

reason, the ventilation systems stopped several times, which diverted 

airborne radionuclides to unintended areas.381 After the accident, the water 

in the reactor building’s basement was heated by residual heat from the 

reactor vessel, evaporated, condensed on the cooler walls, and drained down 

onto the floors and back into the basement.382

Cleanup of the auxiliary and fuel handling building started shortly after the 

accident. Airborne releases contaminated surfaces on the upper-level floor

(at the 328-foot elevation) and mid-level floor (at the 305-foot elevation) of 

the auxiliary building, and liquid releases to the drain system contaminated 

surfaces on the basement-level floor (at the 280-foot elevation). The interior 

of the building, including 26 piping systems, was contaminated by 

radioactive material, though less severely than the interior of the reactor 

building; because most of the interior surfaces, such as walls and floors, 

were constructed of uncoated concrete, radioactive materials penetrated the 

surfaces to varying depths. Approximately 510,000 square feet of surface in 

the auxiliary and fuel handling building required decontamination when 

cleanup operations began.383, 384

The decontamination experience at TMI-2 differed from past experience in 

the nuclear industry in that cleanup of the reactor building did not begin 

immediately. During the time between the accident and the start of the 
132

cleanup, the humidity in the reactor building was 100 percent, causing 

precipitation in the form of rain (steadily dripping condensation) inside the 

reactor building. One result of the rain was that dose rates during the initial 

personnel entries into the reactor building 14 months after the accident were 

lower than expected because radionuclides had been rinsed downward. 

A second result was that radionuclides permeated into porous surfaces such 

as uncoated concrete, were incorporated in corrosion layers as iron surfaces 

rusted, and were trapped in paint layers. Re-cleaning of previously cleaned 

areas was still required during the cleanup period, with concomitant 

exposure of workers.385

Reactor coolant containing core debris and fission products was discharged

on the reactor building’s basement floor through the stuck-open pressurizer 

pilot-operated relief valve when the block valve was periodically opened to 

control pressure. Consequently, these areas were covered with radioactive 

sediments, consisting primarily of river water sediment from a leaking air 

cooler inside the reactor building, concrete dust, and dirt.

386

Decontamination Objectives. The overall objectives of the TMI-2 

decontamination efforts were to maintain access to and operability of plant 

systems, to support defueling preparations and operations, and to permit the 

transition of the facility to a long-term storage condition. Shorter-term 

decontamination objectives focused on the removal or stabilization of 

Decontamination using high-pressure spray.
133

contamination in order to reduce occupational exposure and to prevent 

release of contamination to the environment. Longer-term decontamination 

objectives ensured that any remaining contamination was stable and 

sufficiently isolated for long-term storage.

387 

The initial decontamination objectives in the auxiliary and fuel handling 

building were to permit access without restriction because of surface or 

airborne contamination, to reduce radiation exposure from gamma sources 

to ALARA levels, and to prevent recontamination from other cleanup 

activities or system leaks. Decontamination of cubicles was required to 

permit access for inspection and maintenance of plant equipment, which had 

been deferred since the accident. In addition, decontamination of systems 

and equipment in the auxiliary and fuel handling building were required for 

activities to decontaminate the reactor building and defuel the reactor vessel 

and primary coolant system.388 The initial decontamination objective in the 

reactor building working areas (at the 305-foot and 347-foot elevations) was 

to reduce radiological conditions (general area radiation, airborne gaseous 

and particulate activities, and surface contamination levels) to ALARA 

levels and to maintain those conditions in a way that would permit defueling 

operations.389 The objective in the basement was to remove sludge from all 

accessible areas and stabilize contamination during long-term storage.

390 

The final decontamination objective was to stabilize localized radiological 

conditions in the plant, regardless of whether or not access was required for 

cleanup activities. The associated decontamination efforts were focused on 

meeting the specified goals (such as general-area dose rate, surface 

contamination level, and hot-spot dose rate) required to place the facility in

long-term storage after defueling completion. Long-term storage would 

allow decay of the radionuclides that remained in the facility so that workers 

would be exposed to lower levels of radiation during future decontamination 

and decommissioning.391, 392

Decontamination Criteria and Goals. The safety analysis report for 

post-defueling monitored storage (PDMS) provided a set of 

decontamination objectives (see above), baseline radiological criteria, and 

decontamination end-point goals for the decontamination program. Two of 

many prerequisites that were required for entering PDMS included 

(1) contamination reduction consistent with ALARA principles to meet 

established contamination-level goals in each of the areas of the auxiliary 

and fuel handling building and (2) radiation reductions consistent with 

ALARA principles, as necessary, to levels which would allow necessary 

plant monitoring, maintenance, and inspections.393 
134

As discussed previously, decontamination objectives were assigned 

according to general area within buildings, and were based on initial 

contamination levels, the need for personnel access to perform specific

activities inside the area, and the possibility of release of radioactivity to the 

environment. Baseline radiological criteria were established for each 

decontamination objective while considering the anticipated need for 

personnel access during cleanup and post-cleanup activities. Criteria were 

based on access frequency (hourly, daily, weekly, monthly, or quarterly) 

and radiological conditions (general-area dose rate, maximum hot-spot dose 

rate, and smearable surface contamination level). Criteria were intended as 

guidelines rather than absolute requirements.

394

Specific decontamination goals were set for each decontamination objective

(specific area or cubicle) based on baseline radiological criteria. For 

example, the specific decontamination goals for the 347-foot elevation 

operating floor in the reactor building that required infrequent (quarterly) 

access, was less than 30 mR/hr for general-area dose rate, and less than 

50,000 dpm/100 cm2 for surface contamination within 7 feet of the floor. 

The specific goals for corridors and access ways in the auxiliary building 

Strippable coating used in decontamination of floor surfaces.
135

that required routine access (40 hours per week) were less than 2.5 mR/hr 

for the general-area dose rate, 10 mR/hr for the maximum hot-spot dose 

rate, and less than 1,000 dpm/100 cm2 for surface contamination within 7 

feet of the floor. The specific goals for the seal injection filter room in the 

auxiliary building that may require weekly access, were less than 500 mR/hr 

for the general-area dose rate, less than 1,000 mR/hr for maximum hot-spot 

dose rate, and less than 50,000 dpm/100 cm2

 for surface contamination. 

General decontamination criteria for piping systems, equipment, and 

components were based solely on their contribution to area dose rates. 

Specific decontamination goals are listed in the PDMS safety analysis 

report (refer to DVD folder, After Defueling).395

Cleanup Activities. Decontamination of building surfaces, systems, and 

equipment included multiple activities across the following categories:396 

• Loose equipment. Removal of miscellaneous equipment and debris that 

were in the facility at the time of the accident, such as ladders, 

scaffolding, tools, and portable equipment. 

• Installed equipment. Decontamination or removal of installed 

equipment, such as piping systems, air conditioning and exhaust 

equipment, cable trays, and electrical and lighting equipment. 

• Interior surfaces. Decontamination of interior building surfaces 

consisting of metal and concrete materials.

• Sludge. Removal of contaminated sediment (sludge) from tanks and 

sumps in the auxiliary and fuel handling building and from the reactor 

building’s basement floor and sump. The sediment was transferred to 

the spent resin storage tanks for processing. Concentrated solids were 

transferred to a disposal container for cement solidification.397

• Resins. Removal of highly contaminated resins from the makeup and 

purification system demineralizers located in the auxiliary building. 

This system cleaned up the reactor coolant system during normal plant 

operations. During the accident and thereafter, about six kilograms of 

fuel and fission products were deposited in system filters and 

demineralizers.398

• Recovery and cleanup equipment. Decontamination of systems and 

equipment used for cleanup and defueling activities. Gross 

decontamination of refueling tools and cleanup equipment for reuse or 

disposal was performed in two temporary equipment decontamination 
136

facilities located in the auxiliary building and reactor building. In 1987, 

the waste handling and packaging facility became operational, which 

provided a permanent facility for integrated equipment 

decontamination. This new facility also processed and packaged 

discarded equipment for disposal.399 Following defueling operations, 

equipment located in the TMI-2 spent fuel pools was removed and pools 

were emptied of water and decontaminated. Spent fuel pool and the fuel

transfer canal contained fuel debris that was carried out on the surface 

of the defueling canisters.

• Support activities. Various support activities to ensure worker safety 

and to measure the effectiveness of the cleanup.

The decontamination of the reactor building’s atmosphere and 

accident-generated waste water are discussed in previous sections. 

Workers decontaminating the auxiliary building using the manually applied 

scrubbing method.
137

Decontamination Methods. Combinations of well-known methods were

used in the decontamination of building and equipment surfaces in the 

auxiliary and fuel handling building. Methods used for decontamination of 

surfaces inside the reactor building were based on the results of the gross 

decontamination experiment and subsequent experience gained in 

decontamination of the auxiliary and fuel handling building. The following 

decontamination methods have been reviewed by the NRC and saw at least 

limited use:

400, 401, 402 

• Abrasive blasting of steel surfaces with particulate driven at high 

velocity to remove contamination. This method was especially suited 

for small or irregular surfaces that were not compatible with other 

decontamination techniques. 

• Chemical decontamination of external surfaces of pipes, tanks, and 

system internals. The gross use of chemical agents was limited because 

of the potential for drain-off of such agents to damage water processing 

systems.

• Dry vacuuming to remove powdered contaminants and dried residue. 

This method was used especially for water-sensitive components that 

could not be flushed with water.

• Low-pressure water flush at levels between 100 to 1,000 pounds per 

square inch (psi), flow rates up to 25 gallons per minute (gpm), and 

water temperatures up to 170 degrees Fahrenheit. This method was used 

to clean equipment and loose surface debris. Examples included the 

polar crane, steam generator housing structures (“D-rings”), missile 

shields, refueling canal, and refueling bridge. 

• High-pressure water flush to remove unbonded surface coatings at 

pressures between 2,000 to 10,000 psi and flow rates of 4 to 30 gpm. 

• Ultrahigh-pressure water flush to remove rust, scale, nuclear-grade 

coatings, and surface concrete at pressures up to 60,000 psi and flow 

rates of 1 to 2 gpm.

• “Scabbling” of walls and floors to aggressively remove concrete 

surfaces and surface coatings. The scabblers used pneumatically 

operated reciprocating pistons equipped with tungsten carbide bits to 

pulverize the concrete surface. A vacuum system with a high-efficiency 

particulate air filter was attached. A major scabbling campaign began in 

the reactor building in October 1984.403
138

• A steam and vacuum decontamination system was used throughout the 

TMI-2 cleanup to decontaminate painted and uncoated concrete, 

ductwork, diamond deck plates, lead bricks, penetration covers, piping, 

conduit cable trays, and drain covers. The system used a single-head 

machine that minimized the spread of contamination and impairment of 

vision from sprayback.404 The system was first tested in the auxiliary 

building on June 14, 1984, and decontamination in the reactor building 

began in March 1985.405

• Strippable coatings that involved the application of an organic coating 

which contained chemicals to aid in the removal of radioactive 

contaminants from the surface. As the coating dried, it cracked and 

peeled away from the surface.

• Scrubbing to remove loosely held contamination on floors and walls 

using manually applied or mechanically driven rags, absorbent cloths, 

brushes, pads, grit, and chemical agents. 

• Wet vacuum to remove puddles of contaminated cleaning fluids after 

flushing. 

Remote Robotic Equipment. Remote-controlled robotic vehicles and

supporting control equipment were used extensively to perform work in the 

reactor building’s basement, the makeup demineralizer room in the auxiliary 

building, the reactor coolant pump seal injection valve room in the fuel 

handling building, and the reactor vessel. These vehicles were both versatile 

Remote control robotic high-pressure spray washing system.
139

and productive, and proved useful in many different tasks, including video 

camera inspections, radiation monitoring, sediment sampling, acquisition of 

concrete core samples, high pressure water flushing, concrete scabbling and 

scarification, and debris pickup and removal. In addition to the remote 

vehicles, fixed-position, remotely operated tools were developed for work 

inside the reactor vessel. The tools included a plasma arc cutting system to 

remove the stainless steel core support assembly, and several manipulator 

arms for handling damaged fuel and structural components.406 The use of 

robots at TMI-2 did not require any NRC licensing actions; however, 

activities in which robots were used, like most recovery and cleanup 

activities, required safety evaluations by the licensee and the NRC. Key 

robots used at TMI-2 are summarized below: 

• ROVER or remote reconnaissance vehicle (RRV) was used in the 

reactor 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. The RRV was a tether-controlled, six-wheeled work 

platform with multiple attachments or modules to perform the various 

tasks. ROVER was operated by two operators in a control room located 

outside the reactor building.407, 408

• LOUIE I remote vehicle was used to measure the radiation profiles of 

the two makeup demineralizer vessels, and to remove loose pre-accident 

debris and salt deposits on the floor inside the seal injection valve room. 

On loan from DOE, LOUIE I was a tether-controlled, tracked work 

platform with a telescoping boom-mounted manipulator that had been in 

use at DOE’s Hanford Site since the 1950s. The control console was 

setup in the accessible hallway near the entrance to the room.409 

• LOUIE II remote vehicle was used to perform remote floor scabbling 

in the seal injection valve room. An attached 3-piston pneumatic 

scabbler was used to pulverize the floor surface while vacuuming loose 

concrete. LOUIE II was a tether-controlled, six-wheeled work platform

with a heavy frame to withstand the stresses of the scabbler. The control 

console was setup in the accessible hallway near the entrance to the 

room. 410, 411

• WORKHORSE or remote work vehicle (RWV) was a large, heavy-duty 

robot built for decontamination and demolition work in the basement of 

the reactor building. WORKHORSE was a tether-controlled, sixwheeled work platform that was 10 times heavier than the RRV and had 
140

a boom with a seven-meter vertical reach. A two-level control building 

was built in the turbine building for three operators and support staff to 

operate the RWV and manage the work activities. The RWV was 

successfully tested in mockups, but never used due to changes in 

cleanup direction.412

• Mini-Rover was a commercial submarine vehicle modified to remove 

larger fuel debris inside the pressurizer. The Mini-Rover had a color 

camera, pincer arm to break apart debris, and a scope. The submarine 

was fitted with a ballast tank to improve mobility.413

• Remote manipulator performed defueling operations in areas of the 

reactor vessel that were not directly below the defueling work 

platform’s working slots. The manipulator could handle defueling tools 

and the in-vessel video viewing system, and pickup fuel debris. The 

manipulator was mounted to the manual tool positioner on the defueling 

work platform and had a 4-foot reach and six degrees of freedom of 

motion. The manipulator could be controlled from inside or outside the 

reactor building.414

• ACES or automated cutting equipment system was a remote-controlled 

plasma arc torch installed in the reactor vessel to cut the multi-layered 

lower core support assembly following bulk defueling. ACES consisted 

of a support bridge with a carriage and trolley to provide horizontal 

X-Y plane movement. A manipulator arm provided motion in the 

vertical Z-axis direction, including rotation and bending motions. At the 

lower end of the manipulator arm was a pneumatically operated gripper 

with a plasma torch and effector for cutting. A high-velocity stream of 

high-temperature ionized nitrogen gas plasma transferred an electric arc 

to melt the cut area.415, 416 The two control consoles (one for the 

manipulator subsystem and the other for plasma subsystem) were 

located in the turbine building.417

• MANFRED or manipulators for reactor defueling was a robot 

manipulator system designed and built for underwater disassembly and 

defueling of reactor vessel components. MANFRED consisted of a 

manipulator with various work attachments and grabber manipulators. 

The system was tested but never used due to the success of ACES 

(above) and manual handling of component pieces and loose fuel 

debris.418

Additional information on the uses of remote equipment at TMI-2 are 

provided in the following documents: International Atomic Energy Agency 
141

report “Catalogue of Methods, Tools and Techniques for Recovery from 

Fuel Damage Events” (IAEA-TECDOC-627); the DOE report “TMI 2: 

Lessons Learned by the U.S. Department of Energy, A Programmatic 

Perspective” (DOE-ID-10276); the American Nuclear Society’s Nuclear 

Technology journal (Volume 87; not included in the DVDs); and the EPRI 

report “Final TMI-2 Technology Transfer Progress Report” 

(EPRI-TR-100643; not included on the DVDs). 

Key Actions. By the end of the cleanup program, the floor contamination 

levels in most areas of the auxiliary and fuel handling building were reduced 

to those typical of pre-accident conditions. In the reactor building, radiation 

levels in frequently accessed areas were reduced by 85 percent by scabbling 

accessible areas and shielding finite radiation sources. A selection of other 

key actions during the cleanup period is summarized below: 

• First decontamination workshop. On November 27–29, 1979, the 

Facility Decontamination Technology Workshop, sponsored by DOE, 

was held in Hershey, Pennsylvania. The workshop provided those 

involved with the cleanup at TMI-2 a summary of experience regarding

events and incidents at other facilities that necessitated decontamination 

and dose-reduction activities (see GEND-002).419 

• Gross decontamination experiment. In March 1982, the gross 

decontamination experiment was conducted on various levels and 

surfaces inside the reactor building. The objectives of this experiment 

were to evaluate several decontamination techniques and to 

decontaminate the reactor building surfaces. Results of the experiment 

showed that gross decontamination could achieve the goal of reducing 

smearable contamination (see GEND-034).420 In addition, experiments 

showed that surface contamination contributed much less to 

general-area exposure rates, whereas rusty metal surfaces, large 

equipment, and bags of garbage contributed far more. Findings also 

indicated that recontamination was a problem; high-pressure flushing, 

mechanical scrubbing, and strip coating provided higher 

decontamination factors; residual water from flushing must be removed 

(such as by squeegeeing or vacuuming) to prevent recontamination from 

suspended contaminants; and complex surfaces and equipment were

harder to clean.

421

• Gross reactor building decontamination. On September 17, 1982, 

gross decontamination began in the reactor building at the 305-foot 

entry-level elevation, and at the operating floor on the next level above. 

The effort was designed to reduce smearable levels of contamination to 
142

the point that workers would be able to remove much of their bulky 

protective clothing and their respirators.422 

• Worker protection restrictions lifted in areas of the auxiliary building. 

In November 1982, upper-level corridors of the auxiliary building 

became accessible to workers without the need to wear 

anti-contamination clothing. (Respirators had not been required for 

entry into the upper corridors since October 1979.) The following 

month, the auxiliary building’s corridors in the basement (at the 

281-foot elevation) became accessible without the need to wear 

respiratory protection masks.423 Overhead areas, such as ceiling and 

cable trays, were decontaminated only to the extent that they would not 

re-contaminate the floor below. In such cases, a radiation work permit 

was required to access ceiling areas.424

• Respiratory protection restrictions lifted in areas of the reactor 

building. On June 28, 1984, workers entered the reactor building 

without respiratory protection for the first time since the accident, and 

subsequent entries were made without respirators, in accordance with 

ALARA principles.425, 426

The command center for the gross decontamination experiment in March 

1982 contained the control functions for entry into the reactor building with 

positions for safety, radiological engineering, operations, entry coordinator, 

and command center management.
143

• GPU’s Decontamination Task Force Report. On December 18, 1985, 

the licensee issued the Decontamination Task Force Report. The report 

provided a review of the effort required to decontaminate TMI-2 and an 

evaluation of the reduction in occupational exposure during 

post-defueling monitored storage (PDMS). The objective of the task 

force was to arrive at a consensus technical approach to each of the 

major areas of decontamination work: remote-equipment development; 

sludge transfer and disposal; D-ring dose reduction and 

decontamination; recovery of the reactor building’s basement; auxiliary 

and fuel handling building surface decontamination; non-reactor coolant

systems decontamination; reactor building ventilation modifications; 

reactor building surface decontamination; reactor coolant system 

decontamination, and waste management for the decontamination of the 

reactor building. 

The task force estimated that final decontamination, as part of future 

decommissioning, would result in a total occupational exposure in the 

range of 2,710 to 5,770 person-rem, assuming that further 

decontamination was deferred until after a 30-year period of PDMS. 

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 

advances in both remote cleanup 

technology, and chemical 

decontamination methods.427 Results 

from the task force report were 

documented in Appendix 5A of the 

PDMS safety analysis report, 

“Potential Reductions in 

Occupational Exposure Due to PostDefueling Monitored Storage,” (see 

the DVD folder, After Defueling; the 

task force report itself, however, is 

not available on the DVD). 

• Reactor building sludge removal. On March 31, 1987, the robotic 

removal of sediment in the reactor building’s basement began and was 

A radiation work permit (RWP) was 

required to access contaminated 

ceiling areas above the clean areas of 

the auxiliary building main corridors.
144

completed in July 1987.428 Approximately 10,800 pounds 

(4,900 kilograms) of wet sludge, which contained about 4 kilograms of 

fuel debris, was removed from the reactor building’s basement floor, 

pumped into a tank located in the auxiliary building, and solidified for 

burial at a low-level radioactive waste disposal site. A robotic 

desludging system was used to vacuum about 40 percent of the 

basement floor area which was the only area accessible to the robot. The 

removal efficiency of desludging was greater than 90 percent.429 The 

robotic flushing of the basement floor was completed 3 months later.430

• NRC’s PEIS supplement for completing cleanup. In August 1989, the 

NRC issued the final supplement to the Programmatic Environmental 

Impact Statement (PEIS), NUREG-0683, Supplement 3, which

evaluated the potential environmental impacts of alternative approaches 

to completing the TMI-2 cleanup. The supplement examined the 

estimated occupational radiation doses associated with the licensee’s 

proposal for delayed decommissioning and five NRC-identified 

alternatives that were evaluated quantitatively. The dose estimates were 

based on a task analysis of the cleanup work to be performed. (See 

Supplement 3 of the PEIS in the DVD folder, Guidance-PEIS.) 

Highly radioactive sludge on the basement floor of the reactor building as 

viewed from an upper level. A robotic desludging system was used to vacuum 

the 40 percent of the basement floor area that was accessible to the robot.
145

• Plant-entered PDMS. On December 28, 1993, the NRC granted an 

amendment to permit the licensee to place the plant in a long-term 

PDMS; the amendment also provided the PDMS technical 

specifications. The NRC safety evaluation concluded that the routine 

release of any significant quantity of radioactive material during PDMS 

had been minimized, in part, by the decontamination of large sections of 

the reactor building, auxiliary and fuel handling building surfaces, 

equipment, and piping. In addition, the NRC concluded that radiation 

levels within the facility had been reduced to such an extent that plant 

monitoring, maintenance, and inspection could be performed.431 (See 

section on After Defueling.) 

Document Collections. Documents relating to decontamination activities 

are provided in the following DVD folders:

• documents relating to technical evaluation reports of decontamination 

activities; decontamination schedules; NRC safety evaluation reports; 

NRC technical NUREG reports; GEND reports and technical reports 

documenting research results from DOE national laboratories; and other 

correspondence between the NRC and the licensee, such as 

notifications, requests, reviews, and approvals (see the DVD folder,

Decontamination) 

• DOE summary reports, such as annual and lessons-learned reports, and 

periodical publications (see the DVD folder, DOE/National Laboratory 

Status Reports) 

• NRC TMI Program Office (TMIPO) weekly status reports (see the 

DVD folder, TMI Program Office Weekly Status Reports) 

• post-defueling monitored storage safety analysis reports and documents 

relating to decontamination requirements (see the DVD folder, After 

Defueling) 

Project schedules for decontamination activities are provided in the DVD 

folder General Management and Oversight. 

Defueling tool: hook attachment.
146

Technicians on the defueling work platform.
147

8 Defueling

The TMI-2 accident resulted in severe reactor core damage and migration of 

molten core materials onto the reactor vessel’s lower head. Core damage 

and relocation occurred within four hours of the accident initiation, after 

which long-term cooling stabilized the damaged reactor core. Various 

reactor core inspections (see the section on Data Acquisition and Analysis) 

and observations from early defueling activities provided information to 

estimate the end-state core configuration. The INEL report “TMI-2 

Accident Scenario Update” (EGG-TMI-7489) identified four regions within 

the original core volume, as shown in its iconic figure of the damaged 

reactor core (see Section 5): upper cavity void region, debris bed region, 

previously molten region, and partially standing fuel assemblies (or “stubs”) 

region. Molten reactor core broke through the hard crust of the melt region

at 224 to 230 minutes into the accident, penetrated some of the baffle plates, 

flowed through the complex five-layered lower core support assembly, and 

settled onto the reactor vessel’s lower head.

432 The results of inspections and 

early defueling that were cited in the INEL report (1986) and the earlier 

GEND-007 report (1981) provided the technical basis for planning 

defueling approaches and necessary equipment. During the period of 

exploration of the damaged reactor core, information from new inspections 

and observations altered defueling strategies and tool designs.433

The TMI-2 cleanup effort took ten years with a collective manpower effort 

of over 3.6 million person-hours to complete. The reactor vessel defueling 

operations spanned a five-year period from October 1985 through 

January 1990 and involved over two million person-hours. A total of about 

133,000 kilograms of fuel, cladding, structural, and control materials were 

removed from the reactor vessel during the five-year effort. During July and 

August of 1991, the reactor vessel was drained to make final measurements 

of the residual fuel remaining in the vessel. An 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. The total occupational dose resulting from all cleanup 

activities was less than 6,500 person-rem over the first 10-year period. The 

cumulative occupational dose for defueling and defueling support activities 

was below 2,000 person-rem.

434 The exposure rate to defueling workers 

averaged less than 10 millirem per hour.435

Reactor Core Damage. The final defueling report, which was submitted by 

the licensee in 1990 for NRC review, provided the complete picture of the 

end-state configuration of the core. This report was based on actual
148

Reactor vessel with upper plenum removed and 

internals indexing fixture installed.

Lower core support assembly inside reactor vessel. This massive, five-layered 

structure was cut into pieces to allow access to the fuel debris in the lower 

head. Pieces are stored in the reactor building.
149

defueling experience during the five-year defueling period. The final report 

indicated that the original core inventory included about 94,000 kilograms 

of uranium oxide (fuel) and 35,000 kilograms of cladding, structure, and 

control (neutron-absorption) material. The total amount of core debris was 

estimated to be 133,000 kilograms, accounting for oxidation of core 

components and melt from part of the upper plenum grid plate. About 

50 percent of the original core melted. The defueling program had to 

consider how to remove the fuel debris to the extent that inadvertent 

criticality was precluded in each confined area.436 The distribution of core 

material as documented in the final defueling report is summarized below: 

• Upper core void. The upper core void or cavity consisted of only 

42 partially intact fuel assemblies standing at the periphery of the void. 

Only 2 fuel assemblies of the original 177 assemblies still had most of 

their fuel rods intact. The void was about 26 percent of the original core 

volume and measured 1.5 meters deep from the top of the original core 

to the debris bed.437

• Upper debris bed. The debris bed consisted of 26,000 kilograms of core 

material, such as whole and fractured fuel pellets, control rod spiders, 

fuel assembly end fittings, broken fuel rods, and resolidified debris. The 

bed rested on top of the resolidified, hardened mass and was about 

0.6 meter to 1 meter deep.438

• Resolidified mass. The solid metallic and ceramic mass consisted of 

about 33,000 kilograms of core material. The mass rested on partially 

intact fuel assembly “stubs.” The mass measured about 3 meters in 

diameter, 1.5 meters deep in the center, and 0.25 meter deep around the 

edges.439

• Intact assemblies. Partially intact fuel assembly stubs located under the 

resolidified mass and the peripheral standing assemblies comprised 

about 45,000 kilograms of core material. The standing stubs varied in 

length from about 0.2 meters in the center to 1.5 meters at the periphery. 

The stubs extended from the lower grid plate to the bottom of the 

resolidified mass.440 

• Upper core support assembly. The upper core support assembly 

included vertical baffle plates that formed the peripheral boundary of 

the core, horizontal core former plates to which the baffle plates were 

bolted, the core barrel to which the core formers were attached, and the 

thermal shield. The assembly retained about 4,000 kilograms of loose 

debris and resolidified material. Loose debris ranging from 1.5 meters to 
150

a few millimeters deep settled behind the baffle plate circumference. 

A resolidified crust ranging from 0.5 to 4 centimeters thick was attached 

to the bottom of three core former plates.441

• Lower core support assembly. The lower core support assembly 

included five layers (top to bottom): the lower grid rib section that 

supported fuel assemblies; lower grid flow distributor plate; lower grid 

forging (attach point for the complete lower core support assembly to 

the reactor vessel wall); incore guide support plate; and elliptical flow 

distributor head. The layers were supported by the outer circumferential 

shell. The assembly structures retained about 6,000 kilograms of 

resolidified material around the circumference of the structures.

442

• Lower head region. The reactor vessel’s lower head region contained 

about 12,000 kilograms of loose core debris and 7,000 kilograms of 

agglomerated core debris. The debris on the lower head was 4 meters in 

diameter and 0.75 to 1 meter deep. The surface debris had particle sizes 

which varied from those of large agglomerated debris (up to 0.20 meter) 

to those of granular particles. Resolidified material existed on the 

reactor vessel’s lower head, underneath the loose debris. This material 

was approximately 0.5 meters deep in the center and 1.7 meters in 

diameter.443

Reactor building polar crane being inspected for use in defueling 

preparations. 
151

• Fuel debris distribution outside the reactor vessel. A total of about 

228 kilograms of fuel debris were transported through the reactor 

coolant system. About 95 percent of the debris settled in the “B” loop 

where a reactor coolant pump was initially turned on at 174 minutes into 

the accident. Several minutes of pump operation quenched the hot 

oxidized reactor fuel and caused it to shatter. Other pumps were 

operated sequentially until the last pump was secured in response to an 

emergency procedure on April 27, 1979, when the final pressurizer level 

instrument failed. Most of the debris inside the reactor coolant system 

settled on the upper tube sheet of the “B” steam generator (about 

125 kilograms). Smaller amounts settled in the decay heat system 

suction piping or “drop line” (30 kilograms), reactor coolant pumps (a 

total of 20 kilograms), and pressurizer (12 kilograms). Approximately 

10 kilograms total were deposited in the reactor building on the 

basement floor and sump (5 kilograms) and makeup and purification 

system letdown coolers (4 kilograms). A total of 23 kilograms entered 

the auxiliary building, mainly deposited in the three reactor coolant 

bleed tanks (a total of 15 kilograms) and in the makeup and purification 

system (6 kilograms).444

Preparation of the Reactor Vessel. Before core debris could be removed 

from the reactor vessel, preparations were required to allow direct access to 

the damaged reactor core. While many of these activities were routine

during normal refueling operations, the effects of the severe accident on 

reactor vessel components, the reactor building’s environment, and 

occupational radiation exposures presented complex challenges. 

Preparations included consideration of numerous potential safety issues; for 

example, occupational exposures; decay heat removal; criticality control; 

boron dilution; radioactivity releases; hydrogen evolution inside the reactor 

coolant system; pyrophoricity (spontaneous ignition in air) of zirconium 

fines in the reactor vessel; heavy load drops; polar crane failure; reactor 

vessel draining; and fire protection. Technical considerations included 

potential distortion; warping or physical dislocation of the reactor vessel’s

head or upper plenum; reactor coolant cleanup; reactor coolant system 

depressurization; and lowering of reactor coolant level.445, 446, 447, 448 In 

addition, cleanup of the reactor building, which included atmospheric gases, 

basement water, and surface contamination, was an important prerequisite to 

ensure lower radiation exposures. Key milestones for reactor vessel 

preparation activities are summarized as follows: 

• Leadscrew uncoupling. The last of the lead screws attached to the 

61 control rod and 8 axial power shaping rod spider assemblies were 

uncoupled in November 1982. The 22-foot long lead screw was a 
152

component of each control rod drive mechanism. Verification was 

performed in December 1982 to ensure that no partial fuel or control rod 

assemblies were attached to the lead screw.449

• Polar crane refurbishment and testing. The reactor building’s polar 

crane was needed to remove the reactor vessel head and upper plenum. 

The polar crane was refurbished because of the harsh environmental 

conditions inside the reactor building during the accident (such as 

steam, burning hydrogen, and containment spray) and following the 

accident (such as high humidity and high radiation). The NRC safety 

review of the use of the polar crane was delayed several months to 

investigate allegations made by a contractor about the safety of the polar 

crane. Six months later, a NRC investigation cited deficiencies in the 

administrative and procedural aspects of the polar crane repair. The 

NRC concluded that the specific deficiencies cited did not result in a 

significant increase in risk to the public health and safety. On November 

18, 1983, the staff approved the licensee's safety evaluation for the 

refurbishment and use of the reactor building polar crane. Load-testing 

Reactor vessel head and service structure resting on its stand surrounded by 

shielding. Shown are white tubes filled with sand and yellow lead blankets.
153

of the polar crane was successfully completed on February 29, 1984, 

when a test assembly weighing 214 tons was lifted and moved along 

predetermined test paths.450 Various configurations and uses of the polar 

crane were approved by the NRC throughout the defueling and cleanup 

efforts.

• Canal seal plate installation. A modified canal seal plate was installed 

on April 13, 1984, to ensure a long-term positive seal between the 

reactor vessel and refueling canal. The seal allowed the contingency for 

flooding the refueling canal during head removal.451

• Reactor coolant system draindown. The reactor coolant system was 

partially drained and depressurized to allow for the reactor vessel head 

lift on June 24, 1984. The water level was then raised to cover the 

control rod guide tubes after head lift and installation of the internals 

indexing fixture (discussed below).452

• Removal of reactor vessel studs. Sixty reactor vessel studs were 

cleaned, detensioned, removed from the reactor vessel, and stored in 

racks in the reactor building. This activity was completed on 

July 6, 1984. The first of the two detensioning passes had been 

performed months earlier to identify and correct stuck studs.453

• Leadscrew parking. The last leadscrew to the control rod drive 

mechanism was “parked” on July 21, 1984. Parking refers to raising a 

leadscrew to its uppermost position using a heavy-duty lifting tool. The 

leadscrew was then secured in place with a parking tool so that it would

not extend below the head flange level and interfere with lateral 

movement of the reactor vessel head during the lifting operation.454

• Reactor vessel head lift. The reactor vessel head and the attached 

service structure were removed and placed in shielded storage on 

July 25, 1984. The reactor building’s polar crane was used to lift the 

head and place it on its storage stand located in the reactor building. The 

actual lifting operation took over five hours. Sand-filled columns were 

placed around the head stand for shielding. Lead blankets were placed 

around the service structure. Before the lift operation, training exercises 

were performed using a set of mockups.455

• Internals indexing fixture installation. The modified cylindrical 

internals indexing fixture (IIF) with a watertight gasket system was 

placed on the open reactor vessel flange by the polar crane on 

July 26, 1984. The IIF was normally used to guide the upper plenum 
154

and core support assembly during installation and removal. The IIF was 

modified to allow the installation of a shielded work platform above the 

reactor vessel to support future operations. Before the installation of the 

work platform, the IIF was filled with five feet of borated water to 

provide radiation shielding over the exposed upper plenum in the 

reactor vessel.456

• Upper plenum assembly transfer. The last major structural obstacle to 

defueling was removed on May 15, 1985, when the upper plenum 

assembly was lifted from its jacked position in the reactor vessel, raised 

through the water-filled internals indexing fixture, and transferred to its 

storage stand in the deep end of the fuel transfer canal. Before plenum 

transfer, a six-foot-high dam was constructed, allowing the deep end of 

the canal to be flooded to a level sufficient to provide adequate 

shielding for the stored plenum. Major preparatory activities included 

in-vessel inspection of the plenum, separation of damaged fuel 

assemblies hanging from the bottom of the plenum, and the initial lifting 

of the plenum by a system of four hydraulic jacks. The precautionary 

jacking ensured that there was no binding during plenum lift by the 

reactor building’s polar crane. Before the lift operation, training 

exercises were performed using a set of mockups.

457

Upper plenum being transferred from the reactor vessel (shown) to the 

partially flooded deep end of the fuel transfer canal (not shown).
155

• Pyrophoric issue. The issue of pyrophoricity was addressed in the 

Programmatic Environment Impact Statement (PEIS) for defueling 

activities and in the safety evaluation of the underhead characterization 

study. This issue was based on the concern that appreciable amounts of 

zirconium hydrides might have been formed during the accident when 

hot dry hydrogen reacted with zirconium surface. At high temperatures, 

zirconium hydrides react with steam to form zirconium oxide and 

hydrogen gas. In a finely divided form, zirconium hydrides can be 

pyrophoric out of water. The PEIS concluded that it was unlikely a 

zirconium hydride ignition would occur given that defueling operations 

were planned underwater. However, the underhead characterization 

study required lowering the reactor coolant level in the reactor vessel 

and uncovering some of the vessel internals, such as leadscrew support 

tubes, control rod guide assembly tubes, and upper plenum cover plate. 

Given that these internal surfaces and fuel debris samples would be 

exposed to air, the issue of pyrophoricity was addressed by the licensee 

and NRC in the characterization study’s safety evaluation.458 

Internals indexing fixture with its temporary cover mounted on 

the reactor vessel flange. This fixture would evenually be mated 

to the defueling work platform. 
156

NRC determined in its safety evaluation that (1) the presence of steam 

as an oxidizing agent and the temperature condition during the accident 

would not likely produce significant quantities of zirconium hydride in a 

pyrophoric condition; (2) the primary system flow dynamics during the 

accident would not likely transport large quantities of pyrophoric 

material, if formed, to the top of the plenum; and (3) any pyrophoric 

materials in finely divided form would be dispersed and mixed with 

inert materials of core debris which would prevent the development of 

pyrophoric conditions. The results of the underhead characterization 

study supported the conclusions in the NRC safety evaluation that there 

was little potential for a pyrophoric reaction with the plenum cover 

exposed to air. Results were based on visual observations of the reactor 

vessel underhead conditions and laboratory analyses of the chemical

and pyrophoric properties of samples obtained from components within 

the reactor vessel and from solids filtered from the reactor coolant.

459

Defueling Systems and Equipment. Unique systems and equipment were 

designed and installed 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 and placed into fuel 

containers (baskets) or specially engineered defueling canisters. Some 

long-handled tools had various hydraulically actuated fittings to tackle the 

larger pieces and smaller bits of debris. The core bore machine was placed 

Defueling operations from the rotating defueling work platform with a worker 

holding a long-handled defueling tool down through the slit into the reactor 

vessel. Two jib cranes at each end of the slit were used to lower larger tools.
157

back into service to bore holes in the resolidified mass to help break apart 

the previously molten reactor core. Combinations of tools were used to 

assist defueling the lower reactor vessel region, such as the core bore 

machine and plasma arc torch. Three types of defueling canisters and 

associated support equipment were specially designed to remove the fuel 

debris from the reactor vessel and package it for transportation. Water 

cleanup systems were installed to ensure water clarity in the reactor vessel 

for removing debris and in the “A” spent fuel pool for processing defueling 

canisters.

The defueling systems were designed before the extent of core damage and 

radiological conditions were fully understood. The design of the canisters 

was decided upon early in the cleanup and dictated the design of the 

defueling platform and the shipping cask. The design of the canisters was 

based on the removal of intact fuel assemblies. Ultimately, very few if any 

intact fuel assemblies were removed. The narrow inside dimensions of the 

cylindrical canister design necessitated substantial cutting of distorted fuel

debris in order to load through the opening of the canister. A radiation

analysis program460 was undertaken to identify and quantify possible 

radiation sources and to design defueling equipment to achieve dose-rate 

goals in the defueling area. Specific safety issues were addressed in NRC

evaluations of defueling equipment, systems, and operations. Safety issues 

that were considered in the evaluations were similar to those addressed in 

the preparation of the reactor vessel for defueling (as listed above). Lessons 

learned from experience using defueling equipment were documented in the 

Nuclear Technology journal paper “Fuel Removal Equipment for Three 

Mile Island Unit 2” (Vol. 87, No. 3; not available on the DVDs).461

Examples of systems and equipment used for defueling the reactor vessel 

are summarized below (see GEND-INF-073): 

• Defueling work platform. A shielded work platform and support 

structure was installed over the water-filled internals indexing fixture

(IIF) in August 1985. The support structure circumscribed the IIF and 

extended upward from the refueling canal floor to the work platform. 

The shielded work platform was located 9 feet above the reactor vessel 

flange. The work platform included a defueling work platform with a 

rotatable surface 17 feet in diameter with a 6 inch thick steel plate 

shield. 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. 

Various lines for water treatment and air ventilation to control 

off-gassing were routed into and out of the reactor vessel through the 
158

platform support structure. A decontamination spray system flushed 

radioactive debris from the surface of the canisters, long-handled tools, 

and other equipment as each item was removed from the reactor vessel. 

Two jib cranes were mounted on the defueling work platform to aid the 

operators in manipulating the long-handled tools in the long-handled 

tool slot.462

A full-scale mockup duplicated the rotating work platform, the tool 

racks, and the overhead crane required to handle the 40-foot-long tools. 

The defueling crew developed their sense of feel for the tools, the 

balance of the tools, and the effort required to pull or shear fuel while 

working on the mockup.463 Operator training and tool testing on the 

mockup contributed to effective defueling operations.464

• Off-gas system. The off-gas system created airflow through the 

defueling work platform, into the internals indexing fixture enclosure, 

and out to the reactor building’s atmosphere through a filtration unit. 

The airflow through the platform prevented radioactive gases that 

collected under the platform from reaching personnel working on the 

platform.465

• Defueling water cleanup system (DWCS). The DWCS was used to 

process water in the reactor vessel, in the deep end of the fuel transfer 

canal, and in the “A” spent fuel pool. The system was designed to 

Core bore drilling machine, mounted on a platform over the defueling work 

platform, was used to drill holes and break apart the resolidified mass.
159

reduce radioactivity and improve water clarity for defueling operations.

The DWCS was composed of two systems: the reactor vessel cleanup 

system and the fuel transfer canal/spent fuel pool cleanup system. The 

cleanup of the reactor vessel water proved to be more difficult than 

originally anticipated. (See further discussion in the prior section on 

Waste Management.) 

• Long-handled defueling tools. Numerous manual and hydraulically 

powered long-handled tools were used to perform a variety of functions, 

such as pulling, grappling, cutting, scooping, and breaking up the core 

debris. These “pick-and-place” tools were used to load debris into fuel 

canisters positioned underwater in the reactor vessel. The “working 

end” of the long-handled tool was one of many kinds of end effector 

attachments. Powered end effectors were hydraulically operated. The 

operators relied on the video viewing system to monitor their work as 

they manipulated the tools through the tool slot in the defueling work 

platform. Most tools were supported by an overhead service crane to 

provide vertical and lateral motion, although some lightweight tools 

could be handheld for picking smaller debris. End effectors could be 

detached from their handles and stored in a tool rack located under the 

defueling work platform, or in a rack outside the reactor vessel.466 A few 

of the numerous end effector attachments are summarized below. See 

GEND-INF-073 for a catalog of early defueling tools. 

o Hydraulic cutoff saw. An hydraulically driven hacksaw was used 

for sizing structural material or other debris (that is, reducing it to 

pieces of smaller size) to facilitate the loading of debris into fuel 

canisters or debris baskets.

467, 468

o Hydraulic impact chisel. The impact chisel was a hydraulically 

actuated miniature jackhammer-type tool for use in breaking apart 

hard materials. The chisel’s angle of attack could be varied remotely 

to achieve any position from vertical to horizontal. Different bit 

types were provided for the various anticipated chiseling operations.

469, 470

o Hydraulic shredder. The hydraulic shredder was used to reduce fuel 

pins (with and without fuel pellets) and spacer grids to sizes which 

would facilitate their placement into fuel canisters or debris baskets. 

The shredder was suspended below the defueling work platform 

using the support structure attached to the platform. The fuel debris 

was loaded into an inlet hopper and the shredder output was 

discharged to a transfer container which was then emptied into a 
160

fuel canister or debris basket; discharge from the shredder could 

also be directed to the debris bed. Non-shreddable material lodged 

in the shredder was retrieved through manual manipulations using 

other long-handled tools.

471, 472

o Remote manipulator. Refer to the discussion on Remote Robotic 

Equipment in the section on Decontamination. 

o Water-jet cutting system. The water-jet cutting system performed

remote underwater cutting of hard materials with and without the 

use of abrasives. The equipment was designed to be positioned at 

the cutting location with the remote manipulator.

473

• Core bore machine. The core bore machine, which was used to extract 

core stratification samples in 1985, was used during defueling to break 

up the solidified monolith (melted-together core materials) in the core 

region.

474, 475 The machine was used again to drill through and cut 

portions of the lower core support assembly.476, 477, 478 The drill was 

reinstalled on its elevated mount over the defueling work platform.

479

• Automatic cutting equipment system (ACES) or plasma arc torch. 

Refer to the discussion on Remote Robotic Equipment in the section on 

Decontamination. 

• 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 microns 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 long-term storage of core 

debris. Three types of defueling canisters were designed and fabricated: 

a fuel canister, knockout canister, and filter canister. Each canister 

required fixed neutron-absorber material for criticality control; catalytic 

recombiners to control the concentration of combustible gas mixtures 

generated from radiolytic decomposition of water; and appropriate 

process connections for fillings, closing, dewatering, inerting, and 

monitoring. All three canisters were 150 inches long, 14 inches in 

diameter, and 1⁄4 inch thick.480, 481 As a result of a request from DOE, the 

NRC performed inspections of defueling canisters during the fabrication 

process, which included inspectors observing welding, non-destructive 

examinations, and fitting of components.482

o The fuel canister was designed as a receptacle for large pieces of 

core material, which were picked up and placed either directly into 
161

the canisters, or into other containers which would then be inserted 

into the canister.483 One lesson learned was reported on the 

difficulty of placing distorted debris in the fuel canister due to the 

narrow inside dimensions of the canister.

484

o 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 process inlet line 

entered the top of the canister and bent to direct the flow 

tangentially along the inner circumference of the shell, creating a 

swirling action that caused the entrained debris to settle out in the 

canister vessel. The water then exited through an 850-micron screen 

to a process connection in the top of the canister. The knockout 

canister had a capacity of about 1,800 pounds of fuel debris.485

o The filter canister was designed for use in the fuel debris vacuum 

system, the defueling water cleanup system, and the canister 

dewatering system. The filter captured debris fines larger than 

0.5 microns on sintered metal filters.486

• Debris buckets. Debris buckets were used to configure the debris before 

insertion into the fuel canister, to maximize the packing density in a 

canister, and to eliminate many 10-foot vertical trips to load small 

pieces in a canister. Two types of disposal buckets, the top loading 

Three types of defueling canisters for containing large fuel debris (fuel), 

vacuuming fractured fuel pellets (knockout), and filtering fuel fines (filter).
162

debris bucket and side loading debris bucket, were designed to fit into 

the fuel canisters. A reusable debris bucket fitted to the inside of a fuel 

canister also had a remotely operated trap door on the bottom for 

unloading into the canister. Accessories included a funnel, handling 

tool, and stand.487 

• In-vessel vacuum system. The in-vessel vacuum system was designed 

to remove small loose debris as large as the approximate size of a fuel 

pellet maximally. The vacuum system was supported from the underside 

of the defueling work platform and was controlled from the console on 

the south auxiliary work platform. The vacuum pickup nozzle was 

connected to a canister by a flexible hose and was manipulated using a 

long-handled tool. Debris was picked up and passed first through a 

The canister handling bridge was used to stage 

defueling canisters in a submerged storage rack in 

the “A” spent fuel pool. Shown is the canister 

transfer shield and defueling canisters in the pool. 
163

knockout canister; any remaining debris larger than 0.5 microns was 

collected in a filter canister.488 Vacuum defueling began on 

December 31, 1985, in which about 300 pounds of debris was loaded 

into the knockout canister.489

• Airlift vacuum system. The airlift vacuum system was used to lift debris 

as large as 5 centimeters from the core region rubble bed and the lower 

head region of the reactor vessel. The system included an air 

compressor, airlift pipe, and a fuel canister. Compressed air was 

injected at the bottom or suction end of the pipe to draw water into the 

pipe along with entrained fuel debris. Debris was deposited into a fuel 

canister located at the top of the airlift piping. The airlift increased 

packing efficiency in fuel canisters already containing oddly shaped 

pieces of debris, such as partial fuel assemblies and end fittings.490

• Pressurizer defueling system. The pressurizer defueling system was 

designed to remove fine fuel debris from the pressurizer. The system 

included a submersible vacuum pump, a knockout canister, a filter 

canister, and an agitation nozzle. The defueling water cleanup system 

injected water through an agitation nozzle to suspend debris in the 

water-filled pressurizer. Water was pumped from the pressurizer, 

through the canisters, and into the reactor vessel. Larger debris was 

pickup with the use of a robotic submarine.491 

• Once-through steam generator (OTSG) defueling system. The OTSG 

defueling system was designed to vacuum loose fuel debris from the 

upper tube sheets. The dry vacuum system included a vacuum head on a 

handling pole, vacuum canister, high-efficiency particulate air filter, a 

vacuum breaker, and vacuum pump. Larger debris was manually 

removed using long-handled tongs.492

• Canister positioning system. The canister positioning system was a 

rotating carousel installed in the reactor vessel that could hold up to five 

fuel and knockout canisters, including knockout canisters for use with 

the vacuum system. The height of canisters in the canister positioning 

system could be adjusted to three discrete elevations to allow them to be 

placed more closely to the debris bed as the bed got lower.493 

• Canister handling bridges. Existing fuel handling bridges located in the 

reactor building and fuel handling building were modified with canister 

handling trolleys and canister transfer shields. The reactor building’s

canister handling bridge was used to lift loaded canisters into a canister 

transfer shield, and to move the shielded canister from the reactor vessel 
164

through the air to the flooded deep end of the fuel transfer canal. The 

fuel handling building’s canister handling bridge lifted the canister from 

the fuel transfer system to a submerged storage rack or to the canister 

dewatering station in the “A” spent fuel pool.494

• Canister storage racks. The defueling canister storage racks provided 

storage for loaded defueling canisters. Storage for a total of 

263 canisters was available in the racks located within the deep end of 

the fuel transfer canal inside the reactor building (11 canisters) and in 

the “A” spent fuel pool inside the fuel handling building 

(252 canisters).495, 496

• Fuel transfer system. The existing fuel transfer system was modified to 

transfer defueling canisters from inside the reactor building to the 

adjoining fuel handling building. Canisters were handled in a way 

similar to normal fuel assemblies during refueling operations. A canister 

was lowered into a modified “upender” in the flooded deep end of the 

fuel transfer canal, turned to a horizontal position, and moved through 

one of the fuel transfer tubes into the “A” spent fuel pool. A second 

“upender” at the other end raised the canister to a vertical position.

497

• Canister dewatering systems. The canister dewatering systems were 

used to purge water from submerged defueling canisters in order to 

sufficiently expose the recombiner catalysts and prevent the buildup of 

combustible gases in the canisters. Inert-gas injection systems displaced 

water from the canisters. Dewatering stations were installed in the 

reactor vessel and the “A” spent fuel pool. Typically, a loaded canister

was partially dewatered in the reactor vessel using bottled inert gas 

located at the defueling work platform. Water discharged from the 

canister during in-vessel dewatering remained in the reactor vessel. Any 

inert gas released by dewatering process was vented through the off-gas 

system. The fuel handling building’s dewatering station was also 

located underwater to shield workers from radiation.498

• Fuel transfer cask. The fuel transfer cask used to transfer a defueling 

canister from the “A” spent fuel pool to the shipping cask was both 

cylindrical and bottom-loaded; the transfer cask weighed about 

40,600 pounds when loaded with a defueling canister. The cask was 

suspended from the fuel handling building’s overhead crane, whose 

integral grapple and hoisting mechanism engaged a defueling canister 

and lifted the canister out of the water through a shielded platform into 

the shielded transfer cask. The bottom door of the cask was closed and 

the entire unit was moved with the overhead crane to the Model 125-B 
165

shipping cask located in the truck bay where it mated with the shipping 

cask’s loading tower. 499 

• Model 125-B shipping cask. The Model 125-B shipping cask was 

designed specifically to transport the loaded defueling canisters. The rail 

(train) cask could carry up to seven defueling canisters. (See further 

discussion in the prior section on Waste Management.) 

• Core flood tank modifications. The top of the “A” core flood tank was 

removed to allow the storage of cut lower core support assembly 

component pieces. The piping from the “A” core flood tank to the 

reactor vessel was cut and capped to prevent the possibility of fuel 

transport.500

• Containment air control envelope (CACE). The CACE was a new 

building external to the reactor building that enclosed the equipment 

hatch and backup personnel airlock. Access to the CACE from the 

outside was provided by a personnel door and a 27-foot-wide roll-up 

truck door. The building provided an area where cleanup equipment and 

materials could be assembled and staged prior to transfer into the 

reactor building, thus reducing worker stay times in radiation areas, 

resulting in occupational exposure savings. The CACE also functioned 

as a staging area for contaminated material removed from the reactor 

building, but was not designed to be a storage area for radioactive 

wastes. When both equipment hatch personnel airlock doors were

opened, the reactor building purge system induced air flow from the 

outside through the CACE and the CACE aided in controlling and 

confining potential airborne releases from the reactor building. The 

CACE included a HVAC system, consisting of two filtered exhaust 

trains and associated radioactive effluent release monitor, which 

operated to reduce airborne particulate contamination in the building 

and to protect the building from overpressurization. The HVAC system 

was operated to maintain net airflow into the CACE when isolated from 

the reactor building. Normally, the CACE was maintained at a slightly 

negative pressure, relative to the outside, to limit exfiltration from the 

building, except when the roll up door was open for staging equipment 

and materials into or out of the CACE in support of cleanup activities. 

Periodic monitoring of the CACE atmosphere was performed to ensure 

that potential airborne releases were within regulatory limits.501 The 

CACE was intended to support recovery activities through defueling 

only and was not designed to satisfy the criteria for a permanent TMI-2 

facility.502 The CACE was available for use in June 1986.503
166

Reactor Vessel Defueling. Removal of damaged fuel and structural debris 

from the reactor vessel started on November 12, 1985, six and one-half 

years after the accident. Numerous manual and hydraulically 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 were used to disassemble or cut apart 

reactor vessel components and break up resolidified core material, such as 

the core bore machine, plasma arc torch, and water-jet cutting system. After 

breaking up and sectioning of oversized debris, long-handled tools were 

used to manually load debris into defueling canisters positioned under water 

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 ”A” core flood tank. Smaller pieces or 

“fines” were vacuumed into specially designed knockout canisters and filter 

canisters. Other defueling activities included transferring the loaded 

defueling canisters from the reactor building to the fuel handling building, 

dewatering the filled canisters, and placing canisters into the canister storage 

racks located in the “A” spent fuel pool.504

Problems plagued the cleanup systems early because of microorganism 

growth in the reactor coolant. The resulting loss of visibility halted all 

defueling operations for a few months. Shown is a view down into the 

reactor vessel annulus region between the core support assembly and 

the vessel wall. (A light hanging at the lower right.)
167

Removal of fuel and structural material from the reactor vessel was

completed in stages. The plan for each stage incorporated experience gained 

from previous stages and activities, including in-vessel examinations. The 

NRC site office provided reviews and approvals of the licensee’s technical

evaluation reports and procedures. Defueling included the following stages: 

• Preliminary defueling. Preliminary defueling involved the 

rearrangement of core debris material within the reactor vessel to allow 

complete installation and rotation of the canister positioning system and 

to provide access for defueling tools. The preliminary activities also 

included identification and positioning of core debris samples in the 

reactor vessel. General movement of the core debris included loading of 

debris baskets (or containers), but not loading defueling canisters.505, 506

Preliminary defueling operations started on October 31, 1985.507

• Early defueling. Early defueling involved depositing loose core debris 

into defueling canisters and removing loaded canisters from the reactor 

vessel. The loaded canisters were placed in canister storage racks 

located in the fuel transfer canal in the reactor building and the 

“A” spent fuel pool in the fuel handling building. The debris consisted

of partial fuel assemblies, fuel rods, end fittings, structural materials, 

and loose granular fuel and structural fines. The debris was packaged

into fuel canisters by defueling operators using long-handled tools and 

First loading of debris into a fuel canister started on January 12, 1986. 

Shown is an upper fitting of a control rod cluster being dropped into a 
168

the in-vessel vacuum system. During canister loading, defueling 

personnel were supervised by a specially trained and NRC-licensed fuel 

handling senior reactor operator. Physical and administrative controls 

were implemented to prevent the inadvertent lifting of core debris out of 

the reactor vessel.508, 509 On November 12, 1985, the licensee received 

permission from the NRC onsite office to start loading the fuel canisters 

with fuel debris. Using the vise-grip long-handled tool, the defueling 

workers placed the first piece of debris, a broken piece of fuel rod, into 

the fuel canister.

510 On January 12, 1986, the first three loaded fuel 

canisters of core debris were transferred from the reactor vessel to the 

submerged canister storage rack in the “A” spent fuel pool in the fuel 

handling building.511

• Bulk defueling of the core region. Bulk defueling involved 

reinstallation of the core boring machine (which had previously been 

used to obtain core samples) over the reactor vessel to perform 

additional drilling operations into the resolidified mass in order to 

facilitate defueling. The drilling operations were permitted on the 

monolith in the core region above the lower core support assembly.512, 

513, 514, 515, 516 A total of 409 closely spaced holes in the resolidified 

material were started on October 20, 1986, and completed on 

November 14. After completion of the drilling, the drill rig was 

removed and the defueling work platform was reconfigured to support 

manual debris-removal activities.517, 518, 519 Equipment used during bulk 

defueling included “pick-and-place” long-handled tools, air-operated 

Bulk defueling system arrangement. 
169

chisels, and specially designed fuel-assembly pulling and grasping tools. 

Removal of the stub end fuel assemblies from the reactor vessel using 

custom tools started in March 1987 and was completed in 

September 1987. By September 1988, the entire original core region had 

been defueled.520 

• Removal of the lower core support assembly. This activity required 

three years of planning, inspection, research, and development to access 

the debris on the reactor vessel’s lower head. This massive assembly 

included the following layers (from top to bottom): lower grid rib 

section, lower grid flow distributor plate, lower grid forging, incore 

guide support plate, and elliptical flow distributor head. Defueling 

preparations required the use of the core bore machine, plasma arc 

cutting torch (also called the automated cutting equipment system), 

cavitating water jet, and other equipment to dismantle the multi-layered 

lower core support assembly. The core bore machine was used to cut 

assembly support posts and incore instrumentation guide tubes. The 

plasma torch was used to cut straight vertical and horizontal segments in 

the plates.521 Pieces of the assembly plates were stored inside the 

modified “A” core flood tank. Sections that did not contain incore guide 

By September 1988, the entire original core region had been defueled. 

Shown is a view down into the water-filled reactor vessel of the lower 

grid rib section of the lower core support assembly. Vertical baffle plates 

formed the peripheral boundary of the core.
170

tubes were bagged and stored inside the “A” steam generator’s D-ring. 

All pieces will remain inside the reactor building until the future 

decommissioning of TMI-2. Equipment used to remove loose and 

resolidified debris that remained on the sectioned remnants included a 

high-volume low-pressure water flush and a low-volume high-pressure 

cavitating water-jet flush. Displaced debris was collected on the reactor 

vessel’s lower head for eventual removal. Work was started in 

January 1988 and completed by March 1989.522, 523, 524

• Defueling of the upper core support assembly. This defueling activity 

involved the removal of core debris from between the baffle plates and 

core former plates. Defueling preparations required the use of the 

plasma arc torch to cut the baffle plates in order to enable access to the 

core debris on the core former plates. 525, 526 The baffle plates were cut 

into eight pieces. An untorquing tool and a drilling tool were used to 

Fuel debris at the bottom of the reactor vessel. The lower core support 

assembly structures originally retained about 6,000 kilograms of resolidified 

material around the circumference of the structures. The reactor vessel’s 

lower head region below the assembly contained about 12,000 kilograms of 

loose core debris and 7,000 kilograms of agglomerated core debris.
171

remove 864 bolts that held the baffle plates to the core barrel. 

Equipment used to remove loose debris from within the assembly 

included hydraulically powered counter-rotating brushes mounted on a 

pivoting deployment end effector and the in-vessel vacuum system. 

Resolidified debris was removed using mechanical methods and the 

cavitating water jet. Defueling of the assembly was completed in 

October 1989.527

• Defueling of the reactor vessel’s lower head. This defueling activity 

involved removal of core debris from the reactor vessel’s lower head. 

Defueling preparations required the removal of the gusseted incore 

guide tubes and sections of the elliptical flow distributor head. 

Equipment used to dismantle the lower core support assembly was also 

used in this activity.528, 529, 530 Equipment used to remove loose debris on 

the lower head included an airlift, long-handled tools, and the in-vessel 

vacuum system. The cavitating water jet and an impact hammer with a 

chisel point were used to break up the resolidified debris attached to the 

lower head. The removal of about 30 tons of core debris from the lower 

head was completed in November 1989.531 

• Final cleanup of the reactor vessel. Defueling of the reactor vessel was 

completed in December 1989. A video inspection inside the reactor 

vessel, in conjunction with sample analysis, was completed in 

January 1990 to determine the quantity of residual core debris in the 

vessel.

532 Final reactor vessel re-flushing and re-vacuuming for loose, 

dust-like debris was completed in March 1990. The remaining quantity 

of fuel in the reactor vessel was less than 900 kilograms (or less than 

one percent of the original inventory). The residual fuel consisted 

primarily of finely divided sediment with small particle sizes in 

inaccessible holes, crevices, corners, and surface films, as well as

resolidified material either tightly adherent to the reactor vessel’s

components or inaccessible for defueling. The final defueling report 

concluded that the residual fuel was not readily transportable between

locations and, therefore, that criticality would be precluded under 

postulated worst-case conditions.

533 (See further discussion in the 

section on After Defueling.) 

Defueling Outside the Reactor Vessel. The licensee estimated that about 

260 pounds of fuel debris was transported outside the reactor vessel during 

the course of the accident. Additionally, about the same amount was 

relocated outside the reactor vessel because of cleanup and defueling 

operations, mostly in the reactor coolant system. The remaining fuel in 

hard-to-reach locations will be removed during the future decommissioning 
172

of TMI-2. An extensive program of fuel measurements was implemented, 

which included direct measurement by instrumentation, visual inspection, 

and sample collection and analysis. The results from this program were

reviewed by the NRC to ensure that no amount could achieve criticality 

under worst-case conditions. (See further discussion in the section on After 

Defueling.) Defueling outside the reactor vessel included the following 

activities: 

• Reactor coolant system (RCS). The two methods of fuel transport to 

locations outside the vessel were from sequential operation of the four 

reactor coolant pumps and the “burping” phenomenon during natural 

circulation. About 228 kilograms of fuel debris was transported from 

the reactor vessel into the RCS during the accident. An estimated 

170 kilograms was added to the RCS during defueling operations. The 

reactor coolant system defueling activities removed more than 

90 percent of the debris in the pressurizer, decay heat system suction 

piping (also known as the “drop line”), and hot-leg piping; and 

approximately 70 percent of the debris on the steam generator’s upper 

tube sheets.534 The following techniques were used to defuel reactor 

coolant system components:

o Hot-leg piping. The two hot-legs were initially defueled using a 

combination scraper-and-vacuuming tool. Additional residual core 

debris in the “B” hot-leg was scraped, flushed, and vacuumed into 

defueling canisters as part of reactor vessel defueling.535 Defueling 

of both hot-legs was completed on August 10, 1988.536

o Pressurizer. Initially, the pressurizer was defueled using a 

submersible pump, knockout canister, filter canister, and agitation 

nozzle. The second phase of pressurizer defueling used a remotely

operated submersible vehicle equipped with an articulating claw 

and a scoop to remove larger pieces of debris located on the bottom 

head of the pressurizer.537, 538, 539 Defueling of the pressurizer using 

the vacuum system started in November 1987 and was completed 

using the submersible vehicle on June 14, 1988.540

o Pressurizer spray line. The pressurizer spray line defueling system 

was used to flush water from the defueling water cleanup system 

into the pressurizer and reactor coolant system “2A” cold-leg.

541

o Decay heat drop line. The in-vessel vacuum system was used to 

defuel the decay heat system suction piping drop line. A tool was 

developed to guide the vacuum hose into the vertical portion of the 
173

drop line from the “B” hot-leg. A plumber’s drain-cleaning machine 

was used to break apart a hard, compacted region of debris below 

the loose debris so vacuuming could continue. The material was 

airlifted into the “B” hot-leg and was removed as part of the hot-leg 

defueling.542 

o Steam generators’ upper tube sheets. Pick-and-place and 

vacuuming techniques were principally used to defuel the upper 

tube sheets in both steam generators. A vacuum system removed the 

smaller debris.543 The defueling of the “A” and “B” steam generator 

upper tube sheets was completed in September 1987 and October 

1987, respectively.544

• Reactor building. A small quantity of fuel was released to the reactor 

building as a result of leakage through the pressurizer pilot-operated 

relief valve during the course of the accident. The licensee estimated 

that the scarification and desludging activities in the reactor building’s 

basement had removed approximately 4,900 kilograms of sediment 

which contained approximately 4 kilograms of fuel. Sediment included 

mostly river water sediment from a leaking air cooler inside the reactor 

building, concrete dust, and dirt. A robotic desludging system removed 

about 40 percent of the basement floor area that was accessible. The 

removal efficiency of desludging was greater than 90 percent. About 

75 kilograms of fuel remain outside the reactor vessel in the reactor 

building, with the largest quantity attached to dismantled reactor vessel 

Fuel debris on the upper tube sheet of “B” steam generator.
174

components that were relocated for storage inside the “A” steam 

generator’s D-ring shield structure and the modified “A” core flood 

tank.545 

 

• Auxiliary and fuel handling building. A small quantity of fuel debris 

was transported into the auxiliary building, principally through the 

reactor coolant bleed tanks and the makeup and purification system 

during the accident. Additionally, a smaller amount of fuel might have 

been relocated into the auxiliary building as part of the post-accident 

water processing, cleanup, and defueling activities. About 3 kilograms

of fuel was removed from the makeup and purification system 

demineralizers and 370 grams of fuel was removed from the block 

orifice assembly positioned upstream of the demineralizers. The total 

quantity of fuel material remaining in the auxiliary building was 

estimated to be less than 17 kilograms. Removal of resins from the 

makeup and purification system demineralizers started in October 1987 

and completed in September 1988 after limited success.

Remotely operated submersible vehicle equipped with an articulating claw and 

a scoop was used to remove larger pieces of debris that were located on the 

bottom head of the pressurizer.
175

Defueling arrangement: workers on the defueling platform loaded fuel canisters; a fuel canister bridge transferred a canister from the reactor vessel to the fuel transfer tube and mechanism inside the reactor building; another fuel canister bridge in the fuel handling building loaded the canister into the fuel storage rack in the spent fuel pool; the fuel transfer cask and building crane transferred the processed canister from the pool to the shipping cask loading station; and the Model 125-B shipping cask was lowered from its upright position onto the rail car and prepared for shipment.
176

Shipments of Core Debris. The first shipment of core debris to INEL 

began in July 1986. The core debris transportation campaign consisted of 

22 rail shipments by dedicated train which resulted in the transport of a total 

of 49 casks loaded with 342 defueling canisters. The final fuel shipment 

from Three Mile Island to the INEL started on April 15, 1990.546 

The canister handling and preparation for shipment program included all 

activities necessary to prepare and transfer a loaded defueling canister from 

its storage rack in the “A” spent fuel pool to the shipping cask; to insert the 

canister into the shipping cask; and to verify that the shipping cask was 

prepared for transport in accordance with its NRC-issued certificate of 

compliance. The preparation of canisters for shipment required several 

activities using custom-designed equipment. Canister preparations included 

dewatering and purging the defueling canister with an inert cover gas, 

verification of final canister weights, verification that the catalytic 

recombiners installed inside the 

canister were functioning, and 

verification that the canister had been 

dewatered sufficiently to ensure that 

the catalytic recombiners remained 

operable regardless of canister 

orientation.547, 548

Following final preparations and 

checks, the canister handling bridge 

moved the defueling canister to the 

fuel transfer cask (FTC) loading 

station in the “A” spent fuel pool. 

The fuel handling building’s crane 

lowered the FTC onto the loading 

station platform over the pool where 

the FTC lifted the canister from the 

water by a grapple. As the canister 

breached the surface of the pool 

water, a spray of borated 

demineralized water washed the 

canister. The fuel handling building’s

crane then transferred the loaded 

FTC from the spent fuel pool and to 

the fuel handling building’s truck bay

for loading into the shipping cask. 

After verifying its conformance to its 

NRC-approved shipping cask 

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.
177

certificate of compliance, the loaded shipping cask was transported out of 

the fuel handling building and off the island by rail. A dedicated train 

transported up to three casks at a time to INEL.549 One shipment carrying 

two shipping casks experienced a minor accident as the train engine struck a 

car while traveling at slow speed. The engine received minor damage, but 

the shipping casks were not damaged.

550

Once a rail shipment arrived at the Central Facilities Area at INEL, both 

impact limiters were removed from each end of the shipping cask and the 

cask and skid were lifted from the rail car and placed on a tractor trailer. 

The cask was then transported 42 kilometers (26 miles) to wet storage at the 

INEL Test Area North. Here each canister was removed from the shipping 

cask, filled with water, placed into a storage module, and transferred to the 

storage pool in a designated location. A vent tube was installed on each 

canister for continuous venting. The empty shipping cask was surveyed for 

contamination and prepared for rail shipment as regular freight back to 

TMI-2.551, 552 A detailed account of fuel shipments, including public 

outreach, was provided in the DOE report “Historical Summary of the Three 

Mile Island Unit-2 Core Debris Transportation Campaign” 

(DOE-ID-10400). 

The 342 stainless steel fuel canisters of core debris were stored in

underwater storage from 1986 to 2001 at the INEL.553 During the 2000 to 

2001 period, these canisters were transferred to the TMI-2 Independent 

Spent Fuel Storage Installation, also located at INEL, for interim storage 

of the TMI-2 core debris.554 

Document Collections. Those documents discussed above and many other 

documents relating to defueling activities are provided in the DVD folder,

Defueling. The following types of documents are included in this collection: 

• system description, technical evaluation, and safety evaluation reports 

• research reports, NRC technical NUREG reports, GEND reports, and 

technical reports from DOE national laboratories

• other correspondence between the NRC and the licensee relating to 

defueling systems, equipment, and activities, such as notifications, 

requests, reviews, and approvals

Project schedules for defueling activities are provided in the DVD folder 

General Management and Oversight. 
178

Above: The Model 125-B shipping cask being removed from the rail car at the Central 

Facilities Area at INEL. Both impact limiters are shown removed from the ends of the 

cask. Below: Defueling canisters in the temporary storage pool at INEL. A vent tube 

was installed on each canister for continuous venting. 
179

9 After Defueling 

Defueling activities were considered complete with the shipment of the last 

remaining defueling canisters containing core material from the TMI site on 

April 15, 1990. The major objective of TMI-2 post-defueling activities 

focused on preparing the plant for long-term storage. The licensee called the 

period preceding the ultimate disposition (either refurbishment and restart or 

decommissioning) of the plant “post-defueling monitored storage” or 

“PDMS.” Long-term storage was proposed by the licensee on 

December 2, 1986.555 However, the concept of PDMS was first introduced 

by the NRC Advisory Panel for the Decommissioning of TMI-2 on 

April 12, 1984.556 The approach to PDMS was expanded when the licensee 

submitted its environmental evaluation of PDMS on March 11, 1987.557

During PDMS, the TMI-2 facility would be in long-term monitored storage, 

similar to the decommissioning mode SAFSTOR (mothballing with delayed 

dismantling), in which the facility is secured, monitored, and maintained in 

a manner that ensures the protection of the public health and safety for an 

extended period.558

The licensee stated in its December 2, 1986, post-defueling monitored 

storage plan that a monitored storage period would be beneficial for the 

following reasons: (1) occupational dose in the plant would be reduced 

during monitored storage because of natural decay of radioactive 

contamination; (2) a monitored storage period would allow time for 

continued development of decontamination technology; (3) further 

reduction of occupational exposure would be achieved through the use of 

advanced robotic technology, automatic cleaning and chemical cleaning 

techniques, and advanced waste treatment methods; and (4) developing 

technology for radioactive waste packaging and volume reduction could 

result in a reduction in the total volume of radioactive waste generated 

following PDMS. 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-1 

facility.559, 560 

Prerequisites for PDMS. The requirements for transition to PDMS were

contained in the TMI-2 possession-only license and the NRC-approved list 

of PDMS requirements and commitments.561 The basic criterion for 

transition to PDMS was assurance that the health and safety of the public 

was protected by conformance to all applicable NRC regulations.562

Transitioning to PDMS required the following conditions: (1) criticality was 

no longer possible; (2) potential for fission-product movement was 

eliminated; (3) fuel was removed and shipped offsite; (4) radioactive waste 
180

was shipped or stored; (5) radiation levels were reduced commensurately

with the need for access to permit continued plant monitoring and to support 

plant-disposition decisions; (6) water was removed from plant systems and 

spaces, and the potential for reintroduction of water was precluded; and

(7) a safe, monitored plant condition was established.563 

PDMS Environmental Protection Systems. The principle safety concern 

during PDMS was the inadvertent release of radioactive material into the 

environment. For this reason, the NRC identified structures, systems, and 

components that provided reasonable assurance that the facility could be 

maintained in a defueled condition without undue risk to the health and 

safety of the public. These systems, called PDMS environmental protection 

systems, included the following:

564

• Reactor vessel. Maintained residual debris geometry, precluding the 

possibility of an inadvertent criticality. To keep the residual fuel in the 

reactor vessel in the analyzed geometry during PDMS, the PDMS 

technical specifications limited activities that could alter the geometry 

of the fuel debris in the reactor vessel by controlling excess loads over 

the reactor vessel, limiting movement of remaining fuel debris outside 

analyzed geometries, and minimizing the potential for water 

accumulation in the reactor vessel.565 As a conservative measure, the 

licensee added 1,700 pounds of insoluble neutron poison in the form of 

borosilicate glass shards to the bottom of the reactor vessel to ensure

long-term subcriticality of the residual fuel in the reactor vessel.

566

• Containment structure. Ensured containment of the remaining 

radioactive contamination during the PDMS period. To maintain the 

integrity of the environmental barrier, inactive penetrations were closed 

off with isolation valves or with welded or bolted blind flanges. In 

addition, the PDMS technical specifications required routine 

surveillance inspections of containment penetration isolation. Isolation 

valves on active containment penetrations used by the containment 

atmospheric breather and the reactor building’s purge system would 

close on a high containment pressure.567

• Purge, breather, ventilation, and filtration systems. Controlled

radioactive effluents from the reactor building and the auxiliary and fuel 

handling building. Existing systems used during PDMS included the 

reactor building’s ventilation and purge system, the auxiliary building’s

ventilation and filtration system, and the fuel handling building’s

ventilation and filtration system. These systems would not be operated 

continuously, but on an as-needed basis. A passive containment 
181

atmospheric “breather” was installed for PDMS to maintain pressure 

equilibrium between the auxiliary and fuel handling building and the 

reactor building. A differential pressure would develop when the 

auxiliary building’s ventilation system was operating and the reactor 

building’s ventilation system was not operating. The containment 

atmospheric breather also provided a HEPA-filtered pathway for any 

effluent from the containment.568

• Fire protection system. Detected and mitigated any effects of a fire 

within the facility. The original TMI-2 system of fire protection had 

been modified to address the reduced functional requirements for fire 

protection for PDMS. The PDMS fire protection program consists of 

fire detection and alarm capability, manual suppression by fire brigade, 

removal of most flammable and combustible liquids and materials, 

control of transient combustibles, and deenergization of most electrical 

circuits.569

• Flood protection system. Minimized the intrusion of water in the 

facility and movement of radioactive contamination to the environment. 

The existing unit flood protection capabilities were maintained for 

PDMS, such as protective island dikes, flood panels, watertight doors, 

and an early warning system of flooding conditions.570

• Support and monitoring systems. Supported PDMS configuration to 

ensure personnel and environmental protection and surveillance. 

Systems included (1) electrical systems, such as area lighting, fire 

detection, radiation monitoring, PDMS support systems, and 

communications; (2) effluent monitoring systems for the reactor 

building’s purge system and ventilation systems; (3) environmental 

monitoring systems, such as radiation monitoring, sample collection and 

analysis, rodent carcass analysis, and pest control; (4) administrative 

systems, such as organizational structure, staff qualifications, records, 

independent safety reviews, procedures, occupational radiation 

protection, quality assurance plan, and emergency plan; and 

(5) surveillance programs, such as maintenance of reactor vessel

geometry, reactor building isolation, the reactor building’s breather and 

ventilation system, the auxiliary and fuel handling building’s ventilation 

system, fire protection, flood protection, support and monitoring 

systems, and radiological surveys.571

Key Reports. Reports that supported the licensee’s request and the NRC’s

approval of the proposed license amendments for PDMS (issuance of the 

possession-only license and issuance of the PDMS technical specifications) 
182

are summarized below. Issue dates and report conclusions are further 

summarized in the key actions section that follow.

• Defueling completion report. This report was developed by the licensee 

to document the measurements and calculations that were performed to 

ensure that the plant had been defueled to the extent reasonably 

achievable and that the potential for a nuclear criticality had been 

precluded during normal and accident conditions.572 The report was 

required by the plant technical specifications (shown in Table 1.1 of 

Amendment No. 30) to document the basis for the TMI-2 facility 

transition to Mode 2. (For transition from Mode 1 to Mode 2, the 

licensee was required to demonstrate that the reactor vessel and reactor 

coolant system had been defueled to the extent reasonably achievable, 

that the possibility of criticality in the reactor building was precluded, 

and that there were no defueling canisters containing core material 

remaining in the reactor building.)573 In addition, this report and 

subsequent reports provided the criticality safety analyses to support the 

proposed license amendment for the possession-only license and 

issuance of the PDMS technical specifications.574

The contents of the report included the following sections: overviews of 

defueling objectives and guidelines and residual fuel characterization 

(Section 1); detailed discussions of the post-accident fuel transport and 

dispersion conditions (Section 2); survey techniques used for residual 

fuel measurements (Section 3); major fuel-removal accomplishments 

and methods (Section 4); residual fuel quantification, by location, and 

criticality analyses for each fuel location, as appropriate (Section 5); an 

assessment of possible alternatives and projected occupational doses 

associated with attempts to remove the remaining fuel (Section 6);

cumulative occupational exposures during defueling-related activities 

(Section 7); licensee conclusions (Section 8); and criticality safety 

evaluation for the TMI-2 safe fuel mass limit (Appendix B).575 Some 

analyses and data in the report were later revised in the reactor vessel 

criticality safety analysis report and the reactor vessel post-defueling 

survey report.576, 577 

• PDMS requirements and commitments list. This document was 

developed and revised by the NRC and the licensee to list the 

requirements and commitments needed to place the facility into PDMS. 

The document provided requirements for the final PDMS configuration 

to support the proposed license amendment for the possession-only 

license, the PDMS technical specifications, and license conditions 

contained in the requirements for entry into PDMS. This list was 
183

generated from the PDMS safety analysis report that was submitted by 

the licensee in support of their license amendment request, as amended; 

the associated safety evaluation issued by the NRC on 

February 20, 1992; and several public meetings at TMI.578 The NRC 

approved a process to allow changes to the list of requirements and 

commitments in recognition of the difficulties associated with readying 

the facility for long-term storage and the dynamic nature of the effort. 

Deviation requests by the licensee had to include a safety analysis 

evaluating each proposed change.579 The list was incorporated in 

Amendment 45 to the facility license that modified the original 

operating license to a possession-only license.580 

• PDMS safety analysis report. This report was developed and revised by 

the licensee to document the facility description and safety analysis of 

the PDMS configuration. The report provided the basis for long-term 

stability and safety of the proposed facility license amendment for the 

possession-only license and the PDMS technical specifications.581 The 

report is the current licensing-basis document for PDMS and is 

periodically updated to reflect current plant conditions. 

The contents of the original report included: description of the current 

status of the plant after extensive decontamination (Section 1); site 

characteristics (Section 2); regulatory review of conformance of the 

TMI-2 facility to 10 CFR Part 50 (Section 3); description of 

fuel-removal activities and special nuclear materials accountability 

(Section 4); enumeration of the radiological status of the plant and 

radiological goals to be attained before entry into PDMS (Section 5); 

list of deactivated systems and facilities (Section 6); description of 

operational systems and facilities (Section 7); identification and 

quantification of routine and unanticipated releases during PDMS 

(Section 8); description of the proposed changes to the technical 

specifications to permit entry into PDMS (Section 9); and primary 

administrative functions for the management of TMI-2 during PDMS, 

such as quality assurance, security, emergency preparedness, radiation

protection, and organizational responsibilities (Section 10). The original 

submittal reflected the anticipated facility conditions at the beginning of 

PDMS. The report was amended several times based on new 

information, responses to formal questions from the NRC, and changes 

in specifications for the facility. Routine updates are required by 

regulatory requirements.

582 

• Post-defueling survey reports. A series of 10 reports were developed by 

the licensee to document the assessment of special nuclear material
184

(SNM) remaining in the plant following the completion of the defueling 

effort. A report was issued for the following reactor components: upper 

plenum, letdown coolers, pressurizer, reactor vessel head, reactor

building basement, both steam generators, auxiliary and fuel handling 

building, miscellaneous components in the reactor building, reactor 

coolant system, and reactor vessel.583

The geometric model

that was used to 

conduct the criticality 

safety analysis of the 

residual fuel in the 

reactor vessel. An 

annular ring, 

representing the vessel 

internals and 

postulated debris 

accumulations located 

along the outer 

periphery of the 

vessel, was 

conservatively 

assumed to go 360 

degrees around the 

vessel. The analysis 

used in-vessel 

inspections of debris 

locations and some 

conservative estimates 

of the remaining fuel 

to develop a specific 

three-dimensional 

analytical model of the 

reactor vessel endstate configuration. 

Criticality calculations 

were performed by the 

Oak Ridge National 

Laboratory. A 

separate criticality 

assessment was 

performed for 

accident scenarios. 

(See defueling 

completion report for 

additional 

information.)
185

On October 17, 1985, the licensee was granted an exemption from 

certain requirements for periodic inventory and reporting of SNM (e.g., 

uranium-235 and plutonium) balance for TMI-2. As a condition of the 

exemption, the licensee was required to conduct an assessment of the 

SNM remaining at TMI-2 following the completion of the defueling 

effort.584 The compilation of the individual survey reports provided the 

basis for the final assessment of the quantity of residual SNM for 

accountability purposes.585 In addition, the reactor vessel post-defueling 

survey report provided the basis for satisfying the criticality 

determination requirement for entry into PDMS.586 

The contents of each report included: (1) a detailed description of the 

area, system, or component; (2) its role in the accident and/or cleanup 

activities; (3) the methodology used to determine the quantity of SNM; 

(4) the rationale supporting a conclusion as to whether the area, system, 

or component contained residual SNM and, if so, a summary of the 

appropriate SNM engineering calculations; (5) applicable drawings of 

the area; and (6) an assessment of residual fuel.587 Because of the 

complex structure of the components surveyed, the assessment of the 

residual SNM used a combination of direct measurements, sample 

analyses, volumetric measurements, and engineering analyses. The 

survey reports were supported by detailed engineering calculations, 

contractor reports, and research data.588

• Programmatic Environmental Impact Statement (PEIS) 

Supplement 3. The last supplement to the initial PEIS (NUREG-0683) 

was developed by the NRC to document the environmental evaluation 

of the licensee’s proposal to complete the current cleanup effort and 

place the facility into monitored storage for an unspecified period of 

time. This supplement was used by the NRC to evaluate the licensee’s 

proposed facility license amendment for the possession-only license and 

the proposed PDMS technical specifications. The supplement provided 

an environmental evaluation of the licensee’s proposal and a number of 

alternative courses of action from the end of the current defueling effort 

to the beginning of decommissioning. However, the objective of the 

report was not to provide an evaluation of the environmental impacts 

associated with the decommissioning. 

The contents of the report included: status of the cleanup activities and 

conditions in the plant, including radiation-source characteristics 

(Section 2); evaluation of the licensee’s proposal and NRC-identified 

alternatives for potential environmental impacts, such as the offsite 

population’s exposure to radiation from routine and accidental releases, 
186

occupational radiation dose, waste management impacts, transportation 

impacts, socioeconomic impacts, commitment of resources, and 

regulatory considerations (Section 3); description of the environment 

and population that could be affected by the licensee’s proposed action 

and alternatives (Section 4); comparison of environmental impacts of 

the proposal and alternatives (Section 5); discussion of the potential for 

human health effects from both offsite and onsite occupational radiation 

exposures as a result of TMI-2 cleanup (Section 5.2); discussion of 

nonradiological impacts identified, such as the cost of implementation, 

long-term commitment of land and burial-ground space, and 

socioeconomic effects (Section 5.3); discussion of postulated accidents, 

such as radiological impacts resulting from accidents onsite and offsite 

during waste transportation, and nonradiological impacts including 

traffic accidents, injuries, and fatalities (Section 5.4); NRC conclusions 

(Section 6); and NRC responses to public comments on the draft 

supplement (Section 7).

• PDMS technical evaluation report. This report was developed by the 

NRC to document the technical evaluation of the licensee’s proposal to 

place the TMI-2 facility into PDMS. The report provided the basis for 

the requirements and controls to be maintained during PDMS. The 

report also provided the basis for approving amendments to the facility 

license that issued the possession-only license (Amendment No. 45) and 

the PDMS technical specification (Amendment No. 48).589

The contents of the report included a brief regulatory history of the 

TMI-2 facility (Section 2); description of PDMS (Section 3); status of 

the facility before entry into PDMS (Section 4); description of the major 

prerequisites for facility configuration at the start of PDMS (Section 5); 

discussion of the structures, systems, and components for environmental 

protection during PDMS (Section 6); and conclusions (Section 7).590

• PDMS safety evaluation report. This report was developed and revised 

by the NRC to document the safety evaluation of the licensee’s proposal 

to place the TMI-2 facility in PDMS. The report provided the basis for 

approving amendments to the facility license that issued the 

possession-only license (Amendment No. 45) and the PDMS technical 

specifications (Amendment No. 48). The initial safety evaluation report 

was issued in February 1992 in response to the initial licensee 

application in August 1988. This safety evaluation was updated to 

account for revisions to the licensee application.591, 592
187

The contents of the report included: a summary of licensee submittals 

and NRC actions associated with the licensee’s initial amendment 

request in 1988 (Section 2); review of licensee actions that satisfied the 

license condition for entry into PDMS specified by the possession-only 

license (Section 3); and evaluation of the proposed PDMS technical 

specifications, as revised, and comparison to the existing TMI-2 

technical specifications, as amended (Section 4). The technical 

evaluation report (see above) was issued concurrently with this 

document to provide additional details. Both documents were prepared 

by the Pacific Northwest Laboratory (currently Pacific Northwest 

National Laboratory) under the direction of the NRC. 593, 594

Key Actions. In August 1988, the licensee submitted a safety analysis 

report that documented and supported their proposal to amend the TMI-2 

license to a “possession-only” license and to allow the facility to enter 

PDMS. In response to the request, the NRC issued Final Supplement 3 to 

the Programmatic Environmental Impact Statement for the TMI-2 

decontamination and cleanup in August 1989. In February 1992, the NRC 

issued a safety evaluation report regarding the PDMS license amendment 

and a technical evaluation report regarding PDMS. These three NRC 

documents formed the basis for the position on the acceptability of PDMS. 

The NRC issued a possession-only license in September 1993 and approved 

the PDMS technical specifications three months later. Key post-defueling 

activities and actions are summarized as follows: 

• Licensee submitted license amendment request. On August 16, 1988, 

the licensee submitted a request to amend the operating license to a 

possession-only license and to extensively modify the technical 

specifications in ways consistent with the licensee plans for long-term 

storage at the facility. The request included the proposed amended 

facility license for PDMS, proposed PDMS technical specifications, and 

the PDMS safety analysis report.595 The pre-accident and recovery 

technical specifications consisted of two parts: Appendix A, which 

pertained to the facility and recovery, and Appendix B, which pertained 

to the environment. The licensee proposed combining these two 

documents into one set of technical specifications for PDMS. Also, the 

licensee proposed placing the remaining surveillance requirements for 

PDMS, which were in the recovery operations plan, back into the 

technical specifications.596

• NRC issued PEIS Supplement 3. In August 1989, the NRC published 

Final Supplement No. 3 to the Programmatic Environmental Impact 

Statement (NUREG-0683), which dealt with PDMS and subsequent 
188

cleanup. The NRC assessed the licensee’s proposal and six alternatives. 

The licensee’s proposal and one of the alternatives (continuing and 

completing the cleanup without a storage period) were evaluated in 

detail. The NRC concluded that both the licensee’s proposed plan and 

the NRC staff-identified alternative for completion of cleanup were

within the applicable regulatory limits, and each could be implemented 

without significant environmental impact. Neither alternative was found 

to be clearly preferable from an environmental impact perspective.597, 598

• Defueling completed. In April 1990, the TMI-2 facility transitioned 

from Mode 1 to Modes 2 and 3, in accordance with the transition 

requirements set forth in Table 1.1 in the recovery technical 

specification. In Mode 2, it was recognized that defueling was 

completed and, thus, boration of the reactor coolant system and staffing 

of the control room by licensed operators were no longer required. In 

Mode 3, it was recognized that offsite shipment of the fuel was 

completed and boration of the spent fuel storage pools was no longer 

required. The three criteria for changing from Mode 1 to Mode 2 were 

as follows: (1) the reactor vessel and reactor coolant system were 

defueled to the extent reasonably achievable; (2) the possibility of a 

criticality in the reactor building was precluded; and (3) there were no 

defueling canisters containing core material in the reactor building. The 

additional requirement for transition to Mode 3 was that no defueling 

canisters containing core material remained on the TMI site. The NRC 

and consultants from the Pacific Northwest Laboratory performed a 

detailed technical review and inspection to verify that the criteria were 

met. The facility made the transition from Mode 1 to Mode 2 on 

April 26, 1990, and to Mode 3 the following day.599, 600, 601

• Reactor vessel drained down. During July and August of 1991, the 

reactor vessel was drained to make final measurements of residual fuel 

remaining in the vessel. The reactor vessel fuel-measurement program 

was the final step in the special nuclear materials accountability 

program at TMI-2.602 

• Petition to intervene in the PDMS license amendment request. On 

April 25, 1991, the NRC published a notice of opportunity for a prior 

public hearing regarding the license change to implement PDMS. One 

individual petitioned to intervene. The petitioner, the licensee, and the 

NRC staff reached a settlement on September 25, 1992, and the request 

to intervene was withdrawn; on October 16, 1992, the NRC Atomic 

Safety and Licensing Board dismissed the proceeding.603 
189

• NRC issued safety evaluation of PDMS. On February 20, 1992, the 

NRC issued a safety evaluation addressing the license conditions and 

technical specifications necessary to implement PDMS. As part of the 

evaluation, the staff published a technical evaluation report which 

appraised PDMS as an integrated process and assessed licensee 

commitments that were not in the technical specifications. The safety 

and technical evaluation reports and Supplement 3 to the NRC’s Final 

Programmatic Environmental Impact Statement formed the basis for the 

staff’s position on the acceptability of PDMS.604, 605 

• Decision not to restart. On October 26, 1992, the licensee informed 

NRC of its intention to discontinue maintaining the TMI-2 restart list 

that included a list of NRC generic communications (bulletins, generic 

letters, and information notices) that would be reviewed for applicability 

and required action if a decision was made to refurbish TMI-2 for 

operation. The licensee had publicly acknowledged that TMI-2 would 

not be refurbished as a nuclear power generation facility, but instead, 

TMI-2 would be kept in a monitored storage condition until 

decommissioning of Units 1 and 2 simultaneously.606

• Licensee issued PDMS requirements and commitments. On 

January 15, 1993, the licensee provided a proposed list of remaining 

PDMS requirements and commitments that had to be completed before

issuance of the possession-only license and PDMS technical 

specifications. This list was generated from the PDMS safety analysis 

report; the NRC’s February 20, 1992 safety evaluation report; and 

several public meetings at the nearby TMI training facility. The NRC 

approved this list on May 19, 1993, and approved subsequent revisions 

to the list, as well. 

• Licensee issued the reactor vessel post-defueling survey report. On 

February 1, 1993, the licensee notified the NRC in its last 

post-defueling survey report that the current best estimate of the residual 

fuel remaining in the reactor vessel was 925 kilograms with an 

uncertainty of plus or minus 40 percent as one standard deviation. The 

estimate of remaining fuel in the reactor vessel was based on 

underwater video inspections and passive neutron measurements. Video 

inspections were used to divide the reactor vessel into nine zones which 

separated the major fuel deposits by elevation. An array of helium-filled 

detectors were used to measure fast neutrons produced by the residual 

fuel as water was removed from the vessel in stages so that the water 

could be used as a shadow shield to separate the fuel deposits by zone. 

The estimate was derived from calculations made by onsite staff and an 
190

independent review by an offsite group headed by Dr. Norman 

Rasmussen of the Massachusetts Institute of Technology. The total 

residual fuel estimate (925 kilograms) from fast-neutron measurements 

was about 50 percent larger than the less accurate video estimate.607 

For the balance of the facility outside the reactor vessel, earlier licensee 

estimates based on measurements, sample analyses, and visual 

observations indicated that no more than 385 pounds (174.6 kilograms) 

of residual fuel remained.

608, 609

• NRC issued safety evaluation of reactor vessel criticality analysis. On 

July 6, 1993, the NRC issued a safety evaluation confirming earlier 

analyses done by the licensee which indicated that the fuel remaining in 

the TMI-2 reactor vessel would remain subcritical, with an adequate 

margin of safety, during PDMS. The NRC and consultants from Pacific 

Northwest Laboratory performed independent evaluations and made 

independent measurements of earlier fuel measurements in the auxiliary 

and reactor buildings. The NRC concluded that the fuel remaining in the 

reactor vessel would be subcritical under both steady-state and accident 

conditions. The NRC also found that there were insufficient discrete 

quantities of fuel in areas other than the reactor vessel to sustain a 

criticality.610, 611, 612

• NRC issued possession-only license. On September 14, 1993, the NRC 

issued a possession-only license in Amendment 45 to the facility 

operating license without modification of existing recovery technical 

specifications. The PDMS technical specifications would be issued after 

the licensee had completed the final stages of the cleanup effort; the 

NRC had verified the implementation of the PDMS commitments and 

requirements; and the licensee had satisfied a number of PDMS license 

conditions. These license conditions included (1) completion of a 

special study of the ventilation in the auxiliary and fuel handling 

building that monitored airborne levels for one year before termination 

of the continuous operation of the ventilation systems; (2) development 

of an NRC-approved surveillance requirement for the unfiltered 

leak-rate test in the reactor building; and (3) submittal and 

implementation of a site flood protection plan, a site radiation protection 

plan, an offsite-dose calculation manual, a PDMS fire protection 

program evaluation, a PDMS quality assurance plan, and a radiological 

environmental monitoring plan. In addition, the licensee was required to 

submit the results of the completed plant radiation and contamination 

surveys before entry into PDMS.613
191

• PDMS requirements and commitments completed. On 

November 12, 1993, the licensee informed the NRC that all of the 

requirements and commitments for entry into PDMS had been satisfied 

and that they would be ready to transition to PDMS within the next 

30 days. The licensee’s letter to the NRC provided a list of all PDMS 

entry requirements along with references to the licensee letters that 

provided closeout documentation.

614, 615 The NRC independently 

verified that the licensee had satisfied all the PDMS requirements and 

commitments.616

• NRC issued PDMS technical specifications. On December 28, 1993, 

the NRC issued Amendment 48 to the facility operating license to 

incorporate the PDMS technical specifications. The PDMS technical 

specifications replaced the technical specifications that pertained to the 

facility and recovery (Appendix A to the facility operating license); 

technical specifications that pertained to the environment (Appendix B); 

and the recovery operations plan that pertained to surveillance. The 

license amendment also included an update of the safety evaluation 

report for PDMS that was first issued on February 20, 1992. In the 

safety evaluation, the NRC evaluated the potential for the routine and 

accidental releases of any significant quantity of radioactive material 

during PDMS. The NRC performed independent evaluations of eight 

potential accidents: vacuum canister failure, spraying of concentrated 

contamination with high-pressure spray, cutting of contaminated pipe, 

break of contaminated pipe, elevator/stairwell fire in containment, 

D-ring compartment fire in containment, containment penetration 

failure, and the rupture and release of resins from the makeup and 

purification demineralizers.617

During the safety evaluation, the NRC reviewed the final defueling 

completion report (1990) and the PDMS safety analysis report, as 

amended. The following conclusions of the NRC’s safety evaluation 

were based on the information from the licensee’s reports, the NRC’s 

PEIS Supplement 3 (1989), and the NRC’s PDMS technical evaluation 

report (1992):618

o Defueling of the reactor had been accomplished to the extent 

reasonably achievable.

o All fuel and core debris removed from the reactor and associated 

systems had been shipped offsite.
192

o Results of analyses indicated that there was no potential for 

criticality in the fuel remaining in the TMI-2 facility during either 

normal or accident conditions.

o Remaining radioactive waste from the major TMI-2 

decontamination activities had been shipped offsite or packaged and 

staged (prepared) for shipment offsite.

o Radiation levels within the facility had been reduced to the extent 

that plant monitoring, maintenance and inspections could be 

performed. 

o Radiological surveillance of activities during PDMS could be 

conducted in accordance with the approved offsite-dose calculation 

manual and in compliance with the regulatory requirements of 

10 CFR Part 20, “Standards for Protection against Radiation,” 

which would, with the approved radiation protection plan, ensure 

adequate control of occupational exposure and protection of 

workers.

o The surveillance program proposed by the licensee would

adequately monitor the PDMS environmental protection systems. 

Independent spent fuel storage installation (ISFSI) for the dry storage of 

TMI-2 fuel debris located at the INEL. The TMI-2 defueling canisters were 

removed from the storage pool, dewatered, dried, and placed in a dry shielded 

canister. The dry shielded canister with up to 12 defueling canisters was 

placed inside a reinforced concrete horizontal storage module. The ISFSI was 

certified and licensed by the NRC.
193

o Environmental monitoring activities for TMI-2 during PDMS, 

included in the TMI site radiological environmental monitoring 

plan, would ensure adequate environmental surveillance and 

control. 

o Fire prevention, detection, and control as specified by the approved 

fire protection program evaluation would ensure adequate reduction 

of fire potential, as well as detection and control during PDMS.

o Requirements delineated in the proposed PDMS technical 

specifications provided assurance that the facility would be 

maintained in a safe condition that would not negatively impact the 

environment. 

Document Collections. Those documents discussed above and many other 

documents relating to post-defueling activities are provided in the DVD 

folder, After Defueling. 

TMI-2 independent spent fuel storage installation located at INEL for interim 

storage of the TMI-2 core debris.
194

(FirstEnergy, 2015)
195

10 Postscript

In 2001, FirstEnergy acquired TMI-2 from GPU. FirstEnergy contracted the 

monitoring of TMI-2 to Exelon, the current owner and operator of TMI-1. 

The companies plan to keep the TMI-2 facility in long-term, monitored 

storage until the operating license for the TMI-1 plant expires, at which time 

both plants would be decommissioned. On-site NRC resident inspectors and 

inspectors from the regional office in King of Prussia, PA continue to 

monitor the operations at TMI-1 and the conditions at TMI-2. 

This knowledge management volume ends with the licensing of the PDMS 

configuration and the issuance of the PDMS technical specifications at the 

end of 1993. Activities to maintain and oversee the TMI-2 facility in PDMS 

will continue until the decommissioning of both units on Three Mile Island. 

Documents and correspondence pertaining to the maintenance of PDMS, 

plans for decommissioning, and associated regulatory activities can be 

accessed electronically from the NRC public website (www.nrc.gov). 

The Agencywide Documents Access and Management System (ADAMS) is 

the official recordkeeping system through which the NRC provides access 

to collections or “libraries” of publicly available documents. The NRC’s

Publicly Available Records System (PARS) Library provides full-text 

documents dating from about April 2000 onward (and some earlier 

documents). The Public Legacy Library contains bibliographic citations 

(some with abstracts and full text) for earlier documents, including pre- and 

post-accident time periods. Documents relating to TMI-2 can be found in 

the ADAMS libraries by searching the TMI-2 Docket Number 0500302. 

Most documents prior to April 2000 are only available in microfiche format. 

Refer to the NRC Public Document Room webpage for instructions for 

obtaining documents from the microfiche collection. NRC staff can access 

legacy documents from microfiche stations located at headquarters and 

regional offices.

Other sources of documents relating to TMI-2 can be currently found from 

the sources listed below. 

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

collection of their journal articles and proceedings on every aspect of 

the TMI-2 accident. In particular, the proceedings of the topical meeting 

“The TMI-2 Accident: Materials Behavior and Plant Recovery 

Technology,” held in Washington, DC in 1988 were published in the 
196

special volume of the Nuclear Technology journal. The papers in the 

proceedings and other papers may be purchased from their website. 

• Department of Energy's Office of Scientific and Technical 

Information website (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 

provided on the DVDs to this NUREG/KM. 

• Dickinson College maintains the Three Mile Island website

(www.threemileisland.org) that contains a collection of documents 

relating to the accident, including transcripts and audio recordings of 

interviews collected at the time, government and industry documents, 

photographs, and newspaper coverage. 

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

their website many of their research reports that supported the cleanup 

and understanding the accident. Four comprehensive reports on the 

accident, recovery, and cleanup include “Analysis of Three Mile Island 

- Unit 2 Accident” (NSAC-80-1), “The Cleanup of Three Mile Island 

Unit 2, A Technical History: 1979 to 1990” (EPRI-NP-6931), “TMI-2 

Left: The TMI-2 Investigation Center was established on June 2, 1979, at 

the Commission’s Washington, DC office to provide a facility for 

investigators from congressional committees and the president's 

commission to access all pre- and post-accident documentation concerning 

the plant. Right: A technical information assistant uses the first generation 

automated Document Control System (DCS) to search NRC documents 

since 1978, and all TMI-related documents. Each document received or 

generated by the NRC was indexed by bibliographic elements and 

microfiched for storage and use. The early DCS was replaced with the 

Nuclear Documents System (NUDOCS) in 1987 with text search capability 

of abstracts and select full-text documents. The Agencywide Documents 

Access and Management System (ADAMS) replaced NUDOCS in April 

2000 with full-text storage and retrieval of all NRC records and documents 

that were entered into ADAMS.
197

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

• Idaho National Laboratory of the Department of Energy maintains a 

collection of research reports in their INL Knowledge eRepository 

(http://keros.inl.gov). Most of these documents are provided on the 

DVDs to this NUREG/KM.

• International Atomic Energy Agency (www.iaea.org) made available 

on their website technical reports that incorporated lessons from the 

response to the accident and the cleanup. Key reports include 

“Catalogue of Methods, Tools and Techniques for Recovery from Fuel 

Damage Events” (IAEA-TECDOC-627), “Issues and Decisions for 

Nuclear Power Plant Management after Fuel Damage Events” (IAEATECDOC-935), “Management of Severely Damaged Nuclear Fuel and 

Related Waste” (IAEA-TRS-321), and “Cleanup and Decommissioning 

of a Nuclear Reactor after a Severe Accident” (IAEA-TRS-346). 

• National Technical Information Service of the U.S. Department of 

Commerce serves as the largest central resource for government-funded 

scientific, technical, and engineering information. Formal government 

reports not available elsewhere can be purchased from their website

(www.ntis.gov).

• Pennsylvania State University Libraries’ maintains the TMI-2 

Recovery and Decontamination Collection that contains several 

thousand videotapes, reports, and photographs of the recovery and 

cleanup during the 1979–1990 time period (www.libraries.psu.edu). The 

GPU Nuclear, EPRI, NRC, and DOE (GEND) co-sponsored a project 

with the University to catalog and maintain the extensive videotape 

library for future researchers. 

• WM Symposia (www.wmsym.org) holds an annual international Waste 

Management conference covering the management of radioactive waste 

and related topics. Past proceedings of their annual conferences dating 

back to 1975 are available from their website (www.wmsym.org). Many 

papers relating to the TMI-2 cleanup were presented during the 1980s. 

• Books about TMI-2 by NRC staff include “Three Mile Island: A 

Nuclear Crisis in Historical Perspective” by J. Samuel Walker (2006)619

and “TMI 25 Years Later: The Three Mile Island Nuclear Power Plant 

Accident and Its Impact” by Bonnie A. Osif, Anthony J. Baratta, and 

Thomas W. Conkling (2004)620. 
198
199

11 DVD Navigation and Interpretation

This knowledge management guide is an HTML-based collection of 

information, documents, videos, and photos related to the 1979 accident at 

the Three Mile Island nuclear power plant. This interactive feature is 

provided on the DVD to help you navigate through the historical records. 

Refer to the “Readme” file located on any DVD for viewing instructions.

Navigation 

• Main welcome page. From the welcome page, you can navigate the 

different sections of the guide using the blue tabs from the menu on the 

left side of the page. Each tab is linked to a page that contains one or 

more subtabs that appear to the right of the tab. Each tab corresponds to 

a document collection and each subtab links to a folder of documents. 

This interactive feature will prompt you to load the appropriate DVD to 

retrieve the document, if necessary. Additionally, there is a photo 

gallery, interactive timeline, and a document index (green tab). You 

may also click on the text box on the timeline for a short description. 

• Document retrieval. Documents can be accessed from the document 

page (see figure below) linked to the subtab. Next to each document is a 

hyperlink which will display the associated document, photo, or video. 

Because the document collection spans several DVDs, you may be 

prompted to insert a different DVD as you navigate the guide. A sorting 

feature is provide on select columns on the document page.

• Document searches. A simple keyword search feature is provided on 

the welcome page and each document page. A search is applied to a list 

of all documents on all DVDs in this NUREG/KM. This feature 

searches words in the file name.

• Document lists. Several convenient lists of documents on the DVD can 

be viewed from the “Document Folder Index” tab. These lists are also 

provided in spreadsheet format in the “Common” folder on the DVD 

(located in the "Documents" folder). A list of more than 25,000 TMI-2 

records in the Public Legacy Library is included in the spreadsheet.

• Alternative. Documents are titled and arranged in topical folders on the 

DVD (located in the "Documents" folder) to provide a usable alternative 

to the interactive guide. The document can be accessed directly on the 

DVD using a file explorer.
200

Things to Keep in Mind

• Legacy documents. 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 

provided on the DVDs are for historical reference only and are not 

official NRC records. Please refer to the NRC’s public website

(http://www.nrc.gov) for current information on regulations, policy 

statements, regulatory guidelines, regulatory processes, and research 

results.

• Units of measure. The unit of measure (in English units or the 

International System of Units) that was used in the original source 

document was used in this digest. A conversion chart is provided on the 

back cover.

• Abbreviations. A small set of abbreviations was used throughout this 

report in order to improve readability: ALARA, CFR, DOE, DVD, 

EPRI, GEND, GPU, INEL, NRC, NUREG, NUREG/CR, NUREG/KM, 

PDMS, and TMI-2. Others that were less frequently repeated in the 

report were spelled out at the beginning of each subsection or paragraph 

that contained them.

• Recovery vs. cleanup. 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 actions taken to decontaminate and defuel the plant and 

dispose of radioactive waste. These two terms are often used 

interchangeably for certain actions.

• EPRI and GPU documents. Documents generated by EPRI and GPU 

are generally not provided on the DVDs unless the documents were 

submitted to the NRC or funded by DOE. 

• “GPU” and “licensee.” Unless otherwise noted, the GPU Corporation 

and its subsidiaries (including GPU Service Corporation, GPU Nuclear 

Corporation, and Metropolitan Edison Company) are referred to 

collectively in a historical context as “GPU” or the “licensee” in the text 

of this NUREG/KM and document filenames.

• Document accession number. Each document and each enclosure of a 

document that was cataloged in the Agencywide Documents Access and 

Management System (ADAMS) Public Legacy Library was assigned a 
201

unique accession number. This number can be found on the first page of 

each document. Each enclosure to a transmittal letter was typically 

assigned its own accession number.

• Document filename. The filenames used on the DVDs typically contain 

the date of the document (generally the date of the transmittal letter); 

originating organization (e.g., NRC or GPU); document type 

(e.g., safety evaluation or system description); short title; document 

revision, if any; and reference date of previous correspondence, if any. 

Technical reports by the NRC (e.g., NUREGs), DOE, and national 

laboratories start with the report’s identification number, short title, and 

year and month issued. See examples below.

(1990-01-18) GPU, Defueling Completion Report, Rev. 3 (re 08-18, 10-09-1989)

NUREG-0698, Rev. 2, NRC Plan for Cleanup Operations at TMI-2 (1984-03)

• Documents, photographs, and diagrams in this NUREG/KM and 

DVDs were copied from the best available (surviving) sources. 

Photographs were generally taken by GPU and DOE contractors.

A typical document page listing the contents of a document folder. Columns 

can be sorted. Keywords in document file names can be searched. The 

hyperlink directs you to the document, photo, or video. 
202

Computer-aided design cut-away drawing showing the defueling work platform 

support structure and internals indexing fixture (purple), reactor vessel (green), upper 

core support assembly (red), and lower core support assembly (bottom inside vessel).
203

12 Contributions to Supplement 1

 (NUREG/KM-0001, Recovery and Cleanup)

Knowledge

Don Marksberry 

Gordon Skillman 

Michael Masnik

Management 

Don Marksberry 

Felix Gonzalez 

Kenneth Hamburger

Theodore Smith

Anita Aikins-Afful

Robert Norman 

Thomas Wellock 

Andrew Bates

Justin Poole

David Aird

Ian Gifford

Tom Kardaras 

Mark Henry Salley

and 

The NRC Print Shop and Publications Staff

Document Collections 

NRC’s Agencywide Documents Access and Management System

Publications Branch of the NRC’s Office of Administration 

NRC’s Office of the Secretary

U.S. Department of Energy

Idaho National Laboratory 

Electric Power Research Institute

Dickinson College Community Studies Center 

Pennsylvania State University Engineering Library

National Archives and Records Administration 
204

Cross-sectional view of the core bore drilling machine mounted on a platform over the 

defueling work platform. The core bore machine was originally used to extract core 

stratification samples in 1985 and later to break up the solidified monolith in the core 

region during defueling. The machine was used again to drill through and cut portions

of the lower core support assembly.
205

13 Endnotes

Note: Endnote citations are file names of documents on the DVD, except for 

EPRI reports and journal papers.

 

1 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 2 NUREG-1350, Vol. 25, Information Digest, 2013–2014 3 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 4 NUREG-0698, Rev. 2 5 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 6 (1980-03-26) Coordination Agreement, TMI Information and Examination Program (DOENRC-GPU-EPRI) 7 (1982-03-19) DOE-GPU Agreement, Acquisition of Damaged TMI-2 Reactor Core by 

DOE

8 (1982-04-19) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear Waste 

from TMI-2 (SECY-82-165) 9 DOE-ID-10276, TMI-2: Lessons Learned by the U.S. Department of Energy, A 

Programmatic Perspective, March 1990

10 NUREG/CR-1250, Vol. 2, Part 3 (Rogovin Report) 11 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 12 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 13 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 14 NUREG-0698, Rev. 2 15 (1981-08-26) GAO-EMD-81-106, Greater Commitment Needed to Solve Continuing 

Problems at TMI

16 (1986-01-26) Operating License Amendment 26 (Fully incorporates Recovery Mode 

Proposed Technical Specifications)

17 (1981-08-26) GAO-EMD-81-106, Greater Commitment Needed to Solve Continuing 

Problems at TMI

18 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 19 NUREG-0698, Rev. 2 20 NUREG-0698, Rev. 2 21 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 22 NUREG-0698, Rev. 2 23 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications 24 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 25 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 26 (1979-05-26) NRC Policy, NRC to Advise Use of EPICOR-II After Public Comment 27 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS (TMI-2) 28 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 29 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS (TMI-2) 30 (1980-09-26) NRC Policy, State (Pennsylvania) Requirements at TMI 31 1980 NRC Annual Report (NUREG-0774) 32 (1981-04-27) NRC Policy, Programmatic Environmental Impact Statement of the Cleanup 

of TMI-2 (46 FR 24764)
206

 

33 http://www.nrc.gov/about-nrc/policymaking.html (accessed July 28, 2014) 34 (1979-10-22) NRC Order (10-18-1979), Operation of EPICOR II to Decontaminate Rad 

Waste Water in Auxiliary Building

35 (1980-06-13) NRC Order (6-12-1980), Orders Release Krypton-85 from the Reactor 

Building Atmosphere by Controlled Purge 36 http://www.nrc.gov/about-nrc/policymaking.html (accessed July 28, 2014) 37 (1981-01-12) NRC Chairman to GPU, NRC Activities Required to be Performed by the 

TMI Licensee (re 09-12-1980) 38 (1979-07-20) NRC Order, Suspends Facility Operating License No. DPR-73 to Operate 

TMI-2 39 (1979-10-22) NRC Order (10-18-1979), Operation of EPICOR II to Decontaminate Rad 

Waste Water in Auxiliary Building

40 (1979-10-26) NRC Order Amend, Clarifies Order dated 10-26-1979 Concerning the 

Operation of EPICOR-II 41 (1981-03-25) OL Amend 13 42 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications 43 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 44 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications 45 NUREG-0432, TMI-2 Recovery Technical Specifications (1980-02-13) 46 NUREG-0647, Safety Evaluation and Environmental Assessment, TMI-2 (1980-02) 47 (1980-06-13) NRC Order (6-12-1980), Orders Release Krypton-85 from the Reactor 

Building Atmosphere by Controlled Purge 48 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building 

Sump and RCS Water

49 (1986-01-26) OL Amend 26 (Fully incorporates Recovery Mode Proposed Technical 

Specifications)

50 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications 51 (1980-01-14) NRC Review, Proposed Organization Plan (re 12-11-1979) 52 (1980-03-12) OL Amend 10 53 (1980-06-24) OL Amend 11 54 (1981-12-30) OL Amend 18 (concurrent with modification of Order) 55 (1986-01-26) OL Amend 26 (Fully incorporates Recovery Mode Proposed Technical 

Specifications)

56 (1988-05-25) OL Amend 30 57 (1989-09-11) OL Amend 35 58 (1993-09-14) OL Amend 45 (Possession Only License) 59 (1993-12-28) OL Amend 48 (TS for PDMS) 60 (1995-06-21) OL Amend 49 61 (1979-07-02) GPU, Planning Study for Containment Entry and Decontamination 62 (1979-08-13) GPU, Entry and Decontamination of Reactor Building, Highlights of a 

Planning Study

63 (1979-12-12) GPU, Summary Technical Plan for TMI-2 Decontamination and Defueling 64 SECY-80-175 (Interim criteria for radiological effluents from TMI-2 data gathering and 

Maintenance operations, April 4, 1980)

65 SECY-80-175-SRM (Staff Requirements – Discussion of Interim Criteria for Radiation 

Release at TMI, April 14, 1980) 66 SECY-80-105 (Release of Krypton Gas in Reactor Building Personnel Air Lock at TMI-2, 

February 20, 1980)

67 PNO-TMI-80-03 (Primary Water Leak, February 11, 1980) 68 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
207

 

69 (1980-02-26) Report of The Governor's Commission on TMI 70 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 71 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 72 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 73 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10) 74 NUREG-0683, SUPP. 2, PEIS, Disposal of Accident-Generated Water (1987-06) 75 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08) 76 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08) 77 1981 NRC Annual Report (NUREG-0920) 78 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 79 (1986-12-02) GPU, Post-Defueling Monitored Storage Plan 80 (1988-08-16) GPU, PDMS Proposed License Amendment and Safety Analysis Report (wo 

SAR) (re 12-02-1986) 81 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08) 82 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 83 (1985-10-14) GPU, Recovery Quality Assurance Plan, Rev. 4 84 (1980-01-24) GPU, TMI-2 Radiation Protection Plan, Rev. 1 85 (1984-10-30) GPU, General Project Design Criteria, Rev. 5 86 (1987-04-14) GPU, Special Nuclear Material Accountability Plan 87 (1985-10-17) NRC Exempt, 30.51, 40.61, 70.51(d), 70.53 Regarding Reporting Core 

Special Material

88 (1987-04-14) GPU, Special Nuclear Material Accountability Plan 89 (1985-06-06) GPU Procedure, Licensed Operator Requalification Program Unit 2, Rev. 3 90 (1985-05-29) GPU Procedure, Fuel Handling Senior Reactor Operator Training Program 

Unit 2, Rev. 1

91 (1984-02-10) GPU Procedure, Requirements for Certification of Candidates for NRC 

Operator License, Rev. 0

92 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 93 NUREGCR-1250V2PT3, TMI, A Report To The Commissioners And To The Public 

(Rogovin Report), Vol. II, Part 3

94 NUREGCR-1250V2PT3, TMI, A Report To The Commissioners And To The Public 

(Rogovin Report), Vol. II, Part 3

95 NUREGCR-6252, Lessons learned from the Three Mile Island Unit 2 Advisory Panel 

(1994)

96 (1990-03-00) Final Report, Three Mile Island Unit 2 Safety Advisory Board 97 (1981-12-09) NRC Review, Endorses Concept of Technical Assessment and Advisory 

Group

98 GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982) 99 GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982) 100 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980) 101 (1979-04-06) NRC (Internal), Staffing Report and Reorganization at TMI 102 (1980-03-31) NRC (Internal), Name Change of NRC TMI Organization to TMI Program 

Office

103 (1980-04-01) NRC, Change of NRC-TMI Organization 104 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980-03) 105 1981 NRC Annual Report (NUREG-0920) 106 NUREG-0698, Rev. 2, NRC Plan for Cleanup Operations at TMI-2 (1984-03)
208

 

107 1988 NRC Annual Report (NUREG-1145, Vol. 5) 108 (1979-05-15) GPU, Interim Report on TMI-2 Accident (re 04-11-1979) 109 NUREG-0432, Proposed Technical Specifications 110 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)) 111 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 112 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 113 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)) 114 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 115 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)) 116 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)) 117 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)) 118 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 119 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 120 DOE-ID-10276 121 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03)) 122 NUREG/CR-1250, Vol. 2, Part 2 (Rogovin Report) p.348 123 (1979-10-01) GPU, Water Storage Assessment 124 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical 

Specifications

125 (1980-04-04) NRC Review, A Further Evaluation of Risk of Recriticality at TMI-2 (re 03-

06-1980) 126 (1984-04-24) Order Amendment (also ROP-19) 127 (1984-07-17) Order Amendment (also ROP-20) 128 (1984-03-09) GPU Safety Evaluation, Head Removal, Rev. 5 129 (1984-07-17) NRC Safety Evaluation, Reactor Pressure Vessel Head Lift (re -3-09-1984) 130 (1985-04-23) Order Amendment (see ROP-27) 131 (1984-11-08) GPU, Criticality Report, Reactor Coolant System 132 (1985-03-15) NRC Safety Evaluation, Reactor Coolant System Criticality Report (re 11-

08-1984) 133 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 134 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 135 (1980-11-14) Order Amendment (also ROP-4) 136 GEND-056, TMI-2 Instrumentation and Electrical Program Final Evaluation Report 

(1986)

137 (1980-10-15) GPU, TMI-2 Quarterly Report No. 5 (Followup Report No. 8), Period 

Ending 09-30-1980 138 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 139 (1979-05-15) GPU, Interim Report on TMI-2 Accident (re 04-11-1979) 140 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
209

 

141 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), Period 

Ending 01-15-1980 142 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 143 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 144 (1979-05-15) GPU, Interim Report on TMI-2 Accident 145 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 146 (1979-10-15) GPU, TMI-2 Quarterly Report No. 1 (Followup Report No. 4), Period 

Ending 10-15-1979 147 (1979-05-15) GPU, Interim Report on TMI-2 Accident 148 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 149 (1979-05-15) GPU, Interim Report on TMI-2 Accident 150 (1980-03-04) GPU, BOP Diesel Generators 151 (1979-05-15) GPU, Interim Report on TMI-2 Accident 152 (1980-07-15) GPU, TMI-2 Quarterly Report No. 4 (Followup Report No. 7), Period 

Ending 06-30-1980 153 (1979-04-19) NRC, TMI-2 Ground Water Study.pdf 154 (1981-09-28) TMIPO Weekly 155 (2013-04-00) Exelon, TMI Annual Radiological Groundwater Protection Program Report 

for 2012

156 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 157 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), Period 

Ending 01-15-1980 158 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), Period 

Ending 01-15-1980 159 (1980-07-15) GPU, TMI-2 Quarterly Report No. 4 (Followup Report No. 7), Period 

Ending 06-30-1980 160 (1980-06-04) GPU, Plans to Discontinue Use of Supplemental Ventilation System Atop 

of Aux. Building

161 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 162 (1979-05-15) GPU, Interim Report on TMI-2 Accident 163 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 164 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), Period 

Ending 01-15-1980 165 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 166 (1980-08-11) Order Amendment (also ROP-3) 167 NUREG/CR-1250, Vol. 2, Part 2 (Rogovin Report) 168 (1979-10-01) GPU, Water Storage Assessment 169 (1979-04-11) NRC to FDAA, Request to Locate Railway Cars to Store Low Level Waste 

from TMI-2 170 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 171 NUREG-0774, U.S. NRC Annual Report: 1980
210

 

172 NUREG-0774, U.S. NRC Annual Report: 1980 173 (1980-06-13) NRC Order, Orders Release Krypton-85 from the Reactor Building 

Atmosphere by Controlled Purge

174 (1980-06-24) OL Amend 11 175 (1980-10-15) GPU, TMI-2 Quarterly Report No. 5 (Followup Report No. 8), Period 

Ending 09-30-1980 176 NUREG-0774, U.S. NRC Annual Report: 1980 177 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981-03) 178 Langenbach, J. W., Three Mile Island Unit 2 Reactor Building Entry Program,” Nuclear 

Technology, Vol. 87, No. 2, pp. 368-382, October 1989 179 1980 NRC Annual Report (NUREG-0774) 180 (1980-03-21) GPU, Initial Reactor Building Entry Program 181 (1980-07-15) GPU, Planned Reactor Building Entry (re 05-16-1980) 182 Langenbach, J. W., Three Mile Island Unit 2 Reactor Building Entry Program,” Nuclear 

Technology, Vol. 87, No. 2, pp. 368-382, October 1989 183 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 184 GEND-INF-011-VOL-3, REACTOR BUILDING BASEMENT RADIONUCLIDE 

DISTRIBUTION STUDIES (1983-06) 185 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building 

Sump and RCS Water

186 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 187 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup 188 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 189 (1979-07-02) GPU, Planning Study for Containment Entry and Decontamination 190 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building 

Sump and RCS Water

191 (1981-05-11) GPU System Description, SDS, Rev. 2 192 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 193 (1981-02-27) GPU, Cork Seal Contamination Report (re 01-19-1981) 194 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

195 (1994-06-02) GPU Response to NRC (Discussions), Actions Taken to Modify Cork Joint 

at TMI-2 196 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 197 GEND-INF-001, QUICK LOOK REPORT ENTRY 1 (1981) 198 GEND-INF-002, QUICK LOOK REPORT ENTRY 2 (1981) 199 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2, 

Occupational Radiation Dose (1984)

200 GEND-002, Vol. 1, Facility Decontamination Technology Workshop, Presentations 

(1979)

201 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 202 Merchant, D. J., “Worker Exposures During the Three Mile Island Unit 2 Recovery,” 

Nuclear Technology, Vol. 87, No. 4, pp. 1099-1105, December 1989 203 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE 

(1979-08) 204 1990 NRC Annual Report (NUREG-1145, Vol. 7) 205 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE 

(1979-08)
211

 

206 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE 

(1979-08) 207 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE 

(1979-08) 208 NUREG-0774, U.S. NRC Annual Report: 1980 209 Licensee Event Report LER-89-006; NRC Preliminary Notification PNO-I-89-81 210 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 211 International Atomic Energy Agency, “Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events,” IAEA-TECDOC-627, October 1991 212 International Atomic Energy Agency, “Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events,” IAEA-TECDOC-627, October 1991 213 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE 

(1979-08) 214 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE 

(1979-08) 215 NUREG-0640, TMI-2, Radiation Protection Program Report of the Special Panel (1979-

12)

216 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2, 

Occupational Radiation Dose (1984)

217 NUREG-1090, U.S. NRC Annual Report: 1983 218 NUREG-1145, Vol. 2, U.S. NRC Annual Report: 1985 219 NUREG-1145, Vol. 1, U.S. NRC Annual Report: 1984 220 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 221 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10) 222 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991 223 (1989-02-16) NRC IR 88-20 (See Encl. 2, TMI-2 Clean-Up Program Status Update) 224 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 225 EGG, Vol. 1, No. 1, TMI Unit-2 Technical Information and Examination Program, 

Update (4-15-1980) 226 (1980-03-26) Coordination Agreement, TMI Information and Examination Program 

(DOE-NRC-GPU-EPRI) 227 GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982) 228 (1984-04-16) DOE, Agreement with DOE for Japanese Participation in the TMI-2 

Research and Development

229 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.

230 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980) 231 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 232 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 233 NRC Preliminary Notification PNO-TMI-82-015 234 GEND-INF-038, Assessment of the TMI-2 Axial Power Shaping Rod Dynamic Test 

Results (1983-04) 235 NUREG-0998, U.S. NRC Annual Report: 1982 236 GEND-030, Quick Look Inspection Report on the Insertion of a Camera into the TMI-2 

Reactor Vessel (1983)
212

 

237 (1983-06-27) TMIPO Weekly 238 (1983-05-25) GPU Safety Analysis, Radiation Characterization Under the Reactor Vessel 

Head, Rev. 0

239 (1983-05-25) GPU Safety Analysis, Radiation Characterization Under the Reactor Vessel 

Head, Rev. 0

240 1983 NRC Annual Report (NUREG-1090) 241 (1983-08-15) TMIPO Weekly 242 (1983-08-22) TMIPO Weekly 243 (1983-09-12) TMIPO Weekly 244 (1983-10-24) TMIPO Weekly 245 (1983-09-12) TMIPO Weekly 246 (1983-07-20) GPU Safety Evaluation, Underhead Characterization, Core Sampling 

Addendum

247 (1983-09-19) TMIPO Weekly 248 (1983-10-03) TMIPO Weekly 249 EGG, Vol. 4, No. 1, TMI Unit-2 Technical Information and Examination Program, 

Update (12-15-1983) 250 EGG, Vol. 4, No. 2, TMI Unit-2 Technical Information and Examination Program, 

Update (06-15-1984) 251 Marley, A. W., “Sampling and Examination Methods Used for the Three Mile Island Unit 

2,” Nuclear Technology, Vol. 87, No. 4, pp. 845-856, December 1989 252 EGG, Vol. 4, No. 1, TMI Unit-2 Technical Information and Examination Program, 

Update (12-15-1983) 253 (1983-05-26) GPU Safety Evaluation, Underhead Characterization, Core Topography 

Addendum

254 NUREG-1145, Vol. 1, U.S. NRC Annual Report: 1984 255 EGG, Vol. 4, No. 2, TMI Unit-2 Technical Information and Examination Program, 

Update (06-15-1984) 256 NUREG-1145, Vol. 1, U.S. NRC Annual Report: 1984 257 (1984-04-02) TMIPO Weekly 258 EGG, Vol. 4, No. 2, TMI Unit-2 Technical Information and Examination Program, 

Update (06-15-1984) 259 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 260 EGG-TMI-7429, TMI-2 Lower Plenum Video Data Summary (1987) 261 (1985-01-03) GPU Safety Analysis, Measuring Uranium Fuel Material in Lower Reactor 

Vessel Region, Rev. 0

262 (1985-03-25) TMIPO Weekly 263 (1986-06-11) GPU Safety Evaluation, Core Stratification Sample Acquisition, Rev. 4 264 GEND-INF-092-VOL-1, TMI-2 CORE BORE EXAMINATIONS (1990) 265 NUREG-1145, Vol. 3, U.S. NRC Annual Report: 1986 266 GEND-INF-092-VOL-1, TMI-2 CORE BORE EXAMINATIONS (1990) 267 Marley, A. W., “Sampling and Examination Methods Used for the Three Mile Island Unit 

2,” Nuclear Technology, Vol. 87, No. 4, pp. 845-856, December 1989 268 NUREG-1145, Vol. 6, U.S. NRC Annual Report: 1989 269 NUREGCR-6197, TMI-2 Vessel Investigation Project Integration Report (1994-03) 270 NUREG-1145, Vol. 7, U.S. NRC Annual Report: 1990 271 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 272 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
213

 

273 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 274 GEND-001, GEND Planning Report (1980) 275 (1982-03-15) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear 

Waste from TMI-2 (Signed) 276 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS 277 NUREG-0591, Environmental Assessment, Use of EPICOR-II at TMI-2 278 (1979-10-22) NRC Order (10-18-1979), Operation of EPICOR II to Decontaminate Rad 

Waste Water in Auxiliary Building

279 (1980-02-29) City of Lancaster Settlement Agreement 280 (1980-03-26) Coordination Agreement, TMI Information and Examination Program 

(DOE-NRC-GPU-EPRI) 281 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981) 282 (1981-04-27) NRC Policy, Programmatic Environmental Impact Statement of the Cleanup 

of TMI-2 (46 FR 24764) 283 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building 

Sump and RCS Water

284 NUREG-0796, Operation of the Submerged Demineralizer System at TMI-2 (1981) 285 1981 NRC Annual Report (NUREG-0920) 286 (1980-03-26) Coordination Agreement, TMI Information and Examination Program 

(DOE-NRC-GPU-EPRI) 287 (1982-03-15) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear 

Waste from TMI-2 (Signed) 288 EPRI, “TMI-2 Waste Management Experience,” EPRI TR-100640, 1992 289 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign 

(1993)

2901987 NRC Annual Report (NUREG-1145, Vol. 4) 291 NUREG-0683, SUPP. 2, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2, 

Post-Defueling Storage and Cleanup (1987) 292 1988 NRC Annual Report (NUREG-1145, Vol. 5) 293 (1989-09-11) OL Amend 35 294 1991 NRC Annual Report (NUREG-1145, Vol. 8) 295 1993 NRC Annual Report (NUREG-1145, Vol. 10) 296 (1993-08-05) GPU, Accident Generated Water Disposal Completion Report 297 (1993-08-17) GPU, Completion of Accident Generated Water Disposal 298 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 299 NUREG-0591, Environmental Assessment, Use of EPICOR-II at TMI-2 300 NUREG-0591, Environmental Assessment, Use of EPICOR-II at TMI-2 301 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 302 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 303 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 304 (1987-05-14) GPU System Description, EPICOR II, Rev. 5 305 (1989-07-24) GPU System Description, EPICOR II, Rev. 7 306 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

307 GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water (1983-02) 308 GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water (1983-02) 309 GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water (1983) 310 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990
214

 

311 1988 NRC Annual Report (NUREG-1145, Vol. 5) 312 (1981-10-05) TMIPO Weekly 313 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 314 (1983-10-31) GPU System Description, Processed Water Storage and Recycle System, 

Rev. 0

315 (1980-01-24) GPU, Processed Water Storage Tank Dikes (re 12-13-1979) 316 (1980-02-01) NRC Review, Design Criteria for Processed Water Storage Tanks (re 01-24-

1980)

317 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

318 (1985-05-17) GPU Safety Evaluation, Operation of Internal Indexing Fixture Processing 

System, Rev. 3

319 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 320 (1989-05-12) GPU Technical Evaluation, Defueling Water Cleanup System, Rev. 12 321 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 322 Hofstetter, K. J., and B. S. Ausmus, “The Identification and Control of Microorganisms at 

Three Mile Island,” Nuclear Technology, Vol. 87, No. 4, pp. 814-823, December 1989. 323 1987 NRC Annual Report (NUREG-1145, Vol. 4) 324 (1986-05-06) NRC Review, Addition of a Biocide to the TMI-2 Reactor Coolant System 

(re 05-05-1986) 325 Hofstetter, K. J., and B. S. Ausmus, “The Identification and Control of Microorganisms at 

Three Mile Island,” Nuclear Technology, Vol. 87, No. 4, pp. 814-823, December 1989 326 (1986-07-09) NRC Review, Operation of Filter-Aid Feed System, Use of Diatomaceous 

Earth as Feed (re various letter)

327 (1987-04-29) GPU Technical Evaluation, Defueling Water Cleanup System, Rev. 10 328 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 329 (1987-06-11) NRC Review, DWCS Cross-Connect to Reactor Vessel Cleanup System (re 

05-21-1987) 330 (1989-05-12) GPU Technical Evaluation, Defueling Water Cleanup System, Rev. 12 331 (1986-11-17) GPU Safety Evaluation, TMI-2 Temporary Reactor Vessel Filtration 

System, Rev. 3

332 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 333 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 334 (1992-05-12) GPU Technical Evaluation, Processed Water Disposal System, Rev. 4 335 NUREG-0683, SUPP. 2, PEIS, Disposal of Accident-Generated Water (1987-06) 336 (1993-08-05) GPU, Accident Generated Water Disposal Completion Report 337 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981) 338 (1981-04-27) NRC Policy, Programmatic Environmental Impact Statement of the Cleanup 

of TMI-2 (46 FR 24764) 339 (1981-07-15) NRC-DOE MOU, Removal and Disposition of Solid Nuclear Waste from 

TMI-2 (Signed) 340 (1982-03-15) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear 

Waste from TMI-2 (Signed) 341 (1979-05-11) NRC, Design Criteria for Interim Storage or Solid Waste
215

 

342 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 343 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 344 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 345 (1993-06-29) GPU Technical Evaluation, Interim Solid Waste Staging Facility, Rev. 10 346 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 347 (1993-09-13) GPU System Description, Solid Waste Staging Facility, Rev. 6 348 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 349 (1981-08-17) TMIPO Weekly 350 (1992-07-29) GPU Technical Evaluation, Waste Handling and Packaging Facility, Rev. 8 351 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 352 EPRI-TR-100640, TMI-2 Waste Management Experience (1992) 353 NRC Backgrounder, Transportation of Spent Fuel and Radioactive Materials (2014) 354 NUREGBR-0216, Rev. 2, Radioactive Waste Production, Storage, Disposal (2002) 355 U.S. Code of Federal Regulations, “Licensing Requirements for Land Disposal of 

Radioactive Waste,” Part 61, Chapter I, Title 10, “Energy” (10 CFR 61) 356 (1983-05-11) NRC, Final Waste Classification and Waste Form Technical Position Papers 357 (1982-05-21) NRC, Certificate of Compliance 9152, Rev. 1, for Model CNS-1-13C II 

Shipping Cask

358 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 359 (1985-08-14) NRC, Certificate of Compliance 9152, Rev. 8, for Model CNS-1-13C II 

Shipping Cask

360 EGG-TMI-6304, Design Analysis, High-Integrity Container for Disposal of EPICOR-II 

Prefilter Liners (1983)

361 GEND-045, Disposal Demo of High Integrity Container with an EPICOR II Prefilter from 

TMI (1985)

362 (1984-01-24) NRC, Certificate of Compliance 5026, Rev. 10, for Model CNS-14-190 

Shipping Cask

363 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign 

(1993)

364 (1986-06-11) GPU Safety Evaluation, Canister Handling and Preparation for Shipment 

SER

365 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition 

that Apply to Fukushima (2012)

366 EGG-TMI-6304, Design Analysis, High-Integrity Container for Disposal of EPICOR-II 

Prefilter Liners (1983)

367 GEND-045, Disposal Demo of High Integrity Container with an EPICOR II Prefilter from 

TMI (1985)

368 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 369 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 370 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 371 GEND-032, TMI-2 Technical Information and Examination Program Annual Report 1982 

(1983-04) 372 GEND-INF-025, Development of a Prototype Gas Sampler for EPICOR II Prefilter Liners 

(1982-09) 373 GEND-029, Preparations to Ship EPICOR Liners (1983-06) 374 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03)
216

 

375 GEND-052, Hydrogen Control in Handling, Shipping, and Storage of Wet Radioactive 

Waste (1986-02) 376 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 377 GEND-062, Catalyst Tests for Hydrogen Control in Canisters of Wet Rad Waste (1987-

08).pdf

378 (1985-11-05) NRC Safety Evaluation, Defueling Canisters (re various letters) 379 GEND-051, Evaluation of Special Safety Issues Associated with Handling the TMI-2 

Core Debris (1985-06).pdf 380 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting 

from TMI-2 (1989) 381 Ruhter, P. E., W. G. Zurliene, “Radiological Conditions and Experiences in the Three 

Mile Island Unit 2 Auxiliary Building,” Nuclear Technology, Vol. 87, No. 2, pp. 361-366, 

October 1989

382 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting 

from TMI-2 (1989) 383 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2, 

Occupational Radiation Dose (1984)

384 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting 

from TMI-2 (1989) 385 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2, 

Occupational Radiation Dose (1984)

386 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 387 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

388 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting 

from TMI-2 (1989) 389 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes 

Resulting from TMI-2 (1981) 390 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

391 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting 

from TMI-2 (1989) 392 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report 393 (1993-11-12) GPU, Transition to PDMS 394 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

395 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

396 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2, 

Occupational Radiation Dose (1984)

397 (1989-01-10) GPU Safety Evaluation, Sediment Transfer and Processing Operations, Rev. 

4 

398 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 399 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992 400 (1985-03-25) GPU Safety Evaluation, Reactor Building Decontamination and Dose 

Reduction Activities for 1985, Supp. 1

401 (1985-09-25) NRC Safety Evaluation, Reactor Building Decontamination and Dose 

Reduction for 1985 (re 02-08, 03-25-1985) 402 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990
217

 

403 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 404 International Atomic Energy Agency, “Catalogue of methods, tools and techniques for 

recovery from fuel damage events,” IAEA-TECDOC-627, October 1991 405 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 406 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03) 407 Pavelek II, M.D., “Operations and Achievements of Remote Equipment at Three Mile 

Island,” Nuclear Technology, Vol. 87, No. 4, pp. 1122-1133, December 1989 408 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island 

Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989 409 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island 

Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989 410 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991 411 Pavelek II, M.D., “Operations and Achievements of Remote Equipment at Three Mile 

Island,” Nuclear Technology, Vol. 87, No. 4, pp. 1122-1133, December 1989 412 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island 

Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989 413 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, May 1992 414 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 415 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools (1988) 416 (1989-01-31) NRC Safety Evaluation, Criticality Assessment for Plasma Arc Torch to Cut 

Vessel Internals (08-11, 12-27-1988) 417 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island 

Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989 418 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991 419 GEND-002, Vol. 1, Facility Decontamination Technology Workshop, Presentations 

(1979)

420 GEND-034, Gross Decontamination Experiment Report (1983) 421 Lazo, E. N., The Three Mile Island Unit 2 Reactor Building Gross Decontamination 

Experiment: Effects on Loose Surface Contamination Levels,” Nuclear Technology, Vol. 87, 

No. 2, pp. 407-420, October 1989. 422 NUREGCR-3381, Evaluation of the TMI-2 Reactor Building Decontamination Process 

(1983)

423 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 424 Ruhter, P.E., and W.G. Zurliene, “Radiological Conditions and Experiences in the Three 

Mile Island Unit 2 Auxiliary Building,” Nuclear Technology, Vol. 87, No. 2, pp. 361-366, 

October 1989 425 1984 NRC Annual Report (NUREG-1145, Vol. 1) 426 (1984-07-02) TMIPO Weekly 427 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

428 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 429 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 430 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990
218

 

431 (1993-12-28) OL Amend 48 (TS for PDMS) 432 EGG-TMI-7489, TMI-2 Accident Scenario Update (1986) 433 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 434 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 435 Calhoun, G. L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology, 

Vol. 87, No. 3, pp. 587-594, November 1989 436 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 437 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 438 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 439 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 440 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 441 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 442 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 443 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 444 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 445 (1985-05-07) NRC Safety Evaluation, Lifting, Transfer, and Storage of Plenum Assembly 

(re various letters)

446 (1985-05-01) GPU Safety Evaluation, Plenum Removal and Transfer, Rev. 3 447 (1984-11-27) NRC Safety Evaluation, Initial Plenum Lift (re various letters) 448 (1984-03-09) GPU Safety Evaluation, Head Removal, Rev. 5 449 GEND-044, TMI-2 Reactor Vessel Head Removal (1985) 450 1984 NRC Annual Report (NUREG-1145, Vol. 1) 451 GEND-044, TMI-2 Reactor Vessel Head Removal (1985) 452 (1984-03-09) GPU Safety Evaluation, Head Removal, Rev. 5 453 (1984-07-09) TMIPO Weekly 454 GEND-044, TMI-2 Reactor Vessel Head Removal (1985) 455 GEND-044, TMI-2 Reactor Vessel Head Removal (1985) 456 GEND-044, TMI-2 Reactor Vessel Head Removal (1985) 457 GEND-054, TMI-2 Reactor Vessel Plenum Final Lift (1986) 458 (1984-02-17) NRC, Directors Decision Under 10 CFR 2.206 Regarding Pyrophoric 

Reaction of Zirconium Fines

459 (1984-02-17) NRC, Directors Decision Under 10 CFR 2.206 Regarding Pyrophoric 

Reaction of Zirconium Fines

460 Osgood, N.L., et al., “Review of Radiation Shielding Concerns Associated with the Three 

Mile Island Unit 2 Defueling Systems,” Nuclear Technology, Vol. 87, No. 2, pp. 556-56, 

October 1989 461 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology, 

Vol. 87, No. 3, pp. 587-594, November 1989 462 GEND-INF-065, TMI-2 Defueling System Design Description (1985) 463 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman) 464 Calhoun, G. L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology, 

Vol. 87, No. 3, pp. 587-594, November 1989 465 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 466 GEND-INF-073, TMI-2 Defueling Tools Engineering Report (1986-02) 467 (1986-02-06) NRC Review, Use of Hydraulic Cutoff Saw (re 02-03-1986) 468 (1986-02-03) GPU, Plan to Use Hydraulic Cutoff Saw for Sizing Operations in the 

Reactor Vessel

469 (1986-06-04) NRC Safety Evaluation, Heavy Duty Tong Tool, Spade Bucket, Hydraulic 

Impact Chisel (re various letters)

470 (1986-01-22) GPU, Plan to Use Hydraulic Impact Chisel to Separate Fused Material
219

 

471 (1986-04-16) NRC Review, Use of Hydraulic Shredder 472 (1986-03-26) GPU, Plan to Use Hydraulic Shredder to Reduce Core Debris 473 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 474 (1986-10-16) NRC Safety Evaluation, Core Bore Operations (re 09-18-1986) 475 (1986-09-19) GPU, Core Bore Operations 476 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools (1988) 477 (1988-03-25) NRC Review, Use of Core Bore Machine for Dismantling Lower Core 

Support Assembly (re various letters) 478 (1988-03-25) GPU, Use of Core Bore Machine for Dismantling the Lower Core Support 

Assembly Defueling

479 (1985-12-31) GPU Safety Evaluation, Core Stratification Sample Acquisition, Rev. 3 480 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985) 481 (1985-11-05) NRC Safety Evaluation, Defueling Canisters (re various letters) 482 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. 483 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985) 484 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology, 

Vol. 87, No. 3, pp. 587-594, November 1989 485 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985) 486 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985) 487 GEND-INF-065, TMI-2 Defueling System Design Description (1985-03) 488 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 489 (1986-01-06) TMIPO Weekly 490 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991 491 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991 492 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for 

Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991 493 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 494 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 495 (1986-05-29) GPU Technical Evaluation, Fuel Canister Storage Racks, Rev. 1 496 (1986-09-16) NRC Review, Fuel Canister Storage Racks TER (re various letters) 497 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 498 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 499 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 500 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 501 (1984-12-04) NRC Safety Evaluation, Containment Air Control Envelope Design (re 07-

22-1983, 10-30-1984) 502 (1984-10-30) GPU Technical Evaluation, Containment Air Control Envelope Design 

Criteria (Rev. 4), TER (Rev. 2)

503 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 504 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 505 (1985-10-29) NRC Safety Evaluation, Commencement of TMI-2 Preliminary Defueling 

Operations (re 10-24-1985) 506 (1985-10-24) GPU, Commencement of Preliminary Defueling Operations 507 (1985-11-04) TMIPO Weekly 508 (1985-11-12) NRC Safety Evaluation, Early Defueling of the TMI-2 Reactor Vessel (10-

10-1985)
220

 

509 (1985-10-10) GPU Safety Evaluation, Early Defueling of the TMI-2 Reactor Vessel, Rev. 

4 

510 (1985-11-18) TMIPO Weekly 511 (1986-01-13) TMIPO Weekly 512 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10 513 (1986-05-20) GPU Safety Evaluation, Defueling SER, Rev. 10 (Correction) 514 (1986-08-14) NRC Review, Defueling Safety Evaluation Report, Rev. 10 (re 05-15, 07-

24, 02-21-1986) 515 (1986-07-23) GPU, Plan to Use Core Bore Tooling as Defueling Tool 516 (1986-07-24) NRC Review, Core Region Bulk Defueling Operations Limited to the Core 

Region (re 05-15, 07-23-1986) 517 (1986-10-20) TMIPO Weekly 518 (1986-11-17) TMIPO Weekly 519 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 520 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 521 Porter, L.H., and E.E. Austin, “Disassembly and Defueling of the Three Mile Island 

Unit 2 Reactor Vessel Lower Core Support Assembly,” Nuclear Technology, Vol. 87, No. 3, 

pp. 587-594, November 1989 522 (1988-04-01) NRC Safety Evaluation, Lower Core Support Assembly Defueling (re 01-

18-1988) 523 (1988-01-18) GPU Safety Evaluation, Lower Core Support Assembly Defueling, Rev. 2 524 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 525 (1989-04-04) NRC Safety Evaluation, Completion of Upper Core Support Assembly 

Defueling (re 9-15-1988) 526 (1988-09-15) GPU Safety Evaluation, Completion of Upper Support Core Assembly 

Defueling, Rev. 0

527 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 528 (1988-12-01) NRC Safety Evaluation, Lower Core Support and Lower Head Defueling (re 

06-06-1988) 529 (1988-06-06) GPU Safety Evaluation, Completion of Lower Core Support and Lower 

Head Defueling, Rev. 0

530 (1988-06-27) GPU Safety Evaluation, Completion of Lower Core Support and Lower 

Head Defueling, Rev. 1

531 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 532 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 533 1990 NRC Annual Report (NUREG-1145, Vol. 7) 534 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 535 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 536 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 537 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 538 (1987-08-28) GPU Safety Analysis, Pressurizer Defueling System, Rev. 4 539 (1987-10-21) NRC Review, Pressurizer Defueling System, Rev. 4 (re 08-28, 10-13-1987) 540 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990 541 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 542 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 543 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 544 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,” 

EPRI-NP-6931, September 1990
221

 

545 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 546 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign 

(1993)

547 (1986-06-20) NRC Safety Evaluation, Canister Handling and Preparation for Shipment 

Program (re 02-17, 06-11-1986) 548 (1986-06-11) GPU Safety Evaluation, Canister Handling and Preparation for Shipment SER 549 (1986-06-11) GPU Safety Evaluation, Canister Handling and Preparation for Shipment SER 550 PNO-TMI-87-02 551 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign 

(1993-03) 552 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition 

that Apply to Fukushima (2012)

553 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition 

that Apply to Fukushima (2012-07) 554 DOE-ID-10739, Annual Radiological Environmental Monitoring Program Report for the 

TMI-2, ISFS Installation (2002-02) 555 (1986-12-02) GPU, Post-Defueling Monitored Storage Plan 556 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-

1992)

557 (1987-03-11) GPU, Environmental Evaluation for TMI-2 PDMS 558 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 559 (1986-12-02) GPU, Post-Defueling Monitored Storage Plan 560 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 561 (1993-09-14) Operating License Amendment 45 (Possession Only License) 562 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 563 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

564 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-

1992)

565 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 566 (1992-01-23) GPU, Addition of Borosilicate Glass to the Reactor Vessel 567 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 568 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 569 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 570 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 571 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 572 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 573 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 574 (1988-08-16) GPU, PDMS Proposed License Amendment and Safety Analysis Report 

(wo SAR) (re 12-02-1986) 575 (1990-02-22) GPU, Defueling Completion Report, Final Submittal 576 (1992-12-18) GPU, Criticality Safety Analysis for the Reactor Vessel 577 (1993-02-01) GPU, Post-Defueling Survey Report, Reactor Vessel 578 (1993-01-15) GPU, Proposed List of PDMS Requirements and Commitments (re 12-07-1992) 579 (1993-05-19) NRC Review, List of PDMS Requirements and Commitments (re 01-15-1993) 580 (1993-09-14) Operating License Amendment 45 (Possession Only License) 581 (1988-08-16) GPU, PDMS Proposed License Amendment and Safety Analysis Report 

(wo SAR) (re 12-02-1986) 582 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 583 (1994-11-04) NRC Review, Post-Defueling Survey Reports (PNL)
222

 

584 (1985-10-17) NRC Exempt, 30.51, 40.61, 70.51(d), 70.53 Regarding Reporting Core 

Special Material

585 (1993-02-01) GPU, Post-Defueling Survey Report, Reactor Vessel 586 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

587 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety 

Analysis Report

588 (1988-09-30) GPU, Post-Defueling Survey Report, Plenum and Letdown Coolers (letter) 589 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 590 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 591 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 592 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 593 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992) 594 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 595 (1993-09-14) Operating License Amendment 45 (Possession Only License) 596 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 597 1988 NRC Annual Report (NUREG-1145, Vol. 5) 598 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08) 599 1990 NRC Annual Report (NUREG-1145, Vol. 7) 600 (1990-04-20) GPU, Transition to Facility Mode 3 601 (1990-04-26) NRC Review, Transitioning Plant from Mode 2 to Mode 3 on 4-27-1990 (re 

04-12-1990) 602 1991 NRC Annual Report (NUREG-1145, Vol. 8) 603 1992 NRC Annual Report (NUREG-1145, Vol. 9) 604 1993 NRC Annual Report (NUREG-1145, Vol. 10) 605 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 606 (1992-10-26) GPU, Discontinuation of TMI-2 Restart Punchlist 607 (1993-02-01) GPU, Post-Defueling Survey Report, Reactor Vessel 608 1993 NRC Annual Report (NUREG-1145, Vol. 10) 609 (1993-09-14) Operating License Amendment 45 (Possession Only License) 610 1993 NRC Annual Report (NUREG-1145, Vol. 10) 611 (1993-07-06) NRC Safety Analysis, Reactor Vessel Criticality (re various letter) 612 (1993-09-14) Operating License Amendment 45 (Possession Only License) 613 (1993-09-14) Operating License Amendment 45 (Possession Only License) 614 (1993-11-12) GPU, Completed PDMS Requirements and Commitments (re various letters) 615 (1993-11-12) GPU, Transition to PDMS 616 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 617 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 618 (1993-12-28) Operating License Amendment 48 (TS for PDMS) 619 Berkeley CA: University of California Press, 2004. Dr. Walker joined the NRC History 

Program in 1979 and retired as the NRC historian in 2010.

620 University Park, PA: Penn State Press, 2004. Dr. Baratta was appointed the Associate 

Chief Administrative Judge (Technical) on the NRC Atomic Safety and Licensing Board 

Panel in 2003 and retired in 2014.

 
A loaded defueling canister being remotely lifted from the shipping 

cask at the INEL receiving facility. The 12.5-foot long canister was 

filled with water, placed in a storage module, and transferred to the 

storage pool for temporary storage at the INEL.




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 Gy

100 rad = *1 gray (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 factors

NUREG/KM-0001, Supplement 1 

June 2016