ML20083H996
| ML20083H996 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0683, NUREG-0683-S01, NUREG-0683-S01-DRFT, NUREG-683, NUREG-683-S1, NUREG-683-S1-DRFT, NUDOCS 8401090613 | |
| Download: ML20083H996 (50) | |
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k' NUREG-0683 Supplement No.1 Draft Report
- Programmatic Environmental Impact Statement related to decontamination arid disposal of radioactive' wastes-res'ulting from March 28,1979 accident Three Mile Island Nuclear. Station, Unit 2 Docket No. 50-320 Draft Supplement Dealing with
- Occupational Radiation Dose GPU ' Nuclear, Inc.
U.S. Nuclear Regulatory Commission n
- TMI Program. Office December 1983
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
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NUREG-0683 Supplement No.1 Draft Report Programmatic Environmental Impact Statement related to decontamination and disposal of radioactive wastes resulting from March 28,1979 accident Three Mile Island Nuclear Station, Unit 2 Docket No. 50-320 Draft Supplement Dealing with Occupational Radiation Dose GPU Nuclear, Inc.
U.S. Nuclear Regulatory Commission TMI Program Office Dscember 1983
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1 1
I COVER SHEET AND ABSTRACT 1.
Proposed Action and Location:
l DECONTAMINATION AND DISPOSAL OF RADIOACTIVE WASTES RESULTING FROM THE MARCH 28, 1979. ACCIDENT AT THREE MILE ISLAND NUCLEAR l
STATION, UNIT 2,
LOCATED IN LONDONDERRY TOWNSHIP, DAUPHIN j
COUNTY, PENNSYLVANIA 2.
Comments should be filed no later than 45 days after the date on which the Environmental Protection Agency's notice of availability of this draft supplement to the Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979, Accident Three Mile Island Nuclear Station, Unit 2 is published in the Federal Register.
3.
Further information may be obtained from Dr. Ronnie Lo, the Proj ect Manager for this supplement.
He may be contacted at the Three Mile Island Program Office, U.S.
Nuclear Regulatory Commission, Washington, DC 20555 or at 301-492-8335.
4.
In accordance with the National Environmental Policy Act, the Program-matic Environmental Impact Statement Related to Decontamination and Dis-posal of Radioactive Waste for the 1979 Accident at Three Mile Island Nuclear Station Unit 2 has been supplemented.
The supplement was re-quired because current information indicates that cleanup will entail substantially more occupational radiation dose to the cleanup work force than originally anticipated.
Cleanup was originally estimated'to result in from 2000 to 8000 person-rem of occupational radiation dose. Although only 1700 person-rem have 'resulted from cleanup operations performed up to now, current estimates now indicate that between 13,000 and 46,000 person-rem are expected to be required.
Alternative cleanup methods considered in the supplement either did not result in appreciable dose savings or were not known to be technically feasible.
SUMMARY
The Final Programmatic Environmental Impact Statement Related to Decon-tamination and Disposal of Waste Resulting from March 29, 1979, Accident Three Mile Island Nuclear Station, Unit 2 was issued by the U.S. Nuclear Regulatory Commission in March 1981. That document (referred to as the PEIS) stated that the mosc significant environmental impact of cleanup activities at Three Mile Island Unit 2 (TMI-2) would result from the radiation dose to the clear.up a
work force. The purpose of this supplement to the PEIS is to reevaluate the occupational radiation dose and resulting health effects from cleanup and to i
address additional alternative cleanup approaches using information gathered since the PEIS was prepared.
As a supplement to the PEIS, this document should be considered part of the earlier PEIS.
For completeness, reference to the PEIS should be made for all aspects of the NRC's National Environmental Policy Act review of the TMI-2 cleanup, other than the radiation exposures and resultant health effects which are the subject of this supplement.
I When the PEIS was prepared, it was believed that 2000 to 8000 person-rem of occupational radiation dose would be 1983,g during the decontamination incur and defaeling of TMI-2.
Through August about 1700 person-ren have been incurred in cleanup.
When the PEIS was prepared, the reactor building had been entered only five times.
Since then, it has been entered more than 280 times to collect data, conduct tests, perform decontamination tests and decontamination, refurbish the polar crane, remove trash and contaminated equipment, and begin preparation for fuel removal.
These entries have resulted in increased knowledge of the actual conditions in the building and awareness of the penetration of contamination into surfaces and the extent of corrosion, which have greatly. increased the difficulty of the cleanup task.
The temperatures reached during the accident and the time between the accident and the initiation of cleanup are thought ' to be factors in the decreased effectiveness of cleanup procedures.
Based on additional information available, decontamination workers at the plant are expected to receive a total collective radiation dose estimated at between,13,000 and 46,000 person-rem for the whole cleanup program. Using the internationally accepted health effect risk estimators, the NRC staff esti-mates that health effects corresponding to these doses will range from 2 to 6 additional deaths among these workers due to cancer and from 3 to 12 addi-tional genetic effects among their offspring. These ranges are broad because of uncertainties about the plant conditions and the amount of work that will be needed to decontaminate the reactor building and its contents. The occupa-tional dose to each worker will be limited to 3 rea/ quarter, with an average lifetime dose (past age 18) of 5 rea/ year in accordance with 10 CFR 20; the exact dose to any one individual cannot be predicted, but it is expected to be below the limits in all cases.
(a) In order to prepare this draf t supplement, a cutoff date of August 22, 1983, was established for data. The final PEIS Supplement I will include current data, which is not expected to change the conclusions given here.
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In accordance with the requirements of the National Environmental Policy Act, both the current cleanup plan and several alternative approaches were examined for their impact on occupational dose.
The current plan calls for a dose reduction effort prior to defueling of the reactor, with primary-system decontamination and final buildinh cleanup to follow defueling.
Only one of the three additional alternatives considered in the supplement would result in i
an appreciably lower occupational dose than that expected to result from the current plan, but significant disadvantages are associated with this alterna-tive, as discussed-below.
The first alternative considers using approximately the same task se-quence as that considered the most likely approach when the PEIS was origin-ally prepared, that is, extensive cleanup of the reactor building prior to defueling.
The purt ose of evaluating this alternative was to determine how 1
changing the work sequence from that of the current plan affects the occupa-tional radiation dose, given current information.
In evaluating this alterna-tive, it was determined that some reduction in dose, up to approximately 10%,
might be expected; however, the dose reduction is not considered sufficient to i
justify the delays in fuel removal.
Fuel removal delays are considered unde-sirable because the fuel continues to pose a potential risk to workers and the public and because information obtained from examining the fuel is expected to be useful in improving the safety of other nuclear power facilities.
The second alternative considers phased defueling followed by decontami-nation and building cleanup.
Phased defueling would involve removing fuel debris through the reactor pressure vessel head Safore removing the head and i
plenum.
This approach would minimize the possibility thct fuel fines would contaminate equipment and result in personnel exposure during later opera-j tions. However, no net savings in dose to workers would result because of the need for additional work.
This approach would delay fuel removal and all subsequent cleanup activities fo'r a minimum of 18 months.
The third alternative parallels the current plan through fuel removal, but then considers putting the reactor building, and possibly some of the more highly contaminated portions of the auxiliary and fuel-handling building, into a monitored, interim storage until additional decontamination activities could be performed robotica11y. This alternative, if found to be technically feas-ible, is expected to result in the lowest worker dose.
However, there are obstacles associated with this alternative, including uncertainty about when robotic technology will have evolved enough to be feasible for extensive use in completing cleanup; lack of information about the feasibility and safety of r
interim storage; and lack of assurance that funds will be available for ultimate cleanup.
These obstacles preclude the immediate adoption of this alternative; however, it may warrant further consideration nfter defueling is completed.
No decision is required on this alternative until after the fuel has been removed.
Although this supplement's estimate of the dose to the workers who per-form cleanup and the possible resulting health effects are higher than those estimated in the PEIS, it is still the conclusion of the staff, as it was when I
the PEIS was completed, that cleanup should proceed as expeditiously as pos-sible to reduce the potential for release of radioactive materials to the iv
envircsment and to ensure that TMI-2 does not become a long-term radioactive waste disposal site.
If the damaged fuel and radioactive wastes are not re-moved, the Island would, in effect, become a permanent waste disposal site.
The location, geology, and hydrology of Three Mile Island are a:nong the fac-tors that do not meet current criteria for a safe long-term vaste disposal facility.
Removing the damaged fuel and radioactive waste to storage sites that do meet all of the relevant criteria is the only reliable means for eliminating the long-term risk of widespread uncontrolled contamination of the environment by the accident wastes.
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FOREWOD This supplement to the Programmatic Environmental Impact Statement on the decontamination and disposal of waste from Three Mile Island Unit 2 (the PEIS) was prepared by the U.S. Nuclear Regulatory Commission, TMI Program Office, Office of Nuclear Reactor Regulation (the staff), pursuant to the Connaission's April 27, 1981, Statement of Policy related to the PEIS and the requirements of the National Environmental Policy Act of 1969 (NEPA).
Assistance was pro-vided by the Pacific Northwest Laboratory under the direction of the staff.
In the policy statement, the Connaission states that as the licensee pro-poses specific decontamination alternatives for each major cleanup activity, the staff will determine whether these proposals, and associated impacts that are predicted to occur, fall within the scope of those already assessed in the PEIS. The staff may act on each proposal if the proposed activity and asso-1 ciated environmental impacts fall within the scope of those assessed in the PEIS.
If an activity and its impacts fall outside of the scope of those in the PEIS, the staff shall complete necessary reviews in accordance with NEPA.
One of the conclusions of the PEIS was that the most significant envhon-mental impact associated with cleanup would result from the radiation doses received by the entire work force from cleanup activities.
At the time the PEIS was prepared, it was estimated that the cleanup would require 2000 to 8000 person-rem of occupational radiation dose.
Since the issuance of the PEIS (March 1981) and the Commission's Statement of Policy (April 1981), a substantial amount of new information about the conditions inside the reactor building has become available.
Based on the new information and the apparent decrease in decontamination effectiveness due to delays in initiating cleanup, the staff now believes that the total occupational dose to accomplish the entire cleanup could exceed the range predicted in the PEIS.
(To date, 1700 person-rem have been required.)
Therefore, this supplement to the PEIS has been prepared in compliance with NEPA requirements.
Information for the supplement was obtained from the licensee's Environ-mental Report and Pinal Safety Analysis Report (Metropolitan Edison Co.. and Jersey Central Power & Light Co. 1974), from the staff's Pinal Environmental Statement for the operating license (U.S. Nuclear Regulatory Commission 1976),
from the staff's PEIS of March 1981, and from new information provided by the i.
licensee or independently developed by the staff.
The staff met with the licensee to discuss items of information provided, to seek new information from the licensee that might be needed for an adequate assessment, and gen-erally to ensure that the staff had a thorough understanding of the cleanup operations.
In addition, the staff sought information from other sources that would assist in the evaluation, and visited and inspected the project site and vicinity.
On the basis of the foregoing and other such activities or inquiries ae were deemed useful and appropriate, the staff made an independent evaluation l
of the TMI-2 cleanup plans and operations and prepared this draft supplement i
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to the PEIS. The supplement is being circulated to federal, state, and local governmental agencies for comment.
A summary notice of the availability of the draft supplement has been published in the Federal Registe_r. T tion on which the supplement is based is available to the public,gej informa-and any comments on this information by interested persons will be considered by the staff.. Interested persons are also invited to comment on the draf t supplement.
The following federal and state agencies are being asked to comment on this draft supplement to the PEIS:
U.S. Army Corps of Engineers U.S. Environmental Protection Agency U.S. Department of Energy U.S. Department of Health and Human Services U.S. Department of' Labor U.S. Department of Interior U.S. Department of Interior, Geological Survey U.S. Department of Transportation U.S. Nuclear Regulatory Commission, Advisory Panel on TMI Cleanup Maryland Department of Natural Resources Maryland Department of State Planning New Jersey Department of Environmental Protection Pennsylvania Department of Environmental Resources Pennsylvania Department of Health Pennsylvania Department of Labor and Industry Pennsylvania Department of Public Welfare Penney 1vania State Clearing House Copies of this draf t supplement are also being sent to various local govern-ment agencies and members of the public.
After receipt and consideration of comments on the draft supplement, the staff will prepare a final supplement to the PEIS, which will include a-discussion of comments on the draft supplement and the responses to them.
Single copies of this supplement may be obtained by writing the Director, Division of Technical Information sud Document Control, U.S. Nuclear Regula-tory Commission, Washington, DC 20555.
Comments on the supplement should be addressed to:
Dr. Bernard J. Snyder Program Director, Three Mile Island Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Dr. Ronnie Lo is the Project Manager for this project.
He may be reached at the above address or at (301) 492-8335.
(a) NRC Public Document Room, 1717 H Street, Washington, DC 20555, and NRC TMI Program Office, 100 Brown Street, Middletown, PA 17057.
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CONTENTS
SUMMARY
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'. 1 FOREWORD
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1 1.1
1.0 INTRODUCTION
1.1 PURPOSE AND SCOPE
.T 1.1 h
1.2 HISTORY OF OCCUPATIONAL RADIATION DOSES R"SULTING FROM CLEANUP ACTIVITIES 1.2 1.3 REGULATORY AND ADMINISTRATIVE CONTROLS FOR LIMITING OCCUPATIONAL DOSE 1.6 2.0 CURRENT AND ALTERNATIVE PLANS FOR CLEANUP OF REACTOR AND AUXILIARY BUILDINGS.
2.1 2.1 ' BACKGROUND.INFORMATION ON CLEANUP WORK 2.1 2.1.1 Cleanup of the Reactor Building.innd Equipment 2.1 q.
2.1.2 Disassembly and Defueling of the Reactor 2.8
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2.1.3, Decontamination or the Priscry System.
2.11 2.1.4 Cleanup of the Auxiliary and Fuel-Handling Building 2.11 2.2 CURRENT CLEANUP PLAN: DOSE REDUCTION FOLLOWED BY DEFUELING s
AND DECONTAMINATION.
2.14 2.2.1 Tasks and Sequencing of the Current Cleanup Plan 2.15 2.2.1.1 Dose Reduction 2.15
.2.2.1.2 Reactor Disassembly and Defueling 2.17 r
2.2.1.3 Primary-System Decontamination.
2.20
.2.2.1.4 Reactor Building and Equipment Cleanup 2.20 2.2.1.5 Auxiliary and Fuel-Handling Building Cleanup 2.21 2.2.2,0ccuyational Radiation Dose Associated with 3
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. the Current Cleanup Plan 2.22
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2.3 ALTERNATIVE 1: EXTENSIVE CLEANUP POLLOWED BY DEFUELING 2.24 2.3.1 Tasks and Sequencing of Alternative 1.
2.24 2.3.1.1 Reactor Building and Equipment Cleanup 2.24 2.3.1.2 Reactor Disassembly and Defueling, and Primary-System Decontamination.
2.25 2.3.2 Occupational Radiation Dose Associated with Extensive Cleanup Followed by Defueling 2.25 2.4 ALTERNATIVE 2:
PHASED DEFUELING FOLLOWED BY REACTOR BUILDING CLEANUP 2.27 2.4.1 Tasks and Sequencing of Alternative 2.
2.27 2.4.1.1 Fines Removal Prior to Head Removal.
2.28 2.4.1.2 Reactor Disassembly and Defueling 2.28 2.4.1.3 Primary-System Decon*, amination, Auxiliary and Fuel-Handling Building Cleanup, and Reactor Building and Equipment Cleanup 2.29 2.4.2 Occupational Exposure Associated with Phased Defueling Followed by Reactor Building Cleanup 2.29 2.5 ALTERNATIVE 3: DEFUELING FOLLOWED BY DELAYED CLEANUP USING K0BOTICS 2.30 2.5.1 Tas s and Sequencing of Alternative 3.
2.31 2.5.1.1 Reactor Disassembly and Defueling 2.31 2.5.1.2 Interim Storage of the Defueled Reactor 2.31 2.5.1.3 Primary-System Decontamination.
2.31 2.5.1.4 Robotic Cleanup o# Reactor Building and Equipment 2.32 2.5.2 Occupational Radiation Dose Associated with Defueling Followed by Delayed Cleanup Using Robotics.
2.32 2.6 ANALYSIS OF CURRENT CLEANUP PLAN AND ALTERNATIVES.
2.34 2.6.1 Analysis of Public Safety 2.34 2.6.2 Analysis of Occupational Radiction Doea 2.35 X
2.6.3 Analysis of Time Schedule 2.36 2.6.4 Analysis of Technical Feasibility 2.41 2.6.5 Summary Analysis 2.41 3.0 REVISED ENVIRONMENTAL IMPACTS.
3.1 3.1 AFFECTED POPULATION.
3.1 3.2 REVISED OCCUPATIONAL-DOSE ESTIMATES 3.1 3.3 HEALTH EFFECTS.
3.3
4.0 CONCLUSION
S 4.1
5.0 REFERENCES
5.1 APPENDIX A - CONTRIBUTORS TO THE SUPPLEMENT A.1 APPENDIX B - HEALTH EFFECT ESTIMATORS B.1 xi
FIGURES 1.1 ' Doses at TMI-2 Compared with Doses Per Reactor at All Commercial Nuclear Plants in the United States 1.5 1.2 Average Dose Rates at Selected TMI-2 Locations 1.7 1.3 Number of Workers Versus Yearly Occupational Dose for TMI Units 1 and 2 1.9 2.1 Reactor Building 2.2 2.2 305-ft Elevation 2.3 2.3 347-ft Elevation 2.4 2.4 282-ft Elevation 2.5 2.5 Cross Section of 282-ft Elevation Showing Elevator Shaft 2.6 2.6 Cutaway View of TMI-2 Vessel 2.9 2.7 Plan View of Auxiliary and Fuel-Handling Building.
2.12 2.8 Cutaway View of Auxiliary and Fuel-Handling Building 2.13 2.9 Occupational Radiation Dose to Complete Cleanup 2.35 2.10 Conceptual Schedule for the Current Plan 2.37 2.11 Conceptual Schedule for Alternative 1 2.38 2.12 Conceptual Schedule for Alternative 2 2.39 2.13 Conceptual Schedule for Alternative 3 2.40 xiii
TABLES 1.1' Occupational Radiation Doses at TMI-2-from March 28, 1979, to August 22, 1983.
1.4 i
2.16 2.1 Licensee's Goals for Dose Rate Reduction
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2.2 Occupational Radiation Dose Estimates for the Current 2.22 Cleanup Plan 2.3 Estimated Occupational Radiation Dose for Extensive 2.26 Cleanup Followed by Defueling.
2.4 Estimated Occupational Radiation Dose for Phased Defueling 2.29 Followed by Reactor Building Cleanup 2.5 Estimated Occupational Radiation Dose for Defueling 2.33 Followed by Delayed Cleanup Using Robotics 2.42 2.6 Summary Evaluation of Cleanup Alternatives 3.1 Cumulative Occupational Radiation Dose Associated with Each 3.2 Cleanup Option.
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1.0 INTRODUCTION
1.1.
PURPOSE AND SCOPE In. March 1981, the Nuclear Regulatory Commission > (NRC) published the
' Final ' Programmatic - Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Waste Resulting from March 28, 1979, Accident Three - Mile Island - Nuclear Station, Unit 2 (NUREG-0683).
That document, referred to here as "the PElS," was intended to provide ~an overall evaluation
- of the environmental impacts that would result from cleanup activities at Three Mile Island. Unit 2 (TMI-2), beginning when the plant conditions. were stabilized af ter _' the accident; and continuing through the completion of
' cleanup.
The-purpose of this supplement is to reevaluate the impact of the radiation dose. to workers, - based on current information.
The objective of
" cleanup," as the term is ' used in that document. and this one, is decontami-nating and defueling the plant. The affected environment and the impacts that are not discussed here remain substantially as represented in the PEIS. - As a-supplement, this is not a stand-alone. document.
For completeness the. reader should refer to ther PEIS this document supplents.
Since the issuance of the PEIS, numerous activities (cleanup of accident-generated water, ' reactor and. auxiliary building decontamination, reactor underhead characterization, etc.) have been proposed by the licensee.
These activities were evaluated by the NRC staff and determined to fall within the
- scope of the activities assessed in the impact statement.
Their performance has resulted in considerable new information about conditions in the reactor building and in the auxiliary and fuel-handling building and about the effec-tiveness of various decontamination activities.
One conclusion of the PEIS was that the most significant environmental impact associated with the cleanup would result. from the radiation dose received by the entire work force from cleanup' activities. That collective dose was estimated to be in the range of
- 2000 to. 8000 person-rem.
Cleanup activities conducted through August 22, 1983, have resulted in approximately 1700 person-rem. ~Although this occupa-tional dose is still below the predicted range, there is substantial uncer-tainty about- ~ future occupational exposures, primarily because 'the most difficult work remains to be done and in certain areas dose rates have not declined as projected.
Based on cleanup experience to date at TMI-2, it now appears that 'the entire cleanup could result in doses in excess of the 8000 perron-rem previously estimated.
Therefore, this - supplement has been prepared.to. update the estimates of. radiation dose and assess the associated environme_ntal impacts.
The - doses for waste-related tasks that are used - in this. supplement have been taken. directly from the PEIS.
These ' doses are expected to - make.' only a - very. small contribution to the total dose from cleanup.
'This document, like the impact statement it supplements, is programmatic in unture.
That is, the action being considered is the assessment of the
- cleanup, which is subject to NRC approval.
In order to accurately predict the impact of lthe occupational radiation dose. from cleanup, the most probable
. sequences ~ and methods for cleanup are evaluated.
The most likely course of action, presented here as "the current cleanup plan," differs in sequence from 1.1 1
the most likely course of action at the time the-PEIS was prepared.
At that time, the licensee was planning to begin cleanup in the reactor building with an extensive decontamination of the tuilding and equipment. Although progress 4
has been made on building and equipment decontamination, a great deal of addi-tional work still remains. Rather than complete building and equipment decon-tamination before reactor disassembly and defueling as originally planned, the licensee has indicated his intention to remove the damaged reactor fuel as soon as possible.
Therefore, defueling prior to building cleanup is the pre-dominant feature of the current cleanup plan, which is presented and evaluated in Section 2.2 of this document.
I In accordance with the National Environmental Policy Act, alternative courses of action are considered in this document.
These alternatives were selected to be consistent with the conclusion of the PEIS that the TMI site is not suitable as a permanent repository for the accident-generated radioactive waste.
As discussed in Section 2.1.1 of the PEIS, a "no action" alternative, j
the option of not performing cleanup, would have the effect of converting the 1
reactor to a permanent repository.
Therefore, under all alternatives con-sidered, wastes would be removed from the site.
The alternatives were also selected to employ presently available technology, or, in one case, emerging robotic technology, to effect cleanup operations.
Within these two limita-tions, a wide range of cleanup alternatives is not available.
As a result, l
the alternatives considered here differ from each other and from the current cleanup plan primarily in individual task sequence and methodology.
1.2 HISTORY OF OCCUPATIONAL RADIATION DOSES RESULTING FROM CLEANUP ACTIVITIES Cleanup of the TMI-2 reactor building could not begin until af ter the inventory of the noble gas (krypton-85) had been vented.
Therefore, major work in the reactor building did not begin until the latter part of 1980.
When t:te PEIS was being prepared, the reactor building had been entered only five times since the accident (at a total dose of about 13 person-rem), and little specific information was available on the conditions in the building.
Dose estimates included in the PEIS were therefore based on limited data from i
the reactor building, some experience in the auxiliary and fuel-handling building, experience with previous reactor accidents, and certain necessary assumptions.
In addition, the dose estimates were based on the licensee's
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cleanup schedule as of 1980, which was not constrained by funding. Since that time, major delays in cleanup have resulted from lack of funds.
On the pre-vious bases, cleanup was estimated to require between 2000 and 8000 person-rem of occupational dose.
Since the PEIS was issued, the reactor building has been entered more than 280 times.
Entries now typically take place several times each Lweek and involve several workers performing a variety of tasks.
These-entries have provided a significant opportunity to gather information on the conditions in the building.
At the time the PEIS was prepared, it was estimated that "once the sump water has been removed, hot spots shielded, and general area decontamination completed, general area radiation levels should be reduced to 5 mR/hr or less" on the 347-ft elevation (PEIS Appendix 1).
This has not proved to be the The basement has been drained of highly radioactive water, and many hot case.
1.2 l
I-spots in the building have been shielded.
General-area decontamination was begun but was suspended when it was learned that little dose rate reduction was being achieved ind that cleaned areas were becoming recontaminated.
Workers on the 347-ft elevation currently average about 106 mres/hr (Flanigan 1983). Estimates of the effectiveness of water draining, decontamination, and shielding in other areas of the building were likewise overly optimistic.
Other factors are contributing to the diminir.hed effectiveness of cleanup activities.
Workers are still required to wear respiratory protection, which increases fatigue and decreases productivity.
The TMI experience has differed from past experience in the nuclear industry in that cleanup of the reactor building was not begun in-mediately.
During the intervening time, the humidity in the reactor building was 100%,
and it literally rained in containment. One result of the rain was that dose rates at initial entries were lower than expected because radionuclides had been rinsed downward.
A second result was that radionuclides permeated into porous surfaces such as concrete and were incorporated into corrosion layers as iron surfaces rusted.
The humidity in the reactor building is still high and contamination is still being spread through the air; thus, reeleaning of cleaned areas is still required, with concomitant exposure of workers.
Doses from both periodic maintenance work and repairc of breakdowns have also been and continue to be adversely affected by delay. Certain tacks, such as the testing and replacement of fire extinguishers, must be done periodic-ally whether or not any cleanup is in progrese.
Also, the longer cleanup activities are prolonged, the greater is the probability of failure of systems needed for cleanup, such as lighting and other electrical systems.
Experience with the cleanup thus far, coupled with the desirability of removing the damaged fuel as soon as possibic, has led the licensee to re-evaluate plans, strategies, and occupational doses.
On March 30, 1983, the licensee transmitted to the NRC its first formal estimate of the dose needed to complete cleanup (Kanga 1983). This estimate, 16,000 to 28,000 person-rem, was based on defueling as soon as possible and on the assumptions that.little, if any, difficulty would be experienced in plenum removal and that little, if any, concrete removal would be required.
Because the licensee's predicted doses were outside the range given in the PEIS and the assumptions did not appear overly conservative, the staff undertook to independently reassess the cleanup dose.
This supplement pre-sents the results of that reassessment.
The cleanup effort in the reactor building at TMI-2 has focused on the following activities to date:
o mapping of radiation levels, and air sampling e acquisition of data e decontamination of surfaces e placement of shielding e removal of sources of radiation exposure e processing of the sump water e refurbishment of the polar crane e assessment of the extent of core damage.
1.3
Table 1.1 listo the occupational radiation doses received by workers since the accident.
The doses are shown by activity and year, through 1982.
As of August 1983, approximately 1700 person-rem had been received at TMI-2 from the cleanup operation.
Figure 1.1 shows the doses at TMI-2 relative to doses at all commercial nuclear power reactors in the United States (Brooks 1982).
(Throughout this document, doses are rounded to two significant digits, and current doses include those incurred up to August 22, 1983.)
Doses at TMI-2 since the accident have been lower than the average doses experienced at operating reactors.
In 1981, the most recent year for which figures are available, the average collective dose at U.S. pressurized-water reactors (PWRs) was 652 person-rem per reactor (Brooks 1982).
The collective
-annual doses at TMI-2 since the accident were 490 person-rem in 1979 (some of this dose was incurred prior to the accident), 310 person-rem in 1980, 160 person-rem in 1981, 400 person-rem in 1982, and 340 person-rem in 1983 (to TABLE 1.1.
Occupational Radiation Doses at TMI-2 from March 28, 1979, to August 22, 1983 Dose (person-rem)
- 1979 1980 1981 1982 1983 Reactor Building Decontamination 0.040 12 66 44(b) 94(b)
Reactor Disassembly and Defueling 0.12 0
4.3 250(b) 110(b)
(including polar crane)
Reactor Coolant System and 0.22 0.048 3.1 4.9 2.1 Systems Decontamination Radioactive-Waste Management 84 31 26 20 22 Auxiliary and Fuel-Handling 50 13 2.6 16 18 Building Cleanup Maintenance, Safety, and 260 220 60 64 68 Sampling Unspecified 95 32 0.14 0.90 0
TOTAL 490 310 160 400 340 CUMULATIVE TOTAL 490 800 960 1400 1700 (a) From self-reading personnel dosimeters; all doses are rounded to 2 significant figures (Munson 1983).
(b) Several activities, such as polar crane cleanup and refurbishment, support both building cleanup and reactor disassembly and defueling.
1.4
(
4.000 MIDDLE 50% OF BWRs MIDDLE 50% OF PWRs
@ AVERAGE COLLECTIVE DOSE 3,3gg
@ MEDIAN COLLECTIVE DOSE
$ TMI 2 COLLECTIVE DOSE g TMI DOSES FOLLOWING THE ACCIDENT 3.000 RANGE OF DOSES E
2 6
8 2.500 t
1 a
8 0
5 l
2.000 r
g 8
9 g
1.500 "i
8u l
1.000 4
500 eN 2
e g
-I..
I
-l
~
l i
g 1979 1980 1981 1982 1983 TO DATE YEAR FIGURE 1.1.
Doses at TMI-2 Compared with Doses Per Reactor at All Commercial Nuclear Plants in the United States 1.5
August 22). The average dose per worker was also lower. Workers who received measurableJradiation exposure in U.S. PWRs received an average.of 0.61 rem in 1981.
At TMI Units 1 and 2,7a comparable group of workers averaged 0.23 res/ person in 1979, 0.11 rem / person in 1980, 0.16 res/ person in 1981, and 0.45 rea/ person in 1982.
This data was readily. available only for Units 1 and 2 together.
(In each of these years except.1979, more dose was accumu-lated at Unit 1 than at Unit-2.)
Work on large-scale operations - i~n the reactor building that are both labor-intensive and occupational-exposure-intensive is now beginning or is planned for the near future..The primary operations include:
e placement of radiation shields e removal of the pressure vessel head
.o removal of the plenum e removal of fuel and fuel debris e hands-on decontamination.
Because of the increasing amount of work being done in the reactor build-ing, a major effort to reduce dose. rates was initiated by the NRC and the
. licensee in late 1982. The objective was to identify and eliminate or shield as many sources of radiation exposure as possible. The dose reduction program has been focused on the 305-f t and 347-f t elevations of the reactor building because~this is where most of the cleanup and defueling work will take place in the'near future. This effort has shown some significant results, as can be seen in Figure 1.2 and as discussed further in Section 2.1.
1.3 REGULATORY AND ADMINISTRATIVE CONTROLS FOR LIMITING OCCUPATIONAL DOSE
~Before.any cleanup activity at TMI-2 is initiated, the NRC. staff performs an extensive review of the licensee's technical evaluation report, written procedures, safety analyses, and other documentation governing the work to be performed.
Permission for an activity. to begin is ' granted only when the NRC staff has determined that the following conditions are met:
e safety standards are maintained e-the activity is consistent with the TMI-2 operating license and technical
~ specifications e the activity. does not violate NRC radiation protection regulations, including - the requirement-to maintain radiation doses as low as is reasonably achievable (ALARA) e the activity and associated impacts fall within the scope of the PEIS.
Regulations governing occupational exposure to radiation for all NRC
' licensees, including ~ the TMI-2 licensee, are 'given in Title 10 Part 20, of
=the U.S. Code of Federal Regulations (10 CFR 20).
Two types of requirements to protect worke.s against radiation are set forth in 10 CFR 20.
The'first requirement (10 CI'R 20.101) sets numerical limits on the amount of radiation a 1.6
305' ELEVATION 0.5
- 347' ELEVATION TOP OF REACTOR SERVICE STRUCTURE z
5a a5 x
0.4 y
ec
$$2 z25
- ok O.3
<c2 1
.,,2 w w we m zv oz oo w@
0.2 0 S' 9!!
4m' O.1 0.0 FALL FALL SUMMER FALL EARLY SUMMER 1980 1981 1982 1982 1983 1983 FIGURE 1.2.
A.erage Dose Rates at Selected TMI-2 Locations worker may be exposed to in any calendar quarter.
The limit for whole-body external radiation is 1.25 rem (special limits apply to extremities--see 10 CFR 20.101; in any one ruarter unless certain requirements regarding individual lifetime dose limits and dose records are met, in which case the limit is 3 rem per calendar quarter.
The second requirement deals with the fundamental approach to radiation protection.
The principle of maintaining radiation exposures ALARA has long been a basic goal of radiation protection programs, and 10 CFR 20.1(c) requires that NRC licensees follow this principle.
The basic ALARA objective is to ensure that radiation exposures are kept to the lowest levels that are commen-surate with sound economic and operating practices.
The Nuclear Regulatory Commission's Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable," (NRC 1978) expands on the elements of an effective 1.7 f
ALARA program. These elements are an integral part of the TMI-2 program, and include:
- 1) upper-level management responsibility and authority for the ALARA program; 2) appropriate training and instruction for those at all organiza-tional levels who are involved in radiation work; 3) review of the design of new and modified equipment to ensure that the selection of equipment will minimize occupational radiation exposure; 4) control of access to radiation areas; 5) appropriate use of shielding; 6) extensive review of procedures, job prepara tion, and planning to minimize the dose required to perform specific tasks; and
- 7) adequate protective equipment and personnel-monitoring instrumentation.
To promote ALARA and comply with the dose limits, the licensee has estab-lished administrative radiation dose limits for workers. These administrative limits require management approval for all doses in excess of 1 rem / quarter.
(Successively higher doses, up to the regulatory limits, require authorization from successively higher levels of management (GPU Nuclear 1983).
These administrative limits are set below the regulatory limits to ensure that no worker.will be exposed to radiation in excess of the regulations.
Since the accident, the maximum annual radiation doses received by workers at TMI-2 (not necessarily the same person each year) have been 4.5 rem in 1979; 2.1 rem in 1980; 2.1 rem in 1981; and 3.0 rem in 1982.
(See Figure 1.3 for the number of workers versus the yearly occupational dose at TMI since the accident.)
In addition, all operations planned at TMI-2 undergo review by the licensee's health physics and radiological engineering staff to ensure that the work is conducted in accordance with the ALARA principle.
An important part of the NRC's review and approval of cleanup activities is to independently determine that the proposed work will be carried out following good ALARA practices.
The sections that follow deal with the work to be done and alternative approaches to it (Section 2); the most important impact of cleanup, occupa-tional radiation dose (Section 3); and the conclusions reached in preparing this supplement (Section 4).
1.8
sooo W
5400 D
s, en E
2400 w
a:o3 u.O g
18oo m2oz e
1200 600 m
}
I%
f"""I O
O.g,
7.q,G %
%.q>g %g,? Q Q
9 QS 0
OCCUPATIONAL DOSE IN A YEAR (REM)
FIGURE 1.3.
Number of Workers Versus Yearly Occupational Dose for TMI Units 1 and 2 l
1.9
2.0 CURRENT AND ALTERNATIVE PLANS FOR CLEANUP OF REACTOR AND AUXILIARY BUILDINGS Chapters 5 and 6 of the PEIS address in some detail the tasks to be accomplished for cleanup of the reactor building and the auxiliary and fuel-handling building, disassembly and defueling of the reactor, and decon-tamination of the primary system.
These tasks are briefly presented in Sec-tion 2.1 to provide an appropriate background for the descriptions of the plan and alternatives that follow.
Section 2.1 also reflects current knowledge of the tasks to be performed and the methods available to carry them out.
Sec-tion 2.2 presents the licensee's current plan for cleanup, which was evaluated in preparing this supplement.
Three alternatives developed by the NRC staff are also presented.
Alternative 1, discussed in Section 2.3, is an approach similar to that evaluated in the original PEIS, that is, cleaning the building to reduce the dose rate.to 10 mrem /hr or less prior to defueling.
Alterna-tive 2, discussed in Section 2.4, is the removal of fuel fines and particles through the reactor pressure vessel head before head removal.
Alternative 3, discussed in Section 2.5, involves putting the reactor into a monitored interim-care mode af ter defueling until the high-dose work of building cleanup can be performed robotically.
The plan and alte rnatives are compared and evaluated in Section 2.6.
2.1 BACKGROUND
INFORMATION ON CLEANUP WORK Cleanup work to be performed in the reactor building can be subdivided into three principal endeavors:
- 1) cleanup of the reactor building and equip-ment; 2) disassembly and defueling of the reactor; and 3) decontamination of the primary system.
The first two of these may be performed in any sequence or simultaneously.
The third must follow defueling.
It is tha. variation in sequence that is the primary difference between the current plan and the first alternative.
The second and third alternatives utilize slightly different methods of performing the work.
Cleanup of the auxiliary and fuel-handling building is already underway and, under the current plan and Alternatives 1 and 2, would be completed as resources are available.
Under Alternative 3, those portions of the auxiliary and fuel-handling building cleanup that require the greatest dose might be postponed until additional technology is developed.
The physical and radiological conditions that affect these endeavors are discussed briefly below, followed by a description of the tasks involved in each phase of cleanup.
2.1.1 Cleanup of the Reactor Building and Equipment The reactor buildine i= ; ylindrical reinforced-concrete structure with a dome top, as illustrated in Figure 2.1.
Levels within the building are referred to by elevation above sea level.
The building is entered at the 305-ft elevation.
When the building was first entered af ter the accident, radiation dose rates at this elevation averaged 430 mrem /hr. The placement of shielding, the removal of debris, and decontamination of the building have 2.1 l
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Reactor Building l'
2.2
reduced dose rates at this level to an average of approximately 140 mrem /hr in mid-1983.
The dose rate for normal operation, and the target for the total cleanup effort, is 10 mrem /hr (Kanga 1983).
Because radiation sources are distributed throughout the building and are difficult to remove, reducing the dose rate below the current level is expected to require greater effort than that required so far.
A plan view of the 305-ft elevation is shown in Figure 2.2.
Above the 305-f t elevation is the 347-f t elevation, which is reached by an open stairway.
(An elevator and an enclosed stairwell are also present; however, radiation dose rates resulting from the accident have prevented refurbishment of the elevator and minimized use of the stairwell.) The 347-ft elevation is used to gain access to the reactor vessel head and service structure, the fuel transfer canal, and other areas important for reactor disassembly and defueling. Dose rates at the 347-ft elevation averaged REACTOR PERSONNEL COOLANT AIR LOCF.
PUMP ELEVATOR GENERATOR ENCLOSED REACTOR STAIRWELL COOLANI PUMP
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305-ft Elevation 2.3
240 mrem /hr following the accident.
Shielding, debris removal, and dacon-tamination have reduced the average dose rates to approximately 110 mrem /hr in the summer of 1983.
The target dose rate for cleanup of the 347-ft elevation is 10 mrem /hr. A plan view of this elevation is shown in Figure 2.3.
The polar crane, located at the 426-ft elevation, is reached by ladder or hoist from the 347-ft elevation.
(The elevation of the crane's cab is 418 ft, 6 in.)
The polar crane, shown in Figure 2.1, is necessary for numerous activities in support of disassembly and defueling, and will also facilitate the transportation of decontamination equipment, directional radiation measuring devices, and shielding materials within the building. Dose rates at initial access to the polar crane averaged 120 mrem /hr, but through con-siderable work to decontaminate and prepare the crane for use, the dose rates have been reduced to about 80 mrem /hr.
Below the 305-ft entry level elevation is the 282-ft elevation, or basement, shown in Figures 2.4 and 2.5.
The 282-ft elevation contains large numbers of reactor control cables, various pumps and piping systems, the reactor coolant drain tank (in a shielded cubicle), and other equipment. This area contained accident-generated water to a depth of about 8 feet when the building was initially entered after the accident. Since that time, the water has been drained, purified, and recycled for use in decontamination.
Water frem decontamination efforts on the upper levels has flowed into the basement, ELEVATOR ENCLOSED
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COVERED HATCH
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STORAGE OPEN PRESSURIZER STAND STAIRWELL FIGURE 2.3.
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282-ft Elevation dissolving additional contamination in the basement and then removing it as the water was pumped out.
However, the numerous structures and pieces of equipment at this level (see Figure 2.4) make cleanup particularly difficult, and the area remains. highly contaminated, with dose rates in the range of 1 to 1000 rem /hr, depending on location and distance from the floor.
Although samples have been collected, no entries have been made.
The highest measured radiation levels at the 282-ft elevation are in the vicinity of the elevator shaft and inclosed stairwell.
These structures, which are made of hollow concrete blocks, became saturated with the accident' water and absorbed radionuclides from it.
The bottom of the elevator shaft is an enclosed area that until recently contained highly radioactive water.
Radiation from the contaminants in the elevator and enclosed-stairwell area of the 282-ft elevation have prevented use of the stairwell and elevator at upper levels as well.
2.5
//////////b f, ELEVATOR CAR /
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////////////:
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NOT TO SCALE FIGURE 2.5.
Cross Section of 282-ft Elevation Showing Elevator Shaft Because the accident-generated water remained in the reactor building for several years, radionuclides concentrated on vertical surfaces at the water surface level.
This phenomenon, commonly ref erred to as "the bathtub ring,"
continues to affect dose rates on the 282-ft elevation.
Efforts to remove the ring by spraying from above have not been successful in reducing general-area dose rates.
Although some chemicals may have a positive effect, it is expected that decontamination of concrete areas will require removal of the surface coating and some of the concrete.
The reactor building aump is also expected to be highly contaminated.
The sump is inaccessible for dose rate measurement and has not yet been sampled.
There is a thin layer of sludge on the floor of the 282-ft elevation.
The sludge was originally thought to be a major contributor to dose rates; however, it appears that radionuclides from the slut'.ge have leached into decontamination water and have then been removed.
The radiation contribution from the sludge, therefore, is expected to be comparatively small.
The cleanup of the reactor building will entail:
the removal of miscel-laneous equipment and debris that were in the building at the time of the accident (ladders, scaffolding, tools, etc.); the decontamination or removal of reactor-associated equipment (air coolers, cable trays, r'eac to r piping, 2.6
l etc.); the decontamination of building surfaces (both metal and concrete); and various support activities to ensure the safety of workers performing these tasks - and to measure the effectiveness of the cleanup activities.
Cleanup activities in the reactor building have been underway for several years and are continuing.
Considerable debris and equipment have been removed from the 305-ft and 347-ft elevations, and decontaminacion of the building and remain-ing equipment has been attempted on these elevations. Some remote flushing of the 282-ft elevation has been performed. Although decontamination using high-and low-pressure sprays of corated water has reduced the level of smearable contamination on equipment and building surfaces, these techniones have been of limited success in reducing general-area dose rates.
Effective, although
- temporary, dose rate reduction has been achieved by the shielding of certain sources of high-level radiation, including the elevator shaft and stairwell on the 305-ft elevation and certain floor drains.
(Shielding is considered only a temporary measure because final building cleanup will require the elimina-tion of these sources.)
Most tasks involved in the reactor building decontamination, reactor dis-assembly and defueling, and primary-system decontamination can be done without access to the 282-ft elevation; therefore, cleanup of this area will be left until the later stages of the cleanup operation in all options. However, the water being used for building decontamination is apparently continuing to leach radionuclides from sources on this elevation; hence, it is undergoing some continual decontamination.
To the extent possible, preliminary decon-tamination of the 282-f t elevation will be performed remotely or semi-remotely from the 305-ft elevation.
This task will include remote radiation surveys and video examination, water and/or chemical spraying from above through penetrations, and possibly the use of robota for cleaning and removing equip-ment.
When dose rates permit, hands-on decontamination techniques such as those used in the remainder of the building will be employed.
The ultimate cleanup objective for the 282-ft elevation is 10 mrem /hr.
s Since the accident, the level of airborne radioactive material has neces-sitated the wearing of respirators for all activities in the building.
(Air-borne-radionuclide concentrations during work in the building are currently about 6 times the allowable concentration for a 40-hr/wk exposure without respiratory protection (Flanigan 1983).)
These respirators, while protecting the workers, very significantly reduce productivity and hamper mobility.
In addition, in some areas the airborne radioactive material has redeposited on cleaned surfaces, making decontamination only temporarily effective.
The problems of airborne contamination and redeposition appear to be, at least partially, the result of radioactive material associated with boric acid crystals in the gir (Alvarez 1983).
Boric acid comes from the primary coolant and, most importantly, from the decontamination solutions used in the building since the accident.
(The solutions have been made from recycled accident-generated water that has been purified by a selective ion-exchange treatment that removes radionuclides but not boric acid.)
The removal of boric acid from decontamination water is currently being investigated by the licensee.
The principal radionuclides that were identified in the PEIS (pp. 5-26,
- 27) and reconfirmed by subsequent measurements are cesium-137, cesium-134, and strontium-90.
Cesium-137 has a 30-year half-life and is expected to be a A
2.7
major source of whole-body dose throughout cleanup.
Cesium-134 has a 2.0-year half-life and has therefore diminished to about 25% of the accident inventory in the first 4 years following the accident.
Its contribution to whole-body dose rates will continue to decrease.
Strontium-90 has a 28-year half-life.
Therefore, it has decayed very little since the accident.
It is, however, a beta-emitting radionuclide, which means that protective clothing of fers sub-stantial worker protection.
This mix of radionuclides is markedly different from that of other reactors, where these radionuclides are contained within the core.
In those cases, cobalt-58 (71-day half-life) and cobalt-60 (5.3-year half-life) are the principal sources of worker dose, and the dose rate to which workers are exposed can be halved by waiting 5.3 years. At TMI, the same halving of dose rate requires 30 years.
2.1.2 Disassembly and Defueling of the Reactor A cutaway view of a typical pressure vessel for a PWR is shown in Fig-ure 6.1 of the PEIS.
This drawing has been modified, as shown in Figure 2.6 of this report, to show the results of work in progress and what has been learned about the TMI-2 vessel and its contents by video camera examiaation and other exploratory techniques.
Proceeding from top to bottom of the reactor pressure vessel, the conditions are as follows.
Three of the lead screws that were previously attached to control rod drives have been uncoupled and removed to allow examination of the core and internals.
A complete control rod drive assembly has been removed for further examination of the
{
reactor vessel and internals and for characterization of the radiological conditions under the head.
All of the remaining lead screws have been uncoupled.
The upper plenum assembly, the device that positions the control rods in the core, appears to be relatively undamaged.
Clearance between the pressore vessel and plenum is only 50 mils (50 thousandths of an inch), so the ease of plenum removal is still open to question as the plenum may be warped.
There are portions of damaged fuel assemblies adhering to the underside of the plenum.
Beneath the plenum is a 5-foot-deep void wherc fuel and control rods used to be.
At the bottom of the void is a bed of loose rubble to a depth of at least 14 inches.
The Debris Defueling Working Group has estimated (Runion 1983) that there are approximately 45,000 kg (100,000 pounds) of rubble and fines in the TMI-2 reactor core that are 25,000 pm (1 inch) or less in size.
These estimates indicate that 5300 kg are 800 pm or less and 125 kg are 4 um or less.
The conditions below the rubble are not known.
Material may be
' loose or may have been fused by melted nonfuel material.
The lower support structures may be intact or warped.
Fuel may have been deposited in the lower areas of the reactor vessel below the lower support structure.
The tasks to be performed for reactor disassembly and defueling include:
e visual and radiological characterization of the core and the reactor pressure vessel head e preparation for head lift e lifting and storage of the head and installation of the reactor internals indexing fixture 2.8
l k{il d.{l CONTROL ROD
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CONTAINING LE AD L
REACTOR PRESSURE
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SCREWS (3 LEAD e
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DAMAGE APPARENT (65 TONS)
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PLENUM. TYPICAL
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STRUCTURE (112.5 TONS)
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_ hill l FLOW DISTRIBUTOR-I 1
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CONDITION UNKNOWN E l l '*.
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FIGURE 2.6.
Cutaway View of 711-2 Vessel 2.9
e installation of water cleanup systems for the reactor vessel and fuel transfer canal e refurbishme.nt and modification of the fuel-handling system e removal of the plenum o removal of the fuel
- removal of the core support structure and lower internals.
Initial visual and radiological. characterizations of the reactor vessel and core have been accomplished.
Additional underhead characterization, including dose rate measurements, visual inspection (using closed-circuit television), core topography, and water and debris sampling, is in progress.
Preparations for head lift are in progress.
The uncoupling of the 63 lead screws has been. completed.
Handling of the lead screws is important because experience with those removed so far indicates that they may be a significant source of radiation exposure to the workers. A test is planned to measure the radiation contribution from parked lead screws so that it can be determined whether they should be removed and shielded prior to head lift.
Other preparations necessary for head lift include disconnecting and removing cooling and electrical lines and overhead platforms, detensioning and parking (or removing) hend studs and nuts, refurbishing and installing the seal plate, and refurbiching and attaching the hoisting equipment.
The head will be lifted and stored away from the work area.
It may be a significant source of exposure, particularly if the lead screws are parked, and may therefore require special precautions.
Once the head is removed, the condition of the plenum will be further assessed.
There are plans to prcvide water shielding over the plenum by placing a water-filled cylinder, such as the internals indexing fixture, over it, or by filling the fuel transfer canal.
One or more water cleanup systems will be installed to treat the reactor vessel and fuel canal water. These will be located in the fuel transfer canal to use, canal water as shielding.
Bccause of particulate and dissolved radio-nuclides in the primary coolant, cleanup of any water in contact with the reactor core will be important for dose reduction and the control of airborne contamination.
Plans call for refurbishing and modifying the fuel-handling system to accept fuel canisters.
The plenum will be removed intact or, if necessary, in pieces and stored underwater to provide radiation shielding.
Loose, particulate fuel debris will be removed, follov9d by larger fuel pieces.
Fuel is normally handled underwater for radiation shielding.
When the fuel is removed, it will be placed in canisters in the water-filled fuel transfer canal.
These canisters will be tipped horizontally by the modified fuel transfer equipment and passed through the fuel transfer tube into a fuel storage pool in the auxiliary and fuel-handling building.
Once most of the fuel has been removed, the core support structure and lower reactor internals will be removed (intact if possible, otherwise in pieces) and any remaining fuel particles will be removed.
2.10
)
It is. not certain what effort, if any, will be made to mechanically remove fuel particles from the reactor piping system. Any particles that have been swept into the outlet nozzles of the reactor. vessel may be accessible to defueling equipment through the reactor nozzles once the reactor internals are removed.
Once all the fuel accessible through the reactor vesse) has been removed, defueling will be complete and the transfer canal will be drained and decontaminated. Then primary-system decontamination can begin.
2.1.3 Decontamination of the Primary System Directional radiation surveys indicate that reactor fuel and/or fission products are dispersad throughout the primary piping system as finely divided particles and/or as plating on surfaces.
This material must be removed as part of the cleanup.
Section 6.5 of the PEIS contains a discussion of primary-system decontamination.
Since the completion of the PEIS, the Electric Power Research Institute (EPRI) has funded research into the probable distribution of radionuclides in the primary system (Crnane and Nicolosi 1983 and Daniel et al. 1983) and into physical and chemical methods available for decontamination (Card 1983, Sejvar and Dawson 1983, Gardner et al. 1983, and Muns:- et al. 1983). Although information about the distribution and removal of contamination has thus been gained, there is little additional definitive information on which to base a
task description for primary-system decontamination.
Decontamination solutions may transport radionculides from highly ccn-taminated areas to less-contaminated ones.
In some cases, plateout may occur in the decontaminated areas, resulting in increased dose rates.
For this reason, the most highly contaminated portions of the system, such as the reactor vessel and piping to the pressurizer, may require mechanical decon-tamination by grit blasting or other methods befora, or in place of, full-system chemical decontamination.
Whether chemical or mechanical methods are used and whether the system is decontaminated all at once or section by section, primary-system decontamina-tion will entail most or all of the following in-containment activities:
opening the reactor coolant system, making connections to the reactor piping, and introducing and removing decontamination agents or equipment.
2.1.4 Cleanup of the Auxiliary and Fuel-Handling Building The auxiliary and fuel-handling building has two parts that are separated by a common wall. One part contains tanks, pumps, piping, and other equipment for the processing and storage of water for the reactor and primary cooling system and for the treatment of radioactive wastes.
The other part contains fuel-handling and storage equipment and facilities. The general layout of the auxiliary and fuel-handling building is shown in Figures 2.7 and, 2.8.
2.11
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The interior of the auxiliary and fuel-handling building was severely contaminated by radioactive material as a consequence of the accident. Piping systems that interface with the reactor coolant system were also highly contaminated.
There are 26 such systems in the auxiliary end fuel-handling building.
Some flushing has been done, but major decontamination efforts are still required.
Cleanup of the building entails the following activity:
the removal of miscellaneous equipment and debris that were in the facility at the time of the accident (ladders, tools, portable equipment, etc.); the decon-tamination or removal of-installed equipment (piping systems, air conditioning and exhaust equipment, cable trays, electrical and lighting equipment, etc.);
the decontamination of interior building surfaces (both metal and concrete);
and the removal of contaminated sludge and resins.
In addition, various support activities must be performed to ensure worker safety and to measure the effectiveness of the cleanup.
Cleanup activities in the auxiliary and fuel-handling adiding started shortly af ter the accident and are currently-underway.
Considerable debris and equipment have been removed, -and_ decontamination of the building and remaining equipment has begu...
Because most of the interior surfaces (walls, floors, etc.) are composed of uncoated concrete, radioactive materials have penetrated or leached into the surfaces to varying depths.
The use of high-and low-pressure wate,r sprays, wet vacuuming, concrete spalling, and manual wiping has reduced both the level of smearable contamination on building surfaces and the dose rates in halls and normally occupied areas.
Some temporary dose rate reduction has also been achieved by shielding sources of high radiation (e.g.,
floor drains, the elevator shaf t, and various valves, piping, and pipe dead legs).
Internal-decontamination of tanks and piping remains to be done, including the purification demineralizers, where contami-
~ince the, accident.
Cleanup of several of the nated resin has remained s
higher-dose-rate cubicles also remains.
Support activities 'in the, aux'iliary and fuel-handling building include:
perform radiation surveys to measure the progress of the cleanup effort;
~
identify the need for shielding and/or further decontamination; and provide lighting and utilities.
Support activities are also required for the repair and maintenance of equipment used in the cleanup of the facility and for the repair of piping leaks to eliminate sources of additional contamination.
2.2 CURRENT CLEANUP PLAN: DOSE REDUCTION FOLLOWED BY DEFUELING AND DECONTAMINATION The licensee's schedule for cleanup of the TMI-2 reactor building, as presented in Figure 1.4 of the PEIS, assumed extensive decontamination of the reactor building to signific'antly reduce the radiation levels prior to reactor disassembly and d e f ueling^."
This sequence has been revised for several reasons.
First, the reactor building decontamination to date has been less effective in reducing-dose rates than was originally anticipated.
Second, the presence of the damaged fuel in the reactor core constitutes some risk, pri-marily to workers in the reactor building (the risk results from uncertainties in the core configuration and the remote possibility of a boron dilution incident potentially leading to recritice.lity of the core).
Third, the 2.14 w f
4#
inforaation that will be obtained from laboratory examination of the damaged core will be of value for the design of planned facilities and may also be of benefit to the continued safe ' operation of other nuclear power facilities.
Therefore, to avoid further delaying the removal of the core, the licensee has adopted a revised approach to cleanup.
2.2.1 Tasks and Sequencing of the Current Cleanup Plan The revised cleanup schedule entails the same milestones as the initial schedule, but the sequence of tasks has been altered as follows:
e dose reddction--presently underway and to continue during reactor disassem*aly e reactor disassembly and defueling--to begin in the near future e primary-system decontamination--to follow defueling e reactor building and equipment cleanup--to proceed as resources allow, with coupletion following that of other activities o 'cieanup of th'c auxiliary and fuel-handling building--presently underway and to continue, concurrently with reactor building work, until complete.
- 2. 2.'l.1 Dose Reduction
~
The purpose of the dose reduction program is to reduce the radiation dose rates in occupied portions of the reactor building before and during reactor disassembly and defueling.
These activities, which include the installation of temporary shielding and the removal of certain equipment, are well along and have helped reduce from 40 mrem to 18 mrem the average transit dose for each worker entering the building on the 305-ft elevation and traveling to the 347-ft elevation and back. The objective for the dose reduction program is to red Aa the transit dose to 10 mrem.
Future dose reduction plans call for the continued use of shielding, additional source identification, and the removal, decontamination, or shielding of cable trays, air coolers, and other sources of exposure.
For planning purposes, the licensee has assumed a certain level of effort and set certain goals for reducing dose rates.
These goals are presented in Table 2.1.
Assuming sufficient funds, the time periods shown would correspond tu calendar years? with Period 1 corresponding to 1983, Period 2 to 1984, etc.
However, if the funds available are insufficient to permit this pace for cleanup, each period wculd correspond to two years:
Period I to 1983/1984, Period 2 to 1985/1986, etc.
Dose reduction activities should reduce airborne radioactive contamina-tion and the recontamination of cleaned surfaces.
Efforts are currently in progress to begin, decontamination with water that has been deborated.
This should result in a reduction of airborne contamination.
2.15
TABLE 2.1.
Licensee's Goals for Dose Rate Reduction (Kanga 1983) i Actual Average DoseRate(mrem /hry)
Dose Rate Coals (area /hr) in Occupied Locations Average
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I-Location January 1983(b) Sunener 1983 Period 1(c) = Period 2 Period 3 Period 4 Period 5 Period 6 4
347-ft Elevation 120 110 100 50 20 15 10 305-ft Elevation 350 140 300 200 75 20 15 10 IdI 450 315 200 75 10 282-f t Elevation D Rings 450 315 200 75 10 73 *I 200 40 30 20 98 'I I
I Above Reactor i
Vessel for t
Disassembly g
and Defueling III III 110 50 20 15 10 in Locations on
'120 347-ft Elevation, to Support Reactor Disassembly and Defueling (a)
E= fore dose reduction effort began.
j (b) Data from personnel cosimeters.
(c) A period is 1 to 2 years depending on considerations such as cash flow limitations. The goal is the j
dose rate at the end of the period.
'(d) All work to be done remotely from 305-ft elevation.
1 (e) On service structure before head removal.
l (f) Average values for 347-f t elevation.
4 l
2.2.1.2 Reactor Disassembly and Defueling Early in Period 1 or 2 of the dose reduction program, the preparatory activities that are an essential part of reactor disassembly and defueling will begin.
Disassembly and defueling work is expected to continue at least into Period 4 and possibly into Period 5.
The operations leading to and including the removal of the damaged core from the reactor vessel are listed and discussed below in approximate chronological order.
Some will be done concurrently, and some resequencing may be necessary or advantageous as the cleanup effort progresses.
Although planning is still underway, the licensee's current conceptual designs are briefly described below:
o removal of the reactor pressure vessel head e installation of high-volume cleanup systems for the water in the reactor vessel and fuel transfer canal refurbishment of the fuel transfer canal in the reactor building and of a e
fuel storage pool in the auxiliary and fuel-handling building removal of the reactor vessel upper internals (plenum) a e removal of the reactor fuel, followed by its placement in containers and transfer to the fuel storage pool e removal of the reactor vessel lower internals (core support assembly),
followed by removal of remaining debris from the reactor pressure vessel and draindown and decontamination of the fuel transfer canal.
Removal of the Reactor Vessel Head. The removal of the reactor pressure vessel (RPV) head will require extensive preparation.
Preparatory activities directly related to RPV head removal are expected to include:
- 1) controlling the level of the primary-system water, 2) decontaminating and inspecting support equipment and systems needed for head removal, 3) characterizing radiological conditions under the RPV head to ensure that the contamination and dose rates resulting from the head lif t can be safely handled, 4) removing the missile shields (shown in Figure 2.1), 5) detensioning and parking (or removing) the RPV head studs, 6) refurbishing the reactor internals indexing fixture and placing it on the vessel after the RPV head lift, and 7) fabricat-ing a cover pla,te for possible placement on top of the installed indexing fixture.
Also, as part of the underhead characterization, one control rod drive mechanism has been removed. Additional ones may be removed if required.
All lead screws have already been uncoupled and will be either parked in the RPV head service structure or removed.
When preparations are complete, the RPV head will be lifted with the polar crane to gain access to the reactor vessel internals and the fuel.
It will be placed on an appropriate storage stand with shielding as required.
If necessary, jacking equipment will be used to augment the polar crane during head removal.
If dose rates or contamination potential warrants, head removal 2.17
can be performed with the transfer canal filled.
The internals indexing fix-ture and possibly a cover will then be installed on top of the reactor vessel to facilitate water shielding of the plenum and to provide a work platform for plenum removal and defueling activities.
Installation of High-Volume Water Cleanup Systems.
High-volume water treatment capabilities will be needed to clean particulate and dissolved radionuclides from water in the primary system and the fuel transfer canal both before and during the reactor disassembly and defueling.
Although the submerged demineralizer system (SDS) currently in operation at the site is processing primary coolant, it does not have sufficient capacity to support defueling.
Two separate systems are planned, each with a capacity of about 400 gal / min for filtration and 60 gal / min for ion exchange.
Preliminary designs indicate that one of these systems will treat only reactor vessel water, and the other will treat water in the fuel transfer canal (Devine 1983).
The filter for the system servicing the reactor vessel will be de-signed to fit in modified fuel canisters and will be located in the fuel transfer canal for shielding.
The ion exchange columns are expected to be 8
about 100-ft cask liners of mixed-zeolite ion exchange media.
The columns will be shielded underwater in the transfer canal pool, or placed in a shielded cask inside or outside of containment.
The filter for the system ss rvicing the fuel transfer canal will be like those used for the reactor vessel. This entire system, which will use the existing SDS (after modifica-tion) for cesium removal, will be submerged it 3 pent-fuel pool "A" in the auxiliary and fuel-handling building.
Refurbishment of the Fuel Transfer Canal. The refurbishment of the fuel transfer canal will include the installation of the water cleanup system dis-cussed above, the refurbishment and modification of the fuel transfer equip-ment to handle fuel canisters, and the installation of the seal plate to allow filling of the fuel transfer canal.
Fuel storage racks for fuel pool "A" in the auxiliary and fuel-handling building will also be modified.
Plenum Removal.
Af ter head lift and the installation of the indexing fixture, and concurrently with refurbishment of the fuel pool and preparation and filling of the fuel transfer canal, the condition of the plenum will be evaluated.
The clearance between the plenum and reactor vessel vall was very small prior to the accident.
It is not known whether accident conditions dam-aged the plenum in a way that would make conventional plenum removal impossible.
Plenum removal will require the prior or concurrent removal of the dam-aged fuel assemblies adhering to the underside of the plenum.
They may be dislodged remotely through openings in the plenum, or they may be removed with the plenum.
In an undamaged reactor, the removal and storage of the plenum is norm-ally performed underwater in the fuel transfer canal so that the plenum does not contribute significantly to the occupational radiation dose. This is the current plan for TMI-2.
However, if radiation levels permit, the plenum might be lif ted before the modifications of the transfer canal are complete.
In 2.18
m this case, the plenum would be lif ted into air and subsequently stored under water in part of the transfer canal.
Plenum removal is not ordinarily a high-dose. job; however, ie may be at THI-2, particularly if intact removal is not possible.
Sectioning the plenum would require that workers spend considerable time over the reactor vessel attaching lifting devices to the plenum, aligning cutting equipment, etc.
Workers cutting the plenum would receive some radiation dose from sources in the reactor building t.s well as from the plenum and reactor coolant. Hbwever, the additional dose contribution from the plenum and reactor coolant could be fairly small, depending on the depth of water cover and the effectiveness l
of the water cleanup systems.
Fuel Removal.
Once the plenum assembly has been removed, defueling equipment will be installed in the canal area and the fuel will be removed.
'The fuel removal plans have not yet been finalized because investigations of i
fuel conditions are still in progress.
The reactor vessel defueling sequence will involve removing only that i
fuel material within the reactor vessel--not material that may be lodged in other locations within the reactor primary system, such as in the coolant piping.
The removal of fuel and particulates from other portions of the reactor primary system are discussed in Section 2.2.1.3.
The TMI-2 core contained 177 fuel assemblies.
While their exact condi-tion is uncertain, current information indicates that there are no intact fuel assemblies.
The fuel is assumed to be in a combination of the following configurations:
e fused sections--portions of fuel assemolies fused to each other or to i
structural components in such a way that they will have to be mechanically separated l
e core debris--includes relatively large pieces that can.be mechanically handled, and smaller pieces that will have to be hydraulically vacuumed and filtered.
The initial step of defueling will be the removal of the core debris, to clear the working area in preparation for the removal of large pieces of fuel assemblies.
The small debris will be removed first, followed by accessible loose debris that is larger than pellets but small enough to be placed in canisters.
These canisters will be temporarily stored underwater in the transfer canal, then moved underwater through the transfer tube to the underwater spent-fuel storage racks in the fuel-handling building.
This will provide space in the transfer canal for subsequent defueling operations.
Large fuel pieces will then be removed using remote manipulators and/or long-handled tools.
Adjacent pieces may need to be separated in order to be j
removed.
Removal of Lower Internals.
The core support assembly is a large, I
basket-like component in the reactor vessel that supports the fuel elements l
and directs the entering reactor coolant - towards the lower portion of the 1
2.19 i
i
reactor core.
Along with the removal of fuel from the reactor vessel, fuel particles will be removed from the lower internals.
Then the core support structure will be removed using the internals lifting fixture and polar crane, if possible.
If conditions require, it will be cut up for removal.
As the core support assembly is removed, remaining fuel debris will also be removed and placed in transfer containers.
Although the fuel and reactor core material is highly radioactive, the depth of water over the core should shield workers from all but dissolved or very finely divided debris that becomes dispersed in the coolant. The reactor water cleanup system is expected to remove this material and provide cleaned coolant in the vicinity of defueling workers.
Defueling will, however, require that workers spend considerable time in containment, during which they will receive radiation doses from numerous sources.
Because of the time defueling requires, it will be a relatively large contributor to the radiation dose for cicanup.
After the reactor has been defueled, any remaining fuel canisters and particulate filters from the water treatment system will be transferred through the fuel transfer canal to the fuel storage pool. Defueling equipment will be removed and the transfer canal will be drained and decontaminated.
This will complete reactor disassembly and defueling.
2.2.1.3 Primary-System Decontamination Decontamination of the primary system will involve mechanically and/or chemically decontaminating the internal surfaces, as discussed in Sec-tion 2.1.3 of this report and Section 6 of the PEIS.
At the completion of primary-system decontamination, the radionuclide concentrations in the primary piping system should approach those of operating reactors, 2.2.1.4 Reactor Building and Equipment Cleanup The cleanup of the reactor building and equipment will be an extension of the dose reduction effort, with the purpose of reducing radionuclide con-centrations and radiation dose rates to levels approaching those in operating plants.
Chemical and mechanical decontamination techniques will be used on equipment and on building surfaces. The removal of items such as cable trays, insulation, and portable equipment will reduce doses and facilitate cleanup operations. Some concrete removal is expected to be required, particularly on the 282-ft elevation.
The hollow-concrete-block walls on this elevation will also need to be removed.
Reactor building cleanup will involve a continual sequence of identifying the most significant contributor to radiation dose and airborne contamination, decontaminating or otherwise removing that source, then identifying and decontaminating or removing the next most important source, and so on until dose rate objectives are met.
This repeated process is necessary because of the extreme dif ficulty (with availabic instrumentation) of identifying minor contributors to radiation fields in the presence of major contributors.
2.20
L Cleanup will be further complicated because, once a component is cleaned, it may become recontaminated by particulate radioactive material from the air or from equipment removal or decontamination activities in adj acent areas.
For this reason, it will be important to protect cleaned areas with plastic, strippable coatings, or some other covering, and to determine a sequence for cleanup activities that will minimize recontamination.
Dose rates in the reactor building (from equipment and surfaces) will be a function of the effectiveness of the cleanup actions.
It is expected that a relatively large number of person-hours will be required to complete the cleanup and that the dose rates will decrease ever more slowly as cleanup progresses, because removing a single large source has a much greater effect on dose rates (per worker hour expended) than removing numerous smaller sources.
2.2.1.5 Auxiliary and Fuel-Handling Euilding Cleanup The overall objective of the cleanup effort in the auxiliary and fuel-handling building is to permit access to all portions of the building. Access has been limited because of surface and airborne contamination and radiation exposure from confined sources (radionuclides inside pipe runs, resin columns, dead legs, holding tanks, etc.).
Mechanical and chemical decontamination techniques will be used inside tanks and piping and on equipment and building surfaces.
The removal of contaminated items that are still in the building, such as portable equipment, insulation, sludge, resins, and miscellaneous debris, will facilitate cleanup.
Some concrete spalling has been done and more will be required, particularly on the concrete surfaces that were below the accident water level or were otherwise exposed to contaminated liquids.
Hollow-concrete-block walls may have to be removed.
The building will require some additional general cleanup, primarily of overhead areas and of cubicles and their contents.
As in the reactor building, cleanup may be hampered by recontamination, and covering decontaminated areas with protective materials may be important.
The cubicle areas will be the most difficult to decontaminate because of the concentration of equipment (tanks, filters, piping, etc.), the crowded work space, the need for special shielding (e.g., lead blankets), and the high contamination and radiation levels. The makeup and purification demineralizer cubicles may be the most severely contaminated because of radionuclides that were deposited in the in-line filters and demineralizer resins during the accident.
The decontamination plan presented in the PEIS postulated complete decon-tamination of the auxiliary and fuel-handling building using conventional decontamination methods, including water flushing and hydroblasting (high-pressure water flushing).
Experience has indicated that these methods are not effective in reducing dose rates and are not as rapid as originally anticipated.
2.21
2.2.2 Occupational Radiation Dose Associated with the Current Cleanup Plan In order to determine the occupational radiation dose associated with the current cleanup plan, a team of nuclear-operations and decontamination specialists evaluated the work to be performed and the dose required for each task.
Each task was evaluated assuming that the tasks would be performed in the sequence described and that occupational radiation doses would be main-tained ALARA by the proper planning and execution of each task.
A great deal of information and data required for accurate estimates will become available only during the progress of cleanup (e.g.,
conditions inside the reactor, characterization of contamination).
Because of this, the radiation dose estimate is presented as a range.
The upper and lower ends of the estimated j
range represent the corresponding extremes of conditions based on an evaluation of the information presently available.
Table 2.2 lists the estimated range of occupational radiation doses for cleanup performed according to the current plan.
Doses for work performed to date and doses for waste management tasks (taken from the PEIS) are included.
Observations regarding these estimated doses are presented in the following paragraphs.
The occupational doce incurred during performance of the dose reduction sask will effectively reduce the radiation doses to workers performing subsequent tasks.
Eliminating this task would effectively increase the deses for later tasks.
TABLE 2.2.
Estimated Occupational Radiation Dose for the Current cleanup Plan Task Person-rem Dose Reduction Program 2,000-5,100 Reactor Disassembly and Defueling 2,600-15,000 Primary-System Decontamination 56-970 Reactor Building and Equipment cleanup 5,900-21,000 Auxiliary and Fuel-Handling Building Cleanup 500-1,400 Utility and System Maintenance 100-200 Waste Management and Transportation (*)97-485 Dose To Date 1700 13,000-46,000 (a) From the PEIS.
2.22
The range of estimated doses for completing reactor disassembly and defueling (2,600 to 15,000 person-rem) is wide because of many uncertainties involving the removal of the reactor internals and fuel and the effectiveness of the water cleanup systems.
The plenum may be removed intact, or an extensive effort may be needed to section and remove it.
The time required to transfer the fuel to canisters is likewise uncertain.
If the fuel is not fused, a lower number of person-hours and a lower dose would be expected.
However, if much of the fuel is fused, the dose would be much higher.
The transfer canal will contain myriad small particulate sources of radiation that will be handled by the water cleanup system during defueling.
If these sources are kept well underwater and transferred to fuel canisters by the water cleanup system, dose rates will be low.
However, if a significant portion of these particulates forms a film on the surface of the water in the transfer canal, the average dose rate for the workers could be much higher.
The processes for primary-system decontamination have not yet been identified by the licensee. The occupational dose required will be a function of the number and type of dead legs (sample lines and other areas of restricted flow) that workers must flush, the number of repeat processes that must be performed, the occurrence of spills resulting from leaks in the system, and the waste-handling method used.
Cleanup of the reactor building and equipment will result in an estimated 5,900 to 21,000 person-rem of occupational radiation dose.
As much as 80% of this dose is associated with cleanup of the 282-ft elevation.
This estimate assumes that considerable decontamination of this elevation is performed from the 305-ft elevation through floor penetrations prior to entry into the 282-ft elevation.
As an alternative, immersion decontamination, accomplished by filling the basement with water or other decontamination solutions and processing the water on either a batch or a continuous basis, is being considered but was not evaluated due to limited knowledge of its effectiveness.
Extensive use of robotics on the 282-ft level would also reduce the dose to workers.
The robotic option is explored further as Alternative 3.
Final cleanup of cubicals and systems in the auxiliary and fuel-handling building, including the processing of decontamination waste from system and tank cleanup, is estimated to require between 500 and 1400 person-rem.
The maintenance of utilities, communication systems, and other essential servic.cs during the cleanup is expected to require an additional 100 to 200 person-rem, depending on the frequency of breakdowns and the duration of the cleanup effort.
Approximately 1700 person-rem have already been incurred during cleanup operations through August 22, 1983.
In the opinion of the staff, if cleanup goes well, it might be completed at the low estimate of 13,000 person-rem.
However, even if additional problems continue to arise, cleanup should be completed at less than the high estimate of 46,000 person-rem.
2.23
2.3 ALTERNATIVE 1:
EXTENSIVE CLEANUP FOLLOWED BY DEFUELING As mentioned earlier, the initial cleanup plans discussed in the PEIS called for extensive decontamination of the reactor building and equipment prior to defueling.
It was believed at the time the PEIS was being written that such decontamination could be accomplished largely by water flushing and hydroblasting (high-pressure water flushing).
Experience to date has indicated that these activities are less effective at reducing dos 2 rates than had been anticipated, probably because contamination is embedded deeper in surfaces than was expected because of delays in beginning cleanup.
This alternative to the current cleanup plan calls for meeting the initial dose reduction goal of about 10 mrem /hr in occupied areas through a combination of aggressive decontamination, equipment removal, and shielding.
Once this goal is met, the reactor would be disassembled and defueled and the primary system would be decontaminated.
In this section, the procedures and work sequence for decontaminating the building and equipment, disassembling and defueling the reactor, and decontaminating the primary system are outiined, and the impact of this alternative on occupational dose is discussed.
2.3.1 Tasks and Sequencing of Alternative 1 Under this alternative, decontamination of the auxiliary and fuel-handling building would be as described in the discussion of the current cleanup plan.
The sequence of decontamination operations in the reactor building would consist of first removing debris and heavy deposits, and then cleaning the exposed surfaces.
Cleanup efforts would begin at upper levels and proceed downward to minimize recontamination.
The majority of the building-cleaning effort would precede defueling; however, some final cleanup would be required following defueling and primary-system decontamination.
2.3.1.1 Reactor Building and Equipment Cleanup Cable trays, overhead lighting, and electrical conduits are known to be significant sources of occupational radiation exposure.
Water flushing and hydroblasting are not particularly effective at decontaminating these sources.
Unless some alternative method of chemical decontamination, such as foam cleaning or freon cleaning, proves effective, the equipment would have to be rewoved to eliminate these sources.
Removal of the equipment would require the identification and replacement of instrument and control cables required for safety, and the installation of temporary lighting and electrical outlets needed to operate decontamination and defueling equipment.
Chemical decon-l tamination or removal of the reactor building's air coolers would also be l
required.
Floor drains would have to be chemically decontaminated, the surfaces of concrete floors and walls would have to be removed by spalling.
l and other aggressive decontamination actions would be required.
Some shield-ing of primary piping and other sources would also be required to reach the dose rate objective.
Such an extensive cleanup program would require extensive planning, testing, and source identification as well as a substantial number of workers 2.24
in containment.
Large occupational doses would be incurred early in the cleanup effort.
This approach would delay the start of fuel removal for at least 1-1/2 years and possibly considerably
- longer, depending on the difficulties encountered.
2. 3.1.' 2 Reactor Disassembly and Defueling and Primary-System Decontamination Under Alternative 1, disassembly and defueling of the reactor and decon-tamination of the primary system would involve essentially the same tasks as described for the current plan.
The difference would be that these tasks 4
would be performed in lower radiation fields, with only a small dose contri-bution from radiation sources associated with the building and equipment other than the reactor primary system. During building cleanup, the primary coalant would be processed in small batches through the SDS system, as is now being done.
.This additional processing beyond what has already been done is expected to have a negligible effect on the quantity of radioactive material handled during defueling, or on the dose rates from this material.
Theoreti-cally, the longer radioactive materials are in contact with reactor piping, the greater the extent of radionuclide migration into the oxide layer of the piping and the more difficult decontamination becomes.
In view of the con-siderable time between the accident and decontamination of the primary system (under all eptions), the delay required under this alternative to allow for building cleanup would have little effect on the ease or effectivness of primary-system decontamination.
Much of the dose received during primary-system decontamination is from material in the primary system rather than sources in the building.
Therefore, the dose for primary-system decontamina-tion in this alternative is only slightly less than the dose for the same task in the current plan.
Additional building decontamination would be required during and follow-ing both defueling and primary-system decontamination to maintain the dose rates achieved during the initial building and equipment cleanup phase.
This reeleaning would result in additional occupational radiation doses.
2.3.2 Occupational Radiation Dose Associated with Extensive Cleanup Followed by Defueling The occupational radiation dose associated with this alternative was estimated in the same manner as way the dose for the current cleanup plan and is shown, broken down by tasks, in Table 2.3.
The dose reduction task called for in the current plan does not appear in Table 2.3 because any of those activities required as part of Alternative I would be performed as part of the reactor building and equipment cleanup, not as a separate task.
It was assumed that considerable equipment would need to be removed in order to achieve the goals for this alternative.
Because fuel remains in the
[
reactor, certain safety systems are required. The preservation or replacement of these systems would require a very large number of man-hours in containment and a corresponding increase in worker doses.
l 2.25
-TABLE 2.3.
Estimated Occupational Radiation Dose for Extensive Cleanup Followed by Defueling Task-Person-rem Reactor Building and Equipment Cleanup 9,000-30,000 Reactor Disassembly and Defueling 820-6,500 Primary-System Decontamination 39-780 Reactor Building Recleaning 12-630 Auxiliary and Fuel-Handling Building Cleanup 500-1400 Utility and System Maintenance 100-200 Waste Management and Transportation (*)97-485
~
Dose to Date 1700 12,000-42,000 (a) From the PE1S.
Even assuming release from some of these requirements, higher occupa-tional doses were estimated for reactor building and equipment cleanup under this alternative than under the current cleanup plan, for the following reasons:
e Worker time in containment would be required to replace some control and utility cables to ensure that the reactor is maintained in a safe status prior to fuel removal.
e The lack of a dose reduction program preceding cleanup would result in the cleanup work being done at high dose rates and would require more
. worker hours for completion of this operation.
(Under the current plan,.
some source removal is performed as-part of the dose reduction program.)
Even with aggressive building decontamination, there is little assurance that the average 10-mrem /hr target for the reactor building could be met as long as fuel and fission product contamination remained in the primary system.
The goal would certainly not be met inside the D-rings or near primary-system piping and -- components.
An average working dose rate of 10 mrem /hr was, however. assumed as the low dose rate for most reactor disassembly and defueling tasks.
The occupational dose for primary-system decontamination was lower under this alternative than under the current plan because of the lower general-area dose rates. The average dose rate, however, was assumed to be somewhat above
-10 mrem /hr because the workers would be close to the reactor coolant piping for much of this work.
The task of maintaining reactor building cleanliness during defueling and decontamination is new under this alternative. The level of effort that would 2.26
be required is difficult to estimate because it would depend on the nature of the reactor core debris, the contamination control barriers provided, the work practices, the process used for primary-system decontamination, and the number and size of any leaks in the primary system.
Because the dose rates for this task would be low, the total dose involved would be relatively small.
Cleanup of the auxiliary and fuel-handling building would result in the same dose under this alternative as under the current plan because it would be done in the same way.
Utility and system maintenance is estimated to require approximately the same dose under this alternative as under the current plan.
The utilities would be needed for a longer time under this alternative; however, the dose rates involved in maintenance would decrease earlier in the cleanup operation.
If cleanup were performed according to this alternative, fuel removal would not begin for several years.
2.4 ALTERNATIVE 2: PRASED DEFUELING FOLLOWED BY REACTOR BUILDING CLEANUP Alternative 2 differs from the current plan and the other alternatives in that a large portion of the fuel debris would be removed as a slurry before the reactor vessel head was lifted.
Although there are currently no plans to do any defueling before the head lift, this alternative is included because it would minimize the potential for fuel fines to contaminate equipment and
- result in exposure to personnel during later operations.
Also, there may be safety advantages to having the reactor vessel head in place as long as possible because it would provide shielding to the sorkers performing initial defueling tasks.
Drawbacks to this alternative include delays resulting from the design, fabrication, and testing of equipment for phased fuel removal, and additional equipment costs.
2.4.1 Tasks and Sequencing of Alternative 2 Phased defueling would be accomplished by altering the sequence of tasks for reactor defueling.
The major tasks and their general sequence for phased defueling are:
e implementation of the dose reduction program, as described for the current plan (this program would continue throughout reactor defueling) e installation of water vacuum and support equipment for removing the fuel fines, and removal of the fines through a control rod drive mechanism (CRDM) nozzle in the head e preparation for reactor vessel head removal, and removal of the head, plenum, fuel, and reactor vessel internals, as described for the current plan e decontamination of the primary system, as described for the current plan 2.27 1
e completion of the auxiliary and fuel-handling building cleanup and the reactor building and equipment cleanup, as described for the current plan.
2.4.1.1 Fines Removal Prior to Head Lift Under this alternative, a fuel debris removal system would be installed before the reactor vessel head was lifted. This system would have some of the features of the planned system for reactor water cleanup system except that canisters would be provided for the collection of relatively large quantities of fuel debris, and a system would be required for observing and manipulating the vacuum nozzle within the reactor vessel. The time required for the design and fabrication of this system would delay fuel removal and all subsequent cleanup efforts for at least 18 months, perhaps longer.
The debris removal ' system would include a water vacuum probe inserted through a CRDM nozzle (the CRDM was previously removed for the underhead characterization work).
The vacuum would be used to remove accessible fines and small rubble.
Debris removal would be observed by closed-circuit TV (CCTV) inserted in one of the two vacant CRDM lead screw holes (the lead screws were removed for quick-scan and quick-look operations).
The debris removal nozzle would be controlled by a cable system similar to that used for control of the CCTV cameras.
Clarified borated water would be returned to the reactor vessel using a third CRDM lead-screw opening.
Actual debris removal would take only a few months unless nozzle plugging and visibility problems were severe, in which case it could take much longer.
A substantial portion of the estimated 100,000 lb of rubble 1 inch or less in diameter might be removed in this way.
The fuel canieters would require considerable shielding, either by storage underwater (which might be accomplished by filling the fuel transfer canal) or by the use of massive shielding casks.
Filling the fuel transfer canal for shielding in the near future could impede the necessary refurbish-ment of the canal. The availability of adequately shielded casks has not beea investigated.
2.4.1.2 Reactor Disassembly and Defueling Af tcr the modification and refurbishment of the fuel transfer equipment and the removal of accessible fines from the reactor vessel, reactor dis-assembly and defueling would proceed as described for the current plan, with the exceptions noted below. Under the current plan, every effort will be made to perform a dry head lif t because refurbishment of the transfer canal will not be complete.
If the head lift was delayed until the transfer canal refurbishment was complete, as it would be under this alternative, the incentives for dry head lift would diminish.
A wet head lift is expected to require less occupational dose.
Once the head was lifted, there would be much less particulate radio-activity in the reactor coolant and therefore a diminished probability of 1
rapid releases of dissolved cesium from the core contents as it is disturbed.
2.28 e
=
This would lead to lower dose rates. Defueling after head removal would also involve fewer filter changes and fewer worker hours because so much material would have been removed before head lift. Later defueling activities would be identical to those for the current plan, except that under this alternative, the effort required to decontaminate the transfer canal following defueling could be somewhat lessened because of lower contaminant levels in the water.
2.4.1.3 Primary-System Decontamination, Auxiliary and Fuel-Handling Building Cleanup, and Reactor Building and Equipment Cleanup These activities would be unaffected by the defueling method; hence, for these activities, all aspects of Alternative 2 and the current plan are j
identical.
2.4.2 Occupational Radiation Dose Associated with Phased Defueling Followed by Reactor Building Cleanup The occupational radiation dose required to perform phased defueling followed by reactor building cleanup was estimated in the same manner as the dose for the current plan.
The total estimate and the breakdown by task are given _ in Table 2.4.
The occupational dose needed to accomplish the dose reduction program was unchanged from that of the current plan.
The dose range for removing the fuel fines prior to head lift was estimated assuming that either water or solid material would be used as shielding to diminish the dose contribution from the fuel fines.
TABLE 2.4.
Estimated Occupational Radiation Dose fo~ Phased Defueling Followed by Reactor Building Cleanup Task Person-rem Dose Reduction Program 2,000-5,100 Defueling Operation Prior to Head Lift 140-540 Reactor Disassembly and Defueling 2,600-14,000 Primary-System Decontamination 56-970 Reactor Building and Equipment Cleanup 5,900-21,000 Auxiliary and Fuel-Handling Building Cleanup 500-1,400 Utility and System Maintenance 140-280 Waste Management and Transportation (*}
97-485 Dose To Date 1700 13,000-45,000 (a)
From the PEIS.
2.29 I
I
The doses for reactor disassembly and defueling would be only slightly J
lower under this alternative than under the current plan, because the time that would be required for vacuuming the fines represents only a small portion of the time needed for fuel removal, and the dose rates in the building would remain approximately the same.
The greatest advantage of early fuel removal would be the subsequent decrease in the quantity of particulates that could I
contribute to worker dose. This decrease results in the lowering of the upper bound assumed for the dose rates for the balance of defueling.
The early removal of fines might also simplify cleanup of the transfer canal, and this benefit is reflected in the dose estimate.
The doses for primary-system decontamination, reactor building and equip-ment cleanup, and auxiliary and fuel-handling building cleanup would be the same under this alternative as under the current cleanup plan; they would not be affected by the fuel removal procedure considered under this alternative.
The dose required for utility and system maintenance would increase over that of the current plan to account for the additional time that this alternative would prolong the cleanup.
(This additional time would be needed to allow for the design, development, construction, and testing of the equipment needed for phased fuel removal.)
2.5 ALTERNATIVE 3: DEFUELING FOLLOWED BY DELAYED CLEANUP USING ROBOTICS A third alternative for cleaning up TMI-2 would be to clean up most or all of the auxiliary and fuel-handling building and to reduce the dose rates in and defuel the reactor, as described in the current plan; then to place the reactor and containment building in interim, monitored storage, and to perform final building cleanup using robotics sometime in the future, when appropriate technology and devices become available.
While timely removal of the damaged fuel is considered essential, the option of delaying further cleanup was considered worthy of evaluation.
Robotics is a rapidly emerging technology with the potential for eliminating considerable occupational radiation exposure.
Robotics is already being applied to a limited degree in the auxiliary and fuel-handling building, and applications in the reactor building are being evaluated. How much time would elapse before reliable and economical robotic devices could perform a majority of the in-containment cleanup work is unknown.
The most optimistic proj ec-tions for robotic technology indicate that adequate robots will be available before they would be required for building cleanup under the current work sequence. More realistic projections indicate that a storage period of 10 to 20 years may be required before robotic cleanup would be possible.
Although maximizing the use of available robotic devices for high-dose work would be consistent with the ALARA principle, certain assurances would be required before this alternative could be adopted.
The safety of the interim-care phase would require additional study and assessment. There would need to be better assurance that the robotic technology needed to accomplish cleanup would become available.
In addition, provisions for financing future cleanup would need to be made.
2.30
2.5.1 Tasks and Sequencing of Alternative 3 This alternative would include the phases of cleanup discussed for the reactor building in the current cleanup plan and would incorporate an interim-storage phase as well. These are discussed below.
2.5.1.1.
Reactor Disassembly and Defueling The auxiliary and fuel-handling building cleanup, dose reduction program, and reactor disassembly and defueling would proceed concurrently, essentially as described in the current plan.
The areas of the auxiliary and fuel-handling building with the highest dose rates might be lef t untouched.
In the dose reduction program, slightly greater emphasis might be placed on shielding rather than decontamination, and only locations that must be occupied for reactor disassembly and defueling would be subject to dose reduction efforts.
The 282-ft elevation, for example, would probably be left totally untouched to reduce the occupational radiation dose.
Because the safety of the monitored interim storage period has not been evaluated, it is difficult to predict how much radioactive material, particu-larly fuel, might be allowed to remain during this phase.
Although it is clear that fuel inventories should be reduced to a level where criticality is inconceivable, such a criterion would require only that about half the fuel be removed.
The actual quantity permitted to remain during interim storage, if interim storage were allowed, would probably be much less.
Under this alternative, defueling might stop prior to final cleanup of the transfer canal, or some selected mechanical or chemical decontamination might required for those portions of the primary system that contain fuel particles.
2.5.1.2 Interim Storage of the Defueled Reactor Upon the completion of reactor defueling, the auxiliary and fuel-handling building and the containment building would be placed in an interim, monitored storage mode until robotic technology was available to perform the remaining decontamination of cubicles in the auxiliary and fuel-handling building and of the primary system and the reactor building and equipment.
Interim storage would involve the maintenance of essential services (e.g.,
security and radiological surveillance, utilities, ventilation syste=s, and planning and administration), but no active program of building or equipment decontamina-tion would be conducted except as remote or robotic technology became avail-able.
During interim storage, occurational radiation exposures would be restricted to those necessary to maintain the facilities in a safe and secure condition.
Tasks such as repairing the ventilation systems and changing filters would account for most of the dose received.
2.5.1.3 Primary-System Decontamination Except for those activities necessary for the reactor to be considered safe for interim monitored storage, any primary-system decontamination would 2.31
be done by robotics.
Decontamination performed by plant workers before interim storage might include localized chemical or mechanical cleaning, but would involve only a small fraction of the occupational radiation dose incurred for complete primary-system decontamination under the current plan.
Further primary-system decontamination might or might not be undertaken following interim storage of the reactor, depending on the anticipated future use of the reactor, waste disposal limitations in effect at that time, the capabilities of available robotic devices, and other factors.
If decontamina-tion were undertaken by rebotics, the only occupational radiation dose incurred would be from decontaminating and maintaining the robots, and possibly from htadling and transporting the waste generated; however, some of these tasks might also be done by robotics.
2.5.1.4 Robotic Cleanup of the Reactor Building and Equipment It is somewhat premature to envision in detail what tasks might be involved in robotic cleanup of the TMI-2 reactor building because most*
present-generation robots are severely limited in mobility, dexterity, strength, or logic.
The tasks of equipment removal, building and equipment decontamination, shielding removal, and decontamination and building survey would have to be performed to complete the cleanup.
The principal difference between this alternative and the current plan is that these tasks would be performed without workers routinely being in the reactor building.
Occupational doses incurred during robotic cleanup of the reactor building, like those incurred during primary-system decontamination using robotics, would primarily be those from decontaminating and servicing robots and from waste-packaging, waste-handling, and waste transportation activities that were not done robotically.
2.5.2 Occupational Radiation Dose Associated with Defueling Followed by Delayed Cleanup Using Robotics The occupational radiation dose associated with this alternative was estimated in the same manner as the dose for the cleanup plan and the other alternatives.
The total and task-breakdown estimates are presented in Table 2.5.
The dose reduction program and reactor disassembly and defueling would be performed in the same way and require the same dose as under the current plan.
The primary-system cleaning performed by plant workers before interim storage would consist only of the localized c1 caning required for the plant to be considered defueled.
The extent of this activity was arbitrarily chosen becauce the criteria for interim storage have not been established. A dose of 20% of that required for the full-system decontamination considered in the current plan was used.
In reality, any value between zero and th? maximum dose of 970 person-rem under the current plan might be possible.
2.32
TABLE 2.5.
Estimated Occupational Radiation Dose for Defueling Followed by Delayed Cleanup Using Robotics Task Person-rem Dose Reduction Program 2,000-5,100 Reactor Disassembly and Defueling 2,600-15,000 Primary-System Cleaning 11-190 Utility and System Maintenance 80-160 Interim Care of Reactor Building and Auxiliary and Fuel-Handling Building (1.7-31 person-rem per year) 0-620 ")
Auxiliary and Fuel-Handling Building Cleanup 97-1,400 Robotic Primary-System Decontamination, 300-3,500 Reactor Building and Equipment Decontamina-tion, and Final Auxiliary and Fuel-Handling Building Cleanup Waste Management and Transportation ( )97-485 Dose To Date 1700 6,900-28,000 (a) Based on 0 to 20 years of interim care.
(b) From the PEIS.
Utility and system maintenance would be required only until defueling, including any primary-system decontamination, was complete; therefore, doses associated with this task are lower under this alternative than under the current plan.
However, a new task, interim care during the storage period, would be required.
The dose incurred in maintaining the reactor building during this time would be 1.6 to 30 person-rem per year.
This interim-care period might not be required, or it could continue for as long as 20 years.
It is this difference that accounts for the wide' range of doses presented.
Cleanup of the auxiliary and fuel-handling building would be much the same under this alternative as t is under the current plan, except that areas where there are high dose rates (e.g.,
the insides of tanks and piping systems) might, remain untouched until robotic technology was available.
The elimination of a few high-dose jobs involving a relatively large uncertainty accounts for the difference between the low end of the dose range estimated for this alternative and that presented for the current plan. The high end of the dose range was estimated assuming the same treatment as under the current plan.- The dose incurred for interim care of the auxiliary and fuel-handling building is estimated to be 0.1 to 1.0 person-rem per year.
l Primary-system decontamination, reactor building and equipment decon-l tamination, and cleanup of remaining hot spots in the auxiliary and fuel-2.33
handling building would all be done robotica11y under this alternative.
Robotic activities are, however, expected to result in some radiation dose to workers mainte.ining the robots and performing other activities. This dose was assumed to be between 5% of the low dose and 15% of the high dose from manual performance of the activities.
2.6 ANALYSIS OF THE CURRENT CLEANUP FLAN AND ALTERNATIVES Sections 2.2 through 2.5 described four approaches to accident cleanup at THI-2 and presented estimates of the occupational radiation dose associated with each approach.
The approaches that were selected would use available or emerging technology and would be consistent with the conclusion of the PEIS that the TMI-2 site is not suitable as a permanent repository for the accident-generated waste. This section is intended to summarize the strengths and weakness of the current cleanup plan and the three alternatives and to provide an additional basis for the environmental impact discussed in Section 3.
The criteria against which the licensee's current plan and each alterna-tive were evaluated include:
e public safety
- occupational radiation dose
- time schedule for fuel removal and completion of cleanup
- technical feasibility.
In the following discussion, the four cleanup options are compared using these four criteria.
2.6.1 Analysis of Public Safety The safety concerns of the TMI-2 reactor are presented in the PEIS and have not changed.
Therefore, they are not discussed here.
However, the safety concerns will be substantially reduced when the fuel is removed.
The current plan and Alternative 3 (defueling followed by delayed cicanup using robotics) are therefore preferable according to this criterion. Alternative 2 (phased defueling followed by reactor building cleanup) was evaluated because it appeared to have some advantages for the safety of the public and the workers. The staff now feels that any advantages of Alternative 2 are offset by the fact that it would delay defueling by at least 1-1/2 years.
The public safety of the monitored, interim-storage phase that is envis-ioned as part of Alternative 3 would require additional evaluation.
Although the possible release modes and affected enivronment are well known, the radionuclide inventories, the type of care that would be provided, cnd the duration of the care period are unknown.
An evaluation of the safety of this phase would therefore be premature at this time.
2.34
2.6.2 Analysis of Occupational Radiation Dose As illustrated in Figure 2.9, the estimated dose associated with cleanup of the TMI-2 site under the current plan is considerably higher than the dose associated with cleanup under Alternative 3 (defueling followed by delayed cleanup using robotics), and slightly higher than that for Alternative 1 (extensive cleanup followed by defueling). The estimated dose for the current plan is equivalent to that for Alternative 2 (phased defueling followed by reactor building cleanup).
Although the lowest occupational radiation dose is associated with Al-ternative 3, the tasks that would be performed under this alternative, through the reactor disassembly and defueling phase, are the same as those under the current plan.
Therefore, it is not necessary to make a decision for or against Alternative 3 on the basis of radiation dose at the present time.
The second lowest dose is estimated for Alternative 1, extensive decon-tamination followed by defueling.
The implementation of Alternative I would preclude the use of robotics to perform the high-exposure job of reactor building cleanup because the building would be decontaminated in the very near future, before adequate robotic technology became available.
50,000
"" ~
40,000 s 30.000 E
i O
u) h 20,000
=
=
10,000
=
O CURRENT ALTERNATIVE ALTERNATIVE ALTERNATIVE PLAN ONE TWO THREE FIGURE 2.9.
Occupational Radiation Dose to Complete Cleanup 2.35
On the basis of occupational dose, Alternative 2 (phased defueling fol-loved by reactor building cleanup) is essentially equivalent to the current plan.
2.6.3 Analysis of Time Schedule The prompt removal of fuel and cleanup of the reactor building affects worker dose, both directly because of routine maintenance and indirectly because of ease of cleanup.
An attempt was therefore made to determine the relative effect of the current plan and the alternatives on the timing of fuel removal and the completion of cleanup.
To do this, four schedules (presented as Figures 2.10, 2.11, 2.12, and 2.13) were prepared to reflect the plan and the alternatives. These schedules are presented in periods rather than years.
The periods used here correspond roughly to those used by the licensee in establishing dose reduction goals (Kanga 1983).
If resources were unlimited, a period could correspond to 6 to 9 months.
Under the best conditions of available resources, it probably represents 1 year; under less favorable conditions, 2 years.
These schedules show the earliest probable start time and the latest start time for each activity.
The duration of major tasks in the various approaches to cleanup is discussed below.
1 Under all options, reactor disassembly and defueling must await the re-qualification of the polar crane.
Under Alternative 1 (extensive cleanup followed by defueling), disassembly and defueling must also await the comple-tion of reactor building cleanup.
Under Alternative 2 (phased defueling followed by reactor building cleanup), disassembly and the completion of defueling must avait the design, fabrication, and operation of a system to remove fines through the reactor head.
For all approaches, disassembly and defueling (from head removal through transfer canal cleanup) was estimated to require a minimum of 2-1/4 periods and a maximum of 4-1/2 periods, illustrated in detail in Figure 2.10.
Reactor building cleanup was estimated to require betveen 2 and 3 periods under the current plan and Alternative 2 (phased defueling followed by reactor building cleanup).
Under Alternative 1, when building cleanup would precede defueling, it was estimated to require between 2-1/2 and 4 periods because of the need to maintain some safe.y systems in operable condition.
In addition, under Alternative 1, the reactor building would require some additional clean-ing following both defueling and primary-system decontamination.
Primary-system decontamination was estimated to require 1/4 to 1/2 period following defueling for all cases in which it would be performed.
Cleanup of the auxiliary and fuel-handling building was estimated to require from 1-1/4 periods to 4 periods, and utility and system maintenance is required under all options for as long as work is going on.
As shown in Figures 2.10 and 2.13, the current plan and Alternative 3 (defueling followed by delayed cleanup using robotics) provide for the earliest defueling, completed in 3-1/4 to 6 periods.
Alternative 2 (phased defueling followed by reactor building cleanup) would delay the completion of l
2.36 i
PERIOD (a) 1 2
3 4
5 G
7 8
9 10 11 DOSE REDUCTION O
PROGRAM REACTOR DISASSEMBLY n
n AND DEFUELING HEAD REMOVAL dHD PLENUM REMOVAL M
FUEL & LOWER-I ;
INTERNALS REMOVAL o
()
jq p CANAL CLEANING J
PRIMARY SYSTEM
,,,,,,,j DECONTAMINATION REACTOR BUILDING AND EQUIPMENT O
O""""2i N
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CLEANUP UTILITY & SYSTEM J(
i a
MAINTENANCE
~
6 EARLY START O EARLY FINISH A LATE START S LATE FINISH LIGURE 2.10.
Conceptual Schedule for the Current Plan (a) See text - Section 2.6.3.
l l
l l
l 2.37 l
l l
l
PERIOD (a) 1 2
3 4
5 6
7 8
9 10 11 REACTOR BUILDING AND EQUIPMENT d1 O
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O CLEANUP UTILITY AND SYSTEM X
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MAINTENANCE O EARLY START O EARLY FINISH A LATE START e LATE FINISH FIGURE 2.11.
Conceptual Schedule for Alternative 1 (a) See text - Section 2.6.3.
I 1
2.38
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PERIOD 1
2 3
4 5
6 7
8 9
10 11 DOSE REDUCTION m
PROGRAM if i f PARTIAL DEFUELING
/W PRIOR TO HEAD j g LIFT REACTOR DISASSEMBLY N
l l s
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j '
s 6
MAINTENANCE T
6 EARLY START O EARLY FINISH A LATE START e LATE FINISH FIGURE 2.12.
Conceptual Schedule for Alternative 2 (a) See text - Section 2.6.3.
2.39
PERIOD (a) 1 2
3 4
5 6
7 DOSE REDUCTION m
PROGRAM l'
REACTOR DISASSEMBLY g
i es
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AND DEFUELING INITIAL PRIMARY.
}'
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SYSTEM CLEANING INITIAL AFHB CLEANUP O
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INTERIM MONITORED Jh j
STORAGE ROBOTIC PRIMARY-SYSTEM DECONTAMINATION, REACTOR BUILDING c
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AND EQUIPMENT 2-5 YEARS 2-5 CLEANUP AND YEARS AFHB CLEANUP UTILITY AND SYSTEM A
l 6
MAINTENANCE d EARLY START O EARLY FINISH A LATE START 9 LATE FINISH FIGURE 2.13.
Conceptual Schedule for Alternative 3 (a) See text - Section 2.6.3.
w 2.40
defueling to 4 to 6-1/2 periods. Alternative 1 (extensive cleanup followed by defueling) would have the greatest impact, delaying the completion of defuel-ing to between 4-1/2 and 8-1/2 periods.
The completion of cicanup also varies with the alternatives.
The current plan and Alternative 1 are comparable in this area, with cleanup completed between 5-3/4 and 9-3/4 periods.
Alternative 2 (phased defueling followed by reactor building cleanup) would extend the cleanup time to between 6-1/2 and 10-1/4 periods.
Under Alternative 3 (defueling followed by delayed cleanup using robotics), final cleanup might not be completed for more than 30 years.
2.6.4 Analysis of Technical Feasibility The technical feasibility of the various alternatives was also evaluated.
Alternative 3, involving delayed cleanup by robotics, would clearly have some drawbacks in this area.
Current models have suffered from reliability prob-lems.
In addition, there is no assurance that robotic technology will pro-gress to the point at which robots could perform all phases of cleanup.
However, current models are capable of some cleanup tasks, and the development of more versatile models appears to be progressing rapidly.
Under Alterna-tive 1, the ability of the licensee to meet the goals set for building and equipment decontamination prior to defueling is subject to some doubt.
Fuel in the primary system might preclude meeting these goals.
The current plan and Alternative 2 (phased defueling followed by reactor building cleanup) were both judged to be technically feasible.
2.6.5 Summary Analysis The staff has determined that, in terms of the nature of the activities involved, the current cleanup plan, Alternative 1, and Alternative 2 all fall within the scope of the PEIS. The interim-storage phase of Alternative 3 does not.
All of the options have advantages and drawbacks (summarized in Table 2.6), and all would involve an occupational radiation dose beyond that estimated in the PEIS.
The current plan is equal or superior to the alternatives with respect to all criteria except occupational dose; Alternative 3 would result in a lower occupational dose, but currently the technical feasibility of Alternative 3 is not assured.
Alternative 1 (extensive decontamination followed by defueling) has the drawback of delaying fuel removal.
There is also some question regarding the feasibility of meeting the 10-mrem / hour decontamination goal prior to defuel-ing and primary-system decontamination.
Alternative 2 (phased defueling followed by reactor building cleanup) is equivalent to the current plan with respect to public safety and technical feasibility.
It has the drawback of delaying both fuel removal and final building cleanup.
Alternative 3 (defueling followed by delayed cleanup using robotics) is expected to be superior to the current plan with respect to occupational dose and equivalent with respect to the time for fuel removal.
It would, however, result in an undefined, but possibly very long, delay in the time required to 2.41
TABLE 2.6.
Summary Evaluation of Cleanup Alternatives Criterion
-Current Plan Alternative 1 Alternative' 2 Alternative 3 Public Safety No change (*}
No change ("}
No change (*}
Safety of interim storage not evaluated Occupational Dose Equivalent (b) Equivalent (b) Equivalent ( }
Lower Time for Fuel Early Latest Later Early Removal Time for. Cleanup Early Early Later Not completed in a defined Completion time Technical Feasible Feasible with Feasible Feasibility not assured Feasibility some reservations
~
(a) No significant change from that assessed in the PEIS.
(b) The current plan and Alternatives.1 and 2 were assessed to be equivalent in terms of occupational dose.
complete cleanup.
The safety of the monitored, interim-storage phase could not be evaluated at the present time, but some increased risk to the public is expected to result from aelaying final cleanup.
The major difficulty in assessing Alternative 3 sas in regard to technical feasibility.
There is little doubt that the majority of building cleanup could not reasonably be accomplished using robotic technology at the present time.
One can only speculate on what the state of robotic technology will be in the O to 20 years following defueling. The staff prefers to present Alternative 3 as an alter-native that may warrant further consideration af ter defueling is complete, but cannot be considered feasible at the present time.
2.42
s 3.0 REVISED ENVIRONMENTAL IMPACTS s
The most significant environmental impact defined in the PEIS was the radiation dose to workers during cleanup operations and offsite dose is not going to be significant.
The revision of the estimated' occupational dose was calculated for this supplement to the PEIS, based on new information regarding the difficulty of hieaning up the reactor building and the auxiliary and fuel-handling building.
x In Section 2 of this document, various alternatives for the cleanup of TMI-2 were descVibed. Occupational radiation doses were_ estimated for reactor building cleanup, a uiliary and fuel-handling building cleanup, primary-system decontamination, reactor disassembly and defueling, and dose reduction efforts.
In all cases *, a range of values was given for the occupational dose, represent-ing the uncertainty of the estimates.
This section of the supplement dis-cusses the revised occupational-dose estimates and resulting health effects.
The discussion is divided into three sections.
Section 3.1 discusses the population that would receive the occupational f dose from the cleanup.
Sec-tion 3.2 summarizes the estimated occupational doses that would result from cleanup.. Section 3.3 discusses the potential-health effects associated with those estimated occupational doses.
\\_
3.1 AFFECTED POPULATION y
The only, population group considered in this supplement is composed of members of the work force who enter radiation zones at TMI-2 while conducting cleanup operations. These workers are 18 to 70 years old (average age is 42),
in good healtlG~ and primarily male, w
a 3.2 REVISED OCCUPATIONAL-DOSE ESTIMATES Tha cumulative occupational radiation dose to complete cleanup of TMI-2 is presented in Table 3.1 for each of.the four cleanup options. As discussed in Section 2.6, the current plan e nd-Alternatives 1 and 2 are considered acceptable at this time.
Of these, the current plan represents the most probable course of action for the licensee.
Regardless of which option is chosen, three opdrations 'are responsible for 90% or more of the total occupational dose associatedN!th cleanup. These three operations are:
i y
e reactor building and equipment cleanup e reactor dis' assembly and defueling e dose reduction.
T N
g.
w The highest percentage Sof the-total dase will result from reactor building.tnd equipment cleanup.
This operation is necessary to meet the cleandps objectives.
w N
v 3.1
+
s mm g
5
~
1 TABLE 3.1.
Cumulative Occupational Radiation Dose Associated with Each Cleanup Option i
l Current Cleanup Plan Alternative 1 Alternative 2 Alternative 3 Reactor Building-5,900-21,000 9,000-30,000 5,900-21,000 300-3,500(*}
and Fquipment-Cleanup Reactor Disassembly 2,600-15,000 820-6,500 2,600-14,000 2,600,46n000 and Defueling hr, Primary-System 56-970 39-780 56-970 11-190 Decontamination 49 Dose Reduction 2,000-5,100 2,000-5,100 2,000-5,100
. Auxiliary and 500-1,400 500-1,400 500-1,400 97-1,400 Fuel-Handling Building Cleanup Utility and System 100-200 100-200 140-280 80-160 Maintenance Radioactive Waste 97-485 97-485 97-485 97-485 Management and(b)
Transportation Other
- 630(*)
140-540(d)-
0-620(*)
Dose Received To 1700 1700 1700 1700
'Date in Cleanup 13,000-46,000 12,000-42,000 13,000-45,000 6,900-28,000 l
.(a)' Includes dose to robotically complete primary-system' decontamination and to complete cleanup of the auxiliary and fuel-handling building.
(b) ~ Based on information from the PEIS.
(c). For reeleaning of the reactor building.
(d) For defueling operation prior to head lift.
l(e) For interim care of reactor building and suxiliary and fuel-handling 1
building for up to 20 years.
t 3.2 i
t
. t
i Reactor disassembly and defueling will lead to the next largest portion of the total dose. This operation is essential to the cleanup effort because it assures public safety and provides for removal of the largest quantity of radioactive material from the site.
The dose reduction program is associated with approximately 10% of the total occupational dose for the current cleanup plan and Alternative 2.
There is no separate dose reduction program under Alternative 1 because any dose reduction work done as part of this option would be included in reactor building and equipment cleanup.
For the current plan and Alternative 2, the dose reduction program will result in lower total occupational dose for cleanup than if the program were not carried out.
The dose reduction program is part of the licensee's effort to maintain occupational radiation doses ALARA.
3.3 HEALTH EFFECTS The work force for the TMI-2 cleanup will be exposed predominantly to penetrating radiation distributed over the whole body, so that any conse-quences wil! not be restricted to a particular area or organ of the body.
A great deal of data on the biological (health) effects of radiation has been accumulated on a worldwide basis over the past several decades.
These data have been analyzed by international and national organizations responsible for radiation protection (United Nations Scientific Committee on the Effects of Atomic Radiation '1977, National Academy of Sciences 1972).
The up-to-date findings of these organizations are the basis for estimating radiation-related human health effects in this document.
The occupational doses from routine operations during the course of the TMI-2 cleanup may result in somatic and genetic effects.
The somatic effect (to the body of the worker) of greatest concern is the possibility of inducing a fatal cancer; the genetic effects include a variety of inheritable changes that may affect future generations.
A wide range of factors is utilized in the nuclear industry to estimate health effects.
Although there is a finite probability of no adverse health effects, the following internationally accepted factors were used:
e 131 fatal cancers in the exposed workers per one million person-rem, e 260 genetic effects among the offspring of the work force per one million person-rem.
More detailed information on the health effect risk estimators used by the staff is contained in Appendix Z of the PEIS (Volume 2) and is reproduced as Appendix B of this document.
It should be stressed that these risks, or probabilities, are increments above or additions to those risks to which the entire population currently is exposed.
Current public health statistics show that, for the entire U.S.
population, there is a 1 in 5 probability that death will be due to some form of cancer.
The normal occurrence of hereditary disease in the offspring of the present U.S.
population is about 1 in 9.
It is expected that the 3.3
occupational dose to the work force cleaning up TMI-2 will increase the workers' risk of death from cancer, but this added risk is relatively small in comparison with the existing risk.
In addition, the risk of genetic changes can be expected to increase for the offspring of the work force, but this increment is also very small compared to existing risks.
The health effects from occupational exposure to radiation were calcu-1ated for the work force on the basis of radiation doses ranging between 13,000 and 46,000 person-rem.
For the minimum-collective-dose case (13,000 person-rem), it is expected that 2 additional fatal cancers would be caused.
For the maximum-dose case (46,000 person-rem), 6 addition:1 cancer fatalities would result.
Although it is possible to compute a range of probabilities for cheer induction among average individual workers based on the above figures, the results of such a calculation may not bear a close relationship to actual risks since the work force size and collective dose associated with the various tasks can differ by large factors, rendering inapplicable the concept of an average individual worker.
The licensee applies administrative controls for doses to its employees in order to ensure compliance with the regulations given in 10 CFR 20.
These centrols result in keeping most doses to less than 1 rem / quarter. Most of the workers involved in the cleanup can be expected to be in this category.
The regulations of 10 CFR 20 limit the highest quarterly dose that an individual worker may received to 3 rem / quarter.
Individuals are not allowed to receive exposures in excess of 1 rem / quarter unless. there are special circumstances.
For example, a complex task that would normally be done by a single worker might require several workers if the 1-rem / quarter administrative control were imposed.
In such situations, the total exposure to the work force can often be reduced if one worker is allowed to exceed 1 rem / quarter (but not the 10 CFR 20 limits) in order to complete the task.
For an individual worker who gets 1 rem / quarter throughout an assumed 9-year cleanup period, the total dose would be 36 rem.
For a person of age 30, the probability of dying of cancer from normal causes is expected to be 1 in 5.
Ihe added probability of a premature death from cancer as a result of receiving a radiation dose of 36 rem would be 1 in 210.
Thus, for the decontamination workers, the overall probability of death from cancer would be 1 in 4.9.
The equivalent decrease in life expectancy from a 36-rem dose would be about 23 days.
The risk for a younger worker would be greater, and for an older worker it would be less.
For the minimum-collective-dose case, the expected number of genetic l
effects among the offspring of the work ferce would be 3.
For the maximum-collective-dose case, the expected number would be 12.
The normal (exclusive) of occupaticnal dose)' incidence rate for a hypothetical work force of 10,000 persons would be about 1,100.
3.4
.m
4.0 CONCLUSION
S In this supplement to the Prograusnatic Environmental Impact Statement, the NRC staff has reevaluated the occupational radiation dose and the health affects associated with the proposed cleanup of Three Mile Island Unit 2.
As a result of this evaluation, the staff has reached the following conclusions:
e All options for the TMI-2 cleanup evaluated in this supplement involve occupational radiation doses higher than those predicted more than 3 years ago in the PEIS.
The basis for these revised estimates is increased knowledge of the conditions inside.the reactor building and of the effectiveness of decontamination and dose reduction efforts.
e The costs of the cleanup, in terms of environmental impacts, is in the radiation exposures and potential health effects among the cleanup j
workers.
Despite the possible increase in radiation exposures to the i
workers, the benefits of cleanup still exceed the drawbacks.
The major i
benefit of the cleanup will be the elimination of the continuing risk of potential uncontrolled releases of radioactivity to the environment from damaged fuel.or from the radioactive contamination which is distributed throughout the primary system, the reactor building, and the auxiliary and fuel-handling buildings.
It is the staff's judgment that the conclusion of the 'PEIS that " cleanup should proceed as expeditiously as reasonably possible to reduce the potential for uncontrolled releases of radioactive materials to the environment" remains valid.
e Another benefit the cleanup would accure is the additional knowledge that would be useful for reducing the risks and consequences of possible future accidents at nuclear power plants.
This earlier PEIS conclusion remains valid.
While considerable information has already been obtained in the cleanup to date, much more data remains to be obtained as the focus of the cleanup is directed towards reactor disassembly and defuel-ing.
The information to be obtained increases the understanding of fis-sion product behavior resulting from severe accidents, the metal-water reaction and the corresponding generation of hydrogen, the management of very highly contaminated liquid and solid radioactive waste, the manage-ment of gaseous radioactive vaste, decontamination methodology and techniques, radiological and physical protection of workers in highly contaminated areas, and radiation and environmental effects on materials and equipment.
It could be applied to current and planned nuclear power facilities in a variety of areas including plant and equipment layout and design, accident mitigation system design, instrument location and design, radioactive waste processing system design, surface coatings for contamination control and mitigation of fission product releases from severe accidents.
e The only means identified in this supplement for substantially reducing the occupational dose is the extensive use of robotic technology.
Under any cleanup plan that makes use of this technology, the feasibility of completing the cleanup will depend on developments in robotics, which are 0
uncertain at this time.
Because the highest dose is associated with 4.1 4
5
- - -. - - - -. +.
i reactor building and equipment cleanup, which is expected to follow defueling, a decision to adopt this approach can be deferred until there are further developments in robotic technology.
e Decontamination workers at the plant will receive a total collective radiation dose estimated at between 13,000 and 46,000 person-rem for the whole cleanup program.
Using the internationally accepted health effect risk estimators, the health effect estimates corresponding to these doses range from 2 to 6 additional deaths among these workers due to cancer and from 3 to 12 additional genetic effects among their offspring.
These ranges are broad because of uncertainties about the plant conditions and about the amount of work that will be needed to decontaminate the reactor building and its contents.
e The occupational radiation dose to an individual worker will be limited to less than 3 rem / quarter in accordance with 10 CFR 20.
Based on current experience, most workers will receive radiation doses substan-tially below that limit.
e The most dose-intensive task is reactor building and equipment cleanup, unless this task is done using robotic technology.
An early decision to use robotics is not necessary as long as the licensee defuels the reactor before reactor building cleanup.
e The current plan provides the most likely path for early fuel removal.
Extensive building cleanup before defueling, or the r.odification of defueling methods, would cause substantial, unwarranted delays in fuel removal, with attendant risks.
e The dose reduction program has substantial potential for lowering the total radiation dose to workers during the cleanup. ALARA considerations indicate that a significant commitment of funds and managerial emphasis should continue to be placed on this effort.
o Reactor building cleanup concurrent with defueling can also 1 expected to reduce the occupational dose by removing sources of radiation exposure from the work place.
Other conclusions of the PEIS that do not pertain to occupational radiation dose remain valid.
The staff concludes that the cleanup should proceed as expeditiously as possible while ensuring the health and safety of the workers and the public. All work performed as part of the cleanup should be done in a manner that keeps occupational doses as low as is reasonably achievable.
4.2
~
5.0 REFERENCES
Alvarez
.J. L.
April 1, 1983. " Air Sampling in TMI Containment for Estimates of Recontamination." JLA-2-83, Interoffice memo to T. E. Cox, EG&E Idaho.
- Brooks, B.
G.
1982.
Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1981.
Annual Report.
NUREG-0713, Vol. 3, U.S.
Nuclear Regulatory Commission, Washington, D.C.
Card,- C.
J.
1983.
Postaccident Decontamination of Reactor Primary Syalems and Test Loops. EPRI NP-2842, Electric Power Research Institute, Palo Alto, ICalifornia.
Code of Federal Regulations.
1982. Title 10 Part 20, " Standards for Protec-tion Against Radiation."
'Cunnane, J.
C., and Nicolosi, S. L.
1983. Characterization of the Contamina-tion in the TMI-2 Primary Coolant System.
EPRI NP-2722, Electric Power Research Institute, Palo Alto, California.
- Daniel, J.
A.,
T.
L.
- McVey, E.
A.
- Schlomer, D.
G.
Keefer and J.
E.
Cline.
1983.
Characterization of Contaminants in TMI-2 Systems.
EPRI NP-2922, Electric Power Research Institute, Palo Alto, California.
- Devine, J.
C.
1983.
Planning Study:
Defueling Water Cleanup System.
TP0/TMI-046 Rev 0.
- Flanigan, J.
A.
1983.
TMI Unit II Reactor Building Radiological Status.
July 28, 1983.
GPU Nuclear.
August 25, 19d3.
Radiological Controls Procedure 4015 Revision 4.
Gardner, H.
R.,
R.
P.
- Allen, L.
M.
- Polentz, W.
E.
Skines and G.
A.
Wolf.
1983. Evaluation of Non-Chemical Decontamination for Use on Reactor Coolant Systems.
EPRI NP-2690, Electric Power Research Institute, Palo Alto, California.
Kanga, B. K.
March 30, 1983.
"TMI-2 Program Reassessment: Man-Rem Estimate, January 1983." Letter report to B. J. Snyder, TMI Program Office.
Metropolitan Edison Co. and Jersey Central Power & Light Co.
1974.
Final Safety Analysis Report, Three Mile Island Nuclear Station, Unit 2.
- Munson, L.
F.
September 29, 1983.
" Doses Received in Cleanup of TMI-2."
Letter to R. Lo, TMI Program Office.
Munson, L.
F.,
C. J. Card and J. R. Divine.
1983.
An Assessment of Chemical Processes for the Postaccident Decontamination of Reactor Coolant Systems.
NP-2866, Electric Power Research Institute, Palo Alto, California.
5.1
National Academy of Sciences, Advisory Committee on the Biological Effects of Ionizing Radiation (BEIR).
1972. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation.
National Research Council, Washington,_
i D.C.
Runion, T. C.
1983.
Proposed Methods for Defueling the TMI-2 Reactor Core.
Report from Debris Defueling Working Group to GPU Nuclear.
- Sejvar, I.,
and P.
H.
Dawson.
1982.
Evaluation of Abrasive Grit--High-Pressure-Water Decontamination.
EPRI NP-2691, Electric Power Research Institute, Palo Alto, California.
United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR).
1977. Sources and Effects of Ionizing Radiation. Report to the U.N. Cencral Assembly, New York.
U.S. Atomic Energy Commission.
1976.
Final Environmental Statement Related to Operation of Three Mile Island Nuclear Station Units 1 and 2.
AEC Docket Nos. 50-289 and 50-320.
U.S. Nuclear Regulatory Commission.
1978.
Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable.
Regulatory Guide 8.8, NRC Office of Standards Development, Washington, D.C.
U.S. Nuclear Regulatory Commission.
1981.
Final Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting from March 28, 1979, Accident Three Mile Island Nuclear Station, Unit 2.
NUREG/0683 Vol. 1. Washington, D.C.
l 1
l l
5.2
APPENDIX A CONTRIBUTORS TO THE SUPPLEMENT l
APPENDIX A CONTRIBUTORS TO THE SUPPLEMENT The overall responsibility for the preparation of this statement was assigned to the Three Mile Island Program Office of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission.
The statement was prepared by members of the TMI Program Office with substantial assistance from other NRC components, the Pacific Northwest Laboratory, and other consultants indicated below. The individuals who were major contributors are listed below with the affiliations or expertise:
MME AFFILIATION FUNCTION OR EXPERTISE NRC Ronnie lo TMI Program Office Project Manager Bernard J. Snyder TMI Program Office Director Lake Barrett TMI Program Office Deputy Director Richard Weller TMI Program Office Nuclear Engineering John Nehemias Radiological Assessment Branch Radiological Effects Frank Congel Radiological Assessment Branch Radiological Effects Kimberly Barr Inspection and Enforcement Radiation Specialist Barry O'Neill Inspection and Enforcement Radiation Specialist Michael Wangler Radiation Assessment Branch Radiological Effects Pacific Northwest Laboratory (a)
Glenn R. Hoenes Radiological Sciences Program Manager (PNL)
Linda F. Munson Radiological Sciences Project Leader (PNL)
Leo H. Munson Radiological Sciences Health Physics l
(a) The Pacific Northwest Laboratory is operated for the Department of Energy l
by the Battelle Memorial Institute.
I l
A.1
1.
NAME AFFILIATION FUNCTION OR EXPERTISE Pacific Northwest Laboratory (continued)
George J. Konzek
, Energy Systems Decontamination Jolene C. Juneau Radiological Sciences Engineering.
Greg F. Martin Radiological Sciences Health Physics Carl M. Unruh Radiological Sciences Senior Reviewer Edwin C. Watson Radiological Sciences Environmental Science Thomas H. Essig Radiological Sciences Health Physics John G. Meyers Consultant Health Physics Other Consultants-Valmore Bouchard VIKEM Decontamination Rudolph Nelson VIKEM Decontamination A.2
I-APPENDIX B HEALTH EFFECTS ESTIMATORS (*}
(a) From PEIS, Appendix Z.
APPENDIX B i
HEALTH EFFECTS ESTIMATORS (*
In estimating the number of health effects resulting from both offsite and occupational radiation exposures during the cleanup, the NRC staff used best estimate somatic (cancer) and genetic risk estimators based on widely accepted scientific. information.
Specifically, the staff's estimates are i
based on information compiled by the. National Academy of Science's Advisory Committee on'the Biological Effects of Ionizing Radiation (BEIR).I Although a detailed discussion of the literature available on this subject is outside the scope of this document, this appendix includes:
- 1) information rhat details the bases for health effect estimators used by the NRC staff, and 2) perspec-tive on the uncertainty associated with estimating radiation-induced health effects.
l
-The base data used for the fatal cancer risk estimators (expressed as deaths per _ million person-rem) used by the NRC staf f can be found in Sec-tion 9.3.2, " Upper Bound for Latent Cancer Fatalities," in the NRC staff's Reactor - Safety Study, WASH-1400, October 1975.2 Specifically, the data.on
" Upper Bound. Risk Coefficients for Latent Cancer Fatalities," Table VI 9-2, a
for specific - age groups were used.
Tables B.1 through B.8 contain, for different cancer types, the age, specific data and calculations which, when summed in Table B.9, yield the cancer risk estimators used by the NRC staff (131 and 135 deaths per million person-rem for workers and individual members of the public, respectively)..-The WASH-1400 (October 1975) coefficients are based on BEIR, 1972, and on new data made available since the issuance of BEIR, 1972. Table B.10 is. presented to provide perspective on the uncertainty involved in making estimates of radiation-induced health effects.
The basis
_ for each of these estimates can be found in greater detail in the listed references.
The N'RC staff's genetic risk estimator (260 genetic effects per million total-body person-rem in the future generations of the exposed population) was derived from the 1972 BEIR report.
Specifically, this value can be calcu-I lated by summing the geometric means of the distributions given in Table 4, Chapter V, of that report. The geometric mean is an appropriate technique for obtaining a representative value for the 1972 BEIR report's range of 60 to
~
1500 genetic effects _ per million person-rem.
The 1980 BEIR report listed a
' range of 60 to 1100 genetic effects in offspring per million person-rem.3 In summary,.several points should be emphasized. The values utilized by the NRC staff for estimating potential health effects associated with the decontamination of TMI-2 are based on widely accepted scientific information.
Even if. upper-range BEIR, 1980, risk estimators were used to characterize potential health effects from the cleanup, those health effects would remain small compared to natural incidence.
4 (a) From PEIS, Appendix Z.
i B.1 a-
.-,,.,n.
~-1+
References 1.
"The Effects on Populations of Exposure to Low-Levels of Ionizing Radia- -
tion," National Academy of Science, Advisory Committee on the Biological
_ Effects of Ionizing Radiation (BEIR), November 1972.
2.
- " Reactor : Safety Study," U.S.
Nuclear. Regulatory Commission, WASH-1400, October 1975.
3.
"The Effects on-Populations of Exposure to Low-Levels of Ionizing Radia-tion," National Academy of Science, Advisory Committee on the Biological Effects of Ionizing Radiation (BEIR), 1980.
4.
" Sources and Effects of Ionizing Radiation," United Nations Scientific Committee on. the Effects of Atomic Radiation (UNSCEAR), 1977 Report. to the General Assembly, New York, NY, 1977.
-TABLE B.1.
Calculation of Expected Leukemia Deaths for External Exposure
-(A)
(B)
(C)
(D)
(E)
(F)
Age Life Laten Years Risk Expected Period {,)
Cohort Fraction of Expectancy at Factor Deaths (years) Population (years)
(years)
Risk (10 6/ rem) (b)
(F=AxDxE)
In utero 0.011 71.0-0 10 15 1.65 0-0.99 0.014 71.3
_2 25 2
0.70 L
1-10 0.146.
69.4 2
25 2
7.3 11-20 0.196 60.6 2
25 1
4.9 21-30 0.164
'51.3 2
25 1
4.10 31-40 0.118 42.0 2
25 1
2.95 41-50 0.109 32.6 2
25 1
2.75 51-60 0.104 24.5 2
22.5 1
2.34 61-70 0.080 17.1 2
15.1 1
1.21 71-80 0.044 11.1 2
9.1 1
0.40 80+
0.020 6.5 2
4.5 1
0.09 28.4 (23.2)(C)
(a) The latent period is that period of time between radiation exposure and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes
._ consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
. population age group distribution.
B.2 r
.. ~.
.-.--e..-
. _ - - -. -, ~. - - - _
TA3LE B.2.
Calculation of Expected Lung Cancer Deaths for External Exposure (A)
(B)
(C)
(D)
(E)
(F)
Laten Years Risk Expected Age Life Period {)
at Factor Deaths Cohort Fraction of Expectancy (years) Population (years)
(years)
Risk (106/ rem)(
(F=AxDxE)
Ir1 utero 0.011 71.0 0-0.99 0.014 71.3 1-10 0.146 69.4 11-20 0.196 60.6 15 30 1.3 7.64 21-30 0.164 51.3 15 30
~1.3 6.40 31-40 0.118 42.0 15 27 1.3 4.14 41-50 0.109 32.6 15 17.6 1.3 2.49 51-60 0.104 24.5 15 9.5 1.3 1.28 61-70 0.080 17.1 15 2.1 1.3 0.22 71-80 0.044 11.1 15 0
1.3 0
80+
0.010 6.5 15 0
1.3 0
22.2 (25.3)( )
(a) The latent period is that period of time between radiation exposure and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
population age group distribution.
B.3
TABLE B.3.
Calculation of Expected Stomach Cancer Deaths for External Exposure (A)
(B)
(C)
(D)
(E)
(F)
Age Life Laten Years Risk Expected Period {)
Cohort Fraction of Expectancy Factor (b)
Deaths at (years) Population (years)
(years)
Risk (10 6 / rem)
(F=AxDxE)
Ir1 utero 0.011 71.0 0-0.99 0.014 71.3 1-10 0.146 69.4 11-20 0.196 60.6 15 30 0.6 3.53 21-30 0.164 51.3 15 30 0.6 2.95 31-40 0.118 42.0 15 27 0.6 1.91 41-50 0.109 32.6 15 17.6 0.6 1.15 51-60 0.104 24.5 15 9.5 0.6 0.59 61-70 0.080 17.1 15 2.1 0.6 0.10 71-80 0.044 11.1 15 0
0.6 0
80+
0.020 6.5 15 0
0.6 0
10.2 (11.7)(c)
(a) The latent period is that period of time between radiation exposure and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70).
The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
population age group distribution.
B.4
l TABLE B.4.
Calculation of Expected Pancreas and Alimentary Canal Cancer Deaths for External Exposure (A)
(B)
(C)
(D)
(E)
(F)
Laten Years Risk Expected Age Life Period {,)
Deaths Factor ( }
Cohort Fraction of Expectancy at (years) Population (years)
(years)
Risk (106/ rem)
(F=AxDxE)
Irt utero 0.011 71.0 0-0.99 0.014 71.3 1-10 0.146 69.4 4
11-20 0.196 60.6 15 30 0.2 1.18 21-30 0.164 51.3 15 30 0.2 0.98 31-40 0.118 42.0 15 27 0.2 0.64 41-50 0.109 32.6 15 17.6 0.2 0.38 51-60 0.104 24.5 15 9.5 0.2 0.20 61-70 0.080 17.1 15 2.1 0.2 0.03 71-80 0.044 11.1
-15 0
0.2 0
80+
0.020 6.5 15 0
0.2 0
3.4
_(3.9)(*)
(a) The latent period is that period of time between radiation exposure and the manifestation of a cancer.-
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
population age group distribution.
l B.5
TABLE B.S.
Calculation of Expected Breast Cancer Deaths for External Exposure (A)
-(B)
(C)
(D)
(E)
(F)
Age Life Laten Years Risk Expected Period {,)
Cohort Fraction of Expectancy Factor (b)
Deaths at 6/ rem)
(F=AxDxE)
(years) Population (years)
(years)
Risk (10 In utero 0.011~
71.0 0-0.99 0.014 71.3 1-10 0.146 69.4 11-20 0.196 60.6 15 30 1.5 8.82 21-30 0.164 51.3 15 30 1.5 7.38 31-40 0.118 42.0 15 27 1.5 4.78 41-50 0.109 32.6 15 17.6 1.5 2.88 51-60 0.104 24.5 15 9.5 1.5 1.48 61-70 0.080 17.1 15 2.1 1.5 0.25 71-80 0.044 11.1 15 0
1.5 0
80+
0.020 6.5 15 0
1.5 0
25.6 (29.2)(c)
(a) The latent period is that period of time between radiation expo ore and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk ectimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years).
The offsite population estimator is based on the 1970 U.S.
population age group distribution.
T B.6
TABLE B.6.
Calculation of Expected Bone Cancer Deaths for External Exposure (A)
(B)
(C)
(D)
(E)
(F)
Laten Years Risk Expected Age Life Period {)
Deaths Factor ( }
Cohort Fraction of Expectancy at (years) Population (years)
(years)
Risk (106/ rem)
(F=AxDxE)
~In utero 0.011 71.0 0-0.99 0.014 71.3 10 30 0.4 0.17 1-10 0.146 69.4-10 30 0.4 1.75 11-20 0.196 60.6 10 30 0.4 2.35 21-30 0.164 51.3 10 30 0.2 0.98 31-40 0.118 42.0 10 30 0.2 0.71 41-50 0.109
'Mr.9L 10 22.6 0.2 0.49 51-60 0.104 24.5 10 14.5 0.2 0.30 61-70 0.080 17.1 10 7.1 0.2 0.11 71-80 0.044 11.1 10 1.1 0.2 0.01 80+
0.020 6.5 10 0
0.2 0
8 6.9 (4.5)(")
(a) The latent period is that period of time between radiation exposure and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
population age group distribution.
i i
B.7 I
TABLE B.7.
Calculation of Expected Thyroid Cancer Deaths for External Exposure (A)
(B)
(C)
(D)
(E)
(F)
Age Life Laten Years Risk Expected Period {,)
Cohort Fraction of Expectancy Factor (b)
Deaths at (years)
Population (years)
(years)
Risk (10 6 / rem)
(F=AxDxE)
~In utero 0.011 71.0 0-0.99 0.014 71.3 10 30 4.3 x 1.0 0.181 1-10 0.146 69.4 10 30 4.3 x 1.9 3.58 11-20 0.196 60.6 10 30 4.3 x 1.6 4.05 3
21-30 0.164 51.3 10 30 4.3 x 1 2.12 31-40 0.118 42.0 10 30 4.3 x 1 1.53 41-50 0.109 32.6 10 22.6 4.3 x 1 1.06 51-60 0.104 24.5 10 14.5 4.3 x 1 0.648 I
61-70 0.080 17.1 10 7.1 4.3 x 1 0.244 71-80 0.044 11.1 10 1.1 4.3 x 1 0.021 80+
0.020 6.5 10 0
4.3 x 1 0
12.4 (*)
(9.7)
(a) The latent period is that period of time between radiation exposure and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
population age group distribution.
B.8
TABLE B.8.
Calculation of All Other Expected Cancer Deaths for External Exposure (A)
(B)
(C)
(D)
(E)
(F)
Age Life Laten Years Risk Expected Period {,)
Cohort Fraction of Expectancy at Factor Deaths (years) Population (years)
(years)
Risk (10 6/ rem)(b)
(F=AxDxE)
Ijl utero 0.011 71.0 0
10 15 1.65 0-0.99 0.014 71.3 15 30 0.6 0.25 1-10 0.146 69.4 15 30 0.6 2.63 11-20 0.196 60.6 15 30 1
5.88 21-30 0.164 51.3 15 30 1
4.92 31-40 0.118 42.0 15 27 1
3.19 41-50 0.109 32.6 15 17.6 1
1.92 51-60 0.104 24.5 15 9.5 1
0.99 61-70 0.080 17.1 15 2.1 1
0.17 71-80 0.044 11.1 15 0
1 0
80+
0.020 6.5 15 0
1 0
21.6 (19.5)(c)
(a) The larent period is that period of time between radiation exposure and the manifestation of a cancer.
(b) Error in labeling corrected from the original PEIS Appendix Z.
(c) Values in parentheses represent risk estimates for occupational workers (ages 20-70). The risk estimator (see Table B.9) for workers includes consideration of the workers' age grouping (estimated between 20 and 70 years). The offsite population estimator is based on the 1970 U.S.
population age group distribution.
l' B.9
TABLE B.9.
Maximum Expected Latent Cancer Deaths Per Million Person-rem 6 Person-Rem Expected Deaths /10 Type of Cancer General Public Occupational Leukemia 28.4 23.2 Lung 22.2 25.3 Stomach 10.2 11.7 Alimentary Canal 3.4 3.9 3.4 3.9 Pancregg)
Breast 25.6 29.2 Bone 6.9 4.5 Thyroid (b) 13.4 9.7 All others 21.6 19.5 135 131 (a) Assumes 50 percent mortality / case.
(b) Assumes 10 percent mortality / case; all other types assume 100 percent mortality.
B.10
TABLE B.10.
Comparison of Fatal Cancer Risk Estinators Cancer Mortality Estimators Source (deaths /106 person-rem)
NRC staff (PEIS) 135(*}
BEIR, 1980(b)67-169 BEIR, 1972(C) 115-568 UNSCEAR, 1977(d)75-175 (a) Risk es6imator used for members of the public.
For workers, a risk estimator of 131 deaths /
106 person-rem was used. This value accounts for worker age-specific (20-70) radiosensitivity.
(b) Linear-quadratic dose-response model for absolute and relative projection models. These values represent the BEIR committee's stated best estimate. However,'the committee also pointed but that the linear and pure qusdrate effects models also fit observed data nearly as well.
Projected health effects from those models would range from about 10 to 500 deaths per million person-rem. An update of BEIR, 1972.
(c) Values obtained from Table V-4, BEIR, 1980, are an updata of values obtainable in Table 3-3 and 3-4 of BEIR, 1972. Range attributable to dif-ferences between absolute and relative projection models.
(d) Range of estimates for low-dose, low-LET radia-tion (UNSCEAR 1977).
- U.S. Gvv aeu.n.a s PRINTING OFFICE: 1983-421-299:192 B.11
NRCe R" 335 1, REPOR7 NUMTE R #Assees ey CoCJ u.s. NUCLEAR RecutAToRv commissio" Nureg - 0683 BIBLIOGRAPHIC DATA SHEET Supplement 1 4 TITLE ANo susTiTLE <Aaa voume lvo. er eearer'e=> PROGRAMMATIC ENVIRONMENTAUte=e *'*
IMPACT STATEMENT Related to Decontamination and Disposal Accident Three Mile Island Nuclear Stationof Radioactive Wastes Resulting from March 28,it 2 docket 1979
- 3. RaciP:ENT S AccESSICW No.
Un 50-320 DRAFT SUPPLEMENT DEAT TNC WTTH nerffbATinMAY
- 7. AUTHORIS, MAULATION DOSE
- 5. oATE REPORT coMPLE TED TMI Program Office Ev'e"mber I'I983 9 PERFORMING ORGANIZATION NAME AND MAILING AoORESS (lacAvar I,o Codel D ATE REPORT ISSUED TS*6i&ber I"
T*fI Program Office 1983 s < tee., a,so a stes e wa..
12 SPONSCRING ORGANIZ ATiON N AME AND M AILING ADORESS (tactwee 2,o Codes pq TMI Program Office U. S. Nuclear Regulatory Commieeion
" "N NO-Washington, D. C. 20555 13 TYPE OF REPORT et R800 coV E =f D Itac'esme ese:A Draft Supplement to EIS 7/83 - 10/83 15 $UPPLEVENTARY NOTES 14 ltene **'4 J 16 AeSTR ACT 1200 naves or sesso In accordance with the National Environmental Policy Act, the Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Waste for the 1979 Accident at Three Mile Island Nuclear Station Unit 2 has been supplemented. The supplement was required because current information indicates that cleanup will entail substantially more occupational radiation dose to the cleanup work force than originally anticipated. Cleanup was originally estimated to result in from 2000 to 8000 person-rem of occupa-tional radiation dose. Although only 1700 person-rem have resulted from clean-up operations performed up to now, current estimates now indicate that between 13,000 and 46,000 person-rem are expected to be required. Alternative cleanup methods considered in the supplement either did not result in appreciable dose savings or were not known to be technically feasible.
17 (EY WOROS ANo DOCUMENT AN ALYSIS 17e CESCRIPTORS 17b IDENTIFIERS CPEN ENCEO TE4VS 16 Av AIL A8iLITY ST ATEVENT E
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