Text

Provides Subcommittee on Nuclear Regulation W/Responses to Questions in Re Facility Decontamination Program Plan.Compares Four Decontamination Program Cases,Developed by Nrc,Based on Util & Bechtel Preliminary Plans
	ML19322D802
Person / Time
Site:	Crane
Issue date:	02/04/1980
From:	Ahearne J; NRC COMMISSION (OCM)
To:	Hart G; SENATE, ENVIRONMENT & PUBLIC WORKS
Shared Package
ML19322D803	List: ML19322D802; ML19322D804;
References
	NUDOCS 8002290165
	Download: ML19322D802 (33)
	v • d • e

{{#Wiki_filter:_ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. / UNITED STATES s' ' t, NUCLEAR REGULATORY COMMISSION E {pkh [ j WASHINGTON. D. C. 20555 g\\,,,,./ E February 4,1980 g CHAIRMAN 91 5 il r b The Honorable Gary Hart, Chairman E Subcommittee on Nuclear Regulation i Committee on Environment and Public Works United States Senate Washington,.D.C. 20510

Dear Mr. Chairman:

F" This letter is in response to your November 20, 1979 request for fj information about the program plan for the decontamination of TMI-2; a listing of equipment and technologies needed to accomplish the g plan; the appropriate regulatory guidelines to be used; steps necessary j to reestablish public confidence in the cleanup operation; and NRC's ability to regulate these operations. Enclosed are the responses to your specific questions. I hope this provides you with the information you desire. If you [ should require additional information or clarification of the informa- [ tion presented, please do not hesitate to contact me. t V \\ n'erely, j Si c ( i A John F. Ahearne

Enclosures:

[ .1. Response to Questions 2. TMI, Unit 2 Radiation Protection Program Report { Report of the Special Panel J NUREG-0640 f .cc: Senator Alan Simpson i k = l \\ (c5 8002290 A ' -J

o-RESPONSES TO QUESTIONS IN NOVEMBER 20, 1979 LETTER FROM SENATOR GARY HART AND SENATOR ALAN K. SIMPSON L. g

-00ESTION 1: Please provide us with your best-estimate plan for the clean-1 up. The plan should identify cquipment or techniques that have been selected or that are being considered, along with the rationale used in their selection. The plan should also be coordinated with and reflect the views of the General Public Utility Service Corp. ANSWER: Plans for the cleanup Because of the inpact on schedule that could result from environmental reviews and subsequent equipment and operational restrictions, four decontamination program cases were compared to bound the likely duration of the decontamina-tion process. These cases were developed by the NRC staff based on appro-priate parts of the preliminary plans prepared by General Public Utilities (GPU) and its contractor, Bechtel Power Corporation. Phase 1 covers the decon-mination and cleanup of the reactor building. Phase 2 covers the removal of fuel and the decontamination of the reactor coolant system. The following four cases are compared in Enclosure 1: 1. The first or base case was developed from preliminary schedules provided by GPU and its contractor, Bechtel Power Corporation. For this case, we ~ assumed no delays attributed to environmental reviews by the NRC. 2. The second or Environmental Assessment case was developed by assuming it will take five months for preparation of the Environmental Assess-ment, for obtaining public corcents, and for reaching a decision regarding the acceptability of the methods proposed by the licensee for the removal of the contaminated gas from the reactor building. - Also assumed for this case is that the proposal for venting the gases would meet the Commissions regulations in 10. CFR 50, Appendix I, anc the existing effluent technical specifications for normal operations and would, th'erefore, be found acceptable. (See the_ answer to Question 2 for further information on regulatory standards and criteria).

o We further assumed that a commitment of funds for purchase and installation of any necessary facilities for the processing of the contaminated liquid would not be made until the proposed processing method was approved. 3. The third or cryogenic case studied assumed that venting of the waste gas would not be acceptable and that, at the conclusion of the Environmental Assessment, the licensee would be directed to use alternate means of removing the gases from the reactor building. For this study, we assume that cryogenic processing of the gases would be the preferred method. s 4. The fourth, or Environmental Impact Statement case, assumed that any work on decontaminating the. containment building would cease pending preparation and approval of the Programmatic Environmental Impact Statement. For this case, we assumed that the Commission has determined that it is in the best interest of the public to conduct an Environmental Review concerning the installation of the contaminated liquid processing facilities and on that basis would approve that action. However, we assumed no contaminated liquid or gases would , be allowed to be removed from the reactor building until the Final ' Environmental Impact Statement was published. If the removal of ~ the contamin ed gases and liquids from the reactor building is based only on approval of their respective Environmental Assessments, then the Enviornmental Assessment case a'nd the Environmental + Impact Statement (EIS) case would be identical. As seen in Enclo-sure 3, the Er.vironmental Assessment case adds about four months t'o the 'l ~ base schedule and the EIS case adds an additional five months to that. The cryogenic case adds about, twentyfour renths to the base schedule. ! In developing the plan and schedule, we reviewed the major work efforts to find the critical path work items, i.e., the ones that appear to con-trol the schedule. Many other items have to proceed at the same time, and these have to be completed.in a timely manner to' support the main effort. It appears to us that the first critical path item required for expeditious cleanup is early entry into the reactor building to. determine to what extent the interior is contaminated and damaged. This step is necessary in order to plan for decontamination and modifications necessary to remove the reactor head. Since the contaminated gases currently emit a high level of beta radiation, removal of these gases-is necessary before personn'el can remain in the containment structure for prolonged periods, even with protective clothing. Therefore, the initial step is removal of the gases from this structure. In November the licensee provided a proposed plan for the removal of the gases by venting to the atmosphere. We have started our evaluation of this proposal.and of the alternatives to 'venhing.. ~ When the gas removal method 'is approved, the gases would be removed and. entry- 'to 'the r'eactor ' building for limited periods would be made to take radiation \\ ~ measurements and. samples hnd to assess the damage. ~ t a Y 9 Tnis would be followed by an evaluation of the results and by the development of the detailed methods for reactor builoing decontamination. A TV camera has provided limited visual examination of the reactor build-ing. Localized radiation measurements have also been made. The results of these examinations are currently being evaluated. In order for personnel to work for extended periods of time inside the reactor building, most of the ccitaminated sump water would also have to be removed. We expect the licensee to submit a proposal on the decontami-nation and removal of the water from the reactor building sump early in 1980. Following cleanup of the atmosphere and sump water, the next critical path item would be the decontammination of the interior walls and equipment surfaces within the reactor building. Some flushing of the system's piping may be necessary. These actions complete Phase 1 of Enclosure 1. The next step would require the installation of lo.alized shielding and' rigging for removal of the head, and repair of any damage to structures required to support fuel removal. Before work proceeds, a Safety Evaluation wi]l be prepared that discusses proposed methods for removing the vessel head,'for the fuel removal, and for fuel packaging. Following NRC approval .of this activity, the head would be removed and the fuel inspected. The fuel would then be removed, canned if necessary, and placed in the fuel 4

, pool. The reactor coolant systtm would then be decontaminated. These actions complete Phase 2 of Enclosure 1. These controlling items would necessarily proceed in conjunction with other i tems. For exarple, water storage facilities must be designed, fabricated, and installed. A service structure must be designed and constructed that permits entrances and exits to the reactor building. The licensee is proceeding with the planning and engineering for these items and is currently preparing a detailed decontamination and recovery plan. As notea above, GPU has submitted to NRC a Summary Technical Plan, which will be expanded in the detailed technical plan. The licensee expects to have the plan of the first two phases, the preparation of which we will closely follow, completed by the second quarter of 1980. A number of assumptions have been made in the preparation of our plan. The major ones are outlined in Enclosure 2. Such factors as: the capabili-ity of the licensee to fund the full extent of authorized decontamination activity; the applicability of current regulatory criteria in meeting the demands of the public interest in the locale affected by the TMI-2 accident; and the physical condition of the reactor building and of the fuel, could have a major impact on the schedule and the way in which ~ decontamination could be performed. It was also assumed that no hearings would be held' for any steps during Phase 1 and Phase 2. This assumption, does not in any way affect the likelihood of whether or not a hearing will be held. Rather, it is a simplifying assumption made in order to make ' scheduling estimates independent of the time uncertainties inherent in a- . hearing. A hearing on all or on some portion of the plan could have a substantial effect on the schedule. Further, not all of the intrerciated work effort has be'en evaluated. Other items could have an it. pact that will become evident as the detailed planning proceeds. +

, In summary, our best estimate of the cleanup schedule for these four cases is as follows: 1. Base Case 38 months 2. Environmental Assessment Case 42 months 3. Environmental Impact Statement Case 47 months 4. Cryogenic Case 62-months Equipment and technioues considered For decontaminating the contaminated water currently in the reactor and auxiliary building, decontamination systems based on filtration and ion exchange have been selected. These types of systems provide very, effective means of removing radioactive materials from water. For example, Metropolitan Edison Company has been using the EPICOR-II system, a system based on filtration and de-ionization, to decontaminate the waste water in the auxiliary building. To date, decontamination has proceeded better than expected and the resultant exposure.s to workers and the off-site population have been lower than those predicted in our Environmental Assessment of the system. The licensee is considering use of an evaporator to decon-taminate waste water produced during the decontamination of the reactor building because checicals and detergents that may be used during decon-ta'mination would plug and generally interfere with a de-ionizer. These - types of contaminants and radioactivity can be removed in an evaporator. i More detailed information on these systems is provi~ded in Enclosure 5. 6 0

, On November 13, 1979, the licensee proposed that at:20 spheric venting be used to re' move the contaminated gases (mainly, krypton 85) from the reactor building. The applicant considered various alternatives anc concluced that the venting operation can be done without significant hazaros to persons on or off the site. One advantage to venting the gases is that it can be accomplished in a much shorter time thar. the.other alternatives considered. We are currently preparing an Environmental Assessment of the licensee's proposals. Additional information on the alternatives is provided in. The above discussion covers the items that are of immediate concern, namely, the decontamination of water and the release of contaminated gases from the reactor building. The design and evaluation of methods for the decontamination have progressed to the point where systems have either been approved, proposec, or are in the final stages of evaluation prior to s J submittal to us. For other decontamination requirements, such as the decontamination of the reactor building, preliminary studies have been prepared. However, selection of methods and techniques for these cannot be made until the reactor building is entered and data collected. The tech-niques currently under consideration for conducting this type of decon-timination are discussed in Enclosure 5.

Views of G:neral Public Utilities Corp. GPU submitted to NRC on December 12, 1979, a Summary Tech 6ical Plan for the iMI-2 Decontamination on Defueling (see Enclosure 6). The plan is more detailed than this NRC-prepared plan and provides GPU views on all items in the recovery. We have reviewed the "GPU Sumary Technical Plan for the THI-2 Decontamination and Defueling" and find that it is generally compatible ".A.with the NRC plan in its technical approach and timing, altho 6gh the GPU plan does not include the enviroemental review of options considered by the NRC. PUESTION 2: Please include the regulatory guidelines necessary for the cleanup. For instance, if needed, set standards for the release of radioactive effluents and emissions to the local environment. State whether existing radiation worker dose limit criteria will be employed or whether new limits will be set. The Subcomittee sho~ ld be kept advised of the pro-u gress being made toward satisfying the requirements of the i National Environmental Policy Act, and any impact that this j may have on the Comission's ability to take actions neces-sary to protect the public health. Reculatory cuidelines necessary for cleanup The regulatory criteria and guidelines that will be used for the recovery program at TMI are those embodied in the regulations listed below. We intend to solicit public coment, within the context of the. draft programmatic environmental impact statement for the TMI decontamination and cleanup activities, on ) whether these limits, which were developed for effluents resulting from 1. normal operations,.are appropriate for the TMI cleanup activities in light of dhe ' differences in the volume and duration of the release of j s'uch' effluents. 10 CFR Part 20, Appendix B, " Concentration in dir and Water Above Natural Background i 10 CFR Part 20, Paragraph 50.34a, " Design Objectives for Equipment to. Control Releases of Radioactive Material in Effluents - Nuclear - Power Reactors" ,~c

-g-10 CFR Part 50, Paragraph 50.36a, " Technical Specifications on Effluents from Nuclear Power Reactors" 10 CFR Part 50, Appendix A, ' General Design Criteria for Nuclear Power Plants" 10 CFR 'Part 50, Appendix I, " Numerical Guides for Design Objectives and Limiting Conditions to Meet tre Criterion As low As Is Reasonably Achieveable for Radioactive Material in Light-Water Cooled Nuclear Power Reactor Effluent." 40 CFR Part 190, " Uranium Fuel Cycle Standard" (Environmental Protection' Agency Regulations) Regulatory Guide 8.8 (Rev. 3), "Information Relevant to Ensuring that Occupational Radiation Exposure's at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition" Radiation worker dose-limit criteria ~ As a minimum, we vill require the licensee to adhere to the radiation dose limits specified in 10 CFR Part 20, Paragraph 20.101, " Exposure of Individuals to Radiation in Re;tricted Areas." In addition, we require the licensee to make every reasonable effort to maintain radiation exposures "as low as is reasonably achievable " (ALARA). Our on-site staff has and will continue to review and approve, if satisfactory, all design modifications and operating procedures developed as part of the recovery program to reassure that the licensee has considered the ALARA principles and concepts in these documents.

. Because of deficiencies that had been identified in the TMI-2 radiation protection program, the Director of the Office of Nuclear Reactor Regulation appointed a Special Panel to ccnduct an independent review of 'the radiation protection program, with special emphasis to be given to the unique radiation protection requirements of the decontamination program. On December 7,1979, the Special Panel presented its finding and recommendations to the NRC. In the Executive Summary the Panel stated tnet; There have been well deserved criticisms of the Radiation Safety Program supporting the recovery of Unit 2 at Three Mile Island. However, major GPU/ Met Ed commitments and actions have recently demonstrated a major change in management attitude. Although the Panel concluded that exposures to employees can be maintained as low as reasonably achievable while the initial preparations for recovery continue, further improvements in the radiation safety program will be able to support the major recovery effort. The progress of GPU/ Met Ed in expanding and develop-ing its radiation safety program must be consistent with the time schedule planned for major recovery activities. Enclosed for your information is a copy of "Three Mile Island, Unit' 2, Radia-tion Protection Program - Report of the Special Panel" (NUREG-0640). The NRC is instituting technical specification requirements in the TMI-2 license which will require the-licensee to provide quarterly reports on the progress of implementing the Panel's recommendations. Until such time as the Licensee's health physics program is found to be adequate by the NRC onsite staff, major radiological recovery efforts will not ~ be permitted. )

. Satisfying NEPA recuirements On November 21, 1979, the Commission published in the Federal Register a Statement of Policy and Notice of Intent to Prepare a Programmatic Environmental Impact Statement on the decontamination and disposal of radioactive wastes resulting from the March 28, 1979, accident at TMI-2. -We believe this' action is in keeping with the purposes of the National Enviionmental Policy Act to engage the public in the Commission's decision-making process. We will keep you and members of the Subcommittee fully informed on activities associated with the decontamination program. QUESTION 3: Please include regulatory guidelines that will remain stable over the period of the cleanup, consistent with the Commission's responsibilities to protect the public health and safety and the environment. ANSWER: ~ The staff anticipates that existing Commission regulations, guidelines and criteria applicable to a normally operating facility, will continue to be applied to cleanup activities at TMI-2. However, we recognize that although certain activities would otherwise be permitted at a normally operating facility, it may be warranted, in the public interest, to prohibit them at TMI-2 even though they could be conducted in full compliance with existing effluent limitations in the operating license or NRC regulations, until further evaluation of them is undertaken. At this time, we have identified several such activities: disposal of water decontaminated by EPICOR-II system, treatment and disposal of high-level contaminated water now in the reactor building, and venting or other treatment of the reactor 1 l building atmosphere. An example of how such a further restriction affects I

the cleanup. plan is shown in our analysis of the cryogenic case (see ~ response to Question 1). For the other three cases we assumed that if any proposed method of treatment and disposal meets existing Commission regulations, guidelines, and criteria for operating reactors, the method would be acceptable. For the cryogenic case we assumed that even if venting the reactor building would meet criteria, it would be prohibited and that an alternative method, which would further reduce radioactive effluents (such as cryogenic processing of gases), would be ,necessary. OUESTION 4: please describe the steps that you plan to take to re-establish public confidence in the cleanup operations and in NRC's ability to properly regulate those operations. ANSWER: Since October '2,1979, NRC senior staff members responsible for the TMI recovery program have participated in biweekly briefings on the cleanup j activities at Three Mile Island. These briefings are sponsored by the pennsylvania Department of Environmental Resources to advise public officials, the press, and the public on various phases of the cleanup operation. It is expected that these briefings will continue for the foreseeable future. In addition, we expect to have public meetings with the licensee in the area near TMI on significant aspects of the recovery program, thereby giving the public an opportunity to observe the regulatory process in action. Our senior staff members have met numerous times with Satte and local officials, made presentations to various professional, civic and. concerned citizens groups in the Harrisburg/Middletown area in an effort to explain NRC's role in overseeing cleanup activities. We have opened an office in the Middletown area to make NRC staff members more ~ ~ readily available to the public and to provide a place for public inspection of bocuments related to the recovery program. At the present time, we have

a full-time staff of fifteen professionals and three support staff members on duty at TMI providing round-the-clock coverage of all activities ' associated with the recovery program. It is our intention to maintain this level of staff effort during the entire cleanup period. t j We believe these efforts, together with the preparation of a Programmatic Environmental Impact Statement on the decontamination and disposal of radioactive wastes resulting from the accident, will greatly aid in re-establishing public confidence in the NRC's ability to properly regulate these operations. i 4 i i l 1 5 e.. ,,-,,e

A i i f 1 J l 4 ) LIST OF ENCLOSURES l. 1. Containment Entry and Cleanup Schtdule 2. Assumptions j 3. Cases Evaluated 4. Decontamination and Cleanup Phases i 5. Cleanup e Fquipment or Techniques 6. Met Ed/GPU " Summary Technical Plan for TMI-2 Decontaminated and Defueling" t 7. Three Mile Island, Unit 2, Radiation Protection Program (NUREG-0640) i' w e ,L --.e w =*, -,. - = g

TABLE I COMPARISION OF ALTERNATIVES FOR !)ECOVrialINATION OF Tile GASSES IN TMI-2 P.EACTOR BUILDING Charcoal Compression Cryogenic Freon Status of Technology Known Known Known Pilot scain only System Complexity Simple Simpic, but Complex Complex Static System, un %r pressure no pressure -

Cost of installation

$120-160 $50-75 $10-15 $4-10 million million million million

Time to Install' 30-40 months 25-35 months 20-30 months
18-36 months
Preliminary cost and schedule estimates 4

ENCLOSURE 2 ASSUMPTION ON WHILH PREPARATION OF THE FOUR CASES 0F TMI-2 CLEANUP PLANS WERE BASED 1. Effluents from the plant are acceptable if they meet the require-ments of 10 CFR 50, Appendix I, and the TMI-2 technical specifications for normal operation. Therefore, except for gas venting in the cryogenic case any proposed method of treatment and disposal that meets j these requirements would be found acceptable. For the cryogenic case 4 gas venting would not be found acceptable even if it meets these criteria. j 2. No NRC hearings would be hel.d for any steps during Phase 1 and Phase 2. 3. The licensee will have available the funds necessary to proceed with the effort as scheduled. 3 1 4. The gas within the reactor building is too radioactive to pernit early entry to the b'uilding prior to its removal. 4 i 5. Most of the contaminated liquid within the reactor building must be removed before decontamination and other preparations for removal of fuel can begin. 6. Prior to proceeding with each operation, detailed procedures will be prepared and reviewed to assure protection of the health and safety. of the public and on-site personnel. CASES EVAt0ATED Case 1 - Base Case: ~ Based on preliminary schedules prepared by the Licensee and its contractor, the Bechtel Power Corporation. 4 Case 2 - Environmental Assessment Case: Based on the Licensee preparing a proposal for the decontamination and/or disposal.of the contaminated air and water inside the containment. Approxi-mately five months was allowed for preparation of the Environmental Assessment, obtaining public coment and arriving at a decision as to the acceptability of the proposed method. The Licensee on November l'3,1979, provided a Reactor Building Purge Program. Safety Analysis and Environment Report in which he proposed using venting as the preferred method for disposal of the contam-inated gases. For this case, we assumed that his proposal would be accepted. Case 3 - Cryogenic Case: Based on the NRC finding that reactor building (containment) vcnting is not an acceptable method for decontamination and disposal of the contaminated gases inside the containment. We further assumed that we directed the Licensee to utilize one of the alternate methods discussed in the Licensee's Reactor Building Purge Program and Safety Analysis be used and that no addi- ,tional Environmental Assessment be required. This case was based on the assumption that cryogenic processing of the contaminated gases would be s el ected. The Licensee estimated that twenty to thirty months would be required to design and construct the building and to design, procure, install and test the equipment. We used twenty-four months in our study. We also assumed that entrance into the building would be delayed until the contam-inated gases were re:Nved. This would delay early entry into the containment m-

. and therefore delay observations and data gathering. This voulo result in a delay in preparation of detail plans for the cecontamination of the containment and the preparation for fuel removal. Case 4 - Programmatic Environmental Impact Statement (EIS) Case: i . Based on the assumption that processing and/or release of the contaminated gas and liquid within the containment will not be allowed until the Final Programmatic Environmental Impact Statement (FPEIS) is completed and re-leased. This delays early entry into the containment and therefore delays observations and data gathering, and hence delays preparation of detailed plans for the decontamination of the containment and the preparation for fuel removal. The preliminary schedule data for completion of the Draft Environmental.c,tatement is June 1980, and Final Environmental Statement is Dec encer 1980. This case was also based on the assumption that once Environmental Assessment for the removal of the decontaminated liquid from the containment had been approved, procurement and installation of the necessary equipment would be allowed to proceed. I a e i i

ENCLOSURE 4 DECONTAMIt:ATION AND CLEANUP PHASES Phase 1: a. Removal of contaminated gases and liquids from the reactor buil di ng. b. Initial entry into the reactor building for data gathering. c. Reactor building decontamination. Phase 2: a. Preparation for removal reactor head, b. Removal of fuel. c. Decontamination of the Reactor Coolant System.

ENCLOSURE 5 CLEANUP EQUIPMENT OR TECHNIQUES I. Water 4 A. Decontamination (1) Auxiliary Building On August 14,1979, the NRC staff issued an Environmental Assessment on the use of an EPICOR-II System to decontam-j inate intermediate level waste stored in the auxiliary building. The EPICOR-II system consists of filtration i and ion exchange equipment for the removal of radioactive particles and ions from contaiminated water. On October 16, 1979, after considering public comments on the staff's Environmental Statement, the Commission authorized use of EPICOR-II. To date, the decontamination of water processed through the system has been better than predicted and the resultant exposures to workers and the offsite public has been lower than exposures predicted in the Environmental Assessment. 4 (2) Reactor Building Sump and Primary Coolant The licensee to date has not made a formal proposal to NRC on the method it expects to use for decontamination of the contaminated water in the reactor building sump and of th2 primary coolant in the reactor coolant system. We expect i the licensee to submit a proposal early in 1980. Our program

_.2-plan is based on receiving a formal proposal in February,1980. 'As soon as a formal proposal is received, we will prepare an Environmental Assessment. However, the licensee indicated to NRC its intention to propose the use of a submerged demineralizer system for. decontamination of water in the reactor building sump and in the primary coolant system. The oe-mineralizer system will consist of filtration and ion exchange equipment submerged in the Unit 2 spent fuel pool. The filters and ion exchange equipment are a very effective way of removing radioactive species from water, as demonstrated by the performance of the EPICOR-II system. Placing the system in the fuel pool will provide shielding for onsite personnel from r'adioactivity that would accumulate in the system during the processing of the water. (3) Reactor Building Decontamination Water ~ Contaminated water produced during decontamination, of the reactor buildings, equipment within the buildings, and other decontamination procedures will contain deter-gents and a chemical that would plug fHters and -interfere

. i with the de-ionizing process. Therefore, the licensee has indicated it intends to utilize an evaporator to remove most of the contaminants from these types of liquids and a demineralizer for polishing the evporator condensate. To date, the licensee has not submitted a formal proposal for the use of this system. The evaporator system, in conjunction with a solidifi. cation T system to solidify evaporator concentrate, will probably be located in a new facility to be constructed on site. The solidification system will be discussed later. B. Water Disposal i i The licensee has not submitted a proposal for disposal of water decontaminated by any of the methods described above. We expect a proposal to be suumitted Early in 1980. A number of options are being considered by the licensee. The most direct and cheapest route is to release water to the Susquehanna River after processing reduces the radioactivity concentrations to those below the limits specified in O e w. 4 , = ~

10 CFR 20 and the dose objective in 10 CFR 50, Appendix I. Examples of alternative approaches to river discharge are: (1) Store onsite as liquid in large storage tanks. Approximately 3 to 8 million Sailons of water would have to be stored for the life of the plant. The cost would be higher than river disposal and would not resolve the problem of ultimate disposal of the decontaminated water. 1 (2) Allow the decontaminated water to evaporate from a j holding area. This method would avoid liquid releases but would result in the airborne release o? tritium. This is a low-cost approach but it is still more expensive than the release to the Susquehanna.

However, among the problems associated with this approach are overflow due to storms and the disposition of nuclides left behind after the water has evaporated.

L (3) Solidify as concrete and either st ce onsite or ship ) offsite for retention. This is an expensive option that involves forming a large quantities of concrete. l For exarple the size of a concrete slab that would l l be formed would be 6 inches thick by 235 feet on a side. j Approximately 900 shipments would be required to move ^ the concrete offsite for disposal.

5 ~. I. Reactor Building Atmosphere On November 13, 1979, the licensee submitted its proposal for the disposal of the contaminated atmosphere in the reactor building in a report titlec, "Three Mile Island Reactor Building Purge-Program Safety analysis and Environmental Report" (Novemb.er 12, 1979). The licensee examined four options for removal. and disposal of the contaminated gases, mace up primarily of radioactive krypton within the reactor ouilding: (li atmospheric purge, l (2) charcoal adsorption and storage, (3) gas compression and storage and, (4) cryogenic processing and storage. In addition to these options, the NRC staff examined freon absorbtion and storage. (1 ) Atmospheric Venting The licensee proposed that atmospheric venting be the means used for removing the contaminated gases from the reactor building. Atmospheric venting consists of re-leasing the contaminated gases from the reactor building through the plant vent stack (located 160 feet above grade) at times when the wind and other meteorological conditions are most favorable for atmosphe:ic dispersion. The licensee concluded that the venting can be done with no significant hazards to site personnel or to the general population. One advantoge of venting.is that it is less expensive and faster than the other alternatives considered. w

(2) Charcoal Adsorption and Storage l This method (snsists of passing the contaminated gases from i the reactor building through' beds of charcoal where the radioactive krypton would remain adsorbed to the charcoal. Once decontaminated, the air would be vented to the atmosphere. The charcoal with the adsorbed krypton could be stored in-definitely. To decontaminate the large quantity of air in the reactor building would require 34,000 tons of charcoal stored in 450 tanks approximately 12 feet in diameter and 50 feet long. More details on the system, including the estimated cost and construction schedule, are shown in Table 1. 4 (3) Gas Storge and Compression 4 This method consists of compressing the contaminated gas from the reactor building and storing the compressed gas in piping. At a pressure of 340 psig, about 150,000 feet of 36-inch pipe would be required to store the contaminated gases. Storing the contaminated gases at this pressure for long periods of time increases the likelihood of uncontrolled releases. ) 1 I I i I

, 4 (4) Cryogenic System The contaminated air would b,e removed from the reactor i building, passed through a recombiner to remove the oxygen, and then passed through a cryogenic system cooled by liquid nitrogen. Most of the radioactive gases, mainly krypton, would be liquidified and retained in the cryogenic c j sy stem. The purified gases would then be discharged from the plant via the reactor building vent. Periodically, l the liquified krypton would be vaporized, ano stored as a gas. The gas would have to be storea for a long period of time. The cryogenic systems, if they operate as designed, will remove about 99.9% of the krypton gas. Therefore, a small amount of radioactive krypton will be vented with the purified gas. In addition, con..entrated krypton will accumulate within the system during operation and be stored.- Any uncontrolled release of this radioactivity due to failures could result in higher offsite doses than i the controlled venting. Additional details for this system and an estimated cost construction schedule appear in Figure 1. (5) Freon Absorption In this system, the contaminateo air from the reactor building is brought into contact with freon in a packed column where the- -freon absorbs and removes the radioactive krypton from the air. ~w

l _8 The purified air is vented from the column through the reactor building vent. The krypton is then stripped from the freon in the same column and stored. 'As in the cryogenic system, a I small amount of radioactive krypton would be vented with the purified gases. The krypto.n will require long-term si.orage. Il As with the cryogenic system, some of the radioactive krypton will be released with the purified gases. The potential also exists for higher doses to offsite populations in the event of uncontrolled releases caused by failures. Additional details for the system and its estimated cost and construction schedule are shown in Table 1. j III. Solid Waste I The applicant has indicated it expects to utilize a system to solidify evaporator concentrate and spent ion exchange resins from the ion exchangers in TMI-2. The system will be a standard cement solidification system capable of solidifying the evaporator concentrate or spent ion exchange resins into 55-gallon drums. The drums can either_ be stored onsite or safely shipped offsite for burial. i

.g-l IV. Reactor Building Decontamination Some preliminary studies have been made on the methods that could be used for decontaminating the rdactor building and the external surfaces of equipment located within the building. Howeve r, selection of methods and techniques for these cannot be made until the reactor building has been entered an data gathered. Two methods of decontamination of the reactor building were examined. If the radiation levels within the reactor building after the removal l of the contaminated gases and sump water are too high for decon-l tamination personnel to work, remote, followed by hands-on, decon-l tamination would probably be proposed. On the other hand, if the radiation levels within the building are such that access to the containment for work is feasible after removal of the contaminated gases and sump water, then only hands-on decontamination will probably be utilized. It appears at present that remote decontamination will not be necessary. Remote decontamination will consist of utilizing existing sprays in i the reactor building to flush first with deionized' water and then 4 with detergent solution. This flushing would be followed by the admission of steam that would be allowed to condense on the walls. If by the use of the above two cycles, decontamination is not suf ' ficient to allow access to the building, a chemical-solution flush would be utilized. ~ f 4. e .-.e y -q

+ _,

s

. Tor hands-on decontamination, a detergent solution war' 'own is considered the prime method for initial decontamination. The detergent would be sprayed on with hand held steam nozzles. Hot spot decontamination could be accomplished by hand scrubbing with detergents. Chemicals could be utilized for decontamination if the conditions warrant this. Actual methods and detergents or chemicals to be used will be determined after the building can be entered for data gathering. In addition decisions on protective clothing to be worn by personnel, equipment to be used, visual aid such as TV cameras, etc., to be used will be made after the containment has been entered for data gathering. V. Remove Fuel and Decontaminate Reactor Systems As' the schedule indicates, at the earliest this effort will not begin for at least two years. Therefore, very little effort has been expended in these areas to date.

O b a, Ih i ! _t.; ( e 6 [ 1. .d l I e b i l ,f .b 3 g l l i g , ) l l,,. +,. - i ! l t _, 1l 3 I

i

_i.l ~ t li T ii i. l _e..i. 1I 1 l t n i g .. i di e L g;,y .a t 4 .at 4, f l I i I b l i 's m

r

_l y i 1 4 F ] u. i i t 2 J ,t. / I l 4 '9 l l l e g s. mll_ - f l + I l j i I 1 C-l d --4 4 _L 3 O ll { b j p I _4 3 g l_. 1 +

t. t.,}

f, i i 2 ii e i l g ' t l l I + g } i r. A-(- 4 o I t f 6 l_ _ _. _. _ ._ I. l f

i. l..i.

s i i i 4 a O I I t i I l i I I ! o s- - - -- -. - - - - - - - .-l-z l i i i l 4 g t t => a. f_ _. _ _. ! ! f : I l I l l l t i [- I I 8 k P'* , - r'. I i i. i +. l l b ! ' I i fW [ g, ; l l i l I' .-I l i w l i a 1 - i-q - -t--+- r- [ y p,- e i r l-p . I g O 11 ~. - , _. _ _., l ' 4o-i i p i .e Q O 5! If.I p 6 i il 6 . g',f ' i I i + 7 } l 2 -p* II I I l ' * ( y l.. l ,s g_. _ + + r a

I

_t_{ i i ..[ s r l i . i s i4 3 j i 8 4 _. _6 =.t 1[.a _ L_._ _ f I f' i l l e .L, s I .. ~ 7 t l 'l i k t _ a_ W } 1 l ; i + t 4 '3 i ..._ _ _.[. q. L.- _. i l i t i i i -j l [ >' !. ] l f l ' 3 i fb l I I t ' i I I I t 1 i j o u-r-r I t j n J I f 1 l ' l '1 s ',g i ? le i f I e, pi +i 1 j .'L, i . i t a e i i i a g llg {l i i i I ii i i + ~ I. I i i, ,,l I I 4 E * [ a. e g a i lI i l i i i y e i I I i 3 D g h { l l j i '..[ l ll i i 1 1 1; I 'I. i i :l 1 i e i 7 i. a. i a 2-l'

i l j; i, +

.I r - 3 i l 4.!..i%j ' 1 i

l!, i. E '

a .]. '}i j I ' cl a i ai

4

( I > 3..!. dl. ? l-e. ,I l. i

o.

. 5 i c, si e, ! fg I I IIl: !It 'E. ' [' b;.,I , q.. %,h,. r.f ',r t r l*}:: l 5 j: C i i

r :

i p' v., a .i q2: o ' t : l ! ' l' 1,: ! ',, r e,. v,.i. ' 1..i.!.;,. I, s ! ' : ;I 3-. I s.l '.f <t 2 r t - s : : t-2 t p<.t,1. ',6 r !y: 1-1

a *, w al oo <-

- - o; - t .i ; g i .: g; < go; ; t r e i .....: i lE. i i :C l t ' ($. t t: !E' I I L e i i .s . ) t:) t f .o if' ( .r

~._ I. . ~. e rg tis eq i l l i ij !I {.h _ __ _ _ --. l f ll, -i { ]. l - - ---- p e i!. i ~1 [ l ' * ^ l l 1 i l-l l'!- (l .t. l f, j R b _. - { ,-r' i i i cf ..f .i e i

1. D': '

l l. YI ', .\\ r U [ i l l c5 i i i r '\\ '\\ i ~ (h l i a. \\ (d(, =' l .n

9 I

I II t 6 + I ii j 11 i lp'.E) i r gj i ,c . ~7 m t. C._ 5-i =' T* i i i 1 c - g e. ._L_L.__l f {:3 ${:..<j b. I. .y ' ~ r7 ).. I ti j c-l j 1_ __g -!L___...a I .,sr l 3. 1 !? l, p f i a-

t.

s.4 1.. I l l t .l.,i i 1

cs i

, ii 11 . n'. . _.,. -_. _ _4 _u ]w.'._.: _ _ s. n ,A ) (- l{ I I i _I a. s t _i lI ] 1 ~,._I ' l._:_ a. LF l il [ i l 2 q _il - - )- w + . }- i r,., i 1 a I i p . ' ' i O l _p__._ ii! 3 b, M is li,.-- - 3 M 2,

e "n,

I Ig - t - -- q 6 ,t e v.,

-+t-t

,O p 3-g - + - - - - .g., F- _a,__.__. t : ,,,i 3, j 4 z j q

l 1

Il_a_l_ q L_..'s %_ i.W _.. i i, t I i F n .._. __ p_. f. ; > + g g o g i3 2 IE l' I 1 r--- w

F g

.d. .A _.72.{ _. - f l [ {[ h-i- u I: _ 2 u..p ~i_. s il e i 3 l ~~._i g.._ r 3'. -.a.._. .. d.d._ a [e _4 l EE .O. _., 6 i i l g - - -. g 3, p ,9 .__._._-y_ g q l 4- --~. 1 if i i li 4 g i o I !._ - m._. __ - [ _. l g S l' l f l-e 4 i ,.I. i &s l E! .I ,g j . ~. _ _ - - -. ~ a '4 l 8

5. $ - 3 I

1 E 3 y i i '. I 5E ,I I i D !S l 4

E r

u u s< t-4 i t g-i s; t <! : ,' i <I. sig i i l u 5 v '? s-a Es - J =iO' I <$ .f g.i <,o i, L' if ' O 1 I ;L 4 '. aiIy 3ls: z-e 4 R e !e 3 u u $55.I.I..E 5 82 a N . s,. ! E i. I 3 I..I.! $. s - i li (

!EI.I I$ ! !5:

I m- - f -s . j ,o --e l., e r ,j) a~E { o p gE< I d e +[7 l 34E ?if j l l n ' 2? v o ~ 3 e ie y .t H +

EflCLOSURE 6 Sl$1f33$UWl Summary Technical Plan for TMI-2 Decontamination and Defueling Metropolitan Edison Company December 12,1979 7 792 Z Z 7 sps 39ff'

4 c

SUMMARY

TECIINICAL PLAN FOR TMI-2 DECONTAMINATION AND DEFUELING TABLE OF CONTENTS

1.0 INTRODUCTION

AND SUMJ1ARY 2.0 REACTOR 3.0 DECONTAMINATION OF AUXILIARY AND FUEL HANDLING BUILDINGS 4.0 DECONTAMINATION OF CONTAIN!ENT AND REACTOR COOLANT SYSTEM 5.0 REACTOR EXAMINATION AND DEFUELING 6.0 RADI0 ACTIVE WASTE PROCESSING 7.0 SOLID RADI0 ACTIVE WASTE MANAGEMENT 8.0 FACILITIES 9.0 RADIOLOGICAL CONTROL APPENDICES A. GENERAL SCIEDULE AND ASSUMPTIONS B. NEY RECOVERY DECISIONS C. REQUIRED NRC APPROVALS D. APPLICABILITY OF NRC REGULATORY GUIDANCE E. PERIPIERAL SITE ACTIVITIES F. RESEARCl! AND DEVELOPMENT l

= _ TMI-2 DECONTAMINATION AND DEFITELING

1.0 INTRODUCTION

AND

SUMMARY

Since the March 28, 1979 accident at TMI-2, the primary technical activity has been to achieve cold shutdown of the reactor, maintain reactor system stability, and protect the health and safety of the public. Other technical activities have focused on obvious near-term problems, which include cleaning up the radioactive water, auxiliary building decontami-nation, and gathering of sufficient data in order to complete a compre-hensive plan for the decontamination and defueling. 4 i To proceed with the decontamination and defueling in an orderly manner, formulation of an integrated technical plan has been in progress and is continuing. The plan will address the engineering, construction, and operational aspects of the decontamination and cleanup. It will specify technical activities to be performed. The scope of this summary document is limited to containment entry and decontamination (Phase I) and fuel removal and reactor coolant system decontamination (Phase II). The nature of the recovery necessitates a continually evolving technical plan as additional technical data and information are gathered, or as performance of implemented plans is assessed. New plan activities l will be implemented as new information becomes available or as new options are developed or as other previously recognized options are foreclosed. i It is intended that the technical plans be flexible and the planning 2 f effort be ongoing to recognize and accommodate this dynamic situation. i The major objectives of the TMI-2 decontamination and defueling plan are to: o Maintain the reactor in a safe state, o Decontaminate the plant, i l o Process and immobilize dispersed fission products, l o Remove and dispose of the reactor core, and do so with j maximum assurance of public health and safety. Figure 1-1 presents an overview of the key activities which-are individually summarized in other sections of this report. The technical effort and planning to date concludes that TMI-2 can be decontaminated and defueled, and that resources and technology are j available within the United States to perform this effort. The effort does represent, however, a major manegement and resource coordination challenge. This decontamination and defueling can be accomplished within a time - span of approximately 2 to. 2-1/2 years from working entry to con-tainment, given no unusual technical, regulatory, political, or financial constraints. Radiological. control planning and preliminary environmental assessments concluded to date indicate no significant public health and safety impact' arising from' decontamination and defueling. 1-1 i

This decontamination and defueling can be accomplished within a time span of approximately 2 to 2-1/2 years from working entry to con-tainment, given no unusual technical, regulatory, political confidence, or financial constraints. P.adiological control planning and preliminary environmental assessments conclude to date no significant public health and safety impact arising from decontamination and defueling. 1-2

F S OtC Pr l O f I I S 1 NN 0 A. 1 E 1 l T P. 1 a if A t L t L At rl 0 P F f I I o

0 I $1 S IG A

L A U[A l s rC i t' r

r 0AF f l s

E $ Wl% i N I t 'D t fi O t C I Nt I I Dm I L O,A S N 0 i l t, g f A 4 f F NS N 5t 4 I C Ag O 0 MR l Wg l A l g f F e, T C t g C 0 N t t F O S 5 0 3, C p 5 =wd f P t u 0n I L L 1 F a n D t t n 40 9, 3 1 N N 0 g N' l O 3I I I T M 6 C 7 F gA f N $, A I Mt t Ar l R $ G H OI R f P O L S F O f T R S I t 4 A N H W. C' Ol I D t l L f AAl N S( I H C T 0 f Gt t N ly G i Dw f E N G I 4 AM MA! N lCf $ t 1 N D$ N % I S 0 I l AT 1IU A$ DA 18 DC I C A A I WO C RR N C .I t D IRO A R S f T i T E S OC R P H N C F T I F O Eily 0 D E iN l f I ht ts 0 DA r I*. 1l. f N L-T is 5 1 A 4 V W 0o P N 0_ IL N f 4 0& O Y 7 T Y S 1AR ) R T R 5 0 N C 1 R N IMt l ( f NA N 1 OI 1 O V t0 t 0 Ax" A f 1 TO f t i t i I t tU C t C c D $ cS I i IN i U A R s A" S O F CFO' E D

c TMI-2 DECONTAMINATION ~ AND DEFUELIEG 2.0 P.EACTOR The-reactor is stable, under control, and imposes no immediate safety hazard. Decay heat is being generated by the core, the struc-tural integrity of which is unknown. Cooling is by steaming through the "A" steam generator, with ultimate heat removal through the normal plant circulating water systams and cocling tower. -Criticality control j is by coolant boron concentration being maintained at greater than 3500 ppm. The scope of the reactor plan encompasses the long-term reactor cooling and criticality control, the primary system, the containment integrity, and auxiliary systems associated with maintaining and monitoring integrity. Reactor plan objectives include: o Remove decay heat in a manner compatible with decontamination d and defueling plans, and with high reliability; o Maintain reactor pressure and water inventory maintenance; I ) o Maintain reactor chemistry control; i o Eliminate or minimize structural disturbances to the core; o Provide assurance of adequate reactor suberitical margin; 1 o Maintain emergency fallback operating modes for cooling and 4 water inventory; o Monitor for uncontrolled containment leakage. The reactor decay heat generation rate is shown in Figure 2-1. Temperatures in the reactor coolant system are-being kept as low as practica1'and still maintain adequate heat transfer characteristics under the current natural circulation cooling using "A" steam generator as a heat sink; the average reactor coolant temperature is between 160 and 170 F. I As decay-power decreases, the natural circulation-mode will .become less stable and subject to increasing hydraulic fluctuations. At ] that point in the future when the reactor vessel head will be removed, natural circulation will not be a viable means of' cooling. It is desir-able, therefore, that the' reactor be placed on a long-term cooling mode, in which temperatures and pressures can be individually adjusted and suitable for all operations through defueling. A special system (mini 1 decay heat removal, NDHS) has been designed and is being installed-for i this function..The MDHS transfers the reactor heat to the nuclear service water system and removes all thermal dependency from equipment in the turbine building or the secondary plant circulating water systems. l Fallback or: emergency cooling modes exist through the long-term "B" ~ i steam generator' cooldown system, the normal in plant decay heat system, and reversion to natural circulation. 4 . i 2-1 v- ~, - e

j ' Reactor pressure is maintained by balancing the supply and dis-charge'from the reactor coclant system in a closed cycle operation. The standby pressure control (SPC) system, installed following the accident, is available as a backup. Water inventory can be maintained by makeup f rom either' the normal supply system or from the SPC. Before the MDHS or other long-term cooling system is placed in operation, reactor pres-sure will be reduced from the current 275 to 290 psig range in steps to about 100 psig, as illustrated in Figure 2-1. Pressure reduction is desirable to reduce system-leakage and necessary prior to going on a i decay heat system. The precise pressure reduction schedule is yet to be specified. Chemistry control has as primary objectives:

1) Maintaining boron concentrations greater than 3500 parts per million while monitoring j -

of reactor coolant system and all water introduced to the reactor coolant system; 2) Maintaining oxygen concentrations as low as possible to minimize corrosion;

3) pH maintenance greater than 7.5; and 4) Con-1 trolling chlorides and other potentially harmful elements to the extent possible given other constraints on the reactor coolant system.

i Prevention of significant flow forces from disturbing the core is accomplished by not operating main coolant pumps, and by using natural circulation cooling or decay heat removal systems with very low flow rates. Containment pressure has been maintained slightly subatmospheric since March 28. The building has remained isolated, with only controlled openings for hydrogen recombiner operation, sampling of the atmosphere and the sump, and insertion of a television camera and radiation monitors. z As presented in Section 6, it is intended to purge the Krypton-85 from the reactor building to permit personnel wcrking access. Should contain-ment ecoling fans fail, the containment may revert to a positive pressure with resultant uncontrolled Krypton-85 leakage and higher site and offsite radiation exposures as compared to controlled purge. L j The reactor and containment integrity is monitored by changes in: l ? o Reactor and containment temperature and pressure L l o Containment sump' water level ) Ground water radioactivit, (wells surrounding containment to o be installed) o In-containment TV and radiation detectors i f o Reactor coolant system water inventory balance i

o Source neutron level o

Reactor coolant system chemistry. ~ .The general reactor plan is. illustrated in Figure 2-2. 2 I 2-2 L ~ ,,_<-,y,-[ ,E

,R 4

i ~ r'.-

M e MC w> W i* E i w b u b U E o< m m DE or E g. a h

== R a s i i i I I I i. I I i I I I I I ~ I .t l I = E i j l r _ ...a i I ~ I I I I E

5 8

=~ t i 01 I BE g s= 2E I I I I E I E I i t I i i 5 g ~ r I u e I ~ i 1 I "s E [ i 5 E ri r==J I I .I ~ ,I E .I = o 3 e==d I s, I e I .E m-N 4 I ~ vio -r = = = 4 ./ E5" Eg I

Eg El

.I muu a l l l l _ f = E. R E. 2 4 E B B 24 ~ n ~ Ee 5 1 G 2.C lE Sr-se E i

M 1 N R 1 G 0I E UL 1 0f 0 RU N 4 F 0 1 F 1 1 O J*A 2 L + + 2 P E R R O [5III5 ,g3IIIg U T G C I A F E R t O mi CE 4 MA C nf O f T F A F L G MU N C it R Ait 0SI yC C g IM f f f N C MA 0 Rl 4 f O of H T f A fAt E J G C f,C N f R A OsI H Y C C s E E T T ) N Y l At MN A e 4 NMN 8I At0 M f 0 A il L1t T 5i Sl O O 4 YC vC Sf Sf C 0 g s E t C ( C E D D m E I D T N D N y u "N l T fN 4 A E MA C D C R t 0 I rO H Of R1 I O M tI T OA AgsT F. A T 7 Y L F L yC G AMU "8 C U NgsI O N CEC S IT R ' R C A 8 ( H YC S DS I GIC 0 C S rI]III4 gaIIIIB s e m g.g I 1 I I l

Y 8

D Eg NI S R A y A T U g S ES g R S L g CE A E R N OP l0 i PO NA L P G t 4L o LS S AN A R C lMR N 1 M0 0 R A Ul 4 R 1f ET f 0 N OP 0 1 AA C t NO N. U f Dt L R D 0 e n R ue OGC U sun TSR S f C* I t l me AA C = n t su E* R R P

TMI-2 DECONTAMINATION AND DEFUELING 3.0 DECONTAMINATION OF AUXILIAPJ AND FUEL HANDLING BUILDINGS Decontamination of the auxiliary and fuel handling buildings encompasses cleanup of the interior building surfaces, the exterior surfaces of the equipment, and the interior of ventilation and piping systems and their connected equipment, such as tanks. The primary objective of the auxiliary and fuel handling building decontamination plan is to allow access without restriction because of surface or airborne contamination. Additional objectives are to minimize radiation exposure from gamma sources contained within piping and components and to eliminate beta activity from within piping and components to prevent recontamination in the event of leaks. These objectives will be considered achieved when the following criteria are satisfied: 2 o Smearable contamination is less than 1,000 DpM/100 cm o Airborne contamination is within 10CFR20 limits General area radiation levels are gt plant design values. o If the above criteria cannot be met, tne levels will be reduced to as low as reasonably achievable and normal radiological control practices will be implemented. The open areas, passageways, stairwells, and other general access areas of the auxiliary and fuel handling buildings have been decontaminated to levels which allow unrestricted access. In order to decontaminate equipment areas, tank cubicles, and other individual areas, radiation sources internal to piping systems and tanks will first be removed in order to reduce the area dose rate from these sources. The sequence is shown in Figure 3-1. Removal of sludge from tanks and sumps, changeout of filters, and flushing of piping systems will be conducted. The schedule for these operations must be integrated witn the processing of water as discussed in Section 6. A number of decontamination techniques have been used in the auxiliary and fuel handling buildings. These include: o Abrasive scrubbing combined with solvents and followed by wet-dry vacuuming for floors o High pressure water jets on metal surfaces o Manual wiping and dry vacuuming of electrical and other selected equipment Sandblasting or otherwise removing a layer of surfaces that o have adsorbed contamination o Coating of surfaces to fix and shield adsorbed beta sources. 3-1

The decontamination operation is being conducted in accordance with apprcved procedures that have been reviewed with respect to: o Satisfying radiological control requirements o Minimizing resultant radwaste volume o Coordination with plant operations o Compatibility with waste processing Effectiveness of techniques and solvents to be used. o m 3-2

>C*C E <Z E = e sa m. a.n , 3 ,m 4

m O

C S y :5 - E5:e EE = =z = 3 m@ EC 2 w" ygi g ee C C em o-l p MG O di "Z C E,*. E *d EU <D e. 2" 2 'e =

0 e

g a c ,D,*=" -m=N Y~E - = v2 3 O F.g d 64 o -.E E eE*O C osso k O O g j o. I. mG7 s 4. a i W 4 W w e = N_ *...X m. E.w -E w ,4 U 4 no yF EE C b = ! \\ m 3 'gue y E ~ O E w= .G db l e M -r a 3*;* C n,- = - yw a E.- w --= WOOM g Ea

gj N E -

e.CG < m 6 t d I w-4 M W w *-

to 94 w 4. o.W a o W-.

" > em, a = y 4Y k us_ = .m = G C aw E r G ~: E1~=:ow s w e r_ a.. g * = 0 s o o= U I _ f oh .E. 3r 5 _m -e C oE E1 C4 e.

== NN O E w 0, e s s _U hl m or C Q N z=< r we g we ob EmM th O <*.d. a 4 9 4eW= e-ik ewS EE ou 4 4 y gme EE E= -one a = ' z ", E E 3 E m =. - c w En.O =3 = ena g _an*ge u m o -8.:= c e = e. I --- w M - = E - e 5" E N,,, =.a w

m. o e w

g W O 4 u gg em i B am O l b O b y-

TMI-2 DECONTAMINATION AND DEFUELING 4.0 DECONTAMINATION OF CONTAIN'IENT AND REACTOR C001.Ah"I SYSTEM This portion of the technical plan addresses major in-containment cleanup work other than reactor defueling, which is covered in Section 5. The objectives of this work are twofold: o To establish and maintain radiological conditions (i.e., general area radiation, airborne gaseous and particulate activities, and surface contamination levels) which will permit reactor defueling activities to proceed. o To effect, upon completion of reactor defueling, decon-tamination of the reactor coolant system itself. These two objectives are distinct and, in effect, will comprise two separate elements of containment recovery. The sequence and inter-relationship of activities associated with each of these phases are shown graphically in Figures 4-1 and 4-2, respectively. It is important to note that detailed planning and execution of work will be largely dependent on information developed in prior elements. This portion of the technical plan, at this point, can only be conceptual. However, it does represent a logical approach to the problem and provisions for methodical refinement and development as the recovery effort proceeds. As shown on Figure 4-1, the initial steps in the containment decontamination are associated with the determination of radiological and physical conditions inside the building and improvement of those conditions to the extent necessary to permit access by the decon-tamination forces. The program for determination of conditions inside the building has included the following: o Analytical reconstruction of the accident Obtaining and analyzing gas and liquid samples from inside o the building, via remote sampling devices through existing penetrations in the building walls o Obtaining and assessing direct radiation data from various sources and locations, both inside (via wall penetrations) and cutside the building o Visual examination of interior conditions via television camera inserted through an existing wall penetration o Surface contamination samples. These steps are essentially complete, and general area dose rates inside the containment are now estimated as shown on Table 4-1. With this information, the next step in th'e plan is to collect more compre-hensive information via human entry into containment. Detailed plans for the initial cntry are well under way and include selection and training of team personnel, preparation of procedures, determination of life-support 4-1 d

equipment (clothing, breathing apparatus, communication equipment, .etc.), and development of data gathering techniques. After this initial entry,-it is expected that other exploratory' entries will be planned and executed to collect data to aid in developing detailed recovery plans which minimize exposure to workers. On Figure 4-1, containment purge (i.e., the controlled release of radioactive gases, primarily Krypton-85, currently in the building) is shown as a prerequisite.to initial entry, with the option of entry without purge. While the latt'er option is physically possible, it is -considered highly undesirable in that it would result in an additional a j radiation exposure to the entry team. Moreover, even if the purge is not accomplishd prior to the inital entry, the radioactive gas must be 3 removed from the building before large-scale entry by decontamination-forces. The current technical plan presumes that.these gases will be removed via the controlled purge method, since that method is the simplest, j safest, and only permanent solution available. After access has been gained to the building, overall decontami-nation work will be done in two parts, first a gross decontamination and cleanup effort to decrease exposures from major sources as quickly and efficiently as possible, and then a local, more thorough " hands-on" 4 decontamination to reduce radiation levels to a point wh'ich will allow defueling and subsequent recovery work. There are several techniques for performing gross decontami-nation. The preferred techniques are those that involve the fewest personnel, the most directional coverage, and highest deccntamination effectiveness. Steam jets, water cannons and sprinklers are among the techniques which are being evaluated. Final decisions as to the applica-tion of these techniques will be made in the detailed planning phase, based on information gathered by entry teams. Some consideration has been given to accomplishing gross de-contamination remotely (i.e., controlled from outside of containment) by spraying the containment with large volumes of water, and possibly detergents, chemicals, and steam, via the installed containment spray system. This method is shown as an option in Figure 4-1, but at this l point such an approach is considered unlikely, in light of lower radia-tion levels as. reflected in Table 4-1, uncertain effectiveness, and the large volumes of waste as well as possible' equipment damage that could j result. Following gross decontamination, overall radiation levels will have been reduced, but more thorough manual decontamination techniques will,be employed to further reduce radiation levels and to eliminate hot spots. The following raanual techniques are being evaluated: 4 o semi-remote fire hose sprays-i o hand-held steam nozzles i 'o hydrolasers ~4-2 y __ ~- ,m#,.

grinding and/or needle guns o o manual or power scrubbing o electropolishing of metal surfaces o crushed ice impact sprays o water cannons o confined liquid freon spraying. Because of the wide variety of techniques available, efforts are under way to determine which process will yield the highest decontami-nation factors with the least personnel exposure and with a minimum amount of waste. Final selection of techniques to be used will depend on the results of these evaluations, as well as on assessment of radiation and contamination survey data collected from various containment entries. With respect to equipment and components inside the containment building, some may be completely decontaminated in place, using the techniques outlined above, while others will require decontamination in place followed by dismantling, further decontamination, and either disposal or refurbishment. Techniques chosen for each component must consider the potential for reusing the component. Again, final selection of techniques will be based on detailed survey information available after building entry. The containment decontamination work will lead directly into the reactor examination and defueling effort described in Section 5.0. At the completion of that work, reactor coolant system (RCS) decontamination can proceed, as outlined graphically in Figure 4-2. A major aspect of RCS decontamination, but one which cannot be defined or planned in detail until actual physical conditions inside the primary system are ascertained, is the cleanup of fuel debris. At this point it is assumed that some fuel material has been physically separated from the core and has been deposited in the reactor vessel or distributed elsewhere around the system. Furthermore it is presumed that some of this fuel debris will remain in the system after defueling, and must be removed. Techniques employed may be mechanical (such as vacuuming) or chemical, and they will probably require special development or adaptation to specific conditions encountered. Following cleanup of fuel debris, it is expected that some removal of activity deposited on or absorbed into the system's corrosion film will also be required. Based on existing industry experience, it is anticipated that chemical techniques will prove to be the most effec-tive for the work. A number of mechanical decontamination techniques, such as ultrasonics, hydrolasers, and ice-blasting, are also being considered and, if found useful, will be integrated into the overall RCS decontamination plan as applicable. Final selection of techniques will be based on analysis and testing with actual specimens of RCS materials, such as steam generator manway covers and control red drive mechanisms. 4-3

In general, RCS decontamination will present a variety of technical problems, and it is anticipated that a number of organizations with specialized experience or capabilities will be called on to assist in their resolution. Both containment and reactor coolant system decontamination efforts will require utilization of support facilities, such as the containment recovery service building and personnel access facility, as described in Section 8.0. Also, both of these activities will result in generation of liquid and solid radioactive waste material, to be pro-cessed and disposed of as described in Sections 6.0 and 7.0. The completion of the RCS decontamination will permit subsequent containment recovery work, not covered by this technical plan, to proceed. 4-4

TABLE 4-1 TMI-2 Containment General Area Gamma Dose Rates (Rads /hr)* (Normalized for decay to Lecember 1, 1979, assuming sump has been drained and Krypton-85 purged) Dose Points Estimated in Initial Estimate Based Planning Study on Currently (July 1979)** Available Information Elevation 282' 2.2-19 1.2-9.9 Elevation 305' 6.6 0.26 Elevation 347' 320 0.51-0.7

Does not include local hot spots.
- Initial planning study ke, red on radiation levels measured by containment dome monitor.

Later alternate measurements show that dome monitor was in a failed condition. l 1 i

4 f >F e .*at enn -- E 6.; y em>

N d

o-e nH g 2 >W EU II* = >Z s

== 4 E y U CC Z *< FT a w G 4 "o $2 c:: 2 <i Bme 9 WD*f a=g WM =s Y,m. DO a=.w k c w .I D e*G 6*@ 1 g as-U R i y 0% r** era r f e, q g O =E50 en d 4 ese I I C =c e-gE em p w - C dl E I { = n I FE -s .r I

o. -
.., o I V e $,c E-s-g> D Em co.o E*wa 5 ee I c5 ' OC -c e a3 E -rms g b* EW "w-l r I i (, 1 .I
E r
Es_ n-e se. E -c. w4w Ske ers
W S E*
meg E I esm>
%d e
E.E f q .E$ N" m m- -TE e = em u & y a.s
C h E c U OEt em g rw=
w
n. C
-. =n e e P F C Ak .t 1 F [ m* 4 C E W et E =c 5 o=
=
E* gmm 6 m E - at mC (E"> De ,4 =w Raus ( E% C Og U q w .R s a a d - o O* 4 4 u .E,,,*r 62 e db E er .S se== ne uma ano an s. ~n on w3 m 3E g Cw sc. I 4(
G
- B Eg 4w a p aos ,g g~ = wu E g ~ 3.E r ec ew -m-E.h M k W O El E b em o Eo C n'l SI 9mg r ee r6 m l wNg 5*w E &.E i <acEE y e G.= l 9.- l 1 i
.m a 4 d r W 0hM0fCON OtStCN % IN3f Att 1103 CtE ANuPop ruf t SELECr ar? SECari DECON SUPPORT SY5f tMS/ Ot 9PIS UilNG Mf CH. UI NCI USING pUAIE UI TECHNtGut(s) i I 3 W[CH ANIC AL EDUll' MINI fi IECHNiants h I I frCnNtoui$ e I g g g I I 3 I I I I I I I I II I I I I Pit 0N PE RFORM DECON I Dif Ut tlNG I O Cif ANilP DF FUtt l COMrtEff l E=Of fl0N
5. - - >
of stCS U$tNG .,3 =& USING tht mCAt b ',E UIE (Sitit0N 51 I[f.HNinuf t I FIGURE 4 2 nrACIOR COOL ANI SYSITM DFCON T AMINAllON & Ct.F ANOP l
TMI-2 DECONTAMINATION AND DEFUELING 5.0 REACTOR EXAMINATION AND DEFUELING The examination of the reactor internals and removal of the fuel may represent the most complex operation of the recovery. The planning complexity is heightened by the uncertainty surrounding the actual physical configuration of the core and reactor vessel upper internals. The scope and primary objective of reactor examination and defueling are to: o Provide through analysis and inspection information assur-ance that the reactor vessel head and the upper internals can be removed without disturbing the existing core con-figuration make the core accessible by removing the reactor vessel head o and upper internals o remove the fuel and encapsulate it for transfer to the spent fuel pool. Other activities that will be required in order to meet these objectives are the creation of special inspet' ion and handling mechanisms, preparation of the area around the reactor vessel head, preparation of the reactor internals for decontamination, and modification of the spent fuel pool to hold encapsulated fuel prior to shipment. Figure 5-1 shows the overall sequence of these activities. Because of the uncertainty regarding the actual core condition, planning activities must develop several alternatives for the foregoing activities. As the results of examinations become available and the preparatory activities are completed, the optimum approach will be selected and developed in detail sufficient to establish final designs and procedures. In general, after preparatory activities are complete, and prior to the reactor pressure vessel (RPV) head lift, a thorough evaluation to verify the methods for uncoupling of control components will be conducted. The RPV head lift will be made and continue to a height sufficient to permit additional inspection of the upper plenum area. When the head is removed, it will be lifted out of the refueling canal and placed on its storage stand on the operating deck. Additional temporary shielding will be installed to reduce the radiation levels associated with the head and service structure. Prior to upper internals removal, inspection will be performed through the 69 control rod guide tubes and through the lattice area of the upper internals to assess the core conditions. In order to detect any mechanical binding of the upper internals a load cell will be used to determine the force being exerted by the crane during the lif t. Visual inspections and hold points will occur throughout the lift to obtain the maximum information regarding the status of the core. When the upper internals lift has proceeded to about 2 feet, larger underwater cameras .5-1
with better lighting will be inserted into the annulus between the internals and the core support assembly. Video inspection of the entire top of the core will be conducted. The lift will then continue until the upper internals clear the reactor vessel. With removal of the internals, the top of the core will be exposed for a thorough inspection. This excmination will provide a basis for the final selection of the optimum fuel removal technique and a sequence for removal of the fuel assemblies. It is anticipated that the first assembly removed will be on the core periphery, the most likely location of fuel assemblies which can be lifted intiet. Complete video scans of the top of the fuel and the sides of each assembly when it is removed from the core will ascertain and record conditions of the fuel assemblies. Once the first peripheral fuel assembly is removed, a camera can be lowered into the vacated location to determine the con-dition of adjacent fuel assemblies. Removal of peripheral assemblies, which are anticipated to be intact, but structurally weakened, requires new fuel handling tools be designed which will provide means for lifting and transporting an assembly in a manner which generates no tensile forces in the assembly. Custom designed equipment will be utilized for removal of the more centrally located assemblies which are anticipated to have been geometrically reconfigured. This equipment may include vacuum or other debris extraction devices. Failed fuel cans to limit the leaching and spread of contamination will be used. The actual procedure for movement of the fuel into the cans will depend on many factors which will not be known until the condition of the fuel is assessed. The fuel vill then be staged for shipment to a fuel examina-tion facility for detailed inspection and experimental activities. The primary method for reactivity control will be by maintaining boron concentration in the reactor coolant system greater than 3500 ppm. Special instrumentation will be installed for reactivity measurement. A special materials accountability program will be implemented for fuel accountability. 5-2
I iI 1l I1 i t 8 t I i M3N SJ. O0W i H Ct P tCt N AIt O NP $ l ISSt i F'l A G A N 1 4 I f 5 t L ME 7 En AV f U Xr EE o G D Sl I R E3 RA41 r Cl OD Vt NM TN OAp MI t M8 CA 0 M NU f0A1 A RU I C F F f EN R NE RTF OS R &Nt ffC I i IP rV{ T fir if N ,O l N 0 N O i vL C rA C AR T R V RUD F O U AT fE f S U& f C l f T E R f t A G P rDA u A R oF O C HT S gig es D e A a n VE us NA I H frH n 0t t & Mee f t E S 0m CVA C L t EP v AA fo WR f o t N On f 0u ER ee I 1 t m t ft t RE t f T Ul (N e 0 e F n su r na jgIg y E Rt R u Ot I C3 % F l' A Y T O R A M 0 I I r H R ru S E iN S LR I f D AOf S t Y N NI i I rf O R t T R Ai I t T FD 0 Ni C EI NA 0I C F M A N MOf 1 I 't AF S. A f T P CV f x ML L r Hf A P E f S i R$ SWD E G Ht 2 V I T O1 R F P O t oS jS I g L I M R S E T A tK iS N H S F OF L E L Y V C L A 8 A S i E t L S G VHU VH T T N R AA NN OG GGfH N Cet I t U OL g MUR uUuC A I K I E Cl T g EOE fort 1 S i t RRtA 0E 0 W 4 U RR F S lHu1 0R l F r LHN R 01 CP o3 F A u fTA tT i1 Do U R U RR 0C E r l R c F T F O0 A A f iO T f RR F CC E F AA E I H f E N RR T4 gII g
=
f =U SV f CP 0 RR N ) 8 G N N 0 I 1 7 A 1 g N I^ C g N E A ( C S
TMI-2 DECONTAMINATION AND DEFUELING 6.0 RADIOACTIVE WASTE PROCESSING Radioact.ive waste processing activities addressed in this section include the collection, treatment, handling, and solidification of liquid radioactive waste. Subsequent on-site staging and off-site disposal of resultant solidified material are discussed in Section 7. Section 6 also addresses the program for removal of radioactive gas from containment. The primary objective of liquid radioactive waste processing is to reconcentrate radioactive fission products which are dispersed in liquids and as surface contamination throughout the plant. This proces-sing will result in waste forms suitable for safe handling, storage, and disposal consistent with applicable regulatory requirements. With respect to radioactive gas processing, the primary objective is to remove radioactive gaseous material (primarily Krypton-85) from the containment in a manner which is safe, expeditious, and consistent with applicable regulatory requirements and technical specifications. There are two general categories of radioactive water which will require processing: o accident water, i.e., water which was contaminated with fission products during the accident and is now retained within the reactor coolant system, containment sump, or in auxiliary building tanks decontamination (decon) water, i.e., water which will be used o in cleanup of systems, structures, and equipment contaminated during the accident, and which will become contaminated in the process. Quantities and characteristics of accident water are presented in summary form in Table 6-1. Quantities, chemical, and radionuclear charac-teristics of decon water are not yet well defined. ] Reconcentration of fission products contained in accident and decon j water will be accomplished by a variety of systems specially designed and installed at TMI-2 for that purpose. These treatment systems can be de-scribed briefly, as follows: EPICOR-II - This system employs a series of filters and ion-exchangers (or "demineralizers") to remove suspended and dis-solved impurities (both ra':.oactive and non-radioactive) from contaminated water. EPICOR-II has been specifically designed for treatment of " intermediate level" accident water, contami-nated to a level between 1 pc/cc and 100 pc/cc. The major source of this class of water is that which was released from the primary plant and transported to the auxiliary building early in the ac-cident. Fission products removed from water treated by this system l are captured via ion exchange on organic resin materials in steel liners. When depleted these liners are removed from service, stored, and will ultimately be disposed of. The resultant water 6-1 1
effluent from the system is essentially non-radioactive except for tritium content which is unaffected by the ion exclusive process. The EPICOR-II system has been in operation since early October and as of December I has successfully processed about 65,000 gallons (about 15*a of the intermediate water in the auxiliary building). Evaluations are under way which consider modifications of the EPICOR-II system to permit its use fot other processing require-ments, such as the water in the reactor coolant system (RCS). Submerged Demineralizer System (SDS) - The SDS is an ion exchange system conceptually similar to the EPICOR II system, but designed to accommodate much higher levels of radioactive waste water, such as that presently retained in the RCS and containment sump. There are two major differences between the SDS and the EPICOR II system. The SDS will utilize inorganic ion exchange materials (Zeolites) which permit far higher radiation loadings than organic resins. The SDS system will be located underwater in the TMI-2 spent fuel pool, to provide shielding from high radiation levels to be encountered during operation. Effluent materials from the SDS include contaminated ion exchange materials in liners, and processed water which contains tritium and only trace amounts of other radioactive isotopes. The SDS system is being fabricated and should be operational in the latter half of 1980. Because of possible schedule problems, particularly the competing needs for the fuel pool by the SDS and preparations for fuel storage, some alternatives to the SDS are being evaluated. These include modifications to the system to simplify it, thus making it available sooner, and other major design changes which would permit processing in locations other than the spent fuel pool. Both of these alternative concepts would require some combined use of this system with EPICOR-II. Evaporator / Solidification System Since ion exchange systems may not be suitable for processing of decon solutions containing detergents or other chemical cleaning agents, it may be necessary to provide other means of reconcen-trating fission products from decon water. An evaporator / solidi-fication facility has been selected for this purpose. This facil-ity is in the detailed design phase and will contain a large capacity radwaste evaporator, associated support systems including tankage, feed treatment, filtration, process control, polishing, solidification of concentrates, and storage and handling capa-bilities. i 6-2 i
Since this system is inherently quite complex, the total instal-lation will require at least two years. Once installed, however, the system will be useful not only for treatment of decon solu-tions, but also for treatment of any residual accident water. Low Activity Waste Processing System At the present time, TMI-2 low activity waste water (water not generated by the accident and having fission product concentra-tions less than 1 pCi/cc) is being processed by an icn-exchange system called EPICOR-I. In time, this system will be reserved exclusively for TMI-1 use, and a replacement system will be provided for TMI-2. Such a system is in the conceptual design stage now. Plans are being made to provide solidification capability for the concentrated radioactive materials resulting from EPICOR II, the SDS, and the evaporator. Solidification of radioactive evaporator concentrates is normally required as a prerequisite to shipment and burial, and the re-quired equipment will be provided as part of the evaporator solidification facility. Solidification of contaminated ion exchange materials has not normally been required in the nuclear industry, but such a requirement has been formally invoked by NRC for EPICOR II resins and is expected for SDS ion exchange material as well. As a result of this action by KRC, plans are under way to provide solidification capability for EPICOR II and SDS ion exchange material. All processing systems are being designed to produce effluent water which meets all established discharge quality standards. At this time, however, TMI-2 is prohibited by court order from discharging any accident water, even if processed, into the Susquehanna River. Because of the 1cng term uncertainty of this issue, large processed water storage tanks are being installed on-site and additional methods of disposing of processed water (such as evaporation, and solidification) are being examined. Also, it is intended that processed water be recycled for cleanup or other plant use to the maximum extent possible. It is necessary to remove the radioactive Krypton-85 gas from the containment. It is intended to accomplish this via a controlled release of the gas to the atmosphere. This method is a safe, simple, and permanent solution to the problem, presents no safety hazard to the public, and is in compliance with all applicable regulations and technical specifications. Technical and safety evaluations have shown this method to be superior to any alternates which have been proposed. As discussed in Section 4,0, con-tainment purge is a prerequisite to containment and RCS decontamination. 6-3
TABLE 6-1 Radioactive Water Status LOCATION APPROXIMATE DEGREE OF CONTAMINATION QUANTITY (Activity, pCi/ml) (Gallons) Tritium Gross Activity 1. Auxiliary and Fuel Handling Building Tanks and Sumps 350,000 <0.3 10-70 1 2. Reactor Coolant System 90,000 <0.3 200 (approximate) 3. Containment Sump 700,000 1.0 250 (approximate) 4. Future Decontamination Water Unknown Variable Variable Note: 65,000 gallons of water has currently been processed through EPICOR II and is stored in the EPICOR II facility.
9, I N 0 l l C f$, 3 A M t Af R. 5 rf H E D T T RPF S 0 A L 0 O3 1 Wl Rf f O Ei f A S Nl E VS OfP C H lE i ttt U cC i 4 I AO AA G l r 1 0I GT F ll 0 I u' W T 1 M Nl 0S A E I 0 f 1 I R C fW N DC R l 0 R l A S I $[ l AA i 0 sAll 'I i 1 p RR G i E 0 0A f A C I r 3W C T f WS i M( f s Nf H N A$ I I U D t {\\ f 0 7 h 1 ^ g [ 4 0 D 1 E 1 IS RC N A S f 4 I C S I Ef 8 EF CA 0 R1 OW1 0 R t P t 0 3 pIIlIIge ,j l-b m m b ll[ j l 'l l f A n i l 15 t m. 1 '. w 0 N N / N 0 I E l R 0 T 1 0 1 T Uf L t ( 8 M A" A N H I F r T CI vI E R O C t I M RAI NI A I $g E RI R I R S O T l A I C S fI OI A I MI Y W WF P B N i T P D l S A f. R T I S 04 I SUM A V I E CS D 4 S N nm C a l i s r a O s m tj II E II gE {'< N 0 T f T l 4 ) N f N R f f t A iG R Et A i E 1 f N0 O MR M ! E f NR V RD 0A A f t T EE N MT t f I E W 0 B. TT I W C C f C M t f A SA Ar AA A L CX A R E Y I T W W SW M W f CU R N N O M fV A A O O U i E S L O ft L ( C C N f E D
TMI-2 DECONTAMINATION AND DEFUELING 7.0 SOLID RADI0 ACTIVE WAUTE MANAGEMENT The objectives of solid waste management are to safely accumulate, package, stage, and make available for transport offsite all solid radio-active waste material. This is to be accomplished in a manner which does not create personnel hazard, spread of contamination, satisfies packaging, shipping, and disposal regulations. Disposal of the reactor fuel is specifically excluded from this section and is discussed in Section 5.0. The largest source of solid radioactive waste results from cleanup materials expended in the decontamination efforts. Another major source of solids includes the products of processing water contaminated as a result of the accident and used in decontamination operations, including demineralization material, filter elements, and evaporator concentrates. Plant equipment and materials for which decontamination is not feasible or effective from the standpoint of cost or personnel dose also contribute to the solid radioactive waste inventory. The management of solid radioactive wastes primarily consists of inventory control and radiological protection. The major engineering requirement is determining the criteria for size, type, and operational dates of required staging facilities. After waste quantities are projected, the staging facilities can be s' zed and constructed. This is shown in Figure 7-1. Special technical requirements will apply to handling highly radioactive solids such as demineralizer liners and evaporator bottoms. The movement, storage, and disposition of solid waste must be monitored by a suitable inventory tracking system. Facil-ities are outlined in Table 8-1. 7-1
T N T M E G 1 A J 7 rA E I M U E IG T S T A W D I t O S T R1 O1 P 1$ SNf Af RO I G N t G A f ,t l } N t G G o 4Gt N N G t t f i i I H N t i RA R f l GD ll tc C EGl AN A Uii I Ai N f r KAW C A t 14T C P t0 SA A A F O H - t 0 5 1-i ^ I i J s f, N 0 g l LE i 5, C Al i i N I t F A 4, l N 5, A Nt F f ED O1 I, N f AWi TN M pl 3 r t H iO g ML Ut t EA l g 3 l r 1 D Ltu T 1 UiA t H AP N Q Apt 0 I R3 1 A0D A0 0 G! T R lME Eyg i 1 Ig VVE OT T t R i S F eeei P S A ( S O W S L O E R SA 1 4 l NF 1 t'i' Ct Lt AA CI H Gi l 0W N t 0 0S m T G l fDC l Nt l R A AA t RR i 0t R f i t (f D ff t AO D i NI l f 3 E' i IL 5 I O N 2 l
TMI-2 DECONTAMINATION AND DEFUELING 8.0 FACILITIES TMI-2 recovery operations require support facilities in addition to those existing prior to the accident. These additional facilities include: those which directly support recovery technical activities o those which result from indirect or peripheral requirements o Direct recovery facilities are needed to support the significant increase in the number and diversity of personnel working on the site, support decou". amination and the increase in radioactive waste processing and staging, and to maintain conditions safe for workers and the general public. A mixture of temporary and long-term facilities will result. All recovery support facilities, including those required for radioactive waste processing, are summarized in Table 8-1. Radioactive waste processing facilities are further discussed in Sections 6 and 7. Figure 8-1 is a facilities plan identifying specific locations on the TMI-2 site for each major facility. 8-1
__. _ _... ~ _ a TABI.E 8-1 Facilities Rettuired for THI-2 Recovery FACILITY. DESCRIPTION PURp0SE STATUS L - Containment Recovery Concrete and steel structure. Provide control of contamination Detailed criteria being Service Puilding Includes llVAC systems with during recovery, contain e<guipment ~ defined; preliminary 1 -particulate filters, deconta-and systems used during cleanup, design initiated, mination capability, mise. provide staging f or removal, decon-service equip., heavy handling tamination, packaging, and shipment capability, and truck access of equipment and materials removed from the containment building. . Personnel' Access Steel construction. Includes Provide personnel control with Detailed criteria 1cing
, +ack change areas and changing, cleaning, and defined;6 preliminary showers, and personnel moni-monitoring facilitics adequate design initiated.
torir.g equipment. to support the large numbers of personnel expected to participate in containment cleanup operations. . Ilealt h Physics Undefined at this time. Provide capability for t.alibrating Detailed criteria health physics instruments, health-being developed physics measurement technique development. Chemistry & Radio-Steel intilding. Includes Analytical and radiochemistry Detailed criteria chemistry 1.aboratory internal contamination con-analysis to support processing being deve loped, trols and shielded counting and decontamination, fluid areas. samples, surface samples from low to high activity levels. flinimize/climinate dependency ou outside laboratory efforts.
tI e d - e d n t n u a g a a n n d e g i i l s. e v e n l v b o i a e r i a p f na e g i p o e r n r a e s S l n i U c n e t n e u g T n b i o d a A y o r i T r c e c t i t n a h s o S d r l c i e a i i r s t e t e d g t e f o u e t i i s s i n i f o i lp s - i a i c i f r t f e t r m n f e e e p u a o O o d D d S o V c i t n d ) t m , n a l t n , o s e d n l n a o f n r a e t r o r e e m y 2 l u r e o a u r m p o r o v t c n u t f f i , i c a d e y t lm a c l t l f o r n t t s g uin a d a a n t o e a e d b a a i e n t i i s t e u n r c r n l i u o l l j e t a t n a l I q i s m t h a d n E i t f n o c d m n p s o r s e b e m S t o a c a e o s i s t l e s d a c o s h c c n r n a t t a r i c n O n l u v c o P a c t l e n n a g i e e n i i e d r a i U q n w t i p i i d m in n n m p c i s R u i a h o a l l d e t i d s o a P o t s e a k a l e r d e i s e r a a r m r a s e n e v n t r , c u t a t o a t e g t r m c o n ( o e f i a t w l i o a r r s p r a o i ) a n t t m y n a r h r r p y o n o d l i a e l a r t a e c e e l a t p o i l o t o p o i p i 2 e m i n i r n v n p r d d o t e o l d u n a i n c t u r e t c u t n a t p u v t o o . e o c d d m n d p n v h p t s a s c t i. e n s o o n c c t i ~ t l o e r r o e e u h r e n n o o o n a f u nU n C c r a P c d r b t P t a a c c T a c o f o C ( 1 8 g E g n e ig L n n a i l . c r g e i s e d l i h l o n i o a t f t l e A d t l i T l i a t o n t w b z i i i N i w m c i u i e s d n i O u s e t n n s n U l l u b m l c n o ,d m e a I u o m T e l n e u r l c a m t f e P e t n o r c a I m s o c t d i s e a n c g t r , i R a y i s d n C r s t e n n a n r g d t s a g m S f a t o a o e n l n y p n e E C r s C s i m b a S e i d i D r A t a V m t m d l s l ) o V l w e T e a o l l p p l i d m t t c i l o m e c l y f a a g s n u a t g t a r t r s r r n y e d b m n n s e e i s ( h i s m r s e a e t F s d d u a e m d n p r f o mu i t u s r d c t a o 2 s r p l a u c l n t n i c a r i o e ~ m c h q e c o s a f s e g t & m t i t n i n t f i r p a t a e n x T I e l S I L i S o H t u c I a w n e , g g n n d s o i i Y o l i s r e n l i s s g T r n g i r n e I a c I e o n t a D. a c h w t L t i i a u 1 p o t e e I n t d c G T x r u S C e a a i E P o A C r r f g S , t n y g y e n F t g g i i y d s n p l i t n t , g e d S n d l d i i i r a m U o r n i i d r e n t u l f u t i n m i l e l i l a u m m i t d c c i c n a e n n a lu e e r r a. o d lu i S C D T F I C A I. ' S a i l i 4 [ ,f 4
s ^ d n e e o g t 1 d t i a i n l e i d s k u 2 a p s s e n k n t t o a e n n n a s - o l l u i i a t e t e n g v e y l l t t c a o l n a e t t l. i t l n e l l i a c i . a i e s r p t g r U n i b i a o b s d g o S n n g n n e n n p i a i t s i o n i n m T n n / e 0 d n e a b c f f 0 i t c n e i e A i i t t T t n n d 0, S a a g g a m l n d e l s u i p a i n r i i i t 0 e d i x i e r s s p 0 o e x r e n p e c e e 5 r i m l i e s o t l. d c e u u s t e c n o d q e d i m i i n r n n o w r e n o f r o t C I I C T o r O m o C s i I l l m m d / / - e f i s y i C c i t d t p 1 ip d u n iv t r e 0 S r i s e m o t ) t s p r r f c g d a e i m e e a n e w t h o d t t e i i a s r d s e a a l h e f n r f m n u t E w w b g b i o t r o a o c S a i l i n o s r r e a O d t t n p P e n i h s o t e f r o f n m a c e n 1 a m R t e u y o n n e n s 1 l o U a d s l i e i o e r l c h. b m c bl i P n i e g e R e n t l r c r v p a a a r n O c u a i m c e i e y t y t n r e i C s r a a t t a n o t l I t a a r m o i u s i a P i d t t i i m n n n w l e r d s e s w n E a e o o e h m c i r i f v c n d m r t e d l y d s o d i 1 ) e O t / o r y 1 r e e e t 1 d s s s s ( s s s o f a s r e t c e s s s e s y s / f i R r s R g c a u e e e g e r s e t d O o e e O a a o n c c c r c e c a a i C i o o o a o t d o e l I p c r C r p i m o o I S o m d t r r r h r a n r r n o P e r t P I S c r I t o a n P P P c P w a P T i S E T p S E S o C ( 1 B n o e e s E. g t s i t t a e e l l a n r r c A c o o c c T c t n a i f s o N w e c k i O d e r r c I m m m i n u e d u T e e e l t t e r P t t t o n c a d t I s s s s i u w l R y y y r e n C S S S d d t y i s g o S n e s t h t n i E e e e a s i s l i t D g g g u l c u d c n n n r o e a , a l e a a a o h e d p d v i t h h h t t e a n u o c c c a , s n c u e b r x x x r m / i o g p E E E o e c f e s r a l p t t e g k g r e e n n n a s e d r n - o e r o o o v y r n a a n t t i E s c U L t I s S f l I l g g n n y ) i i t S f i g ig i r D i f a a Y v c S d i t t T i n ( i d r S S I t i l i e L c m m o l t s r e A c e S o a k e t I C D t / S W n n s A w s r a i. a F 1 1 1 e S t 1 e I W n. d y o 1 d T g a s e m im R 2 m r r r n l o e a r r t n s g i O . e e u o o o C t t m p i C i c r e e I i s b l a t I t o o t t P n y u l v a P a r t n n E U S S a E c E c P S I I s
TAflI.E 8-1 (Continucil) FACILITY DESCRIPTION PURPOSE STATUS Equipment and flaterial Steel / concrete structure Stage, prior to shipping, all Criteria being defined Staging packageal radwaste net accom-modateil in other facilities. Provide interim storage to decouple from tight depenilency on off-site disposal ifaintenance llntiefined Support normal plant maintenance Criteria being defined and balance-of plant la yt.p
_n-a h--- 2l a w ..e =,- e w s, I \\ 9 /,d l' l // t ' 1 i e i O Y t ) E W IE I .E : y Y I !E N ./ i m 1 ten AindOIsd** , - = + = d es I EE = s t:: 5 Ej! { t E. 3 i l \\ G 1 I 'I E I i.. E a_ is i: ir i j i,i ,E#ri! < i E " s I
-l
= E I..: y- ' i u 3r y' l . __._m i_ _ y _ ~~ - N u m rin r3. i L i fj p J i a: m is, .p \\. .a
.i Ni 7 s p
-l n - 5 i t i, (155-g;; s. E 1ESE awf a3 o D 9 r if'35El 3:"; i i { '/ / I i i f- .{ ,_L '
=.
s / p / '._ / _... _. { l! ,I,l ; s, 1 s 4 t [ i g f =j 1 ,. g i \\h, e. C, I . //. 2 = I \\' = + e s N. f[. > P - =s 2 (l ; B i\\'il N j t N .f.;W.. 3 N\\l 'u li ) e L'I U:
l h
se ,d ex-n., n, - .= i i 'Il ( 1. l i = f k' g F g \\'N'1\\) l' I pr kT9Jj i .i o s Mi O ll l r3 o a 1 \\i . pl'iiJ gi .,i 'If e l 1 =i A f ~ ~! ( f-1 ~ aEr . ";, gr e n:" 'j. me 4 E. 1 7 -f '- -~- l3 g.g - r. T.. c '-/- i o m ^ 50 I f l' lE 4 n -...... . It ( g 4 P e. s f l - j~.
TMT-2 DECONTAMINATION AND DEFUELING 9.0 RADIOLOGICAL CONTROL The accident which occurred at TMI-2 has created an environment of radiological conditions which is unique to che commercial nuclear power industry. These conditions include high levels of contamination and work in high radiation fields. The accident has focused a tremendous amount of attention on the subject of radiological control as it relates to both worker exposures and releases to the environment. The radiological control program, shown on Figure 9-1, must be fully integrated into the recovery effort. As such, elements of the program will have an impact on activities associated with the technical plan. Specific objectives of the radiological control program are the following: o maintain individual and cumulative external exposure to as low as reesonably achievable (ALARA) o prevent significant internal exposure to radioactivity o prevent uncontrolled release of radioactive material to unrestricted areas. The bases for radiological control shall be the NRC Radiation Protection Plan and the plant technical specifications. Some specific regulations invoked by the above include: 10CFR 19, which addresses worker protection o 10CFR 20, which provides radiation protection criteria o 10"FR 50, which outlines emergency planning requirements o and requires adherance to the "As Low As Reasonably Achievable" (ALARA) principle with respect to occupational exposures and to releases to the environment 10CFR 71, which provides packaging and shipping criteria. o The radiological control functions that are applicable to this technical plan are: o occupational exposure control o in plant contamination control o prevention of uncontvolled releases to the environment o effluent control and monitoring o environmental monitoring. 9-1
Occupational Exposure Control Limiting radionuclide ingestion by personnel is normally accom-plished by engineering controls including process, containment, and ventilation. When such controls are not feasible, respiratory protection is required. Monitoring and air sample analysis provides warning of the presence of airborne radioactivity. External radiation exposure is limited by measures such as decontamination, processing to remove sources, engineering design (including modificatio," such as temporary shielding), administrative controls such as work planning and rehearsal, access control, and administrative exposure authorization requirements. Considerations of external occupational exposure are vital in developing the overall recovery sebedule, equipment, facilities, and sequence. In-Plant Contamination Control Contamination control is exercised by maintaining the integrity of systems and components that contain radioactive material, as well as by administrative measures. When operations require opening contaminated systems or moving contaminated items, con-tamination control methods shall be used to prevent the uncon-trolled spread of radioactive material. Contamination control considerations shall be incorporated into the design of facil-ities and process systems and the criteria for operations and maintenance activities to prevent the inadvertent release of radioactive contaminaticn. Prevention of Uncontrolled Releases to the Environment Systems, facilities, and procedures which are being developed in support of TMI-2 decontamination and defueling reflect the prin-ciple that releases to the environment must not occur in an un-controlled fashion. Effluent Control and Monitoring Effluent control includes all components and procedures (such as filters, processing systems, etc.) which are designed to control releases to the environment "As Low As Reasonably Achievable" (ALARA) as prescribed in Appendix I to 10CFR 50 and implemented via the plant environmental technical specifications. Verifica-tion of compliance with these specifications is achieved by the ef fluent monitoring program which measures the release of radio-active material from the plant via air and water pathways. The environmental monitoring program provides additional verification by measuring the impact on the environment. 9-2
Environmintal Monitorang A comprehensive sampling bioassay and analysis program is in operation to assess the effect, if any, of the accident on the environment currounding TMI-2. 9-3
TriI-2 DECONTAMINATION ANT, DifUELING APPENDIX A General Schedule and Assumptions The schedule shown in Figure A-1 represents the first two major phases of the overall recovery effort at TMI-2. The schedule addresses significant activities of Phase I and Phase II and reflects the logic set forth in this report. Phase I, Containment Entry and Decontamination, commencee at the time of the accident, 3/28/79, with plant cooldown. This phase is complete after containment decontamination. The key events of Phase I include Krypton-85 purge, containment entry, accident water processing, site facilitics completion, auxiliary building decontamination, and con-tainment decontamination. The containment decontamination activity will extend in time past the start of Phase II. Phase II, Fuel Removal and Reactor Coolant System Decontamination, commences with preparation for reactor pressure vessel head removal. This phase is complete after reactor coolant system decontamination. The primary milestones for Phase II are reactor pressure vessel head removal, fuel removal, and reactor coolant system decont.mination com-plete. This schedule will be significantly influenced by many factors which cannot be defined precisely at this time. The following major assumptions and qualifications are reflected in the development of this schedule. There are several planning studies and option evaluations o currently being conducted which will undoubtedly result in schedule changes. The radiochemical status has not been completely defined at o this time. o Financial limitations may further impact these scheduled
dates, o
It has been assumed that required NRC approvals will be obtained as shown in Appendix C. It is assumed that a stable regulatory environment exists o throughout the recovery schedule. o Extraordinary political or legal actions are assumed not to impact TMI-2 recovery. It is assumed that the recovery schedule will not be impacted o due to craft labor and material availability, Off-site radwaste dirposal will be continuously available. o A-1
M AnG O R P L 1 On g 9 g i gG E N g$ n g yU t O R t C g G ,f egtI I L gN F A S g O N l yM LC g ? g G f I g u O M"a f A L I C i N' "R J O 'c t I o I A ri U"R D N 8 IY'r A t I r N5 R I" I L D E t " A G r T N l s V N U : MAf i le N R CI DC f L H 0 ,Y IUOO A l I I N 10 f R C N f I i I N O A f5 H I f O -~ M R C IIiIIIi !i1gggIIl lllllllIl E T ) v M 7 0 i t M 4 f f Ct 0 tt As f t 00A G I r y A C M $ny N F A t t A f. A f f M ( 4 0 1 I C f f n R P N N I g O l T T L ,y l O E NAO N tA ANR g 0 A Pt W3 V g' i D ,t l t r g^ ( N C AO n NI T C g t S N g U l t N A I s 0 G ) C Rf 5 G i Ol f 0 N o N O R I ADl t } t CNNI iC 0 N l AWAl F D R A l t I O i C S A x D f ( R N L 0 M T Nt E A f E f0 I ) l NE t AMf S 4 O P o NNC Y N I us I I A I S G T s r MA ,i i [ 1 AT I A F oN P r R N f Ux O T N A C C NOD C E I T S C OCN D O CFA O ( fO C C 4D 0NG ffANI3 ) AYLG 3 NRDM N MIANf l0 IA D AL L i iH I C T x N tU T OutR s CAu t EFF DO
o Unique capabilities of industries or government agencies can be made available as needed for TMI-2 recovery. o Research and development are assumed to not significantly impact the recovery schedule. I e I 4 0 t y 4 l 4 \\ A-2
TMI-2 DECONTAMINATION AND DEFUELING APPENDIX B Key Recovery Decisions 1 There are several decisions which will have a significant impact on the planning of recovery technical activities; some of these decisions have supporting studies under way. These decisions are presented in Table B-1 for Phases I and II and are referenced to the appropriate sections of this report. Required KRC approvals are treated separately in Appendix C. B-1
I (i ,i F A i ' t6 [ i 1. - l e i:- > I y______.' ' a i e .g ,e! g .a i t t z i i i 1I. , l 8, ..g ,e l, l' !il a .... g ...__.j-.l g g ,l, .) ...I ( Iif l 8 s E is. i l i g l
p s
lss I
es

4 l
l l . 8 l i t ,,1 o r . s, I, t ..t l s .i i ll t c = r 3 g l f s. l 5 s g s. p l{ a I. ,4 s jl> .J o= Dw !; ,t l r g} 0 0I WW t -{ ~ 8 i g* ,..U y n u ~. l-0lo s g :* C M c. - 2 g w>c C=g Dwg-i l l e 8 f 0> s_._. 8 -ot. r, r II w UW l i s s 1 w=<- 5 ~ .I 20' il:p l i ll s e;; tri i 8 g g i i s. r I l 5 s s 8 I I .3. e r, tb 2 .'l:lII. I Il l l el I l!. l'I i.i l - l jil .! l r '. t i t I, } .,i ,:1. i, ::; ri, = i g ; 1 ,a s !
i t.
e: ri I ..i 3; !! I. li l g : :- 2 is. l, , i.gt;:Gi[': Ig,. e ' l ! z,. h t
lt
+
eg si
{!
. i jaj!gi g
o
.

s t r

E*
i E 1:ig{ 6l6; I, $. 2 ; .a
:. l. =-
[8
i. t r i 18'".t *.;. 1 -.,.-
.r I 8
. e. s 1 -l'
. :sp. s '. s ' ..I =te : ap r -e at i
r. s*
.r..) r. .i mir: . =. :i = ji - i,t. .s fis -:g ! j. El!: a
:,0 t
r !s j r N :.
i. 3 r i
.g:i.,.;;:i.,... it n, it : etsi r L .e r i.s t.i .s i: 1.g si !. + r. s l 5,'. 1 ' r u: E.l.a.,p r'
l.'g er 0,:::;g ;
r j :a !... i tg tia,. ...i igie :::
lsi;fI cire
= 2 5 : a ! ;. ;: e g.l-5sp!llI glg i,:E t. l r, & ; s isi i;.eis p ir r. - r i s- !['.: i .t i eliirlitig
j a $.

1,5
-> 3 8 5
t i.
s:
:i g
I a I :. 3 s. r'8j! 8 is.,l leisinti
,slpt.-ip
..;jg ea 8 a.- l- =i-p r e z.E :fle ! l: gp:s.: a. !l; rs! . el q.
,

!r g. a ;,. ;.; -
.1.
eini8 i.4 .v itsj-.! ..,::e!r r s ;:
c;c citi 8
.. I
i. = s

1.s
.. g E. .,g, u.n
3 g
r. g
s

.
.. c..
v..e g gs z.[
=.- .l. c.,. c. 2e ..gr .r: i i ti I !* E' iii .i ii ': 5 i
O n m m o e e - n e - it t a o - t a t s l s pn a st a a a yao y w ynyne cw i c sci t r o riS ev i l t m mi f d eg t ne pn l na et i i t ma ev nel mo o et .l c l u si d f nt pa a q r e at u nc
e r e l
s c iel t oimo ip ge rl i oyr ara o ncfI i i I osi t v t eee d e nn eu fd c c nt o md n a rf i i ee r o l g S g onr l n e e l nd rh ebd R r nl ce p a i O oia up i at el y r y oee pc qa t t ot r at ut r r s T t l r C cou ye n nmnini t i t di s A a ot r sC e oeauol u1 nnu i d F eca tbR s c r c qi i f i oaq n et b b c e ee r n nun s i T e es e mf f ri a ri sr ct N ,tf g oii p ot el ax lh b oe l e td n s r ne< a f a. A yao C t n c nni i. f e gb ac ps rb p I i r et eat g t i l mm yt - y nn nasyeg ao e a ny na ed e m F i l m I i t u nna oi ou at n i ct moa i i n i l f mai t s cm ow N bl t, i G a ais ai a t n t ao t s n ,y y t s I l ne t t i a i v ncs et s t l s nn ,gi na ; sl s eeoie t n i at eo S i bid c oe g vcn mi a gat ons i p i vnmi cie u uo t o pmn ii e ht mi qy qt ncyr pi t gm sp ai l i o s,i s at c r c aua p sis cop ms yb t t i ae ad cr m t us al o b eet a cnv v e t e cqe ol r r t cit eoi ao ad ns nt aec aie ue s ol s f ct t c t eonos p o td v f h yri n f ec ae a eioia mf r cae et op 1 rd R o S pbi Ad a D r D nt ct w 1 i sno i s i ) 1 c t ) 0 e a 1 D e y 0 8 r 1 h l 8 g9 a E, y r 9 n1 l r y e a n t1 i o i l e a g e e nn cy p l A v c r ( m i ay ul T o e. u e l l d r f d or ea o c d) p y l e 0 r p eoa r e i 8 t t m s ce ( n R_ n9 n n i u ( r o y, i1 e e r oe i e m m n son f r t K y n t o l t o u a aci ss c N el i n i O h r a e t cat d o i I t a t m a. iea op f S e n n n) mr n h x i. I f( o i i0 e i t e l ) C o c a m8 hd m e a0 E m t a9 cna mn u8 n t 1 at o q9 D ee f. st o) o n f n f i e1 us 0 c od ot o ot r y n8 ci nc a d f s o9 f em nee ni f i o d( omd od om o i 1 l t i n i a ( en t im t r d ga ay g nv il n) t o ca e c oe io t r i0 oi et t ea h n l t t a ns i a m8 mt mm l i e ne i 9 ea eoy ee cr c T r I ( T1 Rt S cs Sb ? NO I TC 6 E S 2 4 4 4 4 9 5 T R O 2 4 4 P ER
TABI.E B-1 (Continued) REPORT SECTION DECISION SIGNIFICANT_ FACTORS 6 Optimum system configuration for Installation of the SDS in the spent cleanup of containment sump fuel pool, alternate processing water (early 1980). options, gross decontamination schedule dependency. 6 Definition of solidificat ion Transport ation rer triction, facilities (early 1980). acceptability of solidification techniques, facility availability. .7 Definition of radioactive waste Processing and decontaminat. ion staging facilities (early 1980). methodology, ability to estimate waste quantities, ability to estimate site staging buffer time, etc.
TMI-2 DECONTAMINATION AND DEFUELING APPENDIX C Eeauired NRC Approvals I Identified and itemized below are dates for significant regulatory approvals upon which this technical plan is based. It is assumed that the NRC site office remains fully cognizant of the status of work leading to the preparation of major NRC..amittals. 1. Purge of Nr"pton-85 from containment January 1980 2. Initial centainment entry plans February 1980 3. TMI-2 radiological protection plan February 1980 4. Transfer of reactor to long term cooling mode March 1980 5. NRC TMI-2 environmental impact statement July 1980 available 6. Discharge of processed and cleaned accident January 1981 water (within technical specification limits) 7. Design basis for all processing systems and 30 days following facilities installed on site submission of criteria documents 8. Operaticn of all processing systems and At the time of facilit4 s installed on site system availability 9. Summary plan for containment decontamination 30 days after submittal 10. Summary plan for reactor defueling 30 days after submittal 11. Summary plan for reactor and reactor 30 days after submittal coolant system decontamination and cleanup 12. Planning document for reactor fuel trans-30 days after submittal portation offsite for examination 13. Site procedures 3 days after submittal 14. NRC approval for the interim storage onsite 30 days after submittal of projected quantitics cf radioactive waste c-1
TABI.E B-1 (Continued) hjy Recovery Decisions REPORT SECTION DECISION SIGN 1FICANT FACTORS 4 Method for reactor coolant system Long lead times may be required. decontamination after fuel removal Chemical decontamination, if (mid 1981). chosen, will require more extensive preparations. 4 !!cthoil for retrieval of fuel Confidence t hat the method selected debris from reactor coolant and will assure complete removal. other systems (late 1980). 4-Procedures for reactivity control Special monitoring i n s t. rume n-during in-vessel act.ivity (early tat. ion is required, control 1981), methodology. 5 IIethods for removal of RPV head lincertainty regarding condi t ion and upper internals (late 1980). of reactor internals and cfIcet of these operat. ions. Special unique tooling wi11 be required. 5 llandling method for fuel removal lincertainty regarding condition of (early 1981). the core and retention of integrity when removed. 6 Disposition of t ritiated water Storage capability, water management (early 1981). flexibility, alternate disposal methodology. 6 Opt.imum system configuration for Timing of cooling by mini decay cleanup of reactor coolant system heat removal system, maintainability (early 1980). of cooling systems, auxiliary building cleanup. u
TMI-2 DECONTAMINATION MO DEFUELING APPENDIX D Applicability of NRC Reculatory Guidance The recovery effort involves three raajor concerns that directly influence design and operations. These concerns are environmental impact, public health and safety, and occupational dose reduction. Many of the recovery activities contain first-of-a-kind operations that require in-novative solutions and are, therefore, beyond the main stream of typical power reactor design. Each of these activities will be carefully evaluated by Metropolitan Edison against current regulations to ensure minimal envi-ronmental impact, lowest public risk, and occupational exposures meeting "As Low as Reasonably Achievable" guidelines. The current regulations are sufficient to cover the breadth of recovery activities at the TMI-2 site. During the design of the facilities and services for the recovery effort, regulatory documents (e.g., Regulatory Guides, Standard Review Plans, and General Design Criteria) will be reviewed for applicability, taking into account the low stress condition of the plant and temporary nature of many of the facilities. When applicable, current sections of the appropriate documents will be considered part of the design criteria. It is the intent that facilities and systems constructed solely for the recovery period will not be designed to regulatory guidance based on hypothesis of accidents at power. Rather, the low pressure, low tempera-ture condition of the recovery facilities will be used as the bases for design and safety evaluation. This will result in simplification of the design, improved schedule, lower occupational exposure, and cost savings, without additional public risk or environmental hazard. Structural design codes will be determined in part by the temporary or permanent nature of a particular building or system, and in part by the hazard imposed by the failure of the structure. Guidance for design and operatien of facilities to minimize occupational exposure will be developed primarily'by adhering to the "As Low As Reasonably Achievable" principle outlined in Division 1 and 8 Regulatory Guides. Radiation protection procedures and practices are being implemented to maintain occupational exposures within the require-ments of 10CFR20. Existing radiological effluent limits of the TMI Unit 2 Technical Specifications will be used as upper bound design limits for effluent products and the TMI Unit 2 Environmental Technical Specifica-tions will be used as a design objective. These limits are consistent with the existing license and operational " Final Environmental Impact Statement." In general, most facilities and services constructed for the recovery effort only will be separate from existing facilities and s e rvi ces. This approach minimizes the impact on existing facilities and services and thus minimizes the possibility of compromising their original design bases. Permanent additions to plant facilities will be designed to provide the maximum long-term compatibility with the existing plant facilities while fulfilling the objectives of the recovery program. D-1
TMI-2 DECONTAMINATION AND DEFUELING APPENDIX E Peripheral Site Activities During the recovery period various activities will be initiated or continued which are not directly related to recovery. These activities can be categorized as indirect support activities, generally administrative in nature, or routine plant maintenance or layup. Some of these activities were planned or evaluated prior to the TMI-2 accident, while others would not have been necessary or cost-effective had the accident not occurred. A summary description of these activities follows. Separation of Unit 1 from Unit 2 - To enable Unit I restart opera-tions to proceed unimpeded by Unit 2 recovery operations, Unit I will be separated from Unit 2 as completely as is practicable. The major shared structure is connection of adjoining fuel handling buildings. The major shared system is the low-level radwaste processing system; an evaluation is under way to determine the best means of achieving separation of this system. Several service functions such as the fire main, potable water, sewage treatment, industrial waste, and others do not require separation nor directly impact plant operation. Turbine and Auxiliary System Layup - Several auxiliary and power producing systems will not be used during the recovery. In order to preserve them for future use, a program of protective layup will be conducted. Administration Building - A permanent administration building is desired to accommodate approximately 300 persons. The building will house site administrative services and technical support personnel. Guard Facility - As part of the program to separate Unit 1 from Unit 2, a separate access control facility will be constructed. TLD (Dosimetry) Building Expansion / Security Processing Center - The existing TLD building will be expanded to house facilities for personnel clearance and indoctrination with respect to health physics and security requirements. Upgrading of the South Bridge - In order to implement recovery cperations with minimum impact on Unit 1, it will be necessary to upgrade the south bridge to provide full capacity access to the Unit 2 end of the island. Upgrading of Other Site Support Services - To accommodate-the large nitmbers of personnel expected to be involved in recovery activities, such services as sewage treatment, parking lot accommodations, and site drainage will be expanded or upgraded as necessary. E-1
TMI-2 DECONTAMINATIO" AND DEFUELING APPENDIX F Research and Development The TMI-2 accident was the largest, single integral safety test of a complete PWR reactor and associated systems. As undesirable as the accident was, the existence of the plant in its current condition presents opportunities for significant extension of the industry's safety knowledge. In addition, the decontamination and cleanup activities themselves provide opportunities for the development or testing of new techniques and new systems wh!,:h can have generic industry-wide benefit and importance to the nation. It is recognized that the industry, governmental research and development organizations, and regulatory agencies will desire to extract all available information from TMI-2. To facilitate this research and development effort, the GPU System, the Department of Energy, the Nuclear Regulatory Commission, and the Electric Power Research Institute are developing a joint cooperative program. Through this program it is expected the reactor core, selected equipment frcm within containment, and the broad scope of cleanup and decontamination data will be made available to all interested groups. Off-site fuel and equipment exami-nation will be facilitated and coordinated. Installation of demonstra-tion facilities at the TMI-2 site, as they relate to decontamination and waste processing development, could be important for the country as a whole. The detailed technical planning for research and development has just begun. The plans reflected in this report have not as yet integrated the results cf proposed research and development tasks. The GPU System will attempt to accommodate this research and development within the TMI-2 recovery, recognizing that customers of the Metropolitan Edison system cannot be expected to bear the cost burden of development effort nor recovery schedule perturbations. F-1
NUREG 0640 Taree Mile Island, Unit 2, Radiation Protection Program Report of t7e Special Panel C. B. Meinhold, Brookhaven National Lab. T. D. Murphy, NRC D. R. Neely, NRC R. L. Kathren, Battelle Pacific Northwest Lab. B. L. Rich, Exxon Nuclear Idaho Co., Inc. G. F. Stone, Tennessee Valley Authority W. R. Casey, Brookhaven National Lab. Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission p'* coq Nh"$h) D + "b l spcpt b 8 bO49 -}}

ML19322D802

Contents

Text

Dear Mr. Chairman:

Enclosures:

SUMMARY

1.0 INTRODUCTION

1.0 INTRODUCTION

SUMMARY

Navigation menu

ML19322D802

Text

Dear Mr. Chairman:

Enclosures:

SUMMARY

1.0 INTRODUCTION

1.0 INTRODUCTION

SUMMARY

Navigation menu

Search