ML19256G004

Summary Technical Plan for TMI-2 Decontamination & Defueling

ML19256G004
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Site:	Crane
Issue date:	12/12/1979
From:	Metropolitan Edison Co
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NUDOCS 7912270202
Download: ML19256G004 (51)

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Summary Technical Plan forTMI-2 Decontamination and Defueling Metropolitan Edison Company December 12,1979 1646 344 Y%\\\\

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SUMMARY

TECHNICAL PLAN FOR TMI-2 DECONTAMINATION AND DEFUELING TABLE OF CONTEhTS

1.0 INTRODUCTION

AND SLHMARY 2.0 REACTOR 3.0 DECONTAMINATION OF AUXILIARY AhT) FUEL HANDLING BUILDINGS 4.0 DECONTAMINATION OF CONTAINMENT AND REACTOR COOLAhT SYSTEM 5.0 REACTOR EXAMINATION AND DEFUELING 6.0 RADI0 ACTIVE WASTE PROCESSING 7.0 SOLID RADI0 ACTIVE WASTE.%NAGEMEh7 8.0 FACILITIES 9.0 RADIOLOGICAL C0hTROL APPENDICES A.

GENERAL SCHEDULE AND ASSUMPTIONS B.

KEY RECOVERY DECISIONS C.

REQUIRED NRC APPROVALS D.

APPLICABILITY OF NRC REGULATORY GUIDANCE E.

PERIPHERAL SITE ACTIVITIES F.

RESEARCH AND DEVELOPMENT 1646 345

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TMI-2 DECONTAMINATION AND DEFUELING

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.

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 will be implemented as new information becomes available or as new options are developed or as other previously recognized options are foreclosed.

It is intended that the technical plans be flexible and the planning effort be ongoing to recognize and accommodate this dynamic situation.

The major objectives of the TMI-2 decontamination and defueling plan are to:

o Maintain the reacter in a safe state, o

Decontaminate the plant, o

Process and immobilize dispersed fission products, o

Remove and dispose of the reactor core, and do so with 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 available within the United States to perform this effort. The effort does represent, however, a major management 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 uausual 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 defaeling.

1646 346 1_,

This decontamination and defueling can be accomplished within a time span of approximately 2 to 2-1/2 years from working entr'f to con-tainment, given no unusual technical, regulator'/, political confidence.

or financial constraints. Radiological control planning and preliminar';

environmental assessments conclude to date no significant public health and safety impact arising from decontamination and defueling.

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TMI-2 DECONTAMINATION AND DEFUELING 2.0 REACTOR The reactor is stable, under control, and imposes no immediate safety hazard. Decay heat is being generated by the core, the struc-

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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 systems and cooling tower. Criticality control 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 Remcve decay heat in a manner compatible with decontamination and defueling plans, and with high reliability; o

Maintain reactor pressure and water inventory maintenance; Maintain reactor chemistry control; o

Eliminate or minimize structural disturbances to the core; o

Provide assurance of adequate reactor suberitical margin; o

Maintain emergency fallback operating modes for cooling and o

water inventory; Monitor for uncontrolled containment leakage.

o The reactor decay heat generation rate is shown in Figure 2-1.

Temperatures in the reactor coolant system are being kept as low as practical 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.

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 decay heat removal, MDH91 has been designed and is being installed for this function. The MLPS transfers the reactor heat to the nuclear service water system and removes all thermal dependency from equipment in he turbine buildn.g or the secondary plant circulating water systems.

Fallback or emergency cooling modes exist through the long-term "B" steam generator cooldown system, the normal in plant decay heat system, and reversion to natural circulation.

1646 349 2-1

Reactor pressure is maintained by balancing the supply and dis-charge from the reactor coolant system in a closed cycle operation. The standby pressure control (SPC) system, installed following the accident, is available e-a backup. Water inventory can be maintained by makeup from 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 nacessary prior to going on a 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 of reactor coolant system and all water introduced to the reactor coolant system; 2) Maintaining oxygen concentrations as low as possiF'.e to minimize corrosion; 3) pH maintenance greater than 7.5; and..) Con-trolling chlorides and other potentially harmful elements to ;he extent possible given other constraints on the reactor coolant system.

Prevention of significant flow forces from disturbing the core is accomplished by not operating main coolant pumps, and by using -atural 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.

As presented in Section 6, it is intended to purge the Krypton-85 f rom the reactor building to permit personnel working access. Should contain-ment cooling 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.

The reactor and containment integrity is monitored by changes in:

o Reactor and containment temperature and pressure o

Containment sump water level Ground water radioactivity (wells surrounding containment to o

be installed) o In-containment TV and radiation detectors o

Reactor coolant system water inventory balance o

Source neutron level o

Reactor coolant system chemistry The general reactor plan is illustrated in Figure 2-2.

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TMI-2 DECONTAMINATION AND DEFUELING 3.0 DECONTAMINATION OF AUXILIARY AhT) 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:

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Smearable contamination is less than 1,000 DPM/100 cm o

Airborne contamination is within 10CFR20 limits o

General area radiation levels ace at plant design values.

If the above criteria cannot be met, the 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 with 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

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Sandblasting or otherwise removing a layer of surfaces that have adsorbed contamination o

Coating of surfaces to fix and shield adsorbed beta sources.

1646 353 3-1

The decontamination operation is being conducted in accordance with approved 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 o

Effectiveness of techniques and solvents to be used.

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TMI-2 DECONTAMINATION AND DEFUELING 4.0 DECONTAMINATION OF CONTAINMENT AND REACTOR COOLANT SYSTEM This portion of the technical plan addresses major in-containment cleanup work other than reactor defueling, which is covered in Section

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5.

The objectives of this work are twofold:

To establish and maintain radiological conditions (i.e.,

o general area radiation, airborne gasecus and particulate activities, and surface contamination levels) which will pe rmit 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

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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 outside 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 the plan is to collect more compre-hensive information via human entry into containment.

Detailed plans for the initial entry are well under way and include selection and training of team personnel, preparation of procedures, determination of life-support 1646 a56 4-1

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, uith the option of entry without purge.

While the latter option is physically possible, it is considered highly undesirable in that it would result in an additional radiation exposure to the entry team. Moreover, even if the purge is not accomplishd prior to the inital entry, the radioactive gas must be removed from the building before large-scale entry by decontamination forces. The current technical plan presumes that these gases will be

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removed via the controlled purge method, since that method is the simplest, safest, and only permanent solution available.

After access has been gained to the building, overall.fecontami-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" decontamination to reduce radiation levels to a point which 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 decontamination 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 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 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 eliminata hot spots. The following manual techniques are being evaluated:

o semi-remote fire hose sprays o

hand-he:1 steam nozzles o

hydrolasers 1646 357 4-2

o grinding and/or needle guns 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

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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 definea 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 rod drive mechanisms.

1646 358 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 mater:al, 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.

1646 359

TABLE 4-1 TMI-2 Containment General Area Gamma Dose Rates (Rads /hr)*

(Normalized for decay to December 1, 1979, assuming sump has been drained and Krypton-85 purged)

Dosa 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

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Elevation 347' 320 0.51-0.7

• Does not include local hot spots.

• • Initial planning study keyed on radiation levels measured by containment dome monitor. Later alternate measurements show that dome monitor was in a failed condition.

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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 o

make the core accessible by removing the reactor vessel head and upper internals

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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 inspection and handling mechanisms, preparation of the area r.round 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 appr:ach will be selected and developed in detail sufficient to establish final designs and procedures.

In general, af ter 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 lift. 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 camera s 5-1 1647 003

with better lighting will be inser ted 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 examination 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 intact.

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 will 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.

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TMI-2 DECONTAMINATION AND DEFUELING 6.0 RADIOACTIVE WASTE PROCESSING Radioactive 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 object.

+f liquid radioactive waste processing is to reconcentrate radioan s.

tission 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 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 radioactive 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 are captured via ion exchange on organic resin materials in steel liners. When depleted these liners are removed from service, stored, and will ultimately 'ae disposed of.

The resultant water 1647 006 6-1

effluent from the system is essentially non-radioactive except for tritium content wnich is unaffected by the ion exclusive process.

The E"ICOR-II system has been in operation since early October and as of December 1 has successfully processed about 65,000 gcllots (about 15% of the intermediate water in the auxiliary building).

Evaluations are under way which consider modifications of the EPICOR-II system to permit its use for 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 SLS 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 scener, 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 suitab cessing of decon solutions containing detergents or oth al cleaning agents, it may be necessary to provide other m.sas 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-b i.l itie s.

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6-2

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 '.ow activity waste water (water not generated by the accident and having fission product concentra-tions less than 1 Ci/cc) is being processed by an ion-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 radioactivc 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 NRC, 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 Susquebanna River. Because of the long term uncertainty of this issue, large processed water storage tanks are being installed on-site and additional methods of disposing of processed water (sucn 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 1647 008

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 2.

Peactor Coolant System 90,000

<0.3 200 (approximate) 3.

Containment Sump 700,000 1.0 250 (approxin. ate) 4.

Future Decontamination Water Unknown Va riable Variable Note:

65,000 gallons of water has currently been processed through EPICOR II and is stored in the EPICOR II freility.

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TMI-2 DECONTAMINATION AND DEFUELING 7.0 SOLID RADIOACTIVE WASTE 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 sized 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.

1647 011 7-1

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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

o those which result from indirect or peripheral requirements Direct recovery facilities are needed to support the significant increase in the number and diversity of personnel working on the site, support decontamination 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 requirea 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.

1647 013 8-1

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FAClllTY l)ESCRIPTION PilRPOSE STATilE E'lu i pmen t anil Haterial Steel / concrete strrcture Stage, prior to shipping, all Cri teria lieing slefineal Staging packageal radwaste not accom-moitateil in other facilities.

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!!nde f ined Support normal plant maintenance Criteria being elefined anil balance-of plant laynp 4

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TMI-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 the 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:

maintain individual and cumulative external exposure to as o

low as reasonably achievable (ALARA) o prevent significant internal exposure to radioactivity o

prevent uncontrolled release of radioactive material to unrestricted a.eas.

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

10CFR 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 uncontrolled releases to the environment effluent control and monitoring o

o environmental monitoring.

1647 019 9-1

Occupational Exposure Control Limiting radionuclide ingestion by persocnel 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 modifications 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 schedule, 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. Centamination control considerations shall be incorporated into the design of f acil-ities and process systems and the criteria for operations and maintenance activities to prevent the inadvertent release of radioactive contamination.

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 effluent 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.

!647 020 9-2

4 Environmental Monitoring A comprehensive sampling bioassay and analysis program is in operation to assess the effect, if any, of the accident on the environment surrounding TMI-2.

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TMI-2 DECONTAMINATION AND DEFUELING APPENDIX A General Schedule and Assumptiens 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. commences at the time of the accident, 3/28/79, with plant cooldown. This phase is complete after centainment decontamination. The key events of Phase I include Krypton-85 purge, containment entry, accident water processing, site facilities 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 decontamination 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.

Finar,cial limitations may further impact these scheduled o

dates.

it has been assumed that required NRC approvals will be o

obtained as shown in Appendix C.

o It is assumed that a stable regulatory environment exists 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 disposal will be continuously available.

o 1647 023 A_1

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.

O A-2 1647 024

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1647 025

TMI-2 DECONTAMINATION AND DEFUELING APPENDIX B Key Recovery Decisions 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 NRC approvals are treated separately in Appendix C.

1647 026 B-1

b TAlli.E B-I Key itecoverL ecisio_ns D

1(Erol(T SECTION DECISION SIGNIFICANT FACTORS 2

Timing of use of tiie mini-decay licat System avai taliility, reactor coolant system removal system (early 1980).

processing, alternate cooling system capa-leility, magnitude of natural circulation i ns t alii l i t i es, etc.

2,4,6 Initiation of containment purge Aficcts containment entry, containment (early 1980).

decontamination and sulisequent recovery activities; awaiting NRC approval.

4 Timing of cc,ntainment entry (early Data acquisition is essential to follow-on 1980).

recovery planning.

4 Itemote decont aminat ion implemen-Data acquisition from containment entry tation (mid 1980).

ne ded to evaluate if remote decontamina-tion can lie of significant lienefit. System const ruct ion u.ay lie required.

Implementa-tion may dictate additional railioactive waste processing capaliitity.

4 Select ion of cliemicals used in impacts potential for future requalification containment and reactor coolant of equipment, compat ilaility of liquid waste system decontaminalion (early 1980).

processing systems.

4, 9 Select ion of methoils for reducing lie t a activity may lie cont rolling; special

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hela radiation exposure (early 1980).

radiological controls and engineering development s may lie requi red.

4 N

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crane (mid 1980).

other options may extend schedule.

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SY H"L inie ry Dec i s i_onn itEPollT SECTION l)ECISION SIGNil'ICANT FACToltS 4

tiethoil for reactor coolant system 1.ong leail times may be requ i reil.

elecontamination af ter f uel memoval Chemical ilecontaminatir,n, if (miil 1981).

chosen, will resguire more extensive preparations.

4 tiethoil for retrieval of fuel Confiilence that the methoil select eil itchris from reactor coolant anil will assure complete removal.

other systems (late 1980).

4 Proceilures for reactivity control Special monitoring instrumen-

< haring in-vessel activity (early tation is reipai reil, cont rol 1981).

me t hoitol ogy.

5 IIc t hoils for removal of itPV heaal lincertaint y rega riling conilit ion anal upper internals (late 1980).

of reactor internals anil effect of these operations. Special uningue tooling will be reilu i reil.

5 llanilling methoil f or f uel removal llucertainty regariling conilition of (early 1981).

the core anil retention of integrity when removeil.

6 Ilisposit ion at t ritiateil water Storage capability, water management 3

(early 1981).

flexibility, alternate ilisposal A

methociology.

N 6

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of cooling systems, auxiliary buililing g

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HEl'URT SECTIUN 1)ECISION SIGNIF1 CANT FACTORS 6

Optimum system configuration for installation of the SDS in the spent cleanup of containment sump tiel pool, alternate piocessing water (early 1980).

options, gross decontamination sclicif ule ilepenitency.

6 lief init ion of solieli f icat ion Transportation restriction, facilities (early 1980).

acceptability of soliilification t echn i epics, facility availability.

7 fefinition of railioactive waste Processing anil decontamination staging facilities (early 1980).

methoitology, ability to estimate waste quantities, ability to estimate site staging bnifer time, etc,

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TMI-2 DECONTAMINATION AND DEFUELING APPENDIX C Required NRC Approvals 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 submittals.

1.

Purge of Krypton-85 from containment January 1980 2.

Initial containment 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.

Operation of all processing systems and At the time of facilities 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 quantities of radioactive waste 1647 0x0 C-1

e TMI-2 DECONTAMINATION AND DEFUELING APPENDIX D Applicability of NRC Regulatory Guidance The recovery effort involves three major 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 tsr 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 construct >d 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 cf the structure.

Guidance for design and operation of facilities to minimize occupational exposure <ill 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 efflu-n' products and the TMI Unit 2 Environmental Technical Specifica-

h. as 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 se rvices. 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 1647 031

..d 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 durir.g 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 I 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 f acilities for personnel clearance and indoctrination with respect to health physics and security requitements.

Upgrading of the South Bridge - In order to implement recovery operations 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 numbers 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.

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9 TMI-2 DECONTAMINATION 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 which 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 from 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 rill 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 of 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 certurbat.ons.

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