ML20136B809

Safety Evaluation Approving Proposed Early Defueling Activities

ML20136B809
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Site:	Three Mile Island
Issue date:	11/12/1985
From:	Office of Nuclear Reactor Regulation
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ML20136B791	List: ML20136B788 ML20136B809
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NUDOCS 8511200291
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, x THREE MILE ISLAND PROGRAM 0FFICE SAFETY EVALUATION OF EARLY'DEFUELING 0F.THE TMI-2 REACTOR VESSEL

' By letter dated May 20, 1985, GPU Nuclear submitted the safety evaluation report (SER) for early defueling of the TMI-2 reactor vessel.

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This SER was subsequently revised by . letter dated October 10, 1985 (Reference.1). Following our meeting with GPU staff on July 25, 1985, we requested additional = information on early defueling activities in

' -letters dated July 29 and August 13, 1985. GPU responded to those requests in letters dated October 3 and October 23, 1985. Our review of early defueling activities also incorporated our reviews of related defueling documents, including the technical evaluation reports (TER) for the Defueling Water Cleanup System (DWCS) and Defueling Canisters,

the Boron Dilution Hazards Analysis, and the SER for Heavy Load

-Handling. NRC evaluations of those documents are referenced, as appropriate,'in this report.

J By letter dated October 24, 1985, GPU requested NRC approval to conduct

-limited preliminary defueling activities inside the reactor vessel.

lhis approval was granted in a letter dated October 29, 1985, with a supporting safety evaluation. The contiusions of that safety evaluation still apply to all reactor vessel defueling (in-vessel) activities which are covered in this Safety Evaluation. Based on the above information, discussions with GPU Nuclear staff, and our first-hand observation of training, equipment checkout and system testing, we have completed our

< review'of the proposed early defueling activities. This safety evaluation documents the results of our review of the proposed activities.

Description of Early Defueling Activities As described. in Reference 1, the proposed early defueling activities

-involve the ' removal of' accident-generated core debris from the TMI-2 reactor vessel and the storage of that debris in defueling canisters placed in storage racks located in the spent fuel pool of the Fuel Handling Building (FHB) and the fuel transfer canal (FTC) in the Reactor Building (RB). The debris, consisting of partial fuel assemblies, fuel

+ rods, end fittings, structural materials, and loose granular fuel and structural fines, will be deposited into the specially designed defueling canisters by operators using long-handled: tools and a vacuum system. During these canister loading operations, personnel supervised by a specially trained and licensed Fuel Handling Senior Reactor Operator.(FHSRO), will be working from a shielded rotating work platform positioned over the reactor vessel. All in-vessel defueling activities-will be conducted under water. The water level in the modified internals indexing fixture (IIF) will be maintained approximately five feet above the reactor vessel flange to~ provide additional radiation shielding. Physical and' administrative controls will be implemented to prevent the inadvertent lifting of core debris out of the vessel.

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Three types of canisters will be used in defueling operations: fuel, knockout and filter canisters. Fuel canisters will be filled with larger core debris and rubble that can be picked up with the manual or hydraulic long-handled tools. Knockout and filter canisters will be used in the vacuuming system to collect loose fines and other small debris. Filter canisters will also be used in the DWCS and dewatering-systems. The canisters will be supported in the vessel by the Canister

~ Positioning System (CPS), the single canister support bracket (SCSB) or the vacuum system support structure. The defueling canisters are described in more detail in the Defueling Canister TER and in the staff

, safety evaluation for defueling canisters (References 2 and 3).

Other in-vessel early-defueling activities will consist of movement of large debris to allow equipment installation and vacuuming of small

' debris, loading of debris into baskets for more efficient packing of fuel canisters, and weighing and partial dewatering of filled canisters prior to removal from the reactor vessel. Although partial dewatering of canisters in'the vessel is planned, it is not a requirement for their transfer and storage. A second dewatering system will be available in the Fuel Handling Building (FHB) for dewatering stored canisters prior to shipment.

. The reactor building canister handling bridge (CHB) will be.used 'to remove loaded canisters from the reactor vessel. The CHB and canister transfer shield (CTS) will be positioned above the appropriate canister removal port in the defueling platform and the canister will be lifted into the CTS. The CHB will then transfer the shielded canister through air to the flooded deep end of the fuel transfer canal (FTC), where it will be placed in the canal fuel storage racks or in one of the two modified fuel transfer system upenders for transfer into the FHB.

Following transfer into the FHB, a second canister handling bridge will lift the canister from the upender and lower it into a submerged storage rack location or the. FHB dewatering station in spent fuel pool "A". The CHB will also be used to transfer canisters b6een the storage racks and the FHB dewatering station. The fuel canister storage racks are described in detail in GPUN Fuel Canister Storage Rack Technical Evaluation Report and NRC's Safety Evaluation Report (references 4 and 5). Eventually, the defueling canisters will- be removed from storage and shipped offsite; however, that activity is outside the scope of early defueling operations which are addressed in this safety evaluation.

Early'Defueling Equipment The following sections describe some of the unique systems and equipment that have been specially designed for the defueling of TMI-2.

Additional details are provided in Reference 1 and in the appropriate design descriptions and drawings.

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i Defueling Work Platform The rotatable defueling work platform, positioned directly over the reactor vessel, is approximately 17 feet in diameter and includes 6 inches of stainless steel shielding, a full diameter, 18-inch wide tool slot with a 24-inch wide T-shaped extension, and three canister transfer ports. The platform is supported at a height of 9 feet above the vessel flange by a support structure erected around the IIF on the refueling canal floor. The support structure also provides shielding, piping penetrations and an off-gas seal. The platform supports #

defueling operators and the long-handled tools, including two jib cranes to assist the operators in manipulating the tools. The platform is rotated by a cable drive system which limits its speed of rotation and is equipped with a manual disc-type brake. An alignment hole and pin are provided to assure proper platform orientation for canister removal and a torque limiter will prevent platform rotation when the brake is engaged or the alignment pin installed. A cable management system is provided to prevent binding of cable and hose assemblies used in defueling. Various electrical, hydraulic and pneumatic lines are routed through a powertrack which can be reeled in or let out as the defueling platform is rotated. The decontamination spray system, mounted to the underside of the defueling work platform, provides the ability to flush radioactive debris from canisters and tools as they are removed from the reactor vessel, while confining the flush water and debris to the vessel and IIF. Borated flush water will be provided from the FTC fill manifold through a hose connection to a manifold on the defueling work platform. An off-gas system, consisting of a 4000 scfm filtration unit equipped with HEPA filters and a moisture-separating prefilter, will be operated as needed to prevent radioactive gases that could collect under the defueling platform from affecting defueling personnel. The system can maintain an airflow of 150 fpm throuch the work platform support structure via two flexible ducts into the filtration unit, which discharges the filtered effluent to the reactor building atmosphere, away from the defueling platform.

Auxiliary Work Platforms Stationary auxiliary work platforms are located at the defueling work platform elevation, on the north and south sides, to provide additional staging area for defueling personnel and equipment. The south platform also supports the control consoles for defueling equipment and is provided with shielding to allow continuous manning throughout defueling activities. The control system provides hydraulic and electrical power and controls and instrumentation for the viewing system, defueling work platform drive system, and the hydraulic system for defueling tools.

The defueling viewing system consists essentially of the same video equipment used for plenum inspections. Operators can use long-handled tools to manipulate cameras and lighting for best viewing during defueling activities.

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Long-Handled Tools Long-handled tools will be used extensively during early defueling for movement or cutting of core debris, loading debris into fuel canisters, and positioning of defueling equipment in the reactor vessel. The tool handles will be up to 30 feet long and will be sectioned to provide easy handling and assembly. Several different end effectors will be attached to the ends of the handles to perform a variety of functions including pulling, cutting, scooping, and grappling the core debris. Powered end effectors will be hydraulically operated. The operators will rely on the viewing system to monitor their work as they manipulate the tools through the tool slot in the defueling work platform. Most tools will be supported by an overhead service crane to provide vertical and lateral motion, although some can be hand-held.

Canister Positioning System The canister positioning system (CPS) is a rotating carousel installed in the reactor vessel that can hold up to five fuel and knockout canisters. The CPS will interface with the vacuum system when supporting knockout canisters for use with that system. The height of canisters in the CPS can be adjusted to three discrete elevations to allow the canisters to be positioned closer to the debris bed as it gets lower. The CPS is manually rotated to allow canisters to be lifted or lowered by the canister handling bridge through the transfer port in the defueling work platform.

Single Canister Support Bracket The Single Canister Support Bracket (SCSB) may be used to support a single fuel canister prior to CPS installation and could also be used to support a knockout canister in conjunction with vacuum system operation, if necessary. The SCSB, if used, will be mounted en rails below the tool slot in the defueling work platform, which will allow it to be manually moved to any position along the slot. A brake will be provided to prevent movement of the SCSB when positioned. Canisters supported by the SCSB can also be repositioned at other elevations to facilitate loading of debris.

Fines / Debris Vacuum System The fines / debris vacuum system is composed of a pump, piping, valves and knockout and filter canisters. The system is designed to remove small loose debris up to the approximate size of a fuel pellet. The vacuum system is supported from the underside of the defueling work platform and is controlled from the console on the south auxiliary work platform.

The vacuum pickup nozzle is connected to a canister by a flexible hose and is manipulated using a long-handled tool. Debris is picked up and passed first through a knockout canister, then any remaining debris

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larger than 0.5 microns is. collected in a filter canister. Knockout and filter canisters will be weighed in the vessel to assure that they do not ' exceed the weight limitations for loaded canisters.

Canister Handling Bridges

< lima canister handling bridges have been provided for canister transfer ,

, during defueling activities, the reactor building CHB and the fuel 2: handling building CHB. These CHB's were constructed using existing fuel handling bridges and modified canister handling trolleys with canister

-transfershields(CTS). The modified trolleys also include controls for bridge, trolley and hoist movement as well as hoists, grapples, grapple

, guiding tools, cable and hose reels, and load cells. The CTS is made of

. stainless steel . encased lead and has a retractable collar to provide

. shielding when canisters are raised or lowered, as well as during transfer. .The CHB grapples are air-operated single-point devices which engage a machined socket in the upper head of the defueling canister to lift the canisters into the CTS. During canister transfer, redundant canister retention mechanisms on the bottom of the CTS provide a ,

secondary means of supporting canisters. The CHB's are designed in i accordance with ANSI B30.2 - 1983 and B30.16 - 1978 and were load tested to meet the requirements of ANSI 830.2 and the TMI-2 Lifting,an.d Handling Program.

z. Fuel Transfer System The Fuel Transfer' System (FTS) has been modified to handle the transfer of defueling canisters weighing up to 3355 lbs. from the caactor  ;

building to the fuel handling building. Canisters will be handled similarly to normal fuel assemblies; canisters will be lowered into a modified upender in the flooded deep end of the fuel transfer canal, turned into a horizontal position and moved through one of the fuel transfer tubes into the spent fuel pool, where a second upender will raise the canister to a vertical position. The fuel handling building CHB will then be used to lift the canister from the FTS and position it j in a. storage rack location or in the FHB dewatering station. The FTS will be fully tested prior to use.

-Defueling Water Cleanup System (DWCS)

The DWCS will be used to process water in the reactor vessel during early defueling and subsequently will be available to process the deep- .

end of the FTC and in spent fuel pool "A". The system was designed to i reduce activity and improve water clarity through the removal of radioactive ions and particulate matter. The DWCS is composed of two '

subsystems; the Reactor Vessel Cleanup System and the FTC/ Spent Fuel ,

Pool Cleanup System. In the event that these systems are not operational at the start of early defueling activities, water processing 6

could be accomplished by using the Submerged Demineralizer System; however, DWCS operation will provide the capability for continuous and more efficient proces:ing. More detail on the DWCS is provided in References 6 and 7. .

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Dewatering Systems The reactor vessel and FHB dewatering systems are designed to purge water from submerged defueling canisters so that sufficient recombiner catalyst is exposed to prevent the buildup of combustible gases in the canisters. Typically, loaded canisters will be partially dewatered in the reactor vessel using bottled inert gas located on or near the defueling work platform. Water discharged from the canisters during in-vessel dewatering will remain in the vessel and be processed by DWCS.

Any inert gas released by dewatering process will be vented through the off-gas system. The FHB dewatering station is located under water to provide shielding from radiation.

Cranes Cranes that will be used during early defueling include the reactor building polar crane, reactor building service crane, and the two jib cranes installed on the defueling work platform. The polar crane will be used to lift heavy loads during installation of defueling equipment, but will not lift loaded canisters. The service crane will be used to handle tools, equipment, shielding and empty canisters as well as weighing loaded fuel canistgrs, as needed. The jib cranes will be used to assist operators in manipulating long-handled tools in the reactor vessel. The two jib cranes are oriented to lift or support tools at any position along the tool slot.

Safety Issues The safety issues relevant to the proposed early defueling activities are similar to those issues addressed for previous cleanup activities.

Specific safety issues relating to early defueling are addressed in staff evaluations of defueling components and systems, including defueling canisters and the defueling water cleanup system. The conclusions of other staff safety evaluations, as they apply to early defueling safety issues, will be summarized in this evaluation.

Criticality The potential for a criticality event during early defueling activities involves fuel in the reactor vessel and fuel transferred and stored in canisters.

The safety evaluation for preliminary defueling activities, dated October 28, 1985 (Reference 8), referenced the staff's approval of GPUN's Reactor Coolant System Criticality Report (Reference 9).

In References 8 and 9, the staff concluded that during all reactor disassembly and defueling activities, at an RCS boron concentration of 4350 ppm, the damaged core will remain subcritical with a shutdown margin of at least one percent for any postulated fuel configuration.

The actual RCS baron concentration during defueling activities will be

at least 4950 ppm, the administrative limit, thereby providing a substantially greater shutdown margin to preclude the potential for recriticality of the core. The staff also concluded in Reference that, based on GPUN's analysis and administrative controls, the potential for introduction of foreign m;terials into the RCS during early defueling is very small and that the resulting effect on RCS reactivity will not significantly reduce the existing shutdown margin. Therefore, based on our evaluations in References 8 and 9, we conclude that adequate safety margins exist to ninimize the potential for recriticality of the fuel in the reactor vessel during early defueling activities.

The NRC staff has independently verified the calculations performed by GPU to demonstrate that fuel in canisters will remain subcritical during normal and postulated accident conditions. Using conservative assumptions, the cases analyzed included all three types of canisters; alone, in a stored array, and deformed from a worst-case drop accident.

Two fuel configurations were analyzed for knockout and filter canisters, and the effect ci the canister transfer shield on the reactivity of undamaged, loaded canisters was also analyzed. Calculations were also performed to analyze the case where the entire contents of a loaded canister was dropped around a filled, stored canister. All of the cases analyzed yielded K values below 0.95. Based on the conservative -

assumptionsemploy8bfand our independent verification of the calculations, we concur with the licensee's criticality analysis for fuel in defueling canisters. Therefore, we conclude that the potential for a criticality event in the reactor vessel or defueling canisters during early defueling activities is acceptably low.

Boron Dilution Recriticality of the fuel in the reactor vessel is pievented by main-taining a high boron concentration in the RCS. GPU recently submitted a revised boron dilution hazards analysis (Reference 10) to demonstrate that the potential for a boron dilution event leading to recriticality of the core is extremely low during the proposed early defueling activities. This analysis evaluates all potential RCS dilution pathways and their isolation barriers and describes provisions for RCS sampling and level monitoring for early detection and mitigation of a potential dilution event. Other precautions will be taken to minimize the potential for dilution, including the isolation of potential sources of unborated or underborated water using multiple barriers, and the boration of hydraulic fluid used in the operation of defueling equipment. Based on our review of Reference 10 and other actions taken to prevent a dilution event, we conclude that the potential for a boron dilution event during early defueling is small; early detection of a dilution event is likely, due to sampling and monitoring capability and the large margin provided by the operating RCS boron concentration; and that effective remedial action can be taken to terminate the dilution and provide borated makeup water to the RCS.

t Decay Heat Removal The decay heat in the damaged reactor core (currently about 12 Kw) continues to be removed in the loss-to-ambient mode. Defueling activi-ties are not expected to cause a significant increase in the current RCS j temperature of 85'F; however, the RCS temperature will be monitored in accordance with the Technical Specifications and the Recovery Operations l Plan.

The concsusions of Reference 8 with respect to decay heat removal are applicable throughout early defueling activities and the loss-to-ambient mode will adequately dissipate the small amount of decay heat generated.

hjavy Load Handling The handling of heavy loads during early defueling activities are addressed by GPU in two SER's: the SER for handling of heavy loads over the reactor vessel, and the SER for handling of heavy loads inside containment (References 11 and 12). In the staff's safety evaluation for heavy load handling over the reactor vessel (Reference'13), the worst-case accident identified for all loads anticipated to be lifted over the vessel through completion of defueling, was the postulated drop of the plenum assembly (PA). For this bounding case, it was postulated that a drop of the PA would cause the simultaneous failure of all 52 incore instrumentation tubes resulting in a total RCS leakage rate of 20 gpm. As described in Reference 8, the staff :oncludes that reliable sources of borated makeup water will be available to substantially exceed the worst case RCS leakage rate and that adequate leak detection capability is provided. GPU also performed a bounding analysis to calculate offsite doses due to an instantaneous release of all unaccounted for Kr-85 resulting from a load drop. That conservative analysis yielded offsite doses several orders of magnitude below the accident limits specified in 10 CFR Part 100. In addition, the licensee has developed a lift height / weight matrix to control the lifting of defueling equipment so that a load drop will not cause the defueling platform to collapse. As specified in Reference 13, the licensee must analyze alternative load paths in determining that the best pathway for movement of a heavy load is over the vessel. On the bases of our evaluations provided in References 8 and 13, we conclude that the licensee has met the requirements of NUREG-0612 and has implemented adequate measures to prevent a heavy load drop over the reactor vessel

-and to mitigate the consequences of a potential load drop accident over the vessel during early defueling activities.

The movement of heavy loads in other parts of the reactor building and in the fuel handling building is addressed in GPUN's SER for heavy load handling inside containment, currently under staff review. The movement of heavy loads over loaded defueling canisters will require NRC approval of that SER.

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Hydrogen Control

- During in-vessel early defueling activities, including dewatering of canisters, any hydrogen or other gases generated or released will collect in the air space below the defueling work platform. The 1 defueling off-gas system will be operated as needed to create an_ airflow '

through the work platform, into the IIF enclosure, where the air will

_ mix with collected gases, and out to the reactor building after passing *

- through a filtration unit. ,0peration of the off-gas system will . dilute i any accumulated hydrogen gas, preventing a combustible concentration from being reached. Provisions for controlling. hydrogen gas buildup in

- loaded defueling canisters are addressed in detail in References 2 and 3. Each type of canister is designed with catalytic recombiners to limit the concentration of hydrogen gas in order to prevent combustion or canister overpressurization. Dewatering of canisters is necessary to

, . expose sufficient catalyst for recombination to be effective. Loaded

, defueling. canisters will be provided with two relief valves, with t

!! setpoints at 25 psig and .150 psig, in preparation for their removal and storage. If the canisters are not dewatered prior to transfer to the

- FHB, the _radiolytic decomposition of water will result in the generation of hydrogen'and oxygen; however, the relief valves will open to prevent

. canister over-pressurization. Except for a brief time during transfer,

- the canisters will be under water, thus the opening of the relief valves would release small quantities of combustible gas to the water in the ,

'p spent fuel pool or the FTC. These small quantities of gas'will be.  :

, diluted by the large volumes of air ~in the reactor building and FHB and '

therefore will not reach a combustible concentration. In Reference 3,

. the staff concluded that the design of defueling canisters is acceptable for the control of combustible' gases. The staff also concluded that the <

- consequences of an unlikely combustion event inside a canister or in the CTS during canister; transfer will not pose an unacceptable risk.

Therefore, we conclude that acceptable methods of combustible gas

! control' will ~ exist during early defueling activities.

Fire Protec; ion F Fire protection during early defueling activities will be provided in accordance with the current TMI-2-five protection program and associated procedures for control of combustible materials. Fire extinguishers and i- detection equipment will be available in the reactor _ building to

, mitigate' the consequences of a ~ pci.ential fire. It is not anticipated that defueling operations will significantly increase the potential for l' a fire.-

- Pyrophoricity

; Based on tests conducted on core debris samples and the experience of-earlier cleanup activities, the staff concluded in previous safety evaluations that' the potential for submerged core debris to sustain a pyrophoric reaction was extremely remate. During early defueling I

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activities, core debris collected in canisters could be exposed to gases following dewatering. Argon, an inert gas, will be used in the dewater-ing process to purge excess water and act as a cover gas in the canister. Some hydrogen and oxygen is likely to be generated in canisters due to radiolytic decomposition of water. Despite the fact that some debris may be exposed to oxygen, the potential for a pyrophoric reaction is still very small for the following reasons:

significant quantities of potentially pyrophoric material (zirconium hydride) are not postulated to exist in sizes small enough to spon-taneously ignite ( 10 microns); unoxidized surfaces must be newly exposed to an oxygen environment to undergo a pyrophoric reaction and any new surfaces exposed in the course of defueling will be in contact with water, thus oxidizing before canister dewatering occurs; and the rate of oxidation must exceed the heat transfer rate of the material for ignition'to occur. We conclude that the potential for a pyrophoric event during early defueling activities is extremely unlikely. If a pyrophoric event were to occur in a canister, there is a high probability that the canister will be submerged, providing additional protection against the consequences of such an event.

Occupational Exposure

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In Supplement 1 to the PEIS (NUREG-0683, Supplement No.1 drafted in December 1983, final published in October 1984), the staff estimated that under the current cleanup plan, Reactor Disassembly and Defueling could result in 2,600 to 15,000 person-rem. Although not separately listed, defueling activities alone would account for over half of the estimated occupational exposure. Since 1983, the licensee has made substantial progress through the Dose Reduction Program to reduce the radiation levels in the reactor building, especially in areas where defueling workers will spend most of their time. For example, the present dose rate at the defueling platform is less than 10 mrem /hr.

The licensee has a continuing program to further reduce the ambient dose rate in the reactor building and the staff expects that while defueling is in progress, the background radiation levels will continue to decrease. Considering this improvement, along with the ALARA program the licensee will implement during defueling, the staff now estimates that the occupational dose resulting from defueling operations is likely to be close to or fall below the lower range estimated in the PEIS Supplement 1.

Along with the Dose Reduction Program to. reduce ambient dose rate levels, the licensee has a program to maintain dose rates ALARA during defueling. This ALARA program is to be achieved through design features, operator training and operating procedures.

Except during the transfer of loaded canisters from the reactor vessel' to the fuel transfer canal (FTC), the fuel and debris will be shielded by submergence under water. Dose to defueling workers would mainly l result from Cs-137 activity in the water. This activity, currently at about 0.05 uCi/ml, will be kept low by processing through the DWCS/SDS i.

systems. Additional shielding for workers at the defueling platform will be provided by the shield plates in the shielded work platform (SWP) and the closure heads of loaded canisters. Other design features provide shielding during the transfer of loaded canisters to the FTC. Examples of these are the shield boot under the SWP, the canister transfer shield with an extendable shield collar, and the shielded canister handling bridges in the reactor and fuel buildings.

The licensee has a full scale mockup outside of the reactor building, the Defueling Test Assembly (DTA), to provide training for every defuel-ing worker. For each defueling tool, a duplicate is available at the DTA and the use of all defueling tools is practiced at the DTA. Through this training, where actual in-vessel situations are simulated, the operators will be able to perform the defueling manipulations more efficiently, resulting in reduced radiation exposure.

Operating procedures for defueling incorporate considerations to promote the ALARA concept. For example, procedures preclude the raising of fuel debris outside of the four foot water coverage zone unless Radiological Controls personnel have appropriately monitored the situation to determine that such action is ALARA. Procedures also require the flushing of debris from defuelir.g tools as they are withdrawn from the vessel to prevent the spread of contamination. More importantly, the operating procedures were developed, in part, based on the experience of defueling tools operations at the DTA. This feedback promotes efficiency and shortens overall stay times in the radiation area.

Based on the review of the above ALARA considerations, the staff concludes that the licensee has an acceptable program to maintain the collective defueling occupational dose ALARA and that the occupational dose incurred during early defueling should be within or below the range discussed in-the staff's PEIS Supplement 1. Average dose rates at the defueling work platform are expected to be relatively low (approximately 15-25 mrem /hr).

The staff has reviewed the radiation monitoring system that will be in place during defueling. The staff has determined that this system will provide adequate data and appropriate alarms should radiation levels be significantly higher than those expected. The staff has calculated radiation levels at worker occupied areas during unplanned events and/or accident circumstances (e.g., raising of the fuel debris above the normal four feet of water coverage, canister drop over the work platform). The staff has determined that the radiation monitoring system and the continuous monitoring by the Radiation Controls personnel should enable the workers to properly respona to those situations and that the estimated radiation levels are such that the workers will be able to exit the reactor building without endangering their health and safety.

c Release of Radioactivity Reference 1 describes the potential for radionuclide release to the reactor building atmosphere, the FHB atmosphere, and to the environment for normal and accident conditions during early defueling activities.

Potential releases to the environment would be in the form of gaseous effluents as early defueling activities will not create pathways for liquid effluents. All gaseous release pathways to the environment from the reactor building and tne FHB will be filtered and monitored and building ventilation controls will be maintained in accordance with the Technical Specifications.

During in-vessel defueling activities, defueling canisters, tools, and other equipment will be flushed as they are removed from the water to prevent the spread of radioactive contamination. The off-gas system will.be operated to filter particulates and to disperse radioactive gases that may collect under the defueling platform. The DWCS will be operated as needed to limit particulate and ionic activity in the RCS, spent fuel pool "A" and FTC water. The filter systems in the reactor building and' fuel handling building will prevent a significant release of particulate activity to the environment. New sources of tritium will not be produced by defueling activities, but a slight increase in tritium concentrations in the reactor building and in tritium releases to the environment may result from an increased evaporation rate. These slight increases will not cause a significant increase in radiation doses to workers or the public. GPUN has estimated the offsite dose contribution due to a postulated release of Kr-85 during normal defueling activities. The calculated doses are several orders of

. magnitude below the dose limits required by the Technical Specifications. GPUN also analyzed potential offsite doses for two bounding accident scenarios; the instantaneous release of all unaccounted for Kr-85 from the reactor vessel, and a canister drop accident where the entire canister contents are spilled on the dry canal floor. The instantaneous release of all unaccounted for Kr-85 (31,300 C1) was discussed earlier in this evaluation and in previous safety evaluations. The-resulting offsite doses to the whole body are less

- than 1% of the accident dose guidelines of 10 CFR Part 100. The canister drop accident also yielded offsite doses well below the guidelines of 10 CFR Part 100. In addition, the likelihood of a canister drop accident is extremely small for the following reasons:

1) When a canister is transported over the dry portion of the FTC, it will be held in place by the CHB grapple and by redundant canister retention mechanisms on the bottom of the CTS. The possibility of failure of both supporting mechanisms is remote.

2) The design of the canister transfer shield, the canister itself, and the lift height involved make it unlikely that a postulated drop would result in both the failure of the canister pressure boundary and the entire contents of a canister being spilled onto the canal floor.

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6 However, this postulated accident represents the worst-case since other postulated canister drops would occur over water and the consequences would be less severe. Based on our review, we conclude that: adequate methods will be-implemented to minimize the. release of radioactivity during normal early defueling activities; the likelihood of potential accidents will be low; and the offsite radiological consequences of postulated accidents will be within the guidelines specified in 10 CFR Part 100. ,

10 CFR 50.59 Evaluation e In Reference 1, the licensee concluded that the proposed early defueling activities do not constitute an unreviewed safety question, nor do they involve changes to the plant technical specifications. This evaluation is based on the comparability of the proposed defueling activities and activities analyzed in the TMI-2 FSAR. The worst-case accidents analyzed in Reference 1; a danister drop on the dry FTC and the release of the entire inventory of Kr-85, are compared to a fuel handling accident and the rupture of a waste gas decay tank, as analyzed in the FSAR. The dose consequences for both of the postulated defueling accidents are less severe than those for the corresponding accident analyzed in the FSAR. In addition, the current condition of the TMI-2 core and its associated fission product inventory reduces the consequences of most postulated accidents in comparison to those postulated for operating reactors. The RCS is highly borated and protected against dilution to prevent recriticality of the fuel, radioactive source tenns are reduced, decay heat is so low that only passive heat removal is necessary, and the RCS is depressurized.

Based on our review of the licensee's evaluation, we conclude that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety (ITS) as previously evaluated in the FSAR will not be increased. The proposed activities also do not create the possibility for an accident or malfunction of a different type than any evaluated previously.s The consequences of the postulated worst-case defueling accidents were shown to be bounded by accidents analyzed in the FSAR. The special design of defueling equipment and systems and extensive worker training will tend tg i

. decrease the probability of an accident or malfurction of ITS equ pmerA.

Also, no margins of safety will be reduced as defined in the basis for any technical specification as a result.of defueling activities.

l Conclusions In our review of the proposed early defueling activities, we have considered health and safety issues including criticality, boron dilution, decay heat removal, pyrophoricity, hydrogen control, heavy load handling, fire protection, releases of radioactivity and occupational exposure. We also considered whether the proposed activities constituted an unreviewed safety question per 10 CFR 50.59.

Based on our review, we find that; 1) there is little potential for core en

recriticality due to fuel reconfiguration or a boron dilution event, and that the design of defueling canisters provides adequate assurance that loaded fuel will remain subcritical with a substantial shutdown margin for all postulated canister conditions; 2) " loss-to-ambient" cooling of the RCS will provide adequate decay heat removal; 3) there is little potential for a pyrophoric reaction; 4) adequate methods of combustible gas controland dewatering areoff-gas provided through system defueling) operation; 5 canister design andappropriate meas handling of heavy loads have been implemented to minimize the probability and consequences of postulated accidents; 6) existing fire protection measures are acceptable; 7) there is little potential for a release of radioactivity significantly greater than the trace amounts which have been typically discharged throughout the cleanup; 8) there is little potential for worker overexposure and GPU has implemented a program to maintain occupational exposures ALARA; and 9) early defueling activities'do not constitute an unreviewed safety question. We also find that the proposed activities fall within the scope of those analyzed in the PEIS. We, therefore, conclude that the proposed early defueling activities can be safely conducted with minimal risk to the health and safety of the onsite workforce and offsite public. As described in this evaluation, our review of early defueling activities is based in part on concurrent staff reviews of defueling-related equipment, systems, and analyses. Our approval of early defueling activities incorporates any conditions or restrictions imposed by the staff in the supporting safety evaluations.

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T

References 1.

Early(Defueling 1985 GPUN) Safety Evaluation Report - Revision 4, October 10,

2. Defueling Canister Technical Evaluation Report, September 10, 1985 (GPUN)

3. Defueling Canister Safety Evaluation, November 5,1985 (NRC)

4. Fuel Canister Storage Rack. Technical Evaluation Report, February 27, 1985 (GPUN)

5. Fuel Canister Storage Rack Safety Evaluation Report, May 3, 1985 (NRC)

6. Defueling Water Cleanup System Technical Evaluation Report -

Revision 7, October 1985 (GPUN)

7. Defueling Water Cleanup System Safety Evaluation, November 1985 (NRC)

8. Preliminary Defueling Activities Safety Evaluation, October 29, 1985(NRC)

9. Reactor Coolant System Criticality Report Evaluation, March _15, 1985(NRC)

10. Boron Dilution Hazards Analysis - Revision 2, September 1985 (GPUN)

11. - Safety Evaluation Report for Heavy Load Handling Inside Containment, September 11, 1985 (GPUN)

12. Safety Evaluation Report for Heavy Load Handling Over the TMI-2 Reactor Vessel, April 19, 1985 (GPUN)

13. Safety Evaluation for Heavy Load Handling Over the Reactor Vessel, May 2, 1985 (NRC) '

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