ML19345B107

Application for Amend to License DPR-77,authorizing Storage of Low Level Radwaste Generated by Onsite Plant Operation

ML19345B107
Person / Time
Site:	Sequoyah
Issue date:	10/31/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19345B100	List: ML19345B099 ML19345B107
References
NUDOCS 8011260161
Download: ML19345B107 (60)

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SEQUO YAH NUCLEAR PL ANT UNITS 1 AND 2 LOW LEVEL RADIOACTIVE WASTE STORAGE e

TENNESSEE VA LLE Y A UTHORITY REQUES T FOR A LICENSE A MENDMEN T OCTOBER 1980 8031260 % {

ENCLOSURE 1 REQUEST FOR AMENDMENT TO FACILITY OPERATING LICENCE NO. DPR-77 SEQUOYAH NUCLEAR PLANT UNIT 1 (DOCKET No. 50-327)

It is requested that facility operating license DPR-77 for operation of the Sequoyah Nuclear Plant unit 1 be amended by adding a license condition to read as follows:

The licensee may store low-level radios cive waste onsite in the facility as described in and subject to all conditions as stated in the application for license amendment from licensee (Letter from L. M. Mills to H. R. Denton) dated November 24, 1980.

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9 ENCL 1-1

ENCLOSURE 2 LONG-TERT!, LOW-LEVEL RADIOACTIVE WASTE STORAGE FACILITY SEQUOYAll NUCLEAR PLANT 9

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TABLE OF CONTENTS 1.0 INT RODUCTION 1.1 Purpose

1.2 Background

1.3 Need 1.4 Scope 2.0 FACILITY DESIGN DESCRIPTION 2.1 Structural Design 2.2 Security 2.3 Crane 2.4 Fire Protection 2.5 Radiation Monitoring & Protection 2.6 Quality Assurance 2.7 Electrical Requirements 2.8 Equipment Codes 30 FACILITY OPERATION 3.1 Handling and ",torage Operations - Description 3.2 Monitoring Operations 4.0 HADIOLOGICAL CONSIDERATIONS 4.1 Hadiological Assessment 4.2 Incremental Occupational Exposures 4.3 Doses to Unrestricted Areas 5.0 ENVIRONMENTAL ASSESSMENT 5.1 Environmental Impacts of the Proposed Action 5.2 baavoidable Adverse Environmental Impacts 53 Irreversible and Irretrievable Commitments of Resources 6.0 SAFETY ANALYSIS 6.1 Handling and Storage Accidents 6.2 Environmental Design Accidents 6.3 Summary 7.0 DECOMMISSIONING ENCL 2-2 t

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General Plan and Location of Structu.'es l

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2 Structural-Storage Modules and Gatehouse-I Outline and' Reinforcement f

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Structural-Trash Storage Module-Outline and Reinforcement 1

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4 Structural-Resin Storage Module-Outline and Reinforcement 4

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1.0 INTRODUCTION

1.1 PURPOSE This document is provided as supporting information for a proposed amendment to the operating license for each unit of the Sequoyah Nuclear Plant. The proposed amendment is to allow onsite storage of low-level radioactive waste e.t Sequoyah. This document is intended to be fully adequate fci the staff of the U.S. Nuclear Regulatory Commission to make a decision regarding the acceptability of the action authorized by the proposed amendment.

1.2 BACKGROUND

The Tennessee falley Authority (TVA) owns and operates one unit at Sequoyah Nuclear Plant located in Hamilton County, Tennessee. This unit was licensed far a thermal power level of 3,411 megawatts on September 17, 1980.

Operation of Sequoyah results in planned and controlled generation of low-level radioactive waste.

This waste primarily consists of ion exchange resins, and evaporator bottoms and miscellaneous trash as described in Table 1.2-1.

At the present time, TVA is s'.#.pping the waste offsite to a licensed radioactive waste burial fa :ility. However, for the reasons outlined in section 1.3, Nef, TVA is seeking authorization by way of d

o amendment to th ? existing facility operating licenses to store the low-level radioactive waste onsite at Sequoyah for the operational life of the 11cility.

Long-term storage on-site until the end of the plant's operational life has a nun 1er of distinct benefits.

It can provide a management method for low-level radioactive waste in lieu of using increasingly scarce and uncertain space at commercial disposal facilities. These facilities are subject to unexpected shutdowns for indefinite periods of time for reasons such as commercial or regulatory concerns. Because of the uncertainty of available disposal facilities, TVA's proposed action will prevent undesirable impacts on plant operation. This action will also allow time for the Federal government or commercial firms to open new disposal facilities for low-level radioactive waste.

Another benefit of long-term storage comes as a result of the dramatic decrease in the amount of radiation emitted from containers at the time of disposal because of the radioactive decay. Much of the radioactive waste contains radionuclides with half-lives of one year or less. For storage times of 10 to 30 years, radioactive decay removes essentially all radionuclides ENCL 2-4

SQNP

.except' cesium-137,' strontium-90, Land cobalt-60 (all of which'are minor contributors. to: the initial overall activity and container dose rate).. Retaining. the waste Lonsite for-this decay period may result in lower exposure of individuals.in unrestricted areas during transportation 'of' waste for-ultimate offsite disposal.

' Lowericurie contents could ~ result in less-radiological. impact on disposal facilities -and possibly the useL of. less' restrictive

' disposal areas.

(The-consequences of a transportation accident j

may be ' reduced because aor the lower curie-content.)

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1.3~ NEEE Since the.startup. of Sequoyah Nuclear Plan't Unit 1, TVA has

' packaged and shipped low-level radioactive waste (LLRW) generated-at Sequoyah to' Chem-Nuclear Systems, Inc.'s (CNSI) commercial 4

l radioactive waste disposal site near Barnwell, South Carolina.

In the pasti few months, however, significant restrictions have l

been placed on'the amount of packaged LLRW that CNSI will accept for disposal.

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CNSI'has announced a policy that will result in further restrictions al the volume that TVA can send to CNSI in the very near future, and it now appears that acceptable disposal space will become increasingly scarce and uncertain within the next 10 years. The problem of the lack of sufficient available i

disposal < space at CNSI for the LLRW generated at Sequoyah will progressively intenr,1fy as other TVA nuclear plants come on lir.e because the anno"r.ced disposal restrictions are being i

applied co-eafP. utility as opposed to each plant..This situation a

could worsea L over the next two years because of-cutbacks in TVA's monthly allocation of disposal space. Even without these restrictions it is likely that additional disposal options-would be needed because no other waste disposal facilities are currently b'ing planned in the southeast or midwest regions of the-nation.

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The need to develop alternatives to disposing of LLRW at CNSl's i

facility is 'immediate. The' intent of the proposed action is to ensure that the uncertain availability of commercial disposal space will~ not adversely affect future el 'tric power j

generation at Sequoyah', which will be a major contt Abutor to the TVA electric power system and' add significantly to the:

reliability of_the system. Operation of Sequoyah. will increase 4

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' TVA's ' ability. to comply with the Nation's policy of attaining energy indepedence and could continue to minimize future i

. dependence on foreign oil. -Implementation of the proposed action lwill make TVA's operations at Sequoyah essentially immune to:

4 outside restrictions on disposal of.- LLRW for the foreseeable fu tu re.

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ENCL 2-5

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SQNP The need for immediate action twquires a LLRW management plan that can be initiated promptly. The continuing nature of the problem requires a solution that will extend into the foreseeable fu ture. Therefore, TVA's proposed action is to take administrative action to reduce the volume of waste being generated followed by plant design modifications to provide long-term onsite storage and volume reduction of LLRW.

Delaying action at this time would offer TVA no advantages in rese.lving the present and future LLRW storage needs at Sequoyah.

Delaying action now would only make the situation more dtfricult when action is mandatory. There are no foreseeable occurrences which would help alleviate the situation in the short term that could justify TVA's waiting before taking any action.

Therefore, delaying action would have the same effect as taking no action. TVA's assessment indicates that taking no action or delaying action could severely curtail electric power generating capability at Sequoyah during a period in which use of domestic energy sources must be maximized.

1.4 SCOPE The scope of this document is limited to the Long-Term Onsite Waste Storage Facility. The information provided consists of the facility design criteria, the environmental and radiological assessments, and a safety or accident analysis, as well as information regarding facility eperation and decommissioning.

The design basis for the Sequoyah long-term low-level radioactive waste storage facility as given in this document is based on USNRC Regulatory Guide 1.143 Regulatory Guide 1.143 was utilized by TVA as a minimum design basis because it was determined to be the most applicable to the nature of the facilities, although it was not specifically prepared and issued for this purpose. The actual design parameters employed by TVArin the facilities' design are in some cares more conservative than those required oy Begulatory Guide 1.143 This is done in order to facilitate the development of a standardized design acceptable for use at all TVA nuclear power plants.

ENCL 2-0

SQNP TABLE 1.2-1 I Scintillation Liquids IScintillation Vials 10il/ Water mix t' rom lubrication and diesel oil PVC's Polyethylene Boots Rubber Shoe Covers Ion Exchange Resins Evaporator Concentrates Rubber Hose Plastic Hose (Nalgene).

Cotton Gloves, Inserts, Coveralls Surgical Masks Paper Coveralls Pine Crates Oak Crates Plywood Crates Scrap Iron and Steel Copper Wire Small Hand Tools Chains Cables Maps Brooms Wood used for scaffolding and ladders Calle Insulation Laboratory Equipment (vials, glassware, plastic bottles)

HEPA Filters Other wood and small metal objects I These substances are produced in relatively small quantities.

They are stored on site because CNSI will not accept these substances for disposal.

F.NCL2r7

-SQNP 2.0 FACILITY DESIGN DESCRIPTION 2.1 STRUCTURAL DESIGN 2.1.1 General aximum number of storage modules to be constructed at Sequoyah is 18.

This includes a minimum of 5 resin storage modules (20 compartments) and a minimum of 9 trash storage modules (36 compartments).

The resin storage modules and trash storage modules will be above-ground, safety-related structures constructed of reinforced concrete.

All sto: age modules will be designed to withstand the design basis events specified in 2.1.2 and will be designed and constructed for the normal loads, severe environmental loads, and extreme environmental loads specified ir 2.1.3 Details of the site facility layout, the storage modules, and the gatehouse are provided in Figures 1 through 4.

2.1.2 Design Basis Events 2.1.2.1 Design Basis Earthquake Each storage module will be designed to withstand the design basis earthquake (DBE).

The DBE is defined as a top-of-ground motion with three statistically independent orthogonal components. The peak acceleration will be a minimum of 0.09G.

Response

spectrum for this event will be taken in accordance with Regulatory Guide 1.60.

2.1.2.2 Design Basis Flood Lach storage module will be located above the design basis flood elevation. The design basis flood elevation will be above the 500-year flood elevation at the site.

2.1.2.3 Design Basis Wind Each storage module will be designed to withstand the forces exerted by a wind having ENCL 2-8

SQNP a maximum speed of 95 mi/h and a recur rence interval of 100 years.

2.1.2.4 Design Basis Precipitation Normal rainfall of 4 inches / hour has a precipitation frequency value of once every 100 years. Grading for the facility will be such - that buildup of water around the structure during and after precipitation will be minimized.

2.1 3 Loads, Definition, Nomenclature 2.1 3.1 Definition of Load Terms for Safety-Related Structure The following terms are used in the load combination equations for safety-related struc tures:

Normal Loads - those loads to be encountered during normal facility operation and include:

Dead loads or their related internal moment and forces including any permanent equipment load s.

Live loads or their related internal moment and forces including any movable equipment loads.

Thermal effect loads during normal operating conditions based on the most critical transient or steady-state condition.

Severe Environmental Loads include:

Loads generated by the design basis wind and earthquake specified for the facility.

Other Lorqs Construction live loads.

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SQNP 2.1 3.2 Minimum Live Loads 2

The minimum roof live load is 50 lb/ft,

'2.1 3.3 Precipitation Loads The maximum snow load and glaze ice load for the facility is less than the minimum roof live load specified.

No rainfall buildup on the roof la anticipated since there will be no parapets and the roof will be sloped and free draining.

2.1.4 Load Combination 2.1.4.1 Methodology For these concrete structures, the required section strength used in design is the maximum value among the several values determined for the required loading combinations of ACI 318-77, " Building Code Requirements for Reinforced Concrete."

Situations occur where one or more loads in a loading combination have opposite ' signs from the other loads in the same combination. The following situations will be investigated for possible reversal of net effects and for determination of maximum moments nr.d forces:

A.

Area distribution for live load.

B.

Maximum value for live load.

C.

Zero value for live load.

Other loads will be combined with these live load situations.

2.1.4.2 Load Combinations for Concrete Structures l

For service load conditions, the strength design method will be used. The required section strength will be at least equal to the greatest of the load combinations given ir ACI 318-77. All storage modules will meet the requirementa for watertight structures.

ENCL 2-10

SQNP 2.1."

Foundation 2.1.5.1 Foundation Design Each storage mcdule foundation will be a structure composed of concrete base slab and walls placed cn either in situ soil or compacted fill. The foundation of the module will be designed to withstand normal, severe, and extreme environmental loading conditions.

2.1.5.2 Soil Properties The ultimate bearing capacity for the storage module foundation will be determined by standard methods. The maximum allowable bearing capacity will have a factor of safety of 2.5 with respect to the ultimate capacity. The minimum factor of safety for sliding and overturning for the storage modules will be 1.3 for normal conditions b"t may be reduced to 1.1 for severe and cacreme environmental conditions.

2.1.5.3 Settlement The storage modules will be designed for the anticipated total and differential settlement.

F 2.1.6 Concrete 2.1.6.1 Structural Concrete All structural cast-in-place concrete and precast concrete beams and caps will have a specified minimum comprensive strength of 2

3000 lb/in,

2.1.7 Steel i

2.1.7.1 Reinforcing Steel Reinforcing steel will be grade 60 deformed bars per specification ASTM A 615.

2.1.7.2 Structural Steel Rolled shapes, plates, and bars will be

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SQNP per ASTM specification A 36.

2.1.8 Drainage and Sampling Each of the compartments of a storage module will be provided with internal liquid collection and drainage capability routed to an external point for sampling and collec tion. The collection point consists of a stainless steel 2-inch drain valve and a smaller sampling valve at the low point of the drain. The external collection point is surrounded by a covered concrete sump connected to the modale. The concrete sump will be utilized to collect any liquid and provided with a decontaminable coating on the interior surface.

2.1.9 Decontaminable Coatings The interier surfaces of each storage module (excluding the cap) will be coated using an approved decontaminable coating' system in accordance with TVA's General Consti uction Sp(cification G-14, Part No. :i-935.

2.1.10 Wall Thickness and Radiation Shielding The storage modules will be shielded using concrete.

The outer walls of each re. sin storage medule will be a minimum of 42 inches thick while the outer walls of each trash stora e module will be a minimum of 24 c

inches thick. The concrete caps for resin and trash storage modules will be 24 inches thick. The resin storage modules (including foundations, walls, beams, and 24-inch caps) will be designed to support an additional 18-inch (maximum) concrete cap for additional shielding if needed.

The concrete beams which shield the joint between the concrete caps will be a mini ua of 24 inches thick (vertically).

O 13CL2-12

SQNP 2.2 SECURITY The storage facility will' be surrounded by a wire fabric fence with three strands of barbed wire, totaling eight feet in height with a 20-foot isolation zone on each side.

An intrusion detection system and Closed Circuit TV (CCTV) system are provided and terminate in a gatehouse. The gatehouse is designed for 24-hour operation with communications with the nuclear plant via radio and telephone.

Yard lighting of 0.2 foot-candle and a patrol road within the fenced area is provided for surveillance purposes. Two points of access through the fence will be d

provided. The primary source of power for all electrical escurity equipment is the nuclear plant.

2 CRA NE The crane to be used at the LLRW facility will be a rubber-tired, diesel-powered, mobile gantry crane.

It will have two cross beams, a 15-ton capacity trolley on the front beam and two 30-ton capacity trolleys on the rear beam. The 15-ton hoist will be used to handle the LLRW containers and the 30-ton hoists will be used to handle the storage module caps.

In order to facilitate movement from one module to another, the crane will be driven and steered by the same wheels and these wheels will be capable of turning 90 in either direction.

In addition to its standard features, the crane will be equipped with an AC generator, an air compressor, eight SCO-watt lights, a c.ble reel and a hose reel (to provide air and e.' actric power to the 15-ton hook), and a CCTV monitoring system. The CCTV monitoring system will be designed to allow remote handling of the LLRW containers beyond the line of sight of the operator. The CCTV monitors, the CCTV controls, and all crane controls will be mounted in a cab. Three special lifting devices will be furnished.

Handling of the resin liners will be accomplished using a rigid frame with air-actuated lifting lugs. The SS-gallon drums will be handled using a standard gravity-actuated barrel grapple. A magnetic lifting system will be provided to handle the support grating that is to be used for stability between levels of drums or liners.

2.4 FIRE PROTECTION The only significant potential for fire at the storage facility is an external exposure fire. The facility is of noncombustible construction and d. signed to provide a 3-hour fire resistance rating from external exposure fires.

The fire proteccion water supply is taken from the nuclear plant yard fire main.

Hydrants and hydrant houses are provided around the perimeter of the storage facility in accordance with NFPA Standard No. 24 Two points of entry are provided through the ENCL 2-13 m

SQNP security fence to accommodate standard fire department pumpers. Each storage module compartment has been aized to collect and contain that quantity of water used for manual fire tighting from two 2 1/2-inch hose streams simultaneously for a duration of at least one hour. The storage facility will also be provided with multipurpose dry chemical fire extinguishers in accordance tith NFPA Standard No. 10.

All fires will be fought by specially assigned personnel with support from the SNP fire brigade.

2.5 RADIATION MONITORING AND PROTECTION l

2.5.1 Radiation Monitoring Radiation monitors will be permanently installed only at the security gatehouse. All other necessary l

radiation monitoring will be performed by the plant Health Physics Section using portable equipment.

Monitoring wells in clusters will be provided and placed outside the security fence. The initial well of the cluster was core drilled under the supervision of a l

geologist. Representative ground water samples will be collected before waste is stored in the modules. The design and number of additional wells in the cluster should be determined on the basis of the initial core.

All wells will be fully developed, grouted, sealed, and j

capped to prevent the introduction of any extraneous material. Monitoring well identification markers will be erected above ground.

2.5.2 Radiation Protection Fxcept for trash, the design basis radioactivity levels rf the LLRW will be based on plant operation with expected radioactivity concentrations in the reactor coolant. Because there is no actual data on L RW levels at Sequoyah, the design basis radioactivity levels for trash will be a factor of 10 higher than l

average levels measured at Browns Ferry Nuclear Plant l

through June 30, 1979.

The facility will be designed for implementation of the l

control measures for radiation and high radiation areas j

as defined in 10 CFR 20.

The facility will also be designed such that the probability is small that any l

person would receive a dose equivalent greater than 500 l

mrem during any calendar year in unrestricted areas J

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from storage facility operation (which includes handling operations).

As part of the provisions to ENCL 2-14

SQNP implement this restriction on dose equivalent in unrestricted area, the dose equivalent rate at the storage facility security fence will not exceed 0.6 mrem per hour except during handling operations.

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2.6 QUALITY ASSURANCE To ensure the storage module structures will perform as intended, a quality assurance program will be established and documented.

As a minimum, this program shall conform to the requirements of Regulatory Position 6 of USNRC Regulatory Guide 1.143 2.7 ELECTRICAL REQUIREMENTS The Long Term Onsite Storage Facility (LTOSP) will be provided with electrical power from offsite by the nuclear plant.

8 ENCL 2-156

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-_2.8 FJt{1PMENT CODES "All storage facility equipment snell be designed,' procured, constructed, and inspected in accordance with the codes and standards identified below:

EQUIPMENT CODES DESIGN &

INSPECTION COMPONENT' FABRICATION MATERIALS WELDING

& TESTING PIPING AND VALVES t

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Storage Module ANSI B31.1 (QA 304-L or ANSI B31.1 ANSI B 31.1 Drains-shall be in 316-L accordance with attachment RW of MEB E.P.23.5.5) h.

Storm' AASHTO AASHTO Drains c.

Potable water' National _

National National National and sewers Plumbing Plumbing Plumbing Plumbing Code-Code Code Code d.

Fire NFPA Code NFPA Code NFPA Code -

NFPA Code Protection Standard 24 Standard 24 Standard 24 Standard 24 CRANE Joint Industrial ASTM AWS AISC, ASTM, Council & AISC

& AWS 4

l Decontamin-TVA Spec. C-14 ANSI N512 TVA Spec.

able Coatings G-55 Electrical IPCEA Standards ASTM AWS Security, and Industry Industry Industry Industry Radiation-Standards-Standards Standards Standards Monitoring NENA Standards ANSI-N13.1-1969 Equipment RDT Standards ANSI-N13.10-197'4 RDTCl-lT 4

FIRE PROTECTION A.

Extinguishers NFPA Code NFPA Code NFPA Code Standard 10 Standard 10 Standard 10 B.

Hydrants, NFPA Code NFPA Code NFPA Code NFPA Code Houses, Hoses, Standard 24 Standard 24 Standard 24 Standard 24 etc.

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SQNP 30 FACILITY OPERATION 3.1 HANDLING AND STORAGE OPERATIONS - DESCRIPTION Steel liners containing dewatered resins and 55-gallon steel containers with low-level radioactive waste will be transported to the storage site in shielded casks or in van-type trailers depending on dose rates. Some steel liners containing evaporator concentrates' may also be stored in the storage facility. All shipments will be in compliance with applicable DOT and NRC requirements before transport. Access to the storage facility will be through the facility's main gate. The vehicle will be taken to either a resin module or a trash module depending on the type of radioactive waste to be stored. Each module will be identified by a sign denoting its intended contents and storage status. A gantry crane will then be positioned over the module cell to be loaded, and the top hatch of the module removed and set aside. The vehicle will then be parked under the gantry

~

c rane.

For shipments in a shielded cask, the cask cover bolts will be removed using an air wrench. The cask cover will then be removed from the cask using the gantry crane and set aside. An air-actuated remote lif ting device will remove the liner from the cask and place it into a predetermined space in the module. All operations will be observed on closed-circuit television to reduce esployee exposures. The remote lifting device will then be unhocked. The cask cover will then be replaced and bolted and the cask returned to the plant.

For container shipments in a van-type trailer, the trailer will be parked under the gantry crane. <For trailers with removable tops, the containers will be unloaded directly from the trailer using a remote container handling device and placed into the module. For trailers with rear doors, a movable ramp and forklif t will be used to unload the containers to the outside of the trailer. The containers will then be lif ted into the module using a remote container handling device. All vehicles will be monitored for contamination and excessive dose rates before they are returned to the plant.

Steel grating will be placed between each layer of containers to provide stability. Containers will be stored up to 3-layers high. Steel liners will be stored up to 2-layers high.

When all containers in the shipment have been stored, the cell hatuh will then be placel back on the module. When containers are removed for final disposition or for the contents to be volume reduced, this procedure will be reversed.

ENCL 2-16

SQNP Records kept by the plant will indicate the placement of each container. These records will also indicate the container identification number, curie content, dose rate, type of radioactive waste, and whether the contents are volume-reducible.

All laborers, crane operators, and truck drivers will be furnished by the plant. All operations at the storage facility will be monitored by plant health physics employees. Monitoring activities include vehicle and container surveys before and during shipment module loading and unloading.

Periodic surveys shall also be conducted in the area outside the modules as well as r. gular checks to determine the presence of liquid at the external sampling collection points. Sampling wells in the 4

storage area will be checked on a regular basis to indicate if any radioactive contamination has reached the underlying aquifer.

Because of the potentially high radiation dose rate emitted from inside the storage module, all loading of containers into the module will be done remotely utilizing a closed-circuit television monitor to observe the placement of the container in the module. The monitor will be used to ensure that a container is placed in the correct storage cell without damaging either the container, the storage module or other containers in storage.

The CCTV system consists of two monitors and four cameras, all completely independent of each other except for their power source. Each monitor is equipped with manual control capabilities to select display from any of the four cameras. The cameras are equipped with individual pan and tilt control.

Should a camera fail, no major interruption of the system would result.

Cameras are paired and placed in such a manner that loss of a camera would not affect the operational procedure. The crane is equipped with permanent scaffolds to provide easy access to all cameras.

Replacement of a camera shall be accomplished with a minimum of personnel radiation exposure.

Since each monitor; can survey the area through all four cameras, independent of the other monitor, the loss of one monitor would not demobilize the system. Hcwever, should a monitor fail, activities may be slowed.

Should a total failure occur of either both monitors or all cameras (low probability but possible in the case of a total power failure), one of the following courses of action would be taken, depending on the position of the container:

(1)

If CCTV capability is lost while the container is ENCL 2-17 t

r

SQNP outside of the storage module, the container will be immediately returned to the trailer or shipping cask from which it was removed, and additional security precautions

{,

will be taken until the CCTV is repaired.

~

(2) If OCTV capability is lost while the container is i

inside the module (but not in a stored position),

several options exist.

In any case, the crane and i

container will not be moved laterally. One option involves the use of a portable camera system available from the nuclear plant which can be rigged to observe the container'. At the operations supervisor's option, j

the container can then be either retrieved and placed j

in terporary storage until the CCTV is repaired or j

lowered into place in the module under observation t

utilizing the portable camera system.

At the Operations supervisor's option, container storage may I

be continued utilizing the portable camera.

Another option 13 to repair the CCTV' (if possible) while the crane remains in the position where it was when the CCTV was lost. This may not be possible if the failed equipment is in a high radiation field.

In all instances, the cell cap is to be 4

placed atop the open cell as soon as possible to minimize radiation exposure.

The design of the mobile crane will allow storage operations at night. However, because of the dccre 4 sed visibility, storage operations will normally be carried out only during daylight hours. Night operations will not be undertaken unless plant operations will be affected using only daylight storage operations. Extra lights will be used to increase visibility and to ensure that the CCTV system can be used. Low-level radioactive waste storage operation is not expected to require night operations. Storage operations will not be conducted during inclement. weather, such as rain or enow storms.

Should the cables of the mobile crane lock due to motor or power failure making it impossible for the trolley to transfer the container to its storage position, the container can be remotely lowered into the cask or into the module, if it is directly over j

an open module compartment, by manually releasing the brake.

If the trolley locks in a position that is not directly above a cask

~

l or open module compartment, the container can be moved laterally by driving the crane to a safe positinn and the container lowered

. manually by releasing the brake. The crane is then moved back to an area where repairs can be-made. The container would remain isolated, shielded locally, and guarded until repairs are completed and the crane could return to safely place it in its ENCL 2-18 i

4 j

SQNP above a cask or module is to position the cask trailer in an area clear of the modules such that the crans can lower the liner into the shielded cask. The trailer would then be taken to a secure site for storage until repairs on the crane were completed.

Security Operations Closed-circuit television monitors will be used to detect and observe potential intrusion into the storage area. Monitor screens will be located in the gatehouse.

In the event of the loss of a CCTV, security persont.el will call for immediate repair. A spare CCTV will be put into service, if available.

During the time that the CCTV is out of service, the area normally covered by CCTV observation shall be continuously patrolled by security personnel to ensure that the security of the area is not compromised. Additional security personnel shall be utilized when necessary.

ALARA All employee exposures will be kept as low as reasonably achievable (ALARA). When containers with excessively high radiation dose rates are handled, remote methods will be used.

Only employees required to handle the shipment will be allowed in the area where containers are being handled. All container and vehiale dose rates and contamination levels will be within DOT limits for shipment, if applicable.

Employees and contciners will be monitored during all operations by health physics employees

'.o ensure that dose limits are not exceeded and that good work practices are followed. All operations will be conducted in accordance with written procedures.

3.2 MONITORING OPERATI0tlS 3.2.1 Module Interior Before low-level radioactive waste is packaged, all containers will be visually examined and/or pneumatically tested to ensure that the container is not daEJged and is Ir.aktight. Modules will not be opened during inclement weather, such as rain or snow storms, to prevent unnecessary introduction of water into the module. Hatch sealing surfaces will be examined to ensure that they are in proper condition.

The sump in each module will be sampled periodically to detect the presence of water and/or radioactive releases in the module compartment.

Detection of water or radioactive releases in a module compartment will require an intensive check of all containers and the ENCla-19 4

SQNP inside of the module to determine the source.

Corrective actions, including repackaging of a leaking container or repair of a defective hatch seal, will be undertaken.

Radioactive liquids will be collected and transported to the plant for processing by the radwaste system. Nonradioactive liquids will be disposed of in accordance with established plant disposal operations.

3.2.2 Environment Sampling wells in the storage area will be checked on. a regular basis to indicate if any radioactive contamination has reached the underlying aquifer. The area outside the module walls will be checked routinely to detect leakage.

i 3-5 4

1 j

ENCL 2-20 1

l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _I,

SNP 4.0 RADIOLOGICAL CONSIDERATIONS 4.1 RADIOLOGICAL ASSESSMENT TVA has performed a radiological assessment of the long-term waste storage facility, covering the operational releases expected as a result of operator error or equipment malfunction and the releases from an accidental fire.

The attached Tables 4.1-1 and 4.1-2 present the major assumptions used in the assessment and the results of the assessment, respectively.

9 9

D ENCL 2-21

_=-.- -........_

-.. - -- - -- - -. ~. -.,

TABLE 4.1-1 i

SEQUOYAH NUCLEAR PLANT MAJOR ASSLRIPTIONS FOR RADIOLOGICAL ASSESSitCNT OF LONC-TERM t

ONSITE LOW-LEVEL RADIOACTIVE WASTE FACILITY j

+

t General Type of vaste LLRW - miscellaneous trash, spent resins, concentrate liquids, and spent regenerates.

i i

I Activity 54,500 Ci/yr of resins, regenerates, and concentrate liquids.

44 Ci/yr of trash.

a Isotopic breakdown About 89 percent Cs-137 and Ba-137 m, 1 percent Co-60, 2 percent

^

Co-58, 7 percent Cs-134, and 1 percent other fission, activation, and corrosion products for all low-level vaste except for trash.

i.

I OSF Operational Releases

.g Mg Under normal operation, any potential Iccchate in the storage modules will be collected and sampled

,i, prior to release.

However, it is postulated that due to operator error or equipment malfunction, a certain portion of the esticated annual leach reaches the river via ground water.

N I

i Maximum stored activity Releases are assumed to occur when the activity in the LTOSF reaches a 7.axinum (at 40 { cars). The 40-year activity is estimated at about 1.3x10 Ci composed of essentially all i

Cs-137.

Annual leach fraction One percent of the 40-year activity is assumed to leach out 4

of the storage container per year.

Travel distane.e to river 91 m (300 ft).

Groural water velocity 1.5 m/d (5 ft/d).

b

TABLE 4.1-1 (continued) i Total soil porosity 50 percent Bulk soil density 1.6 g/cn' Distribution coefficient (K )

500 cra'/g for Cs d

River dilution Spillage is as'sumed to mix with one tenth of the average rive-3 3

flow (3.1x10 ' cm /yr) before reaching potential receptor.pattways.

LTOS'F Accidental Fire Release Non-volume reduced LLW trash from one compartmcut of a storage module (about one fourth of one year's waste) is assu:7.cd to catch fire due to an unspecified incendiary event.

Activity in module section About 10 C1 Co-60 Fractional release from fire 0.01 for particulates" E

~

9

>]Q (fif ty-percentile, I hour, 1.6x10 ' s/m' Y

ground level U$

Distance to site boundary 1,000 m (3,280 ft) a.

1.'AS H-12 3 3.

1 e

i e

I

. ~..

TABLE 4.1-2 SrQUOYAll NUCLEAR Pl. ANT SU101ARY OF RADIOLOGICAL ASSESS > TENT FOR LONG-TERM O!1S TE LOW-LEVEL RADIOACTIVE WASTE FACILITY LTOSF Operational Releases Leach to river 3.0 meen/yr (whole body)

(10 CFR 50 Appendix I

.04 mren/yr (thyroid) guidelines:

6 mren/yr -

whole body; 20 mrem /yr - organ LTOSF Acc.idental Fire E

Air submersion dose at

.09 mrem (whole body) site boundary 5

Inhalation doses at site

.08 mrem (whole body) boundary (10 CTR 20 guide-30 mrem (lung) lines:

503 neem /yr to whole body and 1,500 mrem /

yr to individual organs other than the thyroid)

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4.2-INCREMENTAL OCCUPATIONAL EXPOSURE

~

Annual occupational, personnel exposures have been estimated for i

the handling and placement of low-level waste in the proposed long-term' waste storage facility.

Dose estimates are given in Table 4.2-1 and ! include exposures. to waste handlers, health physics monitors, crane. operators, nuclear plant employees, and transport personnel.

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1 ENCL 2-25

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4-TABLE 4.2-1 Sij.QUOYAll NUCT. EAR PLANT OCCUPATIO::AL DOSE ESTIMATES FOR LONG-TER?! LLW STORACE FACILITY General Assuqptions 1,

Maste is volene reduced into 55-gallon drums with, reduction factor of 15 for resins and 36 for combustible trush.

2.

About 5.4 drums of resins, 149 ' drums of concentrate liquids, 530 drums of spent regenerates, 68 drums of combustible trash, i.nd 817 drums of noncombustibic trash are stored per year.

f rom individual drums are 1.0x10 R/h fo*: resins, 1.4xlO? R/h for concentrate 3.

txposure ra:es at contact 3.2 R/h for combustil.le tr.sh, and 0.13 n/h for noncutbustible trash.

liquids, 2.3 R/h for spent regenerates, i'

4.

The number of curies in the facility is about 1.3x106 Ci.

5.

The aversee nurber of curies in a resin module cell is about 11,500.

6.

ILodules have 3.5-ft concrete walls (trash odules have 2.0-ft walls) and 2.0-ft concrett ap (cap is removed for waste placement).

Person-r.s r

t 5

p Transport Personnel Y

0.1 y

One driver exposed to 2 mR/h, 50 h/yr.

Crane Operator Two operator.. exposed to.156 R/h", 141 h/yr; sad 1.7x10 R/h", 323 h/yr.

45.1

~3 Waste Handlers

-2 b

Two handlers and one health physics technician exposed to: 3.2x10 R/h. 78.5 hr/yr; 28.0

' y29 i

O.01 R/h", 44 h/yr; 3.9x10 R/h, 141 h/yr; and 3.6x10 R/h", 323 h/ yrs G

3 2

3 R

Honitoring Personnel M

One health physics staff member exposed to 1.5x10 R/h, 156 h/yr.

2.4

-2 O

a.

Includes direct and skyshine radiation from facility during waste placement; crane operator is ggyg to module wall, vaste handling individuals assumed to be 40 feet from the facility.

adjacent from drum.

bg c) b.

Average c <pesare rate during remote handling of drums assuming workers at 40 feet Exposure during cask removal fcc only one worker.

,<<2 d c.

d.

Includes direct and skyshine radiation from facility with module caps in place; monitoring personnel assumed to bc 10 feet from facility.

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TABLE '. 2-1 (continued) 1;ucicar plant Employees

__ _ Person-rem /yr 22.2 1.

Distance is approximately 610 m (2,000 feet) to plant.

2.

2,500 persons exposed,. ssu:nin;; no shieldtu;- by building.

j Exposed to 1.7x10 ' 1:/h, 2,033 h/yr.

Exposed to 3.9x10 It / h",

141 h/yr.

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SNP 4.3 DOSES TO UNRESTRICTED AREAS The doses in unrestricted areas due to waste handling and storage in the long-term waste storage racility have been calculated and are given in Table 4.3-1.

These doses include direct and skyshine radiation from the facility to the site boundary, nearest resident, and nearest onsite non-nuclear facility.

All assumptions are the same as for Sections 4.1 and 4.2, except as noted in Table 4.3-1.

v 9

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ENCL 2-28

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TABLE 4,3-1 SEQliOYAll NUCLEAR PLANT - DOSES IN UNRESTRICTED AREAS mrem /yr Site Boundary 1.

Distance to nearest site boundary is about 1,000 m (3,230 feet) from nearest 5

I codule.

2.

Exposed to 9.2x10 cR/h",141 h/yr and exposed to 4.1x10 ' mR/h, 8,766 h/yr

-3 Nearest Resident 1.

Distance to nearest resident is also about 1,000 n '(3,280 feet) from nearest 5

module.

2.

Exposed to 9.2x10 mR/h", 141 h/yr and exposed to 4.1x10 ' mR/h, 8,766 h/yr.

-8 x

5

[

a.

Includes direct and skyshinc radiation during vaste placement.

b.

Includes direct and skyshine radiation from facility with cap in place.

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SNP 5.0 ENVIRONMENTAL ASSESSMENT 5.1 ENVIRONMENTAL' IMPACTS OF THE PROPOSED ACTION 5.*

Construction-Related Impacts Construction impacts associated with this project include fugitive dust, gaseous emissions, siltation, noise, socioeconomic and potential impact on existing structures at Sequoyah.

Construction began in March 1980.

Present planning schedules indicate that the first LLRW storage module will be available by December 1980, with additional resin and trash modules to be completed by December 1981.

Air Quality The construction activitics associated with the radwaste storage facilities will result in some temporary degradation of local air quality.

Air pollutants generated from this activity will primarily include (1) fugitive particulate emissions from various activities including cleaning of steel and concrete, drilling, painting and mixing concrete (in a batch plant); (2) fugitive dust from earth excavation and grading; (3) particulate emissions

^

from the open burning of small amounts of wood scraps; (4) small amounts of particulates, hydrocarbons, nitrogen oxides and carbon monoxide emissions from fossil-fuels construction and construction employee vehicles.

The construction site mitigation program will consist of fugitive dust suppression, by methods such as water sprinkling, which will substantially reduce this problem.

Open burning will be conducted in accordance with all applicable Federal, State and local regulatory requirements.

Concrete production during construction of the LTOSF will be approximately 150 cubic yds per hour from the existing onsite concrete batch plant.

Fugitive emissions frem the concrete batch plant will be controlled through the use of filters.

Land Use Impacts The construction of the LTOS/ as currently conceived will require approximately 20 acres of land all within the Sequoyah reservation boundary.

Approximately 1,160,000 yds 3, vet of a total of approximately 1,320,000 yds 5, of spoil has been deposited in spoil areas (approximately 35 acres total area at the base elevation).

The proposed ENCL 2-30

SNP action involves no offsite land use conflicts.

The proposed action is compatible with the land-use plan within the Sequoyah reservation for the nuclear plant and its support facilities.

Siltation During construction of this facility, construction runoff will be drained in a manner that will prevent erosion and minimize the amount of sediment reaching local bodies of water.

All construction runoff will be treated by rock filter dams, coffer dams, straw, etc.,

in accordance with best management practices developed by the Environmental Protection Agency (EPA) pursuant to the Federal Water Pollution Control Act ( Guidelines for Erosion and Sediment Control Planning and Implementation, EPA Environmental Protection Technology Series--EPA-R2-72-015, August 1972).

The existing topography of the construction site preclude the use of sedimentation ponds other than the existing yard drainage pond.

All runoff will be discharged through EPA approved NPDES permit discharge locations.

With these precautions, construction activities are not expected to have a significant impact on water quality.

Noise The usual sources of noise associated with construction activity will be present.

However, these impacts ara expected to be minor and limited to the site area.

Solid Waste There will be a small amount of solid waste generated due to the construction of the LTOSF.

Solid wastes generated during construction will be handled in accordance with applicable State and Federal regulations.

S.nitary Waste During the construction period, ;;ortable chemical toilets will be provided for use by construction personnel.

There will be no liquid effluent from these facilities.

Sanitary wastes generated during construction will be handled in accordance with applicable Federal, State, Regional and local regulations.

l 1

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ENCL 2--31 l

- ~,

SNP Cultural Since the proposed action will be constructed entirely within areas the Sequoyah reservation that have been previously di

' bed during the course of construction of Sequoyah the pt

sed action will have no further ef fect on any known archaelogical or cultural resources.

Endangered or Threatened Species No known population of endangered, threatened, or otherwise sensitive species will be impacted by the development of the proposed project.

Floodplains and Wetlands The site for the proposed action is not located in a floodplain nor is it expected to directly or indirectly support or encourage floodplain development.

There are no wetlands which will be affected by the project.

Socioeconomic The proposed action will require a significant construction effort in view of the urgency of the situation.

There is now and will continue to be significant ongoing construction of Sequoyah unit 2 and there is manpower, housing, and services available in the area to fill the construction and labor skill requirements for the LTOSF.

As a result of an adequate supply of manpower, no overall population increase is expected as a result of this construction activity and because this plant is near an urban area (Chattanooga, Tennessee), no significant socioeconomic impacts are expected.

5.1.2 Operation of the LTOSF Air Quality Operation of the LTOSF will have no significant effect on air quality.

Water Quality The operation of the LTOSF will not result in a liquid waste stream including during e.7ergency conditions (i.e.,

fire).

There will be no discharge of a liquid effluent into the surrounding aquatic environment.

Shnitary facilities will be provided in the LTOSF ENCL 2-32

SNP gatehouse, but the liquids will be trea'ed either by piping them directly to the existing sanitary system at the Squoyah, or by installing a small subsurface sanitary waste treatment and disposal system dedicated to the LTOSF.

The small flows expected from the LTOSF sanitary facil) ties (normal occupancy--two people) would not impact operation of not hinder operation of Sequoyah's sanitary waste treatment system.

Subsurface treatment 6 -

.23F sanitary waste would not result in a liquid effluent to be discharged.

No NPDES permit or permit modification will be required for operation of the LTOSF.

1pise Noise, impacts onsite or offsite, from the operation of the LTOSF will be minimal and should not have any significant impact on the site area.

Solid Waste Management The Resourae Conservation and Recovery Act of 1976 (RCHA) specifically exludes nuclear material regulated under the Atomic Energy Act of 1954, as amended (which covers LLRW).

Because the operation of the LTOSF will result in no significant additional amounts of solid waste to be handled, other than LLRW, the proposed action for these facilities does not have solid waste management impacts associated with it.

Should solid and hazardous wastes other than LLRW be generated, they would be managed in accordance with applicable EPA regulations for solid and hazardous wastes.

5.2 UN AVOIDABLE A DVERSE ENVIRONMENT AL IMPACTS There are no significant environmental impacts associated with the construction and operation cf the LTOSF.

During construction, some siltation may occur and the relecse of small amcunts of gaseous and particulate pollutants can be expected.

8 No significant cumulative impacts have been identiftec.

5.3 IRREVERSIBLE AND IRRETRIEVABLE COMMITMENTS OF RESOURCES Irreversible and irretrievable commitments of resources will include fuel oils involved in the construction of the proposed facilities along with materials used for the construction of the LTOSF.

j FNCL2_33

SNP j

6.0 SAFETY ANALYSIS 1

LTOSF MODULE DAMAGE 6.1.1 Dropping a Storage Module Cap Into a Module i

The worst case incident would be dropping a storage cap into a recently filled resin liner storage module about te be closed.

The cap could only drop into the module compartment'if at least two of four suspension cables failed and the cap were suspended higher than approximately five feet above the open module compartment.

The cap could then conceivably fall into the open module, rupturing LLRW containers and causing abrasive damage and possible minor fracture to the walls l

of the modules.

The cap could also be damaged.

No release of radioactivity would occur since all material

-would be contained within the module.

Plant personnel can remove the low level waste, repackage it, and transfer it i

to an undamaged module cell, leaving the damaged cell to i

be deco

• Laminated and repaired as required.

)

Two design features are employed in the prevention of 1

dropping a module cap into a module, making this accident event highly improbable.

Crane lifting cables

+

and lifting lugs are designed to withstand five times.the maximum operating load expected and the cap lifting heights never exceed 5 feet above an open module.

By designing the lifting cables and. lugs to withstand five times the maximum operating load, failure of a cable or a lifting lug during the handling operation is highly improbable.

Therefore, the dropping of a cap is also j

highly improbable.

However, should the cables or lugs fail, an additional operational measure is employed to prevent the. cap from falling into the storage. module.

The i

c a'p is 9 feet 5 inches wide.

For i; to rotate into a position from which it could conceivably fall into the module, would require it to be in excess of 5 feet hight than the upper most rim of the module.

By not lifting the cap higher than 5 feet above the module, it becomes impossible-for the cap.to fall into the module compartment should the cables or lugs fail.

f.s noted previously, the failure of a cable or a' lug is highly improbable.

Therefore, the combined probability of this accident is very low, thus, the associated risk is very low.

6.1.2 Dropping A Storage Module Cap Onto a Module The worst case incident would be the dropping of a cap onto a recently filled open module compartment about to be I

ENCL 2-34 i

.m..

S SNP closed.

Da mage could occur to both the cap and the module walls, consisting of possibly abrasive damage and fracture.

It has been determined that neither the module salls nor the cap would collapse into the module's interior.

Because LLRW containers are spaced away from the module walls, only slight damage to the containers from falling bits of concrete would be encountered and the consequential release of radioactivity would be minimal.

Plant personnel can gain access to the containers and LLRW, remove the LLRW and repackage where needed, remove the und L-2Eed containers to an undamaged module and leave the damaged containers to be decontaminated and/or repaired.

The probability for this accident is low since the only source of the accident is the crane's lifting lugs and cables which have been designed to withstand five times more stress than the maximum operating stresses expected.

However, to further reduce damage to the module by dropping a cap, a cap lift height of 5 feet maximum for the modules will be administratively imposed on daily crane operations.

Up to this height, dropping a cap would cause only.tinor abrasive damage to the module and cap,

~

leaving the integrity of the cap seals unimpaired.

6.1.3 Dropping a Storage Module Cap Onto Another Cap See 6.1.2 6.1.4 Dropping a Storage Module Cap Onto the Ground Should a storage module cap fall to the ground, only the cap would suffer damages.

No radiation health hazard would result.

The cap could be replaced with either a spare or a cap taken from an unused storage module.

To eliminate the possibility of the cap falling to the i

ground, the overground time of transporting the cap shall be kept to a minimum.

6.1.5 Collision of the Mobile Crane ar Transport Vehicle with Storage Module Due to the slow speeds, involved collision of either the mobile crane or transport vehicle with a storage module would result in only minor abrasive damage to the module wall.

No release of radioactivity would be involved.

In the LTOSF, the transport vehicle will be moving no faster than 20 mph and the mobile crane moves no faster than 5 ENCL 2-35

..s SNP 4

nph.

With the combined resistive forces of the 8-inch curb and the steel reinforced concrete module wall, an impact at less than 20 mph with a storage module by a fully loaded transport vehicle would result in only abrasive damage to the module wall with no significant i

impact on the structural' integrity of the module.

6.2 LLRW CONTAINER DAMAGE i

6.2.1 Dropping a LLRW Container Into a Module The worst case incident would be dropping a dewatered resin liner intn an open module with a resulting 100 i

percent spillage.

Since the release would te contained within the module, plant personnel can remove the radioactive material through the module drainage connections to a new liner and cask assembly and locally decontaminate the storage module.

The potential for this accident event to occur is highly 1

improbabic due to design considerations.

Lifting cables and lugs are oenigned to withstand five times their maximum 'perating stress a n d,- therefore, are not expected 4

to fail.

Additionally, liner and 55-gallon-drum lifting devices are designed not to release unless driven to do so i

by a pneumatic or mechanical force delivered by the operator.

The switches that deliver these forces are totally segregated from the centrols that position the trolley and crane, thus reducing confusion.

In the case of the liner lifting device, should the pneumatic system 4

fail, the device is designed such that its center of gravity and configuration lets it keep a firm grip on the liner.

As for the 55-gallon drum lifting device, the drum l

is held basically by its own downward force.

The only way to-release a drum is to completely remove its downward force on the lifting device by setting the drum down on i

the ground or other surface and mechanically driving the gripping claws apart.

6.2.2 Dropping a LLHW Container Outside'of a Storage Module The worst case incident would be the dropping and the j

sutlequent release onto the LTOSF grounds of the entire-

' contents of a 'dewatered resin liner.

Appropriately J

protected LTOSF personnel could collect and repackage all spilled LLRW and contaminated soil to locally decontaminate the area.

Should the rupturing of a dewatered resin liner with a subsequent loss of all LLRW onto the LTOSF grounds be followed by a rainfall, no significant amounts 1of radioactivity would be expected to ENCL 2-36

SNP enter potential drinking water sources due to the long distance to the river and sorptive soil properties.

6.2.3 Dropping a Shield Cask Lid Onto Its Open Cask No damage to the liner within the cask can occur nor would there by any damage to the cask itself.

Precautionary design considerations given to the crane's lifting cables are noted in 6.1.1.

6.2.4 Collision Between a Transport Vehicle and the Mobile Crane The likelihood of a collision between a transport vehicle and the mobile crane is reduced since both machines are not in motion at the same time.

The only time that such an accident has the potential to occur is when the crane straddles the module and transport vehicle.

The wheel of the crane is kept from the transport vehicle by an 8-inch-high curb.

Should the crane override the curb, its speed is too slow to do any damage to either the LLRW containers or to the transport vehicle.

6.2.5 Loaded Transport Vehicle Fire or Explosion We have determined that a liner in its transportation cask is safe from fire or explosion through a review of the sa fety performance history of interstate transit of LLRW.

Explosion or fire of a transport vehicle carrying the 55-gallon drums is likewise considered to be highly unlikely.

6.2.6 Sabotage of LTOSF Damage to LLRW containers inside shipping casks, as well as inside the storage modules, as a result of sabotage is highly improbable due to the high quality of the design and security measures employed.

Sabotage of the LTOSF is c or.s id e r e d a highly improbabic occurence for the following reasons:

(1)

A security fence is provided around the facility.

The grounds and surrounding area are monitored by CCTV and patrolled by security personnel on an around-tae-clock basis.

The facility is completely illuminated at all times, including tops and sides of all structures within the security fence.

The fence

+

totals 8-feet tall topped with 3 strands of barbed wire and is bounded by a total of 40 linear feet of ENCL 2-37

f s

i L

SNP i

cleared land entirely around its perimeter.

It is j

also equipped with an intrusion detection system and has a tamper indication attached to an alarm in a security.gatehouse.

The security gatehouse will i

contain the CCTV monitors and facilities for telephones and PSS radio.

All locking devices for doors and' gates will be key-controlled, high-security 4

locks.

Altogether, the security system provides 1

advanced ~ intruder detection,. penetration determent and rapid communication and alert capabilities that j

make' sabotage virtually impossible.

1 (2)

The wastes stored at the LTOSF are low in radioactivity, making them unlikely targets of.

sabotage.

(3)

Storage modules and resin liner transportation casks are extremely shock resistant because of their design to withstand transportation accidents and seismic activity.

4 i

6.2.7 Liner Breech ~Due to Freezing of Resins I

The breech'or a' liner by the crystallization expansion due to freezing of water mixed with resins inside the liner is improbable Experimental results submitted to the NRC by CPU Service Corporation dated November 17, 1979, proved 1

that resin liners, filled to capacity with dewatered i

resins, do not rupture when freezing of the dewatered

]

resin contents occurs.

6.3

SUMMARY

In summary, to provide long term LLRW storage facility integrity and longevity, the storage modules, associated facilities and equipment will be designed, operated and maintained in order to 4

minimize the consequences of, if not totally eliminate, the potential for these highly unlikely accidents to occur.

i 5

I l

ENCL 2-38

SNP t

7.0 DECOMMISSIONING Low-level radioactive waste scorage in the proposed facility is planned to begin as needed (but not earlier than December 1980) and end at the end of the operational life of the plant.

Near the end of plant life, a final decision will be made as to the method for decommissioning the storage facility.

There are ri r re n t ly several options under consideration for decommissioning.

These options are:

1.

Placing the storage facility in an inactive state and providing a security and monitoring force for an indefinite time.

2.

Sealing all radioactive material inside the storage facility (utilizing a material such as concrete) in a technique known as entombment.

3 Retrieving all radioactive waste containers and transporting all of this material to another facility.

The storage site can then be decontaminated as necessary, leaving the area in as close to its original state as possible.

This method may also involve dismantling and removing the storage facility.

No specific method will be selected at this time since actual deccamissioning for the storage facility will not be necessary for approximately 30 years.

Other methods may be developed in this time period which are more advantageous than the above methods.

It is TVA's intention at this,ime to retrieve all stored radioactive waste.

If, for some reason, TV A cannot remove the stored material for offsite disposal, the material can remain in the storage facility.

Security and environmental monitoring precautions will be continued until either the material is disposed of offsite or the facility is released for unrestricted use.

The viability of the latter option depends en container dose rates and specific activities as well as a regulatory definition of the radiation and concentration IcVels which are considered acceptable.

Although the exact decommissioning method will not be determined until needed, the third method above is preferred by TVA at this time.

After all containers of low-level radioactive waste have been removed and the structure is no longer needed as a radioactive waste storage area, the facility will be decommissioned in accordance with the following guidelines:

1.

TVA will make a reasonable effort to eliminate residual contamination.

ENCL 2-39

I i

r SNP 2.

No radioactive surface will be covered by painting, plating, j

or other method until it is known that contamination levels, to be determined by a survey and docume.;ted, are below a reasonably achievable limit.

3.

TVA.will use the limits presented in Table I of USNRC Regulatory Guide 1.86 as guidelines.

4.

Before release of the premises for unrestricted use, TVA will make a comprehensive survey establishing that the level of residual contamination is as low as reasonably achievable.

A report describing the survey and its results will be filed with the Director, Office of Nuclear Reactor Regulation, U.S.' Nuclear Regulatory Commission, with copies to the i

Director, Office of Inspection and Enforcement Headquarters and to the Director, Region II OIE office.

The survey report will be filed at least 30 days before the planned date of abandonment of the storage area.

5.

TVA may use the storage facility and surrounding area for other plant operations following decommissioning.

TVA is also implementing the following guideline criteria and additional considerations related to facility decommissioning:

1.

Decontaminable coatings will be used in the onsite storage facility to facilitate decontamination.

2.

Materials that cannot be decontaminated to the unrestricted levels identified in Table I of Regulatory Guide 1.86 will be disposed -of by transporting to a permanent disposal site,.the same as for the radwaste containers.

3 Materials that meet-the unrestricted levels of Regulatory Guide 1.86 will be disposed of in routine fashion.

1 1

i

.e ENCL 2-40 1

w ENCLOSURE 3 CONSTRUCTION SCHEDULE LONG-TERM, LOW-LEVEL RADI0 ACTIVE WASTE STORAGE FACILITY SEQUOYAH NUCLEAR PLANT e

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v FACILITY CONSTRUCTION SCHEDULE An evaluation was performed pursuant to 10 CFR Part 50.59, in which it was concluded that construction of the storage facility did not constitute an unreviewed safety auestion. Therefore, NRC's approval was not needed before initiation of construction activities.

Initial phases of construction began in March 1980.

The present schedule calls for completion of one year's storage capacity by December 1, 1980. Completion or all modules is targeted for December 1982.

The accompanying figure shows the major milestones of constructions. Five years worth of storage of a9-produced waste represents long-term storage for volume reduced waste for the remaining life of the plant.

ENCL 3-2

e e

LO2-LEVEL i'ASTE STORAGE FACILIP' CONSTRUCTION SCFEDULE 80 81 S2 83 JAN JA's JA';

JAN LONG TER't ONSITE CS CC ST01 AGE FACILITY (LTOSF)

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"~~~

1*

YEAR CAPACITY 5/1/81 9

2nd YEAR CAPACITY f

~

G 3rd YEAR CAPACITV 8/1/82

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4th YEAR CAPACITY 1/1/83 5th YEAR CAPACITY 6

CS Denotes start of construction (March 1980) 1 CC Denotes corpletion of construction (January 1,1983)

Denotes one year storage module capacities complete o

NOTE: The LTOSF may be expanded in the future if the need arises; however, the total number of storage modules shall not exceed 18.

ENCLOSURE'4

SUMMARY

VOLUME REDUCTION AND SOLIDIFICATION SYSTEMS FOR SOLID AND LIQUID LOW-LEVEL RADIOACTIVE WASTE (LLR")

PROPOSED FOR TVA NUCLEAR PLANTS INTRODUCTION The radioactive waste volume reduction and solidification systems proposed for TVA nuclear plants combin0 the processes of evaporation and incineration to process a variety of liquid and/or chemical wastes, spent ien exchange resins, filter treatment sludges, contaminated lubricating oils, and miscellaneous contaminated combustible solida such as paper, rags, protective clothing, and wood.

The volume reduction system can reduce the volume of the.LLRW which is produced at a particular nuclear plant by an overall factor of 10 or more.

Equally important to reducing the volume of the LLRW generated at a nuclear plant is the final form of the LLRW to be stored or disposed of.-

The solidification system will immobilize in a suitable binding agent the granular solids and ashes that result from the volume reduction of LLRW by evaporation and incineration.

DEFINIYIONS (1)

Low-Level Radioactive Waste - That waste which is contaminated with radioactivity such that it must be packaged for safe handling and transportation to a USNRC or state licensed commercial disposal facility.

e (2)

Evaporation - To remove liquid or moisture by heating so as to make dry or re1uce to a denser state.

(3)

Incineration - To Lurn or reduce to ashes.

i VOLUME REDUCTION SYSTEM The volume reduction system to evapor-te and incinere.te the LLRW arising from power generation at TVA nuclear plants will result in an overall reduction in the volume of LLRW by a factor of 10 or more.

. Incineration takes placa inside of a metal process vessel that is i,

normally 24 to 48 inches in diameter and 20 to 30 feet tall.

The vessel is constructed of acceptable corrosion-resistant materials and i

2 may be either re fractory lined or provided with an exterral cooling shroud in order to maintain external surface temperatures at acceptable IcVels.

Combustible materials such as spent ion exchange resins, filter treatment sludges, contaminated lubricating-oils, and miscellaneous,_ contaminated combustible solids are reno *,ely fed to the incinerator vessel' at a controlled rate to maintain process i

temperatures and are subsequently reduced'to ashes or dry powder.

The gases resulting from the incinerator process which exit the incinerator vessel are_ treated for removal of entrained particulates ENCL 4-1 4

r,

~' '

' ^ ~ ~ ' ' '

and hazardous gases hy an efficient off-gas cleanup system which consists of wet scrubbing solution (caustic), a high efficiency particulate filtration system followed by continuous radiation monitoring and control.

If preset acceptable radiation levels are exceeded, the incineration process is automatically shut down and the gaseous release terminated.

Water, not released as water vapor to the atmosphere, is collected in the wet scrub solution tank and can be recycled to the evaporation portion of the volume reduction system.

Several types of equipment are availabic to process the liquid and/or chemical wastes which are collected at a nuclear plant.

Included in this list are fluid bed dryers, wiped film type evaporators, and extruder-evaporators.

In some processes, the aqueous slurries are reduced to dryness and immobilized in an acceptable binding agent in one step whereas in other processes such as the fluid bed dryer, the water is driven off to leave behind the dry solids in the form of a free-flowing granular material which is s u b s e q u e n t ly immobilized.

Each of the processes result in a reduction in volume with the fluid bed dryer achieving the maximum volume reduction attainable for, liquid wastes.

SOLIDIFICATION SYSTEM The free-flowing granules and ashes resulting from the volu-reduction system must be USNRC regulations be immobilized and packaged in a suitable container prior to their being transported to a commercial disposal facility.

TVA believes that these same requirements should be applied to the solid granular residues before they are stored onsite.

Thus a solidification system will provide a solidification agent tn A 111 immobilize the granular solids and ashes resulting from 1e volume reduction pcucess.

Several solidification agents are available; however, the properties that TVA considers of primary importance in the choice of the particulac binding agent are:

1.

low leachability of radioactivity 2.

high thermal conductivity 3.

chemical stability 4.

resistance to radiation effects 5.

mechanical ruggedness 6.

noncorrosiveness to container 7.

minimum volume 8.

flexibility in its application 9.

lowest cost to achieve acceptable results VOLUME REDUCTION FACILITY STRUCTURE TVA proposes to construct a new and separate building to house the volume reduction system and solidification system equipment at each nuclear plant site, where existing building space is not available.

The new structure will be approximately 80 x 140 x 70 feet tall and will be constructed adjacent to auxiliary plant facilities in the immediate vicinity of existing LLRW processing areas.

The design of the structure will be in accordance with all applicable regulatory ENCL 4-2

requirensnts and-as a-ninigun will-satisfy NRC Regulatory Guide 1.143,

" Design Guidance for Radioactive Waste Management Systems, Structures, and Components. Installed in ' Light-Water-Cooled Nuclear Power Plants."

Construction of the facility will be conducted in such a way as to not adversely impact the safety or continued operation.of the nuclear plant's operating units.

RADIOLOGICAL AND NONRADIOLOGICAL IMPACTS Of primary concern to TVA in deciding-to pursue the volume reduction of LLRW was the potential radiological ard nonradiological impacts on the overall plant. operation due to the incorporation of the volume reduction system.

These impacts were investigated by TVA for the different volume reduction systems available.

Topical safety analysis reports for several different. volume reduction systems are currently.

-undergoing review or have been approved by the USNRC.

In particular, the topical. reports for the Newport News Industrial and Aerojet volune reduction systems have been-under review by the USNRC since June 1977 and October 1979, respectively.

A review by TVA of these topical reports reveals that on a generic basis.the impacts of these systems on overall-nuclear plant operations from a radiological and nonradiological standpoint are minimal and well within acceptable l

limits.

EXISTING PROCESSING SYSTEMS The conventional type LLRW processing systems which are presently installed at TV A's Browns Ferry or Sequoyah and Watts Bar Nuclear Plants were procured, designed, and installed several years ago.

Neither of these installed systems produce a final solidified product that will satisfy the dispo.al facility requirements and federal regulations that exist today or that will be implemented in the near i

future.

As such, TVA has been evaluating the alternatives available j

and considering implementation of new systems and equipment.

The 4

proposed volume reduction and solidification systems will produce in a safe, efficient, and cost effective manner a final solidified product that will satisfy all current and pending regulations known to TVA.

THE NEED Since the startup of Sequoyah Nuclear Plant Unit 1, TVA has packaged and shipped the LLRW generated at Sequoyah to Chem-Nuclear Systems, Inc.'? (CNSI), commercial radioactive waste disposal site near Barnwell, South Carolina.

In the past few months, however, I

significant restrictions have been placed on the amount of packaged LLRW that CNSI will accept for disposal.

Depending on the amount.of TVA's' monthly allocation, without volume reduction, Sequoyah may generate more additional LLRW than CNSI will accept from TVA.

CNSI has announced a policy that will result in further restrictions on the volume that TVA can send to the facility in the very near future, and it now appears that acceptable disposal space will become exceedingly scarce and expensive within the next 10 years.

The problem of available disposal' space at CNSI for the LLRW generated.at ENCIA-3

'.)

l

Sequoyah willfbe significantly worsened as other TVA nuclear plants come on line, because_the announced

posal restrictions are being i

applied.to each utility as' opposed"t; each plant. Even without these

-restrictions it is likely that volume reduction will be needed because no other waste disposal facilities are being currently planned in the

-southeast or midwest regions of'the nation.

LThe need to ensure the ability of disposing,of LLRW at CNSI's facility i*

is.immediate and represents.a potentially serious impact to the future i

operation of.TVA's nuclear plants.

The intent of the proposed ac tion is to ensure that the unavailability of commercial disposal space will not: restrict future electric power generation at TV A's nuclear plants.

TVA's existing and planned nuclear plants will contribute significantly to the TVA electric power system anc will increase the a

reliability of the system.

Continued operation of these plants will increase TVA's ability to comply with the Nation's policy of attaining energy independence and'could continue to minimize future dependence on foreign oil._

Implementation of-the proposed action will make operation of TVA's nuclear plants considerably less dependent on restrictians on disposal of LLRW for the foreseeable fu tu re.

The need for.immediate action requires that TV A action be initiated promptly.

If no action is taken to provide volume reduction of LLRW a t TVA nuclear plants, TVA's ability to continue nuclear plant operaticns may be severely hampered.

Operating costs and the potential for health effects will also increase substantially.

The availability of commercial disponal apace in the fu ture is at best uncertain.

At the present time the public concerns regarding nuclear waste disposal makes it u nl ik e ly that new LLRW disposal facilities will be o

commissioned by private or governmental groups soon enough to prevent significant impact on TVA generating capacity.

If no action is taken at this time,-TVA would-have to consider limiting or ceasing opbration of its nuclear plants at some time in the future.

Furthermore, if power generations are halted, the nuclear plant would still continue to generate small amounts of LLRW'due to plant maintenance and decommissioning' activities.

Delaying action at this time would offer TVA no advantages'in resolving the present and future LLRW needs.

Delaying action now would.only make-the situation that much more difficult when action is mandatory.

There a.*e no foreseeable occurrences which would help

alleviate the. situation in the short term that could jus tify TVA 's waiting before taking any action.

Therefore, delaying action would have.the same effect as taking no action.

TVA's assessment indicates that taking no action or _ delaying action - would severely curtail electric power. generating capability at a period in which use of i

domestic energy sources must be maximized.

Jeopardiration of the operation at TVA's nuclear plants must be avoided, because of the need for power.

Therefore, neither the no action nor delayed action-alternative is 'accep table.

I ENCL 4-4 I,

SUMMARY

The overall concept of LLRW volume reduction, solidification, and packaging provides a method of LLRW management which is very cost effective and amenable to a variety of LLRW management concepts, such as long-term onsite storage, regional waste disposal / storage facilities, shallow land disposal, and deep geological dier7 sal.

The combined use of volume reduction systems and solidification systems provide substantial improvement over present methods of LLHW treatment and disposal, such as:

1.

Significant reduction of LLRW volumes.

2.

Reduction of disposal costs.

3.

Solidified waste forms with physical properties superior to those forms produced by present methods.

4 Flexibility to meet the criteria of a variety of ultimate waste disposal concepts.

S.

Potential for retrievability and further processing, if desired by future needs.

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