ML23104A051

DOE-OR-01-2560aD2 Remediation Strategy

ML23104A051
Person / Time
Site:	05000610
Issue date:	01/13/2023
From:	Oak Ridge
To:	Office of Nuclear Material Safety and Safeguards, US Dept of Energy, Office of Environmental Management
References
DE-SC-0004645 DOE/OR/01-2560&D2
Download: ML23104A051 (1)

Text

DOE/OR/01-2560&D2 Remediation Strategy Plan for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee This document is approved for public release by review by:

DOE/OR/01-2560&D2 Remediation Strategy Plan for the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee Date Issued: January 2013 Prepared for the U.S. Department of Energy Office of Environmental Management URS l CH2M Oak Ridge LLC Managing and Safely Delivering the Department of Energys Vision for the East Tennessee Technology Park Mission under contract DE-SC-0004645 ii

CONTENTS EXECUTIVE

SUMMARY

.......................................................................................................................... 1

1. INTRODUCTION AND BACKGROUND ........................................................................................... 2

2. SELECT APPROACHES FOR LONG TERM S&M............................................................................ 5

3. LONG TERM S&M ............................................................................................................................. 10

4. PLANNING FOR SALT DISPOSITION ............................................................................................ 11

5. THE SALT DISPOSITION PROJECT EXECUTION ........................................................................ 12

6. REFERENCES..................................................................................................................................... 13 FIGURES Figure 1: Projected Thallium-208 Gamma Fields at MSRE ......................................................................... 9 ATTACHMENTS Flowchart Illustrating the MSRE Remediation Strategy .A Remediation Strategy Schedule ................................................................................................................... B TRU Determination [Checklist]................................................................................................................... C iii

ACRONYMS ASME American Society of Mechanical Engineers CERCLA Comprehensive Environmental Response, Compensation, and Liability Act D&D Decontamination and Demolition DOE U.S. Department of Energy DOT Department of Transportation EMWMF Environmental Management Waste Management Facility FFA Federal Facilities Agreement FTIR Fourier Transform Infrared Spectrometer HQ Headquarters MSR Molten Salt Reactor MSRE Molten Salt Reactor Experiment NDA Nondestructive Assay ORNL Oak Ridge National Laboratory TRU Transuranics (isotopes numbered higher than U with a half-life over 20 yrs.)

UCOR URS l CH2M Oak Ridge LLC WAC Waste Acceptance Criteria WIPP Waste Isolation Pilot Plant iv

EXECUTIVE

SUMMARY

Although not yet formally accepted, the defueled salt from the Molten Salt Reactor Experiment (MSRE) appears to be eligible for permanent isolation at the Waste Isolation Pilot Plant (WIPP). The MSRE remediation strategy described in this plan is based strictly on the assumption that the salt will be accepted by WIPP. The remediation strategy involves making and implementing a series of decisions that will result in safe long term storage of the defueled salt and surveillance and maintenance (S&M) of the MSRE facilities. Long term S&M will continue until twelve years prior to the scheduled completion of salt disposition at WIPP, at which time salt disposition actions will commence. The initiation of salt disposition will be determined from the finalized WIPP closure date and will consider U.S.

Environmental Protection Agency and the Tennessee Department of Environment and Conservation input regarding risk-based funding of cleanup priorities on the Oak Ridge Reservation. In parallel to making and implementing the S&M decisions, the TRU and Defense Waste Determinations will be obtained that are essential to gaining WIPP acceptance of the defueled salt.

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1. INTRODUCTION AND BACKGROUND The Molten Salt Reactor Experiment (MSRE) site is located at the Oak Ridge National Laboratory (ORNL), a US Department of Energy (DOE) facility in Oak Ridge, Tennessee. The MSRE facility included a graphite-moderated, liquid-fueled molten salt reactor that operated from June 1965 through December 1969. The reactor employed a molten fuel salt, an integral component of which was UF4 whose uranium content acted as the reactor fuel.

In 1994, significant migration of uranium from the stored fuel and flush salts was discovered. High concentrations of fluorine (F2) and uranium hexafluoride (UF6) gases were present in the off-gas system piping, and a significant deposit of uranium was found in the auxiliary charcoal bed. The migration of the uranium posed significant safety concerns, and therefore the following three response actions were taken pursuant to the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA):

A time-critical removal action (DOE 1995) removed reactive gases from the drain tanks and connected piping systems and equipment using vacuum. This removal action, found in Removal Action Report on the Molten Salt Reactor Experiment Time-Critical Removal Action at Oak Ridge National Laboratory (DOE/OR/01-1623&D2) was completed in February 1999 other than the continued removal of newly generated gases.

The uranium fluoride deposit and carbon fluoride compounds were removed from the auxiliary charcoal bed as documented in Action Memorandum for Uranium Deposit Removal at the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, (DOE/OR/02-1488&D2) and Removal Action Work Plan for Uranium Deposit Removal at the Molten Salt Reactor Experiment Facility at Oak Ridge National Laboratory (DOE/OR/01-1735&D2). This removal action was completed in January 2002 as documented by Removal Action Report for Uranium Deposit Removal at the Molten Salt Reactor Experiment at Oak Ridge National Laboratory, (DOE/OR/01-1918&D2).

A third remedial action was taken that is described in the following three documents:

- Record of Decision for Interim Action to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory; (DOE/OR/02-1671&D2)

- Remedial Design Work Plan to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory (DOE/OR/01-1722&D2)

- Remedial Design Report and Remedial Action Work Plan for the Removal of Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory (DOE/OR/0l-1810&D2) 2

Within this third remedial action, the following four tasks were completed:

(1) the salts were melted and chemically treated, (2) the molten salts were fluorinated to remove uranium, (3) the uranium was condensed into cold traps and transferred to chemical (NaF) traps, and (4) NaF traps loaded with the uranium were transferred to ORNL Building 3019A for storage.

Explanation of Significant Differences for the Record of Decision for Interim Action to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory (DOE/OR/01-2088&D2) deleted the requirement to convert the separated 233U to an oxide form, substituting storage of the removed uranium fuel in Building 3019A as the remedial action. Thus uranium defueling as a remedial action was completed when the uranium was delivered to Building 3019A in NaF traps, as described in the Phased Construction Completion Report for the Removal and Transfer of Uranium from the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory (DOE/OR/01-2256&D1).

After uranium was removed from the salt by fluorination in 1997, 1998, 2006, and 2008, final defueling was declared to be sufficient as of May, 2008. The defueled salt resides today in three tanks fabricated from Hastelloy'-N alloy. Hastelloy'1 is a registered trademark name of Haynes International, Inc.

Once the salt had been defueled, actions to transfer the defueled salts to shielded canisters for storage at ORNL Solid Waste Storage Area 5 were attempted but then delayed due to an inability to transfer the salt using the original transfer process design and equipment. The remaining uranium concentration in the defueled salt was re-evaluated in 2011 by gamma spectroscopy as described in Nondestructive Assay Measurements of Defueled Salts at the Molten Salt Reactor Experiment (DOE/OR/01-2513&D1) which further verified that the salt is defueled and suggests that the salt is fully eligible for permanent isolation at the Waste Isolation Pilot Plant in Carlsbad, NM (WIPP). Engineering Evaluation of Options for Molten Salt Reactor Experiment Defueled Coolant Salts (DOE/OR/01-2496&D1) estimated the cost of a Salt Disposition Project for transferring the salt to WIPP at over 100 million 2010 dollars. Several variant alternatives were planned in DOE/OR/01-2496&D1 to the extent of detail required for estimating their costs and identifying their positive and negative aspects. The Engineering Evaluation concluded that the most favored option is Thermal Transfer of the salt.

Current plans are to store the salt in place until the mid-to-late 2020s for the following reasons:

a receiver site has not yet been formally approved, the salt has now been defueled so that it will no longer support significant UF6 generation, the salt is a dry solid stored in a safe and easily monitored configuration, the salt tanks are stored in an extremely robust underground structure, and the substantial cost of the Salt Disposition Project would require deferral of other critical remediation work that poses a greater and more immediate risk to workers and to the environment.

1 Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof or its contractors or subcontractors.

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Although not yet formally accepted, the salt appears to be eligible for permanent isolation at WIPP. For disposability at WIPP, the transuranic (TRU) concentration must exceed 100 nCi/g. Preliminary calculations indicate that the MSRE defueled carrier salt in two of the tanks contains a minimum of 13,517 nCi/g TRU and that the flush salt stored in the third tank contains a minimum of 274 nCi/g TRUconcentrations that clearly meet the acceptance criteria regarding their TRU component. The TRU isotopes in the salt include 241Am, 237Np, 239Pu, 240Pu, and 242Pu.

Waste disposed of at WIPP must also originate from defense-related sources. In 1969 239PuF3 not produced in the MSRE but manufactured from defense plutonium was added to the carrier salt as part of an experiment that demonstrated the effect of 239Pu on the MSR operations. MSRE was operated to demonstrate that this type of reactor could breed weapon-usable plutonium for defense programs. Because of the comingling of defense plutonium with the fuel, the defense-related purpose of the successful plutonium experiment, and because the defueled salt does not meet the definition of spent nuclear fuel or High Level Waste according to the Nuclear Waste Policy Act, WIPP appears to be the most eligible destination for the long term isolation of the defueled salt.

This report assumes the salt will be accepted at WIPP and that the closure date for WIPP will be 2033.

MSRE salt removal is currently projected to start in the mid-2020s to support disposition of the salt before WIPP closure. The activities described in this remediation strategy plan are intended to improve performance of long term salt storage, which includes reducing the radiation fields and removing contaminated equipment that will not be needed at MSRE. This plan does not address routine infrastructure upgrades or major maintenance items for the MSRE Buildings that are needed to maintain habitability. Routine items will be addressed during annual budgeting and on an as needed basis. For example, 2012 roof refurbishment is not part of this plan.

Final disposition of the defueled salts will be addressed in a future Remedial Action Report. Demolition of the facility will be performed under a response action separate from salt disposition.

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2. SELECT APPROACHES FOR LONG TERM S&M This brief remediation strategy plan describes what safety management related and salt management related actions need to be evaluated and resolved, without deciding how to accomplish those actions.

These tasks and decision points are shown as a flowchart in Attachment A, and as a schedule in Attachment B. The decisions, components of work, and how the decisions and components interface are explained in the text below. Please refer to Attachments A and B when reviewing the following narrative.

Some items may be completed as they are being evaluated, while the more difficult ones will be deferred until a safe process has been defined and waste characterization and transportation issues have been resolved.

Determine the headspace gas treatment method. Salt in the tanks generate helium, fluorine, and three isotopes of radon. In theory, very low concentrations of molybdenum hexafluoride, uranium hexafluoride, and hydrogen fluoride could be generated as well, but none of these fluoride compounds have been detected since defueling was completed. The primary post defueling gases of concern that may cause corrosion are fluorine and hydrogen fluoride. Both of these gases are destroyed upon contact by any one of several compounds that could be employed either inside or outside of the salt tanks in a variety of process schemes. The gases of concern are currently being trapped in very small vessels containing sodium fluoride and activated alumina as was done during defueling. Headspace gases are pumped through these traps twice each year.

Since the current process was designed specifically with defueling in mind and defueling is now completed, other process options will be considered that may have significant advantages for long term deployment. Tests or field demonstration may be performed to verify effectiveness prior to installation of a new process.

Determine the need for FTIR or an alternate monitoring system. The Fourier Transform Infrared Spectrometer (FTIR) functioned well for its intended purpose during defueling, when covalent bonded gases such as uranium hexafluoride, molybdenum hexafluoride, hydrogen fluoride or chlorine trifluoride were normally present. FTIR systems are incapable of detecting several of the gases that could potentially be present in the MSRE salt tank headspace at various times, such as diatomic fluorine, oxygen, nitrogen, helium, radon, or argon because FTIR systems rely on covalent bonds as their target of detection. The information that can be gained by monitoring headspace gases, such as indication of corrosion or leaks, will be evaluated with alternate methods of monitoring the gas, e.g. in situ monitoring technologies or periodic sample collection for analysis. Alternative systems that can provide the same or similar information are operated by UT-Battelle, Materials and Chemistry Laboratory, Inc. and B&W Y-12 on the Oak Ridge Reservation, and samples could also be sent out to commercial laboratories off the reservation for analysis.

Detection of significant quantities of molybdenum hexafluoride would be a strong indicator of corrosion, but due to the physicochemical potentials, the absence of detection cannot be conclusively interpreted as absence of corrosion. Detection of large or greatly increased concentrations of hydrogen fluoride upstream of the traps would imply an inleakage of air into the system. The MSRE FTIR can measure covalent bonded gases such as HF upstream and downstream of the sodium fluoride and activated alumina traps, thus confirming trap effectiveness. Gas monitoring and gas trapping options will be considered together in selecting the best overall long term headspace gas management system.

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Evaluate the need for tank integrity measurements. The extent of tank corrosion has been evaluated using molybdenum hexafluoride levels taken from the FTIR data during defueling and projecting thinning of the tank walls with time through operations log records, and by laboratory analyses of salt and probe plugs for their nickel sulfide (or Ni metal) content. The results of evaluations by these means were inconclusive, although there is good evidence that corrosion has been very low.

The salt was defueled of 235U fuel during August 23-29, 1968 (Haubenreich 1969). At the highest process temperatures and fluorination rates encountered, the corrosion rate was estimated to be about 0.1 mil per fluorination processing hour (Lindauer 1969). With a total 235U defueling time of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> (ibid.), tank thinning would therefore be only 4 mil from the 1969 defueling. A similar total thinning could be expected from the 2006 and 2008 defueling, since process conditions were very similar. As a different approach, if the molybdenum hexafluoride trapping records based on FTIR are used from the 2008 defueling of Fuel Drain Tank No. 1, then 4.15 mil of uniform thinning would result in 2008, which is the very same conclusion. The same calculation for Fuel Drain Tank No. 2 defueling in 2006 produces an even lower result.

The tank walls and thermal wells were measured by ultrasonic thickness equipment in February 2000 (BJC 2000), which established their thickness at that time. These measurements confirmed the very limited effects of the first defueling in 1969 plus years of warming the tanks for salt annealing, plus years of cold fluorination from radiolytic decay. In 2000, corrosion thinning was not detectable on the outer walls and heads, but unfortunately, the amount of thinning had to be based on comparison of the measured thicknesses to specification tolerances, since no baseline measurement was available from the time of purchase of the tanks. What is known is that the thickness of the thermal wells was within 10% of their specification thickness when measured in 2000 and the walls were within their specification tolerance for new tanks. Thus the measured thicknesses compared to the original specifications indicate negligible thinning, and this is in close agreement with Lindauers assessment--which again concludes that defueling results in low corrosion. Lindauers assessment was based on corrosion product levels measured in the salt itself during the 1969 defueling. These historic corrosion evaluations provide some confidence of current tank integrity because while the latter defuelings are not identical to the original in time, temperature, and chemistry, they are very similar.

No thickness measurements have been made after the 2006 and 2008 233U defueling. The corrosion effects of different chemistry insofar as hydrofluorination and chlorine trifluoride feeding to relieve plugging in 1998, 2006, and 2008 are unknown, but anomalous chlorine-fluoride compounds remained visible on FTIR signals in 2012. While the precise effects of differences in chemistry and thermal cycle times between 1969 and latter defueling are not known, it appears likely based on the previous experience that the tank walls and heads remain very near their original specification thicknesses today.

No ultrasonic determinations have been made that verify current tank wall thickness.

Consideration will be given to the feasibility and need for more ultrasonic data, especially at the salt/headspace interface zone of the thermal wells, if there is any indication that corrosion rates have somehow increased, such as detection of more than trace MoF6 in the headspace gas.

Detection of significant MoF6 levels is not expected and should trigger investigation, since the source would have to be corrosion.

Gas leakage from tank thermal well thinning is considered to be the highest risk at MSRE. The point of failure from corrosion thinning is expected to be at the salt-headspace interface of a 6

thermal well. Leakage at this location would allow headspace gas to escape resulting in some redox damage to air filters downstream but there would be no significant release exposing workers and no impact whatsoever off site. The thermal wells are specified as INOR-8 Schedule 40 pipe by the drawing used for fabrication (M20794RF006D). INOR-8 is the original name of Hastelloy'-N alloy. Hastelloy'-N Schedule 40 pipe has a nominal outside diameter of 1.9 and is nominally 0.145 thick. The outside diameter specification tolerance is +0.008 -0.015 while the wall thickness tolerance is + 12.5% or + 0.018125 (Haynes 2012).

The allowable tolerances indicate the MSRE Hastelloy'-N Schedule 40 welded pipe thermal wells are required to be between 0.12688 and 0.163125 thick per specification which is much thinner than the tank walls and heads the walls are nominally 0.5 thick and the heads are nominally 0.75 thick (M20794RF008D). This means if there ever are corrosion leaks, and none are expected, then the thermal wells would be most likely to leak first, especially at the salt-headspace interface where attack would chemically be the greatest and pitting corrosion cannot be ruled out. If this leakage ever occurred it is very likely that the thermal wells could be plugged at the top with expansion plugs using long handled tools. Salt would not be dispersed from a tank as a result of leakage at the thermal well. The salt inside the tanks is a high density solid that behaves very much like a monolithic rock. Prior to heating or pressurizing the tanks for salt transfer, ultrasonic thickness measurements will be made if recent data is not already available at that time.

Determine long term waste management plan. Several items stored at MSRE are growing thallium-208 that emits a very penetrating 2.6 MeV gamma ray photon. The gamma levels on some of the high thallium items at MSRE will peak in 2018 and will not decay to 2012 levels again until the year 2061, so there is an opportunity to limit personnel exposure by removing items from high occupancy areas and either disposing of them or relocating them to areas not frequented by workers. See Figure 1 for an illustration of how the thallium-driven component of the gamma radiation field will vary over time.

Probes- there are five probes and four probe casks at MSRE. The two caskless probes are stored inside plastic pipes. One of the probes stored in a cask in the MSRE High Bay is growing significant thallium gamma levels that require routinely adding shielding to control the growing radiation field. Disposition options to be considered include offsite burial or onsite storage in a more remote location at ORNL. Characterization currently underway should better define all possible solutions. Preliminary calculations indicate that the 208Tl-contaminated probe may contain a Department of Transportation Type B quantity for which no packaging is currently likely to be authorized for shipment in commerce. Waste disposal site acceptance options and onsite storage options will be evaluated when additional characterization data is available.

Piping - there are thallium sources growing in piping, especially at the Cajon VCR type fittings, on the walls shared with offices and with other high occupancy areas near the MSRE High Bay. The radiation levels are low but growing in the high-occupancy areas. These pipes are not needed to execute the Thermal Transfer option when the Salt Disposition Project is performed in the future. Removal of these pipes and disposal if practical, or relocation of the thallium sources to a more remote area as a backup option will be evaluated. Pipes not near high occupancy areas can remain until the facility is decommissioned after salt transfer and charcoal removal.

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Charcoal Canister - The possibility of relocating the charcoal canister to Building 3019, to the basement area beneath the large shield plugs in the MSRE High Bay of Building 7503, or to other locations at ORNL will be evaluated. By relocating the charcoal canister, savings could result through downgrading the facility categorization, reducing the level of nuclear safety rigor, improving security, and reducing facility maintenance cost.

Other Waste - There are containers of legacy waste outside the buildings at MSRE and legacy waste stored in the MSRE basement areas, and there is piping and equipment that are not needed for salt transfer that contribute to the elevated and growing background radiation in the High Bay. Some waste may require additional characterization to make permanent disposal viable. In January 2013, MSRE waste was added to the ORNL S&M Waste Profile at the Environmental Management Waste Management Facility (EMWMF), facilitating the permanent disposal of additional legacy and defueling-related wastes stored at MSRE. Disposition or relocation will be evaluated for the higher gamma source piping and equipment items near high occupancy areas as well as for the stored legacy waste items.

Complete TRU and Defense Waste Determinations. The WIPP site is authorized to receive only TRU Defense Waste. A formal Defense Waste Determination and TRU waste determination have not been completed for the MSRE defueled salt. Because the salt cannot be sent to WIPP without being formally classified as TRU Defense Waste, it will be necessary to complete these determinations prior to approving the disposition project funding in FY21.

The unique homogeneous reactor design at MSRE makes application of the laws and regulations that were intended to be applied to conventional reactors challenging. For example, the extent of uranium removal required in order to declare the MSRE salt defueled is not defined by standard or statute. For a difficult waste determination case like MSRE salt, where various interpretations could appear plausible, a checklist has been formulated to support key DOE participants reaching a consensus through a formal deliberative process. In order for the salt to be disposed of at WIPP the checklist should indicate that the salt is classified as "Transuranic Waste" and is neither "High-level Waste" nor "Spent Nuclear Fuel" pursuant to their statutory definitions. The key assertions required for WIPP acceptance are listed on the first page of the checklist provided as Attachment C to this strategy plan, and the approval signatures that attest to those assertions are found on the second page of the checklist. Before signing, any approver may request additional information on which to base a decision. The approvals made in Oak Ridge are segregated into the left column of the checklist and are to be followed by subsequent approvals of counterpart authorities at the DOE Carlsbad Field Office, shown in the right hand column, with final approvals by the DOE Headquarters Office of Logistics and Waste Disposition Enhancements and by the Assistant General Counsel for Environment at DOE Headquarters.

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Considerable time may be required to complete the formal TRU Waste determination so this item will be addressed in conjunction with early planning of long term S&M, as follows:

- A TRU Waste Determination report (LA-UR-10-07278) will be finalized by Los Alamos National Laboratory during FY13; this report will provide the technical basis supporting completion of the TRU waste determination checklist (Attachment C).

- Concurrence by the DOE Carlsbad Field Office will be obtained during FY14 documenting that the TRU Waste Determination report is sufficient regarding defense waste determination and that a formal Defense Determination under authority of the Secretary of Energy will not be required for the defueled salt.

- If additional characterization measurements will be required in order for the DOE Oak Ridge Office Environmental Management to complete and execute a TRU Waste Determination, then these measurements will be completed during FY13 through FY15.

- The TRU Waste Determination checklist (Attachment C) will be completed by the DOE Environmental Management Oak Ridge Office and submitted to CBFO by the end of FY14, with a request for counterpart approvals to be completed by the end of FY19.

- During FY15 - FY19, additional characterization measurements and information will be obtained if needed to complete their review.

- DOE Headquarters, including General Counsel if necessary, will approve the TRU Waste Determination by the end of FY19 to facilitate preliminary planning in FY20 and budgeting the salt disposition project in FY21.

Figure 1. Projected Thallium-208 gamma fields at MSRE.

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3. LONG TERM S&M Long Term S&M will be based on implementation of the approaches selected for the facility changes and waste management actions described above. Referring to Attachment A, Implement Selected Approaches will improve the configuration at MSRE for a more long range view in recognition of MSREs new safe salt storage function. Long term S&M will continue until twelve years prior to the scheduled completion of salt disposition, which will be determined from HQ input regarding WIPP closure date and TDEC and EPA input regarding funding of cleanup priorities.

DOE will obtain EPA and TDEC input regarding priorities when funding ORR cleanup.

The salt disposition has been forecast to be very expensive, more than 100 million 2010 dollars, and there is not a formally approved receiver site for the salt. The salt is stored in a safe well-monitored storage location where it poses little threat of release. Realizing this, DOE plans to evaluate which cleanup work will create the most reduction in risk to the public for the fixed level of available funds, in risk-based prioritization that involves EPA and TDEC input.

DOE ORO will solicit DOE HQ input as to actual WIPP closure date. A firm date for WIPP closure will be needed to effectively schedule the Salt Disposition Project. There needs to be adequate time between project initiation and WIPP closure because the salt transfer will be a complex undertaking. A five year schedule contingency has been built into this remediation strategy plan to provide a safety net beyond the minimum time determined by the Engineering Evaluation Thermal Transfer schedule.

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4. PLANNING FOR SALT DISPOSITION Because the salt disposition actions will take years to complete, some actions need to take place twelve years before WIPP closure. It is assumed that the Defense Waste Determination will have been completed when this twelve year period begins.

Preliminary Actions Prior to Salt Disposition. Once project completion has been scheduled for twelve years in the future, the long lead items will begin first to ensure meeting the Salt Disposition Project schedule. These early start items include making calculations based on the applicable decay and ingrowth time and evaluating DOT Type B Container planning to meet DOT and WIPP acceptability and ensure transport will be accomplished on schedule. Fissile gram equivalents, inner container design, and content authorization will be ensured and shielding calculations will be performed, resulting in a compliant design and successful authorization of the current content. While the thermal option is currently the preferred choice, other options, including new technologies that could be developed, will be considered.

Salt Disposition Project funding will need to be approved about twelve years prior to the scheduled completion of the project which is assumed to be at WIPP closure. The project can be initiated eleven years prior to completion per the schedule and tasks described for the Thermal Transfer Option in DOE/OR/01-2496&D1 can be accomplished as described. A different preferred option may be selected based on changing conditions or newly developed technologies.

When program requirements significantly change, this remedial action strategy and the Thermal Transfer schedule will be re-evaluated to preserve credibility.

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5. THE SALT DISPOSITION PROJECT EXECUTION The Salt Disposition Project tasks listed below form the basis of planning and cost estimating the Thermal Transfer Option, as described in DOE/OR/01-2496&D1. The initial four tasks below will begin eleven years prior to WIPP Closure assuming the Defense Waste Determination has been completed and WIPP acceptance has been obtained Salt Disposition Project Execution - 5 Year Schedule Contingency Re-evaluate Salt Transfer Options Staffing Project Complete Additional Design Dose Calculations and Controls Design Remedial Design and Action Work Plan Procedure Preparation and Approval Equipment & Material Procurement Safety Management Program Document and Calculations Safety Basis Document Preparation Emergency Planning Update DOE Approval of Safety Basis Safety Basis Implementation and IVR Facility Cleanup and Upgrade for Field Work Construct Hot Cell Instrument Calibration, Equipment Installation and Repairs Build Mock-Up Training Readiness Assessment Perform Thermal Transfer of Salt Salt Removal Formal Notification FFA Milestone Ship Type B Containers to WIPP Demolish Hot Cell Remedial Action Report Project Closeout After shipping the salt to WIPP, detailed planning of the final D&D of the MSRE infrastructure can proceed.

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6. REFERENCES BJC/OR-445. Final Report of the Molten Salt Reactor Experiment Drain Tank Qualification, Oak Ridge National Laboratory, February, 2000, Bechtel Jacobs Company LLC, Oak Ridge, TN.

DOE 1995. Oak Ridge National Laboratory Molten Salt Reactor Experiment Time Critical Removal Action Memorandum Report, Lockheed Martin Energy Systems, Inc., Oak Ridge, Tennessee.

DOE/OR/02-1488&D2. Action Memorandum for Uranium Deposit Removal at the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1996, US Department of Energy Office of Environmental Management, Oak Ridge, Tennessee.

DOE/OR/01-1623&D2. Removal Action Report on the Molten Salt Reactor Experiment Time-Critical Removal Action at Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1998, Lockheed Martin Energy Resources, Oak Ridge, Tennessee.

DOE/OR/01-1735&D2, Removal Action Work Plan for Uranium Deposit Removal at the Molten Salt Reactor Experiment Facility at Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1998, Lockheed Martin Energy Resources, Oak Ridge, Tennessee.

DOE/OR/02-1671&D2. Record of Decision for Interim Action to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee. 1998, Jacobs Engineering, Oak Ridge, Tennessee.

DOE/OR/01-1722&D2. Remedial Design Work Plan to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1999, Lockheed Martin Energy Resources, Oak Ridge, Tennessee.

DOE/OR/0l-1810&D2. Remedial Design Report and Remedial Action Work Plan for the Removal of Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, 1999, Lockheed Martin Energy Resources, Oak Ridge, Tennessee.

DOE/OR/01-1918&D2. Removal Action Report for Uranium Deposit Removal at the Molten Salt Reactor Experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee. 2001, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee.

DOE/OR/01-2088&D2. Explanation of Significant Differences for the Record of Decision for Interim Action to Remove fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, 2006, , US Department of Energy Office of Environmental Management, Oak Ridge, Tennessee.

DOE/OR/01-2256&D1. Phased Construction Completion Report for the Removal and Transfer of Uranium from the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, 2008, US Department of Energy Office of Environmental Management, Oak Ridge, Tennessee.

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DOE/OR/01-2513&D1. Nondestructive Assay Measurements of Defueled Salts at the Molten Salt Reactor Experiment, Oak Ridge Tennessee, August, 2011, URS l CH2M Oak Ridge LLC,.

DOE/OR/01-2496&D1. Engineering Evaluation of Options for Molten Salt Reactor Experiment Defueled Coolant Salts, Oak Ridge, Tennessee, December, 2012, Bechtel Jacobs Company LLC, Oak Ridge, Tennessee.

Haubenreich, P.N. and J.R. Engel 1970. Experience with the Molten Salt Reactor Experiment, Nuclear Applications and Technology, Vol. 8, Feb., URL:

http://moltensalt.org/references/static/downloads/pdf/NAT_MSBRrecycle.pdf Haynes 2012. Hastelloy' and Haynes Pipe and Tubular Products, 2002, Haynes International, Kokomo, IN, URL http://www.haynesintl.com/ accessed 2/28/2012 at 10:23 EST.

Lindauer, R.B. 1969. Processing of the Fuel and Flush Salts, ORNL-TM-2578, August, Oak Ridge National Laboratory, Oak Ridge, TN.

M20794RF006D. Fuel Drain Tank Upper Head Assembly & Details, Renumbered M20794RF006D R13 from Original No. D-FF-A-40458, June 1961, Reactor Division, Union Carbide Nuclear Company, Oak Ridge National Laboratory, Oak Ridge, TN.

M20794RF008D. Fuel Drain Tank Shell and Lower Head Assembly & Details, Renumbered M20794RF008D R13 from Original No. D-FF-A-40460, June 1961, Reactor Division, Union Carbide Nuclear Company, Oak Ridge National Laboratory, Oak Ridge, TN.

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Flowchart Illustrating the MSRE Remediation Strategy Attachment A

Remediation Strategy Schedule Attachment B

Flowchart Illustrating the MSRE Remediation Strategy Remediation Strategy Schedule Attachment B

Attachment C Attachment C DOE/OR/01-2560&D2 RECORD COPY DISTRIBUTION FileDMCRC