ML23104A049

DOE-OR-01-2496aD1 Salt Options

ML23104A049
Person / Time
Site:	Abilene Christian University
Issue date:	12/31/2010
From:	Bechtel Jacobs Co
To:	Office of Nuclear Material Safety and Safeguards, US Dept of Energy, Office of Environmental Management
References
DE-AC05-98OR22700 DOE/OR/01-2496&D1
Download: ML23104A049 (1)

Text

DOE/OR/01-2496&D1 Engineering Evaluation of Options for Molten Salt Reactor Experiment Defueled Coolant Salts, Oak Ridge, Tennessee INTERNAL USE ONLY Caution: This document was prepared for internal use only (e.g., BJC, BJC subcontractors, DOE, and DOE contractors).

Therefore, the document must be reviewed by the Classification and Information Control Office and approved for public release prior to reproduction for external distribution.

DOE/OR/01-2496&D1 Engineering Evaluation of Options for Molten Salt Reactor Experiment Defueled Coolant Salts, Oak Ridge, Tennessee Date IssuedDecember 2010 Prepared for the U.S. Department of Energy Office of Environmental Management BECHTEL JACOBS COMPANY LLC managing the Environmental Management Activities at the East Tennessee Technology Park Y-12 National Security Complex Oak Ridge National Laboratory under contract DE-AC05-98OR22700 for the U.S. DEPARTMENT OF ENERGY

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iii CONTENTS TABLES....................................................................................................................................................... v FIGURES...................................................................................................................................................... v ACRONYMS..............................................................................................................................................vii EXECUTIVE

SUMMARY

.........................................................................................................................ix

1. INTRODUCTION.................................................................................................................................. 1

2. HISTORY OF ACTIONS/INTENTIONS/OUTCOMES...................................................................... 6

3. PURPOSE OF THIS REPORT............................................................................................................ 10 3.1 DISPOSAL AT WIPP................................................................................................................. 11 3.2 STORAGE AT ORNL................................................................................................................ 12 3.3 INITIAL ALTERNATIVES CONSIDERED............................................................................. 12

4. NATURE/CHARACTERISTICS OF MSRE SALT AND TANKS.................................................... 14 4.1 FUEL SALT, DRAIN TANKS, AND THE DRAIN TANK CELL........................................... 14 4.2 RADIONUCLIDE INVENTORY.............................................................................................. 16 4.3 SAMPLING AND ANALYSIS.................................................................................................. 17 4.4 MSRE SALT CONTINUING REACTION / RADIOLYSIS / OFF GAS.................................. 18 4.5 MSRE DRAIN AND FLUSH TANK CORROSION................................................................. 19

5. ALTERNATIVES FOR DISPOSITION OF SALT IN DRAIN TANKS AND FLUSH TANK........ 20 5.1. ALTERNATIVES FOR LEAVING TANKS IN PLACE.......................................................... 21 5.1.1 Maintain for 50 Years at MSRE Facility.......................................................................... 21 5.1.1.0 Assumptions..........................................................................................................21 5.1.1.1 Technical Feasibility.............................................................................................21 5.1.1.2 Discussion.............................................................................................................22 5.1.1.3 Cost and Schedule.................................................................................................22 5.1.2 Entombment at MSRE Facility......................................................................................... 25 5.1.2.1 Hydrogeologic Setting...........................................................................................26 5.1.2.2 Technical Feasibility.............................................................................................26 5.1.2.3 Cost and Schedule.................................................................................................29 5.1.2.4 Discussion.............................................................................................................29 5.2 ALTERNATIVES FOR SALT REMOVAL FROM MSRE...................................................... 30 5.2.1 Geometries and Volumes of Drain and Flush Tanks and Defueled Coolant Salt............. 30 5.2.2 Intact Tank Removal......................................................................................................... 30 5.2.2.1 Technical Feasibility.............................................................................................33 5.2.2.2 Assumptions..........................................................................................................35 5.2.2.3 Cost and Schedule.................................................................................................35 5.2.2.4 Discussion.............................................................................................................35 5.2.3 Thermal Removal............................................................................................................. 36 5.2.3.1 Filling and Handling of Disposal Containers........................................................42 5.2.3.2 Assumptions..........................................................................................................43 5.2.3.3 Technical Feasibility.............................................................................................47 5.2.3.4 Cost and Schedule.................................................................................................47

iv 5.2.3.5 Discussion.............................................................................................................47 5.2.4 Mechanical Removal........................................................................................................ 48 5.2.4.1 Hammer and Vacuum to Loosen and Transfer Salt..............................................48 5.2.4.2 Cutting Tanks with a Diamond Wire Saw.............................................................48 5.2.4.3 Assumptions..........................................................................................................53 5.2.4.4 Technical Feasibility.............................................................................................53 5.2.4.5 Cost and Schedule.................................................................................................53 5.2.4.6 Discussion.............................................................................................................53 5.2.5 On-site Storage of Salt in an Approved Type B Container............................................... 54 5.2.5.1 Technical Feasibility.............................................................................................55 5.2.5.2 Cost and Schedule.................................................................................................55 5.2.5.3 Discussion.............................................................................................................55

6. SHIPPING TO WIPP........................................................................................................................... 57 6.1 WASTE ACCEPTANCE CRITERIA......................................................................................... 61 6.1.1 Container Properties......................................................................................................... 61 6.1.2 Radiological Properties..................................................................................................... 61 6.1.3 Physical Properties............................................................................................................ 62 6.1.4 Chemical Properties.......................................................................................................... 63 6.1.5 Data Package Contents..................................................................................................... 63 6.2 RH TRU WASTE RADIOLOGICAL CHARACTERIZATION............................................... 63 6.3 CH CONTAINER GEOMETRIES............................................................................................. 65 6.4 TRANSPORTATION DOSE RATE.......................................................................................... 65 6.5 WIPP HANDLING AND DOSE RATE..................................................................................... 65 6.6 PAYLOAD WEIGHT................................................................................................................. 65 6.7 RH-TRU PAYLOAD CANISTER CONTENTS........................................................................ 67

7.

SUMMARY

OF EVALUATION........................................................................................................ 69

8. REFERENCES..................................................................................................................................... 72 APPENDIX A. BASIS OF COST ESTIMATES.....................................................................................A-1 APPENDIX B. COST ESTIMATE DETAILS.........................................................................................B-1 APPENDIX C. SCHEDULES..................................................................................................................C-1 APPENDIX D. ATTACHMENTS...........................................................................................................D-1

v TABLES Table i. Evaluation of Alternatives.............................................................................................................xii Table 1. MSRE Salt Related Studies and Removal Actions.........................................................................7 Table 2. Alternatives Estimated for Cost....................................................................................................12 Table 3. Characteristics of MSRE Salt Tanks and Contents.......................................................................16 Table 4. MSRE Salts: Radiological Constituents.......................................................................................17 Table 5. Compositions of Hastelloy N (%).................................................................................................19 Table 6. Comparison of MSRE Salt Radiological Properties and WIPP WAC Limits..............................62 Table 7. RH TRU 72B Package Component Weights................................................................................66 Table 8. Payload Canister Contents............................................................................................................67 Table 9. Evaluation of Alternatives............................................................................................................72 FIGURES Figure 1. MSRE Facility Showing the Reactor and Drain Tanks.................................................................2 Figure 2. Drain Tank Cell, Showing Fuel Drain Tanks and Flush Tank......................................................3 Figure 3. MSRE Salt Tanks in the Drain Tank Cell...................................................................................10 Figure 4. Alternatives for Disposition of MSRE Defueled Coolant Salt....................................................11 Figure 5. Fuel Drain Tank...........................................................................................................................15 Figure 6. Sustainment Cost Trend...............................................................................................................24 Figure 7. Major Repair & Replace Trend...................................................................................................24 Figure 8. 50-Year Cumulative O&M Costs................................................................................................25 Figure 9. Entombment Overview and Detail..............................................................................................27 Figure 10. Ten Drum Over Pack (TDOP)...................................................................................................32 Figure 11. Licensed Shipping Containers for a TDOP...............................................................................34 Figure 12. Contact Handling of Ten Drum Over Packs at WIPP...............................................................34 Figure 13. Salt Melt and Transfer Process..................................................................................................39 Figure 14. Molten Salt Mixer with a Folding Propeller..............................................................................41 Figure 15. Double Wall Disposal Container...............................................................................................42 Figure 16. Filling and handling equipment.................................................................................................45 Figure 17. Mechanical Removal Details and Views...................................................................................51 Figure 18. Salt Tank Sliced into Nine Pieces and a Slice in a WIPP Payload Canister..............................52 Figure 19. Considerations for Off-Site Shipment and Disposal at WIPP...................................................59 Figure 20. WIPP Remote Handling Disposal Process................................................................................68

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vii ACRONYMS AK Acceptable Knowledge BJC Bechtel Jacobs Company LLC CBFO Carlsbad Field Office CERCLA Comprehensive Environmental Response, Compensation, and Liability Act CM Corrective Maintenance CH Contact-Handled DOE U.S. Department of Energy DOT U.S. Department of Transportation D&D Decommissioning and Demolition DNFSB Defense Nuclear Facilities Safety Board DQO Data Quality Objective DTC Drain Tank Cell EPA U.S. Environmental Protection Agency ESD Explanation of Significant Differences FGE Fissile Gram Equivalent FDT Fuel Salt Drain Tank FFT Fuel Flush Tank ft msl feet mean sea level FSDS Fuel Salt Disposal System GSF gross square feet MSRE Molten Salt Reactor Experiment O&M Operations and Maintenance ORIGEN Oak Ridge Isotope Generation ORNL Oak Ridge National Laboratory PM Preventive Maintenance RGRS Reactive Gas Removal System RH Remote-Handled ROD Record of Decision TDEC Tennessee Department of Environment and Conservation SBTM Sustainable Building Technical Manual S&M Surveillance and Maintenance SWSA Solid Waste Storage Area TDOP Ten Drum Over-Pack TRU transuranic UDR Uranium Deposit Removal WAC Waste Acceptance Criteria WIPP Waste Isolation Pilot Plant

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ix EXECUTIVE

SUMMARY

The Molten Salt Reactor Experiment (MSRE) facility is located at Oak Ridge National Laboratory (ORNL), a U.S. Department of Energy (DOE) facility in Oak Ridge, Tennessee. The MSRE facility was a graphite-moderated, liquid-fueled reactor operated from June 1965 through December 1969. Since reactor shutdown, several studies and removal actions have been performed to address the contaminated fuel and flush salts in the facility. In 1998, the Record of Decision for Interim Action to remove fuel and flush salts from the MSRE facility (DOE 1998) was approved by DOE, EPA, and TDEC. The remedy defined in the Record of Decision (ROD) was partially completed by 2008 through heating and melting the salt, adding fluorine, circulating off-gas from the fluorinated salt through NaF traps and collecting the uranium. The removal and disposition of the residual salt was deferred to allow further study and evaluation of alternate methods.

The alternatives considered in this study are assessments of potential approaches for the future disposition of the MSRE salt, but are not necessarily comparable options since they do not all achieve the same objective or have identical outcomes. The alternatives described and evaluated in this report include:

Maintain As-is for 50 years; Entombment of salt tanks in place; Removal of the intact tanks; Mechanical removal of salt and disposition at Waste Isolation Pilot Plant (WIPP);

Thermal removal of salt and disposition at WIPP; and, On-site storage of salt in an approved type B container using one of the previously listed removal methods.

While several options for salt disposition were determined to be viable and their costs have been estimated, the base case (50 Year As-Is) is the recommended as the overall preferred near term option, whether employed for the entire 50 years or not, because it provides a means of making funds available for competing ORR environmental projects whose actions, if not taken quickly, could result in a far greater risk to the environment than the salt. The salt presents low environmental risk because it is stored in closely monitored tanks in a massive reinforced concrete, stainless steel lined subterranean basement, where MSRE has complex programs in place and is staffed by highly trained and qualified personnel that together ensure safe operations, monitoring and responding to changes in parameters associated with the defueled salt, physicochemical treatment of salt tank headspace gases, maintenance of the infrastructure that sustains building habitability and protection of the key electrical systems from the weather, control of contamination, maintaining safe access for all the measurements and inspections needed for planning of decontamination and decommissioning, performing sampling, characterization, packaging, storage, and disposal of radioactive and hazardous waste and responding to emergencies. To the extent that these programs, personnel, equipment, and facilities are maintained, the permanent disposition of the salt can be deferred for years without increasing risk to the environment or creating far greater cleanup costs in the future. Therefore the base case of "Maintain as is" for fifty years is recommended as the best means of making funds available in near term years that could accomplish more critical cleanup work. Once the more critical environmental risks on the ORR have been reduced and the disposal of the salt has been scheduled the discussions of and comparisons between other options in this report will apply.

The alternatives are listed in Table i. along with their performance merits, issues and projected cost and schedule. The alternatives are grouped as two primary options: keep salt at MSRE or remove salt from MSRE. Of the keep salt at MSRE alternatives, entombment in place is preferred over no action as it provides additional shielding from the contaminated salt and provides a pathway for the reduction of the

x nuclear safety categorization (by analysis) for MSRE. The entombment design can be adapted to allow the removal of the entombed tanks at some time in the future for subsequent storage or disposition at an alternate location. Of the remove salt from MSRE alternatives, the thermal removal of salt and disposition at WIPP is preferred because the technical and regulatory issues are significantly less challenging than the other removal alternatives, thermal salt transfer has been successfully demonstrated in the past, albeit in a less technically challenging environment and configuration, and the use of new pumps and piping introduced from the top of the tank would not require pressurization of the tank.

Disposal of MSRE salt at WIPP will require a determination that the salts are defense related transuranic (TRU) waste to meet the licensing requirements for WIPP. All alternatives that involve the removal of salt from MSRE option will allow nuclear facility re-categorization (i.e., downgrading) of the MSRE facility. On-site storage of removed salt/tank at one of the Solid Waste Storage Areas (SWSA) in Melton Valley would require modification to the safety basis for the SWSA and possibly re-categorization.

If MSRE salt is confirmed to be defense-related waste and disposal at WIPP is permissible and a disposal alternative there is pursued, the thermal removal method, transfer to new disposal containers, and highway transport of those containers is the preferred alternative for getting the salt to WIPP. Thermal transfer of the salt is technically challenging, and some of the challenges and possible solutions to those challenges are discussed below. If, for any reason, the salt does not qualify for disposal at WIPP and must be stored indefinitely at ORNL but cannot remain in place at MSRE, then removal of the tanks intact, placement in unlicensed shield containers and storage of those unlicensed containers at ORNL is recommended.

While thermal removal of the salt is the preferred alternative for permanent disposal, previous studies and experience with the MSRE salt have generated the following set of concerns related to thermal removal and repackaging.

Recent fluorination activities may have accelerated corrosion of the hastelloy tanks, and tank integrity may not now be sufficient for pressurizing the tanks for salt removal.

A previous attempt to heat and transfer the salt to new containers was unsuccessful, possibly due to a clogged transfer line.

Melt probes used to initiate salt melt and introduce fluorine into the molten salt experienced clogging problems.

The salt and the radioactive materials included in the salt may be stratified and unevenly distributed in the existing tanks. During transfer and placement of salt in new containers, high radioactivity zones in the drain tanks could result in hot spots in the disposal containers that would not meet dose-rate limits.

Metal fluorides or other non-salt materials present in the tanks may not melt and transfer.

Concepts for tools, disposal containers and methods for thermal salt transfer have been developed that address these concerns.

Tank integrity testing is recommended prior to thermal transfer, but instead of pressurizing tanks to remove the salt, molten salt pumps capable of pumping material at 600 degrees C would be used and the tanks would not need to be pressurized to transfer the salt.

Instead of relying on systems and equipment installed in the 1960s, new heat-traced pipelines would be used to transfer salt.

Melt probes would be re-designed based on lessons learned for previous activities.

xi Table i. Evaluation of Alternatives Option Alternative Performance Merits Issues Project Duration /

Project Cost In Millions (Ref#)

End State After Entire Project Duration Comparative 50-year Total Cost (Maintenance plus Project) in Millions First Year Annual Cost (Maintenance plus Project) in Millions Maintain As-is for 50 years Technically feasible and easy to implement Change in facility maintenance strategy and increase in long term costs MSRE remains a nuclear facility Not a project, maintenance for 50-yr duration (1)

$497 Cost (1) 100 year-old facility has been maintained safe, operable, and habitable

$497

$9.90 Keep salt at MSRE Retrievable Entombment Technically feasible Reduces hydrogeologic and shielding concerns Off-gassing and corrosive effects of HF in an aqueous environment unless getter is added to the salt to address fluorine and tanks are sealed in epoxy grout and lifted above the water table.

62 Month Schedule (2)

$77 Cost (2)

Salt is easily retrievable but must be transported to a hot cell, transferred to disposal containers, and shipped to a final disposal site.

$442

$24.8 Intact-Tank Removal No salt transfer required Reduces tank corrosion concerns Very difficult to package tank and contents for shipment/disposal; therefore, a type-B like container would be used to store the tank and contents at ORNL.

Waste package with shielding would be significantly overweight for off-site shipment/disposal.

63 Month Schedule (3e)

$79 Cost (3e)

Salt is easily retrievable but must be transported to a hot cell, transferred to disposal containers, and shipped to a final disposal site.

$443

$25.0 Thermal Salt Removal Salt stratification and impurities must be addressed Remaining tank and associated residual contamination will remain 71 Month Schedule (4)

$111 Cost (4)

Salt is permanently disposed of at WIPP

$478

$28.6 Mechanical Salt Removal (Needle Scaler)

Meets the removal requirements in ROD Previously demonstrated as feasible Once packaged, shipment and disposal at WIPP is reasonably implementable Removal and packaging is difficult Remaining tank and associated residual contamination will remain 69 Month Schedule (5a)

$101 Cost (5a)

Salt is permanently disposed of at WIPP

$468

$27.5 Remov e salt from MSRE On Site Storage in Type B Container Reduces the nuclear categorization of MSRE Once packaged, shipment and disposal at WIPP is reasonably implementable Requires modification of safety basis at storage location.

62 Month Schedule (6b)

$95 Cost (6b)

Salt is easily retrievable and ready for transport to disposal without transfer

$460

$28.3 Endnotes:

The baseline annual maintenance cost ($9.9 Million) is included in the first year annual cost of all options. The first year annual cost column on the far right is provided for convenience in evaluating competing priorities and selecting projects for funding that achieve the best reduction of environmental risk on the Oak Ridge Reservation. Deferral of all options other than the baseline maintenance will provide roughly 14 - 18 million dollars available for executing projects that are likely to result in greater environmental risk reduction than that achievable hrough permanent disposal or alternative management of the MSRE salt. The comparative 50 year cost is for operating MSRE through and beyond the salt option elected (on to 50 years) and is provided for easy comparison onlyit does not include costs incurred at the facility receiving the salt for applicable cases.

xii The following strategies are included to address the possibility of stratified materials and radioactive hot spots:

o An agitator would be used to mix the molten salt prior to pump transfer; o A new disposal container concept is proposed that consists of a container within a container.

The inner tube would be used for the high radioactivity Drain Tank salt and the outer tube would be filled with lower radioactivity Flush Tank salt and getter (sodium iodide, to capture free fluorine).

o Twelve tanks would be used for the Drain and Flush Tank salt to distribute the radioactive materials and limit the dose rate per container; and, o A sequential container filling method would distribute the radioactive materials evenly between tanks to further limit the potential for hot spots.

The agitator would be more robust than the sparging methods previously used and should result in better mixing of metal fluorides and other non-salt materials into the molten salt.

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1

1. INTRODUCTION The Molten Salt Reactor Experiment (MSRE) is located in Building 7503 at Oak Ridge National Laboratory (ORNL), a U.S. Department of Energy (DOE) facility in Oak Ridge, Tennessee. Building 7503 (Reactor Building) was built in the early 1950s and began operations in 1954, first by the Aircraft Reactor Experiment (operational 1954-1956) and then by MSRE (1965-1969). Building 7503 modifications for the MSRE began in 1962 and ended in 1964 when installation of the salt systems was completed. The MSRE was a graphite-moderated, liquid-fueled reactor operated from June 1965 through December 1969 as a demonstration of the technology needed to develop a large-scale molten salt breeder reactor.

The liquid fuel for the reactor was a fuel salt composed of UF4 dissolved in a carrier salt of LiF, BeF2 and ZrF4. The fuel salt was circulated by a fuel salt pump through the reactor vessel and a primary heat exchanger and was eventually drained into two fuel salt drain tanks (FDTs) located within the stainless steel-lined, concrete-shielded drain tank cell (DTC) adjacent to the reactor cell (Figures 1 and 2). A flush salt similar in composition to the fuel salt (without the uranium fuel) was used twice to flush the reactor system to remove residual pockets of fuel salt. The flush salt, which became contaminated with a small amount of residual fuel salt after each flush, was drained to the Fuel Flush Tank (FFT), also located in the DTC. The radioactivity of salts within the FFT accounts for approximately 1 to 2% of the total radioactivity of the three tanks within the DTC. Between 2004 and 2008, the uranium fuel was removed from the salts in the tanks, converting it to defueled coolant salt.

In 1996, the Reactive Gas Removal System (RGRS) was added to the process inventory to remove reactive gases containing uranium material and other reactive gases (such as fluorine, hydrogen fluoride, and molybdenum hexafluoride) and capture them on sodium fluoride and alumina traps.

This over 50-year old facility has been shutdown since reactor operations ceased in 1969. The facility has been scheduled for clean-up and demolition since the mid-1990s. In response to the Defense Nuclear Facilities Safety Board Recommendation 94-1 (DNFSB 1994), DOE noted for MSRE a comprehensive plan was established and put into place to initiate interim measures (drain water from the Auxiliary Charcoal Bed (ACB) cell, partition the off-gas system, and eliminate water sources), remove the uranium deposits, and dispose of the fuel salt. The interim measures will be completed by November 1995. The uranium deposits will be removed by February 1998, and the fuel salts by May 2000.

In June 1998, a Record of Decision (ROD) for Interim Action to Remove Fuel and Flush Salts from MSRE was issued that included a schedule for removal of uranium and salts from MSRE to be completed by February 2003. The ROD also noted that decontamination and demolition of Building 7503 and the MSRE reactor components will be performed as part of a later separate Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) final action. In 1998, a project was initiated to implement the remedy defined in the Record of Decision for Interim Action to remove fuel and flush salts from the MSRE facility (EPA 1998). The remedy defined in the ROD and subsequent Explanation of Significant Differences (EPA 2005) includes the following steps:

Melt and chemically treat the salts; Fluorinate the salt to remove uranium; Trap the uranium on cold traps and transfer the uranium to chemical traps; Ship the uranium loaded traps to ORNL Building 3019A for storage; Transfer residual salts to shield canisters designed for transportation and storage; and Ship shielded salt canisters to ORNL SWSA 5 for interim storage.

2 Figure 1. MSRE Facility Showing the Reactor and Drain Tanks.

3 Figure 2. Drain Tank Cell, Showing Fuel Drain Tanks and Flush Tank.

4 Between 2004 and 2008, the project completed steps 1 through 4 of the defined remedy. As a result, the uranium fuel was substantially removed from the salts in the tanks. Steps 5 and 6 were deferred pending an engineering/study evaluation to examine the approach for salt removal.

In May 2008, DOE issued a Phased Construction Completion Report (DOE 2008a) following the completion of steps 1 through 4 of the ROD and the removal of approximately 7.5 kilograms of uranium from the facility. In July 2008, DOE issued an Engineering Evaluation Work Plan (DOE 2008b) that described a thermal method for removing salts. In September 2009, an Engineering Evaluation report (BJC 2009a) recommended not proceeding with salt removal until the structural integrity (for both corrosion and stress-cracking concerns) of the drain tanks and flush tank are confirmed, the continued generation of fluorine inside the tanks considered, and the possible existence of multiple phases (stratification) of material in the salt be determined.

This report presents and evaluates six alternatives to addressing the remaining radioactive salts in the FFTs and FDT. Alternatives are considered that would either leave the tanks and contents in place or remove the salt from the MSRE facility. Removal options considered include intact tank removal, mechanical removal, and a thermal removal approach that does not pressurize the tanks during removal and provides mechanical agitation to address the possible stratification of the salt. The ultimate goal of the three removal alternatives is long-term geologic disposition at an appropriate facility such as the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. Several issues must be considered in relation to transport and disposal at WIPP or another long-term storage facility based on the physical, chemical and nuclear characteristics of the salts and their current regulatory classification. Additional discussion of issues related to disposal of the salts at WIPP is provided in Section 6.

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6

2. HISTORY OF ACTIONS/INTENTIONS/OUTCOMES Since reactor shutdown, MSRE has been the subject of a vast array of studies and reports, and there is a substantial collection of documents on the operation of the reactor, nuclear chemistry, safety, high temperature behavior of materials, decay products, the stored fuel salt, and much more. During the 40+ years since the reactor was shut-down, however, there have only been a few actions that changed the nature of the fuel salt or demonstrated tools and techniques for directly addressing the fuel salt. These actions, summarized in Table 1, are especially relevant to a review of the current conditions of the MSRE salt, evaluation of alternatives, and recommendation of future actions. In addition to the short descriptions of these key activities provided in the following table, the reports that describe these activities and other pertinent documents have been included with this report as attachments for convenience.

7 Table 1. MSRE Salt Related Studies and Removal Actions Action Description and Intent of Action Results or Conclusions Reference Non-Destructive Testing of Integrity of Drain and Flush Tanks (1998 - 1999)

A testing contractor performed non-destructive tests of the Drain and Flush tanks to determine if corrosion had weakened the tanks and threatened tank integrity.

Tests determined that the tanks had sufficient integrity to be pressured for transferring the salt. The salt transfer was attempted but did not occur.

Report on Molten Salt Reactor Experiment Drain Tank Qualification, BJC/OR-445 (BJC 2000)

Salt Melt and

Transfer, Demonstration (1999)

An existing tank of 5,754 lbs of MSRE Coolant Salt (not Fuel Salt) was melted and transferred into 5 storage vessels.

Transfer was performed by pressurizing the tank with argon gas and forcing molten salt out through a dip tube to new storage tanks.

Performed successful melt and transfer of salt to a set of 5 smaller storage tanks. Melt took 2 months, transfer took 2.5 hrs.

Transfer effectively emptied the coolant salt tank.

MSRE Remediation Project: Salt Melt, Transfer, and Process Demonstrations (ORNL 1999a)

Demonstration of a Melt Probe to Initiate Salt Melt (1999)

A process of progressively melting coolant salt with an internal thermal probe was developed to demonstrate a technology of safe melting and transfer of the MSRE fuel and flush salts.

The conclusion of the authors was that the prototype probe demonstrated basic suitability of design and performance and that is was possible and practical to establish and follow a progressive melt generated from a

central melting probe.

MSRE Remediation Project: Salt Melt, Transfer, and Process Demonstrations (ORNL 1999a)

Hydrofluorination Demonstration (1999)

A sample salt chemically identical to the fuel salt (but without activation of fission products) was used to demonstrate that fluorine can be restored to the salt by hydrofluorination and can be followed by analysis of the off-gas for HF or H2.

Restoring fluorine to the salt is required to completely melt the salt while avoiding the undesirable reduction of uranium and zirconium.

Excellent correspondence of the FTIR results and the titration values was obtained and the number of equivalents of HF consumed during treatment was very close to that associated with fluoride deficiency estimated to be in radiolyzed salt.

MSRE Remediation Project: Salt Melt, Transfer, and Process Demonstrations (ORNL 1999a)

Direct examination of 50 year old coolant salt (1999)

After the hydrofluorination process described above was complete and the salt cooled, the resulting solid salt was examined.

The final state of the salt was a

complete and homogeneous melt based on visual inspection.

Black graphite deposits (remnants from preparative crucibles) were seen on top of the salt boule. A superficial black film was apparent on the side walls and to a lesser extent, on the bottom, which was judged to be a mixture of reduced metals and salt. The empty salt container was found to be free of significant corrosion.

MSRE Remediation Project: Salt Melt, Transfer, and Process Demonstrations (ORNL 1999a)

Table 1. MSRE Salt Related Studies and Removal Actions (cont.)

8 Action Description and Intent of Action Results or Conclusions Reference Evaluation of Fluorine trapping agents for use with the MSRE Fuel Salt (1999)

An agent that traps fluorine without the generation of gaseous products or unwanted by-products is needed for long-term storage of MSRE salt. Ten different agents were evaluated for efficient trapping of radiolytic fluorine.

Sodium Iodide was the most successful candidate agent tested.

It rapidly and completely converted free fluorine to NaF without significant pressure accumulation and produced no deleterious by-products as long as the quantity of NaI was not exhausted.

Evaluation of Fluorine-Trapping Agents for Use During Storage of the MSRE Fuel Salt (DOE 1999b)

Removal of Uranium from Coolant Salt (Between December 2004 and April 2008)

Uranium was removed from the MSRE Fuel Salt.

Removal of uranium was performed by heating and melting the

salt, adding

fluorine, circulating the fluorinated salt through NaF traps and collecting the uranium. The quantity of uranium removed:

FFT: 468 grams FDT-1: 3,449 grams FDT-2: 3,496 grams Total: 6,413 grams The salts were processed beginning with the FFT, then FDT 2 and finally FDT 1.

Over 99% of the uranium was removed from the salt. With only trace amounts of uranium remaining, the salt is now coolant salt with TRU and fission product constituents. It should not be considered fuel salt or spent fuel.

Phased Construction Completion Report (DOE 2008a)

Attempted removal of coolant salt from MSRE Flush Tank (2008)

After the removal of uranium from the coolant salt described above, an attempt was made to transfer the defueled coolant salt to shielded canisters for on-site storage at ORNL.

Transferring of the defueled coolant salt from FFT was unsuccessful due to a

blockage in the transfer line, inadequate pressure or other cause.

Phased Construction Completion Report (DOE 2008a)

Evaluation of Fluorine Corrosion of Hasteloy (2009)

An engineering review of a plan to remove the residual salt from the drain and flush tanks included a

metallurgical, chemical and system analysis to in part, determine potential structural integrity of the

tanks, continued generation of fluorine, and the likelihood of more than one phase of material within the tanks.

The authors reported a

concern with the structural integrity of the tank due to the corrosion that likely occurred during the fluorination phase of the uranium removal process from 2004 through 2008. The authors further concluded that fluorine generation would continue, albeit at a lower rate after uranium removal and this would need to be considered when deciding the ultimate fate of the salts. Finally, the authors concluded that there is likely more than one phase of material in the tank, either suspended in the salt matrix or layered in the bottom of the tank, which may contribute to plugging during a planned transfer.

Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment (BJC 2009a)

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10

3. PURPOSE OF THIS REPORT This report describes alternatives for disposition of approximately 134 cubic feet of radioactive salt associated with the MSRE. The salt is currently located in three tanks in a steel-lined underground vault beneath the MSRE facility at ORNL (Figure 3).

Figure 3. MSRE Salt Tanks in the Drain Tank Cell.

This report will address two primary options, to keep the salt at MSRE or to remove the salt from MSRE.

Two alternatives are evaluated for keeping the salt at MSRE and three alternative methods are considered for removal of the salt from MSRE. A sixth alternative was developed to evaluate the removal of salt from MSRE and storing it on-site at ORNL. Once removed from MSRE, the salt will be sent to the Waste Isolation Pilot Plant in Carlsbad, New Mexico, for disposal, or the salt may be stored at ORNL pending future disposal at an undetermined off-site location. The alternatives presented in this report are represented by Figure 4.

11 Figure 4. Alternatives for Disposition of MSRE Defueled Coolant Salt.

The alternatives considered in this study are assessments of potential approaches for the future disposition of the MSRE salt, but are not necessarily comparable options since they do not all achieve the same objective or have identical outcomes. The first alternative, Maintain MSRE Facilities As-Is for 50 Years, is obviously not a disposal alternative, does not directly address the defueled salt, and is not a final solution, but is included as a benchmark to provide a description and cost of activities required to defer actions on the salt. Other alternatives may utilize various salt removal methods or container alternatives, or result in different storage and disposal destinations. Storage of salt at ORNL is not considered comparable to final disposal, but is included in the study because it is a potential temporary solution.

Disposal of the salt at the Waste Isolation Pilot Plant (WIPP) is a key element of this study because WIPP is currently considered the only viable disposal location for this type of material, and if the salt is placed in appropriate containers it meets the waste acceptance criteria for disposal WIPP, provided it is considered defense-related waste.

3.1 DISPOSAL AT WIPP Disposal of the MSRE Salt at WIPP is considered a desirable option because it is a permanent solution and it eliminates a significant inventory of radioactive material from ORNL. Removal of the salt is the key to decommissioning and demolition (D&D) of the MSRE facilities because other D&D work cannot be done without remote tools until the extremely high dose-rate salt is addressed. In order to facilitate transport to and disposal at WIPP, the alternatives for removal of salt from MSRE are evaluated in terms of the steps required to dispose of the salt at WIPP, including:

WIPP Waste Acceptance Criteria; Licensed and approved containers; Compatibility of container and waste form geometry; Radiation dose rates for transportation and handling at WIPP; and, Handling methods and equipment for disposal at WIPP.

12 3.2 STORAGE AT ORNL Storage of MSRE Salt at ORNL is considered a less desirable option than disposal at WIPP because it would be an interim measure of unknown duration. This would consist of removal of salt from MSRE, placement in a Type B or other container and on-site storage at an ORNL location. Entombment of salt in removable containers would also be considered storage at ORNL. Either removal of salt from the MSRE facility or entombment within the drain tank cell would allow D&D work on the facility to commence.

Storage at ORNL would require shielding, protection from weather, security and a monitoring program.

Input from regulatory agencies would be needed as no formal requirements are currently in place addressing this option.

3.3 INITIAL ALTERNATIVES CONSIDERED If the salt is removed from MSRE, containers will be required. Containers could be existing licensed Type B containers, new Type B containers (capable of being licensed), or custom shielding containers that would never be licensed and could only be used for transport within the Oak Ridge Reservation. The initial alternatives described above have been estimated with different container options.

The engineering evaluation requested by DOE listed six alternatives to address MSRE defueled coolant salt disposition. The alternatives highlighted in Table 2 below (in bold) generally correspond to the six alternatives requested for evaluation by DOE. In order to address all of the combinations of removal methods, container options and storage/disposal locations, additional variations of these alternatives were considered. Table 2 lists all of the alternatives for which cost estimates were prepared.

Table 2. Alternatives Estimated for Cost.

Alternative Description of Alternative 1

50 Years As-Is Maintain MSRE facilities, keep the salt in the Drain Tank Cell for 50 years 2

Entombment Salt tanks kept in Drain Tank Cell, tanks placed in retrievable shielded containers 3a Intact Tank New Type B containers designed, licensed, loaded, shipped to WIPP 3b Intact Tank New Type B containers designed, licensed, loaded, stored locally at ORNL 3c Intact Tank Salt tanks modified to fit existing Type B containers, shipped to WIPP 3d Intact Tank Salt tanks modified to fit existing Type B containers, stored locally at ORNL 3e Intact Tank Salt tanks placed in non-licensed shield containers, stored locally at ORNL 4

Thermal Removal Salt melted, transferred to RH-72B containers, shipped to WIPP 5a Mechanical Removal Salt removed with needle scaler, transferred to RH-72B containers, shipped to WIPP 5b Mechanical Removal Salt tanks sliced into sections, placed in RH-72B containers, shipped to WIPP 6a Thermal Removal Salt melted, transferred to RH-72B containers, stored locally at ORNL 6b Mechanical Removal Salt removed with needle scaler, transferred to RH-72B containers, stored locally 6c Mechanical Removal Salt tanks sliced into sections, placed in RH-72B containers, stored locally at ORNL

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14

4. NATURE/CHARACTERISTICS OF MSRE SALT AND TANKS 4.1 FUEL SALT, DRAIN TANKS, AND THE DRAIN TANK CELL During operation of the Molten Salt Reactor, the fuel salt circulated through a reactor vessel, fuel salt pump, and a primary heat exchanger at temperatures above 600° C (1,200° F). Heat was transferred from the fuel salt to a secondary coolant salt in the primary heat exchanger (BJC/DOE 2002). The nonradioactive coolant salt was composed of LiF (66%) and BeF (34%) and was drained to a separate tank. The nonradioactive coolant salt was transferred from its drain tank in 1999, as part of a demonstration of methods that were to be used in the thermal transfer of the radioactive defueled coolant salt.

After reactor shut-down, the molten fuel salt was drained into two fuel salt drain tanks (FDTs) located within the stainless steel-lined, concrete-shielded drain tank cell (DTC) adjacent to the reactor cell. A flush salt of similar composition to the coolant salt was used to flush the reactor system to remove residual pockets of fuel salt. The flush salt was circulated through the reactor system twice, and became contaminated with a small amount of residual fuel salt after each flush. It was drained to the Fuel Flush Tank (FFT), also located in the DTC. The radioactivity of salts within the FFT accounts for approximately 1 to 2% of the total radioactivity of the three tanks within the DTC.

The DTC is located below the high bay floor (elevation 852 ft). The DTC is a heavily reinforced concrete cell with a stainless-steel liner covering the cell floor and walls. The cell is approximately 18 ft x 21 ft and extends from the high bay to a flat floor at the 814-ft elevation. The top of the cell originally consisted of two layers of reinforced concrete block shielding. A stainless-steel membrane (seal pan) between the two layers of blocks was welded to the cell liner in order to seal the cell. A section of this shield was subsequently removed and a portable maintenance shield installed to facilitate access to the drain tanks. The portable maintenance shield (6-in.-thick steel plate) was designed with an eccentric plug and tooling plug in order to access each drain tank as required for the uranium removal and salt transfer campaign that took place between 2004 and 2008.

The FDTs, as shown on Figure 5, are supported by scales that provide continuous measurements of the weight of the tank and contents. The cut-away view shows how congested with pipes the FDTs are with their array of thermal wells, inlet and outlet piping, and steel rod bracing to hold all of the piping in place.

The FFT is similar to the FDTs but does not contain any thermal wells. FDT 1, FDT 2, and FFT are approximately 44%, 40% and 85% filled with salt, respectively.

15 Figure 5. Fuel Drain Tank.

Between 2004 and 2008, most of the uranium in the fuel salt was removed by melting and fluorinating the salt to produce gaseous UF6, and drawing the gasses through chemical filters to trap the uranium (DOE 2008a). The uranium remaining in the salt is estimated to be between 10 and 50 ppm, and accounts for approximately 4.3 curies of the total 12,782 curies of radioactive material. The characteristics of the MSRE salt tanks and salt are summarized in Table 3.

16 Table 3. Characteristics of MSRE Salt Tanks and Contents Parameter Fuel Drain Tank 1 (FD1 or FDT-1)

Fuel Drain Tank 2 (FD2 or FDT-2)

Flush Drain Tank (FFT)

Height 86.125 (86-1/8)a 79.8125 84.125 (84-1/8)

Height less hemispherical top and bottom (in) 64.6875 58.625 (58-5/8)c 59.0d Height Hemispherical Top (in) 12.5 (12-1/2)a 12.25 (12-1/4)c 12.5625 (12-9/16)d Height Hemispherical Bottom (in) 8.9375 (8-15/16)a 8.9375 (8-15/16)c 12.5625 (12-9/16)d Tank Wall Thickness (in) 0.75 (3/4)b 0.75 (3/4)c 0.75 (3/4)d Tank Diameter (O.D., in) 49.993a 49.990 50.051d Tank Volume

~80 ft3 (2,265-L)f

~80 ft3 (2,265-L)f

~80 ft3 (2,265-L)f Salt Volume 35.3 ft3 (1,000 L) 30.9 ft3 (875 L) 67.9 ft3 (1,923 L)

Total Weight (contents only) 5,465 lbs (2479 kg) 4,786 lbs (2171 kg) 9,402 lbs (4265 kg)

LiF (lbs)e 2,324 lbs (1054 kg) 2,034 lbs (923 kg) 4,828 lbs (2190 kg)

BeF2 (lbs)e 1,986 lbs (901 kg) 1,739 lbs (789 kg) 4,489 lbs (2036 kg)

ZrF4 (lbs)e 1,131 lbs (513 kg) 992 lbs (450 kg) 84 lbs (38 kg)

UF4 (lbs)e (Year 2000 Inventory)

~23 lbs (10.9 kg)

~20.5 lbs (9.3 kg)

~1.5 lbs (0.7 kg)

UF4 f (Year 2008 Inventory) 10 to 50 ppm 10 to 50 ppm 10 to 50 ppm Metal Oxide (lbs)e Trace Trace Trace Density of Salt @ 25 Ce 2.48 2.48 2.22 a Fuel Drain Tank 1 Cooling System Assembly, M20794RF003E10, 6/30/1961.

b Fuel Drain Tank Details, M20794RF007D10, 6/28/1961.

c Fuel Drain Tank Cooling System Assembly, FDT2 M20794RF016D2, 6/30/1961.

d Fuel Salt System Flush Tank Assembly and Details, D-FFA40462, M20794RF010D11, 12/29/1961.

e Melting-Hydrofluorination and Fluorination Process Flow Diagram, Drawing J3E020794-A005..

f Documented Safety Analysis for the Molten Salt Reactor Experimental Facility, ORNL, Oak Ridge, Tennessee, DSA-OR-7503-007R20.

4.2 RADIONUCLIDE INVENTORY The radionuclide inventory and characteristics of the salt meet the definition of TRU waste, which is waste that has been contaminated with alpha-emitting transuranic radionuclides, possessing half-lives greater than twenty years, and in concentrations greater than 100 nCi/g. The constituents of the remaining radioactive salt significant to the Safety Basis (BJC 2009d) are summarized in Table 4. A more complete list of all MSRE salt radionuclides is included in Attachment 4, Project Calculations: Radionuclide Inventory and Neutron/Gamma Source Term, CAN-02MSRE-A010, Rev.1 (BJC 2003d).

17 Table 4. MSRE Salts: Radiological Constituents Nuclide Source Emit.

Half Life FFT (Ci)

FDT-1 (Ci)

FDT-2 (Ci)

U-232 Thorium Fuel Cycle 68.9 yrs 0.428 0.468 0.3976 Th-228 U-232 Decay Prod.

1.912 yrs 0.856856 26.7696 22.74272 Ra-224 Th-228 daughter 3.66 days 0.856856 26.7696 22.74272 Po-216 Rn-220 daughter 0.145 sec 0.856856 26.7696 22.74272 Pb-212 Po-216 daughter 10.64 hrs 1.713712 53.5392 45.48544 Bi-212 Th-228 daughter 60.6 min 0.856856 26.7696 22.74272 Po-212 Po-216 daughter 0.3 sec 0.545272 17.0352 14.47264 Tl-208 Bi-212 daughter 3.05 min 0.305592 9.5472 8.11104 U-233 Thorium Fuel Cycle 160,000 yrs 0.856 1.17 0.994 Pu-238 Added as a Tracer1 87.7 yrs 0.05778 2.808 2.3856 Pu-239 Added as a Tracer1 24,200 yrs 0.8988 25.74 21.868 Pu-240 Added as a Tracer1 6,563 yrs 0.16478 9.828 8.3496 Pu-241 Added as a Tracer1 14 yrs 1.2412 17.55 14.91 Pu-242 Added as a Tracer1 373,300 yrs 0.00002782 0.001521 0.0012922 Sr-90 Fission Product 28.8 yrs 109.14 3510 2982 Cs-137 Fission Product 30.07 yrs 92.02 3042 2584.4 Total Curies of all Nuclides 210.8 6796.8 5774.4 1Minor amount from neutron activation.

4.3 SAMPLING AND ANALYSIS The radiation emitted by the three salt tanks is extremely high. 137Cs is the predominant gamma emitter and accounts for nearly 3 million R/hour. Direct measurements and sampling of the salts have been limited. The inventories of the salt constituents have largely been calculated or inferred from measurements of off-gasses. The calculated inventory applies defensible methodology in deriving the radionuclide specific activities and should be acceptable to WIPP in lieu of direct sampling and laboratory analysis. However, if WIPP insists upon direct sampling and laboratory analysis for acceptance of the waste material, a sample port can be installed in the collection manifold for the removal options, and samples can be collected during the removal/containerization phase.

The industry standard code for calculating burn-up, radioactive decay, and radioactive source terms is the OakRidgeIsotopeGeneration (ORIGEN) code, which was developed at ORNL. The principal use of the ORIGEN code is to calculate the radionuclide composition and other related properties of nuclear materials. The materials most commonly characterized include spent fuels, radioactive wastes (principally high-level waste, such as MSRE salt), recovered elements (e.g., uranium, plutonium), uranium ore and mill tailings, and gaseous effluent streams (e.g., noble gasses). Several ORIGEN code analyses have been performed to characterize the MSRE salt. All of the radionuclide quantities and curies presented in this study were determined by ORIGEN analysis (BJC 2003d) modified by non-destructive analyses and reductions of inventory due to other MSRE operations (BJC 2002b).

Measurement of radiation fields adjacent to the MSRE tanks and thermal wells is underway but not completed. Measurements of radiation fields are performed at different locations on the outside of the Drain and Flush Tanks using tools that are readily available. A more extensive direct sampling and

18 analysis of MSRE salt has been discussed, but is not currently planned due to technical challenges. If the purpose of sampling the salt is to characterize the salt constituents, a composite sample would be adequate. However, if the objective is to develop a profile of salt strata, several core samples would be needed to provide a representative characterization of the salts. To extract core samples of the salt, tools would have to be developed that would be able to obtain appropriately sized cores. Sample size would need to be small enough to not require excessive shielding or produce such a large radiation field that no lab could accept the samples, but large enough to develop a profile of the salt strata.

As noted, extensive characterization of the salt through direct sampling is technically challenging, and for the purposes of meeting acceptance criteria at WIPP, may not be necessary. The primary acceptance criteria that are pertinent to the salts are related to dose rates, both for shipping to and handling at WIPP.

The salt constituent that is primarily responsible for elevated radiological dose rates is 137Cs. Quantities of 137Cs in the salt are relatively stable, due to its long half life and the fact that it is a fission product and not produced during radioactive decay of other elements. Therefore, additional characterization of the salts through sampling prior to any attempt to transfer the salt would likely not be required to assure acceptance at WIPP from a radiological standpoint.

The issue of stratification of the salt, mentioned above, is important for performing exposure rate calculations and for the developing methods for salt removal, and transfer to new containers. Exposure rate calculations assume that the large gamma contributor, 137Cs, is distributed uniformly throughout the salt. However, theory dictates that the components of the pure salt itself will separate and freeze to a solid in a very non-uniform predictable manner as the molten salt cools, based on their different melting points (ORNL 2002). The chemical form, melting point, and affinity of the 137Cs for any of these various non-homogeneous salt components are unknown. As a result, efforts are underway to profile the gamma dose spatially by inserting a detector down a thermal well penetrating one of the de-fueled salt tanks. A previous attempt to profile the dose failed because radiation levels exceeded the upper limit of instrumentation. The dose profile, if acquired, would give some indication of the uniformity of distribution of the 137Cs. If the 137Cs is limited to a particular location in the salt, this could strongly impact the exposure rate in specific containers or near specific locations around a container, favoring designs and container filling methods that are more likely to mix or evenly distribute the 137Cs.

4.4 MSRE SALT CONTINUING REACTION / RADIOLYSIS / OFF GAS One issue that has been a concern since shutdown of the MSRE reactor is the extent to which the MSRE salt continues to react, the mechanisms responsible for ongoing reaction, and off-gas produced. Before the removal of the uranium between 2004 and 2008, radiolysis of fluorine salts produced HF and free molecular fluorine F2 in sufficient quantity that the pressure in the tanks could have reached 100 psi unless the tanks were periodically vented to the facility off-gas system. Since uranium removal (to trace levels), radiolysis has been reduced to the extent that only a minute amount of HF and F2 are produced and the tank headspace gases are still pumped to the RGRS based on the tank pressure readings. For long term disposal, these two materials can be captured by the addition of a sodium iodide getter. The radioactive constituents of the salt and their decay chain daughter products are continuing to react, but those reactions do not produce any gasses other than de minimis amounts of thoron and radon.

In 1999, an evaluation was performed of ten candidate materials for fluorine-trapping agents (ORNL 1999b). Sodium iodide, NaI, was determined to be the best material for trapping free fluorine while minimizing any gas pressure rise. The NaI converts to NaF and captures all of the free fluorine until all of the available NaI has reacted. As long as an excess quantity of NaI is available to react to free fluorine, the formation of corrosive interhalogens (IF5) can be avoided. Granular unsintered NaI is preferable to NaI that has been exposed to high temperatures, but both will work.

19 4.5 MSRE DRAIN AND FLUSH TANK CORROSION The stainless-steel alloy first known as INOR-8, now known by the trade name Hastelloy N, was developed at ORNL for MSRE. The alloy was determined to be the most promising material for containment of molten fluorides at high temperatures (800 degrees C). Hastelloy N was used for the MSRE reactor, piping, pumps, coolant tanks, drain tanks, flush tank and other facility components that would either be in contact with fluorine salts or extreme heat. Extensive studies were performed to determine the corrosion potential of Hastelloy N exposed to fluorine salts, HF, and F2, and to develop more resistant formulations of Hastelloy N. Table 5 shows several compositions of Hastelloy N that were used at MSRE and a current composition of Hastelloy N.

Table 5. Compositions of Hastelloy N (%)

Alloy Ni Cr Mo Fe Si Mn C

Nb Ti Co Cu W

Regular Hastelloy N (Haynes SP-19) 70.5 7.4 16.7 4.8 0.13 0.48 Regular Hastelloy N (Heat Y-8460) 73.9 7.3 15.9 2.4 0.15 0.31 Hastelloy N - 2% Nb (ORNL Welding Rod) 70.6 7.5 15.4 3.9 0.54 2.1 Hastelloy N (Haynes Intl. 2010) 68.7 7

16 5

1 0.8 0.08 0.5 0.2 0.35 0.5 In 2000, prior to uranium removal and anticipated salt removal, the drain and flush tanks were examined by means of nondestructive testing, including visual examination, ultrasonic thickness measurements, eddy-current-wall-loss measurements, thermal analysis and finite element stress analysis (BJC 2000). The conclusion of the testing was that the mechanical integrity of the tanks was more than adequate to withstand pressures up to 80 psia at 1100°F. Since then, all three tanks have been exposed to both HF and F2 gas during the uranium removal steps while the salt was in a molten state (~450°C). Previous evaluations concluded that even short exposures to fluorine gas under elevated temperatures can result in tank corrosion of 0.1 mil/hour or more (BJC 2009a, Haubenreich 1970, ORNL 1969, ORNL 1985).

Recommendations were made in an engineering evaluation to conduct a similar evaluation to the one conducted in 2000 before heating or pressurizing the tanks, and conduct additional nondestructive tests to evaluate any pitting or stress-corrosion cracking damage to the tanks that may have occurred as a result of fluorination (BJC 2009a). The report further recommended a step-wise approach to heating and pressurizing the tanks, initially isolating and pressurizing a tank for a period of time prior to heating it, to demonstrate leak tightness. Then, heat the tank, insert the probe and apply pressure to the headspace in a gradual manner so as to not overstress the tank structure (BJC 2009a).

20

5. ALTERNATIVES FOR DISPOSITION OF SALT IN DRAIN TANKS AND FLUSH TANK Several alternatives were considered in this study for disposition of the defueled coolant salt in the drain and flush tanks. Generally, the options fall into two categories; those involving leaving the tanks in place and those involving removing the tanks and/or the salt contained inside the tanks.

Options involving leaving the tanks in place include:

1. Maintaining the salt and tanks in their current state for the next 50 years, and

2. Entombing the tanks within the drain tank cell by pouring flowable fill, engineered grout, or other containment media into the cell. Sleeves would be placed around the tanks in the cell prior to filling with the intent of making the three tanks retrievable from the entombed cell. Also, the tanks in shield sleeves will be lifted up to the level of the lower shield floor to be above the local water table.

Options involving tank and salt removal include:

3. Removal of Intact Tanks 3a. Intact tank removal utilizing a new Type B container to accommodate the tanks and ship them to WIPP. A complete container design, fabrication and licensing program would be required.

3b. Intact tank removal utilizing a new Type B container to accommodate the tanks and store them locally at ORNL. The container would be licensed for shipping to a remote site, but shipping and disposal are not included.

3c. Intact tank removal utilizing an existing licensed Type B Ten Drum Over Pack (TDOP) container to accommodate the tanks, and ship them to WIPP. WIPP handling methods and equipment would have to be modified to receive and dispose of this container.

3d. Intact tank removal utilizing an existing licensed Type B TDOP container to accommodate the tanks, and store them locally at ORNL. The container would be licensed for shipping to a remote site, but shipping and disposal are not included.

3e. Intact tank removal utilizing new and unlicensed shielded Type-B - like containers to safely remove the tanks and move them to a storage location on the ORNL site (i.e., SWSA #5).

4. Thermal removal and transfer of the salt from the Drain and Flush Tanks to a new disposal container designed to fit into a RH-72B Payload Canister, and ship to WIPP.

5. Mechanical Removal of salt and disposition at WIPP 5a. Mechanical removal of the salt using a needle scaler tool to break the hardened salt into particles, transferring the particles to disposal containers that would fit in RH-72B Payload Canisters, and shipping to WIPP.

5b. Mechanical removal of the salt by cutting the tanks into vertical slices, transferring the slices to RH-72B Payload Canisters, and shipping to WIPP.

6.

On-site Storage of salt in an approved type B container 6a. Thermal removal and transfer of the salt from the Drain and Flush Tanks to a new disposal container designed to fit into a RH-72B Payload Canister, and store locally at ORNL.

21 6b. Mechanical removal of the salt using a needle scaler tool to break the hardened salt into particles, transferring the particles to disposal containers that would fit in RH-72B Payload Canisters, and storing locally at ORNL.

6c. Mechanical removal of the salt by cutting the tanks into vertical slices, transferring the slices to RH-72B Payload Canisters, and storing locally at ORNL.

5.1. ALTERNATIVES FOR LEAVING TANKS IN PLACE 5.1.1 Maintain for 50 Years at MSRE Facility Maintaining the salts in their current tanks for the next 50 years will require maintenance and operation of the systems needed to safely store the salts in the absence of any removal activities during the 50-year duration. This alternative would continue to manage the MSRE as a Nuclear Category 2 facility and the associated surveillance and maintenance (S&M) activities and programmatic requirements (i.e., conduct of operations, inspections, technical safety controls, etc.) would continue through the 50-year period; however, a significant increase in major repairs and replacement of systems and facility components would be required since the facility and many of its components are already over 50-yers old and were not intended to support this extension in the life of the facility.

This alternative presents the cost benchmark for deferral of decontamination activities and models a No Action scenario.

5.1.1.0 Assumptions This alternative assumes that there will be no salt removal activities performed over the 50-year period of study, with those removal activities being delayed or deferred until a later date.

Additional assumptions include:

For analysis and comparison to other alternatives, a high-level parametric forecast is adequate for determining 50-year costs, using practices and techniques that are generally acceptable to DOE.

The MSRE S&M budget is not responsible for parking lot, landscaping or utilities repairs or replacements.

5.1.1.1 Technical Feasibility MSRE maintenance over the 50-year term is technically feasible, but will require a new funding strategy to cover anticipated and forecasted costs for anticipated major repairs or replacement activities that are currently dealt with as change items to the Interim S&M baseline budget.

Maintaining the facility and its processes (i.e., the drain tanks, the Fuel Salt Disposal System (FSDS), the RGRS, exhaust ventilation and monitoring systems) as-is will require a change in maintenance strategy from primarily corrective maintenance (defined in DOE O 430-1B as the repair of restoration of failed or malfunctioning equipment, systems, or facilities to their intended functions or design conditions. It does not result in an significant extension of the expected useful life.) to a focus on life extension of this over 50-years old facility and sustainment (defined in DOE O 430-1B as maintenance and repair activities necessary to keep the inventory of facilities in good working order. This includes regularly scheduled maintenance as well as anticipated major repairs or replacement of components that occur periodically over the expected life of the facilities.). Figures 6 through 8 illustrate the cost trends associated with this shift in approach.

22 5.1.1.2 Discussion This alternative would maintain MSRE as a Nuclear Category 2 facility with the associated controls remaining in place. Considering the issues identified in the 2009 Engineering Evaluation (BJC 2009a) with respect to structural integrity and fluorine generation, changes to the monitoring program may be required to ensure that further degradation of the tanks does not increase risk over the 50-year period to human health and the environment.

From a CERCLA perspective, the ROD (EPA 1998) would be amended through an Explanation of Significant Differences (ESD) document with approval from the U.S. Environmental Protection Agency (EPA) and the Tennessee Department of Environment and Conservation (TDEC). The amended ROD would essentially rescind the requirement to remove the salt from MSRE and interim storage of the shielded salt canisters at an alternate location in ORNL. Maintaining the tanks and facility in place may also delay the final CERCLA ROD for Melton Valley. If this alternative is selected, the final action for MSRE may be deferred to the final Bethel Valley ROD, such that MSRE actions do not impact the final Melton Valley ROD which is scheduled to be prepared and approved in advance of Bethel Valley. This strategy of moving to Bethel Valley would need to be negotiated with EPA and TDEC during the negotiations for the ESD and can be incorporated into the ESD.

5.1.1.3 Cost and Schedule The schedule for the as-is maintenance alternative is 50-years. During that time, the surveillance activities will continue, and the facilities and surveillance equipment will be maintained in a sustainment mode.

The cost for this maintenance is forecasted to average $9.939 million annually in 2010 dollars, using the forecast methodology described below. The average annualized forecast consists of $2.744 million for surveillance and routine maintenance (i.e., preventive maintenance & minor repair) activities, plus

$5.041 million for major repair and replacement for the facility components and $2.154 million for the process inventory.

The cumulative total cost over the 50-year duration is $496.9 million in 2010-dollars, or $971.5 million in escalated dollars (using the current BJC escalation rate of 2.3% per annum).

Comparative analyses of the projected annualized forecast show the following:

The forecast annual average of $2.744-million for surveillance and routine maintenance compared to the FY10 Actual Cost of Work Performed of $2.681-million shows a variance of +2.34%, which indicates that the forecast reasonably aligns with the current comparable work.

The forecast annual average for the total of surveillance and routine maintenance routine plus non-routine maintenance, such as major repairs and replacement compared a conformed version of the original As Is condition maintenance costs forecast in 1996 for the Feasibility Study for Fuel and Flush Salts Removal from the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee. Feasibility Study Alternate 1 (FSA1) for 30-year duration of S&M activities in the absence of any remedial action was prepared on October 30, 1996 and had a 30-year then-current value of $150,938,000 (equivalent to $224.3-million in 2010$, adjusted using ENR BCI factors to extrapolate current value), with an average annual value of $5.031-million (equivalent to $7.477-million in 2010$, adjusted using ENR BCI factors to extrapolate current value). Since the FSA1 forecast did not include the maintenance costs related to the FSDS system, we added an average annual value of $2,266,353 (10% of the asset value) to our maintenance model basis to account for the FSDS (which shows on BJC property records as having an acquisition date of October 1, 2001 and a then-current value of $16,601,614 that converts to a

$22,663,534 value million in 2010$, adjusted using ENR BCI factors to extrapolate current value).

23 Adding 2010$ value of $2.266-million average annual maintenance cost to the 2010$ average annual $7.477-million forecast results in an adjusted annual value of $9.743-million. Our forecast annual average of $9.939-million average annual maintenance cost compared to the updated and conformed FSA1 annual average maintenance forecast of $9.743-million shows a variance of

+2.62%, which indicates that our forecast reasonably aligns with the comparable 1996 FSA1 forecast in 2010$.

The detailed basis of estimate and estimate for this alternative are provided as Appendices A and B to this document.

24 50-Year Sustainment Cost Trend

$200.00

$400.00

$600.00

$800.00

$1,000.00

$1,200.00

$1,400.00

$1,600.00

$1,800.00 1

5 9 13 17 21 25 29 33 37 41 45 49 Year Maintenance Costs (2010$/gsf)

Yearly Total Cost Ave Annual PM & UR Figure 6. Sustainment Cost Trend.

50-Year Major Repair & Replace Trend

$200.00

$400.00

$600.00

$800.00

$1,000.00

$1,200.00

$1,400.00

$1,600.00

$1,800.00 1

5 9 13 17 21 25 29 33 37 41 45 49 Year Maintenance Costs (2010$/gsf)

Yearly Total Cost Ave Annual MR&R Figure 7. Major Repair & Replace Trend.

25 50-Year Cumulative O&M Costs

$200,000,000

$400,000,000

$600,000,000

$800,000,000

$1,000,000,000 1

5 9 13 17 21 25 29 33 37 41 45 49 Year Maintenance Costs ($ total)

Current $ Cum Escalated $ Cum Figure 8. 50-Year Cumulative O&M Costs.

5.1.2 Entombment at MSRE Facility Previous investigations have raised a number of concerns regarding entombment as a permanent disposal option for the salts (ORNL1996). Many were related to either the hazards associated with long-term entombment or the regulatory pathway to permanent disposal of spent nuclear fuel and/or TRU waste (EPA 1998 and ORNL 1996). Prior to fuel removal, the long lifetime of the 233U isotope was a primary concern when considering that the regulatory requirements for entombment would likely be performance based. While removal of the uranium fuel has reduced the anticipated radioactive longevity of total fission products and daughters from tens of thousands of years to approximately 1000+ years of activities above 10 Ci (ORNL 1988), a number of technical and regulatory hurdles remain.

The entombment approach developed for this alternative includes the following key items:

Placing sleeves around the tanks in the cell prior to filling, with the intent of making the three tanks retrievable from the entombed cell.

Lifting the tanks in shield sleeves up to the level of the lower shield floor to be above the local water table.

Filling the drain tank cell with flowable fill, engineered grout, or other containment media.

Figure 9 shows the DTC filled with flowable fill material and the three tanks in shield sleeves located above the water table elevation and in a retrievable configuration.

26 5.1.2.1 Hydrogeologic Setting Bedrock beneath the MSRE facility is a dark calcareous shale, which is part of the Conasauga Group that underlies the Melton Valley. The Conasauga shale formation is heterogeneous and relatively low in permeability, although flow velocities of a few feet per week are possible in the soil/weathered overburden. The overburden averages 20 ft thick and consists of a blanket of topsoil generally less than 1 ft thick on top of weathered shale. Groundwater levels in the vicinity of the reactor are currently suppressed by pumping a 3-ft by 3-ft sump located to the southeast of the reactor cell. The elevation of the sump is approximately 811 ft msl (ORNL 1965). A previous failure of the pumps resulted in a rise in groundwater levels to an elevation corresponding to approximately 5 ft in the sump room (ORNL 1996).

This corresponds to an elevation of approximately 823 feet mean sea level (ft msl). Previous evaluations have determined the water levels in the vicinity of MSRE would naturally reach approximately 835 ft msl in the absence of pumping (ORNL 1996). An elevation of 835 ft msl corresponds to an elevation slightly above the bottom of the lower drain tank shield block. The elevation of the bottom of the DTC is approximately 814 ft msl.

5.1.2.2 Technical Feasibility There are several technical considerations regarding entombment as a long-term disposal option. These include hydrogeologic or seismic considerations, potential chemical interactions that may occur over time between the groundwater and soils at the site and the entombing media and the effects of corrosive gas(es) produced by continued radiolysis of the salts. The entombing media should be resistant to the effects of radiation from TRU waste and fission products in the salts and external corrosion during long-term submersion in groundwater. The technical feasibility of removing obstacles from the DTC and sealing openings through the cell walls, if it were to be used for secondary containment, would need to be considered during the design phase.

One example of a material that may potentially address many of the technical considerations regarding entombment is an epoxy grout developed by ChemCo Systems, a manufacturer of epoxy and polyurea products. One of their grouts that has previously been used for long-term sealing of radiological materials is, after curing, essentially inert and non-leachable. It is stable in temperature to at least 200° F, is not sensitive to any level of alkalinity and is only sensitive to acidic environments when the pH is well below 1. Chlorinated solvents and some aromatic solvents can swell the grout and ultimately cause damage if left in contact for an extended period. While this material appears to meet several of the objectives for long-term containment, additional testing of the grout would be necessary to verify performance.

27 Figure 9. Entombment Overview and Detail.

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29 Some of the technical concerns are reduced if entombment is approached as a short-term storage alternative rather than a long-term disposal option. Entombment could be performed in a manner that leaves the drain and flush tanks in the DTC, but lifts them above the water table elevation and encloses them in a shield sleeve that is retrievable. Getter could also be added to the tanks and/or shield sleeves to absorb any free fluorine. The primary advantages to the storage approach are that it would place the highly radioactive salt in a safe storage configuration until a permanent disposal approach is developed and Building 7503 could be demolished without removal of the salt.

5.1.2.3 Cost and Schedule The schedule for the entombment alternative is 62 months. The estimated cost for this alternative is

$77,087,758. The detailed schedule, basis of estimate, and estimate for this alternative is provided as appendices to this document.

5.1.2.4 Discussion Many of the same steps in removing piping and other infrastructure to accommodate intact tank removal would also be required for entombment. As the local water table is shallow at MSRE, the interactions of primary or secondary containment materials with groundwater must be considered if the salt is not relocated to an elevation above the water table. In addition, the containment facility could be designed to specifications more appropriate for long-term monitoring and possible removal, if other long-term disposal options, such as WIPP, become viable. If future removal becomes viable, the shielded container would be lifted from the sleeve, transported to an ORNL hot cell and the salt would be removed and placed into containers for transport and off-site disposal. The cost for future removal and disposal is not included in the estimate for this alternative, but would be comparable to a removal packaging and disposal alternative.

Although this alternative does not reduce the radiological inventory for MSRE, entombment may allow for a change in the categorization of the facility by analysis. In order to support the analysis, the radiological inventory at MSRE that is not associated with the FFT and FDT (i.e., charcoal canisters, etc.)

would be removed from the facility and disposed of appropriately. Once the non-salt related contamination is removed, the inventory of the tanks would be within the threshold for a Nuclear Category 3 facility.

With the approval of the revised safety basis, the facility S&M and associated controls may be reduced slightly until facility D&D is complete. Considering the issues identified in the 2009 Engineering Evaluation (BJC2009a) with respect to structural integrity and fluorine generation, changes to the monitoring program may be required to ensure that further degradation of the tanks does not increase risk over the 50-year period to human health and the environment.

From a CERCLA perspective, the ROD (EPA 1998) would be amended through an ESD document with approval from EPA and TDEC. The amended ROD would essentially rescind the requirement to remove the salt from MSRE and interim storage of the shielded salt canisters at an alternate location in ORNL and add the in place entombment of the tanks and salt. A risk assessment would be necessary to verify that risks of entombment to human health and the environment would be acceptable. Once entombment is approved as part of the interim action defined in the ROD, the interim action would need to be assessed as part of the CERCLA final action to address the D&D of Building 7503 and the MSRE reactor components.

30 5.2 ALTERNATIVES FOR SALT REMOVAL FROM MSRE 5.2.1 Geometries and Volumes of Drain and Flush Tanks and Defueled Coolant Salt The defueled coolant salt is currently contained in two drain tanks and one flush tank. The tank exterior dimensions are approximately 50 inches in diameter and 86 inches high. The tanks are 3/4 inch thick Hastelloy-N with domes top and bottom. The total volume of each tank is approximately 80 cubic feet.

Each drain tank has 32 thin-walled heat transfer tubes, thermal wells, that extend through the top domes of the tanks to approximately 15 inches above the bottom of the tank. The flush tank does not contain thermal wells.

Drain Tank 1 contains approximately 35 cubic feet of salt, which fills 44% of the tank volume. It contains approximately 6,800 curies of radioactive material of which 3,000 curies are 137Cs.

Drain Tank 2 contains approximately 31 cubic feet of salt, which fills 39% of the tank volume. It contains approximately 5,700 curies of radioactive material of which 2,500 curies are 137Cs.

The Flush Tank contains approximately 68 cubic feet of salt, which fills 85% of the tank volume. It contains approximately 200 curies of radioactive material of which 90 curies are 137Cs.

The gamma radiation given off by Drain Tanks 1 and 2 is extremely high and will require all work on or near the tanks to be performed with remote tools and significant shielding. Even the flush tank that contains only 2% of the total radioactive material produces a radiation field that requires special care.

5.2.2 Intact Tank Removal This salt removal alternative includes a variety of options for removal, packaging, and disposal or storage of the three tanks, FDT 1, FDT 2, and the FFT, as intact tanks. For purposes of this discussion, intact tank means either the entire tank (approximately 85 inches tall) or the tank section including the salt with an upper portion of the tank removed. The advantages of removal of the tanks intact are:

The salt would not have to be melted/removed and transferred to new containers.

Removal of the intact tanks could be a relatively simple method of removing the salt and tanks from MSRE, if a non-Type B container option is selected. Removal of the tank intact by lifting each tank into a shield sleeve would not require cutting of tanks, heating of tanks, pressurization, removal of salt, additional canisters or extensive tank integrity testing.

Corrosion of the tanks would be less of a concern with intact tank removal than with other removal alternatives due to the fact that the tanks would not be mechanically stressed or require heating or pressurization. Also, in their current configuration, the tanks are suspended by a series of hangers attached to a yoke, and lifting the tanks by the yoke would not change their means of support.

Removal and disposal of the salt and tank together would remove all or part of the tank such that the contaminated tank would not have to be dealt with later. (All alternatives leave radioactively contaminated piping, traps, and charcoal filters to be addressed later.)

The disadvantages include:

Piping to the tanks would have to be remotely cut to make the tanks removable from the DTC where they are currently located.

There is no disposal container currently licensed for use at WIPP that matches the geometry or payload weight of an intact tank.

31 There are two general approaches for removal and storage/disposal of intact tanks. The first would be to simply lift each tank out of the DTC and place it in a container large enough to receive the entire tank.

The second would be to cut an intact tank to a size that would fit in an existing licensed Type-B, Ten Drum Over Pack (TDOP) container. Five variations of these two approaches have been considered, including:

Intact tank removal utilizing a new Type B container to accommodate the tanks and ship them to WIPP. A complete container design, fabrication and licensing program would be required. (Cost Estimate Alternative 3a)

Intact tank removal utilizing a new Type B container to accommodate the tanks and store them locally at ORNL. The container would be licensed for shipping to a remote site, but shipping and disposal would not be included. (Cost Estimate Alternative 3b)

Intact tank removal utilizing an existing licensed Type B TDOP container to accommodate the tanks, and ship them to WIPP. WIPP handling methods and equipment would have to be modified to receive and dispose of this container. (Cost Estimate Alternative 3c)

Intact tank removal utilizing an existing licensed Type B TDOP container to accommodate the tanks, and store them locally at ORNL. The container would be licensed for shipping to a remote site, but shipping and disposal would not be included in the cost estimate. (Cost Estimate Alternative 3d)

Intact tank removal utilizing new and unlicensed shielded containers (Type B - like containers) to safely remove the tanks and move them to a storage location on the ORNL site. (Cost Estimate Alternative 3e)

If the intact tanks are to be placed un-cut in large containers, either a new Type-B container large enough in diameter and tall enough to accommodate an intact tank or a large Type-B - like container would have to be developed. A new Type-B container would be designed, fabricated, drop tested, fire tested, and licensed for radioactive material transport. The new container would include enough lead shielding that it could meet Department of Transportation (DOT) dose rate limits for shipping. It would also have to be shielded adequately for contact handling at WIPP, but would require WIPP to modify their disposal program to accommodate larger and heavier containers than are currently handled. A shielded container that is similar to a Type-B container but not tested and licensed could be used to contain an intact tank, move it around on the Oak Ridge Reservation outside of commerce, and store it, but it would not qualify for shipping off-site inside of commerce.

TDOP containers are licensed for contact handled (CH) disposal at WIPP. The TDOP is the only container currently licensed for WIPP that is large enough in diameter for an intact tank. A TDOP (Figure 10) is larger in diameter but shorter than an intact tank, and TDOPs do not include any layers of shielding.

32 Figure 10. Ten Drum Over Pack (TDOP).

The inside diameter of the TDOP is 71.5 inches and the outside diameter is 50 inches. In the TDOP there is room for the intact tank plus a 10 inch annular space for shielding and bracing. The interior height of the TDOP is 70.5 inches, but the overall height of the tanks is approximately 90 inches. The TDOP is, therefore, too short for the tanks unless the tops of the tanks are cut off and the overall tank height shortened. A diamond wire saw would be used to cut the tank above the level of the salt. For FDT 1 and FDT 2, the approach would be as follows:

Modify a TDOP to accommodate an intact tank by filling the annular space with shielding and other fill material, leaving an empty space in the center of the TDOP to receive the intact tank.

Locate the modified TDOP in the DTC, (suspended between the tanks and the shield floor) where an intact tank can be lifted and placed into it with minimal shielding.

In the drain tanks, fill one foot above the top of the salt with epoxy grout through the existing 3-inch ball valve access port. Epoxy grout would limit salt exposure to air (oxygen and moisture) during removal of the top of tank described below. The drain tanks would then be approximately 55% full. Epoxy grout would not be placed in the flush tank, due to lack of headspace.

Either lift each tank to a clear space above the heaters, piping, etc., or provide work space around the tanks in the DTC, by cutting and removing equipment that is in the way of the diamond wire saw. Cut piping can be dropped to the floor of the vault.

Currently, the tanks are suspended from support frames that would no longer support the tanks if the tops are cut off with the saw. Whether tanks are lifted clear of other equipment or left in place, a temporary base or other means of support will be required to replace the existing supports.

33 To shorten a drain tank and make it fit in a TDOP, cut completely through the top of the tank with a diamond wire saw, cutting all of the thermal wells, piping, etc., just above the level of the epoxy grout. To shorten the flush tank and make it fit in a TDOP, cut through the tank at the top of the salt reduce the length to 70 inches.

Set the cut off top of the tank and piping in the floor of the DTC.

Lift the tank and place it in the modified TDOP.

Fill all of the void space around the tanks in the TDOP with NaI getter.

Add shielding over the top of the intact tank.

Seal the TDOP with its bolt-on lid.

Future disposal at a cost comparable to other disposal options.

5.2.2.1 Technical Feasibility It is feasible to create a new licensed Type-B container or a new unlicensed Type-B-like container. A licensed Type-B container would have to be larger and heavier than any previously licensed. The objective of creating a new Type-B container would be off-site disposal. A new unlicensed Type-B-like container would be used to shield and contain intact tanks for removal from MSRE and storage on the ORNL site.

Use of a TDOP is feasible but problematic. First, the weight of an intact tank, salt, getter, a layer of grout, and the TDOP is approximately 16,000 lbs and the weight limit of the TDOP plus contents is 7,000 lbs if shipped in a TRUPACT-II container or 14,500 lbs if shipped in a 10-160B container (Figure 11). So, with either shipping container the TDOP is over its licensed weight limit. Second, the intact tank in an unshielded TDOP is so radioactive it exceeds the dose-rate limits for contact handling at WIPP and therefore the container must be remote-handled (RH). The TDOP would be under the shipping dose rate limits if shipped in the 10-160B shipping container, which has three inches of steel and 2 inches of lead shielding, but could not be handled at WIPP. At WIPP, TDOP containers are removed from the 10-160B containers and contact handled. WIPP does remotely handle TDOP containers (Figure 12).

Cutting the tanks with a diamond wire say is required to shorten the tanks enough to fit in a TDOP.

However, cutting remotely with a diamond wire saw is technically challenging. Difficulties include the following:

Fatigue of diamond wire cable; Debonding of diamond wire abrasives; Potential chemical interaction of cutting fluids with tank contents; and, Difficulties of replacement of diamond wire saw cable.

If an alternative requiring diamond wire saw cutting is proposed, significant testing and development of diamond wire saw technique with Hastelloy would be required to develop procedures to address these concerns.

34 Two licensed shipping containers used to ship waste materials to WIPP. Both shipping containers can accommodate a Ten Drum Over Pack.

Figure 11. Licensed Shipping Containers for a TDOP.

Figure 12. Contact Handling of Ten Drum Over Packs at WIPP.

35 5.2.2.2 Assumptions A new Type-B container could be designed and fabricated that could accommodate the weight of an intact tank, shielding, and getter, and pass a drop test.

A new Type-B container or a TDOP would always have to be contact handled at WIPP. WIPP currently has one Remote Handling (RH) approach, including one size of RH payload container and horizontally drilled wall borehole. The capability exists at WIPP to perform different size wall boreholes but there is currently only one size of RH lifting and transport devises. It is not likely that WIPP would develop new sizes of equipment and new procedures to remotely handle and dispose of only three containers.

Storage at ORNL would be in shielded containers, located above ground, protected from the weather with a simple enclosure such as a sea-land container or small metal building. The storage would be temporary. Future maintenance of the storage facility or disposal off-site is not included in the estimated costs.

5.2.2.3 Cost and Schedule Estimated costs and schedules for Alternatives 3a through 3e are summarized below. The detailed schedule, basis of estimate, and estimate for these alternatives is provided as appendices to this document.

3a. Intact tank removal utilizing a new Type B container to accommodate the tanks and ship them to WIPP is estimated to cost $105,607,312 and require 77 months.

3b. Intact tank removal utilizing a new Type B container to accommodate the tanks and store them locally at ORNL is estimated to cost $99,876,888 and require 70 months. (The container would be licensed for shipping to a remote site, but shipping and disposal is not included in the cost estimate.)

3c. Intact tank removal utilizing an existing licensed Type B TDOP container to accommodate the tanks, and ship them to WIPP is estimated to cost $83,836,722 and require 66 months.

3d. Intact tank removal utilizing an existing licensed Type B TDOP container to accommodate the tanks, and store them locally at ORNL is estimated to cost $82,941,310 and require 65 months.

(The container would be licensed for shipping to a remote site, but shipping and disposal would not be included in the cost estimate.)

3e. Intact tank removal utilizing new and unlicensed shielded containers (Type B - like containers) to safely remove the tanks and move them to a storage location on the ORNL site is estimated to cost

$79,223,896 and require 63 months.

5.2.2.4 Discussion Intact tank removal provides the following benefits:

Stratification of salt, hard lenses that may not melt, tar-like substances or other issues that must be addressed in thermal or mechanical transfer of salt are not a concern if the salt remains in the drain and flush tanks.

Intact tank removal would not require heating of salt, pressurizing of the tanks, tank integrity evaluation, creation of special tools or other activities needed to remove the salt from the tanks.

Intact tank removal would result in the removal from MSRE of both salt and tanks.

In spite of the benefits of intact tank removal, it should only be considered a favorable alternative if the decision is made to store the salt at ORNL in an unlicensed shielded container. Cutting the tanks to fit in currently licensed Type-B containers or creating a new Type-B container have too many technical

36 challenges (described below) to be reasonable alternatives. The decision to store the salt at ORNL would be a logical conclusion if permanent geologic disposal at WIPP or a similar facility is not available as an option, and the best approach to temporary storage would be in an unlicensed shield container. It would not be a permanent solution but it would be an easier and less costly temporary solution than other alternatives.

Development of a new Type-B container is a lengthy and expensive proposition, generally requiring a minimum of 18 months and $5 million dollars to design, fabricate, test and qualify a new container. It would not be reasonable to go through a lengthy and expensive design and licensing program for new containers to ship only three intact tanks. Also, there is no assurance that a large and heavy container once licensed for shipping would ever be handled at WIPP.

Cutting the intact tank down to a size that would fit in a licensed WIPP container would be difficult to implement. The existing Type-B waste container (TDOP) with the intact tank would either exceed allowable radiation dose rates if shipped unshielded or, if shielding is added, would be over the payload weight limits. In either case, the effort to remove and place the intact tank in a licensed container would be significant, but would not result in disposal of the tanks at WIPP and would only be eligible for storage on the Oak Ridge Reservation.

As any scenario involving the removal of the intact tank from MSRE would result in great challenges to off-site shipment and disposal either because of excessive weight or radiation exposure issues, initial alternative 3e, packaging in a Type-B like container for storage at ORNL, will be carried forward as Alternative 3 because it is the most favorable with respect to cost and schedule. The remaining intact removal options will be removed from further evaluation.

Although this alternative would significantly reduce the radiological inventory for MSRE by removing the radioactive salt from the facility, the inventory would not be reduced to less than Category 3 based on inventory. However, following the removal of the salt, the argument could be made that due to the form and distribution of the remaining radioactive material, primarily located in charcoal in containers, the facility may be re-categorized to Less than Cat 3. If this argument gains acceptance, the basis would be related to design analysis rather than radiological inventory. With the approval of the revised safety basis, the facility S&M and associated controls would be reduced accordingly.

From a CERCLA perspective, the ROD (EPA 1998) would be amended through an ESD document with approval from EPA and TDEC. The amended ROD would essentially replace the requirement to remove the salt by thermal means and store the salt canisters on site with intact removal and off-site disposal.

Once intact removal and disposal is approved as part of the interim action defined in the ROD, the decisions associated with the CERCLA final action to address the D&D of Building 7503 and the MSRE reactor components are simplified.

5.2.3 Thermal Removal Thermal removal of the salt from the two FDTs and the FFT would consist of melting the salt and transferring it from the existing drain and flush tanks to new containers. Thermal removal was the method envisioned by the designers of the reactor, who included piping and equipment to melt and transfer salt from the Drain Tanks to transfer tanks for storage or disposal.

In 1999, a demonstration of salt melt and transfer was performed on a single tank of non-radioactive coolant salt from the MSRE Secondary Cooling loop. The demonstration project successfully melted the salt and transferred it by using nitrogen gas pressure to force the molten salt out of its tank into 5 smaller

37 containers (ORNL 1999a). One of these smaller containers was later used to develop and demonstrate the use of a melt probe device (ORNL 1999a).

Between 2004 and 2008, BJC conducted a campaign to remove the uranium from the Fuel Salt. The salt in the Flush Tank and Drain Tanks was melted using a melt probe, and fluorine was introduced through the probe. The free fluorine converted liquid UF4 to gaseous (at elevated temperature) UF6. The off-gas was then circulated through NaF traps and 99% of the uranium removed, to a level of approximately 10 to 50 ppm. During the uranium fluorination and removal campaign, an attempt was made to utilize the existing system to transfer salt to new storage tanks. New storage tanks (designed to fit into RH-72B Payload Canisters) were connected to the existing facility piping, the tanks were pressurized to 50 psi, and the transfer initiated, but the molten salt could not be transferred possibly due to a clog in the transfer line (DOE 2008a). After attempts to clear the piping proved unsuccessful, transfer efforts were discontinued, the heaters turned off, and the salt (with uranium removed to trace amounts) was allowed to cool and solidify in tanks FDT-1, FDT-2, and FFT.

The thermal removal alternative, as illustrated in Figure 13, would utilize existing heating equipment, introduce new equipment and methods where appropriate, and build on lessons learned from previous efforts to melt and transfer the salt. Key elements of thermal removal would include:

Use of the existing heating panels to melt the salt in tanks FDT-1, FDT-2, and FFT; Use of a melt probe to initiate the melt from the top down in each tank; Agitation of the molten salt to break-up and suspend any solids at melt temperature, if possible; Removal of the molten salt by pumping it from the existing tanks; New heat-traced and insulated piping through which to pump the salt; New disposal tanks (suitable for use with a RH-72B shipping cask inner vessel) to receive the molten salt; and, A valve or manifold system to distribute the molten salt to an array of new disposal tanks.

For thermal salt removal, a new tool assembly consisting of a melt probe, an agitator, and pump suction would have to be developed. This cluster of tools would be combined in a heat-traced steel sleeve that would be introduced into the tanks through the existing 3-inch diameter access port above the center of each tank. The tool sleeve would provide the stiffness needed by the long, slender tools and would provide the heat-tracing to keep the molten salt at high temperature while it is being pumped from the Drain and Flush Tanks to new disposal containers. The tool cluster would consist of an off-the-shelf agitator and pump combined with a melt probe. A solid tip melt probe would be used initiate the melt from the top down in a manner similar to that conducted during the Drain Tank uranium removal. Design and fabrication of tools would consist only of development of the heat-traced steel sleeve and modification of existing tools to fit together into the relatively compact steel sleeve cross-section.

The pump would be a commercially available pump for high temperature viscous liquid. The pump drive would be in the high bay above the Drain Tank Vault, and an impeller or pump suction would be inserted into the tank. By using a pump to transfer the molten salt instead of transferring the salt by the use of gas pressure, there would be no reason to increase the pressure in the tanks. Thus, the risk of tank failure due to the combination of increased pressure and internal corrosion of the tank is minimized.

The agitator would be an adaptation of a commercially available 55-gallon drum mixer designed for mixing high viscosity material, such as the following example from Neptune Mixer Company, Lansdale, Pennsylvania (Figure 14). It is designed with folding propeller that is made to fit through a 2 inch diameter 55-gallon barrel bung. The propeller and bottom of the shaft would likely have to be replaced with high-temperature and corrosion resistant materials and the 3/4 inch shaft would have to be lengthened

38 and adapted with u-joints to offset the motor out of the way of the other tools. The agitation provided by this type of equipment should be sufficiently robust to break up and mix metal fluorides into the molten salt. A black film or tar-like material reported to be present on the bottom of the tanks (ORNL 1999a) might prove more difficult to mix into the molten salt.

39 Figure 13. Salt Melt and Transfer Process.

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41 Figure 14. Molten Salt Mixer with a Folding Propeller.

Other equipment required for thermal salt removal would include:

A glove-box enclosure or hot cell (for operating the tools) located in the high bay above the Drain and Flush Tanks. This would be larger than the current glove box at the site, and designed for performing functions specific to thermal removal.

A heat traced pipeline to location of the disposal containers; Meters and valves to direct a measured quantity of melted salt into each of 12 containers; and, 1/4-inch Hastelloy disposal containers similar to the ones fabricated for the 2004-2008 attempted salt transfer.

Two key features of the thermal transfer approach are the design of a double wall disposal tank and the salt mixing and transfer method. A new disposal tank (Figure 15) would be designed with a 13.5-inch diameter inner tube for fuel salt and a 23-inch diameter outer tank for flush salt and NaI getter. The fuel salt in the inner tube would be shielded by the flush salt and getter in the outer tank. An optimum number of twelve containers has been selected to meet fissile gram equivalent and dose rate requirements for transport and disposal at WIPP. The distribution of fuel salt and flush salt into twelve containers would leave a void volume of approximately 9 cubic feet to be filled with NaI getter. Per the 1999 study, the minimum quantity of NaI getter needed for each container was estimated to be 0.8 cubic feet. Therefore,

42 the volume of getter proposed for inclusion in the disposal containers to fill the void volume and provide shielding is over ten times the minimum volume required (excess getter volume is considered beneficial).

The disposal tank would fit in a RH-72B Payload Canister with an annular space for lead shielding. The salt would be melted, mixed with an agitator and pumped slowly to a set of disposal tanks. The pump and fill rate would be slow enough that salt homogenized by mixing in the drain and flush tanks could be introduced into the disposal tank and would solidify in successive thin layers and minimize stratification.

By using the low-dose flush salt and getter for shielding and keeping the drain salt from stratifying, the 137Cs would be evenly distributed and high radiation hot spots eliminated.

Figure 15. Double Wall Disposal Container.

5.2.3.1 Filling and Handling of Disposal Containers A concept has been developed for filling the disposal containers with defueled coolant salt, flush salt and getter, and for placing the filled containers in RH-72B Payload Canisters, Inner Vessels, and Outer Casks.

The container filling and handling equipment would be remotely operated and completely shielded. The equipment could be used for thermal or mechanical transfer of salt, and consists of the following features:

A simple conveyor or carousel placed around the perimeter of the DTC above the lower shield floor. The carousel would accommodate 12 disposal containers.

43 The carousel would sequentially move the containers around the perimeter to a fill position for filling with defueled coolant salt, flush salt, and then getter.

The tanks would then be moved sequentially to a lidding position where lids would be applied.

Filled tanks with lids would then be sequentially moved around to a loading position where each tank would be raised within a shielded tower to be placed in RH-72B containers.

Figure 16 shows the Filling and Handling equipment concept.

5.2.3.2 Assumptions Prior to heating the tanks for thermal removal of the salts, the structural integrity of the tanks should be verified in accordance with the testing described in Section 4.5. This testing will also ensure that pitting and corrosion as a result of the fluorination and fuel removal has not compromised the structural integrity of the tanks.

Based on information from past activities, the Drain and Flush Tank heaters are assumed to work well. Their performance was reliable in the past, and there is no reason to assume they will not heat the tanks adequately for salt melting, 500-600 degrees C.

The 1999 Coolant Salt Transfer demonstrated that the thermal process could transfer 99% of the tank contents. It is assumed that the heel remaining in the Drain and Flush Tanks would be minimal.

If lenses of metal fluorides or other non-salt materials are present in the salt, it is assumed that the agitation will break up and mix those materials into the molten salt. Any tar-like substances that may be in the tanks are assumed to adhere to the surfaces of the tank and agitation may not mix those into the molten salt.

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45 Figure 16. Filling and handling equipment.

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47 5.2.3.3 Technical Feasibility Thermal transfer to new disposal containers is feasible. Similar melt and transfer activities have been performed at MSRE in the past. In 1999, the non-radioactive coolant salt was successfully melted and transferred to disposal containers. Between 2004 and 2008, radioactive drain and flush tank salt was heated using a melt probe and the uranium was removed. After uranium removal, an attempt was made to transfer the flush salt to new disposal containers. Salt transfer was ultimately unsuccessful, possibly due to a clogged pipeline, but the heaters and melt probe worked, indicating that a thermal approach to transfer is feasible. Problems of clogging would be addressed by new pipelines and heat-tracing.

It is also important to note that the thermal removal process described above does not require pressurization of the tanks during removal. By removing the molten salt with pumps, this alternative does not adversely stress the existing tank walls during the operation.

5.2.3.4 Cost and Schedule Thermal removal and transfer of the salt from the Drain and Flush Tanks to a new disposal container designed to fit into a RH-72B Payload Canister, and ship to WIPP is estimated to cost $110,545,020 and require 71 months. The detailed schedule, basis of estimate, and estimate for these alternatives is provided as appendices to this document.

5.2.3.5 Discussion Thermal removal and repackaging the salt in 12 disposal containers is the recommended alternative for removal of the salt from the Drain and Flush tanks and transfer to WIPP for disposal. The disposal container concept is intended to conform to the geometry, dose-rate, and weight limits for remote handled waste at WIPP, and the concentric tube design will provide shielding and minimize dose-rate if the salt is non-homogeneous. By using a carousel to remotely move disposal containers sequentially to a filling station, Drain Tank salt would be distributed incrementally in thin layers to each of the disposal containers and limit the possibility of high levels of radionuclides being concentrated in a single tank or in a single layer within a few tanks. If the Drain Tank salt is stratified and radionuclides are concentrated in portions of the salt, both the disposal container design and the filling method would limit high dose rates at the exterior of the disposal container.

As with all removal alternatives, this alternative would significantly reduce the radiological inventory for MSRE; however, the inventory would not be reduced to less than Cat 3 levels. Once the salt is removed, the argument could be made that due to the form and distribution of the remaining radioactive material, primarily located in charcoal in containers, the facility may be re-categorized to Less than Cat 3. If this argument gains acceptance, the basis would be related to design analysis rather than radiological inventory. With the approval of the revised safety basis, the facility S&M and associated controls would be reduced accordingly.

From a CERCLA perspective, the ROD (EPA 1998) would be amended through an ESD document with approval from EPA and TDEC. The amended ROD would essentially replace the requirement to store the salt canisters on site with off-site disposal.

The Disposal Containers are conceptually designed to minimize the radiation from the salt and to fit into RH-72B Payload Canisters. Once the Drain and Flush Tank Salt and Getter have been placed in 12 Disposal Containers, which are then inserted into RH-72B Payload Canisters, the containers will meet DOT transportation rad-dose limits and will meet the WIPP Waste Acceptance Criteria for Remote Handled waste.

48 5.2.4 Mechanical Removal Mechanical removal of the salt may be performed by one of two methods:

Using a remotely manipulated pneumatic hammer to loosen the solidified salt and make it available for a vacuum to transfer it to containers for disposal, or Cutting an entire tank and its salt contents into vertical sections that would fit in a RH-72B payload container.

5.2.4.1 Hammer and Vacuum to Loosen and Transfer Salt.

A type of pneumatic hammer, a needle scaler, would be used to loosen the salt. Needle scalers are typically used for removing rust, welding slag, paint or other unwanted coatings from metal parts.

Pneumatically driven needle scaler / vacuum tool combinations are commercially available and relatively inexpensive. The needle scaler / vacuum would be placed on the end of a manipulator arm with a camera for visual guidance of tools (Figure 17). This type of tool was previously used at ORNL to break up and remove solidified charcoal and uranium as part of the Uranium Deposit Removal (UDR) Project, completed in May 2001 (DOE 2010). On the UDR project, the needle scaler was not able to break up some of the solidified charcoal and uranium, and a long handled chisel and sledge-hammer were used to initially break up the charcoal such that the needle scaler could reduce it to a small enough size to vacuum transport (BJC/DOE 2002a). It is thought that the Drain and Flush Tank salts would be more easily broken down to granules than the charcoal and uranium. If needed, the needle scaler could be temporarily replaced on the end of the manipulator arm by a pneumatically driven chisel tool.

Mechanical removal by this method would require a very maneuverable manipulator arm to work around the interferences in the tanks. Interferences in the drain tanks include 32 vertical thermal well pipes tied together with a lattice of 1/2 inch steel rods and other pipes and steel rod pipe supports embedded in the salt. The manipulator arm, needle scaler, vacuum and camera assembly would be inserted into the tanks through the 3-inch access port and would be operated from above the shield floor. The needle scaler and vacuum would remove the salt by breaking it up and vacuum transferring it to disposal containers sized to fit into RH-72B Payload Canisters for shipping to WIPP. The same disposal containers described for thermal transfer could be used for mechanical transfer.

5.2.4.2 Cutting Tanks with a Diamond Wire Saw The objective of cutting an entire tank and contents into sections is to make slices that can fit licensed WIPP disposal containers. The only workable method would be to cut vertical slices that would fit into RH-72B Payload Canisters. Horizontal slices would fit in a TDOP, but could not be remotely handled at WIPP and thus would require some shielding or the slices to be very thin (to minimize the handling dose rate). Four vertical slices with a diamond wire saw would cut a tank and its salt into nine sections that would fit into RH-72B Payload Canisters, as shown in Figure 18. Salt Tank Sliced into Nine Pieces and a Slice in a WIPP Payload Canister.

A cutting container would be designed and fabricated to contain the tanks during cutting and catch all metal and salt saw dust that is produced by the four diamond wire saw slices. A cone bottom on the cutting container would be connected to a shielded drum that would catch all of the cuttings. Each of the FDTs and the FFT would be cut free from the piping currently connected to them and lifted one-by-one out of their current location and placed into the cutting container. Lifting lugs and support cables would be attached to the top of the tank such that each of the nine slices would be supported from above by a cable that could lift each slice separately from the container. After attaching lifting cables to the tank, a containment lid would be lowered to seal the cutting container. Four diamond wire saws mounted inside

49 the cutting container on vertical slides would slice through the Hastelloy tanks and the salt. The resulting nine slices would fit into separate RH-72B Payload Containers.

Twenty-seven Payload Containers would be required to handle all of the slices. The radiation field from the slices would vary from very low to high depending on the shape of the slice and whether it is a slice of a high radiation Drain Tank or a slice of the lower radiation Flush Tank. The highest radiation field slice would be below the remote handling dose rate limit of 200 R/hr. The space around the slices in the Payload Containers would be filled with getter and other inert filler material. The total weight of each slice in its Payload Container, including the 1,100 lb Payload Container, the section of Hastelloy tank, the salt, the getter, and any inert fill material used to hold slices together through the cutting process, would be under the 8,000 pound limit for the RH-72B payload.

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51 Figure 17. Mechanical Removal Details and Views.

52 Figure 18. Salt Tank Sliced into Nine Pieces and a Slice in a WIPP Payload Canister.

53 5.2.4.3 Assumptions If lenses of metal fluorides or other non-salt materials are present in the salt, it is assumed that the material will break up and can be vacuumed out of the tanks. Any tar-like substances that may be in the tanks are assumed to be in very small quantity and adhering to surfaces. If tar-like substances cannot be broken loose through hammering and remain in the tanks, the volume of that material will be small enough to be inconsequential.

It is assumed that the pneumatic hammer removal of drain tank salt could be performed at rate of 3 to 5 cubic feet per day.

For a cutting application, it is assumed the salt will remain relatively intact with pieces of the tank for containerization.

5.2.4.4 Technical Feasibility Mechanical transfer to new disposal containers is feasible. Execution through the pneumatic hammer approach is likely to be less complex than the mechanical slicing approach. However based on past experience at ORNL, the success of the pneumatic hammer would be dependent on the hardness of the material encountered. Previous experience during the 2004-2008 uranium removal and the uranium deposit removal project indicates that the solid salt is likely harder than concrete.

The mechanical slicing approach would require four vertical slices top to bottom through the Hastelloy tanks and the defueled salt. The same technical challenges that would be encountered for horizontally slicing the tank for insertion into a TDOP would apply to vertical slices, including the following:

Fatigue of diamond wire cable Debonding of diamond wire abrasives.

Chemical interaction of cutting fluids with tank contents.

Difficulties of replacement of diamond wire saw cable.

In addition, these challenges are compounded by the need to work in a controlled environment remotely and slice through the salt. All cutting would need to be performed within a sealed containment structure to maintain an inert atmosphere, collect all cuttings and fluids and minimize spread of contamination. If diamond wire saw cutting is proposed, significant testing and development of diamond wire saw technique with Hastelloy would be required to develop procedures to address these concerns.

5.2.4.5 Cost and Schedule Estimated costs and schedules for Alternatives 5a and 5b are summarized below. The detailed schedule, basis of estimate, and estimate for these alternatives is provided as appendices to this document.

5a. Mechanical removal of the salt using a needle scaler tool to break the hardened salt into particles, transferring the particles to disposal containers that would fit in RH-72B Payload Canisters, and shipping to WIPP is estimated to cost $101,115,524 and require 69 months.

5b. Mechanical removal of the salt by cutting the tanks into vertical slices, transferring the slices to RH-72B Payload Canisters, and shipping to WIPP is estimated to cost $111,527,248 and require 71 months.

5.2.4.6 Discussion Of the mechanical removal options, the pneumatic hammer/needle scaler alternative is preferred because it is easier to implement and is more favorable with respect to cost and schedule. Therefore, initial

54 alternative 5a will be carried forward as Alternative 5. The mechanical slicing option will be removed from further evaluation.

The primary disadvantages of needle scaler mechanical salt removal are the difficulty of working around the array of pipes and rods in the solidified salt and the air or gas filtration that would be required for containment of airborne particulates.

If adequately maneuverable tools are combined with the needle scaler, the method has several advantages over other methods:

The needle scaler approach does not require the prolonged heating/melting required for thermal methods, resulting in cost and schedule benefits. However, actual salt removal with the needle scaler would not necessarily be fast. It is estimated that removal of drain tank salt could be performed at rate of 3 to 5 cubic feet per day. The slow speed of removal means that a minimal amount of salt is involved at any time, which reduces risk.

The needle scaler approach does not have to be performed in a continuous campaign. If something happens during loosening of the salt that requires the process to be shut down for a while, the system can simply be shut down and easily restarted. In fact, the transfer of granular salt loosened with tools could be performed over a period of weeks with the equipment started in the morning, operated during the day, shut down at night and the work picked up the next day where it left off.

The work can be performed at room temperature, not 1,100 degrees F. Heat-tracing, melt probes, insulation, heater panels, and heated vaults would not be required.

Lenses of materials that would not melt or black film substances could be removed from the drain tanks as long as they could be chipped to a size less than approximately 3/4 inch.

As with all removal alternatives, this alternative would significantly reduce the radiological inventory for MSRE; however, the inventory would not be reduced the less than Cat 3 levels. Once the salt is removed, the argument could be made that due to the form and distribution of the remaining radioactive material, primarily located in charcoal in containers, the facility may be re-categorized to Less than Cat 3. If this argument gains acceptance, the basis would be related to design analysis rather than radiological inventory. With the approval of the revised safety basis, the facility S&M and associated controls would be reduced accordingly.

From a CERCLA perspective, the ROD (EPA 1998) would be amended through an ESD document with approval from EPA and TDEC. The amended ROD would essentially replace the requirement to remove the salt by mechanical means and store the salt canisters on site with mechanical removal and off-site disposal. Once removal and disposal is approved as part of the interim action defined in the ROD, the decisions associated with the CERCLA final action to address the D&D of Building 7503 and the MSRE reactor components are simplified.

5.2.5 On-site Storage of Salt in an Approved Type B Container Storage at ORNL involves removal of the salt from the MSRE using one of the removal techniques described previously. Three initial alternatives were considered:

Thermal removal of salt to a new disposal container designed to fit into a RH-72B Payload Canister for storage at ORNL (Cost Estimate Alternative 6a).

Mechanical removal of salt using a needle scaler tool to a new disposal container designed to fit into a RH-72B Payload Canister for storage at ORNL (Cost Estimate Alternative 6b).

Mechanical removal by cutting the tanks into sections that would fit in a RH-72B payload container for storage at ORNL (Cost Estimate Alternative 6c).

55 5.2.5.1 Technical Feasibility The removal options described for the on-site storage alternative have been addressed earlier and are considered feasible. On-site storage at an alternate ORNL location would require modifications to the safety basis of the location.

5.2.5.2 Cost and Schedule Estimated costs and schedules for Alternatives 6a through 6c are summarized below. The detailed schedule, basis of estimate, and estimate for these alternatives is provided as appendices to this document.

6a. Thermal removal and transfer of the salt from the Drain and Flush Tanks to a new disposal container designed to fit into a RH-72B Payload Canister, and store locally at ORNL is estimated to cost $104,544,646 and require 64 months.

6b. Mechanical removal of the salt using a needle scaler tool to break the hardened salt into particles, transferring the particles to disposal containers that would fit in RH-72B Payload Canisters, and storing locally at ORNL is estimated to cost $95,212,161 and require 62 months.

6c. Mechanical removal of the salt by cutting the tanks into vertical slices, transferring the slices to RH-72B Payload Canisters, and storing locally at ORNL is estimated to cost $107,353,227 and require 66 months.

5.2.5.3 Discussion Of the on-site storage of a Type B container, the pneumatic hammer/needle scaler alternative is preferred because it is implementable and is more favorable with respect to cost and schedule. Therefore, initial Alternative 6b will be carried forward as Alternative 6. The remaining options will be removed from further evaluation.

As with all removal alternatives, this alternative would significantly reduce the radiological inventory for MSRE; however, the inventory would not be reduced the less than Cat 3 levels. Once the salt is removed, the argument could be made that due to the form and distribution of the remaining radioactive material, primarily located in charcoal in containers, the facility may be re-categorized to Less than Cat 3. If this argument gains acceptance, the basis would be related to design analysis rather than radiological inventory. With the approval of the revised safety basis, the facility S&M and associated controls would be reduced accordingly.

From a CERCLA perspective, the ROD (EPA 1998) would be amended through an ESD document with approval from EPA and TDEC. The amended ROD would essentially update the description of how the salt would be removed from the tanks and packaged. As the existing ROD allows for on-site storage, this component of the ROD would not be impacted. Once the ESD is approved as part of the interim action defined in the ROD, the decisions associated with the CERCLA final action to address the D&D of Building 7503 and the MSRE reactor components are simplified.

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57

6. SHIPPING TO WIPP WIPP is a DOE facility located in southeastern New Mexico, 26 miles east of Carlsbad, NM. Congress designated WIPP as the nations defense-related transuranic waste repository in the Land Withdrawal Act of 1992. It is part of an overall TRU waste program that provides an integrated management system from generation through transportation and disposal. WIPP was designed to accept and dispose of both Contact Handled-and Remote Handled-TRU waste.

The alternatives for salt removal described in this report were, in part, formulated based on the goal of ultimate disposal at WIPP. The processes of salt removal, container geometries and sampling options discussed in the alternatives section considered the need to meet WIPP waste acceptance criteria, in addition to DOT and NRC requirements on transport of radiological material. Details on some of the more pertinent regulatory requirements with respect to MSRE salt are discussed in this section. This includes a comparison of the MSRE salt radiological characteristics and WIPP WAC requirements.

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59 Figure 19. Considerations for Off-Site Shipment and Disposal at WIPP.

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61 6.1 WASTE ACCEPTANCE CRITERIA The WIPP Waste Acceptance Criteria (WAC) document (DOE 2010b) provides TRU waste generator and storage sites with the minimum criteria that TRU waste shipments must meet prior to transport to and disposal at WIPP. Included in the WAC are specific requirements for containers, radiological, chemical and physical waste properties, in addition to information on data package characterization and certification. All of these requirements will need to be addressed by site-specific plans and procedures describing the details of processes, controls, techniques and other actions to be applied to each TRU waste container. The following sections provide an overview of some of the more pertinent WIPP criteria in relation to the configuration of salts within payload containers.

6.1.1 Container Properties Container properties addressed by WIPP include general description/configuration, weight limit, assembly configuration, removable surface contamination, identification/labeling, dunnage and filter vents. Additional discussion of weight limits is provided in Section 6.5. Currently, the only authorized payload containers for shipment of RH TRU waste to WIPP are 55-gallon drums and the RH-TRU waste canisters (either removable lid or fixed lid) (DOE 2010c). Type B containers currently certified by the NRC for transportation of RH TRU waste include the 10-160B and RH-TRU 72B Casks. The only approved payload configuration for the 10-160B cask is 2 stacked pallets, each holding five 55-gallon drums in a circular arrangement. The RH-TRU 72B is designed to hold a single payload canister, approximately 26 in. o.d. and 120.5 inches long. RH-TRU waste may be directly loaded into the canister or packaged within waste containers (e.g., 55-gallon drums or metal cans) in the canister (DOE 2006a).

Payload Canisters greater than 4 liters in volume require venting, except for metal containers packaging solid inorganic waste materials that do not generate any flammable gas. Compliance with this requirement must be verified by one or a combination of the following methods:

Review of records and database information, which may include knowledge of process; Radiography; Visual observation; and, Sampling program.

6.1.2 Radiological Properties Radiological properties for packages required by WIPP may be divided into two groups. The first group includes the activities and masses of the ten WIPP-tracked radionuclides, which are 241Am, 238Pu, 239Pu, 240Pu, 242Pu, 233U, 234U, 238U, 90Sr, and 137Cs. The total activity of these elements must be quantified and tracked to ensure compliance with the Land Withdrawal Act limits of 23 Ci per canister, 5.1 million Ci disposed, and surface dose rates of 1000 rem/hr. Estimates of activities and masses may be derived from Acceptable Knowledge (AK), computations, measurements, sampling or other methods (DOE 2010b).

The second group includes the remaining radionuclides contributing to the 239Pu fissile gram equivalent (FGE), the 239Pu equivalent Curies (PE-Ci), and the decay heat of the payload container. Acceptable 239Pu FGE values are determined based on the Be/BeO content of the waste (Table 4). The PE-Ci limit for solidified/vitrified waste is 1,800 per RH-TRU waste canister or 55-gallon drum. Decay heat is calculated by site personnel using the US DOE RadCalc program. Compliance with the decay heat limit may also be used to meet the gas generation limit of 5% or lower hydrogen concentration by volume during transportation. The 5% limit on hydrogen concentration may be converted to limits on the decay heat for each RH-TRU content code since radiolysis of the TRU materials is the primary mechanism by which hydrogen is generated.

62 Additional general requirements include a minimum of 100 nanocuries of alpha-emitting transuranic isotopes per gram of waste, with half-lives greater than 20 years, and an external dose rate of individual containers 200 mrem/hr and 1,000 Rem/hr.

Table 6 provides a comparison of many of the estimated values for salt radiological properties to WIPP waste acceptance criteria limits. In each case, the estimated values for the salts in each of the drain tanks are within the WAC. The surface dose rate limit for RH waste is 1,000 rem/hr. A calculation in MicroShield software resulted in a surface dose rate of 82.5 rem/hr for each canister, assuming 1/4 inch of lead shielding and the waste is homogenized.

Table 6. Comparison of MSRE Salt Radiological Properties and WIPP WAC Limits WIPP Limit FDT-1 FDT-2 Flush Salt Volume of Salt (m³)

1.00 0.875 1.923 Mass of Salt (kg) 2479 2171 4265 Specific activity (Ci/L) 23 6.80 6.60 0.11 239Pu equivalent activity (PE-Ci) a 1,800 38.8 32.9 1.36 239Pu fissile gram equivalents (g) 100 31b 27b 3b Thermal Power (watts) 300 c

c c

Surface dose rate (rem/hr) 200 mrem/hr and 1,000 rem/hr 82.5 rem/hrd 82.5 rem/hrd 82.5 rem/hrd Transuranic content (nCi/g) 100 nCi/g 23,033 22,344 755 a - Based on the sum of 233U, 238Pu, 239Pu, 240Pu, 241Pu and 242Pu.

b - 239Pu FGE is taken from the draft WIPP TRU Determination Report, accounting for distribution among 12 anticipated payload containers (thermal and mechanical - needle scaler options).

c - The thermal power for each tank is calculated using the RadCalc program, but thermal power generation from the salts in the tanks is not anticipated to be above WIPP limits.

d - Output from MicroShield assuming homogenized waste in a RH-72B Payload Canister with 1/4 inch of lead shielding.

As noted, WIPP is authorized to accept defense-related TRU waste. The WAC notes that the DOE and its predecessor agencies were engaged in a broad range of activities that fall under the heading of atomic energy defense activities. However, currently the MSRE facility is not listed in the WIPP 2009 Annual Transuranic Waste Inventory Report. A 1995 letter from the chairman of the Defense Nuclear Facilities Safety Board also states that MSRE is not a DOE nuclear defense facility. A more recent report (BJC 2001) states that although MSRE was a civilian reactor demonstration, the radionuclides reported in the wastes are clearly of defense origin. A final determination of the defense-related status of the MSRE salts will need to occur prior to a decision on the ultimate disposal of the salts.

6.1.3 Physical Properties Physical properties regulated by WIPP include the quantity of liquid in the waste and the form of the waste. Liquid waste is not accepted at WIPP and shall be less than 1% by volume of the outermost container at the time of radiography or visual examination (WAC 2010). The physical form of the waste refers to the Summary Category Group of the waste and must be reported to WIPP through the WIPP Waste Information System. Sealed containers that are greater than 4 liters are prohibited except for metal containers packaging inorganic waste.

63 Approved waste forms include solids with particle sizes ranging from fine powder to concrete or debris, and large, bulky dense objects with sharp and obtrusive members or components, such as steel pipe or concrete blocks. The MSRE fuel and flush salts, emplaced after melting or mechanical pulverization would comply with these requirements, as would a portion of a tank and salt that had been cut to fit inside a container. Sharp or heavy objects must be blocked, braced or suitably packaged as necessary to provide puncture protection for the Payload Canister containing these objects. In the case of mechanical cutting of the tanks for removal, additional blocking or bracing would be required for a cut portion of the tank if sharp edges were present. Compliance with this requirement must be verified by one or a combination of the following methods:

Review of records and database information, which may include knowledge of process; Radiography; Visual observation; and, Sampling program.

6.1.4 Chemical Properties Chemical properties regulated by WIPP include hazardous waste classification, pyrophoric classification, chemical compatibility, explosives, corrosives or compressed gases, headspace gas concentrations and polychlorinated biphenyls. Radioactive pyrophoric materials are limited to residual amounts (<1 percent by weight) in payload containers. Radioactive pyrophorics in concentrations 1 percent by weight and all nonradioactive pyrophorics shall be reacted (or oxidized) and rendered nonreactive prior to emplacement in the payload container (DOE 2010b). Hazardous waste, if present, must occur as co-contaminants with TRU waste for acceptance at WIPP. TRU waste with materials chemically incompatible with payload container and packing materials are prohibited. No explosives, corrosives or compressed gases are acceptable. Headspace gas requirements of the payload canisters are specified in the site-specific TRAMPAC or related documents. Verification of adherence to chemical WAC may require sampling during or prior to MSRE salt transfer to transport containers.

6.1.5 Data Package Contents Data package content information includes characterization and certification data and shipping data. Sites are required to prepare a Waste Stream Profile Form (WSPF) for each waste stream. Sites are also required to prepare either a bill of lading or a uniform hazardous waste manifest for RH TRU waste shipments as required by the transportation requirements.

6.2 RH TRU WASTE RADIOLOGICAL CHARACTERIZATION The DOE and WIPP Carlsbad Field Office have identified waste characterization methods for RH TRU waste that meet the program DQOs. The RH TRU Waste Characterization Program Implementation Plan provides the methods and requirements used to characterize the radiological composition of the RH TRU waste (DOE 2010b). The primary method is the compilation and qualification of Acceptable Knowledge (AK) information (DOE 2009b). AK includes, but is not limited to, information about the physical form of the waste, the base material composing the waste, the radiological characteristics of the waste, and the process that generated the waste (DOE 2009b). This may include process knowledge or previous examinations and measurements (DOE 2009b). Confirmation alternatives for radiological properties are:

100% nondestructive assay or Dose-to-curie of waste containers; Destructive assay used to establish activity per unit volume or mass for homogeneous waste; Modeling (e.g., ORIGEN) used to confirm isotopic ratios derived from sampling and analysis; and, Analysis of a representative number of samples to confirm isotopic ratios derived from modeling.

64 It is possible to satisfy the program data quality objectives (DQO) with AK information (with the exception of the defense waste determination). AK information must be qualified by one or a combination of the following four methods:

Peer review, conducted in a manner compatible with NUREG-1297, Peer Review for High-Level Nuclear Waste Repositories, February 1988; Corroborating data; Confirmatory testing; or, Evidence of a QA program that is equivalent in effect to ASME NQA-1-1989 edition, ASME NQA-2a-1990 addenda, part 2.7, of ASME NQA-2-1989 edition, and ASME NQA-3-1989 edition (excluding Section 2.1 (b) and (c) and Section 17.1).

The Waste Characterization Implementation Plan provides the following confirmatory testing methods that could be proposed, but are not limited to:

Qualification of existing VE or radiography audio/videotapes by the review of a percentage of the tapes by qualified operators; Qualification of existing radiological characterization data by analyzing representative samples of the waste; Qualification of existing waste container packaging records by VE or radiography of a representative subpopulation of the waste; and, Qualification of existing radiological sampling and analytical information by the use of confirmatory modeling (e.g., ORIGEN).

The Waste Characterization Implementation also requires generator sites that propose to use confirmatory testing to qualify AK information as characterization data to submit a confirmatory testing plan to the Carlsbad Field Office (CBFO) for review and approval. This plan must include:

A description of the waste stream or waste stream lots to which the plan applies; An explicit description of the waste characterization DQOs and QAOs that will be satisfied with the data being qualified; A description of the DQOs and QAOs that will not be confirmed with the data being qualified and an explanation of how compliance with those DQOs and QAOs will be demonstrated; A description of the confirmatory testing proposed, including the percentage of waste containers that will be subject to confirmatory testing; A description of how the tested subpopulation will be representative of the waste stream or waste stream lot; and Quantitative acceptance criteria for determining that the AK information in question can be qualified as characterization information.

Based on the historical knowledge and extensive study/modeling of the composition and characteristics of MSRE salts, it is possible that direct sampling of the salts would not be required during transfer if AK could be demonstrated in accordance with the site specific plans and the requirements of the CBFO.

However, the potential presence of unexpected or undocumented material such as metal fluorides or tar-like substances as discussed in Section 5.2.1.4 may necessitate additional sampling. Additional discussion of sampling is provided in Section 4.3.

65 6.3 CH CONTAINER GEOMETRIES WIPP will accept a variety of CH containers including 55, 85 and 100 gallon drums, a Standard Waste Box with approximate dimensions of 37 in. x 71 in. x 54 in. and a TDOP. As noted in Section 5, some of the larger CH containers could accommodate a drain or flush tank with a portion of the top section removed. However, CH containers provide little additional radiological shielding, and the addition of shielding sufficient to comply with allowable dose rates for CH containers would result in containers with weights exceeding the stated limits at WIPP. Therefore, the MSRE salts will require remote handling for most practical configurations, based on the anticipated dose rates of the RH-72B Payload Canister (for handling at WIPP) and the RH-72B Outer Vessel (for transportation).

6.4 TRANSPORTATION DOSE RATE Transportation of radioactive material is primarily covered by DOT regulations (49 CFR) and NRC regulations (10 CFR Parts 30 and 71). DOT 49 CFR 173.441 specifies the current allowable transportation dose rate for closed transport, which is 1,000 mrem/hr on the surface of the package, 200 mrem/hr at the surface of the vehicle, 10 mrem/hr at 2 meters from any surface of the vehicle and 2 mrem/hr in the vehicle cab. Transportation dose rates were calculated based on the proposed configuration of the salts within an RH-TRU 72B package as outlined in Section 5.2.2 (Figure 15). Using preliminary output from the MicroShield program, the surface dose rate would be 7.4 mrem/hr with 1/4 inch of lead shielding, 3.75 mrem/hr with 1/2 inch of lead shielding and 1.9 mrem/hr with 3/4 inch of lead shielding. The dose rates for 1/4 inch, 1/2 inch, and 3/4 inch of shielding at 2 meters from the package surface would be 1.2, 0.6 and 0.3 mrem/hr, respectively.

6.5 WIPP HANDLING AND DOSE RATE Figure 20 shows the transport, arrival and disposal process at WIPP for an RH-TRU container, in this case an RH-TRU 72-B shipping container. RH-TRU waste is handled at WIPP with adequate shielding to protect workers. After arrival at WIPP, impact limiters are removed from the ends of the package and the remaining assembly, consisting of the payload canister, inner vessel and outer vessel, is rotated into an upright position. The payload canister is removed from the inner and outer vessels and inserted into a facility cask. The cask is lowered approximately 2,150 ft to the disposal rooms in a 2,000 ft thick, Permian-age salt formation. A 41-ton forklift moves the facility cask to the disposal room, where the payload canister is pushed out of the cask and into a horizontal borehole, which is drilled 4 feet from the floor on 8 foot centers. The borehole is capped with a 70 inch waste shield plug to limit the radiation surface dose rate to the disposal room to less than 5 mrem/hr. CH-TRU waste barrels and boxes are stacked in rows on the floor of the same rooms.

To qualify as RH TRU waste, the current external surface radiation dose equivalent rate of individual containers must be greater than or equal to 200 mrem/hr and less than 1,000 rem/hr. Using preliminary output from the MicroShield program, the predicted surface dose rate was calculated for the payload canister. Assuming 1/4 inch of lead shielding, the surface dose rate would be 82.5 rem/hr. Using 1/2 inch of lead shielding, the dose rate is 56 rem/hr and with 3/4 inch of lead shielding, the dose rate is 38 rem/hr.

Each of these fall within the range of surface dose rates specified for RH TRU waste at WIPP and although not calculated, it is likely the dose rate of an unshielded container would not exceed the upper limit of 1,000 rem/hr.

6.6 PAYLOAD WEIGHT Gross shipping weight of a RH-TRU 72B package is 45,000 pounds, maximum. Component weights for RH-TRU 72B are shown in Table 7 (RH TRU 72B SAR). The weight limit of a loaded canister is

66 8,000 lbs. The estimated weight of a single removable-lid payload canister with 3/4 inch of lead shielding, loaded with fuel and flush salt and getter is nearly 8,000 lbs. Estimated weights of canisters with less lead shielding fall well below the loaded Payload Canister weight limit.

Table 7. RH TRU 72B Package Component Weights Weight Component (lbs)

(kg)

Outer Cask 27,883 12,647 Inner Vessel 4,023 1,825 Impact Limiters 5,094 2,311 Loaded Payload Canister 8,000 3,629 TOTAL 45,000 20,412

67 6.7 RH-TRU PAYLOAD CANISTER CONTENTS RH-TRU waste for disposal at WIPP may be directly loaded into Payload Canisters or packaged in waste containers (e.g., 55-gallon drums or metal cans) in the canister. Table 8 identifies material content forms authorized for transport within the canister. The Disposal Container proposed for the Thermal and Mechanical (needle scaler) Salt removal alternatives would qualify as a Waste Material 3, a metal can containing waste.

Table 8. Payload Canister Contents Waste Material Waste Material Description 1

Direct Load: Solids, any particle size (e.g.,

fine powder or inorganic particulates) 2 Direct Load: Solids, large particle size (e.g.,

sand, concrete or debris) 3 Direct Load: Solids, large objects (e.g.,

Metal cans containing waste) 4 Direct Load: Large, bulky dense objects with sharp and obtrusive members or components with dispersible Form 1 and 2 (e.g., steel plate, Electric motors, steel pipe, or concrete blocks)

68 Figure 20. WIPP Remote Handling Disposal Process.

69

7.

SUMMARY

OF EVALUATION The alternatives considered in this study are assessments of potential approaches for the future disposition of the MSRE salt, but are not necessarily comparable options since they do not all achieve the same objective or have identical outcomes. The conclusions of this study and recommendations as to which alternatives are preferred depend on funding priorities, technical issues, legal restrictions, and other constraints.

The alternatives are listed in Table 9. along with their performance merits, issues and projected cost and schedule. The alternatives are grouped as two primary options: keep salt at MSRE or remove salt from MSRE. Of the keep salt at MSRE alternatives, entombment in place is preferred over no action as it provides additional shielding from the contaminated salt and provides a pathway for the reduction of the nuclear safety categorization (by analysis) for MSRE. The entombment design can be adapted to allow the removal of the entombed tanks at some time in the future for subsequent storage or disposition at an alternate location. Of the remove salt from MSRE alternatives, the thermal removal of salt and disposition at WIPP is preferred because the technical and regulatory issues are significantly less challenging than the other removal alternatives, thermal salt transfer has been successfully demonstrated in the past, albeit in a less technically challenging environment and configuration, and the use of new pumps and piping introduced from the top of the tank would not require pressurization of the tank.

While several options for salt disposition were determined to be viable and their costs have been estimated, the base case (50 Year As-Is) is the recommended as the overall preferred near term option, whether employed for the entire 50 years or not, because it provides a means of making funds available for competing ORR environmental projects whose actions, if not taken quickly, could result in a far greater risk to the environment than the salt. The salt presents low environmental risk because it is stored in closely monitored tanks in a massive reinforced concrete, stainless steel lined subterranean basement, where MSRE has complex programs in place and is staffed by highly trained and qualified personnel that together ensure safe operations, monitoring and responding to changes in parameters associated with the defueled salt, physicochemical treatment of salt tank headspace gases, maintenance of the infrastructure that sustains building habitability and protection of the key electrical systems from the weather, control of contamination, maintaining safe access for all the measurements and inspections needed for planning of decontamination and decommissioning, performing sampling, characterization, packaging, storage, and disposal of radioactive and hazardous waste and responding to emergencies. To the extent that these programs, personnel, equipment, and facilities are maintained, the permanent disposition of the salt can be deferred for years without increasing risk to the environment or creating far greater cleanup costs in the future. Therefore the base case of "Maintain as is" for fifty years is recommended as the best means of making funds available in near term years that could accomplish more critical cleanup work. Once the more critical environmental risks on the ORR have been reduced and the disposal of the salt has been scheduled the discussions of and comparisons between other options in this report will apply. In the development of alternatives for the disposition of the fuel and flush salts at MSRE, two primary options were considered: keep the salt at MSRE or remove the salt from MSRE. Of the keep salt at MSRE alternatives, entombment in place is preferred over no action as it provides additional shielding from the contaminated salt and provides a pathway for the reduction of the nuclear safety categorization (by analysis) for MSRE. The entombment alternative can be designed with the tanks entombed above the water table at the MSRE site and encased in retrievable casks that would facilitate tank removal at some time in the future for subsequent storage or disposition at an alternate location.

Of the remove salt from MSRE alternatives, the thermal removal and mechanical removal of salt with a needle scaler both lead to disposition at WIPP and are preferred because the technical and regulatory issues are significantly less challenging than the other removal alternatives. Thermal salt transfer has been

70 Table 9. Evaluation of Alternatives Option Alternative Performance Merits Issues Project Duration /

Project Cost In Millions (Ref#)

End State After Entire Project Duration Comparative 50-year Total Cost (Maintenance plus Project) in Millions First Year Annual Cost (Maintenance plus Project) in Millions Maintain As-is for 50 years Technically feasible and easy to implement Change in facility maintenance strategy and increase in long term costs MSRE remains a nuclear facility Not a project, maintenance for 50-yr duration (1)

$497 Cost (1) 100 year-old facility has been maintained safe, operable, and habitable

$497

$9.90 Keep salt at MSRE Retrievable Entombment Technically feasible Reduces hydrogeologic and shielding concerns Off-gassing and corrosive effects of HF in an aqueous environment unless getter is added to the salt to address fluorine and tanks are sealed in epoxy grout and lifted above the water table.

62 Month Schedule (2)

$77 Cost (2)

Salt is easily retrievable but must be transported to a hot cell, transferred to disposal containers, and shipped to a final disposal site.

$442

$24.8 Intact-Tank Removal No salt transfer required Reduces tank corrosion concerns Very difficult to package tank and contents for shipment/disposal; therefore, a type-B like container would be used to store the tank and contents at ORNL.

Waste package with shielding would be significantly overweight for off-site shipment/disposal.

63 Month Schedule (3e)

$79 Cost (3e)

Salt is easily retrievable but must be transported to a hot cell, transferred to disposal containers, and shipped to a final disposal site.

$443

$25.0 Thermal Salt Removal Salt stratification and impurities must be addressed Remaining tank and associated residual contamination will remain 71 Month Schedule (4)

$111 Cost (4)

Salt is permanently disposed of at WIPP

$478

$28.6 Mechanical Salt Removal (Needle Scaler)

Meets the removal requirements in ROD Previously demonstrated as feasible Once packaged, shipment and disposal at WIPP is reasonably implementable Removal and packaging is difficult Remaining tank and associated residual contamination will remain 69 Month Schedule (5a)

$101 Cost (5a)

Salt is permanently disposed of at WIPP

$468

$27.5 Remov e salt from MSRE On Site Storage in Type B Container Reduces the nuclear categorization of MSRE Once packaged, shipment and disposal at WIPP is reasonably implementable Requires modification of safety basis at storage location.

62 Month Schedule (6b)

$95 Cost (6b)

Salt is easily retrievable and ready for transport to disposal without transfer

$460

$28.3 Endnotes:

The baseline annual maintenance cost ($9.9 Million) is included in the first year annual cost of all options. The first year annual cost column on the far right is provided for convenience in evaluating competing priorities and selecting projects for funding that achieve the best reduction of environmental risk on the Oak Ridge Reservation. Deferral of all options other than the baseline maintenance will provide roughly 14 - 18 million dollars available for executing projects that are likely to result in greater environmental risk reduction than that achievable hrough permanent disposal or alternative management of the MSRE salt. The comparative 50 year cost is for operating MSRE through and beyond the salt option elected (on to 50 years) and is provided for easy comparison onlyit does not include costs incurred at the facility receiving the salt for applicable cases.

71 successfully demonstrated in the past, and the use of new pumps and piping introduced from the top of the tank does not require pressurization of the tank. Mechanical salt transfer would use tools similar to those developed and used successfully for UDR Project, and would not require pressurization of the tank or heating of the salt. Either the thermal or mechanical needle scaler method will result in an optimal payload configuration for transport to WIPP, and meet all of the dose rate, payload weight and WIPP RH requirements.

All alternatives that involve the removal of salt from MSRE option will allow nuclear facility re-categorization (i.e., downgrading) of the MSRE facility. Disposal of MSRE salt at WIPP will require a determination that the salts are defense related TRU waste to meet the licensing requirements for WIPP.

On-site storage of removed salt/tank at one of the SWSAs in Melton Valley would require modification to the safety basis for the SWSA and possibly re-categorization.

The estimated costs developed for this report are based on direct cost of labor and materials for some of the tasks involved in each specific alternative, but the major cost driver is the cost of labor over the significant length of time required to plan activities, prepare documents, train personnel, and perform the work. For example, planning, design and preparation for fluorination and removal of uranium from the Drain and Flush Salt Tanks and associated tasks began FY1999, and the execution of in the work was performed between 2004 and 2008.

Project durations for these alternatives range from 5.1 years for Alternative 6b to 6.4 years for Alternative 3a, with the majority of the schedule for each alternative concentrating on design, development, training and operational readiness. Executing removal activities for alternatives ranged from 8 weeks for Alternative 6b to 26 weeks for Alternatives 5b and 6c.

72

8. REFERENCES BJC BJC 2010. Bechtel Jacobs Company LLC Quality Assurance Program Plan Oak Ridge, Tennessee, BJC/OR-43/R10, Bechtel Jacobs Company LLC, Oak Ridge, TN.

BJC 2009a. Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment Oak Ridge National Laboratory Oak Ridge, Tennessee, BJC-OR-3301, Bechtel Jacobs Company LLC, Oak Ridge, TN.

BJC 2009b. Addendum to Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the Molten Salt Reactor Experiment Oak Ridge National Laboratory Oak Ridge, Tennessee, DOE/OR/01-2445&D1/A1, Bechtel Jacobs Company LLC, Oak Ridge, TN.

BJC 2009c. Technical Specification for an Engineering Evaluation of the Proposed Alternative Salt Transfer Method for the MSRE at ORNL, SPJ-7503-A749, Bechtel Jacobs Company LLC, Oak Ridge, TN.

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73 BJC/DOE 2002a. Gilliam, B.J., Chapman, J.A., Jugan, M. R. ALARA Controls and the Radiological Lessons Learned During the Uranium Fuel Removal Project at the Molten Salt Reactor Experiment Project, Paper by Bechtel Jacobs Company LLC, WM 2002 Conference, February 24-28, 2002, Tucson, AZ.

BJC/DOE 2002b. Haghighi, Mahmoud H., Mark K. Ford, Robert M. Szozda, Michael R. Jugan. Waste Stream Generated and Waste Disposal Plans for Molten Salt Reactor Experiment at Oak Ridge National Laboratory, Paper by Bechtel Jacobs Company LLC, WM 2002 Conference, February 24-28, 2002, Tucson, AZ.

BJC 2001a. Trabalka, et. al., ORNL TRU Waste Historical Survey, BJC/OR-395-V/2, Bechtel Jacobs Company LLC, Oak Ridge, TN.

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BJC Calculations and Drawings BJC 2010b. Project Calculations: An Evaluation of the Impact of Radiolytic Fluorine on HEPA Media Selection for the MSRE CVS HEPA Filters, CAH-7503-A913, Rev. 0. Bechtel Jacobs Company LLC, Oak Ridge, TN.

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BJC 2003d. Project Calculations: Radionuclide Inventory and Neutron/Gamma Source Term, CAN-02MSRE-A010, Rev. 1, Bechtel Jacobs Company LLC, Oak Ridge, TN.

BJC 2002b. Project Calculations: Inventory of the MSRE fuel and flush salts constituents, CAJ-02MSRE-A014, Rev. 2, Bechtel Jacobs Company LLC, Oak Ridge, TN.

74 BJC 2002c. Project Calculations: MSRE Salt Disposition, Salt Transfer Design Basis Calculation, CAJ-02MSRE-A005, Bechtel Jacobs Company LLC, Oak Ridge, TN.

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Drawing J3E020794-A006. Fluorination Gas Recycle and Cold Trap Process Flow Diagram, Bechtel Jacobs Company LLC, Oak Ridge, TN.

Drawing J3E020794-A007. UF6 Recovery (RGRS) Process Flow Diagram, Bechtel Jacobs Company LLC, Oak Ridge, TN.

Drawing J3E020794-A008. Salt Packaging Process Flow Diagram, Bechtel Jacobs Company LLC, Oak Ridge, TN.

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Drawing J3E020794-A029 Rev. 2. Salt Melter System Detail P&I Diagram, Bechtel Jacobs Company LLC, Oak Ridge, TN Drawing J3E020794-A054 Rev. 1. Fuel Drain Tank System for Maintenance and Surveillance Program Process Flow Sheet, Bechtel Jacobs Company LLC, Oak Ridge, TN.

Drawing P3E020794C149 Rev. 1. Flush Salt Tank Alternate Transfer Line, MSRE Remediation Project, Bechtel Jacobs Company LLC, Oak Ridge, TN, October 2005.

Drawing P3E020794C151, Rev. 1. Fuel Drain Tank 2 Alternate Transfer Line, MSRE Remediation Project, Bechtel Jacobs Company LLC, Oak Ridge, TN, October 2005.

Drawing P3E020794C153 Rev. 1. Fuel Drain Tank 1 Alternate Transfer Line, MSRE Remediation Project, Bechtel Jacobs Company LLC, Oak Ridge, TN, October 2005.

Drawing P3E020794-P001. Fuel Salt Disposition General Arrangement, MSRE Remediation Project, Lockheed Martin Energy Systems, Oak Ridge, TN.

Drawing M20794RF003E10. Fuel Drain Tank 1 Cooling System Assembly (6/30/61)

Drawing M20794RF007D10. Fuel Drain Tank Details (6/28/61).

75 Drawing M20794RF016D2, Fuel Drain Tank Cooling System Assembly, FDT2 (6/30/61).

M20794RF010D11, Fuel Salt System Flush Tank Assembly and Details, D-FFA40462 (12/29/61).

DOE/EPA DOE 2010a. 2010 Remediation Effectiveness Report for the U.S. Department of Energy, Oak Ridge Reservation, Oak Ridge, Tennessee, DOE/OR/01-2256&D1, U.S. Department of Energy Oak Ridge Office, Oak Ridge, TN.

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DOE 2000. Removal Action Work Plan for Uranium Deposit Removal at the Molten Salt Reactor Experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee, DOE/OR/01-1732&D4, U.S. Department of Energy Oak Ridge Office, Oak Ridge, TN.

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DOE 1998. 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, DOE/OR/01-1722&D2, U.S. Department of Energy Oak Ridge Office, Oak Ridge, TN.

DOE 1997. Removal Action Report on the Molten Salt Reactor Experiment Time-Critical Removal Action at Oak Ridge National Laboratory, Oak Ridge, Tennessee, DOE/OR/01-1623&D2, U.S.

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76 DOE 1997. Feasibility Study for Fuel and Flush Salt Removal from the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, DOE/OR/02-1559&D2, U.S.

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EPA 2006. 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, DOE/OR/01-2088&D2, U.S. Environmental Protection Agency Region IV, Atlanta, GA.

http://www.epa.gov/superfund/sites/rods/fulltext/e2007040001897.pdf EPA 2005. Explanation of Significant Differences for the Record of Decision for Interim Actions for the Melton Valley Watershed at Oak Ridge National Laboratory, Oak Ridge, Tennessee Deletion of MSRE Ancillary Facilities from the Selected Remedy, EPA/ROD/R04-98/018 1998, U.S.

Environmental Protection Agency Region IV, Atlanta, GA.

EPA 1998. Record of Decision for Interim Action to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment at the Oak Ridge National Laboratory, Oak Ridge, Tennessee, EPA/ROD/R04-98/018 1998,, U.S. Environmental Protection Agency Region IV, Atlanta, GA.

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Chemical Sciences Division, Radioactive Materials Analysis Laboratory, UT-Battelle, Oak Ridge, TN.

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77 ORNL 1985. Notz, K. J., Extended Storage-In-Place Of MSRE Fuel Salt And Flush Salt, ORNL/TM-9756, Union Carbide Corporation, Oak Ridge, TN.

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ORNL/CF-77-391, Union Carbide Corporation, Oak Ridge, TN.

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ORNL 1972. Koger, J. W., Evaluation Of Hastelloy-N Alloys After Nine Years Exposure To Both A Molten Fluoride Salt And Air At Temperatures From 700 to 560 OC, ORNL/TM-4189, Union Carbide Corporation, Oak Ridge TN.

ORNL 1972. Kedl, R. J. The Migration Of A Class Of Fission Products (Noble Metals) In The Molten-Salt Reactor Program, ORNL/TM-3884, Union Carbide Corporation, Oak Ridge, TN.

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78 Other Crowell, B. S, et al. Accelerator-Based Conversion (ABC) of Weapons Plutonium: Plant Layout Study and Related Design Issues, LA-UR-95-1096, Los Alamos National Laboratory, Los Alamos, NM.

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79 http://www.wipp.energy.gov/library/RHsar/R4/Documents/RH-TRU%20PAYLOAD%20APPENDICES.pdf DOE 2006d. Quality list, PE-006-001 RH-TRU 72-B Quality List, Rev 0, U.S. Department of Energy Carlsbad Field Office, Carlsbad, NM.

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Cost Estimating

References:

DOE 2008. Real Property Asset Management, Order 430.1B, U.S. Department of Energy, Washington, D.C.

DOE 2001. Nuclear Facility Maintenance Management Program Guide for Use with DOE O 433.1, DOE G 433.1-1, U.S. Department of Energy, Washington, D.C.

DOE 1997. Cost Estimating Guide, DOE G 430.1-1, Guide 4301-1 U.S. Department of Energy, Washington, D.C. (1) Chapter 9, Operating Costs (2) Chapter 18, Use of Cost Estimating Relationships (3) Chapter 20, Estimating Specialty Costs (4) Chapter 23, Life-Cycle Cost Estimate PNNL 2004. O&M Best Practices-A Guide to Achieving Operational Efficiency, PNNL-14788, Pacific Northwest National Laboratory, Richland, WA.

NIST 2010. Energy Price Indices and Discount Factors for Life-Cycle Cost Analysis-2010, NISTIR 85-3273-25, U.S. Department of Commerce Technology Administration National Institute of Standards and Technology (Prepared for the U.S. Department of Energy Federal Energy Management Program).

APPENDIX A.

BASIS OF COST ESTIMATES

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A-4

APPENDIX B.

COST ESTIMATE DETAILS

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B-3 INSERT APPENDIX HERE

B-4

APPENDIX C.

SCHEDULES

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APPENDIX D.

ATTACHMENTS

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D-3 LIST OF ATTACHMENTS 1

DOE Salt Study Directive Letter 2

DNFSB Defense Related Waste Letter 3

Origins and Characteristics of Contact-Handled Transuranic Wastes 4

Salt Canister Radionuclide Inventory and Neutron/Gamma Source Term 5

Salt Melt, Transfer, and Process Demonstrations 6

Final Report of the Molten Salt Reactor Experiment Drain Tank Qualification 7

Evaluation of Fluorine-Trapping Agents for Use During Storage of the MSRE Fuel Salt 8

ALARA Controls and the Radiological Lessons Learned during the Uranium Fuel Removal Project 9

Completion Report for the Removal and Transfer of the Uranium 10 Waste Stream Generated and Waste Disposal Plan 11 Coolant Drain Tank Lessons Learned 12 Specification for Fuel Salt Vessels, 2001 13 Specification for Fuel Salt Vessel, 2003

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DOE/OR/01-2496&D0 RECORD COPY DISTRIBUTION FileDMCRC