ML23104A052

Msre Rod

ML23104A052
Person / Time
Site:	05000610
Issue date:	06/30/1998
From:	Jacobs
To:	Office of Nuclear Material Safety and Safeguards, US Dept of Energy, Office of Environmental Management
References
DE-AC05-98OR22700 DOE/ORl02-1671&D2
Download: ML23104A052 (1)

Text

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DOE/ORl02-1671&D2 Record of Decision for Interim Action to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee

DOE/ORl02-1671&D2 Record of Decision for Interim Action to Remove Fuel and Flush Salts from the Molten Salt Reactor Experiment Facility at the Oak Ridge National Laboratory, Oak Ridge, Tennessee Date Issued-June 1998 Prepared by Jacobs EM Team 125 Broadway Avenue Oak Ridge, Tennessee Prepared for the U.S. Department of Energy Office of Environmental Management Environmental Management Activities at the Oak Ridge National Laboratory Oak Ridge, Tennessee 37831 managed by Bechtel Jacobs Company LLC for the U.S. Department of Energy under contract DE-AC05-980R22700

PREFACE This Record of Decisioll for illferim Actioll to ReI/lOve Fuel alld Flush Salts from the Moltell Salt Reactor Experimellf Facility at the Oak Ridge Natiollal Laboratory, Oak Ridge, Tellllessee (DOE/ORf02-1671 &D2) was prepared in accordance with requirements under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980. The U.S . Department of Energy, U .S. Environmental Protection Agency, and the state of Tennessee agree here to select the action for removing fuel and flush salts and placing the salt in a more controlled storage condition until final disposition of the salt is arranged. Work on this task was performed under Work Breakdown Structure 1.4.12.6.2.01 (Activity Data Sheet 3700, "Molten Salt Reactor Experiment D&D Support").

This document presents .a description of the selected remedy, which includes removing flush salt and fuel salt from their respective storage containers in the Molten Salt Reactor Experiment facility, removing uranium from the salts, treating the uranium to form an oxide for safer storage, placing the uranium oxide into storage, containerizing the fuel and flush salts without uranium, and temporarily storing this salt at the Oak Ridge National Laboratory until final disposition of the salt. This document relies on and is consistent with information in the Feasibility SllIdy for Fuel alld Flush Salt Removal from the Moltell Salt Reactor Experimellt at the Oak Ridge Natiollal Laboratory, Oak Ridge, Tellllessee (DOE/ORf02-1559&D2), the illferim Actioll Proposed Plallfor Fuel alld Flush Salt Dispositioll from the Moltell Salt Reactor Experimellf, Oak Ridge Natiollal Laboratory, Oak Ridge, Tellllessee (DOE/ORf02-1601&D3), and Evaluatioll of the U.S. Departmellf of Ellergy's Alternatives for the Removal alld Dispositioll of Moltell Salt Reactor Experimellf Fluoride Salts prepared by the National Research Council in 1997.

ACRONYMS AND ABBREVIATIONS ARAR applicable or relevant and appropriate requirement Be beryllium CERCLA Comprehensive Environmental Response, Compensation, and Liability Act of 1980 Ci curie D&D decontamination and decommissioning DOE U .S. Department of Energy EPA U .S. Environmental Protection Agency FFA Federal Facility Agreement

. FS feasibility study ft foot g gram HF hydrogen nuoride kg kilogram km kilometer lb pound Li lithium m meter MSRE Molten Salt Reactor Experiment NEPA National Environmental Policy Act of 1969 ORNL Oak Ridge National Laboratory ORR Oak Ridge Reservation ppm parts per million ROD record of decision TDEC Tennessee Department of Environment and Conservation TRU transuranic U uranium UF, uranium tetranuoride WIPP Waste Isolation Pilot Plant Zr zirconium ITOO869709.IBH fClE iii June 3, 1998

PART 1. DECLARATION rrOO869109.lflWCJE June 3, 1998

SITE NAME AND LOCATION U.S. Department of Energy Oak Ridge Reservation Molten Salt Reactor Experiment Facility-Building 7503 Molten Salt Reactor Experiment Decontamination and Decommissioning Support Oak Ridge, Tennessee STATEMENT OF BASIS AND PURPOSE This record of decision (ROD) presents the selected interim remedial action for addressing fuel and flush fluoride salts from three drain tanks formerly used as part of the Molten Salt Reactor Experiment (MSRE). The tanks are located in the MSRE facility (Building 7503) at the Oak Ridge National Laboratory (ORNL) on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR) . Remediating the MSRE facility is a high priority because of the unacceptable risk associated with the highly radioactive salt stored in the drain tanks. The location, condition, and age of the equipment connected to the tanks and the chemistry of the salt make control of safety factors difficult. The objective of this interim action is to reduce potential on- and off-site risk from the salt.

This interim action was chosen in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA), as amended by the Superfund Amendments and Reauthorization Act of 1986 (42 United States Code, Sect. 9601 et seq.) and, to the extent practicable, the National Oil and Hazardous Substances Pollution Contingency Plan (40 Code of Federal Regulations 300). The ROD is based on the Administrative Record for this site.

DOE issues this document as the lead agency . The U.S. Envirorunental Protection Agency (EPA) and the Tennessee Department of Environment and Conservation (TDEC) are support agencies as parties to the Federal Facility Agreement (FFA) for this response action.

DOE and EPA have jointly selected the remedy for the MSRE fuel and flush salts removal.

TDEC concurs with the selected remedy.

rrOO869709.IDHlCJE 1-3 June 3, 1998

ASSESSMENT OF THE STUDY AREA/OPERABLE UNIT A streamlined risk assessment was conducted to determine whether current or future remedial actions are necessary to protect human health and the envirorunent if current institutional controls are removed. The scenarios considered include on- and off-site receptors. The risk assessment demonstrates that without institutional controls the salts in the MSRE drain tanks pose an unacceptable risk to human health and the environment now and in the future. Thus a response action is required to address the salt stored in the three drain tanks at the MSRE facility.

The objective of this interim action is to reduce current potential on- and off-site risk from the salts. pending final action.

Actual or threatened releases of hazardous substances from the MSRE facility that are not addressed by implementing the response action selected in this ROD may present an unacceptable risk to public health. welfare. and the envirorunent.

DESCRIPTION OF SELECTED REMEDY The selected interim remedial action includes melting and chemically treating the salt in the drain tank cell. separating the uranium from the salts. transferring the uranium to the 2llU repository at ORNL. packaging the residual salt. and placing the salt in interim storage at ORNL until arrangements are made for final disposition. Specific details and methods for this interim remedial action will be included in the remedial design and remedial action plans. As the salt melts in a drain tank. the molten salt will be treated with hydrogen fluoride (HF) to balance salt chemistry. The uranium in the salts will then be removed from the salt and converted to an oxide that is chemically stable and compatible with long-term storage at the 2llU repository at ORNL Building 3019 and managed as a part of the existing 2llU repository inventory. The residual salt will be stabilized/packaged to control fluorine gas generation and the containers placed in interim storage. The location of interim storage will be at an existing storage facility at ORNL.

Placement of the salt for its final disposition will be documented in a subsequent ulm! CERCLA decision document ~nq!;lm ~!RRr9Pfl~\,~;;;i~ National Envirorunental Policy Act of 1969 (NEPA) decision document. These future decisions will incorporate full public participation and will be based on the existing feasibility study (FS).

fTOO869709 .1BHlCJE 1-4 June 3, 1998

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After removal of salts from the MSRE drain tanks , the tanks and associated equipment will be managed in place as part of the facility maintenance program. The storage tanks and reactor components will be addressed as part of a subsequent decontamination and decommissioning (0&0) action of the building .

ST ATUTORY DETERMINATIONS This interim action protects human health and the environment, complies with federal and state requirements that are legally applicable or relevant and appropriate requirements (ARARs),

and is cost-effective. Within its limited scope, this interim action uses permanent solutions and alternative treatment technologies to the maximum extent practicable by removing the salts from the MSRE drain tanks, treating the salts to remove the uranium, and stabilizing/packaging the sal ts for final d ispos it ion . ill!J~i~(9t~i Ilh$ i~~!§~I1@1jj'!~t\inl:f§Hj~qX§9JJ~fl~§;1In~ln~m\2rx Rr~(sr~hg~j;f§,tl E~m~gl~~I;~~mR!8X!Q~L!tl*!ffi~Rj§;;!Mtj;f~g!lg~ i!§81§lI¥;!Im§R1!J~imigEIyg!llmaj(&' 2

!ll:ilIs!R!!i~!~mslI!i Disposal and, if necessary, further treatment of MSRE salts after the uranium has been removed will be performed as part of another action. This interim action addresses the principal threat from criticality or release of contaminants into the environment posed by the salts stored in the MSRE drain tanks . Removal of radioactive salts will permit the remaining structures to be included in a later action. Because this is an interim action ROD , review of this facility will continue as DOE develops final remedial alternatives for 0&0 of Building 7503 .

rrOO869709. I BUfCJE 1-5 June 3, 1998

APPROVALS Rodney R. Nelson, Assistant Manager Date V.S. Department of Energy Oak Ridge Operations I '

Date Richard D. Green, Director Date Waste Management Division V.S . Envirorunental Protection Agency-Region 4 JTOO&69709.1 all/CJE 1-6 JMe 3, 1998

PART 2. DECISION

SUMMARY

rrOO869709.IBWCJE June 3, 1998

SITE NAME AND LOCATION The MSRE site is located in Roane County, Tennessee, on the DOE ORR approximately I km (0.6 miles) south of the ORNL main plant across Haw Ridge in Melton Valley. The ORNL main plant is approximately ' 24 km (15 miles) west of Knoxville, Tennessee, and 16 km (10 miles) southwest of the Oak Ridge, Tennessee, business center (Fig. 2.1).

The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The MSRE site with Building 7503 and other support buildings are located at the intersection of Melton Valley Road and High Flux Isotope Reactor Access Road (Fig. 2.2).

SITE DESCRIPTION, HISTORY, AND ENFORCEMENT ACTIVITIES Building 7503 was constructed in 1951 to contain the Aircraft Reactor Experiment and expanded in 1955 for the Aircraft Reactor Test, which was canceled in September 1957. In 1961, experimentation on a molten salt reactor was revived at MSRE to develop a conullercial molten salt breeder reactor. Adjacent buildings supported the MSRE operation. The reactor, using 2J5U as fuel, achieved criticality on June i, 1965. In August 1968, the "'u fuel was replaced with mU. The reactor operation permanently shut down December 12, 1969.

The MSRE reactor loop consisted of a reactor vessel, primary heat exchanger, pump ,

associated piping, and an off-gas system (Fig. 2.3). During operation, the fluoride salt mixture containing uranium fuel was heated to a liquid state. The molten salt was transferred from the fuel drain tanks into the reactor circuit and criticality would occur in the reactor vessel. Fuel salt, further heated by the nuclear reaction, exited the reactor vessel to the heat exchanger to

. transfer excess heat to a secondary fluoride coolant salt. When the reactor was shut down, fuel salt was removed from the reactor circuit by allowing it to drain by gravity back into the fuel drain tanks. To remove residual fuel salt from the reactor circuit , molten flush salt was circulated through the reactor circuit and returned to the flush salt drain tank. At the time operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks.

The fluoride salt used for the fuel and flush salts in MSRE is generally similar except for the uranium fuel and other radionuclide content differences . After shutdown, the fluoride fuel salt and possibly the flush salt released fluorine and uranium hexafluoride gases into the drain JTOO869709.IDIlfCJE 2-3 June: 3, 1998

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MODIF IED FROM, DOE 1993.

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~II Fig. 2.1 DOE Oak Ridge Reservation and vicinity DOE ORNL. Mollen San Reactor Ellperiment Oak Rodge, TCnllcSScc DOCUMENT 10: 3SH330 00116 * .& 0 / MSRE DRAWING 10, 97,2311.DWG DRAWING DATE ,

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RADIATOR COOLANT DRAIN TANK

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;:7' 3. FUEL PUMP 9. FANS

4. FREEZE FlANGE 10. DRAIN TANKS

5. THERMAL SHIELD 11. FLUSH TANK S. COOLANT PUMP 12. CONTAINMENT VESSEL

13. FREEZE VALVE MODIFIED FROM: ORNL DWG 63*1209R mo Fig. 2.3 Simplified MSRE flow diagram of !'rimary and secondary reacfor circUits DOE - ORNl, Molten Salt Reactor Experiment

• Oak Ridge. Tennessee DOCUMENT 10: 35H830 008S-SO/ROO I ORAWING to; 97-154n.CDR DRAWING CATE:

FEBRUARY 10, 1998 S8

tank head spaces and associated off-gas system. Fluorine generation was expected based on knowledge about the chemical stability of fluoride salt. An annealing process was part of shut-down procedures between 1971 and 1989. This process heated fuel salt to below melting temperatures to force the fluorine in the salt matrix to recombine before it would migrate from the salt. It appears that during the annealing process, unknown to operators, uranium hexafluoride gas was formed and liberated from the salt.

In 1994, investigation of the MSRE site indicated that anomalous levels of uranium hexafluoride and fluorine gases were present throughout the off-gas piping connected to the fuel and flush salt drain tanks. In addition, uranium had migrated through the off-gas system to an auxiliary charcoal bed that resulted in a criticality concern because of the quantity of uranium detected. Interim corrective measures were immediately taken to ensure the safety of workers and personnel. Shortly afterwards, documentation of actions taken and continuing actions were included in a CERCLA time-critical removal action memorandum. A plan was then developed for remediating the MSRE site to reduce the risk presented by the continuing presence of the fuel and flush salts in storage at MSRE. Planners organized mitigation of the migrated MSRE uranium (as uranium hexafluoride) and fluorine gas into three separate CERCLA actions.

Time-Critical Removal Action. This CERCLA action, approved in July 1995 (DOE 1995), is completed. The interim corrective measures provided risk reduction for employees and workers at MSRE by addressing various aspects of containnlent, nuclear criticality control, and chemical reaction prevention. A reactive gas removal system, installed in 1996 as part of the time-critical action, continues to remove and trap uranium hexafluoride and fluorine gases from MSRE off-gas piping.

Non-Time-Critical Removal Action. Removal of the uranium deposit and associated fluorine contaminated charcoal from the auxiliary charcoal bed was approved as a CERCLA non-time-critical removal action (DOE 1996). Removal of uranium and fluorine contaminated charcoal is planned for completion in February 1999. This action will eliminate the potential of a criticality accident or chemical reaction in the charcoal bed cell and reduce the risk to human health and environment from exposure to the toxic and radioactive uranium.

Remedial Action. This ROD for interim action focuses on removal of fuel and flush salts from the MSRE drain tanks to eliminate the major source of contaminants for the MSRE site.

Potential sources of uranium hexafluoride and fluorine gases will be eliminated from the drain tanks thereby reducing the risk to workers, employees, and the public. Contaminants that remain at the MSRE site following this interim action and their associated risks will be addressed in a rrOO869709. IBHfCJE 2-8 June 3, 1998

subsequent CERCLA action . The fuel and flush salts from MSRE will be treated to reduce risks during storage while awaiting shipment for final disposition.

HIGHLIGHTS OF COMMUNITY PARTICIPATION The interim action proposed plan for the MSRE site was released to the public in December 1997. This document is part of the Administrative Record for this decontamination and decommission action, which is maintained at the DOE Information Resource Center, 105 Broadway Avenue, Oak Ridge, Tennessee 37830. Notice of availability for this plan and

. other documents in the Administrative Record was published in The Knoxville News*SellIinei December 22, 1997, The Oak RidgeI' December 22, 1997, The Roane COllllty News December 24, '1997, and T71e Ciilllon Courier*News December 24, 1997. The public comment period was held between December 23, 1997, and January 30, 1998. A public meeting held January 14, 1998, to discuss the proposed plan resulted in verbal comments. Two written conUllents were received during the public comment period. Responses to the written conunents and verbal comments from the public meeting relating to this interim action are presented in Part 3, "Responsiveness Summary," of this document.

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SCOPE AND ROLE OF THE SITE INTERIM REMEDIAL ACTION The scope of this interim remedial action is to remove the fuel and flush salts from the drain tanks, separate the uranium from the fuel and flush salts, convert the uranium to an oxide for storage as part of the existing 2JlU repository inventory, stabilize/package the residual salt, and place the residual salt in interim storage until an end*point location is selected for final disposal. This interim action will eliminate the risk of a criticality incident and the hazards rrOO869709. IBHlCJE 2*9 June 3, 1998

associated with uranium hexafluoride and fluorine gas release at the MSRE site. Decontamination and demolition of Building 7503 and the MSRE reactor components will be performed as part of a later, separate CERCLA final action. Ongoing management and final disposition of the uranium oxide will be determined pursuant to the program for managing the existing 2JJU repository inventory (rather than further CERCLA action).

SUMMARY

OF SITE CHARACTERISTICS This remedial action addresses the two contaminated waste salts at the MSRE site-fuel salt and flush salt. The fuel and flush salts are stored in tanks in the drain tank cell below the floor of Building 7503. The fuel salt is divided between two drain tanks, and the flush salt is stored in one flush drain tank. All three tanks are similarly constructed; however, the fuel drain tanks are equipped with steam domes and thimbles to remove heat produced by radioactive decay.

Heat production within the fuel salt is no longer a concern .

Both salts are composed of Li, Be, and Zr fluoride salts. The fuel and flush salts differ in the amount of fuel and fission products contained in each, and the fuel salts have a higher percentage of zirconium. The flush salt contains a small amount of the fuel and fission products because it was used to flush residual fuel salt out of the reactor and 'the associated piping system after the fuel salt was drained into the storage drain tanks. It is estimated that the flush salts contain approximately 500 g (1.1 Ib) or 2.9 Ci of uranium and 13 g << 0.1 Ib) or I Ci of plutonium. Figure 2.4 describes the proportions of salts constituents at the end of reactor operation. Table 2.1 lists the salt weight, volume, and density, and Table 2.2 lists the principal isotopes in the salts after irradiation in the reactor. The mass of uranium in the fuel and flush salts shown in Table 2 .2 [approximately 37.5 kg (82Ib)] represents the amount of uranium

[1.1 percent of the fluoride salts as uranium tetrafluoride (UF,)] that was transferred to the drain tanks at the end of reactor operation. Since reactor shutdown, uranium has migrated from the fuel salt to the drain tank head space, off-gas system, and an auxiliary charcoal bed in the form of uranium hexafluoride. The current mass of uranium in the fuel salts is calculated to be approximately 20 kg (44 Ib) (0.6 percent of the fluoride salts as UF,) .

Fluorine liberation from the salts has left metallic Li, Be, and Zr in the salt and created a net reducing condition in the salt. As a result the potential exists for uranium to precipitate during the melting process. The present reducing potential of the stored salt is latent because the metal is essentially immobile; however, once the salt is heated to melting temperatures, the reduction reaction may proceed . During melting, the reducing potential could cause up to 12 kg (26 Ib) of uranium metal to precipitate and/or diffuse into the tank wall. This could result in a rrOO869709.1BIIICJE 2-10 June 3, 1998

Fuel Drain Tanks Fuel Flush Tank

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Fluoride Salls Fluoride SaUs lll 'UF (51.3%)

D 'UF (42.3%)

/I BeF, (47.8%)

rJl BeF, (36.0%)

. ZrF, (0.89%)

. ZrF, (20.6%)

DUF, (0.015%)

OUF, (1.1%) and PuF, (0.0004%)

PuF, (0.02%)

UranIum D"'U (83.9%)

II"' U (7.5%) e mu (39.4%)

. "'U (2.6%) .. '"U (3.6%)

DU (5.9%) . '"U (17.4%)

• "'U(0.02%) D'"U (39.4%)

and "'u (0.1%) II "'U (0.008%)

and .,u (0.2%)

Plulonlum IlI m pu (94.7%)

Plulonlum o mpu (90.1%) ""'PU (4.80%)

. Olher Pu (0.500%)

""'Pu (9.50%)

. Olher Pu (0.350%)

~JJ Fig. 2.4 Nole: Does nollnclude the -2.7kg of fission products.

DOCUMENT 11},3SH8lO DRA'.',um 10: ORAW IIIO OATE:

008&SOJROD 91*1S473.COR fEBRUARVI1 , IS?8SR 2-11

Table 2.1. Primary inventory of stored fuel and flush salts, MSRE site, ORNL, Oak Ridge, Tennessee Fuel Drain Tank 1 2,479 1.0 44 Fuel Drain Tank 2 2.172 0.9 39 2.48 Total fuel salt in drain tanks 4.650 1.9 NA All three tanks in the DTC NA Source: Table 3 of Williams, D. r ., 0: D. Del Cui , and L. M. Toth. 1996. A Descriptive Model oflhe 'dollell Salt Reactor Experiment After Shutdoll'n: Rel'iew of FY 1995 Progress. ORNLlTM-13142. Oak Ridge National Laboratory, Chemical Technology Division, Oak Ridge, TN., and Table I of ORNL. 1993 . Request for Nuclear Safety Rel'jew and Approm/, MSRE FlIel alld Fillsh Salt Storage , Committee NSR No. 0039WMOOOI3A. Oak Ridge, TN . The weight and volume estimates shown are those Ihal best correspond to process hi story. ORNL (1993) provides a range of weights for the fuel and nush salls. the minimum of which correspond s to the weights in the above table. The maximum weight for the fuel sa lt is < 5 percent higher than the minimum; the maximum for the flush sa lt is

< I percent higher.

"See Table D.2 of U.S. DepaT1ment of Energy. 1997b. Feasibil/r), SlUdy/or Fuel and Flush Satt Relllomtfrom Ihe Molten Satt Reactor

£tperimelU allhe Oak Ridge National LaboraroT)', Oak Ridge, Tel/nessee. DQE/ORl02* 1559&D2 . Oak Ridge, TN.

"See also Table 8.1 of Thoma. R. E. 1971 . Moltell SaIl Reactor Program: Chemical Aspects of MSRE Opuatiolls, ORNL-4658, UC*80*Reactor Technology. Oak Ridge, TN.

°C = degrees Celsius m = mete r cm = centimeter MSRE = Molten Salt Reactor Experiment DTC = drain tank cell NA = not applicable g = gram ORNL = Oak Ridge National Laboratory kg = kilogram  % = percent

< = less than nuclear criticality and Ihe inability to remove the uranium from the drain tanks. The presence of zirconium in the salts may lessen the amount of uranium that is .reduced. To prevent the uranium from precipitating and/or diffusing inlo Ihe tank walls , the previously liberated fluorine will be replaced by bubbling HF Ihrough the salt during a gradual melting of the salt.

fT00869709 .! BIIICJE 2-13 June 3. 1998

Table 2.2. Activity of principal isotopes in the fuel and flush salts, MSRE site, ORNL, Oak Ridge, Tennessee

~

g,

~

1j 38 Strontium 90 28.5 years 7.550 81 Thallium 208 3.05 m 50 39 Ynrium 90 2.7 days 7.550 82 l<:ad 209 3.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 0.7 40 Zirconium 93 1.5 E6 years 0.3 212 10.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 139 43 Technetium 99 2.1 E5 years 0.5 83 Bismuth 212 1.01 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 139 51 Antimony 125 2.73 years 1.0 213 45 .6m 0.7 52 Tellurium 125 58 days 0.3 84 Polonium 212 4S seconds 89.1 55 Cesium 137 30 years 6.290 213 4"" 0.7 56 Barium 137m 2.6m 5.940 216 150 ms 139 61 Promethium 147 2.62 years 50.3 85 Astatine 217 32 ms 0.7 62 Samarium 151 90 years 121 86 Radon 220 55.6 seconds 139 63 Europium 152 13.3 years 1.5 87 Fn.ncium 221 4.9 m 0.7 154 8.8 years 4.7 88 Radium 224 3.66 days 139 N

155 4.96 years 9 .3 225 14.8 days 0.7 89 Actinium 225 10 days 0.7 90 Thorium 228 1.9 days 139 229 7,300 years 0.7 92 Uranium 232 70 years 135 94 Plutonium 238 87.7 years 0.92 233 1.59 E5 years 302 239 24,110 years 41.7 234 2.45 E5 years 17.4 240 6.540 years 15.3 24lh 14.4 years 270 95 Americium 241 433 years 21.5 Total for uranium isotopes (37.548 21 I 454.4 Total for transuranics (737 2l I 349.4

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Table 2.2. (continued)

I'" Souru: Table 6 of Williams, D. F .* G. D. Del Cui , and L. M. Toth . 1996. A D~scriplive Model a/the Mo{un Salt Reactor Expen'ment After Shutdown: Revi(W of FY ]995 Progress.ORNUTM*13142. Oak Ridge National Laboratory. Chemical Technology Div ision. Oak Ridge, TN . The principal isotopes listed are those wilh a current activity > 0.1 Ci. The total activity and weight for each isotope grouping includes other iSQ(opes not listed here .

6en "Uranium and plutonium inventory values (except l.l~U) are derived from isotopic analysis and are 3 [05 percent lower than those calculated by Bell. M. J. 1970. CalcuUJted Radioactivity a/1M Molten Salt Reactor Experiment Fuel Salt. ORNUTM-2970. Oak Ridge National Laboratory . Oak Ridge, TN . All other projections are derived from the Bell discharge inventory .

I'Plutonium-241 is not a TRU waste element because its half-life is < 20 years.

Ci = curie ms = millisecond g = gram MSRE = Molten Salt Reactor Experiment

> = greater than no. = number

< = less than ORNL = Oak Ridge National Laboratory m = meter TRU = transuranic

$iS = microsecond U = uranium tv V>

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SUMMARY

OF SITE RISKS Analysis shows that actual or threatened releases of hazardous substances from this site, if not addressed by the preferred alternative or another active measure, present a current or potential threat to public health, welfare, or the environment.

HUMAN HEALTH RISK The streamlined risk assessment for the MSRE site evaluated two scenarios. A near-term scenario postulates an exposure that could occur in the next 100 years after institutional controls are lost. The other scenario postulates an exposure that could occur beyond 100 years. Included on the risk assessment are only contaminants of potential concern with a credible exposure pathway and long enough half-life to cause significant exposure if released. For the near-term scenario, a release to the environment (air) from a failure in the off-gas piping connected to the drain tanks was postulated. Contaminants of potential concern evaluated for this scenario included fluorine gas, uranium hexafluoride gas, and HF gas. For the second scenario, a criticality event was assumed to occur because of a t~iiure in the drain tank cell and drain tanks.

Contaminants of potential concern were postulated as being fission-product gases generated by a criticality event. Both scenarios evaluated the consequences to:

• an on-site receptor 100 m (328 ft) from the MSRE site and

• an off-site receptor 1,200 m (3,900 ft), the distance to the nearest public road, from the MSRE site.

The exposure pathways quantified in this assessment were based on the conceptual site model. The pathways included (I) a release of fluorine, uranium hexafluoride, and HF gases because of an off-gas piping failure, which results in passerby exposure through the inhalation and immersion pathways (near-term scenario) and (2) a criticality accident caused by a failure of the drain tank cell and drain tanks resulting in passerby exposure from inhalation and immersion in a cloud of radioactive gas (long-term scenario). No other exposure pathways were evaluated.

Based on EPA guidance for streamlined risk assessments, there is no need to evaluate all pathways when risk is clearly exceeded by one exposure pathway.

The streamlined risk assessment showed that most of the estimated risks were above the I x 10-1 limit and were therefore unacceptable. For the near-term scenario, estimated risk for IT00869709.IBH/CJE 2-16 June J, 1998

the on-site receptor is 5 X 10" and ranges from 3 x 10'3 to 2 X 10'2 for the off-site receptor.

For the long-term scenario, the estimated risk for the criticality pathway is 1 X 10'2 for the on-site receptor and 3 x 10" for the off-site receptor.

ECOLOGICAL RISK The ecological risk assessment evaluated the potential for adverse effects on the environment from exposure to contaminants in the MSRE drain tank cell . In the future, a potential breach in a drain tank and a failure of the drain tank cell could contaminate groundwater and surface water at nearby unnamed tributaries to White Oak Creek. The contaminated groundwater would adversely affect terrestrial plants and wildlife. Thus failure of the fuel flush tank or fuel drain tanks and the drain tank cell would adversely impact terrestrial plants and wildlife. This scenario would also pose a risk to aquatic communities in nearby tributaries.

Aquatic receptors could be directly exposed by contact with and ingestion of contaminated water and sediment. Terrestrial wildlife could also ingest contaminated surface water . Terrestrial flora could be exposed to contaminated groundwater through root uptake.

DESCRIPTION OF ALTERNATIVES An interim action alternative to reduce the risk posed by the fuel and flush salts at t\1e MSRE facility was developed and presented in the interim action proposed plan (DOE 1997a).

Use of this interim action will result in (1) reducing the risk at the MSRE facility and (2) completing an action that is common in the alternatives that consider the ultimate disposition of the salt for disposal.

The alternatives developed in the FS were prepared for an action that ideally would be carried to completion with no delays. However, the locations identified in each alternative for final salt disposition are currently not operational. Decisions about waste acceptance cannot be made until locations for salt disposition are operational. As a result, none of the alternatives developed in the FS can be fully implemented at this time. Selection of a disposal location for MSRE salts must wait until one or both of the disposal facilities are opened and questions about the acceptance of MSRE salts for disposal can be evaluated. In the interim, fuel and flush salts will be removed from the MSRE facility. Uranium will also be removed from the salts and managed as part of the existing "'u repository at ORNL. The salt remaining after the uranium removal process will be stored until it is shipped to a disposal location .

JT00869709. J DIIICJE 2-17 JUlIe 3, 1998

Five alternatives were developed in the FS to remove and dispose of the fuel and flush salts (DOE 1997b). The alternatives consisted of a no further action alternative and four action alternatives. The alternatives as presented in the FS are:

• Alternative I: No Further Action,

• Alternative 2: Disposal at Waste Isolation Pilot Plant as Transuranic Waste,

• Alternative 3: Disposal at the National Repository as Spent Nuclear Fuel,

.

• Alternative 4: Disposal at the National Repository as High-Level Nuclear Waste, and

• Alternative 5: Disposal at a Combination of Sites as High-Level Nuclear Waste and Low-Level Nuclear Waste.

The no further action alternative was evaluated as not meeting the purpose and the objectives of this remedial action and therefore was not considered further . The four action alternatives (Alternatives 2-5) each began by removing the salts from the MSRE facility and then taking the actions necessary to transfer the salts to the designated end point for disposal. The end-point locations for disposal of the salts or components of the salts are either the Waste Isolation Pilot Plant (WIPP) in New Mexico as a defense-related transuranic (TRU) waste or a national repository as either spent nuclear fuel or high-level nuclear waste. A decision now to select a location for disposal of the MSRE salts could not be made with certainty that waste acceptance criteria would be met. Evaluation and selection of a location for disposal of the MSRE salt will be documented subsequently when an end-point location for disposal of the salt is identified.

Another consideration for the MSRE site interim remedial action to remove salt from the fuel and flush salt drain tanks is that removal can be completed without precluding the ultimate disposal options. As indicated in each action alternative, removal of the fuel and flush salt from the storage cell drain tanks is the first activity necessary for ultimate dispos~1 of the salt. This remedial action will include the salt in all three drain tanks, starting with the flush salt drain tank which contains less radionuclides than either of the fuel salt drain tanks. Melting the salt in a drain tank will start with a small volume and increase slowly until all the salt is molten. To chemically rebalance the salt, HF will be introduced into the molten salt as it melts. Uranium will be separated from the molten salt using to the extent possible the same process and equipment used to remove 2"U in 1968. Fluorine gas will be added to the molten salt to oxidize UF, into uranium hexafluoride gas which will be trapped as it passes through vertical columns 1TOO869709. 1BH/CJE 2-18 June 3, 1998

packed with sodium fluoride. The salt with the uranium removed will be moved from the drain tanks into storage containers. The salt, which still contains a large quantity of radionuclides , will then be stabilized/packaged to capture fluorine gas which may be generated. (The waste containers will be placed in shielded casks for interim storage.) The casks will be set in an existing storage facility at ORNL and managed there until final disposition is arranged.

INTERIM ACTION ALTERNATIVE The MSRE interim remedial action activities are consistent with the FS salt disposal alternatives . This action reduces risk and at the same time proceeds toward the end point of fuel and flush salts disposal. Implementation of this interim action will not preclude any of the four action alternatives from future consideration.

The ARARs developed in the FS have been reviewed and those pertinent to the interim action are identified and presented in Tables 2 .3 and 2.4.

SUMMARY

OF COMPARATIVE ANALYSIS OF ALTERNATIVES Implementation of the interim action would address the identified risks associated with current conditions at the MSRE site. By separating uraniunl from the fuel and flush salts, converting it to an oxide, packaging it in criticality-safe containers, and storing ii in a facility designed for the storage of 2JJU, risks associated with the release of uranium hexafluoride are eliminated and risks of a nuclear criticality are managed in accordance with applicable standards .

By stabilizing/packaging the residual salt, fluorine gas generation can also be managed. This action would allow DOE to defer decisions regarding further treatment and disposal of the salt to a later date.

The comparative analysis using the nine CERCLA criteria for this interim remedial action includes the no further action alternative and the interim action . Table 2.5 sununarizes the evaluation of the no further action alternative and this interim action (i.e., removal of salt, separation of uranium, and interim storage of salt).

JTOO869709.IBHlCJ E 2-19 June J, 1998

Table 2.3. ARARs for proposed activities, MSRE site, ORNL, Oak Ridge, Tennessee I'"

(j m

Alteration/destruction of Action(s) that will affect such resources must adhere to the DOE-ORO Any action that will impact historic National Historic Preservation historic resources Memorandum of Agreement (May 6, 1994). When alteration or resources-applicable if there will Act of 1966 (16 USC 457a-w);

desnuction of the resource is unavoidable, steps must be taken [Q be alteration or modification Executive Order 11593; minimize or mitigate the impacts and to preserve data and records of 36 CFR 800; the resource Release of radionucIides DOE will carry out all DOE activities to ensure that radiation dose to Release of radionuclides into the DOE Order 5400 .5(1.4) during removal and storage individuals will be ALARA environment-TBC (proposed as 10 CFR 834) activities Exposures to members of the public from all radiation sources shall not DOE Order 54OO.5(ll.1 a) cause an EOE to be > 100 mrem (1 rnSv)/year (proposed as 10 CFR 834)

N Management of TRU waste shall be conducted in such a manner as to Handling and management of TRU 40 CFR 191.03(b)

~ provide reasonable assurance that the combined annual dose equivalent waste-relevant and appropriateH,h to any member of the public in the general environment resulting from discharges of raaionuclide material and direct radiation from such management shall nO( exceed 25 mrem/year to the whole body and 75 mremJyear to any critical organ Exposures to members of the public from all radiation sources released Point source discharge of 40 CFR 61.92; into the atmosphere shall not cause an EDE to be > 10 mrem radionucIides int~ the air from a Rules of the TDEC 1200-3 (0.1 mSv)/year DOE facility-applicable .08 Radiological emission measurements must be perfonned at all release 40 CFR 61.93; points that have a potential to discharge radionuclides into the air in Rules of me TDEC 1200-3 quantities which could cause an EDE in excess of I % of the standard .08 (0.1 mremlyear). All radionucIides which could contribU(e > 10% of the standard (1 mremlyear) for me release point shall be measured

,..~

~

Table 2.3. (continued)

I Q

'"m (j

Characterization of TRU TRU waste must be evaluated to determine the kinds and quantities of Generation of TRU waste-TBC DOE Order 5820.2A (IIl.3b) waste TRU radionuclides pres!!nt before storage Radionuclide-contaminared External exposures [0 [he waste and concentrations of radioactive Storage of uranium after separation DOE Order 5820.2A (11.3.)

material: on-site storage material which may be released into the environment must not exceed from sah-TBC an EDE of 25 mremlyear 10 any member of the public Temporary storage of fuel! TRU waste shall be segregated or clearly identified 10 avoid Temporary storage of TRU wastes DOE Order 5820.2A (II.3.e) nush salts as a TRU waste commingling of the waste with high-level. low-level waste or other at generating sites-TBC pending disposal noncertified TRU waste TRU waste storage areas must be protected from unauthorized access TRU waste must be monitored periodically to ensure:: that wastes are not releasing their radioactive constiruents TRU waste storage areas must be designed. constructed. maintained.

N, N and o~rated with a contingency plan to minimize the possibility of fire. explosion. or accidental release of radioactive components TRU waste storage areas must be operated in a way to maintain radiation exposures to ALARA Management of TRU waste shall be conducted in such a manner as to Handling and management of TRU 40 CFR 191.03(b) provide reasonable assurance that the combined annual dose equivalent waste-relevant and resulting from discharges of radionuclide material and direct radiation appropriate-**

from such management shall not exceed 2S mremlyear to the whole body and 75 mremlyear to any critical organ Interim storage/disposal of Compliance with the peninent WAC for the storage facility Storage/disposal of LLW-TBC DOE Order 5820.2A (1II.3.e)

LLW generated from the separation process (i.e .* PPE. wipes, contaminated hardware)

,.,f

~

Table 2.3. (continued) i

~

"10 CFR 834.109 (proposed rule) requires that management of radioactive waste not exceed an EDE of 25 mTem/year from all exposure pathways. When promulgated. this rule will be legally applicable.

"DOE Order 5400.5, Chapter 1I1(c)(I), requires that TRU waste management and storage activities at facilities other than disposal facilities not cause members of the public to a'" receive. in a year. a dose equivalent> 25 mrem to me whole body or a commined dose equivalent> 75 mrem to any organ.

'" ALARA = as low as reasonably achievable mSv = millisieven ARAR = applicable or relevant and appropriate: requirement ORNL = Oak Ridge National Laboratory CFR = Code of Federal Regulations ORO = Oak Ridge Operations DOE = U.S. Department of Energy  % = percent EDE = effective dose equivalent PPE = personal protective equipment

> = greater than TBC = to be considered

< = less than TDEC = Tennessee Depamnent of Environment and Conservation LLW = low~ level (radioactive) waste TRU = transuranic mrem = millirem USC = Unired Srates Cod~

MSRE = Molten Salt Reactor Experiment WAC = waste acceptance criteria IV IV IV

,..~

~

Table 2.4. Evaluation of the no further and preferred alternatives using the nine CERCLA criteria, I

MSRE site, ORNL, Oak Ridge, Tennessee 6

Removes the principal threat fro m the MSRE facility by appropriately packaging the sailS and in an appropriate: facility. Removal of me salt is a permanent action to separate the uranium from the salts reduces toxicity of the sailS and mobility is by convc:ning uranium hexafluoride (0 uranium oxide. Volume is amy incrementally reduced it is a small percenta~e of the total volume of the salt During activities of mis alternative. risks from radiation and contamination exposure N with potential release will increase to worke~ and the public as the salt is heated. removed.

N comainerized: however. safety analysis and appropriate precamions will be implemented to reduce

'" control the risks acceptance The interim action proposed plan was presented to the public for review between December 23 .1997.

and January 30. 1998. and no changes in the plans resulted based on the comments that were received.

Comments tended to support the proposed interim action. StakehOlders also participated in review of the documents ARAR = applicable or relevant and appropriate DOE = U.S. Department of Energy ORNL = Oak Ridge National Laboratory requirement EPA = U.S. Environmental Protection Agency ROD = record of decision CERCLA = Comprehensive Environmental F: = fluorine UF" = uranium hexafluoride w

.w Response. Compensation. and Liability Act of 1?80 FFA = Federal Facility Agreement

$ = dollar MSRE = Molten Salt Reactor Experiment

~

Table 2.5. Estimated uranium In the salts before and after separation, MSRE site, ORNL, Oak Ridge, Tennessee Total uranium 117 0.5 673 50 0.214 289 g = gram ORNL = Oak Ridge National Laboratory kg = kilogram ppm = parts per million MSRE = Molten Salt Reaclor Experimenl U = uranium nCi = nanocurie THE SELECTED REMEDY The interim action remedy selected for the MSRE fuel and flush salts remediation is to remove the salt in a chemically stable form, separate the uranium from the salts and store it separately as part of the existing 2lJU repository inventory, place the salt in containers, and store the containerized salt until disposal is arranged. This action will employ the activities common to the first steps in the removal and disposition of the fuel and flush salts for the four action alternatives presented in the FS. The final action required for salt disposal will be documented in a subsequent !m~1. CERCLA decision document and, ~~;; *ijijr§i1r!g~ii'gW!l NEPA decision document.

Removal of salt from the drain tank cell will require new corrosive resistant equipment to add heat and control the salt chemistry. To the extent possible, existing drain tanks and other equipment will be examined and repaired for reuse, but requirements for operating the apparatus remotely and adding HF to the melting salt exceed the original equipment capability. The goal of the project is to remove 99 percent of the salts from each drain tank. This will reduce the uranium mass left in each tank to below criticality safe limits .

JT00869709.1 DH/CJE 2-24 June l, 1998

The separation of uranium from the fuel and flush salts will use the same process and, to the extent practicable, the same equipment used to remove 23>U in 1968. This process involves adding fluorine to the molten salts. Uranium hexafluoride gas is liberated from the salts and then trapped on vertical columns packed with sodium fluoride. The goal is to reduce the residual uranium concentration in the salts to below 50 ppm. Depending on salt chemistry, it may be possible to reproduce the results achieved in 1968 (26 ppm). Table 2.6 shows the estimated 2JJU and total uranium concentrations before and after the separation process .

Uranium must be converted to uranium oxide to be placed in storage at the ORNL repository . Although this conversion process is common in the uranium industry, a modification

. tailored to a small scale, remote chemical operation will be applied to this application. The chemically stable converted uranium will be packaged in suitable containers and prepared for storage with similar packages in a 2llU repository in Building 3019. Storage of this separated uranium will result in approximately 17 kg (37Ib) of 211 U added to the 500 kg (1 ,100 Ib) of 2JJ U currently stored at the facility.

Once the uranium is separated from the salts, the residual salts will be poured into storage containers (approximately 48 containers for the fuel and flush salt) , and chemically stabilized!

packaged to capture fluorine gas which may be generated and to meet transportation requirements for eventual shipment to a disposal area. Because a disposal facility is not available to make waste acceptance determinations or to receive waste, the waste packages will be loaded into shielded casks for interim storage. These casks will be placed in interim storage at an Q!{Nli 911§f~nng storage faci IitYll;t\!)RtM§m;if4'§.UJt!~§!IiIiit#m9i§!iw.1I!s~i}yq!\~i!nsli[~§!!lgSH11lRY' RqnR~t§I!U~g§llg~§t :~\lg7§§~li§m#m§4!:~tgt~g~W~W(~ig;;I!§%;?)i!w9I~tlt~14WJJ99!l8t~!§ y~~!!~

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\It:$&\§?1i§~p~sif!~!\l!M;lqttJjg~\9£~,~~£ifID'§!m]!n!~~gjH~ *wl!§S~§~~):.':: !1m~Iiq~!!mU§n§tm~

§m~I~e9£~>~E;A!!g;~\§&[~g;~!!$;}YJl!Eim£qnl'(5,t~lgq~~!Il.~mqf~i!$;tgll1~gl~!!~~~!M,i Total capital cost (present worthj to implement these interim activities is $39.3 million and the annual operation and maintenance cost (present worth) are expected to be zero. The total capital cost includes only the activities discussed in this section. Costs associated with interim storage are not borne by this project; the $10,000 yearly costs are borne by other DOE-funded programs. Other activities such as transportation to an end point disposal location identified in the original four action alternatives are not included in this cost. Table 2.6 presents the schedule for these activities .

Decisions concerning treatment and disposal of the salt is delayed to a later date . This has the advantage that these decisions could be based on better information as waste acceptance TT00869709. 1BHfCJE 2-25 June 3, 1998

Table 2.6. Interim remedial action schedule, MSRE site, ORNL, Oak Ridge, Tennessee Melt and transfer salts for 2000 2oo2 uranium from salt October 2000 February 2003 Transfer uranium to "'u October 2000 February 2oo3 Stabilize and package salt OClOber 2000 February 2oo3 Imerim storage of salts October 2000 Undetermined Remedial action report February 2oo3 May 2oo3 Notes: Dales include operations. The durations do nOI include design . construction. elc .

MSRE = Mollen Sail Reactor Experiment U = uranium ORNL = Oak Ridge National Laboratory crileria are developed and finalized for the national repository and WIPP, -new treatmen!

technologies emerge, and further development is completed for existing treatment technologies presented in the FS.

STATUTORY DETERMINATIONS Section 121 of CERCLA establishes several statutory requirements and preferences, including compliance with ARARs . CERCLA requires the remedy (1) be cost-effective; (2) be protective of human health and the environment; (3) use permanent solutions and alternative treatment technologies or resource recovery technologies to the maximum extent practicable; and (4) use treatment that permanently reduces tlie toxicity, mobility, or volume of hazardous substances. Interim remedial actions under CERCLA are required to attain only those ARARs specific to the action being implemented, and the above criteria apply to the selection of a final remedy. The selected interim action satisfies the above criteria.

This in!erim action provides short- and long-term protection of human health and the environment through removal of a contaminant source and limitation of the potential spread of contamination. This action will comply with all ARARs. The action is cost-effective. The action uses treat men! to remove and stabilize uranium for storage in the 2l3U repository at ORNL rrOO869109.IBHfCJE 2-26 June J, 1998

and is permanent within the scope of the action because it removes the fuel and flush salts from the MSRE facility. The proposed action also reduces the potential contaminant release and is therefore appropriate as an interim action .

EXPLANATION OF SIGNIFICANT CHANGES A review of all conunents resulted in no significant changes to the remedy originally identified in the proposed plan as the interim action alternative.

REFERENCES DOE (U .S. Department of Energy). 1997a. Illterim Actioll Proposed Plallfor Fuel alld Flush Salt Dispositioll from the Moltell Salt Reactor Experimellt, Oak Ridge Natiollal Laboratory, Oak Ridge, Tellll essee, DOE/OR/02-1601&D3 . Prepared by Jacobs Environmental Management Team, Oak Ridge, TN .

DOE. 1997b. Feasibility Study for Fuel alld Flush Salt Removal from the Moltell Salt Reactor Experimelll at the Oak Ridge Natiollal Laboratory, Oak Ridge, Tellllessee, DOE/ORl02-1559&D2. Prepared by Jacobs EM Team, Oak Ridge, TN .

DOE. 1996. Actioll Memoralldum for Urallium Deposit Removal at the Moltell Salt Reactor Experimelll, Oak Ridge Natiollal Laboratory, Oak Ridge, Tellllessee, DOE/ORl02-1488&D2. Prepared by Jacobs EM Team, Oak Ridge, TN .

DOE. 1995 . Oak Ridge Natiollal Laboratory Moltell Salt Reactor Experimellt Facility Time Critical Removal Actioll Memoralldum Report. Prepared by Lockheed Martin Energy Systems, Inc., Oak Ridge, TN .

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PART 3. RESPONSIVENESS

SUMMARY

ITOO869709.I BH/CJE June J. 1998

RESPONSIVENESS

SUMMARY

The Illterim Action Proposed Plan/or Fllel and Fillsh Salt Disposition/rom the Molten Salt Reactor Experimellt, Oak Ridge National Laboratory, Oak Ridge, Tennessee (DOE 1997a) was released for public review December 22, 1997. The comment period for the public to consider the alternatives developed for interim remediation of MSRE was announced in local newspapers to begin December 23, 1997, and end January 3D, 1998. The notice of availability for this plan and other documents in the Administrative Record was published daily in 171e Knoxville News-Selltillel and 171e Oak Ridger December 23, 1997, and biweekly and weekly in 171e Roalle COllllty News and 171e Clilltoll COllrier-News December 24, 1997. A public meeting was held in Oak Ridge January 14, 1998. This public meeting was also announced in newspapers January 11 and 12, 1998.

Through newspaper announcements and other public relations efforts, DOE invited the public to participate in the review of plans being recommended for interim remediation of MSRE.

The interim action proposed plan and other related documentation in the Administrative Record were made available for review at the DOE Information Resource Center, 105 Broadway Avenue, Oak Ridge, Tennessee. Written comments from the public could be received at the Information Resource Center or sent to Ms. Margaret Wilson, DOE FFA Manager. DOE also accepted written comments at the public meeting and responded to verbal comments. A transcript of the public meeting is included in the Administrative Record.

DOE received two written comments during the public conmlent period . Responses to these comments are included here . In addition, verbal comments that address the current remedial action plan are included here to supplement the initial DOE response made at the public meeting. Public comments and DOE responses that were made at the public meeting and which do not address the plan for interim action are not included here.

LETTER 1 COI/lmelll: DOE and ORNL have approached the plan for MSRE fuel and flush salt disposition in a thonghtful, forthright and honorable way.

Response: The support of the proposed plan is appreciated.

rrOO869709.1 BH/CJE 3-3 June 3, 1998

LEITER 2 Commellt: After review of the doclmlents concerning the interim action proposed plan for fuel and flush salt disposition and attending the public meeting. I fully concur ,vith the

• decision to select the preferred limited alternative which includes removal and interim storage of the fuel and flush salts. I also studied the National Research Council report that evaluated the alternatives for MSRE fuel and flush salts removal and disposition. Tltis report only solidified my ophtion that the proposed plan was the correct one.

I was pleased that TDEC and EPA approved the proposed plan. I am concerned that the r~gulatory process for approvals is not open to the public like the DOE decision process.

I would like to be part of the regulatory process to gain knowledge of their reasoning and have the opportunity to discuss the reasons for decisions with the regulators.

Respollse: The support of the proposed plan is appreciated. XQUfiaMlf~HorI'greafei

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SUMMARY

OF COMMENTS FROM THE PUBLIC MEETING Commel/t 1: Three meeting participants commented that the proposed interim action plan is appropriate and includes a reasonable approach for removing the ' salt from the MSRE. In addition, even though the proposal does not include a recommendation for final disposal of the salt, it is the correct action to take because it reduces the risk of a release of contaminants to the envirolUnent; and that the plan provided for due precautions to solve a complex problem.

Respol/se 1: The support of the proposed plan is appreciated.

Commel/t 2: Three meeting participants raised concerns abont an alleged nuclear criticality accident at the MSRE and alleged past releases/contamination incidents.

Respol/se 2: Previous investigations determined that there has not been a criticality accident at the MSRE, and that contamination incidents were minor and limited to two workers in the facility. It is acknowledged, however, that there is the risk for a nuclear criticality accident and substantial releases to the environment/public of fluorine gas and radioactive contamination associated with the salts in the MSRE drain tanks. This is the reason that instead of the No Action alternative, the proposed plan is to remove the salt from the drain tanks, remove the uranium from the salt, stabilize/package the salt to control fluorine generation, and place the salt containers in interim storage.

Commel/t 3: Suggestions for alt'ernate remediation options were stated during the public meeting by different conmlenters. These various options are presented with a brief response.

(A) Has inCluding the salt in the privatization initiative for transuranic waste treatment after it is removed from MSRE been considered?

(B) Suggest melting the salt and placing it into containers for storage as spent nuclear fuel. TIllS would get it out of the way so you can go ahead and decontaminate and decommission the MSRE building. But you will still have the fluorine problem wherever you store the salt, and that may not be a job you want to do.

(C) Suggest fluorination to remove the uraniUlll from the reactor and mix tillS uralllwn with depleted uranilml from K-2S, denature the uralllum, and make the uralllum ITOO869709. JDII ICJE 3-5 June 3, 1998

safe. Then after that precipitate the uranium with either anunonia or sodium hydroxide and make orange cake, and dispose of the orange cake in the burial grounds.

(D) [This idea was presented as not necessarily practical.] Suggest placing one or two hundred tons of crushed limestone in the cell (containing the fuel and flush salt storage tanks) to fill it. That would take care of uranium hexafluoride, excess fluorine, and probably would take care of a rising water table.

Response 3:

(A) Yes. inquiries about including the MSRE salts in the privatization proj ect have been made; however. because the salts are unique in their chemical make-up with very little similarity to other wastes at ORNL. inclusion of the salts is no longer considered.

(B) The suggestion to containerize and store the material as SNF implies not removing the uranium before containerization. This was evaluated in the FS and discussed with the state of Tennessee and EPA. It was determined that removing the uranium from the salt during the current operations would be a small incremental cost to the project. Not removing the uranium, however, may prevent future disposal at WIPP or prevent processing at INEEL for future disposal at the National Repository. (Note: the work plan will address generation of fluorine during interim storage.)

(C) The quantity of uranium (2JJU) that will be removed from the MSRE fuel and flush

.salts is a very small amount compared with the quantity already stored in the 2llU repository.

The process required to complete the suggested blending is not insignificant. Application of the suggested process to address only the uranium from the fuel and flush salt would be inordinately complicated and costly. The more appropriate implementation of this suggestion is to address all of the 2llU in the repository . Treatment of the repo~itory inventory is beyond the scope of this action.

(0) This interim remedial action is interim in part because it is only the first action for the 0&0 of Building 7503, and this is the first action in removing, storage and disposition of the fuel and flush salts. Before Building 7503 and MSRE can be decontaminated and decommissioned, the fuel and flush salts must be removed. The salts cannot be left in place not only because uranium hexafluoride and fluorine gases are liberated, but also because of the hazards associated with and the regulatory guidance for disposition of spent nuclear fuel and/or TRU waste. Leaving the fuel and flush salt in Building 7503 is not a .viable option under these circumstances, even if crushed limestone would be an effective temporary or permanent cover.

IT00869709.1n.'/CJE 3-6 June 3, 1998