ML20209J547

the Potential for Criticality Following Disposal of Uranium at LOW-LEVEL-WASTE Facilities.Containerized Disposal

ML20209J547
Person / Time
Issue date:	06/30/1999
From:	Coltenbradley, Hopper C, Parks C, Toran L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), OAK RIDGE NATIONAL LABORATORY
To:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
CON-FIN-L-1376 NUREG-CR-6505, NUREG-CR-6505-V02, NUREG-CR-6505-V2, ORNL-TM-13323, NUDOCS 9907210228
Download: ML20209J547 (100)

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NUREG/CR-6505, Vol. 2 ORNL/TM-13323/V2 l

The Potential for Criticality l Following Disposal of Uranium at Low-Level-Waste Facilities Containerized Disposal 4

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U.S. Nuclear Regulatory Commission / ^*s Office of Nuclear Regulatory Research I, .

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NUREG/CR-6505, Vol. 2 ORNL/TM-13323/V2 l

The Potential for Criticality Following Disposal of 4

Uranium at Low-Level-Waste {

Facilities Containerized Disposal Manuscript Completed: August 1998 Date Published: June 1999 Prepared by L. E. Toran, C. M. Hopper, C.V. Parks, ORNL V. A. Colten-Bradley, NRC Oak Ridge National Laboratory Managed by Lockheed Martin Energy Research Corporation Oak Ridge National Laboratory l Oak Ridge, TN 37831-6370

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I Prep red for Divisi:n of Waste Managernent Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC Job Code L1376 m g% l g

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ABSTRACT l

The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, '

geology, geochemistry, soil chemistry, and criticality safety was formed to develop and test some reasonable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM) and to use these seznarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some possible scenarios for uranium migration and concentration increase at LLW disposal facilities,(2) model groundwater transport and subsequent concentration increase via precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM 2

was restricted to "U in the present scope of work. The work documented in this report indicates that the potential for a criticality safety concern to arise in an LLW facility is extremely remote, but not impossible.

Theoretically, conditions that lead to a potential criticality safety concern might arise. However, study of the hydrogeochemical mechanisms, the associated time frames, and the factors required for an actual criticality event indicate that proper emplacement of the SNM at the site can eliminate practical concerns relative to the occurrence and possible consequences of a criticality event.

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' NUREG/CR-6505, iii Vol. 2

-CONTENTS l

l P.Agt AB STRA CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . i i i LI ST OF FIG URES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . , , vii j LI ST OF TAB LES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i x l

EXEC UTIVE

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xi ACKNO WLEDG M ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xiii 1

PURPOSE...........................................................................I 2 PREVIOUS WORK . . . . . . . . . . . . . . . . . . . . . .............................. ........ ...3 3 SITE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5 4 APPROAC H . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

~ 4.1 NUCLEAR CRITICALITY EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.1.1 Code Description and Validation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.1.2 Analytical Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.1.3 Param eters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 4.2 HYDROGEOCHEMICAL MODELING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 {

4.2.1 Conceptual Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 '

4.2.2 M ode l s Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.2.3 Parameters and Model Grid . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 4.2.4 Sensitivity Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 5 A S S UM PTI ON S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 6 CRITICALITY SAFETY EVALUATION RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 6.1 2"U ENRICHMENT INFLUENCE ON CRITICAL MASS OF URANIUM . . . . . .. . . . . . . . . . 23 6.2 -TRENDS FOR U(10)-H 2 0-SIO 2 MIXTURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 6.3 COUPLING OF NUCLEAR CRITICALITY AND HYDROGEOCHEMICAL MODELING . . 28 1 7 HYDROGEOCHEMICAL MODELING RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.1 REDUCING ZONES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.2 SEN SITIVITY ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 7.3 PRECIPITATION OF SILICATE MINERALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 7.4 3-D HYDROGEOCHEMICAL MODEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 8 DI SC U S S I ON , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 8.1 EN RI C HM ENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... ..... .......... 33 8.2 SOU RC E TERM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 8.3 GEOCHEMICAL PROCESSES . . . . . . . . . . . . . . . . . . . .. .. ..... ... ........ 39 NUREG/CR-6505, y Vol. 2 L'

8.4 HORIZONTAL-VS-VERTICAL FLOWPATHS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 8.5 OTHER MITIGATING FACTORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 9 CONC LU S ION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 7 10 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 APPENDIX A: Criticality Study Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 APPENDIX B: Suberiticality Evaluation for Chem-Nuclear Systems, Inc., Trench 23 . . . . . . . . . . . . . . . . . . 79 NUREG/CR-6505, Vol. 2 Vi

LIST OF FIGURES 1

Eigurr Page I 3.1 Typical construction of waste trenches .. .... . ........ . . . . . . . . . 6 l 3.2 Histograms of source material in trenches . . . . . . . . . . . . . . . . . . ... .. ....... . 8 3.3 Example of redox zonation in a landfill cross section ............. . ... .. .. .. . .. I1 4.1 Conceptual configuration for nuclear criticality evaluations . . . . .. ..... . ...... .... . 14 4.2 Schematic of model grids showing oxidized injection into reducing zone . . . ... . 17 6.1 Critical mass "5U vs H/X for UO22 F -H 2O in spherical H 20-reflected systems . . .. . .. . 24 6.2 Critical mass U vs H/X for UO2 F2 -H 2O in spherical H 20-reflected systems .. .. ..... . .. 25 2

6.3 Infinite slab areal density (kg "U/m2 ) vs g H2 0/g SiO2 and log scale of g2 "U/g SiO2 ..........27 7.1 Reaction progress for hydrogeochemical model . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 31 8.1 Relationship between enrichment and source term . . . . . . . . . . . . . ... . .. ..... .. .. . 34 8.2 Critical (km,. 2 0.95) masses of spheres and " pseudo-infinite" slabs and cylinders as a function of water content for different densities of 2"U in waste matrix ... . ... ... . 38 l

8.3 Histograms of calculated concentrations of 2"U in old and new disposal trenches '

using reported SNM or 2"U mass and reported disposal volumes . .. . ..... .. . . 40 l

8.4 Histograms of calculated areal densities of 2"U in old and new disposal trenches 2

using reported SNM or "U mass and reported disposal volumes . . . . . . . . .... .... ..... 42 8.5 Cross sections showing horizontal flow (a) within the waste zone due to preferential flow paths and (b) at the base of the trench due to less-permeable sediments at the bottom bcundary and trench drainage . .............. ... .............. .. .. ... ............ .. . . . 43 2

A.1 Infinite media neutron multiplication factor (k ) vs g H 2 0/g SiO 2 and g "U/g SiO2 ... . .. . 57 A.2 Infinite media neutron multiplication factor (k.) vs g H2 0/g SiO2 and log scale of g2 "U/g SiO2 . . 58 A.3 Infinite slab thickness (cm) vs g ILO/g SiO2 and g2 "U/g SiO 2 ....... .... . .. ..... .. 59 A.4 Infinite slab areal density (kg2 "U/m 2) vs g H2 0/g SiO2 and g2"U/g SiO 2 .. .... ....... 60 2

A.5 Ir: finite slab areal density (kg "U/m2) vs g H2 0/g SiO2 and log scale of g2 "U/g SiO2 .... . . 61 2

A.6 Infinite cylinder diameter (cm) vs g H O/g SiO 2 and g "U/p siO2 . . . . ... . .... 62 NUREG/CR-6505, vii Vol. 2

A.7 Infinite cylindir line:r density (kg 2"U/m) vs g H2 0/g SiO2 and g2 "U/g SiO2 ......... ... ... . 63 A.8 Infinite cylinder linear density (kg 2"U/m) vs g H 20/g SiO 2and log scale of g2 "U/g SiO2 ....... . . 64 2

A.9 Sphere diameter (cm) vs g H 2 0/g SiO 2 and g "U/g SiO2 ........... .......... ........ ... 65 A.10 Sphere mass (kg 2 "U) vs g H2 0/g SiO, and g2 "U/g SiO2 ...................... .. .......... 66 A.11 Sphere mass (kg2 "U) vs g H2 0/g SiO2 and log scale of g2 "U/g SiO 2 .... ................ . . . . 67 NUREG/CR-6505, Vol. 2 viii l

I

b-LIST OF TABLES Table East 3.1 Concentration ofdissolved nonmetals and metals in trench water samples taken at the LLW burial site near Barnwell, S.C. .................................. .......................9 4.1 Components and reactants (minerals and aqueous complexes) used in ParSSim . . . . . . . . . . . . . ....I8 6.1 U(10) plus H2 O plus SiOgoil (S-S) results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 7.1 Parameter variation and results of hydrogeochemical modeling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 30 8.1 Disposal records from Barnwell, S.C., and calculated enrichments and density . . . . . . . . . .. . . . . . . . . 35 A.I U(10) plus H O2 plus SiO -soil 2 (S-S) results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 B.1 Suberiticality evaluation assumptions and ramifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 B.2 Raw and transformed data from Autry, 1998............................................... 84 B.3 . Concentration factors for criticality concern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 l

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NUREG/CR-6505, ix Vol. 2 l

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EXECUTIVE

SUMMARY

This work is Volume 2 of a two-volume study to address the potential for nuclear criticality resulting from increasing uranium concentrations in low-level-waste (LLW) disposal facilities. In contrast to Volume 1, which 2

focused on "U blended with soil, this report focuses on containerized waste with 10 weight percent (wt %) "U in uranium, U(10). Hydrogeochemical modeling assumed precipitation of uranium in reducing zones instead of concentration on sorption sites as a means of forming a critical mass. As in the earlier study, several assumptions were made in developing the hydrogeochemical and criticality models. These are discussed in Section 5.

Th: criticality safety calculations showed that higher concentrations of 2"U were needed for U(10) than U(100),

as expected. Using the minimum concentration values necessary for a potential criticality, the mass of 2"U as 2

U(10) was 1.6 times greater than the "U as U(100). Differences in :"U concentration have not been analyzed on a point-by-point basis and could be larger and smaller than the 1.6 value observed at the minimum concentration posing criticality concerns.

The mechanism of precipitation for increasing the concentration of uranium in assumed reducing zones undcr saturated conditions has been evaluated. These reducing zones formed very efficient barriers to uranium transport, precipitating nearly 100 wt % of the uranium in solution. The results of the geochemical modeling indicated that the reducing zone did not become oxidizing despite the influx of oxidized water. The source of reducing agents is postulated to be steel drums or wooden crates, thus serving as plausible locations for a buildup of uranium. Further study of the geometry of these reducing zones would be needed to evaluate the potential for concentrating relatively small critical masses (e.g., spherical masses). Other limitations may be identified through the evaluation of reducing zones.

Nonetheless, disposal practices at the Chem-Nuclear Systems, Inc., disposal facilities at Barnwell, S.C., restrict the possibility of criticality safety concerns in several ways. Very low-average 2"U enrichments have been l reported for most trenches, below the I wt % limit to produce a criticality concern under typical disposal I conditions. For most trenches with higher 2"U enrichments, the source term (e.g., mass) for uranium is too low to produce a slab of sufficient size with the required increase in concentration of:"U needed for criticality concern. One exception is Trench 23, which has a high enrichment (greater than 80 wt %) and a large source (175 kg)of "U.

l 2

Even for the limited examples that potentially have sufficient "U, very long times are needed to accumulate a I critical mass. For the most conservative travel time, assuming one-dimensional (1-D) flow and no dispersion, tens to hundreds of thousands of years are needed. Three-dimensional (3-D) modeling indicates even longer times are needed when dispersion is incorporated. The flow paths would need to funnel the uranium from large trench volumes to relatively small reducing zones in order to increase the concentration of 2"U to the level that would pose a criticality safety concern. Iflarger reducing zones form, either the2 "U will be too diffuse to pose a criticality safety concern, or larger sources of 2"U than are reported to be present in the trench are required.

Uranium travel times are long enough to allow monitoring and possible mitigation of conditions that could pose criticality safety concerns.

This study results in the following recommendations for consideration oflicense reviews of LLW facilities:

1. Minimize those factors that enhance SNM accumulation.

= Reduce groundwater infiltration

. Reduce enrichment NUREG/CR-6505, xi Vol. 2

Executive Summ:xy

- Minimize opportunities to create isolated zones of reducing conditions. Avoid organic matter in waste cells

- Design trench to minimize focused flow

2. Limit the areal density of the fissile materials.

3. Model trench performance using site-specific condit:ons on a scale that addresses the potential for criticality.

Consequently, the observation that the average enrichn;ent of a trench is less than I wt % 2"U in the urtnium does not necessarily eliminate a criticality concern for the trench. Burial reports may suggest that localized regions of a trench contain quantities of fissile material that greatly exceed the average enrichment.

4. Continue to use sumps in disposal trenches to monitor for the presence ofiron, organics, and uranium as indicators of mobility in the trenches. If uranium is observed in the sumps, determine its enrichment.

Changes in redox conditions may be monitored by changes in different iron species. Even though it may take many years for sufficient buildup of uranium, early detection of mobile iron and uranium would indicate changes in the trench water chemistry.

2 Disposal trenches at the Barnwell, S.C., LLW facility have waste materials containing uranium with average "U cnrichments less than 1 wt %, insufficient masses of 2"U at enrichments larger than I wt %, or distributions and mass proportions of 2"U and 2"U such that criticality safety concerns are not a realistic issue. For the single disposal trench, Trench 23, having a large mass of waste material containing highly enriched uranium, subcriticality is ensured by the physical distribution and commingling of the material with substantial quantities of" source material," which is typically normal or depleted uranium (e.g.,0.7 or 0.2 wt %2 "U in uranium respectively). Uranium concentration factors (i.e., hydrogeochemical relocations and densification of uranium) larger than 10 are required to pose a potential criticality safety concern. As demonstrated by the evaluation of the re:ction for hydrogeochemical cumulative uraninite precipitation for long time frames, it requires a minimum of 2 2 7,000 years to increase the "U density from about 0.002 g/cm to 0.02 g/cm'. It isjudged that the same hydrogeochemical processes will redistribute and concentrate the commingled source material that is present, th reby further ensuring suberiticality through isotopic dilution of the SNM with normal or depleted uranium to 2"U enrichments less than I wt %.

NUREG/CR-6505, Vol. 2 xii

i ACKNOWLEDGMENTS i

This work was supported by the NRC under Task 14," Reconcentration of SNM in Low-Level Disposal l Ftcilities," of JCN L1376, TechnicalSupportfor Design, Construction. Operation, and Performance Reviewsfor l Low-Level Waste. V. Colten-Bradley of the NRC provided technical direction and contributed substantially to the work throughout the duration of this project. The cooperation received from the staff at the Chem-Nuclear Systems, Inc., disposal facilities at Barnwell, S.C., and the State of South Carolina, Department of Health and Environmental Control, was greatly appreciated by the authors.

The authors acknowledge the helpful assistance provided by Patricia B. Fox and Lester M. Petrie, Jr., for performing numerous neutronic calculations and producing representations of the computational results. We also teknowledge thoughtful reviews of this manuscript by Gary Jacobs and John McCarthy.

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NUREG/CR-6505, xiii Vol. 2

1 l

1 PURPOSE Th: purpose of this study was to evaluate the potential for hydrogeochemical processes to redistribute special nuclear material (SNM) in containerized low-level-waste (LLW) disposal facilities such that there is sufficient concentration increase and geometry reconfiguration to permit nuclear criticality. This particular evaluation is restricted to criticality safety concerns and geochemical processes associated with uranium. The approach was (1) to identify some reasonable scenarios for uranium migration and increase in concentration at LLW disposal facilities, (2) to model coupled groundwater transport and geochemical speciation of uranium, and (3) to evaluate the potential for nuclear criticality in terms of passive geometry configurations and increases in uranium concentration. A combination of hydrogeochemistry and criticality safety experts worked together to perform the evr.luation.

This study extends the previous work reposted in The Potentialfor Criticality Following Disposal of Uranium at Low-Level Waste Facilities, Volume 1: Uranium Blended With Soil (Toran et al.,1997). The present work (mphasizes the disposal of containerized uranium instead of disposal of uranium blended in soil. Some different scenarios and concentrations are evaluated. In particular, the emphasis in this report is on the mobilization of uranium under oxidizing conditions, with immobilization under reducing conditions. In the previous report, urznium was desorbed from the soil and resorbed in a zone of high-sorption sites. This process did not increase the uranium concentration enough to cause nuclear criticality safety concerns, and the scenario will not be reexamined here. In addition, the previous criticality safety analysis assumed the uranium was enriched to 100 wt % 2"U [ referred to as U(100)]. In this report, additional calculations are presented for 10 wt % 2"U m l uranium [ referred to as U(10)], and some comparisons are drawn between these U(10) calculations and the 4 previous U(100) computational results reported in Vol.1.

Tha Chem-Nuclear Systems, Inc. (CNSI), LLW disposal facility at Barnwell, S C., was used as an example of site conditions for containerized disposal. Specific disposal practices at the site were evaluated as they related to the ,

potential for the 2"U concentration to increase.  !

However, the models were process-oriented rather than site-specific. That is, the models emphasized processes that could occur in disposal settings, rather than being a detailed construction of site conditions such as might be used in a performance assessment. Some details could not be addressed without a site-specific model that incorporates transient soil moisture conditions, or without additional data such as packing configurations and weathering rates. Assumptions were selected based onjudgment regarding the potential conditions that would increase the possibility for criticality.

The questions addressed in this study are:

o Is there sufficient inventory for the available geometries requisite for criticality?

o How does the concentration of 2"U needed for criticality compare with systems containing U(10) vs U(100)?

o What chemical conditions and physical aspects of trenches are conducive to increasing uranium concentration?

o Can reducing zones, which precipitate uranium, be sustained to enable critical masses to accumulate?

o How could disposal practices, in particular at Barnwell, S.C., enhance or mitigate the development of critical masses?

NUREG/CR-6505, 1 Vol. 2

Purpose Section 1 The questions addressed in this study reflect the important processes that could be evaluated with the available data, hydrogeochemical models and criticality safety analyses. The results provide bounds on conditions that could raise nuclear criticality safety concerns. Insights gained from the hydrogeochemical modeling and criticality safety analyses will be used as a basis for recommendations concerning future disposal practices for SNM.

NUREG/CR-6505, Vol. 2 - 2

f 1

l l 2 PREVIOUS WORK 1

l In the previous report, Vol.1 (Toran et al.,1997a), nuclear criticality evaluations and hydrogeochemical scenarios l w;re based on licensed soil-contamination limits specified for Envirocare of Utah, Inc. The maximum average

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' concentration of 2"U permitted in disposed waste under the State of Utah license (UT 2300249) is 770 pCi 2"U/g soil, which equates to about 0.0006 g of 2"U per em' of soil given a soil bulk density of about 1.6 g soil /cm'.

If disposal occurred at this maximum concentration, there is the theoretical possibility of a nuclear criticality i

recident, given assumptions about hydrogeochemical influences on reconfiguration of the uranium. Only a n:rrow range of conditions resulted in sufficient increase in uranium concentration, and the length of time required to increase the concentration of uranium is expected to be many thousand ofyears. This slow approach to criticality will further mitigate consequences that occur with rapid approaches to critical conditions.

However, it is important to note that reviews of disposal records from Envirocare of Utah, Inc., indicate that cone:ntrations of 2"U in the waste material are more than a factor of 10 less than allowed by the license, and that 2

tha c.verage site "U enrichment is below the minimum I wt % (Pruvost and Paxton,1996) required to achieve nucl:ar criticality. Thus the likelihood of a criticality accident is vanishingly small.

B: cruse of the numerous combinations of parameters that could be considered in nuclear criticality evaluation and hydrogeochemical modeling, bounding and simplifying assumptions were used in the analysis. Nuclear criticality evaluations were performed for simple geometries using two generic soil types: SiO2 soil (the more conservative medium because pure SiO 2 is the least likely soil composition to absorb neutrons, thereby enhancing the potential for criticality) and a " nominal soil" composed of minerals and secondary phases representative of a world-average soil composition.

Potential, direct radiation exposures were estimated for two postulated types of criticalities: one with a high concentration factor (large increase in2 "U concentration) and one with a low concentration factor. The locations i of the determined radiation exposures were for two positions 1 m above grade. One position was directly above I the d: posit, and the other position was 90 m away from the deposit. The assumed fission yields from both type l

syst:ms were based upon the tission energy release necessary to remove the quantity of water that is necessary to '

mod: rate neutrons to sustain nuclear criticality throughout an over-moderated condition. These assumptions were predicated upon a geologically slow approach to a non-idealized critical geometry, thereby permitting localized steam generation and self regulation and shutdown of the fission-chain reaction. Alternative idealized tssumptions have been postulated by others (Bowman and Venneri,1994; Greenspan, Armel, Ahn and Vujic, l 1997) that present more severe consequences.

The criticality evaluation showed that the SiO2 -soil results are similar to the nominal soil results. In terms of the hydrogeochemical processes that can increase uranium concentration, the critical slab configurations are more re dily achieved than cylindrical or spherical configurations (i.e., lower concentration factors are required).

The criticality evaluation also provided a minimum concentration needed to achieve criticality safety concerns,  !

which was the target concentration for hydrogeochemical modeling. i Simplifying assumptions in the hydrogeochemical modeling included one-dimensional (1-0) transport and saturated conditions. The hydrogeochemical scenario was postulated based on the geometry of the Envirocare site with disposal in soil containing sorbed uranium, then mobilization to a zone of higher-sorption capacity below the disposal trench. This sorption zone could potentially contain uranium in a zone of higher concentration. A reducing zone to capture uranium was also hypothesized, but not explicitly modeled, in the previous study. A reducing zone was difficult to define given the limited supply of reducing agents in the trenches and the unsaturated conditions in soil, which would keep the system oxygenated. A sensitivity analysis NUREG/CR-6505, 3 Vol. 2

Previous Work Section 2 was performed to evaluate various factors, such as concentration of completing agent, quantity ofinitial uranium source term, and groundwater velocity on the potential to increase uranium concentration.

l The previous work noted that the concentration of complexing agent and the size of the source term were limiting factors in the reconfiguration of uranium. For most scenarios, once sufficient ure.i=n was mobilized, the I

concentration of a complexing agent was important because it outcompeted sorption sites in the high-sorption aore and prevented increases in uranium concentration. The possibility ofimmobilizing uranium in reducing ze s was presented as a more-likely scenario (to be evaluated in the present work). Furthermore, if the initial c c ;entration of uranium could be limited during disposal (e.g., by limiting disposal thickness), it would not be possible to increase the uranium concentration sufficiently along a 1-D flow path to pose a criticality safety concern. Much uncertainty exists in the estimates of the time frame for the increase in uranium concentration, but analogs from soil-forming processes suggest that these processes can require thousands of years.

Volume 1 of this report provided the following recommendations for consideration during a license review of LLW facilities having uranium blended with soil:

1. Minimize the factors that enhance the increase in the concentration of uranium. For example, reduce water infiltration, dilute the 2"U by reducing the enrichment, and minimize opportunities to create zones of reducing potential that precipitate uranium readily (e.g., by maintaining unsaturated conditiv s, and avoiding organic matter in waste cells to prevent methanogenesis).

2. Limit the areal density of uranium to a safe value by limiting the licensed depth of the disposal cell and the licensed disposal concentration. Results suggest that criticality safety concerns can be reduced or eliminated even under worst-case hydrogeochemical transport by reducing the disposal cell depth.

NUREG/CR-6505, Vol. 2 4 l

3 SITE DESCRIPTION The parameters used in the models are based on site conditions at the CNSI LLW disposal facility in Barnwell, S.C., although not all physical and chemical conditions were explicitly modeled. A variety of geologic and l hydrologic information on the site is available from previous studies (e.g., Weiss and Colombo,1979; Cahill, j 1982; Dennehy and McMahon,1987; and data provided by CNSI). i The Barnwell facility was opened by CNSI in 1971. The facility receives approximately 8490 nf (300,000 ft')

to 11320 m' (400,000 ft') of LLW per year. Barnwell receives Class A, B, and C waste. The majority of the SNM is contained in Class A waste. These waste classifications are defined in 10 CFR 61.55.

1 About two-thirds of the waste disposed at the Barnwell facility comes from nuclear power plants, with the other third coming from other industry and government sources (such as the U.S. Army). The SNM consists primarily of SU. The *U enrichment is typically less than 10 wt % for trenches with enrichment data available.

Approximately 90 wt % of the SNM is dry active waste. Other waste includes resins, which have been separated from the dry waste since the mid-1980s. Scintillation vials and other organic liquids were banned in 1979.

Originally, waste was received and buried at the Barnwell facility in containers, such as cardboard boxes and drums. During the 1980s, shipments more commonly came in plywood boxes. Then disposal requirements bec:me more stringent, and steel drums (then high-density polyethylene containers) became the standard in the mid to late 1980s. These containers are now overpacked in concrete vaults or cylinders. The containers are disposed in trenches that today are typically 30,480 cm (1000 ft) long,6,096 cm (200 ft) wide, and 762 cm (25 ft) deep. Smaller trenches were used in the past and are used today for high-concentration radioactive waste (Class C).

The trenches are excavated through the surficial sand into clayey sand. Prior to waste placement, buffer sand is placcd on the bottom of the trench. The bottom of the trench is sloped, and a French drain is used to move water to sumps, where it can be sampled for detection of contaminant movement (Fig. 3.1). In the past, the French drains were constructed on one side of the trenches and consisted of gravel. In the late 1980s to early 1990s, slott:d plastic pipes within high-permeability material were used to collect water on both sides of the trench.

When the trenches are full, they are backfilled with sand, then covered with earthen caps. New high-density polyethylene caps have been emplaced on some older trenches to further inhibit water from infiltrating the trenches.

From January 1970 to 1984, there were 12 amendments to the original SNM disposal specifications that influ:nced nuclear criticality safety. The original license included a 200-g possession limit for all SNM with no package limit. The license was amended to increase the possession limit and included package limits of 15 g to 50 g. Later amendments included mass and spacing limitations for accumulations of packages based on disposals from specific generators. In 1981, an areal density limit of 200 g/ft2 was imposed. In addition, the waste form requirements and practices varied over time until 1984 when the 10 CFR Part 61 criteria were implemented.

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l Section 3 Site Description CNSI's last Radioactive Material License (No. 12-13536-01) with the NRC allowed 350 g of*U per package.

Th3 350-g limit pertains to *U, which represents the vast majority of the SNM waste. Other SNM, such as "U and isotopes of plutonium, make up less than I wt % of the total grams of SNM. Shipments commonly contain 1:ss than 100 g per package. Any shipment with I g or more SNM is reportable. Shipments of SNM are presently placed only at the bottom of the trench if they contain at least 30 g of SNM. In the past, configuration of the SNM in trenches was determined by " operational randomness"; that is, packages of waste containing SNM w:re not placed on top of packages already emplaced with SNM. These blocks were typically 304.8 cm (10 ft) x 304.8 cm (10 ft) x 365,76 cm (12 ft) in size, then later 762 cm (25 ft) x 762 cm (25 ft) x 1524 (50 ft).

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Disposal records provided for this project by CNSI vary in detail. Specifically, the records for Trenches 1-35 (herein referred to as the "old" trenches) provide only total SNM mass, whereas the records for "new" trenches (38-87), specific isotopes of uranium have been identified, so grams of*U are available and average trench enrichment can be calculated. No data on disposal amounts were provided for Trenches 36 and 37. The disposal rcenrd data indicate that, in general, smaller quantities of SNM were disposed ofin the "old" trenches. Most trenches have less than 40 kg, with masses ranging between 0.5 kg and 175 kg (Fig. 3.2). In the new trenches, disposed quantities of"U range from just a few grams to 1600 kg, with most trenches containing less than 300 kg.

The Barnwell facility is located on the Atlantic Coastal Plain. The surficial deposit at the site is known as the Tobreco Road Formation (formerly the Hawthorn Formation). The deposit is approximately 1828.8 (60 ft) to 2438.4 cm (80 ft) thick. and contains dominantly a sandy clay (e.g.,85 wt % quartz (Pietrzak et al.,1982)] with cotrse sand occasionally present near the base. The present-day water table is within this deposit, typically J

rround 1066.8 cm (35 ft) below the land surface. Beneath the Tobacco Road Formation is the Dry Branch Formation (formerly the Barnwell Formation), which is a massive medium-grained sand. The permeability of the ,

surficial deposit varies from 3 x 10r' to 2 x 104 cm/s (Cahill,1982, p. 38) based on laboratory tests on core collected in the region from a variety of depths. Dennehy and McMahon (1987, p. 28) analyzed shallow cores nr.r experimental trenches and found a permeability range from 7 x 10' to 7 x 104 cm/s. Field measurements using slug tests tend to the upper end of the hydraulic conductivity (Cahill,1982, p. 38). Porosity of the deposit is estimated to be around 40%. Effective porosity typically is somewhat lower, around 30%, but the porosity of the waste matrix could be higher, up to 40 to 50%, as reported at other waste sites (Spalding,1987).

Soil moisture above the water table is typically high due to the humid, wet climate. Cahill (1982) reports soil moisture measurements made over 1.5 years, with values typically greater than 90%. The annual rainfall is about 114.3 cm (45 in.) per year, but only 35.56 cm (14 in.) to 43.18 cm (17 in.) per year is expected to infiltrate the regional flow system (Dennehy and McMahon,1987). Estimates ofinfiltration in the disturbed area around the trenches with earthen caps have not been reported but may be higher due to runoff from the caps.

Groundwater velocity in the trenches has been estimated through tracer tests and groundwater modeling.

Dennehy and McMahon (1987) constructed experimental trenches similar to waste disposal trenches and monitored water levels, soil moisture, and a salt tracer to estimate groundwater travel times from and within the trenches. Salt granules (Nacl) were placed at the bottom of the experimental trenches and at or near the land surface. Detection of the tracer in monitoring points was used to estimate the velocity of 3 x 10' cm/s in the cap tmd t.round 6 x 10-5 cm/s in the backfill material. Cahill (1982) estimated a similar lower vertical velocity of 4

2.5 x 10 cm/s in a regional groundwater flow model in the area. Both the upper and lov.er ranges of values were used in modeling here.

NUREG/CR-6505, 7 Vol. 2

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Section 3 - Site Description Monitoring of the sumps and other wells surrounding the waste facilities has detected a tritium plume about 914.4 m (3000 ft) in length. Tritium is likely to travel at the velocity of groundwater, being an irotope of water, cnd provides an early warning system ofleakage in the trenches. Tritium was first detected outside the trenches in 1978 and measured in an off-site well in 1990 along the base of the Tobacco Road Formation. Additional caps were installed over the oldest trench area first to help minimize future migration of tritium. Ichimura et al. (1994) r: port that the travel time of the tritium is similar to estimates of velocity in the fast horizontal flow zone of

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l x 10d cm/s beneath the trenches. Cobalt-60 and organics have also been detected below the site (Cahill,1982).

Trench water chemistry reported by Weiss and Colombo (1979) is presented in Table 3.1. Although the chemistry is variable, it indicates moderately oxidizing conditions. However, this does not mean that reducing zones will not form. Zonation of redox species in landfills has been reported elsewhere, and analogues from other sites are useful to consider here.

Table 3.1 Concentration of dissolved nonmetals and metals in trench water samples taken at the LLW burial site near Barnwell, S.C. (Weiss and Colombo,1979)

Dissolved Trenches component (mg/L) 3 5 6 8 25/21' Total alkalinity (as CACO3 ) 100 200 40 600 80 Inorganic carbon 24 -

11 130 38 Dissolved organic carbc,? (DOC) 7 -

2 170 12 Chloride 7 10 90 85 42 Nitrogen (N)(ammonia) 0.3 -' l.4 59 25 Nitrogen (N) (NOi + NOi) <0.04 <0.1 23 8.0 15 Silica' 4.3 7.6 5.8 6.0 5.0 Sulfate <5 7 18 34 56 Total anions (meq/L) 23 4.4 4 16 5.7 Calcium 4.0 3.2 16 34 21 Iron 0.15 1.5 0.4 1.2 0.2 Magnesium 2.5 3.3 1.0 18 3.3 Manganese 0.24 0.34 0.45 0.72 0.32 Potassium 1.0 4.6 1.4 12 3.5 l Sodium 2.3 20 29 87 37 Total cations'(meq/L) 0.55 2.9 2.3 12 4.8

• Trenches 25 and 21 are reported together.

' Insufficient sample for analysis.

' Includes nitrogen as .NHl NUREG/CR-6505, 9 Vol. 2

Site Description Section 3 Relatively few studies detail the redox conditions in landfills because of the problematic nature of obtaining reliable measurements. One of the main difficulties is that not all redox couples are in equilibrium, so a given measurement may not be relevant to all redox couples (Lindberg and Runnells,1984). A series of redox reactions may occur in zones around landfill leachate (Baedecker and Back,1979; Christensen et al.,1994). The zones can be identified by pattems in water chemistry, such as concentration of redox couples, evolving from the cerobic zone, to nitrate-reducing, iron-reducing, sulfate-reducing, then methanogenic (Fig. 3.3). The range of r:dox variation in landfills is quite large and controls the mobilization and concentration of many redox-scasitive species. The observed values range from -200 to +600 mV over distances of hundreds of meters. Similar redox vuiation has also been observed in natural systems (Champ et al.,1979) on the scale of hundreds of meters to kilometers.

Although the zones may be extensive horizontally, they are often quite thin vertically, limited by the lack of vertical mixing in the landfill plume. These zones develop over decades, but may initiate on the order of 5 years (e.g., the Bemidji spill (Baedecker et al.,1993)]. Some of the key factors in the development of zones are the mineralogy of the sediments [in particular, availability of iron minerals, according to Heron et al. (1994)], the organic carbon content, microbial degradation rates, and the moisture content. The size of the different zones also depends on the landfill size, vertical mixing, and rates of groundwater flow.

NUREG/CR-6505, Vol. 2 10

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4 . APPROACH 4,1 NUCLEAR CRITICALITY EVALUATION 4.1.1 Code Description and Validation  ;

Th3 SCALE (1995) code system was used to calculate the k,of the designated systems. SCALE is a modular I system of codes that provides criticality safety analysis sequences (CSAS) to calculate the neutron multiplication ,

factor of a system. Problem-dependent processing of the cross sections to account for temperature effects and l

resonance self-shielding is performed using the NITAWL and BONAMI codes. For this study, the XSDRNPM l cods was executed by the CSAS module to provide the k, values. XSDRNPM is a deterministic code that solves the Boltzmann equation for neutron transport in a 1-D mathematical system using a discrete-ordinates approach.

SCALE was used because ofits historic and recognized success in the performance of benchmark and tpplications analyses for licensing activities.

The stationary system of the SCALE codes used for this study and validation, CSAS, BONAMI, NITAWL, XSDRNPM, and KENO V.a, were created May 30,1995. The Brookhaven Evaluated Nuclear Data File B Version V (ENDF/B-V) point cross-section library, which was collapsed to a 238-neutron-energy group library (Greene,1994), named REF01.XN238, was created May 26,1995, and resided on the same hardware platform as the SCALE suite of codes during the period of this study. The 238-energy-group library was used because ofits currency of evaluation, testing, and benchmarking. The hardware platform, the SCALE computational codes, and the 238-energy-group library used were validated through the computation of verification and validation 2

benchmarks involving "U systems before and after the evaluations performed for this study. The verification and validation benchmark calculations provided identical results for calculations performed before the study as those performed after the study, thereby demonstrating the stability of the software and data throughout the study.

The bias and uncertainties of the benchme.rk calculations were within ~0.5% of the experimental values; that is, the calculated k, values of the 14 critical experiment benchmarks were between 0.9954 and 1.0064. See Vol. I for more details.

4.1.2 Analytical Approach  ;

The analytical approach taken for the nuclear critical evaluation was performed in two segments. The first segment was to evaluate the infinite-media multiplication constant, Iq., of a fixed-density SiO 2 soil matrix having differing degrees of 2"U and water contents or densities within the soil. These results provided indications of the combinations of 2"U, soil, and water that could support self-sustaining nuclear fission chain reactions in an l

essentially infinite sea of material (i.e., k., t 0.95). The second segment involved examining three geometries thtt have relevance to the evaluation: spheres, cylinders ofinfinite length, and slabs ofinfinite extent (Fig. 4.1).

In Fig. 4.1, the dimension r + 4 m refers to the determined critical radius plus 4 m of uncontaminated SiO2 and water, and the dimension h + 8 m refers to the thickness of the determined critical slab plus 8 m of SiO2 and water The evaluations of the infinite slabs approximate the effects of the2 "U, contaminating the soil-like waste, settling vertically onto a waste-cell floor and are consistent with previous evaluations (Hopper et al.,1995) performed for reviewing LLW facilities.

4.1.3 Parameters Consultations among Oak Ridge National Laboratory (ORNL) and U.S. Nuclear Regulatory Commission (NRC) staff, evaluating the CNSI disposal records, permitted the inference of a representative uranium enrichment for NUREG/CR-6505, 13 Vol. 2

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the nuclear criticality computational enluations. Though CNSI has been licensed to receive and dispose Jany enrichment of uranium, the assumed arbitrary representative uranium enrichment for the computational evaluations was 10 wt %2 "U in uranium.

The nuclear criticality computational studies that are reported in Vol. I were performed for 100 wt % 2"U in uranium, U(100), and water in one of two hypothetical waste matrixes. Both waste matrixes without uranium or water had identical bulk densities (i.e.,1.6 g/cm') and equivalent 0.4 void fractions. The basic matrix called

" Nominal Soil (N-S)" consisted of" average" weight fractions of earthen elements and the matrix called "SiO 2-Soil,(S-S)" consisted of only SiO 2. The nuclear criticality computational results in Vol. I demonstrated thit the ratio of critical uranium areal densities for the S-S matrix, divided by the critical uranium areal densities for the N-S matrix, was generally on the order of 0.7. The S-S matrix was chosen for this study to provide conservative estimates for the lesser uranium enrichment of 10 wt % 2"U. Furthermore, the sand at Barnwell may be more like S-S than N-S.

The reported results from this study are for various concentrations of 10 wt % 2"U in uranium, U(10), and water in a 1.6-g SiO2 /cm' hypothetical waste matrix as it may relate to the LLW facility operated by CNSI near Barnwell, S.C. The results are presented in tabular and graphic form, followed by discussions about the relevance of the results to practical initial disposal conditions. Some comparisons are drawn between the U(10)-H 20-SiO and 2 previous U(100)-H 0-SiO 2 computational 2 results that were reported in Vol.1. l The lowest 2"U concentration in the nuclear criticality evaluations is the concentration of 2"U used in Vol.1 of 2

this study, which was the permissible State of Utah license limit for "U. Although this concentration is not relevant to disposal as containerized waste, it is below the level of concern for nuclear criticality safety and can be used as a starting point. However, it is no longer a reference point for a concentration factor, as used in Vol.1.

4.2 HYDROGEOCHEMICAL MODELING 4.2.1 Conceptual Model As stated previously, a process-oriented model was developed to evaluate hydrogeochemical mechanisms that would increase the concentration of uranium in disposal settings. The model was not intended to provide site-specific predictions, but data (where available and applicable) from the Barnwell site were used in the model.

The specific process modeled was the reduction and precipitation of oxidized uranium in reducing zones. The interplay between reducing zones and oxidizing water that infiltrates requires a coupled transport and geochemistry model because, instead ofjust a single component (uranium), the transport of multiple components (e.g., uranium, oxygen, complexing agents, competitive electron acceptors such as Fe'*) is involved.

This study focused on an increase in concentration of uranium at a hypothetical boundary between oxidized and reduced zones, rather than the development of zones. The reducing agent was assumed to be either elemental iron, which represented the 55-gal drums, or methane (CH.), which represents the organics contained in the waste and the cardboard, plastic, or wood containers. This study did not evaluate kinetic aspects or time variant infiltration.

One-dimensional (1-D) transport of uranium through the trench was assumed. Even though three-dimensional (3-D) transport is more realistic, dispersion will reduce the concentration of uranium transported from one point to another, as shown in selected runs. By using 1-D flow, the results of the modeling will be conservative: the NUREG/CR-6505, 15 Vol.2

Approach Section 4 tuvel times will be shortest, and the concentrations will be maximized. Transport pathways that would mimic vertical flow through the trench and horizontal flow along the drainage systems can be modeled with two different 1-D legs.

In summary, conditions that tend to enhance the potential for increasing uranium concentration were modeled, but only ifjudged to be within reasonable bounds, or the limitations could be specified. A detailed description of model assumptions and further discussion are provided in Sect. 5.

4.2.2 Models Used Preliminary hydrogeochemical modeling was conducted using two codes: PHREEQC and ParSSim. PHREEQC (Parkhurst,1995) is a chemical speciation code that models 1-D transport using mixing cells. It has a fairly complete geochemical database, but neglects transport effects such as dispersion, which can reduce concentrations. ParSSim (Wheeler et al.,1997) runs on a supercomputer and incorporates full transport and user-defined geochemical reactions in a multidimensional. multispecies transport code. This code permits more realistic simulations, but is also less stable numerically and more time-consuming to use (in terms of run time and output analysis). Not all cases that run for PHREEQC were run successfully with ParSSim. However, a representative 3-D problem has been run successfully with ParSSim, which provided some error bounds on the simplified PHREEQC modeling.

Both codes were tested by comparison with a field and modeling problem involving nitrate removal by oxidation of pyrite, which creates a sharp redox front (Engesgaard and Kipp,1992). Numerical methods in the codes were selected to help code stability over the large concentration ranges resulting from redox problems (Toran et al.,

1997b).

4.2.3 Parameters and Model Grid The model grid represented a 1-D flow field. A continuous input of oxidized water containing dissolved uranium was introduced into a reduced zone and allowed to flow through the reduced zone at velocities of 1 x 104 and I x 104 cm/s. The model grid was 5 m long and represented a reduced zone (Fig. 4.2). The 5-m length was selected arbitrarily to monitor movement of the reduction boundary during the injection. In most cases, the boundary stayed within the model grid. The grid began at the redox boundary because this is the location where most of the geochemical alterations are presumed to occur. This approach neglects some interactions within the oxidized zone in the natural flow field, but these interactions are likely to be less important than the reactions ta'.ing place at the redox boundary.

The oxidized water assumed in this modeling exercise was geochemically similar to trench water at the CNSI site (Trench 25/21, Table 3.1) with an assumed (rather than measured) concentration of uranium. The uranium was input as a dissolved species. The uranium concentration was varied in the different runs, with values ranging from 1 to 20 mg/L. The initial uranium concentration is limited by mineral solubility. For example, in this water, dissolution of schoepite results in an equilibrium concentration of about 20 mg/L (based on PHREEQC modeling). Furthermore, tens of mg/L of dissolved uranium has been observed in water running off of uranium mill tailing piles in oxidizing environments. A lower limit of 1 mg/L uranium was selected as the input concentration for most runs, a reference concentration used in previous studies (Sims et al.,1993). Infiltrate water in the model column is also represented by the geochemistry of Trench 25/21, with a low uranium concentration of 0.01 mg/L.

NUREG/CR-6505, Vol. 2 16 l

l t

l

Section 4 Approach A , Redudag Zone i Oxidized J \

U 8 eating H20 / /

5. C.

Redudag Redudng Zone Zone

~

O -

/ /

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Figure 4.2 Schematic of model grids showing oxidized injection into reducing zone.

A. One-dimensional model used in PHREEQC and 1-D ParSSim evaluation. B. Three-dimensional "vSSim model with small source term, allowing dispersion in three dimensions.

C. Three-dimensional ParSSim model with larger source term, limiting 3-D dispersion along centerline of plume i

l I

i NUREG/CR-6505, 17 Vol. 2

Approach Section 4 The oxygen content of the trench water is not well known; Eh and dissolved oxygen were not reported in CNSI monitoring data or the hydrologic summary of Cahill (1982). A few (5) dissolved oxygen measurements were reported by Weiss and Colombo (1979), ranging from 0.1 to 1.5 mg/L. Calculations of the redox potential of the Trench 25/21 data also indicate an oxidizing Eh (on the order of 400 mV). The undersaturated conditions of the trench also suggest dissolved oxygen is present in the trench water. A base-case value of 2 mg/L dissolved oxygen was selected, and the value was varied in the sensitivity analysis (see below).

Ten components from this background water were selected to form the basic components for the geochemical modeling (Table 4.1). From these components, equilibrium species form and minerals can be selected for equilibration. For PHREEQC, the database contains an extensive list of complexed species. For ParSSim, the user identifies key complexes for equilibration, and a more limited number of complexes is favorable for faster convergence. The complexes selected for equilibration in ParSSim (Table 4.1) were based on PHREEQC modeling, which indicated those complexes formed in significant concentrations (at least one order of magnitude greater concentration than other complexes of the same component).

Table 4.1 Components and reactants (minerals and aqueous complexes) used in ParSSim" Components mg/L Products log (Ky ) Minerals / Phases log (K,,)

H' 5.9 (pH) OH- -14.00 FEMETAL -84.0 COj' 80 Hg -44.67 CH, -130.9 Ca+2 21 HCOs 10.35 UO2 -27.72 Na* 37 H2CO3 16.68 Fe2O3 -*

Mg*2 3.3 -41.018 Fe(OH)3,3 -*

U(OH)%

Fe ' O.2 UO2(CO3)j2 17.00 FeS 2 -*

Ow 2 (varied) UO2 (CO3 )% 9.63 SOl2 9.0 Fe.2 -7.76 UO'2 l.0 (varied) FeHCO3' -5.76 Cl- 42 Fe(OH)2 -5.67

• For PHREEQC components and complexes, see PHREEQC database.

• Used for PHREEQC only. Ksp not calculated for ParSSim database.

2

" Iron input as Fe'+ in ParSSim equilibrates to Fe +,

in the models PHREEQC and ParSim, minerals will not dissolve or precipitate unless they are specifically selected for equilibration with the model solutions. Minerals selected for equilibration were uraninite (the reduced uranium mineral) for the case of a carbon-reducing agent (CH.), and uraninite, hematite, amorphous iron hydroxide, and pyrite for the case of an iron-reducing agent (FEMETAL). These iron minerals resulted in a NUREG/CR-6505, Vol. 2 18

Srction 4 Approach solution that was not supersaturated with respect to any remaining minerals; however, other combinations of iron-mineral equilibration could have been used.

The reducing agents, FEMETAL (Fe') and CH., were selected to represent waste containers in the disposal site.

FEMETAL represents 55-gal dmms and other steel containers commonly used for waste, and Cit is a surrogate for organic matter, such as wood crates used in disposal. The amounts of these reducing agents were varied in the model, but were typically less than the molar content of a single barrel or wooden crate.

The FEMETAL was introduced by setting an equilibration constant with respect to elemental iron (zero oxidation st:te). The simulations were conducted with FEMETAL undersaturated in the system. When saturation with respect to FEMETAL was assumed, the model predicted a large release of reduced iron in the system, along with very low pe, and high pH (> 10). More likely, a sicwer release ofiron occurs, and this was represented in the model by setting the FEMETAL equilibrium to subsaturated conditions. A saturation index of-15 produced a rate of release of reduced iron in solution that was reducing at near-neutral pH. Although the source concentration of FEMETAL in the model is well below the total iron content of containers in the trenches, it is not known if the iron in the trenches would become localized in a zone to form a redox boundary or what the rate of reduction would be.

At Bamwell, there are fewer sources of organic matter than at municipal or industrial landfills, but early disposal practices included wood and cardboard boxes for containers. In addition, small amounts of scintillation solutions were included in the waste prior to 1976. Thus there are potential sources oforganic matter in the waste trenches that could serve as reducing agents or sorbing surfaces to increase the concentration of uranium.

Organic matter is very important in the concentration and reduction of sedimentary uranium ores (Spirakis,1996; Landais,1996; Wood,1996). In particular, sources (including tree trunks) are believed to be important (Wood, 1996). Kinetic inhibition of reduction may be overcome by heat or microbial activity. Organic matter is also common in landfills and is associated with methanogenic reducing zones.

Their dynamic data are not available for wood or cardboard, so a surrogate species must be used. The use of CH, as a reducing agent represents an end-member composition for these materials, although it is probably more reducing than wood or cardboard, and is readily dissolved in water. A partial pressure of CH. gas was miintained such that there was 2 mg/L in solution.

Precipitation in the reducing zone was controlled by the mineral uraninite, a commonly observed uranium mineral in reducing zones of natural ore deposits. Model runs were typically 500 L of pore fluid, which represented 80 years at present-day infiltration rates. Runs representing times up to 1,000 and 10,000 years were also conducted. The longer times were run only with PHREEQC.

4.2.4 Sensitivity Analysis A series of runs were conducted to help provide bounds on conditions that could limit or enhance the potential for nuclear criticality safety concerns. The parameters selected for evaluation were the initial concentration of dissolved uranium, the concentration of oxygen, the amount and size of reducing zones, additional uranium complexes, precipitation of uranium silicate minerals, dispersion, and 3-D source terms. Most of the sensitivity anilyses were conducted with PHREEQC. A more limited number of runs to evaluate dispersion, dimension, and size were done with ParSSim (Fib. 4.2). The range in parameters was selected to span likely field conditions rather than worst-case scenarios. Thus the uranium concentration was varied from 1 to 20 mg/L (an upper limit of observed values around waste sites); the oxygen concentration varied from 1 to 8 mg/L (the upper limit being I l

NUREG/CR-6505, 19 Vol.2

Approach Section 4 the solubility limit for oxygen dissolved in water); the leachate complexes were chosen based on observed values for the anions nitrate and fluoride at the Barnwell site.

NUREG/CR-6505, l Vol. 2 20 l

5 ASSUMPTIONS All models are simplifications of reality. Some of the processes not explicitly modeled in the simulations are evaluated here. He scenarios modeled typically represent bounding calculations.

l 0 Saturated, steady-state flow The models assumed saturated, steady-state flow to approximate conditions over the long time frame. These l conditions represent worst-case scenarios because unsaturated, transient flow likely involves longer travel times.

I o Geochemical complexes and solid phases The geochemical complexes and solid phases that were selected for equilibration were based on a knowledge of important uranium species and what is known about site chemistry. The decision was made to focus on the r:duced form of the uranium (i.e., uraninite). Reactions among species were assumed to be in equilibrium.

Although additional complexes and phases could change the mobility ofreactants, and kinetic considerations could change the rates, the calculations represent likely bounds and best available data.

o Simplistic deposit geometrie; Simplistic deposit geometries, having no density gradients, were used in the criticality assessment. That is the "S

U and source material were assumed to be uniformly distributed over the volume of the trenches. Smaller quantities of fissile material in equivalent volumes may be required to reach criticality for certain density gradients. An extreme, but actual, critical experiment performed by Morfitt (1953) was the assembly of five cone:ntric cylindrical uranyl fluoride solution regions having variable densities of 93 wt % enriched uranium.

Solution uranium densities were selected to produce a nearly uniform thermal neutron core flux. Doing so produced a critical system with 1061 g"SU as compared with a homogeneous core mass of 1162 g"5U in an equal volume.

O Enrichment For the nuclear c.iticality calculations, a "5U enrichment of 10 wt % is assumed. Based on disposal records (see Section 8 and Table 8.1), this represents an arbitrary upward bound on likely enrichments in the trench.

Bec:use ofincomplete records and a lack of regulatory limits on enrichment, a theoretical upper bound of 100 wt % enrichment is conceivable, but not likely. The case of 100 wt % enrichment was modeled previously in Vol.1, and the calculations presented there are applicable to issues of this study. In evaluating individual trenches, the average enrichment was assumed to be appropriate.

1 Enrichment does not affect hydrogeochemical mobility of uranium. Relative enrichments of"5U are assumed to i rem:in constant between the source and the precipitated secondary phases. The geochemical model accounts for  !

total uranium, irrespective of the "5U isotopic enrichment. However, because the amount of dissolved uranium associated with the waste can span an order of magnitude from 1 to 10 mg/L, one can interpret the I mg/L as  !

100 wt % enrichment and the 10 mg/L as 10 wt % enrichment that have equivalentusU densities (i.e.,1 mg "S l U/L). The geochemical models deal only with chemical species and not isotopes.

NUREG/CR-6505, 21 Vol. 2

Assumptions - Section 5 o Reducingzones It was assumed in this analysis that reducing zones exist within the disposal environment. The development and

' longevity of reducing zones was not specifically modeled in this analysis because of a lack of site-specific information on weathering rates of containers, amount of organics in specific trenches, and 3-D data on chemical zonation within and around the trenches. Reducing zones also require a water-saturated environment, which is not currently present in the trenches. Thus the reducing zones represent a worst-case scenario of future conditions, which would enhance precipitation of uranium.

o Containerdegradation For the reasons given above, the degradation of the waste containers was not modeled explicitly in this analysis; all of the source term was assumed to be exposed to migrating fluids. Realistically, degradation is fast relative to the time required for significant chemical migration. Steel can degrade in tens ofyears, concrete can crack, and high-density polyethylene (HDPE) degrades in hundreds of years.

o Horizontal and preferential flow paths An important assumption regarding the mobilization of uranium in trenches is that multiple vertical flow paths can be funneled into a demobilization zone. This funneling occurs through a horizontal component of flow.

Although the areal density disposal limits and the package limits mitigate development of a critical mass along a vertical path, horizontal flow along the base of the trench may result in the concentration of material within the drainage system or the sumps. In this exercise, two-dimensional (2-D) flow was modeled as a combination of 1-D segments.

NUREG/CR-6505, Vol. 2 22

6 CRITICALITY SAFETY EVALUATION RESULTS 6,1 235 U ENRICHMENT INFLUENCE ON CRITICAL MASS OF URANIUM 2

The critical mass of 2"U is inversely proportional to the "U enrichment of the uranium. That is, all else being equal (e.g., degree of neutron water moderation, volume, chemical composition, temperature, relative shape, etc.)

2 as the weight percent "U enrichment in uranium decreases, the critical mass of 2nU increases. The following figures are taken from an illustrative study (Jordan and Turner,1992) of this effect relative to the degree of 2

neutron water moderation, expressed as the ratio of hydrogen atoms to "U atoms (EUX), for fully water-reflected spheres of homogeneous UO F 2 2 -H 2 O solutions. The figures are provided to illustrate the difference between2 "U m .ss response to moderation and uranium mass response to moderation at various2nU enrichments. Figure 6.1 2

demonstrates that for a given H/X of 500 the critical 2"U mass is estimated to be about 0.76 kg "U for U(100),

2 whereas the critical 2"U mass is estimated to be about 1.2 kg "U for U(10). Figure 6.2 demonstrates that for a given H/X of 500 the critical uranium mass is also estimated to be about 0.76 kg uranium for U(100), whereas the critical uranium mass is estimated to be about 10.2 kg uranium for U(10). Thus for an H/X = 500, the ratio of the minimum critical masses of2"U for U(100) and U(10) is about 0.63 (0.76/1.2). As can be observed in Figs. 6.1 and 6.2, for water moderation less than an IVX = 500, the optimum for minimum2"U mass results in significantly smaller ratios of U(100):U(10) critical masses. For instance, at a poorly moderated H/X of 20 for U(100), the critical mass is estimated to be about 6.2 kg2 "U, whereas for U(10) the critical mass is estimated to 2 2 be about 31.0 kg "U, a "U mass ratio of 0.2 or uranium mass ratio of 0.02. These effects are the direct result of the slowing-down power of water for high fission-energy neutrons escaping the neutron resonance capture ch:racteristic of 2"U in the U(10). With less water present in the mixture, neutrons cannot be thermalized as r pidly as when present in optimum ratios of water to 2"U, about H/X = 500, and therefore the neutrons are captured by 2"U resonance capture.

6,2 TRENDS FOR U(10)--H2 0-SIOz MIXTURES Similar trends for increased uranium and 2"U mass are present for the mixtures of U(10)-H 2 0-SiO2 presented in this report. The obvious differences between the above-referenced illustrative study and this study are the assumption of elemental uranium (i.e., no chemical compounds were considered) and the presence of silicon dioxide at a fixed density of 1.6 g SiO2 /cm'. The presence of 90 wt % 2"U in the uranium in these calculations increased the importance of SiO2 as a poor moderator of neutrons and a poor capturer of neutrons. For example, 2

in the studies with U(10) having no water, the resonance capture of neutrons by the "U was important and suppressed the infinite-media neutron multiplication factor, k., from 0.955 for U(100) to 0.713 for U(10) at a2 "U 2

concentration of 0.000886 g "U/g SiO2 . A slight introduction of water to the matrix (i.e.,0.01813 g H2 0/g SiO2 )

for both systems decreased the k,. for the U(100) to 0.867, whereas the k. was increased for the U(10) to 0.779.

This response is the direct result of the SiO2 affording excess neutron slowing down for the U(100) matrix, thereby allowing hydrogen neutron capture to suppress k,.. In the case of the U(10), the SiO2 Provides excess neutron slowing down for the U(10) matrix, but these neutrons are forced to slow in energy into the2nU resonance capture region and be captured. The slight addition of water to the U(10) matrix can slow neutrons by large amounts and thereby permit neutrons to be moderated or slowed to energies that are less than the2nU

]

resonance capture energies, thereby offsetting some of the negative effects of the SiO2 over moderation. l l

I Ttble 6.1 provides extracted results of the computations to highlight the extremes of this study. The complete listing of the results of this study are provided in Table A.1 of the Appendix to this report. The results are provided in the same format that was used in Vol.1, Appendix C. Figure 6.3 provides an interpolated and smoothed surface plot for the critical infinite slab areal density vs H2 O and 2"U concentrations relative to SiO2 -

NUIEG/CR-6505, 23 Vol.2

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Section 6 Criticality Results 9

g1120/g SiO2 1

.15

.25 ii','c. ,c cc #

Y',' ,.'

l', ' '

y Slab Areal Densky 10 (kg U235/m^2) 0 0.003 j l

e.01 log scale of(g U235/gSiO2)

\ .

I 2

Figure 63 Infinite slab areal density (kg235U/m ) vs g H 20/g SiO and 2 log scale of g 235U/g SiO2 NUREG/CR-6505, 27 Vol. 2

Criticality Results Section 6 Figures A.1 through A.11 provide interpolated and smoothed surface plots of the data provided in Table A.l.

These plots are provided in the same format that was used in Vol.1, Appendix D. In future comparisons, care should be exercised to recognize that even though there are similarities between the general appearances of the surface plots of the U(100) and the U(10), the significantly larger"'U values are for uranium enriched to only 10 wt % n5U in the uranium.

6.3 COUPLING OF NUCLEAR CRITICALITY AND HYDROGEOCHEMICAL MODELING The nuclear criticality safety calculations were used to establish the configurations and associated uranium concentration increases that are required to reach a level of concern. Then, using the hydrogeochemical modeling approach and assumptions discussed in Sections 4 and 5, the time required to increase the uranium concentration from the reported disposal values to one of the minimum concentration levels of concern was determined. Several minimum values of the critical mass were selected to provide a conservative scenario as benchmarks for evaluating the potential of developing a critical mass of"5U under the conditions of SiO2 matrix, water, and 10 wt % enrichment "'U. The lowest uranium concentration that could support a nuclear reaction was in a nearly dry system: 0.0024761 g"5U/cm' and 0.029 g 2H 0/cm'(Table A.1,line entry 69). This concentration requires a slab 464.2 cm thick, or 1379 kg, ifin a spherical configuration. By contrast, concentrations of 0.0224 gn5U/cm2 and 0.2817 gn5U/cm2under almost fully saturated conditions would require slabs of 19.7 cm and 9 cm thick, respectively. The hydrogeochemical model assumes a water-filled porosity of 40%. If uranium precipitates under wet conditions, then the system would need to dry out in order to achieve the minimal configuration necessary to support criticality. In order to develop the higher uranium concentrations, funneling of source material by horizontal transport is necessary.

i NUREG/CR-6505, Vol. 2 28

7 HYDROGEOCHEMICAL MODELING RESULTS Hydrogeochemical modeling results are presented as a comparison of simulations with different parameters.

The evolution of uranium precipitation over time is used as the basis for comparison (Table 7.1). PHREEQC modeling is conducted in terms of pore volumes, which can be translated into years by assuming a groundwater velocity.

4 A pore volume in the model was assumed to be 1 L of water. Using an assumed groundwater velocity of 10 cm/s (1 m/ year), most simulations were conducted for 80 years (500 pore volumes) with selected simulations up to thousands ofyears. Relevance of the modeling to site conditions at Barnwell is discussed in Section 8.

7.1 REDUCING ZONES Th2 reducing zones maintained a low redox state over the duration of modeled influx of oxidized water (up to 30,000 L for the longest simulation; typically 500 L). The CH. created a reducing zone that had a pe of -2.7 with a pH of 6, whereas FEMETAL (Fe') affected a reducing zone with a pe of -1.9 and a pH of 6. The rate of release of the reducing agents was on the order c 3 mg/L per year for methane (Fig. 7.1) and 114 mg/L per year for FEMETAL (a small fraction of a barrel). Th values are derived from the modeled conditions and are not meant to represent the actual release rates in a disposal setting. These modeled release rates are not likely to consume the supply of wood from disposal crates or metal from weathering drums. However, the kinetics of weathering of these disposal containers is not well known. Both types of reactions could be mediated by bacteria, providing accelerated weathering rates and formation of reducing zones.

The reducing zones as modeled capture very close to 100% of the uranium input (e.g.,99.998% for Run 0),

producing a linear increase in the amount of uranium precipitated over time (Fig. 7.1). One reason the reducing zones are efficient at capturing uranium is that reduced uranium minerals, such as uraninite, have low solubility.

Another important factor is that there were modest amounts of complexing agents present in the observed trench water.

i 7.2 SENSITIVITY ANALYSIS For the conditions simulated, the most significant parameter in predicting the amount of uranium precipitation was the initial source of uranium or the dissolved concentration associated with wastewater filtering through dry waste.

The uraninite precipitation was directly related to the initial concentration of uranium, and complete precipitation occurred for initial concentrations up to 20 mg/L (maximum modeled). The uraninite precipitation was not sensitive to the oxygen concentration in the infiltrating water, which has the potential to compete with uranium for electron acceptor sites. The concentration of oxygen dissolved in water can vary only over a small range; so it is not a sensitive parameter. The model predicted essentially the same amount of precipitation far cases of half the concentration of the reducing agent. Small, but realistic, amounts of complexing agents were added with no significant effect. In particular, increasing the bicarbonate concentration from 80 to 400 mg/L did not inhibit uraninite precipitation. The uranyl-carbonate complex is not dominant under reducing conditions. Sensitivity to the precipitation of silicate minerals and hydrodynamic dispersion is discussed in the following sections.

1 NUREG/CR-6505, l 29 Vol. 2 l

Hydrogeochemical Modeling Results Section 7 Table 7.1 Parameter variation and results of hydrogeochemical modeling Run No. Amendment Value U PPT CH4 U PPT FE 0 Base case (one set of runs for each of See below 2 x 10d 2 x 10" two reducing agents: FeMetal and CH4 )'

d 1 Vary uranium in leachate 2 mg/L 4 x 10 4 x 10d 2 10 mg/L 2 x 10-' 2 x 10-'

• 20 mg/L 4 x 10-2 4 x 10-'

4 Vary oxygen in leachate 2 mg/L 2 x 10d 2 x 10" 5 8 mgL 2 x 10d 2 x 10" 6 Vary reducing zone I mg/L CH. or 1 M 1.4 x 10d 2 x 10" d

FeMetal (1.4 x 10 )

d 7 I cell reducing 2 x 10 2 x 10d 8 Vary carbonate concentration 240 mg/L HCO; 2 x 10" 2 x 10" 400 mg/L HCO; 2 x lod 2 x 10" 9a Vary leachate complexes Add 2 mg/L F 2 x 10" 2 x 10d d

9b Add 20 mg/L NO3 2 x 10 2 x 10" d

10 Vary precipitate (ppt in zone of higher Si, ppt. coffinite 2 x 10 2 x 10" equilibrated with quartz) d 11 ppt. soddyite 1.0 x 10 12a incorporate transport (dispersion) 1-D transport in ParSSim 1.8 x 10" 12b l-D, slower velocity 3.8 x 10-5 12c l-D reducing zone halfway 1 x 10" down grid 13 3-D transpon with small source 2.6 x 10-5 area 14 3-D transport with large source 1.3 x 10-5 area 15 Larger dispersivity 2.9 x 10 d 5

(1.4 x 10 )

8 Base Case: 1 mg/L U in leachate Reducing Agent 1: 10 M FeMetal(equilibrated, but undersaturated to create a slow release)

Reducing Agent 2: 2 mg/L CH., constant source 2 mg/L 0, 80 mg/L ffCO; 5 cells reducing Equilibration minerals: uraninite, also for FeMetal: pyrite, hematite, Fe(OH)3 U PPT = g U-mineral precipitated per em' soil using CH, or Fe' as reducing agents.

For runs with number shown in parentheses, the run did not go to completion. Value should be compared with value shown in parentheses for equivalent timestep.

NUREG/CR-6505, Vol. 2 30 l

l

i 4

Section 7 Hydrogeochemical Modeling Results o.s4 d o.12 o'

- i 8 ,

e .., .-

h '

o 4

I O

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7 7

3 0.002 O

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/ 1000 20'00 3000 4000 5000 6000 7000 8000 9dOO 10000 Twne, years (b)

Figure 7.1 Reaction progress for hydrogeochemical model: (a) Cumulative CH4 release vs time for short time frame;(b) Cumulative uraninite precipitation for long time 4

frame. Times shown are based on a velocity of 3 x 10 cm/s and a porosity of 0.4 1

l l

I

)

NUREG/CR-6505, i 31 Vol. 2

Hydrogeochemicci Modeling Results Section 7 7.3 PRECIPITATION OF SILICATE MINERALS Some alternative uranium mineral precipitates were considered by including silica in the model. The silica in the infiltrating water is expected to be relatively low, given that the source area contains waste rather than soil.

- However, when the leachste reaches the drains at the base of the trench they will encounter a zone of backfill j containing sand. This is likely to result in water in equilibrium with an amorphous silicon dioxie such as q chalcedony. The scenario modeled was that of water infiltrating into a zone in which chalcedony dissolution occurs, increasing the silica concentration up to about 6 mg/L. Precipitation of uranyl silicates was modeled with and without reducing agents present. When the uranium infiltrates this zone, uranium silicates precipitate. He minerals considered in this analysis were soddyite, an oxidized uranium silicate, and coffinite, a partially reduced wanium silicate. The soddyite does not completely precipitate all of the available uranium because it has a higher solubility than uraninite. The coffinite captured the uranium as efficiently as the uraninite.

7.4 3-D HYDROGEOCHEMICAL MODEL The main error induced by using a mixing cell model, such as PHREEQC, instead of a coupled geochemistry and transport code, such as ParSSim, is that dispersion is neglected in the mixing celt model. Dispersion is a function of the dispersivity and the velocity. It can occur in three dimensions, although the two transverse directions dispersivity tends to be smaller in the transverse directions rather than the longitudinal direction (in the direction of flow). The effects of dispersion on transport and reconcentration were evaluated and compared in a variety of ways: (1) 1-D vs 3-D dispersion, representing an infinite source and a point source; (2) dispersion along the flow p;.th before the reducing zone is encountered; (3) longitudinal and transverse dispersion (ratio 3:1) after release from a 1-m by 1-m source area; (4) longitudinal and transverse dispersion after release from a larger source,3 m by 1 m; and (5) fast vs slow velocities. Rese runs were conducted with ParSSim, using the same background solutions and the same equilibration minerals as in the CH reducing zone case.

He results (Table 7.1) indicated that 1-D dispersion reduces the amount of uraninite precipitated but that the amounts are reduced less than an order of magnitude, assuming typical scaled values for dispersivity (0.2 m).

The errors due to neglecting dispersion in one dimension are relatively small compared with other uncertainties,

.,uch as the kinetics of weathering, mineral dissolution, and precipitation, and the size and distribution of source t:rm. Three-dimensional dispersion results in a decrease in the amount of uramum precipitation of approximately an order of magnitude. This decline in concentration of precipitated uranium indicates that isolated sources of uranium that can disperse in three dimensions (Fig. 4.2) will require (1) much longer times to reach levels of concern for nuclear criticality safety, (2) larger source terms, or (3) may never develop concentrations sufficient to pose a criticality safety concern. Table 7.1 indicates that the model results are not highly sensitive to certain parameters (e.g., velocity (Run Nos.12a and 12b), source term (Run Nos.13 and 14) and flow path length (Run Nos.12a and 12c)}.

NUREG/CR-6505, Vol. 2 32

8 DISCUSSION Several conditions restrict the potential for a criticality safety concern in specific disposal trenches at the Btrnwell LLW disposal facility. Among these are the enrichment, the source term, the chemistry of water within the trenches, and the potential flow paths. Because of uncertainties associated with the disposal conditions, the uncertainties associated with the results of this analysis need to be underscored. The basic geochemical process, precipitation in reducing zones, does not appear to be a limiting factor based on the current model assumptions.

However, long times may be needed. Uncertainties and other mitigating factors are also discussed.

8,1 ENRICHMENT The average enrichment of homogenized material within each trench is available only for "new trenches," Nos.

38 through 87. For the trenches with enrichment data (Fig. 8.1), the average2 "U enrichment is 0.5 wt %, and only six trenches have more than the minimum I wt % 2"U enrichment needed to reach a critical configuration.

Trench 66 has a reported 2"U enrichment of 100 wt %. It isjudged from historic and current disposal practices and data (see Section 3 and Appendix B) that SNM and source material disposals are comingled in the trenches.

B: cruse of this comingling and dispersion of SNM, very few of these trenches need to be considered as criticality safety concerns. Again, this is based upon the reasonable assumption that2 "U and nonfissile 2nU behave in a similar geochemiet.1 fashion.

For the older trenches (trench numbers less than 38), only grams of SNM (Table 8.1) were reported. Since uranium is the most common SNM, to simplify the calculations, it was assumed that all of the SNM is2 "U even though minor amounts of plutonium may have been disposed ofin the trenches. No data were available on amounts of nonfissile uranium, so again the conservative assumption is that the uranium is all2"U and the enrichment is 100 wt % (Fig. 8.1). Two of these trenches have no reported SNM, which leaves 35 trenches with uncertain enrichment. Most of these trenches are unlikely to have high enrichments, although Trench 23 is an exception. The response to a specific NRC staffinquiry resulted in the identification of relative quantities and i

proportions of SNM, source material (SM) and byproduct material activity (BPM) that were placed in Trench 23.

The reference to and summary of that response are provided in Appendix B, Suberiticality Evaluation for Chem-Nuclear Systems, Inc., Trench 23.

8,2 SOURCE fERM The minimum mass of2"U to achieve criticality was calculated and compared with the total mass disposed in each trench. The minimum mass for a sphere provides a lower bound on source calculations, and these are shown in Table A.I. For slabs, the minimum mass for the source can be calculated assuming the length of each side of an " infinite" slab is ten times the calculated critical thickness. These finite volumes are multiplied by the critical uranium concentration to give total mass required fcr a potential criticality concern. These uranium mass values were used to determine the time frames needed to achieve three selected uranium densities: the minimum cone:ntration to achieve criticality (0.0024 g/cm'), approximately 10 times that density (0.022 g/cm'), and cpproximately 100 times that density (the maximum calculated in Table A.1,0.28 g/cnf).

For spheres, the critical masses range from 3 kg to 2000 kg over these density ranges (Fig. 8.2). For slabs, the critical masses vary from 20 kg to 6000 kg (Fig. 8.2). With the exception of Trench 66, the "new" trenches that contain SNM with 2"U enrichments greater than I wt % have less than 30 g of 2"U (Table 8.1), which is well below the minimum critical masses calculated. Trench 66 has a reported source term of 2.3 kg of 100 wt %

NUREG/CR-6505, 33 Vol. 2

i Discussion Section 8 L

8-k 5 - o U-235

• SNM O

g to -

c W

b E

C uJ , _

O O

cy - 0 0

0 O 0 a _

0 g eOh 0 0 0 200 400 600 800 1000 U-235 kg or SNM disposed per trench kg Figure 8.1 Relationship between enrichment and source term NUREG/CR-6505, Vol. 2 34 i

l

Section 8 Discussion l

l Table 8.1 Disposal records from Barnwell. S.C., and calculated enrichments and density" 23 235 SNM mass 235U U U Enrichment Disposal Density' No. (g) (R) _ (n) (wt %) volume (ft')* g/cm' 1 1.32e+04 4.2e+04 1.11e-05 2 3.73e+04 1.le+05 1.20e-05 3 1.0le+04 2.3e+04 1.52e-05 4 4.66e+04 1.3e+05 1.26e-05 5 1.71e+04 1.8e+05 3.36e-06 6 2.53e+04 2.4e+05 3.66e-06 7 3.35e+04 2.4e+05 5.00e-06 ,

8 2.45e+04 2.0e+05 4.29e-06 9 4.10e+04 1.9e+05 7.47e-06 10 2.91e+04 2.0e+05 5.18e-06 11 2.91e+04 1.5e+05 6.81e-06 12 1.73e+04 2.0e+05 3.09e-06 13 1.49e+04 2.le+05 2.50e-06 14 2.46e+04 2.2e+05 3.95e-06 15 3.43e+04 2.8e+05 4.32e-06 16 1.62e+04 2.6e+05 2.20e-06 17 1.30e+04 1.8e+05 2.57e-06 18 1.78e+04 1.7e+05 3.63e-06 19 2.51e+04 2.4e+05 3.69e-06 20 1.07e+04 2.4t. F05 1.61e-06 21 1.42e+04

22. 8.04e+04 9.2e+05 3.10e-06 23 1.75e+05 9.8e+05 4.91e-06 24 5.0le+02 2.6e+05 6.73e-08 25 1.42c+04 2.7e+05 1.83e-06 26 7.3 ie+04 8.9e+05 2.90e-06 27 1.44e+04 2.le+05 2.39e-06 j 28 8.30e+04 8.9e+05 3.29e-06 29 7.31e+04 1.2e+06 2.24e-06 "O 6.86e+04 4.8e+05 5.07e-06 31 1.39e+04 2.5e+05 1.93e-06 32 5.56e+04 7.le+05 2.76e-06 33 7.60e+04 6.6e+05 4.07e 06 34 3.78e+04 5.4e+05 2.47e-06 35 7.04e+04 7.2e+05 3.47e-06 36 NUREG/CR-6505, 35 vol.2

Discussion Section 8 Table 8.1 (continued)

SNM mass 2"U "' U 2"U Enrichment Disposal Density' No. (c) (g) (a) (wt %) volume (ft'f g/cm2 37 38 9.18e*07 2.59e+05 9.16e+07 0.28 5.le+05 1.79e-05 39 1.10e+08 4.16e+05 1.09e+08 0.38 5.5e+05 2.66e-05 40 9.16e+04 9.16e+04 1.8e+04 41 1.68e+08 4.17e+05 1.68e+08 0.25 5.8e+05 2.55e-05 42 2.93e+08 2.87e+05 2.92e+08 0.10 6.le+05 1.67e-05 43 3.88e+08 ' 3.02e+05 3.88e+08 0.08 7.8e+05 1.37e-05 44 2.81e+08 3.79e+05 2.81e+08 0.13 7.2e+05 1.85e-05 45 3.42e+03 4.63e+00 3.42e+03 0.14 1.0e+05 1.61e-09 46 1.14e+08 3.66e+05 1.14e+08 0.32 5.7e+05 2.28e-05 47 5.65e+02 5.65e@2 4.1e+04 48 1.96e+03 2.78e+01 1.93e+03 1.42 6.4e+04 1.54e-08 49 6.7e+04 50 9.4e+04 51 9.93e+07 4.70e+05 9.88e+07 0.47 6.0e+05 2.74e-05 52 7.95e+06 7.95e+06 8.2e+04 53 2.61e+08 1.78e+05 2.61e+08 0.07 5.3e+05 1.19e-05 54 6.02e+08 1.90e+05 6.02e+08 0.03 6.4e+05 1.05e-05 55 2.13e+02 4.63e+00 2.08e+02 2.17 8.6e+04 1.91e-09 56 3.87e+02 3.87e+02 9.9e+04 57 6.30e*03 6.30e+03 8.3e+04 58 5.68e+08 3.94e+05 5.68e+08 0.07 7.3e+05 1.90e-05 59 60 1.le+03 61 9.90e+0S 1.76e+05 9.90e+08 0.02 7.7e+05 8.05e-06 62 5.14e+08 7.27e+05 5.13e+08 0.14 7.5e+05 3.40e-05 63 5.11e+08 1.07e+06 5.10e+08 0.21 7.7e+05 4.90e-05 64 2.42e+08 7.57e+05 2.41e+08 0.31 5.le#05 5.22e-05 65 2.2e@4 66 2.32e+03 2.32e+03 0.00e+00 100.0 2.0e+04 4.12e-06 67 1.24e+02 4.63e+00 1.19e+02 3.74 1.4e+04 1.15e-08 68 4.30e+02 1.39e+01 4.16e+02 3.23 1.9e+04 2.61e-08 69 5.7e+03 70 1.22e+08 2.61e+04 1.22e+08 0.02 2.2e+05 4.23 e-06 71 3.00e+08 3.30e+05 3.00e+08 0.11 6.le+05 1.90e-05 72 3.05e+08 1.83e+05 3.05e+08 0.06 1.0e+06 6.30e-06 NUREG/CR-6505, Vol. 2 36

Section 8 Discussion Table 8.1 (continued) 23 SNM mass 2"U U 2"U Enrichment Disposal Density' No. (a) (2) (g) (wt %) volume (ft')* g/cm' 73 2.51e+07 ' 3.08e+04 2.51e+07 0.12 3.7e+05 2.91e-06 74 4.00e+07 - 8.59e+04 4.00e+07 0.21 3.8e+05 8.06e-06

.75 5.4e+03 76 1.3e+04 77- 6.41e+01 4.63e+00 5.95e+01 7.22 1.2e+04 1.36e-08 78 6.0e+03 79 1.95e+03 1.85e+01 1.93e+03 0.95 3.3e+04 2.00e-08 80 81 2.13e+04 1.06e+02 2.12e+04 0.50 3.le+04 1.20e-07

~82 3.68e+05 6. Ole +01 3.68e+05 0.02 3.0e+04 7.16e-08 83 2.21e+05 1.14e+03 2.20e+05 - 0.52 2.6e+04 1.54e-06 84 )

85 4.52e+06 2.0le+03 4.52e+06 0.04 2.le+05 3.46e-07 86 9.88e+05 3.16e+03 9.84e+05 0.32 5.7e+04 1.96e-06 l 87 1.95e+04 1.95e+04 2.5e+04 l

• SNM masses for Trench 38 and greater are determined by summing uranium isotope masses.

Trench numbers less than 38 are reported SNM. Blank indicates data not reponed or could not be calculated (enrichment or density) from available data.

,

• Values reported in units of cubic feet. Cubic meters can be determined by multiplying cubic feet by 0.0283.

' Density for Trench 38 and greater is the density of 2"U in g/cm'. For trench numbers less than 38, the density is the g/cm' of SNM. Blank values for the density indicate the volume or required mass was not reported.

d 4 Read as 1.32 x 10.

4 i

l l

i NUREG/CR-6505, 37 Vol. 2

Discussion Section 8 I

l l

I 10' ' '  !  ! ' ' ' '

"" " stab: 4 SE-3 g"Ulcm'

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/ O stab: 2.8E-1 g"U/cm'

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I i i i t 0 0.15 0.2 0.25 0.3 0.35 04 Water contents (g H2 0/cm')

Figure 8.2 Critical (k, 2 0.95) masses of spheres and " pseudo-infinite" slabs and cylinders as a function of water content for different densities of 235U in waste matrix NUREG/CR-6505, Vol. 2 38

j Section 8 Discussion l cnriched 2"U, within the range for critical spheres. Because it is difficult to form a sphere under the transport conditions existing in a disposal trench, this small mass does not pose a criticality safety concern. The "old" trenches with unreported enrichments contained 10 to 175 kg of SNM (Trench 23), with a mean value of 30 kg.

These masses are within the range of critical masses for spheres and for some slab configurations, assuming2 "U enrichments of 10 wt %. Note that larger masses will be needed for enrichments between 1 and 10 wt %.

Distribution of the source term within a given trench is an important component in the potential to develop a critical mass. License conditions in effect between 1970 and 1981 limited the mass of 2"U in a package, as well as gave some constraints on disposal. Prior to 1977, package limits varied between 15 and 50 g. The areal density of the material in the "old" trenches is less than 0.05 kg/m2 , less than one-third of the minimum "new" 2

trench areal density of 0.16 kg/m (see Fig. 8.3). Although the enrichment of the early disposed material is not known, the smaller masses of SNM, the lower areal densities, and lower package limits indicate that the SNM is l more dispersed in the "old" trenches (1-37) than in the "new" trenches (38-87). Therefore, while the older I trenches have an inventory of SNM that theoretically would be sufficient to form a critical mass at 100 wt %

enrichment, it is likely that the SNM is dispersed enough that significant funneling would be required to j reconcentrate the2"U into a critical mass. 1 Tr:nch 23 is an exception. Trench 23 contains 175 kg of highly enriched material and sufficient quantity to form cither a ghere or slab with a critical mass. However, a subsequent evaluation of Trench 23 was perfonned (see I Appendix B, "Suberiticality Evaluation for Chem-Nuclear Systems, Inc. Trench 23") that demonstrates that criticality would not be achieved because of:

(1) the limited thickness of the disposal burials (i.e., approximately 4 to 11 ft),

2 (2) the conservatively evaluated low "U densities (e.g., between a maximum of about 0.00168 g2 "U/cm' for 350 g SNM in a single 55-gal container to a generally expected global density ofless than about 2.5 x 105 g

> 2"U/cm' within the trench), and j (3) the significant quantity of source material (i.e.,210 metric tons of normal or depleted uranium or thorium),

that is, codisposed with the 175 kg SNM that results in an homogenized trench averaged 2"U enrichment of l about 0.08 wt %. I 83 GEOCHEMICAL PROCESSES l To increase the concentration of uranium, a geochemical process is needed that will capture uranium in a solid phise. Evidence from ore bodies indicates that uranium concentrations can increase through sorption onto a substrate or by precipitation as reduced uranium minerals. Because sorption was evaluated under a variety of conditions in Volume 1 of this study, this analysis assumed precipitation of uranium minerals.

One important result of the geochemical modeling is that given reasonable amounts of reducing agents, reducing zones are stable in a water-saturated environment (with limited oxygen). A more realistic approach to modeling the redox conditions in the trench would be to model transient wet / dry cycles, and weathering of drums and wood cr:tes. It is known that such containers weather fast enough to be breeched in recent disposal history. However, the weathered containers are still present, not completely weathered away. The remnants of such containers will serve as surfaces for sorption or redox-driven precipitation of transported uranium.

NUREG/CR-6505, 39 Vol.2

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! amount of water transporting dissolved uranium. Because of solubility limitations, the concentration of uranium dissolved in pore water is significantly less than the initial concentration on the waste. Thus the amount of fluid required to flush through the waste matrix in order to dissolve sufficient uranium for developing a critical mass is quite large.

3 To mobilize the 4.4-kg requisite to form a slab with a critical density of 0.0224 g/cm (line 190), with a solubility of 10 mg U/L (or 1 mg 235U/L, assuming 10 wt % enrichment),3.5 x 10' L is needed. For a flow rate of 1 m/ year, 2

the volume of water to pass through 1 m (1-m footprint) of waste with 40% porosity is 400 L per year.

2 The critical slab has a footprint of approximately 4 m and a thickness of approximately 20 cm. Assuming a flow rate of 1 m/ year, almost 11,000 years are needed to mobilize the 4.4 kg. Unsaturated flow, or intermittent flow through the trench will also lengthen this estimated mobilization time.

The amount of reducing agent available was not a limitation in this analysis. Only a few grams are consumed in uranium reduction per pore volume, much less than the mass ofiron- or organic-rich containers present in t',

trenches. Because the reductant was not a limiting reactant, the modeled geochemical reactions were not able to provide a constraint on the conditions for mobilizing and inunobilizing 235 U. However, other changes in hydrogeochemical conditions may occur over the time period of disposal.

8.4 HORIZONTAL-VS-VERTICAL FLOWPATHS The initial areal densities within the waste trenches were estimated from the disposal records. The areal densities 2

were calculated by assmning that the waste concentration is projected on a 1-m footprint (Fig. 8.4). The average areal densities for the Bamwell facility trenches are significantly less than the minimal areal density required for criticality in a slab configuration (Table A.1, Appendix A), indicating that simple 1-D vertical flow paths will be insufficient to produce a critical areal density. Therefore, funneling of vertical flow paths or channeling of flow paths into horizontal flow (drainage system) must be considered.

He potential for horizontal flow in unsaturated soils has been evaluated by Kung (1990a,b). Kung (1990s,b) dJineates three types of preferential flow in the unsaturated zone: short circuiting, fingering, and funneling. '

Shert circuiting is the concentration of flow in macropores or fractures; fingering is the splitting of flow paths due to instabilities; and funneling is the horizontal movement and combining of flow paths caused by heterogeneities.

l Kung gives field evidence for funneling using dye tracers in sandy soils. Typically, a coarse unit will form a barrier to movement because of the large suction required to penetrate the larger pores; the penetration of this barrier is related to the hydraulic conductivity, the slope of the unit, and the height of the available water column. If funneling occurs, there is no need for saturated flow to develop. If flow occurs in 50% of the matrix, this amount is equivalent to increasing the capture zone for potential source material by a factor of 2, and if flow occurs in 1% of the matrix, the capture zone could be nearly a factor of 100 larger than assumed for strictly 1-D flow.

In general, the flow through the trenches at Barnwell is vertical. However, due to local heterogeneity and the transient nature of vertical groundwater flux, some horizontal flow may occur. Horizontal flow could occur within the waste zone or at the base of the trenches (Fig. 8.5). Horizontal flow is anticipated in the drainage systems along the base of the trench.

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I l

Discussion Section 8 Direct evidence indicates that horizontal flow occurs in the Barnwell trenches. Tracer tests were conducted by the U.S. Geological Survey in experimental trenches (constructed like the actual trenches) at Barnwell. Dennehy and McMahon (1987) report a bromide tracer experiment in which "the data from the probes . . indicated the response . was due to ponded water in the trench moving laterally into the undisturbed sediments." They attribute lateral flow to the impermeable nature of the " clayey" sand barrier, and also repon evidence for i temporary perched water, which has potential for saturated lateral flow. l 1

A second piece of evidence for lateral flow is the appearance of tritium-contaminated water in the sumps that monitor trench discharge (Ichimura et al.,1994; see their Fig. 5). These sumps collect water at the base of the trenches from lateral movement along a sand buffer, which is pan of the trench design (Fig. 3.1). Quantitative estimates of flow in the sumps were not available for this analysis. Much of the sump data are suspect because (1) video logs indicate that sump monitoring wells have filled in, and (2) discontinuities in water levels within the same trench indicate that monitoring wells act as conduits for surface water infiltration.' In either case, once the j sump monitoring well is damaged, neither wet nor dry monitoring data can be used to understand water movement at the location. Thus, accumulation of wastewater in the sumps cannot be ruled out.

An estimate of the potential amount of horizontal flow at the base of the trenches was conducted by Mark Thaggard2 of the NRC using the HELP code, a 1-D water-routing code developed to analyze percolation into waste facilities (Schroeder et al.,1994). Thaggard used three layers to represent a waste trench and input regional precipitation data from Cahill (1982). The three layers represented waste backfill, a drainage zone at the l base of the trench (used to collect water at sumps) and undisturbed soils beneath the trench. The hydraulic conductivity of the basal unit was an important variable in determining the amount of horizontal flow that would occur. The hydraulic conductivities measured in four samples of undisturbed clayey sand vary from 6.7 x 10' to 4.8 x 104(Dennehy and McMahon,1987). Using hydraulic conductivity of I x la' cm/s, Thaggard estimated 90% of recharge moves horizontally at the base of the trench and using Ig5 cm/s,0.03% of the flow is horizontal.

These values might be considered bounding estimates, although Thaggard points out that the higher conductivity value affects horizontal flow, which corresponds to recharge estimates that better match regional recharge. It is not known whether recharge beneath the trenches should match regional recharge, however. The calculation leaves a large range in uncertainty for quantifying horizontal flow. However, the sensitivity of the calculation to hydraulic conductivity suggests that more in situ characterization of hydraulic conductivity in the clayey sand would help to better define the nature of flow in the bottom of the Barnwell facility trenches.

The extent oflateral flow needed to increase the areal density of the current configuration of 235U in the Barnwell trenches is unknown and is difficult to estimate given the uncertainties concerning distribution of the source term in the trench, weathering of containers and vault materials, and flow through the vaults. In the new trenches (38-87) if one vault contains SNM, then the adjacent vaults do not contain SNM. The areal densities of 235U in the trenches are significantly less than the areal densities necessary to sustain a nuclear reaction. In order to achieve a minimum areal density of 4 kg/nf (Appendix A; Table A.1), significant lateral flow and/or funneling must occur. For new trenches, the lateral flow would have to be directed along the bottom of the trench in the drainage system, directly below the SNM. It is not known to what extent this occurs. The areal densities for the older trenches suggest that the distribution of SNM in the trenches is probably more diffuse than in the newer

'Vernon Ichimura, Chem-Nuclear Systems, Inc., personal communication to Laura Toran, September 10,1996.

2 l Mark Thaggard, NRC, personal communication to Laura Toran, October 25,1996.

1 NUREG/CR-6505, Vol. 2 44

. S:ction 8 Discussion trenches (Figure 8.4). However, disposal practices during operation of the older trenches were less restrictive, t suggesting that SNM in the older trenches may not be homogeneously distributed. The primary collector for l w:ter in the older trenches is the French drain system, which runs along the side of the trench. In order for a significant increase in concentration of 2"U to occur, a combination oflateral and funneled flow will be n:cessary.

In summary, it will be important to continue to monitor the reliable sumps draining the waste trenches for evidence of waste mobilization. Although the sumps are meant to provide a collection point for wastewater and miy enhance herizontal flow, they also can serve as a means of detecting uranium transport. If possible it would be useful to evalcitte potential formation of reducing zones and the size of such zones. If the drainage system to the sumps becomes water-saturated, there is the potential for reducing conditions to occur. Since reducing zones would not form instantaneously, monitoring of the trenches would indicate changes in water chemistry over time.

By keeping the trenches unsaturated and oxidized, it is possible to prevent buildup of uranium precipitation to thicknesses that would be of concern for nuclear criticality safety.

8.5 OTHER MITIGATING FACTORS Other factors in the disposal setting may mitigate the reconfiguration of fissile uranium into critical masses.

Because these factors cannot be quantified with the present information and scope ofwork, the extent to which thsy limit criticality is not presently known. Nonetheless, these considerations can be discussed in general terms and used to identify any concerns that need to be addressed.

In all trenches at the Barnwell facility, there are multiple barriers specifically included to physically separate the SNM. These barriers, which include disposal containers, vaults, and fill material, will inhibit reconfiguration of uranium.

i' The reducing zone), if they form, will require a water-saturated environment without dissolved oxygen.

Presently, the trenches are not saturated with respect to water and the water chemistry appears to be slightly oxidizing. The effects of changing rainfall rates and trench capping could be modeled to evaluate the potential for saturating all or part of the trenches. Several trenches have high-density polyethylene caps, and future capping is planned. Another effect of assuming saturated conditions is faster estimates of travel time.

The potential distribution (amount and location) of reducing zones in this setting is a large source of uncertainty in this analysis. As stated previously, if the reducing zones are extensive, the precipitated uranium will be disperse and the concentration will be lower than needed to provide a criticality safety concern. If reducing zones do not form, the mechanism for increasing the concentration of uranium needs to change. Sorption on corroded waste packages was not modeled in this analysis. Sorption on mineral surfaces was considered in Volume I of this study, and did not result in sufficient increase in uranium concentration.

Small masses of 2"U may be considered a criticality safety concern (Figure 8.2); however, they must form a sphere, which is difficult to achieve given the generally dispersive natuie of flow. Nonspherical geometries, with the same uranium density, require considerabiy greater masses of 2"U. Even though it may be easier to form a slab or cylinder by flow paths in a trench, the masses required to form such geometries are, in many cases, greater than the inventory of an entire trench.

NUREG/CR-6505, 45 Vol.2

I f

9 CONCLUSIONS l

l This study presented calculations and assumptions to evaluate the potential for a nuclear criticality accident at an LLW site where containerized waste is buried. Many operational conditions were considered. Data needs for further evaluation are suggested. Even though conditions specific to Barnwell were used in the calculations, the work was also intended to provide an example for types of analyses for determmmg the potential for criticality at other disposal sites. The answers to questions posed in this study are as follows:

o Is there sufficient inventony for a criticality safety concern?

2 He inventory of 35U in the trenches is sufficient to form a critical geometry. However, the235U inventory, 235U enrichment and the spatial distribution of the uranium must be considered in combination to judge whether there is a criticality safety concern in a LLW trench. He critical mass of 235U for pseudo-infinite slabs (length and width at least 10 times the thickness) comprised of silica soil (1.6 g SiO2 /cm3 with variable densities of enriched uraruum and water) ranged from a minimum of about 11.3 kg to well over 2000 kg of 235U, depending upon the concentration of water and uranium contamination in the available spaces of the soil porosity. This broad range of critical masses coincides with a broad range of uranium density and system volumes or sizes. The smaller critical masses require a greater concentration (i.e., densification) of enriched uranium than the larger critical masses. For example, a 60 x 60 x 6 m pseudo-infinite slab of silica soil contammated with 0.00155 g 235 U/g of silica soil requires about 53,000 kg235U as 10 wt % enriched uranium to be critical, whereas a 1.31 x 1.31 x 0.131 m pscudo-infinite slab of silica soil contammated with 0.032 g 235 U/g of silica soil requires about 11.3 kg 235 U as 10 wt % enriched uranium to be critical. This differential is a nearly 21-fold increase in uranium density and a nearly 5,000-fold decrease in uranium mass.

The available mass of 235U in the Bamwell trenches ranges from less than 1 kg to about 1070 kg with assumed homogenized enrichments of 100 and 0.21 wt %. Taken together with the available information on the spatial distribution of the uranium in the trenches (including average235U density), these combinations of mass and enrichment were not judged to present a criticality safety concern for the Barnwell LLW site. However, for a single Barnwell disposal trench, Trench 23, the 175 kg of highly enriched uranium contained in the trench caused suflicient concern that a trench-specific investigation was performed (see Appendix B). The resulting investigation indicated suberiticality was ensured by the physical distribution of the fissile material and the commingling of the material with substantial quantities of SM. Without site-specific information regarding the spatial distribution and mass of disposals, such as for the Barnwell Trench 23 disposal, there could be concem for disposal sites hasing sufficient inventory to pose a potential criticality safety concern.

O How does U(10) compare to U(100)? How much more concentration of uranium is needed for U(10) than for U(100)?

The criticality safety calculations showed that higher concentrations of 235U were needed for U(10) than U(100), as ,

expected. At the minimum concentration that achieved safety concern, the concentration ratio of 235U for U(10) to ,

that for U(100) was 1.6. Variances in this ratio have not been analyzed on a point-by-point basis and could be larger and smaller than the value at the minimum concentration. Furthermore, the differences in the nummum potential critical mass of 235 U for U(10) and U(100) are small. Only when enrichments drop below 10 wt % do minimum masses increase significantly.

NUREG/CR-6505, 47 Vol. 2 l

\

Conclusions Section 9 I

o What chemical conditions and physical aspects of trenches are conducive to increasing uranium concentration?

Some trenches have sufficient disposal masses to pose a criticality safety concern, and the conditions are potentially available to mobilize and precipitate uranium. At the Barnwell LLW disposal facility, the water is slightly oxidizing, which enhances the potential for the transport of uranium. The presence of wood and iron containers within the trenches increases the potential for reducing zones to form. Transport could occur as water flushes through the waste and forms uranyl carbonate complexes. Degradation of waste containers can produce a reducing zone in the waste trench, if a water-saturated environment is present to limit oxygen concentrations.

These hypothetical reducing zones can precipitate uranium, as discussed below. The zones are hypothetical because they have not actually been observed at the site.

The hypothetical nature of the reducing zones is an important limitation to de current work because the mass required to cause a criticality safety concern is dependent upon the geometry of the fissile material, which in turn is dependent on specific geometries of the reducing zones. Multiple flow paths would need to funnel the uranium from large trench volumes (if the SNM is distributed across the trench volume) to very small reducing zones in order to increase the concentration of "5U to a criticality safety concern. Iflarger reducing zones form, either the "SU will be too diffuse, or larger sources of 2"U than are reported to be present in the trench are required.

It is difficult to calculate travel time for the accumulation of uranium. Since kilogram quantities of 2"U are j needed, and transporting waters have only mg/L of 2"U, millions ofliters of water need to flush through a reducing zone. Speculation about flow rates and conditions is needed to make estimates, but travel times are ,

long. Tens of thousands ofyears are needed, assuming a I-m/ year velocity, a 40-vol % porosity, and no dispersion.

o Can reducing zones, which precipitate uranium, be sustained to enable critical masses to accumulate? i i'

Yes. In the analysis, reducing zones formed very efficient barriers to uranium transport, precipitating nearly 100% of the uranium in solution. The reducing zones did not become oxidizing despite the influx of oxidized w ter over the range of pore volumes modeled: 500 to 30,000 L. The source of the reducing agent was postulated to be steel drums or wooden crates.

o How could disposal practices, in particular at the Barnwell LLW disposal facility, enhance or mitigate the development of critical masses of SNM?

Some disposal practices and trench designs can increase the potential for criticality and others can reduce it.

Trenches dug in impermeable material tend to retain the fissile material within their confines, increasing the potential for criticality. Trenches that allow for contaminant release reduce the potential for criticality by increasing the volume of the geologic medium into which the uranium can spread. At Bamwell, tritium release has occurred, but uranium might or might not be released from these trenches, depending on the geochemical conditions. Trenches whose floors are sloped can enhance focused flow, increasing the potential for accumulation of uranium. Consequently, French drains and sumps can be sites for SNM accumulation, indicating the desirability to restrict (when possible) their geometries and size to avoid potential critical geometries.

Maintaining relatively impermeable caps on the trenches should tend to reduce the uranium migration by limiting 2

the quantity of water available a-d the potential for locally reducing conditions where "U could accumulate by promoting aeration of the trench due to its unsaturated conditions. Commingling of SNM and SM may reduce the enrichment of the mobilized uranium if the two are leached similarly and well blended. To assume similar l NUREG/CR-6505, Vol. 2 48

1 l Section 9 Conclusions I

leaching and thorough blending, it would be necessary to consider the types of packages containing SNM and SM and their locations in the trench to estimate the time when packages start leaking and the rate ofleaching. The Tr:nch 23 records that were examined in detail indicate that homogenization of SNM and SM would likely occur bec use of the relatively uniform intermingling of SNM and SM emplacements. The presence of materials that can act as effective neutron absorbers can reduce the potential for criticality, Materials that sorb cranium can slow the transfer process and disperse the SNM if the sorbing material is dispersed. On the other hand, materials such as bulk carbon and beryllium can act as low-neutron-absorbing moderators, thereby reducing the needed i density, but increasing the needed mass, of uranium for criticality.

l Although conditions that permit criticality safety concerns are not impossible, disposal practices have limited the i pot:ntial concern. This study results in the following recommendations for consideration oflicense reviews of LLW facilities:

1. Minimize those factors that enhance SNM accumulation.

. Reduce groundwater infiltration

. Reduce enrichment Minimize opportunities to create isolated zones of reducing conditions. Avoid organic matter in waste cells I . Design trench to minimize focused flow

2. Limit the areal density of the fissile materials.

3. Model trench performance using site-specific conditions on a scale that addresses the potential for criticality. )

Consequently, the observation that the average enrichment of a trench is less than I wt % 2"U in the uranium i does not necessarily eliminate a criticality concern for the trench. Burial reports may suggest that localized regions of a trench contain quantities of fissile material that greatly exceed the average enrichment.

4. Continue to use sumps in disposal trenches to monitor for the presence ofiron, organics, and uranium as indicators of mobility in the trenches. If uranium is observed in the sumps, determine its enrichment.

Changes in redox conditions may be monitored by changes in different iron species. Even though it may take many years for sufficient buildup of uranium, early detection of mobile iron and uranium would indicate changes in the trench water chemistry.

In summary, the concentration of dissolved uranium in reducing zones is a possible mechanism for reconcentrating SNM within LLW disposal facilities. Further study of the geometry of these reducing zones is needed to evaluate the potential for concentrating relatively small critical masses (e.g., spherical deposits) and the cssociated predicted consequences. Further evaluation of flow through the trenches would better define the effect of the source term and its distribution within the trench on the development of a critical mass. Evaluation of mitigating factors, such as multiple-barrier disposal in the new trenches and the effect of sorption by organics cnd iron oxyhydroxides to inhibit transport out of the waste containers, will also better define the analyses presented here.

NUREG/CR-6505, 49 Vol. 2

10 REFERENCES Autry, V. R., Director, Division of Radioactive Waste Management, Bureau of Land and Waste Management, South Carolina Department of Health and Environmental Control, letter to T. E. Harris, Project Manager, Low-level Waste and l Regulatory Issues Section, Low-Level Waste and Decommissioning Projects Branch, Division of Waste Management, f Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, l d:ted January 30,1998.

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Baedecker, M. J.,1. M. Cozzarelli, R. P. Eganhouse, D. L Siegel, and P. C. Bennett,"Cmde Oil in a Shallow Sand and Gravel Aquifer-Ill. Biogeochemical Reactions and Mass Balance Modeling in Anoxic Groundwater," AppliedGeochemistry 8,569-86 (1993).

Bowman, C. D., and F. Venneri, Underground Autocatalytic Criticalityfrom Plutonium and Other Fissile Material, LA-UR-94-4022, Los Alamos National Laboratory,1994.

CahilI, J., Hydrology ofthe Low-Level Radioactne Solid-Waste Burial Site and Vicinity Near Barnwell, South Carolina, U.S. Geological Survey Open File Report 82-863,101 pp.,1982.

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Christensen, T. H., P. Kjeldsen, H.-J. Albrechtsen, G. Heron, P. H. Nielsen, P. L. Bjerg, and P. E. Holm, " Attenuation of Landfill Leachate Pollutants in Aquifers," Critical Reviews in Environmental Science and Technology 24,119-202 (1994).

Dennehy, K. F., and P. B. McMahon, Water Movement in the Unsaturated Zone at a Low-Level Radioactive-Waste Burial Site Near Barnwell, South Carolina, U.S. Geological Survey Open File Report 87-46, 66 pp.,1987.

Engesgaard, P., and K. L. Kipp, "A Geochemical Transport Model for Redox-Controlled Movement of Mineral Fronts in Groundwater Flow Systems: A Case of Nitrate Removal by Oxidation of Pyrite," Water Resour. Res. 28,2829-43 (1992).

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1

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l Submitted to the American Geophysical Union Spring Meeting, Baltimore, MD, May 27-30,1997b. )

l NUREG/CR-6505, Vol. 2 52

i Section 10 References Weiss', A. J., and P. Colombo, Evaluation offsotope Migration--Land Burial Water Chemistry at Commercially 'V rated Low-Level Radioactive Waste DisposalSites Status Report Through September 30,1979, NUREGlCR-1289,19^19 Wheeler, M. F., T. Arbogast, S. Bryant, C. N. Dawson, F. Saaf, and C. Wang,"New Computational Approaches for Chemically Reactive Transport in Porous Media," Next Generation EnvironmentalModels and Computational Methods.

Proceedings ofthe U.S. EnvironmentalProtection Agency Workshop (NGEMCOM), G. Delic and M. F. Wheeler, SIAM, Philadelphia, pp. 217-26,1997.

Wood, S. A.,"The Role of Humic Substances in the Transport and Fixation of Metals of Economic Interest (Au, Pt, Pd, U, V)," Ore Geology Reviews 11,1-31 (1996).

I l

I l

NUREG/CR-6505, 53 Vol.2 l

E .

I l

i APPENDIX A Criticality Study Results NUREG/CR-6505, 55 Vol.2

l j Appendix A Criticality Study Results

+

_ -T . _ ,

, _ -.e r '

2 , , ...

s 2 A

-@ Ll l

O ' =

d;3  ;

j

, 4 Yfh?$$

l

'!,2 l

" 3

na. n. ,

41. 3 si:wn sm "*' c (:2mc8602 0

Figure A.1 Infinite media neutron multiplication factor (k.) vs g 2H 0/g SiO 2 and g2 "U/g SiO2 NUREG/CR-6505, 57 Vol. 2

Criticolity Study Results Appendix A l

e 0.9 g ICGg NKJ2 0.1 0.1F

" /l 1 I

8 25 i

f 2.5 K inflaity te sio2

, ,k u s o'

o,1 OA) 0.01 IDR sult ofi2 CUi's SiO2)

GAM 13 Figure A.2 Infinite media neutron multiplication factor (k.) vs g H 20/g SiO 2and log scale of g "U/g SiO2 2

l NUREG/CR.-6505, Vol. 2 58

Appendix A Criticality Study Results 1

1 I

g C.Wx $i02 l

0.01 0.02 -

lj )

c.e jy d f k ': 1 o.tn , j.

3

^

4 i rg g 'y g,,1:,

IT

~

l fs-l*"

J. '. 4

. ~ . '

+, .

i s, _ ,A - - - _.

i p.

3 I ,

oo

.( '

s g i ano ,c, ,

7

o k 'm/

/

n O.15

11. 2 g 1120i: sto2 0.25 2

Figure A.3 Infinite slab thickness (cm) vs g 40/g SiO 2 and g "U/g SiO 2 1

1 NUREG/CR-6505, 59 V01. 2 )

l

Criticality Study Results Appendix A 4

J

• a 1 ,

{ s , j-~

..  % q (y (l .. , 3a j br% i I -

2 m b A rc i t>.-.. ir, ikg U235ien*2

In

! o <

o 0.01 o.os Gt .02 0.15 4.00 g il2n'g SiO2 0.2 0.04 g C235/:: 5 eO2 a25 Figure A.4 Infinite slab areal density (kg2 "U/m 2) vs g H2 0/g SiO2 and g zuU/g SiO2 NUREG/CR-6505, l

Vol. 2 60 i

i

Appendix A Criticality Study Results 1 0

g II20/g SiO2

.1 15 4

\, a

\

.25

\ 'yf,a,?

40 m' ' VP s ma f3a j 20

'\ 1

'\ 3 Simb trral Deusit3 )

\ J10 s kg r2.Wm^21 0

0.(HD

\

0.01 3 log scale of (g LZ35/g SiO2)

I D.03 235 2 Figure A.5 Infinite slab areal density (kg U/m ) vs g H2 0/g SiO2 and log scale of g 2'5U/g SiO2 NUREG/CR-6505, 61 Vol.2

Criticality Study Results Appendix A g UIM/= SiO2 o II) yn 0.03 - --

stu4 t

r .g l .

l A

?<,,, ,,'

n 7

(LOS 0.I (1.15

= 1I t % $602 Figure A.6 Infinite cylinder diameter (cm) vs g 40/g SiO2 and g 2"U/g SiO 2 NG/CR-6505,

Appendix A Criticality Study Results l

,, s in<va sio:

os e.1

!e 8 'S j i l t ji '

x .

(j .2 5 (w; J

j .

?

[ 2s 1"  ; . ii 4.s ' ^

t , . s ..

1 { h .. .

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4 g .

7

, :t

g 4 ;

e  : - - -

i s ;

e

- -- ~

l^

g

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' :l:

1 I- '

twiimi.r I I t..n.L, r>. . s.t.

k\\\\\\\(\\\(L AttAsi;s*ss' L au t 2mnu 5"

j\ggg\\\\\\\\\\\\\

\

\\

l s

s l

\

l 9

41AI

11. 0 2 cm O_M 3 l'2.Wa %Q2 Figure A.7 Infinite cylinder linear density (kg2 "U/m) vs g H2 0/g SiO2 and g 2"U/g SiO 2 NUREG/CR-6505, 63 Vol. 2

Criticality Study Results Appendix A

' 1(4 75 Qlinder 30 f.lararLk s hv an L'2Mim.

1

\\us ', 15 1

0 0

0.DIL1 0.1 0.15

1. I

.: I g 112(kg SiO2 0.03 l'1: $cate or(g 1:239g SiO2 D.25 2

Figure A.8 infinite cylinder linear density (kg 235U/m) vs g 2H 0/g SiO 2 and log scale of g "U/g SiO 2 i

NUf.EG/CR-6505, Vol. 2 64 l .. .

Appendix A Criticality Study Results g U2.lL*r,SiO2 0.01 a.02 0,0 G.M llMI 150 100-

$pberv iHanwteriem:

I sc, ti e.Os 0.1 0.15 0.2 xH2ns902 c.25 235 Figure A.9 Sphere diameter (cm) vs g H2 0/g SiO2and g U/g SiO2 NUREG/CR-6505, 65 Vol. 2

Criticality Study Results Appendix A 0

g il2Wg (102

~

l

]

1

,  ; 1 ,

e '; is 4

1. ! U '

1 9

~

l; h h  ;

l  :. 1 e >  ; ;. a e  ; ,

k i [h si ifl_d -

{ :00 1

1 54' t r,a l Spk reSlam ILtL2 M .

t n

4.461 0.02 0 03 g U235.'g SiO2 0.4LI Figure A.10 Sphere mass (kg 2"U) vs g H 0/g 2 SiO and 2 g 235U/g SiO 2 NUREG/CR-6505, Vol. 2 66

Appendix A Criticality Study Results i

o y$ g I[20.'g SiO2 a3

.(' * .15

'g . u t . . . . . '

].i . ' f '

i O.2

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e . i nk i L( in f i n de r t e' i ir t e )m m c C i( a D t y) isn/'m b e a l s d "U l2 ) e a d i t eg r e n A(k u i . f n in i t n de s s o i t e) n ( c i r km C icc 1 h( A T l e b rf 7 6 t 5 7 . o. n-1 T -i 9 3 4 6 6 1 k k 1 2 2 2 2 2 1 1 1 1 1 1 S-S /g 5 8 8 8 5 t 2 3 3 8 5 2 n O,- 6 4 4 6 2 6 t e I I 3 7 1 5 0 1 n 0 0 1 1 2 2 o g . 0 0 0 0 0 0 c r t e a W .'m. /c 0 8 9 3 1 4 6 h 5 0 1 8 5 2 4 t 1 1 2 3 3 g 0 0 0 O 0 0 S--- Sg / U t n " 5 2 5 2 5 2 5' 2 5 2 5 2 t e g 6 6 6 6 6 6 n 0 0 0 0 0 0 o c U - 8 'm /c U - 8 g 1 1 1 I 1 1 ey. r nt 6 7 8 9 0 1 in 1 1 2 2 Le 1 1 2 2 2 2 2 2 d b $eo@m$y _ <Su l l l I APPENDIX B Suberiticality Evaluation for Chem-Nuclear Systems,Inc., Trench 23 NUREG/CR-6505, 79 Vol. 2 c': l l APPENDIX B Suberiticality Evaluation for Chem-Nuclear Systems. Inc, Trench 23 Introduction A criticality safety review of the Barnwell, S.C., Chem-Nuclear Systems, Inc. (CNSI) Trench 23 disposal information (Autry,1998) was performed to evaluate the suberiticality of the burial trench as compared with recent nuclear criticality safety studies (Toran et al.,1997). The information summarized bidl informatior N Trench 23 (100 feet wide, by 20 feet deep, by approximately 992 feet long) for the moncs ciS /,c::M e' through April 1978. The monthly " Burial Activity Report (s)" included: Abbreviation used Information in this evaluation Byproduct Material, millicuries'd BPM Source Material, pounds 5 SM(Ibs.) Special Nuclear Material, grams SNM(g) Total Volume Buried, cubic feet 3 Tot Vol.(ft )
1. SNM packages #SNM Pkgs.
I Location of materials in trench by I SNM Shipments (i.e., no. shipment groupings) #SNM Grps II Source Material Shipments (i.e., no. shipment groupings) #SM Grps Burial position (ft) of Shipments along the length of Trench 23 from beginning of month to Stan ft end of month End ft No information was provided regarding specific individual burial volumes of SNM, Source Material, or Byproduct Material. Additionally, no information was provided regarding the mass or placement of Byproduct and matrix Material within the trench. It was necessary to apply some assumptions to evaluate the subcriticality of Trench 23. Those evaluation assumptions and their effects on the evaluation are provided below. 8 Title 10 of the U.S. Code of Federal Regulations Part 30, {30.4 Definitions, " Byproduct Material means any radioactive material (except special nuclear material) yielded in or made radioactive by exposure to the radiation incident to the process of producing or utilizing special nuclear material,' December 31,1997. ' Title 10 of the U.S. Code of Federal Regulations Part 40, (40.4 Definitions, " Byproduct Material means the tailings or wastes produced by the extraction or concentration of uranium or thorium from any ore processed pnmarily for its source material content, including discrete surface wastes resulting from uranium solution extraction processes," December 31,1997. ' Title 10 cf the U.S. Code of Federal Regulations Pan 70, (70.4 Definitions, " Source Material means: (1) Uranium or thorium, or any combination thereof, in any physical or chemical form, or (2) ores which contain by weight one-twentieth of one percent (0.05%) or more or: (1) Uranium, (ii) thorium, or (iii) any combination thereof," December 31,1997. NUREG/CR-6505, 81 Vol. 2 1 Suberiticality Evaluation Trench 23 Appendix B Assumptions In order to perform the comparative evaluation of Trench 23, relative to information provided in NUREG-6505, Vol. I or the Nuclear Criticality Safety Guide (Pruvost and Paxton,1996), it was necessary to assume that the containerized waste matrix, contaminated with SNM, was either a hydrogenous material like plastic, water, wood, and paper or was a relatively ineffective neutron-absorbing material like SiO 2. Other assumptions, and their effects on the suberiticality evaluations, are provided in Table B.1 below. Burial Information The Trench 23 information used for the comparative evaluation was derived from Autry,1998, and is summarized in Table B.2. Footnotes to the table provide explanations as to how the reference data were used to derive the values used in the comparative evaluation. The derived values were then compared with information published in Toran et al.,1997 (Vol.1) and Hopper et al.,1995 to demonstrate suberiticality of Trench 23. The primary values ofinterest for the comparative evaluation were2g "U/cm' (or g SNM/cm') and kg2 "U/m2 (or kg 2 SNM/m ). Results of Comparative Evaluation Information from Table C-2 SiO2 -soil (S-S) results provided in Toran et al.,1997 (Vol.1); and guidance from Hopper et al.,1995, was compared with information extracted from Table B.2 above (highlighted cells) to determine the magnitude of SNM density increase (i.e., concentration factor) to approach a criticality concern. The following information was extracted from Table B.2 for the comparative evaluation:
• maximum density is 2.6 x 10' g SNM/cm',
e maximum " infinite" slab areal density is 5.0 x 102 kg SNM/m2 , e 2 maximum single package SNM mass is less than 350 g "U (by license) but calculated to be less than 45 g SNM/Pkg, e minimum package volume calculated is 60.6 ff, l
• maximum single "SNM Shipments" burial is 52369.5 SNM(g), and
= total SNM mass in Trench 23 is 174.93 kg SNM. The information presented in Table B.3 below provides comparisons between the maximum extracted values from Table B.2 above and information from Toran et al.,1997 (Table C-2) and Hopper et al.,1995 (Table 1). The concentration factor necessary to alter the Trench 23 values to the reference values is also provided. As determined from information in Table B.2, the ratio of #SNM Grps to #SM Grps is about 1.8 to 1. However, 2 the Eff wt % of 2"U is abcut 0.08%. Given the reported maximum calculated value of 45 g "U per package 2 (typically larger than 4 ft x 4 ft x 4 ft), significant concentration of "U (via hydrogeochemical processes after breach of package walls) will necessarily involve the very large masses of SM (nearly 1200 g SM per g SNM) in the trench. The concentration and migration of SNM and SM will be effected through repeated dissolution and reduction of the uranium, thereby significantly reducing the effective enrichment of the uranium. NUREG/CR/-6505, Vol. 2 82 >o3EWW - mg. 9g? tgig" , u r.d89 U - ht 8l s 7 aie e d d i ei g e e d h w n 9 r eh m el s t yu lg ng) t 1 lg g s a ic on n ihne t te inn e e t bl ui r s s m mgur - a n 7Mah e er vo nim s imi i ) d d t e h u al ah 6 9luie nd u t 7 mis r v r vgo cb wici 9 oe et h e e oh e ond n vpy r en t r 9 cr ss t,h e e yur t ri t n ri t 1 t pd M r m l t b n 1 s ueMt a inpw t e a mbe e s ol lydnoivS i i e ,l f o i y c n a l si rord ihS eNh t lydnd e r e o s a u t s b S l a s l a a ph t f a u o bu (c he vi gad n ar r ht a le m,tc Hl t ,r o in l ei eUe p reU - f t gt h c r a ( a )U p r muh ev t n" e 2" e o n n n ye - ) t oi"a l t uu o ri e2 is g vn lei cve e 1 t u di rwotf a gfd fi o e ev eefr2 b t 8 eo e gh M s rv r ot e oai t l v e yno eyr uv 9 r s ai e u wi h a oU t e e ") t f oa t 1 t at S wis s wie t t s sl c ima v C;e h v r b rfMty" e o t isf 2 ca ofouon k r ms se nt s pen wa spenni o n x r o es r RUs o%cuenoy pU l N"2 n t r" ma ov (c m rd G eM G e n t, rd c he pn po o c8ddc is t e2 r r s p o a c t O9 -o rpes n pf geni t shN sh nr e n s n so % (c r U o r aiS oh t n e gS t t n g/ v t d o o mt . e8 y e s spr e et ne e n e nn e w "U t f a y( 7Bdud " pl a e gmi a mih miv e ipci a/e e s u aow i 2 l t a c c52 n n is a% ( r e ln "e e a r es gm mgue r ipct hd a ut hd vh t a v ir i ua t i f t ol c mer 0d ui sn sa t S eh S e r r es e  ; b v nht a myuc 5 el d aM r t i m ot w l 3 t c oivve Myn Myni) v f re a is t s0 dn ol mCloi n fi r ot s la r t oh e c Ne bi e Nb ot c a n en o s r pe9l a c u gm e v e neupn u r S rei v S re e gri vo _ d mf f n et p u c mo uoo heht, a h 8 or t yi o t n s 3 a sF in u it a yer nc u ab eS s v eh t, it s r l g s a s r ae e s s e wnrl s o nv oo t e si "re t t t a pe s r t t a sppos e n o on vi n o e d cr h "t Nge iml l h ei t h np h r sn h r mna cb t w pe )w c n e Go e Goo t t i iat a t ivo x n e "g h s c n n ho mo i i e) l e r t s c e nU e r c t t t p a vh r n e g e s ns etnhMd yNMl a le t at ( c n t at ( c n s ro s ol l m e n be2 rP bevn r t i" Tei i eU ieUs .i t u ly (c has u s o foe e a it id esl ic ek S eef SNm d nm" d n m"rpen n oh . e oetr o% n m mr vl u s n s i p2 i p2 c l s i o t r s ih h ds h h Gr o drs no i t a d y c ad w u e si n Chls an t mele drSt n ys e 0 iol ol u Rfi rStiMf omomr i io len t t o nt Not riki cei a w ol ol c a wS de c v e or9 aVv t m ou e i gml r n ei e l o ri a cb , r r e nf t a pn el a n S s e r r s r o t s me a for lci p u is r oU Uuge ei l iar at p te i l t ioonu l a u e r e ea a ed pteM rh a at e l a eh t n yu n" ck iuMrpmc o ib d e v e o n c icr e xr ase nl en s iCb/2 ef h ft b a c mef S uMr b e s o r b e sl a ouc r uc e y it n e3l2 ly pu na yo o e e ci v o )n od puh ot I cb r I cb r on bi f r ic o u a vu I a t nr logm S u a S t S em amhl eaedo )p v ac i gr i la ic f lyu uF l a vi r vp t i t s u n a Non Nonge Np e c i i c a d la a ei r a l CS o r CS ond Cs ab m enn ic et s ob s n c t o u rii v ateh o aebs eciMt u t eh t u eh t t hi eiuu o i a k a ep r e l eS c e d er t _ i iLuTty r r hi e t r h rl Tf ow b r c R B" t c r a B p( ar p Twn Twndl p i u S d e d 1 e r o e fot"snhisd aen s . g a n a t he er eTs g b s r s f B o n pe p e 2 t ig Mt b m p . gk Ge e a r v r ht 0 g e d s Nonmip a c G l e a uh a d r 1 k b St n nS mc p a k Men Meh o t f a m w f uo T u l "r e el ta a pM Ni a Nt e o s ba no pe mh ir lpm S S n a e St n t s n a o s i r t l e a s a f i i t e s c oe e at ahScf h e lat t e . lade w l o an t tp i a t m glpnMn t t e oy ha edh c hta nlgo f si m M a e s e m s N "U u s S2 t cl u ia kca )slapvr ect r nb ei od d t h t en r t ih d mtp kt n e r a f t d t pGriu uht aie et t a e r e e c e n ou on r e M( q e Sot niu n p mrhe mt d id h o s o pu u et un eg t e S uir et omc i sh yq Nsps" sht r si s e lat h t a ai t b es se ro c sae of ar a u mw f o no f op pnt ng n e er i r r o la dU r a seee oe iowv buf v i e" e if ec n r ogme md m iot s e "s t ge e u ad a t t t s a e de g y cf 2 mi u bt a ka at chlov u e c w au pdr e am g mnlpcl laGn n h. l it e o l e a af r u . e v s ps on u s vl po v i la ic s e m vi e nmh a ptih pcn ps mc e et h eMn c a :s. 2m = / i t r ng a r h eb s hh eMrpr en er e ei ne r ep eS nl e nt e e es b c Ta TStrSGpt Owd Olad tr n aw2 "U u ) ) ) ) ) ) S 1 ( 2 3 4 5 6 ( ( ( ( ( i ygONn$F ,w IC Suberiticality Evaluation Trench 23 Appendix B Table B.2 Raw and transformed data from Autry,1998. Month Sep-77 Oct-77 Nov-77 Dec-77 Jan-78 Feb-78 Mar-78 Apr-78 Sums / Averages! SM(lb) 14.0 79807.2 9746.5 18716.6 56251.4 196346.9 34121.3 67625.1 462628.91 SNM(g) 1102.0 1 % 46.7 29131.4 28986.4 27127.8 52369.5 7228.7 9337.7 174930.2l Eff wt %(1) 14.779 % 0.054 % 0.653 % 0.340 % 0.106 % 0.059 % 0.047 % 0.030 % 0.083 % Tot Vol. (ft') 17101.5 125224.8 180667.7 159599.2 110198.8 129574.8 104820.7 158087.7 985275.1 CSNM Pkgs. 64 1273 1026 1339 1153 1176 366 886 7283 Ig SNM/Pkg. 17.2 15.4 28.4 21.6 23.5 44.5 19.8 10.5 24.0 10SNM Grps 3 29 49 53 36 33 20 23 246 CSM Grps 3 16 17 19 20 27 11 18 131
1. SM Pkgs. (2) 64 702 356 480 641 962 201 693 4100 CPkgs. (3) 128 1975 1382 1819 1794 2138 567 1579 11383 ft'/Pkg. (4) 133.6 63.4 130.7 87.7 61.4 60.6 184.8 100.1 86.6 g SNM/cm'(5) 4.6E-06 8.6E-06 7.7E-06 8.7E-06 1.4E-05 2.6E-05 3.8E-06 3.7E-% 9.8E-06 g SM/cm3 (6) 2.6E-05 2.9E-02 3.4E-03 7.lE-03 2.3E-02 5.4E-02 1.5E-02 1.6E-02 2.lE-02I g SNM/cm'-T (7) 2.3E-06 5.5E-06 5.7E-06 6.4E-06 8.7E-06 1.4E-05 2.4E-06 2.lE-06 6.3E-06 l g SM/cm'-T(8) 1.3E-05 1.0E-02 8.7E-04 1.9E-03 8.2E-03 2.4E-02 5.2E-03 6.9E-03 7.5E-03 l l Start ft (9) 0 0 105 280 300 550 756 750 0.0l 30 105 280 456 525 754 850 964 964 lEnd ft (10) l Depth ft (l1) 5.7 11.9 10.3 9.1 4.9 6.4 11.2 7.4 10.2 kg SNM/m2 (12) 7.9E-03 3.1E-02 2.4E-02 2.4E-02 2.0E-02 5.0E-02 1.3 E-02 8.4E-03 3.1E-02!
kg SNM/m2 (13) 3.7E-04 1.9E-03 1.7E-03 1.6E-03 1.2E-03 2.6E-03 7.7E-04 4.4E-04 1.8E-03 l Notes: (1) One hundred times the mass of SNM divided by the sum of the SNM plus SM masses in grams. (2) The #SM Gr cs times the #SNM Pkgs. divided by the #SNM Grps assuming equivalent number of packages per group irrespective of SNM or SM. (3) The sum of #SNM Pkgs. plus #SM Pkgs. (4) The Total Volume Buried, cubic feet, divided by the #Pkgs. (5) The mass of SNM divided by the product of #SNM Pkgs. times ft'/Pkg. (6) The mass of SM divided by the product of f!SM Pkgs. times ft 2/Pkg. (7) The mass of SNM divided by the trench volume (i.e., the trench width times the burial depth times the difference of the End minus the Start), expressed in g SNM/cm'. (3) The mass of SM divided by the trench volume (i.e., the trench width times the burial depth times the difference of the End minus the Stan), expressed in g SM/ cm 2. (9) Starting position within the trench for a burial, in feet. (10) Ending position within the trench for a burial, in feet. (11) Effective depth of a burial determined from Total Volume Buried divided by the product of the trench width (100 ft) times the difference of the Start minus the End of the burial, expressed in feet of depth. (12) The mass of SNM divided by the volume of SNM Pkgs. (i.e., #SNM Pkgs, times d'/Pkg) times the burial Depth, expressed in kg 2"U/m 2. (13) The mass of SNM divided by the Total Volume Buried, expressed in kg2 U/m2, NUREG/CR/-6505, Vol. 2 84 Appendix B Suberitical Evaluation Trench 23 Table B.3 Concentration factors for criticality concern + Subcritical reference values and concer.tration factors for Trench 23 Information extracted from Table B.2 above Table C-2 Concentration Table B.1 Concentration (Trench 23 study) (Toran et al,1997) factor (Hopper et al.,1995) factor 2.6 x 10-5 g SNM/cm 2 ~1.4 x 10-2 g 2"U/cm 3 ~54 1.16 x 10 2 g2 "U/cm' 446 5.0 x 10-2 kg SNM/m 2 3.1 x 10' kg2 "U/m2 62 4.0 x 10' kg2 "U/m 2 80 45' g SNM/Pkg (unit) 2.02 x 102 g2 "U 44 7.6 x 102 g uSU 17.3
• 350 g SNM/Pkg 2.02 x 102 g2 "U 5.7 '.M x 102 g2 "U 2.2 (unit) by License Hydrogeochemical Potential for Uranium Concentration The results of this study for determining the potential for criticality following the disposal of uranium at low-level-waste facilities as containerized waste provide information regarding the cumulative uraninite precipitation for long time frames in an environment consistent with the CNSI Barnwell, S.C. site (see Figure 7.1). The results, as shown in Figure 7.1, show that increasing cumulative uraninite precipitation from 2
about 0.002 g/cm to about 0.02 g/cm' (a tenfold increase) would require about 7000 years. It is judged that, though somewhat preferential, the concentration of the SNM within Trench 23 will become blended with the SM by the same mechanisms and in the same uranium proportions placed in the trench. Conclusions We conclude that the areal density of the buried SNM (disregarding the commingling of SM) in Trench 23 is sufficiently small (i.e., one-eightieth or one-sixty-second of the areal density of concern for criticality) that criticality cannot be achieved as placed in the trench. Even though SNM concentration factors of potential concern could develop over approximately 7,000 years,'M ane hydrogeochemical mechanisms that could cause vertical and horizontal migration of SNM will also migrate SM, thereby further reducing the potential for criticality by reducing the effective 2"U enrichment in the blended materials to well below I wt % 2"U (-0.08 wt % 2"U). NUREG/CR-6505, 85 Vol.2 i a pony 336 U.S. NUCLEAR REGULA10R Y COMMi'SION 1. R TNU ER ,,,,c,,, m m.= n mi. nn ' BIBLIOGRAPHIC DATA SHEET g_w ,,,,,,,, NUREG/CR-6505, Vol. 2

2. TITLE AND SUSTITLE ORNL/TM-13323/V2 The Potential for Criticality Following Disposal of Uranium at Low-Level-Waste Facilities '- ATE REPORWBLISHED

_ l containerized Disposal d','"O GRANT NuMeeR L1376

6. AUTHOR (S) 6. TYPE OF REPORT L. E. Toran, C. M. Hopper, C. V. Parks, ORNL V. A. Colten-Bradley, NRC

7. PERl00 COVERED tincemne Doroso m mes monum, sement er seasrarser. svo.sesr

8. -.

PERFORMI.NG.O.RGANIZAT6ON aa - NAME AND ADDRESS tif ##AC.swowsso Omeme. Otrace er aspose, us memon Assumessy C 0AK RIDGE NATIONAL LABORATORY P. O. Box 2008 Oak Ridge, Tennessee 37831-6370

9. SPONSORING ORGANIZAT10N - NAnsE AND ADOR ESS t/t WAC. ryse '5 mew as ecose;it comvsesor. smweeuw ##C Ommes. Otr e,er nepos, c.A mammer asosassary r mis masser assumJ Division of Waste Management Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washingion, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. AERACT (2priesneerau The purpost of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soll chemistry, and criticality safety was formed to develop and test some reesonable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM),

end to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach wr.s to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some possible scen:rios for uranium migration and concentration increase at LLW Gsposal facilities,(2) model grcundwater transport and subsequent concentration increase via precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potentialincreases in uranim concentration 2 over disposallimits. The analysis of SNM was restricted to '5U in the present scope of won h. The work documented in this report indicates that the potential for a criticality safety concern to arise in an LLW freility is extremely remote, but not impossible. Theoretically, conditions that lead to a potential criticality safety concern might arise. However, study of the hydrogeochemical mechanisms, the associated time frames, and the factors required for an actual criticality event indicate that proper emplacement of the SNM at the site can eliminate practical concerns relative to the occurrence and possible consequences of a criticality event.

12. KE Y WORDS/DESCR:PTORS ttssa ourse orsansess smar amaramar m ascasm, ww meerr.s 13. AvAeLAssuiV si ATEuteni Unlimited urculum, low-level waste (LLW), special nuclear material (SNM), nuclear criticality, i secuar r c'awicaw. -

uranium migration, hydrogeochemical modeling, Barnwell 'raa '** . Unclassified Irn. . r., Unclassified

15. NUM8ER OF PAGES

16. PRICE esAC 7081w 335 (2498

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