ML22286A244

Facility Decommissioning Plan, Rev 3: Appendix N - Criticality and Uranium Loading Calculations

ML22286A244
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Site:	07000925
Issue date:	10/07/2022
From:	Burns & McDonnell Engineering Co, Cimarron Environmental Response Trust
To:	Office of Nuclear Material Safety and Safeguards
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ML22287A079	List: ML22286A217 ML22286A225 ML22286A227 ML22286A230 ML22286A235 ML22286A244 ML22286A246 ML22286A247
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Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

Introduction

Groundwater contaminated above release criteria will be extracted from subsurface aquifers, piped to treatment facilities, and treated to remove the contaminants by ion-exchange and/or bio-remediation processes as needed. Effluent water from treatment processes will be either reinjected into the ground or discharged to surface water in accordance with an Oklahoma Pollutant Discharge Elimination System (OPDES) permit. Resulting contaminated waste (spent resin) will be processed, packaged and disposed in accordance with applicable requirements. The resin matrix will consist of a hydrocarbon based resin (DOWEX 1); it becomes loaded with low-enriched uranium during the groundwater treatment process.

Prior to packaging the resin matrix, non-resin material will be mixed with the resin as needed to absorb free liquid to satisfy transportation and waste disposal requirements. Since the waste will contain enriched uranium, consideration has been given to the design and conduct of the processing operations to ensure that an inadvertent nuclear criticality incident is not credible.

Waste processing operations and storage of packaged waste were evaluated in three separate areas, operating on two different criticality safety limits. Ther e are two treatment systems and a separate secure storage facility for packaged waste. The two processing locations, although separated by over 1/2 mile, will be treated as a combined safe mass unit with a limit of 1,200 grams U-235 and a maximum enrichment of 5% U-235. The secure storage facility for packaged waste will be operated on a safe concentration limit basis to assure nuclear criticality safety, in which th e packaged waste stored in this building (awaiting shipment for disposal) will not exceed the fissile exem pt concentration limit of 1 gram of U-235 per 2,000 grams of non-fissile material. The possession of U-235 fo r the entire site will be further constrained to a total mass limit of 0.5 effective kilograms of Special Nuclear Material (SNM).

A series of calculations were performed based on the assumption that the administrative controls to maintain the safe mass limit for the processing operations and the concentration limit for packaged waste are not effective. The calculations, assuming a uranium enrichment of 7.33% and utilizing an Upper Safety Limit (USL) of [keff plus 3 sigma] <0.9, demonstrate that the maximum allowable safe fissile concentration is 8 g U-235/kg Resi n. This Appendix describes the basis for these input values, presents the calculations performed, and explains why the maxi mum allowable safe fissile concentration cannot be attained.

This evaluation concludes that it not conceivable that any combinations of upset conditions could occur that would result in a [keff plus 3 sigma] exceeding 0.9. Therefore, an inadvertent criticality incident is not credible.

1 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

Criticality Calculation Overview The criticality calculations prepared for the study incorporate the following conservative assumptions:

1) All fissile material on the site is located in one contiguous area even though there are three physically separate locations planned.

2) The enrichment of the uranium is assumed to be 7.33% U-235, even though on average the enrichment is at a maximum of approximately 4.5% U-235 in two areas of the site and about 1.5% in another area.

3) The criticality model is based on an infinite slab model with a thickness of 7 feet.

4) The composition of the fissile material mixture is low-enriched uranium (maximum of 7.33% U-235) adsorbed on a hydrocarbon resin material.

5) No credit is taken for the administrative controls that will be implemented to maintain the safe mass limit and concentration limits specified in the license.

The analysis is based on a highly conservative valu e of 7.33% enriched uranium and a conservative Upper Safety Limit (USL) of keff + 3 sigma > 0.9. The analysis establishes an allowable safe fissile concentration of 8 g U-235/kg Resin.

Groundwater Processing Overview

Concentration of fissile material will occur on the io n-exchange resin as the groundwater is treated to remove the uranium. The relationship between the gr oundwater concentration of uranium and the uranium concentration on the resin is based on tests that we re conducted using the selected resin material and using actual groundwater samples from the site.

The maximum enrichment value of 7.33% U-235 was also conservatively established using historic groundwater monitoring data based on analytical measurements over a 16 year period by the alpha spectrometry (HASL-300) method.1 It was based on the maximum enrichment measured plus 2 sigma (95% Upper Confidence Level (UCL)) for the highest single historic measurement of any well that will

1A comparative evaluation of alternate analytical methods concluded that the ICP-MS (EPA 200.8) would be used for project measurements of the U-235 and U-238 mass concentrations for liquid and solids. An evaluation of groundwater measurement data by IC P-MS since December 2016 demonstrated that the maximum enrichment at the 95% confidence level for the site is 4.76%. Since this value is less than the 7.33% value used for the criticality safety evaluati on, it was determined that the criticality safety evaluation is conservative and no change is requi red. (

Reference:

Enercon Technical Memo titled Determination of Maximum Conservative U-235 Enrich ment Levels for Groundwater at Cimarron Site Utilizing ICP-MS Data Collected December 2016 Thr ough the 1st Quarter 2018 dated April 12, 2018.)

2 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

feed the three treatment trains. Samples where the uranium concentration was less than 30 pCi/L were not included because the uncertainty associated with laboratory analysis of isotope measurement value significantly increases and is not reliable.

During operations, the groundwater feed to each treatme nt train will consist of a blend of groundwater from a many different extraction wells. An evaluation was made of the historic values of calculated enrichment values for the wells that are located within each of the three areas that will feed the treatment trains. Table 1 presents the results for the average and maximum enrichment at the 95% upper confidence level (UCL) anticipated in each of the three treatment trains. Note that the current design of the Treatment System will combine the feed from Areas 1 and 2 into one storage tank which will result in an averaging of the groundwater concentration and enrichment fo r Trains 1 and 2. The analysis presented in this Appendix is based on not combining th e two areas into one feed stream.

Table 1: Projected Maximum Enrichments During Initial Operations Average Average Maximum Concentration (pCi/L) Enrichment Propagated Enrichment Uncertainty at 95% UCL

Train 1 30 3.48% 0.71% 4.19%

Train 2 30 2.69% 1.14% 3.82%

Combined flow to Trains 1 & 2 30 3.09% 0.93% 4.01%

Train 3 30 1.51% 0.37% 1.88%

These maximum enrichments are less than the enrichment assumed for the criticality calculations.

Criticality Calculation Methodology and Results

Criticality calculations were performed for the following 3 cases:

Infinite 7-foot Slab of Homogenous Resin-UO 2 Mixture

o Enrichment of 7.33 wt% U-235 o Fissile Concentrations ranging from 1 to 10 g U-235/kg resin

Infinite 7-foot Slab of Homogenous Resin-UO 2 Mixture for the conditions expected during initial operations.

Transportation Model from NUREG/CR-4382 (see Appendix I of the NUREG) o Enrichment of 7.33 wt.% U-235

o Fissile Concentration of 0.5 g U-235/kg matrix material

3 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

o Matrix Material is modeled as both resin and SiO 2

Table 2 provides the uranium enrichments and fiss ile concentrations used in the three cases.

Table 2: Enrichments and Fissile Concentrations used for Criticality Safety Evaluations

Cases for Nuclear Enrichment Fissile Concentration Criticality Safety (g U-235/kg)

Evaluation (% U-235)

Bounding Case 7.33 1 to 10

Train 1 4.19 2.3 Operational Train 2 3.82 1.5 Cases Train 3 1.88 4.2

Transportation Case 7.33 0.5

For the bounding case, the infinite slab is modeled in Monte Carlo N-Particle Transport Code (MCNP) by filling a 7-foot tall rectangular prism with the homogenized resin-UO 2 mixture. The x and y dimensions are 100 cm in width and have reflecting boundary conditions to simulate an infinite slab. There is a 1-foot water reflector modeled in the z dimension. The fissile concentration was varied from 1 to 10 g U-235/kg Resin. The results for these calculations are provided below in Table 3 and Figure 1.

The following additional modeling assumptions are ma de for the infinite 7-foot slab calculations:

1. The resin is assumed to be composed of carbon and hydrogen with an atomic ratio of 1.

2. The resin is assumed to have a theoretical density of 1.1 g/cm 3 with a 70% packing fraction.

3. The resin is assumed to be dry and there is no additional groundwater present. Previous calculations have shown that the resin-UO 2 mixture is over-moderated and additional groundwater reduces the reactivity.

For the operational cases, the keff value was calculated for each of the three treatment trains using the maximum expected enrichment (95% UCL) and th e maximum expected uranium concentration (95%

UCL) on the resin. These calculated values are provide d in Table 4 and are shown in Figure 1. At the Upper Safe Limit (USL) of keff plus 3 sigma of 0.9, the interpolated value for the fissile concentration on the resin is 7.978 g U-235/ kg resin. This value has been rounded to 8 g U-235/ kg resin.

4 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

As shown above in Table 2, the maximum expected U-235 loading on a resin bed is 4.2 g U-235 per kg of resin. This fissile uranium concentrations on the resin is approximately half of the maximum fissile concentration of 8 g U-235 per kg of resin est ablished as the conservative limit by the bounding case evaluation.

The results for the third case (the transporta tion evaluation) are presented in Appendix I.

Table 3: keff Results for an Infinite 7-foot thick Slab of Resin and UO 2 at 7.33 wt% U-235

g U-235/kg g U/kg Resin Resin keff keff + 3

1 13.6 0.20069 0.00009 0.20096

2 27.3 0.36111 0.00013 0.36150

3 40.9 0.49222 0.00018 0.49276

4 54.6 0.60176 0.00021 0.60239

5 68.2 0.69380 0.00024 0.69452

6 81.9 0.77332 0.00026 0.77410

7 95.5 0.84140 0.00029 0.84227

8 109.1 0.90034 0.00031 0.90127

9 122.8 0.95089 0.00031 0.95182

10 136.4 0.99842 0.00035 0.99947

5 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

Table 4: keff Results for Resin in Three Trains

Enrichment Fissile Concentration Train keff keff + 3 (wt. % U-235) (g U-235/kg Resin)

1 4.19 2.3 0.39689 0.00016 0.39737

2 3.82 1.5 0.28120 0.00012 0.28156

3 1.88 4.2 0.57345 0.00025 0.57420

Figure 1. keff Results for an Infinite 7 Slab of Resin and UO 2 at 7.33 wt% U-235

1.10 1.00 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 01234567891011 FissileConcentration(gU235/kgResin)

CriticalitySafetyEvaluation Train1 Train2 Train3

Possible Upset Conditions for Groundwater Extraction and Process Operations

The following process upset condition could potentially occur, either individually or in conjunction:

1. Major resin spill

2. Major groundwater spill

3. Equipment rearranged to consolidate all within one building including waste containers from storage location

6 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

4. External event such as earthquake or high winds disrupts building integrity and process equipment integrity and location

5. Operational errors during the operation of th e process equipment such as misaligned valves The geometric model for the criticality calculations assu mes that the configuration of the fissile unit is a 7-foot thick infinite slab at the maximum concentrat ion allowable on the resin matrix. None of the above events would result in a configuration of fissile ma terial outside the model used in the evaluation.

Possible Upset Conditions for Resin Loading

The normal operational condition is that the groundwater feeds to each treatment train as a combined flow from a number of extraction wells such that the uran ium mass concentration and the U-235 enrichment is a composite average of many extraction wells.

In order to address an upset cond ition in which all of the groundwater comes from a single well location at the highest uranium mass concentration, the well sample data was reviewed, and the highest mass concentration sample was identified. The value is further increased by the 2-sigma uncertainty to obtain the maximum groundwater concentration at the 95% c onfidence level. The concentration of the uranium on the resin is then calculated using the Upper Bound equation for the loading of the resin. This Upper Bound value is then further increased by adding the 2-sigma value to obtain the maximum uranium concentration on the resin at the 95% confidence leve l. The U-235 enrichment for this groundwater stream is taken as the maximum enrichment at the 95% confidence level for the particular well. These calculation results are presented in Table 5 for each of the three areas that feed the three Trains.

7 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

Table 5: Resin Loading Calculation Results

Treatment Maximum Maximum Maximum 95% UCL Maximum U-Area and Well Influent Influent Uranium Uranium 235 Loading Uranium Uranium Loading on Enrichment on resin at Concentration Concentration Resin 95% (% U-235) 95% UCL at 95% UCL UCL (g/kg) (g/kg)

(µg/l)

Train 1 Well 562 593 55 5.50% 3.0 MWWA-03

Train 2 Well 127 139 39 3.40% 1.3 T-63

Train 3 Well 4,560 4,841 221 1.55% 3.4 TMW-13

This approach is conservative because it adds the 2-sigma uncertainty to each measured and calculated parameter. These U-235 loadings are bounded by the assu mption in the criticality calculations in that the U-235 enrichment is 7.33% and the safe resin concentration is 8 g U-235/kg resin.

The results show that for all three treatment trains both the maximum fissile uranium concentration on the resin and the maximum U-235 enrichment are well within the conservative assumptions utilized in the criticality calculations. The postulated upset conditio n has an extremely low probability of occurring and the extended time frame over which it would have to c ontinue without detection, but regardless the upset condition would not exceed the bounding assumptions utili ze in the criticality calculations. Based on this information, it is concluded that no process operati ons equipment or management measures are required to be identified as items relied on for safety (IROFS).

==

Conclusions:==

The criticality calculations described here demonstrate that the process will remain subcritical for loadings of up to 8 grams U-235 per kg of resin. A review of potential upset conditions indicates that it not considered credible that a combination of upset conditions could occur that would exceed both the U-235 enrichment and the uranium mass loading utilized in the calculations. Therefore, the probability of an inadvertent criticality incident is not credible. The following list summarizes the primary reasons that support the conclusion stated.

8 Facility Decommissioning Plan - Rev 3 Appendix N - Criticality and Uranium Loading Calculations

1. The enrichment in the groundwater would have to exceed 7.33% U-235 which is significantly higher than that measured on the site.

2. The uranium fissile concentration in the resin would have to exceed the limits calculated.

Historical data of uranium concentration from groundwater sampling data and the tests conducted on the resin materials provides the information to show this is not feasible.

3. The infinite slab geometry of the model bounds any possible configuration of SNM on the site.

4. The resins provide a limiting concentration of fiss ile buildup dependent on the concentration of uranium in the groundwater.

5. The higher enrichment in the groundwater is limite d to a different and physically separate area from that where the higher groundwater concentr ations of uranium are present. It is not physically possible to introduce these two sep arate groundwater sources into one treatment system.

6. No specific operations systems or management measures are IROFS.

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