CY-05-061, Supplemental Information Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002

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Supplemental Information Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002
ML051020376
Person / Time
Site: Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date: 03/31/2005
From: Gerard van Noordennen
Connecticut Yankee Atomic Power Co
To:
Document Control Desk, NRC/FSME
References
CY-05-061
Download: ML051020376 (31)
v • d • e


Text

  • CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD
  • EAST HAMPTON, CT 06424-3099 MAR 31 2005 CY-05-061 Docket No. 50-213 RE: 10 CFR 20.2002 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Haddam Neck Plant Supplemental Information Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002 In a letter dated January 4, 2005 1,Connecticut Yankee Atomic Power Company (CYAPCO) proposed to transfer certain portions of its solid waste from decommissioning of the Haddam Neck Plant (HNP) facilities (e.g., structures and buildings) to a disposal facility. Specifically, CYAPCO proposed to dispose of demolition debris from decommissioning of the HNP facilities to the Waste Control Specialist (WCS) Facility, located in Andrews, Texas.

CYAPCO has performed a conservative radiological assessment of the demolition debris material and determined that the potential dose to workers involved in the transportation and placement of the waste at the site and to members of the public after closure of the facility will be no more than a few millirem per year Total Effective Dose Equivalent (TEDE) and a small fraction of the NRC limits for exposure to members of the public of 25 milliremlyr TEDE.

This assessment was provided to the NRC by letter dated January 4, 2005.1 The purpose of this letter is to provide supplemental information requested by the NRC Staff in a teleconference with CYAPCO on February 22, 2005.

1 G. H. Bouchard (CYAPCO) letter to the US NRC Document Control Desk, dated January 4, 2005, "Request for Approval of Proposed Procedures in accordance with 10 CFR 20.2002", CY-05-002.

Document Control Desk CY-05-061 / Page 2 The above mentioned information requested by the NRC staff is in two subject areas:

1. Additional characterization information that was not available for inclusion in the original submittal of this request, and
2. On-site survey limits for various shipping containers, other than the Intermodal-type, which CYAPCO intends to utilize to ship waste to the WCS site.

These information needs are addressed as follows:

Characterization Information:

The original submittal for this request provided a significant amount of characterization information for most of the areas that will generate building debris to be shipped to WCS. The two areas for which characterization information was limited were the Containment Building interior walls and floors and the Spent Fuel Building.

Containment Building Interior Walls and Floors:

The characterization data in the original submittal was limited for the radionuclides H-3 and C-14. The contamination mechanism for these radionuclides was suspected of being gaseous diffusion into the concrete rather than due to leakage of contaminated liquids, which is the mechanism for the other radionuclides of interest. For the radionuclides other than H-3 and C-14, using the floor concentrations for both the walls andfloors (which was done in the original submittal) was considered to be very conservative as the wall contamination levels are normally much lower then the floor concentrations.

To increase the understanding of the H-3 and C-14 concentrations in concrete inside Containment, eight (8) additional concrete cores were taken at six (6) new locations in various areas of the Containment interior. As can be seen in the enclosed Figures, these cores when combined with the original 4 cores (ones that were analyzed for C-14 including 2 analyzed for H-3) cover all three floor levels of Containment and the two interior wall levels. The additional characterization data for these 8 samples is provided in the revised Table 3 (for the significant radionuclides: H-3, C-14, Co-60 and Cs-137) along with the samples results from the original submittal for which H-3 and/or C-14 was analyzed.

It can be seen in the revised Table 3, that the levels of C-14 are higher on the inside (side nearest the reactor) of the Containment interior walls. The inside samples were not analyzed for two internal wall and the two Charging Floor

Document Control Desk CY-05-061 / Page 3 cores. To ensure that the higher concentrations for these surfaces are adequately weighted in the waste average, four additional values (corresponding to the average of the two C-14 inside wall samples) have been added to Table 3 to ensure conservatism.

Table 10 provides an illustration with respect to the method used to determine the average waste concentrations. This table shows how the dilution factors used for the Containment interior concrete were determined and a sample calculation to further demonstrate the calculation method.

This data was reviewed against the conclusions made in the original submittal.

The following was determined:

1. Except for H-3, when significant contamination is present, the core wafer closest to the surface contains contamination that is at least an order of magnitude higher than that in the next sample. This confirmed the original conclusion in this respect. To characterize the waste, the surface wafer concentration is diluted by the total thickness of the floor or for the internal walls (as they are contaminated on both sides), by half of the thickness of the walls and the Charging Floor(underside also conservatively assumed to be contaminated).
2. As can be seen in the revised Table 3, the ratio of C-14 to Co-60 shows significant variability across the samples taken inside the Containment Building. This would be expected, as discussed earlier, due to the different contamination mechanisms for C-14 and Co-60. The use of a scaling factor using Co-60 as a surrogate as discussed in the original submittal is not appropriate for C-14.
3. The concentrations of C-14 in the waste, diluted over the appropriate depth of concrete, are generally consistent. As the underside of the Charging Floor is expected to be contaminated with C-14 at levels consistent with inside wall samples, only half the of the Charging Floor thickness is assumed in the dilution calculation. This data is consistent with that contained in the original submittal.
4. The concentrations of H-3, Co-60 and Cs-1 37 in the waste (surface concentrations diluted over the appropriate depth for the later two) are lower then those presented in the original submittal.

Based upon the above, the following modifications are made to the original submittal of this request:

1. For the Containment Building internal walls and floors, the C-14 concentration to be used to determine the post closure dose will be that contained in the revised Table 3 using actual characterization data in lieu of using a scaling factor to the Co-60 concentration. This change results in a change to the weighted average C-14 concentration for all the waste

Document Control Desk CY-05-061 / Page 4 proposed for disposal at WCS (revised Table 8 enclosed). As can be seen on the revised Table 9 attached, this increase has no affect on the projected post closure dose since C-14 in the waste does not result in any dose for the WCS site. Considering this, the above changes do not alter the conclusion of the original submittal in that "the potential dose to workers involved in the transportation and placement of the waste at the site and to members of the public after closure of the facility will be no more than a few millirem per year Total Effective Dose Equivalent (TEDE) and a small fraction of NRC limits for exposure to members of the public of 25 millirem/yr TEDE".

2. Although the average waste concentrations for H-3, Co-60 and Cs-1 37 determined for the Containment Building interior concrete walls and floors are lower then those contained in the original submittal, the original higher values will be retained for conservatism.

Spent Fuel Building:

Due to the operable status of the Spent Fuel Building, characterization has not been undertaken. Recently, on March 30, 2005, all the spent fuel and GTCC waste was transferred from the spent fuel pool to the Independent Spent Fuel Storage Installation (ISFSI). The characterization of the Spent Fuel Building will be performed in upcoming months once spent fuel racks are removed and the spent fuel pool is drained. Specifically, 20 samples (evenly spaced with 4 in the walls and 8 in the floors below elevation 17'6" and an additional 8 samples from the walls of the spent fuel pool above elevation 17'6") will be taken to provide enough characterization data to confirm radionuclide waste concentrations and scaling factors. The samples taken will be analyzed so that the profile with the depth of the concrete can be confidently shown. The samples will include analysis of concrete from inside and outside surfaces and areas inside the wall with at least 15% of the wall/floor thickness characterized. To adequately assess the volumetric contamination of concrete, a wafer from at least 20% of 20 samples will be analyzed for all 20 nuclides listed in Table 2-12 of the HNP License Termination Plan. The results of these samples will be compared to the waste concentrations assumed in this request. If the results show higher waste concentrations (i.e., higher worker or post-closure dose) the NRC will be asked to review and approve the effect of these differences on the conclusions of this submittal. If the waste concentrations are below the values that have been presented, the sample results will be submitted to the NRC for information.

On-site Survey Limits for Additional Tvpes of Waste Containers:

B-25 Boxes:

In addition to the Intermodal-type containers discussed in our January 4 th letter',

CYAPCO may be using B-25 type boxes to ship waste to WCS. An action level

Document Control Desk CY-05-061 / Page 5 has been developed to identify when it is appropriate to transport waste in a B-25 box to WCS or to an alternate disposal site if the container dose rates exceed the alternate waste disposal procedure criteria of 10 CFR 20.2002. This action level is a bounding value developed using the same assumptions and methodology as were used to determine the actions levels for Intermodal-type boxes in the January 4, 2005, submittal.

Using totally filled B-25 boxes containing the highest density material allowed by the package weight, a 1 meter dose rate of 5 pr/hr is selected as a reliable and conservative action level for determining compliance with the alternate disposal procedure survey criteria. It is considered that containers exhibiting dose rates below the action level may be shipped to WCS and those exhibiting higher dose rates need to be investigated further to determine radionuclide concentrations or shipped to alternate facilities.

As this limit is 50 % of the limit for Intermodal containers, the corresponding dose to transportation workers will be less that that shown acceptable in the January 4 th letter'. Should some very dense material be shipped in B-25 boxes such that the boxes are not completely filled to the maximum allowable weight, the 1 meter dose rate limits shown on Graph 1, below multiplied by a factor of 2.5, will be followed.

Graph 1: Action Levels for B-25 Containers at Maximum Loading Dose Limits for Transfer to RACE for IOK+ lb. boxes 7 -- - - _-_

6-3 --

40% 50% 60% 70% 80% 90% 100% 110%

-CcntactDoseRate(ur/O _-DcseRate at e(ur r  % Full Dump Trucks/Gondola Rail Cars:

The transportation plan for moving the subject waste material to the WCS facility, included in our original January 4, 2005 submittal, consisted of the following:

Document Control Desk CY-05-061 / Page 6

  • Load the material into intermodal containers at the HNP site and transport these containers by truck to a rail loading station within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> trip of the HNP facility.
  • At the rail loading facility, the intermodal containers would be transferred from the truck onto flat bed rail cars for the remainder of the trip to WCS.

CYAPCO is now planning to utilize a different transportation plan as follows:

  • Load the material into trailer dumps at the HNP site and transport to a Transload Facility within a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> trip from the HNP site.
  • Dump the waste material into the gondola rail cars within the enclosure at the Transload Facility.
  • Transport by rail to the WCS Texas Facility
  • Unload at WCS using the landfill site typical practices We have performed a revised run of the TSD-DOSE Dose Calculation (See Attachment 2) to determine the dose to workers transporting the waste to and receiving/burying the waste at the WCS facility. The results of this calculation are shown in Table 11 below along with the per-truckload doses from the original submittal.

Table 11 Transportation/Disposal Worker Yearly Dose (At the Average Waste Concentrations)

January This

_ Submittal _ Submittal Work Dose/ Trips per mrem/ Dose! Trips per mreml Task Truckload year year Truckload year year Driver _ 0.000072 mrem 250 0.018 0.00044 mrem 250 0.055 Transload N/A N/A N/A 0.0002 625 0.125 Facility Worker _ _

Receiving 0.000091 mrem 625 0.057 0.000091 mrem 625 0.057 Worker Landfill 0.00015 mrem 313 0.047 0.00015 mrem 313 0.047 Worker _ _

Offsite 4.1 E-9 mrem 1250 5.1 E-6 4.1 E-9 mrem 1250 5.1 E-6 Individual _ _

Worker 4.6 E-4 mrem 1250 0.575 1.0 E-3 mrem 1250 1.26 Population Average Average Over N/A 0.0575 Average over 14 N/A 0.09 Worker 10 workers _I _ workers

Document Control Desk CY-05-061 / Page 7 A review of Table 11 shows that with the new transportation plan, only the Transload Facility Worker receives more yearly dose (0.125 mrem) then the highest calculated dose in the January submittal (assuming waste at the average concentrations). Using the same methodology stated in the January submittal, if all of the waste was at the maximum allowable concentration (40 times the average concentration) the yearly dose to the Transload Facility Workers would not exceed 5 mrem. As waste concentrations corresponding to 40 times the average waste concentration was used in the original submittal to determine the on-site waste container survey limit and the geometry of the trailer dumps is essentially the same as an Intermodal box, a 1 meter dose rate of 1Opr/hr over background remains a reliable and conservative action level for determining compliance with the alternate disposal procedure. Also as discussed in the original January submittal, it is expected that few if any shipments will have this maximum concentration and the yearly average dose to drivers, transload facility workers and WCS site workers will correspond to the much lower dose (highest value of 0.125 mrem/yr) from material having the average radioactivity concentrations defined in Table 8.

The option to use the transload facility for this waste involves the transfer of aggregate materials from intermodals or other containers to railcars in an enclosed facility. The waste materials would be unloaded directly into a gondola railcar through an open directional hopper. The loads would then be groomed and adjusted as needed. No material would ever be unloaded onto the floor of the facility and the hopper would not be an accumulation bin. Once loaded, the railcars would be transferred to the WCS site in Texas.

Since the option to use the transload facility involved the direct transfer of waste material from one container to another, the potential for transload employees to receive internal exposures from this process was evaluated. A bounding calculation was performed using the average radionuclide concentrations from Table 9 and an estimate of annual shipping volumes to determine the maximum internal dose to a theoretical worker in the transload facility. The calculation assumed that 50 million pounds of material was transferred to a railcar during the course of one year and a maximally exposed theoretical worker was involved in transferring half of the material (25 million pounds or 625 truckloads per year) in an enclosed facility. Using conservative values for airborne release fractions, respirable fractions, and ventilation rates, the dose to this theoretical worker was calculated to be less than 1.5 millirem for the entire process. The actual dose to a real worker in the transload facility is expected to be much lower than this bounding value based on the following considerations:

  • The calculation assumed median airborne release fractions and respirable fractions for material in a powder state. The actual particle size of airborne material from the HNP waste is expected to be large (i.e. 10 pm particle size or greater) whereby the respirable amount would tend to be considerably less.

Document Control Desk CY-05-061 / Page 8

  • The calculation assumed that workers remained in the area until all airborne particulates were removed by normal air transfer within the facility. This is highly unlikely to occur Whereas workers are more likely to leave the facility when after transfer operations are complete.
  • The calculation assumes a "free-fall" scenario where dust generation would be maximized. In reality, dust generation would be limited by the presence of the hopper and receiving container.
  • No positive airborne activity has been detected at the HNP during the aggressive demolition of concrete structures with similar radionuclide concentrations.

This evaluation concluded that yearly average total internal and external dose to workers at the transload facility would be less then 1.625 mrem which is well within the guidelines established of 'a few millrem per year".

In order to demonstrate that the impacts of the transfer operation are as expected, controls will be set up to check airborne and contamination levels at the transload facility for waste coming from portions of the plant that exhibit contamination levels at the higher end of the range shown in Table 8.

Although not expected, should other sized containers be used, appropriate dose rate limits will be determined using the same basis as shown above.

Summary In summary, the conclusion reached in our January 4,2005 letter that the calculated potential dose to members of the public, as a consequence of the proposed waste disposal from the decommissioning activities at the HNP at the Waste Control Specialists, Andrews, Texas Facility, are as follows:

  • Workers involved in the transportation to and placement of the waste in the disposal cells at WCS will receive doses that are a fraction of the 5 mrem/yr dose allowable for this type of activity; and
  • The projected dose to residents after closure of the site is an insignificant fraction of the 25 millirem per year limit.

Therefore, CYAPCO concludes that the proposed request for approval in accordance with 10 CFR 20.2002 will not have a significant impact on the workers, public, or the environment and that it is, therefore, acceptable.

CYAPCO hereby requests expedited review and approval of this request to support our decommissioning activities at the HNP.

Document Control Desk CY-05-061 / Page 9 If you should have any questions regarding this submittal, please contact me at (860) 267-3938.

Sincerely, Gerard P. van Noordennen Date Regulatory Affairs Manager Attachments cc: S. J. Collins, NRC Region 1 Administrator T. B. Smith, NRC Project Manager, Haddam Neck Plant R. R. Bellamy, Chief, Decommissioning and Laboratory Branch, NRC Regionl E. L. Wilds, Jr., Director, CT DEP Monitoring and Radiation Division

CY-05-061 Docket No. 50-213 Attachment 1 Haddam Neck Plant Figure 1 Containment Floor Samples Figure 2 Containment Internal Wall and Charging Floor Samples Three (3) Attachments of Diagrams from the Survey/Sampling Work Plan - SSWP Reactor Containment Building Elevation 1'-6" Reactor Containment Building Elevation 22' Reactor Containment Building Elevation 48'-6" Three (3) Revised Tables Table 3, Containment Floor and Wall Samples, Revision 1 Table 8, Average Waste Concentration Calculation, Revision 1 Table 9, Post Closure Dose Calculation, Revision 1 Table I 0, Basis for Dilution Factors Used in Determining Table 3 Waste Concentrations March 2005

Document Control Desk CY-05-061 / Attachment 1 / Page 1

Document Control Desk CY-05-061 / Attachment 1 / Page 2 Figure 2 Containment Internal Wall and Charging Floor Samples Outside Wall or Charging Floor Outside Wall Surface 2.5" ] 2.5" 0 2.5" 2.5" 2.5" 1 C-02 2.5" 2.5" I IfX IC I 7" IC 7" 7C 4.5" Typical core for locations 4" 4" 188, 190, 191 and 192 4.5" 7"

] 2.5"

] 2.5" 4C-03 2.5"

] 2.5" 54j 2.5" Inside Wall Surface Lower Level Wall Mid Level Wall Location 187 Location 189 Note: I` 1 wafers analyzed for radionuclides All dimensions are approximate

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GENERAL ARRANGEMENT DRAWING ROAOWAKR ITERSCI REACTOR CONTAINMENT BUILDING REMSION

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GENERAL ARRANGEMENT DRAWING -- -06/06/-i RO#RWAU.1NIESECDON RESNDATE: - 06/06/01 REACTOR CONTAINMENT BUILDING EMSioN#K_: 7 SURVEYAREABOUNDAR --

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CONNECTICUTYANKEEATOMICPOWERCOMPANY rnALCiEA1ONDATE 06118197 LEGENDO GENERAL ARRANGEMENT DRAWING 060 RtOORWAU1E'RSECoN -

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Document Control Desk CY-05-061 / Attachment I / Page 8 Table 9 (Revision I dated 3124105)

Post Closure Dose Calculation Dose Equivalent per Concentration Weighted Post Closure Radio- of Radionuclide - Average of Dose for Avg nuclide Resident Farmer All Waste of All Waste (mrem/yr per (pCi/g) (mremlyr) pCi/g)

H-3 4.20E-08 261.88 1.100E-05 C-14 O.OOE+00 9.69 O.OOOE+OO Mn-54 O.OOE+00 1.67E-03 O.OOOE+00 Fe-55 O.OOE+00 0.14 O.OOOE+00 Co-60 O.OOE+00 0.28 O.OOOE+0O Ni-63 O.OOE+00 1.69 O.OOOE+00 Sr-90 O.OOE+O0 0.0277 O.OOOE+00 Nb-94 O.OOE+00 1.25E-03 O.OOOE+00 Tc-99 O.OOE+00 6.49E-03 O.OOOE+00 Ag-1 08m O.OOE+O0 2.04E-03 O.OOOE+0O Cs-1 34 O.OOE+00 4.89E-03 O.OOOE+00 Cs-1 37 O.OOE+00 0.97 O.OOOE+00 Eu-152 O.OOE+00 5.01 E-03 O.OOOE+00 Eu-154 O.OOE+00 3.81 E-03 O.OOOE+00 Eu-155 O.OOE+00 3.85E-03 O.OOOE+00 Pu-238 4.30E-07 3.69E-03 1.587E-09 Pu-239 O.OOE+00 1.23E-03 O.OOOE+00 Pu-241 O.OOE+00 5.09E-02 O.OOOE+00 Am-241 O.OOE+00 6.58E-03 O.OOOE+00 Cm-243 O.OOE+00 1.11 E-03 0.000E+00 Total Post Closure Dose (mremlyr) 1.101E-051 Notes: 1.Values in Bold Type are based on Minimum Detectable Activity (MDA)

(i.e. Radionuclide was not detected at the MDA concentration)

2. Information changed from the original submittal shown in shaded cells

Document Control Desk CY-05-061 / Attachment 1 / Page 9 TABLE 10 Basis for Dilution Factors Used in Determining Table 3 Waste Concentrations A. Dilution Factors Wafer Thick. Total Thick. Dilution (in.) (in.) Factor 1.Floor Samples # 175 & 176: 2.5 24 9.6

2. Sump Floor Samples # 185 & 186: 1.0 11 11 (Dilution Factor used for Co-60 in Sample # 185 and Cs-137 in both samples. For all other results, actual value used as sump will have been totally remediated after 1" removal for the other radionuclides)
3. Duratek Floor Sample# 1: 1.5 24 16
4. Duratek Wall Sample #2: 1 14(1/2 Wall) 14
5. Duratek Wall Sample #3: 1.5 14(1/2 Wall) 9.3
6. Internal Wall Samples # 187 thru 190: 2.5 14(1/2 Wall) 5.6
7. Charging Floor Samples # 191 & 192: 2.5 12(1/2 Floor) 4.8 B. Sample Calculation for C-14 Concentration in Containment Floor and WNalls Sample # Sample Dilution Factor Waste From Table 3 Concentration (pCi/g) From "A" Above Concentration (pCi/Q)
1) 175-IC-01 720 divided by 9.6 equals 75.00
2) 176-1C-01 350 " 9.6 " 36.46
3) 185- 1C-02 (Use 2nd wafer as first 1" of concrete to be disposed elsewhere) 0.5
4) 186-1 C-02 (Use 2nd wvafer as first 1" of concrete to be disposed elsewhere) 0.57
5) 187-IC-01 131 " 5.6 " 23.4
6) 187-4C-05 450 " 5.6 " 80.4
7) 188-IC-01 35 " 5.6 " 6.3
8) Inside Sample Core #188- Use average of #s 6 & 10 ((450+516)/2)/5.6 = 86.3
9) 189-IC-01 10 " 5.6 " 1.8 10)189-4C-04 516 " 5.6 " 92.1 11)190-iC-01 187 " 5.6 " 33.4 12)Inside Sample Core #190: Use average of #s 6 & 10 ((450+516)/2)/5.6 = 86.3 13)191-lC-01 217 " 4.8 " 45.2 14)Inside Sample Core #191: Use average of #s 6 & 10 ((450+516)/2)/4.8 = 100.6 15)192-IC-01 7 " 4.8 " 1.4 16)Inside Sample Core #192: Use average of #s 6 & 10 ((450+516)12)14.8 = 100.6 Average Waste Concentration = 770.3/16=48.1

Docket No. 50-213 CY-05-061 Attachment 2 Haddam Neck Plant Updated TSD-Dose Calculation

("Dose to Non-Radiation Workers During Transport, Receipt, and Disposal for Transload/Gondola Car Transportation Option - dated March 29, 2005")

March 2005

Dose to Non-Radiation Workers During Transport, Receipt, and Disposal In order to assess the impact to non-radiation workers from the transport, receipt, processing, and disposal of low activity radioactive waste, an analysis was performed using the TSD-DOSE model (V 2.22)'.

TSD-DOSE is a program developed by Argonne National Laboratory for estimating doses to facility workers and the surrounding public at Treatment, Storage, and Disposal (TSD) facilities from shipments of hazardous waste that may contain small amounts of radioisotopes.

The steps and parameters used to model the operations were chosen to be conservative yet realistic. In other words, engineering judgment and knowledge from several site visits was used to develop a model which could be applied to most TSD facilities and would produce conservative dose estimates in almost all cases (almost all because not every TSD facility was visited such that the conservatism of the model may not cover a site that has characteristics outside of the model). The default values were chosen to bound the TSD facilities visited.

TSD-DOSE estimates worker and public doses from seven operations. These operations can be turned on or off to reflect the actual TSD facility operations. In addition, many of the parameters used to model the typical operations can be adjusted to fit the actual facility. A dose is calculated for each operation based on radionuclide activities, waste characteristics, and any site-specific information entered by the user.

Doses to various receptors are calculated by summing the doses from those operations that would potentially contribute to the exposure.

Ver 2.2 of the TSD-DOSE model was used to calculate the dose to the truck driver, the non-radiation worker at the TSD facility, and the public during transport and handling of the low activity material.

The worst-case scenario that will maximize the dose to the driver and the non-radiation worker at the TSD facility is a rolloff container with a 25 cubic yard capacity. In order to calculate the maximum dose to the non-radiation worker at the TSD facility for this worst-case scenario, the following assumptions for the seven operations in the TSD-DOSE model will be made. These assumptions are based on the actual Waste Control Specialists experience in handling similar waste streams.

Transport to the TSD facility.

The steps in this operation are:Load andsecure shipment priorto transport;Drive loadedtnrck to TSDfacility; Rest in back of cab en route to TSDfacility; Maintenance (i.e. checking tires or refueling) of truck en route to TSD facility.

For trans-shipment from the generating facility to the rail transload facility, it is assumed that the driver is exposed for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at a distance of 5 feet from the waste. the dose is insignificant to the railroad worker.

Offloading at Transload facility (Mix waste in mixing pit)

1. "TSD-DOSE: A Radiological Dose Assessment Model for Treatment, Storage, and Disposal Facilities", Argonnc National Laboratory, ANLJEAD/LD-4 (Rev. 1), September 1998.

The offloading to a gondola car is modeled by using the offloading to mixing pan module in TSD-DOSE and assuming that the offloading to the railcar conservatively takes one hour.

Two non-radiation worker for one hour at a distance of 5 feet.

For bulk shipments the dose is insignificant to the railroad worker.

Receiving and sampling at WCS Weigh and survey truck and inspect manifest:

One non-radiation workers for 30 minutes at the default distance of 5 feet.

Unload dnumsfor inspection, sampling andstoragepriorto treatment:

This operation was not included Inspect andsample dnims.

This operation was not included Transfer dnrms to storageawaitingtreatment.

This operation was not included Punp dnummed liquids to storage.

Not applicable.

Storage.

Work in solid storage area.

This operation was not included Transfer drumns out of storage arefortreatment.

This operation was not included Work in liquidstorage area.

Not applicable.

Incineration.

This operation was not included.

Treatment and on-site landfill at WCS.

Unload waste to mixingpit.

This operation was not included Mix waste in mixingpit.

This operation was not included Load trick and transportto landfill.

Two non-radiation worker for thirty minutes at a distance of 5 feet.

Unloadtnuck at landfill.

Two non-radiation worker for 15 minutes at a distance of 5 feet.

Transport to off-site landfill.

This operation was not included.

Incinerator maintenance.

This operation was not included.

It is also likely that at least 10 different TSD facility workers could be exposed to any one shipment.

The results of the TSD dose calculation are attached. The first TSD-DOSE report models the shipment to and offloading at the transload facility. The second TSD-DOSE report models shipment, offloading, and disposal at the WCS facility.

Prepared by: Date: March 29, 2005 William P Dornsife, Corporate Radiation Safety Officer, Waste Control Specialists

TSD-DOSE: A Radiological Dose Assessment Model for Treatment, Storage, and Disposal Facilities Version 2.22 - September 1998 Site: WCS Shipment Transport and offload at transload facility User dorsife TOTAL EXTERNAL INTERNAL Dose to:

Driver: 22E-04 mrem 2.2E-04 mrem O.OE+00 mrem Receiving worker not applicable not applicable not applicable Incineration worker not applicable not applicable not applicable Landfill worker: 2.OE-04 mrem 2.0E-04 mrem 8.OE-08 mrem Offsite Individual: 4.1E-09 mrem Offsite population: I.4E-08 p-rem Worker Population: 62E-07 p-rem 62E-07 p-rem 1.6E-1 0 p-rem Dose from: Tf17O SLOFAD FARCGUary Transport to T8CD4adil: 2.2E-04 mrem 2.2E-04 mrem not applicable Receiving and sampling waste: O.OE+00 mrem O.OE.00 mrem O.OE+00 mrem Storage before processing: O.OE+00 mrem O.OE400 mrem not applicable Incineration of waste: not applicable not applicable not applicable "OPLOAD AT u t: 2.OE-04 mrem 2.OE-04 mrem 8.OE-08 mrem

-rI2AA)SL'PZV Transportto offsite landfill: not applicable not applicable not applicable Incinerator maintenance: not applicable not applicable not applicable Doesdue to each Iotope (mrern - populatin dose In p-rem).

Isotope Co60 CsI37+D FeSS Actvity 5.4E-06 Ci 19E-05 Ci 2.7E06 Cl Rekase Fraction 1.0DE-02 2.00E-03 5.00E.03 Drtver 1.3 E-04 9.5 E-05 0.0 E+00 Receivng worker 0.0 E+00 0.0 E400 0.0 E+00 kmnveration worker not appkable Landfihl worker 1.1 E-04 8.4 E-05 33 E-10 OfftithiddWual 1.0 E-09 3.1 E-09 1.1 E-13 Offslte population 3.5 E-09 1.1 E-8 3.7 E-1 3 Worker population 3.5 E07 26 E-07 6.5 E-13

Site Description Operations included: Operations excluded:

Transport to TSD facility Incineration of waste Receiving and sampling waste Transport to offsite landfill Storage before processing Incinerator maintenance Burial at onsite landfill Parameters The follwing are the adqustable parameters used to model each operation.

A (D)after a value Incicates the default value was used.

Fraction solid waste = 1.000 Fraction liquid waste = 0.000 Pre-processed waste density = 1.0 E+00 g/cc Post-processed waste density = 1A E+00 g/cc TASt.oAt, FACILJ-Vy Transport to T4aei*ity (4 steps)

Nunber of Workers: 1.OE+00 (D)

Truck bed dimensions (for all steps) length: 2.00E+01 feet width: 7.SOE+00 fe height 4.50E+OO feet Step A: Load and secure shipment average distance: 3.00E400 fet (D) duration: O.OOE+00 hours shielding tickness: 625E.02 inches (D)

StepW: Drive average distance: 5.OOE+00 feet duration: 2.00E+00 hours shielding thickness: 1,251201 inches (D)

Step C: Rest average distance: 2.00E+00 feet (D) duration: .002E400 hours shieldin thicess: 125E-01 riches (D)

Step D. Maintenance In transit average distance: 3.00E+00 feet (D) durtion: 0.00E+00 hours shieling thickness: 6.25E-02 inches (D)

Recelving and sampling waste (5 steps)

Number of Workers: 1.0E+4 Step A WeIght truck, Inspect manifest average distance: 5.OOE+OO feet (D) duration: O.OOE400 hours 2

Receiving and sampling waste (contd)

Step B: Unload drums average dstance: 3.00E400eet (D) tme per drun or pailet: 0.OOE40O hours Step C: Inspect and sample drums average distance: 5.00-E01 feet (D) time per drum. O.OOE+00 hours eirborne respkia dust concenraion: 1.0E+01 mvhn3 (D) resrtxy proteceion factor 1.0E+01 (D)

Step D: Transfer sofldsto storage average distance: 3.ODE+00 feet (D) tme per drum or pallet O.O0E+00 hours Step E: Pump drummed oll to storage tank average distance: 5.00E-01 feet (D) time per drum: 0.002+00 hours Storage before processing (3 steps)

Step A: Workers In solid waste storage area average dstance: 3.00E+00 feet (D) duration: 0.00+00 ours Step B: Transfer solids out average distance: 3.00E400 feet (D) time per dnrn or pallet O.OOE+00 hours Step C: Workers In liquid waste storage area average distance: 3.00E400 feet (D) duration: O.OO2+0D hours shielrg thickness: 12E-01 inches (D)

Storage tank dimensions:

length: 7.002+00 feet (D) width: 7.00E+00 feet (D) height: 120D401 feet (D)

Burial at onsite landfill (4 steps)

NumnberdWorkers: 2OE400 Dump truck bed dimensions for steps A, C, and D):

length: 200+401 feet width: 7.50E+00 feet height: 4.50E+00 feet AT TOzAZSLDAS D FRACILTf Step A: tUnload wasta t=xl veragedistance: 5.00E+00 feet (D) duration: 1.00E+00 hours shielding thickness: 125E-01 kiches (D) airbome respirable dug concentbration .DE400 rngn3 (D) respitory protection factor I .02E00 (D)

Step B: Mix waste In mixing pit average distance: 1.OOE2+01 feet (D) duration: O.OOE+00 hours cower thckness: 2002400 inches (D)

Mi1n pit dimensions:

lengt 1.OOE+01 feet (D) veictv 1.00E+01 feet (D) depth: 1.00E+01 feet (D) cover thikkness: 200E400 inches (D) 3

Burial at onsite landfill (cofnrd)

Step C: Load truck and transpiort to landfill average fftance: 5.OOE+0 feet (0) duratoe O.DOE.O0 hours sNeldn Thicness: 125E1 Aids (D)

Step D: Unload truck at landfil averagedstanoe: 5.00E+00 feet (D) duration: O.OOE+00 hours eldir Hickns: 125E-01 inches (D) 4

TSD-DOSE: A Radiological Dose Assessment Model for Treatment, Storage, and Disposal Facilities Version 2.22 - September 1998 Site: WCS Shipment Transport and Disposal at WCS User. dornsife TOTAL EXTERNAL INTERNAL Dose to:

Driver: not applicable not applicable not applicable Receiving worker: 9.1 E-05 mrem 9.1E-05 mrem O.OE+00 mrem Incineration worker: not applicable not applicable not applicable Landfill worker: 1.5E-04 mrem 1.5E-04 mrem O.OE+00 mrem Offsite individual: 4.1 E-09 mrem Offsite population: t.4E-08 p-rem Worker Population: 3.9E-07 p-rem 3.9E-07 p-rem O.OE+00 p-rem Dose from:

Transport to TSD facilit. O.OE+00 mrem O.OE+00 mrem not applicable Receiving and sampling waste: 9.1E-05 mrem 9.1 E-05 mrem O.OE+00 mrem Storage before processing: O.OE+00 mrem O.OE+00 mrem not applicable Incineration of waste: not applicable not applicable not applicable Burial at onsite landfill: 1.5E-04 mrem 1.5E-04 mrem O.OE+00 mrem Transport to offsite landfill: not applicable not applicable not applicable Incinerator maintenance: not applicable not applicable not applicable Doses due to each Isotope (mrern - popuaton dose In p-ren).

Isotope Co6O Csl37+D Fe55 Actvity 5.4E.06 Cl t.9E05 Ci 2.7E.06 CI Rekease Fraction t .OOE.02 2.00E-03 5.OOE.03 Driver 0.0 E+00 0.0 E+00 0.0 E+00 Receiving worker 5.2 E.05 3.9 E.05 0.0 EOO inineration worker not arppble Landfill worker 8.5 E-05 6.3 E-05 0.0 E+OO Olkste Indhidual 1.0 E-09 3.1 E-09 1.1 E-13 Offslte population 3.5 E.09 1.1 E-08 3.7 E-13 Worker populaton 22 E-07 1.7 E.07 0.0 E+0O

Site Description Operations included: Operations excluded:

Transport to TSD facility Incineration of waste Recelivng and sampling waste Transport to offslte landfill Storage before processing Incinerator maintenance Burial at onsite landfill Parameters The fodrosng are the acustable parameters used to model each operation.

A(D)after a value indicates the default value was used.

Fraction solid waste = 1.000 Fraction liquid waste 0.000 Pre-processed waste density = 1.0 E+00 g/cc Post-processed waste density = 1.4 E+DO g/cc Transport to TSD facility (4 steps)

NumberofWorkers.A0E+00 (D)

Truck bed dimensions (for all steps) length: 2.00E+01 feet vidth: 7.50E400 fed heht 4.50+W0 fed StepA: Load and secure shipment average distance: 3.04E+0 feet (D) duration: O.0+OE00 hours shiekling thickness: 6.25E-02 inches (D)

Step B: Drive average distance: 1.00E+01 feet duration: O.OOE+00 hours shieiding thickness: 125E-01 inches (D)

StepC: Rest average distance: 200E+00 feet (D) duration: O.OOE+00 hours shielding tickneue 125E-01 inches (D)

Step D: Maintenance In transit average distance: 3.00E+00 feet (D) duration: 0.002400 hours shieling thckness: 6.25E-02 inches (D)

Receiving and sampling waste (5 steps)

Nurrber of Workers: 1.0E+00 Step A: Weight truck, Inspect manifest average distance: 5.00E+W0 feet (D) duration: 5.00E-01 hours 2

Receiving and sampling waste (conrd)

Step B: Unload druns average disance: 3.00E4OO feet (D) time per drum or pallet: O.OOE+OO hours Step C: Inspect and sample drmns average distance: S.OOE-01 eet (D) time per drm: O.OOE4OD hours airborne respirable dust corcerttion: I.0E+01 mshn3 (D) respiratory protedion factor: I.OE+01 (D)

Step D: Transfer solids to storage average distance: 3.00E+00 feet (D) time per drum or palet QO.OE+0 hours Step E: Pump drurnmed oil to storage tank average distance: &OOE-O feet (D) tme per drum: O.OOE+O0 hours Storage before processing (3 steps)

Step A. Workers In solid waste storage area average distance: 3.00E+00 feet (0) duration. O.OOE+OO hours Step B: Transfer solids out average distance: 3.00E+00 feet (D) time per drn or paet O.OOE+00 hours Step C: Workers In liquid waste storage area average distance: 3.OOE+OO feet (D) duration: D.OOE+00 hours shWing thickness: 125E-01 inches (D)

Storage tank dimensions:

length: 7.00E+00 feet (0) wvith: 7.001E2+0 feet (D) heightL 120E+01 feet (0)

Burial at onsite landfill (4 steps)

Number of Workers: 20OEO0 Durnp truck bed dmernslons for steps A.C, and D):

lemgth: 200E+01 feet width: 7.50E+00 feet eight. 4.50E+00 feet Step A: Unload waste to mixing pit average ditance: 5.00E00 feet (D) duratin: O.OOE+0O hours shiefing thickness: 125E-01 Inches (D) airbone respable dust concenttort 1.0E+0O mghn3 (D) respiratory protection factor: 1.0E+00 (D)

Step B: Mix waste In mixing pit averagedistance: 1.OOE+01 feet (D) duration: O.W0E+00 hours cover thickness: 2.004(E+1iches (D)

Mxing pit dimensions:

length: 1.00E+01 feet (D) wvM I.00E+01 feet (D) deph: 1.OOE+01 feet (0) cover thickness: 2.OOE+00 inches (D) 3

Burial at onsito landfill (cont'd)

Step C: Load truck and transport to landfill average distance: 5.O0E+0 fee (D) duation: 5.00E01 hours shiing thknesss: 125E01 indies (D)

Step D: Unload truck at landfill average distance: 5.00E400 feet (D) duration: 250E-01 hours (D) shiedngthc ss: 1.25E01 inches (D) 4

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