CY-05-057, Supplemental Information Regarding the Request or Approval of Proposed Procedures, in Accordance with 10 CFR 20.2002

Supplemental Information Regarding the Request or Approval of Proposed Procedures, in Accordance with 10 CFR 20.2002

ML050680216
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/01/2005
From:	Gerard van Noordennen Connecticut Yankee Atomic Power Co
To:	Document Control Desk, NRC/FSME
References
CY-05-057, FOIA/PA-2005-0203
Download: ML050680216 (14)

Text

,' CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD

• EAST HAMPTON, CT 06424-3099 MAR - 1 2005 CY-05-057 Docket No. 50-213 RE: 10 CFR 20.2002 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D C 20555 Haddam Neck Plant Supplemental Information Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002 In a letter dated September 16, 20041,Connecticut Yankee Atomic Power Company (CYAPCO) proposed to transfer certain 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 US Ecology Idaho Facility, located in Grand View, Idaho.

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 NRC limits for exposure to members of the public of 25 millirem/yr TEDE. This assessment was provided to the NRC by letter dated September 16, 2004.1 By letter dated December 17, 20042, provided an on-site survey limit for the disposition of waste in Intermodal-type containers that can be shipped to US Ecology Idaho disposal site.

' G. H. Bouchard (CYAPCO) letter to the US NRC Document Control Desk, dated September 16, 2004, "Request for Approval of Proposed Procedures in accordance with 10 CFR 20.2002", CY-04-168.

2 G. van Noordennen (CYAPCO) letter to the US NRC Document Control Desk, dated December 17, 2004 "Supplemental Information Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002", CY-04-252.

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U. S. Nuclear Regulatory Commission CY-05-057 / Page 2 The purpose of this letter is to provide supplemental information requested by the NRC Staff in a teleconference with CYAPCO on February 22, 2005.

The above mentioned information requested by the NRC staff is in two subject areas:

1. Additional characterization information was not available for inclusion in the original submittal of this request.

2. On-site survey limits for various shipping containers other than the Intermodal-type which CYAPCO intends to utilize to ship waste to the US Ecology Idaho site.

These information needs are addressed as follows:

Characterization Information:

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

Containment Building Internal Walls and Floors:

The characterization data in the original submittal was considered limited only 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 and floors (which was done in the original submittal) was felt 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 have been 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 displayed in the revised Table 3 (for the significant radionuclides: H-3, C-14, Co-60 and Cs-1 37) along with the samples results from the original submittal for which H-3 and/or C-14 was analyzed. This data was reviewed 'against the conclusions made in the original submittal. The following was determined:

U. S. Nuclear Regulatory Commission CY-05-057 / Page 3

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.

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 to Co-60 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. 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) are lower then those presented in the original submittal.

Considering 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 waste Co-60 concentration. This change results in a change to the weighted average C-14 concentration for all the waste proposed for disposal at US Ecology (revised Table 8 enclosed) and a change in the projected total post closure dose calculation (revised Table 9 enclosed). These changes do not alter the conclusion of the original submittal 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. After movement of all the spent fuel out of the building, characterization will be done. The results of these samples will be compared to

U. S. Nuclear Regulatory Commission CY-05-057 / Page 4 the waste concentrations assumed in the original submittal. If the results show higher waste concentrations, the NRC will be informed as to the effect of these differences on the conclusions of this submittal. If the waste concentrations are below the values that have been presented, the samples results will be retained and be available for NRC inspection.

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

In addition to the Intermodal-type containers discussed in our December 17th letter, CYAPCO may be using B-25 type boxes to ship waste to US Ecology Idaho. An action level has been developed to identify when it is appropriate to transport waste in a B-25 box to US Ecology 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 December 17th letter 2 .

Using totally filled B-25 boxes containing the highest density material allowed by the package weight, a 1 meter dose rate of 2 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 US Ecology Idaho and those exhibiting higher dose rates need to 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 December 17th letter2. 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 will be followed.

Graph 1: Action Levels for B-25 Containers at Maximum Loading Dose Limits for Transfer to US Ecology for IOK+ lb. boxes 8

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12 40% 50% 6% L 70%

, 80% 90% 100%., 110%,,

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U. S. Nuclear Regulatory Commission CY-05-057 / Page 5 Although not expected, should other sized containers be used, appropriate dose rate limits will be determined using the same basis as shown above.

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

Ifyou should have any questions regarding this submittal, please contact me at (860) 267-3938.

Sincerely,

( ___

_ 3-/-°5 Gerard P. van Noordennen Date Regulatory Affairs Manager Attachment 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-05-057 Docket No. 50-213 Attachment I (Total of 8 pages)

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

Document Control Desk CY-05-057 / Attachment I / Page 1 Figure 1 Containment Floor Samples Floor Level 2.5" 2.5" 7,, Top of Concrete in Sump IC 1"

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2C " 2C Containment liner Typical core for floor locations Sump Location Sump Location 175 - 180 Core 185 Core 186 Note: _ wafers analyzed for radionuclides All dimensions are approximate C(I

Document Control Desk CY-05-057 / Attachment 1 / Page 2 Figure 2 Containment Internal Wall and Charging Floor Samples Outside Wall or Charging Floor Outside Wall Surface 2.5" 1 2.5" U 2.5" 2.5" 2.5" 2.5" 2.5" 7" 7"'

IC 4.5" Typical core for locations 4" 4"'

188, 190,191 and 192 4.5" 7"

2.5" 2.5" 2.5" 2.5" 2.5" Inside Wall Surface Lower Level Wall Mid Level Wall Location 187 Location 189 Note: _ wafers analyzed for radionuclides All dimensions are approximate

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, Document Control Desk CY-05-057 / Attachment 1 / Page 8 Table 9 (Revision 1)

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 (mremlyr per (pCi/g) (mremlyr) pCl/g)

H-3 1.045E-0' 261.88 2.737E-03 C-14 3.060E-01 - 5.53 1.692E+OG Mn-54 6.286E-25 1.67E-03 1.052E-27 Fe-55 O.OOOE+Ot 0.14 O.OOOE+Ot Co-60 1.653E-21 0.28 4.692E-22 Ni-63 O.OOOE+O0 1.69 O.OOOE+Ot Sr-90 O.OOOE+00 0.0277 O.OOOE+00 Nb-94 9.961 E-01 1.25E-03 1.246E-03 Tc-99 2.221 E-01 6.49E-03 1.441 E-03 Ag-108m 5.764E-01 2.04E-03 1.176E-03 Cs-134 5.881 E-2t 4.89E-03 2.875E-2E Cs-1 37 6.850E-27 0.97 6.674E-27 Eu-152 1.567E-2 5.01 E-03 7.854E-2t Eu-1 54 5.997E-23 3.81 E-03 2.286E-25 Eu-1 55 O.OOOE+00 3.85E-03 O.OOOE+Ot Pu-238 2.004E-06 3.69E-03 7.398E-09 Pu-239 O.OOOE+00 1.23E-03 O.OOOE+Ot Pu-241 O.OOOE+00 5.09E-02 O.OOOE+OC Am-241 O.OOOE+00 6.58E-03 O.OOOE+OC Cm-243 O.OOOE+00 1.11 E-03 O.OOOE+OC Total Post Closure Dose (mremlyr) ..*.1.699E+OO 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