CY-05-090, Additional Information Regarding Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002

Additional Information Regarding Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002

ML050960492
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/29/2005
From:	Bourassa J Connecticut Yankee Atomic Power Co
To:	Document Control Desk, NRC/FSME
References
+sispmjr200504, CY-05-090, FOIA/PA-2005-0203
Download: ML050960492 (8)

Text

I i CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD

• EAST HAMPTON, CT 06424-3099 CY-05-090 MAR 29 2005 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 Additional 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 In a letter dated December 17, 20042, CYAPCO provided an on-site survey limit for the disposition of waste in Intermodal-type containers that can be shipped to the US Ecology Idaho disposal site.

1 G. H. Bouchard (CYAPCO) letter to the US NRC, "Request for Approval of Proposed Procedures in accordance with 10 CFR 20.2002", dated September 16, 2004.

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

HMSS0 I

-Document Control Desk CY-05-090 / Page 2 In a letter dated March 1, 20053, CYAPCO provided supplemental information in two subject areas:

1. Additional characterization information (e.g., Containment Building walls and floors inside the containment liner) which 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.

The purpose of this letter is to provide additional information requested by the NRC staff in a teleconference with CYAPCO on March 23, 2005. The following modifications are made to the submittal of March 1, 2005:

Containment Building 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 enclosed Revision 2 of 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 (Revision 2 of Table 8 enclosed) and a change in the projected total post closure dose calculation (Revision 2 of 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". In addition, a new Table 10 is included which provides the basis for dilution factors used in determining Table 3 waste concentrations and a sample calculation for C-14 in containment floors and walls.

Spent Fuel Building Due to the operable status of the Spent Fuel Building, characterization has not been undertaken. Once all the spent fuel and GTCC waste is transferred from the spent fuel pool to the Independent Spent Fuel Storage Installation (ISFSI),

characterization of the Spent Fuel Building will be performed. 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 3 G. van Noordennen (CYAPCO) letter to US NRC, "Supplemental Information -

Request for Approval of Proposed Procedures in Accordance with 10 CFR 20.2002", dated March 1, 2005.

Document Control Desk CY-05-090 / Page 3 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 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.

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

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

Sincerely, 3)a'l OS o ph F. Bourassa Date irector, Nuclear Safety/Regulatory Affairs 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-05-090 Docket No. 50-213 Attachment 1 (Total of 4 pages)

Table 3 Containment Floor and Wall Samples, Revision 2 Table 8 - Average Waste Concentration Calculation, Revision 2 Table 9 - Post Closure Dose Calculation, Revision 2 Table 10 - Basis for Dilution Factors Used in Determining Table 3 Waste Concentrations March 2005

. Document Control Desk 6Y-05-090 t Attachment 1 /Page 3 Table 9 (Revision 2 dated 3/22105)

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

H-3 1.045E-05 261.88 2.737E-03 C-14 3.060E-01 9.69 2.964E+OO Mn-54 6.286E-25 1.67E-03 1.052E-27 Fe-55 O.OOOE+00 0.14 O.OOOE+00 Co-60 1.653E-21 0.28 4.692E-22 Ni-63 O.OOOE+00 1.69 O.OOOE+O0 Sr-90 O.OOOE+OO 0.03 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-26 4.89E-03 2.875E-28 Cs-137 6.850E-27 0.97 6.674E-27 Eu-1 52 1.567E-23 5.01 E-03 7.854E-26 Eu-154 5.997E-23 3.81 E-03 2.286E-25 Eu-1 55 O.OOOE+00 3.85E-03 O.OOOE+OO Pu-238 2.004E-06 3.69E-03 7.398E-09 Pu-239 O.OOOE+O0 1.23E-03 O.OOOE+0O Pu-241 O.OOOE+00 5.09E-02 O.OOOE+00 Am-241 O.OOOE+O0 6.58E-03 O.OOOE+00 Cm-243 O.OOOE+O0 1.I1 E-03 O.OOOE+00 Total Post Closure Dose (mrem/yr)[ 2.971E+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 italics

r.

• Document Control Desk CY-05-090 / Attachment 1 / Page 4 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 Walls Sample # Sample Dilution Factor Waste From Table 3 Concentration (pCi/g) From "A" Above Concentration (pCi/g)

1) 175-lC-01 720 divided by 9.6 equals 75.00

2) 176-IC-Ol 350 " 9.6 " 36.46

3) 185-IC-02 (Use 2nd wafer as first 1" of concrete to be disposed elsewhere) 0.5

4) 186- I C-02 (Use 2nd wafer as first 1" of concrete to be disposed elsewhere) 0.57

5) 187-IC 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-IC-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)/2)/4.8 = 100.6 Average Waste Concentration = 770.3/16=48.1