CY-01-084, Response to the Request for Additional Information (RAI) Regarding the Haddam Neck Plant License Termination Plan

Response to the Request for Additional Information (RAI) Regarding the Haddam Neck Plant License Termination Plan

ML012390116
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	08/22/2001
From:	Fetherston N Connecticut Yankee Atomic Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
+sispmjr200601, -nr, -RFPFR, CY-01-084, TAC MA9791
Download: ML012390116 (91)

Text

• CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT 362 INJUN HOLLOW ROAD - EAST HAMPTON, CT 06424-3099 August 22, 2001 Docket No. 50-213 CY-01 -084 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Response to the Request for Additional Information (RAI) Regarding the Haddam Neck Plant License Termination Plan (TAC NO. MA9791)

This letter forwards Connecticut Yankee Atomic Power Company's (CYAPCO's) response to NRC letter dated March 19, 20011. This request for additional information (RAI) consisted of twenty-seven questions arising from the NRC Staff's review of LTP Sections 6 through 8. The RAI questions and CYAPCO's responses to each question are provided as Enclosure 1 to this letter. Please note that the attached response to Question 2 is similar to that provided as the response to Question 51 (c) in the previous RAI. This response contains a table ("Listing of Radionuclide Sample Results from 60 Collected Samples") that was inadvertently omitted from the previous response and was subsequently faxed to the NRC Staff.

As a result of feedback received from the NRC Staff on our responses to the first RAI, we have taken more time to review our enclosed responses to the second RAI to ensure that all questions have been completely answered. This agreement was documented in CYAPCO's letter of August 15, 20012.

CYAPCO proposes to revise the LTP to address changes noted in the responses to the February 1, 2001 and the March 19, 2001, RAIs. Enclosure 2 to this letter is a matrix indicating the sections of the LTP proposed for revision, as a result of the attached RAI response. The LTP revision will be submitted within 30 days of receipt of NRC notice that our responses to the requests for additional information and proposed changes to the LTP are acceptable. The revised pages will contain the information submitted in these responses.

1.) NRC letter, "Haddam Neck Plant - Request for Additional Information Regarding The License Termination Plan (TAC NO. MA9791)," dated March 19, 2001.

2.) CYAPCO letter, "Response to the Request for Additional Information Regarding the Haddam Neck Plant License Termination Plan (TAC No. MA9791)," dated August 15, 2001.

Document Control Desk CY-01 -084/Page 2 The response to question 26, regarding LTP Section 7, Update of Site-Specific Decommissioning Costs, will be supplied by September 6, 2001.

CYAPCO is available to meet with the NRC Staff at your earliest convenience to discuss our responses to this request for additional information. Please let us know if you would like to meet with us to discuss the attached response. If you should have any questions, please call Mr. Gerry P. van Noordennen at (860) 267-3938.

Sincerely, Noah W Fetherston Site Manager

Enclosures:

1. Response to RAI Questions.

2. Matrix of LTP Sections Proposed for Revision.

Note: CD-ROM for Attachment 2 to Enclosure 1 provided as noted below.

cc: H. L. Miller, NRC Region I Administrator J. E. Donoghue, NRC Senior Project Manager, Haddam Neck Plant, w/CD-ROM R. R. Bellamy, Chief, Decommissioning and Laboratory Branch, NRC Region I, w/CD-ROM J. T. Greeves, Director, NRC Division of Waste Management C. L. Pittiglio, NRC NMSS Project Manager, Decommissioning, w/CD-ROM J. J. Cherniack, Office of Ecosystem Protection, US EPA Region I, w/CD-ROM E. L. Wilds, Jr., Director, CT DEP Monitoring and Radiation Division, w/CD-ROM

Enclosure 1 to CY-01-84 Response to RAI Questions

Enclosure 1 to CY-01-084 LTP SECTION 5, FINAL STATUS SURVEY PLAN

1. Section 5.4.6. Selection of derived concentrationguideline levels (DCGLs), page 5-8

- Please addressthe following issues to permit review of the information on operationalDCGLs for a survey unit:

a. Pleaseprovide text explaining how the operationalDCGLs developed for a given survey unit will be applied to the site as a whole. According to the descriptionprovided in Section 5.4.6.1 of the HNP License Termination Plan (LTP) on developing operationalDCGLs, it appearsthat the operational DCGLs are restrictive, but the licensee could have different soil and building DCGLs for each survey unit. It appearsthat the site as a whole could have many different soil and building DCGLs. Explain how the licensee would make sure in the future, when the site is releasedfor unrestricteduse, that the survey classificationused in developing DCGLs would still be valid.

The discussion contained in Section 5.4.6.1 does not specify how the balance between building surface and soil DCGLs will be applied to the different survey units. It was the intent, however, to apply this concept to the same areas as those impacted by the existing groundwater tritium as depicted on Table 6-11, and Section 5.4.6.1 of the LTP will be modified to clarify this intent. In these areas, the operational surface DCGL will be chosen at a value equal to or less than the base case building surface DCGL. This selection will then result in the calculation of the operational soil DCGL, which will be less than the base-case soil DCGLs. The final base case DCGLs will be based on results of testing to determine the site specific Kd values used to accurately calculate site-specific DCGLs.

Survey area classification is an integral part of survey design and is dependent on the DCGLs applied. Therefore, the Final Status Survey process described by the LTP includes verifying that a survey unit is properly classified (Section 5.3) and includes a classification history in the Final Status Survey report (Section 5.9).

Section 5.4.6.1 will be revised to clarify this point during a future change to the LTP.

b. Pleaseprovide a map or table identifying which type of DCGLs will be used in each survey unit at the site.

Attachment 1 identifies which type of DCGLs will be used in each survey unit at the site. This table will be added to LTP Section 5.4.6.1.

2. Section 5.4.6.3. Surrogate Ratio DCGLs. page 5 This section discusses surrogateratio DCGLs. Provide information on the specific areas at the site where these DCGLs would be used and explain how the activity ratios between difficult-to-detect radionuclidesand the easy-to-measure radionuclideswould be developed for a survey area.

The areas where surrogate ratio DCGLs will be used are limited to Class 1 survey units such as the Containment Building and the Primary Auxiliary Building I

Enclosure 1 to CY-01-084 (PAB). Class 2 and Class 3 building surface survey units are likely to be free from measurable activity. For these areas, a conservative DCGL corresponding to Co-60 will be applied.

The first set of RAI responses dated June 14, 2001, from CYAPCO to the NRC included a description of the process for developing surrogate ratio DCGLs in response to question #51(c). This response is effectively repeated below with some clarification provided.

Current data indicates that there is variability in the radionuclide mix at the HNP site, and accordingly there is not "a representative radionuclide mix". CYAPCO anticipates that, after some remediation is performed, the variability of radionuclide mix will be reduced. The table below shows the relative distribution of radionuclides and the applicable gross DCGL as calculated using Equation 5-2 of the LTP for 61 samples collected within the facility without excluding radionuclides using the 5% and 10% rule for elimination of radionuclides, as discussed on page 5-12 in the LTP. (The Am-241 fractions indicated in this example data were not used in the calculation of the corresponding DCGL since other transuranic radionuclides would be present, and would be accounted for by either direct measurements or by a surrogate relationship as discussed below.)

These samples include area contamination smears, and smears and samples of systems/components. Since the anticipated contamination on building surfaces originated from contamination within systems and components, it is reasonable to expect that samples from contaminated systems are representative of the profile of the facility during the current phase of the decommissioning process. As indicated in this table, 29 of the 61 samples were directly from building surfaces.

An analysis of this data indicates that the beta gamma radionuclide mix results in a DCGL ranging from a low of 11,700 dpm/1OOcm 2 for 100%

Co-60 to a high of 23,300 dpm/1 00cm 2 for nearly 100% Cs-1 37. This data also shows that the presence of the other radionuclides have little impact on the gross DCGL where the variability is principally dominated by the relative fractions of Co-60 and Cs-1 37. A statistical analysis of this data shows that the average DCGL is 16,133 dpm/1OOcm 2 with a coefficient of variance (%CV) of 22.1%. This is a reasonably precise distribution that will be generally applied for gross beta/gamma DCGLs during the Final Status Surveys (FSS) using the following process.

a. Prior to performing an FSS, the beta gamma radionuclide mix will be established for a survey unit using a representative number of samples (typically six or more) from the unit or from other unit(s) if the source of the potential contamination from other unit(s) can be reasonably demonstrated to be the same as for the FSS unit.

b. If the average gross beta/gamma DCGL from the above samples is within +/-20% of 16,133 dpm/1OOcm 2 then a gross beta/gamma DCGL of 16,133 dpm/1 00cm 2 will be applied to the survey unit.

2

Enclosure 1 to CY-01-084

c. If the above criterion is not met, one of the following actions will be taken:

i. Following a more detailed spatial analysis of the radionuclide mix distribution, the unit may be subdivided into separate survey units based on the spatial distribution. (For example, if one portion of an area is dominated by Co-60, unlike the remainder, this portion may be segregated into a separate unit.)

ii. The lowest DCGL from the observed radionuclide mix will be applied to the entire survey unit.

iii. Additional samples will be collected and analyzed to allow for a detailed analysis and documented evaluation of the radionuclide distribution resulting in the use of a specific DCGL for the survey unit.

iv. A DCGL of 11,700 dpm/100cm2 (corresponding to Co-60) will be applied.

d. In all cases, the applied gross beta/gamma surface contamination 2 DCGL will be bounded by that for Cs-1 37, or 23,300 dpm/100cm (from RAI #17 response).

e. For cases where the radionuclide mix cannot be readily established because of the absence of measurable contamination and the source of potential contamination cannot be identified with nearby areas, the gross beta gamma DCGL corresponding to Co-60 of 11,700 dpm/1OOcm 2 will be applied.

3

Enclosure 1 to CY-01-084 Table 1: Listing of Radionuclide Sample Results from 61 Collected Samples Radionuclide Fractions Location Mn-54 Co-60 Nb-94 Ag-110m Sb-125 Cs-134 Cs-137 Eu-154 Eu-155 Am-241 'Gross DCGL,

_____dpmllI O0cm2 AWJ Gun Smear 0.000 1.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11700 PZR Spray Line 0.000 0.998 0.000 0.000 0.000 0.000 0.002 0.000 0.000 0.000 11712 Resin Drum Smear 0.002 0.994 0.000 0.000 0.002 0.000 0.000 0.001 0.000 0.002 11738 Loop #4 I/S CH Pipe 0.000 0.973 0.000 0.000 0.000 0.000 0.007 0.005 0.000 0.015 11774 MDM Cut Tool 0.011 0.989 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11784 AVAN Pump Skid 0.012 0.988 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11790 MDM Bridge Smear 0.012 0.988 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11796 AWJ I/S Hose 0.017 0.983 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11833 AWJ Work Area 0.018 0.982 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11837 AWJ Smears 0.020 0.980 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11857 AWJ Head 0.021 0.979 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11863 PCI Bridge 0.022 0.978 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11873 PAB Cut #90 0.000 0.943 0.000 0.000 0.000 0.000 0.021 0.014 0.000 0.022 11910 Charging Floor Composite 2 0.029 0.971 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 11923 Loop #2 SI Pipe I/S 0.008 0.956 0.000 0.007 0.010 0.000 0.003 0.002 0.000 0.014 12006 CCW Surge Tank 0.000 0.911 0.000 0.000 0.000 0.011 0.079 0.000 0.000 0.000 12222 RHR Pit2 0.000 0.865 0.000 0.000 0.000 0.007 0.118 0.000 0.000 0.010 12467 ADHUT Water 0.002 0.857 0.000 0.001 0.002 0.007 0.128 0.001 0.000 0.003 12588 ADHUT Water 0.002 0.856 0.000 0.001 0.001 0.007 0.129 0.001 0.000 0.003 12590 ADT Sludge 0.002 0.846 0.000 0.002 0.002 0.010 0.135 0.000 0.000 0.001 12672 ADT Sludge 0.002 0.842 0.000 0.002 0.002 0.011 0.140 0.000 0.000 0.001 12694 ADT Sludge 0.001 0.820 0.000 0.002 0.003 0.012 0.158 0.001 0.000 0.003 12848 Drain Header Pipe Smear 0.000 0.614 0.000 0.000 0.000 0.000 0.183 0.062 0.007 0.135 13730 PAB Pipe Trench 2 0.000 0.637 0.000 0.000 0.003 0.019 0.323 0.002 0.000 0.015 14162 CTMT Purge Filter Piece 0.027 0.609 0.000 0.000 0.000 0.000 0.364 0.000 0.000 0.000 14608 Loop #3 M/L - L/L 0.005 0.524 0.000 0.000 0.000 0.003 0.467 0.000 0.000 0.001 15335 Composite2 ADTs Smears 2 0.000 0.505 0.000 0.000 0.000 0.026 0.463 0.001 0.000 0.005 15415 Loop #4 LUL #42 0.000 0.473 0.000 0.000 0.000 0.000 0.527 0.000 0.000 0.000 15864 Loop #3 M/L #392 0.000 0.457 0.000 0.000 0.000 0.007 0.530 0.000 0.000 0.007 15978 Loop #2 LUL #132 0.000 0.446 0.000 0.000 0.000 0.000 0.554 0.000 0.000 0.000 16155 CTMT Purge Filters 0.000 0.444 0.000 0.000 0.000 0.000 0.556 0.000 0.000 0.000 16183 Boron Recovery LL2 0.000 0.423 0.000 0.000 0.000 0.000 0.516 0.016 0.000 0.046 16186 CTMT L/L O/S Column 0.000 0.419 0.000 0.000 0.000 0.015 0.566 0.000 0.000 0.000 16402 10-122 LLOA Elec Penetration 2 0.000 0.415 0.000 0.000 0.000 0.016 0.563 0.002 0.000 0.004 16432 PAB Pipe Trench Water 0.000 0.292 0.002 0.000 0.000 0.035 0.654 0.003 0.000 0.014 17839 Loop #4 M/L #42z 0.000 0.299 0.000 0.000 0.000 0.011 0.690 0.000 0.000 0.000 17928 LLOA Tent Overhead 0.000 0.295 0.000 0.000 0.000 0.004 0.698 0.000 0.000 0.003 17999 WDB LL Overhead 2 0.000 0.149 0.000 0.000 0.000 0.000 0.173 0.202 0.000 0.476 18038 Smear #16 Loop I2 0.000 0.276 0.000 0.000 0.000 0.009 0.713 0.000 0.000 0.002 18243 Loop 1 LL2 Smears 0.000 0.255 0.000 0.000 0.000 0.011 0.734 0.000 0.000 0.000 18545 (Percon )

Loops M/L2- L/L 0.000 0.255 0.000 0.000 0.000 0.006 0.739 0.000 0.000 0.000 18574 Composite Loop 1 Floore 0.000 0.225 0.000 0.000 0.000 0.008 0.765 0.000 0.000 0.003 19003 RCP #1 Bolt Smears 0.000 0.224 0.000 0.000 0.000 0.008 0.765 0.001 0.000 0.002 19011 Loop #1 L/L #19 2 0.000 0.216 0.000 0.000 0.000 0.008 0.773 0.000 0.000 0.003 19135 Loop #4 M/L & LL 0.000 0.216 0.000 0.000 0.000 0.009 0.772 0.001 0.000 0.002 19139 4

Enclosure 1 to CY-O1-084 Radionuclide Fractions 1

Location Mn-54 Co-60 Nb-94 Ag-110m Sb-125 Cs-134 Cs-137 Eu-154 Eu-155 Am-241 Gross DCGL

________dpm/1 O0cm2 Compositez Loop #3 L/L #732 0.000 0.212 0.000 0.000 0.000 0.000 0.788 0.000 0.000 0.000 19246 Smear #15 Loop 1 0.000 0.201 0.000 0.000 0.000 0.015 0.780 0.000 0.000 0.005 19344 Loop #1 Floor2 0.000 0.179 0.000 0.000 0.000 0,008 0.808 0.000 0.000 0.004 19729 (Interference) 2 Loops M/L 0.000 0.171 0.000 0.000 0.000 0.000 0.820 0.000 0.000 0.009 19889 WDB 18 6 Smears" 0.000 0.120 0.000 0.000 0.000 0.117 0.763 0.000 0.000 0.000 20144 2

WDB 35 6 Smears 0.000 0.121 0.000 0.000 0,000 0.108 0.763 0.000 0.000 0.007 20152 Loop #4 Grating 0.000 0.149 0.000 0.000 0.000 0.006 0.840 0.000 0.000 0.005 20255 WDB 02 0.000 0.139 0.000 0.000 0.000 0.031 0.831 0.000 0.000 0.000 20308 Loop #2 M/L #372 0.000 0.144 0.000 0.000 0.000 0.000 0.856 0.000 0.000 0.000 20393 Smear #12 Loop 12 0.000 0.118 0.000 0.000 0.000 0.010 0.869 0.000 0.000 0.003 20787 Loop #2 M/L - L/L 0.000 0.117 0.000 0.000 0.000 0.005 0.879 0.000 0.000 0.000 20860 Composite2 LLOA Conduit Box 0.001 0.067 0.000 0.000 0.000 0.132 0.800 0.000 0.000 0.000 21022 Loop #1 M/L #48 0.000 0.098 0.000 0.000 0.000 0.009 0.893 0.000 0.000 0.000 21176 Loop #1 M/L - L/L 0.000 0.079 0.000 0.000 0.000 0.004 0.917 0.000 0.000 0.000 21581 Composite I RCP #1 Bolts and Seal 0.000 0.052 0.000 0.000 0.000 0.074 0.874 0.000 0.000 0.000 21658 Loop #1 Smear 0.000 0.050 0.000 0.000 0.000 0.004 0.946 0.000 0.000 22166 Average= 16133

%CV= 22.1 1

Am-241 fractions were not used in determining this DCGL.

2 These samples were from building surfaces.

In order to address the potential for transuranic radionuclide contamination for gross surface contamination measurements, one of two processes will be employed; the use of a surrogate relationship to the beta gamma contamination, or direct measurement of alpha contamination. Equation 5-4 of the LTP will be applied to establish a surrogate DCGL for each sample used to determine the radionuclide mix for a survey unit. This equation is provided here for completeness as:

DCGLDTD DCGLsuoga(e =1DCGLETD x (fDTD:ETD x DCGLETD )+ DCGLDTD where DCGLETD = the DCGL for the easy-to-detect radionuclide; DCGLDTD = the DCGL for the difficult-to-detect radionuclide; and fDTD:ETD = the activity ratio of the difficult-to-detect radionuclide to the easy-to-detect radionuclide; or the inverse of the beta gamma to alpha ratio, R, in the case of transuranic radionuclides.

The table below provides values of fDTD:ETD for beta-gamma to alpha ratios ranging from 10:1 up to 10,000:1. These fractions are then used with the formula above to assess the impact on the surrogate DCGL using the range of possible values of DCGLETD and a conservative value of 5

Enclosure 1 to CY-Ol-084 DCGLDTD for all alpha emitting radionuclides corresponding to Am-241 of 437 dpmllOOcm 2 (See RAI #17 response). As indicated in this table, a beta-gamma to alpha ratio of 500:1 or more does not change the surrogate DCGL by more than approximately 10% for all ranges of the gross DCGL. For the most likely DCGL of 16,133 dpm/1OOcm 2, the surrogate DCGL is 15,023 which represents a change of less than 7%.

DCGLETD = 11,700 DCGLETD = 16,133 DCGLETD = 23,300 3eta-Gamma to Alpha Activity FDTD:ETD DCGLsur,,,at  % DCGLsu,-,rate  % DCGLs,,,,ogate  %

Change Ratio, R (l/R) DCGLsurrogate DCGLETD Change DCGLsun,oge DCGLETO Change DCGLsur,,gate /DCGLErD 10 1.00E-01 3182 0.272 267 3439 0.213 369 3680 0.158 533 25 4.OOE-02 5650 0.483 107 6514 0.404 147 7438 0.319 213 75 1.33E-02 8622 0.737 35.7 10811 0.670 49.2 13619 0.584 71.1 100 I.00E-02 9229 0.789 26.8 11782 0.730 36.9 15197 0.652 53.3 200 5.OOE-03 10319 0.882 13.4 13618 0.844 18.5 18396 0.790 26.7 400 2.50E-03 10966 0.937 6.69 14769 0.916 9.23 20560 0.882 13.33 500 2.00E-03 11105 0.949 5.35 15023 0.931 7.38 21055 0.904 10.67 700 1.43E-03 11269 0.963 3.82 15324 0.950 5.27 21651 0.929 7.62 1000 1.00E-03 11395 0.974 2.68 15558 0.964 3.69 22121 0.949 5.33 1500 6.67E-04 11495 0.982 1.78 15745 0.976 2.46 22500 0.966 3.55 2000 5.OOE-04 11545 0.987 1.34 15840 0.982 1.85 22695 0.974 2.67 10000 1.00E-04 11669 0.997 0.27 16073 0.996 0.37 23176 0.995 0.53 I

This analysis shows that a beta-gamma to alpha ratio of 500:1 or higher would allow for the elimination of the TRU nuclides based on the 5% and 10% elimination rule discussed in the LTP. Therefore, the following process will be applied to assess the presence of TRU radionuclides for final status surveys (FSS).

a. Prior to performing an FSS, the beta gamma to alpha ratio will be established for a survey unit using a representative number of samples (typically six or more) from the unit or from other unit(s) if the source of the potential contamination from other unit(s) can be reasonably demonstrated to be the same as for the FSS unit.

These samples will be analyzed for TRU (using gross alpha or alpha spectroscopy techniques) and for beta gamma activity (using gross beta analysis and/or gamma spectroscopy techniques).

b. If all sample results indicate that the average beta-gamma to alpha ratio is greater than 500:1, then a surrogate DCGL will not be determined, otherwise, the following process will be applied.

c. A surrogate DCGL will be determined from the samples collected for the survey unit as described above. If this data indicates that the %CV of the average surrogate DCGL has a value of 20% or 6

Enclosure 1 to CY-01-084 less, the average surrogate DCGL will be applied to the survey area. If this criterion is not met, the following steps will be applied.

Following a more detailed spatial analysis of the radionuclide mix distribution, the unit may be subdivided into separate survey units based on the spatial distribution.

ii. The lowest surrogate DCGL from the observed radionuclide mix will be applied to the entire survey unit.

iii. Additional samples will be collected and analyzed to allow for a detailed analysis and documented evaluation of the radionuclide distribution resulting in the use of a specific DCGL for the survey unit.

It is anticipated that the above processes will be approved for use in the final LTP thereby eliminating the need for providing time for regulatory review prior to conducting an FSS for a survey unit.

The process described above will be included in a future change to the LTP.

3. Section 5.4.6.4. Elevated Measurement Comparison (EMC) DCGLs, page 5-16 This section calculates area factors for both the resident farmerand the building occupancy scenario. This section states that the dose decreases more slowly with decreasingarea for the direct exposure pathway than for otherpathways (e.g., inhalation,ingestion) in either scenario and, therefore, only the direct exposure pathway was included in calculationsof areafactors. Further,this section states that since Co-60 represents the limiting radionuclidein terms of direct exposure per unit activity at the site, only Co-60 needed to be used in calculatingarea factors for both scenarios.

a. Providejustification, by calculating area factors for all radionuclidesof concern at the site, that the area factor calculatedfor Co-60 by using only the external exposure pathway active would result in most conservative area factors for a survey unit. If this turns out not to be true, then provide area factors for all radionuclidesof concern by consideringall pathways active for both scenarios.

LTP Section 5.4.6.4 calculated area factors based upon Co-60 because it is one of the most dominant radionuclides expected to be detected at the HNP site. As requested, area factors for both the resident farmer and building occupancy scenarios have been re-calculated for all radionuclides of concern at the site considering all potential pathways of exposure. The re-calculated area factors differ slightly from the area factors presented in LTP Tables 5-7 and 5-8, which were based only on the Co-60 external pathway.

For the resident farmer scenario, the most limiting radionuclide is Eu-1 55. For the building occupancy scenario, the most limiting radionuclide is Mn-54. These two radionuclides have the smallest area factors because nearly all of their dose comes from the external pathway. Other radionuclides which also yield an internal dose component (ingestion or inhalation) have higher area factors. That is, their dose is reduced more by decreases in the contaminated area. For example, LTP Table 6-8 shows that Co-60 has an external pathway dose 7

Enclosure 1 to CY-01-084 contribution of 94.1%, while its internal pathway dose contributions from plant, meat, and milk is 5.9%. Considering all pathways, the soil area factor for Co-60 at 1 m 2 is 10.1. Considering only the external pathway would yield a slightly smaller area factor for Co-60 (9.8).

The area factors listed in the following table represent the most conservative radionuclides and will be applied to all applicable surveys. For comparison, the area factors from LTP Tables 5-7 and 5-8 are also listed.

Area Factors Soil: Resident Farmer Building Surfaces: Building Occupancy Minimum of All LTP Table 5-7 Co Minimum of All LTP Table 5-8 Co Contaminated Radionuclides (All 60 (External Radionuclides (All 60 (External Area (m2) Pathways) Pathway) Pathways) Pathway) 10000 1.0 1.0 7500 1.0 1.0 5000 1.0 1.0 2500 1.0 1.0 1000 1.0 1.1 750 1.1 1.1 500 1.1 1.1 250 1.1 1.1 -

100 1.2 1.2 1.0 1.0 75 1.2 1.3 1.1 1.1 50 1.3 1.4 1.2 1.2 25 1.5 1.6 1.6 1.6 10 1.9 2.1 2.4 2.4 8 2.2 2.4 2.8 2.7 6 2.5 2.9 3.3 3.3 4 3.2 3.7 4.3 4.2 2 4.9 5.8 7.1 7.1 1 8.0 9.8 12.7 12.7 8

Enclosure 1 to CY-01-084 The most restrictive area factors are presented graphically below. This figure illustrates that area factors will generally be applied only to relatively small survey units (smaller than 10 M2 ).

Limiting Area Factors 14.0 13.0 12.0 11.0 10.0 9.0 8.0 0I 7.0 1.

U!

a=

a=

6.0 5.0 4.0 3.0 2.0 1.0 0.0 10000 1000 100 10 Area of Contaminated Zone (m2)

The area factor tables in the LTP (Tables 5-7 and 5-8) and Section 5.4.6.4 will be updated to reflect the recalculated values. This will be included in a future change to the LTP.

b. To make the building occupancy scenario consistent in the approach used in deriving DCGL and elevated measurement comparison DCGL values, use RESRAD-BUILD in computing area factors.

Area factors for the building occupancy scenario have been re-calculated using RESRAD-BUILD 2.37 to be consistent with the derivation of DCGLs. LTP Section 5.4.6.4 will be revised to indicate the use of RESRAD-BUILD 2.37. This will be reflected in a future change to the LTP.

LTP SECTION 6, COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION

4. Section 6.1. Site Release Criteria,page 6 According to NUREG-1727, "NMSS Decommissioning StandardReview Plan," the direct (extemal-gamma) dose from the pipes will be in addition to the total effective dose equivalent from the residual radioactivityon the surfaces in the room. To verify that the dose to the average member of the criticalgroup is not likely to exceed the 25 mrem annual dose criterionin 10 CFR 20.1402, please provide text explaining whether all the associatedstructuralcomponents (embedded piping) from a building would be 9

Enclosure 1 to CY-01-084 removed. If not, please provide text explaining how the dose from the structural (ornon-structural)components that may not be removed during decommissioning would be calculated and incorporatedinto the dose estimation from other media (soil, building, and ground water).

CYAPCO considers non-structural systems and components to be anything not attached to, or not an integral part of, a building or structure (LTP Section 5.6).

Examples of non-structural systems and components include pumps, motors, heat exchangers, and piping between components. Experience to date at HNP has typically shown that non-structural systems and components will probably be removed during the decommissioning process. For instance, many non-structural systems and components have already been removed from the PAB and will be ultimately processed for disposal as radioactive waste.

CYAPCO anticipates that the amount of non-structural systems and components remaining in an area for FSS will be minimal. Non-structural systems and components that remain for Final Status Survey will be surveyed to Regulatory Guide 1.86 residual activity surface contamination limits (LTP Sections 5.6 and 6.1). Those non-structural systems and components that might remain would either have no history of radioactive contamination (e.g., Circulating Water) or would have a high probability of meeting the proposed release criteria for non structural systems and components without extensive remediation. Surveys to date (e.g., 1999 HNP Composite Dry Active Waste Analysis) indicate that the residual radioactivity in these non-structural systems and components will be beta-gamma emitting with Cs-137 and Co-60 being the predominate radionuclides by activity fraction.

The Regulatory Guide 1.86 release criterion for total beta-gamma activity is 5000 dpm/1OOcm 2 (LTP Table 5-9). The potential dose from residual activity residing on non-structural systems and components would result primarily from external radiation since resuspension of radioactivity at levels at or less than Regulatory Guide 1.86 release criteria would not result in significant airborne concentrations.

Furthermore, burial of non-structural systems and components on-site is not an option described in the LTP, so modeling other scenarios (e.g., resident farmer) for potential dose pathways is not necessary. Assuming that the total beta gamma activity is entirely based on Co-60 contamination yields a conservative value for direct (external gamma) radiation. The Regulatory Guide 1.86 release criterion of 5000 dpm/1 00cm 2 total beta-gamma activity is 43% of the Co-60 surface DCGL value of 11700 dpm/1 00cm 2 (LTP Table 6-24). Assuming the Co-60 surface DCGL represents the bounding gross activity DCGL it is highly unlikely that a non-structural system or component meeting the Regulatory Guide 1.86 release criteria would contribute more than 50% to-the annual dose criterion specified in 10CFR20.1404.

The Final Status Survey (FSS) package for each survey unit will describe the non-structural systems and components remaining and document the measurements and/or assessments used to demonstrate compliance with Regulatory Guide 1.86 release criteria including remedial actions performed. If the Final Status Survey for the survey unit demonstrates that residual radioactivity is less than 50% of the DCGL then the potential dose from the remaining non-structural systems and components will not be modeled and 10

Enclosure 1 to CY-01-084 accounted for towards compliance with the annual dose criteria. If the FSS for the survey unit shows that residual radioactivity exceeds 50% of the DCGL, then the contribution of dose from remaining non-structural systems and components will be evaluated. The resulting dose from remaining non-structural systems and components will be accounted for towards compliance with the Total Effective Dose Equivalent whenever the resulting dose exceeds one (1) mrem.

Like non-structural systems and components, embedded piping will either be removed or decontaminated to meet site release criteria. The first set of RAI responses dated June 14, 2001, from CYAPCO to the NRC included an approximation of the amount of piping expected to remain embedded in concrete at the time of the Final Status Survey. The length of piping is expected to be less than 200 feet as stated in response to RAI #34 (b). Much of the remaining embedded piping will consist of wall or floor penetrations (e.g., cable conduit) that are not expected to be very long (i.e., no longer than the thickness of the wall or floor that the pipe penetrates). These sections of embedded piping will be surveyed or assessed for residual radioactivity as described in Section 5.7.3.1.4 of the License Termination Plan. The FSS package for each survey unit will describe embedded piping left behind and document the measurements and/or assessments used to demonstrate the release criteria including remedial actions performed. The same method of dose accounting described above for remaining non-structural systems and components will be applied to this embedded piping.

This information will be included in a future change to the LTP.

5. Section 6.4. Development of DCGLs for Soil, page 6 To permit review of the input parametervalues used in dose evaluationsand the derivationof DCGL values for soil in the L TP, please provide electronic files of the inputs and outputs of the RESRAD code used in deriving the base case and special case DCGL values.

The input data files and program output files are included on the enclosed CD-ROM. Also included is the original RESRAD v 5.91 file. Attachment 2 contains details regarding the file structure on the CD-ROM and instructions on how to install the RESRAD code. No change to the LTP is considered necessary in response to this RAI.

6. Section 6.4.2, ConceptualModel, ContaminantCharacteristics,pages 6-6 to 6-7

- Table 6-2 lists distribution coefficient values for radionuclidesof concern at the site. The justification for use of these values is that most are taken from NUREG/CR-5512, "ResidualRadioactivity Contaminationfrom Decommissioning." These default Kd values have no meaning in terms of their relative conservatism or appropriatenessoutside of the context of the DandD code. Provide analysis using site-specific distributioncoefficient values, or demonstrate that the NUREG/CR-5512 default values are appropriatein this approach.

The soil distribution coefficient, Kd, is an intrinsic property of soil, which represents the ratio of solute concentration in soil to that in solution under equilibrium conditions. In DandD (Reference 1) and RESRAD (Reference 2), Kd is used to determine the retardation factor according to the following definitions:

11

Enclosure I to CY-01-084 R=l+PbKd (DandD) n R =1 + pbKd (RESRAD) 0 In the above equations, R = retardation factor; Pb = soil bulk density; Kd = soil distribution coefficient; 0 = volumetric water content (nxf); n = total porosity; and f

= saturation ratio. [Note: According to Reference 1, "Evaluation of the retardation factor is based on total porosity, n, rather than the volumetric water content, 0, for conservatism... because the total porosity, and thus all sorption sites, comes into play as the pulses of moisture move through the surface and the unsaturated layers." Hence, use of the volumetric water content, assuming a saturation ratio of unity, as was done in the LTP, results in an identical retardation factor for both the RESRAD and DandD calculated values.]

The retardation factor itself is an integral part of the radionuclide leach rate, L.

This leach rate is represented the same way in both DandD (Reference 1, Equation 4.10, page 4.9) and RESRAD (Reference 2, Equation E.3, page 198).

Therefore, given that Kd is an intrinsic property of the soil and that this parameter is used the same way in both DandD and RESRAD, any valid input value that is applicable to DandD should also be a valid input to RESRAD.

The default soil Kd values given in NUREG/CR-5512, Volume 2 (Reference 3, Table D.22, page D-18) were adopted for use in the dose modeling. These values were adopted based on the following recommendation included in this NUREG (Reference 3, page 1-1):

"The default parameter values in DandD Version 1.0 have been defined through a systematic process of assessing the variability of each parameter across the U.S. and then defining default values that produce generic dose estimates that are unlikely to be exceeded at any real site."

Based on the above statement, use of these default values in RESRAD should therefore lead to conservative dose estimates.

It should be further noted that results of dose modeling for soil indicate that 17 of 20 radionuclides, potentially present at the site at the anticipated time of Final Status Survey, derive most of their respective doses from water independent pathways (Reference 4, Table 6-8), indicating no sensitivity to the adopted Kd value. Water dependent pathways determined the dose for only C-14, Pu-238, and Pu-239. Therefore, the adopted Kd values only influence the soil DCGLs for these three radionuclides.

Based upon the above discussion, the Kd values adopted for modeling from NUREG/CR-5512 are appropriate for use and should result in conservative dose estimates.

12

Enclosure 1 to CY-01-084

7. Section 6.4.2. ConceptualModel. Contaminant Characteristics,page 6 Table 6-2 lists 20 radionuclidesfor which DCGLs are developed. According to NUREG/CR-3474, "Long-Lived Activation Products in Reactor Material," and NUREG/CR-0130, "Technology, Safety and Cost of Decommissioning,"many more radionuclidescould be associatedwith power plant operations.To verify that the list of radionuclidesfor which DCGL values were developed is complete, please provide text explaining why DCGLs were developed for only 20 radionuclidesandjustify the characterizationused for selecting these radionuclidesand not others.

CY developed Table 6-2 using available waste characterization data. The list of 20 radionuclides represents the significant radionuclides which potentially contribute to the residual dose from HNP.

The list is consistent with the regulatory guidance for radionuclides of concern in bio-shield wall concrete, rebar and surface contamination. NUREG/CR-3474, "Long-Lived Activation Products in Reactor Material", Tables 5.4 and 5.6, and NUREG/CR-01 30, "Technology, Safety and Cost of Decommissioning", Tables 7.3-5, 7.3-11 and 7.3-14 are used as references for this determination. Many radionuclides in the regulatory guidance are not applicable to the HNP radionuclide suite based on short half-life, low relative abundance, or inconsequential dose contribution.

Table 6-2, as provided in the License Termination Plan, represents a complete and conservative list of radionuclides for which DCGLs should be established for the HNP Final Status Survey. However, as a result of question 7 above, CYAPCO has reconsidered the list of radionuclides provided in Table 6-2.

Table 1, is a matrix of the radionuclides that appear in the subject NUREGs (with the exception of the noble gases) and those included in Table 6-2 of the LTP.

The following discussion provides justification for the selection of the radionuclides provided in Table 6-2.

13

Enclosure 1 to CY-01-084 Table 1 Radionuclides Listed in the Referenced NUREGs NURE:G/CR-3474 NURPE.OU/CR-0130 IY LVT Isotope Half Life (y) I able .5.4 I able O.3 Iladle 7.3-5.1 aole 7.3-11 1ladle 7'.3-14 1 1ade 6 H-3 1.24E+01 Y C-1 4 5.73E+03 y Na-22 2.60E+00 N P-33 6.96E-02 N

S-35 2.40E-01 N

CI-36 3.01E+05 N Ca-41 1.40E+05 N

Ca-45 4.47E-01 N Sc-46 2.30E-01 N Cr-51 7.59E-02 N Mn-53 3.70E+06 N Mn-54 8.56E-01 Y Fe-55 2.70E+00 Y

Fe-59 1.22E-01 N Ni-59 7.51 E+04 Y Co-57 7.42E-01 N Co-58 1.94E-01 N Co-60 5.27E+00 Y Ni-63 9.61E+01 Y Zn-65 6.68E-01 N Se-79 6.50E+04 N

Sr-89 1.38E-01 N Sr-90 2.91 E+01 Y

Y-90 7.31E-03 N Nb-92m 3.60E+07 N Nb-95m 9.89E-03 N Zr-93 1.53E+06 N Mo-93 3.50E+03 N Nb-94 2.03E+04 Y Zr-95 1.75E-01 N Tc-99 2.13E+05 Y Ru-103 1.08E-01 N Ag-108m 1.27E+02 N Ag-1 10m 6.85E-01 N Sn-121m 5.50E+01 N Sb-124 1.65E-01 N Sb-125 2.77E+00 N Te-129m 9.21E-02 N 1-129 1.57E+07 N

1-131 2.20E-02 N 1-133 2.37E-03 N

Ba-133 1.07E+01 N Cs-134 2.06E+00 Y Cs-135 2.30E+06 N Cs-136 3.59E-02 N Cs-137 3.00E+01 Y Ba-140 3.49E-02 N La-140 4.60E-03 N Ce-141 8.90E-02 N Ce-143 3.77E-03 N Pm-145 1.77E+01 N Sm-146 1.03E+08 N Sm-151 9.01E+01 N Eu-152 1.33E+01 Y Eu-154 8.81E+00 Y Eu-155 4.96E+00 Y Tb-158 1.50E+02 N Ho-166m 1.20E+03 N Hf-178m 3.10E+01 N Pb-205 1.43E+07 N U-233 1.59E+05 N Pu-238 8.78E+01 Y Pu-239 2.41E+04 Y 14

Enclosure I to CY-01-084 Screening based on Half-life:

Considering the radionuclides listed in one or both of the referenced NUREGs, 25 radionuclides (shown in Table 2) have half-lives of less than one year. In this group,Y-90 is somewhat of a special case. Y-90 is the short lived daughter of a longer lived parent, Sr-90. Although no DCGLs are developed for Y-90, DCGLs for Sr-90 are included in the LTP, and the dose associated with the Y-90 daughter is included in the dose conversion factor associated with Sr-90. Mn-54 has a half-life less than one year, but since Mn-54 has been identified in several waste streams, it is listed in Table 6-2 of the LTP. Based on the time since the reactor shut down and the projected date for the completion of decommissioning, the short lived (i.e. T112 < 1 year) radionuclides present at shutdown will have gone through at least 5 half-lives. Any remaining concentrations of these short lived radionuclides, will result in negligible dose. In accordance with guidance in Appendix E of NUREG-1727 (Reference 8a), radionuclides contributing less than 10 percent of the dose limit can be screened out.

15

Enclosure 1 to CY-01-084 Table 2 NUREG Radionuclides with Half-lives less than one year (sorted by half-life)

NUREG/CR-3474 NUREG/CR-0130 CY LTP Isotope Half Life (y) f able-5.4 I able lAF 7.3-5 Table r -2 I able (.3-11 1able 7.3-14 lable 1-133 2.37E-03 N N N N Y N Ce-143 3.77E-03 N N N N Y N La-140 4.60E-03 N N N N Y N Y-90 7.31 E-03 N N N N Y N Nb-95m 9.89E-03 N N N N N N 1-131 2.20E-02 N N N Y Y N Ba-140 3.49E-02 N N N N Y N Cs-1 36 3.59E-02 N N N Y Y N P-33 6.96E-02 N N Y N N N Cr-51 7.59E-02 N N Y Y Y N Ce-141 8.90E-02 N N N N Y N Te-129m 9.21 E-02 N N N Y N N Ru-103 1.08E-01 N N N N Y N Fe-59 1.22E-01 N N Y Y Y N Sr-89 1.38E-01 N N N Y Y N Sb-124 1.65E-01 N N N N Y N Zr-95 1.75E-01 N N N Y Y N Co-58 1.94E-01 N N Y Y Y N Sc-46 2.30E-01 N N Y N N N S-35 2.40E-01 N N Y N N N Ca-45 4.47E-01 N N Y N N N Zn-65 6.68E-01 Y Y Y N N N Ag-110m 6.85E-01 N N N N Y N Co-57 7.42E-01 N N N N Y N Mn-54 8.56E-01 Y Y Y Y Y Y 16

Enclosure 1 to CY-01-084 Screening Based on Low Relative Abundance (Limited Presence on Site):

Many radionuclides identified in the referenced NUREGs were not detected in any of the waste streams at HNP and thus were not included in LTP Table 6-2.

However, CYAPCO has reconsidered the longer lived radionuclides that were listed in the referenced NUREGs. (It may also be noted in the table below that some of the transuranic radionuclides listed in the LTP are not included in the NUREGs. DCGLs for Pu-238, Pu-239/240, Pu-241, Am-241 and Cm-243/244 as parent radionuclides or precursors to others in the actinide series are included in the LTP.) Insufficient time has passed since the nuclear fuel first arrived at HNP for a significant quantity of the daughter radionuclides to have accumulated based on serial transformation (e.g., various isotopes of Pb and U). The dose contribution from the daughter radionuclides has been accounted for as discussed in the response to RAI #20.

17

Enclosure 1 to CY-01-084 Table 3 NUREG/LTP Radionuclides with Half-lives greater than one year (sorted by half-life)

NUREG/CR-3474 NUREG/CR-0130 CY LTP Isotope Half Life (y) Table 5.4 Table 5.6 Table 7.3-5 Table 7.3-11 Table 7.3-14 Table 6-2 Cs-1 34 2.06E+00 Na-22 2.60E+00 Fe-55 2.70E+00 Sb-125 2.77E+00 Eu-1 55 4.96E+00 Co-60 5.27E+00 Eu-154 8.81E+00 Ba-133 1.07E+01 H-3 1.24E+01 Eu-152 1.33E+01 Pu-241 1.44E+01 Pm-145 1.77E+01 Cm-243 2.85E+01 Sr-90 2.91E+01 Cs-137 3.00E+01 Hf-178m 3.10E+01 Sn-121m 5.50E+01 Pu-238 8.78E+01 Sm-151 9.01E+01 Ni-63 9.61E+01 Ag-1 08m 1.27E+02 Tb-158 1.50E+02 Am-241 4.32E+02 Ho-166m 1.20E+03 Mo-93 3.50E+03 C-14 5.73E+03 Nb-94 2.03E+04 Pu-239 2.41 E+04 Se-79 6.50E+04 Ni-59 7.51E+04 Ca-41 1.40E+05 U-233 1.59E+05 Tc-99 2.13E+05 Cl-36 3.01E+05 Zr-93 1.53E+06 Cs-135 2.30E+06 Mn-53 3.70E+06 Pb-205 1.43E+07 1-129 1.57E+07 Nb-92m 3.60E+07 Sm-146 1.03E+08 18

Enclosure 1 to CY-01-084 Although previously only identified in minute concentrations (Ag-1 08m only), the presence of Ag-1 08m and Ag-1 1 Om has been recently identified in selected components in the Spent Fuel Building. The presence of Ag-1 08m and Ag-i 1Om is readily identifiable based on their characteristic photo-peak, and there is currently no indication that these radionuclides exist in significant concentrations in waste streams beyond the selected components in which they were identified.

At the time of Final Status Survey, the material in the Spent Fuel Building, including the spent fuel pool liner will have been removed, and it is not expected that either of these isotopes of silver will remain.

1-129 is routinely part of the analysis of various waste streams. It is a specific radionuclide of concern in waste disposal facility licensing and it is identified on manifests for wastes in accordance with the guidance of 10 CFR 20, Appendix G.

However, 1-129 has not been identified above the analysis MDA in any waste stream for the last several years. Therefore, a DCGL was not established for 1-129.

Screening based on Dose Contribution:

Based on the data in the subject NUREGs, the radionuclides with half-lives greater than one year, that are postulated to exist in relative concentrations greater than 0.1% of the Co-60 concentration and not currently included in the LTP Table 6-2, are Sb-1 25, Na-22, Ba-1 33, Hf-1 78m, Sm-1 51, and Ca-41.

Ca-41 and Na-22 are postulated to exist in concentrations of as much as 5% and 3%, respectively. However, in the final evaluation of the potential impact of any radionuclide, both the relative concentration and the relative dose impact must be considered.

The radionuclides shown in Table 4 below were identified as being potential radionuclides of concern in one of the referenced NUREGs, and having half-lives greater than one year, and not currently being included in the LTP. The relative concentration for each of these nuclides was determined based on Co-60 in the bio-shield concrete.

19

Enclosure 1 to CY-01-084 Table 4 NUREG Radionuclides not included in LTP with T112 > 1 year NU I-LU/(.-3.4 NUKL-/.;-U130 Y LCI Isotope Half Life (y) I able !.4 lable 5.5 I able (.J-b I able t.3-1 1 1able 1.3-14 able 5-2 Na-22 2.60E+00 N N -7 N N N Sb-125 2.77E+00 N N N N Y N Ba-133 1.07E+01 Y Y N N N N Pm-145 1.77E+01 Y Y N N N N Hf-178m 3.10E+01 Y Y N N N N Sn-121m 5.50E+01 Y Y N N N N Sm-151 9.01 E+01 Y Y N N N N Ag-108m 1.27E+02 Y Y N N N N Tb-158 1.50E+02 Y Y N N N N Ho-166m 1.20E+03 Y Y N N N N Mo-93 3.50E+03 Y Y Y N N N Se-79 6.50E+04 Y Y N N N N Ca-41 1.40E+05 Y Y Y N N N U-233 1.59E+05 Y Y N N N N CI-36 3.01 E+05 Y Y Y N N N Zr-93 1.53E+06 Y Y N N N N Cs-135 2.30E+06 Y Y N N N N Mn-53 3.70E+06 Y Y N N N N Pb-205 1.43E+07 Y Y N N N N 1-129 1.57E+07 Y Y N N N N Nb-92m 3.60E+07 Y Y N N N N Sm-146 1.03E+08 Y Y N N N N The relative concentration for concrete was multiplied by the effective dose conversion factors to determine what fraction of the potential internal dose may be contributed from these radionuclides individually and collectively. Dose conversion factors for both inhalation and ingestion pathways were obtained from Federal Guidance Report 11. Table 5 shows that these radionuclides existing in the postulated relative concentrations collectively contribute less than 5% of the dose contributed by Co-60. Na-22 and Ca-41 each contributed less than 2% of the dose contributed by Co-60.

Concrete was selected as the basis for this determination because of the large quantity of concrete remaining at the time of the Final Status Survey and the fact that concrete would be the primary source for the activation products of interest.

These relative concentrations are representative of the activation products that may be evident in the concrete surrounding the core. They should represent the worst case for activated concrete.

20

Enclosure 1 to CY-01-084 Table 5 Internal Dose Impact Relative to Co-60 for Radionuclides with T112 > 1 year and not listed in the LTP Table 7-2 FURK11 -"-GR 11 Relative Activity' Table 2.1 Table 2.2 Inhalation Ingestion Fraction of Co-60 Dose

( 10 cm Inhalation Ingestion CEDE CEDE Inhalation Ingestion Inner Edge uCi X/ uCi Co-6OjCi X / uCi Co-6() (Sv/Bq) (Sv/Bq) (Sv/Bq Co-60) (Sv/Bq Co-60) Pathway Pathway Isotope Half-Life N/A 2.07E-09 3.10E-09 7.38E-11 1.10E-10 1.25E-03 1.52E-02 Na-22 2.6 3.56E-02 N/A 3.30E-09 7.59E-10 6.60E-12 1.52E-12 1.12E-04 2.09E-04 Sb-125 2.77 2.OOE-03 5.56E-03 1.68E-10 9.19E-10 1.06E-12 5.81E-12 1.80E-05 7.98E-04 Ba-133 10.7 6.32E-03 2.72E-05 8.23E-09 1.28E-10 2.24E-13 3.48E-15 3.79E-06 4.79E-07 Pm-145 17.7 2.68E-05 5.00E-04 6.65E-07 5.89E-09 1.03E-09 9.10E-12 1.74E-02 1.25E-03 Hf-178m 31 1.55E-03 2.95E-06 1.44E-06 3.11E-09 4.19E-10 9.19E-15 1.24E-15 1.55E-07 1.70E-07 Sn-121m 55 3.89E-03 8.10E-09 1.05E-10 5.15E-11 6.68E-13 8.72E-04 9.18E-05 Sm-151 90.1 6.36E-03 1.64E-05 8.89E-06 8.14E-09 2.06E-09 1.33E-13 3.37E-14 2.25E-06 4.63E-06 Ag-108m 127 2.67E-07 6.91E-08 1.19E-09 6.91E-14 1.19E-15 1.17E-06 1.63E-07 Tb-158 150 1.00E-06 1.50E-04 1.08E-04 2.09E-07 2.18E-09 3.14E-11 3.27E-13 5.30E-04 4.49E-05 Ho-166m 1200 1.56E-06 7.68E-09 3.64E-10 4.19E-14 1.99E-15 7.09E-07 2.73E-07 Mo-93 3500 5.45E-06 7.78E-09 2.66E-09 2.35E-09 2.78E-17 2.46E-17 4.71 E-10 3.37E-09 Se-79 65000 1.05E-08 5.28E-02 3.64E-10 3.44E-10 1.99E-11 1.88E-11 3.36E-04 2.58E-03 Ca-41 1.40E+05 5.45E-02 1.27E-05 3.33E-05 3.66E-05 7.81E-08 1.22E-09 2.60E-12 2.06E-02 3.58E-04 U-233 159000 4.17E-04 5.93E-09 8.18E-10 2.64E-12 3.64E-13 4.47E-05 5.01E-05 CI-36 3.01 E+05 4.45E-04 8.06E-08 8.67E-08 4.48E-10 1.18E-14 6.11E-17 2.OOE-07 8.39E-09 Zr-93 1530000 1.36E-07 3.89E-09 1.23E-09 1.91E-09 5.31E-18 8.25E-18 8.99E-11 1.13E-09 Cs-135 2.30E+06 4.32E-09 7.73E-08 2.06E-08 1.35E-10 2.92E-11 1.04E-17 2.26E-18 1.77E-10 3.10E-10 Mn-53 3.70E+06 7.50E-1 1 1.06E-09 4.41E-10 1.35E-18 5.61E-19 2.28E-11 7.71E-11 Pb-205 1.43E+07 1.27E-09 6.11E-11 4.69E-08 7.46E-08 2.98E-18 4.75E-18 5.05E-11 6.52E-10 1-129 1.57E+07 6.36E-11 3.33E-13 n/a n/a n/a n/a n/a n/a Nb-92m 3.60E+07 6.36E-12 4.72E-13 2.23E-05 5.51 E-08 3.95E-17 9.77E-20 6.69E-10 1.34E-1 I Sm-146 1.03E+08 1.77E-12 Co-60 1.OO1+00 1.UF+O0 5,91[-O8 3-WB 7,BE0-09 5.9+ 7.251:-09 Reference all long-lived radionuclides not Included in the LTP: 4.12E-02 2.06E-02 Total Committed Effective Dose Fraction of 7.3-5 of NUREG CR-0130 Activity is relative to 1 uCi Co60, the concentration for Sb-125 was taken from Table 7.3-14, Na-22 from Table 1

All other relative activities are taken from NUREG/CR-3474 Table 5.4 Table 2.2-Ingestion Federal Guidance Report 11-Committed Effective Dose Equivalent Conversion Factors Table 2.1-Inhalation, Based on Table 5.8 of NUREG/CR-3474, some radionuclides may exist in higher concentrations in the imbedded rebar than in the concrete. Mn-53, Mn-54, Fe-55, Ni-59, Co-60, Ni-63, Zn-65, Se-79, Mo-93 and Pb-205 are each identified as potentially existing in concentrations of 2-25 times higher in rebar near the core axial mid-plane, than in concrete. Five of these ten radionuclides are listed in Table 6-2 of the HNP LTP. Zn-65 was eliminated based on short half-life, and the potential dose impact of Mn-53, Se-79, Mo-93 and Pb-205 can be inferred from Table 5 as negligible even at these higher concentrations.

In accordance with guidance in NUREG-1727, radionuclides contributing less than 10 percent of the dose limit can be screened out. The radionuclides in Table 5 collectively contribute less than 5% of the Co-60 dose and thus can be screened out.

In order to evaluate the significance of the radionuclides listed in Table 5 to external exposure potential, a similar analysis for external exposure was performed using dose conversion factors from Federal Guidance Report 12 for soil contaminated to a depth of 15 cm. This analysis is provided in Table 6 and indicates that if these nuclides were present, the collective exposure would also be less than 5% as compared to Co-60.

Table 7 provides a summary of the disposition of the radionuclides addressed in the above paragraphs.

21

Enclosure 1 to CY-01-084 Table 6 External Internal Dose Impact Relative to Co-60 for Radionuclides with T112 > 1 year and not listed in the LTP Table 7-2 FGR 12 Relative Activity' Table 111.6 Fraction of Co-60 External Dose InnerEdge @10cm External CEDE External Isotope Half-Life, (uCi X I uCi (uCi X / uCi Sv-s (Sv-s-m3/Bq Pathway Co-60) Co-60) m3/Bq Co-60) yrs Na-22 2.6 3.56E-02 N/A 6.31 E-17 2.25E-1 8 3.1 OE-02 Sb-1 25 2.77 2.00E-03 N/A 1.18E-17 2.36E-20 3.26E-04 Ba-1 33 10.7 6.32E-03 5.56E-03 9.88E-1 8 6.24E-20 8.61 E-04 Nb-93m 13.6 4.14E-05 N/A 5.57E-22 2.31 E-26 3.18E-10 Pm-145 17.7 2.68E-05 2.72E-05 1.57E-19 4.27E-24 5.90E-08 Hf-1 78m 31 1.55E-03 5.OOE-04 6.42E-17 9.92E-20 1.37E-03 Sn-121m 55 2.95E-06 1.44E-06 1.05E-20 3.1 OE-26 4.28E-1 0 Sm-151 90.1 6.36E-03 3.89E-03 5.27E-24 3.35E-26 4.63E-10 Ag-108m 127 1.64E-05 8.89E-06 4.61 E-1 7 7.54E-22 1.04E-05 Tb-1 58 150 1.OOE-06 2.67E-07 2.19E-17 2.19E-23 3.02E-07 Ho-166m 1200 1.50E-04 1.08E-04 4.90E-1 7 7.35E-21 1.01 E-04 Mo-93 3500 5.45E-06 1.56E-06 3.16E-21 1.72E-26 2.38E-10 Se-79 65000 1.05E-08 7.78E-09 1.01E-17 1.06E-25 1.46E-09 Ca-41 1.40E+05 5.45E-02 5.28E-02 0.OOE+00 0.OOE+00 0.OOE+00 U-233 159000 1.27E-05 3.33E-05 7.24E-21 2.41 E-25 3.33E-09 CI-36 3.01 E+05 4.45E-04 4.17E-04 1.22E-20 5.43E-24 7.50E-08 Zr-93 1530000 1.36E-07 8.06E-08 0.OOE+00 0.OOE+00 0.OOE+00 Cs-135 2.30E+06 4.32E-09 3.89E-09 2.05E-22 8.85E-31 1.22E-14 Mn-53 3.70E+06 7.73E-08 2.06E-08 0.OOE+00 0.OOE+00 0.OOE+00 Pb-205 1.43E+07 1.27E-09 7.50E-11 3.78E-23 4.81 E-32 6.64E-16 1-129 1.57E+07 6.36E-1 1 6.11E-11 6.93E-20 4.41 E-30 6.08E-14 Nb-92m 3.60E+07 6.36E-12 3.33E-1 3 n/a n/a n/a Sm-146 1.03E+08 1.77E-12 4.72 E-13 0.OOE+00 0.OOE+00 O.OOE+00 Reference Co-60 I.OOE+00 I.OOE+00 7.25E-17 7.25E-1 7 Total Committed Effective Dose Fraction of all long-lived radionuclides not included in the LTP: 3.37E-02 Activity is relative to 1 uCi Co60, the concentration for Sb-i 25 was taken from Table 7.3-14, Na-22 from Table 7.3-5 of NUREG CR-0130 All other relative activities are taken from NUREG/CR-3474 Table 5.4 Federal Guidance Report 11-Committed Effective Dose Equivalent Conversion Factors Table 2.1-Inhalation, Table 2.2-Ingestion 22

Enclosure 1 to CY-01-084 Table 7 Summary Disposition of Radionuclides in NUREGs Versus LTP Isotope In LTP? Reason for exclusion from LTP Table 7-2 H-3 Y 5 C-14 Y 5 Na-22 N 3 P-33 N 1 S-35 N 1 CI-36 N 2 Ca-41 N 3 Ca-45 N 1 Sc-46 N I Cr-51 N 1 Mn-53 N 2 Mn-54 Y 5 Fe-55 Y 5 Fe-59 N 1 Ni-59 Y 5 Co-57 N 1 Co-58 N 1 Co-60 Y 5 Ni-63 Y 5 Zn-65 N 1 Se-79 N 2 Sr-89 N 1 Sr-90 Y 5 Y-90 N 1 Nb-92m N 3 Nb-95m N 1 Zr-93 N 2 Mo-93 N 2 Nb-94 Y 5 Zr-95 N 1 Tc-99 Y 5 Ru-103 N 1 Ag-108m N 2 Ag-110m N 1 Sn-121m N 2 Sb-124 N 1 Sb-125 N 3 Te-129m N 1 1-129 N 2 1-131 N 1 1-133 N I Ba-133 N 3 Cs-134 Y 5 Cs-135 N 2 Cs-136 N 1 Cs-137 Y 5 23

Enclosure 1 to CY-O1-084 Isotope In LTP? Reason for exclusion from LTP Table 7-2 Ba-140 N 1 La-140 N 1 Ce-141 N 1 Ce-143 N 1 Pm-145 N 2 Sm-146 N 2 Sm-151 N 3 Eu-152 Y 5 Eu-154 Y 5 Eu-155 Y 5 Tb-158 N 2 Ho-1 66m N 2 Hf-1 78m N 3 Pb-205 N 4 U-233 N 4 Pu-238 Y 5 Pu-239 Y 5 Pu-241 Y 5 Am-241 Y 5 Cm-243 Y 5 1 T 1 /2 <1year 2 Relative concentration < 0.1% Co-60, relative dose contribution < 5% Co-60 3 Relative dose contribution < 5% Co-60 4 Insufficient accumulation based upon serial transformation 5 Included in LTP list of radionuclides for HNP.

DCGLs are listed in the LTP for Pu-239 and Cm-243. DCGLs were not listed for Pu-240 and Cm-244. A single DCGL was established for each pair (Pu-239/240 and Cm-2431244), primarily because laboratory radiochemical analyses do not report concentrations of these radionuclides separately. Typically, the concentration results for each pair is reported as a single value. Establishing a single DCGL for each of these radionuclide pairs precludes the need to estimate each separately, and allows direct comparison of radiochemical analysis results for a radionuclide pair to a single DCGL established for that pair. Basing the DCGL for Pu-239/240 on Pu-239, and for Cm-243/244 on Cm-243 is conservative. Based on Federal Guidance Report 11, dose conversion factors associated with Pu-239 and Pu-240 are essentially the same and the dose conversion factors associated with Cm-243 are somewhat more limiting than those associated with Cm-244.

Therefore, based on this additional review the only recommended change to the LTP Table 6-2 is to indicate that Pu-239 will be assessed based on Pu-239/240 sample results and Cm-243 will be assessed based on Cm-243/244 sample results. This change will be reflected in a future revision to the LTP.

24

Enclosure 1 to CY-01-084

8. Section 6.4.4. Base Case DCGLs, RESRAD Analysis, pages 6-8 to 6 This section discusses the RESRAD analysis and lists input parametersused in the analysis. To permit review of this analysis, provide the following:

a. Table 6-3 lists area and thickness of the contaminatedzone and states that these are conservative assumptions based on characterizationdata. The comprehensive site characterizationis not complete at this time; therefore, verification is needed to support the assumptions that the contaminated 2

area and thickness would not exceed 10,000 m and I m, respectively.

The LTP (Section 6.4, page 6-4) acknowledges that soil DCGLs are based on preliminary site characterization data. The LTP also indicates that site characterization will continue as part of the decommissioning process, and that if future site characterization data indicate contaminant characteristics are (non-conservatively) different from those stated in the LTP, then soil DCGLs will be revised as appropriate.

Data available at the time of LTP preparation included the results of site characterization activities completed in 1998 and 1999, where over one hundred subsurface soil samples, in some cases down to six feet in depth, were collected in Survey Areas 9302, 9304, 9306 and 9308 in support of plant modifications and site characterization activities. During the same time period, over two hundred soil samples, in some cases down to six feet in depth, were collected inside the RCA in Survey Areas 9307, 9310, 9312 and 9227. As provided in the response to RAI #13 that follows, generally samples taken during characterization activities in 1998 and 1999 have shown that the soil contamination is usually limited to the top 0.3 to 0.6 meters. A summary of data resulting from these site characterization activities will be provided in a supplement to the initial Site Characterization Report. The supplement will be made available to the NRC to allow verification of the assumptions that the contaminated area and thickness would not exceed 10,000 m 2 and 1 m, respectively.

Note that a contaminated zone area of 10,000 m is roughly the same amount of 2

area as is encompassed by the Industrial Area. According to Reference 2, an area of 10,000 m 2 is considered the minimum areal requirement to supply 50%

meat and milk diet. A value of 100% meat and milk diet obtained directly from the site was assumed, providing additional conservatism. Additionally, quantitative sensitivity analysis indicated that measurable sensitivity was observed for only 2 radionuclides (C-14 and Pu-238). Because preliminary site characterization data do not show contamination to be present on a contiguous basis over the entire Industrial Area, an assumed contaminated zone area of 10,000 m2 is believed to be a conservative assumption. Thus no changes to the LTP are considered necessary in response to this RAI.

b. Tables 6-3, 6-4, and 6-5 list soil density, total porosity, effective porosity, hydraulic conductivity, and soil-specific b parametervalues for the contaminated,unsaturated,and saturatedzones. According to the tables, the values used representsilty sand lithology and are taken from the data collection handbook. Explain why soil density is for sand, total porosityand effective porosity values are the upper bound values for silt stone, and 25

Enclosure 1 to CY-01-084 hydraulic conductivity and soil-specific b parametervalue is for loamy sand.

Explain why site-specific measurements were not made for these parameters.

As discussed in the Ground Water MonitoringReport (Reference 5), the overburden lithology within the Industrial Area can be generally described as a silty (loamy) sand. The conceptual site model uses this description to define the hydrologic characteristics of the contaminated, unsaturated, and saturated zones. Values for soil density, total porosity, effective porosity, hydraulic conductivity, and soil-specific exponential parameter were based on this general lithologic description. Table 1 summarizes the parameter, value used, and correct reference to the source. Please note that the LTP cites the incorrect reference for the values of hydraulic conductivity, total porosity, the soil-specific exponential parameter. The LTP will be revised to incorporate this correction.

Table 1 Hydrologic Properties Summary Parameter Value Source Notes Soil density 1.52 Reference 6, Table 2.1, "Sand" Hydraulic 4930 Reference 2, Table E.2, conductivity "Loamy Sand" Total porosity 0.41 Reference 2, Table E.2, Saturated water content "Loamy Sand" assumed equal to total porosity for conservatism (see RAI#6 response above)

Effective 0.33 Reference 6, Table 3.2, porosity "Sand (fine)"

Soil-specific 4.38 Reference 2, Table E.2, exponential "Loamy Sand" parameter Site-specific measurements were not made for the hydrologic parameters listed in Table 1 above based on dose modeling results and sensitivity analysis. Dose modeling results summarized in Table 6-8 of the LTP shows that water dependent pathways are dominant only for 3 of the 20 radionuclides (C-1 4, Pu-238, Pu-239) that are potentially present in site soils. Results of a deterministic sensitivity analysis, summarized in Tables 6-9 and 6-10 of the LTP, show that the sensitivity of the soil DCGLs to these hydrologic parameters is limited. Therefore, site-specific measurements for these parameters is not warranted.

c. Table 6-7 lists the mass loading for inhalationparametervalue from NUREG/CR-5512, Volume 2. The methods used to compute the inhalation pathway doses by the two codes are completely different. Justify why the value from NUREG/CR-5512 was used for RESRAD.

26

Enclosure 1 to CY-01-084 A value of 3.14x10 6 g/m 3 was adopted for mass loading for inhalation, based on recommendations provided in PreliminaryGuidelines for Dose Assessment (Reference 7). The stated purpose of this guideline is to provide a consistent approach for staff to evaluate dose assessments conducted to demonstrate compliance with the current license termination rule. Guidance is provided for both DandD and RESRAD analyses. In discussing analyses using RESRAD, the following guidance is provided:

"To ensure consistency between the critical group used in DandD analyses, analysts are encouraged to use the parameters listed in Table 2."

3 The Table 2 value for mass loading for inhalation value is 3.14xl0e g/m . This value was therefore adopted and used as input to RESRAD (v. 5.91) for the soil DCGL determination.

More recent RESRAD-specific guidance is included in ParameterDistributions for Use in RESRAD and RESRAD-BUILD Computer Codes (Reference 8). This document provides the basis for the RESRAD default probabilistic parameter set described in Appendix C of NUREG-1727. Argonne National Laboratory and NRC staff jointly developed a default parameter input set that may be used to perform probabilistic dose assessments using the RESRAD code.

Reference 8 includes a cumulative distribution function for mass loading for inhalation that is applicable to the RESRAD code. The mass loading input to RESRAD provides the time-averaged respirable concentration of contaminated soil and dust. The respirable portion of resuspended material can be represented by the PM-1 0 fraction of airborne particulate matter (particles < 10 pjm in diameter). Five years of annual average ambient PM-1 0 air concentration data for approximately 1,790 air monitoring stations across the United States and its territories were used to develop a cumulative distribution function. Based on the wide distribution of monitoring locations, these data are representative for the residential (versus commercial) farming scenario that is used in the LTP to develop soil DCGLs.

To assess the appropriateness of the soil DCGLs included in Table 6-8 of the LTP in light of more recent NRC guidance (Reference 9), RESRAD was used to recalculate soil DCGLs using a mass loading for inhalation derived from 3

Reference 8. The 90% quantile value (3.5xl 0-5 g/m ) was selected from Table 4.6-1 of Reference 8 to conservatively represent this parameter. Table 1 summarizes the results, which includes a comparison to the DCGLs included in the LTP.

Note that during the preparation of the response to this RAI, a minor error in the average wind speed input to the soil DCGLs was discovered and corrected. Use of the corrected wind speed value did not significantly change the soil DCGLs from those previously provided in LTP Table 6-8. These corrected soil DCGL values are provided as Column 3 of Table 1. LTP Table 6-8 will be revised to reflected the correct soil DCGLs.

27

Enclosure 1 to CY-01-084 Table 1 Mass Loading for Inhalation DCGL Comparison Radionuclide Recalculated LTP DCGL Recalculated DCGL (pCi/g) DCGL/

(pCi/g) LTP DCGL H-3 9.97E+02 9.97E+02 100.00%

C-14 4.26E+00 4.26E+00 100.00%

Mn-54 1.52E+01 1.52E+01 100.00%

Fe-55 5.35E+04 5.35E+04 100.00%

Co-60 3.49E+00 3.49E+00 100.00%

Ni-59 4.91E+03 4.91E+03 100.00%

Ni-63 1.79E+03 1.79E+03 100.00%

Sr-90 3.1OE+00 3.1OE+00 100.00%

Nb-94 7.1OE+00 7.1OE+00 100.00%

Tc-99 2.37E+01 2.37E+01 100.00%

Cs-1 34 5.74E+00 5.74E+00 100.00%

Cs-1 37 1.16E+01 1.16E+01 100.00%

Eu-1 52 8.24E+00 8.24E+00 100.00%

Eu-154 7.61E+00 7.61E+00 100.00%

Eu-155 3.21E+02 3.21E+02 100.00%

Pu-238 4.86E+00 4.86E+00 100.00%

P, 9,_Q ?1=E+nn 2.30E+00 100.00%

Pu-241 1.67E+03 98.50%

Am-241 4.42E+01 97.85%

Cm-243 3.98E+01 98.74%

As is evident from Table 1 above, there is no change in the soil DCGLs for 17 of 20 radionuclides when recalculated using a the mass loading for inhalation of 3.5x105 g/m 3. The recalculated DCGLs for Pu-241, Am-241, and Cm-243 are approximately 2% lower than the corresponding values included in the LTP.

Because these changes are small, because conservative assumptions have been used in the dose modeling, and because Pu-241, Am-241, and Cm-243 are expected to represent small fractions of the radionuclide mix, revising the soil DCGLs in the LTP is not warranted.

9. Section 6.4.4. Base Case DCGLs. RESRAD Analysis, Water, page 6-9

a. Table 6-5 summarizes the assignment of the water group parameters.

Hydrologicalmodeling was performed using the HELP code to determine site-specific values for the runoff coefficient and the evapotranspiration coefficient These coefficients were then used as input values for RESRAD.

To permit review of these parameters,please provide the following: (1)Text explaining the HELP code methodology and its appropriatenessin this case; and (2) electronic and hard copies of input and output of the HELP code used in this analysis.

28

Enclosure 1 to CY-01-084 The Hydrological Evaluation of Landfill Performance (HELP) model (Reference

10) is a quasi two-dimensional hydrological model used for conducting water balance analysis for both natural soil layers as well as landfills and other solid waste containment facilities. This model accepts weather and soil data that account for the effects of surface storage, snowmelt, runoff, infiltration, evapotranspiration, vegetative growth, soil moisture content, lateral subsurface drainage, and unsaturated vertical drainage.

The HELP model was used to simulate the water balance in the site soil only.

Parameters associated with the Landfill aspect of the model (i.e. leachate collection, liner leakage, etc) were not incorporated into the simulation. The surface hydrology simulation uses aspects of several other standard hydrologic models (Reference 10).

Specific local weather input for the model was obtained from various towns (Bridgeport and Windsor Locks, CT) located in general proximity to the site, where available (Note: the weather data is contained in a database which is part of the HELP model). This data included evapotranspiration data (Bridgeport, CT), Precipitation Data (Windsor Locks, CT), and temperature and solar radiation data (Bridgeport, CT). Site specific data included parameters such as soil type (silty sand), site area (15 acres), site slope (approximately 0%), and slope length (360-ft).

The code is considered appropriate due to the amount of both local and site specific data that was used as inputs.

The input data files and program output files have been included on the enclosed CD-ROM. Also included is the original HELP v. 3 file. Attachment 2 contains details regarding the file structure on the CD-ROM and instructions on how to install the HELP code.

No changes to the LTP are considered necessary in response to this RAI.

b. If the wateryield from the shallow aquifer cannot sustain the needs of farming and household use, the farmermay elect to locate his well in the deep aquifer in the bedrock region. Therefore, the possibility of getting groundwaterfrom the deep aquifercannot be excluded. In order to examine all the likely situations, the LTP should includejustification to exclude the possible location of a well in the deep aquifer (i.e., the bedrock).

As stated in the LTP, it was assumed that a water supply well completed in the shallow aquifer would more likely be the site water source than the deep bedrock aquifer. Currently, the water supply wells for the plant are located in the unconsolidated sediments near the river (Reference 11). These wells produce a yield of approximately 300 gpm.

Bedrock at the industrial area ranges in depth from approximately 10-ft below ground surface at the hillside, or northern border, to over 80-ft below ground surface at the Connecticut River (Reference 5). According to Reference 11:

29

Enclosure 1 to CY-01-084 "All of the metamorphic rocks described above are generally considered relatively poor aquifers. The transmitting capacity of these rocks is through fractures in the bedrock and not between individual grains, as would be the case in unconsolidated materials. The yield of 28 wells completed in open boreholes in the bedrock, located within a two-mile radius of the plant, ranges from less than 1 gpm to 30 gpm. The average reported yield of these wells is 11 gpm." A well producing about 300 gpm would be preferable for use by a farmer, to a well producing only 11 gpm.

In addition, the maximum ground water radioactivity has generally been observed in the shallow aquifer (Reference 5). (Note: The highest ground water radioactivity initially observed was in the shallow aquifer. This concentration, along with generally all other ground water radioactivity, in both shallow and deep aquifers, have decreased over time - however at different rates. The result is that at any given time higher ground water radioactivity may be observed in a deep bedrock well than in the shallow aquifer. It should be noted that the highest observed ground water radioactivity, in all wells, will be considered at the time of final status survey). Also, because of its closer proximity to the probable source of contamination, it is likely that a higher groundwater radiation dose would be obtained from a shallow water supply well, thus making this scenario the bounding condition.

Sensitivity analysis for the radionuclides shown to dose predominately from water-dependent pathways (C-14 and Pu-238), as summarized in Table 6-10 of the LTP, indicate the following:

" DCGLs for C-14 and Pu-238 had limited sensitivity to saturated zone physical properties (density, total porosity, effective porosity, conductivity, and hydraulic gradient), which would likely be different for the bedrock aquifer.

" The DCGL for Pu-238 had high sensitivity (sensitivity coefficient of 0.8) to well pump intake depth. This value would likely be greater than the value of 10 m used if a bedrock aquifer were considered. However, the 0.8 value indicates direct proportionality, that is, the higher value (or in this case, the deeper the well), the higher the DCGL.

This information concerning the basis for assumption of a shallow well will be added to Section 6.4.4 of the LTP. This will be reflected in a future change to the LTP.

10. Section 6.4.4. Base Case DCGLs, Sensitivity Analysis, pages 6-23 to 6 This section discusses sensitivity analysesperformed to evaluate impacts of the input parameterson the DCGLs for soil. Sensitivity of an input parameteris affected by several factors, including the perturbationrange used for the parameter. To permit evaluation of the conclusions about the importantparameters,please provide the following:

a. Text and references supporting the selection of perturbationranges. The potential range of distribution for certain parameters(such as Kd and hydraulic conductivity) can span several orders of magnitude.

30

Enclosure I to CY-01-084 According to Appendix C of NUREG-1727, "NMSS Decommissioning Standard Review Plan" (Reference 8a):

"..deterministic sensitivity analysis, calculates the change in the output result (i.e.

peak dose) with respect to a small change in the independent variables, one at a time" A normalized sensitivity coefficient, as used in the LTP is further defined as follows:

Sk-d dPk /P datk / atk where dP/dak is the marginal sensitivity of P to ak. Sk describes the percentage change of performance measure P to a 1% change of parameter ak. The derivative that defines the marginal sensitivity can calculate to second order accuracy using a central difference scheme, allowing Sk to be represented as Sk (P, - -,)A/aPo 2Aak akO The calculation of each individual Sk requires two model runs: one with ak = ak+1 to determine P+I, and one with ak = ak-I to determine P-1. This equation is essentially identical in form to the sensitivity equation given in Reference 9.

The above equation was used to generate normalized sensitivity coefficients, as discussed in Section 6.4.4 of the LTP. This equation was solved using a central finite difference scheme to determine the sensitivity about a point, or in this case the input parameter. Preliminary iterations were used for the range in perturbation to ensure that the sensitivity coefficient was approximate less than or equal to 1, and to ensure that a measurable response was observable. Table 1 summarizes the perturbation ranges for each input parameter examined.

31

Enclosure 1 to CY-01-084 Table 1 Sensitivity Input Parameter Perturbation Summary Parameter (units) LTP Value Perturbation Range +/

Contamination Contaminated zone thickness (m) 1 0.15 Contaminated zone area (m2) 10000 2500 Contaminated zone density (g/cm3) 1.52 0.52 Soil Unsaturated zone thicknessa (m) 2 0.2 Unsaturated zone total porositya (unitless) 0.41 0.03 Water Runoff coefficient (unitless) 0.11 0.06 Precipitation (m/yr) 1.1 0.1 Irrigation (m/yr) 0.2 0.1 Evapotranspiration coefficient (unitless) 0.50 0.05 Watershed areaa (m') 593,000 100,000 Saturated zone densitya (g/cm') 1.52 0.2 Saturated zone effective porositya (unitless) 0.33 0.03 Saturated zone total porositya (unitless) 0.41 0.05 Saturated zone hyd. Conductivitya (m/year) 4390 2000 Saturated zone hydraulic gradienta 0.017 0.002 (unitless)

Well pump intake depth (m) 10 5 Ingestion Fruit, vegetable, grain consumption (kg/yr) 112 50 Leafy vegetable consumption (kg/yr) 21.4 10 Meat and poultry consumption (kg/yr) 65.1 50 Milk consumption (l/yr) 233 100 Fish consumptiona (kg/yr) 20.6 2 Drinking water consumption (I/year) 478.5 200 Soil ingestion (g/yr) 18.26 10 Livestock fodder intake for meat (kg/d) 27.1 10 Livestock fodder intake for milk (kg/d) 63.25 40 Livestock soil ingestion (kg/d) 0.50 0.25 Occupancy External gamma shielding (unitless) 0.5512 0.3 Fraction time spent indoors onsite 0.6571 0.3 (unitless)

Fraction time spent outdoors onsite 0.1101 0.1 (unitless) a for C-14 and Pu-238 only 32

Enclosure ] to CY-O1-084 The perturbation ranges for the Kd sensitivity coefficient analysis were unique for each radionuclide analyzed, due to the large span of the Kd values. Table 2 summarizes the perturbation ranges.

Table 2 Sensitivity Soil Kd Perturbation Summary Radionuclide Value Perturbation (cm3 /g) Range +/

(cm3 lg)

C-14 4.34 4 Fe-55 535 500 Co-60 1510 1000 Ni-63 37 37 Sr-90 31.4 20 Tc-99 7.37 7.37 Cs-137 10.5 10.5 Eu-152 -

Pu-238 13.6 5 Am-241 1430 1000 Note: RESRAD treats Eu-1 52 as an activation product and does not allow changes in Kd.

No change to the LTP is considered necessary in response to this RAI.

b. Text explaining why a sensitivity value of 0.5 was selected to identify key parameters. This sensitivity value may have eliminated some sensitive parameterssuch as plantand meat transferfactors. Following the guidance in NUREG-1 727, the review should focus on the uncertaintyresulting from the physical parametervalues used in the analysis.

According to Section 6.4.4 of the LTP, a preliminary sensitivity analysis, using the RESRAD sensitivity graphical utility, was performed on 10 radionuclides selected based on their relative abundance at the site. The parameters selected for preliminary analysis were based on the dose fraction from each pathway (water independent and water dependent), as described in Table 6-10 of the LTP.

A quantitative sensitivity analysis was then performed on select parameters based on the preliminary results, as described in the response to RAI #10(a) above, to determine a normalized sensitivity coefficient, as described Reference 7. A value of 0.50 (and greater) was selected as the sensitivity coefficient value for which additional justification would be necessary, as summarized in Table 6-10 of the LTP. This value represents a 0.5% change in DCGL for every 1% change in the parameter analyzed.

33

Enclosure 1 to CY-01-084 Based on the 0.5 criteria, additional justification was performed on select parameters. However, if the criteria of any measurable, non-zero value for the sensitivity coefficient is applied, the following points are observable:

"* Any sensitivity coefficients relating to behavioral and metabolic parameters (i.e. ingestion, occupancy, etc) do not require any additional justification, as inherently stated in Reference 7 and Section 6.4.4 of the LTP.

" Sensitivity coefficients for parameters under the subheading of "Soil" and "Water" in Table 6-10 are generally based on site specific data (see RAI #8(c) regarding physical parameters and RAI #9(b) regarding well intake depth),

rather than assumptions, and therefore further justification is not required.

"* Sensitivity coefficients for parameters under the subheading of

""Contamination" in Table 6-10 are based on site specific and literature values (see RAI#7 and RAI#6 above) and further justification is not required.

No input parameters were eliminated from the analysis - the analysis was based in part on relative radionuclide abundance at the site and percent dose from water dependent/water independent pathways.

No change to the LTP is considered necessary in response to this RAI.

11. Section 6.4.6.1. Affected Survey Areas, page 6 Table 6-11 lists survey areas affected by existing radioactivityin ground water. Survey areas with underlying ground water contamination with H-3 or Cs-137 were defined as affected areas.

This designationincluded survey areas within 100 m of the existing contaminated plume. Please providejustification, including analysis of the hydraulic conductivity of the waterbearing units, for the assumption that the survey areas at distances greaterthan 100 m from the existing plume, especially down gradientareas,would not be impacted in the future (i.e., the ground waterplume would not expand to the regions that are not impacted at the presenttime).

As presented in Reference 5, the general direction of ground water flow is in an approximately southern direction, towards the Connecticut River. As shown in Figure 6-5 of the LTP, the affected areas encompass the entire plume area, downgradient to the river, and a 100-meter buffer zone on each lateral perimeter of the plume. Existing ground water monitoring data has shown that the plume is not expanding laterally.

However, as previously indicated in response to NRC RAI #24 dated June 14, 2001, additional plans for groundwater characterization and monitoring work have been provided in the "Phase 2 Hydrogeologic Investigation Work Plan" to support the existing data. No changes to the LTP are considered necessary in response to this RAI.

12. Section 6.4.6.2. Dose Modeling Approach, Step 1. page 6 To permit review of the code used to determine ground water dose contribution,please insert the RESRAD code version used in calculatingthe water dependent dose contributionsfor H-3 and Cs-137.

34

Enclosure 1 to CY-01-084 RESRAD v. 5.91 was used in calculating water dependent dose contributions. This will be reflected in a future change to the LTP.

13. Section 6.4.6.2. Dose Modeling Approach, pages 6-30 to 6 This section provides steps involved in the approach used to calculatethe special case DCGLs. The first step listed is to determine the ground water dose contribution from H-3 and Cs-137 on the basis of ground water concentrationcurrently observed. However, in assessing the potential radiationexposure for the ground water source, it is importantto: (1) use the maximum concentrationswithin the assessmenttime frame (1, 000 years into the future) (i.e., the releasingpotential of the deeper soil must be considered);and (2) include all the radionuclidesthat exist in the deeper soil. To facilitatereview of the approach,please provide the following:

a. Justificationthat the currently observed ground water concentrationsused in the LTP for Cs-137 and H-3 are the maximum values (i.e., in the future, ground water concentration would not increase).

b. Justificationthat the monitoring well from which the ground water concentrationsof H-3 and Cs-137 were obtainedis properly located,so that the maximum concentrationsin the ground water plume can be detected.

Current ground water radioactivity trends, based on existing monitoring well data, indicate a general downward trend in H-3 and Cs-137 concentrations, where present (see the following table for data). As previously indicated in response to NRC RAI #24 dated June 14, 2001, additional groundwater characterization and monitoring work has been proposed in the "Phase 2 Hydrogeologic Investigation Work Plan" to support the existing data. Specifically, additional monitoring wells will be installed and sampled to ensure that the ground water plume has both been bounded spatially, and that the maximum ground water radioactivity has been or will be detected.

It should also be noted that as stated in Section 6.4.6.3 of the LTP, the special case soil DCGLs considering ground water radioactivity will be re-evaluated based on the most recent groundwater radiological data at the time of the Final Status Survey.

35

Enclosure 1 to CY-01-084 Groundwater Radiological Analysis Data Well [ Round1 Round 2 Round 3 Round 4 Number T pCiLL pCiVL pC &iL pCUiL Tritium Analysis Data 0ooS ND ND Not Sampled Not Sampled 1OOD ND ND Not Sampled <MDC loiS ND ND Not Sampled <MDC 101D ND ND Not Sampled Not Analyzed for H-3 102D 2740 3160 2640 2470 102S ND ND Not Sampled 5540 103D 22180 17550 19660 20900 103S 2580 9260 2980 1230 104S ND ND Not Sampled Not Sampled 105D 4590 2450 3030 2150 105S 138700 67400 23480 15900 106D 3320 1590 5830 1810 106S 24290 16370 Not Sampled Not Sampled 107S ND ND Not Sampled <MDC 107D ND ND Not Sampled <MDC 108S ND ND Not Sampled Not Sampled 109D 33070 31600 21230 15800 109S ND ND Not Sampled <MDC 110D 27630 23280 27230 18300 110S 3090 ND 2470 2360 111s ND ND Not Sampled <MDC 112S ND ND Not Sampled Not Sampled 113S ND ND Not Sampled Not Sampled 114S ND 1180 2850 2760 115S ND ND Not Sampled 5550 117S ND ND Not Sampled Not Sampled Mat Sump 2630 2320 Not Sampled 2890 AST1 ND ND Not Sampled Not Analyzed Gamma Spec. Analysis Date 103S* 76(Cs-137)** I 33(Cs-137)** I 29(Cs-137)** 72 (Cs-137)**

• All other well samples <MDC for Plant Related Gamma Emitting Isotopes

• • All other Plant Related Gamma Emitting isotopes <MDC

c. Justification that no radionuclidesother than H-3 and Cs-137 occur in deepersoil (deeperthan I m as assumed in the base case).

A review of subsurface samples taken to date indicates that, in addition to H-3 and Cs-137, Co-60 has been identified, although infrequently, in soils at depths greater than one (1) meter. A summary of the data reviewed is provided in the following table. If contamination below one meter exists after remediation, either the current 36

Enclosure 1 to CY-01-084 assumptions will be confirmed to be bounding or the DCGLs will be recalculated using a revised value for the contamination depth.

Subsurface Soil Radionuclide Data Survey Area Location Sample ID Depth Concentration (pCi/g) Reference Document Co-60 Cs-1 37 2228 RCA CS501 36" - 48" 0.46 4.6 Reference 12 9302 RCA CS701 36" - 48" <0.10* <0.09*

CS702 36" - 48" <0.05* <0.06*

CS703 36" - 48" <0.06* <0.06*

CS705 36" - 48" <0.06* <0.06*

9304 PA CS702 36" - 48" <0.05* <0.05* Reference 12 CS703 36" - 48" <0.08* <0.06*

CS705 36" - 48" <0.06* <0.07*

CS706 36" - 48" <0.05* <0.06*

9307 RCA CS701 36" - 48" <0.09* <0.08* Reference 12 CS702 36" - 48" <0.05* <0.04*

CS703 36" - 48" <0.10* <0.08*

CS706 36" - 48" <0.08* <0.05*

9308 PA CS706 36" - 48" <0.08* <0.06* Reference 12 CS806 48" - 60" <0.09* <0.09*

CS906 60" - 72" <0.09* <0.08*

CS707 36" - 48" <0.09* <0.09*

CS708 36" - 48" <0.13* <0.11*

CS709 36" - 48" <0.10* 0.04 CS71 0 36" - 48" <0.09* <0.08*

CS711 36" - 48" <0.05* 0.03 CS712 36" - 48" <0.10* <0.11*

CS713 36" - 48" <0.06* <0.06*

CS813 48" - 60" <0.10* <0.09*

CS913 60"- 72" <0.10* <0.07*

CS714 36" - 48" <0.09* <0.07*

CS814 48" - 60" <0.06* <0.08*

CS914 60" - 72" <0.05* <0.04*

CS715 36" - 48" <0.08* <0.08*

CS815 48" - 60" <0.05* <0.06*

CS716 36" - 48" <0.06* 0.04 CS816 48" - 60" <0.07* <0.06*

9310 RCA CS518 36" - 48" 0.25 Reference 13 CS537 36" -48" 0.12 CS549 36" - 48" 0.10 CS456 36" - 48" CS461 36" - 48" 0.22 CS490 36" - 48" 0.08 CS425 36" - 48" CS442 36" - 48" 0.38 9312 RCA CS702 36" - 48" <0.08* <0.07* Reference 12 CS705 36" - 48" <0.11* 0.12 37

Enclosure I to CY-01-084 Survey Area Location Sample ID Depth Concentration (pCilg) Reference Document Co-60 Cs-137 9312 (cont.) CS709 36" - 48" <0.07* <0.07*

CS710 36" - 48" <0.05* 0.16 9520 PA CS501 36" - 48" <0.06* 0.72 Reference 12 CS502 36" - 48" <0.09* <0.08*

CS504 36"- 48" <0.13* <0.08*

CS505 36" - 48" <0.06* <0.05*

CS506 36" - 48" <0.11* <0.08*

CS507 36" - 48" <0.09* <0.09*

CS508 36" - 48" <0.09* <0.06*

CS509 36" - 48" <0.08* <0.09*

CS510 36"--48" <0.10* 6.3 CS610 48" - 60" <0.06* 0.39 CS710 60" - 72" <0.05* 0.21 CS71 0 72" - 84" <0.08* <0.09*

CS51 1 36" - 48" <0.09* <0.07*

CS512 36" - 48" <0.07* <0.08*

Sample concentration less than MDC value indicated.

14. Section 6.4.6. Special Case DCGLs, page 6 Table 6-13 provides DCGLs for soil adjusted by currentground water concentration. The concentrationlevels are used in calculatingthe adjusted DCGL levels. To confirm the direction of ground water flow from past contamination sources and to verify that the highest ground water contaminationis monitored by the existing or new wells, please provide text supporting that the monitoring wells with the highest contamination levels are located in the area(s)of highest impact.

Based on current knowledge of existing ground water radioactivity and historical trends (see ground water concentration table in RAI #13), it appears that the radioactive plume has been bounded horizontally. However, to ensure that an area of radioactivity in the ground water higher in concentration than currently, or previously observed, is captured, additional work will be performed as indicated in RAIs #11 and #13.

15. Section 6.5. Development of DCGLs for Building Surfaces, page 6 To permit review of the input parametervalues used in dose evaluations and the derivation of DCGL values for the building surfaces in the LTP, please provide electronic files of the input and output of the RESRAD-BUILD and RESRAD codes used in deriving the DCGL values for the base case and check cases #1 and #2.

The input data files and program output files have been included on the enclosed CD-ROM. Also included is a copy of RESRAD-BUILD 2.37. Attachment 2 contains details regarding the file structure on the CD-ROM.

16. Section 6.5.1, Base Case DCGL Development, Building Occupancy Scenario, page 6 This section discusses the base case DCGL development and includes the applicablepathways for the building occupancy scenario. Indicate whether the chemical form of tritium compounds was investigated. Because 38

Enclosure 1 to CY-01-084 exposure to tritium (H-3) is a concern, the chemical form of tritium compound needs to be investigated. When H-3 is in the chemical form of HTO, which can be releasedmore easily than other solid chemical compounds from the concrete, inhalationof the HTO vapor releasedwill need to be considered.Indicate whether exposure from inhaling HTO vapor was included. If this pathway was included in the dose estimation, include it in the list of the pathways provided on page 6-33 of the LTP.

Inhalation of tritium vapor (HTO) is considered in the derivation of the building surface DCGL for tritium. As with all radionuclides considered in the building occupancy scenario, the tritium vapor is concentrated into a surface source. No absorption of tritium into the depth of the concrete is assumed, allowing for maximum release of tritium vapor from the surface. Parameters used to evaluate this source and dose pathway are listed in LTP Table 6-14 and are equal to the parameters used to evaluate the inhalation pathway of all other radionuclides.

As shown in LTP Table 6-24, the DCGL for tritium is limited by the resident farmer (concrete debris) scenario, not the building occupancy scenario.

The dose conversion factor incorporated into RESRAD 5.91 and RESRAD BUILD 2.37 (References 14 and 15) for the gaseous form of tritium is 6.3E-8 mrem/pCi. This is equivalent to the value listed in Federal Guidance Report 11 (References 16) for tritium vapor.

LTP Section 6.5.1 will be revised to clarify that inhalation of tritium is included in the applicable pathways. This will be included in a future change to the LTP.

17. Section 6.5.1.2. Assignment of Code Input Parameters,pages 6-34 to 6 This section discusses the values of input parametersused in the RESRAD-BUILD calculation to derive DCGLs for the building occupancy scenario.Surface contaminationwas assumed for the concrete walls, and an area source was specified in the calculation. Since the building scenarioshould considerall types of contamination,please provide text justifying this surface source assumption. If the assumption is that all buildings with activationpotential would be demolished, please provide text explaining the assumption.

This RAI and RAI #21(b) are related and raise two similar issues. The first is whether the use of surface sources rather than volumetric sources is conservative for deriving DCGLs for the building occupancy scenario. The second is whether using the concrete debris DCGLs to address activated concrete would produce an estimated dose less than the release criteria for the building occupancy scenario. To derive the DCGLs, both surface and volumetric residual radioactivity were considered and the more conservative value was used in selection of bounding DCGLs. CYAPCO does not assume that all buildings with activation potential are demolished, although that potential is included in the evaluation, as it is for all buildings. The potential for activation is only considered likely in the Containment Building and the consideration of volumetric residual radioactivity allows for its presence in selection of the building surface DCGLs proposed for use at HNP.

By way of clarification, there is a distinction between building surface DCGLs and DCGLs derived using the building occupancy scenario. Building surface DCGLs 39

Enclosure I to CY-01-084 are derived by comparing the DCGLs from two separate scenarios and selecting the lower (more conservative) value for each radionuclide. The building occupancy scenario is just one of these two scenarios.

The process of determining DCGLs for building surfaces is shown in revised LTP Figure 6-6. This figure is being revised to clarify the process of determining building surface DCGLs using two scenarios: building occupancy and resident farmer (concrete debris). The two scenarios are evaluated separately and then the more restrictive DCGL for each radionuclide is selected. (Note: The excavator scenario will be deleted from the LTP. See the response to RAIs #22-25.)

Buildings at HNP will be surveyed for residual radioactivity while standing. The building occupancy scenario is evaluated to determine DCGLs for the option of plant structures left standing. The building occupancy scenario is described in Section 6.5.1. Once the buildings are released, CYAPCO may choose to demolish the structures and bury the concrete debris on site. To ensure that the building surface DCGLs are conservative for the case of the buried concrete, another scenario is evaluated. This scenario is the resident farmer (concrete debris), discussed in LTP Sections 6.5.2.1.

As shown in revised LTP Figure 6-6, the two scenarios are evaluated separately to derive two sets of building surface DCGLs.

"* DCGLs for the building occupancy scenario are derived using surface sources, which is consistent with Reference 1, Section 6.2.1. DCGLs for this scenario apply to surface sources and are listed in LTP Table 6-17 in units of 2 2 pCi/m 2 and disintegrations per minute per 100 cm (dpm/1 00cm ).

"* DCGLs for the concrete debris (resident farmer) scenario are derived using volumetric sources. DCGLs for this scenario apply to volumetric sources and are listed in LTP Table 6-20 in units of pCi/g.

40

Enclosure 1 to CY-01-084 Revised LTP Figure 6-6 Process for Determining Building Surface DCGLs Historical Site Assessment

& Site Characterization Unit Concentration of Each Individual Radionuclide Evaluate/Select Pathways Plant Structures for Radionuclides Plant Structures Left Standing Demolished Resident Farmer Building Occupancy Scenario (Concrete Debris) Scenario RESRAD 5.91 Using Model Default and RESRAD-BUILD 2.37 Site Specific Data Using Model Default and Site Specific Input Data (Table 6-14) Determine Resident Farmer (Concrete Debris) DCGLs that Result in 25 mrem/yr (Table 6-20)

Determine Building Occupancy DCGLs that Convert Volumetric DCGLs to Pre Result in 25 mrem/yr Demolition Surface DCGLs (Table 6-17) (Table 6-20) 1 Compare DCGLs (Table 6-23):

Use Lower to Bound Both Scenarios (Table 6-24) 1 Use Area Factors for DCGL EMC (Tables 5-7 and 5-8) 41

Enclosure 1 to CY-01-084 In order to compare the two sets of DCGLs using consistent units, a method of converting the volumetric DCGLs to surface DCGLs is needed. The process of converting volumetric DCGLs to surface DCGLs is described in LTP Section 6.5.2.1.5 for buried concrete debris. LTP Equation 6-14 is used to convert 2

volumetric DCGLs (pCilg) to surface DCGLs (dpm/1 00cm ). For the HNP site, the assumption is made that the entire quantity of radioactivity contained within the volume occupied by the available fill area is distributed on the internal surface area of the Containment Building. LTP Table 6-20 (which will be revised to reflect the responses to RAIs #19 and # 20) lists DCGLs for the resident farmer (concrete debris) scenario in volumetric units (pCi/g) and also for the 2

extrapolated pre-demolition condition in surface units of dpm/1 00cm .

The next step in the process is to compare the two sets of building surface DCGLs and choose the minimum value for each radionuclide. Revised LTP Table 6-23 presents the two sets of building surface DCGLs. The limiting value (minimum) is selected and the controlling scenario is identified for each radionuclide. This process ensures that the annual dose to an average member of the critical group does not exceed 25 mrem, whether the buildings remain standing or are demolished and buried.

Revised LTP Table 6-24 again presents the limiting building surface DCGL and controlling scenario for each radionuclide. A column has been added so that the DCGLs are also listed as equivalent volumetric DCGLs, using LTP Equation 6-14 in reverse to convert from surface to volumetric values. These volumetric DCGLs will apply to all activated concrete sources found throughout the plant structures.

(Note: The footnote under LTP Table 6-20 will be removed. DCGLs from revised Table 6-24 will be used for activated concrete.)

42

Enclosure 1 to CY-01-084 Revised Comparison of Building Surface DCGLs from All Postulated Scenarios (LTP Table 6-23)

Revised Resident Farmer Building Bounding Controlling Scenario (Concrete Debris) DCGL Occupancy (Minimum)

(RAI 19 & 20) DCGL DCGL (dpm/1 00cm )2 (dpm/1 00cm 2)

DCGL (pCi/g) DCGL (dpm/1 00cm')

H-3 4.03E+02 5.19E+06 7.89E+08 5.19E+06 Resident Farmer (Concrete Debris )

C-14 2.78E+01 3.58E+05 2.57E+07 3.58E+05 Resident Farmer (Concrete Debris )

Mn-54 7.42E+02 9.56E+06 3.40E+04 3.40E+04 Building Occupancy Fe-55 5.42E+03 6.98E+07 5.94E+07 5.94E+07 Building Occupancy Co-60 7.47E+01 9.62E+05 1.17E+04 1.17E+04 Building Occupancy Ni-59 5.33E+03 6.87E+07 6.05E+07 6.05E+07 Building Occupancy Ni-63 1.94E+03 2.50E+07 2.55E+07 2.50E+07 Resident Farmer (Concrete Debris)

Sr-90 9.55E-01 1.23E+04 1.11 E+05 1.23E+04 Resident Farmer (Concrete Debris)

Nb-94 6.86E+00 8.84E+04 1.74E+04 1.74E+04 Building Occupancy Tc-99 2.03E+01 2.61 E+05 1.61 E+07 2.61 E+05 Resident Farmer (Concrete Debris)

Cs-1 34 1.37E+00 1.76E+04 1.81 E+04 1.76E+04 Resident Farmer (Concrete Debris)

Cs-137 1.81 E+00 2.33E+04 4.83E+04 2.33E+04 Resident Farmer (Concrete Debris)

Eu-152 3.73E+03 4.80E+07 2.42E+04 2.42E+04 Building Occupancy Eu-154 2.89E+03 3.72E+07 2.27E+04 2.27E+04 Building Occupancy Eu-1 55 6.49E+04 8.36E+08 4.34E+05 4.34E+05 Building Occupancy Pu-238 3.64E+01 4.69E+05 4.96E+02 4.96E+02 Building Occupancy Pu-239 2.72E+01 3.50E+05 4.51 E+02 4.51 E+02 Building Occupancy Pu-241 1.55E+03 2.OOE+07 2.47E+04 2.47E+04 Building Occupancy Am-241 3.19E+01 4.11 E+05 4.37E+02 4.37E+02 Building Occupancy Cm-243 4.68E+01 6.03E+05 6.45E+02 6.45E+02 Building Occupancy 43

Enclosure 1 to CY-01-084 Revised Bounding Building Surface DCGLs (LTP Table 6-24)

Limiting (Minimum) DCGL Surface DCGL Volumetric DCGL Controlling Scenario (dpm/1 00cm 2) (pCi/g)1 H-3 5.19E+06 4.03E+02 Resident Farmer (Concrete Debris)

C-14 3.58E+05 2.78E+01 Resident Farmer (Concrete Debris)

Mn-54 3.40E+04 2.64E+00 Building Occupancy Fe-55 5.94E+07 4.61 E+03 Building Occupancy Co-60 1.17E+04 9.08E-01 Building Occupancy Ni-59 6.05E+07 4.70E+03 Building Occupancy Ni-63 2.50E+07 1.94E+03 Resident Farmer (Concrete Debris)

Sr-90 1.23E+04 9.55E-01 Resident Farmer (Concrete Debris)

Nb-94 1.74E+04 1.35E+00 Building Occupancy Tc-99 2.61 E+05 2.03E+01 Resident Farmer (Concrete Debris)

Cs-134 1.76E+04 1.37E+00 Resident Farmer (Concrete Debris)

Cs-137 2.33E+04 1.81 E+00 Resident Farmer (Concrete Debris)

Eu-1 52 2.42E+04 1.88E+00 Building Occupancy Eu-1 54 2.27E+04 1.76E+00 Building Occupancy Eu-1 55 4.34E+05 3.37E+01 Building Occupancy Pu-238 4.96E+02 3.85E-02 Building Occupancy Pu-239 4.51 E+02 3.50E-02 Building Occupancy Pu-24i 2.47E+04 1.92E+00 Building Occupancy Am-241 4.37E+02 3.39E-02 Building Occupancy Cm-243 6.45E+02 5.OOE-02 Building Occupancy Equal to Surface DCGL divided by 12,881 per LTP Equation 6-24 44

Enclosure 1 to CY-01-084 With the building surface DCGL derivation process clarified and the proper volumetric DCGLs assigned for activated concrete sources, the issue of surface sources vs. volumetric sources for the building occupancy scenario can be addressed. To verify that the equivalent volumetric DCGLs listed in revised LTP Table 6-24 are conservative with respect to volumetric DCGLs derived specifically for the building occupancy scenario, RESRAD-BUILD was run with varying thicknesses of concrete sources. As the source thickness increases, the dose to a building occupant increases and DCGLs consequently decrease. Beyond a certain depth, which depends on the radionuclide and its emitted radiation, the thickness of the source becomes effectively infinite and no further decrease in DCGLs occurs. It was determined that a 12" thick source models an infinitely thick source for all radionuclides of interest and yields the most conservative (lowest) volumetric DCGLs. These values are compared to those derived above from equivalent surface DCGLs in the following table.

Comparison of DCGLs Using Volumetric and Surface Sources for Building Occupancy Scenario Radionuclide Building Limiting Scenario Ratio of Limiting Occupancy DCGL (pCi/g) Scenario DCGL to Scenario: 12" 12" Volumetric Volumetric Source Source DCGL DCGL (pCi/g)

H-3 1.71 E+03 4.03E+02 2.35E-01 C-14 5.52E+08 2.78E+01 5.04E-08 Mn-54 9.69E+00 2.64E+00 2.72E-01 Fe-55 1.27E+09 4.61 E+03 3.63E-06 Co-60 3.09E+00 9.08E-01 2.94E-01 Ni-59 1.30E+09 4.70 E+03 3.63 E-06 Ni-63 5.45E+08 1.94E+03 3.56E-06 Sr-90 2.56E+03 9.55E-01 3.73E-04 Nb-94 5.15E+00 1.35E+00 2.62E-01 Tc-99 3.46E+08 2.03E+01 5.87E-08 Cs-1 34 5.27E+00 1.37E+00 2.60E-01 Cs-137 1.46E+01 1.81 E+00 1.24E-01 Eu-152 7.16E+00 1.88E+00 2.62E-01 Eu-154 6.53E+00 1.76E+00 2.70E-01 Eu-1 55 3.45E+02 3.37E+01 9.77E-02 Pu-238 1.04E+04 3.85E-02 3.71 E-06 Pu-239 9.16E+03 3.50E-02 3.82E-06 Pu-241 4.55E+04 1.91 E+00 4.20E-05 Am-241 1.33E+03 3.39E-02 2.55E-05 Cm-243 8.77E+01 5.01 E-02 5.71 E-04 It can be seen that the limiting scenario volumetric DCGLs derived from equivalent surface sources using LTP Equation 6-14 are conservative (i.e., DCGLs are lower or all ratios of the limiting scenario DCGL to the 12" volumetric source DCGL are less than one) with respect to the volumetric DCGLs calculated by RESRAD-BUILD 45

Enclosure I to CY-01-084 for building occupancy with 12" volumetric sources. Thus, CYAPCO has committed to decontaminating buildings to levels lower than those that would result using volumetric sources for the building occupancy scenario. The use of surface sources is conservative for HNP.

The LTP will be revised in a future change to include the previously identified changes.

18. Section 6.5.1.2, Assignment of Code Input Parameters,ParameterJustification, pages 6-39 to 6 This section discusses the selected values for the important parametersidentified by sensitivity analyses. To enable staff review of parameterjustification, please provide text and references supporting the selection of the perturbationranges.

As indicated in the response to RAI #10 above, according to Appendix C of the NMSS Decommissioning SRP (Reference 8a):

"..deterministic sensitivity analysis, calculates the change in the output result (i.e.

peak dose) with respect to a small change in the independent variables, one at a time" A normalized sensitivity coefficient, as used in the LTP is further defined as follows:

dPk I P Sk-=d~/

dak lak where dPl/dak is the marginal sensitivity of P to ak. Sk describes the percentage change of performance measure P to a 1% change of parameter cik. The derivative that defines the marginal sensitivity can calculate to second order accuracy using a central difference scheme, allowing Sk to be represented as s = (P+, - 1ý-) / PO 2

Aak I ako The calculation of each individual Sk requires two model runs: one with ak = ak+1 to determine P+1 , and one with ak = ak.l to determine P-1. This equation is essentially identical in form to the sensitivity equation given in Reference 9.

The above equation was used to generate normalized sensitivity coefficients, as discussed in Section 6.4.4 of the LTP. This equation was solved using a central finite difference scheme to determine the sensitivity about a point, or in this case the input parameter.

Preliminary iterations were used for the range in perturbation to ensure that the sensitivity coefficient was approximate less than or equal to 1. The ranges of values chosen to evaluate the sensitivity of key RESRAD-BUILD input parameters for the building occupancy scenario were relatively small, in 46

Enclosure 1 to CY-01-084 accordance with the guidance of References 3 and 8a, and are listed in the following table.

Range of Parameter Values in Sensitivity Analysis Perturoation Parameter (units) LTP Value Range +1 Receptor Height (m) 1 0.25 Indoor Fraction (unitless) 0.25 0.1 Room Dimension (m) 10 5 Removable Fraction (unitless) 0.1 0.05 Lifetime (days) 365 182.5 Air Fraction (unitless) 0.1 0.05 Deposition Velocity (m/sec) 1.00E-02 5.00E-03 Resuspension Rate (sec"1) 5.OOE-07 2.50E-07 Breathing Rate (mi/day) 18 10 Ingestion Rate (m 2/hr) 1.OOE-04 5.OOE-05 HVAC Removal Rate (hr-1) 0.8 0.2

19. Section 6.5.2.1.3. Applicable Pathways,pages 6-46 to 6 Table 6-18 lists the applicablepathways for concrete debris under the resident farmerscenario.

Only the external radiation,inhalation, and water-dependentingestion pathways are included in the analysis. The reasons cited in the LTP for not considering the water-independentcomponents of the plant, meat, and milk pathways, as well as the soil ingestion pathway cannot be accepted. Given a time frame of 1,000 years, as well as leveling the concrete debris with soil after the decontaminationactivities, radiation exposures resultingfrom farming on the contaminatedarea should not be excluded from the dose assessment.Provide analyses including the active pathways of soil ingestion and water-independent ingestion.

As indicated in the LTP, the following soil independent pathways were not used for calculation of water-independent doses:

"* Meat

"* Milk

"* Plant

"* Soil Ingestion The LTP assumed that there would be no soil cover over the buried concrete debris - therefore farming of these areas was not considered practical (Note: it was assumed that plant could not be grown in concrete debris alone). Therefore, no plant growth equates to no livestock or humans eating plants containing residual radioactivity. Consequently, radioactivity in plants, meat and milk was not considered. Also, direct ingestion of concrete was not considered credible since no plants would be consumed.

47

Enclosure 1 to CY-01-084 However, in view of the concern that the assumption of no farming activities on the concrete debris may not be justified given the 1,000-year timeframe and in order to determine if the assumption was, in fact, non-conservative, the simulation has been re-analyzed assuming that farming activities occur on the concrete debris. An important key assumption that is implied is that cover soil will be necessary to allow farming activities.

Based on the 1000-year timeframe, there is the potential for three distinct scenarios for the water-independent pathways. They are as follows:

Scenario 1: Assume no cover soil and use only exposure and inhalation pathways, as presented in the LTP. This scenario represents the potential conceptual model where there is no cover soil present over the concrete debris zone. No other water dependent pathways (i.e. ingestion) are considered.

Scenario 2: Assume cover soil depth of 1-m and root depth of 0.5-m. This scenario represents the conceptual model where no roots from plants are in the concrete debris zone. All water dependent pathways are considered for this scenario.

Scenario 3: Assume cover soil depth of 0.5-m and root depth of 1-m. This scenario represents the conceptual model where all roots from plants are in the concrete debris zone. All water dependent pathways are considered for this scenario.

In addition, the following assumptions and/or changes to the input data file are made:

"* The cover material is assumed to have identical physical properties to the soil properties previously used (Reference 4).

"* A Soil mixing layer of 0.15-m is assumed (Reference 8)

"* Ground water fraction for livestock watering and irrigation is set at zero (for the purpose of reviewing water independent pathways).

Based on area of the containment building (1533 M2 ) where the debris would be located, it was acknowledged that obtaining 100% of the critical group's diet was not feasible, as that area is 15% of minimum area of 10,000 M2 recommended by Reference 2 to support 50% of the dietary requirements for the critical group.

Therefore, the actual percentage was calculated by RESRAD (based on the model output, the percentages for food derived onsite are 50% for plant food and 8% for meat and milk).

The calculated water-independent doses, considering ground cover and the three previously discussed exposure pathways, are summarized in Table 1 below (the highest doses are shaded for each radionuclide).

48

Enclosure 1 to CY-01-084 Table 1 Concrete Water Independent Dose Summary I Scenario 3, Radionuclide Scenario 1 Scenario 2 Dose, mrem/yr Dose, mremlyr Dose, mremlyr 0.OOE+00 1.85E+00 H-3 6.82E-02 3.30E-02 1.62E-23 870E+01 C-14 2.39E-04 2.1 OE+00 Mn-54 1.50E+02 5.39E-07 0.OOE+00 1 .87E-03 Fe-55 Ni-59 6.14E-07 4.32E-02 4.73E-02 8.73E-05 1I29E-01 Ni-63 1.43E-06 4.52E-03 8.96E+00 Co-60 6.46E+02 1.13E+00 1.24E-08 1.57E+02 Sr-90

4. 09E+02 3.64E+02 3.64E+02 Nb-94 2.95E-03 2.99E-1 7 1 .39E+01 Tc-99 3.35E+02 2.54E-04 1.01 E+01 Cs- 134 7.86E-05 7.90E+00 Cs-137 1)

-40E+02 8.42E-04 5.52E-01 Eu-1 52 2,90E+02 1.21 E-03 6.94E-01 Eu-1 54 3.15E+02 1.72E-12 1.20E-02 Eu-1 55 7.42E+00 9.57E-02 1.02E-02 1.06E+01 Pu-238 1.11E-01 2.75E+01 2.75E+01 Pu-239 5.97E-02 2.17E-01 2.33E-01 Pu-241 2.01 E+O0 6.31 E+00 1.21E+01 Am-241 Cm-243 2.51E+01 3.25E-02 8.28E+00 Note: Dose is based on initialparent radionuclideconcentrationin concrete as 100 pCi/g Because the availability of cover is uncertain over the expanse of 1000 years and all three scenarios above cannot be maintained concurrently, the DCGLs associated with Scenario 3 were selected for use. In general, the changes to the bounding DCGLs are minor, as can be seen by comparing the DCGLs provided in the "Concrete DCGL Comparison" table (see the response to RAI #20) to the DCGLs provided in the LTP for the building occupancy and resident farmer (concrete debris) scenarios. The following are noted:

" With the use of the assumption of cover (with roots growing into the concrete debris), the controlling scenario for one radionuclide (Ni-63) changes. The controlling scenario for Ni-63 becomes the resident farmer (concrete debris) where it was previously building occupancy. With that said, however, the change in the actual DCGL for Ni-63 is minor 2 2.50E+07 dpm/100 cm2 now compared to 2.55E+07 dpm/100 cm in the LTP.

"* For areas in which significant increases in the DCGL are realized with the use of Scenario #3 (i.e., Mn-54, Co-60, Eu-152, Eu-154, Eu-155) the 49

Enclosure 1 to CY-01-084 controlling scenario remains the building occupancy scenario. Thus, while the resident farmer (concrete debris) DCGLs will increase, the bounding DCGLs applied onsite will not change, as they are based upon the building occupancy scenario.

" The DCGLs for Cs-1 34 and Cs-1 37 increase slightly with the use of Scenario #3 (Cs-134 +17%; Cs-137 +10%); however, the controlling scenario continues to be the resident farmer (concrete debris) scenario.

" Four radionuclides (H-3, C-14, Sr-90, and Tc-99) whose DCGLs are currently controlled by the resident farmer (concrete debris) will remain controlled by this scenario; however, the DCGLs associated with this scenario will decrease.

" The remaining radionuclide DCGLs (Fe-55, Ni-59, Nb-94, Pu-238-41, Am 241, Cm-243) continue to be controlled by the building occupancy scenario and thus are unchanged.

Section 6.5.2.1 of the LTP will be updated to reflect the use of the conditions of Scenario #3 in calculation of the DCGLs for the resident farmer (concrete debris) scenario. In addition, current LTP Tables 6-20, 6-23, and 6-24 will be updated, as appropriate, to reflect the values calculated using the assumptions associated with Scenario #3.

the

20. Section 6.5.2.1.4. Dose Modeling Approach, pages 6-47 to 6 Step 3 of concentration of the parent dose modeling approach computes the ground water radionuclide at the beginning of the assessment period. The maximum radiation exposure may not be observed at this time. When calculating radiation exposure, dose contributionsfrom both the parentradionuclidesand the progeny radionuclidesneed to be taken into account. The ingrowth of progeny radionuclidesmay shift the occurrence of the maximum exposure from the beginning of the period to a later time. In that case, DCGLs for the parent radionuclidesshould be derived using the maximum radiationdoses at the later time instead of the radiationdoses at the beginning of the period. If the radiation that doses at the beginning time are used to derive DCGLs, it cannot be assured the dose limit of 25 mrem/yr will not be exceeded in the future. Please include analysis. If potential dose contributionsfrom progeny radionuclidesin the progeny radionuclidesare not included, then provide justification that the future radiationexposure would not exceed the exposure incurredat the beginning of assessmentperiod.

dose Initially, radioactive progeny were not considered when the water-dependent calculation was performed (Note: The RESRAD model considers radioactive that a progeny when calculating the water-independent dose). However, it is feasible This daughter product may produce a higher water-dependent dose than its parent.

50

Enclosure 1 to CY-01-084 would result in a lower parent radionuclide DCGL. Based on the radioactive decay series presented in Reference 2, the following radionuclides required additional analysis:

"* Eu-152

"* Pu-238

"* Pu-239

"* Am-241

"* Pu-241

"* Cm-243 Additionally, inspection of the decay series for each radionuclide of interest shows that only certain progeny are of relative importance - that is, the location of the progeny in the series and long-lived progeny are important. Based on these two criteria, the following "adjusted" decay series are proposed:

152E 152G 63 Eu-- 64Gd(28%)

23 8 234 23 22 6 21 0 Pu-> u-), 0, _ý R8 Pb 94 92 90 88 a--> 82' 243 M- .2939"

> 235r u >231 _2ý27 96 * - 92 89 239PU 235u -).231 'P _227 2 2 94 u-9 2 U9 9 8a--9AC 241P-i 2941A* 2937Np 233rr .229,"rq 94*P ' 95 m'- 93!V " 92 U'-*90-t 951 237* _ 233ur 229,7,/

95Am'- 93Np- 92 U-* 90 According to Reference 17, a 1000-year time span is required to calculate the maximum dose for unrestricted free release. Therefore, the maximum concentrations for each radioactive progeny are required for this timespan.

Several progeny were eliminated, based on their activities and/or half-lives (less than 6 months). The maximum progeny activities for each remaining decay series are summarized in Table 1.

51

Enclosure 1 to CY-01-084 Table 1 Radioactive Progeny Maximum Concentration Progeny T1/2, years1 Time of Maximum Maximum Activtiy=,

Concentration, pCi years Pu-241 Series (14.4 years)

Am-241 432.2 100 2.82xl 0T Cm-243 Series (28.5 years)

Pu-239 2.41x10" 250 1.14x10l 1Reference 2 2 Based on initial radionuclide concentration of 100 pCi/g, or 9.73xl 011 pCi total for volume of concrete in containment building The distribution coefficient (Kd) values were obtained from a literature search.

Table 2 summarizes the observed Kd range as well as the value adopted for dose modeling. As Table 2 illustrates, this parameter exhibits significant variability.

Most of these values were obtained from the preferred Kd database for water cement mixtures published in NUREG/CR-6377 by Krupka and Serne (Reference 18).

Table 2 Kd Values for Concrete Debris (cm 3lg)

Isotope Minimum Maximum Adopted Source Pu-241 Series Am-241 1 2,000 35,000 5,000 NUREG/CR-6377 by Krupka and Serne (1996)

Cm-243 Series Pu-239 - 390 17,000 5,000 NUREG/CR-6377 by Krupka and Serne (1996)

Isotope concentrations for ground water in contact with concrete debris are determined for each progeny at time "t" for the observed maximum activity.

The ground water concentration associated with concrete debris using total activity of the concrete/ground water mix can be calculated from:

• 1000 C= (Total activity) 1+Kdpb where: pb = bulk density of concrete; and n = porosity of the concrete debris. This equation assumes that C is expressed in units of pCi/I and Total activity is expressed in units of pCi. The calculation of C requires values of pb and n for the concrete debris as well as Kd for each radionuclide. Values of n=0.30 (Reference 6 - value for coarse 52

Enclosure 1 to CY-01-084 g/cm 3 gravel) and Pb = 1.68 g/cm (Note: pb =(1 -n) pco,,cete, where n=0.30 and Pcete =2.4 3

[Reference 6]) have been assigned. The assumptions previously discussed in the LTP regarding sorption also considered valid in this discussion. Table 3 summarizes the groundwater concentrations resulting from the use of the previous equation.

Table 3 Water Dependent Progeny Concentration Summary Radionuclide Total Kd, Cw, Activity, cm 3/g pCi/I1 pCi Pu-241 Series Am-241 4 2.82x101 u 5000 0.58 Cm-243 Series Pu-239 4 1.14x10 5000 0.02 1 calculated by above equations Based on Table 3, the progeny water concentrations for Pu-241 (Am-241) and Cm-243 (Pu-239) are both less than 1 pCi/l. Therefore, the original ground water dose for each radionuclide, as presented in Table 6-12 of the LTP, represents the maximum dose over a 100-year time period. Therefore, no change to the LTP is required.

53

Enclosure 1 to CY-01-084 Table 4 Concrete DCGL Comparison (Considers Results of RAI#1 9 and #20)

Scenario I Scenario 11 ScenarioI Isotope Water Water Total Dose, Concrete Water Water Total Dose, Concrete Water Water Total Do Independent Doe Dose, (nui Dependent Dose r .) (G (...

Debris, Independent Depend 'enit Do se, Dose Debris, Independent Dependent Dose Debris, Dose, eISetO~e Dose 2 (inreni/yr.) (unrern/yr.) DCGLc,,, Dose, Dowe, (irnire/yr DC'GLcffc Dose, oe melr) DGC, Dose I (nirjemI/ilyr.) (pOi/g) Dose, (rnrernfyr.) Dose2 (pCilg) Dose, (rnreml/yr.) (pC'ilg)

(marem/yr.)______

')(snrem/yr.) 5.73E+02 1.85E+00 4.36E+00 6.21F+00 4.03E+02 H-3 6.82E-02 4.36E+00 4.43E+00 5.65E+i02 0.OOE+00 4.36E+00 4.36E+00 2.97E+00 3.00E+00 8.33E+02 1.62E-23 2.97E+00 2.97E+00 8.42E+02 8.70E+01 2.97E+00 9.OOE+01 2.78E+01 C-14 3.30E3-02 1.27E+00 1.51E+02 1.65E+01 2.39E-04 1.27E+00 1.27E+00 1.97E+03 2.10E+00 1.27E+00 3.37E+00 7.42E+02 Mn-54 1.50E+02 4.59E-01 4.59E-01 5.45E+03 O.00E+00 4.59E-01 4.59E-01 5.45E+03 1.87E-03 4.59E-01 4.61 E-01 5.42E+03 Fc-55 5.39E-07 2.45E+01 6.7 1E+02 3.73E+00 4.52E-03 2.45E+01 2.45E+01 1.02E+02 8.96E+00 2.45E+01 3.35E+01 7.47E+01 Co-60 6.4613+02 4.22E-01 5.92E+03 4.32E 02 4.22E-01 4.65E-01 5.37E+03 4.73E-02 4.22E-01 4.69E-01 5.33E+03 Ni-59 6.14E-07 4.22E-01 1.16E+00 1.16E+00 2.16E+03 8.73E-05 1.16E+00 1.166E+00 2.16E+03 1.29E-01 1.16E+00 1.29E+00 1.94E+03 Ni-63 1.43E-06 Sr-90 1.13E+00 2.46E+03 2.46E+03 1.02E+00 1.24E-08 2.46E+03 2.46E+03 1.02E+00 1.57E+02 2.46E+03 2.62E+03 9.55E-01 5.54E-01 4.1OE+02 6.10E+00 3,64E+02 5.54E-01 3.651,+02 6.86E+00 3.64E+02 5.54E- 01 3.65E+02 6.86E+00 Nb-94 4.09E+02 1.09E+02 1.09E+02 2.29E+601 .39EE-1 1.09E1+02 1.23E+02 2.03E+01 Tc-99 2.95E-03 3.35E+02 1.811E+03 2.15E+03 1.17E+00 2.54E-04 1.81E+03 1.81E+03 1.38E+00 1.01E+01 1.81E+03 1.82E+03 1.37E+00 Cs-134 1.37E+03 1.511E+03 1.66E+00 7.86E-05 1.37E+03 1.37E+03 1.82E+00 7.90E+00 1.37E+03 1.38E+03 1.81E+00 Cs-137 1.40E+02 2.90E+02 8.62E+OO 8.42E-04 1.18E-01 1.19E-01 2.10E+04 5.52E-01 1.18E-01 6.70E-01 3.73E+03 Eu-152 2.90E+02 1.18E-01 Eu-154 3.15E+02 1.711E-01 3.15F+02 7.93E+00 1.211E-03 1.71E-01 1.72E-01 1.45E+04 6.94E-01 1.711E-01 8.65E-01 2.89E+03 u- 155 7.42E+00 2.65E-02 7.45E+00 3.36E+02 1.72E-12 2.65E-02 2.65E-02 9.43E+04 1.20E-02 2.65E-02 3.85E-02 6.49E+04 Pu-238 9.57E-02 5.801E+01 5.811E+01 4.30E+01 1.02E-02 5.80E+01 5.80E+01 4.31E+O1 1.06E+01 5.80E+01 6.86E+01 3.64E+01 Pu-239 1.1 IE-01 6.45E+01 6.46E+01 3.87E+01 2.75E+01 6.45E+01 9.20E+01 2.72E+01 2.75E+01 6.45E+01 9.20E+01 2.72E+01 Pu-241 5.97E-02 1.38E+00 1.44E+00 1.74E+03 2.17E-01 1.38E+00 1.60E+00 1.57E+03 2.33E-01 1.38E+00 1.61E+00 1.55E+03 Am-241 2.O1E+00 6.62E+01 6.82E+01 3.67E+01 6.316+00 6.62E+01 7.25E+01 3.45E+01 1.21E+01 6.62E+01 7.83E+01 3.19E+01 Cm-243 2.51E+01 4.511E+01 7.02E+01 3.56E+01 3.25E-02 4.511E+01 4.51E+01 5.54E+01 8.28E+00 4.51E+01 5.341E+01 4.68E+01 54

Enclosure 1 to CY-01-084

- This section

21. Section 6.5.2.1.5. Dose Modeling Results for Concrete Debris, page 6-51

1. 1, resident farmer (concrete discusses the dose modeling results from check case debris) scenario.

a. To convert the derived DCGLs for concrete debris to the DCGLs for concrete walls, area of the the ratio between the total volume of the concrete and the total surface 3/16764 m 2 is the concrete is needed. It is stated in the LTP that a ratio of 5790 m walls.

minimum and its use would result in more conservative DCGLs for concrete to confirm the Please provide the source document for this minimum ratio soundness of this choice.

Building Surface The source document for this information is BCY-HP-001 8, "Concrete been provided as Attachment 3.

Area and Volume Determination," which has

b. Table 6-20 footnote states that volumetric DCGLs (derived for the concrete debris),

Suppose that for relevant radionuclides,will be used to addressactivated concrete.

is volumetrically the building is not demolished, that the building concrete structure instead of contaminated, and that a worker may be exposed to the volume source occupancy the surface contamination as assumed in the base case of the building (for example, using RESRAD-BUILD) to show scenario. Please provide analysis address activated concrete would produce that using the concrete debris DCGLs to an estimated dose less than the release criteria for the building occupancy scenario.

DCGLs listed in This RAI is addressed in the response to RAI #17. The volumetric the lowest revised LTP Table 6-24 apply to activated concrete. The table presents DCGLs: building DCGLs for the two scenarios evaluated to determine building surface debris). As discussed in the response to occupancy and resident farmer (concrete revised LTP Table 6-24 in a building RAI #17, using the volumetric DCGLs from less than the release criteria.

occupancy scenario will produce an calculated dose in a future change The footnote to LTP Table 6-20 will be deleted. This will be reflected to the LTP.

insert the RESRAD

22. Section 6.5.2.2. Check #2-ExcavatorScenario, page 6 Please the DCGLs for the excavator scenario, in order BUILD code version used in calculating staff to review the scenario.

2.37. As The excavator scenario was evaluated using RESRAD-BUILD Version is being discussed in the responses to RAIs #23 through 25, the excavator scenario the limiting values.

proposed for deletion from the LTP as none of the DCGLs were indoortime fraction

23. Section 6.5.2.2.1. Conceptual Dose Model Approach, page 6 An on the licensee's assumption of 0.25 is assumed. The selection of this value was based is also likely, and that an office building was constructed. In fact, construction of a house radiation exposure.

a largerindoortime fraction should be used to bound the potential the excavator scenario using an indoor time Therefore, please provide analysis for fraction representativeof a resident.

55

Enclosure 1 to CY-01-084 The excavator scenario was one of a total of three scenarios (the other two being building occupancy and resident farmer-concrete debris) evaluated to verify that DCGLs determined were sufficiently low that the criteria for unrestricted release would be met if the structures standing or if the buildings demolished and the debris buried on site. The verbiage in the LTP is somewhat misleading, as the intent of the excavator scenario was to model only the uncovered concrete debris, not digging the debris up and reusing the debris for building material. Construction of a house (or any other structure) from the buried concrete debris is not considered credible. The indoor time fraction of 0.25 applies to the construction worker standing on the uncovered debris, not to occupancy of a building.

The excavator scenario determined DCGLs such that an unsuspecting construction worker would not be exposed to an annual dose exceeding 25 mrem if the concrete debris were uncovered. The most conservative source configuration is an infinite planar source with all the residual radioactivity facing upward. An occupancy factor (RESRAD-BUILD uses the terminology "indoor fraction") of 0.25 was assumed which corresponds to a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> workweek for an entire year. Increasing the occupancy factor beyond the value of 0.25 for a worker is not reasonable. As intended, the excavator scenario tested whether exposure from an infinite slab is greater than exposure from all surfaces of a room (floor, ceiling, four walls). It was demonstrated that standing on an infinite slab gives a lower external dose than occupancy of a building of the same concentration. In other words, DCGLs were higher (less restrictive) for the excavator scenario than for the building occupancy scenario. This is shown in the original LTP Table 6-23 which compares DCGLs for the three scenarios evaluated to determine building surface DCGLs. None of the DCGLs for the excavator were the limiting values.

CYAPCO used the excavator scenario to conclude which configuration (infinite slab versus surfaces of a room) gave the greatest external exposure. Since the building occupancy scenario gave consistently more restrictive DCGL values than the excavator scenario, the excavator scenario was not used as input to any bounding DCGLs. For this reason, CYAPCO is proposing to delete the excavator scenario from the LTP. Changes to the LTP to remove the excavator scenario will be reflected in a future revision to the LTP.

24. Section 6.5.2.2.2. Applicable Pathways,page 6 Only the direct external dose pathway is consideredfor the excavator scenario. The considerationof only the external radiationexposure is not comprehensive and, therefore, cannot be accepted.

Radionuclides from the planarsource have the potential of being suspended into the air and being inhaled by the excavator. Furthermore,direct incidentalingestion of radionuclidesfrom the planarsource and secondary incidental ingestion of radionuclides deposited from the airbornesource cannot be excluded. Provide analysis for the excavatorscenario including all other relevant pathways.

As stated in the response to RAI #23, the excavator scenario was intended only to model the uncovered concrete debris, not its reuse. Thus, only the direct external dose pathway is considered. The intent of the excavator scenario is to verify that exposure from an infinite planar source is less than that from occupancy of a contaminated building (where the sources include finite-sized floor, ceiling, and walls) and from a farmer living above the debris. Including the exposure pathways of inhalation and ingestion in effect changes the excavator scenario back to the building occupancy scenario. As described in LTP Section 6.5.1, the building occupancy scenario considers external exposure, inhalation of resuspended contamination, and inadvertent ingestion 56

Enclosure 1 to CY-01-084 of contamination. Because of the inclusion of these pathways, the DCGLs for building occupancy are lower (more restrictive) than those calculated for the excavator scenario.

None of the DCGLs for the excavator are the limiting values, as shown in LTP Table 6-23 which compares the three scenarios evaluated for building surface DCGLs.

LTP Section 6.5.2.2 will be revised to delete the excavator scenario, as the building occupancy scenario bounded the excavator scenario in all cases. This will be reflected in a future change to the LTP.

25. Section 6.5.2.2.4. ExcavatorScenario DCGL Results, page 6 This section summarizes the results of the dose modeling calculations. The derived DCGLs are for the surface source that was formed by digging up the volume source lying underground (i.e., the concrete debris) and should not be applied directly as the concentration limits for the concrete walls priorto the demolishing. Pleaseprovide justificationfor using the plane source DCGLs as the building surface DCGLs.

LTP Section 6.5.2.2.4 summarizes the results of the excavator scenario only, not the results of all three scenarios evaluated for building surface DCGLs. As shown in the original LTP Figure 6-6, the process of determining building surface DCGLs consisted of the evaluation of three separate scenarios to address the options of plant structures left standing and plant structures being demolished and buried. The DCGLs listed in LTP Table 6-22 were derived as a method of ascertaining if exposure from the uncovered infinite planar source of contaminated concrete is less than from occupancy of a contaminated building and from a resident farmer living above the buried concrete debris. When compared to the DCGLs derived for the building occupancy and resident farmer (concrete debris) scenarios in the original LTP Table 6-23, it can be seen that none of the DCGLs for the excavator scenario were the limiting values. Therefore, the DCGLs presented in LTP Table 6-22 will not be used when surveying building surfaces, as they are not the most restrictive. The DCGL values in revised LTP Table 6-24 (either surface or volumetric, depending on the sources of contamination) will be used to establish compliance with site release criteria for buildings.

LTP Section 6.5.2.2 will be revised to delete the excavator scenario (as the building occupancy scenario bounded the excavator scenario in all cases), and this will be reflected in a future change to the LTP.

LTP SECTION 7, UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS

26. L TP Section 7. Update of Site-Specific Decommissioning Costs The response to this question will be supplied later.

LTP SECTION 8, SUPPLEMENT TO THE ENVIRONMENTAL REPORT

27. Section 8.1. Introduction, pages 8-1 to 8 In orderto verify the conclusionspresented in this section and determine the adequacyof the supporting information, please address the following:

57

Enclosure 1 to CY-01-084

a. FirstBullet, page 8 To determine that ground water is or is not a source of potential radiationexposure for the public, please provide text describing the following aspects of site ground water:

(1) Information on the location of the nearest residentialwell.

The nearest residential well is approximately 0.5 miles northwest (in the opposite direction of the groundwater plume travelling southwest of the RWST) of HNP stack.

(2) Details on the quality of water in the deep system (e.g., identification of measurable contaminants and theirconcentrations).

Current ground water radioactivity trends, based on existing monitoring well data, indicate a general downward trend in H-3 concentrations, where present (see the following table for data).

Sampling Data for Deep Monitoring Wells Well Round I I Round2 Round 3 Round 4 Number H-3 (pCi/L)* H-3 (pCilL)* H-3 (pCilL)* H-3 (pCilL)*

Tritium Analysis Data 100D ND ND Not Sampled <MDC 101D ND ND Not Sampled Not Analyzed for H-3 102D 2740 3160 2640 2470 103D 22180 17550 19660 20900 105D 4590 2450 3030 2150 106D 3320 1590 5830 1810 107D ND ND Not Sampled <MDC 109D 33070 31600 21230 15800 101D 27630 23280 27230 18300

• <MDC for plant related gamma-emitting isotopes (3) The locations of monitoring wells used to assess the quality of deep ground water.

The locations of existing ground water monitoring wells used to assess the quality of deep ground water are shown on the "Water Table and Analyte Contour Map-March 2, 1999 (Sht 2 of 5)" located in the figure section of Reference 5 and are designated by a "D" indicator (e.g., MW-1 10D). This report was provided to the NRC Staff as an attachment to CYAPCO letter CY-99-077, dated November 17, 1999.

b. Third Bullet, page 8 Cessation of pumping water from the shallow ground water system will allow the water level in the shallow system to increase. If the current direction of ground water flow is from the deep to the shallow, cessation of pumping 58

Enclosure 1 to CY-01-084 may allow contaminatedwater to move vertically downward and enterthe deep ground water system. Please provide text describing the watertable elevations for the shallow and deep aquifers at the site and what impacts may result from not pumping the shallow system.

Text and figures describing water table elevations for the aquifers at the site are described in Section 2.2.2 of the September 1999 Groundwater Monitoring Report and the May 2001 Phase 2 Hydrogeologic Investigation Work Plan. Both of these reports have been docketed and provided to the NRC Staff. Work is ongoing with the CT DEP to further characterize the groundwater flow and to integrate groundwater with a fate and transport model to develop a subsurface hydrogeologic model.

The external containment sump (ECS) in not a groundwater monitoring well, and is not representative of a specific location and depth interval. Rather, the ECS is a structure connected to the mat drain, which controls groundwater levels beneath the reactor containment throughout the original bedrock excavation. This excavation is very large in diameter, intersects numerous fractures as well as the overburden water table, and is pumped cyclically. The ECS is never at equilibrium as the pumps cycle on and off and the water level in the ECS fluctuates. The yield of individual bedrock fractures to the mat drain, their respective tritium concentration and travel time to the ECS varies. Although the ECS could be presumed to have a large averaging effect one could similarly surmise that the sampling time relative to pump cycle time could have a significant impact upon measured tritium concentration as the mat drain is dewatered and new flow is introduced. Therefore, cessation of pumping water from the ECS will likely remove the vertical gradient to the ECS and allow the ground water flow to reestablish an equilibrium water table elevation with flow toward the Connecticut River.

CYAPCO has proposed to discontinue the routine pumping of the external containment mat sump. We have committed to provide the CT DEP a detailed description of the planned groundwater monitoring of the sump and nearby wells prior to shut down of the sump pump. Monitoring results from this sampling program will be analyzed to determine if there is any impact from discontinuing routine pumping of the sump. A copy of the plan and the report summarizing the monitoring results and analysis will be provided to the CT DEP, NRC and EPA.

59

Enclosure 1 to CY-01-084 References

1. US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, "Residual Radioactive Contamination from Decommissioning". Vol. 1. NUREG/CR-5512, October 1992.

2. Yu, C, et al., "Manual for Implementing Residual Radioactive Material Guidance using RESRAD, Version 5.0"; US Department of Energy - Argonne National Laboratory, September 1993

3. US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, "Residual Radioactive Contamination from Decommissioning". Vol. 2. NUREG/CR-5512, May 1999.

4. Connecticut Yankee Atomic Power Company, "License Termination Plan", July 2000.

5. Groundwater Monitoring Report - Connecticut Yankee Atomic Power Company, Malcolm Pirnie, July 1999.

6. Yu et al., "Data Collection Handbook to Support Modeling Impacts of Radioactive Material in Soil," Argonne National Laboratory, April 1993.

7. U. S. Nuclear Regulatory Commission, "Preliminary Guidelines for Evaluating Dose Assessments in Support of Decommissioning."

8. Yu, C, et al., "Parameter Distributions for Use in RESRAD and RESRAD-Build Computer Codes", US Department of Energy - Argonne National Laboratory, March 2000.

8a. NUREG-1 727, NMSS Decommissioning Standard Review Plan, September 2000.

9. NUREG/CR-5512, Vol. 3 "Residual Radioactive Contamination From Decommissioning Parameter Analysis" Draft Report for Comment, dated October 1999

10. Schroeder et al., "The Hydrological Evaluation of Landfill Performance (HELP) Model:

Engineering Documentation for Version 3," EPA/600/R-941168b, U. S. Environmental Protection Agency Office of Research and Development, September 1994.

11. Leggette, Brashears & Graham, Inc. "Completion Report for the Hydrogeological Study of the Water-Supply Potential at the Connecticut Yankee Atomic Power Plant, Haddam, Connecticut". July 1985.

12. "GTS Duratek, Inc. "Augmented Characterization Survey Report for Connecticut Yankee Atomic Power Plant, January 1999"

13. "Results of Scoping Surveys For CYAPCO's HNP"

14. ANL/EAD/LD-2, "Manual for Implementing Residual Radioactive Material Guidelines Using RESRAD, Version 5.0," Environmental Assessment Division, Argonne National Laboratory, September 1993, Table B.1: Committed Effective Dose Equivalent Conversion Factors for Inhalation.

60

Enclosure 1 to CY-01-084

15. ANL/EAD/LD-3, "RESRAD-BUILD: A Computer Model for Analyzing the Radiological Doses Resulting from the Remediation and Occupancy of Buildings Contaminated with Radioactive Material," Environmental Assessment Division, Argonne National Laboratory, November 1994, Table D. 1: Committed Effective Dose Equivalent Conversion Factors for Inhalation.

16. Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestions.

FGR 11

17. NUREG/CR-6377, "Effects of Radionuclide Concentrations by Cement/Groundwater Interactions in Support of Performance Assessment of Low Level Radioactive Waste Disposal Facilities," November 1996.

18. US Nuclear Regulatory Commission, NUREG-1549, "Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination," July 1998.

61 to Enclosure 1 of CY-01-84 "DCGL Type by Survey Unit" Response to RAI 2-1

Attachment 1 to Enclosure 1 of CY-01-084 DCGL Type by Survey Unit Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 1102 Fuel Building Laydown Area 0001 Floor Area, Walls and Ceiling - Section 1 Bldg 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 0003 Floor Area, Walls and Ceiling - Section 3 Bldg 1104 Fuel Building Fuel Cask Decon Area 0000 Floor Area, Walls and Ceiling Bldg 1106 Fuel Building Skimmer Pump and Sump 0000 Floor Area, Walls and Ceiling Bldg Area 1202 Fuel Building New Fuel Storage Area 0000 Floor Area, Walls and Ceiling Bldg 1204 Fuel Building Exhaust Filters and Fan Area 0000 Floor Area, Walls and Ceiling Bldg 1302 Fuel Building Patio Area 0001 Roof Area Bldg 0002 Roof Area Bldg 1304 Fuel Building New Fuel Storage Area 0000 Floor Area, Walls and Ceiling Bldg 1306 Fuel Building Cask Laydown Area 0000 Floor Area, Walls and Ceiling Bldg 1308 Fuel Building Spent Fuel Pool Pit 0001 Floor Area, Pool Sides and Bottom, Walls Bldg and Ceiling - Section 1 0002 Floor Area, Pool Sides and Bottom, Walls Bldg and Ceiling - Section 2 1404 Fuel Building Roof Area 0001 Roof Area - Section 1 Bldg 0002 Roof Area - Section 2 Bldg 0003 Roof Area - Section 3 Bldg 0004 Roof Area - Section 4 Bldg 0005 East & West Exterior Walls Bldg 0006 North & South Exterior Walls Bldg 2002 Auxiliary Building RHR Pump Room A 0000 Floor Area, Walls and Ceiling Bldg 2004 Auxiliary Building RHR Pump Room B 0000 Floor Area, Walls and Ceiling Bldg 2006 Auxiliary Building RHR Heat Exchanger 0000 Floor Area, Walls and Ceiling Bldg Room 2008 Auxiliary Building Primary Drain Tank 0000 Floor Area, Walls and Ceiling Bldg Pump Room Page 1 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 2010 Auxiliary Building Primary Drain Tank 0000 Floor Area, Walls and Ceiling Bldg Room 2012 Auxiliary Building Aerated Drain Tank 0000 Floor Area, Walls and Ceiling Bldg Room 2104 Auxiliary Building Pipe Chase Under 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Hallway 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 2106 Auxiliary Building Pipe Chase Under Valve 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Room 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 2108 Auxiliary Building Boric Acid Evaporator 0000 Floor Area, Walls and Ceiling Bldg Area TK EVI-lA, EV2-lA 2110 Auxiliary Building Pipe Chase East & West 0000 Floor Area, Walls and Ceiling Bldg Outside 2202 Auxiliary Building Hallway 0001 Floor Area, Walls and Ceiling - Section 1 Bldg 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 2204 Auxiliary Building Component Cooling Area 0000 Floor Area, Walls and Ceiling Bldg 2206 Auxiliary Building Boric Acid Evaporator 0000 Floor Area, Walls and Ceiling Bldg Area 2208 Auxiliary Building Boric Acid Mix Tank 0000 Floor Area, Walls and Ceiling Bldg Area 2210 Auxiliary Building "B" Charging Pump Area 0000 Floor Area, Walls and Ceiling Bldg 2212 Auxiliary Building "A" Charging Pump Area 0000 Floor Area, Walls and Ceiling Bldg 2214 Auxiliary Building Metering Pump Area 0000 Floor Area, Walls and Ceiling Bldg 2216 Auxiliary Building Purification Pump Area 0000 Floor Area, Walls and Ceiling Bldg 2218 Auxiliary Building Primary Water Transfer 0000 Floor Area, Walls and Ceiling Bldg Pump Area 2220 Auxiliary Building Sample Room 0000 Floor Area, Walls and Ceiling Bldg 2222 Auxiliary Building Steam Generator 0000 Floor Area, Walls and Ceiling Bldg Blowdown Room 2224 Auxiliary Building HPSI Cubicle Area 0000 Floor Area, Walls and Ceiling Bldg 2226 Auxiliary Building LPSI Cubicle Area 0000 Floor Area, Walls and Ceiling Bldg 2228 Auxiliary Building Drumming Room 0001 Floor, Walls and Ceiling - Section 1 Bldg Page 2 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 0002 Floor, Walls and Ceiling - Section 2 Bldg 2302 Auxiliary Building Component Cooling Area 0000 Floor Area, Walls and Ceiling Bldg 2304 Auxiliary Building Boric Acid Evaporator 0000 Floor Area, Walls and Ceiling Bldg Area 2306 Auxiliary Building Boric Acid Mix Tank 0000 Floor Area, Walls and Ceiling Bldg Area 2308 Auxiliary Building Volume Control Tank 0000 Floor Area, Walls and Ceiling Bldg Room 2310 Auxiliary Building Purge and Dilution Fan 0000 Floor Area, Walls and Ceiling Bldg Areas 2312 Auxiliary Building Service Water Strainer 0000 Floor Area, Walls and Ceiling Bldg Area 2314 Auxiliary Building HEPA Filter and Hall 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Area 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 0003 Floor Area, Walls and Ceiling - Section 3 Bldg 0004 Floor Area, Walls and Ceiling - Section 4 Bldg 0005 Floor Area, Walls and Ceiling - Section 5 Bldg 0006 Floor Area, Walls and Ceiling - Section 6 Bldg 0007 Floor Area, Walls and Ceiling - Section 7 Bldg 0008 Floor Area, Walls and Ceiling - Section 8 Bldg 2316 Auxiliary Building Boric Acid Storage Room 0001 Floor Area, Walls and Ceiling - Section 1 Bldg 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 2402 Auxiliary Building Roof Area 0001 Roof Area by Ventilation Duct- Section 1 Bldg 0002 Roof Area by Ventilation Duct- Section 2 Bldg 0003 Roof Area by Ventilation Duct- Section 3 Bldg 0004 Remaining Roof Area - Section 1 Bldg 0005 Remaining Roof Area - Section 2 Bldg 0006 Exterior Walls Bldg 3002 Containment Enclosure Under Reactor 0000 Floor Area and Walls Bldg Vessel Page 3 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 3004 Containment Enclosure Sump Area Under 0000 Floor Area and Walls Bldg Reactor Vessel 3101 Containment Enclosure #4 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Lower Level NE 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3102 Containment Enclosure #1 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Lower Level NW 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3103 Containment Enclosure #2 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Lower Level SW 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3104 Containment Enclosure #3 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Lower Level SE 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3105 Containment Enclosure Containment Sump 0000 Floor Area, Walls and Ceiling Bldg

_Area 3107 Containment Enclosure Cable Vault Outside 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Containment 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3111 Containment Enclosure Loop #1 Inner 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Annulus Lower Level NE 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3112 Containment Enclosure Loop #2 Inner 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Annulus Lower Level NW 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3113 Containment Enclosure Loop #3 Inner 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Annulus Lower Level SW 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3114 Containment Enclosure Loop #4 Inner 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Annulus Lower Level SE 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3201 Containment Enclosure #1 Outer Annulus 0000 Floor Area, Walls and Ceiling Bldg Ground Level NE Page 4 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 3202 Containment Enclosure #2 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Ground Level NW 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3203 Containment Enclosure #3 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Ground Level SW 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3204 Containment Enclosure #4 Outer Annulus 0001 Floor Area, Walls and Ceiling - Section 1 Bldg Ground Level SE 0002 Floor Area, Walls and Ceiling - Section 2 Bldg 3205 Containment Enclosure Containment Foyer 0000 Floor Area and Walls Bldg Area Ground Level 3206 Containment Enclosure Containment Hatch 0000 Floor Area and Walls Bldg Area Ground Level 3211 Containment Enclosure Loop #1 Inner 0000 Floor Area and Walls Bldg Annulus Mid Ground NE 3212 Containment Enclosure Loop #2 Inner 0000 Floor Area and Walls Bldg Annulus Mid Ground NW 3213 Containment Enclosure Loop #3 Inner 0000 Floor Area and Walls Bldg Annulus Mid Ground SW 3214 Containment Enclosure Loop #4 Inner 0000 Floor Area and Walls Bldg Annulus Mid Ground SE 3301 Containment Enclosure #1 Outside Crane 0001 Floor Area and Containment Enclosure Bldg Charging Floor Wall up to el. 56' 6" - Section 1 0002 Floor Area and Containment Enclosure Bldg Wall up to el. 56' 6" - Section 2 3302 Containment Enclosure #2 Outside Crane 0001 Floor Area and Containment Enclosure Bldg Charging Floor Wall up to el. 56' 6" - Section 1 0002 Floor Area and Containment Enclosure Bldg Wall up to el. 56' 6" - Section 2 3303 Containment Enclosure #3 Outside Crane 0001 Floor Area and Containment Enclosure Bldg Charging Floor Wall up to el. 56' 6" - Section 1 0002 Floor Area and Containment Enclosure Bldg Wall up to el. 56' 6" - Section 2 Page 5 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 3304 Containment Enclosure #4 Outside Crane 0001 Floor Area and Containment Enclosure Bldg Charging Floor Wall up to el. 56' 6" - Section 1 0002 Floor Area and Containment Enclosure Bldg Wall up to el. 56' 6" - Section 2 3311 Containment Enclosure #1 Inside Crane 0000 Floor Area Bldg Charging Floor 3312 Containment Enclosure #2 Inside Crane 0000 Floor Area Bldg Charging Floor 3313 Containment Enclosure #3 Inside Crane 0000 Floor Area Bldg Charging Floor 3314 Containment Enclosure #4 Inside Crane 0000 Floor Area Bldg Charging Floor 3315 Containment Enclosure Removable Grating 0000 Floor Area Bldg for RX Head Staging 3320 Containment Enclosure CTMT Rx Refuel 0000 Floor and Walls Bldg Canal to Spent Fuel Pit 3322 Containment Enclosure CTMT Reactor 0000 Floor and Walls Bldg Refueling Cavity 3324 Containment Enclosure CTMT Reactor 0000 Wall Area and Supports Bldg Vessel Area 3326 Containment Enclosure Upper Core Package 0000 Floor Area Bldg Storage Area 3403 Containment Enclosure Inside Surfaces 0001 Dome - Quadrant 1 Bldg 0002 Dome - Quadrant 2 Bldg 0003 Dome - Quadrant 3 Bldg 0004 Dome - Quadrant 4 Bldg 0005 Shell (el. 56' 6" and up) - Section 1 Bldg 0006 Shell (el. 56' 6" and up) - Section 2 Bldg 0007 Shell (el. 56' 6" and up) - Section 3 Bldg 3502 Containment Enclosure Outside Surfaces 0001 Dome - Quadrant 1 Bldg 0002 Dome - Quadrant 2 Bldg 0003 Dome - Quadrant 3 Bldg 0004 Dome - Quadrant 4 Bldg Page 6 of 15

Attachment I to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 0005 Shell - Quadrant 1, East Section Bldg 0006 Shell - Quadrant 1, North Section Bldg 0007 Shell - Quadrant 2 Bldg 0008 Shell - Quadrant 3 Bldg 0009 Shell - Quadrant 4 Bldg 4102 Turbine Building North Floor Area 0000 Floor Area and Walls Bldg 4104 Turbine Building Oil Room, Heater Drains, 0000 Floor Area, Walls and Ceiling Bldg Emergency Power 4106 Turbine Building Air Compressor Area 0000 Floor Area, Walls and Ceiling Bldg 4108 Turbine Building Steam Generator Feed 0000 Floor Area, Walls and Ceiling Bldg Pump Area 4110 Turbine Building Chemistry/Closed Cooling 0000 Floor Area, Walls and Ceiling Bldg Water Area 4112 Turbine Building Water Treatment Area 0000 Floor Area, Walls and Ceiling Bldg 4114 Turbine Building Condenser Pump and South 0000 Floor Area, Walls and Ceiling Bldg Floor Area 4116 Turbine Building Hoist/Equipment Laydown 0000 Floor Area and Walls Bldg Area 4118 Turbine Building Condenser "A" Water Box 0000 Floor Area, Walls and Ceiling Bldg "A & B" Area 4120 Turbine Building Condenser "B" Water Box 0000 Floor Area, Walls and Ceiling Bldg "C & D" Area 4121 Turbine Building Secondary Chem Lab 0000 Floor Area, Walls and Ceiling Bldg 4202 Turbine Building North End Open Area 0000 Wall Area and Supports Bldg 4204 Turbine Building Oil Reservoir Area 0000 Structure Area and Walls Bldg 4206 Turbine Building S/G Feedwater Heater 2A 0000 Floor Area, Walls and Ceiling Bldg and 2B Area 4208 Turbine Building S/G Feedwater Heater 1A 0000 Floor Area, Walls and Ceiling Bldg and 1B Area 4210 Turbine Building Steam Generator 0000 Floor Area, Walls and Ceiling Bldg Feedwater Control Valve Area 4212 Turbine Building South End/Turbine Hall 0000 Wall Area and Supports Bldg Page 7 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 4216 Turbine Building S/G Feedwater Heater 6B 0000 Floor Area, Walls and Ceiling Bldg and 5B Area 4218 Turbine Building S/G Feedwater Heater 6A 0000 Floor Area, Walls and Ceiling Bldg and 5A Area 4302 Turbine Building 30" Main Steam Line Area 0000 Floor Area, Walls and Ceiling Bldg 4304 Turbine Building 24" Main Steam Line Area 0000 Floor Area, Walls and Ceiling Bldg 4306 Turbine Building MSRHR IA and lB Area 0000 Floor Area, Walls and Ceiling Bldg Reheater 4308 Turbine Building MSRHR IC and 1D Area 0000 Floor Area, Walls and Ceiling Bldg Reheater 4402 Turbine Building Laydown Area North Floor 0000 Floor Area and Walls Bldg 4404 Turbine Building Steam Generator 0000 Floor Area and Walls Bldg Feedwater Heater 3A Area 4406 Turbine Building Steam Generator 0000 Floor Area and Walls Bldg Feedwater Heater 4A Area 4408 Turbine Building Steam Generator 0000 Floor Area and Walls Bldg Feedwater Heater 3B Area 4410 Turbine Building Steam Generator 0000 Floor Area and Walls Bldg Feedwater Heater 4B Area 4412 Turbine Building H.P. Turbine Area 0000 Floor Area Bldg 4414 Turbine Building L.P. #1 Turbine Area 0000 Floor Area Bldg 4416 Turbine Building L.P. #2 Turbine Area 0000 Floor Area Bldg 4418 Turbine Building Generator Area 0000 Floor Area Bldg 4420 Turbine Building Exciter Area 0000 Floor Area Bldg 4422 Turbine Building Laydown Area South Floor 0000 Floor Area and Walls Bldg 4424 Turbine Building Open Hoist Area 0000 Wall Area Bldg 4502 Turbine Building Overhead Crane Area 0001 Ceiling Area - Section 1 Bldg 0002 Ceiling Area - Section 2 Bldg 0003 Ceiling Area - Section 3 Bldg 0004 Ceiling Area - Section 4 Bldg 4603 Turbine Building Roof Area 0001 Roof Area - Section 1 Bldg 0002 Roof Area - Section 2 Bldg Page 8 of 15

Attachment I to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 0003 Roof Area - Section 3 Bldg 0004 Roof Area - Section 4 Bldg 0005 Exterior Walls Bldg 5102 Service Building "A" Diesel Generator Area 0000 Floor Area, Walls and Ceiling Bldg 5104 Service Building "B" Diesel Generator Area 0000 Floor Area, Walls and Ceiling Bldg 5106 Service Building Clean Locker Room Area 0000 Floor Area, Walls, and Ceiling Bldg 5108 Service Building Hot Locker Room Area 0000 Floor Area, Walls, and Ceiling Bldg 5110 Service Building HP Control Point and 0000 Floor Area, Walls, and Ceiling Bldg Office Areas 5112 Service Building Woman's Locker Room 0000 Floor Area, Walls, and Ceiling Bldg Area 5114 Service Building Hot Chemistry Area 0001 Floor Area, Walls, and Ceiling - Section 1 Bldg 0002 Floor Area, Walls, and Ceiling - Section 2 Bldg 5118 Service Building Maintenance Decon Area 0000 Floor Area, Walls, and Ceiling Bldg 5120 Service Building Machine Shop Clean Area 0000 Floor Area, Walls, and Ceiling Bldg 5122 Service Building Machine Shop Hot Area 0001 Floor Area, Walls, and Ceiling - Section 1 Bldg 0002 Floor Area, Walls, and Ceiling - Section 2 Bldg 5124 Service Building Maintenance Clean Shop 0000 Floor Area, Walls, and Ceiling Bldg

,Area 5126 Service Building "A" Auxiliary Boiler Area 0000 Floor Area, Walls, and Ceiling Bldg 5128 Service Building "B" Auxiliary Boiler Area 0000 Floor Area, Walls, and Ceiling Bldg 5130 Service Building East Hallway 0000 Floor Area, Walls, and Ceiling Bldg 5132 Service Building Health Physics Facility 1st 0000 Floor Area, Walls, and Ceiling Bldg Floor 5134 Service Building Health Physics Facility 2nd 0000 Floor Area, Walls, and Ceiling Bldg Floor 5202 Service Building Switch Gear Area 0000 Floor Area, Walls, and Ceiling Bldg 5302 Service Building Control Room Area 0000 Floor Area, Walls, and Ceiling Bldg 5304 Service Building Computer, Operations, 0000 Floor Area, Walls, and Ceiling Bldg Security Area 5306 Service Building Machine and Equipment 0000 Floor Area, Walls, and Ceiling Bldg Area Page 9 of 15

Attachment I to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 5308 Service Building Instrument & Controls 0000 Floor Area, Walls, and Ceiling Bldg Shop 5402 Service Building Roof 0001 Roof Area - Section 1 Bldg 0002 Roof Area - Section 2 Bldg 0003 Roof Area - Section 3 Bldg 0004 Roof Area - Section 4 Bldg 0005 Exterior Walls Bldg 5502 CW System Trench 0001 Unit 1 Discharge Tunnel Surface Area Bldg 0002 Unit 2 Discharge Tunnel Surface Area Bldg 6002 Waste Disposal Building Hall Area Lower 0000 Floor Area, Walls and Ceiling Bldg Level 6004 Waste Disposal Building Area Outside 0000 Floor Area, Walls and Ceiling Bldg Reboiler Room 6006 Waste Disposal Building Bottoms Pump and 0000 Floor Area, Walls and Ceiling Bldg Reboiler Area 6008 Waste Disposal Building Sump Trench Area 0000 Floor Area, Walls and Ceiling Bldg Lower Level 6010 Waste Disposal Building-Waste Decay Tank 0000 Floor Area, Walls and Ceiling Bldg A, B, C Area 6012 Waste Disposal Building Surge Tank Area 0000 Floor Area, Walls and Ceiling Bldg Lower Level 6102 Waste Disposal Building Hall Area 0000 Floor Area, Walls and Ceiling Bldg 6202 Waste Disposal Building Hallway Area 0000 Floor Area, Walls and Ceiling Bldg 6304 Waste Disposal Building Evaporator Area 0000 Floor Area, Walls and Ceiling Bldg 6306 Waste Disposal Building Radwaste Liquid 0000 Floor Area, Walls and Ceiling Bldg Evaporator 6308 Waste Disposal Building Degassifier 0000 Floor Area, Walls and Ceiling Bldg Transfer Pump Area 6312 Waste Disposal Building Degassifier and 0000 Floor Area, Walls and Ceiling Bldg Associated Valves 6404 Waste Disposal Building Evaporator Area 0000 Floor Area, Walls and Ceiling Bldg 6406 Waste Disposal Building Liquid Evaporator 0000 Floor Area, Walls and Ceiling Bldg Area Page 10 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 6408 Waste Disposal-Waste Gas Compressor 0000 Floor Area, Walls and Ceiling Bldg A&B Area 6412 Waste Disposal Building Degassifier Area 0000 Floor Area, Walls and Ceiling Bldg and Associated Valves 6502 Waste Disposal Building Roof Area 0001 Roof Area - Section 1 Bldg 0002 Roof Area - Section 2 Bldg 0003 Exterior Walls Bldg 7002 CW Circ Pump A&B Head Area 0000 Floor Area, Walls and Ceiling Bldg 7004 CW Circ Pump C&D Head Area 0000 Floor Area, Walls and Ceiling Bldg 7102 CW Circ Pump Motor A&B Area 0000 Floor Area, Walls and Ceiling Bldg 7104 CW Circ Pump Motor C&D Area 0000 Floor Area, Walls and Ceiling Bldg 7106 CW Hypochloride Tank Area 0000 Floor Area, Walls and Ceiling Bldg 7108 CW Intake and Screen Area 0000 Floor Area Bldg 7202 CW Roof Area 0000 Roof Area and Exterior Walls Bldg 8100 FW/STM Penetration Building Upper Level 0000 Floor Area, Walls and Ceiling Bldg 8200 FW/STM Penetration Building Mid Level 0000 Floor Area, Walls and Ceiling Bldg 8300 FW/STM Penetration Building Lower Level 0000 Floor Area, Walls and Ceiling Bldg 9102 YD 115KV Switchyard Area 0001 Trench and Adjoining Land Area S.C. Soil 0002 Land Area S.C. Soil 9104 YD Main Transformer Area 0000 Land Area S.C. Soil 9106 Discharge Canal 0001 Land Area From Discharge Structure to Old S.C. Soil Security Guard Shack and Canal Road 0002 Land Area From Old Security Guard Shack S.C. Soil and Canal Road To the CT River 9108 YD North Tank Farm Area 0000 Land Area S.C. Soil 9110 YD South Tank Farm Area 0000 Land Area S.C. Soil 9112 YD Boron Storage Tank Area 0000 Land Area S.C. Soil 9114 YD Ion Exchange Area 0000 Land Area S.C. Soil 9116 YD Resin Slurry Area 0000 Land Area S.C. Soil 9118 YD Fuel Oil Tank Area 0000 Standing Structure and Land Area S.C.Soil 9120 YD Primary Vent Stack 0000 Structure and Land Area Bldg, S.C.

I I Soil Page 11 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 9122 YD Primary Water Storage Tank Area 0000 Land Area S.C. Soil 9124 YD Backup Primary Water Storage Tank 0000 Land Area S.C. Soil Area 9126 YD Large Yard Crane Area 0000 Land Area S.C. Soil 9128 YD Demin Water Storage Tank Area 0000 Land Area S.C. Soil 9202 Switchgear Building "B" 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9208 Administration Building 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9214 Shutdown Auxiliary Feed Pump House 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9226 Radwaste Reduction Facility 0001 Floor, Ceiling and Walls From the Eastern Bldg Entrance to a Point 15 Foot West 0002 Floor, Ceiling and Walls From Point 15 Bldg Foot West to a Point 30 Foot West 0003 Floor, Ceiling and Walls From Point 30 Bldg Foot West to a Point 45 Foot West 0004 Floor, Ceiling and Walls From Point 45 Bldg Foot West to West Wall 0005 Floor, Ceiling and Walls From the Bldg Southwestern Rollup Door to the Opposite Shield Wall 0006 Roof Bldg 0007 Exterior Walls Bldg 9227 Bus 10 Pad and Ground Underneath 0001 Busi 0 Pad and Ground Underneath Bldg, Soil 0002 Bus 13 Structure Bldg 9228 Unconditional Release Facility 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and External Walls 9234 HP Project Trailer 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Walls 9236 HP Count Module 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Walls 9302 Northwest Protected Area Grounds 0000 Land Area S.C. Soil Page 12 of 15

Attachment 1 to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 9304 Southwest Protected Area Grounds 0000 Land Area S.C. Soil 9306 South Central Protected Area Grounds 0000 Land Area S.C. Soil 9307 PAB / Service Building Alleyway 0000 Land Area S.C. Soil 9308 Southeast Protected Area Grounds 0000 Land Area S.C. Soil 9310 East Protected Area Grounds 0001 Land Area From the Spent Fuel Building to S.C. Soil the RadWaste Reduction Facility 0002 Land Area From the RadWaste Reduction S.C. Soil Facility to the East RCA Boundary 9312 Northeast Protected Area Grounds 0001 Land Area From the North RCA Gate to S.C. Soil Security Fence 0002 Land Area From Security Fence to Spent S.C. Soil Fuel Building 9313 Central Site Grounds 0000 Land Area S.C. Soil 9402 Emergency Operations Facility 0000 Structure Including Floor, Walls, Ceiling Bldg and Exterior Walls 9403 Emergency Operations Center Roof 0000 Structure Roof Bldg 9404 North Warehouse 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9406 South Warehouse 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9408 Miscellaneous Trailer Complex 0000 Structure Including Floor, Walls, Ceiling, Bldg I_ Roof and Exterior Surfaces 9410 Steam Generator Mockup Building 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9412 Training Stores Office Building 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9414 Warehouse #1 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9416 Warehouse #2 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces 9418 Office Building #3 and PAP 0000 Structure Including Floor, Walls, Ceiling, Bldg Roof and Exterior Surfaces Page 13 of 15

Attachment I to Enclosure 1 of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 9420 Office Trailer 0000 Structure Including Floor, Walls and Bldg Ceiling 9422 Information Center 0000 Structure Including Floor, Walls, Ceiling Bldg and Exterior Walls 9423 Information Center Roof 0000 Structure Roof Bldg 9424 All Buildings Contained in the Southwest 0000 Structure Including Floor, Walls, Ceiling, Bldg Site Storage Area Roof and Exterior Surfaces 9502 Northeast Site Grounds (Non-Protected 0000 Land Area S.C. Soil Area) 9504 Bypass Road / Secondary Parking Lot 0000 Land Area Soil 9506 North Site Grounds (Non-Protected Area) 0000 Land Area Soil 9508 Pond 0000 Land Area and Pond Sediment Soil 9510 Access Road 0000 Paved Road Soil 9512 Northwest site Grounds (Non-Protected 0000 Land Area S.C. Soil Area) 9514 Primary Parking Lot 0000 Paved Lot S.C. Soil 9518 Southwest Site Grounds (Non-Protected 0000 Land Area S.C. Soil Area) 9520 Southwest Site Storage Area 0001 Land Area From Security Fence to Load S.C. Soil Distribution Tower 0002 Land Area From Load Distribution Tower S.C. Soil East to 150m 0003 Land Area 150m East of Load Distribution S.C. Soil Tower to Gate 3 9521 Southeast Pond 0000 Land Area and Pond Sediment Soil 9522 Southeast Site Grounds (Non-Protected 0001 Land Area Above Transmission Towers S.C. Soil Area) 0002 Land Area Below Transmission Towers S.C. Soil 9523 Southeast Wetland Area 0000 Land Area Soil 9524 South Site Grounds (Non-Protected Area) 0000 Land Area Soil 9525 Southeast Site Road 0000 Paved Road Soil 9526 Northeast Mountain Side 0000 Land Area Soil Page 14 of 15

Attachment I to Enclosure I of CY-01-084 Survey Survey Area Survey Unit Survey Unit DCGL Area Code Description Code Code Description Type Code 9527 East Mountain Side 0001 Land Area From Contour Line 120' to Soil Upper Fence and Eastern Ridge Edge 0002 Land Area From Contour Line 120' to Soil Lower Fence and Eastern Ridge Edge 0003 Land Area From Ridge Edge to 9522A Soil Boundary 9528 Southeast Mountain Side 0000 Land Area Soil 9530 Central Peninsula Area 0001 Land Area Bounded by and Immediately Soil Adjacent to the Road 0002 Western Half of Diked Area and Immediate Soil Surrounding Sides 0003 Eastern Half of Diked Area and Immediate Soil Surrounding Sides 0004 Remaining Land Area Soil 9531 South End of Peninsula 0000 Land Area Soil 9532 East Site Grounds (Non-Protected Area) N/A N/A Soil 9535 South East Landfill Area 0000 Land Area Soil 9536 Construction Piles Near Rifle Range 0000 Land Area Soil 9537 Permitted Landfill Area 0000 Land Area Soil 9538 Material Storage Area 0000 Land Area Soil "Bldg" indicates DCGL developed for buildings (either in units of surface or volumetric contamination, depending upon the application)

"Soil" indicates DCGL developed for soils "S.C. Soil" indicates special case soil DCGL as described in LTP Section 6.4.6 Page 15 of 15

Attachment 2 to Enclosure I of CY-01-84 CD ROM with RESRAD and RESRAD-Build Runs, and HELP Files Response to RAIs 2-15 and 2-20

Attachment 2 to Enclosure 1 of CY-01-084 CD-ROM File Structure

.*5 *.SoiID4l W '

Item RESRAD Input File RESRAD Output File Comment Base all.RAD Base all.TXT Soil DCGLs input/output Resident Farmer Scenario (Base Case)

Groundwater CSH3Dose.RAD Cs-H3#1.TXT (Dose) Groundwater dose Radioactivity Cs-H3#2.TXT input/output (Special Case) (Concentration Report)

RESRAD 5.91 NA NA To run, double-click on Res591 .exe Item HELP Input File HELP Output File Comment - Base Case CYPRE.D4 CYOUT.OUT HELP Input/Output Files CYTEM.D7 CYOUT.DOC CYSOL.D13 CYEVE.D 11 CYSOI.D10 - CY1PRE.D4 CY1OUT.OUT HELP Input/Output Files Sensitivity Study I CY1TEM.D7 CYOUT2.DOC CYlSOL.D13 CYlEVE.Dl1 CY1SOI.D10 - CY2PRE.D4 CY2OUT.OUT HELP Input/Output Files Sensitivity Study II CY2TEM.D7 CY2OUT.DOC CY2SOL.D13 CY2EVE.D 11 CY2SOI.D 10 HELP NA NA To run, copy zhelp3w.exe to a local drive; double click on zhelp3w.exe; then double-click on install.exe Item Input File Output File Comment Building Occupancy RESRAD-BUILD input and output files begin with a Thirty radionuclides are Scenario numeral I followed by the name of the radionuclide. considered in CY Input files have an extension of.inp. Output files have Calculation 24265-000 an extension of. rpt. A summary text file is also MOC-9000-0001-000. The included, called L.out, which combines pertinent result LTP includes 20 of these sections (pathway detail of doses and receptor doses 30.

received for the exposure duration) for each of the individual output files.

Intruder/Excavator RESRAD-BUILD input and output files begin with the Thirty radionuclides are Scenario letters IN followed by the name of the radionuclide. considered in CY Input files have an extension of.inp. Output files have Calculation 24265-000 an extension of rpt. A summary text file is also MOC-9000-0001-000. The included, called IN.out, which combines pertinent result LTP includes 20 of these sections (pathway detail of doses and receptor doses 30.

received for the exposure duration) for each of the individual output files.

IN tý_

RESRAD-BUILD 2.37 INA Setup.exe.

__________________________________________________________________________________ I ________________________________________

Attachment 2 to Enclosure 1 of CY-01-084 Item RESRAD Input File RESRAD Output File Comment Con-inda.RAD Con-inda.TXT Concrete water independent dose Resident Farmer (Concrete Scenario I input/output Debris Scenario) Con-indb.RAD Con-indb.TXT Concrete water independent dose Water Independent Dose Scenario II input/output Scenario I,11,III Con-indc.RAD Con-indc.TXT Concrete water independent dose Scenario III input/output Conwat#1.RAD Cwaterla.txt (Dose) Concrete water dependent Cwaterlb.txt dose input/output #1 Resident Farmer (Concrete (Concentration Report)

Debris Scenario) Conwat#2.RAD Cwater2a.txt (Dose) Concrete water dependent Cwater2b.txt dose input/output #2 Water Dependent Dose (Concentration Report)

Conwat#3.RAD Cwater3a.txt (Dose) Concrete water dependent Cwater3b.txt dose input/output #3 (Concentration Report)

Conwat#4.RAD Cwater4a.txt (Dose) Concrete water dependent Cwater4b.txt dose input/output #4 (Concentration Report)

Conwat#5.RAD Cwater5a.txt (Dose) Concrete water dependent Cwater5b.txt dose input/output #5 (Concentration Report)

Resident Farmer (Concrete Conch#1.RAD Conch#1.TXT Concrete DCGL check#1 Debris Scenario) (water independent)

Conch#2.RAD Conch#2.TXT (Dose) Concrete DCGL check#2 Concrete DCGL Check Conch2a.TXT (water dependent)

(Concentration Report) I _I

Attachment 3 to Enclosure 1 of CY-01-84 Technical Support Document BCY-HP-018 "Concrete Building Surface Area and Volume Determination" Response to RAI 2-20

BCY-HP-0018, Rev 0 BECHTEL HEALTH PHYSICS DOCUMENT No.

TECHNICAL SUPPORT DOCUMENT 24265-000-G65-GEHH-P0O18-000 Connecticut Yankee Decommissioning Health Physics Department Technical Support Document HP Number: BCY-HP-0018 Revision #: 0

Subject:

Concrete Building Surface Area and Volume Determination Date: 453/2000 Date: 3AZ&

Date:

Approved By: W Date: 93O 6/ a Page 1 of 4

BCY-HP-0018, Rev 0

1. Purpose This document provides the basis for determining the concrete surface area and the available volume for concrete debris backfill to provide input into the Derived Concentration Guideline Levels (DCGL) calculations within the LTP (License Termination Plan).

2. Discussion The available volume for containment building debris backfill primarily consists of the full diameter of the building from 1'6" above meansea level to 3' below grade. The areas below the reactor vessel also provide volume for concrete debris. These areas extend to a depth of-19'6" below mean sea level but have a smaller cross section. The available containment building volume for concrete debris backfill was determined using construction: drawings and the PIB (Plant Information Book). See the references for a list of construction drawings used in this determination. The available volume for fill is provided in Table I below.

Table I - Available coatainuit building volume to be fdled by concente debris.

Description Volume ) Volume ()

Inner Containment 193,237 5,472 Cavity 3,491 99 Under the Reactor Vessel 7,726 219 Total Volum (mW): 5,790 The surface area ofthe containment building was determined using construction drawings and the PIB (Plant Information Book). See the references fbr a list of construction drawings used in this determination. Areas were assumed to be empty space with all interference and components removed. Only the inner smufaces ofthe containment building were included in this determination.

The surface area for the containment building is included in Table 2 below.

Page 2 of 4

BCY-HP-0018, Rev 0 Tahle 2 - Snrfe are for the cotainmet building Description Floor Area R OverheadArea Total Area Reactor Cavity, Transfer Canal and Under 142 498 18 658 Reactor Vessel Elevation 1'6" 1,021. 2,313 1,021 4,355 Elevation 22'0,'" 570 3,497 961 5,028 Elevation48'6" 1,088 2,975 2,660 6,723 Dme.

Total Area ra  : 16,764

3. Results and Conclusions The determination of concrete surface area and available volume for concrete debris: backfll has been performed to provide input into the DCGL calculations within the LTP. The resulting volume to surface ratio of the containment building is determined to be approximately 0.35.

Section 6 of the LTP describes the use of the volume to surface area ratio.

Initial :evaluations of the: Primary Auxiliary Building and the Fuel Handling Building indicate that the containment building will be the bounding analysis. Only the Residual Heat Removal (RHR) pit within the PAB extends below the water table. The RUR pit has an elevation of-I9' below mean sea level, and a cross section of 170 square meters. The available volume to fill to 3' below gradeis approximately 1750 cubic meters. The volume of concrete used as fill material is a small fraction of that of the containment building, approximately 3.5 times more concrete will be used to fill the containment volume than the PAB volume. For radioactivity levels associated with the DCGLs for concrete debris, the containment DCGLs will result in a substantially greater total dose and therefore bound the PAB.

The Fuel storage area is the only area within the fuel Handling Building that extends below the water table. The depth of the fuel storage pit is 13'6" above mean sea level. The pit has a cross section of 200 square meters. The available volume to fi to 3' below grade is approximately 90 cubic meters. No concrete fill will be placed below the water table. Therefore, the containment building volume will again be bounding

4. Attachments None Page 3 of 4

BCY-HP-0018, Rev 0

5. References 10899-FM-2E ARRGT - Primary Plan Section AA&BB 10899-FM-2F ARRGT - Primary Plan Section CC&DD 10899-FM-20D ARRGT - Neutron Shield and Neutron Instrumentation 16103-27090 SHEET 5 Decommissioning Work Areas - Containment El. 1'6" & Cable Vault El. 5'6" 16103-27090 Sheet 6 Decommissioning Work Areas - Containment El. 22'0" 16103-27090 Sheet 7 Decommissioning Work Areas - Containment El. 46'6" 16103-27090 Sheet 9 Decommissioning Work Areas - Spent Fuel Building 16103-27090 Sheet 10 Decommissioning Work Areas - Primary Auxiliary Building 16103-50082 Sheet 1 FDN Mat:Dets -El 0"61/4" SH. 1 Reactor Containment 16103-50085 Sheet I Hatch Access Dets -SH. 1 Reactor Containment 16103-50094 Interior Cone. Dets -El 1'6" Reactor Containment 16103-50095 Interior Conc. Dets -El 22'0" Reactor Containment 16103-50096 Interior Cone. Outline -El 48'6" Reactor Containment 16103-50098 Sheet 4 Interior Concrete. Details - Reactor Containment 16103-50098 Sheet 5 Interior Concrete. Details - Reactor Containment Plant Information Book Volume 1:Chapter 2.3 Steam Generators Plant Information Book Volume I Chapter 2.4 Pressurizer and Pressurizer Relief Plant Information Book Volume 3 Chapter 7 Containment Page 4 of 4

Enclosure 2 to CY-01-84 Matrix of LTP Sections Proposed for Revision

Enclosure 2 to CY-01-084 RAI Number Affected Section(s) of Summary of Proposed Change to LTP the LTP 1la Section 5.4.6.1 Clarify intent of use of building verus soil DCGLs lb Section 5.4.6.1 Table to indicate which DCGL will be used in which area 2 Section 5.4.6.2 Add the general process for determining the radionuclide mix--given in the response to RAI 51 (c).

3a Section 5.4.6.4 Add new area factor values to tables and Tables 5-7 and 5-8 change text to reflect new calculational method.

3b Section 5.4.6.4 Change text to reflect use of RESRAD BUILD 2.37 4 Various Revise text to clarify treatment of non structural components and embedded piping.

5 N/A N/A 6 N/A N/A 7 Section 6.4.2 Add clarification of use of only Pu-239 and Cm-243 of the Pu-239/240 pair and Cm 243/244 pair, respectively.

8a N/A N/A 8b Table 6-3 Correct references for soil density, hydraulic conductivity, total prorsity, and soil-specific exponential parameter.

8c N/A N/A 9a N/A N/A 9b Section 6.4.4 Add clarifying information why the well chosen is conservative.

10a N/A N/A 10b N/A N/A 11 N/A N/A 12 Section 6.4.6.2 Insert version of RESRAD Code used (v 5.91) 13 a/b N/A N/A 13c N/A N/A 14 N/A N/A 15 N/A N/A 16 Section 6.5.1 Add clarification that inhalation of tritium is included as an applicable pathway.

Page 1of 2

Enclosure 2 to CY-01-084 RAI Number Affected Section(s) of Summary of Proposed Change to LTP the LTP 17 Figure 6-6, footnote to Revise figure to show modified process for Table 6-20, Table 6-23, determing DCGLs, revise text to describe Table 6-24, accompanying the conversion from surface to volumetric text sources, revise tables to show new values.

18 N/A N/A 19 Section 6.5.2.1 Update section text and Tables 6-20, 6-23, and 6-24 to reflect new methodology and DCGLs associated with the use of Scenario

1. 3 for resident farmer (concrete debris).

20 N/A N/A 21 a N/A N/A 21b See RAI #17 See RAI #17 22 Section 6.5.2.2 Insert Version of RESRAD Build code used (v 2.37) 23 Various (including 6.5.2.2) Delete text and table references to the excavator scenario.

24 See RAI #23 See RAI #23 25 See RAI #23 See RAI #23 26a N/A N/A 26b N/A N/A 26c N/A N/A 26d N/A N/A 26e N/A N/A 26f N/A N/A 26g N/A N/A 26h N/A N/A 27a N/A N/A 27b N/A N/A Page 2of 2