ML20070L128

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Environmental Properties Management, LLC - Revised Evaluation of the Need to License Tc-99
ML20070L128
Person / Time
Site: 07000925
Issue date: 03/10/2020
From: Lux J
Environmental Properties Management
To: Davis P, Robert Evans, Kenneth Kalman
Document Control Desk, Office of Nuclear Material Safety and Safeguards, State of OK, Dept of Environmental Quality (DEQ)
References
Download: ML20070L128 (31)
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environmental properties management. LLC March 10, 2020 Mr. Ken Kalman U.S. Nuclear Regulatory Commission 115 5 5 Rockville Pike Rockville, MD 20852-2738 Mr. Paul Davis Oklahoma Department of Environmental Quality 707 North Robinson Oklahoma City, OK 73101 Mr. Robert Evans U.S. Nuclear Regulatory Commission 1600 East Lamar Blvd; Suite 400 Arlington, TX 76011-4511 Re: Docket No.70-925; License No. SNM-928 Revised Evaluation of the Need to License Tc-99

Dear Sirs:

Solely as Trustee for the Cimarron Environmental Response Trust (CERT), Environmental Properties Management LLC (EPM) submits herein a revision of the evaluation of the need to specifically list Tc-99 in license SNM-928. The original evaluation was submitted to the US Nuclear Regulatory Commission (NRC) and the Oklahoma Department of Environmental Quality (DEQ) on January 14, 2020.

Table 7.2 of the calculation memorandum listed the potential activity concentrations of various radionuclides in the biomass generated during biodenitrification. These concentrations were based on an assumption that both uranium and Tc-99 pass through the ion exchange system and are captured by the biomass. In a March 3, 2020 e-mail, the DEQ notified EPM that Table 7.2 identified those activity concentrations as "per gram of resin".

Enercon Services, Inc. (Enercon) conducted additional review of the calculation memo, and discovered that the same "per gram of resin" appeared in Table 7.3, which listed the potential activity concentrations of various radionuclides in the sediment that is filtered out of the influent prior to treatment. Enercon therefore revised the calculation memo, and it is attached hereto.

EPM maintains that the calculations provided in this paper, which employs extremely conservative assumptions, demonstrates that Tc-99 in groundwater and treatment media and wastes presents "no unique or significant radiation hazards to workers or the public".

Consequently, Tc-99 need not be specifically licensed.

9400 Word Parkway* Kansas {ity, MO 64114 Tel.* 40f-o42-flf2

  • jlux@envpm.com

~

environmental properties management. LLC Mr. Ken Kalman, et. al.

U.S. Nuclear Regulatory Commission March 10, 2020 Page2 EPM requests that the NRC review the attached calculations and provide written concurrence that Tc-99 need not be specifically licensed. If you have questions or comments, please contact me at 405-641-5152 or atjlux@envpm.com.

Sincerely, y~.

Jeff Lux, P.E.

Trustee Project Manager Attachment cc: Michael Broderick, Oklahoma Department of Environmental Quality (electronic copy only)

NRC Public Document Room (electronic copy only)

~

environmental properties management. LLC ATTACHMENT 1 LICENSING CONSIDERATIONS FOR THE PRESENCE OF Tc-99 REVISION 1

CALCNO. EPM027-CALC-001 0 ENERCON Exce1/e11a-f vrry pro}t'rl. Eve, y day.

CALCULATION COVER SHEET REV. 1 PAGE NO. 1 of 21 Cimarron Environmental Licensing Considerations for the Presence of Tc- Client:

Title:

Response Trust 99 Project Identifier: EPM027 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary D ~

infonnation, that require confirmation? (If YES, identify the assumptions.)

2 Does this calculation serve as an "Alternate Calculation"? (If YES, identify the D ~

design verified calculation.)

Design Verified Calculation No.

3 Does this calculation supersede an existing Calculation? (If YES, identify the ~ D design verified calculation.)

Superseded Calculation No. EPM027-CALC-OO 1, Revision 0 Scope of Revision:

Correct the unit designations for the uranium and progeny activity concentrations in the tables for Sections 7.2 and 7.3.

Revision Impact on Results:

No impact on calculated results Study Calculation D Final Calculation ~

Safety-Related LJ Non-Safety-Related ~

4-V ~ J (Print Name and Sign)

A I*

Originator: A. Joseph Nardi Date: 3/S/2020

. l /'

Design Reviewer: Jay Maisler

( - I Digitally signed by Jay Maisler, CHP Date:

Jay Maisler, CHP Date: 2020.03.06 10:06:43 -05'00' 6 March 2020 Date:

Ap:W.Joc~

bf e,/2o1c>

CALC NO. EPM027-CALC-001 jiENERCON CALCULATION Excellence- Every project Every doy. REV. 1 REVISION STATUS SHEET PAGE NO. 2 of 21 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 0 12/16/2019 Initial Issue 1 3/5/2020 Correction to Tables 7 .2 and 7 .3 PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION ll Corrections to Table in Section 7.2

.Ll. Corrections to Table in Section 7.3 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO.OF REVISION ATTACHMEN NO.OF REVISION PAGES NO. TNO. PAGES NO.

Attachment A 7 0

CALCNO. EPM027-CALC-001 TABLE OF CONTENTS JI ENERCON REV. 1 Excellence- Every project Every day.

PAGE NO. 3 of21 1.0 PURPOSE AND SCOPE 5 2.0

SUMMARY

OF RESULTS AND CONCLUSIONS 5

3.0 REFERENCES

5 4.0 ASSUMPTIONS 6 5.0 DESIGN INPUTS 7 6.0 METHODOLOGY 9 6.1 Approaches 9 6.2 Materials Considered 10 6.2.1 Spent Resin Material 10 6.2.2 Biomass Material 10 6.2.3 Filtered Solids Material 10 6.3 Radiation Exposure Pathways Considered 10 6.4 Comparison of Tc-99 and Uranium Pathways 10 7.0 CALCULATION OF Tc-99 DOSE CONTRIBUTION 11 7.1 Radionuclide concentrations in the Resin Material 11 7.2 Radionuclide concentrations in the Biomass Material 12 7.3 Radionuclide concentrations in the Filtered Solids Material 13 7.4 Summary of Calculated Radionuclide Activity Concentrations 13 7.5 Comparison of External Dose Rates for Resin Material 14 7.6 Comparison of External Dose Rates for Biomass Material 15 7.7 Comparison of External Dose Rates for Filtered Solids Material 16 7.8 Calculation of Percent Contribution to Dose for Inhalation Pathway 17

CALCNO. EPM027-CALC-001 TABLE OF CONTENTS JI ENERCON REV. 1 E cellence- Every project. Every day.

PAGE NO. 4 of21 7.9 Calculation of Percent Contribution to Dose for Oral Ingestion Pathway 18 7.10 Summary of Calculated Percent Contributions 19 7.11 Conclusions 19 8.0 CALCULATION OF CONSERVATIVE ANNUAL DOSE FOR Tc-99 19 8.1 General Conservatism in Analysis 20 8.2 External and Skin Dose Pathways 20 8.3 Inhalation and Oral Ingestion Dose Pathways 20 9.0 COMPARISON WITH OTHER EVALUATIONS 20 9.1 External Dose Rate for Resin Bed Columns 20 9.2 Calculation of Potential Intake 20 10.0 CONSIDERATION OF EXPOSURE TO THE PUBLIC FROM THE PRESENCE OF Tc-99 21 11.0 COMPUTER SOFTWARE 21 List of Attachments # of pages Calculation Preparation Checklist 7

CALCNO. EPM027-CALC-001

.:d ENERCON Licensing Considerations for the Presence of Tc-99 REV. 1 Excellence- Every project Every day PAGE NO. 5 of21

1.0 Purpose and Scope

The purpose of this calculation is to estimate the significance of Tc-99 to various routes ofradiation exposure associated with materials that will be present during the operation of the Groundwater Treatment Facility at the Cima1rnn Site. The methodology used is to show that the relative significance of the radiation dose received from Tc-99 exposure is insignificant. The potential for radiation exposure from Tc-99 is based on: 1) calculating the contribution to a radiation exposure pathway for various materials in comparison to the Uranium plus Progeny contribution and 2) a conservative estimate of the annual dose attributable to the Tc-99 present in that pathway. Contribution from the Tc-99 that is less than 10% of the total dose or a small fraction of the applicable dose limit can be considered insignificant.

2.0 Summary of Results and Conclusions In all cases evaluated, the contribution to the total radiation exposure from the Tc-99 that may be present is less than 1% of the total exposure, and the associated dose rate is insignificant in magnitude. It is therefore concluded that the presence of the Tc-99 on the site is not a significant contributor to the radiation exposure potential or the licensing considerations for the site in accordance with the guidance provided in Reference 3 .10, Section 3 .3.

3.0 References 3 .1 10 CFR 20, Appendix B, Annual Limits on Intake (ALis) and Derived Air Concentrations (DACs) ofRadionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage 3.2 EPA 402-R-93-081, External Exposure To Radionuclides In Air, Water, And Soil, Federal Guidance Report No. 12, September 1993 3 .3 Burns & McDonnell Memorandum No. BMCD-GWREMED-TM004, Impact of Sediment, Technetium-99 and Bioreactor Sludge on Waste Generated by Cimarron Remediation Water Treatment Systems, Rev. D, September 10, 2019 3.4 EPM028-CALC-001, Potential Intake Calculation, Rev. 2 3.5 DOE-STD-1136-2004, Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities", December 2004 3.6 "A Review And Verification Of The Isotopic Distribution Of Enriched Uranium And The Impact On Decommissioning Considerations", A. J. Nardi, Tracy Chance and John F. Conant, Presented at the HPS 2007 Midyear Topical Meeting, Jan. 21-24, 2007.

3.7 EPMOl 7-CALC-001, Dose Rate near Uranium Treatment Train, Rev 0, Dec. 21, 2015 3.8 Spreadsheet "VNS-EPM-004-CALC-D-001 RB-Mass Balance Excel.xlsx", Sheet "WATF Consumable Usage", Cell K7

CALCNO. EPM027-CALC-001

,. I ENERCON Licensing Considerations for the REV. 1 Excellence- Every project. Every day. Presence of Tc-99 PAGE NO. 6 of21 3 .9 The Health Physics and Radiological Health Handbook, Revised Edition edited by B. Shleien; 1992 3.10 NUREG-1757, Vol. 2 Revision 1, "Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria", September 2006 4.0 Assumptions

1) Tc-99 is present in groundwater only in the Western Alluvial Area (WAA). The U-235 enrichment of uranium in groundwater in this area was calculated to be 2.6% by weight (at a 95%

confidence level).

2) The specific activity and isotopic distribution for 2.6% enriched uranium: (Reference 3.6) a) Specific Activity of2.6% enrichment- l.41E-06 Ci/g b) Isotopic Activity Distribution
  • U-234- 72.8%
3) The enriched uranium is considered to have returned to secular equilibrium with the progeny appropriate for chemically separated uranium. The following radionuclides were included in the analysis:

j..--T-c--9-9~-U---2-34~~U--2-3-5~-T-h--2-3-l~-U--2-3-8~-T-h--2-3-4~P-a--2-3_4_m~

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CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day. Presence of Tc-99 PAGE NO. 7 of21 5.0 Design Inputs

1) The ALI used to compare the inhalation and oral ingestion pathways for internal dose were based on Reference 3 .1. For this evaluation the most conservative ALI was utilized (Stochastic or Non-stochastic). These values are:

Inhalation Oral Ingestion Annual Limit of Annual Limit of Radionuclide Intake (ALI) Intake (ALI)

(See Note 1) (See Note 1)

(microCi) (microCi)

Tc-99 7.E+02 4.E+03 U-234 4.E-02 l.E+Ol U-235 4.E-02 l.E+Ol Th-231 6.E+03 4.E+03 U-238 4.E-02 l.E+Ol Th-234 2.E+02 3.E+02 Pa-234m 7.E+03 2.E+03 Note 1: 10 CFR 20, Appendix B, Annual Limits on Intake (ALis) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage. This reference only provides a listing for Pa-234 which has been used for the values for Pa-234m.

(The rest of this page was left blank intentionally.)

CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day. Presence of Tc-99 PAGE NO. 8 of21

2) The Dose Conversion Factors used to compare the Effective and Skin Dose rates were based on Reference 3 .2. These values are:

Dose Conversion Dose Conversion Coefficient Coefficient Radionuclide (See Note 1) (See Note 2) 3 Sv/(Bq-s-m- ) (mrem/y)/(µCi/g)

Effective Dose Tc-99 6.72E-22 l.26E+02 U-234 2.15E-21 4.02E+02 U-235 3.86E-18 7.21E+05 Th-231 1.95E-19 3.64E+04 U-238 5.52E-22 l.03E+02 Th-234 l.29E-19 2.41E+04 Pa-234m 4.80E-19 8.97E+04 Skin Dose Tc-99 9.09E-22 1.70E+02 U-234 5.99E-21 l.12E+03 U-235 4.40E-18 8.22E+05 Th-231 2.56E-19 4.78E+04 U-238 3.55E-21 6.63E+02 Th-234 1.50E-19 2.80E+04 Pa-234m 8.27E-18 l.54E+06 Note 1: Dose Conversion Coefficients - Exposure to soil contaminated to an infinite depth (FGR Report Number 12, Table III. 7)-(Ref.3.2)

Note 2: The SI value is multiplied by 1.868E23 to obtain these values.

(The rest of this page was left blank intentionally.)

CALCNO. EPM027-CALC-001 JIENERCON Licensing Considerations for the REV. 1 Excellence- Every project Eve1y day. Presence ofTc-99 PAGE NO. 9 of21 6.0 Methodology 6.1 Approaches Two approaches are considered in this evaluation.

1) The first approach is to consider the percent contribution to dose of the Tc-99 in comparison to the percent contribution of the uranium (including its progeny). This analysis demonstrates that the Tc-99 is insignificant in comparison to the uranium component. These calculations are based on the composition of three materials considered in the analysis. The analysis does not consider a specific exposure scenario that incorporates realistic geometry or release fractions and therefore is not intended to estimate the absolute dose rate associated with the materials considered. The results are presented in te1ms of the percent contribution of the Tc-99 in comparison to the combined contribution of all the above-mentioned nuclides for four occupational exposures; external exposure to the whole body, external exposure to the skin, inhalation and ingestion.

These results are presented in Section 7.0.

2) The second approach translates the results of the first approach to dose rate values in a conservative manner. For the effective dose and the skin dose, the analysis from the first approach provides conservative annual doses for the three materials. Those annual doses are not based on realistic exposure conditions because they conservatively represent the annual dose to an individual above an unshielded infinite plane of the material which is modeled as contaminated soil rather than the actual material being considered in the stored or handled condition.

For the inhalation and oral ingestion dose calculations, the assumption is made that the individual has had an annual intake of one ALI (equivalent to an occupational dose of 5,000 mrem) of the resin mixture. This is a bounding analysis without regard to consideration of how that intake might have occurred. Thus, this approach represents a bounding condition for the occupational annual dose from the Tc-99. These results are presented in Section 8.0.

For comparison, two other evaluations are discussed that are based on more realistic models for radiation exposure. One model is the dose rate near a loaded resin column (Ref. 3. 7) and the second is the potential intake for the quantity of a radionuclide handled in a year (Ref. 3 .4). These comparisons are presented in Section 9.0.

Considerations regarding the dose to the public is discussed in Section 10.0.

CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Evety day. Presence of Tc-99 PAGE NO. 10 of21 6.2 Materials Considered Three materials were considered for this evaluation:

1. Spent resin material
2. Biomass material
3. Filtered solids material 6.2.1 Spent Resin Material The isotopic concentration of this material is based on:
  • All the Tc-99 in the groundwater that passes through the column is captured by the resin material,
  • The uranium loaded on the resin material is based on the uranium loading for the initial resin column for treatment of groundwater from the WATF area. (Reference 3.8) 6.2.2 Biomass Material The isotopic concentration of this material is based on:
  • None of the Tc-99 is captured by the resin and all of it is captured by the biomass material,
  • The uranium concentration of this material was taken from Reference 3 .3.

6.2.3 Filtered Solids Material The isotopic concentration of the Tc-99 and the uranium in this material is based on Reference 3.3.

6.3 Radiation Exposure Pathways Considered The following exposure pathways are considered to compare the contribution of the Tc-99 and Uranium for that pathway:

  • External exposure to the whole body
  • External exposure to the skin
  • Inhalation pathway
  • Oral Ingestion pathway 6.4 Comparison of Tc-99 and Uranium Pathways To compare the effective dose rate and skin dose rate for each material, the dose rate for each radionuclide was calculated (Dose Conversion Coefficient X Radionuclide Concentration). The%

contribution to the total dose rate for each radionuclide was then calculated.

To compare the inhalation and oral ingestion pathways for each material, the ratio of the concentration to the ALI was calculated for each material. This is a relative value of the importance of each radionuclide to the Sum-of-Fractions of the total ALI and is a direct measure of the

CALCNO. EPM027-CALC-001

.)J ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day. Presence of Tc-99 PAGE NO. 11 of21 importance of that radionuclide for that pathway for that material. Radionuclides that provide a total contribution of less than 10% to the total Sum-of-Fractions are considered insignificant.

7.0 Calculation of Tc-99 Dose Contribution 7.1 Radionuclide concentrations in the Resin Material l-------il-'---------1 Initial (and maximum) Tc-99 Groundwater concentration (Ref. 3.3, Sec 2.2.2) 1-------i--=-----1Mas~ of Spe~ r~ in in Colu!Pn (!?~_B-! v. l , S~ction 1_3._l. l) _____

Flow rate to Column (Ref. 3.3, Table 1) (See Note 1) i---49_2___ 05--,......L-/m-in---F lo~ i:_ate to Colu-;;m- - [ - -- - T -- -

129600 min Time until resin column is spent (Based on 90 day cycle for column)

.-97_E_+_l_O_p_C_i_ _ _ 'I~t;l gCi Tc-99 ciptur~d by colum~ - ,_ 1 3.96E+04 pCi/g Tc-99 activity per gram of resin I 1 Uranium I I I I J 3.09E-04 Ci U-235 U-235 activity in starting column for WATF resin (Ref. 3.8) 412 pCi/g U-235 ~~ncentrati on ofU-235 in WA_T_F re;~ - - J _-_- -

1---------1~---=---- - - -- --- -- --- --

2.6% ercent Weight percent ofU-235 in Uranium (Ref. 3.3, Sec 4.1) 72.8% percent Acti\;ity % ofU-_ii4 _!11 _!Jranium (Ref 3.~:1 J_

4.0% percent Activity% of U-235 in Uranium (Ref. 3:.6) 1 23.2% percent Activity% of U-238 in Uranium (Ref. 3.6) I 7.50E+03 4.12E+02 4.12E+02 Ci/g Ci/g Ci/g U-234 activity per grc!m of resin U-235 activity per gram of resin Th-231 activity per gram ofresin tI I 1

2.39E+03 pCi/g U-238 activity per graII! of resin 1 2.39E+03 pCi/g Th-234 a_ctivity per gram of resin I J 2.39E+03 pCi/g Pa-234 activity per gram of resin\ I J Note 1 : The Table list 250 gpm as the flow from the WATF field. Since this flow feeds two treatment j

!t_!"a!!l~~a value 9f 130 gpr:!! was used as the fee9 to one train._

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CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1

[xcellence - Every project Eve1y day. Presence of Tc-99 PAGE NO. 12 of 21 7.2 Radionuclide concentrations in the Biomass Material Tc Assuming all Tc-99 is captured by Biomass and none by Resin 466 Ci/L !_aj!ial (and ma~um) Tc-~~ Groundwater cog_ceng*ation (Ref. 3.3, Sec 2.2.2

- - - - - - - - - - - - - - :t\_::!ass of Spent Biomass at 150 ppm N!!!:ate (Ref. 303, Table 7)

Flow rate from both columns Column (Ref. 3.3, Table 1)

- - - - - - - - - F l o w rate to Colu~ l I Time until resin column is spent (Based on 90 day cycle for column)

~-----+-------1 Total pCiJ'c-99 passed by columnf __ I _

Tc-99 activ* er am of Biomass Uranium -Assuming Uranium Concentration in Biomass per BMCD-GWREMED-TM004, Table 7@ 150 ppm Nitrate (Ret: 3.3) 0.113 Ci/ Uranium concentration in Biomass (Ref. 3.3, Table 7) 2.6% Weight percent of U-235 in Uranium (Ref. 3.3, Sec 4.1) 72.8% Activity% of U-234 in Uranium (Ref. 3.6) I 4.0%

23.2%

8.23E-02 4.52E-03 4.52E-03 Activity% of U-235 in Uranium (Ref. 3.6)

Activity % of U-238 in Uranium (R~f. 3.6)

U-~34 activity per gra_!ll-of biomass material U-235 activity per gram of biomass mat~ ial Th-231 activity per gram of biomass _material

+ *t1*--~GC--

I

_ _2_.6_2E_-_0_2______ U-238 ~c!_ivity per gram of biomass materiaj_ r

_ _2_.6_2E_-_0_2______ _Th-234 activity per gr~m of biomass milterial 1 I 2.62E-02 Pa-234 activi er *am of biomass material I (The rest of this page was left blank intentionally.)

CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day Presence of Tc-99 PAGE NO. 13 of 21 7.3 Radionuclide concentrations in the Filtered Solids Material Tc Filtered Solids From WATF per BMCD-GWREMED-TM004, Table 3 0.05 Ci/ Initial and Maximum Tc-99 in Filtered Solids ef. 3.3, Section 2.2.2 Uranium - Filtered Solids from WATF per BMCD-GWREMED-TM004, Table 2 2.0 Ci/ Uranium c~ncentration in Filtered Solids {~ef.3.3, Sec. 2.2.1) 2.6% Weight percent of U-235 in Uranium (Re_f. 3.3, Sec 4.1) .

72.8% AEtivity % of U-234 in Uranium (Ref. 3.6) 4.0% Activity% of {-!-~ 5 in Uranium (Ref. 3.~

23.2% Activity% of U-238 in Uranium (Ref. l§)

1.46E+o0 U-234 activity per*gram of filtered solids 8.00E-02 U-235 activity per gram of filtered soligs 8.00E-02 Th-231 activity per gram of filtered solids 4.64E-01 U-238 activity per_gr~m of ft!_tered solids 4.64E-01 Th~234 activity per gra~ of fgtered ~olids 4.64E-01 Pa-234 activi er am of filtered solids 7.4 Summary of Calculated Radionuclide Activity Concentrations Activity Concentrations for Materials Resin Filtered Biomass Radionuclide Activity Solids pCi/g pCi/g pCi/g Tc-99 3.96E+04 l.34E-01 5.00E-02 U-234 7.50E+03 8.23E-02 1.46E+OO U-235 4.12E+02 4.52E-03 8.00E-02 Th-231 4.12E+02 4.52E-03 8.00E-02 U-238 2.39E+03 2.62E-02 4.64E-01 Th-234 2.39E+03 2.62E-02 4.64E-01 Pa-234m 2.39E+03 2.62E-02 4.64E-01

CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence - Every projecc. Eve1y day. Presence of Tc-99 PAGE NO. 14 of 21 7.5 Comparison of External Dose Rates for Resin Material r------ - --****- -- ------- -- ---- - ---

External Dose Rate Calculations for Resin Material I Effective Dose from Resin Material Percent Dose Radionuclide Percent of Distribution Contribution Radionuclide Conversion Concentration Contribution of Activity to Dose Rate Coefficient in Resin to Dose Rate in Resin (mrem/y)/

pCi/g  % mrem/yr  %

(µCi/g)

Tc-99 l.26E+02 3.96E+04 71.9% 4.97E+OO 0.84%

U-234 4.02E+02 7.50E+03 13.6% 3.0lE+OO 0.51%

U-235 7.21E+05 4.12E+02 0.7% 2.97E+02 50.2%

Th-231 3.64E+04 4.12E+02 0.7% l.50E+Ol 2.5%

U-238 l.03E+02 2.39E+03 4.3% 2.46E-Ol 0.042%

Th-234 2.41E+04 2.39E+03 4.3% 5.76E+Ol 9.7%

Pa-234m 8.97E+04 2.39E+03 4.3% 2.14E+02 36.2%

Total 5.51E+04 100.0% 5.92E+02 100.0%

Contribution to Effective Dose Rate from Uranium Isotopes plus Progeny 99.16%

I Contribution to Effective Dose Rate from T c-99 0.84%

Skin Dose from Resin Material Percent Dose Radionuclide Percent of Distribution Contribution Radionuclide Conversion Concentration Contribution of Activity to Dose Rate Coefficient in Resin to Dose Rate in Resin (mrem/y)/

pCi/g  % mrem/yr  %

(µCi/g)

Tc-99 l.70E+02 3.96E+04 71.9% 6.73E+OO 0.163%

U-234 l.12E+03 7.50E+03 13.6% 8.39E+OO 0.20%

U-235 8.22E+05 4.12E+02 0.7% 3.39E+02 8.2%

Th-231 4.78E+04 4.12E+02 0.7% l.97E+Ol 0.48%

U-238 6.63E+02 2.39E+03 4.3% l.58E+OO 0.038%

Th-234 2.80E+04 2.39E+03 4.3% 6.70E+Ol 1.6%

Pa-234m l.54E+06 2.39E+03 4.3% 3.69E+03 89.3%

Total 5.51E+04 100.0% 4.13E+03 100.0%

Contribution to Skin Dose Rate from Uranium Isotopes plus Progeny 99.84%

I Contribution to Skin Dose Rate from T c-99 0.163%

JNote 1: Dose Conversion Coefficients - Exposure to soil contaminated to an infinite depth (FOR Report Number 12, Table III. 7)-(Ref.3.2). See Section 5.0

CALCNO. EPM027-CALC-001

,. I ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day. Presence of Tc-99 PAGE NO. 15 of21 7.6 Comparison of External Dose Rates for Biomass Material r------* -External 1

. Dose Rate Calculations for Biomass Material Effective Dose from Biomass Material Percent Dose Radionuclide Percent of Distribution Contribution Radionuclide Conversion Concentration Contribution of Activity to Dose Rate Coefficient in Biomass to Dose Rate in Biomass (mrem/y)/

pCi/g  % mrem/yr  %

(µCi/g)

Tc-99 1.26E+02 l.34E-01 44.1% 1.68E-05 0.26%

U-234 4.02E+02 8.23E-02 27.0% 3.30E-05 0.51%

U-235 7.21E+05 4.52E-03 1.5% 3.26E-03 50.5%

Th-231 3.64E+04 4.52E-03 1.5% 1.65E-04 2.5%

U-238 1.03E+02 2.62E-02 8.6% 2.70E-06 0.042%

Th-234 2.41E+04 2.62E-02 8.6% 6.32E-04 9.8%

Pa-234m 8.97E+04 2.62E-02 8.6% 2.35E-03 36.4%

Total 3.04E-01 100.0% 6.46E-03 100.0%

Contribution to Effective Dose Rate from Uranium Isotopes plus Progeny 99.7%

I Contribution to Effective Dose Rate from Tc-99 0.26%

Skin Dose from Biomass Material Percent Dose Radionuclide Percent of Distribution Contribution Radionuclide Conversion Concentration Contribution of Activity to Dose Rate Coefficient in Biomass to Dose Rate in Biomass (mrem/y)/

pCi/g  % mrem/yr  %

(µCi/g)

Tc-99 l.70E+02 l.34E-01 44.1% 2.28E-05 *o.oso%

U-234 l.12E+03 8.23E-02 27.0% 9.20E-05 0.20%

U-235 8.22E+05 4.52E-03 1.5% 3.72E-03 8.2%

Th-231 4.78E+04 4.52E-03 1.5% 2.16E-04 0.5%

U-238 6.63E+02 2.62E-02 8.6% 1. 74E-05 0.038%

Th-234 2.80E+04 2.62E-02 8.6% 7.35E-04 1.6%

Pa-234m l.54E+06 2.62E-02 8.6% 4.05E-02 89.4%

Total 3.04E-Ol 100.0% 4.53E-02 100.0%

Contribution to Skin Dose Rate from Uranium Isotopes plus Progeny 99.95%

I Contribution to Skin Dose Rate from T c-99 0.050%

CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Eve,y p1oject Eve,y day. Presence of Tc-99 PAGE NO. 16 of 21 7.7 Comparison of External Dose Rates for Filtered Solids Material Effective Dose from Filtered Solids Material Percent Radionuclide Dose Distribution Percent of Concentration Contribution Radionuclide Conversion of Activity Contribution in Filtered to Dose Rate Coefficient in Filtered to Dose Rate Solids Solids (mrem/y)/

pCi/g  % mrem/yr  %

(µCi/g)

Tc-99 l.26E+02 5.00E-02 1.6% 6.28E-06 0.006%

U-234 4.02E+02 1.46E+OO 47.6% 5.85E-04 0.5%

U-235 7.21E+05 8.00E-02 2.6% 5.77E-02 50.6%

Th-231 3.64E+04 8.00E-02 2.6% 2.91E-03 2.6%

U-238 l .03E+02 4.64E-Ol 15.2% 4.78E-05 0.042%

Th-234 2.41E+04 4.64E-01 15.2% l .12E-02 9.8%

Pa-234m 8.97E+04 4.64E-01 15.2% 4.16E-02 36.5%

Total 3.06E+OO 100.0% l.14E-01 100.0%

Contribution to Effective Dose Rate from Uranium Isotopes plus Progeny 99.994%

I Contribution to Effective Dose Rate from T c-99 0.006%

Skin Dose from Filtered Solids Material Percent Radionuclide Dose Distribution Percent of Concentration Contribution Radionuclide Conversion of Activity Contribution in Filtered to Dose Rate Coefficient in Filtered to Dose Rate Solids Solids (mrem/y)/

pCi/g  % mrem/yr  %

(µCi/g)

Tc-99 l.70E+02 5.00E-02 1.6% 8.49E-06 0.001%

U-234 l.12E+03 l.46E+OO 47.6% l.63E-03 0.2%

U-235 8.22E+05 8.00E-02 2.6% 6.58E-02 8.2%

Th-231 4.78E+04 8.00E-02 2.6% 3.83E-03 0.5%

U-238 6.63E+02 4.64E-01 15.2% 3.08E-04 0.038%

Th-234 2.80E+04 4.64E-01 15.2% 1.30E-02 1.6%

Pa-234m 1.54E+06 4.64E-01 15.2% 7.17E-01 89.5%

Total 3.06E+OO 100.0% 8.0lE-01 100.0%

Contribution to Skin Dose Rate from Uranium Isotopes plus Progeny 99.999%

I Contribution to Skin Dose Rate from T c-99 0.001%

Note 1: Dose Conversion Coefficients - Exposure to soil contaminated to an infinite

!depth (FGR Report Nlli!lbe_! 12, Table III. 7)-(Ref.3 ) ). ~ee Section 5.0

CALCNO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Exe 1/ence- Every projec Eve,y day. Presence of Tc-99 PAGE NO. 17 of21 7.8 Calculation of Percent Contribution to Dose for Inhalation Pathway Contribution to Dose Calculation for Airborne Activity of Resin Material Annual Limit of Percent Radionuclide Intake (ALI) Activity Contribution to (See Note I) Concentration Relative Ratio Dose (microCi) (microCi/g) (Ratio) (Percent)

(a) (b) (c) (d) = (c)/(b) (e)=[(d)ffotal ]* 100 Tc-99 7E+02 3.96E-02 5.66E-05 0.022%

U-234 4E-02 7.50E-03 l.87E-Ol 72.78%

U-235 4E-02 4.12E-04 l.03E-02 4.00%

Th-231 6E+03 4.12E-04 6.87E-08 0.00003%

U-238 4E-02 2.39E-03 5.97E-02 23 .19%

111-234 2E+02 2.39E-03 l.19E-05 0.005%

Pa-234m 7E+03 2.39E-03 3.41E-07 0.0001%

Total 2.58E-O I 100%

Contribution to Airborne Dose from Uranium Isotopes+ Progeny 99.98%

I Contribution to Airborne Dose from Tc-99 0.02%

Contribution to Dose Calculation for Airborne Activity of Biomass Material Annual Limit of Percent Radionuclide Intake (ALI) Activity Contribution to (See Note I) Concentration Relative Ratio Dose (microCi) (microCi/g) (Ratio) (Percent)

(a) (b) (c) (d) = (c)/(b) (e)=ffd)ffotall* 100 Tc-99 7E+02 l.34E-07 l.92E-10 0.007%

U-234 4E-02 8.23E-08 2.06E-06 72.79%

U-235 4E-02 4.52E-09 l.13E-07 4.00%

111-231 6E+03 4.52E-09 7.53E-13 0.00003%

U-238 4E-02 2.62E-08 6.55E-07 23 .20%

111-234 2E+02 2.62E-08 l.31E-10 0.005%

Pa-234111 7E+03 2.62E-08 3.75E- 12 0.0001%

Total 2.83E-06 100%

Contribution to Airborne Dose from Uranium Isotopes + Progeny 99.993%

I Contribution to Airborne Dose from Tc-99 0.007%

Sum ofFractions Calculation for Airborne Activity of Filtered Solids Material Annual Limit of Percent Radionuclide Intake (ALI) Activity Contribution to (See Note I) Concentration Relative Ratio Dose (microCi) (microCi/g) (Ratio) (Percent)

(a) (b) (c) (d) = (c)/(b) (e)=[(d)ffotal]* 100 Tc-99 7E+02 5.00E-08 7.14E-l l 0.0001%

U-234 4E-02 l.46E-06 3.64E-05 72.80%

U-235 4E-02 8.00E-08 2.00E-06 4.00%

111-231 6E+03 8.00E-08 l.33E-l l 0.00003%

U-238 4E-02 4.64E-07 l.16E-05 23.20%

111-234 2E+02 4.64E-07 2.32E-09 0.005%

Pa-234111 7E+03 4.64E-07 6.63E-l l 0.0001%

Total 5.00E-05 100%

Contribution to Airborne Dose from Uranium Isotopes+ Progeny 99.9999%

I Contribution to Airborne Dose from Tc-99 Note I : 10 CFR 20, Appendix B, Annual Limits on Intake (ALis) and Derived Air Concentrations 0.0001%

I(DACs.) ofRadionuclides for Occupational Exposure;. Effluent Concentrations; Concentrations for I

IR~le~ e t~~we ~a_ge_(~ef. 3: 1) _ _ _ ____ _ _

CALCNO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day. Presence of Tc-99 PAGE NO. 18 of21 7.9 Calculation of Percent Contribution to Dose for Oral Ingestion Pathway Contiibution to Dose Calculation for Oral Ingestion of Resin Material Annual Limit of Percent Radionuclide Intake (ALI) Activity Contribution to (See Note 1) Concentration Relative Ratio Dose (microCi) (microCi/g) (Ratio) (Percent)

(a) (b) (c) (d) = (c)/(b) r(d)/fotal]* 100 Tc-99 4E+o3 3.96E-02 9.91E-06 0.94%

U-234 lE+ol 7.SOE-03 7.SOE-04 71.47%

U-235 lE+ol 4.12E-04 4.12E-05 3.93%

Th-231 4E+o3 4.12E-04 l .03E-07 0.0098%

U-238 lE+ol 2.39E-03 2.39E-04 22.78%

Th-234 3E+o2 2.39E-03 7.97E-06 0.76%

Pa-234m 2E+o3 2.39E-03 l . 19E-06 0.114%

Total l.OSE-03 100%

Contribution to Oral Ingestion Dose from Uranium Isotopes+ Progeny 99. 1%

I Contribution to Oral Ingestion Dose from Tc-99 0.94%

Contribution to Dose Calculation for Oral Ingestion of Biomass Material Annual Limit of Percent Radionuclide Intake (ALI) Activity Contribution to (See Note 1) Concentration Relative Ratio Dose (microCi) (microCi/g) (Ratio) (Percent)

(a) (b) (c) (d) = (c)/(b) [(d)/fotal]* 100 Tc-99 4E+o3 l.34E-07 3.35E-l l 0.29%

U-234 lE+ol 8.23E-08 8.23E-09 71.94%

U-235 lE+ol 4.52E-09 4.52E-10 3.95%

Th-231 4E+o3 4.52E-09 l.13E-12 0.00988%

U-238 lE+ol 2.62E-08 2.62E-09 22.93%

Th-234 3E+o2 2.62E-08 8.74E-l l 0.76%

Pa-234m 2E+o3 2.62E-08 l.3 lE-11 0. 1146%

Total 1.14E-08 100%

Contribution to Oral Ingestion Dose from Uranium Isotopes+ Progeny 99.7%

I Contribution to Oral Ingestion Dose from Tc-99 0.29%

Contribution to Dose Calculation for Oral Ingestion of Filtered Solids Material Annual Limit of Percent Radionuclide Intake (ALI) Activity Contribution to (See Note 1) Concentration Relative Ratio Dose (microCi) (microCi/g) (Ratio) (Percent)

(a) (b) (c) (d) = (c)/(b) [(d )If otal]

  • 100 Tc-99 4E+o3 5.00E-08 l .25E-l l 0.006%

U-234 lE+ol l.46E-06 l.46E-07 72. 15%

U-235 lE+ol 8.00E-08 8.00E-09 3.96%

Th-231 4E+o3 8.00E-08 2.00E-11 0.0099%

U-238 lE+ol 4.64E-07 4.64E-08 22.99%

Th-234 3E+o2 4.64E-07 l .55E-09 0.77%

Pa-234m 2E+o3 4.64E-07 2.32E-10 0. 115%

Total 2.02E-07 100%

Contribution to Oral Ingestion Dose from Uranium Isotopes+ Progeny 99.994%

I Contribution to Oral Ingestion Dose from Tc-99 0.006%

Note 1: 10 CFR 20, Appendix B, Annual Limits on Intake (ALis) and.Derived Air IConcentrations (DACs) of Radionuclides fo, Occupational Exposuce; Effluent Conc_entration~ Co_!lcentratio1'.s f<_?r Releas~~ S~wer~ge (Ref. 3: 1) _._ .

CALCNO. EPM027-CALC-001

,. I ENERCON Licensing Considerations for the REV. 1 Excellence - Every project Eve1y day. Presence of Tc-99 PAGE NO. 19 of 21 7 .10 Summary of Calculated Percent Contributions Radiation Percent Contribution Exposure Material Uranium+

Pathway Tc-99 Progeny Resin 99.16% 0.84%

Effective Biomass 99.74% 0.26%

Dose Filtered Solids 99.994% 0.006%

Resin 99.84% 0.163%

Skin Dose Biomass 99.95% 0.050%

Filtered Solids 99.999% 0.0011 %

Resin 99.978% 0.022%

Inhalation Biomass 99.993% 0.0068%

Pathway Filtered Solids 99.9999% 0.00014%

Resin 99.06% 0.94%

Oral Ingestion Biomass 99.71% 0.29%

Pathway Filtered Solids 99.994% 0.0062%

These calculations demonstrate that the contribution of Tc-99 to the occupational dose rate, relative to the uranium+ progeny dose rate, is not significant. In all cases the contribution of the Tc-99 is less than 1% and can therefore be ignored (Reference 3.10, Section 3.3). The external exposure calculations are conservative for Tc-99 because no credit is taken for the shielding that would be present in the form of protective clothing and containers.

7.11 Conclusions The summary table provided in Section 7 .10 demonstrates that, relative to the dose rate from the uranium+ progeny, the Tc-99 represents an insignificant contribution. In accordance with guidance provided in Reference 3 .10, Section 3 .3, the Tc-99 meets the criterion for an insignificant radionuclide. Reference 3 .10 states that for radionuclides or pathways for which the sum of the dose contributions from all radionuclides and pathways are no greater than 10 % of the dose criteria are considered insignificant. Such radionuclides or pathways are insignificant contributors to dose and may be eliminated from further detailed consideration.

8.0 Calculation of Conservative Annual Dose for Tc-99 The following provides a more detailed analysis for each dose pathway that demonstrates that not only is the Tc-99 insignificant in comparison to the uranium+ progeny contribution but that the dose considerations are also insignificant.

CALC NO. EPM027-CALC-001 ENERCON Licensing Considerations for the REV. 1 Excellence- Every project Every day. Presence of Tc-99 PAGE NO. 20 of21 8.1 General Conservatism in Analysis Assumptions are made in this analysis of the starting activity concentrations to be used for each of the three materials selected for analysis (Sections 7.1 through 7.3). In general, the assumptions used to generate these material compositions are intended to maximize the Tc-99 concentrations. Even if the uranium concentrations are reduced, the impact on the conclusion that the Tc-99 is an insignificant contributor will not change.

8.2 External and Skin Dose Pathways The contribution of Tc-99 to the external and skin dose pathways are provided in Sections 7.5 through

7. 7. The highest external and skin dose rates are for the resin materials. For the resin material (Section 7 .5) the contribution of the Tc-99 to the annual dose rate is 5 mrem/yr for the external dose pathway and 7 mrem/yr for the skin dose pathway. Therefore, the conservatively estimated dose rate contribution from the Tc-99 is insignificant for both the external and skin pathways.

8.3 Inhalation and Oral Ingestion Dose Pathways The contribution of Tc-99 to the inhalation and oral ingestion dose pathways are provided in Sections 7.8 and 7.9. The highest inhalation and oral ingestion contributions are for the resin materials. For the resin material the contribution of the Tc-99 to the ALI is 0.008% and for ingestion (Section 7.8) and 0.36% for oral ingestion (Section 7.9). This is equivalent to 0.4 mrem/yr for the inhalation pathway and 18 mrem/yr for the oral ingestion pathway assuming the individual had an intake of one ALI, equivalent to a dose of 5,000 mrem/yr. This assumption is without consideration for the realistic potentials for intake. Even with this conservative assumption, the estimated contributions from the Tc-99 is insignificant for both the inhalation and oral ingestion pathways.

9.0 Comparison with other Evaluations Two results of this evaluation can be compared with two other evaluations that have been separately conducted.

9.1 External Dose Rate for Resin Bed Columns Reference 3. 7 provides the calculation of the external dose rate at 1 foot from a resin column that has been loaded with enriched uranium. The dose rate is given as 0.024 mrem/hr which equates to 210 mrem/yr. Section 7.5 gives the dose rate as 1,540 mrem/yr. The dose conversion factors (Sec. 5.0) of this evaluation are based the dose rate of an infinite plane of contaminated soil without intervening shielding. Reference 3. 7 is based on the physical dimensions of the resin columns including the shielding provided by the steel walls of the resin vessels. Thus, it can be concluded that the results presented in this evaluation are consistent with the separate evaluation.

9.2 Calculation of Potential Intake

CALCNO. EPM027-CALC-001

.ii ENERCON Licensing Considerations for the Presence of Tc-99 REV. 1 E cellence- Eve,y p,o/ecL Eve1y day.

PAGE NO. 21 of21 Reference 3 .4 provides an evaluation of resin material of the potential airborne intake of Uranium and Tc-99 using the methodology ofNUREG-1400, "Air Sampling in the Workplace, September 1993". That evaluation concludes that the potential intake of Tc-99 is 0.22% ALI. Section 8.3 calculates that under the conservative assumptions used the annual intake of Tc-99 would be 0.008%

ALI. Thus, it can be concluded that the results presented in this evaluation are reasonably consistent with the separate evaluation given the different approaches used.

10.0Consideration of Exposure to the Public from the Presence of Tc-99 Tc-99 emits a low energy beta particle that would be fully shielded by the containers used when transporting any of the waste materials. The thickness of steel or wood necessary to stop the beta emissions from Tc-99 is less than 118th inch and the range of the Tc-99 beta in air is approximately 4 feet. (Ref. 3 .9, page 51) Thus, there is no credible exposure to the public from the transportation of waste material due to the Tc-99 present in the waste.

Airborne effluents of the three operational materials considered in Section 7.0 would result in the same percent contribution for the Tc-99 as shown above for the occupational inhalation pathway. The resin waste is the limiting material. The effluent concentration limit for effluent air is 9E-10 for Tc-99 or 330 times lower than the occupational DAC. In Section 8.3, the conservative occupational dose for resin waste is given as 0.4 mrem/yr, thus, the conservative effluent dose to the public from Tc-99 is less than 0.4/330 = 0.001 mrem/yr. Thus, this dose pathway to the public is negligible.

The Tc-99 dose from the water effluent can also be bounded. Section 4.0 ofRef.3.3 estimates the water effluent concentration is 466 pCi/L. Using the NRC effluent concentration limit for Tc-99 from Ref. 3.1 of 6E-5 microCi/ml (equivalent to 50 mrem/yr), the dose is 3.9 mrem/yr from the Tc-99.

Thus, this dose pathway to the public is negligible. Any actual public exposure would be further

. reduced by the fact that the discharge is delivered directly to the river which would further dilute the activity concentration prior to public consumption.

There is no credible exposure pathway for oral ingestion of the waste materials by the public.

11.0Computer Software A Microsoft Excel spreadsheet was used to perform calculations discussed in this calculation.

CALCNO. EPM027-CALC-001 ENERCON CALCULATION PREPARATION REV. 0 Excellence- Every projecr Every day. CHECKLIST PAGE NO. 1 of7 CHECKLIST ITEMS 1 YES NO NIA GENERAL REQUIREMENTS I. If the calculation is being performed to a client procedure, is the procedure being used the latest revision?

Client procedure is not used in this calculation. ENERCON QA procedures used throughout D D [8'.J this project.

2. Are the proper forms being used and are they the latest revision? Same format matching EPMOl 7-CALC-001 was used for internal consistency D [8'.J D
3. Have the appropriate client review forms/checklists been completed?

Client procedure is not used in this calculation. ENERCON QA procedures used throughout [8'.J this project.

D D

4. Are all pages properly identified with a calculation number, calculation revision and page number consistent with the requirements of the client's procedure?

Client procedure is not used in this calculation. ENERCON QA procedures used throughout D D [8'.J this project.

5. Is all information legible and reproducible? [8'.J D D
6. Is the calculation presented in a logical and orderly manner?

[8'.J D D

7. Is there an existing calculation that should be revised or voided?

D [8'.J D

8. Is it possible to alter an existing calculation instead of preparing a new calculation for this situation?

No current ENERCON calculations exist that are similar to this calculation.

D [8'.J D

9. If an existing calculation is being used for design inputs, are the key design inputs, assumptions and engineering judgments used in that calculation valid and do they apply to the calculation D [8'.J D revision being performed.
10. Is the format of the calculation consistent with applicable procedures and expectations? [8'.J D D
11. Were design input/output documents properly updated to reference this calculation?

No ENERCON design inputs or outputs are affected by this calculation. D D [8'.J

CALCNO. EPM027-CALC-001 ENERCON CALCULATION PREPARATION REV. 0 Excellence- Every project. Every doy. CHECKLIST PAGE NO. 2 of7 CHECKLIST ITEMS 1 YES NO NIA

12. Can the calculation logic, methodology and presentation be properly understood without referring back to the originator for clarification? ~ D D OBJECTIVE AND SCOPE
13. Does the calculation provide a clear concise statement of the problem and objective of the calculation? ~ D D
14. Does the calculation provide a clear statement of quality classification?

~ D D

15. Is the reason for performing and the end use of the calculation understood? ~ D D
16. Does the calculation provide the basis for information found in the plant' s license basis?

This calculation applies to a remediation site. No work performed in this calculation is D D ~

applicable to a licensing basis.

17. If so, is this documented in the calculation?

D D ~

18. Does the calculation provide the basis for information found in the plant' s design basis documentation? D D ~
19. If so, is this documented in the calculation? D D ~
20. Does the calculation otherwise support information found in the plant's design basis documentation? D D ~
21. If so, is this documented in the calculation? D D ~
22. Has the appropriate design or license basis documentation been revised, or has the change notice or change request documents being prepared for submittal? D D ~

DESIGN INPUTS

23. Are design inputs clearly identified?

~ D D

24. Are design inputs retrievable or have they been added as attachments?

~ D D

CALCNO. EPM027-CALC-001

rJENERCON CALCULATION PREPARATION REV. 0 Excellence- Every project. Every day. CHECKLIST PAGE NO. 3 of7 CHECKLIST ITEMS 1 YES NO NIA
25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable? MS Excel spreadsheet was used to perform the calculation. All equations are D D ~

provided in the calculation.

26. Are design inputs clearly distinguished from assumptions?

~ D D

27. Does the calculation rely on Attachments for design inputs or assumptions? If yes, are the attachments properly referenced in the calculation? D ~ D
28. Are input sources (including industry codes and standards) appropriately selected and are they consistent with the quality classification and objective of the calculation? ~ D D
29. Are input sources (including industry codes and standards) consistent with the plant's design and license basis? D D ~
30. If applicable, do design inputs adequately address actual plant conditions?

D D ~

31. Are input values reasonable and correctly applied?

~ D D

32. Are design input sources approved?

The Cimairnn design is currently at 60% completion. D ~ D

33. Does the calculation reference the latest revision of the design input source?

~ D D l

34. Were all applicable plant operating modes considered? D D ~

ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective? ~ D D
36. Is adequate justification/basis for all assumptions provided?

~ D D

37. Are any engineering judgments used? D ~ D
38. Are engineering judgments clearly identified as such?

D D ~

CALCNO. EPM027-CALC-001 ENERCON CALCULATION PREPARATION REV. 0 Excellence- Every projecc. Every doy. CHECKLIST PAGE NO. 4 of7 CHECKLIST ITEMS 1 YES NO NIA

39. If engineering judgments are utilized as design inputs, are they reasonable and can they be quantified or substantiated by reference to site or industry standards, engineering principles, D D ~

physical laws or other appropriate criteria?

METHODOLOGY

40. Is the methodology used in the calculation described or implied in the plant's licensing basis? D D ~
41. If the methodology used differs from that described in the plant's licensing basis, has the appropriate license document change notice been initiated? D D ~
42. Is the methodology used consistent with the stated objective?

~ D D

43. Is the methodology used appropriate when considering the quality classification of the calculation and intended use of the results? ~ D D BODY OF CALCULATION
44. Are equations used in the calculation consistent with recognized engineering practice and the plant's design and license basis? ~ D D
45. Is there reasonable justification provided for the use of equations not in common use?

Equations applied in this evaluation are in common use in the industry. D D ~

46. Are the mathematical operations pe1formed properly and documented in a logical fashion? ~ D D
47. Is the math performed correctly?

~ D D

48. Have adjustment factors, uncertainties and empirical correlations used in the analysis been correctly applied? ~ D D
49. Has proper consideration been given to results that may be overly sensitive to very small changes in input?

Results generated by calculations performed in this evaluation are not significantly affected by D D ~

minor perturbations of variables.

SOFTWARE/COMPUTER CODES

50. Are computer codes or software languages used in the preparation of the calculation?

D ~ D

51. Have the requirements of CSP 3.09 for use of computer codes or software languages, including verification of accuracy and applicability been met? D D ~

CALCNO. EPM027-CALC-001 ENERCON CALCULATION PREPARATION REV. 0 Excellence- Every project. Every day. CHECKLIST PAGE NO. 5 of7 CHECKLIST ITEMS 1 YES NO NIA

52. Are the codes properly identified along with source vendor, organization, and revision level? D D ~
53. Is the computer code applicable for the analysis being performed? D D ~
54. If applicable, does the computer model adequately consider actual plant conditions?

D D ~

55. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation? D D ~
56. Is the computer output clearly identified?

D D ~

57. Does the computer output clearly identify the appropriate units?

The output units are not identified in the output document. Tallies have been modified through multipliers and dose response functions. This process has been adequately documented within D D ~

this calculation.

58. Are the computer outputs reasonable when compared to the inputs and what was expected?

Only basic functions and operations in Microsoft Excel-Office 16 were applied in this

~ D D calculation.

59. Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? D D ~

RESULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified?

No acceptance criteria required for this evaluation. D D ~

61. Are the stated acceptance criteria consistent with the purpose of the calculation, and intended use? D D ~
62. Are the stated acceptance criteria consistent with the plant's design basis, applicable licensing commitments and industry codes, and standards? D D ~
63. Do the calculation results and conclusions meet the stated acceptance criteria?

D D ~

64. Are the results represented in the proper units with an appropriate tolerance, if applicable? ~ D D

CALCNO. EPM027-CALC-001 ENERCON CALCULATION PREPARATION REV. 0 Excellence- Every projecr. Every doy. CHECKLIST PAGE NO. 6 of7 CHECKLIST ITEMS 1 YES NO NIA

65. Are the calculation results and conclusions reasonable when considered against the stated inputs and objectives? ~ D D
66. Is sufficient conservatism applied to the outputs and conclusions? ~ D D
67. Do the calculation results and conclusions affect any other calculations?

No ENERCON calculations are affected by this evaluation. D D ~

68. If so, have the affected calculations been revised?

D D ~

69. Does the calculation contain any conceptual, unconfirmed or open assumptions requiring later confirmation? D ~ D
70. If so, are they properly identified?

D D ~

DESIGN REVIEW 7 l. Have alternate calculation methods been used to verify calculation results?

D D ~

Note:

1. Where required, provide clarification/justification for answers to the questions in the space provided below each question.

An explanation is required for any questions answered as "No' or "NIA" .

CALC NO. EPM027-CALC-OO I F..:d ENERCON CALCULATION PREPARATION REV. 0 Excellence- Eve ry proJect. Every day. CHECKLIST PAGE NO. 7 of7 Originator:

12/16/2019 Print Name Date

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