ML17363A086

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Duane Arnold Energy Center, Calc No. NEE-323-CALC-001, Primary Containment Radiation EAL Threshold Determination
ML17363A086
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 12/15/2017
From:
NextEra Energy Duane Arnold
To:
Office of Nuclear Reactor Regulation
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ML17363A067 List:
References
NG-17-0235 Calc No. NEE-323-CALC-001
Download: ML17363A086 (17)
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Development of EAL Threshold values from NEE-323-CALC-001 Calculated values were added to the typical background readings of these monitors, and then rounded to aid in evaluator use of the EALs. Resulting values used in the DAEC Fission Product Barrier chart are shown below:

  • Fuel Clad Barrier: o Fuel Clad Barrier LOSS 4.A = Drywell Monitor (9184A/B) reading taken from NEE-323-CALC-006 (V30) instead o Fuel Clad Barrier LOSS 4.B = Torus Monitor (9185A/B) reading greater than 200 R/hr.
  • RCS Barrier LOSS 4.A = Drywell Monitor (9184A/B) reading greater than 5 R/hr after reactor shutdown. (minimum serviceable threshold value accounting for scale of monitor) o
  • Calculated Torus Monitor (9185A/B) response is below scale of monitor and not used.
  • CTMT Barrier LOSS 4.A = Drywell Monitor (9184A/B) reading greater than 5000 R/hr. o
  • CTMT Barrier LOSS 4.A = Torus Monitor (9185A/B) reading greater than 500 R/hr.

CALC NO. NEE-323-CALC-001 ENERCON CALCULATION COVER I SHEET REV. 00 Ex ce llen c e-E very p roj ect. E very d ay. PAGE NO. 1 of 10 Primary Containment Radiation EAL Threshold Client: Duane Arnold Energy Center Title: Determination Project Identifier:

NEE-323 Item Cover Sheet Items Yes No 1 Does this calculation contain any open assumptions, including preliminary D [8J information , that require confirmation? (If YES , identify the assumptions

.) 2 Does this calculation serve as an " Alternate Calculation

"? (If YES , identify the D [8J design verified calculation.)

Design Verified Calculation No. --3 Does this calculation supersede an existing Calculation? (If YES, id entify the D [8J design verified calculation

.) Superseded Calculation No. --Scope of Revision:

Initial Issue Revision Impact on Results: Initial Issue Study Calculation D Final Calculation Safety-Related D Non-Safety-Related (Print Name and Sign) Originator:

Aaron Holloway Date: 12/12/17 Design Verifier 1 (Reviewer if NSR): Jay Bhatt Date: 12/12/17 Approver:

Zachary Rose Date: 12/1 2/17 Note 1: For non-safety-related calculation , design verification can be substituted by review.

JI ENERCON CALCULATION CALC NO. NE E-323-CALC-OO 1 Excellence-Every project. Eve r y day. REVISION STATUS SHEET REV. 00 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 00 12/12/17 I n itial Is sue PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION All 00 APPENDIX/ATTACHMENT REVISION STATUS APPENDIX NO. NO.OF REVISION ATTACHMENT NO.OF REVISION PAGES NO. NO. PAGES NO. A 1 00 1 4 00 B 1 00

." ENERCON Excellence-Every project. Every day. TABLE OF CONTENTS Section 1.0 Purpose and Scope 2.0 Summary of Results and Conclusions 3.0 References 4.0 Assumptions 5.0 Design Inputs 6.0 Methodology 7.0 Calculations 8.0 Impact Assessment List of Appendices Appendix A -Calculation Spreadsheet Appendix B -Calculation Spreadsheet Formulas List of Attachments Attachment 1 -Calculation Preparation Checklist Page 3 of 9 CALC NO. NEE-323-CALC-001 REV. 00 Page No. 4 4 5 5 6 8 1 0 10 # of Pages 1 1 # of Pages 4 ENERCON NEE-323-CALC-001 CALC Primary Containment Radiation NO. EAL Threshold Determination r-------------------1 Excellence-Every projecr. Every day. REV. 00 1.0 Purpose and Scope The purpose of this calculation is to determine the site-specific threshold for primary containment radiation in the event of a loss or potential loss of the three fission product barriers (fuel clad, Reactor Coolant System, containment).

These specific values can be used to determine the Emergency Action Level (EAL) (FA 1 , FS1, or FG1) in accordance with Table 9-F-2 of NEI 99-01, Rev. 6. This calculation is nonsafety-related as it intended for emergency classification and not design basis purposes.

There are no acceptance criteria associated with this calculation since the purpose is only to determine site-specific radiation thresholds.

2.0 Summary of Results and Conclusions The calculated primary containmen t radiation readings for each of the three fission product barriers are listed below. Note that the results presented below are calculated dose rates and do not account for background radiation or any installed detector check sources. Table 1 -Calculated Containment Atmospheric Monitoring System (CAMS) radiation readings for a release into the drywe/1 Failure Reactor Coolant System (Loss) Fuel Clad (Loss) Containment (Potential Loss) Drywell Monitor (9184A/B)

Reading (R/hr) 1.3 3 2000 5130 Table 2 -Calculated CAMS radiation readings for a release into the torus Failure Reactor Coolant System (Loss) Fuel Clad (Loss) Containment (Potential Loss) Torus Monitor (9185A/B)

Reading (R/hr) Page 4 of 10 0.125 (not on s cale) 188 484 ENERCON NEE-323-CALC-001 Excellence-Every project. Every day. CALC Primary Containment Radiation NO. EAL Threshold Determination f----------------1 REV. 00 3.0 References 3.1 NG-88-0966, " Nuclear Generation Division Office Memo , G.E. Fuel Damage Documentation/

Dose Rate Ca l culations", dated 03/18/88 3.2 IPOI 8, "Outage and Refueling Operations

", Rev. 91 3.3 Bech-M115, "Reactor Vessel Instrumentation P&ID", Rev. 62 3.4 Duane Arnold Energy Center Facility Operating License Appendix A -Technical Specifications, as revised through Amendment No. 297 3.5 NEI 99-01 , " Development of Emergency Action Levels for Non-Passive Reactors", Rev. 6 3.6 Shultis, J.K., " Fundamentals of Nuclear Science and Engineering

", 2002 3.7 Lindeburg , M.R., "Mechanical Engineering Reference Manual for the PE Exam", Twelfth Edition , 2006 3.8 Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation , Submersion, and Ingestion", 1989 3.9 I.RIM-V115-01 , "Victoreen Model 876A Containment Radiation Monitor Calibration", Rev. 10 Page 5 of 10 ENERCON NEE-323-CALC-001 Excellence-Every projecr. Every day. CALC Primary Containment Radiation NO. 1---'-'__:..__:_-

EAL Threshold Determination REV. 00 4.0 Assumptions 4.1 Reactor Pressure Vessel (RPV) water level is 535 inches above vessel zero for the purposes of calculating the total Reactor Coolant System (RCS) water volume. This corresponds to the middle of the band between the high and low RPV water level alarm points from Table 4 (Design Input 5.7) and represents the most realistic water inventory during normal operation. 4.2 The fission product isotopic distribution in the reactor coolant will be similar to that of the fission product gap inventory.

This is reasonable since , in the event of a fuel cladding failure , the isotopes of concern (iodines) would be released to the reactor coolant at the same time and distribution. 4.3 All reactor coolant mass is assumed to be released into the primary containment.

This is consistent with the NEI 99-01 Rev. 6 developer notes. 5.0 Design Inputs 5.1 Duane Arnold's license thermal power limit is 1912 MWth , taken from Reference 3.4. 5.2 The specific volume of saturated liquid water at 1000 psia is 0.02160 ft 3/lbm per Appendix 23.B of Reference 3.7. 5.3 The following unit conversions are used within this calculation:

Table 3 -Applicable Unit Convers i ons Ba s e Unit 1 Sievert (Sv) 1 Curie (Ci) 1 Gallon l lbm/ft 3 Equivalent l.O E 5 mr e m 3.7 E 10 B q 3 785.4 cubic c e nt i mete rs (cc) 0.016018 g/cc R e ferenc e 3.6 3.6 3.7 3.7 5.4 The Technica l Specifications limit for RCS activity is 0.2 µCi/gm Dose Equivalent 1-131 (DEi) per LCO 3.4.6 of Reference 3.4. 5.5 The RCS volume at the centerline of the Main Steam lines is 72 , 000 gallons per Reference 3.2. 5.6 The change in RCS volume per unit change in height is 100 gallons/inch per Reference 3.2. 5.7 The following elevations are taken from Reference 3.3: Table 4 -Pertinent RPV elevations relative t o Vessel 0 Point Nozzle N3A,B,C,D (centerline of Main Steam Lines) Height abo v e V e ss el O (inches) 6 2 0.5 Page 6 of 10 ENERCON N EE-323-CALC-OO 1 Excellence-Every projecr. Every day. CALC Primary Containment Radiation NO. EAL Threshold Determination i-----------------1 REV. 00 Point High Level Alarm Low Level Alarm Height above Vessel O (inches) 539.5 530.5 5.8 The drywell dose rate at the CAMS monitor location for 100% gap release into the drywell of a 1691 MWth core at 0.01 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> decay time is 2.27E4 R/hr , per Table 3 of Reference 3.1. 5.9 The torus dose rate at the CAMS monitor location for 100% gap release into the torus of a 1691 MWth core at 0.01 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> decay time is 2.14E3 R/hr , per Table 4 of Reference 3.1. 5.10 The drywell (9184 A/8) and torus (9185 A/8) radiation monitor ranges (1 to 10 7 R/hr) are taken from Reference 3.9. 5.11 The fission product gap inventories for Iodine isotopes are taken from Table 1 of Reference 3.1. These inventories correspond to a core with a rated thermal power of 1691 MW t h. Table 5 -Iod i ne Gap Inventor i es for 1691 MWt h Core Nuclide 1691 MWth Gap Inventory (Ci) 1-130 7.2 5 E+03 1-131 5.34E+0 5 1-132 8.45 E+0 4 1-133 3.6 3E+05 1-134 8.19E+04 1-135 1.93E+0 5 5.12 Dose conversion factors for effective dose due to inhalation are taken from Reference 3.8 , Table 2.1. Table 6 -Dose Conversion Factors fo r To t al Effective Dose from I nha lation Nuclide 1-130 1-131 1-132 1-133 1-134 1-135 Dose Con v ersion Coefficient (S v/Bq) 7.14 E-10 8.89 E-0 9 l.03 E-10 l.58 E-0 9 3.55 E-1 1 3.32E-1 0 Page 7 of 10 ENERCON Excellence-Every project. £very day. 6.0 Methodology NEE-323-CALC

-001 CALC Primary Containment Radiat i o n NO. EAL Threshold Determination t----------------, REV. 00 The approach of this calculation is to scale the results of a prev i ous calculation (NG-88-0966 , Reference 3.1) based on the specific RCS activities as specified in NEI 99-01 Rev. 6. Scaling factors are determined for each of the fission product barrier failure thresholds specified in NEI 99-01 (i.e. Loss of RCS , Loss of Fuel Cladding , and Potential Loss of Primary Containment).

The RCS ac t ivit y concentrations are taken from the NEI 99-01 Rev. 6 developer notes and are listed below. Tab l e 7 -RCS Activities for f i ss i on produc t b a rr ier fa i l ur e s F ailure Reactor Coolant System (Loss) Fuel Clad (Lo ss) Containment otential Loss R CS A ct ivity Te chnic a l Spe ci fica t ion L i m it 30 0 µCi/g D ose Equivalent I-131 20% fue l cl add i n fail ur e These scaling factors are then applied to the CAMS radiation response from calculation NG-88-0966 to determine the site-specific va l ues for t h e three thresholds. Additionally , the previous calculation determined t he CAMS radiation monitor response for an assumed release of 100% gap activity from the core with a power level of 1691 MWth. However , Duane Arnold's licensed power level is now 1912 MW t h. This does not impact the Reactor Coolant System and Fuel Clad barrier thresholds because the radiation responses are scaled based on the DEi levels. However , the difference in licensed power l evel will need to be accounted for in the Potential Loss of Containment threshold because this threshold is related to the total gap inventory. Scaling of the gap inventory based on power level is consistent with calculation NG-88-0966 as Table 1 of NG-88-0966 provides gap inventory per megawatt.

6.1 Determination of RCS water volume and mass The NEI 99-01 Rev. 6 primary containment radiation thresholds are based on specific RCS radioactivity concentrations.

However , the correspond i ng total RCS activity must be known in order to compa r e these t h r esholds to the gap release assumed in calculation NG-88-0966. Therefore , the total mass of water in the RCS must first be determined so that t he total RCS activity can be calculated for each threshold.

The total RCS water volume at normal operation is determined by taking the RCS volume when filled to the centerline of the main steam lines , and then subtracting the difference in volume between the centerline of the main steam lines and the normal water level. This i s presented in Equation 1 below: V orma l = V M SL -aH [Equation 1] Where: V nonnat = The RCS water volume at normal operation (gallons) Page 8 of 10 ENERCON NEE-323-CALC-001 Excellence-fvery project Every day. CALC Primary Containment Radiation NO. EAL Threshold Determination f---------

~ REV. 00 V MsL = The RCS water volume when filled to the centerline of the main steam lines (gallons) a= The change in RCS wate r vo l ume per inch change i n vessel heigh t (gallons/inch) H = T he distance between the centerline of the main steam l i nes and the normal RCS water level (inches) It should be noted that calculating the RCS volume as shown above does not include the volume of the steam lines. However , the volume of the steam lines filled to the centerline of the nozzle is very sma ll compared to the total RCS volume, and therefore does not significantly impact the results of the calculation.

The total mass of water in the RCS can then be determined based on the water density, as outlined in Equation 2 below: M = V n o rma! n o*rmai v [Equation 2] Where: M normat = The mass of water i n the RCS at normal operat i on (grams) v = The specific volume of water at normal ope r ation (grams/gallon) 6.2 Determination of Scaling Factors The scaling factors for the fuel clad and RCS barrier thresholds are determined by comparing the corresponding dose at each RCS activity concentration to the DEi of the fission product gap inventory. This is presented in Equation 3 below: DCF1-1 3 1 A 1-131 M n o rm a l F = '\'i=135 I DCF ~i=1 30 i i [Equation 3] Where: F = The scaling factor for a given RCS activity concentrat i o n t hreshold.

DCF;= The dose convers i on factor for isotope "i" in mrem/C i. These values are developed from Table 6 above. A1-131 = The 1-131 concentration in the RCS fo r a given threshold in Ci/gram. These va l ues are developed from Table 7 and Des i gn Input 5.4 above. Mn o rm at = The mass of RCS water at normal operation in grams. This value i s determined from Equation 2 above. /;=The gap inventory of iodine isotope " i" a t a power leve l of 1691 MW th i n C i. These values are taken from Table 5 above. Page 9 of 10 ENERCON Excellenco-fvery project. fvery day. Primary Containment Radiation EAL Threshold Determination CALC NO. REV. NEE-323-CALC-001 00 For the potential loss of containment threshold , NEI 99-01 specifies that 20% of fuel cladding has failed, rather than giving a specific RCS activity concentration.

Therefore, the scaling factor for this case is simply 0.2 (20% of the 100% gap release case) multiplied by the ratio of the new to previous licensed power levels (1912/1691) to account for the increased gap inventory. 6.3 Determination of CAMS Detector Response Once the scaling factors have been determined for each of the three RCS activity concentration thresholds, they can be applied to the results of calculation NG-88-0966 to determine the CAMS detector response for each threshold.

Specifically, the CAMS detector response can be obtained from Equation 3 below: Where: D* 1 [Equation 3] Di = The dose rate at the CAMS detector for an RCS activity concentration of "j" in R/hr F i= The scaling factor for an RCS activity concentration of "j", determined from Equation 2 above DGAP = The dose rate at the CAMS monitor location for 100% gap release of a 1691 MWth core in R/hr 7.0 Calculation All calculations were completed using Microsoft Excel. The calculation results and spreadsheet formulas are presented in Appendix A and B, respectively.

8.0 Impact Assessment This calculation is based on "realistic" assumptions for the purpose of declaring EALs, rather than typical conservative "bounding" type design basis analyses.

The calculation results are intended to provide order of magnitude dose rates to assist Operations and Emergency Response personnel in determining the state of the three fission product barriers in accordance with NEI 99-01 Rev. 6. Page 10 of 10

I ENERCON CALC N EE-323-CALC-OO 1 Appendix A NO. Excellence-Every project. Every day. Calculation Spreadsheet REV. 00 A B C D E 1 1 Sievert 100000 mrem 2 lCi 3.70E+10 Bq 3 4 Isotope DCF (Sv/Bq) DCF (mrem/C i) 1691 MW Gap Inventory (Ci) Dose (mrem) 5 1-130 7.14E-10 2.64E+06 7.25E+03 l.92E+l0 6 1-131 8.89E-09 3.29E+07 5.34E+05 1.76E+l3 7 1-132 l.03E-10 3.81E+05 8.45E+04 3.22E+l0 8 1-133 l.58E-09 5.85E+06 3.63E+05 2.12E+12 9 1-134 3.55E-11 l.31E+05 8.19E+04 l.08E+10 10 1-135 3.32E-10 1.23E+06 l.93E+05 2.37E+ll 11 12 Total 2.00E+l3 13 V M S L 72000.0 gal 14 a 100.0 gal/inch 15 Elevation of the Ma i n Steam Lines 620.5 inches above vessel 0 16 Elevation of the Normal Water Level 535.0 inches above vessel 0 17 H 85.5 inches 18 VNorma l 63450 gal 19 VNormal 240184264.5 cc 20 Specific Volume @ 1000 psia 0.0216 ft 11 3/lbm 21 water density @ 1000 psia 0.7416 g/cc 22 MN o rm a l 178114424 grams 23 24 Drywell Dose Rate for 100% Gap Re l ease (169 1 MWth) 2.27E+04 R/hr 25 Torus Dose Rate fo r 1 00% Gap Release (1691 MWth) 2.14E+03 R/h r 26 27 Threshold Scaling Factors (F) Drywell Dose Rate (R/hr) Torus Dose Rate (R/hr) 28 0.2 µCi/gm {TS Limit) S.86E-OS 1.33E+OO 1.2SE-01 29 300 µCi/gm 8.79E-02 2.00E+03 1.88E+02 30 20% Failed Fuel 2.26E-01 S.13E+03 4.84E+02 Page 1 of 1 ENERCON Exce l lence-Every p roject. E ve ry d ay. A 1 1 Sievert 2 1 Ci 3 4 Isotope 5 1-130 6 1-131 7 1-132 8 1-133 9 1-134 10 1-135 11 12 13 VMSL 14 a 15 Elevation of the Main Steam Lines 16 Elevation of the Normal Water Level 17 H 18 VNorm o l 19 VNormal 20 Specific Volume @ 1000 ps ia 21 water density@ 1000 psia 22 MN o rm a l 23 24 Drywell Dose Rate for 100% Gap Release (1691 MWth) 25 Torus Dose Rate for 100% Gap Release (1691 MWth) 26 27 Threshold 28 0.2 µCl/gm (TS Limit) 29 300 µCl/gm 30 20% Failed Fuel Appendix B Calculation Spreadsheet Formulas CALC NO. REV. 8 C 100000 mrem 37000000000 8q DCF (Sv/Bq) DCF (m rem/Ci) 0.000000000714

=85*$8$1 *$8$2 0.00000000889

=86*$8$1 *$8$2 0.000000000103

=87*$8$1 *$8$2 0.00000000158

=88*$8$1 *$8$2 0.0000000000355

=89*$8$1 *$8$2 0.000000000332

=810*$8$1*$8$2 72000 gal 100 gal/inch 620.5 inches above vessel 0 535 inches above vessel O =B15-816 inches =Bl3-814*817 gal =818*3785.41 cc 0.0216 ft"3/lbm =0.016018/820 g/cc =819.821 grams 22700 R/hr 2140 R/hr NEE-323-CALC-001 00 D 1691 MW Gap Inventory (Ci) 7250 534000 84500 363000 81900 193000 Total Scaling Factors (Fl Drywell Dose Rate (R/hr) Torus Dose Rate (R/hr) =$C$6"'0.0000002"'$8$22/($E$12)

=828"'$8$24

=828"'$8$25 =$C$6"'0.0003"'$8$22/($E$12)

=829*$B$24

=B29"'$8$25 =0.2 "'1912/1691

=830"'$8$24

=830"'$B$25 Page 1 of 1 E Dose (mrem) =D5*c5 =D6*C6 =D7*C7 =D8*CB =D9*C9 =D1o*c10 =SUM(E5:E10)

ENERCON Excellence-Every p rojecr. Eve ry d ay. Attachment 1 CALCULAT I ON PREPARAT I ON CHECKLIST CHECKLIST ITEMS 1 CALC NO. REV. GENERAL REQUIREMENTS

1. If the calculation is being performed to a client procedure , is the procedure being used the latest rev i sion? The calculation is being prepared to ENERCON's procedures. 2. Are the proper forms being used and are they t h e latest revis i o n? 3. Have the appropriate client review forms/checklists been completed?

The ca l culation is being prepared to ENERCON's procedures. 4. Are a ll pages proper l y identified with a calcu l ation number , calculation revision and page number consistent with the requirements of the client's procedure?

5. I s all information legible and reproducible?
6. I s the calculation presented in a logical and order l y manner? 7. I s there an existing calculation that should be revised or voided? This is a new calculation to support implementing NEI 99-01 Rev. 6 8. Is it possible to alter an existing calculation instead of preparing a new calculation for th i s situation?
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 app l y to the ca l culation revision being performed.
10. I s the format of the calculation consistent with app l icable procedures and expectations?
11. Were design inpuUoutput documents properly updated to reference this calculation?
12. Can the calculati o n logic , meth o dology and presentation be properly understood without referring back to the or i ginator for clarification?

OB JE CT IVE A N D S C O PE 13. Does the calculation provide a clear concise statement of the problem and objective of t h e calcu l a t ion? 14. Does the calcu l at i on provide a clear statement of quality classification?

15. Is the reason for performing and the end use of the calcu l ation understood?
16. Do es the ca l cu l ati o n provide t h e b asis for i nfor m ation found in the plant's l icense basis? 17. If so, is this documented in the calculation?
18. Do es the cal c ulati o n provide the basis for information found in the plant's design basis documentation?

Page 1 of 4 NEE-323-CALC-001 00 YES NO N/A D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D D

19. 20. 21. 22. ENERCON Excellence-Every proj e ct. E ve ry day. Attachment 1 CALCULATION PREPARATION CHECKLIST CHECKLIST ITEMS 1 If so , is th i s documented in the calculat i on? CALC NO. REV. Does the calculation otherwise support informati on found in the plant's design basis documentation?

If so , is this documented i n the calculation?

Has the appropriate design or license basis documentation been revised , or has the change notice or change request documents being prepared for subm i ttal? DESIGN INPUTS 23. A r e design inp uts clearly identified?

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

? 25. If Attachments are used as design inputs or assumptions are the Attachments traceable and verifiable?

26. Are design inputs clearly distinguished from assumptions?
27. Does the calculation rely on Attachments for design i nputs or assumptions?

If yes , are the attachments properly referenced in the calculation?

28. Are input sources (including industry codes and standards) appropriately selected and are they consistent with the quality classification and objective of the calculation?
29. Are input sources (including industry codes and standards) consistent with the plan t's design and license basis? 30. If applicable , do design inputs adequately address actual plant conditions?
31. Are input values reasonable and correctly applied? 32. Are design input sources approved?
33. Does the calculation reference the latest revision of the design input source? 34. Were all applicable plant operating modes considered?

ASSUMPTIONS

35. Are assumptions reasonable/appropriate to the objective?
36. Is adequate justification/basis for all assumptions provided?
37. Are any engineering j udgments used? 38. Are engineering judgments clearly identified as such? 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 , physical laws or other appropriate criteria?

Page 2 of 4 NEE-323-CALC-001 00 YES NO N/A D D 0 D 0 D D D 0 D D 0 0 D D 0 D D D D 0 0 D D D 0 D 0 D D 0 D D 0 D D 0 D D 0 D D 0 D D 0 D D 0 D D 0 D D D 0 D D D 0 D D 0 ENERCON Exc e llence-Every p roject. £ very d ay. Attachment 1 CALCULAT I ON PREPARATION CHECKLIST CHECKLIST ITEMS 1 CALC NO. REV. METHO DOL OGY 40. Is the methodology used in the calculat i on described o r i mplied in the p l ant's licens i ng basis? 41. I f the methodology used differs from that described in the plant's licensing basis , has the appropriate license document change notice been i n i t i ated? 42. Is the methodology used consistent with the stated objective?

43. Is the methodo l ogy used appropriate when considering t he quality classificat i on of the calculation and intended use of the resu l ts? BOD Y OF C ALCULA TION 44. Are equations used i n the calculation consistent with recognized eng i neering pract i ce and the p l an t's design and l i cense bas i s? 45. I s t h ere reasonab l e justification pro vi ded for the use of equations not i n common use? 46. Are the mathematical operations performed properly and documented in a logica l fashion? 47. Is the math performed correctly?
48. Have adjustment factors , uncertainties and emp i rical correlations used i n the analys i s been correctly app l ied? 49. Ha s proper considerat i on been given to results that may be overly sensitive to very sma ll changes in input? S OFTWARE/C O MPUTE R CODES 50. Are computer codes or software languages used in the preparation of the calculation?
51. Have the requirements of CSP 3.09 for use of computer codes or software languages , including ver i fication of accuracy and app li cability been me t? 52. Are the codes properly i dentified along with source vendor , organi z at i on , and revisio n level? 53. Is the computer code appl i cable for the analysis being performed?
54. I f applicable, does t h e comp u ter model adequately consider actua l plant conditions?

5 5. Are the inputs to the computer code clearly identified and consistent with the inputs and assumptions documented in the calculation?

56. I s t h e c o mputer ou tp ut clearly i d e n t i fied? 57. Does the computer output clearly ident i fy the appropriate un i ts? , Page 3 of 4 NEE-323-CALC-001 00 YES NO NIA 0 D 0 D 0 D 0 D D D 0 D 0 D 0 D 0 D 0 D 0 D 0 0 0 0 0 0 0 0 0 D 0 D 0 0
58. 59. ENERCON Excelle n ce-Every project. E ve ry d ay. Attachment 1 CALCULATION PREPARATION CHECKLIST CHECKLIST ITEMS 1 CALC NO. REV. Are the computer outputs reasonable when compared to the inputs and what was expected?

Was the computer output reviewed for ERROR or WARNING messages that could invalidate the results? R E SULTS AND CONCLUSIONS

60. Is adequate acceptance criteria specified?
61. Are the stated acceptance cr i teria consistent with the purpose of the calculation, and i ntended use? 62. Are the stated acceptance criteria consistent with the plant's design basis , applicable licensing commitments and i ndustry codes , and standards?
63. Do the calculation results and conclusions meet the stated acceptance criteria?
64. Are the results represented in the proper un i ts with an appropriate tolerance , if app li cable? 65. Are the calculation results and conclusions reasonable when considered against the stated inputs and object i ves? 66. Is sufficient conservat i sm applied to the outputs and conclusions?
67. Do the calculation results and conclusions affect any other calculations?
68. If so, have the affected calculati o ns b e en revised? 69. Does the calculation contain any conceptual , unconfirmed or open assumptions requiring later confirmation?
70. I f so, are they properly identified?

DESI GN R E VIEW 71. Have alternate calculation methods been used to verify calculation results? No, a Desig n Re v iew was performed. Note: NEE-323-CALC-001 00 YES NO N/A D D D D D D D D D D D D D D D D D D D D D D D D D D D D 1. Where required , provide clarification/justification for answers to the questions in the space prov i ded below each question. An explanation is required for any questions answered as " No' or " N/A". O ri g i n a tor: Aar o n Holloway 12/12/17 Print Name and Sign Date Page 4 of 4

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