NG-17-0235, Duane Arnold Energy Center, Calc No. NEE-323-CALC-002, Dose Rate Evaluation of Reactor Vessel Water Levels During Refueling for EAL Thresholds

From kanterella
Revision as of 03:40, 19 July 2018 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Duane Arnold Energy Center, Calc No. NEE-323-CALC-002, Dose Rate Evaluation of Reactor Vessel Water Levels During Refueling for EAL Thresholds
ML17363A085
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
Shared Package
ML17363A067 List:
References
NG-17-0235 Calc No. NEE-323-CALC-002
Download: ML17363A085 (38)
v • d • e


Text

Development of EAL Threshold values from NEE-323-CALC-002 Due to elevated background radiation levels on these monitors during plant operation (10-12 R/hr), the calculated threshold value was rounded to 5 (minimum serviceable threshold value accounting for scale of monitor) for ease of use by the EAL evaluator , and the " in Mode 5 only" caveat is added to the EAL usage. The resultant EALs are: RA2.2 Reading greater than 5 R/hr on ANY of the following radiation monitors (in Mode 5 only):

  • NW Drywell Area Hi Range Rad Monitor , RIM-9184A
  • South Drywell Area Hi Range Rad Monitor , RIM-91848 CS1/CG1 Core uncovery is indicated by ANY of the following:
  • Drywell Monitor (9184A/B) reading greater than 5.0 R/hr CALC NO. NEE-323-CALC-002 I ENERCON CALCULATION COVER I SHEET REV. 00 E,ceflence

-Evtry p r o1ec1. Every doy, PAGE NO. 1 of 28 Dose Rate Evaluation of Reactor Vessel Client: Duane Arnold Energy Center Title: Water Levels During Refueling for EAL Thresholds Project Identifier:

NEE-323 Item Cover Sheet Items Yes No 1 Does th i s calculation contain any open assumptions , inclu ding preliminary D information , that require confirmation? (If YES , identify the assumptions

.) 2 Does this calculati o n serve as an " Alternate Calculation

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

Design Verified Calculation No. 3 Does this calculation supersede an existing Calculation? (If YES , identify the design D 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 Superseded Calculation No. --Scope of Revision: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:

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

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

I ENERCON CALCULATION CALC NO. NEE-323-CALC-002 I Exc,11,na,-Every project fv*ry day. REVISION STATUS SHEET REV. 00 CALCULATION REVISION STATUS REVISION DATE DESCRIPTION 00 12/12/17 Initial Issue 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 5 00 B 2 00 C 1 00 Page 2 of 28 ENERCON TABLE OF CONTENTS Excellence

-Every project. £very d ay. 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 Computer Software 9.0 Impact Assessment List of Appendices Appendix A -Electronic File Listing Appendix B -DAEAL.xlsx Sheets Appendix C -SCALE Input List of Attachments Attachment 1 -Calculation Preparation Checklist Page 3 of 28 CALC NO. NEE-323-CALC-002 REV. 00 Page No. 4 4 5 6 8 13 14 27 28 # of Pages 1 2 1 # of Pages 5

1 During Refueling for EAL Thresholds REV. 00 1.0 Purpose and Scope The purpose of this calculation is to evaluate dose rates with water at the top of active fuel in the reactor vessel during cold shutdown or refueling operations in o r der to set Emergency Action Level (EAL) thresholds (RA2 , CS1 , CG1) per NEI 99-01 [Reference 3.5]. The dose rates are calculated at the locations of the drywell monitors 9184A/B so that dose rate measurements by these devices can be correlated to the water level in the core, upon failure of other water level detection systems. This calculation is nonsafety-related as the results of the calculation do not affect the design basis or safety-related systems structures or components.

These results are best estimates based on as-built conditions and provide information to operators with respect to classifying an emergency, therefore no acceptance criteria is required.

2.0 Summary of Results and Conclusions The dose rates just prior to the core being uncovered (i.e. water at the top of the active fuel) are shown in the table below. Note that the results presented below are calculated dose rates and do not account for background radiation or any installed detector check sources. Model Description Head Off Head On 1 This value is off scale low. Table 1 -Dose Rate at Top of Active Fuel Drywell Monitor 9184A Reading (R/hr) 1.81 1.11 Drywell Monitor 9184B Reading (R/hr) 1.68 7.41 E-0 1 1 Page 4 of 28 Drywell Monitor (9184A/B) Range (R/hr) 1 to 1 E+7 1 to 1 E+7 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels f-----------------1 ENERCON Excelle n ce-Ev e ry p roJect. E ve ry day. 3.0 References During Refueling for EAL Thresholds REV. 00 3.1 "Standard Composition Library ," ORNL/NUREG/CSD-2N1/R6 , Volume 3 , Section M8 , March 2000 3.2 CGDG-SCALE-6.1

.2 , Rev 00 , Commercial Grade Dedica ti on SCALE Version 6.1.2 3.3 CGDG-MCNP6-V1

.0 , Rev 00 , Commercial G r ade Dedication MCNP6 Version 1.0 3.4 ANSI/ANS 6.1.1-1977 , Neutron and Gamma Flux-To-Dose Conversion Factors 3.5 NEI 99-01 , Rev. 6 , " Development of Emergency Act i on Levels for Non-Passive Reactors" 3.6 I.RIM-V115-01 , Rev. 10 , " Victoreen Model 876A Containment Radiation Monitor Calibration" 3.7 NUREG 1940 , " RASCAL 4: Descriptions of Models and Methods" 3.8 CAL-ROO-PUP-008 , Rev. 03 , " Non-L OCA Radio l ogical Consequence Dose with Alternate Source Term" 3.9 RFP 110 , Rev. 45 , " Refueling Procedure-Reactor Pressu r e Vessel Disassembly" 3.10 Technical Specifications , Section 1.1 3.11 Technical Specifications, Section 4.2.1 3.12 NUREG 1754 , " A New Comparative Analysis of LWR Fuel Designs" 3.13 BECH-M009 , Rev. 14 , " Equipment Locations Reactor Build i ng Section-GG" 3.14 BECH-C405 , Rev. 14 , " Reactor Building Floor Plan@ El. 757'-6"" 3.15 NG-17-0156 , Proprietary Data Transmittal to ENERCON 3.16 BECH-M405 , Sh 04, Rev. 24 , "Instrument Poi n ts and Rack Locations Diagram Plans at Elevs 812'-0" & 833'-6"" 3.17 NG-88-0966 , " G.E. Fuel Damage Documentation/Dose Rate Calculations" 3.18 C003-029, Rev. 0 , " Drywell Cylindrical Shell & Cone" 3.19VS-01-06, Rev. 4 , " Top Head Assembly" 3.20 BECH-C511 , Rev. 5 , " Reactor Building RPV Ped Dev. Elev. & Sect's" 3.21 BECH-C514 , Rev. 1 , " Drywell Interior Biological Shield Wall Reinforcing Sections" 3.22 BECH-C-516 , Rev. 6 , " Drywell Interior Biological Shield Wall Plans El. 816'-3 %" to El 779'-1 Yi"" 3.23 BECH-M405 , Sh 02 , Rev. 71 , " Instrument Points & Lines Diagrams Plan at Elev 757'-6"" 3.24APED-B-31-2816-001 , Rev. 5 , " Outline Reactor Recirculating Pump" Page 5 o f 28 Excellence-Every pro;ecr. Every day. Dose Rate Evaluation of CALC NO. N EE-323-CALC-002 Reactor Vessel Water Levels i--------------During Refueling for EAL 00 Thresholds REV. /, ENERCON 3.25 FSAR Section 4.3.2.1 , and Section 9.1 3.26 CAL-M98-058 , Rev. 1 , " ADS Accumulator Size Verification" 4.0 Assumptions 4.1 The core is homogenized based on the typica l 1 Ox10 fuel assembly dimensions , taking into account the fuel rods and space between. Any small variations in fuel parameters will have a negligible effect on containment dose rates. The cladding is modeled as Zircaloy 4 in lieu of ZIRLO; this is acceptable due to t he s i mila r ity o f the materials.

4.2 Any non-fuel hardware, including rod end plugs , is ignored in the act i ve fuel region. This is acceptable since the primary self-shielding occurs in the fuel itself , and there may be some unknown streaming effects through the non-fuel hardware.

This homogenization takes into account the presence of water when calculating the isotopic weight fraction and ho m ogenized density. For t he case with the reactor vessel head in place , the region between the head and the active fuel region is homogenized based on the actual mass of the u pper internals over the entire region. Homogenization of source regions and s h ields is acceptable due to the insignificant effects on the detector response given the model geometry. 4.3 The composition of the containment structure and componen t s are based on the values in the SCALE standard composition library [Reference 3.1]. These material properties are commonly used in s hi elding applications , and are acceptable for modelling the structures and components used to determine the best estimate response at the detector locations.

4.4 The min i mum per i od of decay after reactor shutdown before moving fuel is 60 hou r s [Reference 3.8 , Section 4.3.8]. This calculation assu m es a decay t ime of 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> to allow EAL thresholds to be determined for reactor vessel conditions that exist prior to the commencement of fuel movement which is representative of the applicable operating modes (cold shutdown , refueling). This decay time is appropriate to produce best estimate results for both the head on and head off configurations. 4.5 The hardware in the upper internals region between the active fuel region , reactor recirculating pumps and reactor vessel head is assumed to be stainless steel type 304. While the actual composition of the hardware may vary slightly , small variations in the material will have a negligible effect on the dose rate response at the detectors.

4.6 It is assumed that the water below the active fuel region is liquid at a constant temperature.

Using a density of 0.9982 g/cm 3 is common in shielding Page 6 of 28 ENERCON Excellence-Every project. Every day. Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levelsf----------

< During Refueling for EAL Thresholds REV. 00 applications.

Any water above this region would be steam with little shielding value. 4.7 The source term is generated shortly after shutdown, therefore , the fuel gamma source term will predominate and the neutron-gamma and hardware activation can be neglected.

4.8 The high range detectors read out in roentgen per hour (R/h) which is a measurement of exposure rate, while the MCNP output is provided in mrem/h which is a measurement of the equivalent dose rate that represents the biological effects of ionizing radiation.

It is assumed that 1 R is approximately 1000 mrem. This is acceptable as only the gamma source term is considered.

4.9 The roof of the Reactor Building is modeled as 0.5 inches of stainless steel. This will account for any scattering interactions that may contribute to the response at the detector.

The magnitude of the detector response due to scattering off of the roof will be small due to the geometry and amount of shielding in the model , and is therefore acceptable.

4.10 Automatic Depressurization System Accumulators 1 R003A/B/C located on the 775'-11 Yi" elevation are not included in the model. The size of the accumulators are 200 gallons [Reference 3.26]. This is relatively small compared to the geometry of the model , and the corresponding scatter interactions will not have a significant impact on the detector response.

Page 7 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levelsi-----------------1 ENERCON Excellence

-Every pro1ecr. E*ery da y. 5.0 Design Inputs During Refueling for EAL Thresholds 5.1 Fuel Assembly Parameters REV. 00 The following fuel assembly parameters are used to homogenize the core in the MCNP model. They are based on typical fuel assembly values for 1Ox10 fuel. Table 2 -Design Input Fuel Assembly Parameters Parameters Fuel type # of Assemblies in Core # Fuel rods per assembly Pitch Density (% of theoretical)

Fuel pellet OD Fuel rod OD Clad thickness Active length 5.2 Model Dimensions Value lOxlO 368 92 0.51 95 0.336 0.395 0.026 144 Unit Reference

[in] [in] [in] [in] [in] 3.25 3.11 3.12 3.12 3.12 3.12 3.12 3.12 3.12 The following elevations and dimensions are based on the associated drawings or other reference.

Some parameters are estimated using drawing scales when exact dimensions are not provided. Table 3 -Design Input Dimensions Dimension ft in cm Reference Pedestal inner radius 8 243.84 3.20 Pedestal outer radius 12 365.76 3.20 Reactor vessel inner diameter 185.3 75 470.85 3.15 Reactro vessel thickness 5 12.70 3.15 Drywell spherical portion radius 31.5 960.12 3.17 Figure 2 Concrete around drywell spherical 36 9 1120.14 3.14 portion(x and y directions radius) Drywell cylindrical portion radius 17 518.16 3.16 Drywell liner thickness 0.75 1.91 3.18 Concrete around drywell cylindri-22 9 693.42 3.16 cal portion (x and y directions)

Reactor Building (x and y direc-140 4267.20 3.14 tions) Reactor Building Roof Thickness 0.5 1.27 Assumption 4.9 Height of active fuel 144 365.76 3.12 Vessel Height 704.5 1789.43 3.15 Page 8 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 ENERCON Reactor Vessel Water Levels

  • During Refue l ing for EAL Excelle n ce-Every project. Every day. REV. 00 Thresholds Dimension ft in cm Reference Reactor vessel head thickness 3.9375 10.00 3.19 Distance from vessel O to bottom of 200.94 510.39 3.15 , 3.12 active fuel Bio shield inner radius 9 6.25 290.20 3.21 Bio shield outer radius 11 8.25 356.24 3.21 Reactor recirculating pump height 17 2 523.24 3.24 Reactor recirculating pump radius 2 9 83.82 3.24 Detector RE-9184A distance from 121.92 3 .23 [Scaled] origin (x plane) Detector RE-9184A distance from 13.33 406.29 3 .23 [Scaled] origin (y plane) Detector RE-9184B distance from 6 182.88 3.23 [Scaled] origin (x plane) Detector RE-9184B distance from 365.76 3.23 [Scaled] origin (y plane) Reactor Recirculating Pump IP-12 365.76 3 .23 [Scaled] 201A distance from origin (x plane) Reactor Recirc u lating Pump IP-12 365.76 3 .23 [Scaled] 201A distance from origin (y plane) Reactor Recircu l ating Pump IP--12 -365.76 3 .23 [Scaled] 201B distance from origin (x plane) Reactor Recirculating Pump IP--12 -365.76 3 .23 [Scaled] 201B distance from origin (y plane) Table 4 -Design Input Elevations 2 Dimension:

ft. in cm Reference Drywell Equator 766 0.5 0.00 3.13 Vessel 0 772 5.5 195.58 3.15 Bottom of pedestal elevation 742 9 -709.93 3.13 Top of cylindrical portion of drywell 855 2711.45 3.13 concrete Top of Reactor Building 897 6 4006.85 3.13 Detector elevation 760 -184.15 3.17 Top of pedestal/

bottom of bio shield 770 10.5 147.32 3.20 Top of bio shield 816 3.25 1530.99 3.22 Reactor recirculatin 748 8.5 -528.32 3.13 2 All elevati o ns listed in centimeters are relative to the equator of the drywell elevation of 766' 0.5" ence 3.13]. Page 9 of 28 ENERCON Excellence

-Every project. Every day. Dose Rate Evaluation of CALC NO. N EE-323-CALC-002 Reactor Vessel Water Levels i-----------

1 During Refueling for EAL Thresholds REV. 00 5.3 Core Is o t o pic Inve n t o ry Co re isotopic activities in Ci/MWt are taken from Refe r ence 3.7 Table 1-1. A ta bl e o f the input values is shown in Table 5 , below. The activities in Ci are dete r m in ed by multip l ying by the rated thermal power of 1912 MWt taken from Ref e rence 3.10. Tab l e 5-Core So u rce Term Is otope Ci/MWt Ci Is otope C i/MWt C i Ba-139 4.7 4E+04 9.06 E+07 Rh-105 2.8 1E+0 4 5.37E+0 7 Ba-140 4.76E+0 4 9.10E+07 Ru-103 4.34E+04 8.30E+07 Ce-141 4.39E+04 8.3 9 E+07 Ru-105 3.0 6E+0 4 5.85E+0 7 C e-1 43 4.00E+0 4 7.65E+07 Ru-106 1.55E+04 2.96E+0 7 Ce-144 3.54E+04 6.77 E+07 S b-127 2.39E+0 3 4.57E+0 6 Cm-242 1.12E+0 3 2.14E+06 S b-129 8.68E+03 1.66E+07 Cs-134 4.70 E+0 3 8.9 9 E+06 S r-89 2.41E+04 4.61E+07 C s-136 1.49E+0 3 2.85E+0 6 S r-90 2.39E+03 4.57E+06 C s-137 3.2 5 E+0 3 6.2 1 E+06 S r-91 3.01E+0 4 5.76E+07 1-13 1 2.6 7 E+04 5.l 1 E+07 S r-92 3.24E+04 6.19E+0 7 1-132 3.88 E+04 7.42E+07 Tc-99m 4.3 7 E+0 4 8.36E+07 1-133 5.42E+04 1.04E+08 T e-127 2.36E+03 4.51E+06 1-134 5.9 8 E+04 1.14E+08 Te-127m 3.9 7 E+02 7.59E+0 5 1-1 35 5.18E+0 4 9.90E+0 7 T e-1 29 8.26E+03 1.58E+07 Kr-83m 3.05E+0 3 5.8 3E+06 Te-129m 1.6 8 E+03 3.21E+0 6 Kr-85 2.7 8E+0 2 5.32E+05 Te-131m 5.41E+03 1.03E+0 7 Kr-85m 6.1 7 E+03 1.18 E+07 Te-132 3.8 1E+04 7.28E+07 Kr-87 1.23E+04 2.3 5 E+07 Xe-131m 3.65E+0 2 6.98E+0 5 Kr-88 1.70 E+04 3.2 5 E+07 Xe-133 5.4 3E+04 l.0 4E+08 La-140 4.91E+04 9.39E+07 Xe-133m 1.72E+03 3.29E+0 6 La-141 4.33E+0 4 8.2 8 E+07 Xe-135 1.42E+04 2.72E+0 7 La-142 4.21E+0 4 8.05E+07 X e-135m 1.15E+04 2.20E+0 7 Mo-99 5.30E+04 1.0 1E+08 Xe-138 4.56E+04 8.72E+07 Nb-95 4.50E+0 4 8.60E+07 Y-90 2.45E+03 4.68E+0 6 Nd-147 1.75 E+04 3.3 5 E+07 Y-91 3.1 7 E+04 6.0 6E+07 Np-239 5.69E+05 1.09 E+09 Y-92 3.26E+04 6.23E+07 Pr-143 3.96E+04 7.57 E+07 Y-93 2.52 E+04 4.8 2E+07 Pu-241 4.2 6 E+0 3 8.15E+06 Zr-95 4.44 E+04 8.49E+07 Rb-86 5.29 E+Ol 1.01 E+05 Zr-97 4.2 3 E+04 8.0 9E+07 Page 10 of 28 Dose Rate Eva lu ation of CA LC N O. NEE-323-CALC-002 Reacto r Vessel W at er Levels f--------------~ ENERCON Excellence-Every p ro1ect. E ve ry day. During Refueling for EAL Thresholds 5.4 Mate r ial Com p osi t ions REV. 00 The following compositions used in t he MCNP mo del ar e t aken or developed from the SCALE standard composition library [Refere nc e 3.1] an d are show n i n Table 6. Table 6 -Scale S t andard Com p o si t i ons us e d in M CNP Mo del M at eria l Isotope We i ght Fra c tion Z ry-4 Zr 0.9823 (6.5 6 g/c m 3) Sn 0.0145 Cr 0.0010 Fe 0.0021 Hf 0.0001 U02 U-235 0.0348 (1 0.4 12 g/cm 3) U-238 0.8466 0 0.1186 Air C 0.0001 (1.21E-0 3 g/c m 3) N 0.7651 0 0.2348 Water H 0.1111 (0.9982 g/c m 3) 0 0.8889 SS-3 04 Fe 0.6838 (7.94 g/cm 3) Cr 0.1900 Ni 0.0950 Mn 0.0200 Si 0.0100 C 0.0008 p 0.0004 C oncret e 0 0.5320 (2.3 0 g/c m 3) Si 0.3370 [KEN O Regula r Ca 0.0440 Con cret e S t a nd a rd Al 0.0340 Mix] Na 0.0290 Fe 0.0140 H 0.0100 Car bon S te e l C 0.0100 (7.82 g/cm 3) Fe 0.9900 P a g e 1 1 of 2 8 ENERCON Excellence-Every project. Every riay. 5.5 Upper Internals Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels r----------------l During Refueling for EAL Thresholds REV. 00 The following weights are used in the MCNP model for the region between the active fuel and the reactor vessel head [Reference 3.9 , Appendix 8.9]:

  • The weight of stainless steel for the moisture separator is 83 , 000 lbs.
  • The weight of stainless steel for the steam dryer is 50 , 000 lbs. 5.6 The drywell (9184 A/8) and torus (9185 A/8) radiation monitor ranges (1 to 10 7 R/hr) are taken from Reference 3.6. 5.7 ANSI/ANS-1977 Flux to Dose Factors Flux to dose conversion factors are taken from ANSI/ANS-6.1.1-1977

[Reference 3.4] and are shown in Table 7. Table 7-ANSIIANS-6

.1.1-1977 Flux to Dose Factors MeV mrem/hr/(y/cm 2/s) MeV mrem/hr/(y/cm 2/s) 0.01 3.96E-03 0.8 l.68E-03 0.03 5.82E-04 1 l.98E-03 -,~ -0.05 2.90E-04 2.2 3.42E-03 0.07 2.58E-04 2.6 3.82E-03 Page 12 of 28 ENERCON Excellence-Every proiecr. Every day. 6.0 Methodology Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels ____________ During Refueling for EAL Thresholds REV. 00 The reactor source terms are decayed to 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> with ORIGEN-S of the SCALE 6.1 code package , Reference 3.2. The results are used to bin design input isotope specific activities into energy dependent photon bins. These energy speci fi c photon emission bins are used as i nput for the energy distribution described by the MCNP source definitions.

The MCNP6 , Reference 3.3 , Monte Carlo transport code is used to determine the dose rates via the flux to dose conversion factors i n Table 7 , wh i le accounting for sh i elding and particle transport.

The detailed engineering drawings are converted into MCNP surface and cell cards in the dimensions shown in Table 3 and Table 4. The radiation monitors of interest are modeled as point detectors to determine the expected dose rate for those detectors.

The dose rates are calculated for two reactor refueling conditions

1. With Head -the reactor is modeled with a 3.9375 inch carbon steel plate as indicated in Table 3 , which is addit i onal attenuation between the source and detector. The mass of the moisture separator and steam dryer is homogenized between the active fuel region and the vessel head. 2. Without head -the reactor is modeled with air between the active fuel zone and containment.
3. A sensitivity case is run with a mirror surface at the top of the drywell to ensure the modeling of the drywell cap would not significantly affect the r esponse at the tector locations due to scattering.

Variance reduction is accomplished with a geometric importance map that is imposed on the homogenized core. In addition , cell based importance weighting and source biasing (see Section 7.5) are utilized to improve the variance reduction of the simple geometric scheme. A superimposed weight window mesh is utilized where necessary to improve variance.

The weight windows are iteratively generated u sing the MCNP weight windows generator card. All final dose rates presented in this calculation include weight windows variance reduction.

Page 1 3 of 28 ca ENERCON Excellence

-Every proiecr. Every da y. 7.0 Calculation 7.1 Source Terms Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels t-----------------i During Refueling for EAL Thresholds REV. 00 The ORIGEN-S input deck , DAECEAL.inp , is provided in Appendix C. This input produces a simple case where the isotopic composition from Table 5 is decayed. The isotope is specified in the 73$$ card using the special identifier described i n Section F7.6.2 of the ORIGEN-S manual , and the activity in curies is specified in the 74** card. The time steps for the decay are given on the 60** card i n hours. Although multiple time steps are calculated , the source term with 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> decay time is used in this calculation to model the core shortly after shutdown. The output of the decay i s given in terms of photons/s/Energy-Group , which is automat i cally normal i zed in th e MCNP input. The results of this calculation are summarized below in Table 8. These values are used in the MCNP input source definition.

Table 8 -Binned To t a l Core Source Term E nergy Group E nergy Bounda r ies (Me V) Ph otons/se c 1 0.01-0.05 2.028E+19 2 0.05-0.1 6.57 2E+18 3 0.1-0.2 l.557E+19 4 0.2-0.3 9.672 E+18 5 0.3-0.4 3.582E+1 8 6 0.4-0.6 7.837 E+ 18 7 0.6-0.8 l.373 E+ 19 8 0.8-1 2.13 2E+18 9 1-1.33 4.942E+l 7 10 1.33-1.66 3.579 E+18 11 1.66-2 6.5 7 6E+1 6 12 2-2.5 7.518 E+16 13 2.5-3 l.llOE+l 7 14 3-4 8.689 E+14 15 4-5 l.553E+1 0 16 5-6.5 2.568E+0 8 17 6.5-8 3.7 92E+0 7 18 8-10 8.041 E+0 6 19 10-1 1 4.352E+05 total s 8.37 E+19 Page 1 4 of 28

.-" ENERCON Excellence

-Every pro1ect Every day. Dose Rate Evaluation of CALC NO. N EE-323-CALC-002 Reactor Vessel Water Levels1----

, During Refueling for EAL Thresholds REV. 00 7.2 MCNP Model Core Homogenization The source term is given for the entire core, therefore, the self-shielding from the assemblies is an important part of the dose rate response.

For simplicity, the core is modeled as a three dimensional cylinder with a uniformly distributed spatial particle distribution.

The calculations for determining the mass of fuel , cladding and water for the core and the resulting density are shown below. The inputs are based on the dimensions in Table 2. Assembly Width= (Array Size -1) x pitch+ Rod OD = (10 -1)(0.51in)

+ 0.395in = 4.985 in Active Fuel Region Area= (Assembly Width)2 x Number of Assemb lies in Core = ( 4.985in)2 x 368 = 914 4.883 in 2 At. F lE . l tR d' IActiveFuelRegionArea;

/9144883in2/

c we ue quwa en a zus = '1 rr = '1

  • rr = 53.953 in Rod Volumeu 0 2 = rr(Pellet Radius)2 x Active Length= rr(0.168 in)2 (144 in) = 12.768 in 3 g cm 3 Rod Massuo = p x V = (10.412-)

(12.7682 in 3) (2.54-.-) = 2178.54 g 2 a m Number of Fuel Rods Assembly Massuo = Rod Mass x A bl = (21 78.54 g)(92) 2 ssem y = 200.43 kg (OD2 JD 2) Clad Volume = rr 4 -4 x Active Length [(0.395 in)2 (0.343 in)2] ( . ) . 3 = (rr) 4 -4 144 m = 4.34 m g cm 3 Rod MaSSzry-4 = p x V = ( 6.56 cc) ( 4.34 in 3) ( 2.54 in) = 466.5 g Page 15 of 28 ENERCON Excellence

-Every p ro1ecr. Every day. Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levelsi-----

l During Refueling for EAL Thresholds REV. 00 Number of Fuel Rods Assembly MaSSzry-4 = Rod Mass x A bl = (466.5g)(92)

= 42.92 kg ssem y Assembly H 2 0 Volume = [(Assembly Width)2 -rr(Rod Radius)2 x Number of Fuel Rods] x A ctive Length = [(4.985 in)2 -(rr)(0.1975 in)2(92)](144 in)= 1955 in 3 ( g)(1955)( cm)3 Assembly MassH O = p x V = 0.9982-. 3 2.54-.-= 31.98 kg 2 cc m in Assembly Volume= Active Length x (Assembly Width)2 = (144 in)(4.985 in)2 = 3578.4 in 3 . Total Mass 1000g/kg(200.43

+ 42.92 + 31.98) kg Density= = 3 = 4.70 g/cc Volume 3578.4 in 3 ( 2.54 ~r;:J The corresponding isotopic composition for the homogenized active fuel region is calculated based on the compositions in Table 6. An example calculation for the mass fraction of U-235 is included below. Assembly Massuo Mass Fraction U235 = T I M 2 x weight fraction U235 ota ass 200.43 kg = (200.43 + 42.92 + 31.98) kg X 0.0 348 = 0*0253 The remaining calculations for the homogenization are done in the worksheet Compositions of the EXCEL workbook DAEAL.xlsx and are shown in Appendix B. The isotopic compositions are calculated with the water level above the top of the fuel. Note that the EXCEL workbook uses additional significant figures. Page 16 of 28

,"' ENE RCON Excellence-Every pro;ect. Every d ay. ZAIDNumber 92235 92238 8016 40000 50000 2400 0 26000 72000 1001 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels 1-----------------1 During Refueling for EAL Thresholds REV. 00 Table 9 -Homogenization of Active Fuel Region Atom Mass Fraction Active Fuel Region Homogenized U-235 0.0253 U-238 0.6163 0 0.1896 Zr 0.1531 Sn 0.0023 Cr 0.0002 Fe 0.0003 Hf 0.0000 H 0.0129 7.3 MCNP Model Upper Internals Homogenization For the case with the reactor vessel head in place , the steam dryer and moisture separator region are modeled as a discrete cylinder with a uniformly d i stributed homogenized material to account for the mass of stainless steel between the active fuel height and reactor vessel head. The homogenization accounts for the mass of metal from Section 5.5 (assumed stainless steel type 304 per Assumption 4.5) distributed evenly across the volume between the active fuel height (2=1071.73 cm) and the head (2=1985.01 cm). Mass Upper I nterna l s= (83000 lb+ 500 00 lb) ( 453.59 ~) = 6.033 x 10 7 g The mass is divided by the volume of the region between the act iv e fuel height and the reactor vessel head to determine the density. Density Uppe r Internals

= Mass Upper Internals+

V = 6.033 x 10 7 g + (913.28cm x (rr(235.43cm)2)

= 0.3792._ cc 7.4 MCNP Model Geometry The following MCNP model geometry is based on the containment dimensions summarized in Table 3 and Table 4. The model only focuses on the primary systems and c0mponents that provide shielding or reflection from the core to the radiation monitors.

These components include the reactor vessel, recirculation pumps , pedestal , biological shield and drywell. VISED plots of the model geometry are provided in Figures 1-3. The MCNP surface cards with the model dimensions (cm) are shown in Figure 4 , and the cell cards are shown in Figure 5 for the cases with no reactor vessel head. A VISED plot of the model with the reactor vessel head is shown in Figure 6. Areas that are not of interest Page 17 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels i-----------------1 ENERCON Excellence

-Evtry pro;ecr. Every day. During Refueling for EAL Thresholds REV. 00 are given an importance of zero (white areas) so MCNP will not track particles in locations that will not contribute to the detector response. F ig ure 1 X-Z V I SED Plot of Reactor Vesse l (No Head) __. +-12.70 cm 1789.43 cm + 470.85cm Homogenized Core --Water Page 1 8 of 28 Air R eact o r Vessel Biolog ic al Shield ENERCON Excellence-Every pro1ecr. Every day. Dose Rate Evaluation of CALC NO. N EE-323-CALC-002 Reactor Vessel Water Levelsf-------

1 During Refueling for EAL Thresholds REV. 00 Figure 2 Vised Plot of Drywe/1 and Reactor Building 3 Radiation Monitor Pump Reactor Building Drywell Pedestal Radiat i on Mon i tor Pump 3 Radiation monitors are not on the same plane shown above. They are included for visualization purposes only. The-VISED Plot was rotated around the Z axis unt i l the Recirculating Pumps were visible. Page 19 of 28 Dose Rate Evaluation of CALC NO. N EE-323-CALC-002

.:d ENERCON Reactor Vessel Water Levels Exce ll enc<-Every pro1ecr. Every da y. During Refueling for EAL REV. 00 Thresholds Figure 3 X-Y Vised Plot of Detectors and Reactor Recirculating Pumps at Elevation 760'-0" 4 4 Detectors are included for visualization purposes only. Page 20 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels 1----------

l ENERCON Excellence-Every pro1ecr. Every day. During Refueling for EAL Thresholds REV. 00 Figure 4 MCNP Model Surface Cards 5 c surfaces 1 rec 0 0 705.97 0 0 365.76 1 37.045 2 rec 0 0 208.28 0 0 1776.73 235.43 3 rec 0 0 1 95.58 0 0 1789.43 248.13 4 rpp -1120 .14 1120.14 -1120.14 1120 .14 5 so 960.12 6 so 958.21 7 pz -709.93 8 rec O O -709.93 0 0 857.25 243.84 9 rec O O -709.93 O O 857.25 365.76 81 rec O O 147.32 0 0 1383.67 290.20 91 rec O O 147.32 O O 1383.67 356.24 -1120.14 82 rec 365.76 365.76 -528.32 0 0 523.24 83.82 92 rec -365.76 -365.76 -528.32 0 0 523.24 83.82 10 pz 1 95.58 11 pz 821.86 12 rec O O 821.86 0 0 1889.59 518.16 13 rec O O 821.86 O O 1889.59 516.25 821.86 14 rpp -693.42 693.42 -693.42 693.42 821.86 2711.45 15 pz 1071. 73 16 pz 1985.01 17 rpp -4267.2 4267.2 -4267.2 4267.2 2711.45 4006.85 18 rpp -4267.2 4267.2 -4267.2 4267.2 4006.85 4008.12 19 pz 147.32 20 pz 1530.99 28 rec O O 1985.01 0 0 10.00 248.13 1 01 pz 742.546 102 pz 779.122 1 03 pz 8 1 5.698 104 pz 852.274 1 05 pz 888.85 106 pz 925.246 107 pz 962.002 108 pz 998.578 109 pz 1035.154 110 pz 1071. 73 $ Active Fuel Region $ Reactor Pressure Vessel Inner Surface $ Reactor Pressure Vessel Outer Surface $ Concrete Spher port drywell outer $ Spher portion of drywell outer surface $ Spher portion of drywell liner surface $ Bottom of Pedestal Elevation

$ Pedestal Inner Surface $ Pedestal Outer Surface $ Bio Shield Inner Surface $ Bio Shield Outer Surface $ Recirc Pump IP-201A $ Recirc Pump IP-201B $ Vessel O $ Transition Spherical to Cylindrical

$ cylin port drywell concrete surface $ cylin port drywell liner surface $ Concrete cylin port drywell outer $ Water Elevation Surface $ Top of RPV (head level) $ Reactor building above drywell $ Reactor building roof $ Top of Ped Elevation/Bottom Bio Shield $ Top of Ped Elevation/Bottom B i o Shield $ Reactor Head 5 The surface card for the MCNP model without the reactor vessel head does not have surface 28. Page 21 of 28 ENERCON Excellence-Every pro1ect. Every day, C cells 101 1 -4.49 101 102 1 -4.49 -1 101 -102 103 1 -4.49 -1 102 -103 104 1 -4.49 -1 103 -104 105 1 -4.49 -1 104 -105 106 1 -4.49 -1 105 -106 107 1 -4. 49 -1 106 -107 108 1 -4.49 -1 107 -108 109 1 -4. 49 -1 108 -109 llO 1 -4.49 -1 109 -llO 2 2 -0.9982 1 15 3 3 -l.21E-03 15 -2 4 4 -7.94 2 16 7 5 -2.3 5 -4 8 5 -2.3 -14 12 9 5 -2.3 -9 8 7 -19 91 5 -2.3 -91 81 19 -20 10 5 -2.3 7 ll 3 -l.21E-03 -8 12 3 -l.21E-03 -6 7 -ll 9 #18 #19 #91 13 3 -l.21E-03 -13 3 #91 14 3 -l.21E-03 -17 15 4 -7.94 2 -18 16 4 -7.94 6 ll 17 4 -7.94 13 -12 18 4 -7.94 -82 1 9 4 -7.94 -92 999 0 1 #2 #3 #4 #7 #8 #9 #15 #16 #17 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels r--------------------1 During Refueling for EAL Thresholds REV. 00 Figure 5 MCNP Model Cell Cards (No Head) imp: p;256 $ Active Fuel Region imp: p;l28 $ Active Fuel Region imp:p;64 $ Active Fuel Region imp:p;32 $ Active Fuel Region imp: p;l6 $ Active Fuel Region imp: p;8 $ Active Fuel Region imp: p;4 $ Active Fuel Region imp:p;3 $ Active Fuel Region imp: p;2 $ Active Fuel Region imp: p;l $ Active Fuel Region imp:p;256 $ Water Region imp:p;256 $ Air Region inside vessel imp:p;256

$ RPV Shell imp:p;256

$ Concrete Surrounding RPV spherical imp:p;256

$ Concrete Surrounding RPV cylindrical imp: p;256 $ Pedestal imp: p;256 $ Bio Shield imp: p;256 $ Concrete at bottom of pedestal imp: p;256 $ Inside Pedestal Air 3 imp: p;256 $ Inside Spherical portion Air imp: p;256 $ Inside Cylindrical portion Air imp: p;256 $ Reactor Building above drywell Air imp: p;256 $ Reactor Build Roof Stainless Steel imp: p;256 $ Containment Liner Spherical portion imp: p;256 $ Containment Liner Cylin portion imp:p;256

$ Recirc Pump IP-201A imp:p;256

$ Recirc Pump IP-2018 #10 #ll #12 #13 #14 #18 #19 #91 imp: p;O $ Problem Boundary Page 22 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water

  • ENERCON During Refueling for EAL Thresholds Exceflenc~-Ew:ry project. Every day. REV. 00 Figure 6 X-Z V SED Plot of Reactor Vessel (With Head) Reactor Vessel Head Cell for the homogenization of the Upper Internals stainless steel 0.379 g/cm 3 7.5 MCNP Source Definition Homogenized Core The core source term is modeled as uniformly distributed throughout the homogenized core, and has an energy spectra based on the decayed core inventory (Section 7.1). Only the gamma source term is taken into account for this evaluation.

The source term is generated shortly after shutdown, therefore, the fuel gamma source term will predominate , and the neutron-gamma and hardware activation source terms can be neglected (Assumption 4.7). The source is defined on the MCNP sdef card using Page 23 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels 1------------------1 ENERCON Excellence-Every proJect. Every day. During Refueling for EAL Thresholds REV. 00 distributions to define the particle location and energy. The radius o f the core is defined with the rad parameter , which automatically creates a uniform distribution based on a cylindrical geometry.

The ext and axs parameters define the direction and distance of the cylinder axis. These parameters combined define the core where the particles can be born. The erg parameter defines the energy spectrum of source part i cles , and i s based on the results of the ORIGEN-S calculation discussed previously.

This distribution is a histogram of energies represented by activities.

These are automatically normalized by MCNP to create a probability distribution.

The total activity is preserved in the tally multiplier.

The MCNP source definition cards are shown below in Figure 7. The sb card is a source biasing card , which in this case biases the particle generation to the lower end of the core. This is a variance reduction technique to improve the statistical certainty in the results. Figure 7 MCNP Source Definition Ca r ds sdef rad=dl ext=d2 axs=O O 1 erg=d8 sil 137.045 si2 h O 742.546 779.122 815.698 852.274 888.85 925.246 962.002 998.578 1035.154 1071.73 sp2 0 1 1 1 1 1 1 1 1 1 1 sb2 O 1 1 0.1 0.1 0.1 0.01 0.01 0.01 0.001 0.001 c Fuel Gamma Spectra f-Source Definition Card -Radius= dl -Extent= d2 -Axis= +Z -Energy= d8 f-Core Radius Distribution f-Core Axial Distribution f-Actual Uniform Distribution f-Biased to Bot Distribution si8 h l.OOOe-002 5.000e-002 l.OOOe-001 2.000e-001 3.000e-001 4.000e-001 f-Source Energy Groups 6.000e-001 8.000e-001 l.OOOe+OOO l.330e+OOO l.660e+OOO 2.000e+OOO 2.500e+OOO 3.000e+OOO 4.000e+OOO 5.000e+OOO 6.500e+OOO 8.000e+OOO l.OOOe+OOl l.lOOe+OOl sp8 O.OOE+OO 2.028E+l9 6.572E+l8 l.557E+l9 9.672E+l8 3.582E+l8 7.837E+l8 f-source Emission on Energy Basis l.373E+l9 2.1 32E+l8 4.942E+l7 3.579E+l8 6.576E+l6 7.518E+l6 l.110E+l7 8.689E+l4 l.553E+l0 2.568E+08 3.792E+07 8.041E+06 4.352E+05 Page 24 of 28 Dose Rate Evaluation of CALC NO. N EE-323-CALC-002 Reactor Vessel Water Levels t-------------------i ENERCON Excelfenc~-Every pro1ecr. Every day. During Refueling for EAL Thresholds 7.6 MCNP Tally Specification REV. 00 The tallies used in this evaluation are point detectors placed at approximate locations of radiation monitors RE-9184A , and RE-91848.

Point detectors are chosen because they use quasi-deterministic dose calculations that will provide better results than surface or cell based tallies that require the particles to enter those regions. The inputs to this card are the coordinates of the dose points followed by an exclusion zone to reduce variance , as well as a multiplier card , which represents the total core activity in photons/sec.

The tally cards are shown in Figure 8. F i gure 8 MCNP Tally Cards f5c RE-9184A , and 9184B f5:p -1 21.92 406.29 -184.15 20 182.88 -365.76 -1 84.15 20 fm5 8.370E+l9 f-Tally Comment Card f-Tally 5 (point detector) x y z exclusion f-Tally Multiplier (Total Activity)

In addition , the flux is multiplied by ANSI/ANS flux-dose conversion factors [Reference 3.4). This is specified in MCNP using the de/df cards. These are shown in Figure 9. Figure 9 ANSI/ANS-6.1.1-1977 Gamma Flu x to Dose Conversion Factors C ------------------------

C ANSI/ANS-6.1.1-1 977 c Gamma Flux to Dose Conversion Factors c (mrem/hr)/(photons/cm2

-s) C ---------------------------

de0 .01 .03 .05 .07 .10 .15 .20 .25 .30 .35 .40 .45 .50 .55 .60 .65 .70 .80 1. 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5. 5.25 5.75 6.25 6.7 5 7.5 9. 11. dfO 3.96E-03 5.82E-04 2.90E-04 2.58E-04 2.83 E-04 5.0lE-04 6.31E-04 7.59E-04 8.78E-04 9.85E-04 l.17E-03 1. 27E-03 l.36E-03 l.44E-03 l.52E-03 l.98E-03 2.51 E-03 2.99 E-03 3.42 E-03 3.82E-03 4.4 1E-03 4.83E-03 5.23E-03 5.60E-03 5.80E-03 6.37E-03 6.74E-03 7. llE-03 7.66E-03 8. 77E-03 Page 25 of 28 3.79E-04 l.08E-03 l.68E-03 4.0lE-03 6.0lE-03 l.03E-02 f-Energy Bins for Flux to Dose Conversion f-Energy Dependent Flux Multipliers ENERCON Exce lle n c~-Every pro j ecr. E very da y. Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water During Refueling for EAL Thresholds REV. 00 7.7 MCNP Material Cards The MCNP material cards are provided in Figure 9. These are based on the compositions described in Table 6 or calculated in Section 7.2. ml m2 m 3 m4 m5 m 6 92235 -0.0253 92238 -0.6163 8016 -0.1896 40000 -0.1531 50000 -0.0023 24000 -0.0002 26000 -0.0003 1001 -0.0129 100 1 2 8016 1 6012 -0.00 0126 7014 -0.76508 8016 -0.234793 6000 -0.0008 14000 -0.0 1 15031 -0.00045 24000 -0.19 25055 -0.02 26000 -0.68375 28000 -0.095 26000 -0.014 1001 -0.01 13027 -0.034 20000 -0.044 8016 -0.532 14000 -0.337 11 023 -0. 029 6012 -0.01 26056 -0.99 Figure 10 MCNP Material Cards $ Homogenized Active Fuel Region $ Water $ Air $ ss 304 $ Reg-Concrete $ Carbon Steel Page 26 of 28 Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levels f---------

l ENERCON Excellence

-Every pro1ect. Every day. 7.8 Results During Refueling for EAL Thresholds REV. 00 The dose rates are provided in Table 10 for the water level at the top of the fuel assemblies.

The dose rate is slightly above the detectable response of 1 R/h (1 E+03 mrem/h) for the no head configuration , and below the detectable response for the configuration with the r eactor vessel head in place for one of the detectors. The sensitivity case shows that there is no sign i ficant impact due to reflection from t he drywell cap. Table 10-Dose R at e Response (mr em l h) Configuration Dose Rate 1 fs d 6 D ose R ate 2 fs d Tally Fi l e RE-91 84 A RE-9184 B No H e ad l.8 1E+0 3 10.81% 1.68 E+03 7.3 1% d O ndm Wit h Head l.11 E+0 3 1 0.16% 7.41 E+02 8.2 4% dOhgm Wi t h H e ad (Se n-l.0 7 E+0 3 15.2 7% 7.67 E+02 15.51% d Or dm s iti v i ty C ase} 8.0 Computer Software This calculation uses ORIGEN-S of the SCALE Version 6.1.2 code package [Reference 3.2] and MCNP Version 6.1.0 [Reference 3.3] in accordance with CSP 3.09. 6 Fraction standard deviation. Page 27 of 28 ENERCON Excellence-Every project. Every da y. 9.0 Impact Assessment Dose Rate Evaluation of CALC NO. NEE-323-CALC-002 Reactor Vessel Water Levelsl---

___________ ____, During Refueling for EAL Thresholds REV. 00 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 determination of core uncovery in accordance with NEI 99-01 Rev. 6. Page 28 of 28 ENERCON Excellence-Every projecr. Every day. Origen o u tput: 07/26/2017 04: 19 PM MCNP output: Directory of \No head\ 08/16/2017 09: 13 AM Directory of \With Head\ 08/1 6/2017 10: 01 AM D i rectory of \sensitivity\

08/16/2017 03: 54 AM Appendix A Electronic File Listing 82 , 114 DAECEAL.OUT 327 , 680 dOnao 1 , 269 , 760 dOhgo 286 , 720 dOrdo Page 1 of 1 CALC NO. REV. N EE-323-CALC-002 00 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 A ENERCON Excellenc e-Every p roject. Ev ery doy. B C D Material Isotope Wei&ht Fraction Zr;-4 Zr 0.98 2 3 {6.5 6 g/cm.3) Sn 0.0 1 45 Cr 0.00 1 Fe 0.002 1 Hf 0.000 1 U0 1 U-23 5 0.034 8 U-23 8 0.8466 0 0.1186 Air C 0.000 1 {1.2 1E-0 3 N 0.7 651 g/crrt3) 0 0.2 3 4 8 Water H 0.1111 {0.998 2 g/crn 3) 0 0.8889 SS-304 Fe 0.6838 {7.94 g/crn 3) C r 0.19 N i 0.095 M n 0.02 S i 0.0 1 C 0.0008 p 0.0004 Concrete 0 0.5 32 {2.30 g/cm.3) Si 0.3 37 Ca 0.044 Al 0.034 N a 0.02 9 Fe 0.0 1 4 H 0.0 1 Carbon Steel C 0.0 1 {7.8 2 g/cm.3) Fe 0.99 Appendix B DAEAL.xlsx Sheets E F G H Mas Reference Material s (KG) [1] U0 2 200.4 Zry-4 42.9 2 Water 3 1.9 8 [1] [1] [1] [1] T [1] t [1] Pa g e 1 o f 2 l CALC NO. REV. ZAIDNumber 92235 92238 8 01 6 40000 50000 2 4 000 26000 72000 100 1 t NEE-323-CALC-002 00 K L Mass Fraction Active Fuel Atom Region Homogenized U-235 0.0253 U-238 0.6 1 63 0 0.1 896 Z r 0.153 1 S n 0.0023 C r 0.0002 F e 0.0003 H f 0.0000 H 0.0129 1.0000 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 A JJ B Material Zry-4 (6.56 gl cm 3) U0 2 Air (1.2 IB-03 g/cni3) Water (0.998 2 g/cm 3) SS-304 (7;94 g/cm 3) Concrete (2.30 g/cm 3) Carbon Steel (7.82 g l cm 3) ENERCON Excellence-Every projecr. Every day. C D E Isotope Weight Reference F ract i on Zr 0.98 2 3 [l] Sn 0.0 1 45 C r 0.00 1 F e 0.0021 Hf 0.000 1 U-23 5 0.034 8 [1] U-23 8 0.8 466 0 0.1186 C 0.0001 [1] N 0.7 6 5 1 0 0.234 8 H 0.1111 [1] 0 0.8 88 9 Fe 0.6838 [l] C r 0.19 Ni 0.095 : M n 0.02 S i 0.0 1 C 0.0008 p 0.0004 0 0.532 [l] Si 0.337 Ca 0.044 Al 0.034 N a 0.02 9 Fe 0.0 1 4 H 0.0 1 C 0.01 [l] F e 0.9 9 CALC NEE-323-CALC-002 Appendix B NO. DAEAL.xlsx Sheets REV. 00 F G H K l 1\fa t erial Mass (KG) ZAI D Mass Fraction Active Fuel Region Homogenized Number Atom 0 0 2 200.42 92235 U-235 =(H 3/S UM(H 3:H 5))*08 Zry-4 42.9 2 92238 U-238 =(H 3/S UM(H 3:H 5))*09 Water 3 1.98 80 1 6 0 =((H3/(S UM(H 3:H 5)))*D l0)+((H5/(S UM(H 3: H5))))*D15 40000 Zr =($H$4/S UM($H$3:$H$5))*D 3 50000 S n =($H$4/S UM($H$3:$H$5))*D4 2 4000 C r =($H$4/S UM($H$3:$H$5))*D 5 2 6000 F e =($H$4/S UM($H$3:$H$5))"0 6 72 000 H f =($H$4/S UM($H$3:$H$5))*D7 1001 H =(H 5/S UM(H 3:H 5))*D 14 =S UM(l3:Ul)

+-' t L f l l I t t t Page 2 of 2 ENERCON Appendix C SCALE Input CALC NO. Excellence-Every pra j ocr. Every da y. REV. =o r igen s 0$$ all 71 e t BW R S o urc e T erm D AE C E AL Ana l y si s 3$$ 2 1 1 1 a4 27 al6 4 a33 19 et 35$$ 0 t 54$$ a8 0 all 2 e 56$$ 0 6 a6 1 a l O 0 a l 3 63 3 3 0 2 0 e 57** 0 a3 1-16 e 95$$ 0 t DA E CEA L Ci Source Te r ms 60** 0 2 4 40 50 60 70 61** 5r l-8 1+6 1+4 65$$ 'GRAM-A T OMS GRAMS CURIES WATTS-A LL WATTS-GAMMA 3 Z 0 1 0 1 0 0 1 0 0 3Z 6Z 3Z 1 1 1 1 0 1 1 1 1 3Z 6Z 3 Z 1 1 1 1 1 1 1 1 1 3Z 6Z 81$$ 2 0 26 1 e 82$$ f2 83** l.10E+07 l.OOE+07 8.00E+06 6.50E+06 5.00 E+06 4.00 E+06 3.00E+06 2.50E+06 2.00E+06 l.66E+06 l.33E+06 l.OO E+06 8.00E+05 6.00E+05 4.00E+05 3.00E+05 2.00E+05 l.OO E+05 5.00 E+04 l.OO E+04 e 84** 2.00E+07 6.43E+06 3.00E+06 1. 85 E+06 l.40E+06 9.00E+05 4.00E+05 l.OO E+05 l.70 E+04 3.00E+03 5.50E+02 l.OOE+02 3.00E+O l l.OO E+Ol 3.05E+OO l.77E+OO l.30E+OO l.13E+OO l.OOE+OO 8.00E-01 4.00E-01 3.25 E-0 1 2.25 E-01 l.OOE-0 1 5.0 0E-02 3.00 E-02 l.OO E-02 l.OOE-05 e 73$$ 561390 561400 581410 581430 58 1 440 962420 551340 551360 55 1 370 5 31310 531320 531330 531340 53 1 350 360831 360850 360851 360870 360880 571400 571410 571420 420990 410950 60 1 470 932390 591430 942410 370860 451050 441030 441050 44 1 060 511270 511290 380890 380900 380910 380920 430991 521270 521271 521290 521291 521311 521320 541311 541330 541331 541350 541351 541380 390900 390910 390920 390930 400950 400970 74** 9.06E+07 9.1 0E+07 8.39E+07 7.65E+07 6.7 7 E+07 2.1 4E+06 8.99 E+06 2.85 E+06 6.2 1 E+06 5.11E+07 7.42E+07 l.04E+08 l.14E+08 5.83 E+06 5.32E+05 l.18E+07 2.35 E+07 3.25 E+07 9.39E+07 8.05E+07 l.0 1 E+08 8.60E+07 3.35 E+07 l.09E+09 7.57E+07 1. 0 1E+05 5.3 7E+07 8.30E+07 5.85 E+07 2. 96 E+07 4.57 E+07 4.61 E+07 4.57 E+06 5.76E+07 6.19 E+07 8.36 E+07 4.51E+06 l.5 8 E+07 3.21 E+06 l.03E+07 7.28 E+07 6.98 E+05 l.04E+08 2.7 2E+0 7 2.20 E+07 8. 72E+07 4.68 E+07 6.06 E+07 6.23E+07 8.49 E+07 8.09E+07 75$$ 3 3 3 3 3 2 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 2 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 t 56$$ fO t e n d Page 1 of 1 9.90E+07 8.28 E+07 8.15E+06 l.66 E+07 7.59E+05 3.29E+06 4.82E+07 3 2 3 3 3 3 N EE-323-CALC-002 00 CALC NEE-323-CALC-002 Attachment 1 NO. I,_ ENERCON CALCULAT I ON PREPARATION

  • CHECKLIST Excellence-Every project. Every day. REV. 0 CHECKLIST ITEMS1 YES NO N/A G ENE RA L R EQUIREMENTS
1. If the calculation is being performed to a client procedure , is the procedure being D D used the latest revision?

The Calcu l ation is performed in accordance with ENERCON procedures.

2. Are the proper forms being used and are they the latest revision? D D The Ca l cu l ati o n is performed in accordance with ENERCON procedures.
3. Have the appropriate client review forms/checkl i sts been completed?

D D OAR will b e performed after calculation subm i ttal 4. A r e all pages properly identified with a calculation number , calculation revision and D D pa g e number consistent w i th the requirements of the client's procedure?

5. Is all information l egible and reproducible? D D 6. I s the calculation presented in a logical and orderly manner? D D 7. Is there an existing calculation that should be revised or vo i ded? D D There is n o existing ca l cu l ation that sh o u ld be revised o r v oided. 8. Is it possible to a l ter an exist i ng calculation instead of preparing a new ca l culation for D D th i s s ituation?

N o existing ca l culati o n would be applicab l e. 9. If an existing calculation is being used for design inputs , are the key design inputs , assumptions and engineering judgments used in that calculat i on valid and do they D D appl y to the calculation revision being performed.

No existing ca l culation is used for design i nputs 10. I s the format of the calculation consistent with applicable procedures and D D expe c tations? 11. W ere design inp u t/o utput doc u ments proper l y updated to reference this calculation?

D D There a r e no design output documents. 12. Can the calcula t ion logic , methodol o gy and presentation be properly understood D D without referring back to the orig i nator for clarification?

OBJ E CTIVE A ND S COP E 13. Does the calculation provide a clear concise s t atement of the problem and objective D D o f the calculation?

14. D o es the ca l c ul ation provide a c l ear statement of quality class i fication? D D 15. Is the reas o n for performing and the end use of the calculation understood? D D 16. Does the calcu l ation provide the basis for information found in the plant's license D D basi s? This does n o t provide basis for license basis 17. If s o , is this d o c u mented in the ca l culation?

D D Page 1 of 5 C A LC NEE-323-CALC-002 Att a chment 1 NO. cl._ ENERCON CALCU LAT ION PREP ARATI ON I CHECKLIST Excellence-Every profecr. Every day. REV. 0 CHECKLIST ITEMS 1 YES NO NIA See above 18. Does the calculation provide the basis for information found in the plant's design D D basis documentation?

This does n o t provide basis for design basis 19. I f so , is this documented in the calculation?

D D See above 20. Does the calculation otherwise support information found in the plant's design basis D D d ocumentation?

This does n o t provide support for information found in design basis documentation

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

D D See above 22. Has the appropriate design or license basis documentation been revised , or has the D D change notice or change request documents being prepared for submittal?

See above DESIGN INPUT S 23. Are d esign inp u ts cl early identified? D D 24. A r e design inputs retrievable o r h a v e they been added as attachments? D D 25. I f Attachments are used as design inputs or assumptions are the Attachments D D traceable and verifiable?

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

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

The D esign Information Transmittal is included as an Attachment is properly referenced in the calculation

28. Are input s o urces (including in du st ry code s a nd standards) app ro pr i ately se l ected D D and are they consistent with the quality classification and objective of the ca l cu l ation? 29. Are input sources (including industry c o des and standards) consistent w i th the plant's D D design and license basis? 30. I f a p p l icable , do des i gn inputs adequately address actual plant conditions? D D 31. Are input values reasonable an d correctly a p p li ed? D D 32. Are d e s ign in pu t sou rces appr ov e d? D D The Des ig n I nformat io n Transmittal contains informat io n fr o m a superse d ed calculation. 33. D o es the c alcu l ation reference the latest revision of the design input source? D D The ca l cu l at i on uses information from a superseded calculation. This information is provided in a Design Information Transm i ttal. 34. Were all applicable plant operating modes cons i dered? D D ASSUMPTIONS Page 2 of 5 C A LC NEE-323-CA L C-002 Attac h ment 1 NO. ENERCON CALCULATION PREPARATION I CHECKLIST

&cellence-Every pro j ect. Every da y. REV. 0 CHECKLIST ITEMS1 YES NO N/A 35. Are assump t ions reasonable/appropriate to the objective?

[gl D D 36. I s adequate justification/basis for all assumptions provided?

[gl D D 37. Are any engineering judgments used? D [gl D Engineering judgement not used as design input. 38. Are engineering judgments clearly identified as such? D D [gl Engineering Judgement is not used as a design input. 39. If engineering judgments are uti l ized as design inputs , are they reasonable and can they be quantified or substantiated by reference to site or industry standards , D D [gl engineering princ i ples , physical laws or other appropr i ate cr i teria? Engineering Judgement is n o t used as a design i nput. MET H ODOLOG Y 4 0. I s t he methodo lo g y used in the ca l c u lation desc r ibed or imp l ied i n the plant's D D [gl l i censing basis? The scope of calculation is outs i de of p l ant lic e nsing basis 41. I f the methodo l ogy used differs from that descr i bed in the plant's licensing basis , has D D t h e appropriate li ce n se document c h ange notice been initiated?

see abo v e. 42. Is the methodology used consistent with the stated objective? D D 43. Is the methodology used appropriate when considering the qua l ity class i fication of D D the calculation a nd i n tended use o f t h e resu l ts? BODY OF C A LCULA TIO N 44. Are equations used in the calculation consistent with recogn i zed engineering practice D D and t h e plant's design and license basis? 4 5. I s t h ere reasona b le j ustificat i on pro v ided for the use of equat i ons not in common D D use? There are no uncommon equat i ons used in the calcu l ation. 46. Are the mathematical operations performed properly and documented in a logical D D fashion? 47. I s the math performed correctly? D D 48. Ha v e adjustment factors , uncertainties and empirical correlations used i n the analysis [gl D D been c o rrectl y applied? 49. Has proper cons id eration been g iv en to res ul ts that may be overly sensitive to very D D small changes in input? SOFTWARE/COMPUTER CODES 50. Are c o mputer c o des or software l anguages used in the preparation of the D D calculation?

MCNP an d Scale are use d Page 3 of 5 CALC NEE-323-CALC-002 Attachment 1 NO. I, ENERCON CALCULATION PREPARATION

  • CHECKLIST Exce lfe nc e-Ev.ry pr oj ecr. E v ery d ay. REV. 0 CHECKLIST ITEMS 1 YES NO N/A 51. Have the r equirements of CSP 3.09 for use of computer codes or software [8J D D l anguages , includ i ng verification of accuracy and applicability been met? 52. Are the codes properly identified along w i th source vendor , organization , and revision [8J D D level? 53. Is the computer code applicable for the analysis being performed?

[8J D D 54. If applicable , does the computer mode l adequately consider actual plant conditions?

[8J D D 55. Are the inputs to the computer code clearly identified and consistent with the inputs [8J D D and assumptions documented in the calcu l ation? 56. Is the computer output clearly identified?

[8J D D 57. Does the computer output clear ly i d entify the appropriate units? [8J D D 58. Are the computer outputs reasonable when compared to the inputs and what was [8J D D expected?

59. Was the computer output reviewed for ERROR or WARNING messages that could [8J D D i n v alidate the results? RESU L T S AND CONCL USIO NS 60. I s a d equate acceptance criteria s p ecified? D D [8J There is no acceptance criteria as discussed in calc. 61. Are the stated acceptance criteria consistent with the purpose of the calculation , and D D [8J intended use? See above 62. Are the stated acceptance criteria cons i stent with the plant's design bas i s , applicable D D [8J l icen s ing commitments and industry codes , and standards?

See abo v e 63. Do the calculat io n results and conclusi o ns meet the stated acceptance criteria?

D D [8J See ab o ve. 64. Are t h e results represented in the proper units with an appropriate tolerance , if [8J D D a p plicable?

65. Are the calculation results and conclusions reasonable when considered against the [8J D D stated inputs and objectives?
66. Is sufficient conservatism applied to the outputs and conclusions?

[8J D D Page 4 of 5 CALC N EE-323-CALC-002 Att achm ent 1 NO. I, ENERCON CALCULATI ON PREPARATION I CHECKLIST

&ceflence-Evtry pro fec r. Every day. REV. 0 CHECKLIST ITEMS 1 YES NO NIA 67. Do the calculation results and conclusions affect any other calculations?

D D No other calculat i ons are affected by this calculat i on. 68. If so , have the affected calculations been revised? D D No other calculat i ons are affected by this ca l culat i on. 69. Does the calculation contain any conceptual , unconfirmed or open assumptions D D requiring later confirmation?

There are n o o pen assumpti o ns requiring c o nfirmation later. 70. I f s o , are they properly identified?

D D There are no open assumptions requiring confirmation later. D ES IGN RE V I EW 71. Have alternate ca l culation methods been used to verify calcu l ation results? D D No a Design Review was performed.

Note: 1. Where required , prov i de clarification~ustification for answers to the questions in the space provided below each ques t i o n. An explanati o n is required for any questions answered as " No' or " N/A". O r igin at or: J ay Bhatt Print Name and Sign Date Page 5 of 5

Retrieved from "https://kanterella.com/w/index.php?title=NG-17-0235,_Duane_Arnold_Energy_Center,_Calc_No._NEE-323-CALC-002,_Dose_Rate_Evaluation_of_Reactor_Vessel_Water_Levels_During_Refueling_for_EAL_Thresholds&oldid=905171"