Text

License Amendment Request to Revise Emergency Action Level Scheme: Attachment 4, Supporting Calculations for EAL Thresholds, Part 5 of 5
	ML17095A120
Person / Time
Site:	Monticello
Issue date:	03/31/2017
From:	; Northern States Power Company, Minnesota, Xcel Energy
To:	; Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML17095A107	List: L-MT-17-012, License Amendment Request to Revise Emergency Action Level Scheme: Emergency Action Level Matrix, Hot Conditions (Attachment 5); ML17095A112; ML17095A113; ML17095A114; ML17095A115; ML17095A118; ML17095A120; ML17095A121;
References
	L-MT-17-012 CA-04-202
	Download: ML17095A120 (80)
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CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" without Attachments (80 total pages) QF-0549 QF-0547 QF-0528 TOC Calculation Description 04-202 Rev 1 TABLE OF CONTENTS Page 1 of 1 Calculation Signature Sheet External Design Document Suitability Review Checklist . . Design Review Comment Form

Table of Contents Sargent & Lundy Calculation 2004-07061' Rev 2 Pages 5 2 5 1 2599 Total 2612

____________ "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ISSUE

SUMMARY

Form SOP-0402-07.

Revision 8 DESIGN CONTROL

SUMMARY

CLIENT: NSP UNIT NO.: 1 PAGE NO.: 1 PROJECT NAME: Monticello Nuclear Generating Plant PROJECT NO.: 12400-045 I S&L NUCLEAR QA PROGRAM I CALC.NO .. : 2004-07061 APPLICABLE 181 YES 0 NO TITLE: Dose Rates to CHRRM Detectors Due to Drop in RPV Waler Level EQUIPMENT NO.: IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED

& REVIEW METHOD This is the initial issue of this calculation (Le., Revision 0) INPUTS/ ASSUMPTIONS 121 VERIFIED 0 UNVERIFIED REVIEW METHOD: Alternate Method REV.: 0 STATUS: [gJ APPROVED 0 SUPERSEDED BY CALCULATION NO. DVOlD DATE FOR REV.: 1012112004 PREPARER:

Anthony G. Klazura {Signa1ure on File) DATE: 10/21/2004 REVIEWER:

John M. Rich {Signature on File) DATE: 1012112004 APPROVER:

W. J. Jonnson (Signature on File) DATE: 1012112004 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED

& REVIEW METHOD This is a complete revision of the calculation.

The calculation is being revised lo incorporate the Alternative Source Term INPUTS/ ASSUMPTIONS (AST) that was calculated for the Montioello Nuclear Generating Plant in Calculation 2004.07600.

18] VERIFIED 0 UNVERIFIED REVJEW METHOD: Detailed REV.: 1 STATUS: [gJ APPROVED 0 SUPERSEDED BY CALCULATION NO. OVOID DATE FOR REV.: 9/.26/2007 PREPARER:

Anthony G. Klawra (Signature on File) DATE: 9/26/2007 REVIEWER:

D. L. Marsh (Signature on File) DATE: 9/26/2007 APPROVER:

W. J. Johnson (Signature on File) DATE: 912612007 JDENTIFICATION OF PAGES ADDEDIREVISED/SUPERSEDEDNOIDED

& REVIEW METHOD This is a complete revision of the calculation.

Pages 1 through 79 and Attachmen!s 1 through 21 (pages 1-1 to 1-3, 2-1 lo 2-1. 3-1 lo 3*9, 4* 1 to 4-3, 5-1 to 5-2, 6-1 to 6-2, 7 -1 to 7-81, 8-1 to 8-176, 9-1 to 9-109, 10-ito 10-137, 11-1 to 11-257, 12*1 to 12-183, 13*1 lo 13-200, 14-1 to 14-INPUTS/ ASSUMPTIONS 148, 15-1 to 15-158, 16-1 to 16-264, 17-1 to 17-249, 18-1 to 18-120, 19-1 to 19-127, 20-1to20-160, and 21-1 to 21-131, f81 VERIFIED Respectively).

Total of 2599 pages. 0 UNVERIFIED REVIEW METHOD: Detailed REV.: 2 STATUS: [gJ APPROVED

0 SUPERSEDED BY CALCULATION NO. OVOID DATE FOR REV.: PREPARER:

-*Anthony G. Klazura / .,LI, DATE: S, /// 'J.Ji// REVIEWER:

D. L. Marsh

(/ DATE: t:; LfoL/[ APPROVER:

W. J. Johnson / /.,,./A/J/"/

DATE: -:;: h:z..L LI ; I --:r .. !JOT!=* PRINT ANO Slr.N IN THF Slr.lt.JATI IRF ARJ=:AR SOP040207 .DOC Page 1of79 RAV n:>tA* n1-?i:J-::>n10 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 2 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date TABLE OF CONTENTS 1.0 Purpose and Scope ................................................................................................................

4 1.1 Purpose ..........................................................................................................................

4 1.2 Scope .............................................................................................................................

4 2.0 Acceptance Criteria ..............................................................................................................

5 3.0 Design Inputs ........................................................................................................................

6 3.1 Reactor Core Source Term ............................................................................................

6 3.2 Reactor Core Operating Parameters

..............................................................................

8 3.3 Reactor Core and Fuel Design Data ..............................................................................

9 3 .4 Radiation Detector Information

....................................................................................

10 3.5 Elevations

......................................................................................................................

10 3 .6 Component Materials and Dimensions

.........................................................................

11 3.7 Material Compositions and Densities

..........................................................................

.14 3.8 Gamma Flux to Dose Rate Conversion Factors ...........................................................

.15 3 .9 Cold Shutdown and Refueling Times ...........................................................................

16 3 .10 Atomic Masses ..............................................................................................................

16 3.11 Conversion Factors .......................................................................................................

16 4.0 Assumptions

.........................................................................................................................

17 5.0 Methodology and Calculations

.............................................................................................

21 5.1 Methodology

.................................................................................................................

21 5.2 Calculations

...................................................................................................................

30 5.2.1 Deleted .................................................................................................................

30 5.2.2 MCNP Boundaries

...............................................................................................

30 5 .2.3 RPV Head Surfaces ..............................................................................................

34 5.2.4 Drywell Bulb Surfaces .........................................................................................

37 5.2.5 Drywell Head Geometry ......................................................................................

37 5.2.6 Drywell Cone Surfaces .......................................................................................

40 5.2.7 Material Compositions

.........................................................................................

44 5.2.8 Radiation Source Region Boundary Definition

...................................................

63 5.2.9 Dose Point Locations

...........................................................................................

64 6.0 Results ..................................................................................................................................

68 7.0 Conclusions

..........................................................................................................................

74 8.0 References

............................................................................................................................

77 9.0 Attachinents

..........................................................................................................................

79 Attachment 1: Identification of RUNT-PC and MCNP Computer Code Input and Output Files ........ 3 Pages Attachment 2: Letter LTR-EP-10-106 (Reference 8.9 to this Calculation)

........................................

1 Page Attachment 3: Design Information Transmittal (DIT), No. 918, Rev 1,

Subject:

Design Inputs for Revision to S&L Calculation 2004-07061 Rev 1 (MCNP Calculation 04-202 Rev 1) to Update Calculation for EPU Source Term and New Steam Dryer ..........................

9 Pages CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 3 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date TABLE OF CONTENTS (Continued)

Attachments (Continued)

Attachment 4: Table 2 of Monticello Nuclear Generating Plant Ops Manual 8.01.01-06, Rev 12 .... 3 Pages Attachment 5: Monticello Nuclear Generating Plant Ops Manual 8.04.01-06, Revision 2 ................

2 Pages Attachment 6: Radiation Monitor Locations on the Refueling Floor Elevation

......................

............

2 Pages Attachment 7: RUNT-PC Output File (RUNT EPU Src 2004-07061 R2_0UTPUT.TXT)

.................

81 Pages Attachment 8: MCNP File (NWCHA) .................................................................................................

176 Pages Attachment 9: MCNP File (WTAFC) ..................................................................................................

109 Pages Attachment 10: MCNP File (WTAFSAC)

...........................................................................................

137 Pages Attachment 11: MCNP File (WTAFARM)

..........................................................................................

257 Pages Attachment 12: MCNP File (NOHDCH8)

..........................................................................................

183 Pages Attachment 13: MCNP File (NOHDA8V)

...........................................................................................

200 Pages Attachment 14: MCNP File (NOHARM3)

.............................. , ...........................................................

148 Pages Attachment 15: MCNP File (NODRYCH)

..........................................................................................

158 Pages Attachment 16: MCNP File (NODRYAR)

..........................................................................................

264 Pages Attachment 17: MCNP File (NOMSA 12) ...........................................................................................

249 Pages Attachment 18: MCNP File (NOMSCHA)

..........................................................................................

120 Pages Attachment 19: MCNP File (NOMSCC) ............................................................................................

127 Pages Attachment 20: MCNP File (NOMSDH) ............................................................................................

160 Pages Attachment 21: MCNP File (NOMSCHD)

..........................................................................................

131 Pages Total Number of Pages: 2,599 Pages CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 JDate X J Safety Related I I Non-Safety Related Page 4 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 1.0 Purpose and Scope 1.1 Purpose This evaluation determines dose rates, due to gamma radiation, at radiation detector locations associated with a drop in the water level within the reactor pressure vessel during cold shutdown and refueling conditions.

The radiation detectors are the Containment High Range Radiation Monitors (CHRRMs) and the radiation detectors located on the refueling floor elevation of the reactor building.

The objective of this calculation is to determine whether these radiation detectors can be used to identify a loss of water in the reactor pressure vessel during cold shutdown and refueling modes. Revision 1 to the calculation determined dose rates at the detectors' locations for the revised core inventory used for the Alternative Source Term (AST). Revision 2 to this calculation determines dose rates at the detectors' locations for conditions associated with an Extended Power Uprate (EPU) to 2004 MWt. Specifically, revision 2 incorporates EPU related changes associated with the reactor core initial heavy metal mass, the reactor core fission product radionuclide inventories, changes in the reactor vessel steam dryer mass and changes in the steam dryer material from type 304 stainless steel to type 316 stainless steel. The initial heavy metal mass in the reactor core increased from 82.067 to 84.432 metric tons. The reactor vessel steam dryer mass increased from 22 to 31 tons. In addition to parameter changes due to EPU, revision 2 incorporates a change in the reported reactor vessel moisture separator mass (from 33 tons to 41 tons) and a change in the reported minimum thickness in the drywell liner in the spherical portion of the drywell (from 1 and 1 /16 inch to 11 /16 inch). 1.2 Scope The intent is to use radiation detectors to satisfy NEI 99-01 [Reference 8.21] Emergency Action Levels (EALs). Dose rates at the radiation detector locations will be determined for NEI 99-01 EALs related to an unplanned loss of water in the Reactor Pressure Vessel (RPV) when irradiated fuel is present and that identify use of CHRRM detectors as an indicator of RPV water loss. These EALs are Site Area Emergency condition CS2 1.b. and 2.b. and General Emergency condition CG1 2.b. Conditions CS2 apply during refueling operability mode. EAL CS2 1.b. applies when containment closure is not established and EAL CS2 2.b. applies with containment closure established.

EAL CG1 2.b. applies for cold shutdown and refueling operating modes, with or without containment closure established.

The CHRRM detectors are located below and to the side of the RPV and are shielded from the reactor core by the water in the RPV, the RPV wall, and the sacrificial shield (Figure 1). Due to the location of the CHRRM detectors with respect to the reactor core and due to the radiation shielding between the core and the detectors

.* it is possible that radiation detectors on the refueling elevation of the reactor building could respond .to lower water levels more quickly than the CHRRM detectors.

As such, this evaluation also determines the dose rates to radiation detectors on the CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s..ree§Lun-'" Cales. For Dose Rates tO CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 5 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date refueling floor elevation of the reactor building.

Dose rates are determined at the detector locations for an RPV water level at the top of the active fuel and for RPV configurations that include or exclude combinations of the moisture separator, the steam dryer, the RPV head, and the drywell head. The dose rate is also determined at the location of the CHRRM detector for the RPV configuration that includes the moisture separator and steam dryer in the RPV with the RPV and drywell heads on and the RPV at normal water level. 2.0 Acceptance Criteria Since this calculation is determining dose rates at radiation detector locations in order to determine whether the detectors can be used to identify a loss of water level in the RPV, the calculated dose rates must be greater than the lower limit of the radiation detector range. As such, the minimum acceptance criterion is that the calculated dose rates at the radiation detector locations be greater than the lower limit of the radiation detector dose rate range. The lower dose rate limit for each detector's dose rate range follows. Minimum dose rates are obtained from Design Input 3.4. Radiation Detector Lower Limit of Detector Dose Rate RanQe CHRRM Channel A 1.0 R/hr CHRRM Channel B 1.0 R/Hr Spent Fuel Monitor Channel A 0.1 mR/Hr Spent Fuel Monitor Channel 8 0.1 mR/Hr Area Radiation Monitor A-1 on Refueling Elevation 0.1 mR/hr Area Radiation Monitor A-2 on RefuelinQ Elevation 1.0 mR/Hr Area Radiation Monitor A-3 on RefuelinQ Elevation 0.1 mR/Hr In addition to the dose rate at the detector location being greater than the lower limit of the radiation detector's dose rate range, the dose rate at the detector location must be greater than the background radiation in the area (i.e., radiation that does not originate from radiation sources in the reactor core).

i_ CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level X Safety Related Non-Safety Related Client NSP Prepared by Project MNGP Reviewed by Proj. No 12400-045 Equip. No. Approved by 3.0 Design Inputs 3.1 Reactor Core Source Term Cale No. 2004-07061 Rev. 2 Date Page 6 of 79 Date Date Date Reactor core radiation source terms are obtained from Appendix A of EPU Task Report T0802 [Reference 8.1]. The radiation source terms consist of a list of reactor core radionuclide inventories that were developed for use in Monticello Nuclear Generating Plant (MNGP) EPU radiological analyses.

The reactor core radionuclide inventories are End of Cycle (EOC) values based on an average burnup of the fuel at 35 GWD/MT across the reactor core. Core radiation source terms are presented in units of Curies per Megawatt Thermal (Ci/MWt).

More than seven hundred radionuclides are presented in Appendix A of Task Report T0802. The fission products from Appendix A of Reference 8.1, that are prevalent at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after reactor shutdown, are included in this evaluation.

These radionuclides are included in Table 3.1-1 and account for more than 99.8% of the total fission product activity at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months decay. The impact associated with ignoring the remainder of the fission products (and ignoring activation products) is underestimation of the dose rate at the detector locations, which is conservative for this evaluation.

Reactor core radionuclide inventories are presented in Table 3.1-1. Table 3.1-1 Reactor Core Radionuclide Inventories (EPU Sources) (1> (EOC Radionuclide Inventories for an* AveraQe Core Burn up of 35 GWO/MT) Core Core Core 10#(2) Inventory 10#(2) Inventory 10#(2) Inventory Nuclide (Ci/MWt) Nuclide (Ci/MWt) Nuclide (Ci/MWt) Noble Gasses Tellurium Group (cont'd) Lanthanides (cont'd) KR83M 45 3.46E+03 TE123M 261 5.45E-01 LA144 375 3.73E+04 KR85 55 3.33E+02 TE125M 270 1.05E+02 NB93M 103 5.71E-03 KR85M 54 7.38E+03 TE127 279 . 2.84E+03 NB94 111 4.03E-03 KR87 65 1.42E+04 TE127M 278 3.70E+02 NB94M 110 ----KR88 70 2.01E+04 TE129 289 8.38E+03 NB95 119 4.87E+04 KR89 74 2.46E+04 TE129M 288 1.24E+03 NB95M 118 3.45E+02 KR90 80 2.43E+04 TE131 299 2.39E+04 NB96 123 8.55E+01 KR91 86 1.81E+04 TE131M 298 3.84E+03 NB97 131 5.03E+04 XE131M 301 2.98E+02 TE132 304 3.82E+04 NB97M 130 4.73E+04 XE133 312 5.48E+04 TE133 309 3.24E+04 NB98 135 4.67E+04 XE133M 311 1.71E+03 TE133M 308 2.05E+04 NB98M 134 4.32E+02 XE135 325 1.82E+04 TE134 315 4.65E+04 NB99 138 4.73E+04 XE135M 324 1.07E+04 TE135 322 2.42E+04 NB100 143 2.62E+04 XE137 334 4.83E+04 Barium, Strontium Group NB101 147 4.30E+04 XE138 339 4.60E+04 BA135M 327 2.96E+OO ND147 387 1.81E+04 XE139 343 3.61E+04 BA137M 336 3.29E+03 ND149 396 1.04E+04 XE140 347 2.38E+04 BA139 345 4.97E+04 ND151 401 5.25E+03 HaloQens BA140 349 4.77E+04 PM147 388 4.33E+03 BR80 32 ----BA141 354 4.51E+04 PM148 394 7.83E+03 BR80M 31 ----BA142 360 4.29E+04 PM148M 393 1.13E+03 BR82 40 1.62E+02 BA143 367 3.77E+04 PM149 397 1.60E+04 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s......,.,;.__..no0y"'

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 7 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 3.1-1 Reactor Core Radionuclide Inventories (EPU Sources) <1> (EOC Radionuclide Inventories for an AveraQe Core Burnup of 35 GWD/MTJ Core Core Core 10#<2> Inventory ID# <2> Inventory 10#<2> Inventory Nuclide (Ci/MWt) Nuclide (Ci/MWt) Nuclide (Ci/MWt) BR83 44 3.45E+03 BA144 374 2.87E+04 PM151 402 5.25E+03 BR84 49 6.03E+03 SR87M 67 1.72E-01 PM152 405 3.67E+03 BR84M 48 2.00E+02 SR89 76 2.68E+04 PM153 410 2.39E+03 BR85 53 7.29E+03 SR90 82 2.64E+03 PM154 413 1.32E+03 BR86 58 5.27E+03 SR91 89 3.37E+04 PR142 363 2.09E+03 BR87 64 1.20E+04 SR92 95 3.62E+04 PR143 370 4.13E+04 BR88 69 1.30E+04 SR93 100 4.08E+04 PR144 377 3.70E+04 BR89 73 9.22E+03 SR94 107 3.84E+04 PR145 380 2.88E+04 1128 285 3.38E+02 SR95 115 3.57E+04 PR146 383 2.29E+04 1129 290 9.98E-04 Noble Metals PR148 391 1.42E+04 1130 295 8.94E+02 M099 139 5.12E+04 SM151 403 1.66E+01 1131 300 2.68E+04 M0101 148 4.58E+04 SM153 411 1.23E+04 1132 305 3.90E+04 M0102 151 4.29E+04 SM155 417 9.78E+02 1133 310 5.51E+04 M0104 158 3.11E+04 SM156 420 6.02E+02 1134 316 6.09E+04 M0105 164 2.17E+04 Y89M 77 4.27E-04 1135 323 5.17E+04 PD107 177 3.01E-03 Y90 84 2.81E+03 1136 329 2.46E+04 PD109 187 8.06E+03 Y90M 83 4.22E-01 1137 333 2.42E+04 PD109M 186 3.34E+03 Y91 91 3.44E+04 1138 338 1.21E+04 PD111 195 1.74E+03 Y91M 90 1.95E+04 Alkali Metals PD111M 194 2.57E+01 Y92 96 3.64E+04 CS134 319 5.35E+03 PD112 200 9.25E+02 Y93 101 4.18E+04 CS134M 318 1.47E+03 PD113 203 7.38E+02 Y94 108 4.19E+04 CS135 326 1.89E-02 PD114 209 5.15E+02 Y95 116 4.49E+04 CS136 331 1.86E+03 PD115 214 4.93E+02 Y96 121 4.26E+04 CS137 335 3.47E+03 PD116 222 3.77E+02 Y97 128 3.66E+04 CS138 340 5.09E+04 RH103M 156 3.65E+04 ZR93 102 7.89E-02 CS139 344 4.82E+04 RH104 162 2.58E+04 ZR95 117 4.85E+04 CS140 348 4.35E+04 RH104M 161 1.68E+03 ZR97 129 4.99E+04 CS141 353 3.24E+04 RH105 168 2.46E+04 ZR98 133 4.60E+04 RB86 61 6.35E+01 RH105M 167 7.58E+03 Cerium Group RB86M 60 6.46E+OO RH106 172 1.52E+04 CE141 356 4.53E+04 RB87 66 -----RH106M 171 5.15E+02 CE142 362 -----RB88 71 2.04E+04 RH107 176 1.52E+04 CE143 369 4.23E+04 RB89 75 2.62E+04 RH108 180 1.04E+04 CE144 376 3.68E+04 RB90 81 2.54E+04 RH109 185 6.63E+03 CE145 379 2.88E+04 RB91 88 3.15E+04 RU103 155 4.05E+04 CE146 382 2.28E+04 RB92 94 2.74E+04 RU105 166 2.71E+04 CE147 385 1.76E+04 RB93 99 2.08E+04 RU106 170 1.41 E+04 CE148 390 1.28E+04 RB94 106 1.08E+04 RU107 175 1.51E+04 NP239 506 5.22E+05 Tellurium Group RU108 179 1.03E+04 PU238 505 9.04E+01 SB122 255 8.15E+01 TC99 141 4.43E-01 PU239 507 1.09E+01 SB122M 254 6.50E-01 TC99M 140 4.54E+04 PU241 509 4.09E+03 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Saree§ L*m"'y"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I , Safety Related I I Non-Safety Related Page 8 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 3.1-1 Reactor Core Radionuclide Inventories (EPU Sources) (1> (EOC Radionuclide Inventories for an' Averaae Core Burnuo of 35 GWO/MT' Core Core Core 10#121 Inventory 10#121 Inventory 10#(2) Inventory Nuclide (Ci/MWt) Nuclide lCi/MWO Nuclide (Ci/MWt) SB124 265 3.64E+01 TC100 145 1.24E+04 Actinide Group 131 SB124M 264 2.87E-01 TC101 149 4.59E+04 U-237 504 3.38E+04 SB125 269 4.93E+02 TC102 152 4.29E+04 Others SB126 274 3.21E+01 TC103 154 4.07E+04 GE77 18 2.73E+01 SB126M 273 1.25E+01 TC104 159 3.32E+04 AS77 19 7.37E+01 SB127 277 2.79E+03 TC105 165 2.66E+04 AS78 23 1.68E+02 SB128 283 3.75E+02 TC107 174 9.92E+03 AG109M 188 8.05E+03 SB128M 282 4.72E+03 Lanthanides 1" 1

AG110M 191 1.02E+02 SB129 287 8.52E+03 AM241 510 4.61E+OO AG111 197 1.74E+03 SB130 293 2.75E+03 AM242 513 2.01E+03 AG111M 196 1.74E+03 SB131 297 2.26E+04 AM242m 512 6.02E-01 AG112 201 9.28E+02 SB132 303 1.35E+04 CM244 514 5.24E+01 AG113 205 6.65E+02 SB133 307 1.60E+04 EU152 408 3.78E-01 CD115 218 4.85E+02 SE77M 20 2.00E-01 EU152M 407 4.37E+OO CD115M 217 4.57E+01 SE79 27 1.39E-02 EU154 415 2.90E+02 CD117 230 2.74E+02 SE79M 26 4.08E+02 EU155 418 2.04E+02 CD117M 229 1.49E+02 SE81 36 1.51E+03 EU156 421 4.01E+03 IN115M 219 4.90E+02 SE81M 35 4.21E+01 EU157 423 5.57E+02 IN117 232 2.53E+02 SE83 43 1.34E+03 EU158 425 2.29E+02 IN117M 231 3.20E+02 SE83M 42 2.05E+03 EU159 427 1.21E+02 SN125 268 4.62E+02 SE84 47 5.84E+03 EU160 430 5.54E+01 SN128 281 4.36E+03 SE85 52 3.45E+03 LA140 350 4.91E+04 GD159 428 6.14E+03 SE86 57 6.83E+03 LA141 355 4.53E+04 TB160 432 1.39E+03 SE87 63 5.42E+03 LA142 361 4.39E+04 TB161 435 9.21E+02 LA143 368 4.21E+04 Notes: (1) Core inventories are the values at a time of 0 seconds after reactor shutdown and are from Appendix A of Task Report T0802 [Reference 8.1]. Core inventories for reactor core fission product radionuclides that are also present in Appendix A of Reference 8.1 as reactor core activation products, include the activation inventories.

(2) The ID# is the nuclide identification number used in RUNT-PC. (3) Radionuclide activity for actinides included in the lanthanide and actinide categories represent more than 99% of total actinide activity in Appendix A of Reference 8.1 at a time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after reactor shutdown.

3.2 Reactor

Core Operating Parameters

Reactor Power Level: 2004 MWt. This is the EPU reactor power level (Table 1 of Attachment 1 to Reference

8.2 which

is included as Attachment 3 to this calculation).

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Sareei Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 9 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date

Metric Tons of Initial Heavy Metal: 84.432 (Table 1 of Attachment 1 to Reference

8.2 which

is included as Attachment 3 to this calculation).

Reactor Core and Fuel Design Data The length of the active core is 12.1033 ft as indicated in the Supplemental Table which is included in the DIT (Reference 8.2, included as Attachment 3 to this calculation).

The active fuel length of 12.1033 ft (i.e., 145.2396 inches) corresponds to GE14 fuel as indicated in Table 2 of Ops Manual B.01.01-06 (Reference 8.l), included as Attachment 4 to this calculation).

The following reactor core and fuel design parameters are obtained from Reference 8.2 and from Table 2 of Reference 8.5 and are for GE14 fuel. Table 3.3-1 Reactor Core & Fuel Desian Data Fuel Parameter Value FUEL ASSEMBLY Number of Assemblies 484 Water-Spacer Capture Rods 2 Rod Pitch (inches) 0.510 Water to Fuel Volume Ratio 2.84 FUEL ROD Active Fuel Lenath (inches) 145.24 Gas Plenum Lenath (inches) 10.1 Pellet Diameter (inches) 0.344 CLADDING Material Zr-2 Thickness (inches) 0.026 Outside Diameter (inches) 0.404 FUEL CHANNEL Material Zircalov Outside Dimension (inches) 5.478 Wall Thickness (inches) 0.100/0.065*

Channel Lenath linchesl 162.44 MOVABLE CONTROL RODS Number 121 Shape Cruciform Pitch (inches) 12.0 Width (inches) 9.75 Control Lenath (inches) 143 REACTOR CORE Circumscribed Core Dia linches) 160.1 Core Diameter leauivalentl linches) 149

The larger fuel channel wall thickness will be used in this evaluation (i.e., 0.100 inches thick). Use of the larger thickness will increase the mass of radiation attenuating material in the reactor core which reduces the dose rate at the radiation detector locations.

Smaller dose rates at the detector locations are conservative for this evaluation.

"Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ,,_...,..s_;.....,n-"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 10 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 3.4 Radiation Detector Information The radiation monitors of interest are the Containment High Range Radiation Monitors (CHRRMs) and the radiation monitors on the refueling floor elevation of the reactor building (i.e., Elevation 1027'). CHRRM The CHRRM detectors are associated with radiation monitors RM7860 Channel A and Channel B and are located in the drywell. Channel A detector is located at 0 degrees Azimuth and Channel Bis at 180 degrees Azimuth (Reference 8.7). The detectors are approximately 6" to 8" from the drywell wall inside surface and are at an approximate elevation of 947'-9" (Reference 8.7). Each detector is 2.5 inches in diameter and 9.9 inches long (Reference 8.8). The low end range of the detectors is 1 R/hr (Reference 8.2). Radiation Detectors Located on Reactor Building El 1027' There are 5 radiation detectors located on elevation 1027' of the reactor building.

Two of these detectors are for the spent fuel pool monitors and 3 are area radiation monitors.

The radiation detectors are identified and their locations and elevations are shown in Attachment 6 to this calculation.

Radiation detector distances presented in Attachment 6 were obtained via field measurements.

A summary of the detectors follows.

Channel A Spent Fuel Pool Monitor: Located adjacent to a crane support column along the east reactor building wall at building coordinate Ng. Detector elevation is 9'-9.5" above the floor. The low-end range of this detector is 0.1 mR/hr (Reference 8.2).

Channel B Spent Fuel Pool Monitor: Located adjacent to a crane support column along the north reactor building wall at building coordinate

6.9. Detector

elevation is 9'-8" above the floor. The low-end range of this detector is 0.1 mR/hr (Reference 8.2).

Area Radiation Monitor A-1: This is the Refueling Floor Low Range Monitor, located 5'-4" west of building coordinate 7.9 and along the north reactor building wall. Detector elevation is 9'-3" above the floor. The low-end range of this detector is 0.1 mR/hr (Reference 8.2).

Area Radiation Monitor A-2: This is the Refueling Floor High Range Monitor, located adjacent to a crane support column along the north reactor building wall at building coordinate

6. Detector elevation is 6'-6" above the floor. The low end range of this detector is 1 mR/hr (Reference 8.2).

Area Radiation Monitor A-3: This is the Refueling Floor Stairway Monitor, located adjacent to a crane support column along the west reactor building wall at building coordinate Ph. Detector elevation is 6'-7.5" above the floor. The low-end range of this detector is 0.1 mR/hr (Reference 8.2). 3.5 Elevations Elevations that are used in the computer model are presented in Table 3.5-1. Table 3:5-1. Elevation

1n ut Data Elevation 920'-6" 8.12.2 942'-5" 8.12.2 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

..... n-" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 11 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 3.5-1 Elevation Input Data Description Elevation References Inside Bottom of Reactor Vessel 949'-5" 8.12.4 Top of Active Fuel Region of Reactor Core 978'-8.5" 8.12.4 Top of Assemblies (Also Top of Upper Grid) 367.125" Above Inside 8.12.4 Bottom of Reactor Vessel Moisture Separator Flanqe 981'-9.25" 8.12.4 Normal RPV Water Level 35" Above 477.5" Above 8.12.4 Bottom Inside Surface of RPV Top of Moisture Separator 554 and 5/8" Above Inside 8.12.4 Bottom of Reactor Vessel Top of Steam Dryer 643.5" Above Inside Bottom 8.12.4 of Reactor Vessel Top of RPV Head (Underside Surface El) 1012'-6.5" 8.12.4 Top of Sacrificial Shield 993'-7" 8.12.12 Bottom of Drywell Cone Section (i.e., bottom of ledge that 995'-9" 8.12.1 protrudes from drvwell wall) Top of Drvwell Cone Section 1001' 8.12.1 Top of Drywell Wall Extension (that rests on drywell wall 1003'-7.25" 8.12.1 cone extension)

Flanqe Elevation for RPV Head 1004'-0" 8.12.4 Bottom of Fuel Storaqe Pool 988'-11" 8.12.5 Bottom of Drver Separator Storaqe Pool 1002'-8" 8.12.5 Refuelinq Floor Elevation 1027'-8" 8.12.6, 8.12.7 Top of Roof Over Refuelinq Floor Elevation 1073'-2" 8.12.8 3.6 Component Materials and Dimensions Relevant dimensions and identification of component materials are presented in Table 3.6-1. Table 3.6-1 Relevant Dimensions and Identification of Component Materials Description Value Reference Comments DIMENSIONS FOR STRUCTURAL FEATURES Sacrificial Shield Outer Diameter 24'-9&7/8" 8.12.3 Sacrificial Shield Thickness 2'-3" 8.12.6 Internal Radius of Spherical Portion 31'-3" 8.12.2 This is the internal radius to of Drvwell the drvwell concrete wall. Drywell Diameter:

for drywell 33'-6" 8.12.9 cylindrical section below concrete ledqe at El 995'-9" Minimum Inner Diameter:

to drywell 28'-2" 8.12.1 wall conical protrusion.

Inner Diameter:

to drywell wall 30'-7" 8.12.1 concrete ledge between El 1001' and 1003'-7.25" Drywell Radius: for drywell cylindrical 16'-0.5" 8.12.1 section between 1003'-7.25" and the bottom of the shield plugs at El. 1021 '-7" CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 12 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 3.6-1 Relevant Dimensions and Identification of Component Materials Description Value Reference Comments Inner Radius of Opening for 16'-6" + 2" + 4.5" + 2" = 8.12.10 Intermediate Row of Drywell Shield 17'-2.5" Blocks Inner Radius of Opening for Lowest 16'-0.5" + 5.5" = 16'-6" 8.12.10 The opening radius varies Row of Drywell Shield Blocks from 16'-6" to 16'-8". A radius of 16'-6" is used. Inner Diameter of Opening for top 35'-6&5/8" 8.12.11 row of Drywell Shield Blocks Thickness of Concrete Shield Blocks 2' Each Leve.I 8.12.6 There are 3 layers of shield Located between the Drywell and the blocks. The top of the upper Refueling Floor of the Reactor layer of shield blocks is at the Building.

refueling floor elevation (i.e., 1027'-8")

Drywell Radius: cylindrical section 16'-0.5" 8.12.10 between top of concrete ledge projection at El 1003'-7.25" to drywell concrete shield pluQ elevation.

Thickness of Stainless Steel Plate 0.25" 8.12.10 that lines the drywell wall from Elevation 1001' and above. Drywell Wall Thickness Varies between 4' and 8.12.5 The 4' wall thickness lies 5.5' between the drywell and fuel storage pool. The pool is normally filled with water and provides additional shielding for drvwell sources. Reactor Bldg Refueling Floor See Comments 8.12.7 North-South:

Dimensions at El 1027'-8" 23', 22'-3", 22'-3", 35'-3"

North -South Direction 8.12.6 for floor between bldg grid coordinates

East -West Direction thickness S-R, R-P, P-N, and N-M respectively.

There is 15.5" between bldg grid coordinate S and the south end of the reactor building and between bldg coordinate M and the north end of the reactor building East-West:

22'-9", 22'-3", 22'-3", 22'-3", 22'-3", & 22'-9" between bldg grid coordinates 3.1-4.1, 4.1-5.1, 5.1-6, 6-6.9, 6.9-7.9, and 7.9-8.9 respectively.

There is 18" between coordinate 3.1 and the west end of the react bldg and 18" between coordinates 8.9 and the east end of react bldg. There is a 9" thick concrete floor slab at refueling floor El CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s..,._§._..n_

... Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I

I Non-Safety Related Page 13 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date .. ; Table3.6-1 . *. .* Relevant Dimensions.

and* Identification of Component Materials . . -Description

Value . *. * . Reference

'* .. Comments * .. .. .* 1027'-6".

Thickness of Reactor Building Outer 1' 8.12.5 Wall (El 1001' to 1073') DRYWELL HEAD Radius: Cylindrical portion of the 13'-5.75" 8.2 and 8.6 This is the radius to the inside drvwell head. surface of the drvwell head. Height: 13'-2 & 5/8" 8.2 and 8.6 This is the height to the underside surface of the drvwell head top. Elevation Where Drywell Head 1004'-0" 8.12.12 Attaches to the Drvwell Shell: Wall Thickness (inches): 1 & 5/16" 8.2 Drywell liner top head thickness Material:

SA-516-70FBX steel plate 8.2 DRYWELL LINER Liner Thickness

Spherical Shell Portion: 11/16" to 2.5" 8.2

Spherical Shell to Cylinder Neck: 2.5"

Cylinder Neck: 0.635" to 1.5"

Bottom Head 1.25" REACTOR PRESSURE VESSEL Material:

Carbon Steel Thickness Supplemental In addition to the values

Top Head Dollar Plate Region: 2 & 11/16" min. Table to Ref 8.2 presented in column 2, each

Plates Closer to Flange: 3.25" min. thickness includes a 3/16"

Cylindrical Portion Near Flange: 5 & 3/16" min. nominal SS weld overlay.

Center of Cylindrical Portion: 5 & 1/16" min.

Bottom Head, Near Cylindrical Section: 5 & 1/16" miri.

Bottom Head Center Section 5 & 15/16" min. RPV Inner Diameter 205" 8.12.4 CORE SHROUD Thickness:

1.75" Supplemental Table to Ref 8.2 Material:

Stainless Steel Supplemental Table to Ref 8.2 Shroud Outer Diameter 167.25" 8.12.4 REACTOR PRESSURE VESSEL MOISTURE SEPARATOR Weight: (tons) 41 Supplemental The moisture separator is Table to Ref 8.2 supported on the RPV core shroud as depicted in Reference 8.12.4. Material:

Nearly All 304 SS Supplemental GE Spec for Vessel Internals Table to Ref8.2 STEAM DRYER Weight: (tons) 31 Supplemental The final steam dryer weight (See Comments)

Table to Ref 8.2 is 29.0 metric tons [Reference 8.9]. (29 metric tons x 1000 kg/metric ton x 1000 gm/kg) I CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 14 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date . _ ___ Table 3:6-1 -.. -" .. ----. ****;--Relevant Dimensions andJdentificatioh of Comoonent Materials

< ;* --Description

--_-* ; -* Value----; .*;'--1***

Reference Comments.

(453.5 gm/lb x 2000 lb/ton)= 32 tons. Where 453.5 gm/lb is from Design Inputs 3.11. The preliminary weight was 31 tons. A steam dryer weight of 31 tons was used in this evaluation.

As indicated in Table 6-3 and Section 7.0 (Conclusions) of this calculation, the dose rates at all radiation detector locations are below the lower limit of the detector when the steam dryer (with a weight of 31 tons) remains within the reactor vessel. Increasing the steam dryer weight from 31 to 32 tons will reduce the dose rates at the detector locations and thus the dose rates will still be below the lower limit of the detectors.

As such, use of 31 tons (in lieu of 32 tons) as the steam dryer mass does not have an impact on the conclusions of this calculation.

Length: (inches) 189.5 Supplemental Table to Ref 8.2 Diameter: (inches) 201 Supplemental Table to Ref 8.2 Material:

316 SS Supplemental Table to Ref 8.2 ASSEMBLY AREA ABOVE ACTIVE FUEL REGION Plenum Springs Table 5.2 of Plenum springs material is Mass 1.1 kg/assembly 8.22 302 SS which is treated in this Material 304 Stainless Steel evaluation as 304 SS. Top End Fittings Table 5.2 of Mass 2.0 kg/assembly 8.22 Material 304 Stainless Steel 3. 7 Material Compositions and Densities Table<37"1 Steel Steel Densit 8.18 N/A Steel is treated as ure iron. Concrete Concrete Densit 8.20 minimum densit = 140 lb/ cu ft CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 15 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 3.7-1 Material Composition Data and Densities Material Value Ref Comment Material Composition (Density Fractions) grams/cc (Based on a density of 2.35 grams/cc)

H 0.013 glee 8.20 0 (in Mix) 0.103 glee 8.20 0( in ore and cement) 1.068 q/cc 8.20 Si 0.742 q/cc 8.20 Ca 0.194 q/cc 8.20 Na 0.04 q/cc 8.20 Mo 0.006 q/cc 8.20 Al 0.107 g/cc 8.20 s 0.003 glee 8.20 K 0.045 glee 8.20 Fe 0.029 q/cc 8.20 Air Air Density 1.293E-3 q/cc 8.19 Po 377 of Reference 8.19 N 75.5 w/% 8.19 0 23.2 w/% 8.19 Ar 1.3 w/% 8.19 Water Water Density 1. glee Stainless Steel SS Type 304SS 316 SS SS Density 7.9 q/cc 8.0 q/cc 8.19 Table 11.3 of Reference 8.19. c 0.08w/% max 0.08 w/% max 8.19 Table 11.2 of Reference 8.19 Mn 2.00 w/% max 2.00 w/% max 8.19 Si 1.0 w/% max 1.0 w/% max 8.19 Cr 18-20 w/% 16-18 w/% 8.19 Ni 8-12 w/% 10-14 w/% 8.19 Fe Difference w/% Difference w/% 8.19 Zirconium (Zr) Density 6.56 glee 8.22 This is the density for zircaloy-2 as indicated in Table 5.1 of Reference 8.22. Zircaloy is approximately 98% zirconium (Table 5.1 of Reference 8.22). Zr 100 w/% NA Pure Zirconium

3.8 Gamma

Flux to Dose Rate Conversion Factors Gamma flux to dose rate conversion factors are presented in Table 3.8-1. Dose rate conversion factors are obtained from Table 2-2 of the RUNT-PC User's Manual (Reference 8.15). The dose rate conversion factors are based on the equations given in Section C1 of Reference 8.17. Dose rate conversion factors are in units of mrad (air)/hr per MeV/cm 2-sec. Conversion factors for mrad (air) are used in this analysis since dose rates are being determined for radiation detectors with readouts in units of R/hr. To obtain dose rates in units of R/hr, the dose rates (in rad (air)/hr) are divided by 0.877 rad (air)/R.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S.....,eiLun-<<*

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 16 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date " Table3.8-1

- - .. ... Gamma Flux to Dose Rate Conversion Factors . . Energy ... : . Dose Rate Conversion Factor Energy Dose Rate Conversion Factor** ::* (MeV) <mrad (air)lnf)Z<MeV/cm 2-sec) (MeV) *:. <mrad 0.015 7.315E-02 0.6 1.705E-03 0.02 2.949E-02 0.8 1.665E-03 0.03 8.525E-03 1.0 1.613E-03 0.04 3.853E-03 1.5 1.480E-03 0.05 2.339E-03 2.0 1.371E-03 0.06 1.757E-03 3.0 1.221E-03 0.08 1.400E-03 4.0 1.117E-03 0.1 1.348E-03 5.0 1.048E-03 0.15 1.440E-03 6.0 1.002E-03 0.2 1.544E-03 8.0 9.331 E-04 0.3 1.659E-03 10.0 8.986E-04 0.4 1.699E-03 15.0 8.525E-04 0.5 1.711 E-03 3.9 Cold Shutdown and Refueling Times

The following information is obtained from Table 1 of Reference 8.2. Cold Shutdown Scheduled for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after reactor scram. 3.10 Atomic Masses Element Atomic Mass Reference HydroQen (H) 1 8.18 OxvQen (0) 16 8.18 Uranium (U) 238 8.18 3.11 Conversion Factors The following conversion factors were used for numerical analyses performed in this evaluation.

1 Kg = 2.205 lb therefore 1 lb = (1 Kg x 1000 g/kg) / 2.205 = 453.5 grams (pg B-10 of Reference 8.23)

2.54 cm per inch

0.877 rad = 1 Roentgen (R) (Pg 132 of 8.25)

1000 mrad = 1 rad CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S-.,es;Lo.,,_

... Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I I Non-Safety Related Page 17 of 79 Cluent NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 4.0 Assumptions Validated Assumptions

4.1 Refueling

Floor Walls and Roof No credit is taken for radiation attenuation or scatter by the material in the walls and roof on the refueling floor elevation.

The walls and roof are composed of thin metal siding. The siding is assumed to be 0.25" thick. The volume occupied by the siding is filled with air. This is acceptable since the thin metal siding will not provide significant scatter or attenuation of the radiation.

4.2 The RPV Head Thickness The RPV head thickness varies between 2&14/16" (head dollar plate region) and 3&7/16" thick closer to the flange (These thicknesses include 3/16" thick stainless steel weld overlay -Table 3.6-1 of Design Input 3.6). For this evaluation an RPV head thickness of 3&7/16" will be used for the entire RPV head. Applying the thicker RPH head value for the entire RPV head surface will minimize the dose rates at the dose point locations.

Since the dose rates are intended to be used to develop radiation monitor setpoints, the smaller dose rates that will result from applying the thickest part of the RPV head to the entire RPV head surface are conservative.

4.3 Drywell

Head Tangent As indicated in Table 3.6-1, the height of the drywell head is 13'-2&5/8" (to the underside of the drywell head). From the figure in Reference 8.6, it is observed that the tangent for the elliptical top portion of the drywell head is slightly less than half the height. For this evaluation, the tangent of the elliptical portion of the drywell head is set at 6' below the bottom surface of the drywell head. Any deviation between the assumed and actual drywell head tangent distances will affect the shape of the top of the drywell head but will not affect the drywell head thickness

or top elevation and thus there will not be a significant impact on dose rate results. 4.4 Drywell Head Inner Radius The inner radius of the drywell head is 13'-5.75" (Table 3.6-1 of Design Input 3.6). To simplify the geometry model used in the MCNP computer code, the outer radius of the drywell head is assumed to be equal to the minimum radius to the drywell wall concrete conical protrusion (i.e., half of the diameter of 28'-2" per Table 3.6-1 of Design Input 3.6). This geometry model simplification does not change the thickness of the drywell head wall between the reactor vessel and the radiation detectors on the refueling floor elevation of the reactor building.

As such, this geometry model simplification will not affect calculated dose rates at the radiation detector locations on the refueling floor elevation of the reactor building.

4.5 Drywell

Steel Liner Thickness The drywell steel liner thickness varies as indicated in Table 3.6-1 of Design Input 3.6. The liner thickness at the spherical shell portion varies from 11/16" to 2.5". The liner thickness at the drywell cylinder neck varies from 0.635" to 1.5". For the purposes of this evaluation average CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

... Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 18 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date thickness values over the spherical portion and over the cylindrical neck portion will be assumed. The drywell steel liner at the spherical portion of the drywell and at the drywell cylinder neck is a radiation scattering surface for reactor core gamma radiation that contributes to the dose rates at the CHRRM detectors that are located in the spherical portion of the drywell. The steel liner in the cylindrical neck portion of the drywell is also a scattering surface for reactor core gamma radiation that contributes to the dose rates at radiation detectors on the refueling floor elevation of the reactor building.

The dose rate contribution to radiation detectors due to radiation that scatters off the spherical portion or the cylindrical neck portion of the drywell is expected to be insignificant because the radiation must pass through the reactor vessel wall and the sacrificial shield wall and will therefore undergo significant radiation attenuation.

As such, use of average drywell wall liner thicknesses would not have a significant impact on calculated dose rates at the radiation detector locations.

4.6 Drywell Wall Thickness The drywell wall thickness varies between 4' and 5.5' for elevations above 1001 '. The 4' thickness is adjacent to the refueling pool. The water in the refueling pool provides additional shielding.

For the purposes of this evaluation, a drywell concrete wall thickness of 4.5' will be assumed for elevations above 1001'. This is acceptable because the dose rate contribution due to gamma radiation that penetrates through the 4.5' thick concrete wall will be negligible.

The drywell concrete wall thickness for elevations below 1001' is assumed to be equal to the difference between the drywell inner wall surface and the reactor building outer walls. This overestimates the actual concrete thickness.

This assumption is acceptable because gamma radiation that would penetrate the full depth of the actual wall thickness would not contribute to the dose rates at the radiation detector locations and thus modeling these areas as solid concrete has no impact on dose rates at the radiation detector locations.

4.7 Moisture

Separator The RPV moisture separator is supported on the RPV core shroud (Design Input 3.6, Table 3.6-1). Drawing NX-7831-197-1 (Reference 8.12.4) shows that the diameter of the RPV moisture separator tube region is approximately equal to the outer diameter of the RPV core shroud. The outer diameter of the RPV moisture separator is assumed to be equal to the outer diameter of the core shroud. A diameter of 167" is assumed for the RPV moisture separator.

The diameter of the RPV moisture separator is used to obtain a homogenized density for the moisture separator that is used in the MCNP computer model. Differences between the actual diameter of the moisture separator region and the approximated diameter would result in small differences in the homogenized density of the moisture separator however, there would not be a significant impact on calculated dose rates to the radiation detectors.

4.8 Concrete

Density Concrete density is assumed to be 140 lbs/ft 3 (Design Input 3.7, Table 3.7-1). This is the minimum concrete density recommended in Reference 8.20 for use in radiation shielding evaluations.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s--..§L=oOy"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 19 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 4.9 Zirconium Alloys The alloys, Zircaloy and Zr-2 are assumed to be pure zirconium.

This is a reasonable assumption since the majority of these alloys is zirconium, e.g., Zr-2 is 98% zirconium (Table 3.7-1 of Design Inputs Section 3.7). 4.10 Fuel Assembly Elemental partial homogenized densities, due to fuel and water rod cladding, are determined in this calculation assuming there are 74 fuel rods per assembly and 2 water rods per assembly within a 9x9 array. There are actually 92 fuel rods and 2 water rods for the 1 Ox1 O GE14 assembly used at Monticello per USAR Figure 3.4-1 (Reference 8.3). Although the water rods have a larger diameter than the fuel rods, for this evaluation, when determining the amount of zirconium in the fuel region, the water rods are assumed to be the size of fuel rods. The increased amount of water in the active fuel region (associated with a 9x9 array compared to a 1 Ox1 O array) and conservatism used in this calculation (e.g., overestimation of the control rod mass in the active fuel region) compensate for the underestimation of cladding material (due to use of an assembly with a 9x9 array) and thus the results in Table 6-3 of this calculation are applicable for the GE14 fuel. assembly.

4.11 Control Rods The control rods are cruciform in shape with control blades that are located between the fuel assemblies (Monticello USAR Figure 3.4-1). The control blades contain stainless steel tubes filled with a neutron absorber.

The stainless steel tubes are held in a stainless steel sheath (Monticello USAR Figure 3.5-1). It is assumed that the width of the control blades is half the gap size between assemblies (Monticello USAR Figure 3.4-1). When determining the partial densities in the active fuel region of the core, it is assumed that the control rods are composed entirely of stainless steel, i.e., the neutron absorber material is ignored. The neutron absorber material has a smaller density than stainless steel and thus assuming the control rods are composed entirely of stainless steel maximizes the control rod mass which maximizes the reactor core active fuel region density. Using a maximum density for the homogenized active fuel region minimizes the calculated radiation detector dose rates which is conservative when determining radiation monitor setpoints for this evaluation.

4.12 Fuel Assembly Constituents Above Active Fuel Region Table 5.2 of Reference 8.22 provides assumed mass distributions for BWR fuel assembly structural materials.

It is assumed that the masses and material compositions specified for the fuel rod plenum springs and the fuel rod top end fittings are applicable to Monticello.

The rod plenum springs and the top end fittings materials and masses are presented in Table 3.6-1 of Design Input 3.6 and are used to calculate a homogenized density for the fuel assembly region above the active fuel region. 4.13 CHRRM Detectors The CHRRM detectors (Channels A and B) are located 6" to 8" from the drywell wall (Design Input 3.4). For this evaluation, the CHRRM detectors are assumed to be 6" from the drywell wall. The dose rate at detector locations that are 6" from the drywell wall are expected to be slightly smaller (but not significantly smaller) than dose rates at a detector location 8" from the wall due to CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

.....

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 20 of 79 Client Project Pl'oj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date the greater distance from the radiation source and due to greater protection from scatter radiation that the closer distance to the wall provides.

Since this is a setpoint calculation, the smaller dose rates are desired. 4.14 Significant Digits A large number of significant digits are carried through when performing calculations in Section 5 of this evaluation and were used in the computer code input files. The large number of significant digits are used in an attempt to prevent inconsistencies (i.e., gaps) between boundaries in the MCNP computer model and to obtain the sum total of material weight fractions as close to 1.0 as possible.

The number of significant digits used in the code inputs does not correlate to accuracy in calculated dose rates. Unvalidated Assumptions None CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s.a.--§L<m"'Y"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related **I I Non-Safety Related Page 21 of 79 Client Project Proj. No ,. 0 5.1 NSP MNGP 12400-045 Equip. No. Methodology and Calculations Methodology Prepared by Date Reviewed by Date Approved by Date The MCNP computer code (Reference 8.16) is used to obtain the dose rates at the CHRRM detector locations and at the radiation detector locations on the refueling floor elevation of the reactor building, due to radiation sources in the reactor core. The RUNT-PC computer code (8.15) is used to obtain the gamma radiation energy spectrum that will be used for the source term in the MCNP computer code. The RUNT-PC input file uses the reactor core radionuclide inventories described in Design Input 3.1. Generally, the radiation source inventory is increased by 2% to account for instrument uncertainties.

Since the purpose of this analysis is to determine dose rates at radiation detector locations with the intent of establishing radiation monitor setpoints, the source term will not be increased by 2% in this calculation.

By not increasing the source term by 2%, the dose rates associated with a defined water level in the reactor vessel are minimized as are the radiation monitor setpoints.

Use of lower setpoints is conservative since there would be earlier indication of an event. An output of the RUNT-PC computer code is a gamma energy spectrum for the radiation source. The energy spectrum will be used as input to MCNP. The reactor enters the cold shutdown mode 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after reactor shutdown (Design Input 3.9). Since the objective of this calculation is to use the radiation detectors to indicate loss of water in the reactor vessel during cold shutdown and refueling modes, the gamma spectrum with 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of decay after reactor shutdown is used. The RUNT-PC computer code is also used to determine the dose rates at 1 ft from the sacrificial shield, along the reactor core midplane and along the bottom of the reactor core. RUNT-PC dose rates are compared to dose rates at the same locations determined by the MCNP code (as a check of the MCNP model). MCNP Code: The MCNP computer code (MCNP4C3, Reference 8.16) is used to determine dose rates at the locations of the CHRRM Monitors (Channels A and B) in the drywell; at the locations of Area Radiation Monitors A-1, A-2, and A-3 on the refueling floor elevation of the reactor building, and at the locations of the spent fuel pool monitors (Channel A and B) on the refueling floor elevation of the reactor building.

MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first and second-degree surfaces.

Continuous-energy cross section data are used. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation.

The MCNP computer code is used in this CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

... Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. . 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 22 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date analysis because the geometry between the radiation source (i.e., reactor core) and the radiation detectors is complex. The direct line of sight between the reactor core and the detector locations is well shielded and therefore, the majority of the dose rate at the detector locations is expected to be due to scatter radiation.

MCNP is capable of handling the complex geometry and determining scattering dose rates while other codes, such as RUNT-PC, are not. Geometry One of the first tasks performed when creating the MCNP input file, is to construct a three dimensional model of the reactor vessel, the drywell and the reactor building.

This model is shown in Figures 1 and 2 to this calculation.

The model specifies regions, regional boundaries, and material composition within regions. The model is simplified to the extent practical, while striving to minimize the effect of model simplifications on the calculated doses. The following simplifications were made to the model developed in this calculation. (Model simplifications are visible from Figures 1 and 2).

Solid concrete is used below the drywell floor at Elevation 920'-6" (i.e., the torus area is not included in the model). Also, solid concrete is used between the drywell and the reactor building outer walls between Elevations 920'-6" and 1001' (Assumption 4.6). The concrete thickness between the drywell and the rooms or areas adjacent to the drywell is sufficient such that any gamma radiation that makes its way into these areas is not likely to contribute to the dose rate at the detector locations.

As such, modeling these areas as solid concrete has no impact on the dose rates at the detector locations.

The 2' -3" thickness of the sacrificial shield is extended all the way down to the drywell floor at Elevation 920'-6". The sacrificial shield is actually thicker than 2'-3" from the drywell floor Elevation at 920'-6" to an elevation that is several feet below the bottom of the reactor vessel. There is already massive shielding between the reactor core and the area outside the wider bottom portion of the sacrificial shield. This shielding is due to the water and components in the lower portion of the RPV, due to the reactor core shroud and RPV wall, and due to the sacrificial shield. The shielding, and the relative location of the thicker portion of the sacrificial shield with respect to the CHRRM detector locations make it unlikely that radiation that passes through the thicker portion of the sacrificial shield will contribute to the dose rate at the detector locations.

The reactor building area from Elevation 1001' to the concrete floor at the refueling floor elevation at 1027'-8" is treated as an air space around the cylindrically shaped drywell wall. This includes the fuel storage pool and the dryer separator storage pool. A 9" thick concrete floor was placed over the air space. The 9" thick concrete floor is the thickness of the floor at the refueling floor elevation (Table 3.6-1 of Design Input 3.6). Water in the storage pool was ignored. This simplification is acceptable since the thickness of the drywell wall will effectively eliminate the dose rate contribution due to direct radiation shine from the core, through the drywell wall to the detectors on the refueling floor elevation.

The metal walls and roof at the refueling floor elevation (i.e., above Elevation 1027'-8")

of the reactor building were ignored. These walls and roof are constructed of thin sheet metal and will not provide significant attenuation or scatter of the gamma radiation (Assumption 4.1).

The RPV volume below the bottom of the active fuel was modeled as being filled with water without consideration of lower vessel internals (such as support structures and control rod drives) that could increase the overall density of the region. Due to the location of the vessel CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 23 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date bottom with respect to the CHRRM detector locations, radiation that passes through the bottom of the vessel is not likely to contribute to the dose rates at detector locations and thus ignoring vessel internals is acceptable.

Material Compositions:

Materials used in the MCNP model include water, uranium, zirconium, iron, stainless steel, concrete and air. The active region of the reactor core is an homogenization of the uranium fuel, oxygen (within the U02), zirconium cladding, control rods (treated as stainless steel), and water. Core homogenization is discussed in Section 5.2.7 of this calculation.

Assembly support structures in the core active region were ignored. The upper assembly area above the active fuel region is an homogenization of the fuel rod extensions, plenum springs, and top end fittings.

The mass of zirconium in the fuel rod extension cladding is calculated using the calculated volume and zirconium density. The material and mass for top end fittings and plenum springs are obtained from BWR average values. The homogenized densities in the upper assembly region are determined in Section 5.2.7 of this calculation.

The moisture separator is composed of stainless steel. The moisture separator region contains stainless steel and air. The stainless steel mass is homogenized by dividing the moisture separator mass by the volume occupied by the moisture separator.

The volume of air is determined by subtracting the volume occupied by the stainless steel from the total volume. The air volume is then multiplied by the air density to obtain the air mass in the region. The homogenized air density is then calculated by dividing the air mass by the total volume occupied by the moisture separator region. The homogenized volume for the steam dryer is determined similarly.

The top portion of the moisture separator overlaps the bottom of the steam dryer. The homogenized density for this region is also determined in a similar manner. The homogenized densities in the moisture separator and steam dryer regions are determined in Section 5.2. 7 of this calculation.

Radiation Source Terms The radiation source term used by MCNP is obtained from RUNT-PC output. The source term is divided into 27 energy groups. A source strength in units of total photons/sec is provided for each energy group. The gamma source strength source term is presented in Table 6-2 and includes bremsstrahlung radiation.

Dose Rates at Detector Locations In order to be an effective means for determining loss of reactor coolant in the reactor vessel during cold shutdown and refueling modes, the CHRRM and radiation detectors on the refueling floor elevation of the reactor building would have to be able to detect dose rate changes for a number of reactor vessel and drywell configurations.

This calculation determines dose rates at the radiation detector locations for the following reactor vessel and drywell configurations.

The moisture separator and steam dryer are in the RPV, RPV head in place, drywell head in place, and RPV water at normal water level. This configuration is used to establish normal 8-hour post shutdown dose rates at the radiation detector locations for comparison with other CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s.....,...i W..n-"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 24 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date configurations in which the water level is at the Top Of the Active Fuel (TOAF).

The moisture separator and steam dryer are in the RPV, RPV head and drywell head in place, and water level at the TOAF. The water level is set at the TOAF per guidance given in NEI 99-01 (Reference 8.21).

The moisture separator and steam dryer are in the RPV, RPV head and drywell head are removed, and water level at the TOAF.

The steam dryer is removed from the RPV, the moisture separator remains in the RPV, RPV head and drywell head are removed, and water level at the TOAF.

The moisture separator and steam dryer are removed from the RPV, RPV head and drywell head are removed, and water level at the TOAF. All the dose rates in this calculation are determined for the drywell configuration in which the drywell shield blocks between the reactor cavity and the refueling area are removed. If the shield blocks were not removed, the dose rates at the radiation detector locations on the refueling floor elevation of reactor building would be insignificant for all reactor vessel configurations.

MCNP input and output files are identified in Attachment 1 of this calculation.

RUNT-PC Computer Code: RUNT-PC calculates time-dependent radionuclide activities and performs gamma and beta shielding calculations for a variety of shield geometries.

Gamma attenuation calculations are performed using Gaussian quadrature integration of point and line kernel dose rate equations.

Both predefined and ad hoc geometries are available.

Energy dependent linear attenuation coefficients and buildup factors, for each integration, are derived from input parameters and internal databases.

Note: The primary purpose for running RUNT-PC is to obtain the radiation source photon energy spectrum at the desired time after reactor shutdown.

The photon energy spectrum is used for the MCNP input. RUNT-PC was also used to determine dose rates at the top of the drywell head and at the side of the reactor pressure vessel (opposite the sacrificial shield). Dose rates calculated by RUNT-PC are compared to dose rates calculated by MCNP. Due to RUNT-PC limitations associated with number of shield regions, several shield regions were combined in the RUNT-PC model. RUNT-PC code inputs are described in Table 5-1. RUNT-PC input and output files are identified in Attachment 1 of this calculation.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X I Safety Related I I Non-Safety Related Page 25 of 79 Client NSP Prepared by Date *-Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date RUNT-PC Inputs Table 5-1 RUNT-PC lnout Parameters

- - - - ..

Input Parameter I Vall.le Description

.. ... Radiation Source Term Input List of Reactor Core Radionuclide Inventories (253 radionuclides per Table 3.1-1) Bypass & T This indicates that the source term will consist of a list of radionuclides and Activation radionuclide inventories.

A list of 253 nuclides and nuclide activities in the core are provided as input (in units of curieslMWt).

The nuclide activities are from Table 3.1-1 of this calculation.

IDEPLT 2004 This is core power level in MWt (Section 3.2). This factor adjusts the radionuclide activities from curieslMWt to curies for the entire core. The followina Parameters are used to aae the radiation source term. DELTAT 2.88E+04 This parameter and the next (NUMBER) establish a time loop (seconds) (seconds) over which the source term will be decayed. DEL TAT is the time in seconds and NUMBER is the number of DEL TAT time intervals in the time loop. DEL TAT of 2.88E+04 sec is 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . (DEL TAT (time steps) of 3600 sec, 14400 sec, and 86400 sec are also included in the RUNT-PC input but are not used in this calculation.)

NUMBER 1 One time loop is used with DEL TAT of 2.88E+04 (See Discussion above). ISOSHLD Input Parameters (To Define Geometry, Material Composition and Dose Point Locations)

Parameters to Determine Dose Rates Outside the Sacrificial Shield at the Core Midplane and Outside the Sacrificial Shield at the Bottom of the Active Core Region NEXT 1, 5 NEXT = 1 indicates a new problem. NEXT= 5 is used to indicate new dose point coordinates are being specified.

The first dose rate calculation in an input file uses NEXT = 1 and subsequent calculations use NEXT = 5. NCMP 5 Indicates number of material compositions used in the problem. MATS 7 Indicates number of materials used to define the compositions.

Material Com12osition Partial Densities (glee} Material ID# Material Descri12tion 1 2 § 103 Concrete 0.0 0.0 0.0 0.0 2.242 101 Air 0.0 0.0 1.293E-03 0.0 0.0 102 Water 0.7175* 1.0 0.0 0.0 0.0 92 Uranium 1.7622 0.0 0.0 0.0 0.0 8 Oxygen 0.2369 0.0 0.0 0.0 0.0 40 Zirconium 0.70937 0.0 0.0 0.0 0.0 26 Iron 0.48333**

0.0 0.0 7.86 0.0 The material compositions are: composition 1 is the fuel region, 2 is the water between the fuel and the RPV wall, 3 is air, 4 is iron for the RPV wall, and 5 is concrete for the sacrificial shield wall. *Note: Water in the fuel region (i.e., Region 1) is located within the fuel assemblies and in the gap area between assemblies.

The fuel region water density used in the RUNT-PC model is 0.7175 glee and is calculated in Section 5.2. 7 of this calculation and used in the MCNP runs. **This is the sum of the SS component densities and iron density presented in Table 5.2-7. (3.8666E-4

+ 9.6666E-3 + 4.8333E-3

+ 9.18327E-2

+ 4.8333E-2

+ 3.28278E-1

= 0.48333).

!SPEC 3 !SPEC = 3 directs to include Bremsstrahluna radiation in aamma source.

,-CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level -** Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 26 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table *.

lnputParameters --Input Parameter

' Value Description NCON 1 NCON=1 directs to use flux to dose rate conversion factors for air (mrad-air/hr) per MeV/cm 2 -sec) !GEOM 11 This defines the source shield region geometry.

!GEOM = 11 is a cylindrical source with cylindrical and slab shields on cylinder side. NSHLD 5 #of Shields JBUF 22 Determine Buildup in Concrete Material # Length Units (T} Shield Thickness Description 1 1 203.327 This is the core radius (160.1"/2).

Length Units= 1 =cm 2 1 52.578 This is the water thickness between the core and the RPV. It is the RPV inner radius (205"/2) minus the circumscribed core radius (160.1"/2) minus the core shroud thickness (1.75"). 3 1 35.71875 This is air. Air was placed next because this region requires a non-cylindrical region that surrounds the first 2 cylindrical regions and air surrounds the vessel. The length is calculated as sacrificial shield outer radius ((24'-9&7/8")/2) minus the sum of the sacrificial shield wall thickness (2'3"), the RPV inner radius (205/2"), the RPV wall thickness (5&3/16"), and the RPV wall SS cladding thickness (3/16"). 4 1 18.10 This is the combined thickness of RPV wall iron, the SS cladding for the RPV wall, and the shroud thickness.

The combined thickness for these materials is (5-3/16" + 3/16" + 1.75" = 7.125" or 18.10 cm). 5 1 68.58 This is the sacrificial shield thickness.

SLTH 368.9086 cm This is the active fuel region height. Dose Point Coordinates:

x 408.78 cm X = sum of material number lengths 1 through 5 plus 30.48 cm for an approximate distance of 1 ft from the sac shield wall. y 184.4543 cm Y is half the ht of the active fuel region since the dose point is along the core midplane.

New Dose Point Location Parameters to Determine Dose Rate 1 ft from Sac Shield, alonq Core Bottom. NEXT 5 New Dose Point Coordinates Dose Point x 408.78 cm X, as defined above. y 0.0 cm Y is at bottom of the active fuel reqion.

CA-04-202 , "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s-ee s:;'-'m.Oy"

Client NSP Project MNGP Proj. No 12400-045 Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level X j Safety Related I j Non-Safety Related Prepared by Reviewed by Equip. No. Approved by FIGURE 1 MCNP Geometry Model Section View of Drywell and Refueling Floor Elevation (Sectional View is through the 'Y' Axis) Cale No. 2004-07061 Rev. 2 \Date Page 27 of 79 Date Date Date CA-0 4-202, "Dose Rates t o CHRRM Detectors Due to Drop in RPV Water Level" S-ree§...._,n .. y'" C l ient N SP P r o ject MN GP P r o]. No 12400-045 Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level X I Safety R elated Equip. No. I I Non-S afety Related Prepared by Reviewed by Approved by FIGURE 2 MCNP Model Section View of React o r Vessel Reactor Co r e Cale No. 2004-07061 Rev. 2 I Date Page 28 of 79 D ate D ate D ate CA-04-202 , "Dose Rates to CHRRM Detectors Due t o Drop in RPV Water Level" S..re8§0.....n"'y

"' Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 Cl i ent NS P P r o ject MN G P P r o j. No 12400-045 Mh Ng Ph In RPV Water Level Rev. 2 I Date X I Safet y Related I I Non-Safet y Related Page 29 Prepared by D ate Reviewed by D ate Equip. No. Approved by Date FIGURE 3 Reactor Building Refueling Floor , Elevation 1027'-8" Dryer Separator Storage Pool ARM-A3 4.1 5.1 6 t N Spent Fuel Pool M o n itor , C hannel B ARM A-1 , i New Fuel !.. ................................

'!::::;..../St orage Vau l t 6.9 Spent Fuel Pool Monitor , Channe l A 7.9 of 79 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 JDate X I Safety Related I I Non-Safety Related Page 30 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 5.2 Calculations

5.2.1 Deleted

5.2.2 MCNP Boundaries Boundaries used in the MCNP model are determined using elevation information presented in Design Input 3.5 and dimensions presented in Design Input 3.6. The origin (x=O, y=O, z=O) is established along the drywell centerline, at the surface of the concrete floor at the drywell bottom (i.e., elevation 920'-6").

Dimensions used in the MCNP model are in centimeters.

A conversion factor of 2.54 cm/inch is used (Design Input 3.11). Midpoint (Equator) of Drvwell Spherical Section Z = 942'-5"-920'-6" = 263" (668.02 cm) Bottom of RPV Z = (949'-5" -5&15/16" -3/16") -920'-6" = 340.875" (865.8225 cm) RPV Bottom Head Tangent Elevation Z = (949'-5" + (205"/2))-920'-6" = 449.5" (1141.73 cm) Bottom of Active Fuel z = (978'-8.5" -145.24") -920'-6" . = 978.708333'

-12.1033' -920.5' = 46.10503'

= 553.26036" (1405.2814 cm) Top of Active Fuel Region of Reactor Core Z = 978'-8.5"-920'-6" = 698.5" (1774.19 cm) Upper Grid & Top of Assemblies Z = (949'-5" + 367.125")-920'-6" = 714.125" (1813.8775 cm) Moisture Separator Flange Elevation Z = 981'-9.25" -920'-6" = 735.25" (1867.535 cm) Bottom of Steam Dryer Z = (949'-5" -920'-6") + 643.5" -189.5" = 801" (2034.54 cm) Top of Sacrificial Shield Z = 993'-7" -920'-6" = 877" (2227.58 cm) Top of Moisture Separator Z = (949'-5" -920'-6") + 554&5/8" = 901.625" (2290.1275 cm) Bottom Elevation of Drvwell Wall Cone Z = 995'-9" -920'-6" = 903" (2293.62 cm)

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" -...... § .... "-'" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X I Safety Related I \ Non-Safety Related Page 31 of . 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Top of Drvwell Wall Cone Z = 1001' -920'-6" = 966" (2453.64 cm) Top of Steam Drver Z = (949'-5" -920'-6") + 643.5" = 990.5" (2515.87 cm) Top of Concrete Ledge Above Drvwell Wall Cone Z = 1003'-7.25" -920'-6" = 997.25" (2533.015 cm) RPV Flange Elevation Z = 1004' -920'-6" = 1002" (2545.08 cm) Drvwell Head Tangent Z = (1004' + 13'-2&5/8" -920'-6") -6' = 1088.605" (2765.1075 cm) The 6' distance is the distance of the drywell head tangent below the underside of the drywell head (Assumption 4.3). Bottom of Lowest Drywell Shield Plug Z = (1027'-8" -3*2' -920'-6") = 1214.0" (3083.56 cm) Bottom of Middle Drvwell Shield Plug Z =Elevation of Lowest Shield Plug+ 2' = 1214.0" + 24" = 1238.0" (3144.52 cm) Bottom of Upper Drvwell Shield Plug Z =Elevation of Lowest Shield Plug+ 4' = 1214.0" + 48" = 1262.0" (3205.48 cm) Bottom of Refueling Floor Z = 1027'-8" -9" -920'-6" = 1277.0" (3243.58 cm) (1027'-8" is the top of the Refueling Floor El floor and 9" is the floor thickness, Table 3.6-1) Top of Refueling Area Floor Z = 1027'-8" -920'-6" = 1286.0' (3266.44 cm) Bottom of Refueling Area Roof Z = 1073'-2" -0.25" -920'-6" = 1831.75" (4652.645 cm) (0.25" per Assumption 4.1) Top of Refueling Area Roof Z = 1073'-2" -920'-6" = 1832.0" (4653.28 cm) Inner Surface of Reactor Bldg East Wall Between El 1001'-2" to 1026'-11" X = 22'-3" + 22'-3" + 22'-9" + 18" -12" = 813.0" (2065.02 cm) Inner Surface of East Refueling Floor Area Wall This wall is assumed to be 0.25" thick (Assumption 4.1 ).

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

..... y"' Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 32 of 79 Client Project Proj. No NSP -Prepared by MNGP Reviewed by 12400-045 Equip. No. Approved by X = 813" + 12" -0.25" = 824.75" (2094.865 cm) Outer Surface of Reactor Building East Wall X = 824.75" + 0.25" = 825" (2095.5 cm) Inner Surface of Reactor Bldg North Wall Between El 1001'-2" to 1026'-11" Y = 35'-3" + 15.5" -12" = 35.25' + 1.2917' -1' = 35.5417' (1083.311 cm) Inner Surface of North Refueling Floor Area Wall This wall is assumed to be 0.25" thick (Assumption 4.1 ). Y = 35.5417' + 1' -0.02083'

= 36.52087' (1113.156 cm) Outer Surface of Reactor Building North Wall Y = 35.5417' + 1' = 36.5417' (1113.791 cm) Inner Surface of Reactor Bldg West Wall Between El 1001'-2" to 1026'-11" X = -22'-3" -22'-3" -22'-9" -1.5' +12" = -67.75' (-2065.02 cm) Inner Surface of West Refueling Floor Area Wall This wall is assumed to be 0.25" thick (Assumption 4.1). X = -67.75' -1' + 0.02083' = -68.72917'

(-2094.865 cm) Outer Surface of Reactor Building West Wall X = -67.75' -1' = -68.75' (-2095.5 cm) Inner Surface of Reactor Bldg South Wall Between El 1001'-2" to 1026'-11" Y = -22'-3" -22'-3" -23' -15.5" +12" = -67.7917'

(-2066.291 cm) Inner Surface of South Refueling Floor Area Wall This wall is assumed to be 0.25" thick (Assumption 4.1 ). Y = -67.7917'

-1' + 0.02083' = -68.77087'

(-2096.136 cm) Outer Surface of Reactor Building South Wall Y = -67.7917'

-1' = -68.7917'

(-2096.771 cm) Active Core Region Outer Radial Dimension Radius (R) = (0.5x160.1")

= 80.05" (203.327 cm) Thermal Shroud Outer Radial Dimension R = (0.5 x 167.25") = 83.625" (212.4075 cm) Thermal Shroud Inner Radial Dimension R = 83.625" -1.75" = 81.875" (207.9625 cm) Date Date Date CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

'-'-'"°'Y'" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 jDate X I Safety Related I I Non-Safety Related Page 33 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Steam Drver Radial Dimension R = 0.5 x 201" = 100.5" (255.27 cm) RPV Inner Radial Dimension R = 0.5 x 205" = 102.5" (260.35 cm) Sacrificial Shield Inner Radial Dimension R = (0.5 x 24'-9&7/8")

-2'-3" = 121.9375" (309.72125 cm) Sacrificial Shield Outer Radial Dimension R = (0.5 x 24'-9&7/8")

= 148.9375" (378.30125 cm) Drvwell Head Inner Radius (Cylindrical Portion) Assume the drywell head outer radius is equivalent to the minimum radius of the concrete conical protrusion from the drywell wall (Assumption 4.4) R = (0.5 x 28'-2") -1&5/16" = 167.6875" (425.93 cm) Drvwell Head Outer Radius (Cylindrical Portion) R = (0.5 x 28'-2") = 169.0" (429.26 cm) Inner Radius of Drvwell Steel Liner (from drywell bulb to conical protrusion at El 995'-9") The drywell liner thickness in this section is assumed to be the average liner thickness in the cylindrical neck region (Assumption 4.5). Average thickness

= 0.5 x (0.635" + 1.5") = 1.0675" (use 1 "). R = (0.5 x 33'-6") -1" = 200" (508.0 cm) Outer Radius of Drvwell Steel Liner (from drywell bulb to conical protrusion at El 995'-9") This is also the inner radius of the concrete R = (0.05 x 33'-6") = 201" (510.54 cm) Inner Radius of Drywell Liner (cylindrical sect above conical protrusion, El 100"-7.25")

R = (0.5 x 30'-7") -.25" = 183.25" (465.455 cm) Outer Radius of Drvwell Liner (cylindrical sect above conical protrusion, El 100"-7.25")

This is also the inner radius of the concrete.

R = (0.5 x 30'-7") = 183.5" (466.09 cm) Inner Radius of Drvwell Liner (cyl sect below shield plugs, El 100"-7.25" to 1021'-8")

R = 16'-0.5" -.25" = 192.25" (488.315 cm) Outer Radius of Drywell Liner (cyl sect below shield plugs, El 100"-7.25" to 1021'-8")

This is also the inner radius of the concrete.

R = 16'-0.5" = 192.5" (488.95 cm)

L CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" .,..,,...,e§ Lun .. y*** Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 34 of 79 Client Project Proj. No NSP MNGP 12400-045 Equip. No. Drvwell Wall Outer Radius Above El 1001' R = 16'-0.5" + 4.5' = 246.5" (626.11 cm) Prepared by Reviewed by Approved by Inner Radius to Liner for Support of Bottom Row of Drvwell Shield Blocks R = 16'-6" -0.25" = 197. 75" (502.285 cm) Outer Radius to Liner for Support of Bottom Row of Drywell Shield Blocks This is also the inner radius of the concrete.

R = 16.5' = 198" (502.92 cm) Inner Radius to Liner for Support of Intermediate Row of Drvwell Shield Blocks R = 17'-2.5" -0.25" = 206.25" (523.875 cm) Outer Radius to Liner for Support of Intermediate Row of Drvwell Shield Blocks This is also the inner radius of the concrete.

R = 17'-2.5" = 206.5" (524.51 cm) Inner Radius to Liner for Support of Top Row of Drywell Shield Blocks R = (0.5 x 35'-6&5/8")

-0.25" = 213.0625" (541.17875 cm) Outer Radius to Liner for Support of Top Row of Drywell Shield Blocks This is also the inner radius of the concrete.

R = (0.5 x 35'-6&5/8")

= 213.3125" (541.81375 cm) Date Date Date 5.2.3 RPV Head Surfaces The RPV upper and lower heads are modeled as hemispheres.

The hemisphere is centered along the Z axis. To describe the hemisphere, the values for variables SZ and R must be provided.

SZ is the elevation of the midpoint for the hemisphere and R is the radius from the midpoint to a point on the hemisphere.

The RPV Upper and Lower Heads are shown in Figure 4. Upper RPV Head: The midpoint (SZ) is along the RPV flange elevation (2545.08 cm as calculated above). The radius, R, is the distance from the midpoint to the RPV head. For the inside surface of the RPV head: R = (1012'-6.5" -920'-6") -2545.08 cm= 260.35 cm The RPV head thickness is 3& 7 /16". At the RPV head outer surface, R = 260.35 cm+ 3&7/16" = 269.08125 cm Lower RPV Head: The midpoint (SZ) is along the elevation where the bottom head hemisphere meets the cylindrical portion of the RPV vessel (449.5", i.e., 1141.73 cm as calculated above). The radius, R, is the distance from the midpoint to the RPV head.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Client NSP Project MNGP Proj. No 12400-045 Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level X Safety Related Non-Safety Related Prepared by Reviewed by Equip. No. Approved by For the inside surface of the RPV head: R = 449.5" -(949'-5" -920.'-6")

= 102.5" (260.35 cm) The RPV head thickness is 6&2/16". At the RPV head outer surface, R = 260.35 cm + 6&2/16" = 275.9075 cm Cale No. 2004-07061 Rev. 2 Date Page 35 of 79 Date Date Date CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S.....,e§'--1>

..... y"' Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 36 of Cl.ient NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date RPV Upper Head Inner Radius (R = 260.35 cm) RPV Upper Head Outer Radius (R=269.08125 cm) RPV Lower Head Outer Radius (R=275.9075 cm) RPV Lower Head Intermediate Radii (R= 262.89 cm, 265.43 cm, 267.97 cm, 270.51 cm, and 273.05 cm) Figure 4 RPV Upper and Lower Head Models RPV Upper Head Sectional View RPVUpper Head Midpoint RPV Lower Head Sectional View RPV Upper Head Intermediate Radii (R= 262.89 cm, 265.43 cm, and 267.97 cm) I RPVFlange Elevation (SZ) At 2545.08 cm RPV Lower Head Midpoint Elevation where RPV Bottom Head Hemisphere Meets the Cylindrical Portion of the RPV. (SZ = 1141.73 cm) RPV Lower Head Inner Radius (R=260.35 cm) 79 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s--e8s;._._.,,.r Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 37 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 5.2.4 Drywell Bulb Surfaces The bulb portion of the drywell is a spherical opening having a midpoint along the equator at 942' -5" (Table 3.5-1 of Design Input 3.5). The bulb is truncated at the bottom by the drywell floor at elevation 920'-6". The bulb is truncated at the top by the cylindrical neck of the drywell. The internal radius of the drywell bulb is 31'-3" (Table 3.6-1 of Design Input 3.6). The cylindrical neck of the drywell is 33'-6" in diameter (Table 3.5-1 of Design Input 3.5). The bulbous portion of the drywell is shown in Figure 1 of this calculation.

The height, above the drywell bulb equator, where the cylindrical neck meets the drywell bulb is determined using the drywell bulb and cylinder neck radii. The angle where the cylindrical neck meets the drywell bulb is: Arcsin((33.5'/2)/31.25')

= 32.41 degrees. The height, above the equator where the cylindrical neck meets the drywell bulb is: 31.25' x cos(32.41°)

= 26.38' above the equator elevation of 942'-5" which is: 942'-5" -920'-6" + 26.38' = 579.56" (1472.0824 cm) above the drywell floor elevation of 920'-6". To describe the drywell bulb (i.e., sphere), the center of the sphere and the radius are required.

The sphere is centered along the Z axis and thus: SZ = 942'-5" -920'-6" = 263.0" (668.02 cm) The drywell sphere has a stainless steel liner. The liner varies in thickness between 11/16" and 2.5". The average thickness will be used (Assumption 4.5) and is: 0.5 x (11/16" + 2.5") = 1.59375".

A thickness of 1.5" will be used in this analysis.

The sphere radii are: Drywell Liner Inner Radius: R = 31 '-3" -1.5" = 373.5" (948.69 cm) Drywell Liner Outer Radius (Inner Radius for Drywell Concrete Wall) R = 31'-3" = 375" (952.5 cm) 5.2.5 Drywell Head Geometry The top 6 feet of the drywell head is modeled as half an ellipsoid, centered on the Z axis, with the x and y coordinate values equal. A sketch of an ellipsoid, with the parameters used to define the ellipsoid, is presented in Figure 5. The standard equation for an ellipsoid is: (x -x')2 + (y -y')2 + (z -z')2 = 1 a2 b2 c2 Equation 1

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" . "'-9es;._._,....,y"*

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 38 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date The MCNP equation used to define an ellipsoid is provided in Table 3.1 of Reference 8.16. The equation follows: A(x -x')2 + B(y -y 1)2 + C(z -z')2 + 2D(x -x') + 2E(y -y') + 2F(z -z') + G = 0 Equation 1 is solved to obtain the coefficients A, B, C, D, E, F, and G that are required for the MCNP equation.

The equation is solved for 3 drywell head ellipsoid surfaces, i.e., the underside surface of the drywell head, the outer surface of the drywell head, and an intermediate surface at half an inch into the drywell head thickness.

The drywell head is 1 &5/16" thick (Table 3.6-1 of Design Input 3.6). The center of the ellipsoid is at coordinates x', y', and z' which is in the tangent plane that is located 6 ft below the underside surface of the drywell head, along the z axis (Assumption 4.3). The tangent plane for the drywell head ellipsoid is 1088.605" (2765.1075 cm) above 920'-6" elevation, as calculated in Section 5.2.2 of this calculation.

The parameter, c, is the distance between the ellipsoid's center (in the tangent plane) and the ellipsoid's surface along the Z axis. Since the ellipsoid is centered along the Z axis, the parameters x' and y' are 0. The parameters a and b are the distances from the ellipsoid center along the x and y axes respectively.

Since the cross section of the ellipsoid is a circle in the x-y plane, the parameters a and b are equal. Setting x' and y' equal to 0 and a equal to b, Equation 1 can be simplified as follows: Equation 2 Equation 2 can be rearranged into the format used in the MCNP equation by multiplying both sides of Equation 2 by a 2 c 2. Since there are 3 ellipsoid surfaces (drywell head underside surface, drywell head outer surface and drywell head intermediate surface), there are 3 values for c. The values for parameter, c, are calculated as follows:

Ellipsoid Surface at Drywell Head Underside Surface (c is 6' above the tangent plane) c = 6' = 182.88 cm

Ellipsoid Surface at Half Inch into the Drywell Head Wall Thickness c = 6' + 0.5" = 72.5" (184.15 cm)

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S.....,e§._..._,, .. Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date. X I Safety Related I I Non-Safety Related Page 39 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date

Ellipsoid Surface at Drywell Head Outer Surface c = 6' + 1&5/16" = 73.3125" (186.21375 cm) The parameters a (and b) are the distances from the ellipsoid center to the ellipsoid surface in the tangent plane. Since there are 3 ellipsoid surfaces, there are 3 values for the parameters a and b (i.e., distance to drywell head inside surface, distance to drywell head outside surface, and distance to drywell head intermediate surface).

The distance to the inside surface and the outer surface are 425.93 cm and 429.26 cm respectively and are calculated in Section 5.2.2 of this calculation.

The distance to the intermediate ellipsoid surface is set at half the thickness of the drywell head, which is added to the drywell head inner surface, (i.e., 425.93 cm+ (0.5

1&5/16"

2.54 cm/inch) = 427.59687 cm). The surface coefficients for the MCNP equation can now be defined using the coefficients from the ellipsoid equation.

Specifically, A= c 2 , B = c 2 , C = a 2 , D = 0, E = 0, F = 0, and G = -a 2 c 2 The surface coefficients for the MCNP ellipsoid equation are calculated in Table 5.2-2 . ***** *\ Table 5.2-2 * . .* ... .. . ... . Prywell Head coefficients for MCNP

.... Dr'yWell Head Surfaces *. Coefficiel'lt

.. ** ... Outer Surface . ll'ltermediate Surface Inner Surface .. a 429.26 427.59 425.93 b 429.26 427.59 425.93 c 186.21375 184.15 182.88 z' 2765.1075 2765.1075 2765.1075 A(1) 3.467556E+04 3.391122E+04 3.344509E+04 8(1) 3.467556E+04 3.391122E+04 3.344509E+04 c'1) 1.842641 E+05 1.828332E+05 1.814164E+05 G(1) -6.389463E+09

-6.200098E+09

-6.067 487E+09 Notes (1) A= c 2 , B = c 2 , C = a 2 , and G = -a 2 c 2 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" sa...,ei._._,n..,,"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 40 of 79 Client NSP Project MNGP Proj. No 12400-045 Prepared by Reviewed by Equip. No. Approved by FIGURE 5 Drywell Head Ellipsoid Geometry Drywell Head Ellipsoid Origin (x', y', in Drywell Tangent Plane ZAxis Date Date Date XAxis 5.2.6 Drywell Cone Surfaces The drywell wall protrudes into the drywell volume between elevations 995'-9" and 1001' (Table 3.5-1 of Design Input 3.5). This is a height of 5'-3" or 63" (i.e., 1001' -995'-9").

The drywell diameter at the base of this protrusion is 33'-6" (concrete to concrete) and the diameter at the top of the protrusion is 28'-2" (Table 3.6-1 of Design Input 3.6). The depth of the protrusion at the top of the protrusion is 32" (i.e., 0.5 * (33'.:6" -28'-2")).

The vertical surface of this concrete protrusion is defined using circular conical geometry, centered on the Z axis (Refer to Figure 6). The MCNP equation which describes a cone centered on the Z axis is presented in Table 3.1 of Reference 8.16. The equation is repeated below. +y 2 -t(z-z')=O The MCNP inputs t, z, and +/-1 are calculated as follows. There are 2 cone surfaces (i.e., the face of the concrete protrusion from the drywell wall and the surface of the steel plate that lines CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 41 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date the face of the concrete protrusion.

The liner thickness is assumed to be the average thickness at the drywell cylinder neck (Assumption 4.5). The liner thickness at the cylinder neck varies between 0.635" and 1.5" (Table 3.6-1 of Design Input 3.6). The average liner thickness at the cylinder neck is: 0.5 x (0.635" + 1.5") = 1.0675" (Use 1 ") The parameter, t, is the tangent of the cone angle which is the protrusion length divided by the protrusion height. (Note: the MCNP equation indicates t, but the MCNP input parameter is t2). Table5.2-3 Calculation of tha Param'eter; t2, Used to Define .. a Cone Surface Protrusion f . Ht Concrete Wall 32" <1) 63" <1) (32"/63")

= 0.5079365 0.2579995 D ell Liner 2 2) 0.5079365

<2> 0.2579995 Notes: (1) The value is calculated in the first paragraph of this section. (2) The triangle formed by a vertical slice through the concrete protrusion forms a similar triangle with a vertical slice through the concrete protrusion and the steel liner. The values oft (i.e., the ratio of the protrusion length to protrusion height) are therefore the same for the cone formed by the concrete protrusion surface and the cone formed b the liner to the concrete rotrusion surface. Z is the elevation of the cone vertex. The length, Z, includes the height of the cone and the height from the drywell floor at elevation 920'-6" to the base of the cone. The cone's angle is the arctangent of the protrusion length divided by the protrusion height. For the concrete portion of the protruding wall, the angle is arctangent(32"/63")

= 26.927677°.

A vertical slice through the cone formed by the concrete protrusion and the cone formed by the wall liner on the surface of the concrete protrusion indicates that the cones form similar triangles and thus the cone angles are the same. For the cone defined by the concrete protrusion, the cone height is: Cone Height = (0.5 x 33'-6")/(tan(26.928°)) . = 395. 71324" (1005.1116 cm) Where 33'-6" is the drywell diameter at the base of the drywell wall protrusion.

For the drywell wall protrusion that includes the 1" thick wall liner, the base of the protrusion extends below the 995'-9" elevation (based on maintaining a drywell diameter of 33'-6" at the base of the protrusion and the top of the protrusion at 1001 '). The length of the protrusion is thus: 33"/(tangent (26.928°)

= 64.9678" Since the height of the concrete protrusion is 63", the protrusion which includes the liner extends down 1.9678". The cone vertex is lowered by this same amount. The base of the concrete protrusion is at El 995'-9". This is 903" (2293.62 cm) above the drywell CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s...__§Lun-***

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 42 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date concrete floor at elevation of 920'-6" (i.e., 995'-9" -920'-6").

Table 5.2-4 Calculation of the Parameter, Z, Used to Define a Cone Surface .. Height to Base of Cone Height{cm)

Z(cm). Protrusion (cm) : Concrete Wall 2293.62 1005.1116 3298.7316 Drvwell Liner ------3293.7333 (j) Notes (1) The height (Z) for the drywell liner is the height (Z) for the concrete wall minus 1.9678" as described above. Note: the MCNP input used a value of Z for the drywell liner cone of 3293.7449 cm. The difference is attributed to round-off errors used in the calculation, e.g., an angle of 26.927677 degrees was used when determining the value input into the code. There is no impact on the MCNP model or the results.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 43 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date FIGURE 6 Drywell Wall Cone Geometry Cone Vertex (concrete wall) "cone Vertex (wall liner) Drywell Radius = 28'-2"/2 Wall Liner ..........................................................

T...................................................

.. .................................

.. ** ........................................................... . Drywell Radius = 33'-6"/2 Client Project Proj. No CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" NSP MNGP 12400-045 Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level x Safety Related Non-Safety Related Prepared by Reviewed by Equip. No. Approved by Cale No. 2004-07061 Rev. 2 Date Page 44 of 79 Date Date Date 5.2.7 Material Compositions Material Regions used in this calculation consist of RPV regions and regions outside the RPV. Regions outside the RPV are composed of 1 of 4 material compositions (i.e., air, concrete, iron, or stainless steel). RPV material regions consist of the active core fuel region, fuel assembly extensions above the active core region, RPV water volumes, the moisture separator region, an overlap region that includes the moisture separator and steam dryer, a steam dryer region, a stainless steel region for the core shroud, and the RPV wall. Material inputs for MCNP consist of a material region's density and elemental compositions in weight percents.

Material densities and elemental compositions (wt%) are provided in Table 3.7-1 of Design Input 3.7 for steel (iron), air, and stainless steel. Water Water is composed of hydrogen (H) and oxygen (0). Its chemical composition is H20 and its density is 1.0 gram/cc. Hydrogen has an atomic mass of approximately 1 and oxygen has an atomic mass of approximately

16. The weight% of hydrogen in water is: ( 2 ) x 100 =11.1 wt% = 0.111 2+16 ' The weight% of oxygen in water is: ( 16 ) x 100 = 88.8 wt% = 0.889 2+16 Stainless Steel Elemental compositions for Cr and Ni in stainless steel are presented as a range of weight percents (Table 3.7-1 of Design Input 3.7). For type 304 SS, Cr is given a range between 18 and 20 wt% and Ni is given a range between 8 and 12 wt%. Similarly, for type 316 SS, Cr is given a range between 16 and 18 wt % and Ni is given a range between 1 O and 14 wt %. For this analysis, an average over the range of the Cr and Ni weight percents will be used (i.e., 19 wt % will be used for Cr and 10 wt % will be used for Ni for 304 SS and 17 wt % will be used for Cr and 12 wt% will be used for Ni for 316 SS). After accounting for all identified elemental wt%, the remainder is assigned to iron (Fe). The wt % of iron in stainless steel is thus: 304 SS: (1 -0.0008 -0.02 -0.01 -0.19 -0.1) = 67.92 wt% (Fe)= 0.6792 316 SS: (1 -0.0008-0.02-0.01-0.17-0.12)

= 67.92 wt% (Fe)= 0.6792 Where: 0.0008, 0.02, and 0.01 are weight fractions for C, Mn, and Si (Table 3.7-1 of Design Input 3.7), and 0.19 and 0.1 and 0.17 and 0.12 are weight fractions for Cr and Ni in 304 SS and 316 SS, respectively, as indicated above.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

... Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 45 of 79 Ciient Pmjeet Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Ordinary Concrete The elemental compositions for concrete are presented in Table 3.7-1 of Design Input 3.7. Elemental compositions for concrete are provided in units of density fractions.

The density fraction values are converted to units of elemental weight percents in Table 5.2-5. Table 5.2-5 Elemental ht Percents for Concrete Partial Densities Element Atomic# (glee) '1 l WeiQht Fraction (2) H 1 0.013 5.5E-03 O(inMix) 8 0.103 4.383E-02 (3 l 0 (in Ore & Cement) 8 1.068 4.545E-01 (3) Si 14 0.742 3.16E-01 Ca 20 0.194 8.255E-02 Na 11 0.04 1.70E-02 Mg 12 0.006 2.60E-03 Al 13 0.107 4.55E-02 s 16 0.003 1.30E-03 K 19 0.045 1.91 E-02 Fe 26 0.029 1.23E-02 Notes: (1) Partial densities are obtained from Table 3.7-1 of Design Input 3.7 and are for a concrete density of 2.35 glee. (2) Weight fractions are obtained by dividing the density fraction values in Column 3 by 2.35 glee. (3) The total weight fraction of oxygen is the sum of 4.383E-02

+ 4.545E-01

= 4.98E-01.

Active Core Region The active core region consists of fuel. (i.e., uranium dioxide), fuel assemblies, control rods, and water. The volume of the active fuel region is: Volume= 3.14159 x ((160.1"/2) x 2.54 cm/inch)2 x (145.24" x 2.54 cm/inch)=

4.7913724E+07 cc Where:3.14159

=Pi 160.1" = active core region diameter (Table 3.3-1) 145.24" =active core region height (Table 3.3-1) Note: The active core region volume is determined using the circumscribed diameter in Table 3.3-1 and not the equivalent diameter.

The circumscribed diameter was used since the core region is being homogenized over the entire volume occupied by the fuel assemblies.

Uranium Fuel Homogenized Densities in Active Core Region The initial mass of heavy metal (i.e., uranium) in the reactor core is 84.432 metric tons (Design CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" -.... ..... ""'"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 IDate X I Safety Related I I Non-Safety Related Page 46 of 79 Client Project Pr1Jj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Input 3.2). The homogenized uranium density in the active fuel region is: (84.432 metric tons x 1000 kg/ton x 1000 grams/kg)/

4.7913724E+7 cc= 1.76217 glee. The fuel is in the form of Uranium Dioxide (U02) and thus for every uranium atom there are 2 oxygen atoms. The atomic mass of oxygen is 16 and the atomic mass of uranium is 238 (Design Input 3.10). The homogenized density of oxygen in the active fuel region is determined as follows: The homogenized density of uranium in the active fuel region is 1. 76217 glee. The density of U02 fuel can be determined by solving for p in the following equation. ( 238 J ( ) xp=l.76217g/cc 2 x16 +238

Solving for the density, p, gives a homogenized U02 density of 1.9991004 glee The homogenized partial density of oxygen in the fuel is: ( 2 x16 J ( ) x 1.9991004

= 0.236930 g/cc 2x16 +238 Zirconium Homogenized Densities in Active Core Region The fuel cladding material is Zr-2 (zircaloy

2) and the material for the fuel channel is Zircaloy (Table 3.3-1 of Design Input 3.3). Both Zr-2 and Zircaloy are treated as pure zirconium (Zr) with a density of zircaloy, i.e., 6.56 glee (Assumption 4.9 and Table 3.7-1). The total volume of Zr in the cladding within the active fuel region is: [ (0 404" 2 54 ;* )2 l r ((0.404"-2(0.026"))

x 2.54 cm/in)2 "'68 9096 76 4841 n

x; """ x (368.9096cm)x 76 x 484 -l" 2 "

cmx x = 2.7030953E+6 cc Where: 0.404" is the cladding outside diameter (Table 3.3-1) 368.9096 cm is the active fuel length= 145.24 inches* 2.54 cm/in (Active fuel length is given in Table 3.3-1) 76 = number of fuel cladding tubes per assembly Assumption 4.1 O 484 =number of assemblies (Table 3.3-1) 0.026" is the fuel cladding thickness (Table 3.3-1) The volume of Zr in the square shaped fuel channel is: [(5.478" x 2.54 cm/inf-((5.478" -2 x (0.1")) x 2.54 cm/in)2] x 484 x 368.9096 cm CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 47 of 79 Client Project Proj. No NSP Prepared by MNGP Reviewed by 12400-045 Equip. No. Approved by .. = 2.478E+06 cc Where: 5.478" is the fuel channel outer dimension (Table 3.3-1) 0.1" is the fuel channel wall thickness (Table 3.3-1) 484 and 368.9096 are as previously defined (above) Date Date Date The total volume of Zr is the sum of the cladding volume and the fuel channel volume. Zr vol= 2.7030953E+06 cc+ 2.478E+06 cc= 5.1811E+06 cc The mass of zirconium in the active core region is: 6.56 glee x 5.1811 E+06 cc = 3.3988E+07 grams The homogenized density is: 3.3988E+07 grams/4.7913724E+7 cc= 0.70937 glee Water Homogenized Densities in Active Core Region Water occupies the area within fuel assemblies and the area between fuel assemblies.

The water to fuel (H 2 0/U0 2) volume ratio within the fuel assemblies is 2.84 (Table 3.3-1 of Design Input 3.3). The water volume in an assembly is determined using the U02 cross sectional area and the water to fuel volume ratio. The water cross sectional area for assemblies in the reactor core is: 2.84 x n( o.344" x cm/in r x 484 x 74 = 60991.543 sq cm Where: 0.344" is the fuel pellet diameter (Table 3.3-1 of Design Input 3.3) 484 =#of fuel assemblies in the core (Table 3.3-1 of Design Input 3.3) 7 4 = # of fuel rods per assembly (Assumption 4.10) The water volume is the water cross sectional area multiplied by the active fuel length of 145.24" (Table 3.3-1 of Design Input 3.3). Water Vol in Assemblies=

60991.543 cm 2

145.24"

2.54 cm/in = 2.25E+ 7 cc The homogenized water density in the active fuel region (due to water in assembly region) is: 1 glee x (2.25E+ 7 cc I 4. 7913724E+

7 cc) = 0.46959405 glee The partial density of Hydrogen = (2/18)

0.46959405 glee = 5.2177E-02 glee The partial density of oxygen is (16/18)

0.46959405

= 4.1741693E-01 glee CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s.....,§....._,n,.y***

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 48 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date The active fuel region cross sectional area is: n(1 6 0.1" x cm/in r 1.29879E + 05 cm 2 The cross sectional area of 484 assemblies is: 484 x (5.478" x 2.54 cm/in)2 = 9.37037E+04 cm 2 Where: 5.478" is the fuel channel outside dimension (Table 3.3-1) The gap size between fuel assemblies is: ((12" x 2.54 cm/in)-2 x (5.478" x 2.54 cm/in)) I 2 = 1.32588 cm Where: 12" is the pitch between control rods (Table 3.3-1) 5.478" is the fuel channel outside dimension (Table 3.3-1) The cross sectional area of 121 control rods is: 2 x (9.75" x 2.54 cm/in) x 0.5 x (1.32588 cm) x 121 rods= 3.9731 E+03 cm 2. Where: 9.75" =control rod width (Table 3.3-1) 121 =the number of control rods (Table 3.3-1) 1.32588 cm is the gap size between assemblies (calculated above) the factor of 0.5 accounts for the fraction of the gap width between fuel assemblies that is occupied by the control rod blades (Assumption 4.11). The volume of water between assemblies is: 368.909 cm x [1.29879E+05 cm 2 -9.37037E+04 cm 2 -3.9731 E+03 cm 2] = 1.18797E+07 cm 3 Where: 368.909 cm is the active core region height (145.24", Table 3.3-1) 1.29879E+05 cm 2 is the active fuel region cross sectional Area (previously calculated) 9.37037E+04 cm 2 is the cross sectional area of 484 assemblies (previously calculated)

The homogenized water density in the active fuel region (due to water in gaps between assemblies) is: 1 glee x (1.18797E+7 cc I 4.7913724E+7 cc)= 0.2479394 glee CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S--.,es;._.,.__,y'" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I I Non-Safety Related Page 49 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date The partial density of hydrogen in the water within the gaps between assemblies is: (2.) x 1 glee x ( 1*18797E + 07 em 3 J = 2.7549E-02 glee 18 4.7913724E

+ 07 em 3 The partial density of oxygen in the water within the gaps between assemblies is: x 1 glee x ( 1*18797 E+07 em 3 J = 2.2039E-01 glee 18 4.7913724E

+ 07 em 3 Control Rods Homogenized Densities in Active Core Region T,he volume of control rods in the active fuel region is: 121 rods x 2 x (9. 75" x 2.54 cm/in) x 1.32588 cm x (145.24" x 2.54 cm/in) = 2.9314E+06 cc Where: 121 rods, 9.75", 1.32588 cm, and 145.24" are as previously defined. The control rods are assumed to be stainless steel (Assumption 4.11 ). The mass of stainless steel in the active fuel region, associated with the control rods is: 2.9314E+6 cc x 7.9 glee= 2.3158E+07 grams Where 7.9 glee is the density of 304 stainless steel (Table 3.7-1 of Design Input 3.7). The homogenized density of stainless steel in the active fuel region is: (2.3158E+07 g / 4.7913724E+07 cc)= 0.48333 glee The partial densities for the elemental constituents of stainless steel are determined by multiplying the elemental weight percents provided in Table 3.7-1 of Design Input 3.7 by the homogenized density of 0.48333 glee. Elemental partial densities follow:

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s...-..ei....._.....,,,,..

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 50 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Table 5.2-6 Partial Densities of Stainless Steel Elements in Active Fuel Reaion Element Weiaht % (1 l Partial Densitv (2 l c 0.08 3.86664E-04 Mn 2.00 9.6666E-03 Si 1.0 4.8333E-03 Cr 19 9.18327E-02 Ni 10 4.8333E-02 Fe 67.92 3.28278E-01 Notes (1) Weight% for C, Mn, and Si in stainless steel are obtained from Table 3.7-1 of Design Input 3.7. Weight% for Cr and Ni are the average values from Table 3.7-1. The weight% for Fe is the remainder as calculated above. (2) Partial density =the homogenized stainless steel density (i.e., 0.48333 glee) multiolied bv the wt % value in column 2 and divided bv 100. Note: The homogenized stainless steel density associated with the control rods in the active core region was determined assuming that the width of the control rod blades is equal to the width of the gap between assemblies instead of half the gap width as stated in Assumption 4.11. Using a control rod blade widtl1 that is equal to the assembly gap width results in a homogenized control rod stainless steel density in the active core region that is twice what the homogenized density would be if the control rod blades were set at half the gap width (i.e., 0.483 glee versus 0.242 glee). The difference in densities (0.242 glee) is approximately 6% of the total homogenized density in the active core region. The total density in the active core region is 3.85 glee as calculated in Table 5.2-7. The slightly larger density in the active core region will result in slightly lower dose rates at the detector locations.

Since this calculation determines dose rates at radiation detectors for the purpose of establishing setpoints, the slightly lower dose rates are conservative.

A summary of elemental partial densities for all materials that were considered in the active core region is presented in Table 5.2-7.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 51 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date ' *, ' ' Table 5.2-7 *,******'

- . ' HomoqenizedElemental

Fractions for the Active* Core Reil ion .,,,. *.* .. *. ',' ..**.* ', Homoqenized Partial'Densities (qfcc) ,* I , '*' * ... ,' ,' ' ,' Material' inActive fueLReqion

.*. ... , ' ** .. * ,' ,, Water ' ', -Total .* Hohlhgenized Element*.*****

Fue1**

Water (In (Between*

Fuel Clad & Control Rods I Densities

, .** .. *,, .. *.*.Weight

*

Fractions<

1 l ' '* . ,

Assemblies)<

_ Assemblies)*

. Fuel Channel (glee) .1 ' , .. , u 1.76217 1.76217 4.508E-01 0 0.236930 4.17424E-O1 2.20392E-01 8.74746E-01 2.238E-01 H 5.21780E-02 2.75490E-02 7.97270E-02 2.039E-02 Zr 7.09368E-01 7.09368E-01 1.815E-01 c 3.86664E-04 3.86664E-04 9.891E-05 Mn 9.66664E-03 9.66664E-03 2.473E-03 Si 4.8333E-03 4.8333E-03 1.236E-03 Cr 9.18327E-02 9.18327E-02 2.349E-02 Ni 4.8333E-02 4.8333E-02 1.236E-02 Fe 3.28278E-01 3.28278E-01 8.397E-02 Total 3.90934 Notes: (1) The weight fraction is the elemental density in column 7 divided by the total density (3.90934 glee). For U, the homogenized weight fraction is (1.76217 q/cc / 3.90934 a/cc)= 4.508E-01 Assembly Region above Active Core Region Material densities in this region are determined based on fuel rod extensions (above the active core), the fuel rod plenum spring mass and the top end fitting mass. The region extends from the top of the active fuel region at elevation 978'-8.5" (Table 3.5-1 of Design Input 3.5) to the top of assemblies which are 367.125" above the top of the inside bottom of the reactor vessel (Table 3.5-1 of Design Input 3.5). Due to the limited size of this region, elemental densities associated with air are ignored. Height of Region= 367.125" -(978'-8.5" -949'-5") = 15.625" (39.6875 cm) Where: 949'-5" is the inside bottom surface of the reactor vessel (Table 3.5-1 of Design Input 3.5) This region has the same diameter as the active fuel region (i.e., 160.1" per Table 3.3-1 of Design Input 3.3). . . . (160.l"x2.54cm/in) 2 68 5 155 06 3 The volume of this region 1s thus: n 2 x 39. 75 cm= . E + cm The mass of zirconium due to fuel rod cladding in this region is: . " [ ( 0.404" x cm/in r -( (0.4 o 4" -)) x 254 cm/in)' l x 484 x 76 x 39.687 5 cm x 6.56 glee = 1.90765E+06 grams CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level Cale No. 2004-07061 Rev. 2 Date X Safety Related Non-Safety Related Page 52 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date The mass of zirconium due to the fuel channels in this region is: (5.4 78" x 2.54 cm/in)2 -(5.4 78" -2(0.1 ") x 2.54 cm/in)2 x 484 x 39.6875 cm x 6.56 glee = 1.749E+06 grams The homogenized density of Zr in this region is: (1.90765E+6 grams+ 1.749E+06 grams)/ 5.155E+06 cc= 0.7093 glee The mass of stainless steel in this region is due to the plenum springs and top end fittings.

The total mass of stainless steel in this region is: (1.1 + 2.0 ) kg/assembly

= 3.1 kg/assembly x 484 assemblies

= 1.50E+06 grams The homogenized stainless steel density for this region is: 1.50E+06 grams/5.155E+06 cc = 0.291 glee Total homogenized density in this region= 0.7093 + 0.291 = 1.00 glee Weight fractions for the elements present in this region are calculated in Table 5.2-8. When water in the reactor vessel is at the normal water level, the assembly region above the active core region is filled with water. The volume of the assembly region above the active core that is not occupied by the fuel rod extensions (with plenum springs within), the top end fittings, or assembly fuel channel material is filled with water. The water volume is: ( 2 1* [2 kg SS x 1000 x 484 assyl 5.155E+06cc-n(0.40 4"x 2*54 cm/in) x76x484x39.6875cm

-(1.749 E+0 6 gmsZrJ-assy kg 2 6.56 glee 7.9 glee = 3.559E+06 cc (volume occupied by water) The hydrogen density in the assembly region water above the active core region is: (-3._) x 1 glee x (3*559 E+0 6 cc)= 7.671E-02 glee 18 5.155E + 06 cc The oxygen density in the assembly region water above the active core region is: x 1 g/cc x (3*559 E+0 6 cc)= 6.137E-01 glee 18 5.155E+06cc CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 53 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Weight fractions for the elements present in the assembly region above the active core region when the region is filled with water are calculated in Table 5.2-8. Table 5.2-8 Homogenized Weight Fractions in the Assembly Region Above Active Fuel Region Homogenized Partial Densities (grams/cc)

Element Elemental Wt Material in Reqion Total Densities Homogenized

% For 304 SS Cladding & Plenum Springs & (g/cc) Weight Fuel Channels End Fittings Fractions

<1 l .. Elemental Weight Fractions With No Water in the Assembly Region Above the Active Fuel Region Zr 7.093E-01 7.093E-01 7.09E-01 c 0.08 2.328E-04 2.328E-04 2.33E-04 Mn 2.00 5.820E-03 5.820E-03 5.82E-03 Si 1.0 2.910E-03 2.910E-03 2.91E-03 Cr 19 5.529E-02 5.529E-02 5.53E-02 Ni 10 2.910E-02 2.910E-02 2.91E-02 Fe 67.92 1.976E-01 1.976E-01 1.98E-01 Total 1.00 Homoqenized PartialDensities (qrams/cc)

Material in Region Total Densities Homogenized Element Elemental Wt Clad & Springs & (g/cc) Weight %For 304 SS Water Fuel End Fractions

<1 l Channels Fittings Elemental Weight Fractions With Water in the Assembly Region Above the Active Fuel Reqion H 7.671E-02 7.671 E-02 4.54E-02 0 6.137E-01 6.137E-01 3.63E-01 Zr 7.093E-01 7.093E-01 4.19E-01 c 0.08 2.328E-04 2.328E-04 1.38E-04 Mn 2.00 5.820E-03 5.820E-03 3.44E-03 Si 1.0 2.910E-03 2.910E-03 1.72E-03 Cr 19 5.529E-02 5.529E-02 3.27E-02 Ni 10 2.910E-02 2.910E-02 1.72E-02 Fe 67.92 1.976E-01 1.976E-01 1.17E-01 Total 1.6910 Notes: (1) The homogenized weight fraction is the elemental density in column 5 (column 6 for the case with water) divided by the total density (1.00 glee no water and 1.691 glee with water). For Zr (and no water), the homogenized weight fraction is (0.7093 q/cc I 1.00 o/cc) = 0.7093 Steam Drver and Moisture Separator Regions The steam dryer and moisture separator regions are divided into 3 regions as shown in Figure 7. These are the top portion of the steam dryer region, the bottom portion of the moisture separator

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop . Cale No. 2004-07061 In RPV Water Level Rev. 2 Date X Safety Related Non-Safety Related Page 54 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date region, and an overlap region where the steam dryer overlaps the moisture separator.

The steam dryer and moisture separator are constructed of stainless steel (Table 3.6-1 of Design Input 3.6). Elemental weight fractions in these regions are determined for stainless steel and air. The steam dryer extends 189.5" downward from a height of 643.5" above the inside bottom of the reactor vessel (Table 3.5-1 of Design Input 3.5 and Table 3.6-1 of Design Input 3.6). The moisture separator extends from the moisture separator flange on El 981'-9.25" to 554&5/8" above the inside bottom of the reactor vessel. The bottom of the steam dryer is: 643.5" -189.5" = 454" (1153.16 cm) above the inside bottom of the reactor vessel. The top of the moisture separator is 554&5/8" x 2.54 cm/in= 1408.7475 cm above the inside bottom of the reactor vessel. The length of the moisture separator region is: 554&5/8" -(981'-9.25" -949'-5") = 554&5/8" -388.25" = 166.375" (422.5925 cm) The overlap region is: 1408.7475 cm -1153.16 cm= 255.5875 cm. The length of the steam dryer region above the overlap region is: (643.5" x 2.54 cm/in) -1408.7475 cm= 225.7425 cm The length of the moisture separator region below the overlap region is: 422.5925 cm -255.5875 cm= 167.005 cm.

l -CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ........

Cales. For Dose Rates to CHRRM Detectors Due to Drop In RPV Water Level X I Safety Related I I Non-Safety Related Cluent Project Proj. No NSP Prepared by MNGP Reviewed by 12400-045 Equip. No. Approved by FIGURE 7 Steam Dryer & Moisture Separator Regions El 643.5" Above Inside Bottom ofRPV El 554 & 5/8" Above Inside Bottom ofRPV El 454" Above Inside Bottom ofRPV 167.005 cm 100.5" (255.27 cm) Steam Dryer Upper Region Steam Dryer and Moisture Separator Overlap Region Moisture Separator Lower Region Cale No. 2004-07061 Rev. 2 I Date Page 55 of 79 Date Date Date CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ,,,.....,8s_;

...........

,,,.. Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 56 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Steam Dryer Upper Region The steam dryer region volume is: :ir( 2 0l"x 2;4 cm/in r x189.5"x2.54cm/in=9.8535417E+07 cc The volume of the steam dryer upper region is: (225.7425 cml(225.7425 cm+ 255.5875 cm)) x 9.8535417E+07 cc= 4.6213E+07 cc The homogenized stainless steel density is: (31tonsx2000 lb/ton x 453.5 g/lb ) = 0_28535 g/cc 9.8535417E

+ 07 cc The fraction of volume occupied by the stainless steel is: 0.28535 glccl8.0 glee= 0.03567 Therefore, the fraction of the volume occupied by air is : 1 -0.03567 = 0.96433 The air mass in this volume is: 1.293E-03 glee x 0.96433 x 4.6213E+07 cc= 5.7622E+04 grams The homogenized air density is: 5.7622E+04 gramsl4.6213E+07 cc= 1.2469E-03 glee The total homogenized density in the region is: (1.2469E-03

+ 0.28535) glee= 0.28660 glee. The homogenized weight fractions for this region are determined in Table 5.2-9.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

..... Y'"' Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X I Safety Related I \ Non-Safety Related Page 57 of 79 CIUent Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Table 5.2-9 Homoqenized Weiqht Fractions in the Steam Drver Upper Reqion Weiaht Fractions (wt%) Homogenized Wt Element Atomic#; Air 31688 Fractions

<1 l c 6 0.08 7.97E-04 N 7 75.5 3.28E-03 0 8 23.2 1.01 E-03 Si 14 1 9.96E-03 Ar 18 1.3 5.66E-05 Cr 24 17 1.69E-01 Mn 25 2 1.99E-02 Fe 26 67.92 6.76E-01 Ni 28 12 1.19E-01 Notes: (1) The homogenized wt fraction is calculated as follows:

For elements in air, the homogenized weight fraction is the elemental wt % from column 3 (divided by 100) and then multiplied by the homogenized air density for the region and divided by the total homogenized density for the region. For N, (75.5/100) x (1.2469E-03/0.2866)

= 3.28E-03

For elements in stainless steel, the homogenized weight fraction is the elemental wt % from column 4 (divided by 100) and then multiplied by the homogenized stainless steel density for the region and divided by the total homogenized density for the reqion. For C, (0.08/100) x (0.28535/0.2866)

= 7.97E-04 Moisture Separator Lower Region The moisture separator region volume is: n (167" x 2;4 cm/in r x (167.005 cm+ 255.5875 5.97189E + 07 cc The diameter of the moisture separator region is assumed to be 167" (Assumption 4.7). The volume of the moisture separator lower region is: (167.005 cm/(167.005 cm+ 255.5875 cm)) x 5.97189E+07 cc= 2.36E+07 cc The homogenized stainless steel density is: (41tonsx2000 lb/ton x 453.5 g/lb ) = 0_6227 glee 5.97189E + 07 cc The fraction of volume occupied by the stainless steel is: 0.6227 g/cc/7.9 glee= 0.0788 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 58 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Therefore, the fraction of the volume occupied by air is: 1 -0.0788 = 0.9212 The air mass in this volume is: 1.293E-03 glee x 0.9212 x 2.36E+07 cc= 2.811 E+04 grams The homogenized air density is: 2.811 E+04 gramsl2.36E+07 cc= 1.191 E-03 glee The total homogenized density in the region is: (1.191 E-03 + 0.6227) glee= 0.6239 glee The homogenized weight fractions for this region are determined in Table 5.2-10. . Table 5.2-10 ' Homogenized Weight Fractions ihthe MolstureSeparc:itorLoWef.Regipn I Weioht Fractions (wt%) Hbmogenized Wt. i :*,:! . .., * . .! 304 SS Fractions

<1 l ******** Element Atomic#! Air *Ii . .. *. . . . c 6 0.08 7.98E-04 N 7 75.5 1.44E-03 0 8 23.2 4.43E-04 Si 14 1 9.98E-03 Ar 18 1.3 2.48E-05 Cr 24 19 1.90E-01 Mn 25 2 2.00E-02 Fe 26 67.92 6.78E-01 Ni 28 10 9.98E-02 Notes: (1) The homogenized wt fraction is calculated as follows:

For elements in air, the homogenized weight fraction is the elemental wt% from column 3 (divided by 100) and then multiplied by the homogenized air density for the region and divided by the total homogenized density for the region. For N, (75.5/100) x (1.191E-03/0.6239)

= 1.44E-03

For elements in stainless steel, the homogenized weight fraction is the elemental wt % from column 4 (divided by 100) and then multiplied by the homogenized stainless steel density for the region and divided by the total homogenized density for the reoion. For C, (0.08/100) x (0.6227/0.6239)

= 7.98E-04 Steam Dryer and Moisture Separator Overlap Region The diameter of the overlap region is the same as the diameter of the steam dryer region, i.e., 201" and the height is 255.5875 cm.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s...._i .....

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 59 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date The volume of the overlap region is: 11:( 201" x 2;4 cm/in J x 255.5875 cm= 5.23226E + 07 cc The overall length of the moisture separator region is 422.5925 cm. The overall length of the steam dryer region is: (225.7425

+ 255.5875)cm

= 481.33 cm The mass of steam dryer stainless steel (type 316) in the overlap region is: 31 tons x 2000 lb/ton x 453.5 g/lb x (255.5875/481.33)

= 1.49302E+07 grams The mass of moisture separator stainless steel (type 304) in the overlap region is: 41 tons x 2000 lb/ton x 453.5 g/lb x (255.5875/422.5925)

= 2.2491 OE+07 grams Total mass of stainless steel in the region is (1.49302E+07

+ 2.2491 OE+07) = 3.7421 E+07 grams. The homogenized steam dryer stainless steel (type 316) density in this region is: 1.49302E+07 grams/5.23226E+07 cc = 0.28535 grams/cc The homogenized moisture separator stainless steel (type 304) density in this region is: 2.2491 OE+07 grams/5.23226E+07 cc= 0.42985 grams/cc The fraction of region volume occupied by the stainless steel (types 304 and 316) is: 0.28535 g/cc/8.0 glee+ 0.42985 g/cc/7.9 glee = 0.090080 Therefore, the fraction of the volume occupied by air is: 1 -0.090080 = 0.90992 The air mass in this volume is: 1.293E-03 glee x 0.90992 x 5.23226E+07 cc = 6.1559 E+04 grams The homogenized air density is: 6.1559E+04 grams/5.23226E+07 cc = 1.1765E-03 glee The total homogenized density in the region is: (1.1765E-03

+ 0.28535 + 0.42985) glee= 0.71638 glee CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ...,...,..s;.....,n-"" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 60 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date The homogenized weight fractions for this region are determined in Table 5.2-11. -***'** .. Table 1-Homogenized Weight.Fraetions-inthe Overlap Region kit ' ; ' ', Weight Fractions (wt%)' ,, -:Element Atomic#* Air Stainless Steel Homogenized Wt ' 304 i ' _-316 ,i, -Fractions

<1 J L ; ' ,;-,-_ c 6 0.08 0.08 7.99E-04 .. 1.24E-03 N 7 75.5 0 8 23.2 3.81 E-04 Si 14 1 1 9.98E-03 Ar 18 1.3 2.13E-05 Cr 24 19 17 1.82E-01 Mn 25 2 2 2.00E-02 Fe 26 67.92 67.92 6.78E-01 Ni 28 10 12 -1.08E-01 Notes: (1) The homogenized wt fraction is calculated as follows:

For elements in air, the homogenized weight fraction is the elemental wt % from column 3 (divided by 100) and then multiplied by the homogenized air density for the region and divided by the total homogenized density for the region. For N, (75.5/100) x (1.1765E-03/0.

71638) = 1.24E-03

For elements in stainless steel, the homogenized weight fraction is the sum of the product of the elemental wt% for moisture separator (type 304 SS from column 4) and the homogenized type 304 SS density in the region and the product of the elemental wt% for the steam dryer (type 316 SS from column 5) and the homogenized type 316 SS density divided by 100 and divided by the total homogenized density for the region. For Cr, [(19/100) x (0.42985/0.

71638 +(17/100) x (0.28535)/( 0.71638)]

= 1.82E-01.

Homogenized Densities when Water Level is at Normal Water Level The normal RPV water level is 35" above 477.5" above the bottom inside surface of the RPV which is at El 949'-5" (Table 3.5-1 of Design Input 3.5). The water level is: 35" + 477.5" = 512.5" (1301.75 cm) above RPV bottom The normal water level is below the top of the moisture separator which is 554&5/8" (i.e., 1408.7475 cm above the inside bottom of the reactor vessel). The water level is (1301.75cm

-1153.16 cm) = 148.59 cm above the bottom of the steam dryer. Elemental Weight Fractions in Moisture Separator Lower Region The volume fraction occupied by stainless steel in the moisture separator region is 0.0788 and the water volume fraction is 0.9212. The homogenized density of stainless steel in this region is CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" se...,..s:;._..n-" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 61 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date 0.6227 glee. The mass of water in the moisture separator lower region is: 1.0 glee x 0.9212 x 2.36E+07 cc= 2.174032E+07 grams The homogenized water density is: 2.174032E+07 gramsl2.36E+07 cc= 0.9212 glee. The total homogenized density in this region is: (0.9212 + 0.6227) = 1.5439 glee The homogenized weight fractions in the moisture separator lower region are determined in Table 5.2-12. Table 5.2-12 HomoQenized WeiQht Fractions in the Moisture Separator Lower ReQion WeiQht Fractions (wt%) Homogenized Wt Element Atomic# Water 304 SS

Fractions

<2 l H 1 11.11 (1) 6.63E-02 c 6 0.08 3.23E-04 0 8 88.89 <1) 5.30E-01 Si 14 1 4.03E-03 Cr 24 19 7.66E-02 Mn 25 2 8.07E-03 Fe 26 67.92 2.74E-01 Ni 28 10 4.03E-02 Notes: (1) The weight fraction of H in water is (2/18) = 11.11 % and the weight fraction for 0 is . (16/18) = 88.89%. (2) Homogenized wt fractions are calculated as follows:

For elements in water, the homogenized weight fraction is the elemental wt % from column 3 (divided by 100) and then multiplied by the homogenized water density for the region and divided by the total homogenized density for the region. For H, (11.11/100) x (0.9212/1.5439)

= 6.63E-02

For elements in stainless steel, the homogenized weight fraction is the elemental wt % from column 4 (divided by 100) and then multiplied by the homogenized stainless steel density for the region and divided by the total homogenized density for the region. For C, (0.08/100) x (0.6227/1.5439)

= 3.23E-04 Elemental Weight Fractions in Overlap Region which Contains Water The volume for the overlap region that contains water is: .*

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s ...... ei LLon .. y' ,. Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 62 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date r{ 201" x 2;4 cm/in )2 x 148.59 cm= 3.042E + 07 cc The volume fraction occupied by stainless steel in the overlap region is 0.09008 and the water volume fraction is 0.90992. The homogenized density of stainless steel in this region is 0.71520 glee (i.e., 0.28535 glee for 316 SS of the steam dryer + 0.42985 glee for 304 SS for the moisture separator).

The mass of water in the overlap region is: 1 glee x 0.90992 x 3.042E+07 cc= 2.76798E+07 grams The homogenized water density is: 2. 76798E+07 gramsl3.042E+07 cc = 0.90992 glee. The total homogenized density in this region is: (0.90992 + 0.71520) = 1.62512 glee The homogenized weight fractions in the water overlap region are determined in Table 5.2-13.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" s...-..eiL.un"'v

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X \ Safety Related I \ Non-Safety Related Page 63 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 5.2-13 HomoQenized WeiQht Fractions i.n the Water Overlap ReQion Weioht Fractions(wt

%) Homogenized Wt Element Atomic#.**

Water Stainless Steel Fractions

<2 l .* . 304** 316 H 1 11.11 <1 l 6.22E-02 c 6 0.08 0.08 3.52E-04 0 8 88.89 <1 J 4.98E-01 Si 14 1 1 4.40E-03 Cr 24 19 17 8.01 E-02 Mn 25 2 2 8.80E-03 Fe 26 67.92 67.92 2.99E-01 Ni 28 10 12 4.75E-02 Notes: ( 1) The weight fraction of H in water is (2/18) = 11.11 % and the weight fraction for 0 is (16/18) = 88.89%. (2) Homogenized wt fractions are calculated as follows:

For elements in water, the homogenized weight fraction is the elemental wt % from column 3 (divided by 100) and then multiplied by the homogenized water density for the region and divided by the total homogenized density for the region. For H, (11.11/100) x (0.90992/1.62512)

= 6.22E-02

For elements in stainless steel, the homogenized weight fraction is the sum of the product of the elemental wt% for moisture separator (type 304 SS from column 4) and the homogenized type 304 SS density in the region and the product of the elemental wt% for the steam dryer (type 316 SS from column 5) and the homogenized type 316 SS density divided by 100 and divided by the total homogenized density for the region. For Cr, [(19/100) x (0.42985/1.62512)

+ (17/100) x (0.28535/1.62512)]

= 8.01E-02.

5.2.8 Radiation

Source Region Boundary Definition The radiation source is homogenized over the active fuel region. The source extends from the bottom of the active fuel region to the top of the active fuel region. The radiation source region is centered along the z axis. The source region radius extends from the z axis to the outer radius of the active fuel region. Range of radiation source height along the z axis is: 1405.2814 cm (bottom of active fuel) to 177 4.19 cm (top of active fuel). The lower and upper limits for the active fuel region are calculated in Section 5.2.2 of this calculation.

The radial range of the radiation source is: 0.0 cm (at z axis location) to 203.327 cm (outer radius of active fuel region).

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 64 of 79 Client NSP Prepared by Date Project MNGP .. Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date The outer radius of the active fuel region is calculated in Section 5.2.2 of this calculation.

The volume of the radiation source region is: n (203.327 cm)2 x (1774.19 cm -1405.2814 cm)= 4.7913595E

+ 07 cc 5.2.9 Dose Point Locations The MCNP point detector option is used to determine dose rates at the locations of the 3 area radiation monitor detectors and the 2 spent fuel pool monitor detectors that are located on the refueling floor elevation of the reactor building.

The dose rates at the locations of the CHRRM detectors in the drywell are determined using the MCNP point detector option and the ring detector option. Detector Locations The refueling floor elevation of the reactor building is at elevation 1027'-8" (Table 3.5-1 of Design Input 3.5). The origin for the coordinate system used to develop the MCNP model is at elevation 920'-6" (Table 3.5-1 of Design Input 3.5). Distances between the drywell centerline and the detector locations on the refueling floor elevation of the reactor building are obtained from information in Table 3.6-1 of Design Input 3.6 and from Reference 8.12.7 and the sketch in Attachment 6 to this calculation.

Calculations of the coordinates for the detector locations follow. Channel A Spent Fuel Pool Monitor This detector is located at the crane support column, adjacent to the east reactor building wall, and is along building coordinate Ng (Design Input 3.4). Detector Elevation X Coordinate Y Coordinate

= (1027'-8" -920'-6") + 9'-9.5" above the floor (1286.0" + 117.5") x 2.54 cm/in= 3564.89 cm = 22'-3" + 22'-3" + 22'-9" -1' (267" + 267" + 273" -12") x 2.54 cm/in= 2019.3 cm The 1' distance puts the detector at approximately the face of the support column. = -10' -5'-6"-1.5'

= (-120"-66"-18")

x 2.54 cm/in= -518.16 cm 18" is an approximate distance to center the detector along the face of the support column Channel 8 Spent Fuel Pool Monitor This detector is located at the crane support column, adjacent to the north reactor building wall, and is along building coordinate 6.9 (Design Input 3.4). Detector Elevation

= (1027'-8" -920'-6") + 9'-8" above the floor (1286.0" + 116.0") x 2.54 cm/in= 3561.08 cm CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 loate X I Safety Related I I Non-Safety Related Page 65 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date X Coordinate

= (22'-3") = (267") x 2.54 cm/in = 678.18 cm Y Coordinate

= 35'-3" -3' = (423.0" -36") x 2.54 cm/in = 982.98 cm Area Radiation Monitor A-1 This detector is located 5'-4" west of building coordinate 7.9 and along the north reactor building wall (Design Input 3.4). Detector Elevation X Coordinate Y Coordinate

= (1027'-8" -920'-6") + 9'-3" above the floor (1286.0" + 111.0") x 2.54 cm/in= 3548.38 cm = (22'-3" + 22'-3" -5'-4") = (470") x 2.54 cm/in = 1193.8 cm = 35'-3" -3' = (423.0" -36") x 2.54 cm/in = 982.98 cm Area Radiation Monitor A-2 This detector is located at the crane support column adjacent to building coordinate 6 and along the north reactor building wall (Design Input 3.4). Detector Elevation X Coordinate Y Coordinate = ( 1027'-8" -920'-6") + 6'-6" above the floor (1286.0" + 78.0") x 2.54 cm/in= 3464.56 cm =O = 35'-3" -3' = (423.0" -36") x 2.54 cm/in = 982.98 cm Area Radiation Monitor A-3 This detector is located at the crane support column adjacerit to building coordinate Ph and along the west reactor building wall (Design Input 3.4). Detector Elevation X Coordinate Y Coordinate CHRRM Detectors = ( 1027'-8" -920'-6") + 6'-7 .5" above the floor (1286.0" + 79.5") x 2.54 cm/in= 3468.37 cm = -22'-3" -22'-3" -22'-9" + 1' = (-795") x 2.54 cm/in= -2019.3 cm = -22'-3" -22'-3" + 2'-8.5" = (-534" +32.5") x 2.54 cm/in = -1273.81 cm The CHRRM detectors (Channels A and B) are located at elevation 947'-9" in the drywell and are 6" to 8" from the drywell wall (Design Input 3.4). For this evaluation, the detectors will be placed 6" from the wall. The detectors are in the spherical portion (i.e., the bulbous portion) of the drywell. The equator for the drywell spherical portion is at elevation 942'-5" (Table 3.5-1 of Design Input 3.5). The radius of the spherical portion of the drywell is 31 '-3" (Table 3.6-1 of Design Input 3.6). The angle between the drywell spherical portion equator and the CHRRM detector locations CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Sare§L<>,,_".

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 jDate X I Safety Related I I Non-Safety Related Page 66 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date is: e . ((947'-9"-942'-5")) . ((11373"-11309"))

998798940

= arcs1n = arcs1n = . (31'-3"-6")

375"-6" The horizontal distance (D) between the reactor vessel centerline and the CHRRM dose point locations is: D = (31 '-3" -6") x cos(9.9879894°)

= (375" -6") x 2.54 cm/in x cos(9.9879894°)

= 923.055 cm The elevation of the dose point with respect to the origin at 920'..:6" is: (947'-9" -920'-6")

= 327" x 2.54 cm/in= 830.58 cm CHRRM Channel A is located at 0° Azimuth which is in the positive direction along the y axis, and Channel Bis located at 180° Azimuth (Design Input 3.4). The detector dose point location coordinates are: Channel A: x = 0.0, y = 923.055 cm, z = 830.58 cm Channel B: x = 0.0, y = -923.055 cm, z = 830.58 cm In addition to the dose point location, MCNP input also requires a radius of a sphere of exclusion around the dose point. The sphere of exclusion is determined by converting the detector volume into a sphere of equivalent volume. * ) The CHRRM detectors are 2.5" in diameter and 9.9" long (Design Input 3.4). The volume of the CHRRM detector is: { Z.5" x cm/in r x 9.9" x 2.54 cm/in = 796.354 cm 3 The volume of a sphere is (4/3) n r 3. The radius of a sphere with a volume of 25.146 cm 3 is: 1 1 V J3 =[(:)x(!)

x 796.354cm 3 ]' =5.75 cm The radius for the sphere of exclusion is 5.75 cm (use 6 cm). A 6 cm sphere of exclusion is also applied to the dose points for the detectors on the refueling floor elevation.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

... Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 I Date X I Safety Related I I Non-Safety Related Page 67 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Ring Detectors Due to the symmetry of the locations of the CHRRM, Channel A and B detectors, the detector locations can be described as a ring detector.

The MCNP inputs required to describe a ring detector are the elevation where the ring plane intercepts the z axis, the radius of the ring, and the sphere of exclusion around a point selected on the ring. The elevation of the ring is the elevation of the CHRRM detectors with respect to the z axis origin at 920'-6" elevation.

This elevation is 830.58 cm as previously calculated.

The radius of the ring is the radius to the drywell wall at the detector elevation minus the distance that the detector is located away from the drywell wall (i.e., 6"). The radius of the ring is 923.055 cm as previously calculated.

The sphere of exclusion is 6 cm as calculated above.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 68 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 6.0 Results The results consist of output from the RUNT-PC and MCNP computer codes. RUNT-PC Output The RUNT-PC input file uses the reactor core radionuclide inventories presented in Design Input 3.1. These radionuclide inventories are the MNGP EPU core inventories for an end of cycle core with an average burnup of 35 GWD/MT. The BAFFLE subroutine of RUNT-PC is used to decay the core inventories for times greater than core shutdown.

The ISOSHLD portion of the RU PC computer code was used to define the geometry and calculate dose rates. A summary of RUNT-PC calculated dose rates is presented in Table 6-1. The RUNT-PC computer code is maintained under Sargent & Lundy's QA program and therefore independent verification and validation of the code is not required.

The code was run on PC# ZD3818. The RUNT-PC output file is identified and described in Attachment 1 to this calculation.

RUNT-PC results (i.e., the output file) are presented in Attachment 7 to this calculation.

Dose rates presented in Table 6-1 are determined at 8-hours post shutdown which is the time it takes to get the reactor to cold shutdown conditions (Design Input 3.9). Table 6-1 RUNT-PC and MCNP Dose Rates Dose Point 8 Hour Post Shutdown Dose Rates 7mrad-air/hr)

RUNT-PC Results MCNP Resultsl'I Along Core Midplane, 1 ft Outside Sacrificial Shield 7.66E+01 5.79E+01 Core Bottom Elevation, 1 ft Outside Sacrificial Shield 3.838E+01 2.90E+01 Notes: (1) MCNP dose rates are for the configuration where the steam dryer and moisture separator are in the reactor pressure vessel with the reactor pressure vessel head and drywell head on and the water level at the top of active fuel. This configuration most closely represents the model used in the RUNT-PC analysis since this configuration, minimizes the scatter dose rate contribution to the dose points. Dose rates in column 3 are average (i.e., mean) MCNP calculated dose rates which are 1000 times the product of the MCNP dose rate from column 2 of Table 6-3 (rad-air/hr per gamma/sec) and the total gamma source strength (i.e., 1.566E+20 gammas/sec) from column 4 of Table 6-3. The MCNP calculated dose rate along core midplane is 5.79E+01 mrad-air/hr (i.e., 1000 x 3.70049E-22 rad-air/hr per gamma/sec x 1.566E+20 qammas/sec).

As indicated in Table 6-1, RUNT-PC calculated dose rates (at 1 foot outside the sacrificial shield -at the core midplane elevation and core bottom elevation) are larger than the MCNP calculated dose rates. MCNP dose rates in Table 6-1 are obtained from Table 6-3 and are for the configuration in which the moisture separator and steam dryer are in the RPV and the reactor head and the drywell head are in place. This configuration is used to compare MCNP calculated dose rates to RUNT-PC calculated dose rates because RUNT-PC calculated dose rates do not include scatter radiation.

The inclusion of the moisture separator and steam dryer within the RPV CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 \Date X I Safety Related I I Non-Safety Related Page 69 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date and the RPV and drywell heads in place minimizes scatter radiation contribution to the dose point locations.

The RUNT-PC code is used to obtain the reactor core gamma energy spectrum (i.e., energy dependent gamma source strength) for use in the MCNP code. The reactor core gamma energy spectrum at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months following reactor shutdown, is provided in Table 6-2. Table 6-2 Reactor Core Gamma Source StrenQth at 8 Hours After Reactor Shutdown Photon Average Decay Gammas Bremsstrahlung Total Gammas Energy Energy (Photons/sec)

Gammas (Photons/sec)

Group (MeV) (Photons/sec) 1 1.500E-02 2.404E+19 9.004E+18 3.304E+19 2 2.500E-02 2.750E+18 2.271E+18 5.021E+18 3 3.500E-02 5.119E+18 1.467E+18 6.586E+18

4. 4.500E-02 1.651E+18 1.089E+18 2.740E+18 5 5.500E-02 3.829E+17 8.526E+17 1.235E+18 6 6.500E-02 4.207E+17 6.766E+17 1.097E+18 7 7.500E-02 1.530E+16 5.391 E+17 5.544E+17 8 8.500E-02 1.644E+18 4.456E+17 2.090E+18 9 9.500E-02 5.543E+18 1.202E+18 6.745E+18 10 1.500E-01 2.665E+19 1.352E+18 2.800E+19 11 2.500E-01 1.630E+19 5.620E+17 1.686E+19 12 3.500E-01 4.838E+18 2.980E+17 5.136E+18 13 4.750E-01 1.068E+19 1.877E+17 1.086E+19 14 6.500E-01 1.538E+19 9.375E+16 1.548E+19 15 8.250E-01 1.091E+19 4.603E+16 1.095E+19 16 1.000E+OO 2.502E+18 2.823E+16 2.530E+18 17 1.225E+OO 1.783E+18 1.625E+16 1.799E+18 18 1.475E+OO 4.651E+18 7.602E+15 4.658E+18 19 1.700E+OO 4.431E+17 3.436E+15 4.465E+17 20 1.900E+OO 1.799E+17 1.762E+15 1.817E+17 21 2.100E+OO 2.003E+17 9.442E+14 2.012E+17 22 2.300E+OO 1.580E+17 4.995E+14 1.585E+17 23 2.500E+OO 1.631E+17 2.636E+14 1.634E+17 24 2.700E+OO 8.797E+15 1.787E+14 8.976E+15 25 3.000E+OO 1.776E+16 1.024E+14 1.786E+16 26 6.143E+OO 3.389E+14 O.OOOE+OO 3.389E+14 27 7.112E+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO TOTAL 1.364E+20 2.014E+19 1.566E+20 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S..r-§.......-.-***

Cales. For Dose Rates to CHRRM Detectors Due to Drop Cale No. 2004-07061 In RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 70 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date MCNP Output MCNP output gives the dose rate (in rad-air/hour) per gamma particle per second. The MCNP generated dose rates are multiplied by the number of gamma particles per second in the source region in order to obtain the total dose rate (in units of rad-air/hour).

The output from the radiation detectors is in Roentgen (R)/hour.

To convert MCNP dose rate output from units of rad-air/hour to units of R/hour, the MCNP generated dose rate is divided by 0.877 (Design Input 3.11). Since MCNP is a statistical radiation transport code, the 95% confidence interval is calculated for all dose rate calculations.

MCNP provides 2 dose rate numbers. The first number is the mean (i.e., expected) dose rate on a per particle basis. The second number is the relative error of the mean. Equation 2.19a of the MCNP User's Manual (Reference 8.16) describes the relative error as the ratio of the standard deviation of the mean to the mean. The mean +/-. 2 times the standard deviation is the 95% confidence interval.

Since this calculation is determining the dose rates at radiation detectors with the intent of establishing a monitor setpoint, the dose rate at the lower range of the 95% confidence interval is presented in the dose rate results. The reported dose rate is thus the mean dose rate minus 2 times the standard deviation.

Table 6-3 presents MCNP generated mean dose rates (on a particle basis) and the relative errors of the means. Also presented in Table 6-3 are the calculated dose rates over all source particles and the dose rate converted to units of R/hr and mR/hr. The MCNP computer code is maintained under Sargent & Lundy's QA program and therefore independent verification and validation of the code is not required.

The code was run on desktop PCs (ZD2318, ZD2840, ZD2841, ZD3818, ZD6188, and ZD6710). A list of MCNP files is presented in Attachment 1 to this calculation.

MCNP results (i.e., output files) are presented in Attachments 8 through 21 of this calculation.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level Cale No. 2004-07061 Rev. 2 Date X Safety Related Non-Safety Related Page 71 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 6-3 8-Hour Post Shutdown Dose Rates at Detector Locations:

MCNP Output (5) MCNP Dose Rate Relative Error Source (2 l Total Dose Rate Adjusted Dose Rate Dose Point Location (Rad-air/hr per Gamma/sec)

(1) Fraction (1 l (Gamma/sec) (Rad-air/hr)

(3 l (R/h) (4 l (mR/hr) (s) Steam Dryer and Moisture Separator in RPV, RPV and Drvwell Heads on, Normal RPV Water Level CHRRM Channel A 5.68862E-24 3.07856E-02 1.566E+20 8.35E-04 9.53E-04 9.53E-01 Steam Drver and Moisture Separator in RPV, RPV and Drywell Heads on, RPV Water Level at Top of Active Fuel Dose Rate Alonq Core Midolane, Outside Sac Shield 3.70049E-22 1.56285E-02 1.566E+20 5.61E-02 6.40E-02 6.40E+01 Dose Rate at Core Bottom Elevation, Outside Sac Shield 1.85360E-22 1.90461 E-02 1.566E+20 2.79E-02 3.18E-02 3.18E+01 CHRRM 6.13097E-24 4.47629E-02 1.566E+20 8.74E-04 9.97E-04 9.97E-01 Refuelinq Elevation Monitor ARM A-1 2.56259E-26 3.83347E-02 1.566E+20 3.70E-06 4.22E-06 4.22E-03 Refuelinq Elevation Monitor ARM A-2 5.91478E-26 3.72692E-02 1.566E+20 8.57E-06 9.77E-06 9.77E-03 Refuelinq Elevation Monitor ARM A-3 1.21512E-26 4.04159E-02 1.566E+20 1.75E-06 1.99E-06 1.99E-03 Spent Fuel Pool Monitor, Channel A 1.56473E-26 3.85409E-02 1.566E+20 2.26E-06 2.58E-06 2.58E-03 Spent Fuel Pool Monitor, Channel 8 4.58567E-26 3.84385E-02 1.566E+20 6.63E-06 7.56E-06 7.56E-03 Steam Dryer and Moisture Separator in RPV, RPV and Drywell Heads Removed, RPV Water Level at Top of Active Fuel Ds Pt 1' Above Drvwell Head 4.29319E-23 5.19380E-02 1.566E+20 6.02E-03 6.87E-03 6.87E+OO Spent Fuel Pool Monitor, Channel A 2.17713E-25 4.90536E-02 1.566E+20 3.07E-05 3.51E-05 3.51E-02 Spent Fuel Pool Monitor, Channel 8 6.23725E-25 5.00737E-02 1.566E+20 8.79E-05 1.00E-04 1.00E-01 Refueling Elevation Monitor ARM A-1 3.58566E-25 5.10113E-02 1.566E+20 5.04E-05 5.75E-05 5.75E-02 Refueling Elevation Monitor ARM A-2 8.19462E-25 5.26801 E-02 1.566E+20 1.15E-04 1.31E-04 1.31 E-01 Refueling Elevation Monitor ARM A-3 1.74759E-25 4.46604E-02 1.566E+20 2.49E-05 2.84E-05 2.84E-02 Separator in RPV, Stearn Dryer RemovedfromRPV, RPV and Prywell Heads Removed, RPV Water Level at Top of Active Fuel Dose Rate Alona Core CL, 1' Above Drvwell Head El. 7 .10257E-20 3.07048E-02 1.566E+20 1.04E+01 1.19E+01 1.19E+04 CHRRM 6.42062E-24 1.34006E-02 1.566E+20 9.78E-04 1.12E-03 1.12E+OO Refuelinq Elevation Monitor ARM A-1 2.61213E-22 2.23491 E-02 1.566E+20 3.91E-02 4.45E-02 4.45E+01 Refueling Elevation Monitor ARM A-2 5.20230E-22 2.48124E-02 1.566E+20 7.74E-02 8.83E-02 8.83E+01 Refueling Elevation Monitor ARM A-3 1.33243E-22 2.22941E-02 1.566E+20 1.99E-02 2.27E-02 2.27E+01 Spent Fuel Pool Monitor, Channel A 1.69730E-22 2.21461E-02 1.566E+20 2.54E-02 2.90E-02 2.90E+01 CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Wpter Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level Cale No. 2004-07061 Rev. 2 JDate x I Safety Related I I Non-Safety Related Page 72 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date Table 6-3 8-Hour Post Shutdown Dose Rates 13t Detector Locations:

MCNP Output (B) MCNP Dose Rate Relative Error Source (Z) iotal Dose Rate Adjusted Dose Rate Dose Point Location (Rad-air/hr per Gamma/sec)

(1) Fraction (1 l (Gamma/sec) (Rad-air/hr)

(3 l (R/h) (4 l (mR/hr) (5 l Spent Fuel Pool Monitor, Channel B 4.10581 E-22 2.33923E-02 1.566E+20 6.13E-02 6.99E-02 6.99E+01 Moisture Separator and Steam Dryer Removed from RPV, RPV and Drywell Heads Removed, RPV Water Level at Top of Active Fuel Dose Rate Alonq Core CL, 1' Above Drvwell Head El, 5.61274E-17 1.04909E-02 1.566E+20 8.60E+03 9.81 E+03 9.81E+06 Dose Rate Along Core Midplane, Outside Sac Shield 2.88732E-21 3.92153E-02 1.566E+20 4.17E-01 4.75E-01 4.75E+02 Dose Rate at Core Bottom Elevation, Outside Sac Shield 1.72940E-21 4.04373E-02 1.566E+20 2.49E-01 2.84E-01 2.84E+02 CHRRM 4.98002E-23 3.21462E-02 1.566E+20 7.30E-03 8.32E-03 8.32E+OO Refueling Elevation Monitor ARM A-1 3.94492E-20 5.23158E-02 1.566E+20 5.53E+OO 6.31E+OO 6.31 E+03 Refueling Elevation Monitor ARM A-2 8.48189E-20 4.83569E-02 1.566E+20 1.20E+01 1.37E+01 1.37E+04 Refueling Elevation Monitor ARM A-3 1.85924E-20 4.68273E-02 1.566E+20 2.64E+OO 3.01E+OO 3.01E+03 Spent Fuel Pool Monitor, Channel A 2.66429E-20 4.70696E-02 1.566E+20 3.78E+OO 4.31E+OO 4.31E+03 Spent Fuel Pool Monitor, Channel B 6.66121 E-20 4.59401 E-02 1.566E+20 9.47E+OO 1.08E+01 1.08E+04 Notes: (1) Values are from MCNP Output. (2) The total gamma source is the total value from column 5 of Table 6-2 and is the sum total of columns 3 and 4 of Table 6-2. (3) The total dose rate is obtained by subtracting 2 times the product of the value in column 2 and the value in column 3 from the value in column 2 and multiplying the result by the gamma source strength value in column 4. The product of the value in column 2 (i.e., the mean dose rate per gamma per second) and the relative error value in column 3 gives the standard deviation from the mean. Two times the standard deviation represents the 95% confidence range. Subtracting 2 times the standard deviation from the mean gives the lower value of the 95% confidence range. The dose rate value (per gamma per second) at the lower end of the 95% confidence range is multiplied by the total gamma source strength value in column 4 to obtain the dose rate over all gamma particles.

For example, the total dose rate at Spent Fuel Pool Monitor (Channel B) in the last row of the table is: 1.566E+20

(6.66121 E-20 -(2

4.59401 E-02

6.66121 E-20)) = 9.47E+OO (4) The adjusted dose rate (in units of R/hr) is the dose rate in Rad-air/hr of column 5 divided by 0.877 Rad-air/R (Design Input 3.11 ). The adjusted dose rate (in R/hr) at the Spent Fuel Pool Monitor (Channel B) in the last row of the table is: 9.47 Rad-air/hr I 0.877 Rad-air/R

= 1.08E+01 R/hr. (5) The dose rate in units of mR/hr is the dose rate in units of R/hr multiplied by 1000 mR/R. (6) Dose rates are due to radionuclide activity that originates in the active fuel portion of the reactor core.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ........

Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 73 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date The dose rates at the CHRRM detector locations increase with removal of the drywell and RPV heads, with removal of the steam dryer and moisture separator from the RPV, and with a drop in the water level in the RPV. Table 6-3 indicates that the calculated dose rates at the CHRRM detector locations vary from approximately 0.9 mR/hr to approximately 8 mR/hr. The smallest dose rate is for the configuration in which the moisture separator and steam dryer are within the RPV and the RPV head and the drywell head are in place. The largest dose rate is for the configuration in which the moisture separator and the steam dryer are removed from the RPV and the RPV head and drywell head are removed which maximizes the radiation that is available for scatter and therefore maximizes the scatter contribution to the dose rate.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ,,......,§.__..

... ,, ... Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 74 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date .0 Conclusions This evaluation determines dose rates at CHRRM locations in containment and at radiation detector locations on the refueling floor elevation of the reactor building in order to determine whether the radiation monitors can be used to detect a loss of water inventory in the reactor vessel during reactor shutdown conditions.

The concrete shield blocks over the reactor cavity are removed and not a consideration in this evaluation.

Since dose rates at the radiation detector locations are dependent on the water level in the reactor vessel and on the presence or exclusion of the moisture separator, the steam dryer, and the reactor vessel and drywell heads; dose rates were determined for different configurations of the reactor vessel. Dose rates were determined at 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after reactor shutdown, which is the start of cold shutdown and the minimum time after reactor shutdown when reactor vessel disassembly begins. Use of CHRRM Detectors The Containment High Range Radiation Monitor (CHRRM) detectors (RM7860, Channels A and B) are located in the spherical portion of the drywell and are approximately 20 feet below the bottom of the active fuel region of the reactor core. The detectors are within inches of the drywell wall and are shielded from direct radiation shine from the reactor vessel and hidden from much of the scattered radiation.

Calculated dose rates at the CHRRM locations do not exceed 1 O mR/hr when the water level in the RPV is at the top of active fuel, the moisture separator and steam dryer are removed from the reactor vessel and the RPV head and the drywell head are removed. The calculated dose rates are due solely to direct and scattered radiation from radionuclide activity within the reactor core. The lower range of the CHRRM detectors is 1 R/hr (Design Input 3.4). Since the calculated dose rate at the CHRRM detector locations is less than 1 R/hr, the calculated dose rates do not meet the acceptance criteria established in Section 2.0 of this calculation.

Use of Detectors on Refueling Floor Elevation Area Radiation Monitors A-1, A-2, and A-3 and Spent Fuel Pool Radiation Monitors Channel A and B are located on the refueling floor elevation of the reactor building (see Figure 3). Area Radiation Monitor A-2 and Spent Fuel Pool Monitor (Channel B) are closest to the reactor cavity (Figure 3). The dose rates at these detectors are expected to be greater than the dose rates to the locations of the other detectors on the refueling floor elevation due to their proximity to the reactor cavity. The calculated dose rates at the locations of these radiation detectors are significantly less than 0.1 mR/hr when the moisture separator and steam dryer are in the reactor vessel, and the RPV head and drywell head are in place. Calculated dose rates in this evaluation are due solely to radioactivity in the reactor core and are based on a water level that is at the top of active fuel. Acceptance criteria for the dose rates at the radiation detector locations on the refueling floor elevation of the reactor building are provided in Section 2.0 of this calculation.

The minimum acceptance criteria are that the dose rates at the radiation detector locations be greater than the lower dose rate range for the detectors.

The lower range of monitor A-2 is 1 mR/hr. The lower range of the other detectors is 0.1 mR/hr. The calculated dose rates at the detector locations are CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" S..rees:;._....._,,.

Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 RPV Water Level Rev. 2 JDate X J Safety Related I J Non-Safety Related Page 75 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date less than the acc*eptance criteria when the moisture separator and steam dryer are in the reactor vessel, and the RPV head and drywell head are in place. The calculated dose rates at the detector locations on the refueling floor elevation of the reactor building are less than 1 mR/hr when the moisture separator and steam dryer remain within the RPV, the water level is at the top of active fuel, and the RPV head and drywell heads are removed. None of the calculated dose rates are greater than the acceptance criterion dose rate, i.e., greater than the lower dose rate range for those detectors.

With the water level at the top of active fuel and the steam dryer, the RPV head, and drywell head removed (but the moisture separator remaining in the RPV), the calculated dose rates at the detector locations on the refueling floor elevation range from 22 to 88 mR/hr. With both the steam dryer and moisture separator removed from the RPV, the calculated dose rates to the detector locations range from approximately 3,000 mR/hr (3 R/hr) to 13,000 mR/hr (13 R/hr). These dose rates are due to radiation source terms in the active fuel region of the reactor core and represent the situation when there is no water in the reactor cavity between the top of active fuel and the refueling floor elevation.

Before the steam dryer or moisture separator are removed from the reactor vessel, the steam dryer and separator pool is filled _with water and so is the reactor cavity (above the top of the RPV). With water in the reactor cavity, the dose rates at the radiation detector locations on the refueling floor elevation, due to radiation shine from the reactor core, will be negligible.

The calculated dose rates at the detector locations are greater than the acceptance criteria (i.e., greater than the lower limit of the radiation detector range) when the RPV and drywell heads are removed and the steam dryer and/or moisture separator are removed from the reactor vessel and there is no water in the reactor cavity. Dose Rate Impact due to Sources Associated with Moisture Separator and Steam Drver The moisture separator and steam dryer will themselves be radioactive due to neutron activation and contamination.

Neutron activation is greater the closer the material being activated is to the reactor core. As such, dose rates associated with activation of the moisture separator and steam dryer are not homogenous over the components but are significantly larger towards the bottom of the components (i.e., the portion closest to the reactor core). The moisture separator and steam dryer are massive structures.

The upper portions of these structures can provide significant attenuation of the radiation associated with the most highly activated portions of these structures, i.e., the lower portions.

This calculation does not determine the dose rates at the CHRRM detector locations or at the locations of radiation monitors on the refueling floor elevation of the reactor building due to radioactivity associated with the moisture separator or steam dryer. The dose rates associated with the moisture separator or the steam dryer are orders of magnitude less that the dose rates due to the active core fuel region. Table 6-3 indicates that the calculated dose rate above the RPV flange area, due to the active core region, is thousands of R/hr when the moisture separator and steam dryer are removed from the RPV and the RPV water level is at the top of active fuel. When the steam dryer and moisture separator are within the RPV they attenuate much of the dose rate from the core and the dose rate above the RPV flange elevation is orders of magnitude less than the dose rate when the steam dryer and moisture separator are removed (and would CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 76 of 79 Client Project Proj. l)lo NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date remain orders of magnitude less, even after accounting for the dose rate contributions attributed to activity on the steam dryer and moisture separator).

Thus, the dose rates at the CHRRM locations, due to radioactivity on the steam dryer and moisture separator would be significantly less than the lower dose rate range of the CHRRM detectors.

CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" ..........

§.....-.-***

Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 77 of 79 Client NSP Prepared by Date Project MNGP Reviewed by Date Proj. No 12400-045 Equip. No. Approved by Date 8.0 References

8.1 Project

Task Report T0802, "Core Source Term," GE-NE-0000-0064-6767-TR-RO (ORF 0000-0064-6760), Prepared for Nuclear Management Company, LLC (NMC) Monticello Nuclear Generating Plant Extended Power Uprate, GE Energy Nuclear, Rev O 8.2 Design Information Transmittal (DIT), No. 918, Rev 1 (EC 13638),

Subject:

Design Inputs for Revision to S&L Calculation 2004-07061 Rev 1 (MNCP Calculation 04-202 Rev 1) to Update Calculation for EPU Source Term and New Steam Dryer, (Included as Attachment 3 to this Calculation)

8.3 Monticello

Nuclear Generating Plant USAR Section 3 Figures, Revision 26 8.4 Deleted 8.5 Monticello Nuclear Generating Plant Ops Manual B.01.01-06, Revision 12 (Included as Attachment 4 to this Calculation)

8.6 Monticello

Nuclear Generating Plant Ops Manual B.04.01-06, Revision 2 (Included as Attachment 5 to this Calculation)

8.7 Drawing

NF-36685 (E-333), "Reactor Building -Drywell Conduits and Trays, Above El933'," Rev G 8.8 Drawing 0360-2062, "Gamma Ionization Chamber Assy RD-23," General Atomic Co, Rev D. This Drawing is included in MNGP Technical Manual NX-19774, "High Range Gamma Radiation Monitoring System," Revision 1. 8.9 M. R. Sivack to Mr. James Brownell, "Monticello Steam Dryer Replacement Project, Transmittal of Monticello Replacement Steam Dryer Weight," Letter L TR-EP-10-106, dated 12/13/2010 (Included as Attachment 2 to this Calculation) 8.1 O Deleted. 8.11 Deleted. 8.12 Monticello Nuclear Generating Plant Drawings 8.12.1 Drawing NF-36651 (C-367), "Drywell Shield Cone. Sections -Sht. 6," Revision 3 8.12.2 Drawing NF-36169 (C-374), "Reactor Bldg. Drywell Vessel Fdn-Sht.1," Revision 2 8.12.3 Drawing NF-36171 (C-378), "Drywell Interior Access Platforms-Sht.1," Revision 76 8.12.4 Drawing NX-7831-197-1, "Reactor Vessel & Internals," Rev D 8.12.5 Drawing NF-36210 (C-63), "Radiation Shielding Requirements, Reactor Bldg. -Plans CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level" Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 Client Project Proj. No RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 78 of NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date Below El. 1027'-8 11 , Rev 2 8.12.6 Drawing NF-36204 (C-69), 11 Radiation Shielding Requirements, Reactor, Turb, & Radwaste Bldgs.-Sect.

A 11 , Rev 3 79 8.12. 7 Drawing NF-36058 (M-5), "Equipment Location -Reactor Bldg. Plan at El 1027'-8", 11 Rev c 8.12.8 Drawing NF-36062 (M-9), 11 Equipment Location -Reactor, Turb, & R. W. Bldgs, Section A-A," Rev C 8.12.9 Drawing NF-36405 (C-365), "Reactor Building Drywell Shield Cone. Sections -Sht.4," Rev4 8.12.1 O Drawing NF-36500 (C-354), "Reactor Building Dryer & Separator Pool Floor Plan and Sections," Rev A 8.12.11 Drawing NF-36392 (C-351 ), "Spent Fuel & Dryer Separator Pool Shield Plug Details," Rev A 8.12.12 Drawing NF-36358 (C-380), "Reactor Building Drywell Int-Key Plan & Sect," Rev 3 8.13 Deleted. 8.14 Deleted. 8.15 RUNT-PC, S&L Computer Program No. 03.7.705-1.1, "A Computer Program for Modeling Radionuclide Buildup and Decay and for General Purpose Isotope Shielding Analysis Using Gaussian Quadrature Integration of Point and Line Kernel Dose Rate Equations for Predefined and Ad Hoc Source and Shield Geometries," Revision 1 8.16 MCNP, S&L Computer Program No. 03.7.511-4.0C, "MCNP-4C3 Monte Carlo N-Particle Transport Code System," July 2002 8.17 The Photon Shielding Manual, Anthony Foderaro, Second Edition 8.18 Nuclides and Isotopes (Chart of the Nuclides), General Energy Group, Fourteenth Edition 8.19 Principles of Radiation Shielding, Arthur B. Chilton, et al., Prentice-Hall, Inc., 8.20 ANSl/ANS-6.4-1997, "American National Standard for Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants," Tables 5.2 (Typical Concrete Properties) 8.21 NEI 99-01, "Methodology for Development of Emergency Action Levels," Revision 4, January 2003 8.22 ORNL/TM-11018, "Standard-and Extended-Burnup PWR and BWR Reactor Models for the CA-04-202, "Dose Rates to CHRRM Detectors Due to Drop in RPV Water Level"

.. Cales. For Dose Rates to CHRRM Detectors Due to Drop in Cale No. 2004-07061 RPV Water Level Rev. 2 !Date X I Safety Related I I Non-Safety Related Page 79 of 79 Client Project Proj. No NSP Prepared by Date MNGP Reviewed by Date 12400-045 Equip. No. Approved by Date ORIGEN2 Computer Code," S. 8. Ludwig and J.P. Renier, Chemical Technology Division, December 1989 8.23 Flow of Fluids Through Valves. Fittings, and Pipe, Crane Technical Paper No. 410, 1981 8.24 Deleted 8.25 Principles of Radiation Protection.

K. Z. Morgan and J. E. Turner, Reprinted 1973 9.0 Attachments Attachment 1: Identification of RUNT-PC and MCNP Computer Code Input and Output Files Attachment 2: Letter L TR-EP-10-106 (Reference 8.9 to this Calculation)

Attachment 3: Design Inputs for Revision to S&L Calculation 2004-07061 Rev 1 (MNGP Calculation 04-202 Rev 1) to Update Calculation for EPU Source Term and New Steam Dryer Attachment 4: Table 2 of Monticello Nuclear Generating Plant Ops Manual 8.01.01-06, Revision 12 Attachment 5: Monticello Nuclear Generating Plant Ops Manual 8.04.01-06, Revision 2 Attachment 6: Radiation Monitor Locations on the Refueling Floor Elevation Attachment 7: RUNT-PC Output File (EPU Src 2004-07061 R2_0UTPUT.TXT)

Attachment 8: MCNP File (NWCHA) Attachment 9: MCNP File (WTAFC) Attachment 10: MCNP File (WTAFSAC)

Attachment 11: MCNP File (WT AFARM) Attachment 12: MCNP File (NOHDCHB)

Attachment 13: MCNP File (NOHDA8V)

Attachment 14: MCNP File (NOHARM3)

Attachment 15: MCNP File (NODRYCH)

Attachment 16: MCNP File (NODRYAR)

Attachment 17: MCNP File (NOMSA12)

Attachment 18: MCNP File (NOMSCHA)

Attachment 19: MCNP File (NOMSCC) Attachment 20: MCNP File (NOMSDH) Attachment 21: MCNP File (NOMSCHD) -Final-

ML17095A120

Text

SUMMARY

SUMMARY

Subject:

3.2 Reactor

8.2 which

8.2 which

6.9. Detector

3.8 Gamma

4.1 Refueling

4.3 Drywell

4.5 Drywell

4.7 Moisture

4.8 Concrete

5.2.1 Deleted

5.2.8 Radiation

8.1 Project

Subject:

8.3 Monticello

8.6 Monticello

8.7 Drawing

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