ML18128A054

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License Amendment Request Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 - Containment Doses for Eals, Calculation G13.18.9.4-045
ML18128A054
Person / Time
Site: River Bend Entergy icon.png
Issue date: 04/30/2018
From:
Entergy Operations
To:
Office of Nuclear Reactor Regulation
References
RBG-47847 G13.18.9.4-045
Download: ML18128A054 (45)
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ENCLOSURE 6 RBG-47847 Containment Doses for EALs, Calculation G13.18.9.4-045

0 AN0-1 0 AN0-2 0GGNS 0 IP-2 0 IP-3 0PLP DJAF 0PNPS ~RBS ow 0W3 0 NP-GGNS-3 0 NP-RBS-3 1 2 CALCULATION <> EC# 74027 < >Page 1 of 44 COVER PAGE c3i Design Basis Cale. DYES ~NO (4) ~ CALCULATION DEC Markup 10

\:i 1 Calculation No: G13.18.9.4-045 ' Revision: 1 (7) 10

Title:

Containment Doses for Emergency Action Levels (EALS) ' Editorial DYES ~NO (9)

System(s): 511 / RMS \lUJ Review Org (Department): BE3 11 12

< > Safety Class: < > Component/Equipment/Structure Type/Number:

~ Safety I Quality Related RMS-RE16A RMS-RE16B D Augmented Quality Program D Non-Safety Related RMS-RE20A RMS-RE20B 13

< > Document Type: F43.02 14

< > Keywords (Description/Topical Codes):

Emergency Plan, Emergency Action Levels, EAL REVIEWS 15 15 17

< > Name/Signature/Date < > Name/Signature/Date < J. Name/Signature/Date Greg Broadbent/ See AS I See AS I See AS I See Melissa Litherland / See AS/ See AS AS See AS Responsible Engineer ~ Design Verifier i Supervisor/Approval D Reviewer D Comments Attached D Comments Attached

\

CALCULATION NO: G 13.18.9.4-045 CALCULATION REVISION: 1 REFERENCE SHEET PAGE: 2 OF 44 I. EC Markups Incorporated (N/A to NP calculations) 813.18.9.4-045 EC 16369 II. Relationships: Sht Rev Input Output Impact Tracking No.

Doc Doc Y/N

1. G13.18.9.5*061 N/A 002 0 D N
2. G13.18.9.4-049 N/A 000 D 0 y EC 74027
3. G13.18.9.5*065 N/A 001 0 D N
4. GE-22A3130AW N/A *004 0 D N
5. PID-25-01 B N/A 007 0 D N
6. 0221.110-000-235 - N/A 300 0 D N
7. 0221 .110-000-236 N/A 300 0 D N
8. 0221.110-000-232 N/A 300 0 D N
9. 0221.120-000-070 N/A 300 0 D N
10. 0221.120-000-011 N/A 300 0 D N 11 . 0221 .120-000-021 N/A 300 0 D "N
12. 0221.180-000-076 N/A 300 0 D N*
13. 7222.251-000-001A N/A 300. 0 D N
14. 6244.400-912-001 B N/A 300 0 D N
15. RBS-SA-08-00002 N/A 000 0 D N
16. PN-228 N/A 003 0 D N
17. PN-330 N/A 001 0 D N
18. G13.18.9.4-051 N/A 000 D 0 y EC 74027
19. IA-RMS*1 N/A 005 0 D N
20. PR-C-358 N/A 001 0 D N
21. EV-001A N/A 010 0 D N 22.' EK-015A N/A 003 0 D N
23. EK-015V N/A 004 0 D N
24. ES-059A N/A 006 0 D N Ill. CROSS

REFERENCES:

1. NEI 99-01, Rev. 6, Development of Emergency Action Levels for Non-Passive Reactors, November 2012
2. RBS Technical Specifications '
3. Crane Technical Paper 410, 25th Printing, 1991
4. Handbook of Tables for Applied Engineering Science, 2nd Edition
5. Radiation Detection and Measurement, Glenn F. Knoll, 1979
6. Perry's Chemical Engineer's Handbook, 50th Edition, 1984
7. Federal Guidance Report No. 12 (FGR-12) , EPA-402-R-93-081, September 1993
8. Radiological Health Handbook
9. Regulatory Guide 1.3 Rev. 2, 1974 Assumptions used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors

CALCULATION NO: G13.18.9.4-045 CALCULATION REVISION: 1 REFERENCE SHEET PAGE: 3 OF 44

10. Regulatory Guide 1.195, May 2003, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors
11. RBC-49932, RBS Issuance of Amern;:lment re: 1.7 Percent Increase in Licensed Power Level (TAC NO. MB5094) / Amendment No. 129 TO NFP-47 / Safety Evaluation/ RBF1-03-0022 IV. SOFTWARE USED:

Title:

Microsoft Excel Version/Release: 2010 Disk/CD No. N/A

Title:

Version/Release: Disk/CD No. -

V. DISK/CDS INCLUDED:

Title:

None Version/Release Disk/CD No.

VI. OTHER CHANGES:

Supersede Calculation G13.18.9.4-049 Rev. 000 Supersede Calculation G13.18.9.4-051 Rev. 000

CALCULATION NO: G13.18.9.4-045 Record of Revision REVISION: 1 PAGE: 40F 44 Revision Record of Revision 0 Initial issue.

This calculation is revised to reflect the current EN-DC-126 calculation format, utilize the current requirements of NEI 99-01 Rev. 6 EAL recommendations and basis, include the 24 month cycle source term, and account for the instrument accuracy of the containment and drywell radiation monitors. The extended duration dose rate reading determination portion has been deleted eliminating the need for use of the RADTRAD program.

This calculation also incorporates and updates the information previously 1

contained in calculation G13.18.9.4-049, Rev. 0, Drywell Doses for Emergency Action Levels (EALs). As a result, G13.18.9.4-049 will be superseded.

This calculation also addresses and revises the information previously contained in calculation G13.18.9.4-Q51, Rev. 0, Drywell Doses for Emergency Action Levels (EAL-RC4) and no longer utilizes Grand Gulf information. As a result, G13.18.9.4-051 will be superseded.

TABLE OF CONTENTS CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 5 OF44 Section Description Page 1.0 CALCULATION COVER PAGE 1 2.0 CALCULATION REFERENCE SHEET 2 3.0 RECORD OF REVISION 4 4.0 TABLE OF CONTENTS 5 5.0 PURPOSE 6

6.0 CONCLUSION

7 7.0 INPUTS AND DESIGN CRITERIA 7 8.0 ASSUMPTIONS 13 9.0 METHOD OF ANALYSIS 15 10.0 CALCULATIONS 25 Attachments None

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 6 OF44 5.0 PURPOSE The purpose of this calculation is to determine the dose rate indicated by the ORMS Containment and Drywell Post Accident Monitors (PAMs) in the primary containment (RMS-RE16A, B) and drywell (RMS-RE20A, B) given the following scenarios:

I

1. An instantaneous release of all reactor coolant mass into primary containment, assuming the reactor coolant activity equals Technical Specification Allowable Limits. This should be determined using the reactor coolant noble gas and iodine inventory with RCS activity at Technical Specification allowable limits released inst~ntaneously into the primary containment atmosphere. This corresponds to NEI 99-01 (Ref. 111.1 page 90), RCS Barrier Threshold, Primary Containment Radiation, Loss 4.A.
2. An instantaneous release of all reactor coolant mass into the primary containment assuming that reactor coolant activity equals 300 uCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. This corresponds to NEI 99-01 (Ref. 111.1, page 86), Clad Barrier Threshold, Primary Containment Radiation, Loss 4.A.
3. An instantaneous release of all reactor coolant mass into the primary containment assumihg that 20% of the fuel cladding has failed. This corresponds to NEI 99-01 (Ref.

111.1, page 95, 96), Primary Containment Barrier Threshold, Primary Containment Radiation, Loss 4.A.

This is being done to support the revision of the Emergency Plan updating the Emergency Action Levels (EALs) to meet the guidelines of NEI 99-01 Rev. 6 (Ref. 111.1 ).

r

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 7 OF 44

6.0 CONCLUSION

The analysis results, including consideration of instrument uncertainty, is provided in Section 10, pages 43 and 44. The analysis and associated discussion provides the basis for the following recommended dose rate threshold values for the Containment and Drywell Post Accident radiation monitors RMS~

RE16A, Band RMS.-RE20A, B for the three scenarios listed in Section 5.

Containment Drywell RMS-RE 16A, B RMS-RE20A, B R/hr R/hr

.:::59 .:::38 Case 1, 4 uCi/g Dose Equivalent 1-131

(.::: 5.90E+01) (.::: 3.80E+01)

> 3000 > 2000 Case 2, 5% Clad Damage

(.::: 3.00E+03) (.::: 2.00E+03)

.::: 12,000 .::: 8,000 Case 3, 20% Clad Damage

(.::: 1.20E+04) (.::: 8.00E+03) 7.0 INPUT AND DESIGN CRITERIA Definitions AST- Alternative (or Alternate) Source Term (synonymous with "Revised Source Term" /

or RST)

DBA- Design Basis Accident EAL- Emergency Action Level EIV- Early In-Vessel Phase of AST GAP- Gap Phase of AST LOCA- Loss of Coolant Accident RCPB- Reactor Coolant Pressure Boundary RG- Regulatory Guide ST- Source Term

CALCULATION No.: G13.18.9.4-045 REVISION: .1 PAGE 8 OF44 General Inputs Drywell Free Volume = 236,136 ft 3 - 11.811 ft 3 - 71.6 ft 3 - 513 ft 3

= 235,539.589 ft 3 use 235,540 ft 3 (Ref. 11.16 and EC Markups EC 5000047947, EC 108, EC 23493, EC 25424)

Containment Free Volume = 1,191,590 ft 3 - 15.27 ft 3 - 0.74 ft 3 1,191,573.99 ft 3 use 1,191,574 ft 3 (Ref. 11.17 and EC Markups EC 5000048884, EC 5000050413, EC 23493, EC 62305)

RMS-RE16A Detector Location = RB Floor Elevation 186' 3" (Ref. 11.23)

Detector Centerline Elevation 194' 3" Azimuth 315 RMS-RE168 Detector Location = RB Floor Elevation 186' 3" (Ref. 11.23)

Detector Centerline Elevation 194' 3" Azimuth 45 RMS-RE20A Detector Location= OW Floor Elevation 114' O" (Ref. 11.22)

Detector Centerline Elevation 125' 3" Azimuth 163 RMS-RE20B Detector Location = DW Floor Elevation 114' O" (Ref. 11.22)

Detector Centerline Elevation 143' 4" Azimuth 201 ° 55' Containment Inner Diameter= 120' (Ref. 11.21)

Containment Monitor Floor Elevation = 186' 3" (Ref. 11.23)

Elevation of Containment Dome Bend Line= 226' 3" (Ref. 11.21)

Containment Gross Volume Above Dome Bend Line= 226,195 ft 3 (Ref. 11.17 pg. 10)

Equipment Volume 186'-3" Elevation= 4984.7 ft 3 (Ref. 11.17 pg. 16)

Drywell Inner Diameter = 68' 11.25" (Ref. 11.21)

Reactor Shield Wall Outer Radius = 14' 11" (Ref. 11.24)

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 9 OF 44 Rated Thermal Power= 3100 MWt This is consistent with the rated thermal power of 3091 MWt in the RBS operating license (Reference 111.2 Operating License) including a power level uncertainty of 0.3% which was approved by the NRG in the SER for the Thermal Power Optimization project (Reference 111.11).

Case 1 Coolant Mass (Steam and Water) Determination Inputs Maximum reactor coolant weight at normal water level (includes vessel and recirculation loops)@ 75 deg. F = 685,000 lb 1 (Ref. 11.4, pg. 3)

Normal water level (nominal instrument zero)=

520.62" above vessel "O" (Ref. 11.5)

Head Flange level= 842" above vessel "O" (Ref. 11.5)

Nominal Vessel Radius= 109" (Ref. 11.6 and II. 7)

Vessel Head height above flange surface= 109" (Ref. 11.8)

/

Weight of Shroud Head and Separators= 88,000 lbs (Ref. 11.9)

Shroud Head and Separator Material= Austenitic Stainless Steel . (Ref. 11.9)

Weight of Steam Dryer= 64,000 lbs (Ref. 11.10)

Steam Dryer Material= 304 Series Stainless Steel (Ref. 11.11) p Water at 70°F = 62.305 lb/ft3 (Ref. 111.3 pg. A6)

Nominal Specific Gravity (s.g.) 304 Stainless Steel= 8.02 (Ref. 111.4, pg. 118)

Main Steam Line Volume to Outboard MSIV including RCIC =1221 ft 3 (Ref. 11.12)

Reactor Operating Pressure= 1070 psia (Ref. 11.13)

Specific Volume of Steam at 1050 psia = 0.42224 ft 3/lb (Ref. 111.3 pg. A-15)

Specific Volume of Steam at 1100 psia = 0.40058 ft 3/lb (Ref. 111.3 pg. A-15)

I CALCULATION No.: G13.18.9.4-045 REVISION: 1' PAGE 10 OF 44 Case 1 Source Term Inputs Steady State Dose Equivalent 1-131 Limit = ~ 0.2 uCi/gm (Ref. 111.2, TS 3.4.8)

Action Level Limit Dose Equivalent 1-131 = ~ 4 uCi/gm (Ref. 111.2, TS 3.4.8)

Steady State Noble Gas Gross Gamma Activity Limit

~ 290 mCi/s after decay of 30 minutes (Ref. 111.2, TS 3.7.4)

Reactor Coolant System Isotopic Concentrations (Ref. II .20 Image pg. 11, 12)

Table 7.1 AST Isotope Reactor Reactor Steam Group Coolant Concentration Concentration uCi/gm uCi/gm (release rate 303.9 mCi/s) 1 Kr-85 3.00E-05 1 Kr-85m 9.?0E-03 1 Kr-87 3.30E-02 1 Kr-88 3.30E-02 1 Xe-133 1.30E-02 1 Xe-135 3.60E-02 2 1-131 2.10E-02 3.40E-04 2 1-132 3.1 OE-01 ,. 4.?0E-03 2 1-133 2.80E-01 4.40E-03 2 1-134 4.90E-01 9.?0E-03 2 1-135 2.?0E-01 4.30E-03 3 Cs-134 1.?0E-04 1.?0E-07

.. 3 Cs-136 1.1QE-04 1.1 OE-07 3 Cs-137 4.40E-04 4.40E-07

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 11 OF 44 1 Bq = 2. 703E-11 Ci (Ref. 111.5 pg. 4) 1 Sv = 100 Rem (Ref. 111.5 pg. 78)

Iodine dose conversion factor, Sv/Bq from NUREG/CR-6604, Table 1.4.3.3-2 contained in Ref.

11.14, image- page 384 and 385, Thyroid line, Inhaled Chronic column for each isotope below.

Table 7.2 Isotope Dose Conversion Factor, Sv/Bq 1-131 2.92E-07 1-132 1.74E-09 1-133 4.86E-08 1-134 2.88E-10 1-135 8.46E-09 Cs-134 1.11E-08 Cs-136 1.73E-09 Cs-137 7.93E-09 Case 2 and 3 Source Term Release Fractions Table 7.3 AST Radionuclide Groups and BWR Gap Release Fractions Group* Title Elements Gap Release 1 Noble Gases Xe, Kr 0.05 2 Halogens I, Br 0.05 3 Alkali Metals Cs,Rb 0.05 4 Tellurium Group Te, Sb, Se 0 5 Barium, Strontium Ba,Sr 0 7 Noble Metals Ru,Rh,Pd,Mo,Tc,Co 0 8 Cerium Group Ce,Pu,Np 0 9 Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, 0 Y, Cm,Am

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 12 OF 44 Nuclide Inventory Table 7.4 BWR Core Inventories .

1-ialf-life Gamma Energy, 24 Month Cycle AST (sec) MeV Concentration, Isotope Group (Ref. 11.14, (Ref. Ill. 7, Table A.1, Ci/MWt pg. 81-82) pg. 200-211) (Ref. 11.15) ,

1 Kr-85 3.38E+08 2.00E-03 3.66E+02 1 Kr-85m 1.61E+04 1.58E-01 7.02E+03 1 Kr-87 4.58E+03 7.93E-01 1.35E+04 1 Kr-88 1 '.02E+04 1.96E+OO 1.89E+04 1 Xe-133 4.53E+05 4.60E-02 5.26E+04 1 Xe-135 3.27E+04 2.49E-01 1.99E+04 2 1-131 6.95E+05 3.82E-01 2.70E+04 2 1-132 8.21 E+03 2.28E+OO 3.92E+04 2 1-133 7.49E+04 6.07E-01 5.52E+04 2 1-134 3.16E+03 2.63E+OO 6.06E+04 2 1-135 2.38E+04 1.58E+OO 5.17E+04 3 Rb-86 1.61E+06 ' 9.50E-02 6.31 E+01 3 Cs-134 6.51 E+07 1.55E+OO 6.11E+03 3 Cs-136 1.13E+06 2.17E+OO 2.00E+03 3 Cs-137 9.51E+08 6.62E-01 3.95E+03

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 13 OF 44 8.0 ASSUMPTIONS Case 1 8.1 Case 1 requires that all reactor coolant mass be released into the primary containment. For this analysis, reactor coolant volume is assumed to be composed of the volume of water in the reactor vessel and recirculatjon loops and the volume of steam out to the outboard MSIV's including the RCIC steam supply line to the outer isolation valve. For simplicity, vessel instrument zero (520.62 inches Above Vessel O cJ'r AVO) is assumed to be the dividing line between the vessel water and steam volumes. The steam separator and steam dryer are assumed to completely reside in the steam volume portion of the reactor.

It should be noted that the mass of water and steam used in this analysis are significantly greater than that used for design basis events, such as the Main Steam Line Break analyzed in G13.18.9.5*065. In design-basis events, the core shroud maintains a floodable volume to the level of the top of the jet pumps and only one of two recirculation loops (in a recirculation line double ended rupture) would lose volume. This analysis interprets "all" to imply 100% of the water and steam in the reactor coolant boundary as opposed to "all postulated water and steam released".

8.2 The steam volume is assumed to contain saturated steam at a normal reactor operating pressure of 1070 psia. Assuming normal operating pressure and saturated conditions maximizes the mass per cubic foot of steam and corresponding source term.

8.3 The vessel steam volume will be determined by treating the reactor vessel as a right circular cylinder with a height from vessel instrument zero level to the top head flange level. The volume of the separator ~md dryer will be subtracted from this volume.

8.4 The reactor vessel top head i& treated as a hemisphere with a radius equal to the vessel radius and height equal to the vessel radius.

8.5 Case 1 requires the coolant iodine concentration and noble gas concentration to be at the Technical Specification limit. The coolant iodine concentration will have two cases, one* at the TS Steady State Limit of 0:2 uCi/g dose equivalent 1-131 and the second at 0.4 uCi/g dose equivalent 1-131. The design basis liquid source term isotopes from Ref. 11.20 will be normalized to the dose equivalent concentration. For both dose equivalent iodine cases, the design basis noble gas source term will be used. The design basis noble gas concentration is based upon a release rate of approximately 304 mCi/sec with 30 minutes d~cay which exceeds the TS limit of a release rate of~ 290 mCi/sec after 30 minutes decay.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 14 OF 44 8.6 The activity concentrations for the r~actor coolant and main steam are based upon Normal Water Chemistry concentrations rather than Hydrogen Water Chemistry concentrations. This is due to the implementation of Noble Metals chemistry and the substantial reduction in hydrogen injection rates from the historical full power HWC application rate of 77 scfm hydrogen to the current application rate of approximately 10.5 scfm hydrogen a The current addition rate more closely reflects Normal Water Chemistry conditions than HWC conditions. Other conservatisms present in the calculation more than accommodate any non-conservatisms resulting from the slight changes in steam line iodine concentration due to HWC.

Case 2 and 3

8. 7 As Case 2 and 3 scenarios involve release of gap activity, the source term will be calculated based upon 3100 MWt consistent with the LOCA and Main* Steam Line Break analysis'. Refer to input item for rated thermal power on page 9.

All Cases 8.8 The radiation monitors can only detect the portion of the release that is within their "line of*

sight". Therefore the containment volume used in the geometry factor will be limited to the containment volume above the floor elevation where the detectors are located (elevation 186' 3"). The containment 186' 3" floor elevation is the refueling floor and has thick concrete floors*

surrounding the upper pool area. The shielding provided by the pools and the surrounding concrete pool support structure and area concrete floors reduce the dose contribution from the elevations below to insignificant levels. Similarly, geometry factor volume for the drywell will reduce the drywell volume to account for the area shielded by the reactor vessel and primary shield wall (bioshield).

8.9 The release is instantaneous and uniformly mixed in the containment and drywell. A flashing fraction of 100% is assumed for the isotopes in the liquid phase (i.e. 100% of the halogen and cesium activity present in the coolant is released).

8.10 There is no dilution, plate out or scrubbing of the release.

8.11 Only the AST fuel gap inventory isotopes are considered.

8.12 The chemical form of the release is 95% cesium iodine (Csl) (which is a particulate or aerosol), 4.85% elemental iodine, and 0.15% organic iodine 8.13 Containment leakage is O scfm

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 15 OF 44 9.0 METHOD OF ANALYSIS Case 1 An instantaneous release of all reactor coolant mass into primary containment, assuming the reactor coolant activity equals Technical Specification Allowable Limits. This should be determined using the reactor coolant noble gas and iodine inventory with RCS activity at Technical Specification allowable limits released instantaneously into the primary containment atmosphere. This corresponds to NEI 99-01 (Ref. 111.1 page 90), RCS Barrier Threshold, Primary Containment Radiation, Loss 4.A.

The reactor coolant mass is defined in Assumption 8.1.

The mass of water is obtained from Reference II .4.

The total volume of steam, Vst is calculated with the following formula:

Vst ft 3= Vrs ft 3 - Vsd ft 3 + Vh ft 3 + Vms ft 3 Where:

Vrs = Volume of the vessel between instrument zero and the head flange, ft 3 Vrs, ft 3 = n * (vessel radius, in)2 * (Head flange level, in AVO - Normal water level, in AVO) /

1728 in 3/ft 3 \

Vsd = Volume displaced .bY the steam separator and steam dryer, ft 3 Vsd ft 3 = (Separator Weight, lb+ Dryer Weight, lb)/ (p water lb/ft3

  • s.g. 304SS)

Vh = Volume of the*vessel head, ft 3 Vh ft 3 = 2/3

  • n
  • r2 *h
  • 1728 in 3/ft3 (Ref. 111.6 pg. 2-12)

Where: r = head radius= vessel radius, in h = head height, in Vms = Volume of the main steam lines to the outboard MSIV's, ft 3 , from Reference 11.12.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 16 OF 44 The mass of steam, Ms, is determined based.on the total volume and steam specific volume at the conditions of Assumption 8.2.

Ms, lb = Vst ft 3 / SV1070, ft3/lb Mw, lb = Maximum reactor coolant weight at normal water level The source term is based upon the source term for normal operating conditions from PR-C-358 (Ref. 11.20). The dose equivalent 1-131 concentration is calGulated based upon the activity concentration in the reactor coolant (water) only. The coolant activity released will be determined for both the normal equilibrium Dose Equivalent 1-131 concentration of 0.2 uCi/gm and maximum transient limit of 4 uCi/gm. As the cesium is released primarily in the form of Csl, the cesium species present in'the reacto coolant will also be included in the normalization process.

The activity released in the steam phase, including iodine and cesium isotopes will be determined in the same manner as the noble gases, i.e. concentration of each isotope multiplied by the mass of steam released. The activity in the steam phase is equal to the activity of each isotope in the vessel meeting or exceeding the value correspondi.ng to the TS release rate of,:::

290 mCi/sec or The detector reading will be based upon the sum of the coolant and steam activity releases of each isotope.

The equations used are generally the same as those in Ref. 11.3, however some are in a slightly different format with differing variable names for clarity purposes.

Dose Equivalent RCS Iodine Activity Concentration Ci= CDi (L

  • DCF 1-131 ) / I/ (CDi
  • DCFi) (Ref. 11.3 pg. 17)

Where: Ci= reactor coolant activity concentration for isotope I for Technical Specification reactor coolant dose equivalent 1-131 activity limit L, uCi/gm.

L = Technical Specification dose equivalent 1-131 reactor coolant activity limit

= 0.2 uCi/gm dose equivalent 1-131 steady state limit

= 4.0 uCi/gm dose equivalent 1-131 action level limit CDi = design basis reactor coolant activity c,pncentration for isotope i from Table 7.1, uCi/gm , *,

DCF 1_131 = thyroid inhalation dose .conversion factor for 1-131, from Ta.ble 7.2, converted from units of Sv/Bq to Rem/Ci

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 17 OF 44 DCFi = thyroid inhalation dose conversion factor for isotope i, from Table 7.2, converted from units of Sv/Bq to Rem/Ci Equivalent Iodine Activity Released in Water Ci* CF= FCi Where: Ci= reactor coolant activity concentration for Iodine isotope I for Technical Specification reactor coolant dose equivalent 1-131 activity limit L, uCi/gm.

CF = Conversion Factor, 453.6 g/lb

  • Mw Where: Awi = reactor coolant water activity released for isotope i, Ci FCi = Reactor Coolant Iodine Concentration, Ci/lb Mw = Mass of Reactor Water released, lbm Activity Released in Steam Ci *CF= FCi Where: Ci= reactor steam activity concentration for isotope i ~ the TS limit of 290 uCi/sec with 30 minutes decay, uCi/gm.

CF = Conversion Factor, 453.6 g/lb

  • 1E-6 Ci/uCi = 4.54E-4 g*Ci / lbmuCi

(

FCi = Reactor Steam isotope i Concentration, Ci/lb

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 18 OF 44 Asi = FCi

  • Ms Where: Asi = reactor coolant steam activity releas'3d for Iodine isotope i, Ci FCi = Reactor Steam Iodine Concentration, Ci/lb Ms= Mass of Steam released, lbm I

Total Iodine Activity Released Ai =Awi +Asi Where: Ai= total activity released for Iodine isotope i, q Awi = reactor coolant water activity released for Iodine isotope i, Ci Asi = reactor coolant steam activity released for Iodine isotope i, Ci The total isotopic activity determined for 0.2 uci/gm Dose Equivalent 1131 and 4 uCi/gm Dose Equivalent 1131 is assumed to be released instantaneously into the combined containment and drywell free volume. The. combined containment and drywell volume will be used to determine the concentration of each isotope present in each area.

The formula used to manually convert the curies of each isotope released to the combined containment and drywell volume is from Reg. Guide 1.3:

Dose, y, Rads/sec= Z:i (0.25* Eavei

  • Xi) (Ref. 111.9 pg. 1.3-2)

I Where: Dose, y, = gamma dose, Rads/sec Eave = average gamma energy for isotope i , MeV Xi = concentration in Ci/m 3 for isotope i Where: Ci = isotopic concentration for isotope i, Curies Vtot = Containment/Drywell volume m 3

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 19 OF 44 Because the radioactive material released is assumed to be distributed throughout the combined containment and drywell volume, an adjustment factor is required to distribute the dose rate for each isotope to the individual containment and drywell area volumes. In addition, only a portion of the containment and drywell volumes in the vicinity of each detector will contribute to the detector reading. To adjust the total volume to the volume of the individual containment and drywell spaces and account for the detector location, a Geometry Factor is used. This converts the infinite cloud dose to a volume specific cloud dose.

0 338 GF = 352 / V * (111.10, pg. 11)

Where: GF is Geometry Factor, unitless Vis the volume of the source cloud, m 3 The containment volume used in this factor will be the volume only in the vicinity of RMS-RE16A and B, which is the free volume of containment above the 186'3" floor elevation. This includes the cylindrical volume between the elevation 186'3" floor and the elevation of the dome bend line at 226' 3", and the volume of the dome area above the bend line at elevation 226' 3". The free air portion is determined by subtracting the volume of equipment on or above 186'3" from the total volume above the, 186'3" floor. Equipment volume is obtained from Ref. 11.17.

Volume of Containment above 186'3", ft 3 = Cylindrical Volume between 186'3" and dome bend line elevation, ft 3 + containment dome volume, ft 3 Cylindrical Volume between 186'3" and dome bend line elevation = n

  • r2
  • h Where: r = inner radius of containment, ft(= inner diameter/ 2, ft) h = height between 186'3" and dome bend line elevation, ft Ve, Geometry Factor Volume of Containment above 186'3, ft 3 =

Volume of Containment above 186'3", ft 3 - Equipment Volume above 186',3, ft 3 Ve, m 3 = Geometry Factor Volume of Containment above 186'3, ft 3

  • 2.832E-2 m 3/ft 3 Tbe drywell geometry factor volume will be determined to only include the volume in the line of sight for RMS-RE20A, B. The detectors in the drywell cannot "see" the portion shielded by the reactor shield wall area. As the reactor and reactor shield wall is a vertical cylinder within the cylindrical drywell, the volume of the drywell hidden can be calculated by determining the hidden and visible detector areas. The hidden drywell volume is determined by multiplying the full drywell open volume by the ratio of hidden area to seen area.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 20 OF 44 In order to determine the area the drywell detectors can see, lines (chords) are drawn from the detector location tangent to the reactor shield wall on each side. (Refer to the figure below.)

The tangent points are on either side of the reactor shield wall diameter. These chords enclose areas on either side of the reactor shield wall that can be seen by the detector (Areas C and D).

The height of the chord, h, is equal to the drywell radius minus the reactor shield wall radius.

Area C is equal to Area D.

Area A and B enclose the triangular area directly in front of the detector extending to the centerline of the reactor shield wall. Triangle AB has a height equal to the drywell radius and a base equal to the reactor shield wall diameter. Area B is half of the area encircled by the reactor shield wall. Area A is determined by subtracting area B from the area of Triangle AB.

The total area that can be seen by the detector= Area A+ Area C + Area D

CALCULATION No.: G13.18.9.4-045 REVISION:'-- 1 PAGE 21 OF 44 rsw, ft = Reactor shield wall outer radius, ft 2

Asw, ft = Reactor Shield Wall Area= TC* rsw2 ddw, ft= Drywell Inner Diameter, ft rdw, ft= Drywell Radius, ft= Drywell Inner Diameter, ft/ 2 Adw = Drywell Area = TC

  • rdw2 t, ft= rdw, ft - rsw, ft Area of Triangle AB, ft 2 = %
  • b
  • h Where b = reactor shield wall inner diameter, ft h = radius of the drywell, ft Area of 8, ft 2 = % *TC* rsw2 Area of A, ft 2 = Area of Triangle AB - Area of B 8, radians = 2 cos-1 ( rsw / rdw) (Ref. 111.6 pg. 2-11)

Area of Sector C or D, ft 2 = %

  • rdw2 , ft2
  • 8, radians (Ref. 111.6 pg. 2-11)

As, ft2, area seen by the detector= Area A+ Area C + Area D Def, Drywell Volume Correction Factor= As/ Adw Vo, Geometry Factor Drywell Volume, m3 = Drywell Free Volume, m3

  • Def 0 338 For Containment GFc = 352 / Ve
  • Where: Ve is the Geometry Factor volume of containment, m 3 0 338 For Drywell GFo = 352 / Vo
  • Where: Vo is the Geometry Factor volume of the drywell, m 3

CALCULATION No.: G13.18.9.4-045 REVISION: 1.

PAGE 22 OF 44 The units of dose used are R/hr rather than R/sec. Incorporating these factors into the previous Gamma Dose formula provides the following formula for the gamma dose in containment:

0 338 Dose, y, Containment, R/hr = Li (0.25* Eavei

  • Ci iVtat) *3600 sec/hr/ (352 / Ve * )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies Vtot = Containment/Drywell volume m 3 Ve is the Geometry Factor volume of containment, m 3 The total gamma dose in the drywell is similar:

0 338 Dose, y, Drywell, R/hr = Li (0.25* Eavei

  • Ci / Vtot) *3600 sec/hr/ (352 / Vo * )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies

  • Vtot = Containment/Drywell volume m3 V 0 is the Geometry Factor volume of the drywell, m3 I

The total gamma dose for both containment and drywell is determined for both 0.2 uCi/gm and 4 uCi/gm .Dose Equivalent 1131.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 23 OF 44 Case 2 ~'

An instantaneous release of all reactor coolant mass into the primary containment assuming that reactor coolant activity equals 300 uCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. This corresponds to NEI 99-01 (Ref. 111.1, page 86), Clad Barrier Threshold, Primary Containment Radiation, Loss 4.A.

River Bend is currently an AST plant, therefore, the AST source term will be used in this calculation. The EAL requires doses for 5% clad damage (i.e., no fuel melt assumed). The AST gap activity will be used in this evaluation. This phase begins when the fuel cladding begins to fail. During this phase the fission gases contained in the plenum and between the fuel pellet and cladding are released. Typically, these gases include the noble gases, the halogens, and the alkali metals. The bulk of the fission products are retained in the fuel pellets. Table 7.3 contains the NU REG 1465 release groups and the gap release fractions for these groups.

GE provided a generic core source term as part of the transition to 24 Month Cycles (Ref. 11.15).

The source term for nuclides in the AST groups 1, 2, and 3 are listed in Table 7.4 The curies of each isotope for a complete gap release of the noble gases, the halogens, and the alkali metals is determined as follows:

Gap Releasei, Ci = Ci/MWt i

  • MWt
  • Gap Release Fractioni The damage fraction (DF 5 ) associated with 5% clad damage is 5% of the complete gap release.

Damage Releasei, Ci= Gap Releasei

  • DFs By definition DF 5 = 0.05 As with Case 1, 100% of the activity is assumed to be instantaneously released to the combined containment and drywell volume. The containment and drywell dose associated with each isotope will be calculated using the RG 1.3 methodology including conversions*and geometry adjustment factors from Case 1. The formulas for the dose in containment and drywell are reproduced below. *
  • 0 338 Dose, y, Containment, R/hr = ~i (0.25* Eavei
  • Ci / Vtot) *3600 sec/hr I (352 I Ve * )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies Vtot = Containment/Drywell volume m 3 Ve is the volume of containment, m3

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 24 OF 44 Dose, y, Drywell, R/hr = ~i (0.25* Eavei

  • Ci / Vtot) *3600 sec/hr/ (352 / Vo 0*338 )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies Vtot = ContainmenUDrywell volume m3 Vo is the volume of the drywell, m 3 Case 3 An instantaneous release of all reactor coolant mass into the primary containment assuming that 20% of the fuel cladding has failed. This corresponds to NEI 99-01 (Ref. 111.1, page 95, 96),

Primary Containment Barrier Threshold, Primary Containment Radiation, Loss 4.A.

As with case 2, this case involves the release of gap activity associated with clad damage. The formula's for analysis of this case are identical to case 2 with the exception of the damage fraction and damage release.

The damage fraction (DF 20 ) associated with 20% clad damage is 20% of the complete gap release.

Damage Releasei, Ci = Gap Releasei

  • Df 20 By definition DF20 = 0.20 Instrument Accuracy Per Ref. 11.19 pg. 28 and 42, the overall accuracy of RMS-RE16A, B and RMS-RE20A, B is

+100%, - 50% of reading. All other allowances are included in the overall accuracy.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 25 OF 44 10.0 CALCULATIONS The total volume of steam, Vst is calculated with the following formula:

Vst ft 3= Vrs ft 3 - Vsd ft 3 + Vh ft3 + Vms ft 3 Where:

Vrs = Volume of the vessel between instrument zero and the head flange, ft 3 Vrs, ft 3 = n * (vessel radius, in)2 * (Head flange level, in AVO - Normal water level, in AVO) /

1728 in 3/ft 3 Nominal Vessel Radius = 109" (Ref. 11.6 and 11. 7)

Head Flange level = 842" above vessel "O" (Ref. 11.5)

Normal water level (nominal instrument zero) =

520.62" above vessel "O" (Ref. 11.5)

Vrs, ft3 = (n * (109")2 * (842" - 520.62")) / 1728 in 3/ft 3 = 6941.9 ft3 Vsd = Volume displaced by !he steam separator and steam dryer, ft 3 Vsd ft 3 = (Sep'arator Weight, lb+ Dryer Weight, lb)/ (p water lb/ft3

  • s.g. 304SS)

I Weight of Shroud Head and Separators = 88,000 lbs (Ref. 11.9)

Weight of Steam Dryer= 64,000 lbs (Ref. 11.10) p Water at 70°F = 62.305 lb/ft 3 (Ref. 111.3 pg. A6)

Nominal Specific Gravity (s.g.) 304 Stainless Steel= 8.02 (Ref.111.4, pg. 118)

Vsd ft 3 = (88,000 lb+ 64,000 lb)/ (62.305 lb/ft3

  • 8.02) = 304.2 ft 3

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 26 OF 44

)

3 Vh = Volume of the vessel head, ft Vh ft 3 = 2/3

  • n
  • r2 *h
  • 1728 in 3/ft3 (Ref. 111.6 pg. 2-12)

Where: r = head radius = vessel radius, in h = head height, in Nominal Vessel Radius = 109" (Ref. 11.6 and 11.7)

Vessel Head height above flange surface = 109" (Ref. 11.8)

Vh ft 3 = ( 2/3

  • n * (109")2
  • 109") / 1728 in 3/ft 3 = 1569.6 ft3 Vms = Volume of the main steam lines to the outboard MSIV's, ft 3 , from Reference 11.12.

Main Steam Line_ Volume to Outboard MSIV including RCIC = 1221 ft 3 (Ref. 11.12)

Vms = 1221 ft3 Vst ft 3 = Vrs ft 3 - Vsd ft 3 + Vh ft 3 + Vms ft 3 Vst ft 3= 6941.9 ft 3 - 304.2 ft 3 + 1569.6 ft: 3 + 1221 ft 3 = 9428.3 tt3 Interpolate specific volume of steam at 1050 psia and 1100 psia to obtain specific volume of steam at 1170 psia.

Specific Volume of Steam at 1050 psia = 0.42224 ft 3/lb (Ref. 111.3 pg. A-15)

Specific Volume of Steam at 1100 psfa = 0.40058 ft 3/lb (Ref. 111.3 pg. A-15) 3 Specific Volume of Steam at 1070 (by interpolation)= 0.41358 ft /lb 3

Ms, lb = Vst ft 3 / SV10 7o, ft /lb Ms, lb = 9428.3 ft3/ = 0.41358 ft 3/lb = 22,796.8 lb use 22797 lb Mw, lb= Maximum reactor coolant weight at normal water level from Ref. 11.4 Mw = 685,000 lb (Ref. 11.4, pg. 3)

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 27 OF 44 Case 1 Dose Equivalent RCS Iodine Activity Concentration The following equations are placed into an Excel spread sheet.

/

Dose Equivalent RCS Iodine Activity Concentration Ci= CDi (L

  • DCF 1-131 ) I "Z; (CDi
  • DCFi) (Ref. 11.3 pg. 17)

\

Where: Ci= reactor coolant activity concentration for isotope I for Technical Specification reactor coolant dose equivalent 1-131 activity limit L, uCi/gm.

L = Technical Specification dose equivalent 1-131 reactor coolant activity limit

= 0.2 uCi/gm dose equivalent 1-131 steady state limit

= 4.0 uCi/gm dose equivalent 1-131 action level limit CDi = design basis reactor coolant activity concentration for isotope i from Table 7.1, uCi/gm DCF 1_131 = thyroid inhalation dose conversion factor for 1-131, from Table 7.2, converted from units of Sv/Bq to Rem/Ci DCFi = thyroid inhalation dose conversion factor for isotope i, from Table 7.2, converted from units of Sv/Bq to Rem/Ci Equivalent Iodine Activity Released in Water Ci* CF= FCi Where: Ci= reactor coolant activity concentration for Iodine isotope I for Technical Specification reactor coolant dose equivalent 1-131 activity limit L, uCi/gm. J CF = Conversion Factor, 453.6 g/lb

  • 1 E-6 Ci/uCi = 4.54E-4 g*Ci / lbm*uCi FCi = Reactor Coolant Iodine Concentration, Ci/lb Jodine Activity Released in Water Awi = FCi

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 28 OF 44 FCi = Reactor Coolant Iodine Concentration, Ci/lb Mw = Mass of Reactor Water released, lbm Activity Released in Steam Ci *CF= FCi Where: Ci= react_or steam activity concentration for isotope i ~ the TS limit of 290 uCi/sec with 30*minutes decay, uCi/gm.

CF = Conversion Factor, 453.6 g/lb

  • 1 E-6 Ci/uCi = 4.54E-4 g*Ci / lbmuCi FCi = Reactor Steam isotope i Concentration, Ci/lb Asi = FCi

Ms= Mass of_ Steam released, lbm Total Iodine Activity Released Ai =Awi +Asi Where: Ai= total activity released for Iodine isotope i, Ci Awi = reactor coolant water activity released for Iodine isotope i, Ci Asi = reactor coolant steam activity released for Iodine isotoP,_e i, Ci T~e following pages include the results of the above equations performed in an Excel spread sheet for both 0.2 uCi/g 1-131 dose equivalent and 4 uCi/g dose equivalent.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 29 OF 44 0.2 uci/gm 1-131 Dose Equivalent RCS Isotope Design Basis RCS Thyroid Dose CDi *DCFi L, uCi/gm DCF1-131 S CDi*DCFi Ci, uCi/gm Concentration, CDi Conversion (uCi/gm) Factor, DCFi (Rem/Ci) 1-131 2.10E-02 1.08E+06 2.27E+04 0.2 1.08E+06 8.40E+04 5.40E-02 1-132 3.1 OE-01 6.44E+03 2.00E+03 0.2 1.08E+06 8.40E+04 7.97E-01 1-133 2.80E-01 1.80E+05 5.03E+04 0.2 1.08E+06 8.40E+04 7.20E-01 1-134 4.. 90E-01 1.07E+03 5.22E+02 0.2 1.08E+06 8.40E+04 1.26E+OO 1-135 2.?0E-01 3.13E+04 8.45E+03 0.2 1.08E+06 8.40E+04 6.94E-01 Cs-134 1.?0E-04 4.11 E+04 6.98E+OO 0.2 1.08E+06 8.40E+04 4.37E-04 Cs-136 1.10E-04 6.40E+03 7.04E-01 0.2 1.08E+06 8.40E+04 ' 2.83E-04 Cs-137 4.40E-04 2.93E+04 1.29E+01 0.2 1.08E+06 8.40E+04 1.13E-03 TOTAL 8.40E+04 "

RCS Isotope Equivalent RCS Conversion Factor Reactor water Mass Water Activity Released Water, Concentration, Ci, uCi/gm g-Ci / lbm-uCi concentration Fci Released, Mw, lbm Awi, Curies (Ci/lbm) 1-131 5.40E-02 4.54E-04 2.45E-05 6.85E+05 1.68E+01 1-132 7.97E-01 4.54E-04 3.62E-04 6.85E+05 2.48E+02 1-133 7.20E-01 4.54E-04 3.27E-04 6.85E+05 2.24E+02 1-134 1.26E+OO 4.54E-04 5.72E-04 6.85E+05 3.92E+02 1-135 C 6.94E-01 4.54E-04 3.15E-04 6.85E+05 2.16E+02 Cs-134 4.37E-04 4.54E-04 1.99E-07 6.85E+05 1.36E-01 Cs-136 2.83E-04 4.54E-04 1.28E-07 6.85E+05 8.80E-02 Cs-137 1.13E-03 4.54E-04 5.14E-07 6.85E+05 3.52E-01

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 30 OF 44 Reactor Steam Isotope Design Basis Reactor Conversion Factor Reactor Steam Mass Steam Activity Released Steam, Steam Concentration, CDi g-Ci / lbm-uCi concentration Fci Released, Ms, lbm Aws, Curies (uCi/gm) (Ci/lbm)

Kr-85 3.00E-05 4.54E-04 1.36E-08 2.28E+04 3.10E-04 Kr-85m 9.70E-03 4.541::-04 4.40E-06 2.28E+04 1.00E-01 Kr-87 3.30E-02 4.54E-04 1.50E-05 2.28E+04 3.42E-01 Kr-88 3.30E-02 4.54E-04 1.50E-05 2.28E+04 3.42E-01 Xe-133 1.30E-02 4.54E-04 5.90E-06 2.28E+04 1.35E-01 Xe-135 3.60E-02 4.54E-04 1.63E-05 2.28E+04 3.73E-01 1-131 3.40E-04 4.54E-04 1.54E-07 2.28E+04 3.52E-03 1-132 4.?0E-03 4.54E-04 2.13E-06 2.28E+04 4.86E-02 1-133 4.40E-03 4.54E-04 2.00E-06 2.28E+04 4.55E-02 1-134 9.?0E-03 4.54E-0_4 4.40E-06 2.28E+04 1.00E-01 1-135 4.30E-03 4.54E-04 1.95E-06 2.28E+04 4.45E-02 Cs-134 1.?0E-07 4.54E-04 7.72E-11 2.28E+04 1.76E-06 Cs-136 1.10E-07 4.54E-04 4.99E-11 2.28E+04 1.14E-06 Cs-137 4.40E-07 4.54E-04 2.00E-10 2.28E+04 4.55E-06 Isotope Activity Released Water, Awi, Curies Activity Released Steam, Aws, Curies 0.2 uCi/gm 1-131 Equivalent, Curies Kr-85 3.10E-04 3.10E-04 Kr-85m 1.00E-01 1.00E-01 Kr-87 3.42E-01 3.42E-01 Kr-88 3.42E-01 3.42E-01 Xe-133 1.35E-01 1.35E-01 Xe-135 3.73E-01 3.73E-01 1-131 1.68E+01 3.52E-03 1.68E+01 1-132 2.48E+02 4.86E-02 2.48E+02 1-133 2.24E+02 4.55E-02 2.24E+02 1-134 3.92E+02 1.00E-01 3.92E+02 1-135 2.16E+02. 4.45E-02 2.16E+02 Cs-134 1.36E-01 1.76E-06 1.36E-01 Cs-136 8.BOE-02 1.14E-06 8.BOE-02 Cs-137 3.52E-01 4.55E-06 3.52E-01

CALCULATION No.: G13.18.9;4-045 REVISION: 1 PAGE 31 OF 44 4 uci/gm 1-131 Dose Equivalent RCS Isotope Design Basis RCS Thyroid Dose CDi *DCFi L, uCi/gm DCF 1_131 [, CDi*DCFi Ci, uCi/gm Concentration, CDi Conversion (uCi/gm) Factor, DCFi (Rem/Ci) 1-131 2.10E-02 1.08E+06 2.27E+04 4 1.08E+06 8.40E+04 1.08E+OO 1-132 3.1 OE-01 6.44E+03 2.00E+03 4 1.08E+06 8.40E+04 1.59E+01 1-133 2.80E-01 1.80E+05 5.03E+04 4 1.08E+06 8.40E+04 1.44E+01 1-134 4.90E-01 1.07E+03 5.22E+02 4 1.08E+06 8.40E+04 2.52E+01 1-135 2.?0E-01 3.13E+04 8.45E+03 4 1.08E+06 8.40E+04 1.39E+01 Cs-134 1.?0E-04 4.11 E+04 6.98E+OO 4 1.08E+06 8.40E+04 8.75E-03 Cs-136 1.10E-04 6.40E+03 7.04E-01 4 1.08E+06 8.40E+04 5.66E-03 Cs-137 4.40E-04 2.93E+04 1.29E+01 4 1.08E+06 8.40E+04 2.26E-02 TOTAL .* 8.40E+04 RCS Isotope Equivalent RCS Conversion Factor Reactor water Mass Water Activity Released Water, Concentration, Ci, uCi/gm g-Ci I lbm-uCi concentration Fci Released, Mw, lbm Awi, Curies (Ci/lbm) 1-131 1.08E+OO 4.54E-04 4.90E-04 6.85E+05 3.36E+02 1-132 1.59E+01 4.54E-04 7.24E-03 6.85E+05 4.96E+03 1-133 1.44E+01 4.54E-04 6.54E-03 6.85E+05 4.48E+03 1-134 2.52E+01 4.54E-04 1.14E-02 6.85E+05 7.84E+03 1-135 1.39E+01 4.54E-04 6.31 E-03 6.85E+05 4.32E+03 Cs-134 8.75E-03 4.54E-04 3.97E-06 6.85E+05 2.72E+OO Cs-136 5.66E-03 4.54E-04 2.57E-06 6.85E+05 1.76E+OO Cs-137 2.26E-02 4.54E-04 1.03E-05 6.85E+05 7.04E+OO

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 32 OF 44 Reactor Steam Isotope Design Basis Reactor Conversion Factor Reactor Steam Mass Steam Activity Released Steam, Steam Concentration, CDi g-Ci / lbm-uCi concentration Fci Released, Ms, lbm Aws, Curies (uCi/gm) (Ci/lbm)

Kr-85 3.00E-05 4.54E-04 1.36E-08 2.28E+04 3.10E-04 Kr-85m 9.?0E-03 4.54E-04 4.40E-06 2.28E+04 1.00E-01 Kr-87 3.30E-02 4.54E-04 1.50E-05 2.28E+04 3.42E-01 Kr-88 3.30E-02 4.54E-04 1.50E-05 2.28E+04 3.42E-01 Xe-133 1.30E-02 4.54E-04 5.90E-06 2.28E+04 1.35E-01 Xe-135 3.60E-02 4.54E-04 1.63E-05 2.28E+04 3.73E-01 1-131 3.40E-04 4.54E-04 1.54E-07 2.28E+04 3.52E-03 1-132 4.70E-03 4.54E-04 2.13E-06 2.28E+04 4.86E-02 1-133 4.40E-03 4.54E-04 2.00E-06 2.28E+04 4.55E-02 1-134 9.70E-03 4.54E-04 4.40E-06 2.28E+04 1.00E-01 1-135 4.30E-03 4.54E-04 1.95E-06 2.28E+04 4.45E-02 Cs-134 1.70E-07 4.54E-04 7.72E-11 2.28E+04 1.76E-06 Cs-136 1.10E-07 4.54E-04 4.99E-11 2.28E+04 1.14E-06 Cs-137 4.40E-07 4.54E-04 2.00E-10 2.28E+04 4.55E-06 Isotope Activity Released Water, Activity Released 4 uCi/gm 1-131 Awi, Curies Steam, Aws, Equivalent, Curies Curies Kr-85 3.10E-04 3.10E-04 Kr-85m 1.00E-01 1.00E-01 Kr-87 3.42E-01 3.42E-01 Kr-88 3.42E-01 3.42E-01 Xe-133 1.35E-01 1.35E-01 Xe-135 3.73E-01 3.73E-01 1-131 3.36E+02 3.52E-03 3.36E+02 1-132 4.96E+03 4.86E-02 4.96E+03 1-133 4.48E+03 4.55E-02 4.48E+03 1-134 7.84E+03 1.00E-01 7.84E+03 1-135 4.32E+03 4.45E-02 4.32E+03 Cs-134 2.72E+OO 1.76E-06 2.72E+OO Cs-136 1.76E+OO 1.14E-06 1.76E+OO Cs-137 7.04E+OO 4.55E-06 7.04E+OO

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 33 OF 44 A summary of the Case 1 0.2 uCi/g and 4 uCi/g source terms is provided in the following table.

0.2 uCi/gm 1-131 4 uCi/gm 1-131 Isotope Equivalent, Equivalent, Curies Curies Kr-85 3.10E-04 3.10E-04 Kr-85m 1.00E-01 1.00E-01 Kr-87 3.42E-01 3.42E-01 Kr-88 3.42E-01 3.42E-01 Xe-133 1.35E-01 1.35E-01 Xe-135 3.73E-01 3.73E-01 1-131 1.68E+01 3.36E+02 1-132 2.48E+02 4.96E+03 1-133 2.24E+02 4.48E+03 1-134 3.92E+02 7.84E+03 1-135 2.16E+02 4.32E+03 Cs-134 1.36E-01 2.72E+OO Cs-136 8.SOE-02 1.76E+OO Cs-137 3.52E-01 7.04E+OO Determine the Containment and Drywell Dose Rate Readings Containment/Drywell Volume, ft 3 = Containment Volume, ft 3 + Drywell Volume, ft 3 Drywell Volume =236,136 ft 3 -11.811 ft 3 -71.6ft 3 -513ft3

= 235,539.589 ft 3 use 235,540 ft 3 (Ref. 11.16 and EC Markups EC 5000047947, EC 108, EC 23493, EC 25424)

Containment Free Volume = 1,191,590 ft 3 - 15.27 ft 3 - 0.74 ft 3 1,191,573.99 ft3 use 1,191,574 ft 3 (Ref. 11.17 and EC Markups EC 5000048884, EC 5000050413,EC 23493, EC 62305)

Containment/Drywell Volume, ft 3 = 1,191,574 ft 3 + 235,540 ft 3

= 1,427,114 ft 3 OR 1.427E6 ft 3

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 34 OF 44 0

ft 3

  • 2.832E-2 = m 3 (Ref. 111.8 pg. 25)

Containment Free Volume, m3 = 1,191,573.99 ft3

  • 2.832E-2 = 33,745.38 m3 Drywell Free Volume, m3 = 235,539.589 ft 3
  • 2.832E-2 = 6,670.48 m 3 Containment/Drywell Free Volume, m3 = 1,427,114 ft 3
  • 2.832E-2 = 40,416 m 3 Containment and Drywell Volumes for Geometry Factor Elevation of Containment Dome Bend Line = 226' 3" (Ref. 11.21)

Containment Inner Diameter= 120' (Ref. 11.21)

Containment Gross Volume Above Dome Bend Line= 226,195 ft 3 (Ref. 11.17 pg. 10)

Equipment Volume 186' 3" Elevation= 4984.7 ft 3 (Ref. 11.17 pg. 16)

Drywell Inner Diameter= 68' 11.25" (Ref. 11.21)

Reactor Shield Wall Outer Radius = 14' 11" (Ref. 11.24)

Volume of Containment above 186'3", ft 3 =

Cylindrical Volume between 186'3" and dome bend line elevation, ft 3 3

+ containment dome volume, ft Cylindrical Volume between 186'3" and dome bend line elevation, ft 3 = n

  • r2
  • h Where: r = inner radius of containment, ft(= inner diameter/ 2, ft) h = height between 186'3" and dome bend line elevation, ft r = inner radius of containment, ft (= inner diameter/ 2, ft) = 120' I 2 = 60' h = 226' 3" -186' 3" = 40' Cylindrical Volume between 186'3" and dome bend line elevation

= n

  • 60' 2
  • 40' = 452,389.342 ft 3

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 35 OF 44 Volume of Containment above 186'3", ft 3 =

452,389.342 ft 3 + +226, 195 ft 3 = 678,584.342 ft 3 Free Air Volume of Containment above 186'3, ft 3 =

Volume of Containment above 186'3", ft 3 - Equipment Volume above 186'3, ft 3 Ve, Geometry Factor Free Air Volume of Containment above 186'3, ft 3

= 678,584.342 ft 3 - 4984. 7 ft 3 = 673,599.642 ft 3

  • Ve, Geometry Factor Free Air Volume of Containment above 186'3, m 3

= 673,599.642 ft 3

  • 2.832E-2 =19,076.342 m 3 rsw, ft= Reactor shield wall outer radius, ft= 14' 11" or 14.917' 2

Asw, ft = Reactor Shield Wall Area= re* r5 w2 Asw, ft2 = Reactor Shield Wall Area = re* 14.917 2 = 699.057 ft 2 ddw, ft = Drywell Inner Diameter, ft = 68' 11.25" or 68.938' rdw, ft= Drywell Radius, ft= Drywell Inner Diameter, ft/ 2 = 68.938 / 2 = 34.469 ft Adw = Drywell Area= re* rdw2 Adw =re* 34.469 2 = 3732.564 ft2 t, ft = rdw, ft - rsw, ft t, ft = 34.469 ft - 14.917 ft = 19.552 ft Area of Triangle AB, ft 2 = %

  • b
  • h Where b = reactor shield wall inner diameter, ft h = rndius of the drywell, ft Area of Triangle AB, ft 2 = 14.917'
  • 34.469 = 514.174 ft2

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE

  • 36 OF 44 Area of B, ft2 = Yz
  • n
  • rsw2 Area of B, ft 2 = Yz
  • n
  • 14.9172 = 349.529 ft 2 Area of A, ft 2 = Area of Triangle AB - Area of B Area of A, ft 2 = 514.174 ft 2 -349.529 ft2 = 164.645 ft 2 8, radians = 2 cos-1 ( rsw I rdw) 8, radians = 2 cos-1 ( 14.917 / 34.469)

= 2 cos-1( 0.433) = 2.246 radians (128.714 deg)

Area of Sector C or D, ft2 = Yz

  • rdw2 , ft2
  • 8, radians

= Yz

  • 34.469 2 , ft2
  • 2.246 radians= 1334.25 ft 2 As, ft2, area seen by the drywell detector= Area A+ Area C + Area D As, tt2 = 164.645 ft 2 + 1334.25 ft 2 + 1334.25 ft 2 = 2833.145 ft2 DcF, Drywell Volume Correction Factor= As / Adw

= 2833.145 ft2 / 3732.564 ft 2 = 0. 759 3

V 0 , Geometry Factor Drywell Volume, m 3 = Drywell Free Volume, m

  • DcF V 0 , m 3 = 6,670.48 m 3
  • 0. 759 = 5062.894 m 3 0 338 For Containment GFc = 352 / Ve
  • Where: Ve is the Geometry Factor volume of containment, m 3 GFc = 352 / 19,076.342 °* 338 = 12.58 0 338 For Drywell GFo = 352 / Vo
  • Where: Vo is the Geometry Factor volume of the drywell, m3 GFo = 352 / 5062.894 °* 338

= 19. 7

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 37 OF 44 The units used are R/hr rather than Rads/sec. Incorporating these factors into the previous Gamma Dose formula provides the following formula for the gamma dose in containment:

0 338 Dose, y, Containment, R/hr = Li (0.25* Eavei

  • Ci / Vtot) *3600 sec/hr/ (352 / Ve * )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies Vtot = Containment/Drywell volume m3 Ve is the Geometry Factor volume of containment, m 3 03 8 Dose, y, Drywell, R/hr = Li (0.25* Eavei

  • Ci / Vtat) *3600 sec/hr/ (352 / Vo
  • i3 )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration torJsotope i, Curies Vtot = Containment/Drywell volume m 3 V 0 is the Geometry Factor volume of the drywell, m3 An Excel spread sheet is used to calculate the total containment dose rate and total drywell does rate for 0.2 uCi/gm and 4 uCi/g dose equivalent 1-131. The results are shown on the following tables:

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 38 OF 44 0.2 uCi/gm 1-131 Equivalent Containment and Drywell Dose Rate Isotope 0.2 uCi/gm 1- Gamma GFc Containment GFo Drywell Dose 131 Equivalent, Energy, MeV Dose Rate, Rate, R/hr Curies R/hr Kr-85 3.10E-04 2.00E-03 12.58 1.1 OE-09 19.70 7.02E-10 Kr-85m 1.00E-01 1.58E-01 12.58 2.81 E-05 19.70 1.79E-05 Kr-87 3.42E-01 7.93E-01 12.58 4.79E-04 19.70 3.06E-04 Kr-88 3.42E-01 1.96E+OO 12.58 1.18E-03 19.70 7.57E-04 Xe-133 1.35E-01 4.60E-02 12.58 1.10E-05 19.70 6.99E-06 Xe-135 3.73E-01 2.49E-01 12.58 1.64E-04 19.70 1.05E-04 1-131 . 1.68E+01 3.82E-01 12.58 1.14E-02 19.70 7.25E-03 1-132 2.48E+02 2.28E+OO 12.58 1.00E+OO 19.70 6.39E-01 1-133 2.24E+02 6.07E-01 12.58 2.41 E-01 19.70 1.54E-01 1-134 3.92E+02 2.63E+OO 12.58 1.82E+OO 19.70 1.17E+OO 1-135 2.16E+02 1.58E+OO 12.58 6.04E-01 19.70 3.86E-01 Cs-134 1.36E-01 1.55E+OO 12.58 3.73E-04 19.70 2.38E-04 Cs-136 8.80E-02 2.17E+OO 12.58 3.38E-04 19.70 2.16E-04 Cs-137 3.52E-01 6.62E-01 12.58 4.12E-04 19.70 2.63E-04 TOTAL 3.68E+OO 2.35E+OO

  • 4 uCi/gm 1-131 Equivalent Containment and Drywell Dose Rate Isotope 4 uCi/gm 1-131 Gamma GFc Containment GFo Drywell Equivalent, Energy, Dose Rate, Dose Rate, Curies MeV R/hr R/hr Kr-85 3.10E-04 2.00E-03 12.58 1.1 OE-09 19.70 7.02E-10 Kr-85m 1.00E-01 1.58E-01 12.58 2.81 E-05 19.70 1.79E-05 Kr-87 3.42E-01 7.93E-01 12.58 4.79E-04 19.70 3.06E-04 Kr-88 3.42E-01 1.96E+OO 12.58 1.1 SE-03 19.70 7.57E-04 Xe-133 1.35E-01 4.60E-02 12.58 1.1 OE-05 19.70 6.99E-06 Xe-135 3.73E-01 2.49E-01 12.58 1.64E-04 19.70 1.05E-04 1-131 3.36E+02 3.82E-01 12.58 2.27E-01 19.70 1.45E-01 1-132 4.96E+03 2.28E+OO 12.58 2.00E+01 19.70 1.28E+01 1-133
  • 4.48E+03 6.07E-01 12.58 4.81 E+OO 19.70 3.07E+OO 1-134 7.84E+03 2.63E+OO 12.58 3.65E+,01 19.70 2.33E+01 1-135 4.32E+03 1.58E+OO 12.58 1.21 E+01 19.70 7.71 E+OO Cs-134 2.72E+OO 1.55E+OO 12.58 7.46E-03 19.70 - 4.76E-03 Cs-136 1.76E+OO 2.17E+OO 12.58 6.76E-03 19.70 4.32E-03 Cs-137 7.04E+OO 6.62E-01 12.58 8.25E-03 19.70 5.27E-03 TOTAL <

,'\ I

,.,, 7.36E+01 *,  ; . '

4.70E+01

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 39 OF 44 Case 2 An Excel spread sheet with the isotopes from Table 7.4 is used with the following formulas to calculate the containment and drywell doses for this case. The tabular results are provided on the following page.

Gap Releasei, Ci = Ci/MWt i

  • MWt
  • Gap Release Fractioni MWt = 3100 Gap Release Fraction= 0.05 Table 7.3 Gap Releasei, Ci = Ci/MWt 'i
  • 3100 MWt
  • 0.05 Damage Releasei, Ci = Gap Releasei
  • DFs Damage Releasei, Ci = Gap Releasei
  • 0.05 0 338 Dose, y, Containment, Rem/hr= Li (0.25* Eavei
  • Ci / Viot) *3600 sec/hr/ (352 / Ve * )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies V101 = Containment/Drywell volume m3 Ve is the volume of containment, m3 Vtot = 40,426 m3 Ve = 19,076.342 m 3 0 338 Dose, y, Drywell, Rem/hr= Li (0.25* Eavei *Ci/ V101) *3600 sec/hr/ (352 / Vo * )

Where: Dose, y,Containment = Containment gamma dose, R/hr Eave = average gamma energy for isotope i , MeV Ci = isotopic concentration for isotope i, Curies Viot = Containment/Drywell volume m 3 Vo is the volume of the drywell, m 3 Vtot = 40,426 m3 Vo, m3 = 5062.894 m 3

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 40 OF 44 Containment and Drywell Dose Rates with 5% Clad Damage

  • 5% Clad Gamma 24 Month Gap Containment Drywall Half-life, Total Damage Isotope Energy, Cycle Cone., Release, GFc Dose Rate, GFo Dose Rate, (sec) Cone., Ci Release, MeV Ci/MWt Ci R/hr R/hr Ci Kr-85 3.38E+08 2.00E-03 3.66E+02 1.13E+06 5.67E+04 2.84E+03 12.58 -1.ooE-02 19.70 6.41 E-03 Kr-85m 1.61 E+04 1.58E-01 7.02E+03 2.18E+07 1.09E+06 5.44E+04 12.58 1.52E+01 19.70 9.71 E+OO Kr-87 4.58E+03 7.93E-01 1.35E+04 4.19E+07 2.09E+06 1.05E+05 12.58 1.47E+02 19.70 9.38E+01 Kr-88 1.02E+04 1.96E+OO 1.89E+04 5.86E+07 2.93E+06 1.46E+05 12.58 5.08E+02 19.70 3.24E+02 Xe-133 4.53E+05 4.60E-02 5.26E+04 1.63E+08 8.15E+06 4.08E+05 12.58 3.32E+01 19.70 2.12E+01 Xe-135 3.27E+04 2.49E-01 1.99E+04 6.17E+07 3.08E+06 1.54E+05 12.58 6.80E+01 19.70 4.34E+01 1-131 6.95E+05 3.82E-01 2.70E+04 8.37E+07 4.19E+06 2.09E+05 12.58 1.41 E+02 19.70 9.03E+01 1-132 8.21E+03 2.28E+OO 3.92E+04 1.22E+08 6.08E+06 3.04E+05 12.58 1.23E+03 19.70 7.83E+02 1-133 7.49E+04 6.07E-01 5.52E+04 1.71 E+08 8.56E+06 4.28E+05 12.58 4.59E+02 19.70 2.93E+02 1-134 3.16E+03 2.63E+OO 6.06E+04 1.88E+08 9.39E+06 4.70E+05 12.58 2.19E+03 19.70 1.40E+03 1-135 2.38E+04 1.58E+OO 5.17E+04 1.60E+08 8.01 E+06 4.01 E+05 12.58 1.12E+03 19.70 7.15E+02 Rb-86 1.61 E+06 9.50E-02 6.31 E+01 1.96E+05 9.78E+03 4.89E+02 12.58 8.22E-02 19.70 5.25E-02 Cs-134 6.51 E+07 1.55E+OO 6.11 E+03 1.89E+07 9.47E+05 4.74E+04 12.58 1.30E+02 19.70 8.29E+01 Cs-136 1.13E+06 2.17E+OO 2.00E+03 6.20E+06 3.1 OE+05 1.55E+04 12.58 5.95E+01 19.70 3.80E+01 Cs-137 9.51E+08 6.62E-01 3.95E+03 1.22E+07 6.12E+05 3.06E+04 12.58 3.59E+01 19.70 2.29E+01 TOTAL 6.13E+03 3.91E+03

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 41 OF 44 Case3 An instantaneous release of all reactor coolant mass into the primary containment assuming that 20% of the fuel cladding has failed. This corresponds to NEI 99-01 (Ref. 111.1, page 95, 96),

Primary Containment Barrier Threshold, Primary Containment Radiation, Loss 4.A.

As with case 2, this case involves the release of gap activity associated With clad damage. The formulas for analysis of this case are identical to case 2 with the exception of the damage fraction and damage release. '

The damage fraction (DF20) associated with 20% clad damage is 20% of the complete gap release.

Damage Releasei. Ci = Gap Releasei

  • DF20 By definition DF 20 = 0.20 An Excel spread sheet with the isotopes from Table 7.4 is used with the formulas listed in Case 2 and the formula above to calculate the containment and drywell doses for this case. The tabular results are provided on the following page.

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 42 OF 44 Containment and Drywell Dose Rates with 20% Clad Damage Isotope Half-life, Gamma 24 Month Total Cone. Gap 20% Clad GFc Containment GFo Drywall (sec) Energy, Cycle Ci Release, Damage Dose Rate, Dose Rate, MeV Cone., Ci Release, Ci R/hr R/hr Ci/MWt Kr-85 3.38E+08 2.00E-03 3.66E+02 1.13E+06 5.67E+04 1.13E+04 12.58 4.02E-02 19.70 2.56E-02 Kr-85m 1.61 E+04 1.58E-01 7.02E+03 2.18E+07 1.09E+06 2.18E+05 12.58 6.08E+01 19.70 3.89E+01 Kr-87 4.58E+03 7.93E-01 1.35E+04 4.19E+07 2.09E+06 4.19E+05 12.58 5.87E+02 19.70 3.75E+02 Kr-88 1.02E+04 1.96E+OO 1.89E+04 5.86E+07 2.93E+06 5.86E+05 12.58 2.03E+03 19.70 1.30E+03 Xe-133 4.53E+05 4.60E-02 5.26E+04 1.63E+08 8.15E+06 1.63E+06 12.58 1.33E+02 19.70 8.48E+01 Xe-135 3.27E+04 2.49E-01 1.99E+04 6.17E+07 3.08E+06 6.17E+05 12.58 2.72E+02 19.70 1.74E+02 1-131 6.95E+05 3.82E-01 2.70E+04 8.37E+07 4.19E+06 8.37E+05 12.58 5.66E+02 19.70 3.61 E+02 1-132 8.21 E+03 2.28E+OO 3.92E+04 1.22E+08 6.08E+06 1.22E+06 12.58 4.90E+03 19.70 3.13E+03 1-133 7.49E+04 6.07E-01 5.52E+04 1.71 E+08 8.56E+06 1.71 E+06 12.58 1.84E+03 19.70 1.17E+03 1-134 3.16E+03 2.63E+OO 6.06E+04 1.88E+08 9.39E+06 1.88E+06 12.58 8.74E+03 19.70 5.58E+03 1-135 2.38E+04 1.58E+OO 5.17E+04 1.60E+08 8.01 E+06 1.60E+06 12.58 4.48E+03 19.70 2.86E+03 Rb-86 1.61 E+06 9.50E-02 6.31 E+01 1.96E+05 9.78E+03 1.96E+03 12.58 3.29E-01 19.70 2.1 OE-01 Cs-134 6.51 E+07 1.55E+OO 6.11 E+03 1.89E+07 9.47E+05 1.89E+05 12.58 5.19E+02 19.70 3.32E+02 Cs-136 1.13E+06 2.17E+OO 2.00E+03 6.20E+06 3.10E+05 6.20E+04 12.58 2.38E+02 19.70 1.52E+02 Cs-137 9.51 E+08 6.62E-01 3.95E+03 1.22E+07 6.12E+05 1.22E+05 12.58 1.43E+02 19.70 9.16E+01 TOTAL 2.45E+04 1.57E+04

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 43 OF 44 Dose Rate Summary with Accident Instrument Uncertainty Range Containment Maximum Minimum RMS-RE16A, B Reading Reading

, I a ' . ',,

Calculated (+100%), (-50%),

"' ,, Reading, R/hr R/hr R/hr Case 1, 0.2 uCi/g Dose Equivalent 1-131 3.68E+OO 7.36E+OO 1.84E+OO Case 1, 4 uCi/g Dose Equivalent 1-131 7.36E+01 1.47E+02 3.68E+01 Case 2, 5% Clad Damage 6.13E+03 1.23E+04 3.06E+03 Case 3, 20% Clad Damage 2.45E+04 4.90E+04 1.23E+04 Drywell Maximum Minimum

,, " RMS-RE20A, B Reading Reading Calculated (+100%) (-50%)

Reading, R/hr Case 1, 0.2 uCi/g Dose Equivalent 1-131 2.35E+OO 4.70E+OO 1.18E+OO Case 1, 4 uCi/g Dose Equivalent 1-131 4.70E+01 9.40E+01 2.35E+01 Case 2, 5% Clad Damage 3.91E+03 7.83E+03 1.96E+03 r

Case 3, 20% Clad Damage 1.57E+04 3.13E+04 7.83E+03 The above table provides the ranges of indications for the containment and drywell radiation monitors including the accident environment instrument uncertainty. It should be noted under normal conditions that the radiation monitors are required to perform to a +/- 20% of reading uncertainty per STP-511-4249, 4250, 4289, and 4290.

Case 1 - using the 4 uCi/g Dose Equivalent 1-131 value gives a Drywell Post Accident Monitor .

reading range from 23.5 R/hr to 94 R/hr and a Containment Post Accident Monitor reading range of 36.8 to 147 R/hr considering accident instrument uncertainty. Under normal uncertainty conditions the range would be 37.6 R/hr to 56.4 R/hr for the Drywell Post Accident Monitor reading and 58.9 R/hr to 88.3 R/hr for the Containment Post Accident Monitor reading.

I

_)

CALCULATION No.: G13.18.9.4-045 REVISION: 1 PAGE 44 OF 44 In order to ensure abnormal conditions exist, it is recommended that the readings used for Case 1 entry use the lower value of the normal uncertainty band, rather than the accident uncertainty band, with values greater than or equal to 38 R/hr for the Drywell and 59 R/hr for Containment.

Case 2 - The case 2 reading range for the Drywell monitors is 1.96E+03 R/hr to 7.83E+03 R/hr The reading range for Containment Monitors are 3.06E+03 R/hr to 1.23E+04 R/hr. Using the lowest reading of the band and rounding to the nearest decade gives greater than or equal to 2000 R/hr for the Drywell and 3000 R/hr for Containment. '

I Case 3 - The case 3 reading range for the Drywell monitors is 7.83E+03 R/hr to 3.13E+04 R/hr.

The reading range for Containment Monitors are 1.23E+04 R/hr to 4.90E+04 R/hr. Using the lowest reading of the band and rounding to the nearest decade gives greater than or equal to 8,000 R/hr for the Drywell and 12,000 R/hr for Containment.

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