Text

Rev 0 to Engineering Analysis (EA) EA-A-NL-92-012-03, Offsite Doses & CR Habitability from MHA Accounting for Plant Mods to Occur by Cycle 12
	ML18058A404
Person / Time
Site:	Palisades
Issue date:	04/22/1992
From:	Harden P; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18058A401	List: ML18058A399; ML18058A404; ML18058A406; ML18058A408;
References
	EA-A-NL-92-012, EA-A-NL-92-012-03-R0, EA-A-NL-92-12, EA-A-NL-92-12-3-R, NUDOCS 9205050234
	Download: ML18058A404 (94)
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ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 OFF-SITE DOSES AND CONTROL ROOM HABITABILITY FROM THE MHA ACCOUNTING FOR PLANT MODIFICATIONS TO OCCUR BY CYCLE 12 (EA-A-NL-92-012-03)

April 29, 1992

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET EA-A-NL-92-012-03 Total Nud>er of Sheets 93 T;t1e OFFSITE DOSES AND CONTROL ROOM HABITABILITY FROM THE MHA ACCOUNTING FOR PLANT MODIFICATIONS TO OCCUR BY CYCLE 12 INITIATION AND REVIEW Calculat;on Status Preliminary Pe~ing Final Superseded D

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The purpose of this EA is to demonstrate that the radiological consequences of the Maximum Hypothetical Accident (MHA), using the Regulatory Guide 1.4 source term, will be within the limits of 10 CFR 100 and 10 CFR 50 Appendix A, accounting for plant modifications to occur by Cycle 12. The two major forecasted plant modifications affecting the radiological consequences of the MHA are changes to the current method of post-LOCA sump pH control to ensure a sump pH~ 7 by RAS (recirculation actuation signal) and expected elimination of unfiltered air inleakage into the control room while the HV AC system is in the emergency mode.

SUMMARY

OF RESULTS:

The offsite and control room doses from the MHA were calculated, accounting for sump pH control ~ 7 by RAS and assuming zero unfiltered air inleakage to the control room when pressurized in the emergency mode of the HV AC system due to expected plant modifications. The analysis followed the assumptions and guidelines of Regulatory Guide 1.4 and the Standard Review Plan. CV-3027 & CV-3056 were assumed to leak sump water at a rate of 0.1 gpm to the SIRW Tank after RAS, continuing throughout the event. The resultant offsite doses were calculated to be 15.908 rem thyroid and 0.287 rem whole body for 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at the site boundary, and 9.501 rem thryoid and 0.053 rem whole body for 30 days at the low population zone distance. These calculated doses are well within the 300 rem thyroid and 25 rem whole body dose limits of 10 CFR 100. The control room operator doses were calculated to be 9.052 rem thyroid, 6.014 rem skin, and 0.695 rem whole body over 30 days, including shine dose contribution from containment and the SIRW Tank. These calculated doses are well within the limits of 10 CFR 50 Appendix A as interpreted by Standard Review Plan 6.4.

Section PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET TABLE OF CONTENTS EA-A-NL-92-012-03 Sheet _2 _ Rev # -~o __

1.0 Objective......................................................................................................................................................... 3 2.0 References...................................................................................................................................................... 3 3.0 Background.................................................................................................................................................... 6 4.0 Analysis Input................................................................................................................................................ 7 5.0 Assumptions................................................................................................................................................... 12 6.0 Analysis........................................................................................................................................................... 14 6.1 MHACALC Input....................................................................................................................................... 14 6.2 Offsite Doses............................................................................................................................................... 20 6.3 CONDOSE Input....................................................................................................................................... 21 6.4 Control Room Doses................................................................................................................................. 24 7.0 Summary......................................................................................................................................................... 28 8.0 Conclusion...................................................................................................................................................... 30 9.0 List of Attachments...................................................................................................................................... 31

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _3_ Rev #

0.

OFFSITE DOSES AND CONTROL ROOM HABITABILI1Y FROM THE MHA ACCOUNTING FOR PLANT MODIFICATIONS TO OCCUR BY CYCLE 12 1.0 OBJECTIVE The purpose of this EA is to demonstrate that the radiological consequences of the MHA, using the Regulatory Guide 1.4 source term, will be within the limits of 10 CFR 100 and 10 CFR 50 Appendix A, accounting for plant modifications to occur by Cycle 12.

The two major forecasted plant modifications affecting the radiological consequences of the MHA are changes to the current method of post-LOCA sump pH control to ensure sump pH ~ 7 by RAS and expected elimination of unfiltered air inleakage into the control room while the HV AC system is in the emergency mode.

2.0 REFERENCES

2.1 Regulatory Guide 1.4 Rev 2, "Assumptions Used For Evaluating The Potential Radiological Consequences of a Loss of Coolant Accident For Pressurized Water Reactors," June 1974.

2.2 EA-P-LOCA-881024, "Calculation of Offsite Doses Due to the Palisades MHA Including the Effect of the CWRT Vent," October 1988.

2.3 E-PAL-90-035, Event Report "RT-88A Test Failure," September 1990.

2.4 D-PAL-91-178, Deviation Report "Post-WCA Sump pH Control," November 1991.

2.5 NUREG-0800, USNRC Standard Review Plan. Section 6.4 Rev 2, "Control Room Habitability System," July 1981. Section 6.5.2 Rev 2, "Containment Spray as a Fission Product Cleanup System," December 1988. Section 15.6.5 Appendix A Rev 1, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident Including Containment Leakage Contribution," July 1981.

Section 15.6.5 Appendix B Rev 1, "Radiological Consequences of a Design Basis Loss-of-Coolant Accident: Leakage From Engineered Safety Feature Components Outside Containment," July 1981.

2.6 Letter from G.B. Slade to the NRC dated January 10, 1992.

Subject:

"Unreviewed Safety Question - Potential For Leakage of Containment Sump Water to the SIRW Tank During an MHA - Revision 1."

2.7 ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers," July 1978.

2.8 EA-A-NL-92-012-02 Rev 1, "Offsite Doses and Control Room Habitability Following a Large Break LOCA to Justify Continued Operation Until Cycle 12," March 1992.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _4_ Rev # -~o __

2.10 Palisades Plant Final Safety Analysis Report.

2.11 EA-GCP-91-04, "Maximum and Minimum Containment Sump Volume and Boron Concentration Following a Large Break LOCA," November 1991.

2.12 Palisades Plant Technical Specifications.

2.13 Letter from A. Schwencer (NRC) to D. Bixel (CPCo) dated November 1, 1977.

Subject:

Transmittal of Amendment No. 31 with Safety Evaluation. Cart/Frame: 2511/1751.

2.14 EA-D-PAL-89-222A, "LOCA FSAR Update," December 1989. Cart/Frame: C290/1032.

2.15 R0-119 Rev 0, Technical Specification Surveillance Procedure, "Inservice Testing of Engineered Safeguards Valves CV-3027 and CV-3056," February 1992.

2.16 EA-PAH-91-05, "Benchmarking of the MHACALC Code," March 1992.

2.17 Drawing C-38 Rev 5, "Field Erected Tanks Sheet 2," January 1989.

2.18 Isometric Drawing M-107 Sheet 2201 Rev 2, "Bldg Location: ESS thru CCW, MSIV RM, to SIRW," November 1988.

2.19 Level Settings Diagram M-398 Sheet 18 Rev 4, "Safety Injection & Refuelling Water Tank T-58."

2.20 EA-PAH-91-06, "Iodine Removal Coefficients for Containment Sprays Based on Standard Review Plan 6.5.2, Revision 2," December 1991. Cart/Frame: F005/2454.

2.21 NED0-24782, "BWR Owner's Group NUREG-0578 Implementation: Analysis and Positions for Plant Unique Submittals," General Electric, August 1980.

2.22 NUREG/CR-1413 ORNL/NUREG-70, "A Radionuclide Decay Data Base - Index and Summary Table," Oak Ridge National Laboratory, May 1980.

2.23 Drawing M-6 Sheet 1Rev13, "Equipment Location - Reactor Bldg. Sections A-A, B-B, C-C, D-D, & E-E," August 1968.

2.24 Drawing C-78 Rev 5, "Control Room Details," December 1990.

2.25 Drawing M-4 Rev 17, "Equipment Loe. - Reactor+ Aux. Bldg. Radwaste Modifications Plan of El. 625'-0"," December 1990.

2.26 EA-A-PAL-90-018-01, "DBA Sequencer Timing Study," July 1991.

2.27 Internal Correspondence LTP91 *011, from LTPhillips to PAHarden dated August 30, 1991.

Subject:

"DBD 1.06 Open Item #12 Control Room HV AC - Fresh Air Make-Up."

.,,.,.,ln,..WSI PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _5_ Rev # -"""'"O __

2.29 DBD-1.06, "Palisades Design Basis Document - Control Room HV AC System," Rev. 0 -

December 1990.

2.30 Letter from J.G. Kovach (Bechtel) to B.L. Harshe (CPCo) dated March 29, 1990.

Subject:

"Palisades Nuclear Plant CPCo Contract CPll-7723, Bechtel Job 20592-001, Transmittal of Bechtel Calculation 001-N-002, Rev l."

2.31 Letter from J.G. Kovach (Bechtel) to B.L. Harshe (CPCo) dated December 28, 1990.

Subject:

"Palisades Nuclear Plant CPCo Contract CPll-8045, Bechtel Job 20592-010, Transmittal of New Control Room x/Q Values." Bechtel Cale. No. 001-N-001Rev1..

2.32 EA-PAH-92-01, "Verification of Control Room Atmospheric Dispersion Calculations For Releases From The SIRW Tanlc, Performed By Bechtel Power Corporation," February 1992.

2.33 EA-A-NL-92-012-01, "Benchmarking of the CONDOSE Code For Control Room Habitability Calculations," March 1992.

2.35 Pal. Spray Notebook, November 1989.

2.36 EOP 4.0 Rev 2, Emergency Operating Procedure "Loss of Coolant Accident Recovery," July 1990.

2.37 Internal Correspondence WLR *004 from WLRoberts to PMDonnelly dated February 5, 1992.

Subject:

"Palisades Plant - Continuation of Conversations with the NRC on CRHAB and MHA Issues."

2.38 Regulatory Guide 1.25, "Assumptions Used For Evaluating The Potential Radiological Consequences of a Fuel Handling Accident in The Fuel Handling and Storage Facility For Boiling and Pressurized Water Reactors," March 1972.

2.39 ASME Steam Tables, Fourth Edition, 1979.

2.40 Piping Class Sheet M-260 Sheet 9 Rev 5, "Class HC."

2.41 NUREG /CR-5732 ORNL/NUREG-11861, "Iodine Chemical Forms in L WR Severe Accidents,"

Oak Ridge National Laboratory, Draft Report July 1991.

2.42 Crane Technical Paper No 410, "Flow of Fluids Through Valves, Fittings, and Pipe," 1988.

2.43 "The Engineers' Manual," Second Edition, Ralph G. Hudson, John Wiley & Sons, Inc.

2.44 NUREG/CR-4697 ORNL/TM-10135, "Chemistry and Transport of Iodine in Containment,"

Oak Ridge National Laboratory, October 1986.

2.45 "Consumers Power Company Palisades Plant NUREG-0578 Design Review Study of Plant Shielding for Post-Accident Operations," Commonwealth Associates Inc. January 15, 1982.

Cart/Frame: 2687 /1277.

I I

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _6_ Rev # --=o __

2.46 Commonwealth Associates Calculation 0350-1-270-63-23-1-04 Rev 0, "Control Room Doses,"

December 1980. Cart/Frame: 2687 /1819.

2.47 "MICROSHIELD Version 3 User's Manual," Grove Engineering, Inc. October 1987.

2.48 Drawing C-78 Rev 5, "Control Room Details," December 1990.

2.49 Drawing C-539 Rev 0, "Cellular Slab Repair Plan of Control Room Roof El 643'-0"," February 1986.

2.50 R0-28 Rev 2, Technical Specification Surveillance Procedure, "Control Room/TSC Ventilation,"

October 1990.

2.51 Letter from E.C. Beahm (Martin Marietta Energy Systems Inc. at ORNL) to Jay Y. Lee (NRC) dated February 5, 1992.

Subject:

Technique for calculating iodine partitioning in a Safety Injection Refueling Water Tank. (Attached)

3.0 BACKGROUND

The MHA (Maximum Hypothetical Accident) analysis is the bounding radiological consequence analysis for a large break LOCA For the MHA, the source terms of Regulatory Guide 1.4 [Ref. 2.1] and Standard Review Plan (SRP) 15.6.5 [Ref. 2.5] are used. This source term amounts to the release of 25 % of the core iodine inventory to the containment atmosphere, 50 % of the core iodine inventory to the containment sump, and 100 % of the core noble gas inventory to the containment atmosphere.

The resultant doses from the MHA analysis must be demonstrated to be within the limits of 10 CFR 100 for the site boundary and low population zone, and within the limits of 10 CFR 50 Appendix A for the control room.

The previous MHA analysis of record [Ref. 2.2] was found to be no longer bounding after the discovery of several problems with the plant configuration and procedures. One problem was that the method and procedures for post-LOCA sump pH control did not meet current regulatory requirements for the time at which a neutral sump pH must be achieved [Ref. 2.4]. The MHA analysis of record did not account for the deviation from the requirements as directed by SRP 6.5.2 [Ref. 2.5]. Another problem was the discovery of a previously unanalyzed potential leak path of sump water outside of containment

[Ref. 2.3]. Because of these problems and the little margin that existed for control room thyroid doses in the analyses of record, it appeared questionable whether the control room thyroid dose limit of 30 rem could be met for the MHA. Therefore, exemption from performing an MHA analysis considering the above mentioned problems until the end of Cycle 11 was requested from the NRC [Ref. 2.6]. To justify operating until all of the necessary plant modifications can be performed, an analysis using plant

specific fuel failures for a large break LOCA was completed to demonstrate that the offsite and control room doses would be within the limits [Ref. 2.8]. It is anticipated that the modifications relative to mitigating the consequences of a LOCA will be completed prior to the start-up for Cycle 12. The two major forecasted modifications are to ensure that the post-LOCA sump pH is controlled above 7 by

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _7_ Rev # ___ o __

RAS (recirculation actuation signal) and attempt to eliminate unfiltered air inleakage into the control room while the HV AC system is pressurized in emergency mode.

This MHA analysis is being completed to account for the modifications that will be performed prior to Cycle 12. To perform this analysis, three computer codes are used; the MHACALC code [Ref. 2.16],

the CONDOSE code [Ref. 2.33], and the MICROSHIELD code [Ref 2.47]. The MHACALC code is used to the model the radionuclide behavior in and transport from the containment building and SIR W Tank. The MHACALC code specifically calculates radionuclide release rates from the containment building and SIRW Tank, and the resultant doses at the site boundary and low population zone. The radionuclide release rates given from the MHACALC code are then used as input to the CONDOSE code to model the transport of the radionuclides into and out of the control room, and to calculate the resultant control room operator doses. Both of those codes are plant specific codes on the Reactor &

Safety Analysis Engineering Department's super mini IBM 3091 computer, VMS. The codes use the methodology and dose conversion factors of ICRP30 [Ref. 2. 7] to calculate doses due to inhalation of iodine and exposure to noble gas. The MICROSHIELD code is a PC based general shielding code and is used to calculate the shine dose rate in the control room from the iodine activity in the SIR W Tank that is given in the output of the MHACALC code.

4.0 ANALYSIS INPUT 4.1 The breathing rates for offsite and control room doses are 3.47E-04 ml /sec from 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 1.75E-04 rrf /sec from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2.32E-04 rrf /sec from 24 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months in accordance with Reference 2.1.

4.2 The rated core thermal power, 2530 MWu the containment design leak rate, 0.1 %/day, and the containment net free air volume, 1.64E + 06 ft3, are from the plant FSAR [Ref. 2.10, sections 1.1, 1.2, & 5.8]. The atmospheric dispersion factor for 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at the site boundary is 1.55E-04 sec/ml and the atmospheric dispersion factors at the low population zone are 1.09E-05 sec/ml from 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 6.94E-06 sec/ml from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 2.58E-06 sec/ml from 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , and 6.25E-07 sec/rrf from 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months , also from Reference 2.10 [Tables 2-17 & 2-18].

4.3 The minimum sump volume after recirculation is based on a sump mass of 2381842 lbm at 25.4 psia and 238.3°F from Reference 2.11.

4.4 The limit for out of containment ESF leakage into the safeguards rooms is 0.2 gpm from the plant Technical Specifications [Ref. 2.12, TS 4.5.3].

4.5 The iodine reduction factor, 2, for automatic isolation of the ventilation in the safeguards rooms upon high radiation signal is from Reference 2.13.

4.6 The peak sump temperature after recirculation is 232°F at a containment pressure of 34.33 psia, occurring at 1550 seconds in the current LOCA containment analysis [Ref. 2.14].

4.7 The acceptance criteria for leakage through CV-3027 & CV-3056 is 0.1 gpm [Ref. 2.15].

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _8_ Rev # ____ o __

4.8 The volatile fraction of the iodine in the sump water that enters the SIRW Tank is 3.0E-04 [Ref.

2.51].

4.9 The total SIRW Tank volume is calculated from the dimensions given and measured from Reference 2.17.

4.10 The volume of pipe leading to the SIRW Tanlc from CV-3027 & CV-3056 is calculated from the dimensions given on Reference 2.18.

4.11 The volume of water present in the SIRW Tanlc after RAS is calculated based on the low level setting of 24" from Reference 2.19.

4.12 The first-order spray removal coefficients for particulate iodine are 4.43 ht1 initially, changing to 0.443 ht 1 after 98 % of the particulate iodine has been removed [Ref. 2.20]. The spray removal coefficient for elemental iodine is 21.3 ht1 and that for organic iodine is 0.0, also from Reference 2.20.

4.13 The long term iodine partition coefficient in containment is estimated to be 1250 from Reference 2.44 [Figure 6], which is referenced by SRP 6.5.2 [Ref. 2.5, pg. 6.5.2-11 ].

This partition coefficient is based on a pH of 7, a sump temperature of 132°F [Ref. 2.14], and 50 % of the total core iodine inventory being deposited in the containment sump as will be discussed later in the analysis.

4.14 The activity source term values (S.) listed in Table 1 for each of the radionuclides of interest are from Reference 2.21 [App. B, pg. B-27]. (Values in Ref. 2.21 for iodine are listed as 50 % of the total inventory as noted on page B-2.) Although Reference 2.21 was generated for BWR's, distribution of the fission products of main interest would be approximately the same for PWR's.

Use of this document for source terms was discussed during a meeting with the NRC on January 15, 1992, and was agreed to as acceptable.

4.15 The radioactive decay constants ()..1) listed in Table 1 are calculated from the half-life values listed in Reference 2.22, using the equation). = ln(2)/(Half-Life).

4.16 The thyroid and whole body dose rate conversion factors for noble gas isotopes are listed in Table 1 for a semi-infinite cloud and in Table 2 corrected for a 1000 m3 room. These values were taken from Reference 2.7 and converted to units of (Rem/sec)/(Ci/m 3 ). The thyroid and whole body inhalation dose conversion factors for iodine isotopes are also listed in Table 2. These values also came from Reference 2.7, and are converted to units of Rem/Ci-inhaled. The use of the noble gas dose rate conversion factors corrected for a 1000 m 3 room is justified by the calculated control room air volume shown later in this analysis. It should be noted that the whole body dose and dose rate conversion factors for each iodine and noble gas radionuclide, respectively, are the sum of the weighted dose or dose rate factors for all organs and tissues listed in Reference 2. 7 for the radionuclide.

Nuclide Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 1-131 1-132 1-133 1-134 1-135 PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet _9 _ Rev #

0 TABLE 1 ISOTOPE DEPENDENT PARAMETERS Dose Rate Conv Factors (semi-infinite cloud)

(Ci/M\\Vi)

(miif1)

(Rem/sec)/(Ci/rrr)

Thyroid Whole Body 2.998E+03 6.313E-03 O.OOOE-00 3.649E-06 6.498E+03 2.579E-03 3.083E-02 3.031E-02 2.999E+02 1.230E-07 O.OOOE-00 4.738E-04 1.155E+04 9.084E-03 1.439E-01 1.447E-01 1.690E+04 4.068E-03 3.803E-01 3.690E-01 1.993E+04 2.194E-01 0.000E-00 O.OOOE-00 1.760E+02 4.065E-05 O.OOOE-00 1.324E-03 1.954E+03 2.198E-04 O.OOOE-00 5.375E-03 5.648E+04 9.177E-05 7.297E-03 6.259E-03 1.698E+04 4.513E-02 O.OOOE-00 7.647E-02 9.781E+03 1.268E-03 O.OOOE-00 4.676E-02 4.705E+04 1.810E-01 O.OOOE-00 O.OOOE-00 4.433E+04 4.906E-02 1.953E-01 1.969E-01 2.938E+04 5.987E-05 N/A N/A 4.160E+04 5.023E-03 N/A N/A 4.808E+04 5.554E-04 N/A N/A 6.218E+04 1.318E-02 N/A N/A 4.922E+04 1.748E-03 N/A N/A

TABLE 2 IODINE DOSE CONVERSION FACTORS l ROOM-CORRECTED NOBLE GAS DOSE RATE CONVERSION FACTORS Dose Rate Conversion Factors Corrected for a 1000 m 3 room (Rem/sec)/(Ci/m 2

)

Nuclide Thyroid Lungs B Surface B Marrow Skin Eye Lens Whole Bod Kr-83m O.OOOE-00 O.OOOE-00 6.475E-06 5.653E-6

1. 747E-04 1.747E-04 3.649E-06 Kr-85m 1.233E-03 1.131E-03 l.953E-03 1.850E-03 5.139E-02

1. 542E-03 1.269E-03 Kr-85 O.OOOE-00 2.056E-05 3.083E-05 2.878E-05 4.728E-02 3.906E-05 2.314E-5 Kr-87 5.550E-03 5.756E-03 6.886E-03 6.269E-03 3.392E-Ol l.007E-Ol 5.684E-03 Kr-88

1. 439E-02 l.336E-02

1. 542E-02 1.336E-02 9.456E-02 2.878E-02 1.402E-02 Kr-89 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 Xe-13lm O.OOOE-00

1. 233E-04 3.597E-04 3.289E-04

1. 542E-02 5.344E-04 l.915E-04 Xe-133m O.OOOE-00 2.775E-04 6.167E-04 5.653E-04 3.083E-02 7.503E-04 3.823E-04 Xe-133 4.317E-04 2.672E-04 6.886E-04 6.269E-04 1.131E-02 6.989E-04 3.361E-04 Xe-135m O.OOOE-00 3.392E-03 4.522E-03 4.214E-03

2. 672E-02 4.522E-03 3.618E-03 Xe-135 O.OOOE-00

1. 850E-03 2.981E-03 2.775E-03 6.578E-02 2.467E-03 2.086E-03 Xe-137 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 Xe-138 7.811E-03 8.119E-03 9.558E-03 8.736E-03 l.644E-Ol 3.l86E-02 7.801E-03 Inhalation Dose Conversion Factors (Rem/Ci-inhaled) 1-131 1.073E+06 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 3.256E+04 1-132 6.290E+03 9.990E+02 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 3.367E+02 1-133 l.813E+05 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 5.550E+03 1-134 l.073E+03 5.180E+02 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO l.106E+02 1-135 3.145E+04 l.628E+03 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 1.121E+03 Vl

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PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 11 Rev #

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4.17 The volume of the control room is calculated from the dimensions given in References 2.23, 2.24,

& 2.25.

4.18 From Reference 2.26, when loss of offsite power is accompanied by a safety injection signal, a 0.5 second time delay exists before generation of a containment high pressure ( CHP) signal [Ref.

2.26, pg. 6] accompanied by a time delay of 67.5 seconds for diesel generator sequencing and starting the control room air handling units V-95 and V-96 [Ref. 2.26, pg. 13].

4.19 The time to pressurize the control room to the required 0.125" WG (water gauge) after start of V-95 and V-96 is 7.34 seconds (from data taken by LTPhillips as documented in Attachment 1 of Reference 2.8).

4.20 The fresh air make-up flow rate for the control room HV AC emergency mode is 1000 cfm when the system flow rate is 3200 cfm as documented in Reference 2.27.

4.21 The total filtered air flow rate of the control room HV AC system when in emergency mode (make-up plus recirculation) is 3200 cfm +/- 10 % [Ref. 2.50].

4.22 The efficiency of the control room HV AC charcoal filters for the remote intake air and recirculation air is 99.0 % for iodine [Ref. 2.12, Table 4.2.3]. The charcoal is actually tested to a higher efficiency.

4.23 The atmospheric dispersion factors for the normal control room air intake and the aux. bay roof doors from releases at the location of the stack are listed in Table 3, from Reference 2.30. The atmospheric dispersion factors listed in Table 3 for the control room remote, or emergency air intake from releases at the location of the stack are from Reference 2.31. The atmospheric dispersion factors listed in Table 3 for the control room normal and remote air intakes from releases at the SIR W Tank vent are from Reference 2.32.

It should be noted that the atmospheric dispersion factors for the location of the stack are not based on an elevated release.

TABLE 3 ATMOSPHERIC DISPERSION FACTORS FOR CONTROL ROOM INTAKES SIRW Tank Vent x/O Containment Stack Location x /Q TIME (sec/rrr)

(sec/m 3

)

(Hrs)

Normal Remote Normal Remote Aux. Bay 0-8 1.14E-02 8.06E-04 1.84E-03 1.22E-03 3.85E-03 8-24 8.lOE-03 5.60E-04 l.65E-03 l.02E-03 3.SOE-03 24-96 6.38E-03 4.35E-04 1.30E-03 6.80E-04 2.85E-03 96-720 4.60E-03 2.85E-04 9.00E-04 3.98E-04 2.15E-03

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 12 Rev # _....:;.o __

4.24 The control room occupancy factors to account for different control room man power requirements over time are 1.0 from 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 0.6 from 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , and 0.4 from 96 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months in accordance with Standard Review Plan (SRP) 6.4 [Ref. 2.5, Table 6.4-1 ].

4.25 The whole body dose from containment shine at the highest dose point in the control room is obtained from References 2.45 & 2.46. The highest dose point is at the access doors to the control room viewing gallery, which receive a dose of 400 mrem (0.4 rem) over 30 days [Ref.

2.45, pg. 20]. Using this value is conservative since, otherwise, the maximum dose point actually in the control room is the southwest comer of the room which receives 13.3 mrem over 30 days

[Ref. 2.46, pgs. 5-6]. Occupancy factors were also conservatively ignored in calculating this value.

4.26 From Reference 2.17, there is al" minimum of sand and a 10" concrete pad under the SIRW tank.

4.27 From Reference 2.49, the control room roof is a cellular design consisting of two 1' slabs of concrete separated by a 4' air space.

5.0 ASSUMPTIONS 5.1 All assumptions inherent in the methodology of the MHACALC code as documented in Reference 2.16.

5.2 All assumptions inherent in the methodology of the CONDOSE code as documented m Reference 2.33.

5.3 The core is assumed to be at 102 % of rated power, or 2580.6 M\\Vi.

5.4 The event begins with instantaneous release of the source term to the containment building.

5.5 Remaining consistent with Regulatory Guide 1.4 [Ref. 2.1] and SRP 15.6.5 Appendices A & B

[Ref. 2.5], 100 % of the core noble gas inventory is assumed to be released to the containment atmosphere.

Also, 25 % of the core iodine inventory is assumed to be released to the containment atmosphere and 50 % of the core iodine inventory is assumed to be released to the sump solution.

5.6 Again following Regulatory Guide 1.4 and SRP 15.6.5 Appendix A, 91 %*of the iodine released to the containment atmosphere is assumed to be elemental, 5 % is assumed to be particulate, and 4 % is assumed to be organic.

5.7 Loss of offsite power occurs coincident with the event.

5.8 Full containment spray flow is conservatively assumed to be achieved at 1 minute after initiation of the event, accounting for diesel generator sequencing and full flow delivery. Actual time is less than 1 minute for all cases of available spray pumps [Ref. 2.35].

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 13 Rev # -"""'o __

5.9 RAS (recirculation actuation signal) is generated in the minimum time of 19 minutes.

5.10 The recirculation volume is at minimum.

5.11 In the MHACALC code, iodine in the air volume of the SIRW Tank exits the tank at the rate at which air is being displaced from the tank times a multiplication factor to account for diffusion out the vent and changes in air density. This multiplication factor is assumed to be 2, which is consistent with the treatment of ESF leakage in SRP 15.6.5 Appendix B [Ref. 2.5]. This factor seems conservative and was informally agreed to as acceptable in telephone discussions with the NRC as documented in Reference 2.37.

5.12 For calculation of the iodine partition coefficient for the volatile species of iodine in the SIRW Tank, the air-water interface temperature is assumed to be 100°F.

5.13 A sump solution pH of 7 is achieved by the onset of recirculation and maintained throughout the event. This assumption is to be addressed as one of the modifications to be completed prior to Cycle 12.

5.14 Containment sprays are assumed to be terminated at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5.15 The source of radionuclide ingress into the control room when depressurized is through the Aux.

Bay roof doors [Ref. 2.30].

5.16 There is no unfiltered air inleakage when the control room HV AC system is pressurized in the emergency mode. This assumption is also to be evaluated as one of the modifications to be completed prior to Cycle 12.

5.17 The air flow rates of the control room HV AC system are assumed to be at the maximum allowed, which is 3200 cfm + 10 % = 3520 cfm. Assuming the make-up and the recirculation flow rates are proportional, 1000/3200 = 0.3125 and 2200/3200 = 0.6875, the make-up and recirculation air flow rates are 0.3125*3520 = 1100 cfm and 0.6875*3520 = 2420 cfm, respectively.

5.18 Following SRP 6.4 [Ref. 2.5, pg. 6.4-8], the base infiltration rate of air into the control room when depressurized is assumed to be one-half the leakage from the control room when pressurized to Ya 11 water gauge. The leakage from the control room when pressurized to Ya 11 water gauge is equal to the make-up air flow rate to maintain steady pressurization. The base air infiltration rate is then 1100/2 = 550 cfm. A contribution from opening and closing doors is not accounted for since vestibules have been installed on the control room entrances [Ref. 2.5, SRP 6.4, pg. 6.4-9].

5.19 Of the total air volume in the control room, viewing gallery, and technical support center, 5 %

is assumed to be occupied by equipment and walls.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 14 Rev # _.,:..o __

5.20 For control room shine dose calculations from the SIRW Tank, all of the iodine that leaked through CV-3027 & CV-3056 and has not been released from the tank is assumed to be in the water volume of the SIRW Tank. The 6" piping leading to the SIRW Tank is therefore ignored since the activity in that volume is assumed to be in the tank.

5.21 The activity in the water volume of the SIRW Tank for the shine dose calculations is assumed to be evenly distributed, and acting as a self-shielding source.

5.22 The SIRW Tank is conservatively assumed to be directly over the control room for shine dose calculations. It is actually centered closer to the viewing gallery.

5.23 The aluminum wall of the SIRW Tank is conservatively assumed to be insignificant for shielding gamma radiation to the control room.

5.24 For calculations of shine dose from the SIRW Tank, the dose point is taken as 6' off the control room floor, assuming the average operator is 6' tall and stands under the tank for the amount of time dictated by the occupancy factors.

5.25 The concrete between the SIR W Tank and the control room is ordinary concrete with a density of 2.35 g/ cc, which is given as the default density in the MICROSHIELD code. Default density for all other materials are also used.

5.26 Any material in the control room drop ceiling is conservatively ignored for the shine dose calculation from the SIRW Tank.

5.27 The Taylor method for calculating buildup factors is appropriate for the MICROSHIELD calculations. The SIRW Tank water volume in which the activity is dispersed is used as the material basis for buildup calculations. The water volume is used as the material basis for the buildup calculations since it is the single most dominant shield.

5.28 The control room dose rate due to shine from the SIRW Tank is conservatively assumed to remain constant over the time interval for each dose calculation.

6.0 ANALYSIS 6.1 MHACALC INPUT The values used for each input of the MHACALC code are described in detail below. A description of the input requirements and the structure for the input deck can be obtained from Reference 2.16, which also describes the capability and limitations of the code.

The first line of the input deck is the title or description of the case to be executed. Line 2 is the debug option, which is specified as 1 for this case since debugging the output is not desired. The third line is the duration of the analysis, which is 43200 minutes (30 days).

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 15 Rev # --=-0 __

The fourth line of the input deck contains the reactor thermal power, containment design leak rate, and the recirculation water volume, respectively. The reactor thermal power will be taken as 102 % of the rated value of 2530 M\\Vi [Ref. 2.10], which is 2580.6 M\\Vi. The design leak rate of the containment building is 0.1 %/day [Ref. 2.10, 5.1-6], which the MHACALC code automatically reduces by a factor of 2 after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to remain consistent with the guidance of SRP 15.6.5 [Ref. 2.5]. Based on a sump mass of 2381842 Ihm at 25.4 psia and 238.3°F [Ref. 2.11], the containment sump volume is taken as minimum, 40324.6 ft 3

The fifth line contains the percentages of the core inventory of noble gas and iodine released to the containment atmosphere and iodine released to the containment sump, respectively.

Following Regulatory Guide 1.4 [Ref. 2.1] and SRP 15.6.5 [Ref. 2.5], 100 % of the core inventory of noble gas is assumed to be released to the containment atmosphere, 25 % of the core inventory of iodine is assumed to be released to the containment atmosphere, and 50 % of the core inventory of iodine is assumed to be released to the containment sump solution.

The sixth line of the input deck contains the fractions of the iodine released to the containment atmosphere that is released in elemental, particulate, and organic forms, respectively. Again from References 2.1 & 2.5, the percentages are 91 % elemental iodine, 5 % particulate iodine, and 4 %

organic iodine.

The seventh line of the input deck contains the following parameters:

1) the retention factor for iodine released into the safeguards rooms,

2) the* iodine partition coefficient for the sump water leakage into the safeguards rooms,

3) the iodine partition coefficient for sump water leakage into the SIRW Tank,

4) the multiplication factor for the rate at which iodine is released from the SIRW Tank,

5) the total volume of the SIRW Tank and recirculation line leading to it,

6) the volume of water in the SIR W Tank at RAS and recirculation line leading to it,

7) the percentage of iodine in the sump water reaching the SIR W Tank that is in volatile form.

The retention factor for iodine released in the safeguards rooms is 2, as was accepted by the NRC in Reference 2.13 since the ventilation dampers in the safeguards rooms automatically isolate upon a high radiation signal in the rooms. This retention factor is used to account for the plateout of iodine onto surfaces in the safeguards rooms since the ventilation rates would be very low after isolation of the dampers.

For the iodine partition coefficient in the safeguards rooms, SRP 15.6.5 Appendix B [Ref. 2.5] states that the fraction of the iodine released to the rooms from ESF leakage should be taken as 10 %

(partition coefficient = 1/.1 = 10) or the flashing fraction, whichever is greater, unless a higher value can be justified based on actual sump pH. Since the leakage would be at the same pH as the sump solution, greater than 7, the partition coefficient used for the containment sump solution would be appropriate. However, it must be ensured that the flashing fraction would not be dominating. Using the peak post-RAS sump temperature of 232°F and containment pressure of 34 psia from Reference 2.14, and the corresponding enthalpy values from Reference 2.39 including that for 212°F and 14.7 psia, the fraction of ESF leakage flashing to steam is (200.38 - 180.17)/970.3 = 0.021 or 2.1 %. This would correspond to an instantaneous partition coefficient of 1/.021 = 47.6, which bounds the iodine partition coefficient for the sump solution. The sump solution iodine partition coefficient, which will be shown

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 16 Rev # _...::.o __

later in this analysis, is much higher resulting in less iodine release. This partition coefficient for the ESF leakage is considered conservative since sump temperature decreases throughout the event, and would eventually be bounded by the sump solution partition coefficient once the leakage stops flashing to steam. The fact that only a fraction of the iodine in the sump water is in a volatile form, as is accounted for in the SIR W Tank, is conservatively ignored for the ESF leakage since the flashing fraction is assumed to be controlling the amount of iodine released. Therefore, 47.6 is used as the partition coefficient for ESF leakage.

The third value on line 7 of the input deck is the iodine partition coefficient for the SIRW Tank. The iodine partition coefficient for 12, the volatile species of iodine in the sump water, can be calculated using Equation (1) below. This equation is from Reference 2.41 [pg. 29], for which the applicability is verified in Reference 2.51(Attachment1) since the SIRW Tank pH would be approximately 5 and the radiation dose in the SIRW Tank would be relatively low.

where p

T log10 P= 6.29-0. 0149 T (1)

= ratio of 12 in liquid phase to 12 in the gas phase, or the iodine partition coefficient

= air-water interface temperature, 0K.

To bound the case, since the SIR W Tank sits outdoors and the iodine partition coefficient decreases with increasing temperature according to Equation (1), the interface temperature is assumed to be 100°F, corresponding to 310.8°K This temperature results in an iodine partition coefficient of 45.6 for the volatile iodine in the SIR W Tank.

The fourth value on line 7 of the input deck is a multiplication factor for the rate at which iodine in the air volume of the SIRW Tank exits through the tank vent. In the MHACALC code, the SIRW Tank is modeled such that iodine in the air volume of the tank exits at the same rate as the air being displaced from the tank due to the water leaking in. The multiplication factor is used to conservatively encompass any possible diffusion of iodine out of the vent or density changes that force air, and iodine in the air, out of the vent. This factor is specified as 2, being somewhat consistent with SRP 15.6.5 Appendix B [Ref. 2.5] treatment of ESF leakage, which must also be multiplied by a factor of 2. The use of a factor of 2 was mutually agreed to as acceptable during telephone conversations with the NRC, as discussed in Reference 2.37.

The SIRW Tank total volume and volume at RAS, both including the mini-flow recirculation line between CV-3027 & CV-3056 and the SIRW Tank, are fifth and sixth values on line 7 of the input deck for the MHACALC code. The volume of the SIRW Tank is calculated in two parts, the cylindrical portion of the tank and the conical top of the tank. From Reference 2.17, the cylindrical portion of the tank has a diameter of 46' and height of 24', which yields a volume of 1tr2L = 39885.66 ft3. From Reference 2.17, the conical top section of the tank has a slope of 1A" per foot, indicating a vertical distance of 5.75" for the 23' radius. The volume of the conical top section of the tank, using the following equation given in Reference 2.43 [pg. 18, #54], is Va7t r 2h = 265.443 ft3. This results in a total volume of 40151.103 ft3 for the SIRW Tank. The volume of the line between CV-3056 and the SIRW Tank must be added to this.

To calculate the volume of the line between CV-3056 and the SIRW Tank, Reference 2.18 is first used to determine the length of pipe between the valve and the tank. Reference 2.40 is then used to

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 17 Rev # -""""o __

determine the pipe schedule, and Reference 2.42 to determine the inner diameter of the pipe. From the length and diameter, the volume can then be calculated. As can be seen from the calculations listed below, the volume of pipe between CV-3056 and the SIRW Tank is 23.678 ft 3

The total volume of the SIRW Tank and recirculation line leading to it is then 23.678 + 40151.103 = 40174.8 ft 3

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PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 18 Rev # -~o __

For the water volume after RAS, the volume of pipe between CV-3056 and the SIRW Tank is added to the volume of water in the SIRW Tank at RAS. The low level transfer setting in the SIRW Tank for RAS is 24" [Ref. 2.19]. The volume of water in the tank at 24", using the 46' diameter, is 1t r 2L =

3323.805 ft3. Adding the volume of water in the tank to the volume of the pipe between CV-3056 and the tank results in 3323.805 + 23.678 = 3347.5 ft3

The last value on the seventh line of the input deck is the fraction of the iodine in the sump water reaching the SIRW Tank that is in volatile form.

From Reference 2.51, which is included as to this analysis, the fraction of iodine leaking into the SIR W tank that would be in volatile form is 3.0E-04 since the sump pH will be controlled above 7. This fraction is based upon radiolytic conversion, which would dominate the amount of volatile iodine, as 12, released in the containment according to Reference 2.41 [pg. 35].

Line 8 of the input deck contains the appropriate breathing rates for 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 1 to 30 days. From Reference 2.1, these values are 3.47E-04 m3 /sec, 1.75E-04 m3 /sec, and 2.32E-04 m

3 /sec, respectively.

The 0 to 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary atmospheric dispersion factor is listed on line 9 of the input deck. This value is 1.55E-04 sec/m3, choosing the maximum value for 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from Reference 2.10 [Table 2-17]. The low population zone atmospheric dispersion factors for 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 1 to 4 days, and 4 to 30 days are listed on the tenth line. Also from Reference 2.10 [Table 2-18], the low population zone atmospheric dispersion factors are 1.09E-05 sec/m3, 6.94E-06 sec/m3, 2.58E-06 sec/m3

, and 6.25E-07 sec/m3, respectively.

The spray removal coefficients for particulate iodine are specified on line 11. Two coefficients are used in accordance with SRP 6.5.2 [Ref. 2.5]: an initial removal coefficient and a long term removal coefficient for particulate iodine. The code automatically changes to the long term removal coefficient after 98 % of the particulate iodine has been removed from the containment atmosphere. These values are 4.43 ht1 and 0.443 hr-1 from Reference 2.20.

Lines 12 through 16 of the input deck are for specifying times at which spray removal of elemental iodine starts or changes, and for the corresponding removal coefficients. Orily one value for elemental iodine spray removal is used in accordance with SRP 6.5.2 [Ref. 2.5]. The other lines for specifying values were added to the code for versatility. Full containment spray flow is conservatively assumed to be achieved at 1 minute accounting for diesel generator sequencing and flow delivery, but actual full spray flow would be achieved in less than 1 minute for any combination of spray pumps [Ref. 2.35].

Therefore the first value on line 12 is 1 minute. The corresponding removal coefficient for elemental iodine on line 12 is 21.3 ht1 from Reference 2.20. Lines 13 through 16 are specified as 0 since only one removal coefficient for elemental iodine is used.

The maximum iodine decontamination factor for elemental iodine and the time at which containment sprays are terminated are on line 17 of the input deck. The maximum iodine decontamination factor is calculated using the following equation from SRP 6.5.2 [Ref. 2.5]:

where PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET VH DF= 1+-

8 Vc

= containment sump liquid volume, ft 3

= containment net free volume less V5, ft 3

= effective iodine partition coefficient.

EA-A-NL-92-012-03 Sheet 19 Rev # --=-o __

(2)

The iodine partition coefficient is a function of sump solution temperature, iodine concentration, and pH and is obtained from Reference 2.44. The pH of the sump is assumed to be controlled above 7.

For the iodine concentration, the SRP 15.6.5 [Ref. 2.5] assumption of 50 % of the core iodine inventory being released to the sump is used. From Reference 2.21, which gives the iodine source term as 50 %

of the total release, the total iodine amount in the sump would be 0.03062 g-at./MW0 which is 79.018 g-at for 2580.6 MWi-Dividing by the sump volume, 40324.6 ft 3 or 1141865.6 L, results in 6.92E-05 g-at./L. The partition coefficient seems to increase with increasing temperature according to Figure 6 of Reference 2.44, so the lowest sump temperature given in Reference 2.14 is evaluated, 132°F or 328.6°K Considering the temperature and iodine concentration, a point between the two plots on Figure 6 of Reference 2.44 would be appropriate. For conservatism, a value of 3.1 is chosen from the figure, which corresponds to a partition coefficient of -1250 since the figure is on logarithmic scale.

Inserting this partition coefficient into Equation (2), along with the sump volume of 40324.6 ft3 and the containment net free air volume of 1.64E+06 ft 3 [Ref. 2.10], results in a maximum decontamination factor of 32.51.

The time at which containment sprays are terminated, on line 17, is assumed to be 1440 minutes (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). The emergency operating procedures [Ref. 2.36] instruct the use of containment sprays for reducing containment pressure and for containment atmosphere iodine removal. By 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the containment pressure would be below the criteria for terminating sprays, 3.0 psig, and it is conservatively assumed that the need for continued spray removal of iodine is not recognized. Since the maximum decontamination factor is reached before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , terminating containment sprays stops removal of particulate iodine from the containment atmosphere.

The times at which ESF leakage into the safeguards rooms begins and/ or changes, the corresponding ESF leak rates, the times at which sump water leakage into the SIR W Tank begins and/ or changes, and the corresponding leak rates through CV-3027 or CV-3056 are listed on lines 18 through 21. On line 18, the time at which ESF leakage starts is 19 minutes, which is the minimum time to RAS. The corresponding Technical Specification leak rate is 0.2 gpm [Ref. 2.12, TS 4.5.3] on line 19. The time, on line 18, for leakage into the SIRW Tank to begin is also 19 minutes. The corresponding leak rate through CV-3026 & CV-3057 is 0.1 gpm [Ref. 2.15] on line 18. All values on lines 19 through 21 of the input deck are specified as 0 since the ESF and SIR W Tank leakage is assumed to remain constant after RAS throughout the incident. It should be noted that the MHACALC code automatically multiplies the ESF leak rate by a factor of 2 to remain consistent with the guidance of SRP 15.6.5 Appendix B

[Ref. 2.5].

On lines 22 through 39 of the input deck are the activity source term, the radioactive decay constant, the thyroid dose or dose rate conversion factor, and the whole body dose or dose rate conversion factor for each radionuclide. The radionuclides are specified in the same order as listed in Table 1. For the noble gas isotopes, all four values are listed in Table 1. For iodine, the activity source term and the

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 20 Rev #

0 radioactive decay constants are listed in Table 1, and the thyroid and whole body dose conversion factors are obtained from Table 2.

The number of time intervals for which release rates are to be calculated and written to output files is on line 40. For this analysis 10 time intervals are used for the release rates. These time intervals are specified on lines 41 and 42. The times for the intervals to change were chosen such that a new time interval begins at each point in time that the release rates could change by a significant amount, plus a few time points in between. The major points in time to consider for the release rates changing are:

the time at which full spray is assumed to be achieved, 1.0 minute; the time at which the control room achieves pressurization in the emergency mode, 0.5 + 67.5 + 7.34 = 75.34 seconds or 1.26 minutes from Reference 2.26 & Attachment 1 of Reference 2.8; the approximate time at which the spray removal effectiveness for elemental iodine ends, 12.0 minutes; the time ESF leakage and leakage into the SIRW Tank begin, 19.0 minutes; the first time at which atmospheric dispersion factors and other parameters change, 480.0 minutes; the time at which containment sprays are assumed to be terminated and the containment atmosphere leak rate changes, 1440.0 minutes; and the last time at which atmospheric dispersion factors and other parameters change, 5760.0 minutes. Several other somewhat arbitrary time points are also specified.

Line 43 of the input deck contains the number of points in time that the containment and SIRW Tank activities of each of the radionuclides are to be printed in the output file. The values of the total iodine activity in the SIRW Tank are to be used for calculating the shine dose to the control room. The number of times to be printed is specified as 23. Lines 44 through 46 of the input deck are the corresponding time points for the activities to be printed in the output file. These times were somewhat arbitrarily chosen as 19, 30, 60, 120, 240, 480, 720, 1440, 2880, 4320, 5760, 8540, 10080, 11520, 12960, 14400, 17280, 20160, 23040, 25920, 28800, 36000, and 43200 minutes.

The input deck constructed with the above listed parameters is listed on the attached microfiche under the filename 1994MHA DATA.

6.2 OFFSITE DOSES The MHACALC code, as described in Reference 2.16, was executed using the 1994MHA DATA input deck in VM8. Four output files from the program execution are listed on the attached microfiche:

1994MHA LISTING, STAK_MHA DATA, SIRW_MHA DATA, and PLOT_MHA DATA The 1994MHA LISTING file contains the containment atmosphere, sump, and SIRW Tank activities at the input specified points in time and the offsite doses from the incident. The STAK_MHA DATA file contains the radionuclide release rates from the containment atmosphere and ESF leakage for the input specified time intervals. The SIRW_MHA DATA file contains the radionuclide release rates from the SIRW Tank for the input specified time intervals. The fourth output file, PLOT_MHA DATA, contains the dose equivalent 1-131 activities at numerous points in time and is set up for direct input to the RETRAN code PLOTER module for creating plots of activity versus time. This file was executed in

the RETRAN code PLOTER module, with two resultant plots given as Attachment 2 to this analysis.

The resultant doses at the site boundary (SB) and low population zone (LPZ) from the incident are obtained from the 1994MHA LISTING file.

Since the NRC has not accepted the full ICRP30 methodology for calculating doses from design basis accidents, the doses taken from the listing file are

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 21 Rev #

O the thyroid doses from inhalation and the whole body doses. The other listed doses incorporate the ICRP30 methodology of adding contributions for internal and external doses to obtain dose equivalent values. The resultant offsite doses from the incident from each release path and the total, as obtained from the MHACALC output, are listed in Table 4 below. As can be seen, these doses are well within the 300 rem thyroid and 25 rem whole body limits of 10 CFR 100.

TABLE 4 SITE BOUNDARY AND LOW POPULATION ZONE DOSES Ctmt ESF SIRW Total Leakage Leakage Release Release (Rem)

(Rem)

(Rem) 0-2 Hr SB Thyroid 11.869 4.039

<0.001 15.908 Whole Body

o.287 N/A N/A 0.287 0-30 Day LPZ Thyroid 5.391 4.111

<0.001 9.501 Whole Body 0.053 N/A N/A 0.053 6.3 CONDOSE INPUT The CONDOSE code requires two input decks for each execution. The first input deck contains all of the physical parameters of the control room and radionuclides and the second input deck contains only the radionuclide release rates. Input decks for the radionuclide release rates are given as output from the MHACALC code: STAK_MHA DATA and SIRW_MHA DATA for the containment plus ESF leakage release and SIRW Tank release, respectively. The input deck with the physical parameters for the CONDOSE code is described by line below. A full description of the required inputs and the structure for the input deck can be obtained from Reference 2.33. Separate input decks are constructed for the release from the containment plus ESF leakage (assumed release point is the location of the stack, ignoring stack height) and the release from the SIR W Tank since different atmospheric dispersion factors are used for each.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 22 Rev #

0 The first line of the input deck is simply the title or case description. The second line of the input deck is the air volume of the control room envelope. This volume includes the control room, the viewing gallery, and the technical support center since the control room habitability system services all of those areas. The dimensions of these areas were obtained from References 2.23, 2.24, & 2.25. Room heights were taken from the floor to the drop ceiling, conservatively ignoring the space above the drop ceiling since smaller volume increases concentration of radionuclides in the control room. The diffiensions of the control room are 46'x 48'x 10' which yields a volume of 22080 ft3. The viewing gallery and adjacent offices are 23.5'x 12.5'x 9' and 22'x 14.5'x 9' not including the stairway to the mechanical equipment room, which results in a volume of 5514.75 ft3

The technical support center and office area dimensions are 25'x 22.5'x 7.5' and 37'x 23'x 7.5' from which the volume of an electrical chase, 9'x 3'x 7.5', and a HV AC chase, 6'x 4'x 7.5', must be subtracted resulting in a volume of 10218.75 ft 3

Summing the volumes from the three areas and assuming 5 % of the air volume is occupied by equipment and walls yields a total volume of 35923 ft 3

This volume, which converts to 1017 m3, also justifies the use of noble gas submersion dose rate conversion factors corrected for a 1000 m 3 room as given in Table 2.

The third line of the input deck is the number of breathing rates to be used, which is 3 following the guidance of Reference 2.1. Lines 4 through 6 contain the times at which each breathing rate starts followed by the corresponding breathing rates. From Reference 2.1 these values are as follows: at 0.0 minutes the breathing rate is 3.47E-04 m3 /sec, at 480.0 minutes the breathing rate changes to 1.75E-04 m3 /sec, and at 1440.0 minutes the breathing rate changes to 2.32E-04 m3 /sec at which it remains throughout the 30 days.

The number of control room occupancy factors to be used is listed on line 7 of the input deck. In accordance with SRP 6.4 [Ref. 2.5, Table 6.4-1 ], 3 occupancy factors are used. The times at which each of the occupancy factors start and the corresponding occupancy factors are on lines 8 through 10 of the input deck. To start, 0.0 minutes, the occupancy factor is 1.0. At 1440.0 minutes the occupancy factor changes to 0.6, and at 5760.0 minutes the occupancy factor changes to 0.4.

Line 11 of the input deck contains the number of atmospheric dispersion factors that are to be used.

As can be seen in SRP 6.4 [Ref. 2.5, Table 6.4-1 ], it is standard practice to use 4 atmospheric dispersion factors. Lines 12 through 15 of the input deck contain the times that each atmospheric dispersion factor starts and the corresponding atmospheric dispersion factors for the normal air intake, the remote air intake, and a location of unfiltered air inleakage, respectively. All of these atmospheric dispersion factors must correspond to the same point of release of radionuclides. The values used are listed in Table 3. During the time that the control room is depressurized after loss of offsite power, air inleakage into the control room is assumed to occur. The closest point for source term inleakage is assumed to be through the aux bay roof doors at the end of the hallway from the control room entrance

[Ref. 2.30]. The input deck used for the release from the containment and ESF leakage has the atmospheric dispersion factors from the stack to the control room air intake locations as listed in Table 3, since the release point is assumed to be the location of the stack. The input deck used for the release from the SIRW Tank has the atmospheric dispersion factors from the SIRW Tank to the control room air intakes as listed in Table 3, with the factors for the unfiltered air inleakage location set to 0.0 since the leakage through CV-3027 & CV-3056 does not begin until 19 minutes and the control room is pressurized by then.

The number of points in time for which the control room HV AC system flow rates are to be specified is on line 16 of the input deck. For this analysis, 2 points in time are used: the initial flow rates while

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 23 Rev # _....:,o __

the control room is depressurized due to loss of offsite power, and the flow rates once pressurization in the emergency mode is achieved since a LOCA would generate a CHP (containment high pressure) signal which in tum would automatically switch the control room HV AC to emergency mode.

Lines 17 and 18 of the input deck contain the times that the control room HV AC system flow rates are specified for and the corresponding flow rates for the normal air intake, the emergency air intake, the recirculation air, and an unfiltered air inleakage flow rate, respectively. At time 0.0 minutes, on line 17, the flow rates for the normal intake, the remote intake, and the recirculation air are 0.0 since the

, control room is assumed to be depressurized due to the loss of offsite power. The rate of unfiltered air inleakage while the control room is depressurized is based upon the gross air leakage from the control room when pressurized to Va" water gauge, as recommended by SRP 6.4 [Ref. 2.5]. To maintain steady state pressurization, the leakage from the control room must equal the make-up to the control room, which is 1100 cfm: One half this value, 550.0 cfm, is assumed to be the base infiltration rate following SRP 6.4.

The time on line 18 of the input deck is the time at which pressurization in the emergency mode is achieved, 1.26 minutes. This time corresponds to a 0.5 second CHP signal generation delay followed by a 67.5 second delay for starting the diesel generators and sequencing fans V-95 and V-96 [Ref. 2.26],

and a delay of 7.34 seconds for pressurizing the control room to~ Va" water gauge (from Attachment 1 of Reference 2.8). The make-up and recirculation air flow rates for the emergency mode of the control room HV AC are based on the total system flow rate of 3200.0 cfm + 10 %, or 3520 cfm, since the test procedure [Ref. 2.50] allows a tolerance of +/- 10 % for the air flow rate when in emergency mode. Assuming the air flow rate is + 10 % is conservative since more make-up air is brought into the control room. Since tile make-up air flow rate is 1000 cfm at system flow rate of 3200 cfm [Ref. 2.27],

at 3520 cfm the make-up air flow rate is 1100 cfm and the recirculation air flow rate is the remainder, assuming that the make-up and recirculation air flow rates are always proportional. The corresponding flow rates on line 18 are then 0.0 cfm for the isolated normal intake, 1100.0 cfm make-up air flow from remote intake, 2420.0 cfm recirculation air flow rate, and 0.0 for unfiltered air inleakage from other sources.

Line 19 of the input deck is the number of points in time for which filter efficiencies have been specified. Since the filters are required by Technical Specifications to remain ~ 99 % efficient for iodine [Ref. 2.12], only 1 time is specified. Line 20 is the start time of the efficiencies, 0.0 minutes, and three corresponding filter efficiencies for the normal intake followed by three filter efficiencies for the remote intake make-up air and recirculation air. The three filter efficiencies for each air intake are for the three possible types of radionuclides: noble gas, halogens (iodine), and solids, respectively. For the normal intake, since no filter exists for the current plant configuration, all three efficiencies are 0.000.

For the remote intake make-up air and recirculation air, the filter efficiencies are 0.000 for noble gas, 0.990 for iodine, and 0.990 for solids or particulates. The efficiency for noble gas is 0.000 since it will simply pass through charcoal and HEP A filters.

On line 21 of the input deck is the number of points in time that the radionuclide concentrations in the control room and the accumulated operator doses are to be printed in the output file. This value is specified as 0 for this analysis since only the total dose in the control room over the 30 day period is of interest.

~-

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 24 Rev # _...;:.o __

Line 22 of the input deck is the number of time intervals for which the radionuclide release rates will be given in the release rate input deck. This value is 10, corresponding to the value on line 40 of the input deck for the MHACALC code. Lines 23 and 24 are the beginning and end times of the time intervals. These lines correspond to the values on lines 41 and 42 of the 1994MHA DATA input deck for the MHACALC code. These values are 0.00, 1.00, 1.26, 12.00, 19.00, 480.00, 1440.00, 5760.00, 14400.00, 28800.00, and 43200.00.

Line 25 of the input deck contains the number of radionuclides for which release rates are specified, and an identifier that specifies the units in which the radionuclide release rates are given. The number of radionuclides is 18 since that is the number of radionuclides that the MHACALC code considers and provides release rates for. The identifier that specifies the units of the radionuclide release rates is 1, corresponding to µCi/Hr.

The total number of radionuclides to be considered, in case daughter products are to be considered that were not given release rates~ is given on line 26. For this analysis, only the radionuclides for which release rates are given are considered, so the value on line 26 is 18.

Lines 27 through 98 of the input decks contain the radionuclide constant data. The radionuclide data must be given in the same order that the radionuclides are given for the release rates, which is the same order as listed in Table 1. For each radionuclide, there are four lines of data. The first line for each radionuclide contains the radionuclide name followed by an identifier for the radionuclide given for the consecutive order that they are specified in, starting with 1. The second line for each radionuclide contains the inhalation dose conversion factors from Table 2 for the thyroid, lungs, bone surface, bone marrow, skin, eye lens, and whole body, respectively. Since noble gas does not result in an inhalation dose, all of these values are 0.0 for the noble gas isotopes, on lines 27 through 78. For the iodine isotopes on lines 79 through 98, these values are listed in Table 2. The third line for each radionuclide contains dose rate conversion factor for submersion in a radioactive cloud. For the noble gas isotopes, these values are listed in Table 2. For the iodine isotopes, since they do not result in a significant submersion dose per ICRP30, these values are specified as 0.0. The fourth line for each radionuclide contains an identifier for the form of the radionuclide, the radioactive decay constant, the identifier of the primary daughter product if applicable, and the production fraction for the daughter product if applicable. The identifier for the form of each radionuclide is 1 for noble gas and 2 for iodine [Ref.

2.33]. The radioactive decay constants for each radionuclide are listed in Table 1. The primary daughter product identifier and production factor are specified as 0 since daughter products are not being considered.

The two input decks constructed with the above listed parameters are listed on the attached microfiche under the filenames CRl_MHA DATA for the containment plus ESF leakage release, and CR2_MHA DATA for the SIRW Tank release.

6.4 CONTROL ROOM DOSES The CONDOSE code, as described in Reference 2.33, was executed once for each of the two release paths. The first execution was with the CRl_MHA DATA and STAK_MHA DATA input decks, which resulted in the CRl_MHA LISTING file with the control room operator doses from the containment atmosphere and ESF leakage. The second execution was with the CR2_MHA DATA and SIRW_MHA

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 25 Rev #

0 DATA input decks, which resulted in the CR2_MHA LISTING file with the control room operator doses from the SIRW Tank release. The listing files are also listed on the attached microfiche. The output files provided by the CONDOSE code for plotting committed effective dose equivalent to the thyroid and total effective dose equivalent to the control room operators versus time were discarded since they were of no interest for this analysis.

As with the listing file from the MHACALC code, the doses on the listing files from the CONDOSE code executions must be taken from that listed for inhalation for the thyroid dose and from that listed for submersion for the whole body and skin doses.* Since all of the doses are calculated using the methodology of ICRP30, the other values combine contributions from internal and external dose which is not required for accident analyses. The resultant control room operator doses due to containment atmosphere and ESF leakage, as can be seen in CRl_MHA LISTING, are 9.052 rem thyroid, 6.014 rem skin, and 0.253 rem whole body. The shine dose from containment must then be added to the whole body dose. Most of the containment shine dose to the control room is from the purge lines. This shine dose was calculated to be 0.400 rem whole body in Reference 2.45 [pg. 20]. However, this value is conservative since it is taken at the access doors to the control room viewing gallery, which is not actually in the control room. This shine dose was also conservatively calculated ignoring control room occupancy factors.

As can be seen in CR2_MHA LISTING, the control room operator doses due to releases from the SIRW Tank are less than 0.0001 rem thyroid, skin, and whole body, remaining negligible. These doses are very low since most of the iodine is retained in the large volume of water in the SIR W Tank.

However, since the tank is on the control room roof, there is a resultant shine dose to the operators.

The SIRW Tank shine dose is calculated using the PC based MICROSHIELD code [Ref. 2.47]. The MICROSHIELD code provides the dose rate, in mrem/hour, for the inputs specified. Inputs for the code are specified interactively. The major inputs for the MICROSHIELD code are the type and geometry of the source and shields, the activities of the radionuclides present in the source, dimensions of the source, the thicknesses of materials between the source and the receptor point, the material composition and density of the source and shields, the buildup factor method to be used, and the material for which the buildup factor calculations will be based. The activities of the radionuclides present in the SIRW Tank air and water volumes and in the piping between CV-3056 and the SIRW Tank are obtained from the output of the MHACALC code. For the shine dose calculations, it is assumed that all of the* activity (in the piping, tank's air volume, and tank's water volume) is in the water volume of the SIR W Tank and evenly distributed.

First, the geometry to be used to model the source of shine must be specified. The MICROSHIELD geometry used is a cylindrical source with shine from the end and slab shields. The dimensions of the source and shielding materials then need to be specified, followed by a case title. From Reference 2.17 it can be seen that there is al" minimum of sand under the tank, followed by a 10" concrete slab on the roof above the control room. As can be seen on Reference 2.49, the control room roof is a cellular design consisting of 1' of concrete followed by a 4' air space, followed by another l' of concrete. The 10" concrete slab and the 1' concrete top of the cellular roof are assumed to be a single shield. Thus, the shields between the tank and the control room are l" of sand, 22" of concrete, 48" of air, and 12" of concrete. As can be seen from Reference 2.24, the control room has a total height of 12', ignoring the drop ceiling. Assuming the average operator has a height of 6' leaves 6' of air space between the control room ceiling and the receptor point. The MICROSHIELD code accommodates up to five shields, including the source if it is a volume source, and will automatically insert an air space if the

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 26 Rev # -~o __

distance to the receptor is greater than the sum of the shield thicknesses. The SIRW Tank is considered a self-shielding source since the activity is assumed to be distributed in the 24" of water left in the tank after RAS. Including the water height in the SIRW Tank, and conservatively ignoring the rising water level as valve leakage enters the tank, the total distance between the top of the source and the receptor point is 179". The model used for the MICROSIIlELD geometry inputs is shown in Figure 1. The shields are labeled as they are on the interactive screen for the MICROSHIELD input.

FIGURE 1 SIRW TANK SHINE MODEL USED IN MICROSHIELD Cyllndrlcal Source (SIRW Tank)

Tl (Water) = 24' T3 (Concrete)

22"

~

x.. 179' T5 (Concrete)

12' Receptor Point After specifying the geometry and dimensions of the source and shields, the method for buildup factor calculations, the material on which the buildup calculations are based, and the point-kernel integration parameters must be specified. The Taylor method was chosen, and the water volume of the SIR W Tank was used as the material basis for the buildup calculations. The water was chosen since it is the dominant shield and results in the highest buildup. This can be proven by simply executing the same

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet

~7 Rev #

0 case choosing a different shield for the material basis of the buildup calculations each time and examining the results. For the point-kernel integration parameters, the number of angular and radial segments were both chosen as 11. These parameters were chosen by starting with the default values and increasing the values, executing the code each time, until the results stopped changing which indicated the closest convergence. Reference 2.47 [pg. 5-6] also states that specifying the integration parameters as 11 should be sufficient for most cases.

The MICROSHIELD code was executed for each of the times at which the activity in the SIRW Tank was listed in 1994MHA LISTING. For each of the executions of the MICROSHIELD code, all parameters except the iodine activity and the case title were left the same. During the executions of the MICROSHIELD code, some problems were discovered with the manner in which the code sorts the photon energy groups. As long as each of the iodine isotopes had an activity listed, the photon energy groups chosen by the code remained fairly similar. However, when some of the iodine isotopes were specified to have no activity present, MICROSHIELD sorted the photons into completely different energy groups giving peculiar.results. This problem was noticed when the activity of 1-132 was specified as 0.0. When the activity of 1-132 is specified as 0.0, the resultant dose rate given by the code increases slightly. Only one explanation could be surmised: since 1-132 has so many different photon energies (approximately 69), it biased the photon energy grouping performed by the code. Drastically changing the energy grouping then changes the calculated dose rates by a small amount. For the five iodine isotopes used in the MICROSHIELD executions for this analysis, the dose rate change due to the energy grouping change was at most seen to be 0.002 mrem/horir, which would not introduce much error in the results. However, for consistency between MICROSHIELD executions, it is desirable to keep the photon energy groups fairly consistent.

To keep the photon energy groups chosen by MICROSHIELD for each execution fairly consistent, the activity of each isotope is specified as 1.0E-60 Curies when the 1994MHA LISTING gives it as 0.0.

Specifying the activity of an isotope as 1.0E-60 Curies rather than 0.0 will not contribute any recognizable amount to the calculated dose rates for this analysis, but will keep the photon energy groups fairly consistent. To demonstrate that no recognizable amount is added to the calculated dose rate by specifying 1.0E-60 rather than 0.0, a test case was executed with the activity of each of the iodine isotopes specified as 1.0E-60 Curies, and all other parameters specified as discussed previously. This test case had a calculated dose rate of 4.996E-64 mrem/hour, as shown on the printout in pages 1 and 2 of Attachment 3. This demonstrates that specifying an activity as 1.0E-60 Curies rather than 0.0 does not introduce any recognizable change in the results of this analysis.

The printouts of the MICROSHIELD executions for each of the times that SIRW Tank activities are provided in 1994MHA LISTING are listed in Attachment 3, with the results of each execution occupying two pages. Examining the results of each execution, the dose rate in the control room can be seen to increase for almost the first 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months due to the increasing iodine concentration in the SIR W Tank. The dose rate then decreases after 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months due to decay of the high energy gamma emitters that have short half-lives. Since the dose rate only increases with time for the first 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and decreases with time for the following 708 hours0.00819 days 0.197 hours 0.00117 weeks 2.69394e-4 months , the dose rate calculated at each time point is assumed to be constant until the next time point at which the dose rate is calculated. For the time from 19 minutes, when leakage into the SIR W Tank began, to 60 minutes, the calculated dose rate at 30 minutes is used since no earlier time point was specified in the_ MHACALC code for the SIRW Tank activity to be listed. The calculation of the dose received, using the results of the MICROSHIELD executions, is shown below. The dose rate is multiplied by the time interval, an appropriate units conversion factor,

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 28 Rev #

0

-~-

and the control room occupancy factor to obtain the total dose received by a control room operator over a time interval. The results are then summed as shown.

Time Interval 19...

60 min 60...

120 min 120...

240 min 240...

480 min 480...

720 min 720...

1440 min 1440...

2880 min 2880...

4320 min 4320...

5760 min 5760...

8640 min 8640... 10080 min 10080... 11520 min 11520... 12960 min 12960... 14400 min 14400... 17280 min 17280... 20160 min 20160... 23040 min 23040... 25920 min 25920... 28800 min 28800... 36000 min 36000... 43200 min Cmrem) 0.1025(1.0)(60-19)/60 =

0.070 0.3263(1.0)(120-60)/60 =

0.326 0.6221(1.0)(240-120)/60 =

1.244 0.9464(1.0)(480-240)/60 =

3.786 1.188(1.0)(720-480)/60 =

4.752 1.167(1.0)(1440-720)/60 =

14.004 0.700(0.6)(2880-1440)/60 =

10.080 0.1705(0.6)(4320-2880)/60 =

2.455 0.06414(0.6)(5760-4320)/60 =

0.924 0.03833(0.4)(8640-5760)/60 =

0.736 0.02223(0.4)(10080-8640)/60 = 0.213 0.01955(0.4)(11520-10080)/60 = 0.188 0.01833(0.4)(12960-11520)/60 = 0.176 0.01782(0.4)(14400-12960)/60 = 0.171 0.01762(0.4)(17280-14400)/60 = 0.338 0.01739(0.4)(20160-17280)/60 = 0.334 0.01698(0.4)(23040-20160)/60 = 0.326 0.01631(0.4)(25920-23040)/60 = 0.313 0.01544(0.4)(28800-25920)/60 = 0.296 0.01444(0.4)(36000-28800)/60 = 0.693 0.01173(0.4)(43200-36000)/60 = 0.563 r = 41. 988 mrem As can be seen, the total shine dose in the control room from the SIR W Tank over the 30 days is 41.988 mrem, or 0.042 rem. Adding this shine dose to the whole body dose from the SIRW Tank release and the containment and safeguards room release, and including the shine from containment results in a total whole body dose in the control room of 0.042 + 0.000 + 0.253 + 0.400 = 0.695 rem. As mentioned earlier, the thyroid dose for the containment and safeguards room release is 9.052 rem and the skin dose is 6.014 rem, with the release from SIRW Tank resulting in negligible thyroid and skin doses. These values are well within the 30 rem thyroid, 30 rem skin, and 5 rem whole body dose limits of GDC 19 as interpreted in SRP 6.4.

7.0

SUMMARY

This analysis was performed to demonstrate that the radiological consequences of the MHA would be within the limits of 10 CFR 100 and 10 CFR 50 Appendix A for the proposed modifications to the plant expected to occur by Cycle 12. The two major forecasted plant modifications affecting the radiological consequences of the MHA are changes to the current method of post-LOCA sump pH control and changes to the control room HV AC system. The post-LOCA sump pH control system is expected to be modified such that a sump solution pH ~ 7 is reached by RAS and maintained throughout the event.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-03 Sheet 29 Rev # --=-o __

The control room HV AC system is expected to be modified in an attempt to eliminate unfiltered air inleakage while operating in the emergency mode.

For the offsite doses and radionuclide releases, the MHACALC code was used. The Regulatory Guide 1.4 and Standard Review Plan 15.6.5 source terms were used for 102 % power operation. Standard Review Plan 15.6.5 was also used as the basis for the iodine chemical forms in the containment atmosphere, the containment leak rate, and the assumptions regarding ESF leakage. The spray removal coefficients for iodine and the maximum decontamination factor for elemental iodine were calculated based on Standard Reveiw Plan 6.5.2, Revision 2. Sump solution leakage of 0.1 gpm through CV-3027

& CV-3056 into the SIRW Tank was also accounted for after RAS. For iodine release from the SIRW Tank, 0.03 % of the iodine was assumed to be in volatile form, and a partition factor based on an assumed air-water interface temperature of 100°F was used to account for the amount of volatile iodine that would become airborne. Airborne iodine in the SIR W Tank was then assumed to exit through the tank vent at twice the rate at which air was displaced from the tank from the water leaking in.

Radionuclide release rates are calculated by the MHACALC code for use in control room habitability calculations with separate files created for the containment plus ESF release rates and for the SIR W Tank release rates. The resultant doses from the event were calculated at the site boundary and the low population zone using the breathing rates specified in Regulatory Guide 1.4 and the dose conversion factors of ICRP30. The total calculated doses are 15.908 rem thyroid and 0.287 rem whole body at the site boundary, and 9.501 rem thyroid and 0.053 rem whole body at the low population zone. These results are shown in Table 5 below. The calculated values are well within the 300 rem thyroid and 25 rem whole body dose limits of 10 CFR 100.

For the control room doses, the CONDOSE code was used to model the radionuclide transport into and out of the control room and to calculate the operator doses from exposure to the radionuclides.

The control room was assumed to be initially depressurized due to loss of offsite power, with a base air infiltration rate in the control room that was calculated in accordance with Standard Review Plan 6.4.

After accounting for diesel generator sequencing and control room pressurization time, the maximum emergency mode fresh air make-up flow rate was assumed for the control room HV AC system, taking credit for the charcoal filters being 99.0 % efficient. Zero unfiltered air inleakage to the control room was assumed when pressurized in the emergency mode of the HV AC system since modifications are expected to eliminate the inleakage. The resultant operator doses due to inhalation of and exposure to the radionuclides in the control room were calculated using the breathing rates specified in Regulatory Guide 1.4, the occupancy factors of Standard Review Plan 6.4, and the dose conversion factors of ICRP30. Two executions of the CONDOSE code were made to account for the difference between the atmospheric dispersion factors for radionuclide release from the SIR W Tank and from the containment plus safeguards rooms, using the appropriate radionuclide release rates provided by the MHACALC code. The calculated operator doses from exposure to radionuclides in the control room are 9.052 rem thyroid, 6.014 rem skin, and 0.253 rem whole body. The whole body dose contribution from containment shine of 0.400 rem, as calculated in the NUREG-0578 design review study of plant shielding, was then added to the operator dose. Shine dose from the SIRW Tank, which is on the roof above the control room, was then accounted for using the MICROSHIELD code. The whole body dose contribution from the SIRW Tank shine was calculated to be 0.042 rem. Therefore, the total calculated doses to the control room operators from the event are 9.052 rem thyroid, 6.014 rem skin, and 0.695 rem whole body. These results are also shown in Table 5. These calculated values are also well within the 30 rem thyroid, 30 rem skin, and 5 rem whole body dose limits of 10 CFR 50 Appendix A as interpreted by the Standard Review Plan.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET TABLE 5 EA-A-NL-92-012-03 Sheet 30 Rev # --=-o __

SUMMARY

OF TOTAL DOSES FROM THE MHA I

I I

I THYROID I

SKIN I WHOLE BODY I

I I

I Location I

(Rem)

I (Rem)

I I

I SB (0-2 Hr) 15.908 N/A 0.287 LPZ 9.501 N/A 0.053 Control Room 9.052 6.014 0.695

8.0 CONCLUSION

The resultant doses offsite and in the control room from the MHA are well within the limits of 10 CFR 100 (300 rem thyroid and 25 rem whole body) and 10 CFR 50 Appendix A as interpreted in the Standard Review Plan (30 rem thyroid, 30 rem skin, and 5 rem whole body). This assures that the expected modifications will bring the plant into conformance with all MHA dose limits if completed as planned.

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET 9.0 LIST OF ATTACHMENTS EA-A-NL-92-012-03 Sheet 31 Rev # _....:.O __

1.

Letter from E.C. Beahm (Martin Marietta Energy Systems, Inc.) to Jay Y. Lee (NRC) dated February 5, 1992, 3 pages.

2.

Plots of containment atmosphere dose equivalent 1-131 activity versus time, 2 pages.

3.

Printouts of results from execution of the MICROSHIELD code, 46 pages.

4.

Design Review Forms Including:

Form 3698 9-89, Engineering Analysis Checklist, 1 page.

Proc. No. 9.11 Attachment 5, Technical Review Checklist, 1 page.

Form 3110 1-82, NOD Document Review Sheet, __2_ pages.

5.

Microfiche containing input and output from MHACALC and CONDOSE codes, 1 fiche.

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ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 37 Rev # 0 Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File Run date:

SIRWTEST.MSH March 16, 1992 3:58 p.m.

Date: _!_!_

Run time:

By:

Checked:

CASE: TEST CASE FOR MINISCULE ACTIVITY OF EACH IODINE ISOTOPE GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector *****************.**.**** X Source cylinder radius ******..*.********.**** R Source cylinder length..***...*.************* Tl Thickness of second shield *****.*.******..*** T2 Thickness of third shield *...************.*** T3 Thickness of fourth shield **..***********.*** T4 Thickness of fifth shield.*...**********..*** T5 Microshield inserted air gap ***************** air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.*

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 38 Rev # 0 Page 2 File: SIRWTEST.MSH CASE: TEST CASE FOR MINISCULE ACTIVITY OF EACH IODINE ISOTOPE BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.0000e-60 I-132 1.0000e-60 I-133 1.0000e-60 I-134 1.0000e-60 I-135 1.0000e-60 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.1181

1. 6189 1.1659

.8490

.6581

.5192

.3715

.2778

.2132

.1355 TOTALS:

2.469e-51 2.271e-50 4.308e-50 1.142e-49 6.811e-50 5.223e-50 3.575e-50 5.390e-51

1. 230e-51

1. 867e-51 3.471e-49 7.655e-62

1. 567e-61 3.613e-62

1. 731e-62 2.017e-63 3.701e-64

1. 527e-65

1. 441e-67 2.414e-69 9.737e-72 2.891e-61

1. 221e-64 2.708e-64 6.733e-65 3.445e-65 4.188e-66 7.587e-67 3.140e-68 2.865e-70 4.526e-72

1. 618e-74 4.996e-64

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037)

Page 1

File Run date: March 11, 1992 Run time: 9:34 a.m.

EA-A-NL-92-012-03 Sheet ~

Rev # 0 File Ref:

Date: ~-/~_/~

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 30 MIN GEOMETRY 10: Cylindrical source from end -

Distance to detector..**.*..*................ X Source cylinder radius.................... ~** R Source cylinder length....................... Tl Thickness of second shield................... T2 Thickness of third shield.................... T3 Thickness of fourth shield.*...*............* T4 Thickness of fifth shield *................... TS Microshield inserted air gap................. air slab shields 454.660 701. 040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet.!t.f2_ Rev # 0 Page 2 File:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 30 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide Curies I-131 1.3800e+02 I-132

1. 6880e+02 I-133 2.2260e+02 I-134 1.9830e+02 I-135 2.2oooe+o2 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.1290 4.790e+ll 1.515e+Ol 2.411e-02 2

1.6244 4.673e+12 3.291e+Ol 5.683e-02 3

1.1658 8.992e+12 7.538e+OO 1.405e-02 4

.8509 2.157e+13 3.309e+OO 6.583e-03 5

.6577 1.186e+13 3.500e-Ol 7.268e-04 6

.5196

1. 095e+13 7.789e-02 1.597e-04 7

.3733 5.290e+12 2.368e-03 4.869e-06 8

.2766 9.411e+ll 2.406e-05 4.781e-08 9

.2150 2.524e+ll 5.338e-07 1.003e-09 10

.1354 3.677e+ll 1.893e-09 3.146e-12 11 12 13 14 15 16 17

. 18 19 20 TOTALS:

6.538e+13 5.934e+Ol

1. 025e-Ol

Page File Run date:

Run time:

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037) 1 60MI:N.MSH March 11, 1992 12:39 p.m.

EA-A-NL-92-012-03 Sheet..!fL_ Rev # O File Ref:

Date: ~-/~_/~

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 60 MIN GEOMETRY 10: Cylindrical source from end -

Distance to detector......................... X Source cylinder radius........*.*.*.......... R Source cylinder length *...*.................. Tl Thickness of second shield................... T2 Thickness of third shield....**.............. T3 Thickness of fourth shield.......*..........* T4 Thickness of fifth shield.................... T5 Microshield inserted air gap....*.*.......... air slab shields 454.660 701.040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet #2 Rev # 0 Page 2 File: 60MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 60 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide curies I-131 5.1340e+02 I-132 5.4110e+02 I-133 8.1580e+02 I-134 4.9780e+02 I-135 7.7790e+02 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.1311 1.6224 1.1702

.8487

.6587

.5204

.3724

.2786

.2164

.1355 TOTALS:

1.560e+12 1.489e+13 2.927e+13 6.038e+13 3.656e+13 3.775e+13 1.877e+13 3.182e+12 8.210e+ll 9.333e+ll 2.041e+14 4.955e+Ol

1. 041e+02 2.504e+01 9.129e+OO 1.089e+OO 2.711e-01 8.196e-03 8.803e-05

1. 836e-06 4.902e-09 1.892e+02 7.882e-02 1.798e-Ol 4.661e-02 1.817e-02 2.261e-03 5.558e-04 1.685e-05

1. 751e-07 3.455e-09 8.148e-12 3.263e-01

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet #3 Rev # O Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File Run date:

120MIN.MSH March 11,. 1992 12:44 p.m.

Date: ____ / ____ / ___ _

Run time:

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 120 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector......................... X Source cylinder radius *...*....**............ R Source cylinder length.*..................... Tl Thickness of second shield........*.......... T2 Thickness of third shield **.................. T3 Thickness of fourth shield *.................*

T4 Thickness of fifth shield.*........**........ T5 Microshield inserted air gap...******.*****.* air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet

~# Rev # 0 Page 2 File: 120MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 120 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide Curies I-131 1.2600e+03 I-132 9.8610e+02 I-133 1.9440e+03 I-134 5.5610e+02 I-135 1.7260e+03 RESULTS:

Group Energy Activity Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 (MeV)

(photons/sec) 2.1371

1. 6209 1.1766

.8444

.6603

.5214

.3711

.2815

.2183

.1359 TOTALS:

3.030e+12 2.843e+13 5.794e+l3 8.862e+13 6.361e+13 8.247e+13 4.339e+13 6.738e+12

1. 631e+12

1. 076e+12 3.769e+14 9.732e+Ol

1. 976e+02 5.099e+Ol

1. 303e+Ol

1. 924e+OO 5.996e-Ol

1. 830e-02 2.091e-04

3. 953e*-o6 5.974e-09 3.615e+02 1.546e-Ol 3.414e-Ol 9.482e-02 2.597e-02 3.994e-03

1. 230e-03 3.764e-05 4.164e-07 7.457e-09 9.939e-12 6.221e-Ol

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet.!L5._ Rev # O Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File Run date:

240MIN.MSH March 11, 1992 12:47 p.m.

Date: _/_/_

Run time:

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 240 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector **....................... X 454.660 Source cylinder radius....................... R 701. 040 60.960 2.540 Source cylinder length....................... Tl Thickness of second shield................... T2 Thickness of third shield.................... T3 55.880 Thickness of fourth shield.................*. T4 121. 920 Thickness of fifth shield.................... TS 30.480 Microshield inserted air gap.*...*........*** air 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet.!fR._ Rev # O Page 2 File: 240MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 240 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)..........*

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 2.7380e+03 I-132 1.1810e+03 I-133 3.9790e+03 I-134 2.5020e+02 I-135 3.0610e+03 RESULTS:

Group Energy Activity Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 (MeV)

(photons/sec) 2.1515

1. 6236 1.1822

.8394

.6621

.5224

.3698

.2844

.2199

.1374 TOTALS:

4.486e+12 4.330e+13 9.262e+13 8.562e+13 7.691e+13 1.536e+14 8.950e+13

1. 210e+13 2.655e+12 5.488e+ll 5.613e+14 1.480e+02 3.040e+02 8.361e+Ol

1. 219e+Ol 2.366e+OO 1.129e+OO 3.649e-02 4.219e-04 6.852e-06 3.732e-09 5.513e+02 2.345e-Ol 5.251e-Ol 1.553e-Ol 2.431e-02 4.908e-03 2.316e-03 7.505e-05 8.413e-07

1. 295e-08 6.235e-12 9.464e-Ol

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet.!L.:z._ Rev # O Microshield 3.13 (Consumers Power Company -

037) File Ref:

Page File 1

480.MSH Date: _/_/_

Run date:

Run time:

March 11, 1992 12:50 p.m.

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 480 MIN GEOMETRY 10: Cylindrical source from end -

Distance to detector......................... X Source cylinder radius....................... R Source cylinder length.....*................. Tl Thickness of second shield................... T2 Thickness of third shield.*.................. T3 Thickness of fourth shield................... T4 Thickness of fifth shield...................* T5 Microshield inserted air gap...............*. air slab shields 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet ~

Rev I O Page 2 File: 480.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 480 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide curies I-131 5.6300e+03 I-132 7.3790e+02 I-133 7.2640e+03 I-134 2.2070e+Ol I-135 4.1980e+03 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.1766

1. 6327 1.1845

.8447

.6626

.5228

.3687

.2862

.2190

.1433 TOTALS:

5.064e+12 5.418e+13 1.223e+14 6.394e+13 6.550e+13 2.60le+14 1.781e+14 2.009e+13 3.668e+12 1.039e+ll 7.730e+14

1. 750e+02 3.933e+02 1.115e+02 9.423e+OO 2.024e+OO 1.922e+OO 7.067e-02 7.519e-04 9.127e-06 1.592e-09 6.933e+02 2.759e-Ol 6.785e-Ol 2.070e-Ol 1.877e-02 4.198e-03 3.943e-03

1. 454e-04 1.501e-06 1.723e-08 2.. 702e-12 l.188e+OO

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet #9 Rev # O Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File Run date:

720MIN.MSH March 11, 1992 1:01 p.m.

Date: ~-/~_/~

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 720 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector **..*...*.*...*....*..... X Source cylinder radius..*..*....***.....*.*.. R Source cylinder length...***.....**.....****. Tl Thickness of second shield...**.............. T2 Thickness of third shield ***..*..**....*....* T3 Thickness of fourth shield.*...*.**.......... T4 Thickness of fifth shield..........*....**... T5 Microshield inserted air gap......*.*...**.*. air 454.660 701. 040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron.

Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet so Rev # 0 Page 2 File: 720MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 720 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide Curies I-131 8.4380e+03 I-132 3.3610e+02 I-133 9.6680e+03 I-134 1.4200e+OO I-135 4.1960e+03 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.1919 4.610e+12 1.638e+02 2.576e-Ol 2

1.6379 5.268e+13 3.899e+02 6.722e-Ol 3

1.1860 1.242e+14 1.141e+02 2.117e-Ol 4

.8556 5.183e+13 8.192e+OO 1.628e-02 5

.6615 5.724e+l3

1. 751e+OO 3.634e-03 6

.5229 3.358e+14 2.484e+OO 5.095e-03 7

.3681 2.630e+14 1.028e-Ol 2.115e-04 8

.2867 2.643e+13 1.012e-03 2.021e-06 9

.2165 3.914e+12 8.786e-06 1.654e-08 10

.1474 3.198e+10 8.566e-10

1. 470e-12 11 12 -

13 14 15 16 17 18 19 20 TOTALS:

9.198e+14 6.804e+02 1.167e+OO

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet...kL_ Rev # 0 Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File Run date:

1440MIN.MSH March 11, 1992 2:52 p.m.

Date: _/_/_

Run time:

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 1440 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector.........................

Source cylinder radius...*...................

Source cylinder length.......................

Thickness of second shield...................

Thickness of third shield...........*........

Thickness of fourth shield *..................

Thickness of fifth shield.....**.............

Microshield inserted air gap..............*.*

x R

Tl T2 T3 T4 T5 air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet.52 Rev # O Page 2 File: 1440MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 1440 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)......*.*..*...

11 Number of radial segments (Nradius) *.*........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131

1. 6380e+04 I-132

1. 8310e+Ol I-133

1. 3140e+04 I-134 2.1760e-04 I-135 2.4160e+03 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2057 2.460e+12 8.967e+Ol 1.406e-Ol 2

1. 6422 2.975e+13 2.237e+02 3.854e-Ol 3

1.1954 8.375e+13 8.025e+Ol 1.486e-Ol 4

.8682 4.088e+13 7.009e+OO

1. 390e-02 5

.6581 7.130e+13 2.112e+OO 4.386e-03 6

.5229 4.428e+14 3.275e+OO 6.717e-03 7

.3673 5.004e+14

1. 916e-Ol 3.94le-04 8

.2874 4.219e+13

1. 662e-03 3.320e-06 9

.2042 3.373e+12 4.592e-06 8.513e-09 10

.1484

1. 605e+09 4.981e-ll 8.572e-14 11 12 13 14 15 16 17 18 19 20 TOTALS:

1.217e+15 4.062e+02 7.oooe-01

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet...:S..3_ Rev # O Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File

2880MIN.MSH March 11, 1992 2:55 p.m.

Date: _/_/_

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 2880 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector...............*......... X Source cylinder radius....................... R Source cylinder length....**...*...........** Tl Thickness of second shield *.................* T2 Thickness of third shield *....**............* T3 Thickness of fourth shield.....*............* T4 Thickness of fifth shield...........*.....**. T5 Microshield inserted air gap...*.*........... air 454.660 701.040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 5~ Rev # 0 Page 2 File: 2880MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 2880 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 3.0260e+04 I-132 2.6620e-02 I-133 1.1890e+04 I-134 2.5070e-12 I-135 3.9250e+02 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.2072

1. 6427 1.2252

.8689

.6548

.5228

.3670

.2882

.1838

.1484 TOTALS:

3.964e+ll 4.823e+12 3.042e+13 3.008e+13 1.159e+14 3.970e+14 9.164e+14 7.036e+13 3.252e+12 2.333e+06 1.569e+15 1.449e+Ol 3.632e+Ol 3.330e+Ol 5.181e+OO 3.327e+OO 2.933e+oo 3.478e-Ol 2.862e-03 1.250e-06 7.243e-14 9.591e+Ol 2.272e-02 6.258e-02 6.127e-02 1.027e-02 6.913e-03 6.014e-03 7.155e-04 5.719e-06 2.257e-09

1. 246e-16

1. 705e-Ol

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet 55 Rev I O File Ref:

Page File

4320MIN.MSH March 11, 1992 3:00 p.m.

Date: _/_/_

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 4320 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector *...***......**....*...*. X Source cylinder radius...**..*.*......*...... R Source cylinder length *****....**.*...*.**... T1 Thickness of second shield **..**.***.*****... T2 Thickness of third shield....**.**....**.*... T3 Thickness of fourth shield ***.**..*.**....*** T4 Thickness of fifth shield ***...**.**.*...*... TS Microshield inserted air gap...******...*..*. air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES {g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

Air gap

.001220'

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet..:5..i._ Rev # O Page 2 File: 4320MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 4320 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131 4.1730e+04 I-132 2.8910e-05 I-133 8.0320e+03 I-134 2.1560e-20 I-135 4.7620e+Ol RESULTS:

Grc;>Up Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.2072

1. 6427

1. 2433

.8688

.6531

.5226

.3670

.2887

.1801

.1484 TOTALS:

4.809e+10 5.851e+ll

1. 505e+13 1.957e+13 1.513e+14 2.704e+14

1. 262e+15 9.491e+13 4.124e+12 2.534e+03 1.818e+15

1. 758e+OO 4.407e+OO 1.807e+Ol 3.368e+OO 4.275e+OO

1. 993e+OO 4.784e-Ol 3.925e-03

1. 230e-06 7.866e-17 3.436e+Ol 2.756e-03 7.592e-03 3.315e-02 6.678e-03 8.887e-03 4.088e-03 9.842e-04 7.845e-06 2.212e-09 l.354e-19 6.414e-02

ATTACHMENT 3 Microshield 3.13

=====

EA-A-NL-92-012-03 Sheet 57 Rev # 0

{Consumers Power Company -

037)

Page

1 File

5760MIN.MSH Run date: March 16, 1992 Run time: 4:07 p.m.

File Ref:

Date: ____ / ____ / ___ _

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 5760 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector *...*....*.**.***......** X Source cylinder radius...*...**..*........*.* R Source cylinder length...*....*..***.**...*** Tl Thickness of second shield....*..*.******.*.. T2 Thickness of third shield ****.*****.*....*.** T3 Thickness of fourth shield.. ~.*.......*..*..* T4 Thickness of fifth shield *.*..*.*....*..*..*. T5 Microshield inserted air gap *.*...**..**.*.*. air 454.660 701.040 60.960 2.540 55.880

.121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 5B Rev # 0 Page 2 File: 5760MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 5760 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 5.1110e+04 I-132 2.7870e-08 I-133 4.8180e+03 I-134 1.0000e-60 I-135 5.1290e+OO RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.2072

1. 6427

1. 2481

.8688

.6523

.5223

.3670

.2889

.1797

.1484 TOTALS:

5.180e+09 6.302e+10 8.439e+12 1.166e+13

1. 803e+14

1. 656e+14 1.545e+15 1.153e+14 5.012e+12 2.443e+OO 2.031e+15 1.893e-Ol 4.747e-Ol

1. 038e+Ol 2.006e+OO 5.053e+OO 1.216e+OO 5.855e-Ol 4.808e-03 1.459e-06 7.583e-20 1.991e+01 2.969e-04 8.177e-04 1.902e-02 3.978e-03 l.051e-02 2.494e-03 1.205e-03 9.610e-06 2.622e-09

1. 305e-22 3.833e-02

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 59 Rev # O Microshield 3.13

=====

(Consumers Power Company -

037) 1 File Ref:

Page File Run date:

8640MIN.MSH March 16, 1992 4:09 p.m.

Date: _/_/_

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 8640 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ***************.*********

Source cylinder radius ******..*.**..**.******

Source cylinder length.*..*****..*..*....****

Thickness of second shield ***.***...*....***.

Thickness of third shield ***.**.**.**********

Thickness of fourth shield **.***..*.*********

Thickness of fifth shield *********..*...*.***

Microshield *inserted air gap...... *...........

x R

Tl T2 T3 T4 TS air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND Source Shield 2

1. 0 2.0 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet ~

Rev # 0 Page 2 File: 8640MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 8640 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments {Npsi)...............

11 Number of radial segments {Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131 6.4590e+04 I-132 2.1830e-14 I-133

1. 4620e+03 I-134 1.ooooe-60 I-135 5.0150e-02 RESULTS:

Group Energy Activity Dose point flux Dose rate

{MeV)

{photons/sec)

MeV/{sq cm)/sec

{mr/hr) 1 2.2072 5.065e+07 1.851e-03 2.903e-06 2

1.6427 6.162e+08 4.641e-03 7.996e-06 3

1.2492 2.523e+12 3.120e+OO 5.716e-03 4

.8688 3.532e+12 6.079e-Ol 1.205e-03 5

.6516 2.233e+14 6.218e+OO 1.293e-02 6

.5206 5.679e+13 4.090e-Ol 8.386e-04 7

.3670 1.951e+15 7.396e-Ol

1. 522e-03 8

.2890 1.449e+14 6.078e-03

1. 215e-05 9

.1797 6.329e+12 1.838e-06 3.303e-09 10

.1484 1.913e-06 5.940e-26

1. 022e-28 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.389e+15

1. llle+Ol 2.223e-02

ATTACHMENT 3 Microshield 3.13 (Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet.fiL_ Rev # 0 File Ref:

Page File Run date:

10080MIN.MSH March 16, 1992 4:10 p.m.

Date: _/_/_

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 10080 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector *.******..**.************

Source cylinder radius **********************.

Source cylinder length..*.***.***.*..*.*.****

Thickness of second shield *.*.**.**....*****.

Thickness of third shield ********************

Thickness of fourth shield *******************

Thickness of fifth shield **.*******....*.****

Microshield inserted air gap.***.**.*.*******

x R

Tl T2 T3 T4 T5 air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet..b..2_ Rev # 0 Page 2 File: 10080MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 10080 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide Curies I-131 6.9150e+04 I-132

1. 8400e-17 I-133 7.6660e+02 I-134 1.0000e-60 I-135 4.7220e-03 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2072 4.769e+06

1. 743e-04 2.733e-07 2

1.6427 5.802e+07 4.370e-04 7.528e-07 3

1.2492 1.322e+12 1.635e+OO 2.996e-03 4

.8688 1.852e+12 3.187e-Ol 6.319e-04 5

.6515 2.383e+14 6.628e+OO 1.378e-02 6

.5189 3.448e+13 2.433e-Ol 4.987e-04 7

.3670 2.089e+15 7.917e-Ol 1.629e-03 8

.2890 1.550e+14 6.508e-03 1.301e-05 9

.1797 6.776e+12

1. 968e-06 3.536e-09 10

.1484

1. 613e-09 5.007e-29 8.616e-32 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.527e+15 9.625e+OO 1.955e-02

ATTACHMENT 3 Microshield 3.13 (Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet...eJ_ Rev # O File Ref:

Page File

11520MIN.MSH March 16, 1992 4:11 p.m.

Date: ___ / ___ / __ _

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 11520 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector......................... x 454.660 Source cylinder radius........................ R 701.040 Source cylinder length....................... Tl 60.960 Thickness of second shield................... T2 2.540 Thickness of third shield.................... T3 55.880 Thickness of fourth shield................... T4 121.920 Thickness of fifth shield.................... T5 30.480 Microshield inserted air gap..........*...**. air 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc):

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet..i.!!__ Rev # O Page 2 File: 11520MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 11520 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 7.2520e+04 I-132

1. 5190e-20 I-133 3.9380e+02 I-134 1.0000e-60 I-135 4.3560e-04 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2072 4.399e+05 1.608e-05 2.521e-08 2

1. 6427 5.352e+06 4.031e-05 6.945e-08 3

1.2492 6.792e+ll 8.401e-Ol 1.539e-03 4

.8688 9.512e+ll

1. 637e-Ol 3.246e-04 5

.6514 2.495e+14 6.937e+OO 1.442e-02 6

.5165 2.265e+13 1.553e-Ol 3.183e-04 7

.3670 2.191e+15 8.303e-Ol 1.708e-03 8

.2891 1.625e+14 6.825e-03 l.364e-05 9

.1797 7.106e+12 2.064e-06 3.709e-09 10

.1484

1. 331e-12 4.133e-32 7.113e-35 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.634e+15 8.933e+OO 1.833e-02

ATTACHMENT 3 Microshield 3.13 (Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet t5 Rev*# O File Ref:

Page File

12960MIN.MSH March 16, 1992 4:12 p.m.

Date: ___ / ___ / __ _

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 12960 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector **********.**.*.*.*******

Source cylinder radius..*.*..******.*.*.**.**

Source cylinder length *.*.*************..*.**

Thickness of second shield.****.********.**.*

Thickness of third shield **.*****************

Thickness of fourth shield..**..*.*********.*

Thickness of fifth shield ***...*..***.**.****

Microshield inserted air gap *****.*...... ~ ***

x R

T1 T2 T3 T4 T5 air 454.660 701. 040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND Source Shield 2

1. 0 2.0 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet ~

Rev # O Page 2 File: 12960MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 12960 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide curies I-131 7.4860e+04 I-132 1.0000e-60 I-133 1.9920e+02 I-134 1.0000e-60 I-135 3.9550e-05 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.2072

1. 6427 1.2492

.8688

.6514

.5138

.3670

.2891

.1797

.1355 TOTALS:

3.994e+04 4.859e+05 3.436e+11 4.812e+11 2.573e+14 1.655e+13 2.261e+15

1. 677e+14 7.336e+12 1.867e-51*

2.711e+15 1.460e-06 3.660e-06 4.249e-01 8.283e-02 7.153e+OO

1. lOOe-01 8.571e-Ol 7.046e-03 2.130e-06 9.737e-72 8.635e+OO 2.289e-09 6.306e-09 7.786e-04

1. 642e-04 1.487e-02 2.253e-04 1.763e-03 1.408e-05 3.828e-09

1. 618e-74 1.782e-02

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet ~

Rev # O Microshield 3.13

=====

(Consumers Power Company -

037) File Ref:

Page File 1

14400MIN.MSH March 16, 1992 4:14 p.m.

Date: ____ / ____ / ___ _

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 14400 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector **......*.*.*.....*..***.

source cylinder radius..****..***..**.....*..

Source cylinder length *.****..**....**.......

Thickness of second shield...**..**....*...**

Thickness of third shield...*.*..**....**..**

Thickness of fourth shield *.*.*....*.*...***.

Thickness of fifth shield..*.*..**.***....**.

Microshield inserted air gap..*...****.*.....

x R

T1 T2 T3 T4 T5 air 454.660 701.040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet ""11.._ Rev # 0 Page 2 File: 14400MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 14400 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide Curies I-131 7.6320e+04 I-132 1.0000e-60 I-133 9.9470e+Ol I-134 1.0000e-60 I-135 3.5460e-06 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.2072

1. 6427 1.2492

.8688

.6514

.5115

.3670

.2891

.1797

.1355 TOTALS:

3.581e+03 4.357e+04 1.716e+ll 2.403e+ll 2.622e+14

1. 346e+13 2.306e+15 1.709e+14 7.479e+12 1.867e-51 2.760e+15 1.309e-07 3.282e-07 2.122e-01 4.136e-02 7.289e+OO 8.701e-02 8.738e-Ol 7.183e-03 2.172e-06 9.737e-72 8.SlOe+OO 2.052e-10 5.654e-10 3.888e-04 8.199e-05 1.516e-02 1.782e-04 1.798e-03

1. 436e-os 3.903e-09

1. 618e-74 1.762e-02

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet..£.j__ Rev # O File Ref:

Page File

17280MIN.MSH March 16, 1992 4:15 p.m.

Date: _/_/_

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 17280 MIN

.GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector......................... X Source cylinder radius.**..*.*******..**....* R Source cylinder length *..*..*****..**.*****.* Tl Thickness of second shield *..*.******..****** T2 Thickness of third shield *.*......**...*..*** T3 Thickness of fourth shield..*....*.**.**..*** T4 Thickness of fifth shield *..*......*......*** T5 Microshield inserted air gap.**************** air 454.660 701. 040 60.960 2.540 55.880 121.920 30.~480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 7o Rev # O Page 2 File: 17280MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 17280 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 I

SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 7.7090e+04 I-132

1. ooooe-60 I-133 2.4110e+01 I-134 1.ooooe-60 I-135 2.7720e-08 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2072 2.799e+01

1. 023e-09

1. 604e-12 2

1.6427 3.406e+02 2.565e-09 4.420e-12 3

1.2492 4.158e+10 5.143e-02 9.423e-05 4

.8688 5.824e+10

1. 002e-02 1.987e-05 5

.6514 2.648e+14 7.359e+OO 1.530e-02 6

.5089 1.108e+13 6.946e-02

1. 422e-04 7

.3670 2.329e+15 8.826e-01 1.816e-03 8

.2891

1. 726e+14 7.256e-03

1. 450e-05 9

.1797 7.554e+12 2.194e-06 3.942e-09 10

.1355 1.867e-51 9.737e-72

1. 618e-74 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.785e+15 8.380e+OO 1.739e-02

ATTACHMENT 3 Microshield 3.13 (Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet ::zj__ Rev # 0 File Ref:

Page File Run date:

20160MIN.MSH March 16, 1992 4:16 p.m.

Date: ~-/~_/~

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 20160 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector *** ~ *******.*.*****..***. X Source cylinder radius *********.*.***.*.*..** R Source cylinder length ***************.***.**. Tl Thickness of second shield ******************. T2 Thickness of third shield *..***********.***** T3 Thickness of fourth shield.**.*********.***** T4 Thickness of fifth shield ******.******.***.** T5 Microshield inserted air gap ************.*.*. air 454.660 701.040 60.960 2.540 55.880

.121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 7~ Rev # 0 Page 2 File: 20160MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 20160 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131 7.5710e+04 I-132 1.0000e-60 I-133 5.6840e+OO I-134 l.OOOOe-60 I-135 2.1060e-10 RESULTS:

Group Energy (MeV)

Activity (photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 2.2072

1. 6427 1.2492

.8688

.6514

.5081

.3670

.2891

.1797

.1355 TOTALS:

2.127e-Ol 2.588e+OO 9.803e+09

1. 373e+10 2.600e+14 1.029e+l3 2.287e+15

1. 695e+14 7.419e+12 1.867e-51 2.734e+l5 7.774e-12 1.949e-11

1. 212e-02 2.363e-03 7.227e+OO 6.389e-02 8.668e-Ol 7.126e-03 2.155e-06 9.737e-72 8.179e+oo

1. 219e-14 3.358e-14 2.222e-05 4.685e-06

1. 503e-02

1. 308e-04 1.783e-03 1.424e-05 3.872e-09 1.618e-74 1.698e-02

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet..z.J.... Rev # 0 File Ref:

Page File Run date:

23040MIN.MSH March 16, 1992 4:17 p.m.

Date: _/_/_

Run time:

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 23040 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ************************* X source cylinder radius **************.***.**** R Source cylinder length.****.*.**.*.*.****..** T1 Thickness of second shield ********.*.****...* T2 Thickness of third shield *********.*****..*** T3 Thickness of fourth shield **********.***..*** T4 Thickness of fifth shield *************.*..*.* TS Microshield inserted air gap *******...**..*** air 454.660 701.040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc):

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet 7~ Rev # 0 Page 2 File: 23040MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 23040 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide Curies I-131 7.2830e+04 I-132 1.0000e-60 I-133

1. 3120e+OO I-134 1.ooooe-60 I-135

1. 5670e-12 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2072 1.582e-03 5.785e-14 9.069e-17 2

1. 6427

1. 925e-02 1.450e-13 2.498e-16 3

1. 2492 2.263e+09 2.799e-03 5.128e-06 4

.8688 3.169e+09 5.455e-04 1.081e-06 5

.6514 2.501e+14 6.952e+OO 1.446e-02 6

.5079 9.758e+12 6.046e-02

1. 238e-04 7

.3670 2.200e+15 8.338e-Ol 1.715e-03 8

.2891

1. 631e+14 6.855e-03

1. 370e-05 9

.1797 7.137e+12 2.073e-06 3.724e-09 10

.1355 l.867e-51 9.737e-72

1. 618e-74 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.630e+15 7.856e+OO

1. 631e-02

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet 75 Rev # 0 File Ref:

Page File Run date:

25920MIN.MSH March 16, 1992 4:18 p.m.

Date: _/_/_

Run time:

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 25920 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector.*...****.***............

Source cylinder radius *.......*****..**.**.**

Source cylinder length...*******...*.........

Thickness of second shield **.**.**.....*.....

Thickness of third shield ********..*..**..*.*

Thickness of fourth shield..*.....***********

Thickness of fifth shield *......***.*....****

Microshield inserted air gap..*..*.**.....*.*

x R

Tl T2 T3 T4 T5 air 454.660 701.040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet ~

Rev # O Page 2 File: 25920MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 25920 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 6.8960e+04 I-132

1. ooooe-60 I-133 2.9820e-Ol I-134 1.ooooe-60 I-135 1.1480e-14 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2072 1.159e-05 4.238e-16 6.644e-19 2

1.6427

1. 411e-04 1.062e-15 1.830e-18 3

1.2492 5.143e+08 6.361e-04 l.166e-06 4

.8688 7.203e+08 1.240e-04 2.458e-07 5

.6514 2.368e+14 6.583e+OO

1. 369e-02 6

.5078 9.208e+12 5.702e-02 l.167e-04 7

.3670 2.083e+15 7.895e-Ol 1.624e-03 8

.2891 1.544e+14 6.490e-03 l.297e-05 9

.1797 6.757e+12 1.963e-06 3.527e-09 10

.1355 1.867e-51 9.737e-72 l.618e-74 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.490e+15 7.436e+OO 1.544e-02

ATTACHMENT 3 Microshield 3.13

=====

EA-A-NL-92-012-03 Sheet 77 Rev # O (Consumers Power Company -

037)

Page

1 File

28800MIN.MSH Run date: March 16, 1992 Run time: 4:19 p.m.

File Ref:

Date: ~-/~_/~

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 28800 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector.****........****........ X Source cylinder radius.****.*....*...***..*** R Source cylinder length.*****.........***....* Tl Thickness of second shield..******..*......** T2 Thickness of third shield ***....**...***...** T3 Thickness of fourth shield....*...**......**. T4 Thickness of fifth shield *.....*..*....*.***. T5 Microshield inserted air gap......*.*...*..*. air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc):

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350

1. 0 2.0 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet..zfi_ Rev # 0 Page 2 File: 28800MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 28800 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)...............

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 6.4490e+04 I-132

1. ooooe-60 I-133 6.6930e-02 I-134

1. ooooe-60 I-135 8.3060e-17 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr/hr) 1 2.2072 8.388e-08 3.066e-18 4.807e-21 2

1. 6427

1. 021e-06 7.687e-18 1.324e-20 3

1.2492 1.154e+08

1. 428e-04 2.616e-07 4

.8688

1. 617e+08 2.783e-05 5.517e-08 5

.6514 2.215e+14 6.156e+OO 1.280e-02 6

.5078 8.604e+12 5.327e-02 1.091e-04 7

.3670

1. 948e+15 7.383e-Ol 1.519e-03 8

.2891

1. 444e+14 6.070e-03

1. 213e-05 9

.1797 6.319e+12

1. 835e-06 3.298e-09 10

.1355

1. 867e-51 9.737e-72

1. 618e-74 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.329e+15 6.954e+OO

1. 444e-02

ATTACHMENT 3 Microshield 3.13 (Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet 7~ Rev # 0 File Ref:

Page File

36000MIN.MSH March 16, 1992 4:22 p.m.

Date: _/_/_

Run date:

By:

Run time:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 36000 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector.........................

Source cylinder radius ****.************..**.*

source cylinder length **.*******.*******.****

Thickness of second shield *******************

Thickness of third shield ********************

Thickness of fourth shield ******.**.*******.*

Thickness of fifth shield ***.***..****.**.**.

Microshield inserted air gap ************.****

x R

Tl T2 T3 T4 TS air 454.660 701.040 60.960 2.540 55.880 121.920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc):

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet../lQ__ Rev # 0 Page 2 File: 36000MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 36000 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments {Npsi)...............

11 Number of radial segments {Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide Curies I-131 5.2390e+04 I-132

1. ooooe-60 I-133 1.5340e-03 I-134 1.0000e-60 I-135

1. ooooe-60 RESULTS:

Group Energy Activity Dose point flux Dose rate

{MeV)

{photons/sec)

MeV/(sq cm)/sec

{mr/hr) 1 2.1181 2.469e-51 7.655e-62

1. 221e-64 2

1.6189 2.271e-50 1.567e-61 2.708e-64 3

1.2492 2.646e+06 3.272e-06 5.996e-09 4

.8688 3.705e+06 6.378e-07 1.264e-09 5

.6514 1.799e+14 5.00le+OO

1. 040e-02 6

.5078 6.988e+12 4.326e-02 8.858e-05 7

.3670

1. 583e+15 5.998e-Ol

1. 234e-03 8

.2891 1.173e+14 4.931e-03 9.857e-06 9

.1797 5.134e+12 1.491e-06 2.679e-09 10

.1355 1.867e-51 9.737e-72

1. 618e-74 11 12 13 14 15 16 17 18 19 20 TOTALS:

1.892e+15 5.649e+OO 1.173e-02

ATTACHMENT 3 Microshield 3.13

=====

(Consumers Power Company -

037) 1 EA-A-NL-92-012-03 Sheet /iL__ Rev # O File Ref:

Page File Run date:

Run time:

43200MIN.MSH March 16, 1992 4:23 p.m.

Date: ~-/~_/~

By:

Checked:

CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 43200 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector...****************.***** X Source cylinder radius ******.*...********..** R Source cylinder length *******..**..*.**.*.**. Tl Thickness of second shield ***.****.*..***.*** T2 Thickness of third shield ***..*************** T3 Thickness of fourth shield *****.**.*****.**** T4 Thickness of fifth shield *..***************.. T5 Microshield inserted air gap ******.*...*.**** air 454.660 701.040 60.960 2.540 55.880 121. 920 30.480 182.880 Source Volume: 9.41197e+7 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium SAND MATERIAL DENSITIES (g/cc) :

Source Shield 2 Shield 3 Shield 4 Shield 5

.001220 2.350 2.350 1.0 2.0 cm.

II II II II II II II Air gap

.001220

ATTACHMENT 3 EA-A-NL-92-012-03 Sheet.fi..d_ Rev # 0 Page 2 File: 43200MIN.MSH CASE: EA-A-NL-92-012-03 SIRWT SHINE TO CONTROL ROOM AT 43200 MIN BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)..*..*....*.*..

11 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 4.0860e+04 I-132 1.0000e-60 I-133 3.3760e-05 I-134 1.0000e-60 I-135 l.OOOOe-60 RESULTS:

Group Energy Activity Dose point flux Dose rate (MeV)

(photons/sec)

MeV/(sq cm)/sec (mr /hr) 1 2.1181 2.469e-51 7.655e-62

1. 221e-64 2

1. 6189 2.271e-50 1.567e-61 2.708e-64 3

1.2492 5.822e+04 7.202e-08

1. 320e-10 4

.8688 8.155e+04

1. 404e-08 2.783e-11 5

.6514 1.403e+14 3.900e+OO 8.llOe-03 6

.5078 5.450e+12 3.374e-02 6.908e-05 7

.3670

1. 234e+15 4.678e-Ol 9.624e-04 8

.2891 9.150e+13 3.846e-03 7.688e-06 9

.1797 4.004e+12 1.163e-06 2.090e-09 10

.1355 1.867e-51 9.737e-72

1. 618e-74 11 12 13 14 15 16 17 18 19 20 TOTALS:

1. 4 76e+15 4.406e+OO 9.149e-03

Form 3698 9-89 PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CHECKLIST Items Affected By This EA Other EAs

2.

Design Documents Elec E-38 through E-49

3.

Design Documents Mech M259, M664, M665 4.0 LICENSING DOCUMENTS 4.1 Final Safety Analysis Report (FSAR) 4.2 Technical Specifications 4.3 Standing Order 54 5.0 PROCEDURES 5.1 Administrative Procedures 5.2 Working Procedures 5.3 Tech Spec Surveillance Procedures 6.0 OTHER DOCUMENTS Q-List

.2 Plant Drawings Equipment Data Base 6.4 Spare Parts (Stock/MMS) 6.5 Fire Protection Program Report (FPPR) 6.6 Design Basis Documents 6.7 Operating Checklists 6.8 SPCC/PIPP Oil and Hazardous Material Spill Prevention Plan 6.9 EEQ Documents Affected Revision Yes No Required D g D

D '

D g D

18 D

rgi D

~

D 0

D..l8l D

~

D 181 D t3 D

~

D 3 y'

I Identify*

3...7, 9{z Do any of the following documents need to be generated as a result of this EA:

Yes No

1.

Corrective Action Document?

Reference ~~~~~~~~~~-

Closeout 0-'o/e Z..fi~

14;z-,;.,7eq' v/ex Vt~ re &ii<<,

2.

Safety Evaluation?

3.

EEQ Evaluation Sheet?

D 3

~ D D g Reference/; ~P,4/t'i4?ecl deA tev,Sct.I z; '4~1, Reference ~~~~~~~~~~-

Is PRC Review of this EA Required?

D

~

ompleted By _

__._~"'-et 1/0<....*..:.../---.A~-,.____._!la........,t.....

fe.""'".;,i_,__ ___________

Date %/cz

Identify Section, No, Drawing, Document, etc.

TECffNICAL REVIEW CffECgLISJ EA -.A-A/t-12-ot2-e3 REV.

0 Proc No 9.11 Attachment S Revision 5 Page 1 of 1 This checklist provides guidance for the review of engineering analyses.

Answer questions Yes or No, or N/A if they do not apply.

Document all co11111ents on a 3110 Form. Satisfactory resolution of conments and completion of this checklist 1s noted by the Techn1eally Reviewed signature on the Initiation and Review record block of Fon11 3619.

1. Have the proper i.nput codes, standards and design principles been specified?

2.

Have the input codes, standards and design principles been properly applied?

3. Are all inputs and assumptions valid and th* basis for their use documented?

4.

Is Vendor information used as input addressed correctly in the analysis?

(Y, N, N/A) y y

5. If the analysts argument departs frOll Vendor Infonutton/Recomendations, ts the departure

/V/ A Justiftcatton documented?

/'/

6. Are assu.pttons accurately described and reasonable?

7. Has*'the use of engineering judg&Mnt been docu.ntlCI and y

Justified?

8. Are all constants, variables and for11Ulas correct and properly applied?

y

9. Have any *tnor (tnstgntftcant) errors beea id1ntiffld? If

~,,,,....3i10 yes; Identify on the 3110 Fona ancl justify their

=lf=c.{.J

Y.

instgntftcance.

(

10. Does analysts involve welding? If Yes; vertfy the following tnfor111tton ts accurately repres1ntlCI on the I\\/.

analysts drawing (OUtput docU111nt).

Type of Veld

Size of Veld *

Mate'ial Betng Jofnld

11ttclmess of Material Bef ng Joined

LocattOlt of Vald(s)

Appropriate Veld S)'llbology

11. Has the objective of the analysts been.. t?

12. Have 1dlltntstr1tfv1 requtr... nts such as nUllbertng and fonaat been sattsff ed?

y y

NUCLEAR OPE NS DEPARTMENT Document Title Or,;:::-5/TE V'0.5.£S

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/?CCO~.A/77/1/i.

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NUCLEAR OPE-NS DEPARTMENT Docum~view Sheet Document Number Comments Date Review Coordinator Date Revision Revision Number Page 3 of ?

Response or Resolution 1'Z

FIRST DISK IS CRl_MHA DATA Wl 3/20/92 11:58:47 SECOND DISK IS CRJCOl DATA Wl 4/14/92 8:02:58

1.

1 EA-A-NL-92-012-03, MHA ANALYSIS ACCOUNTING FOR MODS, STACK RELEASE

2.

1 EA-A-NL-92-012-02, JCO ANALYSIS WITH LOCA FUEL FAILURES, STACK RELEASE

1.

17 0.00 0.0 0.0 0.0 550.0

1.

18 1.26 0.0 1100.0 2420.0 0.0

2.

17 0.00 0.0 0.0 0.0 511.6

2.

18 1.26 11.6 1000.0 2200.0 0.0

1.

22 10

1.

23 0.00*

1.00 1.26 12.00 19.00 480.00 1440.00

1.

24 14400.00 28800.00 43200.00

2.

22 11

2.

23 0.00 1.00 1.26 12.00 19.00 480.00 667.00

2.

24 5760.00 14400.00 28800.00 43200.00 FILEl:

6 INSERTS, 98 RECORDS; FILE2:

6 INSERTS, 98 RECORDS.

0 CHANGES.

FIRST DISK IS CR2_MHA DATA Wl 3/20/92 12:21:26 SECOND DISK IS CRJC02 DATA Wl 4/14/92 8:04:05

1.

1 EA-A-NL-92-012-03, MHA ANALYSIS ACCOUNTING FOR MODS, SIRW RELEASE

2.

1 EA-A-NL-92-012-02, JCO ANALYSIS WITH LOCA FUEL FAILURES, SIRW RELEASE

1.

17 0.00 0.0 0.0 0.0 550.0

1.

18

1. 26 0.0 1100.0 2420.0 0.0

2.

17 0.00 0.0 0.0 0.0 511.6

2.

18 1.26 11.6 1000.0 2200.0 0.0

1.

22 10

1.

23 0.00 1.00 1.26 12.00 19.00 480.00 1440.00

1.

24 14400.00 28800.00 43200.00

2.

22 11

2.

23 0.00 1.00 1.26 12.00 19.00 480.00 667.00

2.

24 5760.00 14400.00 28800.00 43200.00 FILEl:

6 INSERTS, 98 RECORDS; FILE2:

6 INSERTS, 98 RECORDS.

0 CHANGES.

Microshield 3.13

=====

(Consumers Power Company -

037)

Page 1

File Ref:

File Date: ~-/~_/~-

Run date: April 21, 1992 By:

Run time: 7 : 3 5 a. m.

Checked:

CASE: BENCHMARK TEST CASE ONE GEOMETRY 1: Point source - slab shields Distance to detector....*..*..*.....*........ X Lateral displacement of dose point........... Y Microshield inserted air gap.**....**........ air Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium MATERIAL DENSITIES (g/cc) :

Air gap

.001220 6096.

1737.360 6096.

cm.

II II

Page 2 File:

CASE: BENCHMARK TEST CASE ONE BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

Group 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 INTEGRATION PARAMETERS:

None - analytically integrated.

SOURCE NUCLIDES:

Source was entered by energy groups.

RESULTS:

Energy Activity (MeV)

(photons/sec) 6.2000 1.000e+OO TOTALS:

1.000e+OO Dose point flux MeV/(sq cm)/sec 1.109e-08 1.109e-08 Dose rate (mr/hr) 1.208e-11

1. 208e-ll eo~7~/'e./ co

/,Ot,'E-/C ~

N;v;;* -;;iJJe '11-2 bl'./11/c,"dsl/u{/.3 -~""va /,

Microshield 3.13

=====

(Consumers Power Company -

037)

Page 1

File Run date: April 21, 1992 Run time: 7:52 a.m.

CASE: BENCHMARK TEST CASE TWO File Ref:

Date: ~-/~_/~

By:

Checked:

GEOMETRY 7: Cylindrical source from side - cylindrical shields Distance to detector.........................

Source length................................

Dose point height from base *..*....*******.**

Source cylinder radius *..........*...********

Microshield inserted air gap *********......*.

x L

y Tl air 6096.

1067.

91.4 183.

5913.

Source Volume: 1.12258e+8 cubic centimeters Material Air Aluminum Carbon Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water Zirconium MATERIAL DENSITIES (g/cc) :

Source Air gap

.001220

1. 0 cm.

II II II II

\\)\\..

f\\

\\

Page

~

2 File:

CASE: BENCHMARK TEST CASE TWO BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 2.

INTEGRATION PARAMETERS:

Number of lateral angle segments (Ntheta).....

15 Number of azimuthal angle segments (Npsi).....

31 Number of radial segments (Nradius)...........

11 SOURCE NUCLIDES:

source was entered by energy groups.

RESULTS:

Group Energy Activity (MeV)

(photons/sec)

Dose point flux MeV/(sq cm)/sec Dose rate (mr/hr) 1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20

.8000 4.210e+09 6.720e-Ol 1.352e-03 TOTALS:

4.210e+09 6.720e-Ol 1.352e-03 Con/fl"'~/ ~ /,/#C-0..? 71' _r,,,,,,, 7;J/e_ ~-~

o./ ~(Yd*A',;e//.3 /1.:n.va I.

ML18058A404

Contents

Text

SUMMARY

2.0 REFERENCES

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

3.0 BACKGROUND

SUMMARY

SUMMARY

8.0 CONCLUSION

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ML18058A404

Text

SUMMARY

2.0 REFERENCES

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

3.0 BACKGROUND

SUMMARY

SUMMARY

8.0 CONCLUSION

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