Text

Rev 1 to Engineering Analysis (EA) EA-A-NL-92-012-02, Offsite Doses and CR Habitability Following Large Break LOCA to Justify Continued Operation Until Cycle 12.
	ML18058A992
Person / Time
Site:	Palisades
Issue date:	03/30/1992
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18058A991	List: ML18058A990; ML18058A992;
References
	EA-A-NL-92-012, EA-A-NL-92-012-02-R1, EA-A-NL-92-12, EA-A-NL-92-12-2-R1, NUDOCS 9208030172
	Download: ML18058A992 (113)
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ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 Off-Site Doses and Control Room Habitability Following A Large Break LOCA to Justify Continued Operation Until Cycle 12 (EA-A-NL-92-012-02 Rev. 1)

July 28, 1992

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ENGINEERING ANALYSIS COVER SHEET Total Nlallber of Sheets I~

Title OFFSITE DOSES AND CONTROL ROOM HABITABILITY FOLLOWING A LARGE BREAK LOCA TO JUSTIFY CONTINUED OPERATION UNTIL CYCLE 12 /

INITIATION ANO REVIEW Calculation Status Preliminary Pending Final Superseded D D ~ D Initiated Init Review Method Technically Reviewed Revr Rev Appd Appd CPCo I Description By Detail Qual By Appd By Date Alt Cale Review Test By Date f)t;~rk Original Issue / 3/q~

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v v PURPOSE:

The purpose of this EA is to demonstrate that the radiological consequences of a large break LOCA would be within the limits of 10 CFR 100 and 10 CFR 50 Appendix A for current plant configuration. This analysis will account for leakage of recirculated sump water through valves CV-3027 and CV-3056 and takes credit for addition of NaOH to the sump at the beginning of recirculation. The results of this analysis will provide justification for continued operation of the plant until the end of cycle 11, at which time the post-LOCA sump pH control system can be modified to meet current regulatory requirements and any other necessary modifications can be performed to meet the dose limits of 10 CFR 100 and 10 CFR 50 Appendix A.

SUMMARY

oF* RESULTS:

This analysis,~v:aluated the radiological consequences of a large break LOCA for the current plant configurationt~onsidering two cases for the leakage through CV-3027 & CV-3056. For the case of the leakage going directly into the SIRW Tank, the calculated offsite doses are 7.641 rem thyroid and 0.057 rem whole body at the site boundary, and 4.499 rem thyroid and 0.011 rem whole body at the low population zone. For the same case the calculated control room doses are 8.394 rem thyroid, 1.205 rem skin, and 0.463 rem whole body. For the case of the alternate recirculation path routing the leakage to the spent fuel pool tilt pit, the calculated offsite doses are 7.645 rem thyroid and 0.057 rem whole body at the site boundary, and 4.504 rem thyroid and 0.011 rem whole body at the low population zone. For the same case the calculated control room doses are 8.410 rem thyroid, 1.205 rem skin, and 0.451 rem whole body. For both cases, all of the doses are well within the Standard Review Plan & 10 CFR 50 limits of 30 rem thyroid, 30 rem skin, and 5 rem whole body in the control room, and the 10 CFR 100 limits of 300 rem thyroid and 25 tern whole body offsite. This justifies operation of the plant with its current post~LOCA sump pH control system and control room habitability system until the end of Cycle 11.

-1 PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 2 Rev # 0 TABLE OF CONTENTS Section 1.0 Objective .....................................*................................................................................................................... 3 2.0 References ...................................................................................................................................................... 3 3.0 Background ............................................................................................... :..................................................... 6 4.0 Analysis Input ................................................................................................................................................ 8 5.0 Assumptions ................................................................................................................................................... 14 6.0 Analysis ........................................................................................................................................................... 16 6.1 MHACALC Input ...................................................................................................................................... 16 6.2 Offsite Doses ............................................................................................................................................... 21 6.3 CONDOSE Input ....................................................................................................................................... 25

6.4 Control Room Doses ........................................................................................ ~ ........................................ 28 7.0 Summary ......................................................................................................................................................... 31 8.0 Conclusion ....................................................................................................................................................... 33 9.0 Llst of Attachments ...................................................................................................................................... 34

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 3 Rev # --=o__

OFFSITE DOSES~AND CONTROL ROOM HABITABILITY FOLLOWING A LARGE BREAK LOCA TO JUSTIFY CONTINUED OPERATION UNTIL CYCLE 12 1.0 OBJECTIVE The purpose of this EA is to demonstrate that the radiological consequences of a large break LOCA would be within the limits of 10 CFR 100 and 10 CFR 50 Appendix A for current plant configuration.

This analysis will account for leakage of recirculated sump water through valves CV-3027 and CV-3056 and takes credit for addition of NaOH to the sump for pH control at the start of recirculation. The results of this analysis will provide justification for continued operation of the plant until the end of cycle 11, at which time the post-LOCA sump pH control system can be modified to meet current

regulatory requirements and any other necessary modifications can be performed to meet the dose limits of 10 CFR 100 and 10 CFR 50 Appendix A.

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-LOCA Sump pH Control," November 1991.

2.5 NUREG-0800, USNRC Standard Review Plan. Section 6.4 Rev 2, "Control Room Habitability System," ~WY: 1981. Section 6.5.2 Rev 2, "Containment Spray as a Fission Product Cleanup System," Ol'cember 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 EA-P-CRAVS-881028, "Evaluation of Palisades Control Room Radiological Habitability Following the MHA," November 1988.

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

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 4 Rev # -~-

0 2.8 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."

2.9 EMF-91-177, "Palisades Large Break LOCA/ECCS Analysis with Increased Radial Peaking and Reduced ECCS Flow," Siemens Nuclear Power Corporation, October 1991.

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 65.2, Revision 2," December 1991. Cart/Frame: F005/2454. -

2.21 NED0-24782:-o,;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 Lab_oratory, 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 Drawmg 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.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 5 Rev # -~-

0 2.26 EA-A-P.AL-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 HVAC - Fresh Air Make-Up."

2.28 EA-DBD-1.06-01, "Control Room HVAC Isolation Damper Leakage During Emergency Mode,"

August 1990.

2.29 DBD-1.06, "Palisades Design Basis Document - Control Room HVAC 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 1."

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-001 Rev 1.

2.32 EA-PAH-92-01, "Verification of Control Room Atmospheric Dispersion Calculations For Releases From The SIRW Tank, 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.

2.34 ANF-90-079(P) Rev 1, "Mechanical Licensing Report for Palisades High Thermal Performance Fuel Assemblies," Advanced Nuclear Fuels Corporation, September 1990.

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 Co.g::~spondence WLR *004 from WLRoberts to PMDonnelly dated February 5, 1992.

Subject:

"Yalisades 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.

239 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 inLWR Severe Accidents,"

Oak Ridge National Laboratory, Draft Report July 1991.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet _6_ Rev # -~o __

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 dperations," Commonwealth Associates Inc. January 15, *1982.

Cart/Frame: 2687/1277.

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.

3.0 BACKGROUND

The necessity for this analysis revolves around several problems discovered relating to the radiological consequence analysis of the Maximum Hypothetical Accident (MHA). The MHA analysis is just the bounding analysis for the consequences of a LOCA using the source terms of Regulatory Guide 1.4

[Ref. 2.1], which requires the assumption of almost 100 % release of core iodine and noble gas.

Analysis using the Reference 2.1 source terms is required regardless of any plant specific fuel failure analyses for a LOCA The problems discovered with the current MHA analysis [Ref. 2.2] are three fold: a previousiy, un-analyzed potential surrip water leak path outside of containment, failure to consider the imp~~fof not meeting current regulatory requirements for post-LOCA sump pH control in the analysis, a.ii(f"Improper methodology for control room dose calculations.

The potential leak path of sump water outside of containment [Ref. 2.3] involves valves in the mini-flow

. recirculation lines from the safeguards pumps to the Safety Injection & Refueling Water (SIRW) Tank that had never been leak tested. Since the SIR W Tank is vented, iodine entering the tank could be released to the environment increasing the consequences of a LOCA. The increase in the consequences could be intolerable when using the MHA source terms if leakage into the tank is high. Although modifications are being performed during the current outage to leak test CV-3027 & CV-3056 and to eliminate* the leak path through MV-3225 by installing a spectacle flange, no formal MHA analysis to date has accounted for the incre.ased consequences of sump water leakage into the SIRW Tank. At present, an alternate recirculation path has been established in the Emergency Operating Procedures that routes this leakage away from the SIRW Tank, into the spent fuel pool tilt pit where the fuel pool

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 7 Rev # 0

-~-

area filters are pfaced in service to filter iodine released from the leakage [Ref. 2.36, Step 48.g & 1].

For the post-LOCA sump pH control system, inadequacies in the current design were discovered [Ref.

2.4] during an evaluation of the feasibility of retiring T-102, the hydrazine addition (or iodine removal) tank. The original design of the containment spray chemical additive system was for sodium hydroxide (NaOH) to be placed in T-102 and T-103. T-102 was intended to gravity feed NaOH by automatic opening of the discharge valves. T-103 was to be used for post-RAS (recirculation actuation signal)

NaOH addition. During discussions with the NRC [Ref. 2.1], plant staff committed to substituting hydrazine into and pressurizing T-102. These commitments were made because, at the time it was thought that a greater iodine removal efficiency could be achieved with trace levels of hydrazine in the containment sprays. It was also stated in a letter to the NRC [Ref. 2.2] that NaOH addition and a corresponding neutral sump water pH could be achieved within approximately one hour post ..LOCA Prior to discovery of the current problems, it would have taken several hours to initiate and add the required amount of NaOH to the sump. It would have taken several hours since the Emergency Operating Procedures (EOPs) directed NaOH to be added upon Chemistry request, which would be after PASM samples are taken. The Standard Review Plan 6.5.2 [Re[ 2.5] requires that a pH ~ 7 be achieved by the onset of recirculation and maintained throughout the incident. If the pH is not maintained above 7, it must be assumed that volatile iodine evolves into the containment atmosphere.

This evolution of iodine into the containment atmosphere has also not been considered in any formal MHA analyses to date. For this reason, the EOPs are being changed to add a step at RAS to add the .

required amount of NaOH that would adjust the sump pH above 7 for the expected range of sump volumes and boron concentrations.

Problems with the control room habitability analysis methodology occurred in revision 3 of the CRH.AB computer code [Ref. 2.6] that was written to perform all control room habitability calculations. The CRH.AB computer code was revised to use the methodology of ICRP30 [Ref. 2.7], which had not yet been accepted by the NRC. Control room doses were calculated in terms of whole body equivalent, but were recently found to have done so improperly. The whole body equivalent terminology seemed to be consistent with General Design Criterion (GDC) 19, but masked the thyroid dose which has a limit of 30 rem per Standard Review Plan 6.4 [Ref. 2.5]. It appears that the control room thyroid doses, if they had been calculated using the ICRP30 methodology properly, would either have been very close to or exceeded tlie 30 rem thyroid dose limit. The largest contributor to the thyroid dose is unfiltered air inleakage intoc:..the control through the normal intake. Although the NRC has accepted the ICRP30 thyroid dose limiF6f50 rem for occupational exposure in the most recent revision to 10 CFR 20, the NRC has not accepted 50 rem as an acceptable limit for control room habitability after an accident.

When considering the other problems mentioned above, it appeared very questionable whether the control room thyroid dose limit of 30 rem could be met when using the MHA source term with current plant configuration and procedures.

The above mentioned problems result in needed i;nodifications to the plant to leak test valves in recirculation lines leading to the SIRW Tank, to achieve post-LOCA sump pH~ 7 by the onset of recirculation, and possibly to reduce iodine intake into the control room or prevent iodine intake by the operators. Due to the extent of the needed modifications, exemption for performing an analysis using the MHA source terms and considering the above mentioned problems until the end of Cycle 11 has been requested from the NRC [Ref. 2.8]. This allows time to evaluate alternatives for post-LOCA sump pH control and to implement modifications. This also allows time to perform an MHA analysis in

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 8 Rev # 0 accordance with current methodology and regulatory requirements. That MHA analysis will be used to determine the* need for implementing various plant modifications.

This analysis is intended to justify operation without a current MHA analysis on record until the modifications are performed. Temporary relief from using the standard MHA source terms in this analysis to justify continued operation has been agreed to during telephone conversation§ with the NRC.

The source term for this analysis will not be from Reference 2.1, but will be based on the plant specific 10 CFR 50 Appendix K analysis [Ref. 2.9] which justifies that no fuel melting would occur during a large break LOCA The analysis will be performed for two situations: with the current EOPs directing the valve leakage to the spent fuel pool tilt pit, and with the leakage going to the SIR W Tank. Since the alternate recirculation path directing leakage through CV-3027 & CV-3056 to the spent fuel pool tilt pit is being kept in the EOPs at present time, this release path is considered in this analysis. The alternate recirculation path is not being removed from the EOPs due to time constraints with operator training since the requalification exams for the operators are only a few weeks away. However, it is desirable to remove alternate recirculation path from the EOPs to eliminate the additional operator actions.

Therefore, this analysis also considers the leakage going to the SIRW Tank. By also considering the leakage going into the SIRW Tank, the alternate recirculation path can be removed from the EOPs at a later date. The results will be presented for both cases to justify continued plant operation with the leakage through CV-3027 & CV-3056 being routed to either location.

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 SIRW Tank, and to calculate the radionuclide release rates from the containment building and SIRW Tank and the resultant offsite doses. The radionuclide release rates given from the MHACALC code are then used in the CONDOSE code to model the transport of radionuclides into and out of the control room, and to calculate the resultant control room operator doses. Both codes use the methodology and dose conversion factors of ICRP30 [Ref. 2.7] to calculate doses. The MICROSHIELD code is used to calculate the shine dose in the control room from the iodine activity in the SIRW Tank that is given in the output of the MHACALC code. For the ,case of the alternate recirculation path in place with the leakage goinl[-to the spent .fuel pool tilt pit, LOTUS 123 spreadsheets are used since the MHACALC cod~h-_was written such that it specifically models the SIRW Tank. The release rates calculated on sptladsbeets can then be used in the CONDOSE code to calculate the control room doses, and with the leakage going to the tilt pit there would be no expected significant shine dose in the control room.

4.0 ANALYSIS INPUT 4.1 The breathing rates for offsite and control room doses are 3.47E-04 m3 /sec from 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 1.75E-04 m3 /sec from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2.32E-04 m3 /sec from 24 to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months in accordance with Reference 2.1.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 9 Rev # 0 4.2 The basis for no fuel melt occi.Irring from a large break LOCA is the cladding not exceeding 2200 °F from Reference 2.9.

4.3 The rated core thermal power, 2530 MWt, the containment design leak rate, 0.1 %/day, and the containment net free air volume, 1.64E+ 06 ft.3, 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 1.55£-04 sec/m3 and at the low population zone are 1.09E-05 sec/m3 from 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 6.94E-06 sec/m3 from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 2.58E-06 sec/m3 from 24 to 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months , and 6.25E-07 sec/m3 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].

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

4.5 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].

4.6 The iodine reduction factor, 2, for automatic isolation of the ventilation in the safeguards rooms

upon high radiation is from Reference 2.13.

4.7 The sump temperature is taken as the peak after recirculation from the current LOCA containment analysis [Ref. 2.14]. The temperature is 232°F at a containment pressure of 34.33 psia and it occurs at 1550 seconds.

4.8 The time .at which containment pressure is reduced below 3.0 psig is approximately 40000 seconds, or 667 minutes from Reference 2.14.

4.9 The limit for the leak rate through CV-3027 & CV-3056 is 0.1 gpm from Reference 2.15.

4.10 The fraction of the iodine in the sump solution that is in a volatile form is 3.0E-04 [Ref. 2.1~,

Attachment 6].

4.11 The total. SIRW Tank volume is calculated from the dimensions given and measured from Reference~;2~ 17~.

4.12 The volunf~Mcpipe leading to the SIRw Tank from the potential leaking valves is calculated from the dimensions from Reference 2.18.

4.13 The volume of water present in the SIR W Tank after RAS is calculated based on the low level setting of 24" from Reference 2.19.

4.14 The efficiency of the fuel pool area filters is 99 % for inorganic iodine and 94 % for organic iodine [Ref. 2.12, Table 4.2.3].

4.15 The first-order spray removal coefficients for particulate iodine are 4.43 ht1 initially, changing to 0.443 ht1 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.

(@ consumers Power NWElllllS MIClflU#'S NOlilUSI L

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-02 Sheet 10 Rev # -~-

0 4.16 The long term iodine partition coefficient in containment is conservatively estimated as 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.

4.17 The activity source term values (~) 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.)

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

4.19 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. They were taken from Reference 2.7 and converted to units of (Rem/sec)/(Ci/m3 ). The thyroid and whole body inhalation dose conversion factors for iodine isotopes are listed in Table 2. They 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 m3 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 for the radionuclide.

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

& 2.25.

4.21 From Reference 2.26, when loss of qffsite 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 un-its V-95 and V-96 [Ref. 2.26, pg. 13].

4.22 The time to pressurize the contr61 room to the required 0.125" WG (water gauge) after start of V-95 and. \{;96,is 7.34 seconds from data taken by LTPhillips as documented in Attachment 1.

4.23 The fresh ait:'make-up flow rate for the control room HVAC emergency mode has been changed

to 1000 cfm as documented in Reference 2.27.

4.24 The rate of unfiltered air inleakage into the control room across the normal intake isolation dampers when in emergency mode is calculated to be 11.6 cfm from Reference 2.28.

4.25 The total filtered air flow rate of the control room HVAC when in emergency mode _is 3200 cfm

[Ref. 2.29, Appendix A].

4.26 The efficiency of the control room HVAC HEP A/ charcoal filters for the remote intake air and recirculation air is 99 % for iodine [Ref. 2.12, Table 4.2.3].

J

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 11 Rev # _o=----_

TABLE 1 ISOTOPE DEPENDENT PARAMETERS Dose Rate Conv Factors

~ J.., (semi-infinite cloud)

Nuclide (Ci/MWt) (min-1 )

Thyroid Whole Body Kr-83m 2.998E+03 6.313£-03 O.OOOE-00 3.649E-06 Kr-85m 6.498E+03 2.579E-03 3.083E-02 3.031E-02 Kr-85 2.999E+02 1.230E-07 O.OOOE-00 4.738E-04 Kr-87 l.155E+04 9.084E-03 l.439E-01 l.447E-01 Kr-88 l.690E+04 4.068E-03 3.803E-01 3.690E-01 Kr-89 l.993E+04 2.194E-01 O.OOOE-00 O.OOOE-00 Xe-131m 1.760E+02 4.065E-05 O.OOOE-00 l.324E-03 Xe-133m l.954E+03 2.198E-04 O.OOOE-00 5.375E-03 Xe-133 5.648E+04 9.177E-05 7.297E-03 6.259E-03 Xe-135m l.698E+04 4.513E-02 O.OOOE-00 7.647E-02 Xe-135 9.781E+03 1.268E-03 O.OOOE-00 4.676E-02 Xe-137 4.705E+04 1.SlOE-01 O.OOOE-00 O.OOOE-00 Xe-138 4.433E+04 4.906E-02 1.953E-01 1.969E-01

. I~131 2.938E+04 5.987E-05 N/A N/A L-132.

,;:.7::-,...,~ -

4.160E+04 5.023E-03 N/A N/A 1-133 4.808E+04 5.554£-04 N/A N/A 1-134 - 6.218E+04 1.318£-02 N/A N/A 1-135 4.922E+04 1.748E-03 N/A N/A

TABLE 2

*i<0 IODINE DOSE CONVERSION FACTORS & ROOM-CORRECTED NOBLE GAS DOSE RATE CONVERSION FACTORS Nuclide Kr-83m Dose Rate Conversion Factors Corrected for a 1000 m"' room {Rem/sec)/{Ci/m"')

Thyroid Lungs O.OOOE-00 ,. Q.OOOE-00 B Surface 6.475E-06 B Marrow 5.653E-6 Skin l.747E-04 Eye Lens l.747E-04 Whole Bod 3.649E-06 IPI

~r"~';. Y* ,,

Kr-85m l.233E-03 :1 l .131E-03 l.953E-03 l.850E-03 5.139E-02 l.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

?i: "'C

)lo Kr-88 l.439E-02 l.336E-02 1. 542E-02 l.336E-02 9.456E-02 2.878E-02 1.402E-02 r- )>

Kr-89 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 O.OOOE-00 "'

-V .....l en l>

Cl n ,,,

Xe-13lm O.OOOE-00 1. 233E-04 3.597E-04 3.289E-04 1. 542E-02 5.344E-04 l.915E-04 OVl z

-I :z Xe-133m O.OOOE-00 2.775E-04 6.167E-04 5.653E-04 3.083E-02 7.503E-04 3.823E-04 -c zn c:r-Xe-133 4.317E-04 2. 672E-04 6.886E-04 6.269E-04 l.131E-02 6.989E-04 3.361E-04 l:lolTl

-1)>

0- ;:o Xe-135m O.OOOE-00 3.392E-03 4.522E-03 4.214E-03 2.672E-02 4.522E-03 3.618E-03 z "'C r-(11 )>

Xe-135 O.OOOE-00 1.850E-03 2.981E-03 2.775E-03 6.578E-02 2.467E-03 2.086E-03 :CZ

,.,, -I 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 3.186E-02 7.801E-03 Inhalation Dose Conversion Factors (Rem/Ci-inhaled) 1-131 l.073E+06 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 3.256E+04 Vl m O.OOOE+OO :::r >

CD I CD )>

I-132 6.290E+03 9.990E+02 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 3.367E+02 C"+

z I

r-I-133 l.813E+05 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 5.550E+03 ...... I N '° N

I I-134 l.073E+03 5 .180E+02 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO l.106E+02 0

o N 1-135 3 .145E+04 l.628E+03 O.OOOE+OO O.OOOE+OO O.OOOE+OO O.OOOE+OO 1.121E+03 CD

< 0 I

N

~

0

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 13 Rev # 0

--=---

4.27 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].

4.28 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 SIRW Tank vent are from Reference 2.32.

TABLE 3 ATMOSPHERIC DISPERSION FACTORS FOR CONTROL ROOM INTAKES Release From SIR W CTMT or ESF Room Release From TIME Tank Vent x/Q (sec/m3 ) Stack Location x /Q (sec/m3 )

(Hrs)

Normal Remote Normal Remote Aux. Bay 0-8 1.14£-02 8.06£-04 1.84£-03 1.22E-03 3.85£-03 8-24 8.lOE-03 5.60£-04 l.65E-03 1.02E-03 3.50£-03 24-96 6.38£-03 4.35E-04 l.30E-03 6.SOE-04 2.85E-03 96-720 4.60£-03 2.85E-04 9.00E-04 3.98E-04 2.15E-03 4.29 The whole body dose from containment shine at the highest dose point in the control room is obtained: fr()m: References 2.45 & 2.46. The highest dose point is at the access doors to the control roqµ:Lyiewing gallery, which receive a dose of 400 mrem (0.4 rem) over 30 days [Ref.

2.45, pg. 20]f"'"tJsing this value is conservative since the highest dose point that is actually in the control room is in the southwest comer of the room which receives 13.3 mrem over 30 days [Ref.

2.46, pgs. 5-6]. This shine dose is also based on the Regulatory Guide source term.

4.30 From Reference 2.17, there is a 1" minimum of sand and a 10" concrete pad under the SIRW tank.

4.31 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.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 14 Rev # -"""'"o__

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 MWt.

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

5.5 It is assumed that 100 % of the core experiences DNB and consequently has cladding failure, but no fuel melting occurs per the predictions of Reference 2.9.

5.6 The fraction of iodine and noble gas assumed ro be in the pellet-clad gap is 20 %, whereas peak

predicted fission gas release is approximately 18 % from Reference 2.34 [Figure 5.12].

5.7 Remaining consistent with SRP *15.6.5 Appendices A & B [Ref. 2.5], 100 % of the noble gas released from the fuel will be assumed to be released to the containment atmosphere. For iodine released from the fuel, 50 % will conservatively be assumed to be released to the containment atmosphere, and 50 % will be assumed to be released to the sump solution.

5.8 Again following SRP 15.6.5 Appendix A [Ref. 2.5] and adding conservatism, 91 % of the iodine released to the containment atmosphere is assumed to be elemental and 9 % is conservatively assumed to be organic with 0 % particulate since only pellet-clad gap release occurs. For evaluating leakage through CV-3027 & CV-3056 into the spent fuel pool tilt pit, the sump water is assumed to be composed of 99 % elemental iodine, with 1 % organic [Ref. 2.41, pg. 30].

5.9 Loss of offsite power occurs coincident with the event.

5.10 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 is less tha.nLt~i:ninute for all cases of available spray pumps [Ref. 2.35].

-'*E~*

5.11

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

5.12 The recirculation volume is at minimum.

5.13 Operators terminate containment sprays after containment pressure is reduced to less than 3.0 psig following Reference 2.36, conservatively assuming the decision to terminate them is not based on iodine activity.

5.14 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 f~ctor 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

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 15 Rev # --=o__

seems conServative and was informally agreed to as acceptable in telephone discussions with the NRC as documented in Reference 2.37.

5.15 Iodine re-evolution from the sump solution is insignificant since the post-RAS pH is controlled to a value of 7 by addition of NaOH.

5.16 The source of radionuclide ingress into the control room when the control room is depressurized is through the Aux. Bay roof doors [Ref. 2.30].

5.17 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 11 pressurized to Ya water gauge plus the leakage across the normal intake isolation dampers when in emergency mode. The leakage from the control room when pressurized to Ya water gauge is 11 equal to the make-up air flow rate to maintain steady pressurization. This results in an air infiltration rate of (1000/2) + 11.6 = 511.6 cfm [Refs. 2.27 & 2.28]. A contribution from opening and closing doors does not need to be accounted for since vestibules have been installed on the entrances Reference 2.5 [SRP 6.4, pg. 6.4-9].

5.18 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.

11 5.19 Control room shine dose from the 6 piping leading to the SIRW Tank is insignificant compared to the shine from the tank since for the shine dose calculation all of the iodine activity in the tank water and air volumes and in the piping is assumed to be in the tank's water volume.

5.20 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.

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

5.22 For calculations of shine dose from the SIRW Tank, the dose point is taken as 6' off the control room floorj,*aSsuming the average operator is 6' tall and standing under the tank for the 30 days.

5.23 The concrete~T>etween 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 MICROSHIELD.

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

5.25 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.

5.26 For consideration of the* alternate recirculation path to the spent fuel pool tilt pit, the release rates of :iodine will be conservatively based on the sump activity at specific points in time and assumed to remain constant until the next point in time of interest. This is conservative since the sump iodine concentration is continuously decreasing with time due to radioactive decay.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 16 Rev # -~-

0 6.0 ANALYSIS 6.1 MHACALC INPUT The inputs for the MHACALC code are described in detail below. The structure for the input deck is obtained from Reference 2.16.

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).

The fourth line of the input deck contaills the reactor thermal power, containment design leak rate, and the recirculation water volume. The reactor thermal power will be taken as 102 % of the rated value of 2530 MWt [Ref. 2.10], which is 2580.6 MWt. The containment design leak rate is 0.1 %/day [Ref.

2.10], 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 lbm at 25.4 psia and 238.3°F [Ref. 2.11], the containment sump volume is taken as a minimum of 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. Since the melting temperature of the fuel is not predicted to be exceeded [Ref. 2.9], it is assumed that all of the core experiences DNB with subsequent cladding failure. Reference 2.34 predicts that the peak fission gas release to the pellet-clad gap of the peak rod will be less than 20 %, so 20 % is assumed for conservatism. Following References 2.1 & 2.5, 100 % of the noble gas released is assumed to be released to the containment atmosphere, which for this case is 20 % of the total noble gas core inventory. For iodine, the regulatory guidance is based on 50 % of the iodine being released to the containment atmosphere with 25 %

plateout, and 50 % of the iodine being released to the containment sump. For this case, initial plateout will be conservatively ignored so that 50 % of the iodine released is to.the containment atmosphere and 50 % is to the sump, which corresponds to 10 % of the total iodine core inventory released to each since 20 % of the total is released.

The sixth line of the input deck contains the fractions of the iodine released to the containment atmosphere thatJs,released in elemental, particulate, and organic forms. References 2.1 & 2.5 give the percentages as 9{.o/it elemental, 5 % particulate, and 4 % organic. However, those percentages are based on fuel m~t'occurring. For pellet-clad gap release, no particulates would be expected.

Reforence 2.38 gives gap release as being composed of 99.75 % inorganic and 0.25 % organic. For conservatism since organic iodine is not removed by containment sprays, it is assumed the 91 % of the iodine released to the containment atmosphere is in elemental form and 9 % is in organic form.

The seventh line of the input deck contains: the retention factor for iodine released into the safeguards rooms, the iodine partition coefficient for the sump water leakage into the safeguards rooms, the iodine partition coefficient for sump water leakage into the SIRW Tank, the multiplication factor for the rate at which iodine is released from the SIRW Tank, the total volume of the SIRW Tank and recirculation line leading to it, the volume of water in the SIRW Tank at RAS and recirculation line leading to it, and the percentage of iodine in the sump water reaching the SIRW Tank that would be 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 in those rooms automatically isolates upon high radiation. This

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 17 Rev # -~-

O retention factor is used to account for plateout of iodine onto surfaces in the safeguards rooms since the ventilation rates would be very low after isolation.

For the iodine partition coefficient in the safeguards 'rooms Reference 2.5 [SRP 15.6.5, Appendix B]

states that the fraction of iodine released to the rooms from ESF leakage should be taken as 10 %

(partition coef. = 1/.1 = 10) or the flashing fraction, whichever is greater, unless a higher value can be justified based on actual sump pH. Using the maximum 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 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 SIRW Tank, is conservatively ignored for the ESF leakage since the flashing fraction would be controlling the amount of iodine released for a period of time. For use in the spreadsheets when evaluating leakage through CV-3027 & CV-3056 with the alternate recirculation path in place, the same iodine partition .coefficient is applicable since the leakage will be going into the spent fuel pool tilt pit which would not contain water and a fraction would be flashing to steam for a period of time.

The third value on line 7 of the input deck is the iodine partition coefficient for the SIR W Tank. The iodine partition coefficient for the volatile specie of .iodine expected in the SIRW Tank can be calculated using the following equation from Reference 2.41 [pg. 29], for which the applicability is verified in Reference 2.16 [Ref. 2.16, Attachment 6]:

log 10 P = 6. 29 - o. 0149 T (1) where p = ratio of 12 in liquid phase to I 2 in the gas phase, or the iodine partition coefficient T = air-water interface temperature, °K.

To bound the case,* since the SIRW Tank sits outdoors and the iodine partition coefficient decreases with increasing te~p~rature; the interface temperature is assumed to be 100°F, corresponding to 310.8°K. Using Equation (1) with this temperature results in an iodine partition coefficient of 45.6 for the volatile iodine in the SIRW Tank.

The fourth value. on line 7 is a multiplication factor for the rate at which iodine in the air volume of the SIRW Tank exits through the vent. The SIR W Tank is modeled such that iodine in the air volume exits the tank at the rate at which air is displaced due to the water leaking in. The multiplication factor is added to conservatively encompass any possible diffusion out the vent or density changes that force air out of the vent. This factor is chosen 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 two. This factor was informally agreed to as acceptable during telephone conversations with the NRC, as discussed in Reference 2.37.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 18 Rev # O

-~-

The SIRW Tank rotal volume and volume at RAS, including the recirculation line leading to the tank are now calculated. The volume of the line between CV-3056 and the SIRW Tank is first determined using Reference 2.18 to calculate the length of pipe, Reference 2.40 to determine the pipe schedule, and Reference 2.42 to determine the inner diameter of the pipe. As can be seen from the below listed calculations, the volume of pipe between CV-3056 and the SIRW Tank is 23.678 ft 3

G/cvltiti'il'1 cl fli'e ~/v.,.,e ~efw4eA CV-.JQS'

Sl.RW 7iv1..t NC -1 J-" tf S*l~S lfd'. .Z38} J'G/o/e.s3 .SZ-e.a/

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. /~'-(gH .3~.S"'

,~ , .i..v 11*

7J,.('IS 0) ~ I - '¥ $/_.(~' ~ /~""" H El. ""'-~N Ve"t)01I: 6"'

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/~~! 2'-6"'

2'-IJ* £/, '.J2!~"'

2'-sH H~ll';zo,,(4/: 6 "-t9.1" El. '""'- 3 h~"' 1'-41"'

lllJ'iZIJllt4/: 1 '- 41 q K!rZ'l~I: 7!..'1~,..

11'-3"'

'111~/!' ,~_ .....

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 19 Rev # -~-

0 ~

The total 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 1tr 2 L = 39885.66 ft 3

From Reference 2.17, the conical top section of the tank has a slop of 1-4" 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 equation given in Reference 2.43 [pg. 18], is Ya7t r 2 h = 3185.313 ft3. The total volume of the SIR W Tank and recirculation line leading to it is then 23.678 + 39885.66 + 3185.313 = 43094.7 ft3. 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" is 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 as the volume of water in the SIRW Tank and piping after RAS.

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 Attachment 6 of Reference 2.16, the fraction of iod1.1e that would be in volatile form is 3.0Em04 since the sump pH will be controlled at or above 7.

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 m3 /sec, respectively. The same breathing rates are to be used in the spreadsheets for evaluating the release from the spent fuel pool tilt pit for the alternate recirculation path.

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. These atmospheric dispersion factors are also to be used in the spreadsheets for evaluating the release from the spent fuel pool tilt pit for the alternate recirculation path.

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 partj.culate iodine. The code automatically changes to the long term removal coefficient after 98 % of tb~rn~iculate iodine has been removed from the containment atmosphere. Although no iodine is assum~; to be in particulate form for this analysis, the values of. the spray removal coefficients are sPfcmed anyway. These values are 4.43 ht1 and 0.443 ht1 from Reference 2.20.

Lines 12 through 16 of the input deck are for specifying times at which spray removal or re-evolution of elemental iodine starts or changes, and for the corresponding removal or re-evolution coefficients.

Only one value for elemental iodine spray removal is used in accordance with SRP 6.5.2 [Ref: 2.5], and 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 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 O since no iodine re-evolution from the sump is assumed with the addition of NaOH at RAS. . .

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 20 Rev # -~-

O The maximum iodine decontamination factor for elemental iodine and the time at which containment sprays are termirtated 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]:

VH5 DF= 1 + - - (2) vc where = containment sump liquid volume, ft 3

= containment net free volume less V5 , ft3

= effective iodine partition coefficient.*

The iodine partition coefficient is a function of sump solution temperature, iodine concentration, and pH and can be obtained from Reference 2.44. The pH of the sump is taken as 7 since the required amount of NaOH to achieve this is to be added at RAS. Since partition coefficient decreases with increasing iodine concentration, the SRP 15.6.5 [Ref. 2.5] assumption of 50 % of the core iodine inventory being released to the sump will be used even though only the pellet-clad gap fraction is released for this analysis. From Reference 2.21, which gives the iodine source term as 50 % of the total release, the totaI iodine amount in the sump would be 0.03062 g-at./MWu which is 79.018 g-at for 2580.6 ~- Dividing by the sump volume, 40324.6 ft3 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. Considering the temperature and iodine concentration, a point between the two plots on Figure 6 of Reference 2.44 would be appropriate. For coruervatism, 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 ft 3 and the containment net free air volume of l.64E+06 ft3 [Ref. 2.10], results in a maximum decontamination factor of 32.51.

The time at which containment sprays are terminated, on line 17, is 667 minutes. *This is based on the containment pressure as obtained from Reference 2.14, and the emergency operating procedures [Ref.

2.36]. From Reference 2.36, operators are instructed to terminate containment sprays when the containment pressure falls below 3.0 psig unless needed for fodine removal. It will be conservatively assumed that the. sprays are terminated as soon as containment pressure falls below 3.0 psig, which from Reference 2.14, iS~iat approximately 667 minutes.

The times at whiaYEsF leakage into the safeguards rooms begins and/ or changes, the corresponding ESFleak rates, the times at which sump water leakage into the SIRW Tank begins and/or changes, and the corresponding leak rates through CV-3056 or CV-3027 are listed on lines 18 through 21. On line 18, the time at which ESF leakage starts is 19 minutes (minimum time to RAS) and the corresponding Technical Specification leak rate is 0.2 gpm [Ref. 2.12, TS 4.5.3]. The time for leakage into the SIRW Tank to begin on line 18 is also 19 minutes, and the corresponding leak rate is 0.1 gpm [Ref. 2.15]. All values on lines 19 through 21 of the input deck are specified as 0 since the ESF and SIRW Tank leakage is assumed to remain constant after RAS throughout the incident. It should be noted that the 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

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 21 Rev # -~-

0 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 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 11 time intervals are used for the release rates. These time intervals are specified on lines 41 and 42. The times for the intervals to were chosen such that a new time interval begins at each point in time that the relea.Se 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; 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 Tan.k 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, 667.0 minutes; the time at which 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 SIR W activities of each of the radionuclides are to be printed in the output file. This value is specified as 15.

Lines 44 and 45 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, 60, 120, 480, 720, 1440, 2880, 4320, 5760, 7200, 14400, 21600, 28800, 36000, and 43200 minutes.

The input deck constructed with the above listed parameters is listed on the attached microfiche under the filename LOCAJCO DATA 6.2 OFFSITE DOSES The MHACALC'code, as described in Reference 2.16, was executed using the LOCAJCO DATA input deck in VMS. Three, output files from the program execution are listed on the attached microfiche:

LOCAJCO USTING; JCOSTACK DATA, and JCOSIRW DATA The fourth output file from the code- for creating plots of the containment atmosphere iodine activity versus time was discarded since it was not of interest for this analysis. The LOCAJCO 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, including the contribution from the SIRW Tank release. The JCOSTACK DATA file contains the radionuclide release rates from the containment atmosphere and ESF leakage for the inpu.t specified time intervals. The JCOSIRW DATA file contains the radionuclide release rates from the SIRW Tank for the input specified time intervals.

For consideration of the leakage through CV-3027 & CV-3056 going to the SIRW Tank, all of the resultant offsite doses at the site boundary (SB) and low population zone (LPZ) from the incident are obtained from the LOCAJCO 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 EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 22 Rev # _....:,o__

the thyroid doses-irom inhalation and the whole body doses only. The other listed doses incorporate the ICRP30 methodology of adding contributions for internal and external doses to obtain dose equivalents. As can be seen on the attached microfiche, the resultant offsite doses from the incident from each release path and the total are listed in Table 4 below, assuming the leakage through CV-3027

& CV-3056 goes into the SIRW Tank. These doses are well within the 300 rem thyroid and 25 rem whole body limits of 10 CFR 100. As also can be seen, allowing the leakage through CV-3027 & CV-3056 to go into the SIRW Tank results in negligible doses since the iodine is diluted in a large amount of water and only a small amount is in a volatile form.

TABLE 4 SITE BOUNDARY AND LOW POPULATION ZONE DOSES Ctmt

ESF SIRW Total Leakage Leakage Leakage Release (Rem) (Rem) (Rem) (Rem) 0-2 Hr SB Thyroid 6.833 0.808 -0 7.641 Whole Body 0.057 N/A N/A 0.057 0-30 Day LPZ Thyroid 3.676 0.822 -0 4.499 Whole Body 0.011 N/A N/A 0.011 For calculation of the offsite doses with the alternate recirculation path in the EOPs, the doses from the containment atmosphere leakage and the ESF leakage listed in Table 4 will be added to the calculated resultant doses from iodine releases from the spent fuel pool tilt pit. To calculate the doses due to releases from the spent fuel pool tilt pit, the sump activities given in LOCAJCO LISTING are used. At each point in time that the sump activities are listed in LOCAJCO LISTING, the sump iodine concentration is calculated by dividing the sump activity by the sump volume. To account for the ESF leakage, which is multiplied by a factor of 2 as mentioned previously, and the leakage into the tilt pit, the sump volume as it changes with time is represented by the following equation:

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 23 Rev # _....;:.o__

where Vs(10) = initial sump volume = 40324.6 ft 3 LResF = ESF leak rate = 0.2 gpm LRvALve = leak rate through CV-3027 or CV-3056 = 0.1 gpm The calculated iodine concentrations in the sump at each point in time are shown in Table 5.

TABLE 5 SUMP SOLUTION IODINE CONCENTRATIONS 19 min 60 min 120 min 480 min 720 min 1440 min 2880 min 4320 min (Ci/ft') (Ci/ft3) (Ci/ft') (Ci /ft'} (Ci /ft3) (Ci/ft') (Ci /ft3) (Ci /ft3) 1-131 187.80 187.34 186.67 182.68 180.07 172.48 158.25 145 .17 1-132 241.99 196.94 145.69 23.89 7.16 0.19 0.00 0.00 1-133 304.53 297.61 287.96 235.69 206.27 138.28 62.14 27.92 1-134 309.74 180.45 81.82 o. 71 0.03 0.00 0.00 0.00 1-135 304.78 283.72 255.47 136. 13 89.48 25.43 2.05 0.17 5760 min 7200 min 14400 min 21600 min 28800 min 36000 min 43200 min (Ci/ft') (Ci/ft3) (Ci/ft') (Ci/ft') (Ci/ft') CCi /ft') (Ci /ft')

1-131 133.17 122.17 79.39 51.59 33.51 21.79 14.* 16 1-132 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1-133 12.55 5.64 0.10 0.00 0.00 0.00 0.00 1-134 0.00 0.00 0.00 0.00 0.00 0.00 0.00 1-135 0.01 o.oo 0.00 0.00 0.00 o.oo 0.00 To calculate the rel~ase rates from the fuel pool area, assuming all of the iodine that becomes airborne exits directly to tb:~l~p.vironment, the chemical forms of the iodine, the fuel pool area filter efficiencies, and the iodine partitfon coefficient need to be accounted for. For consideration of the iodine chemical forniS, Reference 2.41 [pg. 30] indicates that generally, less than 1 % of the iodine will be converted to an organic form. Although this conversion was based on gas phase reactions, it is conservatively applied to sump solution, even though most of the organic iodine produced would evolve into the containment atmosphere. Therefore, 99 % of the iodine released from the spent fuel pool tilt pit is assumed to be elemental, with 1 % organic. The iodine partition coefficient to account for the fraction of the iodine in the leakage that is released is assumed to be 47.6, which is the same as that for the ESF leakage since the flashing fraction would again be limiting for some period of time. The fuel pool area filter efficiencies are 94 % for organic iodine (methyl iodide) and 99 % for all other forms per the plant Technical Specifications [Ref. 2.12, Table 4.2.3]. By assuming that the iodine in the leakage comes out of solution instantaneously, which would happen if water is flashing to steam, the release rate of iodine to the fuel pool area is the sump solution concentration multiplied by the leak rate through CV-3027

& CV-3056 and divided by the partition coefficient. For the release rate from the fuel pool area to the

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 24 Rev # 0 environment, the fraction of the iodine in each chemical form and the filter efficiencies for the chemical forms are taken into account as shown in the following equation:

q. ( t) = C( t) LRvalve [

pp

)

foxg(l -1'loxg + finoxg(l - 1'linoxg)]

(60. min) (1E+06 µC.i) (3)

Hr Ci where q = release rate of each iodine isotope, µCi/Hr c = concentration of each iodine isotope in the sump solution from Table 5, Ci/ft 3 LR* = leak rate through CV-3027 & CV-3056 = 0.013368 ft 3 /min PF = iodine partition coefficient = 47.6

~ = fraction of iodine in organic form = 0.01

~g = fraction of iodine in elemental form = 0.99 11org = fuel pool area filter efficiency for organic iodine = 0.94

,, lnorg = fuel pool area filter efficiency for elemental iodine = 0.99 The release rates, calculated using Equation (3), are shown in Table 6 for each of the times that the sump solution iodine concentrations are shown in Table 5. The release rates are conservatively assumed to be constant from each listed point in time to the next, with the last listed release rate at 36000 minutes remaining constant until 43200 minutes (30 days).

TABLE 6 IODINE RELEASE RATES FROM SPENT FUEL POOL TILT PIT 19 min 60 min 120 min 480 min 720 min 1440 min 2880 min C11Ci/hr) C11Ci/hr) C11Ci/hr) C11Ci/hr) C11Ci /hr) C11Ci/hr) C11Ci/hr) 1-131 3.323E+04 3.315E+04 3.303E+04 3.232E+04 3.186E+04 3.052E+04 2.800E+04 1-132 4.281E+04 3.484E+04 2.578E+04 4.226E+03 1.266E+03 3.402E+01 2.457E-02 1-133 S.388E+04 5.266E+04 5.095E+04 4.170E+04 3.649E+04 2.447E+04 1.099E+04 1-134 5.480E+04 3.193E+04 1.448E+04 1.259E+02 5.324E+OO 4.028E-04 2.204E*12 1-135 S.392E+04 S.020E+04 4.520E+04 2.408E+04 1.583E+04 4.499E+03 3.629E+02 4320 min 5760 min 7200 min 14400 min 21600 min 28800 min 36000 min C11Ci/hr) C11Ci/hr) C11Ci/hr) C11Ci/hr) C11Ci/hr) C11Ci/hr) C11Ci/hr) I 1-131 2.568E+04 2.356E+04 2. 162E+04 1.405E+04 9.128E+03 5.930E+03 3.855E+03 1-132 1.775E*05 1.282E*08 9.325E*12 0.0 0.0 0.0 0.0 1-133 4.491E+03, 2.221E+03 9.982E+02 1.830E+01 3.356E-01 6.156E*03 1.128E-04 1-134 1.318E~20' o.o 0.0 0.0 0.0 0.0 0.0 I *135 2.928E+01 2.363E+OO 1.906E*01 6.522E*07 2.275E*12 7.635E-18 0.0 Since only iodine is being released from the leakage into the spent fuel pool tilt pit, only the thyroid doses for the site boundary and low population zone need to be calculated. To calculate the offsite thyroid dose over a time interval, the total release over the time interval (release rate times the length of the time interval) is multiplied by the breathing rate, the iodine inhalation dose conversion factor for the thyroid, and the atmospheric dispersion factor [Ref. 2.16, Equation 26]. The breathing rates, atmospheric dispersion factors, and the iodine thyroid inhalation dose conversion factors are the same

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 25 Rev # 0 as those used in the LOCAJCO DATA input deck. The breathing rates and atmospheric dispersion factors are also changed at the appropriate points in time. The calculated resultant thyroid doses due to the alternate recirculation path directing the leakage through CV-3027 & CV-3056 to the spent fuel pool tilt pit are 0.004 rem at the site boundary and 0.005 rem at the low population zone. The total doses from a LOCA with the alternate recirculation path in place are calculated by adding these calculated values to those shown in Table 4 for the containment atmosphere leakage and ESF leakage.

This results in 7.645 rem thyroid and 0.057 rem whole body at the site boundary, and 4.504 rem thyroid and 0.011 rem whole body at the low population zone. These doses for the alternate recirculation path in place are slightly higher than if the leakage through CV-3027 & CV-3056 is allowed to enter the SIRW tank because a higher fraction of the iodine in the leakage is assumed to be released, mainly from part of the leakage flashing. However, as can be seen, these doses are also predicted to be well within the limits of 10 CFR 100.

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 from the containment plus safeguards rooms and for the release rates from the SIRW Tank are given as output from the MHACALC code: JCOSTACK DATA, and JCOSIRW DATA. An input deck with the radionuclide release rates from the spent fuel pool tilt pit for the alternate recirculation path is constructed from the data presented in Table 6. The input deck with the physical parameters for the CONDOSE code is described by line below. However, separate input decks must be constructed for the release from the containment plus safeguards rooms (assumed release point is the stack) and the release from the SIRW Tank since different atmospheric dispersion factors are used for each. A separate input deck for the release from the spent fuel pool tilt pit is also constructed. The structure for the input decks is obtained from Reference 2.33.

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 dimensions of the control room is 46'x 48'x 10' which yield 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 HVAC 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 ft3. This volume, which converts to 1017 m3 , also justifies the use of

.noble gas submersion dose rate conversion factors corrected for a 1000 m3 room as given in Table 2.

The third line of the input deck is the number 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

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 26 Rev # 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 to be 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 safeguards rooms has the atmospheric dispersion factors for the stack to the control room air intake locations as listed in Table 3, since the release point is assumed to be the stack. The input deck used for the release from the SIRW Tank has the atmospheric dispersion factors for 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 only air inleakage then is through the normal intake. The input deck used for the release from the spent fuel pool tilt pit also has the atmospheric dispersion factors for the stack to the control room air intake locations since the fuel pool area exhausts through the stack, with the factors for the unfiltered air inleakage location set to 0.0 since the leakage through CV-3027 & CV-3056 does not start until 19 minutes and the only air inleakage then is through the normal intake.

The number of points in time for which the control room HVAC 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 the control room is depressurized due to loss of offsite power, and the flow rates once pressurization in the emergency niode 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 HVAC 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 inleakage flow rate respectively. At time 0.0 minutes, 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 unfiltered air inleakage while the control room is depressurized is assumed to be 511.6 cfm, as discussed in section 5.0 of this analysis.

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 ~ 1/a" water gauge (from Attachment 1). The corresponding flow rates on line 18 are 11.6 cfm inleakage across the normal intake isolation

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 27 Rev # -~--

O damper [Ref. 2.28], 1000.0 cfm make-up air flow from remote intake [Ref. 2.27], 2200.0 cfm recirculation flow rate since the total emergency mode flow rate is 3200.0 cfm [Ref. 2.29], 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 HEPA filters.

On line 21 of the input deck is the number of points in time that the radionuclide concentrations and the accumulated doses in the control room 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 in,terest.

Line 24 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 11 in the input decks for the release from the containment plus safeguards rooms and for the release from the SIRW Tank, corresponding to the value on line 40 of the input deck for the MHACALC code. Lines 25 and 26 are the beginning and end times of the time intervals. In the input decks for the release from the containment plus safeguards rooms and for the release from the SIRW Tank, these lines correspond to the values on lines 41 and 42 of the input deck for the MHACALC code. These values are 0.00, 1.00, 1.26, 12.00, 19.00, 480.00, 667.00, 1440.00, 5760.00, 14400.00, 28800.00, and 43200.00 in the input decks for the release from the containment plus safeguards rooms and for the release from the SIRW Tank. In the input deck for the release from the spent fuel pool tilt pit for the alternate recirculation path in place, the value on Line 24 is 15, corresponding to an initial release rate of 0 until leakage through CV-3027 & CV-3056 begins plus each of the 14 release rates listed in Table 6. The values on lines 25 and 26 in the input deck for the release from the spent fuel pool tilt pit are 0.00, 19.00, 60.00, 120.00, 480.00, 720.00, 1440.00, 2880.00, 4320.00, 5760.00, 7200.00, 14400.00, 21600.00, 28800.00, 36000.00, and 43200.00.

Line 27 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 in the input decks for the release from the containment plus safeguards rooms and for the release from the SIRW Tank since that is the number of radionuclides that the MHACALC code considers. The number of radionuclides is 5 in the input deck for the release from the spent fuel pool tilt pit since only iodine isotopes are released from the leakage. The identifier for the units of the release rates is 1 to specify µCi/Hr [Ref. 16] in all three of the input decks.

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 28. For this analysis, only the radionuclides for which release rates are given are considered. Therefore, the value on line 28 is 18 in the input decks for the releruie from the containment plus safeguards rooms and for the release from the SIRW Tank. The value on line 28 of the input deck for the release from the spent fuel pool tilt pit is 5.

MKlllliHlS l'llO&IUSI PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-02 Sheet 28 Rev # -~-

0 Lines 29 through 100 of the input decks for the release from the containment plus safeguards rooms and for the relea5e from the SIRW Tank contain the radionuclide constant data. In the input deck for the release from the spent fuel pool tilt pit, the lines are 29 through 48. The radionuclide data must be given in the same order as the radionuclides are given for the release rates, which is the same order as 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 1 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 29 through 80. For the iodine isotopes, on lines 81 through 100, 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, 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. In the input deck for the release from the spent fuel pool tilt pit, only the data for the 5 iodine isotopes are given, in ascending order.

The three input decks constructed with the above listed parameters are listed on the attached microfiche under the filenames CRJCOl DATA for the containment plus safeguards rooms release, CRJC02 DATA for the SIRW Tank release, and CRJC03 DATA for the spent fuel pool tilt pit release.

Since an input deck for the CONDOSE code containing the iodine release rates from the spent fuel pool tilt pit was not created as part of the output from the MHACALC code, a separate input deck is

_needed. The input deck is constructed from the release rate data in Table 6 for the time intervals on lines 25 and 26 of the CRJC03 DATA deck. The file is 15 lines long, with one line of data for each of the time intervals listed on lines 25 and 26 of the CRJC03 DATA deck. On each line, the release rate for each of the five iodine isotopes is listed in ascending order of the isotopes (ie I-131, I-132, etc.)

The format for the data is the same as that in the JCOSTACK DATA and JCOSIRW DATA files, which is exponential form with three decimal places. The first line is all O.OOOE + 00 since it represents the release rate from 0 to 19 minutes. On the following lines the values from Table 6 are specified.

The filename of this file is ALTSIRW DATA, and it is listed on the attached microfiche.

6.4 CONTROL ROOM DOSES The CONDOSE code, as described in Reference 2.33, was execµted once for each release path. The first execution was with the CRJCOl DATA and JCOSTACK DATA input decks, which resulted in the CRJCOl LISTING file with the control room doses from the containment atmosphere and ESF leakage.*

The second execution was with the CRJC02 DATA and JCOSIRW DATA input decks, which resulted in the CRJC02 LISTING file with the control room doses from SIR W Tank release. The third execution was with the CRJC03 DATA and ALTSIRW DATA input decks, which resulted in the CRJC03 LISTING file with the control room doses from the spent fuel pool tilt pit release for the

PALISADES NUCLEAR PLANT EA~A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 29 Rev # _...:;.o__

alternate recirculation path. All of the listing files are also listed on the attached microfiche. The output files for plotting the control room air concentration versus time were discarded since they were of no interest in this analysis.

As with the listing file from the MHACALC code, the doses* on these listing files must be taken from that listed from inhalation for the thyroid dose and from that listed for submersion for the whole body and skin doses. The resultant control room doses due to containment atmosphere and ESF leakage, as can be seen in CRJCOl LISTING, are 8.394 rem thyroid, 1.205 rem skin, and 0.051 rem whole body.

The shine dose from containment must also be added to this whole body dose. Most of the containment shine dose to the control room is from the purge lines. This dose was calculated to be 0.400 rem whole body in Reference 2.45. However, this value is conservative since it is based on the Regulatory Guide 1.4 source term, and is taken at the access doors to the control room viewing gallery which is not actually in the control room. Including the shine dose results in 8.394 rem thyroid, 1.205 rem skin, and 0.451 rem whole body from the contain..rnent and safeguards rooms release. These doses must be added to each of the two release paths for the leakage through CV-3027 & CV-3056 to obtain the total control room dose for each of the two cases.

The resultant control room doses due to releases from the SIRW Tank, as can be seen in CRJC02 LISTING, are low enough to approximate as 0.0 rem thyroid, skin, and whole body. These doses are very low since most of the iodine is retained in the large volume of water in the SIR W Tank. However, there would be a resultant shine dose in the control room from the activity in the SIRW Tank since the tank is on the roof above the control room. The SIRW Tank shine dose is calculated using the pc based MICROSHIELD code [Ref. 2.47] with the SIRW Tank activity values given in LOCAJCO LISTING from the MHACALC code execution. The major inputs for the MICROSHIELD code are the activities of the radionuclides present in the shine source, dimensions of the source, the materials and distances between the source and the receptor point, and the buildup factor method to be used.

To perform the shine dose calculations, the materials between the SIRW Tank and the control room were first determined. From Reference 2.17 it can be seen that there is a 1" 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 l' of concrete followed by a 4' air space, followed by another l' of concrete. As can be seen from Reference 2.24, the control room has a total height of 12', jgnoring the drop ceiling. Assuming the average operator has a height of 6' leaves 6' of air space in the control room. Since the MICROSHIELD code only accommodates up to five shields, all of the air space is considered a continuous space and all of the concrete is considered a continuous shield. It is therefore assumed that the shields are as follows: l" of sand, 34" of concrete, and 120" of air. The SIRW Tank is also considered as a self-shielding source since there is 24" of water left in the tank after RAS. Including the water height in the SIR W Tank, and conservatively ignoring the rising water level as water enters the tank, the total distance between the top of the source and the receptor point is 179". This is conservatively assuming all of the activity in the tank and piping is in the water volume of the tank.

To determine which method for buildup factors to use, the activity of each iodine isotope in the SIR W Tank was entered for the first time at which they were listed in LOCAJCO LISTING, and the code was executed using each buildup factor method, and choosing each shielding material as the basis for the buildup. The Taylor method and the water volume of the SIR W Tank resulted in the highest dose rate so they were chosen. For the number of angle and radial segments, various values were specified, but

@ consumn Power NWEIUllS PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-02

MKMJM#S NOfilUSI Sheet 30 Rev # 0 the default values of 5 and 11 respectively resulted in the highest dose rate so they were used. The MICROSHIELD code was executed for most of the times at which the activity in the SIRW Tank was listed in LOCAJCO LISTING, leaving all parameters except the activity the same for each execution.

The printouts for each of the executions are listed in Attachment 2, with the results of each execution occupying two pages. The dose rate in the control room can be seen to increase for approximately the first 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months due to the increasing iodine concentration, but 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. 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 SIRW Tank began, to 120 minutes, the dose rate at 60 minutes is used since it was not calculated at any earlier point in time. The calculation of the dose received, using the results of the MICROSHIELD executions, is shown below.

Pose (Ml'Sll'J)

I/me //ltera14/

19-+ 1219 tr11°A: 6'.'3S£-2 (i2s-11)4'8 = &.112 12m ~8~ A11it: /, 2tf9E-I (~Bl!J-/261)/~ISI = &. 7S8

'1861~ 72~A1l11: 2,, 'IJS"G-1 (720-1aexv~a = 6l.. '166 12~-+ l+'~llAUA: 2 .. 371£-1 (1'1'18-7261)/68 =2 .. B'IS l'i'ID_.28811 ,,,,;, : /.,4'.29£-1 (2888-/l;~t9)/~8 = 3.. 918 2886' + 'l.32 8111/n: 3. ~86£-2 ('1329-2881>)/~8 : Q,$3.,

~32~-., 724JSM;,,: /,. 329£-2 (12Q6' - ~311J)/tf~ ~., = 3'

'12618 -+/'1'118 "",.,,: S. 7SSE-3 (l'l'lc9s- '1289)/6~ = t9. 6'18

/'1~81/1 _. 21189 mi,,: 3. 'I S/F-3 (21116> - /~~4161')/~&..: 9. "~'

2119~ +288~~m/n: 3. 752£-3 (288ff-21.lt99)/lt9~~ .. 'ISIJ 288'9/g _.. 36&'Sil ,,,,;, : 3. 2~1JE*3 (3,elJIJ-288&'~)/d'.t?= ~" 3'/9 3'~~~ ~ '132'94' A1/...,: .l . . ~E-3(~328&'-3l4J19tJVtfa ~P.. 111 lot-al Pose = //,. 872 mf9em As can be seen, the total shine dose in the control room from the SIR W Tank over the 30 days is 11.872 mrem, or 0.012 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 0.012 + 0.000 + 0.451 = 0.463 rem. As mentioned earlier, the thyroid dose for the containment and safeguards room release is 8.394 rem and the skin dose is 1.205 rem, and the release from SIRW Tank is 0.000. These values are well within the 30 rem thyroid, 30 rem skin, and 5 rem whole body limits of SRP 6.4, and hence GDC 19.

For the alternate recirculation path, the release from the spent fuel pool tilt pit results in control room doses of 0.016 rem thyroid, 0.000 rem skin, and 0.000 rem whole body, as can be seen from the CRJC03

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 31 Rev # O LISTING file. Since the leakage is going into the spent fuel pool tilt pit, no shine dose to the control room would be expected. Therefore, adding this dose to that from the containment and safeguards room release and shine results in 0.016 + 8.394 = 8.410 rem thyroid, 1.205 rem skin, and 0.451 rem whole body. These doses are also well within the limits of SRP 6.4 and GDC 19.

7.0

SUMMARY

This analysis was performed to verify that the offsite and control room doses from a LOCA with the current plant configuration and procedures would not exceed the established limits of 10 CFR 100 or GDC 19. Sin~e this analysis is intended to justify continued operation until the end of cycle 11, the standard Regulatory Guide 1.4 source term was not used. Instead, the 10 CFR SO Appendix K analysis was used to justify that no fuel melting would occur from a large break LOCA, and the conservative assumption that 100 % of the core experiences DNB, and hence has cladding failure, was made.* The fission gas release fraction from the fuel to the pellet-clad gap was assumed to be 20 %, which is slightly higher than the actual peak predicted value. For noble gas, all of the fission gas in the pellet-clad gap was assumed to- be released to the containment atmosphere. For iodine, half of the fission gas in the pellet-clad gap was assumed to be released to the containment atmosphere and half to the containment sump solution. For the iodine in the containment atmosphere, 91 % was assumed to be in elemental form and 9 % in organic form. Since the Emergency Operating Procedures (EOPs) are being changed to start the addition of the required amount of NaOH at RAS, the sump pH was assumed to be controlled with no iodine re-evolution from the sump occurring. ,

The leakage of sump water through CV-3027 & CV-3056 was assumed to be 0.1 gpm, which is the current acceptance criteria of test procedure R0-119. The leakage through CV-3027 & CV-3056 was evaluated for two cases: for the alternate recirculation path redirecting the leakage to the spent fuel pool tilt pit as in the current EOPs, and for the leakage going directly into the SIR W Tank. For the iodine entering the SIRW Tank, 0.03 % was assumed to be of a volatile form, and a partition factor based on an assumed temperature of 100°F was used to account for the amount of the volatile iodine that would become airborne. Airborne iodine in the SIR W Tank was then assumed to exit the tank at twice the rate at which air is displaced from the tank. For the leakage through CV-3027 & CV-3056 going to the spent fuel pool tilt pit, the initial flashing fraction was used to account for the amount of iodine going airborne throughout the incident, and credit was taken for the Tech. Spec. filter efficiencies for the fuel pool area exhaust of 94 % for methyl iodide and 99 % for all other iodine.

The MHACALC code was executed using the data for the containment, ESF, and SIR W Tank releases.

Spreadsheets were used to calculate the release and offsite doses for the case of the leakage going to the spent fuel pool tilt pit. The offsite doses for the leakage going into the SIRW Tank were given in the output of the MHACALC code to be 7.641 rem thyroid and 0.057 rem whole body at the site boundary, and 4.499 rem thyroid and 0.011 rem whole body at the low population zone distance. The offsite doses for the alternate recirculation path directing leakage to the spent fuel pool tilt pit were calculated to be 7.645 rem thyroid and 0.057 rem whole body at the site boundary, and 4.504 rem thyroid and 0.011 rem whole body at the low population zone. All of these resultant doses are well within the 300 rem thyroid and 25 rem whole body limits of 10 CFR 100, for both cases of the leakage through CV-3027 & CV-3056.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 32 Rev # 0 The output from the MHACALC code also provided the radionuclide release rates for calculating control room dos*es for the containment plus safeguards room release and for the SIR W Tank release.

Two separate files containing release rates were provided: . one containing release rates from containment and ESF leakage in the safeguards rooms which are assumed to be released from the stack, and one containing the release rates from the SIRW Tank. The release rates from the spent fuel pool tilt pit for the case of the alternate recirculation path were calculated using spreadsheets and used to construct a release rate file. These three files of release rates were used in three executions of the CONDOSE code, using the appropriate atmospheric dispersion factors for the release location in each execution. The output of the CONDOSE code provided the doses from radionuclides that entered the control room. The containment atmosphere and ESF leakage resulted in a doses of 8.394 rem thyroid, 1.205 rem skin, and 0.051 rem whole body. To this was added a contribution to the whole body from

containment shine. The maximum containment shine dose calculated during the NUREG-0578 shielding study was actually determined to be at the entrance doors to the viewing gallery. The resulting shine dose at the doors over 30 days is 0.400 rem. Although this value is based on Regulatory Guide 1.4 source terms and is not actually in the control room, it was conservatively used as a bounding value.

This brought the control room doses to 8.394 rem thyroid, 1.205 rem skin, and 0.451 rem whole body, not including any contribution from leakage through CV-3027 & CV-3056.

For the case of the leakage through CV-3027 & CV-3056 going directly to the SIRW Tank, the small release from the tank resulted in negligible doses. However, since the tank is located on the roof above the control room, SIRW Tank shfoe had to be added. For the shine dose from the SIRW Tank, the I

MICROSHIELD code was executed using the iodine activity in the SIR W Tank at the time points listed in the output from the MHACALC code. The resulting dose rates were then conservatively assumed to be constant over the time intervals, and used to calculate the total resultant shine dose of 0.012 rem.

Therefore, the resultant total control room doses from the incident with leakage going into the SIR W Tank were calculated to be 8.394 rem thyroid, 1.205 rem skin, and 0.463 rem whole body.

For the case of the leakage through CV-3027 & CV-3056 being rerouted to the spent fuel pool tilt pit, the CONDOSE code calculated the control room dose for the release path to be 0.016 rem thyroid, 0.000 rem skin, and 0.000 rem whole body. With the leakage going into the spent fuel pool tilt pit, no shine dose to the control room from the leakage would result. Therefore, the total control room doses for the case of the alternate recirculation path were calculated to be 8.410 rem thyroid, 1.205 rem skin, and 0.451 rem whole body.

For both cases ofthe leakage through CV-3027 & CV-3056, the resultant control room doses are well within the 30 rem thyroid, 30 rem skin, and 5 rem whole body limits of 10 CFR 50 Appendix A, GDC 19.

A summary of the resultant total offsite and control room doses for both cases of the leakage through CV-3027 or CV-3056 are shown in Table 7.

/F':..\=mn

~ ~Elllllll PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 MEmA#'S NOP' ANALYSIS CONTINUATION SHEET Sheet 33 Rev # 1

~~~

TABLE 7

SUMMARY

OF DOSES FOR THE 'IWO PATHS OF LEAKAGE THROUGH CV-3027 OR CV-3056 LEAKAGE GOING IN SIRW TANK

_____________ "T" __________ "T" ________________ LEAKAGE GOING IN TILT PIT

_____________ "T" __________ "T" _______________

I I I I Thyroid I I Skin I I Whole Body Thyroid I I Skin I I Whole Body I I I I (Rem) I (Rem) I (Rem) (Rem) I (Rem) I (Rem)

SB (0-2 Hr) 7.641 N/A 0.057 7.645 N/A 0.057 LPZ 4.499 N/A 0.011 4.504 N/A 0.011 Control Room 8.394 1.205 0.463 8.410 1.205 0.451 I

NOTE: A case including sump water leakage into the SIRW Tank through the main discharge lines due to an idle containment spray pump has been evaluated and is included as Attachment

3. The results of that case are shown in Table 8 on page 67 of this analysis.

8.0 CONCLUSION

The radiological consequences of a large break LOCA were evaluated using a source term based on the 10 CFR 50 Appendix K analysis. The source term is therefore less conservative than the Regulatory Guide 1.4 source term, but is considered a more realistic estimate for which no fuel melting occurs.

Leakage of sump water through CV-3027 and CV-3056 was accounted for in two separate cases: with the alternate recirculation path in the EOPs rerouting the leakage to the spent fuel pool tilt pit, and for the leakage going directly to the ~IR W Tank, with both cases using the current established acceptance criteria for the leak rate. Also, neutral sump pH was credited since the EOPs are being changed to add the required amount of NaOH at RAS. The MHACALC code was used to calculate the containment atmosphere and SIRW Tank activities, the radionuclide release rates, and the offsite doses from containment, ESF leakage, and SIRW Tank releases. The doses from releases from the spent fuel pool tilt pit were calculated using spreadsheets. The CONDOSE code was then used to calculated the control room doses from radionuclides entering the control room. The MICROSHIELD code was also used, for calculating shine dose in the control room from the SIRW Tank.

The resultant doses from the incident for both cases of the leakage through CV-3027 & CV-3056 were found to be well within the established limits of 10 CFR 50 and 10 CFR 100 for the control room and offsite doses, respectively. Therefore, operation until the end of Cycle 11 with the current plant configuration is justifiable.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 34 Rev # --=-I__

9.0 LIST OF ATfACHMENTS

1. Form 40 from LTPhillips to PAHarden dated February 6, 1991, 1 page.

2. Printouts of results from execution of the MICROSHIELD code, 26 pages.

3. Determination of the magnitude of the allowable leak rate through SIRW Tank main discharge lines using the LOCA source term, 34 pages.

4. Design Review Forms Including:

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

Proc. No. 9.11 Attachment 5, Technical Review Checklist for Rev 0, 1 page.

Form 3110 1-82, NOD Document Review Sheet for Rev 0, ___]_ page(s).

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

Form 3110 1-82, NOD Document Review Sheet for Rev 1, _1 page(s).

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

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{Consumers Power Company - #037)

_Page 1 File Ref:

File SIRW60.MSH Date: _/_/_

Run date: February 22; 1992 By:

Run time: 12:40 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 60 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ****.**....*.*........*.* X 454.660 cm.

Source cylinder radius .....*...*............. R 701. 040 II Source cylinder length ..*****.*.....*........ Tl 60.960 II Thickness of second shield **.........**.. ~ ... T2 2.540 II Thickness of third shield .*.*................. T3 86.360 II Thickness of fourth shield .*....*...*........ T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

Page 2 File: SIRW60.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 60 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) . . . . . . . . . . . . . . . 5 Number of radial segments (Nradius) . . . . . . . . . . . 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131 1.0270e+02 I-132 1.0820e+02 I-133 l.6320e+02 I-134 9.9560e+Ol I-135 1.5560e+02 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1312 3.120e+ll 9.983e+OO l.588e-02 2 1.6224 2.979e+12 2.114e+Ol 3.652e-02 3 1.1702 5.855e+12 5.148e+OO 9.585e-03 4 .8487 1.2ose+13 1.897e+OO 3.777e-03 5 .6587 7.311e+12 2.285e-Ol 4.743e-04 6 .5204 7.552e+12 5.725e-02 1.174e-04 7 .3724 3.755e+12 1. 743e-03 3.585e-06 8 .2786 6.364e+ll 1. 876e-05 3.731e-08 9 .2164 1.642e+ll 3.903e-07 7.344e-10 10 .1355 1.867e+ll 1.02oe-09 1.695e-12 11 12 13 14 15

16 17 18 19 20 TOTALS: 4.083e+13 3.846e+Ol 6.635e-02

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Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File SIRW120.MSH Date: ~-/~_/~

Run date: February 22, 1992 By:

Run time: 12:55 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 120 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . . . X 454.660 cm.

Source cylinder radius . . . . . . . . . . . . . . . . . . . . . . . R 701.040 II Source cylinder length . . . . . . . . . . . . . . . . . . . . . . . Tl 60.960 II Thickness of second shield ................... T2 2.540 II Thickness of third shield *......*............ T3 86.360 II Thickness of fourth shield ...*.....*......... T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

r.;.r -.*t /.., .. '"" -v/""' -(..- ...

Page 2 File: SIRW120.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 120 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) ............... 5 Number of radial segments (Nradius) ..........* 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide curies I-131 2.5200e+02 I-132 l.9720e+02 I-133 3.8880e+02 I-134 1.1120e+02 I-135 3.4510e+02 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1371 6.060e+ll 1. 960e+Ol 3.114e-02 2 1.6209 5.685e+12 4.012e+Ol 6.931e-02 3 1.1766 1.159e+l3 1.048e+Ol l.949e-02 4 .8444 1.772e+l3 2.709e+OO 5.397e-03 5 .6603 1.272e+l3 4.037e-Ol 8.377e-04 6 .5214 1. 649e+l3 1. 266e-Ol 2.596e-04 7 .3711 8.678e+12 3.893e-03 8.006e-06 8 .2815 1. 348e+12 4.455e-OS 8.872e-08 9 .2183 3.26le+ll 8.400e-07 l.585e-09 10 .1359 2.152e+ll 1. 243e-09 2.067e-12 11 12 13 14 15 16 17 18 19.

20 TOTALS: 7.53Be+13 7.344e+Ol 1. 264e-Ol

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : SIRW120.MSH Date: ~-/~_/~

Run date: February 22, 1992 By:

Run time: 12:59 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 480 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector *****.......*.*.......... X 454.660 cm.

II Source cylinder radius **...*......*.........* R 701. 040 Source cylinder length *****...*.****...*..... Tl 60.960 II II Thickness of second shield *...*.*.***.....**. T2 2.540 Thickness of third shield **........*......**. T3 86.360 II II Thickness of fourth shield *.*...****.....**.* T4 304.8 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

tif"i-d-lfl.:. - '/ 2 - O/;Z - O:Z

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Page 2 File: SIRW120.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 480 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) ............... 5 Number of radial segments (Nradius) ........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.1260e+03 I-132 1.4760e+02 I-133 . 1.4530e+03 I-134 4.4150e+OO I-135 8.3960e+02 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1766 1.013e+12 3.523e+Ol 5.556e-02 2 1. 6327 1.084e+13 7.985e+Ol 1.377e-Ol 3 1.1845 2.446e+13 2.292e+Ol 4.255e-02 4 .8447 1.279e+13 1.959e+OO 3.903e-03 5 .6626 1. 310e+13 4.245e-Ol 8.806e-04 6 .5228 5.203e+13 4.059e-Ol 8.325e-04 7 .3687 3.562e+13 1.503e-02 3.092e-05 8 .2862 4.019e+12 1.602e-04 3.198e-07 9 .2190 7.336e+ll 1. 940e-06 3.663e-09 10 .1433 2.079e+10 3.330e-10 5.652e-13 11 12 13 14 15 16 17 18 19 20 TOTALS: 1 ..546e+14 1.408e+02 2.415e-Ol

l.i:,-/-/;-;V'- - I ,i- ~/.2-C'.2 5/, e e r tf,J. -fev 61 Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File SIRW480.MSH

Date: __ / __ / _

Run date: February 22, 1992 By:

Run time: 1:01 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 720 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ..*...................... X 454.660 cm.

II Source cylinder radius. . . . . . . . . . . . . . . . . . . * . . . R 701.040 II Source cylinder length . . . . . . . . . . . . . . . . . . . . . . . Tl 60.960 II Thickness of second shield ..*..............*. T2 2.540 II Thickness of third shield ..*........*.....*.. T3 86.360 II Thickness of fourth shield ...............*... T4 304.8 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

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5A~er '13 R~O?

Page 2 File: SIRW480.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 720 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.6880e+03 I-132 6.7220e+Ol I-133 1.9340e+03 I-134 2.8390e-Ol I-135 8.3920e+02 RESULTS:

Group Energy Activity Dose point flux Dose rate

{MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1919 9.220e+ll 3.299e+Ol 5.187e-02 2 1.6379 1.054e+13 7.916e+Ol 1. 364e-Ol 3 1.1860 2.485e+13 2.345e+Ol 4.352e-02 4 .8556 1. 037e+13 1. 703e+OO 3.384e-03 5 .6615 1.145e+13 3.674e-Ol 7.623e-04 6 .5229 6.718e+13 5.246e-Ol 1. 076e-03 7 .3681 5.261e+13 2.187e-02 4.SOOe-05 8 .2867 5.288e+12 2.157e-04 4.307e-07 9 .2164 7.828e+ll 1.867e-06 3.515e-09 10 .1474 6.397e+09 1.796e-10 *3. 082e-13 11 12 13 14 15 16 17 18 19 20 TOTALS: 1.840e+14 1. 382e+02 2.371e-Ol

/f-A-AI~ * 'l.2-~/.z-02 sJ eeC l/'I -fev 6(

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : SIRW720.MSH Date: ___ / ___ / ___

Run date: February 22, 1992 By:

Run time: 1:03 p.m. Checked:

CASE*: JCO LOCA SIRW TANK SHINE DOSE, AT 1440 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector .*...*.........*......... X 454.660 cm.

Source cylinder radius ...*.*........*........ R 701.040 II Source cylinder length ..**.**.*....*......... Tl 60.960 II Thickness of second shield ***.*....**........ T2 2.540 II Thickness of third shield *...*.*............* T3 86.360 II Thickness of fourth shield .**.......*..*...*. T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND . 2. 0

Ch"-/.1-A/L - 7.C-0/2 -c; ;2.

SA '6!et""° 15 £ev 62 Page 2 File: SIRW720.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 1440 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of* the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) ............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 3.2770e+03 I-132 3.6620e+OO I-133 2.6280e+03 I-134 4.3530e-05 I-135 4.8320e+02 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2057 4.920e+ll 1.805e+Ol 2.831e-02 2 1.6422 5.949e+12 4.540e+Ol 7.822e-02 3 1.1954 1. 675e+13 1.649e+Ol 3.053e-02 4 .8682 8.175e+12 1.456e+OO 2.886e-03 5 .6581 1. 426e+13 4.433e-Ol 9.204e-04 6 .5229 8.856e+13 6.914e-Ol 1. 418e-03 7 .3673 1.001e+14 4.076e-02 8.386e-05 8 .2874 8.441e+12 3.542e-04 7.076e-07 9 .2042 6.747e+ll 9.751e-07 1.soae-09 10 .1484 3.210e+08 1.045e-ll 1.798e-14 11 12 13 14 15 16 17 18 19 20 TOTALS: 2.434e+14 8.257e+Ol 1.424e-Ol

£A-4-Alt.. - ?.2- Cl/.:l- ~.z 5)ee 'l 7'6 _.fev CJ Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File SIRW1440.MSH Date: ___ / ___ / ___

Run date: February 22,

1992 By:

Run time: 1:05 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 2880 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . . . X 454.660 cm.

Source cylinder radius .............*...*..... R 701.040 "

Source cylinder length. . . . . . . . . . * . . . . . . . . . . . . Tl 60.960 II Thickness of second shield .*........*.....*.* T2 2.540 II Thickness of third shield *.*................* T3 86.360 "

Thickness of fourth shield .*..............**. T4 304.8 "

Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

el?-'4-Alt-12-012 -02

.5iee't '17 Rev!;

Page 2 File: SIRW1440.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 2880 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide curies I-131 6.0520e+03 I-132 5.3250e-03 I-133 2.3780e+03 I-134 5.0130e-13 I-135 7.8510e+Ol RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 7.929e+lO 2.918e+OO 4.574e-03 2 1. 6427 9.647e+ll 7.374e+OO l.270e-02 3 1.2252 6.083e+12 6.836e+OO 1. 258e.;..02 4 .8689 6.017e+12 1. 076e+OO 2.133e-03 5 .6548 2.318e+13 6.982e-Ol 1.451e-03 6 .5228 7.940e+13 6.191e-Ol 1. 270e-03 7 .3670 1. 833e+14 7.398e-02 1.522e-04 8 , . 2882 1.407e+13 6.098e-04 1. 218e-06 9 .1838 6.505e+ll 2.645e-07 4.779e-10 10 .1484 4.667e+05 1. 520e-14 2.616e-17 11 12 13 14 15 16 17 18 19 20 TOTALS:

.. .,.t;----* --

3.137e+14 1.960e+Ol 3.486e-02

cA-A~111£ -"1;i.-c;12-r;,i

54eer 'I B * ..eev. 5J Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File SIRW2880.MSH Date: ~-/_-~/~

Run date: February 22, 1992 By:

Run time: 1: 07 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 4320 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . . . X 454.660 cm.

Source cylinder radius ****................**. R 701. 040 II Source cylinder length *..***....*.....*....** Tl 60.960 II Thickness of second shield .................*. T2 2.540 II Thickness of third shield *....*....**..*..... T3 86.360 II Thickness of fourth shield *.........*.*...**. T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water i. o Zirconium SAND 2.0

£"9-4-11/t.. -1.2-()/_2 -02 sA<!'d t/'I .(~v 6>

Page 2 File: SIRW2880.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 4320 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius) **********~ 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide Curies I~131 8.3470e+03 I-132 5.7820e-06 I-133 1.6060e+03 I-134 o.ooooe+oo I-135 9.5230e+OO RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1423 1.165e+lO 3.806e-Ol 6.042e-04 2 1. 6393 1.150e+ll 8.679e-Ol 1. 496e-03 3 1.2452 2.982e+12 3.708e+OO 6. 799e-03

4 .8772 3.668e+12 6.917e-Ol 1.369e-03 5 .6542 3.054e+13 9.146e-Ol 1.901e-03 6 .5229 5.387e+13 4.208e-Ol 8.631e-04 7 .3673 2.511e+14 1. 020e-Ol 2.098e-04 8 .2919 2. 026e+13 1.014e-03 2.030e-06 9 .2568 2.195e+ll 2.696e-06 5.302e-09 10 .1797 8.179e+ll 2.512e-07 4.514e-10 11 12 13 14 15 16 17 18 19 20 TO~~S: 3.636e+14 7.087e+OO 1. 324e-02

£/1-,4-.Nt,-<1,,t.-t:J/Z-0,2 5JeeZ- 50 ~ ~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File SIRW4320.MSH Date:~-/~_/~-.

Run date: February 22, 1992 By:

Run time: 1:10 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 7200 MINUTES

.GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . . . x 454.660 cm.

Source cylinder radius .....*...............*. R 701.040 "

Source cylinder length *....*................. Tl 60.960 "

Thickness of second shield ...*..*.....*...*** T2 2. 540 ' "

Thickness of third shield .........*.......*.* T3 86.360 "

Thickness of fourth shield .*.**..**...*....** T4 304.8 "

Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium water 1.0 Zirconium SAND 2.0

f::."f-/.~-;11" -r,,z-0;.;L-~2 sAeer 51 Rey ~

Page 2 File: SIRW4320.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 7200 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius) .....**...* 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 l.1730e+04 I-132 5.0360e-12 I-133 5.4170e+02 I-134 o.ooooe+oo I-135 1. 0350e-Ol RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr}

1 2.1423

1. 267e+08 4.137e-03 6.566e-06 2 1. 6393 1. 250e+09 9.433e-03 1.626e-05 3 1.2491 9.365e+ll 1.187e+OO 2.175e-03 4 .8760 1. 218e+l2 2.279e-Ol 4.511e-04 5 .6521 4.092e+13 1. 20le+OO 2.497e-03 6 .5220 1. 935e+l3 1.495e-Ol 3.066e-04 7 .3672 3.524e+14 1. 430e-Ol 2.941e-04 8 .2919 2.838e+13 1. 423e-03 2.849e-06 9 .2578 7.154e+l0 9.134e-07 1.797e-09 10 .1797 1.149e+12 3.530e-07 6.344e-10 11 12 13 14 15 16 17 18 19 20 TOT~S:

4.444e+14 2.924e+OO 5.750e-03

E/1-1'1-lfll- 92-012-02 5.f~er 5:2 -f'ev &

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : SIRW7200.MSH Date: ___ / ___ / ___

Run date: February 22, 1992 By:

Run time: 1:12 p.m. Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 14400 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ............*............ X 454.660 cm.

II Source cylinder radius .........*.. ~ .......... R 701. 040 II Source cylinder length .**..*.....**.......... Tl 60.960 II

.Thickness of second shield ..*.*..**......**.* T2 2.540 II Thickness of third shield ..*....**........*** T3 86.360 II Thickness of fourth shield ...*.**.**..*..*.*. T4 304.8 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

£4-/1-tl/i- '12 - 0/2-1,).2 5Aet::?r 53 Rey 6)

Page 2 File: SIRW7200.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 14400 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)................ 5 Number of radial segments (Nradius) . . . . . . . . . . . 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide Curies I-131 l.5260e+04 I-132 o.ooooe+oo I-133 l.9890e+Ol I-134 o.ooooe+oo I-135 7.0930e-07 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 7.163e+02 2.636e-08 4.132e-11 2 1. 6427 8.715e+03 6.662e-08 l.148e-10 3 1.2829 2.923e+10 4.394e-02 8.003e-05 4 .9061 3.954e+10 8.947e-03 1.760e-05 5 .7267 1. 021e+13 6.016e-Ol 1. 230e-03 6 .6316 4.288e+13 1. 065e+OO 2.210e-03 7 .5077 2.052e+12 1. 340e-02 2.743e-05 8 .3670 4.610e+14 1.858e-Ol 3.823e-04 9 .2891 3.417e+13 1.530e-03 3.059e-06 10 .1797 1.495e+12 4.593e-07 8.253e-10 11 12 13 14 15 16 17 18 19 20 TOW~S: 5.519e+14 1. 920e+OO 3.951e-03

EA--4-Ht.-'I~ -012-0.z S~e<!Z- .51/ Rev 6)

Microshield 3.13

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I Page File Run date:

Run time:

1 (Consumers Power Company - #037)

SRW14400.MSH February 22, 1992 1:13 p.m.

File Ref:

Date: ____ / ____ / ____

By:

Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 21600 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ..*...................**. X 454.660 cm.

Source cylinder radius *..........**.......*.. R 701.040 II Source cylinder length ............*...*...... Tl 60.960 II Thickness of second shield ........*......**.. T2 2.540 II Thickness of third shield .................**. T3 86.360 II Thickness of fourth shield .*..**...*......**. T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen I Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0 I

A-4-t11t - '/.l-~.t',2.-0.2 S.i~~r SS ~Y &

I Page 2 File: SRW14400.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 21600 MINUTES BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) ............... 5 Number of radial segments (Nradius) ........... 11 SOURCE NUCLIDES:

Nuclide curies Nuclide Curies Nuclide curies I-131 1.4880e+04 I-132 o.ooooe+oo I-133 5.4740e-Ol I-134 o.ooooe+oo I-135 3.6410e-12 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 3.677e-03 1.353e-13 2.121e-l6 2 1.6427 4.474e-02 3.420e-13 5.89le-16 I 3 4

5 6

7 1.2829

.9061

.7266

.6332

.5078 8.046e+08 1.088e+09 9.925e+12 4.120e+13 1.985e+l2

1. 209e-03 2.462e-04 5.845e-Ol

1. 036e+OO

1. 298e-02 2.203e-06 4.845e-07 1.195e-03 2.152e-03 2.658e-05 8 .3670 4.495e+14 1. 812e-Ol 3.727e-04 9 .2891 3.332e+13 1.492e-03 2.983e-06 10 .1797 1.458e+12 4.478e-07 8.047e-10 11 12 13 14 15 16 17 18 19 20 TO~ALS: 5.374e+14 1.818e+OO 3.752e-03 I

EA-#-AIL -f.,2-~/.2-02 5)ee.t' S 6 .Rev 19 Microshield 3.13

=====

I Page File Run date:

Run time:

1 (Consumers Power Company - #037)

SRW21600.MSH February 22, 1992 1:21. p.m.

File Ref:

Date: ~-/~_/~

By:

Checked:

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 28800 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ................*........ X 454.660 cm.

Source cylinder radius ..............*........ R 701.040 II Source cylinder length ..........****......... Tl 60.960 II Thickness of second shield ...**..*.*.......*. T2 2~540 II Thickness of third shield *.**....*..*......*. T3 86.360 II Thickness of fourth shield ..*..*....*.......* T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum carbon Concrete 2.350 Hydrogen I Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0 I

ii.19-/1-#l -'/.2 -012 -02

~ie.et"S7 ~ev. lQ I Page 2 File: SRW21600.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 28800 MINUTES BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.2900e+04 I-132 o.ooooe+oo I-133 1~3390e-02 I~134 o.ooooe+oo I-135 1.6610e-17 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 1.677e-08 6.172e-19 9.677e-22 1.6427 2.041e-07 1.560e-18 2.688e-21 I

2 3 1.2829 1.968e+07 2.958e-05 5.388e-08 4 .9061 2.662e+07 6.023e-06 1.185e-08 5 .7266 8.603e+12 5.066e-Ol 1. 036e-03 6 .6333 3.570e+13 8.984e-Ol 1. 865e-03 7 .5078 1. 721e+12 1.125e-02 2.304e-05 8 .3670 3.897e+14 1. 571e-Ol 3.231e-04 9 .2891 2.889e+13 1.294e-03 2.586e-06 10 .1797 1.264e+12 3.883e-07 6.976e-10 11 12 13 14 15 16 17 18 19 20 TOTaLS: 4.659e+14 1. 575e+OO 3.250e-03 I

£A-4-.tllt:. - '1.l-Gll.Z -e:J.2

.5)eet" 58 .Rev~

Microshield 3.13 I Page File Run date:

Run time:

1 SRW28800.MSH (Consumers Power Company - #037)

February 22, 1992 1:23' p.m

File Ref:

Date: ~-/~_/~

By:

Checked:

. CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 36000 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . . . x 454.660 cm.

Source cylinder radius ....*....*....**....**. R 701. 040 II Source cylinder length **.**...*..*........**. Tl 60.960 II Thickness of second shield ..*.....***......*. T2 2.540 II Thickness of third shield *........*......***. T3 86.360 II Thickness of fou~th shield .......*..*.......* T4 304.8 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum carbon Concrete 2.350 Hydrogen I Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

t:A-.;/-.YL. ~?.L-C/.l-~,,z Skec S'f Rev&

I Page 2 File: SRW28800.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 36000 MINUTES BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.0480e+04 I-132 o.ooooe+oo I-133 3.0680e-04 I-134 o.ooooe+oo I-135 o.ooooe+oo RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 1. 2829 4.509e+05 6.778e-07 1. 235e-09 2 1. 0216 1. 023e+05 4.545e-08 8.733e-11 I 3 4

5 6

7

.8673

.6514

.5078

.4245

.3671 7.170e+05 3.599e+13

1. 398e+12 5.242e+04 3.156e+14

1. 269e-07 1.0SOe+OO 9.141e-03 8.016e-11 1.277e-Ol 2.517e-10 2.183e-03 l.872e-05 1.642e-13 2.627e-04 8 .2903 2.444e+13 1.150e-03 2.301e-06 9 .2578 4.047e+04 5.174e-13 1. 018e-15 10 .1797 1. 027e+12 3.154e-07 5.668e-10 11 12 13 14 15 16 17 18 19 20 TO'J?.~S: 3.785e+14 1.188e+OO 2.466e-03 I

£ ,4-,,,P-11/i. - f..2-t.:J/.Z-0.2

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5ieer 60 ~&

Microshield 3.13

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I Page File Run date:

Run time:

1 (Consumers Power Company - #037)

.SRW36000 .MSH February 22, 1992 1:24 p.m.

File Ref:

Date:

By:

Checked:

_!_/_

CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 43200 MINUTES GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ..*....*...*...**........ X 454.660 cm.

II source cylinder radius *........*............. R 701.040 II Source cylinder length . . . . . . . . . . . . . . . . . . . . . . . Tl 60.960 II Thickness of second shield ....*...**......... T2 2.540 II Thickness of third shield .....**..........**. T3 86.360 II Thickness of fourth shield .**..*......*...... T4 304.8 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Air .001220 Aluminum Carbon Concrete 2.350 Hydrogen I Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium water 1.0 Zirconium SAND 2.0 I

6;/1-r,-/tlt- -r,;Z-e;,;,,c-~;L

.5AeeT b/ .<ev &

Page 2 File: SRW36000.MSH CASE: JCO LOCA SIRW TANK SHINE DOSE, AT 43200 MINUTES

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments {Npsl)............... 5 Number of radial segments {Nradius)........... 11 SOURCE NUCLIDES:

Nuclide curies Nuclide curies Nuciide curies I-131 8.1720e+03 I-132 o.ooooe+oo I-133 6.7520e-06 I-134 o.ooooe+oo I-135 o.ooooe+oo RESULTS:

Group Energy Activity Dose point flux Dose rate

{MeV) {photons/sec) MeV/{sq cm)/sec {mr/hr) 1 1.2829 9.924e+03 1.492e-08 2.717e-11 2 1. 0216 2.251e+03 1.000e-09 l.922e-12 3 .8673 1. 578e+04 2.793e-09 5.539e-12 4 .6514 2.807e+13 8.185e-Ol 1.702e-03 5 .5078 1. 090e+12 7.128e-03 1.459e-05 6 .4245 1.154e+03 1. 764e-12 3.613e-15 7 .3671 2.461e+14 9.957e-02 2.048e-04 8 .2903 1. 906e+13 8.968e-04 1.794e-06 9 .2578 8.906e+02 1.139e-14 2.241e-17 10 .1797 a.ooae+ll 2.460e-07 4.420e-10 11 12 13 14 15 16 17 18 19 20 TOTALS:

2.951e+14 9.261e-Ol 1.923e-03

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 62 Rev # -- I -

ATTACHMENT 3 DETERMINATION OF THE MAGNITUDE OF THE ALLOWABLE LEAK RATE THROUGH SIRW TANK MAIN DISCHARGE LINES USING THE LOCA INTERIM SOURCE TERM Scenario:

To set up the scenario, there must be failure of one train of containment spray pumps either by loss of offsite power accompanied by a diesel generator failure or by .failure of one of the containment sump isolation valves to open at RAS causing one train to lose suction. Since all of the spray pumps *are tied into a common header, an operating spray pump could force sump water through the discharge check valve of an idle pump, through the idle pump, through the SIRW Tank discharge isolation valve, and through the SIRW Tank discharge check valve into the SIRW Tank. The leak path is a torturous leak path and it requires that all of the above mentioned valves leak and that the containment sump outlet check valve and isolation valve do not leak or leak less than the SIRW Tank discharge check valve and isolation valve. It would also take a considerable amount of time for the operating spray pump to pressurize the path for small leakage, and the containment spray pumps could be expected to be shut off within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when containment pressure is low enough.

To evaluate the magnitude of the allowable leakage through this leak path, the same assumptions, methods, *and computer codes as were used previously in this analysis for calculating resultant doses from a LOCA are used. Assumptions specific to evaluating this scenario are listed below.

Assumptions:

1. It is assumed that all of the necessary failures to set up the scenario occur, and all of related valves !~~ or do not leak as necessary to cause the leak path to exist.

2. To demonstrate the magnitude of the allowable leak rate for this scenario using the LOCA interim source term, the leakage is assumed to be 20 gpm in addition to the 0.1 gpm leakage through CV-3027 & CV-3056. .

3. The leakage through the main discharge lines is conservatively assumed to reach the SIRW Tank at RAS. This is conservative since it would take some time to pressurize the line from the spray pumps back to the SIRW Tank, and there is a good distance of piping filled with water through which the iodine must travel.

4. It is assumed that containment sprays operate for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from initiation of the LOCA, after which no leakage through this path occurs. Sprays would probably be terminated in less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , when containment pressure drops below 3.0 psig.

)

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 63 Rei # -~-

I ATIACHMENT 3

5. Since the containment sump temperature would be below 212°F in about 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months post-LOCA [Ref. 2.14], the SIRW Tank water volume does not heat up enough to permit flashing of leakage entering the tank. This assumption is made taking into consideration that in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at 20 gpm, roughly 642 ft 3 would enter the SIRW Tank which has a RAS volume of 3347.5 ft 3 (calculated previously) at ambient temperature.

6. For calculation of the partition coefficient for the volatile specie of iodine in the SIRW Tank, the air-water interface temperature in the tank is assumed to reach 200°F. This is conservative since the partition coefficient of the volatile specie decreases with increasLng temperature, and the sµmp water temperature at 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months would be approximately 160°F from Reference 2.14.

7. 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. Previously in this analysis, a multiplication factor of 2 was used. Since the SIRW Tank is likely to heat up with the large amount of sump water assumed to be entering for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the multiplication factor is assumed to be 3 to account for an increased air density change which could force more iodine out of the SIRW Tank vent.

8. The water initially in the SIRW_ Tank main discharge lines at RAS is not accounted for in the SIRW Tank volume for any dilution effect.

9. For calculation of the shine dose in the control room from the SIRW Tank, control room occupancy factors are conservatively ignored as was done previously.

10. For calculation of the shine dose in the control room from the SIRW Tank, the rising water

level accompanying the. increasing iodine activity in the SIRW Tank due to the large amount of sump,~ater entering is not accounted for. This is conservative since as the water volume in the SfR.W Tank rises above the RAS volume, the self-shielding effect of the water would increase thereby decreasing the dose rate in the control room.

11. All activity in the SIRW Tank main discharge lines and the recirculation line between CV-3056 and the SIRW Tank is conservatively assumed to be in the SIRW Tank for shine dose calculations in the control room.

12. As was done previously, the shine dose from the SIRW Tank is conservatively assumed to remain constant over the time interval for each dose calculation.

~~

~

~ws NMND

,............. PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-02 Sheet 64 Rev # I

~--~

ATIACHMENT 3 Method:

The method for calculating the doses for this scenario is the same as that used previously in this analysis for leakage through CV-3027 & CV-3056 going to the SIRW Tanlc, with the additional leakage taken into account. The MHACALC code is used to calculate the offsite doses and the radionuclide release rates for the event. The CONDOSE code is then used to calculate the doses to the operators from radionuclides entering the control room. The MICROSHIELD code is then used to calculate the shine dose in the control room from the activity in the SIRW Tank.

The MHACALC input deck for this case is basically the same as the LOCAJCO DATA deck used previously, with a few changes relative to the SIRW Tank inleakage. The only lines of the input deck for this case that are different from LOCAJCO DATA are lines 1, 7, 17, 18, 19, 41, and 43. The changes made to these lines of the input deck are described below.

The title on line 1 was just changed to reflect the case being executed. On line 7, the partition factor for the volatile iodine in the SIRW Tank and multiplication factor for the rate at which iodine in the air volume of the SIRW Tank exits through the vent are changed. These are the third and fourth values, respectively, on line 7. For calculation of the partition coefficient for the volatile spede of iodine in the SIRW Tank, the air-water interface temperature is conservatively assumed to be 200°F, or 366.33°K, as mentioned in the assumptions above. This is conservative considering that the sump temperature is down to approximately 160°F after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months [Ref. 2.14]. Using Equation (1) w:lth this temperature to solve for the partition coefficient as was done previously results in a partition coefficient of 6.8 for the volatile specie of iodine in the SIRW Tank for this case. The multiplication factor for the rate at which iodine exits the SIRW Tank is taken as 3 for this case. This is to account for the possibility of increased iodine rele~e due to the density change of the SIRW Tank air volume as the large amount of high temperature sump water leaks in for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The time at which containment sprays are assumed to be terminated, the second value on line 17, is changed to 2880 minutes (48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ) for this case since it increases the leakage into the SIRW Tank through the main diScharge lines. On line 18, the fourth value is changed to 20.1 gpm to reflect the 20 gpm leakage of sump water into the SIRW Tank through the main discharge lines assumed for this case in addition to the 0.1 gpm leakage through CV-3027 & CV-3056. On line 19, the third value is changed to 2880 minutes to reflect the time that the leakage through the main discharge lines is assumed to stop, and the fourth value is changed to 0.1 gpm to reflect the continued leakage through CV-3027 & CV-3056 to the SIRW Tank.

Line 41 was changed to reflect the addition of a time interval so that radionuclide release rates are calculated for 12 time intervals. This is necessary since the MHACALC code calculates an averaged release rate over each of the time intervals specified. For this case, containment sprays and the 20 gpm leakage stop at 2880 minutes. The radionuclide release rates could therefore change by a significant amount at that time, and a time interval should end there. Line 41 was changed to add 2880 minutes as a time at which one of the time intervals for radionuclide release rate calculations should change.

@:: PALISADES NUCLEAR PLANT EA-A-NL-92-012-02

M*rm*~N*vrBI ANALYSIS CONTINUATION SHEET ATIACHMENT 3 Sheet 65 Rev # -~-

1 The input deck for this case with the above listed changes is listed on the attached microfiche under the filename JCOCASE2 DATA. The MHACALC code was executed with this input deck. The output files are JCOCASE2 LISTING, C2STACK DATA, and C2SIRW DATA, and are also listed on the attached microfiche. A fourth output file created by the code for plotting the containment atmosphere activity versus time was discarded since it was of no interest for this analysis. The JCOCASE2 LISTING file contains the offsite doses from the event and the SIRW Tank and containment activities at the input specified time points. The C2STACK DATA file contains the radionuclide release rates. from the containment atmosphere and ESF leakage for input to the CONDOSE code. The C2SIRW DATA file contains the radionuclide release rates from the SIRW Ta!l_k for input to the CONDOSE code. From JCOCASE2 LISTING, the offsite doses for this case are. 7.648 rem thyroid and 0.057 rem whole body at the site boundary, and 4.550 rem thyroid and 0.011 rem whole body at the low population zone. As can be seen, there is still a significant amount of margin between the calculated offsite doses and the .

10 CFR 100 limits (300 rem thyroid and 25 rem whole body) assuming 20 gpm leakage into the SIRW Tank through the main discharge lines using the plant specific LOCA source term.

For the control room doses, two executions of the CONDOSE code must again be made. The radionuclide release rate input decks to be used are those provided by the MHACALC code: C2STACK DATA and C2SIRW DATA The CONDOSE code input decks with the plant parameters and other information for this case are identical to the CRJCOl DATA and CRJC02 DATA decks used previously, with two changes other than the title which is obvious. In each of the two input decks, the number of time intervals for which release rates are to be specified, on line 22, and the end points for the time intervals on line 24 are changed to correspond to lines 41 and 43 of the JCOCASE2 DATA deck. Line 22 of both files is changed to 12 and line 24 of both files has 2880 minutes added to it. The .

filenames of the two files changed for this case are CR1CASE2 DATA and CR2CASE2 DATA and are listed on the attached microfiche.

The CONDOSE code was executed once for each of the two release paths. The first execution was with the CR1CASE2 DATA and C2STACK DATA decks, which resulted in the CR1CASE2 LISTING file with the control rooJD. doses from the containment atmosphere and ESF leakage. The second execution was with the CR1CASE2 DATA and C2SIRW DATA decks, which resulted in the CR2CASE2 LISTING file with the control room doses from the SIRW Tank release. Both listing files are listed on the attached microfiche. The output files created for plotting the control room air concentration versus time were discarded since they were of no interest ill' this analysis.

As can be seen on CR1CASE2 LISTING, the resultant control room operator doses due to containment atmosphere and ESF leakage are 8.387 rem thyroid, 1.205 rem skin, and 0.051 rem whole body. After adding the conservative 0.400 rem shine dose from containment [Ref. 2.45], the whole body dose is 0.451 rem. To these doses must be added the calculated dose from the release from the SIRW Tank, which from CR2CASE2 LISTING can be seen to be 0.253 rem thyroid, 0.000 rem skin, and 0.000 rem whole body. Therefore, the total dose to the operators before considering SIRW Tank shine is 8.640 rem thyroid, 1.205 rem skin, and 0.451 rem whole body.

PALISADES NUCLEAR PLANT EA-A-NL-92-012-02 ANALYSIS CONTINUATION SHEET Sheet 66 Rev #

1 ATIACHMENT 3 The shine dose in the control room from the SIRW Tank is calculated using the MICROSHIELD code as was done previously in this analysis. All inputs to the MICROSHIELD are the same as were used previously, as listed in section 6.4 of this analysis, with the exception of the case titles and the source activity at each time point for which the dose rate is calculated. The MICROSHIELD code was executed for each of the times at which the activity in the SIRW Tank was listed in JCOCASE2 LISTING, changing the source activity to the appropriate value for each execution. The increasing SIRW Tank water level due to the large amount of inleakage was conservatively ignored. The printouts of each of the executions are on the following pages. As was done previously, the dose rate calculated at each point in time is assumed constant until the next calculated point. The calculation of the dose received in the control room using the MICROSHIELD results is shown below. Control room occupancy factors are conservatively ignored.

Time Interval (mrem) 19 ... 120 min 13.34 (120-19)/60 = 22.456 120 ... 480 min 25.42 (480-120)/60 = 152.520 480 ... 720 min 48.53 (720-480)/60 = 194.120 720 ... 1440 min 47.66 (1440-720)/60 = 571.920 1440 ... 2880 min 28.62 (2880-1440)/60 = 686.880 2880 ... 4320 min 7.007 (4320-2880)/60 = 168.168 4320 ... 5760 min 1.772 (5760-4320)/60 = 42.528 5760 ... 7200 min 0.7998 (7200-5760)/60 = 19.195 7200 ... 14400 min 0.4640 (14400-7200)/60 = 55.680 14400 ... 21600 min 0.1612 (21600-14400)/60 = 19.344 21600 ... 28800 min 0.1033 (28800-21600)/60 = 12.396 28800 ... 36000 min 0.06785 (36000-28800)/60 = 8.142 36000 ... 43200 min 0.04168 (43200-36000)/60 = 5.002 E = 1958.35 mrem As can be seen above, the total shine dose in the control room from the SIRW Tank over 30 days is 1958.35 mrem, or l.958 rem. This results in a total control room whole body dose of 0.451 + 1.958 =

2.409 rem for the case of 20 gpm leakage of sump water into the SIRW Tank through the main discharge lines* for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . As mentioned above the total control room thyroid and skin doses from the event are 8.640 rem and 1.205 rem, respectively. These doses 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.

A summary of the offsite and control room doses for this case of 20 gpm leakage into the SIRW Tank through the main discharge lines for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months are shown in Table 8. As can be seen, there is still a significant amount of margin between the calculated doses and the 10 CFR 50 and 10 CFR 100 limits using the plant specific LOCA source term.

,____**-*n .....as PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET EA-A-NL-92-012-02 Sheet 67 __,

Rev # --=-I__

ATTACHMENT 3 TABLE 8 DOSES WITH 20 GPM LEAKAGE THRU MAIN DISCHARGE FOR 48 HOURS I I I I

I I

THYROID I I

I SKIN l WHOLE BODY (Re~)

I I I LOCATION I (Rem) I I (Rem)

SB (0-2 Hr) 7.648 N/A 0.057 LPZ 4.550 N/A 0.011 Control Room 8.640 1.205 2.409

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev I ~l~

Microshield 3.13

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(Consumers Power Company - #037)

Page 1 File Ref:

File 60MIN.MSH Date: _ / _ / _

Run date: March 18, 1992 By:

Run time: 3:31 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 60 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ......**......*.****..*** X 454.660 cm.

Source cylinder radius .......*..........*.*** R 701.040 II Source cylinder length ....*..**......*.****** Tl 60.960 II Thickness of second shield ***..*...***.*****. T2 2.540 II Thickness of third shield **.....*.....**..*.* T3 55.880 II Thickness of fourth shield **.***......**.**** T4 121. 920 II Thickness of fifth shield .**...**.*........** TS 30.480 II Microshield inserted air gap *..*..*********** air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air ~001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _§j__ Rev # ~l~

Page 2 File: 60MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 60 MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) ..*.....*...... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 2.0640e+04 I-132 2.1750e+04 I-133 3.2800e+04 I-134 2.0010e+04 I-135 3.1270e+04 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1312 6.271e+13 2.006e+03 3.191e+OO 2 1.6224 5.986e+14 4.249e+03 7.339e+OO 3 1. 1702 1.177e+15 1. 035e+03 1. 926e+OO 4 .8487 2.427e+15 3.813e+02 7.591e-Ol 5 .6587 1. 469e+15 4.592e+Ol 9.533e-02 6 .5204 1.518e+15 1.151e+Ol 2.359e-02 7 .3724 7.547e+14 3.503e-:Ol 7.204e-04 8 .2786 1. 279e+14 3.770e-03 7.498e-06 9 .2164 3.300e+13 7.843e-05 1. 476e-07 10 .1355 3.752e+13 2.049e-07 3.406e-10 11 12 13 14 15 16 17 18 19 20 8.206e+15 7.729e+03 1. 334e+Ol

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _IQ_ Rev # ~l~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File 60MIN.MSH Date: ~-/~_/~

Run date: March 18, 1992 By:

Run time: 3:34 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 120 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ....****..******.*.****** X 454.660 cm.

Source cylinder radius ****....*****..******** R 701.040 II Source cylinder length *..**..****.***.******* Tl 60.960 II Thickness of second shield **.**************** T2 2.540 II Thickness of third shield ****.**...*****..*** T3 55.880 II Thickness of fourth shield **.******.****.**** T4 121. 920 II Thickness of fifth shield *****.************** TS 30.480 II Microshield inserted air gap ..******..**.**** air 182.880 II Source Volume: 9.41197e+7 cubic centimeters M..~TERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air

.001220

.001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium

SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ....Il_ Rev # ~l~

Page 2 File: 60MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 120 MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

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

5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131 5.0660e+04 I-132 3.9640e+04 I-133 7.8140e+04 I-134 2.2350e+04 I-135 6.9370e+04 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1371 1.218e+14 3.940e+03 6.260e+OO 2 1. 6209 1.143e+15 8.064e+03 1. 393e+Ol 3 1.1766 2.329e+15 2.107e+03 3.917e+OO 4 .8444 3.562e+15 5.445e+02 1.085e+OO 5 .6603 2.557e+15 8.114e+Ol 1.684e-Ol 6 .5214 3.315e+15 2.544e+Ol 5.217e-02 7 .3711 1. 745e+15 7.825e-Ol 1.609e-03 8 .2815 2.709e+14 8~956e-03 1.784e-05 9 .2183 6.554e+13 1.688e-04 3.185e-07 10 .1359 4.325e+13 2.498e-07 4.155e-10 11 12 13 14 15 16 17 18 19 20 TOTALS: 1.515e+16 1.476e+04 2.542e+Ol

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ....1.1,_ Rev # _1_

Microshield 3~13

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(Consumers Power Company - #037)

Page 1 File Ref:

File : 120MIN.MSH Date: ~-/~_/~

Run date: March 18, 1992 By:

Run time: 3:38 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 480 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ...*...*..***.....*..**** x 454.660 cm.

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 ......****.********* TS 30.480 Microshield inserted air gap ................ . air 182.880 source Volume: 9.41197e+7 cubic centimeters M..ATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _].1_ Rev # _l_

Page 2 File: 120MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 480 MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide curies Nuclide curies Nuclide curies I-131 2.2630e+05 I-132 2.9660e+04 I-133 2.9200e+05 I-134 8.8740e+02 I-135 1. 6870e+05 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1766 2.035e+14 7.080e+03

1.116e+Ol 2 1.6327 2.177e+15 l.604e+04 2.768e+Ol 3 1.1845 4.915e+15 4.606e+03 8.550e+OO 4 .8447 2.570e+15 3.937e+02 7.842e-Ol

5 6

7 8

9 10 11

.6626

.5228

.3687

.2862

.2190

.1433 2.633e+15 l.046e+16 7.159e+l5 8.076e+14 1.474e+14 4.179e+l2 8.531e+Ol 8.157e+Ol J.021e+oo 3.220e-02 3.898e-04 6.69le~os 1.770e-Ol 1.673e-Ol 6.214e-03 6.427e-05 7.360e-07 l.136e--10 12 13 14 15 16 17 18 19 20 TOTALS:

3.107e+l6 2.829e+04 4.853e+Ol

ATTACHMENT 3 EA-A-NL-92-012-02 File Ref: Sheet __.I!_ Rev # ~l~

File : 480MIN.MSH Date: _ / _ / _

Run date: March 18, 1992 By:

Run time: 3:40 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 720 MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ..*.*.*.*.......*.*.*.*** x 454.660 cm.

Source cylinder radius .*..**.*.....*.*.*.***. R 701. 040 "

Source cylinder length *..*..**..*..****.*.*** Tl 60.960 II Thickness of second shield *.*....*...*...*.** T2 2.540 II Thickness of third shield ..*..*...*.*...*.*.. T3 55.880 II Thickness of fourth shield *.*.*.*.***.*...**. T4 12i. 920 II Thickness of fifth shield *.....*.*.*.*.*.***. .TS 30.480 "

Microshield inserted air gap ................ . air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # ~l~

Page 2 File: 480MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 720 MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius) ***.**...** 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide curies I-131 3.3920e+05 I-132

1. 3510e+04 I-133 3.8860e+05 I-134 5.70~0e+Ol I-135 1.6870e+05 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) {photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1919 1.853e+14 6.632e+03 l.043e+Ol 2 1. 6379 2.118e+15 1. 591e+04 2.743e+Ol 3 1.1860 4.995e+15 4.714e+03 8.748e+OO 4 .8556 2.084e+15 3.422e+02 6.801e-Ol 5 .6615 2.301e+15 7.384e+Ol 1.532e-Ol 6 .5229 1.350e+16 1. 054e+02 2.162e-Ol 7 .3681 1.057e+16 4.396e+OO 9.042e-03 8 .2867 1.063e+15 4.335e-02 8.656e-05 9 .2165 1.573e+14 3.754e-04 7.065e-07 10 .1474 1.286e+12 3.610e-08 6.194e-11 11 12 13 14 15 16 17 18 19 20 TOTALS:

3.697e+16 2.778e+04 4.766e+Ol

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _1.§_ Rev # _1_

Microshield 3.13

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{Consumers Power Company - #037)

Page 1

File Ref:

File 720MIN.MSH Date: _ / _ / _

Run date: March 18, 1992 By:

Run time: 3:43 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 1400MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ..***........**......**.* X 454.660 cm.

Source cylinder radius .....**..**....***..*** R 701. 040 II Source cylinder length ***.*.....****.**...*** Tl 60.960 II Thickness of second shield **.....**.***.*.*** T2 2.540 II Thickness of third shield *...*.*..*.***..*.*. T3 55.880 II Thickness of fourth shield ***.*..*****...**** T4 121. 920 , II Thickness of fifth shield ****....*******.**** TS 30.480 II Microshield inserted air gap ****.***.****.*** air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES {g/cc) :

Material Source Shield 2

Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350

~. Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _))_ Rev # ~1~

Page 2 File: 720MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 1400MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide curies Nuclide curies Nuclide Curies I-131 6.5860e+05 I-132 7.3600e+02 I-133

5. 2810e+05 I-134 8.7490e-03 I-135 9.7130e+04 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) - MeV/ (sq cm)/sec (mr/hr) 1 2.2057 9.890e+13 3.629e+03 5.691e+OO 2 1.6422 1.196e+15 9.126e+03 1.572e+Ol 3 1.1954 3.367e+15 3.314e+03 6.137e+OO 4 .8682 1.643e+15 2.925e+02 5.800e-Ol 5 .6581 2.866e+15 8.909e+Ol 1.850e-Ol 6 .5229 1.780e+16 1.389e+02 2.850e-Ol 7 .3673 2.012e+l6 8.193e+OO l.685e-02 8 .2874 1.696e+15 7.119e-02 1.422e-04 9 .2042 1. 356e+14 1. 960e-04 3.634e-07 10 .1484 6.452e+10 2.lOOe-09 3.614e-12 11 12 13 14 15 16 17 18 19 20 TOTALS:

4.892e+16 1.660e+04 2.862e+Ol

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # _l_

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : 1440MIN.MSH Date: __ / __ / __

Run date: March 18, 1992 By:

Run time: 3:45 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 2880MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ..*.*............*......* x 454.660 cm.

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 *.......*....*...*.* TS 30.480 Microshield inserted air gap ................ . air 182.880 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum carbon 2.350 2.350 Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet -11_ Rev # _l_

Page 2 File: 1440MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 2880MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) ...*..******... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131

1. 2160e+06 I-132

1. 01ooe+oo I-133 4.7790e+05 I-134 1.0080e-10 I-135 1. 5780e+04 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 1.594e+13 5.864e+02 9.194e-Ol 2 1.6427 1.939e+14 l.482e+03 2.553e+OO 3 1.2252 1.223e+15 1.374e+03 2.528e+OO 4 .8689 l.209e+15 2.162e+02 4.287e-Ol 5 .6548 4.657e+15 1.403e+02 2.915e-Ol 6 .5228 1. 596e+16 1.244e+02 2.552e-Ol 7 .3670 3.682e+16 1.486e+Ol 3.058e-02 8 .2882 2.827e+15 1. 225e-01 2.448e-04 9 .1838 1. 307e+14 5.316e-05 9.603e-08

. 10 .1484 9.378e+07 3.054e-12 5. 256e 11 12 13 14 15 16 17 18 19 20 TOTALS:

6.304e+l6 3.938e+03 7.007e+OO

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _]Q_ Rev # _1_

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File 2880MIN.MSH Date: _ / _ / _

Run date: March 18, 1992 By:

Run time: 3:47 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 4320MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector .....*....*.............. x 454.660 cm.

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 ......***..***....** TS 30.480 Microshield inserted air gap .....**..***...** air 182.880 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _§]__ Rev # ~l~

Page 2 File: 2880MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 4320MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.1190e+06 I-132 7.7500e-04 I-133 2.1530e+05 I-134 5.7790e-19 I-135 1.2760e+03 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072*

1.289e+12 4.742e+Ol 7.434e-02 2 1.6427 1.568e+13 1.198e+02 2.065e-Ol 3 1.2433 4.035e+14 4.969e+02 9.115e-Ol

.8688 5.245e+14 9.376e+Ol 1.859e-Ol 4

5 .6531 4.057e+15 1. 203e+02 2.5ooe-01 6 .5226 7.247e+15 5.640e+Ol 1.157e-Ol 7 .3670 3.383e+16 1. 364e+Ol 2.807e-02 8 .2887 2.545e+15 l.121e-Ol 2.241e-04 9 .1801 1.106e+14 3.489e-05 6.273e-08 10 .1484 6.793e+04 2.212e-15 3.807e-18 11 12 13 14 15 16 17 18 19 20 TOTALS:

4.874e+16 9.484e+02 1.772e+OO

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # ~l~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : 4320MIN.MSH Date: ~-/~_/~

Run date: March 18, 1992 By:

Run time: 3:48 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 5760MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . * . X 454.660 cm.

Source cylinder radius .........*.........*.** R 701.040 II Source cylinder length .*....*...**.*...*.**.. Tl 60.960 II Thickness of second shield .*...**........**.* T2 2.540 II Thickness of third shield ..*..*..........**** TJ 55.880 II Thickness of fourth.shield **...*....*...*.*.. T4 121. 920 II Thickness of fifth shield .*....*..........**. TS 30.480 II Microshield inserted air gap. * * * . . . . . . . . . * . *. . air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet __§!__ Rev # ~1~

Page 2 File: 4320MIN.MSH CASE: EA-A-NL~92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 5760MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide curies I-131 1.0290e+06 I-132 5.6110e-07 I-133 9.7010e+04 I-134 o.ooooe+oo I-135 1.0330e+02 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2 .*1423 1.264e+ll 4.129e+OO 6.554e-03 2 1. 6393 1. 247e+12 9.415e+OO 1. 623e-02 3 1.2485 1.696e+14 2.143e+02 3.928e-Ol 4 .8762 2.186e+14 4.096e+Ol 8.108e-02 5 .6528 3.647e+15 1.078e+02 2.24le-Ol 6 .5226 3.323e+15 2.586e+Ol 5.303e-02 7 . 3672 . 3.093e+16 1.255e+Ol 2.582e-02 8 .2919 2.492e+15 1.249e-Ol 2.500e-04 9 .2576 1.288e+l3 1.634e-04 3.216e-07 10 .1797 1.008e+14 3.097e-05 5.565e-08 11 12 13 14 15 16 17 18 19 20 4.089e+16 4.152e+02 7.998e-Ol

ATIACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev I ~1~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File Run date:

7200MIN.MSH March 18, 1992 Date:

By: ~-'~-'~

Run time: 3:50 p.m

Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 7200MIN GEOMETRY lO:_cylindrical source from end - slab shields Distance to detector ......................*.* X 454.660 cm.

Source cylinder radius ....................*.. R 701.040 II Source cylinder length ..........*.*.....*.*.. Tl 60.960 II Thickness of second shield ................*.. T2 2.540 II Thickness of third shield .....*.........*.**. T3 55.880 II Thickness of fourth shield ................**. T4 121. 920 II Thickness of fifth shield ..*....*....*....**. T5 30.480 II Microshield inserted air gap *..**...**.* ~**** air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shfeld 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # ~l~

Page 2 File: 7200MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE- 7200MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius) ........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide curies I-131 9.4630e+05 I-132 4.0630e-10 I-133 4.3710e+04 I-134 o.ooooe+oo I-135 8.3530e+OO RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.1423 1.022e+10 3.339e-Ol 5.299e-04 2 1.6393 1.008e+ll 7.613e-Ol. l.312e-03 3 1.2491 7.557e+13 9.580e+Ol 1.755e-Ol 4 .8760 9.825e+13 1. 839e+Ol 3.640e-02 5 .6521 3.301e+15 9.691e+Ol 2.015e-Ol 6 .5220 1.561e+15 1.206e+Ol 2.474e-02 7 .3672 2.843e+16 1.153e+Ol 2.373e-02 8 .2919 2.289e+15 1.148e-Ol 2.298e-04 9 .2578 5.772e+12 7.370e-05 1.450e-07 10 .1797 9.273e+13 2.848e-05 5.118e-08

. 11 12 13 14 15 16 17 18 19 20 TOTALS:

3.585e+16 2.359e+02 4.640e-Ol

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ___§_[__ Rev # ~1~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : 14400MIN.MSH Date: ~-/~_/~

Run date: March 18, 1992 By:

Run time: 3:52 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-14400MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector . . . . . . . . . . . . . . . . . . . . . . . . . x 454.660 cm.

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 .*.*****...*.**...** TS 30.480 Microshield inserted air gap ................ . air 182.880 Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap ,

Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # ~l~

Page 2 File: 14400MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-14400MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 6.2260e+05 I-132 o.ooooe+oo

. I-133 8.1140e+02 I-134 o.ooooe+oo I-135 2.8930e-05 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 2.922e+04 ----------------

1.075e-06 ---------

1.685e-09 2 1.6427 3.555e+05 2.717e-06 4.681e-09 3 1.2829 1.193e+12 1.792e+OO 3.265e-03 4 .9061 1.613e+12 3.650e-Ol 7.18le-04 5 . 72-67 4.164e+14 2.455e+Ol 5.019e-02 6 .6316 1.749e+15 4.344e+Ol 9.018e-02 7 .5077 8.373e+13 5.466e-Ol 1.119e-03 8 .3670 1.88le+16 7.581e+OO 1. 560e-02 9 .2891 1.394e+15 6.243e-02 1.248e-04 10 .1797 6.101e+13 l.874e-05 3.367e-08 11 12 13 14 15 16 17 18 19 20 TO!l'AL.S:

2.252e+16 7.834e+Ol ---------

1.612e-Ol.

tft.;~~~~-

ATIACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev I ~1~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : 21600MIN.MSH Date: _ / _ / _

Run date: March 18, 1992 By:

Run time: 3: 5::3' p.m. Checked:

I CASE: EA-A~NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-21600MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector .**....**..*........***** X 454.660 cm.

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 ....*****.....*.**** TS 30.480 "

Microshield inserted air gap ******..****.**** air 182.880 "

Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350

Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # ~l~

Page 2 File: 21600MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-21600MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide Curies I-131 4.0950e+05 I-132 o.ooooe+oo I-133 1.5060e+Ol I-134 o.ooooe+oo I-135 1.002oe-10 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 1.012e-Ol 3.723e-12 5.838e-15 2 1.6427 1. 231e+OO 9.411e-12 1. 621e-14 3 1.2829 2.213e+10 3.327.e-02 6.060e-05

4 5

6 7

8 9

10

.9061

.7266

.6332

.5078

.3670

.2891

.1797 2.994e+10 2.731e+14 1.134e+15 5.464e+13

1. 237e+16 9.170e+14 4.013e+l3 6.774e-03 1.608e+Ol 2.852e+Ol 3.573e-Ol 4.986e+OO 4.106e-02

1. 232e-05

1. 333e-05 3.289e-02 5.921e-02 7.315e-04

1. 026e-02 8 .. 208e-05 2.21se-08 11 12 13 14 15 16 17 18 19 20 TOTALS: 1. 4 79e+16 5.Q03e+Ol 1. 033e-Ol

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~*Rev# ~l~

Microshield 3.13 (Consumers Power Company - #037)

Page 1 File Ref:

File 21600MIN.MSH Date: ~-/~_/~

Run date: March 18, 1992 By:

Run time: 3:55 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-28800MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ...........*..*....*....* X 454.660 cm.

Source cylinder radius . . . . . . . . . . . . . . . . . . . . . . . R 701.040 "

Source cylinder length ......*..*......*...... Tl 60.960 II Thickness of second shield .....*............. T2 2.540 "

Thickness of third shield ..*...*......**.**.. T3 55.880 II Thickness of fourth shield ....**...*..*.*...* T4 121.920 II Thickness of fifth shield ..................** TS 30.480 II Microshield inserted air gap ...**....*..***** air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon 2.350 2.350 Concrete Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet __il_ Rev # ~1~

Page 2 File: 21600MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-28800MIN

BUILDUP FACTOR: based on TAYLOR method

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide curies Nuclide curies I-131 2.6930e+05 I-132 o.ooooe+oo I-133 2.7950e-Ol I-134 o.ooooe+oo I-135 3.4690e-16 RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 2.2072 3.503e-07

1. 289e-17 2.021e-20 2 1.6427 4.262e-06 3.258e-17 5.613e-20 3 1.2829 4.108e+08 6.174e-04 l.125e-06 4 .9061 5.556e+08 l.257e-04 2.474e-07 5 .7266 1. 796e+l4 1. 058e+Ol 2.163e-02 6 .6333 7.453e+l4 l.875e+Ol 3.894e-02 7 .5078 3.592e+l3 2.349e-Ol 4.810e-04 8 .3670 8.135e+15 3.279e+OO 6.746e-03 9 .2891 6.030e+14 2.700e-02 5.398e-05 10 .1797 2.639e+13 8.lOSe-06 1.456e-08 11 12 13 14 15 16 17 18 19 20

TOTALS:

9.725e+15 3.287e+Ol 6.785e-02

~:£-*.-;;,.:_-

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet _j_g_ Rev # ~1~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File : 28800MIN.MSH Date: _ / _ / _

Run date: March 18, 1992 By:

Run time: 3:56 p.m. Checked:

CASE: EA-A~NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-36000MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector. * . . . . . . . . . . . . . . . . . . . . * * . X 454.660 cm.

Source cylinder radius .......*...*.*..*...*.. R 701. 040 II source cylinder length ...............**..**.* T1 60.960 II Thickness of second shield *.......**...*.**.* T2 2.540 II Thickness of third shield .........**......*** T3 55.880 II Thickness of fourth shield .*.....***.*..*.**. T4 121.920 II Thickness of fifth shi~ld .....**.......*...** TS 30.480 II Microshield inserted air gap ...***....*..*..* air 182.880 II Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc):

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap*

Air .001220 .001220 Alwninum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1.0 Zirconiwn SANO 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet --21_ Rev # ~l~

Page 2 File: 28800MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRO DISCHARGE-36000MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi)............... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies I-131 1.7710e+05 r-132 o.ooooe+oo I-133 5.1860e-03 I-134 o.ooooe+oo I-135 o.ooooe+oo RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 1.2829 7.622e+06 1.146e-05 2.087e-08 2 1. 0216 1.729e+06 7.683e-07 1.476e-09 3 .8673 1. 212e+07 2.145e-06 4.255e-09 4 .6514 6.083e+14 1.774e+Ol 3.689e-02 5 .5078 2.362e+13 1.545e-Ol 3.163e-04 6 .4245 8.861e+05 l.355e-09 2.775e-12 7 .3671 5.334e+15 2.158e+OO 4.439e-03 8 .2903 4.130e+14 1.944e-02 3.888e-05 9 .2578 6.841e+05 8.746e-12 1.721e-14 10 .1797 1. 735e+13 5.330e-06 9.578e-09 11 12 13 14 15 16 17 18 19 20 6.396e+15 2.007e+Ol 4.168e-02 I

_I

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet ~ Rev # ~l~

Microshield 3.13

=====

(Consumers Power Company - #037)

Page 1 File Ref:

File 36000MIN.MSH Date: ~-/~_/~

Run date: March 18, 1992 By:

Run time: 3:58 p.m. Checked:

CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-43200MIN GEOMETRY 10: Cylindrical source from end - slab shields Distance to detector ...........*......*...*.* X 454.660 cm.

Source cylinder radius .......*........**....* R 701. 040 II Source cylinder length ...*...*.*..........*.* Tl 60.960 II Thickness of second shield .*..........*.***** T2 2.540 II Thickness of third shield .*...*.......*.*..** T3 55.880 II Thickness of fourth shield ..**. ~~ *..*.....*** T4 121.920 II Thickness of fifth shield .**.*...**...**..*.. TS 30.480 "

Microshield inserted air gap .*.......**.*..** air 182.880 "

Source Volume: 9.41197e+7 cubic centimeters MATERIAL DENSITIES (g/cc) :

Material Source Shield 2 Shield 3 Shield 4 Shield 5 Air gap Air .001220 .001220 Aluminum Carbon Concrete 2.350 2.350 Hydrogen Iron Lead Lithium Nickel Tin Titanium Tungsten Urania Uranium Water 1. 0 Zirconium SAND 2.0

ATTACHMENT 3 EA-A-NL-92-012-02 Sheet --2.§_ Rev # ~l~

Page 2 File: 36000MIN.MSH CASE: EA-A-NL-92-012-02 Rev 1, 20 GPM LEAK THRU DISCHARGE-43200MIN

BUILDUP FACTOR: based on TAYLOR method.

Using the characteristics of the materials in shield 1.

INTEGRATION PARAMETERS:

Number of angle segments (Npsi) .***........... 5 Number of radial segments (Nradius)........... 11 SOURCE NUCLIDES:

Nuclide Curies Nuclide Curies Nuclide Curies 1.1650e+05 I-132 o.ooooe+oo I-133 9.6220e-05 I-131 I-134 o.ooooe+oo I-135 o.ooooe+oo RESULTS:

Group Energy Activity Dose point flux Dose rate

(MeV) (photons/sec) MeV/(sq cm)/sec (mr/hr) 1 1.2829

1. 414e+05 2.126e-07 3.872e-10 2 1.0216 3.208e+04 1. 425e-08 2.739e-ll 3 .8673 2.249e+05 3.981e-08 7.894e-11

'4 .6514 4.001e+14 1.167e+Ol 2.426e-02 5 .5078 1. 554e+13 1.016e-Ol 2.08le-04 6 .4245 1. 644e+04 2.514e-11 5.149e-14 7 .3671 3.509e+15 1. 419e+OO 2.920e-03 8 .2903 2.717e+14 1. 279e-02 2.557e-05 9 .2578 1. 269e+04 1.623e-13 3.193e-16 10 .1797 1.142e+13 3.506e-06 6.300e-09 11 12 13 14 15 16 17 18 19 20 TOTALS:

~::,.,.--*~--*

4.207e+15

1. 320e+Ol 2.742e-02

Form 3698 9-89 PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS CHECKLIST Affected Revision Items Affected By This EA Yes No Required Identify* Closeout

1. Other EAs 0 E(

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 0 -)(]_

5.2 Working Procedures 0 g 5.3 Tech Spec Surveillance Procedures 0 @

6.0 OTHER DOCUMENTS 6.1 Q-List 0 ~

6.2 Plant Drawings 0 jg' 6.3 Equipment Data Base 0 ~

6.4 Spare Parts (Stock/MMS) 0 ~

6.5 Fire Protection Program Report 0 [81 (FPPR) 6.6 Design Basis Documents 0 l8f 6.7 Operating Checklists 0 g 6.8 SPCC/PIPP Oil and Hazardous

-Material Spill Prevention Plan 0 ~

6.9 EEQ Documents 0 £(

Do any of the followingJ;ig(:uments need to be generated as a result of this EA:

Yes No

1. Corrective Action Document? 0 ~ Reference

2. Safety Evaluation? 0 ~ Reference

3. EEQ Evaluation Sheet? 0 ~ Reference Is PRC Review of this EA Required? 0 .er Completed By ---+J?.~"+'/_._.£_1};.t--"'--~,-J/a,µ.tl--'/.'-'J-==e--=-.P L - > = - - - - - - - - - - - - - - Date -'2=,L,b-L.C'-7_,_d--'-7C.z=---

Identify Section, No, Drawing, Document, etc.

Proc No 9.11 TECHJIJQL REYI[V CHECKLIST Attachment 5 Revision 5 EA - !l-41L - UtJ12_-4!KEV. §< 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 is noted by the Techn;cally Reviewed signature on the Initiation and Review record block of Fon11 3619.

(Y, N, N/A)

1. Have the proper input 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 the basis for their use documented? I l 3/q/q

4. Is Vendor info~t1on used as input addressed correctly in 1c.'i) th* analysis? .J

5. If th* analysts argW11nt departs frOll Vendor Info...atton/R1conmendations, is the departure just1f1cat1on docWDented? '/_

6. Art assumptions accurately d1scribed and rt1sonabl1? '/

7. Has th1 use of 1ngin11ring judge111nt bttn docU111nted and 1 justified? *

8. Art all constants, var1ablts and for'9Ulas correct and y properly applied?

9. Have any *inor (1ns1gn1ftcant) errors been 1dent1f1ed? If see ~ ".?1 IG yes; Id1nt1f1 on the 3110 Fon1 ancljust1f1 their 1ns1gn1f1cance. (

10. Does analysis involve tiMldtng? If Yes; vertfy the following 1nfo1"91tion ts accurately represented on the analysts drawing (OUtput docU111nt).

Ty119 of Velcl

St-za, of V.lct

"'*-111 Being Joined .

r.t~ss of Material Being Joined

Locat.foa of Veld(s)

Approprtata Veld S,Yllbology

11. Has the objective of the analysis been ..t? f

12. H*ave adll1n1strath* requ1rtMnts such as n..a.rtng and fo...at been satisfied? I

DocumentTitle O~F;5"/rc NUCLEAR OPE.NS DEPARTMENT Docume

.AA/£' ce-,c,recL r?i!?tl.-if //"'98 /T~8/L I view Sheet ry . Document Number Revision Revision Number E.-9-#~ - >'.Z-e'J/Z-<=>.2 Page of ~

Item Page and/or Number Section Number Re1pon1& or Resolution Tfif" .STEPS ff<i)Jl?f'D ltJ Sl&H'T 6tJ INP\Jfj Bf. u>Eo '~ ii-IE AtJ /4.L Y .n..r ,.

)(E -131 /11\

@) TFW:.Lc I ~'2.

.s.... -.. ..-f +k A ,_)UC. L U)~ *

~~~-1--~-'----'-""°-~~+--~~~~

wr: ~ iseo To

~~~--~~~~~~~~~Tl.J-~*~12~E&~-Gru10£

( 'J.z<.) f/M~ (-1;)(1.. Tl.JES:-E:

-~_;_;_:-==~~~~~~-'---'-=-~--'---=-'--=-~~~

FlLTE2.S, <££[ SK\? ~f'QJ1tte:S Reviewer Organization Review O>ordinator Date Date

-r.e. 'DvF"FY Rt-SA ?.. A. /./A~D£,J '-"}/~

form 3110 1-82

Nudear Consulting Services, Inc. TELEPHONE:

OUTSIDE OHIO:

(614) 846-5710 1-800-992-5192 F'.O. BOX 29151 7000 HUNTLEY AOAO TELEX: 697441 5

. COLUMBUS, OHIO 43229 U.S.A. FAX: (614) 431--0858 IODINE-131 REMOVAL EFFICIENCY DETERMINATION OF ADSORBENT SAMPLES PERFORMED ON SAMPLE:

VF-26 S/N 722863 March 4, 1992 PERFORMED FOR:

CONSUMERS POWER CO., PALISADES 27780 BLUE STAR MEM HWY COVERT, MI 49043 P. O. No. G0020386 DISTRIBUTION: # COPIES

EXTERNAL l LARRY PHILLIPS INTERNAL l PROJECT MASTER FILE l I-LAB FILE l QA FILE PARAMETERS PER:

ASTM 03803-1989.

I-LAB ID# G362 NU CON* l3CI759/03

*REGISTERED TRADEMARK OF NUCLEAR CONSULTING SERVICES, INC.

Page l Date Performed: 3/04/92

.mwcoN l3CI759/03

~allenge Agent concentration: CH3I,l.75 mg/m~3 Sample: VF-26 S/N 722863 Conditions: Temperature Relative Hwnidity Loading 30.0 c ( 86.0 F) 70 %

Elution 30.0 c ( 86.0 F) 70 %

R E S U L T S Bed No. Bed Depth (in) Efficiency (%) Penetration (%) St. Dev. (%)

l 2 99.981 0.019 0.002

Performed by~~::::....4..::.~~~;.:::=:;;:2:i:;:;:~~~-=:...L~.!.....~~-===~~=--~~~~~_,,....,.....,~-

Nuclear consulting Services, Inc.

Approved by________ ~~~~-~~"-oQl~.....:-~~.;.Ao'.P.l~~~~~~-=:.------=--~---.---------------

w. Peter Freeman For Nuclear consulting Services, Inc.

The results are the best obtainable under current experimental techni~es and according to our best knowledge. Counting data availaJt!,,~ upon request

Page 2 Date Performed: 3/04/92 l3CI759/03 Sample: VF-26 S/N 722863 C O N D I T I O N s:

Temperature Relative Humidity (%) Test Times Equilibration 30.0 c 86.0 F) 70 18.0 Hrs.

Loading 30.0 c ( 86.0 F) 70 60.0 Mins.

Elution 30.0 c ( 86.0 F) 70 60.0 Mins.

Face Velocity 12.20 m/min ( 40. o fpm)

.iJliiallenge Agent Concentration: CHJI,l.75 mg/m"3

._,essure: 101.30 kPa (l.OO atm)

Bed Bed Depth Number (in) (mm) 1 2 so.a

Page 3 C E R T I F I C A T E O F C OMP L I A N C E

Client: CONSUMERS POWER co., PALISADES 27780 BLUE STAR MEM HWY COVERT, MI 49043 Date: March 4, 1992 Project: l3CI759/03 Purchase Order: G0020386 P. o. Date: 2/03/92 NUCON Product/Service: Iodine-131 testing performed on VF-26 S/N 722863 This is to certify that NUCON Product/Service conforms to Specification(s) listed below Of NOCON: NUCON QA Manual Rev 13 Dated: 25 February 1991 Of Client: ASTM 03803-1989 *

Exceptions: TEST RELATIVE HUMIDITY TO BE 70%.

Reviewed in accordance with checklist of NUCON Procedure 95 {QA38 Form attached)

Reviewed For Quality Assurance OepartJnent Nuclear Consulting Services, Inc .

QA38 Page 4 QA Checklist for Review of Inspection Test DocUlllents

l.

2.

Is the report signed by person performing inspection or test?

Has the report been reviewed and signed by an authorized person?

N N

3. Are acceptance criteria specified?

3a. If so, does it meet the customer's requirements?

4. Are inspections or test results satisfactory? N 4a. If not, was a non-conformance report generated?

s. If a non-conformance report was not generated, explain below.

6* comments __~~~~~I.=--~LA ___G.......,.;:,_~---'-----~~~~--------------~~~-

........&.__:r._u_:d:

Signature in Review Section of document indicates satisfactory review

Nudecr Consulting Services, Inc. TELEPHONE: (614) 846-5710 OUTSIDE OHIO: 1-800-992-5192 P.O. BOX 29151 7000 HUNTLEY ROAD TELEX: 6974415 COLUMBUS, OHIO 43229 U.S.A. FAX:_(614) 431-085.8_

IODINE-131 REMOVAL EFFICIENCY DETERMINATION OF ADSORBENT SAMPLES PERFORMED ON SAMPLE:

VF-26 S/N 722855 CONTROL ROOM IN-STOCK February 5, 1992 PERFORMED FOR:

CONSUMERS POWER CO., PALISADES 27780 BLUE STAR MEM HWY COVERT, MI 49043 P. 0. No. - G0020386 DISTRIBUTION: # COPIES

EXTERNAL l LARRY PHILLIPS INTERNAL l PROJECT MASTER FILE l I-LAB FILE l QA FILE PARAMETERS PER:

ASTM D3803-l989.

I-LAB ID# G321 NUCON* l3CI759/0l

*REGISTERED TRADEMARK OF NUCLEAR CONSULTING SERVICES, INC.

-, Paqe l Date Performed: 2/05/92 CON l3CI759/0l allenqe Aqent Concentration: CHJI,l.75 mq/mAJ

' Sample: VF-26 S/N 722855 CONTROL ROOM IN-STOCK Conditions: Temperature Relative Humidity Loadinq 30.0 c ( 86.0 F) 70 %

Elution 30.0 c ( 86.0 F) 70 %

R E S U L T S Bed No. Bed Depth (in) Efficiency (%) Penetration (%) St. Dev. (%)

1 2 99.974 0.026 0.001

Performed by~~~~~::::::)l~~:,.,_f4,.:_::::J~::::.=::_~~~~~=--~,__~~~~__,,,.......,~-

For Nuclear Consultinq services, Inc.

Approved by~~~~,.--=--""~~---~-~~---~-~~--

w. Peter Freeman

~=--~~~-=-~..,..._-......~~~~--~

For Nuclear Consultinq Services, Inc.

The reaQl,ts are the bast obtainable under current experimental techni~. and accordinq to our best knowledqe. Countinq data availaJ:t!~~upon request

Page 2 Date Performed: 2/05/92

. .CON 13CI759/0l Sample: VF-26 S/N 722855 CONTROL ROOM IN-STOCK C 0 N D I T I 0 N s:

Temperature Relative Humidity (%) Test Times Equilibration 30.0 .c ( 86.0 F) 70 18.0 Hrs.

Loading 30.0 c ( 86.0 F) 70 60.0 Mins.

Elution 30.0 c ( 86.0 F) 70 60.0 Mins.

Face Velocity 12.92 m/min (42 .4 fpm)

Challenqe Aqent concentration: CHJI,l.75 mq/m"'J

~ssure: 101.30 kPa (l.OO atm)

Bed Bed Depth Nwnber (in) (:mm)

l. 2 50.8

-~'t~Jt_:

Page 3 C E R T I F I CAT E 0 F C O MP L I A N C E

Client: CONSUMERS POWER co. , PALISADES 277JO BLUE STAR MEM HWY COVERT, MI 49043 Date: February 6, 1992 Project: 13CI759/0l PUrchase order: G0020386 P. o. Date: 2/03/92 NUCON Product/Service: Iodine-131 testing performed on VF-26 S/N 722855 CONTROL ROOM IN-STOCK This is to certify that NUCON Product/Service conforms to

Specification(s) listed below Of NUCON: NUCON QA Manual Rev 13 Dated: 25 February 1991 Of Client: ASTM 03803-1989.

Exceptions: TEST RELATIVE HUMIDITY TO BE 70%.

'Re!:.1b.:.ivc.s:

O. 2 3 6 sec MSGUMtOE TIME.

~ SMAUr4 1°'1'i?.

Reviewed in accordance with checklist of NUCON Procedure 95 (QA38 Form attached)

Reviewed For Quality Assurance Department Nuclear Consulting Services, Inc

QA.38 Page 4 QA Checklist tor Review of Inspection Test Documents

1.

2.

Is the report siqned by person perf orm:ing inspection or test?

/

Has the. report been reviewed and siqned by an authorized person?

© N

N

~

s. If a non-conformance report was not generated, explain below.

6. Comments l:-LAB "'&~~ G3Z.!

Siqnature in Review Section of document indicates satisfactory review *

.mers Power NUCLEAR OPERA.JS DEPARTMEl\IT Document Review Sheet company Document Title c>FF.S/rE PC'S<!:.5 .4AIR' cavr£t::J~ ~~ #,,'f8/7/fB/~/rr roLLt.?W/.c/G ,If IDocument Number ,.~Revision I IRevisio/umber LJ1.RoE AR~Av ~OCA -r~ :Tt/.;,,,--;..=r ,- - '/ J1U:-£) - . v..vr/L rVL.L.E /:2 fi.4*,,f-A/.l.-f.?-~~,,~-o IPage J of I Item Page and/or Number Section Number Comments Response or Resolution 0 A1f* 3, ~tJ~ ~ e.~_f'.(I(/; -h~ {h~,.. t~ile ..... f{_. ~ /1/JTe .6ee/7, dd'cld

~ }]~1i j!~ ,. -l*.*,

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~--------------

-- ~--------*

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f-------- f-------------- ~-- - ---- - - - - - - - - - * - * -*------------.

Reviewer T-c.'b VF~-(

IOrga~:i~A IDate1 hc/c:n.

I Review Coordinator *

~. A. H A12.\l~ IDate/ /c

~ 1-:. 9l

'Do~on~ IDateY~A~

L J=rirrn :l110 1 82 .

Proc No 9.11 TECllllCAL REJIEJI CHECKLIST Attachment 5 Revision 5 Page 1 of 1 This checkl~st provides guidance for the review of engineering analyses.

Answer q~*st1ons Yes or No, or N/A if they do not apply. OocW11nt all co11111ents"on, a 3110 Fornr. Satisfactory resolution of conments and completion of this--dlecklist 1s noted by the Technically Reviewed signature on the Initiation and Review record block of Font 3619.

{Y, N, N/A)

1. Have the proper input codas, standards and design y principles been specified?

2. Have the input codes, standards and design principles been properly applied? y I

3. Are all inputs and assumptions valid and ~h* basis for their use documented? y I

4. Is Vendor information used as input addressed correctly in th1 analysis?

5. If the inalysis argWllnt departs frOll Vendor Infon11tion/R1co1m11ndations, is the departure justification docU111ntld? - ,

6. Are assumptions accurately described and reasonable? y

7. Has --the use of enginHring judg...,.t been doc...,.tlCI anct v justified? * ' ___r_ __

a. Are all constants. variables and formlu correct and properly applied?

9. Have any *inor (insignift_cant) errors blH idmttftld? If JS.\!: Q.*-+... d.~J 1 , 1<:;)

yes; Identtfy on the 3110 Fol"ll ancl Justtfy thetr 1nsigntftcance. i

10. Does 'analysts tnvolve wldtng'I If Yes; vertfy the following tnforut1-01t ts accurately reprtsutlCI Oft th*

analysts drawing (OUtput doc...,.t).

T~ of V.td: -*,

- sta1nf'V.lct,_': '

., Miiif.lal Bet Jof nlCl-

~/illCD.il* ot"latartal Blfng Jo1nld

tiliM* _of Vald(s) 11.

a11111c

. °'-

-rt-ate'Velcl s....au.1 1- 091 Has the obJecttve of th* analysts been *t? __t ___

12. H*ave adllfntstrative requ1reMnts such as numbering ancl forut been sathfted? - __ r___

. NUCLEAR OPE.ONS DEPARTMENT Docum eview Sheet Document Number

E.;p--V.:.. - r.Z-t:!/z-o.2 Revision Revision Number Page of :3 0/'R~TJ(')._,.- V¢n'L Number Comment* Response or Resolution

'11*/. ~le... 1°., G~AMC.

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. <! * "Duf=" FY Rt-SA ?.. A. /./A1W£,J Form 3110 1-82

. NUCLEAR OPE.ONS DEPARTMENT Docume .eview Sheet I

~

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Date Form 3110 1-82

ML18058A992

Contents

Text

SUMMARY

2.0 REFERENCES

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

3.0 BACKGROUND

SUMMARY

SUMMARY

8.0 CONCLUSION

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ML18058A992

Text

SUMMARY

2.0 REFERENCES

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

3.0 BACKGROUND

SUMMARY

SUMMARY

8.0 CONCLUSION

=====

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