ML20217G528
ML20217G528 | |
Person / Time | |
---|---|
Site: | FitzPatrick |
Issue date: | 09/24/1999 |
From: | Golshani M, Re G POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
To: | |
Shared Package | |
ML20217G500 | List: |
References | |
JAF-CALC-RAD-00, JAF-CALC-RAD-0002-R1, NUDOCS 9910210310 | |
Download: ML20217G528 (400) | |
Text
{{#Wiki_filter:. CALCULATION CONTROL SHEET Page 1 of 2 Calculation No.: g ( JAF-CALC-RAD-00023 Rev. 1 MO% IE IVillbl Y Nod Prepared by: JAF Vendor WPo(x) James A. FitzPatrick Nuclear Power Plant ha*M or Teek No:
- Power Uprate Program - Technical Support Center Post-Accident Radiological Habitability Study Discipline: System (number and name): DBD No.
1 n ' REV. Description & Status Prepared by: Name Checked by: Name Approved by: N?n.i D.V. (F,P,1,v or S) Signature /Date Signature /Date Signature /Date Y/N
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1 DISTRIBUTION: c= CONTROLLED i= INFO NAME DEP1 LOC C i c DCM-2A Preparation and Control of Manual ATTACHMENT 1 Calculations and Analyses
- Page 36 of 47 i
Rev. No. 3_ . 9910210310 991014 PDR ADOCK 07109268 B PM
CALCULATION CONTROL SHEET Pp 2 W 2 Q P81 ATED COMPONENTS N/A i 9 _RELATED STRUCTURES N/A RELATED ENGINEERING DOCUMENTS I
.FSAR, TS and DBD I
l RELATED PROCEDURES N/A' l YES O COMPUTER YES S SECURITY: PRINTOUT: No 0 No- @ i No. of Attachments: -1 No. of Calculation pages: 119 9 Preparation and Control of Manual ATTACHMENT 1 DCM-2A Page 37 of 47 Rev.'No. 3 Calculations and Analyses
- I i
F. l
, l '} r, NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 ~'
PAGE NO.2 / PROJECT: JAF NJIT PAGE NO. 2 . PREPARED BY:. #C DATE 9/JFA9 CHECKED BY: /g DATE 9/s v/ff
. TITLE: Power Uprate Program - ' Technical Support . Center 'Pont-Accident Radiological Habitability Study STATM Erf OF PROBLEM - This calculation was prepared by Corporate Radiological - Engineering (CRE) to update'the Technical' Support Center (TSC) radiological habitability analyses documented as Rev. O of this calculation (Refs. 3,'4, 5, 10 & 15). The reasons for the re- .
analysis are as follows: (a) Revision of the atmospheric dispersion factors for elevated releases as documented in Reference 2, (b) Assuming a lowered stand-by gas treatment system (SGTS) charcoal filter efficiency of 90% for halogens, (c) Crediting a pre-isolated TSC under spiked RCS concentration of 2 pCi/gm I-131 DE, (d). Use of the ICRP-30 (Ref. 12) dos's conversion factors for the calculation of thyroid doses, (e) Result of an 8-second unfiltered release through the reactor building exhaust system during a
-Refueling Accident (RA), .(f) Result of a brief release via reactor building exfiltration following an RA, (g) Results including the effect of uncertainty in filtered intake rates in the TSC, (h) The I-131 release fraction was increased from 10%
to 12% and the I-129 release frrction was decreased from 30% to 10% during a Refueling Accident (RA), (i) Results of changing the total number of in-core fuel rods from 36472 to 33600 for a Control Rod Drop Accident (CRDA) and a Refueling' Accident (RA) to ensure consistency with GE 8x8 fuel designs, and (j) Use of Figure 14.6-12 of FSAR for mass discharge rates from an MSLB.
NYPA - CALC.# JAF-CALC-RAD-00023
~~
REV 1 PAGE NO.: 1
. PROJECT: JAF NEXT PAGE NO.: s ePREPARED.BY: Nf(r DATE 9/tdAq CHECKED BY:- // DATE -f TITLE: Power Uprate. Progiam -; Technical- Shpport Center /ty/pf Post-Accident. Radiological Habitability. Study- ' Calculation Use Limitations This calculation was. developed to address .the specific issue (s) described in the above Statement'of Problem. Information provided in this . ~
calculation should not be used to support conclusions, recommendations, decisions or procedure development / revision unrelated to the above issue (s) .
~
For related issues, information' provided in this calculation should be
-used only by qualified staff, and only in conjunction with relevant references--(e.g.; related calculations, FSAR, Technical Specifications, Design Basis Documents, Licensing commitments, design drawings, etc.) , as
- appropriate.
If this calculation is used to change the plant's Design Basis, the
' Responsible Engineer should notify the Corporate Radiological Engineering- )
Group (WPO Nuclear Generation Department) to ensure.these changes
. accurately reflect the information provided in the calculation.
l I
+
1 l 1 l l I i i
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j NYPA -' CALC.# JAF-CALC-RAD-00023_'REV 1
' PAGE NO.: _J ' PROJECT: JAF .
3J 4.- PREPARCD BY: M(r - lDATE 9/AdM CHECKED BY: f/jBIT PAGE DATENO.:- pv/# a
' TITLE: Power Uprate Program ~ - - - Technical Support Center Pdst- I Accident-Radiological Habitability Study REFERENCES
- 1. CRE Calculation JAF-CALC-RAD-03031, Rev. 1,
- Offsite, j Control Room and TSC Domes Resulting from Unfiltered Releases I of Gap Activity.from the JAF Reactor Building during a Design )
Basis Refueling Accident" (9/19/98) I l
- 2. CRE Calculation JAF-CALC-RAD-00007, Rev. 2, " Power Uprate Program - Onsite and Offsite Post-Accident Atmospheric.
Dispersion Factors" (4/8/97) 3.' JAF ACTS Item #17875 - Perform habitability study for the TSC which accounts for TSC ventilation booster fan shaft leakage (prompted by NRC Information Notice 93-06, Ref. 37 below)
- 4. JAF ACTS Item #18268 - Evaluate and Resolve DDOI-JAF-HVAC-072-007 (TSC Dose Calculation)
- 5. JAF ACTS Item #18820 - Review ongoing radiological analyses
.for power uprate and revise proposed Tech Spec change as necessary. Also, review'these analyses to determine if NRC submittal on CR habitability (JPN-95-059) needs revision.
- 6. NUREG-0578, " TMI-2_ Lessons Learned Task Force Status Report and' Short-Term Recommendations" , (7/79)
- 7. NUREG-0696,
- Functional Criteria for Emergency Response Facilities"'(2/81) .
=8. NUREG-0737,
- Clarification of TMI Action Plan Requirements" (11/80)
- 9. JAF Emergency Plan Implementing Procedure EAP-14.1,
- Technical Support Center Activation" '(Rev. 16, 2/16/96)
- 10. .JAF Emergency Plan Implementing Procedure EAP-14.6,
" Habitability of Emergency Facilities" (Rev. 11, 5/6/96)
- 11. ~NYPA Memorandum JAG-93-245 addressed to J. Lazarus, from J.
Gray, titled " Control Rod Drop Accident (CRDA) Assumption" 1 (9/24/93) _ [See JAF-CALC-RAD-00026 for a copy of this ref.] 12 . - International Commission on Radiological Protection (ICRP) Publication 30,
- Limits for Intake by Workers" , 1982
- 13. CRE Computer Code DORITA-2,
- A Computer Code for the Determination of Radioactivity and Radiation Levels in Various Areas of a Nuclear Power Station and Offsite Following Accidental Releases of Gaseous Fission Products" ,
NYPA ' CAT.C.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: # PROJECT: JAF- ' NEET PAGE NO.: S. PREPARED'BY: M (v- DATE 9/M/9f CHECKED BY: // DATE .M r/// TITLE: Power Uprate Program - Technical Support Center ' Post-Accident Radiological Habitability Study RAD-001, Release 1.5.1.2 (1/22/97) l
- 14. GE letter addressed to Richard Chau, NYPA, from C. H. Stoll, GE Plant Performance Engineering, titled
- J. A. FITZPATRICK (JAF 'PP) Power Uprate Program - Formal Transmittal of Final Source Term Analysis Results" (5/2/91) [See JAF-CALC-RAD-00008 (Ref. 10) frr a copy of this 7:sference.]
- 15. J. DiNunno, F. Anderson, R. Baker and R. Waterfield,
- Calculation of Distance Factors for Power and Test Reactor Sites," AEC, Divisica of Licensing and Regulation, TID-14844 (March 1962)
- 16. US NRC Regulatory Guide 1.3,
- Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors" (Rev. 2, June 1974)
- 17. US NRC Regulatory Guide 1.5,
- Assumptions Used for ,
Evaluating the Potential Radiological Consequences of a Steam 1 Line Break Accident for Boiling Water Reactors" (March 1971)
- 18. US NRC 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" (3/23/72)
- 19. US NRC Regulatory Guide 1.49,
- Power Levels for Nuclear Power Plants" (Rev. 1, December 1973)
- 20. US NRC Regulatory Guide 1.77,
- Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors" (May 1974)
- 21. US NRC NUREG-0800,
- Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants"
- 22. General Electric Report NEDO-31400,
- Safety Evaluation for Eliminating the BWR Main Steam Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor" (May 1987) (See JAF-CALC-RAD-00013 for copy)
- 23. Specification for Nuclear Facility Purchase Order No. Apo-10 for Furnishing and Delivery of Condenser Vacuum Pump Equipment.
- 24. US NRC NUREG-0123, Rev. 3,
- Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)" (Fall 1980)
~
NYPA - CALC.# JAF-CALC-RAD-00023
~'
REV 1 PAGE NO.: 8 PROJECT: JAF - NEXT PAGE NO.: /, PREPARED BY: #2/ DATE st/M/99 ' CHECKED BY: // DATE 9 /,2 r/ f f TITLE: Power Uprate Program - Technical Support Center Post-Accident Radiological Habitability Study
- 25. NYPA Memorandum No. MHM-91-6, addressed to J. Lafferty, from M. Mozzor, titled " Charcoal Filter Efficiencies for Use in Accident Analyses Associated with JAF Power Uprate Program" (10/2/91)
- 26. K. G. Murphy and K. M. Campe,
- Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19 " ' 13th Air cleaning conference, pgs 401-430 (1974)
- 27. JAF-SE-96-071,
- Impact of Increased Isolation Time of Reactor Building Ventilation System on FSAR Analyzed Events" , Rev. 1 (10/2/98)
- 28. CRE Calculation JAF-CALC-RAD-00042, Rev. 2,
- Control Room Post-Accident Radiological Habitability - Assessment of Current Ventilation System Configuration" (10/97)
- 29. CRE Calculation JAF-CALC-RAD-00041, Rev. O,
- Radiological Assessment of a Control Rod Drop Accident Without MSIV Closure at pre-Uprate Conditions" (2/9/95)
- 30. GE Technical Report NEDE-31152P, " GE Fuel Bundle Designs,"
Rev. 6
- 31. CRE Calculation JAF-CALC-RAD-00013, " Radiological Justification for Modification of the Main Steam Line Radiation Monitor Trip Functions" (3/13/92)
- 32. US NRC Regulatory Guide 1.52,
- Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" (Rev. 2, March 1978)
- 33. JAF Operating Procedure OP-59B,
- Administration Building Ventilation and Cooling" (Rev. 8, 1/10/96)
- 34. Johnson Service Company, Test Report TLP-774-448 (02-4925-72),
- FitzPatrick NPP Damper Leakage (D-1300 Series)
(11/29/72) (NYPA Microfiche No. 60067" , frames 025-030) (See JAF-CALC-RAD-00028 for a copy of this reference.]
- 35. Design Basis Document DBD-072,
- Administration Building Heating Ventilation and Air Conditioning Systems" (Rev. O, 12/14/94)
- 36. NYPA Modification Control Manual MCM-6A,
- Structure and Component Classification and System Safety Function Control (JAF)" (Rev. 9, 12/13/95)
NYPA - CALC.# JAF-CALC-RAD-00023 PROJ1L'r JAF REV 1 PAGE NO.: [ NEXT PAGE NO.: 7, 4 PREPARED BY: M[ DATE 4/8 N CHECKED BY: /f/ DATE y/Sr/// { TITLE: Power Uprate Program - Technical -Support Center Pos't-Accident Radiological Habitability Study
- 37. GPU Nuclear Corporation letter 5450-95-0006, addressed to M. ]
Karasulu,'from N. G. Trikouros, titled
- FitzPatrick Nuclear
{ Plant Turbine Building EiELB Analysis Results" (2/17/95) 1 [ Sea JAF-CALC-RAD-00042 Ref. 28 for a copy of this reference.] q
- 38. NRC Information Notice 93-06, " Potential Bypass Leakage Paths Around Filters Installed in Ventilation Systems" !
(1/22/93)
- 39. CRE Computer Code QAD-CGGP,
- A Combinatorial Geometry l Version of QAD-P5A, A Point Kernal Code System for Neutron '
and Gamma-Ray Shielding Calculations Using the GP Buildup Factor," RAD-006, Release 1.5.1.2 (1/07/98)
- 40. CRE Calculation-Specific Computer Code MATILDA, documented int (a) CRE Calculation JAF-CALC-RAD-00003,
- Power Uprate Program - Reactor Duilding Post-LOCA EQ Radiation Levels Due to Buildup of Halogen Activity on Air Filtration Systems" (November 1991)
(b) CRE Calculation JAF-CALC-RAD-00015,
- Equipment Qualification Radiation Exposures Following A Control Rod Drop Accident" (7/28/92)
- 41. GE letter addressed to Richard Chau, NYPA, from C. H. Stoll, GE Plant Performance Engineering, titled
- J. A. FITZPATRICK (JAFNPP) Power Uprate Program - Transmittal of Nuclear Boiler Parameters and Fina". Reactor Heat Balance" (2/11/91) [See JAF-CALC-RAD-00008 for a copy of this reference.]
- 42. CRE Calculation JAF-CALC-RAD-00001,
- Impact of Ofgas High Flow Rates on FSAR Radiological Evaluation" (5/7/91)
- 43. CRE Computer Code ELISA,
- A Computer Code for the i Radiological Evaluation of Licensing and Severe Accidents at !
Light-Water Nuclear Power Stations," RAD-005, Release 1.5.1.2 (1/12/98) i
- 44. R. G. Jaeger, Ed., " Engineering Compendium on Radiation Shielding," Springer-Verlag, NY (1975)
- 45. Mine Safety Appliances Co. letter addressed to Stone &
Webster Engineering regarding the JAF Standby Technical Support Center Filtration system, NYPA P. O. #19786-80 (MSA Job No. B-508042, 2/26/81)
NYPA - CALC.9 JAF-CALC-RAD-00023 REV 1 PAGE NO.: 7
. PROJECT: JAF ~ NJXT PAGE NO.: W ",
PREPARED'BY: Mk DATE 9/AFM CHECKED BY: #f DATE eMr/// TITLE: Power Uprate Program - Technical Support Center ' Posit-Accident Radiological Habitability Study
- 46. Mine Safety Appliances Co., Filter Products Division, Technical Document titled
- Installation, operation and Maintenance Manual for the Standby Filtration System - Non-Seismic Category I - for the Power Authority of the State of New York" , NYPA P. O. #19786-80, MSA Job B-508042 (March 1981)
- 47. Calculation JAF-CALC-RBC-03042, Rev. 0,
- Reactor Building Exfiltration During Transition from Normal to Isolation Mode" (8/21/98) 48.
Memo from N. Kurul to G.S. Grochowski
- Mass Outflow Resulting from HELB at Full Power v.s. Hot Standby at JAF" RE-99-459, (8/13/199) l
- 49. Report No: JAF-RPT-PC-02342, Rev. 2,
- Primary Containment Leakage Rate Testing Program Plan" (1/11/99) 50.- NUREG/CR-5009,
- Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactor" , February 1988 51.- Sargent & Lundy,
- Fuel Handling Accident (FRA) Radioiodine Transport Removal and Dilution" , (5/29/98) 52.- USNRC,
- Safety Evaluation by the Office of Nuclear Reactor Regulation Related-to Amendment No. 239 to Facility Operating License No. DPR-59" , Docket No. 50-333. (12/6/96) l l
l l l
c= NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: [ PROJECT: JAF p PAGE NO. 9, PREPARED BY: 4(' DATE 4/A$li t CHECKED BY: g/ DATE , g/Jf/f/ TITLE:- Power. Uprate Program . Technical Support Center Poist-Accident Radiological Habitability Study LIST OF COMPUTER PROGRAMS EMPLOYED The_following Corporate Radiological' Engineering (CRE) computer ! codes _were,used in the1 analyses documented in this calculation: l Program Code Ref. Release Release Computer ) Name , Number Version Date System . l s DORITA-2 RAD-001 1.5.1.2 01/22/97 RS/6000 i QAD-CGGP RAD-006 1.5.1.2 01/07/98 RS/6000 I ELISA RAD-005 1.5.1.2 01/12/98 RS/6000 MATILDA ----- ----- 07/28/92 DG AViiON t (a) MATILDA was developed for use in conjunction with.QAD-CGGP (Ref. 39) and the gamma spectra produced by DORITA-2 (Ref. 13), ELISA (Ref. 43) and ALLEGRA (another CRE code) for the computation of radiation exposures. It is documented in Refs. 40 (a) and 40 (b) . e
, NYPA'- CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 9 PROJECT: JAF '~ PAGE NO.: /o PREPARED BY 46" DATE 9/s.4M CHECKED BY: DATE eVfp/e l TITLE:- Power Uprate Program ; - ' Technical Support Center "Podti p Accident Radiological Habitability Study TABLE OF CONTENTS Page
- STATEMENT OF PROBLEM ...................................... 1 REFERENCES ................................................ 3
' LIST OF COMPUTER PROGRAMS EMPLOYED ........................ 8 W
TABLE OF CONTENTS ........................'..............'... 9 1,. INTRODUCTION ......................................... , 11
- 2.
SUMMARY
.OF RESULTS ................................... 13 2.1 Inumersion Exposures ............................. 14 2.2 Direct Shine Radiation-Fields ................... 14 2.2.1 Shine from Post-LOCA Radioactivity Accumulating on the Refueling Level ...... 15 2.2.2 Shine from Post-LOCA Overhead Clouds ..... 16 2.2.3 shine from TSC Charcoal Filters .......... 17 2.3 Conclusions ..................................... 17 4 i
- 3. METHODS.OF ANALYSIS _.................................. 42 ;
i
- 4. RADIATION EXPOSURES FROM A LOSS OF COOLANT ACCIDENT .. 45 i 4.1 Drywell' Leakage ................................. 46 )
4.1.1 Basic Data and Assumptions ............... 46 4.1.2 Results .................................. 52 4.2 ESF Conqronent Leakagu . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 4.2.1 Basic Data and Assumptions ............... 54 4.2.2 Results .................................. 56 4 . '3 Total LOCA Do'se ................................. 58 l 1
l l 1 NYPA - CALC.#'JAF-CALC-RAD-00023' REV'l PAGE NO.: /d PROJECT: JAF ' NEXT PAGE NO. // A/f/r DATE 9/JF/IM CHECKED BY: // 9//pArf
- PREPARED BY: DATE TITLE: Power Oprate Program' - Technical Support Center Pont-Accident Radiological Habitability Study TABLE OF CONTENTS (Cont.)
Page 1 5.- RADIATION' EXPOSURES FROM A MAIN STEAM LINE BREAK . . . . . - 60 5.1 -Basic Data and Assumptions ...................... 60 5.2 Results.......................................... 66'
- 6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT . 70 6.1' Basic Data and Assumptions ...................... 70 6.2 .Results ......................................... 74
. 7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT ........ 76 7.1 Basic Data and Assumptions ...................... 76 7.2 Results ......................................... 81 .8. RADIATION EXPOSURES FROM EXTERNAL SOURCES ........... 84 )
8.1 Direct' Shine from Post-LOCA Airborne Radioactivity in the' Reactor Building .......... 85 8.1.1 Basic Data and Assumptions .............. 85 8.1.2 Results ................................. 88
'8.2- Direct-Shine'from Post-LOCA Overhead Clouds .... 98 8.2.1 Basic Data'and Assumptions .............. 98 8.2.2 Results ................................. 99 8.3 Direct Shine from Halogens Accumulating on the TSC Charcoal Filters ........................... 115 8.3.1 Basic Data and Assumptions .............. 116 8.3.2 Rasults ................................. 119 I
l l l ATTACHMENTS A. Copies of Computer Outputs j 1 l
)
Q , l' NYPA '. - - CiALC . # JAF- CALC-RAD- 00023 REV 1 PAGE NO.: //
~ . PROJECT: JAF /1_ , PREPARED BY:- M,k DATE 4h,V/M QXT PAGE NO.:
CHECKED BY: // DATE w /cc
' TITLE:- Power Uprate -Program -
Technical Support Center @6fsti-Accident Radiological Habitability Study I 1.' ' INTRODUCTION i The requirements for a Technical Support Center were established I l under the task. force'on' lessons learned from TMI-2 (Ref. 6, Sec. 2.2.2). The habitability criteria were set forth in Reference 7 L (Sec. 2.6) ,- as follows : I
- Since the.TSC is to provide direct management and technical j support to the control room during an accident, it shall have the same radiological habitability as the control room under accident f conditions. TSC personnel'shall be. protected from radiological l
' hazards, including direct radiation and airborne radioactivity j from inplant' sources under accident conditions, to the same degree as control room personnel. Applicable criteria are specified in j j (10 CFR 50, Appendix A) General Design Criterion (GDC) 19, in the l Standard Review Plan (Ref. 19, Sec. 6.4), and in NUREG-0737 (Ref.
8, Item II.B.2). ..... If the TSC becomes uninhabitable, the TSC ! ! plant-management function shall be transferred.to the control l' r o o m .*' This-calcul'ation analyzes the post-accident radiological j habitability in the TSC. Results of the present study are' summarized in Sec. 2, and the methods of solution are presented in Sec. 3. Sections 4 through 7 present the data, assumptions and antalytical details associated with each of the four design-basis accidents. Section 8 looks at l the radiation fields-from external sources, namely, gamma i f radiation emanating-from post-LOCA airborne radioactivity accumulating in the refueling level of the reactor building, from overhead clouds, and from halogens accumulating on the TSC charcoal filtration' system. Copies of the computer outputs appear in Attachment A'. {
.NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 11 PROJECT: JAF- ~ NEXT PAGE NO.: gA PREPARED BY: A (r . DATE ~ 81/vMtf CHECKED . BY /// DATE v>f/91 , ' TITLE: Power Uprate Program
- Technical support Center "Po6b l -Accident Radiological Habitability Study Revision 1 - Remarks Revision 1.of this calculation was undertaken to address the TSC -habitability for the purpose of including recent revisions, as -follows:
- Use of post-accident atmospheric dispersion factors for elevated releases as documented in Reference 2.
- Assuming a lowered stand-by gas treatment system (SGTS) charcoal filter efficiency of 90% for halogens.
- Crediting a pre-isolated TSC under spiked RCS concentration of 2 Ci/gm I-131 DE.
- Use of ICRP-30 (Ref. 12) dose conversion factors for the calculation of thyroid doses.
- Assuming an 8-second unfiltered releases through the reactor building exhaust system during a refueling accident.
- LResult of a brief release via. reactor building exfiltration following an RA.
- Assuming a filtered air intake rate in the isolated mode of
^
3000 scfm 10% uncertainty and unfiltered air intake rate of 1500 scfm to account for vent. Measurement Testing l uncertainty.
- The I-131' release fraction was increased from 10% to 12% per NUREG/CR-5009 and the'I-129 release fraction was decreased
'from 30% to 10% per Reg. Guide 1.25 during a Refueling Accident (RA). j
- Results of changing the total number of in-core fuel rods from 36472 to 33600 for a control Rod Drop Accident (CRDA)
]
and a Refueling Accident (CR) to ensure consistency with GE 8x8 fuel designs.
- Use Figure 14.6-12 of FSAR for mass discharge rates from an l MSLB.
NYPA CALC.#'JAF-CALC-RAD-00023 REV 1 PAGE NO.: W PROJECTS'JAF> .. NpKT PAGE NO.:. i4. I PREPARED.BY: ## k DATE _4f g f CHECKED BY: -X[ DATE 7/,yr/r9 TITLE: Power Uprate Program - Technical Support Center ' Post-Accident Radiological Habitability Study
- 2. StRRERY OF RESULTS l The TSC ie. located in the Administration Building. The pressure
-boundary.of 4.ts ventilation system extends from El. 272' to El. .300'-(see Figs.-2.1 - 2.3). The total floor area and volume ]
within the pressur O boundary are approximately.12,500 f t* and ): l'.5E+05 ft*, respectively. The TSC proper is on El. 286', the primary location being in the Lunch Assembly area. The area w'ithin the TSC pressure' boundary on El. 300' (approximately 2500 l f t') -is designated as the Alternate Operational Support Center. i
.The TSC HVAC system is designed to supply 3000 scfm filtered air, plus or minus 10% uncertainty, to maintain (1/8)" water gauge (WG) pressure within the TSC pressure boundary. A simplified flow
- diagram is shown in Fig. 2.4. Since the system is non-safety related, there are no redundant components, and single-failure ;
criteria are not applicable. ! When operation of the TSC is required, the Administration Building
-Office Area normal HVAC system operation is modified to serve only j the area within the.TSC pressure boundary. This is accomplished through Operating Procedure OP-59B (Ref. 33) by manually turning off exhaust fans 72FN-20 and 72FN-21 (which exhaust air from the ,
conference room and from the print area), and turning on booster fan 72FN-47 (for the charcoal filtration system). Concurrent with the startup of.72FN-47,.outside air damper 72 MOD-106 opens (to supply air to the charcoal filter train), isolation dampers 72 MOD-107,'108 and 109 for various office areas close, and exhaust air dampers 172AOD-146 and 72AOD-71 close. The subsections which follow address the radiation fields due to immersion ~in' airborne radioactivity and direct shine fror external sources.
(: j i NYPA -' CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: I al-PROJECT: JAF NEXT PAGE NO.: i _iq;
-PREPARED ~BY: 4M .DATE 4/M/M ?
- CHECKED BY: 8 DATE TITLE:' , Power Uprate Program. --Technical Support Center o t-Accident-Radiological Habitability Study 2 .1. Iammersion Esoosures..
Thefollowing design-basis accidents.were considered in the re- ] 8
. assessment 1of'the JAF TSC radiological, habitability- I 4 - (a) ~ Loss of coolant' accident (LOCA) (drywell leakage and ESF l component. leakage pathways),
f (b) Main Steam Line Break outside containment (MSLB), (c) , Control Rod Drop Accident (CRDA), and j 4 (d) Refueling Accident (RA). The. basic' data and assumptions in each of the above accident se'enarios are consistent with the current licensing basis and the models'in the regulatory \ guides-(Refs. 16 - 20, and 32) and the , i
~ Standard Review Plan (SRP, Ref. 21). Complete details for each I accident!are presented in Secs. 4 through 7. A summary of the . principal assumptions for each.DBA appear in Table 2.2'.
For each accident, the.TSC.HVAC system was assumed to be operating-
,without HVAC system failures, since the TSC HVAC system is non-safety'related. Details of the ventilation / filtration flows as a 'funetion of time:are shown in Table 2.1.
The immersion exposures associated with each accident are presented in Tables 2.2'and 2.3. Since the receptor was assumed to beLexposed to a radioactive cloud having the same volume as the entire area within;the TSC pressure boundary, the whole body and
'bkin exposures in these tables conservatively apply to all locations within'the pressure boundary. .'2.2 . Direct Shine Radiation Fields Post-accident external-sources of radiation affecting the habitability of.the JAF TSC include the following: - (a) Airborne radioactivity accumulating on the refueling level of the reactor building (RB),
NYPA - CALC.# JAF-CALC-RAD-00023. REV 1 PAGE NO.: /S PROJECT: JAF ~ NEXT PAGE NO.: /g 4 DATE 9/st4ff CHECKED BY: 8 PREPARED BY: TITLE: N/ Power Uprate Program - DATE 7/M/ff Technical support Center 'Pont-Accident Radiological Habitability Study (b) Overhead radioactive clouds, and (c). Halogens accumulating on the TSC charcoal filters. The analyses carried out in the present calculation were limited to the worst-case accident scenarios, namely a LOCA for sources (a) and (b) above, and a MSLB for source (c). Results are presented in the subsections which follow. For the sake of completeness, radiation fields were determined for 15 receptors within the TSC pressure boundary, 11'on El. 286' and 4 on El. 300', as shown in Figs. 2.5 and 2.6. Note that some of the receptors are heavily shielded from radiation emanating from the RB refueling level, while others have no shielding. The effective concrete shield-slab thicknesses protecting the various receptors are as follows: Receptors 1 - 5 (Fig. 2.5) (El. 286' - General Offices) : 0" Receptors 6 - 8 (Fig. 2.5) (El. 286' - Gen. Of f. & part of Lunch area) : 6" Receptors 9 - 11 (Fig. 2.5) (El. 286' - Most of Lunch area & Elevator) : 36" Receptors 12 - 15 (Fig. 2.6) (El. 300' - Alternate TSC) : 30" 2.2.1 shine freet Post-LOCA Radioactivity Accumulating on the Refueling Level Direct-shine radiation doses and cumulative doses to receptors within the TSC pressure boundary due to gamma radiation emanating from Post-LOCA airborne radioactivity within the refueling level
.of the reactor building are shown in Tables 2.3 and 2.4. An overall summary of the results is as follows:
I l
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.s J/ PROJECT: JAF NEXT PAGE NO.: Iy. PREPARED BY: M/r DATE 4/t4M9 CHECKED BY: 8 DATE GhpM TITLES ~ Power Uprate Program -
~ Technical Support Center '@ost-Accident Radiological Habitability Study Peak WB Dose 31-day WB Dose Recentor(s) Rate (mrad /hr) ** Cont. Occ. (rad) 1- 5 1900 - 4800 110 - 350 6- 9 240 -
610 7 - 22 10 - 11 <0.08 <0.001 12 - 14 <0.32 <0.0035 15 29 1.9
- At 8 hrs after the postuinted LOCA for Receptors 1-5 and 15, and at 2 hrs for Receptors 6-14.
It is clear that a limited area within the TSC pressure boundary Eis radiologically habitable under worst-case conditions, namely, only that area which is well _ shielded by the 30" concrete roof. Even receptors close to the edge of the 30" roof (such as Receptor
#9) can be exposed to significant radiation fields. Taking credit
.for partial occupancy does not alter the-above conclusions (e.g., receptor 6 would still exceed 5 rads). 2.2.2 Shine frostPost-LOCA Overhead Clouds Cloud-shine radiation doses and cumulative doses to receptors ,within the Tsc pressure boundary due to gamma radiation amanating from Post-LOCA radioactive clouds are shown in Tables 2.5 and 2.6. An overall' ausanary of the results is as follows: Peak WB Dose 31-day WB Dose Recentor(s) Rate (mrad /hr) ** Cont. Occ. (rad) 1- 5 45 - 75 0.17 - 0.29 6- 9 5 - 25 0.012 - 0.075 10 --11 0.12 - 0.83 <0.002 12 - 15 0.15 - 1.5 <0.005
- At t = 0 hrs after the postulated LOCA
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1- .PAGE NO.: I7
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~ TITLES. . Power. Uprate Program - Technical Support Center Post-Accident Radiologic'al Habitability Study l These radiation fields and doses are small and negligible in . comparison to those due to shine from the refueling level.
l l 2.2'.3 shine from Tsc charcoal Filters j l Worst-case direct shine radiation fields from halogens accumulating on the TSC charcoal filter are shown in Tables 2.7 and 2.8. The worst-case design-basis accident for this source is a p . MiBLB, and the' receptor locations are as follows: (a) Worst-case. receptor on El. 286' of the Administration Building-(between Receptors #4 and #6 in Fig. 2.5), and (b) At the filter casing (El.-300'). The peak dose rates are 4.7'arad/hr at'the worst-case receptor on 1 El. 286', and 1.4 rad /hr in contact with the filter casing. The
-corresponding worst-case 31-day, dose within the TSC is 0.044 rad I for continuous occupancy, which is insignificant.
i . !- 2.3.. Conclusions
.As m' result'of. shine from airborne radioactivity accumulating of the refueling level of the reactor building, only a part of the . TSC area on El. 286' (about 1100 f t 8 ) may remain habitable following a design. basis LOCA. To meet the NM requirement regarding the. minimum size of TSCs (1875 f t' for 25 persons with '75 'f t'/ person) , it would be necessary to make use of part of the
- j. area on El. 300' within the TSC pressure boundary, namely, part of l what is currently designated as the Alternate TSC.
In! summary, a TSC area which meets the GDC-19 exposure guidelines and the NRC requirements for size, corresponds to the areas in Figs. 2.5 and 2.6 represented by Receptors 10 - 14. The area around Receptor #15 (on El. 300') is not suitable since the dose I. l
l 'NYPA' . CALC.# JAF-CALC-RAD-00023
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DATE 9/a4/99 CHECKED BY: 8 Power. Uprate Program - DATE g/,tf/ff Technical Support Center Podit-Accident Radiological Habitability Study rate frca post-LOCA RB shine could be as high as 30 mrad /hr*. The thyroid, whole body and skin doses resulting from the four design-basis accidents appear in Table 2.2. As is seen, the results for'the RA are below the'GDC-19 limits for the pre-isolated and unisolated modes. The LOCA'and CRDA doses are also below GDC-19. limits.
.por the MSLB accident, doses from operating at the TS limit for I-131 DE are within GDC-19 limits. However, for the case of iodine spiking to 2 Ci/gm I-131 DE, doses remain below the limits for a pre-isolated TSC only if the TSC is not occupied for the first hour. The earlier the TSC is occupied, the higher the dose, which reaches a maximum of over 50 rem thyroid for occupancy at t=0 (see Table 5.2).
Note: The current TSC radiation monitor alarm setpoint is 5 (mrad /hr) (Ref. 8).
NYPA - CALC.9 JAF-CALC-RAD-00023 REV 1 FAGE NO.: \9 PROJECT: JAF ' UEXT PAGE NO.: PREPARED BY: 4 & DATE 9/a4/# CHECKED BY: Af DATE . a TITLE: Power Uprate Program - . Technical Support Center [F6st-Accident Radiological ~ Habitability Study Table 2.1
.TSC Ventilation Systest operating Conditions and Flows As a Function of Post-Accident TiEhe-Description Value Without HVAC System Failures Pre-activation air intake rate (scfm) 25800"'
A'tivation c time (min) 60/12(*) Post-activation filtered air intake rate (scfm) 3000"' 10% Post-activation unfiltered air intake rate (scfm) 1500
'(a) Maximum flow that can be provided by the 72-AHU-4 unit (b) Manual activation in 60 minutes for a'LOCA and a CRDA,.and the worst-case time for an MSLB and an RA (namely, 12 min, f rom Sec . 5.1 (j ) ] .
(c) 90% charcoal filtration efficiency for all halogen species (d) Arbitrarily selected to accommodate the followings (1) Leakage through AOD-171, which could be as high as 800 scfm if the damper leakage is at the maximum value provided in Ref. 34-[i.e., 25 (scfm/f t*) ] (2) Potential ventilation booster fan shaft leakage, to cover the concerns of NRC Information Notice 93-06 (Ref.
- 38) and (3) 10 scfm inleakage due to ingress / egress (from Ref. 21, Sec.-6.4) 4
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 1 p_ PROJECT: JAF NJXT PAGE NO.: 2/ PREPARED BY: Adr- DATE 9/24//4f CHECKFD BY: /Gf DATE g/fr/99 TITLE: Power Uprate Program - Technical Support Center Pois ti-
' Accident Radiological Habitability Study Table 2.2 C
JAF TSC Personnel Radiation Doses D u e t o I n s a e r s t. o n i n A i r b o r n e R a d i o a c t i v i t y [3300 scfm filtered (3000+10% uncertainty)] Design-Basis Thyroid Wh. Body Skin Accident J rea) (ran) _ (resa) LOCA (Drywell Leak) 9.840E+00 1.345E-02 1.433E-01 LOCA (ESF Comp. Leak) 1.070E+00 2 . 953 E '14 2.543E-03 LOCA (Total) 1.091E+01 1.375E-02 1.458E-01 Main Steam Line Break i (0.2 pCi/gm I-131 DE)* 1.993E+01 2.104E-02 1.192E-01 (2 Ci/gm I-131 DE) ** 5.328E+01 6.332E-02 3.686E-01 (2 pCi/gm I-131 DE) *** 1.657E+01 2.419E-02 1.496E-01 Control Rod Drop 1.647E+01 6.995E-02 6.766E.01 Refueling Accident (pre-isolated) 8.290E+00 1.890E-02 2.652E-01 (12 min. isolated) 3.435E+01 3.101E-02 4.298E-01 (Not-isolated) 2.181E+01 2.006E-02 2.762E-01 GDC-19 limits 30 5 30 12 min. isolated (worse case) pre-isolated
*** pre-isolated, but not occupied until 1 hour after accident onset.
Note: Continuous occupancy in all cases. Thyroid doses based on the ICRP-30 dose conversion factors. I
NYPA - CALC.# JAF-CALC-R'U)-00023 REV 1 PAGE NO.: ll PROJECT: JAF. ' NJXT PAGE NO.: 1 2. PREPARED BY: x( DATE UN/qi CHECFED BY /G' DATE TITLE: Power Uprate Program - Technical Support 7/fr/5f Center 'Po'st-Accident Radiological Habitability Study Table 2.2 (Continued) JAF TSC Personnel Radiation Doses Due to Insterston in Airborne Radioactivity 72700 scfm filtered (3000-10% uncertainty)] Design-Basis Thyroid Wh. Body Skin Accident frem) (rem) (resa) LOCA (Drywell Leak) 1.083E+01 1.337E-02 1.430E-01 LOCA (ESF Comp. Leak) 1.180E+00 2.889E-04 2.530E-03 LOCA (Total) 1.201E+01 1.366E-02 1.455E-01 Main Steam Line Break (0.2 pCi/gm I-131 DE)* 2.206E+01 2.265E-02 1.284E-01 (2 Ci/gm I-131 DE) ** 5.875E+01 6.572E-02 3.827E-01 (2 pCi/gm I-131 DE) *** 2.066E+01 2.543E-02 1.577E-01 Control Rod Drop 1.804E+01 6.965E-02 6.762E-01 ! l Refueling Accident l (pre-isolated) 9.158E+00 1.884E-02 2.645E-01 (12 min. isolated) 3.655E+01 3.289E-02 4.563E-01 (Not-isolated) 2.181E+01 2.006E-02 2.762E-01 GDC-19 limits 30 - 5 30 I 11 min. isolated (worcm case) pre-isolated
***- pre-isolated, but not occupied until 1 hour after accident onset.
Note: Continuous occupancy in all cases. Thyroid doses based on the ICRP-30 dose conversion factors.
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 2.2. l PROJECT: JAF- ~ NJXT PAGE NO.: 2a I PREPARED BY: N/, - DATE 3 4t//49 CHECKED BY: Jt'f TITLE: -Power Uprate Program - DATE Technical Support Center
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P6st-Accident Radiological Habitability Study Table 2.2 (Continued) BASES i LOCA (Drywell Leakage) (a) A LOCA takes place at full power (2535.8 MWt + 2% uncertainty).- (b) All the noble gases and 25% of the halogens remain airborne within the drywell at the time of the accident and are available for release. (c) Leakage from the drywell is at the rate of 1.5%'per day. ' (d) All the noble gases and halogens leaking from the drywell are exhausted to the atmosphers via the Standby Gas Treatment System (SGTS) and the main stack without holdup or mixing in the reactor building.- j
~
(e) The SGTS filter efficiency is 90% for the removal of all halogen species. LOCA (ESF Leakane) (a) 50% of the total halogen activity present in the core mixes uniformly with the coolant in the RER system (113,400 cu f t) . -(b) The ESF leakage rate.is 5 gym, and is constant from the start of the LOCA through the duration of'the accident. -(c) An additional 30-minute leakage of 50 gym (due to gross failure of a passive component) is conservatively assumed to
-begin'at the time of the accident.
(d) 10% of the halogens in the leaking fluids bacomes airborne and mixes uniformly with the reactor building atmosphere. (e) Release from the reactor building is through the SGTS and the main stack at the rate of 6000 scfm.
F
=NYPA -' CALC.#'JAF-CALC-RAD-00023 REV 1' PAGE NO. 23 PROJECT: JAF NEXT PAGE NO.: res PREPARED BY: pg k DATE -9/a,Mff CHECKED BY: /[ DATE ch p/f' TITLE: Power Uprate Program .- Technical Support Center @otti
,- Accident Radiological Habitability Study Table 2.2 (continued) Main Steam Line Break (a) A line break occurs in the main steam lines during full power operation. [ Note-that there is a potential for a higher
' liquid release of 0%' power, but this is bounded-by the value given in . (c) below.)
(b)' The MSIVs close in 10.5 seconds after the break.
-(c) The total discharge through the break prior to isolation is i equal to 20,000 lb of steam and 120,000 lb of liquid.
(d) The ensuing high fuel temperatures do not lead to any fuel damage. ! (e) The noble gas fission product concentrations in the steam correspond to the design values which would yield the , standard release rate to the~ atmosphere during normal ' operation (i.e., 100,000 pCi/sec folleving a 30-minute decay). One.hundred percent of all noble gases leaving the resactor vessel during the 10.5-see MSIV closure time (via all four steam lines) are released through the break. The l halogen source term in the discharged liquid was selected to represent the limit for the maximum permissible reactor coolant system (RCS) - activity under power uprate conditions, namely 0.2' Ci/gm I-131 Dose Equivalent. In addition, the halogen source ~ term was selected under the spiked RCS concentration of 2..pC1/gm I-131 DE when the TSC is pre-isolated. (f)-'100% of the radioactivity discharged into the turbine building becomes airborne and la released to the atmosphere at ground level over a period of 2 hours. The release rate
-was. selected to be equivalent to 3 air changes per hour.
(g) Activation of the TSC. ventilation system takes place at the post-accident time (less than 1 hr) which maximizes the TSC radiation exposures; this time was determined to be 12 minutes.
i NYPA CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 2,4-PROJECT: JAF NEXT PAGE NO. 2A PREPARED BY: M(y DATE 81/ p/ A f C H E C K E D B Y : 8 DATE . qbr/Sy TITLE: Power Uprate Program - Technical S'upport Center #Pi5s'ti-Accident Radiological Habitability Study Table 2.2 (Continued) Control Rod Dron Accident (a) The reactor has been operating at full power for an extended period of time. It is shut down, taken critical, and brought back to the initial temperature and pressure conditions within 30 minutes of the departure from design power. i (b) A CRDA takes place leading to the failure of 850 fuel rods at a core location with a radial power peaking factor of 1.5. (c) All activity within the gaps of the failed fuel rods is released to the reactor coolant and is instantaneously and uniformly mixed with the coolant in the pressure vasnel at the time of the accident. The released activity corresponds to 10% of all halogens and 10% of all noble gases (mccept Kr
- 85) in each failed rod, and to 30% of the Kr 85 inventory.
(d) 10% of the iodines and 100% of the noble gases released in the pressure vessel reach the turbine and. condensers. (e) As a result of elimination of the MSIV-closure and reactor-shutdown functions of the main steam line radiation monitors, the pathway of post-CRDA atmospheric releases at JAF has changed. Under the new CRDA scenario, the MSIVs stay open and the release is to the offgas system. (f) As a result of plant shutdown following a CRDA, or as a result of offgas system automatic isolation (following a 15-minute delay, which is not considered in the analysis) due to high radiation fields at the offgas monitors, the released radioactivity is retained within the turbine, condensers and the offgas system. Release to the environs is due to leakage from the various contaminated systems into the turbine building. (g) .90% of the iodines plate out on system internal surfaces. (h) The leakage rate from contaminated systems into the tu.;4.no building amounts to 1% per day and lasts for 24 hours. The release to the atmosphere is at ground level and there is no holdup within the turbine building.
> n 'NYPA - CALC.# JAF-CALC-RAE ^7023 ~
REV 1 PAGE NO.: 2.5I PROJECT: JAF tT PAGE NO.: 24 PREPARED BY: 4,(r- DATE ' 4/2.Mff CHECKED BY: Np/ X DATE .oh/6 TITLE: Power Uprate Program - Technical Support Center *P6st'- Accident Radiological Habitability Study Table 2'.2 (continued) Refuelina Accident (a) .The reactor has been operating at full power for an extended period of time. (b) The reactor is shutdown, refueling operations are initiated and an RA takes place at 24 hours after shutdown. (c) 'The accident involves the-dropping of a fuel assembly and the ensuing rupture of-125 fuel rods (a conservative estimate). (d) The failed fuel rods were at a core location with a radial power peaking factor of 1.5. (e) _All activity in the gaps of the failed fuel rods is released to the fuel pool water. The released activity is conservatively assumed to correspond to 10% of all halogens (except I 131) and 10%'of all noble gases (except Kr 85) in each failed rod; 30% of the Kr 85 (Reg. Guide 1.25) and 12% of the I-131 (NUREG/CR-5009). (f) The halogen composition ' (inorganic, organic and particulate species). and the pool halogen retention factors are such that 99% of all released halogens are assumed to be retained by the water in the fuel pool. The retention of noble gases by the pool water is negligible. (g). The exhaust duct is isolated prior to significant unfiltered release with the radioactive material being routed through the SGTS. An isolation delay time of 18 seconds occurs between the time that a Refueling Accident (RA) is detected by the radiation monitoring system and releases from the reactor building exhaust to the SGTS and the stack. The reactor building exhaust system ducting provides a 10-second delay. The difference between the system response time of 18 seconds and the 10-second delay implies that an unfiltered release of 8 seconds in duration could result. In addition, a potential air exfiltration release from the reactor building has been found to occur during the transition from -l the normal mode of operation to the isolation mode. l l (h) The_ Reactor Building air exchange rate was set at the 1 conservative value of 72 air changes per day after isolation. ! l (i) The halogen-removal filter efficiency of the SGTS is 90% for l all halogen species. J l 1
)
NYPA - CALC.# JAF-CALC-RAD-00023 .REV 1 PAGE NO.: 2/ PROJECT: JAF NEXT PAGE NO.: 2 -f PREPARED BY: M/,- DATE 4/M//d CHECKED BY: 8 DATE . /ar/yr TITLE: Power Uprate Progran - Technical Support Center Post-Accident Radiological Habitability Study Table 2.3 _ Direct-shine Dose Rates (rad /hr) to TSC Receptors freutpost-LOCA Airborne Radioactivity Accumulating in j the Refueling Level of the Reactor Building l 1 Time
.ihal. Rec. #1 Rec. #2 Rec. #3 Rec. #4 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 O.5 8.700E 1.893E+00 1.946E+00 9.615E-01 1.0 1.326FA00 2.905E+00 2.986E+00 1.471E+00 2.0 1.821E+00 4.033E+00 4.145E+00 2.033E+00 8.0 1.907E+00 4.613E+00 4.736E+00 2.233E+00 24.0 9.817E-01 2.793E+00 2.861E+00 1.265E+00 48.0 4.729E-01 1.489E+00 1.520E+00 6.583E-01 96.C 2.345E-01 8.127E-01 8.265E-01 3.536E-01 168.0 1.466E-01 5.181E-01_ 5.268E-01 2.242E-01 240.0 1.030E-01 3.579E-01 3.642E-01 1.550E-01 744.0 1.060E-02 3.262E-02 3.337E-02 1.428E-02 i
? Time _(hg). Rec. #5 Rec. #6 Rec. #7 Rec. G8 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 9.513E-01 2.991E-01 1.172E-01 1.450E-01 1.0 1.455E+00 4.518E-01 1.763E-01 2.188E-01 2.0 2.011E+00 6.132E-01 2.388E-01 2.967E-01 8.0 2.211E+00 5.624E-01 2.097E-01 2.686E-01 24.0 1.254E+00- 2.081E-01 6.717E-02 9.538E-02 l 48.0 6.514E-01 8.268E-02 2.436E-02 3.698E-0.2 ! 96.0 3.494E-01 _3.418E-02 9.457E-03 1.501E-02 l l
.168.0 '2.216E-01 1.993E-02 5.331E-03 8.660E-03 ; -240.0 1.533E-01 1.425E-02 3.792E-03 6.181E-03 i 744.0 1.416E-02 1.694E-03 4.517E-04 7.359E-04 I
I i i i
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- Posit-Accident Radiological Habitability Study Table 2.3 (Continued)
Direct-Shine Dose Rates (rad /hr) to TSC Rece;. tors from post-LOCA Airborne Radioactivity Accumulating in the Refueling Level of the Reactor Building Time
,{hg1 Rec. G9 Rec. # 1,Q Rec. # 11 Rec. 912 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 n.5 1.767E-01 3.931E-05 2.960E-06 1.582E-04 1.0 2.670E-01 5.753E-05 4.273E-06 2.310E-04 2.0 3.625E-01 7.589E-05 5.435E-06 3.032E-04 8.0 3.346E-01 4.994E-05 3.063E-06 1.927E-04 24.0 1.261E-01 5.666E-06 2.296E-07 1.927E-05 48.0 5.058E-02 6.692E-07 1.641E-08 1.903E-06 96.0 2.103E-02 8.565E-08 7.908E-10 1.775E-07 168.0 1.229E-02 2.367E-08 9.925E-11 3.674E-08 240.0 8.787E-03 1.420E-08 3.660E-11 1.946E-08 744.0 1.045E-03 1.656E-09 3.799E-12 2.214E-09 i Time ,(h31 Rec. #13 Rec. 914 Rec. #15 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 1.634E-04 3.691E-05 1.297E-02 1.0 2.386E-04 5.369E-05 1.981E-02 2.0 3.132E-04 6.971E-05 2.730E-02 8.0 1.993E-04 4.228E-05 2.933E-02 24.0 2.000E-05 3.747E-06 1.593E-02 48.0 1.982E-06 3.231E-07 7.995E-03 96.0 1.858E-07 2.356E-08 4.145E-03 168.0 3.859E-08 4.004E-09 2.616E-03 240.0 2.047E-08 1.889E-09 1.823E-03 744.0 2.330E-09 2.096E-10 1.768E-04 -
i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 2.Y PROJECT: JAF p PAGE NO. sA PREPARED BY: M(, - DATE 4/2,V/fi CEECKED BY: X[ DATE TITLE: Power Uprate Program - 7/a/Sf Technical Support Center /P6dti-Accident Radiological Habitability Study 1 Table 2.4 Direct-Shine Doses (rad) to TSC Receptors fresa. post-LOCA Airborne Radioactivity Accumulating in the Refueling Level of the Reactor Building l Time' M Rec. #1 Rec. #2 Rec. #3 Rec. #4 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 O.5 2.175E-01 4.732E-01 4.865E-01 2.404E-01 1.0 7.586E-01 1.655E+00 1.701E+00 8.395E-01 1 2.0 2.319E+00 5.093E+00 5.235E+00 2.576E+00 8.0 l'.350E+01 3.099E+01 3.184E+01 1.536E+01 24.0 3.580E+01 8.903E+01 9.136E+01 4.261E+01 48.0 5.252E+01t 1.388E+02 1.423E+02 6.491E+01 96.0 6.883E+01 1.924E+02 1.969E+02 8.844E+01 168.0 8.231E+01 2.395E+02 2.448E+02 1.089E+02 l 240.0 9.120E+01 2.707E+02 2.765E+02 1.224E+02 744.0 1.117E+02 3.392E+02 3.463E+02 1.521E+02 Time Mut). Rec. G5 Rec. G6 Rec. #7 Rec. G8 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 2.378E-01 7.478E-02 2.930E-02 3.624E-02 1.0 8.305E-01 2.599E-01 1.017E-01 1.259E-01 2.0 2.549E+00 7.883E-01 3.077E-01 3.817E-01 8.0 1.521E+01 4.313E+00 1.651E+00 2.076E+00 24.0 4.220E+01 1.002E+01 3.654E+00 4.753E+00 48.0 6.428E+01 1.32EE+01 4.668E+00 6.232E+00 96.0 8.755E+01 1.591E+01 5.424E+00 7.402E+00 168.0 1.078E+02 1.781E+01 5.942E+00 8.233E+00 240.0 1.211E+02 1.903E+01 6.267E+00 8.762E+00
.744.0 1.505E+02 2.200E+01 7.058E+00 1.005E+01
1 l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 2.9 l l- PROJECT: JAF NEXT PAGE NO. so PREPARED BY: et [y- DATE 9/pl[ih CHECKED BY: /// DATE g/sb l TITLE: Power Uprate Program - Technical Support Center Todt-Accident Radiological Habitability Study l TWble 2.4 (Continued) ! 1
. Direct-Shine Doses (rad) to TSC Receptors l from post-LOCA Airborne Radioactivity Accumulating in the Refueling Level of the Reactor Building Time-Jhgl Rec . - 59 Rec. # 10 Rec. # 11 Rec. 812 ,
0.0 0.000E+00 0.000E+00 0-300E+00 0.000E+00 O.5 4.416E-02 9.828E-06 5.400E-07 3.955E-05 1.0 1.535E-01 3.375E-05 '2.528E-06 1.357E-04 2.0 4.659E-01 1.000E-04 7.358E-06 4.011E-04 8.0 2.556E+00 4.721E-04 3.218E-05 1.864E-03 24.0 5.974E+00 7.976E-04 4.968E-05 3.069E-03 48.0 7.958E+00 8.537E-04 f.162E-05 3.249E-03 96.0 9.574E+00 8.674E-04 5.186E-05 3.284E-03 168.0 1.075E+01 i.708E-04 5.189E-05 3.290E-03 240.0 1.150E+01 8.722E-04 5.189E-05 3.292E-03 744.0 1.333E+01 8.751E-04 5.190E-05 3.296E-03 Time Ihrl Rec. #13 Rec. #14 Rec. #15 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 4.084E-05 9.228E-06 3.244E-03 1.0 1.401E-04 3.162E-05 1.132E-02 2.0 4.143E-04 9.297E-05 3.467E-02 8,0 1.926E-03 4.221E-04 2.045E-01 24.0 3.174E-03 6.765E-04 5.557E-01 48.0 3.361E-03 7.101E-04 8.319E-01 96.0 3.398E-03 7.155E-04 1.113E+00 168.0 3.405E-03 7.163E-04 1.352E+00 240.0 3.407E-03 7.165E-04 1.511E+00 744.0 3.411E-03 7.169E-04 1.866E+00
l
- NPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO. %
PROJECT: JAF~ NEXT PAGE NO. st PREPARED BY: M/, - DATE 4/2sM49 CHECKED BY: /// DATE thi/M ' TITLE: Power Uprate Program - Technical S6pport Center 'toiti '
' Accident Radiological Habitability Study !
Table 2.5 Direct-shine Dose Rates -(rad /hr) to TSC Receptors fresa post-LOCA Released Radioactive Clouds Tinne
.[hrl. 'mee. e1 mee. e2 mee. e3 mee. e4 0.0 4.484E-02 7.536E-02 6.112E-02 6.174E-02 0.5 1.747E-02 .2.938E-02 2.381E-02 2.406E-02 l 1.0' 1.347E-02 2.267E-02 1.833E-02 1.852E-02 l 2.0- '1.068E-02 1.803E-02 1.450E-02 1.465E-02 8.0 5.850E-03 9.993E-03 7.844E-03 7.899E-03 24.0 1.851E-03 3.193E-03 2.456E-03 2.474E-02 48.0 2.645E-04 4.612E-04 3.457E-04 3.516E-04 96.0 8.307E 1.474E-04 1.054E-04 1.095E-04 168'.0 2.169E-05 3.853E-05 2.746E-05 2.857E-05 240.0 1.396E-05 2.481E-05 1.768E-05 1. 8'J 9 E - 05 744.0-- 7.090E-07 1.253E-06 8.986E-07 9.321E-07 Time .Ihr1 mee. es mee. e6 mee. e7 . e. e8 0.0 4.'911E-02 2.427E-02 7.213E-03 5.172E-03 0.5 1.912E-02 9.439E-03 2.800E-03 2.00)E-03 1.0 1.469E-02 7.178E-03 2.106E-03 1.509E-03 2.0 1.157E-02 5.516E-03 1.573E-03 1.121E-03 ;
8.0 6.084E-03 2.513E-03 5.870E-04 3.991E-04 24.0 1.867E-03 6.839E-04 1.257E-04 8.056E-05
.48.0 2.591E-04 8.638E-05 1.212E-05 7.707E-06 96.0 7.710E-05 2.123E-05 8.320E-07 5.360E-07 168.0 2.005E-OS 5.436E-06 1.627E-07 1.059E-07 240.0 1.291E-05 3.501E-06 1.054E-07 6.870E-08 744.0 6.584E-07 1.833E-07 8.810E-09 5.700E-09
[ NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.2 3l L PROJECT: JAF N p PAGE NO.t 32 PREPARED BY: A d. DATE oh.i/M CHECKED BY: JE2f DATE 7/M_.7 TITLE: Power Uprate Program - Technical Support Center 'Pept' - Accident Radiological'Babitability Study Table 2.5 - (Continued) Direct-shine Dose Rates (rad /hr) to TSC Receptors fresa post-LOCA Released Radioactive Clouds Time Jhgl Rec. G 9- Ree. 410 Rec . - Gil Rec. 812 0.0 6.725E-03 8.340E-04 1.240E-04 1.487E-04 0.5 2.614E-03 3.230E-04 '4.640E-05
]
5.570E-05 ) 1.0 1.960E-03 2.419E-04 3.480E-05 4.160E-05 2.0 1.453E-03 1.779E-04' 2.513E-05 2.980E-05 8.0 5.054E-04 5.769E-05 6.770E-06 7.240E-06 24.0 9.835E-05 1.017E-05 8.500E-07 6.400E-07 48.0 9.104E-06 9.300E-07 8.200E-08 6.200E-08 96.0 4.380E-07 4'000E-08
. 9.000E-09 1.000E-08 168.0 7.140E-09 6.000E-09 2.000E-09 2.000b-09 '240.0 4.670E-08 3.900E-09 1.300E-09 1.600E-09 744.0 5.250E-09 4.800E-10 8.001E-11 8.001E-11 Time Jhgl Rec. 813 Rac. #14 Rec. 415 0.0 '4.320E-04 1.471E-03 1.105E-03 0.5 1.665E-04 '5.714E-04 4.295E-04 1.0 1.239E-04 4.366E-04 3.214E-04 2.0 8.860E-05 3.377E-04 2.348E-04 8.0 2.108E-05 1.608E-04 7.132E-05 24.0 1.620E-06 4.650E-05 1.191E-05 48.0 1.310E-07 6.576E-06 1.525E-06 96.0 1.000E-08 2.041E-06 4.450E-07 168.0 2.000E-09 5.329E-07 1.159E-07 240.0 1.300E-09 3.432E-07 7.470E-08 744.0- 1.000E-10 1.724E-08 3.890E-09
i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.s 32-PROJECT: JAF NJET PAGE NO.: sa PREPARED BY: #fe TITLE: DATE -4/Whi Power Uprate Program - CHECKED BY: /c'/ Technical Support DATE 7hp/ff Center / Post'- ' I Accident Radiological Habitability Study Table 2.6 Direct-Shine Doses (rad)1t o TSC Receptors from post-LOCA Released Radioactive Clouds Time _ _(hr). Rec. #1 Rec. ~# 2 Ree. #3 Rec. #4 l 0.0 0'.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 1.452E-02 2.441E-02 1.979E-02 1.999E-02 1.0 2.221E-02 3.735E-02 3.026E-02 3.058E-02 l 2.0 3.423E-02 5.761E-02 4.660E-02 4.709E-02 ) 8.0 8.239E-02 1.393E-01 1.116E-01 1.127E-01 24.0 1.380E-01 2.347E-01 1.859E-01 1.875E-01 48.0 1.576E-01 2.687E-01 2.118E-01 2.136E-01 96.0 1.651E 2.819E-01 2.215E-01 2.236E-01 168.0 1.684E-01 2.878E-01 2.257E-01 2.280E-01 l 240.0 1.697E-01 2.900E-01 2.273E-01 2.296E-01 l 1744.0 .1.719E-01 2.940E-01 2.301E-01 2.326E-01 I Time _(hg). Rec. #5 Rec. #6 Rec- #7 Rec. G8 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 1.590E-02 7.852E-03 2.332E-03 1.d73E-03 1.0 2.430E-02 1.198E-02 3.550E-03 2.546E-03 1 2.0 3.737E-02 1.829E-02 5.377E-03 3.852E-03 8.0_ 8.858E-02 4.125E-02 1.143E-02 8.089E-03 24.0 1.457E 6.378E 02 1.628E-02 1.132E-02 48.0 1.653E-01 7.076E-02 1.748E-02 1.209E-02 96.0 1.725E-01 7.303E-02 1.773E-02 1.22SE-02 168'.0 1.756E-01 7.387E-02 1.776E-02 1.227E-02 240.0 1.768E-01 '.418E-02 1.777E-02 1.227E-02 744.0 '1.788E-01 /.4754-02 1.779E-02 1.228E-02
t-NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 33 PROJECT: JAF ' NEJT PAGE NO. : 34 PREPARED BY: Ad (p' DATE A/2t//94 CHECKED BY: #f DATE 9hr/97 TITLE: Power Uprate- Program - Technical Sdpport Center "PoWt'- Accident Radiological Habitability Study Table 2.6 -(Continued) Direct-Shine Doses (rad) to TSC Receptors frostpost-LOCA Released Radioactive Clouds Time Jhgl Rec. G9 Rec. 610 Rec. Gil Rec. 412 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 2.175E-03 2.693E-04 3.950E-05 4.730E-05 1.0 3.311E-03 4.096E-04 5.970E-05 7.160E-05 2.0 5.006E-03 6.180E-04 8.940E-05 1.070E-04 8.0 1.047E-02 1.278E-03 1.779E-04 2.090E-04 24.0 1.453E-02 1.733E-03 2.270E-04 2.590E-04 48.0 1.547E-02 1.829E-03 2.350E-04 2'.650E-04 96.0 1.565E-02 1.849E-03 2.380E-04 2.670E-04 168.0 1.567E-02 1.851E-03 2.380E-04 2.670E-04 240.0- 1.567E-02 1.850E-03 2.380E-04 2.670E-04 744.0 1.568E-02 1.852E-03 2.380E-04 2.680E-04 Time Jhrl Rec. 413 Rec. 414 Rec. 915 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 1.392E-04 4.757E-04 3.575E-04 1.0 2.114E-04 7.261E-04 5.440E-04 2.0 3.167E-04 1.111E-03 8.201E-04
-8.0 6.170E-04 2.549E-03 1.674E-02 24.0 7.590E-04 4.030E-03 2.230E-03 48.0 7.740E-04 4.527E-03 2.354E-03 96.0 7.760E-04 4.717E-03 2.397E 168.0 7.770E-04 4.798E-03 2.414E-03 240.0 7.780E-04 4.829E-03 2.421E-03 744.0 7.780E-04 4.884E-03 2.433E-03'
t NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 3 'f-PROJECT: JAF NJDCT PAGE NO. : MS PREPARED BY: p2/r - DATE_ M 4di CHECKED BY: /5 DATE /prArr TITLE: Power Uprate Program . Technical support Center f'Po6 c'- l Accident Radiological Habitability Study l Table 2.7 Worst-Case (MSLB) Direct Shine Dose Rates from Gamma Radiation Emanating frca the TSC Charcoal Filter Worst-Case TSC Receptor
- on E1. 286' i
Time WB Dome Rate Cum. Dose
.(hrs) (rad /hr) (rad) l 0.0 0.000E+00 0.000E+00 )
{ 1.0 4.675E-03 2.337E-03 1.5 4.156E-03 4.543E-03 2.0 3.568E-03 6.470E-03 5.0 1.828E-03 1.428E-02 8.0 1.220E-03 1.879E-02 12.0 8.151E-04 2.280E-02 18.0 5.054E-04 2.66.9E-02 24.0 3.416E-04 2.920E-02 36.0 1.846E-04 3.226E-02 60.0 8.138E-05 3.529E-02 96.0 3.771E-05 3.733E-02 372.0 7.935E-06 4.260E-02 744.0 2.085E-06 4.423E-02 (a) Between Receptors #4 and #6 in Fig. 2.5 Note: Charcoal Filtration System is on E1. 300'
r l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 3I l PROJECT: JAF p t PAGE NO.: 3g r2G DATE 81/pl[%9 CHECKED BY: Nf PREPARED BY: DATE offp/f9 i TITLE: Power Uprate Program - Technical support CenterYPdst' Accident Radiological Habitability Study l Table 2.8 l Worst-Case (MSLB) Direct Shine Dose Rates from l Gesuna Radiation Ennanating front the Charcoal Filters Receptor in Contact with TSC Charcoal Filter Casing Tinne WB Dose Rate Cum. Dose
- thrs) (rad /hr) (rad) 0.0 0.000E+00 0.000E+00 1.0 1.428E+00 7.139E-01 1.5 1.268E+00 1.387E+00 2.0 1.088E+00 1.975E+00 5.0 5.539E-01 4.349E+00 8.0 3.703E-01 5.716E+00 12.0 2.503E-01 6.9421>00 18.0 1.598E-01 8.152E+00 24.0 1.116E-01 8.95'8E+00 36.0 6.408E-02 9.9863+00 60.0 3.054E-02 1.107E+01 l 96.0 1.527E-02 1.187E+01
.372.0 3.559E-03 1.408E+01 744.0 9.353E-04 1.482E+01 l
? I I . I
I l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 3[ PROJECT: JAF NEXT PAGE NO.: 37 l PREPARED BY: M(,. DATE 9 /sV/4i CHECKED BY: 8 DATE /sp , ! TITLE: Power Uprate Program - Technical fiupport Center 7'Po'st' 7 Accident Radiological Habitability Study Fig. 2.1 - TSC Ventilati'on System Pressure Boundary (from JAF Drawing 11825-SK-92980 - Sheet 1 of 3) PRESSURE SOUNDARY 0\ , 0 h w vssTIBut.E Rb o~t
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NYPA - CALC.#'JAF-CALC-RAD-00023 REV 1 PAG.E NO.: 37 PROJECT: JAF NJXT PAGE NO.: w PREPARED BY: AT& DATE 9/t.4//19 CHECKED BY: /& DATE cp/jp/ff TITLE: Power Uprate Program - Technical Support Center '/P6st-Accident Radiological Habitability Study Fig. 2.2 - TSC Ventilation System Pressure Boundary (from JAF Drawing 11825-SK-92980 - Sheet 2 of 3) M -- Y .h $ @N r l 3 arrier i 2 , 3 4 s .
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i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 39( l PROJECT: JAF XT PAGE NO.: 59 PREPARED BY: 4 DATE 9/z#/49 CHECKED BY: DATE 7/pr/f7 TITLE: Power Uprate Program - Technical Support Center / Post-Accident Radiological Habitability Study Fig. 2.3 - TSC Ventilation System Pressure Boundary (from JAP Drawing 11825-SK-92980 - Sheet 3 of 3) kh k) kh (i?) c=C) N _.i . - - -
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I NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: .39 PROJECT: JAF NEXT PAGE NO.: 440 PREPARED BY: M(r- DATE 9/M/[19 CHECKED BY: /f/ DATE 9/,apd TITLE: Power - Uprate Program - Technical Support Center ' Post' f Accident Radiological Habitability Study Fig. 2.4 - TSC Ventilation System Flow Disgram Following TSC Activation (from Ref. 33) l l l l
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I NYPA - CALC.#..JAF-CALC-RAD-00023 REV 1 PAGE NO.: MO PROJECT: JAF NEXT PAGE NO.: 4// PREPARED..BY: # DATE f/WNi CHECKED BY: // DATE vipMf TITLE: Power. Uprate Prog.am - Technical Support Center "Po'it-Accident Radiological Habitability Study Fig. 2.5 -- Receptor Locations for Direct Shine Radiation Fields from External Sources - TSC El. 286' j l l s .N< f ol . i o2 3o o4 ~50 y l o6 P- . e : I o7 Bo ,,
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k NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 9 2. - PROJECT: JAF7~ NEXT PAGE NO.: vs PREPARED BY: A /r - DATE q /JN/44 CHECKED BY: 8 DATE 9/jp/y TITLE: -Power Uprate Program - ' Technical Support Center 'tojrt f 1 Accident Radiological Habitability Study
- 3. METRODS OF AM& LYSIS - Post-accident radiation _ fields and exposures in the JAF TSC were computed using the following l
(a) The methodology and assumptions'in the regulatory guides (Refs. 16 through 20, and 32) and the pertinent sections l~ of the Standard Review Plan (Ref. 21), (b)- Appropriate source terms, release pathways, decontamination factors and other assumptions, as described-in the sections which follow, j (c) Post-accident atmospheric dispersion factors based on 8-l years' worth of hourly meteorological data collected on l site by Niagara Mohawk, from JAF-CALC-RAD-00007, Rev. 2 i (Ref. 2), and L ! (d) The following CRE Computer Codes: - DORITA-2 (Ref. 13) Computation of radiation exposures, and definition of gn=mm spectra j associated with post-LOCA airborne radioactivity within the reactor building, and with the halogens accumulating on the TSC charcoal filtration system. QAD-CGGP (Ref. 39) Determination of the relative gn=== fluxes at the locations of interest
- (in terms of MeV/sec-cm* per MeV/sec l emitted by a source, as a function of gamma energy), for gn=mm radiation amanating from the reactor building and overhead. clouds.
MATILM. (Ref. 40) Computation of dose rates (and cumulative doses) at the receptors of interest as a function of post-p u i --
m i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: M.5 l PROJECT: JAF'~ NEXT PAGE NO.: 44k- l PREPARED BY: M /r -- DATE 9/gV/11 CHECKED BY: 8 DATE g/w/// TITLE: Power Uprate Program - Technical Support Center Post- , Accident Radiological Habitability Study accident time, using the gamma spectra generated by DORITA-2 and the relative gamma fluxes produced by QAD-CGGP. Sections 4 though 7 which follow present details of the assumptions, data and results associated with each of the design-basis accidents analyzed. ! l l For the MSLB and refueling accidents, the release of radioactivity to the atmosphere and into the TSC is very fast (relative to the l time allotted for. manual activation of the TSC HVAC system), and l therefore contamination of the air entering the TSC will drop l significantly in a very short time. Thus, TSC personnel exposures would tend to increase if the air-exchange rate of the TSC is reduced after the radioactivity has already entered the TSC (thue trapping the radioactivity within the TSC). In view of this ! possibility,.and in similarity with the analysis for the control room, iterative analyses were carried out using various TSC HVAC activation times to determine the worst-case radiation exposures for an MSLB and a refueling accident. The documentation in this calculation includes the results for three different TSC HVAC l activation times in the MSLB analyses which were used to define l the worst-case scenario. The worst-case TSC HVAC activation time ! for an MSLB was then used in the analysis of a refueling accident.
' Note that whole body and skin doses due to immersion in airborne j radioactivity within the TSC pressure boundary were calculated using the assumption of an immersion volume equal to the entire free air volume within the TSC pressure boundary (including all 3 elevations of the Administration Building) . As a result, the calculated doses may be conservatively applied to any receptor within the TSC pressure boundary. The thyroid doses (due to L
\
I NYPA - CALC.#'JAF-CALC-RAD-00023 REV 1 PAGE NO.: ##- l PROJECT:'JAF '~. NJIT PAGE NO.: 4 PREPARED BY: /6 DATE Jhpdf CHECKED BY: #/ DATE ehr/// ' TITLES.' Power Uprate Program - Technical -Support Center Post- ) Accident Radiological Babitability Study inhalation)lwould also be the same since uniform mixing was assumed._ Doses from external sources,.however, are different.at different elevations of the Administration building.
.As a final remark, note that the TSC radiation exposures were computed under the assumption of continuous occupancy for the duration'of the-accident.. The regulatory models governing the habitability of control rooms require 100% occupancy during only the first 24 hours of a postulated accidents for days 2, 3 and 4 1 the occupancy factor is reduced to 60%, and for periods beyond 4 days 40% occupancy is allowed. . However, to simplify this analysis , credit.for partial occupancy was not taken. In addition, the TSC was' assumed to be occupied throughout the entire duration of an accident.(i.e., from t = 0 hr, before the time of TSC . activation) with the single exception of a MSLB accident with iodine _ spiking to 2 pci/g. In this case, a preisolated TSC is not habitable innaediately af ter 'M dent onset, but beccmes habitab?'
af ter' about 30-45 minutes -(1 hr is assumed in this case) . Note that-NUREG-0654 requires activation of Emergency Response Facilities within 1 hour of the declaration of a significant event. Taking- credit for partid occupancy would not significantly affect the results given in.Section 2. Essentially all of the doses for the MSLBA/ CRDA and RA are received within the first 24 hours; thus'there would be a negligible decrease in doses for these scenarios. The post-LOCA thyroid dose would decrease by about 30% from 12: rem. -There would also be a reduction in direct shine doses, but this would not significantly improve habitability (see Section 2.2.1).
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; .p .i NYPA'- CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 4/f PROJECT: JAF ~ ~ NJXT PAGE NO. vg PREPARED BY: M/,- ' TITLE:
DATE 9/1#b4 CHECKED BY: /(o' Power -Uprate Program . DATE Technical. Support Center g/fr/9f Po's h-l Accident Radiological Habitability Study 4 .- RADIATIQat IIPOSURES FROM A LOSS OF COOLhMT ACCIDENT Release pathways and contributing radiation sources which are typically addressed-in the analysis of a LOCA are the followings (a) Drywell leakage,- (b) . ESF Consponent leakage, l (c) Direct ganssa radiation from airborne radioactivity
. accumulating on.the refueling floor of the reactor building, (d) . External' cloud exposure, (e) Halogen activity accumulating on intake / exhaust filters,
, and i j (f) MSIV leakage. f L j For the JAP TSC, release pathways (a) and (b) are of primary concern and are addressed in the subsections which follow. Pathways: (c) , - (d) and (e) are addressed in'Sec. 8, and pathway (f) is not applicable, as discussed below. With respect.to MSIV leakage, JAF is equipped with a Main Steam l Leakage Collection System (MSLCS) whose safety objective is to collect and process leakage past the MSIVs following a LOCA. The affluent.of the MSLCS is processed by the Standby Gas Treatment System (SGTS) ' and is exhausted through the. stack. I The.drywell-leakage pathway is assumed to include MSIV leakage,
.and' separate assessment of MSIV leakage is'not needed (see Ref. 49 l
for calculating of the drywell leak rate).
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NEXT PAGE NO.: /,/ y I PREPARED BY: //(f DATE .4h4Fhf CHECKED BY: // DATE ehy/ff TITLE: ; Power Uprate Program - Technical Support Center 'Pont-Accident Radiological Babitability Study 4.1' Drywell Leakage 4.1.1 Basic Data and Assumptions The'following data and' assumptions were used in the computation of
-the;TSC radiation exposures as.a result of post-LOCA drywell leakage ,I .(a)
A LOCA takes place at full' power (2535.8 MWt + 2%
-uncertainty, i.e. , 2586.5 MWt) (Ref. 41 and Reg. Guide 1.49 ~
- . (Ref.'19)].
i < (b) The core inventory for the radionuclides of interest is shown l-in Table 4.1 (based on information from Ref. 14). (c) 100% of the noble gases and 25% of the halogens present in the core remain airborne within the drywell where they are available as an aerosol for leakage to the secondary containment (Reg. Guide 1.3 (Ref. 16)]. (d) The halogen composition airborne within the drywell is as follows: 91%. elemental, 4% organic and 5% particulate [ Reg. Guide 1.3 (Ref. 16)].- (e). Leakage from the drywell is at the rate of 1.5% per day
-(UFSAR Secs.'14.8.1.5 and 14.8-22). This rate accounts for both-drywell-containment leakage and MSIV leakage (Ref. 49).
l This rate is assumed to be constant for the accident l j; duration. The design-leak rate is 0.5% per day of containment volume (Technical Specifications Sec. 4.7.A.2.8, and.UFSAR Secs. 11.5.3.10 and 14.6.1.3.5). Use of the 1.5% p per. day value is conservative.
.(f)
All:the noble gases and halogens leaking from the drywell are avhausted to the' atmosphere via the Standby Gas Treatment System (SGTS) and the~ main stack without mixing in the reactor building (Ref. 25). (g) The SGTS charcoal filter efficiency for the removal of halogens is 90% for-~all halogen species (verified per plant procedures).
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Power _ Uprate Program - Technical Support Center Nott-Accident Radiological Habitability Study (h) The atmospheric dispersion factors associated with the transport of released radioactivity to the TSC intake are ar follows (in ' (sec/m*) , from Ref. 2) Interval: 0-8 hrs 8-24 hrs 1-4 days 4-31 days j (X/Q).. 9.26E-07 6.75E-07 3.39E-07 1.26E-07 Note that the prescribed assumption for fumigation conditions prevailing at the site at the time of an accident is not applicable to the TSC outside air intake. (i) The TSC is located in the Administration Building and the pressure boundary of its ventilation system extends from El. 272' to El. 300' (see Figs. 2.1 - 2.3). Based on these figures and on the assumption that 20% of the space is taken up by equipment, the free air volume within the TSC pressure boundary is as follows: El. 272's 670 f t* x 13. 5 f t x 0. 8 = 7,520 ft 3 El 286's 9340 f t' x 13.5 f t x 0.8 = 100,900 ft' El. 300's 253 0. f t' x 19. 5 f t x 0. 8 = 39,500 ft' Total = 147,640 ft 8 This volume was conservatively increased to 1.5E+05 ft 8. Note that the tchal floor area within the pressure boundary is 12,540 ft'. The minimum required area, as specified in NUREG-0596 (Ref.,7), is 1875 ft* (for 25 persons at 75 ft* per person). (j) When operation of the TSC is required, the Administration Building Office Area normal HVAC system operation is modified to supply air only to the TSC. As described in Operating i Procedure OP-59B (Ref. 33) and in the Design Basis Document for the Administration Building HVAC system (Ref. 35),
I i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PROJECT: JAF ^ PAGE NO. fY NEXT PAGE NO. 4/9 PREPARED BY: pfk_, DATE 4 / MI/1 1 C H E C K E D B Y : /f/ DATE f/7p/q TITLE: Power Uprat.e Program. - Technical Eiupport Center '4 9 (t y
= Accident Radiological Habitability Study exhaust fans 72FN-20 (from the conference room) and 72FN-21 (from the . print. area) are manually tus.aed off, and booster fan 72FN-47 (for the charcoal filtration system) is turned on. ' Simultaneously with the startup of 72FN-47, outside air damper 72 MOD-106 opens (to supply air to the filter train),
isoletion dampers 72 MOD-107, 108 and 109 for various Office areas close, and erAaust air dampers 72AOD-146 and 72AOD-71 close.- See Tig. 2.4.for a simplified flow diagram of the TSC HVAC system. The TSC HVAC system is designed to supply between 3300 and 2700 - scfm (3000 i 10% uncertainty) filtered air to maintain (1/8)" WG pressure (Ref. 35 and UFSAR Sec. 9.9.3.10). The filter train consists of profilters, electric heater, HEPA profilter, charcoa?. absorber, and HEPA after-filter. The TSC HVAC system is non-safety related (per MCM-6A, Ref. 36); hence, there are no redundant components, and single-failure criteria are not applicable. Based on the above, the TSC air exchange rate and filtration were modelled as follows in the present analysis: Unfiltered air intakes (1) 25,800 scfm prior to activation of the TSC (the maximum flow that can be provided by the 72-AHU-4 unit, from Drawing 11825-FB-32G), (2) 1500 scfm when the TSC ventilation system is isolated and there are no system fsilures*. This value was arbitrarily selected to accommodate the following: (a) Leakage through AOD-171, estimated to be less than 25 (scfm/ft'") x (9'x3.5') ( f t ) = 790 scfm via the closed AOD, where 25 (scfm/ft') is the worst-case damper leakage provided in Ref. 34. (b) Potential ventilation booster fan shaft leakage, to
'NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.9 Yf PROJECT: JAF~ N p PAGE NO. 50 PREPARED BY: M (, - DATE 9/24f/99 CHECKED BY: X[ DATE . g/>fA,y ~
TITLE: Power Uprate Program - Technical Support Center PWt-Accident Radiological Habitability Study l Post-isolation filtered j air intake rate: 3300 or 2700 scfm (3000 i 10% uncertainty) without system failures. ) i
- Intake filter off.: *" % for all halogens species (2" {
charcoal beds with humidity control) l (Refs. 45 and 46; 90% verified per plant procedure). TSC isolation times 1 hour after the postulated LOCA.
. (Note: Ventilation Isolation time is assumed to be from the onset c,f the accident. A small delay (< 1 minute) resulting from the difference in onset of the LOCA and the alarming of .
the TSC will have a negligible effect on the dose to TSC l personnel). j In summary, the case was analyzed, as follows: Without TSC HVAC system failures 0 - 1 hr 25800 scfm unfiltered outside air : 1 - 744 hr 3300 or 2700 scfm filtered (3000 i 10% uncertainty) + 1500 scfm unfiltered. 1 (Note,t The intake rate into the TSC is provided as input to the DORITA-2 code in units of TSC volumes per day, and includes both filtered and unfiltered flows. The split between the filtered and unfiltered portions is accomplished through use of the filter bypass fraction. For instance, for a filtered flow of 3300 scfm and an unfiltered flow of 1500 scfm, the fractional air intake rate into the TSC is equal to (4800 (sefm) x 1440 (min / day) / 1.5E+05 ft']
= 46.08 air volumes per day, and the filter bypass fraction is (1500 scfm / 4800 scfm) = 0.3125, as shown in the DORITA-2 input files in Attachment A.]
cover the concerns of NRC Information Notice 93-06 (Ref. 37)' and JAF ACTS Item #17875, and (c) 10 scfm inleakage due to ingress / egress (from Ref. 21, Sec. 6.4) h
1 i l l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: Io PROJECT: JAF NEXT PAGE NO.: r/ PREPARED BY M[r- DATE 9/24/4f CHECKED BY: /ff DATE - /af4F TITLE: Power Uprate Program - Technical Support Center Post- j l Accident Radiological Habitability Study j (k) Radiation exposures were determined for a 31-day interval and for continuous occupancy, from the time of the accident 8. Whole body exposures were based on immersion in a finite spherical cloud having a volume equal to that of the TSC. The breathing rate was set at-3.47E-04 (m*/sec) for the duration of the accident (Refs. 16 and 26), and use was made of the ICRP-30 dose conversion factors for thyroid exposure (Ref. 12). i l 8 The time delay between the accident and activation of the TSC (about 1. hour) was conservatively ignored, and the TSC was assumed to be occupied at the time of the accident.
l I l NYPA - CALC.# JAF-CJWF-3AD-00023 REV-1 PAGE NO. I/ PROJECT: JAF PAGE NO.: 62_ PREPARED BY M(, - DATE q/W/ff CHECKED Br: .' DATE #ffApr TITLE: Power Uprate Program - Technical Support Center "Ponti ' Accident Radiological Habitability Study Table 4.1 Full-Core Inventory; 2586.5 38tt Nuclide Activ.(Ci) Nuclide Activ. (C1) Br 83 8.078E+06* Kr 83m 8.114E+06 Br 84 1.432E+07 Kr 85m 1.742E+07 Br 85 1.717E+07 Kr 85 7.798E+05 Kr 87 3.342E+07 I 129 2.254E+00 Kr 88 4.733E+07 I 130 2.705E+06 Kr 89 5.887E+07 I 131 6.805E+07 I 132 9.945E+07 Xe 131m 4.092E+05 I 133 1.423E+08 Xe 133m 5.962E+06 i I 134 1.566E+08 Xe 133 1.430E+08 l I 135 1.344E+08 Xe 135m 2.695E+07 I 136 6.479E+07 -Xe 135 1.847E+07 Xe 137 1.255E+08 Xe 138 1.192E+08 I l 3.123E+03 (Ci/MWt from Ref. 14) x 2586.5 (MWt)
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PAGE NO. 53. PREPARED BY: pffy DATE 4/sA//fi CHECKED BY: N W p/ DATE 7//p/59 TITLE: Power Uprate Program - Technical Support Center 'Po'dtil Accident Radiological Habitability Study-l 4.1.2 Results TSC radiation ~ exposures due to drywell leakage following a design-i basis LOCA were calculated using the DORITA-2 computer code and tha' data and assumptions listed above. Copies of the DORITA-2 outputs appear in Attachment A to this calculation (Computer Run Cases # 1 & 2).
.T'able 4.2 which follows presents the time-dependent thyroid, whole body and skin doses in the TSC (due to post-LOCA drywell leakage) I for continuous occupancy. Refer to Sec. 2 for a summary of the exposures. ,
I i 1
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: II PROJECT: JAF Np T PAGE NO.: 5 44.- l PREPARED BY: M/,,- DATE 4/sA//91 CHECKED BY: g ,' DATE 7/pf/Sy TITLE: Power. Uprate Program - Technical support Center /Podst!' Accident Radiological Habitability Study Table 4.2 TSC Radiation Exposures Due to Drywell Leakage Following A Design-Basis Loss-of-Coolant Accident [3300 scfm filtered (3000+10% uncertainty)] Time Thyroid Whole Body Skin (huurs) Dos, (resa) Dome fram) Dose frem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.103E-01 7.916E-04 8.432E-03 1.000E+00 4.684E-01 1.449E-03 1.428E-02 8.000E+00 1.935E+00 5.839E-03 5.361E-02 2.400E+01 3.853E+00 9.511E-03 9.423E-02 9.600E+01 6.989E+00 1.217E-02 1.269E-01 7.440E+02 9.840E+00 1.345E-02 1.433E-01
)
[2700 scfm filtered (3000-10% uncertainty)] Time Thyroid Whole Body Skin (hours) Dose frem) Dose frem) Dose frami 1 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.103E-01 7.916E-04 8.432E-03 1.000E+00 4.684E-01 1.449E-03 1.428E-02 8.000E+00 2.084E+00 5.811E-03 5.345E-02 2.400E+01 4.207E+00 9.426E-03 9.387E-02 9.600E+01 7.678E+00 1.208E-02 1.266E-01 7.440E+02 1.083E+01 1.337E-02 1.430E-01 Note: Continuous occupancy in all cases
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Technical Support CenterNPoistG Accident Radiological Habitability Study - l 4.2 ESF Component leakage 4.2.1 Basic. Data and Assumptions The_following data and assumptions were used to calculate the post-LOCA dose contribution from ESF component leakage: (a) A'LOCA takes place at full power (2586.5 MWt). (b) The core inventory for the radionuclides of interest (halogens in.this case) is as shown in Table 4.1 above. ('c) 50% of the total halogen activity present in the core mixes
~
uniformly with the coolant in the RER system, which hac a total . fluid mass of 3.21 x 10' grams. This is equal to approximately 113,400 cu ft, consisting of (431190 lbs / 62.4 lbs/cu ft) = 6,900 cu ft of cold RCS coolant (from JAF
-Drawing 5.01-101A), 105,600- cu ft of torus water (from UFSAR . Table 5.2-1),.and 900 cu_ft of water from other sources.
(d) Total ESF component leakage rata is 5 gpm,- corresponding to a fractional ~ rate from the recirculating water system of 0.00849 volume per day.
, (e) The'ESF component' leakage.of 5 gpm is assumed to be constant from the start of the LOCA through the duration of the accident.
(f) -An additional leakage contribution due to a gross failure of a-passive compon6nt with an assumed leak rate of 50 gpm is included'in the model (Ref. 21, SRP, Sec. 15.6.5, Appendix
-B). This leakage is assumed to begin at the time of LOCA onset and lasts for a period of 30 minutes. (This assumption is more conservative than-the SRP model which assumes that the additional leakage begins at 24 hours after the LOCA). - (g) It is further assumed that 10% of the halogens contained in the water from ESF component leakage become airborne within the. Reactor Building (Ref. 21, SRP, Sec. 15.6.5, Appendix B),
and mix uniformly with the RB atmosphere. 4
l l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: (( PROJECT: JAF NJXT PAGE NO. : gg ; PREPARED BY: M /r- - DATE 9/tt//1f CHECKED BY: /// TITLE: Power Uprate Program - DATE 7Atyh5 Technical Support Center /P66 M , Accident Radiological Habitability Study l (h)' Release from the reactor building is throug'x ;he SGTS and the main stack at the rate of 3.3 air changes per day (based on l an SGTS flow of 6000 scfm with one fan operating (UFSAR Sec. 5.3.3.4)]. (i) The SGTS filter efficiency for the removal of halogens is 90% for all halogen species (verified per plant procedures) . ! (j) Transport of the released radioactivity to the TSC, the TSC characteristics and other exposure-related parameters are described under Items (h) through - (k) in Sec. 4.1.1. 4 l l l 1
- i l
I. l t l
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REV 1 PAGE NO. $[ NJtXT PAGE NO. - s7 PREPARED BY: DATE 9/s#/d CHECKED BY: /// DATE TITLE:
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Power Uprate Program - Technical , y/sy/9f Support Center 'P6'st-l
)
Accident Radiological Habitability Study I 4.2.2 Results Table 4.5 presents the post-LOCA time-dependent' thyroid, whole
. body and skin doses in the TSC resulting from post-LOCA ESF component leakage. Refer to Sec. 2 for a sununary of the ,
exposures. Also,. refer to Attachment A for copies of the DORITA-2 outputs (Computer Run cases # 1 & 2). l-I
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Technical Support Cente P - Accident Radiological Babitability Study I i Table 4.3 r. TSC Radiation Exposures
- Due to ESF Component Leakage Following A Design-Basis Loss of Coolant Accident (33001 sofa filtered (3000+10% uncertainty)]
Tinne Thyroid whole Body Skin thours) Dose frem) Dose (ren) Dose (rma) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.0005-01 6.817E-03 2.757E-06 '1.610E-05 , 1.000E+00 2.576E-02 1.010E-05 5.982E-05 8.000E+00 1.540E-01 8.364E-05 6.217E-04 2.400E+01 3.602E-01 2.212E-04 1.808E-03 9.600E+01 7.203E-01 2.826E-04 2.407E-03 , 7.440E+02 1.070E+00 2.953E-04 2.543E-03 , [2700 sefa filtered (3000-10% uncertainty)] Time Thyroid whole Body Skin thours) Dose frem) Dose (rea) Dose (rea)
'0.000E+00 '0.000E+00 0.000E+00 0.000E+00 ~5.000E-01 6.817E-03 2.757E-06 1.G10E-05 t 1.000E+00 2.576E-02 1.010E-05 5.982E-05 8.000E+00 1.667E-01 8.254E-05 1.184E-04 2.400E+01 3.948E-01 2.151E-04 1.787E-03 9.600E+01 7.934E-01 2.756E-04 2.388E-03 .
7.440E+02 1.180E+00 2.889E-04 2.5302-03 Note: Continuous occupancy in all cases a
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REV 1- PAGE NO.: PAGE NO.t (( ! S9 ! PREPARED BY: M (r DATE 9/M//91 CHECKED BY: DATE- p g /50P TITLES. Power Uprnte Program - Technical Support CenteF P6dti ' Accident Rt4diological Habitability Study
/4.3 Total LOCA Dose-The total LOCA radiation doses due to both drywell and ESP component leakage are shown in Table 4.4. The table was prepared by summing,the results in Tables 4.2 and 4.3. l l
l l 1 1 , n ! ) l o i l
NYPA - CALC.# JW-CALC-RAD-00023 REV 1 PAGE NO.: 89 PROJECT: JAF T PAGE NO.: #6 PREPARED BY: Af & DATE %f//d CHECKED BY: DATE p/OfA. >' TITLE: Power Uprate Program - Technical Support Center /Pogt' - Accident Radiological Habitability Study Table 4.4 TSC Radiation Exposures Due to Drywell and ESF Consponent Leakage Following A Design-Basis Loss of Coolant Accident [3300 sofa filtered (3000+10% uncertainty)] Time Thyroid Whole Body Skin thours) Dose frem) Dose frem) Dose frem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 ! 5.000E-01 2.171E-01 7.944E-04 8.448E-03 1.000E+00 4.942E-01 1.459E-03 1.434E-02 8.000E+00 2.088E+00 5.923E-03 5.423E-02 2.400E+01 4.213E+00 9.732E-03 9.604E-02 9.600E+01 7.709E+00 1.245E-02 1.293E-01 7.440E+02 1.091E+01 1.375E-02 1.458E-01
)
[2700 scfm filtered (3000-10% uncertainty)] Time Thyroid Whole Body Skin (hours) Dose frem) Dose (rem) Dose (rsa) ] 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.171E-01 7.944E-04 8.448E-03 1.000E+00 4.942E-01 1.459E-03 1.434E-02 8.000E+00 2.251E+00 5.894E-03 5.407E-02 2.400E+01 4.602E+00 9.641E-03 9.566E-02 9.600E+01 8.471E+00 1.236E-02 1.290E-01 l 7.440E+02 1.201E+01 1.366E-02 1.455E-01 I I I Note: Continuous occupancy in all cases ! l l j
f NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: [6 j PROJECT: JAF NJXT PAGE NO.: // PREPARED BY: Mh DATE 4/29/f/CHECKEDBY: E[ DATE . jfMy , TITLE: Power Uprate Program - Technical Support Cente P'cist'
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Accident Radiological Habitability Study l j
- 5. RADIATION EIPOSURES FROM A MAIN STEAM LINE BREAK 5.1 Basic. Data and Assumptions As was the case with all accident analyses documented in this calculation, the computation of TSC radiation exposures associated with a postulated MSLB outside containment was based on data and assumptions consistent with the ragulatory guidelines, specifically, Ref. 17 (Regulatory Guide 1.5), and the Standard R'eview Plan (Ref. 21, Sec. 15.6.4). The following data and assumptions were used in the dose calculation for the MSLB accidents i
(a) A main steam line break occurs outside containment during i full power operation. [ Note: A 0% power MSLB accident could release more mass (Ref. 48). However, the release given in ! (c) below would still be bounding.] (b) The main steam isolation valves close in 10.5 seconds after l l the break (UFSAR Sec. 14.6.1.5.1.e, pg 14.6-29). (Note: j Actual closure time is approximately 3 to 5 seconds.) (c) The releace through the break consists of 20,000 lb of steam ; and 120,000 lb of water (Ref. 48) nnd Figure 14.6-12 of FSAR. I I (d) The ensuing high fuel temperatures do not lead to any fuel i damage. ' (e) The noble gas fission product concentrations in the steam ! correspond to the design values which wot:ld yield the standard release rate to the atmosphere during normal j operation (i.e., 100,000 pCi/sec fol2 awing a 30-minute decay) ; (Ref. 14). 100% of all noble gaser leaving the reactor vessel during the 10.5-sec MSIV closure time are released via ; the break. The total noble gar releases following an MSLB are as follows: l 1 4 i I
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' Source Term at t=0 MSLB Release Nuclide fuci/mee) 1l;i).
Kr 83m 3.4E+03 3.57E-02* Kr 85m 6.1E+03' 6.41E-02
-Kr 85 2.0E+01 2.10E-04 Kr 87 2.0E+04 2.10E-01 Kr 88 2.0E+04 2.10E-01 . Kr 89 1.3E+05 1.37E+00 Xe 131m 1.5E+01 1.58E-04 Xe 133m 2.9E+02 3.05E-03 Xe 133 8.2E+03 8.61E-02 Ze 135m 2.6E+04 2.73E-01 Xe 135 2.2E+04 2.31E-01 Xe 137 1.5E+05 1.58E+00 Ze 138 8.9E+04 9.35E-01
- 3.40E+3 (pci/sec) x 10.5 (sec) x 1.0E-06 1. ,J. / Ci)
The balogen inventory in the steam was determinod to be insignificant in comparison to that in the discharged liquid, and was not considered. [ Note: the steam-to-water halogen { concentration ratio is of the order of 3x10~5 (UFSAR Sec. 14.6.1.5.2.b).] i The halogen source term in the discharged liquid was selected to j represent the Tech Spec limits for the following: ; The maximum RCS concentration under power uprate
- 1) equilibrium conditions (0.2 pCi/gm I-131 DE)', and ii) The maximum RCS concentration under Limiting Conditions of l ._
Operation' (2 ' pCi/gm I-131 DE) This is~the GE Standard Technical Specification limit (Ref. 24),
l l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: PROJECT: JAF DATE 4 /s.Mff CHECKED BY: Kgf W T PAGE NO.: d9h PREPARED BY: /1 b DATE S/Sp TITLE: Power Uprate Program - Technical Support Center '/ toft'- Accident Radiological Habitability Study l Note that there is a 10-fold increase in the RCS 1 concentrations between the two cases. The releases of l J radioactivity presented below are for the equilibrium RCS conditions. For the spiked-RCS case, multiply the released I activities by 10. The relative coolant activities employed in the determination of the post-MSLB halogen releases were based on the data in Ref. 14, which are as follows: Nuclide Primary Coolant Activity (uci/erm) Br 83 0.025 Br 84 0.041 Br 85 0.021 I 131 0.027 l I 132 0.21 l I 133 0.18 I I 134 0.38 l I 135 0.25 l Under equilibrium conditions, the halogen concentrations in the RCS, and the total activities discharged into the turbine building, would be as shown in the table which follows.
l NYPA - CALC.# JAF-CALC-RAD-00023 PROJECT: 'JAF REV 1 PAGE NO.: b PAGE NO.: #4t.-. PREPARED BY: M /,- DATE 4/84/9f CHECKED BY: DATE . TITLE: Power Uprate Program - Technical Support Center /'/#ffr7 Post-
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Accident Radiological Habitability Study 1 Reactor Coolant System Activities and Post-MSLB Releases (Based on RCS concentration of 0.2 pCi/gm I-131 DE) P r l .' b y ICRP-30 Coolant Activity. Activity Nuclide DCF Ac.tivity DCF x ACT 4.2 uci/gm DCF x ACT Release (Rem /Ci) (uCi/gn) (Ci) I-131 1.07E+06 2.700E-02 2.889E+04 8.130E-02 8.699E+04 4.425E+00 I-132 6.29E+03 2.100E-01 1.321E+03 6.323E-01 3.977E+03 3.442E+01* I-133 1.81E+05 1.800E-01 3.258E+04 5.420E-01 9.810E+04 2.950E+01 I-134 1.07E+03 3.800E-01 4.066E+02 1.144E+00 1.224E+03 6.227E+01 I-135 3.15E+04 2.500E-01 7.875E+03 7.528E-01 2.371E+04 4.098E+01 I-131 DE = 6.642E-02 pCi/gm 2.000E-01 pCi/gm
- 6.323E-01 (pci/ga) x 120,000 (1b) x 453.6 (gm/lb) x 1.0E-06 pCi/Ci)
The dose conversion factors (DCFs) in the above table were j extracted from ICRP-30 (Ref. 12). The listed primary coolant concentrations were adjusted'to yield the limit of.0.2 (pci/gm) I-131 DE according to the following expression: I-131 D.E. = I Qi x DCF3 / DCF.1322 1 Where, Qi = Coolant concentration in pCi/gm DCFi = Dose conversion factor for i* species The above concentrations were also used to compute the total bromine activities which would be released to the turbine l building. These are as follows: l
)
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- 4.425 (I-131 Ci released, from above) x 0.025 (Br-83 Conc.) / 0.027 (I-131 Conc.)
l Activation products and other particulates in the coolant were neglected since they would not become airborne. l l (f) 100% of the coolant halogens discharged in the turbine ] building are. assumed to become airborne and released to the
-atmosphere at ground level over a period of 2 hours. The l selected release rate was equivalent to 72 air changes per day, and the cumulative releases to the atmosphere as a '
function of time would be as follows: Post MSLB Time Cwmulative , (min) Release (%) 0 0.0 5 22.1 10 39.3 15 52.8 20 63.2 30 77.7 45 89.5 60 95.0 90 98.9 120 99.8 (g) The atmospheric dispersion factors associated with the transport of released radioactivity to the TSC intake are as follows (in (sec/m') , from Ref. 2]:
i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: M PROJECT: JAF W PREPARED BY: /Y/, - DATE 4/sN/49 CHECKED BY: 13/6 EXT PAGE NO. DATE . pp/fy { l l' TITLE: Power Uprate Program _- Technical Support Cente P'6s ti- i l Accident Radiological Habitability Study i l l Interval: 0-8 hrs 8-24 hrs 1-4 days 4-31 days I (X/Q) _ 3.56E-03 3.03E-03 2.14E-03 1.29E-03 (h) The TSC characteristics and other exposure-related parameters j are as described under Items (i) - through (k) in Sec. 4.1.1, t 1 with one exception: Isolation of the TSC ventilation system ! ! was assumed to take place at a post-accident time (without a j delay for TSC isolation) - which would maximize the TSC radiation exposures (12 min.). l As demonstrated under Item (h) above, the release of l radioactivity to the atmosphere and into the TSC following an MSLB is very fast relative to the time allotted for manual activation of the TSC HVAC system. As a result, contamination of the air entering the TSC will drop
- -significantly in a very short time. Thus, TSC personnel exposures would increase if the air-exchange rate of the TSC l is reduced after the majority of the released radioactivity has already entered the TSC (thus trapping the. radioactivity l within the TSC). In view of this possibility, iterative l analyses were carried out using various TSC HVAC isolation times to determine the worst-case radiation exposures. The documentation in this calculation includes the analyses with TBC HVAC isolation times equal to 9, 12 and 15 min.
I Isolation in'12 minutes was determined to maximize the TSC dose. In addition, the TSC dose was calculated when the TSC is pre-isolated at the sp!.ced RCS concentration of 2 pCi/gm I-131 DE.
1 i
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PREPARED BY: 'M M DATE %V[d DATE 9/M /SS TITLE: Power, Uprate.. Program - Technical Shpport Center @ c4tM Accident Radiological Habitability Study 5.2 Results. 1 TSC radiation exposures following a design-basis MSLB were calculated using the DORITA-2 computer code and the_' data and assumptions listed above. Copies of the DORITA-2 outputs appear in Attachment A to this calculation (Computer Run Case #1 and #2). l The DORITA-2 results of interest (namely the thyroid, whole body and skin exposures.as a function of post-MSLB time) are summarized in Tables 5.1.and 5.2. . Table 5.1 corresponds to the three different TSC HVAC isolation times identified under Item (h) in l l Sec. 5.1. The worst-case exposures occur with the spiked RCS concentration of 2 Ci/gm I-131 DE. However, for the case of iodine spiking to 2 p Ci/gm I-131'DE, doses romain below the limits for a pre-isolated TSC only if the TSC is not occupied for the first hour. I The earlier the TSC is occupied, the higher the dose, which reaches a maximum of over 50 rem: thyroid for occupancy at t=0 (see Table 5.2). 1 o
l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.S [7 PROJECT: JAF @ PAGE.NO. d'q( PREPARED BY: MP DATE AMh1 CHECKED BY: /6' DATE _yf1jv/ / TITLE: Power Uprate Program' - Technical Support Center Post-Accident Radiological Habitability Study Table 5.1 TSC Radiation Exposures Following A Design-Basis Main Steam Line Break . (Equilibrium RCS activity at 0.2 pCi/gm I-131 DE) [3300 scen filtered (3000+10% uncertainty)] Time Thyroid Whole Body skin (hours) Dose frami Dose frem) Dose frem) I Case'#1A 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.500E-01 2.701E+00 3.382E-03 1.922E-02 8.000E+00 1.924E+01 2.009E-02 1.136E-01 2.400E+01 1.924E+01 2.039E-02 1.156E-01 9.600E+01 1.924E+01 2.046E-02 1.161E-01 7.440E+02 1.924E+01 2.047E-02 1.161E-01 Case #1B 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E-01 3.990E+00 4.932E-03 2.791E-02 8.000E+00 1.993E+01 2.072E-02 1.170E-01 2.400E+01 1.993E+01 2.098E-02 1.188E-01 9.600E+01 1.993E+01 2.104E-02 1.192E-01 7.440E+02 1.993E+01 2.104E-02 1.192E-01 Case #1C 0.000E+00 0.000E+00 0.000E+00 0.000E+00 l 2.500E-01 5.230E+00 6.390E-03 3.606E-02 1 8.000E+00 1.996E+01 2.069E-02 1.168E-01 1 2.400E+01 1.996E+01 2.091E-02 1.183E-01 9.600E+01 1.996E+01 2.096E-02 1.187E-01 7.440E+02 1.996E+01 2.097E-02 1.187E-01 Cases #1A-1C: Without TSC HVAC system failures, and TSC HVAC activation times as follows: Case 1A: 9 min Case 1B: 12 min (1bniting. case for WB & skin) < Case 1C: 15 min (negligibly higher than l 1B for Thyroid) ! I Note: Continuous occupancy in all cases I l
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Table 5.1 (Continued) TSC Radiation Exposures Following i A Design-Basis Main Steam Line Break (Equilibrium RCS activity at 0.2 pCi/gm I-131 DE) [2700 sefa filtered (3000-10% uncertainty)] l l Time Thyroid Whole Body Skin (hours) Dose frem) Dose frem) Dose frem) Case #1A 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.500E-01 2.701E+00 3.382E-03 1.922E-02 8.000E+00 2.144E+01 2.181E-02 1.234E-01 2.400E+01 2.144E+01 2.208E-02 1.252E-01 9.600E+01 2.144E+01 2.214E-02 1.256E-01 7.440E+02 2.144E+01 2.214E-02 1.257E-01 Case #1B l 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E-01 3.990E+00' 4.932E-03 2.791E-02 l 8.000E+00 2.206E+01 2.237E-02 1.265E-01 l 2.400E+01 2.206E+01 2.260E-02 1.280E-01 9.600E+01 2.206E+01 2.265E-02 1.284E-01 7.440E+02 2.206E+01 2.265E-02 1.274E-01
- Case #1C 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.500E-01 5.230E+00 6.390E-03 3.606E-02 8.000E+00 2.194E+01 2.220E-02 1.254E-01 2.400E+01 2.194E+01 2.239E-02 1.268E-01 9.600E+01 2.194E+01 2.244E-02 1.271E-01 7.440E+02 2.194E+01 2.244E-02 1.271E-01 l
Cases #1A-1C: Without TSC HVAC system failures, and TSC HVAC activation times as follows: Case 1A: 9 min Case 1B: 12 min (limiting case) Case 1C: 15 min Note: Continuous occupancy in all cases
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PAGE NO.: [9 NEXT PAGE NO.: 70 PREPARED BY: Af4- DATE 9/svAP/ CHECKED BY: /(( TITLE: Power Uprate Program - DATE .pfa d Technical Support Center /Pontil Accident Radiological Habitability Study Table 5.2 TSC Radiation Exposures Following A Design-Basis Main Steam Line Break (Equilibrium RCS activity at 2 pCi/gm I-131 DE) [3300 acfm filtered (3000+10% uncertaintv)] Time Thyroid Whole Body Skin
. (hours) Dose (rea) Dose frem) Dose (rea) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E+00 3.671E+01 3.913E-02 2.190E-01 8.000E+00 5.328E+01 5.743E-02 3.295E-01 2.400E+01 5.328E+01 6.219E-02 3.609E-01 9.600E+01 5.328E+01 6.332E-02 3.686E-01 !
7.440E+02 5.328E+01 6.332E-02 3.686E-01 I [2700 acfm filtered (3000-10% uncertainty)1 1 Time Thyroid Whole Body Skin (hours) nose fr==) Dose frem) Dose fram) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E+00 3.809E+01 4.029E-02 2.250E-01 8.000E+00 5.875E+01 6.055E-02 3.474E-01 2.400E+01 5.875E+01 6.472E-02 3.757E-01 9.600E+01 5.875E+01 6.572E-02 3.826E-01 l 7.440E+02 5.875E+01 6.572E-02 3.827E-01 l l Note: At the spiked RCS concentration of 2 pCi/gm I-131 DE and the TSC pre-isolated. The duration (time to peak RC concentration) of an .i iodine spike has been observed to last any here from two to more than 12 hours. The above spiked RCS doses are below the GDC-19 limits if the TSC is not occupied for the first hour after the MSLB. l
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-6.- BADIATZ W EEPOSURES FRW A C M TROL RQD DROP ACCIDENT 6.1 Basic Data and Assumptions All me=ptioni 'and ' data employed in the analysis of a CRDA are ;
consistent lwith the guidance in.the Standard Review Plan (Ref. 21, , Sec. 15.4.9 & Ref. 11), applicable portions of Regulatory Guide 1.77 (Ref. 20), the updated UFSAR, and JAF-CALC-RAD-00041 (Ref. 29).- They are as~follows: (a) The reactor has been operating at full power until 30 minutes before the-CRDA. As described in the JAF UFSAR, Sec. 14.6.1.2.3, this assumption means that the reactor was shut down from design power, taken critical, and brought to the initial temperature and pressure conditions within 30 minutes of the departure from design power. (b) The reactor power'was at the level for design-basis accident analyses (i.e., 2586.5 MWt, from Sec. 4.1.1). The core inventory for the' radionuclides of interest at the end of a 1000-day continuous operation is as shown in Table 4.1 of this calculation. (c) A CRDA takes place that leads to fuel failure and subsequent release of 2.53 percent of the core inventory. The fractional inventory released is based on the failure of 850 fuel rods for a standard 8x8 GE fuel, which is evaluated as a bounding fuel design (Ref. 22, Sec. 6.2.1, and Ref. 30, Sec. 3.7). The total number of fuel pins in a GE-8 core (for JAF) is.33,600~(60 pins per assembly x 560 assemblies = 33,600 fuel pins). The fractional core inventory is based on the ratio of the failed fuel pins to the total number. It is noted that although the actual number of failed fuel pins would increase for GE-11 and other fuel types, the total number of fuel rods in the core-also correspondingly increase'. The fractional core inventory that is released is
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Technical Support ! Accident Radiological Habitability Study however bounded by the GE-8 analysis. Furthermore, the JAF . core contains fuel of different fuel types (GE-11, GE-12, j etc.) auut the total numbers of fuel rods vary with core load. ) Use of the fractional core inventory release based on the GE-8 fuel type bounds the other fuel types (Ref. 30, Sec. 3.7). (d) The failed' fuel rods were at a core location with a radial
. peaking factor of 1.5 (Ref. 21, SRP pg 15.4.9-7).
('e) All activity within the gaps of the failed fuel rods is released to the reactor coolant and is instantaneously and uniformly mixed with the coolant in the pressure vessel at a the time of the accident. The released activity is I conservatively assumed to correspond to 10% of all halogens and 10% of all noble gases (30% for Kr 85) in each failed rod (Ref. 18, as. recommended in the SRP). (f) Based on the above-information, and without taking credit for j the pre-accident decay time of 30 min. referred to under' Item (a), the noble gas and halogen inventories which are released to the coolant are as shown below. They were computed by applying the following multiplying factors to the core inventory data given in Table 4.1 of this calculation: Multiplying factor for all noble gases except Kr 85 ; I 1.5 (peaking factor) x (850 failed rods / 33600 rods) { x 0.1 (gap fraction) = 3.795E-03 l Nultiplying factor for Kr 85: 1.5 (peaking factor) x (850/33600) x 0.3 (gap fraction) !
=.1.138E-02 !
Multiplying factor for all halogens: 1.5 (peaking factor) x (850/33600) x 0.1 (gap fraction) x 0.1 (fraction reaching turbine / condensers) = 3.795E-04 i l i
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NEXT PAGE NO.: 74 , PREPARED BY: // M DATE 4/2,%)9 d CHECKED BY: /6' DATE ,ZTr//f TITLE:- Power ~Uprate Program - Technical Support Center 'Pos't - Accident Radiological Habitability Study Post-CRDA Balogens and Noble Gases Reaching the Condenser Activity Activity Nuclide ,igil Nuclide .[gil Br 83 3.066E+03* Kr 83m 3.079E+04 Br 84 5.434E+03 Kr 85m 6.611E+04 Br 85 '6.516E+03 Kr 85 8.874E+03 Kr 87 1.268E+05 I 129 8.554E-04 Kr 88 1.796C+05 I 130 1.027E+03 Kr 89 2.234E+05 I 131 2.582E+04 I 132 3.774E+04 Xe 131m 1.553E+03 I 133 5.400E+04 Xe 133m 2.263E+04 I 134 5.943E+04 Xe 133 5.427E+05 I 135 5.100E+04 Xe 135m 1.023E+05 I 136 2.459E+04 Xe 135 7.009E+04 Xe 137 4.763E+05 Xe 138 4.524E+05 8.078E+06 (from Table 4.1) x 3.795E-04 (g) As a result of. elimination of the MSIV-closure and reactor-shutdown functions of-the main steam line radiation monitors (modification No. F1-93-086) the l pathway of post-CRDA atmospheric releases at JAF has ] changed. Under the new CRDA scenario, the MSIVs stay open and the activity is transported to the condenser. Pathways for release from the condenser are discussed below. (h) With the offgas system operating in the startup mode, two potential release pathways exist: (i) Activity is released via a 30 minute holdup time (conservatively assumed as 21.8 minutes per Ref. 31), EEPA filters and the system flow rate of 229 cfm (Ref. 42). Under this scenario the release is through the main stack (an elevated release) with atmospheric w
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 73 PROJECT: JAF NEXT PAGE NO. f# oJfy/f/ PREPARED BY: MG DATE- 9/M M CHECKED BY: 8 DATE TITLE: Power Uprate Program - Technical Support Center Fost- j Accident Radiological Habitability Study j dispersion factors for transport of activity to the i TSC ventilation intake that are lower by more than four orders.of magnitude. Although the activity release rate is higher then the condenser leakage pathway by a factor of about 200 [ system flow rate (229 cfm) / (condenser volume of 1.51E+5 f t3 front Ref
- 23) / (1% per day x 60 hrs x 60 min.) = 218], the combined effect will yield doses that are significantly lower than the condenser leakage pathway.
(ii) The offgas Hi-Bi alarm is tripped and the offgas system is automatically isolated following a 15 minutes delay. Since the 15 minutes delay is less than the 21.8 minutes transit time, no unfiltered release is assumed under this scenario (UFSAR Sec. 11.4.4.2). Activity is retained in the turbine, condenser and offgas system. Release to the environs is due to leakage from the various contaminated systems into the turbine building. (i) Plateout and partitioning of the halogens in the turbine, condensers and other internal surfaces is conservatively assumed to be equal to 90% [Ref. 21 (SRP pg 15.4.9-7), Ref. 21 (Sec. 6. 3 .1.1) , and Ref. 9]. [ Note: The 90% halogen depletion due to plateout and partitioning was numerically accounted for in the DORITA-2 runs by assuming filtration of the release.] (j) The leakage rate amounts to 1% per day and lasts for 24 hours (Reg. Guide 1.77, Ref. 20). The release to the atmosphere is at ground' level and there is no holdup within the turbine building. (k) Transfer of the released radioactivity to the TSC is governed by the accident dispersion parameters described under Item (i) in Sec. 5.1 of this calculation.
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REV 1 PAGE NO.: 7 #- PROJECT: JAF NEXT PAGE NO.: 75 PREPARED BY: //G DATE $ 4bi CHECKED BY: /// DATE y/u/// TITLE: Power _Uprate Program . Technical Support Center 'Poist- l Accident Radiological-Rabitability Study (1) The TSC characteristics and other exposure-related parameters are as described under Items (i) through (k) in Sec. 4.1.1. 6.2 Results l TSC' radiation exposures following a design-basis CRDA were calculated using the DORITA-2 computer code and the data and I assumptions listed above. Copies of the DORITA-2 outputs appear
~in Attachment A to this calculation (Computer Run Cases #1 and #2).
The DORITA-2 results of interest (namely the thyroid, whole body and skin exposures as a function of post-CRDA time, for continuous occupancy) are summarized in Table 6.1. Refer to Sec. 2 for a susanary of the exposures. 1 I
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TSC Radiatica Exposures Following A Design-Basis' Control Rod Drop Accidant [3300 sefa ' filtered (3000+10% ==aerhaintv)] Time Thyroid M ole Body Skin 1 (hours) Dose frem) Dose (ren) Dose frem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E+00 1.822E+00 1.263E-02 1.303E-01 8.000E+00 7.533E+00 4.582E-02 4.218E-01 2.400E+01 1.621E+01 6.816E-02 6.616E-01 9.600E+01 1.647E+01 6.994E-02 6.765E-01 7.440E+02 1.647E+01 6.995E-02 6.766E-01 [2700 acfm - f: itered (3000-10% uncertainty)1 Time Thyroid mole Body Skin (hours) Dose frem) Dose frem) Dose fram) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E+00 1.822E+00 1.263E-02 1.303E-01 8.000E+00 8.111E+00 4.576E-02 4.216E-01 2.400E+01 1.771E+01 6.789E-02 6.608E-01 9.600E+01 1.804E+01 6.964E-02 6.761E-01 j 7.440E+02 1.804E+01 6.965E-02 6.762E-01 l l Notes Continuous occupancy in all cases TSC' isolation time is 12 min. ll
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- 7. RADIATION EIPOSURES FRCM & REFUELING ACCIDENT 7.1 Basic Data and Assumptions As was the case with all accident analyses documented in this calculation, the computation of TSC radiation exposures associated with a postulated refueling accident was based on data and assumptions consistent with the Standard Review Plan (Ref. 21, Sec. 15.7.4) ~, Regulatory Guide .1.25 (Ref. 18), and the UFSAR.
'Ehey are as follows: '(a) The reactor has been operating at full power (2586.5 MNt) for an extended period of ts.me (1000~ days).
(b)- The core inventory for-the radionuclides of interest at the end of such an operating period is as shown in Table 4.1 of this calculation. (c) The reactor is shutdown, refueling operations are initiated and a. refueling accident takes place at 24 hours after shutdown (Ref. 21, SRP pg 15.7.4-3).. (d) The accident involves a fuel assembly dropping from the maximum height allowed by the fuel handling equipment resulting'in the release of 0.37 percent of the core inventory. The fractional inventory released is based on
- rupturing 125 fuel rods in a standard GE 8x8 fuel assembly, based on information in Ref. 30, Sec. 3.8; also, according to the UFSAR, Sec. 14.6.1.4.2, the total number of fuel rods that fail during a refueling accident is 125. The total >
number of fuel rods in a GE-8 core is equal to 33600 (60 pins per assembly x 560 assemblies = 33,600 fuel pins) . The fractional core inventory release is based on the ratio of the number of failed fuel pins to the total number of pins
-(125/33600 = 0.37). See section 6.1. (c) for discussion on other fuel types.
(e) . The failed fuel rods were at a core location with a radial peaking' factor of 1.5 (Reg. Guide 1.25, Ref. 18).
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,(f) All activity within the gaps of the failed fuel rods is released to the fuel pool water. The released activity is conservatively assumed'to correspond to 10% of all halogens (except I 131) and 10% of all noble gases (except Kr 85) in each failed rod, 30% of Kr 85 (Reg. Guide 1.25) and 12% of the I-131 (NUREG/CR-5009, Ref. 50).
(g) The noble gas and halogen inventories released to the fuel poci (prior to adjustment for decay from the time of reactor shutdown, which is handled by the DORITA-2 computer code) are i as shown in the table which follows. They were computed by multiplying the core inventory in Table 4.1 of this calculation by the following factors: j l Mult.iplying factor for all noble gases except Kr 85: 1.5 (peaking factor) x (125 failed rods / 33600 rods) x 0.1 (gap fraction) = 5.580E-04 Multiplying factor for Kr 85: 1.5 (peaking factor) x (125/33600) x 0.3 (gap fraction)
= 1.674E-03 Multiplying factor for all halogens except I 131:
1.5 (peaking factor) x (125/33600) x 0.1 (gap fraction)
= 5.580E-04 Multiplying factor for I 131:
1.5 (peaking factor) x (125/33600) x 0.12 (gap fraction)
= 6.696E-04
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- 8.078E+06 (from Table 4.1) x 5.580E-04 (h) The halogen composition (inorganic, organic and particulate species)~ and the pool halogen retention factors are such that 99% of all released halogens are assumed to be retained by the water in the fuel pool (Ref. 18). This is equivalent to an overall decontamination factor (DF) of 100. The halogen composition of the remaining (airborne) halogens is equal to 75% inorganic and 25% organic (Ref. 18).
(i) The retention of noble gases by the pool water is negligible (i.e., noble gas DF = 1) .
.( j) The exhaust duct is isolated prior to significant unfiltered release with the' radioactive material being routed through the SGTS. An isolation delay time of 18 seconds (Ref. 27) occurs between the time that a Refueling Accident (RA) is . detected by the radiation monitoring system and releases from the reactor building exhaust to the SGTS and the stack. The reactor building exhaust system ducting provides a 10-second -delay (Ref. 27). .The difference between the system response L___________
NYPA - CALC.# JAF-CALC-RAD-00023- REV 1 PAGE NO. 7f PROJECT: JAF PAGE NO. Yo . PREPARED BY: Nk ' DATE 9/M/Mf CHECKED BY: DATE g/u//f TITLE: Power Uprate Program - Technical Support Center ' Post-Accident Padiological Habitability Study time'of 18 seconds and the 10-second delay implies that an unfiltered release of 8 seconds in duration could result
-(Ref.1). In addition, a potential air exfiltration release from the reactor building has been found to occur during the transition from the normal mode of operation to the isolation mode (Ref. 47). The Reactor Building' air exchange rate was set at the conservative value of 72 air changes per day after isolation.
(') k The halogen-removal filter efficiency of the SGTS is 90% for all halogen species (verified per plant procedures) . (1) The atmospheric dispersion factors associated with the transport of released radioactivity are as follows (in (sec/m') , from Ref. 2]: Interval: 0-18 sec. 18 sec.-8 hrs 8-24 hrs (X/Q). 3.56E-03 9.26E-07 - 6.75E-07 Interval: 1-4 days 4-30 days (X/Q),,, 3.39E-07 1.26E-07 (m) The dilution volume was calculated by taking one-half the distance to the louvers adjacent to those above the reactor vessel, multiplied by a distance extending to 9 feet outside I of the ducts, times the height above the floor. A dilution volume of 101,435 ft8 was thus obtained (Ref. 51). This represents approximately 10% of the reactor building volume, which is two times more conservative than the assumption of , 20% used by the NRC in their Safety Evaluation related to amendment No. 239 (Ref. 52). From Ref. 51, a total of 12 louvers in the refuel floor exhaust system are divided equally among the two exhaust ducts. In this calculation the activity is' assumed to be drawn into the four louvers directly above the core; this provides a dilution factor of
l 1 NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.s 9[ PROJECT: OAF PAGE NO.: 9/ , PREPARED BY: 4(.4- DATE 9/s/MP/ CHECKED BY: DATE f//fArf TITLE: Power Uprate Program - Technical Support Center 7Po4t-Accident Radiological Habitability Study 4/12 =0.33. flow rate of 80,000 cfm was assumed for the exhaust system (Ref. 51) . Therefore, the release rate will be (80,000/101,435). x 60 (min /hr) x 24 (hr/ day) x (4/12) (dilution factor) = 378.6 volumes per day for an 8-second unfiltered release. (n) The maximum possible air exfiltration from the Reactor Building during the transition phase from the normal mode of operation to the isolation mode is 16,628 ft' (Ref. 47) or about 20,000 ft'. Therefore, the release rate will be [20,000 (f t') /2. 6E+6 (RB Vol . f t')] x [60 (sec/ min) /1 (sec)]
.x 60 (min /hr) x 24 (hr/ day) = 664.8 volumes per day for an 1- ;
second exfiltration release. l
3 1 NYPA - CALC.# ,JAF-CALC-RAD-00023 REV 1 PAGE NO. 8/ PROJECT: JAF, ~' 2 2. PREPARED BY: MG- -DATE 4/A W#/ CHECKED BY: NEJ[tT f( PAGE f/pp/ff DATE NO.: TITLE: Power Uprate Program - Technical Support Center / Pc4t'- Accident Radiological Habitability Study 7.2 Results TSC radiation exposures following a design-basis refueling
. accident were calculated using the DORITA-2 computer code and the '. data and assumptions li-* T above. Copies of the DORITA-2 outputs appear in Attanh= ant A to this calculation (Computer-Run Case #1 ,
and case # 2). T'he DORITA-2 results of interest (namely the thyroid, whole body
~
and skin exposures as a function of post-accident time) are summarised in Table 7.1. Refer to Sec. 2 for a summary of the exposures.
-As is.seen, the results for.the RA that are below the GDC-19 limits for the pre-isolated and unisolated n. odes and above the GDC-19 limits for the 12-min.-isolation mode.
l
I NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: T2. PROJECT: JAF T PAGE NO.: Wg
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Technical Support Center gPost f Accident Radiological Habitability Study Table 7.1 TSC Radiation Exposures Fol3owing A Design-Basis Refueling Accident [3306 scea filtered (3000+10% uncertainty)] pre-isolated Time Thyroid Whole Body Skin (hours) Dose fram) Dose (remi Dome (rem) O.000E+00 0.000E+00 0.000E+00 0.000E+00 2.800E-03 0.000E+00 0.000E+00 0.000E+00 5.000E-03 1.764E-02 4.141E-05 5.734E-04 8.000E+00 8.290E+00 1.884E-02 2.648E-01 2.400E+01 8.290E+00 1.888E-02 2.651E-01 9.600E+01 8.290E+00 1.889E-02 2.651E-01 7.440E+02 8.290E+00 1.890E-02 2.652E-01 12 min. isolated Time Thyroid Whole Body Skin (hours) Dose fram) Dose frem) Dose (rem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.800E-03 0.000E+00 0.000E+00 0.000E+00 5.000E-03 2.471E-01 2.277E-04 3.122E-03 2.000E-01 1.893E+01 1.732E-02 2.386E-01 8.000E+00 3.435E+01 3.099E-02 4.296E-01 2.400E+01 3.435E+01 3.100E-02 4.298E-01 9.600E+01 3.435E+01 3.101E-02 4.298E-01 7.440E+02 3.435E+01 3.101E-02 4.298E-01 Not-isolated (24 hrs-decav) Time Thyroid Whole Body Skin (hours) Dose fram) Dose (remi Dose (rem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 l 2.800E-03 0.000E+00 0.000E+00 0.000E+00 5.000E-03 2.471E-01 2.277E-04 3.122E-03 l 8.000E+00 2.181E+01 2.004E-02 2.760E-01 2.400E+01 2.181E+01 2.006E-02 2.762E-01 l 9.600E+01 2.181E+01 2.006E-02 2.762E-01 7.440E+02 2.181E+01 2.006E-02 2.762E-01 l i
1 [.3
~
NYPA - CALC.# JAF-CALC-RAD-00023
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REV 1' PAGE NO.: PROJECT: JAF PAGE NO. W4 PREPARED BY: W 'DATE 9Mdf CHECKED BY: DATE . - TITLE: Power Uprate Program -
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Accident Radiological Habitability Study Table 7.1 (Continued) TSC Radiation Exposures Following A Design-Basis Refueling Accident [2700 sefa filtered (3000-10% uncertainty)] nre-isolated Time' Thyroid whole Body Skin (hours) Dose (ren) Dose (ren) Dose frem) I
~
0.000E+00 0.000E+00' O.000E+00 0.000E+00 l 2.800E-03 0.000E+00 0.000E+00 0.000E+00 I 5.000E-03 1.706E-02 3.629E-05 5.022E-04 8.000E+00 9.158E+00 1.879E-02 2.641E-01 ' 2.400E+01 9.158E+00 1.883E-02 2.644E-01 9.600E+01 9.158E+00 1.884E-02 2.645E-01 7.440E+02 9.158E+00 1 884E-02 2.645E-01 12 min. isolated Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rea) Dose (rem) 0.000E600 0.000E+00 0.000E+00 0.000E+00 2.800E-03 0.000E+00 0.000E+00 0.000E+00 5.000E-03 2.471E-01 2.277E-04 3.122E-03 2.000E-01 1.893E+01 1.732E-02 2.386E-01 8.000E+00 3.655E+01 3.287E-02 4.561E-01 2.400E+01 3.655E+01 3.289E-02 4.563E-01 9.600E+01 3.655E+01 3.289E-02 4.563E-01 7.440E+02 3.655E+01 3.289E-02 4.563E-01 Not-isolated (24 hrs-decav) Time Thyroid whole Body Skin (hours) Dome frem) Dose frem) Dose (ren) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.800E-03 0.000E+00 0.000E+00 0.000E+00 5.000E-03 2.471E-01 2.277E-04 3.122E-03 8.000E+00 2.181E+01 2.004E-02 2.760E-01 2.400E+01 2.181E+01 2.006E-02 2.762E-01 9.600E+01 2.181E+01 2.006E-02 2.762E-01 7.440E+02 2.181E+01 2.006E-02 2.762E-01
NYPA - CALC.# JAF-CALC-RAD-00023' REV 1 PROJECT ,JAF
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PAGE NO.s k j NJET PAGE NO. : J PREPARED BY: M(e- DATE 9/*VMf CHECKED BY: f[ DATE TITLE: Power Uprate Program - Technical support Center - g Accident Radiological Habitability Study J
- 8. RADIATION IIPOSURES PROM EXTERMhL SOURCES ;
1 This'section addresses the post-accident sources of external l radiation affecting the habitability of the JAF TSC, specifically: I (a) -Airborne radioactivity accumulating on the refueling level of the reactor building,
.(b) Overhead radioactive clouds, and (c) Eslogens accumulating on the TSC charcoal filters.
The approach employed'in this assessment was not to provide 1 detailed analysis for each accident scenario, but rather to ' evaluate the worst-case impact of the above sources on the habitability of the TSC. The TSC analysis was limited to post-LOCA drywell leakage for sources (a) and (b) above, and a~MSLB for source (c). For the sake of completeness, radiation fields were determined for
' receptors located within the TSC proper (El. 286'), the General ' Offices-(El. 286'), and the Alternate TSC (El. 300').
The radiation dose rates and cumulative doses documented in this section are to' air.(rad).- They may be conservatively applied to the whole body (rem) . i h
l
\ . Nn X CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO. IS PROO4CT: JAF ~
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- J Accident Radiological Habitability Study.
8'1. . Direct Shine from Post-LOCA Airborne Radioactivity in the Reactor Building
~
8.1.1 Basic Data and Assumpticas Of interest here are the definition of the gasuna spectra associated.with the post-LOCA airborne radioactivity within the refueling level,'and the source / receptor geometry. These are addressed below. Source Term The basic data and assumptions for a LOCA are as describcd in Sec. ) 4 of this calculation. The items associated with definition of the. airborne radioactivity within the reactor building and the associated time-dependent gn=== spectra are as follows: (a)- A LOCA takes place at full power (2586.S MWt) . (b) The core inventory for the radionuclides of interest is shown in Table 4.1. (c) 100% of the core-inventory noble gases and 25% of the halogens become instantly airborne within the drywell atmosphere and are available for leakage to the secondary containment. (d) The halogen composition airborne within the drywell is as follows: 91% elemental, 4% organic.and 5% particulate. (e) Leakage from the drywell is at the rate of 1.5% per day. (f) As a result of the ventilation system, airborne radioactivity leaking from the drywell becomes uniformly distributed within the air volume of the reactor fs building (2.6E+06 f t') . (g) Release from the reactor building is through the SGTS and the main stack at the rate of 6000 scfm (or 3.3 air changes per day) .
IWPA'- CALC.# JAF-CALC-RAD-00023' REV'l PAGE NO.t PROJECT: JAF ~ N p PAGE NO.: 7 '7 PREPARED BY: M DATE 9/#/8 CHECKED BY: N/ DATE g/S//9f TITLE: ' Power .Uprate Program - Technical Support Center 'P64 M Accident Radiological Habitability Study l Source / Receptor Geometry The' general-arrangement of the various buildings in the area of interest and the floor plans at El. 286' and 300' are shown in Figs.18.1 - 8.3 The receptor geometry for use with QAD-CGGP is shown in Figs. 8.4 - 8.6. The primary components and corresponding body numbers:in the QAD-CGGP representation are as
.follows:
Reactor Buildings j El. 272.0' - 367.5' - Bodies 2 and 3 El. 367.5' - 429.5' Bodies 4 and 5 MG Set and Fan Rooms: El. 272.0' - 322.0': Bodies 6 and 7 Administration Building - Cable Room, Relay Room and CR El. 272.0' - 322.0's Bodies 8, 9, 10 and 11 Administration Building - El. 272' exclusive of Cable Room El. 272.0' - 206.0's Bodies 46 and 47 Administration Building - TSC proper / Gen. Offices (El. 286') El. 286.0' - 300.0': Bodies 12-45, 48, 69, 70 and 71 (in part) Administration Building - Alternate TSC (El. 300') El.'300.0' - 322.0' - Bodies 49-68, and 71 (in part)' Outer body - Body 1 The body dimensions are presented in Table. 8.1. Note that the RB concrete floors beneath the refueling level source were conservatively represented by a single 18" concrete slab, and that 7: the Auxiliary Boiler Rm was not included in the geometry. The receptor. locations [11 on El. 286' (the TSC proper), and 4 on
r;- j NYPA - CALC.# JAF-CALC-RAD-00023; REV 1 PAGE NO.: [7 PROJECT: JAF N p PAGE NO.: W~ PREPARED BY: /8f b DATE 9/s#8/ CHECKED BY: A'o' DATE . TITLE: Power Uprate Program - Technical S'pport u Center . o - Accident Radiological Habitability Study El. 300'- (the alternate TSC)] are shown in Figs. 2.5 and 2.6. Note that some'of the receptors are heavily shielded from radiation ==mnating from the RB refueling level, while others have no shielding whatsoever. The effective concrete shield-slab thicknesses protecting the various receptors are as follows: Receptors 1 - 5 (El. 286' - General Offices) : 0" Receptors 6 - 8 (El. 206' - Gen. Off. E part of Lunch area) : 6" Receptors 9 - 11 (El. 286' - Most of Lunch area & Elevator) : 36" Receptors 12 - 15 (El. 300' - Alternate TSC) : 30"
]
The density for poured concrete was set at 2.35 g/cc and had the following composition (in weight percent, from Ref. 44, Vol. II, , Teble 9.1.?2-77): I Fe 1.19 H : 0.85 0 : 50.64 Mg 0.23 Ca: 8.03 Nas 1.66 ' Sit 30.49 Als 4.44 S : 0.12 K :- 1.87 The density for block concrete was set equal to approximately J (2.35/3) = 0.78 g/cc and had the same composition. f l i 1 1
NYPA - CALC.# JAF-CALC-RAD-00023 PROJECT: JAP, REV 1 PAGE NO.s N g PAGE NO.: W
#9, !
PREPARED BY: Afk DATE 9/Af/M CHECKED BY: //J DATE f Sp/Sr f TITLE: Power Uprate Program - Technical Support Center [P64t!- I Accident Radiological ~ Habitability Study l 8 '.~ 1. 2 ' Results The time-dependent gewed spectra needed for the shine analysis to l TSC receptors were extracted from the DORITA-2 run case # 3.in
Attachment:
A.'of~the present calculation. l
. Post-LOCA direct-shine radiation dose rates and cumulative doses ~
at the various receptors identified in Sec. 8.1.1 due to post-LOCA radioactivity accumulating on the RB refueling level are shown in Tables 2.3 and'2.4. These were: extracted from the QAD-CGGP and MATILDA Run Case #1 in Attachment A. Refer to Sec. 2.2 for overall summaries and discussion. l i E o
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO. [/ , PROJECT: JAF ~ PAGE NO.: 90 PREPARED BY Mb DATE 9/49/f/ CHECKED BY: DATE TITLE: Power Uprate Program - Technical Support Center o' - Accident Radiological Habitability Study Table 8.1 Locations and Thicknesses of Bodies Employed in the QAD-CGGP Representation of the RB Refueling Level Source and TSC Receptors Body X-min X length Y-min Y length Elev. Height No. Mat. (ft) ( f t's (ft) (ft) (ft) (ft) 1 1- -2624.7 -5741.5 -3248.0 5872.7 0.0 2952.8 2* 2 0.0 114.0 0.0 151.0 272.0 95.5 3 1- 2.0 110.0 2.0 147.0. 272.0 95.5
'4 2 -5.5 125.0 -5.5 162.0 367.5 2.0 5 1 -5.5 125.0 -5.5 162.0 369.5 60.0 6* 2 124.0 37.0 0.0 130.0 272.0 50.0 7 1 114.0 37.0 0.0 130.0 272.0 49.5 8* 2 151.0 81.5 -66.0 66.0 272.0 50.0 9 1 152.0 78.5 -64.0 64.0 272.0 47.5 10 2 152.0 78.5 -64.0 64.0 285.5 0.5 11 2 152.0 78.5 -64.0 64.0 299.5 0.5 12* 2 151.0 2.0 2.0 142.5 286.0 13.5 13 2 151.0 81.5 0.0 2.0 286.0 13.5 14* 2- 151.0 17.0 48.5 5.0 286.0 13.5 15 2 162.5 5.5 53.5 28.0 286.0 13.5 16* 2 151.0 17.0 81.5 5.0 286.0 13.5 17* 2 230.5 2.0 2.0 166.5 286.0 13.5 18* 2 218.0 14.5 51.0 5.0 286.0 13.5 19 2 218.0 5.5 56.0 25.5 286.0 13.5 20* 2 218.0 14.5 81.5 5.0 286.0 13.5 21 2 179.5 1.0 20.5 74.5 286.0 13.5 22 2 179.5 33.5 20.0 0.5** 286.0 13.5 ;
23 3 212.5 0.5 20.5 25.0 286.0 13.5 i 24 3 180.5 32.5 48.5 0.5 286.0 13.5 25 3 212.5 0.5 58.0 28.5 286.0 13.5 26 3 168.0 0.67 2.0 46.5 286.0 13.5 l 27 3 218.0 0.67 86.5 25.0 286.0 13.5 ' 28 2 153.0 26.5 95.0 1.0 286.0 13.5 29 3 179.5 33.5 95.3 0.67 2'86.0 13.5 1 30 3 153.0 60.0 111.0 0.5 286.0 13.5 j 31 3 179.5 0.5 96.0 15.0 286.0 13.5 ' 32 3 194.5 0.5 96.0 15.0 286.0 13.5 I
l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 96 PROJECT: JAF NEy' PAGE NO. f/ PREPARED BY: & DATE 9/#/ff CHECKED BY: M DATE p//f/ff TITLE: Power Uprate Program - Technical Sdpport Center /Pofst' Accident Radiological Habitability Study Table 8.1 (Continued) Locations and Thicknesses of Bodies Employed in the OhD-CGGP Representation of the RB Refueling Level Source and TSC Receptors Body X-min X length Y-min Y length Elev. Height
.No. Mat. (ft) (ft) (ft) (ft) (ft) (ft)
- 33. 3 212.5 0.5 96.0 15.0 286.0 13.5 34 3 166.0 0.67 111.5 19.5 286.0 13.5 35 3' 153.0 13.7 131.0 0.67 286.0 13.5 36 3 213.0 17.5 115.0 0.5 286.0 '13.5 37 3 213.0 0.5 115.5 33.0 286.0 13.5 38 3 213.0 17.5 148.5 0.5 286.0 13.5 l 39* 3 151.0 0.33 144.5 24.0 286.0 13.5 I 40 3 151.3 67.3 155.2 0.5 286.0 13.5 41 3 166.0 0.5 155.7 12.8 286.0 13.5 42 3 179.0 0.5 155.7 12.8 286.0 13.5 43 3 192.0 0.5 155.7 12.8 286.0 13.5 44 3 205.0 0.5 155.7 12.8 286.0 13.5 l 45 3 218.0 0.5 155.7 12.8 286.0 ' 13.5 46 1 151.0 81.5 0.0 168.5 272.0 13.5 47 -2 151.0 81.5 0.0 168.5 285.5 0.5 48 2 151.0 81.5 0.0 129.0 299.5 0.5 49 2 151.0 81.5 0.0 89.5 319.5 1.0 50 2 151.0 81.5 0.0 89.5 320.5 1.5 51 2 151.0 2.0 1.0 89.7 300.0 19.5 l 52 3 151.0 1.0 90.7 38.3 300.0 19.5 53 2 230.5 2.0 1.0 128.0 300.0 19.5 54- 2 151.0 81.5 0.0 1.0 300.0 19.5 55 3 179.5 1.0 1.0 19.0 300.0 19.5 56 2 179.5 1.0 20.0 46.0 300.0 19.5 57 3 206.5 1.0 1.0 19.0 300.0 19.5 58 2 180.5 50.0 20.0 1.0 300.0 19.5 59 3 214.3 1.0 21.0 29.0 300.0 19.5 60 3 180.5 33.8 49.0 1.0 300.0 19.5 61 3 153.0 12.8 66.0 1.0 300.0 19.5 62 3 174.1 32.4 66.0 1.0 300.0 19.5 63 2 153.0 12.8 84.5 1.0 300.0 19.5 64 =2 174.1 32.4 84.5 1.0 300.0 19.5
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 7/ i PROJECT: JAF ' PAGE NO.: 42_ l PREPARED BY: M[p. DATE 4h#/91 CHECKED BY: DATE TITLE: Power Up rate Program - Technical Suppor' Center g@/,rrh ost-l Accident Radiological Habitability Study l Table 8.1 (Continued) Locations and Thicknesses of Bodies Employed in the QAD-CGGP Representation of the RB Refueling Level Source and TSC Receptors Body X-min X length Y-min Y length Elev. Height No Mat. (ft) (ft) (ft) (ft) (ft) (ft) 65 3 206.5 1.0 66.0 19.5 300.0 19.5 l 66 2 179.5 1.0 78.0 6.5 300.0 19.5 3 179.5 1.0 67.0 300.0 19.5
,67 4.5 68 2 213.8 16.7 84.5 1.0 300.0 19.5 69 1 151.0 11.5 53.5 28.0 286.0 13.5 70 1 223.5 9.0 56.0 25.5 286.0 13.5 71* 1 151.0 81.5 0.0 168.5 286.0 36.0 Material: 1 = Air 2 = Poured concrete (density 2.?s5 g/cc) 3 = Unfilled concrete block (eff. den. 0.78 g/cc)
- Zone Definitions:
Zone 1: Bodies +1 - (2, ... 11) -46 -47 -71
-3 Zone 2: Bodies +2 Zone 6: Bodies +6 -7 Zone 8: Bodies +8 -9 Zone 12: Bodies +12 -69 Zone 14: Bodies +14 -12 Zone 16: Bodies +16 -12 Zone 17: Bodies +17 -70 Zono 18: Bodies +18 -17 Zone 20: Bodies +20 -17 Zone 39: Bodies +39 -12 Zone 71: Bodies +71 -(12, ... 45) - (4 8, ... 70) ** Wall thickness of Body #22 was inadvertently assigned a value of 12" instead of 6" ; has no impact at the receptor locations analyzed.
Note: All office walls were conservatively defined as 6" (instead of 8" ) unfilled block walls.
NYPA - CALC.# JAF-CALC-RAD-00023 -REV 1 PAGE NO.: 92. PROJECT: JAF 93 I My NJ.XT PAGE NO. : PREPARED BY: DATE f/JV/f/ CHECKED BY: /(/ DATE /gy/yf l TITLE: Power Uprate Program - Technical Support Center Postl l Accident Radiological Habitability Study ? t Fig. 8.1 - Building Arrangement in the General Area of Interest j (from JAF Drawing 11825-FA-2G) A
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1 NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 93 PROJECT: JAF NEXT PAGE NO.* *t s/- , l PREPARED BY: ##[, TITLE: Power Uprate Program DATE 9 4 F/9 9 C H E C K E D B Y : /[f Technical Support Center Post-DATE g/,ty/P/ l Accident Radiological Habitability Study Fig. 8.2 - Adminstration Building El. 286' (from JAF Drawing 11825-FA-16A) I
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NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: f[ PROJECT: JAF- NEXT PAGE NO.: ~ 99 . PREPARED.BY: O DATE . 4/Ap/ff CHECKED BY: _ /// DATE 7 w/ff
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TITLES.> Power Uprate Program - Technical Support Center ' Post-Accident Radiological Habitability Study 8.2 -Direct Shine from Post-LOCA Overhead Clouds
-Dose rates and cumulative doses inside the TSC pressure boundary due to gassna radiation emanating from overhead clouds are presented in'the subsections which follow.
8.2.1-' Basic Data and Assumptions Clodd' Gamma Spectra' The post-LOCA' gamma spectra in the released radioactive cloud were extracted from ELISA Run Case #1 in Appendix A of the present calculation. Note that these are for an equivalent semi-infinite cloud with uniform concentration, and were arrived at by making use of-the finite-cloud gamma (X/Q)s (Ref. 28). It should be noted that, the dose rates and doses in Rev 0 of this calculation were very small and negligible in comparison to those due to shine from the refueling' level. Therefore, the gamma spectra and QAD- i
.CGGP input data used in this revision are the same as Rev. O of this calculation.
Source / Receptor Geometry The-QAD-CGGP body geometry employed in this part of the calculation is identical to that described above in Sec. 8.1.1. l The only difference is the source region.- To simplify the analysis, two sets-of QAD-CGGP runs were carried out for each receptor, as follows:
- Part IALsource region covering.the entire space of interest and which includes all solid bodies in the geometry (namely, a box extending from -100 to 600 m along the x and y axes, and from ground level to 517 m above ground).
- Part-IIA source region covering only the MG set and Fan rooms, and the Administration building.
. E-
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO. 8l$ PROJECT: JAF NJtX T P A G E N O . \ea , PREPARED BY: TITLE: set (a. DATE 9/2Mfi Power Uprate Program - CHECKED BY: #[ DATE Technical support Center Godt-
.g/pdf7 Accident Radiological Babitability Study Radiation fields were then computed for the two source cases, and the not. dose rates were obtained by subtracting from Part I the contribution from the assumed airborne radioactivity within the MG set end Fan rooms and the Admin. Building (Part II).
8.2.2 Results I i As noted above, the time-dependent gamma spectra needed for the ! cloud-shine dose analysis of TSC receptors were extracted from the I ELISA-Run Case #1 in Appendix A of.this calculation. A listing of the spectra may be found in the MATILDA output for Run Case #2 in Attachment A.of the present calculation. Post-LOCA cloud-shine radiation dose rates and cumulative doses at the various receptors identified in Sec. 8.1.1 due to post-LOCA drywell leakage are shown in Tables 8.2 through 8.16 for each of the 15 receptors. These were prepared using the results in QAD-CGGP and MATILDA Runs cases #2 - (for Part I) and #3 (for Part II) in Attachment A. The not doses and cumulative doses are equal to the differences between the Part I and Part II entries. Refer to -Sec. 2.2 for overall summaries and discussion.
1 NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: /00 PROJECT: JAF NEfr PAGE NO.: /o / , 1 PREPARED BY />[bL. DATE 9/M( CHECKED BY: A DATE .gaf /f / TITLE: Power Oprate Program - Technical Support Center 4ogt
- Accident Radiological Habitability Study l Table 8.2 l
Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary fram post-LOCA Drywell Releases 1 Receptor #1
~
Dose Rates- (rad /hr) Time Net Total (br) Part'I Part II (I - II) 0.0 4.936E-02 4.517E-03 4.484E-02 l 0.5 1.926E-02 1.790E-03 1.747E-02 L 1.0 1.486E-02 1.397E-03 1.347E-02 1 2.0 1.181E-02 1.130E-03 1.068E-02 8.0 6.525E-03 6.744E-04 5.850E-03 l 24.0 2.086E-03 2.349E-04 1.851E-03 48.0 3.068E-04 4.228E-05 2.645E-04 96.0 1.017E-04 l'.859E-05 8.307E-05 168.0 2.663E-03 4.935E-06 '2.169E-05 240.0 1.714E-05 3.172E-06 1.396E-05 l 744.0 8.748E-07 1.659E-07 7,090E-07 l Cumulative Doses (rad) Time Net Total (hr) Part I Part II (I - II) l 0.0 0.000E+00 0.000E+00 0.000E+00 l 0.5 1.599E-02 1.473E-03 1.452E-02 1.0 2.448E-02 2.266E-03 2.221E-02 2.0 3.776E-02 3.525E-03 3.423E-02 8.0- 9.121E-02 8.822E-03 8.239E-02
.24.0- 1.535E-01 1.549E-02 1.380E-01 48.0 1.758E-01 1.819E-02 1.576E-01 96.0 1.847E-01 1.957E-02 1.651E-01 168.0 1.887E-01 2.031E-02 1.684E-01 240.0 1.903E-01 2.060E-02 1.697E-01 744.0 1.930E-01 2.111E-02 1.719E-01 L
I' l i NYPA - CALC.# JAF-CALC-RAD-00023 PROJECT: JAF REV 1 PAGE NO. /d/ PAGE NO.: /o 2
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TITLE:~. Power Uprate Program - Technical Support Center //Po4t- l Accident Radiological Habitability Study. Table 8.3 Cloud-Shine Dose Rates and Cumulative Doses l Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #2 Dose Rates (rad /hr) Time Net Total (br) Part I Part II (I - II) O '. 0 8.439E-02 9.035E-03 7.536E-02 ) 0.5 3.303E-02 3.646E-03 2.938E-02 1 1.0 2.555E-02 2.877E-03 2.267E-02 2.0 2.039E-02 2.366E-03 1.803E-02 l 8.0 1.150E-02 1.503E-03 9.993E-03 24.0 3.756E-03 5.633E-04 3.193E-03 48.0 5.799E-04 1.187E-04. 4.612E-04 I 96.0 2.081E-04 6.076E-05 1.474E-04 1 168.0 5.478E-05 1.625E-05 '3^.853E-05 240.0 3.526E-05 1.045E-05 2.481E-05
.744.0 1.797E-06 5.436E-07 1.253E-06 Cumulative Doses (rad)
Time Net Total (br) Part I Part II (I - II) ! 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 2.738E-02 2.969E-03 2.441E-02 1.0 4.194E-02 4.593E-03 3.735E-02 2.0 6.482E-02 7.206E-03 5.761E-02 8.0 1.579E-01 1.862E-02 1.393E-01 24.0 2.686E-01 3.394E-02 2.347E-01 48.0 3.095E-01 4.079E-02 2.687E-01 96.0 3.269E-01 4.494E-02 2.819E-01 168.0' 3.351E-01 4.737E-02 2.878E-01 l 240.0 3.383E-01 4.832E-02 2.900E-01 l 744.0 3.440E-01 5.001E-02 2.940E-01 1
l NYPA - CALC.# JAF-CALC-RAD-00023 'REV 1 PAGE NO.: 1o 2. cntOJECT: JAF - N g PAGE NO.: )o3 ' PREPARED BY:' M (p- -DATE R/$FMi CHECKED BY: // DATE 9/gy TITLE:' Power Uprate' Program - Technical Support Center 4ost-Accident Radiological Habitability Study 1 1 Table 8.4 l l Cloud-Shine Dose Rates and Cumulative Doses
~
Within the TSC Pressure Boundary I from post-LOCA Drywell Releases Receptor #3 Dose Rates (rad /hr) ;
' Time Net Total (hr) Part I Part II (I -
II) 0.0 6.891E-02 7.783E-03 6.112E-02 0.5- 2.691E-02 3.108E-03 2.381E-02 1.0 2.077E-02 2.440E-03 1.833E-02 2.0 1.650E-02 1.994E-03 1.450E-02 8.0 9.085E-03 1.242E 7.844E-03 24.0 2.907E-03 4.508E-04 2.456E-03 48.0- 4.3323 '8.745E-05 3.457E-04 96.0 1.469E-04 4.149E-05 1.054E-04 168.0 3.853E-05' 1.106E-05 '2.746E-05 240.0 2.479E-05 7.111E-06 1.768E-05' 744.0 1.268E-06 3.692E-07 8.986E-07 Cumulative Doses (rad) Time- Net Total lhgl Part I Eart II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 2.233E 2.546E-03 1.979E-02 1.0: 3.419E-02 3.926E-03 3.026E-02 2.0 5.274E-02 6 .13 6E 4.660E-02 8.0 1.273E-01 1.567E-02 1.116E-01
- 24.0 2.140E '2.815E-02 1.859E-01 148.0 2.452E-01 3.347E-02 2.118E-01 96.0 2'.579E-01 3.643E-02 2.215E-01 168.0 2.638E 3.809E 2.257,E-01
-240.0- 2.660E 3.873E-02 2.273E-01 744.0; - 2.700E-01 3.988E-02 2.301E-01
7 NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: /d3 PROJECT: JAF . NF PAGE NO.: /d#- 9 PREPARED BY: //ds. DATE 4/21//94 CHECKED BY: DATE TITLE: Power Uprate Program - Technical Support Center f/#pPost~ / Accident Radiological Habitability Study Table 8.5 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #4 Dose Rates (rad /hr) l Time Net Total ! (hr) Part I Part II (I - II) 0.0 6.995E-02 8.214E-03 6.174E-02 l 0.5 2.737E-02 3.308E-03 2.406E-02 l 1.0 2.113E-02 2.606E-03 1.852E-02 2.0 1.679E-02 2.137E-03 1.465E-02 8.0 9.239E-03 1.340E-03 7.899E-03 l 24.0 2.972E-03 4.981E-04 2.474E-03 48.0 4.556E-04 1.040E-04 3.516E-04 96.0 1.623E-04 5.284E-05 1.095E-04 168.0 4.269E-05 1.413E-05 '2.857E-05 240.0 2.748E-05 9.080E-06 1.839E-05 744.0 1.405E-06 4.728E-07 9.321E-07 Cumulative Doses (rad) Time Net Total (br) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 2.269E-02 2.697E-03 1.999E-02 1.0 3.475E-02 4.169E-03 3.058E-02 2.0 5.362E-02 6.533E-03 4.709E-02 8.0 1.294E-01' 1.678E-02 1.127E-01 24.0 2.179E-01 3.039E-02 1.875E-01 48.0 2.501E-01 3.643E-02 2.136E-01
'96.0 2.637E-01 4.006E-02 2.236E-01 168.0 2.702E-01 4.217E-02 2.280E-01 240.0 2.726E-01 4.299E-02 2.296E-01 744.0 2.771E-01 4.446E-02 2.326E-01
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: /47 e' PROJECT: JAF N p PAGE NO.: /of. PREPARED BY: 8(s. DATE 9/f#/4( CHECKED BY: M[ DATE g/gp/f TITLE: Power Uprate Progrs= - Technical Support Center Gos(t / Accident Radiological Habitability Study Table 8.6 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #5 Dose Rates (rad /hr) Time Net Total (hr) Part I Part II (I - II) 0.0 5.615E 7.039E-03 4.911E-02 0.5 2.193E-02 2.805E-03 1.912E-02 1.0 1.689E-02 2.198E-03 1.469E-02 2.0 1.336E-02 1.791E-03 1.157E-02 8.0 7.184E-03 1.100E-03 6.084E-03 24.0 2.262E-03 3.950E-04 1.867E-03 48.0 3.346E-04 7.555E-05 2.591E-04 96.0 1.125E-04 3.538E-05 7.710E-05
~
168.0 2.948E-05 9.426E-06 2.005E-05 240.0 1.897E-05 6.059E-06 1.291E-05 744.0 9.733E-07 3.149E-07 6.584E-07 Cumulative Doses (rad) Time Net Total lhgl Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 1.820E-02 2.301E-03 1.590E-02 1.0 2.785E-02 3.546E-03 2.430E-02 2.0 4.290E 5.533E-03 3.737E-02 8.0 1.126E-01 1.404E-02 8.858E-02 24.0 1. '.' 0 8E - 01 2.505E-02 1.457E-01 48.0 1.930E-01 2.968E-02 1.653E-01 96.0 2.008E-01 3.223E-0? 1.725E-01 168.0 2.092E-01 3.364E-02 1.756E-01 240.0 2.109E-01 3.419E-02 1.768E-01 744.0 2.140E-01 3.517E-02 1.788E-01
E l i NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: MI PROJECT: JAF.. NJXT PAGE NO.: /0 #
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Accident Radiological Habitability Study Table 8.7 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Bc,undary from post-LOCA Drywell-Releases Receptor #6 Dose Rates (rad /hr) Time Net Total (hr) Part I Part II (I - II) 0.0 3.064E-02 6.370E-03 2.427E-02 0.5 1.199E-02 2.556E-03 9.439E-03 l 1.0 9.182E-03 2.004E-03 7.178E-03 2.0 7.145E-03 1.629E-03 5.516E-03 8.0 3.496E-03 9.829E-04 2.513E-03 l 24.0 1.038E-03 3.544E-04 6.839E-04 I 48.0 1.580E-04 7.162E-05 8.638E-05 i 96.0 5.668E-05 3.545E-05 2.123E-05 , 168.0 1.490E-05 9.467E-06 ~5.436E-06 240.0- 9.586E-06 6.085E-06 3.501E-06 744.0 5.011E-07 3.177E-07 1.833E-07 Cumulative Doses (rad) Time- . Net Total (hr) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 9.941E-03 2.088E-03 7.852E-03 1.0 1.520E-02 3.223E-03 1.198E-02 2.0 2.332E-02 5.033E-03 1.829E-02 8.0 5.395E-02 1.271E-02 4.125E-02 24.0 8.635E-02 2.256E-02 6.378E-02 48.0 9'.757E-02 2.681E-02 7.076E-02 96.0 1.023E-01 2.928E-02 7.303E-02 168.0 1.046E-01 3.069E-02 7.387E-02 240.0 1.054E-01 3.125E-02 7.418E-02 744.0 1.070E-01 3.223E-02 7.475E-02 1
NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: /ddf PROJECT: JAF /a ~ PREPARED BY: Af/a_ DATE 9/g81 CHECKED BY:13}EET
/([ PAGE NO.:
DATE TITLE: Power' Uprate Program - Technical Support Center - Accident Radiological Habitability Study Table 8.8 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #7 I Dose Rates (rad /hr) j I Time Net Total j (hr) Part I Part II (I - II) I 0.0 1.123E-02 4.017E-03 7.213E-03 j 0.5 4.418E-03 1.619E-03 2.800E-03 i 1.0 3.376E-03 1.270E-03 2.106E-03 2.0 2.605E-03 1.032E-03 1.573E-03 8.0 1.204E-03 6.174E-04 5.870E-04 24.0 3.483E-04 2.226E-04 1.257E-04 48.0- 5.810E-05 4.598E-05 1.212E-05 96.0 2.406E-05 2.323E-05 8.320E-07 168.0 6.369E-06 6.207E-06 ~1.627E-07 240.0 4.094E-06 3.988E-06 1.054E-07 744.0 2.184E-07 2.096E-07 8.810E-09 Cumulative Doses (rad) Time Net Total ' (hr) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 3.651E-03 1.319E-03 2.332E-03 1.0 5.588E-03 2.038E-03 3.550E-03 2.0 8.562E-03 3.185E-03 5.377E-03 8.0 1.946E-02 8.027E-03 1.143E-02 24.0 3.050E-02 1.422E-02 1.628E-02 48.0 3.438E-02 1.691E-02 1.748E-02 96.0 3.624E-02 1.851E-02 1.773E-02 168.0 3.720E-02 1.943E-02 1.776E-02 240.0 3.757E-02 1.980E-02 1.777E-02 744.0 3.823E-02 2.044E-02 1.779E-02
1 i-NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: /07, PROJECT: JAF NJXT PAGE NO.: /0 7, PREPARED BY: /6/(p- DATE 4hy/P4 CHECKED BY: #[ DATE 7/pp/ff TITLE: Power Uprate Program - Technical Efupport Center 'P66ti ' Accident Radiological Habitability Study l Table 8.9 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #8 Dose Rates (rad /hr) 1 Time Net Total lhrl Part I Part II (I - II) 0.0 9.097E-03 3.925E-03 5.172E-03 l 1 0.5 3.571E-03 1.562E-03 2.009E-03 1.0 2.726E-03 1.217E-03 1.509E-03 4 2.0 2.101E-03 9.803E-04 1.121E-03 8.0 9.679E-04 5.689E-04 3.991E-04 24.0 2.782E-04 1.977E-04 8.056E-05 48.0 4.531E-05 3.761E-05 7.707E-06 96.0 1.816E-05 1.763E-05 5.360E-07 168.0 4.800E-06 4.694E-06 ~1.059E-07 240.0 3.085E-06 3.017E-06 6.870E-08 744.0 1.641E-07 1.585E-07 5.700E-09 Cumulative Doses (rad) Time Net Total (br) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 2.955E-03 1.282E-03 1.673E-03 1.0 4.520E-03 1.973E-03 2.546E-03 2.0 6.920E-03 3.068E-03 3.852E-03 8.0 1.569E-02 7.604E-03 8.089E-03 24.0 2.454E-02 1.322E-02 1.132E-02 48.0 2.762E-02 1.554E-02 1.209E-02 96.0 2.905E-02 1.680E-02 1.225E-02 168.0 2.977E-02 1.751E-02 1.227E-02 240.0 3.005E-02 1.778E-02 1.227E-02 744.0 3.055E-02 1.827E-02 1.228E-02
NYPA- CALC.# JAF-CALC-RAD-00023 EEV 1 PAGE NO.: /O [ Ao 9 , PROJECT: JAF PAGE NO.: PREPARED BY: M/w DATE 4/2@ff CHECKED BY: DATE 9/pp/f/ TITLE: Power Uprate Program - Technical Support Center ' Posit-Accident Radiological Habitability Study Table 8.10 Cloud-Shine Dose Ratss and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #9 Dose Rates-(rad /hr) Time Net Total (hr) Part l Part II (I - II) 0.0 1.256E-02 5.837E-03 6.725E-03 0.5 4.965E-03 2.350E-03 2.614E-03 1.0 3.805E-03 1.845E-03 1.960E-03 2.0 2.954E-03 1.501E-03 1.453E-03 8.0 1.412E-03 9.061E-04 5.054E-04 24.0 4.273E-04 3.290E-04 9.835E-05 48.0 7.732E-05 6.822E-05 9.104E-06 96.0 3.499E-05 3.455E-05 4.380E-07 168.0 9.305E-06 9.234E-06 ~7.140E-08 240.0 5.981E-06 5.934E-06 4.670E-08 744.0 3.161E-07 3.109E-07 5.250E-09 Cumulative Doses (rad) Time Net Total (hr) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 4.092E-03 1.9163-03 2.175E-03 1.0 6.272E-03 2.960E-03 3.311E-03 2.0 9.633E-93 4.627E-03 5.006E-03 8.0 2.217E-02 1.170E-02 1.047E-02 24.0 3.534E-02 2.081E-02 1.453E-02 48.0 4.026E-02 2.479E-02 1.547E-02 96.0 4.282E-02 2.717E-02 1.565E-02 168.0 4.422E-02 2.855E-02 1.567E-02
'240.0 4.476E-02 2.908E-02 1.567E-02 744.0 4.573E-02 3.005E-02 1.568E-02
~ NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: /df PROJECT: JAF .
NEXT PAGE NO.: //O , PREPARED BY: g[s - DATE d/zFAf CHECKED BY: /// DATE 7/p/Py
' TITLE: Power Uprate ~ Prog' ram - Technical su~pport Center Todt-Accident Radiological Habitability Study Table 8.11 Cloud-Shine Dose Rates and Cumulative Doses i Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #10 Dose Rates (rad /hr)
Time Net Total (hr) 'Part I Part II (I - II) 0.0 6.868E-03 6.034E-03 8.340E-04 0.5 2.757E-03 2.434E-03 3.230E-04 1.0 2.155E-03 1.913E-03 2.419E-04 2.0 1.739E 1.561E-03 1.779E-04 8.0 1.011E-03 9.534E-04 5.769E-05 24.0 3.595E-04 3.494E-04 1.017E-05 48.0 '7.407E-05 7.314E-05 9.300E-07 96.0 3.735E-05 3.731E-05 4.000E-08 168.0 9.980E-06 9.974E-06 ~6.000E-09 240.0 6.414E-06 6.410E-06 3.900E-09 744.0 3.358E-07 3.353E-07 4.800E-10 Cumulative Doses (rad) Time Net Total (hr) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 2.252E-03 1.983E-03 2.693E-04 j 1.0 3.474E-03 3.064E-03 4.096E-04 l 2.0 5.413E-03 4.795E-03 6.180E-04 1 8.0 1.347E-02 1.219E-02 1.278E-03 24.0' 2.355E-02 2.182E-02 1.733E-03 48.0 2.788E-02 2.606E-02 1.829E-03 96.0 3.046E-02 2.861E-02 1.849E-03 168.0 3.195E-02 3.010E-02 1.851E-03 240.0 3.253E-02 3.068E-02 1.850E-03 744.0 3.357E-02 3.172E-02 1.852E-03 1 4
NYPA - CALC.#:JAF-CALC-RAD-00023 REV 1 PAGE NO.: //O ' PROJECT: JAF NEXT PAGE NO.: /// , I PREPARED BY: ' M[s- DATE 4h(//(f ' CHECIED BY: //,' DATE ./gp/Ff TITLE: Power Uprate Program - Technical support Center 4Fefdt-Accident' Radiological Habitability Study Table 8.12 Cloud-Shine Dose Rates and Cumulative Doses l Within the TSC Pressure Boundary-from por,t-LOCA Drywell Releases Receptor #11 Dose Rates (rad /hr) Time Net Total (br) Part I Part II (I - II) 0.0 3.747E-03 3.623E-03 1.240E-04 0.5 1.515E-03' 1.469E-03 4.640E-05 1.0 1.192E-03 1.157E-03 3.480E-05 2.0 9.715E-04 9.464E-04 2.513E-05 8.0 5.893E-04 5.825E-04 6.770E-06 24.0 2.167E-04 2.159E-04 8.500E-07 48.0 4.655E-05 4.647E-05 8.200E-08 96.0- 2.427E-05 2.426E-05 9.000E-09 168.0 6.495E-06 6.493E-06 ~2.000E-09 240.0 4.174E-06 4.172E-06 1.300E-09 744.0 2.185E-07 2.184E-07 8.001E-11 Cumulative Doses (rad) Time Net Total (hr) Part I Part IJ, (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 1.232E-03 1.193E-03 3.950E-05
.1.0- 1.906E-03 -1.846E-03 5.970E-05 ~2.0 2.984E-03 2.894E-03 8.940E-05 8.0 7.571E-03 7.393E 1.779E-04 24.0 1.353E-02 1.330E-02 2.270E-04 j 48.0 1.619E-02' 1.595E-02 2.350E-04 l 96.0 1.783E-02 1.759E-02 2.380E-04 168.0 1.880E-02' 1.856E-02 2.380E-04 240.0 .
1.918E 1.894E-02 2.380E-04 744.0 '1.985E-02 1.961E-02 2.380E-04 l l
NYPA'- CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.s /// PROJECT: JAF . NJIT PAGE NO. N1. . I- PREPARED BY: M P DATE 9/2#Af ' CEECKED BY: #/ DATE e/> FAT / TITLE: Power .Uprate Program - Technical S'upport Center '@ oft'- Accident Radiological Habitability Study Table 8.13 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases l Receptor #12 Dose Rates (rad /hr) Time- Net Total l lhg) Part I. Part I.T. (I - II) 0.0 6.047E-03 5.898E-03 1.487E-04 0.5 2.436E-03 2.381E-03 5.570E-05 1.0' 1.916E-03 1.874E-03 4.160E-05 2.0 1.563E-03 1.533E-03 2.980E-05 8.0 9.563E-04 9.491E-04 7.240E-06 j 24.0 3.513E-04 3.506E-04 6.400E-07 4 48.0 7.373E-05 7.366E-05 6.200E-08 96.0 3.766E-05 3.765E-05 1.000E-08 168.0 1.007E-05 1.007E-05 2.000E-09 240.0' 6.471E-06 6.469E-06 1.600E-09 744.0 3.382E-07 3.381E-07 8.001E-11 Cumulative Doses (rad) Time Net Total (hr) Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5. 1.986E-03 1.939E-03 ' 4.730E-05 1.0 3.069E-03 2.997E-03 7.160E-05 2.0 4.802E-03 4.695E-03 1.070E-04 8.0 1.221E-02 1.200E-02 2.090E-04 24.0 2.188E-02 2.162E-02 2.590E-04 48.0 2.614E-02 2.588E-02' 2.650E-04 4 96.0- 2.872E-02 2.846E-02 2.670E-04 L 168.0 3.023E-02 2.996E-02 2 . 67 0.E- 04 l- 240.0 3.081E-02 3.055E-02 2.670E-04 i 744.0 3.186E-02 3.159E-02 2.680E-04 l
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< . 1 -NYPA - CALC.#~JAF-CALC-RAD-00023 REV 1 PAGE NO.: //2 PROJECT: JAF. NEXT PAGE NO.: // 9 .
PREPARED BY: /f /, - D A T E __4 /3,9 4 9 CHECKED BY: /// DATE e/a,Mf TITLE: Power Uprate Program - . Technical Support Center W5st-
' Accident Radiological Habitability Study s Table 8.14-l' Cloud-Shine Dose Rates and Cumulative Doses '
Within the TSC Pressure Boundary i from post-LOCA Drywell Releases Receptor #13 Dose Rates (rad /hr) l Time Not Total _{}gi 'Part I Part II (I - II) l ,
- . 0.0 5.844E-03 5.412E-03 4.320E-04 i 'O.5 2.339E-03 2.173E-03 1.665E-04 l
- 1.0 1.827E-03 1.703E-03 1.239E-04 l 2.0 1.472E-03 1.383E-03 8.860E-05 l l 8.0 8.502E-04 8.291E-04 2.108E-05 1 24.0 2.996E-04 2.980E-04 1.620E-06 48.0 6.052E-05 6.039E-05 1.310E-07 l 96.0 3.001E-05 3.000E-05 1.000E-08 l
~ 2.000E-09 l 168.0 -8.011E-06 8.009E-06
! 240.0 5.148E-06 5.147E-06 1.300E-09 l' 744.0 2.702E-07 2.701E-07 1.'000E-10 l Cumulative Doses (rad) ,, Time. Net Total I-
.(hr). Part I Part II (I -
II_). ! 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 1.914E-03 1.775E-03 1.392E-04 l
-1.0 2.950E-03 2.739E-03 2.114E-04 I 2.0 4.593E-03 4.276E-03 3.167E-04 !
l . i L . 8 .~ 0 - 1.139E-02 1.077E-02 6.170E-04 i l- 24.0 ~ 1.983E-02 1.908E-02 7.590E-04 { 48.0 2.342E-02 2.265E-02 7.740E-04 j 96.0 2.551E-02 2.473E-02 7.760E-04 l 168.0 2.671E-02 2.593E-02 7.770E-04 !
.240.0 2.717E-02 2.640E-02 7.780E-04 l 744.0 2.801E-02' 2.723E-02 7.780E-04 l
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NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.2 83 PROJECT: JAF p PAGE NO.: //sc PREPARED BY: Afh DATE 4/zf/(Y CHECKED BY: K[ DATE .,/a rMa> TITLE: Power - Uprate Program - Technical Support Center Post-Accident Radiological Habf_tability Study Table 8.15 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #14 Dose Rates (rad /hr) Time- Net Total (br) Part I Part II (I - II) 0.0 -6.265E-03 4.795E-03 1.471E-03 0.5 2.521E-03 1.95CE-03 5.714E-04 1.0 1.978E-03 1.541E-03 4.366E-04 2.0 1.605E-03 1.267E-03 3.377E-04 8.0 9.590E-04 7.982E-04 1.608E-04 24.0 3.479E-04 3.014E-04 4.650E-05 48.0 7.291E-05 6.633E-05 6.576E-06 96.0 3.724E-05 3.520E-05 2.0415-06 168.0 9.960E-06 9.427E-06 ~5.329E-07 I 240.0 6.4022-06 6.058E-06 3.432E-07 1 744.0 3.338E-07 3.166E-07 1.724E-08 l Cumulative Doses (rad) Time Net Total ' (br) Part I Part II (I - II) l / 0.0 0.000E+00 'O.000E+00 0.000E+00 1 0.5 2.057E-03 1.581E-03 4.757E-04 1.0 3.176E-03 2.450E-03 7.261E-04 2.0 4.961E-03 3.850E-03 1.111E-03 8.0 1.249E-02 9.939E-03 2.549E-03 24.0 2.213E-02 1.810E-02 4.030E-03 48.0 2.635E-02 2.183E-02 4.527E-03 96.0 2.890E-02 2.419E-02 4.717E-03 168.0 3.039E-02 2.559E-02 4.798E-03 240.0 3.097E-02 2.614E-02 4.829E-03 744.0 3.201E-02 2.712E-02 4.884E-03
l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: 8M PROJECT: JAF NEXT PAGE NO.: //S PREPARLD BY: /9/a_ DATE 4h///M CHECKED BY: /f' DATE 9/afd TITLE: Power Uprate Program - Technical Support Center '@ost 7 Accident Radiological Habitability Study Table 8.16 Cloud-Shine Dose Rates and Cumulative Doses Within the TSC Pressure Boundary from post-LOCA Drywell Releases Receptor #15 Dose Raten (rad /hr) Time Net Total (hr) Part I Part II (I - II) 0.0 5.490E-03 4.385E-03 1.105E-03 0.5 2.210E-03 1.781E-03 4.295E-04 1.0 1.728E-03 1.406E-03 3.214E-04 2.0 1.390E-03 1.155E-03 2.348E-04 8.0 7.943E-04 7.230E-04 7.132E-05 24.0 2.831E-04 2.712E-04 1.191E-05 48.0 6.054E-05 5.901E-05 1.525E-06
'96.0 3.149E-05 3.104E-05 4.450E-07 168.0 8.427E-06 8.311E-06 1.159E-07 240.0 5.416E-06 5.341E-06 7.470E-08 744.0 2.831E-07 2.792E-07 3.890E-09 Cumulative Doses (rad)
Time Net Total (br). Part I Part II (I - II) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 1.802E-03 1.445E-03 3.575E-04 1.0 2.782E-03 2.238E-03 5.440E-04 2.0 4.335E-03 3.515E-03 8.201E-04 8.0 1.072E-02 9.047E-03 1.674E-03 24.0 1.865E-02 1.642E-02 2.230E-03 48.0 2.211E-02 1.976E-02 2.354E-03 96.0 2.424E-02 2.185E-02 2.397E-03 168.0 2.550E-02 2.309E-02 2.414E-03 240.0 2.599E-02 2.357E-02 2.421E-03 744.0 2.687E-02 2.444E-02 2.433E-03
1 1 I NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: //f PROJECT: JAF
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NEXT PAGE NO.: ~ PREPARED BY: AfL DATE 9/268)9 CEECKED BY: /// DATE TITLE: Power Uprate LProgram - Technical Support Center Pofs t'- Accident Radiological Habitability Study 8.3 Direct Shine from Balogens Accumulating on the TSC Charcoal Filters The Tsc charcoal filtration system is lucated in the Administration Building, on El. 300' (see Fig. 8.7). Dose rates and cumulative doses from halogens accumulating % the charcoal filters were computed for the worst-case accident scenario for this source' term, namely a'MSLB. Analytical details and results are addressed in the subsections which follow.
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.-NYPA'. . CALC.# JAF-CALC-RAD-00023 REV 1. PAGE NO. // / . PROJECT: C_7 ~
' N y PAGE NO.: // -7,
. PREPARED BY: 14 /s. - DATE 4/AFM 9 ' CHECKED BYa X[ DATE gar /Ff TITLE: Power Uprate Program , Technical' Support Center Podf4 Accident. Radiological Habitability Study 8.3.1 , Basic Data and Assusptions Filter Gamma Spectra The. data and acuumptions associated with the buildup of halogen
- radioactivity.on the TSC charcoal filters are identical to those presented in Sec. 5.1 for a.NSLB. Refer tio that section for ,
details.- A new DORITA-2 computer run was carried out to. expand the time array for the sake of improving the accuracy of the-n'userical integration for the calculation of the cumulative doses. Source / Receptor Gecentry Ths QAD-CGGP source / receptor geometry for computation of the l ~ direct shine radiation fields from the TSC charcoal filters is shown in Fig. 8.7. The filter dimensions were extracted from Ref. 46 (Sec. 4.2.4), and the filter casing was ignored. The density and composition of the concrete floor are as given in.Sec. 8.1.1. L Two receptor locations were considered,'as follows: l (a) Worst-case receptor on El. 286' of the Administration Building (between . Receptors #4 and #6 in Fig. 2.5) , and (b) At tho' filter casing (El. 300'). Receptor'(b) is outside-the TSC pressure boundary and was selected for informational purposes only.
p t l l NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: < PROJECT: JAF DATE 4/g#A4 CHECKED BY: // NEXT PAGE NO.: //// [7 I Centerg/1r/f PREPARED BY: M( DATE l TITLE: Power Uprate Program - Technical Su'p port /PoVt / ! Accident; Radiological Habitability Study Fig. 8.7 - The TSC Charcoal Filtration System (From JAF Drawing 11825-FB-32G) l 1l i*
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i l NYPA - CALC.#'JAF-CALC-RAD-00023 REV 1 PAGE NO.: //[ . PROJECT: JAF PAGE NO.: //9 PREPARED BY: M/r DATE 9/2F/99 CHECKED BY: N X p[ DATE 9/pp/7/ TITLE: Power Uprate Program - Technical SOpport Center Most'- Accident Radiological Habitability Study Fig. 8.8 - QAD-CGGP Source / Receptor Geometry ' (Direct Shine from the TSC Charcoal Filters) l l 1
= axis Charcoal filter (24"W, 30"L, 6.25" H, from Ref. 45, Sec. 4.2.4) , Receptor #1 l 6" Concrete ' Floor '
i El.304'- 10" I l i i - El. 300' I j e Receptor #2
<H El. 292' (x=0, y=0) ;
i I
7 'NYPA - CALC.# JAF-CALC-RAD-00023 REV 1 PAGE NO.: //8
. PROJECT: JAF NEXT PAGE NO.: A/ue l PREPARED BY: & DATE 9/J#/19 CHECKED BY: /[ DATE- 9/>p/ @ost y Technical Support Center - TITLE s - Power Uprate Program -
l Accident Radiological Habitability Study l 8.3.2 Results Direct shine radiation levels due to gamma radiation emanating from the TSC charcoal filters following a MSLB were determined though use of the gamma spectra extracted from DORITA-2 Run Case # 4, and the QAD-CGGP geometry described in Sec. 8.3.1. Summaries of the results, extracted from QAD-CGGP & MATILDA Run Case #4 are presented in Tables-2.7 and 2.8 in Sec. 2.2.3. Refer to that s'oction for details and discussion. I i
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l NYPA/CRE CALCULATION No. JAF-CALC-RAD-00023, Rev. 1 L POWER'UPRATE PROGRAM - TECHNICAL SUPPORT CENTER POST-ACCIDENT RADIOLOGICAL HABITABILITY STUDY l ATTACHMENTS A.- COPIES OF COMPUTER OUTPUTS Included in this attachment are copies of the computer outputs pertinent to this calculation. They appear in the following orders. DORITA-2 l Case'#1 TSC radiation exposures fort (a) Loss of coolant accident (LOCA) (drywall leakage) , (b) Loss of coolant accident (LOCA) (ESF Component leakage) (c) Main Steam Line Break accident (MSLB), with TSC HVAC activation'at 9 minutes after the accident (d) Main Steam Line Break' accident (MSLB), with TSC HVAC activation at'12 minutes after the accident l (e) Main Steam Line Break accident (MSLB), l with TSC HVAC activation at 15 minutes after the accident (f) Main Steam Line Break accident (MSLB), 2 pCi/gm with TSC pre-isolated ' (g) Control Rod Drop Accident (CRDA), (h) Refueling Accident, pre-isolated TSC j i (i) Refueling Accident, 12 min. Isolated TSC ; l _(j) Refueling Accident, not-isolated TSC ! 4
, TSC ventilation system operating conditions: I Pre-activation flow 25800 sofm Activation time (LOCA/ESF/CRDA) : 1 hr Activation time (MSLB, 3 cases): 9, 12 & 15 min Post-isolation filtered flows 3300 seta c Post-Isolation unfiltered flow: 1500 scfm Charcoal filter efficiency: 90 %
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NYPA/CRE CALCULATION No. JAF-CALC-RAD-00023, Rev. 1 POWER UPRATE PROGRAM'- TECHNICAL SUPPORT CENTER POST-ACCIDENT RADIOLOGICAL HABITABILITY STUDY Case #2 -TWC radiation exposures for: (a) Loss of coolant accident (LOCA) (drywell leakage) (b) Loss of coolant accident (LOCA) (ESF Component leakage) (c) Main Steam Line Break accident (MSLB), with TSC HVAC activation at 9 minutes after the accident (d) Main Steam Line Break accident (MSLB),
^
with TSC HVAC activation at 12 minutes after the accident (e): Main Steam Line Break accident (MSLB), with TSC HVAC activation at 15 minutes after the' accident (f) Main Steam Line Break accident (MSLB), 2 pCi/gm with TSC pre-isolated (g) Control Rod Drop Accident (CRDA),. (h) _ Refueling Accident, pre-isolated TSC i (i) Refueling Accident, 12. min. Isolated TSC (j) Refueling Accident, not-isolated TSC .(24 hrs decay) TSC ventilation system operating conditions: Pre-activation flows 25800 scfm Activation time (LOCA/ESF/CRDA): 1 hr Activation time (MSLB, 3 cases): .9 , 12 & 15 min Post-isolation filtered flows 2700 scfm Post-Isolation unfiltered flow: 1500 scfm Charcoal filter efficiency: 90 % Case #3 Gamma spectra associated with post-LOCA airborne radioactivity accumulating in the Refueling Level of the Reactor Building (for use with QAD-CGGP and MATILDA to_ compute the direct shine radiation levels in the TSC). Case #4 Gamma spectra associated from the TSC charcoal filter following a MSLB.
NYPA/CRE CALCULATION No. JAF-CALC-RAD-00023, Rev. 1 I POWER UPRATE PROGRAN - TECHNICAL SUPPORT CENTER i POST-ACCIDENT RADIOLOGICAL HABITABILITY STUDY ELISA Case #1 Post-LOCA ganana spectra associated with the released radioactive cloud-(for use with QAD-CGGP and NATILDA) 5 QAD-CGGP &'NATILDA case #1 Direct shine radiation levels from post-LOCA radioactivity accumulating in the RB Refueling Level - 11 receptors on El. 286'and 4 on El. 300' Case #2 Direct shine radiation levels from post-LOCA radioactivity in the released cloud - Part I (entire space, assuming cloud has penetrated the TSC and adjacent areas) - Eame 15 receptors as in case #1. Case #3 Similar to Case #2, but Part II - source region covering only the TSC and' adjacent areas (Note: Net radiation field from external cloud
= Part I - Part II)
Case #4 Direct shine radiation levels from halogens accumulating on the TSC charcoal filters (based on the gasuna spectra defined under DORITA-2 Case #4) (2 receptors: contact with casing, and on El. i 286', between Receptors 5 and 6 in Case #1.
)
5 For - the sake of reducing the size of Attachment A, only selected output pages.have been retained for Run Cases #1, #2 and
#3; . these include the entire inputs and the results for each receptor location.
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