ML20247G064
| ML20247G064 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/29/1997 |
| From: | Golshani M POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20247F730 | List:
|
| References | |
| [[::JAF-CALC-RAD|JAF-CALC-RAD]], JAF-CALC-RAD-00, JAF-CALC-RAD-00042, JAF-CALC-RAD-42, NUDOCS 9805200124 | |
| Download: ML20247G064 (128) | |
Text
{{#Wiki_filter:.. g,y DESIGN VERIFICATION COVERSHEET page 1 of 1 0 Verification of: DocumentTitle: cnntroi nonm naa4nine4ca' naki+ah4'4+v Under Poder Uprate Conditions .nd CREVASS Reconfiguration Document Number: m -calc-Rao-ooo42, nev. 2
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O ELECTRICAL MECHANICAL INSTRUMENT & E)NTT40L CML/ STRUCTURAL RRE PROTECTION SIMULATOR Radiologi cal /h ' O_ 9805200124 980226 PDR ADOCK 05000333 P PDR DcM 4 l Rev. No. 3
)^Oq DESIGN VERIRCATioN ATTACHMENT 4.1
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f DEStGN VERlRCATION CHECKLIST Q . JAF page 1 of 4 rh IDENTIRCATION: DISCIPUNE:
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- DocurnentTitle:
Control Room Radiological Habital igtD (} l&C l Under Power Uprate Conditions and CREVASS [] MECH (} RreProtect Recon 11guration JAF CALC RAD 00042 Rev. , () C/S Simula Doc. Number. Doc. Revision: - N Other.LSPec'fyl" h af A I QA Category. METHOD OFVERlRCATION: N DesignReview [] Altemate Calculations [] Qualification Test Selected Verifiet, MN 00WkW HM. Gen 0< pnnt nerne.oeparureent_, phone es:.
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(/ g 8 Design verification Questionnaire All questions shall be explained in the space provided.
- 1. Were the inputs correct and incorporated into the desegn? eMca/Mf1 Explanation: 1,,Ps/s ial;;' [ fir fi:;,,, A ) I {3ar rnit - 2AD Ocs 9 2- bdI
- 2. Are the physical and functional characteristics of the proposed design within the approved design basis of the system (s) structure (s) or component (s)? ///j) exo!anation: a lc).L, s L.,.tsj t < k,41& - J n b h.id, t pQ "
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- 3. Does the proposed design incorporate license Commitments? g//)
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- 4. Are assumptions necessary to perform the MEty adequateb/ desenbed and reasonable: Where necessary, are the assumptions identified for subsequent reverification when the detailed design activities are completed? $
Explanation: -f"/e ads,.//M sid %[,., fr;,,, ,/.,,,) S Are the appropriate quality and quality assurance requirements specified? e.g safety classification? hk I 5. Explanation: Arr ch ) }li senwl1V$ NC,,JJ.4 J/,) /LW sar;r lelow //u e
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- 6. Are the applicable codes, standards and regulatory requirements including issue and addenda properiy identified ang! are theiy requirements for design met? Q ,
Explanation: 8 (Yddhd & $_'&) ./h ~)b u,ff ltk. J Ar 291slA/G lN i 8 // I v J i l DCM 4 DESIGNVERIFICATION ATTACHMENT 4.2 Rev. No. 3 Page 9_, of _21_
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DEstGN VERIFICATION CHECKUST /O page 2 of 4
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- 7. Have applicable construction and operating experience been considered ( /t//A. ,
Explanation: Al e%,pp$ewj//? fy- (%;//Nir //,/r[g/,'
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- 8. Have the design interface requirements for mechanical, electrical /l&C. and cM1/ structural engineering been satisfied? .g/g ,
, o Explanation: G h / M b '/'ef e/ / N M M 4
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- 9. Wasthe appropriateMe'f$od used? jef Explanation: As2.) ori /b F;/?'$) y ci, , SAN lb -
- 10. Is the output reasonable compared to inputs? M Explanation: BM off geg / ,/ & M y
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- 11. Are the specified parts, equipment and processes property suited for the fire protection Appendix R. QA, and g
EQ classifications required for the application? Q Explanation- IL9.) off A d, / ,,) 2_
- 12. Are the specified materials compatible with each other and the design environmental conditions to which the materialwillbeexposed? Q _
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. Explanatiore 4 .Le.; J lah,1+ ./La n Jtc1.wcd "
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- 13. Have personnel requirements and limitations for maintenance. testing, and inspection been satisfied? g//)
Explanation: A w k M . 9 eob' _ M '
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- 14. Are accessibility, maintenance repair, and inservice inspection requirements for the plantincluding the plant conditions under which these will be performed been consi,dered? g Explanatiore %.4,,)ik $c.ft of/45 (2kbfiis.
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- 15. Has adequate accessibility been provided to perform the inservice inspection expe::ted to be required during q
the plantlife? g/g Explanation: Qc%) h/f fa.jfr_ / M3 C/MM/. G I f' h - - - - - _ _ _ mmm_ _ e h
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#5 PDwer I DESIGN VERlRCATION CHECKUST l page 3 of 4
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1 All questio 1s shall be explained in the space provided
- 16. Has the design properly considered radiation exposure to the public and plant personnel? (ALARA/ cobalt reduedon) MM Explanation: Ak 5) f' -/4 f,ep &,,M/foorrt, .*V/ e /AV'w c di'V; C u
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- 17. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have satisfactorily accomplished? ,fg -
Explanation: bkr,&af d' /L d.m //o~,,, isJie/,1,Lf1,R /s///,'C
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- 18. l Have adequate preoperational and subsequent periodic test requirements been appropriately specified? p/' 4 Explanation: aowJ ,ck S't'We d //b Ge{d.d3bh. -fd<.' cc.,ff LA)., s 'a /*,., 4,/
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- 19. l Are adequate handling, storage, cleaning and shipping requirements specified? ,,r/A Explanation: a, t e L J L.s Ec,,P c 4 .(LIA @c8=.fru .
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- 20. l Are adequateidentification requirements specified? _ M g. i P Explanation: % -t..,i Akt 9c,,pc J & 01/cSd%,,.
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- 21. Are the conclusions drawn in the Safety Evaluation fully supported by adequate discussion in the test or Safety Evaluationitself? s /A _ m .
I Explanation: \ tm.h, ,i h .4b h,, p Co. arm,rece fdi'okrti<3 haLdvliI C. y
- 22. l Are necessary procedural changes specified, and are responsibilities for such changes clearfy delineated? d/A -
Explanation: 'b k 1 A Scac J L edIdh'm rM2 c./j4,_ syt/ah ~
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- 23. l Are requirements for record preparation, review, approval, retention etc. adequately specified? g//).
Explanation: % .(, ,,1 4k.Chp,,,1 JLfA OdIdbJr . u * *
- 24. Have supplemental reviews by other engineering disciplines (seismic, electrical, etc.) been performed on the integrated design package? ,.r/A n ,
Explanation: -f]' h m//c).f, a)Wc-Lh .sL- c..MN rec ,,. <n.h'kMc.Y L l ')H,, IrV l AE
- 25. l Have the drawings sketches. calculations, gtc.. included in the integreted design package been reviewed? ,v/g s/n t a & ,, .s c /t Vf l . *
& rde$/a., C>ppf/A N p) q_
Explanation: n nsp4&wMr
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G i DCM 4 DESIGNVERIFICATION ATTtDWENT 4.2 l Rev. No.3 Page 11 of _211 j
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# Design Verification Questionnaire A!! questions shall be explaired in the space provided
- 26. Have reviews been performed to identily any offeet on the Check Valve Maintenance Program? g//)
Esplanatlon: f& c)./),, aw/8 jL 66/r;(in J"Y' hlrirYLldhUh a),/~~~ Aw Mr t"~er ),h.h . 7+-) J . b > 'Aed**"J9 l 1 st./h d m-u knen AG ftGrt.
- 27. Does the design ior check valves n" wet the intents of INPO SOER 86 037 //A Explanation: --fW5 e<}QW ,:S,,,/---fei s .)lsy,[4 ./Aa dj yj, rrL L .L . n , f.
f) G ,r % T r,Y7 ) $ b a $ V d $ fr. cheSc% .
- 28. Is the plant reference simulator physical and functional fidelity offected and it's design change been f actored / into the co Explanation: 85 cj/J,yg & ll,j / } ff,c:,s d g,.,,r,,f,/n,) k,]),f, & i G
- 29. Are all references 9nsd (including design calcatationlanalysis) that wers used as part of the design rey'sw?/4 /g Explanation: d W W ir ar2Sk1 ( llln/rl Y r ,d ),,ra-,..J.d;-
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i i .i I U REMARKSICOMMENTS: f I ? l i [ DCM 4 DESIGN VERIFICATION ATTACHMENT 4.2 l Dev w 1 Paoe 12 of 21_
COMPUTER CALCULATION CONTROL SHEET Page 1 of 110 plus attachments
- CALC. NO. JAF-CALC-RAD-OOO42 REV. 2 IP3 JAF MOD / TASK HO.
QA CATEGORY OF CALCULATION: I CA!CULATIOTAL TYPE: PRELIMINARY: FINAL: X PROJECT / TASK: l SYSTEM NO./NAME: ! TITLE: Control Room Padioloolcal Habitability Under Power Uprate Conditions and CREVASA Reconfiguration _, PREPARER: A. Ramacha van 9M CHECKER: c.c. ne* hu 6 47 /n/MN7 VERIFIED: N/A O M. coisaan4 W // N ' ?n/ar/o 7 APPROVED: o.c. ne* ~ f/L f f', Ii ,L 7,9'f 7 sv- ; 7 PROBLEM / OBJECTIVE / METHOD See pages 2, 11-12, and 27-29
,C) DESIGN BASIS / ASSUMPTIONS V
See pages 30-110 , i
SUMMARY
/ CONCLUSIONS See pages 13-26 REFERENCES .
See pages 3-7 I AFFECTED SYSTEMS / COMPONENTS / DOCUMENTS This calculation supersedes JAF-CALC-RAD-00042, Rev. 1 in its antirety r ! O VOIDED D SUPERSEDED BY: - J (CAI4 NO.) Supersedes those portions of JAF-CALC-RAD-00008 which deals with the Control Room doses. NYPA FORM DCM-14, ATTACHMENT 4.1 (REVISION 1) Page 1 of 1
i i NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 2_ OF 11 O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/3c/97-Q TITLE: FINAL [X] CHECKED BY // DATE /of,j/f7 Control Room Radiological Habitability Under Power Ifprate _ Conditions and CREVASS Reconfiguration Statement Of Problem This calculation updates the control room radiological l habitability analyses documented in Reference 1 (JAF-CALC-RAD-
~
00042, Rev. 1) and provides a consolidated document of the , assumptions, scenarios and results for control room habitability t under various postulated design basis accidents. This calculation specifically addresses the concern identified under ACTS Item 23847 (Ref. 2): a) evaluation of the loss of coolant accident (LOCA) and refueling accident (RA) assuming a lowered stand-by gas treatment system (SGTS) charcoal filter efficiency (assumed efficiency of 90% for halogens). l p ' l In addition, the present calculation incorporates the following changes to the assumptions and methods employed in Reference 1: _ b) revision of the atmospheric dispersion f actors for elevated releases as documented in Reference 5, and c) use of the ICRP 30 (Ref. 6) dose conversion factors for the I determination of thyroid doses. Calculation Use Limitations , This calculation was developed to address the specific issue (s) described in the above Statement of Problem. Information provided in this calculatiran 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 l Documents, Licensing commitments, design drawings, etc.) , as appropriate. - If this calculation is a nd 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. O .. I l i
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 3 OF //O PROJECT: JAF PRELM [] PREPARED BY 89. DATE 9/MF FINAL [X] CHECKED BY _jf DATE ,m/,_,/f 7 TITLE: Control Room Radiological Habitability Dnder Power dprate ~ Conditions and CREVASS Reconfiguration i References l l 1. JAF-CnLC-RAD-00042, Rev. 1, " Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration". -
- 2. ACTS Item No. 23847, " Revise Calculations Which Use SBGT Efficiency From 99% to 95%".
- 3. ACTS Item No. 23842, " Formalize Main Steam Line Break Calc at 2.0 Micro-Curies per Gram I-131 Dose Equivalent to Accommodate an Iodine Spike".
- 4. US NRC NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" _
- 5. JAF-CALC-RAD-00007, Rev. 2, " Power Uprate Program - Onsite and offsite Post-Accident Atmospheric Dispersion Factors".
- 6. International Commission on Radiological Protection (ICRP)
Publication 30, " Limits for Intake by Workers" (Various l p Parts and Supplements, 1979-1982)
- 7. CRE Calculation JAF-CALC-RAD-00008, " Radiological Consequences of Design Basis Accidents at JAF" (11/27/91) _
- 8. NYPA Memorandum addressed to J. Hamawi, B. Young, and R.
Young, from D. Lochbaum, titled " Design Document Open Items on JAF Control Room Ventilation System DERs" (2/28/94) [See JAF-CALC-RAD-00028 (Ref. 18) for a copy of this reference.]
- 9. SWEC letter J.O. 04508.24 addressed to A. Ettlinger, from H. Wessinger, titled "NYPA/FitzPatrick DBD Development Program - Control Room and Relay Room Ventilation and _
Cooling Systems (CRRRVAC) DBD - Revision A for NYPA Review" (2/16/94) (DDOI-JAF-CRRRVAC-070-21 and DDOI-JAF-l CREVASS-070-021)
- 10. NYPA Memorandum No. NED-M-93-WP51-RY922, addressed to J.
Lazarus, from R. Young, titled "JAFNPP - Control Room Ventilation System Normal Outside Air Intake" (11/9/93) [See JAF-CALC-RAD-00028 (Ref. 18) for a copy of this l reference.] _
- 11. NYPA Memorandum No. NED-M-93-WP51-RY923, addressed to J.
Lazarus, from W Young, titled "JAFNPP - Control Roo... Ventilation System Normal Outside Air Intake" (11/9/93) [See JAF-CALC-RAD-00028 (Ref. 18) for a copy of this reference.] Os 12. Nucle;ar Safety Evaluation JAF-SE-94-042, " Revision of FSAR Section 11.5.3.9 & 14.8.1.5, Return of Control Room L
I NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /4 OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/)p M FINAL [X) CHECKED BY /U DATE fo/,/f7
'~i TITLE: Control Room Radiological Habitability Under Power dprate Conditions and CREVASS Reconfiguration l Ventilation System to Normal Mode of Operation Following 1994 Maintenance Outage" (4/20/94) [See the Notes section retained by CRE for JAF-CALC-RAD-00028 (Ref. 18) for a copy of this reference.)
- 13. NYPA Memorandum CM-DBDM-94-046, addressed to D. Holliday, from N. Mathur, titled "JAFNPP - DDOI-JAF-CRRRVAC-070-021 Resolution" (3/24/94) [See JAF-CALC-RAD-00028 (Ref. 18) ~
for a copy of this reference.] l 14.
- Stone & Webster Engineering 'culation No. 12966-PE(N)-
019-0,"High Energy Line Br nalysis in the Turbine Building for Class IE Electraval Equipment Qualification in Response to IE Bulletin 79-01B" (6/9/81) 15 .
- GPU Nuclear Corporation letter 5450-95-0006, addressed to M. Karasulu, from N. G. itikouros, titled "FitzPatrick Nuclear Plant Turbine Building HELB Analysis Results" -
(2/17/95)
- 16. MOD F1-93-86: MSIV Closure and Scram Function 3 17. 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)
- 18. CRE Calculation JAF-CALC-RAD-00028, Re.v. O, " Control Room Post-Accident Radiological Habitability - Assessment of _
Current Ventilation System Configuration" (4/19/94)
- 19. 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)
- 20. US NRC Regulatory Guide 1.5, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors" -
(March 1971)
- 21. 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)
- 22. US NRC Regulatory Guide 1.49, " Power Levels for Nuclear Power Plants" (Rev. 1, December 1973) -
- 23. US NRC Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized
( . 24. Water Reactors" (May 197^) US NRC Regulatory Guide 1.52, " Design, Testing and
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE f OF _/ / O PROJECT: JAF PRELM [] PREPARED BY /hC. DATE 9/3c[1')- FINAL [X] CHECKED BY // DATE ,M.2/f7
^
TITLE: Control Room Radiological Habitability'Under Power Dpr' ate Conditions and CREVASS Reconfiguration Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorpt.on Units of Light-Wster-Cooled Nuclear Pr"er i Plants" (Rev. 2, March 1978)
- 25. 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," RAD-001, Release 1.3.1.2 (8/17/92)
- 26. CRE Computer Code ELISA, "A Computer Code for the
, Radiological Evaluation of Licensing and Severe Accidents l at Light-Water Nuclear Power Stations," RAD-005, Release 1.3.1.1 (3/9/92)
- 27. CR" Computer Code QAD-CGGP, "A Combinatorial Geometry Version of QAD-P5A, A Point Kernel Code System for Neutron l and Gamma-Ray Shielding Calculations Using the GP Buildup Factor," RAD-006, Release 1.3.1.1 (3/26/92) _
- 28. CRE Calculation IP3-CALC-RAD-00002, " Post-Accident High-Range Containment Radiation Monitor Response", 12/2/1993.
- 29. 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 Final Reactor Heat Balance" (2/11/91) [See JAF-CALC-RAD-00004 for a copy of this reference.]
- 30. GE letter addressed to Richard Chau, NYPA, from C. H.
l Stoll, GE Plant Performance Engineering, titled "J. A. ! FITZPATRICK (JAFNPP) Power Uprate Program - Formal l Transmittal of Final Source Term Analysis Results" (5/2/91) (See JAF-CALC-RAD-00008 (Ref. 7) for a copy of this reference.] l 31. US NRC Regulatory Guide 1.96, " Design of Main Steam l Isolation Valve Leakage Control Systems for Boiling Water - Reactor Nuclear Power Plants" (Rev. 1, June 1976)
- 32. 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) [See JAF-CALC-RAD-00008 (Ref. 7) for a copy of this reference.]
- 33. NYPA Letter JPN-96-055 addressed to the NRC, titled "JAFNPP - Additional Information Regarding Analyses at -
() 34. Power Uprate Conditions" (12/23/96) Stone & Webster Engineering, Calculation CC-70-04, c [____-__--___ _ _ _ _ _ _ . ___ _ _ _ __ __ _ - . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - _ _ _ _ - _ _ _ - _ _ - _ - _ _ - _ _ -
r-NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE S OF //O PROJECT: JAF PRELM [] PREPARED BY /81C. DATE 9/Jo/D FINAL [X] CHECKED BY // DATE /o/f 7/f 7 TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration
" Calculation for Air Conditioning System Cooling Load" (9/2/70) (DSR #249252)
- 35. Johnson Service Company, Test Report TLP-774-448 (02-4925- -
72), "FitzPatrick NPP Damper Leakage (D-1300 Series) t (11/29/72) (NYPA Microfiche No. 60067, frames 025-030) {See JAF-CALC-RAD-00028 (Ref. 18) for a copy of this reference.]
- 36. J. DiNunno, F. Anderson, R. Baker and R. Waterfield,
" Calculation of Distance Factors for Power and Test Reactor Sites," AEC, Division of Licensing and Regulation, TID-14844 (March 1962)
- 37. K. G. Murphy and K. M. Campe, " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19", 13th Air Cleaning Conference, p. 401-430 (1974)
- 38. JAF Emergency Plan Implementing Procedure EAP-44, " Core Damage Estimation" (Rev 4, 2/14/97)
- 39. NYPA letter JPN-94-059, addressed to the US NRC, titled -
" James A. FitzPatrick Nuclear Power Station - Docket No.
50-333 - Response to NUREG-0737, Item III.D.3.4, Control Room Habitability" (11/16/94) [See the Notes section retained by CRE for JAF-CALC-RAD-00028'for a copy of this reference.]
- 40. JAF Original FSAR, Supplement 25, " Effects of High Energy Piping System Breaks Outside of Primary Containment,"
(7/22/74) _
- 41. US NRC NUREG-0123, Rev. 3, " Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)" (1980)
- 42. Proposed Technical Specification Changes - Power Uprate (JPTS-91-025), and NYPA Letter to NRC JPN-92-028 (6/5/92)
- 43. JAF Procedure AP-08.02, " Fuel Reliability Action Plan" (Rev. 4, 7/5/95) [Previously named " Failed Fuel Action ,
Plan") -
- 44. 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 a copy)
.f es 45. GE Technical Report NEDE-31152P, "GE Fuel Bundle Designs,"
( Rev. 3 (February 1993) ,
- 46. NYPA Memorandum JAG-93-245 addressed to J. Lazarus, from
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE ? OF //O g PROJECT: JAF PRELM [] PREPARED BY M DATE f/3b/ft Q TITLE: Control Room Radiological Habitability Under Power dprate FINAL [X] CHECKED BY // DATE g ,y/r/ , Conditions and CREVASS Reconfiguration J. Gray, titled " Control Rod Drop Accident (CRDA) Assumption" (9/24/93) [See JAF-CALC-RAD-00026 for a copy of this reference.]
- 47. SWEC Engineering Calculation #12966-RP-76-004, "LOCA Six-Month Gamma Doses for IE 79-01B Equipment Qualifications" (9/29/80)
- 48. R. G. Jaeger, Ed., " Engineering Compendium on Radiation I Shielding," Springer-Verlag, NY (1975) 4 9.
- SWEC Engineering Calculation 12966-RP-60-23, " Maximum Post-Accident Iodine Load in Standby Gas Treatment System Charcoal" (6/11/80)
- 50. JAF Original FSAR, Supplement 20, Response to AEC _
Questions of 1/12/1972, Q.11.9-1.
- See Attachment A for a copy of this reference.
O .
)
l I 1
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 6 OF //O PROJECT: JAF PRELM [] PREPARED BY /hC DATE 1/3C/D FINAL [X] CHECKED BY 8 DATE /o/f7/ff TITLE: Control Room Radiological Habitability Under Power dprdte ~ Conditions and CREVASS Reconfiguration List of Computer Procrams Employed The following CRE computer programs and data libraries were used in the analyses documented in this calculation: Program Reference Release Date of Computer Name Number Number Release System DORITA-2 RAD-001 1.5.1.2 01/22/97 RS/6000
~
ELISA RAD-005 1.3.1.1 03/09/92 DG AViiON QAD-CGGP RAD-006 1.3.1.1 3/26/92 DG/AViiON MARIO* --- --- --- DG/AViiON O .... -
- Computer code MARIO was developed for computing dose rates using the results from QAD-CGGP and DORITA-2/ELISA/ ALLEGRA calculations. The program is documented in CRE calculation IP3-CALC-RAD-00002 (Ref. 28), and was used in the current calculation without any modification.
w O . M
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 9 OF //O PROJECT: JAF PRELM [] PREPARED BY /ffl. DATE 9/M99 FINAL [X) CHECTJ!:D BY /f DATE /p//7/r;7 TITLE: Control Room Radiological Habitability Vnder Power 8pr' ate Conditions and CREVASS Reconfiguration l Table of Contents CALCULATION CONTROL SHEET ................................. 1 STATEMENT OF PROBLEM ...................................... 2 REFERENCES ...................................... ......... 3 LIST OF COMPUTER PROGRAMS EMPLOYED ........................ 8 TABLE OF CONTENTS ....:. .................................. 9
- 1. INTRODUCTION ......................................... 11 -
- 2.
SUMMARY
OF RESULTS ................................... 13
- 3. METHODS OF ANALYSIS .................................. 27
- 4. RADIATION EXPOSURES FROM A LOSS OF COOLANT ACCIDENT .. 30 4.1 Drywell and MSIV Leakage ........................ 31 4.1.1 Basic Data and Assumptions ............... 31 4.1.2 Results .................................. 36 4.2 ESF Component Leakage ........................... 39 4.2.1 Baeic Data and Assumptions ............... 39 4.2.2 Results .................................. 40 4.3 Total LOCA Dose ................................. 42 -
- 5. RADIATION EXPOSURES FROM A MAIN STEAM LINE BREAK ..... 44 5.1 Basic Data and Assumptions ...................... 44 5.2 Results ......................................... 52
- 6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT . 57 6.1 Basic Data and Assumptions ...................... 57
~
6.2 Results ......................................... 61
- 7. RADIATION EXPOSUPES FROM A REFUELING ACCIDENT ....... 65 7.1 Basic Data and Assumptions ..................... 65 7.2 Results ........................................ 69
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /O OF // O PROJECT: JAF PRELM [] PREPARED BY /?)ff DATE 9/M99-i FINAL [X] CHECKED BY // DATE M,787 TITLE: Control Room Radiological Habitability Und.er Power dprdte Conditions and CREVASS Reconfiguration TABLE OF CONTENTS (Cont.) Page
- 8. RADIATION EXPOSURES FROM EXTERNAL SOURCES ........... 71 8.1 Direct Shine from Post-LOCA Airborne Radioactivity in the Reactor Building .......... 71 8.1.1 Basic Data and Assumptions .............. 72 l
8.1.2 Results ................................. 75 8.2 Direct Shine from Post-LOCA Overhead Clouds .... 79 8.2.1 Basic Data and Assumptions .............. 80 8.2.2 Results ................................. 83 8.3 Direct Shine from Halogens Accumulating on the - Control Room Charcoal Filters .................. 85 8.3.1 Basic Data and Assumptions .............. 85 l( 8.3.2 Results ................................. 88 '\
- 9. POST-LOCA SGTS FILTER IODINE LOADING ................ 97 9.1 Basic Data and Assumptions ..................... 97 9.2 Results ........................................ 101 -
ATTACHMENTS A. Excerpts from References Pertinent to this Calculation B. Copies of Computer Outputs
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE ll OF // O . PROJECT: JAF PRELM [] PREPARED BY /dfL DATE _9/Df9} FINAL [X] CHECKED BY // DATE /d/n Control Room Radiological Habitability Under Power 8pfate
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TITLE: Conditions and CREVASS Reconfiguration 1.0 Introduction The JAF conurol room habitability under power uprate conditions was originally analyzed in Reference 7. The following changes, subsequently identified and implemented, required re-analyses of the control room habitability from design basis accidents (DBAs): a) re-configuration of the Control Room Emergency Ventilation Air Supply System - CREVASS (hefs. 8-13), b) revision to the MSLB release rates from a break in the 16" bypass line instead of the break in the 24" main steam line as postulated in Reference 6 [the release from the 16" break
~
was found to be more limiting than the break in the 24" line (References 14 and 15)], c) evaluation of the main steam line break (MSLB) accident with an RCS activity concentration equal to 2 pCi/gm I-131 Dose
~
Equivalent (The Technical Specification 1imit for Limited Condition of Operation), and d) changes to the release pathway for a CRDA through the elimination of the reactor scram and MSIV closure functions of the main steam line radiation monitors (Ref. 16) References 17 and 18 addressed the above changes for the pre-uprate conditions. Reference 1 documented the analyses for _ uprate conditions. The current analyses refine those of Reference 1 to address the following: , e) evaluation of the loss of coolant accident (LOCA) and i refueling accidert (RA) assuming a lowered stand-by gas treatment system (SGTS) charcoal filter efficiency (assumed - efficiency of 90% for halogens), O' . f) revision of the atmospheric dispersion factors for elevated __m____-_-.-__-_-____m-______________________.___________-___m. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__ . _ - __ _ _ _ _ _ _ _ _ _ _ _ _ . . - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ J
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE J7 OF //O _
,% PROJECT: JAF PRELM [] PREFARED BY /9ff DATE_9/Mi?-
( FINAL [X] CHECKED BY // DATE /0/>//7 f TITLE: Control Room Radiological Habitability Under Power dpr6te Conditions and CREVASS Reconfiguration releases as documented in Reference 5, and g) use of the ICRP 30 (Ref. 6) dose conversion factors for the . determination of thyroid doses. i I l b ~ l I
i NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 13 OF //D - PROJECT: JAF PRELM [] PREPARED BY 47C. DATE 9/M4 ! I FINAL [X] CHECKED BY // DATE fo//7/f7 TITLE: Control Room Radiological Habitability Dnder Power 6pr' ate Conditions and CREVASS Reconfiguration 2.0 Summarv of Results The following design-basis accidents were considered in the re- i assessment of the JAF control room radiological habitability: (a) Loss of coolant accident (LOCA) (drywell, MSIV and ESF component leakage pathways), (b) Main Steam Line Break outside containment (MSLB), with equilibrium and spiked RCS iodine concentrations, (c) Control Rod Drop Accident (CRDA), and (d) Refueling Accident (RA). l In each case, with the exception of an MSLB with spiked RCS ! activity, CREVASS is assumed to be initially in the " normal" operating mode. The system is then manually placed in the - '( ) " isolate" mode at 30 minutes after a LOCA and a CRDA, and at the l worst-case time of 12 min after an MSLB and an RA. MOV-107 and l MOD-109 in the exhaust duct, and MOD-105 in the supply duct close, while MOV-108 in the supply duct fails to close (see Fig. 2.1). In addition, bypass damper DMPR-105, which is in parallel with MOD-105, is assumed to be manually closed within 12 hours ~ after the accident, as discussed further below. 2 Bypass damper DMPR-109 in the exhaust duct was determined to be " inaccessible" for post-accident response and has been permanently closed (Ref. I 12); CREVASS was balanced with this damper closed. The CR ventilation flows as a function of time are shown in Table 2.1. The basic data and assumptions in each of the four accident l scenarios are consistent with the current licensing basis and the Current plant procedure OP-55B includes closure of DMPR-105 in the list of operator actions for placing the CR in the f3 isolate mode, within 30 minutes. In the current analysis, ( closure of this damper was assumed to take place within 12 hrs - for ALARA reasons.
I NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /(/ OF //O - PROJECT: JAF PRELM [] PREPARED BY //f_ DATE 9/3a/ft ' i FINAL [X] CHECKED BY // DATE jo// //f7 TITLE: Control Room Radiological Habitability Under Power dprkte Conditions and CREVASS Reconfiguration I models in the regulatory guides (Refs. 19 - 24 and 31) and the Standard Revie.: Plan (SRP, Ref. 4). Complete details for each - accident are presented in Secs. 4 through 7. A summary of the principal assumptions associated with each DBA and the doses due to immersion appear in Table 2.2. The worst-case immersion doses and the regulatory exposure limits (from 10 CFR 50 Appendix A, General Design Criterion 19, and Sec. - 6.4 of the SRP (Ref. 4)] are for a CRDA and are as follows: Design-Basis Accident Thyroid Wh. Body Skin and conditions (rem) (ram) (rem) i CRDA with DMPR-105 18.76 0.0514 0.5673 closure at 12 hrs GDC-19 limits 30 ( 5 30 1 Note that the manual closing DMPR-105 is not. required for limiting the dose rates within the design criteria. The delay time of 12 hours was selected for ALARA reasons and used for the analysis of all accidents analyzed in this calculation. - The doses from external sources are shown in Table 2.3, and are relatively insignificant. The worst-case 31-day dose is about 127 mrad, in the Operations Office, for an MSLB with spiked RCS activity. For accessibility of the CR ventilation equipment room, note that the worst-case radiation dose rate in this area, due to gamma radiation emanating from halogen buildup on the charcoal filters, could be as high as 4.3 rad /hr. This is for an MSLB with spiked RCS activity, at a time immediately following the accident and to ()- a receptor in contact with the filter casing. Time-dependent dose rates from this source are shown in Table 2.4; it is seen i I i
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /I OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 87/3c/9'd. ~ FINAL [X] CHECKED BY Af D A T E J g /j p / f 7 TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration that at i day after the accident, the dose rate drops to about 317 mr/hr. For all other accidents, the worst-case dose rcte is less than 70 mrad /hr. Note that, since the ventilation equipment room forms part of the CR pressure boundary, the immersion doses in the equipment room will be the same as those in the control room. The final conclusion is that all radiation doses to CR personnel would be within the regulatory limits and that auto-isolation of _ the control room is not necessary for any design-basis accident. The current OP-55B requirement for closing DMPR-105 within 30 minutes may be relaxed, if desired. C1'osure within 12 hours is recommended for ALARA reasons, but is not required. () This calculation also addresses the post-LOCA iodine loading of - the Standby Gas Treatment (SGTS) filtsrs. The design-basis maximum loading is 817 grams [ based on 720 lb of activated charcoal and a loading ratio of 2.5 mg of total iodine per gram of carbon (Ref. 24)]. It was determined that the filter would be saturated beyond 400 hours following the accident. The filter loading as a function of post-LOCA time is shown in Table 2.5. - ! As a final remark, note that the post-LOCA drywell leakage rate of 1.5% per day used in this analysis includes a leakage of 0.23
% per day by the 4 MSIVs, the latter leakage being collected by the main steam leakage collection system and fed directly to the SGTS.
~
l Hence, the present calculation supports drywell l containment leakage rates up to (1.5 - 0.23) = 1.27 % per day. O l l l {
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /b OF //D _ PROJECT: JAF PRELM [] PREPARED BY /)11P._ DATE f/Ja/M - FINAL [X] CHECKED BY /f DATE'k6h/rP _ TITLE: Control Room Radiological Habitability Under Power Up' rate Conditions and CREVASS Reconfiguration Table 2.1 CR Ventilation System Operatirg Conditions and Flots As a Function of Post-Accident Time - Description Value 4 Pre-isolation air intake rate (scfm) 15000 Manual isolation time (min) 30/12* Post-isolation filtered air intake rate (scfm) 1000"' Post-isolation unfiltered air intake (scfm) 2100 Bypass damper DMPR-105 closure time (hr) 12 '*) Post DMPR-105 closure unfiltered air intake (scfm) 300'" Unfiltered air intake with CR isolated and no MOV 1 failures (scfm) 10 0 (a) Maximum possible intake rate under certain outside air conditions (about 15% higher than can b~e provided by the supply fans). (b) Manual isolation in 30 minutes for a LOCA and a CRDA, and the worst-case time for an MSLB and an RA [namely, 12 min, from Sec. 5.1 (j ) ] . _ (c) One operating charcoal filter train; 90% charcoal filtration efficiency for all halogen species, a conservative value for two 2" charcoal beds in series without humidity control. (d) Failed MOV-108; bypass damper DMPR-105 open. (e) Conservatively selected post-accident time for accessing and closing the damper from ALARA considerations; closure of the damper is not required to limit exceeding exposure limits. - (f) Failed MOV-108; bypass damper DMPR-105 closed. (g) MSLB with spiked RCS activity only (2 pCi/gm I-131 DE) . O
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /t OF //O PROJECT: JAF PRELM [] PREPARED BY /)lf DATE 9/3e/79-FINAL [X] CHECKED BY // DATE / /gp TITLE: Control Room Radiological Habitability Under Power Uprate ~ Conditions and CREVASS Reconfiguration Table 2.2 JAF Control Room Personnel Radiation Dases Due to Immersion in Airborne Radioactivity (Continuous Occupancy) Desian-Basis Accidents Thyroid Whole Body Skin (rem) (rem) (rem) Case 1: LOCA (Dryw/MSIV Leak) 1.01E+01 1.15E-02 1.40E-01 LOCA (ESF Leakage) 1.06E+00 2.34E-04 2.40E-03 LOCA (Total) 1.11E+01 1.17E-02 1.42E-01 MSLB with equil. RCS 1.60E+01 1.42E-02 8.92E-02 ) Control Rod Drop 1.88E+01 5.14E-02 5.67E-01 _ Refueling Accident 2.06E-02 2.~28E-04 3.44E-03 l Case 2 MSLB with Spiked Activity 1.78E+01 1.79E-02 1.31E-01 Note: l Case 1 - DMPR-105 closure at 12 hours after the accident ) Case 2 - Control Room pre-isolated (administrative control, l implemented via the Fuel Reliability Action Plan, Ref. 43); MSLB only. d _____.-.2 . - _ . _ . - -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE lf OF //O p PROJECT: JAF PRELM [] PREPARED BY % DATE 9/M 94 FINAL [X] CHECKED BY // DATE M/f7/f/ 3
- TITLE: Control Room Radiological Habitability Under Power Up/ ate ~
I Conditions and CREVASS Reconfiguration l Table 2.2 (Continued) BASES LOCA (Drywell and MSIV Leakage) (a) A LOCA takes place at full power (2535.8 MWt + 2% uncertainty). (b) All the noble gases and 25% of the halogens become airborne within the drywell at the time of the accident and are l available for release. ! (c) Leakage from the drywell is at the rate of 1.5% per day, consisting of 1.27 % per day due to containment leakage, and 0.23 % per day due to MSIV leakage. (d) All the noble gases and halogens leaking from the drywell , () (_s/ are exhausted to the atmosphere via the Standby Gas Treatment System (SGTS) and the main stack without holdup or l mixing in the reactor building. . (e) The SGTS filter efficiency is 90% for the removal of all halogen species. LOCA (ESF Corponent Leakage) (a) 50% of the total halogen activity present in the core mixes uniformly with the coolant in the RHR system (113,400 cu _ ft). (b) The ESF leakage rate is 5 gpm, and is constant from the , start of the LOCA through the duration of the accident. I i (c) An additional 30-minute leakage of 50 gpm (due to gross l 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 becomes airborne and mixes unifn".ly with the reactor building atmosphere. l (e) Release from the reactor building is through the SGTS and the main stack at the rate of 6000 scfm. O .
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE [T OF //O
. PROJECT: JAF PRELM [] PREPARED BY /M DATE 9/30/M FINAL [X] CHECKED BY // DATE /o/f7/f7 TITLE: Control Room Radiological Habitability Under Power 0 prate Conditions and CREVASS Reconfiguration Table 2.2 (Continued)
Main Steam L.no Break (a) A line break occurs in the 16" bypass line leading to the turbine steam chest outside containment during full power operation. (Note: A break in one of the 24" main steam lines is less restrictive.) (b) The MSIVs close in 10.5 seconds after the break. (c) The total discharge through the break prior to isolation is equal to 18,179 lb of steam and 87,118 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 following a 30-minute decay). Fifty percent of all noble gases leaving the reactor vessel during the 10.5-sec MSIV. closure time (via _
all four steam lines) are released through the break. The halogen source term in the discharged liquid was selected to represent the Technical Specification Limits for the following:
- 1. The maximum RCS concentration under power uprate equilibrium conditions (0.2 pCi/gm I-131 DE), and
- 2. The maximum RCS concentration under Limiting Conditions -
of Operation (2 pCi/gm I-131 DE). (f) 100 % of the radioactivity discharged into the turbine building becomes airborne and is 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) For the case with the RCS concentration at equilibrium , (
~
conditions, placement of CREVASS in the " isolate" mode ' akes ! place at the post-accident time (less than 30 minutes) which j maximizes the CR radiation exposures; this time wat l determined to be 12 minutes. For the spiked RCS
/~') concentration, the CR is assumed to be pre-isolated.
(_/ l
- (
l -
- \
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 20 OF //O PROJECT: JAF PRELM [] PREPARED BY / /- 8 DATE W30 M l FINAL [X] CHECKED BY ff DATE/c//y/f/
. TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration Table 2.2 (Continued)
Control Rod Drop 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. (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 coc' ant and is instantly and uniformly mixed with the coolant in the pressure vessel at - the time of the accident. The released activity corresponds to 10% of all halogens and 10% of all noble gases (except Kr
- 85) in each failed rod, and to 30% of the Kr 85 inventory.
G (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 condenser and offgas system. (f) As a result of plant shutdown following a CRDA, or as a result of offgas system automatic isolation
- due to high l radiation fields at the offgas monitors, the released _
i 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. 1 Offgas system isolation takes place automatically following a 16-minute delay. This delay time is shorter than the transit time (22 min, UFSAR Sec. 11.4.4.2) of offgas effluent to O the main stack via conditions. the 24" hold-up pipe, under startup J
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAG.9 2J OF _//O - PROJECT: JAF PRELM [] PREPARED BY @ DATE 9/3p/M FINAL [X] CHECKED BY M DATE /q47/f7 TITLE: Control Room Radiological Habitability Under Power'Uprate Conditions and CREVASS Reconfiguration (h) The leakage rate from contaminated systems into the turbine building amounts to 1% per day and lasts for 24 hor- s. The release to the atmosphere is at ground level and there is no holdup within the turbine building. l Refueling 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 I 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 within the gaps of the failed fuel rods is released to the fuel pool water. The released activity is
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conservatively assumed to correspond to 10% of all halogens (except I 129) and 10% of all noole gases (except Kr 85) in each failed rod, and to 30% of the I 129 and Kr 85 inventories. (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) Radioactive gases which escape the pool are released to the atmosphere via the SGTS and the main stack over a 2 hour period. The selected release rate and the time of CR - isolation are as described for the MSLB accident. I O - 1
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 21 OF // C ( PROJECT: JAF PRELM [] PREPARED BY //vC. DATE 9/M99 l FINAL [X] CHECKED BY // DATE /o//p//^P - TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration l Table 2.3 JAF Control Room Personnel Radiation Doses Due to Direct Shine from External Sources Worst-Case Accidents and Receptor Locations (Continuous Occupancy) Post-LOCA Shine from Airborne Radioactivity Accumulating in the RB Refueling Level (Receptor at 12 ft from CR S wall) Time Dose Rate Int. Dose Cum. Dose (rad /hr)
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(hrs) (rad) (rad) 0.0 0.0000E+00 0.0000E+00 0.0000E+00 0.5 6.3697E-04 1.5924E-04 1.5924E-04 1.0 9.3312E-04 3.8782E-04 5.4707E-04 2.0 1.2353E-03 1.0772E-03 .1.6242E-03 8.0 8.1985E-04 6.0806E-03 7.7048E-03
/"'\ 24.0 9.2617E-05 5.3359E-03 1.3041E-02 (ms/ 48.0 1.0505E-05 9.0539E-04 1.3946E-02 96.0 1.2244E-06 2.0725E-04 1.4153E-02 168.0 3.0428E-07 4.7583E-05 1.4201E-02 240.0 1.7522E-07 1.6837E-05 1.4218E-02 744.0 2.0272E-08 3.6208E-05 1.4254E-02 l
Post-LOCA Shine from Overhead Radioactive Clouds 1 Time Dose Rate Cum. Dose Ibrs) (rad /hr) (rad) 0.0 9.6977E-05 0.0000E+00 1 0.5 3.3896E-05 3.0005E-05 1.0 2.5428E-05 4.4735E-05 2.0 1.7901E-05 6.6180E-05 8.0 3.5629E-06 1.1947E-04 - 24.0 6.2400E-08 1.3332E-04 48.0 1.0061E-09 1.3368E-04 96.0 2.5417E-11 1.3369E-04 l 168.0 2.9297E-12 1.3369E-04 (~ 240.0 1.6054E-12 1.3369E-04 l ( ,g) 744.0 3.3788E-13 1.3369E-04 W - - - - _ - - - _ - - _ - - - _ _ _ _ _ _ - _ _ - - , - - . - - - - - _ - - - - - - - - - - - _ - - - - . , _ _ . _ - - - _ _ _ - - . - - - - - . _ - . - _ - _ - - - . - - - - . - - . - . - - - - - - - - - - - - - - _ - - - _ - - - - _ . _ . - - - - - - _ - - - _ _ - , , _ - - - -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 23 OF //O g PROJECT: JAF PRELM [] PREPARED BY /f7C DATE 9/3b/9% ~f] TITLE: FINAL [X] CHECKED BY // DATE 4,/ Control Room Radiological Habitability Under Power Dprate f7 /f/
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Conditions and CREVASS Reconfiguration Table 2.3 (Continued) Post-MSLB Shine from Halogens Accumb.ating on the CR Charcoal Filters (Receptor in CR Operations Of fice) Time Dose Rate Int. Dose Cum. Dose (Hrs) (rad /hr) (rad) (rad)
- 0. 0- 0.000E+00 0.000E+00 0.000E+00 0.5 1.319E-02 3.297E-03 3.297E-03 1.0 1.296E-02 6.537E-03 9.834E-03 _
2.0 9.491E-03 1.114E-02 2.097E-02 4.0 5.804E-03 1.499E-02 3.597E-02 8.0 3.246E-03 1.761E-02 5.357E-02 16.0 1.549E-03 1.835E-02 7.192E-02 24.0 8.949E-04 9.538E-03 8.146E-02
- (~') 96.0 9.445E-05 2.563E-02 1.071E-01
. (ms/ 744.0 5.110E-06 1.985E-02 1.269E-01 . With spiked RCS concentration (2 pCi/gm I131 DE)
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(a) 1 l k I i l
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NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 2,4 OF //O f PROJECT: JAF PRELM [] PREPARED BY /)R. DATE 1/Je/O FINAL [E] CHECKED BY ff DATE g/o/f7 - TITLE: Control Room Radiological Habitability Under Fower/Upfate Conditions and CREVASS Reconfiguration Table 2.4 Direct Shine Dose Rates in CR Ventilation Equipment Rc.m Due to Gamma Radiation Emanating from the Charcoal Filters Worst-Case Accident and Receptor Location [MSLB with Spike RCS Activity (2 Ci/gm I131 DE) and Receptor on Contact with Filter Casing) Time Dose Rate Int. Dose Cum. Dose (Hrs) (rad /hr) (rad) (rad) 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 4.301E+00 1.075E+00 1.075E+00 1.0 4.228E+00 2.132E+00 3.207E+00 2.0 3.089E+00 3.628E+00 6.836E+00 4.0 1.877E+00 4.866E+00 1.170E+01 I 8.0 1.049E+00 5.694E+00 1.740E+01 16.0 5.199E-01 6.031E+00 2.343E+01 _ 24.0 3.169E-01 3.280E+00 2.671E+01 96.0 4.345E-02 9.908E+00 3.662E+01 744.0 2.663E-03 9.466E+00 4.608E+01 1 Note: See Sec. 8.3 for more locations and other accidents l O l
__. l l NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 2. f OF //O PROJECT: JAF PRELN [] PREPARED BY M DATE f/_te/@ ' FINAL [X] CHECKED BY // DATE /s/,h/f/ _ TITLE: Control Room Radiological Habitability Nnder Power 6 prate f Conditions and CREVASS Reconfiguration Table 2.5 i PC.*T-LOCA SGTS FILTER IODINE LOADING (Drywell, MSIV and ESF Component Leakage) Post LOCA Filter Loading Time (hr) (gm of Iodine) ! 0 0.000E+00 0.5 1.094E+00 1 2.226E+00 _ 2 4.495E+00 5 ' 1.132E+01 10 2.274E+01 20 4.554E+01 50 1.129E+02 100 2.215E+02 (\ 200 4.268E+02 400 7.974E+02 450 8.830E+02 - 500 9.662E+02 744 1.338E+03 Note: The maximum permissible iodine loading on the SGTS filter is 817 grams, based on 720 lb of activated charcoal and a tota' iodine loading design basis of 2.5 mg per gram of carbou (Ref. 24). - 'O e
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 2b OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9[Jb/$ t FINAL [E] CHECKED BY /f[ DATE /E/n/rf
, , _ TITLE: Contiol Room Radiological Habitability'Under Power dpr' ate Conditions and CREVASS Reconfiguration Fig. 2.1 -
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NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 2? OF //O q PROJECT: JAF PRELM [] PREPARED BY M DATE 9/39/19 Q TITLE: JINAL [X] CHECKED BY // DATE /sd7/f7 Control Room Radiological Habitability Under Power Dpfate Conditions and CREVASS Reconfiguration - 3.0 Methodoloav . Post-accident radiation exposures in the JAF control room (CR) were computed using the following: (a) The methodology and assumptions in the regul.. tory guides (Refs. 19 through 24) and the pertinent sections of the Standard Review Plan (Ref. 4), (b) Appropriate source terms, release pathways, decontamination factors and other assumptions, as described later, (c) Post-accident atmospheric dispersion factors based on 8- - years' worth of hourly meteorological data collected on site by Niagara Mohawk, from JAF-CALC-RAD-00007, Rev. 2 (Ref. 5), (d) The latest configuration of the control room ventilation (. 7-~)s system, and (e) The following CRE Computer Codes: DORITA-2 (Ref. 25) Computation of radihtion exposures, and - definition of gamma spectra associated with post-LOCA airborne radioactivity within the reactor building and with the halogens accumulating on the CR charcoal filters. ELISA (Ref. 26) Definition of the gamma radiation ' spectra associated with radioactive clouds resulting from post-LOCA drywell/MSIV leakage, for computation of the cloud shine in the control room. QAD-CGGP (Ref. 27) Determination of the relative gamma
~
]
fluxes at the locations of interesc (in l terms of MeV/sec-cm per MeV/s c emitted 2 w by a source, as a function of gamma l s- energy), for gamma radiation emanating 1
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 2.T OF //O PROJECT: JAF PRELM [] PREPARED BY /)(l DATE 9/3Mf7 1 FINAL [X] CHECKED BY Af DATE M/ M9/ TITLE: Control Room Radiological Habitability Under Power Dprate Conditions and CREVASS Reconfiguration I from the reactor building, overhead clouds and the CR charcoal filters. MARIO (Ref. 28) Computation of dose rates (and cumulative doses) at the receptors of interest as a function of post-accident time, using the gamma spectra generated by DORITA-2 and the relative gamma fluxes produced by QAD-CGGP. Sections 4 through 9 which follow present details on the assumptions, data and results associated with each of the design- - basis accidents analyzed. For the MSLB and refueling accidents, the release of ( radioactivity to the atmosphere and into the control room is very fast (relative to the time allotted for manual isolation of the control room), and therefore contamination of the air entering - the control room will drop significantly in a very short time. CR personnel exposures would tend to increase if the air-exchange rate of the control room is reduced after the radioactivity has already entered the control room (thus trapping the radioactivity within the control room). In view of the above discussion,
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iterative analyses were carried out using various CR isolation times to determine the worst-case radiation exposures. The documentation in this calculation includes the results for three different CR isolation times in the MSLB analyses which were used to define the worst-case scenario. The worst-case CR isolation time for an MSLB was also used in the analysis of a refueling
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accident. l l As a final remark, note that the CR radiation exposures were ( computed under the assumption of continuous occupancy for the
I l k NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 27 OF _//C2 PROJECT: JAF PRELN [] PREPARED BY /hf. DATE 9/93/% FINAL [X] CHECKED BY 8 DATE fo/f f/f[ _ TITLE: Control Room Radiological Habitability Under Power Dpr' ate Conditions and CREVASS Reconfiguration - duration of the accident. The regulatory models require 100% occupancy during only the first 24 hours of a postulat:d accident; for days 2, 3 and 4 the occupancy factor is reduced to 60%, and for periods beyond 4 days 40% occupancy is allowed. Credit for partial occupancy was not considered since the calculated radiation exposures are within the regulatory guidelines. i 6 O l 1 E______ __________m_ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ __ _____.___._.___.______________.________________.__________._____________
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 30 OF //O PROJECT: JAF PRELM [] PREPARED BY /)C DATE _9/M7 FINAL [X] CHECKED BY /// DATE fodp#7
- TITLE: Control Room Radiological Habitability Under Power Uprate conditions and CREVASS Reconfiguration
- 4. RADIATION EXPOSURES FROM A LOSS OF COOLANT ACCIDENT Release pathways and contributing radiation sources which are typically addressed in the analysis of a LOCA are the following:
(a) Drywell and MSIV leakage, (b) ESF Component leakage, (c) Direct gamma radiation from airborne radioactivity accumulating on the refueling floor of the reactor building, (d) External cloud exposure, and (e) Halogen activity accumulating on intake / exhaust filters. _ For the JAF control room, 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. With respect to MSIV leakage, JAF is equipped with a Main Steam , Leakage Collection System (MSLCS) whose safety objective is to collect and process leakage past the MSIVs following a LOCA. The effluent of the MSLCS is processed by the Standby Gas Treatment System (SGTS) and is exhausted through the stack. The negative delta-P between the MSLCS and the SGTS is sufficient to provide the required flow through the MSLCS to collect all postulated _ leakage. The MSLCS is designed to meet the requirements of a Class I, Seismic Category I system. Since all MSIV leakage is processed by the SGTS prior to release to the atmosphere, the radionuclides release pathway is identical to that for post-LOCA drywell leakage. In view of the existence of the MSLCS, the l drywell-leakage pathway ir assumed to include MSIV leakage, and - separate assessment of MSIV leakage is not needed. More details can be found in Ref. 13. O-C _m________ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _._ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
l l NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 3I OF // O PROJECT: JAF PRELM [] PREPARED BY /)E DATE 9/30/9$- g
") _ TITLE:
FINAL [X] CHECKED BY Control Room Radiological Habitability 'Under Power Dprate g DATE f ./, 7/f 7 I Conditions and CREVASS Reconfiguration 4.1 Drywell and MSIV Leakage 4.1.1 Basic Data and Assumptions The following data and assumptions were used in the computation of the CR radiation exposures as a result of post-LOCA drywell and MSIV leakage:
.(a) A LOCA takes place at full power (2535.8 MWt + 2%
uncertainty, i.e., 2586.5 MWt) [Ref. 29 and Reg. Guide 1.49 (Ref. 22)]. (b) The core inventory for the radionuclides of interest is shown in Table 4.1 (based on information from Ref. 30) . (c) 100% of the noble gases and 25% of the halogens present in the core are instantly released to the drywell where they
\
are available as an aerosol for leakage to the secondary containment [ Reg. Guide 1.3 (Ref. 19)]. (d) The halogen composition airborne within the drywell is as follows: 91% elemental, 4% organic and 5% particulate [ Reg. Guide 1.3 (Ref. 19)]. (e) Leakage from the drywell is at the rate of 1.5% per day (UFSAR, Rev. O, 7/82, Secs. 14.8.1.5 and 14.8-22). This rate accounts for both drywell containment leakage and MSIV leakage, and is assumed to be constant for the accident duration. Note that the drywell design-basis leakage is 0.5% per day of containment volume (Technical Specifications Secs. 4.7.A.2.B and 6.20, and UFSAR, Rev. O, 7/82, Secs. 11.5.3.10 and 14.6.1.3.5). The Tech Spec limit for MSIV leakage is 11.5 scfh for ) each MSIV when tested at 2 25 psig [ Reg. Guide 1.96 (Ref. 31), and Technical Specifications Secs. 4.7.A.2.B and 6.20].
's./ This corresponds to a maximum leakage of [11.5 scfh] x (14.7
f NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 3 2. OF f /O g PROJECT: JAF PRELM [] PREPARED BY /ht DATE 9/30/M FINAL [X] CHECKED BY 8 DATE /o/n/;f7 TITLE: Control Room Radiological Habitability'Under Power dprate Conditions and CREVASS Reconfiguration psia /(14.7 + 25) psia] x [(460 + 300) 91 / (460 + 60) R] = 6.?24 cfh at drywell atmospheric conditions. For 4 MSIVs, the corresponding leakage rate is 24.89 cfh, or [24.89 ft'/hr] x [24 hr/ day] / [264,000 f t', drywell plus torus free volumes] x [100 %) = 0.226 % per day. Based on the above, the overall leakage from the drywell is 0.5 + 0.226 = 0.726 % per day. In the current analysis, the leakage was conservatively assumed to be 1.5 % per day. (f) All the noble gases and halogens leaking from the drywell are exhausted to the atmosphere via the Standby Gas Treatment System (SGTS) and the main stack without mixing in the reactor building.
~x (g) The SGTS filter efficiency for the removal of halogens is sms )
conservatively assumed as 90% for all halogen species [Ref. 32 documents the case for use of a 99% filter efficiency for a 2" charcoal with humidity control, and test acceptance criteria as specified for 4" beds in Ref. 24. Although the higher efficiency has been accepted by the NRC during discussions on the power uprate analyses (Ref. 33) the present analysis employs a lower filter efficiency to provide some relief for testing acceptance criteria. ACTS Item No. 23847 (Ref. 3) requires re-analyzing the DBAs at 95% filter efficiency). (h) The atmospheric dispersion factors associated with the transport of released radioactivity to the control room intake are as follows (in ( sec/m') , from Ref. 5]: Interval: 0-8 hrs 8-24 hrs 1-4 days 4-31 d.,. (X/Q)c., 9.26e-07 6.75e-07 3.39e-07 1.26e-07 f~% (X/Q),,,, 3.24e-06 2.45e-06 1.34e-06 5.60e-07
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. TITLE: Control Room Radiological Habitability'Under Power Uprate Conditions and CREVASS Reconfiguration 11 Note the following:
(1) The concentration (X/Q)s are for defining the nuclide concentrations at the CR outside air intake. (2) The gamma (X/Q)s are for computing the whole body dose to a receptor at the CR outside air intake due l t'o exposure to finite radioactive clouds above. These factors are employed later in Sec. 8; they are included here for convenience. (3) The prescribed assumption for fumigation conditions prevailing at the site at the time of an accident is not applicable to the CR outside air intakes. See Ref. 5 for details. (4) Dispersion factors for elevated releases were reanalyzed to accommodate possible short-term O- meandering and looping effects by placing the control room intake at that distiince from the stack where concentrations would peakJ See Ref. 5 for details. (i; The control room characteristics are as follows: Free air volume: 1.01E+05 ft' (UFSAR, Rev.0, 7/82, pg 14.8-21) Unfiltered air intake: (1) 15,000 scfm when the ventilation system is initially in the " normal" (pre-isolation) mode, (about 15% higher than the maximum flow that can be provided by the 70-AHU-3A/B units), (2) 2100 scfm when CR ventilation system is in the " isolate" s mode with MOV failure, and with bypass damper DMPR-105
l i NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 3h OF /f 0 PROJECT: JAF PRELM [] PREPARED BY A(L DATE 9/30/% FINAL [I] CHECKED BY /f DATE , c/,7/97 TITLE: Contro? Room Radiological Habitability Under Power Uprate ! Condit::.n s and CREVASS Reconfiguration 1 open (based on a design flow l of 1920 scfm via the (l'x3') I bypass damper, from Ref. 34 and Drawing 11825-FB-35C, plus l 15 (scfm/f t"') x (4 ' x3 ' ) (ft"') = 180 scfm leakage via the closed MOD, from Ref. 35]. (Note: the 10 scfm due to ingress / egress, from SRP Sec. 6.4, is relatively small and was neglected.) (3) 300 scfm when the CR ventilation system is in the ! " isolate" mode with MOV failure, but with DMPR-105 closed [ selected to accommodate 10 scfm for , ingress / egress, and 15 ] (scfm/f t"') x 15 (ft"') = 225 scfm damper leakage, Ref. 35]. Normal op. filtration: None Post-isolation filtered air intake rate: 1000 scfm (UFSAR, Rev. O, 7/82, pg 14.8-22), with one filter train operating. Intake filter eff.: 90 % for all halogens species (4 " charcoal beds, but without humidity control) (Ref. 24) l Delay in switch!; CR i i Ventilation system to the " isolate" mode: 30 minutes (as previously selected O for operator action, Ref. 7) (Note:
NYPA - CALC.# JtLF-CALC-RAD-00042 REV 2 PAGE 3[ OF ll0 l PROJECT: JAF PRELN [] PREPARED BY /H2. DATE 9/Joli'F FINAL [X] CHECKED BY /f f DATE / o// 7/77 TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration SRP Sec. 6.4 allows 20 minutes for operator action). Delay time in closing DMPR-1058 : 12 hours [ Note: This delay time for closing DMPR-105 was selected for ALARA reasons. Failure to close DMPR-105 will not result in exceeding the radiation exposure limits.] [ Note: The intake rate into the CR is provided as input to the DORITA-2 code in units of CR volumes per day, and includes both filtered and unfiltered flows. The split between the filtered and unfiltered portions is accomplished d through use of the filter bypass fraction. For instance, for a filtered flow of 1000 scfm and an unfiltered flow of 2100 scfm, the fractional air intake rate into the control room is equal to [3100 (scfm) x 1440 (min / day) / 1.01E+05 ft'] = 44.20 air volumes per day, and the filter bypass fraction is (2100 acfm / 3100 scfm) = 0.677, as shown in the DORITA-2 input files in Attachment B.] (j) Radiation exposures were determined for a 31-day interval and for continuous occupancy. Whole body exposures were based on immersion in a finite spherical cloud having a volume equal to that of the control room. The breathing rate was set at 3.47E-04 (m*/sec) for the duration of the accident (Refs. 19 and 37), and use was made of the ICRP-30 dose conversion factors for thyroid exposure (Ref. 6). Current- plant procedure OP-55B includes closure of DMPR-105 in the list of operator actions for placing the CCR in the 3 isolate mode, within 30 minutes. In the current analysis, j different closure times were analyzed to provide a basis for relaxing the current OP-55B requirement, if the need arises.
l l NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 3$ OF //O , PROJECT: JAF PRELM [] PREPARED BY /9pC. DATE 9/MS
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l FINAL [X] CHECKED BY /// DATE )c/f'/ 2 f7 TITLE: Control Room Radiological Habitability Under Power 6 prate Conditions and CREVASS Reconfiguration I i l 4.1.2 Results ! CR radiation exposures due to drywell and MSIV leakage following a design-basis LOCA were calculated using the DORITA-2 computer code and the data and assumptions listed above. Copies of the DORITA-2 outputs appear in Attachment B to this calculation (Computer Run Case #1). Table 4.2 which follows presents the time-dependent thyroid, whole body and skin doses in the Control Room (due to post-LOCA drywell and MSIV leakage) for continuous occupancy. Refer to Sec. 2 for a summary of the exposures. See Section 4.3 for a discussion on the damper DMPR-105 closure delay time. . O W 9 f 1
NYPA - CALC . # JA'- CEC-RAD- 0004 2 REV 2 PAGE 1} OF [/O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/4c[99 - ( FINAL [X] CHECKED BY /// DATE //f5Ar7
. TITLE: Control Room Radiological Habitability Under Power prate Conditions and CREVASS Reconfiguration Table 4.1 Full-Core Inventon ; 2586.5 MWt ,
Nuclide Activ. (Ci) Nuclide Activ.(Ci) 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 134 1.566E+08 Xe 133 1.430E+08 I 135 1.344E+08 Xe 135m 2.695E+07 I 136 6.479E+07 Xe 135 1.847E+07 Xe 137 O Xe 138 1.255E+08 1.192E+08 3.123E+03 (Ci/MWt from Ref. 30) x 2586.5 (MWt) O .
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 38 OF [/O p PROJECT: JAF PRELM [] PREPARED BY M DATE 9/3c[99-
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l, FINAL [X] CHECKED BY /// DATE M'#v7 TITLE: Control Room Radiological Habitability Under Power Dplate Conditions and CREVASS Reconfiguration Table 4.2 Control Room Radiation Expos _ses Due to Drywell and MSIV Leakage Following A Design-Basis Loss-of-Coolant Accident Time Thyroid Whole Body Skin
.(hours) Dose (rem) Dose (rem) Dose (rem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.027E-01 6.637E-04 7.917E-03 8.000E+00 2.888E+00 5.146E-03 5.285E-02 1.200E+01 3.874E+00 6.218E-03 6.530E-02 2.400E+01 5.161E+00 8.112E-03 9.157E-02 9.600E+01 7.729E+00 1.035E-02 1.237E-01 7.440E+02 1.005E+01 1.146E-02 1.399E-01 Note:
- Closure of DMPR-105 is not required for meeting regulatory limits. However, closure at 12 hours after the accident is recommended for ALARA reasons
- Continuous occupancy
{ t D (V
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 37 OF //O 7/M /9'). PROJECT: JAF PRELM [] PREPARED BY AG. DATE
) FINAL [X] CHECKED BY /( DATE /cdf//7 TITLE: Control Room Radiological Habitability Under Power Dpr'a'e t Conditions and CREVASS Reconfiguration 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 RHR system, which has a total fluid mass of 3.21E+09 grams (Ref. 38). 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, Rev. O, 7/82, Table 5.2-1), and 900 cu ft of water from other sources.
(d) Total ESF component leakage rate is 5 gpm (Tech. Specifications, Sec. 3.6.D, for unidentified leakage inside the containment, and UFSAR, Rev. 1, 7/83, Sec. 4.10.3.2, for maximum allowable leakage rate from unidentified sources in the reactor coolant pressure boundary [both inside and outside the primary containment and systems essential to safe plant' shutdown, i .e . , ECCS] ) ; it corresponds to a fractional rate from the recirculating water system of 0.00849 vol/ 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. l (f) An additional leakage contribution due to a gross failure of l a passive compcnent with an assumed leak rate of 50 gpm is 7- included in the model (Ref. 4, SRP, Sec. 15.6.5, Appendix (_j. B). This leakage is assumed to begin at the time of LOCA 1
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 66 OF // O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/3c/f} FINAL [X) CHECKED BY /f/ DATE /0/,7/p TITLE: Control Room Radiological Habitability'Under Power dprate Conditions and CREVASS Reconfiguration onset and inst for a period of 30 minutes. (This assumption is more cone-rvative 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. 4, SRP, Sec. 15.6.5, Appendix B), and to mix uniformly with the RB atmosphere. (h) Release from the reactor building is through the SGTS and the .ain stack at the rate of 3.3 air changes per day (based on an SGTS flow of 6000 scfm with one fan operating (UFSAR, Rev. O, 7/82, Sec. 5.3.3.4)]. (i) The SGTS filter efficiency for the removal of halogens is 90% for all halogen species (Refs. 3, 24,32, and 33). (j) Transport of the released radioactivity to the control room, s ,/ the control room characteristics and other exposure-related parameters are described under Items (h), (i) and (j) in Sec. 4.1.1. 4.2.2 Results Table 4.3 presents the post-LOCA time-dependent thyroid, whole body and skin doses in 'he Control Room resulting from post-LOCA ESF component leakage. Refer to Sec. 2 for a summary of the exposures. Also, refer to Attachment B for copies of the DORITA-2 outpute (Computer Run Case #1). See Section 4.3 for a l discussion on the damper DMPR-105 closure delay time. l O 1 l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE h[ OF //O PROJECT: JAF PRELN [] PREPARED BY M DATE 7/M9-FINAL [E] CHECKED BY // DATE /s/N/r7 TITLE: Control Room Radiological Habitability 8nder Power Uprate _ Conditions and CREVASS Reconfiguration Table 4.3 Control Room Radiatlon Exposures Due to ESF Component Leakage Following A Design-Basis Loss of Coolant Accident Time Thyroid Whole Body Skin (hours) Dose (rem) Dose frem) Dose (rem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 6.458E-03 2.285E-06 1.484E-05 8.000E+00 2.365E-01 8.032E-05 6.674E-04 1.200E+01 3.371E-01 1.174E-04 1.019E-03 2.400E+01 4.770E-01 1.804E-04 1.737E-03 9.600E+01 7.716E-01 2.234E-04 2.275E-03 7.440E+02 1.056E+00 2.335E-04 2.400E-03 t V Note:
- Closure of DMPR-105 is not required for meeting regulatory limits. However, closure at 12 hours -
after the accident is recommende~d for ALARA reasons
- Continuous occupancy j I
O l
NYPA - CALC.# jar-CALC-RAD-00042 REV 2 PAGE hk OF MC PROJECT: JAF PRELM [] PREPARED BY /9(d DATE 9/ M } i FINAL [X] CHECKED BY // DATE /A/ f Mf7 Control Room Radiological Habitability Under Power (Jprate l TITLE: Conditions and CREVASS Reconfiguration 4.3 Total LOCA Dose t j The total LOCA radiation doses due to both drywell and ESF component leakage are shown in Table 4.4. \'
- The table was prepared l by summing the results in Tables 4.2 and 4.3. Note that, for the 31-day exposures, drywell and MSIV leakage contribute 90.5% of '
the total thyroid dose, and about 98% of the total whole body and skin doses. In addition, the total LOCA radiation dose (drywell'and ESF components) as a function of DMPR-105 closure time is presented in table 4.5. Based on the results of Table 4.5 and the radiation exposure results for a CRDA (presented in Section 6) it i was determined that closure of the damper is not required for l limiting doses to within the GDC-19 criterion of 30 rem to the Thyroid. A maximum DMPR-105 closure delay time of 12 hours was selected from ALARA considerations.
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l l i
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/)# FINAL [X] CHECKED BY //f DATE fofff7 TITLE: Control Room Radiological Habitability Under Power dprate Conditions and CREVASS Reconfiguration Table 4.4 Control Room Radiation Exposures Due to Drywell and ESF Component Leakage Following A Design-Basis Loss of Coolant Accident Time Thyroid Whole Body Skin -
(hours) Dose (rem) Dose (rem) Dose (rem) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 5.00E-01 2.09E-01 6.66E-04 7.93E-03 8.00E+00 3.12E+00 5.23E-03 5.35E-02 1.20E+01 4.21E+00 6.34E-03 6.63E-02 2.40E+01 5.64E+00 8.29E-03 9.33E-02 9.60E+01 8.50E+00 1.06E-02 1.26E-01 7.44E+02 1.11E+01 1.17E-02 1.42E-01 b v Note:
- Closure of DMPR-105 is not required for meeting regulatory limits. However, closure within 12 hours after the accident is recommende'd for ALARA reasons (see Table 4.5 for additional details)
- Continuous occupancy Table 4.5 Control Room Radiation Exposures versus DNPR-105 Closure Delay Time -
DMPR-105 Closure Total Control Room Time (hrs) Thyroid Dose Rate (rem) 0.5 9.12 4.0 9.77 12.0 11.11 24.0 12.71 ~ 168.0 17.14
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744.0 19.54 {
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TITLE: Control Room Radiological Habitability Under Power'dprate Conditions and CREVASS Reconfiguration
- 5. RADIATION EXPOSURES 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 CR radiation exposures associated with a postulated MSLB outside containment was based on data and assumptions consistent with the regulatory guidelines
[specifically, Ref. 20 (Regulatory Guide 1.5), and the Standard Review Plan (Ref. 4, Sec. 15.6.4)]. With the exception of the break location and source term, the analyses are also consistent with those in the updated UFSAR and in the revised response to NUREG-0737, Item III.D.3.4 (Ref. 39). This calculation also addresses an MSLB with an RCS activity (,) concentration equal to 2 pCi/gm I-131 Dose Equivalent (the Technical Specification limit for Limited condition of Operation), with the control room pre-isolated. The following data and assumptions were used to calculate the post-MSLB accident dose to the Control Room: (a) A main steam line break occurs outside containment during full power operation. (b) The main steam isolation valves close in 10.5 seconds after the break (UFSAR, Rev. O, 7/82, Sec. 14.6.1.5.1.e, pg 14.6-29). (Note: Actual closure time is approximately 3 to 5 seconds.) (c) According to the SRP, the potential break locations include the steam tunnel (as in Reg. Guide 1.5) and the entire , turbine building complex. Fecently identified information , ) (Refs. 14 and 15) shows that a break in the 16" bypass line leading to the turbine bypass steam chest would release more Os reactor coolant into the turbine building than a break in 1
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE hT OF #0 PROJECT: JAF PRELM [] PREPARED BY /ht. DATE 9/30/94-( FINAL [X] CHECKED BY j/[ DATE fo/r?/D TITLE: Control Room Radiological Habitability'Under Power 6 prate Conditions and CREVASS Reconfiguration one of the 24" main steam lines in the steam tunnel (see Figs. 5.1 and 5.2). In the present calculation, use was made of the 16" MSLB scenario. (d) As shown in the mass flow vs. time figure (Fig. 5.2) for the 16" line break, there is a change from 100 percent quality steam to two-phase flow at about 2.75 sec after the break. The total mass of steam discharged through the break during the first 2.75 seconds is 4226 (lb/sec) x 2.75 (sec) = 11,621.5 lbs. After the onset of two-phase flow, the total mass of steam and liquid which would be released until the I MSIVs close (at 10.5 seconds after the break) is equal to 13844 (lb/sec) x [(8.533 - 2.75) + (10.5 -8.533)/2] (sec) = 93,675.4 lbs; with a quality of approximately 7% (Ref. 40),
,~, this corresponds to 87,118 lbs of liquid and 6,557.3 lbs of steam. Hence:
Total steam released through the break = 18,179 lbs Total liquid released through the break = 87,118 lbs (e) The ensuing high fuel temperatures do not lead to any fuel damage (UFSAR, Rev. O, 7/82, Pg 14.6-32). (f) The noble gas fission product concentrations in the steam correspond to the design values which would yield the l
)
standard release rate to the atmosphere during normal operation (i.e., 100,000 Ci/sec following a 30-minute l decay). The assumption is made that 50% of all noble gases leaving the reactor vessel during the 10.5-sec MSIV closure time would be released via the break. 1 1 l From Ref. 30 (Table 1), the design-basis release rates for I the roble gases of interest, and the total releases following an MSLB are as follows: 1 l 1 L_________________________...-----
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'~
) FINAL [X] CHECKED BY /3' DATE afi/f7 TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration Source Term Nuclide at t=0 ? *E Release (uci/sec) (Ci)
Kr 83m 3.4E+03 1.82E-02* Kr 85m 6.1E+03 3.27E-02 Kr 85 2.0E+01 1.07E-04 Kr 87 2.0E+04 1.07E-01 Kr 88 2.0E+04 1.07E-01 Kr 89 1.3E+05 6.96E-01 Xe 131m 1.5E+01 8.03E-05 Xe 133m 2.9E+02 1.55E-03 Xe 133 8.2E+03 4.39E-02 Xe 135m 2.6E+04 1.39E-01 Xe 135 2.2E+04 1.18E-01 Xe 137 1.5E+05 8.03E-01 Xe 138 8.9E+04 4.77E-01 3.40E+3 (pCi/sec) x 10.5 (sec) x 50% (flow ( through break) x 1.02 (power adjustment) - x 1.0E-06 (Ci/pCi) (Note: The 1.02 power adjustment factor is not required; it is conservative and was included for consistency with the analysis in Ref. 7) The halogen inventory in the steam was determined 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, Rev. 1, 7/89, Sec. 14.6.1.5.2.b). See also item (g) which follows for a conservative assumption in the amount of iodine released during the two-phase flow.] (g) The halogen source term in the discharged liquid was selected to represent the Tech Spec limits for the g following: (_) i) The maximum RCS concentration under power uprate l II
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE hk OF //O PROJECT: JAF PRELM [] PREPARED BY /rf._ DATE f/Mf)- i FINAL [X] CHECKED BY ~
//[ DATE Mfb/f/
TITLE: Control Room Radiological Habitability Under Power bidate Conditions and CREVASS Reconfiguration equilibrium conditions (0.2 pCi/gm I-131 DE)', ano ii) The maximum RCS concentration under Limiting Concitions of Operation (2 pCi/gm I-131 DE) Note that there is a 10-fold increase in the RCS concentrations between the two cases. The releases of radioactivity presented below are for the equilibrium RCS conditions. For the spiked-RCS case, multiply the released activities by 10. ) i With respect to the release calculation, it was conservatively assumed that the total two-phase flow release , through the break (93,675.4 lbs, or 4.250E47 gm of liquid and steam) would contain iodines at the concentrations equal to those for the liquid phase. The relative coolant activities employed in the determination of the post-MSLB halogen releases were based on the data in Ref. 30 (Table 2), which is as follows: I Nuclide Primary Coolant Activity (uci/am) . Br 83 0.025 Br 84 0.041 Br 85 0.021 I 131 0.027 I 132 0.21 I 133 0.18 I 134 0.38 I 135 0.25 This is the GE Standard Technical Specification limit Q (Ref . 41) , as also used in the original radiological analyses for b/ the power uprate project (Ref. 7) and its associated Technical Specification change request (Ref. 42, pg. 21 of 52). _ 1 m_________ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . _ _ _ ___ __ _ _ _ _ . _ _ . . _ - _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _
i NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE h8 OF //8 PROJECT: JAF PRELM [] PREPARED BY A12 DATE 9/34[9} , FINAL [X) CHECKED BY /2( DATE /o/A/f'7 TITLE: Control Room Radiological Habitability Under Power prhte Conditions and CREVASS Reconfiguration Un/'r equil.4trium conditions, tbe halogen con antrations in the RCS, and the total activities discharged into the turbine building, would be as shown in the table which follows. Reactor Coolant System Activities and Post-MSLB Releases (Based on RCS concentration of 0.2 pCi/gm I-131 DE) Primary ICRP-30 Coolant Activity Activity Nuclide DCF Activity DCF x ACT e.2 pCi/gm DCF x ACT Release - (rem /Ci) (pCi/gm) (Ci) J I-131 1.07E+06 2.700E-02 2.889E+04 8.130E-02 8.699E+04 3.455E+00 I-132 6.29E+03 2.100E-01 1.321E+03 6.323E-01 3.977E+03 2.687E+01* f, I-133 1.81E+05 1.800E-01 3.258E+04 5.420E-01 9.810E+04 2.303E+01 4.066E+02 1.144E+00 1.224E+03 4.863E+01 I-134 1.07E+03 3.800E-01 I-135 3.15E+04 2.500E-01 7.875E+03 7.528E-01 2.371E+04 3.199E+01 _ I-131 DE = 6.642E-02 pCi/gm 2.000E-01 pCi/gm
- 6.323E-01 (pCi/gm) x 4.250E+07 (gm) x 1.0E-06 (Ci/pCi) l The dose conversion factc. (DCFs) in the above table were extracted from ICRP-30 (Ref. 6). The listed primary coolant _
concentrations were adjusted to yield the limit of 0.2 (pCi/gm) I-131 DE according to the following expression: , 1 - 131 D.E. = [ Q, x DCF' DCF,.m i i Where, - 1 Q1 - Coolant concentration in Ci/gm DCFi = Dose conversion factor for i*h species The above concentrations were also used to compute the total
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE h9 OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE TMD FINAL [X) CHECKED BY //f DATE /Mk/f/ TITLE: Control Room Radiological Habitability Under Power 6 prate Conditions and CREVASS Reconfiguration bromine activities which would be released to the turbine building. These are as follows: Nuclide Activity Release (Ci) Br 83 3.199* ~ Br 84 5.247 Br 85 2.687 3.455 (I-131 Ci released, from above) x 0.025 (Br-83 Conc.) / 0.027 (I-131 Conc.) Activation products and other particulate in the coolant
~
were neglected since they would not become airborne. l (h) 100 % of the coolant halogens discharged in the turbine building are assumed to become airborne and released to the g-w atmosphere at ground level over a period of 2 hours. The {
\s%
selected release rate was equivalent to 72 air changes per day, and the cumulative releases to the atmosphere (ignoring buildup and decay) as a function of time would be as follows: Post MSLB Time Cumulative (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 - 1: 0 99.8
,_ (i) The atmospheric dispersion factors associated with the
)
i s transport of released radioactivity to the control room l - intake are as follows [in ( sec/m') , from Ref. 5] : E______________----------
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE P OF //O q PROJECT: JAF PRELN [] PREPARED BY M DATE 1/Je/ff Q TITLE: FINAL [X) CHECKED BY //f DATE M,h/rf Control Room Radiological Habitability Under Power 6 prate Conditions and CREVASS Reconfiguration Interva'- 0-8 hrg o-24 hrs 1-4 davs 4-22 days (X/Q) .e 3.29e-03 2.81e-03 2.00e-03 1.22e-03 ( X / Q ) ,,,,, 4.06e-04 3.48e-04 2.49e-04 1.54e-04 Note the following: i) The concentration (X/Q)s are for defining the radionuclides concentrations at the CR outside air intake. ii) The gamma (X/Q)s are for computing the whole body dose to a receptor at the CR outside air intake due to exposure to finite radioactive clouds above. These _ factors are employed later in Sec. 8 and are included here for convenience. (j ) The control room characteristics and other exposure-related parameters are as described under Items (i) and (j) in Sec. 4.1.1, with the following exception: For the spiked RCS activity, the CR is assumed to be - already in the isolate mode at the time of the accident. This is accomplished through procedural controls implemented through the Fuel Reliability Action Plan (Ref. 43), which require isolation of the control room whenever the RCS I-131 activity exceeds 0.01 pCi/gm. - l For the equilibrium RCS conditions, placement of the ventilation in the " isolate" mode was assumed to take place at a post-accident time (less than 30 minutes) which wo'lld maximize the CR radiation exposures. As demonstrated under
~
Item (h) above, the release of radioactivity to the atmosphere and into the control room following an MSLB is } very fast (relative to the time allotted for manual l
NYPA - CALC.# J'LF-CALC-RAD-0004 2 REV 2 PAGE Cl OF /JD PROJECT: JAF PRELM [] PREPARED BY /hf DATE '7M/9} ( FINAL [X] CHECKED BY /2/ DATE ,jn//7 TITLE: Control Room Radiological Habitability Under Power 6pr/ ate Conditions and CREVASS Reconfiguration isolation of the control room), and therefore contamination i of the air entering the control room will drop significantly in a very short time. Thus, CR personnel exposures would ! increase if the air-exchange rate of the control room is reduced after the majority of the released radioactivity has l already entered the control room (thus trapping the radioactivity within the control room). In view of this possibility, iterative analyses were carried out using various CR isolation times to determine the worst-case ; radiation exposures. The documentation in this calculation I l includes the analyses with CR isolation times equal to 9, 12 l and 15 min. The worst-case exposure occurs for a 12-minute _ isolation time. 1 '{ ) -
%..) b l
l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE O OF '/O - PROJECT: JAF PRELM I ) PREPARED BY M DATE 9/b/9') FINAL [X] CHECKED BY /// DATE fM-5/'y TITLE: Control Room Radiological Habitability Under Power Upfate Conditions and CREVASS Reconfiguration 5.2 Results CR 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 B to this calculation (Computer Run Case #2 for the equilibrium RCS conditions, and Case #3 for the spiked RCS i activity). The DORITA-2 results of interest (namely the thyroid, whole body and skin exposures as a function of post-MSLB time) are summarized in Table 5.1 for the equilibrium RCS conditions, and in Table 5.2 for the spiked RCS case.
~
I On (_) For the equilibrium RCS conditions, the worst-case exposures occur when the CR is isolated 12-minutes after the accident. l Refer to Sec. 2 for a summary of the results. The spiked RCS case yields radiation exposures which are marginally higher than those for the equilibrium case. I - 1 o
l NYPA - CALC.# JAF-C.' LLC-RAD-00042 REV 2 PAGE C3 OF //O g PROJECT: JAF PRELM [] PREPARED BY M DATE Q/.30/f7 FINAL [X] CHECKED BY /f DATE jMn/f7 TITLE: Control Room Radiological Habitability dnder Power 8p/ ate i Conditions and CREVASS Reconfiguration l Table 5.1 l Control Room Radiation Exposures Fol)owing A Design-Basis Main Steam Line Break (Equilibrium RCS activity at 0.2 pCi/gm I-131 DE) 1 1 Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rem) Dose frem) Case A: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.500E-01 1.782E+00 1.951E-03 1.225E-02 8.000E+00 1.565E+01 1.377E-02 8.624F-02 1.200E+01 1.565E+01 1.381E-02 8.652E-02 2.400E+01 1.565E+01 1.389E-02 8.718E-02 9.600E+01 1.565E+01 1.392E-02 8.749E-02 7.440E+02 1.565E+01 1.392E-02 8.749E-02 _ lp Case B () 0.000E+00 2.000E-01 0.000E+00 2.670E+00 0.000E+00 2.885E-03 0.000E+00 1.806E-02 8.000E+00 1.602E+01 1.408E-02 8.812E-02 1.200E+01 1.602E+01 1.411E-02' 8.835E-02 2.400E+01 1.602E+01 1.418E-02 8.893E-02 9.600E+01 1.602E+01 1.420E-02 8.919E-02 7.440E+02 1.602E+01 1.420E-02 8.919E-02 . Case C: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.500E-01 3.541E+00 3.782E-03 2.361E-02 8.000E+00 1.593E+01 1.399E-02 8.752E-02 1.200E+01 1.593E+01 1.401E-02 8.773E-02 2-400E+01 1.593E+01 1.407E-02 8.822E-02 l 9.600E+01 1.593E+01 1.409E-02 8.845E-02 7.440E+02 1.593E+01 1.409E-02 8.845E-02 NOTE:
- Delay time for switching ventilation to accident mode l Case A: 9 minutes Case B: 12 minutes Case C: 15 minutes
+ Continuous occupancy in all cases O
l l
1 NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 5% OF //O . l PROJECT: JAF PRELM [] PREPARED BY ##- DATE 9/Mf I FINAL [X] CHECKED BY //f DATE /J7//f TITLE: Control Room Radiological Habitability Under Power' dpr6 tie ) Conditions and CREVASS Reconfiguration Table 5.2 - Control Room Radiation E;.posures l'owing A Design-Basis Main Steam Line Break ' (Spiked RCS activity at 2 pCi/gm I-131 DE) I l Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rem) Dose (rem) 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.614E+00 2.673E-03 1.658E-02 1.000E+00 6.452E+00 6.063E-03 3.789E-02 2.000E+00 1.194E+01 1.013E-02 6.462E-02 4.000E+00 1.628E+01 1.307E-02 8.652E-02 _ 8.000E+00 1.768E+01 1.496E-02 1.034E-01 1.600E+01 1.778E+01 1.660E-02 1.189E-01 ) O O 2.400E+01 1.778E+01 1.731E-02 1.256E-01 9.600E+01 1.778E+01 1.787E-02 1.311E-01 7.440E+02 1.778E+01 1.788E-02 1.311E-01
~
i I Basis: CR pre-isolated; 1000 cfm through 90% filters, plus 100 cfm unfiltered inleakage m (^) v W___-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ . - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . - - . _ _ _ _ _ _ - _ - - _ - - - - . _ _ . _.___________..-.___-__________________-a
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE CT~ OF f/O PROJECT: JAF PRELN [] PREPARED BY M DATE 9/taf9 7' FINAL [X) CHECKED BY // DATE j/f787 TITLE: Control Room Radiological Habitability Nnder Power dprdte Conditions and CREVASS Reconfiguration Fig. 5.1 - JAF Main Steam Piping System Locations of Postulated 24" and 16" Line Breaks FLOW RESTRICTOR BREAK CHOKES FLOW LOCATION - 10" PIPING CNOKES FLOW x O xx x o 10" PIPING 24" STEAMLINES MAIN STEAM ISOLATION VALVES ' 'q BREAK 16" STOP
' LOCATION
, PIPIN VALVES TUR8INE BYPASS , , STEAM CHEST-l y TO CONDENSERS l O
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 78 OF /lO PROJECT: JAF PRELM [] PREPARED BY /dvil DATE 9/k/Q
,C FINAL [X] CHECKED BY // DATE ,/,$/97 TITLE: Control Room Radiological Habitability Under Power Op/ ate Conditions and CREVASS Reconfiguration Fig. 5.2 -
JAF Main Ste1un Piping System Break Flows from a 24" and a 16" Line Break 13844 lb/sec (16" break) l l\ l i1 l i.1 t l ., -- l
' l i
\
,\
\ l\
l ( _ 8492 lb/sec (24" break) l '
\ ,
' i 8 l I
\ l 2r l
= '
\
I E \ C l i \ j g l 1 . E '
\
4226 lb/sec I I I I I i I 2548 lb/sec l l l l ' l ' 8 l i i i i i i _ 0 2.75 7.5 8.533 10.5 Post Accident Time (sec)
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE N OF //0 7-~ PROJECT: JAF PRELM [] PREPARED BY NL DATE 9/3D/91F () TITLE: FINAL [X] CHECKED BY Control Room Radiological Habitability Under Power'U rate
//[ DATE /0//7/f2 p
Conditions and CREVASS Reconfiguration
- 6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT 6.1 Basic Data and Assumptions All assumptions and data employed in the analysis of a CRDA are consistent with the guidance in the Standard Review Plan (Ref. 4, Sec. 15.4.9), applicable portions of Regulatory Guide 1.77 (Ref.
23), the JAF UFSAR, and JAF-CALC-RAD-00041 (Ref. 17). 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, Rev. O, 7/82, Sec. 14.6.1.2.4, this assumption means that the reactor was shut down from design power, taken critical, and brought to the initial temperature and pressure 7-s conditions within 30 minutes of the departure from design (_s/ power. (b) Prior to the accident, the reacto'r power was at the level for design-basis accident analyses (i.e., 2586.5 MWt, from Ss.c. 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 and leads to the failure of 850 fuel rods (Ref. 44, Sec. 6.2.1, and Ref. 45, Sec. 3.7). The total number of fuel rods in the core ir equal to 36472 (Ref. 7); this number varies slightly from reload to reload, as the ! l assemblies are being gradually switched from 8x8 to 9x9 designs, and has little impact on the final results. (Note: According to Sec. 14.6.1.2.4 of the UFSAR, Rev. O, 7/82, the total number of fuel rods that were assumed to fail in the original a..a_ysis was 330.) (d) The failed fuel rods are at a core location with a radial peaking factor of 1.5 (Ref. 4, SRP, Sec. 15.4.9). (~ (_/ (e) All activity within the gaps of the failed fuel rods is t { l E_-_________________--_-- _ _ _ - - - - - - - - _ _ _ - - - - _ _ j
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE % OF //0 PROJECT: JAF PRELM [] PREPARED BY #L DATE 9/30/9 FINAL [1] CHECKED BY /6' DATE ./o///[7 f/ TITLE: Control Room Radiological Habitability Under Power prate Conditions and CREVASS Reconfiguration III I IIII I I III EE I I I I released to the reactor coolant and is instantly and uniforn'y mixed with the coolant in the pressure vessel at the time of the accident. The released activity is conservatively assumed to correspond to 10% of all halogens and 10% of all noble gases (30% for Kr 85) in each failed rod (Ref. 23, as recommended in the SRP) . (f) Based on the above information, and without taking credit for the pre-accident decay time of 30 minutes 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: () Multiolvino factor for all noble cases exceot Kr 85 1.5 (peaking factor) x (850 failed rods / 36472 rods) x 0.1 (gap fraction) = 3.496E-03 Multiolvino factor for Kr 85 1.5 (peaking factor) x (850/36472) x 0.3 (gap fraction)
= 1.049E-02 Multiolvina factor for all halocens 1.5 (peaking factor) x (850/36472) x 0.1 (gap fraction) x 0.1 (fraction reaching turbine / condensers)
= 3.496E-04 O
l i l I
l 1 i ! NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 77 OF llO l PROJECT: JAF PRELM [] PREPARED BY /4(C DATE SMf). ! FINAL [X] CHECKED BY //f DATE / d 7 /f ;F !
\~ TITLE: Control Room Radiological Habitability Dnder Power Upra'te Conditions and CREVASS Reconfiguration j Post-CRDA Haloaens and Noble Gases Reachina the Condensers Activity Activity Nuclide (Ci) Nuclide (Ci)
Br 83 2.824E+03* Kr 83m 2.836E+04 Br 84 5.007E+03 Kr 85m 6.089E+04 Br 85 6.001E+03 Kr 85 8.179E+03 I Kr 87 1.168E+05 l I 129 7.881E-04 Kr 88 1.655E+05 I 130 9.458E+02 Kr 89 2.058E+05 I 131 2.379E+04 l I 132 3.477E+04 Xe 131m 1.430E+03 l I 133 4.975E+04 Xe 133m 2.084E+04 l I 134 5.476E+04 Xe 133 4.998E+05 _ l I 135 4.697E+04 Xe 135m 9.422E+04 i I 136 2.265E+04 Xe 135 6.456E+04 Xe 137 4.387E+05 Xe 138 4.168E+05 O v
- 8.078E+06 (from Table 4.1) x 3.496E-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 (Ref. 16)] 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. . s I (h) As a result of plant shutdown following a CRDA, or as a l result of offgas system automatic isolation5 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 5 offgas system isolation takes place automatically following a 15-minute delay. This delay time is shorter than the transit time (22 min, UFSAR Sec. 11.4.4.2) of offgas effluent to
/' the main stack via the 24" hold-up pipe, under startup
(_)/ conditions. _s____ _ _ _ _ _ _ . . _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ . _ _
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE bO OF fl O PROJECT: JAF PRELM [] PREPARED BY /8(E. DATE 9/30/9h FINAL [X] CHECKED BY // DATE n// /f/ TITLE: Control Room Radiological Habitability'Under Power prate Conditions and CREVASS Reconfiguration 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. 4 (SRP, Sec. 15.4.9), Ref. 44 (Sec. 6.3.1.1), and Ref.46). [ 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. 23). The release to the atmosphere is at ground level and there is no holdup within the turbi'ne building. (k) Transfer of the released radioactivity to the control room () is governed by the accident dispersion parameters described under Item (i) in Sec. 5.1 of this calculation. (1) The control room characteristics and other exposure-related parameters are as described under Items (i) and (j) in Sec. 4.1.1. As documented in the results of this section, closure of DMPR-105 is not required for keeping the control room personnel exposure below the GDC-19 exposure limits. As was the case with all other accidents analyzed in this calculation, closure of DMPR-105 was assumed to be implemented at 12 hrs after the accident. The 12-hour delay time was selected based on ALARA considerations. I O
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 6l OF NO PROJECT: JAF PRELM [] PREPARED BY /#L DATE 9/)D[99 FINAL [X) CHECKED BY /ff DATE fM 787 TITLE: Control Room Radiological Habitability'Under Power Dp/ ate Conditions and CREVASS Reconfiguration 6.2 Results I CR radiation exposures following a design-basis CRDA were calculated using the DORITA-2 computer code and the data and assumptions listed above. Copies of the DORITA-2 outputs appear in Attachment B to this calculation (Computer Run Case #4). 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 summary of the exposures. Table 6.1 includes CR dose as a function of DMPR-105 closure time.
~
The results show that closure of the ventilation damper is not required to limit the
. post-CRDA exposures within the GDC-19 criterion of 30 rem to the is,/ Thyroid. However, based on ALARA considerations, a maximum DMPR-105 closure delay time of 12 hours was selected.
The Control Room dose as a function of DMPR-105 closure time for j the CRDA and a LOCA are shown in Figure 6.1. It can be seen that the 31 day Control Room dose not exceed the design criterion of l 30 rem for both accidents, even when the damper is not closed for f the entire 31 day period. Ii
)
l i i l 1 O 4 L__----------------_-------------- -- -- -- -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 62- OF //O PROJECT: JAF PRELM [] PREPARED BY A8. DATE 9/fdD FINAL [X] CHECKED BY /[f DATE jd 7// TITLE: Control Room Radiological Habitability Under Power 8pr'a'e/ t Conditions and CREVASS Reconfiguration Table 6.1 Control Room Radiation Exposures Following ~ A Design-Basis Control Mod Drop Accident Time Thyroid Whole Body Skin (hours) Dose frem) Dose frem) Dose frem) Case 1: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E-01 5.860E-02 6.923E-04 1.133E-02 - 5.300E-01 6.715E-01 5.061E-03 6.283E-02 1.000E+00 1.418E+00 9.096E-03 1.054E-01 8.000E+00 5.904E+00 3.304E-02 3.441E-01 2.400E+01 1.195E+01 4.844E-02 5.399E-01 9.600E+01 1.239E+01 4.985E-02 5.573E-01 7.440E+02 1.239E+01 4.986E-02 5.574E-01 Caue 2: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E-01 5.860E-02 6.923E-04 1.133E-02 O 5.000E-01 1.000E+00 4.000E+00 6.715E-01 1.435E+00 5.078E+00 5.061E-03 9.115E-03 2.361E-02 6.283E-02 1.056E-01 2.487E-01 8.000E+00 7.825E+00 3.362E-02 3.484E-01 2.400E+01 1.390E+01 4.902E-02 5.442E-01 9.600E+01 1.434E+01 5.043E-02 5.617E-01 7.440E+02 1.434E+01 5.044E-02 5.618E-01 Case 3: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E-01 5.860E-02 6.923E-04 1.133E-02 5.000E-01 6.715E-01 5.061E-03 6.283E-02 1.000E+00 1.435E+00 9.115E-03 1.056E-01 8.000E+00 9.571E+00 3.416E-02 3.519E-01 1.200E+01 1.331E+01 4.005E-02 4.167E-01 2.400E+01 1.832E+01 4.999E-02 5.498E-01 9.600E+01 1.876E+01 5.141E-02 5.673E-01
~
7.440E+02 1.876E+01 5.141E-02 5.673E-01 Case 4: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E-01 5.860E-02 6.923E-04 1.133E-02 5.000E-01 6.715E-01 5.061E-03 6.283E-02 1.000E+00 1.435E+00 9.115E-03 1.056E-01 8.000E+00 9.571E+00 3.416E-02 3.519E-01 2.400E+01 2.337E+01 O 9.600E+01 7.440E+02 2.438E+01 2.438E401 5.101E-02 5.246E-02 5.246E-02 5.560E-01 5.737E-01 5.738E-01
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE f.3 OF t/O PROJECT: JAF PRELM [] PREPARED BY % DATE 9/Jo/N FINAL [X] CHECKED BY [8 ' DATE /44h/f7 TITLE: Control Room Radiological Habitability Under Power 8p/ ate Conditions and CREVASS Reconfiguration Table 6.1 (Continued ...) Control Room Radiation Exposures Follo..Ang A Design-Basis Control Rod Drop Accident Time Thyroid Fhole Body Skin
.(hours) Dose frem) Dose (rem) Dose (rezel Case 5:
0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E-01 5.860E-02 6.923E-04 1.133E-02 5.000E-01 6.715E-01 5.061E-03 6.283E-02 1.000E+00 1.435E+00 9.115E-03 1.056E-01 8.000E+00 9.571E+00 3.416E-02 3.519E-01 2.400E+01 2.337E+01 5.101E-02 5.560E-01 9.600E+01 2.379E+01 5.230E-02 5.683E-01 - 1.680E+02 2.379E+01 5.230E-02 5.684E-01 7.440E+02 2.379E+01 5.230E-02 5.684E-01 Cane 6: 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.000E-01 5.860E-02 6.923E-04 1.133E-02 5.000E-01 6.715E-01 5.061E-03 6.283E-02 1.000E+00 1.435E+00 9.115E-03 1.056E-01 8.000E+00 9.571E+00 3.416E-02 3.519E-01
~
2.400E+01 2.337E+01 5.101E-02 5.560E-01 9.600E+01 2.379E+01 5.230E-02 5.683E-01 7.440E+02 2.37,_-01 5.230E-02 5.684E-01 Note: DMPR-105 closure at various delay times after the . accident: Case 1 - 0.5 hr closure delay time Case 2 - 4 hr closure delay time Case 3 - 12 hr closure delay time Case 4 - 24 hr closure delay time Case 5 - 168 hr closure delay time Case 6 - 744 hr closure delay time Continuous occupancy in all cases - O
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1 NYPA - CALC.# JAl-CALC-RAD-00042 REV 2 PAGE df OF // O PROJECT: JAF PRELM [] PREPARED BY /M2. DATE 9/3D/D FINAL (Z) CHECKED BY DATE f7/n/97
///
TITLE: Control Room Radiological Habitability Under Fower Uprate Conditions and CREVASS Reconfiguration 7.0 RADIATION EXPOSURES FROM A REFUELING ACCIDENT 7.1 Basic Pata and Assumptions The assumptions and data listed below were used in the analysis of a design-basis refueling accident. All assumptions are consistent with the guidance in the Standard Review Plan (Ref. 4, Sec. 15.7.4), Regulatory Guide 1.25- (Ref. 21), and the UFSAR. (a) The reactor has been operating at full power (2586.5 MWt) for an extended period of time (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 O d shutdown (Ref. 4, SRP, Sec. 15.7.4). (d) The accident involves a fuel assembly dropping from the maximum height allowed by the fuel handling equipment. A total of 125 fuel rods are ruptured. This is a conservative number based on information in Ref. 45, Sec. 3.8; also, according to the UFSAR, Rev. O, 7/82, Sec. 14.6.1.4.2, the total number of fuel rods that fail during a refueling accident is 111. The total number of fuel rods in the core is equal to 36472 (Ref. 7). 1 (e) The failed fuel rods are at a core location with a radial peaking factor of 1.5 (Ref. 21, Reg. Guide 1.25). (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 129) t..J 10% of all noble gases (except Kr BS) in each failed rod, and to 30% of I 129 and Kr 85 (Re2. 21). (g) The nuble gas and balogen inventories released to the fuel pool (prior to adjustment fo2. decay from the time of reactor shutdown, which is handled by the DORITA-2 computer code) , t
NYPA - CALC.# OAF-CALC-RAD-00042 REV 2 PAGE Sh OF //C - [, PROJECT: JAF PRELM [] PREPARED BY //R DATE 9/ M 9-FINAL [X] CHECKED BY ff DATE f3-//f7 TITLE: Control Room Radiological Habitability 'Under PoweE ITplate Conditions and CREVASS Reconfiguration are as shown in the table which follows. They were computed by multiplying the core inve ntory in Table 4.1 of this calculation by the following factors: Multiplying factor for all noble gases except Kr 85: 1.5 (peaking factor) x (125 failed rods / 36472 rods) x 0.1 (gap fraction) = 5.141E-04 Multiplying factor for Kr 85: 1.5 (peaking factor) v (125/36472) x 0.3 (gap fraction)
= 1.542E-03 Multiplying factor for all halogens except I 129:
~
1.5 (peaking factor) x (125/36472) x 0.1 (gap fraction)
= 5.141E-04 Multiplying factor for I 129:
1.5 (peaking factor) x (125/364.72) x 0.3 (gap fraction)
= 1.542E-03 V
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE d} OF //O _ { s PROJECT: JAF PRELM [] PREPARED BY M DATE f/M($
) FINAL [X] CHECKED BY g DATE j,6 // 7 TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration i
Pre-Decay Refuelina Accident Source Term
~
l i Nuclide Activ.(Ci) Nuclide Activ. (Ci) I Br 83 4.153E+03* Kr 83m 4.171E+03 Br 84 7.363E+03 Kr 85m 8.954E+03 Br 85 8.825E+03 Kr 85 1.203E+03 Kr 87 1.718E+04 I 127 3.477E-03 Kr 88 2.433E+04 I 130 1.391E+03 Kr 89 3.026E+04 I 131 3.498E+04 i I 132 5.113E+04 Xe 131m 2.104E+02 ' I 133 7.316E+04 Xe 133m O I 134 8.053E+04 Xe 133 3.065E+03 7.351E+04 I 135 6.908E+04 Xe 135m 1.386E+04 I 136 3.331E+04 Xe 135 9.494E+03 Xe 137 6.452E+04 Xe 138 6.130E+04 8.078E+06 (from Table 4.1) x 5.141E-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. 21). 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. 21). f-~ (i) The retention of noble gases by the pool water is negligible !
\_)g (i.e., noble gas DF = 1).
~
l l l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE (( OF //O PROJECT: JAF PRELM [] PREPARED BY /8VC. DATE 9/a/M. FINAL [X] CHECKED BY Af DATE ,2 5/// - TITLE: Control Room Radiological Habitability Under Power Dprate Conditions and CREVASS Reconfiguration (j) Radioactive material that escapes the pool is released to the atmosphere via the SGTS anc main stack over a 2 hour period (Ref. 21). The RB air exchange rate was arbitrarily set at the conservative value of 72 air changes per day. At this release rate, 99.8 % of all radioactivity would be released to the SGTS within the assumed 2 hour period. (Refer to Sec. 5.1, Item (h) for tabulation of the cumulative release as a function of time.) The actual RB air exchange rate (at the nominal SGTS flow of 6000 scfm) is
~
only 3.3 volumes per day. (k) The halogen-removal filter efficiency of the SGTS is 90% for all halogen species. The Control Room doses were also evaluated assuming no SGTS filtration of halogens, for informational purposes. q\m/ (1) All releases to the atmosphere are via the main stack. Transport of the released radioactivity to the control room intake is dictated by the atmospheric dispersion factors presented under Item (h) of Sec. 4.1.1. (m) The control room characteristics and other exposure-related parameters are as described in Items (i) and (j ) of Sec. 4.1.1, along with the exceptions listed in Sec. 5.1 (MSLB), Item (j) for placing the ventilation system in the " isolate" mode. Note that the fractional release rates to the atmosphere following an MSLB and a refueling accident are the same (72 air changes per day). The refueling-accident analyses documented in this calculation were limited to the l worst-case scenario determined for the MSLB, namely with a _ 12-minute CR isolation. O 4 w____--__-__ . - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ , - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . - . . . _ . _ . . _ _ _ - - . _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - - _ - _ . - - - . _ _ . _ . . _ _ _ _ . _ . _ - _ _ _ _ - _ _ _ - _ _ - _ -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE [7 OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/ M FINAL [X] CHECKED BY // DATE,2,fM7 - TITLE: Control Room Radiological Habitability Under Power 6 prate Conditions and CREVASS Reconfiguration 7.2 Results CR radiation exposures following a design-basis refueling accident were calculated using the LORITA-2 computer code and the data and assumptions listed above. Copies of the DORITA-2 outputs appear in Attachment B to this calculation (Run Cases #1 and #2). The DORITA-2 results of interest (namely the thyroid, whole body and skin exposures as a function of post-accident time) are summarized in Table 7.1. Refer to Sec. 2 for a summary of the radiation exposures. O . 4 u
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE % OF //O g PROJECT: JAF PRELM [] PREPARED BY M DATE 9/gfp i Q . TITLE: FINAL [X] CHECKED BY A DATE fMd/f7 Control Room Radiological Habitability'dnder Power dp/ ate Conditions and CREVASS Reconfiguration Table 7.1 Control Room Radiation Exposures Following A Design-Basis Refueling Accident Time Thyroid Whole Body Skin (hours) Dose frem) Dose (rem) Dose (rem) Case 1: - 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E-01 3.378E-03 3.015E-05 4.772E-04 8.000E+00 2.060E-02 2.157E-04 3.311E-03 1.200E+01 2.060E-02 2.207E-04 3.358E-03
^
2.400E+01 2.060E-02 2.259E-04 3.417E-03 9.600E+01 2.060E-02 2.275E-04 3.439E-03 7.440E+02 2.060E-02 2.276E-04 3.440E-03 _ Case 2: 0.0008+00 0.000E+00 0.000E+00 0.000E+00 2.000E-01 3.378E-02 3.140E-05 4.902E-04 8.000E+00 2.060E-01 2.230E-04 3.387E-03 1.200E+01 2.060E-01 2.279E-04 3.434E-03 2.400E+01 2.060E-01 2.331E-04 3.493E-03 _ 9.600E+01 2.060E-01 2.348E-04 3.515E-03 7.440E+02 2.060E-01 2.348E-04 3.515E-03 Note:
- Case 1: SGTS filter efficiency of 90% (base case)
Case 2: No SGTS filtration credit assumed 1
- DMPR-105 closure at 12 hours after the accident -
(recommended maximum delay time)
- Continuous occupancy in al] cases j O
V
I NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE T/ OF //O {m PROJECT: JAF PRELM [] PREPARED BY /ff DATE 9[ $ FINAL [X) CHECKED BY _gf DATE ((/fh/pf'7 TITLE: Control Room Radiological Habitability Under Power'Up/ ate - Conditions and CREVASS Reconfiguration j l
- 8. RADIATION EXPOSURES FROM EXTERNAL SOURCES This section addresses the post-accident external sources of radiation affecting the habitability of the JAF Control Room, _
specifically: a) Airborne radioactivity accumulating on the refueling level of the reactor building, b) Overhead radioactive clouds, and c) Halogens accumulating on the CR charcoal filters. _ l l The approach employed in this assessment was not to provide a detailed analysis for each accident scenario,.but rather to I n demonstrate that these sources are relatively insignificant in b comparison to the immersion sources. For instance, doses from overhead clouds were first determined for a receptor located on l the roof of the control room for each of the' design-basis accidents; the worst-case accident was then evaluated further to compute the radiation doses to CR personnel. The radiation dose rates and u_mulative doses documented in this section are to air (rad) and may be applied to the whole body (rem). 8.1 Direct Shine from Post-LOCA Airborne Radioactivity in the Reactor Building Airborne radioactivity within the refueling level of the reacror building can result from either post-LOCA drywell leakage, post-LOCA ESF component leakage, or from a Refueling Accident. Since the latter two are relatively insignificant with respect to post-J LOCA drywell leakage (See Table 2.2), they were not considered in I L________________________________.__________________________________-.________________
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE # 2.-. OF //O 3 PROJECT: JAF PRELM [] PREPARED BY #d DATE 4/fo/99-FINAL [X) CHECKED BY ff DATE /o(Mf7
,- TITLE: Control Room Radiological Habitability Under Power #Uprate Conditions and CREVASS Reconfiguration the assessment of radiation emanating from the RB refueling level.
8.1.1 Basic Data and Assumptions Of interest here are the definition of the gamma 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 described in Sec. 4 of this calculation. The items of interest in the q definition of the airborne radioactivity within the reactor
\s / building and the associated time-dependent gamma spectra are as follows: .
(a)
~
A LOCA takes place at full power (2586.5 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; this conservatively includes the 0.23 % per day MSIV leakage component, which in fact goes directly to the stack via the _ SGTS without mixing in the reactor building. (f) As a result of the ventilation system, airborne ("%g radioactivity leaking from the drywell becomes uniformly
\w / distributed within the air volume of the reactor building k
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE Y.3 Or //o l PROJECT: JAF PRELM [] PREPARED BY A12. DATE f/Je/N
) FINAL [X] CHECKED BY /F DATE fc/f7/f7 \~'l ' TITLE: Control Room Radiological Habitability nder Power Dphate Conditions and CREVASS Reconfiguration (2.6E+06 ft', Ref. 47).
(g) Releas from the reactol building is through the SGTS and the main stack at the rate of 6000 scfm (or 3.3 air changes per day). l l Source / Receptor Geometry: The source / receptor geometry for use with QAD-CGGP is shown in Fig. 8.1. The primary components are the refueling level and the control room. The refueling level was represented by a box with no roof and no side walls; the box dimensions are 125' (W) x 162' (L) x 60' (H) and the total volume is 1.215E+06 ft', or 3.440E+04 m'. - The concrete floors beneath the refueling level source were l (I'h l
,/ conservatively represented by a single 18" concrete slab. The thickness of the RB wall below the refueling level is 24" and the CR walls and roof are 30" thick. The concrete density was set at 2.35 g/cc and had the following composition (in weight percent, j from Ref. 48, Vol. II, Table 9.1.12-77):
Fe: 1.19 H : 0.85 0 : 50.64 Mg: 0.23 Ca: 8.03 Na: 1.66 Si: 30.49 A3 : 4.44 S : 0.12 K : 1.87 It is clear from Fig. 8.1 that the direct-shine dose rate from the refueling level to a receptor in the control room would increase with distance from the S wall of the CR, will reach a maximum at some point, and will then decrease. Based on preliminary scoping nulyses, a set of 6 receptors was finally selected for presentation in this calculation, as shown in Fig. 8.1. For added conservatism, the receptors were assumed to be (~'\ ( j along the centerline of the refueling level.
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE D OF //O PROJECT: JAF PRELN [] PREPARED BY M DATE 87/M@ O. FINAL [X] CHECKED BY 8 DATE fod-)/f7 k,) TITLE: Control Room Radiological Habitability Under Power Dpfate Conditions and CREVASS Reconfiguration Fig. 8.1 - QAD-CGGP Source / Receptor Geometry j y axis (Direct Shine from t'2e RB Refueling Floor) y=162' CONTROL ROOM i Receptors (1-6) eooeo y=81' Source Region 1 (
- x axis (0,0) z axis b
z=129.5' REFUELING LEVEL l
~ i Source Region 1 El. 369'_6" z=69.5' l
b18"concretefloor M 24" wall 30" roof and walls z=22' i
~
Receptors l El. 300' a o o o e o-z=4.5' - x axis (0,0) m s ~ U
- n L L L L_.___.__ .__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . . . _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
~
i i NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE W OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 930/94 FINAL [E] CHECKED BY // DATE Mf7/f/ TITLE: Control Room Radiological Habitability Under Power Up/ ate Conditions and CREVASS Reconfiguration 8.1.2 Results Post-LOCA direct-shine radiation levels in the CR from radioactivity accumulating on the RB refueling level are shown in Table 8.1. These were extracted from QAD-CGGP and MARIO Run Case
#1 in Attachment B. The gamma spectra were extracted from DORITA-2 Run Case #6. It is seen that the worst-case location is
! Receptor #3, with a maximum dose rate of 1.2 (mrad /hr) and a 31-day continuous occupancy cumulative dose of 14 mrad. 1 i l I ! u l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE D OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/d99 FINAL [X] CHECKED BY Af DATE fc47/f'7 TITLE: Control Room Radiological Habitability Under Power 6 prate Conditions and CREVASS Reconfiguration Table 8.1 Direct-Shine Radiation Levels to CR Receptors from post-LOCA Airborne Radioactivity Accumulating in Refueling Level of the Reactor Building Time Dose Rate Int. Dose Cum. Dose (hrs) (rad /hr) (rad) (rad) Receptor #1 S wall, contact 0.0 0.0000E+00 0.0000E+00 0.0000E+00 0.5 1.9253E-07 4.8133E-08 4.8133E-08 1.0 2.7384E-07 1.1540E-07 1.6354E-07 2.0 3.3652E-07 3.0411E-07 4.6764E-07 8.0 1.7149E-07 1.4688E-06 1.9365E-06 24.0 1.0803E-08 9.2993E-07 2.8664E-06 _ 48.0 6.3746E-10 8.6207E-08 2.9526E-06 96.0 1.7423E-11 O 168.0 240.0 1.5287E-12 3.7416E-13 8.2678E-09 4.7029E-10
- 5. 906 0E -11 2.9609E-06 2.9613E-06 2.9614E-06 744.0 3.2755E-14 7.0647E-11 2.9615E-06 Receptor #2 6 ft from S wall ,
0.0 0.0000E+00 0.0000E+00 0.0000E+00 0.5 2.7704E-04 6.9259E-05 6.9259E-05 1.0 4.0625E-04 1.6877E-04 2.3803E-04 2.0 5.3905E-04 4.6952E-04 7.0755E-04 8.0 3.6289E-04 2.6710E-03 3.3786E-03 24.0 4.2866E-05 2.3971E-03 5.7757E-03 48.0 5.1389E-06 4.2685E-04 6.2026E-03 96.0 6.5160E-07 1.0430E-04 6.3069E-03 - 168.0 1.7370E-07 2.6026E-05 6.3329E-03 240.0 1.0280E-07 9.7318E-06 6.3426E-03 744.0 1.1955E-08 2.1279E-05 6.3639E-03 O
~
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE N OF //O
- f. ~ PROJECT: JAF PRELM [] PREPARED BY /#l DATE f/M94 .
(' . TITLE: FINAL [X) CHECKED BY /// DATE /g6k/[/ Control Room Radiological Habitability Under Power Upr' ate Conditions and CREVASS Reconfiguration Table 8.1 (Continued) Time Dose Rate Int. Dose Cum. Dose (hrs) (rad /hr) (rad) (rad) Receptor #3 12 ft from S wall i 0.0 0.0000E+00 0.0000E+00 0.0000E+00 1 0.5 6.3697E-04 1.5924E-04 1.5924E-04 1.0 9.3312E-04 3.8782E-04 5.4707E-04 - 2.0 1.2353E-03 1.0772E-03 1.6242E-03 4 8.0 8.1985E-04 6.0806E-03 7.7048E-03 24.0 9.2617E-05 5.3359E-03 1.3041E-02 48.0 1.0505E-05 9.0539E-04 1.3946E-02 96.0 1.2244E-06 2.0725E-04 1.4153E-02 168.0 3.0428E-07 4.7583E-05 1.4201E-02 240.0 1.7522E-07 1.6837E-05 1.4218E-02 (~') 744.0 2.0272E-08 3.6208E-05 1.4254E-02
%J ~
Receptor #4 18 ft from S wall 0.0 0.0000E+00 0.0000E+00 0.0000E+00 0.5 5.2428E-04 1.3107E-04 1.3107E-04 1.0 7.6754E-04 3.1910E-04 4.5017E-04 _ 2.0 1.0145E-03 8.8528E-04 1.3354E-03 8.0 6.6740E-04 4.9732E-03 6.3086E-03 ; 24.0 7.3488E-05 4.3071E-03 1.0616E-02 , 48.0 8.0910E-06 7.1136E-04 1.1327E-02 l 96.0 9.0093E-07 1.5723E-04 1.1484E-02 168.0 2.1571E-07 3.4513E-05 1.1519E-02 240.0 1.2230E-07 1.1852E-05 1.1531E-02 744.0 1.4108E-08 2.5248E-05 1.1556E-02 O .
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE N OF //O PROJECT: JAF PRELM [] PREPARED BY /dC DATE_9/M N FINAL [X] CHECKED BY ff DATEjigf/r7
.- TITLE: Control Room Radiological Habitability Under Power Uprate -
Conditions and CREVASS Reconfiguration Table 8.1 (Continued) Time Dose Rate Int. Dose Cum. Dose (hrs) (rad /hr) (rad) (rad) - Receptor #5 24 ft from S wall 0.0 0.0000E+00 0.0000E+00 0.0000E+00 0.5 4.1942E-04 1.0485E-04 1.0485E-04 1.0 6.1362E-04 2.5519E-04 3.6004E-04 2.0 8.0966E-04 7.0712E-04 1.0672E-03 8.0 5.2791E-04 3.9526E-03 5.0198E-03 - 24.0 5.6660E-05 3.3783E-03 8.3981E-03 48.0 6.0569E-06 5.4318E-04 8.9413E-03 96.0 6.4391E-07 1.1592E-04 9.0572E-03 168.0 1.4845E-07 2.4312E-05 9.0815E-03 240.0 8.2786E-08 8.0956E-06 9.0896E-03 (N 744.0 9.5189E-09 1.7072E-05 9.1067E-03 O Receptor #6 30 ft from S wall 0.0 0.0000E+00 0.0000E+00 0.0000E+00 0.5 3.2483E-04 8.1208E-05 8.1208E-05 1.0 4.7492E-04 1.9757E-04 2.7878E-04 2.0 6.2554E-04 5.4678E-04 8.2555E-04 , 8.0 4.0424E-04 3.0412E-03 3.8667E-03 24.0 4.2316E-05 2.5659E-03 6.4326E-03 ~ 48.0 4.3968E-06 4.0192E-04 6.8345E-03 96.0 4.4660E-07 8.2909E-05 6.9174E-03 168.0 9.9214E-08 1.6626E-05 6.9340E-03 240.0 5.4399E-08 5.3694E-06 6.9394E-03 744.0 6.2339E-09 1.1206E-05 6.9506E-03
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NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE M OF //O
, PROJECT: JAF PRELM [] PREPARED BY M DATE 9/W99 FINAL [X] CHECKED BY /U DATE f[A f/f7 TITLE: Control Room Radiological Habitability'Under Power Uprate Conditions and CREVASS Reconfiguration 8.2 Direct Shine from Post-LOCA Overhead Clouds l
l Dose rates and cumulative doses inside the control room due to gamma radiation emanating from overhead clouds were determined in a two-step approach, as follows: (a) Radiation fields were first computed (using DORITA-2) for a receptor located on the roof of the CR (at El. 322') for each of the design-basis accidents. (b) The worst-case accident was then selected and the - radiation fields were calculated using ELISA, QAD-CGGP and MARIO for receptors on the roof of the control room (for comparative analysis and validation purposes) and inside the control room. The basic data and assumptions for a DORITA-2 evaluation of -
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the whole body gamma radiation doses to a receptor on the roof of the control room under the four design-basis accident scenarios are as described in Secs. 4 through 7. Note in particular that use was also made of the finite-cloud gamma (X/0) s listed in Sec. 4.1.1 Item (h) and Sec. 5.1 It .a (i), and that the doses are to air and not to the whole body. From DORITA-2 Computer Run Case -
#7 in Attachment B, the 31-day _ doses to such a receptor are as a follows:
1 l Desian-Basis Accident Air Dose (Rad) LOCa (Drywell and MSIV Leak) 8.958E-01 - LOCA (ESF Component Leakage) 2.979E-02 Main Steam Line Break 2.303E-02 Control Rod Drop 1.353E-01 Refueling Accident 1.038E-02 j
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE QO OF //O PROJECT: JAF PRELM [] PREPARED BY /dd. DATE 9/32[9 ) t FINAL [X] CHECKED BY ff DATE ,'/f W 7 .
, TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration The above results provide the following important information:
(a) The radiation levels are fairly low and of little concern, and (b) The worst-case accident is the LOCA (drywell and MSIV ' leakage components). In view of this determination, and as a bounding case, assessment of the radiation levels in the control room from overhead clouds (using QAD-CGGP) was limited to post-LOCA drywell and MSIV leakage. Details of the analyses are presented in the subsections which follow. 8.2.1 Basic Data and Assumptions O (s ,/ Of interest here are the definition of the gamma spectra associated with the radioactive cloud resulting from post-LOCA drywell and MSIV leakage (using the ELISA code), and the source / receptor geometry (for input to QAD-CGGP) . These are addressed below. Cloud Ge.:mna Spectra The information needed for definition of the gamma spectra associated with post-LOCA drywell-leakage was extracted from Sec. 4.1.1 of this calculation. The data is the same as that used in the DORITA-2 computer run, with one exception. The concentration (X/Q) s were replaced with the finite-cloud gamma (X/Q) s to yield , the gamma spectra (in MeV/sec-m') associated with a semi-infinite i cloud with uniform concentration; use of the concentration (X/Q)s would have yielded the gamma spectra associated with that portion I s of the radioactive cloud in the vicinity of the CR outside air k_) intake. [ Note that DORITA-2 and ELISA make use of the gamma m
W NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE @/ OF //C PROJECT: JAF PRELM [] PREPARED BY M DATE 1/_b/9')- FINAL [X] CHECKED BY gf DATE g/;yf7 r TITLE: Control Room Radiological Habitability Under Power Uprate ! Conditions and CREVASS Reconfiguration - (X/Q)s to convert a finite radioactive cloud to an equivalent semi-infinit cloud that woulm yield the same dose as the finite cloud.] l l Source / Receptor Geometry Through scoping analyses for assessment of convergence, the size
- of the radioactive plume was' selected to be a rectangular box with dimensions 1200 m (L) x 1200 m (W) x 600 m (H). The CR was represented as a centrally-located square box below the cloud.
The CR concrete walls and roof are 30" thick and the concrete density was set at 2.35 g/cc (same composition as given Sec. . 8.1.1).
~g The source / receptor geometry for use with QAD-CGGP is shown in l \, s/ Fig. 8.2. Note that, in view of the selected symmetry, the geometry was reduced to a single quadrant. Receptors were selected on the CR roof and at 4.5' above the CR floor. The -
! first receptor provided the basis for selecting the final cloud size through comparison of the QAD-CGGP & MARIO dose rates with I the DORITA-2 and ELISA results. I I l
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l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 9% OF //0 PROJECT: JAF PRELM [] PREPARED BY M DATE 9/f0/97 FINAL [X] CHECKED BY /2f DATE ,',4wp 7
,, TITLE: Control Room Radiological Habitability'Under Power'6 prate Conditions and CP.EVASS Reconfiguration Fig. 8.2 -
QAD-CGGP Source / Receptor Geometry (Direct Shine from Overhead Clouds) z axis
- z=600m RADI0 ACTIVE CLOUD (1 quadrant, 25% of total volume)
O I I I i l i l
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Receptor #1 Receptor #2 30" roof and walls l : .~ . .. . . . . - . :.
-z=6.71m CR :- l 0 0
/ l
/.
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x or y (0,U) - - - - - - -
; g' axis x=12m x=600m
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f NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE @.3 OF //o
, PROJECT: JAF PRELN [] PREPARED BY /dTf._ DATE f/MN FINAL [X] CHECKED BY 8 DATEj g g 7 TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration 8.2.2 Results Post-LOCA direct-shine radiation levels on the roof of the CR and inside the control room from post-LOCA radioactive clouds resulting from drywell and MSIV leakage are shown in Table 8.2.
These were extracted from DORITA-2 Run Case #7, ELISA Run Case
#1, and QAD-CGGP and MARIO Run Cases #2 and #3 in Attachment B.
The cloud gamma spectra for use with MARIO were obtained from ELISA Run Case #1. Note the following: a) There is very good agreement between the DORITA-2, ELISA and QAD-CGGP & MARIO results for the receptor on the CR roof. The - s MARIO 31-day dose is slightly higher than the DORITA-2 and ELISA results even though the MARIO dose rates are slightly less; this is due to the use of numerical integration in MARIO. b) The maximum dose rate in the control room is less than 0.097 - mrad /hr, and the 31-day cumulative dose for continuous occupancy is about 0.13 mrad, which is insignificantly low. v _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ . _ _ _ _ . _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ . _ _ _ _ _ _ ._.__._.___a
1 NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE N OF //O f ~ PROJECT: JAF PRELM [] PREPARED BY vf/2. DATE 9/3c[ff- . l FINAL [X] CHECKED BY // TITLE: Control Room Radiological Habitability Under Power DATEudf87Uprate ! Conditions and CREVASS Reconfiguration Table 8.2 Direct-Shine Radiation Levels from Overhead Clouds Resulting from Post-LOCA Drywell and MSIV Leakage Time Dose Rate Cum. Dose Description (hrs) (rad /hr) (rad) Receptor on CR roof Maximum Dose Rate DORITA-2 0.0 2.346E-01 ----- ELISA 0.0 2.061E-01 ----- QAD-CGGP & MARIO 0.0 1.952E-01 ----- 31-Day Cum. Dose DORITA-2 744.0 ----- 8.958E-01 ELISA 744.0 ----- 8.427E-01 QAD-CGGP & MARIO 744.0 ----- 9.397E-01 Receptor Inside CR (QAD-CGGP & MARIO) 0.0 9.6977E-05 0.0000E+00 0.5 3.3896E-05 3.0005E-05 1.0 2.5428E-05 4.4735E-05 2.0 1.7901E-05 6.6180E-05 8.0 3.5629E-06 1.1947E-04 24.0 6.2400E-08 1.3332E-04 48.0 1.0061E-09 1.3368E-04 96.0 3.5417E-11 1.3369E-04 - 168.0 2.C297E-12 1.3369E-04 l 240.0 1.8054E-12 1.3369E-04 744.0 3.3788E-13 1.3369E-04
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NYPA - CALC.# JAF-CALC-RAD-00042 PROJECT: JAF PRELM REV 2 PAGE W OF //O [] PREPARED BY #)lt. DATE 7/JZ[N FINAL [X] CHECKED BY // DATE ,$6N477 TITLE: Control Room Radiological Habitability Under Power bprate Conditions and CREVASS Reconfiguration
.3 Direct Shine from Halogens Accumulating on the Control Roo:m Charcoal Filters Dose rates and cumulative doses from halogens accumulating on the control room charcoal filters were computed for each of the design-basis accidents. The source / receptor geometry for QAD-CGGP (with eight different receptor locations) was extracted from Ref. 7. As a simplification, a complete analysis for each of the receptors was limited to the two worst-case accidents (namely, an I MSLB with a spiked RCS activity, and a CRDA). The worst-case receptor was then used to assess the remaining accidents.
Analytical details appear in the subsections which follow. 8.3.1 Basic Data and Assumptions Filter Gamma Spectra The data and assumptions associated with the buildup of halogen radioactivity on the control room charcoal filters are identical to those presented in Secs. 4 through 7 for the various - accidents. Refer to those sections for details. Indeed, the DORITA-2 computer runs employed for the determination of the activity accumulation in the filters were identical to those used for the determination of the post-accident control room habitability analyses. Charcoal filter activity buildup for the various accidents are documented in DORITA-2 Cases 8 to 12. - 6 l Source / Receptor Geometry As noted above, the QAD-CGGP source / receptor geometry for computation of the direct shine radiation fields from the CR charcoal filters was extracted from Ref. 7, and is shown in Fig. (- 8.3. The figure shows the CR filter train located in a shielded - k cubicle adjacent to the Control Room.
I NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE @ OF //O n PROJECT: JAF PRELN [] PREPARED BY M DATE f/Mf) - Q FINAL [X] CHECKED BY /6I DATE /r4 gf 7
.- TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration Note that the shield walls separating the filter room and the Control Room are concrete block walls (not poured concrete). For -
conservatism, the density of the concrete block walls was assumed to be only half the density of a poured concrete shield wall. l In addition, it was conservatively assumed that the entire halogen inventory was deposited on the charcoal filter closest to the control room, i.e., on Source #2 in Fig. 8.3. { l l Eight different receptor locations were used in the analysis. l These locations represent unshielded locations, both axially and transversely, as well as locations in the CR and the operations p office. Location #8 in the NE corner of the ventilation V equipment room is for the evaluation of potential streaming into the CR from a wall opening at that location. t O
[ NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE @% OF PROJECT: JAF PRELM [] PREPARED BY M DATE g/]c)/if 9 FINAL [X] CHECKED BY Af DATE jS/7/f 7 TITLE: Control Room Radiological Habitability Under Power 17pfate Conditions and CREVASS Reconfiguration Fig. 8 3 - QAD-CGGP Source / Receptor Geometry (Dilect Shine from the CF Charcoal F1 :ers' l y axis a All dimensions are in cm Only Source #2 was modelled y=1638.30 . Source dimensions are: Filter Room, .
- 91.44 x 91.44 x 91.44 cm g
Control Room Operations Office b 365.76- N Source 1 274.32-182.88-91.44 - kN Source 2 \* */ ecek r #7
- x axis (0,[)#1 / -
[ 2 unti11ed concrete biocx wall y=-328.3 , \ "T y=-358. 76 N i
#6
/ \ Main CR Area
- e
$8 2 . &5 7 % 4W L L LL
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE %8 OF //C , A PROJECT: JAF PRELM [] PREPARED BY M DATE 9/d FINAL [X] CHECKED BY // DATE n3 8 7 TITLE: Control Room Radiological Habitability'Under Power #Upfate Conditions and CREVASS Reconfiguration _ l l 8.3.2 Results Direct shine radiation levels in the CR venti]ation equipment room and in the CR due to gamma radiation emanating from the CR filters were determined through use of the gamma spectra
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extracted from DORITA-2 Run Cases #8 to #12, and the QAD-CGGP geometry described in Sec. 8.3.1. Summaries of the results, extracted from QAD-CGGP & MARIO Run Case #4 are presented in Table 8.3 (all receptors for a CRDA), Table 8.4 (all receptors for an MSLB with spiked RCS activity), and Table 8.5 (for Receptor #2, in contact with the filter casing, and all the
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design-basis accidents). It is seen that the worst-case ) radiation levels in the control room are in the Operations Office (Receptor #7) for an MSLB with spiked RCS activity; the maximum J dose rate is about 13 mrad /hr (immediately after the accident), and the 31-day dose is about 127 mrad for continuous occupancy. The information in Tables 8.3 through 8.5 may also be used to determine the accessibility of the CR ventilation equipment room. It is seen that the worst-case dose rate is about'4.3 rad /hr at. Receptor #2 (in contact with the filter casing) following an MSLB with spiked RCS activity. For all other accidents, the worst-case dose rate is 1 css than 70 mrad /hr. _ i !I h !L) m
1 NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE D OF //C, PROJECT: JAF PRELM [] PREPARED BY M DATE __9/#/ O ('~' FINAL [X] CHECKED BY /2/ DATE Mhf7 TITLE: Control Room Radiological Habitability nder Power dpfate Conditions and CREVASS Reconfiguration
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Table 8.3 utrect-Shine Radiation Levels from Halogens Accumulating on the Control Room Charcoal Filters Control Rod Drop Accident - All Receptor Locations Time Dose Rate Int. Dose Cum. Dose (Hrs) (rad /hr) (rad) (rad) Receptor 1 0.0 0.000E+00 0.000E+00 0.000E+00 0.1 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 1.0 1.632E-03 4.081E-04 4.081E-04 8.0 7.640E-03 2.725E-02 2.766E-02 - 12.0 8.195E-03 3.166E-02 5.932E-02 24.0 8.694E-03 1.013E-01 1.606E-01 96.0 2.439E-03 3.543E-01 5.149E-01 () V 744.0 2.004E-04 5.805E-01 1.095E+00 Receptor 2 0.0 0.000E+00 0.000E+00' O.000E+00 0.1 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 1.0 1.309E-02 3.271E-03 3.271E-03 8.0 6.125E-02 2.184E-01 2.217E-01 12.0 6.571E-02 2.538E-01 4.755E-01 24.0 6.973E-02 8.124E-01 1.288E+00 96.0 1.959E-02 2.844E+00 4.132E+00 744.0 1.611E-03 4.664E+00 8.795E+00 Receptor 3 0.0 0.000E+00 0.000E+00 0.000E+00 0.1 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 1.0 9.714E-05 2.429E-05 2.429E-05 8.0 4.548E-04 1.622E-03 1.646E-03 l 12.0 4 #78E-04 1.884E-03 3.530E-03 l 24.0 5.175E-04 6.030E-03 9.561E-03 96.0 1.450E-04 2.108E-02 3.064E-02 744.0 1.191E-05 3.451E-02 6.515E-02
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NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PACE 70 OF _ //C
,_ PROJECT: JAF PRELM [] PREPARED BY M DATE 9/MO I FINAL [X] CHECKED BY //[ DATE gpfyg 7 l
TITLE: Control Room Radiological Habitability'Under Power 'Uprate Conditions and CREVASS Reconfiguration Table 8.3 (Continued) Time Dose Rate Int. Dose Cum. Dose
.(Brs) (rad /hr) (rad) (rad)
Receptor 4 0.0 0.000E+00 0.000E+00 0.000E+00 0.1 0.000E+00 0.000E+00 0.000E+00 l 0.5 0.000E+00 0.000E+00 0.000E+00 1.0 1.221E-04 3.051E-05 3.051E-05 8.0 5.713E-04 2.038E-03 2.068E-03 12.0 6,129E-04 2.367E-03 4.436E-03 24.0 6.501E-04 7.576E-03 1.201E-02 96.0 1.822E-04 2.648E-02 3.849E-02 744.0 ; 1.496E-05 4.334E-02 8.184E-02 - Receptor 5 0.0 0.000E+00 0.000E+00 0.000E+00 0 0.1 0.5 1.0 0.000E+00 0.000E+00 1 544E-04 0.000E+00 0.000E+00 3.859E-05 0.000E+00 0.000E+00 3.859E-05 8.0 7.226E-04 2.577E-03 2.615E-03 12.0 7.751E-04 2.994E-03 5.609E-03 - 24.0 8.222E-04 9.581E-03 1.519E-02 96.0 2.303E-OS 3.349E-02 4.868E-02 744.0 1.892E-05 5.481E-02 1.035E-01 Receptor 6 0.0 0.000E+00 0.000E+00 0.000E+00 0.1 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 - 1.0 2.390E-05 5.975E-06 5.975E-06 8.0 1.080E-04 3.904E-04 3.964E-04 12.0 1.122E-04 4.404E-04 8.367E-04 24.0 1.071E-04 1.315E-03 2.152E-03 96.0 2.321E-05 3.949E-03 6.101E-03 I 744.0 1.797E-06 5.423E-03 1.152E-02 I l l l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 9/ OF //O PROJECT: JAF PRELM [] PREPARED BY /)R DATE 9/3Yff G TITLE: FINAL [X) Control Room Radiological Habitability Under Power Dprate CHECKED BY ff DATE /M7/r7 Conditions and CREVASS Reconfiguration Table 8.3 (Continued) ~ Time Dose Rate Int. Dose Cum. Dose
.(Hrs) (rad /hr) (rad) (rad)
Receptor 7 0.0 0.000E+00 0.000E+00 0.000E+00 0.1 0.000E+00 0.000Ev00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 1.0 3.941E-05 9.851E-06 9.851E-06 8.0 1.785E-04 6.446E-04 6.544E-04 12.0 1.859E-04 7.287E-04 1.383E-03 24.0 1.791E-04 2.189E-03 3.572E-03 96.0 3.971E-05 6.661E-03 1.023E-02 744.0 3.090E-06 9.29? ,-03 1.953E-02 Receptor 8 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0 0.1 0.5 1.0 0.000E+00 7.259E-05 0.000E+00 0.000E+00 1.815E-05 0.000E+00 0.000E+00 1.815E-05 8.0 3.398E-04 1.212E-03 1.230E-03 12.0 3.645E-04 1. 4 0 8 E- 0 3- 2.638E-03 24.0 3.868E-04 4.507E-03 7.145E-03 96.0 1.084E-04 1.576E-02 2.290E-02 - 744.0 8.901E-06 2.579E-02 4.869E-02 O O
l NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 92 OF //8 PROJECT: JAF PRELM [] PREPARED BY /NE DATE @/b/9h l ( FINAL [X] CHECKED BY 8 DATE ,M7/f7 TITLE: Control Room Radiological Habitability Under Power 6p/ ate Conditiona and CREVASS Reconfiguration l Table 8.4 Direct-Shine Radiation Levels from Halogens Accumulating on the Control Room Charcoal Filters MSLB with Spiked RCS Activity (2 Ci/gm I131 DE) All Receptor Locations Time Dose Rate Int. Dose Cum. Dose (Hrs) (rad /hr) (rad) (rad) Receptor 1 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 5.366E-01 1.341E-01 1.341E-01 _ I' 1.0 5.275E-01 2.660E-01 4.001E-01 2.0 3.854E-01 4.527E-01 8.528E-01 4.0 2.342E-01 6.071E-01 1.460E+00 (~ ' 8.0 1.309E-01 7.104E-01 2.170E+00 16.0 6.486E-02 7.525E-01 2.923E+00 a 24.0 3.953E-02 4.092E-01 3.332E+00 96.0 5.412E-03 1.235E+00 4.568E+00 744.0 3.314E-04 1.179E+00* 5.746E+00 Receptor 2 , 0.0 0.000E+00 0.000E+00 0.000E+00 l 0.5 4.301E+00 1.075E+00 1.075E+00 1.0 4.228E+00 2.132E+00 3.207E+00 2.0 3.089E+00 3.628E+00 6.836E+00 4.0 1.877E+00 4.866E+00 1.170E+01 8.0 1.049E+00 5.694E+00 1.740E+01 l 16.0 5.199E-01 6.031E+00 2.343E+01 24.0 3.169E-01 3.280E+00 2.671E+01 l 96.0 4.345E-02 9.908E+00 3.662E+01 744.0 2.663E-03 9.466E+00 4.608E+01
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 9.3 or //o PROJECT: JAF PRELM [] PREPARED BY Aff__ DATE f/Mf FINAL [X] CHECKED BY ,/[ DATE jd/rp/97 TITLE: Control Room Radiological Habitability Under Power'Upfate Conditions and CREVASS Reconfiguration Table 8.4 (Continued) Time Dose Rate Int. Dose Cum. Dose (Hrs) (rad /hr) (rad) (rad) Receptor 3 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 3.193E-02 7.982E-03 7.982E-03 1.0 3.139E-02 1.583E-02 2.381E-02 2.0 2.293E-02 2.694E-02 5.075E-02 - 4.0 1.394E-02 3.613E-02 8.688E-02 8.0 7.793E-03 4.228E-02 1.292E-01 16.0 3.861E-03 4.479E-02 1.739E-01 24.0 2.354E-03 2,437E-02 1.983E-01 96.0 3.220E-04 7.354E-02 2.719E-01 744.0 1.969E-05 7.011E-02 3.420E-01 (q ,/ Receptor 4 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 4.011E-02 1.003E-02 1.003E-02 1.0 3.944E-02 1.989E-02 2.992E-02 l I 2.0 2.882E-02 3.385E 6.376E-02 4.0 1.751E-02 4.539E-02 1.092E-01 8.0 9.791E-03 5.312E-02 1.623E-01 16.0 4.851E-03 5.628E-02 2.186E-01 , 24.0 2.957E-03 3.061E-02 2.492E-01 96.0 4.045E-04 9.239E-02 3.416E-01 744.0 2.474E-05 8.806E-02 4.296E-01 Receptor 5 00 0.000E+00 0.000E+00 0.000E+00 0.5 5.074E-02 1.268E-02 1.268E-02 _ 1.0 4.988E-02 2.515E-02 3.784E-02 < 2.0 3.644E-02 4.281E-02 8.065E-02 ) 4.0 2.215E-02 5.741E-02 1.381E-01 1 l 8.0 1.238E-02 6.718E-02 2.052E-01 16.0 6.135E 03 7.117E-02 2.764E-01 24.0 3.740E-03 3.871E-02 3.151E-01 96.0 5.115E-04 1.168E-01 4.320E-01 744.0 3.128E-05 1.114E-01 5.433E-01 L___________--_______-.____________--_ _ _ _ _ _ _ - - - - _ . . -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE kh OF //O PROJECT: JAF PRELM [] PREPARL*D BY M DATE _9/%ip[9 } FINAL [X] CHECKED BY //[ DATE A4287 TITLE: Control Room Radiological Habitability Under Power Upfate Conditions and CREVASS Reconfiguration Table 8.4 (Continued) Time Dose Rate Int. Dose Cum. Dose (Hrs) (rad /hr) (rad) (rad) Receptor 6 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 8.009E-03 2.002E-03 2.002E-03 1.0 7.873E-03 3.971E-03 5.973E-03 2.0 5.767E-03 6.765E-03 1.274E-02 4.0 3.529E-03 9.113E-03 2.185E-02 8.0 1.974E-03 1.071E-02 3.256E-02 16.0 9.389E-04 1.114E-02 4.370E-02 24.0 5.397E-04 5.768E-03 4.947E-02 96.0 5.556E-05 1.533E-02 6.480E-02 744.0 2.971E-06 1.164E-02 7.644E-02 Receptor 7 0.0 0.000E+00 0.000E+00 0.000E+00
\0 0.5 1.0 1.319E-02 1,296E-02 3.297E-03 6.537E-03 3.297E-03 9.834E-03 2.0 9.491E-03 1.114E-02 2.097E-02 4.0 5.804E-03 1. 4 9 9E 3.597E-02 8.0 3.246E-03 1.761E-02 5.357E-02 16.0 1.549E-03 1.835E-02 7.192E-02 24.0 8.949E-04 9.538E-03 8.146E-02 -
96.0 9.445E-05 2.563E-02 1.071E-01 744.0 5.110E-06 1.985E-02 1.269E-01 Receptor 8 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 2.386E-02 5.964E-03 5.964E-03 1.0 2.345E-02 1.183E-02 1.779E-02 - 2.0 1.714E-02 2.013E-02 3.792E-02 4.0 1.042E-02 2.700E-02 6.492E-02 8.0 5.823E-03 3.159E-02 9.651E-02 16.0 2.885E-03 3.347E-02 1.300E-01 l 24.0 1.759E-03 1.821E-02 1.482E-01 i 96.0 2.407E-04 5.496E-02 2.031E-01 744.0 1.472E-05 5.240E-02 2.555E-01 l . O
T
~
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE OF //O O PROJECT: JAF PRELM [] PREPARED BY M DATE j/30[N FINAL [X] CHECKED BY /f/ DATE /MyM7 l TITLE: Control Room Radiological Habitability'Under Power'Up' rate l Conditions and CREVASS Reconfiguration j Table 8.5 Direct-Shine Radiation Levels from Halogens Accumulating on the Control Room Charcoal Filters All Accidents - Receptor Location #2 (Contact with Casing) Time Dose Rate Int. Dose Cum. Dose (hrs) (rad /hr) (rad) - fradl LOCA - Drywell/MSIV Leakage: 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 l 8.0 1.848E-02 6.929E-02 6.929E-02 12.0 1.891E-02 7.477E-02 1.441E-01 24.0 1.902E-02 2.276E-01 3.7162-01 96.0 1.247E-02 1.117E+00 1.489E+00 744.0 2.599E-03 4.079E+00 5.568E+00 ('~) 1 LOCA - ESF Leakage: 0.0 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 8.0 1.655E-03 6.207E-03 6.207E-03 12.0 1.7652-03 6.838E-03 1.305E-02 24.0 1.919E-03 2.209E-02 3.514E-02 _ 96.0 1.359E-03 1.168E-01 1.520E-01 744.0 3.107E-04 4.603E-01 6.122E-01 CRDA: 0.0 0.000E+00 0.000E+00 0.000E+00 0.1 0.000E+00 0.000E+00 0.000E+00 0.5 0.000E+00 0.000E+00 0.000E+00 1.0 1.309E-02 3.271E-03 3.271E-03 - 8.0 6.125E-02 2.184E-01 2.217E-01 12.0 6.571E-02 2.538E-01 4.755E-01 1 24.0 6.973E-02 8.124E-01 1.288E+00 96.0 1.959E-02 2.844E+00 4.132E+00 744.0 1.611E-03 4.664E+00 8.795E+00 0 1
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 7d OF //0 PROJECT: JAF PRELM [] PREPARED BY /)R. DATE 9/W9% _ FINAL [X] CHECKED BY // DATE M7/$C7 TITLE: Control Room Radiological Habitability'Under Power Dpfa'te Conditions and CREVASS Reconfiguration Table 8.5 (Continued) Time Dose Rate Int. Dose Cum. Dose 1 (hrs) (rad /hr) (rad) (rad) j MSLB (Equilibrium RCS) : 0.0 0.000E+00 0.000E+00 0.000E+00 0.2 0.000E+00 0.000E+00 0.000E+00 8.0 5.768E-02 2.250E-01 2.250E-01 12.0 3.902E-02 1.910E-01 4.159E-01 - 24.0 1.741E-02 3.213E-01 7.373E-01 96.0 2.388E-03 5.445E-01 1.282E+00 744.0 1.464E-04 5.203E-01 1.802E+00 RA: 0.0 0.000E+00 0.000E+00 0.000E+00 l r'~' O.2 0.000E+00 0.000E+00 0.000E+00 ~ i 8.0 1.376E-05 5.367E-05 5.367E-05 l 12.0 1.230E-05 5.206E-05 1.057E-04 24.0 9.362E-06 1.292E-04 2.349E-04 96.0 4.281E-06 4.675E 7.024E-04 744.0 3.827E-07 1.04GE-03 1.749E-03 1 MSLB (Spiked RCS): 1 0.0 0.000E+00 0.000E+00 0.000E+00 _ j l 0.5 4.301E+00 1.075E+00 1.075E+00 l 1.0 4.228E+00 2.132E+00 3.207E+00 j 2.0 3.089E+00 3.628E+00 6.836E+00 4.0 1.877E+00 4.866E+00 1.170E+01 8.0 1.049E+00 5.694E+00 1.740E+01 r 16.0 5.199E-01 6.031E+00 2.343E+01 24.0 3.169E-01 3.280E+00 2.671E+01 96.0 4.345E-02 9.908E+00 3.662E+01 744.0 2.663E-03 9.466E+00 4.608E+01 _ v _
i NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE hh OF //O PROJECT: JAF PRELM [] PREPARED BY /#f DATE f/MM _ FINAL [X] CHECKED BY // DATE /of7/f7 TITLE: Control Room Radiological Habitability Under Power 6 plate Conditions and CREVASS Reconfiguration i
- 9. POST-LOCA SGTS FILTER IODINE LORDING In Ref. 49, Stone and Webster Engineering Corporation assessed the adequacy of the SGTS filter charcoal mass to remove the post-LOCA halogens for the duration of the postulated accident. Their conclusion was that the JAF SGTS charcoal mass (720 lb, see below) was undersized by about a factor of 2. An extensive j search did not reveal any evidence of a NYPA response to SWEC's _
concern. I CRE re-assessed the adequacy of the charcoal mass in conjunction with the original radiological analyses for power uprate (Ref. ' 7). That assessment showed that SGTS filter saturation with () iodines will occur at about 500 hours after the postulated accident. The analysis was based on a 1.5% per day drywell leak rate and a 99.9% filter efficiency. Contribution due to the ESF leakage component was not included in this analysis. Reference 1 (Rev. 1 of the present calculation) re-analyzed the post-LCCA SGTS filter loading based on a 0.73% per day drywell leak rate, 100% filter efficiency and included the ESF leakage component. The analysis concluded that saturation of the SGTS did not occur for the entire duration of the accident. The purpose of this section of the present calculation is to ! reanalyze the SGTS loading assuming a 1.5% per day leak rate, I 100% filter efficiency and by including the ESF leakage component. - 9.1 Basic Data and Assumptions 1 ! The basic data and assumptions, which are equivalent to those in - l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE 9@ OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/3D[% O., FINAL [X) CHECKED BY M DATE f oM7/f'7 TITLE: Control Room Radiological Habitability nder Power pfate - Conditions and CREVASS Reconfiguration Sec. 4.1.1 and 4.2.1 of this calculation with the exception of the drywell .eakage rate and .he SGTS filter efficiency, are as follows': (a) A LOCA takes place at full power (2586.5). (b) The core inventory for the radionuclides of interest is as shown in Table 4.1. (c) Drvwell and MSIV Leakace
- 1. 100% of the ncble gases and 25% of the halogens present in the core are instantly released to the drywell and .
are available as an aerosol for leakage to the secondary containment.
- 2. The halogen composition airborne within the drywell is s as follows: 91% elemental, 4% organic and 5%
~ particulate.
- 3. Leakage from the drywell is at the rate of 1.5 % per -
day, and consists of the following'two components: i) Containment leakage, amounting to up to 1.27% per day, and ii) MSIV leakage, amounting to 0.23 % per day (from j Sec. 4.1.1).
- 4. All the noble gases and halogens leaking from the -
drywell are exhausted to the atmosphere via the Standby I Gas Treatment System (SGTS) and the main stack without mixing in the RB. (d) ESF Component Leakace
- 1. 50% of the total halogen activity present in the core mixes uniformly with the coolant in the RHR syste , ~
which has a total fluid mass of 3.21E+09 gm.
- 2. Total ESF component leakage rate is 5 gpm, and Refer to Secs. 4.1.1 and 4.2.1 for the appropriate references. {
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE N OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/tal% FINAL [X] CHECKED BY /[ DATE /cM/97 TITLE: Control Room Radiological Habitability Under Power dpr6te Conditions and CREVASS Reconfiguration corresponds to a fractional leakage rate from the recirculating water system of 0.00849 volume per day.
- 3. The ESF component leakage is 5 gpm and is assumed to be constant through the duration of the accident. -
- 4. An additional leakage of 50 gpm, due to gross failure of a passive component, is assumed to begin at the time of LOCA onset and to last for a period of 30 minutes.
- 5. 10% of the halogens contained in the leaking coolant becomes airborne within the RB and mixes uniformly with its atmosphere. -
- 6. Release from the RB is via the SGTS and main stack at the rate of 3.3 air changes per day (based on an SGTS flow of 6000 scfm).
(e) The SGTS filter efficiency for the removal of halogens is (r- _' conservatively assumed to be 100% for all halogen species (in lieu of 90 % as used in the radiological assessments - [See Section 4.1.1, Item g, for additional details]). (f) From Ref. 49, the SGTS has 18 charcoal cells. Each cell consists of two horizontal charcoal beds, and each bed contains 20 lb of activated charcoal. Hence, the total SGTS charcoal mass is equal to 18 (cells) x2 (beds / cell) x 20 (lb/ bed) = 720 lb = 3.27E+05 gm.
~
(g) The regulatory limits for iodine loading of charcoal filtration systems is 2.5 mg per gram of activated carbon (Reg. Guide 1.52, Ref. 24, Sec. C.3). This refers to the total iodine loading (radioactive plus stable). Hence, the 1 existing charcoal mass can accommodate a total iodine load
~
of [3 . 27E+ 0 F (gm) x 2.5E-03 (gm of total iodine /gm of charcoal)] = 817 gm. (h) The decay constants (1) of the halogens which would tg accumulate on the SGTS filtration system are as follows
\M (from Ref. 11):
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /00 OF //O PROJECT: JAF PRELM [] PREPARED BY K DATE 1/Mk FINAL [X] CHECKED BY /25 DATE gfEY/ TITLE: Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration - Nuclide 1 (hr'*) Nuclide 1 (h '*) BR 83 2.900E-01 I 131 3.592E-03 l BR 84 1.308E+00 I 132 3.014E-01 BR 85 1.451E+01 I 133 3.332E-02 I 134 7.907E-01 I 129 5.037E-12 I 135 1.048E-01 I 130 5.607E-02 I 136 3.006E+01 I These are needed for the conversion of filter radioactivity I to mass. [ O. U 49 0
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /O[ OF //O I PROJECT: JAF PRELM [] PREPARED BY M DATE 9/JO/$ I FINAL [X] CHECKED BY M DATE f /f 7/f7 TITLE: Control Room Radiological Habitability Under Power Dpr' ate Conditions and CREVASS Reconfiguration 9.2 Results The buildup of halogen radioactivity on the SGTS charcoal filtration system was calculated using the DORITA-2 computer code and the data and assumptions listed above in Sec. 9.1. Summaries of the filter activities as a function of post-LOCA time, extracted from DORITA-2 Run Case #13. are presented in Table 9.1 (for drywell and MSIV leakage) and Table 9.2 (for ESF component leakage). The overall filter activities (for all leakage pathways) are presented in Table 9.3. The isotopic activity (Q ] of each nuclide [I] in Table 9.3 i was converted to mass (M ] through use of the following basic i , formula:
/O i #
Q [Ci] 3.7E+10 (dis /sec-Ci] 3600 (sec/hr] A [ Mass No.] i 1 l M i [gm] = 6.023E+23 [ atoms /gm-mole] l i [hr'*] For instance, the 2.355E-01 Ci of I-129 and the 4.913E+05 Ci of I I-131 on the filter at 744 hours after the postulated LOCA (from Table 9.3), correspond to the following masses (as shown in Table 9.4)': 2.355E-01 x 3.7E+10 x 3600 x 129 Mzu, (gm] = = 1333.8 gm _ 6.023E+23 x 5.037E-12 l l l 4.913E+05 x 3.7E+10 x 3600 x 131 Mzur [gm] = = 3.963 gm 6.023E+23 x 3.592E-03 A summary of the filter loading as a function of post-LOCA time is shown in Table 2.5. It is seen that, by 400 hours after Note that the relatively stable I-129 predominates.
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /02-- OF //O ' PROJECT: JAF PRELM [] PREPARED BY M DATE 9/Mk FINAL [X] CHECKED BY @ DATE fody/f7 TITLE: Control Room Radiological Habitability Under Power f3 prate Conditions and CREVASS Reconfiguration the accident, the total iodine on the filter would amount to 797.4 grams. The design-basis maximum loadina is 817 grams, as - given under Item (g) in Sec. 9.1 above. Saturation of the filters will occur beyond about 400 hours after the LOCA. l i i p l I O -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /O.,/ OF //O I PROJECT: JAF PRELM [] PREPARED BY /ht DATE 9/_?O/N FINAL [X] CHECKED BY ffd~ DATE /q37/rp - ! TITLE: Control Room Radiological Habitability Under Power Uprate l Conditions and CREVASS Reconfiguration l Table 9.1 POST-LOCA SGTS FILTER HALOGEN LOADING (Ci) Due to Drywell and MSIV Leakage Isotope 0 0.5 1 2 5 Activity (Ci) versus Time (hrs) BR83 0.000E+00 5.458E+02 9.442E+02 1.413E+03 1.478E+03 BR84 0.000E+00 5.817E+02 6.049E+02 3.270E+02 1.615E+01 BR85 0.000E+00 9.488E-01 1.342E-03 1.343E-09 4.205E-28 l
~
I129 0.000E+00 1.761E-04 3.521E-04 7.039E-04 1.758E-03 I130 0.000E+00 2.055E+02 3.995E+02 7.552E+02 1.594E+03 l I131 0.000E+00 5.306E+03 1.059E+04 2.110E+04 5.214E+04 ' I132 0.000E+00 6.682E+03 1.149E+04 1.700E+04 1.719E+04 , I I133 0.000E+00 1.093E+04 2.150E+04 4.158E+04 9.396E+04 l I134 0.000E+00 8.238E+03 1.109E+04 1.006E+04 2.344E+03 l (s g_) I135 0.000E+00 9.962E+03 1.890E+04 3.403E+04 6.206E+04 I136 0.000E+00 1.499E-03 8.881E-10 1.558E-22 0.000E+00 _ TOTAL 0.000E+00 4.245E+04 7.553E+04 1.263E+05 2.308E+05 Isotope 10 20 50 100 200 Activity (Ci) versus Time (hrs) BR83 6.923E+02 7.595E+01 3.133E-02 3.112E-08 1.535E-20 - BR84 4.662E-02 1.943E-07 4.396E-24 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I129 3.511E-03 7.000E-03 1.734E-02 3.414E-02 6.621E-02 I130 2.405E+03 2.737E+03 1.260E+03 1.504E+02 1.070E+00 l I131 1.023E+05 1.967E+05 4.374E+05 7.197E+05 9.746E+05 I132 7.608E+03 7.449E+02 2.185E-01 1.229E-07 1.946E-20 I133 1.588E+05 2.269E+05 2.068E+05 7.696E+04 5.329E+03 - I134 8.984E+01 6.597E-02 8.162E-12 1.089E-28 0.000E+00 I135 7.337E+04 5.127E+04 5.467E+03 5.693E+01 3.087E-03 I136 0.000E+00 0.000E+00 0.000E400 0.000E+00 0.000E+00 TOTAL 3.453E+05 4.784E+05 6.509E+05 7.969E+05 9.799E+05 A U k
c l NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /0@ or //0 PROJECT: jar PRELM [] PREPARED BY M DATE 9/fo[% ~ l FINAL [X] CHECKED BY /// DATE rom #f7 TITLE: Control Room Radiological Habitability Under Power prate Conditions and CREVASS Reconfiguration ' Table 9.1 (Continued ...) l l Isotope 400 450 500 744 Activity (Ci) versus Time (hrs) BR83 1.401E-45 0.000E+00 0.000E+00 0.000E+00 BR84 0.000E+00 0.000E+00 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I129 1.246E-01 1.381E-01 1.512E-01 2.095E-01 _ I130 2.715E-05 1.823E-06 1.209E-07 1.914E-13 I131 8.945E+05 8.284E+05 7.578E+05 4.371E+05 I132 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I133 1.279E+01 2.679E+00 5.541E-01 2.259E-04 I134 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1135 4.545E-12 2.664E-14 1.542E-16 1.657E-27 ; r~g I136 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I b TOTAL 8.945E+05 8.284E+05 7.578E+05 4.371E+05 M
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE / O,(~ OF MO PROJECT: JAF PRELM [] PREPARED BY /hC_ DATE 7_9/MS FINAL [E] CHECKED BY M DATE f/d)77
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TITLE: Control Room Radiological Habitability Under Power Upr' ate Conditions and CREVASS Reconfiguration Table 9.2 POST-LOCA SGTS FILTER HALOGEN LOADING (Ci) Due to ESF Component Leakage Isotope 0 0.5 1 2 5 Activity (Ci) versus Time (hrs) BR83 0.000E+00 2.075E+01 5.404E+01 9.695E+01 1.185E+02 BR84 0.000E+00 2.211E+01 3.462E+01 2.245E+01 1.295E+00 BR85 0.000E+00 3.607E-02 7.682E-05 9.219E-11 3.373E-29 _ I129 0.000E+00 6.694E-06 2.015E-05 4.832E-05 1.410E-04 I130 0.000E+00 7.811E+00 2.286E+01 5.183E+01 1.279E+02 I131 0.000E+00 2.017E+02 6.062E+02 1.448E+03 4.182E+03 I132 , 0.000E+00 2.540E+02 6.577E+02 1.167E+03 1.379E+03 I133 0.000E+00 4.156E+02 1.230E+03 2.854E+03 7.536E+03 g-~g I134 0.000E+00 3.132E+02 6.349E+02 6.905E+02 1.880E+02 (/ I135 0.000E+00 3.788E+02 1.082E+03 2.336E+03 4.978E+03 - I136 0.000E+00 5.700E-05 5.083E-11 1.070E-23 0.000E+00 TOTAL 0.000E+00 1.614E+03 4.323E+03 8.667E+03 1.851E+04 Isotope 10 20 50 100 200 Activity (Ci) versus Time (hrs) BR83 6.189E+01 7.498E+00 3.374E-03 3.473E-09 1.760E-21 BR84 4.168E-03 1.918E-08 4.733E-25 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I129 3.139E-04 6.910E-04 1.867E-03 3.810E-03 7.595E-03 I130 2.150E+02 2.702E+02 1.357E+02 1.678E+01 1.228E-01 I131 9.141E+03 1.942E+04 4.710E+04 8.032E+04 1.118E+05 I132 6.801E+02 7.354E+01 2.353E-02 1.372E-08 2.232E-21 l 1133 1.420E+04 2.240E+04 2.227E+04 8.588E+03 6.113E+02 l I134 8.031E+00 c.512E-03 8.789E-13 1.215E-29 0.000E+00 1135 6.559E+03 5.061E+03 5.887E*02 6.353E+00 3.541E-04 1136 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 TOTAL 3.086E+04 4.723E+04 7.009E+04 8.893E+04 1.124E+05 l
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /O[ OF [/O PROJECT: JAF PRELM [] PREPARED BY /hC. DATE 9/Ec[4> FINAL [X] CHECKED BY // DATE /d/ f7 '/9 7 TITLE: Control Room Radiological Habitability Under Power'Upkate Conditions and CREVASS Reconfiguration Table 9.2 (Continued ...) Isotope 400 450 500 744 Activity (Ci) versus Time (hrs) - BR83 0.000E+00 0.000E+00 0.000E+00 0.000E+00 BR84 0.000E+00 0.000E+00 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I129 1.477E-02 1.649E-02 1.818E-02 2.599E-02 I130 3. 217E-06 2.176E-07 1.453E-08 2. 374E -14 I131 1.060E+05 9.888E+04 9.108E+04 5.422E+04 1132 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I133 1.516E+00 3.198E-01 6.660E-02 2.802E-05 I134 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I135 5.387E-13 3.180E-15 1.853E-17 2.056E-28 I136 0'.000E+00 0.000E+00 0.000E+00 0.000E+00 f-s) (_/ TOTAL 1.060E+05 9.889E+04 9.108E+04 5.422E+04 , m O 1 C _____ .m_.m.-_____ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _
NYPA - CALC.# JAF-CALC-FAD-00042 REV 2 PAGE /O} OF //O g PROJECT: JAF PRELM [] PREPARED BY /T._ DATE f/M')- FIF.AL [X) CHECKED BY /f/ DATE A/n/5 7 TITLE: Contro) Roem Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration Table 9.3 POST-LOCA SGTS FILTER HALOGEN LOADING (Ci) Due to Drywell, MSIV and ESF Component Leakage Isotope 0 0.5 1 2 5 Activity (Ci) versus Time (hrs) BR83 0.000E+00 5.666E+02 9.982E+02 1.510E+03 1.597E+03 BR84 0.000E+00 6.038E+02 6.395E+02 3.495E+02 1.745E+01 BR85 0.000E+00 9.849E-01 1.419E-03 1.435E-09 4.542E-28 I129 0.000E+00 1.828E-04 3.723E-04 7.522E-04 1.899E-03 I130 0.000E+00 2.133E+02 4 224E+02 8.070E+02 1.722E+03 I131 0.000E+00 5.508E+03 1.120E+04 2.255E+04 5.632E+04 I132 0.000E+00 6.936E+03 1.215E+04 1.817E+04 1.857E+04 I133 0.000E+00 1.135E+04 2.273E+04 4.443E+04 1.015E+05 I134 0.000E+00 8.551E+03 1.172E+04 1.075E+04 2.532E+03 k'~')s _ I135 0.000E+00 1.034E+04 1.998E+04 3.637E+04 6.704E+04 I136 0.000E+00 1.556E-03 9.389E-10 1.665E-22 0.000E+00 TOTAL 0.000E+00 4.407E+04 7.984E+04 1.349E+05 2.493E+05 Isotope 10 20 50 100 200 Activity (Ci) versus Time (hrs) BR83 7.542E+02 8.345E+01 3.470E-02 3.459E-08 1.711E-20 BR84 5.079E-02 2.135E-07 4.869E-24 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00
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I129 3.825E-03 7.691E-03 1.921E-02 3.795E-02 7.381E-02 I130 2.620E+03 3.007E+03 1.396E+03 1.672E+02 1.193E+00 I131 1.114E+05 2.161E+05 4.845E+05 8.000E+05 1.086E+06 I132 8.288E+03 8.184E+02 2.420E-01 1.366E-07 2.169E-20 l I133 1.730E+05 2.493E+05 2.291E+05 8.555E+04 5.940E+03 I134 9.787b+01 7.248E-02 9.041E-12 1.211E-28 0.000E+00 I135 7.993E+04 5.633E+04 6.056E+03 6.328E+01 3.441E-03 I136 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 - gs TOTAL 3.761E+05 5.257E+05 7.210E+05 8.858E+05 1.092E+06 u______-____ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ _
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /Of OF //O PROJECT: JAF PRELM [] PREPARED BY M DATE 9/f2/9% FINAL [X] CHECKED BY 8 DATE f2N/f7 Control Room Radiological Habitability Under Power'6p/ ate TITLE: Conditions and CREVASS Reconfiguration Table 9.3 (Continued ...) j Isotope 400 450 500 744 Activity (Ci) versus Time (hrs) i BR83 1.401E-45 0.000E+00 0.000E+00 0.000E+00 i BR84 0.000E+00 0.000E+00 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 - I129 1.394E-01 1.546E-01 1.694E-01 2.355E-01 I130 3.037E-05 2.041E-06 1.354E-07 2.151E-13 < I131 1.001E+06 9.273E+05 8.489E+05 4.913E+05 I132 0.000E+00 0.000E+00 0.000E+00 0.000E+00 l I133 1.431E+01 2.999E+00 6.207E-01 2.539E-04 I134 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I135 5.084E-12 2.982E-14 1.727E-16 1.863E-27 - O I136 0.000E+00 0.000E+00 0.000E+00 0.000E+00 ( TOTAL 1.001E+06 9.273E+05 8.489E+05 4.913E+05 9 O lO
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NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE /O OF [/6 r PROJECT: JAF PRELM [] PREPARED BY /h2. DATE f/30/f} FINAL [E) CHECKED BY 8 DATE //j/f7 TITLE: Control Room Radiological Habitability 'Under Power'dpr6t'e Conditions and CREVASS Reconfiguration I Table 9.4 POST-LOCA SGTS FILTER HALOGEN LOADING (grams) Due to Drywell, MSIV and ESF Component Leakage l Decay Isotope Constant 0 0.5 1 2 l (1/hr) Halogen Loading (gms) versus Time (hrs) BR83 2.900E-01 0.000E+00 3.586E-05 6.318E-05 9.557E-05 BR84 1.308E+00 0.000E+00 8.576E-06 9.083E-06 4.963E-06 BR85 1.451E+01 0.000E+00 1.276E-09 1.838E-12 1.859E-18 1129 5.037E-12 0.000E+00 1.035E+00 2.108E+00 4.260E+00 I130 5.607E-02 0.000E+00 1.094E-04 2.166E-04 4.138E-04 I131 3.592E-03 0.000E+00 4.442E-02 9.030E-02 1.819E-01 I132 3.014E-01 0.000E+00 6.718E-04 1.177E-03 1.760E-03 I133 3.332E-02 0.000E+00 1.002E-02 2.006E-02 3.922E-02 I134 7.907E-01 0.000E+00 3.205E-04 4.394E-04 4.029E-04 (' I135 I136 1.048E-01 0.000E+00 2.946E-03 5.692E-03 1.036E-02 3.006E+01 0.000E+00 1.557E-12 9.395E-19 1.666E-31 TOTAL 0.000E+00 1.094E+00 2.226E+00 4.495E+00 Isotope 5 10 20 50 100 Halogen Loading (gms) versus Time (hrs) BR83 1.011E-04 4.774E-05 5.282E-06 2.197E-09 2.190E-15 BR84 2.478E-07 7.213E-10 3.032E-15 6.916E-32 0.000E400 BR85 5.885E-37 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I129 1.076E+01 2.166E+01 4.356E+01 1.088E+02 2.149E+02 I130 8.829E-04 1.343E-03 1.542E-03 7.156E-04 8.572E-05 _ I131 4.543E-01 8.988E-01 1.743E+00 3.908E+00 6.452E+00 I132 1.799E-03 8.027E-04 7.927E-05 2.344E-08 1.323E-14 I133 8.960E-02 1.527E-01 2.201E-01 2.022E-01 7.552E-02 I134 9.490E-05 3.668E-06 2.717E-09 3.388E-19 4.537E-36 1135 1.910E-02 2.277E-02 1.605E-02 1.725E-03 1.803E-05 { l I136 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 l . q i TOTAL 1.132E+01 2.274E+01 4.554E+01 1.129E+02 2.215E+02 -
NYPA - CALC.# JAF-CALC-RAD-00042 REV 2 PAGE //O OF f/O PROJECT: JAF PRELM [] PREPARED BY j#L. DATE f/MN FINAL [X] CHECKED BY /// DATE f2d/97 TITLE: Control Room Radiological Habitability Under Power 8p/ ate Conditions and CREVASS Reconfiguration Table 9.4 (Continued . . . ) Isotope 200 400 450 500 744 Halogen Loading (gms) versus Time (hrs) BR83 1.083E-27 8.868E-53 0.000E+00 0.000E+00 0.000E+00 ~ BR84 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 BR85 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I129 4.180E+02 7.894E+02 8.756E+02 9.593E+02 1.334E+03 I130 6.116E-07 1.557E-11 1.046E-12 6.944E-14 1.103E-19 I131 8.762E+00 8.069E+00 7.479E+00 6.847E+00 3.963E+00 I132 2.101E-27 0.000E+00 0.000E+00 0.000E+00 0.000E+00 I133 5.244E-03 1.263E-05 2.647E-06 5.479E-07 2.241E-10 I134 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 ( I135 I136 9.803E-10 0.000E+00 1.448E-18 0.000E+00 8.495E-21 0.000E+00 4.921E-23 0.000E+00 5.306E-34 0.000E+00 TOTAL 4.268E+02 7.974E+02 8.830E+02 9.662E+02 1.338E+03
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i l' 1 i ' !g NYPA Calculation No. JAF-CALC-RAD-00042, Rev. 2
- control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration Attachment A copies of Pertinent References Excerpts from selected references are presented in the pages which follow. References included are
- 14. Stone & Webster Engineering Calculation No. 12966-PE (N) -019-0, "High Energy Line Break Analysis in the Turbine Building for Class IE Electrical Equipment Qualification in Response to IE Bulletin 79-01B" (6/9/81) - Excerpts only.
' ("'T 15. GPU Nuclear Corporation letter 5450-95-0006, addressed
'(_,/ to M. Karasulu, from N. G. Trikouros, titled "FitzPatrick Nuclear Plant Turbine Building HELB Analysis Results" (2/17/95) - Excerpts only.
- 49. SWEC Engineering Calculation 12966-RP-60-23, " Maximum Post-Accident Iodine Load in Standby Gas Treatment System Charcoal" (6/11/80) i l
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1 O =:::7.2::r" JNuolear Puelpperrf,NewJereer07C54 201316M100 TELEX 136402 Vatera Dised DialNumber (201) 316-7124 February 17, 1995 5450-95-0006 nr. Muzaffer Karasulu New York Power Authority 123 Main Street White Plains, NY 10601
SUBJECT:
FITZPATRICK NUCLEAR PLANT TURBINE BUILDING HELB ANALYSIS RESULTS
Dear Muzaffer:
The temperature, pressure and humidity following a 24" break and a 16" break of the Main Steamline in the Turbine Building have been calculated for Electrical Bays 272-1, 272-2 East, 272-2 West, and 272-3. This letter is being fomarded to you so that you can proceed with your evaluations. A formal design verified calculation package will be forwarded next week. 1 24" HSL BEAK RESETS { The initial temperature was assumed to be 109 deg f. The initial pressure was l t 14.7 psia and the initial humidity was 555. These conditions were identified l as the appropriate FSAR initial conditions with the exception of the initial i pressure which was given as 15.0 in the FSAR case specified by your i calculation. We used 14.7 psia because it resulted in more conservative results. Also, if the building were at 15.0 psia, }he smoke ejectors would be constantly open. The analysis assumed that a 600ft area of siding was blown out at 0.69 p id. It was also assumed that'the smoke ejectors would open at ' 0.278 psid (40 psf) with an area of 411 ft . The analysis was done with G0THIC The GOTHIC version 4.1 and was tr-f:;::f::tly analyzed with The RELAPS Mod 3. O noda112ation diagram is shown in Figure 1. blowdown was based upon the original Stone & Webster blowdowns in calcf 12966-PE(N)-019-0. Information , from your calculation entitled "EQ Analysis For JAF's Switchgear
f i Nr. Mazaffer Karasulu February 17, 1995 5450-95-0006 page Two Electric Bays Following HELB in Turbine Bdilding" was also used for geometry and initial condition information. The analysis was run out to 200 seconds which appeared to be sufficient to capture the effects of the break. l The temperatures in each of the four electrical bays are shown in Figures 2 -
- 4. As you can see, EB 272-1 and 272-3 get more than 10 dog f hotter than the initial temperature. These two bays are essentially open to the main Turbine l
Building and tend to follow the conditions there as a result of the large ! mixing area. The temperatures in EB 272-2 East and West on the other hand stayed low with an increase of about 6 deg f from the initial condition. The pressures in each of the compartments were essentially identical and are shown in Figure 5 for EB 272-2 East and West. The building pressure peak occurs at I the time when the siding blows out (approximately 4 seconds) and the peak pressure of about 0.7 psig in all of the electrical bays refleeps the siding failure pressure assumption of 0.69 psig. The use of a 600 ft siding blowout area and 0.69 psid failure pressure for the 24" break is considered to be extremely conservative and a much larger area and lower pressure could be considered. The humidity in each of the bays is shown in Figures 6-9. Only GOTHIC results are available for these because RELAPS Mod 3 could not be initialized with less than 2005 humidity. As can be seen, the vapor pressure in bays EB 272-1 and 272-3 becomes equal to the saturation pressure (humidity equals 1005) early in the transient and remains saturated for the duration of the analysis. However, the vapor pressure in bays EB 272-2 East and West does not reach saturation so the humidity remains below 1005. Because these bays are not strongly affected by the blowdoun, the humidity in the first 200 seconds is very cLose to the initial condition. In conclusion, the results for the 24" HSL Break in the JAF Turbine Building show that the environment in electrical bays EB 272-2 East and West remains mild while the remainder af the Turbine building does not meet the criteria for a mild environment. 18" HEL ME R RESET 5 The GOTHIC results for the 16" MSL break are provided in Figures 13 - 17. The temperatures in each of the electrical bays increase slightly over the 24" case as a result of the much larger blowdoun (twice as much) for this case. The peak temperatures in EB 272-2 East and West increase to about 117 dog F from the 115 dog F for the 24" break case. The humidity in Figures 13 - 16 indicate 1005 for all bays except EB 272-2 East and West which remain well during and after the break. Pressure peaks are the same in O below 1005 magnitude as the 24" break case, but occur at about 3 seconds which slightly earlier than the 24" break results.
A Mr. Ihcaffer Karasulu , V-' February 17, 1995 5450-95-0006 Page Three The conclusions regarding the mild environment in EB 272-2 East and West remain the same as for the 24" MSL break. The RELAP design verification for this case is in progress and will be completed later today. Both the 16" and 24" cases will he included in the final calculation. Very truly yours,
. /.
N. G. Trikouros Manager, Safety Analysis & Plant Control
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