Text

Rev 1 to JAF-CALC-RAD-00048, Power Update Project - Radiological Impact at Onsite & Offsite Outdoor Receptors Following Design-Basis Accidents
	ML20247G116
Person / Time
Site:	FitzPatrick
Issue date:	11/14/1997
From:	Golshani M, Ramachandran A, Re G; POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	;
Shared Package
ML20247F730	List: JPN-98-007, Forwards Application for Amend to License DPR-59,changing Valve of Allowable Containment Rate to 1.5 Percent Day & Assumed Value of Filter Efficiency for SBGT Used in Calculations to Determine Dose Effects to Onsite; ML20247F779; ML20247F798; ML20247F910; ML20247F948; ML20247F971; ML20247G028; ML20247G064; ML20247G116;
References
	[[::JAF-CALC-RAD\|JAF-CALC-RAD]], JAF-CALC-RAD-00, JAF-CALC-RAD-00048, JAF-CALC-RAD-48, NUDOCS 9805200141
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COMPUTER CALCULATION CONTROL SHEET Page 1 of 78 e- plus Attachments JAF lXl

" l l CALC. NO. JAF-CALC-RAD-OOO48 REV. 1 IP3 MOD / TASK HO.

QA CATEGORY OF CALCULATION: II/III CALCULATIONAL TYPE: PRELIMINARY: FINAL: x 1 PROJECT / TASK: )

SYSTEM NO./NAME: l TITLE: Power Uprate Proiect - Radiological Impact at Onsite and Offsite Outdoor Receptors Followina Desion-Basis Accidents PREPARER: A. Ramac ran h ////3k CHECKER: M. Golshani # [f h '

11/13/9 7 VERIFIED: N/A E .

APPROVED: G.C. Re' M,em 8 /[' w//f/f ~7 sw- j ri PROBLEM / OBJECTIVE / METHOD See pages 2, 11-16, and 36-37

. O, DESIGN BASIS / ASSUMPTIONS / ANALYSIS tQ l See pages 38-78

SUMMARY

/ CONCLUSIONS l l

See pages 17-35 l l

REFERENCES .

l See pages 4-7 AFFECTED SYSTEMS / COMPONENTS / DOCUMENTS D VOIDED 0 SUPERSEDED BY: upersedes Revision O of this (CALC NO.) calculation.

Supersedes Those Portions of JAF-CALC-RAD-00008 Which Deals With the Onsite and Offsite Outdoor Receptors.

NYPA FORM DCM-14, ATTACHMENT 4.1 (REVISION 1) Page 1 of 1 9905200141 990226 PDR ADOCK 05000333 u____ (p PDR

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FINAL [1] CHECKED BY Ar(r - DATE t In3 /,1 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Statement Of' Problem This calculation [ prepared by Corporate Radiological Engineering (CRE)] is in support of the power uprate program at JAF, and re-evaluates the analyses documented in Ref. 1 (JAF-CALC-RAD-00008 )

for outdoor receptors at onsite locations, and receptors at the Site Boundary (SB) and the Low Population Zone (LPZ). The reasons for the re-analysis are as follows:

(a) Incorporation of the recently revised atmospheric dispersion factors (Ref. 2),

(b) Revisions to the scenarios for a Main Steam Line Break Accident (Ref. 3), and a Control Rod Drop Accident (Ref. 4), and (c) Revisions to certain accident assumptions, for ON - consistency with those employed in the revised control Room Habitability analysis (Ref. 5).

The analyses documented in this calculation fall under ACTS Item 18820 (Ref. 39).

Revision 1 - Remarks Revision 1 of this calculation was undertaken to address the concern identified under ACTS Item 23847 (Ref. 43):

(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 In addition, the present calculation incorporates the following

, changes to the assumptions and methods employed in the previous

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3 or 9?

PROJECT: JAF PRELM [] PREPARED BY M(2- DATE 11/ I 3 / 9 }

FINAL CHECKED BY M6- DATE ##/I3/9 /

O _, [I]

TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents revision of this calculation:

(b) revision of the atmospheric dispersion factors for elevated releases as documented in revision 2 of Reference 2, and (c) use of the ICRP 30 (Ref. 45) dose conversion factors for the determination of thyroid doses.

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) TITLE: Power Uprate Project - Radiological Impact at Onsite and j Offsite Outdoor Receptors Following Design-Basis Accidents References

1. CRE Calculation JAF-CALC-RAD-00008, " Radiological Consequences of Design Basis Accidents at James A.

FitzPatrick" (11/27/91)

2. CRE Calculation JAF-CALC-RAD-00007, Revs. 1&2, " Power Uprate Program - Onsite and Offsite Post-Accident Atmospheric Dispersion Factors"

3. CRE Calculation JAF-CALC-RAD-00039, " Revised Offsite Radiation Exposures Following a Design-Basis Main Steam Line Break Accident" (12/15/94)

4. 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)

5. CRE Calculation JAF-CALC-RAD-00042, Revs. O, 1 & 2, .

" Control Room Radiological Habitability Under Power Uprate !

Conditions and CREVASS Reconfiguration"

6. CRE Computer Code DORITA-2, "A Computer Code for the i 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 1.5.1.2 (1/22/97)

7. CRE Computer Code QAD-CGGP, "A Combinatorial Geometry Version of QAD-P5A, A Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations Using the GP Buildup Factor," RAD-006, Release 1.3.1.1 (3/26/92)

8. CRE Calculation-Specific Computer Code MATILDA, documented in- i i

a) CRE Calculation JAF-CALC-RAD-00003, " Power Uprate Program - Reactor Building Post-LOCA EQ Radiation Levels Due to Buildup of Halogen Activity on Air Filtration Systems" (November 1991) b) CRE Calculation JAF-CALC-RAD-00015, " Equipment Qualification Radiation Exposures Following A Control Rod Drop Accident" (7/28/92)

9. NYPA Memorandum JAG-93-245 addressed to J. Lazarus, from J. Gray, titled " Control Rod Drop Accident (CRDA) '

Assumption" (9/24/93) [See JAF-CALC-RAD-00026 for a copy of this ref.]

10. US NRC NUREG-0800, " Standard Review Plan for the Review of ,

gg Safety Analysis Reports for Nuclear Power Plants" I Nn ,A l

NYPA - CALC.9 JAF-CALC-RAD-00048 REV 1 PAGE f OF %@

PROJECT: JAF PRELM [] PREPARED BY #il DATE H//3/M FINAL [X] CEECKED BY M 6- DATE 11/f y,7

'A ) TITLE: Power Uprate Project - Radiological Impact at Onsite and ,

Offsite Outdoor Receptors Following Design-Basis Accidents

11. 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.]

12. GE letter addressed to Richard Chau, NYPA, from C. H.

Stoll, GE Plant Performance Engineering, titled "J. A.

FITZPATRICK (JAFNPP) Power Uprate Program - Formal Transmittal of Final Source Term Analysis Results" .

(5/2/91) [See JAF-CALC-RAD-00008 (Ref. 1) for a copy of l this reference.] l

13. J. DiNunno, F. Anderson, R. Baker and R. Waterfield, l

" Calculation of Distance Factors for Power and Test i Reactor Sites," AEC, Division of Licensing and Regulation, TID-14844 (March 1962)

14. 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)

15. 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)

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16. 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)

17. US NRC Regulatory Guide 1.49, " Power Levels for Nuclear Power Plants" (Rev. 1, December 1973)

18. US NRC Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors" (May 1974)

19. US NRC Regulatory Guide 1.52, " Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power 1 Plants" (Rev. 2, March 1978)

20. General Electric Report NEDO-31400, " Safety Evaluation for i Eliminating the BWR Main Steam Isolation Valve Closure Function and Scram Function of the Main Steam Line 1 Radiation Monitor" (May 1987) (See JAF-CALC-RAD-00013 for 0- copy)

I

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Q s> [), TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents

21. Proposed Technical Specification changes - Power Uprate (JPTS-91-025), and NYPA Letter to NRC JPN-92-028 (6/5/92)

22. US NRC NUREG-0123, Rev. 3, " Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5) " (Fall 1980)

23. 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. 1) for a copy of this reference.]

24. Stone & Webster Engineering Calculation No. 12 966-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) (See JAF-CALC-RAD-00042 (Ref. 5) for a copy of this reference.]

25. 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) (See JAF-CALC-RAD-00042 (Ref. 5), for a copy of this reference.]

26. JAF Emergency Plan Implementing Procedure EAP-44, " Core Damage Estimation" (July 1991)

27. JAF Original FSAR, Supplement 25, " Effects of High Energy Piping System Breaks Outside of Primary Containment,"

(7/22/74)

28. GE Technical Report NEDE-31152P, "GE Fuel Bundle Designs,"

Rev. 3 (February 1993)

29. R. G. Jaeger, Ed., " Engineering Compendium on Radiation Shielding," Springer-Verlag, NY (1975)

30. SWEC Engineering Calculation #12966-RP-76-004, "LOCA Six-Month Gamma Doses for IE 79-01B Equipment Qualifications" (9/29/80)

31. JAF Procedure AP-08.02, " Failed Fuel Action Plan" (Rev. O, i

1/29/94)

32. Johnson Service Company, Test Report TLP-774-448 (02-4925-72), "FitzPatrick NPP Damper Leakage (D-1300 Series)

(11/29/72) (NYPA Microfiche No. 60067, frames 025-030) l [See JAF-CALC-RAD-00028 for a copy of this reference.]

33. Stone & Webster Engineering, Calculation CC-70-04,

" Calculation for Air Conditioning System Cooling Load" 1

r~ (9/2/70) (DSR #249252) l

( ,T/ ' 34. JAF Process Surveillance Procedure PSP-1, " Reactor Water

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N ,) TITLE: Power Uprate Project - Radiological Impact at Onsite and FINAL [X] CHECKED BY M dr- DATE it//hM 7 Offsite Outdoor Receptors Following Design-Basis Accidents Sampling and Analysis" (12/4/91)

35. Niagara Mohawk Power Corporation letter # NMP80969 addressed to J. Hamawi, from Tom Galletta, titled "The Validity of the Nine Mile Point (NMP) Meteorological Data (1985-1990) Sent for Use in Updating the Offsite Dose Calculation Manual (ODCM)" (2/23/93) [See JAF-CALC-RAD-00007, Rev. 2, (Ref. 2) for a copy of this reference.]

36. CRE Calculation JAF-CALC-RAD-00025, Rev. 1, " Atmospheric Dispersion and Deposition Parameters for Routine Releases" (5/4/95) I

37. Empire State Electric Energy Research Corporation (ESEERCO) Technical Report No. EP 91-28, " Eastern Lake Ontario - On-Shore Flow Field Study," prepared by Galson Corp. (4/94) [See JAF-CALC-RAD-00007, Rev. 2, (Ref. 2) for a copy of this reference.]

38. CRE Calculation JAF-CALC-RAD-00005, "Drywell Personnel Access Lock Removable Wall Shielding Analysis" (12/23/91)

39. E-Mail Message addressed to D. Burch, from G. Lozier, l titled " Power Uprate NRC Submittals" (12/20/95) l

) 40. JAF Emergency Plan Implementing Procedure EAP-10,

" Protected Area Evacuation" (Rev. 12, 2/6/95)

41. JAF Emergency Plan Implementing Procedure EAP-11, " Site Evacuation" (Rev. 13, 2/6/95)

42. JAF DVP-01.02, " Radiological Effluent Controls and Offsite Dose Calculation Manual" (Rev. O, 12/28/93) 43.* ACTS Item #23847, " Revise Calculations which use SBGT Efficiency from 99% to 95%".

44. US NRC Nuclear Safety Evaluation related to JAF Amendment Modification #239 (Power Uprate) (12/96)

45. International Commission on Radiological Protection (ICRP)

Publication 30, " Limits for Intake by Workers" (Various Parts and Supplements, 1979-1982)

46. NYPA Letter JPN-96-055 addressed to the NRC, titled "JAFNPP - Additional Information Regarding Analyses at Power Uprate Conditions" (12/23/96) l l

l See Attachment A for a copy this reference v-

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~) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents List of Computer Proor=== 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 QAD-CGGP RAD-006 1.3.1.1 03/26/92 DG AViiON MATILDA ----- -----

07/28/92 DG AViiON (a) MATILDA was developed for use in conjunction with QAD-CGGP (Ref. 7) and the gamma spectra produced by DORITA-2 (Ref. 6) and two other CRE codes (namely, ELISA and ALLEGRA), for the computation of radiation exposures. It is documented in l Refs. 8(a) and 8(b). l l

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table of Contents Page CALCULATION CONTROL SHEET ................................. 1 STATEMENT OF PROBLEM ...................................... 2 REFERENCES ................................................ 4 LIST OF COMPUTER PROGRAMS EMPLOYED ........................ 8 TABLE OF CONTENTS ......................................... 9

1. INTRODUCTION ......................................... 11

2.

SUMMARY

OF RESULTS ................................... 17 2.1 Offsite Receptors ............................... 17 2.2 Onsite Outdoor Receptors - Immersion Dose Rates g* and Cumulative Doses ............................ 17

d 2.3 Onsite Outdoor Receptors - Reactor Building Shine ....... ................................... 18

3. METHODS OF ANALYSIS .................................. 36 l

4. RADIATION EXPOSURES FROM A LOSS OF COOLANT ACCIDENT .. 38 4.1 Drywell Leakage ................................. 39 4.1.1 Basic Data and Assumptions ............... 39 4.1.2 Results .................................. 42 4.2 ESF Component Leakage ........................... 44 4.2.1 Basic Data and Assumptions ............... 44 4.2.2 Results .................................. 45 4.3 Total LOCA Dose ................................. 48

5. RADIATION EXPOSURES FROM A MAIN STEAM LINE BREAK ..... 50 {

j 5.1 Basic Data and Assumptions ...................... 50 l 5.2 Results ......................................... 54 f'

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) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents l TABLE OF CONTENTS (Cont.)

Page

6. RADIATION EXPOSURES FROM A CONTROL ROD DROP ACCIDENT . 57 6.1 Basic Data and Assumptions ...................... 57 6.2 Results ......................................... 61

7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT ........ 63 7.1 Basic Data and Assumptions ...................... 63 7.2 Results ......................................... 66

8. RADIATION EXPOSURES FROM OTHER POST-LOCA SOURCES ..... 68 8.1 Direct Shine from Post-LOCA Airborne Radioactivity in the RB Refueling Level ......... 68 8.1.1 Basic Data and Assumptions ............... 68 8.1.2 Results .................................. 74 (Od ~ 8.2 . Direct Shine from Post-LOCA Airborne Radioactivity at El. 272' of the RB ............. 75 8.2.1 Basic Data and Assumptions................ 75 8.2.2 Results .................................. 78 ATTACHMENTS A. Excerpts from References Pertinent to this Calculation B. Copies of Computer Outputs

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) CHECKED BY FINAL N TITLE: Power Uprate Project - Radiological Impact at Onsite and l

Offsite Outdoor Receptors Following Design-Basis Accidents !

1.0 Introduction The radiological consequences of design-basis accidents at onsite and offsite outdoor receptors under power-uprate conditions at i JAF were originally assessed in CRE calculation JAF-CALC-RAD-00008 (Ref. 1). The present calculation re-evaluates the potential accident consequences for a number of reasons, as follows: .

1 (a) Incorporation of the recently revised atmospheric dispersion factors (Ref. 2),

(b) Revisions to the scenarios for a Main Steam Line Break Accident (MSLB, Ref. 3), and a Control Rod Drop Accident (CRDA, Ref. 4), and !

(c) Minor revisions to some of the accident assumptions, j for consistency with those employed in the revised Control Room Habitability analysis (Ref. 5).

The calculation addressing the post-accident atmospheric dispersion factors (Ref. 2) was recently updated to accommodate the following:

(1) The meteorological data base for calendar years 1985-1990 (which was made use of in JAF-CALC-RAD-00007, Rev.

0) was updated by Niagara Mohawk Power Corporation (Ref. 35) to adjust a minor (few-degree) miscalibration in the wind direction sensors.

(2) For consistency with the dispersion data in Ref. 42

[the JAF Offsite Dose Calculation Manual (ODCM)), the meteorological data base was extended to include 8 years' worth of hourly values (1985 through 1992),

(3) New information on onshore flows at the site (Ref. 37) was used to determine that only receptors at the SB and q- the LPZ can be affected by the prescribed assumption of V

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE J 2.- OF 98 PROJECT: JAF PRELN [] PREPARED BY ML DATE ll/J3/9 O

O FINAL [X] CHECKED BY M (.r - DATE 1/h 3/9 7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents fumigation conditions at the time of a design-basis LOCA. The location of the CR with respect to Lake Ontario krd the main stack is such that the potential for having a fumigation condition affecting the CR (as used in JAF-CALC-RAD-00008) is non-existent.

The revisions to the accident scenarios for an MSLB and a CRDA were as follows:

MSLB:

In contrast to Regulatory Guide 1.5 (Ref. 15), which is applicable under pre-uprate conditions at JAF, Sec. 15.6.4 I of the Standard Review Plan (SRP, Ref. 10, the guiding document in this calculation) extends the possible MSLB j

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locations to encompass not only the steam tunnel but all l locations "outside containment." In the original MSLB analysis under power-uprate conditions (Ref. 1), the rupture l location was assumed to be in the steam tunnel, which is not limiting. Recently identified information (Refs. 24 and 25) shows that a break in the 16" bypass line leading to the turbine bypass steam chest would release more reactor coolant than a break in one of the 24" main steam lines.

CRDA:

Implementation of Modification F1-93-086 during the December 1994 refueling outage, which eliminated the reactor-scram and MSIV-closure functions of the main-steam line radiation monitors, changed the release pathway of a design-basis CRDA. Under the old CRDA scenario, the Main Steam Isolation Valves (MSIVs) close and release of radioactivity to the atmosphere was due to turbine / condenser leakage into the turbine building. In the new scenario (without MSIV closure) the release could also be via the offgas system. The old scenario was used in the original power-uprate analysis

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) [I] MM- DATE ////3/q-7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents (Ref. 1); the present calculation uses the new scenario.

Note that the latest CRDA calculation at pre-uprate conditions (Ref. 4) was also based on the new scenario.

1 With respect to item (c) above, the list of other minor changes in the accident assumptions were as follows:

(1) The breathing rate at the LPZ was allowed to vary with time after the accident, in line with the model in Regulatory Guide 1.3 (Ref. 14); in JAF-CALC-RAD-00008, the high breathing rate of 3.47E-04 m'/sec was assumed for the duration of the accident.

(2) The atmospheric release rate of airborne radioactivity resulting from ESF component leakage into the reactor building was reduced from a conservatively selected r' - value (7.2 air changes per hour) to one which reflects the actual flow through the Standby Gas Treatment System (SGTS) (3.3 air changes per hour) .

(3) The release rate to the atmosphere associated with an MSLB was reduced from instantaneous to 3 air changes per hour, for consistency with the CR habitability model (Ref. 5). (Thyroid doses differ by about 3%

between these two cases.)

(4) The post-CRDA iodine plateout fraction within the steam lines and condenser was increased from 50% to 90% [Ref.

10 (SRP Sec. 15.4.9), Ref. 20 (Sec. 6.3.1.1), and Ref.

9] .

In addition to the above, JAF-CALC-RAD-00008 considered the dose ,

l rates to onsite outdoor receptors due to post-LOCA gamma

]

radiation emanating from airborne radioactivity within the RB l I

Refueling Level, and from gamma radiation streaming from the l

'l g drywell through the Personnel Access Lock (PAL). In the current l k- )

4 NYPA - CALC.# JAF-CALC-RAD-00048 KEV 1 PAGE /f OF M PROJECT: JAF PRELM [] PREPARED BY /hL DATE n/rg/97-Oj TITLE:

FINAL [X] CHECKED BY /2/,. DATE _ /V/Micf Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents releases as documented in Reference 2, and (c) use of the ICRP 30 (Ref. 45) dose conversion factors for the determination of thyroid doses.

Details are presented below.

JAF ACTS Item #23847 requested the revision of the power uprate radiological analyses using an SGTS filter efficiency of 95%

(instead of 99%). The intended purpose of this change is to reduce the filter efficiency test acceptance criteria from a penetration of 0.175% (as committed to the NRC in Ref. 46) to the less restrictive value of 1%.

In the present revision, an SGTS filter efficiency for the O- removal of halogens is conservatively assumed as 90% for all halogen species [Ref. 23 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. 33.

Although the higher efficiency has been accepted by the NRC during discussions on the power uprate analyses (Ref. 46) the present analysis employs a lower filter efficiency to provide some relief for testing acceptance criteria.]

Dispersion factors for elevated releases were reanalyzed to accommodate possible short-term meandering and looping effects by placing the control room intake at that distance from the stack where concentrations would peak. See revision 2 of Reference 2 for details.

Thyroid dose analyses based on ICRP-2 (or, equivalently, TID-14844) yield results which are higher than those based on the more up-to-date factors in ICRP-30. Indeed, use of the latter

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FINAL [I] CHECKED BY M 6- DATE _////3/# 7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents can reduce the dose estimates by about 40%. It is CRE's understanding that radiological analyses based on ICRP-30, and submitted to the NRC by various vendors and utilities, are not uncommon. In fact, the NRC confirmatory analyses for the JAF power uprate (Ref. 44) were based on ICRP-30. Following discussions with Licensing, it was decided to employ the ICRP-30 dose conversion factors for the present revision.

Reference 5, Revs. 1 and 2, document the analysis for an MSLB with an assumed pre-accident iodine spike corresponding to the maximum iodine concentration stated in the technical specifications as a Limited Condition of operation (namely, 2 pCi/gm I-131 DE for JAF), in addition to the equilibrium value for continued full power operation (namely, 0.2 pCi/gm I-131 DE).

For offsite receptors, there is no difference in the relative

[}- exposure with respect to the guidelines, since the 10-fold increase in the RCS concentration (from equilibrium to spiked conditions) is offset by the 10-fold increase in the acceptable dose. As a result, only the equilibrium value is analyzed in the present calculation.

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G[ FINAL [Z] CEECKED BY TITLE: Power Uprate Project - Radiological Impact at Onsite and DATE _lyfv97 Offsite Outdoor Receptors Following Design-Basis Accidents 2.0 Summary of Results 2.1 Offsite Receptors The following design-basis accidents were considered in the re-assessment of the radiological consequences at JAF under power uprate conditions:

(a) Loss of coolant accident (LOCA) (drywell leakage and ESF component leakage pathways),

(b) Main Steam Line Break outside containment (MSLB),

(c) Control Rod Drop Accident (CRDA), and (d) Refueling accident (RA). i The basic data and assumptions in each of the four accident scenarios are consistent with the current licensing basis and the

( models in the regulatory guides (Refs. 14 -

19) and the Standard Review Plan (SRP, Ref. 10). Complete details for each accident are presented in Secs. 4 through 7. A summary of the principal assumptions associated with each DBA and the ensuing immersion doses at the site boundary appear in Table 2.1. The doses at the Low Popalation Zone (LPZ) are presented in Table 2.2.

From Tables 2.1 and 2.2, it is seen that the highest immersion dose is 68.7 rem (to the thyroid at the LPZ following a design-basis LOCA) which is about 23% of the regulatory limit.

2.2 Onsite Outdoor Receptors - Immersion Dose Rates and Cumulative Domes Post-accident immersion dose rates and cumulative doses were also calculated for onsite outdoor receptors at grade elevation in the general vicinity of the old administration building. The results are shown in Tables 2.3 and 2.4. The worst-case dose rates are O-

- - - - - - - - - - - - - - - - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ - - J

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) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 31.2 rem /hr to the thyroid following an MSLB, and 0.22 rem /hr to the whole body at the start of a LOCA. The worst-case cumulative doses are 32.4 rem to the thyroid for a CRDA and 0.85 rem to the whole body for a LOCA.

2.3 Onsite Outdoor Receptors - Reactor Building Shine The results presented below are unchanged from revision 0 of this calculation.

Direct shine radiation fields at various onsite outdoor locations due to post-LOCA airborne radioactivity accumulating on the RB refueling level are presented in Table 2.5. The worst-case dose rate amongst the analyzed receptors is at Loc. #1 (due West from

~

the RB) and amounts to 3.4 rad /hr at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the postulated accident. At the farthest receptor (Loc. #10), the dose rate is 0.24 rad /hr. These dose rates are sufficiently high to require swift access to and from the plant to minimize personnel ;

exposures. The radiation fields remain high for several days, j dropping below 10% of the maxima at about 1 week after the accident.2 For informational purposes, the refueling-level direct shine dose rates at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the postulated LOCA (from Table 2.5) are also presented in Fig. 2.2 as a function of distance from the RB centerline. It is seen that the results follow a smooth curve, The peak dose rates in Table 2.5 are lower than those in '

JAF-CALC-RAD-00008 (Ref. 1) by at least a factor of 3. In the current calculation, the source accumulating in the refueling level was based on the actual RB air exchange rate of 3.3 air changes per day (via the SGTS, at 6000 scfm), in lieu of the conservatively selected rate of 1 air change per day in JAF-CALC-N ~~ RAD-00008.

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE /9 OF M PROJECT: JAF PRELM [] PREPARED BY /d(_ DATE ff/L3/19-FINAL [X] CHECKED BY 4/f- . DATE _11//3/97 V _' TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents B

with the exception of shielding edge effects which become apparent at close-in receptors. Thus, the dose rates in Fig. 2.2 can be applied along any direction from the reactor building.

Direct shine dose rates were also calculated at outdoor receptors adjacent to the east side of the reactor building, at distances ranging from contact with the 21" wall, to 21 ft. The source term in this case was post-LOCA airborne radioactivity within El.

272' of the reactor building. The results are summarized in Table 2.6. The worst-case dose rate (in contact with the wall) peaks at about 150 mrad /hr at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the accident. Due to the large size of the source and the closeness of the receptor locations analyzed to the RB, the dose rate drops relatively slowly with distance. At the farthest location analyzed (21 ft i"om the RB wall), the peak dose rate is 106 mrad /hr. However,

{x\ thece dose rates are low in comparison to those due to radiation emanating from the RB refueling level.

In summary, the post-LOCA radiation fields at onsite outdoor receptors are expected to be high. Indeed, plant procedures (Refs. 40 and 41) suggest that the protected area and site may have to be evacuated during the initial period following a LOCA.

The information presented in this calculation may be used to define the most appropriate emergency evacuation route to minimize exposures.

For the corresponding cumulative doses at the onsite receptors analyzed, refer to the computer outputs in Attachment B (MATILDA Run Cases #1 and #2). The worst-case post-LOCA gamma air dose I

due to shine from the refueling level of the reactor building is 237 rads, at receptor location #1.

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FINAL [X] CHECKED BY Md DATE II//J/97 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.1 JAF Power Uprate Project - Site Boundary Doses Following Postulated Design-Basis Accidents l

Design-Basis Thyroid Wh. Body Skin 1 Accident (rem) (rem) (rem)

LOCA l

l Regulatory Limit 300.0 25.0 ---- ;

l Drywell Leak 5.82E+01 2.32E+00 4.06E+00 ESF Leakage 3.99E+00 2.20E-02 3.40E-02 Total 2-hr Dose 6.22E+01 2.34E+00 4.09E+00

% of Reg. Limit 20.7% 9.36% ----

l MSLB Regulatory Limit"' 30.0 2.5 ---- l Total 2-hr Dose 5.573E-01 7.056E-03 1.113E-02

% of Reg. Limit 1.86% 0.28% ----

CRDA Regulatory Limit *' 75.0 6.0 ---- l Total 2-hr Dose 1.855E-01 1.312E-02 2.514E-02

% of Reg. Limit 0.25% 0.22% ----

RA Regulatery Limit"' 75.0 6.0 ----

Total 2-hr Dose 7.376E-01 9.131E-02 2.108E-01

% of Reg. Limit 0.98% 1.52% ----

(a) Ref. 10, SRP, Sec. 15.6.4 (10% of 10 CFR 100)

(b) Ref. 10, SRP, Sec. 15.4.9 (25% of 10 CFR 100)

()^' (c) Ref. 10, SRP, Sec. 15.7.4 (25% of 10 CFR 100)

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.. TITLE: Power Uprate Project - Radiological Impact at Onsite and l Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.1 (Continued)

BASES (Refer to the pertinent sections for references) i I

1 LOCA (Drywell Leakage)

(a) A LOCA takes place at full power (2535.8 MWt + 2%

uncertainty).

s ) All core-inventory noble gases and 25% of the halogens become airborne within the drywell at the time of the -

accident and are available for release.

(c) Leakage from the drywell is at the rate of 1.5% per day, consisting of 1.27% per day due to containment leakage, and 0.23% per day due to MSIV leakage.

(d) Noble gases and halogens leaking from the drywell are l exhausted to the atmosphere via the Standby Gas Treatment I System (SGTS) and the main stack without holdup or mixing in l

(-}g the reactor building.

(e) The SGTS filter efficiency is 90% for the removal of all halogen species. l (f) 4-hour fumigation conditions prevail at the time of the accident.

LOCA (ESF Leakage)

(a) 50% of the core-inventory halogens mix. uniformly with the coolant in the RHR system (113,4 00 f t') .

(b) The ESF leakage rate is 5 gpm, and is constant from the start of the LOCA through the duration of the accident.

(c) An additional 30-minute leakage of 50 gpm (due to gross failure of a passive component) is conservatively assumed to begin at the time of the accident.

(d) 10% of the halogens in the leaking fluids become airborne and mix uniformly with the reactor building atmosphere (2. 6E+06 f t') .

(e) Release from the reactor building is through the SGTS and

.. the main stack at the rate of 6000 scfm.

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.1 (Continued)

(f) 4-hour fumigation conditions prevail at the time of the accident.

Main steam Line Break (a) A line break occurs in the 16" bypass line leading to the turbine steam chest outside containment during full power j 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 amounts 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 limit for the maximum permissible reactor coolant system (RCS) activity under power uprate conditions, namely 0.2 pCi/gm I-131 Dose Equivalent.

(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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The release rate was selected to be equivalent to 3 air changes per hour.

Control Rod Drop Accident I (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

(-')

(_j

+ conditions within 30 minutes of the departure from design power.

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TITLE: Power Up rate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.1 (Continued)

(b) A CRDA takes place leading to the failure of 850 fuel rods at a core location with a radial power peaking factor of I

1.5.

(c) All activity within the gaps of the failed fuel rods is released to the reactor coolant and is instantaneously and uniformly mixed with the coolant in the pressure vessel at the time of the accident. The released activity corresponds to.10% of all halogens and 10% of all noble gases (except Kr l

85) in each failed rod, and to 30% of the Kr 85 inventory. j (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 !

() (f) stay open and the release is to the offgas system.

As a result of plant shutdown following a CRDA, or as a result of offgas system automatic isolation due to high ~

radiation fields at the offgas monitors (following a 15-minute delay, which is not considered in the analysis), the released radioactivity is retained within the turbine, condensers and the offgas system. Release to the environs is due to leakage from the various contaminated systems into the turbine building. [ Note: Without offgas system isolation, releases would be via the charcoal holdup system and the stack and would be significantly less restrictive than the scenario analyzed.)

(g) 90% of the iodines plate out on system internal surfaces.

l (h) The leakage rate from contaminated systems into the turbine building amounts to 1% per day and lasts for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The l release to the atmosphere is at ground level and there is no l holdup within the turbine building.

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Refueling Accident (a) The reactor has been operating at full power for an extended period of time.

(b) The reactor ic shutdown, refueling operations are initiated and an RA takes place at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 conservatively assumed to correspond to 10% of all halogens (except I 129) and 10% of all noble gases (except Kr 85) in O. 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 I atmosphere via the SGTS and the main stack over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period. The release. rate was selected to be equivalent to 3 air changes per hour.

(h) 4-hour fumigation conditions prevail at the time of the accident.

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FINAL [Z) CHECKED BY TITLE: Power Uprate Project - Radiological Impact at Onsite'and Offsite Outdoor Receptors Following Design-Basis Accidents 44- DATE fy/22,7 Table 2.2 JAF Power Uprate Project - Doses at the Low Population Zone Following Postulated Design-Basis Accidents Design-Basis Thyroid Wh. Body Skin Accident (rem) (rem) (rem).

LOCA Regulatory Limit 300.0 25.0 75.0 Drywell Leak 6.317E+01 1.856E+00 3.124E+00 ESF Leakage 5.496E+00 3.648E-02 5.50SE-02 Total 30-day dose 6.87E+01 1.89E+00 3.18E+00

% of Reg. Limit 22.9% 7.56% ----

MSLB v Regulatory Limit 30.0 2.5 ----

Total 24-hr Dose 6.241E-02 8.616E-04 1.330E-03

% of Reg. Limit 0.21% 0.03% ----

CRDA Regulatory Limit 75.0 6.0 ----

Total 24-hr Dose 1.258E-01 4.524E-03 8.417E-03

% of Reg. Limit 0.17% 0.08% ----

RA Regulatory Limit 75.0 6.0 ----

Total 24-hr Dose 2.879E-01 4.277E-02 9.290E-02

% of Reg. Limit 0.38% 0.71% ----

Refer to Table 2.1 for a listing of the basic assumptions and other details l

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.3 JAF Power Uprate Project - Dose Rates at Outdoor Receptors in the General Vicinity of the Old Administration Building Time Thyroid Whole Body Skin (hours) Dose Rate Dose Rate Dose Rate (rem /hr) (rem /hr) (rem /hr)

LOCA Drywell Leak 0.000E+00 5.238E-01 2.242E-01 3.204E-01 5.000E-01 5.190E-01 8.675E-02 1.130E-01 1.000E+00 5.145E-01 6.727E-02 8.748E-02 2.000E+00 5.060E-01 5.327E-02 6.927E-02 4.000E+00 4.907E-01 4.064E-02 5.323E-02 8.000E+00 4.648E-01 2.929E-02 3.928E-02

/~'\ 1.200E+01 3.228E-01 1.775E-02 2.415E-02 (ss/ 1.800E+01 3.028E-01 1.326E-02 1.838E-02 l 2.400E+01 2.860E-01 9.901E-03 1.394E-02 3.600E+01 1.301E-01 3.057E-03 4.404E-03 4.800E+01 1.196E-01 1.906E-03 2.849E-03 7.200E+01 1.038E-01 1.117E-03 1.749E-03 ,

9.600E+01 9.192E-02 8.805E-04 1.396E-03 1.680E+02 2.478E-02 2.330E-04 3.631E-04 3.360E+02 1.217E-02 8.667E-05 1.353E-04 7.440E+02 2.179E-03 8.554E-06 1.466E-05 l

s

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Tchle 2.3 (Continued)

JAF Power Uprate Project - Dose Rates at Outdoor Receptors in the General Vicinity of the Old Administration Building Time Thyroid Whole Body Skin (hours) Dose Rate Deze Rate Dose Rate (rem /hr) (rea/hr) (rem /hr)

LOCA ESF Leak 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 3.901E-02 8.288E-04 1.052E-03 1.000E+00 3.998E-02 8.447E-04 1.077E-03 2.000E+00 4.164E-02 8.998E-04 1.155E-03 4.000E+00 4.407E-02 1.051E-03 1.360E-03 8.000E+00 4.642E-02 1.233E-03 1.604E-03 1.200E+01 3.415E-02 9.205E-04 1.197E-03

1.800E+01 3.342E-02 7.638E-04 9.972E-04 2.400E+01 3.218E-02 5.707E-04 7.482E-04 3.600E+01 1.485E-02 1.536E-04 2.018E-04 4.800E+01 1.373E-02 7.462E-05 9.957E-05 7.200E+01 1.199E-02 2.607E-05 3.604E-05 9.600E+01 1.069E-02 1.644E-05 2.307E-05 1.680E+02 2.939E-03 4.171E-06 5.752E-06 3.360E+02 1.511E-03 1.962E-06 2.678E-06 7.440E+02 3.021E-04 3.552E-07 4.823E-07 MSLB 0.000E+00 3.118E+01 7.587E-02 2.786E-01 5.000E-01 6.829E+00 1.324E-02 4.692E-02 1.000E+00 1.497E+00 2.387E-03 8.542E-03 2.000E+00 7.216E-02 8.367E-05 3.091E-04 1

4.000E+00 1.687E-04 1.289E-07 5.142E-07 l

8.000E+00 9.373E-10 4.606E-13 2.090E-12 1,200E+01 4.507E-15 1.683E-18 8.391E-18 1.800E+01 6.123E-23 1.666E-26 9.234E-26 2.400E+01 8.447E-31 1.772E-34 1.057E-33

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) TITLE: Power Uprate Project - Radiological Impact at Onsite and 1

Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.3 (Continued) l JAF Power Uprate Project - Dose Rates at Outdoor Receptors in the General Vicinity of the Old Administration Building Time Thyroid Whole Body Skin (hours) Dose Rate Dose Rate Dose Rate

_(rem /hr) (rem /hr) (rea/hr)

CRDA 0.000E+00 1.735E+00 6.314E-02 5.627E-01 5.000E-01 1.719E+00 2.252E-02 1.182E-01 1.000E+00 1.704E+00 1.661E-02 8.534E-02 2.000E+00 1.677E+00 1.232E-02 6.089E-02 4.000E+00 1.627E+00 8.295E-03 4.005E-02 8.000E+00 1.542E+00 4.730E-03 2.534E-02 1.200E+01 1.256E+00 2.803E 1.702E-02 1.800E+01 1.179E+00 1.935E-03 1.339E-02 2.400E+01 1.116E+00 1.465E-03 1.110E-02 RA 0.000E+00 3.931E-02 1.707E-02 24 623E-02 5.000E-01 8.734E-03 3.995E-03 6.084E-03 1.000E+00 2.941E-03 1.201E-03 1.756E-03 2.000E+00 9.585E-05 4.279E-04 5.650E-04 4.000E+00 2.338E-07 3.182E-04 4.135E-04 8.000E+00 1.394E-12 2.103E-04 2.736E-04 1.200E+01 6.067E-18 1.052E-04 1.368E-04 1.800E+01 8.880E-26 5.693E-05 7.432E-05 2.400E+01 1.303E-33 3.102E-05 4.075E-05 1

Note: The results in this table were conservatively based on the atmospheric dispersion factors applicable to the control-room outside air intake located on the roof of the old administration building.

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Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.4 JAF Power Uprate Project - Integrated Doses (Continuous Occupancy) at Outdoor Receptors in the General Vicinity of the Old Administration Building i

l Design-Basis Thyroid Wh. Body Skin i Accident (rem) (rem) (rem) i LOCA Drywell Leak 2.45E+01 8.23E-01 1.13E+00 I ESF Leakage 2.72E+00 2.77E-02 3.63E-02 Total 30-day dose 2.72E+01 8.51E-01 1.17E+00 MSLB Total 24-hr dose 1.03E+01 2.17E-02 7.76E-02 O

V CRDA Total 24-hr dose 3.24E+01 1.23E-01 6.91E-01 RA Total 24-hr dose 1.31E-02 9.24E-03 1.34E-02 Note: The results in this table were conservatively based on the atmospheric dispersion factors applicable to the control-room outside air intake located on the roof of the old administration building.

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l JAF Power Uprate project - Onsite Outdoor Dose Rates (rad /hr) due I to Post-LOCA Shine from Airborne Activity in the Refueling Level Time (br) Loc. #1 Loc. #2 Loc. #3 Loc. #4 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 1.342E+00 1.200E+00 1.211E+00 1.230E+00 1.0 2.059E+00 1.840E+00 1.858E+00 1.888E+00 2.0 2.857E+00 2.550E+00 2.579E+00 2.620E+00 4.0 3.375E+00 3.002E+00 3.045E+00 3.095E+00 8.0 3.264E+00 2.886E+00 2.942E+00 2.993E+00 12.0 2.899E+00 2.549E+00 2.611E+00 2.658E+00 18.0 2.387E+00 2.084E+00 2.148E+00 2.188E+00 24.0 1.970E+00 1.709E+00 1.771E+00 1.806E+00 36.0 1.387E+00 1.188E+00 1.244E+00 1.272E+00 I 48.0 1.044E+00 8.854E-01 9.353E-01 9.582E-01 l 72.0 7.157E-01 5.974E-01 6.395E-01 6.570E-01 i

O~ 96.0 168.0 336.0 5.666E-01 3.611E-01 1.559E-01 4.694E-01 2.986E-01 1.306E-01 5.057E-01 3.222E-01 1.394E-01 5.202E-01 3.316E-01 1.431E-01 l

744.0 2.297E-02 1.977E-02 2.062E-02 2.106E-02 Time (hr) Loc. #5 Loc. #6 Loc. #7 Loc. #8 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 7.971E-01 3.881E-01 3.134E-01 1.375E-01 1.0 1.219E+00 5.925E-01 4.782E-01 2.093E-01 2.0 1.683E+00 8.147E-01 6.567E-01 2.863E-01 4.0 1.964E+00 9.423E-01 7.575E-01 3.266E-01 8.0 1.853E+00 8.710E-01 6.956E-01 2.916E-01 12.0 1.610E+00 7.426E-01 5.894E-01 2.405E-01 18.0 1.290E+00 5.828E-01 4.594E-01 1.817E-01 24.0 1.042E+00 4.638E-01 3.639E-01 1.410E-01 36.0 7.064E-01 3.081E-01 2.403E-01 9.097E-02 48.0 5.146E-01 2.210E-01 1.717E-01 6.408E-02 72.0 3.358E-01 1.408E-01 1.087E-01 3.968E-02 96.0 2.592E-01 1.071E-01 8.234E-02 2.957E-02 168.0 1.637E-01 6.708E-02 5.142E-02 1.822E-02

, 336.0 7.374E-02 3.082E-02 2.374E-02 8.532E-03 l 744.0 1.183E-02 5.138E-03 3.995E-03 1.477E-03 l

l Refer to Fig. 2.1 for the receptor locations.

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.5 (Continued)

JAF Power Uprate project - Onsite Outdoor Dose Rates (rad /hr) due to Post-LOCA Shine from Airborne Activity in the Refueling Level Time (hr) Loc. #9 Loc. #10 Loc. #11 Loc. #12 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 3.410E-01 1.019E-01 2.763E-01 2.023E-01 1.0 5.206E-01 1.550E-01 4.215E-01 3.083E-01 2.0 7.157E-01 2.117E-01 5.787E-01 4.225E-01 4.0 8.273E-01 2.405E-01 6.668E-01 4.847E-01 8.0 7.637E-01 2.123E-01 6.108E-01 4.391E-01 i 12.0 6.503E-01 1.731E-01 5.163E-01 3.673E-01 I 18.0 5.097E-01 1.291E-01 4.013E-01 2.821E-01 24.0 4.053E-01 9.934E-02 3.174E-01 2.213E-01 36.0 2.689E-01 6.346E-02 2.092E-01 1.445E-01 j 48.0 1.928E-01 4.446E-02 1.493E-01 1.025E-01 72.0 1.227E-01 2.732E-02 9.430E-02 6.419E-02 :

Ik-)' -

96.0 168.0 9.323E-02 5.837E-02 2.023E-02 1.239E-02 7.131E-02 4.449E-02 4.822E-02 2.993E-02 336.0 2.684E-02 5.823E-03 2.057E-02 1.393E-02 744.0 4.482E-03 1.015E-03 3.472E-03 2.380E-03 Time (hr) Loc. #13 Loc. #14 Loc. #15 Loc. #16 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 5.314E-01 2.861E-01 1.038E+00 5.638E-01 1.0 8.121E-01 4.364E-01 1.589E+00 8.617E-01 2.0 1.118E+00 5.991E-01 2.197E+00 1.187E+00 4.0 1.299E+00 6.902E-01 2.575E+00 1.380E+00 8.0 1.212E+00 6.319E-01 2.450E+00 1.290E+00 12.0 1.043E+00 5.338E-01 2.144E+00 1.111E+00 18.0 8.267E-01 4.147E-01 1.734E+00 8.821E-01 24.0 6.624E-01 3.278E-01 1.410E+00 7.077E-01 36.0 4.440E-01 2.160E-01 9.655E-01 4.750E-01 48.0 3.206E-01 1.541E-01 7.096E-01 3.435E-01 72.0 2.064E-01 9.729E-02 4.693E-01 2.215E-01

96.0 1.580E-01 7.353E-02 3.650E-01 1.697E-01 l 168.0 9.937E-02 4.587E-02 2.313E-01 1.068E-01 i l 336.0 4.528E-02 2.122E-02 1.030E-01 4.860E-02 744.0 7.427E-03 3.584E-03 1.614E-02 7.949E-03 (sv- Refer to Fig. 2.1 for the receptor locations.

(ms l l

l l

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE .32 OF M PROJECT: JAF PRELM [] PREPARED BY /drZ. DATE /t//*4/D l 'O FINAL [X] CEECKED BY /2(f DATE iI/fs/97

) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 2.5 (Continued)

BASIS i

l LOCA (Drywell Leakage) l (a) A LOCA takes place at full power (2535.8 MWt + 2% i uncertainty). 1 (b) All core-invento n noble gases and 25% of the halogens become airborne within the drywell at the time of the accident and are available for release.

(c) Leakage from the drywell is at the rate of 1.5% per day. !

(Note: The contribution of ESF component leakage was determined to be negligible in comparison to the drywell leakage.)

(d)

Noble gases and halogens leaking (2.60E+06 from the dryw')

f t ellandmix 4

uniformly with the RB atmosphere are exhausted to the atmosphere via the Standby Gas Treatment l

g' System (SGTS) and the main stack at a flow rate of 6000 scfm.

CF

1 l

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents '

l l Table 2.6 JAF Power Uprate project - Dose Rates (rad /hr) at Various Distances from the RB East Wall due to Post-LOCA Shine from Airborne Radioactivity in RB El. 272' Time (hr) Contact 3 ft 6 ft 2_it 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 7.247E-02 6.815E-02 6.422E-02 6.079E-02 1.0 1.073E-01 1.009E-01 9.505E-02 8.998E-02 2.0 1.442E-01 1.357E-01 1.278E-01 1.210E-01 4.0 1.542E-01 1.450E-01 1.367E-01 1.294E-01 8.0 1.083E-01 1.019E-01 9.609E-02 9.101E-02 12.0 6.709E-02 6.318E-02 5.960E-02 5.647E-02 18.0 3.366E-02 3.172E-02 2.994E-02 2.838E-02 24.0 1.859E-02 1.754E-02 1.656E-02 1.571E-02 36.0 7.247E-03 6.843E-03 6.468E-03 6.138E-03 48.0 3.608E-03 3.410E-03 3.225E-03 3.062E-03 72.0 1.462E-03 1.383E-03 1.310E-03 1.244E-03 s _'.

96.0 8.277E-04 7.834E-04 7.421E-04 7.053E-04 168.0 3.427E-04 3.245E-04 3.0.77E-04 2.925E-04 336.0 1.519E-04 1.438E-04 1.364E-04 1.297E-04 744.0 2.735E-05 2.590E-05 2.456E-05 2.335E-05 Time (hr) 12 ft 15 ft 18 ft 21 ft 0.0 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.5 5.773E-02 5.491E-02 5.229E-02 4.981E-02 1.0 8.545E-02 8.129E-02 7.741E-02 7.374E-02 2.0 1.150E-01 1.094E-01 1.041E-01 9.922E-02 4.0 1.229E-01 1.170E-01 1.114E-01 1.062E-01 8.0 8.648E-02 8.231E-02 7.843E-02 7.476E-02 12.0 5.367E-02 5.111E-02 4.872E-02 4.647E-02 18.0 2.699E-02 2.572E-02 2.453E-02 2.341E-02 24.0 1.494E-02 1.425E-02 1.359E-02 1.298E-02 36.0 5.844E-03 5.575E-03 5.325E-03 5.090E-03 48.0 2.917E-03 2.785E-03 2.661E-03 2.546E-03 72.0 1.186E-03 1.133E-03 1.084E-03 1.038E-03 96.0 6.723E-04 6.425E-04 6.148E-04 5.888E-04 168.0 2.790E-04 2.667E-04 2.554E-04 2.448E-04 rg 336.0 1.237E-04 1.183E-04 1.133E-04 1.086E-04

( ) 744.0 2.228E-05 2.130E-05 2.039E-05 1.955E-05

1 NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 34 OF M PROJECT: JAF PRELN [] PREPARED BY /)f. DATE J///3/97 ,

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TITLE: Power Uprate Project - Radiological Impact at Onsite and l

l Offsite Outdoor Receptors Following Design-Basis Accidents

- -- . . - ..- - - - - - _.. . -- - - - - . ~. - -

l Receptor locations for the dose rates in Table 2.5 l Fig. 2.1 from airborne radioactivity accumulating in the RB refueling level.

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PROJECT: JAF PRELM FINAL

[]

[X]

PREPARED BY CHECKED BY

/)f_

M & DATE ////Op-f TITLE: Power tiprate Project - Radiological Impact at Onsite and

~

Offsite Outdoor Receptors Following Design-Basis Accidents i

Fig. 2.2 Direct Shine Dose Rates due to Post-IDCA Activity in the RB Refueling level at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after a IDCA (Receptor Locati:ns #1 - #16 from Table 2.5, arranged as a function of distance from the RB) o

~

LEGEND o - 4 Hours of ter accident

. b -

L

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8 e

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NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 34 OF M

, PROJECT: JAF PRELN [] PREPARED BY /4d2 DATE /Ml3/99

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents

3. METHODS OF ANALYSIS i

Post-accident radiation exposures at the locations of interest I were computed using the following:

a) The methodology and assumptions in the regulatory guides (Refs. 14 through 19) and the pertinent sections of the Standard Review Plan (Ref. 10),

b) Appropriate source terms, release pathways, decontamination factors and other assumptions, l

c) Post-accident atmospheric dispersion factors based on 8-years' worth of hourly meteorological data collected (O '~

on site by Niagara Mohawk, from JAF-CALC-RAD-00007, Rev. 2 (Ref. 2), and d) The following CRE Computer Codes:

DORITA-2 (Ref. 6) Computation of radiation exposures, and definition of gamma spectra associated with post-LOCA airborne radioactivity within the RB.

QAD-CGGP (Ref. 7) Determination of the relative gamma fluxes at the locations of interest (in terms of MeV/sec-cm' per MeV/sec emitted by a source, as a function of gamma energy), for gamma radiation emanating from the RB refueling level and from El'.

272'.

MATILDA (Ref. 8) Computation of dose rates (and

(~'g cumulative doses) at the receptors

, (_)

1

l NYPA - CALC.# JAF-CALC-RAD-00048 PROJECT: JAF REV 1 PAGE 33 oP W i PRELM [] PREPARED BY /%2, DATE _ld/Y9}

, FINAL II] CHECKED BY M P DATE tt/n/9 7

,' TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 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.

Refer to Secs. 4 through 8 for further details.

O I

{

O E_________________________

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 3S OF .

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PROJECT: JAF PRELM [] PREPARED BY /JdiL DATE /J//YO l

[ FINAL [X] CEECKED BY M(,. DATE /I//5/,7 l j TITLE: Power Uprate Project - Radiological Impact at Onsite and. j Offsite Outdoor Receptors Following Design-Basis Accidents j

4. RADIATION EXPOSURES FROM A LOSS OF COOLANT ACCIDENT Release pathways and contributing radiation sources which I are typically addressed in the analysis of a LOCA are the following:

(a) Drywell leakage and ESF component leakage, followed by l atmospheric releases and cloud exposures, (b) Direct gamma radiation from airborne radioactivity accumulating on the refueling floor of the reactor building, and (c) MSIV leakage.

This part of the calculation addresses the immersion exposures at onsite and offsite outdoor receptors due to drywell and ESF component leakage. Shine from the RB [ Item (b) above] is addressed in Sec. 8 of this calculation. The MSIV-leakage

p. pathway is not applicable at JAF since the plant 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; see Ref. 5 for more information.

l

I NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 39 or M PROJECT: JAF PRELM [] PREPARED BY [E DATE f//1 Y M FINAL (Z) CHECKED BY d (,- D A T E _ fi// i h ~/

l x- ,) TITLE: Power Uprate Project - Radiological Impact at Onsite and l Offsite Outdoor Receptors Following Design-Basis Accidents 1

4.1 Drywell Leakage 4.1.1 Basic Data and Assumptions The following data and assumptions were used in the computation of immersion exposures at outdoor receptors as a result of post-LOCA drywell leakage:

(a) A LOCA takes place at full power (2535.8 MWt + 2%

uncertainty, i.e., 2586.5 MWt) [Ref. 11 and Reg. Guide 1.49 (Ref. 17)].

(b) The full-power core inventory for the radionuclides of interest is shown in the table wnich follows (based on information from Ref. 12):

Nuclide Activ. (Ci) Nuclide Activ. (Ci)

Br 83 8.078E+06* Kr 83m 8.114E+06

/"N - 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 1.255E+08 Xe 138 1.192E+08 3.123E+03 (Ci/MWt from Ref. 12) x 2586.5 (MWt)

(c) 100% of the noble gases and 25% of the halogens present in the core are released instantaneously to the drywell where they are available as an aerosol for leakage to the secondary containment [ Reg. Guide 1.3 (Ref. 14)].

(d) The halogen composition airborne within the drywell is as follows: 91% elemental, 4% organic and 5% particulate [ Reg. j Guide 1.3 (Ref. 14)]. !

(e) Leakage from the drywell is at the rate of 1.5% per day i

\ms/ i i

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 40 OP Si l PROJECT: JAP PRELN [] PREPARED BY /Mt DATE fiff'A[f ?

O) TITLE: Power Uprate Project - Radiological Impact at Onsite and FINAL [I] CEECKED BY Offsite Outdoor Receptors Following Design-Basis Accidents g/,- DATE JihfA7 (UFSAR, Rev. O, 7/82, Secs. 14.8.1.5 and 14.8-22). This leak rate accounts for both drywell containment leakage and MSIV leakage and is assumed to be constant for the accident duration. The design leak rate is 0.5% per day of containment volume (Technical Specifications Sec. 4.7.A.2.8, and UFSAR, Rev. O, 7/82, Secs. 11. 5. 3.10 and 14 . 6.1. 3. 5) .

Use of the 1.5% per day value is conservative.

(f) All the noble gases and halogens leaking from the drywell are instantaneously exhausted to the atmosphere via the Standby Gas Treatment System (SGTS) and the main stack without mixing in the reactor building.

(g) The SGTS filter efficiency for the removal of halogens is ,

90% for all halogen species (Ref. 19). Although an efficiency of 95 percent for halogen removal could have been

(} employed for the SGTS filters, the present revision conservatively employs an efficiency of 90% to provide some relief in testing.

(h) The atmospheric dispersion factors associated with the transport of released radioactivity to the onsite and offsite outdoor receptors of interest are as follows (from revision 2 of Ref. 2)>

Time Dispersion Parameter (sec/m 8)

Receptor Interval Location (hrs) Conc. Z/Q Gamma I/Q SB 0- 2 5.24E-5 4.75E-5 LPZ 0- 4 2.04E-5 1.90E-5 4 - 8 2.17E-6 3.91E-6 8- 24 9.53E-7 1.52E-6 24 - 96 3.90E-7 5.68E-7 96 - 744 1.08E-7 1.38E-7 Onsite 0 - 8 9.26E-7 3.24E-6 8- 24 6.75E-7 2.45E-6 24 - 96 3.39E-7 1.34E-6 96 - 744 1.26E-7 5.60E-7 O~

t NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 4l OF WI PROJECT: JAF PRELM [] PREPARED BY /44_ DATE ////1[%

(7,

\~)

\ FINAL [I] CHECKED BY N6- DATE //33/y 7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents i Note the following:

1. The concentration (X/Q) s are for computing the inhalation exposures and the beta component of the skin dose.

2. The gamma (X/Q)s are for computing the whole body doses due to exposure to finite radioactive clouds above.

3. The dispersion parameters at the SB and during the first 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at the LPZ are due to the prescribed assumption of fumigation conditions prevailing at the site at the time of the accident.

4. The dispersion parameters listed above for onsite j outdoor receptors are for the CR air intake (from Ref.

2). They were conservatively assumed to apply to onsite outdoor receptors at grade elevation, in the general vicinity of the old administration building.

'[}- Fumigation conditions at these locations are not !

applicable; see Ref. 2 for details.

(i) The breathing rates at the various receptors of interest were as follows (Ref. 14 for the SB and LPZ, and conservative high breathing rate for the onsite outdoor receptors):

Time Breathing Receptor Interval Rate Location (hrs) (m8 /sec)

SB 0- 2 3.47E-4 LPZ 0- 8 3.47E-4 8- 24 1.75E-4 24 - 744 2.32E-4 Onsite 0 - 744 3.47E-4 i

(j) Thyroid exposures were based on the dose conversion factors

/

in ICRP-30 (Ref. 45).

U]

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 4 2._ OF '4S PROJECT: JAF PRELN [] PREPARED BY /hf DATE //h@

FINAL C' DECKED BY M 6- DATE d/n/g7 k]s j s [I]

TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 4.1.2 Results Radiation exposures at the onsite and offsite outdoor receptors of interest due to drywell 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 Cases #1, #2 and #3).

Table 4.1 which follows presents the time-dependent thyroid, l

whole body and skin doses at the receptors of interest due to post-LOCA drywell leakage. Refer to Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to Table 2.3 for the time-dependent dose rates at onsite outdoor receptor locations in the ,

general vicinity of the old administration building.

l 1

I

A NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE Q3 OF %8 PROJECT: JAF PRELM [] PREPARED BY /h? DATE n//3/$h 9 ,

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 4.1 JAF Power Uprate Project - Post-LOCA Time-Dependent Doses at Outdoor Receptors Due to Drywell Leakage Time Thyroid Whole Body Skin (hours) Dose frem) Dose frem) Dose frem) i SB:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 5.824E+01 2.321E+00 4.059E+00 LPZ:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 2.267E+01 9.283E-01 1.609E+00 4.000E+00 4.462E+01 1.470E+00 2.508E400 8.~000E+00 4.910E+01 1.635E+00 2.748E+00 2.400E+01 5.263E+01 1.784E+00 2.984E+00 9.600E+01 5.890E+01 1.842E+00 3.092E+00 7.440E+02 6.317E+01 1.856E+00 3.124E+00 Onsite 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.607E-01 6.142E-02 8.248E-02 1.000E+00 5.190E-01 9.912E-02 1.315E-01 2.000E+00 1.029E+00 1.583E-01 2.085E-01 4.000E+00 2.026E+00 2.507E-01 3.290E-01 8.000E+00 3.935E+00 3.872E-01 5.098E-01 1.200E+01 5.257E+00 4.664E-01 6.165E-01 1.800E+01 7.132E+00 5.587E-01 7.433E-01 2.400E+01 8.897E+00 6.277E-01 8.397E-01 3.600E+01 1.054E+01 6.770E-01 9.094E-01 4.800E+01 1.203E+01 7.059E-01 9.517E-01 7.200E+01 1.470E+01 7.399E-01 1.004E+00 9.600E+01 1.705E+01 7.635E-01 1.041E+00 1.680E+02 1.914E+01 7.846E-01 1.07*E+00 3.360E+02 2.212E+01 8.094E-01 1.113E+00 7.440E+02 2.449E+01 8.230E-01 1.134E+00

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE l/ l/ OF WT PROJECT: JAF PRELM [] PREPARED BY R DATE 11//3/%

O FINAL [X] CHECKED BY M G- DATE E////97 TITLE: Power Uprate Project - Radiological Impact at Onsite and~

Offsite Outdoor Receptors Following Design-Basis Accidents 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 Sec. 4.1.1, Item (b).

(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 3.21 x 10' grams (Ref. 26). This is equal to approximately 113,400 ft', consisting of (431190 lbs / 62.4 lbs/ft') = 6,900 f t' of cold RCS coolant (from JAF Drawing 5.01-101A), 105,600 f't of torus water (from UFSAR, Rev, 0, 7/82, Table 5.2-1), and 900 ft' of water from other sources.

(d) Total ESF component leakage rate into the RB 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 component with an assumed leak rate of 50 gpm is included in the model (Ref. 10, SRP, Sec. 15.6.5, Appendix

~

B). This leakage is assumed to begin at the time of LOCA 1

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE (/ f OF M PROJECT: JAF PRELN [] PREPARED BY /)R. DATE /f /13[f f FINAL [I] CHECKED BY N4- _ DATE Ii//3/97

,) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents onset and lasts for a period of 30 minutes. (This assumption is more conservative than the SRP model which assumes that the additional leakage begins at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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. 10, SRP, Sec. 15.6.5, Appendix B), and mix uniformly with the RB atmosphere.

(h) Release from the reactor building is through the SGTS and the main 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 (Ref. 19); see discussion in ;

previous section for additional details. l (j) The atmospheric dispersion factors associated with the l transport of released radioactivity to the onsite and offsite outdoor receptors of interest, and other exposure-related parameters are described under Items (h), (i) and l (j) in Sec. 4.1.1.

4.2.2 Results Radiation exposures at the onsite and offsite outdoor receptors of interest due to ESF component 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 Cases #1, #2 and #3).

Table 4.2 which follows presents the time-dependent thyroid, g-s -

whole body and skin doses at the receptors of interest due to

\~ J

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.' TITLE: Power Uprate Project - Radiological Impact at Onsite and ,

Offsite Outdoor Receptors Following Design-Basis Accidents post-LOCA ESF component leakage. Refer to Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to Table 2.3 for the time-dependent dose rates at onsite outdoor receptor locations in the general vicinity of the old administration building.

l l

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l

I i

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TITLE: Power Uprate Project - Radiological Impact at Onsite and l

Offsite Outdoor Receptors Following Design-Basis Accidents l

l Table 4.2 l

JAF Power Uprate. Project - Post-LOCA Thme-Dependent Doses at Outdoor Receptors Due to ESF Component Leakage l

l Time Thyroid Whole Body Skin (hours) Dose (rem) pose (rem) Dose frem)

SB:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 5.600E-01 3.151E-03 4.797E-03 2.000E+00 3.988E+00 2.203E-02 3.400E-02 LPZ 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 2.180E-01 1.260E-03 1.907E-03 2.000E+00 1.553E+00 8.812E-03 1.351E-02 4.000E+00 3.445E+00 2.025E-02 3.150E-02 8.000E+00 3.871E+00 2.585E-02 3.935E-02 l 2.400E+01 4.253E+00 3.382E-02 5.088E-02 l 0- 9.600E+01 7.440E+02 4.973E+00 5.496E+00 3.617E-02 3.648E-02 5.451E-02 5.505E-02 Onsite 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 9.896E-03 2.149E-04 2.725E-04 1.000E+00 2.965E-02 6.331E-04 8.044E-04 2.000E+00 7.048E-02 1.503E-03 1.917E-03 4.000E+00 1.564E-01 3.453E-03 4.431E-03 8.000E+00 3.382E-01 8.097E-03 1.046E-02 1.200E+01 4.745E-01 1.184E-02 1.532E-02 1.800E+01 6.777E-01 1.694E-02 2.197E-02 2.400E+01 8.746E-01 2.094E-02 2.720E-02 3.600E+01 1.061E+00 2.364E-02 3.073E-02 4.800E+01 1.232E+00 2.495E-02 3.246E-02 7.200E+01 1.539E+00 2.600E-02 3.388E-02 9.600E+01 1.811E+00 2.648E-02 3.456E-02 1.680E+02 2.057E+00 2.686E-02 3.507E-02 3.360E+02 2.418E+00 2.735E-02 3.574E-02 l

7.440E+02 2.724E+00 2.773E-02 3.626E-02 l /"' ~

l k.))

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 4.3 Total LOCA Dose The total LOCA radiation doses due to both drywell and ESF component leakage are shown in Table 4.3. The table was prepared by summing the results in Tables 4.1 and 4.2. Note that, for the 31-day exposures at the LPZ, drywell leakage contributes 92% of the total thyroid dose and 98.4% of the total whole body dose.

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l NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE (/9 OF STr PROJECT: JAF PRELM [] PREPARED BY A1t DATE ////3/9 )

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TITLE: Power Uprate Project - Radiological Impact at Onsite and 1 Offsite Outdoor Receptors Following Design-Basis Accidents Table 4.3 JAF Power Uprate Project - Post-LOCA Time-Dependent Domes at Outdoor Receptors Due to Drywell and ESF Component Leakage Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rem) Dose (rem)

SB:

0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.00E+00 6.22E+01 2.34E+00 4.09E+00 LPZ 0.00E+00 0.00E+00 0.00E+00 0.00E+00 2.00E+00 2.42E+01 9.37E-01 1.62E+00 4.00E+00 4.81E+01 1.49E+00 2.54E+00 8.00E+00 5.30E+01 1.66E+00 2.79E+00

('s Q,

2.40E+01 9.60E+01 5.69E+01 6.39E+01 1.82E+00 1.88E+00 3.03E+00 3.15E+00 7.44E+02 6.87E+01 1.89E+00 3.18E+00 Onsite 0.00E+00 0.00E+00 0.00E+00 0.00E+00 '

5.00E-01 2.71E-01 6.16E-02 8.28E-02 1.00E+00 5.49E-01 9.98E-02 1.32E-01 2.00E+00 1.10E+00 1.60E-01 2.10E-01 4.00E+00 2.18E+00 2.54E-Oi 3.33E-01 8.00E+00 4.27E+00 3.95E-01 5.20E-01 1.20E+01 5.73E+00 4.78E-01 6.32E-01 1.80E+01 7.81E+00- 5.76E-01 7.65E-01 2.40E+01 9.77E+00 6.49E-01 8.67E-01 3.60E+01 1.16E+01 7.01E-01 9.40E-01 4.80E+01 1.33E+01 7.31E-01 9.84E-01 7.20E+01 1.62E+01 7.66E-01 1.04E+00 l 9.60E+01 1.89E+01 7.90E-01 1.08E+00 l

1.68E+02 2.12E+01 8.11E-01 1.11E+00 3.36E+02 2.45E+01 8.37E-01 1.15E+00 7.44E+02 2.72E+01 8.51E-01 1.17E+00

^

f U

l

1 NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE TO OF T8 PROJECT: JAF PRELM [] PREPARED BY N2_ DATE ///13/D O FINAL [X] CHECKED BY AEG DATE _ ///f3/9'7

) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents t

5. RADIATION EXPOSURES FROM A MAIN STEAM LINE DREAK 5.1 Basic Data and Assumptions As was the case with all accident analyses documented in this calculation, the computation of radiation exposures associated l with a postulated MSLB outside containment was based on data and l assumptions consistent with the regulatory guidelines, specifically, Ref. 15 (Regulatory Guide 1.5), and the Standard Review Plan (Ref. 10, Sec. 15.6.4). A description of the data and assumptions (as extracted from Ref. 5) follows.

(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) The accident involves a break in the 16" bypass line leading to the turbine bypass steam chest; this would release more reactor coolant into the turbine building than a break in one of the 24" main steam lines in the steam tunnel.

(d) The release through the break consist's of 11,621.5 lb of steam during the initial steam-phase flow, and 93,675.4 lb of steam and water during the two-phase flow.

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.

(f) 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 l operation (i.e., 100,000 Ci/sec following a 30-min decay).

/~N' 50% of all noble gases leaving the reactor vessel during the U

1 1

I NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE Cl OF '% T t PROJECT: JAF PREN [] PREPARED BY /ML. DATE ////t/7f l/ FINAL [X] CHECKED BY NF DATE _ ////U9 7

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) TITLE: Power Up' rate Project - Radiological Impact at Onsite and offsite Outdoor Receptors Following Design-Basis Accidents l 10.5-sec MSIV closure time are released via the break.

From Ref. 5 (Sec. 5.1), the total noble gas releases following an MSLB are as follows:

)

Nuclide MSLB Release (Ci)

Kr 83m 1.82E-02 Kr 85m 3.27E-02 Kr 85 1.07E-04 Kr 87 1.07E-01 Kr 88 1.07E-01 Kr 89 6.96E-01 Xe 131m 8.03E-05 Xe 133m 1.55E-03 Xe 133 4.39E-02 Xe 135m 1.39E-01

/T Xe 135 1.18E-01

(_) Xe 137 8.03E-01 Xe 138 4.77E-01 The halogen inventory in the steam was' determined to be insignificant in comparison to that in the discharged liquid, and was not considered.

(g) The halogen source term in the discharged liquid was selected to correspond to the proposed technical specification limit for the maximum permissible reactor coolant activity, namely 0.2 pCi/gm I-131 DE'. This is the GE Standard Technical Specification limit (Ref. 22). The specified RCS concentration is assumed to accommodate the pre-accident iodine spike which would occur as a result of reactor shutdown or depressurization of the primary system.

Also, it is conservatively assumed that the total two-phase I-131 DE (Dose Equivalent) is that concentration of I-131 l which alone would produce the same committed thyroid dose as all

.. the iodines in a given mixture.

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TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents flow release through the break (93,675.4 lbs of liquid and steam) would contain iodines at the concentrations equal to those for the liquid phase. Under these conditions, the total halogen activities discharged into the turbine building would be as follows (Ref. 5, Sec. 5.1) :

Nuclide MSLB Release (Ci)

Br 83 3.199 Br 84 5.247 Br 85 2.687 I-131 3.455 I-132 26.87 I-133 23.03 I-134 48.63 I-135 31.99 Activation products and other particulate in the coolant were neglected since they would not become airborne.

(h) 100 % of the coolant halogens discharged in the turbine building are assumed to become airborne and released to the atmosphere at ground level over a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 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 98.9 120 99.8

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Offsite Outdoor Receptors Following Design-Basis Accidents (i) The atmospheric dispersion factors associated with the transport of radioactivity at ground level to the outdoor receptors are as follows (from Ref . 2) :

Time Receptor Interval Dispersion Parameter (sec/m')

Location (hrs) Conc. X/Q Gamma X/Q SB 0- 2 1.79E-4 1.32E-4 LPZ 0- 8 2.00E-5 1.61E-5 8- 24 1.34E-5 1.06E-5 24 - 96 5.59E-6 4.27E-6 96 - 720 1.60E-6 1.16E-6 Onsite 0- 8 3.29E-3 4.06E-4 8- 24 2.81E-3 3.48E-4 24 - 96 2.00E-3 2.49E-4 96 - 720 1.22E-3 1.54E-4

()' '

Note the following:

1. The concentration (X/Q)s are for computing the inhalation exposures and the beta-component of the skin dose.

2. The gamma (X/Q)s are for computing the whole body doses due to exposure to finite radioactive clouds above.

3. The dispersion parameters listed above for onsite receptors are for the CR air intake (from Ref. 2).

They were conservatively assumed to apply to onsite outdoor receptors at grade elevation, in the general vicinity of the old administration building.

(j) The breathing rates at the various receptor locations are as described under Item (i) of Sec. 4.1.1.

!O^

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b FINAL [X] CEECKED BY M & DATE _pths/97 ,

TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 5.2 Results Radiation exposures at the onsite and offsite outdoor receptors of interest 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 Cases #1, #2 and #3).

Table 5.1 which follows presents the time-dependent thyroid, whole body and skin doses at the receptors of interest. Refer to Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to

' Table 2.3 for the time-dependent dose rates at onsite outdoor receptor locations in the general vicinity of the old administration building.

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NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE Tf' OF QS PROJECT: JAF PRELN [] PREPARED BY /W2_ DATE ////4[N FINAL [X] CHECKED BY A24 DATE I///9/97 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 5.1 JAF Power Uprate Project - Time-Dependent Doses at Outdoor Receptors Following a Main Steam Line Break Accident Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rem) Dose (rem)

SB:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 5.573E-01 7.056E-03 1.113E-02 LPZ 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 6.226E-02 8.606E-04 1.328E-03 8.000E+00 6.241E-02 8.616E-04 1.330E-03 2.400E+01 6.241E-02 8.616E-04 1.330E-03 Onsite 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 8.016E+00 1.786E-02 6.377E-02

( 1.000E+00 9.773E+00 2.102E-02 7.500E-02 2.000E+00 1.024E+01 2.170E-02 7.746E-02

~

4.000E+00 1.027E+01 2.173E-02 7.756E-02 8.000E+00 1.027E+01 2.173E-02 7.756E-02 1.200E+01 1.027E+01 2.173E-02 7.756E-02 1.800E+01 1.027E+01 2.173E-02 7.756E-02 l

2.400E+01 1.027E+01 2.173E-02 7.756E-02

, Note: The results for the "onsite" receptor in this table were conservatively based on the atmospheric dispersion factors applicable to the control-room outside air intake located on the roof of the old administration building.

1 I

l l

l

NYPA CALC.# JAF-CALC-RAD-00048 REV 1 PAGE G OF M PROJEift JAF PRELM [] PREPARED BY Af2,. DATE 1///VO FINAL [X] CEECKED BY M 6-- DATE W/5/97 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Table 5.1 JAF Power Uprate Project - Time-Dependent Doses at Outdoor Receptors Following a Main Steam Line Break Accident Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rem) Dose (rem)

SB:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 5.573E-01 7.056E-03 1.113E-02 LPZ 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 6.226E-02 8.606E-04 1.328E-03 8.000E+00 6.241E-02 8.616E-04 1.330E-03 2.400E+01 6.241E-02 8.616E-04 1.330E-03 Onsite 0.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 8.016E+00 1.786E-02 6.377E-02 0s - 1.000E+00 9.773E+00 2.102E-02 7.500E-02 2.000E+00 1.024E+01 2.170E-02 7.746E-02 4.000E+00 1.027E+01 2.173E-02 7.756E-02 8.000E+00 1.027E+01 2.173E-02 7.756E-02 1.200E+01 1.027E+01 2.173E-02 7.756E-02 1.800E+01 1.027E+01 2.173E-02 7.756E-02 2.400E+01 1.027E+01 2.173E-02 7.756E-02 Note: The results for the "onsite" receptor in this table were conservatively based on the atmospheric dispersion factors applicable to the control-room I outside air intake located on the roof of the old administration building.

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) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents

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 l are consistent with the guidance in the Standard Review Plan (Ref. 10, Sec. 15.4.9), applicable portions of Regulatory Guide l 1.77 (Ref. 18), the updated UFSAR, and JAF-CALC-RAD-00041 (Ref.

4). 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 conditions within 30 minutes of the departure from design power.

[ (b) The reactor power was at the level for design-basis accident analyses (i.e., 2586.5 MWt, from Sec. 4.1.1). The core inventory for the radionuclides of interest at the end of a 1000-day continuous operation is as shown under Item (b) in Sec. 4.1.1 of this calculation.

(c) A CRDA takes place and leads to the failure of 850 fuel rods (Ref. 20, Sec. 6.2.1, and Ref. 28, Sec. 3.7). The total number of fuel rods in the core is equal to 36472 (Ref. 1)'.

(Note: According to the UFSAR, Rev. O, 7/82, Sec.

14.6.1.2.4, the total number of fuel rods that fail following a CRDA is 330.)

UFSAR Table 3.2-1 lists the total number of fuel rods for Cycle 11 as 35784. This number will be revised to 38708 for Cycle 12. The number of fuel rods employed in the current calculation (36472) was selected for consistency with the original calculation (Ref. 1). As the fuel designs continue to change over the next few fuel cycles to a 10x10 bundle, the quantity of interest is the ratio of the number of failed fuel rods to the total number of rods in the core. The fuel failure Q- ratio for the CRDA employed in this calculation (850/36472) is C/ not expected to significantly change with the new fuel designs.

3 h

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_) TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents (d) The failed fuel rods were at a core location with a radial peaking factor of 1.5 (Ref. 10, SRP Sec. 15.4.9).

(e) All activity within the gaps of the failed fuel rods is released to the reactor coolant and is instantaneously and uniformly mixed with the coolant in the pressure vessel at

)

the time of the accident. The released activity is l t

conservatively assumed to correspond to 10% of all halogens 1 and 10% of all noble gases (30% for Kr 85) in each failed rod (Ref. 18, as recommended in the SRP) .

(f) Based on the above information, and without taking credit for 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 Sec. 4.1.1 of this calculation:

Multiplying factor for all noble gases except Kr 85:

1.5 (peaking factor) x (850 failed rods / 36472 rods) x 0.1 (gap fraction) = 3.496E-03 Multiplying factor for Kr 85:

1.5 (peaking factor) x (850/36472) x 0.3 (gap fraction)

= 1.049E-02 Multf. plying factor for all halogens:

1.5 (peaking factor) x (850/36472) x 0.1 (gap fraction) x 0.1 (fraction reaching turbine / condensers) l = 3.496E-04 l

l 1

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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 Kr 87 1.168E+05 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 I 132 3.477E+04 Xe 131m 1.430E+03 I 133 4.975E+04 Xe 133m 2.084E+04 I 134 5.476E'04 Xe 133 4.998E+05 I 135 4.697E+04 Xe 135m 9.422E+04 I 136 2.265E+04 Xe 135 6.456E+04 Xe 137 4.387E+05 Xe 138 4.168E+05

/

8.078E+06 (from Table 4.1) x 3.496E-04 N~s)' \

(g) As a result of elimination of the MSIV-closure and reactor-shutdown functions of the main steam line radiation monitors (modification No. F1-93-086) the 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. (See Ref. 4 for the radiological analysis of the revised CRDA scenario under pre-upiate conditions.]

(h) As a result of plant shutdown following a CRDA, or as a result of offgas system automatic isolation (following a 15-minute delay, which is not considered in the analysis) due to high radiation fields at the offgas monitors, the released radioactivity is retained within the turbine, condensers and the offgas system. Release to the environs is due to leakage from the various contaminated systems into the turbine building. [ Note: Without offgas system

-s g isolation, releases would be via the charcoal holdup system L)

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/

[D

\2

, ) TITLE: Power Up rate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents and the stack and would be significantly less restrictive than the scenario analyzed. See JAF-CALC-RAD-00041, Ref. 4, for a comparison under pre-uprate conditions.]

(i) Plateout and partitioning of the halogens in the turbine, condensers and other internal surfaces is conservatively )

assumed to be equal to 90% [Ref. 10 (SRP Sec. 15.4.9), Ref. l 20 (Sec. 6.3.1.1), and Ref. 9] .

[ Note: The 90% halogen depletion due to plateout and partitioning was numerically accounted for in the DORITA-2 runs by assuming filtration of the release.]

l (j) The leakage rate amounts to 1% per day and lasts for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Reg. Guide 1.77, Ref. 18). The release to the atmosphere is at ground level and there is no holdup within the turbine building.

(k)

(~ )

Transfer of the released radioactivity to the outdoor receptors is governed by the atmospheric dispersion factors listed under Item (i) in Sec. 5.1 of this calculation. The i breathing rates at the various receptor locations are as described under Item (i) of Sec. 4.1.1. I l

e u___--___---- - - _ - - - - - - - - - - - - - - - - - --

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE bl OF WS PROJECT: JAF PRELM [] PREPARED BY /#7 DATE J///4/O FINAL [X] CHECKED BY /t& DATE J///3/9 f

, TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 6.2 Results Radiation exposures at the onsite and offsite outdoor receptors of interest 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 Cases #1, #2 and #3).

Table 6.1 which follows presents the time-dependent thyroid, whole body and skin doses at the receptors of ir;terest. Refer to Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to Table 2.3 for the time-dependent dose rates at onsite outdoor receptor locations in the general vicinity of the old administration building.

O_

i

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l TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 1

Table 6.1 J JAF Power Uprate Project - Time-Dependent Doses at Outdoor Receptors Following a Control Rod Drop Accident Time Thyroid Whole Body Skin (hours) Dose frem) Dose (rem) Dose frem)

SB:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 1.855E-01 1.312E-02 2.514E-02 LPZ 0.0CJE+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 2.073E-02 1.600E-03 2.966E-03 8.000E+00 7.929E-02 3.386E-03 6.102E-03 2.400E+01 1.258E-01 4.524E-03 8.417E-03 Onsite 0.000E+00 0.000E+00 0.000E+00 0.000E+00 O '

5.000E-01 1.000E+00 8.635E-01 1.719E+00 1.667E-02 2.621E-02 1.079E-01 1.574E-01 l

2.000E+00 3.410E+00 4.035E-02 2.287E-01 4

4.000E+00 6.712E400 6.052E-02 3.266E-01 8.00pE+00 1.304E+01 8.539E-02 4.514E-01 1.2098+01 1.819E+01 9.879E-02 5.276E-01 1.86cd+01 2.549E+01 1.127E-01 6.178E-01 2.400E+01 3.237E+01 1.228E-01 6.908E-01 Note: The results for the "onsite" receptor in this table were conservatively based on the atmospheric dispersion factors applicable to the control-room outside air intake located on the roof of the old administration building.

l O .-

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FINAL [I] CEECKED BY M 6 - DATE N//i/f"/

,, TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidente

7. RADIATION EXPOSURES FROM A REFUELING ACCIDENT 7.1 Basic Data 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.

10, Sec. 15.7.4), Regulatory Guide 1.25 (Ref. 16), 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 under Item (b) in Sec. 4.1.1 of this calculation. l (c) The reactor is shutdown, refueling operations are initiated f--

and a refueling accident takes place at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after

( shutdown (Ref. 10, SRP Sec. 15.7.4).

(d) The accident involves a fuel assembly dropping from the l 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. 28, 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 (from Sec. 6.1).

(e) The failed fuel rods were at a core location with a radial peaking factor of 1.5 (Reg. Guide 1.25, Ref. 16).

(f) All activity within the gaps of the failed fuel rods is i released to the fuel pool water. The released activity is conservatively assumed to correspond to 10% of all halogens (except I 129) and 10% of all noble gases (except Kr 85) in each failed rod, and to 30% of I 129 and Kr 85 (Paf. 16).

(g) The noble gas and halogen inventories released to the fuel

_ pool (prior to adjust ant for decay from the time of reactor

s. -

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE dk OF M PROJECT: JAF PRELM [] PREPARED BY & DATE ff//t/M n (f. DATE )//hf7 O,,

FINAL [I] CEECKED BY TITLE: Power Uprate Project - Radiological Impact at Onsite and offsite Outdoor Receptors Following Design-Basis Accidents shutdown, which is handled by the DORITA-2 computer code) are as shown in the table which follows. They were computed by multiplying the core inventory in Sec. 4.1.1 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) x (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: l 1.5 (peaking factor) x (125/36472) x 0.3 (gap fraction) l

= 1.542E-03 l

l i

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE f[ OF WTr PROJECT: JAF PRELM [] PREPARED BY /$f DATE _////1[f4 FINAL [X] CEECKED BY p/(, - DATE lih5/e7

-f_ TITLE: Power U' p rate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Pre-Decay Refueling Accident Source Tema Nuclide Activ. (Ci) Nuclide Activ. (Ci)

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 129 3.477E-03 Kr 88 2.433E+04 I 130 1.391E+03 Kr 89 3.026E+04 I 131 3.498E+04 I 132 5.113E+04 Xe 131m 2.104E+02 I 133 7.316E+04 Xe 133m 3.065E+03 I 134 8.053E+04 Xe 133 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 Sec. 4.1.1) 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. 16). 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, L6) .

(i) The retention of noble gases by the pool water is negligible (i.e., noble gas DF = 1).

(j) Radioactive material that escapes the pool is released to the atmosphere via the SGTS and main stack over a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period (Ref. 16). The Reactor Building air exchange rate was arbitrarily set at the conservative value of 3 air changes per hour. At this release rate, which is the same as that used for releases from the turbine building

(} _,

following an MSLB, 99.8 % of all radioactivity would be released to the SGTS within the assumed 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period.

1 1

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE O OF '"t$

PROJECT: JAF PRELM [] PREPARED BY /)f_. DATE ////_1/@

FINAL [X] CHECKED BY M(, - DATE _ d//$/9 7

,_ TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents I

[ 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 per day. ;

(k) The halogen-removal filter efficiency of the SGTS is 90% for I all halogen species (Ref. 23).

(1) All releases to the atmosphere are via the main stack.

Transport of the released radioactivity to the outdoor receptors of interest is controlled by the dispersion l

factors listed under Item (h) in Sec. 4.1.1. The presence j of a 4-hour fumigation at the time of the accident is i accounted for. l (m) Other exposure-related parameters are as described in Items (i) and (j) of Sec. 4.1.1.

O(_/

7.2 Results Radiation exposures at the onsite and offsite outdoor receptors of interest following a design-basis Refueling Accident 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 Cases #1, #2 and #3).

Table 7.1 which follows presents the time-dependent thyroid, whole body and skin doses at the receptors of interest. Refer to Tables 2.1, 2.2 and 2.4 for a summary of the exposures, and to Table 2.3 for the time-dependent dose rates at onsite outdoor receptor locations in the general vicinity of the old administration building.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ . _ . - - - _ _ - _ - _ - - . - - - - - - - -- -- - - - - - - - - - - - - - - - - - - ' --~~ ' - - ' -

1 i

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE N OF 93 PROJECT: JAF PRELM [] PREPARED BY /)f__ DATE f///j/f} {

FINAL [X] CHECKED BY pt/f - DATE t//n/q7 j s-TITLE: Power Uprate Project - Radiological Impact at Onsite and j Offsite Outdoor Receptors Following Design-Basis Accidents Table 7.1 JAF Power Uprate Project - Time-Dependent Doses at Outdoor Receptors Following a Refueling Accident Time Thyroid Whole Body Skin (hours) Dose (rem) Dose (rem) Dose (rem)

SB:

0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 7.376E-01 9.131E 2.108E-01 LPZ 0.000E+00 0.000E+00 0.000E+00 0.000E+00 2.000E+00 2.872E-01 3.652E-02 8.329E-02 4.000E+00 2.879E-01 4.074E-02 9.003E-02 8.000E+00 2.879E-01 4.200E-02 9.179E-02 2.400E+01 2.879E-01 4.277E-02 9.290E-02 O Onsite O.000E+00 0.000E+00 0.000E+00 0.000E+00 5.000E-01 1.016E-02 4.448E-03 6.825E-03 1.000E+00 1.242E-02 5.576E-03 8.520E-03 2.000E+00 1.303E-02 6.228E-03 9.428E-03 1

4.000E+00 1.307E-02 6.947E-03 1.037E-02 8.000E+00 1.307E-02 7.989E-03 1.172E-02 ;

1.200E+01 1.307E-02 8.510E-03 1.240E-02 l 1.800E+01 1.307E-02 8.982E-03 1.301E-02 ,

i 2.400E+01 1.307E-02 9.238E-03 1.335E-02 Note: The results for the "onsite" receptor in this table were conservatively based on the atmospheric dispersion factors applicable to the control-room outside air intake located on the roof of the old administration building.

lo G

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE $8 OF M PROJECT: JAF PRELM [] P R E P A R E D B Y _ /) R_. DATE ///Ct/f f O FINAL [X] CHECKED BY M & DATE fl//7/97 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents

8. RADIATION EXPOSURES FROM OTHER POST-LOCA SOURCES This section addresses the direct shine radiation fields at various onsite outdoor receptors due to post-LOCA airborne radioactivity accumulating within the reactor building. Section 8.1 looks at the shine dose rates from activity in the refueling level, and Sec. 8.2 looks at the dose rates adjacent to the east side of the reactor building exterior wall.

Airborne radioactivity within the reactor building can result from either post-LOCA drywell leakage, post-LOCA ESF component l leakage, or from a Refueling Accident. Since the latter two are j l

relatively insignificant with respect to post-LOCA drywell

, leakage (See Table 2.3, for instance), they were not considered l k) in this part of the calculation.

The radiation dose rates and cumulative doses documented in this section are to air (rad /hr ani rad). They may be conservatively applied to the whole body (rem /hr and rem).

8.1 Direct Shine from Post-LOCA Airborne Radioactivity in 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

NYPA - CALC.# JAF-CALC-RAD-00048 REY 1 PAGE 6T OF '&T PROJECT: JAF PRELM [] PREPARED BY /ddiL. DATE ////t/Q FINAL [I] CHECKED BY g(p DATE n//7A ~7

,, TITLE: Power Uprate Project - Radiological Impact at Onsite and i Offsite Outdoor Receptors Following Design-Basis Accidents sec. 4.1 of this calculation. The items of interest in the definition of the airborne radioactivity within the reactor building and the associated time-dependent gamma spectra are as follows:

(a) A LOCA takes place at full power (2586.5 MWt). l (b) The core inventory for the radionuclides of interest is as shown under Item (b) in Sec. 4.1.1 of this ;

calculation.

(c) 100% of the core-inventory noble gases and 25% of the halogens become instantly airborne within the drywell atmosphere and are available for leakage to the secondary containment.

(d) The halogen composition airborne within the drywell is as follows: 91% elemental, 4% organic and 5% !

particulate.

(e) Leakage from the drywell is at the rate of 1.5% per day.

(f) As a result of the ventilation system, airborne radioactivity leaking from the drywell becomes uniformly distributed within the air volume of the reactor building (2.6E+06 f t', Ref. 30).

(g) Release from the reactor building is through the SGTS and the main stack at the rate of 6000 scfm (or 3.3 air changes per day).

. Source / Receptor Gecanetry The source / receptor geometry for use with QAD-CGGP is shown in Fig. 8.1. The primary component is the reactor building. 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 3.439E+04 m'.

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE @ OF 9-8 PROJECT: JAF PRELM [] PREPARED BY /jR_ DATE ////3/f')

FINAL (Z] CHECKED BY M6 - DATE fI//3/1~/

TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents The RB volume beneath the refueling level was conservatively represented as a box with 2-ft concrete walls all around. The concrete density was set at 2.35 g/cc and had the following composition (in weight percent, from Ref. 29, Vol. II, Table 9.1.12-77):

Fe: 1.19 H : 0.85 O : 50.64 Mg: 0.23 Ca: 8.03 Na: 1.66 Si: 30.49 A1: 4.44 S : 0.12 K: 1.87 The receptor locations were selected to be identical to those analyzed in Ref. 1, for consistency. They are shown in Fig. 8.2.

L)

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE % OF @'8 PROJECT: JAF PRELM [] PREPARED BY ANP DATE II/(3/D (q) , ~

FINAL [X] CHECKED BY M (.e-- DATE (l//3/9 7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Fig. 8.1 -

QAD-CGGP Source / Receptor Geometri (Direct shine from the RB Refueling Floor) y a;xis j l

y=162' SOURCE REGION l

l l

A (V 8 f(0,0) ,

- x axis z axis SOURCE REGION z=369.5'

( i'2-ft concrete box

. (all around) s s

f.sz=272'

, x ,x1, x x Il !!

O p

- y s E_____._____._____._____

(

. NYPA - CALC.# JTF-CALC-RAD-00048 REV 1 PAGE ~4 2 OF es PRELM [] PREPARED BY /N2'. DATE ip//q/43 t ; . PROJECT: JAF ~

l ' FINAL [X] CHECKED BY M6 DATE //h US7 l (j ~

TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Fig. 8.2 - Receptor Locations (Direct Shine from the RB Refueling Floor) ;

i a s i ys i _ _ _ .

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No. x (cm) y (ca) No. x (ca) y (ca) 1 1500 8000 2 1500 10000 3 0 8000 4 -1500 6000

$ -10000 2300 6 -15000 2300 7 -15000 10000 8 -20000 15000 9 1500 -15000 10 -25000 10000 11 10000 -15000 12 15000 -15000 T 20000 -4000 15000 -4000 14 j (#'-)

I 13 15 10000 8000 16 15000 8000 z = 8400 cm (or E1. 272') at all receptors rN (x=0, y=0 at the SE corner of the RB)

f NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE 93 or %

PROJECT: JAF PRELM [] PREPARED BY AfiL DATE /s//3[99 FINAL [X] CHICKED BY M(, DATE II//3/f 7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents ;

i

(

Fig. 8.2 - Receptor Locations )

l (Direct Shine from the RB Refueling Floor)

No. x (cm) y (ca) No. x (cm) y (cm) 1 1500 8000 2 1500 10000 3 0 8000 4 -1500 6000 5 -10000 2300 6 -15000 2300 7 -15000 10000 8 -20000 15000 9 1500 -15000 10 -25000 10000 11 10000 -15000 12 15000 -15000 13 15000 -4000 14 20000 -4000 15 10000 8000 16 15000 8000 l l

l Note: I z = 8400 cm (or El. 272') at all receptors (x=0, y=0 at the SE corner of the RB)

Os.

r l

t

!O -

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE M OF 9r8 PROJECT: JAF PRELN [] PREPARED BY /d DATE _11//M FINAL [1] CHECKED BY M (f - DATE (f//9/1 7 l TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 8.1.2 Results Post-LOCA direct-shine radiation levels at the various onsite outdoor locations of interest due to radioactivity accumulating i on the RB refueling level are shown in Table 2.5. These were extracted from QAD-CGGP and MATILDA Run Case #1 in Attachment B.

The gamma spectra were extracted from DORITA-2 Run Case #4.

Refer to Sec. 2.3 for general remarks, and to Fig. 2.2 for extrapolation of the worst-case dose rates (at 4 hrs after the postulated LOCA) to other receptors around the reactor building.

Cumulative doses, if of interest, can be found in the MATILDA I output. ;

Comparison of the dose rates presented in this subsection with J corresponding results in Ref. 1 (JAF-CALC-RAD-00008) shows that l I

the former are lower by a factor of approximately 3 to 4. The reason for this is the conservative RB air exchange rate assumed I in Ref. 1 for this part of the analysis, namely 1 air exchange per day in lieu of the more appropriate value of 3.3. Note also that, for the same reason, the dose rates in Ref. 1 peak at a later time, namely at about 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the accident instead of 4.

l 6

l NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE W OF '&&

PROJECT: JAF PRELM [] PREPARED BY /SC DATE ////%/O FINAL [I] CHECKED BY _M A DATE 1t[/3/97 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents a

8,2 Direct Shine from Post-LOCA Airborne Radioactivity at El. 272' of the RB j 8.2.1 Basic Data and Assumptions Direct shine dose rates from the RB were also calculated at outdoor receptors adjacent to the E side of the reactor building.

The source term in this case was post-LOCA airborne radioactivity within El. 272' of the reactor building and is identical to that described in Sec. 8.1.1 for the refueling level.

l The RB area of interest and the source / receptor geometry for use with QAD-CGGP are shown in Figs. 8.3 and 8.4. Note that the RB wall thickness in this area is 21" (from Fig. 8.3). The main components are the source region (with dimensions of 154 ft

() length, 50 ft depth, and 27.2 ft high), and the exterior wall (21" thick). The source volume is equal to 2.09E+05 f t', or 5.93E+03 m'. The receptor distances range from contact with the !

exterior wall to 21 ft, at 3-ft increments.

I l

O -

l NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE '% OF M PROJECT: JAF PRELM [] PREPARED BY /,ht DATE ////f/9'd-l (j s ~

(

\

FINAL [E] CEECKED BY p/(y- DATE i///5/9'/

TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Fig. 8.3 - RB El. 272' - Plant Drawing 11825-Fa-1E) d, (source area and receptors of interest)

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NYPA - CALC.# CAF-CALC-RAD-00048 REV 1 PAGE N OF Gr9 PROJECT: JAF PRELM [] PREPARED BY /d$iL DATE //// t/% 1

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FINAL [X] CHECKED BY /6Z M DATE ////j/f"/ TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Fig. 8.4 - QAD-CGGP Source / Receptor Geometry (Direct Shine th ough the RB E Wall) y axis J SOURCE REGION j (27.2' high) l t 7 154' , O ~ o an i t l x axis i 21" concrete 1 R wall' I 8 receptors (contact to 21') (4' above grade) O4 1Y

NYPA - CALC.# JAF-CALC-RAD-00048 REV 1 PAGE TT OF 48 PROJECT: JAF PRELM [] PREPARED BY /df?_ DATE _f///j/94 O, " FINAL (Z) CHECKED BY MM DATE _ ///f3/f 7 TITLE: Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents 8.2.2 Results Post-LOCA direct-shine radiation levels at the various onsite outdoor locations of interest due to RB shine through the 21" E wall are shown in Table 2.6. These were extracted from QAD-CGGP and MATILDA Run Case #2 in Attachment B. The gamma spectra were extracted from DORITA-2 Run Case #4. Refer to Sec. 2.3 for general remarks, and to the MATILDA output for cumulative doses, if of interest. O O_

NYPA Calculation No. JAF-CALC-RAD-00048, Rev. 1 Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents I l

Attachment A Excerpts from References Pertinent to this Calculation O.. I l O-

l l Printed: 10/30/97 ACTS CHANGE FORM ACTS #: 23847 I~l New Item

   ,w Change DUE DATE from 11/01/1997          to                  (* indic hard date)

O Change DEPARTMENT from PRE to >> Dept Mgr Apvl 0 change INDIVIDUAL from GR to O ACTS to Startup 1 0 CLOSURE ----->> Date Completed: O Startup to ACTS ACTS Type Code: RNRI Due Date: 11/01/1997 Priority: B Department Code: PRE Resp Indiv: GARY RE ACTS /Startup: A Source Document: JLIC-96-220 Search Code: Descriptn: REVISE CALCULATIONS WHICH USE SBGT EFFIENCY FROM 99*5 TO 95'. s l l 1.) For CHANGES, provide reason and current status of item. 2.) For CLOSURE, describe action taken and attach reference documents. Status: l l 'EN RNRI Closure: [wgiorla ] Initiator: Date: Department Manager: Date: Director / Gen Mgr: Date: QA Manager (RQA Only): Date: Sr. Sponsor (RBP Only): Date: Plant Mgr (RNYS & RBP) : Date: _ q PORC/SRC Chmn (RPOR/RSRC): Mtg# Date: AP-03.08 Rev 8 ACTION AND COMMITMENT TRACKING SYSTEM

ATTACHMEh"I' 1

NYPA Calculation No. JAF-CALC-RAD-00048, Rev. 1 v' Power Uprate Project - Radiological Impact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents Attachment B COPIES OF COMPUTER OUTPUTS I I l l O .- E

            -~                                  Included in this attachment are copies of the computer outputs pertinent to this calculation. They appear in the s .s                     following order:

DORITA-2 Case #1 Radiation fields at the Site boundary for the following design-basis accidents: (a) Loss of coolant accident (LOCA) (drywell leakage), (b) Loss of coolant accident (LOCA) (ESF Component leakage), (c) Main Steam Line Break accident (MSLB), (d) Control Rod Drop Accident (CRDA), and (e) Refueling accident (RA) Case #2 Similar to Case #1, for the Low Population Zone Case #3 Similar to Case #1, for a receptor located at the CR outside air intake (on top of the old administration building) [ Note: This location (~D was conservatively assumed to apply to outdoor

         \-)                                                                                    receptors at ground elevation, in the general vicinity of the old administration building.]

1 Case #4 Gamma spectra associated with post-LOCA airborne radioactivity within the Reactor Building (for use l with QAD-CGGP and MATILDA to compute the direct shine radiation levels from the refueling level and from El. 272' of the RB). Note: DORITA-2 Case #4 employs a SGTS filter efficiency of 99% for release from the RB to the outside atmosphere. Although the efficiency was lowered from 99% to 90% in evaluating the radiological consequences, this run was not revised since the change in filter efficiency does not affect the post-LOCA Reactor Building spectrum. QAD-CGGP & MATILDA Case #1 Direct shine radiation levels from post-LOCA l radioactivity accumulating in the RB Refueling Level, based on the gamma spectra defined under

DORITA-2 Run Case #4 (16 receptors all around the p RB)

U s l l L_.___ ____ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ . . _ _ _ . _ _ _ . _

Case #2- Direct shine radiation levels from post-LOCA u- radioactivity in El. 272' of the RB based on the gamma spectra defined under DORITA-2 Run Case #1) (8 receptors along the source centerline, at about 4 ft above grade). Note: The title line in the QAD-CGGP runs erroneously identify the plant as IP3. This does not impact the results.

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NYPA/CRE CALCULATION No. JAF-CALC-RAD-00048, Rev. 1 THIS IS THE LAST PAGE OF THIS CALCULATION E o O o

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ML20247G116

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