RS-05-020, Quad, Units 1 and 2, Additional Information Supporting the Request for License Amendment Related to Application of Alternative Source Term

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Quad, Units 1 and 2, Additional Information Supporting the Request for License Amendment Related to Application of Alternative Source Term
ML050630530
Person / Time
Site: Dresden, Quad Cities  
(DPR-019, DPR-025, DPR-029, DPR-030)
Issue date: 03/03/2005
From: Simpson P
Exelon Generation Co, Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-05-020, TAC MB6530, TAC MB6531, TAC MB6532, TAC MB6533
Download: ML050630530 (49)
v • d • e


Text

{{#Wiki_filter:RS-05-020 March 3, 2005 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 Subject : Additional Information Supporting the Request for License Amendment Related to Application of Alternative Source Term

References:

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S. NRC, "Request for License Amendments Related to Application of Alternative Source Term," dated October 10, 2002

2. Letter from L. W. Rossbach (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC), "Quad Cities and Dresden Nuclear Power Stations -

Request for Additional Information Regarding Alternative Source Term Amendment Request JAC Nos. M136530, M136531, MB6532, and M136533b" dated February 28, 2005

3. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S. NRC, "Additional Information Supporting the Request for License Amendment Related to Application of Alternative Source Term," dated March 28, 2003 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to the facility operating licenses for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed changes support application of an aftemative source term (AST) methodology.

In Reference 2, the NRC requested additional information related to releases from a postulated main steam line break accident. The attachments to this letter provide the requested information.

March 3, 2005 U. S. Nuclear Regulatory Commission Page 2 Additionally, in Reference 3, EGC requested a revision to the implementation period, to support the implementation plan for the proposed amendments, such that the amendments shall be implemented prior to the start of refueling outage D2R1 8 and Q2R1 7 for DNPS and QCNPS, respectively. Since these refueling outages have already occurred, EGC requests to revise the implementation period as follows. Once approved, the amendments shall be implemented within 180 days for both DNPS and QCNPS. EGC has reviewed the information supporting a finding of no significant hazards consideration that was previously provided to the NRC in Attachment C of Reference 1. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. If you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803. I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd day of March 2005. Respectfully, Patrick R. Simpson Manager - Licensing Attachments: 1. Response to Request for Additional Information

2. Calculation DRE02-0035, Revision 2, "Reanalysis of Main Steam Line Break (MSLB)

Accident Using Alternative Source Terms"

3. Calculation QDC-0000-N-1266, Revision 2, "Reanalysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms" cc :

Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

NRC Request Provide the calculation or calculations associated with the expansion of the released reactor coolant and steam associated with a IVISLB accident and the modeling of the resultant hemisphere (plume) past the Quad Cities and Dresden control room intakes. Identify the pressure and temperature conditions of the steam at the time of the assumed break.

Response

ATTACHMENT 1 Response to Request for Additional Information Attachments 2 and 3 provide the requested calculations for Dresden Nuclear Power Station and Quad Cities Nuclear Power Station, respectively. A reactor vessel steam outlet temperature of 5480 F was used in these calculations to validate the flashing fraction. In a saturated system, this corresponds to a pressure of 1028.5 psia. Normal reactor pressure is approximately 1020 psia, which is the pressure used in the accident analysis described in Updated Final Safety Analysis Report section 15.6.4 for the main steam line tweak accident. The use of 5480F is conservative since it would lead to a larger flashing fraction than actual conditions.

CHMENT2 Calculation DRE02-0035, Revision 2, "Reanalysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms"

CC-AA-309 - ATTACHMENT I - Design Analysis Approval Page I of 2 Pop: I of 14 (Printed : 3/11/2003 8:07 AM) E .-For _CC -AA-309-1 vi J for use with CC-AA-309 Revision I and above DESIGN ANALYSIS NO. : DRE02-0035 PAGE NO. 1 Major REV Number: 002 Minor Rev Number: 000 BRAIDWOOD STATION BYRON STATION CLINTON STATION [X] DRESDEN STATION LASALLE CO. STATION QUAD CITIES STATION Unit: 0 1 [X] 2 [X ] 3 DESCRIPTION CODE:(cois) N01, R01, R02 DISCIPLINE CODE: (coil} IN SYSTEM CODE: (CO11) NA TITLE: Reanalysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms [X] Safety Related I Augmented Quality Non-Safety Related ATTRIBUTES (C(l16) TYPE VALUE TYPE VALUE Elevation Soft COMPONENT EPN : {Coi4 Panel) EPN TYPE DOCUMENT NUMBERS : (CO12 Panel) (Design Analyses Redhervenceves) Type/Sub Document Number Num7!!! Input (YIN) Report/Eng GE-NE-A22-00103-64-01, RO y Calc/Eng DRE02-0036, RO y Calc/Eng DRE97-0071, R1 y Calc/Eng DRE97-0150, R2 y

1. j l 3 :5-4 7 c Z-REMARKS : This 10*0 Calculation corrects an error in Dose Conversion Factor utilization in the Attachment A spreadsheet as used for Rev. 0 and Rev. I of this calculation. The impact of the correction is minor, as all resulting doses remain a small fraction of the applicable regulatory limits.

CC-AA-309 - ATTACHMENT I - Design Analysis approval Page 2 of 2 E-Form CC-AA-309-1 v l. Page 2 of 14

02/26/03 5 :28 PM3

-309 Revision I and above. DESIGN ANALYSIS NO. DRE02-0035 REV : 002 PAGE NO. 2 Revision Summary (including EC's incorporated) : Attachment A and Section 7.0 results completely replaced, plus minor editing, clarifications and updates "ce.g_ UFSAR revision) incorporated. Electronic Calculation Data Files : (Program Name, Version, File Name extensionisizet"date,,bour.,inin) 35~20 z Design impact review completed? Yes I X I N/A, Per E # (If yes, attach impact review sheet) zz~ A10113 Prepared by : 'Paul,,.Reichert I Pot Sign Reviewed by : Harold Rothstein Date 1/11 A Print Sign Date Method of Review : [XI Detailed [ j Alternate Test This Design Analysis supersedes : Rev. 0 and Rev. I in its entirety. Stippicnient ;d Review Required? N es 1XI No R'CN ieN% F C.1111 Special Review I cani : (N /.~ for kddifion.rl ReNie~k) ReNie~Ncrs :

3) 11 W WII Approved by :

Harold Rothstein Print Sign Date External Desi--on Analvsis.Review (Attachment 3 Att ed) Reviewed by : 6-t?A IL D fl, 1-tr ; / A f Print Sign Date Approved by. Print Sign -n ate -j - V " /M Awl" SO ~J-Do aNv ASSUMIPTIONS / E.NGLYEERING JUDGEMENTS require later Verification? I Yes JXJ No Tracked BL Al'#. EC# et0

Owner's Acceptance Rev DESIGN

2. Are assumptions compatible with the way the plant is operated and with the ing basis?

3. Do the design 4. Are design inputs correct and reasonable? 5. Are design inputs compatible with the way the plant is operated and with the licensins basis? have sufficient ra have sufficient rationale? Checklist for SIS NO. DRE02-0035 REV: 2 fi. Are Engineering Judgments clearly documented and justified? 7. Are Engineering Judgments compatible with the way the plant is operated and with the licensing basis? 8. Do the results and conclusions satis analysis? 10. Doe documentation? 12. Are there any unverified assumptions? analysis include the applicable design b 13. Do all unverified assumptions have a tracking and closure place? se and objective of the design 9. Are the results and conclusions compatible with the way the plant is operated and with the licensing basis? 11. Have any limitations on the use of the results been identified and transmitted to the appropriate organizations? nal Design Analysis Yes No NIA 0 l3 [3 P6

1.0 PURPOSE/OBJECTIVE....................................................................... .................................... -- 4 2.0 METHODOLOGY AND ACCEPTANCE CRITERIA..................... -- .. 4 2,1 General Description........................................................ ........................................................ 4 12 Source 'term Model.................................................................................................................... 4 2.3 14 2.4.1 2A2 15 Dose Model 2.5.1 EAB and LPZ....... 2.5.2 Control 2.6 Acceptance Criteria.............. .3 .0 ASSUMPTIONS....................... 3.1 LCULATION, TABLE OF CONTENTS del................................................ ................................................................. - 5 .I ........... 5 EAB and LPZ............... .Control ]Zoom.......... Activity Release and TranspoControl Room . Model............ .Site Boundary Model............. 4.0 DESIGN 4.1 Mass Release 4.2 Iodine Distribution................ .4 .3 Control Room Data............... .4 .4 EAB and LPZ Esta....................... 5.0 REFERENCES............................................................................................................................... 10 6.0 CALCULATIONS..................... -............................................... I I 6.1 Cloud Volumes, Masses, and Control Room Intake Transit Times.... ............................... 11 6.2 Dispersion for Offsite Dose Assessment....................................... -.-.................................. - 12 6.3 Release Isotopics and Quantification................ - .......................... - 12 6.4 Dose Assessment...................................................................................................................... 13 TO

SUMMARY

AND CONCLUSIONS........................................................................................... 14 A. Spreadsheet performing MSLB Dose Assessment [pages Al-A6] Computer Disclosure Sheet [pages BI-BIJ ............................ --...................................... -- 5 ..... -............................................... 5 ........................................... 6 ................................................................... 6 ......................................................................... 6 ............. 6 .. --......................................... I ........................................... I .......... 8 rt Models..................................................................................... 8 ................................................................................................ -.. " 8 .................................................................................................... 8 9 9 9 9 9

[PAIC tuaAMON NO....1PRE0 -0035 REV. INO. 2 -j--PAGE N0.-4 of 14 1.0 PURPOSEIOBJECTIVE The purpose of this calculation is to determine the Control Room (CA), Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) doses following a Main Steam Line Break (MSLB) Accident based on the assumptions on the break and resulting radiological releases to the Turbine Building as contained in existing calculations DREOO-0071 Rev. 1 and DRE97-0150 Rev. 2 [References 5.1 and 5.2], and the additional assumptions for use of Alternative Source Terms (AST) contained in Appendix D of Regulatory Guide (R. G.) 1.183 [Reference 5.12]. However, inhalation Committed Effective Dose Equivalent (CEDE) Dose Conversion Factors (DCFs) from Federal Guidance Report No. 11 [Ref. 5.4) are used for calculation of normalized Iodine-131 Dose Equivalent activity. As per UFSAR Section 15.6.4, this event involves the postulation that the largest steam line instantaneously and circumferentially breaks outside the primary containment at a location downstream of the outermost isolation valve, with this event representing the envelope evaluation of steam line failures outside primary containment. Closure of the Main Steam Isolation Valves (MSIVs) terminates the mass loss when the full closure is reached. No operator actions are assumed to be taken during the accident, so the normal air intake into the Room continues unfiltered during the duration of the event. The mass of coolant released during the MSLB was obtained from reference 5.1, which bases analysis on 5.5-second closure of main steam isolation valve. 2.0 METHODOLOGY AND ACCEPTANCE CRITERIA 2.1 General Description The radiological consequences resulting from a design basis MSLB accident to a person at the EAB ; to a person at the LPZ ; and to an operator in the Control Room following an MSLB accident were performed using a Microsoft EXCEL spreadsheet, provided as Attachment A. 2.2 Source Term Model No fuel damage is expected to result from a MSLB. Therefore, the activity available for release from the break is that present in the reactor coolant and steam lines prior to the break, with two cases analyzed. Case 1 is for continued full power operation with a maximum equilibrium coolant concentration of 0.2 uCVgm dose equivalent 1-131. Case 2 is for a maximum coolant concentration of 4.0 uCVgm dose equivalent I-131, based on a pre-accident iodine spike caused bypower changes. In determining 1-131 equivalence, inhalation CEDE DCFs; from Ref. 5.4are used. This accident source term basis meets the guidance in A.G. 1.183 for analysis of this event.

CAL CAKULATION DRE02-0035 I RED, NQ. 2 PAGE N0.5-of N I 2.3 Release Model The release model is identical to that historically used. The previously determined mass of reactor coolant release and mass of steam release, before the break is isolated by IVISIV e, are used. Reactor coolant radioactivity is based on the above reactor coolant concentrations. Reactor steam radioactivity is based on a steam to coolant iodine concentration ratio (carry-over) of 2.4 Dispersion Model d to be instantaneous and no credit is taken for dilution in turbine building X/O determinations are handled differently, but conservatively in both cases. 2.4.1 EAB and LPZ EAB and LPZ X/Q's are determined usi 0033 Q OY U where 2.4.2 Control Room the original methodology in R.G. 1.5. Specifically: Y = horizontal standard deviation of the plume (meters) u = wind velocity (meters/second) Horizontal standard deviations are taken from the PAVAN outputs for the EAB and LPZ include in Calculation DRE02-0036. Per Regulatory Guide 1.5, F stability and a 1 meter/sec ; wind speed is used. For control room dose calculations, the plume was modeled as a hemispherical volume, the dimensions of which are determined based on the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam. elease is conservatively assumed to effectively occur at the Control Room intake ation and, again conservatively, no credit is taken for plume buoyancy. A conservative translation time of the plume over the intake is assumed. of the cloud is based on the total mass of water released from the break, not just the flashes to steam. This assumption is conservative because it considers the maximum release of fission products.

Dose models for both onsite and offsite are simplified and meet R.G. 1.183 requirements. Dose conversion factors are based on Federal Guidance Reports 11 and 12. 2.5.1 EAB and LPZ Doses at the EAB and LPZ for the MSL V -- (ux*n') *Breathing Ram (m'/sec) *Inhalation DCF(remICi inhaled) Dow

finally, 2.5.2 Control Room CR operator doses are determined somewhat differently, because steam cloud concentrations are used, rather that 7

¬ /O time a curie release rate. No CR filter credit is taken and, therefore, for inhalation, a dose for a location outside of the CR can be and is used. For cloud submersion, a geometry factor is used to credit the reduced plume size seen in the control room. This is a conservative implementation of RG 1.183 guidance. The formulas used are : and Dose and finally, Dose = Release (Curl Dose DoseTEDE (rem (rem) = Plume Concentration (Ci/m')

  • Transit Duration (sec) 2.6 Acceptance Criteria

= Release (Curies)* -! (seCIM 3 )

  • Submersion DCF (rem,,, _ m1 Ci _ sec}

(rem) = Plume Concentration (Ci/ are based on the following formulas : (rem) + Dose Dose.,(rem} Rate {m3/sec}

  • Inhalation DCF (rem

/Ci inhaled) ranch Duration (sec)

  • Submersion DCF (rem,,,, - m'/ Ci - sec) criteria are per 1 OCFR50.67 and R.G. 1.183 guidance.

Table 1 lists the regulatory limits for accidental dose to 1) a control room operator, 2) a person at the EAB, and 3) a person at the LPZ boundary.

I CALCULATIONNO. DRE02-0035 I REV, NO. 2 1 PAGE NO. 7 of 14 Table 1. Regulatory Dose Limits (Rem TEDE) 1-131 Dose Equivalent CR (30 days) EAB (2 hours) LPZ (30 davs -Normal Equilibrium 2.5 2.5 Iodine Spike 5 25 25

I CALCULATION NO. DRE02-0035 I RIM MD. 2 - I PAGE N{). 8 of 14 3.0 ASSUMPTIONS 3.1 Activity Release and Transport Models Iodine activity distribution in the coolant was taken from the Quad Cities UFSAR, Section 15.6.4.5 which provides more detail than the Dresden UFSAR. [5.13, 5.7] The two facilities are sister units of the same basic design and operating conditions, as such the iodine activity distribution would be similar. se from the break to the environment is assumed instantaneous. No holdup in ine Building or dilution by mixing with Turbine Building air volume is credited. steam cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam. of the cloud is based on the total mass of water released from the break, portion that flashes to steam. This assumption is conservative because it he maximum release of fission products. The fraction of liquid water contained in steam, which carries activity into the cloud, was assumed to be 2%. Flashing fraction of liquid water released was assumed as 40%, as derived in section 6.1 below. However, all activity in the water is assumed to be released. For offsite dose calculations the release is treated as a point source with ground level dispersion per R.G. 1.145. Buoyancy effect of the cloud was conservative) pored. the control room dose calcula the plume was modeled as a hemispherical volume. This is consistent with the assumption of no Turbine Building credit. It is also reasonable for the more likely release paths through 8 large blowout panels situated around the Turbine Building Main Floor. dispersion of the activity of the plume was conservatively ignored. The cloud was assumed to be carried away by a wind of speed I m/s. Credit is not taken for decay. 3.2 Control Room Model 3.3 Site Boundary Model This model is as discussed in Subsection 2.5.2 above. No credit was taken for the operation of the CREFS during the MSLB. Inhalation doses are determined based on concentrations at the intake, and exposures for the duration of plume traverse. External exposure doses are determined based on concentrations at the intake, exposures for the duration of plume traverse, and a geometry factor credit based on the Control Room proper volume of 64,000 cubic feet.

I CALCULATION NO. DRE02-0035 I REV. 191 2 1 PAGE NO. 9 of 11 4.0 DESIGN INP 4.1 glass Release Data " The mass steam released is 20,000 lb. [5.21 " The mass liquid water released is 10,125 lb. [5.2] 41 thomdWiMne D1s1rh5cudVcwn The Dresden UFSAR provides distribution of fission products only in term of 1-131, 1-133, other halogens, and other fission products (section 15.6). [5.7] Therefore, the distribution of fission products in the coolant was obtained from the UFSAR for Quad Cities, which is a plant similar in design to Dresden. The relative mix of iodine isotopes in the reactor coolant at the onset of the ident, based on Quad Cities UFSAR is given below (section 15.6.4). [5.13) This analysis ores the contribution from other halogens and noble gasses, which have negligible dose consequence. Iodine Isotope Activity (pCVcc) 4131 0,067 1-132 0.38 14133 0.40 1-134 1-135 4.3 Control Room Data Control Room proper volume = 64,000 ft'. [5. 11 No Emergency Filtration Credit taken. 4.4 E48 and LPZ Data " EAB Distance from Release, m 800 (Tech. Spec., section 5) [5.151 " LPZ Distance from Release, m 8,000 (Tech. Spec., section 5) [5.15]

CALCULATION NO. DRE02-0035 REV. NO. 2 PAGE NO. 10 of 14

5.0 REFERENCES

5.1 Calculation DRE97-0071, "Impact of Extended Power Uprate on Site Boundary and Control room Doses for LOCA and Non-LOCA Events", Revision 1. 5.2 Calculation DRE97-0150, "Control Room Habitability Following a Main Steam Line Break", Revision 2. 5.3 Not Used 5.4 Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, Ingestion", 1988. 5.5 Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993. 5.6 Not Used. 5.7 Dresden Nuclear Power Station UFSAR Rev. 5. 5.8 Not Used 5.9 Not Used. 5.10 NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980. 5.11 Regulatory Guides 1.5, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accidents for Boiling Water Reactors," 3/10/71. 5.12 Regulatory Guide 1. 183, "Alternative Radiological Source Terms For Evaluating Design Bats Accidents At Nuclear Power Reactors", July 2000 5.13 Quad Cities Nuclear Power Station UFSAR Rev. 7, Section 15.6.4. 5.14 Code of Federal Regulations: 10 CFR Part 50.67 5.15 Dresden Technical Specifications. 5.16 Calculation DRE02-0036, "Reanalysis of Fuel Handling Accident (FH Alternative Source Terms", Revision 0 ing

I CALCULATION NO. ago-0035 I REV. POD, 2 1 PAGE No. 11 of 14 1 6.0 CALCULATIONS No or minimal fuel damage is expected for the limiting MSLB. As discussed in section 2, two iodine concentrations will be used (0.2,uCi/g and 4.0 IjCVg) when determining the consequences of the main steam line break. All of the radioactivity in the released coolant is assumed to be released to the atmosphere instantaneously as a ground-level release. No credit is taken for plateout, holdup, or dilution within facility buildings. The spreadsheets in Attachment A perform this analysis using data and formulations discussed above. The following summarizes parameters and their treatment in the spreadsheet. 6.1 Cloud Volumes, Masses, and Control Room Intake Transit Times The cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashes to steam. The flashing fraction (FF) is derived as follows : FF x (steam enthalpy at 212 F) + (OFF) x (liquid enthalpy at 212 F) (liquid enthalpy at temperature of steam at reactor vessel outlet) A 548 F vessel outlet temperature is used, with liquid enthalpy of 546.9 BTUAb. At 212 F, a steam enthalpy of 11515 ETUAb and a liquid enthalpy of 180.17 BTU/lb are used (these enthalpies are taken from the ASME Steam Tables). Substituting, FF = (546.9 - 180.17) 1 [(1150.5 -180.17}] =.378 conservatism, a value of.40 or 40% is used. As stated in Section 3.1, the cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam. Mass of water carrying activity into the cloud was calculated as the sum of the fraction of water in the steam and the liquid blowdown. The mass steam released = 20,000 lb The mass liquid water released =10,125 lb Flashing fraction for calculating cloud volume = 4 The mass water contained in steam released = (20,000 lb) = 400 lb The mass of water carrying activity into the cloud = 400 + 10,125 lb =10,525 lb = (10,525 lb)(453.59 g/lb)

= 42706 g The mass of steam in the cloud (20,000 - 400) + 40%*10,125 lb 23,650 lb The release is assumed to be a hemisphere with a uniform concentration. The cloud dimensions (based on to 10,525 lb of steam at 14.7 psi and 212 OF, v', = 26.799 ft3 /lb) were calculated as follows

I CALCULATION NO. DRE02403Y --- IT RED, NO. 2 1 PALE INTO. 12 of 141 The volume of a hemisphere is rc 40.9 am. The period of time required for the cloud to pass over the control room intake, assuming a wind speed of I n-ds is 40.9 s (=(40.9 m)/(l m1s)). Therefore, at a wind speed of 1 m/s, the base of the hemispherical cloud will pass over the control room intake in 40.9 seconds. Volume = (23,650 lb)(26.799 fellb) = 633,796 ft' = (6SI796 ft3)/(35.3 ft3/M3)

= 17,955 6.2 Dispersion for Ofisite Dose Assessment As discussed in Section 2.4.1 the following formulation was used for Offsite Dose X/O assessment, with F Pasquill Stability and a 1 rn/sec wind speed.

X 0.0133 Q a yu where = horizontal standard deviation of the plume (meters) u = wind velocity (meterstsecond) Iculated in the PAVAN run in Reference (5.16], at the 800 meter EAB distance ay is 30.2, and at the 8000 meter LPZ distance ay is 242. The resulting EAB and LPZ X/Os are 4.40E-4 and 5.50E-05 SedM3, respectively. &3 ANN= 1soQpk;s and Quantification Thus, the diameter of the hemispherical cloud is The iodine isotopic distribution given in Section 4.2 is used. The concentrations of this mix are adjusted to 1-131 equivalence, using the inhalation Committed Effective Dose Equivalent (CEDE) Dose Conversion Factors (DCFs) from Federal Guidance Report No. 11 [Ref. 5.4]. This is a more conservative set of DCF assumptions for Control Room and off-site dose calculation than the prior use of ICRP 2 DCFs. It is also more conservative for these calculations than use of R. G. 1.109 or Federal Guidance Report No. 12 [Ref. 5.5) DCFs. This 1-131 equivalent mix is adjusted to the activity yielding the two design basis MSLB accident reactor coolant activities of 0.2 gCVcc and 4.0 gCi/cc. The released activities is these concentrations times the 4.77E+06 grams of activity released, with the assumption that TS activities are based on laboratory temperature and pressure conditions.

I CALCULATION NO. ]DRE02-4010M I REV 141 2 1 PAGE N0.13 of 14 1 6.4 Dose Assessment Doses at the EAB and LPZ distances, and in the Control Room are calculated in Attachment A using the formulas in Section 2.5. Concentrations at the receptor locations are that in the steam plume for the Control Room or based on the release times the applicable X/O for the EAB and LPZ Doses are calculated for inhalation (rem CEDE) and plume submersion (rem EDE) and totaled to yield rem TEDE. The breathing rate of 3.47E-04 m3/sec is per RG 1.183 guidance without the round-off. The resulting calculated doses are in the spreadsheet and in the Summary and Conclusions Section below.

I CALCULATION NO. DRE02-0035 I REV. W 2 7.0

SUMMARY

AND CONCLUSIONS Accidental doses from a design basis MSLB were calculated for the control room operator, a person at EAB, and a person at LPZ. The results are summarized in the Table below. The doses at the Control Room, EAB, and LPZ resulting from a postulated design basis MSLB do not exceed, and are a small fraction of the regulatory limits. Location Case 1 Case 2 (normal equilibrium (iodine spike limit of 0.2 gCi) limit of 4.0 IACI) Dose (rem TIEDE) Dose (mum TEDE) LIMITS CR 5.0; EAB&LPZ 2.5 CR 5.0 ; EAB&LPZ 25 EAB 6.36E-03 1.27E-01 LPZ 7,95E-04 1.59E-02 7 CR 2.51E-02 5.02EW1

Calculation DRE02-0035, Rev. 2 Attachment A Page Al of A6 A SLB Dose Spreadsheet j Case t iReactor Coda it maximum value (DE 1-131 of 0.2 uCi/cc) permitted for continued full power operation 31- ... i cloud {cubic meter value permitted (DE 1-1 of 4.0 .77E+06 ~ ~ Mass s of water in reactor coolant Reactor Coolant at maximum release (grams) I uCi/cc) corresponding to an assumed pre-accident spike 1-1.111 I+/-mactorcoolant density !smeai6r JqTamstcc) j I,, - - 11.11, 40 9 seconds for cloud to pass over CR intake far wind _speed of I m/se 1 64000 Vol m6 of Cop q~ Roam proper {cubic feet} ' I - I I I ~ ' ' I ~ I . I I r I I I r I r I Activity 1 Release. Release j , :, ~ ~ ~ ' ~ 9 9 r ' r r r 6 v 9 0 0 ~ 9 r r r 9 ~ 9 r 9 r 4 9 ~ i,. Release I r r

Normalized 1

Case 1 1 Case 2 1 Release 1 Release Cloud I Cloud OR 11 17911.1-!?~~Flrlj~-~Npr~ Normalized Actual ( not I Distribution ; DCF i Activit 1 Activitv Activity normalized Releases v uCilcc IReMCEOE/Cil CVM' Cilmi I-131' 0.067 3.29E+04 6.70E-02 8.23E-02. , 1.65E 00 3.93 01 llr~7.-1§5E+Oorrrl+.2*19 05.. 3 E0 24 - 2.4$E 03' .75E+00 2.34E +00 4.691: 01 ; 1.31 E -O4 ! E-031 4135 0.49 1 1,23E+03 i 1 83E-02 i 2025E-02 4*50E-0 1 217'~~ 1- "spiked" I non-spIKeU 1 F-C I 1 1111111111111111111 j ^- A a-n 4 flnn^ 'I 2/26/2003

Calculation DRE02-0035, Rev. 2 Attachment A Page A2 of A6 27 28 29 130 31 32 33 35 38 40 41 42 44 47 48 49 50 51 52 54 i 15 i ... _1.. ~, _1.11 11.1-1-11 111-1 1-1-- Curies Released Case 1 Dose (rem C~P~) Case 2 Dose (rem C~D~) T MQ: to ine tnviron ent I I J t

ationp, L I Z Isotope

&AFT-Case 2 1 RemCEDE/Ci 5E ~,100 2~23E+00 1 4.4 tog--li ~,71f7 lwl 33, WAS AbOE+01 5,85E+03 1 IME-02 2.09E-03 1 2.62E-04 I-- I I 1 1 34 621t~dl 2_ 1 2E IF k22&051~~ .7 -.1.1.111--.1-1.11 .11.1 1 11112~ --2 11, ~2~- 1 E~ :2~ 1 1 . 1 .. 1.1-1-1.1.1.1- _117T i jotqpe 6

2.

Inhalation Effective Dose C L o L n " v r, e , rs L i ' o ' n 'Fa~,t?.'JReM_EPE./Curie) from Federal Guidance 96id L e-r j Effective Dose Conversion Factor (rem- /Cude-seconq)!cr ~_~Y!?Tersion from FGR 12 per latory- ~qerj - _o (w*~i TUi~~ rL 3,59E-0 ....... j 'L. 3 1 02 E + 0 PAVAN runs for FHA in 616i P L A V A N r u n s f o r F H A in R e f 5 1 6 1 00E+09 1,1,)~!q~-~pee~ 4 4 0 E 0 4 1 X/Q (seconds/rT~~ ~9~n~y 2 based on RG 1.5 methodology i bne~ cnLLLL 6 j '6 5.50E-05 at Low Population Zone. ~jj;~j 1 . 1.11-1.11.. . 11 r L L 1 L... . r.. Curies Released to the Environment 1 Kem~p~-T-x Case I Dose (rem ERE~ i 1 t~xerr!9'1 I Cape ? P 9 r e T /r.'1A____1% 11 it IT ~Case ~ ~ ~ - d -i-s I econd 1 n 4 F\\ - I-L

rrr, I 'rrrr~Apr

-L]"'L.LL. EA' L, j FQ4 3.93E-01 1 7.85E+00 _6Ypjf - y_ j ~66 I 1 .16E-05 1 1.45E-06 1 1 4.33E-05 191&05 1.51E-03 12 0~ EM 6. E 6.5 ~~-04 _04 5j ME= MAP T ------ 4E-1 7.4 3EW4 EA~ I LPZ 3.95E- ? I ~-93g-03 134ESU" --it§E&3 124E-04 217E_d~j 1- ~-j - 9 E 0 1 1.55E-04 -~~ 1 ME-02 1" .. 6 ~i~ _fil - ~ I 5E-03

Calc DRE02-0035, Rev, 2 Attachment A Case 1 : 0 Page A3 of A6 Case 2 : Reactor Coolant at maximum value perm 5 Volume r 4000 Mass ofv` I . reactor c sec 64666 Volume 0 - 1 111.1.... i I I ICaseI 1-131 DE Normalized ed Normalized (not nor Activity uCilcc uCi1cc u i 1lCi 1 11 -1.1-1 I P0 =DW6.2jb lb i=E14*20 EwsAs4t.6obooi/sA$5 C15 B151C$14 -D15 0.21D$19 -1T15 20 --{C$14IC15} F15 $A$4 0 0000011$A$5 0 =13161C$14 0 =* .2b$19 E16 20 =CWB17/C$14 rozb I ti7*2O-

WMAM

=C18 B181G$14 =DI8*0.2/D$19 '~E18 20 .-(C:$141C1.8} E18 $A$4 0.0000011$A$5 "spiked" "I." ..,.",.",.,.""..",""-"'I WS I to i Environment I DCF 1 I i (inhalation) i r, --- 11 1 D it, ; 1-0 1 chn isotope Case I t....... !=114*$E30*$A$50*$A$6 !=C30*$E30*$A$50*$A$55 ~- 6 r 1 0 r j 9 1 ~ 1 4 r 9 9, I r 6 1 1 J 1 6 r J 9 9 r ]' ' =' j' ~1 1 1 1 1 6 . r 9. 94 9 r ], =9 H I 5 6 j9 6 .. r. 9 ..r -9 r.r - r J999949 r 9. r. 4.. 381 1 =1 1 5 * $E3 1 *$A$50*$A$6 1 =C3 1 * $E3 I *$A$50*$A$55 ~ I ~ I I I I. ~ ~ ~ I. I r I I I.. ' j I ' ' ~ ~ I I r I r I r I r I I I ~ ' I =G16 - - 1, Q 5846 1 r 1 6 1 9 9 ~ 0 r 0 Ir " 0 ~ j j 6 ~ ~ 6 " j 6 6 6 j 0 r ~ j r r r 9 r. r 9. r 4 : : j I- ~.=jj j j I j 6 133 1 !=C32*$E32*$A Pq7$'§Qq QG17 W-117 Q 3~~$A$50*$A$6

  • $A$50*$A$55
  • $E

!=GIB 41-118 1 1230 =ljB34*$A$W$A$6 I=C34*$E34*$A$50*sA$55 otals =SUM(F30 : 2/26/2003 i Activity FqR 11 istr ubol RCF l°132 _ 0.38 381 0.4 5846 170 0.53 131 I-135 0.49. .1230

Calculation DRE02-0035, Rev. 2 Attachment A Page A4 of A6 to the Environment I ReMEDE"M3i 1 i (External) isot i '-,i.. - L' L Curies Released DCF Case j pose (rem EDE) r r " r r rr L L L L ub-Total {rem..TEDE} alation Effective Dc i I I dive Dose Convey i i 0.000347 > Breathing 1 1 y1j,110-Fit.. r r 30 (MI Wind 8pe, =0,0133/AW/A$54 IXtQ (secoj .636iA$53/A$54 XIQ 'r (secol pe =-13 .1. Case 1 ,=G30 I - 9 1 1 6 9 9 9 9 r 9 9 r - ~ 6 ~ - j 9 r. j j. r 6 J 9 9 9 9 9 r - ta e ~!,J C~ 6i- -second 1 7pfp 1OP6734 r - r 9 9 r r 9 1 r 9 9 r ... 9.. . 9.. 9. 9 1 1 1 r I ~ 1 R 1 ~ 1 1 4 -'W4 I Z~AZ5E51 zW$b j 6 I 1 :~t4 -1 Z~A ;~Zn) r 9 1 9 1 r. I I-132. -C31 =D31 $55 I-133. =C32 -D32 '010878 -116*$E43 $A$51 $A$6 =C43*$M*W55 list 1 KC93 -D33 IOA81 1=117*$E44*$A$51*$A$6 !=C44*$E44*$A$55 k135 ---Af9526 1=118*$E45*$A$51*$A$6 !=C45*$E45*$A$55 total 1 I I j=SUM(F41 :F45) =SUM(G41 :G45)

Calculation DRE02-0035, Rev, 2 Attachment A Page A5 of A6 continued full power operation d(DEI-!3jof4. 0 uCilcc} corresponding to ...C ase Case 2 1 1 Case 2 I ---l-I. "9y I Release 1 Release i Release Cloud 1 C ud i alized Releases) ____ Concentration Cancentratian - b6bb6IIW5-- &G1V$A31 44~1:~ 14~~, i i=10A (0$141015} *F15 *§A$4*d.bqbb6i Ut5 i 5i$ [=HIN$AS3 ------ ={C$141G16} F16 $A$4 0.0000011$A$5 =G161$A$3 =M16/$A$3 '(C$14/C17)*F17*$A$4".6666 V$AS5 !--q144 assumed pre-acc pike (Inhalation LPG EAB LPZ

  • D3W$E3o*tAs5O*$ASAbt&SE30*$A$50! ;,;56 tAt 1 =J 15* $E31 *$A$50*

$50*$A$'=D3VtE .3'*50*$A$6§ r W$A$6 I =D32*$E32*sAt5 17 A bnIB3*$A$501A$ 1_=O3*$t IAWIAW j-1X34*$E34*$ MUN(00 :134) 1~=( 3O:J34) j=SUM(K30:K34) I _?_ 2/26/2003

Calculation DRE02-0035, Rev. 2 Attachment A Page A6 of A6 3 (External) ... CR CAB 1-1111.11. ICase 2 Dose (rem EDEN

  • $E41 *$A$56

! -~*tt42IAW =D42-$E42~W 6 =C44*$E44*$A$56 Z~41,71E44*Wpj*$A$6 !=D44*$E44*$A$55 D44*$E44*$A$56 1-J18*$E45*$A$51 -t4 ]=b45*$ 65 !=b4t45 1456 >YWMI UMO4!--, >YMQ95+M9

TION NO. DRE02-0035, Attachment B NO. 2 Computer Disclosure Sheet Discipline Nuclear Client : : Exelon Corporation Date: 2/26/03 Project : Dresden Units 20 MSLB AST Jot) 140. Program(s) used Rev No. Rev Date Calculation Set No. : DRE02-0035, Rev. 2 Attachment A spreadsheet N/A N/A Status Prelim. X Final J Void WGI Prequalification Yes [X] No Run No.

== Description:== Analysis Description : Spreadsheet used to perform dose assessment for MSLB, as described in calculation. The attached computer output has been reviewed, the input data checked, And the results approved for release. Input criteria for this analysis were established. By : On: 2/26/03 Run by : P. Reichert Checked by: H. Rothstein Approved by : H, Rothstein Remarks: WGI Computer Software Control This spreadsheet is relatively straight-forward and was hand checked. Attachment includes the spreadsheet in both normal and formula display mode and so is completely documented.

ATTACHMENT 3 Calculation QDC-0000-N-1266, Revision 2, "Reanalysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms"

CC-AA-309 - ATTACHMENT 1-Design Analysis Approval Pale 1 of 2 Page 1 of 14 (Printed:03/11/03 8:25 AM) E-Form CC-AA-309-1 v1.1 for use with CC-AA-309 Revision land above. DESIGN ANALYSIS NO.: ODC-0000-N-1266 PAGE NO. 1 Major REV Number : 002 Minor Rev Number : 000 [ ] BRAIDWOOD STATION [ BYRON STATION [ ) CLINTON STATION [ ] DRESDEN STATION [ ] LASALLE CO. STATION [X] QUAD CITIES STATION Unit:[ ] 0 [X]l_[X]2 ] 3 DESCRIPTION CODE:(colg) NOl, R01, R02 DISCIPLINE CODE: (coil) N SYSTEM CODE: (COll) NA TITLE: Reanalysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms [X] Safety Related [ ] Augmented Quality [ ] Non-Safety Related ATTRIBUTES (CO16) TYPE ~_ VALUE / TYPE VALUE Elevation _ Software COMPONENT EPN: (COl4 Panel) EPN FPE T DOCUMENT NUMBERS: (CO12 Panel) (Design Analyses References) Type/Sub Document Number Input (YIN) Report/Eng GE-NE-A22-00103-64-01, RO Y Calc/Eng QDC-0000-N-1267, RO Y Calc/Eng QDC-0000-N-1020, RO Y Calc/Eng QDC-9400-M-0365, R2 Y REMARKS: This Rev.2 Calculation corrects an error in Dose Conversion Factor utilization in the Attachment A spreadsheet as used for Rev. 0 and Rev. 1 of this calculation. The impact of the correction is minor, as all resulting doses remain a small fraction of the applicable regulatory limits.

CC-AA-309 - ATTACHMENT 1 - Design Analysis Approval Paee 2 of 2 Page 2 of 14 (Printed: 02/26/03 5:37 PM) E-Form CC-AA-309-1 vl.l for use with CC-AA-309 Revision 1 and above. DESIGN ANALYSIS NO. QDC-0000-N-1266 REV: 002 PAGE NO. 2 Revision Summary (including EC's incorporated) : 331 Attachment A and Section 7.0 results completely replaced, plus minor editing, clarifications and updates (e.g., UFSAR revision incorporated. Electronic Calculation Data Files: (Program Name, Version, File Name extension/size/date/hour/min) Design impact review completed? [ ] Yes [X]N/ Per EC (If yes, attach impact review sheet) )o Prepared by : Paul Reichert / / -"" 1a Print Si Date Reviewed by : Harold Rothstein / Print Sign Date Method of Review: [X] Detailed [ ] Alternate [ ] Test This Design Analysis supersedes : Rev. 0 and Rev. 1 in its entirety. Supplemental Review Required? [ ] Yes [X] No [ ] Additional Review [ ] Special Review Team Additional Reviewer or Special Review Team Leader : / / Print Sign Date Special Review Team : (N/A for Additional Review) Reviewers : 1) / /

2)

/ / Print Sign Date Print Sign Date

3) l l
4)

I l Print Sign Date Print Sign Date Supplemental Review Results : Approved by : Harold Rothstein Print Sign Date External Design Analysis Review (Attachment 3 ~~~~ ed 7 Reviewed by : 6l'ZQC 1) , ( P HTf l ^l 2 T (,t t.4'zoo3 Print Sign Date Approved by : ~~~. ~'~ / h:~...~~ l 3 Print Sign Cp,,IClnl~'~'20Su- / ..~ or T5 ~- /C'.Nae-dhh h~ Gn`E:~ tS Cst~C. Date Nca.~~ y",B1/h5~ L / W'S s b c~~3 y+ a o r Do any ASSUMPTIONS/ENGINEERI JUDGEMENTS require later verification? [ ] Yes [X] No Tracked By : AT#, EC# etc.

Owner's Acceptance Review Checklist for External Design Analysis DESIGN ANALYSIS NO. ODC-0000-N-1266 REV : 2 1. Do assumptions have sufficient rationale? 2. Are assumptions compatible with the way the plant is operated and with the licensing basis? 3. Do the design inputs have sufficient rationale? 4. Are design inputs correct and reasonable? 5. Are design inputs compatible with the way the plant is operated and with the licensing basis?

6. Are Engineering Judgments clearly documented and justified?

7. Are Engineering Judgments compatible with the way the plant is operated and with the licensing basis? M 0 0 8. Do the results and conclusions satisfy the purpose and objective of the design 0 0 analysis?

9. Are the results and conclusions compatible with the way the plant is operated and with the licensing basis?
10. Does the design analysis include the applicable design basis documentation?

11. Have any limitations on the use of the results been identified and transmitted to the appropriate organizations? 12. Are there any unverified assumptions?

13. Do all unverified assumptions have a tracking and closure place?

EXELON REVIEWER: 6eg4u) Yes No NIA 0 0 99 0 0 0 Jl 0 0 (?9 0 0 0 p

CALCULATION NO. QDC-0000-N-1266 - (REV . NO. 2 (PAGE NO. 3 of 14 CALCULATION TABLE OF CONTENTS 1.0 PURPOSE/OBJECTIVE.................................................................................................................. 4 2.0 METHODOLOGY AND ACCEPTANCE CRITERIA................................................................... 4 2.1 General Description.......................................................................................................................... 4 2.2 Source Term Model .......................................................................................................................... 4 2.3 Release Model.................................................................................................................................. 4 2.4 Dispersion Model.............................................................................................................................. 5 2.4.1 EAB and LPZ ........................................................................................................................... 5 2.4.2 Control Room........................................................................................................................... 5 2.5 Dose Model....................................................................................................................................... 6 2.5.1 EAB and LPZ........................................................................................................................... 6 2.5.2 2.6 3.0 3.1 3.2 3.3 4.0 4.1 4.2 4.3 4.4 5.0 6.0 6.1 6.2 6.3 6.4

7.0 Attachments

Control Room........................................................................................................................... 6 Acceptance Criteria.......................................................................................................................... 6 ASSUMPTIONS.............................................................................................................................. 8 Activity Release and Transport Models............................................................................................ 8 Control Room Model ........................................................................................................................ 8 Site Boundary Model........................................................................................................................ 8 DESIGN INPUT............................................................................................................................... 9 Mass Release Data............................................................................................................................ 9 Iodine Distribution............................................................................................................................ 9 Control Room Data........................................................................................................................... 9 EAB and LPZ Data........................................................................................................................... 9 REFERENCES............................................................................................................................... 10 CALCULATIONS.......................................................................................................................... 11 Cloud Volumes, Masses, and Control Room Intake Transit Times............................................... 11 Dispersion for Offsite Dose Assessment........................................................................................12 Release Isotopics and Quantification..............................................................................................12 Dose Assessment............................................................................................................................ 13

SUMMARY

AND CONCLUSIONS ............................................................................................. 14 A. Spreadsheet performing MSLB Dose Assessment [pages A1-A6] B. Computer Disclosure Sheet [pages B 1-B 1

CALCULATION NO. QDC-0000-N-1266 ( REV. NO. 2 PAGE NO. 4 of 14 1.0 PURPOSE/OBJECTIVE The purpose of this calculation is to determine the Control Room (CR), Exclusion Area Boun-dary (EAB), and Low Population Zone (LPZ) doses following a Main Steam Line Break (MSLB) Accident based on the assumptions on the break and resulting radiological releases to the Turbine Building as contained in existing calculations QDC-0000-N-1020, Rev. 0 and QDC-9400-M-0365, Rev. 2 [References 5.1 and 5.2], and the additional assumptions for use of Alternative Source Terms (AST) contained in Appendix D of Regulatory Guide (R. G.) 1.183 [Ref. 5.12]. Inhalation Committed Effective Dose Equivalent (CEDE) Dose Conversion Factors (DCFs) from Federal Guidance Report No. 11 [Ref. 5.4] are used for calculation of normalized Iodine-131 Dose Equivalent activity. As per UFSAR Section 15.6.4, this event involves the postulation that the largest steam line instantaneously and circumferentially breaks outside the primary containment at a location downstream of the outermost isolation valve, with this event representing the envelope evaluation of steam line failures outside primary containment. Closure of the Main Steam Isolation Valves (MSIVs) terminates the mass loss when the full closure is reached. No operator actions are assumed to be taken during the accident, so the normal air intake into the Control Room continues unfiltered during the duration of the event. The mass of coolant released during the MSLB was obtained from reference 5.1, which bases analysis on 5.5-second closure of main steam isolation valve. 2.0 METHODOLOGY AND ACCEPTANCE CRITERIA 2.1 General Description The radiological consequences resulting from a design basis MSLB accident to a person at the EAB; to a person at the LPZ; and to an operator in the Control Room following an MSLB accident were performed using a Microsoft EXCEL spreadsheet, provided as Attachment A. 2.2 Source Term Model No fuel damage is expected to result from a MSLB. Therefore, the activity available for release from the break is that present in the reactor coolant and steam lines prior to the break, with two cases analyzed. Case 1 is for continued full power operation with a maximum equilibrium coolant concentration of 0.2 uCi/gm dose equivalent I-131. Case 2 is for a maximum coolant concentration of 4.0 uCVgm dose equivalent I-131, based on a pre-accident iodine spike caused by power changes. In determining 1-131 equivalence, inhalation CEDE DCFs from Ref. 5.4 are used. This accident source term basis meets the guidance in R.G. 1.183 for analysis of this event. 2.3 Release Model The release model is identical to that historically used. The previously determined mass of reactor coolant release and mass of steam release, before the break is isolated by MSIV

CALCULATION NO. QDC-0000-N-1266 REV. NO. 2 PAGE NO. 5 of 14 closure, are used. Reactor coolant radioactivity is based on the above reactor coolant concentrations. Reactor steam radioactivity is based on a steam to coolant iodine concentration ratio (carry-over) of 2%. Releases are assumed to be instantaneous and no credit is taken for dilution in turbine building air. 2.4 Dispersion Model Onsite and Offsite X/O determinations are handled differently, but conservatively in both cases. 2.4.1 EAB and LPZ EAB and LPZ X/O's are determined using the original methodology in R.G. 1.5. Specifically: _x _ 0.0133 Q a Y u where Horizontal standard deviations are taken from the PAVAN outputs for the EAB and LPZ include in Calculation QDC-0000-N-1267. Per Regulatory Guide 1.5, F stability and a 1 meter/sec wind speed is used. 2.4.2 Control Room 6 Y = horizontal standard deviation of the plume (meters) u = wind velocity (meters/second) For control room dose calculations, the plume was modeled as a hemispherical volume, the dimensions of which are determined based on the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam. Activity release is conservatively assumed to effectively occur at the Control Room intake elevation and, again conservatively, no credit is taken for plume buoyancy. A conservative translation time of the plume over the intake is assumed. The activity of the cloud is based on the total mass of water released from the break, not just the portion that flashes to steam. This assumption is conservative because it considers the maximum release of fission products.

CALCULATION NO. QDC-0000-N-1266 1 REV. NO. 2 1 PAGE NO. 6 of 14 Dosec and 2.5 Dose Model Dose models for both onsite and offsite are simplified and meet R.G. 1.183 requirements. Dose conversion factors are based on Federal Guidance Reports 11 and 12. Doses at the EAB and LPZ for the MSLB are based on the following formulas: and and finally, 2.5.1 EAB and LPZ (rem) = Release (Curies)

  • Q (sec/m 3 )
  • Breathing Rate (M3/Sec) *Inhalation DCF (remCEDE /Ci inhaled)

Dose 2.5.2 Control Room CR operator doses are determined somewhat differently, because steam cloud concentrations are used, rather that X/O time a curie release rate. No CR filter credit is taken and, therefore, for inhalation, a dose for a location outside of the CR can be and is used. For cloud submersion, a geometry factor is used to credit the reduced plume size seen in the control room. This is a conservative implementation of FIG 1.183 guidance. The formulas used are: Dose 2.6 Acceptance Criteria (rem) = Release (Curies)

  • Q (sec/m 3 )
  • Submersion DCF (rem E'DE - m3/ Ci -sec)

DoseTEDE (rem) = Dosec (rem) + Dose EDE (rem) (rem) = Plume Concentration (Ci/m 3)

  • Transit Duration (sec)
  • Breathing Rate (M3 /sec)
  • Inhalation DCF (rem Dose EDE (rem) = Plume Concentration (Ci/m 3)
  • Transit Duration (sec)
  • Submersion DCF (remEDE - m 3/ Ci - sec) and finally, DoseTEDE (rem) = DosecEDE (rem) + DoseEDE (rem)

Dose acceptance criteria are per 10CFR50.67 and R.G. 1.183 guidance. /Ci inhaled) Table 1 lists the regulatory limits for accidental dose to 1) a control room operator, 2) a person at the EAB, and 3) a person at the LPZ boundary.

I CALCULATION NO. QDC-0000-N-1266 1 REV. NO. 2 PAGE NO. 7 of 1104-~- ~ I Table 1. Regulator Y Dose Limits (Rem TEDE) 1-131 Dose Equivahara CR (30 days) EAB (2 hours) LPZ (30 days Normal Equilibrium 5 2.5 2.5 Iodine Spike 5 25 25

CALCULATION_ NO. QDC 0000-N-1266 - -I REV. NO. 2 PAGE NO. 8 o f 14 3.0 ASSUMPTIONS 3.1 Activity Release and Transport Models " Iodine activity distribution in the coolant was taken from the Quad Cities UFSAR, Section 15.6.4.5. " Release from the break to the environment is assumed instantaneous. No holdup in the Turbine Building or dilution by mixing with Turbine Building air volume is credited. " The steam cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam. " The activity of the cloud is based on the total mass of water released from the break, not just the portion that flashes to steam. This assumption is conservative because it considers the maximum release of fission products. " The fraction of liquid water contained in steam, which carries activity into the cloud, was assumed to be 2%. Flashing fraction of liquid water released was assumed as 40%, as derived in section 6.1 below. However, all activity in the water is assumed to be released. For offsite dose calculations the release is treated as a point source with ground level dispersion per R.G. 1.145. Buoyancy effect of the cloud was conservatively ignored. For the control room dose calculations, the plume was modeled as a hemispherical volume. This is consistent with the assumption of no Turbine Building credit. It is also reasonable for the more likely release paths through 8 large blowout panels situated around the Turbine Building Main Floor. A dispersion of the activity of the plume was conservatively ignored. D The cloud was assumed to be carried away by a wind of speed 1 m/s. Credit is not taken for decay. 3.2 Control Room Model " No credit was taken for the operation of the CREFS during the MSLB. " Inhalation doses are determined based on concentrations at the intake, and exposures for the duration of plume traverse. External exposure doses are determined based on concentrations at the intake, exposures for the duration of plume traverse, and a geometry factor credit based on the Control Room proper volume of 58,300 cubic feet. 3.3 Site Boundary Model This model is as discussed in Subsection 2.5.2 above.

CALCULATION NO. QDC-0000-N-1266 REV. NO. 2 PAGE NO. 9 of 14 4.0 DESIGN INPUT 4.1 Mass Release Data " The mass steam released is 20,000 lb. [5.2] " The mass liquid water released is 10,125 lb. [5.2] 4.2 Iodine Distribution The distribution of fission products in the coolant was obtained from the UFSAR for Quad Cities. The relative mix of iodine isotopes in the reactor coolant at the onset of the accident, based on Quad Cities UFSAR is given below (section 15.6.4). [5.13] This analysis ignores the contribution from other halogens and noble gasses, which have negligible dose consequence. 4.3 Control Room Data Control Room proper volume = 58,300 W. [5.1] " No Emergency Filtration Credit taken. 4.4 EAS and LPZ Data " EAB Distance from Release, m 380 (Tech. Spec., section 5) [5.15] " LPZ Distance from Release, m 4,828 (Tech. Spec., section 5) [5.15] Iodine Isotope I-131 Activity (/jCVcc) 0.067 I-132 0.38 I-133 0.40 I-134 0.53 I-135 0.49

I CALCULATION NO. QDC-0000-N-1266 I REV. NO. 2 1 WOE 19D. 10 of 14

5.0 REFERENCES

5.1 Calculation QDC-0000-N-1 020, "Impact of Extended Power Uprate on Site Boundary and Control room Doses for LOCA and Non-LOCA Events", Revision 0. 5.2 Calculation QDC-9400-M-0365, "Control Room and Site Boundary Doses Following a Main Steam Line Break", Revision 2. 5.3 Not Used 5.4 Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion", 1988. 5.5 Federal Guidance Report No. 12,External Exposure to Radionuclides in Air, Water, and Son 1993. 5.6 Not Used. 5.7 Not Used 5.8 Not Used 5.9 Not Used. 5.10 NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980. 5.11 Regulatory Guides 1.5, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accidents for Boiling Water Reactors," 3/10/71. 5.12 Regulatory Guide 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors", July 2000 5.13 Quad Cities Nuclear Power Station UFSAR Rev. 7, Section 15.6.4. 5.14 Code of Federal Regulations: 10 CFR Part 50.67 5.15 Quad Cities Technical Specifications. 5.16 Calculation ODC-0000-N-1 267, "Reanalysis of Fuel Handling Accident (FHA) Using Alternative Source Terms", Revision 0

CALCULATION NO. QDC-0000-N-1266 REV. NO. 2 PAGE NO. 11 of 14 6.0 CALCULATIONS No or minimal fuel damage is expected for the limiting MSLB. As discussed in section 2, two iodine concentrations will be used (0.2,jCi/g and 4.0IjCi/g) when determining the consequences of the main steam line break. All of the radioactivity in the released coolant is assumed to be released to the atmosphere instantaneously as a ground-level release. No credit is taken for plateout, holdup, or dilution within facility buildings. The spreadsheets in Attachment A perform this analysis using data and formulations discussed above. The following summarizes parameters and their treatment in the spreadsheet. 6.1 Cloud Volumes, Masses, and Control Room intake Transit Times The cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashes to steam. The flashing fraction (FF) is derived as follows: FF x (steam enthalpy at 212 F) + (1-FF) x (liquid enthalpy at 212 F) _ (liquid enthalpy at temperature of steam at reactor vessel outlet) A 548 F vessel outlet temperature is used, with liquid enthalpy of 546.9 BTU/lb At 212 F, a steam enthalpy of 1150.5 BTU/lb and a liquid enthalpy of 180.17 BTU/Ib are used (these enthalpies are taken from the ASME Steam Tables). Substituting, FF = (546.9 - 180.17) / [(1150.5 -180.17)] =.378 For conservatism, a value of.40 or 40% is used. As stated in Section 3.1, the cloud is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam. Mass of water carrying activity into the cloud was calculated as the sum of the fraction of water in the steam and the liquid blowdown. The mass of steam in the cloud = (20,000 - 400) + 40%*10,125 Ib = 23,650 Ib The release is assumed to be a hemisphere with a uniform concentration. The cloud dimensions (based on to 10,525 Ib of steam at 14.7 psi and 212 °F, v9 = 26.799 ft3/Ib) were calculated as follows: The mass steam released = 20,000 Ib The mass liquid water released = 10,125 Ib Flashing fraction for calculating cloud volume = 40% The mass water contained in steam released = (20,000 lb) 2% = 400 Ib The mass of water carrying activity into the cloud = 400 + 10,125 Ib = 10,525 Ib = (10,525 Ib)(453.59 g/Ib) = 4.774E6 g

CALCULATION NO. QDC- -N-1266 REV. NO. 2 PAGE NO. 12 of 14 Volume = (23,650 Ib~(26.799 ft3/Ib) = 633,796 ft = (633,796 ft3)/(35.3 ft3/m3) =17, 17,955 m3 The volume of a hemisphere is n d3 /12. Thus, the diameter of the hemispherical cloud is 40.9 m. The period of time required for the cloud to pass over the control room intake, assuming a wind speed of 1 m/s is 40.9 s (=(40.9 m)/(1 rn/s)). Therefore, at a wind speed of 1 m/s, the base of the hemispherical cloud will pass over the control room intake in 40.9 seconds. 6.2 Dispersion for Mite Dose Assessment As discussed in Section 2.4.1 the following formulation was used for Offsite Dose X/Q assessment, with F Pasquill Stability and a 1 n-/sec wind speed. 0.0133 Q O Yu where QY = horizontal standard deviation of the plume (meters) u = wind velocity (meters/second) As calculated in the PAVAN run in Reference [5.16], at the 380 meter EAB distance ay is 15.4, and at the 4828 meter LPZ distance ay is 153. The resulting EAB and LPZ X/Qs are 8.64E-04 and 8.68E-05 sec/m3, respectively. 6.3 Release lsotopics and Quantification The iodine isotopic distribution given in Section 4.2 is used. The concentrations of this mix are adjusted to I-131 equivalence, using the inhalation Committed Effective Dose Equivalent (CEDE) Dose Conversion Factors (DCFs) from Federal Guidance Report No. 11 [Ref. 5.4]. This is a more conservative set of DCF assumptions for Control Room and off-site dose calculation than the prior use of ICRP 2 DCFs. It is also more conservative for these calculations than use of R. G. 1.109 or Federal Guidance Report No. 12 [Ref. 5.5] DCFs. This I-131 equivalent mix is adjusted to the activity yielding the two design basis MSLB accident reactor coolant activities of 0.2 gCVcc and 4.0 ACi/cc. The released activities is these concentrations times the 4.77E+06 grams of activity released, with the assumption that TS activities are based on laboratory temperature and pressure conditions.

I CALCULATION NO. QDC-0000-N-1266 I REV. NO. 2 1 PAGE N0.13 of 14 1 6.4 Dose Assessment Doses at the EAB and LPZ distances, and in the Control Room are calculated in Attachment A using the formulas in Section 2.5. Concentrations at the receptor locations are that in the steam plume for the Control Room or based on the release times the applicable X/Q for the EAB and LPZ. Doses are calculated for inhalation (rem CEDE) and plume submersion (rem EDE) and totaled to yield rem TEDE. The breathing rate of 3.47E-04 m3/sec is per RG 1.183 guidance without the round-off. The resulting calculated doses are in the spreadsheet and in the Summary and Conclusions Section below.

I CALCULATION NO. QDC-0000-N-1266 I REV. NO. 2 ~TPAGE 0.14 of 14 7.0

SUMMARY

AND CONCLUSIONS Accidental doses from a design basis MSLB were calculated for the control room operator, a person at EAB, and a person at LPZ. The results are summarized in the Table below. The doses at the Control Room, EAB, and LPZ resulting from a postulated design basis MSLB do not exceed and are small fractions of the regulatory limits. Location Case l (normal equilibrium limit of 0.2 jiCQ Does (rem TIEDE) Case 2 (iodine spike limit of 4.0 gCi) Dose (rem TEDE) LIMITS CR 5.0 ; EAB&LPZ 2.5 CR 5.0 ; EAB&LPZ 25 EAB 1.25E-02 2.50E-01 LPZ 1.25E-03 2.51 E-02 CR 20 E02 5.02E-01

Calculation QDC-0000-N-1266, Rev. 2 Attachment A Page A1 of A6 2/26/03 0e.. Cities MSLB D.~ se Spreadsheet Reactor Coolant at maximum value (DE 1-131 of 0.2 uCi/cc) © permitted. continued full power operation 3 4 7 17955 4.77E+06 Volume of cloud (cubic meters) Case 2 : Mass of water in reactor coolant release (grams) ]uCi/cc) Reactor Coolant at maximum value permitted (DE I-131 of 4.0 corresponding to an assumed pre-accident spike 5 1 reactor coolant density when activity is measured (grams/cc) 6 4_0_.9 seconds for cloud to pass over CR intake for wind speed of 1 m/second 7 Volume of Control Room proper (cub 8 __58300 Case 1 Case 2 Case 1 Case 2 9 Activity Activity Release Release 10 Normalized Case 1 Case 2 Release Release__ Cloud Cloud 11 Isotope Activity FGR 11 I-131 DE Normalized Normalize Actual (not Concentration Concentration 12 Distribution DCF' Activity Activity Activity normalized © uci/cc Ci Ci Ci/m' Ci/m 14 1-131 0.067 3.29E+04 6.70E-02 8.23E-02 1.65E+00 3.93E-01 7.85E+00 2.19E-05 4.37E-04 © 1-132 0.38 3.81 E+02 4.40E-03 5.40E-03 1.08E-01 2.23E+00 4.45E+01 1.24E-04 2.48E-03 16 I-133 0.4 5.85E+03 7.11 E-02 8.73E-02 1.75E+00 2.34E+00 4.69E+01 1.31 E-04 2.61 E-03 17 1-134 0.53 1.31 E+02 2.11 E-03 2.59E-03 5.18E-02 3.11 E+00 6.21 E+01 1.73E-04 3.46E-03 18 1-135 0.49 1.23E+03 1.83E-02 2.25E-02 4.50E-01 2.87E+00 5.74E+01 1.60E-04 3.20E-03 19 20 - 21 22 23

Calculation QDC-0000-N-1266, Rev. 2 Attachment A Page A2 of A6 2126103 A _13 25 26 27 ies Released Case 1 D Gase 2 ose rte_ 28 7 -r"

the to Environment (Inhalation (Inhalation) 29 Isotope Case 1 Case 2 RemCEDE/Ci EAB LPZ CR EAB LPZ 30 W131 3.93E-01 7.85E+00 3.29E+o4 3.87E-03 3.89E-04 2.04E-01 734ET2 738EW M__1-132 2.23E+00 4.45E+01 311E+02 6.71 E-04 214EM MEM MET2 519003 5.11 E-04 32 1-133 2.34E+00 4.69E+01 5.85E+03 1.08E-02 4
11ET3, 4.13E-04 2.17E-01 8211002 826&03 33 1-134 3.1 1 E+00, 6.21E+01 1.31 E+02 3.22E-04 1.22E-04 1.23E-05 6.43E-03.

2.44E-03 2A5&04 34 '_ V 05 05 ---- 2.87E+001 5.74E++01 01 113E+03 2.79E-03 1.06E-03 116E-04 5.58E-02-02 2.12E-6QjT13E W 35 2.48E-02 9.41 E-03 ME-041 4.97E-01 I 1.88E-01 1.89E-02 37. L 38 Curies Released DCF Case Dose (rem EDE) Case 2 Dose Own EDE) 39 to the Environment ReMEDE -m3/ (External) (External) 40 41 Notop e 1-131 Case 1 313EM1 Case 2 TBIDTO Ci-second 6.73E-02 Ck 2AOE46 EAB 238&05 LPZ 230EM6 CR 4119ET5 EAB ME-0 Z_ AJOE-005 42 43 ____I-1_32 1-133 2.23E+00 2.34E+00 4.4500 4.69E+01 WRE-01 1.09E-01 7.31 E-05 2.02ET5 7.90-04 210EM4 811 E05 2.21 E-05 1.46E-03 4.04ET4 1.59E-02 4AOE-O'- 1.60E-03 4.43E-04 44 45 46-1-134 1-1351 Sub-total 3.11 E+00 2.87E+00 6.21 001 f-k.74E+01 1 2.95E-01 -04 WE= VAN WAS 3 732ET4 -3A-6-E-4 1.30E-04 7W&05 __3.08E-04 2-37E-03 1.34E-03. 5.62E-03 2.58E-02 1.46E-02 6.13E-02 2.59E-03 _104-70 6.16E-03 47 Total 2.51 E.02Lt.25E .25E.03 5A2001 2ZOO01 2.51 E-02 48 Inhalation Effective Dose Conversion Factor (RenaEDE/Curie) from Federal Guidance Report (FGR) 11 per Reg. Guide 1.183 49 Effective Ease Conversion Factor (rem-ni/Cude-second) for Air Submersion from FGR 12 per Reg. Guide 1.183 L50 WE= Breathing rate( /second) per Regulatory Guide 1.183 (without round-off) 911771 FUT-TUMNIM AMIAMM. 55 8.64EM4 XJQ (seconds/rrf) at EA Boundary 2 hours based on RG 1.5 methodology 56 8.68E-05 Q (seconds/ ) at Low Population Zone 2 based on RG 1.5 methodology

Calculation QDC-0000-N-1266, Rev. 2 Attachment A Page A3 of A6 256M3 7-G RMM Reactor Coolant at maximum value 1 17955 Case 2 : Reactor Coolant at maximum value 4 9 4774000 5 1 WOU Case 1 9 Activity 1101 Normalized Case 1 Case 2 Release I I Isotope Activity FGR 11 1-131 DE J-Normalize Normalized Actual (not norm © ~ Activit y Activity Activity uCi/cc uCi/cc uCi/cc Ci 14 1-131 0.067 32900 =C14*B14/C$l =U14i0.2/D$19

  • 20

=(C$14/C 1 4)'E 1 4*$A$4 -0. 00-0 1 1-132 0.38 381 =C 1 5*B 1 5/C$1 =D 1 5*0.2/D$19 =E 1 5*20 =(C$14/C 1 5)-E 1 =(C$14/Cl6)*EI6*$A$4*0.000001i // / 5-$A$4*0. 00000 1 / 16 1-133 0.4 5846 =C 16*13 1 6/C$1 =D 1 6*0.2/D$19 =E 16*20 17 1-134 0.53 131 =C17*B17/C$1=D17*0.2/D$19 =E17*20 =(C$14/C 1 7)*E 1 7-$A$4*0. 000001 18 1-135 0.49 1230 =ClWBIWC$1=Dl8*0.2/D$19 =E18*20 -- =(C$14/C 1 8)*E 1 8*$A$4*0. 00000 1 . 19 1-=SUM(Dl4:Dl =SUM(E14:E18 =SUM(F14:F18) 20 "non-spiked" "spiked-' _____ .21 22 23 24 25 26 27 Curies Released Case 1 Dose (rem CEDE) 28 to the Environment DCF 1 (Inhalation) 29 Isotope Case 1 case 2 RemCEDE/Ci CR EAB 30 1-131 =G14 I=IH14 32900 =114*$E30*$A$50*$A =C30*$E30*$A$50*$A$55 31 1-132 381 =1 1 F$E31 *$A$50*$A =C31 *$E31 *$A$50*$A$55 32 1-133 =G16 j=H16 5846 1=116*$E32*$A$50*$A =C32*$E32*$A.T,50*$A$55 33 1-134 _---] =1 1 7*$E33*$A$50*$A -=C33*$E33*$A$50*$A$55 4 35 36 34 1-135 J=G18 =H18 1230 Totals =118*$E34*$A$50*$A I=SUM(F30:F34) =C34*$E34*$A$50*$A$55 7-SUM(G30*G34)

Calculation QDC-0000-N-1266, Rev. 2 Attachment A Page A4 of A6 2/26/03 A B C D -© 37 38 - Released DCFZ Case 1 Dose (rem EDE) 39 nvironment ReMEDE-m3/ (External) 40 Isotope Case 1 Case 2 Ci-second CR EAB 41 1-131 =C30 =D30 0.06734 =114*$E41 1A$51 *$A =C41 *$E41 *$A$55 42 1-132 =C31 =D31 0.4144 =115*$E42*$A$51 *$A =C42*$E42*$A$55 43 1-133 =C32 =D32 0.10878 =116*$E43*$A$51 *$A =C43*$E43*$A$55 44 1-134 =C33 =D33 0.481 =117*$E44*$A$51 *$A =C44*$E44*$A$55 45 1-135 _ =C34 =D34 0.29526 =118*$E45*$A$51 *$A =C45*$E45*$A$ 55 46 Sub-total--------- =SUM(F41 :F45) =SUM(G41 :G45) 47 Total =SUM(F35+F46) =SUM(G35+G46) 48 ' Inhalation Effectiv 49 2 Effective Dose Co 50 0.000347 Breathing',- IIIIII'll'lll'L 51 =($A$7^0.338)/117' Control R -. .,I 52 15.4 _ EABay ( --~ ©153.2 _ LPZ ay (m~ 54 1 Wind Spe, -~ 55 =0.0133/A$52/A$ X/Q (seco 56 =0.0133/A$53/A$ X/Q (seco

Calculation QDC-0000-N-1266, Rev. 2 Attachment A Page A5 of A6 2/26/03 1 o u i cc permitted or continued full power operation 4 ©ermitted (DE I-131 of 4.0 uCi/cc) corresponding to an assumed pre-accident spike 6 8 9 _Case 2 _ Activity Case 1 Release Case 2 Release 10 Release_ Cloud Cloud 11 lized) Releases Concentration Concentration © Ci Ci/ Ci/m 14 =(C$14/C1 4)*F14*$A$4*0.000001/$=G14/$A$3 =H14/$A$3 15 =(C$14/C15)*F15*$A$4*0.000001/$=G15/$A$3 =H15/$A$3 16 =(C$14/C16)*F16*$A$4*0.000001/$=G16/$A$3 =H16/$A$3 17 =(C$14/C17)*F1 7*$A$4*0.000001/$=G17/$A$3 =H17/$A$3 18 =(C$1 4/C18)*F18*$A$4*0.000001/$=G18/$A$3 =H18/$A$3 19 20 21 23 24 . 25 27 Case 2 Dose (rein CEDE 28 (Inhalation) 29 LPZ CR EAB LPZ 30 =C30*$E30*$A$50*$A$56 =J14*$E30*$A$50*$A$ =D30*$E30*$A$50*$A$5 =D30*$E30*$A$50*$A$56 31 =C31 *$E31 *$A$50*$A$56 =J 15*$E31 *$A$50*$A$ =D31 *$E31 *$A$50*$A$5 =D31 *$E31 *$A$50*$A$56 32 =C32*$E32*$A$50*$A$56 =J16*$E32*$A$50*$A$ =D32*$E32*$A$50*$A$5 =D32*$E32*$A$50*$A$56 33 =C33*$E33*$A$50*$A$56 =J17*$E33*$A$50*$A$ =D33*$E33*$A$50*$A$5 =D33*$E33*$A$50*$A$56 34 =C34*$E34*$A$50*$A$56 =J18*$E34*$A$50*$A$ =D34*$E34*$A$50*$A$5 =D34*$E34*$A$50*$A$56 35 =SUM(H30:H34) =SUM(130:134) =SUM(J30:J34) =SUM(K30:K34) 1361

Calculation QDC-0000-N-1266, Rev. 2 Attachment A Page A6 of A6 2126103 H K 37 38 _. Case 2 Dose (rem EDE) 39 (External) 40 LPZ CR EAB LPZ 41 C41*$E41*$A $56 =J14*$E41*$A$51*$A$ =D41*$E41*$A$55 =1341*$E41*$A$56 42 =C42*$E42*$A$56 =J15*$E42*$A$51*$A$ =D42*$E42*$A$55 =D42*$E42*$A$56 43 =C43*$E43*$A$56 =J 16*$E43*$A$51 *$A$ =D43*$E43*$A$55 =D43*$E43*$A$56 44 =C44*$E44*$A$56 =J17*$E44*$A$51*$A$ =D44*$E44*$A$55 =D44*$E44*$A$56 45 =C45*$E45*$A$56 =J18*$E45*$A$51*$A$ =D45*$E45*$A$55 =D45*$E45*$A$56 46 =SUM(H41 :H45) =SUM(141 :145) =SUM(J41 :J45) =SUM(K41 :K45) 47 =SUM(H35+H46) =SUM(135+146) =SUM(J35+J46) =SUM(K35+K46) 48 49 50 52 --~~1_ 53 55 56

I CALCULATION NO. QDC-0000-N-1266, Attachment B ~7FF REV. NO. 2 1 PAGE NO. Bi of B1 Computer Disclosure Sheet Discipline Nuclear Client :: Exelon Corporation Date: 2/26/03 Project: Quad Cities Units 1&2 MSLB AST Job No. Program(s) used Rev No. Rev Date Calculation Set No. : QDC-0000-N-1266, Rev. 2 Attachment A spreadsheet N/A N/A Status Prelim. X Final Void WGI Prequalification Yes No Run No.

== Description:== Analysis

Description:

Spreadsheet used to perform dose assessment for MSLB, as described in calculation. The attached computer output has been reviewed, the input data checked, And the results approved for release. Input criteria for this analysis were established. By : On: 2/26/03 Run by : P. Reichert Checked by: H. Rothstein Approved by : H. Rothstein Remarks: WGI Form for Computer Software Control This spreadsheet is relatively straight-forward and was hand checked. Attachment includes the spreadsheet in both normal and formula display mode and so is completely documented.}}

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