Text

Calculation LM-0645, Rev 1, Re-analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms, Attachment 009
	ML053330373
Person / Time
Site:	Limerick
Issue date:	09/27/2005
From:	Reichert P, Rothstein H; Exelon Generation Co, Exelon Nuclear
To:	; Office of Nuclear Reactor Regulation
References
	CC-AA-309-1001, Rev 2 LM-0645, Rev 1
	Download: ML053330373 (86)
	v • d • e

ADDITIONAL ATTACHMENTS TO 10-10-05 Letter: Supplement to Request for LAR Application of AST Attachment 009 AST - LM-0645 Rev 1 FHA.

CC-AA.309-1001 Revision 2 ATTACYLMDT 1 Desixn Analysis CoeyfrSheet Design Analysis (Major Revision) Last Page No. e 18/ Att G-3 Analysis No.:' LM-0645 Revislon 2 1 itUe 3 Re-analysis Of Fuel Handling Accident (FHA) Using Alternative Source Terrr ECIECR No_

04-00003 Revision:6 0 Station(s):7 Linerlck Component(s):t Unt No.: ' 1& 2 M/A Discipline 9 MEDC Descrip.

Code/Keyword: '° H84 IAST. FHA SafetyfQA Class:" SR System Code ' 2 912 Structure: " N/A CONTROLLED DOCUMENT REFERESCES "

Document No.arorntlo Document No.: nremTo LGS UFSAR ~romnTo Design Analysis LM-0312 Design Analysis LM-0641 :rrx Tech. Spec. 314.9.4 ror Calcuiaton M-78 01 rom _ >DwNo.M-102, M-107 Fo Dwg No. SIM-M-76, Sheet 2 Prom Is this Design Analysis Safeguards Information?

Yes a No 0 Ifyes. see SY-AA-101-106 Does this Design Analysis contain Unvertfied y 0 yes, TIf Assumptions? 7 es ATiAR#:

This Design Analysis SUPERCEDES:" LM-0645. Rev. 0 entirnty Description of Revislon (list affected pages for partials):"

This revision Incorporates responses to pertinent NRC Requests for Addiional Information (RAis) with respect to all Exelon Nuclear Station Alternative Source Term License Amendment Applications The fuel damage assessment utilized was revised to agree wIth that currently hI the UFSAR, ecept for use of a more conservative peeking factor to provide bounding results.

Bounding results for other previously considered FHA scenarios are also Included. Finally, additional assumptions from Regulatory Guide 1.183 are Included to directly Indicate contormanse with this Regulatory Guide.

Preparer: Harold Rothstein - _ 9127/2005 Print Name Sign Name Date Method of Detailed Review Ateate Calculatins (attace Testing 0 Reviewer. Paul Reichert 9127/2005 Print Name Sign NamD ate PM AAUI 2Review Independent review ID Peer reviqw a Note: a1~f7NJW .LOC : D5> 5(MP Approver?/ 0s Is a SupplemPental Review RNquired? nmYes No im e ttnent 3 Approver. '. .ICk-52 Print Narne Sign Narne Date

I CALCULATION NO. LM-0645 I REV. NO. I I PAGE NO. 2 Table of Contents

1. PURPOSE/OBJECTIVE ........................................................ 3

2. METHODOLOGY AND ACCEPTANCE CRITERIA ....................................................... 4 2.1. Fuel Source Term Model . . . 4 2.2. Gap Activity . . . 5 2.3. Pool Decontamination Factor (DF) . . . 5 2.4. Release Model . . . 6 2.5. Control Room Model . . . 6 2.6. Dose Modeling . . . 6 2.6.1. EAB and LPZ ........................................................ 7 2.6.2. Control Room ........................................................ 7 2.7. Acceptance Criteria ........................................................ 7

3. ASSUMPTIONS.................................................................................................................... I1

4. DESIGN INPUT ......................................................... 12

5. REFERENCES .................................................... 14

6. CALCULATIONS................................................................................................................ 15 6.1. New Dose Analysis Basis ........................................................ . 15 6.2. Margin Assessment for Other Previously Considered FHA Scenarios ..................................... 17

7.

SUMMARY

AND CONCLUSIONS ........................................................ 17

8. OWNER'S ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS..................................................................................................................................8... 18 Attachments A. Source Terms [1 5 pgs.]

B. RADTRAD Runs [24 pgs.]

C. FHA RADTRAD Nuclide Information File [10 pgs.]

D. FHA RADTRAD Release Fraction File [I pg.]

E. LGS Fuel Handling Accident Assessment of Limiting Event [8 pgs.]

F. Computer Disclosure Sheet [I pg.]

G. Evaluation of Bounds for Other Previously Considered FHA Scenarios [3 pgs.]

I CALCULATION NO. LM-0645 I REV. NO. 1 I PAGE NO. 3 l

1. PURPOSE/OBJECTIVE The purpose of this calculation is to apply Alternative Source Term (AST) methodology to the analysis of the Fuel Handling Accident (FHA) for Limerick Generating Station (LGS) Units 1 & 2.

The calculation is based on normal Reactor Building unfiltered exhaust through the South Stack with no Control Room Emergency Filtration (CREF). Therefore, this calculation supports changes to the current LGS I & 2 Technical Specifications (TS) to consider that maintenance of the secondary containment integrity and the operability of emergency filtration systems and subsystems previously required to mitigate the radiological consequences of fuel handling accidents may not be necessary. Not having to consider secondary containment and control room integrity and filtration requirements in support of refueling activities has the potential to significantly improve the flexibility and duration of scheduled plant outage activities. Based on LGS Technical Specification 3/4.9.4 on Decay Time for movement of irradiated fuel in the reactor pressure vessel, in agreement with the discussion in UFSAR [Ref. 1] Sections 15.7.4.5 and especially 15.7.4.5.2.1, movement of irradiated fuel will not occur less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the associated reactor is subcritical, and therefore, a 24-hour delay period is used. This value continues to be a very conservative assumption for BWRs, given the operations necessary before commencing fuel movement.

A "recently irradiated fuel" parameter is considered as the point in time after shutdown when secondary containment integrity features are not required. Therefore, this calculation also justifies 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as this time after shutdown.

Regulatory Guide 1.183 [Ref. 2] is the basis for these evaluations. Concerning the FHA, this AST guidance has the advantage of smaller gap fractions, a larger pool decontamination factor

[DF], and dose criteria that replace both the whole body and thyroid dose limits with a limit on Total Effective Dose Equivalent [TEDE].

Other changes from the current UFSAR calculation are listed below.

An offsite dose limit of 6.3 rem TEDE (Reg. Guide 1.183 & 10 CFR 50.67(b)(2)(iii)) is applied instead of the Standard Review Plan (SRP) 15.7.4 values of 25% of the I OCFR100 limits.

A control room dose limit of 5 rem TEDE (10 CFR 50 Appendix B (II) GDC 19) is applied instead of the SRP 15.7.4 values of 5 rem whole body, or its equivalent.

Analysis is based on NRC Regulatory Guide 1.183.

Secondary Containment automatic isolation and filtration are not credited.

The Control Room and offsite X/a's were recalculated for release from the Unit 1 & 2 Reactor Building (RB) South Stack as the worst case release location with respect to the Control Room intake for no Standby Gas Treatment System (SGTS) filtration, and the new limiting "/a's were applied to this analysis.

CREF operation is not credited.

Dose Conversion Factors (DCFs) for Immersion and Inhalation are taken from Federal Guidance Reports (FGRs) 12 [Ref. 5] and 11 [Ref. 4], respectively. Regulatory Guide 1.183 cites these DCFs as acceptable current estimates for evaluating the radiological impact of nuclear plant accidents.

This calculation also documents the development of core source terms to be used in this and other accident analyses that involve postulated fuel damage and are being reanalyzed using alternative source terms.

I CALCULATION NO. LM-0645 I REV. NO. I PAGE NO.4

2. METHODOLOGY AND ACCEPTANCE CRITERIA Analyses of radiological consequences resulting from a Fuel Handling Accident (FHA) are performed using the guidance for application of Alternative Source Terms to this event in Regulatory Guide (RG) 1.183.

Analyses of radiation transport and dose assessment are performed using RADTRAD v. 3.03.

RADTRAD is a simplified model of RADionuclide Transport and Removal And Dose Estimation developed for the NRC and endorsed by the NRC as an acceptable methodology for reanalysis of the radiological consequences of design basis accidents. The technical basis for the RADTRAD code is documented in NUREG/CR-6604 [Ref. 3]. The methodologies significant to this analysis are the dose consequence analysis (NUREG Section 2.3) and the Radioactive Decay Calculations (NUREG Section 2.4). This version of RADTRAD has been pre-qualified for safety related design analysis by Washington Group International per its 10CFR50 Appendix B Quality Assurance program.

2.1. Fuel Source Term Model The dose assessments in this calculation use the UFSAR (Ref. 1) historical fuel damage assumptions of a total of 212 failed rods based on an 8x8 fuel design containing 62 fuel rods, but with a conservatively higher radial peaking factor (PF) of 1.7 instead of 1.5. As per UFSAR Section 15.7.4, the analytical methodology and licensing bases for determination of fuel damage in a FHA are provided in GESTAR II, and compliance with these bases is verified for each new fuel design. The dose analysis in this calculation applies the UFSAR Section 15.7.4 additional conservative assumptions so that this calculation continues to provide margin.

The assumed accident, per UFSAR Section 15.7.4, is an assembly and mast drop from the maximum height allowed by the refueling platform (a height of 32 feet for the fuel assembly, and 47 feet for the mast) over the reactor well onto fuel in the reactor. Based on fuel damage assessments in Ref. 9 and as shown below, this bounds the damage assessments for various 8x8 and 7x7 array fuel types with 60 and 49 fuel pins per bundle, respectively, and 111 failed pins and a 1.5 PF, as well as GE11 or GE13 9x9 array fuel types with 74 fuel pins per bundle,140 failed pins and a 1.5 PF, and GE12 or GE14 10x10 array fuel types with 172 failed pins per bundle and a 1.7 PF.

Fuel Pins in Damaged CoreDmgdCr Bundle Type Array Bundle Failed Pins Damagdctionre PF Fraction multiplied Fractionby PF Previous UFSAR Basis with 1.7 PF [New Dose Analysis Basis] 8x8 62 212 0.004476 1.7 0.007608 Previous UFSAR Basis 8x8 62 212 0.004476 1.5 0.006713 Various 8x8 60 111 0.002421 1.5 0.003632 Various 7x7 49 111 0.002965 1.5 0.004448 GE11&GE13 9x9 74 140 0.002476 1.5 0.003714 GE12&GE14A 1Ox10 87.33 172 0.002578 1.7 0.004382 A. Bounding Assembly type in current use, with higher peaking factor commensurate with full core application for l0xl0 fuel B. Damaged Core Fraction = Failed Pines I Pins in Bundle / 764 bundles in core

CALCULATION NO. LM-0645 l REV. NO. I PAGE NO. 5 l The associated power of the UFSAR Basis damaged fuel = 3527 MWth

0.007608 = 26.83 MWth.

For bounding dose assessment purposes, the fuel source term is based on the reactor core source terms described in Attachment A. These source terms are bounding for LGS fuel cycle designs as documented on the last two pages of Attachment A.

2.2. Gap Activity This calculation is applicable to fuel whose burnup and power limits are bounded by those specified in RG 1.183, footnote 11. This allows application of the gap activity fractions for LOCA events per RG 1.183, Table 3, which are as follows:

5% of the noble gases (excluding Kr-85) 10% of the Kr-85 5% of the iodine inventory (excluding 1-131) 8% of the 1-131 12% of the Alkali metal inventory Because RADTRAD does not allow for application of isotope specific release fractions, the

'Limerick Generating Station AST Source Term.nif" file is modified to accommodate the differential gap activities among the halogen (1-131) and noble gas (Kr-85) gap fractions dictated by RG 1.183 [Ref. 2] shown above. Therefore, the initial activity of isotope 1-131 and Kr-85 are multiplied by 1.6 and 2.0, respectively, in order to accommodate the respective 10% and 8%

release fractions directed by regulatory guidance [Ref. 2].

2.3. Pool Decontamination Factor (DF)

Attachment E provides assessments of water coverage for FHAs over the reactor well and the spent fuel pool, and demonstrates that the drop over the reactor well is more limiting. This is due to the greater number of fuel rods damaged for the reactor well drop, and the fact that the lower iodine decontamination factor for a drop over the spent fuel pool is not significant enough to overcome the fuel damage difference.

As prescribed in RG 1.183, Appendix B, for the 23 feet or greater water depth, the overall effective DF of 200 is used, with DF1 ,Org = 285.29, which is the inorganic iodine DF that would yield an overall DF of 200 for a 23 foot water depth, when used in conjunction with the chemical fractions listed below. Conservatively, the 500 DF per RG 1.183, Appendix B is not credited.

fraction of inorganic iodine in fuel = 0.9985 fraction of organic iodine in fuel = 0.0015 fraction of elemental iodine above the water = 0.70 fraction of organic iodine above the water = 0.30

CALCULATION NO. LM-0645 lTREV. NO. I I PAGE NO. 6 2.4. Release Model Release modeling uses the RADTRAD computer program. As discussed in Section 2.2 above, the normal Nuclide Inventory File (NIF) representing a LGS core is artificially adjusted to account for the higher than average gap fractions for 1-131 and Kr-85 provided by RG 1.183.

The compartments are the Secondary Containment Refuel Floor Air Space, the Environment, and the Control Room. The refuel floor exhaust rate is set artificially high at 6 air changes per hour. This results in 99.9994% of the contained radioactivity being exhausted within two hours.

The exhaust point under the assumed no filtration condition is the Reactor Building South Stack as per Ref. 12. This release point results in specific dispersion characteristics which are defined by unique dispersion factors, or x/o's, as derived in Ref. 6. The North Stack, which is used for releases filtered by the SGTS, is located closer to the Control Room intake and therefore has higher x/I's, as also derived in Ref. 6. However, the SGTS is designed to remove at least 99%

of the iodine that would otherwise be released; this filtration more than overcomes the effect of the higher 1/0's, as demonstrated herein, so the South Stack release unfiltered is bounding.

Site walkdowns and specific reviews of the LGS General Arrangement Drawings such as M-102, the Plan at El. 217'- 0" (one foot above Grade) [Ref. 15] and M-107 [Ref. 16], the Section showing the North and South Stack, confirmed that there are no potential release pathways that could be worse with respect to the Control Room intake than the analyzed stacks. In particular, there are no hatches or single-door Reactor Building openings leading directly to the outside, and grade openings are considered to have x/0 's that are bounded by the South Stack release point x/O based on the greater distances of travel required for releases from them to the Control Room intake. This includes the large railroad doors at grade elevation, which could be postulated to be open at the same time as the equipment hatch cover on the refueling floor to support a future spent fuel cask move. Any other non-normal opening that could be postulated would be evaluated for its effects on this accident before such simultaneous opening would be allowed.

2.5. Control Room Model The Control Room (CR), as analyzed for this FHA analysis, is unfiltered. Although the normal maximum flow into the CR is 2100 cfm, a Control Room changeover rate of I CR volume per minute is used for conservatism and to allow for any unfiltered inleakage. Flow into the CR is therefore assumed to be 126,000 cfm, and to balance the system for analytical purposes, an equal flow of clean air is considered to leave the CR.

2.6. Dose Modeling Dose models for both onsite and offsite meet RG 1.183 requirements. Dose conversion factors are based on Federal Guidance Reports 11 and 12 [Ref. 4, 5]. RADTRAD uses the following formulations, integrated numerically over the accident duration:

I CALCULATION NO. LM-0645 I REV. NO. I I PAGE NO. 7 2.6.1. EAB and LPZ Doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) for the FHA are based on the following formulas:

DoseCEDE (rem) = Release (Curies)** (sec/M3 )

Breathing Rate (m3 /sec)

Inhalation DCF (remCEDE/Ci inhaled)

Q and DoseEDE (rem)= Release (Curies)

X (sec/M3 )* Submersion DCF (rem EDE -m 3 / Ci - sec)

Q and finally, DoSeTEDE (rem) = DoSeCEDE (rem) + DoseEDE(rem) 2.6.2. Control Room The formulas used by RADTRAD, by time increment, are:

DoseCEDE (rem) = Time Dependent CR Air Concentration (Ci/m3 )

Time Increment Duration (sec)

Breathing Rate (m3/sec)

Inhalation DCF(remCEDE/Ci inhaled)* OccupancyFactorof I and DoseEDE (rem) = Time Dependent CR Air Concentration (Ci/m 3 )

Time Increment Duration (sec)

Submersion DCF (rem EDE - m3 / Ci -sec)

Occupancy Factor of I

CR Geometry Factor and finally, DOseTEDE (rem) = DoseCEDE (rem) + DoseEDE (rem) 2.7. Acceptance Criteria Dose acceptance criteria are per 10CFR50.67 and RG 1.183 guidance.

Table 2.1 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.

Table 2.1. Regulatory Dose Limit (Rem TEDE)*

CR EAB LPZ (30 days) (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) (30 days 5 6.3 6.3 Direct conformance with the relevant sections of the body of Regulatory Guide 1.183 (such as the Acceptance Criteria provided above) and all of the Assumptions in its Appendix B "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident" is provided by this analysis, as shown in the Conformance Matrix Table 2.2.

I CALCULATION NO. LM-0645 I REV. NO. I I PAGE NO. 8-l Table 2.2: Conformance with RG 1.183 Appendix B (Fuel Handling Accident) :_-_-_ -

RG LGS Analysis Section: RG Position Comments -

1 Acceptable assumptions regarding core inventory and the release of Conforms These assumptions are radionuclides from the fuel are provided in Regulatory Position 3 of this guide. utilized; see Section 2 of this calculation 1.1 The number of fuel rods damaged during the accident should be based on a Conforms A conservative fuel conservative analysis that considers the most limiting case. This analysis damage analysis has should consider parameters such as the weight of the dropped heavy load or been performed; see the weight of a dropped fuel assembly (plus any attached handling grapples), Section 2.1 and the height of the drop, and the compression, torsion, and shear stresses on the Attachment E of this irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., calculation.

events over the reactor vessel), should be considered.

1.2 The fission product release from the breached fuel is based on Regulatory Conforms These assumptions are Position 3.2 of this guide and the estimate of the number of fuel rods breached. utilized; see Section 2.2 All the gap activity in the damaged rods is assumed to be instantaneously of this calculation.

released. Radionuclides that should be considered include xenons, kryptons, halogens, cesiums, and rubidiums.

1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool Conforms All iodine added to pool should be assumed to be 95% cesium iodide (Csl), 4.85 percent elemental is assumed to iodine, and 0.15 percent organic iodide. The Csl released from the fuel is dissociate.

assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.

2 If the depth of water above the damaged fuel is 23 feet or greater, the Conforms, the The decontamination decontamination factors for the elemental and organic species are 500 and 1, 500 DF for factor was determined in respectively, giving an overall effective decontamination factor of 200 (i.e., elemental a more conservative 99.5% of the total iodine released from the damaged rods is retained by the iodine is not manner than prescribed water). This difference in decontamination factors for elemental (99.85%) and used. A more in RG 1.183, as organic iodine (0.15%) species results in the iodine above the water being conservative described in Section 2.3 composed of 57% elemental and 43% organic species. If the depth of water is value of 285.29 of this calculation.

not 23 feet, the decontamination factor will have to be determined on a case- is used since it

I CALCULATION NO. LM-0645 I REV. NO. I PAGE NO.9 Table 2.2: Conformance with RG 1.183 Appendix B (Fuel Handling Accident)

RG LGS Analysis Section, RG Position .Comments by-case method. is the value that yields an overall effective DF of 200 for 23 feet of water when combined with the stated initial iodine fractions.

3 The retention of noble gases in the water in the fuel pool or reactor cavity is Conforms These assumptions are negligible (i.e., decontamination factor of 1). Particulate radionuclides are utilized.

assumed to be retained by the water in the fuel pool or reactor cavity (i.e.,

infinite decontamination factor).

4.1 The radioactive material that escapes from the fuel pool to the fuel building is Conforms This assumption is assumed to be released to the environment over a 2-hour time period. utilized. No credit is taken for the SGTS.

4.2 A reduction in the amount of radioactive material released from the fuel pool by Not Applicable No credit is taken for engineered safety feature (ESF) filter systems may be taken into account filtration from the reactor provided these systems meet the guidance of Regulatory Guide 1.52 and building.

Generic Letter 99-02. Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

4.3 The radioactivity release from the fuel pool should be assumed to be drawn Not Applicable Two-hour release to the into the ESF filtration system without mixing or dilution in the fuel building. If environment is mixing can be demonstrated, credit for mixing and dilution may be considered assumed, without on a case-by-case basis. This evaluation should consider the magnitude of the mixing or dilution.

building volume and exhaust rate, the potential for bypass to the environment, the location of exhaust plenums relative to the surface of the pool, recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the pool and the exhaust plenums.

ICALCULATION NO. LM-0645 IREV. NO. I PAGE PAGE NO.

NO.10IO lI I CALCULATION NO. LM-0645 I REV. NO. I Table 2.2: Conformance with RG 1.183 Appendix B (Fuel Handling Accident) -_ -_- _-

RG LGS'Analysis -

Section RG Position - - Comments 5.1 If the containment is isolated during fuel handling operations, no radiological Not Applicable Containment is not consequences need to be analyzed. isolated.

5.2 If the containment is open during fuel handling operations, but designed to Not Applicable Containment is not automatically isolate in the event of a fuel handling accident, the release isolated.

duration should be based on delays in radiation detection and completion of No credit is taken for containment isolation. If it can be shown that containment isolation occurs 'defense in depth U before radioactivity is released to the environment, no radiological actions.

consequences need to be analyzed.

5.3 If the containment is open during fuel handling operations (e.g., personnel air Conforms This 2-hour release lock or equipment hatch is open), the radioactive material that escapes from assumption is utilized.

the reactor cavity pool to the containment is released to the environment over a 2-hour time period.

5.4 A reduction in the amount of radioactive material released from the Not Applicable No credit is taken for containment by ESF filter systems may be taken into account provided that filtration of release from these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter the reactor building.

99-02. Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

5.5 Credit for dilution or mixing of the activity released from the reactor cavity by Not Applicable No credit is taken for natural or forced convection inside the containment may be considered on a dilution or mixing of the case-by-case basis. Such credit is generally limited to 50% of the containment activity released from free volume. This evaluation should consider the magnitude of the containment the reactor cavity.

volume and exhaust rate, the potential for bypass to the environment, the A 2-hour release location of exhaust plenums relative to the surface of the reactor cavity, assumption is utilized recirculation ventilation systems, and internal walls and floors that impede stream flow between the surface of the reactor cavity and the exhaust plenums.

I CALCULATION NO. LM-0645 I REV. NO. I I PAGE NO. 11 l

3. ASSUMPTIONS Assumptions and analyzed conditions regarding the fuel handling accident scenarios are provided below.

1. Movement of recently irradiated fuel will not occur less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the associated reactor shutdown, per LGS Technical Specification 3/4.9.4.

2. Fuel bundle peak burnup will not exceed the RG 1.183 footnotes 10 and 11 limit of 62 GWD/MTU.

3. For fuel exceeding a 54 GWD/MTU burnup, the maximum linear heat generation rate will not exceed the RG 1.183 footnote 11 limit of 6.3 kW/ft rod average power.

4. As shown in Attachment E, the bounding fuel damage assessment scenario associated with a drop over the reactor core is used. For this event the RG 1.183 DF value of 200 is conservative.

5. Spent fuel source terms are based on reactor core source terms as discussed in Attachment A.

6. Activity reaching the refuel floor airspace will essentially all be exhausted within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by using an artificially high exhaust rate. This also provides an allowance for uneven mixing in the refuel floor airspace. (See Section 2.4)

7. The release pathway is unfiltered through the worst case Reactor Building South Stack to the Control Room normal intake. (See Section 2.4)

8. No credit is taken for the operation of the CREFAS system during the FHA. Conservatively high intake and outflow are considered. (See Section 2.4)

9. As the LGS Control Room has no exterior walls or overlying structures that are less than 2 feet thick concrete, this is considered sufficient to eliminate separate consideration of the radiation shine from the external radioactive plume release.

I CALCULATION NO. LM-0645 I REV. NO. I I PAGENO. 12 l

4. DESIGN INPUT The design inputs used for this calculation are summarized in the following Table 4:

Parameters Applicablc to AST Fuel Handling Accident Dose Considerations for Limerick Generatin Station TABLE 4: FHA AST Analvsis Parameter or Method for AST Value Source Documents Limerick Generating Station Reactor Power 3527 MWth Caic. No. LM-312, Rev. 0 Damaged Rods 212 Ref. I Fuel Assembly 8x8 in a 62 fuel pin bundle Ref. I Configuration Peaking Factor 1.7 New bounding analysis value compared to 1.5 in Ref. I Allowable Fuel Burnup and Peak burnup less than 62 RG 1.183, Table 3 non-LOCA gap fractions GWD/MTU FHA Radionuclide Inventory From Attachment A of this See Attachment A Caic. for the 60 isotopes forming the standard RADTRAD library, with decay to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per Ref. I and Technical Specification 3 /

4.9.4, for fuel not recently Gap activities per R.G. 1.183. irradiated Underwater Noble Gases: I RG 1.183 Decontamination Factor Particulate (cesium and RG 1.183 rubidium): infinity Iodine: 200, conservative value RG 1.183 for drop over reactor well. (See Attachment E)

Dose Conversion Factors EPA Federal Guidance Reports Ref. 4 and 5 11 and 12 Offsite Dose Limit 6.3 rem TEDE for the duration RG 1.183 of the accident Control Room Dose Limit 5 rem TEDE for the duration of I OCFR50 App. A, GDC 19 the accident and IOCFR50.67 Secondary Containment Not credited RG 1.183 Automatic Isolation and Filtration Mitigation by CREF System Not credited RG 1.183 Normal Control Room Fresh An artificial bounding value I Air Make-up Rate and CR volume per minute is used Volume for conservatism and to allow for any unfiltered inleakage; therefore flow is assumed to be 126,000 cfm (actual design values are more than an order of magnitude lower).

I CALCULATION NO. LM-0645 I REV. NO. I . I PAGE NO. 13 I TABLE 4: FHA AST Analysis Parameter or Method for AST Value Source Documents Limerick Generating Station Volume 126,000 ft3 Ref. 14 Refuel Floor Normal Approximately 6 air changes Conservative value for Ventilation rate per hour and an artificial value calculation of 100 ft3 is used for simplicity.

This evacuates 99.9994% of all activity within 2-hours.

CR Release Point Basis Reactor Building South Stack Ref. 12 Dispersion Factors 0 - 2 hr 1.26E-03 sec/M 3 Ref. 6 EAB Release Point Basis and Reactor Building South Stack Ref. 12 Distance to EAB and 731 mn Dispersion Factors 0 - 2 hr 3.18E-04 sec/m 3 Ref. 6 LPZ Release Point Basis and Reactor Building South Stack Ref. 12 Distance to LPZ and 2043 mn Dispersion Factors 0 - 2 hr 1.15E-04 sec/m 3 Ref. 6

I CALCULATION NO. LM-0645 I REV. NO. I PAGE NO. 14 l

5. REFERENCES

1. Limerick Generating Station Units 1 & 2, UFSAR, Revision 12.

2. Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors", July 2000.

3. NUREG/CR-6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation", April1998, and Supplements 1, June 1999, and 2, October 2002.

4. Federal Guidance Report (FGR) No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion", 1988.

5. FGR No. 12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.

6. LGS Design Analysis LM-0641, "Calculation of Alternative Source Terms Onsite and Offsite x/0 Values", Rev. 0.

7. LGS Design Analysis LM-0312, "Impact of Power Rerate on Fuel Handling Accident Doses and Activities", Rev. 0.

8. G. Burley, "Evaluation of Fission Product Release and Transport", Staff Technical Paper, 1971.

9. NEDC-32868P, "GE14 Compliance With Amendment 22 of NEDE-2401 1-P-A (GESTAR II)", Rev. 1, September 2000.

10. NEDE-31152P, "General Electric Fuel Bundle Designs" February 1993.

11. GESTAR II, NEDE-2401 1-P-A-11-US, Refueling Accident Analysis.

12. Limerick Generating Station Drawing SIM-M-76, Sheet 2, "Reactor Enclosure and Refueling Area - HVAC (Unit 2)", Rev. A (applied to both Units).

13. NEDE-24011-P-A-14-US, General Electric Standard Application for Reactor Fuel, Licensing Topical Report, June 2000

14. LGS Calculation No. M-78-01, "Control Room Area - Room Volume", Rev. 6.

15. Drawing M-102, "General Arrangement Plan at El. 217'-0", Rev. 10.

16. Drawing M-107, 'General Arrangement Section A-A & B-B", Rev. 8.

17. LGS Design Analysis LM-0656, "Determine Failed Fuel Rods From Collision of Refuel Bridges", Rev. 0.

18. LGS Design Analysis LM-0641, 'Determine Impact of RCWP Hoist Drop on UFSAR Fuel Handling Accident", Rev. 0.

I CALCULATION NO. LM-0645 I REV. NO. I I PAGENO. 15 l

6. CALCULATIONS 6.1. New Dose Analysis Basis This calculation evaluates the radiological dose to an operator in the Control Room and a person at the EAB and LPZ locations following an FHA involving irradiated fuel that has decayed for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after shutdown. This analysis uses Alternative Source Term assumptions per guidance in RG 1.183. The RADTRAD v. 3.03 computer code was used for this Limerick Units 1 & 2 FHA calculation. Releases are treated as unfiltered (no SGTS) through the Reactor Building South Stack (a 99% SGTS filtered release through the North Stack is considered separately in another RADTRAD run to show it is not bounding).

The RADTRAD inputs are summarized below:

A. Compartments

1. Containment - This compartment represents the Reactor Building Air Space, into which fission products leaving the spent fuel pool are released.

a. Compartment type - Other - since it is not the environment or control room.

b. Volume - 1.OOOE+02 ft 3 - This is an artificial RB volume, used to simplify the evacuation of activity from the RB to the environment. An exhaust rate was tailored to this nominal volume in order to model 6 air changes per hour, which ensures that essentially all activity is released within the 2-hour period.

c. Source term fraction - 1.0

d. Compartment features - no compartment removal mechanisms selected.

2. Environment

a. Compartment type - Environment

3. Control Room

a. Compartment type - Control Room

b. Volume - 126,000 ft3 - Ventilated volume.

c. Source term fraction - 0.0

d. Compartment features -none selected B. Transfer Pathways

1. Filtered Flow, Leak to the Environment

a. From Compartment I - Containment

b. To Compartment 2 - Environment

c. Transfer mechanism - "Filter" selected

d. Filter Efficiency Panel - Flow rate - 10 cfm - This is an arbitrary value that was set to ensure the release of 99.9994% of the activity from the Reactor Building within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , based on the nominal value as set above. This flow rate corresponds to 6 air changes per hour.

e. Filter Efficiency Panel - The efficiency used is 0 to represent no SGTS filtation.

2. Environment to Control Room

a. From Compartment 2 - Environment

b. To Compartment 3 - Control Room

c. Transfer mechanism - "Filter" selected

d. Filter Efficiency Panel - 126,000 cfm - Artificially high CR intake flowrate of one air change per minute, to conservatively allow for any unfiltered inleakage, for the duration of the accident.

I CALCULATION NO. LM-0645 I REV. NO. 1i PAGE NO. 16

e. Filter Efficiency Panel - Filter efficiency is entered as 0.0% for all chemical forms of iodine, to show that no CREF is credited.

3. Control Room to Environment

a. From Compartment 3- Control Room

b. To Compartment 2 - Environment

c. Transfer mechanism - "Filter' selected

d. Filter Efficiency Panel -Flow rate - Flow rate - 126,000 cfm for the duration of the accident

e. Filter Efficiency Panel - Filter efficiency is entered as 100.0% iodine chemical for all time periods. This is the exit from the control room to the environment; the filtration prevents a double counting of the iodine release. Note that the noble gas release will still be re-circulated between the control room and the outside environment.

C. Dose Locations

1. Exclusion Area Boundary

a. In Compartment 2 - Environment

b. 1/Q - 3.18E-04 sec/M 3 - this is the 0-2 hr accident X/Q for LGS Reactor Building South Ventilation Stack. This value is entered from time 24-hours to the end of the accident.

c. Breathing Rate - 3.5E-04 m3/sec.

2. Low Population Zone

a. In Compartment 2 - Environment

b. xQ - 1.1 5E-04 sec/M 3 - this is the 0-2 hr accident 1IQ for LGS Reactor Building South Ventilation Stack. This value is entered from time 24-hours to the end of the accident.

c. Breathing Rate - 3.5E-04 m3/sec.

3. Control Room

a. In Compartment 3 - Control Room

b. x/Q - 1.26E-03 sec/M 3 - this is the 0-2 hr accident X/Q for LGS Reactor Building South Ventilation Stack. This value is entered from time 24-hours to the end of the accident.

c. Breathing Rate - 3.5E-04 m3/sec

d. Occupancy Factor- 1.0 - this the RG 1.183 value for the first day.

D. Source Term and Release Fraction Treatment

a. The "Limerick Generating Station AST Source Terms for FHA.nif' file [Attachment C]

reflects the LGS core activities with the 1-131 value multiplied by 1.6 and the Kr-85 value multiplied by 2.0. These changes are made so that the Reg. Guide 1.183, Table 3 differentiation in release fraction can be made.

b. The power level of 26.83 MW is per section 2.1 above and reflects the fraction of the core damaged and the radial peaking factor applied to that fuel.

c. The 24-hour delay time reflects the minimum time after shutdown that fuel movement is expected.

d. The file "Limerick Generating Station AST FHA.rft" [Attachment D] is designed to reflect gap activity fractions per Reg. Guide 1.183, Table 3, with the adjustment for the "niff file described above.

E. Dose Conversion Factors The default FGR-11 and FGR-12 dose conversion factors provided with RADTRAD are used.

I CALCULATION NO. LM-0645 REV. NO. I PAGE NO. 17 l 6.2. Margin Assessment for Other Previously Considered FHA Scenarios Additional FHA scenarios have been considered in Calculations LM-0656 and LM-0657 (Ref. 17 and 18). Ref. 17 considers a collision of 2 refuel bridges carrying 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (since irradiation) spent fuel over the Spent Fuel Pool, causing both spent fuel bundles to drop over the spent fuel racks. Ref. 18 considers a drop of 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (since irradiation) spent fuel over the open reactor vessel onto the core, combined with a simultaneous drop of a jib crane and its suspended load for a total drop load of 1000 pounds from the Reactor Cavity Work Platform (RCWP).

Both of these scenarios are addressed based on application of AST methodology using an assessment of available margin to the New Dose Analysis Basis (as described in Sections 2.1 and 7.0) in the Attachment G spreadsheet.

7.

SUMMARY

AND CONCLUSIONS The RADTRAD code was used to examine the effect of the alternative source term release on offsite and CR doses. Shown below are the results, as provided in the first RADTRAD run in Attachment B, as well as the dose acceptance criteria.

Location Dose (rem TEDE)

LIMITS CR 5.0; EAB & LPZ 6.3 EAB 1.52 LPZ 0.549 CR 4.47 The alternative of treating releases 99% SGTS filtered through the North Stack, with a X/a to the Control Room Intake of 6.88E-03 sec/M 3 , is considered in the second RADTRAD run in Attachment B, with non-bounding results 0.620 rem TEDE to the CR, 0.395 rem TEDE to the EAB, and 0.143 rem TEDE to the LPZ.

These results indicate that the calculated consequences of the postulated Fuel Handling Accident at or after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of shutdown will be within regulatory limits without the requirement of SGTS filtration. Furthermore, Control Room filtration is not required to maintain operator doses within regulatory limits. These results will bound any other fuel handling accident scenarios that are:

- based on the GESTAR II damage analysis assessment methodology of Attachment E,

- have a Damaged Core Fraction with PF (as derived in Section 2.1) up to 0.007608, for either:

o an overall effective DF of 200 (for 23 feet minimum water coverage), or o for an assumed drop over the spent fuel storage racks with less than 23 feet water coverage, the calculated number of damaged rods is adjusted by the factor of [200 divided by the appropriate DF for the water coverage over the damaged rods]

(pages E-5 through E-8 of Attachment E provides an example of such spent fuel pool DF consideration).

As examples, Attachment G shows that both of the scenarios addressed in Section 6.2 are bounded by the margin to the New Dose Analysis Basis described in Sections 2.1, and establishes the limiting number of damaged rods in the struck bundles for each scenario.

CALCULATION NO. LM-0645 l REV. NO. I lPAGE NO. 18l

8. OWNER'S ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS DESIGN ANALYSIS NO. LM-0645 REV: I Yes No N/A

1. Do assumptions have sufficient rationale? So So

2. Are assumptions compatible with the way the plant is operated and with the -

licensing basis? 6 APST')

J;17 0

3. Do the design inputs have sufficient rationale? El

4. Are design inputs correct and reasonable? 0O Are design inputs compatible with the way the plant is operated and with the licensing basis? ( AST)

6. Are Engineering Judgments clearly documented and justified?

7. Are Engineering Judgments compatible with the way the plant is operated and 0 with the licensing basis? (A. JST) 0< S3 So Do the results and conclusions satisfy the purpose and objective of the Design

8. Analysis? EW E Are the results and conclusions co patiblei ith the way the plant is operated 9* and with the licensing basis? C ST) ESt 0 0

10. Does the Design Analysis include the applicable design basis documentation?

I1 Have any limitations on the use of the results been identified and transmitted B;K El M to the appropriate organizations?

el0 Ho

12. Are there any unverified assumptions?

El 2- E

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

14. Have all affected design analyses been documented on the Affected Documents List (ADL) for the associated Configuration Change? LV/ M 0o Do the sources of inputs and analysis methodology used meet current technical requirements and regulatory commitments? (If the input sources or

15. analysis methodology are based on an out-of-date methodology or code, additional reconciliation may be required if the site has since committed to a more recent code)

16. Have vendor supporting technical documents and references (including GE DRFs) been reviewed when necessary?

EXELON REVIEWER: f J, 4 ts? DATE: 7uPz9 /o-C' nrinn

ATTACHMENT A Source Terms (The following Source Terms derivation is from thle Peach Bottom Atomiic Power Station (PBAPS) AST FHA CalculationPM-1059. The results are also applicable to LGS, as conimrmed by Exelon on the last tvo pages of this Attachment]

Introduction The following description and calculation input and results from Robert Jaffa of Exelon Nuclear derives the isotopic inventory (core source terms) to be used as inputs for the PBAPS AST Design Basis Accident analyses. The derivation uses an Exelon Nuclear controlled-version of ORIGEN 2.1, as noted. Although derived for PB Unit 3, they are considered as applicable to PB Unit 2 as well.

The core source terms are initially calculated based on a power level of 3514.9 MWt, which is approximately the Rated Thermal Power of 3514 MWt. The AST analyses utilize a Design Basis Accident power level of 3528 MWt. RADTRAD uses a NIF file in units of Curies/MWt and multiplies this by the specified core power, so the derived Curie quantities for the 60 RADTRAD isotopes in the following Table 2 are divided by 3514.9 to create the NIF file as presented in Attachment C, and 3528 MWt is used as the input specified core power for full core power analyses.

Caic. No. LM-0645, Rev. l, Attachment A, Page A- I of A-15

1.0 PURPOSE To calculate the isotopic core inventory for PB Unit 3 using the ORIGEN2.1 code based on reactor operation at 3514.9 MWt and an equilibrium 711 EFPD two-year cycle design.

2.0

SUMMARY

OF RESULTS The bounding isotopic core inventory for a 711 EFPD two-year cycle design at PB Unit 3 is shown in terms of activity (Curies) and concentration (grams) in Tables 2 and 4, respectively.

This bounding isotopic core inventory was determined for PB Unit 3's expected power uprate power level of 3514.9 MWt, which is approximately equal to the rated thermal power of 3514 MWt.

3.0 REFERENCES

3.1 RSIC Code Package CCC-371, "ORIGEN 2.1, Isotope Generation and Depletion Code Matrix Exponential Method," May 1999.

3.2 Exelon Nuclear memo NFM-MA:03-002 from J. Wolfrom to R. Jaffa, Re: "Peach Bottom Alternate Source Term Information," dated January 9, 2003.

3.3 Memo from R. Jaffa to R. Tropasso, Re: "Installation Verification of ORIGEN2.1 on the IBM PC Platform," dated July 7, 2000.

3.4 ORNfTM-1 1018, "Standard- and Extended-Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer Code," S. Ludwig, J. Renier, December 1989.

4.0 ASSUMPTIONS 4.1 The ORIGEN2.1 code [Ref 3.1] was used to calculate isotopic activities based on the cycle design described in Reference 3.2. The ORIGEN2.1 code was run on an IBM-PC and was confirmed to be controlled per Ref. 3.3.

4.2 Batch-average enrichments and exposures from Ref. 3.2 were used to develop input to ORIGEN2. 1. This is equivalent to performing individual calculations for each sub-batch.

4.3 For fuel burned in more than one cycle, ORIGEN2.1 runs ignored refueling outages.

This has no impact on short-lived isotopes which reach equilibrium concentrations shortly after cycle startup and has a conservative, albeit minimal, impact on long-lived isotopes which continually increase in concentration as a function of exposure; ignoring intermediate decay periods will increase the final concentrations.

4.4 For isotopic activities calculated at BOC, BOC is defined as 100 days into the cycle to ensure that all short-lived isotopes (half lives < I year) are at equilibrium levels.

Caic. No. LM-0645, Rev. I, Attachment A, Page A-2 of A-15

5.0 DESIGN INPUT 5.1 Batch-average information for an equilibrium two-year cycle (Ref. 3.2) was obtained using the FCYCLEOI code starting from the PB3 Cycle 14 design.

5.2 The ORIGEN2.1 cross-section library, BWRUE.LIB, is used in this calculation as this is most representative of current PB3 two-year cycles. The library is based on an "extended cycle" reactor model where fuel achieves 40 GWd/mtU burnup in four cycles.

5.3 Equilibrium cycle lengths are 711 EFPD.

5.4 The power level used for determining exposure data for the equilibrium two-year cycle is 3514.9 MWt, which is approximately the Rated Thermal Power level for PB3 of 3514 MWt after the 1.62% Caldon power uprate.

6.0 OVERALL APPROACH AND METHODOLOGY The isotopic core inventory is a function of the reactor power level and the exposure of the fuel.

The PB Unit 3 isotopic core inventory is calculated using the ORIGEN2.1 code [Ref. 3.1]. A 3514.9 MWt 711 EFPD two-year equilibrium cycle design [from Ref. 3.2] is used as the basis for this calculation 6.1 Computer Code Information The batch file used to execute ORIGEN2.1 (DTSQA Product ID# EX0004724) for this calculation provide the paths and filenames of the executable program and libraries that were called. The batch file used is pb3 ast.bat. The PC-based ORIGEN2.1 code used in this calculation was verified to be controlled by comparing the time/date/size stamp of the executable file to that documented in Ref. 3.3.

Volume in drive D is PER30290 Volume Serial Number is I IF7-OCIF Directory of D:\Origen21\CODE ORIGEN2 EXE 1,267,348 06-10-96 1:09p ORIGEN2.EXE The time/date/size stamps of the library files used in this calculation were verified against those documented in Ref. 3.4.

Volume in drive D is PER30290 Volume Serial Number is I 1F7-OCIF Directory of D:\Origen2l\LIBS BWRUE LIB 173,676 08-01-91 2:1Oa DECAY LIB 278,636 08-01-91 2:10a GXUO2BRM LIB 167,526 08-01-91 2:10a Calc. No. LM-0645, Rev. 1,Attachment A,Page A-3 of A-15

7.0 CALCULATION Equilibrium Twvo-Year Cycle Isotopic Core Inventory An equilibrium two-year cycle design for Peach Bottom Unit 3 based on the FCYCLEOI code using the actual Cycle 14 design as a reference cycle and a cycle length of 711 EFPD is used as the basis for the source term calculation. In addition, batch average burnups were increased to account for a 1.62%

Caldon power uprate. The resulting batch-average burnups for once-burned, twice-burned and thrice-burned fuel batches are shown below.

l Avg. Bumup Avg. Enr. per Cyclc Power Loading U235 wt U238 wt. Oxygen wt.

Batch of FA (w/o U235) (MWd/mtU) (MW1) (MTU) (gms) (gMs) (grns)

- 208 4.105 22800.49 1197.0 37.3256 1,532,215.88 35,793,384.12 5,018,379.87 18256.15 958.4 12116.92 636.1 2 280 4.107 22800.49 1611.5 50.2516 2.063,833.21 48,187,766.79 6,756,264.28 18256.15 1290.3 3 276 4.107 22800.49 1588.5 49.53372 2,034,349.88 47.499,370.12 6.659,746.22 The equilibrium cycle isotopic core inventory is calculated using ORIGEN2.1 and the BWR extended burnup cross-section library BWRUE. The input deck is pb3 ast.inp and the batch file is pb3_ast.bat.

The specific power for a batch in a given cycle is determined by multiplying the batch average burnup for that cycle by the batch loading and then dividing by the number of EFPD in the cycle. For example, the specific power for Batch I in its first cycle of operation is:

(22,800.49

37.3256) / 711 = 1197.0 MW.

The grams of U235 and U238 for each batch were determined by the following formulas:

U235 (gms) = Batch loading * (Avg. Enr./100)

106 U238 (gms) = Batch loading * (I - Avg. Enr./100)

106 The corresponding weight of oxygen in U0 2 pellets for each batch is:

0 (gms) = Total batch U weight (gm U) / 238 (gm U/gm atom U)

2 (gm atom O/gm atom U)

15.9994 gm O/gm atom 0 The ORIGEN2.1 input deck is set up to deplete each fuel batch and write the 100 EFPD and EOC results to temporary storage vectors. Once all batches have been depleted, the results from the temporary vectors are combined to give the results for the entire core. The ORIGEN2.1 core inventory activity and composition results for the equilibrium two-year cycle at 100 EFPD (BOC) and EOC are shown below in Calc. No. LM-0645, Rev. 1,Attachment A,Page A-4 of A-15

Tables I and 3, respectively. The maximum of the 100 EFPD and EOC values for each isotope are selected to generate the bounding isotopic core inventory activity and composition results as shown in Tables 2 and 4, respectively.

Calc. No. LM-0645, Rev. I, Attachment A, Page A-5 of A-15

Table 1 ORIGEN2.1 Isotopic Activity Results for Peach Bottom Unit 3 ll 100 EFPD EOC ll 100 EFPD l EOC Isotope (Ci) I (Ci) Isotope l C) l (Ci)_

KR 83M 1.324E+07 1.158E+07 XE131M 1.015E+06 1.056E+06 BR 84 2.373E+07 2.002E+07 TE132 1.322E+08 1.343E+08 BR 85 2.888E+07 2.404E+07 1132 1.338E+08 1.364E+08 KR 85 8.806E+05 1.387E+06 1133 1.953E+08 1.925E+08 KR 85M 2.922E+07 2.436E+07 XE133 1.904E+08 1.930E+08 RB 86 1.118E+05 2.291E+05 XE133M 5.956E+06 6.007E+06 KR 87 5.739E+07 4.672E+07 1134 2.167E+08 2.118E+08 KR 88 8.096E+07 6.570E+07 CS134 1.335E+07 2.559E+07 RB 88 8.197E+07 6.678E+07 1135 1.825E+08 1.806E+08 SR 89 9.836E+07 8.846E+07 XE135 7.832E+07 7.086E+07 SR 90 6.982E+06 1.117E+07 XE135M 3.636E+07 3.773E+07 Y 90 7.142E+06 1.150E+07 CS136 3.568E+06 7.123E+06 SR 91 1.336E+08 1.110E+08 CS137 9.460E+06 1.595E+07 Y 91 1.212E+08 1.143E+08 BA137M 8.965E+06 1.510E+07 SR 92 1.412E+08 1.205E+08 XE138 1.679E+08 1.589E+08 Y 92 1.416E+08 1.210E+08 CS138 1.841E+08 1.760E+08 Y 93 1.591 E+08 1.404E+08 BA139 1.787E+08 1.719E+08 ZR 95 1.521E+08 1.578E+08 BA140 1.721 E+08 1.661 E+08 NB 95 1.399E+08 1.586E+08 LA140 1.764E+08 1.722E+08 ZR 97 1.637E+08 1.578E+08 LA141 1.631 E+08 1.563E+08 MO 99 1.771E+08 1.785E+08 CE141 1.579E+08 1.575E+08 TC 99M 1.551E+08 1.563E+08 LA142 1.593E+08 1.510E+08 RU103 1.234E+08 1.477E+08 CE143 1.556E+08 1.449E+08 RU105 7.642E+07 1.022E+08 PR143 1.509E+08 1.416E+08 RH105 7.31 OE+07 9.673E+07 CE144 1.012E+08 1.264E+08 RU106 3.856E+07 6.081E+07 ND147 6.459E+07 6.321 E+07 SB127 8.572E+06 1.018E+07 NP239 1.616E+09 1.897E+09 TE127 8.402E+06 1.010E+07 PU238 2.639E+05 6.312E+05 TE127M 1.039E+06 1.355E+06 PU239 3.100E+04 4.218E+04 SB129 2.732E+07 3.036E+07 PU240 2.871 E+04 4.526E+04 TE129 2.679E+07 2.988E+07 PU241 1.365E+07 2.173E+07 TE129M 3.875E+06 4.453E+06 AM241 1.634E+04 3.349E+04 1129 2.774E+00 4.816E+00 CM242 3.645E+06 8.393E+06 TE131M 1.269E+07 1.360E+07 CM244 2.654E+05 9.147E+05 1131 9.139E+07 9.444E+07 Calc. No. LM-0645, Rev. 1, Attachment A, Page A-6 of A-15

Table 2 Bounding Isotopic Core Inventory Peach Bottom Unit 3 Isotopic Isotopic Activity Activity Isotope (Ci) Isotope (Ci) 1.324E+07 XE131M 1.056E+06 KR 83M BR84 2.373E+07 TE132 1.343E+08 BR 85 2.888E+07 1132 1.364E+08 KR 85 1.387E+06 1133 1.953E+08 KR 85M 2.922E+07 XE1 33 1.930E+08 RB 86 2.291 E+05 XE133M 6.007E+06 KR 87 5.739E+07 1134 2.1 67E+08 KR 88 8.096E+07 CS1 34 2.559E+07 RB 88 8.1 97E+07 1135 1.825E+08 SR89 9.836E+07 XE1 35 7.832E+07 SR 90 1.117E+07 XE135M 3.773E+07 Y90 1.1 50E+07 CS136 7.1 23E+06 SR 91 1.336E+08 CS137 1.595E+07 Y91 1.21 2E+08 BA137M 1.51 OE+07 SR 92 1.41 2E+08 XE1 38 1.679E+08 Y92 1.41 6E+08 CS138 1.841 E+08 Y93 1.591 E+08 BA1 39 1.787E+08 ZR 95 1.578E+08 BA140 1.721 E+08 NB 95 1.586E+08 LA1 40 1.764E+08 ZR 97 1.637E+08 LA141 1.631 E+08 MO 99 1.785E+08 CE141 1.579E+08 TC 99M 1.563E+08 LA142 1.593E+08 RU103 1.477E+08 CE143 1.556E+08 RU105 1.022E+08 PR143 1.509E+08 RH105 9.673E+07 CE144 1.264E+08 RU106 6.081 E+07 ND147 6.459E+07 SB127 1.01 8E+07 NP239 1.897E+09 TE127 1.01 OE+07 PU238 6.31 2E+05 TE1 27M 1.355E+06 PU239 4.21 8E+04 SB129 3.036E+07 PU240 4.526E+04 TE129 2.988E+07 PU241 2.173E+07 TE1 29M 4.453E+06 AM241 3.349E+04 1129 4.81 6E+00 CM242 8.393E+06 TE1 31 M 1.360E+07 CM244 9.1 47E+05 1131 9.444E+07 Ca1c. No. LM-0645, Rev. 1, Attachmcnt A, Page A-7 of A-lS

Table 3 ORIGEN2.1 Isotopic Concentration Results for Peach Bottom Unit 3 ll 100 EFPD I EOC l 100 EFPD I EOC Isotope !! (grams) (grams) Isotope rams (grams)

KR83M 6.414E-01 5.611E-01 XE131M 1.212E+01 1.260E+01 BR84 3.370E-01 2.843E-01 TE132 4.352E+02 4.423E+02 BR85 3.741 E-02 3.114E-02 1132 1.296E+01 1.321E+01 KR85 2.244E+03 3.534E+03 1133 1.723E+02 1.699E+02 KR85M 3.550E+00 2.959E+00 XE133 1.017E+03 1.031 E+03 RB 86 1.373E+00 2.814E+00 XE133M 1.328E+01 1.339E+01 KR 87 2.025E+00 1.649E+00 CS133 1.025E+05 1.678E+05 KR88 6.451 E+00 5.235E+00 1134 8.118E+00 7.936E+00 RB 88 6.826E-01 5.561E-01 CS134 1.031 E+04 1.977E+04 SR89 3.384E+03 3.044E+03 1135 5.195E+01 5.140E+01 SR90 5.116E+04 8.183E+04 XE135 3.065E+01 2.773E+01 Y90 1.312E+01 2.113E+01 XE135M 3.990E-01 4.140E-01 SR91 3.683E+01 3.060E+01 CS135 4.502E+04 7.841 E+04 Y91 4.939E+03 4.658E+03 CS136 4.867E+01 9.715E+01 SR92 1.123E+01 9.579E+00 CS137 1.087E+05 1.832E+05 Y 92 1.471E+01 1.256E+01 BA137M 1.666E-02 2.807E-02 Y 93 4.768E+01 4.206E+01 XE138 1.745E+00 1.652E+00 ZR95 7.079E+03 7.341 E+03 CS138 4.348E+00 4.156E+00 NB 95 3.576E+03 4.054E+03 BA139 1.092E+01 1.051 E+01 ZR 97 8.560E+01 8.251E+01 BA140 2.359E+03 2.277E+03 MO 99 3.691 E+02 3.720E+02 LA140 3.168E+02 3.093E+02 TC99M 2.947E+01 2.971E+01 LA141 2.883E+01 2.762E+01 RU103 3.822E+03 4.575E+03 CE141 5.541 E+03 5.528E+03 RU105 1.136E+01 1.519E+01 LA142 1.115E+01 1.056E+01 RH105 8.657E+01 1.146E+02 CE143 2.342E+02 2.181 E+02 RU106 1.152E+04 1.817E+04 PR143 2.241 E+03 2.102E+03 SB127 3.208E+01 3.811E+01 CE144 3.172E+04 3.962E+04 TE127 3.182E+00 3.826E+00 ND147 8.039E+02 7.867E+02 TE127M 1.101E+02 1.436E+02 NP239 6.963E+03 8.173E+03 1127 4.533E+03 8.040E+03 PU238 1.541E+04 3.685E+04 SB129 4.856E+00 5.397E+00 PU239 4.985E+05 6.782E+05 TE129 1.279E+00 1.426E+00 PU240 1.259E+05 1.986E+05 TE129M 1.286E+02 1.478E+02 PU241 1.324E+05 2.108E+05 1129 1.571E+04 2.727E+04 AM241 4.759E+03 9.755E+03 TE131M 1.591 E+01 1.704E+01 CM242 1.102E+03 2.537E+03 1131 7.369E+02 7.615E+02 CM244 3.279E+03 1.130E+04 Calc. No. LM-0645, Rev. I, Attachment A, Page A-8 of A- 15

Table 4 Bounding Isotopic Core Inventory Peach Bottom Unit 3 Isotopic Isotopic Concentration Concentration Isotope (grams) Isotope (grams)

KR 83M 6.414E-01 XE1 31 M 1.260E+01 BR84 3.370E-01 TE132 4.423E+02 BR85 3.741 E-02 1132 1.321 E+01 KR85 3.534E+03 1133 1.723E+02 KR 85M 3.550E+00 XE1 33 1.031 E+03 RB 86 2.814E+00 XE133M 1.339E+01 KR 87 2.025E+00 CS133 1.678E+05 KR 88 6.451 E+00 1134 8.118E+00 RB 88 6.826E-01 CS134 1.977E+04 SR89 3.384E+03 1135 5.1 95E+01 SR90 8.1 83E+04 XE1 35 3.065E+01 Y90 2.11 3E+01 XE1 35M 4.1 40E-01 SR 91 3.683E+01 CS135 7.841 E+04 Y91 4.939E+03 CS136 9.71 5E+01 SR 92 1.123E+01 CS137 1.832E+05 Y92 1.471 E+01 BA1 37M 2.807E-02 Y93 4.768E+01 XE138 1.745E+00 ZR 95 7.341 E+03 CS138 4.348E+00 NB 95 4.054E+03 BA1 39 1.092E+01 ZR 97 8.560E+01 BA140 2.359E+03 MO 99 3.720E+02 LA1 40 3.1 68E+02 TC 99M 2.971 E+01 LA1 41 2.883E+01 RU103 4.575E+03 CE141 5.541 E+03 RU105 1.51 9E+01 LA142 1.115E+01 RH105 1.146E+02 CE143 2.342E+02 RU106 1.81 7E+04 PR143 2.241 E+03 SB127 3.811 E+01 CE144 3.962E+04 TE127 3.826E+00 ND147 8.039E+02 TE127M 1.436E+02 NP239 8.1 73E+03 1127 8.040E+03 PU238 3.685E+04 SB129 5.397E+00 PU239 6.782E+05 TE129 1.426E+00 PU240 1.986E+05 TE129M 1.478E+02 PU241 2.1 08E+05 1129 2.727E+04 AM241 9.755E+03 TE131M 1.704E+01 CM242 2.537E+03 1131 7.61 5E+02 CM244 1.1 30E+04 Cale. No. LM-0645, Rev. 1,Attachment A, Page A-9 of A-15

Input Deck pb3_ast.inp

-1

-1 BAS Grams of Heavy Metal per Fuel Batch RDA PLACE FUEL into vectors -1, -2 and -3 LIP 0 0 0 LIB 0 1 2 3 657 658 659 9 3 0 1 42 PHO 0 0 0 10 RDA READ FUEL COMPOSITION FOR BATCH 3 INP -1 1 1 1 1 RDA READ FUEL COMPOSITION FOR BATCH 2 INP -2 1 1 1 1 RDA READ FUEL COMPOSITION FOR BATCH 1 INP -3 1 1 1 1 RDA TIT IRRADIATION OF TMI-1 CYCLE 1 FULL CORE MOV -3 1 0 1.0 BATCH 1 FRESH HED 1 CHARGE RDA BATCH 1 BURNUP IN CYCLE 1 BUP IRP 50.0 1197.0 1 9 4 2 IRP 100 .0 1197.0 9 2 4 0 IRP 150.0 1197.0 2 9 4 0 IRP 200.0 1197.0 9 3 4 0 IRP 250.0 1197.0 3 9 4 0 IRP 300.0 1197.0 9 4 4 0 IRP 350.0 1197.0 4 9 4 0 IRP 400.0 1197.0 9 5 4 0 IRP 450.0 1197.0 5 9 4 0 IRP 500.0 1197.0 9 6 4 0 IRP 550.0 1197.0 6 9 4 0 IRP 600.0 1197.0 9 7 4 0 IRP 650.0 1197.0 7 9 4 0 IRP 711.0 1197.0 9 8 4 0 BUP OPTL 4*8 5 8 5 17*8 OPTA 4*8 5 8 5 17*8 OPTF 4*8 5 8 5 17*8 OUT -8 1 -1 0 MOV 8 1 0 1.0 BATCH 1 ONCE BURNED HED 1 CHARGE RDA BATCH 1 BURNUP IN CYCLE 2 BUP IRP 761.0 958.4 1 9 4 3 IRP 811.0 958.4 9 2 4 0 IRP 861.0 958.4 2 9 4 0 IRP 911.0 958.4 9 3 4 0 IRP 961.0 958.4 3 9 4 0 IRP 1011.0 958.4 9 4 4 0 IRP 1061.0 958.4 4 9 4 0 IRP 1111.0 958.4 9 5 4 0 IRP 1161.0 958.4 5 9 4 0 IRP 1211.0 958.4 9 6 4 0 IRP 1261.0 958.4 6 9 4 0 IRP 1311.0 958.4 9 7 4 0 IRP 1361.0 958.4 7 9 4 0 Cale. No. LM-0645, Rev. 1, Attachment A, Page A-10 of A-15

IRP 1422.0 958.4 9 8 4 0 BUP OUT -8 1 -1 0 MOV 8 1 0 1.0 BATCH 1 TWICE BURNED HED 1 CHARGE RDA BATCH 1 BURNUP IN CYCLE 3 BUP IRP 1472.0 636.1 1 9 43 IRP 1522.0 636.1 9 2 40 IRP 1572.0 636.1 2 9 40 IRP 1622.0 636.1 9 3 40 IRP 1672.0 636.1 3 9 40 IRP 1722.0 636.1 9 4 40 IRP 1772.0 636.1 4 9 40 IRP 1822.0 636.1 9 5 40 IRP 1872.0 636.1 5 9 40 IRP 1922.0 636.1 9 6 40 IRP 1972.0 636.1 6 9 40 IRP 2022.0 636.1 9 7 40 IRP 2072.0 636.1 7 9 40 IRP 2133.0 636.1 9 8 40 BUP OUT -8 1 -1 0 MOV 2 -9 0 1.0 BATCH 1 100 EFPD PLACED IN TEMP VECTOR -9 MOV 8 -10 0 1.0 BATCH 1 EOC3 PLACED IN TEMP VECTOR -10 RDA BATCH 2 BURNUP IN CYCLE 2 MOV -2 1 0 1.0 BATCH 2 FRESH HED 1 CHARGE RDA BATCH 2 BURNUP IN CYCLE 2 BUP IRP 50.0 1611.5 1 9 4 2 IRP 100. 0 1611.5 9 2 4 0 IRP 150.0 1611.5 2 9 4 0 IRP 200.0 1611.5 9 3 4 0 IRP 250.0 1611.5 3 9 4 0 IRP 300.0 1611.5 9 4 4 0 IRP 350.0 1611.5 4 9 4 0 IRP 400.0 1611.5 9 5 4 0 IRP 450.0 1611.5 5 9 4 0 IRP 500.0 1611.5 9 6 4 0 IRP 550.0 1611.5 6 9 4 0 IRP 600.0 1611.5 9 7 4 0 IRP 650.0 1611.5 7 9 4 0 IRP 711.0 1611.5 9 8 4 0 BUP OUT -8 1 -1 0 MOV 8 1 0 1.0 BATCH 2 ONCE BURNED HED 1 CHARGE RDA BATCH 2 BURNUP IN CYCLE 3 BUP IRP 761.0 1290.3 1 9 4 3 IRP 811.0 1290.3 9 2 4 0 IRP 861.0 1290.3 2 9 4 0 IRP 911. 0 1290.3 9 3 4 0 IRP 961.0 1290.3 3 9 4 0 IRP 1011. 0 1290.3 9 4 4 0 IRP 10 61. 0 1290.3 4 9 4 0 Cale. No. LM-0645, Rev. 1, Attachment A, Page A- Il of A-15

IRP 1111. 0 1290.3 9 5 40 IRP 1161.0 1290.3 5 9 40 IRP 1211.0 1290.3 9 6 40 IRP 1261.0 1290.3 6 9 40 IRP 1311.0 1290.3 9 7 40 IRP 1361.0 1290.3 7 9 40 IRP 1422.0 1290.3 9 8 40 BUP OUT -8 1 -1 0 ADD 2 -9 0 1.0 BATCH 2 100 EFPD ADDED TO TEMP VECTOR -9 ADD 8 -10 0 1.0 BATCH 2 EOC3 ADDED TO TEMP VECTOR -10 MOV -1 1 0 1.0 BATCH 3 FRESH HED 1 CHARGE RDA BATCH 3 BURNUP IN CYCLE 3 BUP IRP 50.0 1588.5 1 9 4 2 IRP 100. 0 1588.5 9 2 4 0 IRP 150.0 1588.5 2 9 4 0 IRP 200.0 1588.5 9 3 4 0 IRP 250.0 1588.5 3 9 4 0 IRP 300.0 1588.5 9 4 4 0 IRP 350.0 1588.5 4 9 4 0 IRP 400.0 1588.5 9 5 4 0 IRP 450.0 1588.5 5 9 4 0 IRP 500.0 1588.5 9 6 4 0 IRP 550.0 1588.5 6 9 4 0 IRP 600.0 1588.5 9 7 4 0 IRP 650.0 1588.5 7 9 4 0 IRP 711.0 1588.5 9 8 4 0 BUP OUT -8 1 -1 0 ADD 2 -9 0 1.0 ADD 8 -10 0 1.0 MOV -9 1 0 1.0 CYCLE 3 @ 100 EFPD MOV -10 2 0 1.0 CYCLE 3 @ EOC HED 1 100 EFPD HED 2 EOC OUT -2 1 -1 0 END 2 !922350 2034349.88 922380 47499370.12 0 0.0 U02 4 ()80000 6659746.22 0 0.0 U02 0

2 t922350 2063833.21 922380 48187766.79 0 0.0 U02 4 t)80000 6756264.28 0 0.0 U02 0

2 t922350 1532215.88 922380 35793384.12 0 0.0 U02 4 t)80000 5018379.87 0 0.0 U02 0

END Calc. No. LM-0645, Rev. 1, Attachment A, Page A-12 of A-15

Job Batch File pb3_ast.bat echo off echo ****************************

echo ****************************

echo **

echo ** O R I G E N 2 **

echo ** Oak Ridge Isotope GENeration and Depletion Code **

echo ** Version 2.1 (8-1-91) **

echo **

echo ***********************************************************************

echo ** **

echo ** Developed by: Oak Ridge National Laboratory **

echo ** Chemical Technology Division **

echo **

echo ** Technical

Contact:

Scott B. Ludwig **

echo ** (615) 574-7916 FTS 624-7916 **

echo **

echo ** Distributed by: Radiation Shielding Information Center (RSIC) **

echo ** Oak Ridge National Laboratory **

echo ** P.O. Box 2008 **

echo ** Oak Ridge, TN 37831 **

echo ** (615) 574-6176 FTS 624-6176 **

echo ***********************************************************************

pause echo ** Execution continuing ...

echo ********************************************************

echo ***********************************************************

echo **

echo ** Version 2.1 (8-1-91) for mainframes and 80386 or 80486 PCs **

echo ** **

copy pb3_ast.inp tape5.inp >nul REM (NOT USED IN THIS CASE) copy samp_2.u3 tape3.inp >nul copy \origen2l\libs\decay.lib+\origen2l\libs\bwrue.lib tape9.inp >nul copy \origen2l\libs\gxuo2brm.lib tapelO.inp >nul

\origen2l\code\origen2 rem combine and save files from run copy tapel2.out+tape6.out pb3_ast.u6 >nul copy tapel3.out+tapell.out pb3_ast.ull >nul ren tape7.out pb3 ast.pch ren tapel5.out pb3_ast.dbg ren tapel6.out pb3 ast.vxs ren tape5O.out pb3_ast.ech rem cleanup files del tape*.inp del tape*.out echo ***********************************************************

echo ******************* OR I G E N 2 - Version 2.1

echo *********************** Execution Completed ***************************

echo **********************************************************

echo on Calc. No. LM-0645, Rev. 1, Attachment A, Page A-13 of A-15

Original Message-----

From: Mscisz, Thomas J. [1]

Sent: Wednesday, October 22, 2003 9:53 AM To: 'paul.reichert@wgint.com'; Reichert, P.T.

Subject:

FW: LGS Source Term Information

Original Message-----

> From: Jaffa, Robert P.

> Sent: Monday, October 20, 2003 3:40 PM

> To: Mscisz, Thomas J.

> Cc: Tusar, James J.

Subject:

RE: LGS Source Term Information

> Tom,

> The attached email indicates that the equilibrium two-year cycle used to

> generate the source term for the Peach Bottom Unit 3 AST analysis will be

> bounding for the cycle designs currently being loaded in the Limerick

> Generating Station (LGS). Since LGS rated power is 1.6% lower than that

> assumed for the PB-3 AST source term, isotopic activities (short and

> long-lived) would be -1.6% lower for the LGS source term (assuming similar

> cycle designs).

> LGS has the same core loading as PB-3 (in terms of mtU of uranium).

> Design cycle lengths for LGS are bounded by the 711 EFPD assumed for the

> PB-3 source term; shorter cycle lengths would result in reduced activities

> from long-lived isotopes. Since the batch sizes, enrichments, and burnup

> distributions are similar between PB-3 and LGS (as noted below), the PB-3

> AST source term would be bounding for current LGS cycle designs.

> Bob

Original

- Message-----

> From: Tusar, James J.

> Sent: Monday, October 20, 2003 10:45 AM

> To: Jaffa, Robert P.

> Cc: Mscisz, Thomas J.

Subject:

LGS Source Term Information

> Action Required: Review Peach Bottom Source Term fuel cycle

> data/assumptions for applicability to the LGS Source Term Calculation

> Recommendation: Forward conclusion to Tom Mscisz

> Bob:

> The Peach Bottom Source Term fuel cycle assumptions are considered

> bounding relative to Limerick Generating Station. The Peach Bottom batch

> sizes, batch average burnups, and reload bundle average enrichments are

> similar to those at Limerick. For example, the bundle average enrichment

> used in the Peach Bottom analysis was approximately 4.11 wt. % U-235. The

> expected bundle average enrichment for LGS is approximately 4.16 wt. %

> U-235. Additionally, the rated thermal power level for the Peach Bottom

> analyses was 3514 MWt (uprated for Appendix K Thermal Power Optimization).

Calc. No. LM-0645, Rev. 1, Attachment A, Page A-14 of A-15

> This is bounding relative to Limerick's 3458 MWt rated thermal power by

> 1.62%. Therefore, the Peach Bottom Source Term fuel cycle data should be

> applicable to the Limerick Source Term calculations. Let me know if you

> have any other questions.

> James J. Tusar, PE

> Manager, BWR Design

> 200 Exelon Way, KSA 2-N

> Kennett Square, PA 19348

> james.tusar@exeloncorp.com

> 610.765.5818 (voice)

> 610.765.5651 (fax)

This e-mail and any of its attachments may contain Exelon Corporation proprietary information, which is privileged, confidential, or subject to copyright belonging to the Exelon Corporation family of Companies.

This e-mail is intended solely for the use of the individual or entity to which it is addressed. If you are not the intended recipient of this e-mail, you are hereby notified that any dissemination, distribution, copying, or action taken in relation to the contents of and attachments to this e-mail is strictly prohibited and may be unlawful. If you have received this e-mail in error, please notify the sender immediately and permanently delete the original and any copy of this e-mail and any printout. Thank You.

Cale. No. LM-0645, Rev. I, Attachment A, Page A- 15 of A- 15

Limerick FHA 24 hr Delay - No Filter Credit 4141441414141##4#4###########414##4#41#4141########4#$#41###$ ###4#41###411#4141414#####

RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:48:13 41#########4141#414141#################4$######41#44#111#1111#11#4111444########

File information Plant file P:\Users\Nuc\Exelon EOC\Discipline Files\Process\AST\Limerick AST\LGS FHA\RADTRAD\LGS FHA 24hr Delay Test - No Filter Credit Revl.psf Inventory file = p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast source terms for fha.nif Release file p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast fha.rft Dose Conversion file = c:\program files\radtrad3-03\defaults\fgrll&12.inp 4#####1 ###1# #41#### # 41 41 4141414141 41 41 4141414141

1 #1 # # # 414 4141 4141 41 41

1 #1 #1 # # #i # 41

4#41 #1144#411 # # I# # # # 41 41 41 41

1 # # #1 #

1 41## #1 #

41# 41# # #1 41

1 Radtrad 3.03 4/15/2001 PBAPS FHA - TB/RB Ventilation Stack to CR Intake, EAB, & LPZ - 24 Hour Delay and No Filtration Credit Nuclide Inventory File:

p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast source terms for fha.nif Plant Power Level:

2.6830E+01 Compartments:

3 Compartment 1:

Containment 3

1.0000E+02 0

0 0

Compartment 2:

Environment 2

0.OOOOE+00 0

0 0

Calc. No. LM-0645, Rev. 1, Attachment B, Page B-i of B-24

Limerick FHA 24 hr Delay - No Filter Credit Compartment 3:

Control Room 1

1.2600E+05 0

0 0

Pathways:

3 Pathway 1:

Leak to Environment 1

2 2

Pathway 2:

Environment to Control Room 2

3 2

Pathway 3:

Control Room to Environment Exhaust 3

2 2

End of Plant Model File Scenario Description Name:

Plant Model Filename:

ACCEPT\TEST1.PMF Source Term:

1 1 1.OOOOE+00 c:\program files\radtrad3-03\defaults\fgrll&12.inp p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast fha.rft 2.4000E+01 1

0.0000E+00 7.OOOOE-01 3. OOOOE-01 1.0000E+00 Overlying Pool:

0 0.0000E+00 0

0 0

0 Compartments:

3 Compartment 1:

0 1

0 0

0 Compartment 2:

Calc. No. LM-0645, Rev. 1, Attachment B, Page B-2 of B-24

Limerick FHA 24 hr Delay - No Filter Credit 0

1 0

0 0

Compartment 3:

0 1

0 0

0 Pathways:

3 Pathway 1:

0 0

0 1

2 2.4000E+01 1.OOOOE+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 4.8000E+C1 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Pathway 2:

0 0

0 1

2 2.4000E+01 1.2600E+05 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 4.8000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Pathway 3:

0 0

0 1

Calc. No. LM-0645, Rev. 1, Attachment B, Page B-3 of B-24

Limerick FHA 24 hr Delay - No Filter Credit 2

2.4000E+01 1.2600E+05 1.OOOOE+02 l.OOOOE+02 l.OOOOE+02 4.8000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Dose Locations:

3 Location 1:

Exclusion Area Bndry 2

1 2

2.4000E+01 3. 1800E-04 4.8000E+01 O.OOOOE+00 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 0

Location 2:

Low Population Zone 2

1 2

2.4000E+01 1.1500E-04 4.8000E+01 O.OOOOE+00 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 0

Location 3:

Control Room 3

0 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 1

2 2.4000E+01 l.OOOOE+00 4.8000E+01 O.OOOOE+00 Effective Volume Location:

1 2

2.4000E+01 1.2600E-03 4.8000E+01 O.OOOOE+00 Simulation Parameters:

5 2.4000E+01 l.OOOOE-03 2.4010E+01 l.OOOOE-02 2.4100E+01 1.OOOOE-01 2.6000E+01 1.OOOOE+00 4.8000E+01 O.OOOOE+00 Output Filename:

Calc. No. LM-0645, Rev. I, Attachment B, Page B4 of B-24

Limerick FHA 24 hr Delay - No Filter Credit P:\Users\Nuc\Exelon EOC\Discipline Files\Process\AST\Limerick AST\LGS FHA\RADTRAD\LGS FHA 24hr Delay Test - No Filter Credit Revl.oO 1

1 1

0 0

End of Scenario File Calc. No. LM-0645, Rev. 1, Attachment B, Page B-5 of B-24

Limerick FHA 24 hr Delay - No Filter Credit RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:48:13 Plant Description Number of Nuclides 60 Inventory Power = 1.0000E+00 MWth Plant Power Level = 2.6830E+01 MWth Number of compartments = 3 Compartment information Compartment number 1 (Source term fraction = 1.OOOOE+00 Name: Containment Compartment volume = 1.OOOOE+02 (Cubic feet)

Compartment type is Normal Pathways into and out of compartment 1 Exit Pathway Number 1: Leak to Environment Compartment number 2 Name: Environment Compartment type is Environment Pathways into and out of compartment 2 Inlet Pathway Number 1: Leak to Environment Inlet Pathway Number 3: Control Room to Environment Exhaust Exit Pathway Number 2: Environment to Control Room Compartment number 3 Name: Control Room Compartment volume = 1.2600E+05 (Cubic feet)

Compartment type is Control Room Pathways into and out of compartment 3 Inlet Pathway Number 2: Environment to Control Room Exit Pathway Number 3: Control Room to Environment Exhaust Total number of pathways = 3 Calc. No. LM-0645, Rev. I, Attachment B, Page B-6 of B-24

Limerick FHA 24 hr Delay - No Filter Credit RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:48:13 Scenario Description Time between shutdown and first release = 2.4000E+01 (Hours)

Radioactive Decay is enabled Calculation of Daughters is enabled Release Fractions and Timings GAP EARLY IN-VESSEL LATE RELEASE RELEASE MASS 0.000001 hr 0.0000 hrs 0.0000 hrs (gm)

NOBLES 5.OOOOE-02 0.OOOOE+00 0.OOOOE+00 3.108E+00 IODINE 2.5000E-04 0.OOOOE+00 0.OOOOE+00 2.795E-03 CESIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 TELLURIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 STRONTIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 BARIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 RUTHENIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 CERIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 LANTHANUM 0.OOOOE+00 0.OOOOE+00 0.000OE+00 0.OOOE+00 Inventory Power = 27. MWt Nuclide Group Specific half Whole Body Inhaled Inhaled Name Inventory life DCF Thyroid Effective (Ci/MWt) (s) (Sv-m3/Bq-s) (Sv/Bq) (Sv/Bq)

Kr-85 1 7.892E+02 3.383E+08 1.19OE-16 o.OOOE+00 0.OOOE+00 Kr-85m 1 8.313E+03 1. 613E+04 7.480E-15 0.OOOE+00 0.OOOE+00 Kr-87 1 1.633E+04 4 .578E+03 4.120E-14 0.OOOE+00 0.OOOE+00 Kr-88 1 2.303E+04 1. 022E+04 1.020E-13 0.OOOE+00 0.OOOE+00 I-131 2 4.299E+04 6.947E+05 1.820E-14 2.920E-07 8.890E-09 I-132 2 3.881E+04 8.280E+03 1.120E-13 1.740E-09 1.030E-10 I-133 2 5.556E+04 7.488E+04 2.940E-14 4.860E-08 1.580E-09 I-134 2 6.165E+04 3.156E+03 1.300E-13 2.880E-10 3.550E-1l I-135 2 5.192E+04 2.380E+04 8.294E-14 8.460E-09 3.320E-10 Xe-133 1 5.491E+04 4.532E+05 1.560E-15 0.OOOE+00 0.000E+00 Xe-135 1 2.228E+04 3.272E+04 1.190E-14 0.OOOE+00 0.OOOE+00 Nuclide Daughter Fraction Daughter Fractiorn Daughter Fraction Kr-85m Kr-85 0.21 none 0.00 none 0.00 Kr-87 Rb-87 1.00 none 0.00 none 0.00' Kr-88 Rb-88 1.00 none 0.00 none 0.00 Sr-90 Y-90 1.00 none 0.00 none 0.00 Sr-91 Y-91m 0.58 Y-91 0.42 none 0.00 Sr-92 Y-92 1.00 none 0.00 none 0.00 Y-93 Zr-93 1.00 none 0.00 none 0.00 Zr-95 Nb-95m 0.01 Nb-95 0.99 none 0.00 Zr-97 Nb-97m 0.95 Nb-97 0.05 none 0.00 Mo-99 Tc-99m 0.88 Tc-99 0.12 none 0.00 Tc-99m Tc-99 1.00 none 0.00 none 0.00 Ru-103 Rh-103m 1.00 none 0.00 none 0.00 Ru-105 Rh-105 1.00 none 0.00 none 0.00 Ru-106 Rh-106 1.00 none 0.00 none 0.00 Calc. No. LM-0645, Rev. I, Attachment B, Page B-7 of B-24

Limerick FHA 24 hr Delay - No Filter Credit Sb-127 Te-127m 0.18 Te-127 0.82 none 0.00 Sb-129 Te-129m 0.22 Te-129 0.77 none 0.00 Te-127m Te-127 0.98 none 0.00 none 0.00 Te-129 I-129 1.00 none 0.00 none 0.00 Te-129m Te-129 0.65 I-129 0.35 none 0.00 Te-131m Te-131 0.22 I-131 0.78 none 0.00 Te-132 1-132 1.00 none 0.00 none 0.00 I-131 Xe-131m 0.01 none 0.00 none 0.00 I-133 Xe-133m 0.03 Xe-133 0.97 none 0.00 I-135 Xe-135m 0.15 Xe-135 0.85 none 0.00 Xe-135 Cs-135 1.00 none 0.00 none 0.00 Cs-137 Ba-137m 0.95 none 0.00 none 0 .00 Ba-140 La-140 1.00 none 0.00 none 0.00 La-141 Ce-141 1.00 none 0.00 none 0.00 Ce-143 Pr-143 1.00 none 0.00 none 0.00 Ce-144 Pr-144m 0.02 Pr-144 0.98 none 0.00 Nd-147 Pm-147 1.00 none 0.00 none 0.00 Np-239 Pu-239 1.00 none 0.00 none 0 .00 Pu-238 U-234 1.00 none 0.00 none 0.00 Pu-239 U-235 1.00 none 0.00 none 0 .00 Pu-240 U-236 1.00 none 0.00 none 0 .00 Pu-241 U-237 0.00 Am-241 1.00 none 0 .00 Am-241 Np-237 1.00 none 0.00 none 0.00 Cm-242 Pu-238 1.00 none 0.00 none 0.00 Cm-244 Pu-240 1.00 none 0.00 none 0.00 Iodine fractions Aerosol = 0.OOOOE+00 Elemental = 7.OOOOE-01 Organic = 3.OOOOE-01 COMPARTMENT DATA Compartment number 1: Containment Compartment number 2: Environment Compartment number 3: Control Room PATHWAY DATA Pathway number 1: Leak to Environment Pathway Filter: Removal Data Time (hr) Flow Rate Filterr Efficiencies (%)

(cfm) Aerosol Elemental Organic 2.4000E+01 1.OOOOE+01 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 4.8000E+01 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Pathway number 2: Environment to Control Room Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerosol Elemental Organic 2.4000E+01 1.2600E+05 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 4.8000E+01 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Pathway number 3: Control Room to Environment Exhaust Calc. No. LM-0645, Rev. I, Attachment B, Page B-8 of B-24

Limerick FHA 24 hr Delay - No Filter Credit Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerosol Elemental Organic 2.4000E+01 1.2600E+05 1.0000E+02 1.0000E+02 1.0000E+02 4.8000E+01 0.0000E+00 0.OOOOE+00 0.0000E+00 0.OOOOE+00 LOCATION DATA Location Exclusion Area Bndry is in compartment 2 Location X/Q Data Time (hr) X/Q (s

m'-3) 2.4000E+01 3.1800E-04 4.8000E+01 0.OOOOE+00 Location Breathing Rate Data Time (hr) Breathing Rate (mA3

secA-l) 2.4000E+01 3.5000E-04 4.8000E+01 0.OOOOE+00 Location Low Population Zone is in compartment 2 Location X/Q Data Time (hr) X/(Q (s

m^-3) 2.4000E+01 1. 1500E-04 4.8000E+01 0.000OE+00 Location Breathing Raite Data Time (hr) Br(eathing Rate (m^3

secA^l) 2.4000E+01 3.5000E-04 4.8000E+01 0.OOOOE+00 Location Control Room is in compartment 3 Location X/Q Data Time (hr) X/(Q

m^-3) 2.4000E+01 1.2600E-03 4.8000E+01 0.000OE+00 Location Breathing Raite Data Time (hr) Br(eathing Rate (mA3

sec^-l) 2.4000E+01 3.5000E-04 4.8000E+01 0.OOOOE+00 Location Occupancy Fa(,tor Data Time (hr) Oc(cupancy Factor 2.4000E+01 1.000OE+00 4.8000E+01 0.OOOOE+00 USER SPECIFIED TIME STEP DIWTA - SUPPLEMENTAL TIME STEPS Time Time -step O.0000E+00 1.0()OOE-03 1.0000E-02 1.0()OOE-02 1.0000E-01 1.0()OOE-01 2.OOOOE+00 1.0()OOE+00 2.4000E+01 0.0()OOE+00 Calc. No. LM-0645, Rev. 1, Attachment B, Page B-9 of B-24

Limerick FHA 24 hr Delay - No Filter Credit RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:48:13 nfIr ##### ##### IO #t #####i

t It It It 4 It It It It It

t 4 It It It It It It It It

t It It #t 44444 It It I

t It It #t #t It It

t It It #t #t It It

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# It4It44It444# 444444# # It444It 4It# # 4## It#

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Dose Output Exclusion Area Bndry Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.2878E-06 1. 0671E-04 4.5732E-06 Accumulated dose (rem) 1.2878E-06 1.0671E-04 4.5732E-06 Low Population Zone Doses:

Time (hi = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 4.6570E-07 3.8589E-05 1.6538E-06 Accumulated dose (rem) 4.6570E-07 3.8589E-05 1.6538E-06 Control Room Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 6.9103E-12 1.2684E-08 3.9743E-10 Accumulated dose (rem) 6.9103E-12 1.2684E-08 3. 9743E-10 Exclusion Area Bndry Doses:

Time (h) = 48.0000 Whole Body Thyroid TEDE Delta dose (rem) 4.2490E-01 3.5529E+01 1.5186E+00 Accumulated dose (rem) 4.2490E-01 3.5529E+01 1.5186E+00 Low Population Zone Doses:

Time (h) = 48.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.5366E-01 1.2848E+01 5.4918E-01 Accumulated dose (rem) 1.5366E-01 1.2849E+01 5.4918E-01 Control Room Doses:

Time (h) = 48.0000 Whole Body Thyroid TEDE Delta dose (rem) 7.6986E-02 1.4274E+02 4.4708E+00 Accumulated dose (rem) 7.6986E-02 1.4274E+02 4.4708E+00 161 Calc. No. LM-0645, Rev. 1, Attachment B, Page 3-IO of B-24

Limerick FHA 24 hr Delay - No Filter Credit I-131 Summary Containment Environment Control Room Time (hr) I-131 (Curies) 1-131 (Curies) I-131 (Curies) 24.000 2.6581E+02 7.9745E-04 5.9748E-05 24.400 2.4080E+01 2.4157E+02 2.0047E-01 24.700 3.9761E+00 2.6166E+02 -3.3101E-02

25. 000 6.5653E-01 2.6497E+02 5.4657E-03 25.300 1.0841E-01 2.6552E+02 9.0250E-04

25. 600 1.7900E-02 2.6561E+02 1.4902E-04

25. 900 2.9557E-03 2.6563E+02 2.4607E-05

26. 200 4.8805E-04 2.6563E+02 4.0631E-06 26.500 8.0587E-05 2.6563E+02 6.7090E-07

26. 800 1.3307E-05 2.6563E+02 1.1078E-07 27.100 2.1972E-06 2.6563E+02 1.8292E-08 27.400 3.6280E-07 2.6563E+02 3.0204E-09 27.700 5.9906E-08 2.6563E+02 4.9873E-10

28. 000 9.8918E-09 2.6563E+02 8.2351E-11 28.300 1.6333E-09 2.6563E+02 1.3598E-11

28. 600 2.6970E-10 2.6563E+02 2.2453E-12

28. 900 4.4533E-11 2.6563E+02 3.7074E-13 29.200 7.3533E-12 2.6563E+02 6.1217E-14

29. 500 1.2142E-12 2.6563E+02 1.0108E-14

29. 800 2.0049E-13 2.6563E+02 1.6691E-15 30.100 3.3105E-14 2.6563E+02 2.7560E-16

30. 400 5.4662E-15 2.6563E+02 4.5507E-17

30. 700 9.0259E-16 2.6563E+02 7.5142E-18

31. 000 1.4904E-16 2.6563E+02 1.2408E-18

31. 300 2.4609E-17 2.6563E+02 2.0487E-19 31.600 4.0635E-18 2.6563E+02 3.3829E-20

31. 900 6.7096E-19 2.6563E+02 5.5859E-21 32.200 1.1079E-19 2.6563E+02 9.2234E-22 32.500 1.8294E-20 2.6563E+02 1.5230E-22

32. 800 3.0207E-21 2.6563E+02 2.5148E-23

33. 100 4.9878E-22 2.6563E+02 4.1524E-24

33. 400 8.2359E-23 2.6563E+02 6.8565E-25 33.700 1.3599E-23 2.6563E+02 1.1321E-25 34.000 2.2455E-24 2.6563E+02 1.8694E-26 34.300 3.7078E-25 2.6563E+02 3.0868E-27 48.000 7.0589E-61 2.6563E+02 5.8766E-63 Cumulative Dose Summary Exclusion Area Bndry Low Population Zone Control Room Time Thyroid TEDE Thyroid TEDE Thyroid TEDE (hr) (rem) (rem) (rem) (rem) (rem) (rem) 24.000 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 24.400 3.2315E+01 1.3822E+00 1.1686E+01 4.9985E-01 1.2817E+02 4.0148E+00 24.700 3.4999E+01 1.4962E+00 1.2657E+01 5.4109E-01 1.4033E+02 4.3956E+00

25. 000 3.5442E+01 1.5149E+00 1.2817E+01 5.4785E-01 1.4234E+02 4.4584E+00 25.300 3.5515E+01 1.5180E+00 1.2843E+01 5.4896E-01 1.4267E+02 4.4688E+00

25. 600 3.5527E+01 1.5185E+00 1.2848E+01 5.4914E-01 1.4273E+02 4.4705E+00

25. 900 3.5529E+01 1.5186E+00 1.2848E+01 5.4917E-01 1.4274E+02 4.4707E+00 26.200 3.5529E+01 1.5186E+00 1.2848E+01 5.4918E-01 1.4274E+02 4.4708E+00

26. 500 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

26. 800 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 Calc. No. LM-0645, Rev. I, Attachment B, Page B-l I of B-24

Limerick FHA 24 hr Delay - No Filter Credit 27.100 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 27.400 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 27.700 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

28. 000 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 28.300 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

28. 600 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

28. 900 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 29.200 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274EE+02 4.4708E+00

29. 500 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

29. 800 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

30. 100 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

30. 400 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

30. 700 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

31. 000 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

31. 300 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

31. 600 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

31. 900 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 32.200 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 32.500 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

32. 800 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

33. 100 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00

33. 4 00 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 33.700 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 34.000 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 34.300 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 48.000 3.5529E+01 1.5186E+00 1.2849E+01 5.4918E-01 1.4274E+02 4.4708E+00 Worst Two-Hour Doses Exclusion Area Bndry Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 24.0 4.2490E-01 3.5529E+01 1.5186E+00 Calc. No. LM-0645, Rev. I, Attachment B, Page B-12 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:50:17 File information

44##########414######14####$##414#$## 4######4######44#######1########14###########14#####

Plant file = P:\Users\Nuc\Exelon EOC\Discipline Files\Process\AST\Limerick AST\LGS FHA\RADTRAD\LGS FHA 24hr Delay Test - With SGTS Credit Revl.psf Inventory file = p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast source terms for fha.nif Release file = p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast fha.rft Dose Conversion file = c:\program files\radtrad3-03\defaults\fgrll&12.inp

1. 1. 1. 14 ####14# #1 # # #####414 # # ##44#

44# 414 1

1 # ## # #1 # #1 1

4# #4# # #1 #1 #

1

1. 1. 1. 1. 41#414 ##$#14 #1 #1 #1 #1 ##### $1 #1

1. 1 #1 #1 #1 # ## 1

1. #1

1 #1 # # ## #1 #

1. 1. 1. 14 #1 # # # #1 Radtrad 3.03 4/15/2001 PBAPS FHA - TB/RB Ventilation Stack to CR Intake, EAB, & LPZ - 24 Hour Delay and SGTS Filtration Credit Nuclide Inventory File:

p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast source terms for fha.nif Plant Power Level:

2.6830E+01 Compartments:

3 Compartment 1:

Containment 3

1.0000E+02 0

0 0

Compartment 2:

Environment 2

0.OOOOE+00 0

0 0

Calc. No. LM-0645, Rev. I, Attachment B, Page B- 13 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q 0

0 Compartment 3:

Control Room 1

1.2600E+05 0

0 0

Pathways:

3 Pathway 1:

Leak to Environment 1

2 2

Pathway 2:

Environment to Control Room 2

3 2

Pathway 3:

Control Room to Environment Exhaust 3

2 2

End of Plant Model File Scenario Description Name:

Plant Model Filename:

ACCEPT\TEST1.PMF Source Term:

1 1 1.OOOOE+00 c:\program files\radtrad3-03\defaults\fgrll&12.inp p:\users\nuc\exelon eoc\discipline files\process\ast\limerick ast\lgs fha\radtrad\limerick ast fha.rft 2.4000E+01 1

0.OOOOE+00 7.OOOOE-01 3.OOOOE-01 1.OOOOE+00 Overlying Pool:

0 0.0000E+00 0

0 0

0 Compartments:

3 Compartment 1:

0 1

0 0

Calc. No. LM-0645, Rev. I, Attachment B, Page B-14 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q 0

Compartment 2:

0 1

a 0

0 0

0 Compartment 3:

0 1

0 0

0 Pathways:

3 Pathway 1:

0 0

0 1

2 2.4000E+01 1.OOOOE+01 9.9000E+01 9.9000E+01 9.9000E+01 4.8000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Pathway 2:

0 0

0 1

2 2.4000E+01 1.2600E+05 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 4.8000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Pathway 3:

0 0

Calc. No. LM-0645, Rev. 1, Attachment B, Page B-15 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q 0

1 2

2.4000E+01 1.2600E+05 l.OOOOE+02 1.OOOOE+02 1.OOOOE+02 4.8000E+01 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 0

0 0

0 Dose Locations:

3 Location 1:

Exclusion Area Bndry 2

1 2

2.4000E+01 3.1800E-04 4.8000E+01 0.OOOOE+00 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 0

Location 2:

Low Population Zone 2

1 2

2.4000E401 1. 1500E-04 4.8000E+01 O.OOOOE+00 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 0

Location 3:

Control Room 3

0 1

2 2.4000E+01 3.5000E-04 4.8000E+01 O.OOOOE+00 1

2 2.4000E+01 1.OOOOE+00 4.8000E+01 O.OOOOE+00 Effective Volume Location:

1 2

2.4000E+01 6.8800E-03 4.8000E+01 O.OOOOE+00 Simulation Parameters:

5 2.4000E+01 1.OOOOE-03 2.4010E+01 l.OOOOE-02 2.4100E+01 l.OOOOE-01 2.6000E+01 l.OOOOE+00 Calc. No. LM-0645, Rev. 1, Attachment B, Page B-16 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q 4.8000E+01 O.OOOOE+00 Output Filename:

P:\Users\Nuc\Exelon EOC\Discipline Files\Process\AST\Limerick AST\LGS FHA\RADTRAD\LGS FHA 24hr Delay Test - With SGTS Credit Revl.oO 1

1 1

0 0

End of Scenario File Calc. No. LM-0645, Rev. 1, Attachment B, Page B-1 7 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:50:17 Plant Description Number of Nuclides = 60 Inventory Power = 1.OOOOE+00 MWth Plant Power Level = 2.6830E+01 MWth Number of compartments = 3 Compartment information Compartment number 1 (Source term fraction 1.OOOOE+00 Name: Containment Compartment volume = l.OOOOE+02 (Cubic feet)

Compartment type is Normal Pathways into and out of compartment 1 Exit Pathway Number 1: Leak to Environment Compartment number 2 Name: Environment Compartment type is Environment Pathways into and out of compartment 2 Inlet Pathway Number 1: Leak to Environment Inlet Pathway Number 3: Control Room to Environment Exhaust Exit Pathway Number 2: Environment to Control Room Compartment number 3 Name: Control Room Compartment volume = 1.2600E+05 (Cubic feet)

Compartment type is Control Room Pathways into and out of compartment 3 Inlet Pathway Number 2: Environment to Control Room Exit Pathway Number 3: Control Room to Environment Exhaust Total number of pathways = 3 Calc. No. LM-0645, Rev. I, Attachment B, Page B-18 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:50:17 Scenario Description Time between shutdown and first release = 2.4000E+01 (Hours)

Radioactive Decay is enabled Calculation of Daughters is enabled Release Fractions and Timings GAP EARLY IN-VESSEL LATE RELEASE RELEASE MASS 0.000001 hr 0.0000 hrs 0.0000 hrs (gm)

NOBLES 5.OOOOE-02 0.OOOOE+00 0.OOOOE+00 3.108E+00 IODINE 2.5000E-04 0.OOOOE+00 0.OOOOE+00 2.795E-03 CESIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 TELLURIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 STRONTIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 BARIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 RUTHENIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 CERIUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 LANTHANUM 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOE+00 Inventory Power = 27. MWt Nuclide Group Specific half Whole Body Inhaled Inhaled Name Inventory life DCF Thyroid Effective (Ci/MWt) (S) (Sv-m3/Bq-s) (Sv/Bq) (Sv/Bq)

Kr-85 1 7.892E+02 3.383E+08 1.190E-16 0.OOOE+00 0.OOOE+00 Kr-85m 1 8.313E+03 1.613E+04 7.480E-15 0.OOOE+00 0.OOOE+00 Kr-87 1 1.633E+04 4.578E+03 4.120E-14 0.OOOE+00 0.OOOE+00 Kr-88 1 2.303E+04 1.022E+04 1.020E-13 0.OOOE+00 0.OOOE+00 I-131 2 4.299E+04 6.947E+05 1.820E-14 2.920E-07 8.890E-09 1-132 2 3.881E+04 8.280E+03 1.120E-13 1.740E-09 1.030E-10 I-133 2 5.556E+04 7.488E+04 2.940E-14 4.860E-08 1.580E-09 1-134 2 6.165E+04 3.156E+03 1.300E-13 2.880E-10 3.550E-11 I-135 2 5.192E+04 2.380E+04 8.294E-14 8.460E-09 3.32OE-10 Xe-133 1 5.491E+04 4.532E+05 1.560E-15 0.OOOE+00 0.OOOE+00 Xe-135 1 2.228E+04 3.272E+04 1.190E-14 0.OOOE+00 0.OOOE+00 Nuclide Daughter Fraction Daughter Fractiorn Daughter Fraction Kr-85m Kr-85 0.21 none 0.00 none 0.00 Kr-87 Rb-87 1.00 none 0.00 none 0.00 Kr-88 Rb-88 1.00 none 0.00 none 0.00 Sr-90 Y-90 1.00 none 0.00 none 0.00 Sr-91 Y-91m 0.58 Y-91 0.42 none 0.00 Sr-92 Y-92 1.00 ' none 0.00 none 0.00 Y-93 Zr-93 1.00 none 0.00 none 0.00 Zr-95 Nb-95m 0.01 Nb-95 0.99 none 0.00 Zr-97 Nb-97m 0.95 Nb-97 0.05 none 0.00 Mo-99 Tc-99m 0.88 Tc-99 0.12 none 0.00 Tc-99m Tc-99 1.00 none 0.00 none 0.00 Ru-103 Rh-103m 1.00 none 0.00 none 0.00 Ru-105 Rh-105 1.00 none 0.00 none 0.00 Ru-106 Rh-106 1.00 none 0.00 none 0.00 Calc. No. LM-0645, Rev. I, Attachment B, Page B-19 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q Sb-127 Te-127m 0.18 Te-127 0.82 none 0.00 Sb-129 Te-129m 0.22 Te-129 0.77 none 0.00 Te-127m Te-127 0.98 none 0.00 none 0.00 Te-129 I-129 1.00 none 0.00 none 0.00 Te-129m Te-129 0.65 I-129 0.35 none 0.00 Te-131m Te-131 0.22 I-131 0.78 none 0 .00 Te-132 I-132 1.00 none 0.00 none 0.00 1-131 Xe-131m 0.01 none 0.00 none 0.00 I-133 Xe-133m 0.03 Xe-133 0.97 none 0.00 I-135 Xe-135m 0.15 Xe-135 0.85 none 0.00 Xe-135 Cs-135 1.00 none 0.00 none 0.00 Cs-137 Ba-137m 0.95 none 0.00 none 0.00 Ba-140 La-140 1.00 none 0.00 none 0.00 La-141 Ce-141 1.00 none 0.00 none 0.00 Ce-143 Pr-143 - 1.00 none 0.00 none 0.00 Ce-144 Pr-144m 0.02 Pr-144 0.98 none 0.00 Nd-147 Pm-147 1.00 none 0.00 none 0.00 Np-239 Pu-239 1.00 none 0.00 none 0.00 Pu-238 U-234 1.00 none 0.00 none 0.00 Pu-239 U-235 1.00 none 0.00 none 0.00 Pu-240 U-236 1.00 none 0.00 none 0.00 Pu-241 U-237 0.00 Am-241 1.00 none 0.00 Am-241 Np-237 1.00 none 0.00 none 0.00 Cm-242 Pu-238 1.00 none 0.00 none 0.00 Cm-244 Pu-240 1.00 none 0.00 none 0.00 Iodine fractions Aerosol -

0.OOOOE+00 Elemental 7.OOOOE-01 Organic 3.OOOOE-01 COMPARTMENT DATA Compartment number 1: Containment Compartment number 2: Environment Compartment number 3: Control Room PATHWAY DATA Pathway number 1: Leak to Environment Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerosol Elemental Organic 2.4000E+01 1.OOOOE+01 9.9000E+01 9.9000E+01 9.9000E+01 4.8000E+01 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Pathway number 2: Environment to Control Room Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerosol Elemental Organic 2.4000E+01 1.2600E+05 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 4.8000E+01 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Pathway number 3: Control Room to Environment Exhaust Calc. No. LM-0645, Rev. 1, Attachment B, Page B-20 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerossol Elemental Org anic 2.4000E+01 1.2600E+05 1.0000E.+02 1.OOOOE+02 1.00 OOE+02 4.8000E+01 0.OOOOE+00 0.0000E.+00 O.0000E+00 0.00 OOE+00 LOCATION DATA Location Exclusion Area Bndry is in compartment 2 Location X/Q Data Time (hr) X/Q (S

MA -3) 2.4000E+01 3.1800E-04 4.8000E+01 0.OOOOE+00 Location Breathing Rate Data Time (hr) Breathing Rate (mA3

secA-1) 2.4000E+01 3.5000E-04 4.8000E+01 0.OOOOE+00 Location Low Population Zone is in compartment 2 Location X/Q Data Time (hr) X/Q (s

mA-3) 2.4000E+01 1.1500E-04 4.8000E+01 0.0000E+00 Location Breathing Rate Data Time (hr) Breathing Rate (mA3

secA^-)

2.4000E+01 3.5000E-04 4.8000E+01 0.OOOOE+00 Location Control Room is in compartment 3 Location X/Q Data Time (hr) X/Q (s

mA^-3) 2.4000E+01 6.8800E-03 4.8000E+01 0.OOOOE+00 Location Breathing Rate Data Time (hr) Breathing Rate (m^3

secA^1) 2.4000E+01 3.5000E-04 4.8000E+01 0.0000E+00 Location Occupancy Factor Data Time (hr) Occupancy Factor 2.4000E+01 1.0000E+00 4.8000E+01 0.0000E+00 USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS Time Time step 0.OOOOE+00 1.0000E-03 1.0000E-02 1.0000E-02 1.0000E-01 1.0000E-01 2.0000E+00 1.0000E+00 2.4000E+01 0.0000E+00 Calc. No. LM-0645, Rev. 1, Attachment B, Page B-21 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q RADTRAD Version 3.03 (Spring 2001) run on 9/23/2005 at 9:50:17

1. 1. 1. 1. 414#$############4414##$1##14###$1#########$##$4##4###$1##$4##1##$4#########4######

1. 1. 1. 14 #1 $ 414141## 4141#41#

# #1 # 41 41 #

4141# 4# 41 41 #

#4 # 4$ 41 ###41# 41 41

4 # 4# 41 41 41 41 4141# # 41 41

1. 1. 1. 14 41#### 41 41 ####14 Dose Output

1. 1. soAe######Ds####$#:##############################################

Exclusion Area Bndry Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.1609E-06 1.0671E-06 1.1938E-06 Accumulated dose (rem) 1.1609E-06 1.0671E-06 1.1938E-06 Low Population Zone Doses:

Time (h) = 24.0000 Whole Body Thyroid -TEDE Delta dose (rem) 4.1982E-07 3.8589E-07 4.3170E-07 Accumulated dose (rem) 4.1982E-07 3.8589E-07 4.3170E-07 Control Room Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.4016E-11 6.9257E-10 5.5339E-11 Accumulated dose (rem) 3.4016E-11 6.9257E-10 5.5339E-11 Exclusion Area Bndry Doses:

Time (h) = 48.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.8384E-01 3.5529E-01 3.9478E-01 Accumulated dose (rem) 3.8384E-01 3.5529E-01 3.9478E-01 Low Population Zone Doses:

Time (h) = 48.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.3881E-01 1.2848E-01 1.4276E-01 Accumulated dose (rem) 1.3881E-01 1.2849E-01 1.4277E-01 Control Room Doses:

Time (h) = 48.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.7983E-01 7.7939E+00 6.1975E-01 Accumulated dose (rem) 3.7983E-01 7.7939E+00 6.1975E-01 161 Calc. No. LM-0645, Rev. I, Attachment B, Page B-22 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR X/Q I-131 Summary Containment Environment Control Room Time (hr) I-131 (Curies) I-131 (Curies) I-131 (Curies) 24.000 2.6581E+02 7.9745E-06 3.2625E-06 24.400 2.4080E+01 2.4157E+00 1.0946E-02 24.700 3.9761E+00 2.6166E+00 1.8074E-03 25.000 6.5653E-01 2.6497E+00 2.9845E-04 25.300 1.0841E-01 2.6552E+00 4.9280E-05

25. 600 1.7900E-02 2.6561E+00 8.1371E-06

25. 900 2.9557E-03 2.6563E+00 1.3436E-06 26.200 4.8805E-04 2.6563E+00 2.2186E-07

26. 500 8.0587E-05 2.6563E+00 3.6633E-08

26. 800 1.3307E-05 2.6563E+00 6.0489E-09

27. 100 2.1972E-06 2.6563E+00 9.9880E-10 27.400 3.6280E-07 2.6563E+00 1.6492E-10 27.700 5.9906E-08 2.6563E+00 2.7232E-11 28.000 9.8918E-09 2.6563E+00 4.4966E-12 28.300 1.6333E-09 2.6563E+00 7.4248E-13

28. 600 2.6970E-10 2.6563E+00 1.2260E-13

28. 900 4.4533E-11 2.6563E+00 2.0244E-14 29.200 7.3533E-12 2.6563E+00 3.3427E-15

29. 500 1.2142E-12 2.6563E+00 5.5194E-16

29. 800 2.0049E-13 2.6563E+00 9.1137E-17

30. 100 3.3105E-14 2.6563E+00 1.5049E-17

30. 400 5.4662E-15 2.6563E+00 2.4848E-18

30. 700 9.0259E-16 2.6563E+00 4.1030E-19

31. 000 1.4904E-16 2.6563E+00 6.7749E-20

31. 300 2.4609E-17 2.6563E+00 1.1187E-20

31. 600 4.0635E-18 2.6563E+00 1.8472E-21

31. 900 6.7096E-19 2.6563E+00 3.0501E-22

32. 200 1.1079E-19 2.6563E+00 5.0363E-23 32.500 1.8294E-20 2.6563E+00 8.3160E-24

32. 800 3.0207E-21 2.6563E+00 1.3731E-24 33.100 4.9878E-22 2.6563E+00 2.2673E-25

33. 400 8.2359E-23 2.6563E+00 3.7439E-26

33. 700 1.3599E-23 2.6563E+00 6.1819E-27 34.000 2.2455E-24 2.6563E+00 1.0208E-27 34.300 3.7078E-25 2.6563E+00 1.6855E-28 48.000 7.0589E-61 2.6563E+00 3.2088E-64 Cumulative Dose Summary Exclusion Area Bndry Low Population Zone Controll Room Time Thyroid TEDE Thyroid TEDE Thyroid TEDE (hr) (rem) (rem) (rem) (rem) (rem) (rem) 24.000 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 24.400 3.2315E-01 3.5971E-01 1.1686E-01 1.3008E-01 6.9983E+00 5.5719E-01 24.700 3.4999E-01 3.8908E-01 1.2657E-01 1.4070E-01 7.6627E+00 6. 0952E-01

25. 000 3.5442E-01 3.9385E-01 1.2817E-01 1.4243E-01 7.7723E+00 6.1808E-01

25. 300 3.5515E-01 3.9463E-01 1.2843E-01 1.4271E-01 7.7903E+00 6.1947E-01

25. 600 3.5527E-01 3.9475E-01 1.2848E-01 1.4276E-01 7.7933E+00 6.1970E-01

25. 900 3.5529E-01 3.9477E-01 1.2848E-01 1.4276E-01 7.7938E+00 6.1974E-01 26.200 3.5529E-01 3.9478E-01 1.2848E-01 1.4277E-01 7.7939E+00 6.1975E-01

26. 500 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

26. 800 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 Calc. No. LM-0645, Rev. I, Attachment B, Page B-23 of B-24

Limerick FHA 24 hr Delay - With SGTS Filter Credit and North Stack CR XIQ

27. 100 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

27. 400 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6. 1975E-01 27.700 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

28. 000 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

28. 300 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

28. 600 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

28. 900 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 29.200 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

29. 500 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6. 1975E-01

29. 800 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

30. 100 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 30.400 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

30. 700 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

31. 000 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

31. 300 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

31. 600 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 31.900 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 32.200 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 32.500 3.5529E-01 3. 9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

32. 800 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

33. 100 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6. 1975E-01

33. 400 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 33.700 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 34.000 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 34.300 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01

48. 000 3.5529E-01 3.9478E-01 1.2849E-01 1.4277E-01 7.7939E+00 6.1975E-01 Worst Two-Hour Doses Exclusion Area Bndry Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 24.0 3.8384E-01 3.5529E-01 3.9477E-01 Calc. No. LM-0645, Rev. 1, Attachment B, Page B-24 of B-24

Limerick Generating Station AST Source Termns for FHA.nif Nuclide Inventory Name: Source Terms per this calculation Limerick Generating Station (LGS) FHA AST - in Ci/MW Power Level:

0.1000E+01 Nuclides:

60 Nuclide 001:

Co-58 7

0.6117120000E+07 0.5800E+02 0.1529E+03 none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 002:

Co-60 7

0.1663401096E+09 0.6000E+02 0.1830E+03 none 0.OOOOE+00 none 0.0000E+00 none 0.0000E+00 Nuclide 003:

Kr-85 1

0.3382974720E+09 0.8500E+02 0.7892E+03 2.0*LOCA Value for FHA none 0.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 004:

Kr-85m 1

0.1612800000E+05 0.8500E+02 0.8313E+04 Kr-85 0.2100E+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 005:

Kr-87 1

0.4578000000E+04 0.8700E+02 0.1633E+05 Rb-87 0.1000E+01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 006:

Kr-88 1

0.1022400000E+05 0.8800E+02 0.2303E+05 Rb-88 0.1000E+01 none 0.OOOOE+00 Casc. No. LM-0645, Rev. I, Attachment C, Page C- I of C-10

Limerick Generating Station AST Source Terms for FHA.nif none O.OOOOE+00 Nuclide 007:

Rb-86 3

0.161222400OE+07 0.8600E+02 0.6518E+02 none O.O00OE+O0 none O.OOOOE+00 none O.OOOOE+O0 Nuclide 008:

Sr-89 5

0.4363200000E+07 0.8900E+02 0.2798E+05 none O.OOOOE+O0 none O.OOOOE+OO none O.OOOOE+O0 Nuclide 009:

Sr-90 5

0.9189573120E+09 0.9000E+02 0.3178E+04 Y-90 0.1000E+O1 none O.OOOOE+OO none 0.OOOOE+00 Nuclide 010:

Sr-91 5

0.3420000000E+05 0.9100E+02 0.3801E+05 Y-91m 0.5800E+00 Y-91 0.4200E+00 none O.OOOOE+O0 Nuclide 011:

Sr-92 5

0.9756000000E+04 0.9200E+02 0.4017E+05 Y-92 O.lOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 012:

Y-90 9

0.2304000000E+06 0.9000E+02 0.3272E+04 none O.OOOOE+0O none O.OOOOE+00 none O.OOOOE+00 Nuclide 013:

Y-91 9

0.5055264000E+07 Calc. No. LM-0645, Rev. 1, Attachment C, Page C-2 of C-l 0

Limerick Generating Station AST Source Terms for FHA.nif 0.9100E+02 0.3448E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 014:

Y-92 9

0. 1274400000E+05 0.9200E+02 0.4029E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 015:

Y-93 9

0.3636000000E+05 0.9300E+02 0.4526E+05 Zr-93 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 016:

Zr-95 9

0.5527872000E+07 0.9500E+02 0.4489E+05 Nb-95m 0.7000E-02 Nb-95 0.9900E+00 none O.OOOOE+00 Nuclide 017:

Zr-97 9

0.6084000000E+05 0.9700E+02 0.4657E+05 Nb-97m 0.9500E+OO Nb-97 0.5300E-01 none O.OOOOE+00 Nuclide 018:

Nb-95 9

0.3036960000E+07 0.9500E+02 0.4512E+05 none O.OOOOE+00 none O.OOOOE+00 none 0.OOOOE+00 Nuclide 019:

Mo-99 7

0.2376000000E+06 0.9900E+02 0.5078+05 Tc-99m 0.8800E+00 Tc-99 0.1200E+00 none O.OOOOE+00 Calc. No. LM-0645, Rev. 1,Attachment C, Page C-3 of C-10

Limerick Generating Station AST Source Terms for FHA.nif Nuclide 020:

Tc-99m 7

0.2167200000E+05 0.9900E+02 0.4447E+05 Tc-99 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 021:

Ru-103 7

0.3393792000E+07 0.1030E+03 0.4202E+05 Rh-103m 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 022:

Ru-105 7

0.1598400000E+05 0.1050E+03 0.2908E+05 Rh-105 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 023:

Ru-106 7

0.3181248000E+08 0.1060E+03 0.1730E+05 Rh-106 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 024:

Rh-105 7

0.1272960000E+06 0.1050E+03 0.2752E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 025:

Sb-127 4

0.3326400000E+06 0.1270E+03 0.2896E+04 Te-127m 0.1800E+00 Te-127 0.8200E+00 none O.OOOOE+00 Nuclide 026:

Sb-129 4

0.1555200000E+05 0.1290E+03 Calc. No. LM-0645, Rev. I, Attachment C, Page C4 of C-10

Limerick Generating Station AST Source Terms for FHA.nif 0.8638E+04 Te-129m 0.2200E+00 Te-129 0.7700E+00 none O.OOOOE+00 Nuclide 027:

Te-127 4

0.3366000000E+05 0.1270E+03 0.2873E+04 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 028:

Te-127m 4

0.9417600000E+07 0.1270E+03 0.3855E+03 Te-127 0.9800E+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 029:

Te-129 4

0.4176000000E+04 0.1290E+03 0.8501E+04 I-129 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 030:

Te-129m 4

0.2903040000E+07 0.1290E+03 0.1267E+04 Te-129 0.6500E+00 1-129 0.3500E+00 none O.OOOOE+00 Nuclide 031:

Te-131m 4

0.1080000000E+06 0.1310E+03 0.3869E+04 Te-131 0.2200E+00 I-131 0.7800E+00 none O.OOOOE+00 Nuclide 032:

Te-132 4

0.2815200000E+06 0.1320E+03 0.3821E+05 I-132 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 033:

Calc. No. LM-0645, Rev. 1, Attachment C, Page C-5 of C- IO

Limerick Generating Station AST Source Terms for FHA.nif I-131 2

0.6946560000E+06 0.1310E+03 0.4299E+05 1.6*LOCA Value for FHA Xe-131m 0.11OOE-01 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 034:

I-132 2

0.8280000000E+04 0.1320E+03 0.3881E+05 none 0.0000E+00 none 0.0000E+00 none 0.0000E+00 Nuclide 035:

I-133 2

0.7488000000E+05 0.1330E+03 0.5556E+05 Xe-133m 0.2900E-01 Xe-133 0.9700E+00 none 0.OOOOE+00 Nuclide 036:

I-134 2

0.3156000000E+04 0.1340E+03 0.6165E+05 none 0.OOOOE+00 none 0.O000E+00 none 0.O000E+O0 Nuclide 037:

I-135 2

0.2379600000E+05 0.1350E+03 0.5192E+05 Xe-135m 0.1500E+00 Xe-135 0.8500E+00 none 0.OOOOE+00 Nuclide 038:

Xe-133 1

0.4531680000E+06 0.1330E+03 0.5491E+05 none 0.OOOOE+00 none 0.0OOOE+00 none 0.OOOOE+00 Nuclide 039:

Xe-135 1

0.3272400000E+05 0.1350E+03 0.2228E+05 Calc. No. LM-0645, Rev. I, Attachment C, Page C-6 of C-1O

Limerick Generating Station AST Source Terms for FHA.nif Cs-135 O.lOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 040:

Cs-134 3

0.6507177120E+08 0.1340E+03 0.7280E+04 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 041:

Cs-136 3

0.1131840000E+07 0.1360E+03 0.2027E+04 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 042:

Cs-137 3

0.9467280000E+09 0.1370E+03 0.4538E+04 Ba-137m 0.9500E+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 043:

Ba-139 6

0.4962000000E+04 0.1390E+03 0.5084E+05 none 0.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 044:

Ba-140 6

0.1100736000E+07 0.1400E+03 0.4896E405 La-140 O.lOOOE+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 045:

La-140 9

0.1449792000E+06 0.1400E+03 0.5019E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 046:

La-141 Calc. No. LM-0645, Rev. I, Attachment C, Page C-7 of C-l 0

Limerick Generating Station AST Source Terms for FHA.nif 9

0.1414800000E+05 0.1410E+03 0.4640E+05 Ce-141 0.1000E+01 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 047:

La-142 9

0.5550000000E+04 0.1420E+03 0.4532E+05 none O.OOOOE+00 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 048:

Ce-141 8

0.2808086400E+07 0.1410E+03 0.4492E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 049:

Ce-143 8

0.1188000000E+06 0.1430E+03 0.4427E+05 Pr-143 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 050:

Ce-144 8

0.2456352000E+08 0.144OE+03 0.3596E+05 Pr-144m 0.1800E-01 Pr-144 0.9800E+00 none O.OOOOE+00 Nuclide 051:

Pr-143 9

0.1171584000E+07 0.1430E+03 0.4293E+05 none O.OOOOE+00 none O.OOOOE+00 none O.OOOOE+00 Nuclide 052:

Nd-147 9

0.9486720000E+06 0.1470E+03 0.1838E+05 Pm-147 0.1000E+01 CaIc. No. LM-0645, Rev. 1, Attachment C, Page C-8 of C-l O

Limerick Generating Station AST Source Terms for FHA.nif none O.OOOOE+00 none O.OOOOE+00 Nuclide 053:

Np-239 8

0.2034720000E+06 0.2390E+03 0.5397E+06 Pu-239 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 054:

Pu-238 8

0.2768863824E+10 0.2380E+03 0.1796E+03 U-234 O.lOOOE+01 none O.OOOOE+00 none 0.OOOOE+00 Nuclide 055:

Pu-239 8

0 .7594336440E+12 0.2390E+03 0.1200E+02 U-235 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 056:

Pu-240 8

0.2062920312E+12 0.2400E+03 0.1288E+02 U-236 0.1000E+01 none O.OOOOE+00 none O.OOOOE+00 Nuclide 057:

Pu-241 8

0.4544294400E+09 0.2410E+03 0.6182E+04 U-237 0.2400E-04 Am-241 O.lOOOE+01 none O.OOOOE+00 Nuclide 058:

Am-241 9

0.1363919472E+ll 0.2410E+03 0.9528E+01 Np-237 0.1000E+01 none 0.OOOOE+00 none O.OOOOE+00 Nuclide 059:

Cm-242 9

Calc. No. LM-0645, Rev. 1, Attachment C, Page C-9 of C-l 0

Limerick Generating Station AST Source Terms for FHA.nif 0.1406592000E+08 0.2420E+03 0.2388E+04 Pu-238 0.1000E+01 none 0.OOOOE+0O none O.OOOOE+00 Nuclide 060:

Cm-244 9

0.5715081360E+09 0.2440E+03 0.2602E+03 Pu-240 0.1000E+01 none O.OOOOE+0O none O.OOOOE+0O End of Nuclear Inventory File Calc. No. LM-0645, Rev. 1, Attachment C, Page C-10 of C-10

Limerick Generating Station AST FHA.rft Release Fraction and Timing Name:

Limerick Generating Station FHA, 8x8 bundle Pool I DF=200, Cs DF=infinity Duration (h):

O.lOOOE-05 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Noble Gases:

5.OOOOE-02 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Iodine:

2.5000E-04 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Cesium:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Tellurium:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Strontium:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Barium:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Ruthenium:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Cerium:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Lanthanum:

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 Non-Radioactive Aerosols (kg):

O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 End of Release File Calc. No. LM-0645, Rev. 0, Attachment D, Page D-1 of D-l

LGS Fuel Handling Accident Assessment of Limiting Event This Attachment benchmarks a fuel damage assessment approach as developed for GESTAR II by the General Electric Company for GE14 10x1O fuel dropping 34 feet over the open reactor vessel, and then applies it to the damage assessment of the same fuel and also an 8x8 fuel bundle for the corresponding drop over spent fuel stored in racks in the LGS Spent Fuel Pool (SFP). The results are then used in an Excel spreadsheet to show that the reactor vessel drop, with more than 23 feet of water covering the dropped and the struck fuel, bounds the results of the SFP drop, with less water coverage.

The analysis associated with the GE14 10x10 fuel is based on the NEDE-24011-P-A-US' analysis and the known result for a 34 ft drop, namely 172 broken rods. The general expression for the number of broken rods is:

drop height *(bundle + mast weight)- 0.5 d clad weight rods in fracthard bundleweight - fuel weight Indropenergy with

- - - - -. - 3 a ., impacted fuel fraction of energy shared between clad and fuel oWLUI UIUhKen LUU = UauppaU 1 bundle 17 5 fit - Ibm I , rod bunl energy per rod failure number broken due to initial impact clad weight bundle length *(mast weight + 0.5 -bundle weight)- 0.5 fraction shared bundle weight - fuel weight with impacted fuel fraction of energy shared between clad and fuel 175 ft - Ibm rod energy per rod failure number broken due to secondary impact where:

Drop Height =34 ft' Bundle Length =160 in' Mast Weight [Wet] =619 lbs' Bundle Weight [Wet] =568 lbs2 Cladding Weight =100.9 Ibs2 Total assembly Weight [Dry] =6451bs 2 Total Pellet Weight =455 lbs3 Energy per rod failure =175 ft-Ibm/rod 2 Fraction of Energy Absorbed by Clad:

100.9 Ibm -0.531 645 Ibm - 455 Ibm Inserting the above values, one obtains the following:

l60in *(619+0.5-568)-0.5-0.531 12in 12-92rods+ 34ft -(6191b + 568 bm)- 0.5-0.531+ fit = 172rods dropped 175- 175 fi-Ibm bundle rod rod initial impact-62 rods secondary impactl 8 rods I NEDE-2401 1-P-A-14-US, General Electric Standard Application for Reactor Fuel, Licensing Topical Report, June 2000 2 Letter dated June 2, 2000 from J. Baumgartner, GNF Fuel Project Manager, to J. Carmody, Exelon 3GE Nuclear Energy Fuel Bundle Data Sheet 24A5424, Rev. 2 FHA Damage Assessment Caic. No. LM-0645, Rev. 1, Attachment E, Page E-I of E-8

Thus, in the case of a 1Ox1 0 bundle the number of failed rods would be 92 from the impacting (dropped) assembly 62 (H/34) from the impacted assemblies when H is the height of the drop 18 from the second impact.

Thus, the desired benchmark for the 34 foot drop height over the reactor core resulting in 172 failed rods is achieved.

The following analysis is for the same GE14 10x1O fuel dropped over the worst case SFP configuration. The drop height is conservatively considered as 2.33 feet for a drop from the Bottom of the Spent Fuel Assembly at Full Uplift to the Top of the Active Fuel in the Spent Fuel Rack, which from page E-7, rows 10 and 15 is 226.5 inches - 198.5 inches (the same drop to the page E-7 row 12 Top of Assembly Bail Handle in the Racks in the Spent Fuel Pool is only 9.2 inches). Again, the general expression for the number of broken rods is:

drop height *(bundle + mast weight) . 0.5

ciaU Welg[lL rods in dropeneegy drop ,icthcrgy ' fractshared wh bundle weight - fuel weight

. . I I I I I -I impacted fuel fraction of energy shared between clad and fuel total broken rods = dropped +

175f -Ibm bundle

% V rod drop pd bundl encrgy per rod failure number broken due to initial impact clad weight bundle length *(mast weight + 0.5 - bundle weight)- 0.5 fraction shared -bundle weight - fuel weight with impacted fuel fraction of energy shared between clad and fuel 1 ft - Ibm rod energy per rod failure number broken due to secondary impact where now:

Drop Height =2.33 ft Bundle Length =160 in Mast Weight [Wet] =619 lbs Bundle Weight [Wet] =568 lbs Cladding Weight =100.9 lbs Total assembly Weight [Dry] =6451bs Total Pellet Weight =455 lbs Energy per rod failure =175 ft-Ibm/rod Fraction of Energy Absorbed by Cla d:

100.9 Ibm 0.531 645 Ibm - 455 Ibm Inserting the above values, one obtains the following:

160.in . (619 + 0.5 -568) *0.5 -0.531 12 in 92rods + 2.33ft -(6191b + 568 ibm). 0.5-0.531 + ft dropped 175 ' + 175 ibm bundle .- ,1 roa rod initial impactx4 rods secondary impact l8 rods FHA Damage Assessment Calc. No. LM-0645, Rev. I, Attachment E, Page E-2 of E-8

Thus, in the case of a 1Ox1 0 bundle the number of failed rods from a worst case SFP drop would be 92 from the impacting (dropped) assembly 4 from the impacted assemblies 18 from the second impact for a total of 114 10x10 rods damaged in the SFP.

A corresponding analysis is performed below for the GE 8x8 fuel (62 fuel rods and 2 water rods) addressed in LGS UFSAR Section 15.7.4 dropped over the worst case spent fuel pool configuration. The drop height is again conservatively considered as 2.33 feet for a drop from the Bottom of the Spent Fuel Assembly at Full Uplift to the Top of the Active Fuel in the Spent Fuel Rack. Using the general expression above for the number of broken rods with the energy per rod failure from LGS UFSAR Section 15.7.4.3.3.2 of 250 ft-lbm/rod, clad weight drop height . (bundle + mast weight)- 0.5 rods in drop lo rgy f _with fractiohared bundle weight - fuet weight

~ ti.~.. .. ~I

- ,impacted fuel fraction or energy sharedbetwcen clad and fuel tLolL UIUKUIL LUUs = U1UppeU 1 bundle 250 ft - Ibm

% V rod dropped bundle energy per rod failure number broken due to initial impact clad weight bundle length *(mast weight + 0.5 -bundle weight)- 0.5 fraction shared bundle weight - fuel weight with impacted fuel fraction of energyshared between clad and fuel 1

2 5 0 ft - Ib rod energy per rod failure number broken due to secondary impact where:

Drop Height =2.33 ft Bundle Length =160 in Mast Weight [Wet] =619 lbs Total Assembly Weight [Dry] =6781bs for worst case (GE7/8 2 water rod) 8x8 fuel from PECO Nuclear Memorandum,

Subject:

GE7, GE8, and GE9 Bundle Masses, from V. K. Aggarwal to A, M. Olson, January 22, 1999.

Bundle Weight [Wet] =597 lbs using 0.88 wet/dry weight ratio of 10x10 fuel above Fraction of Energy Absorbed by Clad = (mass of fuel cladding) / (mass of total assembly - mass of fuel) = 0.504 for worst case (GE9) 8x8 fuel from PECO Nuclear Memorandum,

Subject:

GE7, GE8, and GE9 Bundle Masses, from V. K. Aggarwal to A, M. Olson, January 22, 1999.

Inserting the above values, one obtains the following:

l6Oin 12!Li1 .(619 + o.s5. 597). 0.5 . 0.504 12ht 62rods + 2.33ft . (6191b + 597 Ibm). 0.5 0.504 A+ for 62 + 2 +l12rods 62rods dropped25 2 f t /b- in 250 ff - Nib, bundle rod rod initial impact*2 rods secondary impactel2 rods FHA Damage Assessmcnt Caic. No. LM-0645, Rev. 1, Attachment E, Page E-3 of E-8

Thus, in the case of the worst case 8x8 bundle the number of failed rods for a SFP drop would be 62 from the impacting (dropped) assembly

+ 2 from the impacted assemblies

+ 12 from the second impact, for a total Of 76 8x8 rods damaged.

Comparing the 114 damaged 1Ox10 rods to the 76 damaged 8x8 rods on a number of affected bundles basis, with 87.33 the effective number of rods in a 1Ox10 bundle and 62 rods in a 8x8 bundle, the 1Ox1O fuel bounds, as shown below:

1Ox10: 114/(87.33) = 1.305 8x8: 76/(62) = 1.226 FHA Damage Assessment CaIc. No. LM-0645, Rev. 1,Attachment E, Page E-4 of E-8

A I B I C I D I E I F I G H K L M T LGS Fuel Handling Accident Assessment of Limiting Evenl l 2 The balance of this

Attachment:

3 [a] Evaluates the reduced water coverage for dropped fuel lying horizontally on top 4 of the racks and struck fuel in the racks for FHAs in the Spent Fuel Pool (SFP) 5 compared to over the Reactor Well.

6 [b] Evaluates impact of water coverages of 23 feet and less for determination of 7 conservative SFP pool water Decontamination Factors (DF). _

8 [c] Justifies that a FHA over the Reactor Well is the limiting event.

9.

10 Baseline R.G. 1.183 based Analysis of DFs _

11 RG 1.183 RG 1.183 =

12 Water RG 1.183 RG 1.183 Inorganic Organic; 13 Coverage Inorganic Organic Iodine Iodine DF Overall _

14 (feet) Iodine DF Iodine DF Fraction Fraction DF _

15 23 500 1 0.9985 0.0015 286.0 Inorganic Iodine DF Guidance Controlling_

16 23 285.3 1 0.9985 0.0015 200.0 Overall DF Guidance Controlling _

19 DFs deter ined per Burley Pap r with Inor anic Iodine DF Guidance assumptions 20 RG 1.183 RG 1.183 =

21 Water RG 1.183 RG 1.183 Inorganic Organic 22 Coverage Inorganic Organic Iodine Iodine DF Overall 23 (feet) Iodine DF iodine DF Fraction Fraction DF 24 23 500 i 0.9985 0.0015 286.0 capped at 200 25 22.5 436.8 1 0.9985 0.0015 264.1 capped at 200 26 22 381.6 1 0.9985 0.0015 242.9 capped at 200 27 21.5 333.4 1 0.9985 0.0015 222.5 capped at 200 28 21 291.3 1 0.9985 0.0015 202.9 capped at 200 29 20.5 254.5 1 0.9985 0.0015 184.4 30 20 - 222.3 1 1 0.9985 0.0015 j 166.9 31 19.5 194.2 J 1 0.9985 0.0015 150.6 _

32 19 169.7 1 1 0.9985 0.0015 1 135.4 I 33 All water coverages are more than 21 feet. Therefore, the 200 DF is conservative for all cases.

34 _ I I I 1 I__

35 UFs determined per Burley Paper with Overall DF Guidance assumptions 36 _ l l RG 1.183 RG 1.183 l 37 Water RG 1.183 RG 1.183 Inorganic Organic _ ____

38 Coverage Inorganic Organic Iodine Iodine DF Overall _ T I 39 (feet) Iodine DF Iodine DF Fraction Fraction DF l 40 23 285.3 1 0.9985 0.0015 200.0 _ _ _

41 22.5 252.3 1 0.9985 0.0015 183.2 T I _

42 22 223.1 1 0.9985 0.0015 167.4 43 21.5 197.3 1 0.9985 0.0015 152.4 44 21 174.5 1 0.9985 0.0015 138.5 T_

45 20.5 154.3 1 0.9985 1 0.0015 125.5 46 20 136.5 1 0.9985 0.0015 113.4 47 19.5 120.7 1 0.9985 0.0015 102.3 48 19 106.7 1 0.9985 0.0015 92.1 _ _ _

49 21.611 202.8 1 0.9985 0.0015 155.7 line 19 of page E-7; DF for dropped bundle 50 22.612 259.4 1 0.9985 l 0.0015 186.9 line 21 of page E-7; DF for struck bundles 51 Applying the appropriate DFs above for 92 10xlO rods damaged in the dropped bundle, 52 and 22 10x10 rods damaged in the struck (in-rack) bundles (from page E-3):

53 1 l l Overall weighted DF 161.7 l 54 Resulting overall % of 200 DF for 23 feet coverage 80.8% l 55 i ____ _

56 Fuel failure over SFP vs. over the reactor well = 114/172 66.3%l _ _ _ _

57 Therefore, the drop over the reactor well is bounding when a 23 feet coverage DF of 200 is used. I I Bounding FHA Assessment Calc. No. LM-0645, Rev. I, Attachment E, Page E-5 of E-8

l B IC ID E F I G 1 1LGS FuellI I I I I 2 The balance of this Attachment

[a] Evaluates the reduced water coverage for dropped fuel lying horizontally on top of the racks and struck fuel in the racks for FHAs in the S Pool (SFP) compared to over the Reactor Well.

3 [b] Evaluates impact of water coverages of 23 feet and less for determination of conservative SFP pool water Decontamination Factors (DF).

[c] Justifies that a FHA over the Reactor Well is the limiting event.

7 8

9 10 Baseline R.

11 RG 1.183 RG 1.183 12 Water RG 1.183 RG 1.183 Inorganic Organic 13 Coverage Inorganic Organic Iodine Iodine DF Overall 14 (feet) Iodine DF Iodine DF Fraction Fraction DF 15 23 500 1 0.9985 0.0015 =1/(D15/B15+E15/C15) Inorganic Iodine DF G 16 23 285.3 1 0.9985 0.0015 =11(D16/B16+E161C16) Overall DF Guidance 4 17 18 19 DFs determ 20 RG 1.183 RG 1.183 21 Water RG 1.183 RG 1.183 Inorganic Organic 22 Coverage Inorganic Organic Iodine Iodine DF Overall 23 (feet) Iodine DF Iodine DF Fraction Fraction DF 24 23 500 1 0.9985 0.0015 =1/(D24/B24+E24/C24) capped at 200 25 =A24-0.5 =BS24^(A25/A$24) 1 0.9985 0.0015 =1/(D251B25+E25/C25) capped at 200 26 =A25-0.5 =B$24A(A26/A$24) 1 0.9985 0.0015 =11(D261B26+E26/C26) capped at 200 27 =A26-0.5 =BS24^(A27/A$241 0.9985 0.0015 =1/(D27/B27+E271C27) capped at 200 28 =A27-0.5 =B$24A(A28/A$24) 1 0.9985 0.0015 =1/(D28/B28+E281C28) capped at 200 29 =A28-0.5 =B$24A(A29/A$24) 1 0.9985 0.0015 =1/(D29/B29+E291C29) 30 =A29-0.5 =BS24A(A30/A$24) 1 0.9985 0.0015 =1/(D30/B30+E30/C30) 31 =A30-0.5 =B$24A(A31/AS24) 1 0.9985 0.0015 =1/(D31/B31+E31/C31) 32 =A31-0.5 =BS24A(A32/AS24) 1 0.9985 0.0015 =1/(D32/B32+E32/C32) 33 All water co 34 35 DFs determ _

36 RG 1.183 RG 1.183 37 Water RG 1.183 RG 1.183 Inorganic Organic 38 Coverage Inorganic Organic Iodine Iodine DF Overall 39 (feet) Iodine DF Iodine DF Fraction Fraction DF 40 23 285.3 1 0.9985 0.0015 =1/(D40/B40+E401C40) 41 =A40-0.5 =B$40A(A41/A$40) 1 0.9985 0.0015 =1/(D41/B41+E41/C41) 42 =A41-0.5 =B$40A(A42/A$40) 1 0.9985 0.0015 =1/(D421B42+E421C42) 43 =A42-0.5 =B$40A(A43/A$40) 1 0.9985 0.0015 =1/(D43/B43+E43/C43) 44 =A43-0.5 =B$40A(A44/A$40) 1 0.9985 0.0015 =1/(D44/B44+E44/C44) 45 =A44-0.5 =B$40A(A45/A$40) 1 0.9985 0.0015 =1U(D45/B45+E45/C45) 46 =A45-0.5 =B$40A(A46/A$40) 1 0.9985 0.0015 =1/(D46/B46+E46/C46) 47 =A46-0.5 =B$40A(A47/A$40) 1 0.9985 0.0015 =1/(D471B47+E47/C47) j 48 =A47-0.5 =B$40A(A48/A$40) 1 0.9985 0.0015 =11(D48/B48+E48/C48) 49 21.611 =B$40A(A49/A$40) 1 0.9985 0.0015 =1/(D49/B49+E49/C49) line 19 of page E-7; DF 50 22.6120833: =B$40A(A50/A$40) 1 0.9985 0.0015 =1/(D50/B50+E50/C50) line 21 of page E-7; OF 51 Applying the appropriate UFs above for 92 10x10 rods damaged in the dropped bundle.

52 and 22 10x10 rods damaged in the struck (in-rack bundles (from page E-3):

53 Overall weighted DF =(92-F49+(22)1F50)1114l 54 200 DF for 23 feet coverage =F53/200 55 56 Fuel failure _ 0.663 57 Therefore. U Bounding FHA Assessment Formula Calc. No. LM-0645, Rev. I, Attachment E, Page E-6 of E-8

A I B I C I D I E I F I G I H I K L 1 LGS Damaged Fuel Water Cover Assessment for Fuel Handling Accidents 2 l 3 Reference Points _ II_1 4 MSL Water level SPF Bottom 5 (feet) (inches) (feet) 6 352.000 506 39.250 Refuel Floor 7 351.000 494 38.250 Normal Spent Fuel Pool Water Level 8 350.000 482 37.250 Tech Spec 3.9.8 water level for 22 ft above reactor pressure vessel flange 9 327.958 217.5 15.208 Elevation at Reactor Vessel Flange -

10 328.708 226.5 15.958 Bottom of Spent Fuel Assembly at full uplift 11 328.389 222.7 15.639 Top of Assembly Lying on Bail Handles 12 327.943 217.3 15.193 Top of Assembly Bail Handle in Spent Fuel Pool 13 327.542 212.5 14.792 Top of Fuel Rack Cell II 14 327.388 210.7 14.638 Top of Fuel Rod in Spent Fuel Rack 15 326.375 198.5 13.625 Top of Active Fuel in Spent Fuel Rack 16 312.750 35.0 0.000 Bottom of Spent Fuel Pool 17 296.417 -161.0 -16.333Top of Core 18 l 19 21.611 Coverage over Assembly Lying on Bails With Water Level at T.S. Limil 20 23.625 Coverage over Active Fuel With Water Level at T.S. Limil i__I _

21 22.612 Coverage over Top of Fuel Rod With Water Level at T.S. Limit I _

22 32.292 Drop Distance over Top of Core [Less than GESTAR 34 ft. drop assumption[

23 1.167 Drop Distance over Spent Fuel Racks l l 25

References:

l l _ l l_=

26 Fuel Assembly Dimensions (Inches) Based on GE DWG No. 829E431, Rev. 2 27 as per LGS Drawing SDOC-340-H-VC-00021, Rev. 0 l l l 28 176.16 Maximum Fuel Assembly Length Il lll l 29 5.348 Assembly thickness, lying on its side (from GE DWG No. 107E1593, Rev. 1) - average of 5.226 30 Upper End Fitting width and 5.47 maximum overall bundle width l l_ _

31 l l1 1 lll 1

32 Technical Specification 3.9.8 LCO on Water Level - Reactor Vessel 34 UFSAR Section 9.1.2.2.2.1 for normal water level. @_l__ _

35 l_ l l_ l llll 36 Drawing C-775 for bottom of pool elevation l_lll_

37 _ _ _ _ _ _ __ __ _ _ _I_ _

38 GP-6,"Shutdown Operations - Refueling, Core Alteration and Core Off- - _

39 I loading", Attachment 4 for 9 Inch distance from full up to top of vessel l l 40 flange. l l Water Coverage Assessment Calc. No. LM-0645, Rev. I, Attachment E, Page E-7 of E-8

A l a C D E lF H 1 LGS Damaged Fuel Wate 2T 3 Reference Points 4 MSL Water level SPF Bottom 5 (feet) (Inches) (feet) 6 352 B7+12 .C$10+(AS-ASIO) Refuel Floor 7 351 494 .CS10+(A7-ASIO) Normal Spent Fuel Pool Water Level 8 350 =87.12 *CSI0+(A8-ASI0) Tech Spec 3.9.8 waler level for 22 feet a 9 =A7-(B7-B9y12 217.5 .CS1+0 A9-AS10) Elevation at Reactor Vessel Flange 10 =A90..75 _39+9 gC114A10-AI1 Bottom ofSpent Fuel Assembly atfuluF 11 *AS16 .fC11-C$I6) BS7-(AS7-AlI)1t2 .C12+SA29112 ToP of Assembly Ly in on Ball Handles 12 =A$16+.Ct2-CS16) =B$7-(AS7-A12)'12 *182.32112 ToP ofAssembly Bail Handle inSpent F 13 *AS16+(C13-CS16) B57jAS7-A13)'12 .177.5112 Top of Fuel Rack Cell 14 *AS164(C14-C16) -B57-(AS7-A14)'12 .175.655112 Top of Fuel Rod in Spent Fuel Rack 15 *AS16 (C15-CS16) *BS7 (AS7A15) 12 .163.5112 Top ofActive Fuel InSpent FuelRack 16 312.75 =BS7 (AS7-A16) 12 0 Bottom of Spent Fuel Pool 17 *A16-(B16-B1 p12 -161 .A17-A16 Top ofCore 18 19 *CS-CII11 Coverag over Assemby Ly 20 -CB-C15 Coverage over Active Fuel 21 *C8-C14 Coverage over Top of Fuel F 22 *A1O-A17 Drop Distance over Top of C 23 *A10 A13 Drop Distance over Spent F 24 25

References:

26 Fuel Assembil Dimensionse 27 as per LGS Drawing SDOC __

28 176.16 Maximum Fuel Assembly Leo 29 *5.226+(5.47-5.226p2 Assembly thickness, Ilyingp 30 Upper End Fitting and 5.47 _ _

3i1 32 Technical Specification 3.9.8 34 UFSAR Section 9.1.2.2.2.1 f 36 Drawing C-775 for bottom of 37 _ _

38 GP-6.Shutdown Operations -Refueling, Core Alteration and Core Off-loading", Attachment 4 for 9 Inch distance from full up to top of 39 vessel flange.

401 Water Cov Assessment Formula Calc.No.LM-0645. Rev. I, Attachment E, Page E.8of E-8

A I B I C 1 Attachment G: Evaluation of Bounds for Other Previously Considered FHA Scenarios 2 _

3 Scenario 4 LM-0656 LM-0657 Calculation Number 5 Crane RCWP Jib Scenario Descriptions 6 Collision Crane Drop 7 _ Fuel Array Basis GE14 (10xl 10 with 764 assemblies in core 8 _92 92 [Total Rods / Assembly 9 87.33 I 87.33 IEffective Rods / Assembly with 14 2/3 length rods 10 184 92 Fuel Rods Damaged in Dropped Assemblies (all rods) 11 155.7 200 Pool DF applicable to Dropped Assemblies (from Attachment E) 12 1.7 1.7 Peaking Factor 13 58 206 FuelRods Damaged from being struck by dropped assemblies or loads (limiting) 14 186.9 200 Pool DF applicable to Struck Assemblies (from Attachment E) 15 0.007604 T 0.007593 1Damaged Core Fraction Multiplied byqPF _______ _ _

16 0.007608 New Dose Analysis Basis Damaged Core Fraction Multiplied by PF 17 Therefore, these scenarios are bounded by AST Dose Assessment.

19 Notes: __________________ _

20 __ 1. Fuel Damage Assessment are from the cited calculations unless otherwise indicated below. Both 2 1c calculation use GESTAR II damage assessment methodology.

22 2. The worst case fuel damage for struck assemblies in LM-0656 is 42 rods. However, to determine 23 . the allowable margin for, e.g., future use of a Heavy Mast or fuel weight increases, a 58 rod 24 . damage assumption is used.

25 3. The worst case fuel damage for struck assemblies in LM-0657 is 154 rods with 68 damaged by 26 _ the dropped assembly and 86 by the dropped Jib Crane and Load. As shown, a total of 206 fuel 27 rods in Struck Assemblies can fail before the dose analysis basis is exceeded.

CaIc. No. LM-0645, Rev. 1, Attachment G, Page G-1 of G-3

A I B T _ Attachment G: Evaluation of Bounds for Other Previously Con 3 Scenario 4 LM-0656 LM-0657 5 Crane RCWP Jib 6 Collision Crane Drop 7 __ _ __ _ _____ _ _Fuel Array Basis GE14 (10x10), with 764 assembliE 8 92 [92_

9 87.33 187.33 10 184 92 11 155.7 200 12 1.7 1.7 13 58 _06 14 186.9 200 15 =(A1 0/A9/764)*A1 2*(200/A1 )+(Al 3/A91764)*A1 2*(200/A14) 1(B101B91764)*BI 2*(200IB1I )+(B 13/B9/764)*B1 2*(200/B14) 16 0.007608 178 _ _ __ ___ _ ____ ____

19 Notes: ___ __ _____ __ ___ __-

20 1. Fuel Damage Assessment are from the cited calculations unless 21 methodology.

22 2. The worst case fuel damage for struck assemblies in LM-0656 i 23 _ ___________ _ Heavy Mast or fuel weight increases, a 58 rod damage assumpti 24 _ _ _ _____ __ _

25 _ _ 3. The worst case fuel damage for struck assemblies in LM-0657 is 26 _ ___ Crane and Load. As shown, a total of 206 fuel rods in Struck As 27 CaIc. No. LM-0645, Rev. 1, Attachment G, Page G-2 of G-3

C 1 sidered FHA Scenarios 2

3 4 Calculation Number =_ __ ___

5 Scenario Descriptions 6

7 s in core _ _ __ __ _ -

8 Total Rods/ Assembly_________________ __-

9 Effective Rods / Assembly with 14 2/3 length rods 10Fuel Rods Damaged in Dropped Assemblies (all _rods)__ _ __

11 Pool DF applicable to Dropped Assemblies (from Attachment E)L 12 Peaking Factor 13 Fuel Rods Damaged from being struck by dEopped assemblies or loads (limiting)_

14 Pool DF applicable to Struck Assemblies (from Attachment E) 15 Damaged Core Fraction Multiplied by PF-16 New Dose Analysis Basis Damaged Core Fraction Multiplied by PF ____

17 Therefore, these scenarios are bounded by AST Dose Assessment.

18 19 _

20 otherwise indicated below. Both calculation use GESTAR II damage assessment 21 ____

22 42 rods. However, to determine the allowable margin for, e.g., future use of a 23 Dn is used.

24 __ _ ___ ____

25 154 rods with 68 damaged by the dropped assembly and 86 by the dropped Jib 26 emblies can fail before the dose analysis basis is exceeded.

27 Calc. No. LM-0645, Rev. 1, Attachment G, Page G-3 of G-3

H

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34 i

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ADDITIONAL ATTACHMENTS TO 10-10-05 Letter: SupPlement to Request for LAR Application of AST Attachment 010 AST - Drawing M-01 02 (1 of 2).

ADDITIONAL ATTACHMENTS TO 10-10-05 Letter: Supplement to Request for LAR Application of AST Attachment 011 AST - Drawing M-0107 (2 of 2).

v t e Letter Sequence Request
TAC:MC2295, Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance Supplement, Supplement, Supplement Administration Meeting, Meeting Questions 8 October 2004 30 June 2005 18 August 2005 3 April 2006 27 April 2006 Answers Results Approval, Approval Other:
MONTHYEAR2005-09-21 [Table View]

ML053330373

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