Text

License Amendment Request Pursuant to 10 CFR 50.90: Alternative Source Term Update of Design Basis Analysis for Fuel Handling Accident
	ML023610200
Person / Time
Site:	Perry
Issue date:	12/09/2002
From:	Kanda W; FirstEnergy Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	PY-CEI/NRR-2674L
	Download: ML023610200 (157)
	v • d • e

FENOC PerryNuclear Power Plant 10 CenterRoad FirstEnergyNuclearOperatingCompany Perry Ohio 44081 440-280-5579 William R. Kanda Fax 440-280-8029 Vice President- Nuclear December 9, 2002 PY-CEI/NRR-2674L United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Perry Nuclear Power Plant Docket No. 50-440 License Amendment Request Pursuant to 10 CFR 50.90: Alternative Source Term Update of the Design Basis Analyses for the Fuel Handling Accident Ladies and Gentlemen:

A license amendment is requested for the Perry Nuclear Power Plant (PNPP). The requested change utilizes Alternative Source Term radiological calculations to update the design basis analysis for the Fuel Handling Accident. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", was utilized in the development of this application.

Approval is requested by March 14, 2003, to support preparations for the ninth refueling outage. This application is considered a cost beneficial licensing change due to anticipated cost savings on outage duration, due to redefinition of the term "recently irradiated fuel." If you have questions or require additional information, please contact Mr. Vernon K. Higaki, Manager - Regulatory Affairs, at (440) 280-5294.

Very truly yours,

Enclosures:

1. Notarized Affidavit 2 Evaluation of the changes, including a Summary, Description of the Changes, Background, Technical Analysis, Regulatory Analysis/Commitments, and Environmental Consideration

3. Significant Hazards Consideration

4. Dose Calculation entitled "Fuel Handling Accident Using Alternative Source Term"

5. Information copy of proposed Updated Safety Analysis Report (USAR) changes (mark-up) cc- NRC Project Manager NRC Resident Inspector NRC Region III State of Ohio ACO

Enclosure 1 PY-CEI/NRR-2674L Page 1 of 1 I, William R. Kanda, hereby affirm that (1) I am Vice President - Perry, of the FirstEnergy Nuclear Operating Company, (2) I am duly authorized to execute and file this certification as the duly authorized agent for The Cleveland Electric Illuminating Company, Toledo Edison Company, Ohio Edison Company, and Pennsylvania Power Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.

Subscribed to and affirmed before me, the 9 day of AA f rý , oo 3

/ o6. (2ArAoL 4t(CO.

Enclosure 2 PY-CEI/NRR-2674L Page 1 of 24 Summary The amendment proposed to the Perry Nuclear Power Plant (PNPP) license is based on a new dose analysis for the design basis Fuel Handling Accident, using an Alternative Source Term (AST).

Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", was utilized in the development of this application.

The analysis assumes the event occurs after only one day (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) of radiological decay, rather than seven days as was assumed in License Amendment 102 (issued in March 1999). Even with this 24-hour assumption, the doses from such an event remain within regulatory acceptance limits.

Description of the Changes A license amendment is requested based on 10 CFR 50.67.b(1) and the Nuclear Regulatory Commission (NRC) position in Section 1.1.1 of Regulatory Guide 1.183 that an initial AST implementation should be approved per 10 CFR 50.90. This is the initial use of an alternative source term for the Fuel Handling Accident design basis analysis at PNPP. Therefore, the calculation is provided in Enclosure 4 to this submittal for NRC review and approval of a license amendment per 10 CFR 50.90.

Background

In a Federal Register Notice dated December 23, 1999, the Nuclear Regulatory Commission (NRC) published a new regulation, 10 CFR 50.67, providing a mechanism for licensed power reactors to replace the traditional accident source term used in design basis accident analyses with Alternative Source Terms (ASTs). Regulatory guidance for the implementation of these ASTs is provided in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", dated July 2000. 10 CFR 50.67(b)(1) states that licensees who seek to revise their current accident source term in design basis radiological consequence analyses should apply for a license amendment under 10 CFR 50.90.

Two previous PNPP license amendments have laid the groundwork for the current license amendment request. License Amendment 102, issued in March 1999, introduced the concept that Shutdown Safety administrative controls can be utilized during fuel handling once the dose calculations demonstrate that regulatory limits for the Fuel Handling Accident can be met without credit for filtration systems and the Containment/Fuel Handling Buildings. License Amendment 103, also issued in March 1999, involved a pilot plant application of an alternative source term for a design basis Loss Of Coolant Accident (LOCA). Subsequent to March of 1999, significant consideration was given to the characteristics of the AST for a Fuel Handling Accident (FHA), which is different from a LOCA. The results of these considerations were published by the NRC in Regulatory Guide 1.183, which is the primary regulatory basis document for the currently proposed change. This current request therefore applies the AST characteristics of a Fuel Handling Accident to the PNPP radiological calculations for handling fuel that has been subcdtical for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Technical Analysis The following table summarizes conformance to Regulatory Guide 1.183, to ensure the guidance is adequately addressed. This supplements the actual calculation, which is included as Enclosure 4.

Regulatory Guide 1.183 Guidance Degree of Conformance Section 1 Implementation of AST 1.1.1 Safety Margins "The proposed uses of an AST and the associated Sufficient safety margins are maintained with the

Enclosure 2 PY-CEI/NRR-2674L Page 2 of 24 RiniIutnr, I IR flat, ran af ranfn in, - -

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proposed facility modifications and changes to Alternative Source Term analyses. There are a procedures should be evaluated to determine number of conservatisms in the calculations, which whether the proposed changes are consistent with account for analysis uncertainties. The primary the principle that sufficient safety margins are uncertainties in the calculation consist of the maintained, including a margin to account for inventory released from the fuel, the scrubbing of the analysis uncertainties. The safety margins are nuclides from the water pool over the-fuel, the products of specific values and limits contained in efficiency of filtration provided by plant structures the technical specifications (which cannot be and systems, and dispersion of the release as it changed without NRC approval) and other values, travels away from the plant. The inventory available such as assumed accident or transient initial in the fuel is determined using an NRC accepted conditions or assumed safety system response code (see Section 3.1 below) which is conservative times. Changes, or the net effects of multiple enough to address uncertainties in the inventory, changes, that result in a reduction in safety margins and in the radiological decay process. The fraction may require prior NRC approval. Once the initial of that inventory which is available in the gap of the AST implementation has been approved by the staff fuel rods is assumed to be the same as provided in and has become part of the facility design basis, the Regulatory Guide 1.183 (see Section 3.2 below),

licensee may use 10 CFR 50.59 and its supporting which addresses uncertainties in the gap fraction.

guidance in assessing safety margins related to subsequent facility modifications and changes to The uncertainties of the scrubbing provided by the procedures." water is addressed by using the overall decontamination factor (DF) of 200 documented in Regulatory Guide 1.183. This is a conservative value. Several plants that have submitted Alternative Source Term analyses have shown that the overall DF provided by the water over the fuel is actually greater than the 200 value. The requirements for water coverage over the fuel in the Technical Specifications remain unchanged by this proposal.

The uncertainties in the efficiency of filtration systems to treat the release are addressed by assuming there are no Containment or Fuel Handling Buildings or ventilation/filtration systems present, and the release from the pool to the environment is an instantaneous, undiluted, and unfiltered release. The release is then dispersed by ChVQ values previously approved by the NRC, which were considered to adequately address uncertainties in the actual dispersion of the release.

The dose calculation results remain below the limits of 10 CFR 50.67. Table 7 of Enclosure 4 presents the results of the base calculation (Table 8 presents sensitivities), along with the applicable dose limits for the Control Room, Exclusion Area Boundary (EAB),

and the Low Population Zone (LPZ):

TABLE 7 RESULTS Control Room EAB LPZ RADTRAD Results (rem) 1.03 1.44 0.161 Regulatory limit (rem) 5 6.3 6.3

Enclosure 2 PY-CEI/NRR-2674L Page 3 of 24 fl flf Cnnfrrmn,-e I te HI ato- VI Gud 1 183 Gv irin-*o*, nn%* nan nf Vr lilVIIIIrr iln~

1.1.2 Defense in Depth "The proposed uses of an AST and the associated Adequate defense in depth is maintained through the proposed facility modifications and changes to use of Technical Specification controls over buildings procedures should be evaluated to determine and filtration systems when fuel being handled is whether the proposed changes are consistent with "recently irradiated fuel", and using the controls the principle that adequate defense in depth is described and approved in License Amendment 102, maintained to compensate for uncertainties in when fuel is not recently irradiated. These are in accident progression and analysis data. addition to the natural defenses of radiological decay Consistency with the defense-in-depth philosophy is over time (which reduces the magnitude of any maintained if system redundancy, independence, release) and the scrubbing effect of the water pool and diversity are preserved commensurate with the over the fuel (the controls over water level are not expected frequency, consequences of challenges to being changed as a result of this amendment). As the system, and uncertainties. In all cases, noted above, the radiological calculations show that compliance with the General Design Criteria in Fuel Handling Accident doses remain within Appendix A to 10 CFR Part 50 is essential. regulatory acceptance limits.

Modifications proposed for the facility generally should not create a need for compensatory programmatic activities, such as reliance on manual operator actions ...

Proposed modifications that seek to downgrade or Fuel Handling Accidents are not modeled by the remove required engineered safeguards equipment PNPP Probabilistic Safety Analysis (PSA). However, should be evaluated to be sure that the modification risk is assessed and managed in accordance with does not invalidate assumptions made in facility plant procedure, which uses defense in depth criteria PRAs and does not adversely impact the facility's and other considerations as specified in the controls severe accident management program." put in place by License Amendment 102. Similarly, the Severe Accident Management (SAM) Program would not be affected by the proposed change, based on the maintenance of the controls put in place per Amendment 102 over buildings and filtration systems during Dlant shutdowns.

1.1.3 Integrityof FacilityDesign Basis

"...Although a complete re-assessment of all facility See further discussion under Regulatory Position 1.3 radiological analyses would be desirable, the NRC below, and see the example markups of the Updated staff determined that recalculation of all design Safety Analysis Report (USAR) provided for analyses would generally not be necessary. information in Enclosure 5.

Regulatory Position 1.3 of this guide provides guidance on which analyses need updating as part These note that this application is considered to be a of the AST implementation submittal and which may selective application of the AST. The USAR need updating in the future as additional markups note that the source term assumptions and modifications are performed. This approach would radiological criteria in the previous Fuel Handling create two tiers of analyses, those based on the Accident analyses have been superceded by the previous source term and those based on an AST. new analyses, and future revisions of Fuel Handling

... In either case, the facility design bases should Accident analyses will use the updated source term clearly indicate that the source term assumptions assumptions and radiological criteria.

and radiological criteria in these affected analyses have been superseded and that future revisions of these analyses, if any, will use the updated approved assumptions and criteria. ... "

1.1.4 Emergency PreparednessApplications

"...The AST is not representative of the wide No relief is being requested from emergency spectrum of possible events that make up the planning provisions.

planning basis of emergency preparedness.

Therefore, the AST is insufficient by itself as a basis Procedures are already in place for responding to a

Enclosure 2 PY-CEI/NRR-2674L Page 4 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance for requesting relief from the emergency fuel handling accident as a result of License preparedness requirements of 10 CFR 50.47 and Amendment 102.

Appendix E to 10 CFR Part 50. This guidance does not, however, preclude the appropriate use of the insights of the AST in establishing emergency response procedures such as those associated with emergency dose projections, protective measures, and severe accident management guides."

1.2.1 Full Implementation This application is not considered to be a full implementation. See Section 1.2.2 below.

1.2.2 Selective Implementation "Selective implementation is a modification of the This application is a selective implementation. It facility design basis that (1) is based on one or entails re-evaluation of a limited subset of the more of the characteristics of the AST or (2) entails design basis radiological analyses, specifically the re-evaluation of a limited subset of the design basis Fuel Handling Accident. The only DBA that needs radiological analyses. The NRC staff will allow to be reanalyzed for this amendment is the Fuel licensees flexibility in technically justified selective Handling Accident. The description of how the implementations provided a clear, logical, and design basis for this event is being maintained is consistent design basis is maintained. An example included in the example USAR markups in of an application of selective implementation would Enclosure 5.

be ... a request to remove the charcoal filter media from the spent fuel building ventilation exhaust. For the latter, the licensee may only need to re-analyze DBAs that credited the iodine removal by the charcoal media. Additional analysis guidance is provided in Regulatory Position 1.3 of this guide.

NRC approval for the AST (and the TEDE dose It is understood that since this is a selective criterion) will be limited to the particular selective application of the AST for the Fuel Handling implementation proposed by the licensee. The Accident, NRC approval will be limited to this event.

licensee would be able to make subsequent Use of AST to change the design basis for other modifications to the facility and changes to events such as the Control Rod Drop Accident or procedures based on the selected AST the Main Steam Line Break Outside Containment, characteristics incorporated into the design basis or changes to the approved AST characteristics, under the provisions of 10 CFR 5Q.59. However, would require prior staff approval under 10 CFR use of other characteristics of an AST or use of 50.67.

TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, would require prior staff approval under 10 CFR 50.67..."

1.3.1 Design Basis RadiologicalAnalyses "There are several regulatory requirements for 10 CFR 50.49 Environmental Qualification of which compliance is demonstrated, in part, by the Equipment - No credit is taken for filtration system evaluation of the radiological consequences of OPERABILITY (or OPERABILITY of any other design basis accidents. These requirements system) in the design basis calculations for the Fuel include, but are not limited to, the following. Handling Accident after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of radiological

+ Environmental Qualification of Equipment (10 decay. Therefore, there is not a concern that some CFR 50.49) aspect of the alternative source term could make

+ Control Room Habitability (GDC-19 of Appendix such systems unable to perform a "credited" safety A to 10 CFR Part 50) function.

Emergency Response Facility Habitability GDC 19 Control Room Habitability- For a Fuel (Paragraph IV.E.8 of Appendix E to 10 CFR Handling Accident, a design basis dose calculation Part 50) for the Control Room was performed assuming 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of radiological decay. The base calculation

Enclosure 2 PY-CEI/NRR-2674L Page 5 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance

Alternative Source Term (10 CFR 50.67) showed that doses remained less than the 5 rem

Environmental Reports (10 CFR Part 51) Total Effective Dose Equivalent (TEDE) limit in

Facility Siting (10 CFR 100.11) GDC-19 and 10 CFR 50.67, even assuming no There may be additional applications of the credit for Containment or FHB or filtration systems.

accident source term identified in the Technical 10 CFR 50 Appendix E Emergency Response Specification bases and in various licensee Facility Habitability -The proposed changes do commitments. These include, but are not limited to, not result in changes to Emergency Response the following from Reference 2, NUREG-0737. Facility Habitability. 10 CFR 50 Appendix E does

Post-Accident Access Shielding (NUREG-0737, not contain habitability criteria, however ll.B.2) NUREG-0737 Supplement 1 does. The only facility

Post-Accident Sampling Capability (NUREG with a specific dose criterion is the Technical 0737, Il.B.3) Support Center (TSC). The dose limit in

Accident Monitoring Instrumentation (NUREG Supplement I for this facility is 5 rem whole body, 0737, II.F.1) or its equivalent. The "or equivalent" for this

Leakage Control (NUREG-0737, III.D.1.1) evaluation is considered to be 5 rem TEDE.

Emergency Response Facilities (NUREG-0737, Although the TSC has essentially no response III.A.1.2) function for a Fuel Handling Accident, a scoping

+ Control Room Habitability (NUREG-0737, study for the TSC was performed. The ventilation intakes for the TSC are farther away from the II1.D.3.4)."

containment structure and from ventilation system release points than the Control Room intakes, and the TSC intake is at a lower elevation by more than 60 feet. Since the dispersion of a plume for an intake at a greater distance and lower elevation would be correspondingly better, the scoping evaluation concluded that the 5 rem TEDE limit would be met for the TSC as well. The regulatory guidance does not include specific dose limits for Emergency Operations Facility (EOF) and backup EOF habitability. For the same reasons as discussed for the TSC, these facilities are also considered to not be adversely affected as a result of this change in the source term assumptions.

10 CFR 50.67 Accident Source Term - The acceptance criteria of 10 CFR 50.67 and the attributes of an acceptable alternative source term as described in Regulatory Guide 1.183 are being utilized in this application.

10 CFR Part 51 Environmental Protection Regulations - See the section of this letter entitled "Environmental Consideration" below.

10 CFR 100.11 Facility Siting -As noted in Footnote 5 of Reg. Guide 1.183, the dose guidelines of 10 CFR 100.11 are superceded by 10 CFR 50.67 for applications implementing an alternative source term such as this.

NUREG-0737 Item ll.B.2 Post-Accident Access Shielding - There are no design basis actions credited outside the Control Room for a Fuel Handling Accident. TSC access/dose was addressed above.

NUREG-0737 Item ll.B.3 Post-Accident Sampling Capability - No post-accident sampling inside the I Containment is required for a Fuel Handlina m

Enclosure 2 PY-CEI/NRR-2674L Page 6 of 24 Rcniktnn, (iiidc I IR (iirlnt-i fl nf Renulato- Guide 1 1 AA r."'irinnna r)g% ran rtf rrnfermnnt-n Accident.

Accident Monitoring Instrumentation (NUREG-0737, II.F.1) - No post-accident monitors are required to respond to a Fuel Handling Accident.

NUREG-0737 Item II1.D.1.1 Leakage Control - No post-accident leakage control is required for a Fuel Handling Accident.

NUREG-0737 Item II1.A.1.2 Emergency Response Facilities - Item III.A.1.2 is unaffected, since no dose protection or habitability guidance is included in this TMI item. See discussions above on Emergency Response Facilities.

NUREG-0737, Item II1.D.3.4 Control Room Habitability - Control Room habitability was analyzed and determined to be acceptable, by meeting the radiological dose limits of 10 CFR 50.67. The proposed amendment does not affect protection from toxic gases.

No additional applications of the accident source term for a Fuel Handling Accident were identified in the Technical Specification Bases or in licensee commitments.

1.3.2 Re-Analysis Guidance "Any implementation of an AST, full or selective, and The change consists of a redefinition of the term any associated facility modification should be "recently irradiated fuel" to be consistent with the supported by evaluations of all significant new dose calculations. Compliance with various radiological and non-radiological impacts of the regulations and commitments are addressed in this proposed actions. This evaluation should consider Table. Technical Specification controls exist if the impact of the proposed changes on the facility's "recently irradiated fuel" is handled; the controls put compliance with the regulations and commitments in place per Amendment 102 continue to apply listed above as well as any other facility-specific during other periods of fuel handling.

requirements. These impacts may be due to (1) the associated facility modifications of (2) the differences in the AST characteristics.

The scope and extent of the re-evaluation will The design basis FHA calculation has been updated necessarily be a function of the specific proposed and is included in Enclosure 4 for NRC review. This facility modification and whether a full or selective selective implementation is solely for the Fuel implementation is being pursued. The NRC staff Handling Accident. Other design basis calculations does not expect a complete recalculation of all were determined to not be affected by this proposed facility radiological analyses, but does expect license amendment. Example USAR markups are licensees to evaluate all impacts of the proposed also provided for information in Enclosure 5.

changes and to update the affected analyses and the design bases appropriately. An analysis is considered to be affected if the proposed modification changes one or more assumptions or inputs used in that analysis such that the results, or the conclusions drawn on those results, are no longer valid. Generic analyses, such as those performed by owner groups or vendor topical reports, may be used provided the licensee justifies the applicability of the generic conclusions to the

Enclosure 2 PY-CEI/NRR-2674L Page 7 of 24 T

Rea]ulatorv Guide 1.183 Guid~ntr*. Dnr of Cnnfnrmnr Realaor 11--Gida-e-----o-Cnfrmnc -id specific facility and implementation. Sensitivity analyses, discussed below, may also be an option.

If affected design basis analyses are to be re In the calculation, all affected assumptions and calculated, all affected assumptions and inputs inputs were updated to address AST and TEDE, and should be updated and all selected characteristics of all selected characteristics of the AST and the TEDE the AST and the TEDE criteria should be addressed. criteria are addressed.

The license amendment request should describe the Statements regarding the acceptability of the licensee's re-analysis effort and provide statements proposed amendment against each of the applicable regarding the acceptability of the proposed items identified in Regulatory Position 1.3.1 of the implementation, including modifications, against Reg. Guide were provided above.

each of the applicable analysis requirements and commitments identified in Regulatory Position 1.3.1 The above discussion addressed radiological impact of this guide. ... " of the proposed change. Since there are no physical design modifications being made in conjunction with this proposal, there are also no non-radiological impacts as a result of the proposed change.

i 1.3.3 Use of Sensitivity or Scoping Analyses "It may be possible to demonstrate by sensitivity or No sensitivity evaluations that varied AST scoping evaluations that existing analyses have characteristics were performed.

sufficient margin and need not be recalculated. As used in this guide, a sensitivity analysis is an However, several sensitivity evaluations were evaluation that considers how the overall results performed which varied Control Room ventilation vary as an input parameter (in this case, AST assumptions to show doses remained acceptable.

characteristics) is varied. A scoping analysis is a In the base case, normal ventilation continues to brief evaluation that uses conservative, simple operate throughout the event, which initially brings methods to show that the results of the analysis the undiluted, unfiltered source term directly into the bound those obtainable from a more complete Control Room without any isolation protection. This treatment. Sensitivity analyses are particularly case takes no credit for the Control Room Area applicable to suites of calculations that address Radiation Monitor or the Emergency Recirculation diverse components or plant areas but are otherwise (filtration) system. Two other calculation sensitivity largely based on generic assumptions and inputs. cases were also run, which isolated the control room Such cases might include post-accident vital area at the worst possible time, after the source term access dose calculations, shielding calculations, and available at the intake is introduced into the Control equipment environmental qualification (integrated Room. For these two cases, the isolation exists for dose). It may be possible to identify a bounding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and then was followed by either a case, re-analyze that case, and use the results to subsequent re-initiation of the normal intake flow, or draw conclusions regarding the remainder of the the use of the filtration system. Each case produced analyses. It may also be possible to show that for an acceptable result, showing that following a Fuel some analyses the whole body and thyroid doses Handling Accident, the operators have flexibility on determined with the previous source term would how to operate their ventilation systems without bound the TEDE obtained using the AST. Where exceeding the radiological acceptance criteria. The present, arbitrary "designer margins" may be base case shows that filtration systems are not adequate to bound any impact of the AST and TEDE required. The two sensitivity cases show that criteria. If sensitivity or scoping analyses are used, ventilation/filtration systems can be effectively used the license amendment request should include a to reduce doses to the Control Room operators in discussion of the analyses performed and the the event that the radiation monitor isolates the conclusions drawn. Scoping or sensitivity analyses Control Room intake at the worst possible time after should not constitute a significant part of the available activity is taken into the Control Room.

evaluations for the design basis exclusion area boundary (EAB), low population zone (LPZ). or m 1 r I I 1 * -

A scoping evaluation is also used to show that the

Enclosure 2 PY-CEI/NRR-2674L Page 8 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance control room dose." TSC doses would be lower than the Control Room doses since the TSC inlet is farther away from the plant vents and the Containment itself than the Control Room inlets are, and is lower on the buildings by more than 60 feet, so the dispersion factors would be better than the previously NRC approved dispersion factors for the Control Room intakes.

1.3.4 UpdatingAnalyses Following Implementation "Full implementation of the AST replaces the This is a selective implementation rather than a full previous accident source term with the approved implementation.

AST and the TEDE criteria for all design basis radiological analyses. ... Since [for a full Since the USAR discussions of the Fuel Handling implementation] the AST and the TEDE criteria are Accident will include the AST and TEDE criteria (see part of the approved design basis for the facility, use Enclosure 5), future updates to Fuel Handling of the AST and TEDE criteria in new applications at Accident calculations will continue to use the the facility do not constitute a change in analysis characteristics of the AST and TEDE under the methodology that would require NRC approval. This provisions of 10 CFR 50.59.

guidance is also applicable to selective implementations to the extent that the affected analyses are within the scope of the approved implementation as described in the facility design basis. In these cases, the characteristics of the AST and TEDE criteria identified in the facility design basis need to be considered in updating the analyses. Use of other characteristics of the AST or TEDE criteria that are not part of the approved design basis, and changes to previously approved AST characteristics, requires prior NRC staff approval under 10 CFR 50.67."

1.3.5 Equipment Environmental Qualification "Current environmental qualification (EQ) analyses There are no increased EQ requirements as a result may be impacted by a proposed plant modification of this proposed amendment. Further details on EQ associated with the AST implementation. The EQ are provided in Section 6.

analyses that have assumptions or inputs affected by the plant modification should be updated to address these impacts.

The NRC staff is assessing the effect of increased The cesium issue discussed in this section of the cesium releases on EQ doses to determine whether Regulatory Guide is associated with a LOCA and is licensee action is warranted ... " unrelated to a Fuel Handling Accident.

1.4 Risk Implications "The use of an AST changes only the regulatory As noted in Section 1.1.2, the PNPP PSA model is assumptions regarding the analytical treatment of not affected by the proposed amendment.

the design basis accidents. The AST has no direct effect on the probability of the accident. Use of an AST alone cannot increase the core damage frequency (CDF) or the large early release frequency (LERF). However, facility modifications made possible by the AST could have an impact on risk. If the proposed implementation of the AST involves changes to the facility design that would invalidate assumptions made in the facility's PRA, the impact I

Enclosure 2 PY-CEI/NRR-2674L Page 9 of 24 PnziItnn, (iiirIc I 1R tuir4nrc flDnrca nf (rnfrrmnei RenI*u ulato G[v uid e 1 183 G UiU*innt- nn..i.* rna ^fI ri.r/*nliW*,./ i tId.,II on the existing PRAs should be evaluated.

Consideration should be given to the risk impact of Although risk insights are not being used to support proposed implementations that seek to remove or this change, some risk insights were utilized in NRC downgrade the performance of previously required approval of License Amendment 102, which remain engineered safeguards equipment on the basis of applicable to this proposed change. --

the reduced postulated doses.

The NRC staff may request risk information if there is a reason to question adequate protection of public health and safety. The licensee may elect to use risk insights in support of proposed changes to the design basis that are not addressed in currently approved NRC staff positions .. "

1.5 Submittal Requirements

" ... The NRC staffs finding that the amendment The dose analysis calculation is being provided for may be approved must be based on the licensee's NRC review as Enclosure 4. Additional detail on analyses, since it is these analyses that will become how the NRC guidance in Regulatory Guide 1.183 is part of the design basis of the facility. The being met is provided in this table format.

amendment request should describe the licensee's analyses of the radiological and nonradiological impacts of the proposed modification in sufficient detail to support review by the NRC staff.

The staff recommends that licensees submit affected USAR pages, which include examples of the types of FSAR pages annotated with changes that reflect the changes that will be made, are also included for revised analyses or submit the actual calculation information as Enclosure 5.

documentation.

If the licensee has used a current approved version The Code used in the analysis was RADTRAD 3.02, of an NRC-sponsored computer code, the NRC staff January 5, 2000.

review can be made more efficient if the licensee identifies the code used .... "

1.6 FSAR Requirements

"... The regulations in 10 CFR 50,71(e) require that USAR pages are provided for information in the FSAR be updated to include all changes made in Enclosure 5, which provide examples of how the the facility or procedures described in the FSAR.... licensing basis will be revised as a result of this The affected radiological analysis descriptions in the proposed amendment.

FSAR should be updated to reflect the replacement of the design basis source term by the AST. The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numeric results.

The descriptions of superseded analyses should be removed from the FSAR in the interest of maintaining a clear design basis.

Section 2 Attributes Of An Acceptable AST

"...Regulatory Position 3 of this guide identifies an This application uses the characteristics of the AST that is acceptable to the NRC staff for use at source term outlined in Regulatory Position 3 of operating power reactors. A substantial effort was Reg. Guide 1.183. Therefore the rest of Section 2 is expended by the NRC, its contractors, various considered to be not applicable, since no attempt is national laboratories, peer reviewers, and others in made to define different source term characteristics performing severe accident research and in from those provided in the Reg. Guide.

developinq the source terms provided in NUREG-

Enclosure 2 PY-CEI/NRR-2674L Page 10 of 24 Regulatorv Guide 1.183 Guidance Degree of Conformance 1465 (Ref. 5). However, future research may identify opportunities for changes in these source terms. The NRC staff will consider applications for an AST different from that identified in this guide."...

Section 3 Accident Source Term 3.1 Fission ProductInventory General Electric (GE) used the computer code "The inventory of fission products in the reactor ORIGEN 2 to determine the core inventory for a core and available for release to the containment Fuel Handling Accident. This input was originally should be based on the maximum full power developed to support the power uprate and 24 operation of the core with, as a minimum, current month operating cycle amendments (License licensed values for fuel enrichment, fuel bumup, Amendments 112 and 115). The core inventory and an assumed core power equal to the current provided by GE was performed in Curies per licensed rated thermal power times the ECCS megawatt (CiIMW). The inventory was adjusted by evaluation uncertainty. The period of irradiation an additional 2% to account for evaluation should be of sufficient duration to allow the activity uncertainty.

of dose-significant radionuclides to reach equilibrium or to reach maximum values. The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN 2 (Ref. 17) or ORIGEN-ARP (Ref. 18) ...

For DBA events that do not involve the entire core, The fissionproduct inventory of each of the fuel the fission product inventory of each of the rods was determined by dividing the total core damaged fuel rods is determined by dividing the inventory by the number of fuel rods in the core. To total core inventory by the number of fuel rods in account for differences in power level across the the core. To account for differences in power level core, a radial peaking factor of 2.0 was applied.

across the core, radial peaking factors from the This simulates that the rods in the bundle being facility's core operating limits report (COLR) or dropped and the struck bundles would be the technical specifications should be applied in highest inventory rods in the core. The maximum determining the inventory of the damaged rods ... core wide radial peaking factor of 2.0 is being added into the list of reload analysis parameters that must be re-verified each cycle. [See Commitment 1 at the end of this enclosure].

For events postulated to occur while the facility is For the Fuel Handling Accident analyses performed shutdown, e.g., a fuel handling accident, radioactive for this submittal, radioactive decay from the time of decay from the time of shutdown may be modeled." shutdown was modeled.

3.2 Release Fractions

" ... For non-LOCA events, the fractions of the core Table 3 fractions were applied to the fission product inventory assumed to be in the gap for the various inventory determined as described above for the radionuclides are given in Table 3. The release rods with the maximum core radial peaking factor.

fractions from Table 3 are used in conjunction with These fractions are 8% of 1-131, 10% of Kr-85, 5%

the fission product inventory calculated with the of the other Noble Gases, 5% of the other maximum core radial peaking factor." Halogens, and 12% of the Alkali Metals.

[An applicable footnote is linked to Table 3. For footnote 11, which applies to Table 3, the Footnote 11 states "The release fractions listed provisions in the first sentence of the footnote are here have been determined to be acceptable for met at PNPP. The fuel in use at PNPP is NRC use with currently approved LWR fuel with a peak approved fuel, and the average exposure of the burnup up to 62,000 MWD/MTU provided that the peak fuel rod is maintained below 62,000 maximum linear heat generation rate does not MWD/MTU (= 62 GWD/MTU). Also, the maximum exceed 6.3 kw/ft peak rod average power for linear heat generation rate for the fuel that could burnups exceeding 54 GWD/MTU. As an exceed 54 GWD/MTU by the end of the cycle is

Enclosure 2 PY-CEI/NRR-2674L Page 11 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance alternative, fission gas release calculations maintained at or below 6.3 kwlft (in other words, the performed using NRC-approved methodologies higher burnup fuel is moved to lower power portions may be considered on a case-by-case basis. To be of the core such as the periphery). The burnup limit acceptable, these calculations must use a projected of 62 GWD/MTU on the average exposure of the power history that will bound the limiting projected peak rod, and the LHGR limit of 6.3 kw/ft peak rod plant-specific power history for the specific fuel average power for the higher burnup-fuel (> 54 load...."] GWD/MTU), are both being added into the list of reload analysis parameters that must be re-verified each cycle. [See Commitment 2]

3.3 Timing of Release Phases

"... For non-LOCA DBAs in which fuel damage is For the Fuel Handling Accident, the release from the projected, the release from the fuel gap and the fuel fuel gap is assumed to occur instantaneously with pellet should be assumed to occur instantaneously the impact of the fuel bundle.

with the onset of the projected damage...."

3.4 RadionuclideComposition Table 5 lists the elements in each radionuclide group This guidance is generic for all events. More that should be considered in design basis analyses. specific guidance for a Fuel Handling Accident is Table 5 provided in Appendix B to Reg. Guide 1.183. In Radionuclide Groups summary, only the first three groups in this table are Group Elements considered to be available in the gap for immediate Noble Gases Xe, Kr release (the Noble Gases, the Halogens, and the Halogens I, Br Alkali Metals). However, the Alkali Metals (Cesium Alkali Metals Cs, Rb and Rubidium) are particulates that have an infinite Tellurium Group Te, Sb, Se, Ba, Sr decontamination factor (i.e., they are fully retained Noble Metals Ru, Rh, Pd, Mo, Tc, Co by the water in the fuel pool or reactor cavity).

Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr Twenty of the most significant Noble Gases and Sm, Y, Cm, Am Halogens are used in the calculation for a Fuel Cerium Ce, Pu, Np Handling Accident (see Enclosure 4). The other nuclides in these groups were not included because their core activity was less than 1E-9 Ci/MWt, and were considered insignificant.

3.5 Chemical Form

" ... The accident-specific appendices to this Specific details on Chemical Form for Fuel Handling Regulatory Guide provide additional details." Accidents are in the Appendix B discussions below.

3.6 Fuel Damage in Non-LOCA DBAs

" ... The amount of fuel damage caused by a FHA is See the fuel pin failure discussion below for the addressed in Appendix B of this guide. Appendix B items.

Section 4 Dose CalculationalMethodology "The NRC staff has determined that there is an The Total Effective Dose Equivalent (TEDE) criteria implied synergy between the ASTs and total are utilized in this AST application, which is effective dose equivalent (TEDE) criteria, and performed pursuant to 10 CFR 50.67.

between the TID-14844 source terms and the whole body and thyroid dose criteria, and therefore, they do not expect to allow the TEDE criteria to be used with TID-14844 calculated results. The guidance of this section applies to all dose calculations performed with an AST pursuant to 10 CFR 50.67."

4.1 Offsite Dose Consequences

"...4.1.1 ... TEDE is the sum of the committed The TEDE dose calculations considered the effective dose equivalent (CEDE) from inhalation radionuclides, including progeny from the decay of and the deep dose equivalent (DDE) from external parent radionuclides, which are significant with exposure. The calculation of these two components regard to dose consequences and the released of the TEDE should consider all radionuclides, radioactivity. All the isotopes of bromine, iodine,

Enclosure 2 PY-CEI/NRR-2674L Page 12 of 24 Reaulatorv Guide 1.183 Guidance Denree of Conformance

_____ __________ _____ _____ _____ of--------------

including progeny from the decay of parent I krypton, and xenon with core activity greater than radionuclides, that are significant with regard to dose 1E-9 Ci/MWt (a total of 20) and their daughters, i.e.,

consequences and the released radioactivity." an additional three isotopes of cesium and rubidium, were used.

4.1.2 The exposure-to-CEDE factors for inhalation The conversion factors utilized for the CEDE of radioactive material should be derived from the inhalation component (of TEDE) were obtained data provided in ICRP Publication 30, "Limits for from the 1989 printing of Federal Guidance Intakes of Radionuclides by Workers" (Ref. 19). Report 11.

Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 20),

provides tables of conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the CEDE.

4.1.3 For the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the breathing rate of The recommended offsite Exclusion Area Boundary persons offsite should be assumed to be 3.5 x 10"4 (EAB) and Low Population Zone (LPZ) breathing cubic meters per second. From 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months rates were used, however, considering that the following the accident, the breathing rate should be release would occur instantaneously, the effective assumed to be 1.8 x 104 cubic meters per second. breathing rate used was 3.5E-4 m3 /s.

After that and until the end of the accident, the rate should be assumed to be 2.3 x 104 cubic meters per second."

4.1.4 The DDE should be calculated assuming The conversion factors utilized for the DDE/EDE submergence in semi-infinite cloud assumptions external component (of TEDE) were obtained from with appropriate credit for attenuation by body the MACCS2 computer code, which uses the 1993 tissue. The DDE is nominally equivalent to the version of Federal Guidance Report 12.

effective dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly.

Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE. Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 21),

provides external EDE conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses corresponding to the EDE.

4.1.5 ... The maximum EAB TEDE for any two-hour The activity is conservatively assumed to be period following the start of the radioactivity release immediately released (a puff release rather than a should be determined ... by calculating the 2-hour period). Atmospheric dispersion of postulated dose for a series of small time radioactivity during transport was accounted for by increments and performing a "sliding" sum over the using the PNPP dispersion factors (ChiIQ), but the increments for successive two-hour periods. The release was transported to the EAB and the LPZ maximum TEDE obtained is submitted ... (see also immediately, without delay or deposition on the Table 6). ground. Therefore, it was not necessary to perform sliding sums. Table 6 of Reg. Guide 1.183 identifies the FHA analysis release duration as 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The puff release replaced that assumption,

Enclosure 2 PY-CEI/NRR-2674L Page 13 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance and is considered conservative.

4.1.6 TEDE should be determined for the most The TEDE dose was determined for the most limiting receptor at the outer boundary of the low limiting receptor at the outer boundary of the LPZ.

population zone (LPZ) and should be used in The results, and the 10 CFR 50.67 limits, are determining compliance with the dose criteria in 10 presented in Table 7 of Enclosure 4:

CFR 50.67.

4.1.7 No correction should be made for depletion of No credit was taken in the calculations for the effluent plume by deposition on the ground." deposition of the radionuclides on the ground.

4.2 ControlRoom Dose Consequences All the radioactivity released from the pool is

"...4.2.1 The TEDE analysis should consider all assumed to be immediately transported outside of sources of radiation that will cause exposure to the Containment without dilution. Contamination of control room personnel. The applicable sources will the Control Room atmosphere by the intake of the vary from facility to facility, but typically will include: available radioactive material contained in the

Contamination of the control room atmosphere radioactive plume was modeled. Infiltration in by the intake or infiltration of the radioactive addition to the 6600 cfm of unfiltered intake was not material contained in the radioactive plume incorporated, since there are no in-plant pathways released from the facility, that can transport activity to within the Control

Contamination of the control room atmosphere Room as effectively as via the outside air intake by the intake or infiltration of airborne (additional information is provided in Section radioactive material from areas and structures 3.14.1.2 of the calculation in Enclosure 4). In the adjacent to the control room envelope, event that the Control Room intake isolates and the

Radiation shine from the external radioactive activity is trapped in the Control Room, assuming plume released from the facility, lesser quantities of infiltration is conservative, since

Radiation shine from radioactive material in the subsequent inleakage would dilute/purge the reactor containment, trapped activity.

Radiation shine from radioactive material in systems and components inside or external to Due to shielding of the Control Room, radiation the control room envelope, e.g., radioactive shine from a Fuel Handling Accident is considered material buildup in recirculation filters. to be a negligible dose contributor. More details on the various assumptions for radiation sources and the shielding available to the Control Room is provided in the calculation, Section 3.14.

4.2.2 The radioactive material releases and The radioactive material releases and radiation radiation levels used in the control room dose levels used in the Control Room dose analysis were analysis should be determined using the same determined using the same source term, transport, source term, transport, and release assumptions and release assumptions used for determining the used for determining the EAB and the LPZ TEDE EAB and the LPZ TEDE values. Control Room values, unless these assumptions would result in Chi/Q values were utilized.

non-conservative results for the control room.

4.2.3 The models used to transport radioactive The RADTRAD computer code was used to model material into and through the control room, and the transport of radioactive material into and through shielding models used to determine radiation dose the Control Room. This modeling provides suitably rates from external sources, should be structured to conservative estimates of the exposure to Control provide suitably conservative estimates of the Room personnel.

exposure to control room personnel.

4.2.4 Credit for engineered safety features that The base calculation takes no credit for Control mitigate airborne radioactive material within the Room engineered safety features or isolations, i.e.,

control room may be assumed. Such features may no credit for the Control Room radiation monitor include control room isolation or pressurization, or that can isolate the intake, or for any filtration on the

Enclosure 2 PY-CEI/NRR-2674L Page 14 of 24 Reaulatorv Guide 1.183 Guidance nonroc nf rnnfnrm~nro Reciulatorv Guide 1 183 Guidance intake or recirculation filtration.... In most designs, intake flows. Since no credit was taken for isolation control room isolation is actuated by engineered of the intake by the radiation monitor, the issue of safeguards feature (ESF) signals or radiation whether this monitor might be delayed in monitors (RMs). In some cases, the ESF signal is responding to the radiation is not a concern.

effective only for selected accidents, placing Sensitivity studies were performed to examine the reliance on the RMs for the remaining accidents. flexibility the Control Room operators have in using Several aspects of RMs can delay the control room ventilation, to ensure there were no dose outliers.

isolation, including the delay for activity to build up The studies evaluated what steps could be taken to concentrations equivalent to the alarm setpoint even if the radiation monitor was to isolate the and the effects of different radionuclide accident intake at the worst possible time (after all of the isotopic mixes on monitor response. available activity from the plume had been introduced into the Control Room). The sensitivity studies showed that even if the operators take two hours to take action, they can then either purge or use ventilation filters to remove the activity, and neither method resulted in excessive doses.

Procedural guidance for response to a Fuel Handling Accident will be updated to recommend that the operators evaluate what dose minimization method for the Control Room is best suited for the case at hand (filtration or re-initiation of normal intake), then take the appropriate ventilation measures to minimize dose. [See Commitment 3].

4.2.5 Credit should generally not be taken for the No credit was taken for the use of personal use of personal protective equipment or protective equipment or prophylactic drugs.

prophylactic drugs. Deviations may be considered on a case-by-case basis.

4.2.6 The dose receptor for these analyses is the The dose receptor for these analyses was the hypothetical maximum exposed individual who is hypothetical maximum exposed individual, who is present in the control room for 100% of the time present in the Control Room for 100% of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of the during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of the time between 1 and 4 days, and 40% of the time time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the from 4 days to 30 days. For the duration of the event, the breathing rate of this individual should be event, the breathing rate of this individual was assumed to be 3.5 x 10. cubic meters per second. assumed to be 3.5 x 10.4 cubic meters per second.

4.2.7 Control room doses should be calculated The Control Room doses were calculated using the using dose conversion factors identified in same dose conversion factors as identified in Regulatory Position 4.1 above for use in offsite Regulatory Position 4.1 for use in offsite dose dose analyses. The DDE from photons may be analyses. Also, the RADTRAD computer code corrected for the difference between finite cloud uses the equation provided in Section 4.2.7 for geometry in the control room and the semi-infinite correcting the finite versus semi-infinite cloud cloud assumption used in calculating the dose assumptions.

conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DDEý , to a finite cloud dose ... "

4.3 OtherDose Consequences "The guidance provided in Regulatory Positions 4.1 See Section 1.3.1 above for the responses to each and 4.2 should be used, as applicable, in re of those items. "Design envelope source terms" are assessing the radiological analyses identified in not being changed by the Fuel Handling Accident Regulatory Position 1.3.1, such as those in dose re-calculation. Radiation exposure estimates NUREG-0737 (Ref. 2). Desian enveloDe source to plant personnel for many of the NUREG-0737

Enclosure 2 PY-CEIINRR-2674L Page 15 of 24 Reclulatorv Guide 1.183 Guidance Dearee of Conformance I -

I terms provided in NUREG-0737 should be updated considerations are also not affected by a Fuel for consistency with the AST. In general, radiation Handling Accident. The Technical Support Center exposures to plant personnel identified in doses were addressed through a scoping study Regulatory Position 1.3.1 should be expressed in comparison to the Control Room. Equipment terms of TEDE. Integrated radiation exposure of qualification requirements for plant equipment in the plant equipment should be determined using the Fuel Handling Building are not being revised as a guidance of Appendix I of this guide." result of the new Fuel Handling Accident calculation, consistent with guidance in Regulatory Guide 1.183, Section 1.3.5. In the Containment, the Fuel Handling Accident doses are not bounding for EQ purposes, so the design basis integrated exposure values are unaffected.

4.4 Acceptance Criteria i "The radiological criteria for the EAB, the outer The 5 rem TEDE Control Room dose criterion from boundary of the LPZ, and for the control room are in 10 CFR 50.67 is used. For EAB and LPZ, the 10 CFR 50.67. These criteria are stated for 6.3 rem TEDE dose criterion from Table 6 of evaluating reactor accidents of exceedingly low Regulatory Guide 1.183 is used (-25% of the probability of occurrence and low risk of public 10 CFR 50.67 criterion). The NUREG-0737 item exposure to radiation, e.g., a large-break LOCA. potentially affected by a Fuel Handling Accident is The control room criterion applies to all accidents. TSC dose (if the TSC is activated for such an For events with a higher probability of occurrence, event), which is estimated by a scoping study to be postulated EAB and LPZ doses should not exceed well within the 5 rem TEDE dose. The USAR the criteria tabulated in Table 6. The acceptance markup provided in Enclosure 5 shows how the criteria for the various NUREG-0737 (Ref. 2) items new dose criteria are being updated.

generally reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or RG 1.183 Table 6 also shows an "analysis release specify criteria derived from GDC-1 9. These duration" of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Instead of the 2-hour release criteria are generally specified in terms of whole duration, the calculation conservatively used an body dose, or its equivalent to any body organ. For instantaneous (puff) release assumption.

facilities applying for, or having received, approval for the use of an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10 CFR 50.67(b)(2)(iii)."

Section 5. Analysis Assumptions and Methodology 5.1 General Considerations 5.1.1 Analysis Quality "The evaluations required by 10 CFR 50.67 ... The revised Fuel Handling Accident calculations should be prepared, reviewed, and maintained in were prepared under a 10 CFR 50 Appendix B accordance with quality assurance programs that quality assurance program.

comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

These design basis analyses were structured to The conservative, bounding characteristics of the provide a conservative set of assumptions to test AST that the NRC staff chose to present in the performance of one or more aspects of the Regulatory Guide 1.183 are used in the facility design. Many physical processes and calculations.

phenomena are represented by conservative, bounding assumptions rather than being modeled directly. The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for larme

Enclosure 2 PY-CEI/NRR-2674L Page 16 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

Licensees should exercise caution in proposing Therefore there are no proposed deviations from deviations based upon data from a specific accident the AST characteristics that are based on specific sequence since the DBAs were never intended to accident sequences that would require additional represent any specific accident sequence - the justification to prove they are conservative for other proposed deviation may not be conservative for accident sequences.

other accident sequences."

5.1.2 Credit for EngineeredSafeguard Features "Credit may be taken for accident mitigation No credit for ESF systems or components is taken features that are classified as safety-related, are in the base calculation, which produced acceptable required to be operable by technical specifications, results after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of radiological decay. In the are powered by emergency power sources, and are Control Room base calculation, the normal Control either automatically actuated or, in limited cases, Room ventilation system is considered to continue have actuation requirements explicitly addressed in to run throughout the event without filtration (see emergency operating procedures. The single active discussion in Section 5.1.3 below).

component failure that results in the most limiting radiological consequences should be assumed.

Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequences."

5.1.3 Assignment of Numeric Input Values "The numeric values that are chosen as inputs to the Conservative assumptions were utilized in the analyses required by 10 CFR 50.67 should be analyses.

selected with the objective of determining a conservative postulated dose. In some instances, a As described above, one area in which sensitivity particular parameter may be conservative in one studies were completed is with the Control Room portion of an analysis but be nonconservative in dose. The base case assumes the normal another portion of the same analysis. For example, ventilation system continues to run. This ensures assuming minimum containment system spray flow the intake of activity into the Control Room is is usually conservative for estimating iodine maximized, and ensures no credit is taken for active scrubbing, but in many cases may be functions such as isolations from the radiation nonconservative when determining sump pH. monitor or activation of the Emergency Sensitivity analyses may be needed to determine Recirculation system. This base case the appropriate value to use. As a conservative conservatively assumed intake flow 10% above alternative, the limiting value applicable to each nominal in order to maximize the amount of activity portion of the analysis may be used in the evaluation that enters the Control Room, then conservatively of that portion. A single value may not be applicable assumed exhaust flow 10% below nominal after the for a parameter for the duration of the event, activity has been introduced into the Control Room.

particularly for parameters affected by changes in The sensitivity studies examined actions the density. For parameters addressed by technical operators could take after a period of time in an specifications, the value used in the analysis should isolated, non-filtered mode (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after all the be that specified in the technical specifications. " activity is introduced into the Control Room in the studies), to ensure that their choice of action would not result in a dose outlier. The sensitivities studied the effects of turning on the filtration system, with low carbon filter efficiency (50%), or re-establishing the normal system intake, effectively purging the Control Room. Both cases provided doses within the dose limits.

Enclosure 2 PY-CEI/NRR-2674L Page 17 of 24 Reaulatorv Guide 1.183 Guidance Dearee of Conformance J -

5.1.4 Applicability of PriorLicensing Basis "The NRC staff considers the implementation of an Two items may be considered to be "retained items" AST to be a significant change to the design basis of from the current licensing basis. The first is the the facility that is voluntarily initiated by the licensee. allowance that the water level above the reactor In order to issue a license amendment authorizing vessel flange may be 22 feet 9 inches, less than the the use of an AST and the TEDE dose criteria, the standard 23 foot value. This has been previously NRC staff must make a current finding of compliance reviewed and approved by the NRC based on the with regulations applicable to the amendment. The fact that there is actually no fuel stored at the level characteristics of the ASTs and the revised dose of the flange (there is more than 51 feet of coverage calculational methodology may be incompatible with over the top of the fuel that is down in the reactor many of the analysis assumptions and methods vessel itself). Technical Specification 3.9.6 requires currently reflected in the facility's design basis the 22 foot 9 inch height over the flange of the analyses. The NRC staff may find that new or reactor vessel. As explained in the Bases, a unreviewed issues are created by a particular site dropped bundle would not be striking another fuel specific implementation of the AST, warranting bundle at this level where less than 23 feet of review of staff positions approved subsequent to the coverage exists. This limits the potential damage initial issuance of the license. This is not considered from the strike at this elevation to the pins in just a backfit as defined by 10 CFR 50.109, "Backfitting." one bundle rather than the two or more bundles that However, prior design bases that are unrelated to are involved in the bounding calculation (where a the use of the AST, or are unaffected by the AST, strike occurs in the core with 51 feet of coverage).

may continue as the facility's design basis. By itself, this single versus multiple bundle damage Licensees should ensure that analysis assumptions limits the release at this height, and more than and methods are compatible with the ASTs and the compensates for the coverage being less than 23 TEDE criteria. feet. Also, a bundle dropped at this elevation is falling less than 2 feet, rather than the drop of 34 feet assumed in the evaluation that determines the number of fuel pins that might be damaged by a drop. As already explained in the Technical Specification 3.9.6 Bases, the reduction in this water level over the flange is acceptable. To validate this conclusion, a separate calculation was performed (and is included as Appendix A to Enclosure 4) for the drop of a bundle that strikes the refueling shield. The refueling shield, which sets on the reactor vessel flange during the refueling process, is the highest horizontal surface that a fuel bundle could strike if dropped in the reactor cavity area. As expected, the resultant doses were bounded by the analyses where a dropped bundle hit multiple other bundles (the doses from the drop onto the refueling shield would be less than 75% of the design basis cases). Therefore, despite the water level being less than 23 feet, this does not represent the limiting case. Further details on the drop onto the refueling shield are contained in Appendix A to Enclosure 4.

The second "retained item" from the current licensing basis is that the "Decay Time" specification will remain in the PNPP Operational Requirements Manual (ORM) The Decay Time specification was relocated out of the Technical Specifications as part of Amendment 69, the improved Technical Specifications. The Decay Time specification

Enclosure 2 PY-CEI/NRR-2674L Page 18 of 24 Pnm mIahr, (, 4 '1 %a2 fi urfini-n fl na ran af tanfarw, - n,.n II 4sAlitt* T ," A.itat *. I* D ' iJ* ll/**t ImJ5, Ca U ,UIIU *iIC I .

requires that the plant be subcritical for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before movement of irradiated fuel may begin.

It is proposed that this control remain in the ORM, since the PNPP Decay Time specification was relocated to the ORM as part of the improved Technical Specifications. The NRC Safety Evaluation for Amendment 69 still holds true, where it stated: "Although Criterion 2 of the Final Policy Statement would require [the Decay Time specification] to be retained in the improved TS, the requirement for a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months decay time following subcriticality before commencing movement of irradiated fuel in the reactor vessel will always be met for a refueling outage. ....Therefore, the requirement is unnecessary and has been relocated from the specifications to the ORM."

5.2 Accident-Specific Assumptions "The appendices to this regulatory guide provide Reg. Guide 1.183 Appendix B is the applicable accident-specific assumptions that are acceptable appendix for a Fuel Handling Accident. Each to the staff for performing analyses that are required assumption in that guidance is addressed below.

by 10 CFR 50.67. ... Licensees should analyze the Except for the 23 feet of water over the vessel DBAs that are affected by the specific proposed flange issue discussed above, and the applications of an AST. The NRC staff has instantaneous puff release, also discussed above, determined that the analysis assumptions in the alternatives to the assumptions in Appendix B are appendices to this guide provide an integrated not being proposed.

approach to performing the individual analyses and generally expects licensees to address each assumption or propose acceptable alternatives.

Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration. The assumptions in the appendices are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other. Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency ... "

5.3 Meteorology Assumptions "Atmospheric dispersion values (X/Q) for the EAB, No changes to ChiQ atmospheric dispersion values the LPZ, and the control room that were approved are being proposed. The current USAR ChiIQ by the staff during initial facility licensing or in values for the Control Room, EAB and LPZ were subsequent licensing proceedings may be used in approved in conjunction with License Amendments performing the radiological analyses identified by 102 and 103, in March 1999. The actual values this guide. .. All changes in XIQ analysis used are presented in the calculation attached as methodology should be reviewed by the NRC staff." Enclosure 4.

Section 6. Assumptions for Evaluating the RadiationDoses for Equipment Qualification "The assumptions in Appendix I to this guide are No changes are proposed to equipment

Enclosure 2 PY-CEI/NRR-2674L Page 19 of 24 Raulatorv GoldA 11R3 GuidancA flAnr nf Cnnfnrmanr Realaor 83 Gude1uiane Dnre f Cnfrmnc acceptable to the NRC staff for performing qualification requirements at PNPP-as a result of radiological assessments associated with the reanalysis of the Fuel Handling Accident. Since equipment qualification. The assumptions in most of the particulate radionuclides that escape Appendix I will supersede Regulatory Positions from the fuel rod gap are assumed to convert to an 2.c(1) and 2.c(2) and Appendix D of Revision 1 of elemental form prior to release from the water (see Regulatory Guide 1.89, "Environmental the Appendix B discussions below), the source term Qualification of Certain Electric Equipment composition is not significantly different than before, Important to Safety for Nuclear Power Plants" (Ref. except it has been scrubbed more efficiently by the 11), for operating reactors that have amended their water in the pool (DF of 200 versus 100). This licensing basis to use an alternative source term. more efficient scrubbing reduces the overall release Except as stated in Appendix I, all other above the pools, which is the dose that might be assumptions, methods, and provisions of Revision seen by plant equipment. In the Fuel Handling 1 of Regulatory Guide 1.89 remain effective. The Building, the dose received by equipment from a NRC staff is assessing the effect of increased Fuel Handling Accident would therefore be lower cesium releases on EQ doses to determine whether than the dose using the original assumptions. Also, licensee action is warranted. Until such time as this in both the Containment and the Fuel Handling generic issue is resolved, licensees may use either Building, the LOCA is the event that sets the EQ the AST or the TID 14844 assumptions for requirements for equipment, rather than the Fuel performing the required EQ analyses. However, no Handling Accident. Finally, after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of plant modifications are required to address the radiological decay, no credit is taken in the revised impact of the difference in source term calculations for the operation of any plant characteristics (i.e., AST vs TID14844) on EQ equipment to mitigate the release before it escapes.

doses pending the outcome of the evaluation of the Therefore it is conservative to retain existing EQ Qeneric issue." Drocram requirements.

Appendix B ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OFA FUEL HANDLING ACCIDENT App. B Section 1. Source Term "Acceptable assumptions regarding core inventory The number of fuel rods damaged in a Fuel and the release of radionuclides from the fuel are Handling Accident (151) was determined by the fuel provided in Regulatory Position 3 of this guide. The vendor, Global Nuclear Fuels (GNF). The vendor following assumptions also apply. used a methodology that has been generically 1.1 The number of fuel rods damaged during the reviewed and approved by the NRC as part of accident should be based on a cb'nservative NEDE-2401 1-P-A (GESTAR II). The analysis analysis that considers the most limiting case. This considers the weight of a dropped GE 12 or 14 fuel analysis should consider parameters such as the assembly, including the weight of the triangular fuel weight of the dropped heavy load or the weight of a handling mast. Italso considered the height of the dropped fuel assembly (plus any attached handling drop, and the compression, torsion and shear grapples), the height of the drop, and the stresses on the irradiated fuel rods. Damage to compression, torsion, and shear stresses on the adjacent fuel assemblies was also considered.

irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

1.2 The fission product release from the breached The total core inventory is multiplied by the gap fuel is based on Regulatory Position 3.2 of this fractions from Position 3.2, a radial peaking factor, guide and the estimate of the number of fuel rods and the fraction of fuel rods failed. There are breached. All the gap activity in the damaged rods approximately 64,208 fuel rods in the core. The is assumed to be instantaneously released. percentage of rods breached is 151/64,208 32, Radionuclides that should be considered include which is 0.235% of the core. This activity is xenons, kryptons, halogens, cesiums, and instantaneously released. The radionuclide groups rubidiums. listed in Rea. Guide 1.183 Table 3 were considered.

Enclosure 2 PY-CEIINRR-2674L Page 20 of 24 Reaulatorv Guide 1.183 Guidance Dearee of Conformance 1.3 The chemical form of radioiodine released from No attempt is made to justify a mechanistic the fuel to the spent fuel pool should be assumed to treatment of the halogen release from the pool.

be 95% cesium iodide (Csl), 4.85 percent elemental The non-organic halogens are assumed to re iodine, and 0.15 percent organic iodide. The Csl evolve in an elemental form.

released from the fuel is 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."

App. B Section 2. Water Depth "If the depth of water above the damaged fuel is The only case where the water is not 23 feet or 23 feet or greater, the decontamination factors for greater is over the reactor vessel flange/refueling the elemental and organic species are 500 and 1, shield, as discussed above. The tops of all the respectively, giving an overall effective irradiated fuel assemblies in storage have greater decontamination factor of 200 (i.e., 99.5% of the than 23 feet of coverage. Specifically, the top of the total iodine released from the damaged rods is fuel in the reactor vessel is - 51 feet below the retained by the water). This difference in surface, and the fuel in the Fuel Handling Building decontamination factors for elemental (99.85%) and is - 28 feet below the surface. The fuel in the upper organic iodine (0.15%) species results in the iodine Containment pools is - 27 feet below the surface.

above the water being composed of 57% elemental and 43% organic species. If the depth of water is Therefore, in all the calculations except for the "less not 23 feet, the decontamination factor will have to than 23-foot" refueling shield calculation discussed be determined on a case-by-case method above, an overall effective decontamination (Ref. B-I)." factor (DF) of 200 is used for the halogens (iodines and bromines). It should be noted that this is a conservatism, since a DF of 500 for the elemental species would actually result in a higher overall effective DF than 200, as discussed in several other plant's submittals on this subject.

For the refueling shield calculation, an overall DF

-. was calculated to be 152.4. Despite the reduced DF, this was shown not to be a limiting case, primarily due to fewer pins being damaged in this event as compared to the bounding event over the reactor vessel. Details are provided in Appendix A to Enclosure 4.

App. B Section 3. Noble Gases (and particulates)

"The retention of noble gases in the water in the 100% of the Noble Gases (xenons and kryptons) fuel pool or reactor cavity is negligible (i.e., are assumed to escape the water pool (DF of 1).

decontamination factor of 1).

Particulate radionuclides are assumed to be None of the particulate radionuclides (the alkali retained by the water in the fuel pool or reactor metals - cesiums and rubidiums) are assumed to cavity (i.e., infinite decontamination factor)." escape the water pool (DF of co).

App. B Section 4. Fuel HandlingAccidents Within The Fuel Building "For fuel handling accidents postulated to occur The following section (Section 5) addresses "Fuel within the fuel building, the following assumptions Handling Accidents Within Containment". At PNPP, are acceptable to the NRC staff. a drop within Containment bounds the drop within

Enclosure 2 PY-CEI/NRR-2674L Page 21 of 24 Rea]ulatorv Guide 1.183 Guidance. flrnrPA of cnnfnrminrr Rejltr Guide- 1.83Gudace-----f-onorma 4.1 The radioactive material that escapes from the the Fuel Handling Building. Therefore the next fuel pool to the fuel building is assumed to be section, which addresses the Containment, will released to the environment over a 2-hour time provide more details on the bounding calculation.

period. The Containment drop is bounding because the 4.2 A reduction in the amount of radioactive drop over the reactor vessel would have higher material released from the fuel pool by engineered kinetic energy and therefore a greater number of safety feature (ESF) filter systems may be taken individual fuel rods damaged. The drop distance into account provided these systems meet the over the vessel is - 30 feet (GE used 34 feet in their guidance of Regulatory Guide 1.52 and Generic calculation), whereas in the Fuel Handling Building, Letter 99-02 (Refs. B-2, B-3). Delays in radiation the drop is - 8 feet. Since both analyses then detection, actuation of the ESF filtration system, or assume that the activity which escapes from the diversion of ventilation flow to the ESF filtration pool is treated the same, i.e., it is released system should be determined and accounted for in immediately and directly to the environment, the the radioactivity release analyses. FHA inside Containment will be bounding. There 4.3 The radioactivity release from the fuel pool also is no practical difference between the two should be assumed to be drawn into the ESF buildings at PNPP during handling of fuel that is not filtration system without mixing or dilution in the fuel considered to be "recently irradiated". After License building. If mixing can be demonstrated, credit for Amendment 102, handling of fuel that has been mixing and dilution may be considered on a case subcritical for more than 7 days has been by-case basis. This evaluation should consider the performed using a two building "envelope" magnitude of the building volume and exhaust rate, consisting of the Containment and the FHB. The the potential for bypass to the environment, the equipment hatch between these two buildings is location of exhaust plenums relative to the surface opened, and the Fuel Handling Building Ventilation of the pool, recirculation ventilation systems, and System can draw down the two building envelope.

internal walls and floors that impede stream flow This exhaust is then routed through filters and out between the surface of the pool and the exhaust of the plant vent (note again that the calculations do plenums." not credit this filtration or delay time in the release).

App. B Section 5. Fuel HandlingAccidents Within Containment "For fuel handling accidents postulated to occur within the containment, the following assumptions are acceptable to the NRC staff.

5.1 If the containment is isolated during fuel No credit is taken for Primary or Secondary handling operations, no radiological consequences Containment isolation during fuel handling after 24 need to be analyzed. 1. hours of radiological decay.

5.2 If the containment is open during fuel handling No credit is taken for automatic isolations of the operations, but designed to automatically isolate in Primary or Secondary Containment after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the event of a fuel handling accident, the release of radiological decay.

duration should be based on ... "

5.3 If the containment is open during fuel handling Rather than releasing the activity over a 2-hour time operations (e.g., personnel air lock or equipment period, the release is conservatively considered to hatch is open), the radioactive material that be instantaneous.

escapes from 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 No credit is taken for filter systems after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of material released from the containment by ESF radiological decay, although such systems will filter systems may be taken into account provided continue to be available during fuel handling per that these systems meet the guidance of controls implemented for Amendment 102.

Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection,

Enclosure 2 PY-CEI/NRR-2674L Page 22 of 24 Regulatory Guide 1.183 Guidance Degree of Conformance 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 No credit is taken for dilution or mixing of the released from the reactor cavity by natural or forced activity inside the Containment.

convection inside the containment may be considered on a case-by-case basis. ... "

Requlatorv AnalvsislCommitments The NRC's traditional methods (prior to the AST) for calculating the radiological consequences of design basis accidents are described in a series of regulatory guides and Standard Review Plan (SRP) chapters. That guidance was developed to be consistent with the TID-14844 source term and the whole body and thyroid dose guidelines stated in 10 CFR 100.11. Many of those analysis assumptions and methods are inconsistent with the ASTs and with the Total Effective Dose Equivalent (TEDE) criteria provided in 10 CFR 50.67. Regulatory Guide 1.183 provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST. This guidance supersedes corresponding radiological analysis assumptions provided in the older regulatory guides and SRP chapters when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67. One of the Regulatory Guides that is superceded for PNPP for the Fuel Handling Accident is Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors".

Due to the comprehensive nature of Regulatory Guide 1.183, the matrix (Table) given above was incorporated into this submittal to show how each section of the new guidance is being addressed.

Also, the NRC published a new SRP section to address AST. It is Standard Review Plan Section 15.0.1, Rev. 0, entitled "Radiological Consequence Analyses Using Alternative Source Terms". It provides guidance on which NRC branches will review various aspects of an AST license amendment request, but otherwise is consistent with the guidance found in Regulatory Guide 1.183. The plant specific information provided above to support the license amendment request is believed to adequately address the guidance found in SRP 15.0.1.

Several Regulatory documents other than Regulatory Guide 1.183 are applicable to the proposed change. The following matrix addresses these.

SUBJECT:

Fuel Handling Accident Using Alternative Source Term CLASSIFICATION CATEGORY REFERENCED IN REFERENCED OPEN USARVALIDATION IN ATLAS? ASSUMPTIONS?

[ SAFETY-RELATED [ ACTIVE DATABASE?

I] AUGMENTED QUALITY I] HISTORICAL [ YES [ YES 5 YES 0] NON SAFETY RELATED 5 STUDY [ NO [1 NO [ NO COMPUTER PROGRAM(S)

RADTRAD Mod 3.02, Microsoft@ Word 2000, Microsoft& Excel 2000 REVISION RECORD Rev. Description of Change Affected Preparer/Date El Lead Pages Reviewer/Date Engineer/Date I OVerifierlDate 0 Initial Issue oI. All I I __ I ____ I ____ I ____

iI I v i I I I I-I I 1 4 4-1 4 t 4 .1-I I I I +

i i i I I-1 4 -t 4 *1-1 4 I 4 I I 1 1 4-

_________ L _________________ I ________________ _________________

Enclosure 4 PY-CEI/NRR-2674L Page 2 of 76 RtrstneC U T Page ii Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term OBJECTIVE OR PURPOSE:

The purpose of this calculation is to determine radiological consequences of a design basis fuel handling accident (FHA) at Perry Nuclear Power Plant (PNPP), which occurs at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following reactor shutdown. Total effective dose equivalent (TEDE) at the control room, exclusion area boundary (EAB), and outer boundary of the low population zone (LPZ) are to be calculated using a source term derived from NUREG-1465, Reg. Guide 1.183, NEI 99-03, and the following conservative assumptions: [DIN # 1, 2, 3]

"* No credit for containment/fuel handling building integrity.

"* No credit for Annulus Exhaust Gas Treatment System (AEGTS) or Fuel Handling Area Exhaust Ventilation System (FHAEVS).

"* No credit for filtration of Control Room Emergency Recirculation System.

"* No credit for the isolation of the control room intake.

This calculation will replace the PNPP FHA dose analyses for the EAB, LPZ, and control room, CEI calculations 3.2.8 and 3.2.8.1, which were performed using Reg. Guide 1.25 and TID-14844 methodologies. [DIN # 4, 5, 6, 7]

SCOPE OF CALCULATION/REVISION This calculation performs radiological dose analysis at the control room, EAB, and outer boundary of LPZ for a design basis fuel handling accident using an alternative source term. The scope is limited to calculating TEDE for a given number of fuel rods failed.

SUMMARY

OF RESULTS/CONCLUSIONS:

Table 7 lists TEDE values calculated for the control room, EAB, and LPZ and compares these with regulatory limits. As shown in Table 7, the RADTRAD calculated TEDE values are well below the regulatory limits. Table 8 shows the TEDE values calculated for control room for the two sensitivity cases assuming the initiation of

1) control room fresh air intake and 2) control room recirculation filtering at two hours after a control room isolation assumed to occur after intake of all the activity. Bo3th cases show TEDE values below the regulatory limit for the control room. Appendix A concluded that the dose consequences for an FHA occurring while transiting fuel over the Refueling Chute is bounded by the dose consequences for the design basis FHA.

IMPACT ON OUTPUT DOCUMENTS:

This calculation will be the basis to revise the USAR and Technical Specifications.

Enclosure 4 PY-CEI/NRR-2674L Page 3 of 76 FirstEnety Page iii tegy CALCULATION Perry Nuclear Power Plant PNPP No. 6077 Rev. 217/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 "TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term DOCUMENT INDEX Z I.*-

0 0 Document Number/Title Revision, Edition, Date 0 I NUREG-1465, 'Accident Source Terms for February 1995 0El

[]

Light-Water Nuclear Power Plants" 2 Reg. Guide 1.183, "Alternative Radiological July 2000 N El Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" 3 NEI 99-03, "Control Room Habitability June 2001 ElO :l Assessment Guidance," Nuclear Energy Institute 4 CEI CALC. No. Calculation 3.2.8, FHA Rev. 1 ElO 3l Inside Containment" 5 CEI CALC. No. 3.2.8.1, "Control Room Rev. 0. 12/30/1998 E] 0 1:[

Habitability Following a Fuel Handling Accident" 6 Reg. Guide 1.25 (Safety Guide 25), March 23,1972 0] El E "Assumptions Used for Evaluating Radiological Consequences of a Fuel Handling Accident in a Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors" 7 TID-14844, "Calculation of Distance 1962 OL] l Factors for Power and Test Reactor Sites" 8 NUREG/CR-6604, "RADTRAD: A June 1997 On] l Simplified Model for RADionuclide Transport and Removal And Dose Estimation" 9 NUREG/CR-6604, Supplement 1, June 8, 1999 ID lEl 3

"RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation" 10 Letter from D. R. Rogers to J. B. Balcken, January29, 1996; Revised, March 14,1996 El 0 El "Fission Product Inventories for Perry High Energy Cycles" (Attachment 2 to DIN # 12)

Enclosure 4 PY-CEI/NRR-2674L Page 4 of 76 FirtEne .rgy Page iv "Perry Nuclear CALCULATION PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term z

z s 0 Document Number/Title Revision, Edition, Date 11 Code of Federal Regulations:

10 CFR Part 50.67, "ACCIDENT 01/24/2000 5 [

SOURCE TERM" 10 CFR Part 50, Appendix A, 01/24/2000 0 0 E1

'GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS" 10 CFR Part 50, Appendix K 01/24/2000 0 0 El 12 DI-240, 'Fuel Handling Accident Input Rev. 0, 11-8-01 11 E El Assumptions: Fuels Input" 13 ORNLITM-7175, "A User's Manual for the July 1980 ] E -E ORIGEN 2 Computer Code" 14 Federal Guidance Report 11, "Limiting 2rd Printing, 1989 El 0 0l Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" 15 CCC-652 Oak Ridge National Laboratory V.1.12 Code Package, 1997 0 ED Ell RSICC Computer Code Collection MACCS2 16 Federal Guidance Report 12, "External 1993 El 0 [

Exposure to Radionuclides in Air, Water, and Soil" 17 PNPP Drawings: El 0 [

015-026 411-0103 413-0101 413-0102 414-0102 414-0523 4549-18-1 4549-0194-1 18 Calculation PSAT.08401T.03. "Perry Plant Rev. 5 El N 1:1 TEDE Calculation"

Enclosure 4 PY-CEI/NRR-2674L Page 5 of 76 FirstEnergy Page v Pey Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14. Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 6 o z CL z O_

E Document NumberfTitle Revision, Edition, Date 00 19 License Amendment 102 March 1999 El] [J E 20 License Amendment 103 March 1999 [] I] 0 21 License Amendment 112 June 2000 11 S 0 22 NEDC-32868P, -GE-14 Compliance with Rev. 1, December 11,2000 El 19 0 Amendment 22 of NEDC-2401 1-P-A "GESTAR-II" 23 Perry Technical Specifications 3.9.6 [] [J []

24 Letter from L. R. Conner of Global Nuclear November 5,2001 El 0 El Fuel to P. J. Curran of PNPP, "Fuel Handling Accident - Bounding Fuel Rod Pressure for GE12 and GEI4" (Attachment I to DIN # 12) 25 Calculation CL-M26-01 Rev. 1 E] 9 [E3 26 P&ID 912-610 Rev. CC E] 0 [

27 Periodic Test Instruction PTI-GEN-P001 1 Rev.1 El[ 9 El 28 SCIENTECH Interoffice Memo from 3/14/01 ED [ [El H. A. Wagage to T. Bladen, "RADTRAD Code Verification and Validation" 29 Perry letter PY-CEI/NRR-1510L ED El []

30 G. Burley, "Evaluation of Fission Product October 5, 1971 ED El El Release and Transport for Fuel Handling Accident," Radiological Safety Branch, Division of Reactor licensing

4-Enclosure 4 PY-CEI/NRR-2674L Page 6 of 76 Fir*ElergyPage vi FirstEnerg CALCULATIONPaev "PerryNuclear Power Plant PNPP No. 6077 Rev. 2f7101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 "TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term TABLE OF CONTENTS SUBJECT PAGE COVERSHEET:

OBJECTIVE OR PURPOSE ii SCOPE OF CALCULATION

SUMMARY

OF RESULTS/CONCLUSIONS ii IMPACT ON OUTPUT DOCUMENTS ii DOCUMENT INDEX iii.

CALCULATION COMPUTATION 1 1 M ETHOD OF ANALYSIS ................................................................................................................................. 1 2 ACCEPTANCE CRITERIA .............................................................................................................................. 4 3 ASSUM PTIONS .............................................................................................................................................. 4 4' DETAILED CALCULATIONS ........................................................................................................................ 10 4.1 Developm ent of Input Files ................................................................................................................... 11 4.1.1 Main Input File, pnpp_fha.psf .................................................................................................. 11 4.1.2 Auxiliary Input Files ...................................................................................................................... 12 4.1.2.1 Input File on Release Fraction and Timing, pnppfha.rft ..................................................... 12 4.1.2.2 Input File on Radionuclides Inventory and Decay Data, pnppfha.nif ................................. 13 4.1.2.3 Input File on Dose Conversion Factors, pnppfha.dcf ....................................................... 13 4.2 Running RADTRAD Code .................................................................................................................... 13 4.3 Results of the RADTRAD Run for the Base Case ............................................................................ 13 4.4 Sensitivity Analysis ................................................................................................................................ 16 4.4.1 Sensitivity Case 1: Effect of CR Isolation and Fresh Air Intake .............................................. 16 4.4.2 Sensitivity Case 2: Effect of CR Isolation and Recirculation Filtering ..................................... 16 5 COMPUTER INPUTAND OUTPUT ............................................................................................................. 16 Appendix A. Fuel Handling Accident while Transiting over the Refueling Shield . Main RADTRAD Input File for PNPP FHA, pnpp_fha.psf: Plant Model, Release Scenario, and Output Flags . Auxiliary RADTRAD Input File for PNPP FHA, pnpp_fha.rft: Release Fraction and Timing . Auxiliary RADTRAD Input File for PNPP FHA, pnpp_fha.nif: Radionuclides Inventory and Decay Data .Auxiliary RADTRAD Input File for PNPP FHA, pnppfha.dcf: Dose Conversion Factors . RADTRAD Output File for PNPP FHA, pnpp_fha.out . RADTRAD Output File for Sensitivity Case 1: Effect of CR Isolation and Fresh Air Intake . RADTRAD Output File for Sensitivity Case 2: Effect of CR Isolation and Recirculation Filtering

Enclosure 4 PY-CEI/NRR-2674L Page 7 of 76 FirstEnergy Page vii Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 "TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term LIST OF TABLES Table 1. Dose Conversion Factors [DIN # 14, 15, 16] ......................................................................................... 6 Table 2. Atmospheric dispersion values (X/Q) used (s/M3) ................................................................................ 8 Table 3. Gap Fractions for Fuel Handling Accident [Table 3. DIN # 2] ................................................................ 9 Table 4. Selection of RADTRAD Input Parameters for the Three Compartments ............................................ 11 Table 5. Calculation of FHA Source Term ............................................................................................................. 14 Table 6. Radionuclides Decay Data Used for the Analysis [DIN # 15] ................................................................... 15 Table 7. Comparison of TEDE Calculated for Control Room, EAB, and LPZ with Regulatory Limits ................ 15 Table 8. Comparison of Sensitivity Analysis Results with the Base Case for TEDE Calculated for Control Room 16 LIST OF FIGURES Figure 1. PNPP FHA Release Model ....................................................................................................................... 2

Enclosure 4 PY-CEI/NRR-2674L Page 8 of 76

-irstner"y Page 1 ucCALCULATION Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 32.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term CALCULATION COMPUTATION I METHOD OF ANALYSIS The RADTRAD computer code was used to determine the total effective dose equivalent (TEDE) at the control room, exclusion area boundary (EAB), and outer boundary of the low population zone (LPZ), for the design-basis fuel-handling accident (FHA) inside containment at the Perry Nuclear Power Plant (PNPP) using NUREG-1465 and Reg. Guide 1.183 alternate source terms. [DIN # 8,9, 1,2] When comparing an FHA inside containment with an FHA in the Fuel Handling Building, the inside containment event would have higher kinetic energy and greater number of fuel pins damaged. Both analyses make the equivalent assumption that the activity, which escapes from the pool, is released immediately and directly to the environment. Therefore, the present analysis was performed for an FHA inside containment, which will be bounding. Appendix A analyzed and concluded that the dose consequences for an FHA occurring while transiting fuel over the Refueling Shield is bounded by the dose consequences for the design basis FHA.

Although the RADTRAD computer code consists of standardized source term data, the PNPP-specific, GE-calculated core isotope inventory at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown was used instead, along with guidance provided in Reg. Guide 1.183. [DIN # 10, 2] All the isotopes of bromine, iodine, krypton, and xenon with core activity greater than 1 E-9 CiIMWt (a total of 20) and their daughters, i.e., an additional three isotopes of cesium and rubidium, were used for the analysis. Thus a total of 23 isotopes were used for TEDE analysis. The source terms of isotopes of cesium and rubidium were ignored as they were assumed to be retained completely by the pool. Sprays and natural deposition that may reduce the quantity of radioactive material were not credited. No filters or deposition of radioactive material in any pathway was modeled. The analysis does not credit the isolation of control room intake following a signal from the Rad. Monitor. The analysis also considers the effects of isolating the CR intake after activity is introduced into the control room (i.e., trapping the activity in the control room). The analysis considers the effects of trapping the activity for two hours followed by the cleanup by either the Control Room Emergency Recirculation System or by reopening the control room intake. Radioactive decay and in-growth of radionuclide daughters were also modeled in this analysis.

The RADTRAD model estimates doses in the control room and at EAB and LPZ. (Figure 1 schematically shows the PNPP FHA Release Model.) The model calculates the changes in radioactivity in the containment as a result of releasing radioactivity from the containment to the environment. Radioactive material is assumed to transport from the release point to the control room air intake, EAB, and LPZ without delay or deposition to the ground.

Atmospheric dispersion of radioactivity during transport was accounted by using dispersion factors (XIQ values).

The change in radioactivity in the control room results from radioactivity entering the room with air intake, release of radioactivity with air exhaust, radioactive decay of nuclides in the control room. An additional mechanism of changing activity in the control room is filtering, which was not modeled for the base case calculation but was performed for sensitivity analysis (case 2) (§4A.2).

Equation I shows the modeling of the change in radioactivity in the containment or control room, referred to as a "compartment"in the RADTRAD model. [§2.1.1, DIN # 8]

dN K K T. n,, =Sn, + y-Fi,jNn,j + T2fn,vvNv,i NJ Z VFi, j ,1nNni - 1n,iFn,recNn,i Equation 1 j=1 j=M j=1 j~i 1i j14 j#i

Enclosure 4 PY-CEI/NRR-2674L Page 9 of 76 FistEnerg Page 2 PerryNucear' CALCULATION PNPP No. 6077 Rev. 217/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO. 4 Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term FlHiA Release (Instantaneous) Recirc. Filter CotinetControl Room Control Room Control Room ConainentAir Intake Exhaust EAB LPZ Environment Figure 1. PNPP FHA Release Model' where:

N,,' = number of atoms of nuclide n in compartment i

= source injection rate of nuclide n into compartment i (atoms/s)

K = Number of compartments defined in the RADTRAD model Fj volume-normalized air flow rate from compartmentjto i (s-') (F,1 ;0)

= fraction of nuclide v that decays to nuclide n (dimensionless) jQ,,. = radioactive decay constant of nuclide n (s- I), which is calculated from radioactive half life, (t,1 2),, (s) as shown in Equation 2 17nj = filter efficiency for nuclide n in compartment i (dimensionless)

Fn.rec = volume-normalized recirculation air flow rate in compartment i (s-').

Anf = IZ Equation 2 The terms on the right hand side of Equation 1 models the following:

Injection rate of radioactive source of radionuclide n into compartment i

Intake rate of radionuclide n from all the other compartments into compartment i

Generation rate of radionuclide n by decay of radionuclide v in compartment i

Release rate of radionuclide n as a result of air exhaust from compartment i to all the other compartments.

Note that the net airflow rate into a compartment is equal to the net air exhaust rate.

Note that control room recirculation filtering was not modeled for the base case calculation but used only for sensitivity analysis (case 2) (§4.4.2)

En,'josure 4 PY-CEI/NRR-2674L Page 10 of 76 FirstEner Peny Pgey Perr Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 0 Decay rate of radionuclide n in compartment i 0 Removal rate of radionuclide n from compartment i by recirculation filtering.

The TEDE was calculated as the sum of committed effective dose equivalent (CEDE) from inhalation and the effective dose equivalent (EDE), which is assumed to be equivalent to deep dose equivalent (DDE) from external exposure from each nuclide, as shown in Equation 3 (§3.13.4).

TEDEL n 1=(EDEI

-= +CEDEn) Equation 3 where L represents the location (control room, EAB, or LPZ) and M is the total number of radionuclides used in the analysis.

The EDE from each nuclide at environment (env) (EAB or LPZ) is calculated as given in Equation 4. [§2.3.1, DIN #8]

EDEnenv E~en=DCFEDEfvAn

=TF.E~ And:t dn Equation 4 where:

EDEnenv = EDE (cloudshine dose) due to nuclide n in the environment at given location (rem)

DCF~,,oF = user-provided EDE (cloudshine) dose conversion factor for nuclide n frem.m 3 ci.s )

T = duration of analysis (s)

An = activity release rate of nuclide n (Ci)

(xQ) = user-provided atmospheric relative concentration at EAB or LPZ (sfm 3 )

IQenvy t = time (s)

The activity is related to the number of atoms of nuclide n as given in Equation 5.

An = Nn Zn Equation 5 The CEDE from each nuclide at environment (env) (EAB or LPZ) is calculated as given in Equation 6. [§2.3.1, DIN # 8]

CE=DEnv DCFcEDEn fTAnZ/Q BRenv dt Equation 6 0 "'-'env where:

CEDEenv = CEDE due to nuclide n in the environment at given location (rem)

DCFcEDE., = user-provided CEDE conversion factor for nuclide n (rem/Ci).

BR,,,, = user-provided breathing rate for the hypothetical individual at EAB or LPZ (m 3Is).

Enclosure 4 PY-CEIINRR-2674L Page 11 of 76 F erstffnery Page 4 Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 217101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term The EDE from each nuclide in the control room (CR) is calculated as given in Equation 7. [§2.3.2. DIN #8]

EDECR - DCF,.EnT nGV ACR, n OFdt Equation 7 where:

EDECR = EDE (cloudshine dose) due to nuclide n in the control room (rem)

GF = geometry factor as calculated using Equation 8 (dimensionless)

Ac.z*,, = activity of nuclide n in the control room at time t (Ci)

VCR = Volume of the control room (m 3)

OF = user-provided control room occupancy factor (dimensionless)

GFV= 1173 Equation 8 VCR 0 3 where:

VcR - Volume of the control room (ft 3 ). (Note the difference of units of this variable in Equation 7 and Equation8.)

The CEDE from each nuclide in the control room is calculated as given in Equation 9. [§2.3.2, DIN # 8]

CEDECR = BRCR DCFCEDEn T n OF dt

,ACR~ Equation 9 VCR where:

CEDECR = CEDE due to nuclide n in the control room (rem)

BRcR = user-provided breathing rate for the control room operator (m3/s).

2 ACCEPTANCE CRITERIA Both EAB and LPZ dose limits for FHA are TEDE of 6.3 rem during the analysis release duration of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

[Table 6, DIN # 2] The control room dose limit for FHA is TEDE of 5 rem for the duration of the accident (10CFR50.67(b)(2)(iii)). [DIN # 11]

3 ASSUMPTIONS 3.1 The FHA was assumed to occur at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following reactor shutdown.

3.2 When comparing an FHA inside containment with an FHA in the Fuel Handling Building, the inside containment event would have higher kinetic energy and greater number 6f fuel pins damaged due to the comparative height of the drop. Both analyses make the equivalent assumption that the activity, which escapes from the pool, is released immediately and directly to the environment. Therefore, the present analysis was performed for an FHA inside containment, which will be bounding.

3.3 No integrity of containment/fuel handling building was assumed.

Enclosure 4 PY-CEI/NRR-2674L Page 12 of 76 Fisr-fnery Page 5 Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7101 NEI-O341 CALCULATION TYPE CALCULATION NO.

INITIATING DOCUMENT Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 3.4 No credit was taken for Annulus Exhaust Gas Treatment System (AEGTS) or Fuel Handling Area Exhaust Ventilation System (FHAEVS).

3.5 No credit was taken for filtration of Control Room Emergency Recirculation System during the base case calculation. The effect of control room recirculation filtering following isolation of the control room air intake was studied in sensitivity case 2 (§4.4.2).

3.6 No control room isolation was assumed for the base case calculation. Sensitivity cases were run with RADTRAD to assess the impact of isolating the control room after the activity has entered the control room

(§4.4).

3.7 All failed fuel is assumed to be operating at high peaking factors and maximum exposures although the core operating limits on power density would prohibit high-exposure bundles from being at high peaking factors.

3.8 This analysis assumed a radial peaking factor of 2. [DIN # 12]

3.9 Radionuclide release from FHA was assumed to occur instantaneously.

3.10 Radioactive decay and corresponding in-growth of radionuc!:de daughters were modeled during this analysis.

3.11 Radioactive material is assumed to transport from the release point to the control room air intake, EAB, and LPZ without delay or deposition to the ground.

3.12 Fission Product Inventory (§3.1, Reg. Guide 1.183): [DIN # 2]

Reg. Guide 1.183 states that the inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full power operation of the core with, as a minimum, current licensed rated thermal power times the ECCS evaluation uncertainty. 2 The period of irradiation should be sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN 2. [DIN # 13]

This analysis used the ECCS evaluation uncertainty of 1.02. Fission product inventories were calculated by GE using the ORIGEN 2 computer code assuming 1500 effective full power days (EFPD) of operation.

[DIN # 10]

3.13 Offsite Dose Consequences (§4.1, Reg. Guide 1.183): [DIN # 2]

3.13.1 This calculation determines TEDE, which is the sum of CEDE from inhalation and EDE, which is assumed to be equivalent to DDE from external exposure (Equation 3) (§3.13.4). Impact of daughter products was considered by decaying core radionuclides inventory for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in DIN # 10.

Radioactive decay and corresponding in-growth of radionuclide daughters were modeled during this analysis (§3.10).

3.13.2 This calculation applies the CEDE conversion factors from Federal Guidance Report 11, which are readily available in the dose conversion factors file for 825 radionuclides, Dosdai825.inp, which was provided with the MACCS2 computer code package. [DIN # 14, 15] The CEDE conversion factors for the 23 nuclides that were selected for TEDE analysis, as described in §4.1.2.2, are listed in Table 1.

Note that for BR 82 and BR 83, this analysis used more conservative CEDE conversion factors, which are for the lung clearance class of W (weeks) as given in DIN # 14 than those used in DIN # 15 for the class D (days).

3.13.3 This calculation applies the recommended breathing rates: 3.5E-4 m 3/s for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 1.8E-4 m Is from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2.3E-4 mi3/s thereafter. However, since the release was conservatively modeled as an instantaneous release, the analysis used an effective breathing rate of 3.5E-4 mi3 /s.

3.13.4 The DDE should be calculated assuming submergence in semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally equivalent to the effective 2 The uncertainty factor used in determining the core inventory should be that value provided in Appendix K to 10 CFR Part 50. typically 1.02. [DIN # 111

Enclosure 4 PY-CEI/NRR-2674L Page 13 of 76 Fu'stFletx Page 6 Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7101 NEI-0341 INITIATING DOCUMENT I CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14 Rev. 0 TITLE

SUBJECT:

Fuel Handling Accident Using Alternative Source Term dose equivalent (EDE) from external exposure if the whole body is irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations, EDE may be used in lieu of DDE in determining dose to the TEDE.

This calculation applies the EDE conversion factors from Federal Guidance Report 12, which are readily available in the dose conversion factors file for 825 radionuclides, Dosdat825.inp, which was provided with the MACCS2 computer code package. [DIN # 16, 15] The EDE conversion factors for the 23 nuclides that were selected for TEDE analysis, as described in §4.1.2.2, are listed in Table 1.

Table 1. Dose Conversion Factors [DIN # 14, 15. 16]

EDE CEDE No. Isotope (rem-mý/Ci-s) (Sv-m3 /Bq-s) (ren/Ci) (Sv/Bq) 1 BR82 4.810E-01 1.300E-13 1.528E+03 4.130E-10 2 BR83 1.413E-03 3.820E-16 8.917E+01 2.410E-11 3 KR 83M 5.550E-06 1.500E-18 0 0 4 KR 85 4.403E-04 1.190E-16 0 0 5 KR 85M 2.768E-02 7.480E-15 0 0 6 KR 87 1.524E-01 4.120E-14 0 0 7 KR 88 3.774E-01 1.020E-13 0 0 i 8 RB 87 6.734E-06 1.820E-18 3.234E+03 8.740E-10 9 RB 88 1.243E-01 3.360E-14 8.362E+01 2.260E-1 1 10 1129 1.406E-03 3.800E-16 1.735E+05 4.690E-08 11 1130 3.848E-01 1.040E-13 2.642E+03 7.140E-10 12 1131 6.734E-02 1.820E-14 3.289E+04 8.890E-09 13 1132 4.144E-01 1.120E-13 3.811E+02 1.030E-10 14 1133 1.088E-01 2.940E-14 5.846E+03 1.580E-09 15 1134 4.810E-01 1.300E-13 1.314E+02 3.550E-11 16 1135 3.069E-01 8.294E-14 1.228E+03 3.320E-10 17 XE129M 3.922E-03 1.060E-15 0 0 18 XE131M 1.439E-03 3.890E-16 0 0 19 XE133 5.772E-03 1.560E-15 0 0 20 XE133M 5.069E-03 1.370E-15 0 0 21 XE135 4.403E-02 1.190E-14 0 0 22 XE135M 7.548E-02 2.040E-14 0 0 23 CS135 2.091E-06 5.650E-19 4.551E+03 1.230E-09 3.13.5 For the EAB, the objective of the TEDE analysis is to consider the dose during the worst two hour period. The maximum allowed duration of release from an FHA is 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . [Table 6, DIN # 2] For this analysis, the release was assumed to transport instantaneously from the release location to the receptor. Because the complete release was assumed to be instantaneous, the initial 2-hour gives the maximum dose. Therefore, TEDE was determined for the first two hours and no sliding window calculations were performed.

Enclosure 4 PY-CEI/NRR-2674L Page 14 of 76 firs etpy Page 7 FiZt;n:; CALCULATION Perry Nuclear Power Plant PNPP No. 6077 Rev. 217/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 1ITLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 3.13.6 The CFR states that the TEDE should be determined for the most limiting receptor at the outer boundary of the LPZ and should be used in determining compliance with the dose criteria in 10 CFR 50.67. [DIN #11]

This calculation determined the TEDE for the first two hours for the most limiting receptor at the outer boundary of LPZ. The radioactivity was available at LPZ during the release, which occurred during the initial 1E-4 hours (0.36 s) because no delay was assumed for the transport of activity. Therefore, no TEDE was received by the receptor at LPZ after 0.36 seconds, as no activity was available.

3.13.7 The dispersion factors used in this calculation do not take credit for ground or any other deposition.

3.14 Control Room Dose Consequences (§42, Reg. Guide 1.183): [DIN # 2]

3.14.1 This calculation considers potential radiation sources to the control room operator.

3.14.1.1 Unfiltered intake of the radiation plume into the control room was assumed to occur at 6600 cfm (normal flow rate + 10%) during the release, which occurred during the initial I E-4 hours (0.36 s). Exhaust flow rate was chosen to be equal to the intake flow rate. After radioactivity was taken into the control room, the exhaust flow rate was conservatively chosen to be at 5400 cfm (normal flow rate - 10%) in order to minimize the purging effect of the ventilation system. During this time, the intake flow rate was chosen to be equal to the exhaust flow rate.

3.14.1.2 Intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope:

Other than the 6600 cfm of unfiltered inleakage being assumed to be introduced directly into the control room from outside, infiltration of airborne radioactive material from adjacent areas and structures was considered to be a negligible dose contributor and was neglected.

Normal operation maintains a positive differential pressure between the inside and outside of the control room and thus between adjacent spaces. Control room doors lead to closed chase spaces, closed stairwells, or closed corridor spaces such that neither outside wind conditions nor other ventilation systems can cause infiltration/leakage into the control room.

3.14.1.3 Radiation shine from the external radioactive plume released from the facility:

Radiation shine from the external radioactive plume for the purpose of this calculation was considered to be a negligible dose contributor. The roof of the control complex building consists of a 2'-4.5"-thick concrete slab and the control room ceiling is an 18"-thick concrete slab (PNPP Drawings 015-026 and 414-0523). [DIN # 17] Considering this shielding, the contribution to the control room dose due to the cloud passing by was considered to be negligible and was neglected. This judgment is further supported by the LOCA control room dose calculation, which documents the cloud direct gamma dose for 30 days as being

<0.05% of the total control room LOCA dose. [DIN # 18] The percentage should be lower for a fuel handling accident and its resultant brief plume.

3.14.1.4 Radiation shine from radioactive material in the reactor containment:

The direct line from the containment with the least shielding is through the 3'-thick concrete containment shield building, the 2'-thick concrete control building wall and the 1.5'-thick concrete control room ceiling (PNPP Drawings 015-026, 414-0102, 411-0103). [DIN # 17]

Other direct lines from the containment to the control room would result in additional concrete shielding. Similarly, the direct line from the fuel handling is through a 3'-thick concrete fuel handling area wall, a 3'-thick concrete Intermediate Building wall and a 2'-thick Control Complex building wall (PNPP Drawings 413-0101, 413-0102, 414-0102). [DIN # 17]

Considering this shielding, the contribution to the control room dose due to shine resulting from radioactive material in the containment or fuel-handling building was considered to be negligible and was neglected. This judgment is further supported by the LOCA control room dose calculation which documents the containment direct gamma dose for 30 days as being

-3% of the total control room LOCA dose. [DIN # 18]

Enclosure 4 PY-CEI/NRR-2674L Page 15 of 76 FirstEnery Page 8 Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 "liTLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 3.14.1.5 There are no additional sources of radiation shine from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters. Radioactive material buildup in recirculation filters was not considered as the recirculation filters were not assumed to operate during FHA.

The effect of control room recirculation filtering following isolation of the control room air intake was studied in sensitivity case 2 (§4.4.2). The recirculation filters are located outside the control room envelope. The filter plenum equipment pad as well as the 18"-thick concrete control room ceiling shields the control room envelope. Therefore, radioactivity buildup in control room isolation filters would not affect the results of sensitivity case 2.

3.14.2 The radioactive material releases and radiation levels used in the control room dose analysis were determined using the same source term, transport, and release assumptions used for determining the EAB and LPZ TEDE values.

3.14.3 RADTRAD computer code was used to model transport of radioactive material into and through the control room. This modeling provides suitable conservative estimates of the exposure to control room personnel.

3.14.4 No credit was taken for engineered safety features that mitigate airborne radioactive material within the control room. The effect of trapping the activity in the control room for two hours, followed by cleanup by the emergency filters was studied in sensitivity case 2 (§4.4.2).

3.14.5 No credit was taken for using protective equipment or prophylactic drugs.

3.14.6 The dose receptor for these analyses was the hypothetical maximum exposed individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60%

time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of this of the event, the breathing rate of this individual was assumed to be 3.5E-4 m3/s.

3.14.7 Control room doses were calculated using the same dose conversion factors as the offsite dose calculation given in Table 1. RADTRAD computer code uses Equation 8 to correct the difference between finite cloud geometry in the control room and the semi-infinite cloud assumption used in calculating the dose conversion factors for the EDE from photons.

3.15 Acceptance Criteria (§4.4, Reg. GuideT'I.183): [DIN # 2]

As given in §2, this calculation applies acceptance criteria in Table 6 of Reg. Guide 1.183 and 10CFR50.67 for the offsite and control room doses. [DIN # 2, 11] Instead of the 2-hour release duration that is recommended for FHA in Table 6 of Reg. Guide 1.183, this calculation conservatively used instantaneous release assumption.

3.16 Meteorology Assumptions (§5.3, Reg. Guide 1.183): [DIN # 2]

The analysis uses atmospheric dispersion values (X/Q) used for the EAB, LPZ, and control room, listed in Table 2, that were previously approved by the NRC. [DIN # 19, 20]

Table 2. Atmospheric dispersion values (x/Q) used (slm3 )

Location X/Q (s/m3 ) Reference Control Room 3.5E-4a DIN#5 CEAB 4.3E-4 DIN #4 LPZ 4.8E-5 DIN # 4 DIN # 5 lists X/Q values for different time periods up to 30 days. The initial value, which was for 0 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , was used in this calculation because FHA release, and thus the activity intake to the control room was assumed to be instantaneous.

Enclosure 4 PY-CEI/NRR-2674L Page 16 of 76 FirstEnergy CALCU LATIO N Perry Nuclear Power Plant PNPP No. 6077 Rev. 217101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TIfLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 3.17 Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident (Appendix B, Reg. Guide 1.183): [DIN # 2]

3.17.1 Source Term:

3.17.1.1 The number of fuel rods damaged during an FHA, i.e., 151, was based on the fuel vendor's NRC approved methodology for GE 12 and 14 bundles and triangular fuel handling mast.

[DIN # 22]

3.17.1.2 This calculation used the gap fractions in §3.2 of Reg. Guide 1.183, as listed in Table 3, and assumed that the source terms were instantaneously released. Only bromine, iodine, krypton, and xenon radioisotopes were used in the calculation of FHA source term. Cesium and rubidium radioisotopes were assumed to retain completely by the fuel pool water.

Reg. Guide 1.183 noted that these gap fractions were applicable up to a peak rod average exposure of 62 GWD/MTU provided that the maximum linear heat generation rate did not exceed 6.3-kW/ft peak rod average power for bumup exceeding 54 GWD/MTU. As noted in DIN # 12, PNPP fuel designs satisfy this criterion.

Table 3. Gap Fractions for Fuel Handling Accident [Table 3, DIN # 2]

Isotope/Group Gap Fraction 1-131 8%

Kr-85 10%

Other Noble Gases (Xe, Kr) and Halogens (I, Br) 5%

Alkali Metals- (Cs, Rb) 12%

0 Alkali metals were assumed to be retained completely by the pool

(§3.17.1.2, 3.17.3).

3.17.1.3 The chemical form of-radioiodine released from the fuel to the spent fuel pool was assumed to be 95% cesium iodide (Csl), 4.85% elemental iodine, and 0.15% organic iodide. The Csi release from the fuel was assumed to completely dissociate in the pool water. Because of the low pH of pool water, the iodine is assumed to re-evolve as elemental iodine. This was assumed to occur instantaneously.

As a halogen, bromine isotopes were modeled identical to iodine in terms of chemical form.

3.17.2 Water Depth:

This calculation used the Reg. Guide 1.183 pool overall D F value of 200 for iodine isotopes. (As a halogen, bromine isotopes were modeled identical to iodine in terms of pool DF.) This assumption requires that PNPP pools maintain at least 23 feet of water coverage above damaged fuel.

The PNPP requires that it maintain at least 23' of water coverage above the fuel in the reactor pressure vessel (RPV) or the spent fuel storage pool. The PNPP has approximately 51.5' of water above the core, 27' above the fuel rack in the upper containment pool, and - 28' of coverage over spent fuel in the spent fuel pool (25' above IFTS gate sill). Therefore, if the dropped fuel bundle strikes another irradiated fuel bundle, 23' of water coverage above the damage bundle will be available.

Per Technical Specifications, PNPP requires that only 22'-9" of water coverage above the RPV flange during refueling. If the dropped bundle were to strike the RPV flange versus another bundle, there would be a possibility that only 22'-9" of water coverage is available. [DIN # 23] However, as addressed in the bases for Technical Specifications 3.9.6, such a drop will result in reduced release of fission gasses and it was judged that slight reduction in water level was acceptable. In addition, the drop onto the RPV flange will be on the order of 1.5' to 2', which is significantly less than the drop of 34' that was assumed in the GE analysis, which calculated the number of fuel rods failing during an

Enclosure 4 PY-CEI/NRR-2674L Page 17 of 76 RrstFlneAT Page 10 Perry Nuclear Power Plant PNPP No. 6077 Rev. 217101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term FHA. Therefore, the actual number of fuel rods failing in such an event will be significantly less than that assumed for this analysis (i.e., 151 fuel rods) and thus, the slightly reduced water level would be acceptable.

3.17.3 Noble Gases/Particulates:

The retention of noble gases in the water in the fuel pool or reactor cavity was assumed to be negligible (i.e., decontamination factor of 1). Particulate radionuclides (Cs and Rb) were assumed to be retained by the water in the fuel pool or reactor cavity (i.e., decontamination factor of o0).

3.17.4 Fuel handling Accidents within Containment:

3.17.4.1 It was conservatively assumed that the containment was not isolated during fuel handling operations.

3.17.4.2 It was conservatively assumed that the containment would not isolate in the event of an FHA.

3.17.4.3 For an open containment, Reg. Guide 1.183 recommends assuming that the radioactive material that escapes from the fuel building to be released to the environment over a 2-hour time period. This analysis conservatively assumed that the release took place instantaneously.

3.17.4.4 No credit was assumed for a reduction in the amount of radioactive material released from the containment by engineered safety features filter systems.

3.17.4.5 No credit was assumed for dilution or mixing of the radioactivity released from the reactor cavity by natural or forced convection inside the containment.

3.18 NEI 99-03 Insight on Release Pressure Limit of 1200 psig: (DIN # 3]

For pool overall DF value of 200 for iodine (and bromine) isotopes to be applicable, in addition to a 23' depth in the pool, the release pressure is to be limited to 1200 psig. (See §3.17.2.) In the event of an FHA, the PNPP fuel rod pressure will be below 1200 psig. [DIN # 24]

3.19 General Design Criteria for Nuclear Power Plants (10 CFR 50, Appendix A): [DIN # 11]

General Design Criteria (GDC) 61 and 63 of 10 CFR 50, Appendix A addresses FHA.

3.19.1 GDC 61-Fuel storage and handling and radioactivity control states that the fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed with appropriate containment, confinement, and filtering systems. At PNPP, these systems include the primary and secondary containment, Fuel Handling Building, and AEGTS/FHAEVS. However, this analysis assumes no credit for the presence of these buildings or AEGTS/FHAEVS.

3.19.2 GDC 63-Monitoring fuel and waste storage states that appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions. This analysis assumes no credit for the initiation of appropriate safety actions.

4 DETAILED CALCULATIONS Calculations were performed using the RADTRAD computer code, which requires a main input file and three auxiliary input files. The main input file, named as pnppfha.psf in this analysis, describes 1) the plant model with compartments and release pathways, 2) the release scenario with source term, release rates, mechanisms of reducing radioactive elements, including overlying pools, suppression pool, sprays, filters, and natural deposition, and 3) the code output. The three auxiliary input files, named as pnppfha.rft, pnppfha.nif, and pnpp fha.dcf in this analysis, describe release fraction and timing, radionuclides inventory, and radioactivity to dose conversion factors. Section 4.1 describes the development of input files. Sections 4.2 and 4.3 describe running and results of the RADTRAD code

Enclosure 4 PY-CEI/NRR-2674L Page 18 of 76 FirstEnergy Page 11 "PerryNuclear Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Anatysis 3.2.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 4.1 Development of Input Files 4.1.1 Main Input File, pnppfha.psf The input parameters are described in this section in the same order as they appear in the main input file, pnpp_fha.psf, which is given in Attachment 1. The radionuclides inventory file was defined in the main input file as pnpp_Iha.nif, which is described in §4.1.2.2. Power level was set to 3833.2 MWt (102% of 3758 MWt). [DIN # 21]

Figure 1 schematically shows the PNPP FHA Release Model, which consists of three compartments, the containment, control room, and environment, and three-release/flow paths, release from the containment to the environment, control room air intake, and control room exhaust.

Table 4 shows the selection of RADTRAD input parameters for the three compartments. Each value of the compartment type is that recommended in the RADTRAD input manual. [DIN # 9] Volume of the containment was arbitrarily selected as 1 ft3 because the actual value is unimportant for the assumed instantaneous, complete release of the FHA source term. Following example problems given in the RADTRAD input manual, the volume of the environment was chosen as zero. [DIN # 8]

Table 4. Selection of RADTRAD Input Parameters for the Three Compartments Number Name Type [DIN # 9] Volume (ft 3 )

1 Containment 3 1.0 (arbitrary) 2 Environment 2 0 3 Control Room 1 367,070 [DIN # 5, 251 The three radioactivity release/intake pathways are identified between the compartments as schematically shown in Figure 1. The type of release pathway from the containment to the environment was defined as "air leakage" by choosing number 4 for the release flag. [DIN # 9]

The type of release pathway for 1) air intake from the environment to the control room and 2) air exhaust from the control room to the environment was defined as "filtered pathway" by choosing number 2 for the release flag.

[DIN # 9] Note that the filter efficiencies were set to zero for both of these two paths, later in the input file.

The number I below the line with "Source Term" identifies that the whole source term was placed in only one compartment. Numbers I and I in the next line identify that the whole source term was placed in the containment (compartment # 1).

The auxiliary input files giving dose conversion factors, and release fraction and timing are identified as pnppfha.dcf and pnppjha.rft. These files are described in §4.1.2.3 and 4.1.2.1.

Delay time for the release was chosen as zero in order to model the release as instantaneous. A value of I was chosen for the flag to enable the calculation of radioactive daughter products. Next line shows the fractions of aerosol, elemental, and organic halogens and the fraction of halogens that are radioactive as 0, 0.9985, 0.0015, and 1 (§3.17.1.3). (Note that the fraction of halogens that are radioactive is a redundant input required by RADTRAD because specifying the activity of a nuclide implies that it is radioactive.)

No overlying pools were modeled with RADTRAD because decontamination factors were modeled separately in calculating the FHA source term as described in §4.1.2.2.

Enclosure 4 PY-CEI/NRR-2674L Page 19 of 76 FlirstEC L A Page 12 "PerryNuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2W7101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 32.15.14, Rev. 0 TITLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term The number of compartments is given as 3. The first and second numbers under each compartment give flags to indicate whether detail output was to be given and whether radioactive decay is to be calculated. Detailed output was requested only for the control room by choosing a flag value of 1. For all the three compartments, the containment, environment, and control room, radioactive decay was modeled by choosing a flag value of 1.

The dose calculation model was run for 30 days (720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months ). The release rate from the containment was chosen arbitrarily as 1El 0 %-volumelday. A high value was chosen to ensure complete, instantaneous release (§3.9).

Both control room intake and exhaust flow rates were set to 6,600 cfm (6000 cfm + 10% is based on the design value as shown on DIN # 26 with an allowable operating tolerance as specified in DIN # 27) during activity intake, which occurred during the initial I E-4 hours (0.36 s). After the activity was taken into the control room, both intake and exhaust flow rates were set to 5,400 cfm (6000 cfm - 10%). The filter efficiencies for both of intake and exhaust flow paths were set to zero.

TEDE was calculated at three locations, EAB, outer boundary of LPZ, and control room. Atmospheric-dispersion values (XIQ) of 4.3E-4 and 4.8E-5 s/m3 were used for EAB and the outer boundary of LPZ (Table 2). For both offsite locations the Reg. Guide 1.183 recommended breathing rates were used: 3.5E-4 m 3/s for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , 1.8E-4 m3/s from 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2.3E-4 m3Is thereafter (§3.13.3). However, considering that the release was conservatively modeled as near an instantaneous release, the effective breathing rate used was the highest of the three values recommended, i.e., 3.5E-4 m 3/s.

Atmospheric dispersion value (X/Q) of 3.5E-4 s/m3 was used for the control room intake for 30 days (Table 2). The control room occupancy factors used were 100% during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% between 1 and 4 days, and 40% from 4 to 30 days (§3.14.6). A constant breathing rate of 3.5E-4 m 3/s was used for the control room (§3.14.6).

Simulation time steps were selected as follows: 0.025 h from 0 to 8 h, 0.1 h from 8 h to 24 h, and 0.4 h from 24 h to 720 h (30 days).

Flag value of 1 was selected for each to include plant model, scenario description, and results for every simulation in the output.

4.1.2 Auxiliary Input Files 4.1.2.1 Input File on Release Fraction and Timing, pnpp_fha.rft The auxiliary input file on release fraction and timing, pnpp_fha.rft, given in Attachment 2, shows that 100% of noble gas (Kr and Xe), halogen (Br and I), and alkali metals (Cs and Rb) groups were released to the containment in 1E-4 hours (0.36 s). The source term of Cs and Rb is zero because they were assumed to be retained completely by the pool, thus the timing and fraction of release for alkali metals group is immaterial. (Note that the input file identifies groups, except the noble gases group, by the representative nuclide in each group.

Thus, the halogen group is named as iodine group.)

3 During an FHA, activity in the gap is released into the fuel pool. The radionuclides that are not retained in the pool are immediately released from the top of the pool, in this case to the containment. The FHA source term, as listed in the last column of Table 5. was calculated by accounting for the pool DF. Therefore, the fuel pool was not specifically input to the RADTRAD model, and the complete source term was released to the containment but not to the pool.

Enclosure 4 PY-CEI/NRR-2674L Page 20 of 76 rstEne.r CALCULATIONPage 13 Perry Nuclear Power Plant PNPP No. 6077 Rev. 2/7101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 liTLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 4.1.2.2 Input File on Radionuclides Inventory and Decay Data, pnppfha.nif The FHA source term was determined using GE-calculated core inventory for 641 radioisotopes for different decay times for 1500 EFPD of operation. [DIN # 10] These included 17 bromine, 21 iodine, 15 krypton, and 18 xenon isotopes, amounting to a total of 71 isotopes. Note that isotopes of alkali metals (cesium and rubidium) were ignored as they were assumed to be retained completely in the pool. Of 71 isotopes, 48 isotopes with zero activity at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown were ignored. In addition three isotopes, BR 84, 1128, and XE138, which had activity less than I E-9 CiIMWt at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown, were also ignored. The remaining 20 isotopes of bromine, iodine, krypton, and xenon were chosen for TEDE analysis. Three additional nuclides, RB 87, RB 88, and CS1 35, which are daughter products of radioactive decay of KR 87, KR 88, and XEI 35M that were included in original set of 20 isotopes, were also used for TEDE analysis. Thus a total of 23 radionuclides of bromine, cesium, iodine, krypton, rubidium, and xenon were used for TEDE analysis. Table 5 lists the 23 isotopes used for this analysis and their core inventory at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reactor shutdown. [DIN # 10]

Table 5 shows the calculation of FHA source term. The core inventory per unit power was multiplied b-ythe gap fraction, reactor power, the fraction of fuel rods failed, and radial peaking factor and divided by the pool DF. The reactor power, fraction of fuel rods failed, and radial peaking factor were combined to form a single multiplication factor because each parameter was independent of isotopes. This calculation used a reactor power of 3833.2 MWt (102% of 3758 MWt) and a radial peaking factor of 2 (DIN # 21; §3.8). Of effective 64,208.32 rods in the core, this calculation assumed that 151 rods had failed during FHA (DIN # 12; §3.17.1.1). Using these values, the multiplication factor was calculated as 18.029 (= (3833.2 MWt)*((151 rods failed)/(64,208.32 rods in the core))*(2)).

Radionuclides decay data needed for this input file includes, identification of daughter nuclides and fractions and radioactive half-life as listed in Table 6. These data were obtained from the radionuclides data file for 825 radionuclides, lndexr.inp, which was provided with the MACCS2 computer code package. [DIN # 15]

4.1.2.3 Input File on Dose Conversion Factors, pnppfha.dcf The input file on dose conversion factors, pnpp_fha.dcf, which is given in Attachment 4, was developed using the values listed in Table 1. Note that the units of EDE (cloudshine) and CEDE (inhaled chronic) dose conversion factors to be used for this file are in Sv-m3/Bq-s and Sv/Bq.

4.2 Running RADTRAD Code The RADTRAD computer code was installed and executed on a Dell Latitude computer running on Windows NT Version 4.0 operating system as currently assigned to Hanry Wagage (owned by Matrix Leasing, no. 210158).

Satisfactory operation of the RADTRAD code on this computer has been confirmed by verification. [DIN # 28] The main input file, pnppfha.psf and the three auxiliary input files, pnpp_fha.rft, pnppfha.nif, and pnpp_fha.dcf were used as input to the code. These files are given in Attachment I through Attachment 4. The Output file, pnppfha.out, is given in Attachment 5.

4.3 Results of the RADTRAD Run for the Base Case The detailed results are given in the computer output file, pnppfha.out, which is given in Attachment 5. Table 7 lists TEDE values calculated for the control room, EAB, and outer boundary of the LPZ and compare these with regulatory limits. As shown in Table 7, the RADTRAD calculated TEDE values are well below the regulatory limits.

Enclosure 4 PY-CEI/NRR-2674L Page 21 of 76

.r 'e-y CALCULATIONPage 14 Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2t7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLEI

SUBJECT:

Fuel Handling Accident Using Alternative Source Term Table 5. Calculation of FHA Source Term Core Activity at Power*

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Fraction of Rods Failed)* FHA (Ci/MWt) Gap Fraction (Peaking Factor) Pool DFa Release No. Isotope [DIN # 101 [Table 3] (MWt) [§3.17.2, §3.17.31 Activity (Ci)

I BR 82 1.2390E+02 5% 18.029 200 5.5845E-01 2 BR 83 3.2960E+00 5% 18.029 200 1.4856E-02 3 KR 83M 1.2750E+01 5% 18.029 1 1.1494E+01 4 KR 85 4.1550E+02 10% 18.029 1 7.4911E+02 5 KR 85M 1.6560E+02 5% 18.029 1 1.4928E+02 6 KR 87 2.6830E-02 5% 18.029 1 2.4186E-02 7 KR88 5.1150E+01 5% 18.029 1 4.6109E+01 8 RB 87 N/Ab . I Z '9, 18.029 co 0 9 RB 88 5.7120E+01 & I?01- 18.029 00 0 10 1129 1.3910E-03 5% 18.029 200 6.2696E-06 11 1130 3.0390E+02 5% 18.029 200 1.3698E+00 12 1131 2.5290E+04 8% 18.029 200 1.8238E+02 13 1132 3.2140E+04 5% 18.029 200 1.4486E+02 14 1133 2.5280E+04 5% 18.029 200 1.1394E+02 15 1134 1.3320E-03 5% 18.029 200 6.0037E-06 16 1135 4.1600E+03 5% 18.029 200 1.8750E+01 17 XE129M 2.3050E-01 5% 18.029 1 2.0778E-01 18 XE131M 3.0440E+02 5% 18.029 1 2.7440E+02 19 XE133 5.1070E+04 5% 18.029 1 4.6037E+04 20 XE133M 1.5560E+03 5% 18.029 1 1.4027E+03 21 XE135 1.4060E+04 5% 18.029 1 1.2674E+04 22 XE135M 6.6640E+02 5% 18.029 1 6.0073E+02 , .7,o Z 23 CS135 2.7050E-02c I ,5%0IZ% 18.029 1 0o 0 Notes:

a. Note that the infinite value for pool DF of RB 87. RB 88, and CS135 indicates that cesium and rubidium were assumed to retain completely in the pool (§3.17.3).

b. DIN # 10 does not list the core inventory for RB 87.

c. The core inventory of RB 88 and CS135 are not used for the calculation because of infinite pool DF and are listed only for information. [DIN # 10]

Enclosure 4 PY-CEI/NRR-2674L Page 22 of 76 FirstEnerge Page 15 Perry Nuclear CALCULATION Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term Table 6. Radionuclides Decay Data Used for the Analysis [DIN # 15]

No. Isotope Daughter 1 Dau hter 2 Radioactive Half-life, t1,,

Nuclide Fraction Nuclide Fraction - (s) 1 BR 82 35.3 h 1.2708000E+05 2 BR 83 KR 83M 1 2.39 h 8.6040000E+03 3 KR83M 1.83 h 6.5880000E+03 4 KR 85 _ 10.72 y 3.3806592E+08 5 KR85M KR85 0.211 " 4.48 h 1.6128000E+04 6 KR 87 RB 87 1 76.3 m 4.5780000E+03 7 KR 88 RB 88 1 2.84 h 1.0224000E+04 8 RB 87 4.70E+10y 1.4821920E+18 9 RB 88 17.8 m 1.0680000E+03 10 1129 1.57E+07 y 4.9511520E+14 11 1130 12.36 h 4.4496000E+04 12 1131 XE131M 0.0111 8.04d 6.9465600E+05 13 1132 2.3 h 8.2800000E+03 14 1133 XE133M 0.029 XE133 0.971 20.8 h 7.4880000E+04 15 1134 1 52.6 m 3.1560000E+03 16 1135 XE135M 0.154 XE135 0.846 6.61 h 2.3796000E+04 17 XE129M _ 8d 6.9120000E+05 18 XE131M 11.9 d 1.0281600E+06 19 XE133 5.245 d 4.5316800E+05 20 XE133M XE133 1 2.188 d 1.8904320E+05 21 XE135 Cs-135 1 9.09 h 3.2724000E+04 22 XE135M XE135 0.9999 CS135 4.50E-05 15.29 m 9.1740000E+02 23 CS135 I I 2.30E+06 y 7.2532800E+13 Table 7. Comparison of TEDE Calculated for Control Room, EAB, and LPZ with Regulatory Limits Location " Control Room EAB LPZ RADTRAD results (rem)[Attachment 5] 1.03 1.44 0.161 Regulatory limit (rem) [§2] 5 6.3 6.3 RADTRAD Value Regulatory Limit 20.5% 22.9% 2.6%

Enclosure 4 PY-CEI/NRR-2674L Page 23 of 76

.jlrst* e t l Page 16

""rt~e~ CALCULATION Perry Nuclear Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 AE Analysis 3.2.15.14, Rev. 0 "TITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term 4.4 Sensitivity Analysis 4.4.1 Sensitivity Case 1: Effect of CR Isolation and Fresh Air Intake Sensitivity Case I was run to study the effect of control room isolation and fresh air intake.

The Rad Monitor was assumed to isolate the air intake at the worst possible time o~,c~the available activity is introduced into the control room. This is considered conservative as it maximizes the dMto the control room operators.

No inleakage of air 1.7.cz into the control room was assumed. This is conservative, as additional inl6akage of fresh air would tend to dilute the radioactivity in the control room. At 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , outside air purge at a rate of 5400cfm (6000cfm

-10%) assumed to initiate and continue until the end of dose analysis (30 days). [DIN # 26, 27] The 5400 cfm was considered to be conservative as it kept the activity in the control room longer.

The RADTRAD computer code was run with the main input file, which was changed to reflect the above flow rate data, and the same auxiliary input files that were used for the base case. The computer output is listed in Attachment 6. Table 8 compares the control room TEDE for this case with the base case and sensitivity case 2 results. The control room TEDE calculated for starting fresh air intake after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following control room isolation is 2.81 rem, which is below and 56% of the regulatory limit.

4.4.2 Sensitivity Case 2: Effect of CR Isolation and Recirculation Filtering Sensitivity Case 2 was run to study the effect of control room isolation and Recirculation Filtering.

The Rad Monitor was assumed to isolate the air intake at the worst possible time once the available activity is introduced into the control room. No inleakage of air into the control room was assumed. This is conservative, as additional inleakage of fresh air would tend to dilute the radioactivity in the control room. At 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the control room emergency recirculation was assumed to initiate. A recirculation flow rate of 27,000 cfm (30,000 -10%) and a charcoal efficiency of 50% were chosen to be consistent with assumptions in the LOCA analysis. [DIN

18]

The RADTRAD computer code was run with the main input file, which was changed to reflect the above flow rate data, and the same auxiliary input files that were used for the base case. The computer output is listed in . Table 8 compares the control. room TEDE for this case with the base casedand sensitivity case 1 A/'

results. The control room TEDE calculated for starting recirculation filtering after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following control room '01 isolation is 2.97 rem, which is below and 59% of the regulatory limit.

Table 8. Comparison of Sensitivity Analysis Results with the Base Case for TEDE Calculated for Control Room Control Room Case TEDE (rem)

Base Case Fable 7] 1.03 Sensitivity Case 1: Effect of CR Isolation and Fresh Air Intake [Attachment 6] 2.81 Sensitivity Case 2: Effect of CR Isolation and Recirculation Filtering [Attachment 71 2.97 5 COMPUTER INPUT AND OUTPUT The main input file for the base case, pnppfha.psf, and the three auxiliary input files, pnpp_fha.rft, pnppfha.nif, and pnpp_fha.dcf were used as input to the code. These files are given in Attachment I through Attachment 4.

The output file, pnppfha.out, is given in Attachment 5. Computer output for sensitivity cases 1 and 2 are given in and Attachment 7.

Enclosure 4 PY-CEI/NRR-2674L Page 24 of 76 Calc. No. 3.2.15.14, Rev. 0 Appendix A Page A-1 Appendix A. Fuel Handling Accident while Transiting over the Refueling Shield Once a fuel bundle is removed from the core, it is typically taken to either the Inclined Fuel Transfer Tube or the Upper Containment Fuel Pool racks located in an adjacent pool. In order to accomplish this the fuel must be moved from the Reactor Cavity over the Refueling Shield into the adjacent pool. The refueling shield is a device that is put in physically placed during a refueling outage in order to provide radiological shielding to the drywell area below the reactor cavity. The bottom of the refueling shield contains 8"of lead sandwiched bef*'een stainless steel plates. The shield is set onto locating pins on the Reactor Cavity/Steam Dryer gate opening and sits on the reactor flange.

In performing reviews of the License Amendment Request for Fuel Handling Accident re-analysis, it was determined that the current Technical Specification Bases for the amount of water above the reactor flange during movement of irradiated fuel within the RPV may not have considered all potential fuel handling accident scenarios.

Currently, Technical Specification 3.9.6 require a minimum water level of 22'-9" above the reactor flange during movement of irradiated fuel within the RPV. [DIN # 23] Regulatory Guidance 1.183 specifies a requirement for 23' providing the basis for the iodine decontamination factor used in the analysis. [DIN # 2] The Tech Spec Bases provides an assessment that while the worst case assumption include the dropping of the irradiated fuel assembly onto the reactor core, the possibility exists of the dropped assembly striking the RPV flange and releasing fission products. The Bases goes on to say that dropping an assembly on the RPV flange will result in reduced releases of fission gases. The Bases conclude that the operation with slightly less than 23' is acceptable in the event of C fuel drop on the reactor flange. The NRC in their response to Perry letter PY-CEI/NRR-1510L had accepted this.

[DIN # 29] Perry's letter further noted shielding >23' over the Upper Containment Pool fuel racks and Inclined Fuel Transfer System Upender.

The amount of shielding assumed in the current Technical Specification Section 3.9.6 is 22'-9".

The Refueling Shield bottom height is 9.25" (0.75" thick bottom plate, 8.0" thickness of lead, 0.5" thick upper plate). (Perry Drawing 4549-0194-1) [DIN # 17] Fuel bundle channel square dimension is 5.72" (Perry Drawing 4549-18-1) [DIN # 17]

Therefore, the least amount of shielding for a dropped bundle on the Refueling Shield is - 21'6" (21.5')

(=22.75' - ((9.25" + 5.72")/12)). This is less than the 22'-9" assumed in the Technical Specifications.

The purpose of this appendix is to examine the effect of this reduced water level on radiological doses at control room, EAB, and LPZ.

Overall decontamination factor for halogen species, DF, can be calculated using individual decontamination factors for inorganic and organic fractions of the species as given in Equation 10.

DF = finorg forg Equation 10 DFinorg DForg Equation 10 can be rewritten to obtain the decontamination factor for inorganic halogen species for known overall and organic decontamination factors as given in Equation 11.

S finorg Equation 11 1 org DF DForg Burley calculated the decontamination factor for inorganic (elemental) iodine as given by Equation 12. [DIN # 30]

DF, ,r 0 9 = exp-6 kff H Equation 12

,db vb )

Enclosure 4 PY-CEI/NRR-2674L Page 25 of 76 Calc. No. 3.2.15.14, Rev. 0 Ao5pendix A Page A-2 Where db - Bubble diameter H - Bubble rise height, i.e., the effective depth of water, defined as the water depth between the, top of the damaged fuel rods and the fuel pool surface keff - Mass transfer coefficient Vb - Bubble velocity Equation 12 can be rewritten as Equation 13 using a constant, C, as defined in Equation 14.

DFino* = exp(CH) Equation 13 C = 6keff I Equation 14 Using Equation 13, the decontamination factor for inorganic halogen species for a given pool depth can be expressed as given in Equation 15.

DF1 o,* =exp(CHo) Equation 15 Equation 16 is obtained by substituting for C from Equation 15 in Equation 13.

S= (DFin0 ,'°)Z-* Equation 16 Calculations for the design basis FHA, described in the main report, assumed an overall decontamination factor of 200 for a pool depth of 23', and inorganic and organic fractions of halogen species of 99.85% and 0.15%

(§3.17.1.3 and 3.17.2). Substituting these values in Equation 11, the corresponding decontamination factor for inorganic halogen species was calculated as 285.3. Using this value for a pool depth of 23', the decontamination factor for inorganic halogen species for a reduced pool depth of 21.5' was calculated as 197.3, using Equation 16.

The corresponding value of overall decontamination factor for halogen species was calculated as 152.4. using Equation 10, which is 76.2% of the overall decontamination factor of 200 used in the main report.

Calculations for the design basis FHA, described in the main report, assumed a total number of fuel rods damaged during an FHA as 151 (§3.17.1.1). However, the number of fuel rods damaged in for an FHA occurring while transiting over the Refueling Shield would be limited to 85.84, which is the total number of equivalent full length fuel rods in a fuel assembly. [DIN # 12] Therefore, the number of fuel rods damaged FHA while transiting over the Refueling Shield is 56.8% (=85.84/151) of that assumed for the design basis FHA.

The doses were calculated to be 74.5% (= (56.8%)1(76.2%)) of those calculated for the design basis FHA.

Therefore, the dose consequences for an FHA occurring while transiting over the Refueling Shield will be bounded by the dose consequences for the design basis FHA.

Calc. No. 3.2.15.14, Rev. 0 Attachment 1 Page I of 4 Enclosure 4 Radtrad 3.02 1/5/2000 PY-CEI/NRR-2674L perry fha Page 26 of 76 Nuclide Inventory File:

d:\hwagage\computer codes\radtrad\run batch\perry\pnpp_fha.nif Plant Power Level:

3.8332E+03 Compartments:

3 Compartment 1:

Containment 3

1.0000E+00 0

0 0

Compartment 2:

Environment 2

0.OOOOE+00 0

0 0

Compartment 3:

Control Room 1

3.6707E+05 0

0 0

Pathways:

3 Pathway 1:

Unfiltered Release to Environment 1

2 4

Pathway 2:

Unfiltered Environment to CR 2

3 2

Pathway 3:

Control Room Exhaust 3

2 2

End of Plant Model File Scenario Description Name:

Plant Model Filename:

Calc. No. 3.2.15.14, Rev. 0 Attachment 1 Page 2 of 4 Source 1 Term: Enclosure 4 PY-CEI/NRR-2674L I 1.0000E+00 Page 27 of 76 d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.dcf d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.rft o.OOOOE+00 1

O.OOOOE+00 0.9985E+00 0.0015E+00 1.OOOOE+00 Overlying Pool:

0 o.OOOOE+00 0

0 0

0 Compartments:

3 Compartment 1:

0 1

0 0

0 Compartment 2:

0 1

0 0

0 Compartment 3:

1 1

0 0

0 Pathways:

3 Pathway 1:

0 0

0

Catc. No. 3.2.15.14, Rev. 0 Atcmn 1 Attachment ae3o Page 3 of 4 Enclosure 4 0

PY-CEI/NRR-2674L 0 Page 28 of 76 0

1 2

o .OOOOE+00 1. OOOOE+10 7 .2000E+02 0. OOOOE+00 0

Pathway 2:

0 0

0 1

3 O.OOOOE+O0 6. GOOOE+03 o .OOOOE+00 0.OOOOE+00 0. OOOOE+00 0.000O1E+00 5.4000OE+03 0.OOOOE+00 0. OOOOE+00 o .OOOOE+00 7 .2000E+02 5 .4000E+03 0.OOOOE+00 0.OOOOE+00 o .OOOOE+00 0

0 0

0 Pathway 3:

0 0

0 1

3 0.OOOOE+00 G.G000E+03 o .OOOOE+00 o .OOOOE+00 o .OOOOE+0O 0.OOO1E+00 5.4000E+03 o .OOOOE+00 o .OOOOE+00 0. OOOOE+00 7.2000E+02 5.4000E+03 o .OOOOE+0O O.OOOOE+0O o .OOOOE+00 0

0 0

0 Dose Locations:

3 Location 1:

Exclusion Area Boundary 2

1 2

o0.0000E+00 4. 3000E-04

2. OOOOE+00 o .OOOOE+00 1

3 o .OOOOE+00 3. 5000E-04

8. OOOOE+OO 1. 8000E-04 2 .4000E+01 2. 3000E-04 0

Calc. No. 3.2.15.14. Rev. 0 Attachrment I Page 4 of 4 Location Outer 2:

Boundary of the LPZ Enclosure 4 PYCEI/NRR-2674L 2 Page 29 of 76 1

2 0.0000E+00 4.8000E-05 2.OOOOE+00 O.OOOOE+00 1

3 O.OOOOE+00 3.50OOE-04 8.OOOOE+00 1.8000E-04 2.4000E+01 2.3000E-04 0

Location 3:

Control Room 3

0 2

0.OOOOE+00 3.50OOE-04 7.2000E+02 0.OOOOE+00 1

4 0.0000E+00 1.0000E+00 2.4000E+01 6.0000E-01 9.6000E+01 4.OOOOE-01 7.2000E+02 0.OOOOE+00 Effective Volume Location:

1 2

0.OOOOE+00 3.5000E-04 2.OOOOE+00 0.OOOOE+00 Simulation Parameters:

4 O.OOOOE+00 2.5000E-02 8.OOOOE+00 1.0000E-01 2.4000E+01 4.OOOOE-01 7.2000E+02 0.OOOOE+00 Output Filename:

1 1

1 0

0 End of Scenario File

Caic. No. 3.2.1 5.14, Rev. 0 Atcmn 2 Attachment Page 1 of 1 Release Fraction and Timing Name:

Perry FHA PYn-ClEs1/NrR4R.26 74 L Duration (h) Page 30 of 76 0.0001E+00 0.OOOOE+00 0. OOOOE+00 0.0OOOE+00 Noble Gases:

0. lOOOE+01 0. OOOOE+00 0. OOOOE+00 0.OOOOE+00 Iodine:

0. lOOOE+01 0. OOOOE+00 o .OOOOE+0O 0.OOOOE+00 Cesium:

0.1000OE+01 0. OOOOE+00 0. OOOOE+00 0.OOOOE-,00 Tellurium:

0. OOOOE+00 0.OOOOE+00 0. OOOOE+00 0.OOOOE+00 Strontium:

0. OOCOE+00 0.OOOOE+00 0.OOOOE+00 0. OOOOE+0O Barium:

0.OOOOE+00 0.OOOOE+00 0. OOOOE+00 0. OOOOE+00 Ruthenium:

0.OOOOE+00 o .OOOOE+O0 0. OOOOE+00 0. OOOOE+00 Cerium:

o .OOOOE+00 0. OOOOE+00 0.OOOOE+00 0. OOOOE+00 Lanthanum:

0. OOOOE+00 0. OOOOE+00 o .OOOOE+00 0. OOOOE+00 Non-Radioactive Aerosols (kg):

0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 End of Release File

Calc. No. 3.2.15.14, Rev. 0 Attachment 3 Page 1 of 2 Enclosure 4 Nuclide Inventory Name: Rb-87 0.1000E+01 PY-CEI/NRR-2674L Perry EHA none 0.OOOOE+00 Page 31 of 76 Power Level: none 0.OOOOE+00 3.8332E+03 Nuclide 007:

Nuclides: Kr-88 23 1 Nuclide 001: 1.0224000000E+04 Br-82 0.8800E+02 2 4.6109E+01 1.2708000000E+05 Rb-88 0.1000E+01 0.8200E+02 none 0.0000E+00 5.5845E-01 none 0.OOOOE+00 none 0.0000E+00 Nuclide 008:

none 0.OOOOE+00 Rb-87 none 0.0000E+00 3 Nuclide 002: 1.4821920000E+18 Br-83 0.8700E+02 2 0.OOOOE+00 8.6040000000E+03 none 0.0000E+00 0.8300E+02 none 0.0000E+00 1.4856E-02 none 0.OOOOE+00 Kr-83m 0.1000E+01 Nuclide 009:

none 0.0000E+00 Rb-88 none 0.0000E+00 3 Nuclide 003: 1.0680000000E+03 Kr-83m 0.8800E+02 1 0.0000E+00 6.5880000000E+03 none 0.0000E+00 0.8300E+02 none 0.OOOOE+00 1.1494E+01 none 0.0000E+00 none 0.OOOOE+00 Nuclide 010:

none 0.0000E+00 1-129 none 0.OOOOE+00 2 Nuclide 004: 4.9511520000E+14 Kr-85 0.1290E+03 1 6.2696E-06 3.3806592000E+08 none 0.0000E+00 0.8500E+02 none 0.OOOOE+00 7.4911E+02 none 0.OOOOE+00 none 0.0000E+00 Nuclide 011:

none 0.OOOOE+00 1-130 none 0.0000E+00 2 Nuclide 005: 4.4496000000E+04 Kr-85m 0.1300E+03 1 1.3698E+00 1.6128000000E+04 none 0.OOOOE+00 0.8500E+02 none 0.OOOOE+00 1.4928E+02 none 0.OOOOE+00 Kr-85 0.2110E+00 Nuclide 012:

none 0.OOOOE+00 1-131 none 0.OOOOE+00 2 Nuclide 006: 6.9465600000E+05 Kr-87 0.1310E+03 1 1.8238E+02 0.4578000000E+04 Xe-131m 0.1110E-01 0.8700E+02 none 0.0000E+00 2.4186E-02 none 0.OOOOE+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 3 Page 2 of 2 Enclosure 4 Nuclide 013: 4.5316800000E+05 PY-CEIINRR-2674L 1-132 0.1330E+03 Page32of76 2 4.6037E+04 8.2800000000E+03 none 0.OOOOE+00 0.1320E+03 none 0.OOOOE+00 1.4486E+02 none 0.0000E+00 none 0.0000E+00 Nuclide 020:

none O.0000E+00 Xe-133m none 0.0000E+00 1 Nuclide 014: 1.8904320000E+05 1-133 0.1330E+03 2 1.4027E+03 7.4880000000E+04 Xe-133 0.1000E+01 0.1330E+03 none 0.OOOOE+00 1.1394E+02 none 0.OOOOE+00 Xe-133m 0.2900E-01 Nuclide 021:

Xe-133 0.9710E+00 Xe-135 none 0.OOOOE+00 1 Nuclide 015: 3.2724000000E+04 1-134 0.1350E+03 2 1.2674E+04 0.3156000000E+04 Cs-135 0.1000E+01 0.1340E+03 none 0.0000E+00 6.0037E-06 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 022:

none 0.OOOOE+00 Xe-135m none 0.OOOOE+00 1 Nuclide 016: 9.1740000000E+02 1-135 0.1350E+03 2 6.0073E+02 2.3796000000E+04 Xe-135 0.9999E+00 0.1350E+03 I Cs-135 4.5000E-05 1.8750E+01 none 0.OOOOE+00 Xe-135m 0.1540E+00 Nuclide 023:

Xe-135 0.8460E+00 Cs-135 none 0.OOOOE+00 3 Nuclide 017: 7.2532800000E+13 Xe-129m 0.1350E+03 1 0.OOOOE+00 6.9120000000E+05 none 0.OOOOE+00 0.1290E+03 none 0.OOOOE+00 2.0778E-01 none 0.OOOOE+00 none 0.OOOOE+00 End of Nuclear Inventory File none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 018:

Xe-131m 1

1.0281600000E+06 0.1310E+03 2.7440E+02 none 0.OO0OE+00 none 0.0000E+00 none 0.0000E+00 Nuclide 019:

Xe-133 1

00+aO00,0 oo+aooo-o 00+SOOO'O OO+aOOO*O 00+300010 00+aO00*0 00+300010 ziuvw aau 00+S00010 00+aO0010 00+aO00*0 00+aO0010 00+aO0010 00+aO00*0 00+300010 somm 00+aO0010 00+300010 00+aO0010 OO+HOOO *0 00+300010 00+aO00*0 00+S00010 ,LSVa'da oo+aoooo 00+3000-0 00+aO0010 00+aO00*0 00+aO0010 oo+aooo, o 00+300010 SCIVNOD 00+aO00,0 oo+aooo-o 00+aO0010 00+aO00 *0 00+aO0010 oo+aooo, o 00+S00010 W)E8-IX

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aqid SIHI NI aaNiasa sacaronm Ez (NEW NDIS aAijzaaaa uaaNivwaa aiOUAM uns aNOS uuvw ca'd soxnri

.Lsvaus savNog

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oo+aooolo oo+aooolo 00+3000*0 oo+aooo*o oo+aoooo oo+aooo*o 00+3000 0 SGVNO9 UISET-ax oo+aooolo oo+aoooo 00+300010 oo+aooo *0 oo+aooo*o oo+aooolo 00+3000*0 (UDi) NIXS oo+aoooo oo+aooo*o oo+aooo 10 oo+aooolo oo+aooo*o 00+3000,0 DIT-a061 IT aAIlOadaa oo+aooolo 00+900010 oo+aooo*o oo+aooo o oo+aooo*o oo+aooo *o 00+300010 'da(iNivwau oo+aooolo oO+SooOlo oo+aooo*o oo+aooo 0 oo+aoooo oo+aooo, o oo+aooo*o GIOUXHI oo+aoooo OO+SoOO*o oo+aooolo oo+aooolo oo+aooo*o 00+300010 00+3000*0 uns HNOU oo+aoooo oo+sooo*o oo+aooolo 00+3000*0 oo+aooo*o oo+aooolo oo+aooolo uavw aad 00+300010 oo+aooolo OO+aooo 10 00+3000*0 oo+aooo*o oo+HOoolo oo+aoooo soxnri oo+aooo*o oo+aooo*o 00+3000*0 oo+aooolo oo+aooo*o oo+aooolo oO+aooOlO isva'dg oo+aoooo oo+aooo 10 oo+aooolo oo+aooolo oo+aooolo oo+aoooo oo+aoooo SaVNOD SET-DX oo+aoooo oo+aooo*o oo+aooolo oo+aooo*o oo+aooo*o oo+aooo 10 oo+aooolo (UEM) NIAS Oo+aooolo oo+aooo 0 oo+aoooo ootaooo*o oo+aooo*o 00+300010 SI-aOLE'l 3Aiiosaaa oo+aooo"o oo+aooo 0 oo+aooo*o oo+aooo*o oo+aooo*o oo+aooo, o 00+3000*0 ,daaNivws'd Oo+aOoolO oo+aooo*o oo+aoooo 00+300010 oo+aooo*o oo+aooolo 00+3000*0 aioum.L oo+aooolo oo+aooolo oo+HoOolo 00+3000*0 oo+aooolo oo+aooo 10 oo+aooolo uns aNos oo+aooo*o oo+aooolo oo+aooo *0 oo+aooolo oo+aooo*o oo+aooolo 00+3000*0 uuvw aau oo+aooolo oo+aooo*o 00+3000*0 00+300010 oo+aooo-o oo+aooo*o oo+aoooo SE)NnIi oo+aooolo oo+aooo*o oo+aooolo oo+aooolo oo+aooolo oo+aooo *o oo+aooolo i.svadEt oo+aoooo 00+3000 0 Oo+aoOoO 00+3000,0 oo+aoooo oo+aoooo oo+aooo*o SaVNOD MEET-ax oo+aooo*o oo+aooo 0 oo+aooo*o oo+aooo*o oo+aooo"o oo+aooo*o oO+aoOO*O MW NIXS 00+300010 oo+aooolo oo+aooolo oo+aooolo 00+3000*0 oo+aooolo ST-ao9s*l aAIJZZaa3 00+300010 oo+aooolo oo+aooolo oo+aooolo 00+300010 oo+aooolo oo+aoooo 'daC[NIVWaU oo+aooolo oo+aooolo oo+aooolo oo+aooolo oo+aooolo 00+3000*0 oo+aooo*o aiouxHi Oo+aOoOlo oo+aooo 10 oo+aooolo 00+3000*0 oo+aoooo oo+aoooo oo+aoooo ms aNOH oo+aooolo oo+aooolo oo+aooolo oo+aooolo oo+aooolo 00+2100010 oo+aooo, o UUVW Cad oo+aooolo oo+aooolo oo+aooolo oo+aoooo oo+aooo*o oo+aooolo oo+aooolo somri oo+aooolo oo+aooolo oo+aooo*o oo+aooo*o oo+aooo*o oo+aooo *0 oo+aooo*o .Lclýa oo+aooolo oo+aooolo oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o 00+3000*0 SCIVNOD EET-GX oo+aooolo oo+aooolo 00 +300010 oo+aooolo 00+300010 oo+aoooo oo+aooolo OdDd) NIXs oo+aooo*o oo+aooo"o oo+aooolo 00+200010 oo+aooo*o 00+3000*0 91-3068*E aAI10aada oo+aooo*o oo+aooolo oo+aooolo oo+aooolo 00+3000*0 oo+aooolo oo+HooOlo 00+3000*0 oo+aooolo oo+aooolo 00+300000 oo+aooo 0 oo+aooo*o 00+3000*0 aIOH7,Hl.

oo+aooolo oo+aooo 10 oo+aooolo oo+aoooo 00+3000 0 oo+aooolo oO+SoOO

0 uns aNOEI oo+aooo*o oo+aooolo 00+3000"0 oo+aooo*D oo+aooo*o oo+aooolo oo+aooolo uuvw Gad oo+aooo*o 00+300010 00+300010 oo+aooo*o oo+aooo*o oo+aooo*o oo+aooolo sE)xni oo+aooo*o oo+aooo*o oo+aooo*o 00+3000*0 oo+aoooo oo+aooo, o 00+3000*0 .LSVNHH oo+aooo*o oo+aooo 10 oo+aooolo oo+aooolo 00+3000*0 oo+aooolo oo+aooolo SaVNOD IHIET-aX oo+aooolo oo+aooolo oo+aooo 10 00+3000*0 oo+aooo*o oo+aooo*o oo+aooolo ('d EM ) N rA S oo+aooo*o oo+aooolo oo+aooo*o 00+300000 oo+aooo*o oo+aoooo ST-30901T 3AIlDadda oo+aooolo oo+aooolo oo+aooo*o oo+aooo*o oo+aooo*o 00+300010 00+3000*0 UaaNIVWa'd oo+aooo-o oo+aooolo 00+200010 oo+aooolo oo+aooo, o 00+2000*0 00+300010 aIO'dAlU oo+aooolo oo+aoooo oo+aooolo oo+aooolo oo+aoooo oo+aooo*o oo+aooolo UnS SNOB oo+aoooo oo+aooolo oo+aooolo oO+aOoOlo oo+aooo*o oo+aooo*o OO+aOOolO uuvw cau oo+aoooo oo+aooolo oo+aooo*o oo+aooo*o oo+aoooo oo+aooolo oo+aoooo sE)Nni oo+aoooo oo+aOOO 0 oo+aooo*o oo+aooolo oo+aooo *0 oo+aooo*o oo+aooo*o Is-Ma oo+aooolo 00+2000 0 oo+aooolo oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o SCIVNOD IUGZT-9x oo+aooolo oo+aoOO 0 oo+aooo*o oo+aoooo 00+2000*0 00+3000'0 Oo+aoOOlo OdDd) NDIS oo+aooo 0 oi-aoZE*E 00+3000*0 oo+aooolo oo+aooo*o oo+aooo*o t'TaD,6z* 9 HAuomaa oo+aooo 0 oo+aooo *0 oo+aooolo oo+aooolo oo+aooo *o oo+aoooo oo+aooolo 'daaNivwa'd oo+aooolo oo+aooolo oo+aooolo oo+aooolo oo+aooo*o oo+aooo*o oo+aooolo aiod7m oo+aooolo oo+aooo*o oo+aooolo oo+aooo"o 00+3000'0 oo+aoooo 00+3000*0 UnS HNOU 00+300010 oo+aooo*o 00+300010 oo+aooolo oo+aoooo oo+aoooo oo+aooo *o uuQw ca'a oo+aooo*o oo+aooo 10 oo+aooOlo oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o SE)Nrri oo+aooolo oO+HOoOlO oo+aoou*o oo+aoooo oo+aoooo 00+3000*0 OO+aOoOlO .Lsvada 00+300000 oo+HoOOlO oo+aooolo oo+aooo*o oo+aoooo oo+aooo, 0 oo+aooolo SaVNOD 9L 10 9C 06ed 9 JO t? Obed It7L9Z-1LýN1130-Ad j7 ainsolou3 17 )Uawqoeilv 0 -ON _31eO

Calc. No. 3.2.15.14. Rev. 0 Enclosure 4 Aftachment 4 PY-CEI/NRR-2674L Page 5 of 5 Page 37 of 76 RED MARR O.OOOE+00 O.OOOE+00 0. 000E.06 0. OOOE+00 0.000E+00 O.OOOE+00 O.OOOE+00 BONE SUR O.OOOE+00 O.OOOE-00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 THYROID O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 REMAINDER O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 EFFECTIVE 2.040E-14 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 SKIN(FGR) O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 Cs-135 GONADS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. 000E;:00 RED MARR O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BONE SUR O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 THYROID O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 REMAINDER O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 EFFECTIVE 5.650E-19 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 1.230E-09 O.OOOE+00 SKIN(FGR) O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 1 of 16 Enclosure 4 PY-CEI/NRR-2674L RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48 Page 38 of 76 File information Plant file name = pnpp-fha.psf Inventory file name = d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.nif Scenario file name = NEW SDF.SDF Release file name = d: \hwagage\computer codes\radtrad\run batch\perry\pnppfha.rft Dose conversion file name = d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.dcf 4*4*4* 4*4* 4*4*4*4* 4*4*

ifif 4* 4* 4*4*4*4*4* 4* 4* 4*4*4*4*4*

4* if 4* 4* 4* # # 4* 4* 4* 4*

4* 4* 4* 4*

4* 4* ## 4 4* 4* 4* 4* 4*

4*4*4*4*4* 4*4*4*4* #4*4* if 4* 4*##4 #4*4*4*4* 4* 4*

4* 4*

4*4* 4* 4* 4*4 4* 4* 4* 4*

4* 4*4* 4* 4* 4* 4*

if 4*

4*4*4*4* 4* 4* 4* 4*4*4*4* 4*

Radtrad.3.02 1/5/2000 perry fha Nuclide Inventory File:

d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.nif Plant Power Level:

3.8332E+03 Compartments:

3 Compartment 1:

Containment 3

1.OOOOE+00 0

0 0

Compartment 2:

Environment 2

0.0000E+00 0

0 0

Compartment 3:

Control Room 1

3.6707E+05 0

0 0

Catc. No. 3-2.15.14, Rev. 0 Attachment 5 Page 2 of 16 Pathways 3 : Enclosure 4 PY-CEI/NRR-2674L Pathway 1: Page 39 of 76 Unfiltered Release to Environment 1

2 4

Pathway 2:

Unfiltered Environment to CR 2

3 2

Pathway 3:

Control Room Exhaust 3

2 2

End of Plant Model File Scenario Description Name:

Plant Model Filename:

Source Term:

1 1 1.0000E+00 d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.dcf d:\hwagage\computer codes\radtrad\run batch\perry\pnpp_fha.rft O.OOOOE+00 1

0.OOOOE+00 0.9985E+00 0.0015E+00 1.OOOOE+00 Overlying Pool:

0 O.OOOOE+00 0

0 0

0 Compartments:

3 Compartment 1:

0 1

0 0

0 Compartment 2:

0 1

0 0

0

Caic. No. 3.2.1 5.14. Rev. 0Atchet5Pg3of1 Attachment 5 Page 3 of 16 Compartment 3: Enclosure 4 1 PY-CEI/NRR-2674L 1 Page 40 of 76 0

0 0

Pathways:

3 Pathway 1:

0 0

1 2

0.0000E+00 1. 000OE-i10 7.2000E+02 0.000 OE+00 0

Pathway .2:

0 0

0 1

3 0.OOOOE+00 6 .6000E+03 0.OOOOE+00 0.OOOOE+00 0.0000E+00 0 .0001E+00 5.4000E+03 o .OOOOE+00 0.0000E+00 0.OOOOE+00 7 .2000E+02 5 .4000E+03 0.OOOOE+00 o .OOOOE+00 O.OOOOE÷00 0

0 0

0 Pathway 3:

0 0

0 1

3 o .OOOOE+00 6.6000E+03 o .OOOOE+00 0.000 OE+00 O.OOOOE+O0 o .OOO1E+00 5.4000E+03 o .OOOOE+O0 o .OOOOE+00 0.OOOOE+O0 7.2000E+02 5.4000OE+03 o .OOOOE+00 o .OOOOE+00 0 .OOOOE+00 0

0 0

0

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 4 of 16 Enclosure 4 Dose Locations:

PY-CEIINRR-2674L 3

Page 41 of 76 Location 1:

Exclusion Area Boundary 2

1 2

0.000OE+00 4.3000E-04 2.0000E+00 o.OOOOE+00 1

3 O.0000E+00 3.5000E-04 8.0000E+00 1.8000E-04 2.4000E+01 2.30OOE-04 0

Location 2:

Outer Boundary of the LPZ 2

1 2

o.OOOOE+00 4.8000E-05 2.OOOOE+00 O.0000E+00 1

3 0.0000E+00 3.5000E-04 8.OOOOE+00 1.8000E-04 2.4000E+01 2.3000E-04 0

Location 3:

Control Room 3

0 1

2 O.OOOOE+00 3.5000E-04 i 7.2000E+02 O.0000E+00 1

4 0.0000E+00 1.0000E+00 2.4000E+01 .O0000E-01 9.6000E+01 4.OOOOE-01 7.2000E+02 O.OOOOE+00 Effective Volume Location:

1 2

0.0000E+00 3.5000E-04 2.0000E+00 0.OOOOE+00 Simulation Parameters:

4 0.OOOOE+00 2.5000E-02 8.0000E+00 1.0000E-01 2.4000E+01 4.OOOOE-01 7.2000E+02 0.OOOOE+00 Output Filename:

1 1

1 0

0 End of Scenario File

Calc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 5 of 16 Enclosure 4 PY-CEI/NRR-2674L RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48 Page 42 of 76 Plant Description Number of Nuclides = 23 Inventory Power = 3.8332E+03 MWth Plant Power Level = 3.8332E+03 MWth Number of compartments - 3 Compartment information Compartment number 1 (Source term fraction = 1.OOOOE+00

)

Name: Containment Compartment volume = l.0000E+00 (Cubic feet)

Pathways into and out of compartment 1 Pathway to compartment number 2: Unfiltered Release to Environment Compartment numbcr 2 Name: Environment Pathways into and out of compartment 2 Pathway to compartment number 3: Unfiltered Environment to CR Pathway from compartment number 1: Unfiltered Release to Environment Pathway from compartment number 3: Control Room Exhaust Compartment number 3 Name: Control Room Compartment volume 3.6707E+05 (Cubic feet)

Pathways into and out of compartment 3 Pathway to compartment number 2: Control Room Exhaust Pathway from compartment number 2: Unfiltered Environment to CR Total number of pathways = 3

Caic. No. 3.2.15.14, Rev. 0 Attachment 5 Page 6 of 16 Enclosure 4 PY-CEI/NRR-2674L Page 43 of 76 RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48 Scenario Description Radioactive Decay is enabled Calculation of Daughters is enabled RELEASENAME = Perry FHA Release Fractions and Timings GAP EARLY IN-VESSEL 0.0001 hrs 0.0000 hrs NOBLES 1.0000E+00 0.OOOOE+00 IODINE 1.0000E+00 0.OOOOE+00 CESIUM 1.OOOOE+00 0.0000E+00 TELLURIUM 0.0000E+00 0.0000E+00 STRONTIUM 0.0000E+00 0.0000E+00 BARIUM 0.0000E+00 0.0000E+00 RUTHENIUM O.0000E+00 0.0000E+00 CERIUM o.0000E+00 0.0000E+00 LANTHANUM 0.OOOOE+00 0.OOOOE+00 Iodine fractions Aerosol 0.OOOOE+00 Elemental 9.9850E-01 Organic 1.5000E-03 COMPARTMENT DATA Compartment number 1: Containment Compartment number 2: Environment Compartment number 3: Control Room PATHWAY DATA Pathway number 1: Unfiltered Release to Environment Convection Data Time (hr) Flow Rate (t / day) 0.OOOOE+00 1.0000E+10 7.2000E+02 0.OOOOE+00 Pathway number 2: Unfiltered Environment to CR Pathway Filter: Removal Data

Time (hr) Flow Rate Filter Efficiencies (M)

(cfm) Aerosol Elemental Organic 0.OOOOE+00 6.6000E+03 O.0000E+00 0.OOOOE+00 0.OOOOE+00 1.OOOOE-04 5.4000E+03 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 7.2000E+02 5.4000E+03 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Pathway number 3: Control Room Exhaust Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerosol Elemental Organic

Catc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 7 of 16 o.OOOOE+00 6.6000E+03 o.O000E+00 O.OOOOE+0O Enclosure 4 o .OOOOE+00 1.0000E-04 5.4000E+03 0.0000E+00 O.OOOOE+O0 PY-CEIINRR-2674L 0.0000E+00 O.OOOOE+00 7.2000E+02 5.4000E+03 O.OOOOE+00 Page 44 of 76 o .OOOOE+00 LOCATION DATA

'Location Exclusion Area Boundary is in compartment 2 Location X/Q Data Time (hr) X/Q (s

m^-3 )

0.OOOOE+00 4.3000E-04 2.OOOOE+00 O.OOOOE+00 Location Breathing Rate Data Time (hr) Breathing Rat( W(m3

sec^-l)

O.OOOOE+00 3.5( OOE-04 8.OOOOE+00 1.8( OOE-04 2.4000E+01 2.3( OOE-04 Location Outer Boundary of the LI Z is in comrpartment 2 Location X/Q Data Time (hr) X/Q (s

m^-3) 0.OOOOE+00 4.8000E-05 2.OOOOE+00 0.OOOOE+00 Location Breathing Rate Data Time (hr) Breathing Rate (m^3

sec^-l)

O.OOOOE+00 3.SOOOE-04 8.OOOOE+00 -.8000E-04 2.4000E+01 2.3000E-04 Location Control Room is in compartment 3 Location X/Q Data Time (hr) X/Q (s

m^-3) 0.0000E+00-" 3.5000E-04 2.OOOOE+00 0.0000E+00 i Location Breathing Rate Data Time (hr) Breathing Rate (mW3

sec--l)

O.OOOOE+00 3.5000E-04 7.2000E+02 O.OOOOE+00 Location Occupancy Factor Data Time (hr) OccupanCy Factor O.OOOOE+00 1.0000E+00 2.4000E+01 6.OOOOE-01 9.6000E+01 4.OOOOE-01 7.2000E+02 0.OOOOE+00 USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS Time Time step 0.OOOOE+00 2.5000E-02 8.0000E+00 1.00QOE-01 2.4000E+01 4.OOOOE-01 7.2000E+02 O.OOOOE+00

Calc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 8 of16 Enclosure 4 PY-CEI/NRR-2674L RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48 Page 45 of 76 ft ftft ft ft ftfttftf ftf t#t## ft f ##ftft#

ft ft ft ft ft ft ft #t f ft ft ff ft ft ft ft

f ft ft *f

f ft ft f--

ft ftttff ft ft #t ft ftftftft ftftft# ##ftf #t Dose, Detailed model and Detailed Inventory Output

fl######ff ##*ftftf##ft########ft#####f#############################f####

Exclusion Area Boundary Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.2456E-01 0.0000E+00 1.4377E+00 Accumulated dose (rem) 4.2456E-01 0.OOOOE+00 1.4377E+00 Outer Boundary of the LPZ Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.7393E-02 0.OOOOE+00 1.6048E-01 Accumulated dose (rem) 4.7393E-02 0.0000E+00 1.6048E-01 Control Room Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 1.2078E-06 0.0000E+00 4.5686E-05 Accumulated dose (rem) 1.2078E-06 0.OOOOE+00 4.5686E-05 Control Room Compartment Nuclide Inventory:

Time (h) = 0.0001 Ci " kg Atoms Bq Br-82 6.0733E-04 5.6097E-13 4.1198E+12 2.2471E+07 Br-83 1.6156E-05 1.0227E-15 7.4200E+09 5.9776E+05 Kr-83m 1.2500E-02 6.0583E-13 4.3957E+12 4.6248E+08 Kr-85 8.1467E-01 2.0751E-06 1.4702E+19 3.0143E+10 Kr-85m 1.6234E-01 1.9727E-11 1.3976E+14 6.0067E+09 Kr-87 2.6301E-05 9.2854E-16 6.4273E+09 9.7315E+05 Kr-88 5.0143E-02 3.9989E-12 2.7366E+13 1.8553E+09 Rb-87 4 .4281E-24 5.0613E-20 3.5034E+05 1.6384E-13 Rb-88 1.1716E-05 9.7601E-17 6. 6792E+08 4.3349E+05 1-129 6.8183E-09 3.8601E-08 1.8020E+17 2.5228E+02 1-130 1.4897E-03 7.6381E-13 3.5383E+12 5.5118E+07 1-131 1.9834E-01 1.5999E-09 7.3546E+15 7.3387E+09 1-132 1.5753E-01 1.5262E-11 6. 9627E+13 5.8288E+09 1-133 1.2391E-01 1.0938E-10 4.9529E+14 4.5847E+09 1-134 6.5286E-09 2.4473E-19 I. 0999E+06 2.4156E+02 1-135 2.0391E-02 5.8063E-12 2.5901E+13 7.5446E+08 Xe-129m 2.2597E-04 1.7859E-12 8. 3372E+12 8.3607E+06 Xe-131m 2.9842E-01 3.5627E-09 1. 6378E+16 1.1041E+10 Xe-133 5.0066E+01 2.6747E-07 1.2111E+18 1.8525E+12 Xe-133m 1.5255E+00 3.3997E-09 1. 5394E+16 5.6442E+10

Catc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 9 of 16 Xe-135 1.3783E+01 5.3973E-09 2.4076E+16 Enclosure 4 5.0998E+11 Xe-135m 6.5313E-01 7.1700E-12 3.1984E+13 2.4166E+10 PY-CEI/NRR-2674L Cs-135 4.7418E-14 4.1157E-14 1.8359E+11 1.7545E-03 Page 46 of 76 Control Room Transport Group Inventory:

Overlying Time (h) = 0.0001 Atmosphere Sump Pool Noble gases (atoms) 1.5969E+19 0.OOOOE+00 o.OOOOE+00 Elemental I (atoms) 1.8787E+17 0.00O0E+00 0.OOOOE+00 Organic I (atoms) 2.8223E+14 0.OOOOE+00 0.0000E+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 o.OOOOE+00 Deposition Recirculating Time (h) = 0.0001 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.OOOOE+00 Elemental I (atoms) O.OOOOE+00 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 O.OOOOE+00 Aerosols (kg) O.OOOOE+O0 O. 000E+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) = 0.0001 Filter Noble gases (atoms) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 Control Room Exhaust Transport Group Inventory:

Pathway Time (h) = 0.0001 Filter Noble gases (atoms) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 Exclusion Area Boundary Doses:

Time (h) = 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.3779E-03 0.OOOOE+00 4.7275E-03 Accumulated dose (rem) 4.2594i-01 0.OOOOE+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.5382E-04 0.0000E+00 5.2772E-04 Accumulated dose (rem) 4.7547E-02 0.OOOOE+00 1.6101E-01 Control Room Doses:

Time (h) = 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.1084E-02 0.OOOOE+00 8.5359E-01 Accumulated dose (rem) 2.1086E-02 0.OOOOE+00 8.5363E-01 Control Room Compartment Nuclide Inventory:

Time (h) = 2.0000 Ci kg Atoms Bq Br-82 1.0029E-04 9.2638E-14 6. 8034E+11 3.7108E+06 Br-83 1.5536E-06 9.8342E-17 7.1353E+08 5.7483E+04 Kr-83m 1.0076E-03 4.8836E-14 3. 5434E+II 3.7281E+07 Kr-85 1.3992E-01 3.5639E-07 2. 5250E+18 5.1771E+09

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 10 of 16 Kr-85m 2.0462E-02 2.4t65E-12 1.7616E+13 7.5711E+08 Enclosure 4 Kr-87 1.5187E-06 5.3615E-17 3.7112E+08 5.6191E+04 PY-CEI/NRR-2674L Kr-88 5.2861E-03 4.2156E-13 2.8849E+12 1.9559E+08 Page 47 of 76 Rb-87 9.2626E-21 1.0587E-16 7.3285E+08 3.4272E-10 Rb-88 5.9844E-03 4.9854E-14 3.4117E+11 2.2142E+08 1-129 1.1711E-09 6.6298E-09 3.0950E+16 4.3329E+01 1-130 2.2871E-04 1.1727E-13 5.4323E+11 8.4623E+06 1-131 3.3822E-02 2.7281E-10 1.2541E+15 1.2514E+09 1-132 1.4809E-02 1.4347E-12 6.5453E+12 5.4793E+08 1-133 1.9910E-02 1.7576E-11 7.9582E+13 7.3667E+08 1-135 2.8396E-03 8.0858E-13 3.6069E+12 1.0507E+08 Xe-129m 3.8531E-05 3.0453E-13 1.4216E+12 1.4256E+06 Xe-131m 5.1007E-02 6.0896E-10 2.7994E+15 1.8873E+09 Xe-133 8.5079E+00 4.5453E-08 2.0581E+17 3.1479E+11 Xe-133m 2.5519E-01 5.6873E-10 2.5752E+15 9.4421E+09 Xe-135 2.0357E+00 7.9713E-10 3.5559E+15 7.5319E+10 Xe-135m 9.5488E-04 1.0483E-14 4.6761E+10 3.5331E+07 Cs-135 1.5125E-10 1.3128E-l0 5.8561E+14 5.5963E+00 Control Room Transport Group" Inventory:

Overlying Time (h) = 2.0000 Atmosphere Sump Pool Noble gases (atoms) 2.7427E+18 0.0000E+00 0.0000E+00 Elemental I (atoms) 3.2268E+16 0.0000E+00 0.0000E+00 Organic I (atoms) 4.8474E+13 o.OOOOE+00 0.0000E+00 Aerosols (kg) 0.000OE+00 0.0000E+00 0.0000E+00 Deposition Recirculating Time (h) = 2.0000 Surfaces Filter Noble gases (atoms) 0.0000E+00 0.0000E+00 Elemental I (atoms) 0.OOOOE+00 0.0000E+00 Organic I (atoms) 0.O000E+00 0.0000E+00 Aerosols (kg) 0.0000E+00 0.0000E+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) = 2.0000 Filter Noble gases (atoms) 0.0000E+00 Elemental I (atoms) O.0000E+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.0000E;60 Control Room Exhaust Transport Group Inventory:

Pathway Time (h) = 2.0000 Filter Noble gases (atoms) 0.0000E+00 Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 Exclusion Area Boundary Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.00002+00 0.0000E+00 Accumulated dose (rem) 4.2594E-01 0.0000E+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00

Calc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 11 of 16 Accumulated dose (rem) 4.7547E-02 0.6000E+00 1.6101E-01 Enclosure 4 PY-CEI/NRR-2674L Control Room Doses: Page 48 of 76 Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.6824E-03 O.0000E+00 1.7162E-01 Accumulated dose (rem) 2.4768E-02 0.OOOOE+00 1.0253E+00 Control Room Compartment Nuclide Inventory:

Time (h) = 8.0000 Ci kg Atoms Bq Br-82 4.4677E-07 4.1267E-16 3.0307E+09 1.6530E+04 Br-83 1.3664E-09 8.6496E-20 6.2758E+05 5.0558E+01 Kr-83m 5.2275E-07 2.5337E-17 1.8383E+08 1.9342E+04 Kr-85 7.0120E-04 1.7860E-09 1.2654E+16 2.5944E+07 Kr-85m 4.0529E-05 4.9249E-15 3.4892E+10 1.4996E+06 Kr-88 6.1254E-06 4.8850E-16 3.3430E+09 2.2664E+05 Rb-87 6.9036E-23 7.8908E-19 5.4620E+06 2.5543E-12 Rb-88 7.0418E-06 5.8663E-17 4.0145E+08 2.6055E+05 1-129 5.8690E-12 3.3226E-11 1.5511E+14 2.1715E-01 1-130 8.1872E-07 4.1978E-16 1.9446E+09 3.0293E+04 1-131 1.6589E-04 1.3381E-12 6.1513E+12 6.1379E+06 1-132 1.2168E-05 1.1788E-15 5.3779E+09 4.5021E+05 1-133 8.1699E-05 7.2121E-14 3.2656E+1I 3.0229E+06 1-135 7.5856E-06 2.160OE-15 9.6354E+09 2.8067E+05 Xe-129m 1.8897E-07 "1.4935E-15 6.9721E+09 6.9917E+03 Xe-131m 2.5196E-04 3.0081E-12 1.3828E+13 9.3226E+06 Xe-133 4.1296E-02 2.2062E-10 9.9894E+14. 1.5279E+09 Xe-133m 1.1817E-03 2.6337E-12 1. 1925E+13 4.3725E+07 Xe-135 6.4602E-03 2.5297E-12 1.1285E+13 2.3903E+08 Xe-135m 1.2568E-06 1.3797E-17 6.1547E+07 4.6502E+04 Cs-135 2.4484E-12 2.12SIE-12 9.4797E+12 9.0591E-02 Control Room Transport Group Inventory:

Overlying Time (h) = 8.0000 Atmosphere Sump Pool Noble gases (atoms) 1.3745E+16 0.OOOOE+00 0.0000E+00 Elemental I (atoms) 1.6171E+14 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 2.4294E+11 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 O.0000E+00 0.0000E+00 Depositi6n Recirculating Time (h) = 8.0000 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) o.OOOOE+00 0.OOOOE+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) = 8.0000 Filter Noble gases (atoms) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 Control Room Exhaust Transport Group Inventory:

Pathway Time (h) = 8.0000 Filter Noble gases (atoms) O.OOOOE+00

Calc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 12 of 16 Elemental I (atoms) o .OOOOE-i00 Enclosure 4 Organic I (atoms) O.OOOOE+00 PY-CEI/NRR-2674L Aerosols (kg) 0.OOOOE+O0 Page 49 of 76 Exclusion Area Boundary Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose-(rem) 0.OOOOE+00 0.OOOOE+00 o.0000E+00 Accumulated dose (rem) 4.2594E-01 o.0000E+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.000OE+00 o.0000E+00 Accumulated dose (rem) 4.7547E-02 0.0000E+00 1.6101E-01 Control Room Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.3641E-05 0.0000E+00 8.2602E-04 Accumulated dose (rem) 2.4782E-02 0.OOOOE+00 1.0261E+00 Control Room Compartment Nuclide Inventory:

Time (h) 24.0000 Ci kg Atoms Bq Kr-85 5.1571E-10 1.3136E-15 9.3064E+09 1.9081E+01 1-129 4.3169E-18 2.4440E-17 1.1409E+08 1.5973E-07 1-131 1.1520E-10 9.2926E-19 4.2718E+06 4.2626E+00 1-133 3.5259E-11 3.1125E-20 1.4093E+05 1.3046E+00 Xe-131m 1.7832E-10 2.1289E-18 9.7868E+06 6.5979E+00 Xe-133 2.7883E-08 1.4896E-16 6.7449E+08 1.0317E+03 Xe-133m 7.0400E-I0 1.5690E-18 7.1041E+06 2.6048E+01 Xe-135 1.4044E-09 5.4995E-19 2.4532E+06 5.1963E+01 Cs-135 3.3126E-18 2.8752E-18 1.2826E+07 1.2257E-07 Control Room Transport Group Inventory:

Overlying Time (h) = 24.0000 Atmosphere Sump Pool Noble gases (atoms) 1.0110E+10 0.OOOOE+00 0.0000E+00 Elemental I (atoms) 1.1895E+08 O.0000E+00 0.OOOOE+00 Organic I (atoms) 1.7869E+05 0.OOOOE+00 0.0000E+00 Aerosols (kg) 0.OOOOE+bb 0.OOOOE+00 0.000OE+00 Deposition Recirculating Time (h) = 24.0000 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.0000E+00 Elemental I (atoms) 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 0.000E+00 0.OOOOE+00 Aerosols (kg) o.OOOOE+00 0.OOOOE+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) = 24.0000 Filter Noble gases (atoms) 0.0000E+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) o.OOOOE+00

No. 3.2.15.14. Rev. 0 Attachment 5 Page 13 of 16 trol Room Exhaust Transport Group Invrentory: Enclosure 4 PY-CEI/NRR-2674L Pathway Page 50 of 76 e (h) = 24.0000 Filter le gases (atoms) 0.OOOOE+00 mental I (atoms) 0.000OE+00 anic I (atoms) O.OOOOE+00 osols (kg) o.OOOOE+00 lusion Area Boundary Doses:

e (h) = 96.0000 Whole Body Thyroid TEDE ta dose (rem) 0.OOOOE+00 0.OOOOE+00 O.0000E+00 umulated dose (rem) 4.2594E-01 0.OOOOE+00 1.4424E+00 er Boundary of the LPZ Doses:

e (h) = 96.0000 Whole Body Thyroid TEDE ta dose (rem) o.0000E+00 0.OOOOE+00 0.OOOOE+00 umulated dose (rem) 4.7547E-02 o.OOOOE+00 1.6101E-01 trol Room Doses:

e (h) = 96.0000 Whole Body Thyroid TEDE ta dose (rem) 3.6064E-12 0.000OE+00 3.3485E-l0 amulated dose (rem) 2.4782E-02 0.OOOOE+00 1.0261E+00 trol Room Compartment Nuclide Inventory:

e (h) = 96.0000 Ci kg Atoms Bq

rol Room Transport Group Inventory:

Overlying S(h) = 96.0000 Atmosphere Sump Pool le gases (atoms) 2.5382E-18 0.OOOOE+00 0.000OE+00 nental I (atoms) 2.9862E-20 0.OOOOE+00 0.OOOOE+00 inic I (atoms) 4.4860E-23 0.OOOOE+00 0.OOOOE+00 Dsols (kg) O.OOOOE+00 o.OOOOE+00 O.OOOOE+00 Deposition Recirculating

ý (h) = 96.0000 Surfaces Filter Le gases (atoms) 0.OOOOE+'60 0.OOOOE+00 nental I (atoms) 0.0000E+00 0.OOOOE+00 inic I (atoms) 0.OOOOE+00 0.OOOOE+00

)sols (kg) 0.0000E+00 0.OOOOE+00 iltered Environment to CR Transport Group Inventory:

Pathway S(h) = 96.0000 Filter Le gases (atoms) 0.OOOOE+00 nental I (atoms) 0.OOOOE+00 inic I (atoms) 0.0000E+00

)sols (kg) O.OOOOE+00

rol Room Exhaust Transport Group Inventory:

Pathway S(h) = 96.0000 Filter

.e gases (atoms) o.OOOOE+00 iental I (atoms) O.OOOOE+00 inic I (atoms) 0.OOOOE+00

Calc. No. 32.15.14, Rev. 0 Attachment 5 Page 14 of 16 Aerosols (kg) 0.0000E+00 Enclosure 4 PY-CEIINRR-2674L Exclusion Area Boundary Doses: Page 51 of 76 Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Accumulated dose (rem) 4.2594E-01 0.0000E+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) o.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Accumulated dose (rem) 4.7547E-02 0.OOOOE+00 1.6101E-01 Control Room Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.0169E-40 O.OOOOE+00 4.1186E-38 Accumulated dose (rem) 2.4782E-02 0.OOOOE+00 1.0261E+00 Control Room Compartment Nuclide Inventory:

Time (h) = 720.0000 Ci kg Atoms Bq Control Room Transport Group Inventory:

Overlying Time (h) = 720.0000 Atmosphere Sump Pool Noble gases (atoms) 1.5937-257 0.OOOOE+00 0.OOOOE+00 Elemental I (atoms) 1.8750-259 0.OOOOE+O0 0.OOOOE+00 Organic I (atoms) 2.8167-262 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Deposition Recirculating Time (h) = 720.0000 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.0000E+00 Elemental I (atoms) 0.0000E+00 0.OOOOE+00 Organic I (atoms) 0.0000E+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.0000E+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway" Time (h) = 720.0000 Filter Noble gases (atoms) 0.0000E+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 Control Room Exhaust Transport Group Inventory:

Pathway Time (h) = 720.0000 Filter Noble gases (atoms) 0.OOOOE+00 Elemental I (atoms) 0.00O0E+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 2223

Calc. No. 3.2.15.14. Rev. 0 Attachment 5 Page 15 of 16 Enclosure 4 1-131 Summary PY-CEI/NRR-2674L Page 52 of 76 Containment Environment Control Room Time (hr) 1-131 (Curies) 1-131 (Curies) 1-131 (Curies) 0.000 4.3771E-01 1.8194E+02 1.9834E-01 0.275 0.0000E+00 1.8242E+02 1.5580E-01 0.525 0.OOOOE+00 1.8245E+02 1.2486E-01 0.775 o.OOOOE+00 1.8248E+02 1.0006E-01 1.025 o.OOOOE+00 1.8250E+02 8.0194E-02 1.275 o.OOOOE+00 1.8251E+02 6.4269E-02 1.525 o.OOOOE+00 1.8253E+02 5.1507E-02

"--1.775 O.0000E+00 1.8254E+02 4.1278E-02 2.000 0.OOOOE+00 1.8254E+02 3.3822E-02 2.250 O.OOOOE+00 1.8255E+02 2.7100E-02 2.500 o.OOOOE+00 1.8256E+02 2.1715E-02 2.750 0.000OE+00 1.8256E+02 1.7399E-02 3.000 0.OOOOE+00 1.8256E+02 1.3941E-02 3.250 0.0000E+00 1.8257E+02 1.1171E-02 3.500 0.OOOOE+00 1.8257E+02 8.9506E-03 3.750 0.0000E+00 1.8257E+02 7.1718E-03 4.000 0.OOOOE+00 1.8257E+02 5.7465E-03 4.250 0.OOOOE+00 1.8257E+02 4.6045E-03 4.500 0.OOOOE+00 1.8257E+02 3.6894E-03 4.750 o.OOOOE+00 1.8258E+02 2-9562E-03 5.000 o.0000E+00 1.8258E+02 2.3687E-03 5.250 o.0000E+00 1.8258E+02 1.8979E-03 5.500. 0.0000E+00 1.8258E+02 1.5208E-03 5.750 0.OOOOE+00 1.8258E+02 1.2185E-03 6.000 0.OOOOE+00 1.8258E+02 9.7636E-04 6.250 0.OOOOE+00 1.8258E+02 7.8232E-04 6.500 0.OOOOE+00 1.8258E+02 6.2685E-04 6.750 0.0000E+00 1.8258E+02 5.0227E-04 7.000 0.OOOOE+00 1.8258E+02 4.0245E-04 7.250 o.OOOOE+00 1.8258E+02 3.2247E-04 7.500 o.OOOOE+00 1.8258E+02 2.5838E-04 7.750 0.0000E+00 1.8258E+02 2.0703E-04 8.000 0.0000E+00 1.8258E+02 1.6589E-04 8.400 o.OOOOE+00 1.8258E+02 1.1637E-04 8.700 o.OOOOE+00 1.8258E+02 8.9205E-05 9.000 o.OOOOE+00 1.8258E+02 6.8379E-05 9.300 o.OOOOE+00 "= 1.8258E+02 5.2415E-05 9.600 o.OOOOE+00 1.8258E+02 4.0178E-05 9.900 0.0000E+00 1.8258E+02 3.0797E-05 10.200 0.OOOOE+00 1.8258E+02 2.3607E-05 24.000 0.OOOOE+00 1.8258E+02 1.1520E-10 96.000 0.0000E+00 1.8258E+02 2.2331E-38 720.000 0.0000E+00 1.8258E+02 1.4904-278 Cumulative Dose Summary Exclusion Area Bounda Outer Boundary of the Control Room Time Thyroid TEDE Thyroid TEDE Thyroid TEDE (hr) (rem) (rem) (rem) (rem) (rem) (rem) 0.000 0.OOOOE+00 1.4377E+00 0.OOOOE+00 1.6048E-01 0.OOOOE+00 4.5686E-05 0.275 0.OOOOE+00 1.4415E+00 0.0000E+00 1.6091E-01 0.OOOOE+00 2.2323E-01 0.525 0.OOOOE+00 1.4417E+00 0.OOOOE+00 1.6094E-01 0.OOOOE+00 3.8370E-01 0.775 0.OOOOE+00 1.4419E+00 0.00O0E+00 1.6096E-01 0.OOOOE+00 5.1208E-01 1.025 0.OOOOE+00 1.4421E+00 0.OOOOE+00 1.6097E-01 0.OOOOE+00 6.1479E-01 1.275 0.OOOOE+00 1.4422E+00 0.OOOOE+00 1.6099E-01 0.OOOOE+00 6.9696E-01

Caic. No. 3.2.15.14. Rev. 0 Atcmn 5 Aftachment Page 16 of 16 1.525 o .OOOOE+00 Enclosure 4 1.4423E+00 0.0000E4;00 1.6100E-01 0. OOOOE+00 7.6271E-01 1.775 0.OOOOE+00 PY-CEI/NRR-2674L 1.4423E+,00 0.OOOOE+00 1.6101E-01 0.0000E+00 8 .1533E-01 2.000 0.OOOOE+00 Page 53 of 76

1. 4424E+00 0. OOOOE+00 1.6101E-0l 0.0000E+00 8.5363E-01 2.250 0. OOOOE+00 1 .4424E+00 0.OOOOE+00 I.G101E-01 0. OOOOE+0O 8.880SE-01 2.500 0.000 OE+00 1.4424E+00 0.OOOOE+00 1.GI01E-01 0.0000E+00 9.1564E-01 2.750 0. OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 0. OOOOE+00 9.3770E-01 3.000 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 0. OOOOE+00 9.5535E-01 3.250 0.OOOOE+00 1.4 424E+00 0.OOOOE+00 1.6101E-01 0.OOOOE+00 9.6947E-01 3.500 0. 00OOE+.00 1.4 424E+00 0.OOOOE+00 1. 6101E-01 0. OOOOE+00 9.8077E-01 3.750 0 .OOOOE+00 1 .4424E+00 0 .OOOOE+00 l.6101E-01 0.0000E+00 9.8982E-01 4.000 0. OOOOE+00 1.4424E+00 0.OOOOE+00 1.610 lE-Ol 0.OOOOE+00 9. 970GE-01 4.250 0. OOOOE+00 1.4424E+00 0.0000E+00 1.6101E-01 0. OOOOE+00 1. 0028E+00 4.500 0 .OOOOE+00 1.4424E+00 0.OOOOE+00 1. 6101E-01 0. OOOOE+00 1.0075E+00 4.750 0. OOOOE+00 1. 4424E+00 0.OOOOE+00 1. 6101E-0l 0.OOOOE+00 1.0112E+00 5.000 0.OOOOE+00 1. 4424E+00 0.OOOOE+00 1. 6101E-01 0.OOOOE+00 1.0142E+00 5.250 0.OOOOE+00 1.4424Ei-00 0.OOOOE+00 1. 610IE-01 0.OOOOE+00 1. 0165E+00 5.500 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101lE-01 0. OOOOE+00 1.0184E+00 5.750 0.OOOOE+00 1 .4424E+00 0.0000E+00 1.6101E-01 0.OOOOE+00 1. 0200E+00 6.000 0.OOOOE+00 1 .4424E+00 0. OOOOE+00 1.6101E-01 0.OOOOE+00 1.0212E+00 6.250 0.OOOOE+00 1 .4424E+00 0.OOOOE+00 1.6101E-01 0.OOOOE+00 1. 0222E+00 6.500 0. OOOOE+00 1.4424E+00 0.OOOOE+00 1. 6101E-0l 0.000 OE+00 1. 0229E+00 6.750 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1. 6101E-01 0.OOOOE+00 1.0236E+00 7.000 0. OOOOE+00 1.4424E+00 0.OOOOE+00 I1.6101E-01 0.OOOOE+00 1.0241E+00 7.250 0. OOOOE+00 1.4424E+00 0. OOOOE+00 l.6101E-01 0.OOOOE+00 1. 0245E+00 7.500 0.OOOOE+00 1 .4424E+00 0. OOOOE+00 1. 6101E-01 0. OOOOE+00 1.0248E+00 7.750 0.000 OE+00 1 .4424E+00 0.OOOOE+00 1. 6101E-0l 0. OOOOE+00 1.0250E+00 8.000 0.OOOOE+00 1 .4424E+00 0.OOOOE+00 1.6101E-01 0.OOOOE+00 1.0253E+00 8.400 0. OOOOE+00 1.4424E+00 0.OOOOE+00 l.6101E-0l 0. OOOOE+00 1.02S5E+00 8.700 0. OOOOE+00 1.4424E+00 0. OOOOE+00 1.6101E-01 o .OOOOE+00 1. 0256E+00 9.000 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 0.OOOOE+00 1.0257E+00 9.300 0. OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 0.OOOOE+00 1-0258E+00 9.600 0. OOOOE+00 1.4424E+00 0.OOOOE÷00 1. 6101E-0l 0.00 OOE+00 1.-0259E+00 9.900 0.OOOOE+00 1.4424E+00 0. OOOOE+00 1.6101E-01 0.OOOOE+00 1. 0259E+00 10.200 0.OOOOE+00 1 .4424E+00 0. OOOOE+00 1.6101E-01 0.OOOOE+00 1.0260E+00 24.000 0.OOOOE+00 1 .4424E+00 0.OOOOE+00 1.LIOlE-Ol 0.OOOOE+00 1-0261E+00 96.000 0.OOOOE+00 1.4424E+00 0. OOOOE+00 l.6101E-0l 0.OOOOE+0O 1. 0261E+00 720.000 0.OOOOE+00 1.4424E+00 0. OOOOE+00 1.6101E-01 0. OOOOE+00 1. 0261E+00 Worst Two-Hour Doses Note: All of the dose locations are shown below but the worst two-hour dose is only meaningful for the EAE dose location. Please disregard the two-hour worst doses for the other dose locations Exclusion Area Boundary Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.3779E-03 ).OOOOE+00 4-.7275E-03 Outer Boundary of the LI Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.5382E-04 0 000 OE+00 5.2772E-04 Control Room Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 2. 1084E-02 0).OOOOE+00 8.5359E-01

%,dCl.,. INU. .3.4. IJ. 1'., MUV. U Attacnment b Page 1 of 9 RADTRAD Version 3.02 run on 11/13/2001 at 13:54:31 0 Pathwayss 3

Pathway Is Pile information Unfiltered Release to Environment 1

2 Plant file name - pnppfha.psf 4 Inventory tile name - d:\hwagage\computer codes\radtrad\run Pathway 2:

batch\perry\pnppfha.nif Unfiltered Environment to CR Scenario file name - NEW SDF.SDF 2 Release file name = ds\hwagage\computer codes\radtrad\run 3 batch\perry\pnpptha.rft 2 Dose conversion file name - di\hwagage\computer codes\radtrad\run Pathway 3s batch\perry\pnpp_fha.dcf Control Room Exhaust 3

2 2

End of Plant Model File flff* tff t*# #9 499 #W#if#

9*#

Scenario Description Names A # # ### ## # # #

# " # 4 # # # t 0 Plant Model Filenames o 0#0 0* 000## Nl #

Source Terms

#AN 4 # # #

#A # # ## # #

1

#f# # N # # #9# #

1 1.0000E+00 d.\hwagage\computer codea\radtrad\run batch\perry\pnpp_fha.dcf ds\hwagage\computer codes\radtrad\run batch\perry\pnpptha.rft Radtrad 3.02 1/n/2000 0.00002+00 1

perry fhat sensitivity case 1 0.0000+E00 0.9985E+00 0.0015E+00 1.00008+00 Nuclide Inventory Files Overlying Pool, d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.nif 0 Plant Power Levels 0.0000E+00 3.8332E+03 0 Compartments: 0 3 0 Compartment 1: 0 Containment Compartmentls:

3 3 1.O0000E00 Compartment 1I 0 0 0 1 0 0 0 0 0 0 Compartment 2: 0 Cnvironment 0 2 0 O.O000E+00 0 Compartment 2:

0 0 0 1 0 0 0 0 -0 "o r Compartment 3: 0 Control Room 0 1 0 3.6707E+05 0 0 0 .

4*2 -t 0 Compartment 3t 0)

-i 04 0 0

QflJ OTAVU03 ;o PU3 0 z0 0 0 0 0 V) W LO 0

.23)a o 0) 00+3000010 00+3000010 00+300001*0 (Ot3000VIS r0+3000Z*L a> ca 00+3000010 00+300001*0 00+30000 *0 EO+30001'S 00+300001Z w CL C 00+3000010 00+3000O00 00+300001*0 00+3000010 00+31000.0 t~wv"ITS qndino 00+30000 *0 00+~3000 010 00+300001*0 E0+30009*9 00+3000010 00+3000010 z0+30Q0VLt 10-30000*1 10#3000VIC p 10-9000011 00+30000*9 to03000Stz 00+30000'0 0 0

ls~lawv~vd UOTVlUTnwS 0 00*3000010 00+300001Z 'tAA4V to-3ooSIC 00+3000010 0 0EAM4g 0

SUOp~eOql OwftOA QJTI0Q;3 0 00+3000010 z0+3000VL 0 10-300001t 10+3000916 0 10-3000019 I0*3000tlz 00+3000010 00+300001*0 00+30000 *0 c0+3000WS ZO*3000Z*L 00+3000010 W003000 01*0 00+30000*0 0O+3 00015'* 00+~3000 01Z 00+30000*0 00+3000010 00*3000010 00+3000010 00+3000010 Z0+3000Z-L 00+3000010 00+3000010 00+300000O E0,3000,9* 00+3000010 t0-30005't 00+20000'0 v, T

0 0 0 0

wood jolluoD 0 IC Uopleoo'z 0 0 lz Aepqivd tO-3000ct0 10+3000P'r 0 t0-30009*1 00+30000*9 00+3000010 Z0+3000C'L WO-3000SIC 00+3000010 01+30000*1 00+3000010 E z I

00+3000010 00*300001Z 0

0 0

0 Z4'1 04 ;0 A4zuPUnOg J~nO 0 0

0 0 VO03000t'Z T0+3000P*Z 0 t0-30008*T 00+3000016 0 0

to-3000CtE 00+3000010 I 0 0

0 0

/iepunou eazv uoIsflTox3 0 qI UoT;v~oi 0 E 0 isUQ$luzoq' as0 0 *AEýj'VcjVZC 'O 6 JO Z 6e 9 luaewtpe;i'

. - -. .- I.. , I%-..V /-%Lkd3%tIIIjjUIL U v-age .3 or 9 RADTRAD Version 3.02 run on 11/13/2001 at 131S4:31 tfffl#ff##fff*ffflf~#*f#fftU*t##i#t#####t(#ff##*###n###*#ff#f##f#f#t(*##ff tf#eff#nflflnnfl~ffftfftff~ff###Nfff#N#ff*#f#Mfff*#f#i#tfl#iflnIH#IhInnn*#

Plant Description Number of Nuclides - 23 Inventory Power - 3.8332E.03 MWth Plant Power Level - 3.8332E+03 MWth Number of compartments

3 Compartment information Compartment number 1 (Source term fraction - 1.0000E+00 Names Containment Compartment volume - l.OOOOE+00 (Cubic feet)

Pathways into and out of compartment I Pathway to compartment number 21 Unfiltered Release to Environment Compartment number 2 Name, Environment Pathways into and out of compartment 2 Pathway to compartment number 31 Unfiltered Environment to CR Pathway from compartment number It Unfiltered Release to Environment Pathway from compartment number 3: Control Room Exhaust Compartment number 3 Names Control Room Compartment volume - 3.6707E+05 (Cubic feet)

Pathways into and out of compartment 3 Pathway to compartment number 2: Control Room Exhaust Pathway from compartment number 2: Unfiltered Environment to CR Total number of pathways - 3 0

o r-'

Calc. No. 3.2,15.14, Rev. 0 Attachment 6 Page 4 of 9 (cfm) Aerosol Elemental Organic 0.0000.E00 6.60009+03 0.00000E00 0.0000.E00 0.0100.E+00 RADTRAD Version 3.02 run on 11/13/2001 at 13s54:31 1.0000E-04 0.0000,E00 0.OOOOE.00 0.00005.00 0.0COOEO0 2.0000E÷00 5.4000E503 0.0000Z#00 0.0000.E00 0.010O0E000 7.2000E502 5.4000E+03 0.OOOOE.00 0.0000E+00 0.0000E+00 Scenario Description LOCATION DATA Location Exclusion Area Boundary is in compartment 2 Radioactive Decay is enabled Location X/O Data Calculation of Daughters is enabled Time (hr) X/Q (s

m^-3)

RELEASE NAME - Perry FHA 0.0000.E00 4.3000E-04 Release Fractions and Timings 2.0000E.00 0.0000E00 GAP EARLY IN-VESSEL 0.0001 hrs 0.0000 hra Location Breathing Rate Data NOBLES 1.0000E+00 0.O000OE00 Time (hr) Breathing Rate (m^3

sec'-l)

IODINE 1.0000E+00 0.00002+00 0.00000E+0 3.5000E-04 CESIUM 1.0000E+00 0.00005,00 8.0000E+00 1.8000E-04 TELLURIUM O.0000E+00 0.O000E+00 2.40003*01 2.3000E-04 STRONTIUM 0.0000E*00 0.0000E+00 Location Outer Boundary of the LPZ is in compartment 2 BARIUM 0.0000S÷00 0.0000E+00 RUTHENIUM O.OOOOE+00 O.OOOOE+00 Location X/Q Data CERIUM 0.00005400 0.0000.E00 Time (hr) X/Q (a

m'-3)

LANTHANUM O.OOOOE+00 O.OOOOE+00 0.0000R+00 4.000E-0S 2.0000E+00 0.OOOOE+00 Iodine fractions Aerosol . 0.0000.E00 Location Breathing Rate Data Elemental . 9.98SO0-01 Time (hr) Breathing Rate Wm'3I seo'.l)

Organic - 1.5000E-03 0.0000.E00 3.5000C-04 0.0000E+00 1.80005-04 COMPARTMENT DATA 2.4000E501 2.3000E-04 Location Control Room is in compartment 3 Compartment number It Containment Location X/Q Data Compartment number 2: Environment Time (hr) X/O (a

m^-3) 0.00005400 3.5000E-04 Compartment number 3s Control Room 2.0000E+00 0.00005.00 PATHWAY DATA Location Breathing Rate Data Time (hr) Breathing Rate (m*3

secA-l)

Pathway number It Unfiltered Release to Environment 0.0000E400 3.S000E-04 7.2000E502 0.0000,E00 Convection Data Time (hr) Flow Rate (% / day) Location Occupancy Factor Data 0.OOOOE+00 1.0000E510 Time (hr) Occupancy Factor 7.2000E+02 O.00005.00 O.OOOOE+00 1.0000.E00 2.4000E501 6.00005-01 Pathway number 2t Unfiltered Environment to CR 9.6000Es0l 4.0000E-01 7.2000E+02 0.00005.00 Pathway Filter: Removal Data USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS Time (hr) Flow Rate Filter Efficiencies (t) Time Time step (cfm) Aerosol Elemental Organic 0.0000E+00 2.5000E-02 0.0000E500 6.6000E*03 0.0000E#00 0.0000,+00 0.0000E+00 8.0000E+00 1.00005-01 1.OOOOE-04 0.0000E+00 0.0000E+00 0.0000.E00 0.0000E+00 2.4000E+01 4.0000E-01 2.0000.E00 5.4000E+03 0.0000B÷00 0.OOOOE00 0.0000.E00 7.20005E02 0.OOOOE.00 a) -< :

7.2000E+02 5.4000E+03 0.0000E+00 O.0000S÷00 O.0000+E00 M0 (0 O*

Pathway number 3: Control Room Exhaust - C oz Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (t) -41

-. I~.1 J.~.. ..*. I t I %tav. V ~~tII~~I P'%1tdbllllt11e1t 0 Page 5 of 9 Xe-135 1.3783E+01 5.3973E-09 2.4076E,16 5.09985011 Xe-13Sm 6.S313E-01 7.1700E-12 3.19849.13 2.4166E+10 RADTRAD Version 3.02 run on 11/13/2001 at 13s54,31 Cs-135 4.7418E-14 4.1157E-14 1.8359E011 1.7545-03 Control Room Transport Group Inventory:

Time (h) - 0.0001 Atmosphere Sump Overlying Pool Noble gases (atoms) 1.5969E+19 0.0000E+00 0.0000E+00 00110 II /I

Elemental I (atoms) 1.8787E017 0.0000£+00 0.0000S+00 If If IfS If Organic I (atoms) 2.8223E.14 0.00000.00 0.O000E+00

I Oil # I(ff9 999999 99 f# 99 999 Aerosols (kg) 0.0000E.00 0.00000.00 0.O0000+00

I 990 # # ## # #

I 9991 g Deposition Recirculating 9999 # # # #

Time (h) - 0.0001 Surfaces Filter 11911fl1 051199 Noble gases (atoms) 0.0000E+00 0.00000.00 Elemental I (atoms) 0.0000E+00 0.0000E+00 Organic I (atoms) 0.O000E+00 0.0000E+00 19111*19#IlS II 99 ~n~nn 0999999 #5 09999999999 99999999 99# 99 # 0 IlO##

f#ff099 Aerosols (kg) 0.00000+00 0.0000E+00 Dose, Detailed model and Detailed Inventory Output Unfiltered Environment to CR Transport Group Inventoryi Exclusion Area Boundary Doses, Pathway Time (h) - 0.0001 Filter Time (h) - 0.0001 Whole Body Thyroid TEDS' Noble gases (atoms) 0.0000,E00 Delta dose (rem) 4.2456E-01 0.0000E.00 1.4377E+00 Elemental 1'(atoms) O.OOOOE+O0 Accumulated dose (rem) 4.2456E-01 0.OOOOE.00 1.4377E+00 organic I (atoms) 0.00000.00 Aerosols (kg) 0.0000oo00 Outer Boundary of the LPZ Doses, Control Room Exhaust Transport Group Inventoryt Time (h)

0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.7393E-02 O.OOOOE+00 1.6048E-01 Pathway Accumulated dose (rem) 4.7393E-02 0.0000E000 1.6048E-01 Time (h) - 0.0001 Filter Noble gases (atoms) 0.00000E00 Control Room Doses, Elemental I (atoms) 0.0000E+00 organic I (atoms) 0.0000r+00 Time (h) - 0.0001 Whole Body Thyroid TEDE Aerosols (kg) 0.000OE00 Delta dose (rem) 1.2078E-06 0.0000E+00 4.5986E-05 Accumulated dose (rem) 1.20780-06 0.0000E+00 4.5686E-05 Exclusion Area Boundary Dosess Control Room Compartment Nuclide Inventory: Time (h) - 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.02142-03 0.0000E+00 3.4587E-03 Time (h) - 0.0001 Ci kg Atoms Bq Accumulated dose (rem) 4.2559E-01 O.0000E*00 Br-82 1.4411E+00 6.0733E-04 S.6097E-13 4.11980+12 2.2471E+07 Br-83 1.61560-0$ 1.0227E-IS 7.4200E+09 5.9776E+05 Outer Boundary of the LPZ Doses, Kr-83m 1.2500E-02 6.0583E-13 4.3957E+12 4.6248E008 Kr-85 8.1467E-01 2.07S10-06 1.4702E÷19 3.0143E+10 Time (h) - 2.0000 Whole Body Thyroid Kr-8Sm TEDE 1.6234E-01 1.9727E-11 1.3976E+14 6.0067E+09 Delta dose (rem) 1.1402E-04 0.0000E+00 3.8609E-04 Kr-87 2.6301E-OS 9.2854E-16 6.4273E+09 9.73150E05 Accumulated dose (rem) 4.7507E-02 0.00000.00 1.6087E-01 Kr-88 5.0143E-02 3.9989E-12 2.7366E+13 1.8553E009 Rb-87 4.4281E-24 S.0613E-20 3.5034E+05 1.6384E-13 Control Room Doses, Rb-B8 1.17160-0S 9.7601E-17 6.6792E+08 4.3349E+05 1-129' 6.8183E-09 3.8601E-08 1.8020E+17 2.5228E+02 Time Mh) - 2.0000 Whole Body Thyroid TEDE 1-130 1.4897E-03 7.6381R.13 3.,383E+12 5.5118E007 Delta dose (rem) 4.3741E-02 0.0000E+00 1.8080E+00 1-131 1.9834E-01 1,5999E.09 7.3546E015 7.3387E009 Accumulated dose (rem) 4.3742E-02 0.0000E000 1.8080E000 1-132 1.5753E-01 1.5262E-11 6.9627E+13 S.8288E+09 1-133 1.2391E-01 1.0938E-10 4.9529E+14 4.5847E009 Control Room Compartment Nuclide Inventoryt "-0 "0 m 1-134 6.5286E-09 2.4473E-19 1.0999E,06 2.41560+02 a) -< :,

1-135 2.0391E-02 5.8063E-12 2.5901E+13 7.5446E+08 Time (h) - 2.0000 Ci kg Atomes to - 0

. sq CD06 Xe-129m 2.2S970-04 1.7859E-12 8.3372E+12 8.3607E+06 Br-82 S.8394E-04 5.3937E-13 3.9611E*12 2.1606E+07 Xe-131m 2.9842E-01 3.5627E-09 1.6378E*16 1.1041E+10 Br-83 Mn M Cn Xe.133 5.0066E+01 2.6747E-07 1.2111E+18 1.S52SE+12 9.045SE-06 5.7258E-16 4.15440.09 3.3468E*05 oz*B Kr*83m 5.8665E-03 2.8434E-13 2.0631E012 2 . 1706E008 Xe-133m 1.5255E+00 3.3997E-09 1.5394E+16 5.6442E+10 Kr-8S 8.1466E-01 2.07509-06 1.4701E+19 3.0143E010

-D 0)I-P

,.,aic. No. J.Z.10.14, Rev. 0 Attachment 6 Page 6 of 9 Kr-85m 1.1914C-01 1.4477E-11 1.0257E+14 4.4081E+09 Delta dose (rem) 0.0000Eo0o 0.00000+00 0.0000E,00 Kr-87 8.8422E-06 3.1216E-16 2.1608E+09 3.2716E÷0S Accumulated dose (rem) 4.7507E-02 Kr-88 0.0000.E00 1.6087E-01 3.0777E-02 2.454SE-12 1.6797E÷13 1.1388E+09 Rb-87 5.3930E-20 6.1643E-16 4.2669E*09 1.9954E-09 Control Room Dosest Rb-80 3.4043E-02 2.9027E-13 1.9864E+12 1.2892E+09 1-129 6.8183E-09 3 .801E-08 1.8020E+17 2.5228E+02 Time (h) - 8.0000 Whole Body Thyroid TEDE 1-130 1.3316E-03 6.8277E-13 3.1629E+12 4.9271E007 Delta dose (rem) 2.1440E-02 0.0000E+00 9.9924C-01 1-131 1.96929-01 1.5084E-09 7.30200÷16 7.2861E+09 Accumulated dose (rem) 6.3182E-02 1-132 0.0000E÷00 2.8073E+00 8.6223E-02 8.3532E-12 3.8109E+13 3.1902E.09 1-133 1.1S92E-01 1.0233E-10 4.6335E÷14 4.2892E009 Control Room Compartment Nuclide Inventory, 1-134 1.3431E-09 S.03460-20 2.2626E+05 4.96932+01 1-135 1.6533E-02 4.7078E-12 2.1001E+13 6.11730.08 Time (h) 8.0000 Ci kg Atoms Bq Xe-129m 2.2434E-04 1.7731E-12 8.2773E+12 8.30060÷06 Br-*2 2.6012E-06 2.4027E-15 9.6246E+04 Xe-131m 2.9698E-01 3.$4S6E-09 1.6299E+16 1 0988E+10 Br-83 1.7646E+10 7.9S59E-09 5.0361E-19 3.6540E÷06 2.9437E+02 Xe-133 4.9536E+01 2.6464E-07 1.19832+18 1.8328E012 Kr- 83m Xe-133m 3.04360-06 1.4752E-16 1.0703E+09 1.1261E+05 1.48S8+.00 3.3113E-09 1.4993E+16 5.497E5+10 Kr-S5 4.0826E-03 Xe-13S 1.0399-e08 7.3674E+16 1.5106E+08 1.18M2+.01 4.6412E-09 2.0704E+16 4.3834E0+1 Kr-SSm 2.3S98E-04 Xe-135m 2.8674E-14 2.0315E+11 8.7311E÷06 S.SS97E-03 6.1033E-14 2.7226E+11 2.0372E+08 Kr-87 1.6835E-09 S.9434E-20 Cs-135 8. 8064E-10 4.1140E+05 6.2290E+01 7.643SE-10 3.4096E015 3.2584E+01 Kr-88 3.5664E-05 2.84420-15 1.94640+10 1.3196E+06 Rb-B? 4.0195E-22 4.5943E-18 3.1802E+07 1.4872E-11 Control Room Transport Group Inventoryi Rb-8e 4.1000-.05 3.4156E-16 2.3374E+09 1.5170E+06 r-129 3.4171E-11 1.934SE-10 9.0311C÷14 1.2643E+00 Overlying 1-130 4.7669E-06 2.4441E-15 Time (h) . 2.0000 1.1322E+10 1.76370+05 Atmosphere Sump Poor 1-131 9.6586E-04 7.7908E-12 Noble gases (atoms) 1.$969E+19 0. 0000E00 3.5815E+13 3.S737E÷07 0.0000÷+00 1-132 7. 0844E-05 6.86330-15 Elemental I' (atoms) 1.8787E+17 3.1312E010 2.6212E+06 0.00000+00 0.00000E00 1-133 4.75680-04 4.19910-13 Organic I (atoms) 1.9013E+12 1.7600E+07 2.82230+14 0.000E÷00 0.0000+O00 r-135 4.4166E-05 Aerosols (kg) 1.2576E-14 S.6101E+10 1.6341E006

0. 0000E00 0.O000E+00 O.O000E+00 Xe-129m 1.1002E-06 8.6956E-15 4.0594E+10 4.0708E#04 Xe-131m 1.46700-03 1.7514E-11 8.0514E013 S.4279E*07 Deposition Recirculating Xe 133 2.40448-01 Time (h) - 2.0000 1.2845E-09 5.8162E0+S 8.89610+09 surfaces Filter Xe-133m 6.8805E-03 Noble gases (atoms) I.S3340-11 6.9432E+13 2.54580.08 0.0000E+00 0.00000.00 Xe-135 3.7613E-02 Elemental I (atoms) 1.4729E-11 6.5703E+13 1.3917E+09 0.00000400 0.0000+E00 Xe-135m 7.3176E-06 Organic I (atoms) 8.0332E-17 3.5835E+08 2.707SE0+S 0.0000E+00 0.O0000S+00 Cs-13S 1.4255S-11 1.23730-11 Aerosols (kg) 0. 0000E00 0.O000E+00 5.5194E*13 5.2743E-01 Unfiltered Environment to CR Transport Group Inventory, Control Room Transport Group Inventory, Pathway Overlying Time (h) - 8.0000 Atmosphere sump Pool Time (h) - 2.0000 Filter Noble gases (atoms) 8.0029E016 0.0000E+00 O.O0000+00 Noble gases (atoms) 0.0000E+00 Elemental I (atoms)

Elemental I (atoms) 9.4150£+14 0. 0000E00 0.0000E+00

0. 0000+00 Organic I (atoms) 1.4145E+12 Organic I (atoms) 0.00008+00 O.0000E+00 0.0000+E00 Aerosols (kg) 0.00000E+00 0, 0000E.00 Aerosols (kg) O.0000E+00 0. 0000E00 Control Room Exhaust Transport Group Inventory, Deposition Recirculating Time (h) - 8.0000 Surfaces Filter Noble gases (atoms) O.OOOOE+00 0.00000+00 Pathway Elemental I (atoms) 0.0000E+00 0.0000E+00 Time (h) - 2.0000 Filter Organic I (atoms)

Noble gases (atoms) O.00000E00 O.O000E+00 0.0000+E00 Aerosols (kg) O.0000E+00 0.0000E400 Elemental I (atoms) 0.0000+E00 Organic I (atoms) 0.0000E*00 Unfiltered Environment to CR Transport Group Inventory, Aerosols (kg) 0.0000E+00 Pathway Exclusion Area Boundary Dosest Time (h) - 8.0000 Filter Noble gases (atoms) 0.0000E+00 l u10Tm Time (h) - 8.0000 WhOle Body Thyroid TEDE Elemental I (atoms) 0.O0000E+00 Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 a) -< :

Accumulated dose (rem) 4.2559E-01 Organic I (atoms) 0. 0000E00 EO 1 0 0.O000E+00 1.4411E000 Aerosols (kg) 0. 0000+00 Outer Boundary of the LPZ Doses, Control Room Exhaust Transport Group Inventory, Time (h) - 8.0000 Whole Body Thyroid TEDE 0 z Pathway ;0)C

",4 I-4

CaIc. No. 3.2.15,14, Rev. 0 Attachment 6 Page 7 of 9 Time (h) - 8.0000 Filter Noble gases (atoms) O.0000E+00 Elemental I (atoms) Pathway O.OOOOE+O0 Time (h) - 24.0000 Filter Organic I (atoms) 0.OOOOE+00 Noble gases (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.O00OOE00 Exclusion Area Boundary Doses; Aerosols (kg) 0.O000OE00 Tlime. (1h) - '4. 0000 Whole Body 'Thylohul TWDE Exclusion Area Boundary Dosess Delta dose (rem) 0.OOOOE00 0.OOOOE+00 0.0000E+00 Accumulated dose (rem) 4.2S59E-01 0.OOOE+00 1.4411E+00 Time (h) - 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.OOOOE+00 0.0000E+00 0.0000E,00 Outer Boundary of the LPZ Doses, Accumulated dose (rem) 4.2559E.01 0.0000E+00 1.4411E+00 Time (h) - 24.0000 Whole Body Thyroid TEDE Delta dose (rem) Outer Boundary of the LPZ Doses, O.OOOOE+00 0.O000E+00 0.OOOOE+00 Accumulated dose (rem) 4.7507E-02 0.0000E+00 1.6087E-01 Time (h) - 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E00 0.O000E+00 0.0000E+00 Control Room Doses; Accumulated dose (rem) 4.7507E-02 O.0000E+00 1.6087E-01 Time (h) a 24.0000 Whole Body Thyroid TEDE Control Room Doses; Delta dose (rem) 7.9420E-05 0.OOOOE+00 4.8094E-03 Accumulated dose (rem) 6.S262E.02 0.0000E+00 2.8121E+00 I Time (h) - 96.0000 Whole Body Thyroid TEDE Control Room Compartment Nuclide InventoryM Delta dose (rem) 2.0998E-11 0.0000+E00 1.9496E.09 Accumulated dose (rem) 6.5262E.02 0.OOOOE+00 2.8121E÷00 Time (h) - 24 0000 Ci kg Atoms Bq Control Room Compartment Nuclide Inventory; Kr-85 3.0026E-09 7.6480E-l5 S.4185E+10 . ll10E+02 1-129 2.513SE-17 1.4230E-16 6.6428E+08 9.2998E-07 Time (h) . 96.0000 Ci kg Atoms 1-131 6.7076E-10 5.4104E-18 2.4872E+07 2.4818E+01 1-133 2.0529E-10 1.8122E-19 8.2095E+05 7.5957E#00 Control Room Transport Group Inventory; Xe-131m 1.0383E-09 1.2395E-17 5.6982E+07 3.8415E.01 Xe-133 1.6235E-07 8.6731E-16 3 .9271E+09 6.00682+03 Xe-133m 4.0989E-09 9.1350E-18 Overlying 4 .1363E+07 l.5166E+02 Time (h)

96.0000 Atmosphere Sump Pool Xe-13S 8.1770E-09 3.2020E-18 1.4284E+07 3.0255E+02 Noble gases (atoms) 1.47788-17 0.0000E+00 0.0000E+00 Cs-135 1.9287E-17 1.6740E.17 7.4676E+07 7.1363E-07 Elemental t (atoms) 1.7387E-19 0.0000E,00 0.0000E+00 Organic I (atoms) 2.9119E-22 0.0000Z+00 0.0oo0E,00 Control Room Transport Group Inventory; Aerosols (kg) 0.OOOOE+00 0.0000E400 0.000000 Overlying Deposition Recirculating Time (h) - 24.0000 Atmosphere Sump Pool Noble gases (atoms) Time (h)

96.0000 Surfaces Filter 5.8866E+10 0.OOOOE+00 O.O000E+00 Noble gases (atoms) O.OOOO+00 0.00000+00 Elemental I (atoms) 6.9256E+08 0.0000E+00 0.00009÷00 Elemental I (atoms) 0.0000E,00 0.0000E+00 Organic I (atoms) 1.0404E+06 0.0000S+00 0.OOOOE+00 Aerosols (kg) Organic I (atoms) 0.0000E+00 0.0000E+00 0.0000E+00 0.00OOE400 0.0000E+00 Aerosols (kg) 0.0000+E00 O.0000E+00 Deposition Recirculating Time (h) - 24.0000 Unfiltered Environment to CR Transport croup Inventoryl Surfaces Filter Noble gases (atoms) O.O000E+00 0.0000E+00 Elemental ? (atoms) 0.00000+00 0.00000+00 Pathway Time (h) - 96.0000 Filter Organic I (atoms) 0.0000.+00 O.OOOOE+00 Noble gase3 (atoms) 0.00000E00 Aerosols (kg) O.0000E+00 O.00000E+0 Elemental I (atoms) 0.000E0000 Organic I (atoms) 0.0000o+00 Unfiltered Environment to CR Transport Group Inventoryt Aerosols (kg) 0.0000Ez00 V m Pathway Control Room Exhaust Transport Group Inventory; Time (h)

24.0000 Filter -o CoI0 Noble gases (atoms) 0.0000e+00 Elemental I (atoms) Pathway O.OOOOE+00 Time (h) - 96.0000 I Organic I (atoms) Filter O.O000EO0o Noble gases (atoms) 0.0000o÷00 Aerosols (kg) 0.O0000000 Elemental I (atoms) O.0000E.00 Control Room Exhaust Transport Group Inventory; Organic I (atoms) 0.00000÷00 M Aerosols (kg) 0.0000E+00 0r

-4

'..- . uA11Z1U u V Page 8dot 9 1 Exclusion Area Boundary Doses, Containment Environment Control Room Time (hr) 1-131 (Curies) 1-131 (Curies) 1-131 (Curies)

Time (h) - 720.0000 0.000 4.37712-01 1.8194E#02 1.98342-01 Whole Body Thyroid TEDE 0.275 Delta dose (rem) 0.0000E+00 1.8238E*02 1.9815E-01 O.O0002+00 O.O000E+O0 O.O0000+00 0.525 Accumulated dose (rem) 4.2559E-01 0.O000+00 1.8238E+02 1.9797E-01 0.00002+00 1.4411E200 0.775 O.0000.00 1.8238E402 1.9779E-01 1.025 0.O000E00 1.82382*02 1.97612.01 Outer Boundary of the LPZ Dosest 1.275 0.0000B00 1.8238E202 1.9744E-01 1.525 O. 000 000 1.8238E+02 1.9726E-01 Time (h)

720 0000 Whole Body Thyroid TEDE 1.775 0.0000*+00 1.8238E+02 Delta dose (rem) O.0000E+00 0.O000E+O0 0.0000£+00 1.9708E-01 Accumulated dose (rem) 4.7507E-02 2.000 0.*0000200 1.8238E.02 1.9692E.01 0.00002+00 1.6087E-01 2.250 0.*0000+00 1.8242E402 1.5779E.01 2.500 0.00002+00 1.8245E+02 1.2643E-01 Control Room Doses: 2.750 0.*0000+00 1.82482E02 1.0130E-01 Time (h)

720.0000 3.000 0.00002+00. 1.82SOE+02 8.1171E-02 Whole Body Thyroid TEDE 3.250 Delta dose (rem) 1.75652E39 0.00002+00 0.O000O+0 1.8251E*02 2.3980E-37 3.500 Accumulated dose (rem) 6.5262E-02 0.0000E+00 0.00002 00 1.8252E+02 6.5039E-02 5.2114E-02 2.8121E+00 3.750 0.00002+00 1.8253E802 4.1757E-02 4.000 O.O0000+00 1.82542+02 Control Room Compartment Nuclide Inventory: 3.34582-02 4.250 0. 0000+00 1.8255E+02 2.6809E-02 4.500 0.0000+E00 1.82552,02 2.14812-02 Time (h) - 720.0000 Ci kg Atoms 4.750 0.00002+00 1.8256E+02 1.72122-02 5.000 0. 00002+00 1,8256E202 1.3791E202 Control Room Transport Group Inventory, 5.250 0.00002+00 1.82572+02 1.1050.-02 5.500 0.00002+00 1.8257E+02 8.85442-03 Overlying 5.750 0 . 00002+O0 Time (h) - 720.0000 1.8257E+02 7.0947E-03 Atmosphere Sump Pool 6.000 0.O00002.00 Noble gases (atoms) 1.8257E+02 5.68472-03 9.2790-257 0.00'002+00 0.000E+00 6,250 O.O0000+O0 Elemental I (atoms) 1.8257E202 4.5550E-03 1.0917-258 0.00 00E+00 0.00002+00 6.500 O.O000E+00 Organic I (atoms) 1.6400-261 1.8257E.02 3.6497E-03 0.00 00E+00 0.0000E+00 6.750 0. 0000+00 Aerosols (kg) 0. 0000E00 0.00 000+00 0.00002+00 1.8257E+02 2.9244E-03 7.000 0.00009+00 1.8257E202 2.3432E-03 7.250 0.00002E00 1.82572+02 1.8775E-03 Deposition Recirculating 7.500 0. 0000E00 Time (h)

720.0000 Surfaces 1.8257E+02 1.50442-03 Filter 7.750 0.00002+00 Noble gases (atoms) 1.8257E+02 1.20542-03 0.00002+00 0.0000+E00 8.000 Elemental I (atoms) 0.0000E+00 1.8258E+02 9.6586E-04 0.0000EO0 0.00002+00 8.400 O.0000+EO0 Organic I (atoms) 1.82S8E+02 6.7757E-04 0.00002+00 0.00002+00 8.700 0. 0000E00 Aerosols (kg) 1.82582+02 5.1938E-04 0.00002+00 0.0000E+00 9.000 0.00002400 1*8258E+02 3.9812E-04 Unfiltered Environment to CR Transport Group Inventory, 9.300 0.0000E+00 1.8258E#02 3.05182-04 9.600 0.0000E+00 1.82582+02 2.3393E-04 9.900 0. 0000E00 1.8258E+02 1.7931E-04 Pathway 10.200 0. 0000E00 Time (h) - 720.0000 1.825$E+02 1.37452-04 Filter 24.000 0.00002+00 Noble gases (atoms) 1.82589E02 6.70762-10 0.00002+00 9G.000 0. 0000+00 Elemental I (atoms) 0.00002.00 1.8258E+02 1.3002E-37 720.000 0.00000E200 1.8258E+02 8.6776-278 Organic I (atoms) 0. 0000+00 Aerosols (kg) 0. 0000+00 Control Room Exhaust Transport Group Inventoryt Cumulative Dose Summary Pathway Exclusion Area Bounda Outer Boundary of the Time (h) - 720.0000 Filter Control Room Time Thyroid TEDS Thyroid TEDE Thyroid TEDE Noble gases (atoms) O.0000E00 (hr) (rem) (rem) (rem) (rem) (rem) (rem)

Elemental I (atoms) 0.0000+E00 0.000 0.0000+E00 1.4377E+00 0.0000E+00 1.60482-01 0.00002+00 4.5686E-05 Organic I (atoms) 0. 00 0000 0.275 0.00000E+O 1.4411E+00 0.00002+00 1.6087E-01 0.00002+00 2.5085E-01 Aerosols (kg) 0.0000E+00 0.525 0.0000+E00 1.4411E+00 0.00002+00 1.60872.01 0.00002.00 0.775 0.00002+00 4.7025E-01 -a-urn 1.4411E+00 0.00002E00 1.6087E-01.0.00002+00 7.0504E-01 2223 a) t< 0 1.025 O.0000+E00 1.4411E+00 0.00002.00 1.6087E-01'0.0000E+00 9.31242-01 1.275 0.00002+00 CD 6 1.44112+00 0.0000+E00 1.6087E-01 0.00002+00 1.1569E+00 1.525 0.0000,E00 1.4411E+00 0.00002+00 1.6087E-01 O.0000E+00 1.3819E200 oLo 1-131 Summary 1.775 0.00002+00 1.4411E+00 0.00002+00 1.6087E-01 0.0000E+00 1.60642.00 2.000 0.0000E+00 1.4411E+00 0.0000+E00 1.6087E-01 0.00002+00 1.8080+E00 SC) 2.250 0.00002,00 1.4411E+00 0.0000E+00 1.6087E-01 0.00002.00 2.0086E,00 0z-N I-NI"

NI,.. A 4 4A 11... fl A .....

,101G. NO. J.2. Ij. 14, Kev. u Attachment 6 Page 9 of 9 2.500 0.0000+00 1.4411E+00 0.0000OO00 1.6087Eo01 0.0000E+00 2.1691r.00 2.750 O.0000.E00 1.4411E+00 0.0000+E00 1.6087E-01 O.0000E.00 2.2975E+00 3.000 0.0000E+00 1.44110*00 0.0000E000 1.6087E-01 0.0000E*00 2.4003E+00 3.250 0.0000S+00 1.4411E000 0.0000.E00 1.6087E-01 .O.000E+00 2.48250E00 3.500 0.0000E+00 1.4411E000 0.0000E*00 1.6087E-01 0.0000.E00 2.5483E+00 3.750 0.0000.00 1.4411C.00 0.0000E+00 1.6087C.01 0.0000E.00 2.6010r.00 4.000 O.OoO0oc00 1.4411E+00 0.0000i.00 1.6087E-01 0.0000.E00 2.6431E*00 4.250 0.OOOoc.00 1.4411E÷00 0.0000F.00 1.6087E-01 0.00000,00 2.6768E+00 4.500 0.0000.E00 1.4411E+00 0.0000.*00 1.6087E-01 0.0000E+00 2.7038E.00 4.750 0.00000*00 1.4411E+00 O.0000.E00 1.6087E-01 0.0000.E00 2.72540E00 5.000 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.O0000E00 2.7427E000 5.250 0.0000.E00 1.4411E÷00 0.0000R.00 1.6087E-01 0.0000.E00 2.7566e.00 5.500 0.0000E+00 1.4411E+00 0.0000E.00 1.6087E-01 0.O0000E00 2.76796E+00 5.750 0.00000.00 1.4411E000 O.O000E+00 1.6087E-01 0.00000E00 2.7765E800 6.000 0.0000S.00 1.4411E000 0.00000.00 1.6087E-01 0.00000+00 2.7836E+00 6.2S0 0.00000E00 1.4411E+00 0.00000,00 1.6087E-01 0.0000÷+00 2.7893E000 6.500 0.0000E+00 1.4411E800 0.00000E00 1.6087E-01 0.OoooEo00 2.7938E800 6.750 0.00000E00 1.4411E+00 0.O000E+00 1.6087E-01 0.00000E00 2.7975E+00 7.000 0.00000E00 1.4411E000 0.0000+E00 1.6087E-01 0.0000C.00 2.8004E+00 7.250 0.0000E+00 1.4411E+00 0.0000E400 1.6087E-01 0.0000E+00 2.8027E000 7.500 0.0000E+00 1.4411E*00 O.0000E+00 1.6087E-01 0.0000E+00 2.8046E000 7.750 0.00000.00 1.4411E.00 0.00008E00 1.6087E-01 0.0000E+00 2.8061E+00 8.000 0.0000+E00 1.4411E*00 0.00000E00 1.6087E-01 0.0000,E00 2.8073E+00 8.400 0.00000+00 1.4411C+00 0.0000.E00 1.60870-01 0.0000E+00 2.8087E000 8.700 0.00000.00 1.4411E+00 0.0000.E00 1.6087E-01 0.O0000E00 2.8095E+00 9.000 0.00000E00 1.44110+00 0.0000E+00 1.6087E-01 0.0000.+00 2.8101E+00 9.300 0.0000E+00 1.4411E800 0.0000+E00 1.6087E-01 0.00008.00 2.8106E+00 9.600 0.0000E+00 1.4411E#00 0,0000E000 1.6087E-01 0.0000E800 2.8109E+00 9.900 0.0000E+00 1.4411E+00 O.0000.E00 1.6087E-01 0.0000E+00 2.81129+00 10.200 0.0000E+00 1.4411E000 O.O0000E00 1.6087E-01 0.00000E*0 2.8114E+00 24.000 0.0000E+00 1.4411E+00 0.0000+E00 1.6087E-01 0.0000E*00 2.8121E000 96.000 0.0000.E00 1.4411E+00 0.0000+E00 1.6087E-01 0.00000+00 2.81210.00 720.000 0.00000E00 1.4411E÷00 0.00000400 1.6087E-01 O.00000E00 2.8121E+00 Worst Two-Hour Doses Notet All of the dose locations are shown below but the worst two-hour dose is only meaningful for the EAB dose location. Please disregard the two-hour worst doses for the other dose locations Exclusion Area Boundary Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.0214E-03 0.00000E00 3.4587E-03 Outer Boundary of the LPZ Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.1402E-04 0.0000E800 3.8609E.04 Control Room Time Whole Body Thyroid TEDE e_

(hr) (rem) (rem) (rem) ('a 1 0.0 4.3741E-02 0.0000+E00 1.8080E+00 CD 0o OM"'

MC

",4 r-

"C3 IL'. 44v. 1'ý. 14, 1 NUV. U Auacnment f Page 1 of 9 0

0 RADTRAD Version 3.02 run on 11/13/2001 at 13t56s28 Pathwayst 3

Pathway 1, Unfiltered Release to Environment File information 1 2

4 Plant file name - pnppfha.psf Pathway 2s Inventory file name - dt\hwagage\computer codes\radtrad\run Unfiltered Environment to CR batch\perry\pnppfha.nif 2 Scenario file name - NEWSDF.SDF 3 Release file name - dt\hwagage\computer codes\radtrad\run 2 batch\perry\pnpp fha.rft Pathway 3t Dose conversion file name - ds\hwagage\computer codes\radtrad\run Control Room Exhaust batch\perry\pnppfha.dcf 3 2

2 End of Plant Model File NINONfff IN#i# Iffff If i If NINO If I #Nof##if Scenario Description Name, 9## i 04# NON 1 #fr ###4 f# f 0f Plant Model Filename, HHH# HHH HH Hf H Hf H #HH Hf t( H Source Terms ff gong ff 1 1 1 #f## f I.

I I l.OOOOE+00 dt\hwagage\computer codes\radtrad\run batch\perry\pnpp_fha.dcf dt\hwagage\computer codes\radtrad\run batch\perry\pnpp_fha.rft 0.0000÷E00 Radtrad 3.02 1/5/2000 1 perry fhas sensitivity case 2 0.0000+E00 0.998SE.00 0.O015E+00 1.0000E+00 Nuclide inventory Filet Overlying Pools di\hwagage\computer codes\radtrad\run batch\perry\pnpp fha.nif 0 Plant Power Levels 0.00009+00 3.8332E+03 0 Compartmentsf 0 3 0 Compartment it 0 Containment Compartments, 3 3 1.0000E+00 Compartment Is 0

0 1 0 0 0 0 0 0 Compartment 2, 0 Environment 0 2 0 O.O000E+00 0 0 Compartment 2t 0 0 0

0 00 0

0 0 M Compartment 3. 0 C4 Zi--C Control Room 0 0 rz 1 0 3.6707E+05 0 0) K, 0 0 C) 0 Compartment 31 1 1

-j BITS OT~ua~s 0 u 0 0 0

'To 0 0 0 2z 0 00o3000010 00+30O0000 00+30000 .0 00+30000.0 z0+a000r-L 00+3000010 00+30000.0 00+30000.0 00+30000.0 00+31000oo W 0w 00+3000010 00+30000.0 0+3Q000.Q (0+30009.9 00+30000.0 0(.) I0 C)

D>

wa-a.m 0T~O0000' 10+3000Q~z toaoooo01 00+3000019 0 Z0-30009*Z 00+30000*0 0 0

SS.alWgRItd UOT2VnWJ$ 0 00+3000010 00*30O000, 0 t0Z30005,c 00+30000'0 0

0 IU014900'I RwfltA RATI~a 0 00+30O000O ZO+3000;.L 0 10Z0O000* 10+3000916 0 10-3000019 10.ao00or~ 00#300010 003000.

0+3000.0 003000*0 Q*3 00*

00+3O000Q* 00+3000010 00+3000010 00+30O000. 00+20000 .0 00+3000010 00+3100010 00+30000-0 00+30000.0 OO+30000.0 E0+3000900 00+3100000 00+3000 oo~oooo ooaoooo C.~oo9~ 0t3000 00+30000o0 ZO+a00o*L I t0-3000SC 00+30000'0 0 0 0 0

1z 0mqe 0 0

~o*~ooor H 10302100 00+3000010 ZO+3000ZIL t1030008E 10+300008r 01+300001T 00+3000010 to-3000o1 00+30000os 0

00+30000*0 00+30000Ot 0 60*20008*t 00+30000*0 0 0

0 0

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tO-0.0o06. 00+30000*0 0

00*30000*0 00+300001z 10+30000's 10+30000.s 10+30oo0os rO+3000ZIL WO-300oVW 00+30000'0 10+3000015 10+30000's 10+30000's 00+30000*r 00+300O000 00+30000,0 00+30000*0 00+3000010 P0#1I000t2 z Au sslu vipu~jullsOa2 0

0 0 0 0

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Caic. No. 3.2.15.14, Rev. 0 Attachment 7 Page 3 of 9 RADTRAD Version 3.02 run on 11/13/2001 at 13tS6t28 Plant Description i on nil fi n" i f ""i if no Pog ifi i No" f fif on" f~ifiiffifafan"If i Ron" i No ofiffif no No afNORfi gnu No"f gnu i fi ilfinififif iifofi iiiiii Number of Nuclides 23 Inventory Power - 3.8332E.03 MWth Plant Power Level - 3.8332E+03 tlth Number of compartments - 3 Compartment information Compartment number 1 (Source term fraction - 1.00005.00 Name: Containment Compartment volume - 1.0000.E00 (Cubic feet)

Pathways into and out of compartment 1 Pathway to compartment number 2t Unfiltered Release to Environment Compartment number 2 Namer Environment Pathways into and out of compartment 2 Pathway to compartment number 3t Unfiltered Environment to CR Pathway from compartment number Is Unfiltered Release to Environment Pathway from compartment number 31 Control Room Exhaust Compartment number 3 Name: Control Room Compartment volume - 3.6707E+03 (Cubic feet)

Removal devices within compartment.

Pilter(s)

Pathways into and out of compartment 3 Pathway to compartment number 21 Control Room Exhaust Pathway from compartment number 21 Unfiltered Environment to CR Total number of pathways - 3 0

0 0M M

i. 14U. 3.4.

'dIU. . 14, rev. U Attachment 7 Page 4 of 9 7.2000E+02 0.000DE20 O0.0000200 0.00009+00 O0.O000000 RADTRAD Version 3.02 run on 11/13/2001 at 13:56:28 Pathway number 3: Control Room Exhaust Pathway Filtert Removal Data Scenario Description Time (hr) Flow Rate Filter Efficiencies (%}

1. 88888#888888888f#8888888##f#####f########f#u####### 88###n##

{cfm) Aerosol Elemental Organic 0.0000.E00 6.6000E.03 0.OOOOE+00 0.OOOOE,00 0.0000.E00 Radioactive Decay is enabled 1.0000Eo04 0.0000Z.00 0.0000E+00 Calculation of Daughters is enabled O.0000.E00 0.0000.E00 7.2000E÷02 0.0000.E00 0.OOOOE+00 0.0000.E00 0.OOOOE*00 RELEASENAME - Perry FHA Release Fractions and Timings LOCATION DATA GAP EARLY IN-VESSEL Location Exclusion Area Boundary is in compartment 2 0.0001 hrs 0.0000 hra NOBLES 1.00002E00 O.0000.E00 Location X/O Data IODINE 1.0000E400 O.O0000E00 Time (hr) X/O (a

m* 3)

CESIUM 1.00002+00 0.00002E00 0.0000E400 4.30001-04 TELLURIUM 0.00001300 0.0000+O00 2.0000E+00 0.0000E+00 STRONTIUM O.0000.E00 0.0000+E00 BARIUM 0.0000E+00 0.0000.E00 Location Breathing Rate Data RUTHENIUM O.00002E00 O.0000÷+00 Time (hr) Breathing Rate (m^3

sec^-l)

CERIUM 0.0000÷+00 0.00002E00 0.0000E400 3.50002-04 LANTHANUM 0.00001300 O.O000E+00 8.0000E+00 1.8000E-04 2.4000E+01 2.30002-04 Iodine fractions Location Outer Boundary of the LPZ is in compartment 2 Aerosol . 0.0000z+00 Elemental - 9.98$08-01 Location X/O Data Organic - 1.5000E-03 Time (hr) X/O (a

m^-3) 0.00002+00 4.8000z-0; COMPARTMENT DATA 2.0000.E00 0.OOOE 00 Compartment number It Containment Location Breathing Rate Data Time (hr) Breathing Rate (m'3

sec^-1)

Compartment number 2: Environment 0.0000.E00 3.50002-04 Compartment number 8.0000r200 1,80000-04 3: Control Room 2.4000E÷01 2.3000.E04 Location Control Room Is in compartment 3 Compartment rilter Data Location X/O Data Time (hr) Flow Rate Filter Efficiencies (t) Time (hr) X/1 (s , mA-3)

(cfm) Aerosol Elemental Organic 0.0000E+00 3.:0000.04 0.0000+E00 2.7000E204 0.00002+00 O.00002E00 0.0000E+00 2.OOOOE+00 0.00002+00 2.00002+00 2.7000E+04 5.0000E+01 5.00002E01 5.0000E+01 7.2000E+02 2.7000E+04 5.00002+01 S.0000.E01 5.0000E+01 Location Breathing Rate Data Time (hr) Breathing Rate (m^3

seCA-l)

PATHWAY DATA 0.0000E+00 3.5000E-04 7.2000E202 0.0000E200 Pathway number I: Unfiltered Release to Environment Location Occupancy Factor Data Convection Data Time (hr) Occupancy Factor Time (hr) Flow Rate (% / day) 0.00002.00 I.0000E.00 O.0000E00 1.00002+10 2.4000E+01 6.0000E-01 7.2000E+02 0.0000+E00 9.6000S+01 4.0000E01 7.2000C+02 0.0000E400 Pathway number 2: Unfiltered Environment to CR 0) -< 0 USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STPS to 6a0 Pathway Filter. Removal Data Time Time step TT O.0000.E00 2.5000E-02 0Y)Z C Time (hr) Flow Rate Filter Efficiencies (t) 8.0000E400 1.00002-01 (cfm) Aerosol Elemental Organic 2.4000E201 4.00002-01 0.0000.E00 6.6000E+03 0.00002.00 0.00002+00 0.00002+00 7.2000E+02 0.00002400 o 1.0000E-04 0.00002+00 O.00000.00 O.00002E00 0 CD 0.0000+E00

Calc. No. 3.2.15.14, Rev. 0 Attachment 7 Page 5 of 9 Xe-135 Xe-135m 1.3783E401 6.5313E-01 S.39739:09 2.407gE416 5.0998El1l 7.1700E-12 3.1984E+13 2.4166E.+0 RADTRAD Version 3.02 run on 11/13/2001 at 13:56:28 Cs-135 4.7418E-14 4.1157E-14 1.8359E+11 iffiiffi#ffiiffi#ifh#i#ififfifii#iififfifsfifs##sfifsfiffifiifsfifii#iffiff#fi#f#ffiffifi##iff#ffifii# 1.7545E.03 Control Room Transport Group Inventorys Overlying Time (h) - 0.0001 Atmosphere Sump Pool Noble gases (atoms) 1.s969E+19 0.0000E400 0.OOOOE+00 ling" 11# g#t # N #4000ft Elemental I (atoms) 1.8787E+17 0.OOOOE+00

## # ### # # # 0.0000E.00

I # # # # 9 # Organic I (atoms) 2.6223E+14 0.OOOOE+00 0.OOOOE+00 9 # # 1111111 # # # Aerosols (kg) 0. 0.0000L+00 E.OOOOE+00 0.OOOOE+00 9 # # # # # #

0 # i i i i i f f Deposition Recirculating Time (h) - 0.0001 Surfaces Filter hffif #ifif # # if# if Noble gases (atoms) 0.0000E,00 0.0000E+00 Elemental I (atoms) 0.O000E*00 O.0000E+00 Organic I (atoms) 0.0000E+00 0.0000E+00 Aerosols (kg) 0.0000E+00 0.00009+00 Dose, Detailed model and Detailed Inventory Output Unfiltered Environment to CR Transport Group Inventoryt Exclusion Area Boundary Doses Pathway Time (h) - 0.0001 Filter Time (h) - 0.0001 Whole Body Thyroid TEDE Noble gases (atoms) 0.0000E.00 Delta dose (rem) 4.2456E-01 0.OOOOE+00 1.4377E+00 Elemental I (atoms) 0.OOOOE+00 Accumulated dose (rem) 4,2456E-01 O.0000E+00 1.4377E+00 Organic I (atoms) 0.0000E+00 Aerosols (kg) 0.OOOOE+O0 Outer Boundary of the LPZ Dosest Control Room Exhaust Transport Group Inventory:

Time (h) - 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.7393E-02 0.0000E+00 1.6048E-01 Pathway Accumulated dose (rem) 4.7393E-02 0.0000E+00 1.6048E-01 Time (h) - 0.0001 Filter Noble gases (atoms) 0.0000E+00 Control Room Doses: Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Time (h) - 0.0001 Whole Body Thyroid TEDE Aerosols (kg) 0.0000E+00 Delta dose (rem) 1.2078E-06 0.OOOOE+00 4.5686E-05 Accumulated dose (rem) 1.2078E-06 0.OOOOE+00 4.5686E-05 Exclusion Area Boundary Doses, Control Room Compartment Nuclide Inventoryi Time (h) - 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.0214E-03 0.0000E+00 3.4SM7E-03 Time (h)

0.0001 Ci kg Atoms Bq Accumulated dose (rem) 4.2SS9E-01 0.0000E.00 Br-B2 1.4411E+00 6.0733E-04 5.6097E-13 4.1198E+12 2.2471E÷07 Br-83 1.6156E-0S 1.0227E-15 7.4200E+09 5.9776E+05 Outer Boundary of the LPZ Dosess Kr-83m 1.2500E-02 6.0583E-13 4.3957E+12 4.6248E+08 Kr-85 8.1467E-01 2.0751E-06 1.4702E+19 3.0143E010 Time (h) - 2.0000 Whole Body Thyroid TEDE Kr-85m 1.6234E-01 1.9727E-11 1.39762+14 6.0067E+09 Delta dose (rem) 1.1402E.04 0.OOOOE*00 3.8609E-0, Kr-87 2.63010-05 9.20$48-16 6.4273E÷09 9.7315E+05 Accumulated dose (rem) 4.7507E-02 O.0000E400 1.6087E-01 Kr-88 5.0143E-02 3.9989E-12 2.7366E+13 1.SS53E+09 Rb-87 4.4281E-24 5.0613E-20 3.5034E0.0 1.6384E-13 Control Room Doses, Rb-88 1.1716E-0S 9.7601E-17 6.6792E+08 4.33490+05 1-129 6.8183E-09 3.8601E-08 1.8020E+17 2.S220E+02 Time (h) - 2.0000 Whole Body Thyroid TEDE 1-130 1.4897E-03 7.6381E-13 3.$383E.12 5.5118E007 Delta dose (rem) 4.3741E-02 0.0000E00 1.80801300 1-131 1.9834E-01 1.5999E-09 7.3546E+15 7.3387E+09 Accumulated dose (rem) 4.3742E-02 0.00OOE400 1.8080E,00 1-132 1.5753E-01 1.5262E-1l 6.9627E+13 S.8288E.09 CDo m 1-133 1.2391E-01 1.0938E-10 4.9529E*14 4.5847E#09 Control Room Compartment Nuclide Inventorys 1-134 6.5286E-09 2.4473E-19 1.0999E006 2.4156E+02 1-135 2.0391E-02 5.8063E-12 2.5901E+13 7.5446E.08 Time (h) - 2.0000 CL kg Atoms Xe-129m

Bq 2.2597E-04 1.78S9E-12 8.3372E+12 8.3607E+06 Br-82 5.8394E-04 5.3937E-13 3.9611E+12 2.1606E#07 03 xe-131m 2.9842E-01 3.5627E-09 1.6378E+16 1.1041E+10 Br-83 9.04SSE-06 S.7251E-16 C)M0) 4.1544E+09 3.3468E+05 Xe-133 5.0066E+01 2.6747Eo07 1.21110.18 1.8525E+12 Kr-83m 5.8665E-03 2.84342-13 0 CD 2.0631E+12 2.1706E+08 Xe-133m 1.5255E+00 3.3997E*09 1.5394E+16 5.6442E÷10 Kr-85 8.1466E-01 2.0750.-06 1.4701E019 3.0143E010 M-

%..'ic. 'JO. /.L. Io. 14, N<ev. U Attachment 7 Page 6 of 9 Kr-85m 1.1914E-01 1.4477E-11 1.0257E+14 4.4081E+09 Delta dose (rem) 0.0000E*00 0.0000OE00 0.0000E#00 Kr- 87 8.84222-06 3.1216E-16 2.1608E+09 3.271M05 Accumulated dose (rem) 4.7507E-02 0.0000E.00 1.6087E.01 Kr-88 3.0777r-02 2.454SE-12 1.6797E.13 1. 13 88E+09 Rb-87 5.3930E-20 6.1643E-16 4.2669E409 1.9954E-09 Control Room Doses, Rb-88 3.4843E-02 2.9027E-13 1.9864E+12 1.2892E109 t-129 6.8183E-09 3.8601E-08 1.8020E117 2.5228E102 Time (h) - 8.0000 Whole Body Thyroid TEDE r-130 1.3316E-03 6.8277E-13 3.1629E112 4.9271E107 Delta dose (rem) 9.7343E-02 0.0000E100 4.9289E-01 1-131 1.9692E-01 I.S884E-09 7.3020E115 7.2861E+09 Accumulated dose (rem) 1.4108E-01 0.00001E00 2.3009E+00 1-132 8.6223E-02 8.3532E-12 3.8109E.13 3.1902E+09 1-133 1.1592E-01 1.0233E-10 4.6335E+14 4.2892E+09 Control Room Compartment Nuclide Inventoryt 10-134 1.3431C-09 S.03462-20 2.2626E+05 4.9693E101 t-135 1.6533E-02 4.7078E-12 2.1001E+13 6.1173E+08 Time (h) 8.0000 ci kg 119 Atoms Xe-129m 2.2434E-04 1.7731E-12 8.2773E112 Eq Xe-131m 2.9698E-01 3.5456E-09 1.6299E116 8.3006E+06 Br-B2 9.2289E-10 8.524 5 6.2604E+06 3.4147E+01 1.0988E+10 Kr-83m 6.0467E-04 2.9307E.14 2.124E1+11 Xe- 133 4.9536E101 2.6464E-07 1.1983+118 1. 8328E112 2.2373E*07 Kr-85 8.1463E-01 2.0749E-06 1.4701E+19 3.01411110 Xe-133m 1.48S80+00 3.3113E-09 1.4993E+16 S.4975$+10 Kr-85m 4.7085E-02 5.7215E-12 4.0S36E+13 1.7422E+09 Xe-135 1.1852E+01 4.6412E-09 2.0704E+16 4.3854E+11 Kr-87 3.3592E-07 1.1859E-17 8.2090E+07 1.2429E+04 Xe-135m S.5597E-03 6.10331-14 2.7226E11. 2.0571E+08 Kr-88 7.1163E-03 5.6752E-13 3.8837E+12 2.6330E+08 Cs-135 8.8064E-10 7.6435E-10 3.4096E+15 3.25841+01 Rb-87 3.4993E-22 3.99981E18 2.7686E+07 1.2948E-11 Rb-88 4.0916E-03 3.40861-14 2.3326E+11 1.5139E108 Control Room Transport Croup Inventory, 1-129 1.2123E-14 6.8636E-14 3.2041E+11 4.4857E-04 1-130 1.6912E-09 8.6715E-19 4.0170E+06 6.2576E+01 Overlying 1-131 3.4268E-07 Time (h) - 2 0000 Atmosphere 2.7641E-15 1.2707E+10 1.2679E+04 sump Pool;* 1-132 2.513SE-Q8 Noble gases (atoms) 1.5969E+19 0.0000+E00 1.60772-01 2.43501-18 1.1109E107 9.2999E#02 O.0000E+00 1-133 Elemental I (atoms) 1.8787E÷17 0.0000E.00 1.48981-16 6.74571108 6.2443E+03 0.000O0E100 I.135 1.5670O-08 4.46191-18 1.99048+07 5.79771102 Organic I (atoms) 2.8223E+14 0.0000E+00 0.0000E+00 Xe-129m 2.1953E-04 1.7351E-12 8.0999E+12 B.1227E+06 Aerosols (kg) 0.O00OE.00 0.60000E00 O.O0001E00 Xe-131m 2.9269E-01 3.4944E-09 1.6064E+16. 1.0830E+10 Xe-133 4.79731+01 2.56291-07 1.16051E18 1.7750E112 Deposition Recirculating Xe-133m Time 1.3727E+00 3.0592E-09 1.3852E+16 5.0790E÷10 (h)'- 2.0000 Surfaces Filter Xe-125 Noble gases (atoms) 0.0O00E+00 0.0O000E00 7.5012E+00 2.9373E-09 1.3103E+16 2.7754E+11 Xe-135m 2.S232E-08 1.6722C-19 7.4592E105 S.6359E+02 Elemental I (atoms) 0.0000E+00 0.00001+00 Cs-13S 1.24511-10 1.0807E-10 4 .8206E+14 4.6067E+00 Organic I (atoms) 0.60000E00 0.0O000E00 Aerosols (kg) 0.0000E+00 O.O600E+00 Control Room Transport Group Inventory, Unfiltered Environment to CR Transport Group Inventory, Overlying Time (h) - 8.0000 Atmosphere Sump Pool Pathway Noble gases (atoms)

Time (h)

2.0000 Filter 1.5969E+19 0.0O000E00 0.00000E+0 Noble gases (atoms)

Elemental I (atoms) 3.3405E11. 0.0000E+00 0.0 000+00 0.0000.E00 Organic I (atoms)

Elemental I (atoms) 0.0000S+00 5.0183E+08 0.0000E+00 0.0000E100 Aerosols (kg) 0.0000.E00 0.0000E+00 0.000OE+00 Organic I (atoms) 0.0000E1,00 Aerosols (kg) 0.0000r+00 Deposition Recirculating Time (h) - 8.0000 Surfaces Filter Control Room Exhaust Transport Group Inventorya Noble gases (atoms) 0.00001E00 0.0000E100 Elemental I (atoms) 0.00001E00 1.8787E+17 Pathway Organic I (atoms)

Time (h) - 2.0000 0.0000E+00 2.8223E+14 Filter Aerosols (kg)

Noble gases (atoms) O.O0000+00 0.00003+00 0.00003+00 Elemental I (atoms) 0.0000E+00 Unfiltered Environment to CR Transport Group Inventory, Organic I (atoms) O.O000E+00 Aerosols (kg) 0.0000E400 Pathway Exclusion Area Boundary Doses, Time (h) - 8.0000 Filter Noble gases (atoms) 0.0000E+00 Time (h) - Elemental I (atoms) 0.0000E+00 8.0000 Whole Body Thyroid TEDE 03 Organic I (atoms) O.0000E+00 Delta dose (rem) 0.0000E+00 0.0O00E+00 O.00001+00 Aerosols (kg) O.O00OE+00 CD 05 Accumulated dose (rem) 4.2559E-01 0.00001E00 1.4411E100 03~

Control Room Exhaust Transport Group Inventory:

Outer Boundary of the LPZ Doses, 03'.

Pathway Time (h) - 8.0000 Whole Body Thyroid TEDE Time (h) - 8.0000 Filter -4 M r-

1,t0. 11U. 0 .4.). 1'4, r*ev. U Attachment 7 Page 7 of 9 Noble gases (atoms) 0.0000E+00 Control Room Exhaust Transport Group Inventory, Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.00000400 Pathway Aerosols (kg) 0.0000z+00 Time (h) - 24.0000 Filter Noble gases (atoms) o.ooooE+00 Exclusion Area Boundary Doses, Elemental.? (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Time (h) - 24.0000 Whole Body Thyroid TEDE Aerosols (kg) O.O000E+00 Delta dose (rem) 0.OOOOE+00 0.0000E.00 0.0000E+00 Accumulated dose (rem) 4.2559E-01 0.0000E÷00 1.4411E+00 Exclusion Area Boundary Doses, Outer Boundary of the LPZ Doses, Time (h) - 96.0000 Whole Body Thyroid Delta dose (rem) O0.0000E00 0.00000+00 TEDE00+00 0.000 Time (h) - 24.0000 Whole Body Thyroid TEDE Accumulated dose (rem) 4.2559E-01 0.0000E+00 1.4411E÷00 Delta dose (rem) 0.00000E+0 0.0000E÷00 0. 0000+00 Accumulated dose (rem) 4.7507E-02 0.OOOOE+00 1.6087E-01 Outer Boundary of the LPZ Doses, Control Room Doses, Time (h) - 96.0000 TEDE Whole Body Thyroid Delta dose (rem) O.0000Z+00 0.0000E+00 0.0000E+00 Time (h) - 24.0000 Whole Body Thyroid TEDE Accumulated dose (rem) 4.7507E.02 0.0000E+00 1.60870-01 Delta dose (rem) 1.6709E-01 0.OOOOE00 1.6726E-01 Accumulated dose (rem) 3.0817E-01 O.0000E+00 2.4682Z400 Control Room Doses, Control Room Compartment Nuclide Inventory.

Time (h)

96.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.2611E-01 0.00000+00 2.2611E-01 Time (h)

24.0000 Ci kg Atoms Bq Accumulated dose (rem) 5.3428E-01 0.00000+00 2.69430+00 Kr-83m 1.4111E-06 6.8394E-17 4.9624E+08 5.22112+04 Kr-85 8.1453E-01 2.0747E-06 1.4699E+19 3.0138E+10 Kr-85m Control Room Compartment Nuclide Inventory, 3.9605E-03 4.8129E-13 3.4099E+12 1.4655E+08 Kr-88 1.4332E-04 1.1430B-14 7.8218E÷10 5.3029E+06 Rb-88 Time (h) - 9d.0000 Ci kg Atoms Bq 9.7010E-0S 8.0816E-16 5.53050+09 3.5894E006 Xr-85 Xe-129m 8.14100-01 2.0736E-0.

0 .4691E+19 3.0122E010 2.0721E-04 1.63770-12 7.6453E+12 7.6668E+06 Kr-dSm Xe-131m 5.7S172-08 6.98910.18 4 .9517E+07 2.12810+03 2.81540-01 3.3613E-09 l.S4520+16 1.0417E+10 Xe-129m Xe-133 4.4031E+01 1.59780.04 1.2628E.12 5 .8953E012 5.9119E+06 2.3523E-07 1.0651E+18 1.6292E+12 Xe-131m 2.3640E-01 Xe-133m 1.1114E+00 2.8224E-09 I .2975E+16 8.7470E009 2.47682-09 1.1215E+16 4.11200+10 Xe-133 2.98470+01 Xe-135 2.2144E+00 8.6714E-10 1.5946E-07 7.2201E+17 1.1044E+12 3.86820.1S 8.1934E+10 Xe-133m 4.29640-01 9.5751E-10 4 .3355E+15 Cs-13S 3.9856E-11 3.4593E-ii 1.5431E+14 1.4747E+00 1.5897E+10 Xe-13S 9.13839-03 3.5784E-12 I .5963E,13 3.3812E008 Cs-135 2.2265E-13 1.9325E-13 8 .6205E+11 8.2380E-03 Control Room Transport Group Inventorys Control Room Transport Group Inventory, Overlying Time (h) - 24.0000 Atmosphere Sump Pool Noble gases (atoms) 1.59699E19 0.0000E+00 O.0000E+00 Overlying Elemental I (atoms) Time (h) - 96.0000 Atmosphere Sump Pool 1.5500E-04 0.0000E+00 0.0000E+00 Noble gases (atoms)

Organic I (atoms) 2.3286E-07 0.0000E+00 0.0000E+00 1.5969E+19 0.0000E+00 0.0000E+00 Aerosols (kg) Elemental I (atoms) 1.5479E-73 0.0000E+00 0.0000E+00 o.ooooE+0o 0.0000+E00 0.0000E+00 Organic I (atoms) 2.3253E-76 0.0000E+00 0.0000.E00 Aerosols (kg) 0.00000+00 0.0000E+00 0.0000R+00 Deposition Recirculating Time (h) - 24.0000 Surfaces Filter Deposition Recirculating Noble gases (atoms) 0.0000E+00 0.0000E+00 Elemental I (atoms) Time (h) - 96.0000 Surfaces Filter 0.0000E+00 1.8787E+17 Noble gases (atoms)

Organic I (atoms) 0.0000E+00 2.8223E+14 0.0000E400 0.0000E+00 Elemental I (atoms) 0.0000.E00 1.8787E+17 Aerosols (kg) 0.0000E+00 0.0000E+00 Organic I (atoms) 0.0000E+00 2.8223E+14 Aerosols (kg) 0.0000.E00 0.00000+00 Unfiltered Environment to CR Transport Group Inventory, ;?-Urn Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) - 24.0000 Filter Noble gases (atoms)

Pathway 0.0000E+00 Time (h) - 96.0000 Filter 0D Elemental I (atoms) 0.0000E+00 Noble gases (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Elemental I (atoms)

Aerosols (kg) 0.00000+O0 0.00000E00 Organic I (atoms) 0.0000E+00 Aerosols (kg) 0.0000+00 4

dtC. 11O. J./. Ib.14, Hev. 0 Attachment 7 Page 8 of 9 Control Room Exhaust Transport Group Inventory/

Pathway Time (h) - 720.0000 Filter Pathway Noble gases (atoms)

Time (h) - 96.0000 Filter 0.0000E.00 Noble gases (atoms)

Elemental I (atoms) 0.00000E00 0.O0000E000 Organic I (atoms)

Elemental I (atoms) 0.0000E+00 0.0000E000 Aerosols (kg) 0.0000E+00 Organic I (atoms) 0.0000E+00 Aerosols (kg) O.O000E+00 2223 Exclusion Area Boundary Doses, Time (h) - 720.0000 Whole Body Thyroid TEDE 1-131 Summary Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 Accumulated dose (rem) 4.2559E.01 0.OOOOE+00 1.4411S+00 Containment Environment Control Room Outer Boundary of the LPZ Dosest Time (hr) 1-131 (Curies) 1-131 (Curies) 1-131 (Curies) 0.000 4.3771E-01 1.8194E+02 1.9834E-01 Time (h) - 720.0000 Whole Body 0.275 0.00000+00 1.8238E+02 Thyroid TEDE 1.98150.01 Delta dose (rem) 0.0000E+00 0.0000E+00 0.525 0.00000+00 1.8238E+02 0.0000E+00 1.9797E-01 Accumulated dose (rem) 4.7507E.02 0.0000E+00 0.775 0.0000÷+00 1.8238E+02 1.6087E-01 1.9779E-01 1.025 0.0000E+00 1.8238E+02 1.9761E-01 Control Room Dosest 1.27S 0.0000E+00 1.82380.02 1.9744E-01 1.525 0.0000£+00 1.8238E+02 1.9726E-01 Time (h) - 720.0000 Whole Body 1.775 0.00000.00 1.8238E+02 Thyroid TEDE 1.9708E-01 Delta dose (rem) 2.7824E-01 2.000 0.00000+00 1.82380+02 0.0000E+00 2.7824E-01 1.9692E-01 Accumulated dose (rem) 8.1252E-01 2.250 0.00000+00 1.82380E+02 0.0000E+00 2.9725E*00 1.1332E-01 2.500 0.0000E+00 1.8238E+02 6.5215E-02 Control Room Compartment Nuclide Inventory, 2.750 0.00000+00 1.8238E+02 3.7529E-02 3.000 0.0000£+00 1.8238E+02 2.1597E-02 Time (h) - 720.0000 3.250 0.0000E+00 1.82380+02 Ci kg Atoms Bq 1.2428E-02 Kr-85 3.500 0.00000+00 1.8238E+02 7.1523E-03 8.1036E-01 2.0641E-06 1.4624E+19 2.9983E+10 3.750 Xe-129m 1.67950-05 1.3274E-13 0.0000E+00 1.0238E+02 4.1159E-03 6.1967E0I1 6.2141E+05 4.000 0.0000E+00 Xe-131m 5.1993E-02 6.2073E-10 2.8535E+15 1.9237E+09 1.82380+02 2.3686E-03 Xe- 133 4.250 0.00000E+00 1.8238E02 9.7071e-01 5.1859E-09 2.3481E+16 3.5916E+10 1.3631C-03 Xe- 133m 4.500 0.00000.00 1.82380+02 1.13750-04 2.5352E-13 1.14790+12 4.2089E+06 7.8441E-04 4.750 0.00000+00 1.82380+02 4.5140E-04 Control Room Transport Group Inventory: 5.000 0.00000+00 1.8238E+02 2.5977E-04 5.250 0.0000+E00 1.8238E+02 1.4949E-04 5.500 0.00000+00 1.8238E+02 8.6028E-05 Overlying 5.750 0.00000+00 Time (h) - 720.0000 Atmosphere Sump Pool 1.8238E+02 4.9507E-05 Noble gases (atoms) 6.000 0.0000E400 1.8238E+02 1.5969E+19 0.0000E+00 0.00000+00 2.8490E-05 Elemental I (atoms) 6.250 0.0000E+00 1.82328+02 1.6395E-05 0.0000E+00 0.0000E+00 O.000 00+0 Organic I (atoms) 6.500 0.0000E+00 1.8238E+02 9.4349E-06 0.00000E00 0.0000E+00 0.0000E+00 Aerosols (kg) 6.750 0.0000E+00 1.8238E+02 5.4295E-06 0.00009+00 0.0000E+00 0.0000E+00 7.000 0.00002+00 1.8238E+02 3.12450-06 7.250 0.00 00+00 1.82380+02 Deposition Recirculating 1.79810-06 Time (h) - 720.0000 7.500 0. 0000+00 1.8238E+02 Surfaces Filter 1.0348E-06 Noble gases (atoms) 7.750 0.00000+00 1.8238E+02 5.9547E-07 0.00000+00 0.0000E+00 8.000 Elemental I (atoms) 0. 0000+00 1.82380,02 3.4268E-07 0.00000.00 1.8787E+17 0. 0000+00 Organic I (atoms) 0.0000E+00 2.8223E+14 8.400 1.8238E002 1.4155E-07 Aerosols (kg) 8.700 0.00000+00 1.82380+02 7.29380-08 0.0000E+00 0.0000E+00 9.000 0.0000E+00 1.8238E+02 3.75820-08 Unfiltered Environment to CR Transport Group Inventory, 9.300 0.0000E+00 1.8238E002 1.9365S-08 9.600 0.0000R+00 1.8238E002 9.9780E-09 9.900 0 00000+00 1.8238E+02 S.1413E-09 Pathway 10,200 Time (h) - 720.0000 0,00000+00 1.8238E+02 2.6491E-09 CD 05 Filter 24.000 Noble gases (atoms) 0.0000E+00 0.0000E400 1.8238E002 1.5013E-22 Elemental I (atoms) 96.000 0.0000E+00 1.8238E+02 1.157S5-91 0.0000E+00 720.000 o-c Organic I (atoms) 0.00000+00 1.8238E+02 0.0000S+00 0.0000E+00 Aerosols (kg) 0. 0000+00 Cumulative Dose Summary N3 Control ROOMExhaust Transport Group Inventoryl 0-

,aIc. No. J.2.15.14, Rev. 0 Attachment 7 Page 9 of 9 Exclusion Area Bounda Outer Boundary of the Control Room (hr) (rem) 0.0 1.14020-04 (rem) 0.00000.00 (rem) 3.8609E-04 Time Thyroid TEDe Thyroid TEDE Thyroid (hr) TEDE (rem) (rem) (rem) (rem) (rem) (rem) Control Room 0.000 O.0000E+00 1.4377E+00 0.0000E+00 1.60480-01 0.0000E+00 4.5686E.05 0.275 O.0000.E00 1.4411E000 0.0000E+00 1.6087E-01 0.0000E+00 2.5085E-01 Time Whole Body Thyroid TEDE 0.525 0.00002+00 1.4411E400 0.OOOOE+00 1.6087E-01 0.0000E+00 4.7825E-01 (hr) (rem) (rem) (rem) 0.775 0.0000.E00 1.4411E+00 0.00000E00 1.6087E-01 0.0000.E00 7.0504E-01 0.0 4.3741E-02 O.0000E+00 1.80800,00 1.025 O.0000E+00 1.4411E*00 0.O000E+00 1.6087E-01 0.0000E*00 9.3124E-01 1.275 O.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000Z,00 l.S25 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 1.1569E000 1.77S 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 1.38192400 2.000 0.0000.E00 1.4411E+00 0.0000+E00 1.6087E-01 O.O000E+00 1.6064E+00 2.250 0.0000,E00 1.4411E+00 0.0000E+00 1.6087E-01 0.00000E00 1.8080E+00 2.500 0.00000+00 1.4411E+00 0.O0000E00 1.6087E-01 0.0000E+00 1.9809E000 2.750 0.0000E+00 1.4411E+00 0.00000,00 1.6087E-01 0.0000E+00 2.0823E+00 2.1424E+00 3.000 0.0000E+00 1.4411C+00 0.0000E+00 1.6087E-01 O.00000÷00 3.250 O.0000E+00 1.4411E+00 0.0000+E00 1.6087E-01 0.0000E+00 2.17890+00 2.2017E+00 3.500 0.0000S.00 1.44110*00 0.0000.E00 1.6087E-01 0.00OOE÷00 2.2166E÷00 3.750 0.0000E+00 1.44110+00 0.0000Z+00 1.6087E-01 0.O0000E00 4.000 0.0000E+00 1.4411E+00 0.0000+E00 1.6087E-01 0.0000E+00 2.22700400 2.2347E+00 4.250 0.00000+00 1.4411E+00 0.O000E+00 1.6087E-01 0.O000E+00 4.500 0.0000E+00 1.4411e+00 0.00000+00 1.6087E-01 0.00000+00 2.2410E+00 2.2463E000 4.750 0.0000E+00 1.4411E000 0.0000E+00 1.6087E-01 6.000,E÷00 2.2510E+00 3.000 0.0000t+00 1.44110+00 0.O000E+00 1.6087E-01 0.O000E+00 5.250 O.00000E00 1.4411E+00 0.00000+00 1.60870-01 0.00000+00 2.2555E000 5.500 0.00000E00 1.44110,00 0.0000+E00 1.6087E-01 0.0000+E00 2.2597E,00 5.750 O.0000,+00 1.4411E+00 0.0000E000 1.6087E-01 0.0000E+00 2.2638E+00 2.2677E÷00 6.000 0.0000+E00 1.4411E+00 0.0000E+00 1.6087E-01 0.00000E+0 2.27160+00 6.250 0.0000E+00 1.4411E+00 0.0000+E00 1.6087E-01 0.00000E00 6.500 O.00000E00 1.4411E+00 0.0000E+00 1.6087E-01 0.00000+00 2.2755E+00 6.750 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00000+00 2.2792E+00 2.2830E+00 7.000 0.O0000E00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000+00 7.250 0.00000+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00000+00 2.2866E+00 7.500 0.00009+00 1.4411E+00 0.00000,00 1.6087E-01 0.0000E+00 2.2903Z+00 7.750 0.0000E+00 1.44110B00 0.00000E00 1.6087E-01 0.OOOOE,00 2.2939E,00 2.2974E000 8.000"0.00000E00 1.4411E+00 0.0000E+00 1.6087E-01 0.O0000E00 8.400 0.0000E+00 1.4411E+00 0.00000+00 1.60870-01 0.00002+00 2.3009E,00 8.700 0.0000E+00 1.4411E+00 0.0000Z+00 1.6087E-01 0.00000+00 2.3065E+00 9.000 0.0000E+00 1.4411E+00 0.00000+00 1.60879-01 O.O000E+00 2.3105E+00 2.31460+00 9.300 0.00000E00 1.4411E+00 0.0000+E00 1.6087E-01 0.00000+00 9.600 0.0000E+00 1.4411E000 0.0000E+00 1.6087E-01 0.0000R.00 2.3186E+00 2.3225E000 9.900 0.O000E+00 1.4411E+00 0.0000E+00 1.60870-01 0.00000+00 10.200 0.0000+E00 1.4411E+00 O.00000E00 1.60870-01 0.0000+E00 2.3264E+00 24.000 0.0000E*00 1.4411E000 0.00000E00 1.6087E-01 O.00000E00 2.3302E+00 2.4682E,00 96.000 0.0000E+00 1.4411E+00 0.00000,00 1.60870-01 0.0000,E00 2.6943E400 "720.000 0.0000,E00 1.4411E+00 0.0000E+00 1.6087E-01 0.00000+00 2.9725C+00 Worst Two-Hour Doses Notei All of the dose locations are shown below but the worst two-hour dose is only meaningful for the EAS dose location. Please disregard the two-hour worst doses for the other dose locations

-U m Exclusion Area Boundary Time Whole Body Thyroid 01-<

TEDE CD0m M*

-,4 0 (hr) (rem) (rem) (rem) 0.0 1.0214E-03 0.0000E+00 3.4587E-03 Outer Boundary of the LPZ Time Whole Body "0Z-C z

Thyroid TEDE 0 CD

Enclosure 4 PY-CEI/NRR-2674L Page 72 of 76 Calc 3.2.15.14. Revision 0 ED SAFETY RELATED 0] AUGMENTED QUALITY [] NONSAFETY.RELATED SUPPORTING/REFERENCE DOCUMENTS DESIGN ORIGINATOR: (Printand Sign Name) DATE Hao~nrý A.W &k" 4 3i 1 oa12 VERIFICATION METHOD (Check one)

[ DESIGN REVIEW (Complete Design [I ALTERNATE CALCULATION [] QUALIFICATION TESTING Review Checklist or CalculationReview Checklist)

JUSTIFICATION FOR SUPERVISOR PERFORMING VERIFICATION:

The SCIENTECH manager performed this verification since this is an area for which he stays abreast of the industry positions and is known throughout the industry as an expert in this area.

APPROVAL: (Print and Sign Name) DATE 1--AlrINI %-Jr V IIL Ll U The FHA calculation was completely reviewed, including the methodology, the use of the RADTRAD code, the conformance with current regulatory expectations, etc. This calculation was found to be acceptable and appropriate for inclusion with the LAR to the NRC. In addition, during the course of the calculation development, an analysis was done using an Excel spreadsheet which yielded identical results. Editorial comments and comments on the wording were provided to the preparer and have been incorporated.

COMMENTS, ERRORS OR DEFICIENCIES IDENTIFIED? 2 YES [] NO RESOLUTION: (ForAitemate Calculationor QuahficationTesting only)

RESOLVED BY: (PrintandSign Name) DATE VERIFIER: (Pnnt andSign Name) DATE APPROVED BY: (Printand Sign Name) 1

, "/ T DATE to

FirstEnergy CALCULATION REVIEW CHECKLIST CALCULATION NO.

Page 1 of 2 REV. 6)

NOP-CC-2001-04 Rev. 00 QUESTION UNIT INA Yes NoI COMMENTS RESOLUTION REFERENCES X 1 Does the stated objective/purpose clearly describe why the calculation Is being performed?

2. Are applicable codes, standards, design/licensing basis documents, etc., Including edition and X addenda where appropriate clearly Identified?

3. Do the references reflect the appropriate revision? X INPUTS X

4. Are design inputs clearly Identified and their source documents referenced, Including revision level as appropriate?

5. Are the design Inputs relevant, current, consistent with deslgn/licensing bases and directly

- X applicable to the purpose of the calculation, Including appropriate tolerances and ranges/modes of operation?

6. Are all design inputs retrievable? Ifnot, have they been added as attachments? X

7. Are preliminary or conceptual Inputs clearly identified for later confirmation as open assumptions? X ASSUMPTIONS 0 X

8. Have the assumptions necessary to perform the analysis been adequately documented?

9. Is suitable justification provided for all assumptions (except those based upon recognized X engineering practice, physical constants or elementary scientific principles)?

10. Are all assumptions for the calculation reasonable and consistent with design/licensing bases?

X

11. Have all open assumptions needing later confirmation been clearly Identified on the Calculation X cover sheet, Including when the open assumption needs to be closed?

12. Has a Condition Report been Issued for open assumptions if required? X

13. Have engineering judgments been used? X

14. Are engineering judgments reasonable and adequately documented? X METHOD OF ANALYSIS X

15. Is the method used appropriate considering the purpose and type of calculation?

16. Is the method In accordance with applicable codes, standards, and design/licensing bases?

X IDENTIFICATION OF COMPUTER CODES (Ref: NOP-SS-1001) X

17. Have the versions of the computer codes employed In the design analysis been certified for this application?

18. Are codes properly Identified along with source, Inputs and outputs?

X -...

19. Is the code suitable for the analysis being performed? X

20. Does the computer model, that has been created, adequately reflect actual (or to be modified) X plant conditions (e.g., dimensional accuracy, type of model/code options used, time steps, etc.)? _ "U "_

21. Is the computer output reasonable when compared to Inputs and what was expected? X -

0 COMPUTATIONS X

22. Are the equations used consistent with recognized engineering practice and deslgn/licensIng o

0 bases? oU

23. Is justification provided for any equations notin common use? X -

24. Is the justification reasonable? X 0)

25. Have adjustment factors, uncertainties, empirical correlations, etc., used in the analysis been X correctly applied?

26. Is the result presentedwith proper units and tolerance? X

27. Has proper consideration been given to results that may be overly sensitive to very small X changes In Input?

Page 2 of 2 FirstEnergy CALCULATION REVIEW CHECKLIST CALCULATIONNO.

REV. O NOP-CC-2001-04 Rev. 00 UNIT QUESTION lNA jYes INo COMMENTS RESOLUTION CONCLUSIONS X

28. Is the magnitude of the result reasonable when compared to Inputs?

29. Is the direction of trends reasonable?

I X

30. Are stated conclusions justifiable based on the calculation results?

X "3'1. Are all pages sequentially numbered and marked with a valid calculation number? X

32. Is all Information legible and reproducible? X

33. Have all changes In the documentation been Initialed (or signed) and dated by the author of the X change and all required reviewers?

34. Have all calculation results stayed within existing design/licensing basis parameters? X This calc deviates from current LAR being prepared to adopt design and licensing basis. this calculation for PNPP.

35. If the response to Question 34 Is NO, has Licensing been notified as appropriate? (i.e. UFSAR or X Tech Spec Change Request has been Initiated).

36. Does the calculation meet Its purpose/objective? X

37. Has the calculation vendor used all applicable design information/requIrements provided? X

38. Did the calculation vendor determine if the calculation was referenced In design basis documents X and/or databases?

39. Did the Preparer determine If the calculation was used as a reference In the UFSAR? X

40. If the calculation Is used as a reference In the UFSAR, Is a change to the UFSAR required or an X USAR affected. Once LAR Is approved, the update to the UFSAR Validation Database, Ifapplicable, required? USAR will be updated. LAR includes USAR changes

41. Ifthe answer to Question 40 is YES, have the appropriate documents been Initiated? X See item 40.

42. Is the calculation acceptable for use? X

43. What checking method was used to review the calculation? Check all that apply.

" spot check for math X "complete check for math X

" comparison with tests

" check by alternate method X

" comparison with previous calculation X Review Summary: The FHA calculation, methodology, use of the code, etc. was completely reviewed and found acceptable. In addition, during the course of the calculation development, an analysis was done using an Excel spreadsheet which yielded Identical results. Editorial comments provided to clarify the calculation.

The EDE for 1135 in the calc does not agree with FR12 (available at http:llwww.epa.govlradlation/foderalldocs/fgr12.pdfo. The calc used a value from MACCS2 (DIN15) which is slightly higher than the FR12 value (conservative which Is OK).

Reviewer (Print and Sign Name)

Vt., " I U*I I*VVIV*d V Inuumnnr'@ ,-,,V,,,.*

A n*ea

n ,, .,,"

D ljý ,.u ,,u IO..,-,4 ,u, f-. o.-,culaitonsprepared - by

. a vendog Reviewer (Printand Sign Name) uate Reviewer: (Printand Sign Name) Date David A. Studley Q-C ? I Approver: (Print and Sign Name)

D -

.. Qz

" Date

)'JX

-4 0r 0)

Page 1 of 2 FrstEnery7 CALCULATION REVIEW CHECKLIST CALCULATION NO. 3.2.15.14 REV. 0 NOP.CC-2001-04 Rev. 00 UNIT Porn, QUESTION .NA Yes NoI COMMENTS RESOLUTION REFERENCES X I Does the stated objective/purpose clearly describe why the calculation Is being performed? ..

2 Are applicable codes, standards, design/licensing basis documents, etc., Including edition and x addenda where appropriate clearly Identified?

3 Do the references reflect the appropriate revision? x INPUTS x

4. Are design inputs clearly Identified and their source documents referenced, Including revision level as appropriate?

5. Are the design inputs relevant, current, consistent with design/licensing bases and directly x Not necessarily consistent with N/A applicable to the purpose of the calculation, Including appropriate tolerances and ranges/modes design/license bases. License of operation? amendment request required.

6.' Are all design inputs retrievable? Ifnot, have they been added as attachments? x

7. Are preliminary or conceptual inputs clearly identified for later confirmation as open assumptions? x ASSUMPTIONS x 8 Have the assumptions necessary to perform the analysis been adequately documented? ,,

9. Is suitable justification provided for all assumptions (except those based upon recognized x engineering practice, physical constants or elementary scientific principles)?

10. Are all assumptions for the calculation reasonable and consistent with deslgn/licensing bases? x See comment to Item 5 N/A

11. Have all open assumptions needing later confirmation been clearly Identified on the Calculation X cover sheet, including when the open assumption needs to be closed?

12. Has a Condition Report been issued for open assumptions if required? X

13. Have engineering judgments been used? X_,,,_,,

14. Are engineering judgments reasonable and adequately documented? X METHOD OF ANALYSIS X

15. Is the method used appropriate considering the purpose and type of calculation? ,

16. Is the method in accordance with applicable codes, standards, and design/licensing bases? x See comment to item 5 N/A IDENTIFICATION OF COMPUTER CODES (Ref: NOP-SS-1001) x

17. Have the versions of the computer codes employed in the design analysis been certified for this application'

18. Are codes properly identified along with source, inputs and outputs? x

19. Is the code suitable for the analysis being performed? x Code Is NRC endorsed N/A

20. Does the computer model, that has been created, adequately reflect actual (or to be modified) x plant conditions (eg., dimensional accuracy, type of model/code options used, time steps, etc.)? ............ .

21. Is the computer output reasonable when compared to inputs and what was expected? x COMPUTATIONS x See comment to item 5 N/A

22. Are the equations used consistent with recognized engineering practice and design/licensing bases?

23. Is justification provided for any equations not in common use? x ) -o m no 1 0

24. Is the justification reasonable? x -,4 M CA CZQ) C:

25 Have adjustment factors, uncertainties, empirical correlations, etc., used in the analysis been x o1*'

correctly applied?

26. Is the result presented with proper units and tolerance? x 0)Z

27. Has proper consideration been given to results that may be overly sensitive to very small x Refer to sensitivity analyses N/A ",4 changes in input? I_17T ] M-

Page 2 of 2 FirstEnergy CALCULATION REVIEW CHECKLIST CALCULATION NO. 3.2.15.14 REV. 0 NOP-CC-2001-04 Rev. 00

-.... _ _ UNIT Perry QUESTION NA Yes No COMMENTS RESOLUTION CONCLUSIONS X

28. Is the magnitude of the result reasonable when compared to inputs?

29 Is the direction of trends reasonable? x 30 Are stated conclusions justifiable based on the calculation results?

x 31 Are all pages sequentiully numbered and marked with a valid calculation number? x

32. Is all Information legible and reproducible? x 33 Have all changes in the documentation been initialed (or signed) and dated by the author of the x Two typographical errors corrected N/A change and all required reviewers? as part of this review on page 16.

34. Have all calculation results stayed within existing design/licensing basis parameters? x See comment to item 5 N/A

35. Ifthe response to Question 34 is NO, has Licensing been notified as appropriate? (Ie. UFSAR or x License amendment request has N/A Tech Spec Change Request has been initiated) been prepared.

36. Does the calculation meet its purpose/objective? x

37. Has the calculation vendor used all applicable design information/requirements provided? x

38. Did the calculation vendor determine if the calculation was referenced in design basis documents x FE to prepare USAR Change and N/A and/or databases?

ATLAS database updates

39. Did the Preparer determine if the calculation was used as a reference In the UFSAR? x 40 Ifthe calculation is used as a reference in the UFSAR, is a change to the UFSAR required or an x update to the UFSAR Validation Database, if applicable, required?

41. Ifthe answer to Question 40 Is YES, have the appropriate documents been initiated? x License amendment prepared. N/A

42. Is the calculation acceptable for use? x Pending NRC approval N/A

43. What checking method was used to review the calculation? Check all that apply. X Reviewed inputs to calculation. N/A

- spot check for math X Did not re-run RADTRAD Code.

- N/A

- complete check for math X _ _ _ _ _ _

. comparison with tests X

. check by alternate method X

. comparison with previous calculation X eevIew jummary: i ne UAR Change Request and ATLAS database update will consider the entire amendment request content and will be processed according to procedures upon NRC amendment approval. Owner's acceptance review looked at the calculation, sensitivities, appendix, as well as the attachments for approach, Inputs, and conclusions. It Is noted that Appendix A to the calculation evaluates a scenario that was Identified in Condition Report 01.4224 as a result of a review of the draft amendment request. The associated amendment request was reviewed to ensure that the calculation and request were consistent.

Fi1 Tn,. Ini^I,. 0 -

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L,.J I *.* b I II I II.*l;I I I \1.; V I*; Vy Xi tJWnnr ntn,A Rn',ina, Ilaa,,,,Wn,4 .....j L.. -

Reviewer Siit*n

nn N,-mnel n.,#^, I* Otner

.... S.'rn-a_ _, ., :n. 0***..

... "* l1- I 4 r--LLI ILI OII..LIOLIU I s prepaedu by a vendor)

L0a La Reviewer: (r-nni a d Sign vame) Date A. Widmer ,/ _*'Z/*f--

01-08-02 Approver: (Printand S/en Name) -IF) L Date

- mC 01 C5/0Z 0)

.-4 T; 0)N3 r-

TABLE 1.8-1 (Continued)

USAR Section/I Requlatorv Guide (Rev.;RRRC Cateqorv) Degree of Conformance Reference 1.12 - (Revision 1 - 4/74;RRRC Cat. 4)

Instrumentation for earthquakes PNPP design conforms to this guide with the 3.7.4 exception of Paragraph C.4.b, Response Spectrum Recorder Frequency Range. The Perry Nuclear Power Plant Response Spectrum Recorders have a frequency range of "2 Hz to 25.4 Hz," rather than the recommended 1 Hz to 30 Hz.

1.13 - (Revision 1 - 12/75;RRRC Cat. 4)

Spent fuel storage facility design PNPP design conforms to this guide.4 Wlir rYu basis fXUm-mIDE 1, 1 mzag fwI q./. TuL INvrd7Nroj 9.1, or 2ANDAoCTfIVI r1Arer'ALS.AVAu..AaLeV (6/a L-AJAA6I 9.4.2 1.14 - (Revision 1 - 8/75) GUIDE 1,193.

Reactor coolant pump flywheel Not applicable to PNPP design.

integrity 1.15 - (Revision 1 - 12/72;RRRC Cat. 1)

Testing of reinforcing bars for PNPP design conforms to this guide. 3.8.1, 0 Seismic Category I concrete structures 3.8.3, CD 0 3.8.4, 3.8.5

/t. '.*

K)

-. )

=v 3

Revision 11 1.8-7 September, 2001

TABLE 1.8-1 (Continued)

USAR Section/i Regulatory Guide (Rev.;RRRC Category) Degree of Conformance Reference 1.24 - (Revision 0 - 3/72;RRRC Cat. 1)

Assumptions used for evaluating the Not applicable to PNPP design.

potential radiological consequences of a pressurized water reactor gas storage tank failure 1.25 - (Revision 0 - 3/72;RRRC Cat+/-. 1)

Moj- &pp1;c.&61e- i- PNPP. -&-a- RA5'L\&jV1 akkkA.0, Assumptions used for evaluating the N PNPP d design conforms to this guide with potential radiological consequences of a fuel handling accident in the fuel following exceptions: a. (Regulato 19.lA1 handling and storage facility for Position C.1.j) filter efficle es of 95% 19A.2,I are1P used tio in.1c r0I sst accordance wýý egulatory boiling and pressurized water reactors LGuide P 1.52; C b. (Regul u ry Position C.3.a/c) dose conversion Oooinso6and average gamma I je ar energies 0tf from NRC TACT III aken fP0 and/o r T 5 computer comp t code in lieu of 1e 1 and T Reference r u c :e. 12. i p 1.26 - (Revision 3 - 2/76;RRRC Cat. 1) rWft Quality group classifications and PNPP design complies with this guide. 3.2.1, N'ON standards for water-, steam- and Table 3.2-1, radioactive-waste-containing components 6.2.4, of nuclear power plants 6.5, 6.7, "-0 "0 m M 9.4, 9.5, 0 N-10 10.3.3, (0 :7 a

ýPTl 17.2 m @

0 C I"

Revision 11 1.8-13 September, 2001

TABLE 1.8-1 (Continued)

USAR Section/ I Regulatory Guide (Rev.;RRRC Category) Degree of Conformance Reference 1.163 (Continued) The containment isolation check valves in the Feedwater penetrations are tested per the Inservice Testing Program.

8.1 - (Revision 0 - 2/73)

Radiation symbol PNPP conforms to this guide.

(t93 - C*-;,o

. .o-* *

/*-V-k4.*bp,.

o ,...*\UI*.&j:-c.

KzIIJkjLebr P

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,,,., +e,..N poor .s Z.ssu.,%A, 4a 1Q rI s-7 S. G nn M

,.4 Z'A Revision 11 1.8-59b September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 4 of 53 A radiologicAnalysis, as a result of a fuel drop f 1em-

0* ee Above the

.C~sk Po. racks rac - -6

.ý " " -

A the causes of the accident,

assumptions and starting conditions. The fission product release from the fue , the airborne activity n the fuf the environsjis calculated and the corresponding radiological effects offsite are evaluated using the methods, assumptions and conditions in Regulatory Guide/AL.25 and I.2.

di.cucn qhcPonseq"ences i__;,n-

_n_ &7 of the ot p-1r

£cztion 15 7 4_

9.1.2.3.3 Spent Fuel Rack Design - GE Racks Spent fuel rack design features are as follows (refer also to Figure 9.1-2):

a. Each containment spent fuel pool contains 19 sets of racks which may contain up to 190 fuel assemblies. A maximum of 380 fuel assemblies may be stored in the two spent fuel pools.

b. The storage racks provide an individual storage compartment for each fuel assembly and are secured to the pool wall through associated hardware. The fuel assemblies are stored in a vertical position with the lower tie plate engaged on a captive slot in the lower fuel rac&k support casting. Additional restraints are provided to restrict lateral movement.

c. The weight of the fuel assembly is held by the lower rack support casting.

d. The spent fuel storage racks are made from aluminum. Materials used for construction are specified in accordance with the latest issue of applicable ASTM specifications. The material choice is based on a consideration of the susceptibility of various metal combinations to electrochemical reaction. When considering the 9.1-21

Enclosure 5 PY-CEI/NRR-2674L Page 5 of 53 radiation alarm can be obtained in the control room following the leakage of less than 10 gallons. A plate out factor of 2 is taken into consideration for this evaluation.

With a reactor coolant pressure of 1,000 psi, and a flow restrictor of 1/4 inch, the initial flow rate through the break, assuming 100 percent flashing will be in excess of 10 gpm. Thus a monitor response to the potential hazard of less than 10 minutes is expected.

d. Fuel Handling Building Exhaust:

The drop of a channeled spent fuel bundle has been identified in Section 15.7.4 as a hazard for personnel in this building.

Assuming that .q-x 102 Ci of Kr85 (Table 15.7-k') are released and mixed instantaneously into the whole volume of the fuel handling building, the resulting concentration is expected to be 1.7 x 10-2 pCi/cc. This activity level is well within the range capability of the monitor.

The response time of the monitor is inversely proportional to the activity level and is expected to be negligible at the high anticipated levels which may be reached during this calculable fuel handling accident.

The particulate and iodine filters are removable for laboratory analysis to verify and identify activity levels and to provide a backup to the continual monitoring of the areas of surveillance.

z~

.. ;'" ~~p * * * ,F* t, 12.3-68

Enclosure 5 PY-CEI/NRR-2674L Page 6 of 53

d. Generation of a condition that results in a consequential loss of function of a necessary containment barrier.

15.0.3.1.3 Unacceptable Results for Limiting Faults [Design Basis (Postulated) Accidents] -.

The following are considered to be unacceptable safety results for limiting faults (design basis accidents):

a. Radioactive material release which results in dose consequences that exceed the guideline values of 10 CFR 100 (for the dQ&41j

o*a . --- b-ci T %LOCA

analysis, the offsite dose limit is 25 rem TED,,

b. Failure of fuel cladding which would cause changes in core geometry such that core cooling would be inhibited.

c. Nuclear system stresses in excess of those allowed for the accident c ssification by applicable industry codes.

d. Containmen tresses in excess of those allowed for the accident "classification bapplicable industry codes when containment is required.

e. Radiation exp.sure to plant oper ons personnel in the main control room in excess of 5 rem whole y, 30 rem inhalation and 75 rem skin [5 rem TEDE for the LOCA- ";T

~JA 15.0.3.2 Sequence of Events and Systems Operation Each transient or accident is discussed and evaluated in terms of:

a. A step-by-step sequence of events from initiation to final stabilized condition.

.9 !C-Revision 10 15.0-7 October, 1999

Enclosure 5 PYcCEI/NRR-2674L Page 7 of 53 The Low-Low Set (LLS) Relief Function, armed upon relief actuation of any S/R valve; will cause a greater magnitude blowdown, in the relief mode, for certain specified S/R valves and a subsequent cycling of a single low set valve. The effect of the LLS design on reactor coolant pressure is demonstrated, in Chapter 5, on the MSIV closure event. This is considered bounding for all other pressurization events and, therefore, is not simulated in the analysis presented in this chapter.

A sensitivity study was also performed to support higher analytical limits for relief valve setpoints. The study shows an increase of 20 psi in the relief valve setpoint causes legs than 20 psi increase in reactor peak pressures. However, these reactor peak pressures are still well below the ASME code limit (1,375 psig). Also, the increase of 20 psi in the relief setpoints does not have any effect on the peak surface heat flux, since all safety/relief valves open after the occurrence of MCPR during transients. Therefore, the analytical limits for relief valve setpoints in the Technical Specifications were 20 psi higher than those listed in Table 15.0-1.

The analytical limits used for the relief valve setpoints for the current reload analysis are listed on Table 15B.15.0-1 for the pressurization transients and on Table 15B.5.2-1 for the overpressurization transients. The analytical values are the basis for the deviation of the Technical Specification value.

15.0.3.5 Radiological Consequences ý co r In this chapter, the consequences of radioactivity release during the three types of events: incidents of moderate frequency (anticipated operational transients), infrequent incidents (abnormal operational transients) and limiting faults (design basis accidents) are considered.

For all events whose consequences are limiting a detailed quantitative evaluation is presented. For non-limiting events a qualitative evaluation is presented or results are referenced from a more limiting or enveloping case or event.

Revision 8 15.0-15 Oct. 1996

Enclosure 5 PY-CEI/NRR-2674L Page 8 of 53 For limiting faults (design basis accidents) two quantitative analyses are considered:

a. The first is based on conservative assumptions for the purposes of worst case bounding of event consequences to determine the adequacy of the plant design to meet 10 CFR 100 guidelines~ fort L Ca ASn A nalysis, the licensing basis limit is 25 rem TEDE(. This analysis is referred to as the "design basis analysis." b.%"& +-Or-0 i . .*' Ws*i,& AccAn&"Al -W- 1uut.Ai.

b. The second is ea is ic assumptions to reflect expected radiological consequences. This analysis is referred to as the "realistic analysis." The "realistic analysis" is not performed for the LOC analysiý ,

Results for both are shown to be within NRC guidelines.

Doses resulting from the events in Chapter 15 are determined either manually or by computer code. Doses associated with Offgas System Failure (Section 15.7.1.1) are evaluated using GASPAR II (NUREG/

CR-4653).(8' Time dependent releases are evaluated with the TACT computer code.( 2

)(6) Instantaneous or "puff" type releases are evaluated by methods based on those presented in Regulatory Guide 1.3)(I 3) and NUREG-1465. The General Electric NEDO-31400 analysis (7) also is utilized in determining doses associated with a Control Rod Drop Accident (Section 15.4.9). Dose conversion factors and breathing rates are presented in Table 15.0-4.

15.0.4 NUCLEAR SAFETY OPERATIONAL ANALYSIS (NSOA) RELATIONSHIP Appendix 15A is a comprehensive, total plant, system-level, qualitative failure modes and effects analysis, relative to all the Chapter 15 events considered, the protective sequences utilized to accommodate the events and their effects, and the systems involved in protective actions.

Revision 11 15.0-16 September, 2001

' 051 Ir 4 k*

js J ~~

En:!.nure 5 PY-CEI/NRR-2674L Page 9 of 53

3. "General Electric Company Model for Loss of Coolant Analysis in Accordance with 10 CFR 50, Appendix K," December 1975 (NEDO-20566).

4. ASME Boiler and Pressure Vessel Code, Section III, Class 1, "Nuclear Power Plant Components," Article NB-7000, "Protection Against Overpressure."

5. General Electric Company "General Electric Standard Application for Reactor Fuel" including the "United States Supplement,"

NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved revision).

6. U.S. Nuclear Regulatory Commission Computer Code TACT 5, Computer Code for Calculating Radiological Consequences of Time Varying Radioactive Releases, NUREG/CR-5106, June 1988.

7. General Electric Company "Safety Evaluiation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor" NEDO-31400A, Oct. 1992.

8. U.S. NRC Computer Code GASPAR II, Computer Code to Perform Environmental Dose Analyses for Release of Radioactive Effluents.

NUREG/CR-4653, March 1987.

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

10. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31897P-A, Class III (Proprietary), May 1992.

11. Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion", 2 d Printing, 1989

12. CCC-652 Oak Ridge National Laboratory RSICC Computer Code Collection MACCS2, V.I.12 Code Package, 1997

13. Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993 Revision 11 15.0-17f September, 2001

Enclosure 5 PY-CEI/N RR-2674L Page 10 of 53 TABLE 15.0-3

SUMMARY

OF ACCIDENTS Failed Fuel Pins GE NRC Worst Calculated .- Case Section Title Value Assumption 15.3.3 Seizure of One Recirculation Pump None 15.3.4 Recirculation Pump Shaft Break None 15.4.9 Control Rod Drop Accident <770 770 15.6.2 Instrument Line Break None None 15.6.4 Steam System Pipe Break Outside None None Containment 15.6.5 LOCA Within RCPB None 100%

15.6.6 Feedwater Line Break None None 15.7.1. 1 Main Condenser Offgas Treatment N/A N/A System Failure 15.7.3 Liquid Radwaste Tank Failure N/A N/A 15.7.5 Spent Fuel Cask Drop Accident None None 15.7.6 Fuel Handling Accident Inside ý9: s1,5 _-11U.

Containment(Ch P / f-,,L Q/r/R.'NGuL'Li 15.8 ATWS SPECIAL EVENT Revision 11 15.0-25 September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 11 of 53 TABLE 15.0-4 DOSE CONVERSION FACTORS*

Thyroid Whole Body Isotope (rem/Ci) 0.25xMeV/dis 1-131 1.49E+6 8.72E-2 1-132 5.35E+4 5.13E-1 1-133 3.97E+5 1.55E-1 1-134 2.54E+4 5.32E-1 1-135 1.24E+5 4.21E-1 Kr-83m 5.02E-6 Kr-85 3.72E-2 Kr-85m 5.25E-4 Kr-87 1.87E-1 Kr-88 4.64E-I Kr-89 5.25E-1 Xe-131m 2.92E-3 Xe-133m 8.OOE-3 Xe-133 9.33E-3 Xe-135m 9.92E-2 Xe-135 5.72E-2 Xe-137 4.53E-2 Xe-138 2.81E-1 Breathing Rates Time Period Breathing Rate (hr) (m (3)/sec) 0-8 3.47E-4** 3.5'6- I 8-24 1.75E-4 I,& E-4 24-720 2.32E-4 Z,3

_fnT j.oing dose conversion factors (DCF's) are used in the Q

f u*r*C.A

4-analyq; LDOCA- DCF's for inhalation: EPA Federal Guidance Report ll*(Reference 9)

- This breathing rate was used for the duration of the Control Room radiological consequence analyses.

(APr C-C-SL

>b~/E~t~ ju* we.e-GOA &-'- er4A CI, 12)~

Revision 10 15.0-26 October, 1999

"-aq i 'u *ff ,t -* N D*

Fnclosure 5 PY-CEI/NRR-2674L Page 12 of 53

k. Credit is taken for dilution in the lake to the nearest drinking water intake (0.5 miles ENE of the plant) as presented in Table 5.1-10 of the Perry Nuclear Power Plant Environmental Report (Operating License Stage).

The resulting exposures from liquid releases to the groundwater are presented in Table 15.7-14.

The individual isotopic concentrations and fraction of maximum permissible concentrations (FMPC) for the radionuclides released by a postulated failure of the concentrated waste tank are given in Table 15.7-15a. (Radiological assessments performed prior to October 4, 1993 that were used for the plant design bases as discussed in this USAR were evaluated against the 10 CFR 20 regulations prior to October 4, 1993. Radiological assessments for plant design bases modifications that are performed after October 4, 1993 will be evaluated using the revised 10 CFR 20 dated October 4, 1993.)

A summary of the total isotopic concentration and total FMPC for each component postulated to fail is given in Table 15.7-16.

As indicated by these results the concentrations are well within the 10 CFR 20 effluent concentration limits for unrestricted areas (10 CFR 20, Appendix B). Likewise, the resultant exposures are a small fraction of acceptable limits for this type of event.

15.7.4 FUEL HANDLING ACCIDENT OUTSIDE CONTAINMENT This accident is not reanalyzed as part of the reload analyses as the h1t1,gtt(jcLW ,iiSfOr CONT~irOlLfrU~ (SAE -1:5.1 )

in~~i' Gyele analysis is bounding¢ Radiological exposures were recalculated fCr C3"yle 9 incorporating GE14 fuel resulting in exposures

'zc!! belowt: 1 CFP 100 ' ine1. Th*

7uide&£CLZNSINGBCI* L,-r S 4 6~.3 kerl leoE ecirsac) 4MIO 5 rý-b

, (6*NjkoC goo.-I.)

Revision 11 15.7-21 September, 2001

'I'9 ý- ij, 1V.

3ý-

Enclosure 5 PY-CEI/NRR-2674L Page 13 of 53 15.7.4.1 Identification of Causes and Frequency Classification 15.7.4.1.1 Identification of Causes The fuel handling accident is assumed to occur as a consequence of a failure of the fuel assembly lifting mechanism resulting in the dropping of a raised fuel assembly onto stored fuel bundles.

15.7.4.1.2 Frequency Classification This event has been categorized as a limiting fault.

Revision 6 7

15. -21a March, 1994

Enclosure 5 PY-CEI/NRR-2674L Page 14 of 53 15.7.4.2 Sequence of Events and Systems Operation 15.7.4.2.1 Sequence of Events The most severe fuel handling accident from a radiological release viewpoint is the drop of a.channeled spent fuel bundle onto unchanneled spent fuel in the spent fuel racks in the fuel handling building. The sequence of events which is assumed to occur is as follows:

Approximate Event Elapsed Time

a. Channeled fuel bundle is being handled by a crane over spent fuel pool. Crane motion changes from horizontal to vertical and the fuel grapple releases, dropping the bundle.

The channeled bundle strikes unchanneled bundles in the rack. 0

b. Some rods in both the dropped and struck bundles fail, releasing radioactive gases to the pool water. 0

c. Gases pass from the water to the fuel handling building. 0

d. The fuel handling building ventilation system high radiation alarm alerts plant personnel. <1 Min 15.7.4.2.1.1 Identification of Operator Actions The accident analysis does not assume any operator actions for the mitigation of this event.

Revision 8 15.7-22 Oct. 1996 jrr.Ipr-- ftý V -- v b .

Enclosure 5 PY-CEI/NRR-2674L Page 15 of 53 15.7.4.2.2 Systems Operation Ner-mally, Qpor~atinq p'wnt' and-'-

'="Q -- n'c ;R-,@ ;~'~-Qd to

-function r!though credit- ir tokon cn@1' for the cpzration Rf thze FlhIAES.

Operation of Q plant, reactor protection or ESF systems is not taken into account.

15.7.4.2.3 The Effect of Single Failures and Operator Errors The FHAES is designed to single failure criteria and safety requirements. No C bt[ f*1, TAAK-*.*' FoP- nIH6 FIIAtES.

ReLfer to SOeetien2 1 3 @Qd 9 4 _ 7A-ther--det*_i__F..

peni 15 for Revision 8 15.7-23 Oct. 1996

'QA V

Enclosure 5 PY-CEI/NRR-2674L Page 16 of 53 15.7.4.3 Core and System Performance (6_ rNIlu/IP( r-YCL) 15.7.4.3.1 Mathematical Model The analytical methods and associated assumptions used to evaluate the consequences of this accident are considered to provide a realistic, yet conservative assessment of the consequences for the initial cycle.

The kinetic energy acquired by a falling fuel assembly may be dissipated in one or more impacts.

To estimate the expected number of failed fuel rods in each impact, a conservation of energy approach is used. The fuel assembly is expected to impact on the spent fuel racks at a small angle from the vertical, possibly inducing a bending mode of failure on the fuel rods of the dropped assembly. It is assumed that each- fuel rod resists the imposed bending load by a couple consisting of two equal, opposite concentrated forces. Therefore, fuel rods are expected to absorb little energy prior to failure as a result of bending. Actual bendi-ng tests with concentrated point-loads show that each fuel rod absorbs approximately 1 ft-lb prior to cladding failure. Each rod that fails as a result of gross compression distortion is expected to absorb approximately 250 ft-lb before cladding failure (based on 1 percent uniform plastic deformation of the..rods).

The energy of the dropped assembly is conservatively assumed to be absorbed by only the cladding and other pool structures. Because an unchanneled fuel assembly consists of 76 percent fuel, 19 percent cladding and 5 percent other structural material by weight, the assumption that no energy is absorbed by the fuel material results in considerable conservatism in the mass-energy calculations that follow.

Revision 11 15.7-24 September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 17 of 53 The energy absorption on successive impacts is estimated by considering a plastic impact. Conservation of momentum under a plastic impact shows that the fractional kinetic energy absorbed during impact is:

M1 Ml + M2 where M, is the impacting mass and M2 is the struck mass.

15.7.4.3.2 Input Parameters and Initial Conditions (.ITIALr CCLE)

The assumptions used in the analysis of this accident are listed below:

a. The fuel assembly is dropped from the maximum height allowed by the fuel handling equipment.

b. The entire amount of potential energy, referenced to the top of the spent fuel racks, is available for application to the fuel assemblies involved in the accident. This assumption neglects the dissipation of some of the mechanical energy of the falling fuel assembly in the water above the rack and requires the complete detachment of the assembly from the fuel hoisting equipment. This is only possible if the fuel assembly handle, the fuel grapple or

.the grapple cable breaks.

c. None of the energy associated with the dropped fuel assembly is absorbed by the fuel material (uranium dioxide).

15.7-25

Enclosure 5 PY-CEI/NRR-2674L Page 18 of 53 15.7.4.3.3 Results 14 ITIAL- CYC.LE

a. Energy Available Dropping a fuel assembly onto the spent fuel racks from the maximum assumed height of 10 ft (actual height is 8 ft), results in an impact velocity of 25.4 ft/sec.

The kinetic energy acquired by the falling fuel assembly is less than 8,000 ft-lb and is dissipated in one or more impacts.

b. Energy Loss Per Impact Based on the fuel geometry in the spent fuel rack, two fuel assemblies are struck by the impacting assembly. The fractional energy loss on the first impact is approximately 63 percent.

The second impact is expected to be less direct. The broad side of the dropped assembly impacts approximately 22 more fuel assemblies, so that after the second impact only 88 ft-lb (approximately 2 percent of the original kinetic energy), is available for a third impact. Because a single fuel rod is capable of absorbing 250 ft-lb in compression before cladding failure, it is unlikely that any fuel.,rod will fail on a third impact. In calculating the activity release, however, it is conservatively assumed that one rod fails on the third impact.

If the dropped fuel assembly strikes only one fuel assembly on the first impact, the energy absorption by the fuel rack support structure results in approximately the same energy dissipation on the first impact as in the case where two fuel assemblies are 15.7-26

Enclosure 5 PY-CEI/NRR-2674L Page 19 of 53 struck. The energy relations on the second and third impacts remain approximately the same as in the original case. Thus, the calculated energy dissipation is as follows:

First impact 63 percent Second impact 35 percent Third impact 2 percent (no cladding failures)

C. Fuel Rod Failures (l"Ii*rtuL CLC"CL

1. First Impact Failures The first impact dissipates 0.63 x 8,000 or 5,040 ft-lb of energy. It is assumed that 50 percent of this energy is absorbed by the dropped fuel assembly and that the remaining 50 percent is absorbed by the struck fuel assemblies in rack.

Because the fuel rods of the dropped fuel assembly are susceptible to the bending mode of failure and because 1 ft-lb of energy is sufficient, to cause cladding failure as a result of bending, all 62 rods of the dropped fuel assembly are assumed to fail. Because the 8 tie rods of each struck fuel assembly are more susceptible to bending failure than the other 54 fuel rods, it is assumed that they fail on the first impact. -,Thus 2 x 8 = 16 tie rods (total in 2 assemblies) are assumed to fail.

Because the remaining fuel rods of the struck assemblies are held rigidly in place in the spent fuel racks, they are susceptible only to the compression mode of failure. To cause cladding failure of one fuel rod as a result of compression, 250 ft-lb of energy is required. To cause failure of all the remaining rods of the 2 struck assemblies, 250 x 54 x 2 or 27,000 ft-lb of energy would have to be absorbed in cladding alone. Thus, it is clear that not all the remaining fuel rods 15.7-27 gr U k U1

Enclosure 5 PY-CEI/NRR-2674L Page 20 of 53 of the struck assemblies can fail on the first impact. The number of fuel rod failures caused by compression is computed as follows:

19 0.5 x 5,040 x 19 + 5

= 8 250 Thus, during the first impact, fuel rod failures are as follows:

Dropped assembly 62 rods (bending)

Struck assemblies 16 tie rods (bending)

Struck assemblies 8 rods (compression) 86 failed rods

2. Second Impact Failures Because of the less severe nature of the second impact and the distorted shape of the dropped fuel assembly, it is assumed that in only 2 of the 22 struck assemblies are the tie rods subjected to bending failure. Thus 2 x 8 = 16 tie rods are assumed to fail. The number of fuel rod failures caused by compression on the second impact is computed as follows:

0.35 19 2 x 8,000 x 19 + 5

=5 250 4 f 15.7-28 N 11 O~-&

V

Enclosure 5 PY-CEI/NRR-2674L Page 21 of 53 Thus, during the second impact the fuel rod failures are as follows:

Struck assemblies 16 tie rods (bending)

Struck assemblies 5 rods (compression)._

21 failed rods

3. Total Failures The total number of failed rods resulting from the accident is as follows:

First impact 86 rods Second impact 21 rods Third impact 1 rods 108 total failed rods (initial cycle) 15 151 tetal failod -r-splr:;ro aooumoad for G93,4 fuol.

1wH6 T-DAt FA IEB Nu.--lfae, OF fAIL5 LjagUCLRe?D5' For r*lir 23ý-Vte FUcL HA,/NDULN< /?"Cc.ADL--7Njr -5 /AJ/V 5*I z, .

15.7.4.4 Barrier Performance This failure occurs in the fuel handling building outside the normal barriers (RCPB andcontainment). Therefore, this section is not directly applicable. The transport of fission products to the environment is discussed in the next section.

15.7.4.5 Radiological Consequences

Throe ooiparato radia cnl :nan o oreo nrc:-'--

I --

A r-- '"- 4 .... 4-.

,qXCI Y51 S

The 44 -rt is based on conservative assumptions considered to be acceptable to 'the NRC for the purpose of determining adequacy of the plant design to meet., I C 4T, guideline8.

100 This analysis is "REG,1

,L/.,,Z.Y 6,50 .1,93 DO5 (-RiTCZ iA.

Revision 11 15.7-29 September, 2001

'nV

Enclosure 5 PY-CEI/NRR-2674L Page 22 of 53 referred to as the "design basis analysis," and is based on a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months radioactive decay period of the fuel.

F *b.

The zznd &n&lyzi... 9 15aae4 an eena4:dered - re":-

VA- V

.r-_a

-- iji ZiUIiJcgucnZcz. .JfllZ

-analysis ic referrod to as the "realictic analysis," and is: based

,on a94 h-a-. .... ct*vo decoa p.ri4:vd ef the-ul.

7"

- I1Thj'- IIKJLI '31 UIIZUlLO app

-idose consequences ";hon deroyihts ficured being hanadled oftsi sevenI dny of r~dioocti-' Q dgga' has seeaurrccl. ThIc4 eeft54:6cid C.ip.

i i

o. .oth. b .r.. ti deS19n ba..Ic ;. d ro.... . Qcurc t.r...... When comparing a fuel handling accident inside containment with a fuel handling accident in the Fuel Handling Building, the inside containment event would be bounding due to higher kinetic energy and the greater number of fuel pins damaged. Poth an.ly... make 1 AbE . ..

Ihe equivalen assumption-that the activity which escapes from the pool is released immediately and directly to the environment.

Thus, for this analysis, refer to Section 15.7.6.4 on the bounding analyses for the fuel handling accident inside containment.

Fr all afnalyse, Ofhe fission product inventory in the fuel rods assumed to be damaged is based on operation at 3,833 MWt.

15 7t5.l Doirign; Paris A'nalysis Asscuming 24 ou R.o;dio~acti':c Dcooy of tho Fuel h455u."/PVofqc, R~to He-$'oDg CcA%~JE~D //

The design basis analysis is based on NR.C -tcndrd o.vic;; Plan 1.7.4 az"4- NRC Regulatory Guide 1-.2S-.-. ua__iQcr- -f p. . d ;;4 the evalluation are presented in Table 15.7 17.

Revision 11 15.7-30 September, 2001

Enclosure 5 PY-CEIINRR-2674L Page 23 of 53 15.7.4.5.1.1 Fission Product Release from Fuel "RE refo ,a ,1:5.

1 CI.- I following conditions are assumed applicable for this event:

a. The fu rod gap activity is assumed to consist of 10% of the total halogen an oble gas activity in the rods at the time of the accident, excep or Kr-85 which is assumed to be 30%.

b. Because of the negligible articulate activity available for release from the fuel plena, n e of the solid fission products are assumed to be available for release.

c. It is conservatively assumed that 100 percen f the noble gas plenum activity and 1.0 percent of the halogen ple m activity in the damaged fuel rods is released from the spent fuel 1 to the fuel handling building atmosphere.

Revision 8 15.7-30a Oct. 1996

Enclosure 5 PY-CEI/NRR-2674L Page 24 of 53

-Based en the above conditions the activity airborne in the fue-l--hand144ig b..i..in. + pis....t.d in Table 15.7 18.

15.7.4.5.1.2 Fission Product Transport to the Environment In accordance with the criteria_ pro=nee i Re-ulao-r-r rz-ld-*, 1 9r n-d i.52 it is ~asaiode that the airbei..e etivity 4in the fuol! handling

-buldig

.(..n Table 15.7-12) izrc~caced to the .n.ionment eve-r a 2 houL

-period za a C1 percent-iode fficien FHiAES The total activity released to the enuronm~en 4i prme1-H in Tad-ble 15 7-19.

15.7.4.5.1.3 Results lBO,.-blMG FJO.L fi 4f4L,'C ACC~Ibtd-T The calculated exposures for the doc!gn; ba..r. na!1,,cis are presented in

.. 35" Table 15.7-%"0 and are-4,-l- within the guiAdoinoc o-f 10 C=P. 190. Dos*.C CgjT-E*Qj OF 1?E-18b~ Ii3.

,5.

45-. 2 .ea.4! stir- Anailysis, AssuFRiong L24 Me.urki,- adieaective BeDat- f the Fuel The realistic analysis is based on a realistic but sti conservative assessment of this accident. Specific values of ameters used in the evaluation are presented in Table 15.7-17.

15.7.4.5.2.1 .Iission Product ease from Fuel Fission product release e mates for the fuel handling accident are based on the followi assumptions:

a. An ave ge of 1.8 percent of the noble gas activity and 0 . ercent of the halogen activity is in the fuel rod plena and available for release. This assumption is based on fission product release data from defective fuel experiments.(3)

Revision 8 15.7-31 Oct. 1996 I,,VFr

Enclosure 5 PY-CEI/NRR-2674L Page 25 of 53

b. Because of the negligible particulate activity available for release from the fuel plena, none of the solid fission pro cts are assumed to be released.

c. It is conservatively assumed that 100 percent of the oble gas plenum activity and 1.0 percent of the halogen ple m activity in the damaged fuel rods is released from the spent uel pool to the fuel handling building atmosphere.

Based on the above conditions the activity airb ne in the fuel handling building is presented in Table 15.7-21.

15.7.4.5.2.2 Fission Product Transpo, to the Environment It is conservatively assumed that all activity released to the fuel handling building is released to th environment in the first two hours after the accident via a 95 perce t iodine efficient FHAES. Based on these assumptions, the activity released to the environment is shown in Table 15.7-22.

15.7.4.5.2.3 Results The calculated expos es for the realistic analysis are presented in Table 15.7-23 and e well below the guidelines set forth in 10 CFR 100.

15.7.4.5.3 Design Basis and Realistie Analyses Assuming Seven Day Radioactive Decay -

The radi ogical releases from a fuel handling accident inside contai ent (based on -a seven day decay) are larger than those from a fuel andling accident outside containment (based on a seven day decay).

Th efore, refer to Section 15.7.6-4.3 for details related to the unding analyses for a fuel handling accident inside containment.

Revision 8 15.7-32 Oct. 1996

?;aft;ýN

Enclosure 5 PY-CEI/NRR-2674L Page 26 of 53 15.7.5 SPENT FUEL CASK DROP ACCIDENTS This accident is not affected by the reload analysis.

15.7.5.1 Cask Drop from Transport Vehicle In the unlikely event that the fuel cask falls from the transport vehicle, the maximum height which the cask will drop should be in general less than 10 ft. Since the cask is designed to withstand a 30 ft drop onto a non-yielding surface without failure, the fall from the transport vehicle will cause no failure of the cask.

15.7.5.2 Cask Drop from Crane The Mark III containment design includes a separate fuel handling building. The spent fuel storage pools in'this building are arranged so that the overhead crane which handles the cask cannot possibly move the cask above the spent fuel storage pool. This precludes the possibility of the cask falling on the stored spent fuel bundles. Also, the pools are arranged so that a rupture of the cask loading pool floor will not drain water from the spent fuel storage pool. The cask loading area design and operating procedures are specifically formulated so that a cask drop will not result in failure of the cask.

15.7.6 FUEL HANDLING ACCIDENT INSIDE CONTAINMENT The radiological exposures were recalculated foer.ye4le-9-incorporating GEl4 Fuel and power uprate analysis resulting in exposures 'el' 1zzlo The analysis in Section 15.7.6.4 asumes a decay time prior to the accident occurrin The value forms the definition of "recently irradiated fuel" as identified in the Technical

"--12_4 ours.f

. 7,D =e^- Rervvisionn =II 15.7-33 September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 27 of 53 Sitications. This analysis is reviewed each cycle to verify that the 9;' assumption is valid. As a result of this review th definition of "recently irradiated fuel" may chang * . "a*uz cthcr 15.7.6.1 Identification of Causes and Frequency Classification 15 .7.6.1.1 Identification of Causes Various mechanisms for fuel failure during refueling have been investigated. Procedural controls, backed up by the refueling interlocks, impose restrictions on the movement of refueling equipment and control rods, to prevent an inadvertent criticality during refueling operations. In addition, the reactor protection system is able to initiate a reactor scram in time to prevent fuel damage for errors or malfunctions occurring during planned Revision 11 15.7-33a September, 2001

.lot ~L

Enclosure 5 PY-CEI/NRR-2674L Page 28 of 53 criticality tests with the reactor vessel head off. It is concluded that the accident that results in the release of significant quantities of fission products during this mode of operation with the greatest analyzed radiological consequences is one resulting from the accidental dropping of a fuel bundle onto the top of the core.

The movement of non-fuel items over irradiated fuel inside containment will be administratively controlled such that the radiological consequences associated with their accidental dropping with or without primary or secondary containment will remain bounded by that of a dropped fuel bundle.

15.7.6.1.2 Frequency Classification This event has been categorized as a limiting fault.

15.7.6.2 Sequence of Events and Systems Operation 15.7.6.2.1 Sequence of Events The sequence of events which is assumed to occur is as follows:

Approximate Event Elapsed Time

a. Channeled fuel bundle being removed from f -JoCLW* reactor vessel by'crane. Fuel bundle is dropped from maximum height allowed by the refueling equipment. Fuel bundle strikes core. 0

b. Some rods in both dropped and struck bundles fail releasing radioactive gases to pool water. 0

c. Gases pass from water immediately to building. 0

d. Containment vessel and drywell purge ventilation system isolates due to high radiation signal. Nor C&Enrri_ im 1-Wi 7kl 20 sec Revision 9 15.7-34 April, 1998

Enclosure 5 PY-CEI/NRR-2674L Page 29 of 53 15.7.6.2.1.1 Identification of Operator Actions The accident analysis does not assume any operator actions for the mitigation of this event.

..... .... Il-

, . .... ...... ° Revision 8 15.7-35 Oct. 1996

Enclosure 5 PY-CEIUNRR-2674L Page 30 of 53 15.7.6.3 Core and System Performance The methods used for this evaluation are the same as those presented in r,,

Section 15.7.4.3/ o- THE IA,1ITb2'7... VyLL.c*' 7- ANYi cI_&bUrs /*/"6;,6-(-.

p~ Z'seDON G 6 tf1rllbb. cLY Fo/zc;4 /Y Xb .4 7;'vc:

-qmAM)* R-~.e 15.7.6.4 Radiological Consegquences CAI5ES Three separate radiological analyses are provided for this accident:

a. The first is based on conservative assumptions considered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet 40 CFR. 100 guidelinec. This analysis is

.. r....d to a: the "deij. . .....baso lari ", e.nd i based on a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months radioactive decay period of the fuel. TH1s5 IS Cc*ICP-Z1T TYE 8 C';C Ami) TAK6 NC Ce~bt- F01Z C9IECP6 TAFET-y Or~A"WeES, LCLJTirr.4 (rfo FT '04

&~I

b. 4 he *eeR?**I*4 *n=op4*

,,,.,* considere to prevido a 4..- . .. ..- ...... e t ..

.. ra zllelegieal

.. e

. .zxceý. ThLi S Qn@ýZSS 4 '-gf'-rd to 2r. the reliti analy'cic," and isbae on *_ '4_ hour rndic~cti-e decay poriod of the fue!.

id5C1ZT 1:1 C. -71hc; thi rd an;a!yr.; r '4 r. w'-Mn'+- nfQf- +4-n Mnr rn-d-,- R-J desc eenzzcquenees w'hen i:rradiated fuel ft beaing 1har~dled af-teL 5et-en r-n%'z of -. a.linnr-+-iir, '-*," ' '.. r.d..,h the issumption that the r*lease imudieatcly And directly r .lea..t h environment. Thi anal 1,sls conslioe-z both nonygrir-ti4e design basis and rcalzstic

~fcp *s For all na-yz, the fission product inventory in the fuel rods assumed to be damaged is based on operation at 3,833 MWt. Specific values for parameters used I" the U t t,, analyses are provided in Table 15.7-.

S2-pcEl -Jg f- +-b- thir-d analys'is are pre;4idee in. Table 1S.7? 32.

Revision 11 15.7-36 September, 2001 At V

Enclosure 5 PY-CEI/NRR-2674L Page 31 of 53 Insert #1 (15.7.6.4.b)

The second case is identical to the first case but was performed to determine the effect of control room isolation and fresh air intake. The control room dose was calculated assuming that once the available activity is introduced into the control room, the fresh air intake was isolated without any additional inleakage. At two hours, outside air purge is assumed to initiate and continue for the remainder of the 30-day dose analysis.

Insert #2 (15.7.6.4.c)

The third case is also identical to the first case but was performed to determine the effect of control room isolation and emergency recirculation filtering. The control room dose was calculated assuming that once the available activity is introduced into the control room, the fresh air intake was isolated without any additional inleakage. At two hours, the control room emergency recirculation system was assumed to initiate and continue for the remainder of the 30-day dose analysis. A filtration efficiency for the charcoal of 50% was assumed in order to be consistent with Table 15.6-14.

Insert #2A Note: The second and third cases were performed to examine the flexibility the Control Room operators have in using ventilation, to ensure that there were no dose outliers. The results show that even if the operators take 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to initiate an action, (i.e., re-initiate normal intake or utilize recirculation filtration) the doses remain below the licensing basis limits.

Insert #2B An additional scenario was reviewed in the event a fuel bundle was dropped when travelling through the refueling shield inside containment. This reduces the amount of water coverage resulting in a lower Decontamination Factor. The resultant dose is less than the cases described above when considering that the damage is limited to the number of rods in one fuel bundle.

j.S; i x

Enclosure 5 PY-CEI/NRR-2674L Page 32 of 53 15.7.6.4.1 Design Basis Analysis Assuming 24 Hour Radioactive Decay of the Fuel

a. Fission Product Release from Fuel I,vL~X s g- 3 I -- ----- 3 C fls- hn-l ---- n n---- rjL~

ZQ2111 f- nf n

. -M d. -+ Q -;R@ýAGEI P A P ýwý a I r;1I -" -I -- A- .'r " I -4 "Ww Ifi C+t'S "~

tnot ~ I Revision 11 15.7-36a September, 2001

Enclosure 5 PY-CEI/NRR-2674L Insert 3 (15.7.6.4.1 .a) Page 33 of 53 The fission product activity released from the fuel damaged as a result of a fuel handling accident is calculated using the methods below:

I) The fuel rod gap activity is assumed to consist of 5% of the total halogen and noble gas activity in the rods at the time of the accident, except for KR-85, which is assumed to be 10% and 1-131 which is assumed to be 8%.

2) Twelve percent of the alkali metals are available for release but are retained in the water.

A total of 151 fuel rods fail as a result of the accident given a core loaded with GE14 fuel.

TTý

Enclosure 5 PY-CEI/NRR-2674L Page 34 of 53

b. Fission Product Activity. i4rbc'rr in thg R c B"ilding. R The following assumptions and initial conditions are used in calculating the fission product activity released to the r bui din-. N VI (Z(,WS ELEMENTAL.

1. The iodine gap inventory is composed of .Inorganic species (99.W* percent) and organic species (0.-* percent).

2. The minimum water depth between the top of the damaged fuel rods and the containment pool surface is 23 feet.

F-LEMENTrAL

3. The pool decontamination factors for the i and organic species of iodine are . and 1, respectively, giving an overall effective decontamination factor of .-? (i.e.,

S00 q79.gE 9 percent of the total iodine released from the damaged rods is retained by the pool water). This difference in decontamination factors for *---g,-n- and organic iodine species results in the iodine above the fuel pool being 7EL.Er1tNTA L p ni composed of 1-5 percent ineorgani and-2-& percent organic species.

4. The retention of noble gases in the pool is negligible (i.e.,

decontamination factor of 1). ?4r(T'IU.LULATC- ,4,RE

5. The effects of plateout and fallout are neglected.

Based on these assumptions, AJ the activity released from the pool to the .reactor building is listed in Table 15.7-Z2, IV. li/C. kc5iVITiE.o (,j7*//# r/71 FALc-j) Z'oDS IRE 'k5504-76D 7b /IAVS &CN

7. /RLL Aarkry PdLE-ýi) FiecV-X MEt POOL rd VIC-- Co4RfkWMeArr /S 15 moor Crizi&7Lb) 15.7-37

Enclosure 5 PY-CEI/NRR-2674L Page 35 of 53 sig onnirono n Product Pezecis to The following assumptions and initial conditions are used in c culating the fission products released to the environs:

1. T containment vessel and drywell purge system are in oper ion at the time the accident occurs. These systems are describ d in Section 9.4. It is conservatively assumed that isolation f the containment vessel and drywell purge system will occur 2 seconds after the release of fission product activity from t containment pool due to a high radiation signal in this sys em.

2. All activity during the irst 20 seconds after the accident is assumed to be released to e environs via the containment vessel and drywell purge exha st system filter as a "puff" release.

3. No credit is taken for filtering iodi e during the first 20 seconds after the accident.

4. The activity remaining in containment is rele sed to the environs (via the annulus exhaust gas treatment ystem).

Based on these assumptions, the activity released to the Revision 11 15.7-38 September, 2001

Enclosure 5 PY-CEIINRR-2674L Page 36 of 53

d. Results IoTAL. C-FZ:LT1YC ZDOA& CGQ~IJIALE1I (raoc)

Based on these assumptions, the intogratod 41hcle body dcscs and 4

i" -g'-.-Qe

- -N . . . . .Q thyroid D NXTJ d*o* L Rat e ZOthe exclusion boundary " - - . a. low population zonehare summarized in Table 15.7-24. The doses at these distances are well bele; thOc 19 zenV 100 liir,-,i- LIT14t, 'THE latclcNId 1-5 7 -6 '12 P0@1i4zt-i, nn lyýg ~cinn 9A 1 4,0,r Pn~inn-1-i~,~ Mr'a o-f the Fuel4

a. sion Product Release from uel The fi ion product activity released from the fuel damaged as a result of fuel handling accident is calculated using the methods outlined in ection 15.7.4.5.2.1. As a result of this accident, 124 fuel rods e assumed to fail for the initial cycle. A total of 151 fuel rods e assumed to fail for a core loaded with GE14 fuel.

b. Fission Product Activity leased to Containment The following assumptions and i itial conditions are used in calculating the fission products leased to the containment:

1. The fission product activity relea ed to the containment will be inversely proportional to the remo al efficiency of the water in the upper containment pool. B ause water has a negligible effect on removal of the noble ses, noble gases are assumed to be instantaneously released fr m the pool to the containment.

2. The iodine activity in the fuel rod plena is compose of inorganic species (99.75 percent) and organic species (0.25 percent).

Revision 11 15.7-39^ P*+

P-' September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 37 of 53

-nas*. ul zs s.....tionn, bte, actitvity releasede te t~he conjtainF.fficntt pool and the environment is given in Table 15.7-33 and Table 15.7-34 respectively.

c. Results Based on these assumptions, the integr d whole body doses and integrated thyroid doses at the exclusion bou ry, low population zone, and Control Room are suimmarized in Table 15.7- The doses at these distances are well below the 10CFRIOO limits, and*ti the General Design Criteria 19 limit for the Control Room.

15.

7.7 REFERENCES

FOR SECTION 15.7

1. Nguyen, D., "Realistic Accident Analysis - The RELAC Code,"

October 1977, (NEDO-21142).

2. Bunch, F. D., "Dose to Various Body Organs from Inhalation or Ingestion of Soluble Radionuclides," IDO-12054, AEC Research and Development Report, TID-4500, August 1966.

3. N. R. Horton, W. A. Williams, J. W. Holtzclaw, "Analytical Methods for Evaluating the Radiological Aspects of the General Electric Boiling Water-Reactor," March 1969, (APED 5756).

4. General Electric Company "General Electric Standard Application for Reactor Fuel," including the United States Supplement, S' NEDE-24011-P-A and NEDE-P-A-US (latest approved revision).

5. Reg. Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the purpose of evaluating Compliance with 10CFR Part 50, Appendix I, Rev. 1, October 1977.

6. "Z1E(LATo,7jY LUID15 9 1YQ 1.183,"A-7, V4A1/*_' 7bM.LaC/.4L So60C.Z 7Ci7;o Fog L"VALUvA-0 DT.si'A, 14s/s ,AccS1.,T,%,7-/r ,IT LCAltC, Poi7,6Z- ,Z ,

JuLY 2o-00 Revision 10 15.7-41b October, 1999

Enclosure 5 TABLE 15.7-17 PY-CEI/NRR-2674L Page 38 of 53 FUFLJ, ANDLING ACCIDENT OUTSIDE C _T_*Tt'MZT rAR',.ETEPh TABULATED FOP PcSTULATED ACCTGD'M ANALYSIS Realis c Design Basis Bas's'"

Assumptions Ass ptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level 3,833 MWt 3,833 MWt B. Radial peaking factor 1.7 1.6 C. Fuel damage (GEl4) 151 ods 151 rods

,D. Release of activity by ction Section nuclide 15.7.4.5.1.1 15.7.4.5.2.1 E. Iodine fractions (1) Organic 0.0025 0.0025 (2) Elemental 0.9975 0.9975 (3) Particulate 0 0 II. Data and assumptions used to estimate activity leased A. Fuel handl? g building leak rate 100%/2 hr 100%/2 hr B. -Adsorp ion and filtration effic encies

) Organic iodine 95% 95%

(2) Elemental iodine 95% 95%

(3) Particulate iodine 95% 95%

C. All other pertinent data None None

-.and assumptions Revision 11 15.7-66 September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 39 of 53 TABLE 15.7-32 FUEL HANDLING ACCIDENT INSIDE CONTAINMENT PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSIS (ASSUMING --- B- RADIOACTIVE DECAY) I V' Hu..-t Design Basis -AQ aIi't! ic Care Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level 3,833 MWt 3, 433 MWt B. Bur-R- XJJSLV741 Se tion 1517.6.4 C. Fuel damage (GE14) 151 rods 15 rods D. Release of activity to Section Se tion containment pool by 15.7.6.4ZA 15 7.6.4.3 nuclide, per failed rod E. Iodine fractions - organic

.8 .001n

. Oi 25

.9945 .9175 f7 IR~bAL. ?LAKWCh ;Ae7oRL 2.0 II. Data and assumptions used to estimate activity released A. Primary containment leak Instantaneous In. [tantaneous rate total release tot al release of all activity of all activity leaving pool to lez ving pool to environment eni 1ironment B. Secondary containment leak N/A N1/

rate C. Isolation valve closure N/A N/1 times D. ,Filtration efficiencies N/A N/4 E. Recirculation systems N/A N/IA parameters (flow rates vs.

.time, mixing factor, etc.)

Revision 11 15.7-83 September, 2001

Enclosure 5 Insert #6 (Table 15.7-32, Item I.B) PY-CEI/NRR-2674L Page 40 of 53 B. Bum-up -Peak rod exposure is less than 62,OOOMWD/MT

-Between 54,000 and 62,OOOMWD/MT, the maximum linear heat generation rate does not exceed 6.3 KW/Ft peak rod average exposure.

Enclosure 5 PY-CEI/NRR-2674L Page 41 of 53 TABLE 15.7-32 (Continued)

Design Basis Assumptions Assumfption~s F. Containment spray N/A parameters (flow rate, drop size, etc.)

G. Containment volumes N/A H. Pool Removal (iodine Žee-2oo +/-988 partition factor)

-I. All other pertinent data Section Seet7ei

and assumptions .15.7.6 J. Activity released to Table 15.7-i% Table 15.7 53 environment K. Control Room Parameters See below N./LA 3

1. Volume (ft 3 ) 367,070 ft

2. Dzsigr.n Fla: ý.efcfi) 6, 600 (Instantanetous iNAi4 ni -t mzing + 109 of dc..i g. flLw) ..

3 Filter Efficiencv 0 Ne !&iltz..d re-+/-Laulat on III. Dispersion Data A. Boundary and LPZ 863/4002 83 S,' 40902 distances (m)

B. Offsite X/Q's (Corresponding 4. 3E-4/4 . 8E-5 4,9E; 4,149-5 to 7 year meteorological data for 0-2 hr for SB/0-8 hr for LPZ)

C. Control Room X/Q's {c-3tr) 3,5r--

'qa,ge I,*,l nl f1I n 0 -- , , - A (2) 8 24 hrs 2.1E-4

-;3) 1 4 d .ys +/-.+/-E-4 C4) 4 3*.. s .5E-5 IV. Dose Data A. Method of dose Section Sct ion calculation 15.0.3.5 15.0.3.5 Revision 11 15.7-84 September, 2001 Ct,,: *-Y"

Enclosure 5 PY-CEI/NRR-2674L Insert #4 (Table 15.7-32 Item II.K) Page 42 of 53

2. Design Flow (cfm) Intake Exhaust (Ali- purc6) Emergency Recirculation Case I 6600 (Normal +10%) 5400 (Normal -10%) 0 Case 2 6600 5400 0 Isolated after activity Started after 2 hrs introduced into Control Room Case 3 6600 0 27,000 (Normal -10%)

Isolated after activity Started after 2 hrs introduced into Filter efficiency = 50%

Control Room

Enclosure 5 PY-CEI/NRR-2674L Page 43 of 53 TABLE 15.7-32 (Continued)

Design Basis p.c1istiz Czae Assumptions B.* Dose conversion t5.5.5.5 1:5. &! . 5 I assumptions

1. Dose conversion assumptions (Offsite)

2. Dose Conversion Assumptions (Control Room) ;nternaticna-1 Commrissizn zf CoRnversion Factors (r-em/Ci)

I- 21 1.10E6 TI-132 6.JOE3 I -133 -1.0E5 I1313 1.10E3 T 13* 1 lnlfl C. Peak activity N/A concentrations in containment D. Doses Table 15.7-35 Table 15.7" 35 Revision 11

15. 7-84a September, 2001

Encrosure 5 PEYCEI/NRR-22674L Page 44 of 53 TABLE 15.7-33 FUEL HANDLING ACCIDENT INSIDE CONTAINMENT ASSUMING 7 DAY RADIOLOGICAL DECAY OF THE FUEL DESIGN BASIS AND REALISTIC ANALYSIS ACTIVITY RELEASED TO THE CONTAINMENT POOL (CURIES)*

Design Basis Realis c Source Terms Sourc Terms Isotope Activity A ivit 1-131 2.1E+4 1.7E+3 1-132 **

1-133 2.OE+2 5.1E+0 1-134 ** **

1-135 2.OE-3 2.9E-5 Kr-83m ** **

Kr-85m 7.8E- 4.3E-9 Kr-85 9. +2 6.2E+2 Kr-87 ** **

Kr-88 ** **

Kr-89 ** **

Xe-131m 1.9E+2 4.8E+I Xe-133m 1.4E+3 1.4E+2 Xe-133 2.6E+4 4.3E+3 Xe-135m **

Xe-135 2.8E-1 5.8E-2 Xe-137 /* **

Xe-138 **

GE14 Fu @ 3,833 MWt Negligible levels of activity Revision 11 15.7-85 September, 2001 n' rS3. T4

Enclosure 5 PY-CEI/NRR-2674L

'4ffr/* Page 45 of 53 TABLE 15.7-34 FUEL ANDLING ACCIDENT INSIDE CONTAINMENT ASSUMING'*\*-*A RADIOLOGICAL DECAY OF THE FUEL DESIGN BASIS .N, hisTie

....... ANALYSIS ACTIVITY RELEASED TO THE ENVIRONMENT (CURIES)*

Design Basis Real stic Source Terms Sourcý Terms Isotope Activity Acta vity IZ-,?f &.27E-0 )'yI-131 2.OE+O E+l 1-133 "V:r-,

1-134 E--2 II 1-135

,.&S I 2 .t'OE---5 2.9 E-7 B.,'e "=ý 2 2ý Kr-83m Kr-85 9-~-2-E-+ 2 6.2 9+2 11-9 Kr-85m -I 7 .-ft*- 8 4.3 Kr-87 Kr-88

ý- on r4.-

ir 212.0. 2. 0eP Xe-131m 1 .-9£--'2 4.8] +1 I Xe-133m 1.41 +2 I Xe-133 4.3E+3 Xe-135m to, e'0E 1-,e+-

2 .-8£-E=

I Xe-135 5.8E-2

-- ~ I. n"

-- I n n

GE114 Fuel @ 3,833 MWt I Ncgl:

!-!big 1Q-Q;S of i %er c+/-.i- iti AwJ C.;7LcrfLY, TV(e* ?CnL.

£5E- I R13 8?1 0 &4 RB 86 Revision 11 15-7-86 September, 2001

Enclosure 5 PY-CEI/NRR-2674L TABLE 15.7-35 Page 46 of 53 "FUEL HANDLING ACCIDENT INSIDE CONTAINMENT

ýU~iING t*7-- RADIOLOGICAL DECAY OF THE FUEL DESIGN BASIS ANPD .flEAITTIC ANALYSIS RADIOLOGICAL EFFECTS BESiati DASIS SOUjReE TER Whole Body Inhalation eta Skin Dose (rem) Dose (rem) Dose (rem)

Exclusion Area 1.17E-1 4.71E+l N/A (863 Meters)

Low Population Zone 1.40E-2 5.2 N/A (4,002 Meters)

Control Room 1.27E-2 2.76E+1 3.37E-1 REALISTIC SO CE TERMS W le Body Inhalation ose (rem) Dose (rem)

Exclusion Area 1.87E-2 3.85 (863 Meters)

Low Population one 2.10E-3 0.43 (4,002 Meters Control R om N/A N/A NOTE- The above radiological effects have been updated to reflect the scaled increases associated with Power Uprate to 3,758 MWt.

Revision 11 15.7-87 September, 2001

Enclosure 5 Insert #5 ( Table 15.7-35) PY-CEI/NRR-2674L Page 47 of 53 DESIGN BASIS SOURCE TERM TEDE Dose (rem) Licensing Limit (rem)

Exclusion Area 1.44 6.3 (863 Meters)

Low Population Zone 0.161 6.3 (4002 Meters)

Control Room:

Case 1 1.03 5 Case 2 2.81 5 Case 3 2.97 5

I Enclosure 5 PY-CEI/NRR-2674L Page 48 of 53 TABLE 15A.2-4 UNACCEPTABLE CONSEQUENCES CRITERIA PLANT EVENT CATEGORY: DESIGN BASIS ACCIDENTS Unacceptable Consequences 4-1 Radioactive material release exceeding the guideline values of 10 CFR 100 (for th d S-'*-

g z= P"oq.q'OCA analysis, the licensing basis offsite dose limit is 25 rem TED 4-2(l) Failure of the fuel barrier as a result of excee ng 111 r:5 "Sir mechanical or thermal limits. f s*,

4-3 Nuclear system stresses exceeding that allowed for accidents by applicable industry codes.

4-4 Containment stresses exceeding that allowed for accidents by applicable industry codes when containment is required.

4-5 Overexposure to radiation of plant main control room personnel.

NOTE:

1. Failure of the fuel barrier includes fuel cladding fragmentation (loss-of-coolant accident) and excessive fuel enthalpy (control rod drop accident).

Revision 11 15A.2-16 September, 2001 k'ý kl'ý u~wF,

Enclosure 5 PY-CEI/NRR-2674L Page 49 of 53 Figure 15A.6-35 presents the different protection sequences for the control rod drop accident. As shown in Figure 15A.6-35, the reactor is automatically scrammed and isolated. For all design basis cases, the neutron monitoring, reactor protection and control rod drive systems will provide a scram from high neutron..flux.

After the reactor has been scrammed, core cooling is accomplished by either the RCIC or the HPCS or the normal feedwater system.

b. Event 36 - Fuel Handling Accident Outside Containment Because a fuel-handling accident can potentially occur any time when fuel assemblies are being manipulated in the fuel handling building, this accident is considered in all operating states.

Considerations include mechanical fuel damage caused by drop impact and a subsequent release of fission products. The protection sequences pertinent to this accident are shown in Figure 15A.6-36.

It is important to note that the systems, structures, and cmponents described within Figure Ionlycredited for the fue ad' cdetta nolves dropping of a recently lae udle onto other recently irradiated bundles,.j Revision 10 15A.6-36 October, 1999 IS

Enclosure 5 PY-CEI/NRR-2674L Page 50 of 53

h. Event 44 - Postulated Radioactive Releases Due to Liquid Containing Tank Failures The postulated events that could cause release of the radioactive inventory of a waste tank include a tank failure and/or an operator error. The possibility of a tank failure and consequent release rates receives primary consideration in system and component design. The tanks and piping to the first isolation valve are Safety Class 3, Quality Group C components and are designed to Seismic Category I. The concentrator waste tanks are designed to operate at atmospheric pressure and 200*F maximum temperature so the possibility of failure is considered small. A liquid radwaste release caused by operator error is also considered a remote possibility. Operating techniques and administrative procedures emphasize detailed system and equipment operating instruction. A positive action interlock system is provided to prevent inadvertent opening of a drain valve. Should a release of liquid radioactive wastes occur, floor drain sump pumps in the floor of the radwaste building will receive a high water level alarm, activate automatically and remove the spilled liquid to a contained storage tank.

The protective sequences for this event are provided in Figure 15A.6-42. a

i. Event 45 - Fuel Handling Accident Inside Containment -4 A fuel-handling accident inside containment can only occur when fuel assemblies are being manipulated over the reactor core.

Therefore, this accident is only considered in operating State A.

Considerations include mechanical fuel damage caused by drop impact and a subsequent release of fission products. The protection sequences pertinent to this accident are shown in Figure 15A.6-43.

Revision 11 15A.6-41 September, 2001

Enclosure 5 PY-CEI/NRR-2674L Page 51 of 53 taanntoýnoethat the systems, structures, and cop ne t dds ibedd w*

wl eeý5k, -43 are only credited for Sthe fuel handling aaccident that involves rcnl ated bundle onto other recently irradiated bundes dia Revision 10 15A. 6-41a October, 1999 f1I'r! i ztý IF

I Enclosure 5 PY-CEI/NRR-2674L Page 52 of 53 (Rev. 10 10/99)

Protective Sequences for Fuel Handling Accident Outside Containment Figure 15A.6-36

'ur " I

Enclosure 5 PY-CEI/NRR-2674L Page 53 of 53 (Rev. 10 10/99)

Protective Sequences for Fuel Handling Accident Inside Containment Figure 15A.6-43 N

h3v

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