ML020560226

Part 1 of 2, Perry Nuclear, License Amendment Request Pursuant to 10CFR50.90: Alternate Source Term Update of Previous License Amendment Which Adopted Shutdown Safety Administrative Controls During Fuel Movement

ML020560226
Person / Time
Site:	Perry
Issue date:	02/11/2002
From:	Campbell G FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-nr, PY-CEI/NRR-2609L, RG-1.183
Download: ML020560226 (147)

Text

FENOC 1ý Perry Nuclear Power Plant 10 Center Road FirstEnergyNuclear Operating Company Perry, Ohio 44081 Guy G. Campbell 440-280-5224 Vice President - Nuclear Fax: 440-280-8029 February 11, 2002 PY-CEI/NRR-2609L 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 a Previous License Amendment which Adopted Shutdown Safety Administrative Controls During Fuel Movement.

Ladies and Gentlemen:

A license amendment is requested to the Technical Specifications for the Perry Nuclear Power Plant (PNPP). The requested change utilizes Alternative Source Term radiological calculations to update and expand upon License Amendments 102 and 103, issued in March 1999.

Amendment 102 introduced the concept of utilizing Shutdown Safety administrative controls in place of Technical Specifications to provide defense in depth during fuel handling activities.

Amendment 103 was a pilot plant application of an Alternative Source Term for a loss of coolant accident. New fuel handling accident calculations performed for PNPP will no longer credit OPERABILITY of filtration systems or the Containment/Fuel Handling buildings. As a result, the 10 CFR 50.36 criteria that specify the items that must remain in Technical Specifications no longer apply to these buildings and filtration systems during fuel handling activities, since they are no longer part of the "primary success path" for the Fuel Handling Accident. Shutdown Safety administrative controls over filtration system availability and building closure will be used in place of Technical Specifications to -provide defense in depth during fuel handling activities.

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 December 14, 2002, 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. If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager - Regulatory Affairs, at (440) 280-5305.

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. Proposed Technical Specification Changes (mark-up)

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

6. Information copy of Technical Specification Table of Contents and Bases (mark-up)

7. Information copy of proposed USAR changes (mark-up) cc: NRC Project Manager NRC Resident Inspector NRC Region III State of Ohio

Enclosure 1 PY-CEI/NRR-2609L Page 1 of 1 I, Guy G. Campbell, 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.

! Guy G. Campbell, Subscribed to and affirmed before me, the / '4Lday of 261?2d JANE E. MOTT iry Public, State of Ohio My Comre lewon Expires Feb. 20, 2005 (RQcordd in Lake County)

Enclosure 2 PY-CEI/NRR-2609L Page 1 of 28 Summary The changes proposed to the Perry Nuclear Power Plant (PNPP) Technical Specifications are 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.

Although it is not physically possible for the process of plant cooldown and vessel disassembly to be completed to the point of handling fuel within only one day (three days is the current best estimate),

the analysis assumes the event occurs after only one day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 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, even without credit for OPERABILITY of filtration systems or the Containment/Fuel Handling Building.

Since the filtration systems and Containment/Fuel Handling Buildings are no longer credited as part of the "primary success path" for the Fuel Handling Accident, the 10 CFR 50.36 criteria for including items within Technical Specifications are no longer applicable to these filtration systems and buildings during fuel handling activities. The Technical Specification requirements on OPERABILITY of various buildings and filtration systems during fuel handling are therefore being replaced with Shutdown Safety administrative controls, which maintain closure controls for buildings and availability (rather than OPERABILITY) for ventilation systems. In keeping with this concept, a commitment is included to the recently published Nuclear Energy Institute (NEI) guidance on assessing systems removed from service during the handling of irradiated fuel assemblies.

Description of the Changes As a result of the new design basis calculation, a number of Specifications are deleted or revised.

Since the only Applicability statement in the Fuel Handling Building (FHB) Specification and the FHB Ventilation Exhaust System Specification was "During movement of recently irradiated fuel in the FHB", these two Specifications would be removed from the Technical Specifications in their entirety.

Also, in a number of specifications, various buildings/systems would now only be required to be OPERABLE when the plant is pressurized (MODE 1, 2 and 3), or during Operations with a Potential for Draining the Reactor Vessel (OPDRVs). The APPLICABILITY of these specifications would no longer include "During movement of recently irradiated fuel...". Finally, references to the FHB Ventilation System would be removed from the Ventilation Filter Testing Program in the Administrative Controls portion of the Specifications. The FHB Ventilation System would no longer be classified as an Engineered Safety Feature (ESF) system.

Specifically, the first change is to remove the following two Specifications in their entirety because they were only applicable during movement of recently irradiated fuel in the FHB:

Specification Specification Name 3.7.8 Fuel Handling Building 3.7.9 Fuel Handling Building Ventilation Exhaust System The second change is to delete references to "movement of recently irradiated fuel assemblies" in the Primary Containment and/or the Fuel Handling Building from the APPLICABILITY statements and ACTIONS of the following Specifications:

Specification Specification Name 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation 3.3.7.1 Control Room Emergency Recirculation (CRER) System Instrumentation 3.6.1.2 Primary Containment Air Locks 3.6.1.3 Primary Containment Isolation Valves (PCIVs) 3.6.1.10 Primary Containment - Shutdown

Enclosure 2 PY-CEI/NRR-2609L Page 2 of 28 3.6.1.11 Containment Vacuum Breakers 3.6.1.12 Containment Humidity Control 3.6.4.1 Secondary Containment 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System 3.7.3* Control Room Emergency Recirculation (CRER) System 3.7.4* Control Room Heating, Ventilating, and Air Conditioning (HVAC) System 3.8.2* AC Sources - Shutdown (Applicability during MODES 4 & 5 remains) 3.8.5* DC Sources - Shutdown (Applicability during MODES 4 & 5 remains) 3.8.8* Distribution Systems - Shutdown (Applicability during MODES 4 & 5 remains)

• The NOTE's in these five specifications which state "LCO 3.0.3 is not applicable" are also being deleted. LCO 3.0.3 is a "shutdown statement" which only applies in MODES 1, 2, and 3. The purpose of the "LCO 3.0.3 is not applicable" note was to make it clear that if irradiated fuel was being handled in the Fuel Handling Building fuel pool during a period when the plant was pressurized (MODE 1, 2, or 3), there was no need to shut down the plant if a problem occurred with a system credited with mitigating a Fuel Handling Accident. As a result of the Fuel Handling Accident AST calculations, these systems are no longer credited in the design basis analyses. Since the APPLICABILITY statements for the above specifications will no longer include "movement of recently irradiated fuel assemblies", there is no need for this "LCO 3.0.3 is not applicable" note to remain.

Finally, reference to the Fuel Handling Building Ventilation System is removed from Administrative Control 5.5.7 "Ventilation Filter Testing Program (VFTP)", subsections a, b, c, d and e.

The annotated pages are provided in Enclosure 4. Also included for NRC information in Enclosures 6 and 7 are annotated Technical Specification Bases pages and a sampling of Updated Safety Analysis Report (USAR) pages, respectively.

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) 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 in lieu of Technical Specification controls 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 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, to show that Shutdown Safety administrative controls can be applied to handling of fuel that has been subcritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Enclosure 2 PY-CEI/NRR-2609L Page 3 of 28 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 5.

DAnr nf Cnnfnrm3nr RaI* '"In ' rv Guide

• 1, 1 W" ui'"ceDeof

,. onorm nc 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 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 Chi/Q values previously approved by the NRC, which were considered to adequately address uncertainties in the actual dispersion of the release.

Taking no credit for the Containment or Fuel Handling Buildings or ventilation/filtration systems is a significant penalty, since as detailed below for Item 1.1.2 "Defense in Depth", buildings and filtration systems such as the Fuel Handling Building

Enclosure 2 PY-CEI/NRR-2609L Page 4 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance Ventilation System will remain available (note: the USAR name for this system uses the word "Area" rather than "Building").

One other significant item that the calculation does not fully credit is the amount of radiological decay that the available inventory would undergo before the point in the outage when it is physically possible to be handling fuel and have the accident occur.

The calculation only assumes 24-hours of decay. It is not physically possible to begin handling of irradiated fuel, or any loads over irradiated fuel, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. It takes substantially longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to drain and decontaminate the upper cavity, remove the Drywell and reactor vessel heads, and remove the steam separators and dryers. An estimate of the best time that can be expected at a plant with PNPP's design features is 3 days (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). The best time that has been achieved to date at PNPP was 79 hours9.143519e-4 days <br />0.0219 hours <br />1.306217e-4 weeks <br />3.00595e-5 months <br />, in a planned mid-cycle fuel replacement outage that was dedicated only to fuel movement. Since radiological decay is a natural process, it is 100% reliable in its reduction of the source term available for release. Therefore assuming only 1 day instead of 3 days of decay is a significant penalty, since crediting the 2 additional days of decay in the calculation would result in a lower source term and lower resultant doses. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> value is enforced in the PNPP Operational Requirements Manual (ORM), in the Decay Time specification (a copy is included at the beginning of Enclosure 6). The 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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."

The dose calculation results remain below the limits of 10 CFR 50.67. Table 7 of Enclosure 5 presents the results of the base calculation (Table 8 presents sensitivities), along with the applicable dose limits:

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

Enclosure 2 PY-CEI/NRR-2609L Page 5 of 28 RAnIIlRtnn, (iiirIi Ii R2 A. iirIn, N*nrn,* hf i*hnfnrm*nebn

-- ' .- - i sr -.- -a-v -e -llq.--'. ,,-ll V &. ~-, VI*il , IJ. lS,/ilSiS

• llAS 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 Shutdown Safety administrative controls over procedures should be evaluated to determine building closure and filtration system availability, in whether the proposed changes are consistent with addition to the natural defenses of radiological decay the principle that adequate defense in depth is over time (which reduces the magnitude of any maintained to compensate for uncertainties in release) and the scrubbing effect of the water pool accident progression and analysis data. over the fuel (which is not being changed as a result Consistency with the defense-in-depth philosophy is of this amendment). As noted above, the maintained ifsystem redundancy, independence, radiological calculations show that Fuel Handling and diversity are preserved commensurate with the Accident doses remain within regulatory acceptance expected frequency, consequences of challenges to limits. Therefore, there are no requirements for the system, and uncertainties. In all cases, Engineered Safety Feature (ESF) grade ventilation compliance with the General Design Criteria in systems or Technical Specification controls over Appendix A to 10 CFR Part 50 is essential. Containment/Fuel Handling Building or ventilation Modifications proposed for the facility generally system OPERABILITY. However, monitoring and should not create a need for compensatory filtration of releases from the plant, even following programmatic activities, such as reliance on manual postulated accidents, is still necessary to ensure operator actions ... compliance with:

• 10 CFR 20 and 10CFR 50 Appendix I requirements on effluents to unrestricted areas

+ Technical Specification 5.5.4 "Radioactive Effluent Controls Program"

• Technical Specification 5.6.3 "Radioactive Effluent Release Report"

+ GDC 61"Fuel Storage and Handling and Radioactivity Control"

• GDC 63 "Monitoring Fuel and Waste Storage"

• GDC 64 "Monitoring Radioactivity Releases" Due to the above requirements, although the Fuel Handling Building Ventilation Exhaust System will no longer be classified as an ESF system, and the Technical Specification controls will be removed, Shutdown Safety administrative controls will still remain in place. As part of the Nuclear Regulatory Commission (NRC) resolution of the proposed Shutdown Rule (1997), the Maintenance Rule, 10 CFR 50.65, was revised to require licensees to assess the impact on shutdown safety before removing equipment from service for maintenance.

The industry, through the Nuclear Energy Institute (NEI), developed guidance to implement this new requirement. A recently approved Revision 3 to NUMARC 93-01, Section 11.3.6.5, contains the final approved wording on how the industry is addressing Containment during plant shutdown periods. License Amendment 102 (March 1999), which provided the original approval of handling irradiated fuel under Shutdown Safety controls rather than Technical Specifications at PNPP, contained a commitment to the draft version of this NUMARC (now NEI) document. With this new letter, this commitment is updated to commit to Revision 3 of NUMARC 93-01,

Enclosure 2 PY-CEI/NRR-2609L Page 6 of 28 Reaulatorv Guide 1.183 Guidance Dearee of Conformance Decree of Conformance Section 11.3.6.5. This ensures that a building closure plan is in effect and ventilation systems remain available to monitor and filter a release from a Fuel Handling Accident. [See Commitment 1 at the end of this Enclosure.]

Proposed modifications that seek to downgrade or In order to be sure the Technical Specification remove required engineered safeguards equipment modifications and the downgrade of the Fuel should be evaluated to be sure that the modification Handling Building Ventilation System to a non-ESF does not invalidate assumptions made in facility do not invalidate assumptions in the PNPP PRAs and does not adversely impact the facility's Probabilistic Safety Analysis (PSA), and does not severe accident management program." adversely impact the facility's Severe Accident Management (SAM) program, this submittal was reviewed by subject matter experts for both the above areas. The conclusion was that neither the PSA nor the SAM guidelines were invalidated or adversely affected.

1.1.3 Integrity of 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 markups of the Updated Safety staff determined that recalculation of all design Analysis Report (USAR) in Enclosure 7.

analyses would generally not be necessary.

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

... In either case, the facility design bases should assumptions and radiological criteria.

clearly indicate that the source term assumptions 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 for requesting relief from the emergency fuel handling accident using administrative controls preparedness requirements of 10 CFR 50.47 and for building closure, as a result of License Appendix E to 10 CFR Part 50. This guidance does Amendment 102.

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 manaqement quides."

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 entails re-evaluation of a limited facility design basis that (1) is based on one or subset of the design basis radiological analyses,

Enclosure 2 PY-CEI/NRR-2609L Page 7 of 28 IRe~ .ltnnrv "miidei1 IWA "irhinnna n--nro.= nf (~nnfenrm~ nv--

na re M (1enfdrmnni-n more of the characteristics of the AST or (2) entails specifically the Fuel Handling Accident. There are re-evaluation of a limited subset of the design basis no physical design modifications to the plant being radiological analyses. The NRC staff will allow performed in concert with this amendment. The licensees flexibility in technically justified selective only "modifications" being proposed are to replace implementations provided a clear, logical, and the Technical Specification controls on consistent design basis is maintained. An example OPERABILITY of ventilation/filtration systems and of an application of selective implementation would the Containment/Fuel Handling Buildings during be ... a request to remove the charcoal filter media fuel handling with Shutdown Safety controls, and from the spent fuel building ventilation exhaust. For declassification of the Fuel Handling Building the latter, the licensee may only need to re-analyze Ventilation Exhaust System to a non-ESF status. In DBAs that credited the iodine removal by the order to accomplish this, the only DBA that needs to charcoal media. Additional analysis guidance is be reanalyzed is the Fuel Handling Accident. The provided in Regulatory Position 1.3 of this guide. descriptions of how the design basis for this event NRC approval for the AST (and the TEDE dose is being maintained are included in the sample criterion) will be limited to the particular selective USAR markups in Enclosure 7.

implementation proposed by the licensee. The licensee would be able to make subsequent It is understood that since this is a selective modifications to the facility and changes to application of the AST for the Fuel Handling procedures based on the selected AST Accident, NRC approval will be limited to this event.

characteristics incorporated into the design basis Use of AST to change the design basis for other under the provisions of 10 CFR 50.59. However, events such as the Control Rod Drop Accident or use of other characteristics of an AST or use of the Main Steam Line Break Outside Containment, TEDE criteria that are not part of the approved or changes to the approved AST characteristics, design basis, and changes to previously approved would require prior staff approval under 10 CFR AST characteristics, would require prior staff 50.67.

arpproval 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. Therefore, there is not a

• Environmental Qualification of Equipment (10 concern that some aspect of the alternative source CFR 50.49) term could make such systems unable to perform a

• Control Room Habitability (GDC-1 9 of Appendix "credited" safety function.

A to 10 CFR Part 50) GDC 19 Control Room Habitability - For a Fuel

• Emergency Response Facility Habitability Handling Accident, a design basis dose calculation (Paragraph IV.E.8 of Appendix E to 10 CFR for the Control Room was performed. The base Part 50) calculation showed that doses remained less than

• Alternative Source Term (10 CFR 50.67) the 5 rem Total Effective Dose Equivalent (TEDE)

• Environmental Reports (10 CFR Part 51) limit in GDC-1 9 and 10 CFR 50.67, even assuming

• Facility Siting (10 CFR 100.11) no Containment or FHB integrity, or filtration system There may be additional applications of the OPERABILITY.

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, ll.B.3)

Support Center (TSC). The dose limit in

"* Accident Monitoring Instrumentation (NUREG Supplement 1 for this facility is 5 rem whole body, 0737, II.F.1) or its equivalent. The "or equivalent" for this evaluation is considered to be 5 rem TEDE.

Enclosure 2 PY-CEI/NRR-2609L Page 8 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance

• Leakage Control (NUREG-0737, II1.D.1.1) Although the TSC has essentially no response

• Emergency Response Facilities (NUREG-0737, function for a Fuel Handling Accident, a scoping III.A.1.2) study for the TSC was performed. The ventilation

• Control Room Habitability (NUREG-0737, intakes for the TSC are farther away from the III.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 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 Containment is required for a Fuel Handling 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 I11.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

Enclosure 2 PY-CEI/NRR-2609L Page 9 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance 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 As noted above, there are no design changes being any associated facility modification should be made in conjunction with this proposal. The supported by evaluations of all significant buildings and ventilation/filtration systems are not radiological and non-radiological impacts of the being physically modified. The Technical proposed actions. This evaluation should consider Specification controls during handling of fuel are the impact of the proposed changes on the facility's being replaced with Shutdown Safety administrative compliance with the regulations and commitments controls, since the design basis calculations no listed above as well as any other facility-specific longer need to credit the Containment/Fuel requirements. These impacts may be due to (1) the Handling Building integrity or ventilation system associated facility modifications or (2) the effectiveness. The Fuel Handling Building differences in the AST characteristics. The scope Ventilation System will continue to be available as a and extent of the re-evaluation will necessarily be a non-ESF system, similar to other non-ESF function of the specific proposed facility modification ventilation systems at PNPP, which are and whether a full or selective implementation is maintained/tested consistent with guidance in being pursued. The NRC staff does not expect a Regulatory Guide 1.140. Compliance with various complete recalculation of all facility radiological regulations and commitments were addressed in analyses, but does expect licensees to evaluate all items above.

impacts of the proposed changes and to update the affected analyses and the design bases The design basis FHA calculation has been updated appropriately. An analysis is considered to be and is included in Enclosure 5 for NRC review. This affected if the proposed modification changes one or selective implementation is solely for the Fuel more assumptions or inputs used in that analysis Handling Accident, since fuel handling is the activity such that the results, or the conclusions drawn on that is being removed from the Applicability of the those results, are no longer valid. Generic analyses, various Technical Specifications. Other design such as those performed by owner groups or vendor basis calculations were determined to not be topical reports, may be used provided the licensee affected by this proposed Technical Specification justifies the applicability of the generic conclusions change. Draft Bases and USAR markups are also to the specific facility and implementation. provided for information in Enclosures 6 and 7.

Sensitivity analyses, discussed below, may also be an option. If affected design basis analyses are to In the calculation, all affected assumptions and be re-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. Statements regarding the The license amendment request should describe the acceptability of the proposed Technical Specification licensee's re-analysis effort and provide statements changes against each of the applicable items regarding the acceptability of the proposed identified in Regulatory Position 1.3.1 of the Reg.

implementation, including modifications, against 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 changes. Since there are no physical design changes being made in conjunction with this proposal, there are also no non-radiological impacts as a result of the proposed changes.

Enclosure 2 PY-CEI/NRR-2609L Page 10 of 28 I -

RauIatnrv bJidA I 1W (,uieIn- a .rt Jfl f f4lrn #^rm,.n,

-e ----- uatr -. -.--. 1 11A --. da-- I- N ~E%F I IuIEcuI%.

I5 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, showing that they do not need to largely based on generic assumptions and inputs. be in the Technical Specifications during fuel Such cases might include post-accident vital area handling. This serves as the basis for the removal access dose calculations, shielding calculations, and of the Technical Specification Applicability of "During equipment environmental qualification (integrated movement of recently irradiated fuel..." from several dose). It may be possible to identify a bounding Control Room related specifications. Two other case, re-analyze that case, and use the results to calculation sensitivity cases were also run, which draw conclusions regarding the remainder of the isolated the control room at the worst possible time, analyses. It may also be possible to show that for after the source term available at the intake is some analyses the whole body and thyroid doses introduced into the Control Room. For these two determined with the previous source term would cases, the isolation exists for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and then was bound the TEDE obtained using the AST. Where followed by either a subsequent re-initiation of the present, arbitrary "designer margins" may be normal intake flow, or the use of the filtration system.

adequate to bound any impact of the AST and TEDE Each case produced an acceptable result, showing criteria. If sensitivity or scoping analyses are used, that following a Fuel Handling Accident, the the license amendment request should include a operators have flexibility on how to operate their discussion of the analyses performed and the ventilation systems without exceeding the conclusions drawn. Scoping or sensitivity analyses radiological acceptance criteria. The base case should not constitute a significant part of the shows that filtration systems are not required. The evaluations for the design basis exclusion area two sensitivity cases show that ventilation/filtration boundary (EAB), low population zone (LPZ), or systems can be effectively used to reduce doses to control room dose." the Control Room operators in the event that the radiation monitor isolates the Control Room intake at the worst possible time after available activity is taken into the Control Room.

A scoping evaluation is also used to show that the 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

Enclosure 2 PY-CEI/NRR-2609L Page 11 of 28 RAnoIRtnrv (b.irIA I 1W tiidanr'a Reuatr "id. - I. I---A-- r-- .--. n- 4: f% I-F. I l

-part of the approved design basis for the facility, use Enclosure 7), 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 armroval under 10 CFR 50.67."

1.3.5 Equipment Environmental Qualification "Current environmental qualification (EQ) analyses There are no physical plant modifications associated may be impacted by a proposed plant modification with this AST implementation. There are also no associated with the AST implementation. The EQ increased EQ requirements as a result of this analyses that have assumptions or inputs affected proposed Technical Specification change and by the plant modification should be updated to declassification of the Fuel Handling Building address these impacts. The NRC staff is assessing Ventilation Exhaust System from an ESF status.

the effect of increased cesium releases on EQ Further details on EQ are provided in Section 6.

doses to determine whether licensee action is warranted...." The cesium issue discussed in this section of the Regulatory Guide is associated with a LOCA and is unrelated to a Fuel Handling Accident.

1.4 Risk Implications "The use of an AST changes only the regulatory The Technical Specification changes constitute the assumptions regarding the analytical treatment of proposed "modifications", along with the the design basis accidents. The AST has no direct declassification of the Fuel Handling Building effect on the probability of the accident. Use of an Ventilation Exhaust System to a non-ESF status.

AST alone cannot increase the core damage Shutdown Safety administrative controls over frequency (CDF) or the large early release frequency building closure and filtration system availability (LERF). However, facility modifications made (rather than OPERABILITY) will replace the possible by the AST could have an impact on risk. If Technical Specification controls. Even without credit the proposed implementation of the AST involves for the Containment or Fuel Handling Building changes to the facility design that would invalidate integrity or the ventilation/filtration systems, the dose assumptions made in the facility's PRA, the impact calculations continue to meet the regulatory on the existing PRAs should be evaluated, acceptance criteria. Therefore, there is not an Consideration should be given to the risk impact of impact on the Large Early Release Frequency as a proposed implementations that seek to remove or result of the proposed changes. Probabilistic Safety downgrade the performance of previously required Analysis personnel reviewed this submittal to engineered safeguards equipment on the basis of determine the impact on the existing PSA. They the reduced postulated doses. The NRC staff may concluded that the change did not invalidate request risk information ifthere is a reason to assumptions made in the PSA.

question adequate protection of public health and safety. The licensee may elect to use risk insights in Although risk insights are not being used to support support of proposed changes to the design basis this change, some risk insights were utilized in NRC that are not addressed in currently approved NRC approval of License Amendment 102, which remain staff positions...." applicable to this proposed change.

1.5 SubmittalRequirements

" ... The NRC staff's finding that the amendment The dose analysis calculation is being provided for

Enclosure 2 PY-CEI/NRR-2609L Page 12 of 28 Renulatnrv uidA 1iR "R a ....ltr uid =II 18 V*

(iiidsan, Gua Idanl IVngsY I flonrnn id rJ f t

V (r

nf,rma nlalrlilnngl may be approved must be based on the licensee's NRC review as Enclosure 5. 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 A draft of a USAR change package, which includes impacts of the proposed modification in sufficient examples of the types of changes that will be made, detail to support review by the NRC staff. The staff is also included for information as Enclosure 7.

recommends that licensees submit affected FSAR f pages annotated with changes that reflect the The Code used in the analysis was RADTRAD 3.02, revised analyses or submit the actual calculation January 5, 2000.

documentation. If the licensee has used a current approved version of an NRC-sponsored computer code, the NRC staff 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 A draft USAR change package is provided for the FSAR be updated to include all changes made in information in Enclosure 7, which provides examples the facility or procedures described in the FSAR.... of how the licensing basis will be revised as a result The affected radiological analysis descriptions in the of this 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 maintainina a clear desian 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.

developing the source terms provided in NUREG 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."...

.4 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 burnup, 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 (Ci/MW). 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 The fission product inventory of each of the fuel inventory should be determined using an rods was determined by dividing the total core I inventory by the number of fuel rods in the core. To m

Enclosure 2 PY-CEI/NRR-2609L Page 13 of 28 anu.Itiw, t,iirIa I 1W fnii'4nj.o rf,,nr-c- rnf (C.nnfnrm~n,'g Dim , . ili I. --

S fl.itirl ill narmn n Vi PrDnDV ISI* II appropriate isotope generation and depletion account for differences in power level across the computer code such as ORIGEN 2 (Ref. 17) or core, a radial peaking factor of 2.0 was applied.

ORIGEN-ARP (Ref. 18).... For DBA events that do This simulates that the rods in the bundle being not involve the entire core, the fission product dropped and the struck bundles would be the inventory of each of the damaged fuel rods is highest inventory rods in the core. This maximum determined by dividing the total core inventory by core wide radial peaking factor of 2.0 is being the number of fuel rods in the core. To account for added into the list of reload analysis parameters differences in power level across the core, radial that must be re-verified each cycle. [See peaking factors from the facility's core operating Commitment 2]

limits report (COLR) or technical specifications should be applied in determining the inventory of For the Fuel Handling Accident analyses performed the damaged rods. ... For events postulated to for this submittal, radioactive decay from the time of occur while the facility is shutdown, e.g., a fuel shutdown was modeled.

handling accident, radioactive decay from the time of shutdown may be 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 alternative, fission gas release calculations maintained at or below 6.3 kw/ft (in other words, the performed using NRC-approved methodologies higher bumup 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 31 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 Radionuclide Composition 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

Enclosure 2 PY-CEI/NRR-2609L Page 14 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance 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 5). The other nuclides in these groups were not included because their core activity was less than 1 E-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 FuelDamage 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-1 4844 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-1 4844 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, including progeny from the decay of parent krypton, and xenon with core activity greater than radionuclides, that are significant with regard to dose 1 E-9 CVMWt (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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 I

Enclosure 2 PY-CEI/NRR-2609L Page 15 of 28 nl~nr* nf (*.nnfnrmannv_

Reglintnru~v Tdi Iww Ii WA *'*MdBiVne noroa MlB 1nnfrll~rlano-lg%

should be assumed to be 2.3 x 10. 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 (Chi/Q), 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The puff release replaced that assumption, 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 5.

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 5). In the adiacent to the control room enveloDe.

Enclosure 2 PY-CEI/NRR-2609L Page 16 of 28 Regulatory Guide 1.183 Guidance De-gree of Conformance

• Radiation shine from the external radioactive event that the Control Room intake isolates and the plume released from the facility, activity is trapped in the Control Room, assuming

• Radiation shine from radioactive material in the lesser quantities of infiltration is conservative, since reactor containment, subsequent inleakage would dilute/purge the

• Radiation shine from radioactive material in trapped activity.

systems and components inside or external to the control room envelope, e.g., radioactive Due to shielding of the Control Room, radiation material buildup in recirculation filters. shine from a Fuel Handling Accident is considered 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 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

Enclosure 2 PY-CEI/NRR-2609L Page 17 of 28

---

- *1 I flnr ,f

---

lat-

...... r 1. 1.. fvl

• evui ncnov i r

• nim rVi Vf ilr illf ll m ...........

measures to minimize dose. [See Commitment 4].

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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 -4 cubic meters per assumed to be 3.5 x 10e cubic meters per second.

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, DDE._, to a finite cloud dose ... "

-i 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). Design envelope source to plant personnel for many of the NUREG-0737 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 hicjher probability of occurrence.

Enclosure 2 PY-CEI/NRR-2609L Page 18 of 28 Reaulatorv OuldA 11R2 (iiidsnr.c rrlnrar* nf 1'rinft'rmnnr,-t 1u . 183 uu Guidancel~gl I~li~lli I ---- .. lla ... Guid ...

postulated EAB and LPZ doses should not exceed event), which is estimated by a scoping study to be the criteria tabulated in Table 6. The acceptance well within the 5 rem TEDE dose. The USAR criteria for the various NUREG-0737 (Ref. 2) items markup provided in Enclosure 7 shows how the generally reference General Design Criteria 19 new dose criteria are being updated.

(GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-19. These RG 1.183 Table 6 also shows an "analysis release criteria are generally specified in terms of whole duration" of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Instead of the 2-hour release body dose, or its equivalent to any body organ. For duration, the calculation conservatively used an facilities applying for, or having received, approval instantaneous release assumption.

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 GeneralConsiderations 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 The conservative, bounding characteristics of the Reprocessing Plants," to 10 CFR Part 50. These AST that the NRC staff chose to present in design basis analyses were structured to provide a Regulatory Guide 1.183 are used in the conservative set of assumptions to test the calculations. Therefore there are no proposed performance of one or more aspects of the facility deviations from the AST characteristics that are design. Many physical processes and phenomena based on specific accident sequences that would are represented by conservative, bounding require additional justification to prove they are assumptions rather than being modeled directly. conservative for other accident sequences.

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 large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion. Licensees should exercise caution in proposing deviations based upon data from a specific accident sequence since the DBAs were never intended to represent any specific accident sequence -- the proposed deviation may not be conservative for other accident seauences."

i 5.1.2 Creditfor 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. The normal ventilation system is are powered by emergency power sources, and are considered to continue to run throughout the event.

either automatically actuated or, in limited cases, Since the base case shows that no credit for have actuation requirements explicitly addressed in isolation or filtration is necessary during fuel emergency operating procedures. The single active handling, the Technical Specification controls on component failure that results in the most limiting mitigating systems are being replaced with radiological consequences should be assumed. Shutdown Safety administrative controls.

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

Enclosure 2 PY-CEI/NRR-2609L Page 19 of 28 Sm I I Iieoulatorv cuid9 11R3 (iiirInr

---

.. . r . . .. .. vv- - .w - vm v n~l ll/Ql r~ndf rtf ('nno%

"l'*

til V l luIIJ r

l n

llul 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 filter efficiency (50%), or re-establishing the normal system intake, effectively purging the Control Room. Both cases provided doses within the dose limits.

5.1.4 Applicability of PriorLicensing Basis I "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 backf it 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 I failing less than 2 feet, rather than the drop of 34 I

Enclosure 2 PY-CEI/NRR-2609L Page 20 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance 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 5) 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 5.

The second item is that the "Decay Time" specification was relocated out of the Technical Specifications as part of Amendment 69, the improved Technical Specifications. The Decay Time specification is located in the PNPP Operational Requirements Manual (ORM), and it requires that the plant be subcritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before movement of irradiated fuel may begin. As described more fully in Section 1.1.1 of this matrix above, it is proposed that this control remain in the ORM, since it is physically impossible to disassemble the vessel, remove all the reactor internals, and move fuel within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

Enclosure 2 PY-CEI/NRR-2609L Page 21 of 28 Requlatory Guide 1.183 Guidance Dearee of Conformance 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 MeteorologyAssumptions "Atmospheric dispersion values (X/Q) for the EAB, No changes to Chi/Q atmospheric dispersion values the LPZ, and the control room that were approved are being proposed. The current USAR Chi/Q 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 X/Q analysis used are presented in the calculation attached as methodology should be reviewed by the NRC staff." Enclosure 5.

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 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 TID1 4844 assumptions for requirements for equipment, rather than the Fuel performing the required EQ analyses. However, no Handling Accident. Finally, no credit is taken in the plant modifications are required to address the revised calculations for the operation of any plant impact of the difference in source term equipment to mitigate the release before it escapes.

characteristics (i.e., AST vs TID14844) on EQ Therefore it is conservative to retain existing EQ doses pending the outcome of the evaluation of the program requirements.

generic issue."

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 conservative 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

Enclosure 2 PY-CEI/NRR-2609L Page 22 of 28 Renultorv GuidA 11R2 (..idnr "R a ltr Guid"e

---

.-- 1.. 18 "iin n fnflelwrn ~llljnll l PUUHIV M.v el 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. It also 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 Reg. Guide 1.183 Table 3 were considered.

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 Dool."

App. B Section 2. Water Depth "Ifthe 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. Ifthe 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 limitina case.

Enclosure 2 PY-CEI/NRR-2609L Page 23 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance 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 5.

App. B Section 3. Noble Gases (andparticulates)

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

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 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 - 7 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 Handling Accidents 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, although need to be analyzed. Shutdown Safety administrative controls will be in place for building closure.

Enclosure 2 PY-CEI/NRR-2609L Page 24 of 28 Regulatory Guide 1.183 Guidance Degree of Conformance 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.

the event of a fuel handling accident, the release 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. This ignores the administrative escapes from the reactor cavity pool to the controls that normally keep the two-building containment is released to the environment over a envelope (of the Containment and the Fuel 2-hour time period. Handling Building) closed, and the proceduralized closure plans that are in place to close off pathways out of this two-building envelope if a Fuel Handling Accident would occur. A commitment is made to Revision 3 to NUMARC 93-01, Section 11.3.6.5 to track the continuance of these administrative controls.

5.4 A reduction in the amount of radioactive No credit is taken for filter systems, although such material released from the containment by ESF systems will continue to be available during fuel filter systems may be taken into account provided handling in accordance with Shutdown Safety that these systems meet the guidance of administrative controls.

Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyses.

5.5 Credit for dilution or mixing of the activity 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-bv-case basis. ... "

Regulatory Analysis/Commitments 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.

Enclosure 2 PY-CEI/NRR-2609L Page 25 of 28 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.

I fltkr nm uItnniLL flnru .mntQ flAnr nf CnnformBnr rfhgr 12& "Intr% T nne-"mantc

• vv*wamwavwm*

I Denree of Conformance

• vm*i=vv v= v ..............

VlIISI It*n*ldat45Vl GDC 61, "Fuel storage and handling and The fuel storage and handling systems, including water radioactivity control." The fuel storage and coverage over the fuel, are not affected by the proposed handling ... systems ... shall be designed to changes. These systems can still be periodically inspected assure adequate safety under normal and and tested. Radiation protection shielding by the buildings postulated accident conditions. These is also unaffected. Appropriate containment, confinement, systems shall be designed (1) with a and filtering systems will remain in place under Shutdown capability to permit appropriate periodic Safety administrative controls rather than Technical inspection and testing of components Specification controls. [Note: the guidance on Fuel important to safety, (2) with suitable Handling Accidents has not required that releases be shielding for radiation protection, (3) with contained or confined. They are typically assumed to be appropriate containment, confinement, and released through a ventilation system within no longer than filtering systems, ... " a 2-hour period. The ventilation systems that serve the Containment area (Containment and Drywell Purge System, Annulus Exhaust Gas Treatment System, and -with the Containment equipment hatch removed- the exhaust subsystem of the Fuel Handling Building Ventilation Svstem) all contain filtration systems.1 GDC 63, "Monitoring fuel and waste Radiation monitors will remain part of the plant design to storage." Appropriate systems shall be detect increases in radiation levels, and along with grab provided in fuel storage ... systems and samples, are addressed by the Shutdown Safety associated handling areas (1) to detect administrative controls. Since no credit is taken for filtration excessive radiation levels and (2) to initiate in the calculations, no safety actions are required to be appropriate safety actions. initiated, although Shutdown Safety administrative controls will be in place for buildinq closure.

GDC 64, "Monitoring radioactivity releases." The Fuel Handling Building Ventilation System is a normally Means shall be provided for monitoring the running system, so releases into the combined reactor containment atmosphere, ... , Containment/Fuel Handling Building envelope will continue effluent discharge paths, and the plant to be routed there (or - if they remain running or are started environs for radioactivity that may be up by the operators - into the Containment and Drywell released from normal operations, including Purge System or the Annulus Exhaust Gas Treatment anticipated operational occurrences, and System) for monitoring prior to release. The plant vents are from postulated accidents. monitored for releases. The Shutdown Safety administrative controls include closure plans to block pathways for unmonitored releases from the two-building envelope of the Containment/Fuel Handling Building. [See Commitment 11.

10 CFR 20, 10 CFR 50 Appendix I, See above discussions. Systems and controls remain in Technical Specification 5.5.4 "Radioactive place to meet these requirements.

Effluent Controls Program", and Technical Specification 5.6.3 "Radioactive Effluent Release Report", each require monitoring of releases and limitations on their maqnitude.

Enclosure 2 PY-CEI/NRR-2609L Page 26 of 28 Other Regulatory Documents Degree of Conformance Regulatory Guide 1.13, "Spent fuel storage PNPP design conforms to this guide with the exception of facility design basis", Revision 1. paragraph C.4. The inventory of radioactive materials available for leakage is based on the assumptions given in Regulatory Guide 1.183.

Regulatory Guide 1.25, "Assumptions used No longer applicable to PNPP. See Regulatory Guide 1.183 for evaluating the potential radiological for the fuel handling accident.

consequences of a fuel handling accident in the fuel handling and storage facility for boiling and pressurized water reactors",

Revision 0.

Regulatory Guide 1.183, "Alternative PNPP conforms to this guide for the fuel handling accident, Radiological Source Terms for Evaluating with minor exceptions to Design Basis Accidents at Nuclear Power + Appendix B, Item 2 (23-foot coverage over the reactor Reactors", Revision 0. vessel flange, as addressed above in the discussions for Section 5.1.4 and App. B Item 2), and

+ Table 6, and Appendix B, Items 4.1 and 5.3 (a puff release rather than a 2-hour release, as discussed above for Sections 4.1.5 and 4.4, and for App. B Items 4.1 and 5.3). The original PNPP licensing basis for a Fuel Handling Accident utilized Regulatory Guide 1.25.

Regulatory Guide 1.140, "Design, testing The Fuel Handling Area Ventilation Exhaust Subsystem will and maintenance criteria for normal now be tested and maintained in accordance with this ventilation exhaust system air filtration and Regulatory Guide.

adsorption units of light-water-cooled nuclear power plants", Revision 0.

Regulatory Guide 1.52, "Design, testing and PNPP design and testing conform to this guide as maintenance criteria for postaccident presented in USAR Tables 6.5-1 and 6.5-3. The Fuel engineered-safety-feature atmosphere Handling Building filter plenum has been evaluated for cleanup system air filtration and absorption compliance to R.G. 1.140.

units of light-water-cooled nuclear power plants", Revision 2.

As the Fuel Handling Accident calculations no longer credit the Containment or Secondary Containment or various ventilation systems, there are no design basis calculations remaining that formally credit these Structures, Systems and Components (SSCs) during plant shutdowns.

Therefore, Criterions 1, 2, and 3 of 10 CFR 50.36 (which identify the SSCs which must be retained within the Technical Specifications due to their association with design basis events) no longer apply during shutdown. Although no design basis calculations credit these structures during shutdown, the Technical Specifications for a number of these SSCs still remain applicable during Operations with a Potential for Draining the Reactor Vessel (OPDRVs). Therefore, these SSCs are remaining within the Technical Specifications during OPDRV periods due to Criterion 4 of 10 CFR 50.36, i.e., "A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety."

Additional Licensing/Regulatory information is provided in the USAR markups in Enclosure 7.

The following table identifies the actions that are considered to be regulatory commitments. Any other actions discussed in this document represent intended or planned actions, are described for the NRC's information, and are not regulatory commitments. Please notify the Manager - Regulatory Affairs at the Perry Nuclear Power Plant of any questions regarding this document or any associated regulatory commitments.

Enclosure 2 PY-CEI/NRR-2609L Page 27 of 28 Commitments

1. Due to the above requirements, although the Fuel Handling Building Ventilation Exhaust System will no longer be classified as an ESF system, and the Technical Specification controls will be removed, Shutdown Safety administrative controls will still remain in place.

As part of the Nuclear Regulatory Commission (NRC) resolution of the proposed Shutdown Rule (1997), the Maintenance Rule, 10 CFR 50.65, was revised to require licensees to assess the impact on shutdown safety before removing equipment from service for maintenance. The industry, through the Nuclear Energy Institute (NEI), developed guidance to implement this new requirement. A recently approved Revision 3 to NUMARC 93-01, Section 11.3.6.5, contains the final approved wording on how the industry is addressing Containment during plant shutdown periods. License Amendment 102 (March 1999), which provided the original approval of handling irradiated fuel under Shutdown Safety controls rather than Technical Specifications at PNPP, contained a commitment to the draft version of this NUMARC (now NEI) document. With this new letter, this commitment is updated to commit to Revision 3 of NUMARC 93-01, Section 11.3.6.5. This ensures that a building closure plan is in effect and ventilation systems remain available to monitor and filter a release from a Fuel Handling Accident.

Note: The exact wording from NUMARC 93-01, Section 11.3.6.5 is as follows:

"In additionto the guidance in NUMARC 91-06, for plants which obtain license amendments to utilize shutdown safety administrativecontrols in lieu of Technical Specification requirementson primary or secondary containmentoperability and ventilation system operabilityduringfuel handlingor core alterations,the following guidelines should be included in the assessment of systems removed from service:

"Duringfuel handling/corealterations,ventilationsystem and radiation monitor availability(as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoringof releasesfrom the fuel. Following shutdown, radioactivityin the RCS decays fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay.

The goal of maintainingventilationsystem and radiationmonitor availabilityis to reduce doses even further below that provided by the naturaldecay, and to avoid unmonitored releases.

" A single normal or contingency method to promptly close primary or secondary containmentpenetrationsshould be developed. Such prompt methods need not completely block the penetrationor be capableof resistingpressure. The purpose is to enable ventilation systems to draw the release from a postulatedfuel handling accident in the proper directionsuch that it can be treated and monitored."

2. This maximum 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.

3. The burnup limit of 62 GWD/MTU on the average exposure of the peak rod, and the LHGR limit of 6.3 kw/ft peak rod average power for the higher burnup fuel (> 54 GWD/MTU), are both being added into the list of reload analysis parameters that must be re-verified each cycle.

4. 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

Enclosure 2 PY-CEI/NRR-2609L Page 28 of 28 Commitments 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.

Environmental Consideration A review has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Enclosure 3 PY-CEI/NRR-2609L Page 1 of 2 SIGNIFICANT HAZARDS CONSIDERATION Changes are proposed to the current Perry Nuclear Power Plant (PNPP) Technical Specifications.

These changes reflect use of Shutdown Safety administrative controls in place of Technical Specification requirements on OPERABILITY of various buildings and filtration systems, which are no longer credited in the design basis radiological dose calculations for a fuel handling accident.

The new Fuel Handling Accident dose calculations were performed using assumptions associated with the Alternative Source Term (AST) outlined in Regulatory Guide 1.183 "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors". The revised calculations demonstrate that the regulatory limits for a Fuel Handling Accident can be met without the credit that was previously taken for filtration systems and the Containment/Fuel Handling Buildings. There are no physical changes to the plant associated with the proposed Technical Specification changes. The "modifications" to the plant therefore consist only of the replacement of the Technical Specification controls with Shutdown Safety administrative controls, and the declassification of the Fuel Handling Building Ventilation Exhaust System to a non-Engineered Safety Feature (ESF) system.

The standards used to arrive at a determination that a request for amendment does not involve a significant hazard are included in Commission regulation 10 CFR 50.92, which states that operation of the facility in accordance with the proposed changes would not:

1) involve a significant increase in the probability or consequences of an accident previously evaluated; or

2) create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) involve a significant reduction in a margin of safety.

The proposed amendment has been reviewed with respect to these three factors, and it has been determined that the proposed change does not involve a significant hazard because:

1. This proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes involve implementation of the Alternative Source Term for the Fuel Handling Accident at PNPP. There are no physical changes to the plant associated with the proposed Technical Specification changes. The revised calculations and controls over buildings and ventilation systems do not impact the initiators of a Fuel Handling Accident in any way. They also do not impact the initiators for any other design basis events. Therefore, because design basis accident initiators are not being altered by adoption of the Alternative Source Term analyses or by the revised controls, the probability of an accident previously evaluated is not affected.

With respect to consequences, the only previously evaluated accident that could be affected is the Fuel Handling Accident. The Alternative Source Term is an input to calculations used to evaluate the consequences of an accident, and does not by itself affect the plant response, or the actual pathway of the radiation released from the fuel. It does however, better represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied. For the Fuel Handling Accident, the AST analyses demonstrate acceptable doses, within regulatory limits, without credit for Containment/Fuel Handling building integrity, filtration system operability, or Control Room automatic isolation. Therefore, the consequences of an accident previously evaluated are not significantly increased. Declassification of the Fuel Handling Building Ventilation Exhaust System to a non- ESF system will change the test acceptance criteria used, from Regulatory Guide 1.52 controls to Regulatory Guide 1.140 controls. However, since the results of the design basis calculation for the Fuel Handling

Enclosure 3 PY-CEI/NRR-2609L Page 2 of 2 Accident remained within the regulatory acceptance criteria without crediting the Fuel Handling Building filtration components, the consequences of this accident are not considered to be significantly increased by the test acceptance criteria change. In addition, although the Technical Specification controls over filtration systems and the Containment/Fuel Handling Buildings are being removed by the proposed changes, Shutdown Safety administrative controls will remain in place to ensure other requirements are met for filtration and monitoring of releases should a Fuel Handling Accident actually occur. Thus, appropriate mitigating techniques will still exist to minimize consequences of such an event to levels lower than those postulated in the revised calculations.

Based on the above conclusions, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. This proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed and there are no physical changes to existing equipment associated with the proposed changes). Also, the changes in methods governing plant/system operation during fuel handling do not create any new initiators or precursors of a new or different kind of accident. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed changes.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. This proposed amendment does not involve a significant reduction in a margin of safety.

The proposed changes are associated with the implementation of a new licensing basis for PNPP Fuel Handling Accidents. Approval of the change from the original source term to a new source term taken from Regulatory Guide 1.183 is being requested. The results of the accident analyses, revised in support of the proposed license amendment, are subject to revised acceptance criteria. The analyses have been performed using conservative methodologies, as specified in Regulatory Guide 1.183. Safety margins have been evaluated and analytical conservatism has been utilized to ensure that the analyses adequately bound the postulated limiting event scenario. The dose consequences of the limiting Fuel Handling Accident remains within the acceptance criteria presented in 10 CFR 50.67 "Accident Source Term", and Regulatory Guide 1.183.

The proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the Control Room, are within corresponding regulatory limits. For the Fuel Handling Accident, Regulatory Guide 1.183 conservatively sets the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) limits below the 10 CFR 50.67 limit, and sets the Control Room limit consistent with 10 CFR 50.67.

Since the proposed changes continue to ensure the doses at the EAB, LPZ and Control Room are within corresponding regulatory limits, the proposed license amendments do not involve a significant reduction in a margin of safety.

Therefore, the change does not involve a significant reduction in a margin of safety.

Based on the above considerations, it is concluded that a significant hazard would not be introduced as a result of this proposed change.

Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 Enclosure 4 PY-CEI/NRR-2609L 3.3 INSTRUMENTATION Page 1 of 44 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation LCO 3.3.6.1 The primary containment and drywell isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERLILL.

APPLICABILITY: According to Table 3.3.6.1-1. 0-ACTIONS

---UI E--------- -----------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACT ION COMPLET ION TIME A. One or more required A.1 P1lace channel in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for channels inoperable. trip. Functions 2.b, 5.b, and 5.d AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Functions other than Functions 2.b, 5.b, and 5.d B. One or more automatic BAI Restore isolation I hour Functions with capability.

i sol at ion capabilIity not maintained.

.(continued)

PERRY - UNIT 1 3.3-48 Amendment No. 69

Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 Enclosure 4 PY-CEI/NRR-2609L ACTINS (continued) Page 2 of 44 CONDITION REQUIRED ACTION COMPLETION TIME K. As required by K.1 Isolate the affected Immediately Required Action C.1 penetration flow and referenced in path(s).

Table 3.3.6.1-1.

OR rt- +. ..... n.. m'

. .... .in .. 9+

.. 1 m a1 i4t,3; w

n.c~ccn:y rr*ui tcu K.? Amnitiate action to Immediately suspend operations with a potential for draining the reactor vessel.

L. As required by Required Action C.1 L.1 Initiate actions to suspend operations Immediately I and referenced in with a potential for Table 3.3.6.1-1. draining the reactor vessel.

PERRY - UNIT 1 3.3-52 Amendment No. 102

Primary Containment and Drywell Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of 6) Enclosure 4 Primary Containment and Drywell Isolation Instrumentation PY-CEI/NRR-2609L Page 3 of 44 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOUABLE FUNCTION CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE

2. Primary Containment and Drywetl Isolation

9. Contairment and (d) 2 K SR 3.3.6.1.1 S4.0 above iR/hr Dryweltt Purge Exhaust Plenum SR 3.3.6.1.2 background SR 3.3.6.1.4 Radiation - High SR 3.3.6.1.5 (continued) 2(W

h. Manual Initiation 1,2,3 G SR 3.3.6.1.5 NA (d) 2 K SR 3.3.6.1.5 MA

3. Reactor Core Isolation Cooling (RCIC) System Isolation

a. RCIC Steam Line FLow - High 1,2,3 F SR 3.3.6.1.1 S298.5 inches SR 3.3.6.1.2 water SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5

b. RCIC Stem Line FLow 1,2,3 1 F SR 3.3.6.1.2 2 3 seconds and Time Delay SR 3.3.6.1.4 & 13 seconds SR 3.3.6.1.5

c. RCIC Steam SumpLy Line 1,2,3 1 F SR 3.3.6.1.1 S55 psig Pressure - Low SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5

d. RCIC Turbine Exhaust 1,2,3 2 F SR 3.3.6.1.1 ý 20 psig Diaphragm SR 3.3.6.1.2 Pressure - High SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5

e. RCIC Equipment Area 1,2,3 1 F SR 3.3.6.1.1 S145.9°F Ambient SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.5 SR 3.3.6.1.7

f. Main Steam Line Pipe 1,2,3 F SR 3.3.6.1.1 s 158.9*F Tunnel SR 3.3.6.1.4 Temperature - High SR 3.3.6.1.5 SR 3.3.6.1.7 (continued)

(b) Required to initiate the drywett isolation function.

Cd) Du perations w,_"itPhapote=ni-.lfor draining the reactor vess4ea_

_0_=4 - of rosenti;' rriiat

-nmn L-I PERRY - LW 1 3.3-56 AM W NOID. 102

CRER System Instrumentation 3.3.7.1 Enclosure 4 PY-CEI/NRR-2609L 3.3 INSTRUMENTATION Page 4 of 44 3.3.7.1 Control Room Emergency Recirculation (CRER) System Instrumentation LCO 3.3.7.1 The CRER System instrumentation for each Function in Table 3.3.7.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.7.1-1. y-, L_

ACTIONS

-NOTE Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Enter the Condition Immediately inoperable, referenced in Table 3.3.7.1-1 for the channel.

B. As required by B.1 Declare associated I hour from Required Action A.1 CRER subsystem discovery of and referenced in inoperable, loss of CRER Table 3.3.7.1-1. initiation capability in both trip systems AND B.Z Place channel in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> trip.

(continued)

PERRY - UNIT I 3.3-70 Amendment No. 69

CRER System Instrumentation 3.3.7.1 Enclosure 4 TabLe 3.3.7.1-1 (page 1 of 1) PY-CEI/NRR-2609L Control Room Emergency Recirculation System Instrunentation Page 5 of 44 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION A.1 REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 a SR 3.3.7.1.1 z 14.3 inches Level- Low Low Low, (a) SR 3.3.7.1.2 Level 1 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

2. DryweLl Pressure--High 1,2,3 2 a SR 3.3.7.1.1 s 1.88 psig SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4 SR 3.3.7.1.5

3. Control Room 1, I C SR 3.3.7.1.1 s 800 cpm Ventilation Radiation SR 3.3.7.1.2 Monitor SR 3.3.7.1.4 SR 3.3.7.1.5 (a) During operations with a potential for draining the reactor vesseL.

( ul")+/-c in riaa

_h o' 1 6 !j hnnr~ltpr

!*m-- mn d mi - ato Dcet 3.3-73 Alfdlnt N. 102

Primary Containment Air Locks 3.6.1.2 3.6 CONTAINMENT SYSTEMS Enclosure 4 PY-CEI/NRR-2609L Page 6 of 44 3.6.1.2 Primary Containment Air Locks LCO 3.6.1.2 Two primary containment air locks shall be OPERABLE.

APPLICABILITY: DES 1. 2 an During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS

1. Entry and exit is --------------- NOTES permissible to perform repairs of the affected air lock components.

2. Separate Condition entry is allowed for each air lock.

3. Enter applicable Conditions and Required Actions of LCO 3.6.1.1. nPrimary Containment-Operating." when air lock leakage results in exceeding containment leakage rate acceptance criteria in MODES 1. 2. and 3. overall CONDITION REQUIRED ACTION COMPLETION TIME A. One or more primary ------------ NOTES -----

containment air locks 1. Required Actions A.1.

with one primary A.2. and A.3 are not containment air lock applicable if both doors door inoperable, in the same air lock are inoperable and Condition C is entered.

2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.

(continued)

PERRY - UNIT 1 3.6-3 Amendment No. fg, 80

Primary Containment Air Locks 3.6.1.2 Enclosure 4 PY-CEI/NRR-2609L ACTIONS Page 7 of 44 CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.3 Restore air lock to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A. AND B. or C not met in MODE 1. 2. or 3. D.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> E. Required Action and E.1 -1E~l mC:WTL0 IIII- ~

associ ated Compl eti on Time of Condition A.

Aor not met irig eenta 4nmznet.

iradeiitcd0 fauth or. Zr--60 OPDRVs.

• - wT t' Initiate action to Immedi ately t I suspend OPDRVs.

I h ___________________

PERRY - UNIT I 3.6-6 Amendment No. 85

IPCIVs 3.6.1.3 Enclosure 4 3.6 CONTAINMENT SYSTEMS PY-CEI/NRR-2609L Page 8 of 44 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except containment vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation."

ACTIONS

-NOTES

1. Penetration flow paths except for the inboard 42 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except penetration flow paths penetration flow path for main steam with one PCIV by use of at least line, inoperable except due one closed and de to leakage not within activated automatic AND limit, valve, closed manual valve, blind flange, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main or check valve with steam line flow-through the valve secured.

AND (continued)

PERRY - UNIT I 3.6-9 Amendment No. 69

PCIVs 3.6.1.3 Enclosure 4 PY-CEI/NRR-2609L

.AcrIcf] Page 9 of 44 CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.3 Perform SR 3.6.1.3.6 Once per 92 days for the resilient seal purge valves closed to comply with Required Action D.1.

E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A, AND B. C, or D not met in MODE 1. 2, or 3. E.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Required Action and

- F.1 Suspend movement of Immedi

-associated Completion recently irradiated Time of Condition A. fuel assemblies i B C. or D not met .

primary c .t.

for PCIV(s) required to be OPERAB3LE during movement irradiatedof fu, recentlyj e

assembj~- n the.

IURY - UNIT 1 PErYnen No. 102

PCIVs 3-6.1.3 Enclosure 4 PY-CEI/NRR-2609L

• ACTONS (CMorured) Page 10 of 44 CONDITION REQUIRED ACTION COMPLETION TIME Required Action and

• y associated Completion 1 Initiate OPDRVs.

suspend action to Immediately I

Time of Condition A.

B. C. or D not met OR for PCIV(s) required to be OPERABLE during 2 Initiate action to Immediately MODE 4 or 5 or during restore valve(s) to operations with a OPERABLE status.

potential for I draining the reactor I vessel (OPDRVs).

/

/

I P- (0 sr031 PERRY - UNIT 1 AEerdmrit No. 102

Primary Containment-Shutdown 3.6.1.10 3.6 CONTAINMENT SYSTEMS Enclosure 4 PY-CEI/NRR-2609L 3.6.1.10 Primary Containment-Shutdown Page 11 of 44 LCO 3.6.1.10 Primary containment shall be OPERABLE.

APPLICABILITY: Fecerntl irrcdi ta ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Primary containment A.1 yofflo";r- Af; . 4 A- -I--

inoperable. - U "IJ 4'ucl azzcmbfi644e 4in I 7rnitiate action to Immediately ispend OPDRVs. I PERRY - UNIT 1 Alr*dmnt No. 102

Containment Vacuum Breakers 3.6.1.11 3.6 CONTAINMENT SYSTEMS Enclosure 4 PY-CEI/NRR-2609L 3.6.1.11 Containment Vacuum Breakers Page 12 of 44 LCO 3.6.1.11 Three containment vacuum breakers shall be OPERABLE and four containment vacuum breakers shall be closed.

APPLICABILITY: MODES 1- 2 and 3.

uring operaions W a potential for draining the reactor vessel (OPDRs)

ACTIONS

---

NOTE -------------------------------------------

Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment-Operating" when the containment vacuum relief subsystem leakage results in exceeding overall containment leakage acceptance criteria.

CONDITION REQUIRED ACTION COMPLETION TIME A. ----NOTE-------- A.1 Close the associated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Separate Condition motor operated entry is allowed for isolation valve.

each containment vacuum breaker. AND One or two containment A.2 Restore required containment vacuum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I vacuum breakers not breaker to OPERABLE closed, status.

OR One required containment vacuum breaker inoperable for other reasons.

(continued)

PERRY - UNIT I 3.6-31 Amendment No.111

Containment Vacuum Breakers 11 6 1.11 Enclosure 4 PY-CEI/NRR-2609L AC-GM (continued)

CONDITION JREQUIRED ACTION PaNe 13 of 44 JCOMPLETION TIME B. Required Action and - - -- ------------ -NOTE ---

associated Completion Only applicable in MODE 1. 2 Time of Condition A or 3.

not met.

OR B.1.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Three or more AND containment vacuum breakers not closed. B.1-2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR AND Two or more required - NOTE -----------

containment vacuum 40 1 ap licable durin breakers inoperable 0 1!ýIic ý. lý I for other reasons. 9. enll~~i~ll Pp... A*f in S... .. - ......... ... TI I J

OPOR-Ys.

Sf B.2.1 bu mo'cmn Hi01RR 9;F IlfC1 4¶@Q4 atL il---

rcccntl': I rrdI tcr-

"*.f I

thz priImry Initiate action to suspend OPDRVs.

Immediately I

I _______________

PERY - UNIT 1 3.6-32 A1drt No. 102

Containment Humidity Control 3.6.1.12 3.6 CONTAINMENT SYSTEMS Enclosure 4 3.6.1.12 Containment Humidity Control PY-CEI/NRR-2609L Page 14 of 44 LCO 3.6.1.12 Containment average temperature-to-relative humidity shall be maintained within limits.

APPLICABILITY: .S 1. . and 3.

During operations with a potential for draining the reactor vessel (OPDRVs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO A.1 Restore containment 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> not met. average temperature to-relative humidity to within limits.

I (continued)

PERRY - UNIT 1 3.6-34 Amendment No.111

Containment Humidity Control Enclosure 4 3.6.1.12 PY-CEI/NRR-2609L ACTIONS (continued) Page 15 of 44 CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1. 2. I or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and C.I  ; ..... n t f4 R 1mmed i,÷ ately associated Completion rW..... A*AZYd-i.t Time of Condition A fel  : lic: in movementof- ar.4ently U cr

-. g DUzS iInitiate action to 1suspend:0OPRVs. Immediately SURVEILLANCE REQUIREMENT SURVEILLANCE FREQUENCY SR 3.6.1.12.1 Verify containment average temperature- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to-relative humidity to be within limits.

PERRY - UNIT 1 3.6-35 Amendment No.111

Secondary Containment 3.6.4.1 Enclosure 4 3.6 CONTAINMENT SYSTEMS PY-CEI/NRR-2609L Page 16 of 44 3.6.4.1 Secondary Containment LCO 3.6.4.1 The secondary containment shall be OPERABLE.

APPLICABILITY:

potential for draining the reactor ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Secondary containment A.1 Restore secondary 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable in MODE 1, containment to

2. or 3. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met.

B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

PERRY - UNIT 1 3.6-51 Anerl~elt No. 102

Secondary Containment 3.6.4.1 Enclosure 4 PY-CEI/NRR-2609L Page 17 of 44 CONDITION REQUIRED ACTION COMPLETION TIME C. Secondary C.1 e#.....

^#

f a te; .

!w......

_y-containment i~~radifuzl fue4blz~

=in24 IR

.diat.d I

e C.2DVs. Immediately I

Initiate action to ssuspend OPDRVs.

SUIRVETLANCEQffUITREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1.1 Verify secondary containment vacuum is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> t 0.66 inch of vacuum water gauge.

SR 3.6.4.1.2 Verify the primary containment equipment 31 days hatch is closed and sealed and the shield blocks are installed adjacent to the shield building.

SR 3.6.4.1.3 Verify each secondary containment access 31 days door is closed, except when the access opening is being used for entry and exit.

PIBY - UNIT 1 3.6-52 6a1erdlnt Nl. 102

SCIVs 3.6.4.2 Enclosure 4 3.6 CONTAINMENT SYSTEMS PY-CEI/NRR-2609L Page 18 of 44 3.6.4.2 Secondary Containment Isolation Valves (SCIVs)

LCO 3.6.4.2 Each SCIV shall be OPERABLE.

APPLICABILITY: D 1 2 and3 .

During opera-ions wit a potential for draining the reactor vessel (OPDRVs).

ACTIONS

-NOTES

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by SCIVs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Isolate the affected 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> penetration flow paths penetration flow path with one SCIV by use of at least inoperable, one closed manual valve or blind flange.

AND (continued)

PEIRY - LNIT 1 3.6-53 Amrim(Jllt W. 102

SCIVs 3.6.4.2 Enclosure 4 PY-CEI/NRR-2609L ACf'IS (:axi nued) P~n a 1 C1nf A CONDITION REQUIRED ACTION COMPLETION TIME D.1 c~mn oupn f 4fflffl e44at -

D. Required Action and r...t irm*- I--t associated Completion fuci cszmbl zz in Time of Condition A I Sin .cnt 1nmnt.

Firradi at-d f-Uci' AND I D.2 Initiate action to Immnediately

= Dsuspend OPDRVs.

SURVETL{ANCE REOUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.2.1 ------------------ NOTES ------------

1. Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2. Not required to be met for SCIVs that are open under administrative controls.

Verify each secondary containment 31 days isolation manual valve and blind flange that is required to be closed during accident conditions is closed.

PERRY - UNTI" 1 3.6-55 A1mernt No. 102

AEGT System 3.6.4.3 Enclosure 4 3.6 PY-CEI/NRR-2609L CONTAINMENT SYSTEMS Page 20 of 44 3.6.4.3 Annulus Exhaust Gas Treatment (AEGT) System "

LCO 3.6.4.3 Two AEGT subsystems shall be OPERABLE.

APPLICABILITY: MODES 1Lad3,)

During op s wi a potential for draining the reactor vessel (OPDRVs).

-ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One AEGT subsystem A.1 Restore AEGT 7 days inoperable, subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1. 2.

or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and C.1 Place OPERABLE AEGT Immediately associated Completion subsystem in Time of Condition A operation.

W ROR i

(continued)

PEM - UNIT 1 3.6-56 Atir No. 102

AEGT System 3.6.4.3 Enclosure 4 PY-CEI/NRR-2609L Page 21 of 44 COMPLETION TIME C. (continued) tuel ~mle~i Imeitl eomto 4me4---

Initiate action to Immediately suspend OPDRVs.

D. Two AEGT subsystems Iflmmedi ately inoperable in MODE 1.

2. or 3.

E. Two AEGT subsystems erable NiIno during . re ly i rr ~t t

• L....I ~*.T. ,,,,,a *,. I '

5 ermJb t44 4n _

azzcmbliz in t.

Initiate action to Immedi atel y suspend OPDRVs.

PERRY - UNIT 1 3.6-57 PY N1ADl~ndlent

. 102

CRER System 3.7.3 Enclosure 4 3.7 PLANT SYSTEM PY-CEI/NRR-2609L Page 22 of 44 3.7.3 Control Room Emergency Recirculation (CRER) System LCO 3.7.3 Two CRER subsystems shall be OPERABLE.

APPLICABILITY:

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRER subsystem A.I Restore CRER 7 days inoperable. subsystem to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A AND not met in MODE 1. 2.

or 3. B.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued) 3.7-4 3.7-4AlneIlet No. 102

CRER System Enclosure 4 J.1/.,

ACTIONS (continued) PY-CEI/NRR-2609L CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and ------------ NOTE --------- 2 associated Completion LCO 3.0.3 is n Time of Condition A ............

ogt et durnnE Wm A- aMCondC.1 Afrradhated fui Place OPERABLE CRER subsystem in Immediately I nemergency 4M recirculation mode.

t_ OR OPDRVs.

r.2.3 upn movcmcnt ef 4m4eey

• czntly

• . ,1-i rrad atc fuselzmblicz 4n I the primary ý7 C..2 Initiate action to Immediately suspend OPDRVs.

D. Two CRER subsystems D01 Enter LCO 3.0.3. Immediately inoperable in MODE 1.

2. or 3.

(continued)

PERY - UNIT 1 3.7-5 AnaxEntlNo. 102

CRER System 3.7.3 Enclosure 4 PY-CEI/NRR-2609L Paue 24 of 44 E. Two CRER subsystems

/* *erabieca-durnn go i F-ec*-ent! y irradiated fuel az~cmbli M%4t OPDRVs.. -Initiatet action to t

Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Operate each CRER subsystem for 31 days

Ž 10 continuous hours with the heaters operating.

SR 3.7.3.2 Perform required CRER filter testing in inaccordance accordance with the Ventilation Filter with the VFTP Testing Program (VFTP).

SR 3.7.3 3 Verify each CRER subsystem actuates on an 24 months I actual or simulated initiation signal.

(continued)

PERRY - UNIT 1 3.7-6 Amendment No. 115

Control Room HVAC System 3.7.4 Enclosure 4 PY-CEI/NRR-2609L 3.7 PLANT SYSTEMS Paqe 25 of 44 3.7.4 Control Room Heating. Ventilating, and Air Conditioning (HVAC) System LCO 3.7.4 Two control room HVAC subsystems shall be OPERABLE.

APPLICABILITY: E 2. and 3 During opera ions w6t ap n-ial 'or 'rainingt e reactorI vessel (OPDRVs).

ACTIONS CONDITION REQUIRED. ACTION COMPLETION TIME A. One control room HVAC A.1 Restore control room 30 days subsystem inoperable. HVAC subsystem to OPERABLE status.

B. Two control room HVAC B.1 Verify control room Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> subsystems inoperable, air temperature is

, 90'F.

AND B.2 Restore one control 7 days room HVAC subsystem to OPERABLE status, C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A or AND B not met in MODE 1.

2. or 3. C.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)

PERRY - UNIT 1 3.7-8 N13Aye Nlt O2

Control Room HVAC System

_ 3.7.4 Enclosure 4 PY-CEI/NRR-2609L ACTIN (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and ------------

NO associated Completion LCO 3.00 not applicable.

Time of Condition A -__ -----------------------

ot et duri D.1 Place OPERABLE Immnediately irradiated ass-Mba-6 in the control room HVAC subsystem in I

, 01a eel~ainme-operation.

s. I

.. el......... e in contanmentand fuiel h4andling building;9.

D.2@ Initiate action to suspend OPDRVs.

Irmediately I

(continued)

PEIRY - UNIT 1 3.7-9 AI&ikrent No.102

Control Room HVAC System 3.7.4 Enclosure 4 PY-CEI/NRR-2609L ACTIONS (continued) Pane 27 of 44 Pane 27 of 44 CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and associated Completion LC60 isnot applicable.

Time of Condition B

-. 4

( iprrdia*4ted f'-elth@

3Zcmbi§44:- in r-edntlyo-I pa~d 4ate fuol4e the ~pimffdfi 3z4mbic i

• lIilli*lJmJJf .................

+h diit.ng bldi ng.

4 Initiate action to Immediately suspend OPDRVs.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify each control room HVAC subsystem has the capability to remove the assumed heat 24 months I load.

PERRY - UNIT I 3.7-10 Amendment No. 115

Fuel Pool Water Level 3.7.7 Enclosure 4 PY-CEI/NRR-2609L k .

Page 28 of 44 3.7 PLANT SYSTEMIS 3.7.7 Fuel Pool Water Level j~or LCO 3.7.7 The fuel pool water level shall be 3 ft over the top of 2>

irradiated fuel assemblies seated in the fuel handling building (FHB) and upper containment fuel storage racks.

APPLICABILITY: During movetnent of irradiated fuel asse blies in the associated fuel storage pools.

ACTIONS (

CONDITION TIME A. Fuel pool water level A.I NOTE-- --------.

not within limit.

Suspend movement of Immediately irradiated fuel assemblies in the associated fuel storage pool(s).

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify the fuel pool water level is ? 23 ft 7 days over the top of irradiated fuel assemblies seated in the storage racks.

PERRY - UNIT I 3.7-14(reid :aAsl.j-lj Amendment No. 69

Fuel Handling Building 3.7.8 Enclosure4 Enclosur 4

.7 PLANT SYSTEMS PY-CEI/NRR-2609L Page 29 of 44

3. 8 Fuel Handling Building LCO 3. .8 The fuel handling building (FHB) shall be OPERABL APPLICABILI During movement of recently irradiated fuel the FHB.

serblies in I

ACTIONS L - - ---------------- NOTE---- ----------------------

LCO 3.0.3 isnot applicable.

CONDITION

• REQUIRED ACTI/ COMPLETION TIME A- FHB inoperable. *.1 Suspendl *vement of Immediately rece/ntl~irradi ated the F a Ifuel i n embis I SURVEILLANCF REOUIREMENTS SURVEIL/ C FREQUENCY SR 3.7.8.1 Verify all1F* floor hatches and thhe :shield 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> blocks adj /ent to the shield buil l'ng are instal led-/and the FHB railIroad traL* door0 SR 3.7.8.2 Ver* fy each FHB access door i s closed,. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

/exept when the access opening i en led for entry and exit.

PEWRY

/-LUflNIT 31 3.7-15

ý No. 12

Fuel Handling Building Ventilation Exhaust System 3.7.9 Enclosure 4 PY-CEI/NRR-2609L 7 PLANT SYSTEMS Page 30 of 44 3.-N9 Fuel Handling Building Ventilation Exhaust System LCO .9 Three fuel handling building (FHB) ventilation subsystems shall be OPERABLE.

APPLICABIL: During movement of recently irradiated fuela&s the FHB.

in I ACTIONS

---

l-----

-b-e.

LCO 3.0.3 is not app O/

CONDITION REQUIRED !71N COMPLETION TIME A. One required FHB 7 days ventilation exhaust ventil ion exhaust subsystem inoperable- A. Restore H/HB subsy em to OPERABLE stat s.

I.

B. Required Action and Immediately associated Completion Time of Condition A not met.

Immedi ately I

C. Two or three ventilation subsystems it Inimediat I

I Sntinued)

PERRY UNIT1 3.7-16 No:.102

Fuel Handling Building Ventilation Exhaust Syst

3. .9 Enclosure 4 PY-CEI/NRR-2609L ArA AM #-^n+i;nttarl\ Page 31 of 44t CONDITION REQUIRED ACTION COMPLETI TIME D. FHB v tilation D.1 Obtain and analyze a Every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> exhaus, radiation grab sample of the monitor noble gas) FHB ventilation inoperabi exhaust system effluent.

AND Z.2.1Verify Unit I Plant Every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> vent noble gas monitor is opera e.

OR

.2.2 Place the F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ventil ati.o. exhaust radiation onitor (noble*/coippetr condition.

gAd) inNthe PERRY -NIT 1 3.7-17 Amendment No.

Fuel Handling Building Ventilation Exhaust Sy .7.9 e

/ "

Enclosure 4 PY-CEI/NRR-2609L Page 32 of 44 S EIILLANCE REQUIREMENTS SURVEILLANCE FREQ Cy SR 3.7. 1 Operate each FHB ventilation exhaust 31d s subsystem for Ž 10 continuous hours with heaters operating.

SR 3.7.9.2 Pe form FHB ventilation exhaust filter tes F\ilt ing Testing Program with in accordance Z the Ventilatio (VFTP).

In accordance with the VFTP SR 3.7.9.3 Perform a stem functional test- 24 months I SR 3.7.9.4 Perform a CHANN FUNCTIONAL ST of the 92 days FHB ventilation e haust r!di ion monitor (noble gas)

PERRY UNIT 1 3.7-18 Amendment 115

AC Sources--Shutdown 3.8.2 Enclosure 4 3.8 ELECTRICAL POWER SYSTEMS PY-CEI/NRR-2609L Paqe 33 of 44 3.8.2 AC Sources -Shutdown LCO 3.8.2 The following AC electrical power sources shall be OPERABLE:

a. One qualified circuit between the offsite transmission network and the onsite Class 1E AC electrical power distribution subsystem(s) required by LCO 3.8.8, "Distribution Systems - Shutdown";

b. One diesel generator (DG) capable of supplying one division of the Division 1 or 2 onsite Class YE AC electrical power distribution subsystem(s) required by LCO 3.8.8: and

c. One qualified circuit, other than the circuit in LCO 3.8.2.a. between the offsite transmission network and the Division 3 onsite Class 1E electrical power distribution subsystem, or the Division 3 DG capable of supplying the Division 3 onsite Class 1E AC electrical power distribution subsystem, when the Division 3 onsite Class 1E electrical power distribution subsystem is required by LCO 3.8.8.

APPLICABILITY: MODES 4 andS---, -

P - UNIT 1 3.8-17 3tNo. 102

AC Sources - Shutdown 3.8.2 Enclosure 4 PY-CEI/NRR-2609L ACTIONS Paqe 34 of 44

.CO 3.0.3is--.. i .

CONDITION REQUIRED ACTION JCOMPLETION TIME NOTE--------- -............

A. LCO Item a not met. Enter applicable Condition and Required Actions of LCO 3.8.8. when any required division is de-energized as a result of Condition A.

A.1 Declare required Immedi atel y feature(s) with no offsite power available from a required circuit inoperable.

OR A.2. 1 Suspend CORE Imedi ately ALTERATIONS.

AND I I fueenl Azmlizi I A.2r* ae action to Immedi ately suspend operations with a potential for draining the reactor vessel (OPDRVs).

AND (continued)

PERRY - UNIT I 3.8-18 Aendi~it No. 102

AC Sources - Shutdown 3.8.2 Enclosure 4 PY-CEI/NRR-2609L Page 35 of 44 CONDITION REQUIRED ACTION COMPLETION TIME A. (Continued) A.2 Initiate action to Immediately restore required offsite power circuit to OPERABLE status.

B. LCO Item b not met. B.1 Suspend CORE Immediately ALTERATIONS.

AND u pc,,

n-dm3,.'~mcnt of TInmcdiat"ll fu-c! rccc azzemb!li ntly  : in4t adil rin...M I ate atoto Immediately B. Intiateaction to suspend OPDRVs.

AND

-* Immedi atel y B.0 Initiate action to restore required DG to OPERABLE status.

C. LCO Item c not met. C.1 Declare High Pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Core Spray System inoperable.

PERRY - UNIT 1 3.8-19 ATmt No. 102

DC Sources--Shutdown 3.8.5 Enclosure 4 3.8 ELECTRICAL POWER SYSTEMS PY-CEI/NRR-2609L Page 36 of 44 3.8.5 DC Sources-Shutdown LCO 3.8.5 The following DC electrical power subsystems shall be OPERABLE:

a. One Class 1E DC electrical power subsystem capable of supplying one division of the Division 1 or 2 onsite Class 1E electrical power distribution subsystem(s) required by LCO 3.8.8. "Distribution Systems Shutdown";

b. One Class 1E battery or battery charger, other than the DC electrical power subsystem in LCO 3.8.5.a. capable of supplying the remaining Division 1 or Division 2 onsite Class 1E DC electrical power distribution subsystem when required by LCO 3.8.8: and

c. The Division 3 DC electrical power subsystem capable of supplying the Division 3 onsite Class 1E DC electrical power distribution subsystem. when the Division 3 onsite Class 1E DC electrical power distribution subsystem is required by LCO 3.8.8.

APPLICABILITY: MO 4* 5 .....

- an f PPROW^ ipýAdi'tM f41^1lnn PERRY - UNIT I 3.8-28 Amendment No. 102

DC Sources - Shutdown 3.8.5 Enclosure 4 PY-CEI/NRR-2609L PaQe 37 of 44 NS

---

NOWTI--- --

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Declare affected Immediately DC electrical power required feature(s) subsystems inoperable, inoperable.

OR A.2.1 Suspend CORE Immediately ALTERATIONS.

AND

-I A.2**2 Initiate action to Immediately csuspend operations withea potential for draining the reactor vessel.

AND A.22 Initiate action to Immediately restore requi red DC electrical power subsystems to OPERABLE status.

PERRY - UNIT I 3.8-29 Amendment No.102

Distribution Systems - Shutdown 3.8.8 Enclosure 4 3.8 ELECTRICAL POWER SYSTEMS PY-CEI/NRR-2609L Pacie 38 of 44 3.8.8 Distribution Systems--Shutdown LCO 3.8.8 The necessary portions of the Division 1, Division 2, and Division 3 AC and DC electrical power distribution subsystems shall be OPERABLE to support equipment required to be OPERABLE.

APPLICABILITY: and 5.

f V__

the prmr coti or wnfuie handling building.:

ACTIONS CONDITION I REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Declare associated Immediately AC or DC electrical supported required power .distribution feature(s) subsystems inoperable. inoperable.

OR A.2.1 Suspend CORE Immedi ately ALTERATIONS.

AND A-_ bupei Hi eiei . 9+.-, i...t ;wedqu _40LC 4 .- ,A4 ,,+,%A 2

i

+'m~~ranI ;r ra a~ta I

-the primary z-eeO. cn and fuel (continued)

I. ______________________ I PEFRY - UNIT 1 3.8-38 APr('nc t No. 102

Distribution Systems-Shutdown 3.8.8 Enclosure 4 PY-CEI/NRR-2609L Paqe 39 of 44 ArTTfl*J CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.6 Initiate action to Immediately suspend operations with a potential for draining the reactor vessel.

AND A.2 -0 Initiate actions to Immediately restore required AC and DC electrical power distribution subsystems to OPERABLE status.

AND A.2.**) Declare associated Immediately required shutdown cooling subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct breaker alignments and 7 days voltage to required AC and DC electrical power distribution subsystems.

PERRY - UNIT 1 3.8-39 Amendment No. 69

RPV Water Level-Irradiated Fuel 3.9.6 Enclosure 4 3.9 REFUELING OPERATIONS PY-CEI/NRR-2609L Page 40 of 44 3.9.6 Reactor Pressure Vessel (RPV) Water Level-Irradiated Fuel LCO 3.9.6 RPV water leve 1 shall be _>22 ft 9 inches above the top of the RPV flange APPLICABILITY: During movemen t f irradiated fuel assemblies within the RPV. k ACTIONS "CONDITION REQUIRED ACTION COMPLETION TIME A. RPV water level not A.1 Suspend movement of Immediately within limit, irradiated fuel assemblies within the RPV.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify RPV water level is ? 22 ft 9 inches 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> above the top of the RPV flange.

PERRY - UNIT 1 3.9-8 Amendment No. 69

RPV Water Level-New Fuel or Control 3.9.7 Rods Enclosure 4 PY-CEI/NRR-2609L Paqe 41 of 44 3.9 REFUELING OPERATIONS Fuel or Control Rods 3.9.7 Reactor Pressure Vessel (RPV) Water Level--New irradiated LCO 3.9.7 RPV water level shall be 2 23 ft above the top of fuel assemblies seated within the RPV.

During movement of new uel assemblies or handlingfuel of APPLICABILITY: irradiated control rods within the RPV when assemblies are seated within the RPV.

ACTIONS ,

REQUIRED ACTION COMPLETION TIME CONDITION A.1 Suspend movement of Immediately A. RPV water level not within limit, new fuel assemblies and handling of control rods within the RPV.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify RPV water level is k 23 ft above the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> top of irradiated fuel assemblies seated within the RPV.

3.9-9 Amendment No. 69 PERRY - UNIT 1

No A- ~ - +tA~s Programs and Manuals 5.5 Enclosure 4 5.5 Programs and Manuals (continued) PY-CEI/NRR-2609L Paqe 42 of 44 5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

I

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Required frequencies Addenda terminology for for performing inservice inservice testing activities testinq activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required frequencies for performing inservice testing activities:

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

A 'program shall be established to implement the following required "testing of Engineered Safety Feature ([SF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2.

(continued)

PERRY - UNIT 1 5.0-10 Amendment No. 120

Programs and Manuals 5.5 Enclosure 4 5.5 Programs and Manuals PY-CEI/NRR-2609L Paqe 43 of 44 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI N510-1980 at the system flowrate specified below +/- 10%:

ESF Ventilation System Flowrate a) Control Room Emergency Recirculation 30,000 scfm b) FRuo' HW.-.d-ing BOWl

_.ding e-+K c) Annulus Exhaust Gas Treatment 2,000 scfm

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2 and ANSI N510-1980 at the system flowrate specified below +/- 10%:

ESF Ventilation System Flowrate a) Control Room Emergency Recirculation 30,000 scfm b) Fuel HEadliustuilding tremen 15,000 scfm '

c) Annulus Exhaust Gas Treatment 2,000 scfm

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 °C and equal to the relative humidity (RH)

I specified below:

ESF Ventilation System Penetration RH a) Control Room Emergency Recirculation b) Fuo-Al H-.d.n. Building Ie2 c) Annulus Exhaust Gas Treatment 2.5%

S.........

In~ &a/

J *m[

0.5%

70%

70%

J 0 (-

. r (continued)

I PERRY - UNIT 1 5.0-11 Amendment No. 117

Programs and Manuals 5.5 Enclosure 4 5.5 Programs and Manuals PY-CEI/NRR-2609L Paqe 44 of 44 5.5.7 Ventilation Filter Testing Program (VFTP) (continued)

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52. Revision 2. and ANSI N510-1980 at the system flowrate specified below t 10%:

ESF Ventilation System elta P I__Flowrate a) kiR ren nf c)3nulus aus as a rea ea men en2OA 2.00 s l

e. Demonstrate that the heaters for each of the ESF systems dissipate the value specified below _t101 when corrected to nominal input voltage when tested in accordance with ANSI N510-1980:

ESF Ventilation System Wattacte c) au u uu"as"ýreatmen'r ýion The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.5.8 Explosive Gas and Storage Tank Radioactivity Mpnitoring Program This program provides controls for potentially explosive gas mixtures contained in the main condenser offgas treatment system.

and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen in the main condenser offgas treatment system and a surveillance program to ensure the limits are maintained. Such limits sha 1 be appropriate to the system's design criteria (i.e.. whether or not the system is designed to withstand a hydrogen explosion): and (continued)

PERRY - UNIT 1 5.0-12 Amendment No. 85

Enclosure 5 PY-CEI/NRR-2609L Page 1 of 76 Page i FrstEnery Pr cleary 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 CLASSIFICATION CATEGORY REFERENCED IN REFERENCED OPEN USAR VALIDATION IN ATLAS? ASSUMPTIONS?

[ SAFETY-RELATED [ ACTIVE DATABASE?

E] AUGMENTED QUALITY LI HISTORICAL [ YES [ YES I] YES

[0 NON SAFETY RELATED F] STUDY El NO I NO I E NO COMPUTER PROGRAM(S)

RADTRAD Mod 3.02, MicrosoftWord 2000, Microsoft@ Excel 2000 REVISION RECORD Rev. Description of Change Affected Preparer/Date LI Lead Pages Reviewer/Date Engineer/Date 0uVerifierIDAate 0 Initial Issue All ***/~2-* //2.

Enclosure 5 PY-CEI/NRR-2609L Page 2 of 76 ii arst~nePage Srst~ rgyCALCULATION 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 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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. Both 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 5 PY-CEI/NRR-2609L Page 3 of 76

~Enetgy CALCULATION Page iii 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 W

zZ

• oS z 6 *0 0. 0 ent Num ber/Title Revision, Edition, Date n 1 NUREG-1465, "Accident Source Terms for February 1995 El []

LI Light-Water Nuclear Power Plants" 2 Reg. Guide 1.183, "Alternative Radiological July 2000 El ED [

Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" 3 NEI 99-03, "Control Room Habitability June 2001 E L[

Assessment Guidance," Nuclear Energy Institute 4 CEI CALC. No. Calculation 3.2.8, "FHA Rev. 1 L1 LI[

Inside Containment" 5 CEI CALC. No. 3.2.8.1, "Control Room Rev. 0, 12/30/1998 El 0 EL Habitability Following a Fuel Handling Accident" 6 Reg. Guide 1.25 (Safety Guide 25), March 23, 1972 [] I] I "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 Factors for Power and Test Reactor Sites" 8 NUREG/CR-6604,"RADTRAD: A June 1997 ED E] 11 Simplified Model for RADionuclide Transport and Removal And Dose Estimation" 9 NUREG/CR-6604, Supplement 1, June 8, 1999 Z 0 El "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation" 10 Letter from D. R. Rogers to J. B. Balcken, January 29, 1996; Revised, March 14, 1996 LI ED L "Fission Product Inventories for Perry High Energy Cycles" (Attachment 2 to DIN # 12)

Enclosure 5 PY-CEI/NRR-2609L Page 4 of 76 Page iv 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 C.)

C O z

0) 0 O Document Number/Title Revision, Edition, Date 11 Code of Federal Regulations:

10 CFR Part 50.67, "ACCIDENT 01/24/2000 SOURCE TERM" 10 CFR Part 50, Appendix A, 01/24/2000 0 El El "GENERAL DESIGN CRITERIA FOR NUCLEAR POWER PLANTS" 10 CFR Part 50, Appendix K 01124/2000 E] 0 El 12 DI-240, "Fuel Handling Accident Input Rev. 0, 11-8-01 El 0 r-E Assumptions: Fuels Input" 13 ORNL/TM-7175, "A User's Manual for the July 1980 N El [E ORIGEN 2 Computer Code" 14 Federal Guidance Report 11, "Limiting 2nd Printing, 1989 DI 0 El 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 EL Z 17 RSICC Computer Code Collection MACCS2 16 Federal Guidance Report 12, "External 1993 El N El Exposure to Radionuclides in Air, Water, and Soil" 17 PNPP Drawings: El Z 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 Z R TEDE Calculation"

Enclosure 5 PY-CEI/NRR-2609L Page 5 of 76 rCALCULATION Perry Nuclear PwrPat PNPP No. 6077 Rev. 2r7101 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 Z .G c 0,Number/Title 0 Revision, Edition, Date"Elf 19 License Amendment 102 March 1999 El 1Z [

20 License Amendment 103 March 1999 [] 0 []

21 License Amendment 112 June 2000 El 0 1:1 22 NEDC-32868P, "GE-14 Compliance with Rev. 1, December 11, 2000 El 0 [II Amendment 22 of NEDC-2401 1-P-A "GESTAR-Il"'

23 Perry Technical Specifications 3.9.6 El 0 []

24 Letter from L. R. Conner of Global Nuclear November 5, 2001 El 0 D]

Fuel to P. J. Curran of PNPP, "Fuel Handling Accident - Bounding Fuel Rod Pressure for GE12 and GE14" (Attachment 1 to DIN # 12) 25 Calculation CL-M26-01 Rev. 1 E] 0 El 26 P&ID 912-610 Rev. CC El [9 El 27 Periodic Test Instruction PTI-GEN-P0011 Rev.1 El 0 []

28 SCIENTECH Interoffice Memo from 3/14/01 [E El El H. A. Wagage to T. Bladen, "RADTRAD Code Verification and Validation" 29 Perry letter PY-CEI/NRR-1510L MLILI 30 G. Burley, "Evaluation of Fission Product October 5, 1971 ] L]

EI Release and Transport for Fuel Handling Accident," Radiological Safety Branch, Division of Reactor licensing

Enclosure 5 PY-CEI/NRR-2609L Page 6 of 76 Page vi F "Wner 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 TABLE OF CONTENTS SUBJECT PAGE COVERSHEET:

OBJECTIVE OR PURPOSE SCOPE OF CALCULATION

SUMMARY

OF RESULTS/CONCLUSIONS IMPACT ON OUTPUT DOCUMENTS DOCUMENT INDEX iii CALCULATION COMPUTATION 1 METH O D O F A NA LYS IS ................................................................................................................................ 1 2 AC CEPTANC E C RITER IA .............................................................................................................................. 4 3 A S S UMPTIO NS .............................................................................................................................................. 4 4 D ETAILED CALC ULATIO NS ........................................................................................................................ 10 4.1 D evelopm ent of Input Files ................................................................................................................... 11 4.1.1 M ain Input File, pnppfha.psf ....................................................................................................... 11 4.1.2 A uxiliary Input Files ...................................................................................................................... 12 4.1.2.1 Input File on Release Fraction and Timing, pnpp fha.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 RADTRA D Code .................................................................................................................... 13 4.3 Results of the RADTRAD Run for the Base Case ........................................................................... 13 4 .4 Sensitivity A nalysis ................................................................................................................................ 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 CO MPUTER INPUT AND O UTPUT ............................................................................................................. 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, pnpp_fha.dcf: Dose Conversion Factors . RADTRAD Output File for PNPP FHA, pnppfha.out . RADTRAD Output File for Sensitivity Case 1: Effect of CR Isolation and Fresh Air Intake Attachment 7. RADTRAD Output File for Sensitivity Case 2: Effect of CR Isolation and Recirculation Filtering

Enclosure 5 PY-CEI/NRR-2609L Page 7 of 76 Page vii I= 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 LIST OF TABLES Table 1. Dose Conversion Factors [DIN # 14, 15, 16] ........................................................................................ 6 3

Table 2. Atmospheric dispersion values (X/Q) used (s/m ) ..................................................................................... 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 15 Table 6. Radionuclides Decay Data Used for the Analysis [DIN # 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 2

Figure 1. PN PP FHA Release M odel .......................................................................................................................

Enclosure 5 PY-CEI/NRR-2609L Page 8 of 76 r

Perry Nuclear eCALCULATION COPUTATION anwe. Guide PNPP No. 6077 Rev. 2t7m01 [ArnaFA Nsd NEI-0341 INITIATING DOCUMENT CALCULATION TYPECAULTONO Project 99-001-31 1AE Analysis. 3.2.15.14, Rev. 0 SITLE/

SUBJECT:

Fuel Handling Accident Using Alternative Source Term fulhnln ciet(H)isd otimn Plant(PP)uigNR ttePryNcerPower -45 CALCULATION COMPUTATION I METHOD OF ANALYSIS The RADTRAD computer code was used to determine the toteq room, exclusion area boundary (EAB), and outer boundary of alentent (TEDaE) at the control 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 I E-9 Ci/MWt (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 (X/Q 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) (§4.4.2).

Equation 1 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]

d K K K

-Tn,, = Sni + j=1 X/.F-NnJ

+ j=1 7_ Fi,jl-nn,i Jn,vAvNv, -Nn,i j=1 n,iFn,recn,i Equation 1 j~i j~i

Enclosure 5 PY-CEI/NRR-2609L Page 9 of 76 FirstnergyPage 2

_r Q:nr; CALCU LATIrON Perry Nuclear Power Plant PNPP No. 6077 Rev. 2/7101 NE1-0341 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 Release (instantaneous) Recirc. Filter Con Room--*. LControl Room - Control Room Air Intake C Ro Exhaust EAB LPZ Environment Figure 1. PNPP FHA Release Model' where:

= number of atoms of nuclide n in compartment i S.o

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

K = Number of compartments defined in the RADTRAD model Fre = volume-normalized air flow rate from compartmentjto i (s-) (F 1 >_0)

- fraction of nuclide v that decays to nuclide n (dimensionless)

- radioactive decay constant of nuclide n (s-1 ), which is calculated from radioactive half life, (ti1 2_), (s) as shown in Equation 2 77n.i = filter efficiency for nuclide n in compartment i (dimensionless)

Fn,rec volume-normalized recirculation air flow rate in compartment i (SI).

_= n(2)

Equation 2 TtT1 2T 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).

Enclosure 5 PY-CEI/NRR-2609L Page 10 of 76 TTEUBJECT Perry Nuclear Fe lCALCULATION T PwrPat PNPP No. 6077 Rev. 217/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

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

SUBJECT:

Fuel Handling Accident Using Alternative Source Term

"* Decay rate of radionuclide n in compartment i.

"* 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 = : (EDEn, +CEDEL) Equation 3 n=l1 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]

T.

EDEenv =DCFEDE,nfAn(/Qedt Equation 4 0 Y-'env tEuio4 where:

EDEenv = EDE (cloudshine dose) due to nuclide n in the environment at given location (rem) n rem.mC DCFEDE,, = user-provided EDE (cloudshine) dose conversion factor for nuclide (Ci.s)

T = duration of analysis (s)

An = activity release rate of nuclide n (Ci) 3 (ZQ) = user-provided atmospheric relative concentration at EAB or LPZ (s/m )

lyIenv t = time (s)

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

An = Nn An 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]

CEDEnenv T '

=DCFCEDEn nAn (ZYQ' 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).

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

Enclosure 5 PY-CEI/NRR-2609L Page 11 of 76 4

Q FkrstneWPage CALCULATION 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 TITLE/

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]

ECR = CFEDE,n T~R EDE F JACR n OFdt Equation 7 where:

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

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

ACR,, = activity of nuclide n in the control room at time t (Ci)

VCR = Volume of the control room (mi3)

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

G F - VCRR 1173 0 338 Equation 8 where:

VcR - Volume of the control room (ft3 ). (Note the difference of units of this variable in Equation 7 and Equation 8.)

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

CEER - DCFCEDE'n T Equation 9 CEDEcR 0IACRn OF dt

-BRCR VCR where:

CEDECR = CEDE due to nuclide n in the control room (rem) 3 BRcR = user-provided breathing rate for the control room operator (m I/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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

[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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 of 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 5 PY-CEI/NRR-2609L Page 12 of 76 Page 5

jyCALCULATION Srst~nr Perry Nuclear Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 1AE Analysis ,13.2.15.14, 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 radionuci~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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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, Dosdat825.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 3.13.3 This calculation applies the recommended breathing rates: 3.5E-4 m /s for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 1.8E-4 m 3/s from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 2.3E-4 m 3/s thereafter. However, since the release was conservatively modeled as an instantaneous release, the analysis used an effective breathing rate of 3.5E-4 m 3/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 # 11]

Enclosure 5 PY-CEI/NRR-2609L Page 13 of 76 Page 6 FirstCALCULATION 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 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-m3 /Ci-s) (Sv-m 3/Bq-s) (rem/Ci) (Sv/Bq) 1 BR82 4.810E-01 1.300E-13 1.528E+03 4.130E-10 2 BR 83 1.413E-03 3.820E-1 6 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 8 RB 87 6.734E-06 1.820E-1 8 3.234E+03 8.740E-1 0 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. [Table 6, DIN # 21 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 5 PY-CEI/NRR-2609L Page 14 of 76 Page 7 TITLE/

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 #111 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 1 E-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 (§4.2, 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 1E-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 5 PY-CEI/NRR-2609L Page 15 of 76 TLrs JEC Fe CALCULATION Term Perry Nuclear PwrPat 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 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of this event, the breathing rate of this individual was assumed to be 3.5E-4 m 3 /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. Guide 1.183): [DIN # 21 As given in §2, this calculation applies acceptance criteria in Table 6 of Reg. Guide 1.183 and I OCFR50.67 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 (xIQ) 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 dispers.ion values (T/Q) used (s/m 3)

Location x/Q (s/m 3 ) Reference Control Room 3.5E-4a DIN # 5 EAB 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, was used in this calculation because FHA release, and thus the activity intake to the control room was assumed to be instantaneous.

Enclosure 5 PY-CEI/NRR-2609L Page 16 of 76 9

3.17.1n S~%

W ePage CAL CU LATI"O N Perry Nuclear PwrPat PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

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

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/IMTU provided that the maximum linear heat generation rate did not exceed 6.3-kW/ft peak rod average power for burnup 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 (1, Br) 5%

Alkali Metals' (Cs, Rb) 12%

0Alkali 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 Csl 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 DF 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 5 PY-CEI/NRR-2609L Page 17 of 76 Page 10

""rZn CALCULATION Perry Nuclear PwrPat PNPP No. 6077 Rev. 27/701 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 oc).

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 pnpp_fha.rft, pnppfha.nif, and pnppfha.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 5 PY-CEI/NRR-2609L Page 18 of 76 TITLE/

SUBJECT:

Fl HPage 11 Perry Nuclear PwrPat PNPP No. 6077 Rev. 217/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION, NO.

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

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 pnppfha.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 (ft3 )

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 1 below the line with "Source Term" identifies that the whole source term was placed in only one compartment. Numbers 1 and 1 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 pnpp_fha.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 1 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 5 PY-CEI/NRR-2609L Page 19 of 76 Page 12 Srtnrcl CALCULATION Perry Nuclear Power Plant PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 1AE Analysis .13.2.15.14, Rev. 0 TITLE/

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 <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />). The release rate from the containment was chosen arbitrarily as 1El 0 %-volume/day. 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 (X/Q) of 4.3E-4 and 4.8E-5 s/mi3 were used for EAB and the outer boundary of LPZ (Table 2). For both 3

offsite locations the Reg. Guide 1.183 recommended breathing rates were used: 3.5E-4 m /s for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 1.8E-4 m 3/s from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 2.3E-4 m 3l/s 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.

3 Atmospheric dispersion value (X/Q) of 3.5E-4 s/mi 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% between 1 and 3

4 days, and 40% from 4 to 30 days (§3.14.6). A constant breathing rate of 3.5E-4 m /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, pnppfha.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 3 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 5 PY-CEI/NRR-2609L Page 20 of 76 13 Sr _ Z; FirstneWPage CALCULATION Perry Nuclear Power Plant PNPP No. 6077 Rev. 27/101 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 4.1.2.2 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown were ignored. In addition three isotopes, BR 84, 1128, and XE138, which had activity less than 1 E-9 CiMWt at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 CS135, which are daughter products of radioactive decay of KR 87, KR 88, and XE135M 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown. [DIN # 10]

Table 5 shows the calculation of FHA source term. The core inventory per unit power was multiplied by the 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)*((1 51 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, Indexr.inp, which was provided with the MACCS2 computer code package. [DIN # 15]

4.1.2.3 Input File on Dose Conversion Factors, pnpp_fha.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-m 3/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_Tha.rft, pnpp_fha.nif, and pnpp_fha.dcf were used as input to the code. These files are given in Attachment 1 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 5 PY-CEI/NRR-2609L Page 21 of 76 Page 14 T rI tEne C FulHLdUnAccd Perry Nuclear PwrPat PNPP No. 6077 Rev. 2/7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

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

SUBJECT:

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

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Fraction of Rods Failed)* FHA (Ci/MWt) Gap Fraction (Peaking Factor) Pool DFa Release No. Isotope [DIN # 10] [Table 3] (MWt) [§3.17.2, §3.17.31 Activity (Ci) 1 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 KR 88 5.1150E+01 5% 18.029 1 4.6109E+01 8 RB 87 N/Ab ý.o l-% 18.029 00 0 /4W4 1.7 0z 9 RB 88 5.7120E+01c &/ 1 90, 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 23 CS135 2.7050E-02c W 'zr 18.029 00 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 5 PY-CEI/NRR-2609L Page 22 of 76 TLE/ C: F Perry Nuclear l d CALCULATIONrcegTerm PwrPat PNPP No. 6077 Rev. 2/7101 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 1AE 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]

Isotope Daughter 1 Daughter 2 Radioactive Half-life, t 1,2 No.

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 KR 85M KR 85 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+10 y 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 8.04 d 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 52.6 m 3.1560000E+03 16 1135 XE135M 0.154 XE135 0.846 6.61 h 2.3796000E+04 17 XE129M 8 d 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-1 35 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 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 5 PY-CEI/NRR-2609L Page 23 of 76 S~~~CALCULATIONPae1 Perry Nuclear PwrPat PNPP No. 6077 Rev. 2(7/01 NEI-0341 INITIATING DOCUMENT CALCULATION TYPE CALCULATION NO.

Project 99-001-31 1AE 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 1 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 onc~the available activity is introduced into the control room. This is considered conservative as it maximizes the d4Qto the control room operators. No inleakage of air --1-0Z 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 . 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 wns 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 case/"and sensitivity case 1 A-.

results. The control room TEDE calculated for starting recirculation filtering after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following control room 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 [Table 7] 1.03 Sensitivity Case 1: Effect of CR Isolation and Fresh Air Intake [Attachment 61 2.81 Sensitivity Case 2: Effect of CR Isolation and Recirculation Filtering [Attachment 7] 2.97 5 COMPUTER INPUT AND OUTPUT The main input file for the base case, pnpp_fha.psf, and the three auxiliary input files, pnpp_fha.rft, pnppfha.nif, and pnppfha.dcf were used as input to the code. These files are given in Attachment I through Attachment 4.

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

Enclosure 5 PY-CEI/NRR-2609L Page 24 of 76 Cabc. 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 between 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 z 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 ff 1forg Equation 10 inorg Forg 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.

D1 finorg Equation 11 f

,norg 1 org OF DForg Burley calculated the decontamination factor for inorganic (elemental) iodine as given by Equation 12. [DIN # 30]

DFinorg = exp ke, Equation 12 (b Vb

Enclosure 5 PY-CEI/NRR-2609L Page 25 of 76 Calc. No. 3.2.15.14, Rev. 0 Appendix 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.

DF1,. 1ý = exp(CH) Equation 13 6 1 ke-f 1 Equation 14 C=-

db Vb Using Equation 13, the decontamination factor for inorganic halogen species for a given pool depth can be expressed as given in Equation 15.

DFinorg,0 = exp(CH 0 ) Equation 15 Equation 16 is obtained by substituting for C from Equation 15 in Equation 13.

DFjt.ng = inogo0Ho 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%)/(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 5 Radtrad 3.02 1/5/2000 PY-CEI/NRR-2609L 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.OOOOE+00 0

0 0

0 0

Compartment 2:

Environment 2

0.OOOOE+00 0

0 0

0 0

Compartment 3:

Control Room 1

3.6707E+05 0

0 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:

CaIc. No. 3.2.15.14, Rev. 0 Attachment I Page 2 of 4 Enclosure 5 Source Term: pY-CEI/NRR-2609L 1 Page 27 of 76 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

O.OOOOE+00 0.9985E+00 0.0015E+00 1.0000E+00 Overlying Pool:

0 o.OOOOE+00 0

0 0

0 Compartments:

3 Compartment 1:

0 1

0 0

0 0

0 0

0 Compartment 2:

0 1

0 0

0 0

0 0

0 Compartment 3:

1 1

0 0

0 0

0 0

0 Pathways:

3 Pathway 1:

0 0

0 0

0 0

0

Caic. No. 3.2.15.14, Rev. 0 Atcmn 1 Aftachment ae3o Page 3 of 4 Enclosure 5 0 PY-CEU/NRR-2609L 0 Page 28 of 76 0

.1 2

0. 0000E+00 1. OOOOE+10

7. 2000E+02 0. OOOOE+00 0

Pathway 2:

0 0

0 0

0 1

3

0. OOOOE+00 6.6000OE+03 o .OOOOE+00 o .OOOOE+00 o .OOOOE+00 o0.0001E+00 5. 4000E+03 0. OOOOE+00 o .OOOOE+O0 o.-OOOOE+O0 7 .2000E+02 5. 4000E+03 o .OOOOE+00 o .OOOOE+O0 o .OOOOE+OO 0

0 0

0 0

0 Pathway 3:

0 0

0 0

0 1

3 0.OOOOE+O0 6.6000E+03 o .OOOOE+00 0. OOOOE+00 o .OOOOE+O0 0.OOO1E+OO 5.4000E+03 o .OOOOE+00 o .OOOOE+00 0. OOOOE+00 7.2000E+i02 5-4000E+03 o .OOOOE+0O 0. OOOOE+00 0. OOOOE+00 0

0 0

0 0

0 Dose Locations:

3 Location 1:

Exclusion Area Boundary 2

1 2

0. OOOOE+00 4. 3000E-04

2. OOOOE+00 0. OOOOE+O0 1

3

0. OOOOE+0O 3. 5000E-04

8. OOOOE+O0 1. 8000E-04

2. 4000E+01 2. 3000E-04 0

Calc. No. 3.2.15.14, Rev. 0 Attachment 1 Page 4 of 4 Enclosure 5 Location 2: PY-CEI/NRR-2609L Outer Boundary of the LPZ Page 29 of 76 2

1 2

0.OOOOE+00 4.8000E-05 2.0000E+00 o.0000E+00 1

3 0.OOOOE+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 0.0000E+00 3.5000E-04 7.2000E+02 o.OOOOE+00 1

4 0.OOOOE+00 1.OOOOE+00 2.4000E+01 6.0000E-01 9.6000E+01 4.OOOOE-01 7.2000E+02 0.0000E+00 Effective Volume Location:

1 2

O.0000E+00 3.5000E-04 2.OOOOE+00 O.0000E+00 Simulation Parameters:

4 O.OOOOE+00 2.5000E-02 8.0000E+00 1.0000E-01 2.4000E+01 4.OOOOE-01 7.2000E+02 O.0000E+00 Output Filename:

1 1

1 0

0 End of Scenario File

Caic. No. 3.2.15.14, Rev. 0 Attachment 2 Atcmn Page ae1o I of 1 Release Fraction and Timing Name:

E n c-lo~s u/reR5 -5 Perry FHA Duration (h) Page 30 of 76 o0.0001E+00 o .OOOOE+00 0. OOOOE+00 0 000 OE+OO Noble Gases:

o.1000E+01 0. OOOOE+00 o .OOOOE+00 0 .OOOOE+00 Iodine:

0. l000E+01 o .OOOOE+00 o .OOOOE+00 0. OOOOE+00 Cesium:

o0.1000E+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. OOOOE÷OO o .OOOOE+00 0. OOOOE+00 0. OOOOE+OO Barium:

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

0. OOOOE+00 0. OOOOE+00 0. OOOOE+00 0. OOOOE+00 Cerium:

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

0. OQOOE+0O 0. OOOOE+00 0. OOOOE+00 0. 0000E+00 Non-Radioactive Aerosols (kg):

0.OOOOE+00 0.OOOOE+00 0 . 000E+00 0. OOOOE+00 End of Release File

Calc. No. 3.2.15.14, Rev. 0 Attachment 3 Page 1 of 2 Enclosure 5

0. 1000E+01 PY-CEI/NRR-2609L Nuclide Inventory Name: Rb-87 none 0.OOOOE+00 Page 31 of 76 Perry FHA 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 1.27080000000E+05 Rb-88 0.1000E+01 0.8200E+02 none 0.OOOOE+00 5.5845E-01 none 0OOOOE+00 none 0.OOOOE+00 Nuclide 008:

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

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

none 0.OOOOE+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.OOOOE+00 0.8500E+02 none 0.OOOOE+00 7.4911E+02 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 011:

none 0.OOOOE+00 1-130 none 0.OOOOE+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.OOOOE+00 2.4186E-02 none 0.OOOOE+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 3 Page 2 of 2 Enclosure 5 Nuclide 013: 4.5316800000E+05 PY-CEI/NRR-2609L 1-132 0.1330E+03 Page 32 of 76 2 4.6037E+04 8.2800000000E+03 none 0.OOOOE+00 0.1320E+03 none 0.0000E+00 1.4486E+02 none 0.OOOOE+00 none 0.0000E+00 Nuclide 020:

none O.OOOOE+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.OOOOE+00 6.0037E-06 none 0.OOOOE+00 none 0.OOOOE+00 Nuclide 022:

none 0.0000E+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 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 0.0000E+00 End of Nuclear Inventory File none 0.OOOOE+00 none 0.0000E+00 Nuclide 018:

Xe-131m 1

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

Xe-133 1

Cale. No. 3.2.15.14, Rev. 0 Attachment 4 Page 1 of 5 Enclosure 5 fgr 11 and 12 dcfs PY-CEI/NRR-2609L Page 33 of 76 9 ORGANS DEFINED IN THIS FILE:

GONADS BREAST LUNGS RED MARR BONE SUR THYROID REMAINDER EFFECTIVE SKIN(FGR) 23 NUCLIDES DEFINED IN THIS FILE:

Br-82 D Br-83 H Kr-83m Kr-85 Kr-85m Kr-87 Kr-88 Rb-87 Y Rb-88 1-129 Y 1-130 H 1-131 D 1-132 D 1-133 D 1-134 M 1-135 D Xe-129m Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m Cs-135 Y CLOUDSHINE GROUND GROUND GROUND INHALED INHALED INGESTION SHINE 8HR SHINE 7DAY SHINE RATE ACUTE CHRONIC Br-82 GONADS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+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 O.OOOE+00 O.OOOE+00 O.OOOE+00 0-000E+00 O.OOOE+00 O.OOOE+00 THYROID O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0-000E+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 1.300E-13 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 4.130E-10 O.OOOE+00 SKIN(FGR) O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0-000E+00 0.000E+00 O-OOOE+00 Br-83 GONADS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+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 O.OOOE+00 O.OOOE+00 O.OOOE+00 0-000E+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 3.820E-16 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 2ý410E-11 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 Kr-83m GONADS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0.000R+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0-000E+00 O.OOOE+00 O.OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 RED MARR O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00

Enclosure 5 Calc. No. 3.2.15.14, Rev. 0 Aftachment 4 PY-CEI/NRR-2609L Page 2 of 5 Page 34 of 76 BONE SUR O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. 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 REMAIN-DER O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 1.500E-18 O.OOOE+00 O-OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 SKIN(FGR) O-OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 Kr-85 GONADS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 LUNGS 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 RED MARR 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BONE SUR O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 THYROID 0. OOOE+00 O.OOOE+00 0 . OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 REMAINDER 0 - OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 1. 190E- 16 O.OOOE+00 O-OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 SKIN(FGR) O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O-OOOE+00 0. OOOE+00 O.OOOE+00 Kr-85m GONADS 0. OOOE+00 0. OOOE+00 0 . OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BREAST O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 LUNGS 0 - OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0 - OOOE+00 0. OOOE+00 O.OOOE+00 RED MARR 0 . OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BONE SUR 0 . OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0 - OOOE+00 0. OOOE+00 O.OOOE+00 THYROID 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 REMAINDER O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 7.480E-15 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 SKIN(FGR) 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 Kr-87 GONADS 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BREAST 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 LUNGS 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 RED MARR 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. 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 0. OOOE+00 O.OOOE+00 THYROID 0. OOOE+00 0. OOOE+00 0 . OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 REMAINDER 0. OOOE+00 0. OOOE+00 0 . OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 4.120E-14 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 SKIN(FGR) 0. OOOE+00 0.. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 Kr-88 GONADS 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BREAST 0 . OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 LUNGS 0 . OOOE+00 0 . OOOE+00 0 . OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 RED MARR 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0 . OOOE+00 0. OOOE+00 BONE SUR 0. OOOE+00 0 . OOOE+00 O.OOOE+00 O-OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 THYROID 0. OOOE+00 O.OOOE+00 0 - OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 REMAINDER 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 EFFECTIVE 1.020E-13 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 SKIN(FGR) O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 Rb-87 GONADS O.OOOE+00 0 . OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 BREAST 0. OOOE+00 O.OOOE+00 0 . OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 LUNGS 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 RED MARR 0. OOOE+00 O.OOOE+00 0 . OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 BONE SUR 0. OOOE+00 OýOOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 THYROID 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 REMAINDER O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O-OOOE+00 EFFECTIVE 1.820E-18 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 8.740E-10 O.OOOE+00 SKIN(FGR) O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 Rb-88 GONADS 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 BREAST 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 LUNGS 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 RED MARR 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0 . OOOE+00 BONE SUR 0. OOOE+00 0 . OOOE+00 0. 000E+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 THYROID 0 . OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 REMAINDER O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00

Enclosure 5 Ca1c. No. 3.2.15.14, Rev. 0 Attachment 4 Page 3 of 5 PY-CEI/NRR-2609L Page 35 of 76 EFFECTIVE 3.360E-14 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 2-260E-11 O.OOOE+00 SKIN(FGR) O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0.000E+00 O.OOOE+00 O.OOOE+00 1-129 GONADS 0. OOOE+00 O.OOOE+00 0 - OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O-OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 RED MARR 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 BONE SUR 0. OOOE+00 0. OOOE+OG 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 THYROID O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 REMAINDER O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 EFFECTIVE 3.800E-16 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 4.690E-08 O.OOOE+00 SKIN(FGR) 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 1-130 GONADS 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0 . OOOE+00 O.OOOE+00 0. OOOE+00 BREAST O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0 . OOOE+00 0. OOOE+00 0. OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 RED MARR O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 BONE SUR O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 THYROID 0 - OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 REMAINDER 0-000E+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 1. 040E-13 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 7. 140E-10 O.OOOE+00 SKIN(FGR) O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 1-131 GONADS 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BREAST 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 LUNGS 0. OOOE+00 0-000Z+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 RED MARR 0. 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 0 - OOOE+00 O.OOOE+00 0 . 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 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 EFFECTIVE 1.820E-14 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 8.890E-09 0. OOOE+00 SKIN(FGR) 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 1-132 GONADS O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BREAST 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 LUNGS 0 . OOOE+00 O.OOOE+00 0 . OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 RED MARR 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 BONE SUR O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 THYROID 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 REMAINDER 0 . OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 EFFECTIVE 1.120E-13 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 1.030E-10 O.OOOE+00 SKIN(FGR) 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 1-133 GONADS 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 LUNGS 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 RED MARR 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BONE SUR 0. OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0 . OOOE+00 THYROID 0 . OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 REMAINDER 0. OOOE+00 O.OOOE+00 0-000E+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 2.940E-14 0. OOOE+00 0 . OOOE+00 O.OOOE+00 0. OOOE+00 1. 580E-09 0 . OOOE+00 SKIN(FGR) 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 1-134 GONADS O.OOOE+00 0. OOOE+00 O-OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 BREAST 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 LUNGS O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 RED MARR 0. OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0 . OOOE+00 BONE SUR 0. OOOE+00 0 . OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 0 . OOOE+00 THYROID 0 . OOOE+00 0. OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0 . OOOE+00 0 . OOOE+00 REMAINDER 0 . OOOE+00 O.OOOE+00 0. OOOE+00 0. OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 EFFECTIVE 1. 3OOE-13 O.OOOE+00 0 . OOOE+00 0. OOOE+00 O.OOOE+00 3 .550E-11 O.OOOE+00 SKIN(FGR) 0 . OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 0. OOOE+00 O.OOOE+00 1-135

oo+aoooo oo+aooo*o oo+2ioooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo SDNnIi oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo ISVaUS oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo SCIVNOE)

UlSET -GX oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o (ýIoa)NDIS 00+2[000*0 00+S000*0 00+HOOO'O 00+HOOO"O 00+H000*0 00+aOOO'O VT-ao6l'T sAiLDal4lqa oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo 'dsciNivwszl oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo C[IOd7,H.L oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aooo*o zms aNola oo+aoooo oo+aooo*o oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aoooo 'dUVW GSU oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o sf)mni oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aoooo ISVadEl oo+aooo*o oo+aooo*o oo+aooo*o oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo s(avNoE)

SET-;DX oo+aoooo oo+aooo*o oo+aoooo oo+aooo*o oo+aoooo oo+aoooo oo+sooo*o (IdDi)NEAS oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o 00+aOOO'O Sl-aOLUT SAIjosdda oo+aooo*o oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aoooo 'daciNivwsd oo+aooo*o oo+aoooo oo+aooo-o oo+aoooo oo+aoooo oo+aoooo oo+aoooo CIIOýýH.L oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+zooo*o oo+aoooo ýMS aN06 oo+aooo*o oo+aoooo oo+aoooo oo+aooo-o oo+aoooo oo+aooo*o oo+aoooo duvw cia'd oo+aoooo oo+aooo*o oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aoooo SE)NnIi oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo ilsvaua oo+aoooo oo+aooo*o oo+aoooo oo+aooo-o oo+aooo*o, oo+aoooo oo+aoooo SCIVN00 WEET-ax oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o OdSd)NDIS oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo sT-aogsT aAiioaqlqa oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aooo"o oo+aoooo oo+aooo"o oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo GIOUARL oo+aoooo oo+aooo*o oo+aoooo oo+aooo"o oo+aoooo oo+aooo*o oo+aooo*o ýms aNos oo+aooo*o oo+aoooo oo+aooo*o oo+aoooo oo+aooo*o oo+aoooo oo+aoooo EUVW CIEd oo+aooo"o oo+aoooo oo+aoooo oo+aoooo oo+aooo"o oo+aoooo oo+aoooo Sf:)Nfl'I oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo JSVa'dg oo+soooo oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+Eiooo*o oo+aoooo SCIVNOE)

ZET--GX oo+sooo"o oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aooo*o oo+aoooo (d0d)NIXS oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o qi-ao68,E aAij.:)Eada oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aooo*o 'daaNiVWa-d oo+soooo oo+aoooo oo+aoooo oo+aooo"o oo+aoooo oo+aoooo oo+aoooo CIIOUAHL oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aooo*o ýMS SNOU oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o oo+aooo*o oo+aooo*o oo+aoooo uldvw cia'd oo+aoooo oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aoooo Sf)NnII oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo i-svada oo+aoooo oo+aooo*o oo+aooo*o oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o SaVNOS UITET-@X oo+zoooo oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aooo*o oo+aooo*o (HEUMIAS oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o si-aogoi aAILoadaa oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+Eiooo*o oo+aooo*o oo+aoooo oo+aooo*o oo+aooo*o ClIOUAH1 oo+aoooo oo+aooo*o oo+aoooo oo+soooo oo+aooo"o oo+aoooo oo+aoooo ",tms aN01a oo+aooo*o oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o UdVW Cla'd oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo"o oo+aoooo oo+aooo*o SSNEIII oo+aoooo oo+aooo*o oo+Ecooo*o oo+aooo*o oo+aoooo oo+aooo*o oo+aoooo ILSVSUS oo+aoooo oo+aoooo oo+Eioooo oo+aooo"o oo+aoooo oo+aooo*o oo+sooo"o SCIVNO-D ul6zl-;Dx oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aooo"o oo+aoooo oo+soooo ('df)d)NI)IS oo+aooo*o oT-aozE*E oo+aoooo oo+aoooo oo+aoooo oo+aooo*o 'vT-aV6Z*8 aAIlDa.iaa oo+sooo*o oo+aooo*o oo+Eioooo oo+aooo*o oo+aooo"o oo+aooo*o oo+aoooo uaciNIVNS'd oo+aoooo oo+aooo*o oo+aoooo oo+aoooo oo+aooo*o oo+aooo*o oo+aooo"o CII0'd2Ml oo+aoooo oo+soooo oo+aoooo oo+aooo*o oo+aooo*o oo+aoooo oo+Eioooo UnS SNOZ oo+aooo*o oo+aooo*o oo+aoooo oo+aooo*o oo+Eioooo oo+aoooo oo+aooo*o uzivw cia*d oo+aooo*o oo+aoooo oo+aoooo oo+aoooo oo+aooo"o oo+aoooo oo+aoooo SSNrl'I oo+aoooo oo+aoooo oo+aoou*o oo+aoooo oo+aooo"o oo+aooo*o oo+aoooo Lsva'da oo+aoooo oo+aoooo oo+aoooo oo+aoooo oo+aooo*o oo+aoooo oo+aoooo SCIVNOS 9L 10 96 Of)-Rd 9 jo t, a5ed ý60R-HUN/130-Ad 9 ejnsOl3u3 t, IUGWLloeuv O'A9H '17V9L'Z'S -ON _31eO

Calc. No. 3.2.15.14, Rev. 0 Affachment 4 Enclosure 5 Page 5 of 5 PY-CEI/NRR-2609L Page 37 of 76 RED MARR O.OOOE+00 O-OOOE+00.0.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BONE SUR O.OOOE+00 0.000F+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 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 O.OOOE+00 O.OOOE+00 O.OOOE+00 Cs-135 GONADS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 BREAST O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 LUNGS O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.OOOE+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 O.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 O.OOOE+00 O.OOOE+00 O.OOOE+00 EFFECTIVE 5.650E-19 O.OOOE+00 O.OOOE+00 O.OOOE+00 O.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

Caic. No. 3.2.15.14, Rev. 0 Attachment 5 Page I of 16 Enclosure 5 PY-CEI/NRR-2609L

1. 4##f######4####ftf####################################################ff####

Page 38 of 76 RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. f############################ft###############################

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. f#######################

File information Plant file name = pnppfha.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\pnpp-fha.dcf f#### #####f ft ft ft ###### ft ft #f###

ft ft ft ft ft #f ## # #f ft ft ft ft fttff #tf ft ft ft # #tfft#f ft # #

ff##### ft#ft# ft ft # # ft##

1. t ft # ft ft ft # ft ftt# ft ft ft ft

1. t #f#f# #t Radtrad 3.02 1/5/2000 perry fha 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.OOOOE+00 0

0 0

0 0

Compartment 2:

Environment 2

0.OOOOE+00 0

0 0

0 0

Compartment 3:

Control Room 1

3.6707E+05 0

0 0

0 0

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 2 of 16 Pathways : Enclosure 5 PY-CEI/NRR-2609L 3

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.OOOOE+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

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 0

0 0

0 Compartment 2:

0 1

0 0

0 0

0 0

0

Caic. No. 3.2.15.14, Rev. 0 Attachment 5 Page 3 of 16 Compartment 3: Enclosure 5 PY-CEI/NRR-2609L 1 Page 40 of 76 0

0 0

0 0

Pathways:

3 Pathway 1:

0 0

0 0

0 0

0 0

0 0

1 2

o .OOOOE+O0 1. OOOOE+10

7. 2000E+02 0. OOOOE+00 0

Pathway 2:

0 0

0 0

0 1

3 o .OOOOE+00 6. 6000E+03 o .OOOOE+00 o .OOOOE+00 o .OOOOE+00 o0.0001E+00 5.4000E+03 o .OOOOE+00 o .OOOOEO00 o .OOOOE+00 7.2000E+02 5.4000E+03 o.-OOOOE+00 o .OOOOE+OO o .OOOOE-i00 0

0 0

0 0

0 Pathway 3:

0 0

0 0

0 1

3 o -OOOOE+00 6. 6000E+03 o .OOOOE+00 o .OOOOE+0O o .OOOOE+00 0 .0001E+00 5. 4000E+03 o .OOOOE+00 0. OOOOE+00 o .OOOOE+0O 7 .2000E+02 5. 4000E+03 o .OOOOE+00 0. OOOOE+OO 0. OOOOB+00 0

0 0

0 0

0

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

3 PY-CEI/NRR-2609L Location 1: Page 41 of 76 Exclusion Area Boundary 2

2 o.OOOOE+00 4.30OOE-04 2.OOOOE+00 o.OOOOE+00 1

3 o.OOOOE+00 3.5000E-04 8.OOOOE+00 1.8000E-04 2.4000E+01 2.3000E-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 o.OOOOE+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 0.OOOOE+00 3.5000E-04 7.2000E+02 0.0000E+00 1

4 O.OOOOE+00 1.OOOOE+00 2.4000E+01 6.OOOOE-01 9.6000E+01 4.OOOOE-01 7.2000E+02 O.OOOOE+00 Effective Volume Location:

1 2

O.OOOOE+00 3.5000E-04 2.0000E+00 O.0000E+00 Simulation Parameters:

4 O.0000E+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 5 PY-CEI/NRR-2609L

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. i################ Paqe 42 of 76 RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48 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 1.OOOOE+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

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 6 of 16 Enclosure 5 PY-CEI/NRR-2609L

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 4################################################### 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 RELEASE NAME = Perry FHA Release Fractions and Timings GAP EARLY IN-VESSEL 0.0001 hrs 0.0000 hrs NOBLES .OOOOE+00 0.OOOOE+00 IODINE 1.OOOOE+00 0.OOOOE+00 CESIUM 1.0000E+00 0.OOOOE+00 TELLURIUM 0.OOOOE+00 0.OOOOE+00 STRONTIUM 0.OOOOE+00 0.OOOOE+00 BARIUM 0.OOOOE+00 0.0000E+00 RUTHENIUM 0.0000E+00 0.OOOOE+00 CERIUM 0.OOOOE+00 0.OOOOE+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 (% / day) 0.OOOOE+00 1.OOOOE+10 7.2000E+02 0.OOOOE+00 Pathway number 2: Unfiltered Environment to CR Pathway Filter: Removal Data Time (hr) Flow Rate Filter Efficiencies (%)

(cfm) Aerosol Elemental Organic 0.OOOOE+00 6.6000E+03 0.OOOOE+00 0.OOOOE+00 0.0000E+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

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 7 of 16 Enclosure 5 o.0000E+00 6.6000E+03 o.0000E+00 o .OOOOE+00 o .OOOOE+O0 PY-CE4/NRR-2609L 1.0000E-04 5.4000E+03 0.0000E+00 o .OOOOE+O0 o .OOOOE+OO Paqe 44 of 76 7.2000E+02 5.4000E+03 0.0000E+00 0.0000E+00 o .OOOOE+00 LOCATION DATA Location Exclusion Area Boundary is in compartment 2 Location X/Q Da .ta Time (hr) X/Q (s

• m--3) o.0000E+00 4.3000E-04 2.OOOOE+00 0.OOOOE+00 Location Breathiing Rate Data Time (hr) Breathing Rate (m^3

• sec'-l) 0.0000E+00 3.5000E-04 8.OOOOE+00 1.80OOE-04 2.4000E+01 2.3000E-04 Location Outer Boundary of the LPZ is in compartment 2 Location X/Q Da ta Time (hr) X/Q (s

• mA-3) 0.OOOOE+00 4.8000E-05 2.OOOOE+00 0.OOOOE+00 Location Breathing Rate Data Time (hr) Breathing Rate (mW3

• sec^-1)

O.OOOOE+00 3.5000E-04 8.OOOOE+00 1.8000E-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" 3.50OOE-04 2.OOOOE+00 O.0000E+00 Location Breathing Rate Data Time (hr) Breathing Rate (mW3

• sec^-1)

O.OOOOE+00 3.5000E-04 7.2000E+02 O.OOOOE+00 Location Occupancy Factor Data Time (hr) Occupancy Factor 0.0000E+00 1.0000E+00 2.4000E+01 6.0000E-01 9.6000E+01 4.0000E-01 7.2000E+02 O.OOOOE+00 USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS Time Time step O.OOOOE+00 2.5000E-02 8.OOOOE+00 1.00O0E-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 of 16 Enclosure 5 PY-CEI/NRR-2609L

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. Paae 45 of 76 RADTRAD Version 3.02 run on 11/15/2001 at 16:53:48

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. if i ##############ifi####### ######################

if if

1. # #f if ifif# # #

if if

1. # ## # ##### #f if

1. ifif #if# #if #if i #f if

1. # # # # # #f #f

1. 1. 1. i

1. fff Dose, Detailed model and Detailed Inventory Output Exclusion Area Boundary Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.2456E-01 o.OOOOE+00 1.4377E+00 Accumulated dose (rem) 4.2456E-01 o.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.OOOOE+00 1.6048E-01 Control Room Doses:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 1.2078E-06 0.OOOOE+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 1. 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-!31m 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

Cabc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 9 of 16 Enclosure 5 Xe-135 1.3783E+01 5.3973E-09 2.4076E+16 5.0998E+11 PY-CEI/NRR-2609L Xe-135m 6.5313E-01 7.1700E-12 3. 1984E+13 2.4166E+10 Page 46 of 76 Cs-135 4.7418E-14 4.1157E-14 1.8359E+11 1.7545E-03 Control Room Transport Group Inventory:

Overlying Time (h) = 0.0001 Atmosphere Sump Pool Noble gases (atoms) 1.5969E+19 0.0000E+00 o.0000E+00 Elemental I (atoms) 1.8787E+17 0.OOOOE+00 o.0000E+00 Organic I (atoms) 2.8223E+14 0.OOOOE+00 o.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 o.0000E+00 Deposition Recirculating Time (h) = 0.0001 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 0.0000E+00 Organic I (atoms) 0.OOOOE+00 0.0000E+00 Aerosols (kg) 0.OOOOE+00 0.0000E+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) 0.0000E+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) 4.2594E-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.OOOOE+00 5.2772E-04 Accumulated dose (rem) 4.7547E-02 0.0000E+00 1.6101E-01 Control Room Doses:

Time (h) = 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.1084E-02 0.0000E+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+I1 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+ll 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 Enclosure 5 Kr-85m 2.0462E-02 2.4865E-12 1. 7616E+13 7.5711E+08 PY-CEIUNRR-2609L Kr-87 1.5187E-06 5.3615E-17 3.7112E+08 5.6191E+04 Paqe 47 of 76 Kr-88 5.2861E-03 4.2156E-13 2.8849E+12 1.9559E+08 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-10 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.OOOOE+00 o.OOOOE+00 Elemental I (atoms) 3.2268E+16 0.OOOOE+00 o.OOOOE+00 Organic I (atoms) 4.8474E+13 0.OOOOE+00 o.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 o.0000E+00 Deposition Recirculating Time (h) = 2.0000 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.0DOOE+00 Elemental I (atoms) 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) = 2.0000 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) = 2.0000 Filter Noble gases (atoms) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 Exclusion Area Boundary Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Accumulated dose (rem) 4.2594E-01 0.OOOOE+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 11 of 16 Accumulated dose (rem) 4.7547E-02 0.OOOOE+00 1.6101E-01 Enclosure 5 PY-CEI/NRR-2609L Paqe 48 of 76 Control Room Doses:

Time (h) = 8.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.6824E-03 0.OOOOE+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. 511E+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+11 3.0229E+06 1-135 7.5856E-06 2.1600E-15 9. 6354E+09 2.8067E+05 Xe-129m 1.8897E-07 1.493SE-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 2.1251E-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.OOOOE+00 Elemental I (atoms) 1.6171E+14 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 2.4294E+ll 0.OOOOE+00 o.OOOOE+00 Aerosols (kg) O.OOOOE+00 0.OOOOE+00 0.0000E+00 Deposition 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) 0.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) 0.OOOOE+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 12 of 16 Elemental I (atoms) o.OOOOE+00 Enclosure 5 Organic I (atoms) o.0000E+00 PY-CEI/NRR-2609L Aerosols (kg) o.0000E+00 Paqe 49 of 76 Exclusion Area Boundary Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.OOOOE+00 o.0000E+00 0.OOOOE+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.OOOOE+00 o.OOOOE+00 o.OOOOE+00 Accumulated dose (rem) 4.7547E-02 o.0000E+00 1.6101E-01 Control Room Doses:

Time (h) = 24.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.3641E-05 o.OOOOE+00 8.2602E-04 Accumulated dose (rem) 2.4782E-02 o.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.OOOOE+00 Elemental I (atoms) 1.1895E+08 o.OOOOE+00 0.0000E+00 Organic I (atoms) 1.7869E+05 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Deposition Recirculating Time (h) = 24.0000 Surfaces Filter Noble gases (atoms) 0.OOOOE+00 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 0.000E+00, 0.OOOOE+00 Aerosols (kg) O.OOOOE+00 o.OOOOE+00 Unfiltered Environment to CR Transport Group Inventory:

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

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 13 of 16 Enclosure 5 Control Room Exhaust Transport Group Inventory: PY-CEI/NRR-2609L Pa~qe 50 of 76 Pathway Time (h) = 24.0000 Filter Noble gases (atoms) o.OOOOE+00 Elemental I (atoms) o.OOOOE+00 Organic I (atoms) o.OOOOE+00 Aerosols (kg) o.OOOOE+00 Exclusion Area Boundary Doses:

Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.OOOOE+00 0.OOOOE+00 Accumulated dose (rem) 4.2594E-01 0.OOOOE+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.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) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.6064E-12 0.OOOOE+00 3.3485E-10 Accumulated dose (rem) 2.4782E-02 0.OOOOE+00 1. 0261E+00 Control Room Compartment Nuclide Inventory:

Time (h) = 96.0000 Ci kg Atoms Bq Control Room Transport Group Inventory:

Overlying Time (h) = 96.0000 Atmosphere Sump Pool Noble gases (atoms) 2.5382E-18 0.OOOOE+00 0.OOOOE+00 Elemental I (atoms) 2.9862E-20 0.OOOOE+00 0.0OOOE+00 Organic I (atoms) 4.4860E-23 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Deposition Recirculating Time (h) = 96.0000 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) 0.OOOOE+00 0.OOOOE+00 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) = 96.0000 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) = 96.0000 Filter Noble gases (atoms) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 14 of 16 Enclosure 5 Aerosols (kg) 0.OOOOE+00 PY-CEI/NRR-2609L Paqe 51 of 76 Exclusion Area Boundary Doses:

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.OOOOE+00 1.4424E+00 Outer Boundary of the LPZ Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 Accumulated dose (rem) 4.7547E-02 0.0000E+00 1.6101E-01 Control Room Doses:

Time (h) = 720.0000 Whole Body Thyroid TEDE Delta dose (rem) 3.0169E-40 0.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 o.OOOOE+00 Elemental I (atoms) 1.8750-259 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 2.8167-262 o.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.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 0.OOOOE+00 Organic I (atoms) 0.OOOOE+00 0.OOOOE+00 Aerosols (kg) 0.OOOOE+00 0.OOOOE+00 Unfiltered Environment to CR Transport Group Inventory:

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

Calc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 15 of 16 Enclosure 5

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. -####### PY-CEI/NRR-2609L 1-131 Summary Paqe 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 o. 0000E+00 1. 8242E+02 1.5580E-01 0.525 o. 0000E+00 1. 8245E+02 1.2486E-01 0.775 o. 0000E+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 .OOOOE+00 1.8254E+02 4.1278E-02 2.000 o .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 o.OOOOE+00 1.8256E+02 1.7399E-02 3.000 o.OOOOE+00 1.8256E+02 1.3941E-02 3.250 o.0000E+00 1.8257E+02 1.1171E-02 3.500 o.OOOOE+00 1.8257E+02 8.9506E-03 3.750 o.0000E+00 1.8257E+02 7.1718E-03 4.000 o.0000E+00 1.8257E+02 5.7465E-03 4.250 o.0000E+00 1.8257E+02 4.6045E-03 4.500 o.OOOOE+00 1.8257E+02 3.6894E-03 4.750 o.0000E+00 1.8258E+02 2.9562E-03 5.000 o.OOOOE+00 1.8258E+02 2.3687E-03 5.250 o.OOOOE+00 1.8258E+02 1.8979E-03 5.500 o.OOOOE+00 1.8258E+02 1.5208E-03 5.750 o.OOOOE+00 1.8258E+02 1.2185E-03 6.000 o.OOOOE+00 1.8258E+02 9.7636E-04 6.250 o.OOOOE+00 1.8258E+02 7.8232E-04 6.500 o.OOOOE+00 1.8258E+02 6.2685E-04 6.750 o.0000E+00 1.8258E+02 5. 0227E-04 7.000 o.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 o.OOOOE+00 1.8258E+02 2 .0703E-04 8.000 o.OOOOE+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 o .OOOOE+00 1.8258E+02 3.0797E-05 10.200 o. 0000E+00 1.8258E+02 2.3607E-05 24.000 o. 0000E+00 1.8258E+02 1.1520E-10 96.000 o. 0000E+00 1.8258E+02 2.2331E-38 720.000 o .OOOOE+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 3048E-01 0.OOOOE+00 4.5686E-05 0.275 0.OOOOE+00 1.4415E+00 0.OOOOE+00 6091E-01 0.OOOOE+00 2.2323E-01 0.525 0.OOOOE+00 1.4417E+00 0.OOOOE+00 6094E-01 0.OOOOE+00 3. 8370E-01 0.775 0.OOOOE+00 1.4419E+00 0.OOOOE+00 6096E-01 0.0000E+00 5.1208E-01 1.025 0.00008+00 1.4421E+00 0.OOOOE+00 6097E-01 0.0000E+00 6.1479E-01 1.275 0.OOOOE+00 1.4422E+00 0.OOOOE+00 i.E6099E-01 0.OOOOE+00 6. 9696E-01

CaIc. No. 3.2.15.14, Rev. 0 Attachment 5 Page 16 of 16 Enclosure 5 1.525 0.OOOOE+00 1.4423E+00 O.OOOOE+00 1.6100E-01 o.OOOOE+00 7.6271E-01 PY-CEI/NRR-2609L 1.775 0.OOOOE+00 1.4423E+00 o.0000E+00 1.6101E-01 o.OOOOE+00 8.1533E-01 Paqe 53 of 76 2.000 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1. 61O1E-01 o.OOOOE+00 8.5363E-01 2.250 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 8.8808E-01 2.500 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-0l o.OOOOE+00 9.1564E-01 2.750 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 9.3770E-01 3.000 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 9.5535E-01 3.250 0.OOOOE+00 1.4424E+00 O.OOOOE+00 1.6101E-01 o.OOOOE+00 9.6947E-01 3 .500 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 9.8077E-01 3 .750 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 9.8982E-01 4.000 o.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 9.9706E-01 4 .250 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0028E+00 4.500 0.OOOOE+00 1.4424E+00 O.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0075E+00 4 .750 o0.0000E+00 1.4424E+00 o.OOOOE+00 1.6101E-01 0.OOOOE+O0 1.0112E+00 5.000 o.OOOOE+00 1.4424E+00 O.OOOOE+00 1.6101E-01 o .OOOOE+00 1.0142E+00 5.250 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.61O1E-01 o.OOOOE+00 1.0165E+00 5.500 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.611OE-01 o.0000E+00 1.0184E+00 5.750 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0200E+00 6.000 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.611OE-Ol o.OOOOE+00 1.0212E+00 6.250 0.OOOOE+00 1.4424E+00 O.OOOOE+00 1.6101E-01 o.OOOOE+O0 1.0222E+00 6.500 o.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0229E+00 6.750 0.OOOOE+00 1.4424E+00 0.0000E+00 1.6101E-O1 o.OOOOE+O0 1.0236E+00 7.000 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0241E+00 7.250 0.OOOOE+00 1.4424E+00 O.OOOOE+00 1.611OE-01 o.OOOOE+00 1.0245E+00 7.500 o.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0248E+00 7.750 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 0.0000oE+o0 1.0250E+00 8.000 0.OOOOE+00 1.4424E+00 O.OOOOE+00 1.61.OE-01 o.OOOOE+00 1.0253E+00 8.400 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0255E+00 8.700 o.OOOOE+00 1.4424E+00 O.OOOOE+00 1.6101E-01 0.OOOOE+00 1.0256E+00 9.000 0.OOOOE+00 1.4424E+00 o.OOOOE+00 1.61O1E-01 o.OOOOE+00 1.0257E+00 9.300 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0258E+00 9.600 0.OOOOE+00 1.4424E+00 O.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0259E+00 9.900 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.61.OE-01 o.OOOOE+00 1.0259E+00 10.200 o.OOOOE+00 1.4424E+00 o .OOOOE+O0 1.6101E-01 o.0000E+00 1.0260E+00 24.000 0.000OE+00 1.4424E+00 o.OOOOE+00 1.6101E-01 o.OOOOE+00 1. 0261E+00 96.000 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1. 0261E+00 720.000 0.OOOOE+00 1.4424E+00 0.OOOOE+00 1.6101E-01 o.OOOOE+00 1.0261E+00

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. ff############################

Worst Two-Hour Doses Note: 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 1.3779E-03 0.OOOOE+00 4.7275E-03 Outer Boundary of the LPZ Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.5382E-04 0.OOOOE+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

Caic. No. 3.2.15.14, Rev. 0 Attachment 6 Page 1 of 9 0

0 RADTRAD Version 3,02 run on 11/13/2001 at 13:54:31 0 Pathways:

3 111 11111II II111111 II1111 I11111111 11111I 1 I 111111111 I111Ill11111111llllll1l I 1 11I filllllllll I I11II Pathway 1:

File information Unfiltered Release to Environment 1

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

batch\perry\pnppfha.nif Unfiltered Environment to CR Scenario file name - NEWSDF.SDF 2 Release file name = d:\hwagage\computer codes\radtrad\run 3 batch\perry\pnppfha.rft 2 Dose conversion file name = d:\hwagage\computer codes\radtrad\run Pathway 3:

batch\perry\pnppfha.dcf Control Room Exhaust 2

2 End of Plant Model File Scenario Description Name:

1141111111 # # 11111111"1 It ## # If # If #1##

Plant Model Filename:

4 it 4It I #1111 11111 If # #

S #11 11 Source Term:

1. 1. 1. If #0 # If If # 1

"##N # 0 If # if 11 1 1.OOOOE+00 If If If 1141 111 11 11011 #

1 d:\hwagage\computer codes\radtrad\run batch\perry\pnpp_fha.dcf d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.rft 0.0000E+00 Radtrad 3.02 1/5/2000 1 perry fha: sensitivity case I 0.OOOOE+00 0.9985E+00 0.0015E+00 1.OOOOE+00 Nuclide Inventory File: Overlying Pool:

d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.nif 0 Plant Power Level: 0.0000+E÷00 3.8332E+03 0 Compartments: 0 3 0 Compartment 1: 0 Containment Compartments:

3 3 1.0000E+00 Compartment 1:

0 0 0 1 0 0 0 0 0 0 Compartment 2: 0 Environment 0 2 0 0.OOOOE+00 0 0 Compartment 2:

0 0 0 1 0 0 0 0 Compartment 3: 0 ;Q

-0 Control Room 0 CD 0 1 0 3.6707E+05 0 ,

0 0 0 0 Compartment 3:

Co I--

(0 0 0

1+/- (0 0 Q11a o~aeua~'s go pus 0 0 0

0 bw~f 0

.0 () T 00+R0000,0 004S3000,00 00+30000,0 £0+3000S Z0+2000Z*L IL)>- (7. T 00430000'0 00+30000*0 00+30000*0 c0+3000VIS 00+30000*1 cwa-a 00+3T000*0 00+30000,0 00+30000*0 00+20000*0 00+30000,0 00+30000*0 00+30000,'0 00+30000*0 Et0+30009,9 00+30000*0 00+30000,0 (0+3000z'L 10-30000'V 10+30001'(

10-300001T 00+30000's 0 tO -3000S It 00+30000,0 0 0

0 00+30000*0 00+30000'Z 0

1'0-3000SIE 00+a000010 0

0

uoT)0Doo awfl1OA aApZ3j33 0 00+30000,0 It0+a000ItL 0 10-300001' T0+30009'6 0 10-30000*9 10+30001'(

00+30000,0 00+3000010 00+30000*0 (0+30001'S Z0+3000VL 00+a000011I 00+30000*0 00+30000*0 00+3000010 00+30000*0 E0+3000v'S 00+30000'Z 00+3000010 00+30000*0 00+3000010 00+30000*0 00+3100010 00+3000010 00+3000010 00+30000'0 E0+30009,9 00+3000010 00+3000010 It0+3000( L 1'0-3000S IC 00+30000,0 It 0

0 0 0

0

z A~m4)d 0

0 1'0-300001T 00+E000018 00+30000*0 z0+3000(L4 01+30000*1 00+3000010 1'0-3000S IE 00+30000*0 It 0

00+30000,0 00+30000'Et 0

G0-30008 5' 00+3000010 0

0 0

0 Zd'S a41 go /aipunos aS0)l0

zt UOTIVOO'J 0

0 0

0

!'0-3000It*t T0+a000F'I 0

1'0-3000811 00+30000,9 1'0-M000IE 00430000,0 00+30000*0 00+30000'Z 0

1'0-3000clt1 00+30000*0 0

0 0

kxopunos ealv UO1snlFoXs 0

1 uOT 1000' EUTEOS00 6 JOZ Gbd 9 IUGWq~lj3Wv o'AE)ýj '17V9ZT 'ON -0180

Calc. No. 3.2.15.14, Rev. 0 Attachment 6 Page 3 of 9 1####1#1 #################1###1 ff f###1111ff#1

1. 1. ifif 1i# ifP i i#f##i f## ##f RADTRAD Version 3.02 run on 11/13/2001 at 13:54!31

1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. 1. • #*#*########*###########*##########*###*#####I no" lf1ift 1111"i ff ff# fiifi i fiffiifi ff1i f111iff1 i 11fif1 f fi ii ii f f"Itoi i fi f i ffi fIffifi f 1111111111111111 f9 0 Plant Description if11#11 If IfiItif ifIt"1 fififif *fifif i### Iffl1 f1111111111114# 11 iH* 111111##### 11)111111#111111111111111111111

1. 1. 1 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.O000E+00 Name: Containment Compartment volume = 1.0000E+00 (Cubic feet)

Pathways into and out of compartment 1 Pathway to compartment number 2: Unfiltered Release to Environment Compartment number 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

_0 -< M (D 0 6 o0 0) r-

Caic. No. 3.2.15.14, Rev. 0 Attachment 6 Page 4 of 9 (cfm) Aerosol Elemental Or:ganic 0,0000E+00 6.6000E+03 0.0000E+00 O.0000E+00 0.010OOE+00 1.OOOOE-04 0.0000E+00 0.0000E.00 0.0000E+00 0.010OOE+00 RADTRAD Version 3.02 run on 11/13/2001 at 13:54:31 2.OOOOE+00 5.4000E+03 0.OOOOE+00 0,0000E+00 0.0( 000E+00 7,2000E+02 5.4000E+03 0.OOOOE+00 0.0001E+00 0,0(000E+00 8f88ff 88f#

8nHfilf f8 ftCf fff8 88 ff8 #ft# If f 8fflCH 88f C filltl it fil 8ff lift 8 8 ff8 it81111liltCC Cf LOCATION DATA Scenario Description 2 8ff8888888888888888888888888888888888888f ff888888888888888888888888 Location Exclusion Area Boundary is in compartment Location X/Q Data Radioactive Decay is enabled Time (hr) X/Q Cs

• m^-3)

Calculation of Daughters is enabled 0.0000E+00 4.3000E-04 RELEASE NAME - Perry FHA 2.OOOOE+00 0.OOOOE+00 Release Fractions and Timings CAP EARLY IN-VES.SEL Location Breathing Rate Data 0.0001 hrs 0.0000 hrs Time (hr) Breathing Rate Wm*3 - sec^-l)

NOBLES 1.0000E+00 0.OOOOE+00 0.OOOOE+00 3.5000E-04 IODINE l.O000E+00 0.0000OE00 8.OOOOE+00 1.8000E-04 CESIUM 1.0000E+00 0.O000E+00 2.4000E+01 2.3000E-04 TELLURIUM 0.O000E+00 0.0000E+00 Location Outer Boundary of the LPZ is in compartment 2 STRONTIUM 0.0000+E00 0.00001E00 BARIUM 0.OOOOE+00 O.OOOOE+00 Location X/Q Data RUTHENIUM 0.O000E+00 0.0000E+00 Time (hr) X/Q (s - m^-3)

CERIUM 0.OOOOE+00 0.OOOOE+00 o.OOOOE+00 4.8000E-05 LANTHANUM 0.O000E+00 0.O0001E00 2.OOOOE+00 0.O0001E+00 Iodine fractions Location Breathing Rate Data Aerosol . O.OOOOE+00 Time (hr) Breathing Rate (m^3

• sec*-l)

Elemental = 9.9850E-01 0.0000E+00 3.5000E-04 Organic . 1.5000E-03 8,0000E+00 1.8000E-04 2,4000E+01 2.3000E-04 COMPARTMENT DATA is in compartment 3 Location Control Room Compartment number 1: Containment Location X/Q Data Time (hr) X/Q Cs

• m'-3)

Compartment number 2: Environment 0.0000E+00 3.5000E-04 2.0000E+00 0.OOOOE+00 Compartment number 3: Control Room Location Breathing Rate Data PATHWAY DATA Time (hr) Breathing Rate (m'3

• sec*-l)

O.OOOOE+00 3.50OOE-04 Pathway number 1: Unfiltered Release to Environment 7.2000E+02 0.0000E+00 Convection Data Location Occupancy Factor Data Time (hr) Flow Rate C% / day)

Time (hr) Occupancy Factor 0.OOOOE+00 1.OOOOE+÷0 0.0000E+00 1.0000E+00 7,2000E+02 0.OOOOE+00 2.4000E+01 6.OOOOE-01 9.6000E+01 4.0000E-01 Pathway number 2: Unfiltered Environment to CR 7.2000E+02 0.OOOOE+00 Pathway Filter: Removal Data USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS Time Time step Time (hr) Flow Rate Filter Efficiencies (%C 0.OOOOE+00 2.5000E-02 (cfm) Aerosol Elemental Organic 8.0000E+00 1.0000E-01 0,0000E+O0 6.6000E÷03 0.00001+00 0.00001+00 0.OOOOE+00 2.4000E+01 4.OOOOE-01 1.0000E-04 0.00005+00 0.0000E+00 0.0000E+00 0.0000E+00 5.4000E+03 0.0000E+00 0.00001+00 0.0000E+00 7.2000E+02 0,OOOOE+00 0)-< m 2.00003+00 7.2000E+02 5.4000E+03 0.0000E+00 O.0000E+00 0.0000E+00 (D O0 Pathway number 3: Control Room Exhaust Pathway Filter: Removal Data M(71 0@

Time (hr) Flow Rate Filter Efficiencies (%) I-4

Calc. No. 3.2.15.14, Rev. 0 Attachment 6 Page 5 of 9 Xe-135 1.3783E+01 5.3973E-09 2.4076E+16 5,0998E+11 Xe-135m 6.5313E-01 7.1700E-12 3.1984E+13 2.4166E+10 RADTRAD Version 3.02 run on 11/13/2001 at 13:54:31 nnnn n#nnn finn nnnn nn nnnn nnnnnn nnn~f~nn n n~nn #f nn nn nn nnn nnn nn nn nn nnn

1. ifnn#nnnn#####i Cs-135 4.7418E-14 4.1157E-14 1.8359E+11 1.7545E-03 Control Room Transport Group Inventory:

Overlying Time (h) - 0.0001 Atmosphere Sump Pool Noble gases (atoms) 1.5969E+19 0.0000E+00 0.0002E+00 tif If if if itfit)i I it if 1111 Elemental I (atoms) 1.8787E+17 O.0000E+00 0.O000E+00 If Iif tif if If if Organic I (atoms) 2.8223E+14 O.0000E+00 0.O000E+00 a1 If If II4 11# 11 Aerosols (kg) 0.O000E+00 O.0000E+00 0.O000E+00 If #l # If if funiffif If if if if if If if if if Deposition Recirculating if if if if if If if if Time (h) = 0.0001 Surfaces Filter nnnn nnif lfifI if ltififif if Noble gases (atoms) 0.0002E+00 0.0000E+00 Elemental I (atoms) 0.0000E+00 0.00002+00 Organic I (atoms) 0.0000E+00 0.0000E+00 Aerosols (kg) 0.002E+00 O.0000E+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 0.0000E+00 1.4377E+00 Elemental I (atoms) 0.0000+E00 Accumulated dose (rem) 4.2456E-01 0.0000E+00 1.4377E+00 Organic I (atoms) 0.0000+E00 Aerosols (kg) 0.0000E+00 Outer Boundary of the LPZ Doses:

Control Room Exhaust Transport Group Inventory:

Time (h) = 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.7393E-02 O.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.O0000+00 Control Room Doses: Elemental I (atoms) 0.0000+E00 Organic I (atoms) O.0000E+00 Time (h) - 0.0001 Whole Body Thyroid TEDE Aerosols (kg) 0,0000E+00 Delta dose (rem) 1.2078E-06 0.0000E+00 4.5686E-05 Accumulated dose (rem) 1.2078E-06 0.0000E+00 4.5686E-05 Exclusion Area Boundary Doses:

Control Room Compartment Nuclide Inventory: Time (h) - 2:0000 Whole Body Thyroid TEDE Delta dose (rem) 1.0214E-03 0.0000E+00 3.4587E-03 Time (h) = 0.0001 Ci kg Atoms Bq Accumulated dose (rem) 4.2559E-01 0.00002+00 1.4411E+00 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 Outer Boundary of the LPZ Doses:

Kr-83m 1,25002-02 6.0583E-13 4.3957E+12 4.6248E+08 Kr-85 8.1467E-01 2.0751E-06 1.4702E+19 3.01432310 Time (h) = 2.0000 Whole Body Thyroid TEDE Kr-85m 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-05 9.2854E-16 6.4273E+09 9.73152+05 Accumulated dose (rem) 4,7507E-02 0.00002+00 1.6087E-01 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 Control Room Doses:

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 Time (h) = 2.0000 Whole Body Thyroid TEDE 1-130 1.4897E-03 7.6381E-13 3.5383E+12 5.5118E+07 Delta dose (rem) 4.3741E-02 0.0000E+00 1.8080E+00 1-131 1.9834E-01 1.5999E-09 7.3546E+15 7.3387E+09 Accumulated dose (rem) 4.3742E-02 0.0000E+00 1.8080E+00 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 Coptrol Room Compartment Nuclide Inventory:

1-134 6.5286E-09 2.4473E-19 1.0999E+06 2.4156E+02 1-135 2.0391E-02 5.8063E-12 2.5901E+13 7.5446E+08 Time (h) - 2.0000 Ci kg Atoms Bq Xe-129m 2.2597E-04 Xe-131m 2.9842E-01 1.7859E-12 3.56272-09 8.3372E+12 1.6378E+16 8.3607E+06 1.1041E+10 Br-82 5.83942-04 5.3937E-13 3.9611E+12 2.16062+07 MnLo Br-83 9.0455E-06 5.7258E-16 4.1544E+09 3.34682+05 Xe-133 5.00661+01 2.6747E-07 1.2111E+18 1.8525E+12 Kr-83m 5.8665E-03 2,8434E-13 2.0631E+12 2.1706E+08 Xe-133m 1.5255E+00 3.3997E-09 1.5394E+16 5.64422+10 Kr-85 8,1466E-01 2.0750E-06 1.4701E+19 3,0143E+10 01' 0Y)

Caic. No. 3.2.15.14, Rev. 0 Attachment 6 Page 6 of 9 Kr-85m 1.1914E-01 1.4477E-11 1. 0257E+14 4.4081E+09 Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 8.8422E-06 3.1216E- 16 2.1608E+09 3. 2716E+05 Accumulated dose (rem) 4.7507E-02 0.0000E+00 1.6087E-01 Kr-87 Kr-88 3.0777E-02 2.4545E-12 1.6797E+13 1.1388E+09 Rb-87 5,3930E-20 6.1643E-16 4.2669E+09 1.9954E-09 Control Room Doses:

Rb-88 3.4843E-02 2.9027E-13 1.9864E+12 1.2892E+09 6.8183E-09 3.8601E-08 1.8020E+17 2.5228E+02 Time (h) - 8.0000 whole Body Thyroid TEDE 1-129 1.3316E-03 6.8277E- 13 3.1629E+12 4.9271E+07 Delta dose (rem) 2.1440E-02 0.0000E+00 9.9924E-01 1-130 1-131 1.9692E-01 1.5884E-09 7.3020E+15 7.2861E+09 Accumulated dose (rem) 6.5182E-02 0.0000E+00 2.8073E+00 1-132 8.6223E-02 8.3532E-12 3.8109E+13 3.1902E+09 1.1592E-01 1.0213E-10 4.6335E+14 4.2892E+09 Control Room Compartment Nuclide Inventory:

1-133 1-134 1.3431E-09 5.0346E-20 2.2626E+05 4.9693E+01 1-135 1.6533E-02 4.7078E-12 2.1001E+13 6.1173E+08 Time (h) 8.0000 Ci kg Atoms Bq 1.7731E-12 8.2773E+12 8.3006E+06 Br-82 2.6012E-06 2.4027E-15 1.7646E+10 9.6246E+04 Xe- 129m 2.2434E-04 3.5456E-09 1 6299E+16 1.0988E+10 Br-83 7.9559E-09 5.0361E-19 3.6540E+06 2.9437E+02 Xe-131m 2.9698E-01 2.6464E-07 1 .1983E+18 1.8328E+12 Kr-83m 3.0436E-06 1.4752E-16 1.0703E+09 1.1261E+05 Xe- 133 4 .9536E+01 3.3113E-09 1.4993E+16 5.4975E+10 Kr-85 4.0826E-03 1.0399E-08 7.3674E+16 1.5106E+08 Xe- 133m 1.4858E+00 1.1852E+01 4,6412E-09 2.0704E+16 4 .3854E+11 Kr- 85m 2.3598E-04 2.8674E-14 2.0315E+11 8 .7311E+06 Xe- 135 5.5597E-03 6.1033E-14 2.7226E+11 2.0571E+08 Kr-87 1.6835E-09 5.9434E-20 4.1140E+05 6.22902+01 Xe-135m 7.6435E-10 3.4096E+15 3.2584E+01 Kr-88 3.5664E-05 2.8442E-15 1.94642+10 1.3196E+06 Cs-135 8.8064E-10 Rb-87 4.0195E-22 4.5943E-18 3.1802E+07 1.4872E-11 Rb-88 4.1000E-05 3.4156E-16 2.3374E+09 1.5170E+06 Control Room Transport Group Inventory:

1-129 3.4171E-11 1.9345E-10 9.0311E+14 1.2643E+00 Overlying 1-130 4.7669E-06 2.4441E-15 1.1322E+10 1.7637E+05 Sump Pool 1-131 9.6586E-04 7.7908E-12 3.5815E+13 3.5737E+07 Time (h)( 2.0000 Atmosphere 1.5969E+19 0.O000E+00 0.0000E+00 1-132 7.0844E-05 6.8633E-15 3.1312E+10 2.6212E+06 Noble gases (atoms) 0.0000E400 0.0000E+00 1-133 4.7568E-04 4.1991E-13 1.9013E+12 1.7600E+07 Elemental I (atoms) 1.8787E+17 0.0000E+00 0.0000E+00 1-135 4.4166E-05 1.2576E-14 5.61012+10 1.63412+06 Organic I (atoms) 2.8223E+14 0.0000E+00 0.0000E+00 Xe-129m 1.1002E-06 8.6956E-15 4 0594E+10 4.0708E+04 Aerosols (kg) 0.0000E+00 Xe-131m 1.4670E-03 1.7514E-11 8.0514E+13 5.4279E+07 Xe-133 2. 4044E-01 1.2845E-09 5.8162E415 8.8961E+09 Deposition Recirculating Xe- 133m 6.8805E-03 1.5334E-11 6.9432E+13 2.5458E+08 Time (h) = 2.0000 Surfaces Filter O.O000E+00 Xe- 135 3.7613E-02 1.4729E-11 6. 5703E+13 1.3917E+09 Noble gases (atoms) O.0000E+00 Xe-135m 7.3176E-06 8.0332E-17 3.5835E+08 2.7075E+05 Elemental I (atoms) 0.0002+00 0.O000E+00 Cs-135 1.4255E-11 1.2373E-11 5.5194E+13 5.2745E-01 Organic I (atoms) O.0000E+00 0.00002+00 Aerosols (kg) 0.0000E+00 0.O0000E+00 Control Room Transport Group Inventory:

Unfiltered Environment to CR Transport Group Inventory:

Overlying Atmosphere Sump Pool Pathway Time (h) - 8.0000 Noble gases (atoms) 8.0029E+16 0.00002+00 0.0000E+00 Time (h) = 2.0000 Filter 0.0002E+00 Elemental I (atoms) 9.4155E+14 0.0000E+00 Noble gases (atoms) 0.0000E+00 0.0000E+00 Organic I (atoms) 1.4145E+12 0.0000E+00 Elemental I (atoms) 0.0000+E00 0.0000E+00 Aerosols (kg) O.00002+00 O.0000E+00 Organic I (atoms) 0.0000E+00 Aerosols (kg) O.0000+E00 Deposition Recirculating Time (h) - 8.0000 Surfaces Filter Control Room Exhaust Transport Group Inventory:

Noble gases (atoms) O.0000+E00 0.0000E+00 Elemental I (atoms) 0.0000E+00 0.O000E+00 Pathway Filter Organic I (atoms) 0.0000E+00 0.O000E+00 Time (h) = 2.0000 0.0000E+00 Aerosols (kg) 0.0000+E00 0,0000E+00 Noble gases (atoms)

Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Unfiltered Environment to CR Transport Group Inventory:

Aerosols (kg) O.0000E+00 Pathway Time (h) - 8.0000 Filter Exclusion Area Boundary Doses:

Noble gases (atoms) 0.0000E+00 -u-ur TI -V m Time (h) = 8.0000 whole Body Thyroid TEDE Elemental I (atoms) 0.0000E+00 nea0 Organic I (atoms) 0.0000E+00 Delta dose (rem) 0.00002+O0 0.0000+E00 0.0000E+00 Aerosols (kg) 0.0000E+00 (0 Oz Accumulated dose (rem) 4.2559E-01 0.0000E+00 1.44112+00 Outer Boundary of the LPZ Doses: Control Room Exhaust Transport Group Inventory:

o

-4M 8.0000 Whole Body Thyroid TEDE Pathway Time (h) = 0)

Calc. No. 3.2.15.14, Rev. 0 Attachment 6 Page 7 of 9 Time (h) = 8.0000 Filter Noble gases (atoms) 0.OOOOE.00 Pathway Elemental I (atoms) 0.0000E+00 Time (h) - 24.0000 Filter Organic I (atoms) 0.0O00E+00 Noble gases (atoms) 0.O000E.00 Aerosols (kg) 0.0000E+00 Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Exclusion Area Boundary Doses: Aerosols (kg) 0.OOOOE+00 Ti(me Mt) -  ;!-1.0000 Whole Body ThyI otId Exclusion Area Boundary Doses:

Delta dose (rem) O.OOOOE+00 O.OOOOE+00 0. 0000E+00 Accumulated dose (rem) 4.2559E-01 0.OOOOE00 1.4411E+OQ Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+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 Outer Boundary of the LPZ Doses:

Delta dose (rem) 0.OOOOE+00 0.00000E+0 0.OOOOE+00 Accumulated dose (rem) 4.7507E-02 0.OOOOE+00 1.6087E-01 Time (h) - 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 Control Room Doses: Accumulated dose (rem) 4.7507E-02 0.0000E+00 1.6087E-01 Time (h) = 24,0000 Whole Body Thyroid TEDE Control Room Doses, Delta dose (rem) 7.9420E-05 O.O000E+00 4. 8094E-03 Accumulated dose (rem) 6.5262E-02 0.0000E+00 2. 8121E+00 Time (h) - 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.0998E-11 0,0000E+00 1.9496E-09 Control Room Compartment Nuclide Inventory: Accumulated dose (rem) 6.5262E-02 0.0000E+00 2.8121E+00 Time (h) = 24.0000 Ci kg Atoms Bq Control Room Compartment Nuclide Inventory:

Kr-85 3.0026E-09 7.6480E-15 5.4185E+10 1.1110E+02 1-129 2.5135E-17 1.4230E-16 6.6428E+08 9.2998E-07 Time (h) - 96.0000 Ci kg Atoms Bq 1-131 6.7076E-10 5.4104E-18 2.4872E+07 2.4818E+01 1-133 2.0529E-10 1.8122E-19 8.2055E÷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 6.0068E+03 Overlying Xe-133m 4.0989E-09 9.1350E-18 4 . 1363E+07 1.5166E+02 Time (h) - 96.0000 Atmosphere Sump Pool Xe-135 8.1770E-09 3.2020E-18 1.4284E+07 3.0255E+02 Noble gases (atoms) 1.4778E-17 0.0000E+00 0.0000E+00 Cs-135 1.9287E-17 1.6740E-17 7.4676E+07 7.1363E-07 Elemental I (atoms) 1.7387E-19 0.0000E+00 0.0000S+00 Organic I (atoms) 2.6119E-22 0.00002+00 0.0000E+00 Control Room Transport Group Inventory: Aerosols (kg) 0.0000E+00 0.0000+E00 0.0000E+00 Overlying Deposition Recirculating Time (h) = 24.0000 Atmosphere Sump Pool Time (h)i 96.0000 Surfaces Filter Noble gases (atoms) 5.8866E.10 0.0000E+00 0.0000E+00 Noble gases (atoms) 0.0000E+00 O.0000E+00 Elemental I (atoms) 6.9256E208 O.0000E+00 0.O000E+00 Elemental I (atoms) 0.0000E+00 0.0000E+00 Organic I (atoms) 1.0404E+06 0.0000E+00 0.0000E+00 Organic I (atoms) 0.0000E+00 0.0000E+00 Aerosols (kg) 0.O000E+00 O.0000E+00 0.O000E+00 Aerosols (kg) 0.0000E+00 O.0000E+00 Deposition Recirculating Unfiltered Environment to CR Transport Group Inventory!

Time (h) - 24.0000 Surfaces Filter Noble gases (atoms) 0.0000E+00 0.0000E+00 Pathway Elemental I (atoms) 0.0000E+00 0.0000E+00 Time (h - 96.0000 Filter Organic I (atoms) 0.O000E+00 O.0000E+00 Noble gaseo (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 O.0000E+00 Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.0000E+00 Unfiltered Environment to CR Transport Group Inventory: Aerosols (kg) 0.O000E+00 Pathway Control Room Exhaust Transport Group Inventory; Time (h) - 24.0000 Filter Noble gases (atoms) 0.0000E+00 Pathway Elemental I (atoms) 0.0000E+00 Time (h) = 96.0000 Filter M0 Organic I (atoms) 0.0000E+00 Noble gases (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 o Elemental I (atoms) 0.00002+00 Organic I (atoms) 0.0000E+00 (0 0 3 Control Room Exhaust Transport Group Inventory: Aerosols (kg) 0.0000E+00 r-

Calc. No. 3.2.15.14, Rev. 0 Attachment 6 Page 8 of 9 Containment Environment Control Room Exclusion Area Boundary Doses: 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 Thyroid TEDE 0.275 o.0000E+00 1.8238E+02 1.9815E-01 Delta dose (rem) 0.00002E00 0.0000E+00 0.0000E+00 0.525 O,.00001+00 1.8238E.02 1.9797E-01 Accumulated dose (rein) 4.2559E-01 0.00002+00 1.4411E.00 0.775 O.0000E+00 1.8238E+02 1.9779E-01 1.025 o.OOOOE+00 1.8238E+02 1.9761E-01 Outer Boundary of the LPZ Doses: 1.275 O.O0001E+00 1.8238E+02 1.9744E-01 1.525 .O0000E+00 1.8238E+02 1.97261-01 Time (h) = 720.0000 Whole Body Thyroid TEDE 1.775 o.OOOOE+00 1.8238E+02 1.9708E-01 Delta dose (rem) 0.0000E+00 0.0000E+00 0.00002300 2.000 0.00001+00 1.8238E+02 1.96921-01 Accumulated dose (rem) 4.7507E-02 O.00002+00 1.6087E-01 2.250 O.O0000E+00 1.8242E+02 1.5779E-01 2.500 O.0000E+00 1,8245E.02 1.2643E-01 Control Room Doses: 2.750 o.0000E+00 1.8248E+02 1.0130E-01 3.000 0.0000+E00 1.8250E+02 8.1171E-02 Time (h) - 720.0000 Whole Body Thyroid TEDE 3.250 O.0000B+00 1.8251E+02 6.5039E-02 Delta dose (rem) 1.75652-39 0.00002+00 2.3980E-37 3.500 O.00002300 1.8252E+02 5.21141-02 Accumulated dose (rem) 6.5262E-02 0.00001.00 2.8121E+00 3.750 0.00001+00 1.8253E+02 4.1757E-02 4.000 o.0000E+00 1.8254E+02 3.3458E-02 Control Room Compartment Nuclide Inventory: 4.250 0.0000E+00 1.8255E÷02 2.6809E-02 4.500 0.0000E+00 1.8255E+02 2.1481E-02 Time (h) = 720.0000 Ci kg Atoms Bq 4.750 0.O0000E+00 1.8256E+02 1.7212E-02 5.000 o.0000E+00 1.8256E+02 1,3791E-02 Control Room Transport Group Inventory: 5.250 o.0000E+00 1,8257E+02 1.1050E-02 5.500 O.0000E+00 1.8257E+02 8.8544E-03 Overlying 5.750 o.0000E+00 1,8257E+02 7.09472-03 Time (h) = 720.0000 Atmosphere Sump Pool 6.000 .O0000E+00 1.8257E+02 5.6847E-03 Noble gases (atoms) 9.2790-257 0.O000E+00 O.0000E+00 6.250 0.O00002+00 1.8257E+02 4.55502-03 Elemental I (atoms) 1.0917-258 0.O000E+00 0.O000E+00 6.500 0.0000E+00 1.8257E+02 3.6497E-03 Organic I (atoms) 1.6400-261 0.00002E00 O.0000E+00 6.750 o.0000E+00 1.8257E+02 2.9244E-03 Aerosols (kg) 0.00002+00 0.0000+E00 0.00001+00 7.000 0. 0000E00 1.8257E+02 2.3432E-03 7.250 o.0000E+00 1.8257E+02 1.8775E-03 Deposition Recirculating 7.500 0.0000E+00 1.8257E+02 1.5044E-03 Time (h) = 720.0000 Surfaces Filter 7.750 0.0000E÷00 1.82578+02 1.2054E-03 Noble gazes (atoms) 0.0000E+00 0.0000.E00 8.000 0.0000+E00 1.8258E+02 9.6586E-04 Elemental I (atoms) O.0000E+00 0.OOOOE+00 8.400 o.0000E+00 1.8258E÷02 6.7757E-04 Organic I (atoms) O.0000E+00 0.00002E00 8.700 o.0000E+00 1.8258E+02 5.1938E-04 Aerosols (kg) O.O000E+00 0.0000E+00 9.000 o.0000+E00 1.8258E+02 3.9812E-04 9.300 o.0000E+00 1.8258E+02 3.0518E-04 Unfiltered Environment to CR Transport Group Inventory: 9.600 0.0000E+00 1.8258E+02 2.3393E-04 9.900 0.0000E+00 1.8258E+02 1.79312-04 Pathway 10.200 O.O0000E÷00 1.8258E+02 1.3745E-04 Time (h) = 720.0000 Filter 24.000 0.0000E+00 1.8258E+02 6.7076E-10 Noble gases (atoms) 0.00002+00 96.000 o.0000E+00 1.8258E+02 1.3002E-37 Elemental I (atoms) 0.0000E+00 720.000 o.0000E+00 1.8258E+02 8.6776-278 Organic I (atoms) 0.0000E+00 Aerosols (kg) O.0000E+00 #996#69969#96#9#9##6666######6#####6696######f6f##9#6##6#6######9#966 Cumulative Dose Summary Control Room Exhaust Transport Group Inventory: ########9##################6######################99################

Pathway Exclusion Area Bounda Outer Boundary of the Control Room Time (h) = 720.0000 Filter Time Thyroid TEDE Thyroid TEDE Thyroid TEDE Noble gases (atoms) O.O000E+00 (hr) (rem) (rem) (rem) (rem) (rem) (rem)

Elemental I (atoms) O.O000E+00 0.000 0.00002+00 1.4377E+00 O.0000E+00 1.6048E-01 0.0000E+00 4.5686E-05 Organic I (atoms) 0.00002+00 0.275 0.0000E+00 1.4411E+00 O.0000E+00 1,6087E-01 0,0000E+00 2.50852-01 Aerosols (kg) 0.O000E+00 0.525 0.0000E+00 1.4411E+00 0.0000E+00 1.60871-01 0.0000E+00 4.7825E-01 Q) -<  :

0.775 0.00001+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00001+00 7.0504E-01 2223 1.025 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 9.3124E-01 CD 0~

1.275 0.0000E+00 1.44112+00 0.0000E+00 1.6087E-01 0.00001+00 1.1569E+00 fifff69lf

6) Hug # 66)ffit"H f))6ff)ff) 99)) 1119#11999) 69))"I 9))

696If6666 6f66996 69 99 1.525 0.00001+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00001+00 1.3819E÷00 1-131 Summary 1.775 0.0000E+00 1.4411E+00 0.00001+00 1.6087E-01 0.0000E+00 1.6064E+00 2.000 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00002+00 1.80801+00 0Z 2.250 0.0000E+00 1.4411E+00 O.0000E+00 1.6087E-01 0.0000E+00 2.0086E+00 I-4

Caic. No. 3.2.15.14, Re'v. 0 Attachment 6 Page 9 of 9

2. 500 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00001+00 2.1691E+00 2.750 0.0000E+00 1,4411E200 0.0000E+00 1.6087E-01 0.0000E+00 2.2975E+00 3.000 0.0000E+00 1.4411E+00 0.0000.E00 1.60871-01 O.0000E+00 2.4003E+00 3.250 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.4825E+00 3.500 0.00001300 1.4411E+00 0.00002+00 1.6087E-01 0.0000E+00 2.5483E+00 3.750 0.0000E+00 1.4411C.00 0.O000EO00 1.6087E-01 0.00001+00 2.6010E+00 4 .000 0.0000+E00 1.4411E+00 0.0000+00 1.60875-01 O.0000E+00 2.6431E+00 4 .250 0.00000E+0 1.4411E+00 0.0000E+00 1.6087E-01 0.00001+00 2.6768E+00

4. 500 0.0000E+00 1.4411E+00 0.0000+E00 1.6087E- 01 0.0000E+00 2.7038E+00 4.750 0.0000E+00 1.4411E+00 0.00001+00 1.6087E-01 0.0000E+00 2.7254E+00 5.000 0.00001+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.7427E+00 5.250 0.0000.E00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.7566E+00 5.500 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.7676E÷00 5.750 0.0000E+00 1.4411E+00 O.0000.E00 1.60871-01 0.0000E+00 2.7765E+00 6.000 0.0000E÷00 1.4411E+00 0.0000.E00 1.6087E-01 0.0000E+00 2.7836E+00 6.250 0.0000E+00 1.4411E+00 O.0000+E00 1.6087E-01 0.0000E+00 2.7893E+00 6.500 0.0000E+00 1.4411E+00 O.0000E+00 1.6087E-01 O.0000E+00 2.7938E+00 6.750 0.00000E+0 1.4411E+00 0.O000E+00 1.6087E-01 0,0000E+00 2.7975E+00 7.000 0.O000E÷00 1.4411E+00 0.00001+00 1.6087E-01 0.0000E+00 2.80042+00 7.250 0.00000E+0 1.44111+00 0.0000E+00 1.6087E-01 0.00O0E+00 2.8027E+00 7.500 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.8046E+00 7.750 0.00001+00 1.4411E+00 0.0000E+00 1 . 6087E-01 0.00002+00 2.8061E+00 8.000 0.0000.E00 1.4411E+00 O.0000E+00 1.6087E-01 0.0000E+00 2.80732+00 8.400 0.0000E+00 1.4411E+00 O.0000E+00 1.6087E-01 0.0000E+00 2.8087E+00 8.700 0.0000E+00 1.4411E÷00 0.0000E+00 1.6087E-01 O.0000E+00 2.8095E÷00 9.000 0,0000E÷00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.8101E+00 9.300 0,0000E+00 1.4411E+00 0.00001÷00 1.6087E-01 O.00002E00 2.8106E+00 9.600 0.00002+00 1.4411E+00 0.00000E+0 1.60871-01 0.0000E+00 2.8109E+00 9.900 0.0000E+00 1.4411E+00 0.00001+00 1.6087E-01 0.0000E+00 2.8112E+00 10.200 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.8114E+00 24.000 0.O000E+00 1.4411E+00 O.0000E+00 1.6087E-01 0.00000E+0 2.8121E+00 96.000 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.8121E+00 720.000 0.00000E+0 1.4411E+00 0.00000E+0 1.6087E-01 0.00002+00 2.8121E+00 88888888888888888888888888888888808888888888888888888888 888888888888 Worst Two-Hour Doses Note: 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 868888fill8it8H888888888888888888H#8#8#It88881888I88 18H8#1 #41#1tH#I Exclusion Area Boundary Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.0214E-03 0.0000E+00 3.4587E-03 Outer Boundary of the LPZ Time Whole Body Thyroid TEDE (hr) (rem) (rem) (rem) 0.0 1.14022-04 0.0000E+00 3.8609E-04 Control Room Time Whole Body Thyroid TEDE T-u-ur

13) -:3 (hr) (rem) (rem) (rem) Q 0o 0.0 4.3741E-02 0.0000S+00 1.8080E+00 0)z 33 0)

Co

Calc. No. 3.2.15.14, Rev. 0 Attachment 7 Page 1 of 9 0

0 RADTRAD Version 3.02 run on 11/13/2001 at 13:56:28 Pathways:

IttllItItIft 6illH ifltt IfIfif1166 6 IfIf#It6*6I6It If0figofn fOttJillIf I 0 it"ff#it if l I$IfItItIfiff Ifi If 3 Pathway i:

I6fi6I1Ifit6fi16666666f6If6If 866"IffIffI 66i666i6866fi1It I#f1161111116661616fifll6 6If68866f6fill68inIf"11Iti8111111111f

1. 11116 Unfiltered Release to Environment File information 1

1. if#fififf##### 6########66 f###########i#####8866686666#### 8######

1. 1. 8##6#6## 2 4

Plant file name - pnppfha.psf Pathway 2:

Inventory file name = d:\hwagage\computer codes\radtrad\run Unfiltered Environment to CR batch\perry\pnppfha.nif 2 Scenario file name - NEWSDF.SDF 3 Release file name - d:\hwagage\computer codes\radtrad\run 2 batch\perry\pnppfha.rft Pathway 3:

Dose conversion file name - d:\hwagage\computer codes\radtrad\run Control Room Exhaust batch\perry\pnppfha.dcf 3 2

2 End of Plant Model File Scenario Description Name:

1. 1. 6## ##68# 8#6# If 6 If IfIfi#

If 11 I 66 If If Of If 6 8 Plant Model Filename:

It It If If If IfIf # 8 I If 6 H It #f If 8

1. if

1. If # if Source Term:

1. It If If # ## if 1 If # Iff If 11I t 111116 #

1. 1. 1. 6If It If 1 1.O000E+00 d:\hwagage\computer codes\radtrad\run batch\perry\pnpp_fha.dcf d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.rft o.0000E+00 Radtrad 3.02 1/5/2000 1 perry fha: sensitivity case 2 0.OOOOE+00 0.99851+00 0.00151+00 1.0000E+00 Nuclide Inventory File: Overlying Pool:

d:\hwagage\computer codes\radtrad\run batch\perry\pnppfha.nif 0 Plant Power Level: 0.0000E÷00 3.8332E+03 Compartments:

3 0 Compartment 1: 0 Containment Compartments:

3 3 1.00003E+00 Compartment 1:

0 0 0 1 0 0 0 0 0 0 Compartment 2: 0 Environment 0 2 0 O.OOOOE+00 0 0 Compartment 2:

0 0 0 1 -V ,

0 0 0 0 Compartment 3: 0 Control Room 0 03 1 0 M 0 0 3.6707E+05 0 14:

0 0 (0 @

0 Compartment 3: (7 ,

1 1

-1

0) a11A O~leuaDs jo pus 0 0 0 L0 0 0 z(0 1 0 00+3000010 00+30000*0 00+300000o 00+80000,0 00~s000z*L LOU 00+30000*0 00~a300000 00+H0000,0 00+30000*0 00+3T000'0 00~aoo0000 00+300000 00+30000,0 E0+30009,9 00+30000,0

. au iameual~a ind~no 00+H000010 zo0s3000,L w 11. fL 0 0O-3000010 o0+300oi~

10 -3000011 00+30000'9 0 zo-sooos z 00+30000,0 0 C, 0

SIQ3MW32A UO~~lelWTS 0

E Auzmqlo~

00+3O0000 0 00+230000*Z t'0-3000GIE 00+30000,0 0 0

0

u0J10oo0 atuilIOA aA~j3Qfl3 0 00+3000010 zo+soooo L 0 0

10-s300009 T0+30009'0 00+30000,0 00+3000010 00+30000*0 ,0+30009 L 00+30000*0 00-3000019 00+300000z 00+30000,0 00+HT10000 00+30000*0 00+30000*0 00+20000*0 00+3000010 00+20000*0 00+30000,0 E0+30009*9 00+300000o 0o-aooosT 00+30000,0 V

I 0 0

0 0

0 0 :Z Aemiqled 0

~o-aooo1OUO00e30 c 00+2000010 00+30000 L 0I+30000'T 00+300000

ý'0-30008I 1 0+300008Z V0-2000S £ 00+30000'8 0

00+30000 0 00+30000*0 0 0

9O-30008 C' 00+30000*0 0 0

0 0

Zdq aq ;o Raapunos aolno 0 lz UOX1eo'I 0 0 0 t'03000£0Z 10+3000I'0 ;I R+mtfled I'0-300081I 00+30000*0 v'0-3000slC 00+30000,0 'SAeMT4:1d 0

I 0 00+2000010 00+30000*Z I0+3000O's 10+300009s 10+30000's z00a3000L I'0-3000EV1 00+30000*0 T0+30000SG I0+30000*S I0+30000's 00+30000'E 00+30000,0 00+30000,0 00+30000*0 00+30000'0 0

ktuputiou +01+1 ttolu(oxaI 0

1u U3OTI0300 0

0 0 0 6 10 Z 06d L IuGW40efll 0 0'AGU't7VWZ'C ej 4 L 'ON OJO-01eO

Calc. No. 3.2.15.14, Rev. 0 Attachment 7 Page 3 of 9

1. 1. 1. 1. 4h########*#####*########N############################*##############If RADTRAD Version 3.02 run on 11/13/2001 at 13:56:28

1. 1. 1. R#h#####################################################*f############

f Iif aI fI i t Iffli If0It00 0 0it0H0If0If1H111111110111f 11f111111fill Plant Description Number of Nuclides = 23 Inventory Power = 3.8332E+03 MWth Plant Power Level - 3.83321+03 MWth Number of compartments 3 Compartment information Compartment number 1 (Source term fraction - 1.0000E+00 Name: Containment Compartment volume . .0000E+00 (Cubic feet)

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

Removal devices within compartment:

Filter(s)

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)

CD

0) MZ 0)Z 0)

Calc. No. 3.2.15.14, Rev. 0 Attachment 7 Page 4 of 9 7.2000E+02 O.0000E+00 O.OOOOE+00 0.0000E+00 0.0000E+00 RADTRAD Version 3.02 run on 11/13/2001 at 13:56:28 Pathway number 3: Control Room Exhaust Pathway Filter: Removal Data Scenario Description Time (hr) Flow Rate Filter Efficiencies (M) fifffffftt###huff#f#########*#######*#####4##############################

(cfm) Aerosol Elemental Organic 0.0000E+00 6.6000E+03 0.0000E+00 0.0000E+00 0.0000.E00 Radioactive Decay is enabled 1.OOOOE-04 0.O000E+00 0.00008+00 0.00000E+0 0.OOOOE+00 Calculation of Daughters is enabled 7.2000E+02 0.0000E+00 0.O000E+00 0.O000E+00 0.O000E+00 RELEASE NAME . 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 hrs NOBLES 1.0000E+00 0.00008+00 Location X/Q Data IODINE 1.0000E÷00 0.O000E+00 Time (hr) X/Q (s

• m'-3)

CESIUM 1.OOOOE+00 0.O000E+00 0.0000E+00 4.3000E-04 TELLURIUM O.OOOOE+00 O.OOOOE+00 2.0000E+00 0.0000E+00 STRONTIUM 0.0000E+00 0.0000E+00 BARIUM 0.0000E+00 0.OOOOE+00 Location Breathing Rate Data RUTHENIUM 0.OOOOE+00 O.0000E+00 Time (hr) Breathing Rate (m^3

• sec^-l)

CERIUM 0.0000E+00 0.OOOOE+00 o.0000E+00 3.5000E-04 LANTHANUM 0.0000E+00 0.OOOOE+00 8.0000E+00 1.8000E-04 2.4000E+01 2.3000E-04 Iodine fractions Location Outer Boundary of the LPZ is in compartment 2 Aerosol = 0.00000E+0 Elemental - 9.98508-01 Location X/Q Data Organic - 1.5000E-03 Time (hr) X/O (s

• m'-3) 0.O000E+00 4.8000E-05 COMPARTMENT DATA 2.0000.E00 0.00000E+0 Compartment number 1: Containment Location Breathing Rate Data Time (hr) Breathing Rate (m*3 sec^-l)

Compartment number 2: Environment 0.0000E.00 3.5000E-04 8.0000E+00 1.8000E-04 Compartment number 3: Control Room 2.4000E+01 2.3000E-04 Location Control Room is in compartment 3 Compartment Filter Data Location X/Q Data Time (hr) Flow Rate Filter Effic.iencies (M) Time (hr) X/Q (s

• m'-3)

(cfm) Aerosol Elemsental Organic 0.0000E+00 3.5000E-04 0.0000E+00 2.7000E+04 0.0008E+00 0.000' 0E+00 0.0000E+00 2.OOOOE+00 O.0000E+00 2.OOOOE+00 2.7000E+04 5.OOOOE+01 5.00010E+01 5.0000E+01 7.2000E+02 2.7000E+04 5.0000E+01 5.00010E+01 5.0000E+01 Location Breathing Rate Data Time (hr) Breathing Rate (m'3

• sec-l)

PATHWAY DATA 0.OOOOE+00 3.5000E-04 7.2000E+02 0.00000E+0 Pathway number 1: Unfiltered Release to Environment Location Occupancy Factor Data Convection Data Time (hr) Occupancy Factor Time (hr) Flow Rate (8 / day) 0.OOOOE+00 1.O000E+00 O.0000E+00 1.0000E+10 2.4000E+01 6.0000E-01 7.2000E+02 0.0000E+00 9.6000E÷01 4.0000E-01 7.2000E+02 0.0000E+00 Pathway number 2: Unfiltered Environment to CR 0) -< :3 USER SPECIFIED TIME STEP DATA - SUPPLEMENTAL TIME STEPS 0 Pathway Filter: Removal Data Time Time step 0.OOOOE+00 2.5000E-02 0) M o Time (hr) Flow Rate Filter Effi:ciencies M%) 8.0000E+00 1.0000E-01 (D 5 (cfm) Aerosol Elem:ental Organic 2.4000E+01 4.0000E-01 -4f 0.0000E+00 6.6000E+03 0.0000+E00 0.00010S+00 0.0000E+00 7.2000E+02 0.0000E+00 0).

1.0000E-04 0.0000E+00 0.00008+00 0.000(0E+00 0,0000E+00

Calc. No. 3.2.15.14, Rev. 0 Attachment 7 Page 5 of 9 Xe-135 1.3783E+01 5.3973E-09 2,4076E+16 5.0998E+11 Xe-135m 6.5313E-01 7.1700E-12 3.1984E+13 2.4166E+10 RADTRAD Version 3.02 run on 11/13/2001 at 13:56:28 Cs-135 4.7418E-14 4,1157E-14 1.8359E+ll 1.7545E-03 8866#88868888868886888#8#88##8#888###886888688))888888#88888888## 88##8###

Control Room Transport Group Inventory:

Overlying Time (h) - 0.0001 Atmosphere Sump Pool Noble gases (atoms) 1.5969E+19 0.0000E+00 0.0000E+00 If 6If 666))))I~ Elemental I (atoms) 1.8787E+17 0.0000E+00 0.0000E+00 If 8 8 2.8223E+14 0.0000E+00 0.00001+00

((((6 Organic I (atoms) ff If If If a88 6f If Aerosols (kg) 0.0000E+00 0.0000E+00 0.0000E+00 If If 6 8f If Deposition Recirculating

1. 1. f If It If # 8t 8f 8 If #t 8f~If It 8f Time (h) h 0.0001 Surfaces Filter IfIflif ###"f~ Noble gases (atoms) 0.00001+00 0.0000E+00 Elemental I (atoms) 0.0000E+00 0.OOOOE+00 Organic I (atoms) 0.00002+00 0.00002+00 Aerosols (kg) 0.0000E100 0.0000E+00 Dose, Detailed model and Detailed Inventory Output Unfiltered Environment to CR Transport Group Inventory:

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) 4.24562-01 0.00002+00 1.4377E+00 Elemental I (atoms) 0.00003+00 Accumulated dose (rem) 4.2456E-01 0.00002+00 1.4377E+00 Organic I (atoms) 0.00001+00 Aerosols (kg) .00003E+00 Outer Boundary of the LPZ Doses:

Control Room Exhaust Transport Group Inventory:

Time (h) 0.0001 Whole Body Thyroid TEDE Delta dose (rem) 4.73932-02 0.00002+00 1.6048E-01 Pathway Accumulated dose (rem) 4.7393E-02 O.0000E+00 1.60482-01 Time (h) - 0.0001 Filter Noble gases (atoms) 0.00002+00 Control Room Doses: Elemental I (atoms) .00003E+00 Organic I (atoms) 0.00002+00 Time (hM = 0.0001 Whole Body Thyroid TEDE Aerosols (kg) 0.00001+00 Delta dose (rem) 1.20782-06 0.00002+00 4.56862-05 Accumulated dose (rem) 1.20782-06 0.00002+00 4.5686E-05 Exclusion Area Boundary Doses:

Control Room Compartment Nuclide Inventory: Time (h) - 2.0000 Whole Body Thyroid TEDE Delta dose (rem) 1.0214E-03 0.00002+00 3.4587E-03 Time (h) = 0.0001 Ci kg Atoms Bq Accumulated dose (rem) 4.25591-01 0.OOOOE+00 1.4411E+00 Br-82 6.0733E-04 5.60972-13 4.1198E+12 2.24712+07 Br-83 1.6156E-05 1.02272-15 7.4200E+09 5.9776E+05 Outer Boundary of the LPZ Doses:

Kr-83m 1.25002-02 6.0583E-13 4.3957E+12 4.6248E+08 8.1467E-01 2.0751E-06 1.4702E+19 3.01432+10 Time (h) = 2.0000 Whole Body Thyroid TEDE Kr-85 Kr-85m 1.62342-01 1.9727E-11 1.3976E+14 6.0067E+09 Delta dose (rem) 1.14022-04 0.OOOOE+00 3.8609E-0; 2.63012-05 9.2854E-16 6,4273E+09 9.73151+05 Accumulated dose (rem) 4.7507E-02 0.00002+00 1.6087E-01 Kr-87 Kr-88 5.01432-02 3.9989E-12 2.7366E+13 1.85532+09 Rb-87 4.42812-24 5.06132-20 3.5034E+05 1.6384E-13 Control Room Doses:

Rb-88 1.17162-05 9.76012-17 6.6792E+08 4.3349E+05 1.8020E+17 2,5228E+02 Time (h) - 2.0000 whole Body Thyroid TEDE 1-129 6.8183E-09 3.86012-08 1-130 1.48972-03 7.63812-13 3.5383E+12 5.51182+07 Delta dose (rem) 4.3741E-02 0.00002+00 1.80802+00 1-131 1.98342-01 1.5999E-09 7.3546E+15 7.3387E+09 Accumulated dose (rem) 4.3742E-02 0.0000E+00 1.80801+00 u m 03 1-132 1,57532-01 1.52622-11 6.9627E+13 5.82882+09 1-133 1.23912-01 1.09382-10 4.9529E+14 4.5847E+09 Control Room Compartment Nuclide Inventory: CD06 1-134 6.5286E-09 2.4473E-19 1.09992+06 2.4156E+02 2.03912-02 5.80632-12 2.59011+13 7.5446E+08 Time (h) - 2.0000 Ci kg Atoms Bq 1-135 Xe-129m 2.2597E-04 1.7859E-12 8.3372E+12 8.3607E+06 Br-82 5.8394E-04 5.39372-13 3.9611E+12 2.1606E+07 0)M 1 1.6378E+16 1.1041E+10 Br-83 9.0455E-06 5.7258E-16 4.15442+09 3.34682+05 Xe-131m 2.98422-01 3.5627E-09 5.0066E+01 2.6747E-07 1.21112+18 1.85252+12 Kr-83m 5.8665E-03 2.8434E-13 2.0631E+12 2.1706E+08 Xe-133 I-M Xe-133m 1.52552+00 3.3997E-09 1.5394E+16 5.6442E+10 Kr-85 8.14661-01 2.0750E-06 1.4701E+19 3.01432+10 0

Caic. No. 3.2.15.14, Rev. 0 Attachment 7 Page 6 of 9 Kr- 85m 1.1914U-01 1.4477E-11 1.0257E+14 4.40812E09 Delta dose (rem) 0.0000E+00 0.O000E+00 0.OOOOE+00 Kr- 87 8.84221-06 3.1216E-16 2.1608E+09 3.2716E+05 Accumulated dose (rem) 4.7507E-02 0.00002E00 1.6087E-01 Kr-88 3.077713-02 2.4 545E-12 1.6797E213 1.1388E+09 Rb- 87 5.3930E-20 6.1643E-16 4.2669E+09 1.9954E-09 Control Room Doses:

Rb-88 3.4843E-02 2.9027E-13 1.9864E+12 1.2892E+09 1-129 6.8183E-09 3.8601E- 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 4.9271E+07 Delta dose (rem) 9.7343E-02 0.O000E+00 4.9289E-01 1-131 1.9692E-01 1.5884E-09 7.3020E+15 7.2861E+09 Accumulated dose (rem) 1.4108E-01 0.0000E+00 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 Inventory:

1-134 1.3431E-09 5.0346E-20 2.2626E+05 4.9693E+01 1-135 1.6533E-02 4.7078E-12 2.1001E+13 6.1173E+08 Time (h) - 8.0000 Ci kg Atoms Bq Xe- 129m 2.2434E-04 1.7731E-12 8.2773E+12 8.3006E+06 Br- 82 9.2289E-10 8.5245E-19 6.2604E+06 3.4147E+01 Xe- 131m 2.9698E-01 3.5456E-09 1.6299E216 1.0988E+10 Kr-83m 6.0467E-04 2.9307E-14 2.1264E+11 2.2373E+07 Xe- 133 4.9536E+01 2.6464E-07 1.1983E+18 1.8328E+12 Kr-85 8.1463E-01 2.0749E-06 1.4701E+19 3.0141E+10 Xe-133m 1.4858E+00 3.3113E-09 1.4993E+16 5.4975E+10 Kr- 85m 4.7085E-02 5.7215E-12 4.0536E+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 5.5597E-03 6.1033E-14 2.7226E+11 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 3.2584E+01 Rb-87 3.4993E-22 3.9998E-18 2.7686E+07 1.2948E-11 Rb-88 4.0916E-03 3.4086E-14 2.3326E+11 1.5139E+08 Control Room Transport Group 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 2.7641E-15 1.2707E+10 1.2679E+04 Time (h) = 2.0000 Atmosphere Sump Pool 1-132 2. 5135E-Q8 2.4350E-18 1.1109E+07 9.2999E+02 Noble gases (atoms) 1.5969E+19 0.0000E+00 0.0000E+00 1-133 1.6877E-07 1.4898E-16 6.7457E+08 6.2443E+03 Elemental I (atoms) 1.8787E+17 0.O000E+00 O.0000E+00 1-135 1.5670E-08 4.4619E-18 1.99042+07 5.7977E+02 Organic I (atoms) 2.8223E+14 O.0000E+00 0.O000E+00 Xe-129m 2.1953E-04 1.7351E-12 8.0999E+12 8.1227E+06 Aerosols (kg) 0.0000E+00 0.O000E+00 0.00002E00 Xe-131m 2.9269E-01 3.4944E-09 1.6064E216, 1.0830E+10 Xe-133 4.7973E+01 2.5629E-07 1.1605E+18 1 . 7750E+12 Deposition Recirculating Xe-133m 1.3727E+00 3.0592E-09 1.3852E+16 5.0790E+10 Time (h) = 2.0000 Surfaces Filter Xe-130 7.5012E+00 2.9373E-09 1.31032+16 2.7754E+11 Noble gases (atoms) 0.0000E+00 0.0000E+00 Xe-135m 1.5232E-08 1.6722E-19 7.4592E+05 5.6359E+02 Elemental I (atoms) 0.O000E+00 O.O0000+00 Cs- 135 1.2451E-10 1.0807E-10 4.8206E+14 4.6067E+00 Organic I (atoms) 0.0OO0E+00 O.0000+E00 Aerosols (kg) 0.0O00E+00 0.O0000+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) 1.5969E+19 0.0000E+00 O.0000E+00 Time (h) = 2.0000 Filter Elemental I (atoms) 3.3405E+11 0.0000E+00 0.0000E+00 Noble gases (atoms) 0.O000E+00 Organic I (atoms) 5.0183E+08 0.0000+E00 0.0000+E00 Elemental I (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 0.0000E+00 O.0000E+00 Organic I (atoms) 0. 000E+00 Aerosols (kg) 0.0000E+00 Deposition Recirculating Time (h) - 8.0000 Surfaces Filter Control Room Exhaust Transport Group Inventory: Noble gases (atoms) O.0002E+00 0.0000E+00 Elemental I (atoms) 0.0000+E00 1.8787E+17 Pathway Organic I (atoms) 0.0000E+00 2.8223E+14 Time (h) = 2.0000 Filter Aerosols (kg) 0.0000E+00 0.0000E+00 Noble gases (atoms) 0.0000E+00 Elemental I (atoms) 0.0000E+00 Unfiltered Environment to CR Transport Group Inventory:

Organic I (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 Pathway Time (h) - 8.0000 Filter Exclusion Area Boundary Doses: Noble gases (atoms) 0.0000E+00 -u-am Elemental I (atoms) 0.0000E+00 o')

Time (h) - 8.0000 Whole Body Thyroid TEDE Q0 Qo I Organic I (atoms) 0.0000E+00 Delta dose (rem) 0.O00002+O 0.OOO2E+00 O.00002+00 Aerosols (kg) 0.-0000+00 Accumulated dose (rem) 4.2559E-01 0.0000E+00 1.44112+00 Control Room Exhaust Transport Group Inventory:

Outer Boundary of the LPZ Doses: oz Pathway 0)'

0)

Time (h) - 8.0000 Whole Body Thyroid TEDE Time (h) - 8.0000 Filter 0

(0 r,

Caic. No. 3.2.15.14, Rev. 0 Attachment 7 Page 7 of 9 Noble gases (atoms) o.OOOOE.00 Control Room Exhaust Transport Group Inventory:

Elemental I (atoms) 0.0000E+00 Organic I (atoms) 0.0000+E00 Pathway Aerosols (kg) O.0000E+00 Time (h) - 24.0000 Filter Noble gases (atoms) O.O000E+00 Exclusion Area Boundary Doses: Elemental. I (atoms) O.O000E+00 Organic I (atoms) O.O000E÷00 Time (h) = 24.0000 Whole Body Thyroid TEDE Aerosols (kg) O.O000E+00 Delta doae (rem) 0.0000E+0 0.OOOOE+00 0O.ooooE+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 TEDE Delta dose (rem) 0,0000E+00 0.0000E+00 0.0000+E00 Time (h) = 24.0000 Whole Body Thyroid TEDE Accumulated dose (rem) 4.2559E-01 0.0000E+00 1.4411E+00 Delta dose (rem) 0.0000E+00 0.0000E+00 0.0000E+00 Accumulated dose (rem) 4.7507E-02 0.0000E+00 1.6087E-01 Outer Boundary of the LPZ Doses:

Control Room Doses: Time (h) = 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 0.0000E+00 0.0000+E00 0.O000E+00 Time (h) = 24.0000 Whole Body Thyroid TEDE Accumulated dose (rem) 4.7507E-02 0.0000E+00 1.60872-01 Delta dose (rem) 1.6709E-01 0.0000E+00 1.6726E-01 Accumulated dose (rem) 3.0817E-01 0.0000E+00 2.4682E+00 Control Room Doses:

Control Room Compartment Nuclide Inventory: Time (h)( 96.0000 Whole Body Thyroid TEDE Delta dose (rem) 2.2611E-01 0.0000E+00 2.2611E-01 Time (h) = 24.0000 Ci kg Atoms Bq Accumulated dose (rem) 5.3428E-01 0.0000E+00 2.6943E+00 Kr-83m 1,4111E-06 6.8394E-17 4.9624E+08 5.2211E+04 Kr-85 8.1453E-01 2.0747E-06 1.4699E+19 3.0138E+10 Control Room Compartment Nuclide Inventory:

Kr-85m 3.9608E-03 4.8129E-13 3.4099E+12 1.4655E+08 Kr-88 1.4332E-04 1.1430E-14 7.8218E+10 5.3029E+06 Time (h) - 96.0000 Ci kg Atoms Bq Rb-88 9.7010E-05 8.0816E-16 5.5305E+09 3.5894E+06 Kr-85 8.1410E-01 2.0736E-06 1.4691E+19 3.0122E+10 Xe-129m 2.0721E-04 1.6377E-12 7.6453E+12 7.6668E+06 Kr-85m 5.7517E-08 6.9891E-1l 4.9517E+07 2.1281E+03 Xe-131m 2.8154E-01 3.3613E-09 1,5452E+16 1.0417E+10 Xe-129m 1.5978E-04 1.2628E-12 5.8953E+12 5.9119E+06 Xe-133 4.4031E+01 2.3523E-07 1.0651E+18 1.6292E+12 Xe-131m 2.3640E-01 2.8224E-09 1.2975E+16 8.74702+09 Xe-133m 1.1114E+00 2.4768E-09 1.1215E+16 4.1120E+10 Xe-133 2.9847E+01 1.5946E-07 7.2201E+17 1.1044E+12 Xe-135 2.2144E+00 8.6714E-10 3.8682E+15 8.1934E+10 Xe-133m 4.29642-01 9.5751E-10 4.3355E+15 1.5897E+10 Cs-135 3.9856E-11 3.4593E-11 1.5431E+14 1.4747E+00 Xe-135 9.1383E-03 3.5784E-12 1.5963E+13 3.3812E+08 Cs-135 2.2265E-13 1.9325E-13 8.6205E+11 8.2380E-03 Control Room Transport Group Inventory:

Control Room Transport Group Inventory:

Overlying Time (h)( 24.0000 Atmosphere Sump Pool Overlying Noble gases (atoms) 1.5969E+19 0.O000E+00 O.0000E+00 Time (h) = 96.0000 Atmosphere Sump Pool Elemental I (atoms) 1.5500E-04 0.0000E+00 0.0000E+00 Noble gases (atoms) 1.5969E+19 0.O000E+00 0.0000E+00 Organic I (atoms) 2.3286E-07 0.0000E+00 0.0000E+00 Elemental I (atoms) 1.5479E-73 0.0000+E00 0.0000E+00 Aerosols (kg) 0.00002E00 0.0000E+00 0.0000E+00 Organic I (atoms) 2.3253E-76 0.O0000S+00 0.0000E+00 Aerosols (kg) 0.00002+00 0.0000E+00 O.0000E+00 Deposition Recirculating Time (h) = 24.0000 Surfaces Filter Deposition Recirculating Noble gases (atoms) O.0000E+00 0.O000E+00 Time (h) = 96.0000 Surfaces Filter Elemental I (atoms) O.0000E+00 1.8787E+17 Noble gases (atoms) 0.0000E+00 0.0000E+00 Organic I (atoms) 0.0000E+00 2.8223E+14 Elemental I (atoms) O.O000E+00 1.8787E+17 Aerosols (kg) 0.0000+E00 0.0000E+00 Organic I (atoms) 0.0000E+00 2.8223E+14 Aerosols (kg) O,00002+00 0.0000E+00 *rm Unfiltered Environment to CR Transport Group Inventory:

CD 0 Unfiltered Environment to CR Transport Group Inventory:

Pathway Time (h) - 24.0000 Filter Pathway (0 - C Noble gases (atoms) O.0000E+00 Time (h) = 96.0000 Filter P Elemental I (atoms) 0.0000E+00 0 mo Noble gases (atoms) O.0000E+00 Organic I (atoms) O.0000E+00 Elemental I (atoms) O.0000E+00 Aerosols (kg) O.0000E+00 Organic I (atoms) O.0000+E00 Aerosols (kg) 0.00002+00

Caic. No. 3.2.15.14, Rev. 0 Attachment 7 Page 8 of 9 Control Room Exhaust Transport Group Inventory: Pathway Time (h) = 720.0000 Filter Pathway Noble gases (atoms) o.0000E+00 Time (h) = 96.0000 Filter Elemental I (atoms) 0.0000E+00 Noble gases (atoms) 0.00OOE400 Organic I (atoms) 0.OOOOE+00 Elemental I (atoms) 0.OOOOE+00 Aerosols (kg) o0.0000E00 Organic I (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 2223 Exclusion Area Boundary Doses:

1-131 Summary Time (h) - 720.0000 Whole Body Thyroid TEDE #8#### 8#888#8888 888888 fff#8 # #f 888#888888f8ff8# 8888888f#####8# 88####

Delta dose (rem) 0.OOOOE+00 0.00000+00 0.0000E+00 Accumulated dose (rem) 4.2559E-01 0.0000+00 1.4411E+00 Containment Environment Control Room Time (hr) 1-131 (Curies) 1-131 (Curies) 1-131 (Curies)

Outer Boundary of the LPZ Doses: 0.000 4.3771E-01 1.8194E+02 1.9834E-01 0.275 0.0000E+00 1.8238E+02 1.9815E-01 Time (h) = 720.0000 Whole Body Thyroid TEDE 0.525 O.0000E+00 1.8238E+02 1.97970-01 Delta dose (rem) 0.O000E+00 0.0000E+00 0.OOOOE+00 0.775 0.0000+E00 1.8238E+02 1.9779E-01 Accumulated dose (rem) 4.7507E-02 0.0000E+00 1.6087E-01 1.025 0.0000E+00 1.8238E+02 1.9761E-01 1.275 0.0000E+00 1.8238E+02 1.9744E-01 Control Room Doses: 1.525 O.0000E+00 1.8238E+02 1.9726E-01 1.775 0.0000E+00 1.8238E002 1.9708E-01 Time (h) - 720.0000 Whole Body Thyroid TEDE 2.000 0.0000+E00 1.8238E+02 1.9692E-01 Delta dose (rem) 2.7824E-01 0.0000+E00 2.7824E-01 2.250 0.0000E+00 1.8238E+02 1.1332E-01 Accumulated dose (rem) 8.1252E-01 0.0000E+00 2.9725E+00 2.500 0.0000E+00 1.8238E+02 6.5215E-02 2.750 0.0000+E00 1.8238E+02 3.7529E-02 Control Room Compartment Nuclide Inventory: 3.000 O.0000E+00 1.8238E+02 2.1597E-02 3,250 0.0000E+00 1.8238E+02 1.2428E-02 Time (h) = 720.0000 Ci kg Atoms Bq 3.500 0.0000+E00 1.8238E+02 7.1523E-03 Kr-85 8.1036E-01 2.0641E-06 1.46240+19 2.9983E+10 3.750 0.0000E+00 1.8238E+02 4.1159E-03 Xe-129m 1.6795E-05 1.3274E-13 6.1967E+11 6.2141E+05 4.000 0.00000+00 1.8238E+02 2.3686E-03 Xe-131m 5.1993E-02 6.2073E-10 2.8535E+15 1.92378+09 4 .250 0.00000E00 1.8238E+02 1.3631E-03 Xe-133 9.7071E-01 5.1859E-09 2.3481E+16 3.5916E010 4.500 0.0000E+00 1.8238E+02 7.8441E-04 Xe-133m 1.1375E-04 2.5352E-13 1.1479E+12 4.2089E+06 4.750 0.0000E+00 1.8238E+02 4.5140E-04 5.000 0.0000E+00 1.8238E+02 2.5977E-04 Control Room Transport Group Inventory: 5.250 O.0000E+00 1.8238E+02 1.4949E-04 5.500 0.0000+E00 1.8238E+02 8.6028E-05 Overlying 5.750 0.0000E+00 1.8238E+02 4.9507E-05 Time (h) - 720.0000 Atmosphere Sump Pool 6.000 0.0000E+00 1.8238E+02 2.8490E-05 Noble gases (atoms) 1.5969E+19 0.0000E+00 0.0000E+00 6.250 0.0000E+00 1.8238E+02 1.6395E-05 Elemental I (atoms) 0.0000E+00 O.0000E+00 0.0000E+00 6.500 0.0000E+00 1.8238E+02 9.4349E-06 Organic I (atoms) 0.0000E+00 0.0000E+00 0.0000E+00 6.750 0.0000+E00 1.8238E+02 5.4295E-06 Aerosols (kg) 0.0000E+00 0.0000E+00 0.0000E+00 7.000 0.0000+E00 1.8238E+02 3.1245E-06 7.250 0.0000+E00 1.8238E+02 1.7981E-06 Deposition Recirculating 7.500 0.0000E+00 1.8238E+02 1.0348E-06 Time (h) - 720.0000 Surfaces Filter 7.750 0.0000E+00 1.8238E+02 5.9547E-07 Noble gases (atoms) 0.0000E+00 O.0000E+00 8.000 0.0000E+00 1.8238E+02 3.4268E-07 Elemental I (atoms) 0.O000E+00 1.8787E+17 8.400 0.0000E+00 1.8238E+02 1.4155E-07 Organic I (atoms) 0.0000E+00 2.8223E+14 8.700 0.0000E+00 1.8238E+02 7.2938E-08 Aerosols (kg) 0.O0000+00 0.0000E+00 9.000 0.0000E+00 1.8238E+02 3.7582E-08 9.300 0.0000E+00 1.8238E+02 1.9365E-08 Unfiltered Environment to CR Transport Group Inventory: 9.600 0.0000E+00 1.8238E+02 9.9780E-09 9.900 0.0000E+00 1.8238E+02 5.1413E-09 T TM Pathway 10.200 0.0000E+00 1.8238E+02 2.6491E-09 Time (h) = 720.0000 Filter 24.000 0.0000E+00 1.8238E+02 1.5013E-22 Noble gases (atoms) 0.0000E+00 96.000 0.0000E+00 1.8238E+02 1.1575E-91 Elemental I (atoms) 0.O000E+00 720.000 0.OOOOE+00 1.8238E+02 O.0000E+00 Organic I (atoms) 0.0000E+00 Aerosols (kg) 0.0000E+00 Cumulative Dose Summary 0)

Control Room Exhaust Transport Group Inventory: 8888#888888888888#8#88#88888##8888888888888888888#####888888888888u8*#

0 (0

r-

Caic. No. 3.2.15.14, Rev. 0 Attachment 7 Page 9 of 9 Exclusion Area Bounda Outer Boundary of the (hr) (rem) (rem)

Control Room 0.0 1.1402E-04 Time Thyroid TEDE Thyroid 0.0000E+00 (rem)

3. 8609E-04 TEDE Thyroid TEDE (hr) (rem) (rem) (rem) (rem) (rem) (rem) Control Room 0.000 0.0000E+00 1.4377E+00 0.O000E+00 1.6048E-01 0.0000E+00 4.5686E-05 Time Whole Body Thyroid 0.275 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 TEDE 0.0000E+00 2.5085E-01 (hr) (rem) 0.525 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 (rem) (rem)

O.0000E+O0 4.7825E-01 0.0 4.3741E-02 0.775 0.0000E+00 1.4411E+00 0.0000S+00 1.6087E-01 0.0000+E00 1.80802+00 0.0000E+00 7.0504E-01 1.025 0.0000E÷00 1.4411E+00 0.0000S+00 1.6087E-01 1.275 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O,0000E+00 9.3124E-01 O.0000E+00 1.1569E+00 1.525 0.O000E+00 1.4411E+00 0.OOOOE+00 1.6087E-01 O.0000E+00 1.3819E+00 1.775 0.0000E+00 1.4411E+00 0.OOOOE+00 1.6087E-01 0.0000E+00 1.6064E+00 2.000 0.00000E+0 1.4411E+00 0.O000E+00 1.6087S-01 0.00002+00 1.8080E+00 2.250 0.0000S+00 1.4411E+00 0.OOOOE+00 1.6087E-01 O.0000E+00 1.9809E+00 2.500 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.0823E+00 2.750 0.OOOOE+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.1424E+00 3.000 0.O000E+00 1.4411E+00 0.OOOOE+00 1.6087E-01 O.0000E+00 2.17892+00 3.250 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+O0 2.2017E+00 3.500 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.2166E+00 3.750 0.0000E+00 1.44112+00 0.0000E+00 1.6087E-01 O.0000E+00 2.2270E+00 4.000 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000+E00 2.2347E+00 4.250 0.O000E+00 1.4411E+00 0.00000E+0 1.6087E-01 0.0000S+00 2.2410E+00 4.500 0.0000E+00 1.4411E÷00 0.0002E+00 1.6087E-01 0.0000E+00 2.2463E+00 4.750 0.OOOOE+00 1.4411E+00 0.00000E+0 1.6087E-01 0.0000E+00 2.2510E+00 5.000 0.0000E+00 1.4411E+00 0.OOOOE+00 1.6087E-01 O.0000E+00 2.2555E+00 5.250 0.0000E+00 1.4411E+00 0.O000E+00 1.6087E-01 O.0000E+O0 2.2597E+00 5.500 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.2638E+00 5.750 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.2677E+00 6.000 0.OOOOE+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.2716E+00 6.250 0.OOOOE+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000EO0 2.2755E+00 6.500 0.OOOOE+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.2792E+00 6.750 0.0000E+00 1.4411E+00 0,O000E+00 1.6087E-01 0.0000E+00 2.2830E+00 7.000 0.O000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000+00 2.2866E+00 7.250 0.O000E+00 1.4411E+00 0,00002+00 1,6087E-01 0.0000E+00 2.2903E+00 7.500 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.00002+00 2.2939E+00 7,750 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.2974E+00 8.000 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.3009E+00 8.400 0.0000E+00 1.4411E200 O.0000E+00 1.6087E-01 O.0000E+O0 2.3065E+00 8.700 0.O000E+00 1,4411E+00 0.0000E+00 1.6087E-01 O.0000E+00 2.3105E+00 9.000 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.3146E+00 9.300 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.3186E+00 9.600 0.0000E+00 1.4411E+00 0,0002E+00 1.6087E-01 0.0000E+00 2.3225E+00 9.900 0.0002E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.3264E+00 10.200 0.0000E+00 1.4411E+00 0.0000E+00 1.6087E-01 0.0000E+00 2.3302E+00 24.000 0.00000E+0 1.4411E+00 O.0000E+00 1.6087E-01 O.0000E+00 2.4682E+00 96.000 0.0000E+00 1.4411E+00 0.0000S+00 1.6087E-01 O.0000E+00 2.6943E+00 720.000 0,0000E,00 1.4411r.00 0.0000S.00 1.6087r-01 0.0000E+00 2.9725E+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 EAB dose location. Please disregard the two-hour worst doses for the other dose locations 88#8888##8#88888##888#######8#######8######8ffff####888##############

Exclusion Area Boundar, Time Whole Body Thyroid TEDE 0)

(hr) (rem) (rem) (rem) 0.0 1.0214E-03 0.O000E+00 3.4587E-03 v-0 M00 Outer Boundary of the LPZ Time Whole Body Thyroid 03 TEDE 0) co I-