L-13-306, Request for Licensing Action Pursuant to 10 CFR 50.59. 50.67. and 50.90: Full Implementation of Alternative Accident Source Term Design Basis Accident Analyses. and an Associated Technical Specification Change: Difference between revisions

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{{#Wiki_filter:FENOC'                                                                                      Perry Nuclear Power Plant P.O. Box97 10 Center Road RrstEnergy Nuclear Operating Company                                                                Perry. Ohio 44081 Ernest J. Hartmess                                                                                      440-280-5382 Vice President                                                                                    Fax: 440-280-8029 December 6, 2013 L-13-306                                                            10 CFR 50.59 10 CFR 50.67 10 CFR 50.90 ATTN: Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555-0001
{{#Wiki_filter:}}
 
==SUBJECT:==
 
Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 Request for Licensing Action Pursuant to 10 CFR 50.59. 50.67. and 50.90:
Full Implementation of Alternative Accident Source Term Design Basis Accident Analyses. and an associated Technical Specification Change A license amendment is requested for the Perry Nuclear Power Plant (PNPP), per the Code of Federal Regulations (CFR), Title 10, Sections (§) 50.67 Accident source term, §50.59 Changes, tests, and experiments, and §50.90 Application for amendment of license, construction permit, or early site permit.
FirstEnergy Nuclear Operating Company proposes to revise the PNPP Updated Safety Analysis Report (USAR) to reflect updated radiological calculations using an alternative accident source term for the applicable design basis events addressed by Regulatory Guide (RG) 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. 10 CFR 50.67 requires licensees who seek to revise a radiological source term utilized in design basis radiological consequence analyses to apply for a license amendment under §50.90. Accordingly, updates to the PNPP USAR summaries of these accidents are provided for Nuclear Regulatory Commission (NRC) review and approval.
The proposed USAR changes were evaluated under the criteria of §50.59 and were determined to result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases, which also requires amending the operating license pursuant to §50.90.
Implementation of these USAR changes also requires modification of the definition of DOSE EQUIVALENT IODINE-131 in the PNPP Technical Specifications, per §50.90.
 
Perry Nuclear Power Plant L-13-306 Page2 An evaluation of the proposed revisions is enclosed. Approval of the amendment is requested by February 27, 2015, to support restart following PNPP's spring 2015 refueling outage.
There are no regulatory commitments established in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager- Fleet Licensing, at (330) 315-6810.
I declare under penalty of perjury that the foregoing is true and correct. Executed on December ~ , 2013.
Sincerely,
  ;;i-Ernest J. Harkness
 
==Enclosure:==
Evaluation of the Revised Dose Calculation Request for Licensing Action cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager State of Ohio (NRC Liaison)
Utility Radiological Safety Board
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 1of43
 
==Subject:==
Request for Licensing Action Pursuant to 10 CFR 50.59, 50.67, and 50.90:
Full Implementation of Alternative Accident Source Term Design Basis Accident Analyses, and an Associated Technical Specification Change
: 1.
 
==SUMMARY==
DESCRIPTION
: 2. DETAILED DESCRIPTION
: 3. TECHNICAL EVALUATION
: 4. REGULATORY EVALUATION 4.1    Applicable Regulatory Requirements/Criteria 4.2    Precedent 4.3    Significant Hazards Consideration 4.4    Conclusion
: 5. ENVIRONMENTAL CONSIDERATION Addenda:
: 1. Updated Safety Analysis Report (USAR) Page Changes (Mark Up)
: 2. Associated Technical Specification (TS) Page Change (Mark Up)
: 3. Associated TS Page Change (Re-typed - For Information) *
: 4. Summary of Loss of Coolant Accident (LOCA) Dose Calculation
: 5. Summary of Control Rod Drop Accident (CRDA) Dose Calculation
: 6. Summary of Main Steam Line Break Outside Containment (MSLBOC) Dose Calculation
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 2of43 1.0
 
==SUMMARY==
DESCRIPTION This evaluation supports a request to amend the Operating License No. NPF-58 for the Perry Nuclear Power Plant (PNPP), per the Code of Federal Regulations (CFR),
Title 10, Sections(§) 50.67 Accident source term, §50.59 Changes, tests, and experiments, and §50.90 Application for amendment of license, construction permit, or early site permit.
FirstEnergy Nuclear Operating Company proposes to revise the PNPP Updated Safety Analysis Report (USAR) to reflect updated radiological calculations using an alternative accident source term (AST) for the applicable design basis events addressed by Regulatory Guide (RG) 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. 10CFR 50.67
* requires licensees who seek to revise a radiological source term utilized in design basis radiological consequence analyses to apply for a license amendment under
§50.90, and the application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report... ,... ***
Accordingly, updates to the PNPP USAR summaries of these accidents are provided for Nuclear Regulatory Commission (NRC) review and approval.
The proposed USAR changes were evaluated under the criteria of 10 CFR 50.59 and were determined to result in a departure from a method of evaluation described iri the Final Safety Analysis Report (FSAR) (as updated) used in establishing the design bases, which also requires amendment of the operating license pursuant to §50.90.
Implementation of these USAR changes requires modification of the definition of*
* DOSE EQUIVALENT IODINE-131 in the PNPP Technical Specifications, per §50.90.
2.0 DETAILED DESCRIPTION At PNPP, selective implementations of alternative AST radiological dose calculations were previously approved for the design basis loss of coolant accident (LOCA) and the fuel handling accident (FHA). The LOCA calculations were submitted to and reviewed by the NRC prior to the issuance of 10 CFR 50.67 and RG 1.183. Recently, calculations were performed to support a full implementation of an alternative AST for each of the applicable design basis events addressed by RG 1.183. The design basis events involving postulated fuel failures were analyzed using a source term for Global Nuclear Fuels (GNF) fuel type GNF2. The GNF2 fuel will begin replacing the currently used GE14 fuel bundles during the next fuel cycle at PNPP. Each of the boiling water reactor (BWR) design basis events addressed by RG 1.183, including those that only evaluate releases due to reactor coolant activity (no fuel failures), were re-analyzed using the methodology provided by NUREG/CR-6604, RADTRAD: A Simplified Model for Radionuclide Iransport and Removal And Dose Estimation. Use of the RADTRAD methodology for the design basis events in place of other previously used methodologies will provide consistency in the PNPP calculations, and permit use of the alternative AST approach under 10 CFR 50.59 and its supporting guidance when
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 3 of 43 assessing safety margins for future facility modifications or changes to procedures, as described in Section 1.1.1 of Regulatory Guide 1.183.
Sections of the USAR requiring revision as a result of changes to the input assumptions or the results of the revised dose calculations are included as Addendum 1 to this Evaluation. Summaries of the radiological dose calculations sufficient to identify the input assumptions and end results of the applicable RG 1.183 design basis events are included in Addenda 4 through 6. The USAR markups indicate the areas of change from previous design basis analyses, and the dose
* calculation summaries show that regulatory limits and guidance will continue to be met. Therefore the updated design/licensing basis is acceptable. Information on how the regulatory guidance was met in the development of the revised calculations is included in the Technical and Regulatory Evaluations below and in the attached*
summaries of the calculations.
One change to the Technical Specifications is directly linked to the evaluation ofthe .:
main steam line break (MSLB) outside containment (MSLBOC). This event involves only reactor coolant activity because no fuel failures are postulated; therefore the
* GNF2 source term is not applicable to it. However, the calculation results are proyided for NRC review as part of implementation of the RADTRAD code and to support *a **
* revision to the TS definition of DOSE EQUIVALENT 1-131 found in TS Section 1:t:- **
The definition is revised because the new MSLBOC calculation determines a DOSE
* EQUIVALENT 1-131 value using dose conversion factors from Federal Guidance-:;* .: *
* Report (FGR) 11 rather than from the Technical Information Document (TIO) 14844 document currently listed in the DOSE EQUIVALENT 1-131 definition. The marked*up TS Section 1.1 Definition page is provided in Addendum 2, with a retyped page *iit ***
Addendum 3.
No other TS changes are proposed as part of this request for licensing action, nor* any modifications to the plant other than the planned use of GNF2 fuel beginning with Cycle 16. Associated TS Bases changes to reflect use of the alternative AST and the dose acceptance criteria of 10 CFR 50.67 rather than the values in 10 CFR 100 Reactor Site Criteria will be processed separately per the TS Bases Control Program following NRC approval of the revised licensing basis.
 
==3.0 TECHNICAL EVALUATION==
 
===
Background===
In a Federal Register Notice dated December 23, 1999, the Nuclear Regulatory Commission (NRC) published a new regulation, 10 CFR 50.67 Accident source term, providing a mechanism for licensed power reactors to replace the traditional accident source term used in design basis accident analyses with alternative accident source terms based on improved knowledge about the radiological releases likely to occur during an accident. Regulatory guidance for the implementation of these alternative
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page4of43 ASTs is provided in RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000.
A number of previous PNPP regulatory activities have laid the groundwork for the current license amendment request.
* License Amendment 58, issued in April 1994, approved an analysis of a separate scenario for a control rod drop accident (CRDA) that does not include the Standard Review Plan (SRP) input assumption of a coincident loss of offsite power (LOOP), based on Licensing Topical Report NED0-31400, Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor. This licensing basis is retained as a scenario in the CRDA analysis.
* License Amendment 103, issued in March 1999, involved a pilot plant application of an alternative AST for a design basis loss of coolant accident.
Many aspects of the licensing basis from this initial PNPP alternative AST application are also retained.              .
* License Amendment 112, issued in June 2000., approved a five percent power*
uprate* for PNPP, which looked at four of the design basis events and increased prior dose calculation analysis results as necessary in order to conservatively
* bolind the power uprate.                                                      .
* During development of RG 1.183 (issued in July 2000), significant consideratiori> *.
was given to characteristics of the AST for a fuel handling accident. The        ** *
* RG 1.183 *guidance was utilized in the approval of PNPP Amendment 122 in .*
March 2003, which used RADTRAD to update the FHA radiological calculations to not take any credit for the Containment/Fuel Handling structures or for any ***.
* installed filtration systems (after at. least 24 hours of radiological decay). * *, * *
* Finally, in support of reviews of Generic Letter (GL) 2003-01 Control Room Habitability, it was determined appropriate to identify which design basis event is *most limiting for control room dose. Therefore, it was necessary to perform control room dose analyses for several design basis events that had never previously considered control room doses. Dose calculations were therefore performed for a MSLBOC, an instrument line break inside containment (ILBIC),
* and a CRDA. These calculations were not required to be submitted to the NRC at the time of the Generic Letter reviews, but the MSLBOC calculation and an updated CRDA calculation are now included as Addenda 5 and 6 as part of this full implementation of the alternative AST. The performance of these additional calculations showed that, as expected, the LOCA event was most limiting for control room dose.
The calculations done for the Control Room Habitability GL 2003-01 review of the MSLBOC were performed using RADTRAD Version 3.02. The CRDA calculations were re-performed in 2013 as part of this GNF2 implementation package using RADTRAD 3.03, since some fuel failures are postulated for such an event (if the rod pattern control (RPC) system and use of the banked position withdrawal sequence (BPWS) are not fully credited), requiring the CRDA dose calculations to use
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 5of43 the GNF2 source term. The other dose calculations performed in 2013 also used RADTRAD 3.03.
Technical Analysis Due to the comprehensive nature of Regulatory Guide 1.183, the following Degree of Conformance matrix summarizes conformance to its regulatory positions, to demonstrate the guidance is adequately addressed. This information supplements the summaries of the calculations (provided in Addenda 4 through 6).
Rggulato~      Gulde 1.183 Guidance                    Degree of Confonnance Section 1 Implementation of AST Regulatory Position 1. 1. 1 Safety Margins "The proposed uses of an AST and the associated        The revised radiological dose consequence proposed facility modifications and changes to        analyses using an AST were performed solely in procedures should be evaluated to determine            support of the introduction of the GNF2 fuel type in whether the proposed changes are consistent with      Cycle 16. No other facility modifications are the principle that sufficient safety margins are      proposed at this time. Sufficient safety* margins are maintained, including a margin to account for          maintained with the alternative AST analyses..
analysis uncertainties. The safety margins are        There are conservatisms in the calculations; which products of specific values and limits contained in    account for analysis uncertainties. The nuclide the technical specifications (which cannot be          inventory In the core is determined using an*
changed without NRC '
approval) and other values,  appropriate code (as described in Section 3~ 1 such as assumed accident or transient Initial          below), which is conservative enough to address conditions or assumed safety system response          uncertainties in the inventory and the radiolOgical times. Changes, or the net effects of multiple        decay process. The fraction of that inventory changes, that result in a reduction in safety margins  available in the gap of the fuel rods is assumed to be may require prior NRC approval.                        the same as provided in RG 1.183 (as described in Section 3.2 below), which addresses uncertainties in the gap fraction.* Dose calculation input assumption changes are summarized in the USAR markups and
                                                      *dose calculation summaries provided in the Addenda to this request for licensing action.
Once the Initial AST implementation has been          Consistent with this portion of RG 1.183 Regulatory approved by the staff and has become part of the      Position 1.1.1, since the initial AST implementation facility design basis, the licensee may use 10CFR      of the FHA using RADTRAD has been previously 50.59 and its supporting guidance in assessing        approved by the NRC staff and became part of the safety margins related to subsequent facility          PNPP design basis, the FHA dose calculations modifications and changes to procedures. n            performed in 2013 to support use of the GNF2 fuel type are being processed per 10 CFR 50.59, and do not require NRC review at this time. Therefore, this RG 1.183 degree of conformance matrix does not address RG 1.183 Appendix B Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident.
The margins to the regulatory limits for the other BWR-applicable RG 1.183 design basis events (LOCA, MSLB, and CRDA), are shown below for the control room, exclusion area boundary (EAB), and low population zone (LPZ). The calculation results meet the requirements of 10 CFR 50.67, and the guidelines of RG 1.183.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 6of43 Regulatorv Gulde 1.183 Guidance                        Degree of Confonnance LOCA Results (TEDE. rem)
Control Room EAB        LPZ RADTRAD Result 2.8                21.2      6.5 RG 1.183 limit      5            25        25 MSLB Results - 0.2 mlcrocurles/am (TEDE. rem>
Control Room EAB        LPZ RADTRAD Result 3.11 E-2          5.90E-2 6.59E-3 RG 1.183 limlt      5            2.5      2.5 MSLB Results - 4.0 mlcrocurles/gm <<TEDE. rem>
Control Room EAB        LPZ RADTRAD Result 6.22E-1            1.17      1.31E-1 RG 1.183 llmlt      5            25        25 CRDA Results - with LOOP (TEDE. rem>
Control Room EAB        LPZ RADTRAD R~sult 2.63E-1            1.61E-1 1.62E-1 RG 1.183 llmlt      5            6.3      6.3 CRDA Results - NED0-31400 CTEDE. rem>
Control Room EAB        LPZ RADTRAD Result 1.31E-6            2.80E-6 2.96E-6 RG 1.183 llmlt      5            6.3      6.3
: 1. 1.2 Defense in Depth "The proposed uses of an AST and the associated        The use of an AST In the revised radiological proposed facility modifications and changes to        consequence analyses does not impact the principle procedures should be evaluated to determine            of adequate defense In depth. Adequate defense in whether the proposed changes are consistent with      depth continues to be maintained through the use of the principle that adequate defense in depth is        Technical Specification controls over structures, maintained* to compensate for uncertainties in        systems, and components that are credited In the accident progression and analysis data. ...            design basis analyses, as specified in 10 CFR 50.36.
Proposed modifications that seek* to downgrade or      There are no proposed plant design modifications remove required engineered safeguards equipment        associated with this license amendment request should be evaluated to be sure that the modification  other than the planned use of GNF2 fuel beginning does not Invalidate assumptions made in facility      with Cycle 16, and the PNPP Probabilistic Risk PRAs and does not adversely impact the facility's      Assessment (PRA) model is not affected by the severe accident management program. n                  proposed amendment. Also, the Severe Accident Management (SAM) Program is not adversely Impacted by the proposed amendment.
: 1. 1.3 Integrity of Facility Design Basis
"... Although a complete re-assessment of all facility This application is considered to be a full radiological analyses would be desirable, the NRC      implementation of the alternative AST. The USAR staff determined that recalculation of all design      markups in Addendum 1 show that the source term analyses would generally not be necessary.            assumptions and radiological criteria in the previous Regulatory Position 1.3 of this guide provides        accident analyses have been superseded by the new guidance on which analyses need updating as part      analyses. Future revisions will use the updated of the AST implementation submittal and which may source term assumptions and radiological criteria.
need updating in the future as additional modifications are performed. This approach would create two tiers of analyses, those based on the previous source term and those based on an AST.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 7of43 Regulatory Gulde 1.183 Guidance                        Degree of Confonnance
... In either case, the facility design bases should 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 aooroved assumptions and criteria. ... "
: 1. 1. 4 Emergency Preparedness Applications
"... 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 itseff as a basis for requesting relief from the emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50. This guidance does not, however, preclude the appropriate use of the insights of the AST in establishing emergency response procedures such as those associated with
~mergency dose_projections, protective measures, and severe accident manaaement auides:"
1.2.1 Full Implementation "Full implementation Is a modification of the facility  This application is considered-to be a full design basis that addresses all characteristics of the  implementation, revising the plant licensing basis to AST, that Is, composition and magnitude of the          specify the AST In place of the previous accident radioactive material, its chemical and physical form,  source term, and establishing the total effective dose and the timing of its release. Full Implementation      equivalent (TEDE) dose as the new acceptance revises the plant licensing basis to specify the AST    criteria for design basis accidents. This applies not in place of the previous accident source term and      only to the analyses submitted herein (which consist establishes the TEDE dose as the new acceptance        of a subset of the plant analyses, as detailed below),
criteria. This applies not only to the analyses        but also to all future design basis analyses.
performed In the application (which may only Include    [Note: despite the above statement of Intent, the a subset of the plant analyses), but also to all future PNPP pilot plant adoption of an alternative AST design basis analyses.                                  retained limited applications of the TIO 14844 source term and the associated 10 CFR 100 acceptance criteria, which are again retained In this application.
Therefore, the USAR markups in Addendum 1 do nol entirely delete all references to the original accident source term and 10 CFR 100. For example, the markup of USAR Table 1.8-1 details the PNPP "Degree Of Conformance" to Regulatory Guides, and notes for RG 1.3 that original accident source term assumptions are retained for post-LOCA equipment qualification, vital area access, post-accident sampling system access, control room dose due to shine, and containment purge isolation analyses.
The proposed markups are not considered to be Inconsistent with the philosophy expressed in RG 1.183 Regulatory Position 1.2.1 - the AST could still be applied in the future to other design basis calculations that are not currently in this application.]
At a minimum for full implementations, the OBA          More events than just the design basis LOCA must be re-analyzed using the guidance In          accident (OBA) LOCA are addressed In this full Appendix A of this guide. Additional guidance on        implementation. Three of the BWR events in the analysis is provided in Reaulatorv Position 1.3 of this RG 1.183 Aooendices are included herein for
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 8 of 43 RegulatoOl Gulde 1.183 Guidance                      Degree of Conformance guide.                                                PNPP (LOCA, MSLBOC, and CRDA). As noted in RG 1.183 and again in NRC Regulatory Issue Summary (RIS) 2006-04 Experience With Implementation Of Alternative Source Terms, footnote 2, there are some existing calculations for events classified as accidents or limiting faults that do not need to be reanalyzed as part of a full implementation of the AST. The RIS example is for off-gas or waste gas system release events that do not need to be addressed for a full AST implementation unless a design change is being proposed for the waste gas tank or systems at the same time. No such design changes are proposed, and the PNPP discussions of the gaseous and liquid waste events are not re-analyzed at this time (note that the NRC Standard Review Plan sections addressing *both gaseous and liquid waste system events were deleted in Rev. 1-July1981). Another example is the PNPP USAR discussion of an Instrument Line Break Inside Containment, for Which the current licensing basis detailed in USAR Section 15.6.2.5 Radiological Consequences is that no Design Basis dose analysis is applicable, only a.
                                                      "Realistic Analysis.n The calculation results provided in the USAR are useful however, in that they Identify the total amount of reactor coolant released* from such a small line during a five hour plant shutdown performed to depressurize the reactor coolant system (since the small line is assumed to be unlsolable). The total mass of coolant released Is determined to be approximately one-fifth of the amount released from the Main Steam Line Break Outside Containment accident. That bounding MSLBOC evaluation does not credit any building holdup or ventilation system treatment and is a re-analyzed event per this full implementation alternative AST submittal. Since the MSLBOC event bounds the ILBIC, the USAR markups in Addendum 1 for the ILBIC note that its aReallsticn radiological dose calculation is historical In nature.
The USAR markups for the ILBIC are provided in Addendum 1, for information only, since that "Realisticn analysis does not need NRC review as part of an AST submittal.
Since the AST and TEDE criteria would become part    New (future) uses of the AST at PNPP would not of the facility design basis, new applications of the require prior NRC approval unless stipulated by AST would not require prior NRC approval unless        10 CFR 50.59 Changes, Tests, and Experiments, or stipulated by 10 CFR 50.59, "Changes, Tests, and      unless the new application involved a change to a Experiments," or unless the new application involved  technical specification. However, a change from an a change to a technical specification. However, a    approved AST to a different AST that is not approved change from an approved AST to a different AST        for use at PNPP would require a license amendment that is not approved for use at that facility would  under 10 CFR 50.67.
reauire a license amendment under 10 CFR 50.67.n
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 9 of 43 Regulatorv Gulde 1.183 Guidance                          Degree of Confonnance 1.2.2 Selective Implementation "Selective implementation is a modification of the      This application is not considered to be a selective facility design basis that (1) is based on one or        Implementation, as explained In Section 1.2.1 more of the characteristics of the AST or (2) entails    above. The description of how the design basis for re-evaluation of a limited subset of the design basis    these events is being maintained is included In the radiological analyses. The NRC staff will allow          USAR markups [Addendum 1].
licensees flexibility in technically justified selective implementations provided ... "
1.3.1 Design Basis Radiological Analyses "There are several regulatory requirements for which compliance is demonstrated, in part, by the evaluation of the radiological consequences of design basis accidents. These requirements include, but are not limited to, the following:
* Environmental Qualification of Equipment            10 CFR 50.49 Environmental Qualification of (10 CFR 50.49)                                      Equipment- Discussions of equipment qualification are provided in response to Regulatory Position 6 and Appendix I.
* Control Room Habitability (GDC-19 of                GDC 19 ConflDl Room Habltablllty- No changes Appendix A to 10CFR Part 50)                        to control room design or remote shutdown capabilities are associated with this request for licensing action (RLA). For each of the RG 1.183 BWR design basis events, a dose calculation for the control room was performed. The calculations show that doses remain less than the 5 rem TEDE limit in 10 CFR 50.67 (and GDC 19).
10 CFR 50 Appendix E Emergency Response
* Emergency Response Facility Habitability            Faclllty Habltablllty- The proposed changes do (Paragraph IV.E.8 of Appendix E to 10 CFR          not result in changes to emergency response facility Part 50)                                            habitability. 10 CFR 50 Appendix E does not contain habitability criteria, however NUREG-0737 Supplement 1 does. The only facility with a specific dose criterion Is the technical support center (TSC),
The dose limit In NUREG-0737 Supplement 1 for the TSC is 5 rem whole body, or its equivalent. The "or its equivalenr for this evaluation is considered to be 5 rem TEDE. Dose studies of the TSC were performed. The study is presented in Addendum 4 Sections 6.5, 6.6, and 6.8. The calculated dose values are reasonable since, as noted previously In the fuel handling accident AST submittal, the TSC ventilation intake is farther away from the containment structure and from ventilation system release points than the control room intake, and the TSC intake is at a lower elevation by more than 60 feet. Since the dispersion of a plume at the TSC intake at a greater distance and lower elevation would be correspondingly better than at the control room intake, the evaluations 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) habitability; regardless, in 2013, this facility was re-located more than 10 miles from
 
Evaluation of the Revised Dose Calculation Request For licensing Action Page 10 of 43 Regulato~      Gulde 1.183 Guidance                Degree of Conformance PNPP, so for the same reasons as discussed for the TSC, this facility is considered to not be adversely affected as a result of a change in source term assumptions.
10 CFR 50.67 Accident source term - The acceptance criteria of 10CFR 50.67 and the
* Alternative Source Term (10 CFR 50.67)        attributes of an acceptable alternative source term as described in Regulatory Guide 1.183 are being utilized in this application, as detailed herein.
10 CFR Part 51 Environmental Protection Regulations - Refer to Section 5.0 of this letter
* Environmental Reports (10 CFR Part 51)        entiUed Environmental Consideration, below.
10CFR100.11 Faclllty Siting-As noted in Footnote 5 of Reg. Gulde 1.183, the dose
* Facility Siting (10 CFR 100.11)                guidelines of 10 CFR 100.11 are superseded by 10CFR 50.67 for applications implementing an There may be additional applications of the        alternative source term such as this. The USAR is accident source term identified In the Technical    revised to change applicable references from Specification bases and in various licensee        Part 100 to 10 CFR 50.67.
commitments. These include, but are not limited to, NUREG-0737 Item 11.B.2 Post-Accident Access
.the following from Reference 2, NUREG-0737.        Shielding - This Three Mile Island (TMI) Action
* Post-Accident Access Shielding (NUREG-0737, 11.B.2)
Plan Item is addressed in USAR Appendix 1A, which references the reader to USAR Section 12.6.
This item is unaffected by implementation of the alternative AST, since as noted in the Degree of Conformance discussion above for Section 1.2.1, the original source term is still used for vital area aceess evaluations. In a letter dated January 18, 1999, as part of the approval process for the pilot plant LOCA AST, an integrated assessment was provided of vital area access issues, which concluded that the existing TIO-based radiological design basis for vital area access provides adequate margin, and a change to its design basis was not necessary. To ensure the existing licensing basis remains appropriate, an updated evaluation was performed for one of the post.;accldent vital area access locations as part of this reanalysis of the LOCA. That evaluation reached the conclusion that dose to the operator performing the post-accident task remained acceptable (less than 5 rem).
NUREG-0737 Item 11.B.3 Post-Accident Sampling
* Post-Accident Sampling Capability (NUREG-0737, 11.B.3)
Cspablllty- This TMI Action Plan Item is addressed in USAR Appendix 1A, which references the reader to USAR Section 9.3.6. This item is unaffected by Implementation of the alternative AST, since as noted in the, Degree of Conformance discussion above for Regulatory Position 1.2.1, the original source term Is still used for post-accident sampling capability evaluations. In a letter dated January 18, 1999, as part of the approval of the PNPP pilot plant LOCA AST, an integrated assessment was provided of post-accident
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 11of43 Regulatory Gulde 1.183 Guidance                      Degree of Confonnance sampling access, which concluded that existing TIO-based radiological design bases for post-accident sampling access provide adequate margin, and a change to the design basis was not needed.
That existing licensing basis remains unchanged for this submittal.
NUREG-0737, Item 11.F.1 Accident Monitoring
* Accident Monitoring Instrumentation            Instrumentation - This TMI Action Plan Item is (NUREG-0737, 11.F.1)                            addressed in USAR Appendix 1A, which references the reader to USAR Sections 6.2.5, 7.3, 7.6, and 11.5, along with the Emergency Plan. This new submittal makes no changes to plant systems or components, and does not affect the capability of Installed accident monitoring instrumentation to perform its function.
NUREG-0737ltein111.D.1.1 Leakage Contro/-
* Leakage Control (NUREG-0737, 111.D.1.1)        Technical Specification 5.5.2 Primary Coolant Sources Outside Containment requires a program to minimize leakage from systems outside containment that could carry radioactive fluids post-accident. The program includes administrative controls that limit such leakage to less than half of the 15 gallon per hour (gph) value assumed in the radiological dose calculations. This administrative limit is discussed in USAR Section 15.6.5.5.1.7
[Addendum 1].
NUREG-0737Item111.A.1.2 Emergency Response
* Emergency Response Facilities (NUREG-0737,      Facilities - Item 111.A.1.2 is unaffected, since no 111.A.1.2)                                      dose protection or habitability guidance is included in this TMI item. Additional information is provided above for Appendix E Emergency Response Facility Habitability.
+ Control Room*Habitabllity (NUREG-0737,              NUREG-0737, Item 111.D.3.4 Control Room
* 111.D.3;4)."                                    Habitability- Control Room habitability was analyzed for the LOCA, CRDA, and MSLB, and it was determined to be acceptable by meeting the 5 rem TEDE radiological dose limit for the control room in 10 CFR *50.67. lhe proposed amendment does not affect orotection from toxic aases.
1.3.2 Re-Analysis Guidance "Any implementation of an AST, full or selective, and The change .1s considered to be a full any associated facility modification should be        implementation. Compliance with various supported by evaluations of all significant          regulations, commitments, and other facility-specific radiological and non-radiological impacts of the      requirements are addressed in this Table. No facility proposed actions. This evaluation should consider    modifications other than the planned use of GNF2 the impact of the proposed changes on the facility's  fuel beginning with Cycle 16 are associated with the compliance with the regulations and commitments      proposed RLA, at this time. Since there are no listed above as well as any other facility-specific  physical design modifications being proposed, there requirements. These impacts may be due to (1) the    are also no non-radiological impacts as a result of associated facility modifications or (2) the          the proposed amendment.
differences in the AST characteristics.
The scope and extent of the re-evaluation will        This full implementation of the alternative AST necessarily be a function of the specific proposed    updated the LOCA, CRDA, and MSLB radiological
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 12of43 Regulatoat Gulde 1.183 Guidance                        Degree of Confonnance facility modification and whether a full or selective  analyses, and summaries of the calculations are implementation is being pursued. The NRC staff        included in Addenda 4 through 6 for NRC review.
does not expect a complete recalculation of all        Proposed USAR markups are provided in facility radiological analyses, but does expect        Addendum 1.
licensees to evaluate all impacts of the proposed changes and to update the affected analyses and the design bases appropriately. An analysis is considered to be. affected if the proposed modification changes one or more assumptions or inputs used In that analysis such that the results, or the conclusions drawn on those results, are no longer valid. .. .
Sensitivity analyses, discussed below, may also be    Where appropriate, sensitivity analyses have been an option.                                            performed; more detail is provided in Section 1.3.3.
If affected design basis analyses are to be re-        In the calculations, affected assumptions and inputs calculated, all affected assumptions and inputs        were updated to address ASTand TEDE, and should be updated and all selected characteristics of  selected characteristics of the AST and the TEDE the AST and the TEDE criteria should be addressed. criteria are addressed.
The license amendment request should describe the      Statements regarding the acceptability of the licensee's re-analysis effort and provide statements  proposed amendment against each of the applicable regarding the acceptability of the proposed            items identified in Regulatory Position 1.3.1 of implementation, including modifications, against      RG 1.183 were provided above.
each of the applicable analysis requirements and commitments identified in Regulatory Position 1.3.1 of this guide....
... For full implementation, a complete OBA LOCA      Similar to the discussion in Section 1.2.1 above, analysis as described in Appendix A of this guide      more events than just the design basis accident should be performed, as a minimum. Other design        (OBA) LOCA are addressed in this full basis analyses are updated in accordance with the      implementation.
guidance in this section. n 1.3.3 Use of Sensitivity or Scoping Analyses "It may be possible to demonstrate by sensitivity or  A scoping evaluation estimated the* significance of scoping evaluations that existing analyses have        dose to control room operators due to shine from sufficient margin and need not be recalculated.        filters in the control room emergency recirculation (ventilation) system, as described in the response to Regulatory Position 4.2, below.
As used In this guide, a sensitivity analysis is an    The following sensitivity evaluations were performed evaluation that considers how the overall results      (and are summarized here) to consider how overall vary as an input parameter (in this case, AST          analysis results vary as selected Input parameters in characteristics) Is varied. A scoping analysis Is a    a base case might vary.
brief evaluation that uses conservative, simple methods to show that the results of the analysis      For the LOCA, the results of a sensitivity study bound those obtainable from a more complete            reported in Addendum 4, Section 11.0 "Overall treatment. Sensitivity analyses are particularly      Results," demonstrate that initiation of the hydrogen applicable to suites of calculations that address      mixing system at 30 minutes post-accident results in diverse components or plant areas but are otherwise    slightly lower airborne leakage doses than the case largely based on generic assumptions and inputs.      that assumes the hydrogen mixing system is initiated Such cases might include post-accident vital area      at two hours post-accident. The two hour case is access dose calculations, shielding calculations, and  therefore considered bounding and became the
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 13 of 43 Regulatoa Gulde 1.183 Guidance                        Degree of Conformance equipment environmental qualification (integrated      base case.
dose). It may be possible to identify a bounding case, re-analyze that case, and use the results to    For the MSLBOC, a sensitivity analysis was draw conclusions regarding the remainder of the        performed that varied control room ventilation analyses. It may also be possible to show that for    assumptions to show doses remained acceptable, some analyses the whole body and thyroid doses        as summarized in Addendum 6, Section 6.4 "Control determined with the previous source term would        Room Ventilation Parameters." The base case bound the TEDE obtained using the AST. Where          shows that filtration systems are not required. The present, arbitrary "designer margins" may be          sensitivity shows that ventilation and filtration adequate to bound any impact of the AST and TEDE      systems can be effectively used to reduce doses to criteria. If sensitivity or scoping analyses are used, the control room operators in the event the normal the license amendment request should include a        control room ventilation intake is isolated during the discussion of the analyses performed and the          event, rather than continuing to run. The MSLBOC conclusions drawn. Scoping or sensitivity analyses    calculation also includes an evaluation assuming a should not constitute a significant part of the        larger control room volume, in its Attachment 6, evaluations for the design basis exclusion area        "Results for Increased Control Room Volume" boundary (EAB), low population zone (LPZ), or          [Addendum 6). The results of the larger control control room*dose."                                    room evaluation (for the 4 &#xb5;Ci/g pre-existing iodine spike analysis) were slightly higher than the base case, therefore the control room dose value reported in Addendum 6 Att. 6 is the dose value that is reflected in the USAR markups for the 4 &#xb5;Ci/g event.
For the CRDA, the sensitivities varied control room ventilation to show doses remained acceptable, as summarized in Addendum 5, Section 4.2 "Control Room Isolation and In leakage," and in Addendum 5, Attachment 1 "UFSAR Scenario 2." Again, the base case shows that filtration systems are not required.
The sensitivity shows that ventilation and filtration systems can be effectively used to reduce doses to the control room operators in the event the normal control room ventilation Intake is isolated during the event. rather than continuing to run.
1.3.4 Updating Analyses Following Implementation "Full implementation of the AST replaces the          This is a full Implementation of the alternative AST.
previous accident source term with the approved AST and the TEDE criteria for all design basis        Since the USAR discussions of the design basis radiological analyses. After the implementation is    events will include the AST and TEDE criteria complete, there may be a subsequent need (e.g., a      [Addendum 1], future updates to the associated planned facility modification) to revise these        calculations will use the characteristics of the AST analyses or to perform new analyses. For these        and the TEDE criteria under the provisions of recalculations, the NRC staff expects that all        10 CFR 50.59.
characteristics of the AST and the TEDE criteria incorporated Into the design basis wlll be addressed  A Footnote 7 to this section notes that it may be in all affected analyses on an individual as-needed    desired to have a method for comparing former dose basis.... Since [for a full implementation] the AST    results (previously expressed in terms of both whole and the TEDE criteria are part of the approved        body and thyroid) with the new results (expressed in design basis for the facility, use of the AST and      terms of TEDE). To make such a comparison for the TEDE criteria in new applications at the facility do  CRDA and MSLB, the formula in Footnote 7 is used, not constitute a change in analysis methodology that  since TEDE doses for those events have not been would require NRC approval."                          previously docketed. When the Footnote 7 formula is applied, conclusions can then be reached for all the events regarding whether doses decreased,
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 14of43 Regulatorv Gulde 1.183 Guidance                      pegree of Conformance were negligibly changed, or increased.
Specifically, for the LOCA, the newly calculated TEDE doses at the exclusion area boundary (EAB) and the low population zone (LPZ) boundary are unchanged or lower than previously evaluated.
For the MSLB scenarios (two) and the CRDA scenarios (two), the EAB and LPZ doses are either negligibly changed or are lower than previously evaluated, except for the CRDA Seenario 1 analysis, for which the calculated doses increase, but by less than 2 percent of the margin to the regulatory acceptance criterion. For the CRDA, the acceptance criterion in RG 1.183 Table 6 is 6.3 rem TEDE, which is only 25 percent of the regulatory limit in 10 CFR 50.67.
With regard to control room dose, for the LOCA the newly calculated control room TEDE dose is lower.
For the MSLB and CRDA, control room dose calculations were not previously required, so a comparison Is not possible.
1.3. 5 Equipment Environmental Qualification ucurrent environmental qualification (EQ) analyses    There are no increased EQ requirements as a result may be impacted by a proposed plant modification      of this proposed amendment. There are no associated with the AST implementation. The EQ        proposed plant modifications associated with this analyses that have assumptions or inputs affected    AST implementation other than the planned use of by the plant modification should be updated to        GNF2 fuel beginning with Cycle 16. Further details address these impacts.                                on EQ are provided in response to Regulatory Position 6.
The NRC staff is assessing the effect of increased    The cesium issue discussed in this section of the cesium releases on EQ doses to determine whether      Regulatory Guide is discussed in response to licensee action is warranted .... n                  Regulatory Position 6 below.
1.4 Risk Implications "The use of an AST changes only the regulatory        As noted in Section 1.1.2, the PNPP PRA model is assumptions regarding the analytical treatment of    not affected by the proposed amendment.
the design basis accidents. The AST has no direct effect on the probability of the accident. Use of an AST alone cannot increase the core damage frequency (CDF) or the large early release frequency (LERF). However, facility modifications made possible by the AST could have an impact on risk. If the proposed implementation of the AST involves changes to the facility design that would invalidate assumptions made in the facility's PRA, the impact on the existing PRAs should be evaluated. n
: 1. 5 Submittal Requirements u ... The NRC staffs finding that the amendment      Detailed summaries of the dose calculations are may be approved must be based on the licensee's      being provided for NRC review as Addenda 4 - 6.
analyses, since it is these analyses that will become Additional detail on how the NRC guidance in part of the design basis of the facility. The        Regulatory Guide 1.183 is being met is provided in amendment reauest should describe the licensee's      this table format.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 15 of 43 Regulatory Gulde 1.183 Guidance                        Degree of Conformance analyses of the radiological and nonradiological impacts of the proposed modification in sufficient detail to support review by the NRC staff.
The staff recommends that licensees submit affected    USAR pages, which show the changes that will be FSAR pages annotated with changes that reflect the      made, are included for NRC review and approval as revised analyses or submit the actual calculation      Addendum 1. Addenda 4 through 6 provide documentation.                                          summaries of the calculations, as noted above.
If the licensee has used a current approved version    The code used in the design basis radiological of an NRC-sponsored computer code, the NRC staff        consequence analyses was RADTRAD. The review can be made more efficient if the licensee      addenda summarize the inputs to the calculations.
identifies the code used and submits the inputs that the licensee used in the calculations made with that code. In many cases, this will reduce the need for .
NRC staff confirmatory analyses."                .
1.6 FSAR Requirements
* ... The regulations in 10 CFR 50.71(e) require that  USAR pages are provided in Addendum 1, which the FSAR be updated to include all safety              show how the affected radiological analysis evaluations performed by the licensee in support of    descriptions will be revised as a result of this requests for license amendments... The analyses        proposed amendment.
required by 10 CFR 50.67 are subject to this requirement. The affected radiological analysis descriptions in the 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. .."
Section 2 Attributes Of An Acceptable AST
*... Regulatory Position 3 of this guide identifies an  This application uses characteristics of the source AST that is acceptable to the NRC staff for use at      term outlined in Regulatory Position 3 of RG 1.183.
operating power reactors. A substantial effort was      Therefore the rest of Section 2 Is considered to be expended by the NRC, its contractors, various          not applicable, since no attempt Is made to define national laboratories, peer reviewers, and other8 in    different source term characteristics from those performing severe accident research and in              provided in the regulatory guide.
developing the source terms provided in NUREG-1465. However, future research may                The GNF2 source term is described below. The identify opportunities for changes in these source      MSLB event uses only a reactor coolant activity terms. The NRC staff will consider applications for    level, which is independent of fuel type or failures.
an AST different from that identified In this guide."
Section 3 Accident Source Term
: 3. 1 Fission Product Inventory "The inventory of fission products in the reactor      For the LOCA and CRDA events, which postulate core and available for release to the containment      fuel failure, generation of the GNF2 equilibrium core should be based on the maximum full power              inventory source term for PNPP was performed by operation of the core with, as a minimum, current      GE Hitachi Nuclear Energy, LLC (GEH) using plant licensed values for fuel enrichment, fuel bumup,        specific parameters as input Into ORIGN01 P and an assumed core power equal to the current          (ORIGN01 P is the GEH controlled version of the licensed rated thermal power times the ECCS            Oak Ridge National Laboratory code ORIGEN2, evaluation uncertainty. The period of irradiation      which is the Isotope generation and depletion code should be of sufficient duration to allow the activity  referenced in RG 1.183). GEH assumed a core of dose-significant radionuclides to reach              power equal to the current licensed thermal power.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 16 of 43 Regulatorv Guide 1.183 Guidance                    Degree of Confonnance equilibrium or to reach maximum values. The        For the dose calculations, the inventory provided by core inventory should be determined using an        GEH was increased by an additional 2 percent to appropriate isotope generation and depletion        account for evaluation uncertainty [Addendum 4 computer code such as ORIGEN 2... or                Section 6.2, and Addendum 5 Section 3.2.5).
ORIGEN-ARP ....
For the event that does not involve fuel failure, the MSLBOC, two different levels of reactor coolant activity are assumed [Addendum 6 Section 2). The equilibrium coolant inventory was developed based on infinite exposure time and no decay
[Addendum 6 page 5).
For the OBA LOCA, all fuel assemblies in the core  For the LOCA, the equilibrium core inventory is
* are assumed to be affected and the core average    provided by GEH for the GNF2 fuel [Addendum 4, inventory should be used.                          Section 6.2).
For OBA events that do not involve the entire core, For the CROA, to account for differences in power the fission product inventory of each of the        level across the core, a radial peaking factor of 2.0 damaged fuel rods is determined by dividing the    was applied, rather than the recommended total core inventory by the number of fuel rods in  minimum of 1.5. This simulates that the affected the core. To account for differences in power level rods have operated at a core power level greater across the core, radial peaking factors from the    than the average power level of the core facility's core operating limits report (COLR) or  [Addendum 5 Section 3.2.5). The radial peaking technical specifications should be applied in      factor of 2.0 used in the alternative AST determining the inventory of the damaged rods. ... calculations is a reload analysis parameter provided to the fuel vendor each cycle.
No adjustment to the fission product inventory      No adjustment to the fission product inventory was should be made for events postulated to occur      made for events postulated to occur during power during power operations at less than full rated    operations at less than full rated power or those power or those postulated to occur at the beginning postulated to occur at the beginning of core life.
of core life.... ."
3.2 Release Fractions aThe core Inventory release fractions, by          For the LOCA, RG 1.183 Table 1 values for BWR radionuclide groups, for the gap release and early  core inventory fraction released into containment in-vessel damage phases for OBA LOCAs are listed    were used [Addendum 4 Section 5.11).
in Table 1 for BWRs ... These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.
For non-LOCA events, the fractions of the core      For the CROA, Table 3 release fractions were inventory assumed to be in the gap for the various  modified (Increased) as described in .the last radionuclides are given in Table 3. The release    sentence of Footnote 11 [Addendum 5 Section fractions from Table 3 are used in conjunction with 3.1.5). For the MSLB, no fuel failure is postulated the fission product inventory calculated with the  to occur, consequently the release fractions of maximum core radial peaking factor."                Table 3 are not applicable [Addendum 6 page 5).
[Two footnotes are provided in this section)        For footnote 10, which applies to all of Section 3.2, Footnote 1O states UThe release fractions listed    the GNF2 fuel is an "approved LWR fuel."
here have been determined to be acceptable for use with currently approved LWR fuel with a peak    Discussion of Footnote 11 regarding a peak bumup burnup up to 62,000 MWO/MTU.                        limit of 62,000 MWO/MTU is provided below.
The data in this section may not be applicable to  There are no definitive plans to utilize MOX fuel at
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 17 of 43 Regulatory Gulde 1.183 Guidance                        Degree of Confonuance cores containing mixed oxide (MOX) fuel."              PNPP now or in the foreseeable future.
*Footnote 11 states "The release fractions listed        For footnote 11, which applies to Table 3, the here have been determined to be acceptable for          provisions in the first sentence of the footnote are use with currently approved LWR fuel with a peak        met at PNPP. The GNF2 fuel is an approved fuel, bumup up to 62,000 MWO/MTU provided that the            and the average exposure of the peak fuel rod is maximum linear heat generation rate does not            maintained below 62,000 MWO/MTU exceed 6.3 kw/ft peak rod average power for            (= 62 GWO/MTU). Also, the maximum linear heat bumups exceeding 54 GWO/MTU. As an                      generation rate (LHGR) for the fuel that could alternative, fission gas release calculations          exceed 54 GWO/MTU by the end of the cycle is performed using NRC-approved methodologies              maintained at or below 6.3 kw/ft. The bumup limit may be considered on a case-by-case basis. To be        of 62 GWO/MTU on the average exposure of the acceptable, these calculations must use a projected    peak rod, and the LHGR limit of 6.3 kw/ft peak rod power history that will bound the limiting projected    average power for the higher bumup fuel (> 54 plant-specific power history for the specific fuel      GWO/MTU), are both included in the list of reload load. For the BWR rod drop accident ... , the gap      analysis parameters that are re-verified each cycle.
fractions are assumed to be 10% for iodines and noble gases."
3.3 Timing of Release Phases "Table 4 tabulates the onset and duration of each      For the LOCA, assumptions regarding release sequential release phase for OBA LOCAs at PWRs          fractions and timing are consistent with Tables 1 and BWRs. The specified onset is the time following    and 4 of RG 1.183 [Addendum 4 Section 5.11].
                                          =
the initiation of the accident (i.e., time 0). The early Credit was taken for decay prior to the onset of the in-vessel phase immediately follows the gap release    gap release at 2 minutes [Addendum 4 phase.                                                  Section 6.2.1 ).
The activity released from the core during each        The core source terms are assumed to be released release phase should be modeled as increasing in a at a constant rate (in a linear fashion) such that the linear fashion over the duration of the phase.          release is completed by the end of the specified release period [Addendum 4 Section 5.11).
For non-LOCA OBAs in which fuel damage is              For the CROA, the activity released from the gap and projected, the release from the fuel gap and the fuel  the fuel pellets is assumed to be instantaneously pellet should be assumed to occur instantaneously      mixed in the reactor coolant within the pressure with *the onset of the projected damage.                vessel [Addendum 5, Section 3.1.5).
For facilities licensed with leak-before-break          PNPP is not licensed for leak-before-break methodology, the onset of the gap release phase        methodology so this regulatory position is not may be assumed ... n                                    applicable.
3.4 Radionuclide Composition "Table 5 (below) lists the elements in each            For the LOCA, the isotopes used in the calculation radionuclide group that should be considered in        are based on the isotopes specified in the design basis analyses:                                  RAOTRAD computer code. As stated in the RAOTRAO user's manual, NUREG/CR-6604, the Table 5                              60 isotope nuclide file is based on isotopes selected Radionuclide Groups                          in WASH-1400 with the addition of 6 Isotopes used Group                      Elements                    in the MACCS code [Addendum 4 Section 6.2).
Noble Gases                Xe, Kr Halogens                  I, Br                        For the CROA, the isotopes used are discussed in Alkali Metals              Cs, Rb                      Addendum 5 Sections 3.1.1 and 3.2.5.
Tellurium Group            Te, Sb, Se, Ba, Sr Noble Metals              Ru,Rh,Pd,Mo,Tc,Co            For the MSLB, no fuel damage is postulated. The Lanthanides                La, Zr, Nd, Eu, Nb, Pm, Pr  calculation evaluates the released activity
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 18 of 43 Regulato~      Gulde 1.183 Guidance                  Degree of Confonnance Sm, Y,Cm,Am                associated with the maximum coolant activity Cerium                    Ce, Pu, Np"                allowed by the PNPP Technical Specifications (i.e., 0.2 and 4.0 microcuries/gram (&#xb5;Ci/gm) dose equivalent 1-131) [Addendum 6 Section 4 Item 2).
Calculated values are In Addendum 6 Section 6.6.
: 3. 5 Chemical Form
* ... The accident-specific appendices to this        Specific details on chemical form are in the Regulatorv Gulde provide additional details.*        discussions of the RG 1.183 Aooendices, below.
3.~ Fuel Damage in Non-LOCA DBAs "The amount of fuel damage caused by non-LOCA        For the CRDA, when the bounding fuel enthalpy design basis events should be analyzed to            analysis is applied to GNF2 fuel, approximately determine, for the case resulting in the highest      1,200 fuel rods are determined to reach a fuel radioactivity release, the fraction of the fuel that  enthalpy of 170 cal/gm, at which point they are reaches or exceeds the initiation temperature of fuel assumed to experience cladding failure. For melt and the fraction of fuel elements for which the  conservatism, the PNPP calculation assumed fuel clad Is breached .... n                          additional cladding failures, totaling 1376 fuel rods (specifically, failure of all the fuel rod clads in the 16 bundles nearest to the postulated dropped rod).
A diagram showing this 16-bundle configuration is included in Addendum 5, Section 3.2.3. With respect to the percentage of fuel melt assumed in the CRDA, the maximum mass fraction in the damaged fuel that reaches temperatures in excess of the melting point is 0.0077 [Addendum 5 Section 3.2.31.
Section 4 Dose Calculational Methodology "The NRC staff has determined that there is an        The TEDE criteria are utilized in this AST implied synergy between the ASTs and total            application, which is performed pursuant to effective dose equivalent (TEDE) criteria, and        10 CFR 50.67.
between the TID-14844 source terms and the whole body and thyroid dose criteria, and therefore, they do not expect to allow the TEDE criteria to be used with TID-14844 calculated results. The guidance of this section applies to all dose calculations cerformed with an AST oursuant to 10 CFR 50.67 ... a
: 4. 1 Offsite Dose Consequences
*... 4. 1. 1 The dose calculations should determine  The dose calculations determine the TEDE.
the TEDE. TEDE is the sum of the committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (ODE) from external exposure.
The calculation of these two components of the        Refer to the discussion In Section 3.4, above. Also, TEDE should consider all radlonuclides, including    for the CRDA, as noted in Addendum 5 Section progeny from the decay of parent radionuclides that  3.1.1, although the initial core source terms were are significant with regard to dose consequences      generated with ORIGEN and consider the impact of and the released radioactivity."                      daughter products during operation, the calculation does not include daughter production during the release period. This is conservative because in RADTRAD, a choice to model daughter production during the release period must be coupled with use of "decay." Combining decay and daughter production would result in a lower calculated dose.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 19of43 Regulatorv Guide 1.183 Guidance                            Degree of Conformance
: 4. 1.2 * ... Table 2.1 of Federal Guidance Report 11,      The dose conversion factors utilized for the "Limiting Values of Radionuclide Intake and Air            committed effective dose equivalent (CEDE)
Concentration and Dose Conversion Factors for              inhalation component (of TEDE) were obtained Inhalation, Submersion, and Ingestion" (Ref. 20),          from the 1989 revision of Federal Guidance provides tables of conversion factors acceptable to        Report 11 [Addendum 4 Section 6.4, Addendum 5 the NRC staff. The factors in the column headed            Section 3.1.1, and Addendum 6 Section 6.3].
"effective" yield doses corresponding to the CEDE."
: 4. 1.3 "For the first 8 hours, the breathing rate of        The recommended offsite EAB and LPZ breathing persons offsite should be assumed to be 3.5 x 10-4          rates were used [Addendum 1, markup of USAR cubic meters per second. From 8 to 24 hours                Table 15.0-4). For the MSLB calculation, the values following the accident, the breathing rate should be        for the period after 8 hours appear slightly different assumed to be 1.8 x 10-4 cubic meters per second.          than the values in Regulatory Position 4.1.3, but After that and until the end of the accident, the rate      when rounded to the same number of significant should be assumed to be 2.3 x 10-4 cubic meters            digits, are the same.
per second."
: 4. 1.4 *... Table 111.1 of Federal Guidance Report 12,      The 1993 version of Federal Guidance Report 12 "External Exposure to Radionuclides in Air, Water,          was used [Addendum 1, markup of USAR Table and Soil" (Ref. 21 ), provides*extemal EDE                  15.0-4].
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-                For the LOCA, RADTRAD automatically calculates hour period following the start of the radioactivity        the worst two hour dose for the EAB.
release should be determined ... by calculating the postulated dose for a series of small time                  For CRDA, the release begins immediately and Increments and performing a "sliding0 sum over the          decreases exponentially as the condenser activity Increments for successive two-hour periods. The            decays and is depleted by leakage, so the first two maximum TEDE obtained is submitted.... (see also            hours is the worst case for the EAB dose and no Table 6)."                                                  sliding window calculations are performed
[Addendum 5 Section 3.1.1 Item 5].
For MSLB, the release is assumed Instantaneous, so the first two hours is the worst case for the EAB dose and no sliding window EAB dose calculations are performed [Addendum 6 Section 9].
4.1.6 "TEDE should be determined forthe 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 are in compliance with 10 CFR 50.67 determining compliance with the dose criteria in            limits [Addendum 4 Section 11.0, Addendum 5 10 CFR 50.67."                                              Section 6.0, and Addendum 6 Attachment 6].
: 4. 1. 7 "No correction should be made for depletion        Deposition of radionuclides on the ground is not of the effluent plume bv deoosition on the around. n        modeled in RADTRAD.
s4.2 Control Room Dose Consequences
".. .4.2.1 The TEDE analysis should consider all            For the CRDA, Addendum 5 Section 3.1.2 provides sources of radiation that will cause exposure to            information addressing the guidance in RG 1.183 control room personnel. The applicable sources will        Regulatory Positions 4.2.1through4.2.7.
vary from facility to facility, but typically will include:
+ Contamination of the control room atmosphere              For the LOCA, for Section 4.2.1, contamination of bv the Intake or infiltration of the radioactive      the control room atmosphere is modeled using the
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 20of43 Regulatory Gulde 1.183 Guidance                          Degree of Conformance material contained in the radioactive plume        control room X/Q atmospheric dispersion values in released from the facility,                        Addendum 4 Section 6.5, coupled with the
+ Contamination of the control room atmosphere          unfiltered inleakage assumptions into the control by the intake or infiltration of airborne          room in Sections 5.3 and 8.4. Discussions of radioactive material from areas and structures      radiation shine doses from the external plume and adjacent to the control room envelope,              from the containment are in Addendum 4
+ Radiation shine from the external radioactive        Section 8.4. With respect to radiation shine from plume released from the facility,                  control room emergency recirculation (CRER)
+ Radiation shine from radioactive material in the      system filters, a scoping evaluation was performed reactor containment,                                that concluded the dose to control room operators
+ Radiation shine from radioactive material in          from this source would be minor (less than 0.008 systems and components inside or external to        rem) due to existing protection by distance and the control room envelope, e.g., radioactive        shielding, which would not change the calculated material buildup in recirculation filters.          2.8 rem value for the control room.
For the MSLB Outside Containment, the information presented in Addendum 6 Sections 6 through 11 focuses on dose to the control room due to the first two bullets In RG 1.183 Section 4.2.1 (for the MSLB event, there is no external plume, and no shine from inside containment or from ventilation filters) 4.2.2 "The radioactive material releases and            For the LOCA and MSLB, the radioactive material radiation levels used in the control room dose          releases and radiation levels used in the control analysis should be determined using the same            room dose analysis were determined using the source term, transport, and release assumptions          same source term, transport, and release used for determining the EAB and the LPZ TEDE            assumptions used for determining the EAB and LPZ values, unless these assumptions would result in        TEDE values, except that control room specific X/Q non-conservative results for the control room.*          values were utilized [Addendum 4 Figures 1 and 2, and Addendum 6 Sections 8 and 9).
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, which is an accepted method of rates from external sources, should be structured to    providing suitably conservative estimates of the provide suitably conservative estimates of the          exposure to control room personnel.
exposure to control room personnel. n .
4.2.4 "Credit for engineered safety features that        For the LOCA, credit for engineered safety mitigate airborne radioactive material within the        feature (ESF) systems is taken for high efficiency control room may be assumed. Such features may          particulate air (HEPA) and charcoal filters in the include control room isolation or pressurization, or    control room emergency recirculation (CRER) intake or recirculation filtration .... In most designs, system, but not for the automatic initiation features control room isolation is actuated by engineered        [Addendum 4 Sections 1.0, 5.1, 7.0, 8.4, and 9.0).
safeguards feature (ESF) signals or radiation monitors (RMs). In some cases, the ESF signal is        For the MSLB, the design basis Case 1.takes no effective only for selected accidents, placing          credit for ESF systems such as filtration systems or reliance on the RMs for the remaining accidents.        system isolations. A second case is modeled for Several aspects of RMs can delay the control room        each of the two MSLB scenarios, but only to ensure isolation, including the delay for activity to build up  there were no dose outliers should a control room to concentrations equivalent to the alarm setpoint      isolation occur, due perhaps to a radiation monitor and the effects of different radionuclide accident      signal [Addendum 6 Sections 6.4 and 11).
isotopic mixes on monitor response.*
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 21of43 Reaulatorv Gulde 1.183 Guidance                        Degree of Confonnance 4.2.5 "Credit should generally not be taken for the    No credit was needed, nor taken, for use of use of personal protective equipment or                personal protective equipment or prophylactic drugs prophylactic drugs. Deviations may be considered      for personnel, to meet the control room dose on a case-by-case basis. n                            acceptance criterion.
4.2.6 "The dose receptor for these analyses is the    For the LOCA and MSLB, the NRC-identified hypothetical maximum exposed individual who is        occupancy factors and breathing rates were present in the control room for 100% of the time      assumed [Addendum 4 Section 6.6, and during the first 24 hours after the event, 60% of the  Addendum 6 Section 6.4).
time between 1 and 4 days, and 40% of the time from 4 days to 30 days. For the duration of the event, the breathing rate of this individual should be assumed to be 3.5x10""" cubic meters per second."
4.2. 7 "Control room doses should be calculated        RADTRAD only uses one set of dose conversion using dose conversion factors identified in            factors (DCFs) within an analysis, so the control Regulatory Position 4.1 above for use in offsite      rooms DCFs are the same as those described dose analyses. The DOE from photons may be            above for Regulatory Position 4.1.
corrected for the difference between finite cloud geometry in the control room and the semi-infinite    The recommended semi-infinite to finite cloud dose cloud assumption used in calculating the dose          conversion factor is built into the RADTRAD code.
conversion factors. The following expression may be used to correct the semi-infinite cloud dose, DOE.., , to a finite cloud dose, DDEfintte. where the control room is modeled as a hemisDhere ... "
4.3 Other Dose Consequences "The guidance provided in Regulatory Positions 4.1    Refer to the responses to Regulatory Position 1.3.1 and 4.2 should be used, as applicable, in re-          above for each of those items.
assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those In            Design envelope source terms are not being NUREG-0737. Design envelope source terms              changed by this application.
provided In NUREG-0737 should be updated for consistency with the AST. In general, radiation        Radiation exposure estimates to plant personnel for exposures to plant personnel identified In            many of the NUREG-0737 considerations are also Regulatory Position 1.3.1 should be expressed In      not being changed by this submittal.
terms of TEDE. Integrated radiation exposure of plant equipment should be determined using the        Equipment qualification requirements for plant guidance of Appendix I of this guide."                equipment as a result of the new calculations are discussed further for Reaulatory Position 6.
: 4. 4 Acceptance Criteria "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 10CFR 50.67 is used.
10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low        For the LOCA, and for the MSLB with a pre-existing probability of occurrence and low risk of public      4.0 &#xb5;Ci/gm dose equivalent 1-131 spike, the 25 rem exposure to radiation, e.g., a large-break LOCA.        TEDE offsite dose criterion from 10CFR 50.67 is The control room criterion applies to all accidents. used.
For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed          For the two CRDA analyses, the 6.3 rem TEDE the criteria tabulated in Table 6. The acceptance      offsite dose criterion from Table 6 of RG 1.183 is criteria for the various NUREG-0737 (Ref. 2) items      used (-25% of the 10 CFR 50.67 dose criterion).
generally reference General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or          For the MSLB with a 0.2 &#xb5;Ci/gm dose equivalent specify criteria derived from GDC 19. These criteria 1-131 equilibrium coolant activity, the 2.5 rem TEDE
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 22 of 43 Reaulatorv Gulde 1.183 Guidance                      Degree of Confonnance are generally specified in terms of whole body dose,  offsite dose criterion from RG 1.183 Table 6 is used or its equivalent to any body organ. For facilities  (-10% of the 10 CFR 50.67 criterion).
applying for, or having received, approval for the use of an AST, the applicable criteria should be    The USAR markups provided in Addendum 1 show updated for consistency with the TEDE criterion in    how the applicable criteria are being updated.
10 CFR 50.67(b)(2)(iii)."
Section 5. Analysis Assumptions and Methodology 5.1 General Considerations
: 5. 1. 1 Analysis Quality "The evaluations required by 10CFR 50.67 ...        The revised calculations were prepared, reviewed, should be prepared, reviewed, and maintained in      and will be maintained in accordance with a accordance with quality assurance programs that      10 CFR 50 Appendix B quality assurance program.
comply with Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.
These design basis analyses were structured to        The conservative, bounding characteristics of the provide a conservative set of assumptions to test    AST"included in Regulatory Guide 1.183 are used the performance of one or more aspects of the        In the calculations.
facility design. Many physical processes and phenomena are represented by conservative,
*bounding assumptions rather than being modeled directly. The staff has selected assumptions and models that provide an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensate for large uncertainties in facility parameters, accident progression, radioactive material transport, and
*atmospheric dispersion.
Licensees should exercise caution in proposing        There are no proposed deviations from the AST deviations based upon data from a specific accident  characteristics that are based on specific accident sequence since the DBAs were never intended to        sequences that would require additional justification represent any specific accident sequence - the        to prove they are conservative for other accident proposed deviation may not be conservative for        sequences.
other accident sequences."
5.1.2 Credit for Engineered Safeguard Features "Credit may be taken for accident mitigation          The plant systems that are credited to reduce doses features that are classified as safety-related, are  in the evaluations are safety-related, are required to required to be operable by technical specifications,  be Operable by Technical Specifications, are are powered by emergency power sources, and are      powered by emergency power sources, and are either automatically actuated or, in limited cases,  either automatically actuated or have actuation have actuation requirements explicitly addressed in  requirements explicitly addressed in emergency emergency operating procedures.                      operating procedures.
The single active component failure that results in  Single active component failures are assumed.
the most limiting radiological consequences should
* For the LOCA, it is a failure of all four main steam be assumed. Assumptions regarding the                    shutoff valves (MSSV, a third safety-related, occurrence and timing of a loss of offsite power          seismically qualified valve located downstream of should be selected with the objective of maximizing      the two main steam isolation valves (MSIVs) in the postulated radiological consequences. n              each of the four main steam lines) to close, due to being powered from a single electrical division
[Addendum 4 Section 5.9). Although these MSSVs are the assumed single failure, controls
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 23 of 43 Regulatory Guide 1.183 Guidance                              Degree of Conformance
* over their operability are required by TS 3.6.1.9, as implemented per Amendment 103.
* For the CRDA, the assumed failure is of one division of either the rod control and Information
* system (RC&IS) or one division of the average power range monitor (APRM) scram system.
Both systems are designed per the single failure criteria, such that their function is achieved due to operation of the other division during the event
[Addendum 1, USAR Section 15.4.9.2.3).
* For the MSLB, the assumed failure is of one
* MSIV, such that isolation of that steam line still occurs due to operation of the other MSIV
[Addendum 6 Section 8).
For the ESF ventilation filtration systems, in the CRDAand MSLB calculations, some sensitivity studies are run that consider the effects of unnecessary isolations of the normal control room ventilation inlet, but the control room base calculations, which produced acceptable results, assume the normal control room ventilation system continues to run throughout the events without filtration or isolation. This is done to simulate a large unfiltered inleakage value into the control room [Addendum 5 Sections 4.2 and 6, and Att. 1, and Addendum 6 Sections 6.4 and 111.
: 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 10CFR 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 cases in Addenda 5 and 6 assume another portion of the same analysis. For example,            the normal ventilation system continues to run, to assuming minimum containment system spray flow                maximize the intake of activity Into the control room is usually conservative for estimating iodine                (by maximizing the unfiltered lnleakage rate), and scrubbing, but in many cases may be                          ensure no credit is taken for active functions such nonconservative when determining sump pH.                    as Isolations from the radiation monitor or activation Sensitivity analyses may be needed to determine              ofthe emergency recirculation system. The the appropriate value to use. As a conservative              sensitivity. studies examined an alternative scenario alternative, the limiting value applicable to each            where the normal control room ventilation isolates portion of the analysis may be used in the evaluation        at time zero, and emergency recirculation is not of that portion. A single value may not be applicable        initiated until 30 minutes after the start of the event, for a parameter for the duration of the event,                to ensure that even such a long delay before particularly for parameters affected by changes in            manually initiating the emergency recirculation density. For parameters addressed by technical                system would not result in a dose outlier.
specifications, the value used In the analysis should be that specified in the technical specifications.... n
: 5. 1.4 Applicability of Prior Licensing Basis "The NRC staff considers the implementation of an            Several Items may be considered to be retained AST to be a significant change to the design basis of        items from the current licensing basis. Primarily the facility that is voluntarily initiated by the licensee. these are associated with retention of TIO-based In order to issue a license amendment authorizing            source term values where no benefit to safety or
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 24of43 Regulatory Gulde 1.183 Guidance                        Degree of Confonnance the use of an AST and the TEDE dose criteria, the      efficiency would result from adopting new AST-NRC staff must make a current finding of compliance    based numbers. These were retained in the with regulations applicable to the amendment. The      previous adoption of an AST in Amendment 103, characteristics of the ASTs and the revised dose        such as in the areas of equipment qualification, vital calculational methodology may be incompatible with      area access evaluations, post-accident sampling many of the analysis assumptions and methods            system access, control room dose due to shine, and currently reflected in the facility's design basis      containment purge isolation analyses.
analyses. The NRC staff may find that new or unreviewed issues are created by a particular site-    Another retained item results in higher calculated specific implementation of the AST, warranting          dose from a LOCA. The discussion for Regulatory review of staff positions approved subsequent to the    Position 5.2 below provides details related to initial issuance of the license. This is not considered retention of the assumption of additional ESF a backfit as defined by 10 CFR 50.109, "Backfitting.
* leakage from a failed system component outside of However, prior design bases that are unrelated to      the primary and secondary containments after the use of the AST, or are unaffected by the AST,      24 hours.
may continue as the facility's design basis.
Licensees should ensure that analysis assumptions and methods are compatible with the ASTs and the TEDE criteria."
: 5. 2 Accident-Specific Assumptions "The appendices to this regulatory guide provide        The RG 1.183 Appendices for the LOCA, MSLB, accident-specific assumptions that are acceptable      and CRDA are addressed below.
to the staff for performing analyses that are required by 10 CFR 50.67.... Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST. The NRC staff has determined that the analysis assumptions In the appendices to this guide provide an integrated approach to performing the individual analyses and generally expects licensees to address each assumption or propose acceptable alternatives.
Such alternatives may be justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some cases, a previously approved licensing basis consideration. The assumptions in the appendices are deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with each other. Although licensees are free to propose alternatives to these assumptions for consideration by the NRC staff, licensees should avoid use of previously approved staff positions that would adversely affect this consistency ....
* 5.3 Meteorology Assumptions "Atmospheric dispersion values (X/Q) for the EAB,        Current USAR X/Q values for the Control Room, the LPZ, and the control room that were approved        EAB and LPZ were approved in conjunction with by the staff during initial facility licensing or in    License Amendment 103 in March 1999 [NRC subsequent licensing proceedings may be used in          Safety Evaluation pages 14 and 15). These values performing the radiological analyses identified by      are presented in Addendum 4 Section 6.5, this guide. ... All changes in X/Q analysis              Addendum 5 Section 3.1.4, and Addendum 6 methodology should be reviewed by the NRC staff."        Section 6.2. No changes to XIQ atmospheric discersion values or methodoloav are procosed.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 25 of 43 Regulatoa Gulde 1.183 Guidance                        Degree of Conformance Section 6. Assumptions for Evaluating the Radiation Doses 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 (EQ) requirements at PNPP as a result radiological assessments associated with              of these re-analyses. The current degree of equipment qualification. The assumptions In          conformance to RG 1.89 is maintained with this Appendix I will supersede Regulatory Positions        application. This is based on reviews of EQ that 2.c(1) and 2.c(2) and Appendix D of Revision 1 of    have previously been performed as part of the prior Regulatory Gulde 1.89, "Environmental                adoption of an alternative AST for PNPP Qualification of Certain Electric Equipment          (Amendment 103). The reviews performed Important to Safety for Nuclear Power Plantsn (Ref. specifically for PNPP and those performed 11), for operating reactors that have amended their  generically by the NRC reached similar licensing basis to use an alternative source term. conclusions. Due to the improved knowledge about Except as stated in Appendix I, all other            the form that the released radionuclides take, the assumptions, methods, and provisions of Revision      same or less of the radionuclides remain airborne 1 of Regulatory Guide 1.89 remain effective. The      during the course of the event, because the NRC staff is assessing the effect of increased        particulates that make up more of the radionuclldes cesium releases on EQ doses to determine whether      tend to settle or be scrubbed out of the air. In licensee action is warranted. Until such time as this general, therefore, the dose rate values used in generic issue is resolved, licensees may use either  most areas inside and outside of containment that the AST or the TID14844 assumptions for              were determined using the TIO-based release performing the required EQ analyses. However, no      compositions tend to be the same or conservative plant modifications are required to address the      as compared to the actual compositions and impact of the difference In source term              airborne doses that are expected to result. The characteristics (i.e., AST vs TID14844) on EQ        remaining question when RG 1.183 was issued was doses pending the outcome of the evaluation of the    the significance of the doses to equipment exposed generic Issue. n                                      to sump or suppression pool water, because more particulates (particularly cesium) making their way into the suppression pool water. could lead to higher dose rates from that water. It was noted in RG 1.183 Regulatory Position 1.3.5 that the NRC staff "is assessing the effect of an Increased cesium release on EQ doses to determine whether licensee action is warranted. Until such time as this generic issue is resolved, licensees may use either the AST or the TIO 14844 assumptions fur performing the required EQ analyses."
Regarding this Issue, at PNPP, a conservative decision was made as the plantwas being built, during development of the profiles that equipment was qualified to, which mitigates* this concern. As noted in a PNPP letter dated January 18, 1999, the TIO-based ECCS equipment EQ used a 50 percent cesium source term, versus the 1 percent cesium source term that must be met to conform to RG 1.89. Even if this was not the* case for PNPP, the NRC has subsequently determined that the sump/suppression pool water long-term dose issue is not a significant concern. The candidate Generic Issue 187 The Potential Impact Of Postulated Cesium Concentration On Equipment Qualification In The Containment Sump has since been closed.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 26of43 Regulatoa Guide 1.183 Guidance                      Degree of Conformance As noted by the NRC in the Safety Evaluation for PNPP Amendment 103, the reviews done for Grand Gulf showed that the water pool doses did not become higher than the TIO-based doses until approximately 150 days after a LOCA. The April 30, 2001 final conclusion of the Generic Issue Review Panel noted that longer term equipment operability issues associated with severe fuel damage accidents (with which the AST is associated) could be addressed under accident management or plant recovery actions as necessary. A 150 day (five month) plant recovery period provides time to bring In significant external resources to supplement installed plant equipment.
For the above reasons, it is not necessary to revise the PNPP equipment qualification program to convert to alternative AST assumptions. Any changes to the EQ program or the environmental profiles in the PNPP USAR will be processed separately from this license amendment application, per 10 CFR 50.59.
Appendix A ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A LWR LOSS-OF-COOLANT ACCIDENT Source Term Assumptions App. A; 1. "Acceptable assumptions regarding core  Refer to above discussion for Regulatory Position 3.
Inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide.*
App. A; 2. "If the sump or suppression pool pH Is  As discussed in USAR Sections 3.11.5.1.2, 9;3.5.3, controlled at values of 7 or greater, the chemical  15.6.5.5.1.8, suppression pool pH is maintained form of radlolodlne released to the containment    at 7 or above to minimize conversion of cesium should be assumed to be 95% cesium iodide (Csl),    iodide to elemental iodine [see also Addendum 4 4.85% elemental Iodine, and 0.15% organic iodide. Section 6.3). To accomplish this, the standby liquid
... Evaluations of pH should consider the effect of control (SLC) system is used to provide a pH acids and bases created during the LOCA event,      buffering solution for the reactor vessel and e.g., radiolysis products. With the exception of    suppression pool water, to retain such fission elemental and organic iodine and noble gases,      products in the water. The SLC system is safety-fission products should be assumed to be in        related and Seismic Category I, and was NRC-particulate form. n                                approved for this pH buffering function as part of Amendment 103. As acidic products are added post-LOCA, only a small decrease in pH occurs provided the buffer capacity is not exceeded.
The pH calculation completed ih 2013 in support of this application determined that with a full SLC injection within 12 hours post-LOCA, re-evolution of Iodine from the water during the 30-day post-LOCA period is negligible [note: the graph and table
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 27 of 43 Regulaio~  Gulde 1.183 Guidance                Degree of Confonnance included below (copied from the supporting calculation) illustrate pool pH with SLC injection at one hour post-LOCA. However, it was also determined that the suppression pool pH curve remains acceptable (above 7) during the 30-day post-LOCA period even if SLC injection does not occur until 12 hours post-LOCA).
As summarized in Section 6.3 of Addendum 4, the chemical form of radioiodine can therefore be assumed to be 95% cesium iodide (Csl), 4.85%
elemental iodine, and 0.15% organic iodide.
Sappmsloa .Pool pH darill1 Ille >>day l'Dsl*IA>CA Period wltb SLC lajedloll at I llaar IO    -------------    r'                                  -......
                                                      .. ~f-------** . -*--*                                            \
i*.o
                                                      ..,          I IO
: u. -* ------..*- - - -*. . .                      **---***----
                                                      ...001            DI            I ID        ...        -
Table s.1: SemldvhJ Catcalatlom (.tnactmwrt 3) pH (wltb SLC IDjecdoD al I llolll')
ame      a-llellllll            M211V,...        M lllO"P 0          6.0                  6.0            6.D 2min              6.0                  6.G            6.G 5min              7.4                  7.0            8.1 IOmln              7.9                  7.6            8.6 20mln              8.2                  7.9            9.0 32 min              8.3                  8.0            9.2 45min              8.4                  8.1            9A 60min              8.5                  8.2            9.5 90mln              u                    u                8.7 120mln              8.5                  u                8.7 122mln              8.5                  8.5              8.7 168mln              u                    u                8.6 5br            8.5                  8.5            8.6 12br            8.5                  8.5            8.6 I day            8.5                  8.5            8.6
                                                            ]days            8.4                  8.4            8.6 IOdays              8.3                  8.3            8.4
                                                        *20days                8.1                  8.1            8.2 JOdaya              7.6                  7.6            7.7
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 28of43 Regulatory Gulde 1.183 Guidance                    Degree of Conformance ASSUMPTIONS ON TRANSPORT IN PRIMARY CONTAINMENT App. A; 3. "Acceptable assumptions related to the transport, reduction, and release of radioactive material in and from the primary containment in PWRs or the drywell in BWRs are as follows:n App. A; 3.1 "The radioactivity released from the    Both the gap release that begins at two minutes fuel should be assumed to mix instantaneously and  post-accident and the in-vessel release that begins homogeneously throughout the free air volume of    at 30 minutes are released into the drywall at a the primary containment in PWRs or the drywell in  constant rate over their release periods BWRs as it is released. This distribution should be [Addendum 4 Section 7.0). In RADTRAD, the adjusted if there are internal compartments that    drywell is treated as a single volume, and the code have limited ventilation exchange. The suppression  assumes uniform concentration in each volume.
pool free air volume may be included provided there is a mechanism to ensure mixing between the drywell to the wetwell.
The release into the containment or drywen*should  The source term release from the vessel is be assumed to terminate at the end of the early in- terminated at two hours with the actuation of ECCS vessel phase.*                                      [Addendum 4 Section 7.0).
App. A; 3.2 "Reduction in airborne radioactivity in Due to the use of RADTRAD, the deposition model the containment by natural deposition within the    from NUREG/CR-6189 is utilized [Addendum 4 containment may be credited. Acceptable models      Section 8.3.3).
for removal of iodine and aerosols are described in
... NUREG/CR~189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments" (Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3) . ... :
App. A; 3.3 "Reduction in airborne radiQactivity hi The Powers model for aerosol removal by sprays the containment by containment spray systems that  (from NUREG/CR-5966) that is built into the have been designed and are maintained in            RADTRAD code is used in the analysis accordance with Chapter 6.5.2 of the SRP            [Addendum 4 Section 8.3.3).
(Ref. A-1) may be credited. Acceptable models for the removal of iodine and aerosols are described in Chapter 6.5.2 of the SRP and NUREG/CR-5966, "A Simplified Model of Aerosol Removal by Containment Sprays" (Ref. A-4). This simplified model is incorporated Into the analysis code RADTRAD (Refs. A-1 to A-3).
The evaluation of the containment sprays should    The mixing rate between the unsprayed address areas within the primary containment that  containment and the sprayed containment is are not covered by the spray drops. The mixing      assumed to be 71,400 cubic feet per minute (cfm),
rate attributed to natural convection between      as previously approved by the NRC as part of sprayed and unsprayed regions of the containment    Amendment 103.
building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified ....
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 29of43 Regulatory Gulde 1.183 Guidance                          Degree of Conformance The SRP sets forth a maximum decontamination              These limits on decontamination were applied, factor (OF) for elemental Iodine based on the            specifically the limit on elemental iodine removal by maximum iodine activity in the primary containment        sprays is capped at a OF of 200, and the removal of atmosphere when the sprays actuate, divided by            particulate iodine (aerosols) is reduced by a factor the activity of iodine remaining at some time after      of ten once a OF of 50 is reached, with no cap decantamination. The SRP also states that the            [Addendum 4 Sections 6.13.2 and 8.3.3).
particulate iodine removal rate should be reduced by a factor of 1O when a OF of 50 is reached. .. .        To conservatively calculate the OF for both There is no specified maximum OF for aerosol              elemental iodine and particulate iodine so the removal by sprays. The maximum activity to be            above limits on OF can be applied, the maximum used In determining the OF Is defined as the iodine      activity in the containment at the end of two hours activity in the columns labeled "Total" in Tables 1      was used rather than the amount of activity at the and 2 of this guide multiplied by 0.05 for elemental      time when the containment sprays are actuated (at iodine and by 0.95 for particulate iodine (i.e.,          either 10 or 30 minutes (automatically or manually, aerosol treated as particulate in SRP                    respectively)). This assumption was made because methodology)."                                            at the 30 minute point, the in-vessel release period is just beginning, per Table. 4 of RG 1.183, whereas after two hours, substantially more radionuclides have been released (at a constant rate over the ninety minute in-vessel release period) into the drywall (not a sprayed region). This much larger radionuclide inventory is then mixed with the containment volume by the flushing of the drywall at the two hour point.
App. A; 3.4 "Reduction in airborne radioactivity in      The PNPP design does not include in-containment the containment by In-containment recirculation          recirculation filter systems.
filter systems may be credited if... "
App. A; 3.5 "Reduction In airborne radioactivity in      Even though the evaluation of ECCS leakage the containment by suppression pool scrubbing in          outside containment assumes that all the releases BWRs should generally not be credited. However,          (except noble gases) are dissolved in the the staff may consider such reduction on an              suppression pool; the evaluations of airborne individual case basis ... n                              radioactivity In the containment conservatively does not credit any scrubbing by the suppression pool
[Addendum 4 Sections 5.4 and 8.3.2).
App. A; 3.6 "Reduction in airborne radioactivity in      Only natural deposition and containment sprays are the containment by retention In ice condensers, or        used to reduce airborne radioactivity in the other engineering safety features not addressed          containment [Addendum 4 Section 8.3.3).
above, should be evaluated on an individual case basis....*
App. A; 3. 7 "The primary containment ... should be      TS 5.5.12 explains that the. peak calculated primary assumed to leak at the peak pressure technical            containment pressure (Pa) at PNPP is only specification leak rate for the first 24 houl'S. .. . For 6.4 pounds per square Inch-gauge (psig), but that BWRs, leakage may be reduced after the first              for conservatism, the design basis value for Pa is 24 hours, if supported by plant configuration and        defined as 7.8 pslg. The 7.8 psig value was the analyses, to a value not less than 50% of the            previous value of Pa prior to evaluations performed technical specification leak rate ....                    for power uprate, and the 7.8 psig Pa value was retained for leak rate test program consistency when the lower value of 6.4 psig was calculated.
In the new LOCA dose calculations, the primary containment is assumed to leak at the peak
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 30of43 Regulato~      Gulde 1.183 Guidance                  Degree of Confo[!!!ance pressure technical specification leak rate (La=0.20%
of primary containment air weight per day at the peak containment pressure (Pa)), for the first 24 hours [Addendum 4 Section 5. 7). After 24 hours, the leakage rate was reduced, but not by the maximum 50 percent allowance in RG 1.183.
It was only reduced to 0.69La (in other words, a 31 percent reduction) [Addendum 4 Section 6.11].
The plant configurations and analysis that support this reduction are reasonable, since the 6.4 psig peak pressure value is already nearly 20 percent below the conservatively assumed value of 7.8 psig, and the use of containment sprays for dose reduction would also greatly reduce containment pressures to well below the assumed values. When calculating post-accident pressures to determine P8 , the pressure reduction effect of .
containment sprays is not credited, since the goal of such a calculation is to maximize post-accident pressures. A discussion of the effect of sprays to reduce containment pressures, and therefore their effectiveness to reduce the associated leak rates at those lower pressures, is included in Addendum 4 Section 5.7, and in the PNPP USAR beginning on page 6.5-12.
For BWRs with Mark Ill containments, the leakage      Leakage from the drywall into the primary from the drywall into the primary containment          containment during the two-hour period between should be based on the steaming rate of the:heated    the initial blowdown and termination of the fuel reactor core, with no credit for core debris          radioactivity release is based on the steaming rate relocation. This leakage should be assumed during      of the heated reactor core; as was used in the the two-hour period between the initial blowdown      previously NRC-approved LOCA calculation from and termination of the fuel radioactivity release (gap Amendment 103 [Addendum 4 Sections 5.2, 5.4, and early in-vessel release phases).                  and 8.2.3).
After two hours, the radioactivity is assumed to be    The analysis assumes there is a homogenous uniformly distributed throughout the drywall and the  mixture In the drywall and containment starting at primary containment. n                                two hours [Addendum 4 Sections 5.2 and 8.3.2).
App. A; 3.8 "If the primary containment Is routinely  The primary containment Is not routinely purged purged during power operations, releases via the      during power operations. Per TS Surveillance purge system prior to containment isolation should    Requirement (SR) 3.6.1.3.2, the purge valves are be analyzed and the resulting doses summed with        required to be closed except when opened the postulated doses from other release paths. The    "for pressure control, ALARA or air quality purge release evaluation should assume that 100%      considerations for personnel entry, or Surveillances of the radionuclide inventory In the reactor coolant  or special testing on the purge system that require system liquid Is released to the containment at the    the valves to be open. n Initiation of the LOCA. This inventory should be based on the technical specification reactor coolant  A discussion is included (and retained) In system equilibrium activity. Iodine spikes need not    USAR Section 6.2.4.2.3 "Consideration of NRC be considered. If the purge system is not isolated    Branch Technical Position CSB 6-4, 'Containment before the onset of the gap release phase, the        Purging during Normal Plant Operations'." As release fractions associated with the gap release      noted therein, the purge system containment and early in-vessel phases should be considered as    isolation valves are capable of isolating
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 31of43 Regylato~      Gulde 1.183 Guidance                    Degree of Conformance applicable."                                          containment within five seconds, which is much shorter than the two minute period before onset of a BWR gap release. The containment purge exhaust plenum radiation-high instrumentation to isolate these valves is controlled by TS 3.3.6.1, Table 3.3.6.1-1 Function 2.g. The dose calculation discussed in USAR Section 6.2.4.2.3 was performed using TIO 14844 assumptions, was not updated as part of the Amendment 103 adoption of an alternative AST, and is again retained as part of this application. The reported site boundary doses are 0.9 rem to the thyroid and 162 millirem (mrem) whole body. Use of the RG 1.183 Footnote 7 methodology to convert these values to a TEDE dose ((0.9*0.03) +0.162) results in only 0.19 rem TEDE.
ASSUMPTIONS ON DUAL CONTAINMENTS App. A; 4. "For facilities with dual containment systems, the acceptable assumptions related to the transport, reduction, and release of radioactive material in and from the secondary containment or enclosure buildings are as follows."
App. A; 4.1 "Leakage from the primary containment      Primary containment leakage (except a fraction that should be considered to be collected, processed by    bypasses the secondary containments) is collected engineered safety feature (ESF) filters, if any, and  and processed (HEPA only, no credit for charcoal) released to the environment via the secondary          by the annulus exhaust gas treatment (AEGT) containment exhaust system during periods in          system, an ESF system. During normal operation, which the secondary containment has a negative        the shield building annulus is maintained at a slight pressure as defined in technical specifications.      negative pressure by operation of an AEGT subsystem. Following a OBA, the annulus is expected to remain negative; however, for a short period the analysis assumes it may not be maintained below the design negative pressure value of 0.25 inches water gauge (USAR 6.5.3.2.1 ).
Therefore, it was assumed that primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA (Addendum 4 Sections 5.5, 6.10, 7.0, and 8.3.2).
Credit for an elevated release should be assumed        No credit is assumed for an elevated release.
only If the point of physical release is more than two Releases are assumed to be ground level releases and one-half times the height of any adjacent          (Addendum 4 Section 6.5).
structure. n App. A; 4.2 "Leakage from the primary containment      Refer to the above two discussions.
Is assumed to be released directly to the environment as a ground-level release during any period in which the secondary containment does not have a negative pressure as defined in technical* specifications. n App. A; 4.3 "The effect of high wind speeds on the      The ability of the secondary containment (the abilitv of the secondary containment to maintain a      annulus at PNPP) to maintain a negative pressure
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 32 of 43 Regulatory Gulde 1.183 Guidance                  Dearee of Confonnance negative pressure should be evaluated on an      Is maintained even with the effects of wind on the individual case basis. The wind speed to be      buildings. The annulus negative pressure has assumed is the 1-hour average value .that is      several components. First, is a base margin of exceeded only 5% of the total number of hours in  0.25 inches water gauge (in. w.g.). The discussion the data set. Ambient temperatures used In these  for this component of the negative pressure margin assessments should be the 1-hour average value    Is provided at the end of this paragraph. Second, at that is exceeded only 5% or 95% of the total      PNPP, the AEGT system is normally run during numbers of hours In the data set, whichever is    plant operation (Modes 1, 2, and 3). This starts the conservative for the intended use .. ."          secondary containment (annulus) at a negative pressure before the initiation of a postulated LOCA.
To attain a goal of keeping the annulus at a negative pressure throughout an event, even during diesel generator start and loading periods, the required annulus pressure was increased to 0.40 in. w.g. This additional margin exists even though the dose calculations now assume that the annulus does not stay negative for the first 40 seconds of the LOCA, so this component of the margin provides conservatism. Third, the required negative pressure was increased in Amendment 66 to 0.66 in. w.g. (see current TS SR 3.6.4.1.1) to account for temperature differences as identified in NRC Information Notice 88-76. Branch Technical Position CSB 6-3 to section 6.2.3 of the SRP states that an outward positive differential on the secondary containment wall can be created by wind loads. In this regard, a "positive" pressure is defined as any pressure greater than -0.25 in. w.g.,
to account for wind loads and the uncertainty In the pressure measurements, and the Branch Technieal Position notes that whenever pressure in the secondary containment volume exceeds -0.25 in.
w.g., the leakage-prevention function of the secondary containment is assumed to be negated.
Therefore, wind loads are an integral part of the base -0.25 In. w.g. margin for the AEGT system (the first component of the annulus negative pressure mentioned above).
App. A; 4.4 "Credit for dilution in the secondary The AEGT system extracts and filters a maximum containment may be allowed when adequate          of 2000 cfm from the annulus. During an accident, means to cause mixing can be demonstrated.        the expected discharge to the atmosphere is Otherwise, the leakage from the primary            1000 cfm. The balance of the filtered AEGTS flow eontainment should be assumed to be transported  is routed back to the annulus. However, the directly to exhaust systems without mixing .... n analysis conservatively assumes that 2000 cfm is discharged directly to the environment with no recirculation (holdup) of iodine in the annulus
[Addendum 4 Section 5.6].
App. A; 4.5 "Primary containment leakage that      Secondary containment bypass leakage is currently bypasses the secondary containment should be      limited to 5.04% of La (0.0504La) when pressurized evaluated at the bypass leak rate incorporated In  to -Pa by TS SR 3.6.1.3.9, even though the the technical specifications.... n                previously-NRC-approved LOCA dose calculation assumed bypass of 10.08% of La (0.1008La). The
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 33 of 43 Reaulatorv Gulde 1.183 Guidance                      Degree of Conformance secondary containment bypass leakage assumption is maintained at 0.1008L8 in this analysis to allow for a future increase In the TS allowable leakage limit [Addendum 4 Section 5.10).
App. A; 4.6 "Reduction in the amount of radioactive  The AEGTS Includes HEPA filters, which are material released from the secondary containment    periodically tested to demonstrate compliance with because of ESF filter systems may be taken into      Regulatory Guide 1.52. Particulate removal by the account provided that these systems meet the        HEPA filters in the AEGT system is assumed to be guidance of Regulatory Gulde 1.52 (Ref. A-5) and    99% in accordance with RG 1.52. The system also Generic Letter 99-02 (Ref. A-6)."                    contains 4-inch deep activated charcoal adsorbers to remove elemental and organic iodine; however, the analysis conservatively assumes a removal efficiency of 0% for the charcoal adsorbers to allow operational flexibility (after a future TS change)
[Addendum 4 Sections 5.6, 6.14, and 8.3.2).
ASSUMPTIONS ON ESF SYSTEM LEAKAGE App. A; 5. "ESF systems that recirculate sump water outside of the primary containment are assumed to leak during their intended operation ....
The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of leakage from ESF components outside the primary containment for BWRs and PWRs."
App. A; 5.1 "With the exception of noble gases, all  For the purpose of determining dose due to leakage the fission products released from the fuel to the  from the portions of the ECCS systems that are containment (as defined in Tables 1 and 2 of this    routed outside of the primary and secondary guide) should be assumed to instantaneously and      containments at PNPP, the gap and core activity homogeneously mix In the ... suppression pool (In    released to the drywell ls assumed to be BWRs) at the time of release from the core .... ."  immediately dissolved in the suppression pool (even though other calculations take no credit for suppression pool scrubbing). Only halogens are modeled in this analysis, since noble gases are not soluble [Addendum 4 Section 8.1.1 ].
App. A; 5.2 "The leakage should be taken as two      As discussed In more detail above in Section 1.3.1 times the sum of the simultaneous leakage from all  regarding NUREG-0737 Item 111.D.1.1, PNPP USAR components In the ESF recirculation systems above    Section 15.6.5.5.1.7 commits to administrative which the technical specifications, or licensee      controls that limit ESF leakage to less than half of commitments to item 111.D.1.1 of NUREG-0737          the 15 gallon per hour (gph) value assumed In the (Ref. A-8), would require declaring such systems    radiological dose calculations [also Addendum 4 .
inoperable.                                          Section 8.1.3).
The leakage should be assumed to start at the        For PNPP, ECCS leakage is assumed to begin at earliest time the recirculation flow occurs in these the onset of gap release at two minutes and systems and end at the latest time the releases      continue for the duration of the event. This is a from these svstems are terminated. Consideration    conservative assumption that maximizes the dose
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 34of43 Regulatorv Gulde 1.183 Guidance                    Degree of Conformance should also be given to design leakage through      contribution for this release pathway [Addendum 4 valves isolating ESF recirculation systems from    Sections 5.8, 7.0, 8.1.3).
tanks vented to atmosphere, e.g., emergency core cooling system (ECCS) pump min-flow return to the  In addition to the ESF leakage guidance in refueling water storage tank.*                      RG 1.183, above, an additional source of dose from ESF leakage is included in the PNPP radiological dose calculation, due to previously existing guidance in NUREG-0800 Standard Review Plan (SRP) Section 15.6.5, Appendix B, specifically, leakage from an ECCS system component outside of the primary and secondary containments is assumed to occur at a rate of 50 gallons per minute (gpm) starting 24 hours into the accident and lasting for 30 minutes. Although this additional dose contributor is retained in the calculations, it is not part of the guidance in RG 1.183, so inclusion of this extra leakage into plant buildings with non-ESF ventilation systems provides a substantial conservatism in the PNPP dose calculation.
App. A; 5.3 "With the exception of iodine, all      With the exception of iodine, all radioactive radioactive materials in the recirculating liquid  materials in the recirculating liquid are assumed to should be assumed to be retained in the liquid      be retained in the liquid phase [Addendum 4 phase."                                            Section 8.1.1 ).
App. A; 5.4 "If the temperature of the leakage      Refer to Appendix A Position 5.5, because the exceeds 212&deg;F, the fraction of total iodine in the  suppression pool temperature will not exceed 212&deg;F liquid that becomes airborne should be assumed      [Addendum 4 Section 8.1.4).
equal to ... n App. A; 5.5 "If the temperature of the leakage is  The temperature of the recirculating liquid will be less than 212&deg;F or the calculated flash fraction is less than 212&deg;F, as will the leakage, so 10% of the less than 10%, the amount of Iodine that becomes    iodine in the ECCS leakage is assumed to become airborne should be assumed to be 10% of the total  airborne [Addendum 4 Section 8.1.4).
iodine activity in the leaked fluid .. ."
App. A; 5.6 "The radioiodine that is postulated to  The chemical species of the airborne source terms be available for release to the environment is      is assumed to be 97% elemental and 3% organic assumed to be 97% elemental and 3% organic.        [Addendum 4 Section 8.1.4).
Reduction in release activity by dilution or holdup Natural removal mechanisms and holdup in the within buildings, or by ESF ventilation filtration  auxiliary building are conservatively neglected systems, may be credited where applicable.... n    [Addendum 4 Sections 8.1.2 and 8.1.4). The ECCS leakage is assumed to be released directly to the environment [Addendum 1Section15.6.5.5.1.7 and Addendum 4 Section 7.0).
ASSUMPTIONS ON MAIN STEAM /SOLA TION VALVE LEAKAGE IN BWRS App. A; 6. "For BWRs, the main steam isolation valves (MSIVs) have design leakage that may result In a radioactivity release. The radiological consequences from postulated MSIV leakage should be analyzed and combined with consequences postulated for other fission oroduct
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 35 of 43 Rm1ulato~ Gulde 1.183 Guidance                        Degree of Conformance release paths to determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of MSIV leakage."
App. A; 6.1 "For the purpose of this analysis, the    The flow through the main steam lines is from the
*activity available for release via MSIV leakage        drywall atmosphere [Addendum 4 Sections 6.12, should be assumed to be that activity determined to    8.2.1, and 8.2.3].
be in the drywall for evaluating containment leakage (see Regulatory Position 3). No credit should be assumed for activity reduction by the steam separators or by iodine partitioning in the reactor vessel."
App. A; 6.2 "All the MSIVs should be assumed to        A total main steam line leak rate of 250 standard leak at the maximum leak rate above which the          cubic feet per hour (scfh) is assumed to occur, technical specifications would require declaring the  which is the maximum allowable per Technical MS IVs inoperable."                                    Specification SR 3.6.1.3.10, consisting of:
(1) 100 scfh through the steam line that is postulated to have broken (the initiating event),
(2) 100 scfh through a second (intact) steam line, and (3) the remaining 50 scfh through a third intact steam line.
This is modeled as 100 scfh through the broken steam line and 150 scfh through unbroken steam lines [Addendum 4 Sections 5.9, 6.12, 8.2.2, and 8.2.3].
The leakage should be assumed to continue for the      Releases to the environment via MSIV leakage duration of the accident.                              continue for 30 days [Addendum 4 Section 7.0].
Postulated leakage may be reduced after the first      For the same reasons as detailed above for 24 hours, if supported by site-specific analyses, to a Appendix A Item 3. 7, the postulated leakage value not less than 50% of the maximum leak rate."    through the main steam lines was reduced to 69 percent of the initial values after 24 hours.
App. A; 6.3 "Reduction of the amount of released      The deposition In the main steam lines uses the radioactivity by deposition and plateout on steam      previously NRC-approved aerosol removal system piping upstream of the outboard MSIVs may      efficiencies from the Amendment *103 review be credited, but the amount of reduction in            [Addendum 4 Section 6.15]. These removal concentration allowed will be evaluated on an          efficiencies include a 1Opercent Increase in aerosol individual case basis. Generally, the model should    penetration to add conservatism to the main steam be based on the assumption of well-mixed volumes,      line leakage pathway, as agreed to In a letter dated but other models such as slug flow may be used if      January 18, 1999 (PY-CEl/NRR-2359L).
justified. n App. A; 6.4 "In the absence of collection and          All MSIV leakage past the outboard MSIV is treatment of releases by ESFs such as the MSIV        assumed to be released directly to the environment leakage control system, or as described In            [Addendum 4 Section 8.2.4].
paragraph 6.5 below, the MSIV leakage should be assumed to be released to the environment as an        All releases In the LOCA calculation are ground unprocessed, ground-level release.                    level releases [Addendum 4 Section 6.5].
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 36of43 Regulatorv Gulde 1.183 Guidance                      Degree of Confonnance Holdup and dilution in the turbine building should  No credit is taken for holdup in the auxiliary building not be assumed.*                                    or turbine building [Addendum 4 Section 8.2.4).
App. A; 6.5 "A reduction in MSIV releases that is    The PNPP LOCA dose calculations do not credit due to holdup and deposition in main steam piping    holdup and deposition In the main condenser, as downstream of the MSIVs and in the main              this was not necessary due to the safety-related, condenser, including the treatment of air ejector    seismic Category I piping downstream of the effluent by offgas systems, may be credited if the  outboard MSIVs out to the third isolation valve in components and piping systems used in the release    each main steam line (the MSSVs), as discussed
.Path are capable of performing their safety function above for Regulatory Position 5.1.2. This piping during and following a safe shutdown earthquake      segment is credited in the analysis to support the (SSE). The amount.of reduction allowed will be      single failure assumption, wtiich is that the MSSVs evaluated on an individual case basis.... "          all fall to close as a result of a loss of their divisional power source. If these valves close during a postulated LOCA, an additional length of main steam piping would be available for deposition in each line, which would reduce doses even further.
However, no credit is taken in the calculation summarized in Addendum 4 for such deposition, due to the above described assumed single-failure.
Technical Specification 3.6.1.9 controls are maintained over the MSSVs since they are, in essence, credited, and therefore meet Criterion 3 of 10 CFR 50.36(c)(2)(1i).
ASSUMPTION ON CONTAINMENT PURGING App. A; 7. "The radiological consequences from      PNPP combustible gas and pressure control post-LOCA primary containment purging as a          measures do not require post-LOCA primary combustible gas or pressure control measure          containment purging, and such purging is not should be analyzed. If the installed containment    credited in any design basis analysis, so purging capabilities are maintained for purposes of  radiological consequences of such purging did not severe accident management and are not credited      need to be calculated.
in any design basis analysis, radiological consequences need not be evaluated. If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOCA. Reduction in the amount of radioactive material released via ESF filter systems may be taken into account provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref.
A-5) and Generic Letter 99-02 CRef. A-6).
* Append/xC ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR ROD DROP ACCIDENT App. C; 1. "Assumptions acceptable to the NRC        The source terms applied in this analysis are staff regarding core inventory are provided in        conservatively based on rod clad failures of 16 fuel Regulatory Position 3 of this guide. For the rod      bundles (1376 fuel rods) [Addendum 5 Sections drop accident, the release from the breached fuel is  3.1.5 and 3.2.3). The gap release fraction Is based on the estimate of the number of fuel rods      1O oercent for the noble aases and haloaens
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 37of43 Regulatoot Gulde 1.183 Guidance                        Degree of Conformance breached and the assumption that 10% of the core      [Addendum 5 Table 3-3).
inventory of the noble gases and iodines is in the fuel gap.
The release attributed to fuel melting is based on    The maximum mass fraction in the damaged fuel the fraction of the fuel that reaches or exceeds the  that reaches temperatures in excess of the melting initiation temperature for fuel melting and on the    point is 0.0077 [Addendum 5 Section 3.2.3). The assumption that 100% of the noble gases and 50%        calculation assumes the release of 50 percent of of the iodines contained in that fraction are released the iodine and 100 percent of the noble gases for to the reactor coolant. n                              fuel postulated to reach melt conditions [Addendum 5 Section 3.1.5).
App. C; 2. "If no or minimal fuel damage is            Since fuel damage is postulated for this event, the postulated for the limiting event, the released        impact of coolant source terms is neglected activity should be the maximum coolant activity        [Addendum 5 Section 3.1.5).
(typically 4 &#xb5;Ci/gm DE 1-131) allowed by the technical specifications."
App. C; 3. "The assumptions acceptable to the NRC staff that are related to the transport, reduction, and release of radioactive material from the fuel and the reactor coolant are as follows."
App. C; 3.1 "The activity released from the fuel      The activity released from the fuel from the gap and from either the gap or from fuel pellets is assumed    fuel pellets is assumed to be instantaneously mixed to be instantaneously mixed in the reactor coolant    in the reactor coolant within the pressure vessel within the pressure vessel. a                          [Addendum 5 Section 3.1.5).
App. C; 3.2 "Credit should not be assumed for          No credit Is taken for partitioning in the pressure partitioning in the pressure vessel or for removal by  vessel or for removal by the steam separators the steam separators."                                [Addendum 5 Section 3.1.5).
App. C; 3.3 aof the activity released from the        Of the activity released from the reactor coolant reactor coolant within the pressure vessel, 100% of    within the pressure vessel, 100 percent of the noble the noble gases, 10% of the iodine, and 1% of the      gases, 1O percent of the iodine, and 1 percent of remaining radionuclides are assumed to reach the*      the remaining radlonuclldes are assumed to reach turbine and condensers. a                              the turbine and condensers [Addendum 5 Section 3.1.5].
App. C; 3.4 "Of the activity that reaches the turbine  Of the activity that reaches the turbine and and condenser, 100% of the noble gases, 10% of        condenser, 100 percent of the noble gases,
* the iodine, and 1% of the particulate radionuclides    10 percent of the iodine, and 1 percent of the are available for release to the environment.          particulate radionuclldes are available for release to the environment [Addendum 5 Section 3.1.5).
The turbine and condensers leak to the atmosphere      The turbine and condensers leak to the atmosphere as a ground-level release at a rate of 1% per day      as a ground-level release at a rate of 1 percent per for a period of 24 hours, at which time the leakage    day for a period of 24 hours, at which time the is assumed to terminate.                                leakage is assumed to terminate [Addendum 5 Section 3.1.5].
No credit should be assumed for dilution or holdup      No credit is taken for dilution or holdup within the within the turbine building. Radioactive decay        turbine building. Radioactive decay during holdup during holdup in the turbine and condenser may be      in the turbine and condenser is credited
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 38of43 Regulatory Gulde 1.183 Guidance                      Degree of Confonnance assumed.*                                            [Addendum 5 Section 3.1.5].
App. C; 3.5 "In lieu of the transport assumptions    The transport assumptions provided for Regulatory provided in paragraphs 3.2 through 3.4 above, a      Positions 3.2 through 3.4 were used, so this item is more mechanistic analysis may be used on a case-    not applicable.
by-case basis.... n App. C; 3.6 "The iodine species released from the    Although it is not directly stated in Addendum 5, the reactor coolant within the pressure vessel should be Iodine species released from the reactor coolant assumed to be 95% Csl as an aerosol, 4.85%          within the pressure vessel is assumed in the CRDA elemental, and 0.15% organic.                        calculation to be 95 percent Csl as an aerosol, 4.85 percent elemental, and 0.15 percent organic.
The release from the turbine and condenser should    The iodine release from the turbine and condenser be assumed to be 97% elemental and 3% organic."      is assumed to be 97 percent elemental and 3 percent organic [Addendum 5 Section 3.1.5].
AppendixD ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR MAIN STEAM LINE BREAK ACCIDENT SOURCE TERM App. D; 1. *Assumptions acceptable to the NRC        Refer to App. D Item 2 below, since there is no staff regarding core inventory and the release of    breached fuel for this event.
radionuclides from the fuel are provided in Regulatory Position 3 of this guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. a App. D; 2. "If no or minimal fuel damage is          There is no fuel damage as a consequence of this postulated for the limiting event, the released      accident [USAR Section 15.6.4.3.2), so this item activity should be the maximum coolant activity      applies In lieu of Item 1 above. The iodine allowed by technical specification. The Iodine      concentration in the primary coolant Is assumed to concentration in the primary coolant is assumed to  correspond to the following two scenarios, which correspond to the following two cases in the nuclear correspond to the values specified in PNPP steam supply system vendor's standard technical      Technical Specification 3.4.8, "RCS Specific specifications.
* Activity:"
Scenario 1: the isotopic concentrations are those App. D; 2.1 "The concentration that is the that produce 0.2 &#xb5;Ci/gm dose equivalent 1-131, and maximum value (typically 4.0 &#xb5;Ci/gm DE 1-131) permitted and corresponds to the conditions of an    Scenario 2: the isotopic concentrations use the assumed pre-accident spike, and                      maximum iodine concentration allowed by the technical specifications for short periods, App. D; 2.1 [sic] The concentration that is the      specifically 4.0 &#xb5;Ci/gm dose equivalent 1-131 maximum equilibrium value (typically 0.2 &#xb5;Ci/gm      [Addendum 6 Section 6.1].
DE 1-131) permitted for continued full power operation. "
App. D; 3. "The activity released from the fuel      There Is no fuel damage as a consequence of this should be assumed to mix instantaneously and        accident [USAR Section 15.6.4.3.2).
* homogeneously in the reactor coolant. Noble
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 39of43 Regulatoot Gulde 1.183 Guidance                        Degcee of Confo!!!!ance gases should be assumed to enter the steam phase instantaneously. n TRANSPORT App. D; 4. "Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows."
App. D; 4.1 "The main steam line isolation valves      The main steam lines are assumed to close in the (MSIV) should be assumed to close in the              maximum time allowed by the PNPP technical maximum time allowed by technical specifications. n    specifications (i.e., S5.0 seconds) plus margin for actuation signal delay (1.05 seconds). The total time conservatively assumed in this analysis is 6.05 seconds [Addendum 1 in USAR Section 15.6.4.4; Addendum 6, Section 4, subsection 4.1 (rounded to 6 seconds in this discussion); and Addendum 6, Section 5, subsection 1).
App. D; 4.2 "The total mass of coolant released        The analysis of the MSLB considers both steam should be assumed to be that amount in the steam      and liquid flows to determine the total mass line and connecting lines at the time of the break    released prior to MSIV closure, including that plus the amount that passes through the valves        released during MSIV closure [Addendum 1 in prior to closure."                                    USAR Section 15.6.4.4; Addendum 6 Section 4, subsection 4.2; and Addendum 6 Section 5, subsection 1, which includes an illustration of the conservatism in the assumed mass released].
App. D; 4.3 "All the radioactivity in the released    The main steam release is assumed to be an coolant should be assumed to be released to the        instantaneous ground level release without any atmosphere Instantaneously as a ground-level          retention within the auxiliary building steam tunnel release. No credit should be assumed for plateout,    [Addendum 6 Section 4, subsection 4.3 and holdup, or dilution within facility buildings."        Section 5, subsection 2).
App. D; 4.4 "The iodine species released from the      The analysis uses this iodine species distribution main steam line should be assumed to be 95%.Csl        [Addendum 6 Section 4, subsection 4.4).
as an aerosol, 4.85% elemental, and 0.15%
oraanic."
Append/xi ASSUMPTIONS FOR EVALUATING RADIATION DOSES FOR EQUIPMENT QUALIFICATION "This appendix addresses assumptions associated        As noted above for Regulatory Position 6 with equipment qualification that are acceptable to    "Assumptions For Evaluating The Radiation Doses the NRC staff for performing radiological              For Equipment Qualification," PNPP Is retaining assessments. As stated In Regulatory Position 6 of    conformance to RG 1.89 rather than conforming to this guide, this appendix supersedes Regulatory        the guidance in this Appendix I.
Positions 2.c.(1) and 2.c.(2) and Appendix D of Revision 1 of Regulatory Gulde 1.89, "Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants" (USNRC, June 1984), for operating reactors that have amended their licensing basis to use an alternative source term. Except as stated in this
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 40 of 43
* Regulato~    Gulde 1.183 Guidance                Degme of Confonnance appendix, other assumptions, methods, and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.... "
 
==4.0 REGULATORY EVALUATION==
 
4.1 Applicable Regulatory Requirements/Criteria The traditional methods (prior to the AST) for calculating 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. 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. Therefore RG 1.183 serves as the base guidance for this submittal.
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 is consistent with the guidance found in RG 1.183. The plant-specific information provided above to support the license amendment request adequately addresses the guidance found in SRP 15.0.1.
The Degree of Conformance matrix identifies some exceptions to the regulatory guidance, as do the USAR markups in Addendum 1. These are provided for NRC review as part of this amendment application.
As shown in the Degree of Conformance matrix, the results of the revised radiological dose consequence analyses using an AST meet the.requirements of 10 CFR 50.67 related to individuals located at the exclusion area boundary for any two hour period (a 25 rem TEDE limit), at the low population zone boundary during the entire period of the passage of a plume (a 25 rem TEDE limit), and in the control room for the duration of the accident (a 5 rem TEDE limit). The results also meet guidelines included in RG 1.183, some of which are a fraction of the 10 CFR 50.67 limits.
As no modifications to installed plant systems or reductions in TS requirements are proposed as part of this RLA, the guidance provided in the General Design Criteria in Appendix A of 10 CFR 50 (as incorporated in the PNPP licensing basis) is not challenged by this application.
Also, 10 CFR 20, 10 CFR 50 Appendix I, Technical Specification 5.5.4 "Radioactive Effluent Controls Program," and Technical Specification 5.6.3 "Radioactive Effluent Release Report," each require monitoring of releases and provide limitations on their
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page 41of43 magnitude. The revised radiological calculations do not revise compliance with these regulations or the PNPP Technical Specifications.
4.2 Precedent Although a number of alternative AST submittals have been reviewed and approved by the NRC since RG 1.183 was published, and other licensees have revised the TS definition of dose equivalent 1-131 based on updated dose conversion factors, no specific precedent submittals are referenced herein.
4.3 Significant Hazards Consideration A license amendment is requested for the Perry Nuclear Power Plant (PNPP) per the requirements of the Code of Federal Regulations (CFR), Title 10, Sections (&sect;) 50.67 "Accident source term," &sect;50.59 "Changes, tests, and experiments," and &sect;50.90 "Application for amendment of license, construction permit, or early site permit."
FirstEnergy Nuclear Operating Company proposes to revise the PNPP Updated Safety Analysis Report (USAR) to reflect updated radiological calculations using an alternative accident source term (AST) for the applicable design basis events addressed by Regulatory Guide (RG) 1.183 "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." 10 CFR 50.67 requires*licensees who seek to revise a radiological source term utilized in design basis radiological consequence analyses to apply for a license amendment under
&sect;50.90.
The revised calculations use the NUREG/CR-6604 methodology "RADTRAD: A Simplified Model for Radionuclide Transport and Removal And Dose Estimation," to detennine a total effective dose equivalent (TEDE) dose. The associated USAR changes that reflect the revised calculations were evaluated under the criteria of 10 CFR 50.59 and were determined to result in a departure from a method of evaluation described in the Final Safety Analysis Report (FSAR) (as updated) used in establishing the design bases, which also requires amendment of the operating license pursuant to &sect;50.90.
Implementation of these USAR changes requires modification of the DOSE EQUIVALENT IODINE-131 definition in the PNPP Technical Specifications, to reflect use of Federal Guidance Report (FGR) 11 dose conversion factors in the dose analyses.*
An evaluation of whether a significant hazards consideration is involved with the proposed amendment was performed by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
: 1. This proposed amendment does not Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed amendment involves implementation of the AST for the control rod drop accident (CRDA) and the main steam line break (MSLB) at PNPP. The proposed amendment also updates the methods and assumptions used in the loss
 
Evaluation of the Revised Dose Calculation Request For Licensing Action
* Page 42 of 43 of coolant accident (LOCA) dose calculation, which maintains conformance with Regulatory Guide 1.183, and revises the TS DOSE EQUIVALENT 1-131 definition.
The proposed amendment does not involve any physical design modifications to plant structures, systems or components other than the planned use of GNF2 fuel beginning with Cycle 16, and the revised calculations do not impact any accident initiators. Because design basis accident initiators are not being altered, the probability of an accident previously evaluated is not affected.
With respect to consequences, the AST is an input to calculations used to evaluate the consequences of an accident, and that AST input does not by itself affect the plant response, or the actual path of radiation postulated to be released.
The design basis radiological consequence analyses themselves, which include updates to the core source term, input assumptions, and the methodology used to calculate dose consequences, do not affect the plant response, or the actual pathway of radiation that might be released during an event. Likewise, the DOSE EQUIVALENT 1-131 definition revision does not affect any plant response. For the evaluated events and the definition revision, the analyses demonstrate acceptable doses within regulatory limits. As detailed in the technical evaluation for the amendment request, a comparison of the former dose consequences against the newly calculated dose consequences for the evaluated events showed that the doses at the EAB and the LPZ are either negligibly changed or are lower than previously evaluated, except for the CRDA Scenario 1 analysis, for which the calculated doses increase, but by less than 2 percent of the margin to the acceptance criteria. The acceptance criteria for the CRDA is specified in Regulatory Guide 1.183 Table 6, and is only 25 percent of the regulatory limit specified in 10 CFR 50.67. Control room doses for the LOCA event decrease; for the other events, control room doses were not previously required to be calculated. Therefore, it is concluded that the consequences of previously evaluated accidents are not significantly increased.
Based on the above conclusions, this proposed amendment 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 amendment does not involve a physical alteration of the plant.
No new or different type of equipment will be installed and there are no physical modifications to existing installed equipment associated with the proposed changes. Also, there are no proposed changes to the methods governing plant/system operation, so no new initiators or precursors of a new or different kind of accident are created. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed amendment.
 
Evaluation of the Revised Dose Calculation Request For Licensing Action Page43 of43 Thus, this amendment 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.
Approval is requested for changes that primarily conform with Regulatory .
Guide 1.183 for the CRDA, MSLB, and LOCA analyses, as well as the TS DOSE EQUIVALENT 1-131 definition. The results of the accident analyses, including the use of FGR 11 dose conversion factors, are subject to acceptance criteria specified in 10 CFR 50.67 "Accident source term," and Regulatory Guide 1.183.
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 the analyses adequately bound postulated event scenarios. The dose consequences remain within the acceptance criteria presented in 10 CFR 50.67 "Accident source term," and Regulatory Guide 1.183. The only calculated doses that were determined to increase did so by less than 2 percent of the margin to the acceptance criteria specified in Regulatory Guide 1.183 (the regulatory guide acceptance criteria are 25 percent of the regulatory limits specified in 10 CFR 50.67).
Therefore the proposed license amendment 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 amendment.
4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with NRC regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
 
==5.0 ENVIRONMENTAL CONSIDERATION==
 
An examination of the three criteria provided in 10 CFR 51.22 Criterion for categorical exclusion: identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review Section (c)(9) determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
 
ADDENDUM 1 Updated Safety Analysis Report (USAR) Page Changes (Mark Up) 146 pages follow
 
TABLE 1.3-8 (Continued)
FSAR Section in Which Change Item      Change                Reason for Change        is Discussed Purge System The purge system      Four-second closing      9.4.6 Isolation    isolation valve        time is in accordance Valve's      closing times were    with Branch Technical changed from 1 to      Position 6-4. Offsite 4 seconds for the      releases would valves isolating      exceed <10 CFR the drywell from      guidelines.
the containment Examples showing and from 2 seconds                                retention of 10 CFR 100 to 4 seconds for                                  references due to the valves isolat-                                historical nature of this ing the contain~                                  Table entitled ment from the out-side.
                                                              "Significant Changes from PSAR To FSAR" Annulus      The maximum dis-      The change reflec s Exhaust Gas  charge rate cap-      the actual perfo -        Figure 6.5-1 Treatment    ability was          mance capabili System      revised from 650      the system an to 2,000 cfm.          revised disc rate does n    exceed the guidel' es of
                                    <10 CFR 100>.
Annulus      Delete the cap-        The change permitted      6.5.3, Exhaust Gas  ability to auto-      the use of a non-        Figure 6.5-1 Treatment    matically isolate      saf ety-related System      the active filter-    radiation monitor ing system and        and simplified the automatically start    control system.
the standby fil-tering system on indication of high radiation in the exhaust from this system. Start both units auto-matically follow-ing an accident.
Revision 12
: 1. 3-36                      January, 2003
 
TABLE 1.8-1 CONFORMANCE TO NRC REGULATORY GUIDES USAR Section/
Regulatory Guide (Rev.; RRRC Category)                Degree of Conformance                        Reference
<Regulatory Guide 1.1> - (Revision 0 - 11/70; RRRC Category 1)
Net positive suction head for emergency    PNPP  confo~ms  to t~is guide.                  <Section 5.4.7>,
core cooling and containment heat                                                            <Section 6.3.2>
removal system pumps
<Regulatory Guide 1.2> - (Revision 0 - 11/70; RRRC Category 1)
Thermal shock to reactor pressure          Withdrawn by the NRC June 1991. Super-seded by <10 CFR 50.61>, "Fracture
                                                                                              <Section 3.11.5.2.2>
vessels Toughness Requirements for Protection            <Section 6.2.4.2.3>
Against Pressurized Thermal Shock Events."
analyses
<Regulatory Guide 1.3> - (Revision 2 - 6/74; RRRC Category 1)
Assumptions used for evaluating the        The original licensing        LOCA                        2.3.4>,
potential radiological consequences of      radiological              , which were                    2.3.5>,
a loss-of-coolant accident for              primarily based on <Regulatory Guide 1.3>                  6.5.1>,
boiling water reactors                      and SRP 15.6.5, are now used only for            <Section 9.4.2>,
post-LOCA equipment qualification,              <Section 12.6.1>,
vital area access, ii-RQ PASS access            *<Section 15.0.3>,
analyses. The current LOCA dose                  <Section 15.6.5>
calculations are based 9n the alter source terms and assumptions presen in <WYR~C 14~5>, with modifications as descr ed in the referenced USAR sections.
                                                                            , control room Regulatory Guide 1.183            dose due to          Revision 13
: 1. 8-3                radiation shine,    December, 2003 and containment purge isolation
 
TABLE 1.8-1 (Continued)
USAR Section/
Regulatory Guide (Rev.; RRRC Category)                  Degree of Conformance                          Reference
<Regulatory Guide 1.4> -  (Revision 2 - 6/74)
Assumptions used for evaluating the          Not applicable to the PNPP design.
potential radiological consequences of a loss-of-coolant accident for                                              Not Applicable. Replaced by Regulatory Guide 1.183 pressurized water reactors
<Regulatory Guide L5> -  (Revision 0 - 3/71; RRRC Category 1)
Assumptions used for evaluating the                                                                <Section 2.3>,
potential radiological consequences of                                                            <Section 15.6.4>
a steam line break accident for bo.iling    aae  avera~e  ~ai:FJ'R:a eaer~ies *1ere talEeR water reactors                                fre~ W~C  TACT III aRe/er TACT 5 seae 1 aae at~9S~Reris aiffysieR is as aeseriBea iR Ye~R <eestieR 2.a>.
<Regulatory Guide 1.6> -  (Revision 0 - 3/71; RRRC Category 1)
Independence between redundant standby      The independence among standby power                <Section 7.1.2>,
(onsite) power sources and between          sources and among their distribution                <Section 8.1>,
their distribution systems                  systems is in accordance with this guide.            <Section 8.3.1>
The HPCS system conformance is discussed in <Section 8.3.1>.
<Regulatory Guide 1.7> -  (Revision 2 - 11/78; RRRC Category 1)
Control of combustible gas concentra-        PNPP conforms to this guide.                        <Section  6.1.1>,
tions in containment following a                                                                  <Section  6.2.5>,
loss-of-coolant accident                                                                          <Section  7.3.1>,
                                                                                                  <Section  7.3.2>,
Tech. Specs.
Revision 13
: 1. 8-4                                        December, 2003
 
TABLE 1.8-1 (Continued)
USAR Section/
Regulatory Guide (Rev.; RRRC Category)                Degree of Conformance                  Reference
<Regulatory Guide 1.86> - (Revision 0 - 6/74; RRRC Category 1)
Termination of operating licenses for        PNPP will comply with this guide.
nuclear reactors
<Regulatory Guide 1.87> -  (Revision 1 - 6/75; RRRC Category 1)
Guidance for construction of Class I        Not applicable to PNPP design.
components in elevated-temperature reactors (supplement to ASME Section III Code Classes 1592, 1593, 1~~4, 1595, arid 1596)
<Regulatory Guide 1.88> - (Revision 2 - 10/76; RRRC Category 1)
Collection, storage and maintenance of      See <Chapter 17.2>                          <Section 17.2>
nuclear power quality assurance records
<Regulatory Guide 1.89> -  (Revision 1 - 6/84; RRRC Category 4)
Qualification of Class lE equipment          Class lE equipment is qualified in          <Section 3.10>,
for nuclear power plants                    accordance with IEEE Standard 323-1974,    <Section 3.11>,
as endorsed by <Regulatory Guide 1.89>      <Section 7.1.2>,
No changes proposed.                      with the following specific exceptions:      <Table 8.1-2>,
Provided for context.                                                                  <Section 8.3.1>
: 1. NSSS Class IE equipment located in mild environmental zones was procured and qualified to IEEE Standard 323-1971.
: 2. Regulatory Position C2. The basis for radiological source terms used in discussed in <Section 3.11.5.2.2>.
Revision 17 1.8-37                                October, 2011
 
TABLE 1.8-1 (Continued)
USAR Section/
Regulatory Guide (Rev.; RRRC Category)                Degree of Conformance                Reference
<Regulatory Guide 1.89> - (Revision 1 - 6/84; RRRC Category 4)  (Continued)
: 3. Additional specific guidance for type No changes proposed.                          testing of cables, field splices and Provided for context.                        terminations is provided by IEEE Standard 383-1974, <Table 8.1-2>.
: 4. Specific criteria for assessing the acceptability of the environmental qualification program for safety related electrical equipment in a harsh environmental is provided by
                                              <NUREG-0588> Category I.
: 5. The acceptance criteria for the environmental qualification of safety related equipment located in a mild environment is the following:
: a. The documentation required to demonstrate qualifications of safety related equipment in a mild environmental is the "Design/Purchase" specifications. The specifications contain a description of the functional requirements for its specific environmental zone during normal and abnormal environmental conditions. A well supported maintenance/surveillance program in conjunction with a good preventive maintenance program will ensure that equipment that meets the specifications is qualified for the designed life.
Revision 17
: 1. 8-37a                              October, 2011
 
TABLE 1.8-1 (Continued)
USAR Section/
Regulatory Guide (Rev.; RRRC Category)                Degree of Conformance            Reference
<Regulatory Guide 1.89> - (Revision 1 - 6/84; RRRC Category 4)  (Continued)
: b. The maintenance/surveillance No changes proposed.                            program data and records will be Provided for context.                            reviewed periodically (no~ more than 24 months) to ensure that the design qualified life has not suffered thermal and cyclic degradation resulting from the accumulated stresses triggered by the abnormal environmental conditions and the normal wear due to its service condition.
Engineering judgment shall be used to modify the replacement program and/or replace the equipment deemed necessary.
Revision 17 l.8-37b                          October, 2011
 
evaluating design basis accidents TABLE 1.8-1 (Continued)
USAR Section/
Regulatory Guide (Rev.; RRRC Category)                                                                                      De ree of Conformance                      Reference
                                                                                                                                                                                                      <Section 3.11.5.2.2>
                            <Regulatory Guide 1.183> -                                          (Revision 0 - 7/00)                                                                                  <Section 6.2.4.2.3>
                                                                                                                                                                                                      <Section 6.5>
Alternative Radiological Source Terms                                                                      PNPP conforms to this guide for                      <Section . . >,
                            ~~r Eval~ating Cesig~ Basis Accidents                                                                      r~e:        ~3R~liR~ AGGi~ent Wlth the              <Section at Nuclear Power Reactors                                                                                  following exceptions:                                        on
                                                                                                                                                                                              <Section
: 1. Appendix B, Section 2; water depth                Tech. Specs.
above reactor flange inside containment is less than 23 feet.
                                                                                                                                                                                                          <Section  12.6.1 >
                                                                                                                                                                                                          <Section  15.4.9>
: 2. Table 6, and Appendix B Sections 4.1                          <Section  15.6.4>
and 5.3; the radioactivity that                          <Section  15.6.5>
escapes from the pool is assumed to be released to the environment instantaneously.
                            <Re ulator                    Guide 1.190> -                        (Revision Calculational and dosimetry                                                                                Neutron fluence methodologies                        4.1, 4.3, 5.3 determining pressure vessel                                                                                used by PNPP will conform to this fluence                                                                                                    guide.
                            <Regulatory Guide 1.192>)
Operation and Maintena                                                                                    PNPP conforms to the regulatory guide Acceptability, ASME O Code                                                                                revisions which correspond to the
~3~.""'A_p_p_e-n""'d.,..ix_I_:.,..in-"""tie_u_of.,,..co-n""'i~o-rm--an-ce--w.,..ith.,......Ap,a....p_e_n_d_ix-1-,-co-n.,,..io-rm-a-n-ce-w-ith--.PPli cable code of record. Application RG 1.89 is maintained, with exceptions as noted in its Degree of                                                                          f Code Cases will be evaluated against Conformance column and its listed USAR Sections/References.                                                                              he current revision of this regulatory
: 4. The original licensing basis accident source term is retained for post- uide prior to their use*
LOCA equipment qualification, vital area access, post-accident sampling system (PASS) aceess, control room dose due to radiation shine, and containment purge isolation analyses.                                                                                                                                                            Revision 15
: 5. A RG 1.183-based alternative accident source term is used in                                                                                                                                  October, 2007 1.8-61a radiological dose consequence analyses for LOCA, FHA, CROA, and MSLB; future revisions to other design basis analyses will also utilize a RG 1.183-based alternative accident source term.
 
2.1.3.4        Low Population Zone The Low Population Zone (LPZ) has been defined as a 2-1/2-mile radius from the plant center, which is midway between the Unit 1 and Unit 2 reactor buildings, as established in accordance with <10 CFR 100>.
  <Figure 2.1,;,18> illustrates topographical features characterist        the LPZ.  <Table 2.1-3> denotes the 1978 population distr!bution of which was determined using the methodology described in                        Example
  <Section 2. L 3 .1>. The estimated transient population distribution showing retention of provided in <Table 2.1-10>, along with the peak daily and seasonal            appropriate transient values (Reference 18) (Reference 19). To assist in the          10CFR100 formation of emergency planning, the following discussion of activities and facilities is not limited to the LPZ.
Listed below are the schools located within five miles of PNPP and their associated *1998-99 enrollment figures.
Enrollment as of
* Proximity to Institution                    1998-99 School Year
* NPP (miles)  I Perry Elementary                        643                        2-3      I Perry Middle                            605                        2-3 Perry High*                            677                        2-3
  *New Life Christian Academy                40                        2-3 Redbird Elementary                      483                        3-4
:  Hale Road Elementary                    310                        3-4 Homer Nash Kimball Elementary            428                        4-5 Revision 13 2.1-9                    December, 2003
 
3.1.2.6      Group VI, Fuel and Radioactivity Control (Criteria 60-64) 3.1.2.6.1      Criterion 60 - Control of Releases of Radioactive Materials to the Environment The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences.            Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.
3.1.2.6.1.1      Evaluation Against Criterion 60 Waste handling systems have been incorporated in the plant design_for processing and/or *retention of radioactive wastes from normal plant operations to.ensure that the effluent releases to the environment are as low as reasonably achievable and within the limits of <10 CFR 20>.
The plant is also designed with provisions to prevent radioactivity releases during accidents from exceeding the limits of <10 CFR (superseded by <10 CFR 50.67> for future design basis analyses)
The principal gaseous effluents from the plant during normal operation*
are the noncondensible gases from the air ejectors.          These gases are exhausted through a holdup system and a low temperature offgas treatment system including charcoal absorbers.      The effluent from this system.is continuous!~  monitored and controlled, and the system will be shut down and isolated in the event of abnormally high radiation levels.
Ventilation air from the various plant areas is exhausted through HEPA and charcoal filters, and is continuously monitored and controlled *.
Revision 12 3.1-88                          January, 2003
 
3.2        CLASSIFICATION OF STRUCTURES, COMPONENTS AND SYSTEMS Certain structures, components and systems of the nuclear plant are considered important to safety because they perform safety actions required to avoid or mitigate the consequences of abnormal operational transients or accidents.            The purpose of this section is to classify structures, components and systems according to the importance of the safety function they perform.              In addition, design requirements are placed upon such equipment to assure the proper performance of safety actions, when required.
3.2.1          SEISMIC CLASSIFICATION Plant structures, systems and components important to safety are designed to withstand the effects of a Safe Shutdown Earthquake (SSE) and remain functional if they are necessary to assure:
: a. The integrity of the reactor coolant pressure boundary,
: b. The capability to shut down the reactor and maintain it in a safe condition, or
: c. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of <10 CFR 100>.
Plant structures, systems and components ( ncluding their foundations and supports) designed to remain functiona                in the event of an SSE are designated as Seismic Category I, as indica                  in <Table 3.2-1>.
Structures, components, equipment, and syste s designated as Safety Class 1, Safety Class 2 or Safety                            <Section 3.2.3> for a discussion of safety classes) are                                      Category I except for (1) those noted in                                  (2) those portions of or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)                          Revision 12 3.2-1                        January, 2003
 
3.5        MISSILE PROTECTION 3.5.1        MISSILE SELECTION AND DESCRIPTION 3.5.1.l        Internally Generated Missiles (Outside Containment) 3.5.1.1.1        Criteria The following criteria have been adopted to assess the plant's capability to ensure that, if a generated missile of any postulated type occurs, there is:
: a. No loss of containment function.
: b. No direct loss of reactor coolant.
: c. No loss of function (assuming offsite power is not available during the shutdown of the plant) to systems required to shut down the reactor and maintain it in a safe shutdown condition, or mitigate the consequences of the missile damage, thereby ensuring:
: 1. No equipment will be damaged in one safety-related division, e.g., Division 1, from internally generated missiles originating from another safety-related division, e.g.,
Division 2.
: 2. No damage will occur to any safe shutdown equipment by missiles generated from nonsafety-related equipment.
: d. No offsite exposure will exceed the guidelines of <10 CFR 100>.
: e. No loss of integrity of the spent fuel pool.
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
Revision 12 3.5-1                            January, 2003
 
As discussed in <Section 3.4>, flood missiles are not a consideration.
Therefore, tornado generated missiles are considered as the limiting natural phenomena hazard in the design of all structures which are required for ensuring the integrity of the reactor coolant pressure boundary, ensuring the capability to shutdown the reactor and maintain it in a safe shutdown condition, and those whose failure could lead to radioactive releases which would exceed offsite radiation exposure limits (25% of <10 CFR 100> guideline exposures), as discussed in
            <Regulatory Guide 1.117>.
3.5.1.4.2            Missile Protec          Methods System and component safety cla sification and seismic category are given in <Table 3.2-1>.                      location within the building is provided by the layout drawings                        1.2-2>, <Figure 1.2-3>,
            <Figure 1.2-4>, <Figure 1.2-5>,                        1.2-6>, <Figure 1.2-7>,
            <Figure 1.2-8>, <Figure 1.2-9>,                        1.2-10>, <Figure 1.2-11>,
            <Figure 1.2-12>, <Figure 1.2-13>, < igure 1.2-14>, <Figure 1.2-15>,
            <Figure 1.2-16>, and <Figure                        Those systems or components listed in <Table 3.2-1> that                                ensure the integrity of the reactor coolant pressure boundary,                                                  or prevent release of radiation which exposure limits, are provided with                    do missile protection by location within Seismic Category I                              unique missile barriers, by the shielding of an                                Category I structure, or, they have been analyzed as discussed                n <Section 3.5.1.4.2.1>.
            <Table 3.2-1> also identifies Seismic Categ                                  The exterior walls and roof of these structures                                definition, to withstand the effects of the design                      tornado including tornado missiles.      These elements are two foot                            reinforced concrete having a 28 day compressive strength                        psi. Design approach is discussed in <Section 3.5.3>.                      ms or components located wholly within these structures are considered p otected from external or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)                      Revision 12 3.5-12                      January, 2003
 
structures are evaluated as not requiring unique tornado missile protection barriers.          Two approaches were used in this evaluation:
: a. Certain safety-related systems and components are screened out using the criteria of <Regulatory Guide 1.117> "Tornado Design Classification", including its Appendix, which together detail the items that should be protected from the effects of tornadoes.                The criteria in the Regulatory Guide are summarized as "important" systems and components required to ensure the integrity of the reactor coolant pressure boundary; ensure the capability to shut down the reactor and maintain it in a safe shutdown condition; and those whose failure could lead to radioactive              rele~ses  resulting in calculated offsite exposures greater than 25% of the g&#xb5;ideline exposures of <10 CFR 100> using appropriately conservative analytical methods and ass                        The safety-related systems and components not required to                            <Regulatory Guide 1.117>
guidelines are evaluated as            ot requiring unique tornado missile protection.
: b.    "Important" systems and compone ts (as discussed in <Regulatory Guide 1.117>) are generally prot cted.                The limited amount of unprotected portions of important                          components are analyzed using a probabilistic mi                          analysis as permitted in Standard                                      siles Generated by Natural Phenomena".      This analysis is condu (cumulative) probability per year of              issiles striking important structures, systems                                  to postulated tornadoes.
This information is then utilized to d termine the specific design provisions that must be provided to mai tain the estimate of strike probability below an allowable level ..
The allowable level established for the pr tection of such systems and components at PNPP is consistent with t e acceptance criteria in Standard Review Plan 2.2.3 "Evaluation of Potential Accid~nts",
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 wlll be updated to <10 CFR 50.67>)
Revision 12 3.5-14                        January, 2003
 
i.e., that a probability of occurrence of initiating events (those that could lead to potential consequences in excess of the e.._
                <10 CFR 100>        uidelines) of "approximately 10- 6 per year is acceptable if        when  crn~ined    with reasonable qualitative arguments, the realistic        robability can be shown to be lower." The PNPP-specific                                            the total probability of tornado                simply striking an important system or component must be shown                              less than 10- 6 per year. This PNPP-specific                              the following inherent conservatisms:
* It is assumed th t an important system or component simply being                              missile will result in damage from performing its intended safety function, although                        realistic for all cases.
* It is assumed that          e damage to the important system or component results in                                        to result in conservatively calcula                              release values in excess of the <10                                  although this is not realistic for all
* There are                                            impact on irradiated of the Intermediate Buildin                Any missiles postulated to enter this area either miss              e pools entirely, are stopped by internal walls, or                      far side of the pool above the level of the fuel.
* While residing in the FHB rail              y, the HI-STORM and HI-TRAC transfer cask with spent fuel                vulnerable to a tornado missile strike due to the FHB                            which is identified as not being                                  against a tornado missile per USAR <Table 3.5-6>.                  HI-TRAC transfer cask and HI-STORM storage cask have been                                    missiles while being located on the (Reference 13).                                      sis for overturning of or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)                        Revision 18 3.5-15                        October, 2013
 
3.7      SEISMIC DESIGN 3.7.1        SEISMIC INPUT Geologic and seismologic surveys of the site were conducted to establish two design earthquakes with different intensities of ground motion.
These are the operating basis earthquake (OBE) and the safe shutdown earthquake (SSE).
The OBE is postulated to be the earthquake which could reasonably be expected to affect the plant site during the operating life of the plant. The OBE produces the vibratory ground motion for which the Seismic Category I structures, systems and components are designed to remain operational without undue risk to the health and safety of the public. The OBE is considered to be a modified Mercalli Intensity VI as measured at the site <Section 2.5.2.7>.
The SSE represents the strongest vibratory ground motion earthquake for which these features (as mentioned for OBE) are, as a minimum, designed to remain functional.        The SSE is considered to be a modified Mercal.li Intensity VII as measured at the site <Section 2.5.2.4> and
        <Section 2.5.2.6>.
These Seismic Category I structures, systems and components, and the seismically analyzed systems and components of the plant are necessary to assure:    (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential of fsite exposures comparable to the guideline The design earthquakes, OBE and SSE, for the plant are spe and SSE design response spectra.            These criteria are base          pl.ant site geologic investigations and or <1 OCFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)                Revision 12 3.7-1                      January, 2003
 
definitions) in <Table 3.2-1> and is referred to generally as "safety-related" equipment.            The safety-related mechanical and electrical equipment and its associated supports are classified Seismic Category I, except those portions of the radioactive waste treatment handling and disposal systems, whose postulated.simultaneous failure would not result in conservatively calculated offsite exposures comparable to the guidelin*e exposures of <10 CFR 100>.
For seismic and dynamic qualification, "safety-relate " equipment is categorized in three groups by safety function:
: a. Safety-related      El~ctrical    Equipment designated as "Class IE" per IEEE Standard 279          1971.
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
: b. Safety-related Mechanical Equipment
: 1.    "Active" Mechanical Equipment - is that which must move or change position to perform its safety function (examples are pumps, motor-operated valves, safety-relief valves, or check valves).
: 2.    "Passive" Mechanical Equipment - is that which must only maintain its pressure integrity to perform its safety function (examples are tanks, heat exchangers or manual valves).
A discussion of the listing of the specific equipment by MPL No. that was prepared to demonstrate the seismic and dynamic qualification is provided in <Section 3.10.4>.
Recognizing that the hydrodynamic loads associated with a loss-of-coolant accident (LOCA) and safety/relief valve (SRV) discharge can have a significant effect on the design of structures, systems and equipment, both seismic and hydrodynamic loads are addressed in this section. The hydrodynamic loads are applicable to equipment in the Revision 12 3.10-2                    January, 2003
 
No changes to this page.
Provided for context.
onsite emergency power from their respective units' standby diesel generators should loss of offsite power occur. No single failure can result in loss of ventilation to the ESW pumphouse.
3.11.5      ESTIMATED CHEMICAL AND RADIATION ENVIRONMENT This section presents the justification for the estimated chemical and radiation enviromnents of <Section 3.11.1.1> and <Figure 3.11-11>,
<Figure 3.11-12>, <Figure 3.11-13>, <Figure 3.11-14>, <Figure 3.11-15>,
<Figure 3.11-16>, <Figure 3.11-17>, <Figure 3.11-18>, <Figure 3.11-19>,
<Figure 3.11-20>, <Figure 3.11-21>, <Figure 3.11-22>, <Figure 3.11-23>,
<Figure 3.11-24>, <Figure 3.11-25>, <Figure 3.11-26>, <Figure 3.11-27>,
<Figure 3.11-28>, <Figure 3.11-29>, <Figure 3.11-30>, <Figure 3.11-31>,
<Figure 3 .11-32>, <Figure 3.11-33>, <Figure 3.11-34>, <Figure 3.11-35>,
<Figure 3.11-36>, <Figure 3.11-37>, and <Figure 3.11-38>.
3.11.5.1      Chemical Environment 3.11.5.1.1        Normal Operation Water of the reactor, suppression pool, upper containment pool, fuel storage pools, fuel transfer syste111, residual heat removal system, and emergency core cooling systems is not chemically inhibited and is controlled by ion exchange systems to be compatible with the no:rmal operating limits listed in <Table 3.11-13>.
Sampling capabilities are provided for periodic analysis of this water to assure compliance with operational limits.
3.11.5.1.2        Design Basis Accident Water released from the reactor to the suppression pool, following a design basis accident and used for the conta.inment spray, is calculated on the basis of <Regulatory Guide 1.7> to have a pH range of 4.5 to Revision 13 3.11-40                December, 2003
 
7.0, a conductivity of      ~21  &#xb5;S/CM, oxygen content of SB ppm, a carbon dioxide content of    ~1  ppm, dissolved hydrogen of      ~60  ppb, dissolved salts of  ~2x10- 5 g mole/L, and undissolved solids          ~9  ppm. No significant concentrations of airborne or waterborne deleterious chemicals have been identified due to the post-LOCA fission products.
For the  ~eviseEI  J\.ssiEleRt i;etuse Teri!\ (R.",i;T) design basis LOCA analysis, the suppression pool pH is maintained at 7 or above to minimize the conversion of cesium iodide to elemental iodine.              The SLCS is used following the design basis LOCA for postaccident containment water chemistry management <Section 15.6.5.5.1.8>.
The containment spray system provides demineralized water as described above (for containment depressurization), at 5,250 gpm per train (A and B), 120 psig and 132&deg;F from the containment spray headers.                The train A spray may be initiated in conjunction with the RHR operation 10 minutes after a LOCA signal (drywell high pressure and reactor vessel low Level 1) either manuaJ.ly or automatically on high containment pressure with the high drywell pressure signal still present.              The train B spray initiation logic is identical to train A except that an additional 90 second time delay is used in the design.
3.11.5.2        Radiation Environment 3.11.5.2.1        Normal Operation Radiation sources during normal plant operations are identified in
<Chapter 11> and <Chapter 12> for the various plant systems.
For the neutron and gamma sources, energy spectra information is provided in <Table 12.2-2>, <Table 12.2-3>, <Table 12.2-4>, and
<Table 12.2-5>.      Alpha and beta sources do not contribute to the integrated doses.      The resulting normal radiation environments, Revision 12 3.11-41                        January, 2003
 
integrated over 40 years, are given in <Figure 3.11-11>,
            <Figure 3 .11-12>, <Figure 3 .11-13>, <Figure 3.11-14>, <Figure 3.11-15>,
            <Figure 3 .11-16>, <Figure 3 .11-17>, <Figure 3.11-18>, <Figure 3.11-19>,
            <Figure 3.11-20>, <Figure 3 .11-21>, <Figure 3 .11-22>, <Figure 3.11-23>,
            <Figure 3.11-24>, <Figure 3.11-25>, <Figure 3 .11-26>, <Figure 3.11-27>,
            <Figure 3.11-28>, <Figure 3 .11-29>, <Figure 3.11-30>, <Figure 3.11-31>,
            <Figure 3 .11-32>, <Figure 3 .11-33>, <Figure 3.11-34>, <Figure 3.11-35>,
            <Figure 3.11-36>, <Figure 3.11-37>, and <Figure 3.11-38>
3.11.5.2.2      Design Basis Accident The radiation doses from recirculating fluid lines tised to determine the equipment qualification environmental conditions are in accordance with <NUREG-0588> for comment (dated December 1979) and <Regulatory Guide 1.89>, and are based on the radiation sources given in
            <NUREG-0737>, Section II.B.2. The postaccident radiation doses (for equipment qualification purposes) are based on these more limiting
            <NUREG-0737> sources as opposed to the source terms used for the accident source terms used in the LOCA submittal Equipment qualification environmental conditions are reflected in the Environmental Tables on <Figure 3.11-11>,
alternative <Figure 3 .11-12>, <Figure 3 .11-13>, <Figure 3.11-14>, <Figure 3 .11-15>,
            <Figure 3.11-16>, <Figure 3.11-17>, <Figure 3.11-18>, <Figure 3.11-19>,
            <Figure 3.11-20>, <Figure 3 .11-21>, <Figure 3.11-22>, <Figure 3.11-23>,
            <Figure 3.11-24>, <Figure 3 .11-25>, <Figure 3.11-26>, <Figure 3.11-27>,
            <Figure 3 .11-28>, <Figure 3 .11-29>, <Figure 3.11-30>, <Figure 3.11-31>,
            <Figure 3.11-32>, <Figure 3.11-33>, <Figure 3.11-34>, <Figure 3.11-35>,
            <Figure .3.11-36>, <Figure 3.11-37>, and <Figure 3.11-38>. Accident radiation environments are provided for both gamma and beta integrated over six months.
Revision 12
: 3. 11-42                January, 2003
: 8. Institute of Electrical and Electronics Engineers (IEEE),
    "Standa~d  for Type Test of Class lE Electric Cables, Field Splices and Connections for Nuclear Power Generating Stations,"
Standard 383, dated 1974.
: 9.  ~NUREG-0588>,  "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment."
: 10. Institute of Electrical and Electronics Engineers (IEEE!, "Guide for the Preparation of Test Procedures for the Thermal Evaluation and Establishment of Temperature Indices of Solid Electrical Insulating Materials," Standard 98, dated 1972.
: 11. Institute of Electrical and Electronics Engineers (IEEE!, "Guide for the Preparation of Test Procedures for the Thermal evaluation of Insulation Systems for Electrical Equipment," Standard 99, dated 1970.
: 12. Institute of Electrical and Electronics Engineers (IEEE), "Guide for Statistical Analysis of Thermal Life Test Data," Standard 101, dated 1972.
: 13. Electrical Power Research Institute (EPRI), "A Review of Equipment Aging Theory and Technology" NP-1558, Franklin Research Center.
: 14. Amendment No.                                    No. NPF Perry Nuclear Power Plant Unit 1 (TAC No. MQ~Q~l).
Revision 12 3.11-44                  Jan:Jary, 2003
 
LIST OF TABLES (Continued)
Table                      Title                                  Page 6.5-4  Control Room Emergency Recirculation System Materials List (Design Data)                            . 6~S-43 6.5-5  Fuel Handling Area Exhaust Subsystem Materials List (Design Data)                              6.5-46 6.5-6  Annulus Exhaust Gas Treatment System Materials List (Design Data)                              ~.5-49 6.5-7  Primary Containment Operation Following a Design Basis Accident                                      6.5-52 6.5-8  Design Data for Annulus Exhaust Gas Treatment System Components                                          6.5-53 6.5-9  Input Parameters for the Spray Removal Analysis *6.5-56 6.5-10 (Deleted)                Deposition Removal Factors        6.5-57 6.5-11 Elemental              Partio~late    Roffisval Rate for Cont3iRffiORt g~ray                                    6.5-58 6.9-1  Single Failure Analysis of Feedwater Leakage Control System                                            6.9-7 Revision 17 6-xi                            October, 2011
 
6.2        CONTAINMENT SYSTEMS 6.2.1        CONTAINMENT FUNCTIONAL DESIGN 6.2.1.1        Containment Structure 6.2.1.1.1        Design Bases The pressure suppression containment system is designed to have the following functional capabilities:
: a. The containment and dry1-1ell have the capability to maintain functional integrity during and following peak transient pressures and temperatures which would occur following any postulated loss-of-coolant accident (LOCA).      The LOCA includes the worst single failure (which leads to maximum containment and drywell pressure and temperature) and is further postulated to occur simultaneously with loss of offsite power and a safe shutdown earthquake (SSE) . A detailed discussion of LOCA events is presented in <Section 6.2.1.1.3.3>.      A detailed discussion of mass and energy released is presented in <Section 6.2.1.3>.
50.67
: b.                    , in combination with other accident mitigation product leakage during and following the basis accident to values less than leakage rates that woul          in offsite doses greater than those stated in
      <10 CFR  ~>  (fer tRe aesi&sect;R easis    R:.~eT bQCA aRalysis 1 tRe lieeRsiR&sect; easis  li~it  is  ~a  re~ T~Q~),
: c. The containment system and drywell can withstand coincident fluid jet forces associated with the flow from the postulated rupture of any pipe within the containment or drywell.
Revision 12 6.2-1                      January, 2003
 
6.2.2.4        Tests and Inspections The preoperational test program of the containment cooling system is described in <Section 6.2.2.2>.      *Preoperational testing of the RHR system is discussed in <Section 5.4.7.4>.
6.2.2.5        Instrumentation Requirements The details of the instrumentation are provided in <Section 7.3>.          The suppression pool cooling mode of the RHR system is manually initiated from the control room.
6.2.3        SECONDARY CONTAINMENT FUNCTIONAL DESIGN 6.2.3.1        Design Bases The secondary containment system includes the shield building and the annulus exhaust gas treatment system (AEGTS).      Details of the AEGTS are given in <Section 6.5.3>.      The following are the design bases for the shield building:
LOCA                            50.67
: a.                                        to collect the fission leakage from the primary postulated design basis aosieeRt and delay it until released to the environs after processing through exhaust gas treatment system such that the resu are less than the values set forth in <10 CFR    +Q.G.> aRe <lQ CFR aQ, A~~QRBill  A> CeReral  Qesi~R CritorieR  1~.  (For tAo  eosi~R  easis R.~eT LOCA aRalysis, tAe  liooRsiR~  easis eoso limit is  ~a  rem T~Q~
effsite, aRe a rem    T~Q~ fer tAe oeRtrel room.)
: b. The shield building is designed to withstand the peak transient pressures and temperatures which could occur due to the postulated design basis accident.
Revision 12 6.2-70                      January, 2003
 
post-LOCA
: c. The shield building is                Seismic Category I structure.
: d. The shield building is ma* tained at a slight negative pressure relative to atmospheric  ressure (approximately 0.40 inch water gauge) so any leakage                    building or the containment vessel is into this space.
: e. The design loads on the shield building are discussed in
    <Section 3.8.1>.
: f. The leak tightness of the shield building is continually verified by maintaining the annulus at a vacuum of 0.40 inch water gauge.
This constitutes a continuous testing program. Inspection of the secondary containment structure will not be necessary as long as a vacuum can be maintained through normal operation of plant equipment.
6.2.3.2      System Design The shield building is a cylindrical reinforced concrete structure with a spherical dome enclosing the containment vessel. The internal diameter is 130 feet and the outside diameter is 136 feet. There is an annulus width of five feet between the containment vessel and the inside of the shield building.  <Figure 1.2-3>, <Figure 1.2-4>, <Figure 1.2-5>,
<Figure 1.2-6>, <Figure 1.2-7>, <Figure 1.2-8>, <Figure 1.2-9>,
<Figure 1.2-10>, <Figure 1.2-11>, <Figure 1.2-12>, and <Figure 1.2-13>
show plan and elevation views of the shield building.
There are two doors allowing access to the annulus area, both of which are normally locked. They are provided with position indicators and alarms which annunciate in the control room.
A tabulation of the design and performance data for the shield building is provided in <Table 6.2-28>.
Revision 12
: 6. 2-71                  January, 2003
 
and the control room Th~  performance objective of the shield building is to collect and retain any fission product leakage from the containment vessel during and following a design basis accident and, in conjunction with the annulus exhaust gas treatment system, process and release the fission products to the environs in a controlled manner.        This release is accomplished such that the resultant offsite doses to the general public are within the values given in <10 CFR .;i,.Q.Q.> aRe tho eosos to tho soRtrol CeReral for the ooRtrol rooffi,)
The principal construction codes, standards and guides used in the design of the shield building are described in <Section 3.8.1>.        In order to minimize the amount of radioactive material that leaks to the secondary containment following a design basis accident, primary containment penetrations are provided with redundant, ASME Code, Section III, Class 2, Seismic Category I isolation valves, one inside of the_ primary containment and one outside of the shield building, or some other acceptable configuration such as a closed system outside of containment. The piping out to the outboard containment isolation valve or in the closed system is also ASME Code, Section III, Class 2.        This isolation valve arrangement functions to minimize "through-line" leakage, which is limited by leak rate testing as described below.        The containment isolation system is discussed in more detail in
<Section 6.2.4>. The containment and reactor vessel isolation control system is discussed in more detail in <Section 7.3.l>.
The containment boundary and all penetrations except for penetrations with guard pipes terminate in the annulus.        Therefore, containment shell leakage and penetration leakage are considered to be totally directed to the annulus. The sources listed in <Table 6.2-33> are a sununary of potential leakage paths that could bypass the AEGTS.        The containment design basis accident leakage is 0.2 percent by weight of the contained Revision 12 6.2-72                        January, 2003
 
(although the radiological dose calculations have been revised to assume a secondary containment bypass leakage limit of 10;08 percent of L8 ).
atmosphere in 24 hours.      maximum test leakage rate permitted from the sources                  6.2-33> is 5.04 percent of the total containment leakage. This value will be the technical specification commitment for leakage bypassing the AEGTS as listed in the Technical Specifications. In order to verify that the total amount of potential bypass leakage will be within this limit, a testing and evaluation program will be conducted on isolation valves, personnel airlocks and guard pipes as described in <Section 6.2.4.3.1>.
The expected leakage rates per valve have been calculated and are shown on <Table 6.2-33> for the potential bypass leakage paths.        In these calculations, it was assumed that the onsite leakage limit per valve will be the same as the* shop test limits given in the valve specifications.
The air-filled lines penetrating primary containment that are not entirely contained in the secondary containment, and are potential bypass leakage paths, are identified in <Table 6.2-33>.      The air supply lines to the ADS accumulators are not considered bypass leakage paths since they are safety-related lines that remain pressurized post-LOCA (at a pressure exceeding containment vessel design pressure).
6.2.3.3      Design Evaluation All high energy lines which penetrate the containment wall and the shield wall are protected by guard pipes, with the exception of the control rod drive hydraulic supply <Table 6.2-33>.      The CRD supply is a 2-1/2 inch waLer line with a normal operating pressure of 1,850 psi.
The justifying logic allowing postulation of pipe rupture locations in other high energy lines penetrating the containment which require guard pipe protection does not apply to this line. Guard pipes are used where the energy release rate of the postulated accident is such that the resulting pressurization must be confined to the drywell rather than Revision 12 6.2-73                        January, 2003
 
6.2.4      CONTAINMENT ISOLATION SYSTEM 6.2.4.1        Design Bases                                        50.67 The design objective for the containment isolation systems normal or.emergency passage of fluids through the containm      t boundary.
while preserving the ability of the boundary to prevent or limit the escape of fission products that may result from postulate                so that site boundary dose guidelines specified by <10 CFR      ~>  are not exceeded (for tRe  aesi~A  sasis R.~~T hOCA aRalysis, tfie liseRsiR~ sasis offsite ease limit is  ~&sect;  rem TEQE) . This objective is achieved by provisions for automatic isolation of appropriate lines that penetrate the containment boundary.
The containment isolation systems are automatically actuated with input from the following sLgnals:
: a. Low reactor pressure vessel water level (3 setpoints).
: b. High drywell pressure.
: c. High main steam line tunnel high ambient or differential temperature.
: d. Main steam line high area temperature, turbine building.
: e.    (Deleted)
: f. High main steam line flow.
: g. Low main steam line pressure at turbine inlet.
: h. High radiation in containment and drywell purge exhaust.
Revision 14 6.2-78                    October, 2005
 
ensure that the site boundary dose guidelines specified are not exceeded following a postulated accident TEQE) . The "B" Train of the safety-related                          system provides postaccident makeup                                                  assure leak tightness of the outboard MSIVs. A                for rapid closure of all lines provides a containment barrier                  lines that is sufficient to maintain leakage within permissi le limits.
6.2.4.2      System Design 6.2.4.2.1      General A summary of containment isolation            is presented in
<Table 6.2-32>.                                including penetration number, applicable GDC or regulatory                  system, fluid, line size, valve arrangement, location,    alve type, actuation mode, valve position, various containment isola 'on valve arrangements.
Justification for con                      provisions which differ from the is presented in <Section 6.2.4.2.2>.
Containment isol tion valve closure times <Table 6.2-32> are established to prevent                        from exceeding the guidelines specified by            > (fer tR.e aesi~R  sasis RJ'.eT J;,GCA o3Ralysis 1 tl:i.e lieSRSiR~
sasis effsite aese limit is  ~9  rem TEQE) . A discussion of valve closure times, for those valves thrrx1gh which a direct path from containment to the environment could exist, is provided in <Chapter 15>.            Containment isolation for such lines is accomplished in accordance with NRC Branch Technical Position CSB 6-4, "Containment Purging During Normal Plant Operations."  Additional discussion of Branch Technical Position CSB 6-4 is presented in <Section 6.2.4.2.3>.
Revision 14 6.2-82                        October, 2005
 
bellows, and steel containment penetration, and outside the containment building by the transfer tube, drain line, drain valve, and local leak rate test valve (Reference 35).
: f. Conclusions Concerning General Design Criterion 56 To assure protection against the consequences of accidents involving the release of significant amounts of radioactive material, piping that penetrates containment has been shown to provide adequate isolation capability on a case-by-case basis in accordance with GDC 56.
In addition to satisfying the isolation requirements specified by GDC 56, the pressure retaining components of these systems are designed, fabricated and tested in accordance with the requirements of the ASME Code, Section III. In some cases, provision of a high quality system obviates the need for isolation valves due to the diminished probability of a rupture in such a system. Additional information concerning classification is presented by
    <Table 3.2-1>. The containment and reactor vessel isolation control system is addressed in <Section 7.3>.
6.2.4.2.2.3      (Deleted)                            No changes to this page.
Provided for context.
6.2.4.2.3      Consideration.of NRC Branch Technical Position CSB 6-4, "Containment Purging during Normal Plant Operations" The containment vessel and drywell purge system is designed to achieve the objectives stated in Branch Technical Position CSB 6-4. Purge system containment isolation valves are capable of isolating containment within five seconds. The containment purge system is described in
<Section 9.4.6>.
Revision 12 6.2-107                    January, 2003
 
Example for the initial cycle                          showing retention of 10CFR100
* Radiological consequences of a postulated LOCA during containment p system operation have been evaluated in accordance with Branch Tee Position CSB 6-4. Since drywell purge system (approximately 25,CO capacity) operation is restricted to reactor cold shutdown          condi~  ons and refueling operations, only the containment purge system (5,0 capacity) was assumed to be operating at the start of the calculated site boundary doses are 0.9 rem to the thyroid and                mrem whole body. These doses are a small fraction of the <10 CFR 100>
guideline values. This analysis is separate from the radiological dose analysis performed in <Chapter 15> for the limiting design basis LOCA and is not impacted by implementation of the revised accident source term in Jsioonso l\meRdH10Rt ~la. lQ:il.
                                                          ~
Major assumptions used in the dose analysis are as follow*:
: a. A double ended guillotine break of the recirculation          ine was assumed to occur instantaneously.          This accident was it represents the worst break and, consequently, the            ghest doses.
: b. Purge system isolation valve closure will isolate conta nment within five seconds (includes valve closure time of fou          seconds and an additional maximum time of one second for conserv tism).
During this period reactor coolant blowdown was conserva estimated to be 109,766 pounds <Table 6.2-25>.
(initial cycle)
: c. Forty percent of the blowdown was assumed to flash to steam.            It was conservatively assumed that the entire iodine activity in the flashed fraction of the total blowdown was instantaneously released to the containment atmosphere at the instant the accident occurred.
Plate out of iodine was ignored.        Retention of iodines in the suppression pool was also ignored although, actually, the flashed activity would first be dumped into the suppression pool and would then slowly evolve into containment.
Revision 12 6.2-108                      January, 2003
 
No changes to this page.
Provided for context.
: d. Specific activity in the reactor coolant was conservatively assumed to be 6.56 &#xb5;Ci/g of I-131 and 34.9 &#xb5;Ci/g of Xe-133, with other isotopes in proportionate quantities. This corresponds to spike conditions.
: e. Turbulence resulting from the high blowdown rates and operation of fan coolers in containment was assumed to ensure good mixing in the entire containment volume.
: f. Containment air was assumed to be released through two 18 inch purge lines, one supply and one exhaust, for five seconds.
Constant flow rates through the open purge lines corresponding to the maximum containment pressure of approximately 3.0 psig during the release period <Figure 6.2-2> were used to determine a total flow to the environment of 1,020 pounds. This value is conservative since it ignores lower flow rates due to lower containment pressures and partial closure of the purge isolation valves at times prior to five seconds.
: g. No credit was allowed for iodine removal by the 99 percent efficient charcoal adsorbers in the containment purge exhaust lines.
: h. Site boundary x/O <Table 15.6-12> was used in the dose calculation.
6.2.4.3      Design Evaluation 6.2.4.3.l      General Evaluation To ensure the accomplishment of the design objective stated in
<Section 6.2.4.1>, redundancy is provided in all design aspects of the containment isolation systems. Mechanical components are redundant and each isolation valve is protected, by separation and/or adequate barriers, against the consequences of potential missiles. Also, system Revision 12 "6.2-109                    January, 2003
: c. Capacity The normal occupancy level of the control room is six people following an  ~ccident.
: d. Food, Water, Medical Supplies, and Sanitary Facilities First aid equipment, food and water are provided to sustain seven people for seven days following an accident.                        Chemical toilet facilities are provided for use in the event that normal sanitary facilities become inoperative.
10 CFR 50.67
: e. Radiation Protection Radiat.i on protection, as required by <+l~Q'""""Cr!!li''"I'~,,...~,i.....-Q.Q.Qf'l~'"*-
CriterioR 19, is provided by shield walls on the four exposures, shield slabs at floor and ceiling, radiation monitoring equipment, and emergency filtering systems. The control room atmosphere is monitored for radiation. When required, the control room atmosphere can be recirculated through the emergency filter system to remove contaminants. This filter system consists of roughing, high efficiency particulate air (HEPA) and charcoal filters.
Assumptions and analyses regarding sources and amounts of radioactivity which may surround or leak into the control room and related shielding requirements are discussed in <Chapter 12> and
  <Chapter 15>. The radiation monitoring system is discussed in
  <Section 12.3.4>.
: f. Noxious Gas Protection Smoke detectors located in the control room air supply duct and in the emergency filter system discharge duct actuate alarms to indicate the pr.esence of smoke in these locations.                          Additionally, Revision 12 6.4-2                                            January, 2003
 
The fans, filter elements, coils, and dampers are of standard industrial design manufactured in accordance with the Quality Assurance (QA) requirements of Safety Class 3, Seismic Category I items.      The filter racks and plenums are specially designed to satisfy system space requirements and also meet the above QA requirements.      However, the electric duct heating coils and the humidifiers are  nonsafety~related items. Design information for the major components in this system is listed in <Table 6.4-2>.
6.4.2.2.2      Control Room Emergency Recirculation System This system provides the necessary supplementary particulate and halogen filtration of the air supplied to the control room areas and offices during emergency periods and other abnormal conditions for personnel protection.
This system is automatically activated by an emergency signal (such as LOCA), or high *radiation, or by manually setting the mode selector switch to the emergency recirculation mode. While the system is automatically activated, no credit is taken iA tAe  R.'>~T  aAal.ysis during the first 30 minutes following a design basis LOCA. As a result of a LOOP condition the system will automatically activate in the emergency recirculation mode when power is restored to the emergency busses.      The operator may shut down one of the trains after it has been established that both are operating satisfactorily. In addition, the receipt of the emergency signal causes dampers in the control room HVAC system to be automatically positioned to the emergency recirculation mode of operation (see Note 4 of <Figure 6.4-1 (2)>, for the damper positions) .The vortex damper operator (M25-F260A, Bl of the supply fan (M25-C001A, Bl is then automatically de-energized allowing the operator spring to partially close the variable inlet vanes to reduce the supply air flow to  30~000 cfm. This flow reduction is required so that the supply fan and recirculation fan flow rates are compatible.      Return fan Revision 12 6.4-8                      January, 2003
: 6. 4. 4      DESIGN EVALUATION Each of the operating systems which ensures control room habitability is discussed in detail in other sections.      These systems, and.the section in which they are discussed, are as follows:
: a.      Control room ventilation system, <Section 6.4.2>.
: b.      Fire protection system, <Section 9.5.1>.
: c.      Communications system, <Section 9.5.2>.
: d.      Lighting system, <Section 9.5.3>.
: e.      Offsite power system, <Section 8.2>.
: f.      Onsite power system, <Section 8.3>.
: g.      Radiation monitoring system, <Section 12.3.4>.
A summary evaluation of control room habitability based on selected considerations is presented in <Section 6.4.4.1>, <Section 6.4.4.2>,
<Section 6.4.4.3>, <Section 6.4.4.4>,<Section 6.4.4.5>, and
<Section 6.4.4.6>.
6.4.4.1        Radiological Protection The evaluation of radiological exposures to control room operators following a design basis loss-of-coolant accident is presented in
<Section 15.6.5>.
The analysis in <Section 15.6.5>    assu~es  a constant unfiltered coritrol room inleakage value of 1375    cfm~the      duration of the LOCA accident.
after 30 minutes and I
* Revision 13 6.4-14                      December, 2003
 
The design basis radiological calculations for a Normally, the control room boundary inleakage is maintained at a value consistent with pre-operational testing such that the actual inleakage is substantially less than 1375 cfm.
Throughout the life of the plant, various plant activities may need to be performed which temporarily degrade the control room boundary such that the unfiltered inleakage significantly exceeds 1375 cfm.                                          ~
                                        \!91'.'9  t;o 0601:11'.' l:ll'IQ91'.'  l:A9S9  ElORaitioRS I    Fl31'.'<HR9\;l'.'iEl that it is acceptable to delay the restoration of boundary, provided that once it is restored, the actual inleakage would be                              below 1375 cfm for the remainder accident. This allows                                  degradations of the boundary assume an unfiltered        to occur without                                          accident dose to the control room inleakage of      operators.                              controls are utilized during planned 6000 cfm for                                              boundary can be~estore~ithin the bounding the first 30
                                                          <fi~IH9 9 4 4 (    I          > "3RQ    '~1:11'.'9 e, 4 4 ( ~) >,
minutes, which shows                to at or                                                      !rapidly I    to
: 6. 4. 4. 2 No toxic materials which could interfere with control room occupancy are stored in the plant.            Sodium hypo-chlorite, rather than chlorine, is used as a biocide.            No chlorine is stored on site.                    The potential effects of offsite and onsite hazardous materials are discussed in
                  <Section 2.2.2> and <Section 2.2.3>.                          Protection against offsite toxic gases are detailed in <Section 6.4.1.g>.
: 6. 4. 4. 3      Control Room Emergency Recirculation System The general arrangement and control of the control room emergency recirculation system is as described in <Section 6.4.2.2.2>.                                        Detailed information concerning the emergency filter is presented in
                  <Section 6.5.1>.        The equipment is shielded, housed in a Seismic Category I structure, separated, redundant, and powered from the Revision 14 6.4-15                                  October, 2005
 
1Delete I
.!;"---:-Z---~.:--'----=,!:"-1--~+/----~~;_,_,........J*l----~"---~.----l,l----lo ca UP' 'faoil.&) "AP peon* P>P"O          tMev  U    12'03>>
PERRY NUCLEAR POWER PLANT Control Room Dose at 30 days - Final Unfiltered In-Leakage of 300 cfm Fivure &.t-4 (Sheet 1 of 2J
 
1Delete I
          '                                              :3
                    ~                                    *
                          '                                .c
                                                            -a UI g
                                                          *  :II l
                                                          * .!!*0 J
-- ---                                                    :I i
      -~-- .....
                                                          . I*c
                          --- ---                        :s ~
                            .a        PERRY NUCLEAR POWER PLANT Control Room Dose at 30 days - Final Unfiltered In-Leakage of 100 cfnl Figure 6.4-4 (Sheet 2 of 2)
 
6.5          FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6-. 5 .1        ENGINEERED SAFETY FEATURES (ESF) FILTER SYSTEMS The control room emergency recirculation system (CRERS), the exhaust subsystem of Fuel Handling Area Ventilation System (FHAVS) known as the Fuel Handling Area Exhaust Subsystem (FHAES), and the annulus exhaust gas treatment system (AEGTS) are the ESF filter systems that reduce the concentration of airborne radioactive contaminants following a design basis accident (OBA).
6.5.1.1          Design Bases Design bases for the charcoal adsorber plenums of the CRERS, FHAES and the AEGTS are as follows:
: a.      Design Criteria The CRERS, FHAES and AEGTS are safety-related. System design conforms with the requirements of General Design.Criteria (GDC) 1, 2, 3, 4, 19, 60, and 61 of <10 CFR 50, Appendix A>. To satisfy the requirements of these GDCs, the guidance presented in <Regulatory Guide 1.3>, <Regulatory Guide 1.13>, <Regulatory Guide 1.26>,
        <Regulatory Guide 1.29>, <Regulatory Guide 1.47>, and <Regulatory Guide 1.52> has been considered in the design of these systems.
: b.      Need for Filtration
                                    <10 CFR 50.67>
The remote                airborne radioactive contaminants following a LOCA and the requirements of need for the CRERS for filtration of control room air.          requires, in part, that adequate radiation protection be provided to permit access to, and occupancy of, the control room under accident conditions for the duration of the Revision 12 6.5-1                    January, 2003
 
TEDE.
exposure to personnel in excess of TEQE fer tfie eesi&sect;R Basis  ~GCA  aRe tfie fyel RaReliR&sect; aeeieeRt),
The remote possibility of release of airborne radioactive contaminants due to a fuel handling accident, the requirements of GDC 61, and the recommendations of <Regulatory Guide 1.13>
establish the need for the FHAES to accomplish fuel pool area air filtration. GDC 61 requires, in part, that fuel storage and handling, and radioactive waste and other systems that may contain radioactivity be designed to ensure adequate safety under normal and postulated accident conditions and that appropriate filtering systems be provided. However, no accident dose calculations credit the FHAES <Section 15.7.4> and <Section 15.7.6>.
The AEGTS is provided to reduce the radiological consequences of fission product releases in the containment from a LOCA, or fuel handling accident involving recently irradiated fuel, although credit is no longer taken for AEGTS filtration in the FHA dose calculation <Section 15.7.6>. AEGTS collects and filters the leakage from containment. Also, the AEGTS is designed to maintain a negative pressure in the annulus relative to the outside which minimizes ground level release of airborne radioactivity due to containment exfiltration during normal and postaccident conditions.
: c. Component System Sizing                      <10 CFR 50.67>
Two 100 percent capacity filter              provided for the CRERS.
Air flow rate for the CRERS is          cfm per plenum. Based on this assumed air flow rate and      assumed charcoal adsorber efficiencies and                    in <Section 15.6>, the overall dose to the                      an accident has been shown to satisfy the                      lQ, er tfie a re~ TEQ~ eese  li~it Y89Q fer ER9 eesi&sect;R eaeis  ~GCA eese ealeYlatieRS.
Revision 13 6.5-2                      December, 2003
 
Three 50 percent capacity filter units are provided for the FHAES.
The FHAES provides exhaust flow from the fuel handling area, the fuel pool cooling equipment rooms, the control rod drive maintenance area, and the control rod drive pump areas. Flow is 30,000 cfm. Of this quantity, 15,300 cfm is exhausted directly from the fuel pool area. This air flow rate is based on flow patterns that should entrain contaminants escaping from the fuel pool area.
Two 100 percent capacity AEGTS filter units are provided for each reactor unit. Air flow rate for the AEGTS is 2,000 cfm per plenum.
Based on this flow rate, the negative pressure in the annulus is maintained at -0.25 inches of water gauge minimum continuously.
Components of these filter systems have been sized to handle system air flow based on the recommendations of <Regulatory Guide 1.52>,
ERDA 76-21 and general engineering practice.
: d. Fission Product Removal Capability The fission product removal capability of the activated charcoal adsorber material used in the CRERS, FHAES and AEGTS is based on the recommendations of <Regulatory Guide 1.52>.
The decontamination efficiency of the AEGTS 99 percent for both elemental iodine and                  of iodine.
However, none of the radiological                        'da& tak9R fQF tR9            elemental and consequence organic iodines by the charcoal filters in the AEGTS. The AEGTS analyses charcoal adsorber bed is 4 inches deep with annulus exhaust air maintained at less than 70 percent relative humidity.
The decontamination efficiencies of the CRERS and FHAES charcoal adsorbers are 99 percent for elemental iodine and 95 percent for Revision 13 6.5-3                  December, 2003
 
the design basis LOCA <Section 15.6.5.5.1.9> credits an 80 percent removal efficiency of elemental. and organic iodines by the charcoal filters in the CRERS. The Steam System Piping Break Outside Containment <Section 15.6.4>, Control Rod Drop                        ~
Accident <Section 15.4.9>, and the Fuel Handling Accident <Section 15.7.4> and
<Section 15.7.6>, do not take credit for the charcoal filters in the CRERS.                          __
terffi LOCA aRalysis aRd oRe fHel RaRdliRq aooideRt seRsitivity ease asSYffiOG aR oleffioRtal aRd orqaRio roffioval offioieRoy of oRly 5QI for tRo oAareoal adsoreors.        For the FHAES, the alternative source term FHA analysis took no credit for the charcoal adsorbers.                    The CRERS and FHAES charcoal adsorber beds are 2 inches deep.                  Exhaust air for both plenums is maintained at less than 70 percent relative humidity.
The HEPA filter efficiency of all the plenums is 99.97 percent on particles 0.3 microns and larger.          However,R*-.e-_-G~~-~*~~~-~-;;-..4~~-~1-1+-~~
I I\
tRo  HE~A  filtors iR tRo alterRativo soHroo term aRalysis for tAo fyol RaRdliRq aooieoRt.
Additional bases for the design of the CRERS, FHAES and AEGTS aie presented in <Section 6.4>, <Section 9.4.2>, and <Section 6.5.3 respectively.              the LOCA analysis only credits the HEPA filters in the AEGTS and CRERS at an efficiency of 99 percent. The other design basis radiological 6.5.1.2          System Design    calculations do not take credit for the HEPA filters ....._...
In the AEGTS, CRERS, or FHAES.
The design features of the CRERS, FHAES and AEGTS are compared to the recommendations of <Regulatory Guide 1.52> in <Table 6.5-1>,
          <Table 6.5-2>, and <Table 6.5-3> respectively.
Design of the activated charcoal adsorber plenums used in the CRERS, FHAES and AEGTS follows the guidelines of <Regulatory Guide 1.52> and ERDA 76-21.
Each charcoal adsorber plenum contains the following:
: a. Demisters to remove large particles and water droplets (about 1 micron diameter) .
: b. Roughing filters to remove large particles (about 1 micron).
Revision 13 6.5-4                              December, 2003
 
radioactive contaminants following a OBA to develop decay heat that could ignite the charcoal adsorber material.
The charcoal adsorber plehurns are redundant, physically separated and powered from separate Class lE electrical systems.
In the event smoke from a fire is exhausted through any charcoal filter, the filter will be tested for any degradation in charcoal performance as a result of the smoke. This testing will be performed in accordance with the Ventilation Filter Testing Program. If the testing indicates that degradation has occurred beyond acceptable limits, the charcoal will be replaced. For charcoal filters in systems needed to mitigate the consequences of a LOCA, Ramely  t~e aRnulHs  ex~aHst  ~as treatmeRt system (ngcTg), the filters will be tested and the charcoal replaced, if required, within a period specified in the technical specifications.
6.5.1.4      Tests and Inspections Tests and inspections of the CRERS, FHAES and AEGTS charcoal adsorber plenums are performed prior to startup and on a periodic basis thereafter. Other tests and inspections of these filter systems are discussed in <Section 6.4>, <Section 9.4.2>, and <Section 6.5.3>
respectively.
6.5.1.4.1      Filter and Charcoal Adsorber Tests HEPA filters are individually tested by an appropriate filter test facility at 100 percent and 20 percent of rated flow,    in accordance with the recommendation of <Regulatory Guide 1.52>. Original or replacement HEPA filters used in the CRERS, FHAES and AEGTS are tested as indicated above.
Each batch of charcoal adsorber material satisfies the "acceptable results" recommended by <Regulatory Guide 1.52>. Since the charcoal Revision 12 6.5-7                      January, 2003
 
No changes to this page.
Provided for context.
product of primary concern in the evaluation of a loss-of-coolant accident. The major benefit of the containment spray is its capacity to
. collect and remove particulate iodine from the containment atmosphere and thus reduce its release to the environment. *offsite and control room operator doses are a function of both the rate of removal and the final equilibrium decontamination factor. The .dose calculation assumes *
(non-mechanistically) that the containment spray will operate for up to 24 hours. However, the dose calculations also expand on this assumption, noting the following:
: 1)  The dose calculations assume the sprays are run for the first 24 hours, then are suspended. This is the most important time period for scrubbing of radiation down into the suppression pool.
However, in an actual event, spray use would not necessarily be suspended at 24 hours, if appropriate conditions for their use still existed. Therefore, the phrase "up to" is not intended to be interpreted to stop using sprays after 24 hours.
: 2)  The phrase "up to" is intended to mean that in an actual event, the sprays will be run when it is appropriate, and not necessarily the entire time during the first 24 hours of a LOCA.      This does not invalidate the assumptions in the dose calculations.      The accident guidance to operators must be written to be symptom based, rather than event based. Most postulated LCX:As will not result in large radiation releases. Therefore, it would not be appropriate to run containment sprays for 24 hours following such an event.      Another critical factor in spray use is containment pressu.re.      Use of the sprays will work to reduce containment pressures, due to steam condensation and the containment heat removal function that they provide. In the majority of cases, if a high radiation signal is present from the containment radiation monitor and pressures are elevated in containment, the sprays would be run.      However, if containment pressure gets reduced to near zero and use of the sprays is terminated by the operators, this.does not have an Revision 12 6.5-12                    January, 2003
 
Elemental iodine removal is credited in the drywall and containment volumes. Airborne elemental iodine is removed by deposition to the walls in the drywell and containment. As reported in Section 5.1.2 of NUREG/CR-0009 (Reference 1), this process is driven by the temperature differences between the surfaces and the atmosphere.
pact on offsite doses (or the dose calculations) since pressure for containment and MSIV leakage has been The dose calcs assume that the maximum allowable corresponding to the peak postaccident pressure (Pa).*
during the entire 24      hour~eriod, so if containment re actually gets reduced to substantially less than Pa, a ion in leakage and the resultant offsite doses will follow.
6.5.2.3.        Iodine Removal Performance Evaluation The                              is based on the assumptions presented below and The                      flow associated with only one RHR pump operating in mode. It is conservatively assumed that the system directly sprays approximately 41 percent of the l containment free volume (excluding the drywell).          <Section 15.6.5>
a description of the volumes and flow paths used in the analyses.
          !deposition~                      factors            ~
The calculated iodine removal                  the containment s13rays are given*
in <Table 6. 5-11> fer Ui.e elemeRtal aREI 13artieYlate ieEliRes as '/ell as*
etRer 13artieYlates. QeeaYse ef tRe lar&sect;e syrfaee area ef tRe iRitially aireerRe 13artieylate, tRe elemeRtal ieEliRe is assYmeEI te se aelsereeel 9Rt9 tR9 13arl;]_9ylate aREI te ee rei:Reueel *1itR it.
It has been conservatively assumed in these                      of spray removal effectiveness that organic iodine forms are                      by the sprays.
6.5.2.3.2        Evaluation of Analytical Assumptions          These removal constants are applied until a decontamination 6.5.2.3.2.1        Iodine Retention by Spray Solution          factor (OF) of 200 *has been obtained.
The equilibrium between the concentrations of iodine in the liquid and vapor phases is given by the partition coefficient, H, which is a Revision 12 6.5-13                      January, 2003
 
function of iodine concentration, pH and temperature.            In accordance with (Reference 3) re-evolution of iodine does not have to be considered, (i.e., H will be very large) as long as the pH of the suppression pool .is maintained greater than or equal to 7. 0 postaccident.
6.5.2.3.2.2          Elemental Iodine and Particulate Removal Constant
~
TRo oaloYlatisnal ffisdol Ysoe ts      dotsrffiino tRO oloffiontal and partioHlato iedino roffiova 1 oonstant is    ~olostar  Applied TooRnolo&sect;y's "STARWAUA" ooffipYtor oodo  (~oforonoo  4)  1~ioR  inoorporatos ERO spray reffisval ffiodolin&sect; feat1.uos &sect;iven in ,'\ppendiM g of (Roferonos :g),      TRo inpHt data ro~Hirod  by tRe ooffipHter oodo is &sect;iven in <Tabla      ~.&sect;  9>,  TRo moan drop fall ROi&sect;Rt of &sect;:g 2 foot . *as oalsYlatod by ta){iR&sect; a HOi&sect;Rtod auEHa&sect;O of I
tRo ROi&sect;Rt of oaoR rin&sect; abe"e tRo oporatin&sect; floEH ana tfio assooiatea spray flmr rate. as folle*is:
tJR:ere:
49.7&sect;'
            =      94. 7&sect;'
anEl m - rin&sect; flo*;,. rate - RYffibor of noggles n flo*
* rate por OOi2ii2ile
            =      114  ll 19.22 &sect;pm      =
102  >< a.22 &sect;pm Tfie spatial and temporal distribYtions are derived from analysis Ysin&sect; tfio ygwRC's oOHlpYtor ooee "SPIRT" (Referonoe 1).
Revision 12 6.5-14                      January, 2003
 
Insert for USAR Section 6.5.2~3;2.2 The two Control Room Emergency Recirculation System (CRERS) subsystems each have a high efficiency particulate air filter, charcoal adsorbers, and a post .
HEPA filter. The CRERS is an ESF system that is tested in accordance with R.G. 1.52 (Reference 4). The calculation model (Reference 6) assumed an elemental and organic iodine removal efficiency of 80 percent for the charcoal .
adsorber removal efficiency.
* Each HEPA filter is tested to show a penetration and system bypass of less than 0.05 percent when tested in accordance with Regulatory Guide 1.52 (Reference 4). A penetration and bypass of less than 0.05 percent allows credit for a particulate removal efficiency of 99 percent per Regulatory Guide 1.52. The analysis therefore used a CRERS HEPA filter efficiency of 99 percent for aerosol particulates.
The AEGTS includes HEPA filters and 4-inch deep charcoal filters. Particulate removal by the HEPA filters is assumed to be 99 percent in accordance with Regulatory Guide 1.52. The analysis conservatively assumed a removal efficiency of O percent for the charcoal adsorbers.
A simplified model for estimating the fission product aerosol removal by containment sprays following a postulated LOCA was used in the analysis. The model for aerosol removal by sprays built into the RADTRAD code is the Powers model. The Powers model was derived by correlating the results of Monte Carlo uncertainty sampling analyses.assessing the uncertainties in aerosol properties, aerosol behavior, spray droplet behavior, and the initial and boundary conditions expected to be associated with a postulated LOCA in the containment.
Input parameters for the Powers model are presented in <Table 6.5-9>.
 
TR:e re111sval ssRstaRts fer tl:ie eTAR~IAYA    aRalysis are preseRteEI iR
<Tal3le 8,f3 11>. TR989 re111sval SSRStaRts,    a~~lisal3le  te setR: 9l9Hl9Rtal ieEiiRe aRe partisYlate, 111ay se  se111~areEI  ts a raR&sect;9 ef val.yes ef  ~.a/R:syr ts lQ,Q/R:eY:E ealeYlateEi fsE ele111eRtal ieeiRe sRly HitR: ePIRT, 6.5.2.4        Tests and Inspections The CSS spray nozzles will be verified unobstructed following maintenance which could result in nozzle blockage.            The test may be performed using an inspection of the nozzle or an air or smoke flow test. Further testing and inspection of the CSS is described in
<Section 6.2.2>.
6.5.2.5        Instrumentation Requirements Instrumentation requirements for the CSS are discussed in
<Section 7.3.1.1.4>.
6.5.2.6        Materials No spray additives are used in the CSS other than the pH buffering chemical (boron solution) from the standby liquid control system, which is injected into the reactor vessel and suppression pool following a design basis LOCA.
6.5.3        FISSION PRODUCT CONTROL SYSTEMS 6.5.3.1        Primary Containment The primary containment vessel is a hybrid pressure retaining structure composed of a steel cylinder and ellipsoidal dome secured to a steel lined reinforced concrete foundation mat.          The containment vessel is designed to contain radioactive material that might be released during an accident and to ensure leak tightness during normal operating and Revision 12 6.5-15                          January, 2003
 
6.5.4          ICE CONDENSER AS A FISSION PRODUCT CLEANUP SYSTEM This section is not applicable to PNPP.
6.
 
==5.5          REFERENCES==
FOR SECTION 6.5
: 1. Postma, A. K.; Sherry, R. R.; Tam, P. S.; "Technological Bases for Models of Spray Washout of Airborne Contaminants in Containment Vessels," <NUREG/CR-0009>, October 1978.
: 2. ANSI/ANS-56.3-1979, "American National Standard for PWR and BWR Containment Spray System Design Critecia."
: 3. Electrical Power Research Institute, "Generic Framework for Application of Revised Accident Source Terms to Operating Plants,"
TR-105909, Interim Report, November, 1995.
4r      Peles tar ApplieEi Test.=IRsle&sect;y,  IRe, 1 "eTAR~l.'\YA, A GeEie fer 1!:&#xa5;ah1atiR&sect; eevere AesiEieRt; Aeresel Qet.=lavier iR NHelear      P~~r  PlaRt; CeRt;aiRIReRt;s i *"* CeEie Qeseript;ieR aREi \laliEiat;ieR aREi Herifisat;ieR Repert," PeAT ClQl      Q~,  RevisieR 1 1  Fe~rYary ~~,    199~.
: 4. Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in LightWater-Cooled Nuclear Power Plants", Revision 2, March 1978.
: 5. NUREG-0800, "Standard Review Plan (SRP) *6.5.2, Containment Spray As A Fission Product Cleanup System," Revision 4, March 2007.
: 6. Calculation 3.2.15.16, "Design Basis LOCA Dose Evaluation Using Alternate Source Terms," Revision 0, October 2013.
Revision 12 6.5-24                        January, 2003
 
TABLE 6.5-1 (Continued)
Regulatory Position                    System Design Feature 6.b (Cont.)        conform to this position, with the exception that utilization of adsorbent samples removed from the subject bed may be used to refill the samplers. The new unused activated carbon used to replace a bed on failure to meet the applicable tests of Table 2 will meet the requirements of Table, 5-1 or ANSI N509-1980, which meets or exceeds the requirements of Table 5-1 of ANSI N509-19'76. Adsorbent samples will meet the requirements of the Ventilation Filter Test Program in lieu of the applicable tests of Table 2 of <Regulatory Guide 1.52>.
NOTE:
an 80 percent Ill As discussed in                            the design basis LOCA    ~
analysis) only credits        removal efficiency of elemental and organic iodines by the charcoal filters in the    CRERS. ~
1 The steam system piping break outside containment <Section 15.6.4>,
Control Rod Drop Accident <Section 15.4.9>, and the Fuel Handling      -
Accident <Section 15. 7.4> and <Section 15.7.6>, do nottake credit for the charcoal filters in the CRERS.
Revision 12 6.5-30                        January, 2003
 
TABLE 6.5-3 (Continued)
Regulat6ry Position                    System Design Feature 6.b (Cont.)          that utilization of adsorbent*samples removed from the subject bed may be used to refill the samplers. The new unused activated carbon used to replace a bed on failure to meet the applicable tests of Table 2 will meet the requirements of Table 5-1 of ANSI N509-1980, which meets or exceeds the requirements of Table 5-1 of ANSI N509-1976. Adsorbent samples will meet the requirements of the Ventilation Filter Test Program in lieu of the applicable tests of Table 2 of <Regulatory Guide 1.52>.
NOTE:                          Chapter 15, none of
                                                                ~
basis bQGA  (R.~eT a.Ral&#xa5;sis) Eiees R91' credit any removal of elemental and organic iodines by the charcoal filters in the AEGTS.
Revision 12 6.5-42                    January, 2003
 
TABLE 6.5-9 INPUT PARAMETERS FOR THE SPRAY REMOVAL ANALYSIS
                                                        .,1,.
3
                                                        ~r Total containment net free volume, ft                    1.16&sect; 4 x 10 6 Sprayed containrr{ent volume, F3                          481174 Unsprayed containment volume, F3                          684226 Mean spray fall height, ft Number of spray pumps operating                          1 Spray flow rate, gpm                                      5,250 Spray solution pH                                        7.0 Number sf Elrep si;2;9  ~reups Q, Spray Flux, cfm/ft2                                  0.0621 Alpha, unsprayed/sprayed volume                          1.422 Pct, uncertainty percentile                              10 Csemetris ~eaR Elrep sies fer spatial                      4 I !Hi l[ -i..g.:?
ElistribHtieR 1 om Cssmetrio msaR partials sies fer      iRosmiR~
asrssol, em.
* Ceemstris meaR staRdarEI ElsviatieR Ne Hall esREleRsatieR We seRElsRsatieR SR    *~ter Elreplets We oeRsiEleratieR ef partiele    Ry~reseepieity Revision 12 6.5-56                    January, 2003
 
DEPOSITION REMOVAL FACTORS TABLE 6. 5-11 ELEMENTAL IODINE FroHI t; Q t;o t; 09Q 6000Ad6            Q/ho1:1r FroHI t; 09Q to t        7.~rn 60GOAdS    QI lJ/hOl:i&#xa3; FroHI    t; 72Q to t n4 sooonds            4,J2/ho1:1r rroHI t; 924 t;o t; lJ;!,7 SOGOAdS          J ,QUho1:1r
                                    '"FroHI t; f&#xa3;OHI t; un nm t;o *-- 171Q  SOG9REiS t;o t; 1Q97 SOGOAdS 2 I a2/A01:1&#xa3; 14 I d/hO\,U Fr OHi t; 1Q97 to t 2Q7Q SOGOAdS              Q,  7 a/l:lmi.r FroHI t; 2Q7Q t;o t 21329 seooAds              a.Q7/ho1:1r Fr OHi t 2929 to t JHl SOGOAdS                J, Q4 /l:lowr FroHI t; ;nn to t 4QJQ sooonds                J, 28/l:lowr FroHI t 4QJQ to t aJJ9 sooonds                J, 22/AOl:H FroH1 t; aJJ9 to t 07Q2 SOGORQS                J, JQ/l:lo1:1r FroHI t; 07Q2 to t 7d77 SOOORQS                0. aa/l:lo1u
* FroHI t; 7J77 to t 770Q SOQORQS                J,JQ/R:owr FroHI t 779Q to t 11724 SOOORQS                  l,lQ/R:o1:1r FroHI t 11724 to t ;!,?HiQ SOQORQS                Q, 8Q/R:o1u; FroHI t; 17499 t;o t JQQ2J SOQOAdS                Q,27/ho1:1r FroHI t; JQQ2J to t; 4QQJ9 SOOORSS                Q,2J/howr.
FroHI t; 4QQJQ t;o t &9alJ SOQOAdS                Q,2Q/howr FroHI    t; a9alJ t;o t; Q7QQQ soeonas            Q, l Q,lhowr FroHI t Q7QQQ SOGOAQS to ORS                Q/howr j
RellllrVa1 Volume  Wall Area l'aat:or (ft 1 )    (ft3 )
OaE"'J Drywell            276,500    15,000          0.8'78 Revision 12 Sprayed containment      481,174    29,000          0.975 6.5-58                              January, 2003 unsprayed Containment      684,226    61,000          1.143
 
9.0      AUXILIARY SYSTEMS 9.1      FUEL STORAGE AND HANDLING 9.1.1        NEW FUEL STORAGE 9.1.1.1        Design Bases 9.1.1.1.1        Safety Design Bases - Structural Structural related safety design bases are as follows:
: a. The new fuel storage racks containing a full complement of fuel assemblies are designed to:      (1) withstand all credible static and dynamic loadings,  (2) prevent damage to the structure of the racks, and therefore the contained fuel and (3) minimize distortion of the racks arrangement <Table 3.9-3>.
: b. The racks are designed to protect the fuel assemblies from excessive physical damage which may cause the release of radioactive materials in excess of <10 CFR 20> and <10 CFR requirements under normal or abnormal conditions caused by impacting from either fuel assemblies, bundles or other equipment.
: c. The racks are designed  a~d  _constructed in accordance with the Quality Assurance requirements of <10 CFR 50, Appendix B>.
: d. The new fuel storage racks are categorized as Safety Class 2 and Seismic Category I.
: e. The design of the building containing the new fuel storage vault (new fuel storage facilit~~ conforms to the guidelines of
      <Regulatory Guide  1.13>~    Thus it ensures that any deleterious Revision 12 9 .1-1                    January, 2003
 
full core of fuel assemblies; it .is designed to withstand all cr~dible static and dynamic loadings, thereby preventing damage to the structure of the racks and the contained fuel, and minimizing distortion of the racks arrangement <Table 3.9-2>.
: b. The racks are designed to protect the fuel assemblies from excessive physical damage which may cause the release of radioactive materials in excess of <10 CFR 20> and <10 CFR requirements under normal or abnormal conditions caused by impacting from other fuel assemblies.
: c. The racks are constructed in accordance with the Quality Assurance requirements of <10 CFR 50, Appendix B>.
: d. The spent fuel storage racks are categorized as Safety Class 2 and Seismic Category I.
: e. The spent fuel storage facility is designed in accordance with General Design Criteria 2, 3 and 4, and <Regulatory Guide 1.13>,
    <Regulatory Guide 1.29>, <Regulatory Guide 1.102>, and <Regulatory Guide 1.117>. The design precludes any deleterious effects on spent fuel rack integrity due to natural phenomena such as earthquakes, tornadoes, hurricanes, missiles, and floods.
Compliance with <Regulatory Guide 1.13> is discussed in
    <Section 9.1.2.3.3> and <Section 1.8>.
9.1.2.1.2      Structural - PAR Racks Structural related safety design bases for Programmed and Remote Systems, Inc. (PAR) racks are as follows:
: a. The densified spent fuel storage racks are designed to withstand all credible static and dynamic loadings to prevent damage to the Revision 12 9.1-9                    January, 2003
 
shutdown margin, to ensure complete shut.down from the most reactive condition at any time in core life.
: c. The time required for actuation and effectiveness of the backup control is consistent with the nuclear reactivity rate of change predicted between rated operating and cold shutdown conditions. A fast scram of the reactor or operational control of fast reactivity transients is not specified to be accomplished by this system.
However, its performance also ensures compliance with criteria imposed for postulated anticipated transients without scram.
: d. Means are provided by which the functional performance capability of the backup control system components can be verified periodically under conditions approaching actual use requirements.
Demineralized water, rather than the actual neutron absorber solution, can be injected into the reactor to test the operation of all components of the redundant control system.
: e. The neutron absorber will be dispersed within the reactor core in sufficient quantity to provide a reasonable margin for leakage or imperfect mixing.
: f. The system is reliable to a degree consistent with its role as a backup safety system; the possibility of unintentional or accidental shutdown of the reactor by this system is minimized.
: g. For the ~OCA      analysis, the SLCS is used to maintain the suppression pool pH at 7 or above to minimize the conversion of cesium iodide to elemental iodine following a design basis LOCA
  <Section 15.6.5.5.1.8>.
Revision 12 9.3-24                  January, 2003
 
extremely low probability non-design basis postulated incidents.            The analyses given there demonstrate additional plant safety consideration far beyond conservative assumptions.
Due to the nature of the fission products that are predicted to be released 13y <HYRl!:C 14 89> (Re.,iseel J'.ssiEleR~ Se1,use TerHI) in the event of a LOCA, the SLC system has also assumed a post-LOCA function of providing a pH buffering solution for the reactor vessel and suppression pool water. This will help to retain the fission products in the water post-LOCA <Section 15.6.5.5.1.8>.
9.3.5.4      Inspection and Testing Requirements Operational testing of the SLC system is performed in at least two parts to avoid inadvertently injecting boron into the reactor.
With the valves to the reactor and from the storage tank closed and the valves to and from the test tank opened, demineralized water in the test tank can be recirculated by locally starting either pump from the MCC.
This test can be accomplished with the reactor operating without affecting the ability of the other pump to inject borated water.
During a refueling or maintenance outage, the injection portion of the system can be functionally tested by valving the suction line to the test tank and actuating the system from the control room.            System operation is indicated in the control room.
After functional tests, the injection valve shear plugs and explosive charges must be replaced and all the valves returned to the normal positions as indicated in <Figure 9.3-19>.
After closing a local locked-open valve to the reactor, leakage through the injection valves can be detected by opening valves at a test connection .in the line between the isolation check valves.            Position Revision 12 9.3-35                        January, 2003
 
9A.4.1.l      Unit 1 Reactor Building 9A.4.1.1.1      Fire Zone lRB-la 9A.4.1.1.l.1      Description Fire Zone lRB-la is shown on <Figure 9A-2>, <Figure 9A-5>,
<Figure 9A-10>, <Figure 9A-14>, <Figure 9A-18>, <Figure 9A-22>,
<Figure 9A-25>, <Figure 9A-27>, <Figure 9A-28>, and <Figure 9A-29>.
This zone, referred to as the annulus, is located between the shield building wall and the containment vessel wall. It serves as a secondary barrier for maintaining the radiation doses within the limits specified b'l <10 CFR ~>.
            ~
The outside wall and ceiling (dome) of this fire zone are constructed of reinforced concrete. The inside wall and ceiling (dome) are the steel containment vessel. The outer concrete wall has a 3 hour fire resistance rating. The wall provides separation of redundant trains of safe shutdown equipment. The floor is constructed of reinforced concrete. The annulus at 574'-10" has been filled with concrete from Elevation 574'-10" to Elevation 598'-4". Wall and ceiling penetrations have 3 hour fire rated seals. Access to this zone is through a Class A fire door from the auxiliary building.
The ventilation system for this fire zone operates to maintain a negative pressure in the annulus relative to the outside to minimize exfiltration and ground level release of airborne activity. This system consists of two 100 percent capacity charcoal filter trains with exhaust fans located in the intermediate building. Smoke detectors are located in the discharge ducts of each fan to actuate an alarm in the control room if smoke is detected. Duct penetrations through the shield building wall are provided with 3 hour rated fire dampers with 160&deg;F fusible links.
Revision 12 9A.4-3                    January, 2003
 
11.3      GASEOUS WASTE MANAGEMENT SYSTEMS 11.3.1      DESIGN BASES 11.3.1.1      Design Objective The objective of the gaseous waste management system is to process and control the release of gaseous radioactive effluents to the site environs to maintain as low as reasonably achievable, the exposure of persons in unrestricted areas to radioactive gaseous effluents to
<10 CFR 50, Appendix.I>. This is to be accomplished while maintaining occupational exposure as low as reasonably achievable and without limiting plant operation or availability.
11.3.1.2      Design Criteria The gaseous effluent treatment systems are designed to limit the dose to offsite persons from routine station releases to significantly less than the limits specified in <10 CFR 20> and to operate within the emission rate limits established in the Offsite Dose Calculation Manual.
In addition, the Offgas Treatment System limits the dose to offsite persons from a Control Rod Drop Accident <Section 15.4.9> ~
significantly less than the limits specified in <10 CFR +Q.Q.>.
As a design basis for this system, an annual average noble radiogas source term (based on 30 minute decay) of 100,000 &#xb5;Ci/sec of the "1971 Mixture" will be used.    <Table 11.3-la> indicates the design basis noble radiogas source terms referenced to 30 minute decay with the charcoal temperature at 0&deg;F.  <Table 11.3-lb>, <Table 11.3-lc>, and
<Table 11.3-ld> indicate source terms referenced to 30 minute decay with the charcoal temperature at temperatures, 20&deg;F, 40&deg;F and 70&deg;F, respectively.
Revision 12 11.3-1                    January, 2003
 
12.3.2      SHIELDING 12.3.2.1      Design Objectives The design objectives of the plant radiation shielding are:
: 1. To ensure that during normal operation, including anticipated operational occurrences, the radiation dose to plant personnel and authorized site visitors is as low as reasonably achievable and within the limits set forth in <10 CFR 20>.
: 2. To provide the necessary protection for plant operating personnel following a reactor accident to maintain habitability of the control room as specified in <10 CFR 9Q 1 Appendix      Criterion  l~.
: 3.                                                              the dose requirements of (for tRo dosi&sect;n easis iOCA analysis, tRo li9onsin&sect; easis  li~it  is
    ~a  ram TgQg) and to maintain exposures as low as reasonably achievable, a small fraction of the <10 CFR 20> limits during normal operation.
: 4. To protect certain components from excessive radiation damage or activation.
12.3.2.2      Design Description 12.3.2.2.1      Plant Shielding Description Detailed layout drawings showing all plant structures are shown in
<Figure 1. 2-3>, <Figure 1.2-4>, <Figure 1. 2-5>, <Figure 1.2-6>,
<Figure 1. 2-7>, <Figure 1.2-8>, <Figure 1. 2-9>, <Figure 1.2-10>,
Revision 12 12.3-8                    January, 2003
 
LIST OF TABLES (Continued)
Table                            Title 15.6-4      (Deleted)                                              15.6-50 15.6-5      (Deleted)                                              15.6-50 15.6-6      Sequence of Events for Steam Line Break Outside Containment                                    15.6-51 15.6-7      Steam Line Break Accident (Design Basis Analy-sis) Activity Release to Environment (Curies)          15.6-52 15.6-8      Steam Line Break Accident - Radiological Effects                                                15.6-53 Steam Line Break Accident - Parameters Tabu-lated for Postulated Accident Analyses                  15.6-54 15.6-10      Steam Line Break Accident (Realistic Analysis)
Activity Release to Environment (Curies)                15.6-56 15.6-11      Steam Line Break Accident - Radiological Effects                                                15.6-57 15.6-12a    Loss-of-Coolant Accident - Parameters and Assumptions Used in Radiological Consequence Calculations - Main Steam Isolation Valve Leakage Pathway                                        15.6-58 15.6-12b    Loss-of-Coolant Accident - Parameters and Assumptions Used in Radiological Consequence Calculations - Containment Leakage Pathway              15.6-59 15.6-12c    Loss-of-Coolant Accident - Parameters and Assumptions Used in Radiological Consequence Calculations - Engineered Safety Feature (ESF)
Leakage Pathway                                        15.6-60
: 15. 6-12d    IppJJt Parameters for    the Qrj'Jo'el l N3tpr3l
~          ~/ Sedjmeptati 0 n Apal)csi S                              15.6-61
~Elemental                Iodine and Partic1Jlate Removal Rates fgr  Qrywel 1 Natural  Sedimentation                    15. 6-62 15.6-13      Meteorological Data                                    15.6-63 15.6-14      Control Room Model                                      15.6-64 Revision 12 15-xiv                        January, 2003
 
"anticipated average radiation levels."        The consequences of very unlikely events (faulted events) are compared to the <10 CFR eese limit is ~a Eem  T~Q~). The analyses described in that the consequences for these two types of events are less severe the corresponding 10 CFR limits.
Unless otherwise identified, it is assumed that all equipment (safety grade or nonsafety grade) is available to mitigate the transients described and analyzed in this chapter.        However, only safety grade equipment is assumed to be used to mitigate accidents and safely shut dm*m the reactor* or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10CFR100 will be updated to <10 CFR 50.67>)
15.0.2    ANALYTICAL CATEGORIES Transient and accident events contained in this chapter are discussed in individual categories as required by <Regulatory Guide 1.70>.                The results of the analyses of these events are summarized in
<Table 15.0-2a> and <Table 15.0-2b> for events in the main text of
<Chapter 15>.  <Appendix 150>, <Appendix 15E> and <Appendix 15F> present summary tables for partial feedwater heating, MEOD and single loop operation respectively.    <Appendix 158>, Reload Safety Analysis presents these results for the events analyzed for each reload.              Each event is assigned to one of the following categories:
: a. Decrease in Reactor Coolant Temperature:
Reactor vessel water (moderator) temperature reduction results in an increase in core reactivity.        This could lead to fuel cladding damage.
Revision 12 15.0-3                              January, 2003
 
15.0.. 3.1.2          Unacceptable Results for Infrequent Incidents [Abnormal (Unexpected) Operational Transients]
The following are considered to be unacceptable safety results for infrequent incidents (abnormal operational transients):
: a. Release of radioactivity which results in dose consequences that exceed a small fraction of <10 CFR 100>.
: b. Fuel damage that would preclude resumption of normal operation after a normal restart.
: c. Generation of a condition that results in consequential loss of function of the reactor coolant system.
: d. Generation of a condition that results in a consequential loss of function of a necessary containment barrier.
15.0.3.1.3            Unacceptable Results for Limiting Faults [Design Basis (Postulated) Accidents]
The following are considered to be unacceptable safety results for limiting faults (design basis accidents):
: a. Radioactive material release which results in dose consequences that exceed the guideline values of <10                  t.l:le
: b. Failure of fuel cladding which would cause            in core geometry such that core cooling would be inhibited.
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
Revision 13 15.0-7            December, 2003
 
the relief mode, for certain specified safety/relief valves and a
*subsequent cycling of a single low set valve. The effect of the LLS design on reactor coolant pressure is demonstrated, in <Chapter 5>, on the MSIV closure event. This is considered bounding for all other pressurization events and, therefore, is not simulated in the analysis presented in this chapter.
A sensitivity study was  al~o  performed to support higher analytical limits for relief valve setpoints. The study shows an increase of 20 psi in the relief valve setpoint causes less than 20 psi increase in reactor peak pressures. However, these reactor peak pressures are still well below the ASHE code limit (1,375 psig). Also, the increase of 20 psi in the relief setpoints does not have any effect on the peak surface heat flux, since all safety/relief valves open after the occurrence of MCPR during transients. Therefore, the analytical limits for relief valve setpoints in the Technical Specifications were 20 psi higher than those listed in <Table 15.0-1>.
The analytical limits used for the relief valve setpoints for the current reload analysis are listed on <Table 158.15.0-1> for the pressurization transients and on <Table 158.5.2-1> for the overpressurization transients. The analytical values are the basis for the deviation of the Technical Specification value.
No changes to this page.
15.0.3.5      Radiological Consequences      Provided for context.
In this chapter, the consequences of radioactivity release during the three types of events:  incidents of moderate frequency (anticipated operational transients), infrequent incidents (abnoJ:mal operational transients) and limiting faults (design basis accidents) are conelclered.
For all events whose consequences are limiting a detailed quantitative evaluation is presented. For non-limiting events a qualitative evaluation is presented or results ara rafaranced from a more limiting or enveloping case or event.
Revision 12 15.0-16                    January, 2003
 
typically limiting faults (design basis accidents) two quantitative analyses
: a. The first is based on conservative assumptions for the purposes of worst case bounding of event consequences to determine the adequacy of the plant design to meet <10 CFR 100> guidelines .(fsr                ~Rs res~)  ,  This analysis is referred to as            he "design basis analysis."
: b. The second is based on realistic assumptions to reflect expected radiological consequences . . This analysi              is referred to as the "realistic analysis."          The "realistic an lysis" is not performed for the LOCA analysis, or the Fuel Handl' g Accident.
Results for both are shown to be within NRC gu'delines.
15  are determined either manually or by computer code.                                    with Offgas System Failure <Section 15.7.1.l> a e evaluated using GASPAR I I <NUREG/CR-4653>
(Reference 8).                            releases are ev luated with the TACT computer                        ce  2) (Reference 6).      Instantaneous or "puff" type releases are                                                those presented in
              <Regulatory Guide                                              >r aRe ~WYR~G  1499>. The General Electric NED0-31400 analysis (Referenc                  7) also is utilized in determining doses associated with a Control Ro                  Drop Accident
              <Section 15.4.9>.          Dose conversion factors and breathing rates are presented in <Table 15.0-4>.
or <1 OCFR 50.67> (future revisions to design basis analyses that compare consequences to 10CFR100 wlll be updated to <10 CFR 50.67>)
Revision 13 15.0-17                        December, 2003
: 3.  "General Electric Company Model for Loss-of-Coolant Analysis in Accordance with <10 CFR 50, Appendix K>," December 1975 (NED0-20566).
: 4. ASME Boiler and Pressure Vessel Code, Section III, Class 1, "Nuclear Power Plant Components," Article NB-7000, "Protection Against Overpressure."
: 5. General Electric Company "General Electric Standard Application for Reactor Fuel" including the "United States Supplement,"
NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved revision).
: 6. U.S. Nuclear Regulatory Corrunission Computer Code TACT 5, Computer Code for Calculating Radiological Consequences of Time Varying Radioactive Releases, <NUREG/CR-5106>, June 1988.
: 7. General Electric Company "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor" NED0-31400A, Oct. 1992.
: 8. U.S. NRC Computer Code GASPAR II, Computer Code to Perform Environmental Dose Analyses for Release of Radioactive Effluents.
    <NUREG/CR-4653>, March 1987.
9.
    ~
Fe~eral  GYidaRee ~e~ert We. 11,  "~i~itiR&sect;  ValYes ef  ~adieRYelide IRtalce aRd Air GeReeRtratieR aRd Qese Gei:i.uersieR Faeters fer
: 10. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31897P-A, Class III (Proprietary), May 1992.
Revision 14 15.0-24                      October, 2005
 
No changes proposed.
Provided for context.
: 11. Federal Guidance Report 11, "Licensing Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion", 2nd Printing, 1989.
: 12. CCC-652 Oak Ridge National Laboratory RSICC Computer Code Collection MACCS2, V.1.12 Code Package, 1997.
: 13. Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.
Revision 13.
15.0-24a                    December, 2003
 
TABLE 15.0-3
 
==SUMMARY==
OF ACCIDENTS Failed Fuel Pins GE          NRC Worst Calculated        Case Section                      Title              Value      Assumption
<Section 15.3.3>  Seizure of One Recirculation Pump            None
<Section 15.3.4>  Recirculation Pump Shaft Break                          Non~
<Section 15.4.9>  Control Rod Drop Accident      ~              770
<Section 15.6.2>  Instrument Line Break          None          None
<Section 15.6.4>  Steam System Pipe Break Outside Containment            None          None
<Section 15.6.5>  LOCA Within RCPB              None          100%
<Section 15.6.6>  Feedwater Line Break          None          None
<Section 15.7.1.1> Mairr Condenser Offgas Treatment System Failure      N/A            N/A
<Section 15.7.3>  Liquid Radwaste Tank Failure                        N/A            N/A
<Section 15.7.5>  Spent Fuel Cask Drop Accident                        None          None
<Section 15.7.6>  Fuel Handling Accident Inside Containment (GE12 and GE14 fuel w/triangular mast                            151
<Section 15.8>    ATWS                            SPECIAL EVENT Revision 13 15.0-36                    December, 2003
 
TABLE 15.0-4 DOSE CONVERSION FACTORS 111 Thyroid                      Whole Body Isotope                    (rem/Ci)                    0.25xMeV/dis I-131                      l.49E+6                    8. 72E-2 I-132                      5.35E+4                    5 .13E-1 I-133                      3.97E+5                    l.55E-l I-134                      2.54E+4                    5.32E-1 I-135                      l.24E+5                    4. 21E-l Kr-83m                                                  5.02E-6 Kr-85                                                  3. 72E-2 Kr-85m                                                  5.25E-4 Kr-87                                                  l.87E-l Kr-88                                                  4.64E-1 Kr-89                                                  5.25E-1 Xe-131m                                                2.92E-3 Xe-133m                                                8.00E-3 Xe-133                                                  9.33E-3 Xe-135m                                                9.92E-2 Xe-135                                                  5.72E-2 Xe-137                                                  4.53E-2 Xe-138                                                  2.81E-l Time Period (hr)
Breathing Rates
                                              ~~
Breathing Rate (m 131 /sec)
FHA r      I LOCA, CRDAI 0-8                                    r:i'Cl ~~E-4 121              3. SE-4 121 8-24                                  ~ - 1. 75E-4                  1. 8E-4 24-720                                              2.32E-4            2.3E-4 NOTES:
CRDA, MSLB 111        lowing dose conversion factors (DCF's) are used in the alternative The term analyses;                                                                      1989
          ~    - QCF's fer iRfia!atieR; EPA Federal Guidance Report 11 -
(Reference          11) and EPA Federal Guidance Report 12-1993 (Reference 13).
~FHA          - CEDE: EPA Federal Guidance Report 11 - 1989 (Reference 11)
ODE/EDE: MACCS2 computer code (Reference 12), which used Federal Guidance Report 12 - 1993 (Reference 13).
121 This breathing rate was used for the duration of the Control Room radiological consequence analyses.
Rev.i sion 13 15.0-37                        December, 2003
 
The BWR has precluded this event by incorporating into its design mechanical *equipment which restricts any movement of the control rod drive system assemblies. The control rod drive housing support assemblies are described in <Chapter 4>.
No changes proposed.
15.4.9      CONTROL ROD DROP ACCIDENT (CRDA)    Provided for context.
Certain limiting safety analyses are reperformed each operating cycle to determine and/or verify safety margins. The methods, input conditions, and results for the current cycle for the control rod drop accident are presented in <Appendix 158> - Reload Safety Analysis.
The NRC approved generic bounding Control Rod Drop Accident (CRDA) analysis for Banked Position Withdrawal Sequence (BPWS)  pl~nts  (such as PNPP) described in GESTAR (Reference 9) is applied and therefore, the CRDA is not reanalyzed on a reload - specific basis. As new fuel designs, methodologies or correlations are developed (e.g., GEMINI methods) the applicability of the generic analysis is reverified.      For the second cycle the CRDA was reverified due to GEMINI methods being used. The impact of GEMINI methods on the results of the generic analysis is negligible. Also, the effect of increasing core thermal power to 3,758 MWt on the generic CRDA analysis is negligible due to the considerable margin present in the generic analysis.
15.4.9.1      Identification of Causes and Frequency Classification 15.4.9.1.1      Identification of Causes The control rod drop accident is the result of a postulated event in which a high worth control rod, within the constraints of the banked position rod control and information system (RC&IS), drops from the fully inserted or intermediate position in the core. The high worth rod becomes decoupled from its drive mechanism. The mechanism is withdrawn but the decoupled control rod is assumed to be stuck inplace. At a Revision 12 15.4-23                    January, 2003
 
later moment, thecontrol rod suddenly falls free and drops to the control rod  driv~  position. This results in a localized power excursion.
A more detailed discussion is given in (Reference 3).
It is important to note that, because of Perry's rod pattern control system (RPCS) <Section 7.6.1.5>, the consequences of a rod drop become insignificant. The fission product release would be so small that no reactor isolation is expected to occur. This is based on analysis of similar control systems as described in (Reference 1) . Therefore, due to RPCS a control rod drop accident resulting in                        50.67 consequences even remotely approaching <10 CFR                      is not considered credible at Perry. The radiolo&sect;isal  sale~latioRs  **ish follo*
* are fffoseRted for informatien enly aRd are 13asoa on a hypothetisal desi&sect;n basis rod drop assidont    11itho~t RPCS.
15.4.9.1.2        Frequency Classification The CRDA is classified as a limiting fault because it is not expected to occur during the lifetime of the plant; but if postulated to occur, it has consequences that include the potential for the release of radioactive material from the fuel.
15.4.9.2        Sequence of Events and Systems Operation 15.4.9.2.1        Sequence of Events Before the control rod drop accident (CRDA) is possible, the sequence of events presented in <Table 15.4-9> must occur.      No operator actions are required to terminate this transient.
Revision 12 15.4-24                    January, 2003
 
No changes proposed.
Provided for context.
15.4.9.2.2      Systems Operation The unlikely set of circumstances, referenced above, makes possible the rapid removal of a. control rod. The dropping of the rod results in high.
reactivity in a small region of the core.      For large, loosely coupled cores, this would result in a highly peaked power distribution and subsequent operation of shutdown mechanisms.      Significant shifts in the spatial power generation would occur during the course of the excursion.
The rod control and information system (RC&IS) limits the worth of any control rod which could be dropped by regulating the withdrawal sequence. This system prevents the movement of an out-of-sequence rod in the 100 to 75 percent rod density range, and from the 75 percent iod density point to the preset power level.      The RC&IS will only allow bank position mode rod withdrawals or insertions.
The RC&IS uses redundant input to provide absolute assurance on control rod drive position. If either of the diverse inputs were to fail, the other would provide the necessary information.
The termination of this excursion is accomplished by automatic safety features or inherent shutdown mechanisms.      Therefore, no operator action during the excursion is required. Although other normal plant instrumentation and controls are assumed to function, no credit for their operation is taken in the analysis of this event.
15.4.9.2.3      Effect of Single Failures and Operator Errors Systems mitigating the consequences of this event are RC&IS and APRM scram. The RC&IS is designed as a redundant system and therefore provides single failure protection. The APRM scram system is also designed to single failure criteria.      Therefore, termination of this transient within the limiting results discussed below is assured.
Revision 12
: 15. 4-25.                  January, *2003
 
No changes proposed.
Provided for context.
No operator error (in addition to the one that initiates this event) can result in a more limiting case since the reactor protection system will automatically terminate the transient.
<Appendix 15A> provides a detailed discussion on this subject.
15.4.9.3      Core and System Performance 15.4.9.3.1.      Mathematical Model The analytical methods, assumptions and conditions for evaluating the control rod drop accident are described in detail in (Reference 3),
(Reference 4),  (Reference 5), and (Reference 11). They are considered to provide a realistic yet conservative assessment of the associated consequences. The bounding analyses are presented in (Reference 6).
Compliance checks are made to verify that the maximum rod worth does not exceed 1. percent  ~k, so that the maximum local core power increase cannot cause more than the number of fuel rod failures assumed in the following radiological evaluations.
15.4.9.3.2        Input Parameters and Initial Conditions The core at the time of rod drop is assumed to be at the point in cycle which results in the highest incremental rod worth, to contain no xenon, to be in a hot startup condition, and to have the control rods in sequence A at 50 percent rod density (Groups 1.-4 withdrawn) . Removing xenon, which has a high neutron absorption cross section, increases the fraction of neutrons absorbed, or worth of the control rods.      The 50 percent control rod density which nominally occurs at the hot startup condition, ensures that withdrawal of a rod results in the maximum increment of reactivity.
<Table 15.4-10> provides other input parameters and initial conditions for the CRDA analysis.
Revision 12 15.4-26                    January, 2003
 
Peak fuel enthalpy is the parameter used to determine the severity of a transient and the Control rod worth calculations were          onset offuel pin failure. As reference points, the following design and fuel failure criteria performed for a representative, high          have been established by General Electric:
energy, equilibrium GNF2 core design.        Enthalpy = 170 calories per gram, cladding failure threshold All control rod worths were less than the    Enthalpy= 280 calories per gram, specific energy design limit confirmation criterion of I percent ~k.      Enthalpy = 425 calories per gram, prompt fuel dispersal threshold 15.4.9.3.3          Results res1:1lts of Urn sompliaRoe olrnok oalo1:1lat;;ioR 1 as sl:ie11R iR tl:io A conservative number of pins that could reach the cladding failure threshold of 170 calories per gram is 15.4.9.4          Barrier Performance                    less than 1,200 for all plant operating conditions. The PNPP radiological evaluations are conservatively based on an a'\sumed failure of 1.376 fuel nins for G F2 fuel.
An evaluation of the barrier performance was not made for this accident since this is a highly localized event with no significant change in reactor temperature or pressure.
15.4.9.5          Radiological Consequences Two separate radiological analyses are provided for this accident:
: a.      The first analysis assumes the accident occurs at a low reactor power level with the mechanical vacuum pumps in operation or at any power level with a coincident loss of offsite power.                      This is referred to as Scenario 1.
: b.      The second analysis assumes the accident occurs at a higher reactor power level with the Steam Jet Air Ejectors in operation such that Revision 12 15.4-27                                January, 2003
 
the condenser gases are processed through the Offgas filtration system. This is referred to as Scenario 2.
Each analysis is based on conservative assumptions considered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet <10 CFR      ~>  guidelines and is referred to as the "Design Basis Analysis."        ~
~The exposures were calculated using the computer code RADTRAD (Reference 13). j A schematic of the leakage path for each analysis is shown in
<Figure 15.4-5>.
15.4.9.5.1          Scenario 1 Analysis The Scenario 1 analysis is based on the NRC's Standard Review Plan 15.4.9 (Reference 8) where a loss of offsite power (LOOP) occurs coincident with the CRDA.        The LOOP condition results in the automatic
~losure    of the MSIVs, thereby stopping the transport of the fission products to the condenser.        Radioactive release to the environment follows due to the condenser leakage.          Alternatively, if the mechanical vacuum pumps are in operation (i.e., low reactor power level) at the time of the CRDA and a LOOP does not occur, the high radiation condition will be detected by the MSLRMs causing an automatic isolation of the operating mechanical vacuum pump line.          Once again, the radioactive release to the environment will occur as a function of condenser 001<1pYtOE o*.ial1:1atioR aEo ElossEieoEI iR (ReforoRoo 7) , Specific parametric values used in the evaluation are presented in <Table 15.4-12>.
: 15. 4. 9. 5. 1.1        Fission Product Release from Fuel The failure of      +r+o+
                        ~ fuel pins  is used for this analysis .. The mass fraction of the fuel in the damaged rods which reaches or exceeds the initiation temperature of fuel melting (taken as 2,842&deg;C) is estimated to be 0.0077.
Revision 12 15.4-28                    January, 2003
 
                                                                              , and 25 percent of the alkali metals (Cs, Rb).
Fuel  reach~ng      melt c ~ans    is assumed    t~ r~lea~e noble gas *inventoryt::.i:: percent of the iodine inventor .
100 pe cent of the The remaining fuel in the damaged pins *is assumed to release 10 percent of both the noble gas and iodine inventories                                      .          *
                                                  ~.and 12 percent of the alkali metals.
A maximum equilibrium inventory of fission products in the core is based on 1,000 days of continuous operation at 3,833 MWt.              No delay time is considered between departure from the above power condition and the initiation of the accident.
15.4.9.5.1.2            Fission Product Transport to the Environment r
The transport pathway is shown in <Figure 15.4-5> and consists of carryover with steam to the turbine condenser and leakage from the condenser to the environment.        No credit is taken for decay during retention in the turbine building .
                                      ..-~~~~~~.......,....~~~~~~~~~~~-
                                      ' and 1 percent of the remaining radionuclides Of the activity released from the fuel, 100 percent of the noble gases
~    10 percent of the iodines are assumed to be carried to the                            ~
condenser.
Of the activity reaching the
                                      , and 1 percent of the particulate radionuclldes ondenser, 100 percent of the noble gases v*
~    10 percent of the iodines (due to partitioning and plateout) remain airborne.. The activity airborne in the condenser is assumed to leak directly to the environment at a rate of 1.0 percent Radioactive decay is accounted for during residence in the however, it is neglected after release to the environment.                        , for a period of 24 hours.
The activity airborne in the condenser is presented in <Table 15.4-13>.
15.4.9.5.1.3            Results The calculated exposures from the Scenario 1 design basis analysis are presented in <Table 15.4-14> and are well within the guidelines of
<10 CFR .,1,.Q.Q.>.
            -~                                                        Revision 12 15.4-29                          c1anuary, 2003
 
15.4.9.5.2      Scenario 2 Analysis models and The Scenario 2 analysis is based      n the NRC's Standard Review Plan 15.4.9 (Reference 8) with t e exception that a LOOP is not assumed to occur. The specific ~eaele, assumptions      aR9 tRo pre&sect;ram used      for~
oompytor evaluation are described in (Reference                Specific parametric values used in the evaluation are 15.4.9.5.2.1        Fission Product Release from Fuel The fission product release from fuel is the same as presented in
<Section 15.4.9.5.1.l> for Scenario 1.
15.4.9.5.2.2        Fission Product Transport to the Environment The transport pathway is shown in <Figure 15.4-5> and consists of carryover with steam to the turbine condenser and leakage from the condenser to the environment via the Offgas System.
                                    , and 1 percent of the remaining radionuclides Of the activity released from          fuel, 100 percent of the noble gases~
.a.R4 10 percent of the iodines        assumed to be carried to the condenser.
                                  , and 1 ercent of the particulate radionuclides Of the activity reaching the    ondenser, 100 percent of the noble gases~
.a.R4 10 percent of the iodines (9Yo to partitioRiR&sect; aRa platooyt) remain airborne. The activity airborne in the condenser is processed by the Offgas filtration system prior to release to the environment.
Radioactive decay is accounted for during residence in the condenser and the Offgas System, however, it is neglected after release to the                Only the noble gases are environment.
released from the offgas system.
The activity airborne in the condenser is presented in <Table 15.4-13>.
Revision 12
: 15. 4-30                          January, 2003
 
15.4.9.5.2.3        Results The calculated exposures _from the Scenario 2 design basis analysis are presented in <Table 15.4-16> and are well within the guidelines of
<10 CFR ~>.
          ~
15.4.10      REFERENCES FOR SECTION 15.4
: 1. Paone, C. J., "Banked Position Withdrawal Sequence," January 1977, (NED0-21231).
: 2. Woolley, J. A., "Three Dimensional Boiling Water Reactor Core Simulator," May 1976,    (NED0-20953).
: 3. R. C. Stirn, et al., "Rod Drop Accident Analysis for Large BWRs,"
March 1972, (NED0-10527).
: 4. R. C. Stirn, et al., "Rod Drop Accident Analysis for Large BWRs,"
July 1972,  (NED0-10527 Supplement 1).
: 5. R. c. Stirn, et al., "Rod Drop Accident Analysis for Large BWRs,"
January 1973, (NED0-10527 Supplement 2).
: 6.  "GE BWR Generic Reload Application for 8 x 8 Fuel" (NED0-20360 Supplement 3 to Revision 1).
        ~<<~:~::~
1
: 7. If              ** '" WilU*R1s, K,    w, Woltasl*u, "Aaalytied  N******
fer  Eval~atiR&sect;  tRe  Ra~iele&sect;isal As~ests    sf GeReral Elestrie QeiliR&sect; Water Reastsrs,  II HareR 1989,  (APl!:Q a7ae} I
: 8. USNRC Standard Review Plan, <NUREG-75/087>, Washington, D.C.,
Revision 1.
Revision 16 15.4-31                      October, 2009
: 9. General Electric Company "General Electric Standard Application for Reactor Fuel" including the -"United States Supplement,"
NEDE-24011-P-A and NEDE-24011-P-A-US (latest approved revision).
: 10. GESSAR .II BWR/6 238 Nuclear Island Design, 22A7007, Revision 21, Appendix 158 - BWR/6 Generic Rod Withdrawal Error Analysis (or GE report EAS 69-0687).
: 11. General Electric Company "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor" NED0-31400, May 1987.
: 12. GE Services Information Letter, SIL No. 517 Supplement 1, "Analysis Basis for Idle Recirculation Loop Startup".
S.L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport And Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998.
Revision 12 15.4-32                        January, 2003
 
TABLE 15.4-12 CONTROL ROD DROP ACCIDENT EVALUATION PARAMETERS
                                        .Scenario 1        Scenario 2 Assumptions        Assumptions I. Data and assumptions used to                                    1376 estimate radioactive source from postulated accidents.
A. Power level B. Burnup C. Fuel damaged D. Release of activity by nuclide                        <Table 15.4-13>
E. Iodine fractions, %
( 1)  Organic (2)  Elemental (3)  Particulate F. Reactor coolant activity before the accident.            N/A                N/A II. Data and assumptions used to estimate activity released.
A. Condenser leak rate (%/day)    1.0                N/A
: 8. Turbine building leak rate (%/day)                    N/A                N/A
: c. Valve closure time (sec)        N/A                5 D. Adsorption and filtration efficiencies (1)  Organic iodine            N/A                N/A (2)  Elemental iodine          N/A                N/A (3)  Particulate iodine        N/A                N/A
( 4)  Particulate fission products                  N/A                N/A E. Recirculation system parameters (1)  Flow rate                N/A                N/A (2)  Mixing efficiency        N/A                N/A (3)  Filter efficiency        N/A                N/A F. Containment spray parameters (flow rate, drop size, etc.)                N/A                N/A G. Containment volumes            N/A                N/A Revision 12 15.4-44                  January, 2003
 
TABLE 15.4-12 (Continued)
Scenario 1                Scenario 2 Assumptions                Assumptions H.  ~11 other pertinent data
        . and assumptions.                      None                      Holdup Time in Offgas Pretreat-ment System
                                                                            <Table 11.3-Sa>:
: a. Xe    54.2 days
: b. Kr    59.3 hours III. Dispersion Data A. Boundary and LPZ distances (m)                        863/4,002                  863/4,002 B. X/Q' s (sec/m 3 ) for time intervals of:
_(1)  0-2hr-SB/LPZ                    4.3E-4/4.8E-5              4.3E-4/4.8E-5 (2)  2-8 hr - LPZ                    4.8E-5                    4.BE-5 (3)  8-24 hr - LPZ                  3.3E-5                    3.3E-5 (4)  1-4 days - LPZ                  1. 4E-5                    1. 4E-5 (5)  4-30 days - LPZ                4.lE-6                    4.lE-6 Data of dose calculation          <Section 15.0.3.5>        <Section 15.0.3.5>
B. se conversion assumptions <Section 15.0.3.5>                  <Section 15.0.3.5>
: c. Pea activity concentrations <Table 15.4-13>                      <Table 15.4-13>
D.                                        <Table 15.4-14>            <Table 15.4-16>
C. Control Room XIQ (sec/m3)
NOTE:                                for time intervals of:
(1) 0-2 hours                      3.5E-4    3.5E-4 (2) 2-8 hours                      3.5E-4    3.5E-4 (3) 8-24 hours                      2.1E-4  2.1E-4 (4) 24-96 hours                      1.1E-4    1.1E-4 (5) 96-720 hours                    5.75E-5  5.75E-5 IV. Data and assumptions used to estimate dose to the Control Room Volume (ft3 )                            3.71E+5  3. 71E+5 Flow Rate - unfiltered inleakage (cfm)  6600      6600 Recirculation flow (cfm)                0          0 Recirculation filter efficiencies (%)
Particulate                          0          0 Elemental and organic Iodine        0          0 Revision 12 15.4-45                            January, 2003
 
TABLE 15.4-13 CONTROL ROD DROP ACCIDENT - SCENARIO 1 (DESIGN BASIS ANALYSIS)
ACTIVITY AIRBORNE IN CONDENSER (CURIES) U>
Activity Isotope                Curies
                                                                  ;i,' Ql!:*J J, Jl!:*J  "'      BR-82    3.0B+l J, Ql!:+J          BR-83    5.SB+2
: a. ee;+;;i          BR-84    9.4B+2
: 4. al!:*J          I-128    6.98+1 l-130    1.7B+2 J,;!1!:*4          l-131    4.6E+3
: i. el!:+;;i        l-U2      6.6B+3
                                                                  +.QI!:* 4 I-U3      9.3B+3 L+!!:*a I*U4      1.0B+4
                                                                  ~.JI!:*&sect; J, Q!!:*Hi          1*135    B.7B+3 CS-132    1.SB-2 Xe Blm                LJl!:+;J            CS-134    1.48+1 Xe l;il~m            &sect;. +/-1!:*4            CS-134M  3.2B+O Xe lJJ                J, ~!!:*&sect;            CS-136    4.3B+O Xe lJam              Bdil!:*4            CS-137    8.3B+O Xs lJa                J. Ql!:*4            CS-138    9.9B+l Xe 1J4                4.al!:*a RB-86    l.4B*l Xe lJQ                4. ~!!:*&sect; RB-88    3.6B+l NOTE:                                                                RB-89    4.7B+l KR-83M    5.7B+4 11
                    > GIH4 F1:1e+/- @ 3, 833 MWt.                                      KR-85    6.68+3 IGNF2 Fuel, 1350 EFPD  J                                                              KR-85M KR-87 KR-88 1.28+5 2.2B+S 3.2B+S XB-129M  3.SB+O XB-131M  5.3B+3 XB-133    9.3B+S XB-133M  3.0B+4 XB-135    3.4B+S XB-135M  1.9B+5 XB-138  7.9B+S Revision 12 15.4-46                    *January, 2003
 
TABLE 15.4-14
                  . CONTROL ROD DROP ACCIDENT - SCENARIO 1 (DESIGN BASIS ANALYSIS)
RADIOLOGICAL EFFECTS~                          Licensing Basis Limit (TEPE. rem) d~E~!~
Dose (rem)
Exclusion area                            &#xa3;:-:1E-11 (863 Meters)
Low population zone (4,002 Meters)
                                            ~
: 1. 491!; 2
~Control Room                              2.63E-1<  1>
5 NOTE:    Limiting case presumed no control room isolation or filtration.
(l)
Revision 12 15.4-47                              January, 2003
 
TABLE 15.4-16 CONTROL ROD DROP ACCIDENT - SCENARIO 2 (DESIGN BASIS ANALYSIS)
RADIOLOGICAL EFFECTS Licensing Basis
                                                            &#xa3;9~~:.1*
Limit (TEDE. rem)
Dose (rem)
Exclusion area (863 Meters)                                      ~
a.Q~  a Low population zone (4,002 Meters)
                                                            ~
w.Ml4
          ~Control Room                                    1.31E-612 >        5 NOTES:
111 Iodine releases are negligible because of their retention in the charcoal beds of the Offgas Treatment System as noted in NED0-31400.
:;:; Wl:iolc Elody Qose. at tRe :bl?:!: is Rot oaloHlatea iR tRe ~l~QO al4QQ            Jj\
Limiting case presumed no control room isolation or filtration.          , "Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Isolation Valve Closure Function of the Main Steam Line Radiation Monitor."
Revision 12 15.4-49                        January, 2003
 
15.6      DECREASE .IN REACTOR COOLANT INVENTORY 15.6.1        INADVERTENT SAFETY/RELIEF VALVE OPENING This event is discussed and analyzed in <Section 15.1.4>.
This page provided for 15.6.2        INSTRUMENT LINE PIPE BREAK information only.
This event involves the postulation of a small steam or liquid line pipe break inside containment.      In order to bound the event, it is assumed that a small instrument line, instantaneously and circumferentially, breaks at a location where it may not be isolated and where immediate detection is not automatic or apparent.
                                    "Steam System Piping Break Outside Containment" Obviously, this    vent is far less limiting than the postulated events in
                <Section 15.6.4>, <Section 15.6.5> and <Section 15.6.6 .      Accordingly, this accident was not reanalyzed f r the current reload as it has been determined to be less limiting an                the analy ed accidents described in the previously liste "Feedwater Line Break -
"Loss-of-Coolant Accidents - Inside Containment",
Outside Containment" This postulated event represents the envelope evaluation for sma        ine failure inside containment, relative to sensitivity to detection.
15.6.2.1        Identification of Causes and Frequency Classification 15.6.2.1.1        Identification of Causes There is no specific event or circumstance identified which results in the failure of an instrument line.      These lines are designed to high quality, engineering standards, seismic, and environmental requirements.
However, for the purpose of evaluating the consequences of a small line rupture, the failure of an instrument line is assumed to occur.
Revision 12 15.6-1                  January, 2003
 
This page provided for information only.
Therefore, all information concerning.ECCS models employed, input parameters and detailed results for a more limiting (steam line break) event may be found in <Section 6.3>.
15.6.2.3.3      Consideration of Uncertainties The approach toward conservatively analyzing this event is discussed in detail for a more limiting case in <Section 6.3>.
15.6.2.4      Barrier Performance 15.6.2.4.1      General The release of primary coolant through .the orificed instrument line could result in an increase in containment pressure and the potential for isolation of the normal ventilation system.
                                              ~ , if operating I The following assumptions and conditions are the basis for the mass loss during the 5 hour reactor shutdown period of this event:
: a. Shutdown and depressurization initiated at 10 minutes after break occurs.
: b. Normal depressurization and cooldown of reactor pressure vessel.
: c. Line contains a 1/4-inch diameter flow restricting orifice inside the drywell.
: d. Moody critical blowdown flow model. (Reference 1) is applicable and flow is critical at the orifice.
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This page provided for information only.
The total integrated mass of fluid released into the containment via the break during the blowdown is 25,000 pounds.        Of this total, 6,000 pounds flash to steam.      Release of this mass of coolant results in a containment pressure which is well below the design pressure.
15.6.2.4.2          Containment Effects Following the postulated failure of an instrument line in the containment, the containment pressure will rise due to the release of primary system fluid and will continue until the reactor is depressurized.      The containment pressure increase is evaluated based on the calculated mass release.      The calculation is based on the assumptions outlined above and includes the heat losses to the containment structures that will occur.
15.6.2.5        Radiological Consequences 15.6.2.5.1          Design Basis Analysis While the NRC has developed a standard review plan (Reference 2) for this event, a specific regulatory guide calculation method has not been For ~ reaso~ only ;;; 1.ealistic
                                ~          (!)    ~-
Realistic Analysis Analysis performed for initial plant licensing is analysis is based on a realistic but still conservative accident. The specific models, assumptions and the program used                    are described in <Section 15.0.3.5>.
Specific values of                          the evaluation are presented in
                <Table 15.6-2>.
----t , and as noted above, instrument line breaks              This Realistic Analysis was performed are considered to be bounded specifically by              for initial plant licensing, and is retained lhe steam line break <Section 15.6.4>, which              for historical information (not updated).
releases 141,687 pounds of fluid in a manner that is treated as direct to the environment.
Revision 12 15.6-6                            January, 2003
 
This page provided for information only.
15.6.2.5.2.3        Results The calculated inhalation doses for t"he realistic analysis are 2.03 rem at the exclusion area boundary (863 meters) and 0.249 rem at the low population zone boundary (4,002 meters).
                                                -___,These were calculated for.
initial plant licensing, and 15.6.3      STEAM GENERATOR TUBE FAILURE            are retained for historical information (not updated).
This section is not applicable to the BWR.
15.6.4      STEAM SYSTEM PIPING BREAK OUTSIDE CONTAINMENT The accident was not reanalyzed for the current reload since the original analysis is still applicable.
This event involves the postulation of a large steam line break outside containment. It is assumed that the largest steam line, instantaneously and circumferentially, breaks at a location downstream of the outboard containment isolation valve <Figure 15.6-1>. The plant is designed to immediately detect such an occurrence, initiate isolation of the broken line and actuate the necessary protective features. This postulated event represents the envelope evaluation of steam line failures outside containment.
15.6.4.l        Identification of Causes and Frequency Classification 15.6.4.1.1        Identification of Causes A main steam line break is postulated without the cause being identified. These lines are designed to high quality engineering codes and standards, and to restrictive seismic and environmental Revision 12 15.6-8                      January, 2003
 
No changes proposed.
Provided for context.
15.6.4.2.2      Systems Operation A postulated guillotine break of one of the four main steam lines outside the containment results in mass loss from both ends of the break. The flow from the upstream side is initially limited by the flow restrictor upstream of the inboard isolation valve.      Flow from the downstream side is initially limited by the total area of the flow restrictors in the three unbroken lines. Subsequent closure of the MSIVs further limits the flow when the valve area becomes less than the restrictor area and finally terminates the mass loss when full closure is reached.
A discussion of plant and reactor protection system action and ESF action is given in <Section 6.3>, <Section 7.2>, <Section 7.3>, and
<Section 7.6>.
15.6.4.2.3      The Effect of Single Failures and Operator Errors The effect of single failures has been considered in analyzing this event. The ECCS aspects are covered in <Section 6.3>.      The break detection and isolation considerations are defined in <Section 7.3> and
<Section 7.6>. All of the protective sequences for this event are capable of SACF and SOE accommodation and yet completion of the necessary safety act.ion. Refer to <Appendix 15A> for further details.
15.6.4.3      Core and System Performance Quantitative results (including mathematical models, input parameters and consideration of uncertainties) for this event are given in
<Section 6.3>. The temperature and pressure transients resulting as a consequence of this accident are insufficient to cause fuel damage.
Revision 12
: 15. 6-10                    January, 2003
 
No changes proposed.
I Provided for context.
15.6.4.3.1      Input Parameters and Initial Conditions Refer to <Section 6.3> for initial conditions.
15.6.4.3.2      Results There is no fuel damage as a consequence of this accident.
Refer to <Section 6.3> for ECCS analysis.
15.6.4.3.3      Consideration of Uncertainties
<Section 6.3> and <Section 7.3> contain discussions of the uncertainties associated with ECCS performance and the containment isolation systems, respectively.
15.6.4.4      Barrier Performance Since this break occurs outside the containment, barrier performance within the containment envelope is not applicable.
The following assumptions and conditions are used in determining the mass loss from the primary system from the inception of the break to full closure of the MSIVs:
: a. The reactor is operating at the power level associated with maximum mass release.
: b. Nuclear system pressure is 1,060 psia and remains constant during closure.
: c. An instantaneous circumferential break of the main steam line occurs.
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: 15. 6-11                    January, 2003
: d. Isolation valves start to close at 0.5 seconds on high flow signal and are fully closed at 5 .. ~ seconds.
The analysis conservatively assumes flow through the valves for 6.05 seconds.
: e. The Moody critical flow model (Reference 1) is applicable.
: f. Level rise time is conservatively assumed to be 1.0 second.
Mixture quality is conservatively taken to be a constant 7.0 (steam weight percentage) during mixture flow.
Initially only steam will issue from the broken end of the steam line.
The flow in each line is limited by critical flow at the limiter to a maximum of 200 percent of rated &#xa3;low for each line. Rapid depressurization of the RPV causes the water level to rise resulting in a steam-water mixture flowing from the break until the valves are closed. The total integrated mass leaving the RPV through the steam line break is 141,687 pounds of which 127,376 pounds is liquid and 14,311 pounds is steam.
15.6.4.5      Radiological Consequences Two separate radiological analyses are provided for this accident:
: a. The first is based on conservative assumptions considered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet <10 CFR                      This analysis is referred to as the "design basis                  50.67
: b. The second is based on assumptions considered to provide a realistic conservative estimate of the radiological consequences.
This analysis is referred to as the "realistic analysis."
~The exposures were calculated utilizing the computer code RADTRAD (Reference 21)      I A schematic of the release path is shown in <Figure 15.6-1>.
Revision 12 15.6-12                    January, 2003
 
15.6.4.5.1        Design Basis Analysis 1.183 The design basis analysis                        Standard Review Plan 15.6.4 and NRC <Regulatory Guide              The specific models, assumptions and the program used for computer evaluation are described in
<Section 15.0.3.5>. Specific values of parameters used in the evaluation are presented in <Table 15.6-9>.
15.6.4.5.1.1        Fission Product Release from Fuel There is no fuel damage as a result of this accident.          The only activity available for release from the break is that which is present in the reactor coolant and steam lines prior to the break.          This level of aotiYity is ooRsisteRt *iith aR offEJaS Foloaso Fato of        lQQ  l=!Ci f?OI' iodine concentration in the reactor coolant is thoR EjivoR by (,.Ci poF EjEaHI) I assumed to be 4.0 &#xb5;Ci per gram dose equivalent 1-131:
I-131              7.&sect;1!:    ~
9.02 E-1 I-132              Q. 71!:    1 9.02 E+O I-133              a .al!:    1        6.24 E+O I-134              1,ijl!; ~  Q        1.95 E+1 I-135              Q. 91!:    1        9.02 E+O QooaHso of its shoFt half lifo, W 1~ is Rot        ooRsi~oI'o~  iR tho aRalysis.
15.6.4.5.1.2        Fission Product Transport to the Environment The transport pathway is a direct unfiltere                      the environment.
The MSIV detection and closure time discharge of 14,311 pounds of steam and 127,376 pounds of liquid from the break. Assuming all the activity in this discharge becomes airborne, the release of activity to the environment is presented in
<Table 15.6-7>.
Revision 12 15.6-13                      January, 2003
 
15 .. 6. 4 . 5. 1 . 3  Results The calculated exposures for the design basis analysis are presented in
  <Table 15.6-8> and are a small fraction of the guidelines of
  <10 CFR .l.Jl!l.> .
.                  ~
15.6.4.5.2            Realistic Analysis The realistic dnalysis is based on a realistic but still conservative assessment of this accident.        The specific models, assumptions and the program used for computer evaluation are described in
  <Section 15.0.3.5>.        Specific values of parameters used in the evaluation are presented in <Table 15.6-9>.
15.6.4.5.2.1          Fission Product Release from Fuel There is no fuel rod damage as a consequence of this event, therefore, the only activity released to the environment is that associated with the steam and liquid discharged from the break.
15.6.4.5.2.2          Fission Product Transport to the Environment The activity released from the accident is a function of the coolant activity, valve closure time and mass of coolant released.        A portion of the released coolant exists as steam prior to the blowdown, and as such does not contain the same concentration per unit of mass as does the steam generated as a consequence of the blowdown.        Therefore, it is necessary to subtract the initial steam mass from the total mass released and assign to it only 2 percent of the iodine activity contained by an equivalent mass of primary coolant.
Revision 12 15.6-14                  January, 2003
 
The following assumptions are used in the calculation of the quantity and types of radioactive material released from the reactor coolant pressure boundary.
: a. The amount of coolant discharged is that calculated in the analysis of the nuclear system transient.
: b. The concentrations of biologically significant radionuclides contained in the primary coolant is &sect;iveR ay .&#xb5;Ci. per &sect;rtHR as
    ~easYremeRts    maele eR sYrreRt &sect;eRsratieR QWRs      sR~I tRe astivity ratie setHeeR tRs 1RaiR tYrBiRe seREleRsate aRel reaster seelaRt is SR c.
ts se Q,1 Ci per seseREl 1 aR YRYSYally Ri&sect;R Rermal eliseRar&sect;s rats.
    &sect;as astivity lea1.riR&sect; tRe reaster uessel at tRe time ef tl:le aeeieleRt. Tl:le rss~lt is tl:lat Q,4~ Ci ef Reels &sect;as aeti**ity leaves tl:ls rsaster vessel eliuiR&sect; easl:I seesREl tl:lat tl:le iselatieR valve is e~eRo Revision 12 15.6-15                        January, 2003
: d. Because of the short half-life of Nitrogen-16, the radiological effects from this isotope are of no major concern and are not considered in the analysis.
Based on the above considerations, the amount of activity which is available for atmospheric dispersion is presented in <Table 15.6-10>.
15.6.4.5.2.3          Results The calculated exposures for this event are presented in
<Table 15.6-11>.      As noted, these values are a small fraction of
<10 CFR .l..Oll.>.
            ~
15.6.5          LOSS-OF-COOLANT ACCIDENTS (RESULTING FROM SPECTRUM OF' POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY) - INSIDE CONTAINMENT This accident was evaluated as part of the analyses supporting PNPP operation in various operating modes and/or with equipment out-of-service results of which are presented in the following Chapter 15 appendices:
    <Appendix 15D>        Partial Feedwater Heating Operation Analysis
    <Appendix 15E>        Maximum Extended Operating Domain Analysis
    <Appendix 15F>        Recirculation System Single-Loop Operation Analysis This accident was re-analyzed for cycle 8 using the SAFER/GESTR-LOCA methodology.      Subsequent fuel cycles will be re-analyzed to ensure the new fuel types remain bounding with the initial (Cycle 8)
SAFER/GESTR-LOCA analysis.        The results for these cycles with respect to the loss-of-coolant accident are presented in <Appendix 15B> of this chapter.
Revision 12 15.6-16                  January, 2003
 
design limits. For details and results of the analyses, see
                <Section 3.8>, <Section*3.9> and <Section 6.2>.
Regulatory Guide 1.183 15.6.5.5        Radiological Consequences alternative Two separate radiological analyses are P. ovided for this
: a. The "design basis analysis"  i~ based on the        accident source term                                This conservative analysis is used for the purpose of determining adequacy of the plant design to
.--~~~~~~~~-.meet          the licensing basis limits for offsite consequences and
, control room dose due to radiation      control room consequences (25 rem 'l'EDE and 5 rem TEDE, shine, and            respectively). This analysis is referred to as the "design basis containment purge analysis."
isolation
: b.        ost-LOCA equipment qualification, vital area access, a-R-4 PASS access analyses are based on the "original licensing basis analysis". This "original licensing basis analysis" is based on the source terms and methodology of <Regulatory Guide 1.3>,
Revision 2 and <Regulatory Guide 1.7>, Revision 2 and SRP 15.6.5 (Reference 2).
                ~The exposures were calculated using the computer code RADTRAD (Reference 21 ). I 15.6.5.5.1        Design Basis Analysis
                <10 CFR 100> required, in support of the reactor siting, that a fission product release into containment be postulated and that offsite radiological consequences be evaluated against the guideline dose values specified in that regulation. The fission product releases into containment are used for evaluating the acceptability of both the plant site and the effectiveness of Engineered Safety Feature (ESF) components and systems. As discussed in <Section 15.6.5.1.1>, there are no realistic, identifiable events which would result in a pipe break inside the containment of the magnitude required to cause a loss-of-coolant accident coincident with safe shutdown earthquake plus SACF Revision 12 15.6-21                    January, 2003
 
radiological consequence requirements. In  dition, the analysis in <Section 6.3> demonstrates that even in such a                event, the event does    ot result in failed fuel.                              accident provides an    per limit estimate to the resultant ef                this category of pipe    reaks, it is evaluated without th                being identified. The analysis even assumes (without the              being identified) that ECCS water makeup does not reach the core                hours postaccident. This produces a source term comparable in                to the original licensing basis source term, but different in the              of the releases and the radionuclide composition (Referenc          ). The "original" licensing basis source term
                                                            <Regulatory Guide 1.3>, and
              <Regulatory Guide 1.7>        was taken from information published in 1962 by the U.S. l\tornic Energy Commission in Technical Information Document (TIO) 14844, '                  of Distance Factors for Power and Regulatory Guide 1.183 alternative The                                              which is s ill a very on the advances i    the understanding of the timing, magnitude                              of fis ion product releases from severe reactor acci
!July 2000 ~ehrnaq*      1 995 (Reference                                                research and experience that culminated in the development                                LOCA**
accident source term.      In addition, due to the that uses a spectrum of postaccident isotopes,                        analysis is evaluated using a Total Effective Dose Equivalent (TEDE) methodology.
The acceptance criteria for the LOCA is 25 rem TEDE, offsite, and 5 rem TEDE to the control room operators.        As a pilot application of the RAST there were PO    Reg11latorj' Guides  auajl3bl9  at the time the RAST methadal og)' !alas apprp1red bl' the NBC for  JJSA at pNpp  nt pNpp    the an5T is considered the new design basis analysis fpr the radiological cggse~1epces  gf a  TOCA Revision 12 15.6-22                      January, 2003
 
Thp RAST analysis was pursped jpitj3ll&#xa5; to support an increase in      the main steam ljpp leak rate to 250 scfh qpd t 0    pljminate the MSIV Teakage Captral System
        *.~
The ~        analysis is based on the following:        Regulatory Guide 1.183
* using a reactor accident source term developed from <NTJREG-1465>,
* relying on natural deposition of fission product aerosol in the drywell,
* relying on natural deposition of fission product aerosol in the main steam lines,
* controlling the pH of the water in the containment to prevent iodine re-evolution,
* operating the containment spray system for up to 24 hours
          <Section 6.5.2.3>,
* not crediting iodine removal by charcoal adsorbers in the Annulus Exhaust Gas Treatment System (AEGTS),
* delaying actuation of the control room emergency recirculation system for up to 30 minutes,
      --rre?teasi ng elemental and organic iodine removal efficiencies of~80 percent for tt control room emergency recirculation system charcoal adsorbers
~t~I.  *~ -fr"m  Qc; percent tO 50 percent,.
~~
* increasing the engineered safety feature system leakage outside bypass The BA&#xa3;.:I!. analysis considers the following four potential fission product release pathways following the design basis LOCA:
* main steam isolation valve leakage,
* containment leakage, Revision 12 15.6-23                  January, 2003
 
The most limiting LOGA with respect to offsite and control room radiological consequences is different than the most limiting LOGA for EGGS analysis. The most limiting event for radiological consequences is a guillotine pipe break In one of the four main steam lines upstream of the inboard MSIV, because this break minimizes the amount of steam line length available for particulate deposition.
* secondary containment bypass
* post-LOCA water leakage from                      safety features systems outside containment.
The fission product transport model used consequences is shown in <Figure 15.6-2>.
radiological consequence                                  in <Table 15.6-15>.
  <Table 15.6-12a>, <Table 15.6-12b>,              able 15.6-12c>, <Taele la.9      l~e>,
aRe <Tasle    1~.9  l~e>  summarize the input parameters analysis.
For the                    it is assumed that the fraction of core inventory given in Table .a..:.+ of (Reference      ~)  are released from an equilibrium core operating                              ~      MWt for  ~      days prior to the accident.                                    th  core activit    implies substantial fuel damage.                                      n is inconsi tent with operation 6.3>, it    s assumed ap licable for the Of this    elease, 100  ercent of the noble gases and five perc                  iodine a e gaseous.      he remainder of the in  articulate f rm as stated in also est blishes that the spectrum of as oppos d to the lim* ed isotopes of a 3833          1350 15.6.5.5.1.1      Main Steam Isolation Valve Leakage Pathway There are four main steam lines; each line has an inboard MSIV, an outboard MSIV, and a third isolation valve.            These valves isolate the reactor coolant system in the event of a break in a steam line outside the primary containment, a design basis LOCA, or other events requiring containment isolation.        These MSIVs along with the main steam lines, up
*. to and including the third isolation valve, are designed as Seismic Category I.
Revision 12 15.6-24                          January, 2003
 
The analysis conservatively assumes that the fission product leakage from the main steam lines is released directly into the environment.
The leakage past the MSIVs is conservatively assumed to begin immediately after the accident. In actuality, the three intact steam lines would contain trapped steam which would be relatively cooler and more dense as compared to the atmosphere in the reactor vessel upper head during the overheating of the core. This condition would greatly inhibit mixing between the activity released from the core and the steam leaking through the three intact steam lines and the three associated sets of MSIVs. However, for conservatism, all of the lines are assumed to be leaking contaminated drywell atmosphere.
Other significant conservatisms in the analysis of steam line transport include:
(1)  No  consideration gf rpdpced steam  ljpe mass leak rate With (2) extremely small MSIV leak paths due to particulate deposition at the entrance to or within the leak path as the gas flow accelerates to sonic or near-sonic conditions.
Two configurations were analyzed to cover all single-failure possibilities. In the first configuration (Configuration 1), the inboard MSIV on the affected line was assumed to fail open, and this line was assumed to leak at 100 scfh. The three intact lines were then assumed to leak at 100 scfh, 50 scfh, and 0 scfh to maximize flow rates through the lines, which in turn maximizes the activity release. At 20 minutes after the start of release the third safety-related and seismically-qualified isolation valves (just outboard of the outboard Revision 12 15.6-25                    January, 2003
 
MSIVs) were assumed to be manually closed in all four lines.        This configuration was evaluated to be less limiting than a second configuration (Configuration 2) in which all MSIVs successfully closed, but in which the third isolation valves remained open due either to operator error or a failure .of the common power supply.
In both cases (i.e., Configuration 1 and Configuration 2) particulate deposition is credited in all volumes of the steam line upstream of closed isolation valves. This is in accordance with Sectiori 5.2.3 of (Reference 18). For the more limiting Configuration 2 this means deposition is considered in the space between the closed MSIVs for the affected line and between the reactor vessel and the closed inboard MSIVs as well as between the closed MSIVs in the three intact lines.
with Configuration 1, the affected line is assumed to leak at 100 scfh and the three intact lines at 100, 50, and    o scfh. Therefore, leakage past the outboard MSIV is assumed to be released directly to the environrnen1 For the three intact steam lines, the space between the reactor vessel and the inboard MSIV is assumed to be well-mixed and the Polestar Appljprj Techpglog)1  Ipc  \\STARNArTA" computer code  1g p59d  ta calcplate the pffectiue fjltration pro,,ided by this portion    of the  i*ntact steam ljpes. For all four lines, the space between closed MSIVs is considered to exhibit plug flow as long as unequal cooling of the line does not create the potential for internal circulation (as compared to the magnitude of the plug flow velocity) . As the cooling of the line continues and this potential is approached, the effective length of the line between the MSIVs is assumed to be "shortened" so as to ignore the portion where circulation may be occurring and, therefore, to avoid any potential for overestimating the filtration effect of this portion of the steam line. A simple integrable model (similar to the DEPOSITION computer code, (Reference 16)) is used to treat the plug flow.          For both models, the input, time-dependent particulate flow rates and the
* particle size distributions are taken from the results of the "upstream" STARNAUA  analyse~.
(Reference 15)
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: 15. 6-26                    January, 2003
 
Elemental iodine retention efficiency is based on a comparison .of deposition and resuspension rates from (Reference 17) and is conservatively set at 50%. All of the "unfiltered" iodine is conservatively assumed to be released in organic form.            Additionally, the main steam line aerosol removal efficiency (the ability of the steam lines to retain aerosol fission products) was slightly reduced in the analysis. This aerosol removal efficiency is equivalent to an increase in aerosol penetration of 10 percent. This was done to further increase the dose from the main steam line pathway.
15.6.S.5.1.2    Fission Product Transport in Drywell The most limiting OBA, with respect to the offsite and control room radiological consequences, is considered a large-break LOCA as a result of a double guillotine pipe rupture in one of the four main steam lines upstream of the inboard MSIV. It is further conservatively assumed that all fission products are released directly to the drywell and leaked into the primary containment and into the main steam lines, bypassing the suppression pool. The analysis also assumes that at a point two hours after accident initiation (when the ECCS is assumed to be able to reach the core and reflood it) the fission products are homogeneously distributed between the drywell and the primary containment.            The objective of this well mixed approach is to achieve an appropriate balance for the design of drywell leakage mitigative devices such as the MSIVs as well as containment leakage mitigative features such as the---.
annulus exhaust gas treatment system.  ~R~e-fe_r_e_n_c_e~1-9---.          HEPA filters in the As characterized in in-vessel fission product releases terminate 2 hours after accident initiation. For the fission product releases to terminate, the reactor vessel would need to be reflooded. In lieu of evaluating all of the potential steaming rates due to various reflooding scenarios, the analysis assumes that a substantial amount of fission products will end up in the primary containment as well as in the drywell, and as such, Revision 12 15.6-27                            January, 2003
 
                                                              ~7ust I mitigative features such as the HEPA filters in the annulus  effl pent gas treatment system are designed to accommodate a significant portion of the source term. The 2-hour assumption for the homogeneous mixture of the source term between the drywell and the containment is used since it provides an appropriate balance, because the "worst 2 hoursn are considered for the EAB radiological dose results, as opposed to simply the first 2 hours as was done when the TIO source term was used.
The radiological consequences are dependent upon the drywell bypass leakage prior to the termination of fission product release at 2 hours.
Because of this sensitivity, the analysis uses a steaming rate of an intact core without relocation to the lower head region, on the order of 3,000 cfm. For the period prior to 2 hours, the analysis conservatively does not credit steaming due to relocation, cooling from alternative water sources, or the release of hydrogen gas, all of which would provide a higher steaming rate and remove more of the fission products from the drywell region.
15.6.5.5.1.3    Aerosol Deposition Within the Drywell Activity released to the drywell as a result of the design basis loss-of-coolant-accident is initially airborne and can be removed from the atmosphere in one of four ways:
(1) Convection from the drywell to the containment (2) Natural removal within the drywell (e.g., particulate sedimentation)
(3)  Leakage into the broken steam line and through the MSIVs (4)  Leakage back into the reactor vessel and through the MSIVs Revision 12 15.6-28                  January, 2003
 
Elemental iodine removal is credited in the drywall volume. Airborne elemental iodine is removed by deposition to the walls in the drywall. This process is driven by the temperature differences between the surfaces and the atmosphere. The calculated removal constants are applied until a decontamination factor (OF) of 200 has been obtained. Aerosol removal In the drywall is modeled using the Power's removal model as given in NUREG/CR-6189 (Reference 20). The lower bound decontamination coefficient associated with the 10th percentile uncertainty was used for conservatism.
The leakage cont rj h11t ion is small by design; and therefore, the two cipal mechanisms for depleti6              of activity in the drywell atmosphere by radioactive decay) is convection from the drywell to the w thin the drywell.
Following the fuel release phase              f the accident, the restoration of ECCS (thus arresting further core damage) would quench the core debris, and results in a rapid sweep-out              f the drywell into the containment as discussed in Section 5.2.3 of (Re For the design basis analysis, a              egotiated licensing basis was established for the transport of              ctivity between the containment and the drywell.          The negotiated basi        in effect mixes activity between the regions and does not consider a s                            of the activity after two hours.        The negotiated parame                    in <Table 15.6-12b>.
Natural      r9mgu3l    of  actjujt~c rjpe  to  ph~'_Sical    prgce5sp5  (j  p    other than h&#xa5;. radi  0 act i ''A deca)') can  he  assaci at Ad    lori th many  et feet s  i nc1 udi ng sedimentation          djffpsiophqresis      and    thermophqresjs        Only sedimentation processes (degcrjhed in Sectjop 5              2  3 and Appendix E of        (Referegce    18) are credited in this analysis              The Polestar Applied Technology "STP.QNAIJA"    rpmputer code    !Rpferepcp    _15)  js  used far  the  calcnl3tjap of natural partjcnlate          remoual  in the    dr1rwel l      The  key  iPp!lt 3SSJJmpti 0 PS are gj  3 rep  On    (Table  15 fi-12rj) and the remogal r3te5 ("l3mhdas"                3S calculated      b~1  STARNAllA)  are  shown  on  (Table    15 6-129)      These    remO"al Revision 12 15.6-29                              January, 2003
 
rates aro also assYmse to apply to olomsRtal ioeiRe (sos eeotioR                    9.~.d
          . sf (RsfereRG8 Hl)).        ~late  tlut tRs eTARNAYA      aRal~*sis QQRSiesrs    HOH  8Yt ef tRs ei:ry*1ell aRel seeiRleRtatiaR simYltaRsaYsly,          IR tl:~is  way t'Re remsval rates (IJRisR: iH1f3rave uitR iRsreasiR&sect; partisYlate ssRSsRtratisR) are Rst s',*erestimateei:. TR:e f:!artisYlats release frsRI t'Rs ei:ryusll (assesiale:se uile:A tl:ls eryHell: ts 98RtaiRRl8Rt 88RV88tisR eiSGYSS8Q aBs"e) ssssmss tl:ls iRpYt fer tl:le    eT~RNAY~  saloYlatieR fer le:R:o spraysei:    rs~ioR  of
            ~.44  ts aoosYRt for the faot tl:lat tl:lo sprayeei:      ro~isR  is aRly 41% of tRs
          . ooRtaiRHleRt free volYffis,      Were  a~aiR,  tl:le iRteRt is to eRSYFO tl:lat t'Re partioYlate ooRoeRtratioR (aRe tl:lorofore 1 t'Ro rato of            partio~lato romoval) iR tl:le sprayoei:    re~ioR  of tl:le ooRtaiRHIOAt is Rot OHorostiH1ateei:.
15.6.5.5.1.4      Containment Leakage Pathway The primary containment consists of a drywell, a wetwell, and supporting systems to limit fission product leakage during and following the postulated LOCA with isolation of the containment boundary penetrations.
The design basis leak rate of the primary containment is 0.2 volume percent per day.      The analysis                                  the design basis leak rate                        for the (30 days).      at 24 hours as permitted by                          reduces Ito 0 .69L&  I      ~        <Regulatory Guide 1.183>                  remaining The secondary containment (shield building) which surrounds the primary containment will collect and retain fission product leakage from the primary containment and will release fission products to the environment in a controlled manner through the AEGTS.              AEGTS will maintain the secondary containment pressure negative following a DBA by the time the gap release could migrate outside the containment structure.                  Therefore, if a short period of time exists post-LOCA when the annulus pressure is not negative, the dose calculations would not be affected.
Although the primary containment is enclosed by the secondary containment, there      ar*~ systems that penetrate both the primary Revision 12 15.6-30                          January, 2003
 
No changes to this page.
Provided for context.
containment and the shield building boundaries that could create potential pathways through which fission products in the primary contairunent could bypass the leakage collection and filtration systems associated with the shield building. The analysis conservatively assumes 10.08% of the primary containment leakage bypasses the secondary containment (the Technical Specifications limit bypass leakage to a lower limit).
The analysis assumes 89.92 percent of the primary containment leak rate goes into the secondary containment for its radiological consequence analysis. This leakage is collected in the shield building and processed through the AEGTS HEPA filters before being released into the environment. The remaining 10.08 percent of the primary containment leak rate is assumed to bypass the shield building and to be released directly to the envirorunent for the entire duration of the postulated LOCA.
15.6.5.5.1.5    Annulus Exhaust Gas Treatment System The AEGTS is an engineered safety features system and is designed to collect, process, and release the fission product leakage from the primary containment into the shield building. The AEGTS is a redundant
* system consisting of two 100 percent capacity subsystems. Each subsystem has a design capacity of 2000 cfm and consists of, among other things, a HEPA pre-filter, one 4-inch deep charcoal adsorber, and a HEPA post-filter. The system is designed to Seismic Category I standards and is located in a Seismic Category I structure.
The system is operated continuously during normal plant operation, and it maintains a slight negative pressure in the shield building.      The analysis assumes a 99 percent removal efficiency for fission products in aerosol form for HEPA filters. The analysis however does not consider any fission product removal by the charcoal adsorbers in the AEGTS.
Revision 12 15.6-31                  January, 2003
 
15.6.5.5.1.6                      Containment Spray The containment sprays are an engineered safety feature mode of RHR, designed to provide containment cooling, pressure reduction and fission product removal in the containment following a postulated LOCA.                                      The containment sprays consists of two redundant and independent loops.
Each loop has a design spray water flow capacity of 5250 gpm.                                      The system is designed to Seismic Category I standards and is located in a Seismic Category I structure.                        No chemical additives are used in the containment sprays, other than the pH buffering chemical (boron solution) from the existing Standby Liquid Control System <Section 3.4>
following a LOCA.
Tbp    ap3J yse5        35$!JWA    3  m1 x1pg  rate    gf  6 3  JJ05pr3)'Ad  vqJ !JIDAS* per  hqpr het'c 8 P 1  8      the    spra~ 1 ed    and  ppspra&#xa5;ed    port j ans  qf the  contai pment atmosphere by operation at the containment sprays                              Thjg mjyjng    rate jg hjgher than the    tie1o    tJJTQQJ'Ars      qf    the  JJpspraypd    regi  QD  per  bopr  Speci  fj ed i Q  the St anda*rd R        0 1ri Ala' pl an Section 6 5        2    This miyipg rate was accepted OP the h3sj9 that the pNpp calcplatiODS demonstrated 30 adequate mjvjpg fl Ola'  Tari 1 1 exist      het1&deg; 1een JJPspra~ 1 ed  and  spra~ 1 ed regj  ops b&#xa5; nat11ra 1 cop1rect; op To support the                              the containment sprays will be operated based upon plant emergency guidelines
    <Section monitor            Qthertarj  se                            sprays will automatically initiate 10 minutes following a LOCA signal if containment pressure exceeds the high* pressure setpoint ,
1 See <Section 6.5.2> for additional design basis of the containment sprays.
Otherwise, the sprays will be manually initiated post-accident within the 30-minute time assumed in the analysis, based on readings from the containment high range radiation monitor or other emergency operating procedure guidance.
Revision 12 15.6-32                              January, 2003
 
15.6.5.5.1.7      Post-LOCA Leakage Pathway from Engineered Safety Features Outside Containment
.Any leakage of water from ESF components located outside the primary containment releases fission products during the recirculation phase of long-term core cooling following a postulated LOCA.      The PNPP administrative controls limit this leakage to less than half of the value used in the radiological dose calculations.      The analysis is conservatively based on a leakage rate of 15 gallons per hour of ESF leakage for the entire duration of the accident (30 days).
Additionally, leakage from a gross failure of a passive component is assumed to occur at a rate of 50 gpm starting 24 hours into the accident and lasting for 30 minutes. Ten percent of iodine (all forms) contained in the leakage is assumed to be released directly to the environment and the pH of water leakage is assumed to be above 7.
15.6.5.5.1.8      Postaccident Containment Water Chemistry Management
                                                , and that iodine entering the containment from the reactor                during an accident would be composed of at least 95 percent                            no more th n 5 percent of iodine (I) and                          Once in t e containment, highly soluble cesium                      dissolve in    ater pools forming iodide (I-) in solution.              also consid rs the radiation-induced conversion of iodide in water into elemental io ine (1 2 ,) to be strongly dependent on the p~he      NIJREG identifies that without pH control, a large fraction of iodide dissolved in water pools in ionic form will be conveited to elemental iodine and will be released into the containment atmosphere if the pH is less than 7. On the other hand, if the pH is maintained above 7, very little (less than 1 percent) of the dissolved iodides will be converted to elemental iodine.
The Standby Liquid Control System (SLCS) is used for controlling and maintaining long-term suppression pool water pH levels to 7 or above Revision 12 15.6-33                    January, 2003
 
following the postulated OBA. The SLCS is a safety-related system and designed as a Seismic Category I system. Its primary function is as a reactivity control system to provide backup capability to be able to shut down the reactor if the normal control rods become inoperable. The system is manually initiated* from the main control room to pump a boron * *.
neutron absorber solution into the reactor.
8.357E+5 The SLCS contains a borax-boric acid solution. Such boron solutions act as pH buffers. Buffering will cause only a small de.rease in pH with addition of an acid so long as the buffer capacity i    not exceeded. The analysis used a containment water pool volume (which includes the suppression pool and reactor coolant inventory) of  1 3E+6 gallons and assumed all cesium iodide released into the drywell is directly deposited in the containment water pool. The analysis of pH levels in the containment water pool considered the following factors:
(1) cesium hydroxide formed from the fission products released from the core (basic-raises pH)
(2) the addition of the boron solution from SLCS (buffer)
(3) nitric acid produced by irradiation of water and air in the containment (acid-lowers pH)
.(4) hydrochloric acid generated from electrical cable degradation (acid-lowers pH)
The analyses demonstrate that with the amount of the boron solution provided in the containment, the pH of the postaccident water in the containment will remain abqve 7 for the duration of the postulated LOCA.
Revision 12 15.6-34                  January, 2003
 
15.6.5.5.1.9        Control Room Habitability Upon receipt of an ESF actuation system signal or high radiation, the control room Heating, Ventilation, and Air Conditioning (HVAC) system is designed to automatically switch to the emergency recirculation mode of operation (CRERS).      The analysis conservatively assumes a 30-minute delay in actuation of the CRERS.
The CRERS is a redundant system and each subsystem has a design flow capacity of 30,000 cfm.        The analysis uses a conservative recirculation flow rate of 27,000 cfm.        Each subsystem consists of, among other things, a High-Efficiency Particulate Air (HEPA) filter, charcoal adsorbers, and a HEPA post-filter.        The analysis also uses a conservative HEPA filter efficiency of .Qa. percent for aerosol particulate    and~percent        charcoal    f~~er    removal efficiency for iodine in element l and organic forms.
an 80                        99                  6000 During normal operation, the HVAC system is designed to                          the control room envelope with 45,000 cfm recirculation 6,000 cfm outside makeup air.        During an emergency, operates in the emergency recirculation mode,                                      is isolated and the control room envelope is not adjacent areas. To be conservative, the analysis uses                  cfm ~unfiltered I inleakage to the control room during the emergency recirculation mode fer tR:e eRtire aYratieR sf tR:e aeeiaeRt.          The major parameters and assumptions used in the analysis are lis ed in <Table 15.6-14>.
first 30 minutes, followed by 1,375 cfm unfiltered inleakage in the AR 9Jl9Rl~ti9R  Has EjraRte9 BY tRS    m~C  frSRI tae 99Rtre1 r99RI Q988 aeee~taRee  eriterieR sf <lQ CFR 5Q,      Af'l~eRaix    A>, CeReral QesiEJR Criteria 19 1 "Ceatrel: Ree111,"    TR:e eneRlf;!'eiea f18F111its Yse sf a 5 reRI TEQE aay f1art sf tR:e 8eay 1 " as eYrreRtly statea iR CQC 19 fer tR:e eeatrel:
Revision 12 15.6-35                          January, 2003
 
No changes to this page.
Provided for context.
15.6.5.5.1.10      Atmospheric Relative Concentrations at Control Room, Exclusion Area Boundary and Low Population Zone The atmospheric dispersion factors used in the-Control Room Habitability analysis were determined based on several analyses including NRC ARCON96 calculations in conjunction with the NUS Tracer Gas Study (Reference 8).
The NUS Tracer Gas Study was performed to characterize the atmospheric dispersion within the building complex at PNPP. Prior estimates of atmospheric relative dispersion (X/Q) values had been made for postulated releases to the control room using the Murphy-Campa methodology referenced in Standard Review Plan 6.4. The objective of conducting the tracer gas tests was to demonstrate more site specific/realistic control room air intake X/Q values.
The NRC reviewed and compared the results of the tracer gas study with calculations made using the ARCON96 methodology described in
<NUREG/CR-6631>, Revision 1, "Atmospheric Relative Concentrations in Building Wakes" (Reference 13).
For the postulated release point resulting in the largest X/Q values, the calculated X/O values from the tracer gas study were as much as 50 times lower than the original X/O values calculated using the Murphy-Campe methodology. For the same postulated release point, X/Q values using the ARCON96 methodology were approximately two times lower than the Murphy-Campe values.
The ARCON96 methodology assumes that the effluent travels the shortest distance possible between the postulated release point and the control room air intake. While the model calculates dispersion within building complexes, it is not intended to provide an exact model of postulated scenarios for complex site-specific flow paths around obstructions.
Meander and building effects are implicitly factored in, based on the field test studies used in the development of ARCON96.
Revision 12 15.6-36                  January, 2003
 
At PNPP, the effluent from a release postulated from the plant vent or containment building would need to disperse over or around an obstruction, down the side of a building and around a missile shield to be drawn into the control room air intake. For the limiting case, the X/Q values calculated from field tests performed by the licensee are about a factor of two or three lower for the control room air intake than for measurements made at the top of the building on which the intake is located. Thus, it was determined that results using ARCON96 would overestimate X/Q values for this scenario at PNPP.
The tracer gas field tests were conducted over a period of approximately one week in September 1985. While care was taken to assure that the tests were made under adequately limiting meteorological conditions, there is some likelihood that testing may not have captured the full range of poor dispersion conditions. Also, the field measurements may include some off centerline conditions, and due to solar heating in the building complex, better dispersion may have occurred during the tests than might occur at some other times of the year.
After discussing the tracer gas test limitations with the NRC, the X/Q value~n    <Table  15.6-l~ccepted        as the PNPP design basis.
The~ST    analysis  also~          the X/Q's for the EAB and LPZ. The analyses are based upon the <Regulatory Guide 1.145>. The dispersion factors used in the analyses for the offsite dose analysis are in
<Table 15.6-13>.
15.6.5.5.1.11      Results The calculated exposures for the design basis analysis are presented in
<Table 15.6-15> and are within the licensing basis limits of 25 rem TEDE (offsite) and 5 rem TEDE (control room).
Revision 12 15.6-37                    January, 2003
 
15.6.6      FEEDWATER LINE BREAK - OUTSIDE CONTAINMENT      . No changes proposed.
Provided for context.
This accident  wa~ not reanalyzed for the current reload since the original analysis is still applicable.
In order to evaluate large liquid process line pipe breaks. outside containment, the failure of a feedwater line is assumed to evaluate the response of the plant to this postulated event. The postulated break of the feedwater line, representing the largest liquid line outside containment, provides the envelope evaluation relative to this type of occurrence. The break is assumed to be instantaneous, circumferential and upstream of the outermost isolation valve.
A more limiting event from a core performance evaluation standpoint (feedwater line break inside containment) has been quantitatively analyzed in <Section 6.3>, MEmergency Core Cooling Systems."    Therefore, the following discussion provides only new information not presented in
<Section 6.3>. All other information is covered by cross-referencing to appropriate topics in <Section 6.3>.
15.6.6.1      Identification of Causes and Frequency Classification 15.6.6.1.1      Identification of Causes A feedwater line break is assumed without the cause being identified.
The subject pi~ing is designed to high quality, to strict engineering codes and.standards and to severe seismic environmental requirements.
15.6.6.1.2      Frequency Classification This event is categorized as a limiting fault.
Revision 12 15.6-38                  January, 2003
 
(future revisions to design basis analyses that compare consequences
                  *to 10 CFR 100 will be updated to <10 CFR 50.67>)
Taking no credit for holdup, decay or plateout during transport through the turbine building, the release of activity to the environment is presented in <Table 15.6-21>.      The total release is assumed to take place within 2 hours of the occurrence of the break.
15.6.6.5.2.3      Results The calculated exposures for the realistic analysis are presented i
<Table 15.6-22> and are a small fraction of <10 CFR 100> guidelines.
15.6.6.5.2.4      Sensitivity Analysis As described in <Section 6.2.4.2.2.1.a.l>, should a break occur in a feedwater line, the control closure check valves prevent significant loss of reactor coolant inventory and provide immediate isolation.                  A sensitivity analysis was performed to estimate the amount of leakage that would have to occur through the control closure check_ valves in order for the consequences of a feedwater line break outside containment event to exceed the consequences of the main steam line break outside containment. The results of the sensitivity analysis are that the leakage through the control closure check valves would have to exceed 200 gallons per minute for each feedwater line (400 gallons per minute total) for 2 hours in order for the consequences of the feedwater line break outside containment to exceed the consequences of the main steam line break outside containment <Table 15.6-8> and <Table 15.6-11>.                  The alternative non-Type C testing performed on these check valves per the Inservice Testing Program will verify proper closure of these yalves to prevent significant leakage of this order of magnitude.                The "exercise closed" (EC) testing will include a water leak rate test with an acceptance criterion of  ~200    gallons per minute per Feedwater penetration, when tested at    ~1.1    Pa), with no significant valve seat orifice defects [those large enough to result in leakage greater than the 200 gpm limit (400 gpm total) during a high pressure transient].
Revision 14 15.6-43                            October, 2005
: 2. USNRC Standard Review plan, <NUREG-75/087>.
: 3. Brutschy, F. J., G. R. Hills, N. R. Horton, A. J. Levine, "Behavior of Iodine in Reactor Water During Plant Shutdown and Startup,"
August 1972,    (NED0-105.85).
: 4.    (Deleted)
: 5.    (Deleted)
: 6.    (Deleted)
: 7.    (Deleted)
: 8. NUS Corporation "Results of the Atmospheric Tracer Study Within the Building Complex at the Perry Nuclear Power Plant," March, 1986, (NUS-4 7 92) .
: 9. J,
        ~ Soffer et al    "Accident Soprce    Terms for  I ight-Water  Npclear pglcrer plants" (N[JRFG 1465>    Eehrpar~r 1995
: 10. J. J. Carbajo, "MELCOR DBA LOCA Calculations," ORNL/NRC/LTR -
97/21, Oak Ridge National Laboratory, TN, January 1999.
: 11. . p ~ n  Powers "A Simplified Model Processes in Reactor containments " (NC!RfGICR-6189>          SANQ94-D4Q7 Sandia National I3hgratories      Alhnqperqpe  NM    ,Jpl~r 1996
: 12. D. A. Powers and S. B. Burson, "A Simplied Model of Aerosol Removal by Containment Sprays," <NUREG/CR-5966>, SAND92-2689, Sandia National Laboratories, Albuquerque, NM, June 1993.
Revision 12 15.6-44                        January, 2003
: 13. J. V. Ramsdell, Jr., and      c. A. Simmonen, "Atmospheric Relative Concentrations in Building Wakes," <NUREG/CR-6331>, PNNL-10521, Rev. 1~ Pacific Northwest National Laboratory, May 1997.
: 14. <SECY 96-242>, "Use of the <NUREG-1465> Source Term at Operating Reactors" dated 11/25/1996.
: 15. Polestar Applied Technology, Inc., "STARNAUA, A Code for Evaluating Severe Accident Aerosol Behavior in Nuclear Power Plant Containments:      A Code Description and Validation and Verification Report," PSAT ClOl.02, Revision 1, February 23, 1996.
: 16. Arnad, N. K., McFarland, A. R., Wong, F. S., Kocmoud, C. J.,
    "Deposition:    Software to Calculate Particle Penetration Through Aerosol Transport Systems," <NUREG/GR-0006>, April 1993.
: 17. Cline, J., "MSIV Leakage Iodine Transport Analysis," prepared for USNRC under contract NRC-03-87-029, Task Order 75, March 26, 1991.
: 18. Electric Power Research Institute, "Generic Framework for Application of Revised Accident Source Terms to Operating Plants,"
TR-105909, Interim Report, November 1995.
: 19. Nuclear Regulatory Commission, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at NuclearPower Reactors," Regulatory Guide 1.183, July 2000.
: 20. D. A. Powers, K. E. Washington, S. B. Burson, and J. L. Sprung, "A Simplified Model of Aerosol Removal by Natural Process in Reactor Containments," NUREG/CR-6189, SAND 94-0407, Sandia National Laboratories, Albuquerque, NM, 1995.
: 21. S. L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport And Removal and Dose Estimation," NUREG/CR-6604, USNRC, April 1998.
Revision 12 15.6-45                          January, 2003
 
This page provided for information only.
TABLE 15.6-2 INSTRUMENT LINE BREAK ACCIDENT - PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSES Design Basis Assumptions I. Data and assumptions used to estimate radioactive source from postulated accidents A. Power level                      None      N/A
: 8. Burn-up                          None      N/A
: c. Fuel damaged                      None      None D. Release of activity by nuclide                          None      <Table 15.6-3>
E. Iodine fractions, %
(1) Organic                    None      0 (2) Elemental                  None      100 (3) Particulate                None      0 F. Reactor coolant activity          None      <Section 15.6.2.5.2.1>
before the accident II. Data and assumptions used to estimate activity released A. Primary containment leak rate (%/day)                      None      N/A B. Secondary containment leak rate (%/day)                None      N/A
: c. Valve movement times              None      N/A D. Adsorption and filtration efficiencies (1) Organic iodine              None      90 (2) Elemental iodine            None      90 (3) Particulate iodine          None      90 (4) Particulate fission        None      N/A products E. Recirculation system parameters
( 1) Flow rate                  None      N/A (2) Mixing efficiency          None      N/A (3) Filter efficiency          None      N/A F. Containment spray parameters (flow rate,            None      N/A drop size, etc.)
G. Containment volumes              None      N/A H. All other pertinent data          None      None and assumptions Revision 12 15.6-47                  January, 2003
 
This page provided for information only.
TABLE 15.6-2 (Continued)
Design        Realistic Basis            Basis Assumptions      Assumptions Ill. Dispersion Data A.      Boundary and LPZ distance (m)                          None        863/ tJ002 B. X/Q' s (sec/m 3 ) for time intervals of
( 1) 0-2 hr - SB/LPZ                None        6.7E-4/8.2E-5 (2) 2-8 hr - LPZ                    None        8.2E-5
( 3) 8-24 hr - LPZ                  None        5.2E-5
( 4) 1-4 days - LPZ                  None        l.9E-5 (5) 4-30 days - LPZ                  None        4.7E-6 IV. Dose Data A.      Method of dose calculation            N/A          <Section 15.0.3.5>
B. Dose conversion assumptions            N/A          <Section 15.0.3.5>
: c.      Peak activity                          N/A          <Section 15.6.2.3>
concentrations in containment D. Doses                                  N/A          <Section 15.6.2.5>
    ~Note:
        <1> As  detailed in this table, no Design Basis radiological consequence analysis was performed for initial plant licensing. The Realistic Basis Assumptions presented herein are retained for historical information (not updated).
Revision 12 15.6-48                      January, 2003
 
This page provided for information only.
TABLE 15.6-3 INSTRUMENT LINE FAILURE (REALISTIC ANALYSIS)
ACTIVITY AIRBORNE IN INSTRUMENT LINE BREAK STRUCTURE (CURIES)
Isotope              Activity I-131                7. 46E + 1 I-132                1.15E  + 2 1-133                1.79E  + 2 I-134                1.97E  + 2 I-135                1.70E  + 2 (Historical)
Revision 12 15.6-49                  January, 2003
 
TABLE 15.6-7 STEAM LINE BREAK ACCIDENT (DESIGN BASIS ANALYSIS)
ACTIVITY RELEASE TO ENVIRONMENT (CURIES)
Isotope                      Activity I-131                        llQ"<J;'+(\~
5.259E +1 I-132                        6 2~E      ;I; 1 ' 5.277E +2 I-133                        3 54E      ;I;    3.637E +2 I-134                            J 6E    ;I;    1.148E +3 I-135                        5 72E      -t* 1  5.263E +2 Kr-83m                      7  59E' - 2        1.028E -1 Kr-85m                          331:' -        2.261E -1 Kr-85                        ~
I l 9E - 4        8.447E -4 Kr-87                        4 lH'      -      6.740E -1 Kr-88~                      4 :26E -            6.697E -1 Kr-89                            :z:ZE  ;I; a  4.288E +O Xe-131m                      4  :23E - 4        8.582E -4 Xe-133m                      6  3 31:" - 3      9.007E -3 Xe-133                          :z 8 E: -      2.581E -1 Xe-135m                      5  l 9E    -      9.974E -1 Xe-135~                      4  BOE -          7.042E-1 Xe-137          (1)          :2  34E -~ a        4.873E +O Xe-138                          VE ;I; a        2.840E +O Br-83                                          6.151E +1 Br-84                                          1.337E +2 Br-85<1>                                      9.826E +1
~Note:                                                          *
      <1l  Not included in dose calculation since no dose conversion factors for these isotopes are listed in FGR 11 or FGR 12.
Revision 12 15.6-52                              January, 2003
 
TABLE 15.6-8 (Iodine Concentration                                            uivalent I-131)
Licensing Basis Limit (TEPE. rem)
Exclusion area (863 Meters)
Low population zone (4,002 Meters)
~Control Room(1l                          6.22E-1                5 Normal HVAC flow wino control room isolation.
{l) ahaue radjological effects h3ve hepp updated to reflect the SCal Ad j pcre35e5 3SSOCi ated Jeri th P0 101 Ar (]prate to 3 758 MWt Revision 12 15.6-53                          January, 2003
 
TABLE 15.6-9 STEAM LINE BREAK ACCIDENT - PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSES Design        Realistic Basis            Basis Assumptions      Assumptions I. Data and assumptions used to estimate radioactiv~ source from postulated accidents A. Power level                      N/A            N/A B. Burn-up                          N/A            N/A
: c. Fuel damaged                      None            None D. Release of activity by            <Table          <Table nuclide                          15.6-7>        15.6-10>
E. Iodine fractions~
1::~ I
( 1) Organic                    .(L(            lLk?
(2) Elemental                  .l...          .l...
( 3) Particulate                .(L            .(L F. Reactor coolant activity          <Section        <Section before the accident              15.6.4.5.1.1>    15.6.4.5.2.1>
II. Data and assumptions used to estimate activity released A. Primary containment leak          N/A              N/A rate (%/day)
B. Secondary containment            N/A            N/A leak rate (%/day)
: c. Isolation valve closure          .5....
time (sec)
D. Adsorption and filtration                              6.05 efficiencies
( 1) Organic iodine            N/A              N/A (2) Elemental iodine            N/A              N/A
( 3) Particulate iodine        N/A              N/A
( 4) Particulate fission products                    N/A            N/A E. Recirculation system parameters
( 1) Flow rate                  N/A            N/A (2) Mixing efficiency            N/A            N/A
( 3) Filter efficiency          N/A            N/A F. Containment spray parameters (flow rate,            N/A            N/A drop size, etc.)
Revision 12 15.6-54                    January, 2003
 
TABLE 15.6-9 (Continued)
Design                  Realistic Basis                      Basis Assumptions              Assumptions G. Containment volumes                          N/A                    N/A H. All other pertinent                          None                    None data and assumptions III. Dispersion Data A. Boundary and        LPZ distance (m)                                  863/4002                863/4002
        ~    ;ii;;/Q' 9 Ee;r;: eeea;i, dess e8/bP6 ( ses/m 3 )
Data
(~.
Method of dose                                <Section                <Section calculation                                  15.0.3.5>              15.0.3.5>
.          Dose conversion                              <Section                <Section assumptions                                  15.0.3.5>              15.0.3.5>
Peak activity                                N/A                    N/A concentrations in D.                                                <Table                  <Table 15.6-8>                15.6-11>
B. X/Q for total dose (sec/m 3)
For time intervals of:        EAB      LPZ      CR (1) 0*8 hrs                    4.3E-4    4.BE*S    3.SE-4 (2) 8*24 hrs                  4.3E-4    3.3E*5    2.1E-4 (3) 24*96 hrs                  4.3E-4    1.4E*5    1.1E-4 (4) 96*720 hrs                4.3E-4    4.1E*6  5.75E*5 IV. Data and assumptions used to estimate dose to the Control Room Volume (ft3)                                4.205E+5 4.205E+5 Flow Rate* unfiltered inleakage (cfm)        6600    6600 Recirculation flow (cfm)                    0        0 Recirculation filter efficiencies (%)
Particulate                              0        0 Elemental and organic Iodine            0        0 Revision 12 15.6-55                              January, 2003
 
TABLE 15.6-10 STEAM LINE BREAK ACCIDENT (REALISTIC ANALYSIS)
ACTIVITY RELEASE TO ENVIRONMENT (CURIES)
Isotope                      Activity I-131                          29E    ;I;; a -E- 2.630E +O I-132                          6:ZE  ;I;; 1    2.639E +1 I-133                        9 65E    ;I;; a    1.819E +1 I-134                        3 09E    ;I;;      5.739E +1 I-135                          54E    ;I;; 1    2.632E +1 Kr-83m                      2  53E -      2    1.028E-1 Kr-85m                      4  '13E  -    2    2.261E-1 Kr-85                          :Z 3-E -    4    8.447E-4 Kr-87                            38E  -          6.740E-1 Kr-BB~                          42E    -          6.697E-1 Kr-89        (1)            5 89E    -          4.288E +O Xe-13lm                        '1 J E -    4    8.582E-4 Xe-133m                      2 JJE    -    3    9.007E-3 Xe-133                      5 92E    -    2    2.581E-1 Xe-135m                        :Z3E -            9.974E-1 Xe-135~                        60E -            7.042E-1 Xe-137                      :z BOE -            4.873E +O Xe-138                      5 A9E    -          2.840E +O Br-83                                          3.075E +O Br-84                                          6.687E +O Br-8511)                                      4.913E +O
~Note:
    <1l Not included in dose calculation since no dose conversion factors for these isotopes are listed in FGR 11 or FGR 12.
Revision 12 15.6-56                              January, 2003
 
TABLE 15.6-11 STEAM LINE BREAK ACCIDENT (Iodine Concentration in Coolant = 0.2 &#xb5;Ci/~e - eqriivalent I-131)
Licensing Basis Limit (TEDE. rem)
Dose (rem)
Exclusion area (863 Meters)                        4
                                      ~ ?BF - 2 Low population zone (4,002 Meters)
                                      ~
5 85E -  3
~Control Room(1 l                    3.11E-2          5 NOTE:
Normal HVAC flow wino control room isolation.
Revision 12 15.6-57                  January, 2003
 
                                            *TABLE 15.6-12a LOSS-OF-COOLANT ACCIDENT PARAMETERS AND ASSUMPTIONS USED IN RADIOLOGICAL CONSEQUENCE CALCULATIONS MAIN STEAM ISOLATION VALVE LEAKAGE PATHWAY
                                                                ~*                      '
Parameter                              ~        Value Reactor power
                                                                      .:il.5.8. MWt 5~
!Containment N:;::~~          ~~~~~=
3
: 2. 765 x 10        t  6 1.165 x 1          t Volume of one main steam line between MSIV's                                  146 ft 3
                  ''o} pmetrj C fl Dial rate dr~ne19l 1 to al 1 main steam ljnes        (t 0 tal leakage)      298 cfb      from t = a to t = 7484 seconds
                                                                      ':l47  cfb  from t    =  7484  secoorls to 30 d3)'s llolpmetric    flo1o1 rate  (mayjmpm)  one main steam line to environment                  1 91    cfm
          .____ ___.Volumetric flow rate, drywell to broken steam line 0 to 7484 seconds                      1.987 ft3/min 7484 seconds to 24 hours              1.647 ft3/min 24 hours to 30 days                    1.371 ft3/min Volumetric flow rate, drywell to intact steam lines 0 .to 7484 seconds                    2.98 ft3/min 7484 seconds to 24 hours              2.47 ft3/min 24 hours to 30 days                    2.056 ft3/min Revision 12 15.6-58                        January, 2003
 
TABLE 15.6-12b LOSS-OF-COOLANT ACCIDENT PARAMETERS AND ASSUMPTIONS USED IN RADIOLOGICAL CONSEQUENCE CALCULATIONS CONTAINMENT LEAKAGE PATHWAY Reactor power Parameter
                                                    ~                    Value
                                                                .3.1.5.8. MWt Volume of sprayed region                                        4.81 x 10 5 ft3 Volume of unsprayed region                                      6.84 x 10 5 ft 3 Flow rate from drywell to unsprayed region 0 - 2 hours 2 hours - 30 days Flow rate from unsprayed region to drywell 0 - 2 hours 2 hours - 30 days Flow rate between drywell and sprayed region Flow rate from sprayed region to unsprayed region Flow rate from unsprayed region to sprayed region 40 seconds - 30 da)/S Spray removal rate for particulate                    ~.9.ll percent uncertainty
                                                  ~      ..... distribution Spray fall height                                  ~.5.3.....2. ft Spray removal rate for elemental iodine                          <Table 6. 5-11>
(sprayed region only)
Containment leak rate to environment from 11nspra~ 1 ed  region a -    ao second*s 4Q    seconds - 30 days                                a  135 ft 3 1mjp Copt3jpm9pt      leak rate to 3QQ)]lps  from spra)'ed region a - 4Q seconds                                        a  ft 3 1mjp 40 seconds - 30 days                                  0 603 ft 3 /mip Cqpt3jpmppt leak rate to 3PP1Jlp5 from a - 40 seconds                                        a  ft3tmjp 40 seconds - 30 days                                  1  2Q5 ft3/mjp Annulus volume                                                  1. 96 x 10 5 ft 3 Flow rate from annulus to environment                            2000 ft 3 /min Annulus exhaust gas treatment system filter efficiency particulate                                            99 percent elemental and organic iodine                          0 Revision 12 15.6-59                            January, 2003
 
Insert for Table 15.6-12b Containment leak rate to environment from sprayed region Oto 40 sec            o.13s ft31min 40 sec to 24 homs    o.067 ft3Jmin*
24 hrs to 30 days    o.04623 ft31min Containment leak rate to environment from unsprayed region 0 - 40 seconds        t.046 ft31min 40 seconds - 24 hows  0.096 ft31min 24 hours - 30 days    0.0662 t\3/min Containment leak rate to annulus from unsprayed region 0 - 40 seconds        0 t\3/min 40 seconds - 24 homs  0.9S t\3/min 24 hours - 30 days    0.656 t\3/min Containment leak rate to annulus from sprayed region 0 - 40 seconds        oft3tmin 40 seconds - 24 homs  0.668 t\3tmin 24 hours - 30 days    o.461 ft31min
 
TABLE 15.6-12c LOSS-OF-COOLANT ACCIDENT PARAMETERS AND ASSUMPTIONS USED IN RADIOLOGICAL CONSEQUENCE CALCULATIONS ENGINEERED SAFETY FEATURE (ESF) LEAKAGE PATHWAY ECCS Leakage Model Parameter Regulatory Guide 1.183 Plant power Release fractions and timing                      for BWR in
                                    <          h> (gap and early in-vessel iodine releases only)
Release location                    Directly to suppression pool Suppression pool water volume      114, 379 fe ECCS leak rate 0 - 24 hours                  15 gph 24 - 24.5 hours              15 gph and 50 gpm for 30 minutes 24.5 hours - 30 days          15 gph Partition factor                    10 Revision 13 15.6-60                    December, 2003
 
      !Np[]T Qrywell    pres5pre  psig            o - 630 secccds*                  .l.:Z.-:i 630    J830 seccccis*            J7 3 to 5 3 J830 - :Z333 secccds*            .5.....L
:Z'B3 *- :za s.i secccds*        .l.:Z.-:i
:Z4 54 - 86'200 secocds*          J5 3 86'100    J 06 sa~ggdi*          J 5 3 tc 5 3 Dq0o el l temperat*pre 1                    f            o    1Q800 seconds*
1osoo - 21600 seconds*            320 216QO 8640Q      JQ 6 secgods*          250 to JSO fr9e uol11me    3                                                  765        lQ~
Qr31111el J              ft                                              2            y Qry1,rel l  sedimentation area ft 2 Geometric mean particle size For incomjog aerosol cm                                                        4    '1 x  1 q-5 Geometric mean standard deviation                                        .L.B.L CsQl;I*        3 6:z5
                                                            .c.s.r..:_    .4......5.l.
:r.e..:.....  .6....2.4.
                                                            ~            .5...12.
SJ:.Q.:_      A....i...
CeOi!*        .1....3....
I ai!o3*      6 5]
                                                            .Ru..:....    .J..2......6.
St cuC:t
* 5 6 No wall condensation No condensation on particles No consideration of particle hygroscopicity NQTE*
.u.J-nuerage decsit~r during gap release phase= approx 3 8 grams/cc
_A3cprage depsjt&#xa5; dprjpg fuel release phase - approx 4 J grams/cc Revision 12 15.6-61                              January, 2003
 
TABLE    1~ELETED:
        ..*ET EMENTZ\T IOQINE ANQ PART                    T RATES  E  R QRYWETI  NAT[JBAT    SEP'ENTATTON Fr pm t-Q to t-30 sec 0 nds        -                          a  /bqp  r
**Fram t-30 ta t-66 seconds            -                        a 084/hopr From t-66 to t-1867 s 0 c 0 nds        -                      a  184/hqpr from t-1867 to t-3203 seconds            -                    a  2 5 /hop r Er om t-3203 to      ,. -4 384 seconds                        a  35/hopr Er am t-4384 to t*-5862 seconds          -                    a  45/hopr Er om t-5862 to 1*-7333 seconds                                a  ~.4 /bqp r prqm t-7"333 to t-7484 seconds            -                    a  58/hopr ErQm t-7484 to t-9254 seconds                                  a  54 /hopr Ergm t-9254 to t-15881 sec nds      0
                                              -                  a  a 5 thapr fr gm t-1$881 to t-30669 seconds                              a  35/hqpr fr gm t-30669 to t 51639 seconds                              a  25/hqpr pr gm t-51639 to t JOQQQQ seconds              -              a  16/honr F;c:om t-JQQQQQ to ecd                                        O/hppr Revision _12 15.6-62                        January, 2003
 
TABLE 15.6-13 METEOROLOGICAL DATA Exclusion Area Boundary Time (hr)              X/Q (sec/m3 )
0-720                  4. 3xl0- 4 Low Population Zone Distance Time (hr)              X/Q ( sec/m 3 )
0-8                    4. 8xl0- 5 8-24                  3. 3xl0- 5 24-96                  1.4xl0- 5 96-720                4. lxl0- 6 Control Room Time (hr)              X/Q (sec/m;)
0-8 8-24 24-96 96-720 Revision 12 15.6-63                January, 2003
 
TABLE 15.6-14 CONTROL ROOM MODEL Parameter                  ~    Value Volume                                ~ x 10 5 ft 3
                                      ~ H 3/miR
                                      ~ ft\'miR ulation flow rate
        - 0.5 hour                    0 0 S hour - 30 days              2.7 x 10 4 ft 3 /min Recirc lation filter efficiencies ticulate ental and organic iodine  ~::
Flow rate - unfiltered inleakage 0 - 0.5 hour                    6000 ft 3 /min 0.5 hour - 30 days              1375 ft 3 /min Flow rate - exhaust 0 - 0.5 hour                    4800 ft 3 /min 0.5 hour - 30 days              1375 ft 3 /min Revision 12 15.6-64            January, 2003
 
TABLE 15.6-15 LOSS-OF-COOLANT ACCIDENT (DESIGN BASIS ANALYSJ:&L _
RADIOLOGICAL EFFECT~
Licensing Dose (Expressed              Basis Limit as TEDE, Rem)              (TEDE, Rem)
: 1. Offsite Doses Exclusion area (863 Meters)                                                  25 21~
Low population zone (4,002 Meters)                    J..a....l.                  25
: 2. Control Room Doses (0-30 days)
                                                    ~
A....:.L                    5 NOTE*
(l) Thp  2hq11e r3djgl9gjc3l  effects  h3ue  beep        !lpdated to refl99t the
    *scaled  jpcre35p5 associated    wjth  pqso1er  Uprate to 3 758 MWt Revision 12 15.6-65                            January, 2003
 
MSIVLeak DryweR
                "/
telnmenl fl:::.
Conttol Room ESF (Rev. 12  1/03)
G    PERRY NUCLEAR POWER PLANT Air Leaka9e Flowpath llost-LOCA Fi911ra_1s.&-2
 
Sprayed                                                                          Sprayed ID Sprayed ID AnmWa
                                                                        .                                        Environment (Bypass}
Unspl'BJ9d ID Spayed    f    !SpaJedlDUnsprayed AnmWa
                                                                                                            ~
AEGTS Drywell ID Unspriy8d                                                                    .
Unapl8J8d
                          .-                    UllSPllY8d ID AMulu8 UlllPIQld fD                                              Unsprayed ID
                                                                                              .-                        CR lntak8 Drywall Envlronrmnl (Bypasa)
EnvlnlnllB'lt - a
                                                                                                                                      . Contnll Room OryweO                                                                                                                -
                                                                                                                      - CREdlaust MSL2 Drywell1D MSL21D
                                                      .- MSl3                MSL31D
        ~                                MSl3                              EnvllOnmant MSl.1
                      .-                          MSL1ID Envlrannalt
                                                                                              .L MSL.1
 
15.7      RADIOACTIVE RELEASE FROM SUBSYSTEMS AND COMPONENTS 15.7.1      RADIOACTIVE GAS WASTE SYSTEM LEAK OR FAILURE  No changes proposed.
Provided for context.
This accident is not affected by the reload analysis.
The following radioactive gas waste system components are examined under several failure mode conditions:
: a. Main condenser gas treatment system failure.
: b. Malfunction of main turbine gland sealing system.
: c. Failure of air ejector lines.
15.7.1.1      Main Condenser Offgas Treatment System Failure 15.7.1.1.1      Identification of Causes and Frequency Classification Those events which could cause a gross failure in the offgas treatment system are:
: a. A seismic occurrence exceeding the seismic capabilities of the equipment.
: b. A hydrogen detonation which ruptures the system pressure boundary.
: c. A fire in the filter assemblies.
: d. Failure of adjacent equipment which could subsequently compromise offgas equipment.
The seismic event is considered to be the most probable and is the only conceivable event which could cause significant system damage.
Revision 12 15.7-1                  January, 2003
 
15.7.1.1.3            Core and System Performance The postulated failure results in a system isolation necessitating reactor shutdown. because of loss of vacuum in the main condenser.        This transient has been analyzed in <Section 15.2.5>.
15.7.1.1.4            Barrier Performance The postulated failure is the rupture of the offgas system pressure boundary.      No credit is taken for performance of secondary barriers, except to the extent inherent in the assumed equipment release fractions discussed in the next section.
15.7.1.1.5            Radiological Consequences Two separate radiological analyses are provided for the seismic accident:
: a. The first analysis is based on conservative assumptions.considered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet <10 CFR                        This analysis is referred to as the "design basis analysis."
: b. The second is based on assumptions realistic conservative estimate of                                  This analysis    i~  referred to as the "realistic anal sis."
Both are based on the following equipment charact ristics with respect to retention of radioactive solid                      s during normal operation of the offgas system.
: a. Offgas condenser - 100 percent retained a d continuously washed out with condensate.
(future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
Revision 12 15.7-4                January, 2003
 
(future. revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
release fract'ons given in <Table 15.7-4>.                The inventory of activities (d sign basis values) in each component, before the assumed failu e, is presented in <Table 15.7-3a>.                  The release rates due to  he continued operation of the SJAE are given in
        <Table 15.7-3
: 3. Results Dose consequ nces due to failure of the worst single component (the holdup  ipe) and assuming the SJAE continues to pump for 30 minutes a ter the break are presented in <Table 15.7-5>.
The doses ar  a small fraction of the limits specified in
        <10 CFR 100>.
: b. Realistic Analysis The realistic analysis is still a conservative assessment of this accident. The specific models, assumptions and the program used for computer evaluation are described in (Reference 5) .                  Specific values of parameters used in the evaluation are presented in
  <Table 15.7-6>.
: 1. Fission Product Release Assumptions The activity in the offgas system is based on normal operating conditions of 30 scfm air inleakage and 100,000 &#xb5;Ci/sec noble gas after 30 minutes delay.
The activity stored in the various components before failure is given in <Table 15.7-3a> (Normal/Realistic Values).
Revision 12 15.7-8                            January, 2003
 
Failure of the steam packing exhauster fan results in the escape of clean steam from the high pressure and low pressure shaft seals. The most undesirable result of operating in this condition is that some condensate from the escaping seal steam could leak into the lube oil system.
Excessive pressure in the sealing steam header as a result of a malfunction of the seal steam evaporator or the backup steam supply valve is prevented by a relief valve so that there is no detrimental effect on the operation of the shaft seals.                No changes proposed.
Provided for context.
15.7.1.3      Failure of Main Turbine Steam Air Ejector Lines 15.7.1.3.1      Identification of Causes and Frequency Classification Those events which could cause a failure in the main turbine steam air ejector system are:
: a. Failure of the steam line to the air ejectors.
: b. Failure of the air ejector suction line.
: c. Failure of the air ejector discharge line to the offgas system.
In each of these failures it is assumed that the worst case condition exists and that the failure is in a section of line conunon to both air ejectors so as to negate the use of the standby air ejector.
This event is categorized as a limiting fault.
15.7.1.3.2      Sequence of Events and Systems Operation
<Table 15.7-31> lists the sequence of events.
Revision 12 15.7-12                  January, 2003
 
(future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
: e. It is assumed that an equilibri m* coolant concentration consistent with an offgas release rate of 100,000 &#xb5;Ci/second after 30 minutes exists prior to the accident.
15.7.1.3.5.2                              Transport to the Environment The following assumptions                  in calculating the amount of activity released to the environs:
: a. It is conservatively a sumed that all of the iodine and noble gas activity released fr              break is instantaneously released to the environment via the              building ventilation system where it is treated by a serie        of roughing, HEPA and charcoal filters.
: b. The charcoal fil          efficiency is assumed to be 90 percent for the
: c.                        ptions relating to this event are tabulated in The activity released to the environment is presented i      <Table 15.7-9>.
15.7.1.3.5.3            Results The calculat d exposures for this analysis are presented in
                      <Table 15.7 10> and are a very small fraction of <10 CFR 100>
guidelines.
15.7.2          RADIOACTIVE LIQUID WASTE SYSTEM FAILURES (RELEASE TO ATMOSPHERE)
This accident is not affected by the reload analysis.
Revision 14 15.7-15                    October, 2005
: b. Safety and Power Generation Aspects Matters identified with "safety".classification are governed by regulatory requirements.            Safety function& include:
: 1. The accommodation of abnormal operational transients and postulated design basis accidents.
: 2. The maintenance of containment integrity.
: 3. The assurance of ECCS.
: 4. The continuance of reactor coolant pressure boundary (RCPB) integrity.
Safety classified aspects are related to <10 CFR 100> dose limits, infrequent and low probability occurrences, SACF crit ria, worst case operating conditions and initial assumptions, a tomatic (10 minute) corrective action, significant unaccepta le dose and environmental effects, and the involvement of other              oincident (mechanistic or nonmechanistic) plant and environme              al situations.
Power generation classified considerations are rela ed to continued plant power generation operation, equipment operat' nal matters, component availability aspects, and to long term o fsite public effects.
Matters identified with "power generation" classi ication are also covered by regulatory guidelines.              Power generati n functions include:
: 1. The accommodation of planned operations and anticipated operational transients.
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)
Revision 12 lSA.2-3                      January, 2003
: d. Cohs~rvative      Analysis - Margins The unacceptable consequences established in this appendix relative to the public health and safety aspects are in themselves in strict and conservative conformance to regulatory requirements.
Restrictive Operations on hypothetical limits established by further operational limits (e.g., setpoint margins) leads to disrespect for true safety aspects.
: e. Safety Function Definition First, the essential protective sequences shown for an event in this appendix list the minimum structures and systems required to be available to satisfy the SACF or SOE evaluation aspects of the event. Other protective "success paths" exist in some cases than are shown with the event.
Second, not all the events involve the same natural, environmental or plant conditional assumptions.                For example LOCA and SSE are associated with Event 39.            In Event 35, Control Rod Drop Accident CRDA is not assumed to be associated with any SSE or OBE occurrence.      Therefore, seismic safety function requirements are not considered for Event 35.            Some of the safety function equipment associated with the Event 35 protective sequence are also capable of handling more limiting events, such as Event 39; Third, containment may be a safety function for some event (when uncontained radiological release would be unacceptable) but for other events it may not be applicable (e.g., during refueling).
The requirement to maintain the containment in postaccident recovery is only needed to limit doses to less than <10 CFR effsite 9ese limie ie          ~&sect;  rem  T~Q~)  .
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences to 10 CFR 100 will be updated to <10 CFR 50.67>)                Revision 12 lSA.2-5                      January, 2003
 
TABLE lSA.2-4 UNACCEPTABLE CONSEQUENCES CRITERIA PLANT EVENT CATEGORY:          DESIGN BASIS ACCIDENTS Unacceptable Consequences 4-1          Radioactive material release exceeding the guideline values of
                    <10 CFR                      altorRativo soHroo  tor~ LQCA aRalysis, Failure of                              a result of exceeding mechanical 4-3          Nuclear system stresse          exceeding that allowed for accidents by applicable industry 4-4          Containment stresses exce                that allowed for accidents by applicable industry codes                containment is required.
4-5        Overexposure to radiation of                  main control room personnel.
or <10 CFR 50.67> (future revisions to design basis analyses that compare consequences.to 10 CFR 100 will be updated to
<10 CFR 50.67>)
NOTE:
01 Failure of the fuel barrier includes fuel cladding fragmentation (loss-of-coolant accident) and excessive fuel enthalpy (control rod drop accident).
Revision 13 lSA.2-16                    December, 2003
 
Related Unacceptable                                50.67 -
Consequences Safety Action          Criteria Establish Reactor        4-1            To limit radiol gical effects to Containment                              not exceed th guideline values of <10 CFR    >.
Containment              4-4            To prevent excessive pressure in Cooling                                  the containment when containment is required.
Stop Rod Ejection        4-2            To prevent fuel cladding failure.
Restrict Loss of          4-2            To prevent fuel cladding Reactor Coolant                          failure.
(passive)
Control Room              4-5            To* prevent overexposure to Environmental                            radiation of plant personnel in Control                                  the control room.
Limit Reactivity          4-2            To prevent fuel cladding failure Insertion Rate            4-3            and to prevent excessive nuclear (passive)                                system pressure.
NOTE:
11
  > Failure of the fuel barrier includes fuel cladding fragmentation (loss-of-coolant accident) and excessive fuel enthalpy (control rod drop accident).
15A.6.5.3        Event Definition and Operational Safety Evaluations
: a. Event 35 - Control Rod Drop Accident (CRDA)
The control rod drop accident (CRDA) results from an assumed failure of the control rod-to-drive mechanism coupling after the control rod (very reactive rod) becomes stuck in its fully inserted position. It is assumed that the control rod drive is then fully Revision 12 15A.6-36                  January, 2003
 
withdrawn before the stuck.rod falls out of the core. The control rod velocity limiter, an engineered safeguard, limits the control rod drop velocity. The resultant radioactive material release is maintained far below the guideline values of <10 CFR    ~*      _
        .                    .                                ~
The control rod drop  ~ccident  is applicable only in operating State D. The control rod drop accident cannot occur in State B because rod coupling integrity is checked on each rod to be withdrawn if more than one rod is to be withdrawn. No safety actions are required in States A or C where the plant is in a shutdown state by more than the reactivity worth of one rod prior to the accident.
  <Figure lSA.6-35> presents the different protection sequences for the control rod drop accident. As shown in <Figure 15A.6-35>, the reactor is automatically scrammed and isolated. For all design basis cases, the neutron monitoring, reactor protection and control rod drive systems will provide a scram from high neutron flux.
After the reactor has been scrammed, core cooling is accomplished by either the RCIC or the HPCS or the normal feedwater system.
: b. Event 36 - Fuel Handling Accident Outside Containment Because a fuel-handling accident can potentially occur any time when fuel assemblies are beihg manipulated in the fuel handling building, this accident is considered in all operating states.
Considerations include mechanical fuel damage caused by drop impact and a subsequent release of fission products. The protection sequences pertinent to this accident are shown in
  <Figure 15A.6-36>.
Revision 13 lSA.6-37                  December, 2003
 
FIGURE LIST FOR ALL CHAPTERS
~~~~~F_i_*g_u_r_e~~~~~~-D_r_a_w_i_*n_g~-#~~~~~~~~~-T_i_*t_l_e~o~f~F-i_g_u_r_e~~~~~~~~I Figure 6.4-3                              Location of Panels and Fire Extinguishing Equipment Figure 6.4-4 (1) 101=1 ETEOI  ..
control Room Dose at 30 0 ays llnf'i ,* +-..., ... ...,,-l Tn-T "'" i,,,, _ _ -"  ~nn  ,..,,_
Final Figure 6.4-4 (2) lnELETEDI    . CggtcgJ Bggm Ogsa at 30
                                          ~--
nf'i l +-~.---'          Tn_T ~~*],~~~      ~+
cia~s 1 0()  ,..,,_Eina 1
Figure 6.5-1            D-912-605        Annulus Exhaust Gas Treatment System Figure 6.5-2 ( 1)                        Annulus Exhaust Gas Treatment System Distribution Ductwork Figure 6.5-2  (2)                        Annulus Exhaust Gas Treatment System Distribution Ductwork
*Figure 6.5-3            D-302-661        Containment Spray System Figure 6.5-4                              DELETED Figure 6.5-5                              DELETED Figure 6.7-1 ( 1)                          DELETED F.igure 6.7-1 (2)                          DELETED Figure 6.7-2 ( 1)                          DELETED Figure 6.7-2 (2)                          DELETED Figure 6.7-2 (3)                          DELETED Figure 6.7-3                              DELETED Figure 6.7-4                              DELETED Figure 6.8-1            D-302-271          Safety-Related Instrument Air System Figure 6.9-1            D-302-971          Feedwater Leakage Control System Figure 6.9-1 (2)                          DELETED Cha2ter 7 Figure 7.2-1 ( 1)      D-808-302(1)      Reactor Protection System Figure 7.2-1 (2)        D-808-302(2)      Reactor Protection System Figure 7.2-1 (3)        D-808-302(3)      Reactor Protection System Figure 7.2-1 (4)        D-808-302(4)      Reactor Protection System Instrumentation and Electrical Diagram Figure 7.2-2                              DELETED Figure 7.3-1 ( 1)      D-808-311          High Pressure Core Spray System Figure 7.3-1 (2)        D-808-311          High Pressure Core Spray System Figure 7.3-1 (3)        D-808-311          High Pressure Core Spray System Figure 7.3-2                              DELETED Figure 7.3-3 ( 1)      D-808-303(1)      Nuclear Boiler System Figure 7.3-3 (2)        D-808-303(2)      Nuclear Boiler System Figure 7.3-3 (3)        D-808-303(3)      Nuclear Boiler System Figure 7.3-3 ( 4)      D-808-303(4)      Nuclear Boiler System Revision 18 4 4 of 64                                                    October, 2013
 
ADDENDUM 2 Associated Technical Specification Page Change (Mark Up) 1 Page Follows
 
Definitions 1.1 1.1 Definitions (continued)
CHANNEL FUNCTIONAL TEST    A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify
* OPERABILITY. including required alarm. interlock.
display, and trip functions. and channel failure tri~s. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.
* CORE ALTERATION            CORE ALTERATION shall be the movement of any fuel, sources. or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:
: a. Movement of source range monitors. local power range monitors. intermediate range monitors.
traversing incore probes. or special movable detectors (including undervessel replacement);
and
: b. Control rod movement. provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS      The COLR is the unit specific document that REPORT (COLR)              provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 (continued)
PERRY - UN IT 1                      1.0-2                  Amendment No.
 
ADDENDUM 3 Associated Technical Specification Page Change (Re-typed - For Information) 1 Page Follows
 
Definitfons 1.1 1.1 Definitions (continued)
. CHANNEL FUNCTIONAL TEST        A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify
                              *OPERABILITY. including required alarm, interlock.
display, and trip functions. and channel failure tri~s. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential.
overlapping, or total channel steps so that the entire channel is tested.
CORE ALTERATION                CORE ALTERATION shall be the movement of any fuel.
sources. or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions ate not considered to be CORE ALTERATIONS:
: a. Movement of source range monitors. local power range monitors. intermediate range monitors.
traversing incore probes. or special movable detectors (including undervessel replacement):
and
: b.          Control rod movement. provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
* CORE OPERATING LIMITS          The COLR is the unit specific document that REPORT CCOLR)                  provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in .
accordance with Specification 5.6.5. Plant R-t--d-F---~- -t 1tion within these limits is addressed in
          .1 e-ype - or 1 niorma 1on~idual Specifications.
0 0 0 0 DOSE EQUIVALENT 1-131          DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131. I-132. I-133, 1-134.
and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those 1i sted in Federal Guidance Report (FGR) 11.
                                "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation. Submersion. and Ingestion," 1989.
(continued)
PERRY - UNIT 1                                1.0-2                  Amendment No.
 
ADDENDUM 4 Summary of Loss of Coolant Accident (LOCA) Dose Calculation 44 pages follow This Addendum is considered a summary of the associated calculation - only selected
* pages of the approved document are provided.
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radiological Analyses Using Alternative Accident Source Tenn                                                                                Pagei TABLE OF CONTENTS SUBJECT                                                                                                  PAGE COVERSHEET:                                                                                                                                                                  i T~ble of Contents                                                                                                                                                            ii OBJECTIVE OR PURPOSE                                                                                                                                              v SCOPE OF CALCULATION                                                                                                                                              v
 
==SUMMARY==
OF RESULTS/CONCLUSIONS                                                                                                                                    v LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY                                                                                                          v IMPACT ON OUTPUT DOCUMENTS                                                                                                                                        v DOCUMENT INDEX                                                                                                                                                    vi CALCULATION COMPUTATION (BODY OF CALCULATION):
1.0 PURPOSE **********************************************************************************-*--***********-***********************-****-..**..******************** 1 2.0
 
==3.0 BACKGROUND==
      *********-**********M***************************************************--*--******-******************-****-*--*************************
ACCE:Pr'ANCE. CRIJERIA******- ........................................................- ...........................................- ......................... 2 1
4.0      METHOD OF ANALYSIS ******-**--**********..***************...**-***-****-**-*-.....****-..**-**-*****......_..................................... J 5.0      ASSUMPl'IONS................................................................................................._...................................................... 3 5.1      CONnu>L RooM EMERGENCY REORtulATION SV5TEM (CRERS) ............_,,_,,_......................................_._, __ ,,,,... _............... 3 5.2      HYDROGEN MIXING SVSIEM ...................................................................- .........................................- ................................ 4 5.3      CoNnloL RooM INLEAKAGE ................................................................_.............................................................................. 4 5.4      DRYWEll FlOWS ...............................................................................................................................................................4 5.5      CONTAINMENT l.EAxAGE RATE ............................................................................................................................................. 5 5.6      AEGTS FILTRATION ...........................................................................- ...................................................- ......................... 5 5.7      CONTAINMENTSPRAY.............................................................................- **- ......................................- ............................. 5 5.8      ECCS l.EAICAGE **********************************************************************************-*-************************************************************************** 6 5.9      MSIV LEAKAGE RATE ************************************************************* .-............................................................_***************************** 6 5.10    BYPASiS LEAKAGE ...............................................................................................................................................................6 5.11    SOURa TERM RELEASE....................................................................... _...............................................................- ............. 7 6.0      DESIGN INPU'r  -********************-***-*****M*******...--*************-******-*-*----"*"*-**..**H*****H***-...................................                                  8 6.1      PLANrGRADE ***************************-**************************************'!!****~*********-**************************************-****  ................................... &
6.2      CORE SOURCE TERMS AND REl.EASES ..................................................................................................................................... 8 6.2.1    Onset of Gap Release................................................................- ....- .................................................................... 10 6.2.2    Release fractlans. ............................................................*...............-....................................................................... 10 6.3      SUPPRESSION Pool. loDINE RE-E\IOUl110N........................................................................................................................... 10 6.4      Dose CCHvERSION FACIORS ...................................................................- ........................................................................ 10 6.5      ATMOSPHERIC DISPERSION FACTORS ...............................................................................................- ...- .......;..................... 11 6.6      BREATiilNG RATE AND OccuPANCY FACTORS .............................................- ..... _ ........................................_. ...................... 11 6.7      CoNrAINMENT VOLUMES ..................................................................................................................................................12 6.8      TEOfNICAL SUPPORT CENJEt Doses ................................................................................................................................... 12 6.9      MIXING BElWEEN TiiE UNSPRAYED CONTAINMENT AND SPRAYED CONTAINMENT ........................................................- ............. 13 6.10    CoNrAINMENT l.EAtcRATE ......................................................................-      ......................................................................... 13
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radlologlcal Analyses Using Altematlve Accident Source Tenn                                                                                      Page ii 6.11    lEAlcAGE AFTER 24 HouRS ................................................................................................................................................ 14 6.12 MSIV Ft.ows............................- **- ................................................................................................................................ 15 6.13    NATURAL REMOVAL MECHANISMS...................................................................................~.................................................. 16 6.13.1      Elemental Iodine RentOVOl.................................................................................................................................. 16 6.13.2      Aerosol Removal................................................................................................................................................. 19 6.14 ANNUUJS MODEL ................................................................................................................... :........................................ 23 6.15    DEPOSITION IN MAIN STEAM LINES ..................................................................................................................................... 23 7.0      AcaDENT SCENARIO AND OIRONOt.OGY ................... _,, ................- ..... _..........................- ...*--**********-**-*--~ 24 8.0 8.1 MODEL DEVELOPMENT************-*-*-********-****-******************************-***..***************************..*--**--***-..******-***-.. ZS Source Terms ..........................................................................................................................................................25 8.1.l    Volumes ..................................................................................................................................................................25 8.1.3    Flows....................................................................................................................................................................... 25 8.1.4    Removal Mechanisms ............................................................................................................................................. 26 8.1.S    Model.......................................................................................;.............................................................................. 26 8.1.6    Results ............................................- ....................................................................- .............................................. 26 8.2      MSIV LEAKAGE ...............................................................................................................................................................27 8.2.1    Source Terms ..........................................................................................................................................................27 8.2.2    Volumes ..................................................................................................................................................................27 8.2.3    Flows..........................................................................................................................................- ....................- .... 27 8.2.4    Release Points...................................................................................................................................................- **.* 28 8.2.S    Model.......................................................................................................................................................................28 8.3      CoNTAINMENT & CONTAINMENT BYPASS l.EMAGE................................................................................................................ 28 8.3.J Volumes ....................................................................................................................................- ........................... 28 8.3.2    Flows......................................................................................................................- ......................- ********************** 29 8.3.3    RentOllOI Mechanisms.***********.*****.*- **- ...........................................................................- .............- .................... 30 8.3.4    Release Points......................................................................................................................................................... 31 8.3.S    Model............................................. _......- ............................................................................................................... 31 8.4      CON'TRoL RooM *********************************************************************************!***********-************************-*~********************************** 31 8.4.l Results ...................................................................................................................... _ ........................................... 32 8.5      RADTRAD MODEL........................................................................................................................................................32 9.0      OPEi.ATOR ACl'IONS *************---*-*****-.......................................................__.................................................. 32 10.0    COMPUTATI0" ......................................................................................................................................_............ 15 1LO      OVERAl.L RESUL"IS..................................................................................................- ...- ...................................... 15 ATTACHMENTS:
Attachment 1: PNPP ESF EQ.out                                                                                                                                    19 Pages
~A~tta..;._ch~ment~-2-:PN-=PP---L_OCA___;~EQ-.-out~~~~~~~~~~~~~~~~~-------1244Pages Attachment 3: PNPP LOCA TSC.out                                                                                                                                  32 Pages
~A~tta---ch_m_181_nt_4_:=PN~P=P~e=s~F~T=s~c-.out----------------------------------------------------------~20Pages Attachment 5: Hydrogen Mixing System Actuation Tme Beftliliuly Study                                                                                            223 Pages 1'61 'l*Jl*13
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radiologlcal Analyses Using Alternative Accident Source Tenn      Page iii SUPPORTING DOCUMENTS (ForReconls Copy Only)
DESIGN VERIFICATION RECORD                                                                1 Page CALCUlATION REVIEW CHECKLIST                                                              ,Pages 10CFR50.59 DOCUMENTATION DESIGN INTERFACE
 
==SUMMARY==
 
DESIGN INTERFACE EVALUATIONS
                                                                                        "  Not Applicable 12Pages 0 #Pages
                                                                                                                  /ID 9.21.B DESIGN INPUT RECORD                                                                    0 .aPages EXTERNAL MEDIA? (MICROFICHE, ETC.) (IF YES, PROVIDE LIST IN BODY OF CALCUlATION)          LI YES 181 NO TOTAL NUMBER OF PAGES IN CALCUlATION (COVERSHEETS +BODY+ ATTACHMENTS)                      174Pages
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radlologlcal Analyses Using Altematlve Accident Source Tenn      Page iv OBJECTIVE OR PURPOSE:
This calculation replaces the current loss-of-coolant accident (LOCA) dose ca1cu1aac.n (PSAT 08401T.03, DIN 25) and supports the tra'1sition to GNF2 fuel. In addition, certain excess conservatisms contained In the current LOCA dose calculation will be removed to increase the margin of safety. This calculation will be performed in accordance with the guidance provided in Regulatory Gulde 1.183 (DIN 7) for application of an alternative radiological source term and will demonstrate that the offsite and onsite post-accident doses comply with the requirements and acceptance criteria of 10 CFR Part 50.67.
SCOPE OF CALCULATIONIREYISION PNPP will be transitioning to GNF2 fuel in future outages. The purpose of this calculation Is to prepare a dose analysis supporting this transition and to establish the new design basis LOCA dose analysis using the RADTRAD 3.03 computer program, which was developed for the Nuclear Regulatory Commission (NRC) and Is In common use for this type of application In the nuclear Industry. This calculation may also be used to support a license amendment request (LAR) associated with the GNF2 fuel transition.
 
==SUMMARY==
OF RESULTS/CONCLUSIONS The post-accident offsite, Control Room, and Technical Support Center doses for a postulated design basis LOCA meet*the requirements of 10 CFR Part 50.67. The LOCA dose results, including an leakage pathways, from Table 11..(Zare given below:                                                                Jli' 'l*/113 LOCADoae          Regulatory Umlt (DIN 7, 14)
Location (remTEDEl                fremTEDEl Exclusion Area Boundary (EAB)                      21.2                25 (0.25SV)
Low Population Zone (LPZ)                            6.5                25 (0.25SV)
Control Room                                        2.8                  5 (0.05 SV)
Technlcal Supj>ort Center (TSC)                      0.7                  5(0.05SV)
LIMITATIONS O~ RESTRICTIONS ON CALCULATION APPLICABILITY:
This calculation determines the radiological dose consequences resulting from the reactor coolant release that accompanies a postulated design basis LOCA, which are reported in USAR Section 15.6.5.
This calculation will become the licensing basis LOCA dose analysis after the transition to GNF2 fuel.
IMPACT ON OUTPUT DOCUMENTS The results of this calculation will be incorporated into the USAR following LAR approval.
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radiological Analyses Using Altematlve Accident Source Tenn            Page v DOCUMENT INDEX B
c a                                                                                                            ~
z
    ~                              Document Numbertrdle Revision, Edition, Date  i    !5 Q.
c 0
1    Safety Evaluation by the Ollice of Nuclear Reaclor Regulation Related to      NIA            a  0          0 Amendment No. 103 to Facility Operating License No. NPF-58 2    NUREGICR.e604, RADTRAD: A Simplified Model for Radionuclide Transport        December      a  0      0 and Removal and Dose Estimation                                              1997 3    NUREG-1465, Accident Source Terms for light-Water Nuclear Power Plants        February      a  0      0 1995 4    Federal Guidance Report 11, Limiting Values of Radionuclide Intake and Air    Second        a  0      0 Concentration and Dose Conversion Factors for Inhalation, Submenllon, and    Printing, 1989 Ingestion 5    Federal Guidance Report 12, External Exposure to Radlonudldes In AJr, water,  September      a  0      0 and Soil                                                                      1993 6    NUREGICR-5966, A Simplified Model of Aerosol Removal by Containment          June1993      a  0      0 Sprays 7    Regulatory Gulde 1.183, *Alternative Radiologlcal Source Terms for Evaluating July2000      a  0      0 Design Basis Accidents at Nudear Power Reac:tora*
8    NotUsed                                                                                      0  0      0 9    GEH-KL1WX23P.017, from E. G. Thacker II (GE) to E. S. Tomlinson Ill          May7,2012      0  181    0 (FENOC), PNP GNF2 Fuel Transition: F0802 Source Tenn Output Flies 10  DESl98-0845, Telephone and Confaranca Memorandum by Paul J.                  1212198        0  181    0 Roney/DES, Perry Control Room Almoapherlc Dlsper&lon Factor& (Chl/Q) 11  PSAT 150.01C.03, Dose C&lculatlon Data Base for Applic:atlon of the Revised  Ravl&lon2      0  B      0 DBA Sourat Tenn to the*Peny Power Uprate 12  PSAT 04202H.04, Aerosol Decay Rates (Lambda) In Drywell                      RavlalonO      0  181    0 13  PSAT 04202U.03, Dose Calculation Dais Base for Appllcation of the Revised    Ravl&lon2      0  B      0 DBA Soun:e Tenn to the CEI Peny Nuclear Power Plant 14  10 Code of Federal Regulations 50.67, Aa:idenl Source Terms                  December 23,  a  0      0 1999 15  PNPP Technical Specifications                                                Amelldment    0  181    0 150 16  M26-001, M28, Volume Calculation, Control Room Envelope                      Revlsion2      0  181    0 17  PSAT 04202H.13, Offslte and Control Room Dose Celallatlon                    Revision 1    0  181    0
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn                    Page vi 8c zci                                                                                                                  'S a
c DoaJment Numbertrdle Revision, Edition. Data    j    'S Q.
c fi 18    PERRYUSAR                                                                          Revision 17    a    0    D 19    NEI 99-03, Control Room Habitabluty Assessment Guidance                            Man:h2003      a    D    0 20    Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration Revlaion2. a    D    0 and Adsorption Units of Post-Accident Englneerad-Saraty.Feature Atmosphere          Man:h1978 Cleanup Systems in Light-Water-Cooled Nuclear Power Plants" 21    PNPP Technical Specification Section 5.5. 7, PNPP Ventilation Fitter Testing        Amendment      181  D    D Program (VFTP)                                                                      143 22    .NUREGICR-6604, Supplement 1, RADTRAD: A Simplified Model for                        Juna&.1999    a    0    0 Radionuclide Transport and Removal and Dose Esllmation 23    NUREGICR-6604, Supplement 2, RADTRAD: A Simplified Model for                        October 2002  a    0      0 Radionuclide Transport and Removal and Dose Estimation 24    10 Code of Federal Regulations 20 , Standards For Prutection Against                Oct. 1,2007    a    0      0 Radiation 25    PSAT 08401T.03, Perry Plant Total Effective Dose Equivalent (TEDE)                  Revision&      a    E      0 C&lculatlon 28    PNPP C&k:ulation 3.2.15.17, Containment Water Pool pH Post-Accident                RavlsionO      0    E      0 27    GE letter from D. Braden (GE) to E. Root (CEI), GE.PAIP-651, DRF A22-00084-        NIA            0    E      0 oo, dated 3112/01, "Additional Contaimnent Response curves*
28    10 Code of Federal Regulations Part 50
* Domestic Licensing of Pnxlucllon and      77FR39906,    a    0      0 utmzatlon Facilities                                                                Jul.8,2012 29    10 Code of Federal Regulations Part 100 - Reac:tor Site Criteria                    77FR39910,    181  a a Jul.8, 2012 30    RADTRAD Computer Program Certification, FNOCPP184, CEl-120                                        181  a a 31    CEI C&lallation 3.2.6.4, Post-LOCA Doses with Spray at 10 min for 6 Hours and      RevlslonO      181  0      a
        *Control Room lnleakage of 90 CFM 32    Wolfram Mathworld httD:/lmlthwo~,WDlfram.com/OblaleSDherold,html                    aocaaaad      181  0      0 7/20l2012 33    NUREGICR-0009, Technological Bases for Models of Spray W8ahout of                  0              181  0      D Airborne Contaminants In Containment Vessels 34    NUREG-800, Standard Review Plan (SRP) 8.5.2, Containment Spray As A                Revialon4,      181  D      0 Fission Product Cleanup System                                                      Man:h2007 35    PNPP Drawing 320-0661.00000, Containment Spray System                              RavislonT      0    E      0
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radiological Analyses Using Alternative Accident Source Tenn          Page vii za                                                                              Revision,      I    'S
                                                                                                                'S a
c Document NumberfT"dla                            Edition. Dale  I    @-    ~
36    PNPP Drawing ~14-661, Sheet 3, Containment Vessel Spray Ring "A" Piping    Revision&      0  181    0 37    PNPP Drawing ~14-661, Sheet 8, Containment Vessel Spray Ring "B"-          Revision&      0  181    0 Piping 38    PNPP drawing SS-304-661, Sheet 105.2, Piping lsomebic- Containment Spray    RevislonC      0  181    0 System 39    PNPP Drawing ~14-661, Sheet 7, Containment Vessel Spray Ring "D"            Revision&      0  181    0 40    PNPP Drawing SS-304-661, Sheet 102.2, Piping Isometric-Containment          Revision B    0  181    0 Spray System Reactor Building 41    PNPP Drawing ~14-661, Sheet 6, Containment Vessel Spray Ring "P            Revision&      0  Im      0 42    PSAT 04202H.08, Steamllne: Particulate Deconlamlnatlon Calculation.        Revl8lon 1    0  Im      0 43    PSAT 04212H.02, Orywell S-.<>ut Rate and Related ThennaJ.Hydraulic          Revision 1    Im  D      D Conditions Inside Containment 44    NUREG-0800, Standard Review Plan, 15.6.5, Appendix B, Radiological          Ravlaian 1,    Im  0      D Consequences of a Design Basis Los&-oJ.Coolant Accident: Leakage fn:lm      July 1981 Enolneered Safatv Feature ComP01'18ntB Outside Containment 45    PNPP Drawing 511-001&00000, Reactor Building- Steel Framing Sections        RevlslonO      0  181    0 and Detalla 46    CEI Calculation 3.2.6.3, LOCA Doses as a Function of Spray Initiation Time  RevislanO      0  181    D 47 003.008-001-00, FSAR Figura 3.8-1, "Typical Section of Reactor Building
        ~
Revision 12    Im  0      0 48    NUS letter from A E. Mitchell (NUS) to R. F. Zucker (CEI). CSA--8106187-12, NIA            D  0      181 PY-NUSICEl-1474, dated 312198, "Habitability ChllQs forTSC", Attached to PNPP Cale 5.7.1.2 49    PNPP Calculation.5.7. 1.2, Technical Support Cenlar- Final Dose            RevlslonO      0  0      181 so  PNPP Drawing E-013-011, Final Plant Layout, Section ArA                    Revision&      D  181    0 51  PNPP Drawing E-002-002, Final Plant Layout, Section A                      Revision 15    0    181    0 52  PNPP Drawing D-912-810, Control Room HVAC and Emergency Redn:ulatlon        RevislonFF    0    181    D SVBtam WASH-1400, "Reactor Safety Study: An Assessment of Accident Risks                          Im          D 53 In U.S. Commerclal Nuclear Power Plants,* NUREG 751014, Nuclear 1975                D Reaulatorv Commission YJashlnaton. DC.
D.I. Chanin, J.L Sprung, LT. Ritchie, and H-N .tow, "Melcor Accident 54  Consequence Code System (MACCS) User's Guide," NUREGICR4191, Vol. 1,        1990            Im  D      D Sandia National Laboratortea.  & K          NM.
 
TITLE/
 
==SUBJECT:==
Loss of Coolant Accident Radlologlcal Analyses Using Altematlve Accident Source Tenn      Page I 1.0    PURPOSE The purpose of this calculation Is to prepare a dose analysis supporting the transition to GNF2 fuel and to establish the new design basis loss-of-coolant (LOCA) dose analysis. This calculation makes the current LOCA dose calculation (PSAT 08401T.03, DIN 25) and Technical Support Center calculation (Caleulation 5.7.1.2, DIN 49) historical and removes certain conseNBtisms contained in the current LOCA dose calculation to increase the margin of safety. This calculation will be performed in accordance with the guidance provided in Regulatory Guide 1.183 (DIN 7) for application of an alternative radiological source term and will demonstrate that the offsite and onsite post-accident doses comply with the requirements and dose limits of 10 Code of Federal Regulations (CFR) Part 50.67 (DIN 14). Onsite doses calculated include the Technical Support Center (TSC) dose.
The evaluation of the limiting design basis loss-of-coolant accident will use the RADTRAD 3.03 Code Instead of the proprietary STARDOSE code used in PSAT 08401T.03. RADTRAD 3.03was developed for the NRC and is commonly used in the nuclear power industry for applications of this type. Excess conservatisms removed from the current calculation (PSAT 08401T.03) are given in detail in the following sections and are summarized below:
* Increased Control Room Emergency Recirculation System (CRERS) filter efllclency for elemental and organic iodine from 50% to 80%.
      * *Credit for decay during the two (2) minute onset of the gap release.
* Credit for elemental and aerosol removal In the unsprayed containment region.
* Credit for reduced containment and annulus bypass leakage after 24 hours based on post-accident containment pressure.
* Increased CRERS HEPA fitter efficiency from 95% to 99%.
An additional conservatism added to this calculation is the removal of credit for auto-initiation of the CRERS. Isolation of the nonnal ventilation system and actuation of CRERS is assumed to be performed manually from the control room at 30 minutes post-accident.
 
==2.0    BACKGROUND==
 
The Perry Nuclear Power Plant (PNPP) pilot Altemative Source Term (AST) submittal to the NRC was based on the LOCA analysis presented In PSAT 08401T.03, Revision 5 (DIN 25). This analysis utilized the POLESTAR proprietary computer code STARDOSE to determine the offalte and onsite consequences of a LOCA.
The NRC, In approving the PNPP pilot Attemative Source Term (AST) submittal, perfonned a confirmatory radiological consequence calculation that evaluated potential fi99ion product release pathways following a postulated LOCA. The NRC calculation was documented In the Perry Safety
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn              Page 2 Evaluation by the Office of Nuclear Re8ctor Regulation related to Amendment No. 103 (DIN 1). The NRC staff used the RADTRAD Code. .
* The guidance of Regulatory Gulde 1.183, *Alternative RadlologlCal Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors* (DIN 7), will be used to Identify the conservatisms currently being applied in the Perry design basis LOCA model. Regulatory Guide (R.G.) 1.183 establishes an acceptable Alternative Source Term (AST) and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. This guide also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST. This calculation will.remove some conservatism per the guidance of R.G. 1.183.
3.0    ACCEPTANCE CRITERIA The post-accident offsite and control room doses must meet the requirements of 10 CFR Part 50.67,
  *Accident Source Term.*
10 CFR 50.67 gives the limits applicable to plants revising their accident source terms. The dose limits specified are given in&sect; 50.67, (b)(2)(1), (I~. and (iii) as follows:
(bX2Xi) -An individual located at any point on the boundary of the exclusion mea for any 2-hour period following 1be onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(bX2Xii) - An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting fiom the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of0.25 Sv (2S rem) total effective close equivalent (TEDE).
(b)(2Xiii) - Adequate radiation protection is provided to pennit access to*ancl occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of O.OS Sv (S rem) total effective dose.equivalent (TEDE) for the dmation of the accident.
For plants Implementing the alternative radiological source term methodology, the dose Omits of 10 CFR 50;67, given above, replace the limits given In 10 CFR 100.11, '!Determination of exclusion area, low population zone, and population center distance.* which are expressed In terms of whole body and thyroid dose as follows:
(aXI) An exclusion area of such sim that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 2S rem or a total radiation dose in, excess of300 rem to the thyroid fiom iodine exposure.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn                Page 3 (a)(2) A low population zone of such sii.e that an individual localed at any point on its outer boundary Who is exposed to the radioactive cloud resulting tiom the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of2S rem or a total ndiation dose in excess of 300 rem to the thyroid tiom iodine exposure.
As noted above, the dose limit for control room personnel is specified in 10 CFR Part 50.67 (DIN14).
4.0        METHOD OF ANALYSIS This calculation will evaluate the total effective dose equivalent (TEDE) for the Perry Nuclear Plant design basis radiological accident (LOCA) using the revised accident source term based on Regulatory Guide 1.183, *Alternative Radiological Source Terms for Evaluating Design Basis Acddents at Nuclear Power Reactors* (DIN 7). The TEDE dose Is defined as the sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures) (DIN 24). The RADTRAD Code, Version 3.03, will be used to calculate radiological consequences In terms of TEDE.
RADTRAD (Radionuclide Iransport and Removal And Dose Estimation) calculates fission produd transport and removal along with the resulting radiation doses at selected receptors. The code is described in NUREG/CR-6604, *A Slmprtfied Model for Radionuclide Transport and Removal and Dose Estimation* (DIN 2, DIN 22, and DIN 23). RADTRAD 3.03 W8$ certified for this application (DIN 30) In accordance with the ENERCON computer code certification procedure [ENERCON CSP 3.02).
5.0      ASSUMPTIONS 5.1      Control Room Emergency Recln:ulaaon System (CRERS)
Upon receipt of an ESF aduation system signal or high radiation signal, the Perry control room heating, ventilation, and air conditioning (HVAC) system is designed to automatically isolate and activate the CRERS; however, this analysis conservatively assumes that the normal HVAC system continues to operate with an outside air Intake (6000 cfm) and exhaust to the environment (4800 cfm) until the CRERS Is manually actuated at 30 minutes.
Each .redundant CRERS subsystem has a high efficiency particulate air (HEPA) filter, charcoal adsorbers and a post HEPA filter. Operation of the CRERS fans, charcoal adsorbers, and HEPA fitters are credited In this analyslS. The CRERS Is an ESF system that Is tested In accordance with R.G. 1.52. The current test acceptance criterion for the CRERS charcoal adsorbers requires a penetration of less than 2.5%
(DIN 21 ). Based on the testing requirements, a charcoal adsorber removal efliciency of 95% could be justified; however, for additional operational margin, elemental and organic Iodine removal efficiency Is assumed to be 80%. Technical Specification 5.5.7 (DIN 21) states that each HEPA filter Is tested to show a penetration and system bypass of less than 0.05% when tested in accordance with Regulatory Guide 1.52 (DIN 20). A penetration and bypass of less than 0.05% allows credit for a particulate removal efficiency of 99% per Regulatory Position C.5.c of Regulatory Guide 1.52. This analysis 11888 a HEPA filter efficiency of 99 percent for aerosol particulates.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 4 5.2      Hydrogen Mixing System The hydrogen mixing system Is assumed to be manually Initiated. Based on manual operation, the earliest system initiation Is assumed to be 30 minutes following the onset of an accident. Start of the hydrogen mixing system at 30 minutes Is addressed in a sensitivity case provided in Allachment 2; however, the base case assumes that hydrogen missing system actuation is delayed until two hours to provide additional operator flexlbUity. Additional leakage from the drywall into the primary containment is due to steaming from the heated reactor core in accordance with R.G. 1.183, Appendix A. Assumption 3.7. This leakage is assumed during the two-hour period between the Initial blowdown and termination of the fuel radioactivity release (gap and early in-vessel release phases). The termination of the release from the core is due to core recovery and reflood. Instead of evaluating all of the potential steaming rates due to various refloodlng scenarios, this analysis will assume that there Is a homogenous mixture in the drywell and containment starling at two hours. The assumption of a well-mixed drywall and containment atmosphere at two hours is appropriate because the EAB radiological doses consider the worst two hours as opposed to the first two hours as was done for the previous TIO 14844 source term methodology. The assumption of a well-mixed drywall and containment atmosphere Is implemented by assuming a high mixing flow (2.77E+05 cfrn, approximately one drywall volume per minuta) between the two volumes. The mixing flow is conservatively assumed to continue for the duration of the accident Instead of isolating the drywall after the core Is quenched.
5.3
* Control Room lnleakage As described in Section 6.4 of the PNPP USAR (DIN 18), the control room Is normally maintained at a slightly positive pressure to the surrounding areas from the 6000 cfm fresh air makeup and out leakage of 4800 cfm. In the Isolated mode, there Is no Intake from outside air sources and the control room pressure would eventually reach that of the surrounding areas. After the CRERS Is Initiated, the maxlmu!11 control room unfiltered lnleakage of 1375 cfm, will be used (DIN 11 ). The leakage out of the control room envelope Is also modeled as 1375 cfm to avoid pressurization of the control room envelope.
5.4      Drywall Flows The flow rate from the Drywell to the Wetwell, which bypasses the suppression pool, is given in PSAT 04212H.02 (DIN 43) as follows:
Table 5-1 Drywall and Wetwell Mbdng Flows Time After Gap Release          Flow from DW tD WW      Flow from WW tD DW (houm)                          (cfm)                    (cfm) 0-0.5                            0                          0 0.5-2.0                          3000*                      O*
2.0-720                        2.77E+OS                  2.77E+o5
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn      Page s TllE      M l~Mt:.                                  .                ,1 CJ
            *The base case assumes thathydrogen MiaaiRI system actuation Is delayed unbl two hours. The 9,;1.11 sensitivity study presented in Attachment 2 assumes a 500 cfm hydrogen mixing system flow Into and ~ut of the drywell beginning at 30 minutes In addition to the 3000 cfm flow listed above.
5.5      Containment Leakage Rate The primary containment consists of a drywell, a wetwell and supporting systems to limit fission product leakage during and following the postulated LOCA with rapid Isolation of the containment boundary penetrations. The secondary containment will collect and retain fission product leakage from the primary containment and will release fission products to the environment through the AEGTS. During normal operation, the shield building Is maintained at slight negative pressure. Following a OBA, it is expected to remain negative, however for a short period it may not be maintained below the design negalMt pressure value of 0.25-inch water gauge (USAR 6.5.3.2.1 ). Therefore, it will be assumed that the primary containment leakage is released directly to the environment for the first 40 seconds following the LOCA (DINs 11 and 25).
5.6      AEGTS Fiitration The AEGTS includes HEPA fillers which are periodically tested to demonstrate compliance with Regulatory Guide 1.52. Particulate removal by the HEPA fillers is assumed to be 99% In accordance with Regulatory Gulde 1.52 (DIN 20). The system also contains 4-lnch deep activated charcoal adsorbers to remove elemental and organic Iodine; however, this analysis conservatively assumes a removal efficiency of 0% for the charcoal adsorbers to allow operational flexibility.
The AEGTS extracts and filters a maximum of 2000 cfm from the annulus. During an accident, the maximum *expected discharge to the atmosphere is 1000 cfrn (DIN 18). The balance of the filtered AEGTS flow is routed back to the annulus. This analysis conservatively assumes that 2000 cfm is discharged directly to the environment with no recirculation (holdup) of Iodine In the annulus.
5.7      Containment Spray Manual Initiation of containment spray Is assumed at thirty minutes instead of automatic Initiation on high
*pressures and tow water level per the current design. Containment spray is assumed to end at 24 hours et which time the aerosol removal by containment spray is terminated. USAR Subsection 8.5.2.3 gives a discussion of the non-mechanistic assumption that sprays will operate up to 24 hours. In an actual event, spray use would not necessarily be suspended at 24 hours if appropriate conditions for their use still existed. Therefore, the assumption that sprays stop at 24 hours is not *Intended to be interpreted as a commitment to stop using sprays after 24 hours. In addition, the statement that sprays will operate up to 24 hours implies that the sprays will not necessarily operate continually for 24 hours.
The containment sprays will be run when it Is appropriate, and not necessarily the entire time during the first 24 hours of a LOCA. However, this does not Invalidate the assumption used in this calculation. The accident guidance to operators is symptom based, rather than event based. Most postulated LOCAs will not result in large radiation releases and would not require containment sprays to run for 24*hours for
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn      Page 6 removal of radioactivity from the containment. Sprays are also used to reduce containment pressure, as needed, by steam condensation and containment heat removal. If a high radiation signal is present from the containment radiation monitor and pressures are elevated in containment, the sprays would be operated. However, if containment gauge pressure Is reduced to near zero and use of the sprays is terminated by the operators, this does not have an adverse impact on off-site doses (or the dose calculations) since the driving pressure for containment and main steam Hne leakage has been eliminated. The dose calculations assume that the maximum allowable leakage (La) corresponding to the peak post--accident pressure (Pa) remains during this first 24 hour period, so if containment pressure is reduced to substantially less than Pa, a reduction in leakage and the resultant offslte doses will follow. If containment pressure increases again, and the high radiation signal is present. sprays would be actuated again.
5.8      ECCS Leakage Leakage from ECCS systems located outside the primary containment is estimated to be 5 gph (DIN 48).
For conservatism, this analysis assumes that the ECCS leakage is 15 gph for the entire duration of the accident. Additionally, leakage from a gross failure of a passive component Is assumed to occur at a rate of 50 gpm starting 24 hours into the accident and lasting 30 minutes In accordance with NUREG-0800 SRP 15.6.5, Appendix B (DIN 44)
* Regulatory Gulde 1.183, Appendix A. Section 5.2 states that engineered safeguards feature leakage should be assumed to start at the earliest time that recirculation flow occurs in these systems and end at the latest time the releases from these systems are terminated.
For PNPP, ECCS leakage may begin up to 30 minutes post-accident but Is assumed to begin at the onset of gap release at two minutes and continue for the duration of the event This is a conservative assumption which maximizes the dose contribution for this release pathway.
5.9      MSIV Leakage Rate There are four main steam lines: each line has an Inboard MSIV, an outboard MSIV, and a third Isolation (shutoff) valve. This analysis assumes a double guillotine pipe rupture In one of the four main steam lines upstream of the Inboard MSIV and failure of all four third main steam shutoff valves (1 N11-F0020A, B, C, and D) to close as a result of a common power failure (slngle-fallure criterion). A total of a 250 acfh maximum allowable leakage limit (TS SR 3.6.1.3.10) is assumed to occur: (1) 100 scfh through the broken steam line, (2) 100 scth through a second intad steam line, and (3) the remaining 50 scfh through a third lntad steam line. This is modeled as 100 scfh through the broken steam line and 150 scth through the unbroken steam lines.
5.10 Bypass Leakage Secondary containment bypass leakage is In addition to the containment allowable leakage, L.. The leakage paths Include all pathways which could potentially allow leakage to bypass secondary*
containment. Therefore, any bypass leakage releases would not be.treated by an ESF filtration system
 
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Loss of Coolant Accident Radiological Analyses Using Alternative Accident Source Tenn  Page 7 prior to being released tO the environment The secondary containment bypass leakage is currently limited to 5.04% oU.a. when pressurized to ~P,, by Technical Specification SR 3.8.1.3.9 (DIN 15) even
* though the previous LOCA dose calculation (DIN 25).assumed a leakage of 10.08% of La. The containment bypass leakage will be maintained at 0.1008 La In this analysis to allow for an Increase in the Technical Specifieation allowable leakage limit.
5.11    Source Tenn Release In accordance with R.G. 1.183 (DIN 7), only the gap and In-vessel release phases are considered In this design basis LOCA dose calculation. The core source terms are assumed to be released at a constant rate such that the release is completed by the end of the specified release period. Assumptions regarding release fractions and timing are consistent with Tables 1 and 4 of R.G. 1; 183 (DIN 7).
Table 1 of R.G. 1.183 Is given below:
BWR Core Inventory Fracdon Released Into Containment I~:~.J              ~~:__J
                                . Gro~            Phaiel Phase~
Noble Gases            0.0S          0.95          1.0 Halogens              0.05          0.25          0.3 Alkali Metals          O.OS          0.20          0.25 Tellurium Metals      0.00          O.OS          O.OS Ba, Sr                0.00          0.02          0.02 Noble Metals          0.00          0.0025        0.0025 Cerium Group          0.00          0.0005        0.0005 Lantbanides            0.00          0.0002        0.0002 Table 4 of R.G. 1.183 is reproduced below:
LOCA Release Phases PWRs                        BWRs Phase                Onset        Duration Onset            Duration Gap Release          30 sec        0.5 hr      2 min          0.5 hr Early In-Vessel      o.s hr        1.3 hr    0.5 hr          1.5 hr
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term    Page 8 6.0    Design Input 6.1    Plant Grade The PNPP plant grade elevation ls*620 feet (DIN 50).
6.2    Cont Source Tenns and Releases The PNPP core source term release magnitude, timing and chemical form are based on Regulatory Guide 1.183 (DIN 7). The core source terms were developed by GE Hitachi (DIN 9). The calculated inventories are based on 2-year GNF2 refueling cycles and serve as input for design basis accident analyses based on Regulatory Guide 1.183 source term assumptions. The fission product Inventory calculations were performed using the ORIGEN2 code. The Ci/MW multipliers developed in DIN 9 are applied here to generate the core source terms at the onset of the event A reactor power level of 3833 Mwt will be used based on 102% of the rated thermal power, 3758 Mwt, as defined In Technical Specification 1.1, Definitions, page 1.0-5, Amendment 112 (DIN 15).
The GNF2 fuel source terms are based on the GNF2 equilibrium source activity given below. The source terms include fission products, actinides, and. activation products. The listing of isotopes given in Table 6-1, below, is based on the isotopes used in the RADTRAD computer code. As stated In the RADTRAD User's Manual, NUREG/CR-6604 (DIN 2), the 60 isotope nuclide file is based on isotopes selected In WASH-1400 [DIN 53) with the addition of 6 Isotopes used in the MACCS code [DIN 54).
Table&-1 Source Tenn
                                          .._              GNFZ Equilibrilm Co58                2.847E..o2 CCl60              4.821E..o2 Kr85              3.789E..o2 Kl85m              8.737E..o3 ICl81              t.283E-t04 Kl88              t.804E-t04 Rb88                6.882E..o1 Sr89              2.425Et04 Sl90              3.0t6E..o3 8191              3.084E-t04 6192              3.348E-t04 Y90              3.207E..o3 Y91              3.t53E-t04 Y92              3.383Et04 V93              3.928E-t04 ll95              4.72tE-t04
                                            '1Jfll            4.946E-t04
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn Page 9 GNFZ Equilibrium lsalnDa Nb95 TG99m 4.741E.o4 5.127E.o4 4.470E.o4 Ru103              4.309E..o4 Ru105              3.046E..o4 Ru106              1.750E.o4 Rh105              2.871E..o4 Sb127              3.016E..o3 Sb129              8.906E..o3 Te127              2.997E..o3 Te127m              4.049E402 Te129              8.762E..o:t Te129rn            1.304E..o:t Te13tm              3.965E..o3 Te132              3.85CE.o4 1131              2.714E..o4 1132              3.914E..o4 1133              5.495E..o4 1134              8.025Et04 1135              5.150E-t04 X.133              5.302Et04 X.135              1.934E..OC Cs134              8.926E..o3 Cs138              2.162E..o3 Cs137              4.190E..o:t Ba139              4.877E-t04 Ba140              4.709E..o4 la140              5.CJ02E.e04 la141              4.44CE..OC la142              4.27Et04 Ce141              4.460E.o4 Ce143              4.090E.o4 Cet44              3.870E.o4 Pr143            3.957E..o4 Nd147              1.795E-e04 Nn219              5.819E..o5 Pu238              1.338E"&deg;2 Pu239              1.291E..01 Pu240              1.749E..01 Pu241              5.748E..o:t
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn      Page to GNF2 Equilibrium l!mlaaa              ..
Am241                7.237Et00 Cm242                1.799E..o3 Cm244                1.124E..o2 6.2.1 Onset of Gap Release Table 4 of Regulatory Gulde 1.183 tabulates values acceptable to the NRC for the onset and duration of each sequential release phase for OBA LOCAs at PWRs and BWRs. The specified onset of the gap
                                                                            =
release is the time following the initiation of the accident (I.e., time 0) prior to the start of the gap release. For a BWR the onset is 2 minutes. Credit will be taken for decay prior to the onset of the gap release at 2 minutes.
6.2.2 Release fractions The release fractions used In this analysis are consistent with Table 1 of R.G. 1.183 (DIN 7) which is reproduced In Section 5.11 of this calculation.
6.3      Suppression Pool Iodine Re-evolution The impact of any postulated Iodine nMtvolutlon from the suppression pool hes been evaluated.end shown to be negligible based on the pool pH level. If the pH is maintained above 7, very little (less than 1%) of the dissolved Iodine will be converted to elemental iodine (DIN 1, DIN 7). The Standby Liquid Control SyStem (SLCS) Is used for controlUng and maintaining long-term suppression pool water pH levels .to 7 or above. The pH of post-accident water In the containment will remain above 7 for the entire duration of the postulated LOCA (DIN 28). As such, this analysis will not consider any impact to the offstte or control room doses due to Iodine re-evolution from the suppression pool. Also, in accordance with Appendix A of Regulatory Gulde 1.183 (DIN 7), because the suppression pool pH Is controlled at values of 7 or greater, the chemical form of radlolodine released to the containment can be assumed to be 95% cesium iodide (Csl), 4.85 percent elemental iodine, end 0.15 percent organic iodide.
6.4      Dose Conversion Factors The effective dose conversion .factors for the TEDE calculations are based on FGR 11, *umiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion* (DIN 4) and FGR 12, *External Exposure to Radlonuclldes In Air, watar, and Sor (DIN 5).
These reports tabulate dose coefficients for external exposure to photons and electrons emitted by radlonuclides distributed in air, water, and soil, es well es, dose coefficients for the committed dose equivalent to tissues of the body per unit activity of Inhaled or ingested radionuclide&. These dose coefficients for exposure to radiation are Intended for the use In calculating the dose equivalent to organs and tissues of the body and are endorsed by the NRC In Regulatory Gulde 1.183, Sections 4.1.2 (FGR
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 11
: 11) and 4.1.4 (FGR 12)*. Dose conversion factors for the 60-isotope, 9 element NUREG 1465 (DIN 3) accident source term composition are included in the RADTRAD Input.
6.5    Abnospherlc Dispersion Factors The atmospheric dispersion factors (x/Q values) for the LPZ and EAB are obtained from PSAT
* 04202U.03 (DIN 13) and Calculation 3.2.6.3 (DIN 48). The Control Room atmospheric dispersion factors are documented in DES/98-845 (DIN 10). The Technical Support Center atmospheric dispersion factors are documented in PY-NUSJCEl-1474 (DIN 48) and Revision o of the TSC Dose Evaluation (DIN 49).The x/Q values, based on a ground level release, are given below:
Table&-3
                                                'llQ (sec/mI')
Location Time Interval            EAB                LPZ Oto2 hrs                    4.3E-4            4.BE-5 2 to8 hrs                                      4.8E-5 8to24 hrs                                      3.3E-5 24to96hrs                                      1.4E-5 96to720 hrs                                    4.1E-6 Table&-4 Control Room and TSC 'llQ (sec/m~
Time Interval            CONTROL ROOM                      TSC Oto 8 hrs                              3.SE-4                    5.1E-5 8to24 hrs                              2.1-E-4                  4.1E-5 24to96hrs                              1.1E-4                    3.1E-5 96to720 hrs                          5.75E-5                    1.1E-5 6.6    Breathing Rate and Occupancy FactOra The breathing rates applied In the calculation of the inhalation dose are consistent with those reported In Sections 4.1.3 and 4.2.6 of R.G. 1.183 (DIN 7).
TableM Breathln 1 Rates Im la)
Time Period        EAB              LPZ            Control RoomtrSC Oto 8 hours          3.5E-4          3.SE-4            3.5E-4 8to24hours          1.8E-4          1.BE-4            3.5E-4 1 to30davs          2.3E-4          2.3E-4            3.5E-4
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term Page 12 The control room and TSC occupancy factors are consistent with those reported In Section 4.2.6 of R.G.
1.183 and are tabulated below.
Table6-8 Control Room and TSC Occu nc Factors Time Period      Occu nc Factor O to 24 hours              1.0 6.7    Containment Volumes The volumes of the containment regions are from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31).
Table6-7 Containment Volumes Region                          Volume (ft')
Unsprayed Containment                            684,228 Sprayed Containment                              481,174 Drywall                                          276,500 Note: the above volumes are shown as rounded values in the RADTRAD 8Cl89f1 views but the actual values are used In the RADTRAD input file.
6.8    Technical Support Center Doses Dos8s to personnel In the Technlcal Support Center (TSC) are-calculated In the same manner as the doses to the Control Room operators except for the TSC specific atmospheric dispersion, JIQ, values and TSC data. The additional data needed to determine the TSC doses Is as follows (DIN 49):
Table&-8 TSCData Parameter                                            Value TSC Volume (ft3)                                    113,412 HVAC Flow Rate (dm)                                  37,000 Recirculation Filter Flow Rate (cfm)                  6,000 Charcoal Filter Bed Depth (In)                          2 Filtered Damper lnleakage (cfm)*                        12
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn        Page 13 Parameter                                                  Value Unfiltered lnlealcage (cfm)..                                27.2
                        *Added to recirculation flow through charcoal fitter
                        **After recirculation *is initiated at 60 minutes (Includes 10 cfm for ingress and egress)
For the TSC charcoal removal efficiency, a removal efficiency of 80% will be used to provide margin as was done for the Control Room charcoal adsorber removal efficiency. Note that In the RADTRAD files, the TSC Is labeled as the Control Room. The normal HVAC flow is assumed to operate for the first 60 minutes after which it is isolated and the recirculation filter Is initiated.
6.9      Mixing Between the Unsprayed Containment and Sprayed Containment The. mixing rate between the unsprayed containment and the sprayed containment was determined to be 71,400 cfm in Calculation PSAT 04202U.03, Rev. 0 (DIN 13).
6.10 Containment Leakrate The maximum allowable primary containment lealuate, La. is 0.2 volume percent per day at the peak containment pressure (Pa) of 7.80 pslg perTechnlcal Specification 5;5.12 (DIN 15). Per Assumption 6.4, primary containment leakage Is released direc:tfy to the environment.for the first 40 seconds following the LOCA, when the shield building may not be at a negative pressure. SubsequenUy, the annulus will collect and retain any fission product leakage from the primary containment and will release.fission products to the environment through the AEGTS. The leakrate for the sprayed and unsprayed regions of the containment is calculated below:
[(4.812e + OS f t 3 )
* 0.2'6]
Lealcrate Sprayed to Bn11lronment        =                  mba z4hr*6orr Lea/crate Sprayed to Bn11iromnent        = 0~668 cfm
[(6.842e + 05 /t 3)
* 0.296)
Lea/crate Unsprayed to Bn11ironrnent      =                    min 24hr*60rr Lea/crate *uflSJITayed to Environment      = 0.950 cfm Secondary Containment Bypass leakage is leakage in addition to La. Technical Specification SR 3.6.1.3.9 (DIN 15) limits the secondary containment bypass leakage to equal to or less than 5.04 percent of the *primary containment leak rate. The containment bypass leakage for this calculation is assumed to be 0.1008 L..(Assumption 6.9)"
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term Page 14 Sprayed Region Bypass Leakage = 0.668 cfm
* 0.1008 = 0.067 cfrn Usprayed Region Bypass Leakage      =0.950 cfrn
* 0.1008 =0.096 cfrn Total Bypass Leakage = 0.163 cfm Table6-9 Containment Leakrate Summa Leakrate ft3/min Oto40sec.          40 sec. to 24 hrs 0.668                  0.0 .
0.0                  0.688 0.950                  0.0 0.0                  0.950 0.067                  0.067 0.098                  0.098 6.11  Leakage after 24 Hours Containment leakage depends upon containment p1888ure and will be reduced at 24 hours as allowed by Regulatory Gulde 1.183, Appendix A, Section 3.7. Based on the post-accident containment pressure cuNe for a MSLB (DIN 27), the containment pressure at 24 hours post-accident Is 18.1 psla. This value was obtained by digitizing the containment pressure cuNe and finding the pressure at 24 hours. Atmospheric pressure at the elevation of the PNPP site (620 ft AMSL) Is 14.37 psla. Because the flow rate is proportional to the square* root of the differential pressure, the reduction In flow rate may be estimated as follows (assuming all other* parameters remain constant):
                              . 1- .. ~Ji.        .JlS.1-14.37 =0.691=>0.69 L0      .I!        J1.8
 
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Loss of Coolant Accident Radiological Analyses Using Alternative Accident Source Tenn  Page 15 Because the secondary containment bypass leakage also depends upon containment pressure, this leakage rate will also be reduced to 69% of the Technical Specification alloWable values beginning 24 hours post-accident. Therefore, this analysis will reduce all leakage flows to 69% of the Technical Specification allowable value after 24 hours.
Tablel-10 Containment Leakrate after 24 hours Leakage                            Leakrate ft'lmln 0.461 0.656 0.04623 0.0662 6.12 MSIV Flows The flows given In Section 5.9 are based on MSIV leakage rate testing requirements at standard conditions. The drywall atmosphere will not be at standard conditions after the reactor blowdown.
Calculation PSAT 04202H-04 (DIN 12) determined that the total MSIV flow rate from 0 to 7484 seconds was 298 cfh based on the minimum post-accident drywell pressure of 15.7 psla and minimum temperature of 215&deg;F. From 7484 to 88400 seconds, the MSIV flow rate Is 247 cfh baaed on the minimum pressure of 15.7 psia and temperature of 100&deg;F.
Based on the assumed .ftow split given In Section 6.9, the flow through the broken steam Ona.* is:
298 cfh 100 scfh Qrwo1cenu...Ct S 7484seconds) =      mlr&/hr
* 250scfh = 1.987 cfm 60
                                                .              247 cfh 100 scfh Qrwo1cen Hne{7484 < t S 86400 seconds) =        min/hr* 250 scfh = 1.647 cfm 60 The flow through the Intact steam lines (100 scth through a second Intact steam line and the remaining 50 scfh through a third Intact steam line) Is given below.
                                    .                    298 cfh 150 scfh Q1ntact1m.s(t S 7484 seconds)= 60 mln/hr
* 250 scfh = 2.98c/m 247 cfh 150 scfh QantactHnesC7484 < t S 86400 seconds)= 60 min/hr* 2S0scfh        = 2.47 c/m
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term      Page 16 Consistent with Section 8.2 of Appendix A to R.G. 1.183, this leak rate may be reduced by as much as a factor of 2 after 24 hours, if supported by plant speciftc analysis. As calculated above, this analysis will reduce all initial leakage flows to 69% of the Technical Specification allowable leakage after 24 hrs. The MSIV leakage flows at this time become:
Table 8-11 Steam Lina Leakrate after 24 Hours Leakage Path                            Leakrate Ctt'/mln)
Broken Steam Line                                            1.371 Intact Steam Lines                                          2.056 6.13    Natural Removal Mechanisms Natural removal mechanisms for elemental iodine and aerosols will be applied in this calculation using NRC correlations incorporated into the RADTRAD 3.03 code.
6.13.1 Elemental Iodine Removal Elemental iodine removal is credited in the drywall and containment volumes. Airborne elemental iodine is removed by deposition to the walls In the drywall and containment As reported in Section 5.1.2 of NUREG/CR-0009, DIN 33), this process is driven by the temperature differences between the surfaces and the atmosphere. The removal factor reported in *NUREG/CR-0009 Is given by the following equation.
Kg A A=-
V where:
1 = removal rate constant due to surface deposition, Ir,,= average mass .transfer coefficient 0.137 cm/a (16.18 Mir) from page 17 of NUREG/CR-0009, A = surface area for wall deposition, and V = volume of contained gas.
This formula is also reported In Standard RevieW Plan 8.5.2 (DIN 34) as a method of calculating the total elemental iodine removal capability. These removal constants are applied until a decontamination factor (OF) of 200 has been obtained.
Volume and Area Calculations Drywall Volume For all volume calculations, surfaces other than the Inner and outer building wall will be con881V8tively neglected. The PNPP drywall volume of 278,500 tt3 from CEI Cslculatlon 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31) is used in this calculation.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Altematlve Accident Source Tenn              Page 17 Wall Surface Area Considering the 36'6* Inside radius (DIN 45) of the drywall cylinder and the approximately 68 foot height above*the suppression pool high water level (DIN 51), the area of the Inner drywall wall ls calculated to be 15.oOO ft2. The use of the suppression pool high water level Is conservative because it minimizes the wall surface area available for deposition.
Area = n:Dh = n:
* 73'
* 66' = 15136.2 ft 2 or, 15,000 f t 2 Sprayed Containment Region Volume AlthougJl in some parts of the containment, the containment spray would fall directly to the suppression pool, the refueling floor (grating) at El. 689&deg;-&* would affect a large fraction of the containment spray. As such, the only containment volume credited with spray removal is that area above the refueling floor. The upper containment (sprayed) region volume of 481, 174 ftS from CEI Calculation 3.2.6.4, Revision 0, Page 3A of 33 (DIN 31) is used in this calculation.
Wall Surface Area The surface area Is taken as the containment wall area above the refueling floor at 689'-6* (DIN 47) and below the containment spring line at 727' (DIN 47). Using the containment radius of 60' (DIN 45) the surface area Is calculated below as 14, 137 tt2.
Area  = n:Dh =  n:
* 2
* 60' * (727' - 689.S') = 14,137 f t 2 Tha* surface area of the oblate elliptical spheroid.above the spring line is given by: (DIN 32) c (1+e) 2 S= 2ira2 + ir-ln -
e    1-e Where *a* is the equatorial radius and *c* 1s the polar radius and the elllptlclty, *e* is given by:
                                                        . R2 e= 1 -az-Substituting 60' for *a* and (757' - 727'    =30') for *d' (DIN 47) gives:
(30') 2 e= 1- (60')2 = 0.866
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn      Page 18 S - 2rr (60
                                              ')Z
                                                  + rr (30')
2 (1 + 0~866)- _.
0.8 ln _ 0.866 - 31,218 t f z 66    1 The surface area of the dome Is half of this total area: S  =31,218 ft'l2 =15,809 tt2 The total area available for plateout is therefore 29,746 ft' or, 29,000 ft'.
Unsprayed Containment Region Volume The volume of the unsprayed containment region Is 684,226 ft" per from CEI Calculation 3.2.8.4, Revision 0, Page 3A of 33 (DIN 31 ).
Wall Surface Area Considering the 41'6. outside radius of the drywell (DIN 45) and the approximately 96 foot height (689'-6*
            =
- 593'-4. 96.17') above the suppression pool high water level (DIN 51 ), the area of the outside drywell wall is calculated to be 25,000 ft2. The use of the suppression pool high water level Is conservative because It minimizes the wall surface area available for deposition.
Area = JrDla =  ff. 2. 41.5'
* 96'  = 25,032 /t2 OT, 25,000 /t 1 The radius of the unsprayed containment wall ls 60' giving a surface area of Area = rrDla = n
* 2
* 60'
* 96' = 36,191 /t2 OT,      36,000 /t1 This gives a total surface area of 81,000 ft" Using the above wall areas and volumes, the removal rate constants are given below:
Table&-12 Elemental Iodine De          ltlon Removal Factors Removal Volume      Wall Area          Factor Node                        rt                            hf1 278500                          0.878 481174                            0.976 684226                            1.443 Airbome elemental iodine removal by deposition to the walls in the drywall and containment Is assumed to end when a OF of 200 is reached.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term    Page 19 6.13.2 Aerosol Removal Aerosol removal In an regions of the containment and drywall is modeled using the Power's removal model supplied with RADTRAD 3.03 as discussed in Subsection 2.2.2.1.2 of NUREGICR-6604 (DIN 2).
The realistic 90% lambda values are applied consistent with Section 3.2 of Appendix A to R.G. 1.183.
Using overly conservative removal coefficients is not nec:eSsary considering the already conservative accident scenario and other modeling considerations. Aerosol removal Is considered only In the drywall and unsprayed region of the containment because the containment spray will adversely impact the particle size distribution In the sprayed region of the containment Aerosol Spray Removal A simplified model for estimating the fission product aerosol removal by containment sprays following a postulated LOCA is used. The model for aerosol removal by sprays bultt into the RADTRAD 3.03 code is the Powers model. The model was developed using values of 10, 100, and 2500 cm3 Hz()/ cm2-s for the spray water flux. The model should not be used for spray water fluxes and fall heights outside of these ranges (DIN 2, 22, and 23). The Powers model was derived by correlating the resutts of Monte Carlo uncertainty sampling analyses assessing the uncertainties in aerosol properties, aerosol behavior, spray droplet behavior, and the initial and boundary conditions expected to be associated with a postulated LOCA In the containment The Powers mechanistic model requires that the user specify the following:
: 1. Q, the spray water flux, in cfm/sq ft;
: 2. H, the fall height, In meters;
: 3. ALPHA, the ratio of unsprayed volume to sprayed volume,
: 4. PCT, the uncertainty percentile selected for the model (10th, SOth, 90th percentiles).
                                                  =
The spray alpha is-6.8423E+05/4.8117E+05 1.422. The other two parameters used In this evaluation that are not treated as uncertainty distributions for Perry are (1) spray water flux, and (2) mean spray fall height These parameters are specified based on plant specific design Information. The 1>est estimate*
value Is associated with the 50th percentile, or median values; the lower bound is assoclalad with the 10th percentile; and the reasonable upper bound, or largest dec:ontamination factor (OF), with the 90th percentile. For spray removal, the RADTRAD Powers Model 10th percentile uncertainty dlatrtbutlon for fission product In aerosol form is used in this analysis. Note that the reactor and accident type used In the Powers aerosol model must be reset to *BWR-Design Basis Accident" prior to each execution of the RADTRAD code.
The PNPP LOCA dose analysis credits spray removal of aerosols In the sprayed region of the containment. The Powers spray removal model Implemented In RADTRAD requires the spray flux and spray height as Inputs. The spray flow Is 5250 gpm (D-302-0661-00000, Rev. G, DIN 36) per train.
Technical Specification 3.6.1.7 requires that the spray flow from the RHR system be 2:5250 gpm.
Because the Powers model spray removal is a direct function of spray flow rate, a lower bound of the spray flow (i.e., 5250 gpm) Is conservative.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term              Page 20 S    ay Flwc = Spray Flow = 5250 gpm
* 0.1337 cfm/mnn =                  o.06206 cfm 1 pri -      _ Sprayed Area                    n: * (60')2                            I sq ft The average droplet fall height is dependent on the available train of containment spray. As shown below, the headers for the *A* Train are located above the headers for the *e* Train per drawing D-320-0661-00000, Rev. G~ If the flow rate through all nozzles Is assumed to be equal, the average drop height can be calculated by the nozzle-weighted average of the drop heights. The average drop height Is used because the train operating post-accident is unknown. The drop height is based on the distance above the operating floor at El. 689'.a* (DIN 51).
Table 8-13 p NPP c ontainment Sprav Hela    i htB RHR          Header          Header      Reference Drawing        Haight-Hi        Number of            Nf'Hi Train    Designation      Elevation                                  (ft)      *Nozzlas7 - N.  -
(ft)
A            A          735.250      D-314-661, Sheet 3,        45.75            129            5901.75 Rev. B. IDIN 36) c          744.250      Ss-304-861, Sheet          54.75            113            6186.75 105.2, Rev. C, (DIN 38)
E          750.500      Ss-304-861, Sheet          61.00            102              8222 102.2, Rev. B, (DIN 40)
B            B          737.000      D-314-661, Sheet 8,        47.50            129              6127.5 Rev. B. COIN 37l D          745.750      D-314-661, Sheet 7,        56.25            113            6358.25 Rev. B. COIN 39)
F          752.000      D-314-661, Sheet 6,        62.50            104              6500 Rev. B. lDIN No. 41)
Total            890 Average (ft)          54.05 7
32C>-0661-00000, Rev. G (DIN 35)
H    - I,H,N,
                                                    ""IJ -  l:a N, where:
N1 is the number of nozzles on header I H1 is the height of header I above the operating floor (ft)
The average fall height for both trains combined is therefore 54.05 ft.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Altematlve Accident Source Tenn    Page 21 As discussed in SRP 6.5.2 (DIN 34), beaiuse the removal of particulate material depends markedly upon the relative sizes of the particles and the spray drops, the aerosol spray rem0val lambda is assumed to decrease by a factor of 10 after the aerosol mass has been depleted by a factor of 50.
Elemental Iodine SRP 6.5.2 provides guidance on calculating the spray lambda for removal of elemental iodine. The following fonnula is valid for lambdas greater than 1Oper hour with a maximum of 20 per hour to prevent extrapolation beyond the existing data.
where:
A.= first-order removal coefficient by spray, kg = the gas-phase mass-transfer coefliclent,
                            =
T the fall time of the drops, which may be estimated by the ratio of the average fall height to the tennlnal velocity of the mau-mean drop,
                            =
F volume flow rate of the spray pump,
                            =
V containment building net free volume, and D = mass-mean diameter of the spray drops.
Gas Phase Mass Transfer Coefficient The gas-phase mass-transfer coefficient. kg, can be detennlned by back-calculation from a solved case with slightly different assumptions. Speclfically, the example on Page 108 of NUREG/CR-0009, "Technological Bases for Models of Spray Washout of Airborne Contaminants In Containment Vessels",
1978, (DIN 33) uses the stagnant film mOdel to determine the spray removal coefliclent for a PWR case with a 1713 spray nozzle and the following parameters.
A.= 14.2 hf1 F= 1500gpm V= 1.75E6.ft3 Height = 90 ft Temp=250&deg;F Solving the above equation for kg, gives:
 
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Loss of Coolant Accident Radlologlcal Analyses Using Altematlve Accident Source Tenn    Page 22 To calculate kg, the values of the mass-mean drop diameter, D, and the faU time of the drops, T, are needed. The PNPP spray nozzles are Spraco 1713A nozzles (DIN 35). Recent test results with the Spraco 1713A nozzles presented In F1gure 4 of NUREG/CR-5966 (DIN 6) have a mean droplet size of 234 J,lm (NUREG/CR-5966, page 7) and an upper diameter of about 1500 J,1m. The mass-weighted .
average drop size, however, will be larger than 234 microns since the larger drops have exponentially more mass. This volume-weighted size distribution (which Is directly related to the mass-weighted distribution) is reported In Figure7 of NUREG/CR-5968 which Illustrates an average of the volume weighted distribution to be approximately 1200 microns. A value of 1200 microns will be applied in this calculation. The terminal velocity of 1200 J.lm drops can be found to be approximately 400 cm/s from Figure 16 of NUREG/CR-5966. Conservatively assuming the velocity Is equal to the terminal velocity, a 90 foot (2743 cm) fall height gives a fall time of 6.86 seconds. Using the above data to determine the gas-phase mass-transfer coefficient, kg. gives:
As
* V
* D    14.2 hr-1 60b~l8
* 1.7SE6 ft3
* 1200
* 10-6 m
* 100 cm/m      cin kg= 6*T*F =                    908:            cm                      ft:3 6
                                                                                            = ;;c 6
* 400 em/sec
* 30.48 l t
* 1500 gpm
* 0.1337 jil For PNPP; the average fall height of the spray drops is calculated to be 54.05 ft (1647 cm). The terminal velocity of 1200 J,lm drops can be found to be approximately 400 cm/s from Figure 18 of NUREG/CR-5966. The drop fall time is calculated to be 4.1 seconds. The spray flow Is 5250 gpm from ~-0661-00000, Rev. G, (DIN 35) and the sprayed volume of the containment Is 481, 174 tt3 from CEI calculation.
3.2.8.4, Revision 0, page 3Aof33 (DIN 31). From the SRP equation, below, the PNPP spray lambda.for elemental iodine can be calculated to be 107.68 hr.
cm                              ft3    min 6 *kg
* T
* F    6
* 6sec
* 4.1 sec* 5250 gpm
* 0.1337 pr* 60nr _              _
107 66    1
            .t,, =    V*D      =        481,174 ft3
* 1200
* 10-6 m
* 100 cm/m        -
* hr
* This result is reasonable considering the 14.2 hf1 value calculated for the PWR case described in
* NUREG/CR-0009, the much higher spray flow rate at PNPP, and the smaller sprayed volume.atPNPP.
Since the SRP allows a maximum lambda of 20 hr"1
* this calculation will apply a spray removal lambda of 20 hr"1 for elemental Iodine. As discussed previously, elemental Iodine Is removed by deposition *to the walls in the drywall and containment with a removal coefficient of 1.835 hr"1 for the sprayed region which gives a total elemental Iodine removal coeflicient for the sprayed region of containment as 21.835 hf1*
As discussed In SRP 8.5.2, the maximum decontamination factor is 200 for elemental Iodine. The effectiveness of the spray in removing elemental iodine will be presumed to end at that time, post-LOCA, when the maximum elemental iodine OF is reached.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 23 6.14 Annulus Model The Annulus Exhaust Gas Treatment System (AEGTS) is an engineered safety features system designed to collect, process, and release the fission product leakage from the primary containment into the shield building. The system is operated continuously during normal operation and maintains a slight negative pressure in the shield building. The AEGTS is a redundant system consisting of pre-HEPA filters, charcoal adsorbers and post-HEPA filters. Reduction in release activity by ESF ventilation filtration systems may be credited where applicable if filter systems used in these applications are evaluated against the guidance of Regulatory Guide 1.52 (DIN 20). The AEGTS charcoal adsorbers are not credited for reducing the released activity, so testing in accordance with R.G. 1.52 is not necessary.
The AEGTS HEPA filter is tested in accordance with Regulatory Guide 1.52 to verify a penetration and system bypass of less than 0.05% (DIN 21). Aerosol removal by the HEPA filters is therefore assumed to be 99%. As discussed previously, no credit for charcoal filtration of the annulus exhaust is taken in this calculation.
6.15      Deposition in Main Steam Lines The deposition in the main steam lines will use the aerosol removal efficiencies from PSAT 08401T.03 (DIN 25) which was based on PSAT 04202H.08 (DIN 42). These removal efficiencies include a 10%
increase in aerosol penetration to add conseNatism to the main steam line leakage pathway. The removal efficiencies are given below.
Table 6-14 Main Steam Line Removal Fractions MSL1                  MSL2 failed steamline        i e to Intact steamlines The elemental iodine removal efficiency is 0.45 for all steam lines.
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn  Page 24 7.0      ACCIDENT SCENARIO AND CHRONOLOGY 0 minutes to 40 seconds A design basis double-ended guillotine break occurs in a main steam line upstream of the Inboard MSIV, releasing reactor coolant to the drywall. The dryweD Is pressuriied driving drywall atmosphere out the MSIVs and into containment via the drywall bypass. All MSIV and containment leakage Is Initially directed to the environment. The control room and offslte dose points begin to accumulate dose from the ECCS, MSIV and containment leakage. The control room normal ventilation mode is assumed to run during this 40 second period.
* 40 seconds to 30 minutes The AEGTS system achieves a 0.25-inch vacuum in the secondary containment at 40 seconds (assumption 8.4) and draws 2000 cfm of secondary containment atmosphere through a HEPA filter and charcoal bed before release to the environment. All primary containment leakage is dlrectad to the secondary containment at this time except for the containment bypass leakage, which la assumed to bypass secondary containment and is released directly to the environment No credit for elemental or organic Iodine removal by the AEGTS charcoal adsorbers is taken. Particulate removal by the HEPA filters is assumed to be 99% In accordance with Regulatory Gulde 1.52. The control room normal ventilation mode Is assumed to run during this period.
The gap release begins by releasing the gap source terms Into the drywall at a constant rate over the 30-mlnute release period following the onset of gap release at two minutes post-accident ECCS leakage Is assumed to begin at this time leaking contaminated suppression pool water (10% of Iodine - all forms) directly to the environment even though the postulated core damage Is occurring because no ECCS injection Is assumed to be available during the first two hours.
30 minutes to 2 hours At 30 minutes, the control room normal ventilation syst8m is manually Isolated, and the CRERS is manually Initiated. The CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA filters and charcoal adsorbers. The in-vessel release begins at 30 minutes by releasing the in-vessel source terms Into the drywall at a constant rate over the 90-mlnute release period. Manual Initiation of containment spray is assumed at thirty minutes. Manual Initiation of containment spray at 30 minutes Is reasonable based on Emergency Operating Procedure guidance requiring operation of containment spray based on the *Pressure Suppression Pl888ure* curve contained in the EOP. The containment pressure threshold is met within 30 seconds of the LOCA per DIN 27.
2 hours to 24 hours The source term release from the vessel is terminated at 2 hours with the actuation of ECCS, which results In large amounts of steam evolution and large flows out of the drywall Into the containment. The
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 25 drywell and lower containment region are assumed to become well-mixed at 2 hours. The hydrogen mixing system is assumed to be actuated at 2 hours to increase operator flexibility and will be incorporated in the licensing basis (see AttaChment.5). A sensitivity study provided in Atlachment 2 demonstrates that initiation of the hydrogen mixing system at 30 minutes post-accident results in slightly lower airborne leakage doses (19. 7 rem TEDE at the EAB, 4.9 rem TEDE at the LPZ, and 1.66 rein TEDE in the Control Room).
24 hours to 30 days Releases to the environment via the containment bypass, MSIV leakage, ECCS leakage and AEGTS exhaust continue for 30 days. As discussed above, containment leakage Is reduced at 24 hours based on containment pressure.
8.0    MODEL DEVELOPMENT This analysis considers the following three pathways through which source terms can be released from the containment.
* ECCS liquid leakage outside of containment
* MSIV leakage
* Containment airborne leakage (containment bypass and containment leakage)
These three pathways are discussed below. RADTRAD modeling capabilities allow Incorporation of the MSIV leakage and the containment airborne leakage into one model, therefore; the three release pathways are addressed In two RADTRAD models.
8.1    ECCS Uquld Leakage 8.1.1 Soun:e Tanna The gap and core activity Is released to the drywall atmosphere based on the ralease fractlon8 and timing raported in Tables 1 and 4 of R.G. 1.183 and is assumed to be Immediately dissolved In the suppression pool. Only halogens are modeled In this analysis. Noble gases are not soluble and, with the exception of Iodine, all other radioactive materials in the recirculating liquid should be assumed to be retained in *the liquid phase. This Is consistent with the guidance of RG. 1.183, Appendix A.
8.1.2 Volumes The suppression pool inventory expected during the LOCA Is 114,379 ft' (DIN 11). No credit Is taken for holdup in the Auxiliary Building where the ECCS systems are located.
8.1.3 Flows
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Term    Page 26 The earliest that the containment spray system could potentially be automatically lnltiatad to spray the containment Is 10 minutes post-accident if high containment pressure combined with other LOCA signals Is sensed. For this calculation it is conservatively assum8d that the ECCS system leakage begins immediately after the LOCA (at the beginning of the gap release at two minutes post-accident).
Leakage from ECCS systems located outside the primary containment Is estimated to be 5 gph. For conservatism, this analysis assumes that the leakage is twice the established administrative limit of 7.5 gph (I.e., 15 gph, 0.0334 cfm) for the entire duration of the acctdent. Additionally, leakage from a gross failure of a passive component is assumed to occur at a rate of 50 gpm (6.68 cfm) starting 24 hours into the accident and 18sting 30 minutes in accordance with NUREG-0800 (DIN 44).
8.1.4* Removal Mechanisms Because the suppression pool temperature wlU not exceed 212&deg;F (DIN 11) during the accident, ten percent of the Iodine In the ECCS leakage Is assumed to become airborne consistent wfth R.G. 1.183, Appendix A Natural removal mechanisms and holdup in the auxiliary building are conservatively neglected. Consistent with Section 5.6 of R.G. 1.183, Appendix A, the chemical species of these airborne source terms is assumed to be 97% elemental and 3% organic.
8.1.5 Model The ECCS liquid leakage model is Illustrated In Figure 1.
8.1.6 R*ulls The radiological doses for the ECCS liquid leakage transport path are reported below. The RADTRAD output file, Including the Input summary, Is listed In Attachment 1.
Tablel-1 ESF Uauld Leakaae Dose Resulta Location              Dose fRem TEDE)
EAB                                  0.79 LPZ 1.53 Control Room 0.99
 
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Loss of Coolant Accident Radiological Analyses Using Altemative Accident Source Tenn    Page 27 8.2 . MSIV Leakage 8.2.1 Source Tenns As discussed previously, the PNPP core soun:e tenns have been developed with the ORIGEN2 methodology. These source terms are released Into the drywall based on the release fractions and timing reported in Tables 1and4 of R.G.1.183.
8.2.2 Volumes This analysis assumes a double guUlotlne pipe rupture In one of the four main steam lines upstream of th~ inboard MSIV and failure of all four main steam shutoff valves (1 N11-F0020A, B, C, and D) valves to close as a result of a common power failure (single-failure criterion). The maximum allowable MSIV leakage of 250 scth Is modeled to occur through two pathways: (1) through the broken steam 6ne and, (2) through the second and third Intact steam lines. The volume of the ruptured main steam line between the MSIVs is 146 tt3 (DIN 11 ). Leakage past the second MSIV in this One Is released dlradly to the environment. The volume of the two intact steam lines between the reactor vessel and the Inboard MSIVs Is 440 ft3 (DIN 25). The leakage past the first MSIVs in these lines is released to the volume between the first and second MSIVs which Is 292 ft3, two times the volume between the MSIVs in one steam line (146 ft3) (DIN 11). Leakage past the second MSIVs in these lines is also raleased directly to the environment. This configuration was previously identified In DIN 17 to be limiting with respect to dose consequences.
8.2.3 Flows
.This calculation will apply a maximum MSIV leak rate of 250 acth with the worst-case main steam line leaking no more than 100 scfh. The leakage limit Is assumed to occur: (1) 100 scfh through the broken steam line, (2) 100 scfh through a second intact steam line, and (3) the remaining 50 scfh through a third Intact steam line. As stated above, leakage Is modeled to occur through two paths, one path consisting of the broken steam line and a second path consisting of the second and third intact steam lines. All leakage past the outboard MSIVs Is assumed to be released to the environment.
The drywell atmosphere will not be*at standard conditions after the reactor blowdown. The MSIV leakage rate must be converted to a ftow at the drywall conditions. The MSIV leakage rates at the drywall conditions were datennlned from PSAT 04202H.04 Rev. O(DIN 12). The leakage rates are reduced at 2 hours when well-mixed conditions between the dryweU and primary containment apply.
Additionally, the flow in the main steam lines past the inboard MSIV is represented as wall-mixed. The total MSIV leakage rates (DIN 1 and 11) are 298 cfh for the first two hours and 247 cfh thereafter. The maximum flow rate Is 191 cfh (DIN 11) through any single main steam line to the environment The values used In the modal are given in Sectionl.12.                                                      II~ 'P*l'I
 
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Loss of Coolant Accident Radiological Analyses Using Alternative Accident Source Tenn    Page 28 In addition to the leakage through the MSIVs, the drywall will also continue to leak activity Into the eontalnment over this 2 hour period. This calculation will assume a leakage rate of 3000 cfm for the drywall bypass flow consistent with PSAT 08401 T.03 (DIN 25).
8.2.4 Release Points All MSIV leakage past the outboard MSIV is assumed to be released directly to the environment No credit for holdup in the auxiliary building or turbine building is taken.
8.2.5 Modal The RADTRAD model applied for this leakage path, as well as the containment airborne leakage path, is illustrated in Figure 2.
8.3      Containment & Containment Bypass Leakage 8.3.1 Volumes In addition to the main steam lines, the following volumes are used in the LOCA airborne leakage dose calculation (DIN 11 and DIN 31):
Table8*2 LOCAVolumes Volume                                    Description                  Volume( )
in Modal 1                                                                    2.785E+05 5                                                                    4.812E+05 6                                                                    6.842E+05 7        Annulus                                                    1.96E+05
: 9.        Control                                                      390,020 Room The volume of the Perry control room has recently been nMV&luated. The current volume to be used in Control Room Dose calculations is 390,020 ft3 (DIN 16).
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn  Page 29 8.3.2 Flows From Drywell Volume Into Containment (Suppression Pool Bypass)
The flow rate from the Drywall to the Wetwell are given below, see PSAT 04212H.02 (DIN 43):
Table 8-3 DJryweII Flows Time After Gap          Flow from DW to WW        Flow from WW to DW Release                      (cfm)                        (cfm)
(hours)
                          *o-o.s                        0                          0 0.6-2.0                      .3800,3000                  .eao 2.0-720                    2.77E+06                    2.77E+05 At two hours, the drywall and unsprayed portion of the containment will be assumed to become Instantly well-mixed without credit for suppression pool scrubbing In accordance with Regulatory Gulde 1.183, sec.T 3.S' ANb Section 3. 7 (DIN 7).
From Unsprayed and Sprayed Containment Volumes to the Environment By the time the gap release begins, the containment is completely Isolated and only containment leakage Is assumed. The design basis containment leakage for Perry Is 0.2% per day. Since the AEGTS will not completely draw down the annulus for 40 secbnds, a 40 second positive pressure period Is assumed In which all containment leakage Is assumed to leak directly to the environment.
From Unsprayed and Sprayed Containment Volumes to the Annulus After the 40 second positive pressure period, the majority of the containment leakage Is drawn into the annulus by the AEGTS. Although the primary containment Is enclosed by the secondary containment, there are systems thatpenetrete both the primary containment and the shield building boundaries that could create potential pathways through which fission products In the primary containment could bypass the leakage collection and filtration systems associated with the shield building. The Perry Technical Specification SR 3.6.1.3.9 (DIN 15) limit the secondary containment bypass leakage to equal to or less than 5.04 percent of the primary containment leak rate. This analysis uses a bypass leakrate of 10.08 percent of the primary containment leek rate.
Mixing Between the Unsprayed Containment and Sprayed Containment The mixing rate between the unsprayed containment and the sprayed containment Is 71,400 cfm, Calculation PSAT 04202U.03, Rev. 0 (DIN 13).
 
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Loss of Coolant Accident Radiologlcal Analyses Using Alternative Accident Source Tenn      Page 30 From Secondary Containment The only flow from secondary containment Is via the AEGTS system which draws 2000 cfm through a charcoal-filter unit and HEPA filter unit. The HEPA filters are tested per Regulatory Gulde 1.52 and therefore are credited for a 99% removal efficiency in the analysis; however, no credit Is taken for the charcoal adsorbers in this analysis.
8.3.3 Removal Mechanisms Natural removal mechanisms for elemental iodine and aerosols will be applied in this calculation using NRC correlations. Elemental Iodine removal is credited in the drywall and containment volumes. Aerosol removal is credited only in the drywell and unsprayed region of the containment since containment spray will adversely impact the particle size distribution In the containment.
Fission product removal by containment sprays Is considered. The Perry containment spray system is initiated manually based on high radiation readings or is Initiated automatically approximately 10 minutes following a LOCA based on pressure and low water level. In this calculation, sprays* are assumed to be manually Initiated at 30 minutes. The Powers model for aerosol removal by sprays which Is built Into the RADTRAD code Is used In this analysis. Consistent with the guidance in Section 3.3 of Appendix A to R.G. 1.183, the maximum spray decontamination factors for elemental Iodine ls 200. After the aerosol mass has been depleted by a factor of 50, the spray removal lambda Is assumed to decr8aae by a factor of 10. After the elemental Iodine activity has been depleted by a factor of 200, the elemental Iodine removal Is assumed to end completely.
The foUowlng section determines when these DFs were determined to occur. As discussed In Section 3.3 of Appendix A to R.G. 1.183, these Dfs are based on the Inventories at the end of the In-vessel release phase. Containment spray Is assumed to end at 24 hours and the aerosol removal by containment spray is terminated.
Pecongmlnatlon Factor Reductions As discussed above, the elemental iodine removal by natural deposition Is neglected after a OF of 200 is reached. Based on the elemental iodine lambda of 0.878 hf' In the drywall, a OF of 200 would be reached In approximately six hours without any leakage or decay. The output In Attachment 2 Indicates that the Drywell 2-hour post.gap release (I.e., 2.0333 hr) elemental 1-131 Inventory of 2.29E+22 atoms has reduced to 1.15E+20 at 3.0333 hours after gap release representing a DF of 199. Thia calculation will therefore model the elemental Iodine removal to end at 3.0333 hours In the drywall.
In the containment, the total (sprayed+ unsprayed) elemental 1-131 inventory Is 3.48E+21 atoms after the drywall is flushed at two hours after gap release (I.e., 2.0333 hr). The total activity In both regions of the containment is considered because, if only the activity In the sprayed region of the containment was considered, a longer period of the higher lambdas would be applicable. At 3. 75 hours, the total elemental
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn      Page 31 1-131 Inventory Is 1.98E+19 atoms, representing a OF of 175. This calculation will model spray removal to end at 3.75 hours. Natural deposition of elemental iodine in the unsprayed region of the containment Is also assumed to end at this time.
The particulate removal (Powers Model) in the sprayed region of the containment is reduced by a factor of 10 when the ~erosol activity Is reduced by a OF of 50. In the containment, the total (sprayed+
unsprayed) particulate inventory is 9.05 kg after the drywall is flushed. At 4.85 hours, the total aerosol inventory is 1.21E-1 kg, representing a OF of 48.7. This calculation will model this removal coefficient to be reduced at 4.85 hours. This is accomplished by reducing the spray flow used In the PolNers Model by a factor of ten at this time.
8.3.4 Rel-* Points All source terms released via containment leakage are released through the plant vent.
8.3.5 Model The containment alrbome model is illustrated in Figure 2 which Is based on the time at which the gap release begins. This figure also Includes the MSIV leakage transport pathways.
8.4      Control Room Although the current configuration of the control room HVAC system would automatically Initiate the control room recirculation on a LOCA signal, this analysis assumes that the CRERS Is manually initiated at thirty minutes. Once the CRERS is initiated, CRERS fans recycle 27,000 cfm of control room atmosphere through HEPA fitters and charcoal adsorbers before being returned to the contlol room. The normal control room recirculation air flow Is 45,000 cfm (DIN 52) Including 8,000 cfm of outside air for ventilation. To represent the normal positive pressurization In the control room, the exftltratlon air ftow is modeled as 4,800 cfm before isolation at 30 minutes. The RADTRAD code only allows a single control room recirculation air flow. As a resutt. the normal recirculation air ftow Is not modeled. This is acceptable because the normal recirculation flow does not change the radlonucllde concentration In the control room.
* After isolation, unfiftered inleakage of 1375 cfm is assumed to be drawn from the control room intake duct for the duration of the postulated accident (30 days). The flow from the control room to the environment Is also set at 1375 cfm to avoid pressurization.
Consistent with the requirements of R.G. 1.183, the contribution to the control room dose due to shine from the containment building (0.13 Rem) and release plume (0.002 Rem) must be considered. These dose contributions are given* in PSAT 04202H.13 (DIN 17). These 30-day doses to the control room were generated with the previous TIO 14844 source term that assumed an Instantaneous release to the containment of 100% of the core inventory of noble gases and 50% of the radlolodlnes. These assumptions are conservative compared to the Altemative Source Term methodology (AST) due to
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 32 removal of radlolodines from the containment atmosphere by sprays and deposition thereby reducing the radionuclide concentration in containment In addition, the total halogen release fraction is 0.3 for the AST methodology providing additional margin. Based on these considerations, the previously calculated shine and plume doses are considered bounding for this analysis.
8A.1 Results The radiological doses for the containment airbome pathway (Base Case, Two Hour Hydrogen Mixing System Delay) are reported in the pertinent parts of the RADTRAD output file listed in Attachment 5.
* RADTRAD modeling capabilities allow incorporation of the MSIV leakage and the containment airborne leakage into one model. *The results of the alrbome calculations Including MSIV leakage are summarized below.
Table8-4 Alrbome Leaka1ae Dosa RnuIla.* Baae Caae Location                LOCADoae (remTEDE' EAB                        20.38 LPZ                        4.98 Control Room                1.7 8.5    RADTRAD MODEL The models developed for the analysis are Blustrated In Figures 1 and 2.
9.0    Operator Actions The operator actions assumed In this analysis Include the following:
: 1. Manual lnttlation of containment spray at 30 minutes
: 2. Manual initiation of CRERS at 30 minutes
: 3. Initiation of hydrogen mixing system at two hours
: 4. The pH calculation (DIN 26) assumes initiation of Standby Liquid Control (SBLC) to control suppression pool pH
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 33 Figure 1 Fission Product Transport Model *
(ECCS Leakage Pathway) 1- Pool to Environment SUPPRESSION POOL - 1 ENVIRONMENT-2 114,379 ft3                      end) 8.71 cfm (24-24.5 hrs) 2 - Conlrol Room Intake 6000cfm 1375 cfm afterCRERS lnillatlon 3 - Conlnll Room Exhaust 4800ctn 1375 ctn aftarCRERS lnlllallon CONTROL ROOM - 3 390,020 ft'
 
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Loss of Coolant Accident Radiological Analyses Using Alternative Accident Source Tenn                                            Page 34 Slln&#xa5;ld*5              8* Sprayad ID Anmils 3-UnllprlyadfD 11,400*
                                      ~
4.812t51P I    ~Spr.,.ifD ~
11,400*
                                                                              =
* 9-AEGTS (Arndul ID Envtronmen0 2000dm 1t
____          ...... =~
1.Qiywa1                    ~-a
                    . . . . _ _....  - 8.842e51P
                                      ..                    .., I~---~ ID Amlill T
w,i-----1
                                                                                                            .-                10. CA ln1Bta
                                                                                                                              &OOOdm 1375 dm after CAERS Initiation
* Conlral Roam. 9 390.G201P
                                                                                                                          .... 11.cR Exhlu8l 4800dm          --
1375 dm after CAERS        ---
lnlllallan 18.2*3                          MBIJ*4
                    ...._-~M          440IP        ...__ __.,..:        2921P 13- Oywell  tD                  14"MSL2 11M8.31D
                      ~              - - - - IDMSIJ                                      EINliGUl6t li8.1*2
                    ""'----* 148fP 15
* MSl.1 ID Envlrann8lt 12-0rywall fD
'-------------'MSL.1                .._________,
 
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Loss of CoolantAccldent Radlologlcal Analyses Using Alternative Accident Source Tenn    Page 35 10.0 COMPUTATION The RADTRAD output files, which include the Input summary, are given in Attachments 1 and 2. The RADTRAD input and output files *used for this calculation are identified below:
Description                          ECCS Leakage                  Containment Leakage Plant scenario file                  PNPP ESF EQ.psf                PNPP LOCA EQ.psf Auxiliary RADTRAD Input Files Nucllde* Inventory File              PNPP_ESF EQ.nif                PNPP GNF2_eq.nif Release Fraction and Timing Fiie      pnpp_esf.rft                  PNPP_OBA.rft Dose Conversion Factors              Fgr11&12.inp                  Fgr11&12.lnp Output Fiie                          PNPP ESF EQ.out                PNPP LOCA EQ.out Plant scenario files for the TSC dose calculation are:
PNPP ESF TSC.psf PNPP LOCA TSC.psf The nuclide Inventory files, release fraction and timing files, and the dose conversion factor files for the LOCA and ESF cases are the same as above.
11.0 Overall Results Table 11-1 presents the dose results for Individual leakage pathways for MSIV leakage, containment leakage, containment bypass, ECCSleakage, and shine dose. Control room shine dose Is from DIN 17 and25.
Table11*1 Dose Result&, Sensitivity Case 30 Minute : ._ . n Mbdna Svatem Initiation fram TEDE)
Pathway                      EAB          LPZ        Control Room            TSC Containment & MSIV Leakage                19.7          4.9              1.86              0.35 ECCS Leakage                              0.79          1.53              0.99            0.154 Shine Dose                                                              0.132            0.132*
                  .Total                      20.S          &A                2.8              0.8 Regulatory Limit                  25            25                5                5
                                *Assumed to be the same as the Control Room
 
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Loss of Coolant Accident Radlologlcal Analyses Using Alternative Accident Source Tenn Page 36 Table 11-2 Dc:tae Result&, Base Case (rem TEDE)
Doses with a two hour Hvdroaen Mixlna Svstem Initiation Delay Pathway                      EAB          LPZ        Contlol *Room Containment & MSIV Leakaae              20.38          4.98              1.7 ECCS Leakage                            0.79          1.53            0.99 Shine Dose                                                            0.132 Total                      21.2          6.5              2.8 Regulatory Limit                25            25                5 A separate dose run was not perfonned for the TSC with a two hour hydrogen mixing system initiation time. The TSC dose increases in the same manner as the control room doses given in Tables 11-1 and 11-2. This control room dose increase, 1.66 rem to 1.7 rem, gives a TSC dose of 0.7 rem TEDE.
 
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Losa of Coolant Accident Radiological Analyses Using Alternative Accident Source Term Addendum Page I ofl Operator Action Timing Assumptions Calculation 3.2.15.16 provides an analysis of a stylized LOCA scenario that is Intended to represent a range of possible events, while not purporting to be the only possible sequence of events. To provide a conservative analytical basis, certain assumptions needed to be made concerning the sequence of events and possible operator actions. These operator actions ara not "time crltlcar operator actions, In the sense that lhey need to be codified In plant procedures, or necessarily mandatory actions, but were considered to be typical post-accident actions used to conservatively evaluate event dose consequences. The post-accident operation of the hydrogen mixing system and the containment spray system are addressed In lhls Addendum.
The LOCA dose calculation assumes a Main Steam Une Break LOCA as the limiting, and representative, scenario. This scenario assumes no Injection Into the reactor vessel for the first two hours after the event. At two hours the core is quenched which stops the release from the core.
Regarding the use of the hydrogen mixing compressors, the calculation base case assumes that the hydrogen mixing system Is actuated at two hours. This action was not based on any particular symptom or pivcedural requirement and could be removed.from the calculation without Impacting the calculation results or conclusions.
This conclusion is based on lhe fact lhat the atmospheric mixing between the drywall and containment provided by the mixing compressors at 2 hours Is Insignificant when compared to the ftowrates due to the quenching of the core at 2 hours. Because the hydrogen mixing flow rate Is low (500 dm) compared to the assumed core reflood mixing flow (2.nE+05 cfm, see calculation page 4, which Is approximately one drywall volume per minute),
assumed operation of the hydrogen mixing system at two hours Is Insignificant. Operation of the hydrogen mixing system at two hours post-accident has no Impact on the mixing between the drywall and the containment, the post-accident releases to the environment, or the resulting dose consequences. Therefore, operation of the hydrogen mixing system at two hours Is not considered a "time critical* operator action.
The accident chronology given on page 24 of the calculation states that manual Initiation of the containment sprays at 30 minutes is assumed. As a basis for this timing, the calculation states that *manual Initiation of containment spray at 30 minutes Is reasonable based on Emergency Operating Procedure guidance requiring operation of containment spray based on the *Pressure Suppression Pressure* curve contained In the EOP. The containment pressure threshold Is met within 30 seconds of the LOCA per DIN 27.* The RADTRAD code, used to evaluate the accident consequences, does not model the containment accident pressure response and then Initiate sprays based on. pressure. Instead, spray Initiation Is based on an Input time*whlch Is not necessarily based on any specific physlcal parameter. Given Iha containment response curve for the design basis LOCA (DIN 27) and the PNPP symptom-based EOP guidance to Initiate containment spray when exceeding the Pressure Suppression Pressure curve, a 30 minute time period was deemed a reasonable time frame to Input Into the analysis for determination of conservative dose consequences. However, because the stylized LOCA, as modeled In this calculation, Is not the only possible accident sequence this timing should not be considered to be a -Ume crltlcar operator action.
Section 9 of the calculation provides a listing of assumed operator actions used In the calculation. These operator actions are not "time crtt1ca1* operator actions, but are assumptions made for conservative evaluation of the dose consequences of the postulated event. In an actual accident, operation of these systems will be In ac:cordance with the EOPs.
 
ADDENDUM 5 Summary of Control Rod Drop Accident (CRDA) Dose Calculation
                      *21 pages follow This Addendum is considered a summary of the associated calculation - only selected pages of the approved document are provided.
 
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Control Rod Drop Accident Radiological Analysis with Revised Source Tenns                                            Page I Control Rod Drop Accident Radiological Analysis using Alternate Source Terms TABLE OF CONTENTS SUBJECT OBJECTIVE OR PURPOSE SCOPE OF CALCULATION
 
==SUMMARY==
OF RESULTS/CONCLUSIONS LIMITATIONS OR RESTRICTION ON CALCULATION APPLICABILITY IMPACT ON OUTPUT DOCUMENTS DOCUMENT INDEX CALCULATION COMPUTATION (BODY OF CALCULATION):
1.0 PURPOSE .............................................................................................................................. 1
 
==2.0 BACKGROUND==
..................................................................................................................... 1 3.0 METHODOLOGY ....................... ~........................................................................................... 2 3.1 REGULATORY GUIDE 1.183 COMPLIANCE **...*......*......*..**..*.**.******.........*.*.......*...****.**.***.*..**. 2 3.2 SRP 15.4.9 COMPLIANCE************************************************************************* ......**.*......*.....***.*. 5 3.3 TURBINE AND CONDENSER VOLUMES .*.....................*..........*..*****.....*......*....................*****.... 7 4.0 ASSUMPTIONS ..................................................................................................................... 8 4.1 BROMINE MODELING *****.****..*.***.***.....******....*..***.*.*..****.**....*...****....*****.*......****.**........*******.*. 8 4.2 CONTROL ROOM ISOLATION AND INLEAKAGE ****.*..*...*****....*.**.**....*******..*..*..........***********....... 8 5.0 CALCULATIONS .................................*.............................*...............................~ ...*............. 10 5.1 RELEASE FRACTIONS ......*.*............*............*......*.**......****.......*.****.....*.***........*.......********.... 10 5.2 MODEL .....*.***.*..*...*******.......**.*........*.*...*....*.......**....******..*.......***.*.*..**.*..*....*........******....*.* 12*
5.3 CALCULATIONAL METHODOLOGY *.....*.****.*...*........*******...*.*****..********...*..**.*....*.**...*....*****.*.*. 13 6.0 RESULTS ............................................................................................................................. 13 ATTACHMENTS: : UFSAR Scenario 2 Evaluation                                                                                          58pages : Case 1 output                                                                                                        30pages : Case1a output                                                                                                        44 pages TOTAL NUMBER OF PAGES IN CALCULATION (COVERSHEETS + BODY+ ATTACHMENTS)                                                              152 Pages
 
TITLE/
 
==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenna Page II SUPPORTING DOCUMENTS (For Records Copy Only)
DESIGN VERIFICATION RECORD                                                              1 Page CALCULATION REVIEW CHECKLIST                                                            8 Pages 10CFR50,59 DOCUMENTATION                                                                O Pages DESIGN INTERFACE
 
==SUMMARY==
12 Pages DESIGN INTERFACE EVALUATIONS                                                            2 Pages OTHER                                                                                          Pages EXTERNAL MEDIA? (MICROFICHE, ETC.) (IF YES, PROVIDE LIST IN BODY OF CALCULATION) 0    YES 181  NO
 
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==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenns    Page Ill OBJECTIVE OR PURPOSE:
The purpose of this calculation is to conservatively determine the radiological dose consequences resulting from the reactor coolant release that accompanies a Control Rod Drop accident (CRDA). The total effective dose equivalent (TEDE) at the control room, the exclusion area boundary (EAB), and the outer boundary of the low population zone (LPZ) are calculated using NRC-approved methods. This calculation incorporates the source terms derived from Regulatory Guide 1.183. [DIN# 5)
Revision 1 of this calculation determines the radiological dose consequences resulting from the reactor coolant release that accompanies a CRDA using GNF2 fuel. This calculation will replace the current PNPP CRDA, revision O of this calculation.
SCOPE OF CALCULATION/REVISION This calculation determines the dose consequences in the control room, at the EAB, and at the outer boundary of the LPZ for a design basis control rod drop accident using alternate source terms. This calculation may also be used to support a license amendment request (LAR) associated with the GNF2 fuel transition.
 
==SUMMARY==
OF RESULTS/CONCLUSIONS The offsite and onsite (control room) doses are substantially below the regulatory limits as given in Regulatory Guide 1.183, 10 CFR 50.67, and 10 CFR 50, Appendix A, General Design Criterion 19.
LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY:
This calculation determines the radiological dose consequences resulting from a control rod drop accident using GNF2 fuel.* After NRC approval, this calculation will become the design basis for the CRDA.
IMPACT ON OUTPUT DOCUMENTS Revision 1 ofthis calculation will be the basis for revising the UFSAR.
 
TITLEJ
 
==SUBJECT:==
Control Rod Drop Accident Radfologlcal Analysls with Revised Source Tenns Page Iv DOCUMENT INDEX ci z
z a
Document NumberfTitle                        Revision, Edition, Date
                                                                                                --=
::J c..
                                                                                                      'S a.
0 1    UFSAR Section 15.4.9 D      D 2    NUREG-800, Standard Review Plan (SRP)        DRAFT Rev. 3 - April 1996 15.4.9, "Spectrum of Rod Drop Accidents D D CBWRl*.
3    10 CFR 100                                  67 FR 67101, Nov. 4, 2002 D D 4    10CFR50.67                                  64 FR 72001, Dec. 23, 1999 D            0 5    Regulatory    Guide 1.183, *Alternate July2000 Radiological Source Terms for Evaluating 0            0 Design Basis Accidents at Nuclear Power Plants" 6    10CFR50                                      64 FR 72002, Dec. 23, 1999 0      0 7    10 CFR 50, Appendix A, General Design Criterion 19 D            D 8    TIO 14844, *calculation of Distance Factors  March 23, 1962 for Power and Test Reactor Sites*, J. J.
D D Nunno, etal 9    CEI Calculation 3.2.15.2, "Control Rod Drop  Revision 0, 312012001 Accldenr.
0      0 10    CEI Calculation 3.3, "Control Rod Drop        Revision o, 11126/91 Accident Without MSIV Closure*.
D D 11    DES/98-0845, *perry Control Room            December 2, 1998 Atmosoheric Dlsoersion Factors" 0            D 12    Federal Guidance Report 11, Limiting        Second Printing 1989 Values of Radionuclide Intake and Air 0
Concentration and Dose Conversion Factors for Inhalation. Submersion. and lnaestion 13    Federal Guidance Report 12, External          1993 Exoosure to Radlonuc!ldes in Air. Water.
D      181 D And Soll 14    CEI Calculation CL-M26-01, *M26, Volume      Rev. 2                                            0 Calculations" NED0-31400A, *safety Evaluation for          October 1992 15 Eliminating the Boiling Water Reactor Main 0      181 0 Steam lsolatlon Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor".
16    Regulatory Gulde 1.78, "Assumptions for      June 1974                                        0 Evaluatina the Habitabllitv of a Nudear
 
TITLE/
 
==SUBJECT:==
Control Rod Drop Accident Radfologlcal Analysis with Revised Source Tenns  Pagev Power Plant Control Room during a Postulated Hazardous Chemical Release*.
17    Technical Specification 3.7.3, (CRER)        Amendment No. 161 System      Control  Room    Emergency                                        D    181 D Recirculation 18    Regulatory Guide 1.52, "Design, Testing,    Revision 2, March 1978 and Maintenance Criteria for Post Accident                                      181  0  0 Engineered-Safety-Feature      Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants".
19    K G. Murphy and K M. Campa, "Nuclear          1Jlh AEC Air Cleaning Conference Power Plant Control Room Ventilation 181  0  0 System Design for Meeting General Criteria 19". 1Jlh AEC Air Cleaning Conference 20    NUREG/CR-6604,          "RADTRAD:          A April 1998 Simplified    Model for
* RADlonuclide                                          181  0  0 Transport And Removal And Dose Estimation" 21    Regulatory Guide 1.77, *Assumptions Used      May 1974 for Evaluating a Control Rod Ejection                                            181  0  0 Accident for Pressurized Water Reactors*.
22    Deleted 0    0  0 23    PSAT 04202U.03, "Dose Calculation Data Rev.a Base for Application of the Revised Source D    181 0 Term to the Penv Nuclear Planr.
24    Deleted D    0  0 25    NEOE-31152P General Electric            Fuel  Rev.6 Bundle Destnns.
I 0    181 0 26    PNPP Technical Specifications                Appendix "A" to License No. NPF-58 0    181 D 27    Regulatory Gulde 1.25, *Assumptions Used March 23, 1972 for Evaluating the Potential Radiological                                        181  0 0 Consequences of a Fuel Handling Accident In the Fuel Handling and Storage Facility for Ballina and Pressurized Water Reactors*.
28    *Memo from Jeff Balcken (GE) to Emin          March 14, 1996 Ortalan (PNPP), "Fission Product 181  0  0 Inventories for Perry High Energy Cycles",
JBB96008.
29    GESTAR-11, NEDE-24011-P-A-11-US 0    181 0 30    PNPP Calculation 3.2.14.15, "Fuel Rev.O Handling Accident Using Alternate Source D    181 0 Terms*.
31    Dl-240, "Fuel Handling Accident Input Rev. O, 11/8/01                            0    181 0 Assumotions: Fuels lnout".
 
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==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysfs with Revised Source Tenns Pagevl
* 32    912-0610, "Control Room HVAC and Revision FF D      ~* D Emeraencv Recirculation".
33  NEDC-33270P        "PNP    GNF2      Fuel Revision4 Transition"                                                                      D      ~  D 34  GEH-KL1WX23P-017, Letter from E.G. May 7, 2012 Thacker II to E.S. Tomlinson, "PNP GNF2 D      ~  0 Fuel Transition: F0802 Source Term Output Files"
 
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==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysls with Revised Source Tenns 1.0 PURPOSE This calculation assesses the offsite and control room radiological impacts associated with a design basis control rod drop accident (CRDA) with the Regulatory Guide 1.183 [DIN # 5) source term assumptions and GNF2 fuel source term.                                                      *
 
==2.0 BACKGROUND==
 
The previous PERRY Control Rod Drop Accident (CRDA) radiological analysis, reported in Calculation 3.2.15.2 [DIN# 9) and Calculation 3.3 [DIN# 10), apply the source term assumptions in Appendix A to Standard Review Plan (SRP) Section 15.4.9 [DIN # 2). This guidance suggests a scenario in which fuel failures release source terms into the reactor coolant from which a fraction reach the condenser, which consequently leaks to the environment.
This calculation assumes the identical release scenario; however, the source term release fractions, chemical species distribution, and removal mechanisms are based on those reported in the NRC guidance in Regulatory Guide 1.183 [DIN # 5). The current PNPP FSAR also evaluates an event sequence in which offsite power is not lost (Scenario 2 in FSAR Section 15.4.9.5.2). This scenario is not specifically required by SRP 15.4.9 which requires a LOOP (FSAR Scenario 1). The offsite and onslte doses for this scenario are significantly lower than the Scenario 1 doses because of: 1) complete retention of halogens in the offgas system, and 2) significant holdup time for noble gases (2.47 day decay time for Kr and 54.2 day decay time for Xe). The offsite and onslte doses resulting from FSAR Scenario 2 are evaluated in . Due to the non-limiting nature of Scenario 2 and the lack of any regulatory requirement to evaluate this scenario, the FSAR could be revised to*remove Scenario 2 and only discuss the new limiting design basis calculation. The following discussions and calculations, with the exception of Attachment 1, are based on the limiting Scenario 1 evaluation.
ConseNatisms In this calculation include the following.
: 1. All impacted fuel is assumed to have been operating at high peaking factors and maximum exposures although: (i) the current PNPP core designs do not put peaking fuel bundles in face
* adjacent locations; and (ii) the core operating limits on power density would prohibit high-exposure bundles from being at high peaking factors .
  . 2. Fuel damage assumptions bound those calculated by the fuel vendor.
 
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==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with ReVlsed Source Tenns 3.0 METHODOLOGY 3.1 Regulatory Guide 1.183 Compliance 3..1.1Section4.1 of RG 1.183 This calculation applies an offsite dose calculation methodology consistent with that described in Section 4.1 of RG 1.183.
: 1. This calculation generates the resulting doses in terms of TEDE. Although the initial core source terms were generated with ORIGEN2 and consider the impact of daughter products during operation, this calculation does not include daughter production during the release period. Section 3.2.5 describes the source term assumptions in more detail.
: 2. This calculation applies the Inhalation dose conversion factors from Federal Guidance Report 11 [DIN# 12) consistent with the RADTRAD code as noted in NUREG/CR-6604
[DIN # 20) Table 1.4.3.3-2.
: 3. This calculation applies the recommended offsite breathing rates. Breathing rates of 3.5E-4 m3/s and 1.BE-4 m3/s are applied for the first eight hours and the following 16 hours, respectively.
: 4. This calculation applies the EDE dose conversion factors from Federal Guidance Report 12[DIN#13].
: 5. This calculation considers the impact of a 2-hour sliding EAB window; however, since this release begins immediately and decreases exponentially as the condenser activity decays and is depleted by *leakage, the first two hours would be the worst case for the EAB dose and no sliding window calculations are performed.
: 6. This calculation assesses the LPZ dose for compliance with the acceptance criteria.
: 7. The dispersion factors for this calculation do not consider depletion of the effluent plume for ground (or any other) deposition.
 
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==SUBJECT:==
Control Rod Drop Accldent*Radlologlcal Analysls with Revised Source Tenns 3.1.2 Section 4.2 of RG 1.183 This calculation applies a control room dose calculation methodology consistent with that described in Section 4.2 of RG 1.183.
: 1. This calculation considers all potential radiation sources to the control room operators.
Intake of the radiation plume from the control room ventilation intake is the only credible mode of contamination of the control room atmosphere. Other potential sources of control room dose are infiltration from the Turbine Building atmosphere and radiation shine.
Although the Turbine Building is located adjacent to the control room, infiltration from this area is neglected since holdup and dilution in the Turbine Building is neglected in the analysis. In addition, if dilution and holdup were considered, the source terms would be considerably diluted by mixing in the large PNPP Turbine Building and, considering the lack of any pressure differential between these areas, leakage into the Control Building through doors or sealed penetrations Is expected to be less than that predicted by the assumed immediate dispersion to the control room intake.
Radiation shine from the plume or external areas is insignificant due to the significant thickness of concrete between the control room and the plume or the condenser.
: 2. The control room dose calculation applies the same source term, transport, and release assumptions as the offsite dose calculation.
: 3. This calculation credits only manual actuation of the Control Room Emergency Recirculation System. No reliance is placed on radiation monitors for this isolation.
: 4. No credit is taken for personal protective equipment or prophylactic drugs.
: 5. This calculation applies the recommended occupancy factor and breathing rate.
Occupancy factors of 1.0 for the first 24 hours, 0.6 from 24 to 96 hours, and 0.4 from 96 to 720 hours are used In the control room evaluation. A breathing rate of 3.5E-4 rf'IJ/s is applied in the control room evaluation.
: 6. The control room calculation applies the same dose conversion factors as the offsite dose calculation. The recommended semi-infinite to finite cloud dose conversion factor is also applied to the control room EDE calculation within the RADTRAD code.
3.1.3 Section 4.4 of RG 1.183 This calculation applies the acceptance criteria in Table 6 ofRG 1.183 [DIN# 5] and10CFR50.67 [DIN#
4] for the offsite and control room doses respectively. This calculation also applies the 24-hour release duration from Table 6 of RG 1.183.
3.1.4 Section 5.3 of RG 1.183 The atmospheric dispersion factors (x/O values) for the LPZ and EAB are obtained from CEI Calculation 3.2.6.4 [DIN# 23). The Control Room x/Q values are from the Perry DES/98-0845 [DIN# 22). The x/Q values, based on a ground level release, are given below:
 
Pa119 4 TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenns Table 3-2 1/Q Values (sec/m3 )
Location Time Interval                    EAB                LPZ          CONTROL ROOM Oto 2 hrs                      4.3E-4              4.SE-5          3.5E-4 2 to 8 hrs                                        4.SE-5            3.5E-4 8 to 24 hrs                                        3.3E-5          2.1E-4 24 to96 hrs                                        1.4E-5          1.1E-4 96 to 720 hrs                                      4.1E-6          5.75E-5
: 3. 1.5 Appendix C of RG 1. 183 This calculation applies the acceptable assumptions described in Appendix C of RG 1.183.
: 1. The number of fuel rods damaged during the accident is based on a NRC-approved fuel vendor methodology and is described in detail in Section 3.2.3.
: 2. This calculation applies the gap fractions in Section 3 of RG 1.183 (seeTable 3-3 below) and assumes the release of 50% of the iodine and 100% of the noble gases for fuel postulated to reach melt conditions. Since the fuel gap can also contain the alkali metals per RG 1.183 Table 1, this calculation will apply a gap fraction of 12% per RG 1.183 Table
: 3. Since Appendix C of RG 1.183 does not address the melt release .fraction for alkali metals for a CRDA, this calculation will assume 25% of the alkali metals are released from the melted fuel consistent with RG 1.183 Table 1. Although RG 1.183 Table 1 reports that small fractions of other nuclide groups are also released from the melted. fuel, these source terms are neglected in this calculation due to (i) the small amount of fuel exposed to melt conditions (<1 %), (ii) the small in-vessel release fractions for these nuclide groups, and (Iii) the low volatility of these particulates from both the reactor and condenser.
Table 3-3 Gap Fractions Jor Hi1gh Burnup FueI Group            Gap Release          Melt Release Fracuon              Fraction Noble Gases            10%                  100%
Halooens              10%                  50%
Alkali metals          12%                  25%
: 3. Since fuel damage is postulated for this event, the impact of coolant source terms is neglected.
: 4. The activity released from the fuel from the gap and fuel pellets is assumed to be instantaneously mixed in the reactor coolant within the pressure vessel.
: 5. No credit is taken for partitioning in the pressure vessel or for removal by the steam separators.
: 6. Of the activity released from the reactor coolant within the pressure vessel, 100% of the noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to reach the turbine and condensers. The MSIVs are not assumed to close to maximize the radiation that reaches the condenser. Per Section 6.3.2.1 of NED0-31400A [DIN # 24],
iodine carryovers of only 2% would be expected, indicating that the 10% partition assumption is applicable even without the MSIV isolation on high steamline radiation. The NRC, in the SER to NED0-31400A, has approved this conclusion.
 
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==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenns
: 7. Of the activity that reaches the turbine and condenser, 100% of the noble gases, 10% of the iodine, and 1% of the particulate radionuclides are available for release to the environment. The turbine and condensers leak to the atmosphere as a ground-level release at a rate of 1% per day for a period of 24 hours, at which time the leakage is assumed to terminate. This leakage rate is applicable to PNPP since the mechanical vacuum pumps would be automatically tripped by the high steamline radiation signal associated with the postulated fuel damage as well as by any assumed loss of offsite power. AHhough the mechanical vacuum pump trip and isolation on high steamline radiation is not fully safety-related, the NRC has approved credit for this trip for the CRDA.
No credit is taken for dilution or holdup within the Turbine Building. Radioactive decay during holdup in the turbine and condenser is credited.
: 8. The iodine release from the turbine and condenser is assumed to be 97% elemental and 3% organic.
3.2 SRP 15.4.9 Compliance
: 3. 2. 1 Loss of Offsite Power Appendix A of Standard Review Plan 15.4.9 [DIN # 2) reports that a loss of offsite power (LOOP) should be assumed at the time of the accident. A LOOP would cause the turbine stop and control valves to close, scram the reactor, and trip the condenser offgas system or mechanical vacuum pumps. Non-LOOP scenarios that credit the operation of the offgas system are not limiting due to the significant holdup of both halogens and nobles in the low-temperature PNPP offgas system. This calculation makes the assumption of a coincident LOOP.
3.2.2 Turbine and Condenser Integrity Appendix A of Standard Review Plan 15.4.9 reports that the Integrity of the turbine and condenser is unaffected by this accident. As described later, this calculation makes this assumption and credits the source term holdup in the condenser and turbines. Since, at low reactor power levels, steam may be directed to the condenser via the turbine bypass system, the integrity of the bypass piping is also assumed to be unaffected by this event.
3.2.3 Largest Source Term Appendix A of Standard Review Plan 15.4.9 reports that the combination of operating mode, rod positions, burnup, etc. should be that which resuHs in the largest source term. This calculation applies a fuel damage estimate that is developed based on the worst-case rod worth possible under the Banked Position Withdraw! Sequence (BPWS) system, which is required by Technical Specification 3.1.6 [DIN # 26). In addition, this calculation considered the maximum allowable inoperable rods permitted under Technical Specification 3.1.3 [DIN # 26).
The PNPP BPWS controls rod patterns to minimize the rod worth of any control rod. In Section S.2.2.3.1.4 of GESTAR-11(DIN#29], General Electric has determined that the worst-case CRDA for a 7x7 bundle would result in no more than 850 rods (including a 10% allowance for uncertainties in the calculation) would reach a fuel enthalpy of 170 cal/gm. As discussed in SRP 15.4.9, fuel rods that exceed a deposited fuel enthalpy of 170 cal/gm are assumed to experience cladding failure. NED0-31400A [DIN # 24) applied this value with a bounding rod power level of0.12 MW for a total power level of 102 MW (i.e., 850*0.12) for the failed rods.
 
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==SUBJECT:==
Control Rod Drop Accident Radiological Analysls with Revised Source Terms As reported in Section 3.7 of NEDE-31152P [DIN# 25], 850 fuel rods were calculated based on an 8x8 fuel design and a value of 1000 rods is applicable to the GE11 and GE13 fuel design. For GE12 fuel, 1Ox1 O design, approximately 1200 fuel rods would reach a fuel enthalpy of 170 cal/gm. Since the GE 8x8 design contains 62 full-length rods1, the 850 failed rods represents 13.7 bundles. With an effective 87 full-length fuel rods in the GE14 bundles2 , 13.7 bundles would be equivalent to 1191 rods validating the GE value of 1200 rods. As with other 10x10 designs, when the bounding fuel enthalpy analysis is applied to GNF2 fuel, approximately 1,200 fuel rods are calculated to reach a fuel enthalpy of 170 cal/gm [DIN # 33).
The source terms applied in* this analysis will be conservatively based on the failure of 16 fuel bundles (-
1376 fuel rods) representing the four-bundle cell associated with the dropped control blade and one additional row as illustrated below. This conservatively bounds the 1,200 fuel rods given in *NEDC-33270P
[DIN #33] for GNF2 fuel 3. GNF2 is very similar to the GE14 design, with the same number of full-length fuel rods. Some of the partial-length rods in GNF2 fuel are shorter than that of GE14. The number of effective full-length fuel rods for GNF2 fuel is 85.6, which is less than the number of effective full length rods for GE14.
Dropped DDDG![DDD                    Control D 1111111                    Blade D**                  D D 11111
* IllJ:Y              Failed Bundles DBllllMfD DDDDDD Figure 3-1 Failed Bundles in a CRDA As discussed in Section 6.2.1 of NED0-31400A (DIN# 24), the maximum mass fraction in the damaged fuel that reaches temperatures In excess of the melting point is 0.0077.
3.2.4 Source Term Decay Appendix A of Standard Review Plan 15.4.9 reports that no decay should be credited* prior to accident initiation. This calculation makes this assumption.
3.2.5 Rod Power Level Appendix A of SRP 15.4.9 reports that the rods that are calculated to fail are assumed to have operated at a core power level 1.5 times that of the average power level of the core. Although the CRDA results in significant fuel failure only at very low core powers where the reactivity worth of a control blade is maximized, it is possible, In the event of a quick startup after a scram, that the source term Inventories in the fuel rods could be comparable to the full-power activities. This SRP 15.4.9 power assumption is Although some GE 8x8 designs include 63 fuel rods, this calculation will conservatively assume that the reported 850 failed fuel rods is based on an 8x8 design with 62 fuel rods to maximize the number of failed bundles.
2    GE14 bundles conlain 92 fuel rods, fourteen of which are part-length rods (Reference DCC-002 to Calculation 3.2.15.2).
Assuming that the part length rods are approximately 60% of the length of the full-length rods, this gives an effective 87 full-length rods per bundle.
3    The GNF2 bundles also contain 92 fuel rods, fourteen of which are part-length rods. However, the effective length of the GNF2 part-length rods is less than that of GE14 part-length rods [DIN #33).
 
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==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenns consistent with the 1.5 minimum radial peaking factor applied in the fuel handling accident as required by Regulatory Guide 1.25 [DIN # 27] with a flat local peaking profile. Considering the localized nature of this event, bundle-average source terms are expected to be sufficient for this accident; however, as applied in the fuel handling accident [DIN # 30], this calculation will apply a radial peaking factor of 2.0. The GE ORIGEN2 analysis of fission product inventories was based on a power level of 3758 MWt. The core inventory, expressed in terms of Ci/MWt, is independent of power level and may be used for different assumed power levels. This inventory is adjusted for a 2% measurement uncertainty, and the number of bundles (748) In the core for GNF2 fuel in the following table.
Table 3-4 Bundle Source Term (no decay)
                .. *- -**-*--* --------------* -- .**--* --- -.. ----*** -- r-*-*- ---- -----*---.. ---*---- *r------*--. ------ ------ - ---* ... *-*--- .----- *----
                !----- .~o~~-r_!evel =---->------~8~!*_._                              ------***_L*---- "--------_l Core                  Bundle                                                    Core                  Bundle Activity              Activity                                              Activity                  Activity Isotope        (Ci/MWt)                    (Cl)                    Isotope                (Cl/MWt)                          (Cl)
BR-82            3.534E+02              1.811E+03                    RB-86                    1.376E+02                7.053E+02 BR-83            6.456E+03              3.308E+04                    RB-88                    3.672E+04                1.882E+05 BR-84            1.109E+04              5.683E+04                    RB-89                    4.696E+04                2.406E+05 1-126            6.156E+02              4.160E+03                    KR-83M                    6.472E+03                3.317E+04 1-130            2.056E+03              1.055E+04                    KR-85                    7.578E+02                3.863E+03 1-131            5.428E+04              2.782E+05                    KR-85M                    1.347E+04                6.905E+04 1-132            7.628E+04              4.011E+05                    KR-87                    2.566E+04                1.315E+05 1-133            1.099E+05              5.632E+05                    KR-88                    3.608E+04                1.849E+05 1-134            1.205E+05            6.175E+05                    >CE-129M                  4.034E-01                2.067E+OO 1-135            1.030E+05              5.278E+05                    >CE-131M                  6.082E+02                3.117E+03 CS-132            1.512E+01            7.748E+01                    >CE-133                  1.060E+05                5.434E+05 CS-134            1.385E+04              7.098E+04                    >CE-133M                  3.450E+03                1.768E+04 CS-134M            3.230E+03              1.655E+04                    >CE-135                  3.868E+04                1.982E+05 CS-136            4.324E+03              2.216E+04                    >CE-135M                  2.182E+04                1.118E+05 CS-137            8.380E+03              4.294E+04                    >CE-136                  8.974E+04                4.599E+05 CS-138            9.970E+04              5.109E+05
                                                                                              *--~----*
L.~. *--                              -- ~*-------*--                                -
3.3 Turbine and Condenser Volumes It is recognized that this calculation is not sensitive to the volume of the condenser or turbine when the leak rate is expressed in terms of a percent per day as recommended in SRP 15.4.9. A larger volume would result in a larger leak rate (in cfm) of a more diluted source term, while a smaller volume would result in a smaller leak* rate of a more concentrated source term. However, the condenser volume Is needed if a specific flow from the condenser (in cfm) is specified. The condenser airborne volume is taken from Calculation 3.2.15.2 [DIN # 9), which reports the free volume of the condenser and turbine as 1.26E+05 ft3.
In the event steam Is being directed to the condenser via the bypass system when the CRDA occurs, the volume of the turbine would still be available for source term migration and leakage since the turbine sits directly above the condenser.
 
PaaeB TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenns 4.0 ASSUMPTIONS 4.1 Bromine Modeling As halogens, the bromine isotopes are modeled identical to the iodine isotopes.
4.2 Control Room Isolation and lnleakage Control Room Volume and Unfiltered lnleakaqe Case 1:
This case assumes that the normal HVAC system is operating with a flow of 6,000 cfm +/- 10% [DIN# 32]
and is not isolated for the duration of the accident. For this case, the Control Room Volume is 371, 760 ft3
[DIN # 14], the maximum inflow to the control room of 6,600 cfm is assumed and the emergency recirculation mode is not initiated.
Case 1a:
When the system operates in the emergency recirculation mode, the outside makeup air is isolated.and the control room envelope is not pressurized relative to adjacent areas. For this case, the Control Room Volume is 390,020 ft3 (DIN # 14]. The Control Room Emergency Recirculation System (CRERS) recirculation flow is assumed to begin operation at 30 minutes post-accident. An unfiltered inleakage to the control room of 1,375 cfm during the emergency recirculation mode, beginning at time zero, is assumed for the remaining duration of the accident. This assumption agrees with the design basis LOCA assumption (Reference PSAT 04202U.03 [DIN # 23]). In addition to this unfiltered inleakage, an unidentified unfiltered inleakage of 10 cfm for ingress and egress, in accordance with Regulatory .Guide 1.78 [DIN # 16], is assumed. The total unfiltered in leakage into the control room for this case is therefore 1385 cfm.
Recirculation Flow Rate:
The CRERS is a redundant system and each subsystem has a design flow capacity of 30,000 cfm :I: 10%.
A conservative recirculation flow rate of 27 .000 cfm has been used in previous analyses (Reference PSAT 04202U.03 [DIN# 23]). The Control Room Emergency Recirculation (CRER) subsystem provides, Inter alla, recirculation of the control room air through the CRER filter units.
Filter Efficiencies:
Per Technical Specification 3. 7 .3, (CRER) Control Room Emergency Recirculation System [DIN# 17], the CRER filters are tested in accordance with the Ventilation Filter Testing Program (VFTP) which complies with Reg. Guide 1.52 [DIN# 18]. The CRER HEPAs are credited with removing 99% of the particulate iodine activity while the charcoal adsorber removes 95% of the elemental and organic iodine activity. The filter efficiencies used in this calculation are conservatively assumed to be 80% for all iodine species.
Control Room Parameters Reference
 
Paae9 TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysls with Revised Source Tenns
* Intake and Recirculation Charcoal Adsorber Efficiencies(%)
elemental iodine                          95            RG 1.52 11 organic iodine                            95 particulate iodine                        99            11 Note: This calculation conservatively assumes an HEPA/charcoal adsorber efficiency of 80% for iodines.
* Unfiltered lnleakage (cfm)
Case 1:
No unfiltered inleakage assumed Normal HVAC flow of 6,600 cfm assumed, no control room isolation.
Case  1a:
(1) Unfiltered inleakage (cfm)1375              Assumed (2) In leakage due to door openings      1O cfm      RG 1.78, Position C.1 O Total unfiltered inleakage  = 1385 cfm beginning at time zero.
Emergency Recirculation initiated at 30 minutes.
* Filtered Recirculation Rate (cfm)
During emergency mode                    27,000      CEI Cale. 3.2.6.5, Rev. O Emergency Recirculation is assumed to occur at 30 minutes by Operator action.
(Case 1a)
* Breathing Rates (m3/sec) 1 0- 720 hrs                        3.5-04
* Occupancy Factors2 O - 8 hrs                                  1.00 8-24 hrs                                  1.00 24-96 hrs                                  6.00-01 96- 720 hrs                              . 4.00-01 1
Regulatory Guide 1.183 [DIN # 5) 2K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criteria 19", 131h AEC ftJr Cleaning Conference. [DIN # 19)
 
Paoe 10 TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radiological Analysis with ReVlsed Source Tenna 5.0 CALCULATIONS 5.1 Release Fractions As discussed in Section 3.1.5, various percentages of the noble gases, halogens, and alkali metals are assumed to be released as a result of the gap failure. As discussed in Section 3.2.3, 0.77% of the fuel mass in the failed bundles reach melt conditions. The total release fractions from the 16 affected bundles are calculated below.
Halogens:                Release Fraction= (1-0.0077)*10% + 0,0077*50% = 10.308%
Nobles:                  Release Fraction= (1-0.0077)*10% + 0.0077*100% = 10.693%
Alkali Metals:          Release Fraction= (1-0.0077)*12% + 0.0077*25% = 12.1%
Based on the release fractions calculated above, the bundle activity from Table 3-4, and the release fraction assumptions reported in Section 3.1.5, the activity available for release from the condenser is tabulated below.
 
Paaa 11 TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radiological Analysls with Revised Source Tenns Table 5-1 Activity Available for Release Activity Available Activity in                                      Fraction for Release 16            Release Release                    Available from Bundle          Bundles      FracUon Fraction to for                          Condenser Isotope    Activity (Cl)  (Cl)          from Fuel Condenser Release (Cl)
BR-82      1.81E+03      2.90E+04      10.308%              10%              10%      2.9868E+01 BR-83      3.31E+04      5.29E+05      10.308%              10%              10%      5.4564E+02 BR-84      5.68E+04      9.09E+05      10.308%              10%              10%      9.3n9E+02 1-128      4.18E+03      6.69E+04      10.308%              10%              10%      6.B932E+01 1-130      1.05E+04      1.69E+05      10.308%              10%              10%      1.7394E+02 1-131        2.78E+05      4.45E+06      10.308%              10%              10%      4.5876E+03 1-132        4.01E+05      6.42E+06      10.308%              10%              10%
* 6.6160E+03 1-133      5.63E+05      9.01E+06      10.308%              10%              10%      9.2884E+03 1-134        6.17E+05      9.88E+06      10.308%              10%              10%      1.0184E+04 1-135        5.28E+05      8.45E+06      10.308%              10%              10%      8.7052E+03 CS-132      7.75E+01      1.24E+03      12.10%              1%              1%        1.5001E-02 CS-134      7.10E+04      1.14E+06      12.10%              1%              1%        1.3743E+01 CS-134M      1.66E+04      2.65E+05      12.10%              1%              1%        3.2045E+OO CS-136      2.22E+04      3.55E+05      12.10%              1%              1%        4.289BE+OO CS-137      4.29E+04      6.87E+05      12.10%              1%              1%        8.3138E+OO CS-138      5.11E+05      8.17E+06      12.10%              1%              1%        9.8912E+01 RB-86      7.05E+02      1.13E+04      12.10%              1%              1%        1.3655E-01 RB-88        1.88E+05      3.01E+06      12.10%              1%              1%        3.6430E+01 RB-89        2.41E+05      3.85E+06      12.10%              1%              1%        4.6589E+01 KR-83M      3.32E+04      5.31E+05      10.69%            100%              100%      5.6742E+04 KR-85        3.88E+03      6.21E+04      10.69%            100%              100%      6.6439E+03 KR-85M      6.90E+04      1.10E+06      10.69%            100%              100%      1.1813E+05 KR-87        1.31E+05      2.10E+06      10.69%            100&deg;k            100%      2.2497E+05 KR-88        1.85E+05      2.96E+06      10.69%            100%              100%      3.1633E+05 XE-129M      2.07E+OO      3.31E+01      10.69%            100%              100%      3.5388E+OO
              >CE-131M    3.12E+03      4.99E+04      10.69%            100%              100%      5.3323E+03
              >CE-133      5.43E+05      8.69E+06      10.69%            100%              100%      9.2969E+05 XE-133M      1.nE+04        2.83E+05      10.69%            100%              100%      3.0247E+04 XE-135      1.98E+05      3.17E+06      10.69%            100%              100%      3.3912E+05 XE-135M      1.12E+05      1.79E+06      10.69%            100%              100%      1.9130E+05 XE-138      4.60E+05      7.36E+06      10.69%            100%              100%      7.8678E+05
            ..... --~-----  **---**-*--*.1..--..... _J __ ...- . . . .L....- - - * * ---**---*-- *----**
                                                                                                          --- _ _ _ _ _ .. l
 
Page 12 TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radlologlcal Analysis with Revised Source Tenns 5.2 Model The inlet isolation valve for the mechanical vacuum pumps closes simultaneously with a pump motor trip upon a main steam line high radiation signal (see USAR Sections 7.3.1.1.2 and 10.4.2.6). Additionally, the inlet valves fail closed and the pump motors trip on the loss of offsite power.
The analytical model consists of two volumes, the condenser, from which the release occurs, and the control room. Control room isolation is not credited (Case 1) or is assumed to occur at time zero (Case 1a). An in leakage rate of 1385 cfrn is assumed in the isolated configuration. For the unisolated case (i.e.,
Case 1), the 6,600 cfm outside air intake continues throughout the event. Although the release is terminated at 24 hours, the analytical model is conservatively evaluated for 30 days to consider the radiological impact of any residual source terms in the control room at the end of 24 hours. The release pathway is illustrated in Figure 5-1 *while the associated analytical model is Illustrated in Figure 5-2.
MSIVs (Remain Open)
Vessel Turbine Leakage to Environment
                                                        ...-~-*---        Condenser Figure 5-1 Control Rod Drop Accident Release Pathway 6600 cfm nonnal (Case 1)              Turbine or 1385 cfm unfiltered              Bldg Vent inleakage (Case 1a)                      ...        CondenserfTurblne
                                              +                      1% per day Volume = 1.26E+05 ft3 Control Room                              Activities from Table 5-1 Volume= 3.7176E+05 ffl 27,000cfm (30 minutes, Case 1a)            6600 cfm or 1385 cfm
          *Applies to all iodine species Figure 5-2 Control Rod Drop Accident Model
 
Paae 13 TITLE I
 
==SUBJECT:==
Control Rod Drop Accident Radiological Analysis with Revised Source Tenns 5.3 Calculational Methodology The above data and model will be used in the RADTRAD code [DIN # 20) to determine the resulting onsite and offsite doses. The calculations used a modified RADTRAD nuclide inventory and release fraction files to include only the source terms available for release as reported in Table 5-1. The isotopes In Table 5-1 that are not part of the RADTRAD default files (e.g., Bromines, Kr-83m, etc.) were added to the RADTRAD nuclide inventory and release fraction files. The RADTRAD input and output files used for this Scenario are given below:
Description                            Case 1, No CR isolation          Case 1a, CR isolation Plant scenario file                            Perry CRDA 1.psf                  Perry CRDA 1a.psf Auxiliary RADTRAD Input Flies Nuclide Inventory File            crda.nif                            crda.nif Release Fraction and Timing File          PerryCRDA.RFT                      PerryCRDA.RFT Dose Conversion Factors                crda.inp                          crda.inp Output Fiie                                    Perry CRDA 1.out                Perry CRDA 1a.out The output files for the above cases are given in Attachment 2 and 3.
6.0 RESULTS The results of the CRDA analysis are reported below.
Table 6-1 CRDA Dose Results Dose (Rem TEDEI Location              Case 1, No            Case 1a, Control          TEDE Acceptance Control Room            Room Isolation              Criteria (Rem)
Isolation EAB                    1.61E-01                  1.61E-01                      6.3 LPZ                    1.62E-01                  1.62E-01                      6.3 Control Room*          2.63E-01                  2.22E-01                        5
                  *control Room results are conseNatively rounded up to bound any changes in calculated Control Room volume.
The offsite results are well below the offslte acceptance criteria of 6.3 rem TEDE. In addition, the control room dose is well within the 5 rem limit. As such, it is concluded that a control rod drop accident, when modeled with the alternate source terms, satisfies the NRC's acceptance criteria.
 
ATIACHMENT 1 Page 1 ATTACHMENT 1 UFSAR SCENARIO 2 UFSAR Scenario 2 considers an event sequence In which offsite power is not lost (see Scenario 2 in FSAR Section 15.4.9.5.2). This scenario is not specifically required by SRP 15.4.9 - which requires a LOOP (FSAR Scenario 1). The offsite and onsite doses for this scenario are expected to be significantly lower than the Scenario 1 doses because of: 1) complete retention of halogens in the offgas system, and
: 2) significant holdup time for noble gases (2.47 day decay time for Kr and 54.2 day decay time for Xe).
Due to the non-limiting nature of Scenario 2 and the lack of any regulatory requirement to evaluate this scenario, the FSAR could be revised to remove Scenario 2 and only discuss the new limiting design basis calculation. The offsite and onsite doses resulting from FSAR Scenario 2 are evaluated below.
Release from Offgas System:
From CEI Calculation 3.3 [DIN# 10], the charcoal delay holdup times are: Krypton= 2.47 days, Xenon=
54.2 days. Iodine retention Is 100 %. The activity released from the offgas system is given below:
Charcoal Delay Holdup times:                  Krypton=          2.47 days Xenon=            54.2 days l\WVllJ Activity                Available for Activity in  Release                    Fraction      Available for    Decay    Release from Bundle    l&Bundles Fraction from Release Fraction Available for Release from    ConSlant  Offgas System Isotope    Activity (Ci)      (Ci)      Fuel    to Condenser      Release      Condenser (Ci)    (1/hr)        (Ci)
BR-82        1.811E+03    2.898E-t04  10.308%        10%            10%          2.9868E+01    1.96E-02          0 BR-83        3.308E+04    5.293E+05    10.308%        10%            10%          5.4564E-t02  2.89E-01          0 BR-84      5.683E+04      9.093E+05    10.308%        10%            10%          9.3729E+02    1.31E+OO          0 KR-83M      3.317E+04      5.306E+05    10.693%        100%          100%          5.6742E+04    3.73E-01    1.450E-05 KR-85      3.883E+03      6.213E+04    10.693%        100%          100%          6.6439E+03    7.38E-06    6.641E-t03 KR-85M      6.905E+04      1.105E+06    10.693%        100%          100%          1.1813E+05    1.55E-01    1.228E-t01 KR-87        1.315E+05    2.104E+06    10.693%        100%          100%          2.2497E+05    5.45E-01    2.086E-09 KR-88        1.849E+05    2.958E+06    10.693%        100%          100%          3.1633E+05    2.44E-01    1.647E-01 1-128      4.180E+03      6.687E-t04  10.308%        10%            10%          6.8932E+01    1.66E+OO          0 1-130        1.055E+04    1.687E+05    10.308%        10%            10%          1.7394E+02    5.61E-02          0 1-131      2.782E+05      4.451E+06    10.308%        10%            10%          4.5876E+03    3.59E-03          0 1*132      4.011E+05      6.418E+06    10.308%        10%            10%          6.6160EtG3    3.01E-01          0 1-133      5.632E+05      9.011E+06    10.308%        10%            10%          9.2884EtG3    3.33E-02          0 1-134      6.175E+05      9.880E+06    10.308%        10%            10%          1.0184E-t04  7.91E-01          0 1-135      5.278E+05      8.445E+06    10.308%        10%            10%          8.7052EtG3    1.05E-01          0 XE-129M    2.067E+OO      3.30BE+01    10.693%        100%          100%          3.5368EtGO    3.25E-03    5.143E-02 XE-131M    3.117E+03      4.987E+04    10.693%        100%          100%          5.3323E+03    2.40E-03    2.335E+02 XE-133      5.434E+05      8.694E+06    10.693%        100%          100%          9.2969E+05    5.51E-03    7.204E+02 ltf-133M    1.788E+04    2.829E+05    10.893%        100%          100%          3.0247E+04    1.26E-02    2.415E-03 XE-135      1.982E+05    3.171E-t06  10.693%        100%          100%          3.3912E+05    7.63E-02    2.833E-38
~-135M      1.118E+05    1.789E-t06  10.693%        100%          100%          1.9130E+05    2.67E+OO      O.OOE+OO XE-138      4.599E+05      7.358E+06    10.693%        100%          100%          7.8678E+05    2.95E+OO      O.OOE+OO To determine the offsite and onsite dose consequences of the above release, the two cases analyzed in the main body of this calculation are reevaluated here by changing only the source (release) term. The RADTRAD input and output files used for this Scenario are given below:
 
ATTACHMENT 1 Page2 Description                            Case 2, No CR isolation          Case 2a, CR isolation Plant scenario file                          Perry CRDA 2.psf                    Perry CRDA 2a.psf Auxiliary RADTRAD Input Flies Nuclide Inventory File            crda2.nif                          crda2.nif Release Fraction and Timing File          PerryCRDA.RFT                      PerryCRDA.RFT Dose Conversion Factors              crda.inp                            crda.inp Output Fiie                                    Perry CRDA 2.out                Perry CRDA 2a.out The results are given below:
Scenario 2 CRDA Dose Results Dose IRem TEDE)
Location              Case 2, No            Case 2a, Control            TEDE Acceptance Control Room            Room Isolation              Criteria (Rem)
Isolation EAB                    2.BOE-06                  2.BOE-06                      6.3 LPZ                    2.96E-06                  2.96E-06                      6.3 Control Room            1.31E-06                1.24E-06                        5 Note that the control room doses are nearly identical for case 2 and 2a since the only difference is the use of the control room recirculation filter which is effective only for iodine removal. Since the iodines are retained in the offgas system the only removal is from decay.
 
ADDENDUM6 Summary of Main Steam Line Break Outside Containment (MSLBOC) Dose Calculation 28 pages follow This Addendum is considered a summary of the associated calculation - only selected pages of the approved document are provided.
 
*I TITLE/
 
==SUBJECT:==
Main Steam Une Break Ualng A1temat8 Soun:e Tanna                                                                          Pagei                            I Main Steam Une Break Using Alternate Source Terms TABLE OF CONTENTS SUBJECT Table of Contents OBJECTIVE OR PURPOSE SCOPE OF CALCULATION
 
==SUMMARY==
OF RESULTS/CONCLUSIONS LIMITATIONS OR RESTRICTION ON CALCULATION APPLICABILITY IMPACT ON OUTPUT DOCUMENTS DOCUMENT INDEX CALCULATION COMPUTATION (BODY OF CALCULATION):
: 1. PURMSE ............................................................................................................................................ l
: 2. BACKGROUND.................................................................................................................................. I
: 3. ACCEPI'ANCE CRITERIA*~*.......................................................................................................... 2 3.1    OFFSrm Dose~ ....................................................................................................................... 2 3.2    CoN'IRoL RooM Doss LIMn- .**..*.**..**......*.**...*..**..*..***..**.*.*.***.......**.***.**.*..**.*.**..**.*.**......*.....***.*.** 2
: 4. REGULATORY COMl'LIANCE ..................................................................................................... 2-
: 5. ASSUMP'l10NS ....................................." ........................................................................................... 5
: 6. DATA....................................................................................~.............................................................. 7 6.1 SCBN'ARJOS .ANALVZBD *********** *********** ************* ************************************ ********************************************** 7 6.2 A'IMOSPllBR.Ic DISPBR.SION FAcroRS ********************************-*************************************************************** 7 6.3 DoSB CONVBR.SION FAC'I'ORS *************************************** ~ ******************************************************************* 8 6.4 CON'IROL ROOM VBN'l'llA110N PARAME'l'BRS ****************.*** ~ ****************.******************************************** 8 6.S *IODIN'B SPBCIBS DIS'J'RJBU110N ************************** *********************** ** ********************************************* ******* I 0 6.6 SOURCE 'l'BRMS ************************************.**********************************.***.*******************************.******.************ 11 6.7 ACflVITY Rm..BAsBD *** ** *************** ** *** **. ******* *** ** *********** ***** ** ******* **** *** ****** ************* *** ****** ** *** **** *** ** ***
* 13 6.1.1 Halogen Activity Released ...................................................................................................... 13 6.1.2 Noble Gas Activity Released ......*.........*.................................................................................. 14
: 7.  *l\t:EimODOLOGY............................................................................................................................ 16
* 8. RELEASE PATBWAY MODEL .................................................................................................... 16
 
TIRE/
 
==SUBJECT:==
Main Steam Line Break Using Allemate Source Tenna                                                                  Paee ii                  I
: 9. RA.D'l'RAD MODEL .....................................................................................N***************..*******.......... 17
: 10. CALCULATION ***********************************H***..***************************************************************************...-..... 19 10.1 SCENARIO I - MAxlMuM EQUILIBRRJM IODINE CONCENTRATION ................................................ 19 10.2 SCENARIO 2 - PRE-ACCIDENT IODINE SPIKE ................................................................................. 19
: 11. RESUL'J:'S.......................................................................................~................................................... 20 11.1 SCENARIO 1, MAxlMUM EQUILIBRIUM IODINE REACTOR COOLANT CONCENTRATION .................. 20 11.2 SCBN'ARIO 2, PRB-BXIS1'IN'G IODIN'B SPIKB ********.*********************************************************************.******* 21 ATTACHMENTS: : MSLB 02a Output                                                                                                      20Pages : MSLB 02b Output                                                                                                      22 Pages : MSLB 40a Output                                                                                                      20Pages : MSLB 40b Output                                                                                                      22Pages : TIO dose evaluation                                                                                                  83Pages : Increased Control Room volume evaluation                                                                              88Pages TOTAL NUMBER OF PAGES IN CALCULATION (COVERSHEETS +BODY+ ATTACHMENTS)                                                              280Pages SUPPORTING DOCUMENTS (For Records Copy Only)
DESIGN VERIFICATION RECORD                                                                                                          1 Page CALCULATION REVIEW CHECKLIST                                                                                                s      ~Plilges j/IJ 10CFR50.59 DOCUMENTATION                                                                                                          ~          Pages DESIGN INTERFACE
 
==SUMMARY==
 
DESIGN INTERFACE EVALUATIONS
: 2.  *.t'Page 6 Pages
                                                                                                                                                    ;J&l OTHER                                                  DE.Sl~tJ llJf'uf' 'R*~                                                      'I'      Pages EXTERNAL MEDIA? (MICROFICHE, ETC.) (IF YES, PROVIDE UST IN *eoov OF CALCULATION)
                                                                                                                                  .D      YES 181    NO
 
I TITLE/
 
==SUBJECT:==
Main Steam Line Break Using Alternate Soun:e Terms                    Page iii          I OBJECTIVE OR PURPOSE:
The purpose of this calculation is to conservatively determine the radiological dose consequences resulting from the reactor coolant release that accompanies a Main Steam Line Break Outside Containment (MSLBOC) accident. The total effective dose equivalent (TEDE) at the control room, the exclusion area boundary (EAB ), and the outer boundary of the low population zone (LPZ) are calculated using NRC-approved methods. This calculation incorporates the source terms derived from Regulatory Guide 1.183. [DIN# 5] This calculation also provides thyroid and whole body doses based on TIO 14844 [DIN # 10] for comparison with current design basis results and a limited sensitivity study of control room volume.
This calculation will replace the current PNPP MSLBOC dose analysis given in CEI calculation 3.2.15.4
[DIN# 12) which was performed using Regulatory Guide 1.5 and TID-14844 methodologies. [DIN# 9 and 10]
SCOPE OF CALCULATION/REVISION This calculation determines the dose consequences at the control room, the BAB, and the outer boundary of the LPZ for a design basis main steam line break using alternate source terms. This analysis is one of those required for full implementation of the Alternate Source Tenns in accordance with Regulatory Guide 1.183.
 
==SUMMARY==
OF REmJLTS/CONCLUSIONS The offsite and onsite (control room) doses are substantially below the regulatory limits as given in Regulatory Guide 1.183 [DIN# 5], 10CFRS0.67 [DIN# 4], and 10 CFR 50, Appendix A, General Design Criterion 19 [DIN # 7].
LIMITATIONS OR RESTRICTIONS ON CALCULATION APPLICABILITY:
This calculation determines the radiological dose consequences resulting from the reactor coolant. release that accompanies a Main Steam Line Break Outside Containment (MSLBOC) accident After NRC approval, this calculation will become the design basis for the MSLBOC accident.
IMPACT ON OUTPUT DOCUMENTS This calculation will replace the current PNPP MSLBOC dose analysis given in CBI calculation 3.2.15.4
[DIN # 12] and will be the basis for revising the UFSAR.
 
I TITLE/
 
==SUBJECT:==
Main Steam Un~ Break Using Alternate Soun:e Tanna                            Page iv        I DOCUMENT INDEX 0
z
    ~      Document Numberlfitle                                      Revision, Edition, Date 1    UFSAR Section 15.6.4                                Revision J2, January,2003        D      D 181 2    NUREG-800, Standard Review Plan (SRP) lS.6.4, ''Radiological Consequences of Main DRAFr Rev. 3 -April 1996                D D Steam Line Failure Outside Containment lBWR)".
3    10CFR 100                                          67 PR 67101, Nov. 4, 2002              D 0 4    10 CFR. 50.67                                      64 FR 72001, Dec. 23, 1999      0      181 D 5    Regulatory        Guide      1.183,      "Alternate July2000                        0      181 D Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants" 6    10CFRSO                                            64 PR 7'Jrol, Dec. 23, 1999            0 0 7    10 CPR SO, Appendix A, General Design                                                0      181 0 Criterion 19 8  Technical Specification 3.4.8, "RCS Specific Amendment No. 69                        D      181 0 Aclivitv" 9    Regulatory Guide l .S (Safety Guide 5), 311on1                                              0  0
          ..Assumptions Used For Evaluating the Potential Radiological Consequences of a Steam Line Bn:ak Accident for Boiling Water Reactors".
10  TID 14844, "Calculation of Distance Pacton March23, 1962                                    0  D for Power and Test Reactor Sites", J. J. Nunno, etal 11  UFSAR Section 1.3                                    Revision 12, January, 2003              0  0 12    CBI Calculation 3.2.15.4, "Main. Steam Line          Revision O, 9/15180.            0      0  181 Break".                                              and DCC-001 13    Technical Specification 3.6.1.3, SR 3.6.1.3.7        Amendment 115                    D      181 0 14    PNPP Calculation No. 3.2.15.0, "Design Basis        Revision 0                      0      181 0 Reactor Coolant and Steam Source Terms" 15    PSAT 04202U.03 (Dose Application Data Rev. 0                                          0      181 0 Base) 16    DBS/98-0845,          ''Perry    Conttol      Room December 30, 1998                  0      181 0 Atmospheric Disoersion Factors" 17    Federal Guidance Report 11, ldmitipg Values Second Printing 1989                            D  0 of Radionuclide In!Jlke Mel Air Concentration and Dose Conyeqion Factors for Inhalation.
          - -            J1nd*
18    Federal      Guidance      Report    12,  External  1993                            181    D 0
          -            tn D  **    "'i  in A~. Water. and
 
I TinE/
 
==SUBJECT:==
Main Staam Une BnNlk Ualng AH8mala Soun:e Tenna                              Pagev      I Soil 19  CEI Calculation CI;M26-01, "M26, Volume Rev. 1                                      D    181 D Calculations" 20  Calculation 3.2.lS.14, ''Fuel Handling Accident Rev.O                                D    181 D Usin2 Alternate Source Terms" 21  CBI Calculation No. 3.2.8.1, "Control Room Rev. 0, 12130/1998                        181  0  D Habitability Following a Fuel Handling Accident" 22    Not used                                                                            D    D D 23    Not Used                                                                            D    D D 24    Regulatory Guide 1.78, "Assumptions for          June 1974                          181  0 D Bvaluating the Habitability of a Nuclear Power Plant Control Room during a Postulated Hamrdous Chemical Release".
25  Technical Specification 3.7.3, (CRBR) System
* Amendment 125                      D    181 D Control Room Bmemency Recirculation 26  Regulatory Guide 1.52, "Design, Testing, and      Revision 2, March 1978            181  D  D Maintenance Criteria for Post Accident Engineered-Safety-Feature            Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants".
27  K. G. Murphy and K. M. Campe, "Nuclear            13111 ABC Air Cleaning Conference  181  D  D Power Plant Control Room Ventilation System Design for Meeting General Criteria 19", 13111 ABC Air Cleanina Conference 28  Technical Specification 3.7.S, Majp Conclenser    Amendment 125                      D    181 D Offaas 29  NURBG/CR-6604, ''RADTRAD: A Simplified April 1998                                    181 Model for RADionuclide Transport And D  D Removal And Dose Estimation" 30  PNPP Ventilation Filter Testins Propam Amendment 117                                  D    181 0 lVr.1r1.
31  P&ID 912-610                                      Rev. I                              D    181 D 32  Periodic Test lnsttuction Pn-GBN-POOl 1          Rev. cc                            D    181 D 33  Standard Review Plan 15.0.1, "Radiological Rev,0                                      D    181 D Consequence AnalySis Using Alternate Source Terms" 34  22A2703T, "Radiation Sources".                    Revision6                          D    181 D 35  GB-NB-A2200084-42-0l, "Perry Nuclear RevisionO                                        D    181 D Power Plant Asset Improvement Project", Task Gl-42.
36  Regulatory Guide 1.195, "Methods and May2003                                          D Assumptions for Evaluating Radiological 181 D Consequences of Design Basis Accidents at Uoht-Water Nuclear Power Reactors"
 
Page1 I TITLE I
 
==SUBJECT:==
Main Steam Une Break Using Altemata Source Tenne                                                I
: 1.      PURPOSE The Main Steam Line Break Outside Containment (MSLBOC) accident is classified as a limiting fault in UFSAR section 15.6.4 [DIN # 1). This event is not expected to occur during the life of the plant, however it has consequences that include the potential for release of radioactive material from the reactor vessel and subsequently to the environment. In order to evaluate the possible effects of this event on plant and public safety, a complete circumferential break is postulated to occur on one of the main steam lines immediately downstream of the outermost Main Steam Isolation Valve (MSIV) outside of the primary containment. The purpose of this calculation is to conservatively determine the radiological dose consequences resulting from the reactor coolant release that accompanies the break. The calculation uses NRC-approved methods. The following changes to the current plant design basis calculation are considered:
* Appllcatlon of the Altemate Source Term Methodology of RG 1.183.
* Increased control room lnleakage.
This calculation also provides thyroid and whole body doses based on TIO 14844 [DIN # 10] for comparison with current design basis results and a limited sensitivity study of control room volume.
: 2.        BACKGROUND Standard Review Plan (SRP) 15.6.4 [DIN# 2] presents the NRC criteria for evaluating the radiological consequences of main steam line failures outside containment for BWRs. The fundamental acceptance criterion is that the calculated whole-body and thyroid doses at the exclusion area boundary (EAB) (site boundary) and low population zone (LPZ) boundary do not exceed certain exposure guidelines. Two separate scenarios are to be evaluated: ( l) an assumed pre-accident coolant iodine concentration.
corresponding to the maximum "spiking" value permitted by the Technical Specifications. I; and (2) an assumed pre-accident reactor coolant iodine concentration corresponding to the maximum equilibrium value* permitted by the Technical Specifications. The exposure limits associated with each scenario are, respectively: (1) the exposure guidelines of 10 CFR 100; and (2) a "small fraction" of the exposure guidelines of 10 CFR 100 [DIN# 3]. A "small fraction" has been defined in SRP 15.6.4 to be 10% of 10 CFR 100 guidelines. Applying this to the 25 rem TEDB limit [DIN# A, 5] gives an acceptance criteria of 2.5 rem TEDB for Scenario 1 and 25 rem TEDE for Scenario 2. The exposure guidelines for control room personnel is based on General Design Criteria 19 (ODC 19) of Appendix A to 10 CFR Part 50
* which states that "exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TBDE) as defined in &sect; 50.2 for the duration of the accident". [DIN# 4, 5, 6, 7]
1SRP 15.6.4 actually    states that the iodine values should correspond to those allowed by the "NSSS vendor's standard technical specifications". It was written in this manner because the analyses were originally intended to confinn the adequacy of the proposed technical specification values for each plant. Once these values are in place, the analyses should be based on the existing specifications as is done*in this calculation.
 
Page2 I Tl1lE I
 
==SUBJECT:==
Main Steam Line Braak Ualng Altemale Soun:e Tanna                                        I This calculation examines two basic scenarios in order to determine the limiting scenario. Both of these scenarios involve the release of only the activity present in the reactor coolant at the time of accident.
This is appropriate for PNPP because no fuel failures are predicted for this accident [DIN# l]. Scenario 1 assumes that the maximum amount of iodine allowed by the current PNPP Technical Specifications is present in the coolant. This is the amount, given in Technical Specification (TS) 3.4.8 [DIN # 8). which is less than or equal to 0.2 &#xb5;Ci/gm dose equivalent 1-131. Scenario 2 assumes that the reactor coolant contains the maximum iodine concentration that is allowed during iodine spiking conditions (4.0 &#xb5;Ci/gm dose equivalent 1-131) [DIN# 8). Both scenarios assume that the design basis quantities of noble gases (Kr and Xe) are present
: 3.      ACCEPTANCE CRITERIA 3.1    Offs/te Dose Limits E'Aluilibrium Iodine lnveptoa The offsite dose limits for an equilibrium iodine inventory are "a small fraction" of 10 CPR 100 guidelines. This has been interpreted by the NRC [DIN # 2) to mean 10% of the 10 CPR 100 limits.
or 2.5 rem TEDE.
Maximum Iodine Inyentorv The offsite dose limits specified for the maximum technical specification iodine concentration are the mDB dose limits of 25 rem [DIN # 5).
The above limits agree with those given in Table 1 of Standard Review Plan 15.0.1 [DIN # 33].
3.2    Control Room Dose Limit In addition to the above limits, the doses to control room personnel may not exceed those specified in 10 CFR 50. General Design Criterion 19 (GDC 19) and 10 CFR 50.67. [DIN# 7, 4] *Applying this to the TEDB limits gives a control room limit of S rem 'IBDB.
: 4.        REGULATORY COMPLIANCE The basic regulatory guidance on implementing the Alternate Source Term (AST) methodology for the main steam line break (MSLB) is provided in appendix D of Regulatory Guide 1.183. Compliance with this guidance is detailed below:
 
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==SUBJECT:==
Main Steam Line Break Using Altemate Source Terms                                                                      I SOUBCETERM
: 1. Assumptions acceptable to the NRC staff regarding core inventory and the release of radionuclides from the fuel are provided in Regulatory Position 3 of this guide. The release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.
PNPP Compliance: See discussion of Regulatory Guide Position 3.1 and 3.2 below.
: 2. If no or minimaP fuel damage is postulated for the limiting event, the released activity should be the maximum coolant activity allowed by technical specification. The iodine concentration in the primary coolant is assumed to co"esporrd to the following two cases. in the nuclear steam supply system vendor's standard technical specifications.
2.1 The concentration that is the maximum value (typically 4.0 &#xb5;Ci/gm DE 1-131) permitted and co"esponds to the conditions ofan assumed pre-accident spike, and 2.2 The concentration that is the maximum equilibrium value (typically 0.2 &#xb5;Ci/gm DE 1-131) permitted/or continued full power operation.
PNPP Compliance: No fuel damage is postulated for this event. Therefore, this calculation evaluates the released activity associated with the maximum coolant activity allowed by the PNPP technical specifications (i.e., 0.2 &#xb5;Ci/gm DE 1-131 and 4.011Ci/gm DE 1-131).
: 3. The activity released from the fuel should be assumed to mix instantaneously and homogeneously in the reactor coolant. Noble gases should be assumed to enter the steam phase instantaneously.
PNPP Compliance: No fuel damage is postulated for this event. The maximum normal reactor coolant inventory and the pre-accident "spike" inventories are based on eqllillbrium activities with instantaneous and uniform mixing in the coolant. Noble gases are assumed to be iiistantaneously released to the steam phase.
TRANSPORT
: 4. Assumptions acceptable to the NRC staff related to the transport, reduction, and release of radioactive material to the environment are as follows.
2 The activity assumed in the analysis should be based on lhe activity associated with the projected fuel damage or the maximum technical speclficalion values, whichever maxlmi7.e8 the radiological consequences. In determining dose equivalent 1-131 (DB 1-131), only the radioiodine associated with nonnal operations or iodine spikes should be Included. Activity from projected fuel damage should not be Included.
 
Paga4 JTITLE/
 
==SUBJECT:==
Main Steam Line Braak Using Altemate Soun:e Terms .                                                                    I 4.1            The main steam line isolation valves (MSW) should be assumed to close in the maximum time allowed by technical specifications.
PNPP Compliance: The main steam lines are assumed to close in the maximum time allowed by the PNPP technical specifications (i.e., S 5.0 seconds, TS 3.6.1.3, SR 3.6.1.3.7 [DIN# 13] plus margin for actuation signal delay. The total time conservatively assumed in this analysis is 6 seconds.
4.2            The total mass of coolant released should be assumed to be that amount in the steam line and connecting lines at the time of the break plus the amount that passes through the valves prior to closure.
PNPP Compliance: The main steam line break, with minimum ECCS, has been evaluated and the blowdown results consider both steam and liquid flows. The data from this analysis is used to determine the total mass released prior to MSIV closure. The activity released to the environment is the sum of the activity in the steam and liquid release plus the activity generated during MSIV closure.
4.3            All of the radioactivity in the released coolant should be assumed to be released to the atmosphere instantaneously as a ground-level release.
* No credit should be assumed for plateout, holdup, or dilution within facility buildings.
PNPP Compliance: All released activity is assumed to be an instantaneous ground level release without any retention within the Auxiliary Building Steam Tunnel.
4.4            The iodine species released from the main steam line should be assumed to be 95% Cs/ as an aerosol, 4.85% elementa~ and 0.15% organic.
PNPP Compliance: This analysis uses this iodine species distribution.
Regulatory Guide 1.183, Regulatory Position 3.1, Fission Product Inventory The inventory offission 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 values for fuel enrichment, fuel bumup, and an assumed core power equal to the current licensed rated thermal power times the ECCS evaluation uncertainty.3 The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. 4 The core inventory should be determined using an 3 The uncertafnty factor used in determining the core inventory should be that value provided in Appendix K to 10 CPR Pait SO, typically t.02.
4 Note that for some radionuclides, such as Cs-137, equilibrium will not be reached prior to fuel offload. Thus, the maximum inventory at the end of life should be used
 
Paae5 I TlnE I
 
==SUBJECT:==
Main Steam Une Bruk Using Alternate Soun:e Tenna                                                                      I appropriate isotope generation and depletion computer code such as OR/GEN 2 (Ref. 17) or OR/GEN-ARP (Ref. 18). Core inventory factors (CVMWt) provided in TID/4844 and used in some analysis computer codes were derived for low bumup, low enrichment fuel and should not be used with higher bumup and higher enrichment fuels.
PNPP Compliance: The PNPP plant specific core inventory is not used in the MSLB evaluation which is based only on the maximum allowable reactor coolant inventory since fuel failure is not postulated [DIN # 14]. This equilibrium coolant inventory was developed based on infinite exposure time and no decay.
Regulatory Gulde 1.183, Regulatory Position 3.2, Release Fract1ons5 For non-WCA events, the fractions of the core inventory assumed to be in the gap for the various radionuclides are given in Table 3. The release fractions from Table 3 are used in conjunction with the fission product inventory calculated with the maximum core radial peaking factor.
PNPP Compliance: No fuel failure is postulated to occur as the result of a MSLB, consequently the release fractions of Table 3 are not applicable.
Regulatory Guide 1.195, Appendix D The above guidance agrees with the direction provided in Appendix D of Regulatory Guide 1.195 [DIN#
36] except with regard to the guidance for the iodine species distribution. Appendix D states that the assumed iodine species distribution is 5% particulate, 91% elemental, and 4% organic. Instead of this guidance, the distribution given in Regulatory Guide 1.183, above, will be used.
: 5.        ASSUMPTIONS The analysis is based on the regulatory guidance given in NRC Standard Review Plan (SRP) 15.6.4 [DIN
# 2) and Regulatory Guide (RO) 1.5 [DJN # 9] except where deviation is justified below. The assumptions used are:
: 1. The mass of reactor coolant assumed to be released to the environment is 14,311 lbs. of steam and 127,376 lbs. of liquid from the break. [DIN# 1] This exceeds the value (100,000 lbs.) assumed by the
* NRC in SRP 15.6.4 [DIN# 2] for 238" diameter vessel (PNPP vessel size given in DIN# 11, TABLE 1.3-1 and 22A2703T [DIN# 34]) plants such as PNPP. The assumed total release of 141,687 lbs.
agrees with Calculation 3.2.15.4. [DIN# 12] Note: Section 3.1.2.2 of GE-NE-A2200084-42-0l [DIN
      # 35] documents that there is no change to the release as a result of power uprate.
5 The release fracdons listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak bumup up to 62,000 MWDIM'IU. The data In this secdon may not be appllcable to cores containing mixed oxide (MOX) fuel.
 
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==SUBJECT:==
Main Steam Una Bl98k Using Altemale SoUn:e Tanna                                                  I Note: The mass release data given in the UFSAR and calculation 3.2.15.4 are based on a total isolation time of 6.05 seconds. Also, the total release for each time interval was calculated based on the release rate at the beginning of the time interval. This results in a conservatively high total release. If the average release rate were used for each time interval, the total release would be lower and would agree better with the data given in SRP 15.6.4. This is illustrated below:
I i                        '                                        Total Rlleaae Over Time
(
                                        '                        i                      Interval (Iba) ua1ng Anrage Steam        Liquid        Total                                  Allena and a Atleaaa      Atlaaae    Allena      Time Interval                    Riie**
Rate        Rate          Rate    uaedlnCalc. Rom Cale.      Duration of &.&
Time      (lbs/aec)    (lba/aac)  (lba/aec)      3.2.1&.4    3.2.18A            sec 0      9655          0          9665        0.05        482.75 0.05      11220          0          11220      0.21-0.05    1,795.20          521.88 0.21      8072          0          8072    0.995-.021    8,338.52          1,543.38 0.995      7616          0          7ff16    1.01-0.995    116.14          6,181.09 1.01      988        26860        27648      2.05-1.01    28,753.92          264.93 2.05      1065        28100        27165      3.05-2.05    27,165.00        28,602.76 3.05      1148        25630        2ff178      4.05-3.05    28,778.00        28,970.60 4.05      1171        23960        25131      6.05-4.05    50,262.00        25,953.60 5.5                            21690.615                                    33,945.88 e;os      1096        19290        20088 TGrAL                    _,1~1~~
: 2. The main steam release is assumed to be directly to the atmosphere in accordance with SRP 15.6.4
[DIN # 2). The release is assumed to be instantaneous.
: 3. No fuel failures are assumed. This is consistent with the GB transient analysis for this event. [DIN
      #I] SRP 15.6.4 requires that any fuel failures be accounted for in the dose analysis. Thus, should the transient analysis change such that fuel failures are predicted this calculation must be revised to reflect the additional activity in the release.
: 4. No additional leakage paths for reactor coolant are assumed. Any small amounts of leakage past the MSIVs are not considered to be significant since the source term for this event is limited to routine coolant activity.
S. No credit for plateout of iodine is assumed.
: 6. No credit is t.aken for filtration, holdup, dilution, or decay of iodine once it is released from the break.
No decay is assumed for noble gas isotopes.
 
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==SUBJECT:==
Main Steam Une Break Using Alternate Source Tanna
* I
: 7. MSW closure time is 6.0 seconds. This is based on TS 3.6.1.3 [DIN# 13], which requires the MSNs to close within 5 seconds and an assumed allowance of 1.0 seconds for the actuation signal.
  &.      DATA 6.1    Scenarios Analyzed
  >  Scenario 1 uses the full power equilibrium iodine concentration in accordance with the PNPP Technical Specifications. The isotopic concentrations are those which produce 0.2 &#xb5;Ci/gm dose equivalent 1-131. The isotopic concentrations based on this limit are given in DIN 14.
>  Scenario 2 uses the maximum iodine concentration per technical specifications corresponding to short term increased iodine concentrations. Technical Specification 3.4.8 [DIN # 8] gives the maximum iodine concentration of 4.0 &#xb5;Ci/gm dose equivalent 1-131. This TS allows a higher coolant iodine level for up to 4 hours to account for the short duration concentration "spikes" which can accompany power level changes. A reactor coolant specific activity greater than 4.0 &#xb5;Ci/gm requires isolation of all main steam lines within 12 hours. Isotopic concentrations based on this limit are given in DIN 14.
6.2    Atmospheric Dispersion Factors The atmospheric dispersion factors ('X/Q values) for the LPZ and BAB are obtained PSAT 042020.03 (Dose Application Data Base) [DIN #15]. The Control Room 'l)Q values are from DES/98-845 [DIN#
16]. The 'X/Q values, based on a ground level release, are given below:
Table6-1
                                                  't/O(sedm~
Locadon Time Interval                BAB              LPZ Oto2hrs                      4.3E-4          4.8E-5 2to8hrs                                      4.8E-5 8 to24hrs                                    3.3E-5 24 to96hrs                                    1.4E-5 96 to720brs                                  4.lE-6
 
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==SUBJECT:==
Main Steam Line Break Using Alternate Soun:e Tenna                                      I Table6-2 Control Room' r/O (sec/m3>
Location Time Interval                    CONTROL ROOM 0 to 8 hrs                              3.5E-4 8 to 24hrs                              2.lE-4 24 to96 hrs                            l.1E4 96to720brs                            5.7SE-5 6.3    Dose Conversion Factors The effective dose conversion factors for the TBDE and thyroid calculations are based on FGR 11 [DIN #
17] and 12 [DIN# 18].
6.4    Control Room Ventilation Parameters Control Room Volume 367,070 ft3    CBI Calculation CL-M26-0I, Rev. 1[DIN#19]
Note: Calculation M26-l is cmrently being revised with a preliminary control room volume of 420,494 ft3. The effect of this difference in control room volume is evaluated in Attachment 6. This evaluation demonstrated that a larger Control Room volume does not result in a significant difference in doses to the Control Room-operators.
Control Room Unfiltered* lnleakage For each of the design basis scenarios identified previously, two cases will be executed to quantify the control room doses due* to different control room HVAC operating modes
:case t:
. ;This case assumes that the normal HVAC system is operating with a flow of 6,000 cfm ::t: 10% and is not
  -isolated for the duration of the accident. This flow is based on the design value shown in P&ID 912-610
[I>IN # 31) with an allowable operating tolerance as specified in PrI-GEN-POOll [Din# 32). For this case the maximum inflow to the control room of 6.600 cfm is assumed and the emergency recirculation mode is not initiated.
Case2:
During an emergency, when the system operates in the emergency recirculation mode, the outside makeup air is isolated at Time = 0 and the control room envelope is not pressurized relative to adjacent areas. An
 
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==SUBJECT:==
Main Steam Line Break Using Alternate Soun:e Terms unftltered inleakage to the control room of 1,375 cfm is assumed to begin at time zero and is assumed for the remaining duration of the accident. Emergency recirculation is assumed to begin at 30 minutes.
These assumptions agree with the design basis LOCA assumptions (Reference PSAT 04202U.03 [DIN #
15)). In addition to this unfiltered inleakage, an unidentified unfiltered inleakage of 10 cfm for ingress and egress, in accordance with Regulatory Guide 1.78 [DIN # 24). is assumed. The total unfiltered inleakage into the control room for this case is therefore 1385 cfm.
Recirculation Flow Rate:
The CRERS is a redundant system and each subsystem has a design flow capacity of 30,000 cfm :I: 10%
[DIN # 30]. A conservative recirculation flow rate of 27.000 cfm has been used in previous CBI analyses
[DIN # 15]. The Control Room Emergency Recirculation System (CRER) subsystem recirculates control room air through the CRER filter units.
Filter Efficiencies:
Per Technical Specification 3.7.3, "(CRER) Control Room Emergency Recirculation System" [Din# 25],
the CRER filters are tested in accordance with the Ventilation Filter Testing Program (VFfP) [DIN # 30]
which complies with Reg. Guide 1.52 [DJN # 26]. The CRER HEPA& are credited with removing 99% of the particulate iodine activity while the charcoal adsorber removes 95% of the elemental and organic iodine activity. The filter efficiencies used in this calculation are conservatively assumed to be 50% for all iodine species.
Control Room Parameters Refe!ence
* Intake and Recirculation Charcoal Adsorber Efficiencies (%)
elemental iodine                        95              RG 1.52 [50% assumed for conservatism]
organic iodine                          95              "
particulate iodine                      99              "
Note: This calculation conservatively assumes an adsorber efficiency of 50% for iodines.
* Unfiltered Jnleakage (cfm)
Case 1:
Normal HVAC flow of 6,600 cfm assumed with no control room isolation.
Case2:
(1) Unfiltered inleakage (cfm)          1375          Assumed (2) Inleakage due to door openings 10                  RG 1.78, Position C.10
                                      =
Total unfiltered inleakage 1385 cfm beginning at time zero
 
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==SUBJECT:==
Main Steam Une Break Using Alternate Soun:e Terms                                      I Emergency recirculation initiated at 30 minutes
* Filtered Recirculation Rate (cfm)
During emergency mode          27 ,000    Ventilation Filter Testing Program (VFfP) [DIN# 30)
* Breathing Rates (m3/sec) 1 0-720hrs                                3.SE-04
* Occupancy Factors 1 0-8hrs                                  1.00 8-24brs                                1.00 24-96hrs                                6.00-01 96-720hrs                              4.00-01 1
Reference 26, K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System
* Design for Meeting General Criteria 19", 13lh AEC Air Cleaning Conference.            .
6.5    Iodine Species Distribution The guidance in Regulatory Guide 1.183 [5], suggests the following chemical forms for the released iodine:
Table6-3S        e.s Distribution Form                Fraction %
Csl as aerosol            95 Elemental                4.85 Or anic                  0.15
 
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==SUBJECT:==
Main Steam Line Break Using Alternate Source Terms                                  I 6.6    Source Terms PNPP Calculation 3.2.15.0, Revision 0 [DIN# 14) determined the iodine and noble gas generation rates and the concentrations in the reactor coolant and steam. The calculated values are reproduced below:
Table 6-4 Release Rates at~ Cu.Ci/s)
Release Rate tu.Cl/a) 0.2 J1CUg                    4.0 &#xb5;Cl/g Isotope      1*131 Dose Equivalent          1*131 Dose Equivalent 1-131              2.61E+03                      5.22E+04 1-132              3.61E+04                      7.21E+o5 1-133              1.78E+04                      3.56E+o5 1-134
* 1.02E+05                      2.03E+06 1..135              2.85E+04                      5.69E+05 Br-83                3.88E+03                      7.75E+04 Br-84                1.89E+04                      3.78E+05 Br-85                8.59E+04                      1.72E+06 Kr-83m                1.12E+04                      1.12E+04 Kr-85m                2.47E+04                      2.47E+04 Kr-85                9.22E+01                      9.22E+01 Kr-87                7.36E+04                      7.36E+04 Kr-88                7.31E+04                    7.31E+04 Kr-89                4.68E+05                      4.68E+05 Xe-131m                9.37E+01                      9.37E+01 Xe-133m              .9.83E+02                    9.83E+02 Xe-133              2.82E+04                    2.82E+04 Xe-135m                1.09E+05                      1.09E+05 Xe-135              7.69E+04                    7.69E+04 Xe-137              5.32E+05                      5.32E+05 Xe-138              3.10E+05                    3.10E+05
 
Page 12
*...I 11_n_e_1_su_BJ_ec_r_:Ma_1n_Steam                                          _ _ _ _ _ _ _ _ _ _ _ _ _ _I
* _ _L_1ne_a_rea_ku_a_1n..,a_A1t_*_ma_1e_1ou_rca_Tenna Table 6-5 Reactor Steam Concentrations (u.CU2)
Reactor Steam Concentrations Reactor Steam                    0.2&#xb5;CUg                          4.0pCUg Isotope              Concentrations            1-131 Dose Equivalent            1*131 Dose Eaulvalent 1-131                                                1.27E*03                        2.54E*02 1-132                                                1.76E-02                        3.51E-01 1-133                                                8.66E-03                        1.73E-01 1-134                                                4.95E-02                        9.90E-01 1-135                                                1.39E-02                        2.77E-01 Br-83                                                1.89E-03                        3.78E-02 Br-84                                                9.20E-03                        1.84E-01 Br-85                                                4.18E-02                        8.36E-01 Kr-83m                  5.48E-03 Kr-85m                  1.20E*02 Kr-85                4.49E-05 Kr-87                3.58E-02 Kr-88                3.56E-02 Kr-89                2.28E-01 Xe-131m                  4.56E*05 Xe-133m                  4.79E-04 Xe-133                1.37E-02 Xe-13Sm                  5.29E*02 Xe-135                3.74E-02 Xe-137                2.59E-01 Xe-138                1.51E*01 Table H Reactor Coolant Concentrations Reactor Coolant Concentrations 0.2p,CUg                          4.0pCUg Isotope      1-131 Dose Eauivalent                1-131 Dose Eaulvalent 1*131              4.51E-02                            9.02E-01 1*132              4.51E-01                            9.02E+OO 1-133              3.12E-01                            6.24E+OO 1-134              9.77E-01                            1.95E+01 1*135              4.51E-01                            9.02E+OO Br-83                6.26E-02                            1.0SE+OO Br-84                1.13E-01                            2.25E+OO Br-85                7.14E-02                            1.43E+OO
 
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==SUBJECT:==
Main Steam Una Break Using Alternate Source Terms                                            1 6.7  Activity Released The mass of reactor coolant assumed to be released to the environment is 14,311 lbs of steam and 127,376 lbs of liquid from the break. The total activity released to the environment is the sum of the activity in the steam and liquid release plus the activity generated during MSIV closure.
6.7.1 Halogen ActMty Released The amount of halogens released from the break (Ci) is determined based on the coolant halogen concentrations, &#xb5;Ci/gm and the halogen generation rate, &#xb5;.Ci/sec. PNPP Calculation No. 3.2.15.0 [DIN #
14] gives the predicted maximum equilibrium and spike RCS release rates and concentrations based on TS allowable levels. The concentrations are converted to total activity released based on the amount and form of the coolant escaping prior to MSIV closure. These values are used to calculate the total Ci released during the 6.0 second main steam line isolation.
Given the concentration of each Halogen isotope, the total activity released may be calculated as follows.
Rell  =(c.1, 1*1q *Ml.tq +c.1,stm *Mstm )*4.54B-04 lbm    Ci gm +R1 *6.0sec* ~i
                                                                          &#xb5;Ci              10 &#xb5;.Ci Where:          Rel1  = Total Iodine activity released for isotope i (Ci) qJiq = Reactor coolant halogen concentration (JI.Ci/gm) for each isotope Ct.scm = Steam halogen concentration (&#xb5;Ci/gm) for each isotope Ri = Halogen generation rate (JI.Ci/sec) 4.54E-04 = Conversion from &#xb5;Ci/gm to Ci/lbm Muq = Total Reactor Coolant liquid released, 127,376 lbm Mstm = Total steam release, 14,311 lbm
 
Page 14 TITLE I
 
==SUBJECT:==
Main Steam Une Break Ualng Alternate Source Tenne Table 6-7
                                              ~talHal 0      oaen  RIe ease Total Release (Cl)
Isotope        0.211Cl/g      4.0 JLCl/g 1-131 Dose 1-131 Dose Eaulvalent Eaulvalent 1-131            2.630E+o0 5.259E+01 1-132            2.639E+01 5.277E+02 1*133            1.819E+o1 3.637E+02 1-134            5.739E+01 1.148E+03 1*135            2.632E+01 5.263E+02 Br-83            3.075E+OO 6.151E+01 Br-84            6.687E+OO 1.337E+02 1
Br-85            4.913E+OO 9.826E+01 1
                    . Br-85 is not included in the dose calculation since no dose conversion factors for this isotope is listed in FGR 11 or FGR12.
6.7.2 Noble Gas Activity Released The amount of xenon and krypton activity released (Ci) is based on the release rates (&#xb5;Ci/sec.) and concentrations (&#xb5;Ci/gm.) provided in Reference 14. Given the mass release rates, the amount of activity released prior to MSIV closure can be detennined.
The noble gas reactor coolant activity is based on the maximum allowable offgas activity release reported in Technical Specifications. PNPP Technical Specification 3.7.5 [DIN# 28] requires that the noble gas offgas activity be no more than 358 mCi/s after 30 minutes of decay. This ensures that 10 CFR 100 limits are met should this effluent be discharged directly into the environment (such as from a malfunction of the offgas system). The 30 minutes of decay is intended to account for the delay prior to reaching the offgas system~
* PNPP Calculation No. 3.2.15.0 [DIN# 14] provides the reactor coolant noble release rates for no decay time and for a 30-minute decay prior to release. This analysis will conservatively take no credit for decay of noble gases to better reflect the instantaneous release from the steam line. The release rates are shown in Table 6-8.
 
Page 15 I TITLE
 
==SUBJECT:==
Main Sleam.Une Break Uelng I                                    Alternate Source Tenna                                I Given the concentration of each Noble Gas isotope, the total activity released may be calculated as follows.
Rel 1
                            =(c. stm *Mstm )*4.54E-04 Ihm t,
Ci gm +R *6.0sec* ~i .
                                                                  &#xb5;Ci    1
* 10 &#xb5;Ci Where:          Rel1  = Total Noble Gas activity released for isotope i (Ci)
Ct;stm = Steam Noble Gas concentration (&#xb5;.Ci/gm) for each isotope Ri = Noble Gas generation. rate (&#xb5;Ci/sec) 4.54E-04 = Conversion from &#xb5;Ci/gm to Ci/Ihm
                        =
Mstm Total steam release, 14,311 Ihm Table6-8 Total Noble Gas Release Total Release (Cl}
Isotope Kr-83m          1.028E-01 Kr-85m          2.261E-01 Kr-85          8.447E-04 Kr-87          6.740E-01 Kr-88          6.697E-01 1
Kr-89          4.288E+OO Xe-131m          8.582E-04 Xe-133m          9.007E-03 Xe-133          2.581E-01 Xe-135m          9.974E-01 Xe-135        . 7.042E-01 Xe-1371          4.873E+OO Xe-138          2.840E+OO 1
Kr-89 and Xe-137 are not included in the dose calculation since no dose conversion factors for these isotopes are listed in FGR 11 or FGR12.
 
Page 16 I Tine I
 
==SUBJECT:==
Main Steam Une Break Using Alternate Source Terms                                            I
: 7.      METHODOLOGY The calculation of offsite and onsite doses is performed using the RADTRAD computer code [DIN # 29).
Proper operation of this code was verified by execution of the benchmark BWR cases included with the RADTRAD documentation [DIN # 29).
: 8.      RELEASE PATHWAY MODEL The release pathway for this event is from the failed steam line directly to the atmosphere. The fission product release is modeled as a "puff' release as illustrated in Figure 1. This applies to both Scenarios 1 and 2. No credit is taken for filtration *by the Annulus Exhaust Gas Treabnent System; mixing, holdup, or dilution in the sec0ndary containment; or, decay prior to release. The release is assumed to terminate upon closure of the MSIVs at 6.0 seconds into the event A single failure of one MSIV to close is implicitly assumed, however this does not impact the results due to the redundancy of the MSIVs. No additional leakage other than that directly from the break is assumed for this event.
Figure 1 Release Flow path PRIMARY CClllTAlllllNT
                -----1J AUllLIARY
* AUll.UIRY IUILDMCI                                                BUILDING OolllOI lldgllYAO T
t jc:on:L      3                                                  ,, .......
Ln)
MIOARD  ITl!All RPV
                                                                                ,_  Lllll 8UPPA188IOI POOL
 
Page 17 I TITLE I
 
==SUBJECT:==
Main Steam Una Break Using Alternate Source Terms                                        I
: 9.      RADTRAD MODEL The mode] developed for the analysis is illustrated in Figure 2. It consists of two nodes, the steamJine(s) and the control room. No mixing, dilution, plateout, or decay is assumed for the fission products released.
The control room is modeled with the inleakage equal to the outleakage at 1385 cfm while the recirculation flow rate of the standby fresh air supply system is 27 ,000 cfm. The control room volume was modeled at 3.67E5 cu. ft. and the charcoal filters were modeled to remove iodine at a 50% efficiency.
The '1.}Q values are reported in Table 6-1 and 6-2 for the BAB and Control Room, respectively. All thyroid doses were calculated using a breathing rate of 3.SE-4 m3/s and TEDE Dose Conversion Factors (DCF).
The fission products are released over the first time step to simulate an essentially instantaneous release.
This is accomplished in the model by creating a volume of l cubic foot and establishing a release flow rate (lElO %/day) sufficient to ensure that all of the activity exits this volume during the first time step.
By the nature of the model, this causes the entire drise at the Site Boundary and LPZ to be accumulated in the first time step. This is conservative because longer release times would result in lower doses due to decay during release. The short release time allows essentially zero decay during release for the isotopes of interest. Since this release occurs instantly, the first two hours would be the worst case for the BAB dose and no sliding window BAB dose calculations are performed.
 
Page 18 I TITLE /
 
==SUBJECT:==
Main Steam Una Break Using Alternate Source Tenna                                                                  I Figure2 RADTRAD Model Equivalent d 6.0 sec.
Main Steam Unea flow fran break release cl No fuel fallUl9                                                    .over flrat time Interval.          XfQ, OCF, breathing Design basis noble gas release rate
                                                                          ~~>i--------,~                      rates= Dose at EAB
                                                                                                              &LPZ Scenario 1: o.2E-06atgm OE ladlne-131 (TS E'quilbrlum)                                      ENVIRONMENT
                                                                              -~
Scenario 2: 4.<E.()6 Q'gm DE lodne-131 (TS 8'llkB)                  \
11'11eaka9e flew 1385dm XIO. DCF, breathing rates. filbtn., occ. factois
                                                                                                            =DcselnCR OONIROLAOOM
                                                                      =
vol&.me 3.87E5 a.I.ft.
reclrcflow "l:fZT,fm dm OLdlealcage flaN 1385cfrn
 
Paae 19 I TinE I
 
==SUBJECT:==
Main Steam Una Break Using Altamata Source Tanna                                      I
: 10. CALCULATION 10.1 Scenario 1
* Maximum Equilibrium Iodine Concentration Offsite and onsite doses for Scenario 1 were calculated using the RADTRAD model described above.
The RADTRAD output files are given in Attachments 1 and 2. The RADTRAD input and output files used for this Scenario are identified below:
Description                              Case l, No CR isolation Case 2, CR isolation Plant scenario file                          Perry MSLB 02a.psf          Perry MSLB 02b.psf Auxiliary RADTRAD Input Files Nuclide Inventory File          perry02.nif                pmy02.nif Release Fraction and Timing File          perrymslb.rft              perrymslb.rft Dose Conversion Factors            perry.inp                  perry.inp Output File                                  Perry MSLB 02a.out          Perry MSLB 02b.out 10.2 Scenario 2 - Pre-accident Iodine Spike Offsite and onsite doses for Scenario 2 were calculated using the RADTRAD model described above.
The RADTRAD output files are given in Attachments 3 and 4.
Description                              Case 1, No CR isolation Case 2, CR isolation Plant scenario me                            Perry MSLB 40a.psf          Perry MSLB 40b.psf Awdllary RADTRAD Input Files Nuclide Inventory File          perry40.nif                perry40.nif Release Fraction and Timing File          perrymslb.rft                perrymslb.rft Dose Conversion Factors              perry.inp                  pelT)'.inp Output File                                  Perry MSLB 40a.out        Perry MSLB 40b.out
 
Page20 I TITLE I
 
==SUBJECT:==
Main Steam Una Break Ualng AHemate Source Tanna I
: 11.      RESULTS Since the releases for this accident were modeled as a "puff' release, the two hour sliding dose, as required by Regulatory Guide 1.183, is not app1icab1e. The worst two-hour dose is the initial 2-hour dose for a ''puff' release. The results given in Attachments 1 through 4 are summarized below:
11.1 Scenario 1, Maximum Equilibrium Iodine Reactor Coolant Concentration Main Steam Line Break Scenario 1 Maximum Equilibrium Iodine Reactor Coolant Concentration Case 1 Normal BVAC now of 600 ctm assumed with no control room Isolation)
TEDE Dose              TEDE Dose em            Limit* em ite Boundary (EAB)                                                          S.90E-02                    2.5 863 m, 0-2 hrs ]
w Population 7Atne (LPZ)                                                  6.59B-03                    2.5 4002 m, 0-30 da ONTROL ROOM PERSONNEL DOSES                                                3.1 lE-02                    S    Request for 0-30 DAYS                                                                                                      Licensing Action
            ....,.--====-'---------------------.~--""                                                                        Note:
Main Steam Line Break                                              For Control Room Scenario 1                                                Dose, see Maximum Equilibrium Iodine Reactor Coolant Concentration
* Attachment 6 "Results for Increas(
            ..---~C~;as!!se!.!2~Co~n!!,!trol~roo!!!m!!.!iso!!!!la~U!!oD!!f.ll~38!5~cti~m~unftl2!!!!te!!red~!!lnl~*;!!!l!eL... Control Room TEDE Dose TEDE Dose Volume" ie--......-==----..............---=-==---------11---.....,~em~--...-L*lm.*~t~em....._.(afterpage21) 5.87E-02                    2.S 6.SSB-03                    2.S 5.64B-03                *. s
 
Page21 I TITLE I
 
==SUBJECT:==
Main Steam Una Break Using Alternate Source Tanna                                    I 11.2 Scenario 2, Pre-existing Iodine Spike Main Steam Line Break Scenario2 Pre-existing Iodine Spike Case 1 Normal HVAC Dow of 6 600 cfm assumed with no control room Isolation)
TEDE Dose TEDE Dose em  Limit em ite Boundary (EAB)                                      1.17      25 863 m, 0-2 hrs w Population Zone (LPZ)                            l.31B-01    25 4002 m, 0-30 da s ONTROL ROOM PERSONNEL DOSES                          6.20B-01      5      Request for 0-30DAYS                                                                    Licensing Action lli!:==-====-=======-=---=======----......,----.._.lbqp              __...,.,Note:
For Control Room Main Steam Line Break                          Dose, see Scenarlo2                              Attachment 6 Pre-existing Iodine Spike                                                                nResults for Increase Control Room (Case 2, Control room lsolatio      1385 cfm unf"dtered inl                              Volume" TEDEDose                (next 2 pages) lte Boundary (EAB) 863 m, 0-2 hrs w Population Zone (LPZ)                            1.31B-01 4002 m, 0-30 da
* ONTROL ROOM PERSONNEL DOSES                          1.12E-01      5 0-30DAYS The effect of a larger control room volume is addressed in Attachment 6.
To provide a basis of comparison with the existing PNPP design basis, the offsite and onsite dose consequences are calculated on the basis of whole body and thyroid doses using the TID ICRP2 dose conversion factors given in the RADTRAD code. The results and output files are given in Attachment S.
 
Attachment 6, page 1 A1TAC8MENT 6 Results for Increased Control Room Volume The current design basis control room volume is being revised (Revision 2 of calculation M26-001). The preliminary results of this reanalysis indicate that the control room volume may increase to 420,494 ft3*
To quantify the impact of this change, the Scenario 1 and 2 cases will be rerun with a larger control room volume. To bound the potential increase in volume, a control room volume of 421,000 ft3 will be assumed. The input and output files are as follows:
Scenario 1 - Maximum Equilibrium Iodine Concentration The RADTRAD input and output files used for this Scenario are given below:
Description                              Case l, No CR isolation    Case 2, CR isolation Plant scenario file                        Peny MSLB 02a BIG.psf        Perry MSLB 02b BIG.psf Auxiliary RADTRAD Input Files Nuclide Inventory File            perry02.nif                  perry02.nif Release Fraction and Timing File            perrymslb.rft                perrymslb.rft Dose Conversion Factors              perry.inp                    perry.inp Oa1putFUe                                  Pem MSLB 02a BIG.out          Perry MSLB 02b BIG.out Scenario 2 - Pre-accident Iodine Spike Offsite and onsite doses for Scenario 2 were calculated using the RADTRAD model described above.
The RADTRAD output files are given in Attachments 3 and 4.
Description                              Case 1, No CR isolation    Case 2, CR isolation Plant scenario file                        Perry MSLB 40a BIG.psf        Perry MSLB 40b BIG.psf Awdllary RADTRAD Input Files Nuclide Inventory File            perry40.nif                  perry40.nif Release Fraction and Timing File            perrymslb.rft                perrymslb.rft Dose Conversion Factors              perry.inp                    perry.inp Output File                                Pel'l'J MSLB 40a BIG.out      Perg MSLB 40b BIG.out The results from these cases are given below Cnote; since the offsite doses are unaffected by this change, only the control room doses are Riven];
 
Attachment 6, page 2 Scenario 1, Maximum Equilibrium Iodine Reactor COOiant Concentration Main Steam Line Break Scenario 1 Maximum Equilibrium Iodine Reactor Coolant Concentration 1 Case 1, Normal HVAC now of 6,600 din assumed with no control room isoladon)
TEDEDose TEDEDose (Rem)  Limit<Rem)
CONTROL ROOM PERSONNEL DOSES                      3.llE-02        5 0-30DAYS)
Main Steam Line Break Scenario 1 Maximum Equilibrium Iodine Reactor Coolant Concentration Case 2 Control room isolatio 1385 din unfiltered inleak e TEDE Dose TEDE Dose em    Limit    em ONTROL ROOM PERSONNEL DOSES                    5.317E-03        5 0-30DAYS Scenario 2, Pre-existing Iodine Spike Main Steam Line Break Scenario2 Pre-existing Iodine Spike Case 1 Normal RVAC ftow of 6 '80 elm assumed with no control room isolation)
TEDE Dose TEDE Dose Limit ONTROL ROOM PERSONNEL DOSES
* 5        Request for
        ~0-~30~D:;;:,,;:A~Y.;;;;S'-'---------....i1~---...L.---.;;;;;:;iiililicensing Action Note:
Main Steam Line Break                          Maximum Control Scenario 2                              Room Dose Pre-existing Iodine Spike Case 2 Control room Isolation 1385 elm unfiltered lnl      e TEDE Dose TEDE Dose em    Limit em
* ONTROL ROOM PERSONNEL DOSES                      t.06E-01        S 0-30DAYS}}

Latest revision as of 00:22, 11 January 2025

Request for Licensing Action Pursuant to 10 CFR 50.59. 50.67. and 50.90: Full Implementation of Alternative Accident Source Term Design Basis Accident Analyses. and an Associated Technical Specification Change
ML13343A013
Person / Time
Site: Perry FirstEnergy icon.png
Issue date: 12/06/2013
From: Harkness E
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-13-306
Download: ML13343A013 (292)
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