ML12334A800

Official Exhibit - NYS000287-00-BD01 - NEI 05-01, Rev a, Severe Accident Mitigation Alternatives (SAMA) Analysis (November 2005) (NEI SAMA Guidance)

ML12334A800
Person / Time
Site:	Indian Point
Issue date:	11/30/2005
From:	Nuclear Energy Institute
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 21598, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML12334A800 (79)

Text

{{#Wiki_filter:United States Nuclear Regulatory Commission Official Hearing Exhibit In the Matter of: Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3) ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: Identified: Admitted: Withdrawn: Rejected: Stricken: Other: NYS000287-00-BD01 10/15/2012 10/15/2012 NYS000287 Submitted: December 21, 2011 c.,-VR REGU<..tr l~" 0 ~ ~ ~ ~ ~ i .... ~.., ~c$' NEI 05*01 [Rev A] Severe A~ccident MitigatiCJ.n Alternatives (SAMA) A~nalysis Guidance Document November 2005 OAG 10000585_00001

NEI 05*01 [Rev AI Nuclear El1lergy Institute Severe, Accident Mitigation Alternatives (SAMA) Analysis Guidance Document Novelnber 2005 Nuclear Energy Institute, 1776 I Street N. W, Suite 400, Washington D. C. (202.739.8000) OAGI0000585 00002

ACKNOWLEDGEMENTS The author would like to acknowledge the assistance of the following in the development of this guideline: Fred Polaski, Exelon Nuclear; Kenneth Brune, TVA; Bill Watson, Millstone Power Station License Renewal; Richard Gallagher, Millstone Power Station License Renewal; Jeff Gabor, ERIN Engineering and Research, Inc.; Stanley H. Levinson, AREV A; Alan B. Cox, Entergy License Renewal Services; and Lori Ann Potts, Entergy License Renewal Services. NOTICE Neither NEI, nor any of its employees, members, supporting organizations, contractors, or consultants make any warranty, expressed or implied, or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately owned rights. OAGI0000585 00003

EXECUTIVE

SUMMARY

This document provides a template for completing the severe accident mitigation alternatives (SAMA) analysis in support of license renewal. Its purpose is to identify the information that should be included in the SAMA portion of a license renewal application environmental report to reduce the necessity for Nuclear Regulatory Commission (NRC) requests for additional information (RAls). The method described relies upon NUREGIBR-OI84 regulatory analysis techniques, is a result of experience gained through past SAMA analyses, and incorporates insights gained from review of NRC evaluations of SAM A analyses and associated RAls. OAGI0000585 00004

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TABLE (.F CONTENTS EXECUTIVE SU M MARY *************************************.*****************.*********************************************** I 1 INTRODUCTION *.......**.*........***...........*.*...*......***..........**..*......*.**..............*****.*.........* 1 1.1 PURPOSE...................................................................................................................... 1 1.2 REQUIREMENTS........................................................................................................ 1 2 METHOD...*.......****.*..*...***.**....**.****.......*****..........**.*.*.......***.*..........*.*.**.*.**...*.....***.*** 2 3 SEVERE ACCIDENT RiSK *.****.****.****.*...*************..********.*******.***.****.***..*...****************.*** 4 3.1 LEVEL 1 PSA MODEL............................................................................................... 4 3.1.1 3.1.2 Internal Events................................................................................................ 4 External Events............................................................................................... 5 3.2 LEVEL 2 PSA MODEL............................................................................................. 10 3.2.1 3.2.2 Description Of Level 2 PSA Model............................................................. 11 Level 2 PSA Model Changes Since IPE Submittal *.*......**..*..................*.*. ll 3.3 MODEL REVIEW

SUMMARY

................................................................................ 12 3.4 LEVEL 3 PSA MODEL............................................................................................. 13 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 Population Distribution................................................................................ 13 Economic Data.............................................................................................. 13 Nuclide Release............................................................................................. 14 Emergency Response.................................................................................... 14 Meteorological Data...................................................................................... 15 3.5 SEVERE ACCIDENT RISK ru:SUL TS.................................................................. 15 4 COST OF SEVERE ACCIDENT RISK / MAXIMUM BENEFIT...................................... 16 4.1 OFF-SITE EXPOSURE COST.................................................................................. 16 4.2 OFF-SITE ECONOMIC COST................................................................................ 17 4.3 ON-SITE EXPOSURE COST................*.........*....................................................*.... 17 4.4 ON-SITE ECONOMIC COST.................................................................................. 19 4.5 TOTAL COST OF SEVERE ACCIDENT RISK / MAXIMUM BENEFIT......... 22 5 SAMA IDENTIFICATION..........................................................................*................ 23 5.1 PSA IMPORTANCE.................................................................................................. 23 5.2 PLANT IPE................................................................................................................. 24 5.3 PLANT IPEEE............................................................................................................ 24 5.4 INDUSTRY SAMA CANDIDATES.......................................................................... 24 5.5 LIST OF PHASE I SAMA CANDIDATES.............................................................. 24 III OAGI0000585 00006

6 PHASE I ANALYSIS ************************************************************************************************** 25 7 PHASE II SAMA ANALySiS ************************************************************************************** 27 7.1 SAMA BENEFIT......................................................................................................... 27 7.1.1 7.1.2 7.1.3 Severe Accident Risk \\Vith SAMA Implemented ************************************** 27 Cost of Severe Accident Risk with SAMA Implemented.......................... 28 SAMA Benefit **.*...*.........***...*........*.......*..*....*........*.............*...*.*..***..**.****... 28 7.2 COST OF SAMA IMPLEMENTATION...**......*.....*..*..*......*...*.....*...**....*...*.*.***..*. 28 8 SENSITIVITY ANALYSES ****************************************************************************************** 30 8.1 PLANT MODIFICATIONS........................................................................................ 30 8.2 UN CERTAINTY *...*..............*..*.. ' *.*...........*....*..**.................*...*.*.*....***....................... 30 8.3 PEER REVIEW FINDINGS OR OBSERVATIONS.............................................. 31 8.4 EVACUATION SPEED ***.......*.. ' *..****.*........*....*....*...*..........*..*...**.***..*.**..**.***.*.....*... 31 8.5 REAL DISCO UNT RATE..........*..**.*..*.....*..*...*..*......................*..*.*.*.*.*.*.*.***.....***.... 31 8.6 ANALYSIS PERIOD *......*..*....... ' *..*..*.**....**......*........*.....*..........*......*.*...***...*.....*....*. 32 9 CONCLUSiONS ********************************************************************************************************* 33 10 TABLES AND FIGURES ****************** ~ ************************************************************************* 34 TABLE 1 SAMPLE ACCIDENT CLASS DISTRIBUTION............................................................ 34 TABLE 2 SAMPLE RELEASE SEVERITY AND TIMING CLASSIFICATION SCHEME................ 34 TABLE 3 SAMPLE RELEASE CATEGORY FREQUENCY AND RELEASE FRACTIONS.................. (SOURCE TERM) ******************************************************************************************************************** 35 TABLE 4 SAMPLE ESTIMATED POPULATION DISTRIBUTION WITHIN A 50-MILE RADIUS.36 TABLE 5 SAMPLE MACCS2 ECONOMIC PARAMETERS...................................................... 37 TABLE 6 SAMPLE CORE INVENTORY VALUES..................................................................... 38 TABLE 7 SAMPLE RELEASE CHARACTERISTICS.................................................................. 39 TABLE 8 SAMPLE

SUMMARY

OF SEVERE ACCIDENT RISK RESULTS.................................. 39 TABLE 9 SAMPLE PSA IMPORTANCE REVIEW..................................................................... 40 TABLE 10 SAMPLE LIST OF PHASE I SAMA CANDIDATES..................................................41 TABLE 11 SAMPLE PHASE II SAMA LIST............................................................................ 42 TABLE 12 SAMPLE SENSITIVITY ANALYSIS RESULTS..........................................................43 TABLE 13 STANDARD LIST OF BWR SAMA CANDIDATES.................................................44 TABLE 14 STANDARD LIST OF PWR SAMA CANDIDATES................................................. 56 11 REFERENCES *********************************************************************************************************** 71 IV OAGI0000585 00007

Acronym AC AMSAC ATWS BWR CCW CDF CRD CST CS DC ECCS EDG EOP EPRI FIVE HPCI HRA HVAC IPE IPEEE ISLOCA LERF LOCA LOOP List of Acronyms Definition alternating current A TWS miti!~ation system actuation circuitry anticipated transient without scram boiling water reactor component cooling water core damage frequency control rod drive condensate storage tank containment spray direct current emergency core cooling system emergency diesel generator emergency operating procedure Electric Power Research Institute fire-induced vulnerability evaluation high pressure coolant injection human reliability analysis heating, ventilation, and air conditioning individual plant examination IPE - external events interfacing systems loss of coolant accident large, early release frequency loss of coolant accident loss of off-site power v OAGI0000585 00008

Acronym LPCI MACCS2 MCC MSIV NPSH NRC PSA PWR RAI RCIC RHR RHRSW RPV RWCU SAG SAMA SAMDA SBO SLC SMA SRV SW TBCCW USI Definition low pressure coolant injection MELCOR accident consequence code system motor control center main steam isolation valve net positivE! suction head Nuclear Regulatory Commission probabilistic safety assessment pressurized water reactor request for additional information reactor con3 isolation cooling residual heat removal residual heat removal service water reactor pressure vessel reactor water cleanup severe accident guidelines severe accident mitigation alternatives severe accident mitigation design alternatives station black-out standby liquid control seismic margins analysis safety relief valve service water turbine building closed cooling water unresolved safety issue VI OAGI0000585 00009

NEI 05-01 (Rev A) November 2005 SEVERE ACCIDENT IVIIITIGATION ALTERNATIVES (SAMA.) ANALYSIS GUIDANCE DOCUMENT 1 INTRODUCTION This document provides a template for completing the severe accident mitigation alternatives (SAMA) analysis in support of license renewal. Its purpose is to identify the information that should be included in the SAMA portion of Ii license renewal application environmental report to reduce the necessity for Nuclear Regulatory Commission (NRC) requests for additional information (RAls). The method described relies upon NUREGIBR-0184 regulatory analysis techniques, is a result of experience gaine:d through past SAMA analyses, and incorporates insights gained from review of NRC evaluations ofSAMA analyses and associated RAls. 1.1 PURPOSE The purpose of the analysis is to identify SAMA candidates that have the potential to reduce severe accident risk and to determine if implementation of each SAMA candidate is cost-beneficial. 1.2 REQUIREMENTS OAGI0000585 00010

NEI 05-01 (Rev A) November 2005 2 METHOD The SAMA analysis consists of the following steps. Determine Severe Accident Risk Levell and 2 Probabilistic Safety Assessment (PSA) Model Use the plant-specific PSA model (Section 3.1 through Section 3.3) as input to a Level 3 PSA analysis. Incorporate external event contributions as described in Section 3.1.2. Level 3 PSA Analysis Use Level I and 2 PSA output and site-specific meteorology, demographic, land use, and emergency response data as input for at Level 3 PSA (Section 3.4). Estimate the severe accident risk i.e., off-site dose and economic impacts of a severe accident. Determine Cost of Severe Accident Risk / Maximum Benefit - Use NRC regulatory analysis techniques to estimate the cost of severe accident risk. Estimate the maximum benefit that a SAMA could achieve if it eliminated all risk i.e., the maximum benefit (Section 4). SAMA Identification - Identify potential SAMA candidates (that prevent core damage and that prevent significant releases from containment) from the PSA model, Individual Plant Examination (lPE) and IPE - External Events (lPEEE) recommendations, and industry documentation (Section 5). As has belen demonstrated by past SAMA analyses, SAMA candidates are not likely to prove cost-beneficial if they only mitigate the consequences of events that present a low risk to the plant. Therefore, PSA importance analyses play a key role in the SAMA identification process. Preliminary Screening (Phase I SAMA Analysis) - Screen out SAMA candidates that are not applicable to the plant design, candidates that have already been implemented or whose benefits have been achieved at the plant using other means, and candidates whose roughly-estimated cost exceeds the maximum benefit. PSA insights may be used to screen SAMA candidates that do not address significant contributors to risk in this phase (Section 6). Final Screening (Phase II SAMA Analysis) - Estimate the benefit of severe accident risk reduction to each remaining SAMA candidate and compare to an implementation cost estimate to determine net cost-benefit (Section 7). In an implementation cost estimate, all costs associated with the SAMA should be considered including design, engineering, safety analysis, installation, and long-term maintenance, calibrations, training, etc. that will be required as a result of the change. As has been demonstrated by past SAMA analyses, cost-beneficial SAMAs are most likely limited to procedure changes and minimal hardware changes. 2 OAGI0000585 00011

NEI 05-01 (Rev A) November 2005 Sensitivity Analysis - Evaluate how changes in SAMA analysis assumptions and uncertainties would affect the cost-benefit analysis (Section 8). Identify Conclusions - Summarize results and identify conclusions (Section 9). List potentially cost-beneficial SAMA candidates. The remainder of this document describes these steps in more detail and indicates associated information that should be included in the SAMA portion of the license renewal environmental report. Figure 1 provides a graphical representation ofthe SAMA analysis process. 3 OAGI0000585 00012

NEI05-01 (Rev A) November 2005 3 SEVERE ACCIDENT RISK Describe the PSA models used to calculate severe accident risk. Describe the Level IPSA model (internal and external), the Level 2 PSA model, PSA model review history, and the Level 3 PSA model, as shown in Section 3.1 through Section 3.4. Include results of the severe accident risk calculation as shown in Section 3.5. For multi-unit sites, provide either separate results for each unit or results for a single unit with rationale for why the single analysis is representative or bounding for the other unit( s). 3.1 LEVEL 1 PSA MODEL Level 1 PSA models determine CDF based on initiating event analysis, scenario development, system analyses, and human-factor evaluations. 3.1.1 INTERNAL EVENTS 3.1.1.1 Description of Levell Internal Events PSA Model Identify and describe the Level 1 internal eventsPSA model used for the SAMA analysis, including the model freeze date. If different PSA versions are used for identifying SAMAs (Section 5.1) and for the benefit analysis (Section 7.1), the impact of using a later version should be described. For example, The Levell Internal Events PSA Model used for the SAMA analysis was the most recent internal events risk model (Revision xxx) that contains modeling of all plant changes implemented up to [date), uses failure and unavailability data to the same date, and resolves industry peer review comments on a previous revision of the model. Provide a breakdown of the internal events CDF by major contributors, initiators, or accident classes. Include contributions to core damage frequency from station blackout (single unit and dual unit) and anticipated transient without SCRAM events. Candidate SAMAs should concentrate on these events. Table 1 shows a typical accident class distribution. Provide Level I internal events importance measures. This list may be combined with an evaluation of applicable SAMA candidates as shown in Table 9. If applicable, identify changes to the Level I internal events PSA model made to accommodate the SAMA analysis. 3.1.1.2 Levell PSA Model Changes since IPE Submittal Describe major changes to the Levell internal events PSA model since the IPE submittal and the impact these changes have had on CDF. 4 OAGI0000585 00013

NEI 05-01 (Rev A) November 2005 Discuss changes to the plant, such as power uprate or steam generator replacement that are planned or have occurred since the model freeze date. Indicate if the model used for the SAMA analysis addresses these changes. If the model used for the SAMA analysis does not address these changes, include a qualitative discussion of the impact of the changes on the SAMA analysis. If desired, sensitivity analyses may be performed to support the discussion (Section 8). 3.1.2 EXTERNAL EVENTS The IPEEE identified the highest risk externally initiated accident sequences and potential means of reducing the risk posed by those sequences. Typically, the following external events were evaluated. Internal fires Seismic events Other external events such as high wind events, external flooding, transportation and nearby facility accidents The type of information available for these initiators vanes by the type of risk analysis performed for the IPEEE. For instance, a fire or seismic analysis performed using PSA modeling techniques produces quantitative results. However, due to differences in assumptions, model techniques, uncertainties (e.g., related to initiating event frequencies and human actions), care should be taken when comparing quantified external events with the results of the best-estimate internal events analysis. Furthelmore, seismic margins analysis (SMA) does not produce a CDF (i.e., is a qualitative analysis) and is predicated on the ability to evaluate the seismic durability of equipment required to safely shut the plant down. The results of this kind of analysis do not directly lend themselves to the frequency-based SAMA analysis. Also, a fire analysis using the Electric Power Research Institute (EPRI) Fire-Induced Vulnerability Evaluation (FIVE) method produces fire zone CDF values that are conservatively high and not suitable for comparison with best-estimate internal events CDF values. As a result, each of the external event contributors must be considered in a manner suiting the type of risk analysis performed. For each external event, summarize the risk analysis method and subsequent revisions as shown in Section 3.1.2.1 through Section 3.1.2.3. Discuss recommendations to reduce risk due to each external event, and indicate whether or not they have been implemented. Potential improvements from the IPEEE and improvements to address USI A-46 outliers that have not been implemented should be included in the list of Phase I SAMA candidates (Section 5.3). Describe the method used to quantitatively incorporate external event severe accident risk in the SAMA analysis, as shown in Section 3.1.2.4. OAGI0000585 00014

NEI 05-01 (Rev A) November 2005 3.1.2.1 Internal Fires 3.1.2.1.1 Risk Analysis Provide a brief discussion of the risk analysis method used for the IPEEE. Indicate if a fire PSA model was created, or if the EPRI FIVE method was used. If the EPRI FIVE method was used, identify first-pass assumptions and screening criteria (e.g., 1.0E-06) and discuss methods used to evaluate zones that did not screen on the first pass. Indicate if the fire risk analysis has been updated since the IPEEE. If so, provide revised fire zone CDF values. If the EPRI FIVE method was used, the results are conservative and not comparable to internal events core damage frequencies. If a fire PSA model was created, the results should be less conservative than if the FIVE method had been used, but caution must be exercised when making comparisons to best-estimate values. Discussion of specific conservatisms may be provided, as in the following examples. Initiating Events: System Response: Sequences: Fire Modeling: BRA: Level of Detail: The frequency 0.( fires and their severity are generally conservatively overestimated. A revised NRC fire events database indicates a trend toward lower frequency and less severe fires. This trend reflects improved housekeeping, reduction in transient fire hazards, and other improved fire protection steps at utilities, Fire protection measures such as sprinklers, CO2, and fire brigades may be given minimal (conservative) credit in their ability to limit the spread of afire. Cable routings are typically characterized conservatively because of lack of data regarding the routing of cables or lack of analytic modeling to represent the different routings. This leads to limited credit for balance of plant systems that are important in core damage mitigation. Sequences may subsume a number of fire scenarios to reduce the analytical burden. Subsuming initiators and sequences is done to envelope those sequences included. This results in additional conservatism. Fire damage and fire spread are conservatively characterized. Fire modeling presents bounding approaches regarding the immediate effects of a fire and fire propagation (e.g., all components in a fire zone are failed by a fire in the zone, or all cables in a tray are failed for a cable tray fire). There is little industry experience with crew actions followingfires. This has led to conservative characterization of crew actions in fire analyses. Because CDF is strongly correlated with crew actions, this conservatism has a profound effect on fire results. Fire analyses may have a reduced level of detail in mitigation of the initiating event and subsequent system damage. 6 OAGI0000585 00015

Quality of Model: Recommended Improvements NEI 05-01 (Rev A) November 2005 The peer review process for fire analyses is less well developed than for internal events PSAs. For example, no industry process, such as NEI 00-02, exists for the structured peer review of a fire PSA. Discuss existing fire prevention and mitigation features and recommended hardware or procedure changes (including those from the IPEEE and subsequent fire evaluations) to reduce risk in the dominant fire zones. For example, the dominant fire zones may be monitored by a detection system that alarms in the control room, and they may be equipped with automatic suppression systems. Electrical cabinets in the zones may use rated cables that are difficult to ignite and slow to propagate. Radiant energy shields may be used to prevent a fire on one component from disabling redundant components. Also, hot work permit and transient combustible loading programs reduce possible ignition sources and the fire protection program maximizes the availability of fire protection equipment. If this discussion duplicates in1hrmation provided to NRC for the IPEEE, reference to docketed correspondence may be substituted. Potential improvements to reduce risk in lhe dominant fue zones (including those from the internal fire portion of the IPEEE and subsequent fire evaluations) should be included in the list of Phase I SAMA candidates (Section 5.3). 3.1.2.2 Seismic Events 3.1.2.2.1 Risk Analysis Provide a brief discussion of the risk analysis method used for the IPEEE. Indicate if a seismic PSA model was created, or if the EPRI SMA method was used. Indicate if the seismic risk analysis has been updated since the IPEEE. If so, provide revised results. If a seismic PSA model was created, discuss whether the seismic CDF value is conservative or best-estimate. Discussion of specific conservatisms may be provided as in the examples for internal fires. Recommended Improvements Discuss enhancements (including those recommended in the IPEEE) to ensure equipment on the safe shutdown list is capable of withstanding a review level earthquake. Discuss USI A-46 resolution and whether all identified outliers have been addressed. If this discussion duplicates information provided to NRC for the IPEEE, reference to docketed correspondence may be substituted. 7 OAGI0000585 00016

NEI 05-01 (Rev A) November 2005 Potential improvements to minimize seismic risk (including those from the seismic events portion of the IPEEE, subsequent seismic evaluations, and improvements to address unresolved USI A-46 outliers) should be included in the list of Phase I SAMA candidates (Section 5.3). 3.1.2.3 Other External Events 3.1.2.3.1 Risk Analysis Provide a brief discussion of the risk analysis method used for the IPEEE and indicate if the analysis has been updated since the IPEEE. If so, provide revised results. Discussion of specific conservatisms may be provided as in the examples for internal fires. Recommended Improvements Describe existing prevention and mitigation features and recommended hardware or procedure changes from the IPEEE to reduce risk from external events caused by high winds, external flooding and transportation accidents, as applicable. If this discussion duplicates information provided to NRC for the IPEEE, reference to docketed correspondence may be substituted. Potential improvements to reduce risk from other external events (including those from the other events portion of the IPEEE) should be included in the list of Phase I SAMA candidates (Section 5.3). 3.1.2.4 External Event Severe Accident Risk Discuss the method used to address external event risk. As discussed previously, the preferred method is dependent on the risk analysis methods available for the plant. IPEEE reports typically concluded that the risk from other external events (i.e., not fire and seismic events) is less than IE-06lrx-yr. Therefore, these events are typically not the dominant contributors to external event risk and quantitative analysis. of these events is not practical. Thus, the various combinations of internal fire and seismic risk analysis are discussed below. FIVE and SMA Methods The SMA method does not provide a quantitative result, but resolution of outliers assures that the seismic risk is low and further cost-beneficial seismic improvements are not expected. Therefore, the FIVE results may be used as a measure of total external events risk. Estimate the degree of conservatism for the external events risk. Since a FIVE method fire analysis contains numerous conservatisms, as discussed previously, a more realistic assessment could result in a substantially lower fire CDF. NRC staff has accepted that a more realistic fire CDF may be a factor of three less than the screening value obtained from a FIVE analysis (Reference 1). Technical justification should be provided for selection of a reduction factor. Reduce the fire CDF by an appropriate factor and compare to the internal events CDF to estimate an external events multiplier. 8 OAGI0000585 00017

For example, NEI 05-01 (Rev A) November 2005 Assume that the total of the unscreenedfire zone CDFs from the FIVE analysis is 2.7E-05/rx-yr. Also, assume that the internal event CDF is 8E-06/rx-yr. Given afactor of three reduction, the resulting fire CDF would be about 9E-06/rx-year, which is the same order of magnitude as the internal events CDF. This would justify use of an external events multiplier of two. Use the external events multiplier on the maximum benefit (Section 4.5) and on the upper bound estimated benefits for individual SAMA candidates during the Phase II screening (Section 7). Fire PSA and SMA Method The SMA method does not provide a quantitative result, but resolution of outliers assures that the seismic risk is low and further cost-beneficial seismic improvements are not expected. Therefore, the fire PSA results may be used as a measure of total external events risk. Estimate the degree of conservatism for the external events risk. If the fire PSA analysis contains numerous conservatisms, as discussed previously, a more realistic assessment could result in a substantially lower fire CDF. Technical justification should be provided supporting determination of a reduction factor to obtain a more realistic fire CDF. Use the reduction factor on the baseline firePSA results and compare to the internal events CDF to obtain an external events multiplier as described for the FIVE method. Use the external events multiplier on the maximum benefit (Section 4.5) and on the upper bound estimated benefits for individual SAMA candidates during the Phase II screening (Section 7). FIVE Method and Seismic PSA Since the FIVE method and seismic PSA provide quantitative results, the results may be combined to represent the total external events risk. Estimate the degree of conservatism for the external events risk. Since a FIVE method fire analysis contains numerous conservatisms, as discussed previously, a more realistic assessment could result in a substantially lower fire CDF. NRC staff has accepted that a more realistic fire CDF may be a factor of three less than the screening value obtained from a FIVE analysis (Reference 1). Also, if the seismic PSA analysis contains numerous conservatisms, as discussed previously, a more realistic assessment could result in a substantially lower seismic CDF. Technical justification should be provided supporting determination of reduction factors to obtain more realistic fire and seismic CDF.values Reduce the fire and seismic CDF values by their factors, combine to obtain a total external events CDF, and compare to the internal events CDF to estimate an external events multiplier. 9 OAGI0000585 00018

NEI05-01 (Rev A) November 2005 For example, Assume that the total of the unscreenedfire zone CDFs from the FIVE analysis is 2.7E-05/rx-yr. Assume that the seismic PSA resulted in a CDF of 3E-6/rx-yr; which was estimated to be a factor of four higher than a best-estimate of seismic CDF. Also, assume that the internal event CDF is 8E-06/rx-yr. Given afactor of three reduction, the resultingfire CDFwould be about 9E-06/rx-year. Given afactor offour reduction, the resulting seismic CDFwould be about 8E-7/rx-yr. Thus, the total external events risk would be 9.8E-6, which is the same order of magnitude as the internal events CDF. This would justifY use of an external events multiplier of two. Use the external events multiplier on the maximum benefit (Section 4.5) and on the upper bound estimated benefits for individual SAMA candidates during the Phase II screening (Section 7). Fire PSA and Seismic PSA Since fire PSA and seismic PSA provide quantitative results, the results may be combined to represent the total external events risk. Estimate the degree of conservatism for the external events risk. If the fire PSA analysis contains numerous conservatisms, as discussed previously, a more realistic assessment could result in Ii substantially lower fire CDF. T(~chnical justification should be provided supporting determination of a reduction factor to obtain a more realistic fire CDF. Also, if the seismic PSA . analysis contains numerous conservatisms, as discussed previously, a more realistic assessment could result in a substantially lower seismi,c CDF. Technical justification should be provided supporting determination of a reduction factor to obtain a more realistic seismic CDF. Reduce the fire and seismic CDF values by their factors, combine to obtain a total external events CDF, and compare to the internal events CDF to estimate an external events multiplier (as in the above example). Use the external events multiplier on the maximum benefit (Section 4.5) and on the upper bound estimated benefits for individual SAMA candidates during the Phase II screening (Section 7). 3.2 LEVEL 2 PSA MODEL Level 2 PSA models determine release frequency, severity, and timing based on Levell PSA, containment performance, and accident progression analyses. 10 OAGI0000585 00019

3.

2.1 DESCRIPTION

OF LEVEL 2 PSA MODEL NEI 05-01 (Rev A) November 2005 Identify and describe the Level 2 PSA model used for the SAMA analysis, including the model freeze date. For example, The Level 2 PSA model usedfor the SA.MA analysis was the most recent model (Revision xxx) that contains modeling of all plant changes implemented up to [date}, uses failure and unavailability data to the same date and resolves industry peer review comments on a previous revision of the model. Provide a description of the release severity and timing scheme. This may be in paragraph form or like the example shown in Table 2. Provide a table or matrix describing the mapping of Level 1 accident sequences into Level 2 release categories and a description of the representative release sequences. Provide the release category frequencies and fission product release characteristics (release fractions, timing, and energy). Ifthe sum of release frequencies does not equal the total CDF, an explanation should be provided. Table 3 displays sample release category frequencies and release fractions. Provide Level 2 importance measures. These measures should not only be based on consideration of large early release frequerlcy contributors, but should consider other release categories that are major contributors to population dose. This* list may be combined with an evaluation of applicable SAMA candidates as shown in Table 9. If applicable, identify changes to the Level 2 PSA model made to accommodate the SAMA analysis. 3.2.2 LEVEL 2 PSA MODEL CHANGES SINCE IPE SUBMITTAL Describe changes to major modeling assumptions, containment event tree structure, accident progression / source term calculations, or binning of endstates in the Level 2 PSA model since the IPE submittal and the impact these changes have had on large, early release frequency (LERF). Discuss changes to the plant, such as power uprate or steam generator replacement that are planned or have occurred since the model fre:eze date. Indicate if the model used for the SAMA analysis addresses these changes. If the model used for the SAMA analysis does not address these changes, include a qualitative discussion of the impact of the changes on the SAMA analysis. If desired, sensitivity analyses may be performed to support the discussion (Section 8). 11 OAGI0000585 00020

NEI 05-01 (Rev A) November 2005 3.3 MODEL REVIEW

SUMMARY

Provide a brief description of in-house and peer reviews of the Levelland 2 PSA models that have been perfonned since the IPE. For example, The model has been updated several times since completion of the IPE to maintain it consistent with the as-built plant, to incorporate improved thermal hydraulic results, and to incorporate PSA improvements. The updates have involved a cooperative effort including both licensee personnel and PSA consultant support. In each of the updates, an independent review of revisions to the PSA model is performed. The PSA model and results have been maintained as plant calculations or engineering reports. As part of each major update, in order to ensure adequacy of the updated model, an expert panel reviews the PSA model results. The panel is typically composed of experienced personnel from various plant organizations, including Operations, System Engineering, Design Engineering, Safety Analysis, and PSA. An Owner's Group peer review of the model was conducted in [date). The results of this review are described below. In addition, Nuclear Regulatory Commission (NRC) Staff reviewed results of the prior version of the model as part of the benchmarking of the Significance Determination Program Notebook. The Staff and its contractors conducted the review at the site during [date). The Staff further reviewed the model, primarily the human reliability analysis and fire risk analysis, as part of its review of the risk impact of extended power uprate. This review included a site visit in (date). Provide a brief description of the overall findings of the owner's group peer review. Discuss signific;:mt findings or observations and indicate if resolution was included in the model used for the SAMA analysis. If the model used for the SAMA analysis does not address significant fmdings or observations, include at least a qualitative discussion of the impact of the findings or observations on the SAMA analysis. Sensitivity analyses may be perfonned to support the discussion (Section 8). 12 OAGI0000585 00021

3.4 LEVEL 3 PSA MODEL NEI 05-01 (Rev A) November 2005 Level 3 PSA models determine off-site dose and economic impacts of severe accidents based on Level I PSA results, Level 2 PSA results, atmospheric transport, mitigating actions, dose accumulation, early and latent health effects, and economic analyses. Provide a description of the Level 3 analysis method and input data. In many SAMA analyses, the MELCOR Accident Consequence Code System (MACCS2) (Reference 2) is used to calculate the off-site consequences of a severe accident. Some SAMA analyses have used previous Level 3 analyses such as those included in NUREG/CR-4551. Description of the method may be no more than a reference to the document describing the method. However, the various input parameters and associated assumptions must still be described. The following sections describe input data if MACCS2 (Reference 2) is the analysis tool. If another code is used, similar description of the input parameters must be documented. 3.4.1 POPULATION DISTRIBUTION Provide a predicted population within a 50-mile radius of the site. The predicted population distribution may be obtained by extrapolating publicly available census data. Transient population included in the site emergency plan should be added to the census data before extrapolation. Explain why the population distribution used in the analysis is appropriate and justify the method. used for population extrapolation. Typically, with increasing population, the predicted population is estimated for a**ye.ar within the second half of the period of extended operation. Extrapolation to a later date, and therefore a larger population, adds conservatism to the analysis. Of course, if a population reduction is projected~ extrapolation to an earlier date would be more reasonable. The population distribution should be by location in a grid consisting of sixteen directional sectors, the first of which is centered on due north, the second on 22.5 degrees east of north, and so on. The direction sectors should be divided into a number of radial intervals extending out to at least 50 miles. A sample population distribution is provided in Table 4. 3.4.2 ECONOMIC DATA Provide economic data from publicly available information (e.g., from the u.S. Census Bureau, U.S. Department of Agriculture, or state tax office) on a region-wide basis. Economic data should be expressed in today's dollars (dollars for the year in which the SAMA analysis is being performed), not extrapolated to the end of thc~ period of extended operation. Economic data from a past census can be converted to today's dollars using the ratio of current to past consumer price indices. Describe the values and bases for the following economic estimates. Cost of evacuation Cost for temporary relocation (food, lodging, lost income) Cost of decontaminating land and buildings 13 OAGI0000585 00022

NEI 05-01 (Rev A) November 2005 Lost return on investments from properties that are temporarily interdicted to allow contamination to be decreased by decay of nuclides Cost of repairing temporarily interdicted property Value of crops destroyed or not grown because they were contaminated by direct deposition or would be contaminated by root uptake Value of farmland and of individual, public, and non-farm commercial property that is condemned Sample MACCS2 economic data is provided in Table 5. 3.4.3 NUCLIDE RELEASE Provide a discussion of the source of core inventory values and a list of those values. Table 6 shows sample core inventory values. The aetuallist of radioisotopes may differ from the list in Table 6. MACCS2 default core inventory values are for a reference plant with a power level of 3,412 megawatts-thermal. Since actual core inventory is usually fuel vendor proprietary information, plant-specific core inventory values may be obtained by scaling the MACCS2 default values by the ratio of power level to reference plant power level. Additional adjustment of the core inventory values may be necessary to account for differences between fuel cycles expected during the period of extended operation and the fuel cycle upon which the MACCS2 default core inventory values are based. Also provide a description of the characteristics associated with the release (i.e., elevation of release, thermal content of release). Use of a release height equal to half the height of the containment is acceptable, because it provides adequate dispersion of the plume to the surrounding area. Table 7 shows example release characteristics. 3.4.4 EMERGENCY RESPONSE Discuss emergency response and evacuation parameter assumptions. Provide an evacuation start time delay and a radial evacuation speed based on site-specific information. Since population dose is highly dependent on radial evacuation speed, and uncertainties may be introduced during derivation of a single evacuation speed from emergency plan information, sensitivity analyses should be documented to show that the radial evacuation speed used in the SAMA analysis is reasonable (Section 8.4). Best-estimate values for groundshine and cloudshine shielding factors are acceptable (e.g., Grand Gulfvalues found in Table 3.28 of Reference 3). MACCS2 default values are acceptable for other parameter inputs, such as inhalation and skin protection factors, acute and chronic exposur1e effects, and long-term protective data. 14 OAGI0000585 00023

3.4.5 METEOROLOGICAL DATA NEI 05-01 (Rev A) November 2005 Describe the meteorological data used in the analysis, including wind speed, wind direction, stability class, seasonal mixing heights, and precipitation. Indicate the sources of the data (e.g., site meteorological tower, National Climatic Data Center). Also indicate the span of the data. Examplt~s include, "a foil year (2003) of consecutive hourly values," or "an average offive years (1995-2003) of consecutive hourly values." Explain why the data set and data period are representative and typical. For example, Annual meteorology data sets from 1998 through 2000 were investigated for use in MACCS2. The 1998 data set was found to result in the largest doses and was subsequently used to create the one-year sequential hourly data set used in MACCS2. The conditional dose from each of the other years was within 10 percent of the chosen year. If data is not from the plant meteorological tower, discuss why the data is acceptable. 3.5 SEVERE ACCIDENT RISK RESULTS Provide the mean annual off-site dose and economic impact due to a severe accident for each of the release categories analyzed. Report results for all release categories, including those with normal containment leakage (intact containment). Provide total off-site dose and total economic impact, which are the baseline risk measures from which the maximum benefit is calculated (Section 4). Table 8 provides a sample summary of severe accident risk results. 15 OAGI0000585 00024

NEI05-01 (Rev A) November 2005 4 COST OF SEVERE ACCIDENT RISK / MAXIMUM BENEFIT Using the baseline risk measures from Section 3.5, calculate severe accident impacts in four areas: off-site exposure cost, off-site economic cost, on-site exposure cost, and on-site economic cost (Section 4.1 through Section 4.4). The following descriptions of the severe accident impact calculations are based on the NRC-accepted methods found in NUREGIBR-OI84 (Reference 4). Calculation of severe accident impacts involves an analysis period term, tf, which can be defined as either the period of extended operation (20 years), or the years remaining until the end of facility life (from the time of the SAMA ana.lysis to the end of the period of extended operation) (25 years or more). The value typically used for this term is the period of extended operation (20 years). Since this is a license renewal application, if the analysis determines that an aging-related SAMA is potentially cost-beneficial, the plant is under no obligation to implement the SAMA immediately. Thus, the plant will commit to implementing the SAMA by the beginning of the period of extended operation. Therefore, the benefits of the SAMA are only assured for 20 years. However, NRC has asked several plants to perform a sensitivity analysis using the period from the time of the SAMA analysis to the end of the period of extended operation to determine if SAMAs are potentially cost-beneficial if performed immediately. This sensitivity analysis should be performed to provide the information wanted by the regulator (Section 8.6). Alternatively, the analysis could use the period from the time of the SAMA analysis to the end of the period of extended operation (25 years or more), and a sensitivity analysis would not be needed. This method adds conservatism to the analysis.. Calculation of severe accident impacts also involves a real discount rate, r, which is typically assumed to be 7% (0.07/year) as recommended in NUREGIBR-OI84. A value of 7% is conservative because cost estimates are usually performed by utilities using values between 11 and 15%. Use of both a 7% and 3% real discount rate in regulatory analysis is specified in Office of Management Budget (OMB) guidance (Reference 5) and in NUREG/BR-0058 (Reference 6). The two discount rates represent the difference in whether a decision to undertake a project requiring investment is viewed as displacing either private investment or private consumption. A rate of 7% should be used as a baseline for regulatory analyses and represents an estimate of the average before-tax rate of return on an average investment in the private sector in recent years. A rate of3% should also be used and represents an estimate of the "consumption rate of interest," i.e., the real, after-tax rate of return on widely available savings instruments or investment opportunities. To address this concern, perform a sensitivity analysis using a 3% real discount rate (Section 8.5). Combine the severe accident impacts with the external events multiplier to estimate the total cost of severe accident risk. Since this is the maximum benefit that a SAMA could achieve if it eliminated all risk, it is the maximum benefit (Section 4.5). 4.1 OFF-SITE EXPOSURE COST Convert the baseline off-site dose to dollars using the conversion factor of $2,000 per person-rem, and discount to present value using the following equation. 16 OAGI0000585 00025

= C

• Zpba Where:

W pba off-site exposure cost ($) C [1-exp( -rtr) l/r (years) tf analysis period (years) (see Section 4) r = real discount rate (7% = O.07/year) (see Section 4) NEI 05-01 (Rev A) November 2005 Zpha = value of public health (acddent) risk per year before discounting ($/year) Zpha = $2,OOO/person-rem

• mean annual off-site dose impact due to a severe accident from Section 3.5 For example, Assume the baseline off-site dose from Section 3.5 is 9 person-rem/year.

Then, Zpha = 9 person-rem/year * $2,000lperson-rem = $18,000/year. Assume a 20-year analysis period and a 7% real discount rate. Then, C is approximately 10.76 years. Therefore, off-site exposure cost is 10. 76 years * $18, OOO/year = $193,680. 4.2 OFF-SITE ECONOMIC COST Discount the off-site economic cost to present value using the same equation as in Section 4.1, with Zpba = mean annual economic impact due to a severe accident from Section 3.5. F or example, Assume the baseline off-site economic impact from Section 3.5 is $21.000/year, then Zpha $21,000/year. Assume the same analysis period and real discount rate. Then, off-site economic cost = 10.76 years * $21,000/year = $225,960. 4.3 ON-SITE EXPOSURE COST The values for on-site (occupational) exposure consist of "immediate dose" and "long-term dose." The best estimate value provided in NUREGIBR-OI84 for immediate occupational dose 17 OAGI0000585 00026

NEI05-01 (Rev A) November 2005 is 3,300 person-rem/event, and long-term.occupational dose is 20,000 person-rem (over a ten-year clean-up period). The following equations are used to calculate monetary equivalents. Immediate Dose WIO = Where: W10 R F DIO C r tf For example, R*F*DIO*C = = = = = = = immediate on-site exposure cost ($) monetary equivalent of unit dose ($/person-rem) Level 1 internal events core damage frequency (events/year) immediate on-site (occupational) dose (person-rem/event) [l-exp( -rtf) ]/r (years) real discount rate (7% = 0.07/year) (see Section 4) analysis period (years) (see Section 4) Using the following values from above, R = $2,000/person-rem r 0.07/year DlO = 3,300 person-rem/event 1j = 20 years And assuming the Level 1 internal events core damage frequency, F = 1 E-6 events/year Then, the immediate on-site exposure cost is: Wio = $2,000/person-rem ;f: 1E-6 events/year

• 3,300 person-rem/event
• 10. 76 years

= $71 Long-Term Dose WLTO = Where: WLTO R F DLTO C R

• F
• DLTO
• C * {[I - exp(-rm)]/rm}

long-term on-site exposurc~ cost ($) monetary equivalent of unit dose ($/person-rem) Levell internal events core damage frequency (events/year) long-term on-site (occupational) dose (person-rem/event) = [l-exp( -rtf) ]/r (years) 18 OAGI0000585 00027

r = real discount rate (7% = O.07/year) (see Section 4) tr = analysis period (years) (see Section 4) m years over which long-tenn doses accrue For example, Using the following values from above, R $2,000/person-rem r 0.07/year DLTO 20,000 person-rem/event m 10 years If 20 years F 1 E-6 eventslyear Then, the long-term exposure cost is: NEI 05-01 (Rev A) November 2005 WLTO $2,000/person-rem

• 1 E-6 eventslyear
• 20,000 person-rem/event
• 10.76 years
• {[J -exp(-0.07*IO)}/O.07*1Oj

$310 Total On-site Exposure - Combining immediate and long-term on-site exposure costs results in a total on-site exposure cost, W 0, of Wo = WIO+WLTO For the example, Wo = ($71 + $310) = $381 4.4 ON-SITE ECONOMIC COST On-site economic cost includes cleanup and decontamination cost, and either replacement power cost or repair and refurbishment cost. Cleanup and Decontamination Integrate the net present value of the total cost of clean-up and decontamination of a power reactor facility subsequent to a severe accident over the analysis period. The total cost of cleanup and decontamination of a power reactor facility subsequent to a severe accident is estimated in NUREGIBR-OI84 to be $1.5E+9. Calculate the present value of this cost as follows. PVCD = [CCD/m] * {[I - exp(-rm)]/r} 19 OAGI0000585 00028

NEI05-01 (Rev A) November 2005 Where: PVco Cco m r For example, = = = net present value of a single event ($) total cost of cleanup and decontamination effort ($) cleanup period (years) real discount rate (7% = O.07/year) (see Section 4) Using the following values from above, CCD $1.5E+9 m 10 years r 0.07/year Then, PVCD = $1.5E+9 / 10 years * {[1 -exp(-0.07*1O)]I0.07/year} = $1.08E+9 Integrate this cost over the analysis period as follows. Uco = PVco* C Where: Uco PVco C = total cost of dean up and decontamination over the analysis period ($-years) = net present value of a single event ($) r tf For example, = [ l-exp( -rtf) ]/r real discount rate (7% = O.07/year) (see Section 4) analysis period (years) (see Section 4) Using the following values from above, PVCD

=

r ft $1.08E+9 0.07lyear 20 years Then, the cleanup and decontamination cost is, $1.08E+9

• 10.76 years = 1. 16E+1O $-years Replacement Power Cost Determine the net present value of replacement power for a single event, PVRP, using the following equation.

[B/r] * [1 - exp( -rtf)]2 20 OAGI0000585 00029

NEI05-01 (Rev A) November 2005 Where: PV RP = net present value of replacement power for a single event, ($) r = real discount rate (7% = 0.07/year) (see Section 4) tf = analysis period (years) (see Section 4) B = a constant representing a s.tring of replacement power costs that occur over the lifetime of a reactor after an event (for a 91 OMWe "generic" reactor, NUREGIBR-0184 uses a value of$1.2E+8) ($/yr) For example, Assuming a 1023 MW e plant, and scaling B for power level, B = 1.2E+8$lyr

• 1023/910 = 1.35E+8$/yr Using the following values from above, r

0.07/year tr 20 years Then, PVRP = [1.35E+8$/yrI.07/yr] * [1 - exp(-.07*20J/ = $1.09E+9 Sum the single-event costs over the entire analysis period, using the following equation. [PVRP Ir] * [1 - exp(-rtf)i Where: URP net present value of replacement power over life of facility ($-year) r = real discount rate (7% = 0.07/year) (see Section 4) tf = analysis period (years) (se1e Section 4) For example, Using the following values from above, PVRP $1.09E+9 r 0.07/year tr 20 years Then, the replacement power cost is, [$1.09E+9/0.07/year] * [1 - exp(-0.07*20J/ = 8. 84E+9 $-years Repair and Refurbishment Cost Repair and refurbishment costs may be estimated in accordance with NUREG/BR-0184 as 20% of the cost of replacement power previously discussed. Assuming that replacement power will 21 OAGI0000585 00030

NEI 05-01 (Rev A) November 2005 be required for the remaining life of the plant results in higher benefit estimates and is, therefore, more conservative than assuming the plant will be repaired. Thus, repair and refurbishment costs need not be estimated. Total On-Site Economic Cost Calculate total on-site economic costs by summing cleanup/decontamination costs and replacement power costs, and multiplying this value by the internal events CDF. For example, Using the values from above and assuming an internal events CDF of 1E-6/year, Total onsite economic cost = (1.16E+10 $-years + 8.84E+9 $-years)

• 1E-6/year = $20,440.

4.5 TOTAL COST OF SEVERE ACCIDENT RISK / MAXIMUM BENEFIT Calculate the severe accident impact by summing the off-site exposure cost, off-site economic cost, on-site exposure cost, and on-site economic cost. For the example, the sum of the baseline costs is as follows. Off-site exposure cost $193,680 Off-site economic cost $225,960 On-site exposure cost $381 On-site economic cost $20,440 Severe accident impact $440,461 Combine the severe accident impact with the external events multiplier (Section 3.1.2.4) to calculate the total cost of severe accident risk. Since this is the maximum benefit that a SAMA could achieve if it eliminated all risk, it is the maximum benefit. For example, lfthe external events multiplier in Section 3.1.2.4 is two, Maximum benefit = $440,461

• 2 = $880,922 The maximum benefit is used in the Phase I screening process (Section 6) to eliminate SAMAs that are not cost-beneficial. If the estimated cost of implementing a SAMA exceeds this value, it is excluded from further analysis.

22 OAGI0000585 00031

5 SAMA IDENTIFICATION NEI 05-01 (Rev A) November 2005 Develop a list of SAMA candidates by reviewing the major contributors to CDF and population dose based on the plant-specific risk assessment and the standard BWR or PWR list of enhancements (Table 13 or 14). The following sections provide a more detailed description of the identification process and the necessary documentation. 5.1 PSA IMPORTANCE Identify plant-specific SAMA candidates by reviewing dominant risk contributors (to both CDF and population dose) in the Level I and Level 2 PSA models. Describe how dominant risk contributors, including dominant sequences, equipment failures, and operator actions identified through importance analyses, were used to identify plant-specific SAMA candidates. This should include a review of dominant sequences or cutsets for failures that could be addressed through an enhancement to the plant. It should also include a similar review of dominant equipment and human failures based on importance measures. Past SAMA analyses have shown that SAMA candidates are not likely to prove cost-beneficial if they only mitigate the consequences of events that present a low risk to the plant. The definition of "dominant sequences or cutsets" is open to interpretation. The SAMA portion of the license renewal environmental report should indicate how the dominant sequences were defined and the rationale for the cutoff value. For example, "The top 100 Levell cutsets, representing 62% of the total CDF, were reviewed. Individual cutsets below this point have little influence on CDF and are therefore not likely contributors for identification of cost beneficial enhancements. " Similarly, the definition of dominant equipment and human failures is open to interpretation. The SAMA portion of the license renewal environmental report should indicate how the dominant failures were defined and the rationale for the cutoff value. For example, "Failures with risk reduction worth> 1.005 were identified as the most important failures. Events below this point influence CDF by less than 0.5% and are therefore not likely contributors for identification of cost beneficial enhancements. " Provide a list of equipment failures and human actions that have the greatest potential for reducing risk based on importance analysis. For each dominant contributor describe relevant Phase I SAMAs and list the Phase II SAMA(s) that address that contributor. SAMAs may be hardware changes, procedure changes, or enhancements to programs, including training and surveillance programs. Hardware changes should not be limited to permanent changes involving addition of new, safety-grade equipment, but should also include lower cost alternatives, such as temporary connections using commercial grade equipment (e.g., portable generators and temporary cross-ties). Previous SAMA analyses for similar plants are a prime source for identifying potential low-cost alternatives to address similar risk contributors. If a SAMA was not evaluated for a dominant risk contributor, justify why SAMAs to further reduce the contributor would not be cost-beneficial. A sample partial PSA importance review is provided in Table 9. 23 OAGI0000585 00032

NEI 05-01 (Rev A) November 2005 5.2 PLANT IPE Plant IPE submittals included a list of risk-based insights and potential plant improvements. Identify if potential improvements have not been implemented. Include potential improvements that have not been implemented in the list of Phase I SAMA candidates. 5.3 PLANT IPEEE Potential improvements to reduce the risk in dominant fire zones and to reduce seismic risk and risk from other external events (including those from the IPEEE, subsequent fire and seismic evaluations, and improvements to address USI A-46 outliers) should be included in the list of Phase I SAMA candidates. 5.4 INDUSTRY SAMA CANDIDATES Include the generic BWR or PWR enhancements (Table 13 or 14) in the list of Phase I SAMA candidates. 5.5 LIST OF PHASE I SAMA CANDIDATES. The combined list of potential improvements from Section 5.1 through Section 5.4 is the list of Phase I SAMA candidates. Maintain this comprehensive list of SAMA candidates, with the source of each candidate indicated, in on-site documentation. Due to its size and limited value to NRC reviewers, this list need not be included in the SAMA portion of the license renewal environmental report. A sample partial list of Phase I SAMA candidates is presented in Table 10. The last two columns in this table are part of the Phase I analysis and are discussed in Section 6. 24 OAGI0000585 00033

6 PHASE I ANALYSIS NEI 05-01 (Rev A) November 2005 Perform a preliminary screening of SANlA candidates to eliminate SAMAs from further consideration. This step is taken to limit the number of SAMAs for which detailed analysis in Phase II is necessary. Describe the screening criteria used in the Phase I analysis. The following are examples of screening criteria that may be applied. Not Applicable: If a SAMA candidate does not apply to the plant design, it is not retained. For example, installation of accumulators for turbine-driven feedwater pump flow control valves would not require further analysis at a plant with motor operated turbine-driven feedwater pump flow control valves. Already Implemented: If a SAMA candidate has already been implemented at the plant, it is not retained. For example, installation of motor generator set trip breakers in the control room to reduce the frequency of core damage due to an ATWS would not require further analysis at a plant with a control room actuated diverse scram system. Combined: If a SAMA candidate is similar in nature and can be combined with another SAMA candidate to develop a more comprehensive or plant-specific SAMA candidate, only the combined SAMA candidate is retained. For example, addition of an independent reactor coolant pump seal injection system and use of an existing hydro test pump for reactor coolant pump seal injection provide similar risk-reduction benefits. If the lower-cost alternative is not cost-beneficial, the higher-cost alternative also will not be cost-beneficial. Therefore, the higher-cost alternative would not require further analysis. Excessive Implementation Cost: If a SAMA requires extensive changes that will obviously exceed the maximum benefit (Section 4.5), even without an implementation cost estimate, it is not retained. For example, the cost of installing an additional, buried off-site power source would exceed the maximum benefit from Section 4.5 and would not require further analysis. Consideration should be given to lower cost alternatives, such as temporary connections using commercial grade equipment (e.g., portable generators and temporary cross-ties), procedure enhancements, and training enhancements that could offer much of the potential risk reduction at a fraction of the cost of safety-related modifications. Very Low Benefit: If a SAMA from an industry document is related to a non-risk significant system for which change in reliability is known to have negligible impact on the risk profile, it is not retained. For example, if the instrument air system is not a risk-significant system at the plant, and failure of the air compressors is not on the PSA importance list (Section 5.1), the plant risk profile would be unchanged if the air compressors were made perfectly reliable. Therefore, an improvement to replace the: current air compressors with a more reliable model would not require further analysis. 25 OAGI0000585 00034

NEI 05-01 (Rev A) November 2005 Provide a description of the screening process and its results, in sufficient detail that a reader can understand how the initial set of Phase I SAlv1As was reduced to the more limited set of Phase II SAMAs (e.g., an accounting of the SAMAs eliminated by each criterion.) Table 10 provides sample Phase I dispositions for individual SAMA candidates. Those SAMAs that require detailed cost-benefit analysis are retained for Phase II analysis (Section 7). 26 OAGI0000585 00035

7 PHASE II SAMA ANALYSIS Perform a cost-benefit analysis on each ofthle remaining SAMA candidates. NEI 05-01 (Rev A) November 2005 The benefit is the difference in the baseline cost of severe accident risk (maximum benefit from Section 4.5) and the cost of severe accident risk with the SAMA implemented (Section 7.1). The cost is the estimated cost to implement the SAMA (Section 7.2). If the estimated cost of implementation exceeds the benefit of implementation, the SAMA is not cost-beneficial. For multi-unit sites, assure that the benefits and implementation costs are provided on a consistent basis, e.g., all benefit and all cost estimates are on a per-site basis. Ifbenefit and cost estimates are provided on a per-unit basis, the impact (and efficiencies) associated with implementation of the SAMA at multiple units should be reflected in the estimated implementation costs. 7.1 SAMA BENEFIT 7.1.1 SEVERE ACCIDENT RISK \\VITH SAMA IMPLEMENTED Perform bounding analyses to determine the change in risk following implementation of SAMA candidates or groups of similar SAMA candidates. F or each analysis case, alter the Level I internal events or Level. 2 PSA model to conservatively consider implementation of the SAMA candidate(s). Then, calculate the severe accident risk measures using the same procedure used for the baseline case described in Section 3. For SAMAs specifically related to external events, estimate the approximate benefits through use of the external events PRA, if available, or bounding-type analysis, (e.g., estimating the benefit of completely or partially eliminating the external event risk). Describe the changes made to the PSA models for each analysis case. For example, LBLOCA This analysis case was used to evaluate the change in plant risk profile that would be achieved if a digital large break LOCA protection system was installed. Although the proposed change would not completely eliminate the potential for a large break LOCA, a bounding benefit was estimated by removing the large break LOCA initiating event. This analysis case was used to model the benefit of SAMA 7. 27 OAGI0000585 00036

NEI 05-01 (Rev A) November 2005 DCPWR This analysis case was used to evaluate plant modifications that would increase the availability of Class IE DC power (e.g., increased battery capacity or the installation of a diesel-powered generator that would effectively increase battery capacity). Although the proposed SAMAs would not completely eliminate the potential failure, a bounding benefit was estimated by removing the battery discharge events and battery failure events. This analysis case was used to model the benefit of SAMAs 4, 5, 10, 12, and 24. 7.1.2 COST OF SEVERE ACCIDENT RISK WITH SAMA IMPLEMENTED Using the risk measures from Section 7.1.1, calculate severe accident impacts in four areas: off-site exposure cost, off-site economic cost, on-site exposure cost, and on-site economic cost using the same procedure used for the baseline casl~ described in Section 4. As in Section 4.5, sum the severe accident impacts and combine with the external events multiplier (Section 3.1.2.4) to estimate the total cost of severe accident risk with the SAMA implemented. Use of the external events multiplier is inappropriate for some SAMAs. For example, SAMAs specifically related to external events that would not impact internal events (e.g., enhanced fire detections) and SAMAs related to specific internal event initiators (e.g., guard pipes for main steam line break events). Provide a discussion of SAMAs on which the external events multiplier was not applied. 7.1.3 SAMA BENEFIT Subtract the total cost of severe accident risk with the SAMA. implemented from the baseline cost of severe accident risk (maximum benefit from Section 4.5) to obtain the benefit. List the estimated benefit for each SAMA candidate. Table 11 provides a sample portion of a Phase II SAMA candidate list with estimated benefits listed. 7.2 COST OF SAMA IMPLEMENTATION Perform a cost estimate for each of the Phase II SAMA candidates. Describe the cost estimating process and list the cost estimate for each SAMA candidate. As SAMA analysis focuses on establishing the economic viability of potential plant enhancement when compared to attainable benefit, often detailed cost estimates are not required to make informed decisions regarding the: economic viability of a particular modification. SAMA implementation costs may be clearly in excess of the attainable benefit estimated from a particular analysis case. For less clear cases, engineering judgment may be applied to determine if a more detailed cost estimate is necessary to formulate a conclusion regarding the economic viability of a particular SAMA. Nonetheless, the cost of each SAMA candidate should be conceptually estimated to the point where economic viability of the proposed modification can he adequately gauged. 28 OAGI0000585 00037

NEI 05-01 (Rev A) November 2005 For hardware modifications, the cost of implementation may be established from existing estimates of similar modifications from previously performed SAMA and SAMDA analyses. Costs associated with implementation of a SAMA including procurement, installation, long-term maintenance, surveillance, calibration, and training should be considered. Discuss conservatisms in the cost estimates. For example, cost estimates may not include the cost of replacement power during extended outages required to implement the modifications. They also may not include contingency costs associated with unforeseen implementation obstacles. Estimates based on modifications that were implemented or estimated in the past may be presented in terms of dollar values at the time of implementation (or estimation), and not adjusted to present-day dollars. In addition, implementation costs originally developed for SAMDA analyses (i.e., during the design phase of the plant) do not capture the additional costs associated with performing design modific:ations to existing plants (i.e., reduced efficiency, minimizing dose, disposal of contaminated material, etc.). Table 11 provides a sample portion of a Phase II SAMA candidate list with cost estimates. 29 OAGI0000585 00038

NE105-01 (Rev A) November 2005 8 SENSITIVITY ANALYSES Evaluate how changes in SAMA analysis assumptions would affect the cost-benefit analysis. Perform the following sensitivity analyses, as applicable. Table 12 contains sample sensitivity analysis results. 8.1 PLANT MODIFICATIONS Major changes to the plant, such as power uprate or steam generator replacement, may be planned or may have occurred since the model freeze date, as described in Section 3.1 and Section 3.2. If the Levell or Level 2 PSA model used for the SAMA analysis does not address a major plant change, a sensitivity analysis may be performed to support discussion of the impact of the change on the SAMA analysis results. In this sensitivity analysis, modifY the PSA model (or its results) to simulate incorporation of the plant modification and perform the Phase II analysis with the revised severe accident risk results. Sufficient margin exists in the maximum benefit estimation that the Phase I screening should not have to be repeated in the sensitivity analysis. Discuss the plant modification and how its effects were simulated in the PSA model. Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis. If SAMAs appear cost-beneficial in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimates discussed in Section 7.2), and their impact on the results may be appropriate. 8.2 UNCERTAINTY A discussion of CDF uncertainty, and conservatisms in the SAMA analysis that off-set uncertainty, should be included. For example, use of conservative risk modeling to represent a particular plant change may be used to offsl~t uncertainty in risk modeling; use of conservative implementation cost estimates may be used to offset uncertainty in cost estimates; and use of an uncertainty factor derived from the ratio of the 95th percentile to the mean point estimate for internal events CDF may be used to account for CDF uncertainties. Estimate an uncertainty factor based on this discussion and perform a sensitivity analysis using the uncertainty factor on the results. [Based on analysis to date the ratio of the 95th percentile to the mean point estimate for typical internal events CDF values is 2 to 5 (Reference 1).] Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis. If SAMAs appear cost-beneficial in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimates discussed in Section 7.2), and their impact on the results may be appropriate. 30 OAGI0000585 00039

8.3 PEER REVIEW FINDINGS OR OBSERVATIONS NEI 05-01 (Rev A) November 2005 If the model used for the SAMA analysis does not address significant findings or observations from the PSA peer review discussed in Section 3.3, sensitivity analyses may be performed to support discussion of the impact of the fmdings or observations on the SAMA analysis results. In these sensitivity analyses, modify the PSA model (or its results) to simulate incorporation of the finding or observation and perform the Phase II analysis with the revised severe accident risk results. Sufficient margin exists in the maximum benefit estimation that the Phase I screening should not have to be repeated in the sensitivity analysis. Discuss the finding or observation and how its effects were simulated in the PSA model. Provide pertinent results and discuss how th,ey affect the conclusions of the SAMA analysis. If SAMAs appear cost-beneficial in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimates discussed in Section 7.2), and their impact on the results may be appropriate. 8.4 EVACUATION SPEED Population dose may be significantly affected by radial evacuation speed, and uncertainties may be introduced during derivation of a single evacuation speed from emergency plan information, as discussed in Section 3.4.4. Therefore, perform sensitivity. analyses to show that variations in this parameter would not impact the results of the analysis. This sensitivity analysis should modify the evacuation speed assumed in the Level 3 PSA model and recalculate the baseline severe accident risk results. Multiple speeds may be evaluated as necessary. Discuss uncertainty in the evacuation speed and how the modified speed was selected. Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis. 8.S REAL DISCOUNT RATE Calculation of severe accident impacts also involves a real discount rate, r, which is typically assumed to be 7% (O.07/year) as recommended in NUREGIBR-OI84. A value of 7% is conservative because cost estimates are usually performed by utilities using values between 11 and 15%. Use of both a 7% and 3% real discount rate in regulatory analysis is specified in Office of Management Budget (OMB) guidance (Reference 5) and in NUREG/BR-0058 (Reference 6). The two discount rates represent the difference in whether a decision to undertake a project requiring investment is viewed as displacing either private investment or private consumption. A rate of 7% should be used as a baseline for regulatory analyses and represents an estimate of the average before-tax rate of return on an average investment in the private sector in recent years. A rate of3% should also be used and represents an estimate of the "consumption rate of interest," i.e., the real, after-tax rate of return on widely available savings instruments or investment opportunities. To address this concern, perform a sensitivity analysis using a 3% real discount rate. 31 OAGI0000585 00040

NEI 05-01 (Rev A) November 2005 In this sensitivity analysis, modify the real discount rate in the Level 3 PSA model and perform the Phase II analysis with the revised severe accident risk results. Sufficient margin exists in the maximum benefit estimation that the Phase I screening should not have to be repeated in the sensitivity analysis. Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis. If SAMAs appear cost-beneficial in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimates discussed in Section 7.2), and their impact on the results may be appropriate. 8.6 ANALYSIS PERIOD As described in Section 4, calculation of severe accident impacts involves an analysis period term, tf, which can be defined as either the period of extended operation (20 years), or the years remaining until the end of facility life (from the time of the SAMA analysis to the end of the period of extended operation) (25 years or more). The value that is typically used for this term is the period of extended operation (20 years). However, NRC has asked several plants to perform a sensitivity analysis using the period from the time of the SAMA analysis to the end of the period of extended operation to determine if SAMAs are potentially cost-beneficial if performed immediately. This sensitivity analysis should be performed to provide the information wanted by the regulator. In this sensitivity analysis, modify the analysis period in the calculation of severe accident risk and perform the Phase II analysis with the revised analysis period. The cost of additional years of maintenance, surveillance, calibrations, and training should be included in the cost estimates for SAMAs in this Phase II analysis. Sufficient margin exists in the maximum benefit estimation that the Phase I screening should not have to be repeated in the sensitivity analysis. Provide pertinent results and discuss how they affect the conclusions of the SAMA analysis. If SAMAs appear cost-beneficial in the sensitivity results, discussion of conservatisms in the analysis, (e.g., conservatisms in cost estimat,es discussed in Section 7.2), and their impact on the results may be appropriate. 32 OAGI0000585 00041

9 CONCLUSIONS NEI 05-01 (Rev A) November 2005 Discuss SAMAs that are cost-beneficial aftt:r the Phase II and sensitivity analyses. It may also be useful to discuss the combination of selected SAMAs and their impact on the overall plant risk. In some instances, addressing certain SAMAs may reduce the importance of the remaining candidates. This analysis may not estimate all of the benefits or all of the costs of a SAMA. For instance, it may not consider increases or decreases in: maintenance or operation costs following SAMA implementation. Also, it may not consider the possible adverse consequences of procedure changes, such as additional personnel dose. Since the SAMA analysis is not a complete engineering project cost-benefit analysis, the SAMAs that are cost-beneficial after the Phase II analysis and sensitivity analyses are only potentially cost-beneficial. 33 OAGI0000585 00042

NEI 05-01 (Rev A) November 2005 10 TABLES AND FIGURES TABLEt SAMPLE Accident Class Distribution Class Description TABLE 2 Frequency (per year) Percent of Total SAMPLE Release Severity and Timing Classification Scheme Release Severity Source Release Timing Term Release Fraction Classification Cesium Iodide % Classification Category Release Category Time of Release(l) Extreme (E) greater than 50 Late (L) greater than 6 hours Large (L) 10 to 50 Early (E) less than 6 hours Medium(M) 1 to 10 Small (S) less than 1 (1) Relative to declaration of a General Emergency_ 34 OAGI0000585 00043

TABLE 3 NEI 05-01 (Rev A) November 2005 SAMPLE Release Category Frequency and Release Fractions (Source Term) D Release Categoryn*2) h~-'E~-E~--~--7L-~E'---;;-~L'-'L--~I~~M~-E~~I;-~M~_7L---'---'S~-E~--~--~S'_L---1 in Frequency 2.64E-09 4.20E*06 7.19E-{)6 MAAP Run Case 1 Case 2 Case 3 me aner :Scram When General Emergency IS declared 25 min 30m," 2hrs ission Product Group:

1) Noble Total Release Fraction at 40 Hour.

9.9E-Ol 7.4E~)l 8.5E-Ol Start of Release (hr 0.25 0.8(' 9.00 End of Release (hr) 2.00 4.0(' 10.00 i2)Csl Total Release Fraction at 40 Hours 8.3E-Ol 4.6E~)l 2.8E-Ol Start of Release (hr 0.25 0.8e 9.00 End of Release (hr) 2.00 40.00 40.00 ~)Te02 Total Release Fraction at 40 Hours 6.8E-Ol 2.4E~)l 9.9E-02 Start of Release (hr 0.25 0.80 9.00 End of Release (hr) 2.00 12.00 22.00 4)SrO Total Release Fraction at 40 Hours 1.5E-02 4.7E~)3 2.0E-05 Start of Release (hr 0.25 0.80 9.00 End of Release (hr) 6.00 6.00 9.00 5)Mo02 Total Release Fraction at 40 Hours 2.4E-02 3.7E~)3 4.1E-07 Start of Release (hr) 0.25 0.80 9.00 End of Release (hr) 2.00 2.00 16.00 [e)CsOH Total Release Fraction at 40 Hours 6.9E-Ol 3.1E-Ol 1.9E-Ol Start of Release (hr) 0.25 0.80 9.00 End of Release (hr) 2.00 30.00 18.00

17) BaO Talal Release Fraction at 40 Hours 2.BE-02 6.1E~13 1.6E-05 Start of Release (hr) 0.25 0.80 9.00 End of Release (hr) 2.00 6.00 9.00

~) La203 Total Release Fraction at 40 Hours 6.5E-04 4.8E-04 5.6E-07 Start of Release (hr) 0.25 0.80 9.00 End of Release (hr) 6.00 6.00 9.00 9)Ce02 Total Release Fraction at 40 Hours 4.6E-03 2.0E-03 8.8E-08 Start of Release (hr) 4.00 3.00 9.00 End of Release (hr) 6.00 6.00 9.00 10)Sb Total Release Fraction at 40 Hours 5.9E-Ol 3.8E-COI 1.6E-Ol Start of Release (hr) 0.25 0.80 9.00 End of Release (hr) 2.00 40.00 40.00 11)Te2 Total Release Fraction at 40 Hours 2.3E-03 2.4E-C'2 1.2E-02 Start of Release (hr) 4.00 3.00 9.00 End of Release (hr) 6.00 40.00 20.00 12)U02 Total Release Fraction at 40 Hours 2.0E-05 1.1E-C5 1.BE-07 Start of Release (hr 4.00 3.00 9.00 End of Release (hr) 6.00 6.00 20.00 (I) Puff releases are denoted in the table by those entries with equivalent start and end times. (2) All cases run for 40 hrs 35 8.99E~8 1.09E-06 1.81E-07 3.97E-05 Case 4 Case 5 Case 6 Case 7 2hrs 18 hrs 1 hr 2hr 6.2E-Ol 1.0E+00 1.0E+00 1.0E+00 4.40 34.00 1.00 16.00 6.00 36.00 4.00 18.00 8.9E-02 2.7E-02 5.0E-03 2.7E-03 4.40 36.00 1.00 16.00 14.00 40.00 6.00 18.00 1.2E-Ol 7.5E-03 2.4E-03 9.6E-04 4.40 34.00 1.00 16.00 8.00 40.00 4.00 40.00 2.3E-02 7.4E-06 1.5E-04 5.2E-{)6 4.40 34.00 2.00 16.00 6.00 40.00 6.00 26.00 4.4E-06 6.1E-06 2.7E-04 8.4E-08 4.40 34.00 1.00 16.00 6.00 34.00 4.00 16.00 1.4E-Ol 5.7E-03 3.4E-03 8.7E-04 4.40 34.00 1.00 16.00 8.00 40.00 6.00 18.00 1.0E-02 6.4E-06 3.7E-04 2.8E-{)6 4.40 34.00 1.00 16.00 6.00 40.00 4.00 16.00 1.7E-03 1.3E-07 9.7E-{)6 8.9E-08 4.40 34.00 1.00 16.00 6.00 36.00 8.00 16.00 1.5E-02 3.8E-07 5.9E~5 9.4E-07 4.40 34.00 4.00 16.00 6.00 36.00 6.00 24.00 4.4E-OI 2.0E-04 3.2E-02 3.4E-03 4.40 34.00 1.00 16.00 40.00 36.00 14.00 40.00 2.4E-02 7.8E-06 3.3E-04 1.2E-03 4.40 36.00 5.00 16.00 40.00 40.00 8.00 40.00 7.7E-05 1.3E-IO 3.2E-07 8.0E-09 4.40 36.00 5.00 16.00 6.00 40.00 8.00 40.00 OAGI0000585 00044

NEI05-01 (Rev A) November 2005 TABLE 4 SAMPLE Estimated Population Distribution Within a 50-Mile Radius Sector 0-10 miles 10-20 20-30 30-40 40-50 miles miles miles miles N 1752 3211 6617 3250 1666 NNE 2029 1530 5073 9080 3560 NE 2357 10080 12428 4616 15346 ENE 7797 9726 9548 23262 23199 E 8436 25584 36954 30706 50569 ESE 6243 22217 224818 322317 372411 SE 9976 26461 188697 788711 785680 SSE 3114 12878 45896 179943 150702 S 5132 17275 17036 24134 12217 SSW 1995 6219 9689 8202 13624 SW 2432 5053 9951 11975 16255 WSW 1372 8140 3616 13662 6280 W 1879 4061 5821 6432 8220 WNW 1671 6540 14434 15309 7830 NW 739 10546 130402 9655 6890 NNW 4610 4129 4398 6235 10743 Total 61534 173650 725378 1457489 1485192 36 50-mile total 16496 21272 44827 73532 152249 948006 1799525 392533 75794 39729 45666 33070 26413 45784 158232 30115 3903243 OAGI0000585 00045

TABLE 5 SAMPLE MACCS2 Economic Parameters

• Variable
• Description
• DPRATE
• Property delueciation rate (per yr)
• DSRATE
• Investment rate of return (per yr)
• EVACST
• Daily cost for a person who has been evacuatled ($/person-day)
• POPCST
• Population relocation cost ($/person)
• RELCST
• Daily cost for a person who is relocatfd ($/person-day)
• CDFRMO
• Cost of farm decontamination for various levels of decontamination

($/hectare)

• CDNFRM
• Cost or non-farm decontamination per resident p,erson for various levels of decontamination ($/person)
• DLBCST
• Average cost of decontamination labor f($/person-year)
• VALWFO
• Value of farm wealth ($/hectare)
• VALWNF
• Value of non-farm wealth ($/person) 37 NEI 05-01 (Rev A)

November 2005 Value 0.2 0.12 43 7967 43 897 1992 4781 12754 55793 4547 126108 OAGI0000585 00046

NEI05-01 (Rev A) November 2005 TABLE 6 SAMPLE Core Inventory Values Core inventory Core inventory Nuclide (becquerels) Nuclide (becquerels) Cobalt-58 3.22E+16 Tellurium-131 M 4.67E+17 Cobalt-60 2.47E+16 Tellurium-132 4.66E+18 Krypton-85 2.47E+16 Iodine-131 3.20E+18 Krypton-85M 1.16E+18 Iodine-132 4.72E+18 Krypton-87 2.11E+18 Iodine-133 6.76E+18 Krypton-88 2.86E+18 Iodine-134 7.43E+18 Rubidium-86 1.88E+15 Iodine-135 6.38E+18 Strontium-89 3.58E+18 Xenon-133 6.78E+18 Strontium-90 1.94E+17 Xenon-135 1.27E+18 Strontium-91 4.62E+18 Cesium-134 4.32E+17 Strontium-92 4.80E+18 Cesium-136 1.31E+17 Yttrium-90 2.08E+17 Cesium-137 2.41E+17 Yttrium-91 4.36E+18 Barium-139 6.27E+18 Yttrium-92 4.81E+18 Barium-140 6.21E+18 Yttrium-93 5.45E+18 Lanthanum-140 6.34E+18 Zirconium-95 5.52E+18. Lanthanum-141 5.82E+18 Zirconium-97 5.76E+18 Lanthanum-142 5.61E+18 Niobium-95 5.21E+18 Cerium-141 5.65E+18 Molybdium-99 6.09E+18 Cerium-143 5.49E+18 Technetium-99M 5.25E+18 Cerium-144 3.40E+18 Ruthenium-1 03 4.54E+18 Praseodymium-143 5.38E+ 18 Ruthenium-1 05 2.94E+18 Neodymium-147 2.41E+18 Ruthenium-l 06 1.03E+18 Neptunium-239 6.46E+19 Rhodium-! 05 2.04E+18 Plutonium-238 3.66E+15 Antimony-127 2.79E+17 P1utonium-239 8.25E+14 Antimony-129 9.85E+17 Plutonium-240 1.04E+15 Tellurium-127 2.69E+17 Plutonium-241 1.75E+17 Tellurium-127M 3.55E+16 Americium-241 1.16E+14 Tellurium-129 9.26E+17 Curium-242 4.43E+16 Tellurium-129M 2.44E+17 Curium-244 2.59E+15 38 OAGI0000585 00047

TABLE 7 SAMPLE Release Characteristics Parameter Early-Early-Bypass Rupture Leaks Heat of Release 2.1E+06 1.8E+06 1.0E+06 (W) Height of 30 30 30 Release (m) TABLES SAMPLE Summary of Severe Accident Risk Results Off-Site Dose Economic Release (person-Impact Category rem/year) ($/year) E-E 1.39E-02 6.05E+Ol L-E 1.73E+Ol 1.31E+05 L-L 1.58E+Ol . 1. 17E+05 M-E 2.57E-Ol

1. 79E+03 M-L 4.43E-Ol 4.63E+02 S-E 7.00E-02 5.85E+Ol S-L

4. 13E+OO 3.24E+03 None O.OE+OO O.OE+OO (intact)

Totals 3.80E+Ol 2.54E+05 39 NEI 05-01 (Rev A) November 2005 Late 9.2E+05 30 OAGI0000585 00048

o >> G> o o o o (J'1 00 1(J'1 o o o .j>.. <D NEI 05-01 (Rev A) November 2005 Risk Significant Terms LINER-MELT HPCI ECCS Low Pressure Interlock Depressurization (SRVs and ADS Logic) Loss of feedwater - Initiating event Operator Action: Operator fails to open SRVs for vessel depressurization during transients and small LOCA EXV-STM-EX RRW 9.362 1.4966 1.3472 1.2724 1.1794 1.1109 1.009 TABLE 9 SAMPLE PSA Importance Review Disposition This term represents the probability of sufficient corium leaving the vessel to melt the containment liner. Phase II SAMAs 004 and 009 to increase injection systems and provide a dedicated drywell spray system were examined to reduce the risk of containment liner melt This term represents random failure of the HPCI system. Phase I SAMAs to improve availability and reliability of the HPCI system include raising backpressure trip set points and proceduralizing intermittent operation. Additional improvements were evaluated in Phase II SAMAs 049, 050, 051, 052, and 053. This term represents random failures of the reactor low-pressure transmitters during transients with stuck open SRVs or LOCAs in which random failures prevent all low-pressure injection valves from opening. Phase II SAMAs 065 and 066 were examined to reduce the risk due to the failure of the ECCS low-pressure interlock. This term represents random failures of the SRVs to open on demand to depressurize during transients and small LOCAs. Phase I SAMAs to enhance reliability of the SRVs include adopting symptom based EOPs and SAGs, modifying.AnS logic~ and upgranl'lg S!tV pnen..1!'.atic components...A..dditicna! inlprovements \\vere exa..T.ined hI} Phase II SAMAs 059 and 060. This term represents the initiating event for loss offeedwater. Modifications to significantly reduce or eliminate the potential for loss of feedwater have already been implemented, such as installing a digital feedwater control system, providing a backup water supply, and adding a third feedwater pump. Many of the Phase II SAMAs (e.g. 035,051,052,053, and 054) explored potential benefits for mitigation of this event. No additional SAMAs were recommended for this broad subject. This term represents the operator failing to manually open the SRVs to depressurize during transients and small LOCAs. Improvement of plant procedures and instrumentation to enhance the likelihood of success of operator action in response to accident conditions were examined in Phase I SAMAs during preliminary screening. No additional SAMAs were recommended for this subject. This term represents a steam explosion which fails containment. Phase II SAMAs 014 and 006 to strengthen the drywell and add a diverse injection system were examined to reduce the risk of a steam explosion in containment. 40

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 o PHASE I SAMA ID NUMBE R 1 2 3 4 SAMA TITLE Provide an additional diesel generator Add additional battery charger or portable, diesel-driven battery charger to existing DC system. Provide a portable generator to support SRVs and hard pipe vent Contingency plans during switchyard work TABLE 10 SAMPLE List of Phase I SAMA Candidates SAMA DISCUSSION SOURCE PHASE 1 DISPOSITION This SAMA would help mitigate LOOP events Levell The cost of installing an additional and would reduce the risk of on-line EDG Importance List EDG has been estimated to be greater maintenance. Benefit would be increased if and standard than $20 million in the Calvert Cliffs the additional diesel generator could 1) be listofBWR Application for License Renewal. As substituted for any current diesel that is in SAMA this is greater than the Maximum maintenance, and 2) if the diesel was ofa candidates Benefit, it has been screened from diverse design such that common cause failure further analysis. dependence was minimized. Improved availability of DC power system. Levell and 2 Retain for Phase II analysis. Importance Lists and standard list of BWRSAMA candidates Improved availability of DC power system:. Levell and 2 Retain for Phase II analysis. Importance Lists Assessing likely failures of the off-site AC Levell Retain for Phase II analysis. power supply due to switchyard work and Importance List providing plans for power restoration in the event that such a loss occurs could reduce the time required to recover off-site power. 41 NEI 05-01 (Rev A) November 2005 RETAINED FOR PHASE n ANALYSIS? No Yes Yes Yes

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 NEI 05-01 (Rev A) November 2005 PHASE SAMA IISAMA TITLE ID NUMBE R 010 Use the fire water system as a backup source for the containment spray system 011 Make containment sump recirculation outlet valve motor-operated valves diverse from one another SAMA DISCUSSION Improved containment spray capability. TABLE 11 SAMPLE Phase II SAMA List UPPER BOUND ESTIMATED ESTIMATED COST OF BENEFIT IMPLEMENTA TI ON $178,000 $1,500,000 Replace one of the two containment $520,440 $424,783 sump valves with an air-operated valve. This would reduce the potential for common cause failure of these valves. 42 CONCLUSION BASIS FOR CONCLUSION Not Cost-Beneficial Elimination of all off-site releases results in a benefit of $178,000 (analysis case OFFSITE). In 1993, the cost of implementing a similar SAMA in the Westinghouse-CE System 80+ was estimated to be $i,500,OOO. Since the cost of implementing this SAMA exceeds the attainable benefit, this SAMA is not cost-beneficial. Potentially Cost-Elimination of all core damage due Beneficial to containment sump valve failures results in a benefit of $520,440 (analysis case SUMPMOV). The cost of implementing this SAMA is judged to be $424,783. Therefore, this SAMA is potentially cost-beneficial.

o >> G> o o o o 0'1 00 10'1 o o o 0'1 ~ Phase II SAMAID I 2 3 4 5 6 7 8 9 10 II 12 13 14 15 TABLE 12 SAMPLE Sensitivity Analysis Results Upper Bound Upper Bound Estimate Benefit Estimate Benefit SAMA Title Base line Estimated Cost Sensitivity Case 1 Add a service water pump. $120,000 $5,900,000 S140,000 Provide a redundant train or means of EOG $470,000 $1,000,000 $550,000 room ventilation. Add a diesel building high temperature alarm $160,000 $2500,000 $180,000 or redundant louver and thermostat. Install an independent method of suppression $530,000 $5,800,000 $620,000 pool cooling. Install a filtered containment vent to remove $0 $3,000,000 $0 decay heat. Install an A TWS sized filtered containment $0 >$2,00,000 SO vent to remove decay heat. Create a large concrete crucible with heat S640,000 >$100 million $720,000 removal potential to contain molten core debris Provide a reactor vessel exterior cooling $640,000 SI9,000,000 $720,000 system. Enable flooding of the drywell head seal. $20,000 , >SI,OOO,OOO $20,000 Enhance fire protection system and standby $1,410,000, >2,500,000 $1,610,000 gas treatment system hardware and Iprocedures Create a core melt source reduction system $640,000 >$ 1,000,000 $720,000 Install a passive drywell spray system S530,000 $5,800,000 $620,000 Strengthen primary/secondary containment $530,000 SI2,000,000 $620,000 e.g., add ribbing to containment shell}. Increase depth of the concrete base mat or $640,000 >$ I,000,000 $720,000 use an alternative concrete material to ensure melt-through does not occur. Provide a reactor vessel exterior cooling $640,000 $2,500,000 $720,000 system. 43 Upper Bound Estimate Benefit Sensitivity Case 2 $160,000 $640,000 $220,000 $720,000 $0 SO $890,000 $890,000 $30,000 $1,980,000 $890,000 $720,000 $720,000 $890,000 $890,000 NEI 05-01 (Rev A) November 2005

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 w NEI 05-01 (Rev A) November 2005 SAMAID 001 002 003 004 005 006 007 008 009 010 011 012 013 014 015 TABLE 13 STANDARD List of BWR SAMA Candidates Potential Enhancement (SAMA Title) Result of Potential Enhancement Improvements Related to AC and DC Power Provide additional DC battery capacity. Extended DC power availability during an SBO. Replace lead-acid batteries with fuel cells. Extended DC power availability during an SBO. Add additional battery charger or portable, Improved availability of DC power system. diesel-driven battery charger to existing DC system. Improve DC bus load shedding. Extended DC power availability during an SBO. Provide DC bus cross-ties. Improved availability of DC power system. Provide additional DC power to the I20/240V Increased availability of the 120 V vital AC bus. vital AC system. Add an automatic feature to transfer the 120V Increased availability of the 120 V vital AC bus. vital AC bus from normal to standby power. Increase training on response to loss of two Improved chances of successful response to loss of two 120V AC buses. l20V AC buses which causes inadvertent actuation signals. Reduce DC dependence between high-pressure Improved containment depressurization and high-pressure injection following DC failure. injection system and ADS. Provide an additional diesel generator. Increased availability of on-site emergency AC power. Revise procedure to allow bypass of diesel Extended diesel generator operation. generator trips. Improve 4.I6-kV bus cross-tie ability. Increased availability of on-site AC power. Create AC power cross-tie capability with other Increased availability of on-site AC power. unit (multi-unit site). Install an additional, buried off-site power Reduced probability of loss of off-site power. source. Install a gas turbine generator. Increased availability ofon-siteAC power. 44 Source Reference 1,3,6,10,11, 12,17 6,10 5 1,7 6 3 5 5 1 1,6,10,11,12 15 1,6,11,12 1,7,13 1 1,6

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 .j>.. SAMAID 016 017 018 019 020 021 022 023 024 025 026 027 028 029 030 031 Potential Enhancement (SAMA Title) Result of Potential Enhancement Install tornado protection on gas turbine Increased availability of on-site AC power. generator. Install a steam-driven turbine generator that Increased availability of on-site AC power. uses reactor steam and exhausts to the suppression pool. Improve uninterruptible power supplies. Increased availability of power supplies supporting front-line equipment. Create a cross-tie for diesel fuel oil (multi-unit Increased diesel generator availability. site). Develop procedures for replenishing diesel fuel Increased diesel generator availability. oil. Use fire water system as a backup source for Increased diesel generator availability. diesel cooling. Add a new backup source of diesel cooling. Increased diesel generator availability. Dc;vdup prucc;uurc;s io repair or replace failed 4 Increased probability of recovery from failure of breakers that transfer 4.16 kV non-KV breakers. emergency buses from unit station service transformers. In training, emphasize steps in recovery of off-Reduced human error probability during off-site power recovery. site power after an SBO. Develop a severe weather conditions procedure. Improved off-site power recovery following external weather-related events. Bury off-site power lines. Improved off-site power reliability during severe weather. Improvements Related to Core Cooling Systems Install an independent active or passive high Improved prevention of core melt sequences. pressure injection system. Provide an additional high pressure injection Reduced frequency of core melt from small LOCA and SBO sequences. pump with independent diesel. Raise HPCIIRCIC backpressure trip set points. Increased HPCI and RCIC availability when high suppression pool temperature exists. Revise procedure to allow bypass ofRCIC Extended RCIC operation. turbine exhaust pressure trip. Revise procedure to allow intermittent Extended HPCI and RCIC operation. operation of HPCI and RCIC. 45 NEI 05-01 (Rev A) November 2005 Source Reference 18 6 6 1 1 1 I 1 1 1,3,17 1 5,6 5 15 15 1

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 (J'1 NEI 05-01 (Rev A) November 2005 SAMAID 032 033 034 035 036 037 038 039 040 041 042 043 044 045 046 047 048 Potential Enhancement (SAMA Title) Revise procedure to control torus temperature, torus level, and primary containment pressure to increase available net positive suction head (NPSH) for injection pumps. Revise procedure to manually initiate HPCI and RCIC given auto initiation failure. Modify automatic depressurization system components to improve reliability. Add signals to open safety relief valves automatically in an MSIV closure transient. Revise procedure to allow manual initiation of emergency depressurization. Revise procedure to allow operators to inhibit automatic vessel depressurization in non-ATWS scenarios. Add a diverse low pressure injection system. Increase flow rate of suppression pool cooling. Provide capability for alternate injection via diesel-driven fire pump. Provide capability for alternate injection via reactor water cleanup (RWCU). Revise procedure to align EDG and allow use of essential CRD for vessel injection. Revise procedure to allow use of condensate pumps for injection. Revise procedure to allow use of suppression pool jockey pump for injection. Revise procedure to re-open MSIVs. Improve ECCS suction strainers. Revise procedure to align LPCI or core spray to CST on loss of suppression pool cooling. Remove LPCI loop select logic. Source Result of Potential Enhancement Reference Increased probability that injection pumps will be available to inject coolant into the vessel. I Increased availability of HPCI and RCIC given auto initiation signal failure. 1 Reduced frequency of high pressure core damage sequences. 3,21 Reduced likelihood of SRV failure to open in an MSIV closure transient reduces the 3 probability of a medium LOCA. Improved prevention of core damage during transients, small and medium LOCAs, and 21 ATWS. Extended HPCI and RCIC operation. 5 Improved injection capability. 5,6 Improved suppression pool cooling. 6 Improved injection capability. 5 Improved injection capability. 1 Improved injection capability. 15 Improved injection capability. 15 Improved injection capability. 6 Regains the main condenser as a heat sink. 15 Enhanced reliability ofECCS suction. 22 Improved injection in loss of suppression pool cooling scenarios. 15 Enables use of LPCI A loop for injection in the event of a B injection path failure. 18 46

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 (j) SAMAID 049 050 051 052 053 054 055 056 057 058 059 060 061 062 063 Potential Enhancement (SAMA Title) Result of Potential Enhancement Replace two of the four electric safety injection Reduced common cause failure of the safety injection system. This SAMA was originally pumps with diesel-powered pumps. intended for the Westinghouse-CE System 80+, which has four trains of safety injection. However, the intent ofthis SAMA is to provide diversity within the high-and low-pressure safety injection systems. Improvements Related to Cooling Water Change procedures to allow cross connection of Continued operation of both RHRSW pumps on failure of one train of SW. motor cooling for RHRSW pumps. Add redundant DC control power for SW Increased availability of SW. pumps. Replace ECCS pump motors with air-cooled Elimination of ECCS dependency on component cooling system. motors. Provide self-cooled ECCS seals. Elimination ofECCS dependency on component cooling system. Enhance procedural guidance for use of cross-Reduced frequency of loss of component cooling water and service water. tied component cooling or service water pumps. Implement modifications to allow manual Improved ability to cool RHR heat exchangers. alignment of the fire water system to RHR heat exchangers. Add a service water pump. Increased availability of cooling water. Enhance the screen wash system. Reduced potential for loss of SW due to clogging of screens. Improvements Related to Feedwater and Condensate Install a digital feedwater upgrade. Reduced chance of loss of main feedwater following a plant trip. Create ability for emergency connection of Increased availability of feedwater. existing or new water sources to feedwater and condensate systems. Install an independent diesel for the condensate Extended inventory in CST during an SBO. storage tank makeup pumps. Add a motor-driven feedwater pump. Increased availability of feedwater. Improvements Related to Heating, Ventilation, and Air Conditioning Provide reliable power to control building fans. Increased availability of control room ventilation. Provide a redundant train or means of Increased availability of components dependent on room cooling. ventilation. 47 NEI 05-01 (Rev A) November 2005 Source Reference 5,10 3 3 I I 1 1 6 23 1 5 5 1,3 2 1

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 NEI 05-01 (Rev A) November 2005 SAMAID 064 065 066 067 068 069 070 071 072 073 074 075 076 077 078 079 Potential Enhancement (SAMA Title) . Result of Potential Enhancement Enhance procedures for actions on loss of Increased availability of components dependent on room cooling. HVAC. Add a diesel building high temperature alarm Improved diagnosis of a loss of diesel building HV AC. or redundant louver and thermostat. Create ability to switch HPCI and RCIC room Increased availability ofHPCI and RCIC in an SBO event. fan power supply to DC in an SBO event. Enhance procedure to trip unneeded RHR or Extended availability of required RHR or CS pumps due to reduction in room heat load. CS pumps on loss of room ventilation. Stage backup fans in switchgear rooms. Increased availability of ventilation in the event of a loss of switchgear ventilation. Add a switchgear room high temperature alarm. Improved diagnosis of a loss of switchgear HV AC. Improvements Related to Instrument Air and Nitrogen Supply Provide cross-unit connection of Increased ability to vent containment using the hardened vent. uninterruptible compressed air supply. Modify procedure to provide ability to align Increased availability of instrument air after a LOOP. diesel power to more air compressors. Replace service and instrument air compressors Elimination of instrument air system dependence on TBCCW and service water cooling. with more reliable compressors which have self-contained air cooling by shaft driven fans. Install nitrogen bottles as backup gas supply for Extended SRV operation time. safety relief valves. Improve SRV and MSIV pneumatic Improved availability of SRVs and MSIVs. components. Improvements Related to Containment Phenomena Install an independent method of suppression Increased availability of containment heat removal. pool cooling. Revise procedure to initiate suppression pool Improved containment pressure control and containment heat removal capability. cooling during transients, LOCAs and ATWS. Cross-tie open cycle cooling system to enhance Increased availability of containment heat removal. drywell spray system. Enable flooding of the drywell head seal. Reduced probability of leakage through the drywell head seal. Create a reactor cavity flooding system. Enhanced debris cool ability, reduced core concrete interaction, and increased fission product scrubbing. 48 Source Reference 3 1 1 3 5 5 3 18 5 18 6 6,8,9 6,8,9 8,9 6,8,9 1,7, 11, 12

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 00 SAMAID 080 081 082 083 084 085 086 087 088 089 090 091 092 093 094 Potential Enhancement (SAMA Title) Install a passive drywell spray system. Use the fire water system as a backup source for the drywell spray system. Enhance procedures to refill CST from demineralized water or service water system. Enhance procedure to maintain ECCS suction on CST as long as possible. ModifY containment flooding procedure to restrict flooding to below the top of active fuel. Install an unfiltered, hardened containment vent. Install a filtered containment vent to remove decay heat. OptiuII 1: Gravd Bed Filier Option 2: Multiple Venturi Scrubber Enhance fire protection system and standby gas treatment system hardware and procedures. ModifY plant to permit suppression pool scrubbing. Enhance containment venting procedures with respect to timing, path selection, and technique. Control containment venting within a narrow band of pressure. Improve vacuum breaker reliability by installing redundant valves in each line. Enhance air return fans (ice condenser plants). Provide post-accident containment inerting capability. Create a large concrete crucible with heat removal potential to contain molten core debris. Result of Potential Enhancement Improved drywell spray capability. Improved drywell spray capability. Reduced risk of core damage during station blackouts or LOCAs that render the suppression pool unavailable as ail injection source. Reduced chance of pump failure due to high suppression pool temperature. Reduced forced containment venting. Increased decay heat removal capability for non-ATWS events, without scrubbing released fission products. Increased decay heat removal capability for non-ATWS events, with scrubbing of released fission products. Improved fission product scrubbing in severe accidents. Increased scrubbing of fission products by directing vent path through water in the suppression pool. Improved likelihood of successful venting. Reduced probability of rapid containment depressurization thus avoiding adverse impact on low pressure injection systems that take suction from the torus. Decreased consequences of a vacuum breaker failure to reseat. Reduced probability of containment failure in sao sequences. Reduced likelihood of hydrogen and carbon monoxide gas combustion. Increased cooling and containment of molten core debris. Molten core debris escaping from the vessel is contained within the crucible and a water cooling mechanism cools the molten core in the crucible, preventing melt-through ofthe base mat. 49 NEI 05-01 (Rev A) November 2005 Source Reference 6, 14 4,6 15 15 16 6,8,9 6,8,9, 14 9 6 16 18 6 1 6,7,12 6,8,9

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 <D NEI 05-01 (Rev A) November 2005 SAMAID 095 096 097 098 099 iOO 101 102 103 104 105 106 Potential Enhancement (SAMA Title) Result of Potential Enhancement Create a core melt source reduction system. Increased cooling and containment of molten core debris. Refractory material would be placed underneath the reactor vessel such that a molten core falling on the material would melt and combine with the material. Subsequent spreading and heat removal from the vitrified compound would be facilitated, and concrete attack would not occur. Strengthen primary/secondary containment Reduced probability of containment over-pressurization. (e.g., add ribbing to containment shell). Increase depth of the concrete base mat or use Reduced probability of base mat melt-through. an alternate concrete material to ensure melt-through does not occur. Provide a reactor vessel exterior cooling Increased potential to cool a molten core before it causes vessel failure, by submerging the system. lower head in water. Construct a building to be connected to Reduced probability of containment over-pressurization. primary/secondary containment and maintained at a vacuum. Institute simulator training tor severe accident Improved arrest of core. melt progress and prevention of containment failure. scenarios. Improve leak detection procedures. Increased piping surveillance to identify leaks prior to complete failure. Improved leak detection would reduce LOCA frequency. Install an independent power supply to the Reduced hydrogen detonation potential. hydrogen control system using either new batteries, a non-safety grade portable generator, existing station batteries, or existing AClDe independent power supplies, such as the security system diesel. Install a passive hydrogen control system. Reduced hydrogen detonatiori.'potential. Erect a barrier that would provide enhanced Reduced probability of containment failure. protection of the containment walls (shell) from ejected core debris following a core melt scenario at high pressure. Improvements Related to Containment Bypass Install additional pressure or leak monitoring Reduced ISLOCA frequency. instruments for detection ofISLOCAs. Add redundant and diverse limit switches to Reduced frequency of containment isolation failure and ISLOCAs. each containment isolation valve. 50 Source Reference 13 5,6, 10, 14 10 10 6,10 6 6 5,10 5,10 5 4,7,11,12,15 1

o >> G> o o o o (J'1 00 1(J'1 o o o (j) o SAMAID 107 108 109 110 III 112 113 114 115 116 117 118 119 120 121 Potential Enhancement (SAMA Title) Increase leak testing of valves in ISLOCA paths. Improve MSIV design. Install self-actuating containment isolation valves. Locate residual heat removal (RHR) inside containment Ensure ISLOCA releases are scrubbed. One method is to plug drains in potential break areas so that break point will be covered with water. Revise EOPs to improve ISLOCA identification. Improve operator training on ISLOCA coping. Create cross-connect ability for standby liquid control (SLC) trains. Revise procedures to control vessel injection to prevent boron loss or dilution following SLC injection. Provide an alternate means of opening a pathway to the RPV for SLC injection. Increase boron concentration in the SLC system. Add an independent boron injection system. Provide ability to use control rod drive (CRD) or RWCU for alternate boron injection. Add a system of relief valves to prevent equipment damage from pressure spikes during anATWS. Increase safety relief valve (SRV) reseat reliability. Result of Potential Enhancement Reduced ISLOCA frequency. Decreased likelihood of containment bypass scenarios. Reduced frequency of isolation failure. Reduced frequency of ISLOCA outside containment. Scrubbed ISLOCA releases. Increased likelihood that LOCAs outside containment are identified as such. A plant had a scenario in which an RHR ISLOCA could direct initial leakage back to the pressurizer relief tank, giving indication that the LOCA was inside containment. Decreased ISLOCA consequences. Improvements Related to ATWS Improved availability of boron injection during ATWS. Improved availability of boron injection during ATWS. Improved probability of reactor shutdown. Reduced time required to achieve shutdown concentration provides increased margin in the accident timeline for successful initiation of SLC. Improved availability of boron injection during ATWS. Improved availability of boron injection during ATWS. Improved equipment availability after an ATWS. Reduced risk of dilution of boron due to SRV failure to reseat after standby liquid control (SLC) injection. 51 NEI 05-01 (Rev A) November 2005 Source Reference 1 6 5 14 1 1 1 18 15 18 18 18 1 19 1

o >> G> o o o o (J'1 00 1(J'1 o o o (j) NEI 05-01 (Rev A) November 2005 SAMAID 122 123 124 125 126 127 128 129 130 131 132 133 134 135 Potential Enhancement (SAMA Title) Result of Potential Enhancement Provide an additional control system for rod Improved redundancy and reduced ATWS frequency. insertion (e.g., AMSAC). Install an A TWS sized filtered containment Increased ability to remove reactor heat from A TWS events. vent to remove decay heat. Revise procedure to bypass MSIV isolation in Affords operators more time to perform actions. Discharge of a substantial fraction of steam turbine trip A TWS scenarios. to the main condenser (i.e., as opposed to into the primary containment) affords the operator more time to perform actions (e.g., SLC injection, lower water level, depressurize RPV) than if the main condenser was unavailable, resulting in lower human error probabilities. Revise procedure to allow override of low Allows immediate control oflow pressure core injection. On failure of high pressure core pressure core injection during an ATWS event. injection and condensate, some plants direct reactor depressurization followed by five minutes of automatic low pressure core injection. Improvements Related to Internal Flooding Seal penetrations between turbine building Increased flood propagation prevention. basement and switchgear rooms. Improve inspection of rubber expansion joints Reduced frequency of internal flooding due to failure of circulating water system expansion on main condenser. joints. Modify swing direction of doors separating Prevents flood propagation. turbine building basement from areas containing safeguards equipment. Improvements to Reduce Seismic Risk Increase seismic ruggedness of plant Increased availability of necessary plant equipment during and after seismic events. components. Provide additional restraints for CO2 tanks. Increased availability of fire protection given a seismic event. Modify safety related condensate storage tank. Improved availability of CST following a seismic event. Replace anchor bolts on diesel generator oil Improved availability of diesel generators following a seismic event. cooler. Improvements to Reduce Fire Risk Replace mercury switches in fire protection Decreased probability of spurious fire suppression system actuation. system. Upgrade fire compartment barriers. Decreased consequences of a fire. Install additional transfer and isolation Reduced number of spurious actuations during a fire. switches. 52 Source Reference 18 6 1,20 16 1 1 5 3,10 17 6 1 7 7 18

o >> G> o o o o (J'1 00 1(J'1 o o o (j) ~ SAMAID 136 137 138 139 140 141 142 143 144 Potential Enhancement (SAMA Title) Enhance procedures to use alternate shutdown methods if the control room becomes uninhabitable. Enhance fire brigade awareness. Enhance control of combustibles and ignition sources. Install digital large break LOeA protection system. Enhance procedures to mitigate large break LOeA. Install computer aided instrumentation system to assist the operator in assessing post-accident plant status. Improve maintenance procedures. Increase training and operating experience feedback to improve operator response. Develop procedures for transportation and nearby facility accidents. Result of Potential Enhancement Increased probability of shutdown if the control room becomes uninhabitable. Decreased consequences of a fire. Decreased fire frequency and consequences. Other Improvements Reduced probability of a large break LOeA (a leak before break). Reduced consequences of a large break LOeA Improved prevention of core melt sequences by making operator actions more reliable. Improved prevention of core melt sequences by increasing reliability of important equipment. Improved likelihood of success of operator actions taken in response to abnormal conditions. Reduced consequences of transportation and nearby facility accidents. 53 NEI 05-01 (Rev A) November 2005 Source Reference 6,7 7 7 5 7 6 6 6 7

o >> G> o o o o (J'1 00 1(J'1 o o o (j) w NEI 05-01 (Rev A) November 2005 Table 13 References

1.

NUREG-1560, "Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance," Volume 2, U.S. Nuclear Regulatory Commission, December 1997.

2.

Letter from Mr. M. O. Medford (TVA) to NRC Document Control Desk dated September 1, 1992. "Watts Bar Nuclear Plant (WBN) Units 1 and 2 - Generic Letter (GL) 88 Individual Plant Examination (IPE) for Severe Accident Vulnerabilities - Response - (TAC M74488)."

3.

Appendix D-Attachment F, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Edwin I. Hatch Nuclear Power Plant Units 1 and 2, March 2000.

4.

Letter from Mr. D. E. Nunn (TVA) to NRC Document Control Desk, dated October 7, 1994. "Watts Bar Nuclear Plant (WBN) Units 1 and 2 - Severe Accident Mitigation Design Alternatives (SAMDA) - Response to Request for Additional Information (RAI) - (TAC Nos. M77222 and M77223)."

5.

NUREG-1437, "Generic Environmental Impact Statement for License Renewai of Nuclear Piants," Calvert Cliffs Nuclear Power Plant", Supplement 1, U.S. Nuclear Regulatory Commission, February 1999.*

6.

General Electric Nuclear Energy, Technical Support Document for the ABWR, 25A5680, Revision 1, January 18, 1995.

7.

NUREG-0498, "Final Environmental Statement related to the operation of Watts Bar Nuclear Plant, Units 1 and 2," Supplement No.1, U.S. Nuclear Regulatory Commission, April 1995.

8.

Cost Estimate for Severe Accident Mitigation Design Alternatives, Limerick Generating Station for Philadelphia Electric Company, Bechtel Power Corporation, June 22, 1989.

9.

NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Volume 1, 5.35, Listing of SAMDAs considered for the Limerick Generating Station, U.S. Nuclear Regulatory Commission, May 1996.

10.

NUREG-1462, "Final Safety Evaluation Report Related to the Certification of the System 80+ Design," U.S. Nuclear Regulatory Commission, August 1994.

11.

NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Volume 1, 5.36, Listing of SAMDAs considered for the Comanche Peak Steam Electric Station, U.S. Nuclear Regulatory Commission, May 1996. 54

o >> G> o o o o (J'1 00 1(J'1 o o o (j) .j>.. NEI 05-01 (Rev A) November 2005

12.

Letter from Mr. W. 1. Museler (Tennessee Valley Authority) to the NRC Document Control Desk, dated October 7, 1994, "Watts Bar Nuclear Plant (WBN) Units 1 and 2 - Severe Accident Mitigation Design Alternatives (SAMDAs)."

13.

Letter from Mr. D. E. Nunn (Tennessee Valley Authority) to NRC Document Control Desk, dated June 30, 1994. "Watts Bar Nuclear Plant (WBN) Unit I and 2 - Severe Accident Mitigation Design Alternatives (SAMDAs) Evaluation from Updated Individual Plant Evaluation (IPE)."

14.

Letter from N. J. Liparulo (Westinghouse Electric Corporation) to NRC Document Control Desk, dated December 15, 1992, "Submittal of Material Pertinent to the AP600 Design Certification Review."

15.

NUREG/CR-5474, "Assessment of Candidate Accident Management Strategies", U.S. Nuclear Regulatory Commission, March 1990.

16.

Severe Accident Applicability of BWROG Revision 4, "Emergency Procedure Guidelines", BWROG, September 1988.

17.

Appendix E-Environmental Report, Appendix G, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Peach Bottom Nuclear Power Plant Units 2 and 3, July, 2001.

18.

Appendix F, Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Quad Cities Nuclear Power Plant Units 1 and 2, January 2003.

19.

NEDC-33090P, Rev.O, "Safety Analysis Report for Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate", September 2003.

20.

BWROG EPC Issue 98-07.

21.

Individual Plant Examination for Severe Accident Vulnerabilities - Generic Letter 88-20, U.S. Nuclear Regulatory Commission, November 23, 1988.

22.

NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors", May 1996 U.S. Nuclear Regulatory Commission.

23.

Duke Power Company, Applicant's Environmental Report, Operating Licensing Renewal Stage. Attachment K, "Oconee Nuclear Station Severe Accident Mitigation Alternatives (SAMA) Analysis." Rev. O. Charlotte, North Carolina, June 1998. 55

o >> G> o o o o (J'1 00 1(J'1 o o o (j) (J'1 NEI 05-01 (Rev A) November 2005 SAMAID 001 002 003 004 005 006 007 008 009 010 011 012 013 014 015 016 TABLE 14 STANDARD List of PWRSAMA Candidates Potential Enhancement (SAMA Title) Result of Potential Enhancement Improvements Related to AC and DC Power Provide additional DC battery capacity. Extended DC power availability during an SBO. Replace lead-acid batteries with fuel cells. Extended DC power availability during an SBO. Add additional battery charger or portable, Improved availability of DC power system. diesel-driven battery charger to existing DC system. Improve DC bus load shedding. Extended DC power availability during an SBO. Provide DC bus cross-ties. Improved availability of DC power system. Provide additional DC power to the 1201240V Increased availability of the 120 V vital AC bus. vital AC system. Add an automatic feature to transfer the 120V Increased availability of the 120 V vital AC bus. vital AC bus from normal to standby power. Increase training on response to loss of two Improved chances of successful response to loss of two 120V AC buses. l20V AC buses which causes inadvertent actuation signals. Provide an additional diesel generator. Increased availability of on-site emergency AC power. Revise procedure to allow bypass of diesel Extended diesel generator operation. generator trips. Improve 4.16-kV bus cross-tie ability. Increased availability of on-site AC power. Create AC power cross-tie capability with other Increased availability of on-site AC power. unit (multi-unit site) Install an additional, buried off-site power Reduced probability of loss of off-site power. source. Install a gas turbine generator. Increased availability of on-site AC power. Install tornado protection on gas turbine Increased availability of on-site AC power. generator. Improve uninterruptible power supplies. Increased availability of power supplies supporting front-line equipment. 56 Source Reference I, 3, 6, 10, 11, 12,17 6,10 5 1,7 6 3 5 5 1,6, 10, 11, 12 15 1,6,11,12 1,7,13 1 1,6 18 6

o >> G> o o o o (J'1 00 1(J'1 o o o (j) (j) SAMAID 017 018 019 020 021 022 023 024 025 026 027 028 029 030 031 032 033 Potential Enhancement (SAMA Title) Result of Potential Enhancement Create a cross-tie for diesel fuel oil (multi-unit Increased diesel generator availability. site). Develop procedures for replenishing diesel fuel Increased diesel generator availability. oil. Use fire water system as a backup source for Increased diesel generator availability. diesel cooling. Add a new backup source of diesel cooling. Increased diesel generator availability. Develop procedures to repair or replace failed 4 Increased probability of recovery from failure of breakers that transfer 4.16 kV non-KV breakers. emergency buses from unit station service transformers. In training, emphasize steps in recovery of off-Reduced human error probability during off-site power recovery. site power after an SBO. Develop a severe weather conditions procedure. Improved off-site power recovery following external weather-related events. Bury off-site power lines. Improved off-site power reliability during severe weather. Improvements Related to Core Cooling Systems Install an independent active or passive high Improved prevention of core melt sequences. pressure injection system. Provide an additional high pressure injection Reduced frequency of core melt from small LOCA and SBO sequences. pump with independent diesel. Revise procedure to allow operators to inhibit Extended HPCI and RCIC operation. automatic vessel depressurization in non-A TWS scenarios. Add a diverse low pressure injection system. Improved injection capability. Provide capability for alternate injection via Improved injection capability. diesel-driven fire pump. Improve ECCS suction strainers. Enhanced reliability ofECCS suction. Add the ability to manually align emergency Enhanced reliability ofECCS suction. core cooling system recirculation. Add the ability to automatically align Enhanced reliability ofECCS suction. emergency core cooling system to recirculation mode upon refueling water storage tank depletion. Provide hardware and procedure to refill the Extended reactor water storage tank capacity in the event of a steam generator tube rupture. reactor water storage tank once it reaches a specified low level. 57 NEI 05-01 (Rev A) November 2005 Source Reference I I 1 I I I 1,3, 17 I 5,6 5 5 5,6 5 22 5 5 5,10

o >> G> o o o o (J'1 00 1(J'1 o o o (j) NEI 05-01 (Rev A) November 2005 SAMAID 034 035 036 037 038 039 040 041 042 043 044 045 046 047 048 Potential Enhancement (SAMA Title) Provide an in-containment reactor water storage tank. Throttle low pressure injection pumps earlier in medium or large-break LOCAs to maintain reactor water storage tank inventory. Emphasize timely recirculation alignment in operator training. Upgrade the chemical and volume control system to mitigate small LOCAs. Change the in-containment reactor water storage tank suction from four check valves to two check and two air-operated valves. Replace two of the four electric safety injection pumps with diesel-powered pumps. Provide capability for remote, manual operation of secondary side pilot-operated relief valves in a station blackout. Create a reactor coolant depressurization system. Make procedure changes for reactor coolant system depressurization. Add redundant DC control power for SW pumps. Replace ECCS pump motors with air-cooled motors. Enhance procedural guidance for use of cross-tied component cooling or service water pumps. Add a service water pump. Enhance the screen wash system. Cap downstream piping of normally closed component cooling water drain and vent valves. Source Result of Potential Enhancement Reference Continuous source of water to the safety injection pumps during a LOCA event, since water 10 released from a breach of the primary system collects in the in-containment reactor water storage tank, and thereby eliminates the need to realign the safety injection pumps for long-term post-LOCA recirculation. Extended reactor water storage tank capacity. 5 Reduced human error probability associated with recirculation failure. 5 For a plant like the Westinghouse AP600, where the chemical and volume control system 5 cannot mitigate a small LOCA, an upgrade would decrease the frequency of core damage. Reduced common mode failure of injection paths. 5 Reduced common cause failure of the safety injection system. This SAMA was originally 5, 10 intended for the Westinghouse-CE System 80+, which has four trains of safety injection. However, the intent ofthis SAMA is to provide diversity within the high-and low-pressure safety inieciion sysiems. Improved chance of successful operation during station blackout events in which high area 5 temperatures may be encountered (no ventilation to main steam areas). Allows low pressure emergency core cooling system injection in the event of small LOCA 5,10 and high-pressure safety iniection failure. Allows low pressure emergency core cooling system injection in the event of small LOCA 5 and high-pressure safety injection failure. Improvements Related to Cooling Water Increased availability of SW. 3 Elimination ofECCS dependency on component cooling system. I Reduced frequency of loss of component cooling water and service water. I Increased availabiiity of cooling water. 6 Reduced potential for loss of SW due to clogging of screens. 23 Reduced frequency of loss of component cooling water initiating events, some of which can 5 be attributed to catastrophic failure of one of the many single isolation valves. 58

o >> G> o o o o (J'1 00 1(J'1 o o o (j) 00 SAMAID 049 050 051 052 053 054 055 056 057 058 059 060 061 062 063 Potential Enhancement (SAMA Title) Enhance loss of component cooling water (or loss of service water) procedures to facilitate stopping the reactor coolant pumps. Enhance loss of component cooling water procedure to underscore the desirability of cooling down the reactor coolant system prior to seal LOCA. Additional training on loss of component cooling water. Provide hardware connections to allow another essential raw cooling water system to cool charging pum~ seals. On loss of essential raw cooling water, proceduralize shedding component cooling water loads to extend the component cooling water heat-up_ time. Increase charging pump lube oil capacity. Install an independent reactor coolant pump seal injection system, with dedicated diesel. Install an independent reactor coolant pump seal iniection system, without dedicated diesel. Use existing hydro test pump for reactor coolant pump seal injection. Install improved reactor coolant pump seals. Install an additional component cooling water I pump. Prevent makeup pump flow diversion through the relief valves. Change procedures to isolate reactor coolant pump seal return flow on loss of component cooling water, and provide (or enhance) guidance on loss of injection during seal LOCA. Implement procedures to stagger high pressure safety injection pump use after a loss of service water. Use fire prevention system pumps as a backup seal injection and high pressure makeup source. Result of Potential Enhancement Reduced potential for reactor coolant pump seal damage due to pump bearing failure. Reduced probability of reactor coolant pump seal failure. Improved success of operator actions after a loss of component cooling water. Reduced effect of loss of component cooling water by providing a means to maintain the charging pump seal injection following a loss of normal cooling water. Increased time before loss of component cooling water (and reactor coolant pump seal failure) during loss of essential raw cooling water sequences. Increased time before charging pump failure due to lube oil overheating in loss of cooling water sCiiucnces. Reduced frequency of core damage from loss of component cooling water, service water, or station blackout. Reduced frequency of core damage from loss of component cooling water or service water, but not a station blackout. Reduced frequency of core damage from loss of component cooling water or service water, but not a station blackout. Reduced likelihood of reactor coolant pump seal LOCA. Reduced likelihood of loss of component cooling water leading to a reactor coolant pump seal LOCA. Reduced frequency of loss of reactor coolant pump seal cooling if spurious high pressure injection relief valve opening creates a flow diversion large enough to prevent reactor coolant pump seal injection. Reduced frequency of core damage due to loss of seal cooling. Extended high pressure injection prior to overheating following a loss of service water. Reduced frequency of reactor coolant pump seal LOCA. 59 NEI 05-01 (Rev A) November 2005 Source Reference 5 5 5 5 5 5 5,10 5,10 5 5 5 5 S S 5

o >> G> o o o o (J'1 00 (J'1 10 o o (j) <D NEI 05-01 (Rev A) November 2005 SAMAID 064 065 066 067 068 069 070 071 072 073 074 075 076 077 Potential Enhancement (SAMA Title) Result of Potential Enhancement Implement procedure and hardware Improved ability to coel residual heat removal heat exchangers. modifications to allow manual alignment of the fire water system to the component cooling water system, or install a component cooling water header cross-tie. Improvements Related to Feedwater and Condensate Install a digital feed water upgrade. Reduced chance of loss of main feed water following a plant trip. Create ability for emergency connection of Increased availability of feedwater. existing or new water sources to feedwater and condensate systems. Install an independent diesel for the condensate Extended inventory in CST during an SBO. storage tank makeup pumps. Add a motor-driven feedwater pump. Increased availability of feedwater. Install manual isolation valves around auxiliary Reduced dual turbine-driven pump maintenance unavailability. feedwater turbine-driven steam admission valves. Install accumulators for turbine-driven Eliminates the need for local manual action to align nitrogen bottles for control air following auxiliary feedwater pump flow control valves. a loss of off-site power. Install a new condensate storage tank (auxiliary Increased availability ofthe auxiliary feedwater system. feedwater storage tank). Modify the turbine-driven auxiliary feedwater Improved success probability during a station blackout. pump to be self-cooled. Proceduralize local manual operation of Extended auxiliary feedwater availability during a station blackout. Also provides a success auxiliary feedwater system when control power path should auxiliary feedwater control power be lost in non-station blackout sequences. is lost. Provide hookup for portable generators to Extended auxiliary.feedwater availability. power the turbine-driven auxiliary feedwater pump after station batteries are depleted. Use fire water system as a backup for stearn Increased availability of steam generator water supply. 'generator inventory. Change failure position of condenser makeup Allows greater inventory for the auxiliary feedwater pumps by preventing condensate storage valve if the condenser makeup valve fails open tank flow diversion to the condenser. on loss of air or power. Provide a passive, secondary-side heat-Reduced potential for core damage due to loss-of-feedwater events. rejection loop consisting of a condenser and heat sink. 60 Source Reference 5 I 5 5 1,3 5 5 5,10 5 5 5,10 5 5 5

o >> G> o o o o (J'1 00 1(J'1 o o o o SAMAID 078 079 080 081 082 083 084 085 086 087 088 089 090 091 Potential Enhancement (SAMA Title) Result of Potential Enhancement Modify the startup feedwater pump so that it Increased reliability of decay heat removal. can be used as a backup to the emergency feedwater system, including during a station blackout scenario. Replace existing pilot-operated relief valves Increased probability of successful feed and bleed. with larger ones, such that only one is required for successful feed and bleed. Improvements Related to Heating, Ventilation, and Air Conditioning Provide a redundant train or means of Increased availability of components dependent on room cooling. ventilation. Add a diesel building high temperature alarm Improved diagnosis of a loss of diesel building HV AC. or redundant louver and thermostat. Stage backup fans in switchgear rooms. Increased availability of ventilation in the event of a loss of switchgear ventilation. Add a switchgear room high te~erature alarm. Improved diagllosis of a loss of switchgear HV AC. Create ability to ~witch emen>encv feenwMer --g+---.1 ----.. -..... Continued fan operation in a station blackout. room fan power supply to station batteries in a station blackout. Improvements Related to Instrument Air and Nitrogen Supply Provide cross-unit connection of Increased ability to vent containment using the hardened vent. uninterruptible compressed air supply. Modify procedure to provide ability to align Increased availability of instrument air after a LOOP. diesel power to more air compressors. Replace service and instrument air compressors Elimination of instrument air system dependence on service water cooling. with more reliable compressors which have self-contained air cooling by shaft driven fans. InstaJl nitrogen bottles as backup gas supply for Extended SRV operation time. safety relief valves. Improve SRV and MSIV pneumatic Improved availability ofSRVs and MSIVs. components. Improvements Related to Containment Phenomena Create a reactor cavity flooding system. Enhanced debris cool ability, reduced core concrete interaction, and increased fission product scrubbing. InstaJl a passive containment spray system. Improved containment spray capability. 61 NEI 05-01 (Rev A) November 2005 Source Reference 10 S 1 I S S S 3 18 S 18 6 1,7,11,12 6, 14

o >> G> o o o o (J'1 00 1(J'1 o o o NEI 05-01 (Rev A) November 2005 SAMAID 092 093 094 095 096 097 098 099 100 101 102 103 104 105 Potential Enhancement (SAMA Title) Use the fire water system as a backup source for the containment spray system. Install an unfiltered, hardened containment vent. Install a filtered containment vent to remove decay heat Option I: Gravel Bed Filter Option 2: MUltiple Venturi Scrubber Enhance fire protection system and standby gas treatment system hardware and procedures. Provide post-accident containment inerting capability. Create a large concrete crucible with heat removal potential to contain molten core debris. Create a core melt source reduction system. Strengthen primary/secondary containment (e.g., add ribbing to containment shell). Increase depth of the concrete base mat or use an alternate concrete material to ensure melt-through does not occur. Provide a reactor vessel exterior cooling system. Construct a building to be connected to primary/secondary containment and maintained at a vacuum. Institute simulator training for severe accident scenarios. Improve leak detection procedures. Delay containment spray actuation after a large LOCA. Source Result of Potential Enhancement Reference Improved containment spray capability. 4,6 Increased decay heat removal capability for non-ATWS events, without scrubbing released 6,8,9 fission products. Increased decay heat removal capability for non-A TWS events, with scrubbing of released 6,8,9,14 fission products. Improved fission product scrubbing in severe accidents. 9 Reduced likelihood of hydrogen and carbon monoxide gas combustion. 6,7,12 Increased cooling and containment of molten core debris. Molten core debris escaping from 6,8,9 the vessel is contained within the crucible and a water cooling mechanism cools the molten cere in the crucible, preventing melt-through of the base mat. Increased cooling and containment of molten core debris. Refractory material would be 13 placed underneath the reactor vessel such that a molten core falling on the material would melt and combine with the material. Subsequent spreading and heat removal from the vitrified compound would be facilitated, and concrete attack would not occur. Reduced probability of containment over-pressurization. 5,6,10,14 Reduced probability of base mat melt-through. 10 Increased potential to cool a molten core before it causes vessel failure, by submerging the 10 lower head in water. Reduced probability of containment over-pressurization. 6,10 Improved arrest of core melt progress and prevention of containment failure. 6 Increased piping surveillance to identify leaks prior to complete failure. Improved leak 6 detection would reduce LOCA frequency. Extended reactor water storage tank availability. 5 62

o >> G> o o o o (J'1 00 1(J'1 o o o ~ SAMAID 106 107 108 109 110 III 112 113 114 115 116 117 1I8 119 120 Potential Enhancement (SAMA Title) Result of Potential Enhancement Install automatic containment spray pump Extended time over which water remains in the reactor water storage tank, when full header throttle valves. containment spray flow is not needed. Install a redundant containment spr~_system. Increased containment heat removal ability. Install an independent power supply to the Reduced hydrogen detonation potential. hydrogen control system using either new batteries, a non-safety grade portable generator, existing station batteries, or existing ACIDC independent power supplies, such as the securi!y system diesel. Install a passive hydrogen control system. Reduced hydrogen detonation potential. Erect a barrier that would provide enhanced Reduced probability of containment failure. protection of the containment walls (shell) from ejected core debris following a core melt scenario at high pressure. Improvements Related to Containment Bypass Install additional pressure or leak monitoring Reduced ISLOCA frequency. instruments for detection ofISLOCAs. Add redundant and diverse limit switches to Reduced frequency of containment isolation failure and ISLOCAs. each containment isolation valve. Increase leak testing of valves in ISLOCA Reduced ISLOCA frequency. paths. Install self-actuating containment isolation Reduced frequency of isolation failure. valves. Locate residual heat removal (RHR) inside Reduced frequency of ISLOCA outside containment. containment Ensure ISLOCA releases are scrubbed. One Scrubbed ISLOCA releases. method is to plug drains in potential break areas so that break point will be covered with water. Revise EOPs to improve ISLOCA Increased likelihood that LOCAs outside containment are identified as such. A plant had a identification. scenario in which an RHR ISLOCA could direct initial leakage back to the pressurizer relief tank, giving indication that the LOCA was inside containment. Improve operator training on ISLOCA coping. Decreased ISLOCA consequences. Institute a maintenance practice to perform a Reduced frequency of steam generator tube ruptures. 100% inspection of steam generator tubes during each refueling outage. Replace steam generators with a new design. Reduced frequency of steam generator tube ruptures. 63 NEI 05-01 (Rev A) November 2005 Source Reference 5 5,10 5,10 5,10 5 4,7,11,12,15 1 1 5 14 I 1 1 5,10 5

o >> G> o o o o (J'1 00 1(J'1 o o o w NEI 05-01 (Rev A) November 2005 SAMAID 121 122 123 124 125 126 127 128 129 130 131 132 133 134 Potential Enhancement (SAMA Title) Increase the pressure capacity of the secondary side so that a steam generator tube rupture would not cause the relief valves to lift. Install a redundant spray system to depressurize the primary system during a steam generator tube rupture Proceduralize use of pressurizer vent valves during steam generator tube rupture sequences. Provide improved instrumentation to detect steam generator tube ruptures, such as Nitrogen-l 6 monitors). Route the discharge from the main steam safety valves through a structure where a water spray would condense the steam and remove most of the fission }Jl'oducts. Install a highly reliable (closed loop) stearn generator shell-side heat removal system that relics on natural circulation a11d stored water sources Revise emergency operating procedures to direct isolation of a faulted steam generator. Direct steam generator flooding after a steam generator tube rupture, prior to core damage. Vent main stearn safety valves in containment. Add an independent boron injection system. Add a system of relief valves to prevent equipment damage from pressure spikes during anATWS. Provide an additional control system for rod insertion (e.g., AMSAC). Install an ATWS sized filtered containment vent to remove decay heat. Revise procedure to bypass MSIV isolation in turbine trip A TWS scenarios. Source Result of Potential Enhancement Reference Eliminates release pathway to the environment following a steam generator tube rupture. 5,10 Enhanced depressurization capabilities during steam generator tube rupture. 5, 10 Backup method to using pressurizer sprays to reduce primary system pressure following a 5 steam generator tube rupture. Improved mitigation of steam generator tube ruptures. 5,10 Reduced consequences of a steam generator tube rupture. 10 Reduced consequences 'of a steam generator tube rupture. 5 Reduced consequences of a steam generator tube rupture. 5 Improved scrubbing of steam generator tube rupture releases. 5 Reduced consequences of a steam generator tube rupture. 5,10 Improvements Related to ATWS Improved availability of boron injection during ATWS. 18 Improved equipment availability after an ATWS. 19 Improved redundancy and reduced ATWS frequency. 18 Increased ability to remove reactor heat from A TWS events. 6 Affords operators more time to perform actions. Discharge of a substantial fraction of steam 1,20 to the main condenser (i.e., as opposed to into the primary containment) affords the operator more time to perform actions (e.g., SLC injection, lower water level, depressurize RPV) than if the main condenser was unavailable, resulting in lower human error probabilities. 64

o >> G> o o o o (J'1 00 1(J'1 o o o .j>.. SAMAID 135 136 137 138 139 140 141 142 143 144 145 146 147 148 149 Potential Enhancement (SAMA Title) Result of Potential Enhancement Revise procedure to allow override of low Allows immediate controloflow pressure core injection. On failure of high pressure core pressure core injection during an ATWS event. injection and condensate, some plants direct reactor depressurization followed by five minutes of automatic low pressure core injection. Install motor generator set trip breakers in Reduced frequency of core damage due to an A TWS. control room. Provide capability to remove power from the Decreased time required to insert control rods if the reactor trip breakers fail (during a loss of bus powering the control rods. feedwater A TWS which has rapid pressure excursion). Improvements Related to Internal Flooding Improve inspection of rubber expansion joints Reduced frequency of internal flooding due to failure of circulating water system expansion on main condenser. ~oints. Modify swing direction of doors separating Prevents flood propagation. turbine building basement from areas containing safeguards equipment. Improvements to Reduce Seismic Risk Increase seismic ruggedness of plant Increased availability of necessary plant equipment during and after seismic events. components. Provide additional restraints for CO2 tanks. Increased availability of fire protection given a seismic event. Improvements to Reduce Fire Risk Replace mercury switches in fire protection Decreased probability of spurious fire suppression system actuation. system. Upgrade fire compartment barriers. Decreased consequences of a fire. Install additional transfer and isolation Reduced number of spurious actuations during a fire. switches. Enhance fire brigade awareness. Decreased consequences of a fire. Enhance control of combustibles and ignition Decreased fire frequency and consequences. sources. Other Improvements Install digital large break LOCA protection Reduced probability of a large break LOCA (a leak before break). system. Enhance procedures to mitigate large break Reduced consequences of a large break LOCA. LOCA. Install computer aided instrumentation system Improved prevention of core melt sequences by making operator actions more reliable. to assist the operator in assessing post-accident plant status. 65 NEI 05-01 (Rev A) November 2005 Source Reference 16 5 5 1 5 3,10 17 7 7 18 7 7 5 7 6

o >> G> o o o o (J'1 00 1(J'1 o o o (J'1 NEI 05-01 (Rev A) November 2005 SAMAID 150 151 152 153 Potential Enhancement (SAMA Title) Improve maintenance procedures. Increase training and operating experience feedback to improve operator response. Develop procedures for transportation and nearby facility accidents. Install secondary side guard pipes up to the main steam isolation valves. Source Result of Potential Enhancement Reference Improved prevention of core melt sequences by increasing reliability of important 6 equipment. Improved likelihood of success of operator actions taken in response to abnormal conditions. 6 Reduced consequences of transportation and nearby facility accidents. 7 Prevents secondary side depressurization should a steam line break occur upstream of the 5, JO main steam isolation valves. Also guards against or prevents consequential multiple steam igenerator tube ruptures following a main steam line break event. 66

o >> G> o o o o (J'1 00 1(J'1 o o o (j) Table 14 References NEI 05-01 (Rev A) November 2005

1. NUREG-1560, "Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance," Volume 2, U.S. Nuclear Regulatory Commission, December 1997.

2. Letter from Mr. M. O. Medford (TVA) to NRC Document Control Desk dated September 1, 1992. "Watts Bar Nuclear Plant (WBN) Units 1 and 2 - Generic Letter (GL) 88 Individual Plant Examination (IPE) for Severe Accident Vulnerabilities - Response - (TAC M74488)."

3. Appendix D-Attachment F, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Edwin I. Hatch Nuclear Power Plant Units 1 and 2, March 2000.

4. Letter from Mr. D. E. Nunn (TVA) to NRC Document Control Desk, dated October 7, 1994. "Watts Bar Nuclear Plant (WBN) Units 1 and 2 -

Severe Accident Mitigation Design Alternatives (SAMDA) - Response to Request for Additional Information (RAI) - (TAC Nos. M77222 and M77223)."

5. NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Calvert Cliffs Nuclear Power Plant",

Suppiement 1, U.S. Nuclear Regulatory Commission, February 1999.

6. General Electric Nuclear Energy, Technical Support Document for the ABWR, 25A5680, Revision 1, January 18, 1995.

7. NUREG-0498, "Final Environmental Statement related to the operation of Watts Bar Nuclear Plant, Units 1 and 2," Supplement No.1, U.S.

Nuclear Regulatory Commission, April 1995.

8. Cost Estimate for Severe Accident Mitigation Design Alternatives, Limerick Generating Station for Philadelphia Electric Company, Bechtel Power Corporation, June 22, 1989.

9. NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Volume 1, 5.35, Listing of SAMDAs considered for the Limerick Generating Station, U.S. Nuclear Regulatory Commission, May 1996.

10. NUREG-1462, "Final Safety Evaluation Report Related to the Certification of the System 80+ Design," U.S. Nuclear Regulatory Commission, August 1994.

11. NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants," Volume 1, 5.36, Listing of SAMDAs considered for the Comanche Peak Steam Electric Station, U.S. Nuclear Regulatory Commission, May 1996.

67

o >> G> o o o o (J'1 00 1(J'1 o o o NEI 05-01 (Rev A) November 2005

12. Letter from Mr. W. 1. Museler (Tennessee Valley Authority) to the NRC Document Control Desk, dated October 7, 1994, "Watts Bar Nuclear Plant (WBN) Units 1 and 2 - Severe Accident Mitigation Design Alternatives (SAMDAs)."

13. Letter from Mr. D. E. Nunn (Tennessee Valley Authority) to NRC Document Control Desk, dated June 30, 1994. "Watts Bar Nuclear Plant (WBN) Unit 1 and 2 - Severe Accident Mitigation Design Alternatives (SAMDAs) Evaluation from Updated Individual Plant Evaluation (IPE)."

14. Letter from N. J. Liparulo (Westinghouse Electric Corporation) to NRC Document Control Desk, dated December 15, 1992, "Submittal of Material Pertinent to the AP600 Design Certification Review."

15. NUREG/CR-5474, "Assessment of Candidate Accident Management Strategies", U.S. Nuclear Regulatory Commission, March 1990.

16. Severe Accident Applicability of BWROG Revision 4, "Emergency Procedure Guidelines", BWROG, September 1988.

17. Appendix E-Environmental Report, Appendix G, Severe Accident Mitigation Alternatives Submittal Related to Licensing Renewal for the Peach Bottom Nuclear Power Plant Units 2 and 3, July, 2001.

18. Appendix F, Severe Accident Mitigation Alternatives Analysis Submittal Related to Licensing Renewal for the Quad Cities Nuclear Power Plant Units 1 and 2, January 2003.

19. NEDC-33090P, Rev.O, "Safety Analysis Report for Vennont Yankee Nuclear Power Station Constant Pressure Power Uprate", September 2003.

20. BWROG EPC Issue 98-07.

21. Individual Plant Examination for Severe Accident Vulnerabilities - Generic Letter 88-20, U.S. Nuclear Regulatory Commission, November 23, 1988.

22. NRC Bulletin 96-03, "Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling-Water Reactors", May 1996 U.S. Nuclear Regulatory Commission.

23. Duke Power Company, Applicant's Environmental Report, Operating Licensing Renewal Stage. Attachment K, "Oconee Nuclear Station Severe Accident Mitigation Alternatives (SAMA) Analysis." Rev. O. Charlotte, North Carolina, June 1998.

68

Initial SAMA List Phase I Analysis Figure 1 SAMA Sicreening Process 69 Retain for potential implementation Phase II Analysis No Screened OAGI0000585 00078

SAMA Analysis Guidance Document 11 REFERENCES 1 Pages G-8 and G-28 of Draft NUREG-1437, Supplement 19, Generic Environmental Impact Statement/or License Renewal 0/ Nuclear Plants, Regarding Arkansas Nuclear One, Unit 2, August 2004. 2 NUREG/CR-6613, Vol. 1, Code Manual/or MACCS2, User's Guide, D. Chanin and M.L. Young, Technadyne Engineering Consultants and Sandia National Laboratories for U. S. Nuclear Regulatory Commission and U. S. Department of Energy, SAND97-0594, May 1998. 3 NUREG/CR-4551, Evaluation o/Se1!ere Accident Risks: Quantification o/Major Input Parameters, MACCS Input, J. L. Sprung, et. aI., Sandia National Laboratories for the U. S. NRC, Vol. 2, Rev. 1, Part 7, Dc::cember 1990. 4 NUREGIBR-0184, Regulatory Analysis Technical Evaluation Handbook, U. S. Nuclear Regulatory Commission, 1997. 5 Office of Management and Budget, "Regulatory Analysis," Circular No. A-4, September 17,2003. http://www.wbitehouse.gov/omb/circulars/a004/a-4.pdf 6 NUREGIBR-0058, Revision 4, Regulatory Analysis Guidelines o/the U.S. Nuclear Regulatory Commission, September 2004. 71 OAGI0000585 00079}}