ML042860100

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Revision 2, Probabilistic Safety Assessment, Summary Report.
ML042860100
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 02/23/2004
From: Mikschl T
ABS Consulting
To:
Office of Nuclear Reactor Regulation
References
R05 040921 002
Download: ML042860100 (41)
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R05 040921 002 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT PROBABILISTIC SAFETY ASSESSMENT UNIT 3

SUMMARY

REPORT Revision 2 RISR COESU CoNIsultOg Q RISK CONSULTING DIVISION

TENNESSEE VALLEY AUTHORITY SYSTEMS AND ANALYSIS BROWNS FERRY NUCLEAR PLANT PROBABILISTIC SAFETY ASSESSMENT UNIT 3

SUMMARY

REPORT Revision 2 February 2004

Unit 3 Summary Report Browns Ferry Nuclear Plant Probabilistic Safety Assessment REVISION LOG Unit 3 Summary Report

[ oevn l Description of Revision PPrepared By I l Checked By I Approved By I N j _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _jDate IDa t

e I Date 0 Initial Issue S. Rodgers D.Bidwell D. McCamy 1 Changes from EPU S. Rodgers D.Johnson H.Jones 5-13-03 2 Resolve minor comments identified in SAMA T. Mikschl/ D. H. Johnson/ H.Jones analysis. 2-23-04 2-23-04 9-21-04 i S1059446-1426.022304

Unit 3 Summary Report TABLE OF CONTENTS Section Page 1.0 EXECUTIVE

SUMMARY

..................................... . 1-1

1.1 BACKGROUND

AND OBJECTIVES . . . ................................ 1-1 1.2 PLANT FAMILIARIZATION ................................... 1-1 1.3 OVERALL METHODOLOGY ................................... 1-2 1.4

SUMMARY

OF MAJOR FINDINGS ................................... 1-4 1.4.1 Total Core Damage and Large Early Release Frequency . 1-5 1.4.2 Contributors to Total Core Damage Frequency .1-9 1.4.2.1 Important Core Damage Sequence Groups. 1-9 1.4.2.2 Analysis of Individual Sequences .1-17 1.4.2.3 Important Operator Actions .1-21 1.4.2.4 Important Plant Hardware Characteristics. 1-22 1.4.3 Results for Large Early Release Frequency .1-23 1.4.3.1 Important Plant Hardware Characteristics for Containment Performance .1-24 1.4.4 Comparison with the 2002 Browns Ferry Unit 3 PRA, Revision 0 .1-25 1.5 INSIGHTS ............... 1-25

2.0 REFERENCES

............... 2-1 APPENDIX A Unit 3 Top Ranking Sequences Contributing to CDF BFN_Unit 3 Summary Rpt (Rev 2).doc ii S1 059446-14264022304

Unit 3 Summary Report LIST OF TABLES Table 1-1 Comparison with Other PRAs ........................................................... 1-8 Table 1-2 Unit 3 Initiating Event Group Contributions to Core Damage Frequency ....1-11 Table 1-3 Summary of the Core Damage Accident Sequence Subclasses .................. 1-15 Table 1-4 Breakdown of Core Damage Sequences in Each Frequency Range ........... 1-17 Table 1-5 Browns Ferry Unit 3 Important Operator Actions ............................................. 1-22 Table 1-6 Summary of Revised Human Error Probabilities .............................................. 1-25 LIST OF FIGURES Figure 1-1 Uncertainty Curve for Browns Ferry Unit 3 Core Damage Frequency ......... 1-6 Figure 1-2 Uncertainty Curve for Browns Ferry Unit 3 Large Early Release Frequency........................................................................................................................... 1-7 Figure 1-3 Browns Ferry Unit 3 Core Damage Frequency by Initiating Event Category ........................................................ 1-10 BFN_Unit 3 Summary Rpt (Rev 2).doc ii S1059446-14264022304

Unit 3 Summary Report SECTION 1 EXECUTIVE

SUMMARY

1.1 BACKGROUND

AND OBJECTIVES This documents the performance by the Tennessee Valley Authority (TVA) in updating the Unit 3 PSA. An integrated team of engineers and specialists from TVA and ABS Consulting performed this revision TVAs overall objectives for this revision were to incorporate the Extended Power Uprate into the PSA.

The purpose of this summary is to present the results of the PSA on Browns Ferry Unit

3. These results include an estimate of the total core damage frequency (CDF); data uncertainties in the estimated CDF; and the large early release frequency (LERF) data uncertainties in the estimated LERF. This summary also provides the sequences, systems, and sources of uncertainty that are the significant contributors to the results.

1.2 PLANT FAMILIARIZATION The Browns Ferry Nuclear Plant is located on the north shore of Wheeler Lake at Tennessee River mile 294 in Limestone County, Alabama. The site is approximately 10 miles southwest of Athens, Alabama, and 10 miles northwest of Decatur, Alabama.

The plant consists of three units, with Unit 1 rated power level of 3,293 MWt and Unit 2 and 3 rated at 3,952 MWt. Unit 2 and Unit 3 are the only units currently operating.

Unit 3 is a single-cycle forced-recirculation boiling water reactor (BWR) nuclear steam supply system supplied by General Electric Corporation. Major structures at Browns BFN.Unit 3 Summary Rpt (Rev 2).doc 1-1 S1 059446-1426-022304

Unit 3 Summary Report Ferry Unit 3 include a reactor building with a Mark I drywell containment, a turbine building, a control bay, and an intake pumping station.

A detailed description of the plant site, facilities, and safety criteria is documented in the Browns Ferry Final Safety Analysis Report (Reference 1-2).

1.3 OVERALL METHODOLOGY The Browns Ferry Unit 3 PSA is founded on a scenario-based definition of risk (Reference 1-3). In this application, "risk" is defined as the answers to three basic questions:

1. What can go wrong?
2. What is the likelihood?
3. What are the consequences?

Question 1 is answered with a structured set of scenarios that is systematically developed to account for design and operating features specific to Browns Ferry Unit 3.

Question 2 is answered with a prediction or estimate of the frequency of occurrence of each scenario identified in the answer to question 1. Since there is uncertainty in that frequency, the full picture of likelihood is conveyed by a probability curve that conveys the state of knowledge, or confidence, about that frequency.

Question 3 is answered in two ways. One measure is the core damage frequency. The loss of adequate core cooling is defined as the rapid increase in fuel clad temperature due to heating and Zircaloy-water reactions that lead to sudden deterioration of fuel clad integrity. For the purposes of the Level 1 PSA a surrogate has been developed that can be used as a first approximation to define the onset of core damage. The onset of core damage is defined as the time at which more than two-thirds of the active fuel becomes BFN.Unit 3 Summary Rpt (Rev 2).doc 1-2 S1 059446-1426-022304

Unit 3 Summary Report uncovered, without sufficient injection available to recover the core quickly, i.e., water level below one-third core height and falling. The other measure is the large, early release frequency. The original IPE answered question 3 in a Level 2 PSA, in terms of the key characteristics of radioactive material release that could result from the sequences identified. Consistent with recent PSA practice, BFN does not track the entire spectrum of releases. Instead, it tracks the frequency of large, early releases. A large early release is defined as the rapid, unscrubbed release of airborne fission products from the containment to the environment occurring before the effective implementation of off-site emergency response and protective actions. The results reported here are based on the methods that conform to the NRC guidelines (Reference 1-1, Appendix 1) and the IEEE/ANS "PSA Procedures Guide" (Reference 1-4).

A large fraction of the effort needed to complete this PSA was to develop a plant-specific model to define a set of accident sequences. This model contains a large number of scenarios that have been systematically developed from the point of initiation to termination. A series of event trees is used to systematically identify the scenarios.

Given the knowledge of the event tree structures, accident sequences are identified by specifying:

1. The initiating event.
2. The plant response in terms of combinations of systems and operator responses.
3. The end state of the accident sequence.

The RISKMAN3 PC-based software system (Reference 1-5) was used to construct effectively a single, large tree for Level 1 and LERF. The sequences analysis start with an initiating event and terminate in end states of LERF or no LERF. The sum of these two end states is the CDF.

The initiating events and the event tree branching frequencies are quantified using different types of models and data. The system failures that contribute to these events BFNUnit 3 Summary Rpt (Rev 2).doc 1-3 S1059446-1426-022304

Unit 3 Summary Report are analyzed with the use of fault trees that relate the initiating events and event tree branching frequencies to their underlying causes. These causes are quantified, in turn, by application of models and data on the respective unavailabilities due to hardware failure, common cause failure, human error, and test and maintenance unavailabilities.

The frequencies of initiating events, the hardware failure rates of the components, and operator errors were obtained using either generic data or a combination of generic and plant-specific data.

Dependency matrices have previously been developed from a detailed examination of the plant systems to account for important interdependencies and interactions that are highly plant specific. To facilitate a clear definition of plant conditions in the scenarios, separate stages of event trees are provided for the response of the support systems (e.g., electric power and cooling water), the frontline systems [e.g., high pressure coolant injection (HPCI) and residual heat removal (RHR)], and the containment phenomena; e.g., containment overpressurization failure. A separate tree is used to determine core damage and develop plant damage classes. This tree provides the interface between the Level 1 and Level 2 event trees.

The systematic, structured approach that is followed in constructing the accident scenario model provides assurance that plant-specific features are identified. It also provides insights into the key risk controlling factors.

1.4

SUMMARY

OF MAJOR FINDINGS The major findings of the Browns Ferry Unit 3 Level 2 PSA are presented in this section.

The results delineate the principal contributors to risk, and provide insights into plant and operational features relevant to safety. The presentation describes both the core damage and large early release results.

BFN_Unit 3 Summary Rpt (Rev 2).doc 1-4 S1059446-1426-022304

Unit 3 Summary Report 1.4.1 Total Core Damage and Large Early Release FrequencV The total CDF for Browns Ferry Unit 3 was found to be 3.4 x 10.6 per reactor-year. The results for CDF were developed in terms of a mean point estimate. The CDF data uncertainty curve is shown in Figure 1-1.

The total Large Early Release Frequency (LERF) for Browns Ferry Unit 3 was found to be 4.5 x 10-7 per reactor year. The results for LERF were developed in terms of a mean point estimate. The LERF data uncertainty curve is shown in Figure 1-2.

BFNUnit 3 Summary Rpt (Rev 2).doc 1-5 S1059446-1426-022304

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Unit 3 Summary Report A comparison of this study with other PSAs on other plants that used similar methods, databases, and work scopes is given in Table 1-1. The calculated mean CDF for Browns Ferry Unit 3 is of the same order of magnitude as Quad Cites, Peach Bottom Unit 2 and Grand Gulf Unit 1, and an order of magnitude lower than that reported for Nine Mile Point Unit 2 (which includes external events).

Table 1-1 Comparison with Other PRAs Plant Il__I_Flood Included l Mean CDF I (per reactor-year) l Reference j

l Mean LERF (per year)

Quad Cities Yes 4.6E-6 1-7 3.3E-6 Nine Mile Point Unit 2* Yes 5.7E-5 1-8 1.6E-6 Browns Ferry Unit 3 Yes 3.4E-6 This Study 4.5E-7 Peach Bottom Unit 2 No 4.5E-6 1-9 Not Updated Grand Gulf Unit 1 No 5.5E-6 1-10 Not Updated includes external events.

Factors that contribute to the results for Browns Ferry Unit 3 are summarized below:

  • The increase in core thermal power resulting from the EPU eliminated the use of CRD as an alternative injection source if the vessel remains at high pressure and other injection sources fail. The increase in the CDF estimate from Revision 0 is largely due to the elimination of this success path.
  • The accident sequences that were analyzed are those initated by internal events and internal floods. Sequences initiated by internal fires, seismic events, and other external events have not been modeled in this internal events model.
  • The current results do not reflect any future plant or procedural changes that TVA may decide to make to improve safety.

. This study used plant specific data to update failure rates for selected components and initiating events frequencies. The common cause parameters of the multiple Greek model used in this study were estimated with the benefit of a plant-specific screening of industry common cause event data in accordance BFNUnit 3 Summary Rpt (Rev 2).doc 1-8 S1059446-14264022304

Unit 3 Summary Report with NUREG/CR-4780 (Reference 1-11). The common cause event data was taken from the NRC database (Reference 1-14). Common cause estimates not screened were taken from NUREG/CR-5497.

1.4.2 Contributors to Total Core Damage Frequency In the quantification of the Level 1 event sequence models, the principal contributors to the CDF were identified from several vantage points. The results and contributors are summarized in this section. Causes for individual system failures are listed in each systems analysis notebook.

1.4.2.1 Important Core Damage Sequence Groups The importance of initiating events was examined by determining the contributions of core damage sequences grouped by initiating event. The ranked results are shown in Figure 1-3 and Table 1-2 for major initiating event categories.

The Loss of Offsite Power (LOSP) initiators include station blackout sequences (failure of all diesel generators) and nonstation blackout scenarios in which core damage resulted from other failures. These other failures include battery board failures (resulting in loss of high pressure injection and failure to achieve low pressure injection) and decay heat removal failures. Overall, the LOSP initiated sequences account for 31% of CDF.

Transients with the Power Conversion System (PCS) unavailable as a result of the initiator account for 27.2% of the CDF. Loss of condenser heat sink, which includes closure of the main steam isolation valves and turbine trip without bypass, are specific examples of initiator in this group.

BFN.Unit 3 Sumrnary Rpt (Rev 2).doc 1 -9 S1059446-1426-022304

Unit 3 Summary Report Transients with the PCS not disabled as a result of the initiator contribute 23.8% to the core damage frequency. The turbine trip, in which the main steam isolation valves and turbine bypass are available, is a specific example of an initiator in this group.

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Figure 1-3 Browns Ferry Unit 3 Core Damage Frequency by Initiating Event Category BFNUnit 3 Summary Rpt (Rev 2).doc 1-10 S1 059446-1426402304

a Unit 3 Summary Report Table 1-2 Unit 3 Initiating Event Group Contributions to Core Damage Frequency Initiating Event Category M Mean ya) Percentage of Total Loss of Offsite Power 1.05E-06 31.3 PCS Unavailable 9.15E-07 27.2 Transient with PCS Available 7.99E-07 23.8 Support Systems 2.34E-07 7.0 Internal Floods 1.63E-07 4.8 LOCAs 1.38E-07 4.1 Stuck Open Relief Valves 5.82E-08 1.7 Total 3.4E-06 100 Support system failure initiators (specifically, loss of plant air, loss of raw cooling water, or loss of either l&C bus 2A or 2B failures) contribute 7% to the total CDF.

Scenarios initiated by internal floods contribute 4.8% to the core damage frequency. No internal flooding scenarios lead directly to core damage but require additional hardware failures. Flooding initiators were postulated in the Unit 3 reactor building, in the Unit 1 or Unit 2 reactor building, and in the turbine building (two sizes).

LOCAs and interfacing systems LOCAs (i.e., when the boundary between a high and a low pressure system fails and the lower pressure system overpressurizes) make up 4.1% of the total CDF.

Scenarios initiated by the inadvertent opening of one or more safety relief valves (SRVs) contribute 1.7% the core damage frequency. Two distinct initiators are considered: opening of one SRV, and opening of two or more SRVs.

BFNDUnit 3 Summary Rpt (Rev 2).doc 1-1 1 S1059446-1426-022304

II Unit 3 Summary Report The preceding paragraphs considered the contribution to the total CDF from groups of initiating events. The sequences leading to core damage were also reviewed to identify common functional failures.

An event sequence classification into five accident sequence functional classes can be performed using the functional events as a basis for selection of end states. The description of functional classes is presented here to introduce the terminology to be used in characterizing the basic types of challenges to containment. The reactor pressure vessel condition and containment condition for each of these classes at the time of initial core damage is noted below:

Core Damage RPV Condition Containmentl Functional Class Condition I Loss of effective coolant inventory (includes high and low Intact pressure Inventory losses)

II Loss of effective containment pressure control, e.g., heat Breached or Intact removal IIl LOCA with loss of effective coolant Inventory makeup Intact IV Failure of effective reactivity control Breached or Intact V LOCA outside containment Breached (bypassed)

In assessing the ability of the containment and other plant systems to prevent or mitigate radionuclide release, it is desirable to further subdivide these general functional categories. In the second level binning process, the similar accident sequences grouped within each accident functional class are further discriminated into subclasses such that the potential for system recovery can be modeled. These subclasses define a set of functional characteristics for system operation which are important to accident progression, containment failure, and source term definition. Each subclass contains front end sequences with sufficient similarity of system functional characteristics that the containment accident progression for all sequences in the group can be considered to behave similarly in the period after core damage has begun. Each subclass defines a BFNUnit 3 Summary Rpt (Rev 2).doc 1-12 S1059446-1426-022304

Unit 3 Summary Report unique set of conditions regarding the state of the plant and containment systems, the physical state of the core, the primary coolant systems, and the containment boundary at the time of core damage, as well as vessel failure.

The important functional characteristics for each subclass are determined by defining the critical parameters or system functions that impact key results. The sequence characteristics that are important are defined by the requirements of the containment accident progression analysis. These include the type of accident initiator, the operability of important systems, and the value of important state variables (e.g., reactor pressure) that are defined by system operation. The interdependencies that exist between plant system operation and the core melt and radionuclide release phenomena are represented in the release frequencies through the binning process involving these subclasses, as shown in past PRAs and PRA reviews. The binning process, which consolidates information from the systems' evaluation of accident sequences leading to core damage in preparation for transfer to the containment-source term evaluation, involves the identification of 13 classes and subclasses of accident sequence types.

Table 1-3 provides a description of these subclasses that are used to summarize the Level 1 PRA results.

Published BWR PRAs have identified that there may be a spectrum of potential contributors to core melt or containment challenge that can arise for a variety of reasons. In addition, sufficient analysis has been done to indicate that the frequencies of these sequences are highly uncertain; and therefore, the degree of importance on an absolute scale and relative to each other, depends upon the plant specific features, assumptions, training, equipment response, and other items that have limited modeling sophistication.

This uncertainty means that the analyst can neither dismiss portions of the spectrum from consideration nor emphasize a portion of the spectrum to the exclusion of other BFN.Unit 3 Summary Rpt (Rev 2).doc 1-13 S1059446-1426-022304

Unit 3 Summary Report sequence types. This is particularly true when trying to assess the benefits and competing risks associated with a modification of a plant feature.

This end state characterization of the Level 1 PRA in terms of accident subclasses is usually sufficient to characterize the CET entry states for most purposes. However, when additional refinement is required in the CET quantification, it may be useful to further discriminate among the contributors to the core damage accident classes. This discrimination can be performed through the use of the individual accident sequence characteristics.

BFN-Unit 3 Summary Rpt (Rev 2).doc 1-14 S1 059446-1426-022304

Unit 3 Summary Report Table 1-3 Summary of the Core Damage Accident Sequence Subclasses Acciden ClassWASH--1400 Designator Subclass Definition Designator Desgnaor_________________________________________________ Example Class I A Accident sequences involving loss of inventory TQUX makeup in which the reactor pressure remains high.

B Accident sequences involving a station blackout and TEQUV loss of coolant Inventory makeup.

C Accident sequences involving a loss of coolant inventory induced by an ATWS sequence with TTCMQU containment intact.

D Accident sequences Involving a loss of coolant Inventory makeup In which reactor pressure has been successfully reduced to 200 psi; i.e., accident TQUV sequences Initiated by common mode failures disabling multiple systems (ECCS) leading to loss of coolant inventory makeup.

E Accident sequences Involving loss of Inventory makeup In which the reactor pressure remains high and DC power is unavailable.

Class 11 A Accident sequences Involving a loss of containment heat removal with the RPV Initially intact; core damage TW Induced post containment failure L Accident sequences involving a loss of containment heat removal with the RPV breached but no initial core AW damage; core damage after containment failure.

T Accident sequences Involving a loss of containment heat removal with the RPV Initially intact; core damage N/A induced post high containment pressure V Class IIA or IlL except that the vent operates as designed; loss of makeup occurs at some time following vent initiation. Suppression pool saturated 1W but Intact.

BFN_Urut 3 Summary Rpt (Rev 2).doc 1-15 S1059446-1426-022304

Unit 3 Summary Report Table 1-3 Summary of the Core Damage Accident Sequence Subclasses Accident Class Designator _______ss Subclass Definition-1400 Deiito Desgaor Class iII A Accident sequences leading to core damage LOCA) conditions initiated by vessel rupture where the R containment Integrity is not breached In the Initial time phase of the accident.

B Accident sequences initiated or resulting In small or medium LOCAs for which the reactor cannot be SQUX depressurized prior to core damage occurring.

C Accident sequences initiated or resulting in medium or large LOCAs for which the reactor is a low pressure AV and no effective Injection is available.

D Accident sequences which are Initiated by a LOCA or RPV failure and for which the vapor suppression AD system is inadequate, challenging the containment integrity with subsequent failure of makeup systems.

Class IV A Accident sequences involving failure of adequate (ATWS) shutdown reactivity with the RPV initially Intact; core TTCMC2 damage Induced post containment failure.

L Accident sequences Involving a failure of adequate shutdown reactivity with the RPV Initially breached N/A (e.g., LOCA or SORV); core damage Induced post containment failure.

T Accident sequences involving a failure of adequate shutdown reactivity with the RPV Initially intact; core N/A damage Induced post high containment pressure.

V Class IV A or L except that the vent operates as designed, loss of makeup occurs at some time N/A following vent initiation. Suppression pool saturated but Intact.

Class V Unisolated LOCA outside containment N/A For BFN, functional based plant damage states are used to summarize Level 1 results and to ensure that the Level 2 CETs are sufficient to allow each functional sequence to be addressed.

BFNUnnft 3 Summary Rpt (Rev 2).doc 1-16 S1059446-1426-022304

Unit 3 Summary Report 1.4.2.2 Analysis of Individual Sequences A large number of sequences make up the total CDF. Table 1-4 provides information on the distribution of core damage sequences across the frequency range.

Table 1-4 Breakdown of Core Damage Sequences in Each Frequency Range Frequency Range Number of Sequences Percentage of CDF (events per year)

>1 E-07 1 5

>1 E-08 36 28

>1 E-09 380 57

>1E-10 3179 81

>1 E-11 22,071 96

>1E-12 (base case) 44,066 100 The following presents a brief description of the 15 highest-ranking sequences to the CDF.

A loss of condenser heat sink initiates the first sequence. The initiator directly causes a loss of Feedwater, degrading high pressure injection capabilities. Subsequent failures of HPCI and RCIC eliminate all of high pressure injection. The remaining success path of low pressure injection is not viable because of a failure to depressurize. A lack of inventory causes core damage.

Sequence 2 is a non-minimal version of sequence 1, representing a different path in the LERF event tree.

A general transient initiates the third sequence. A subsequent loss of the main condenser results in a situation identical to the first sequence initiator, a loss of the condenser heat sink. The remainder of sequence 3 is identical to that of sequence 1.

BFNUnit 3 Summary Rpt (Rev 2).doc 1-17 S1059446-1426-022304

Unit 3 Summary Report The fourth sequence is that of an interfacing system LOCA that results in core damage, and is also a LERF sequence. This sequence represents the total contribution from a variety of interfacing system LOCAs. An interfacing system LOCA is initiated by leakage of reactor coolant through valves that separate the nuclear boiler from the RHR or core spray systems.

The fifth sequence is initiated by a total LOSP followed by the failure diesel generators 3A, 3B, 3C, and B. This combination of failures results in the 480V shutdown board 3B, which supplies room cooling to the B and D RHR pumps. Thus all Unit 3 RHR pumps are failed. The LPCI injection path is failed because of 3EB and 3EC diesel generators.

HPCI fails long term because of the failure of diesel generator 3EA, which maintains the charger for long-term 250V DC power. RCIC fails long term because of its dependency on diesel generator A. Diesel Generator A and B are required to maintain the charging for long-term 250V DC power. Hence, there is no high-pressure injection. Suppression Pool cooling is failed due to electrical supports. Core damage occurs because of lack of injection The sixth sequence is initiated by a total LOSP followed by the failure diesel generators 3A, 3B, 3C, and A. This combination of failures results in the loss of 480V shutdown board 3B, which supplies room cooling to the B and D RHR pumps. Thus all Unit 3 RHR pumps are failed. The LPCI injection path is failed because of 3EB and 3EC diesel generators. HPCI fails long term because of the failure of diesel generator 3EA, which maintains the charger for long-term 250V DC power. RCIC fails long term because of its dependency on diesel generator A. Diesel Generator A and B are required to maintain the charging for long-term 250V DC power. Hence, there is no high-pressure injection. Suppression Pool cooling is failed due to electrical supports.

Core damage occurs because of lack of injection.

The sixth sequence is initialized by a total LOSP followed by the failure diesel generators 3A, 3B, 3C, and A. This combination of failures results in the 480V BFNDUnit 3 Summary Rpt (Rev 2).doc 1-18 S1059446-1426-022304

I Unit 3 Summary Report shutdown board 3B, which supplies room cooling to the B and D RHR pumps. Thus all Unit 3 RHR pumps are failed. The LPCI injection path is failed because of 3EB and 3EC diesel generators. There is no suppression pool cooling without the HWWV. Core damage occurs because of lack of injection. HPCI fail long term because of the failure of diesel generator 3EA. RCIC fails long term because of its dependency on diesel generator A.

The seventh sequence is also initialized by LOSP. Diesel generator 3A, 3B, and 3C fail.

The RCIC pump also fails. Because of the dependency on shutdown board 3EA, HPCI fails in the long run. RHR pumps A, B, and C fail because of the failure of diesel generators 3A, 3B, and 3C. RHR pump D fails because of its dependency on 480V shutdown board 3EB. The RHR low pressure injection path fails. Core damage occurs due to lack of injection.

A LOSP initiates sequence 8. Unit 3 diesel generators A, B and C fail. RCIC fails to operate in the short term. The failure DG B fails 480V 3B, which powers the air-handling unit for RHR pump D. Hence, all injection is lost.

The ninth sequence is similar to sequence 3 except that the failure of condensate/feedwater is due to the failure of the turbine bypass valves.

Sequence 10 is a non-minimal sequence 2. The additional failure is Unit 2 at power.

Sequence 11 is initiated by a loss of offsite power to Unit 2. Although the diesel generators are successful, this sequence progress to core damage as HPCI and RCIC fail, followed by a failure to depressurize.

Sequence 12 is the classic SBO following a total LOSP. The Unit 3 diesel generators fail and the Unit 1/2 diesel generators fail by common cause. Offsite power is not recovered before core damage occurs.

BFN.Unit 3 Summary Rpt (Rev 2).doc 1-19 S1059446-1426-022304

Unit 3 Summary Report Sequence 13 is initiated by a loss of condenser heat sink. It is similar to sequence 2 but has additional, non-minimal failures.

A general transient initiates sequence 14. It is similar to sequence 3 but the failure of Feedwater is caused by the failure of the turbine bypass valves.

Sequence 15 is also initialized by a total LOSP. This is followed by a failure of diesel generators B, C, 3EA, and 3EB. This combination of diesel generator failures causes a loss of EECW. Offsite power is not recovered prior to core damage.

Appendix A of this report contains a listing of the top 50 core damage sequences.

The table below shows the frequency, percentage of total, and the cumulative percentage of total for the sequences discussed above.

BFNUnit 3 Summary Rpt (Rev 2).doc 1-20 S1059446-1426-022304

Unit 3 Summary Report Sequence Frequency  % CDF Cumulative%

1 1.58E-07 4.7 4.7 2 5.83E-08 1.7 6.4 3 5.76E-08 1.7 8.1 4 4.63E-08 1.4 9.5 5 4.38E-08 1.3 10.8 6 4.28E-08 1.3 12.1 7 3.85E-08 1.1 13.2 8 3.57E-08 1.1 14.3 9 2.95E-08 0.9 15.2 10 2.77E-08 0.8 16.0 11 2.55E-08 0.8 16.8 12 2.55E-08 0.8 17.5 13 2.43E-08 0.7 18.2 14 2.42E-08 0.7 19.0 15 2.40E-08 0.7 19.7 1.4.2.3 Important Operator Actions The importance of a specific operator action was determined by summing the frequencies of the sequences involving failure of that action and comparing that sum to the total CDF. The importance is the ratio of that sum to the total CDF.

Table 1-5 summarizes the important operator action failures ranked in crder of their impact on the total CDF. The operator actions to recover electric power are not included in Table 1-5 because they are a complex function of the time available and the specific equipment failures involved. No other HEPs are shown because of a dramatic fall off in importance.

BFNUnit 3 Summary Rpt (Rev 2).doc 1-21 S1059446-1426-022304 1

Unit 3 Summary Report Table 1-5 Browns Ferry Unit 3 Important Operator Actions Operator Action l Top Event l Importance Align RHR for Suppression Pool Cooling ORVD 44.8 Control Reactor Vessel Level at Low Pressure using RHR or Core OLP 9.4 Spray Align Alternate Injection to Reactor vessel via the Unit 2 RHR U2 7.7 Crosstie*

Manual Depressurization of the Reactor Vessel using the Safety Relief OSP 4.8 Valves

  • The Importance of Top Event U2 was weighted by the relative contribution of the human action contained In the system analysis.

1.4.2.4 Important Plant Hardware Characteristics An importance analysis of plant system failure modes to the total CDF was also performed. Only hardware failures involving the system itself are considered in Table 1-6, which provides a ranking in order of their impact on the total CDF.

BFNUnit 3 Summary Rpt (Rev 2).doc 1-22 S1050,446-1426-022304

Unit 3 Summary Report Table 1-6 REVISE Browns Ferry Unit 3 Important Systems System  % CDFl HPCI 56 RCIC 55 Diesel Generators 28 Main Steam 15 RHR 14 Feedwater/Condensate 11 RPS 10 RHRSW to RHR Loop II 4 RHRSW to RHR Loop I 4 Core Spray 3 Standby Liquid Control 3 Control Rod Drive 2 The system importance means the fraction of the CDF involving partial or complete failure of the indicated system. These importance measures are not strictly additive because multiple system failures may occur in the same sequence. The importance rankings account for failures within the systems that lead to a plant trip, or failures that limit the capability of the plant to mitigate the cause of a plant trip. Consequential failures resulting from dependencies on other plant systems [e.g., the loss of drywell control air due to failure of reactor building closed cooling water (RBCCW)] are not included in this importance ranking.

1.4.3 Results for Larae Earlv Release Freauencv BFN_Unft 3 Summary Rpt (Rev 2).doc 1-23 S1059446-1426-022304

Unit 3 Summary Report This section summarizes the limited results for the Level 2 analysis, which estimates the large containment failure and subsequent early release of radionuclides known as LERF. In contrast to the IPE submittal, this update concerned itself with two metrics, core damage frequency and large early release frequency. This section presents the LERF results and contributors.

The results indicate that about 13% of the core damage scenarios esult in LERF.

Typically, LERF as a percentage of CDF for BWRs ranges from 10% to almost 50%.

These are generally highly dependent on the level 1 results. BFN Unit 3 falls in the mid-range for BWRs.

This release category represents the most severe source term scenario. Here the containment failures are dominated by drywell shell failures (due to thermal interactions with the molten core debris on the drywell floor). The important post-core damage contributors are drywell shell failures, in-vessel recovery, and the effectiveness of the reactor building in scrubbing the release. With respect to pre-core damage top events, the failure of the RPS system dominates.

1.4.3.1 Important Plant Hardware Characteristics for Containment Performance As discussed in the previous Section 1.4.3.1, the dominant contributor to the most significant release category group (large, early containment failure and large bypasses) is drywell shell thermal attack from corium on the drywell floor. This is representative for most Mark I containments. The likelihood of drywell shell thermal attack failure is significantly reduced if the drywell floor is flooded with water prior to vessel breach.

Drywell spray represents an important hardware component since it is the primary means of flooding the drywell.

BFN-UJit 3 Summary Rpt (Rev 2).doc 1-24 S1 059446-1426-022304

Unit 3 Summary Report 1.4.4 Comparison with the 2002 Browns Ferry Unit 3 PRA, Revision 0 TVA has previously performed an individual plant examination in accordance with the U.S. Nuclear Regulation Commission (NRC) Generic Letter No. 88-20 (Reference 1-1).

The IPE was revised on several occasions. PSA Revision 0 marked the change from IPE to an application and risk informed approach. This Revision 2 reflects plant operations with the extended power uprate. The increase in thermal power eliminated the use of the CRD system as an effective injection source when the vessel remains at high pressure and the other high pressure injection sources have failed. The increase in thermal power also required revisions to some human actions due to the change in sequence timi ng. See Table 1-7.

Table 1-6 Summary of Revised Human Error Probabilities Operator l Current I Previous l Notes Action HEP HEP l HOAD1 4.89E-03 3.45E-03 Inhibit ADS During ATWS with Unisolated Vessel HOAD2 9.52E-03 4.64E-03 Inhibit ADS During ATWS with Isolated Vessel HOAL2 1.29E-01 3.91 E-02 Lower and Control Vessel Level HOSL1 1.61 E-02 6.71 E-03 Initiate SLCS Given ATWS with Unisolated RPV.

HOSL2 7.71E-02 3.50E-02 Initiate SLCS, Given an ATWS with RPV Isolated 1.5 INSIGHTS The power increase eliminated the use of CRD as a viable high pressure injection if the vessel remains at high pressure. The increase in CDF given EPU as compared to the Revision 0 model is almost entirely due to this elimination. The high pressure injection systems and the operator action to depressurize are much more important given EPU.

BFNUrnt 3 Sumnary Rpt (Rev 2).doc 1-25 S1059446-14264022304

Unit 3 Summary Report It is noted that LOSP initiated sequences are of higher frequency for Unit 3 than for Unit

2. This is due to the different board layouts and resulting dependencies between the units. On Unit 3, the failure of 3 DGs (and associated boards) will fail all the RHR pumps. Failure of DGs 3EA, 3EB, and 3EC fail the motive power for RHR pumps A, B, and C, and fails 480V shutdown board 3B. 480V shutdown board 3B supplies room cooling to the Unit 3 B and D RHR pumps. HPCI fails long term because of the failure of DG 3EA, which maintains the charger for long-term 250V DC power. This is the trigger for the higher frequency LOSP sequences on Unit 3. Core damage results with a failure of RCIC.

The fact that RCIC long-term operation requires both DGs A and B aggravates the situation. In contrast, no combination of 3 Unit 2 DG failures will guarantee the failure of Unit 2 RHR. Further, the Unit 2 RCIC does not depend o n the Unit 3 boards.

BFNUnit 3 Summary Rpt (Rev 2).doc 1-26 S1059446-1426-022304

  • ' b Unit 3 Summary Report SECTION 2 REFERENCES 1-1. U.S. Nuclear Regulatory Commission, "Individual Plant Examination for Severe Accident Vulnerabilities", 10CFR50.54(y, Generic Letter No. 88-20, November 23,1988.

1-2. Tennessee Valley Authority, "Browns Ferry Nuclear Plant Final Safety Analysis Report".

1-3. Kaplan, S., and Garrick, B. J., "On the Quantitative Definition of Risk," Risk Analysis, Vol. 1, pp. 11-37, March 1981.

1-4. American Nuclear Society and Institute of Electrical and Electronics Engineers, "PRA Procedures Guide; A Guide, to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants," sponsored by U.S. Nuclear Regulatory Commission and Electric Power Research Institute, NUREG/CR-2300, April 1983.

1-5. PLG, Inc., "RISKMAN - RA Workstation Software", Users Manuals I-IV, Version 5.11,1994.

1-6. Deleted 1-7. Conversation between Shawn S. Rodgers, ERIN Engineering and Research, Inc. and Xavier Polanski, Commonwealth Edison Co., May 17, 2000.

1-8. Conversation between Shawn S Rodgers, ERIN Engineering and Research, Inc. and Leo Kacanik, Niagra Mohawk, May 17, 2000.

1-9. Conversation between Shawn S. Rodgers, ERIN Engineering and Research, Inc. and Greg Kreuger, PECO, May 17,2000.

1-10. Conversation between Shawn S. Rodgers, ERIN Engineering and Research, Inc. and Gary Smith, Entergy, May 17, 2000.

1-11. Mosleh, A., et at., "Procedures for Treating Common Cause Failures in Safety and Reliability Studies," Pickard, Lowe and Garrick, Inc., prepared for U.S.

Nuclear Regulatory Commission and Electric Power Research Institute, NUREG/CR-4780, EPRI NP-5613, PLG-0547, Vols. 1-2, January 1988.

1-12. Deleted.

BFNUnit 3 Summary Rpt (Rev 2).doc 2-1 S1 059446-1426-022304

i' ;

Unit 3 Summary Report 1-13. Deleted.

1-14. U.S. Nuclear Regulatory Commission, "Common-Cause Failure Parameter Estimations", NUREG/CR-5497, October, 1998, INEELUEXT-97-01328 BFNjUnit 3 Summary Rpt (Rev 2).doc 2-2 S1059446-1426-022304

A1 . ;

Unit 3 Summary Report APPENDIX A UNIT 3 TOP RANKING SEQUENCES CONTRIBUTING TO CDF BFNUnit 3 Summary Rpt (Rev 2).doc A-1 S1 059446-14264022304

C. -

Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin 1 41 LOCHS 1.5744E-007 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCI1*HPI NLERF 4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 2 45 LOCHS 5.8312E-008 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCI1*HPI NLERF 4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*FC2*RBEF 3 135 TRAN 5.7555E-008 //SDRECF*OXF*//DWF*//MCDl*RVCO*FWHF*RCIl*HPI NLERF 4*0BDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 4 1 ISLOCA 4.6342E-008 LERF 5 815 LOSP 4.3780E-008 /OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*GF4*EPR303 NLERF

  • A3EAF*RXF*ROF*A3ECF*A3EBF*RYF*RPF*/UB4lAF*U B41BF*U842AF*UB42BF*SHUTlF*SHT2F*GB1*ABF*RSF
  • RHF*DKF'SDRECF*/UB42CF*CBBF*UB43CF*RJ3F*RK3 F*RL3F*/DlWF*/RCWF*EAF*ECF*RBCF*SW2CF*SWlCF*P CAF*DCAF*/IVOF*RVCO*CDF*EPR63*

RCLF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*R48 OF*RPAF*RPCF*RPBF*RPDF*SPF*SPRF*LPCF*OAIF*/N CDF*DWSF*RHSW3F*/NOCDF 6 869 LOSP 4.2797E-008 /OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*GF4*EPR303 NLERF

  • A3EAF*RXF*ROF*A3ECF*A3EBF*RYF*RPF*/U841AF*U B41BF*UB42AF*UB42BF*SHUTlF*SHT2F*GA1*AAF*RQF
  • REF*RMF*SDRECF*/UB42CF*CBBF*UB43CF*RJ3F*DN3 F*RK3F*RL3F*/DWF*/RCWF*EAF*ECF*RBCF*SW2AF*SW lAF*DCAF*/IVOF*RVC0*CDF*EPR63*

RCLF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*R48 OF*RPAF*RPCF*RPBF*RPDF*SPF*SPRF*LPCF*OAIF*/N CDF*RHSW3F*/NOCDF 7 722 LOSP 3.8516E-008 /OGSF*OG16F*UB43AF*UB43BF*GEl*CG2*GF4*EPR303 NLERF

  • A3EAF*RXF*ROF*A3ECF*A3EBF*RYF*RPF*/UB41AF*U B41BF*UB42AF*UB42BF*SHUTlF*SHT2F*SDRECF*/UB4 2CF*CBBF*UB43CF*RJ3F*RK3F*RL3F*/DWF*/RCWF*EA F*ECF*RBCF*DCAF*/IVOF*RVCO*CDF*EPR63*RCL1*HP LF*/FWAF*HRLF*SUFWF*HSF*CDAF*C RDF*R480F*RPAF*RPCF*RPBF*RPDF*SPF*SPRF*LPCF*

OAIF*/NCDF*RHSW3F*/NOCDF 733 LOSP 3.5735E-008 /OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*GF4*EPR303 NLERF

  • A3EAF*RXF*ROF*A3ECF*A3EBF*RYF*RPF*/UB41AF*U B41BF*UB42AF*UB42BF*SHUTlF*SHT2F*SDRECF*/UB4 BFN_Unit 3 Summary Rpt (Rev 2).doc A-2 S1 059446-1426-02304

I C. -

Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin 2CF*CBBF*UB43CF*RJ3F*RK3F*RL3F*/DWF*/RCWF*EA F*ECF*RBCF*DCAF*/IVOF*RVC0*CDF*RCI1*EPR63*HP LF*/FWAF*HRLF*SUFWF*HSF*CDAF*C RDF*R480F*RPAF*RPCF*RPBF*RPDF*SPF*SPRF*LPCF*

OAIF*/NCDF*RHSW3F*/NOCDF 9 231 TRAN 2.9538E-008 //SDRECF*OXF*//DWF*//TB1*RVCO*FWHF*RCI1*HPI4 NLERF

  • OBDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*/

NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 10 238 LOCHS 2.7697E-008 //SDRECF*OXF*//DWF*U2AP1*//IVOF*RVCO*FWHF*RC NLERF I1*HPI4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF

  • CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 11 78 LSOOU2 2.5487E-008 /OG5F*EPR301*/SDRECF*OXF*//DWF*//MCDF*RVCO*F NLERF WHF*RCI1*HPI4*OBDF*ORVD2*/FWAF*HRLF*HR6F*SUF WF*HSF*CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 12 967 LOSP 2.5480E-008 /OGSF*OG16F*UB43AF*UB43BF*GE1*GG2*GF4*GH4*EP NLERF R303*DGC1*A3EAF*RXF*ROF*A3ECF*A3EBF*RYF*RPF*

A3EDF*/UB41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SH T2F*GAF*GDF*GBF*GCF*AAF*RQF*REF*RMF*ABF*RSF*

RHF*DKF*ACF*RRF*RFF*ADF*RTF*RKF*RLF*RIF*RJF*

RNF*DLF*SDRECF*/UB42CF*CBBF*UB 43CF*RJ3F*DO3F*DN3F*RK3F*RL3F*/DWF*/RCWF*EAF

  • EBF*ECF*EDF*RBCF*SW2AF*SW1AF*SW2BF*SW1BF*SW 2CF*SW1CF*SW2DF*SW1DF*PCAF*DCAF*CADF*/OEEF*I VOF*RVCO*CDF*EPR63*RCLF*HPLF*/FWAF*HRLF*SUFW P*HSF*CDAF*CRDF*ORPF*R480F*RPAF*RPCF*U2F*RPB F*RPDF*OSPF*LPCF*OAIF*/NCDF*RH 13 43 LOCHS 2.4334E-008 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCI1*HPI NLERF 4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*RBI2 14 47 TRAN 2.4175E-008 //SDRECF*OXF*//DWF*//BVR1*RVCO*FWHF*RCI1*HPI NLERF 4*0BDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 15 476 LOSP 2.4046E-008 /OGSF*OG16F*UB43AF*UB43BF*GE1*GF2*EPR303*A3E NLERF AF*RXF*ROF*A3EBF*/UB41AF*UB41BF*UB42AF*UB42B F*SHUTlF*SHT2F*GB1*GC2*ABF*RSF*RHF*DKF*ACF*R RF*RFF*SDRECF*/UB42CF*CBBF*UB43CF*/DWF*/RCWF

  • EAF*EBF*ECF*RBCF*SW2BF*SW2CF*SW1CF*PCAF*DCA F*/OEEF*IVOF*RVCO*CDF*EPR63*RC LF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*

BFNUnit 3 Summary Rpt (Rev 2).doc A-3 S1059446-14264022304

Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin RPAF*RPCF*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*/NCDF

  • RHSW3F*/NOCDF 16 399 LOSP 2.2936E-008 /OGSF*OC16F*UB43AF*UB43BF*CEl*GH2*EPR303*A3E NLERF AF*RXF*ROF*A3EDF*/UB41AF*UB41BF*UB42AF*UB42B F*SHUTlF*SHT2F*GD1*CC2*ACF*RRF*RFF*ADF*RTF*R IF*RJF*RNF*DLF*SDRECF*/UB42CF*CBBF*UB43CF*/D WF*/RCWF*EAF*EBF*EDF*RBCF*SW2BF*SW2DF*SW1DF*

PCAF*DCAF*/OEEF*IVOF*RVCO*CDF*

EPR63*RCLF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CR DF*ORPF*RPAF*RPCF*U2F*RPBF*RPDF*OSPF*LPCF*OA IF*/NCDF*RHSW3F*/NOCDF 17 139 TRAN 2.1317E-008 //SDRECF*OXF*//DWF*//MCD1*RVCO*FWHF*RCI1*HPI NLERF 4*OBDF*ORVD2* /FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*FC2*RBEF 18 42 LOCHS 1.7494E-008 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCI1*HPI NLERF 4*0IVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*RBE4 19 32 FLTB2 1.6742E-008 //SDRECF*OXF*//DWF*//MCDF*RVCO*CDF*RCI1*HPI4 NLERF

  • ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*/NCDF*

/NOCDF*NLERFF*ELF*WWBF*WWF 20 22 LOAC 1.6020E-008 //SDRECF*OXF*//DWF*//MCDF*RVC0*CDF*RCI1*HPI4 NLERF

  • ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*/NCDF*

/NOCDF*NLERFF*ELF*WWBF*WWF 21 7 L500U2 1.5926E-008 /OG5F*/SDRECF*OXF*//DWF*//MCDF*RVCO*FWHF*RCI NLERF 1*HPI4*OBDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*

CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 22 7 LOPA 1.4261E-008 //SDRECF*OXF*//DWF*/PCAF*DCAF*/IVOF*RVCO*FWH NLERF F*RCI1*HPI4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF

  • HSF*CDAF*LCF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*W WF 23 17 LRBCCW 1.4259E-008 //SDRECF*OXF*//DWF*/RBCF*DCAF*/IVOF*RVCO*FWH NLERF F*RCI1*HPI4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF
  • HSF*CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 24 92 TRAN 1.3310E-008 //SDRECF*OXF*//DWF*//MCD1*RVCO*FWHF*OBCF*/FW NLERF AF*HSF*HXA1*HXC2*U22*HXB5*HXD7*OSPF*OSDF*OLP 2*/NCDF*/NOCDF 25 55 LOCHS 1.2680E-008 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCI1*HPI LERF 4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*OP3*IVR1*TR BFNUnit 3 Summary Rpt (Rev 2).doc A-4 S 1059446-1 426-022304

n ;

Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin F*FC3*DWIF*RBE5 26 372 TRAN 1.2149E-008 //SDRECF*OXF*//DWF*//RXSl*OSLl*/NAF*FWAF*HRL LERF F*HR6F*SUFWF*CDAF*/NCDF*/NOCDF*NLERFF*CILF*W Wl*IVR10*TR6*FCF*DWIF*RBE8 27 366 TRAN 1.2149E-008 //SDRECF*OXF*//DWF*//RXSl*OSLl*/NAF*FWAF*HRL LERF F*HR6F*SUFWF*CDAF*/NCDF*/NOCDF*NLERFF*CILF*I VR10*TR6*FCF*DWIF*RBE7 28 4 BOC 1.1795E-008 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCIF*HPI hLERF 4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*//NOCDF*NLERFF*ELF*WWBF*WWF 29 633 LOSP 1.1567E-008 /OGSF*OG16F*UB43AF*UB43BF*GEl*GG2*EPR303*A3E NLERF AF*RXF*ROF*A3ECF*RYF*RPF*/UB41AF*UB41BF*UB42 AF*UB42BF*SHUTlF*SHT2F*GA1*AAF*RQF*REF*RMF*S DRECF*/UB42CF*CBBF*UB43CF*R33F*DN3F*RK3F*RL3 F*/DWF*/RCWF*EAF*RBCF*SW2AF*SWlAF*DCAF*/IVOF

  • RVCO*CDF*EPR63*RCLF*HPLF*/FWA F*HRLF*SUFWF*HSF*CDAF*CRDF*RPAF*RPBF*CS6*LPC F*OAIF*/NCDF*/NOCDF 30 235 TRAN 1.0940E-008 //SDRECF*OXF*//DWF*//TBl*RVC0*FWHF*RCI1*HPI4 NLERF
  • OBDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*/

NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*FC2*RBEF 31 587 LOSP 1.0441E-008 /OG5F*OG16F*UB43AF*UB43BF*GEl*GC2*EPR303*A3E NLERF AF*RXF*ROF*A3ECF*RYF*RPF*/UB41AF*UB41BF*UB42 AF*UB42BF*SHUTlF*SHT2F*GBl*ABF*RSF*RHF*DKF*S DRECF*/UB42CF*CBBF*UB43CF*RJ3F*RK3F*RL3F*/DW F*/RCWF*EAF*RBCF*SW2CF*SW1CF*PCAF*DCAF*/IVOF

  • RVCO*CDF*EPR63*RCLF*HPLF*/FWA F*HRLF*SUFWF*HSF*CDAF*CRDF*RPAF*HXCF*RPBF*CS 6*LPCF*OAIF*/NCDF*/NOCDF 32 563 LOSP 1.0415E-008 /OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*EPR303*A3E NLERF AF*RXF*ROF*A3ECF*RYF*RPF*/UB41AF*UB41BF*UB42 AF*UB42BF*SHURrlF*SHT2F*SDRECF*/UB42CF*CBBF*U B43CF*Ra3F*RK3F*RL3F*/DWF*/RCWF*EAF*RBCF*DCA F*/IVOF*RVCO*CDF*EPR63*RCL1*HPLF*/FWAF*HRLF*

SUFWF*HSF*CDAF*CRDF*RPAF*RPBF*

CS6*LPCF*OAIF*/NCDF*/NOCDF 33 35 ISCRAM 1.0344E-008 //SDRECF*OXF*//DWF*//MCDl*RVC0*FWHF*RCIl*HPI NLERF 4*OBDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF BFNUnit 3 Summary Rpt (Rev 2).doc A-5 S105°446-1426-022304

'; It ;

Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin 34 242 LOCHS 1.0258E-008 //SDRECF-OXF*//DWF*U2AP1*//IVOF*RVCO*FWHF*RC NLERF I1-HPI4*OIVF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF

  • CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*FC2*R BEF 35 18 FLTB2 1.0147E-008 //SDRECF*OXF*//DWF*//MCDF*RVCO*CDF*RCI1*HPI4 NLERF
  • /FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*LCl*JC1*/NCD F*/NOCDF*NLERFF*ELF*WWBF*WWF*IVR6 36 543 TRAN 1.0125E-008 //SDRECF*OXF*//DWF*U2AP1*//MCDl*RVCO*FWHF*RC NLERF I1*HP14*OBDF*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF
  • CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 37 9 LOAC 9.7102E-009 //SDRECF*OXF*//DWF*//MCDF*RVCO*CDF*RCI1*HPI4 NLERF
  • /FWAF*HPLF*HR6F*SUFWF*HSF*CDAF*LCl*JCl*/NCD F*/NOCDF*NLERFF*ELF*WWBF*WWF*IVR6 38 569 LOSP 9.6625E-009 /OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*EPR303*A3E NLERF AF*RXF*ROF*A3ECF*RYF*RPF*/UB41AF*UB41BF*UB42 AF*UB42BF*SHUTlF*SHT2F*SDRECF*/UB42CF*CBBF*U B43CF*RJ3F*RK3F*RL3F*/DWF*/RCWF*EAF*RBCF*DCA P*/IVOF*RVCO*CDF*RCI1*EPR63*HPLF*/FWAF*HRLF*

SUFWF*HSF*CDAF*CRDF*RPAF*RPBF*

CS6*LPCF*OAIF*/NCDF*/NOCDF 39 240 LOSP 9.5933E-009 /OG5F*OS16F*UB43AF*UB43BF*GF1*GH2*EPR303*A3E NLERF BF*A3EDF*/UB41AF*UB41BF*UB42AF*UB42BF*SHUT1F

  • SHT2F*GD1*GB2*GC4*ABF*RSF*RHF*DKF*ACF*RRF*R FF*ADF*RTF*RKF*RLF*RIF*RJF*RNF*DLF*SDRECF*/U B42CF*CBEF*UB43CF*/DWF*/RCWF*EBF*ECF*EDF*RBC F*SW2BF*SW2CF*SWlCF*SW2DF*SWlD F*PCAF*DCAF*/OEEF*IVOF*RVCO*CDF*EPR63*RCLF*H PLF*/FWAP*HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*RPAF
  • RPCF*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*/NCDF*RHS W3F*/NOCDF 40 45 LRCW 9.5206E-009 //SDRECF*OXF*//DWF*/RCWF*/MCDF*RVCO*CDF*RCI1 NLERF
  • HP14*ORVD2*/FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*C RDF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF 41 516 LOSP 9.4531E-009 /OG5F*OG16F*UB43AF*UB43BF*GE1*GF2*EPR303*A3E NLERF AF*RXF*ROF*A3EBF*/UB41AF*UJB41BF*UB42AF*UB42B F*SHUTlF*SHT2F*GAl*GB2*GC4*AAF*RQF*REF*RMF*A BF*RSF*RHF*DKF*ACF*RRF*RFF*SDRECF*/UB42CF*QB BF*UB43CF*DN3F*/DWF*/RCWF*EAF*EBF*ECF*RBCF*S W2AF*SWIAF*SW2BF*SW2CF*SWlCF*P BFNUrt 3 Summary Rpt (Rev 2).doc A-6 S1059446-14264022304

- trl Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index initiator Frequency Failed and Multi-State Split Fractions Bin CAF*DCAF*/OEEF*IVOF*RVCO*CDF*EPR63*RCLF*HPLF

  • /FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*RPAF*RP CF*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*/NCDF*RHSW3F
  • /NOCDF 42 82 L500U2 9.4398E-009 /OG5F*EPR301*/SDRECF*OXF*//DWF*//tMCDF*RVCO*F NLERF WHF*RCI1*HPI4*OBDF*ORVD2*/FWAF*HRLF*HR6F*SUF WF*HSF*CDAF*/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF
  • FC2*RBEF 43 504 LOSP 9.4042E-009 /OG5F*OG16F*UB43AF*UB43BF*GEl*GF2*EPR303*A3E NLERF AF*RXF*ROF*A3EBF*/UB41AF*UB41BF*UB42AF*UB42B F*SHUTlF*SHT2F*GDl*GB2*GC4*ABF*RSF*RHF*DKF*A CF*RRF*RFF*ADF*RTF*RKF*RLF*RIF*RJF*RNF*DLF*S DRECF*/UB42CF*CBBF*UB43CP*/DWF*/RCWF*EAF*EBF
  • ECF*EDF*RBCF*SW2BF*SW2CF*SW1C F*SW2DF*PCAF*DCAF*/OEEF*IVOF*RVCO*CDF*EPR63*

RCLF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*ORP F*RPAF*RPCF*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*/NC DF*RHSW3F*/NOCDF 44 1 ELOCA 9.3900E-009 /NCDF LERF 45 429 LOSP 9.1630E-009 /OG5F*OG16F*UB43AF*UB43BF*GEl*GH2*EPR303*A3E NLERF AF*RXF*ROF*A3EDF*/UB41AF*UB4lBF*UB42AF*UB42B F*SHUTlF*SHT2F*GDl*GB2*GC4*ABF*RSF*RHF*DKF*A CF*RRF*RFF*ADF*RTF*RXF*RLF*RIF*RJF*RNF*DLF*S DRECF*/UB42CF*CBBF*UB43CF*/DWF*/RCWF*EAF*EBF

  • EDF*RBCF*SW2BF*SW2CF*SWlCF*SW 2DF*SWlDF*PCAF*DCAF*/OEEF*IVOF*RVCO*CDF*EPR6 3*RCLF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*O RPF*RPAF*RPCF*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*/

NCDF*RHSW3F*/NOCDF 46 141 LOSP 9.0940E-009 /OG5F*OG16F*UB43AF*UB43BF*EPR303*/UB41AF*UB4 NLERF lBF*UB42AF*UB42BF*SHUTlF*SHT2F*GAl*GD2*GB4*G C7*AAF*RQF*REF*RMF*ABF*RSF*RHF*DGG*RCF*DKF*A CF*RRF*RFF*ADF*RTF*CPRECF*RKF*RLF*RIF*RJF*RN F*DLF*SDRECF*OXF*/RB3F*UB42CF*CBBF*UB43CF*/P X2F*NPIIF*NH2F*DWF*/RCWF*EBF*S W2AF*SWlAF*SW2BF*SWlBF*SW2CF*SWlCF*SW2DF*PCA F*DCAF*/IVOF*RVCO*CDF*HPIF*EPR63*RCLF*/FWAF*

HRLF*SUFWF*HSF*CDAF*CRDF*HXAF*HXCF*U2F*RPBF*

OSPF*OSDF*OAIF*/NCDF*/NOCDF BFNUnit 3 Summary Rpt (Rev 2).doc A-7 S10S9446-142640223W4

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Unit 3 Summary Report Model Name: U3EPUB Master Frequency File: MFFALL Sequences for Group: ALL Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin 47 443 LOSP 9.0146E-009 /OG5F*0016F*UB43AF*UB43BF*GEl*GH2*EPR303*A3E NLERF AF*RXF*ROF*A3EDF*/UB41AF*UB41BF*UB42AF*UB42B F*SHUTlF*SHT2F*GAl*GD2*GC4*AAF*RQF*REF*RMF*A CF*RRF*RFF*ADF*RTF*RIF*RJF*RNF*DLF*SDRECF*/U B42CF* CBF*UB43CF*DN3F*/DWF*/RCWF*EAF*EBF*ED F*RBCF*SW2AF*SWlAF*SW2BF*SW2DF

  • SWIDF*PCAF*DCAF*/OEEF*IVOF*RVCO*CDF*EPR63*R CLF*HPLF*/FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*ORPF
  • RPAF*RPCF*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*/NCD F*RHSW3F*/NOCDF 48 46 LOCBS 9.0128E-009 //SDRECF*OXF*//DWF*//IVOF*RVCO*FWHF*RCI1*HPI NLERF 4*OIVF*ORVD2*/FWAF*HRLF*BR6F*SUFWF*BSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*FC2*RBI2*RB EF 49 621 LOSP 8.9982E-009 /OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*EPR303*A3E NLERF AP*RXF*ROF*A3ECF*RYF*RPF*/UB41AF*UB41BF*UB42 AP*UB42BF*SHUTlF*SHT2F*GDl*GB2*GC4*ABF*RSF*R HF*DKF*ACF*RRF*RFF*ADF*RTF*RKF*RLF*RIF*RJF*R NF*DLF*SDRECF*/UB42CF*CBBF*UB43CF*RJ3F*DO3F*

RK3F*RL3F*/DWF*/RCWF*EAF*EBF*E DF*R8CF*SW2BF*SWlBF*SW2 CF*SWlCF*SW2DF*PCAF*D CAF*/IVOF*RVCO*CDF*EPR63*RCLF*HPLF*/FWAF*HRL F*SUFWF*HSF*CDAF*CRDF*RPAF*HXCF*U2F*RPBF*HXD F*OSPF*LPCF*OAIF*/NCDF*/NOCDF 50 51 TRAN 8.9538E-009 //SDRECF*OXF*//DWF*//BVRl*RVCO*FWHF*RCIl*HPI NLERF 4*OBDF*ORVD2*/FWAF*HRLF*R6F*SUFWF* HSF*CDAF*

/NCDF*/NOCDF*NLERFF*ELF*WWBF*WWF*FC2*RBEF BFNUnit 3 Summary Rpt (Rev 2).doc A-8 S1059446-1426-022304

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