ML042330563

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TVA Systems and Analysis Browns Ferry Nuclear Plant, Probabilistic Safety Assessment Unit 3.Summary Report Rev. 1, January 2003
ML042330563
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 03/18/2004
From: Adkins G
Tennessee Valley Authority
To: Palla R
Office of Nuclear Reactor Regulation
References
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U3, RI PSA Summary Report Creation Date: 3/18/04 11:30AM From: "Adkins, Gary M." <gmadkins~tva.gov>

Created By: gmadkins~tva.gov Recipients nrc.gov owf2_po.OWFNDO RLP3 (Robert Palla) tva.gov gmmorrison CC (Morrison, George M.)

Post Office Route owf2_po.OWFNDO nrc.gov tva.gov Files Size Date & Time MESSAGE 167 03/18/04 11:30AM Unit 3 PSA Summary Report Rev 1.pdf 138510 Mime.822 191632 Options Expiration Date: None Priority: Standard Reply Requested: No Return Notification: None Concealed

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[Mic-haelMasnikT

-,RIPSA SummaryRe-port Page From: "Adkins, Gary M." <gmadkins~tva.gov>

To: "Robert Palla" <RLP3@nrc.gov>

Date: 3/18/04 11:59AM

Subject:

U3, R1 PSA Summary Report Bob, Attached is the Unit 3 PSA Summary Report.

Gary M. Adkins, P.E.

BFN License Renewal Manager Phone:(423)751-4363

<<Unit 3 PSA Summary Report Rev 1.pdf>>

CC: "Morrison, George M." <gmmorrisonetva.gov>

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

SUMMARY

REPORT Revision 1 January 2003

Unit 3 Summary Report Browns Ferry Nuclear Plant Probabilistic Safety Assessment REVISION LOG Unit 3 Summary Report Revision Description of Revision Prepared By / Checked By / Approved By I No. Date Date Date 0 Initial Issue D. Bidwell S. Rodgers D.McCamy/

1 Changes from EPU S. Rodgers D. Johnson S1329901-1396-031902

Unit 3 Summary Report TABLE OF CONTENTS Section Paqe 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 .......................... 1-16 1.4.2.3 Important Operator Actions ................................... 1-20 1.4.2.4 Important Plant Hardware Characteristics ............ 1-21 1.4.3 RESULTS FOR LARGE EARLY RELEASE FREQUENCY ....1-22 1.4.3.1 Important Plant Hardware Characteristics for Containment Performance ............................... 1-23 1.4.4 COMPARISON WITH THE 2002 BROWNS FERRY UNIT 3 PRA, REVISION 0.1-24 1.5 INSIGHTS ............... 1-24

2.0 REFERENCES

............... 2-1 APPENDIX A Unit 3 Top Ranking Sequences Contributing to Group S1329901-1396-031902

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 ...1-10 Table 1-3 Summary of the Core Damage Accident Sequence Subclasses ............... 1-14 Table 1-4 Breakdown of Core Damage Sequences in Each Frequency Range ......... 1-16 Table 1-5 Browns Ferry Unit 3 Important Operator Actions ........................................ 1-21 Table 1-6 Browns Ferry Unit 3 Important Systems ..................................................... 1-22 Table 1-7 Summary of Revised Human Error Probabilities ........................................ 1-24 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 1.

ii S1329901-1 396-031 902

Unit 3 Summary Report SECTION 1 EXECUTIVE

SUMMARY

1.1 BACKGROUND

AND OBJECTIVES This documents the performance by the Tennessee Valley Authority (TVA) in revising 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; an estimate of the large early release frequency (LERF); and 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 1 -1 S1329901-1396-031902

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.

Core damage is the loss of adequate core cooling 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-1-2 S1329901-1396-031902

Unit 3 Summary Report thirds of the active fuel becomes 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 RISKMANO 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.

1 -3 S1329901-1396-031902

Unit 3 Summary Report 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 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 1-4 S1 329901-1396-031902

Unit 3 Summary Report plant and operational features relevant to safety. The presentation describes both the core damage and large early release results.

1.4.1 TOTAL CORE DAMAGE AND LARGE EARLY RELEASE FREQUENCY 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.

1-5 S1329901-1396-031902

Unit 3 Summary Report

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Frequency I Figure 1-2 Uncertainty Curve for Browns Ferry Unit 3 Large Early Release Frequency 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 Cities, 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).

1-7 S1329901-1396-031902

Unit 3 Summary Report Table 1-1 Comparison with Other PRAs Plant l Flood Included l Mean CDF l Reference l Mean LERF l l (per reactor-year) l (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 initiated 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 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 1-8 S1329901-1396-031902

Unit 3 Summary Report 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 31.3% of CDF.

Transients with the Power Conversion System (PCS) unavailable as a result of the initiator account for 28.3% 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.

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.

1 -9 S1329901-1396-031902

Unit 3 Summary Report 35 c

UL Loss of Offsite Transients with Transient With Support Systems Internal Floods LOCAs Stuck-Open Relief Power PCS Unavailable PCS Avalable Valves Figure 1-3 Browns Ferry Unit 3 Core Damage Frequency by Initiating Event Category Table 1-2 Unit 3 Initiating Event Group Contributions to Core Damage Frequency Initiating Event Category 1 Mean CDF l Percentage of Total (per reactor-year) l Loss of Offsite Power 1.05E-06 31.3 Transients with PCS Unavailable 9.53E-07 28.3 Transients with PCS Available 8.OOE-07 23.8 Support System Failures 2.35E-07 7.0 Internal Floods 1.63E-07 4.8 LOCAs 1.01 E-07 3.0 Stuck-Open Relief Valves 5.83E-08 1.7 Total 3.4E-06 99.9 (rounding) 1-10 S1 329901-1 396-031 902

Unit 3 Summary Report Support system failure initiators (specifically, loss of plant air, loss of raw cooling water, or loss of either l&C bus 3A or 3B failures) contribute 7.0% 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 2 reactor building, in the Unit 1 or Unit 3 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 3.0% of the total CDF.

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

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:

1-11 S1329901-1396-031902

Unit 3 Summary Report Core Damage RPV Condition Containment Functional Class l Condition Loss of effective coolant inventory (includes high Intact and low pressure inventory losses) 11 Loss of effective containment pressure control, e.g., Breached or Intact heat removal Ill LOCA with loss of effective coolant inventory Intact makeup 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 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 which 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) which are defined by system operation. The interdependencies that exist between plant system operation and the core melt and radionuclide release 1-12 S1329901-1396-031902

Unit 3 Summary Report 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 1-2 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 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.

1-13 S1329901-1396-031902

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

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

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

D Accident sequences involving a loss of coolant TQUV inventory makeup in which reactor pressure has been successfully reduced to 200 psi; i.e., accident 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 II A Accident sequences involving a loss of containment TW heat removal with the RPV initially intact; core damage induced post containment failure L Accident sequences involving a loss of containment AW heat removal with the RPV breached but no initial core damage; core damage after containment failure.

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

1-14 S1329901-1396-031902

Unit 3 Summary Report Table 1-3 Summary of the Core Damage Accident Sequence Subclasses Accident Class Subclass Definition WASH-1400 Designator Designator Example g Class III A Accident sequences leading to core damage R LOCA) conditions initiated by vessel rupture where the containment integrity is not breached in the initial time phase of the accident.

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

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

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

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

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

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

V Class IV A or L except that the vent operates as N/A designed, loss of makeup occurs at some time 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.

1.4.2.2 1-15 S1329901-1396-031902

Unit 3 Summary Report 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)

> 1E-07 1 5

> 1E -08 35 28

>1E -09 378 59

> 1E-10 3179 83

  • I1E-11 22,119 95

>1E-12 [ 22,119 + Notsaved l 100 The following presents a brief description of the 20 highest-ranked sequences to the CDF.

A loss of condenser heat sink initiates the first sequence. The initiator directly causes a loss of reactor 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.

The second sequence is similar to the first. It differs in that the second sequence includes additional failures (e.g., non-minimal).

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 1-16 S1 329901-1396-031902

Unit 3 Summary Report condenser heat sink. The remainder of sequence three is identical to that of sequence one.

The fourth sequence is that of an interfacing system LOCA that results in core damage.

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

The RCIC pump also fails long term. 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 1-17 S1329901-1 396-031 902

Unit 3 Summary Report 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.

Sequence eight is similar to sequence seven with the difference being the RCIC failure is in the injection phase.

A general transient initiates sequence nine. It is similar to sequence three but the failure of feedwater is caused by the failure to trip the turbine.

Sequence ten is a non-minimal version of sequence one. The additional failure is Unit 1/2 at power.

Sequence eleven is initiated by a loss of the 500kV to the unit. The loss causes feedwater to fail. HPCI and RCIC fail. Depressurization fails and core damage follows due to lack of inventory.

Sequence twelve 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 Sequence thirteen is a non-minimal version of sequence two.

A general transient initiates sequence fourteen. It is similar to sequence three but the failure of reactor feedwater is caused by the failure of the turbine bypass valves.

Sequence fifteen 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.

1-18 S1329901-1396-031902

Unit 3 Summary Report Sequence sixteen is similar to sequence fifteen except that different combinations of diesel generators fail such that the EECW success criterion is not met. In this case, the C, D, 3EA, and 3ED diesel generators fail.

Sequence seventeen is similar to sequence three, but has different failures in the containment event tree.

Sequence eighteen is similar to sequence two, but has different failures in the containment event tree.

A flood in the turbine building initiates sequence nineteen. The flood disables feedwater. The remainder of the sequence is the familiar failure of high pressure injection with a failure to depressurize.

Sequence twenty is initiated by a loss of all condensate. This causes a loss of feedwater. HPCI and RCIC fail followed by a failure to depressurize.

Section Appendix A contains a listing of the top 50 core sequences.

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

1-19 S1329901-1396-031902

Unit 3 Summary Report Sequence Frequency  % CDF Cumulative 1 1.57E-07 4.62E-02 4.62E-02 2 5.83E-08 1.71 E-02 6.33E-02 3 5.76E-08 1.69E-02 8.03E-02 4 4.63E-08 1.36E-02 9.39E-02 5 4.38E-08 1.29E-02 1.07E-01 6 4.28E-08 1.26E-02 1.19E-01 7 3.85E-08 1.13E-02 1.31E-01 8 3.57E-08 1.05E-02 1.41 E-01 9 2.95E-08 8.68E-03 1.50E-01 10 2.77E-08 8.15E-03 1.58E-01 11 2.55E-08 7.50E-03 1.66E-01 12 2.55E-08 7.50E-03 1.73E-01 13 2.43E-08 7.15E-03 1.80E-01 14 2.42E-08 7.12E-03 1.87E-01 15 2.40E-08 7.06E-03 1.94E-01 16 2.29E-08 6.74E-03 2.01 E-01 17 2.13E-08 6.26E-03 2.07E-01 18 1.75E-08 5.15E-03 2.12E-01 19 1.67E-08 4.91 E-03 2.17E-01 20 1.60E-08 4.71 E-03 2.22E-01 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 order of their impact on the total CDF. The operator actions to recover electric power are not 1-20 S1329901-1396-031902

Unit 3 Summary Report 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.

Table 1-5 Browns Ferry Unit 3 Important Operator Actions Operator Action Top l Importance Event/Split Fraction Depressurize to Allow Low Pressure Injection ORVD2 43.2 Open the Hardened Wetwell Vent OLP2/314 9.2 Operator Aligns Suppression Pool Cooling OsP 4.8 Align Alternate Injection to Reactor Vessel via the Unit 3 to Unit 2 RHR U22 3.5 Crosstie*

  • The importance of the split fraction U22 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.

1-21 S1329901-1396-031902

Unit 3 Summary Report Table 1-6 Browns Ferry Unit 3 Important Systems System I  % CDF HPCI 55 RCIC 55 Diesel Generators 28 Feedwater/Condensate 13 Main Steam 10 RHR 9 RPS 7 Core Spray 3 RHRSW to RHR Loop II 2 Control Rod Drive 2 RHRSW to RHR Loop I 1 Standby Liquid Control 1 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 LARGE EARLY RELEASE FREQUENCY 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, 1-22 S1329901-1396-031902

Unit 3 Summary Report 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 result 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 group 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.

1.4.4 1-23 S1329901-1396-031902

Unit 3 Summary Report 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 1 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 timing. See Table 1-7.

Table 1-7 Summary of Revised Human Error Probabilities Operator CPPU Current Notes Action HEP HEP 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 7.71 E-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 current 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.

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 1-24 S1329901-1396-031902

Unit 3 Summary Report 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 on the Unit 3 boards.

1-25 S1329901-1396-031902

Unit 3 Summary Report SECTION 2 REFERENCES 1-1. U.S. Nuclear Regulatory Commission, "Individual Plant Examination for Severe Accident Vulnerabilities", 10CFR50.54(?), 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 - PRA Workstation Software", Users Manuals l-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, Niagara 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 al., "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.

2-1 S1329901-1396-031902

Unit 3 Summary Report 1-13. Deleted.

1-14. U.S. Nuclear Regulatory Commission, "Common-Cause Failure Parameter Estimations", NUREG/CR-5497, October, 1998, INEEL/EXT-97-01328 2-2 S1329901-1396-031902

Unit 3 Summary Report APPENDIX A UNIT 3 TOP RANKING SEQUENCES CONTRIBUTING TO CDF S1 329901-1396-031902

Unit 3 Summary Report 1 LOCHS 222 1.5744E-007 SDRECF*OXF*DWF*IVOF*RVC0*FilHF*RCIl*HPI4*OIVF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NoCDF-NLERFF*ELF*WWB F*wwF 2 LOCHS 226 5.8312E-008 SDRECF*OXF*DWF*IVOF*RVCOFnHF*RCI1*HPI4*OIVF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*WWB F*WWF*FC2*RBEF 3 TRAN 632 5.7555E-008 SDRECF*OXF*DWlF*MCD1*RVC0*FWHF*RCI1*HPI4*OBDF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF-NCDF*NOCDF*NLERFF-ELF*WWB F*WWF 4 ISLOCA 1 4.6342E-008 LERF 5 LOSP 5877 4.3780E-008 OG5F*OG16F*UB43AF*UB43BF*GE1*GG2*GF4*EPR303*A3EAF*RXF NLERF

  • ROF*A3ECF*A3EBF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*

SHUTlF*SHT2F*GB1*ABF*RSF*RHF*DKF*SDRECF*UB42CF*CBBF*U B43CF*RJ3F*RK3F*RL3F*DWF*RCWF*EAF*ECF*RBCF*SW2CF*SW1C F*PCAF*DCAF*IVOF*RVCO*CDF*EPR63*RCLF*H PLF*FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*R480F*RPAF*RPCF*RPB F*RPDF*SPF*SPRF*LPCF*OAIF*NCDF*DDWSF*RHSW3F*NOCDF 6 LOSP 6199 4.2797E-008 OG5F*OG16F*UB43AF*UB43BF*GE1*GG2*GF4*EPR303*A3EAF*RXF NLERF

  • ROF*A3ECF*A3EBF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*

SHUTlF*SHT2F*GA1*AAF*RQF*REF*RMF*SDRECF*UB42CF*CBBF*U B43CF*RJ3F*DN3F*RK3F*RL3F^DWF*RCWF*EAF*ECF-RBCF*SW2AF

  • SWlAF*DCAF*IVOF*RVCO*CDF-EPR63*RCLF*H PLF*FWAF*HRLF*SUFWF*HSF*CDAF^CRDF-R480F*RPAF*RPCF*RPB F*RPDF*SPF*SPRF*LPCF*OAIF*NCDF*RHSW3F*NOCDF 7 LOSP 5332 3.8516E-008 OG5F*OG16F*UB43AF*UB43BF*GEl*GG2*GF4*EPR303*A3EAF*RXF NLERF
  • ROF*A3ECF*A3EBF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*

SHUTlF*SHT2F*SDRECF*UB42CF*CBBF*UB43CF*RJ3F*RK3F*RL3F

  • DWF*RCWF*EAF*ECF*RBCF*DCAF*IVOF*RVCO*CDF*EPR63*RCL1*

HPLF*FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*R48 OF*RPAF*RPCF*RPBF*RPDF*SPF*SPRF*LPCF*OAIF*NCDF*RHSW3F

  • NOCDF 8 LOSP 5386 3.5735E-008 OG5F*OG16F*UB43AF*UB43BF*GE1*GG2*GF4*EPR303*A3EAF*RXF NLERF
  • ROF*A3ECF*A3EBF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*

SHUTlF*SHT2F*SDRECF*UB42CF*CBBF*UB43CF*RJ3F*RK3F*RL3F

  • DWF*RCWF*EAF*ECF*RBCF*DCAF*IVOF*RVCO*CDF*RCI1*EPR63*

HPLF*FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*R48 OF*RPAF*RPCF*RPBF*RPDF*SPF*SPRF-LPCF*OAIF*NCDF*RHSW3F

  • NOCDF 9 TRAN 1016 2.9538E-008 SDRECF*OXF*DWF*TB1*RVC0*FWHF*RCI1*HPI4*OBDF*ORVD2*FWA NLERF F*HRLF*HR6F*SUFWF*HSF*CDAF-NCDF*NOCDF*NLERFF*ELF*WWBF
  • WWF 10 LOCHS 1245 2.7697E-008 SDRECF*OXF*DWF*U2AP1*IVOF*RVCO*FWHF*RCI1*HPI4*OIVF*OR NLERF VD2*FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*E LF*WWBF*WWF 11 L500U2 497 2.5487E-008 OG5F*EPR301*SDRECF*OXF*DWF*MCDF*RVC0*FWHF*RCI1*HPI4*0 NLERF BDF*ORVD2*FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NL S1329901-1396-031902

Unit 3 Summary Report ERFF*ELF*WWBF*WWF 12 LOSP 6729 2.5480E-008 OG5F*OG16F*UB43AF*UB43BF*GE1*GG2*GF4*GH4*EPR303*DGC1* NLERF A3EAF*RXF*ROF*A3ECF*A3EBF*RYF*RPF*A3EDF^UB41AF*UB41BF

  • UB42AF*UB42BF*SHUT1F-SHT2F*GAF*GDF*GBF*GCF*AAF*RQF*R EF*RMF*ABF*RSF*RHF*DKF*ACF*RRF*RFF*ADF*RTF*RKF*RLF*RI F*RJF*RNF*DLF^SDRECF*UB42CF*CBBF*UB43C F*RJ3F*DO3F*DN3F*RK3F^RL3F^DWF*RCWF^EAF*EBF*ECF*EDF*R BCF*SW2AF*SWlAF*SW2BF^SW1BF*SW2CF*SWlCF*SW2DF*SWlDF*P CAF*DCAF*CADF*OEEF*IVOF*RVC0*CDF*EPR63*RCLF*HPLF*FWAF
  • HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*R480F*RPAF*RPCF*U2F*RP BF*RPDF*OSPF*LPCF*OAIF*NCDF*RHSW3F*NOC 14 LOCHS 224 2.4334E-008 SDRECF*OXF*DWF*IVOF*RVCOFilHF*RCI1*HPI4*OIVF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF^NOCDF*NLERFF*ELF*WWB F*WWF*RBI2 15 TRAN 262 2.4175E-008 SDRECF*OXF*DWF*BVR1lRVC0*FWHF*RCI1*HPI4*OBDF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*WWB F*WWF 16 LOSP 3970 2.4046E-008 OG5F*OG16F*UB43AF*UB43BF^GE1*GF2*EPR303*A3EAF^RXF^ROF NLERF
  • A3EBF*UB41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SHT2F*GB1G C2*ABF*RSF*RHF*DKF*ACF*RRF*RFF*SDRECF*UB42CF*C8BF*UB4 3CF*DWF*RCWF*EAF*EBF*ECF*RBCF*SW2BF^SW2CF*SWlCF*PCAF*

DCAF*OEEF*IVOF*RVCO*CDF*EPR63*RCLF*HPL F*FWAF*HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*RPAF*RPCF*U2F*RP BF*RPDF*OSPF*LPCF^OAIF*NCDF*RHSW3F*NOCDF 17 LOSP 3515 2.2936E-008 OG5F*OG16F*UB43AF*UB43BF*GE1*GH2*EPR303^A3EAF^RXF*ROF NLERF

  • A3EDF*UB41AF*UB41BF*UB42AF^UB42BF*SHUTlF*SHT2F*GD1*G C2*ACF*RRF*RFF*ADF*RTF*RIF*RJF*RNF*DLF*SDRECF'UB42CF*

CBBF*UB43CF*DWF*RCWF*EAF*EBF*EDF*RBCF*SW2BF*SW2DF*SW1 DF*PCAF*DCAF*OEEF*IVOF*RVCO*CDF*EPR63*

RCLF*HPLF*FWAF*HRLF*SUFWF^HSF*CDAF*CRDF^ORPF*RPAF^RPC F*U2F*RPBF*RPDF*OSPF*LPCF*OAIF*NCDF*RHSW3F'NOCDF 18 TRAN 636 2.1317E-008 SDRECF*OXF*DWF*MCDl*RVC0*FWHF*RCI1*HPI4*OBDF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*Ww8B F*WWF^FC2*RBEF 19 LOCHS 223 1.7494E-008 SDRECF*OXF*DWF*IVOF*RVCO*FWHF*RCI1*HPI4*OIVF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF^WWB F*WWF*RBE4 20 FLTB2 168 1.6742E-008 SDRECF*OXF*DWF*MCDF*RVCO*CDF*RCI1lHPI4*ORVD2*FWAF*HRL NLERF F*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF^WWBF*WWF 21 LOAC 124 1.6020E-008 SDRECF*OXF*DWF*MCDF*RVCO*CDF*RCI1^HPI4*ORVD2*FWAF*HRL NLERF F*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*WWBF*WWF 22 L50OU2 38 1.5926E-008 OG5F*SDRECF*OXF*DWF*MCDF*RVC0*FWHF*RCI1*HPI4*OBDF*ORV NLERF D2^FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*EL F*WWBF*WWJF 23 LOPA 40 1.4261E-008 SDRECF*OXF*DWF*PCAF*DCAF*IVOF*RVCO*FWHF*RCI1*HP14*OIV NLERF F*ORVD2*FWAF*HRLF*HR6F*SUFWF*HSF*CDAF^LCF^NCDF*NOCDF*

NLERFF^ELF*VIWBF*WWF 24 LRBCCW 83 1.4259E-008 SDRECF*OXF*DWF*RBCF*DCAF*IVOF*RVC0*FWHF*RCI1*HPI4*OIV NLERF S1329901-1396-031902

Unit 3 Summary Report F*ORVD2*FWAF*HRLF*HR6F'SUFWF'HSF*CDAF*NCDF*NOCDF^NLER FF*ELF'WWBF*WWF 25 TRAN 433 1.3310E-008 SDRECF*OXF*DWF*MCD1*RVCO*FWHF^OBCF'FWAF*HSF*HXA1*HXC2 NLERF

'U22*HXB5*HXD7*OSPF*OSDF*OLP2*NCDF*NOCDF 26 LOCHS 244 1.2680E-008 SDRECF*OXF*DWF*IVOF*RVC0*FWHF*RCI1*HPI4*OIVF*ORVD2'FW LERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*WWB F*WWF*OP3*IVR1*TRF*FC3*DWIF*RBE5 27 TRAN 1773 1.2149E-008 SDRECF*OXF*DWF*RXS1*OSL1*NAF*FWAF*HRLF*HR6F*SUFWF*CDA LERF F*NCDF*NOCDF*NLERFF*CILF*IVR10*TR6*FCF*DWIF*RBE7 28 TRAN 1782 1.2149E-008 SDRECF*OXF*DWF*RXS1*OSL1*NAF*FWAF*HRLF*HR6F*SUFWF*CDA LERF F'NCDF*NOCDF*NLERFF*CILF*WW1*IVR10*TR6*FCF*DWIF*RBE8 29 BOC 18 1.1795E-008 SDRECF*OXF*DWF*IVOF*RVCO*FWHF*RCIF*HPI4*OIVF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*WWB F*WWF 30 LOSP 4866 1.1567E-008 OG5F*OG16F*UB43AF*UB43BF*GE1*GG2*EPR303*A3EAF*RXF*ROF NLERF

  • A3ECF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SHT 2F*GA1*AAF*RQF*REF*RMF*SDRECF-UB42CF'CBBF*UB43CF*RJ3F
  • DN3F*RK3F*RL3F*DWF*RCWF*EAF*RBCF*SW2AF'SWlAF*DCAF'IV OF'RVCO*CDF*EPR63*RCLF*HPLF*FWAF*HRLF*

SUFWF*HSF*CDAF*CRDF*RPAF*RPBF*CS6*LPCF*OAIF*NCDF*NOCD 31 TRAN 1020 1.0940E-008 SDRECF*OXF*DWF*TB1*RVCO*FWHF*RCIl*HPI4*OBDF*ORVD2*FWA NLERF F*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*ELF*WWBF

  • WWF*FC2*RBEF 32 LOSP 4618 1.0441E-008 OG5F'OG16F'UB43AF*U843BF*GEl*GG2*EPR303*A3EAF'RXF*ROF NLERF
  • A3ECF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SHT 2F*GB1*ABF*RSF*RHF*DKF*SDRECF'UB42CF*CBBF*UB43CF'RJ3F
  • RK3F*RL3F*DWF*RCWF*EAF*RBCF*SW2CF*SW1CF*PCAF*DCAF*IV OF*RVCO*CDF*EPR63*RCLF*HPLF*FWAF*HRLF*

SUFWF*HSF*CDAF*CRDF*RPAF*HXCF*RPBF*CS6*LPCF*OAIF*NCDF

  • NOCDF 33 LOSP 4483 1.0415E-008 OG5F*OG16F*UB43AF*UB43BF'GE1*GG2'EPR303'A3EAF'RXF'ROF NLERF
  • A3ECF*RYF*RPF*UB41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SHT 2F*SDRECF*UB42CF*CBBF'UB43CF*RJ3F*RK3F*RL3F'*DWF*RCWF*

EAF*RBCF*DCAF*IVOF*RVC0*CDF*EPR63'RCL1*HPLF*FWAF*HRLF

  • SUFWF*HSF*CDAF*CRDF*RPAF*RPBF*CS6*LPC F*OAIF*NCDF*NOCDF 34 ISCRAM 201 1.0344E-008 SDRECF*OXF*DWF*MCD1*RVC0*FWHF*RCI1*HPI4*OBDF*ORVD2*FW NLERF AF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF'ELF*WWB F*WWF 35 LOCHS 1249 1.0258E-008 SDRECF*OXF*DWF*U2AP1*IVOF*RVC0'FWHF'RCI1*HPI4'OIVF*OR NLERF VD2*FWAF*HRLF'HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*E LF*WWBF*WWF*FC2*RBEF 36 FLTB2 104 1.0147E-008 SDRECF*OXF*DWF*MCDF*RVCO*CDF*RCI1*HPI4*FWAF*HRLF*HR6F NLERF
  • SUFWF*HSF*CDAF*LC1*JCl*NCDF*NOCDF*NLERFF*ELF*WWBF*'W F*IVR6 37 TRAN 3078 1.0125E-008 SDRECF*OXF*DWF*U2AP1*MCD1*RVC0*FWHF*RCI1*HPI4*OBDF*OR NLERF VD2*FWAF*HRLF*HR6F*SUFWF*HSF*CDAF*NCDF*NOCDF*NLERFF*E LF*WWBF*WWF S1329901-1396-031902

Unit 3 Summary Report 38 LOAC 60 9.7102E-009 SDRECF*OXF'DWF*MCDF*RVCO*CDF-RCI1'HPI4-FWAF-HRLF'HR6F NLERF

  • SUFWF*HSF*CDAF*LC1*JC1*NCDF*NOCDF-NLERFF*ELFWWJBF*WW F*IVR6 39 LOSP 4502 9.6625E-009 OG5F*OG16F-UB43AF*UB43BF*GElGG2*EPR303*A3EAF*RXF*ROF NLERF
  • A3ECF*RYF*RPF*UB41AF-UB41BF-UB42AF*UB42BF*SHUT1F*SHT 2F*SDRECF*UB42CF*CBBF*UB43CF*RJ3F*RK3F*RL3F*DWF*RCWF*

EAF*RBCF*DCAF*IVOF*RVCO*CDF*RCI1*EPR63*HPLF*FWAF-HRLF

  • SUFWF*HSF*CDAF*CRDF*RPAF*RPBF*CS6*LPC F*OAIF*NCDF*NOCDF 40 LOSP 2167 9.5933E-009 OG5F*OG16F*UB43AF*UB43BF*GFl*GH2*EPR303*A3EBF-A3EDF-U NLERF B41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SHT2F*GD1*GB2*GC4*A BF-RSF*RHF*DKF*ACF'RRF*RFF*ADF*RTF-RKF*RLF*RIF*RJF*RN F*DLF*SDRECF*UB42CF*CBBF*UB43CF-DWF*RCWF*EBF*ECF*EDF*

RBCF*SW2BF-SW2CF*SWlCF*SW2DF*SWlDF*PCA F*DCAF*OEEF*IVOF*RVC0*CDF*EPR63*RCLF*HPLF*FWAF*HRLF*S UFWF*HSF*CDAF*CRDF-ORPF*RPAF*RPCF*U2F*RPBF*RPDF*OSPF*

LPCF*OAIF*NCDF*RHSW3F*NOCDF 41 LRCW 247 9.5206E-009 SDRECF*OXF'DWF*RCWF*MCDF*RVCO*CDF-RCI1*HPI4*ORVD2*FWA NLERF F*HRLF*HR6F*SUFWF*HSF-CDAF*CRDF*NCDF*NOCDF*NLERFF*ELF

  • WWBF*WWF 42 LOSP 4194 9.4531E-009 OG5F'OG16F-UB43AF*UB43BF*GE1lGF2*EPR303*A3EAF*RXF*ROF NLERF
  • A3EBF*UB41AF*UB41BF*UB42AF*UB42BF*SHUTlF*SHT2F*GA1*G B2*GC4*AAF*RQF*REF*RMF*ABF*RSF*RHF*DKF*ACF*RRF*RFF*SD RECF*UB42CF*CBBF*UB43CF*DN3F*DWF*RCWF*EAF*EBF*ECF*RBC F*SW2AF*SWlAF*SW2BF*SW2CF*SWlCF*PCAF*D CAF*OEEF*IVOF*RVCO*CDF*EPR63*RCLF*HPLF*FWAF*HRLF*SUFW F*HSF*CDAF*CRDF*ORPF*RPAF*RPCF*U2F*RPBF*RPDF-OSPF*LPC F*OAIF*NCDF*RHSW3F*NOCDF 43 L50OU2 501 9.4398E-009 OG5F*EPR301*SDRECF*OXF*DWF*MCDF*RVCO*FWHF*RCI1*HPI4*0 NLERF BDF*ORVD2*FWAF*HRLF*HR6F*SUFWF*HSF*CDAF'NCDF*NOCDF*NL ERFF*ELF*WWBF*WWF*FC2*RBEF 44 LOSP 4128 9.4042E-009 OGSF*OG16F*UB43AF*UB43BF*GEl*GF2*EPR303*A3EAF*RXF*ROF NLERF
  • A3EBF*UB41AF*UB41BF*UB42AF*UB42BF-SHUTlF*SHT2F*GD1*G B2*GC4*ABF*RSF*RHF*DKF*ACF*RRF-RFF*ADF*RTF*RKF*RLF*RI F-RJF*RNF*DLF*SDRECF*UB42CF-CBBF*UB43CF*DWF*RCWF*EAF-EBF*ECF-EDF*RBCF*SW2BF*SW2CF*SWlCF*SW2 DF*PCAF*DCAF*OEEF*IVOF*RVC0*CDF*EPR63*RCLF*HPLF*FWAF*

HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*RPAF*RPCF*U2F*RPBF*RPDF

  • OSPF*LPCF*OAIF*NCDF*RHSW3F*NOCDF 45 ELOCA 1 9.3900E-009 NCDF LERF 50 LOSP 3671 9.1630E-009 OG5F*OG16F*UB43AF*UB43BF*GE1*GH2*EPR303*A3EAF*RXF*ROF NLERF
  • A3EDF*UB41AF*UB41BF'UB42AF*UB42BF*SHUTlF*SHT2F-GD1*G B2*GC4*ABF*RSF*RHF*DKF*ACF*RRF*RFF*ADF*RTF*RKF*RLF*RI F*RJF*RNF*DLF*SDRECF*UB42CF*CBBF*UB43CF*DWF*RCWF*EAF*

EBF'EDF*RBCF*SW2BF*SW2CF*SWlCF-SW2DF-S WlDF^PCAF-DCAF-OEEF*IVOF*RVC0*CDF*EPR63*RCLF-HPLF*FWA F*HRLF*SUFWF*HSF*CDAF*CRDF*ORPF*RPAF*RPCF*U2F*RPBF*RP DF*OSPF*LPCF*OAIF*NCDF*RHSW3F*NOCDF S1329901-1396-031902

Unit 3 Summary Report S1329901-1396-031902

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