ML042860080

Revision 2, Probabilistic Safety Assessment, Summary Report.

ML042860080
Person / Time
Site:	Browns Ferry
Issue date:	08/31/2004
From:	Rodgers S ABS Consulting
To:	Office of Nuclear Reactor Regulation
References
Download: ML042860080 (40)

Text

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___i TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT I PROBABILISTIC SAFETY ASSESSMENT UNIT 1

SUMMARY

REPORT Revision 2

• R-A Consulting cS RISK CVOISULTING DIVISION

TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT 1 PROBABILISTIC SAFETY ASSESSMENT UNIT I

SUMMARY

REPORT Revision 2 Prepared by:

XABS;Consulting FrSK CONSVUCrI?4 VIVSIONi 300 Commerce Drive, Suite 200 Irvine, CA 92602-1300 (714) 734-4242 August 2004

i i Unit I Summary Report Browns Ferry Nuclear Plant Probabilistic Safety Assessment REVISION LOG Unit I Summary Report Revision Description of Revision Prepared By I Checked By I Approved By I No. Date Date Date 0 Initial Issue S. Rodgers D. Johnson W. D. Crouch 1 Revised model - Review comments incorporated S. Rodgers D. Johnson W. D. Crouch 2 Corrections to Table 3-7 and editorial changes S. Rodgers D.Johnson W. D. Crouch iii

Unit 1 Summary Report TABLE OF CONTENTS Section Page SECTION 1 EXECUTIVE

SUMMARY

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

1.1 BACKGROUND

AND OBJECTIVES .................................................... 1-1 1.2

SUMMARY

OF MAJOR FINDINGS .................................................... 1-3 SECTION 2 OVERALL METHODOLOGY .................................................... 2-1 SECTION 3 RESULTS

SUMMARY

AND DISCUSSION .................................................... 3-1 3.1 CONTRIBUTORS TO TOTAL CORE DAMAGE FREQUENCY .................................. 3-1 3.1.1 Important Core Damage Sequence Groups .................................................... 3-1 3.1.2 Analysis of Individual Sequences .................................................... 3-10 3.1.3 Important OperatorActions .................................................... 3-12 3.1.4 Important Plant Hardware Characteristics .................................................... 3-14 3.2 RESULTS FOR LARGE EARLY RELEASE FREQUENCY ....................................... 3-14 3.3 IMPORTANT ASSUMPTIONS .................................................... 3-16 3.4 SENSITIVITY EVALUATIONS .................................................... 3-17 3.5 INSIGHTS .................................................... 3-18 SECTION 4 REFERENCES .................................................... 4-1 APPENDIX A Unit I Top Ranking Sequences Contributing to CDF iv

Unit I Summary Report LIST OF TABLES Table 1-1 Core Damage Frequency Data Uncertainty Characteristics ...................................... 1-3 Table 1-2 Large Early Release Frequency Data Uncertainty Characteristics ............................ 1-3 Table 1-3 Comparison with Other PRAs ........................................................ 1-4 Table 3-1 Initiating Event Group Contributions to CDF ........................................................ 3-4 Table 3-2 Core Damage Accident Sequence Classes ........................................................ 3-5 Table 3-3 Core Damage Accident Sequence Subclasses.........................................................3-7 Table 3-4 Level 1 Plant Damage States ........................................................ 3-9 Table 3-5 Core Damage Sequences in Each Frequency Range ............................................. 3-10 Table 3-6 Contribution to CDF of the Top Twenty Sequences ................................................ 3-12 Table 3-7 Operator Actions Important to CDF ........................................................ 3-13 Table 3-8 Systems Important to CDF ........................................................ 3-14 Table 3-9 Initiating Event Contribution to LERF ........................................................ 3-15 LIST OF FIGURES Figure 3-1 Core Damage Frequency by Initiating Event Category ............................................ 3-3 v

I it Unit I Summary Report SECTION 1 EXECUTIVE

1.1 BACKGROUND

AND OBJECTIVES This documents the performance by the Tennessee Valley Authority (TVA) in the Unit 1 PSA.

An integrated team of engineers and specialists from TVA, Bechtel Power Corporation (BPC) and ABS Consulting (ABSG) performed this study. The study depended heavily on information and guidance from TVA. The TVA organizations included Unit 1 Engineering and Operations personnel in general, plant EPU personnel, the training simulator, maintenance personnel, and personnel involved in surveillance and testing. The information from interactions with TVA included:

• EOI guidance and interpretation

• Statistics on systems unavailability for Units 2 and 3 (projected to Unit 1)

• Comprehensive list of surveillance and testing requirements that affect system unavailability

• EPU design criteria and analyses

• Operator performance and sequence timing

• Specific Unit 1 information and insights BPC administered the contract for the job and:

• Provided PRA expertise and independent review.

• Identified and made available key reference material such as flow diagrams, electrical one lines and schematics, DCNs and DCAs for Unit 1.

• Acted as liaison between TVA and ABSG.

• Provided the changes to the electrical systems and equipment dependencies.

ABS Consulting performed the technical analyses.

TVA's objective was to provide a Unit 1 PSA based on restart conditions. A starting point for the analysis was the existing Unit 2 and 3 analyses.

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i I Unit 1 Summary Report TVA's requirements for the analysis included:

1. Follow guidelines established by the ASME-RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications for a Category II assessment with the expectation that the Browns Ferry Unit 1 PSA will be technically acceptable and comply with the same requirements of DG-1122 and the associated SRP 19.1.

2. Incorporate all applicable design changes planned for implementation up to Unit 1 restart.

3. The scope will evaluate Unit 1 in an uprated condition (120% of Original Rated Thermal Power with 30 psi reactor dome pressure increase) operation on a 24 month fuel cycles with the same operating flexibility options (e.g., Single Loop Operation, Increased Core Flow, etc) as is licensed for Units 2/3.

4. The results shall be presented as stand alone Unit 1 specific numbers rather than as differences between Units 2/3.

5. Fully integrate the model (fault trees/event trees).

6. Calculations to be prepared in accordance with TVA procedure NEDP 2 'Design Calculation Process Control."

7. Basic event identification must include the Browns Ferry Unit 1 UNID designation as applicable to UNID identified equipment.

The anticipated configuration at Unit 1 restart for each system is described by a combination of the existing plant documentation and a Design Change Notification (DCN) packages. The existing plant documentation along with each issued Unit 1 DCN through the design freeze date of February 6, 2003, was reviewed by the staff performing the PSA for Unit 1 to ensure that the Unit I system models represent the anticipated configuration. Other available Unit I DCNs in preliminary stages of development were reviewed during the initial preparation of the Unit 1 PSA system notebooks to evaluate their affects on the Unit I system models. If the Unit 1 DCN was not available at the time of this review, the Unit 2 system configuration was assumed applicable and is explicitly noted in the individual system analysis notebooks. Unit 1 procedures have not been developed, so the PSA model assumed the Unit 2 procedures are applicable.

Bechtel, ABSG, and TVA personnel participated in a plant walkdown to confirm that the analyzed systems, structures, and components correctly reflected the as-built plant. Also, spatial and environmental aspects of the plant design were observed during the walkdown.

The purpose of this summary is to present the results of the PSA on Browns Ferry Unit 1.

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

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Unit I Summary Report 1.2

SUMMARY

OF MAJOR FINDINGS The major findings of the Browns Ferry Unit 1 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.

The total CDF for Browns Ferry Unit I is 1.86 x 106 per reactor-year. The results for CDF were developed in terms of a mean point estimate. The CDF data uncertainty curve is characterized in Table 1-1.

Table 1-1 Core Damage Frequency Data Uncertainty Characteristics The total Large Early Release Frequency (LERF) for Browns Ferry Unit 1 is 1.87 x 107 per reactor year. The results for LERF were developed in terms of a mean point estimate. The LERF data uncertainty curve is shown in Table 1-2.

Table 1-2 Large Early Release Frequency Data Uncertainty Characteristics A comparison of this study with other PSAs on other plants that used similar methods, databases, and work scopes is given in Table 1-3. The calculated mean CDF for Browns Ferry Unit 1 is of the same order of magnitude as Browns Ferry Units 2 and 3, 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).

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Unit 1 Summary Report Table 1-3 Comparison with Other PRAs Plat Flood Mean CDF Mean LERF Reference an Included (per year) (per year)

Quad Cities Yes 4.6E-6 3.3E-6 7 Nine Mile Point Unit 2* Yes 5.7E-5 1.6E-6 8 Browns Ferry Unit 1 Yes 1.86E-6 1.87E-7 This Study Browns Ferry Unit 2 No 2.7E-06 4.OE-7 Unit 2 Summary (EPU)** Rev 1 Browns Ferry Unit 3 No 3.4E-06 4.5E-7 Unit 3 Summary (EPU)- Rev I Peach Bottom Unit 2 No 4.5E-6 Not Updated 9 Grand Gulf Unit 1 No 5.5E-6 Not Updated 10

• Includes external events. **Unit 1 not operating.

Factors that contribute to the results for Browns Ferry Unit 1 are summarized below:

• Modeling improvements have been made in a number of areas. Among these are battery board failures, HPCI/RCIC common cause, and human error probabilities for emergency depressurization.

• A number of high-ranking sequences are characterized by an initial loss of the power conversion system followed by the random failure of HPCI and RCIC with a failure to emergency depressurize. This is true even considering the modeling improvements noted above.

• Transients and ATWS sequences are among the most important sequences. This is a reflection of the low overall CDF.

• Given a loss of offsite power and the conditions that all four Unit 1/Unit 2 diesels are unavailable but all four Unit 3 diesels are available, it is assumed that 2 of the 4 available Unit 3 diesels will be aligned to service Unit 1 and Unit 2.

• 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. However, enhanced CRD may be sufficient to prevent vessel melt through.

This has not been credited but offers potential for reducing the large early release frequency estimate.

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Unit 1 Summary Report

• The Thermal Hydraulic Analyses performed in support of the Unit 1 PSA and EPU indicated that a single RHR pump and Heat exchanger is sufficient for suppression pool cooling for all Power Conversion System (PCS) isolation events.

• 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 11). The common cause event data was taken from the NRC database (Reference 14). Common cause estimates not screened were taken from NUREG/CR-5497.

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t 1 Unit 1Summary Report SECTION 2 OVERALL METHODOLOGY The Browns Ferry Unit 1 PSA is founded on a scenario-based definition of risk (Reference 4).

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 systematically developed to account for design and operating features specific to Browns Ferry Unit 1.

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 I 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 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. 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 ASME guidelines (Reference 1) and the NRC guidance on the standard (Reference 2).

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 systematically developed from the point of initiation to termination. A series of event trees systematically identifies 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 software system (Reference 5) was used to construct effectively a single set of a linked event trees that can be used for estimating either CDF or LERF. The sequence analyses start with an initiating event and terminate in end states for the Level 1 analysis or the Level 2/LERF analysis. If the Level I analysis is chosen, the end states are plant damage 2-1

Unit 1 Summary Report states (PDS), which are categories describing the status of reactivity, vessel integrity, inventory control, and containment integrity at the time of core damage. These PDSs serve the purpose of transferring information from the Level 1 to Level 2. If the Level 2ILERF analysis is chosen, the end states are LERF or no LERF. In both Level 1 and Level2/LERF the sum of these 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 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 developed from a detailed examination of the plant systems 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. Both the Level 1 analysis and Level 2/LERF analysis utilize the same set of linked event trees.

The systematic, structured approach 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.

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7 I Unit I Summary Report SECTION 3 RESULTS

SUMMARY

AND DISCUSSION 3.1 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.

3.1.1 Important Core Damaqe Sequence Groups Transients with the PCS not disabled as a result of the initiator contribute 29.4% 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.

Transients with failure of reactor scram (ATWS) scenarios contribute 16.9% to the total CDF.

Note that ATWS events are modeled as initiating events in the Unit I PSA. The Unit 2 and Unit 3 PSAs do not model ATWS as initiating events but question the reactor protection system in the event trees. The approach taken for Unit I was to allow distinct event trees for ATWS and non-ATWS events.

Transients with the PCS unavailable as a result of the initiator account for 16.0% 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.

The LOSP initiators include station blackout sequences (failure of all diesel generators) and non-station 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 11.7% of CDF.

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 11.6% to the total CDF.

Scenarios initiated by internal floods contribute 6.1% 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.

LOCAs (other than interfacing systems LOCAs) account for 4.8% of the total CDF.

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 2.7% of the total CDF.

Scenarios initiated by the inadvertent opening of a relief valve contribute 0.8% of the core damage frequency.

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Unit 1 Summary Report 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 3-1 and Table 3-1 for major initiating event categories.

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Unit I Summary Report 30%

25%

20%

5% -15 10%

5%

0%

Transients ATWS Transients Total Loss of Support Internal LOCAs Interfacing Stuck-Open with PCS Initiators with PCS Offsite Power Systems Flooding Systems Relief Valves Available Unavailable LOCAs Figure 3-1 Core Damage Frequency by Initiating Event Category 3-3

Unit I Summary Report Table 3-1 Initiating Event Group Contributions to CDF Initiating Event Category Mean CDF (per Percentage of Total reactor-year)

Transients with PCS Available 5.48E-07 29.4%

ATWS 3.14E-07 16.9%

Transients with PCS Unavailable 2.98E-07 16.0%

Total Loss of Offsite Power 2.19E-07 11.7%

Support Systems 2.17E-07 11.6%

Internal Floods 1.14E-07 6.1%

LOCAs 8.90E-08 4.8%

Interfacing Systems LOCAs 5.OOE-08 2.7%

Stuck-Open Relief Valves 1.42E-08 0.8%

Total 1.86E-06 100%

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 in Table 3-2.

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Unit I Summary Report Table 3-2 Core Damage Accident Sequence Classes Core Damage RVCnionContainment Functional Class RPV Condition Condition 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 III 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 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 3-3 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.

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Unit I Summary Report 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.

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Unit 1 Summary Report Table 3-3 Core Damage Accident Sequence Subclasses Accident Class Subclass Definition WASH-1400 Designator DesigatorExample 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.

0 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 II 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 TW following vent initiation. Suppression pool saturated but intact.

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Unit I Summary Report Table 3-3 Core Damage Accident Sequence Subclasses WASH-1400 Accident Class Subclass Definition Designator Designator Example Class Ill A Accident sequences leading to core damage (LOCA) conditions initiated by vessel rupture where the containment integrity is not breached in the initial time R 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 following vent initiation. Suppression pool saturated N/A but intact.

Class V Unisolated LOCA outside containment N/A 3-8

I 1.

Unit I Summary Report For Browns Ferry, functional based plant damage states shown in Table 3-4 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.

Table 3-4 Level 1 Plant Damage States Bin Frequency (per year) Percentage of CDF CLASS2A 6.552E-07 34.72%

CLASSIA 5.629E-07 29.83%

CLASS4 3.255E-07 17.25%

CLASS11BL 1.859E-07 9.85%

CLASS2L 6.337E-08 3.36%

CLASS5 4.997E-08 2.65%

CLASS3C 1.508E-08 0.80%

CLASS1D 1.112E-08 0.59%

CLASS3A 9.100E-09 0.48%

CLASSIBE 8.217E-09 0.44%

CLASS3B 7.760E-10 0.04%

CLASSIC O.OOOE+00 0.00%

CLASSIE O.OOOE+00 0.00%

CLASS2T O.OOOE+00 0.00%

CLASS2V O.OOOE+00 0.00%

Total 1.887E-06* 100.00%

Notes:

• The CDF for Level 1 is slightly higher than for the Level 2 because of truncation in the Containment Event Trees 3-9

Unit I Summary Report 3.1.2 Analysis of Individual Sequences A large number of sequences make up the total CDF. Table 3-5 provides information on the distribution of core damage sequences across the frequency range.

Table 3-5 Core Damage Sequences in Each Frequency Range Frequency Range Number of Sequences Percentage of CDF (events per year) Saved

>1 E-08 17 17%

>1 E-09 229 46%

>1E-10 1957 70%

>1 E-11 13358 87%

>1E-12 (base case) 100%

The following presents a brief description of the 20 highest-ranking sequences to the CDF:

• A loss of condenser vacuum is the initiating event for the first two sequences. The Level 1 portion of each sequence is identical but the CET path is different. HPCI and RCIC fail and depressurization fails. Core damage is from lack of injection.

• In sequences three and four the initiator is inadvertent closure of the MSIVs. Both reactor feedwater and condensate are lost as a result of the initiator. HPCI and RCIC both fail. No manual emergency vessel depressurization occurs. Core damage results from a loss of inventory.

. The fifth sequence is initiated by a break in the RHR suction piping. This is an interfacing system LOCAs. The break is not isolated and then steams the ECCS equipment rooms. Core damage results because of assumed damage to ECCS equipment. The second sequence is similar but is a large pipe break that has the same effect

• The sixth sequence is initiated by a loss of offsite power. All of the Unit 1 and 2 diesel generators fail as do Unit 3 diesel generators 3EA and 3ED. This results in a failure of EECW and RHRSW. Electric power is not recovered in six hours and core damage occurs.

• Sequence seven in initiated by a loss of Raw Cooling Water. The power conversion system fails as a result of the initiator and there is a common cause of all four RHR heat exchangers. The hardened wetwell vent fails and core damage occurs.

• Sequence eight is initiated by a turbine trip and a subsequently injection with feedwater/condensate fails. Other injection systems are successful but suppression pool cooling and the hardened wetwell vent fails.

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Unit I Summary Report

• Sequences nine, ten, thirteen and fourteen are initiated by a turbine trip with a failure to scram. The cases are all similar in that injection failure is the cause of core damage.

The sequences differ in the containment event tree results. Two of the sequences proceed to LERF.

• Sequence eleven is initiated by a total loss of feedwater. High-pressure injection fails and the vessel is not depressurized. Core damage occurs due to lack of inventory control.

• Sequence twelve is an interfacing system LOCA sequence due to test and maintenance of the RHR injection line. The coolant loss flow rate is small, but the operators fail to isolate the LOCA; this sequence causes a NOLERF end state.

• Sequence fifteen is initiated by a loss of the 500kV grid to the unit. High pressure injection fails and the vessel is not depressurized.

• A flood from the torus initiates sequence sixteen. For this initiator the operators are assumed to initiate a manual shutdown. Such a flood is assumed to disable all of the ECCS injection equipment in the torus room and corner rooms. Given the extremely degraded state of injection capability the only viable sources of level control are feedwater or low-pressure injection with condensate or through the RHR crosstie to Unit

2. In this sequence both condensate and the crosstie fail.

• Sequence seventeen is similar to sequence 8 except the initiator is a miscellaneous trip instead of a turbine trip.

• Sequence eighteen represents a total loss of feedwater event in which both HPCI and RCIC fail.

. Sequence 19 is a non-minimal version of sequence one. That is, it contains an additional failure (top event U2AP) that is not relevant to core damage.

• Sequence 20 is an ATWS scenario initiated by a loss of feedwater. A lack of early injection causes core damage.

Section Appendix A contains a listing of the top 100 sequences.

Table 3-6 shows the frequency, percentage of total, and the cumulative percentage of total for the sequences discussed above.

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Unit I Summary Report Table 3-6 Contribution to CDF of the Top Twenty Sequences Sequence Frequency %CDF Cumulative 1 5.11 E-08 2.74% 2.74%

2 3.72E-08 2.00% 4.74%

3 3.00E-08 1.61% 6.35%

4 2.19E-08 1.17% 7.52%

5 1.88E-08 1.01% 8.53%

6 1.72E-08 0.92% 9.45%

7 1.60E-08 0.86% 10.31%

8 1.50E-08 0.81% 11.12%

9 1.36E-08 0.73% 11.85%

10 1.36E-08 0.73% 12.58%

11 1.31 E-08 0.70% 13.29%

12 1.23E-08 0.66% 13.95%

13 1.19E-08 0.64% 14.59%

14 1.19E-08 0.64% 15.23%

15 1.17E-08 0.63% 15.86%

16 1.12E-08 0.60% 16.46%

17 1.09E-08 0.59% 17.04%

18 9.57E-09 0.51% 17.56%

19 8.96E-09 0.48% 18.04%

20 8.83E-09 0.47% 18.51%

3.1.3 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 3-7 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 included in Table 3-7 3-12

Unit 1 Summary Report because they are a complex function of the time available and the specific equipment failures involved. No other human error probabilities are shown because of a dramatic decline in importance.

Table 3-7 Operator Actions Important to CDF Frequency-Weighted Fractional Database Operator Action Description Importance to Core Variable Damage HPRVD1 OPERATOR FAILS TO INITIATE DEPRESSURIZATION 2.7E-01 HPWWV1 OPERATOR FAILS TO OPEN WETWELL VENT 2.3E-01 HRSPC1 OPERATOR FAILS TO LOCALLY RECOVER SP COOLING FAILURE 1.4E-01 HOU11 OPERATOR FAILS TO ALIGN THE RHR UNIT 1/UNIT 2 CROSSTIE 4.OE-02 HPHPE1 OPERATOR FAILS TO CONTROL LEVEL WITH HPCI/RCIC (EARLY - 2.OE-02 6 HOURS)

HPHPR1 OPERATOR FAILS TO CONTROL LEVEL WITH HPCIIRCIC 1.7E-02 FOLLOWING LEVEL 8 TRIP HOTAF1 OPERATOR FAILS TO CONTOL LEVEL AT TAF DURING ATWS - 1.5E-02 UNISOLATED VESSEL HPSPC1 OPERATOR FAILS TO ALIGN SUPPRESSION POOL COOLING- 1.1 E-02 THIS IS A NON ATWS SCENARIO HODWS1 OPERATOR FAILS TO ALIGN FOR DRYWELL SPRAY. THIS IS A 6.5E-03 NON ATWS SCENARIO.

HPRVD2 OPERATOR FAILS TO INITIATE DEPRESSURIZATION GIVEN 6.2E-03 FAILURE TO CONTROL HIGH PRESSURE LEVEL CONTROL HPTAF2 OPERATOR FAILS TO CONTROL LEVEL AT TAF DURING ATWS - 3.4E-03 ISOLATED VESSEL HREEC1 OPERATOR FAILS TO RESPOND TO INADEQUATE EECW FLOW 2.8E-03 TO DG FOLLOWING LOSP HPSLC1 OPERATOR FAILS TO INITIATE SLC GIVEN RPV ISOLATED 1AE-03 HPSPC2 OPERATOR FAILS TO ALIGN SUPPRESSION POOL COOLING 1.4E-03 DURING ATWS 3-13

Unit I Summary Report 3.1.4 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 3-8, which provides a ranking in order of their impact on the total CDF.

Table 3-8 Systems Important to CDF System %CDF HPCI 55.7%

RCIC 40.7%

RHR 38.7%

Feedwater/Condensate 17.1%

RPS 16.9%

Main Steam 11.1%

Diesel Generators 9.3%

RHRSW (not including EECW) 2.4%

Core Spray 1.6%

Standby Liquid Control System

• 0.5%

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 associated plant event. Consequential failures resulting from dependencies on other plant systems are not included in this importance ranking.

3.2 RESULTS FOR LARGE EARLY RELEASE FREQUENCY This section summarizes the results for the Level 2 analysis, which estimates the large containment failure and subsequent early release of radionuclides known as LERF. This section presents the LERF results and contributors.

The results indicate that about 10% 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. Browns Ferry Unit 1 falls in the lower 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 factors that contribute to the 3-14

Unit 1 Summary Report release of radionuclides 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.

A short summary of the top twenty LERF sequences follows.

The top twenty sequences can be characterized by three groups of functional failures. The highest frequency group is comprised of eighteen sequences of the top twenty. All involve ATWS sequences with failure to control level. The functional failure group with the second highest frequency in the top twenty sequences is an interfacing system LOCA involving a small break in the RHR suction shutdown cooling line. This group comprises just one sequence (the fifth) in the top twenty. The other group is just one sequence and involves a failure of the torus during a medium LOCA.

Table 3-9 presents percentage contribution to LERF from different initiating event categories.

Table 3-9 Initiating Event Contribution to LERF Mean LERF (per Percentage of Initiating Event Category reactor year) Total ATWS 1.7E-07 88.9%

LOCAs 7.1 E-09 3.8%

Interfacing System LOCAs 5.2E-09 2.8%

Transients with PCS Available 3.1E-09 1.7%

Transients with PCS Unavailable 2.7E-09 1.5%

Internal Floods 9.7E-10 0.5%

Total Loss of Offsite Power 9.3E-10 0.5%

Support Systems 4.6E-10 0.2%

Stuck-Open Relief Valves 2.3E-1 0 0.1%

Total 1.9E-07 100.0%

As discussed previously, 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 3-15

Unit I Summary Report water prior to vessel breach. Drywell spray represents an important hardware component since it is the primary means of flooding the drywell.

3.3 IMPORTANT ASSUMPTIONS

1. Analyses indicate that enhanced CRD injection alone is not sufficient to prevent fuel damage if the vessel remains at high pressure. However, as indicated in cases CRD1-5 in the thermal hydraulics notebook, enhanced CRD may be sufficient to prevent vessel melt through.

2. Given a loss of offsite power and the conditions that all four Unit 1/Unit 2 diesels are unavailable but all four Unit 3 diesels are available, it is assumed that 2 of the 4 available Unit 3 diesels will be aligned to service Unit 1 and Unit 2.

3. This is a Unit 1 specific PSA. Given an initiator that may involve multiple units, such as loss of offsite power or loss of plant control air, plant requirements for the number of available RHRSW pumps may be impacted. The current PSA model assumes that three of the eight RHRSW pumps are required and only those the RHR heat exchangers served by those pumps have sufficient cooling.

4. The diesel driven fire pump is represented in the scenario logic model. However, given the limited flow available to the vessel from this source, no credit is given for this source for preventing fuel damage.

5. The initiators considered in the PSA are limited to internal initiators and internal floods.

The process used to identify and select these initiators is rigorous and results in a reasonably complete set.

6. Spatial interaction analyses form an important part of the analysis of internal flooding initiators. Spatial interaction considerations are also included in the evaluation of scenarios in which a common environment could impact the operability of multiple components or where the initiator can cause collateral impacts. Examples include: the assumed loss of all equipment in the torus room and comer rooms (includes HPCI, RCIC, RHR and CS) given a flood of that area; and, the isolation of the main steam lines given the specific break (RCIC steam line break in the steam tunnel) assumed for the Break Outside Containment initiator.

7. To successfully mitigate a medium LOCA, it is assumed that operation of HPCI or the opening of SRVs is required to allow timely injection by the low-pressure systems.

8. The success criteria for sufficient core cooling for large LOCAs is based EPU Task Report T0407 (Reference 16).

9. Water successfully delivered to the drywell floor (for example, via operations of drywell spray) is assumed to prevent ex-vessel debris from failing the drywell liner.

3-16

Unit I Summary Report

10. A simplified common cause model is used for the 12 RHRSW pumps. The model assumes that the 12 pumps can be adequately modeled with three common cause groups with one master group that includes all 8 pumps dedicated to RHR.

11. Following a transient initiator and loss of all high pressure injection, operation of a single SRV will depressurize the vessel in a timely manner allowing successful level control by low pressure injection systems.

3.4 SENSITIVITY EVALUATIONS The data uncertainty presented earlier in the executive summary represent only a portion of the total uncertainty in the model. A number of sensitivity runs have been performed to develop a perspective on the assumptions and characteristics of the model. The sensitivity evaluations are listed below.

1. Assumption 2 above relates to the availability of two EDGs from Unit 3 to support Units 1 and 2 given a total LOSP and all four Unit 3 EDGs successful. The failure to provide a

&spare" EDG to Unit 1will be considered in sensitivity Case 1.

2. Assumption 9 relates to the effectiveness of the water on the drywell floor to cool the ex-vessel debris. The LERF will be evaluated without this assumption as sensitivity Case 2.

3. The modeling of common cause for the RHRSW pumps will be varied to obtain a perspective on the grouping used in the model. This result is presented below as sensitivity Case 3.

4. Common cause is modeled between HPCI and RCIC. This is not necessarily done in the PSA for other BWRs. Sensitivity Case 4 evaluates CDF and LERF assuming no common cause between the HPCI and RCIC pumps.

5. Sensitivity Case 5 evaluates CDF and LERF under the assumption that the operator action to depressurize is guaranteed to succeed.

The results of the six sensitivity cases are compared to the base case CSF and LERF results below.

Case ID CDF LERF Base 1.86E-06 1.87E-07 Case 1 2.07E-06 1.87E-07 Case 2 N/A 2.92E-07 Case 3 1.87E-06 I.87E-07 Case 4 1.79E-06 1.84E-07 Case 5 1.35E-06 1.36E-07 3-17

A Unit 1 Summary Report Short summaries of the conclusions from the sensitivity cases follow.

1. Failure to credit the Unit 3 diesel generators almost doubles the contribution from LOSP.

The diesel generator cross tie capability is an important feature in the design and operation of BFN.

2. As expected, debris bed cooling reduces LERF. This sensitivity may prove useful if the plant considers changes to the high-pressure fire protection system.

3. The sensitivity case expanded the global common cause failure term from the 8 pumps dedicated to RHRSW to all twelve RHRSW pumps. This had little affect on the results, as sequences in which EECW is necessary generally require RHR heat exchanger cooling. However, the value assigned to the common cause failure is important.

4. Modeling common cause between HPCI and RCIC is necessary to ensure completeness because of the significance of sequences characterized by an initial loss of the PCS where HPCI and RCIC fail and the vessel is not depressurized.

5. The sensitivity reinforces the importance of the operator action to depressurize given the loss of high-pressure injection.

3.5 INSIGHTS The diversity in power and cooling at Browns Ferry offers additional mitigation features not available to the majority of BWRs.

Failures of high-pressure injection with a failure to depressurize are an important class of core damage sequences. The most important of these are initiated by a loss of the power conversion system as the initiating event.

The ability of Unit 1 to access backup sources for control and motive power plays a significant role in reducing the core damage frequency. The backup control power capability is manifested in the model as the use of the spare charger for sequences involving failure of batteries 1, 2 or 3. The 250V DC System Notebook is referred to for further details.

The ability to crosstie 4kV shutdown boards A and C to 3EA and 3EC, respectively, reduces core damage in cases where the units have lost offsite power. The Unit 3 boards are only credited if all Unit 3 4kV shutdown boards are powered. The ability to power the 480V RMOV shutdown boards 1A and 1B from 4kV shutdown board B is also significant. The 480V AC Power System Notebook is referred to for further details.

3-18

Unit 1 Summary Report SECTION 4 REFERENCES

1. ASME RA-S-2002, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, April 2002.

2. Draft Regulatory Guide DG-1122, An Approach for determining the technical Adequacy of Probabilistic Risk assessment results for Risk-Informed Activities, November 2002.

3. Tennessee Valley Authority, "Browns Ferry Nuclear Plant Final Safety Analysis Report".

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

5. PLG, Inc., "RISKMAN for Windows Software", Users Manuals l-IV, Version 7.0, 2004.

6. Electric Power Research Institute, Nuclear Safety Analysis Center, "EPRI PRA Repository",

NSAC 152, Vol. 5, Shoreham Unit 1, October 1991.

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

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

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

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

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.

12. Tennessee Valley Authority, "Browns Ferry Unit 1 Probabilistic Risk Assessment,"

September 1987.

13. U.S. Nuclear Regulatory Commission, "Modeling Time to Recovery and Initiating Event Frequency for Loss of Off-Site Power Incidents at Nuclear Power Plants", NUREG/CR-5032, January 1988.

14. U.S. Nuclear Regulatory Commission, uCommon-Cause Failure Parameter Estimations",

NUREG/CR-5497, October 1998, INEELIEXT-97-01328.

4-1

Unit I Summary Report

15. Browns Ferry Nuclear Plant Unit 2 Probabilistic Safety Assessment Individual Plant Examination, Interim Order Number 3, October 1997, RIMS No. R14940812 201.

16. Browns Ferry Nuclear Plant Unit 1, EPU and MELLLA+ Task Report T0407, ECCS - LOCA SAFER/GESTER, March 2004 4-2

.. 1 Unit 1 Summary Report APPENDIX A UNIT I TOP RANKING SEQUENCES CONTRIBUTING TO CDF A-1

..: *. 1.

Unit 1 Summary Report INITIATOR DESCRIPTIONS BOC BREAK OUTSIDE CONTAINMENT ELOCA EXCESSIVE LOCA EXFW EXCESSIVE FEEDWATER FLOW FLRB1 EECW FLOOD IN REACTOR BULIDING - SHUTDOWN UNIT FLRB2 EECW/RHRSW FLOOD IN REACTOR BUILDING - OPERATING UNIT FLRB3C FLOOD FROM THE CONDENSATE STORAGE TANK FLRB3S FLOOD FROM THE TORUS FLTB LARGE TURBINE BUILDING FLOOD FLTB2 (SMALL) FLOOD IN THE TURBINE BUILDING HIPT RCS HIGH PRESSURE TRIP IMSIV INDVERTANT MSIV CLOSURE IOOV INADVERTENT OPENING OF ONE SRV IOOVA INADVERTANT OPEN ONE SRV ATWS ISCRAM INADVERTENT (OTHER) SCRAM LSOOPA LOSS OF 500 KV TO PLANT L500U LOSS OF SOO KV TO ONE UNIT LCV LOSS OF CONDENSER VACUUM LICA LOSS OF I & C BOARD A LICE LOSS OF I & C BOARD B LLCA CORE SPRAY LOOP I LINE BREAK LLCB CORE SPRAY LOOP II LINE BREAK LLDA RECIRC DISCHARGE LINE A BREAK LLDB RECIRC DISCHARGE LINE B BREAK LLO OTHER LARGE LOCA LLSA RECIRC SUCTION LINE A BREAK LLSB RECIRC SUCTION LINE B BREAK LOCHSA LOSS OF CONDENSER HEAT SINK ATWS LOFWA LOSS OF FEEDWATER ATWS LOPA LOSS OF PLANT AIR LOSP TOTAL LOSS OF OFFSITE POWER LOSPA LOSP ATWS LRCW LOSS OF RAW COOLING WATER MLOCA MEDIUM LOCA MLOSP MOMENTARY LOSS OF OFFSITE POWER OT OTHER TRANSIENTS PLCF PARTIAL LOSS OF CONDENSATE FLOW PLFW PARTIAL LOSS OF FEEDWATER SLOCA SMALL LOSS OF COOLANT ACCIDENT (LOCA)

TBU TURBINE BYPASS UNAVAILABLE TLCF TOTAL LOSS OF CONDENSATE TLFW TOTAL LOSS OF FEEDWATER TT TURBINE TRIP WITH BYPASS, OTHER SIMILAR EVENTS TTA TURBINE TRIP ATWS VI CORE SPRAY INJECTION LINE (ONE LOOP) ISLOCA VITM ISLOCA DUE TO TEST AND MAINTENANCE OF CS INJECTION LINE (ONE LOOP)

VLOCA VERY SMALL LOCA (RECIRC PUMP SEAL LOCA)

VR RHR INJECTION LINE ISLOCA (ONE LOOP)

VRTM ISLOCA DUE TO TEST & MAINTENANCE ON RHR INJECTION LINE (ONE LOOP)

VS RMR SUCTION LOCA (BOTH LOOPS)

A-2

'i '.

Model Name: u1040727 Master Frequency File: L2RO Sequences for Group: CDF Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin 1 63 LCV 5.1052E-008 ///DWF*//RVC0*MCDF*RCI1*HPI4*ORVD1*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*/ELF 2 67 LCV 3.721SE-008 ///DWF*//RVCO*MCDF*RCI1*HPI4*ORVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*OI1*IRl*/ELF 3 39 IMSIV 3.0015E-008 ///DWF*//IVOF*RVC0*RCI1*HPI4*0RVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*/ELF 4 42 IMSIV 2.1882E-008 ///DWF*//IVOF*RVC0*RCI1*HPI4*ORVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*OIl*IRl*/ELF 5 1 VS 1.8769E-008 /ISLV43*ISLV51*/ISVF*OPVF NOLERF 6 2113 LOSP 1.7190E-008 /OG5F*OG16F*UB41AF*U341BF*UB41CF*UB42AF*0B42 NOLERF BF*SHUTlF*SHT2F*GAl*GD2*G34*GC4*EPR303*DGC1*

AAF*RQF*REF*RMFT*RZF*CHlF*ABF*RSF*RHF*CH2AF*U B42CF*DIF*DKF*ACF*RRF*RFF*ADF*RTF*RIF*RJF*RN F*DLF*/UB43AF*UB43BF*GEF*A3EAF*RXF*ROF*CH2BF

• CPRECF*DNF*GGF*A3ECF*GFF*A3EB F*RYF*DOF*RPF*GHF*A3EDF*SDRECF*/DWF*/RCWF*EA F*EBF*ECF*EDF*SW2AF*SWlAF*SW2BF*SWlBF*SW2CF*

SWlCF*SW2DF*SWlDF*PCAF*CADF*DCAF*/IVOF*RVCO*

CNDF*EPR61*/FWSDF*EECWF*EECWRF*REPWRF*RFPWRF

• RPAF*HXAF*RPCF*HXCF*RPBF* BBF*RPDF*HXDF*/PC SF*CSF*U2XF*DWSF*SPF*OSPRF*CRD 7 36 LRCW 1.6039E-008 ///DWF*/RCWF*/RVCO*MCDF*CNDF*/FWSDF*FXA1*hXC NOLE2F 2*HXB5*HXD7*/PCSF*PCSRF*DWSF*SPF*OSPRF*OWWVl

• //ELF 8 43 TT 1.4999E-008 ///DWF*//RVCO*MCD1*CNDF*/FWSDF*/PCSF*PCSRF*S NOLERF Pl*OSPRl*OWWV1*//ELF 9 2 TTA 1.3649E-008 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HP12*/FWSDF*/P NOLERF CSF*PCSRF*//IVR2*FCF 10 8 TTA 1.3649E-008 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HP12*/FWSDF*/P LERF CSF*PCSRF*//WW2*IVR2*FCF 11 18 TLFW 1.3123E-008 ///DWF*//RVCO*FWHF*RCI1*HPI4*ORVD1*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*/ELF 12 1 VRTM 1.2340E-008 /ISLV42*ISLV53*/ISVF*OPVF NOLERF 13 4 TTA 1.1900E-008 ///DWF*//RPSN1*OSVF*RVC4*FWHF*8P12*/FWSDF*/P NOLERF CSF*PCSRPF*//OP5*IVR1*FCF 14 10 TTA 1.1900E-008 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P LERF CSF*PCSRF*//WW2*OP5*IVR1*FCF 15 73 L50OU 1.1729E-008 /OGSF*RZF*//DWF*//RVCO*MCDF*CNDF*RCI1*FPI4*0 NOLERF RVD1*/FWSDF*/PCSF*LVPRF*/ALF*/ELF 16 3 FLRB3S 1.1156E-008 ///DWF*//RVCO*CND1*RCIF*HPIF*/FWSDF*RPAF*HXA NOLERF F*RPCF*HXCF*RPBF*HXBF*RPDF*HXDF*/PCSF*CSF*U2 X1*DWSF*SPF*OSPRF*//ELF 17 35 OT 1.0900E-008 ///DWF*//RVCO*MCDl*CNDlF*/FWSDF*/PCSF*PCSRF*S NOLERF Pl*OSPR1*OWWV1*//ELF 18 20 TLFW 9.5668E-009 ///DWF*//RVCO*FWHF*RCI1*HPI4*ORVD1*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*OI1*IRl*/ELF 19 295 LCV 8.9609E-009 ///DWF*U2AP1*//RVCO*MCDF*RCIl*HP14*ORVDl*/FW NOLERF SDF*/PCSF*LVPRF*/ALF*/ELF 20 7 LOFWA 8.8314E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P LERF

z. 9?

Model Name: U1040727 Master Frequency File: L2RO Sequences for Group: CDF Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin CSF*PCSRF*//WW2*IVR2*FCF 21 2 LOFWA 8.8314E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P NOLERF CSF*PCSRF*//IVR2*FCF 22 75 L500U 8.5507E-009 /OG5F*RZF*//DWF*//RVCO*MCDF*CNDF*RCI1*HPI4*0 NOLERF RVDl*/FWSDF*/PCSF*LVPRF*/ALF*OIl*IRl*/ELF 23 385 TT 8.4580E-009 ///DWF*//RVC0*MCDl*CNDF*RCIl*HPI4*ORVDl*/FWS NOLERF DF*/PCSF*LVPRF*/ALF*/ELF 24 3 LOFWA 7.7003E-009 ///DWF*//RPSNl*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P NOLERF CSF*PCSRF*//OP5*IVR1*FCF 25 8 LOFWA 7.7003E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P LERF CSF*PCSRF*//WW2*OP5*IVR1*FCF 26 1 LLDB 7.4722E-009 //////SPI2*SPIIll*SPCF*//ELF NOLERF 27 1 LLDA 7.4722E-009 //////SPI2*SPIIll*SPCF*//ELF NOLERF 28 31 FLTB 7.4559E-009 ///DWF*/RCWF*PCAF*DCAF*/IVOF*RVCO*CNDF*/FWSD NOLERF F*HXA1*HXC2*HXBS*HXD7*/PCSF*PCSRF*DWSF*SPF*O SPRF*OWWV1*//ELF 29 5 LRCW 6.5924E-009 ///DWF*/RCWF*/RVC0*MCDF*CNDF*/FWSDF*/PCSF*PC NOLERF SRF*SPl*OSPRl*OWWVl*//ELF 30 297 LCV 6.5326E-009 ///DWF*U2APl*//RVCO*MCDF*RCIl*HPI4*ORVDl*/FW NOLERF SDF*/PCSF*LVPRF*/ALF*OIl*IRl*/ELF 31 1 VR 6.1678E-009 /ISLV42*ISLV55*/ISVF*OPVF NOLERF 32 387 TT 6.1660E-009 ///DWF*//RVC0*MCDl*CNDF*RCIl*HP14*ORVD1*/FWS NOLERF DF*/PCSF*LVPRF*/ALF*OIl*IRl*/ELF 33 2 VRTM 6.1474E-009 /ISLV22*ISLV5F*/ISVF*OPVF NOLERF 34 316 OT 6.1461E-009 ///DWF*//RVCO*MCDl*CNDF*RCIl*HPI4*ORVDl*/FWS NOLERF DF*/PCSF*LVPRF*/ALF*/ELF 35 7 FLRB3S 6.0573E-009 ///DWF*//RVCO*CND1*RCIF*HPIF*/FWSDF*RPAF*HXA NOLERF F*RPCF*HXCF*RPBF*F 3F*RPDF*HXDF*/PCSF*CSF*OR HXT1*DWSF*SPF*OSPRF*//ELF 36 84 BOC S.9852E-009 ///DWF*//IVOF*RVC0*CNDF*RCIF*HPI4*ORVD1*/FWS NOLERF DF*/PCSF*LVPRF*/ALF*/ELF 37 1 ELOCA 5.9471E-009 ///DWP*////ALF*/ELF NOLERF 38 239 LOPA 5.6772E-009 ///DWF*/PCAF*DCAF*/IVOF*RVCO*CNDF*RCI1*HP14* NOLERF ORVD1*/FWSDF*/PCSF*LVPRF*/ALF*/ELF 39 188 IMSIV 5.2684E-009 ///DWF*U2APl*//IVOF*RVCO*RCIl*HPI4*ORVDl*/FW NOLERF SDF*/PCSF*LVPRF*/ALF*/ELF 40 93 TT 5.1789E-009 ///DWF*//RVC0*MCDl*CNDF*/FWSDF*HXAl*HXC2*HXB NOLERF 5*HXD7*/PCSF*U2Xl*DWSF*SPF*OSPRF*OWWVl*//ELF 41 3 VS 5.1193E-009 /ISLVll*ISLV31*/ISVF*RBVF LERF 42 57 L500PA 4.9522E-009 /OGSF*UB42AF*UB42BF*RZF*//DWF*//RVC0*MCDF*CN NOLERF DF*RCIl*HPI4*ORVDl*/FWSDF*/PCSF*LVPRF*/ALF*/

ELF 43 131 TLCF 4.6293E-009 ///DWF*//RVC0*CNDF*RCIl*HPI4*ORVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*/ELF 44 318 OT 4.4806E-009 ///DWF*//RVCO*MCDl*CNDF*RCIl*HP14*ORVDl*/FWS NOLERF DF*/PCSF*LVPRF*/ALF*OI1*IRl*/ELF 45 1856 LOSP 4.3914E-009 /OGSF*OG16F*U,341AF*UB41BF*UB41CF*UB42AF*UB42 NOLERF BF*SHUTlF*SHT2F*GAl*GD2*G34*EPR303*AAF*RQF*R EF*RMF*RZF*CHlF*ABF*RSF*RHF*CH2AF*UB42CF*DIF

• DKly*ADF*RTF*RIF*RJF*RNF*/UB43AF*UB43BF*CH2B

', rt -.

Model Name: U1040727 Master Frequency File: L2RO Sequences for Group: CDF Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin F*CPRECF*DNF*GG1*A3ECF*RYF*RPF*SDRECF*/DWF*/

RCWF*SW2AF*SWlAF*SWlBF*SW2CF*S WlCF*SW2DF*PCAr*DCAF*/IVOF*RVCO*CNDF*EPR61*/

FWSDF*REPWRF*RPAF*HXAF*RPCF*HXCF*HXBF*RPDF*H XDF*/PCSF*U2XF*DWSF*SPF*OSPRF*CRDF*//ELF 46 1805 LOSP 4.3907E-009 /OGSF*OG16F*UB41AF*UB41BF*UB41CF*UB42AF*UB42 NOLERF BF*SHUTlP*SHT2F*GAl*GD2*GB4*EPR303*AAF*RQF*R EF*RMF*RZF*CHlF*ABF*RSF*RHF*CH2AF*UB42CF*DIF

• DRF*ADF*RTF*RIF*RJF*RNF*/UB43AF*UB43BF*CF2B F*CPRECF*DNF*GH1*A3EDF*SDRECF*/DWF*/RCWF*SW2 AF*SWlAF*SW2CF*SWlCF*SW2DF*SWl DF*PCAF*DCAF*/IVOF*RVCO*CNDF*EPR61*/FWSDF*RE PWRF*RPAF*HXAF*RPCF*EXCF*HXBF*RPDF*HXDF*/PCS F*U2XF*DWSF*SPF*OSPRF*CRDF*//ELF 47 1675 LOSP 4.3640E-009 /OG5F*OG16F*UB41AF*UB41BF*tJB41CF*UB42AF*UB42 NOLERF BF*SHUTlF*SHT2F*GAl*GD2*GC3*EPR303*AAF*RQF*R EF*RMF*RZF*CHlF*UB42CF*DIF*DKF*ACF*RRF*RFF*A DF*RTF*RIF*RJF*RNF*DLF*/UB43AF*UB43BF*CH2BF*

CPRECF*GG1*A3ECF*RYF*DOF*RPF*SDRECF*/DWF*/RC WF*EBF*SW2AF*SWlAF*SW2BF*SWlBF

• SW2DF*DCAF*/IVOF*RVCO*CNDF*EPR61*/FWSDF*REP WRF*RFPWRF*RPAF*HXAFr*HXCF*RPBF*IVCBF*RPDF*HXD F*/PCSF*CSF*U2XF*DWSF*SPF*OSPRF*CRDF*//ELF 48 86 BOC 4.3633E-009 ///DWF*//IVOF*RVCO*CO1DF*RCIF*HPI4*ORVDI*/FWS NOLERF DF*/PCSF*LVPRF*/ALF*OIl*IRl*/ELF 49 1632 LOSP 4.3191E-009 /OG5F*OG16F*UB41AF*UB41BF*UB41CF*Ui342AF*UB42 NOLERF BF*SHUTlF*SHT2F*GAl*GD2*GC3*EPR303*AAF*RQF*R EF*RMF*RZF*CHlF*UB42CF*DIF*DKF*ACF*RRF*RFF*A DF*RTF*RIF*RJF*RNF*DLF*/U343AF*UB43BF*CH2BF*

CPRECF*GF1*A3EBF*SDRECF*/DWF*/RCWF*EBF*ECF*E DF*SW2AF*SWlAF*SW2BF*SW2DF*PCA F*DCAF*/IVOF*RVCO*CNDF*EPR61*/FWSDF*EECWF*RE PWRF*RFPWRF*RPAF*HXAF*EXCF*RPBF*FXBF*RPDF*HX DF*/PCSF*CSF*U2XF*DWSF*SPF*OSPRF*CRDF*//ELF 50 66 LCV 4.2637E-009 ///DWF*//RVCO*MCDF*RCIl*HPI4*ORVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*OI1*/ELF 51 2 LOCHSA 4.2420E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P NOLERF CSF*PCSRF*//IVR2*FCF 52 6 LOCHSA 4.2420E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*FPI2*/FWSDF*/P LERF CSF*PCSRF*//WW2*IVR2*FCF 53 241 LOPA 4.1388E-009 ///DWF*/PCAF*DCAF*/IVOF*RVC0*CNDF*RCI1*HPI4* NOLERF ORVDl*/FWSDF*/PCSF*LVPRF*/ALF*OIl*IRl*/ELF 54 288 LRCW 4.0795E-009 ///DWF*/RCWF*/RVC0*MCDF*CNDF*RCI1*HPI4*/FWSD NOLERF F*EXAl*HXC2*HXB5*HXD7*/PCSF*PCSRF*DWSF*SPF*O SPRF*//ELF 55 442 5T 3.8718E-009 ///DWF*//RVC2*CNDF*/FWSDF*/PCSF*PCSRF*SPl*OS NOLERF PR1*/ALF*/ELF 56 4 LICB 3.8487E-009 ///DWF*/DCAF*/IVOF*RVCO*CNDF*/FWSDF*/PCSF*PC NOLERF SRF*SPl*OSPRl*OWWVl*//ELF 57 190 IMSIV 3.8408E-009 ///DWF*U2APl*//IVOF*RVCO*RCIl*HPI4*ORVDl*/FW NOLERF

VTI Model Name: U1040727 Master Frequency File: L2RO Sequences for Group: CDF Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin SDF*/PCSF*LVPRF*/ALF*OIl*IRl*/ELF 58 318 TT 3.8150E-009 ///DWF*//RVCO*MCDl*CNDF*RCIl*HPI4*/FWSDF*/PC NOLERF SF*PCSRF*SPl*OSPRl*//ELF 59 74 OT 3.7633E-009 ///DWF*//RVCO*t!CDl*CNDF*/FWSDF*FXAl*hXC2*HXB NOLERF 5*hXD7*/PCSF*U2Xl*DNSF*SPF*OSPRF*OWWV1*//ELF 60 328 LRCW 3.7174E-009 ///DWF*/RCWF*/RVCO*MCDF*CNDF*RCIl*HPI4*ORVD1 NOLERF

• /FWSDF*/PCSF*LVPRF*/ALF*/ELF 61 7 LOChSA 3.6987E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HP12*/FWSDF*/P LERF CSF*PCSRF*//WW2*0P5*IVR1*FCF 62 3 LOCHSA 3.69B7E-009 ///DWF*//RPSN1*OSVF*RVC4*FWhF*HPI2*/FWSDF*/P NOLERF CSF*PCSRP*//OP5*IVR1*FCF 63 1359 LOSP 3.6302E-009 /OG5F*OG16F*UB41AF*UB41BF*UB41CF*UB42AF*UB42 NOLERF BF*ShU'lF*SHT2F*GAl*GB2*GC3*EPR303*AAF*RQF*R EF*RMF*RZF*CHlF*ABF*RSF*RHF*Ch2AF*UB42CF*DIF

• DKF*ACF*RRF*RFF*DLF*/UB43AF*U343BF*DNF*GGl*

A3ECF*RYF*DOF*RPF*SDRECF*OXNNF*/DWF*/RCWF*EB F*SW2AF*SWlAF*SW2BF*SWlBF*SW2C F*SWlCF*PCAF*CADF*DCAF*/IVOF*RVCO*CNDF*EPR61

• /FWSDF*REPWRF*RFPWRF*RPAF*hXAF*RPCF*hXCF*RP BF*HXBF*HXDF*/PCSF*CSF*U2XF*DWSF*SPF*OSPRF*C RDF* //ELF 64 1298 LOSP 3.6143E-009 /OGSF*OG16F*UB41AF*UB41BF*UB41CF*Ui342AF*UB42 NOLERF BF*ShU-lF*SHT2F*GAI*GB2*GC3*EPR303*AAF*RQF*R EF*RMF*RZF*CFlF*ABF*RSF*RHF*CFZAF*UB42CF*DIF

• DKF*ACF*RRF*RFF*DLF*/UB43AF*UB43BF*DNF*GH1*

A3EDF*SDRECF*OXNNF*/DWF*/RCWF*EBF*SW2AF*SW1A F*SW2BF*SW2CF*SWlCF*SWlDF*PCAF

• D)CAF* /IVOF* RVC0* CNDF* EPR61 */FWSDF*REPWRF*RF PWRF*RPAF*hXAF*RPCF*hXCF*RPBF*hXBF*hXDF*/PCS F*CSF*U2XF*DWSF*SPF*OSPRF*CRDF*//ELF 65 59 L500PA 3.6103E-009 /OG5F*UB42AF*UB42BF*RZF*//DWF*//RVCO*MCDF*CN NOLERF DF*RCI1*hPI4 *ORVD1*/FWSDF*/PCSF*LVPRF*/ALF*O Il*IRl*/ELF 66 33 LOPA 3.4781E-009 ///DWF*/PCAF*DCAF*/IVOF*RVCO*CNDF*/FWSDF*h5A NOLERF 1*HXC2*HXB5*H'XD7*/PCSF*U2Xl*DWSF*SPF*OSPRF*O WWV1*//ELF 67 5 MLOCA 3.4273E-009 ///DLWF*//HXA1*HXC2*HXB5*EXD7*OSPCF*//ELF NOLERF 68 133 TLCF 3.374BE-009 ///DWF*//RVCO*CNDF*RCIl*HPI4*ORVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*OI1*IR1*/ELF 69 145 FLRB3S 3.1317E-009 /OG51*RZF*//DWF*//RVCo*MCDF*CNDF*RCIF*HPIF*/ NOLERF FWSDF*RPAF*IHXAF*RPCF*hXCF*RPBF*hXBF*RPDF* hXD F*/PCSF*CSF*U2X1*DWSF*SPF*OSPRF*//ELF 70 27 LOPA 3.0881E-009 ///DWF*/PCAF*DCAF*/IVOF*RVCO*CNDF*/FWSDF*HXA NOLERF l*HXC2*HXBS*HXD7*/PCSF*PCSR4*DWSF*SPF*OSPRF*

OWWV1*//ELF 71 5 FLTB 3.0645E-009 ///DWF*/RCWF*PCAF*DCAF*/IVOF*RVCO*CNDF*/FWSD NOLERF F*/PCSF*PCSRF*SP1*OSPRl*OWWV1*//ELF 72 2142 LOSP 3.0171E-009 /OGSF*OG16F*UB41AF*UB41BF*UB41CF*UB42AF*UB42 NOLERF BF*SHUTlF*SHT2F*GAl*GD2*GB4*GC4*EPR303*DGC1*

AAF*RQF*REF*RMF*RZF*CHlF*ABF*RSF*RHF*CH2AF*U

Model Name: U1040727 Master Frequency File: L2RO Sequences for Group: CDF Sorted by Frequency Rank Index Initiator Frequency Failed and Multi-State Split Fractions Bin B42CF*DIF*DKF*ACF*RRF*RFF*ADF*RTF*RIF*RJF*RN F*DLF*/UB43AF*UB43BF*GEF*A3EAF*RXF*ROF*CH2BF

• CPRECF*DNF*GGF*A3ECF*GFF*A3EB F*RYF*DOF*RPF*GHF*A3EDF*SDRECF*/DWF*U2AP1*/R CWF*EAF*EBF*ECF*EDF*SW2AF*SWlAF*SW2BE*SWlBF*

SW2CF*SWlCF*SW2DF*SWlDF*PCAF*CADF*DCAF*/IVOF

• RVCO*CNDF*EPR61*/FWSDF*EECWF*EECWRF*REPWRF*

RFPWRF*RPAF*HXAF*RPCF*HXCF*RPBF*HXBF*RPDF*HX DF*/PCSF*CSF*U2XF*DWSF*SPF*OSP 73 18 TLCF 2.B345E-009 ///DWF*//RVCO*CNDF*/FWSDF*HXA1*HXC2*HXB5*HXD NOLERF 7*/PCSF*U2Xl*DWSF*SPF*OSPRF*OWWV1*//ELF 74 599 LRCW 2.8152E-009 ///DWF*U2AP1*/RCWF*/RVC0*MCDF*CNDF*/FWSDF*HX NOLERF A1*HXC2*HXB5*HXD7*/PCSF*PCSRF*DWSF*SPF*OSPRF

• OWWV1*//ELF 75 362 OT 2.8135E-009 ///DWF*//RVC2*CNDF*/FWSDF*/PCSF*PCSRF*SP1*OS NOLERF PR1*/ALF*/ELF 76 97 TT 2.8120E-009 ///DWF*//RVC0*MCDl*CNDF*/FWSDF*HXAl*HXC2*HXB NOLERF 5*HXD7*/PCSF*ORHXT1*DWSF*SPF*OSPRF*OWWVIl*//E LF 77 257 OT 2.7722E-009 ///DWF*//RVCO*MCDl*CNDF*RCI1*EPI4*/FWSDF*/PC NOLERF SF*PCSRF*SPl*OSPRl*//ELF 78 330 LRCW 2.7100E-009 ///DWF*/RCWF*/RVC0*MCDF*CNDF*RCIl*HP14*ORVDl NOLERF

• /FWSDF*/PCSF*LVPRF*/ALF*OIl*IR1*/ELF 79 25 BOC 2.6996E-009 ///DWF*//IVOF*RVC0*CNDF*RCIF*HP14*/FWSDF*/PC NOLERF SF*PCSRF*SPl*OSPRl*//ELF BO 1060 TT 2.632BE-009 ///DWF*U2APl*//RVCO*MCDl*CNDF*/FWSDF*/PCSF*P NOLERF CSRF*SPl*OSPRl*OWWVl*//ELF 81 14 ISCRAM 2.5629E-009 ///rDWF*//RVCO*MCD1*CNDF*/FWSDF*/PCSF*PCSRF*S NOLERF Pl*OSPRl*OWWVl*//ELF 82 31 LOSP 2.5427E-009 /OGSF*OG16F*UB41AF*UB41BF*UB41CF*U342AF*U342 NOLERF BF*SHUTlF*SHT2F*RZF*UB42CF*/UB43AF*UB43BF*/D WF*/RCWF*/BVRF*RVCO*CNDF*RCI1*HPI4*/FWSDF*HX A1*EXC2*HXB5*EXD7*/PCSF*PCSRF*DWSF*SPF*OSPRF

• //ELF 83 41 IMSIV 2.5067E-009 ///DWF*//IVOF*RVC0*RCIl*HP14*ORVDl*/FWSDF*/P NOLERF CSF*LVPRF*/ALF*OI1*/ELF 84 693 TT 2.5005E-009 ///DWF*/PCA1*DCAF*/IVOF*RVCO*CNDF*/FWSDF*HXA NOLERF 1*1BXC2*HXB5*HXD7*/PCSF*PCSRF*DWSF*SPF*OSPRF*

OWWV1*//ELF 85 179 TTA 2.4804E-009 ///DWF*//RPSNl*OSVF*RVC4*OTAFl*/FWSDF*/PCSF* LERF LVPRF*//WW2*IVR2*FCF 86 175 TTA 2.4804E-009 ///DWF*//RPSN1*OSVF*RVC4*OTAF1*/FWSDF*/PCSF* NOLERF LVPRF*//IVR2*FCF B7 29 MLOCA 2.4107E-009 ///DWF*//TORl*DWSF*/ALF*/WWBl*IVR3*CEF LERF 88 333 TTA 2.3956E-009 ///DWF*U2AP1*//RPSNl*OSVF*RVC4*FWHF*HPI2*/FW LERF SDF*/PCSF*PCSRF*//h'W2*IVR2*FCF 89 329 TTA 2.3956E-009 ///DWF*U2AP1*//RPSNl*OSVF*RVC4*FWHF*HPI2*/FW NOLERF SDF*/PCSF*PCSRF*//IVR2*FCF 90 476 OT 2.3941E-009 ///DWF*//TTPl*IVC1*/FWSDF*/PCSF*LVPRF*/ALF*/ NOLERF ELF

Model Name: U1040727 Master Frequency File: L2RO Sequences for Group: CDF Sorted by Frequency Rank Index Initiator Frequency Failed and Kulti-State Split Fractions Bin 91 22 LCV 2.3132E-009 ///DWF*//RVCO*MCDF*RCIl*OHPCl*OHPRl*HPL6*/FW NOLERF SDF*/PCSF*LVPRF*/ALF*/ELF 92 152 TLFW 2.3034E-009 ///DWF*U2APl*//RVCO*FWHF*RCIl*HPI4*ORVDl*/FW NOLERF SDF*/PCSF*LVPRF*/ALF*/ELF 93 2029 LOSP 2.2789E-009 /OGSF*OG16F*UB41AF*UB41BF*UB41CF*UB42AF*UB42 NOLERF BF*SHUTIlF*SHT2F*GAl*GD2*GB4*GC4*EPR303*AAF*R QF*REF*RMF*RZF*CHlF*ABF*RSF*RHF*CH2AF*UB42CF

• DIF*DKF*ACF*RRF*RFF*ADF*RTF*RIF*RJF*RNF*DLF

• /UB43AF*UB43BF*CH2BF*CPRECF*DNF*GG1*A3ECF*R YF*DOF*RPF*SDRECF*/DWF*/RCWF*E BF*SW2AF*SWlAF*SW2BF*SW1BF*SW2CF*SWlCF*SW2DF

• PCAF*CADF*DCAF*/IVOF*RVCO*CNDF*EPR61*/FWSDF

• REPWRF*RFPWRF*RPAF*HXAF*RPCF*HXCF*RPBF*HXBF

• RPDF*HXDF*/PCSF*CSF*U2XF*DWSF*SPF*OSPRF*CRD F*//ELF 94 42 LRCW 2.2762E-009 ///DWF*/RCWF*/RVC0*MCDF*CNDF*/FWSDF*HXA1*HXC NOLERF 2*HXB5*HXD7*/PCSF*U2Xl*DWSF*SPF*OSPRF*OWWVl*

//ELF 95 1179 LOSP 2.2732E-009 /OG5F*OG16F*UB41AF*U341BF*UB41CF*UB42AF*UB42 NOLERF BF*SHUTlF*SHT2F*GAl*GB2*EPR303*AAF*RQF*REF*R MF*RZF*CHlF*ABF*RSF*RHF*CH2AF*UB42CF*DIF*DKF

• /UB43AF*UB43BF*DNF*GG1*A3ECF*RYF*RPF*GH2*A3 EDF*SDRECF*OXNNF*/DWF*/RCWF*SW2AF*SWlAF*SWlB F*SW2CF*SWlCF*SWlDF*PCAF*DCAF*

/IVOF*RVCO*CNDF*EPR61*/FWSDF*REPWRF*RPAF*HXA F*RPCF*HXCF*HXBF*HXDF*/PCSF*U2XF*DWSF*SPF*OS PRF*CRDF*//ELF 96 1993 LOSP 2.26B9E-009 /OGSF*OG16F*UB41AF*UB41BF*UB41CF*UB42AF*UB42 NOLERF BF*SFTlF*SHT2F*GAl*GD2*GB4*GC4*EPR303*AAF*R QF*REF*RMF*RZF*CHlF*ABF*RSF*RHF*CF:2AF*UB42CF

• DIF*DKF*ACF*RRF*RFF*ADF*RTF*RIF*RJF*RNF*DLF

• /UB43AF*UB43BF*CH2BF*CPRECF*DNF*GH1*A3EDF*S DRECF*/DWF*/RCWF*EBF*SW2AF*SW1 AF*SW2BF*SW2CF*SWlCF*SW2DF*SWlDF*PCAF*DCAF*/

IVOF*RVC0*CNDF*EPR61*/FWSDF*REPWRF*RFPWRF*RP AF*HXAF*RPCF*EXCF*RPBF*HXBF*RPDF*HXDF*/PCSF*

CSF*U2XF*DWSF*SPF*OSPRF*CRDF*//ELF 97 133 TTA 2.2668E-009 ///DWF*//RPSNl*OSVF*RVC4*FWHF*RCIl*HP14*/FWS NOLERF DF*/PCSF*PCSRF*//IVR2*FCF 98 136 TTA 2.2668E-009 ///DWF*//RPSNl*OSVF*RVC4*FWHF*RCIl*HPI4*/FWS LERF DF*/PCSF*PCSRF*//WW2*IVR2*FCF 99 1 TTA 2.2317E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P NOLERF CSF*PCSRF*//RBEF 100 7 TTA 2.2317E-009 ///DWF*//RPSN1*OSVF*RVC4*FWHF*HPI2*/FWSDF*/P LERF CSF*PCSRF*//WW2*RBEF