ML23109A083

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NRC-2022-000160 - Resp 2 - Final, Agency Records Subject to the Request Are Enclosed, Part 6 of 7
ML23109A083
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II I Duke Power Company J. W HAMPTON Oconee Nuclear S11e Vice Presidenl P.0. Bo.r 1439 (864)885-3499 Office Seneca, SC '29679 (81H)885-3564 far

  • DUKEPOWER February 13, 1997 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Oconee Nuclear Site Docket Nos. 50-269, -270, -287 Probabilistic Risk Assessment Individual Plant .Examination

revision o, was issued as NSAC-60. In response to Generic Letter 88-20, Duke Power submitted revision 1 of the Oconee PRA to the NRC as part of the Oconee *Individual ,Plant Examination (IPE) submittal on November 30, 1990. Duke Power periodically evaluates changes to the *plant with respect to the assumptions and modeling in the Oconee PRA.

As a result of the periodic evaluations, Oconee initiated an *.

  • effort in 1995 to .update the Oconee PRA. The update addressed the following functions:
  • Incorporate plant changes since the time of the last update,
  • Continue to improve assumptions and .documentation,
  • Make use of recent plant experience and data,

A copy of the summary report for revision 2 of the Oconee PRA is attached. The summary report contains the high level results of the Oconee PRA up~ate .

A011 *Ii Pnn1ec,o,,~oa,,e,

, I If you have any questions about the attached report, please contact P. M. Abraham at (704)382-4520 or Michael Bailey at (864)885-4390.

Very truly yours,

~ttt!s: Vice President Nuclear Station Mr. L.A. Reyes, Regional Administrator U. S. Nuclear ~egulatory Commission, Region II Mr . D. E. LaBarge, Project Manager (2 copies)

Office of Nuclear Reactor Regulation Mr . M.A . Scott Senior Resident Inspector Oconee Nuclear Site

Conlrolltd Copy # I NRC Hud Quartm I...

Duke Power Company OCONEE NUCLEAR STATION.

PRA Revision 2 Summary Report December 1996

FORWARD This is the 1995 update of the Probabilistic Risk Assessment (PRA) study of Oconee Nuclear Station. 'This update is designated as Revision 2 since it constitutes a* revision to .

two previous studies-- the original Oconee PRA (NSAC-60) and the subsequent update of the original study (designated as Revision 1), which was used to satisfy the NRC's IPE requirement.

For developing the 1995 update, the plant configuration and procedures existing in 1994 fonned* the basis. The system and equipment reliabil.ity and avaiiability data were based on plant-specific data for the period 1988 - 1993 on risk-significant equipment. For initiating events, plant-specific data and certain generic data for the period for the period .

1980 - 1993 fonned the database.

The Revision 2 study employed the traditional event tree-fault tree methodology to conduct the systems analysis. However, a more state-of-the-art integration scheme was used to couple the many event tree end-point sequences with the fault tree models and the human recovery functions. Also, a large number of sensitivity studies have been done to derive insights on risk-significant systems, equipment failures, operator actions, and others. These enhancements were made to make the PRA model more suitable for the current environment of increased risk-informed plant operation and problem-solving. As is the case with the earlier studies, the Revision 2 study *also included the thermal-hydraulics and source term analysis of the reactor coolant and containment systems (known as the Level 2 analysis) and the analysis of the public health risk by considering the population (1990 census data) and meteorological data (known. as the Level 3 analysis). Furthermore, the external event analysis presented in this study is cqnsistent with the ~EEE results reported to the NRC in response to Supplement 4*of NRC GL 88-20.

As is the case with the previous revisions, the reliability data, core melt frequency, and risk results presented in this PRA are for an "average" plant configuration, representing

.the "at-power" and hot shutdown conditions. For the cold shutdown and refueling modes, a separate PRA (based on the ORAM methodology) is utilized at this time.

iii

The data, results, and conclusions presented in this report supersedes those presented in the 1989 PRA/IPE report. Nevertheless, the three volume PRA/IPE report can be used to .

~~tain general information on plant systems and their arrangements.

The 1995 update utilized five PRA engineers and two technical support personnel. In addition, two Project Managers (Steve Deskevich and Mike Barrett) provided the project coordination and technical management functions of this effort.

P. M. Abraham, Section Manager Severe Accident Analysis

. January 1997 iv

TABLE" OF CONTENTS 1.0 SUMM.ARY **!****.*********************************************************************************************************** l-:l 1.1 Introduction ........................................_. .................................................................:..... 1-1 1.2 Core Damage Frequency Development... ................................................................... 1-2 1.3 Containment Perforrnance ..........................'. ............................................................... 1-4

  • 1,4 Public Health ~onsequences .. :...,............................................................................... 1-5 1.5 Results ........................................................................................................................ 1-6

2.0 INTRODUCTION

.................................................................................................... 2-J

  • 2.1 Objective ...................'. ...:.. **************'******'. ................:........................................:.......... 2-l 2.2 Plant Des.cription *****:************** ..************ ..:.......................................:............................ 2-2 2.3 Intended Use Of This Report......................................................................*................ 2-3 2.4 General Methodology ............................:..........................................:......................... 2-4 3.0 CORE DAMAGE FREQUENCY DEVELOPMENT ........................................... 3-1 3.1 Data Development ....................:.......................................................................:......... 3.-1 3.1.1 Equipment Data Collection **:************* .............................................. :.-............. 3-1 3.1.2 Initiator Development...*.......... .

3-2

. 3.1.3 Common Cause Development. ...................................................................... 3-12 3:1.4 Human Reliability Analysis ..........................................................................

3-13 3.1.5 Plant Specific Experience.............................................................................. 3-17 3.2 Internal Event Model Development ......................................................................... 3-18 3.2. l Systems Analysis ........................................................................................... 3-18 3.2.2 Accident Sequence Development.................................................................. 3-26 3.3 External Event Model Development ........................................................................ 3-34 3.3.1 Seismic Analysis ........................................................................................... 3-35 3.3.2 Internal Fire ................................................................................................... 3:-36 3.3.3 Tornado ...................................................*...........................;.......................... 3-36 3.3.4 External Flood ............................................................................................... 3-37 3.4 Integrated Plant Model ............................................................................................. 3-39 3.5 Methodology 3-41 4.0 CONTAINl\fENT PERFORMANCE ..................................................................... 4-1 4.1 Containment Event.Tree (CEn Model Development ********.***************************************4-l 4.2 Data Development ................................... 1**** : ******* *** ********* *** ********** ********* ********* **** ******* 4-5

  • 4.3 Methodology .......................................................:.. :.....:.............................................4-7 5.0 PUBLIC HEALTH CONSEQUENCES .....;.......................................................... 5-1 5.1 Model Development ........... ,........................... :~.......................................................... 5-l V

LIST OF TABLES 3.1.l-l Equipn,ent Failure Rates 3.1.2-1 .Final Initiating Event List 3.1.5-1 Plant-Specific and Bayesian-Updated Failure Rates 3.5-1 Core Melt Bin Designations 3.5-2 Containment Safeguards Designations 4.3-1 Oconee Release Category Cross Reference 6.1-1 . Summary of Core Damage Frequency Results 6.1.3-1 Top 100 CDF Cut Sets for Internal Initiators 6.1 .3-2 Top l 00 CDF Cut Sets for External Initiators

6. l.3-3 Basic Event Importance Listing 6.2-1 Summary of Containment Analysis Results 6.3-1 Summary of Risk Results for Internal I.niriators 6.3-2 Summary of Risk Results for External Initiators 6.3-3 Summary of Risk Results for All Initiators vii

LIST OF FIGURES 3.2.l-l Oconee Reactor Coolant System 3.2.1-2 Oconee Reactor Coolant System - Nonna! Flowpa~

3.3.2-1 Small LOCA Event Tree 3.3.2-2 Medium LOCA Event Tree 3.3.2-3 Large LOCA Event Tree 3.3.2-4 Transient Event,.Tree 3.3.2-5 Steam Generator Tube Rupture Event Tree 4.1-1 Oconee Containment Event Tree 6.1-l Percent Contribution to Core Damage Frequency by Initiator 6.l-2 Probability Density Distribution 6.1-3 Cumulative Probability Distribution

\.

ix

TABLE OF CONTENTS 1.0 SUJ\,WARY**~************************************************************************************************************ 1-:1 1.1 Introduction ..........................................................................................................:..... 1-1 1.2 Core Damage Frequency Development... ................................................................... 1-2 1.3 Containment Performance .......................... :............................................................... 1-4 l ,4 Public Health Consequences .. :***'*********'. ..................................................................... 1-5 1.5 Results ........................................................................................................................ 1-6

2.0 INTRODUCTION

.................................................................................................... 2-1 2.1 Objective ...................*.... '. ****************************************: .................................................... 2-l 2.2 Plant Description *****:*****************************:***************************************:............................ 2-2 2.3 Intended Use Of This Report.. :********************~**********************************************~***************2-3 2.4 General Methodology ............................:..........................................::........*............... 2-4 3.0 CORE DAMAGE FREQUENCY DEVELOPl\-fENT ................................- .......3-1 3.1 Data Development .............................................................................................:......:.. 3-1 3.1. l Equipment Data Collection ..:*****************.. *************** ......................... .'............... 3-l 3.1.2 initiator Devclopment.............. ;.......................................................................

3-2

. 3.1.3 Common Cause Devclopmcnt.. ..................................................................... 3-12 3:1.4 Human Reliability Analysis .............................................................~............ 3-13 3.1.5 Plant Specific Experience.............................................................................. 3-17 3.2 Internal Event Model Development ......................................................................... 3-1 S-3.2. l Systems Analysis ........................................................................................... 3-18 3.2.2 Accident Sequence Development. ................................................................. 3-26 3.3 External Event Model Development ........................................._. .............................. 3-34 3.3.l Seismic Analysis ........................................................................................... 3-35 3.3.2 Internal Fire*................................................................................................... 3-36 3.3.3 Tornado .........................................:...............~ ............................................... 3-36 3.3.4 External Flood .................................................:............................................. 3-37 3.4 Integrated Plant Model ............................................................................................. 3-39 3.5 Methodology ............................................................................................................ 3-41 4.0 CQNT~NT PERFORMAN'CE ******************-**************************************************4-1 4.1 Containment Event.Tree (CET) Model DevelopmeQt ........_. ...................................... 4-1 4.2 Data Development ..............................................................................:*************************4-5

  • 4.3 Methodology ........................................................*........*.............................................. 4-7 S.O PUBLIC IIEALTH CONSEQUENCES ................................................................ 5-1 5.1 Model Development ...........:....................................................................................... 5-l V

5.2 Data Development ................................:************************************************************'********5-2 5.3 Methodology ....................................... :...................................................................... 5-3 6.0 ~SULTS .................................................................................................................. 6-l

6. I Core Damage Frequency ..............................-.............................................*................. 6-2
6. I. l Internal Events.......:................................................................. ,............... .-....... 6-2
6. I .2 External Events ....................:..........................................................................6-7 6.1.3 Total ..............:................................................................................................. 6-9 6.1.4 Sensitivity Studies On CDF Results........_.........................................,. ............ 6-13 6.2 Containment Performance ..........................................................*.............................. 6-18
6.3 Public Health Risks .....................................................
............................................ 6-22

7.0 REFERENCES

.......................................................................~................................. 7-1 Appendix A SYSTEM MODEL SUMMARIES Appendix B EXTERNAL EVENTS ANALYSIS -*

vi

LIST OF TABLES .

3.1 .1-1 Equipn,erit .Failure Rates 3.1.2-1 . .Final Initiating Event List

  • 3.1.5-1 . Plant-Specific and Bayesian-Updated Failure Rates 3.5-1 Core Melt Bin Designat_ions 3.5-2 Containment Safeguards Designations ,

4.3-1 Oconee Release Category Cross Reference 6.1-1 . Summary of Core Damage Frequency Results 6.1.3-1 Top 100 CDF Cut Sets for Internal Initi.ators .

6.1.3-2 Top 100 CDF Cut Sets for External Initiators 6.1.3-3 Basic Event Importance Listing 6.2-1 Summary of Containment Analysis Results

. 6.3-1 Summary of Risk Results for Internal I.~iti_ators 6.3-2 Summary of Risk Results for External Initiators 6.3-3 Summary of Risk Results for All Initiators vii .*;'

LIST OF F1GURES 3.2.1-1 Oconee Reactor Coolant System 3.2.1-2 Oconee Reactor Coolant System - Nonna! Flowpa~

3.3.2-1 Small LOCA Event Tree 3.3.2-2 Medium LOCA Event Tree 3.3.2-3 Large LOCA Event Tree 3.3.2-4 Transient Event_.Tree 3.3.2-5

  • Steam Generator Tube Rupture Event Tree 4.1-1 Oconee Containment Event Tree 6.1-1 Percent Contribution to Core Damage Frequency by Initiator 6.1-2 Probability Density Distribution 6.1-3 Cumulative Probability Distribution ix

Acronyms AC Alter11ating Current ATWS . Anticipated Transient Without Scram BWR Boiling Water Reactor BWST - Borat~d Water Storage Tank * .

CA.FTA Computer Aided Fault Tree Analysis CCDF . Co~plimentary Cumulative Distribution Function ccw Condenser Circulating Water CET Containment Event Tree

~IS Containment Isolation State CMB* Core Melt Bin

EFW Emergency Feedwater EOF Emergency Offsite Facility EP Emerg~ncy Procedure EPRI *Electric Power Research Institute ESFAS .. Engineered Safety Features Actuation System EWST Elevated Water Storage Tank FSAR Final Safety Analysis Repo~

GI Generic Issue GL Generic Letter GSI Generic Safety Issue HPME High Pressure Melt Ejection HPI High Pressure Injection HPR High Pressure Recirculation .

HPSW High Pressure Service W,a~er HVAC

  • Heating, Ventilation, and Air Conditi_oning_

I&C Instrumentation and Control ICS Integrated Control System Xl

Acronyms ( contd. )

-IDCOR Industry Degraded Core Rtilemakihg

  • Individual Plant Examination for Extern~ Events

-ISLOCA * .Interfacing-Systems

. Loss of Coolant Accident*.

kV Kilovolt

.I LL Large LOCA .

LOCA . Loss of Coolant Accident LOOP Loss of Offsite Power LPI

  • Low Pressure Injection

. Service.Water.

MAAP Modular Acddent Analysis Program MFW Main Feedwater

  • ML MediumLOCA MOV Motor-Operated Valve
  • MWt Megawatt thermal
  • NRC . Nuclear Regulatory Conunissiµn NSSS Nuclear Steam Supply System*

osc Operations Support Center Pns* Plant Damage State PORV . Power-Operated Relief Valve PRA Probabilistic Risk Assessment PSIG* Pounds per Square Inch Gage*

PWR Pressurized Water Reactor RBCU Reactor ~uilding Cooling Unit RBS Reactor Building Sprays RC Release Cat~gory RCM Release Category ~atrix

RCS Reactor Coolant System RPS *

Acronyms ( contd'. )

SRV Safety Relief Valve SSE Safe Shutdown Earthquake SSF Safe Shutdown Facility SSS Standby Shutdown System TDP Turbine Driven Pump TMI Three Mile Island TSC Technical Support Center USI Unresolved Safety Issue xiii

. Reactor Pressure Vessel Failure (b)(7)(F), (bX3):16 U.S.C. § 8240-l(d)

Flooding (internal)

The calculated .core damage frequency due to internal flooding is approximately .

9.SE-06/yr. This accounts for approximately 10% ~f the total CDF and 37% of the internal *coF. \

(b)(7)(F), (bX3)16 U.S.C. § 8240-l(d)

I 6.1.2 External Events Approximately 71 % of the calculated annual core damage frequency for Oconee Unit 3 is attributable to external initiating e~ents. The dominant component failures and/or human errors required to produce a core melt for each initiating event are described in this section. The calculated annual core damage frequency due to external events is appr"ximately 6.3E-0S/yr.

6-7

Seismic Events The calculated annual core damage frequency resulting from a seismic event is approximately 3.9E-05/yr, This initiating event accounts for 44% of the total CDF and 62% of the external CDF. \

(bX7)(F). (bX3):16 U .S.C. § 8240-l(d)

I Tornadoes The calculated annual core damage frequency resulting from tomados impacting Oconee is approximately l .4E-05/yr: This initiating event accounts for 16% of the total CDF and 21% of 'the external *CDF. l (b)(7)(F), (bX3):16 U .S.C . § 8240-l(d)

I Flooding (external)

The calculated core damage frequency due to external flooding is approximately

~f P1 ?i~;!i~~> This accounts for approximately 7% of the total CDF and 9% of the external CDF.I I..

(bX7)(F), (bX3)1 6 U S C . § 8240-l (d)

  • Fires The calculated annual core damage frequency due to *fire is approximately 4.SE-06/yr. This accounts for approximately 5% of the total CDF and 7% of the 6~8 .

SEN Si 11 vE SEC0Rl'f i -tt:J!LA'f!!I:> ttqt-OR:MA'ftOM cru 1iCAt f!tqf!ft6 i J!!t!!C'fltICAL UqPM~'fltUC'fURE Uqt-OlUt'IA'ffOfq

Table 6.1-1 Rev. 2

SUMMARY

of CORE DAMAGE FREQUENCY RESULTS Initiating Event Frequency Total Frequency Inte* .....1 2.6E-05 (bX7)(F), (b)(3)16 USC. § 8240-l(d)

Extern~

. 6.3E-05 Seismic 3.9£-05 Tornado l.4£-05 CbX XF), (bX,):

External Flood U.S.C. § 8240-l(d)

Fire 4.SE-06 Total Core Damage Frequency 8.9E-05

Figure 6.1-2 Rev. 2 C:\SS3\0R2RMOS.CUT Page : 2 Oate  : 6126196 PROBABILITY DENSITY DISTRIBUTION M

  • Mean 9.0BE--05 I . so;. 3.18E-05
  • . 50% 6.91E-05 I
  • 95% 2 .12E-04

~tandard Deviation 8.S0E-05 Skewness 7.07E+OO Kurtosls 9.19E+01 Sample Site 5000 D

8 11 s

i I

y 1.0E-04 1.0E-03 Frequency

Figure 6.1-3 Rev. 2 Page . : 3 C~\OR2RMQS.CUT Date : 6126196 CUMULATIVE PROBABILITY DISTRIBUTION M - Mean 9.0SE-05 *

{

  • S'Yo 3.18E-05
  • . 50% 6.91 E-05 I - 95% 2.12E-04 Standard Oevtation 8.SOE-OS Skewness 7.07E+OO Kurtosis 9.19E+01 Sample Size 5000 1.0 C

u 0.9 m

u 0.8 I

a t 0.7 I

V 8 0.6 p

r 0.5 0

b 0.4 a

b i 0.3 I

r I 0.2 y

0.1 0.0 1.0E-04

  • 1.0E-03 Frequency

7.0 REFERENCES

Section 2.0 2.1 PRA Procedures Guide. NUREG/CR-2300, Office of Nuclear Regulatory Research, U.S . Nuclear Regulatory Commission, Washington, D.C., January 1983.

Section 3.0

3. 1 SAROS Generic Equipment Failure Rate Database, SAAG File No. 342, Duke Power Co., November 1995.

3.2 Licensed Operating Reactors Status Summary Report, NUREG-0020, Volume 14.

U.S. Nuclear Regulatory Commission, Washington, DC, January 1990.

3.3 Losses of Off-Site Power at U.S. Nuclear Power Plants -Through 1995. EPRI TR-106306, Electric Power ~esearch Institute, P~lo Alto, CA; April 1996.

3.4

  • Li~nsee Event Report 50-261/75-9, H. B. Robinson Unit 2, May 1975.

3.5 NSAC/60. A Probabilistic Risk Assessment of Oconee Unit 3, Duke Power and the Nuclear Safety Analysis Center. Palo Alto. CA. June 1984.

3.6 Kitzmiller. J. T. and Frost. D. R. Westinghouse Owners Group Trip Reduction and Assessment Program, Westinghouse Inadvertent Plant Trip

  • Experience January 1987 through December 1987, WCAP 11779. Westinghouse Electric Company, Pittsburgh, PA, March 1988.

7-1

3.7 PRA Procedures Guide, NUREG/CR-2300, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, D.C., January 1983.

3.8 More, J. H., et al., Value Impact Analysis of Recommendations Concerning Stearn Generator Tube Degradations and Rupture Events, Science Applications Inc. McLean, Virginia, 1983.

3.9 Advanced Light Water Reactor Requirements Document. Appendix A: PRA Key Assumptions and Groundrules, Rev. 1, Electric Power Research Institute, Palo Alto, CA, August 1990.

3.10 Evaluation of Sta~on Blackout Acciden~ at Nuclear Power Plants, Technical

  • Findings Related to Unresolved Safety Issue A-44, NUREG-1032, U.S. Nuclear Regulatory Commission, Washington, DC, May 1985.

3.11 Proceedings: Main Coolant Pump Diagnostics, EPRI NP-6116, Electric Power Research Institute, Palo Alto, CA, December 198K 3.12 Licensee Event Repon 3~3-80015, Arkansas Nuclear One - 1, May 1980.

3.13 Pipe Break Frequency Estimation for Nuclear Power Plants, NUREG/CR-4407, U.S. Nuclear Regulatory Commission, Washington, DC, May 1987.

3.14 Reactor Safety Study -- An Assessment of Accident Risks in U.S. Comm~rcial

, ~uclear Power Plants, WASH-I ~00, NUREG-75/014, U.S. Nuclear Regulatory Commission, Washington, DC, 1975...

7-2

3.15 Swain, A. D., H. E. Guttmann, Handbook of Human Reliability Analysis with

'\

Emphasis on Nuclear Plant Applications Final Report. Sandia National Laboratories, NUREG/CR-1278, August, 1983.

3.16 Hannaman, G. W., Spurgin, A. J., Lukic, Y. D., Human Cognitive Reliability Model for PRA Analysis, NUS-4531, December 1994:

3.17 Parry, G. W., Lydell B. 0 . Y., An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment. Halliburton NUS Environmental Corp, Gaithersburg, Maryland, EPRI-TR-100259, October 1992.

  • 3.18 Vlahopolus, C:, et al:, A Loss ofOffsite Power Recovery Model. Anticipated and Abnormal Transients in Nuclear Power Plants Topical Meeting, Atlanta, Georgia.

April 1987.

3.19 A. Mosleh, et al., Procedures for Treating Common Cause Failures in Safety and Reliability Studies. Vol. 1, NUREG/CR04780 (EPRI NP-5613), January 1988.

7-3

J APPENDIXB EXTERNAL EVENTS ANALYSIS

This page left intentionally blank B-6

B.O EXTERNAL MODEL DEVELOPMENT The original Oconee PRA/IPE report and its subsequent update performed an evaluation of external events, with four events identified for a detailed review:

  • Seismic Events
  • Fires
  • Tornadoes
  • Floods A variety of methodologies were employed to derive the .overall *frequencies for these events. The analyses are summarized below and are explained in detail in Oconee's Individual Plant Examination for External Events (IPEEE) report (Ref.

B.l).

B-7

LIST OFTABLES i:s.3-1 -Tornado*Data B.3-2 Piping Damaged By Tornado-Induced West Penetration Room Failure B.3-3 Conditional Probability Of Damage By Tornado-Generated Missile B.3-4 . _Keowee:Conditional Failure Probabilities B.4-1 Historical Dam Failure Events B.4-2 Chronological Order Of Dam Failures B-4

8.1.2

  • Plant Information The Oconee systems and components which are essential to the prevention or mitigation of accidents* which could affect the public health and safety were designed to enable the facility to withstand the effects of natural forces including earthqµakes. The plant was designed to withstand both a Design Basis Earthquake (DBE) and a Maximum Hypothetical Earthquake (MHE). The

\

structural design criteria for the MHE was based on 0.10g peak ground acceleration (PGA) for Class I structw:es founded on rock and 0.15g PGA _for Class 1 structures founded on overburden. The structural design criteria for the DBE was 0.05g PGA.

(b)(7)(F), (bX3)16 U S.C § 8240-l(d)

(bX7)(F), (bX3):16 U.S.C . § 8240-l(d)

(bX7XF), (bX3):16 U.S C § 8240-l(d)

B-9

'.bX7)(F), (b)(3) 16 U.S.C. § 8240-l (d)

(bX7)(F), (bX3) 16 U S.C. § 8240-l (d)

~I iel

~

  • I l'I

.i l!I It, It.

!!I It ltJ Ill 1'1111 (bX7)(F), (bX3) 16 U.S.C. § 8240-l (d) r X7)(F), (bX3):16 U.S.C. § 8240-l(d)

B- 10

B:t.3' *Information Sources A structurai anaiysis consultant, Structural Mechanics Associates, was used to

.develop the structural and equipment fragilities for the original fragility analysis.

For the most part, results of existing analyses and evaluations of structures and equipment for the Oconee plant were utilized in this study. As part of the evaluation, some limited analysis based on original design analysis loads was conducted to determine the expected seismic capacities of the important structures. The following bullets list the important sources used for the original fragility analysis:

  • existing design basis dynamic analyses
  • Final Safety Analysis Report
  • design reports
  • United States Corps of Engineers shock test reports
  • specifications on the design or'equipment
  • seismic qualification test reports B.1.4 ~alkdowns Plant walkdowns are considered to be an important part of the seismic risk assessment. In support of this assessment, a number of walkdowns were conducted. Walkdowns were performed to support the developm~t of the initial Oconee PR.A/IPE which included external events. The initial study was completed in 1984. Walkdowns were also conducted to support revision 1 of the Oconee PRA/IPE, submitted in 1990. As a part of the IPEEE effort, extensive
  • . walkdowns were . conducted in -1994 and 1995 consistent

. with

. the intent of the guidelines described in E_PRI NP~604 l (Ref. B.2).

B-11

Defmition of Failure For purpose~ of this study, seismic Category l and non-Category 1 structures are considered to have failed when inelastic deformations of the structure under seismic load potentially interfere with the operability of equipment attached to the structure. These limits on inelastic energy absorption capacity (ductility limits) are estimated to correspond to the onset of significant structural damage,' not

_necessarily structure co~lapse. PiJ>ing, : as well as electrical, mechanical, and electro-mechanical equipment vital to mitigating the effect of earthquakes are considered to fail when they can* no-longer perform their designated functions.

Relay chatter is an example of a functional failure for an electrical component.

Also, rupture of pressure boundaries are considered failures. For active equipment, the functional failure definition usually governs as equipment pressure boundaries are usually very conservatively designed for equipment such as pumps and valves.

Evaluation of Component Fragilities and Failure Mode The seismic capacities of most plant structures and components were developed by Structural Mechanics Associates for the original PRA. That study (Ref. B.3)

  • gives a detailed description of bow the seismic capacities were derived. The seismic capacities of the Keowee and Joc~see Dams were developed by Dr.

oan*iel Ven_eziano of MIT, a consultant to Law Engineering Testing Company.

The results of that study are reported in Reference B.4. References B.3 and B.4 were published in*1981 and included in NSAC 60, the original Oconee PRA/IPE.

References B.5 and B.6 are later revisions to some of the original capacities based on ad~itional data and a, mor~ in-depth evaluation. The seismic capacities are presented ~ the final form of fragility curves, which express the _conditional probability of failure as a function of ground acceleration. Components and structures with a median fragility greater than 2.0g are screened out of the model.

  • Not all components are explicitly modeled in the PRA. Many components were B-12

effects of this PMP on reservoirs and spillways were evaiuated in a study performed by Duke Power Company in 1966 (Ref. 8.4 l ). The results of this study demonstrated that the Keowee and Jocassee reservoirs are* designed to contain and control the floods that could result from a PMP. Thus, in order to flood the plant site, rainfall exceeding the PMP must occur. The frequency of exceeding the PMP was obtained from the analysis presented in Ref. 8.8 for the Oconee site and was used as a bounding estimate of the frequency of core damage due to rain-induced external flooding.

The analysis yields the following frequencies of a PMP associated with the lower, median, and upper bounds of the probability distribution:

Cumulative Probability PMP_Frequency (per yr) 0.05 4.9E-8

  • 0.50 2.9E-7 0.95 8.9E-7 The cumulative probability for a particular frequency interval is to be interpreted as the degree of certainty that the PMP frequency observed over a long period of ti~ will be less than or equal to the upper valu~ of that frequency interval.

_T he SSF is equipped with ] ft.. flood barriers at its two entrances, and has etherwise been made impervious to site floodin . Therefore, the SSF would b~

~vailable to mitigate external flooding sequences .$ '~ fti Applying the SSF as a response to exceeding the PMP results in a calculated core damage frequency approximately an order of magnitude lower than the mean PMP frequency. In addition, the calculated.mean frequency of the PMP is more than an order of magnitude less than that due to a random failure ofJocassee Dam, which

  • would also flood the site. For both these reasons, it is concluded that SENSITPiE SECURITY R£L:it.:TEE> INFOR~fATION CRITICJ,b ENERCWEbECTRICAb INFRl,S+RUCT\::JRe l~*FOR).fATIO~

8-57

precipitation-induced external flooding is a negligible contributor to core damage frequency and public risk.

External Flooding From Dam Failure The Oconee site has a yard grade elevation a few feet below the full-pond level of Lake Keowee, which serves as the source of i(s condenser circulating water. Lake Jocassee has a full-pond elevation about 300 feet above Lake Keowee. If a sudden failure of the Jocassee Dam were to occur, and a rapid enough *release of the impounded water from Lake Jocc1ssee into Lake Keowee resulted, the flood wave generated in Lake Keowee would over-top the Keowee Dam and the Oconee intake dike*, flooding the plant. This section presents ~e ~lysis performed to estimate the frequency of such a flood.

B.4.3 Frequency of Dam Failure The Jocassee Dam is an earth-rockfill structure approximately 400 feet high. The

  • dam was completed in l 972~ and the reservoir was filled by April 1974. The spillway lies* along one of the abutments, about one-quarter of a mile from the

.. dam, and is a concrete structure founded on* granite.

An analysis was performed to**determine an annual frequency of failure for earth, earth-rockfill, and rock.fill ~ s due *to events other than* over-topping and earthquake ground shaking, which were considered in separate analyses. . Also, based on dam design information, structural failure of the spillway during discharge and failure associated with seepage along an outlet works have been eliminated as a possible failure mechanism.

B-58

The following principal modes of failure were considered: .

l. Piping
2. Seepage
3. Embankment slides
4. Structural failure of the foundation *or abutments These failure mechanisms are referred to collectively as random failures. Only failures resulting in the complete collapse of the structure and the uncontrolled release of the reservoir's contents were considered to have the poten ti~l .for flooding the Oconee plant.
  • Previous investigations into the frequency of dam failure indicate that it decreases with later years of construction (Ref. B.42). This is generally attributed to improvements in the methods of design and construction. Therefore, another criterion considered in developing a data base and failure-frequency estimate was the period of construction.

The age of a dam is another factor that has been identified as having an effect on the rate of dam failure. . Approximately half of all dam failures occur during the ftrst 5 years of operation (Ref. 8 .49). Therefore, age was also consi4~red in developing a data base.

Size, type of construction, realistic failure modes, period of construction, _and age were the major considerations used to define a data base for use in estimating the failure frequency of the Jocassee Dam.

B-59

The data base characteristics that were attributed to the Jocassee Dam are:

~baracteristic Jocassee Dam Location (country) United States Year completed 1972 Age (years in operation) 21 Height (ft) 400 Type Earth-rockfill Data Various catalogs were used to develop the data groups that were studied (Ref.

B.43). Each group _consisted oflarge earth, earth-rockfill, or rockfill dams (more than 45 feet high- Ref. B.44) in the United States that were in operation 6 or more years when they failed. Of the references used in this study, none were a complete catalog, and therefore* they were used collectively. At present, these references represent the best readily available information. Fr~m the various listings of failures, cross-checks were made when possible.

A data base uniquely suited in every major respect to the Jocassee Dam was unattainable because of a scarcity of the number of the earth-rockfill type.

The data base ultimately devel_o ped reflects discussions with Duke Power engi-neers familiar with the characteristics of the Jocassee Dam. It was decided that the data base should include only the failure modes that could occur at Jocassee.

The two major failure types excluded from the data set were failures resulting from piping at a conduit passing through the dam and structural failures of the spillway during the flood discharge. Neither_of these failures can occur because the necessary physical conditions do not exist at Jocassee. (Note, however, that B-60

dams in the data base do ,include those that _can fail in either or both of these modes. This is *proper because the experience from these dams represents realizations of non-failure for other failure modes, such as embankment piping, foundation failure, and slope failure.)

Because of limit:ations in the historical.record, it is possible only to develop a data set _that takes into .account a limited number of the specific properties of the Jocassee Darn: . S_ince Jocassee is a structure desi~ed and constructed in recent times, it can be assumed that state-of-the-art technology was used in its design. In addition, because of its role as*a critical facility;of large size and importance, other aspects of the dam, such as seepage monitoring and inspection programs, are important factors that .decrease the likeliho~. ~at the dam will: fail. The following specific characteristics of Jocassee are identified as relevant factors that will affect the frequency of failure:

.* . Quality maintenance and inspection_pro~

.* . Monitoring of die dam (i.e., seepage, settlement, etc.)

  • ~sence ofpersonnel at the site
  • Responsiveness of the owner to poteiltial..problems. (i.e., implementation of

..emergency plans)

.* Detailed geologic inve*gations conducted before site selection *

  • Experience of earth-rockfill dams in.Jocassee's class (with respect to random failures)
  • Design 'techniques These factors notwithstanding, the data were examined, and the best available

.data base for application to Joc~see Dam was extracted and used.

The data base covered the period of dam construction from 1940 to 1987. During

-this period, U.S.

  • dams in operation 6 or more years at the time of failure and 45 B-61

feet or more in height ~ ere conside~el The dams were of three types--earth, earth-rockfill, or rockfill-and only catastrophic* fa;tures were included.

Table B.4.:1 lists the failures considered in the analysis:

The number of dam-ye~ of operation was determined from da_ta on the rate of .

c~nstruction provided 'in *Refs. B.44 ~d* B~4S. Table B.4~2 lists chronologically for the period of construction the year a failure occurred, the interval.*between each failure in years, the cumulative number of dam-years to *the year of failure,

  • and the riumber of dam-years between failures.

Ref. B.46 updated this information by taking into account the additional amount of dam-years from 1988 through 1993. . As shown therein, there were no additioW!-l dam failures meeting the above criteria during this period. The revised analysis indicates that th~ revised random failure frequency for th_e class of large

. U.S. built earthen dams meeting the above criteria is 1.3E-5. This number is a point estimate using the two

  • failures shown in* Table B.4-1 and the estimated 154,380 dam-years of operation from Ref. B.46 and Table B.4-2 (Le:; 2/154,380 =

l.3E,.5). To the e;ic.tent that this class is representativ~ of the Jocassee Dam, these results* can be interpreted as the predicted annual random failure frequency of the Jocassee Dam from causes other than earthquakes or over-topping.

Ii1 order to evaluate the contribution of random failure of J ocassee Dam to the frequency of core damage at_Oconee U~it 3,-~ factors _must be q~antified. The first factor is the frequency of random dam failures, calculated above. The remaining two factors are:

  • The conditional probability of flooding of the Oconee site given a failure of the Jocassee Dam
  • The conditio~l*probability of core damage given flooding at the Oconee site.

I * ~ I *I

  • The conditional orobabilitv of flooding at the Oconee site given a failure of the.

Jocassee Dam is a comolex function of several variables. [The level of th~

1ocassee Lake at the time of failure determines the amount of imoounded watei

~at will be released into Lake Keowee; the level of Lake Keowee determines ho~

much flow can be received from Lake Jocassee before Keowee Dam and th~

Dconee intake dike are over-topped The time available for warning of an

[!npending dam failure will determine the likelihood that effective action C.!Jl b~

taken to lower the levels of the Jocassee and Keowee *Lakes and whether an action ike notchin the east side of the Keowee Dam allowing the flood to bvoass th~

'pconee site. could be conceived and imolemented. The conditional probabili!}'_Qf flooding given a dam failure also depends on the mode of failure, the rate of erosion (time to failure). the vertical depth to which the failure penetrates (fraction pf the impounded water released).. and whether either Keowee Dam or the eas1 intake dike fails ranidlv when over-tonoed (allowin2 the flood to bvoass ~

bconee site}l Previous memos have been writt~ (Ref. B.47) to address this phenomenon.

There is evidence, based on rock-filled dams that have either failed or have been

..saved from failure, that pre-failure damage in the form *of . seepage would be *

  • observable for some time .before significant failure occurs. Even *.though
  • significant warning times would be needed to prevent Keowee over,-topping, the overall resulting flood levels could be reduced by lowering lake levels as quickly as possible via the spillway and the generators.

The data base includes only catastrop_hic failures by modes believed applicable to Jocassee. Most of the catastrophic failures reported in the literature for earth or earth-rockfill dams were total (i.e., the entire reservoir emptied), and most times to failure "'ere in the range of 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. However, the failures in the data base (Table B.4-1) occurred only in earthen dams; thus, there is some question as to B-63

how applicable their rates of erosion and depths of failure are to .the Jocassee

. Dam, an earth-rockfill structure.

The Hell Hole Dam, an earth-rockfill structure, failed during construction in 1964 when extreme precipitation caused ove*r-topping. Since construction had not been completed and since the dam was over-topped, it did not satisfy the data base characteristics and was not included in the failure set. However, its material of construction and size were very similar to those of Jocassee, and its failure behavior can therefore be used as one point of reference in judgments about the possible failure behavior of Jocassee. lbe over-topping of this partly constructed dam lasted for more than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> before a catastrophic failure occurred. This suggests that a long warning time may be available, at least in sonic cases, before the failure of an earth-rockfill dam. Once the dam was breached, however, the breach -propagated the full depth of the dam, down to the foundation, in about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Thus, though more warning time may have been available, once failure began, the time and depth to complete breach were quite similar to those observed for earthen dams.*

  • (bX7)(F), (b)(3):16 U.S.C. § 8240-l(d)

~X7)(F), (b)(3)16 U.S.C. § 8240-l(d)

B-64

(bX7)(F), (bX3):16 U.S.C § 8240-l(d)

B.4.4 Event Tree The external flooding event tree is shown in Figure B.4-1. It is based on the internal initiator tree and contains many of the same functional top events. Refer to the IPEEE for a detailed discussion of these events and the supporting logic for the event tree.

B.4.S Limitations of the Analysis The unpredictability of actual flood levels has been discussed above. I SBIIICE(I (bX7)(F), (b)(3) 16 U .S C . § 8240-l(d)

As also discussed previously, the random failure frequency of the Jocassee Dam is based upon known failures of similarly constructed dams. . To the extent that the set of criteria specified is representative of the Jocassee Dam, this value can be interpreted as the predicted annual random failure frequency of the Jocassee Dam from causes other than earthquakes or over-topping.

B.4.6 Results and Conclusions The estimated external flooding core damage frequency from this analysis is SRltClill 7.SE-6/yr. f (b)(7XF), (b)(3) 16 U .S.C. § 8240-l(d)

Sl!!~SITIVI! Sl!CUltlTY-ltl!tKTl!f) n~POltMKTirn~

CRl'ffC!tL ENERffDELECTRf~ INFRASTRUCTURE UffORt.f!tTIO~~

- r X7)(F), (b)(3):16 U.S.C. § 8240-l(d)

The Oconee external flood core damage frequency of 7.SE-6 is of the same

  • magnitude as other potential severe accidents such as seismic events, fires, tornados, and other events.

B.4.7 Insights

  • 10,)())(F), 0,)(7) 16 U S.C. § 8240-l(d)

SRlfCEII SJ!~SI I I V1! St!Ctittl f I -.llf!LA'fffl HqfiOitMPr'FtOH CBJIJCAT ENEBGY/fI ECIBICAJ INFB ASIBJTCIJTBF INFQBMAIJQN

B.5 REFERENCES B.l EPRI, NP-6041, Rev. 0 and Rev. l, "A Methodology of Assessment of Nuclear Power Plant *Seismic Margin," October 1988 and August 199 l.

8.2 Structural Mechanics Associates, SMA 12904.01, "Conditional Probabilities Of Seismic Induced Failures For Structures And Components For Oconee Generating Station Unit 3," September 1981.

B.3 Veneziano, "Seismic Fragility Curves For Jocassee Dam and Oconee Dikes,"June 1981.

B.4 *. Letter from R. V. Hester, Oconee Engineering Division, to T. F. Wyke, Engineering Support Division, Attention: K. S. Canady, July 16, 1990, File No.

OS-203. .

B.5 Letter from NTS Engineering, Long Beach, CA, to T.F. Wyke, Duke Power Company, August 27, 1986.

B.6 Documentation of the Seismic Eventlmpact Sequence Model (SEISM) Computer Code, PSA-84-17, Duke Power Company, September 1984.

B.7 A Probabilistic Risk.Assessment of Oconee Unit 3, NSAC/60, Electric Power Research Institute and Duke Power Company, Palo Alto, California, June 1984.

B.8 Kazarians, M. , and Apostolakis, G. , Fire Risk Analysis for Nuclear Plants, NUREG/CR-2258, UCLA-ENG-8102, School of Engineering and Applied Science University of California, September 1981.

B.9

  • Sumitra, P. S., Categoriz.ation of Cable Flammability Intermediate - Scale Fire Tests of Cable Tray Installations, EPRI NP-1881, Factory Mutual Research Corporation, Noiwood, Massachusetts, August 1982.

B. l O Berry, D. L , and Minor, E. *E. , Nuclear Power Plant Fire Protection - Fire - *

  • Hazard Analysis (Subsystems Study Task 4), Sandia Laboratories, NUREG/CR-0654, SAND 79-0324, Albuquerque, New Mexico, September 1979..

B.11 Herman, P., Guise, A. , Hall, R. , and MacDougall, E., Turbine Oil fires As Related To Nuclear Power Stations,'Brookhaven National Laboratory, BNL-

_NUREG-23316, Upton, New York, September 1977.

B.12 Duke Power Company McGuire Nµclear Station, Incident Investigation Report No. M87-014-l, Fire in Unit 1 Turbine Piping Insulation, March 1987.

B-67

B.13 Clayton, T. C., and Hall, D. T., "Turbine Generator Five Protection Overview,"

Paper presented at 12th Annual W A.Tiec Energy Conference and Exhibition, Kr,uxviile, Tennessee, February 1985.

B.14 Harbinson, L. T., Oconee Nuclear Station Appendix R LPI & LP2 Analysis Addition to "Memo to File 5/27/87," Memo to File in Duke file OS-72 dated February 7, 1988.

  • 8.15 Advanced Light Water Reactor Utility Requirements Document Volume II, ALWR Evolutionary Plant Chapter I, Appendix A PRA Key Assumptions and Ground Rules, Electric Power Research Institute; Palo Alto, California, 1992.

B.16 Wycoff, H., Losses ofOffsite Power at U. S. Nuclear Power Plants Through 1993, NSAC-203, EPRI, April 1994.

BJ 7 Holmes, Wayne D., A Methodology for the Assessment of Risk of Major Fire Loss in Multi- Unit Turbine generator Buildings,* SFPE TR84-9.

B.18 Duke Nuclear Station Modifications, NSM ON-12885, 22885, 32885.

B.19 "Tornado Occurrences Within 125 NM* Radius Of Oconee Nuclear Station,"

National Severe Storms Forecast Center, Kansas City, MO, August 1994.

B.20 Minor, J. E., "Applications Of Tornado Technology In_ Professional Practice,"

Proceedings of the Symposium on Tornadoes: Assessment of Knowledge and hnplications for Man. June 22-24, 1976;Texas Tech*Univers_ity, Lubbock, Texas, pp375-392.

B.21 .. McDonald, J. R., "Tornado-Generated Missiles and Their Effects," Proceedings of the Symposium on Tornadoes: Assessment of Knowledge and Implications for Man. June 22-24, 1976, Texas Tech University, Lubbock, Texas, pp 331-348.

B.22 Golden, J. H., "Comments in Session 1," Proceedings of the Symposium on Tornadoes: Assessment of Knowledge and Implications for Man, June 22-24, 1976, Texas Tech University, Lubbock, Texas, pp 483.

B.23 Fuji~ T. T., "Comments in Session 7," Proceedings of the Symposium on Tornadoes: Assessment of Knowledge and Implications for Man. June 22-24, 1976, Texas Tech University, Lubbock, Texas, p. 673.

B.24 Oconee Nuclear Station Procedure AP/3/A/l 700/11, Change 11, Loss of Power."

B.25 Oconee Nuclear Station Procedure AP/3/A/1700/19, Change 4, "Loss of Main Feedwater."

B.26

  • Oconee Nuclear Station Procedure AP/3/A/ 17.00/06, Change 0, "Natural Disaster." '* .)'.-*>,: * *. " *:*.'.>'..t;i~'>~*-

B.27 Ramsdell, J.V. and Andrews, G .. i..., "Tornado Climatology of the Contiguous United* States," NUREG/CR-4461, U. S. _Nuclear Regulatory Commission; Washington, DC, May 1986.

B.28 Ramsdell, J. V., et al., "Methodology for Estimating Extreme Winds for Pro~abilistic Risk Assessments," NUREG/CR-4492, U.S. Nuclear Regulatory Commission, Washington, DC, October 1986.

B.29 McDonald, J.R., "A Methodology for Tornado Hazard Probability Assessment,"

NUREG/CR-3058, U. S. Nuclear Regulatory Commission, Washington, DC, October 1983.

B.30 OSS:0254.00-00-1000, "Design Basis Specification for the Emergency Feedwater and Auxiliary Service Water Systems," Rev. 12, May 1, 1995.

B.31 Berkebile, B. H., Memo To File, "Wind Load Capacity of West Penetration Room Exterior Walls," Duke File No: 0S-203, May 31, 1990.

B.32 Kanipe, I;.'. M., "Calculation Of Tornado Strike Probabilities For Oconee Nuclear Station," SAAG File# 175, Duke Power Company, Charlotte, NC, March 1995.

8.33 Mccann, M. W. Jr., Jack Benjamin & Associates, "Wind Capacity of Oconee Nuclear Station Borated Water Storage Tank," July 26, *1982.

8 .34 Twisdale, L. A., et al., "Tornado Missile Simulation and Design Methodology,"

NP.2005, Electric Power Research Institute, Palo Alto, CA, August 1981.

B.35 Twisdale, L. A., et al., "Tornado Missile Risk Analysis," NP-768 and NP-769, Electric Power Research Institute, Palo Alto, CA, May 1978. *

8.36 Deskevich, S. A., "Verification of Computer Program TORMIS," COM-0204.C6-l l-0038 Revision 1, Duke Power Company, Charlotte, NC, October 1993.

B.37 Deskevich, S. A., "Verification of TORMIS ~nbancement," COM-0204.C6-ll-0039 Revision I, Duke Power Company, Charlotte, NC, October 1993.

B.38 Kanipe, L. M., "Damage Frequency of the Oconee Nuclear Station Emergency Feedwater System By _Tornado-Generated Missiles," OSC-3361 Rev. l, Duke

_Power Company, Charlotte, NC, November 1993.

B-69

B.39 Kanipe, L. M., "Keowee Tornado *Path Simulation Model," SAAG File # 174, Duke Power Company, Charlotte,.NC, April 1995.

  • B.40 Duke Power . Company, ~'Flood Study, Joc~see and Keowee Reservoirs,"

Charlotte, NC, 1966.

B.41 Baecher, G. B., M. E. Pate, and R. De Neufville, "Risk of Dam Failure in Benefit-Cost Analysis," Water Resources Research, Vol. 16, No. 3, pp. 449-456, 1980.

B.42 Benjamin, Jack R., and Associates, "Statistical Evaluation of the Frequency of Random Dam Failure," Prepared for NSAC, Palo Alto, CA; 1981 .

B.43 USCOLD, "Lesson from Dam Incidents USA-II," Am~rican *society of Civil Engineers, 1988.

B.44 Nash, J. A., Memo to File, "Update o.f the Random Failure Frequency of the Jocassee Dam," File No. OS-203, October 3,.198~. . ..

B.45 Farish, P. T., Memo to File, "Update of the Jocassee Dam Random Failure Frequency," File No. OS-203, March 15, 1995 B.46 .Lewis, S. R., Memo to File; "Evaluation of Jocassee Dam Failure," 7/2/82.

B.47 Farish, P. T., Memo to File; ~'Jocassee Dam Flooding Factors;" File No: OS-203; December 16, 1994 B.48 Be'njamin, Jack R., and Associates, "A Database for : the Evaluation of the Frequency of Random Dam Failure,. Report 120-010-01, Palo Alto, CA, 1982.

B.49

  • PRA Procedures Guide, NUREG/CR-2300, Volume I, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Co~ission, Washington, D.C.,

January 1983.

B-70

¥abt~ B.4-t Re~:*2 HISTORICAL DAM FAILURE EVENTS Dam Year Completed Year Failed Baldwin Hills a 1951 1963 Walter Boudin b 1967 1975 a Data from Babb and Mermel (1968) and USCOLD (1975).

b Data from Jansen ( 1980).

Table B.4-2 Rev. 2 CHRONOLOGICAL ORDER OF DAM FAILURES Period of Construction: 1940-1993 Year of failure Years between Cumulative Dam-years

. . failures dam-years between failures 1963 23 32,207 32,207 1975 12 74,782 42,575 1987 No failure 126,435 No failure 1993 No failure* 154,380 No failure

Figure B.4-1 Rev. 2 -- OCONEE EXTERNAL FLOOD EVENT TREE

~ .

(h)(7)(F), (h)(3) 16 U. S.C. § 8240-l(<l)

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