ML20209G756
| ML20209G756 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/28/1999 |
| From: | Kazarians M, Khatibrahbar, Sewell R AFFILIATION NOT ASSIGNED, ENERGY RESEARCH, INC. |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20209G748 | List: |
| References | |
| CON-NRC-04-94-050, CON-NRC-4-94-50 ERI-NRC-95-509, NUDOCS 9907190241 | |
| Download: ML20209G756 (68) | |
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ERI/NRC 95 509 TECHNICAL EVALUATION REPORT ON THE
" SUBMITTAL-ONLY" REVIEW OF THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS AT COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 FINAL REPORT l
Completed: December 1996 Final: February 1998 M. Khatib-Rahbar PrincipalInvestigator Authors:
R. T. Sewell, M. Kazarians' and M. Modarres' Energy Research,Inc.
P.O. Box 2034 Rockville, Maryland 20847 Work Performed Under the Auspices of the United States Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, D.C. 20555 Contract No. 04-94-050 9907190241 990708 DR ADOCK 050 45 3
Kazarians and Associates,425 East Colorado Street, Suite 545, Glendale. CA 91205
- University of Maryland, Depanment of Materials and Nuclear Engineering. College Park, MD 20742 Attachment 4
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TABLE OF CONTENTS EXECUTIVE S UMMAR Y............................................................ y PREFACE.........................................................................xi AB B REVIATIONS............................................................
1 INTRODUCTION...........................................................
1.1 Plant Characterization................................................... I 1.2
' Oversiew of the Licensee's IPEEE Process and Important Insights............... 2 1.2.1 S e is mic........................................................ 2 1.2.2' Fire.......................................................... 3 1.2.3 HFO Events.................................................... 3 1.3 Overview of Review Process and Activities................................. 4 1.3.1 Se i s mic.................................................
Fire.........................................................4 1.3.2
..... 5 1.3.3 HFO Ev en ts.................................................... 5 2
. CONTRACTOR REVIEW FINDINGS..................
Seismic..............................................................7 2.1
......... 7 2.1.1 Overview and Relevance of the Seismic IPEEE Process................. 7 2.1.2 Success Paths and Component List.................................. 8 2.1.3 Non-Seismic Failures and Hun p Actions........................... 9 2.1.4 Seismic Input.................................................. 9 2.1.5 Structural Responses and Component Demands......................
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-2.1.6 Screening Criteria.............................................. 10 2.1.7 Plant Walkdown Process........................................ 10 -
2.1.8 Evaluation of Outliers...........................................
12 2.1.9 Relay Chatter Evaluation........................................ 13 2.1.10 Soil Failure Analysis........................................... 13
-2.1.11 Containment Performance Analysis................................ 13 2.1.12 Seismic-Fire Interaction and Seismically Induced Flood Evaluations......
14 2.1.13 Treatment of USI A-45.......................................... 15 2.1.14 Treatment of GI-131............................................
15 2.1.15 Other Safety Issues............................................. 16 2.1.16 ' Peer Review Process............................................ 16 2.1.17 Summary Evaluation of Key Insights............................... 17 2.2 Fire.................................................................17 2.2.1 ~ Overview and Relevance of the Fire IPEEE Process................... 17 2.2.2 Review of Plant Information and Walkdown......................... 18 2.2.3 Fire-Induced Initiating Events.................................... 19 2.2.4 Screening of Fire Zones......................................... 20 2.2.5 Fire Hazard Analysis........................................... 21 2.2.6.
Fire Growth and Propagation..................................... 21 2.2.7-Evaluation of Component Fragilities and Failure Modes................ 22 2.2.8 Fire Detection and Suppression.........-.......................... 23 Energy Research,Inc.
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2.2.9 Analysis of Plant Systems and Sequences........................... 23 2.2.10 Fire Scenarios and Core Damage Frequency Evaluation................ 24 2.2.11 Analysis of Containment Performance.............................. 24 2.2.12 Treatment of Fire Risk Scoping Study Issues........................
2.2.13 USI A-45 Issue..................................
. 26 23 HFO Events........................................................ 27 23.1 High Winds and Tornadoes...................................... 27 23.1.1 General Methodology.................................... 27 23.1.2 Plant-Specific Hazard Data and Licensing Basis............... 28 23.13 Significant Changes Since Issuance of the Operating Li cense......................................... 29 23.1.4 Significant Findings and Plant. Unique Features 23.1.5 Hazard Frequency....................................... 2 29 23.1.6 PRA Analysis.........................................
External Flooding.............................................. 2 23.2
. 30 23.2.1 General Methodology................................... 30 23.2.2 Plant-Specific Hazard Data and Licensing Basis............... 30 23.23 Significant Changes Since Issuance of the Operating Li cen se........................................ 31 23.2.4 Significant Findings and Plant. Unique Features................ 31 23.2.5 Hazard Frequency...................................... 31 233 Transportation and Nearby Facility Accidents........................ 31 233.1 General Methodology.................................... 31 233.2 Plant-Specific Hazard Data and Licensing Basis............... 32 2333 Significant Changes Since Issuance of the Operating
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i License........................................ 32 233.4 Significant Findings and Plant-Unique Features................ 32 233.5 Hazard Frequency...................................... 32 2.4 Generic Safety Issues (GSI-147, GSI.148, and GSI-172)...................... 32 2.4.1 GSI-147," Fire-Induced Altemate Shutdown /ControlPanelInteraction"... 32 2.4.2 GSI-148, " Smoke Control and Manual Fire Fighting Effectiveness"...... 33 2.43 GSI-172; " Multiple System Responses Program (MSRP)".............. 33 3-OVERALL EVALUATION, CONCLUSIONS AND RECOMMENDATIONS.......... 38 3.1 Seismic.......................................................
Fire...............................................................38 3.2
.... 39 33 HFO Events......................................................... 40 4
IPEEE INSIGHTS, IMPROVEMENTS, AND COMMITMENTS..................... 41 4.1 Seismic..........................................................
Fire................................................................41 4.2
.. 41 43 HFO Events......................................................... 4 2 5
IPEEE EVAT.UATION AND DATA
SUMMARY
SHEETS......................... 43 6
REFERENCES............................................................ 51 Energy Research,Inc.
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8 LIST OF TABLES I
Table 5.1 Extemal Events Results.......................
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-Table 5.2 SMM Seismic Fragility......................
................ 45 Table 5.3 PWR Seismic Success Paths.......................................... 46 l
Table 5.4 PWR Accident Sequence Overview Table - For Fire PRA Only............... 47 1
Table 5.5 PWR Accident Sequence Detailed Table - For Fire PRA Only
............... 48 Table 5.6 '
PWR Accident Sequence Overview Table - For Wind (Tornado) PRA Only...... 49
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Table 5.7 PWR Accident Sequence Detailed Table - For Wind (Torando) PRA Only....... 50 4
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EXECUTIVE
SUMMARY
This technical evaluation repon '(TER) documents a " submittal-only" review of the individual plant examination of external events (IPEEE) conducted for the Comanche Peak Steam Electric Station (CPSES),
Units I and 2. This technical evaluation review was performed by Energy Research,Inc. (ERI) on behalf of the U.S. Nuclear Regulatory Commission (NRC). He submittal-only review process consists of the following tasks:
L Examine and evaluate the licensee's IPEEE submittal and directly relevant available documentation.
e Develop requests for additional information (RAls) to supplement or clarify the licensee's IPEEE i
submittal, as necessary.
Examine and evaluate the licensee's responses to RAls.
i Conduct a final assessment of the strengths and weaknesses of'.he IPEEE submittal, and develop i
review conclusions.
his *IER documents ER1's qualitative assessment of the Comanche Peak IPEEE submittal, particularly with L
l respect to the objectives described in Generic Letter (GL) 88-20, Supplement No. 4, and the guidance presented in NUREG-1407.
1 The licensee of CPSES is Texas Utilities Electric Company (TU Electric). The Comanche Peak IPEEE submittal considers the following external initiators: seismic; fire; and high winds, floods, and other (HFO) events. De seismic IPEEE was based on a reduced-scope seismic margin assessment (SMA); the fire IPEEE I
was based on a probabilistic risk assessment (PRA); high winds (tornadoes) were evaluated using PRA methodology; and external floods, transportation, and nearby facility accidents were evaluated using a hazard-based screening approach. The submittal notes the importance of extensive design and construction ver:fication efforts undertaken for plant licensing, as well as of other historical programs, to various facets l
of the IPEEE. The IPEEE was managed by the risk and reliability engineering division of TU Electric, with l
assistance from the civil engineering and stmetural engineering divisions. Much of the work conducted for the IPEEE made use of in-house expertise at TU Electric, however, outside consultants were also used to provide technical support and guidance, as necessary. All aspects of the external events analyses received an independent (internal or external) review.
Licensee's IPEEE Process In NUREG-1407, Comanche Peak is assigned to the reduced-scope seismic review category. For the seismic IPEEE, TU Electric implemented a reduced-scope evaluation, following the Electric Power Research l
Institute (EPRI) seismic margin methodology (SMM). He seismic IPEEE process relied primarily on plant walkdowns that focused on evaluating component anchorage capability and the potential for adverse spatial interactions. Seismic walkdowns were conducted using EPRI SMA walkdown procedures. He submittal l
notes that the seismic review team (SRT) verified that historical programs have been sufficient in j
demonstrating the seismic design and construction adequacy of plant components. A safe shutdown equipment list (SSEL) was developed based on preferred and alternate success paths that assume loss of offsite power (LOSP) and equivalent small break loss ofcoolant accident (LOCA) conditions. De preferred l
success path relies on secondary decay heat removal via the auxiliary feedwater (AFW) system, whereas the Energy Research,Inc.
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attemate success path relies on bleed-and feed cooling. The SSEL was reviewed by the SRT for adequac and SSEL components were either walked down or " walked by." Stmetures were screened out based on their seismic design and their meeting other SMA caveats. For the seismic containment performance analysis, components necessary for containment spray and isolation functions were included in the walkdowns.
(Based on results of the Comanche Peak individual plant examination [IPE) addressing intemal events, containment fan coolers were excluded from the seismic evaluation.) No relay chatter evaluation was performed for the seismic IPEEE. (Comanche Peak is not an Unresolved Safety Issue [USI) A-46 plant.)
With regard to fire-related initiators, the licensee conducted an extensive and detailed analysis of potential fire events and their effects at Comanche Peak. Several databases and related documentation were produced
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to establish, and keep track of, fire-related plant features, including fire zones and areas. IPE models were i
used for some portions of the fire analysis. Several extensive walkdowns of the plant were conducted to support the analysis. De licensee implemented a detailed Level-1 fire PRA framework. Fire frequency data and fire protection system data provided by EPRI were used for fire-scenario quantification. EPRI's fire-induced vulnerability evaluation (FIVE) methodology was used to evaluate fire propagation, detection and suppression, and cable and equipment damage. Special detailed attention was given to human actions and to the performance shaping factors influencing operator effectiveness. For redundant-train failure frequency i
evaluation, the plant models developed in the IPE were used.
De Comanche Peak HFO-events IPEEE was conducted using PRA methods and a hazard-based screening approach. The overall IPEEE process for HFO events involved the following steps:
identification of all potentially important extemal events e
performance of a qualitative screening process performance of a PRA for some external events e
Among all HFO events, tornadoes were further analyzed using a quantitative PRA approach. Detailed event trees and fault trees were developed (primarily by modifying the IPE event and fault trees) and quant'fied.
Historical data were used for determining tomado frequencies and tornado fragilities. External flooding, j
aircraft crashes, land transportation accidents, and nearby hazardous facility events were discussed in the IPEEE submittal, but these events were all screened out due to low probability of occurrence. Rus, no formal PRA or bounding analysis was performed for these initiating events. Walkdown findings were reported in the submittal for high winds, but not for other HFO events.
The Comanche Peak IPEEE was performed for Unit 1. The submittal states that the study for Unit 1 is also applicable to Unit 2.
KeyIPEEE Findings -
De key seismic IPEEE findings are walkdown related; no quantitative insights have been derived from the seismic assessment. Hus, no values for plant-level, system-level, or component high-confidence oflow-probability of failure (HCLPF) capacities have been estimated in 'the seismic IPEEE. He reduced-scope seismic evaluation revealed only two noteworthy outliers for which safety resolutions have been proposed and implemented. Dese outliers were encountered with respect to a review of safe shutdown equipment; no outliers were found specifically as a result of the containment performance evaluation.
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Based on the fire IPEEE, the licensee concluded that there are no significant vulnerabilities at Comanche Peak. The total fire-induced core damage frequency (CDF), under full-power operating conditions, was estimated to be 2.09x10-5 per reactor-year (ry). Given the level of detail of the fire analysis, this value represents.the combined fire CDF attributable to all locations within the plant. This value of fire CDF is within the range of results obtained for other nuclear power plants. With the exception of fire scenarios in j
the control room (which have a combined CDF of 9.04x 10 /ry ), the remaining dominant fire scenarios have 4
' CDF contributions ranging from 10 /ry to 2x10 /ry. De control-room fire scenarios are dominated by 4
4 operator failure to control the plant from the remote shutdown panel. Dese scenarios are modeled by loss of offsite power and failure to align a diesel generator, with a resulting station blackout (SBO). The following events appear in the majority of the dominant fire scenarios:
reactor coolant pump (RCP) seal failure, and resulting small LOCA, following loss of component
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cooling water (CCW) small LOCA due to a stuck-open power-operated relief valve (PORV) or safety relief valve (SRV) e l
operator error in controlling the PORVs e
2 operator failure to isolate safe shutdown equipment from the effects of a fire
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operator failure to properly align or restore core-cooling systems e
For HFO-events, the CDFdue to tornadoes was estimated to be 3.7x 10 /ry, which comprises a comparatively 4
low contribution (about 15%) to the total reported mean CDF due to extemal events (i.e., a CDF value of 4
2.46x10 /ry for fires and tornadoes combined). Tomado events can affect a plant either by direct wind loading or by tomado-induced missiles, ne IPEEE did not consider direct wind loading on the buildings themselves as being important to risk. The analysis assumed that wind loading can cause loss of offsite power; however, such an occurrence was treated as a recoverable event. The tomado PRA revealed that the dominant CDF sequences involve failure of the diesel generators. The core damage cutsets and the precise recovery actions considered in the PRA are, however, not definitively provided in the submittal. External flooding, aircraft crashes, land transportation accidents, and nearby hazardous facility events were all screened out based on their having low occurrence probabilities;' no PRA or bounding analyses were performed for these events.
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- Generic Issues and Unresolved Safety Issues For seismic events, the Comanche Peak IPEEE specifically addressed the following additional issues: USI A-45, " Shutdown Decay Heat Removal Requirements"; Generic Issue (GI)-131, " Potential Seismic
- Intenction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Plants"; and GI-57,
" Effects of Fire Protection Systems Actuation en Safety Related Equipment".
For USI A-45, the seismic IPEEE considers success paths that depend, separately, on the AFW system and on bleed-and-feed cooling. In addition, the submittal notes that the refueling water storage tank (RWST),
reactor make-up storage tank, and condensate storage tank (CST) are all designed as seismic Category I components to resist loads generated by a safe shutdown earthquake (SSE) event. For GI-131, the submittal states that the flux mapping system at Comanche Peak is designed to Category II seismic requirements, and will withstand SSE loads without posing an interaction hazard to the seal table. For GI-57, the submittal Energy Research,Inc.
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e notes that the SRT found the relevant concerns to have been adequately addressed in the seismic des CPSES. The submittal also comments on resolution of the eastern U.S. seismicity issue (Charleston Earthquake issue), and notes that no special reporting was required as part of the IPEEE. No vulnerabilities or concerns were noted with respect to any of the seismic-related GIs/USIs.
For the fire IPEEE, the licensee has addressed Sandia fire risk scoping study issues and USI A-45 concerns.
For both of these topics, the licensee has dealt with the relevant issues in detail, and has concluded that there are no outstanding problem areas. He licensee used reasonable methodologies for addressing these issues, and the conclusions noted in the fire IPEEE submittal concur substantially with results obtained for similar pressurized water reactors (PWRs). The presence of Thermo-Lag in the pia.J is mentioned in the IPEEE submittal, but no discussion is provided concerning the assumed effectiveness of this material.
The HFO-events IPEEE submittal does not describe any formal analyses of other safety issues. The licensee considers USI A-45 as being resolved for Comanche Peak. The submittal also makes the general statement that other generic and unresolved safety issues have been studied in the IPEEE process, and that these issues are considered closed. Additional documentation from the licensee states that GI-103, " Design for Probable Maximum Precipitation (PMP)", is considered to be resolved. No e,xplicit discussion was presented for USI A-17, " System Interactions in Nuclear Plants".
Some information is also provided in the Comanche Peak IPEEE submittal which pertains to generic safety issue (GSI)- 147, GSI-148, and GSI-172.
Vulnerabilities and Plant Improvements Although no plant-specific vulnerabilities were identified, the seismic IPEEE of CPSES did reveal a few minor anomalies and maintenance concems. The two observations that were considered to be significant enough to report, are:
The presence of unanchored, ancillary equipment close to safety related equipment, in the control room An instance ofinsufficient clearance between a motor control center and adjacent cable tray supports ne first of these observations was noted by the seismic IPEEE reviewer,'as opposed to the SRT. He submittal notes that, to address both of these issues, the SRT took appropriate follow-up actions to ensure their satisfactory resolution.
For fire events, no vulnerabilities were identified. No improvements and commitments were found to be necessary to further reduce the fire risk at Comanche Peak.
For HFO-events also, no vulnerabilities were identified. Due to the small value of CDF attributable to tornado events, the licensee did not consider any plant-specific improvements. Hence, no fixes or commitments related to HFO events are planned by the licensee.
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E Observations The seismic IPEEE of CPSES has addressed the major elements recommended in NUREG-1407 for evaluation of a reduced scope plant. Extensive design and construction verification efforts (including walkdowns), conducted earlier as part of the plant licensing process, served as a primary basis in the IPEEE for establishing seismic adequacy at the design-basis level. No major weaknesses of the licensee's seismic IPEEE submittal have been noted in this review.
For the fire IPEEE, considerable effort was expended, a valid methodology (a Level-1 PRA) was implemented, and appropriate databases for fire occurrence and fire suppression system failure quantifications were used. Following are listed the strengths of the fire IPEEE submittal:
De documentation is well written. The overall presentation in the submittal is clear and well organized. Tables and figures provide a considerable amount of information that supports the analysis and conclusions.
State-of-the-art methodology and data were used. The significant attention to detail,in fact, exceeds
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current standards of practice.
The fmal conclusions are reasonable and are within the range of results expected for a PWR power plant.'
The licensee used conservative screening methods and criteria.
The list ofissues addressed by the licensee is extensive and is considered to be complete.
l The licensee has clearly gained some excellent insights, and has realized an important experience, from its investigation of potential fire vulnerabilities at CPSES.
From its HFO-events IPEEE of CPSES, the licensee has gained a quantitative understanding of the effects j
of tornadoes, and a qualitative understanding of the effects of external floods and transportation events.
Since the HFO-events analysis was conducted by the licensee without any outside assistance, the licensee has likely gained significant insights concerning the capability of CPSES to resist potential severe accidents due to HFO initiators. There are, however, a number ofissues that have been raised in the present review, the resolution of which should improve the quality and credibility of the submittal. For example, the CDF evaluation for tornado events has taken credit for recovery of offsite power. His approach results in a reduced tomado CDF (in comparison to results assuming non-recoverability), and is typically applied to tornado PRAs in a different way than internal PRA evaluation of loss of offsite power recovery. The approach should consider the possibility of major damage to power lines or to the switchyard, making recovery actions difficult.
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De treatment of flood and transportation events in the IPEEE is generally considered to be quite adequate.
Notable weaknesses of the HFO-events IPEEE include:
ne submittal does not provide anyjustification for crediting recovery actions in a tornado-induced LOSP event.
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The submittal failed to report any changes in conditions, since the time of issuance of the operating license (OL), that may impact the analysis of HFO events.
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PREFACE The Energy Research. Inc., team members responsible for the present IPEEE review documented herein, include:
Seismic R. Sewell B.tt M. Kazarians Hinh Winds. Floods and Other External Events M. Modarres Review Oversieht. Coordination and Interration M. Khatib-Rahbar, Principal Investigator, Report Review
. A. Kuritzky, IPEEE Review Coordination and Integration R. Sewell, Report Integration Dr. John Lambright, of Lambright Technical Associates, contributed to the preparation of Section 2.4 following the completion of the dmft version of this TER.
This work was performed under the auspices of the United States Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The continued technical guidance and support of various hTC staffis acknowledged.
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ABBREVIATIONS AFW Auxiliary Feed Water CCW
. Component Cooling Water CDF Core Damage Frequency.
CFR Code of Federal Regulations CPSES Comanche Peak Steam Electric Station CST' Condensate Storage Tank EPRI Electric Power Research Institute ERI Energy Research,Inc.
FIVE Fire-Induced Vulnerability Evaluation FPP Fujita-Pearson tornado rating scale FSAR Final Safety Analysis Report GI GenericIssue GL Genericletter GSI Generic SafetyIssue HCLPF High Confidence of Low Probability of Failure (capacity)
HFO High Winds, Floods, and Other (external events)
HRA Human Reliability Analysis HVAC Heating, Ventilation, and Air Conditioning IPE Individual Plant Examination IPEEE Individual Plant Examination of Extemal Events LLNL Lawrence Livermore National Laboratory LOCA less of Coolant Accident LOSP.
Loss of Offsite Power MSLB Main Steam Line Break NRC United States Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center OL Operating License i
PGA Peak Ground Acceleration PMP Probable Maximum Precipitation PORV Power-Operated Relief Valve PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor -
RAI Request for AdditionalInformation RCP.
Reactor Coolant Pump RLE Review LevelEarthquake RWST Refueling WaterStorage Tank SBO Station Blackout-SMA Seismic Margin Assessment SME Seismic Margin Earthquake SMM Seismic Margin Methodology SPLD Success Path Logic Diagram SRP' Standard Review Plan SRT Seismic Review Team SRV' Safety Relief Valve Energy Research,Inc.
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C SSE' Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List.
SSW Station Service Water TER TechnicalEvaluation Report TU Electric Texas Utilities Electric Company
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USI Unresolved Safety Issue VSLOCA Very Small (break) Loss of Coolant Accident i
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INTRODUCTION i
This technical evaluation report (TER) documents the results of the " submittal-only" review of the individual I
plant examination of external events (IPEEE) for the Comanche Peak Steam Electric Station (CPSES), Units 1 and 2 [1]. This technical evaluation review, conducted by Energy Research,Inc. (ERI), has considered various external initiators, including seismic events; fires; and high winds, floods, and other(HFO) extemal events.
The U.S. Nuclear Regulatory Commission (NRC) objective for this review is to determine the extent to which the IPEEE process used by the licensee, Texas Utilities Electric Company (TU Electric), meets the intent of Generic Letter (GL) 88-20, Supplement No. 4 [2]. Insights gained from the ERI review of the IPEEE submittal are intended to provide a reliable perspective that assists in making such a determination.
His reyiew involves a qualitative evaluation of the licensee's IPEEE submittal, development of requests for additional information (RAIs), evaluation of the licensee responses to these RAIs, and finalization of the TER.
The emphasis of this review is on describing the strengths and weaknesses of the IPEEE submittal, panicularly in reference to the guidelines established in NUREG-1407 [3]. Numerical results are verified for reasonableness, not for accuracy; however, when encountered, numerical inconsistencies are reported.
This TER complies with the requirements of NRC's contractor task order for an IPEEE submittal-only review.
The remainder of this section of the TER describes the plant configuration and presents an overview of the.
licensee's IPEEE process and insights, as well as the review process employed for evaluation of the seismic,
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fire, and HFO-events sections of the Comanche Peak IPEEE. Sections 2.1 to 2.3 of this repon present ERI's detailed findings related to the seismic, fire, and HFO-events reviews, respectively. Sections 3.1 to 3.3 summarize ERI's overall evaluation and conclusions from the seismic, fire, and HFO-events reviews, respectively. Section 4 summarizes the overallIPEEE insights, improvements, and licensee commitments.
Section 5 includes completed IPEEE data. summary and entry sheets. Finally, Section 6 provides a list of the references cited in the TER.
1.1 Plant Characterization Comanche Peak is a two-unit power generating facility; each unit is a 4-loop Westinghouse pressurized water reactor (PWR). The two units do not share any common areas or systems except for the control room. The containment building for each unit consists of a steel-lined reinforced concrete structure (large, dry type) having venical cylindrical walls capped with a hemispherical dome. De containment is supponed on a foundation mat having a reactor cavity pit. The plant is located about 40 miles southwest ofFt. Worth, Texas on the banks of Squaw Creek Reservoir. Each of the Comanche Peak nuclear units has a rated full-power core thermal output of 3,425 MWt and a net electrical output of 1,156 MWe. Comanche Peak is a comparatively new plant (Unit 2 achieved commercial operation in 1993), and has undergone extensive design and construction verification.
The safe shutdown earthquake (SSE) peak ground acceleration (PGA) for Comanche Peak is 0.12g for horizontal motion and 0.08g for vertical motion. He SSE spectral shape is the same for both units; each unit Energy Research,Inc.
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was designed for a spectral shape very similar to a Regulatory Guide 1.60 spectrum [4). The plant is founded on rock (limestones and sandstones).
The Comanche Peak IPEEE was started in the fall of 1992, and documentation of the study was completed in June 1995. The IPEEE does not specify cutoffdates that would apply for establishing plant configuration and operating conditions. The IPEEE was performed specifically for Unit 1. After the completion of the Unit I study, TU Electric conducted a comparison study to identify differences between Units 1 and 2, and to evaluate the possible effects of these differences on the IPEEE findings. The differences in plant design and operation were reviewed and physically walked down, and were concluded to have insignificant impact I
on the IPEEE models and final results. De IPEEE submittal thus states that the IPEEE study is applicable to both units.
1.2 Overview of the Licenwe's IPEEE Process and important Insinhts 1.2.1 Seismic NUREG-1407 assigns Comanche Peak to the reduced-scope seismic review category. For the seismic IPEEE, TU Electric implemented a reduced-scope evaluation, using the Electric Power Research Institute (EPRI) seismic margin methodology (SMM). The seismic IPEEE process emphasized plant walkdowns that focused on evaluating component anchorage capability and the potential for adverse spatial interactions.
Seismic walkdowns were conducted using EPRI seismic margin assessment (SMA) procedures. De Comanche Peak seismic IPEEE study relied heavily on reference to historical programs at CPSES. The submittal notes that the seismic review team (SRT) verified that the historical programs have been sufficient in demonstrating the seismic design and construction adequacy of plant components. The submittal notes that the following steps were undertaken for the reduced-scope seismic evaluation of Comanche Peak:
Development of a program phan Selection of the SRT Development of the safe shutdown equipment list (SSEL) and containment systems equipment list Preparatory work prior to the walkdown, including screening and review of seismic design and seismic qualification of various elements Seismic capability walkdown and evaluation Subsequent walkdown and evalation Containment walkdown and evaluation Evaluation of generic issues (GIs) and unresolved safety issues (USIs)
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4 De seismic IPEEE concludes that there are no vulnerabilities at CPSES. A number of maintenance-type issues and two significant anomalies were observed, and actions were implemented for their resolution. The major finding of the Comanche Peak seismic IPEEE, according to the submittal report, is that the structures and equipment needed for safe shutdown, and for containment isolation and cooling given a seismic event, meet design-basis requirements and are adequately installed with respect to anchorage and system interaction considerations.
1.2.2 Fire The licensee conducted an extensive and detailed analysis of potential fire events and their effects at Comanche Peak. Several databases and related documentation were produced to establish, and keep track of, fire-related plant features, including fire zones and areas. Individual plant examination (IPE) models were used for some portions of the fire analysis. Several extensive walkdowns of the plant were conducted to support the analysis. De licensee implemented a detailed Level-1 fire probabilistic risk assessment (PRA). Fire frequency data and fire protection system data provided by EPRI were used for fire-scenario quantification. EPRI's fire-induced vulnerability evaluation (FIVE) methodology was used to evaluate fire propagation, detection and suppression, and cable and equipment damage. Special detailed attention wa.c given to human actions and to the performance shaping factors influencing operator effectiveness. For redundant-train failure frequency evaluation, the IPE models of the plant were used.
Multi-stage screening and detailed analyses were employed to identify the dominant contributors to fire core damage frequency (CDF). For the majority of the analysis, fire screening was based on a CDF threshold of 4
10 per reactor year (ry). The licensee addressed the control room and cable spreading room as fire areas.
He possibility of forming a hot gas layer was analyzed in great detail. The possibility of fire, with hot gas propagation among adjacent compartments, was also addressed.
1 The licensee evaluated the fire CDF for CPSES at 2.09x 10 /ry. Since the CDFcontributions for the leading 4
contributors were all of the same order of magnitude, no single pre-dominant risk contributor was identified.
The licensee concluded that there are no significant fire vulnerabilities at Comanche Peak.
He licensee also considered issues pertaining to the Sandia fire risk scoping study and to USI A-45. For both topics, the licensee addressed the relevant concems, and the IPEEE did not reveal any outstanding problem areas.
1.2.3 HFO Events The licensee conducted a detailed analysis for some HFO events, employing both PRA and hazard-based screening methods. His review has found some strengths of the study, along with a few weaknesses which the licensee may elect to re-visit. The weaknesses are summarized in Section 3.3 of this review.
De CPSES IPEEE has found no vulnerabilities with respect to HFO events. The most significant HFO-events CDF contributoris tornadoes, which contributes only 15% to the total CDF due to external initiators.
The dominant sequences for tomado are reported as those involving failure of the diesel generators.
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IJ Overview of Review Pra=* and Actit4 In its qualitative review of the Comanche Peak IPEEE, ERI focused on the stu % 3completenessin reference to NUREG-1407 guidance; its ability to achieve the intent and objectives of GL 88-20, Supplement No. 4; its strengths and weaknesses with respect to the state-of-the-art; and the robustness ofits conclusions. His review did not emphasize confirmation of numerical accuracy of submittal results; however, any numerical errors that were obvious to the reviewers are noted in the review findings. De review process included the following majoractivities:
e Completely examine the IPEEE and related documents Develop a preliminary TER and RAIs Examine responses to the RAls -
Finalize this TER and its findings Because these activities were performed in the context of a submittal-only review, ERI did not perform a site visit or an audit of either plant configuration or detailed supporting IPEEE analyses and data. Consequently, it is important to note that the ERI review team did not verify whether or not the data presented in the IPEEE matches the actual conditions at the plant, and whether or not the programs or procedures described by the licensee are indeed implemented at CPSES.
1.3.1 Seismic In conducting the seismic review, ERI generally followed the emphasis and guidelines described in the report, Individual Plant Examination of External Events: Review Guidance (5), for review of a seismic margin assessment, and the guidance provided in the NRC report,IPEEEStep-I Review Guidance Document
{6). In addition, on the basis of the Comanche Peak IPEEE submittal, ERI completed data entry tables developed in the Lawrence Livermore National Laboratory (LLNL) document entitled IPEEEDatabase Data Entry Sheet Package [1].
In its seismic review of the Comanche Peak IPEEE, ERI examined the following documents:
Sections 1,2,3,4.9.1,6,7, and 8 of the IPEEE submittal for CPSES (1) the licensee's responses to the RAIs (8) generated as part of the initial submittal review The checklist ofitems identified in Reference (5) was generally consulted in conducting the seismic review.
Some of the primary considerations in the seismic review have included (among others) the following items:
Were appropriate walkdown procedures implemented, and was the walkdown effort sufficient to accomplish the objectives of the seismic IPEEE7 Was the development of success paths performed in a manner consistent with prescribed practices?
Were random and human failures properly considered in such development?
Were component demands assessed in an appropriate manner, using valid seismic motion input and structural response modeling, as applicable? Was screening (including pre-screening) appropriately conducted?
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Were meaningful capacity calculations performed, where necessary, and are the capacity results reasonable?
Does the submittal's discussion of qualitative assessmer
- i (e.g., containment performance analysis, seismic-fire evaluation) reflect reasonable engineering judgment, and have all relevant concerns been addressed?
Has the seismic IPEEE produced meaningful findings, has the licensee proposed valid plant improvements, and have all seismic outliers been addressed?
13.2 Fire During this technical evaluation, ERI reviewed the fire-events portion of the IPEEE for completeness and consistency with past experience. This review was based on Sections 1,2,4,6,7,8 and 9 of Reference [1],
and on the licensee responses to fire-related RAls [8]. The guihace provided in Reference [6] was used to formulate the review process and to organize this technical evaluation report. The data entry sheets used in Section 5 are taken from Reference [7].
The process implemented fc; ERI's review of the fire IPEEE included an examination of the licensee's methodology, relevant data, and results. ERI reviewed the methodology for coni -
vith currently accepted and state-of-the-art methods, focusing on the screening methodology and on e x:edure used for estimating the frequency of occurrence of a fire scenario, to ensure that no fire...:enarios were prematurely eliminated. The data element of a fire IPEEE includes, among others, such items as:
cable routing' fire zone / area partitioning fire occurrence frequencies event sequences a
fire detection and suppression capabilities The conditions described, and information provided, by the licensee were evaluated to determine their reasonableness, and their similarity with other fire PRAs. For a few fire zones / areas that were deemed important ERI also verified the logical development of thejustifications/ arguments (especially in the can of fire-zone screening) and the computations for fire occurrence rates and CDF.
133 HFO Events
'Ihe review process for HFO events closely followed the guidance provided in the report entitled IPEEE Step-1 Review Guidance Document [6). This process involved examinations of the methodology, the data used, and the results and conclusions derived in the submittal. The IPEEE methodology was reviewed for consistency with currently accepted practices and NRC recommended procedures. Special attention was focused on evaluating the adequacy of data used to estimate the frequency of HFO events, and on confirming that any analysis of Standard Review Plan (SRP) conformance was appropriately executed. In addition, the validity of the licensee's conclusions, in consideration of the results reported in the IPEEE submittal, was assessed. Also,in some Instances, computations of frequencies of occurrence of hazards, fragility values, Energy Research,Inc.
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a and failure probabil'ities were spot checked.' Review team experience was relied upon to assess the reasonableness of the licensee's evaluation.
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. CONTRACTOR REVIEW FINDINGS '
2.1 Seismic A summary of the beensee's seismic IPEEE process has been described in Section 1.2. Here, the licensee's seismic evaluation is described in detail, and discussion is provided regarding significant observations encountered in the present review.
-2.1.1
- Overview and Relevance of the Seismic IPEEE Process
' Seismic Review Category and Review-LevelEarrhquake (RLE)
. a.
I Comanche Peak is assigned, in NUREG-1407, to the reduced-scope seismic review category. He review level earthquake (RLE) is described by the design-basis spectrum for the plant (i.e., a shape similar to the Regulatory Guide 1.60 spectral shape, anchored to a PGA value of 0.12g).
^ b.
SeismicIPEEE Process TU' Electric' elected to implement a reduced-scope' evaluation, following the EPRI seismic margin methodology, for conducting the seismic IPEEE of Comanche Peak l
! CPSES is a two-unit PWR plant. The IPEEE evaluation was performed for Unit 1, and the licensee verified j
that the findings were also applicable to Unit 2. The licensee stated that the seismic walkdown did not address only Unit-1 equipment, but that some Unit-2 equipment were also walked down.
c.
Review Findings,
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The Comanche Peak seismic IPEEE essentially addresses :ll major elementr. identified in NUREG-1407 for
. a reduced-scope evaluation, including consideration of containment performance. Comanche Peak is not a'USI A-46 plant, and consequently (consistent with NUREG-1407 guidelines), no relay evaluation was
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' performed for the seismic IPEEE. Neither was an evaluation of potential soil failures conducted, since such evaluation is not requested for a reduced-scope plant, and since Comanche Peak is predominantly a rock site.
' The plant has already undergone extensive design and construction verification programs, including seismic
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. plant walkdowns. Th's objective of a reduced-scope evaluation is to assess seismic adequacy at the design-basis level, and the SRT was able to readily'confinn (though walkdowns and walk-bys) that the historical programt,have been sufficient in demonstrating the seismic design and construction adequacy of plant components. In addition, the SRT investigated the adequacy of equipment anchorage and the potential for seismic interactions.
.The submittal clarifies that differences between Units 1 and 2 were identified in a comparison study. De differences were reviewed and walked dow n, and the IPEEE submittal states that the differences were found to have ins:gnificant impact on the seismic IPEEE. On this basis, the submittal concludes that the seismic L
, IPEEE study is applicable to both plants.
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i The approach implemented for the Comanche Peak seismic IPEEE thus appears to be relevant to the evaluation of seismic severe accident resistance, and is consistent with the approach requested by NUREG-I407.
2.1.2 Success Paths and Component List For the seismic IPEEE of Comanche Peak, TU Electric developed a success path logic diagram (SPLD) that describes the plant functions needed to achieve and maintain a stable shutdown condition for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The SPLD identifies both preferred and alternate success paths. The preferred success path relies on high-pressure injection via the centrifugal charging pumps, with secondary side heat removal via the auxiliary feedwater (AFW) system. The alternate success path relies on intermediate pressure injection via the safety injection pumps, with bleed-and-feed cooling for decay heat removal. The submittal report notes I
that, in general,if one train of a multi-train safety system was used in one success path, then an entirely different system was used in the other success path, rather than simply a different train of the same safety system. Safety systems at CPSES are dual-train systems, where the second train is qualified by similarity considerations; hence, the submittal states that the second trains provide additional success paths.
The success paths chosen for safe shutdown of Comanche Peak, following a seismic margin earthquake (SME), assume both loss of offsite power (LOSP) conditions and small break loss-of-coolant accident (LOCA) conditions. IPE event trees for loss of offsite power and very small break LOCA (VSLOCA) events
- were used in developing the success paths. Although the terminology used by the licensee in the IPE to describe the VSLOCA event appears to suggest that only a very small break LOCA was addressed in the SMA success paths, the IPE definition for VSLOCA (up to an equivalent 2 inch diameter break)is, in fact, more stringent than the SMA definition of small LOCA (equivalent 1-inch break). Hence, the licensee's use of the IPE VSLOCA event adequately addresses the SMA contingency of mitigating a small-break LOCA in at least one success path.
The IPEEE submittal provides the following synopsis of the methodology used to derive the safe shutdown equipment list (SSEL):
Identify the plant-specific critical safety functions for CPSES
_ Develop event trees for LOSP and VSLOCA initiating events Identify the systems that provide the functions identified in the event trees Develop the SPLD, and identify both the preferred and alternate success paths for the initiating events, taking into account both operational and systems considerations The submittal also notes that tSe first phase of the seismic walkdown effort was undertaken to confirm the selected success paths and ve.ify completeness of the SSEL. Following this approach, both the SSEL and a containment systems equipment list were finalized. The submittal indicates that the containment systems equipment list was developed to ensure a success path where early containment failure, given core damage, is presented. Successes of the containment spray system and the containment isolation system were n. quired in the development the containment systems equipment list.
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r The approach described in the Comanche Peak seismic IPEEE for defining success paths and safe shutdown equipment list and containment systems equipment list appears to be reasonable and consistent with the guidelines presented in NUREG-1407.
j 2.13 Non-Seismic Failures and Hu'aan Actions OverallApproach a.
l Non-seismically-caused (random) component unavailabilities and operator errors were addressed in the
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seismic IPEEE, albeit in a qualitative manner. With respect to non-seismic failures and human actions, the Comanche Peak seismic IPEEE notes the following:
that safety systems at Units 1 and 2 are dual-train systems that the preferred and altemate success paths were chosen to closely parallel the paths that the e
operators would use during recovery from an SME With respect to human:1ctions, the submittal states that a review was made of the primary procedures used to respond to a r.eismically induced LOSP or VSLOCA initiating event. A senior reactor operator examined the actions required in the IPE model, and qualitatively assessed whether or not an operator could be expected to perform each action following an earthquake. Availability of control room instruments needed to complete the required humaa actions was also evaluated, considering seismic qualification information.
'Ihelicensee concludes that a seismic event will not adversely affect the performance of human interactions required to mitigate a seismically induced LOSP or VSLOCA initiator.
b.
Screening Criteria No quantitative screening criteria were applied to non-seismic failures and human actions. The screening was performed on a qualitative basis.
c.
Review Findings The fact that safety systems are dual-train systems helps alleviate concerns related to success-path vulnerabilities to non-seismic failures. Also, the fact that success paths were selected to closely parallel the recovery paths expected by operators helps to alleviate concems related to success-path vulnerabilities to human errors. It is, therefore, reasonable to conclude that there are no cut-of-ordinary concems associated with non-seismic and human failures along the chosen SMA success paths. Consequently, the licensee's l
treatment of non-seismic failures and human actions appears to be an appropriate approach for a reduced-scope plant.
2.1.4 SeismicInput The Comanche Peak seismic IPEEE uses the SSE ground design response spectrum as seismic input. The SSE spectral shape is very close to the Regulatory Guide 1.60 spectrum; the SSE spectrum for horizontal motion is anchored at a PGA value of 0.12g. The SSE spectrum for vertical motion is the horizontal l
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spectmm multiplied by two-thirds. Other relevant criteria delineated in the Comanche Peak final safety analysis report (FSAR) were also used by the SRT for defining the input for seismic evaluation.
The seismic input used in the Comanche Peak IPEEE is consistent with the relevant guidelines presented in NUREG-1407.
2.1.5 Structural Responses and Component Demands 1
a.
OverallApproach l
The Comanche Peak seismic IPEEE uses FSAR structural responses to define structural demands, and FSAR in-structure response spectra to define component demands. No new structural models were developed, nor l
wem any new in-structure response spectra generated.
b.
StructuralModels The submittal does not provide a detailed description of the FSAR structural models and parameters used to generate the original (design-basis) structural responses and component demands
]
c.
Review Findings The development of structural responses and component demands used in the Comanche Peak IPEEE is consistent with the relevant guidelines presented in NUREG-1407.
2.1.6 Screening Criteria The Comanche Peak seismic IPEEE made use of the screening criteria and procedures discussed in EPRI NP-6041, Rev.1 [9). He screening column applicable to spectral accelerations less than the 0.8g limit was used (Tables 2-3 and 2-4 of EPRI NP-6041, Rev.1). Caveats of the seismic margin screening tables were used in the screening process for structures, as well as for electrical and mechanical equipment. Structures were screened ouc from further evaluation in consideration of design-basis information. Anchorage and spatial interaction concems were addressed in the walkdown(s). De screening approach helped focus the walkdown effort by eliminating items thought to have high seismic margin, and identifying remaining items with potentially low seismic capacity.
The screening criteria and procedures used in the Comanche Peak seismic IPEEE are consistent with NUREG-1407 guidelines, and are judged to be appropriate for screening SSEL components.
l 2.1.7 Plant Walkdown Process a.
Preparatory Wort Prior to the walkdown effort, the SRT reviewed the documentation of recent seismic qualification and verification programs conducted for Comanche Peak, in order to evaluate their applicability to the IPEEE.
Walkdown checklists were compiled for SSEL and containment-systems components. Drawings indicating the mounting details of equipment were examined, and piping system diagrams were reviewed to collect Energy Research,Inc.
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information on the location and seismic qualification of valves. Components were then grouped by plant area and separate walkdown packages were assembled.
b.
Systems andElement Selection Walkdown ne first phase of the walkdown itself focused on confirming the success paths and determining completeness of the SSEL. The specific objectives of this phase of the walkdown were to:
Review the components and stmetures in the SSEL for any obvious problems related to seismic l
evaluation at the SME level Locate SSEL equipment to determine its accessibility Arrange for SRT access to equipment, and identify plant areas (with the assistance of radiation protection personnel) that were' inaccessible due to high radiation.
his initial walkdown did not reveal any obvious problems with plant SSEL components and structures, c.
Seismic Capabiliry Walkdown ne SRT physically reviewed important attributes of equipmem, according to procedures described in
- Appendix A of EPRI NP-6041, Rev.1. The walkdown review was performed for both SSEL items and
. components related to containment systems.
The seismic capacity walkdown focused on evaluation of potential anchorage problems and spatial
- interaction concerns. Representative components were visually inspected in detail. Among SSEL and containment-systems components that are similar, and similarly anchored, one or two representative components were selected for detailed walkdown. Other SSEL items were field reviewed on the basis of a
" walk by" which focused on identifying obvious outliers. In instances where suspect / poor equipment support details were encountered, information was collected in the field to enable subsequent evaluation of these items. The SRT also looked for instances of poor seismic housekeeping. The submittal notes that a detailed walkdown was performed of the diesel generator room, d.
Subsequent Walkdowns ne submittal does not indicate that a significant effort was required in performing subsequent walkdowns.
- e.
Treatment ofInaccessible Components Where areas and/or components were inaccessible due to high radiation / contamination, documentation reviews were conducted in place of physical walkdowns. Several components were inaccessible for walkdown; the majority of these components were valves. De ~ documentation review addressed both seismic capability and interaction concerns.
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f.
Walkdown Duration and Training ofSeismic Review Team (SRT)
De seismic IPEEE walkdown of Comanche Peak was conducted by trained licensee personnel and experienced consulting seismic review expens. The SRT consisted of six senior personnel, from which two walkdown teams were assembled. Each team consisted of two Comanche Peak engineering personnel and one consultant from EQE International. Each team was accompanied by plant representatives from operations, radiation protection, and other Comanche Peak organizations. The duration of the seismic l
walkdown process is not mentioned in the submittal, however, it is surmised to be substantial based on the submitted walkdown repon.
g.
Review Findings Walkdowns had already been conducted prior to the IPEEE, as part of the extensive design and construction verification efforts undertaken for the licensing process. Documentation from these efforts was reviewed by the SRT and helped focus the seismic IPEEE walkdown effon. For example, the submittal notes that, due to their seismic design, prior walkdown verification, and/or their evaluation based on earthquake experience data, passive components such as piping, cable trays, and heating, ventilation, and air conditioning (HVAC) ducts were not included in the seismic IFEEE walkdown.
The walkdown process conducted for the seismic IPEEE of Comanche Peak appears to be reasonable and appropriate for a reduced-scope evaluation, as well as capable ofidentifying vulnerabilities with respect to safe shutdown and early containment integrity following a seismic margin canhquake.
2.1.8 Evaluation of Outliers a.
OverallApproach Comanche Peak is not a USI A-46 plant; hence, the IPEEE evaluates outliers againe FSAR requirements.
The seismic IPEEE submittal refers to outliers as " anomalies." The submittal mentions that such anomalies take the form of suspect support details (including condition and configuration of anchorage and connections), presence of cracks in concrete foundation pads, seismic interaction issues, etc. The SRT made field notes regarding these items, sufficient for purposes of subsequent evaluation. The evaluations were conducted using seismic qualification documentation to ensure that the observed critical or suspect details j
were adequately addressed.
He submittal contains an itemized list of walkdown observations and their resolutions. Among the 14 anomalies observed and/or evaluated as a result of the screening and walkdown, only two items were i
reponed as requiring corrective actions. The submittal notes that most of the anomalies related primarily to maintenance concems.
b.
Calculations i
The submittal does not present any calculations or calculational results produced from evaluation of i
anomalies / outliers.
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' 2.1.12 J Seismic-Fire Interaction and Seismically Induced Flood Evaluations a.
Seismic-Fire Evaluation and Walkdown
- The Comanche Peak IPEEE included a review of seismic-fire interaction concerns as part of the evaluation of the Sandia fire risk scoping study issues. This review included an assessment of:(1) seismically induced
. fires, (2) seismic actuation of fire suppression systems, and (3) seismic-induced failure of fire suppression systems. Walkdowns, as necessary, were performed by the SRT, primarily to identify potential interaction problems and to verify conformance to earlier design and qualification programs.
b.
Seismic-induced Fires The assessment for seismically induced fires looked at potential breakage of flam:nable liquid / gas vessels as well as other potendal seismically caused ignition hazards. The submittal notes that all significant amounts of flammable liquids are stored in separate fire areas that are isolated from adjacent plant areas by hour rated fire barriers. Fire detectors are provided in each area, and a fixed fire extinguishing system is also usually provided. Bulk flammables are stored in an open structure located outdoors in the plant yard; the licensee states that an explosion or fire in this area would not affect any of the primary plant buildings.
The submittal references plant procedures for the storage and handling of flammable / combustible materials
. and compressed gases, and for the control of transient combustibles and other fire-ignition sources.
The submittal does not discuss the potential for fire caused by seismically induced rupture / leakage of piping
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containing flammable / combustible gases, nor for fires caused by damage to energized high-voltage cabinets.
Seismic Actuation ofFire Protection System Equipment c.
ne submittal references an internal flood design analysis for the plant, as well as the IPE assessment of intemal floods. Seismic qualification of various elements of the fire suppression system (e.g., deluge valves, j
sprinkler system) was reviewed, and the potential for dust-related inadvertent actuation of the Halon system was considered.
d.
Seismic-induced Failure ofFire Protection System Equipment De licensee's evaluation of seismic-induced failure of fire suppression systems looked at the seismic design' i
of the fire suppression sy. tem, the potential for adverse spatial interactions, and the potential for seismic failure of threaded pipe connections. A walkdown was performed by the SRT to review seismic spatial interaction issues, and to confirm that the design of the fire protection system is in accordance with seismic requirements. The emphasis of the evaluation was on survivability of fire protection systems in close proximity to SSEL equipment; the potential of an earthquake to compromise fire suppression capability was not explicitly stressed.
l e.
Seismically Induced Flood Evaluation The licensee addressed seismically induced flooding in its treatment of seismically induced inadvertent actuation of fire protection system equipment. Flooding that might occur due to seismically induced failure of non-SSEL tanks and piping was not explicitly addressed. However, the IPE and an internal flood design Energy Research,Inc.-
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analysis for the plant address the impacts of potential flooding, as may be caused by an earthquake or any
. other event.
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Review Findings De Comanche Peak IPEEE has implemented a seismic-fire interactions evaluation that addresses a number of issues of concern. De evaluation did not explicitly consider the potential and effects of seismically induced electrical cabinet fires; rupture / leakage of gas piping; flooding due to failure of non-SSEL water tanks and piping; and loss of fire suppression capability. The submittal's documentation of extensive design verification programs, combined with the fact that Comanche Peak is a reduced-scope plant and a newer vintage plant, suggest that it is unlikely that any problems exist with these items.
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' 2.1.13 Treatment of USI A-45 i
i De success paths developed for the Comanche Peak seismic IPEEE address decay heat removal requirements, following a seismic margin earthquake, via AFW capability and bleed-and-feed cooling. The i
elements of these decay heat removal fur.ctions have been evaluated as part of the consideration of SSEL components. The SSEL was developed assuming decay heat removal requirements, under conditions ofloss of offsite power and small LOCA, for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following the seismic margin earthquake. The primary water sources (refueling water storage tank [RWST], reactor make-up storage tank, and condensate storage tank [ CST)) that supply decay heat removal systems were included in the evaluation of success-path components, and were all found to have been designed as Category I structures for SSE loads. Based on the
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seismic margin evaluation, the IPEEE submittal concludes that there are no specific seismic vulnerabilities usociated with decay heat removal function capability at CPSES.
De Comanche Peak seismic IPEEE includes a meaningful evaluation of potential vulnerabilities in decay j
heat removal systems, which is judged to adequately address the relevant concems of USI A-45.
l 2.1.14 Treatment of GI-131 i
l As part of the Comanche Peak seismic IPEEE, the SRT reviewed existing documentation from an earlier l plant-specific seismic /non-seismic systems interaction program related to seismic capability of the flux mapping system. The SRT found that the flux mapping system is designed to Category II requirements to be capable of withstanding SSE loads, thus precluding interactions with the seal table or other safety related components at the SSE level.
The licensee's prior treatment of GI-131, as described in the Comanche Peak seismic IPEEE submittal, appears reasonable.
2.1.15 OtherSafetyIssues Eastem U.S. Seismicityissue a.
Probabilistic seismic hazard calculations were performed for the Comanche Peak site, as part of the resolution program for the eastern U.S. seismicity issue (Charleston Earthquake issue), he seismic IPEEE Energy Research,Inc.
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submittal mentions that Comanche Peak is not an outlier plant with respect to this issue, and hence, that no additional reponing is needed.
b..
GI-57 One issue identified in the Sandia fire risk scoping study pertains to the effects of fire protection system actuation on safety-related equipment. Concerns associated with this issue are addressed under GI-57. As part of the consideration ofinternal fires, NUREG-1407 lists GI-57 as a program that may be coordinated with the IPEEE. Even though GI-57 relates primarily to internal fires, the Comanche Peak seismic IPEEE includes a discussion pertaining to its resolution. The submittal notes that the SRT completed a survey of f
various hist.orical programs and design documents that address GI-57 issues, and that the SRTconcluded that the effects of fire suppression system actuation on safe shutdown equipment have been adequately addressed j
in the design of CPSES.
c.
Generic Safetyissues Some seismic-related infonnation having relevance to Generic Safety Issue (GSI)-172 is provided in the submittal, as discussed in Section 2.4.3 of this TER.
d.
Review Findings The Comanche Peak seismic IPEEE contains brief discussions relevant to the bases for resolutions of o seismic safety issues.
2.1.16 PeerReviewProcess
'IU Electric personnel had a substantial involvement in conducting the Comanche Peak seismic IPEEE. The entire seismic IPEEE examination process received an independent peerreview from TU Electric personnel and from outside consultants. Each aspect of the seismic iPEEE was subject to at least one review; many aspects were subject to two reviews. Plant personnel from training and operations were also interviewed regarding the appropriateness of seismic IPEEE findings and assumptions. The peer review process was organized such that individuals were selected on the basis of their having relevant expenise to then.narticular review task.
'Ihe IPEEE submittal describes the general expenise/make-up of the peer review team, but (other than identifying the walkdown reviewer) does not actually list individual review-team members. The submittal report presents a summary of reviewer comments pertaining to the seismic IPEEE, and describes how eac,h review concern was resolved. It is noted in these peer review comments that one of the two outliers / anomalies reported in the IPEEE submittal had not, in fact, been encountered by the SRT, but had actually been found by an external reviewer.
It isjudged that a meaningful peer review was conducted of the Comanche Peak seismic IPEEE, and that the major comments / concerns of the peer review team were substantially addressed.
2.1.17 Sumrmey Ew1uation of KeyInsights Energy Research,Inc.
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L De Comanche Peak seismic IPEEE draws two principal insights: first, the consistent SRTobservation that all equipment, stmetures, and other components meet seismic design requirements and are adequately installed with respect to systems interaction considerations; and second, the identification of two notable anomaliesrequiringfollow upactionforresolution.Desubmittalconcludesthattherearenovulnerabilities L
with respect to safe shutdown capability, containment performance, decay heat removal capability, flux
. mapping system capability, or seismic-fire interactions at Comanche Peak.
Only a small number of anomalies were revealed from the seismic IPEEE, which is not unexpected given the recent vintage of the plant and the extensive design verification pmcess.
2.2 Egg A summary of the licensee's fire IPEEE process has been described in Section 1.2. Here, the licensee's fire evaluation is described in detail, and discussion is provided regarding significant observations encountered in the present review.
2.2.1 Overview and Relevance of the Fire IPEEE Process Merhodology Selectedfor the Fire IPEEE a.
In the Comanche Peak IPEEE, fire events were analyzed using detailed Level-1 PRA methodology, making use of the approaches and data provided in References [10] and [11). De analyses of fire propagation and hot gas layer formation were based on the computational techniques of the FIVE methodology (Reference
[12]). De overall IPEEE methodology followed a sophisticated set of steps, where several different databases were developed, and where ignition frequencies, fire scenarios, fire propagation, fire detection and suppression, and operator actions were addressed and analyzed in detail. De licensee gave special attention to potential fires in the control room and in the cable spreading room. Also, multi-compartmental fire propagation and damage was analyzed in the IPEEE.
De formulations provided in the FIVE methodology (Reference [12]) were used for estimating the behavior of fires, including fire propagation, damage to equipment and cables, and hot gas layer formation. Extensive plant walkdowns were conducted for the various steps of the analyses.
The IPEEE submittal provides extensive discussions regnrding the methods and data used in conducting the fire analysis.
The fire IPEEE was conducted specifically for Unit 1. The licensee states that a comparison study was conducted to identify the differences between the two units, and that it was concluded that the differences are insignificant from the standpoint ofIPEEE issues. Hence, the results obtained from analyzing Unit 1 are said to apply to Unit 2, as well.
b.
Key Assumptions Used in Performing the Fire IPEEE he IPEEE submittal does not provide a separate list of evaluation assumptions; however, during the present review, the following assumptions were identified as potentially having an effect on the final results:
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1.
Reactor subcriticality was assumed to be successful in all cases 2.
A full fire-zone suppression system was assumed to be capable ofpreventing formation of a hot gas layer 3.
Fire barriers / boundaries were taken to be as good as rated 4.
The fire protection system was designed and installed per proper codes and standards Status ofAppendix RModifications c.
ne licensee did not indicated the status of Appendix-R modifications at the time of preparation of the fire PRA. The IPEEE submittal does not discuss Appendix-R issues. However, it can be inferred from the submittal that efforts spent on Appendix-R compliance were marginally useful in the preparation of the IPEEE.
d._
New or Existing PRA The IPEEE is a new fire PRA study. De core damage frequency (CDF) calculations were based on the IPE plant response model (i.e., event trees and fault trees).
2.2.2 Review of Plant Information and Walkdown a.
Walkdown Team Composition Five specific walkdowns are described in Reference [1]. In addition to these walkdowns, the fire analysis team inspected the plant for specific areas and rooms relevant to fire evaluation.
In Section 6 of Reference [1), qualifications of the licensee's in-house IPEEE team are presented. The team was composed of experienced engineers and analysts. In addition, the fire-analysis portion of the IPEEE was "a part of an EPRI tailored collaborative project," and benefitted from reviews by a wide range of experts.
However, no discussion is provided as to when, where, and by whom the walkdowns were conducted. Some information is provided as to the method used for conducting the walkdowns. However, the format of the records taken during the walkdown is not presented. Considerable effort was apparently expended in preparing for the walkdowns. The fire protection and civil engineering groups were interviewed to gain an understanding of the specific features of the plant and safety-related practices followed by the staff at the plant. In addition to these i..terviews, a technical :eview of each compartment was conducted to identify fire barriers and communication paths among fire areas. Special attention was given to the identification of areas where transient fuels may be present, and to the possibility of formation of a hot gas layer.
A separate walkdown was conducted for the control room and cable spreading room. Operators were interviewed to assist understanding of how the operating crew would respond to a fire in the control room.'
b.
Significant Walkdown Findings Energy Research,Inc.
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o The scope of the walkdowns included examinations for fixed and transient ignitions sources, the potential for multi-compartmental fires, control room and cable spreading room features, and plant features relevant to Sandia fire risk scoping study issues. Reference [1] does not indicate that the walkdown team discovered any new fire vulnerabilities from the plant walkdowns. However, some observations are noted regarding presence of transient fuels, features of the control room panels, confinement ofcombustible liquids, etc. The features noted as part of the walkdown observations can be considered as typical for nuclear power plants.
c.
Significant Plant Features The IPEEE submittal does not provide much information regarding Comanche Peak plant-specific features.
No additional information regarding the plant was available for this review. From the IPEEE submittal report, the following plant features could be gleaned:
1.
De plant consists of two identical units; 2.
Each unit is a Westinghouse PWRt and 3.
The two units share a common control room.
2.2.3 Fire-Induced Initiating Events Were initiating Events Other than Reactor Trip Considered?
a.
Components and initiators modeled in the IPE have been included in the fire IPEEE study. From the discussions in Sections 4.4.1,4.6.2 and 4.7, it can be inferred that the licensee addressed initiators other than reactor trip in estimating the fire CDF and the effects of a fire. However, two issues are not explicitly discussed in the IPEEE submittal. First, it is not clear whether or not the list ofIPE components has been expanded to include cables and components that would cause an initiator. Secondly,in nearly all of the discussion, there is no mention of cables as components explicitly considered in the analysis.
b.
Were the initiating Events Analy:ed Properly?
The relevant initiating events (from among the list ofIPE initiators) were addressed in the IPEEE. A list is provided in the submittal of the specific initiating events that were analyzed; the items in this list capture all the initiators that are typically addressed as part of a PWR fire risk analysis. As was mentioned above, no indication is provided in the submittal that all IPE initiators have been analyzed in detail, nor that the IPE component list has been e. panf ed to include the additional components associated with all initiators.
2.2.4 Screening of Fire Zones Was a Proper Screening Methodology Employed?
a.
Screening was performed at several points in the fire analysis. All screening was based ori CDF estimates, which is a proper approach for IPEEE purposes. The initial screening of compartments was based on a CDF threshold of 10 per reactor year (ry). His threshold value has led the analysts to select a large number (as compared to other IPEEEs) of fire zones for detailed analysis. When analyzing multi-compartmental fire scenarios, the threshold CDF value was taken to be 5.0x10-'/ry. Although this choice is less conservative than that used for single-compartmental fires, it is an acceptable level since it is less than 10 /ry (which is 4
the screening threshold typically used in other IPEEEs)-
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Fire scenarios for the cable spreading room and the contic,ow were screened based on a conditional CDF of 5.0x 10"/ry. It may be noted that these two rooms were not screened out; the fire scenarios that ma within these rooms were screened based on the conditional CDF criteria. This conditional screening le implies a (unconditional) CDF ofless than 10 /ry, astning any fire in these rooms could restdt in the worst 4
possible damage.
For the human error analysis, the CDF threshold value was taken to be 10 /ry. For fire scenarios that 4
included human actions, no detailed human error analysis was conducted if the CDF was estimated to be less than this threshold value.
The licensee did not provide a complete list of the fire areas and zones, nor a list of the equipment / system trains that could potentially be affected by a fire in the various companments. Therefore, the present review of screening results is based solely on consideration of the methodology described in the submittal and on consideration'of various ponions of the final results.
b.
Have the Cable Spreading Room and the Control Room Been Screened Out?
The cable spreading room and control room were addressed in a detailed analysis and were not screened out.
The analysis for these two areas followed a detailed state-of-the-an approach. The methodology for fire assessment of these two rooms was based on EPRI's " Fire Risk Analysis Implementation Guide." For cabinet fires, no credit was given to the possibility of fire suppression prior to the loss of the circuits contained within the cabinet. Three types of operator response scenarios were considered. In the first scenario, the control room is evacuated and safe shutdown is achieved solely from the attemate shutdown panel. In the next scenario, safe shutdown is achieved from a combined usage of the main control room panels and the alternate shutdown panel. In the last scenario, despite multiple-cabinet fire damage, the operators remain in the control room and do not use the altemate shutdown panel.
Were There Any Fire Zones / Areas that Have Been Improperly Screened Out?
c.
No improper screening of fire areas / zones was encountered. This finding is based primarily on the described methodology and the final results. The final results appear reasonable and are within the expected range of results for a PWR. Table 4.6-3 of the submittal provides a list of fire zones that were screened out.
However, this table does not include any information regarding the usage of the fire zones, nor does it provide a list of system trains present.
2.2.5 Fire Hazard Analysis The fire initiation data provided in Reference [11) was used in the fire IPEEE Weighting factors were applied, in order to apponion the overall fire frequency to a specific fire zone. The methodology described in the submittal for collecting the necessary data, and for apportioning the fire frequencies,is deemed to be reasonable. However, the submittal provides little informationjustifying the weighting values that were used for this purpose. Final results were presented for those areas that were not screened out. It is not clear how transient combustibles were addressed and included in the fire frequency evaluation..
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r A plant-specific database was not used. However,in Section 4.3.2 of the submittal,it is stated that the plant-specific events were reviewed, and that these events were considered as being consistent with the generic I
data.
l 2.2.6 Fire Growth and Propagation The licensee undertook detailed fire modeling where heat release quantity, heating of target materials, temperature rise of the target materials, and damage were all considered. The possibility of hot gas layer formation, propagation beyond the compartment of origin, and damage to cables near the ceiling were also considered. Some conservative assumptions were made in the analysis. Heat losses by convection in ventilated rooms, or by conduction in cables and equipment, were not credited. Fires were assumed to be fully developed at the time ofignition (in other words, it was assumed that a combustible material, from the moment it ignites, will generate heat at a rate equal to that when it is fully engulfed in fire).
Treatment of Cross-Zone Fire Spread and Associated Major Assumptions a.
Cross-zone fire propagation was considered in the fire IPEEE analysis. An extensive analysis was conducted for multiple-zone effects. Rese fire scenarios contribute about 3 percent to the total fire CDF. The primary modes of fire propagation were hot gas layer penetration th>ough openings, and fire-barrier failures.
Various causes of fire-barrier failure were considered. A probability of 7.4x10-8/ry was used for failure of doors. De potential for fire-barrier failure from fire-fighting activities was not addressed. An example of such an event may include a fire in a room containing train-A equipment and cables, while access to this room is via the adjacent train-B compartment. The barrier-failure probability for dampers was not provided j
in the submittal.
l b.
Assumptions Associated with Detection and Suppression The specific fire detection and suppression characteristics of each fire zone were addressed and analyzed in some detail. The combined detection and suppression time was compared against the time to damage. This is a proper approach for evaluating the effectiveness of the fire protection system. However, the submittal does not provide any examples of the comparison between the two time periods, nor any probabilistic statement pertaining to their comparison.
Although suppression system unreliability is discussed in the submittal, suppression-failure probabilities were taken to be considerably greater than the unreliability values. This approach accounts for the incomplete effectiveness of the suppression system. That is, in some cases, the suppression system rnay function as designed, but may fail to control the fire simply because of peculiar characteristics of the fire zone - a fact that becomes especially evident when the b.'ations of IPE components and cables are l
considered.
Treatment ofSuppression-induced Damage to E'quipment, ifAvailable c.
No discussion of suppression-induced damage was provided in the submittal. Such damage affects cables and equipment, and results from activation of the fire suppression system in extinguishing a small fire in a given area. A somewhat related issue is discussed as part of the Sandia Fire Risk Scoping Study issues;i.e.,
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7, the " Total Environmental Equipment Survival" issue addresses the survivability of safe shutdown equipment in the event of spurious actuation of a suppression system.
d.
Computer Code Used, ifApplicable i
From the submittal, it is inferred that the formulations provided in FIVE methodology (Reference [12]) were i-used for fire. propagation analysis.
l 2.2.7 Evaluation of Component Fragilities and Failure Modes a.
Definition ofFire-induced Failures
- It is inferred from the IPEEE submittal that fire induced failures were properly considered. Reference [1]
does not discuss spurious actuation of valves or of other equipment / Other than for cables, no concise and specific discussion is provided regarding component fragilities and failure modes.
- b.
Method Used to Determine Component Capacities The cable damage threshold criterion was taken to be 662*F, No other damage criteria are presented in the submittal. For electrical cabinets,it was conservatively assumed that cabinet function is lost upon ignition,
. prior to suppression.
c.
Generic Fragilities No specific discussion is provided regarding equipment generic fragilities. Cable fragility was expressed in terms of a threshold temperature.
f
'd.
Plant-Specific Fragilities No plant-specF.; failure fragilities were mentioned in the submittal.
Technique Used to Treat Operator Recovery Actions e.
' Operator recovery actions were addressed in several cases. For human reliability analysis (HRA) under fire conditions, the same methodology as that used in the IPE was employed. The HRA for fire events was recorded in a separate report; only a summary of the analysis is provided in the IPEEE submittal. Detailed decision trees were developed to enumerate different situations that may influence operator actions.
- Ihree types of operator actions were considered: (1) actions in response to a fire in the plant; (2) actions in response to a fire in the control room; and (3) actions in response to a fire in the cable spreading room. The range of values presented for different performance shaping factors is deemed to be reasonable.
-It is unclear whether or not the possibility of smoke egress, affecting the movement of operators to local panels or valves, was addressed. Optimistic results may be obtained if the effects of smoke egress on operators is not accounted for in human action analysis.
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t.-..........
2.2.8 Fire Detection and Suppression.
Fire detection and suppression was modeled explicitly for many fire scenarios. The combined time of detection and suppression was compared with the time to equipment and cable damage. Manual fire fight i
and the effects of fixed fire suppression systems on the formation of a hot gas layer, were considered. In general, the values reported in the submittal for failure to suppress prior to damage, are deemed to be reasonable.
~
1 2.2.9 Analysis of Plant Systems and Sequences Key Assumptions including Success Criteria and Associated Bases a.
Success criteria were taken directly from the IPE and were not modified for the fire analysis.
b.
Event Trees (Functional or Systemic)
Functional event trees are presented in the fire IPEEE submittal for those initiating events that were deemed to be possible for a fire occurrence. A systemic fault tree was presented for differect initiating events, and was terminated at the front-line system level. No support systems were included in the fault-tree model,
. Dependency Matrix, ofit is Diferentfrom thatfor Seismic Events c.
No dependency matrix was provided in the submittal.
d.
Plans Unique System Dependencies ne submittal does not describe any unique system dependencies, except that Units 1 and 2 share the sane control room enclosure.
e.
Shared Systemsfor Multi Unit Plant The submittal does not discuss the existence of any shared systems between the two units.
f Most Signficant Human Actions Human actions analysis was an integral aspect of fire scenario quantification. The most significant group of fire scenarios is associated with the main control room. The most prevalent event of the dominant fire scenarios for the main control room is failure of the operators in taking the proper actions (specifically, use of the remote shutdown panel). All of the other dominant fire scenarios also include operator error as one of the events that leads the plant from a fire incident to core damage. The most common operator failures in these dominant fire scenarios are noted as follows:
Operator failure to manually control the power-operated relief valves (PORVs).
Operator failure to realign or restore the core cooling systems.
Operator failure to isolate fire safe shutdown equipment.
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2.2.10 Fire Scenarios and Core Damage Frequency Evaluation Core damage frequency is the main parameter used for all screening actions and for identifying poten vulnerabilities. Sophisticated plant damage models (derived from the IPE models) were used. Fire ignit propagation, detection, and damage were quantified, and human error probabilities were evaluated. These frequencies and probabilities were combined using formulations derived from the failure models, and a resulting CDF was calculated.
{
i The IPEEE submittal does not provide sufficient detail to enable verification of a chain of computations for the present review. Although numerous tables and figures were provided in the IPEEE submittal, verification of the final results is not possible.
i 2.2.11 Analysis of Containment Performance Signi6 cant Containment Performance insights a.
'Ihe submittal does not include a discussion of containment fires. However, the submittal does include a detailed review of containment perfonnance as part of the evaluation of fire-initiated accident sequences.
The submittal concludes that, other than for fire-caused interfacing-systems LOCA events, containment performance is the same as that analyzed in the IPE. It was inferred that containment failures occur via the failure of support systems and failure ofcontainment-related frontline systems. Section 4.7 of the submittal provides an extensive discussion concerning the possibility of occurrence of an interfacing systems LOCA, I
and concludes that the occurrence rate is significantly less than 10 /ry.
4 b.
Plant-Unique Phenomenology Considered No containment-related event trees were used in any of the fire screening phases, nor in evaluating the unscreened fire zones.
2.2.12 Treatment of Fire Risk Scoping Study Issues Assumptions Used to Address Fire Risk Scoping Study issues a.
All fire risk scoping study issues were addressed and resolved. The licensee presented a detailed discussion for each issue. The following comments are made with respect to the IPEEE treatment of each major issue:
1.
Seismic-fire interactions were addressed by considering: (1) the potential for a fire event during an earthquake; (2) the potential for failure of fire suppression systems, with resulting effects on safety equipment; and (3) the potential for seismic-induced failure of fire protection systems.
Based on a walkdown of the plant, it was concluded that there are no flammable materials in areas where a seismically induced fire couldjeopardize safety-related equipment.
Inadvertent actuation of suppression systems was addressed in detail. No evidence is provided in the submittal that safety equipment would not be affected. However, it is argued in the submittal that the occurrence of such events has already been adequately addressed in the internal flood Energy Research,Inc.
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<e analysis, and that such occurrence represents only a small ponion of the overall flood occurrence frequency. Several provisions expected to minimize the likelihood of inadvenent actuation of suppression systems are described.
I Regarding failure of fire protection systems due to an canhquake, the seismic review team (SRT) conducted a walkdown of the plant using EPRI NP-6041 procedures. From the SRT assessment, it was concluded that fire protection systems are installed properly and comply with the requirements of Regulatory Guide 1.29.
2.
Regarding fire-barrier qualifications, specific procedures, as well as the IPEEE multi-companmental fire analysis, are cited in reference to the design, inspection, maintenance, and adequacy of fire doors, fire dampers, fire barriers, and penetration seal assemblies.
3.
Several procedures are implemented at the plant for fire detection, fire fighting, general personnel training, and fire brigade training and drills. De brigade undergoes extensive drills on a regular basis. The drills are reviewed and critiqued by observers.
4.
For those areas where the potential for safety equipment damage from inadvenent suppression system actuation exists, the submittal states that the system is equipped with pre action and other isolation devices to minimize the likelihood ofinadvertent actuation and spraying of safe shutdown equipment.
Regarding fire impacts on operator effectiveness in carrying out required tasks (from the standpoint of environmental effects), the submittal states that all auxiliary operators are members of the fire brigade, and therefore, are familiar with the use of fire protection and fire fighting equipment, such that potential adverse effects of their actions will be minimized. Also, the submittal states that operator actions were analyzed in detail in the IPEEE fire analysis. However, no information is provided as to how the~ adverse effects of a fire were modeled and included in the analysis of CDF.
5.
Control system interaction was addressed via use of a remote shutdown panel and via capability to isolate the panels from each other. No information is provided regarding the nature and location of the transfer switches.
b.
Significant Findings 1.'
' The fire brigade undergoes sufficient training.
2.
The suppression systems, in safety-related areas, can withstand seismic events, and include provisions to minimize spraying of safe shutdown equipment.
3.
Procedures are available that address fire-related issues.
4.
The remote shutdown panel, and the provision for isolating the panels from one another, together minimize the potential for control-systems interaction.
2.2.13 USI A-45 Issue Energy Research,Inc.
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m-
.o Methods ofRemoving Decay Heat a.
The fire IPEEE analysis used the IPE models, which account for the entire array of decay heat removal capabilities of the plant.
b.
Abiliry of the Plant to Feed andBleed The IPE model used in the IPEEE includes the provision of feed-and-bleed cooling. The licensee used the
' dominant core damage scenarios to demonstrate how decay heat removal is modeled in the fire analys c.
Credit Takenfor FeedandBleed Credit was taken for feed-and-bleed capability.
j d.
Presence of Thermo-Lag i
Reference [)) mentions the presence of Thermo-Lag at Comanche Peak, but the issue is not discussed in
{
detail in Section 7.0 of the submittal, the licensee states that Thermo Lag has been upgraded according to I
NRC accepted standards. A sensitivity analysis was conducted, where the effect of Thermo-Lag on core damage frequency was investigated. The total fire CDF was found to increase by about 13% if the thermal protection is not credited.
2.3 HFO Events The IPEEE found no vulnerabilities with respect to HFO events. The most significant contributor to the extemal-events CDF was found to be internal fires, whereas the most imponant HFO initiator was found to be tornadoes. Tornadoes contribute about 15% to the CDF reponed in the IPEEE submittal. The tornado PRA, as documented in the IPEEE submittal, revealed that the dominant sequences are station blackout (SBO) sequences, where failure of the diesel generators occurs.
The general methodology utilized in the HFO-events IPEEE follows that recommended in NUREG-1407.
The steps implemented in this general methodology are summarized as follows:
1.
High winds, external flooding, and transportation and nearby facility accidents, were considered for analysis.
2.
Some of these HFO events were screened out, so as to identify the significant events requiring funher analysis.
3.
A scoping analysis was performed for the significant events. Only tornadoes were selected for a detailed PRA treatment. External floods were also analyzed funher, but due to their low frequency of occurrence, they were screened out. Transponation and nearby facility accidents were also evaluated, using a hazard-based screening approach.
4.
The overall HFO analysis was documented in the IPEEE submittal repon.
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n L
In performing Steps (2) and (3) above, the following effons were undertaken:
Review of plant-specific hazard data and licensing basis -
~ Determination as to whether or not the hazard frequency is acceptably low
' Completion of a bounding analysis, if necessary.
l I
' Completion of a PRA,if necessary.
Significant changes ('if any) since the time the plant operating license (OL) was issued were not identified
. in the IPEEE. It is unclear from the report whether or not any such changes were ever identified as pan of the progressive screening analysis. No discussion ~ explicitly stating plant compliance with the 1975 Standard Review Plan (SRP) criteria is provided in the submittal. However, Comanche Peak is a new-vintage plant that should be in compliance with these provisions.
2.3.1 High Winds and Tornadoes 2.3.1.1 GeneralMethodology he CPSES tomado analysis was based primarily on a PRA approach, and relies on the event tree and fault-tree models developed for the IPE [13]. The PRA approach involved three stages in estimating the tornado CDF. De first stage was the determination of the frequency and intensity of tornadoes which may strike i
the CPSES site. The second stage involved the determination of the vulnerability of plant structures and components to tornado-generated missiles.~ The final stage involved quantification of the fault trees and event trees.
2.3.1.2 Plant-Specific Hazard Data and Licensing Basis The tornado occurrence data were obtained from the National Severe Storm Forecast Center [14]. T characterization of tornado intensity was based on the Fujita-Pearson (FPP) rating scale, which includes force l
. intensity, path length and width measures. 'Ihe FPP rating system consists of six tomado intensity classifications, ranging from F0 to F5. The IPEEE study added the F6 classification to account for tornadoes in excess of 318 mph. While existing data covers tomadoes with wind speeds of up to 300 mph, no prior experience with F6-level tornadoes exists in the region. Since Category-I buildings are designed to withstand
' wind speeds of up to 300 mph, the study assumed that F0 to F5 winds will not cause any damage to these buildings, nor to any equipment they house. It was conservatively assumed, however, that an F6-class
. tornado would lead to core damage. It was further assumed that F1 to F5 tomadoes would lead to LOSP,
~ with varying degrees of recovery probability, and with subsequent potential consequential damage to
)
equipment. In addition, tomadoes of F4 and F5 categories were assumed to cause a main steam line break (MSLB), and tomadoes of F3 to F5 categories were assumed to cause loss ofinstrument air.
1 In order to calculate the fmquency of tomado strikes for each FPP category, three separate hazard models were used: (a) WASH-1300 [15), (b) Reinhold Point Strike, and (3) Reinhold Aerial Strike [16,17]. Each
. of these models yields approximately the same result. The submittal assumes that the Reinhold Aerial Strike
_ _ model yields the "best estimate," and thus, this model was used in the study. The licensee lists the frequency l
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L i
of the various F-scale tornadoes in Table 5.1-2 of the submittal. The total frequency of a tomado strike was estimated at 5.0x104/yr. This value seems to be a reasonable estimate.
Aside from the direct wind damage to systems and structures, damage due to tornado-induced missiles was I
also considered. However,it is unclear whether or not any formal analysis. for example using a Monte-Carlo i
approach, was performed to calculate the missile strike probabilities reported in Table 5.1-4 of the submittal.
- D e results indicate that, for an F4 or higher tornado, the probability of a missile strike is 0.999 for the main steam line,0.156 for the diesel generator exhaust, and 0.263 for the station service water (SSW) traveling screens. The remaining targets have low strike probabilities.
Category-I buildings at CPSES are designed to withstand design-basis wind loads, for winds of up to 300 mph. For winds exceeding this design basis, the study assumes that core damage will occur. However, the estimate of the frequency for occurrence of winds exceeding 318 mph (i.e., class-F6 tornado)is 1.87x 10 /yr, 4
which appears somewhat low. The submittal offers nojustification for using this frequency.
Tornadoes that strike transmission lines leading from the switchyard to the plant, were assumed to cause a recoverable loss of offsite power. The frequency of such an event was estimated at 5.0x10"/yr. The conditional probability of core damage, given tornado-induced loss of offsite power, was estimated to be 4
about 7.4x10. Di; conditional probabihty is dominated by the random failure of both emergency diesel generators. It is unclear what value of non-recovery probability was used in the analysis, and to what degree the recoverability of offsite power considered effects of tornadoes (e.g., whether major damage to offsite power lines or the switchyard, caused by tornado, are actually recoverable).
- 23.13 Significant Changes Since Issuance of the Operating License De submittal does not provide any discussion regarding significant changes that have occurred since the time the OL was issued.
23.1.4 Significant Findings and Plant-Unique Features he submittal does not report results of any plant design review; however, the results of the plant walkdown are described in general terms. The design-basis tornado, with a maximum wind speed of 300 mph, is considered to be harmless to Category-I buildings. No other significant findings are cited in the submittal.
- 23.1.5 Hazard Frequency Delicensec used three models to estimate tornado frequency. These models include WASH-1300, Reinhold Point Strike, and Reinhold Aerial Strike. The submittal assumes that the Reinhold Aerial Strike model yields the best estimate for the study. Extreme tornadoes were characterized according to F-scale intensities of the Fujita-Pearson rating scale. The frequency of F-scale categories, as developed in the IPEEE, are: 3.08x10'
- /yrforFI;136x10 /yrforF2;4.09x10 /yrforF3;9.42x10 /yrforF4;1.52x10 /yrforF5;and1.87x10 /yr d
4 4
4 4
for F6. De probability that a tomado-induced missile strikes a building was calculated and is reported in Table 4.1-4 of the submittal. However,it is unclear how the missile-strike analysis was performed. No references to, or description of, the applicable methodology were provided.
23.1.6 PRA Analysis Energy Research,Inc.
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The PRA models used in the CPSES IPE for analysis of internal events [13] were used to develop and quantify tomado-induced sequences. The tornado event trees are shown in Figures 5.1-1 through 5.1-5 of the submittal. It was assumed that, if a tornado of class F1, or higher, (80 mph, or higher, wind speed) strikes transmission lines, then loss of offsite power will occur. Also,it was assumed that missiles may cause some trrnsient events, such as main steam line break, or may lead to equipment failures, such as failure of the diesel generator exhaust or the turbine-driven AFW pump exhaust. However, the effects of these failures on the CDP is small. Since the major risk-significant components are located in Category-I buildings, and since these buildings are assumed not to be affected by missiles or winds for F5 (and lower) tomadoes, no equipment failures within these buildings are considered in the PRA. It is assumed that an F6 tornado will cause damage to these buildings, leading to core damage (with probability of unity). No considention of the possibility of a scenario involving combined wind and flood loads is mentioned in the submittal.
The tornado PRA was conducted using a state-of-the-art approach. He LOSP event tree has delineated accident sequences leading to core damage, caused by either a SBO or a LOCA (induced by reactor coolant i
pump [RCP] seal failure, PORV failure to reclose, or opening of a primary safety relief valve [SRV]).
Fifteen tornado accident sequences were identified and quantified in the IPEEE. The specific cut sets for these sequences are not presented in the submittal. De total tornado CDF was calculated to be 3.7x10 /ry.
4 Table 5.1-6 of the submittal summarius the quantification results. About 54% of the tomado CDF is due to tomadoes of scale F1, with scale F2 having a 23% contribution. Other tornado classes have much smaller l
contributions. De dominant contributing sequences to the tornado CDF are SBO sequences, with a lesser contribution for sequences involving induced-LOCA events. Recovery actions were credited for restoration of offsite electric power and for recovery of diesel generator operation. Since recovery actions are usually credited for tornado conditions with the consideration of magnitude of damage,it is important that the basis for such consideration be fully justified.
Direct tornado damage to plant structures, other than to Category.I buildings, was considered forF3 and F4 tornadoes. The dominant sequence of events in these cases is damage to the turbine building, which results in the failure ofinstrument air. However, the PRA analysis indicates that loss ofinstrument air contributes i
only 3% to the tornado CDP. Also, missile-induced main steam line break contributes less than 0.1 % to the tomado CDF. Tomado-induced missile damage to other equipment at the plant was also determined to have a negligible contribution to the tornado CDF.
2.3.2 External Flooding 2.3.2.1 GeneralMethodology De IPEEE systematically considered the various factors that can contribute to external flooding, such as river floods, probable maximum precipitation (PMP), potential dam failures, surges, hurricanes, and tsunamis. Since the results of the IPEEE showed that none of these factors would lead to a flood level exceeding the current elevation of the CPSES site, external flooding was screened out, and no further analysis was performed.
2.3.2.2 Plant-Specific Hazard Data and Licensing Basis De main source of potential river flooding near the plant is from the Brazos River. Flood records for this river show that the highest flood elevation near the plant has been 601.69 feet. De plant is located at an Energy Research Inc.
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elevation of 810 feet. Similar flood data for Paluxy River indicates a maximum historical flood elevation of 637 feet.
De local flood level due to the PMP was estimated to be '/94.7 feet. Again, all plant facilities are above this elevation. Assuming domino-type dam breaks of bo'.h Morris Sheppard and Decordova Bend dams, the submittal notes that the maximum waterlevel nearthe plant site would reach an elevation of about 700 feet.
His flood level still poses no harud to CPSES.
I ne small size of Squaw Creek Reservoir minimizes 0:: pssibilhy of significant seiches and surges. Also, the SSW intake structure is located above the PMP lael, av is protected by the safe shutdown impoundment dam.
It is unclear from the IPEEE study whether or not the turbine h -
or Nfsite pc wer would be affected,
by PMP flooding.
1 Although the CPSES probable maximum flood analysis was conducted prior to the issuance of Regulatory l
Guide 1.59, " Design Basis Floods for Nuclear Power Plants," a detailed comparison performed by the licensee shows that the plant "... complies with this regulatory guide, with a few minor exceptions." With regard to roof loading, the licensee's evaluation demonstrates that the design is adequate. His is the case because CPSES buildings are equipped with roof drainage systems that can effectively collect and discharge the water volume resulting from a rainfall of six inches per hour, with a maximum rainfall intensity of two inches in five minutes. He roofs of all safety-related buildings are designed to withstand an eight. inch water depth. It should be noted that the PMP and resulting one-hour rainfall produces less than the eight-inch maximum water depth. Herefore, the licensee concludes that no further evaluation is required. Based on this treatment, the licensee considers GI-103 to be resolved for CPSES.
23.23 Significant Changes Since Issuance of the Operating License ne submittal does not discuss any significant changes that have occurred since the time the plant OL was issued.
23.2.4 Significant Findings and Plant-Unique Features No significant findings are reported. De licensee concluded that external floods do not pose any threat that would constitute a plant vulnerability. No discussion is provided that describes the walkdown process and its results.
23.2.5 Hazard Frequency External flooding was screened out due to the fact that the plant elevation is above the elevation of expected flood levels associated with all kinds of severe-flood events. As such, no frequency of occurrence of external flooding, nor conditional probability of plant damage, was calculated.
233 Transportation and Nearby Facility Accidents 233.1 Genem! Methodology Energy Research. Inc.
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De Comanche Peak IPEEE sddressed aircraft crashes, water, rail and highway transportation accidents, mishaps involving on site hazardous material inventories, and potential gas pipeline ruptures.
In the analysis of aircraft crashes, the submittal indicates that no airport operates within five miles of the CPSES site. The submittal indicates that, based on data from the National Transportation Safety Board, and j
from other evaluations,it was determined that the probability of an aircraft crashing into the plant site is l-remote. A statistical model was implemented, based on the probability that an aircraft loses control and on the probability of a random collision (given loss of control). As such, the frequency of an aircraft crashing at the Comanche Peak site was calculated to be 1.19x 10 /yr. The submittal does not indicate whether or not 4
SRP methodology for estimating the impact frequency was employed. The IPEEE methodology is based on determination of: (1) the in-flight crash rate per mile, for aircraft accidents in the vicinity of the plant; (2) the total annual number of flights along the airway above the plant; (3) the size of the target area, which includes all structures that are sensitive to aircraft crash; and (4) the width of the airway above the plant.
No heavily traveled highway passes close to the site. Therefore, no hazard from land transportation accidents was considered in the IPEEE.
Similarly, a survey of on-site toxic materials resulted in the conclusion that any hazard from storage of such materials is minimal. The submittal refers to four pipelines traversing the site vicinity; one is a 26" crude oil line, and the other three are gas pipelines. According to the submittal, the combination of elevation and distance (from the plant site) of these pipelines is such that they will not pose any significant threat to the plant.
2.3.3.2 Plant-Specific Hazard Data and Licensing Basis For aircraft crashes, a frequency of 1.19x10 /yr was estimated. This value is based on recent data provided 4
by the National Transportation Safety Board. No licensing basis related to aircraft crash hazard at the plant site was discussed.
For the analysis of land and water transportation events, no licensing basis information, specific events, nor hazard data were cited. Transportation events were considered to be insignificant based on qualitative arguments.
2.3.3.3 Significant Changes Since Issuance of the Operating License The submittal does not discuss any significant changes that have occurred since the time the plant OL was issued.
2.3.3.4 Significant Findings and Plant-Unique Features No significant findings are discussed in the submittal pertaining to transportation and nearby facility accidents. A description of walkdown findings is not provided in the submittal.
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For aircraft crashes at the site, a frequency of occunrence of 1.19x 10 /yr was estimated by the licensee. His 4
value represents the frequency ofimpact, not the frequency of core damage. Since this frequency is small, i
the submittal reports that no additional analysis to assess the conditional probability of core damage, given an aircraft crash; was performed.
No estimates of hazard frequencies were made for any other type of transponation and nearby facility event.
Qualitative judgments were used to screen out all remaining events. De submittal states that the frequency of explosion from on-site storage of hazardous materials is small.
2.4 Generic Safety Issues (GSI 147. GSI-148. and GSI 172)
-2.4.1 GSI-147, " Fire-Induced Altemate Shutdown / Control Panel Interaction" GSI-147 addresses the scenario of fire occurring in a plant (e.g., in the control room), and conditions which could develop that may create a number of potential control system vulnerabilities. Control system interactions can impact plant risk in the following ways:
Electrical independence of remote shutdown control systems Loss of control power before transfer Totalloss of system function Spurious actuation of components
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The licensee evaluated spurious actuations leading to LOCAs and interfacing system LOCAs. The submittal has followed the guidance provided in FIVE concerning control system interactions on page 4-167, where it states that all circuitry associated with remote shutdown are electrically independent of the control room.
2.4.2 GSI 148, " Smoke Control and Manual Fire Fighti~ng Effectiveness" GSI 148 addresses the effectiveness of manual fire-fighting in the presence of smoke. Smoke can impact plant risk in the following ways:
By reducing manual fire-fighting effectiveness and causing misdirected suppression effons By damaging or degrading electronic equipment j
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By hampering the operator's ability to safely shutdown the plant
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By initiating automatic fire protection systems in areas away from the fire
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Reference [18] identifies possible reduction of manual fire-fighting effectiveness and causing misdirected suppression efforts as the central issue in GSI-148. Manual fire-fighting was not credited in the analysis.
Thus, the issue of manual fire-fighting effectiveness is not addressed in this TER.
2.4.3 GSI-172, " Multiple System Responses Program (MSRP)"
Reference [18] provides the description of each MSRP issue stated below, and delineates the scope of information that may be reported in an IPEEE submittal relevant to each such issue. The objective of this Energy Research,Inc.
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l subsection is only to identify the location in the IPEEE submittal where information having potential l
relevance to GSI-172 may be found.
Common Cause Failures (CCFs) Related to Human Errors Descriotion of the Ieme [18]: CCFs resulting from human errors include operator acts of commission or omission that could be initiating events, or could affect redundant safety-related trains needed to mitigate the events. Other human errors that could initiate CCFs include: manufacturing errors in components that affect redundant trains; and installation, mainter. ace or testing errors that are repeated on redundant trains.
In IPEEEs, licensees were requested to address only the human errors involving operator recovery actions following the occurrence of external initiating events.
)
. Section 3.1.2 (page 3-23) of the CPSES IPEEE submittal briefly mentions operator recovery actions, as a l
consideration for success path selection in the seismic analysis; detailed supporting information has been supplied in Reference [8]. In regard to the fire analysis, the submittal provides information on operator.
i recovery actions in Sections 4.1 (page 4-2 and Figure 4.1-1),4.1.6,4.6.1,4.6 (Section 4.6.8 provides a detailed discussion of the human reliability analysis), and 4.8 (pages 4-163 to 4-167). For HFO events, 4
operator actions are discus' ed in Sections 5.1.3 and 5.1.4.
s Non Safety-Related Control SystemfSafety-Related Protection System Dependencies Descriotion of the Issue [18): Multiple failures in non safety-related control systems may have an adverse impact on safety-related protection systems, as a result of potential unrecognized dependencies between control and protection systems. The concern is that plant-specific implementation of the regulations regarding separation and independence of control and protection systems may be inadequate. The licensees
- IPE process should provide a framework for systematic evaluation ofinterdependence between safety-related and non-safety-related systems, and should identify potential sources of vulnerabilities. The dependencies between safety-related and non-safety-related systems resulting from external events -i.e., concerns related to spatial and functional interactions - are addressed as part of " fire-induced alternate shutdown and control room panel interactions," GSI-147, for fire events, and " seismically induced spatial and functional interactions" for seismic events.
l
, Information provided in the Comanche Peak IPEEE submittal pertaining to seismically induced spatial and i
functional interactions is identified below (under the heading Seismically Induced Spatial and Functional i
Interactions), whereas information pertaining to fire-induced alternate shutdown and control panel interactions has already been identified in Section 2.4.1 of this TER.
Heat /Smole/ Water Propagarion Efectsfrom Fires Descriotion of the Issue 118): Fire can damage one train of equipment in one fire zone, while a redundant train could potentially be damaged in one of the following ways:
i 1
4 Heat, smoke, and water may propagate (e.g., through HVAC ducts or electrical conduit) into a second fire zone, and damage a redundant train of equipment.
l A random failure, not related to the fire, could dtmage a redundant train.
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Multiple non safety-related control systems could be damaged by the fire, and their failures could affect safety-related protection equipment for a redundant train in a second zone.
A fire can cause unintended operation of equipment due to hot shorts, open circuits, and shons to ground.
Consequently, components could be energized or de-energized, valves could fail open or closed, pumps could continue to run or fail to mn, and electrical breakers could fail open or closed. The concern of water propagation effects resulting from fire is panially addressed in GI-57, Effects of Fire Protection System Actuation on Safety-Related Equipment." The concern of smoke propagation effects is addressed in GSI-148. He emvem of alternate shutdown / control room interactions (i.e., hot shorts and other items just mentioned)is addressed in GSI-147.
Information provided in the Comanche Peak IPEEE submittal pertaining to GSI-147 and GSI-148 has i
i already been identified in Sections 2.4.1 and 2.4.2 of this TER. Sections 3.2 (pages 3-33 and 3-34) and 4.8 (pages 4-164 to 4-167) of the submittal present information pertaining to this issue.
Efects ofFire Suppression System Actuation on Non Safety RelatedandSafety-RelatedEquipment Descriotion of the Issue [18): Fire suppression system actuation events can bve an adverse effect on safety-related componer.ts, either through direct contact with suppression agents or th ough indirect interaction with i
non-safety related components. His concem is adaressed in GI-57.
Information penaining to suppression-induced damage to equipment, as well as seismically induced inadvertent actuation of fire suppression systems. can be found, respectively, in pages 4-164 to 4-167, and pages 4-159 to 4-161, of Section 4.8 of the IPEEE submittal. Discussion on GI-57 is provided on pages 3-33 and 3-34.
Efects of Flooding and/or Moisture intrusion on Non-Safety-Related and Safery-Related Equipment Descriotion of the Issue [18]: Flooding and water intmsion events can affect safety-related equipment either directly or indirectly through flooding or moisture intrusion of multiple trains of non-safety-related equipment. This type of event can result from external flooding events, tank and pipe ruptures, actuations i
of fire suppression systems, or backflow through parts of the plant drainage system. The IPE process addresses the concems of moisture intmsion and intemal flooding (i.e., tank and pipe ruptures or backflow through pan of the plant drainage system). The guidance for addressing the concern of external flooding is i
provided in Chapter 5 ofNUREG-1407, and the concem of actuations of fire suppression systems is provided in Chapter 4 ofNUREG-1407.
The following information is provided relevant to this issue: the Comanche Peak IPEEE submittal discusses external floods in Section 5.2; discussion is provided on pages 4-164 to 4-167 of Section 4.8 and pages 3-33 to 3-34 of Section 3.2 regarding actuations of fire suppression systems; discussion of seismically induced inadvenent actuation of fire suppression systems is provided on pages 4-159 to 4-161 of Section 4.8; and seismically induced internal and external flooding are discussed, respectively, on page 4-160 of Section 4.8, and Section 5.2.
Seismically induced Spatial and Functional Interactions Energy Rescatch Inc.
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Descriotion of the Issue [18): Seismic events have the potential to cause multiple failures of safety-related systems through spatial and functional interactions. Some particular sources of concern include: ruptures in small piping that may disable essential plant shutdown systems; direct impact ofnon-seismically qualified structures, systems, and components that may cause small piping failures; seismic functional interactions of control and safety-related protection systems via multiple non-safety-related control systems' failures; and indirect impacts, such as dust generation, disabling essential plant shutdown systems. As part of the IPEEE, it was specifically requested that sei;mically induced spatial interactions be addressed during plant walkdowns. The guidance for performing such walkdowns can be found in EPRI NP-6041.
'Ihe Comanche Peak IPEEE has included a seismic walkdown which investigated the potential for adverse physical interactions. The submittal siates that EPRI NP-6041 guidelines were followed in the seismic walkdowns. Relevant information can be found in Sections 3.I.1 (pages 3-3,3-13 to 3-20),3.1.5, and 4.8 (pages 4-158 to 4-161) of the submittal.
Seismicallyinduced Fires Descriotion of the Issue [18): Seismically induced fires may cause multiple failures of safety-related systems. The occurrence of a seismic event could create fires in multiple locations, simultaneously degrade fire suppression capability, and prevent mitigation of fire damage to m altiple safety-related systems.
1 Seismically induced fires is one aspect of seismic-fire intersction concems, which is addressed as part of the Fire Risk Scoping Study (FRSS) issues. (IPEEE guidance specifically requested licensees to evaluate FRSS issues.) In IPEEEs, seismically induced fires should be addressed by means of a focused seismic-fire interactions walkdown that follows the guidance of EPRI NP-6041.
Section 4.8 (pages 4-157 to 4-159) of the Comanche Peak IPEEE submittal provides a discussion of seismically induced fires.
Seismically induced Fire Suppression System Actuation j
i Descriotion of the Issue [18): Seismic events can potentially cause multiple fire suppression system actuations which, in tum, may cause failures of redundant trains of safety-related systems. Analy5es currently required by fire protection regulations generally only examine inadvertent actuations of fire suppression systems as single, independent events, wher9as a seismic event could cause multiple actuations of fire suppression systems in various areas.
Sections 4.8 (pages 4-159 to 4-161) of the Comanche Peak IPEEE subrA. '.1 provides discussion of seismically induced fire suppression system actuation and degradation.
SeismicallyinducedFlooding Descriotion of the Issue [18]: Seismically induced flooding events can potentially cause multiple failures of safety-related systems. Rupture of small piping could provide flood sources that could potentially affect multiple safety-related components simultaneously. Similarly, non-seismically qualified tanks are a potential flood source of concern. IPEEE guidance specifically requested licensees to address this issue.
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'!he Comanche Peak IPEEE submittal includes infonnation on seismically induced flooding in Sections 4.8 i
(page 4-160) and 5.2.
SeismicallyinducedRelay Charter Descriotion of the Issue [18): Essential relays must operate during and after an earthquake, and must meet one of the following conditions:
remain functional (i.e., without occunence of contact chattering);
a be seismically qualified;or be chatter acceptable.
It is possible that contact chatter of relays not required to operate during seismic events may produce some unanalyzed faulting mode that may affect the operability of equipment required to mitigate the event. IPEEE guidance specifically requested licensees to address the issue of relay chatter.
In NUREG 1407, Comanche Peak is identified as a reduced-scopa plant where a relay chatter evaluation is not required (since CPSES is not a USI A-46 plant). Consequently, as noted in Section 2.1.9 of this TER, and as stated on page 3-21 of the submittal, the Comanche Peak IPEEE did not include a relay chatter evaluation.
Evaluation ofEanhquale Magnitudes Greater than the Safe Shutdown Eanhquake Descriotion of the Issue [l 8): The concern of this issue is that adequate margin mcy not have been included in the design of some safety-related equipment. As pan of the IPEEE, all licensees are expected to identify potential seismic vulnerabilities or assess the seismic capacities of their plants either by performing seismic PLAs or seismic margins assessments (SMAs). The licensee's evaluation for potential vulnerabilities (or unusually low plant seismic capacity) due to seismic events should address this issue.
Comanche Peak is designated as a reduced-scope plant in NUREG-1407, and consistent with the relevant guidelines for a reduced-scope plant, the IPEEE has considered seismic input equivalent to the SSE level.
1 m hquake loads in excess of the SSE have not been considered.
Efects ofHydrogen Line Ruptures Descriotion of the Issae [18): Hydrogen is used in electrical generaters at nuclear plants to reduce windage
- losses, and as a heat transfer agent. It is also used in some tanks (e.g., volume control tanks) as a cove, gas.
Leaks or breaks in hydrogen supply piping could result in the accumulation of a combustible mixture of air and hydrogen in vital areas, resulting in a fire and/or an explosion that could damage vital safety-related systems in the plants. It should be anticipated that the licensee will treat the hydrogen lines and tanks as potential fixed fire sources as described in EPRI's FIVE guide, assess the effects of hydrogen line and tank ruptures, and report the results in the fire ponion of the IPEEE submittal.
~ The submittal provides some briefdiscussion of potential seismically induced ruptures of hydrogen lines in Section 4.2.5 (page 4-21).
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' OVERALL EVALUATION, CONCLUSIONS AND RECOMMENDATIONS 3.1 Seismic The approach chosen by the licensee for conducting the seismic IPEEE essentially addresses all relevant issues and concerns for Comanche Peak Steam Electric Station, a reduced-scope site. In addition, the format for documenting the seismic IPEEE, although brief, was generally well structured, and followed the recommendations of NUREG-1407.
j He Comanche Peak seismic evaluation has found and addressed some meaningfulIPEEE-related concerns, he study repons a comparatively small number of outliers (14 anomalies, two of which were considered to be noteworthy). However, the small number of reported anomalies is notjudged to be unreasonable, as l
the plant has recently undergone extensive design and construction verification work, including plant walkdowns. He IPEEE seismic review team was generally able to confirm that these earlier efforts (including seismic qualification work and evaluation of systems interactions) insured that safe shutdown systems and containment systems were properly designed and could withstand loads incurred by an SSE event.
i Based on this submittal-only review, the follo' wing items are identified as the primary strengths and weaknesses of the seismic IPEEE submittal for CPSES
)
SIMEthft 1.
The study implements a relevant, meaningful approach for conducting the seismic IPEEE, and considers all major issues of concem.
2.
De study made extensive use of plant personnel (who received training in seismic evaluation methods), yet utilized outside expertise where needed; thus it was ensured that the study was completed (and reviewed) by qualified personnel, and that the licensee was well familiar with the evaluation.
3.
De study provides a good background description of historical programs conducted at Comanche Peak.
4.
Several components were included in the seismic evaluation.
Wann,=
1.
The seismic-fire evaluation did not explicitly consider the potential and effects of seismically induced electrical cabinet fires; rupture / leakage of gas piping; flooding due to seisrzuc failure of non-SSEL water tanks and piping; and seismically induced loss of fire suppression capability.
It is clear that the licensee has benefitted from the seismic IPEEE effort, that plant anomalies have been identified and addressed,'and that the licensee has improved its understanding of plant behavior in response to potential severe accidents caused by earthquakes.
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With respect to findings of the Comanche Peak seismic IPEEE, it is impon.nt to highlight that existing 1
anomalous conditions as noted in the submittal (e.g., suspect conditions and configumdons of anchorage and connections, presence of cracks in concrete foundation pads, and seismic interaction issues), car. become more significant concerns as time proceeds, and new anomalies (e.g., rusting / cracking of anchorage or suppons) may also be encountered, due to plant degradation with the passage of time. It is thus considered important that such anomalo..onditions be monitored; inasmuch, the benefits of maintaining a living i
seismic IPEEE for Comanche Peak are obviously substantial.
3.2 HM-The licensee expended considerable effon in the preparation of the fire analysis portion of the Comanche Peak IPEEE. The licensee employed a valid methodology (a Level-1 PRA) and used appropriate databases for fire occurrence and fire suppression system failure quantifications. The primary strengths and weaknesses of the fire IPEEE are summarized below.
Streneths The following are considered to be the strengths of the fire IPEEE submittal:
The documentation is well written. The overall presentation in the submittal is clear and well organized. Tables a nd figures provide a considerable amount of in'ormation that suppons the analysis and conclusions.
State-of-the-an methodology and data were used. The significan; attention to detail,in fact, exceeds current standards of practice.
l The fina' conclusions are reasonable and are within the range of results expected for a PWR power plant.
The licensee used conservative screening methods and criteria.
The list ofissues addressed by the licensee is extensive and is considered to be complete.
Weaknesses 1
Weaknesses of the fire IPEEE submittal, which are deemed to be sufficiently important to be brought to the licensee's attention, include:
The submittal provides extensive explanation of the methods used, and of the key issues addressed,
. in the fire IPEEE. However, examples from the analysis were often not provided in support of the discussions. To give one instance,it was stated that the operator error analysis includes the adverse environmental effects of fire; yet, no example was provided from the analysis to show how these effects were included in accident sequence modeling and in core damage frequency evaluations.
l Similar to the preceding note,in several instances, the licensee did not specifically describe the data or the actions that were undenaken during data analysis. For example, it is not clear from the l
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submittal that the IPE initiators were reviewed explicitly to ensure that all possible initiators for fire events were considered. As another example, the licensee did not elaborate on whether the Appendix-R list of cables and equipment was modified to include the IPE list of cables and equipment.
Despite these weaknesses,it may be confidently stated that the licensee has clearly gained some excellent insights, and has realized an important experience, from its investigation of potential fire vulnerabilities at CPSES.
3.3 HFO Events The CPSES IPEEE for HFO events generally followed the submittal guidelines of NUREG-1407. He primary strengths and weaknesses of the submittai are summarized as follows:
Strengths 1.
The analysis of tomado events was detailed and used state-of-the-art methods. De submittal report provides detailed discussions concerning the approaches used to evaluate relevant probabilities.
2.
De HFO analysis was performed and reviewed primarily by TU Electric, using itsin-house expertise. This evaluation strategy has ma'imized the licensee's appreciation of severe-accident behavior and understanding of the most important sequences and contributors to core damage frequency.
Weaknesses l
1.
Significant changes (if any) since issuance of the plant operating license were not discussed. It is unclear from the submittal report whether or not such changes were identified as part of the progressive screening. No discussion concerning plant compliance with 1975 SRP criteria was provided.
2.
He tornado PRA credited operator recovery actions in restoring offsite power; yet, there is no discussion in the submittal report to assure that the applicability and practicality of such recovery actions were considered.
3.
De submittal did not provide the cut sets for the tornado-induced core damage sequences.
Herefore, it is difficult to independently verify the adequacy of the conditional core damage probability, given a tomado event.
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4 IPEEE INSIGHTS, IMPROVEMENTS, AND COMMIDIENTS 4.1 Eglaglg
. He key seismic IPEEE insights are walkdown related; no quantitative insights have been derived from the seismic evaluations. Thus, no values for plant-level, system-level, or component high-confidence of low-probability of failure (HCLPF) capacities have been estimated as a result of the seismic IPEEE.
The licensee makes a general conclusion in the IPEEE submittal that there are no plant-specific
- vulnerabilities to severe accidents from extemal events. However, the reduced-scope seismic evaluation has revealed a few minorplant anomrJies and maintenance concerns. The two observations that were considered to be significant enough to report are:
De presence of unanchored, ancillary equipment close to safety related equipment, in the control room.
An instance ofinsufficient clearance between a motor control center and adjacent cable tray supports (leading to associated concerns with respect to relay chatter).
The first of these observations was noted by the seismic IPEEE reviewer, as opposed to the seismic review team members. The submittal notes that, to address both of these issues, the seismic review team took appropriate follow-up actions to ensure their satisfac ory resolution. These outliers were encountered with respect to a review of safe shutdown equipment; no outliers were found specifically as a result of the containment performance evaluation.
4.2 Eks Overall, the licensee concludes that there are no significant fire vulnerabilities at Comanche Peak. The total fire-induced core damage frequency (CDF), under full-power operating conditions, was estimated to be 2.09x10'5 per reactor-year (ry). Given the level of detail of the fire analysis, this value represents the combined fire CDF attributable to alllocations within the plant. This value of fire CDFis within the range of results obtained for other nuclear power plants.
With the exception of fire scenarios in the control room (which have a combined CDF of 9.04x10 /ry), the 4
remaining dominant fire scenarios have CDF contributions ranging from 10 hy to 2x10 /ry.
4 4
De control-room fire scenarios are dominated by operator failure to control the plant from the remote shutdown panel. Dese scenarios are modeled by loss of offsite power and failure to align a diesel generator, with a resulting station blackout (SBO).
De following events appear in the majority of the dominant fire scenarios:
RCP seal failure, and resulting small LOCA, following loss of component cooling water (CCW).
Small LOCA due to a stuck-open power-operated relief valve (PORV) or safety relief valve (SRV).
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' Operator error in controlling the PORVs.
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Operator failure to isolate safe shutdown equipment from the effects of a fire.
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. Operator faimre to properly align or restore core-coolirg systems.
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'Ihe effort undertaken in performing the fire IPEEE has, of course, provided an excellent eyyudunity for
.. licensee engineers to improve their knowledge of the characteristics of the plant, of how the plant would behave under fire conditions, and of what human actions would be necessary to protect the core from any adverse fire-related effects.
No improvements nor commitments were determined in the fire IPEEE as being necessary to further reduce the fire risk at Comanche Peak.
4J HFO Events
. The HFO-events CDF(due exclusively to tomadoes) was estimated to be 3.7x10 hy, which comprises a 4
comparatively low contribution (about 15%) to the total reported mean CDF due to external events (i.e., a CDF value of 2.46x10 shy for fires and tornadoes combined).
No vulnerabilities associated with HFO events were identified. Due to the small value of CDF attributable to tomado events, the licensee did not consider any plant-specific safety enhancements. Hence, no fixes or commitments related to HFO events are planned by the licensee.
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4 5
IPF.EE EVALUATION AND DATA
SUMMARY
SHEETS Completed data entry sheets for the Comanche Peak Steam Electric Station IPEEE are provided in Tables
{
i 5.1 to 5.7. These tables have been completed in accordance with the descriptions in Reference [7]. Table 5.1 lists the overall extemal events results. Table 5.2 summarizes general seismic data pertaining to the reduced-scope seismic evaluation. Table 5.3 provides the PWR Seismic Success Paths table, which givesf a description of the success paths developed for the reduced-scope seismic evaluation. Tables 5.4 and 5.5, respectively, present PWR Accident Sequence Overview and PWR Accident Sequence Detailed tables for the fire PRA. Tables 5.6 and 5.7, respectively, present PWR Accident Sequence Overview and PWR Accident Sequence Detailed tables for the wind (tomado) PRA.
j De tables for the fire PRA (Tables 5.4 and 5.5) include specific equipment damage and human-acticn scenarios that lead to core damage with a frequency greater than 10~'/ry. In addition to this fonnat for reporting scenarios, the fire IPEEE submittal also catalogs scenarios by fire zones, generally as a collection of specific equipment-damage scenarios. Often, the total core damage frequency for a fire zone will be much greater than indicated by the frequencies provided in the attached tables, though individual core damage frequencies for the specific equipment damage scenarios are always less than 10-'hy. He control room has the largest contribution to the total fire core damage frequency. However, specific information provided in the submittal, regarding the sequence of events leading to core damage, is insufficient to completely fill out Tables 5.4 and 5.5.
i 1
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6-REFERENCES 1.
" Individual Plant Examination of External Events for Severe Accident Vulnerabilities - Comanche Peak Steam Electric Station," Texas Utilities Electric Company, ER-EA-008, June 27,1995.
2.
" Individual Plant Examination of Extemal Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f)," U. S. Nuclear Regulatory Commission, Generic Letter 88-20, Supplernent 4, June 28,1991.
3.
J. T. Chen, et al., " Procedure and Submittal Guidance for the Individual Plant Examination of l_
Extemal Events (IPEEE) for Severe Accident Vulnerabilities," U. S. Nuclear Regulatory Commission, NUREG-1407,May 1991.
l l
4.
" Design Respc,nse Spectra for Seismic Design of Nuclear Power Plants," U.S. Atomic Energy i
Commission, Regulatory Guide 1.60, Revision 1, December 1973.
5.
R.T. Sewell, et al., "Individur.! Plant Examination for Extemal Events: Review Guidance,"
ERI/NRC 94-501, Draft, May 1994.
6.
"IPEEE Step 1 Review Guidance Document," U. S. Nuclear Regulatory Commission, June 1992.
7.
S. C. Lu and A. Boissonade, "IPEEE Database Data Entry Sheet Package," Lawrence Livermore National Laboratory, December 14,1993.
l 8.
" Comanche Peak Steam Electric Station (CPSES), Docket Nos. 50-445 and 50-446, Response to NRC Request for Additional Information on CPSES Individual Plant Examination of Extemal Events (TAC No. M83608)," Letter from C. L. Terry, Texas Utilities Electric Company, to the U.
i S. Nuclear Regulatory Commission, February 1,1996.
9.
"A Methodology for Assessment of Nuclear Power Plant Seismic Margin," Electric Power Research Institute, EPRI NP-6041-SL, Revision 1, August 1991.
10.
" Fire Risk Analysis Implementation Guide," Electric Power Research Institute, Project 3385-01, Draft Report, January 1994.
11.
" Fire Event Database for U.S. Nuclear Power Plants," Electric Power Research Institute, NSAC-178L, Final Report, June 1992.
- 12. ~ -
- Fire Induced Vulnerability Evaluation (FIVE)," Electric Power Research Institute, TR-100370, Revision 1, September 29,1993.
13.
"TU Electric, Individual Plant Examination, Comanche Peak Steam Electric Station," RXE-92-01 Volumes I and H. August / October 1992.
14.
"TORPLOT for 125 NM Radius Around CPSES Site,1950- 1993," National Severe Storms Forecast Center, Kansas City, Missouri.
15.
E. H. Markee, Jr., " Technical Basis for Interim Regional Tornado Criteria," Atomic Energy Commission, WASH-1300, May 1974.
Energy Research,Inc.
54 ERI/NRC 95-609 l
.c s. ie*
i 16.
T. A. Reinhold, and B. Ellingwood, " Tornado Damage Risk Assessment," Center for Building,,,
Technology, National Bureau of Standards, NUREG/CR-2944, September 1982. -
i 17.
T. R. Mcdonald, "A Methodology for Tornado Hazard Probability Assessment lInstituteTor~-
Disaster Research, Texas Tech University, NUREG/CR-3058, October 1983.
(
\\
[
18.
" Staff Guidance ofIPEEE Submittal Review on Resoiution of Generic or Unresolved Safety r==@
Issues (GSI/USI)," U.S. Nuclear Regulatory Commission August 21l1997 '. '- --
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